BY
C. fl. ERDmflN
jlU, KELLY
m. KIRBIYIK
fl. 6.
PREP A RED FOR
osasr-
-------
BIBLIOGRAPHIC DATA
SHEET
1. Report No.
EPA-520/3-75-Q19
3. Recipient's Accession No.
4. Title and Subtitle
Radionuclide Production, Transport, and Release
from Normal Operation of Liquid-Metal-Cooled Fast Breeder
Reactors
5. Report Date ,.T , n-,-7,-
November 1975
Cdate published^
6.
7. Author(s) C. A. Erdman, J. L. Kelly, M. Kirbiyik, A. B. Reynolds
(All affiliated with University of Virginia. Charlottesville')
8. Performing Organization Rept.
No. NE-4146-102-73
9. Performing Organization Name and Address
University of Virginia
Office of Sponsored Programs
P. 0. Box 3901
Charlottesville, Virginia 22903
10. Project/Task/Work Unit No.
11. Contract/Grant No.
68-01-0547
12. Sponsoring Organization Name and Address .
Environmental Protection Agency
Office of Radiation Programs
Washington, D. C. 20460
13. Type of Report & Period
Covered
Final
14.
15. Supplementary Notes
Extracted portions of this report were published in NUCLEAR SAFETY,
Vol. 16-1 (Jan/Feb 1975) and Vol 16-3 (May/June 1975)
16. Abstracts Sources of radioactivity from the normal operation of an LMFBR and the transport
of this radioactivity, were studied. Reliance was placed predominantly on published re
suits although some new calculations were made where needed. Results were normalized t<(> a
1000 MWe LMFBR and compared to values for a 1000 MWe LWR. The radioactivity sources
studied included plutonium and other transuranium elements, fission products, tritium,
corrosion products, and tramp fuel. Radionuclide transport studies included fission pr(
ducts and fuel from failed fuel, behavior of radioactivity in sodium and cold traps, an
operation of gaseous radwaste systems. Operating experience was reviewed, including da
from the fast reactors EBR-II, Fermi, SEFOR, Dounreay, Rapsodie, and BR-5, and limited
data from the thermal reactors SRE, S8ER, and Hallam. The authors completed this study in
September 1973. Since that time, a number of technical papers, reports, and studies ha-re
been published which might serve to extend or refine some of the conclusions of the preuent
study (and in some cases may even refute results in the study). Therefore, the users o
1"hi.q rp.nairt arp rauf'irmpfl t~n kp.cm in mind 1"hp Sppi'pTnbpr 197^ "rnf'nff" date, far t".he ref •
17. Key Words and Document Analysis. 17a. Descriptors
Radioactive Wastes (1807)
Fission Products
Radioactive Isotopes
Radioactive Waste Processing
Radioactivity (1808)
Fission Products
Radioactive Materials
Induced Radioactivity
17b. Identifiers /Open-Ended Terms
erences cited and used in
preparing this study.
Nuclear Reactors (1809)
Breeder Reactors
Fast Reactors (nuclear)
Liquid Metal Cooled Reactors
17c. COSATI Field/Group
MICK SUBJECT TO CHANGE
18. Availability Statement Release unlimited - Limited number
of copies available free from: U0 S. EPA, Technology
Assessment Division (AW-459), 401 "M" Street, SW,
Washington D.C. 20460
19.. Security Class (This
Report)
UNCLASSIFIED
21. No. of Pages
20. Security Class (This
Page
UNCLASSIFIED
22. Price
FORM NTis-35 (REV. 10-73) ENDORSED BY ANSI AND UNESCO.
THIS FORM MAY BE REPRODUCED
USCOMM-DC 826S-P74
-------
INSTRUCTIONS FOR COMPLETING FORM NTIS-35 (Bibliographic Data Sheet based on COSATI
Guidelines to Format Standards for Scientific and Technical Reports Prepared by or for the Federal Government,
PB-180 600).
1. Report Number. Each individually bound report shall carry a unique alphanumeric designation selected by the performing
organization or provided by the sponsoring organization. Use uppercase letters and Arabic numerals only. Examples
FASEB-NS-73-87 and FAA-RD-73-09.
2. Leave blank.
3. Recipient's Accession Number. . Reserved for use by each report recipient.
4. Title and Subtitle. Title should indicate clearly and briefly the subject coverage of the report, subordinate subtitle to the
main title. When a report is prepared in more than one volume, repeat the primary title, add volume number and include
subtitle for the specific volume.
5- Report Date. Each report shall carry a date indicating at least month and year. Indicate the basis on which it was selected
(e.g., date of issue, date of approval, date of preparation, date published).
6. Performing Organization Code. Leave blank.
7. Author(s). Give name(s) in conventional order (e.g., John R. Doe, or J.Robert Doe). List author's affiliation if it differs
from the performing organization.
8. Performing Organization Report Number. Insert if performing organization wishes to assign this number.
9. Performing Organization Name and Mailing Address. Give name, street, city, state, and zip code. List no more than two
levels of an organizational hierarchy. Display the name of the organization exactly as it should appear in Government in-
dexes such as Government Reports Index (GRI).
10. Project/Task/Work Unit Number. Use the project, task and work unit numbers under which the report was prepared.
11. Contract/Grant Number. Insert contract or grant number under which report was prepared.
12. Sponsoring Agency Name and Mailing Address. Include zip code. Cite main sponsors.
13- Type of Report and Period Covered. State interim, final, etc., and, if applicable, inclusive dates.
14. Sponsoring Agency Code. Leave blank.
15. Supplementary Notes. Enter information not included elsewhere but useful, such as: Prepared in cooperation with . . .
Translation of ... Presented at conference of ... To be published in ... Supersedes . . . Supplements
Cite availability of related parts, volumes, phases, etc. with report number.
16. Abstract. Include a brief (200 words or less) factual summary of the most significant information contained in the report.
If the report contains a significant bibliography or literature survey, mention it here.
17. Key Words and Document Analysis, (a). Descriptors. Select from the Thesaurus of Engineering and Scientific Terms the
proper authorized terms that identify the major concept of the research and are sufficiently specific and precise to be used
as index entries for cataloging.
(b). Identifiers and Open-Ended Terms. Use identifiers for project names, code names, equipment designators, etc. Use
open-ended terms written in descriptor form for those subjects for which no descriptor exists.
(c). COSATI Field/Group. Field and Group assignments are to be taken from the 1964 COSATI Subject Category List.
Since the majority of documents are multidisciplinary in nature, the primary Field/Group assignment(s) will be the specific
discipline, area of human endeavor, or type of physical object. The application(s) will be cross-referenced with secondary
Field/Group assignments that will follow the primary posting(s).
18. Distribution Statement. Denote public releasability, for example "Release unlimited", or limitation for reasons other
than security. Cite any availability to the public, other than NTIS, with address, order number and price, if known.
19 & 20. Security Classification. Do not submit classified reports to the National Technical Information Service.
21. Number of Pages. Insert the total number of pages, including introductory pages, but excluding distribution list, if any.
22. NTIS Price. Leave blank.
FORM NTIS-35 (REV. 10-73) USCOMM-DC 8265-P74
-------
This report was prepared as an account of work sponsored
by the Environmental Protection Agency (EPA) of the United
States Government under Contract No. 68-01-0547. The report has
been reviewed and edited by the EPA and approved for publication.
However, approval does not signify that the contents of the report
necessarily reflect the views and policies of the EPA. Neither
the United States nor the EPA makes any warranty, express or
implied, or assumes any legal liability or responsibility for the
accuracy, completeness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use would not
infringe privately owned rights.
-------
FOREWORD
This report was prepared for the Environmental Protection
Agency by the University of Virginia, Department of Nuclear •
Engineering, for the purpose of summarizing the available informa-
tion on radioactivity discharges from liquid metal fast breeder
reactors (LMFBR's). The Energy Research and Development Administration has
underway an extensive effort to perfect this reactor type for
commercial operation to produce electricity. While producing
electricity, the LMFBR will at the same time breed more fissile
material than is consumed by the reactor. This is accomplished by
using excess neutrons from the fission process to convert the abun-
dant isotope of uranium (238y) to the fissile plutonium isotope 239p .
(Smaller quantities of other plutonium isotopes, some fissile and
some non-fissile, are also produced in the process, by successive
neutron absorptions and beta decays.) The plutonium so produced
can then be extracted and used to refuel the LMFBR, or to provide
fuel for other.LMFBR's or other reactor types, Present-day light-
water reactors operate by fissioning 235y, which comprises only
0.7% of natural uranium, the balance being essentially all "8^
By converting this 138-times more abundant "°U isotope to fissile
material the LMFBR will, effect a many-fold increase in the amount
of electrical energy which in principle can be produced from the
available uranium deposits.
This report is being published so that it will be available
as a resource to the scientific community and the general public.
The results of the report will assist EPA in assessing the environ-
tnenal impacts of the LMFBR program as a whole as well as those of
related individual- facilities as they are developed. - The information
also will be used to assist EPA in developing generally applicable
environmental radiation standards for LMFBR-related facilities, and
may be of assistance in our evaluations of LMFBR accidents.
We solicit, and will appreciate receiving, any corrections or
critical comments on the information and conclusions contained in
this report. Please send any such comments to the Environmental
Protection Agency, Office of Radiation Programs (AW-459), Washington,
D.C. 20460.
.P.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs (AW-458)
-------
IV
-------
ABSTRACT
Sources of radioactivity from the normal operation of an
LMFBR, and the transport of this radioactivity, were studied.
Reliance was placed predominantly on published results although some
new calculations were made where needed. Results were normalized to
a 1000 MWe LMFBR and compared to values for a 1000 MWe LWR.
Sources of radioactivity which were studied included plutonium
and other transuranium elements, fission products, tritium,
corrosion products, activation products, and tramp fuel.
The study of the transport of radionuclides included reviews
of transport of fission products and fuel from failed fuel, behavior
of radioactivity in sodium and cold traps, and operation of gaseous
radwaste systems.
Operating experience for liquid metal cooled reactors
relating to radioactivity was reviewed. Included were data from the
fast reactors EBR-II, Fermi, SEFOR, Dounreay, Rapsodie, and BR-5, and
limited data from the thermal reactors SRE, S8ER, and Hal lam.
The authors conpleted this study in September, 1973. Since
that date a number of technical papers, reports, and studies have
been published which might serve to extend or refine some of the
conclusions of the present study (and in some cases may even refute
results in this study). Therefore, the users of this report
are cautioned to keep in mind the September, 1973 "cutoff date"
for the references cited and used in preparing this study.
-------
-------
TABLE OF CONTENTS
Page
1. INTRODUCTION 1
1.1 Objectives of the Present Study 1
1.2 Some General Results of the Study z
2. SUMMARY.... . 3
2.1 Objectives and Methods 3
2.2 Sources 4
2.2.1 Plutonium and Other Transuranium Elements.. 4
2.2.2 Fission Products 5
2.2.3 Tritium 6
2.2.4 Activated Corrosion Products 6
2.2,5 Cladding Activation 7
2.2,6 Sodium Activation. 7
2.2.7 Miscellaneous Activation Products 8
2.2.8 Tramp Fuel 8
2.3 Transport of Fission Products and Fuel from
Failed Pins 8
2.4 Radioactivity in the Sodium System 10
2.5 Gaseous Radwaste Management 11
2.6 Liquid and Solid Radwaste Management at EBR-II.... n
3. PLUTONIUM AND OTHER TRANSURANIUM ELEMENTS 14
3.1 Plutonium Inventories 15
3.2 Isotopic Composition of Plutonium 17
3.3 The Higher Actinides... 17
4. FISSION PRODUCT GENERATION 25
4.1 LMFBR Fission Product Generation 25
4.2 Comparison with LWR Fission Product Generation 33
5. OTHER SOURCES 45
5.1 Tritium and Its Transport 45
5.1.1 Summary. 45
5.1.2 Sources. 46
5.1.2.1 Ternary Fission 46
5.1.2.2 Boron Carbide Control Rods 48
5.1.2.3 Lithium Contamination 53
5.1.3 Transport of Tritium in an LMFBR System 54
5.1.3.1 Escape into Sodium System 54
5.1.3.2 Transport in the Sodium and Steam
Systems (including EBR-II
Experi ence) 55
5.2 Activated Corrosion Products 61
5.2.1 Estimated Corrosion Product Activity in
1000 MWe LMFBR 63
5.2,2 Distribution of Corrosion Products in the
Primary System 69
5.2.3 Calculational Method 70
5.2.4 Corrosion of ^2ya from Tantalum Control
Rods 72
5.2,5 Activated Corrosion Product Experience at
Operating Sodium-Cooled Reactors 73
vn
-------
TABLE OF CONTENTS
(Continued - Page 2)
Page
5.2.5.1 Summary 73
5.2.5.2 EBR-II 75
5.2.5.3 Rapsodie 75
5.2.5.4 SRE 75
5.2.5.5 SEFOR 76
5.3 Activation Products 78
5.3.1 Sodium Activation 78
5.3.1.1 Sodium-24 78
5.3.1.2 Sodium-22 79
5.3.2 Cladding Activation 85
5.3.3 Activation Products 39Afj 41Ar and 23Ne 87
5.3.4 Miscellaneous Activation 89
5.3.4.1 From Fission Products 89
5.3.4.2 From Impurities in Sodium Systems. 89
5.4 Tramp Fuel 90
5.4.1 SEFOR 90
5.4.2 EBR-II..... 90
5.4.3 Rapsodie 90
5.4.4 Extrapolation to 1000 MWe LMFBR 91
TRANSPORT OF FISSION PRODUCTS FROM FAILED FUEL 92
6.1 Introduction 92
6.2 Brief Background Description of Irradation
Experience Relating to Fission Product Release.... 93
6.3 Tritium Release from Fuel Pins (see also
Section 5.1) 95
6.4 Release Fractions for Noble Gases from Oxide
Fuel s 95
6.5 Fuel Failure Rates.. 101
6.6 Leakage of Fission Products from Failed Fuel--
Gaseous and Solid 102
6.6.1 Escape Rates from Plenum to Sodium 102
6.6.2 Transit Time from Failure to Cover Gas 103
6.6.3 Time for Diffusion out of the Fuel 104
6.6.4 Diffusion Direction 104
6.6.5 Experimental Data on Transport of
Specific Fission Products to the Sodium
or NaK Coolant 105
6.6.6 Theoretical Models for Fission Product
Transport 107
6.7 Vented Fuel... 107
6.8 Example Calculations of Releases of a Few Selected
Radionuclides to the Primary Coolant and the
Cover Gas System 110
FISSION PRODUCTS IN SODIUM SYSTEMS 123
7.1 Fission Product Behavior in Sodium. 123
7.1.1 Behavior of Each Fission-Product Type 124
7.1.1.1 Noble Gases 124
7,1.1.2 Iodine 125
7.1.1.3 Alkali Metals (e.g. Cesium) 125
-------
TABLE OF CONTENTS
(Continued - Page 3)
Page
7.1.1.4 Alkaline-Earth Metals (e.g.
Strontium) 126
7.1.1.5 Rare Earths 126
7.1.1.6 Transition Metals 127
7.1.1.7 Noble Metals 127
7.1.1.8 Tritium (see Section 5.1.3) 127
7.1,2 Operating Experience with Fission Products
in Sodium (or NaK) Cooled Reactors
(Excluding Experience with Cold Traps) 127
7.1.2.1 Summary..... 127
7.1.2.2 EBR-II 127
7.1.2.3 BR-5 130
7.1.2.4 Dounreay 130
7.1.2.5 Rapsodie 132
7.1.2.6 SRE 132
7.1.2.7 SEFOR 136
7.2 Cold Traps 136
7.2.1 Brief Description of Cold Trap Technology.. 136
7.2.2 Cold Trap Decontamination Terminology.. 139
7.2.3 Experiments on Cold Trapping of Particular
Radionuclides 142
7.2.3.1 Cesium 142
7.2.3.2 Iodine......... 144
7.2.3.3 Strontium, Barium and Zirconium... 145
7.2.3.4 Tritium (see Section 5.1.3.2) 145
7.2.4 Operating Experience on Cold Trapping of
Fission Products at Sodium-Cooled Reactors. 145
7.2.4.1 Summary 145
7.2.4.2 EBR-II 146
7.2.4.3. BR-5 146
7.2.4.4 SRE 146
7.2.4.5 SEFOR and Fermi (see also
Section 7.2.1) 148
8. GASEOUS RADWASTE MANAGEMENT. 151
8.1 FFTF Gaseous Radwaste Systems.. 151
8.1.1 Primary Sodium System Seals 152
8.1.2 Radioactive Argon Processing System (RAPS). 153
8.1.3 Cell Atmosphere Processing System (CAPS)... 156
8.2 EBR-II Gaseous Radwaste Systems 156
8.2.1 Present Operation 156
8.2.1.1 Normal Operation 158
8.2.1.2 Fast Gas Purge System 158
8.2.2 Proposed Gas Radwaste System 158
8.2.2.1 Criteria 159
8.2.2.2 Cover Gas Cleanup System 160
8.2.2.3 Seven-Day-Delay System 160
8,2.2.4 24-Hour-Delay System 162
IX
-------
TABLE OF CONTENTS
(Continued - Page 4)
Page
8.3 Gaseous Radwaste Experience in Other Operating Fast
Reactors 162
8.3.1 Fermi 162
8.3.2 SEFOR 165
8.3.3 Rapsodie 165
8.3.4 Dounreay 165
8.4 Comparison of LWR and LMFBR Radgas Effluents 165
8.4.1 LWR Gaseous Releases 166
8.4.1.1 ORNL Study 166
8.4.1.2 USAEC Regulatory Study 168
8.4.2 Comparison of LWR and LMFBR Radioactive
Gas Rel eases 168
9. LIQUID AND SOLID RADWASTE MANAGEMENT AT EBR-II 172
9.1 Liquid Radwaste System 172
9.2 Solid Radwaste Management 173
APPENDICES
A. Environmental Operating Data for Fermi, SEFOR, and
EBR-II 177
B. Fission Product Data 224
-------
TABLES
Table Page
2.1 Radionuclides Observed in the Primary Coolant 12
System of Sodium-Cooled Reactors
3.1 1000 MWe Reactor Charges, Discharges, and 16
Inventories of Plutonium
3.2 Isotopic Composition of Plutonium in 18
Discharged Fuels (wt %)
3.3 Average Annual Amounts and Activities of 19
Selected Actinides Discharged from Reactors
3.4 Neutron and Alpha Particle Yields from 22
Selected Actinides and Their Compounds
4.1 Fuel Mass and Fission Product Activity 26
Discharged Annually from an LMFBR and
Fission Product Power of Discharged Fuel
4.2 Operating Conditions for Two 1000 MWe Designs 27
4.3 Fission Product Activity of Core Discharge Fuel 28
from AI 1000 MWe Reference Oxide Design (80,000
MWd/MT Exposure) as a Funciton of Cooling Time
4.4 Fission Product Activity of Axial Blanket 31
Discharge from AI 1000 MWe Reference Oxide
Design, as a Function of Cooling Time
4.5 Fission Product Activity of Radial Blanket Discharge 33
from AI 1000 MWe Reference Oxide Design, as a Function
of Cooling Time
4.6 Fission Product Activity of Core Discharge Fuel 35
from GE 1000 MWe Follow-on Design (100,000 MWd/MT
Exposure), as a Function of Cooling Time
4.7 Energy Generation Rate from Fission Product Decay 36
for the AI 1000 MWe Reference Oxide Design as a
Function of Cooling Time
/
4.8 Total Activity of Noble Gas and Iodine Nuclides 37
During Operation of a 1000 MWe LMFBR
4.9 LWR Fission Product Activities as a Function of 39
Cooling Time
-------
TABLES (Cont'd)
Table Page
4.10 Fission Product Activity Transported Annually 42
From a 1000 MWe LMFBR
4.11 Yields of Selected Fission Products from 43
Thermal and Fast Fission
5.1 Estimated Tritium Production Rates in a 46
1000 MWe LMFBR
5.2 Summary of EBR-II Tritium Data 57
5.3 Tritium Concentrations in EBR-II 58
5.4 Activation Reactions in Stainless Steel 62
5.5 Data for Corrosion Product Calculation 63
5.6 Estimates of Activated Corrosion Products in the 64
Primary System of the 1000 MWe LMFBR After 30
Years Operation
5.7 Corrosion Product Neutron Cross Sections 67
5.8 Comparison of HEDL, AI, and ORNL Cross Sections 68
Averaged Over an LMFBR Core Energy Spectrum
5.9 316 Stainless Steel Composition and Isotopic 69
Abundance
5.10 Fraction of Nuclides Deposited in Primary 70
System Components - HEDL and 6E Results
5.11 Corrosion Products in Sodium-Cooled Reactors 74
5.12 Weight Percent of Impurities in SEFOR Cold Traps 76
5.13 Cladding Activity of Spent LMFBR Core Fuel 86
as a Function of Cooling Time
5.14 Cladding Activity Discharged Annually From 86
1000 MWe LMFBR
6.1 Oxide Fuel Pins Irradiated in LMFBR1s 96
6.2 General Electric F2 Series of Fuel Pin 98
Irradiations
6.3 ANL Irradiations in EBR-II 99
-------
TABLES (Cont'd)
Table Page
6.4 Percent of Fission Gas Released vs Fuel Burnup TOO
6.5 Leaching Results from Grossly Defected Oxide Pin 106
in NaK
6.6 Loss of Fission Products from Grossly Defected 108
Oxide Fuel in Flowing Sodium
6.7 Isotopic Release Fractions from GE Vented Fuel 111
Test
6.8 Example Equilibrium Cover-Gas Activity from 114
Failed Fuel for Various Delay Times after
Birth for Gaseous Radionuclides
6.9 Calculated Annual Activities of Long-Lived 115
Radionuclides Entering Primary Sodium from
Failed Fuel
7.1 Fission Products Observed in the Primary 128
System of Sodium and NaK Cooled Reactors (other
than Tritium)
7.2 Fission Product Activity in BR-5 During 131
First Stage of Operation (1962-64)
7.3 Gamma Activity of Fission Products in DFR 131
Coolant, 6 Days After Sampling
7.4 Typical Radioactivity Levels of SRE Primary 133
Sodium Prior to Run 14
7.5 Initial Fission Product Analysis of SRE 134
Primary Sodium After Run 14
7.6 Fission Product Analysis of SRE Primary 135
Sodium as a Function of Time After Run 14
7.7 Example of SRE Primary Pipe Wall Fission- 137
Product Contamination from HC1 Etch at Pipe
Surface
7.8 Fission Products Observed in Primary Cold 145
Traps of Sodium or NaK Cooled Reactors
7.9 Comparison of Impurity Levels in SRE Cold 147
Trap to those in Sodium Coolant
8.1 Radioactive Argon Processing System Delay
Times
xiii
-------
TABLES (Cont'd)
Table
8.2 Xenon and Krypton Conditions in Delay Beds
8.3 Typical Annual Gaseous Releases from a 1000
MWe PWR Operating with 0.25% Defective Fuel
8.4 Summary of Variables for PWR Gaseous Radwaste
Treatment Systems
8.5 Estimated Annual Releases of Radioactive
Gaseous Effluents from 1000 MWe PWR with
0.25% Defective Fuel
9.1 EBR-II Solid Waste Management
Page
163
167
169
170
175
A.I
through Environmental Operating Data for Fermi,
A.31 SEFOR, and EBR-II
B.I Half Lives and Fission Yields of Fission
Products Listed in Section 4
177
225
xiv
-------
FIGURES
Figure
3.1 Formation Scheme for Important Actinides 20
5.1 Tritium in Primary Sodium (EBR-II) 56
8.1 Radioactive Argon Processing System 154
8.2 Cell Atmosphere Processing System 157
8.3 Seven-Day-Delay Cover Gas Cleanup System 161
for EBR-II
8.4 24-Hour-Delay Cover Gas Cleanup System 164
for EBR-II
9.1 EBR-II Liquid Radwaste System 174
B.I Fission Product Decay Schemes Used for 228
Calculations in Section 4
xv
-------
Page Intentionally Blank
xvi
-------
1. INTRODUCTION
••' Objectives of the Present Study
Radioactivity produced in a Liquid Metal Fast Breeder Reactor
(LMFBR) must either decay or ultimately leave the reactor site (or
remain at the site after decommissioning the plant.) The purpose
of this study is to examine all sources of radioactivity in an
LMFBR power reactor and to determine the ultimate fate of this
activity during the normal operation of the plant. This investigation
is concerned with the quantity and form of the radioactivity that
leaves the site. The "environment" is defined as anything beyond
the site boundary, so all radioactivity leaving the site enters the
"environment". Nearly all of the radioactivity leaving will be
contained; but then this radioactivity becomes the source at the
next stage in an environmental study. An attempt is made to
identify the small amounts of radioactivity that leave the site un-
contained and enter the atmosphere and water directly.
It should De emphasized that the study is limited to normal
operation and therefore does not include accident situations. Normal
operation does assume operation with some failed fuel, however.
The study also includes numerous comparisons between the
operation of an LMFBR and a light water reactor (LWR).
Numerical results are based on a 1000 MWe plant. For the
LMFBR, and efficiency of 40% was assumed, so that the thermal power
was 2500 MW(th). For the LWR an efficiency of 34% was assumed, which
gives a thermal power of 2940 MW(th).
Environmental statements for the LMFBR Demonstration Plant
(WASH-1509) and for the FFTF (WASH-1510) were published in April, 1972
and May, 1972. The statement for the Demonstration Plant contains
a general description of the LMFBR program, including history, the
projected U. S. program, and European, USSR, and Japanese programs.
Tnerefore, this type of discussion is not presented in the present
report.
A study related to parts of the present one, and which also
included a review of accidents, was reported in March, 1969 by
ORNL (G. W. Keilholtz and G. C. Battle, Jr., "Fission Product Release
and Transport in Liquid Metal Fast Breeder Reactors," ORNL-NSIC-37,
March, 1969). Since that time much has beeto repoTted-and further
experience with several Fast reactors has been obtained; this new
material in addition to the old is included in the present report.
The present study i icludes a review of operating experience
with liquid metal cooled fast reactors and also some data from
sodium cooled thermal reactors. The reactorsi reviewed include:
-------
U.S. Fast Reactors:
Experimental Breeder Reactor-II (EBR-II)
Enrico Fermi Fast Reactor (Fermi)
Southwest Experimental Fast Oxide Reactor (SEFOR)
U. S. Thermal Reactors:
Sodium Reactor Experiment (SRE)
Hal lam Nuclear Power Facility (HNPF)
Snap-8 Experimental Facility (S8ER)
UK Fast Reactor:
Dounreay Fast Reactor (DFR)
French fast Reactor:
Rapsodie
USSR Fast Reactor:
BR-5
All of these reactors were sodium cooled except Dounreay and
S8ER5 which were cooled by NaK. It was more useful to treat various
results from the operation of these reactors throughout the report
under the different functional sections rather than to devote a
separate overall section to operating reactor experience.
1.2 Some General Results of the Study
Despite the vast amount of material reviewed and discussed in the
present report it is clear that there is still much to learn
concerning transport of the radionuclides produced in an LMFBR. The
quantities of the radionuclides produced in an LMFBR, which include
the Isotopes of plutonium and other transuranium elements (Section 3)
ana the contained fission products (Section 4) are fairly accurate
and well defined. Their disposition and safeguarding are of
concern with respect to the reprocessing and fuel fabrication
plants, transportation, and the waste disposal systems. The areas
which are still poorly defined concern fuel failure during normal
operation (Section 6), the transport of fission products from failed
fuel (Section 6), and the fate of the non-noble gas fission
products in the sodium (Section 7). The status of knowledge on
tritium and corrosion products (Section 5} is better than that of
the above problems, but still is not adequate. Only long-term
operation of power LMFBR1s (for example, the Clinch River Breeder
Reactor Plant, the Fast Flux Test Facility, and the currently
operating demonstration plants in the U. K., France and the USSR)
wil'i provide experience In these problem areas to replace the early
estimates reported here.
If an efficient caseous radwaste system is used on all fast
power reactors, as proposed for the FFTF, it is expected that during
normal operation the gaseous effluent released to the environment
(i.e. to the atmosphere at the reactor site) can be made as low
as required. In a sodium reactor leakage of coolant to the environment
cannot be permitted during normal operation. The only liquid
waste (other than tritiated water) will come from experiments1 ana
cleanup facilities, which can be made to contain little waste if
-------
necessary as in :ne case of EBR-II (Section 9). The only major source
of radioactivity released to the atmosphere and water at the plant
site which is difficult to eliminate is tritium, but even here the
cold traps in an LMFBR appear to provide a reduce ion below e-fluent
levels typical for light water reactors. Based en EBR-II ex-
perience (which is a very limited basis for extrapolation tc a large
power reactor) and the estimated tritium sources in a large olant,
the annual tritium release rate to the surrounding atmosphere from
1000 MWe LMFBR is of the order of 100 Ci/yr and 'is less than
10 Ci/yr to the condenser water (Section 5.1). This liquid effluent
rate compares to predicted tritium releases of 100 Ci/yr and 600 Ci/yr
in the liquid effluent of a 1000 MWe BWR and PWR respectively
(Section 5.1)
For these reasons, it appears that the principal environmental
problems for normal operation of an LMFBR will involve offsite
handling of the larger amounts of plutonium and transuranium
elements and fission products shipped away from the site in the
irradiated fuel, storage of the 85Kr, storage of the cold traps
shipped away from the plant, and storage of sodium and primary
system equipment after decommissioning of the power plant. The
problems concerning sodium and sodium cold trap disposal do not
exist, of course, for water reactors. The quantities of plutonium
and the decreased 240pu jn radial blanket material lead to a
worsening of the plutonium problem for the LMFBR over the "Pu
Recycle LWR", but not greatly so (Section 3). The other environmental
problems from normal operation of the LMFBR (e.g. fuel reprocessing,
transportation, long-term storage of fission products) are similar
to those of a light water reactor.
2. SUMMARY
2.1 Objectives and Methods
The purpose of this investigation was to examine all sources of
radioactivity in an LMFBR power reactor and to determine the ultimate
fate of this radioactivity. Only normal operation was considered.
There was particular concern with the form and quantity of radio-
activity which leaves the reactor site; this radioactivity was
considered to enter the "environment", meaning that this activity
must be dealt with at a reprocessing plant, a storage facility, or
elsewhere.
The method of the study was predominantly to obtain numbers
from published values in the literature. These were augmented by
some original estimates where needed. Operating experience of liquid
metal cooled reactors w,is reviewed, including EBR-II, Fermi, SEFOR,
Dounreay, Rapsodie, BR-5, SRE, S8ER, and Hallam. Numerical results
-------
are presented on the basis* of the normal operation of a 1000 MWe
LMFBR.
2.2 Sources
Radionuclides which are transported to the environment from
normal operation of an LMFBR include:
• plutonium and other transuranium elements
• fission products
•tritium
• corrosion products
•activation products (24Na, 22Na, 39A, 41A, 23Ne, cladding)
f tramp fuel
Transport to the environment refers to any transport from the
reactor site. Planned transport paths include shipments from the plant
of irradiated fuel, cold traps or other used primary equipment,
bottled gas from the gaseous radwaste system, wastes in water
solutions, solid wastes, and sodium and primary equipment after
decommissioning of the plant. Unplanned leakage to the environment
includes gas leakage through the building ventilation system and
tritium leakage through the secondary sodium system to the steam
systems and thence to the condenser water.
2.2.1 Plutonium and Other Transuranium Elements
Depending on the specific design, a 1000 MWe LMFBR will have
from 1100 to 1850 kg of plutonium loaded into it annually, and from
1540 to 2000 kg of plutonium will be unloaded annually. The maximum
plutonium inventory will be on the order of 3000 to 3300 kg. The
typical isotopic content of the discharged plutonium for the mixed
blankets and core is shown below.
Wt. % of
Isotope Total Pu
238Pu 1
239Pu 69
22
5
3
* In general, extrapolation of data to correspond to the 1000
MWe size LMFBR has been done by assuming proportionality to reactor
thermal power. Most parameters do chenge linearly in such a scale-up,
and the assumption of linearity is generally a good approximation.
However, there are some important parameters which do not change
linearly, and caution should be exercised in making any such
extrapolations.
-------
For comparison, a 1000 MWe LWR will charge 0 to 730 kg of
plutonium annually, will discharge 250 to 4)0 kg annually, and will
have maximum plutonium inventories on the o~der of 500 to 2000
kg, depending on whether they are based on -35y fue] or p]utonium
recycle.
The amounts and activities of uranium and the various transuranium
elements discharged annually from core and Blankets of a 1000 MWe
LMFBR are shown below for a plutonium discharge of 2000 kg.
Element
U
Np
Pu
Am
Cm
At
kg
2x1 04
8
2000
17
1
Discharge
Ci
1.2x109
1.5x10'
3.4x10°
l.SxlO6
Ci after
90 d
7 3
l.SxlO,
1.5x10'
4.6x10^
1.2xlOb
2.2.2 Fission Products
The important fission products still in the fuel one year after
the fuel is removed from an LMFBR (i.e. those of activities still
greater than 104 Ci/year in the fuel discharged from a 1000 MWe
plant) are
85
89
90
91
95
103
Kr
Sr
Sr-Y
Y
Zr-Nb
Ru-Rh
Ru-Rh
137
l>}/
Cs-Ba
141
144
147
151
155
162;
Ce
Ce-Pr
Pm
Sm
Eu
Gd-Tb
The total fission product activity in the fuel discharged
annually from a typical 1000 MWe LMFBR, together with the associated
fission product power, are shown below as a function of cooling time
after discharge.
Cooling Time
Activity (Ci/yr)
Power (MW)
30d
a
4.3x10°
1.9
90d
o
2.4x10°
1.0
150d
R
1.6xlOH
\ 0,7
300d
7
8. 6x1 0'
0.4
30yr
c
3.4x10°
0.01
-------
2.2.3 Tritium
Tritium is produced in ternary fission,.but in an LMFBR nearly
all of this tritium escapes from the fuel. Tritium is also produced
in boron carbide control rods, but it is unclear how much of this
escapes to the sodium. The annual estimated tritium production rates
are summarized as follows:
Source Annual Production Rate
(Ci/yr)
Ternary fissionn 20,000
Control rods: uB(n,t) 2a 7,000
7Li(n,nt)a 2,500
Lithium contamination: Li(n,t)a
In fuel (maximum) 4,000
In sodium 100
TOTAL ^ 30,000
Tritium leaks to the environment, both as a gas to the atmosphere
and as a liquid in the condenser water. Based on direct extrapola-
tion from EBR-II (60 MWth) to a 1000 MWe LMFBR, the annual leakage
rates would be 'v 60 Ci/yr of tritium to the atmosphere and ^ 3 Ci/
yr to the condenser water. This value compares to BWR and PWR
liquid effluent rates of ^ 100 Ci/yr and ^ 600 Ci/yr, respectively.
Most of the tritium in an LMFBR is held up in cold traps, and is
eventually shipped from the site with the cold traps.
2.2.4 Activated Corrosion Products
Activated steel cladding and steel structures in and near the
core are slowly corroded by sodium. The principal activated corrosion
products are 60co, 58c0 and $^Mn. Most of the corrosion products
plate out on the walls of the primary system, while some are held
up in the cold traps and some remain in solution in the sodium.
Estimates of the corrosion products which will enter the sodium and
still be present in the primary sodium system after 30 years of
operation in the 1000 MWe LMFBR, together with the principal reactions
and the half lives, are summarized below.
-------
Nuclide
60r '
Co
58Co
54Mn
59Fe
51Cr
182Ta
Formation
Reaction
59Co(n,Y)
6°Fe(n,p)
58Ni(n,p)
54Fe(n,p)
58Fe(n,Y)
5°Cr(n,Y)
181Ta(n,Y)
r;alf Primary-System Corrosion Product
Life Activity After 30 Years
5.24 yr
71 d
313 d
45 d
28 d
115 d
(cn
19,000
1 ,000
23,000
19,000
1,000
7,000
<6,000
2.2.5 Cladding Activation
The cladding and channel walls are activated and the activity is
shipped to the reprocessing plant with the fuel. Estimates of the
cladding activity shipped from a 1000 MWe LMFBR annually with the
discharged fuel are given below as a function of cooling time:
Cooling Time
30 d 90 d 150 d 300 d 30 y
Activity (Ci/yr) 5xl06 3xl06 2xl06 IxlO6 IxlO3
Two activation products not considered among the activated
corrosion products are 59Ni and 63Ni (half lives of 8 x 104 yr and
92 yr respectively) which contribute 20 Ci and 500 Ci respectively to
the 1000 Ci total at 30 years.
2.2.6 Sodium Activation
Sodium activates to 24^ (by an n^ reaction) and 22^fl (by n,2n).
Sodium-24 has a relatively short half life (15 hr) but 22^a has a 2.6
year half life. At % 8 days after shutdown the 22^ becomes
dominant activity. The equilibrium activities of the primary system
of a 1000 MWe LMFBR are estimated at 2 x 107 Ci for 24Na and 3000 Ci
for 22Na.
-------
2.2.7 Miscellaneous Activation Products
Argon-39 is produced by activation of 39K in sodium and is
significant because of its long half-life (269 yr). Although no
reported observations of 39A were found for operating fast reactors,
the calculated 39/\ production rate for a 1000 MWe LMFBR (assuming
300 ppm potassium in the coolant) is ^30 Ci/yr.
Small amounts of the following activation products are found in
operating fast reactors:
, 23Ne -- Gaseous activation products always present in LMFBR's.
65Zn, Ag, 125$b, 210po -- Activation products observed in
some fast reactors.
Also, ^34cs>154 £Uj and several other isotopes listed in this
report under fission products are actually activation products
produced from activation of fission products.
2.2.8 Tramp Fuel
Tramp fuel will probably not be a concern in large LMFBR's.
Total tramp fuel inventories of the order of half a gram of heavy
metal atoms can be predicted based on other reactor experience.
This results in an equilibrium fission product inventory in the
primary circuit of about 300 Ci of fission products and 20 to 30
Ci of actinide activities. The long-lived isotopic buildup would be
to a few tens of curies of fission products and a few curies of
long-lived actinides over the plant life.
2.3 Transport of Fission Products and Fuel from Failed Pins
Estimation of the transport of activity from failed fuel pins
to the primary circuit is an extremely difficult problem. On the basis
of limited experimental results and operating experience, the following
conservative assumptions were used to calculate releases.
1% failed pins
90% of failed pins are leakers
10% of failed pins have gross cladding failures
75% of the noble fission gases are released from the fuel proper,
i.e., pellets, of failed pins
For gross failures, the escape fractions assumed are
Fuel 1%
Br, I, Cs 15%
Te, Ru, Tc, Mo 5%
All Others 1%
The plutonium fraction in the escaped fuel is assumed to be
-------
only one-tenth of the original Pu fraction because of the observed
preferential leaching of the U and the inward migration of Pu.
The releases calculated from the above assumptions can easily be
adjusted to other failure or release fractions if experience or
judgement dictates.
Annual releases of a few selected fission products and plutonium,
to the primary sodium or cover gas, together with the total released
activity still present one year after a 30 year operating period, are
given below.
Radionuclide
85Kr
90Sr-Y
106Ru-Rh
125Sb
125"Ve
134Cs
137Cs-Ba
144Ce-Pr
147Pm
15V
154Eu
155Eu
241Pu(3)
Pu(a)
Pu
Annual Release
(Ci)
1900
30
4600
—
—
—
1100
900
100
—
—
—
20
0.6
2.5 grams
Activity Present One
Year After a 30 Year
Operating Period
(CI)
25,000
600
6,000
40
10
350
30,000
800
500
50
4
60
400
20
75 grams
-------
A reduction in failed pins from 1% to 0.1% and a corresponding
reduction in gross cladding failures from 0-1% to 0.01% would reduce
all of the above releases, including the 8^Kr release, by a factor
of ten. If the assumption concerning relatively higher leaching
of uranium than plutonium is correct, this reduction in failure
rates would reduce the annual plutoniuiii release to the same order
of magnitude as the plutonium in tramp fuel ( M).l gm.). Also,
the fraction of fuel leached from gross cladding failures may be
significantly less than 1% in most cases, further lowering the
Pu release.
2.4 Radioactivity in the Sodium System
The primary sodium system of an LMFBR becomes highly radioactive.
Fission products and small quantities of fuel enter the sodium from
failed fuel pins, tritium and activated corrosion products enter the
sodium, sodium becomes activated to ^Ha and 22Na, and small amounts
of other activation products and tramp fuel are present. Sodium
leakage from the primary system (other than very small leakage to
the secondary sodium) cannot be tolerated. Hence, coolant leakage
is not a source of fission products in the environment as is the
case with LWR's. Activated sodium causes maintenance problems, but
only to plant personnel. An important environmental concern is the
periodic removal from the site of cold traps which contain radioactive
material. Also, after decommissioning of the reactor, the radioactive
sodium and contaminated primary system components must be removed
from the site and stored somewhere in the environment.
Many experimental studies on fission product behavior in
sodium have been reported, some of which are reviewed in this report.
One of the principal purposes for those studies is to determine the
activity in the sodium in the case of an accidental sodium fire or
sodium release, and these considerations are beyond the scope of
this report.
Cold traps are used in LMFBR's to purify the sodium. Although
the primary function of the cold trap is to maintain a low oxygen
concentration, cold traps also remove much of the radioactivity
from sodium during normal operation. These cold traps must be removed
periodically, shipped from the site, and stored somewhere in the
environment. Iodine and tritium are effectively removed from sodium
by cold traps. Much of the cesium is removed, but according to
EBR-II experience cold traps are not adequate for cesium removal.
At low temperatures, much of the cesium will plate out on the walls
of the primary system. Also, most activated corrosion products
plate out on the primary system. Nobel gases are volatile in
liquid sodium, except for small amounts that remain in solution, and
quickly escape to the cover gas.
A summary table of experience with the sodium (or NaK) system
of operating sodium (or NaK) cooled reactors is given on the next two
10
-------
pages. The table only lists radioactive isotopes that have been
observed. Further details on levels of activity are given in
Section 7 of this report.
2.5 Gaseous Radwaste Management
Information on the gaseous radwaste systems of FFTF, EBR-II,
Fermi, SEFOR, Rapsodie, and Dounreay was reviewed.
The amount of gaseous activity released from FFTF is expected
to be trivial as a result of two factors: (a) a sophisticated
gaseous radwaste system is used, which includes charcoal delay beds
and a cryogenic distillation column, and (b) virtually no coolant
leakage is permitted from a liquid metal cooled reactor. In this
system, 85}(r will be bottled and shipped off site for storage.
Assuming the same type of gaseous radwaste system on an
LMFBR power plant, the °$Kr release to the atmosphere from a 1000
MWe LMFBR would be 3 Ci/year, based on scaleup of the FFTF projected
values and assuming operation with ]% failed fuel. In addition the
30 Ci/yr of A (produced from 300 ppm potassium in the coolant)
would all be released to the environment.
EBR-II has a gaseous waste system that does not presently allow
operation with failed fuel. A proposed system using charcoal
delay beds and 85«r storage, will allow operation with up to twelve
failed test oxide fuel pins.
For comparison, short-lived noble gas and 85Kr leakage to the
atmosphere is much higher for typical LWR's than for an LMFBR with an
FFTF-type gaseous radwaste system. For a typical LWR with 0.25%
defective fuel the 85
-------
Table 2.1
Radionuclides Observed in the Primary Coolant System of Sodium (and NaK) Cooli\i Uc-ictors (otter
than ''"Na -
,__
-*~
Observed in
Primarv
Coolant
Observed or.
Primary
System
Surfaces
,.-•"""'"''
Observed in
Cold
Traps
-,
FfSsipji Products
'-•-.,,_
'**--.. _
Activated
Corrosion Products
Other Activat_Lor.
Product.-
Fission Products
Activated
Corrosion Products
^.-••etfer Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fermi
i'-Ba-La, '• ; :Cs
°9Sr, '-"-I
"'"••>,,
^-^
"•-
10jRu, 9-Zr-Nb
...<*•'"'
BR-5
IUI*Ce, '"'Ce, p'''Pr,
lll0Ba-La, :3';Cs,
'16Cs, n''Ru, '<0Sr,
"Zr-Nb, '-I
-.. 65,
"•--:"
\.,
•~'.~.
^^Ce, • ''Ce, -'''tPr,
• "'"-:Cs, ;'-Uu, 9ISr,
"zr-Nb, :-'l
!"7Cs, !36Cs, :3:,
133I, 95Zr-Nb,
135If l^Ba-La
EBR-II
''"CS, "'H
-IT.
6=Zn, :--Sb, ..-••*'- Sb, • "^.-:,
• ... r.'-^ni • Trv>..
,•••• '--Po,
,
5-Mn, 6-Co
65Zn, :!;Ta
' r:Cs, : -^Cs
:K, I
'^Mn, 60Co
55Zn, I2"Sb
DO'unreay
l"ilCe, ^"Ce,. : 32Te
13!I, 105Ru, :';-Ru
132 1, 137Cs, ";r;Zrftt)
ll<0Ba-La, '3!Cs
. , _
; ; "Cs-,
"""•-'...
Rapsodic
;''7Cs
,,."'-"""
ii',, ' • •• •
-I, --^-L..
^Zr-Nr-, a;V,
«Sr-Y
-"Mn, 58Co,
-Co
:-'~Zn
-"Mr., 58Co,
^Co
"Zn
-------
Table 2.1
Radionuclides Observed in the Primary Coolant System of Sodium (and NaK) Cooled Reactors (other than 2l*Na and 22Na)
no
Observed in
Primary
Coolant
Observed on
.Primary
System
Sin-faces
Observed in
Cold
Traps
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fermi
V^Ba-La, 137Cs
89ffl, 13»I
l"lCe, ""Co, 133I,
103Ru, 95Zr-Nb
BR-5
l^Ce/ lklCe> l^pr/
140Ba-La, 137CS,
136cs, 106Ru, 90sr,
95Zr-Nb, 131I
&52n
i^Ce, ^Ce, "'•Pr,
1It0Ba-La, 137Cs,
136Cs, 106Ru, 9°Sr,
95Zr-Nb, 131I
137Cg, 136CSf 131..
133j, SSzr-jjb,
135j, l^Ba-La
EBR-II
13I+Cs, 3H
51*Mn
65Zn, iz^sb, 125Sb, llonkg,
iJ3Sn, 113mln, 117mSn,
210Po,
i37Cs
5"Mn, 50Co
65Zn, I82Ta
l37Cs, 13"Cs
3Hr I
"Mn, 6°Co
65Zn, I2"sb
Dounreay
141Ce, ^"Ce, 132Te
I3llf 103RU/ 106RU
132I, 137Cs, 95Zr-Nb
ll<0Ba-La, 138Cs
""Ba-La
137Cs
Rapsodie
137Cs
210Po
i-iCe, 137Cs,.131I,
132If l^OBa-^,
95Zr-Nb, 91Y,
90Sr-Y
SI*Mnf 58Co,
6°Co
65Zn
51*Mn, 58Co,
60Co
65Zn
-------
-------
Table 2.1. (continued - page 2)
Radionuclides Observed in the Primary Coolant System of Sodium (and NaK) Cooled Reactors
Observed in
Primary
Coolant
Observed on
Primary
System
Surfaces
Observed in
Cold
Traps
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
SEPOR
86Rb
6°Co
-65Zn, iz^sb
110Ag
(Cu, Fe, Cr,
Ni, Mn)
SRE
131*Cs, 137Cs, 89Sr,
9°Sr, «il, i«Ce,
""Ce, 103Ru, 106Ru,
95Zr-Nb ""Ba-La
51Cr, 5"Mn,
59Fe, 60Co
125Sb
89Sr, 9°Sr, i^Ce,
95Zr-Nb, J06Ru
i37Cs
51Ci, 51*Mn,
59Fe, 60Co
i37Cs, 106Ru,
"^Ce-Pr, n°flg
"Mn, 59Fe,
6°Co,
I25Sb
S8ER
55Mn, 60Co
89Sr, 90Sr, iOiRi,
106Ru-Rh, 1I(1Ce,
11|lfCe, 95Zr-Nb
llt0Ba-La
54Mn, 59Fe,
58Co, 6°Co
51*Mn, 59Fe
Hallam
-
51*Mn, 50Co
5"Mn, 6°Co
-------
Page Intentionally Blank
-------
T.iMo ,\ 1 (oontimicJ - page 2)
Radionuclutes Observed in the Primary Coolant Syste-n of Sodium (and UaK) Cooled Reactors
Opse^rved ir.
Primary ~- ,
Coola-it
Observed on
Prirtarv
Systerp
Surfaces
Observed in
Cold
Traps
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
Fission Products
Activated
Corrosion Products
Other Activation
Products
SHTR
?tRb
= ^Co
'•-Zn, • -Sb
;.;:A,'
(Cu, Pe, Cr,
Ni, Mn)
SRE
lj"Cs, ~~-~Cs, ^Sr,
"Sr, 131I, !"!Ce,
:;;'*Ce, !D'RU, :of'Ru,
••-"""r-Vb 1(f:Ba-Uj
-JCr, c"Mn,
59Fe, "Co
1?5Sb
a9Sr, 93sr, '"-Ce,
95Zr-Nb, 106Ru
! = 7Cs
;!Ci, ---Mn,"
5%e, "Co
137Cs, 106Ru,
""*Ce-Pr, !!0Ag "
5"Mn, 59Fe,
6 "CO,
125Sb
S8ER
?°Mn, --'Co
"Sr, 5cSr, 1;3Ru,
;06Ru-Rh, !"]Ce,
"-"-, 95Zr-Nb
: " :Ba-La
5 'Co, =:Co
5<*Mn, 59Fe
Hallatn
S
z^Mv ' C^-
5l-Mn, -60Co
-------
Both solid and liquid waste quantities from EBR-II are larger
than would be expected from an LMFBR power reactor due to the
extensive use of hot cells for experiments there.
3. PLUTONIUM AND OTHER TRANSURANIUM ELEMENTS
The LMFBR will have larger inventories of plutonium and other
higher actinides than are found in enriched-uraniurn-fueled light
water reactors (LWR's). In comparing LMFBR's and LWR's, it is also
interesting to consider LWR's fueled with recycled plutonium,
because plutonium recycle has already become a reality with the
Big Rock Reactor. All three types of fuql-reactor combinations
are considered in this section in the predictions of transuranium
element production. ''
There are important environmental concerns raised by the
increased plutonium inventories, forecast for planned LMFBR's. -,
These concerns are related to the extremely toxic nature of plutonium ,
the possible diversion of plutonium for clandestine weapons
production2, and the increased production in high-plutoniurn-content
fuels of other transuranium elements in the form of extremely high
specific activity nuclides.
The general conclusions of this section, which address most of
the concerns raised above are as follows:
(1) The average plutonium inventory in an LMFBR, while signifi-
cantly greater than that in a uranium-fueled LWR of the
same size, will be less than an order of magnitude greater
than in the LWR.
(2) The average plutonium inventory in an LWR fueled with
recycled plutonium will be about half the inventory in
an LMFBR, thus posing similar toxicity control problems.
(3) Plutonium derived from reprocessed fuel either from
uranium-fueled LWR's or from LMFBR's can be used to
construct an explosive nuclear weapon. The major differences
are that the LMFBR has "blankets" of 238U to increase
plutonium production, the plutom'um from the radial
blanket of an LMFBR would be exceptionally good for weapons,
and the separate diversion of the radial blankets is
possible because of the normal physical segregation of
radial blanket and core.
(4) The overall implication of conclusions (1), (2), and (3)
above is that the plutonium problem, whether with respect
to toxicity control or fissile material safeguards, is
not created by the LMFBR but rather is aggravated by it.
(5) Of the transplutonium isotopes, only Am and Am will be
produced in kilogram quantities each year in a large LMFBR.
14
-------
242 244
However, the much smaller quantities of ^ Cm and Cm produced
will yield higher alpha and neutron activities1than both the
amen"ciurn isotopes and the vastly more abundant plutonium isotopes
combined. (The plutonium will have a higher overall activity
because of the beta decay of 241Pu). Indeed the spontaneous fission
activity from the curium isotopes will be much greater than that of
the plutonium isotopes in discharged LMFBR fuel despite the huge
difference in inventories. Thus the curium isotopes deserve
special attention in LMFBR fuels and pose even more of a concern in
plutonium recycle fuel from LWR's.
3.1 Plutonium Inventories
Calculated plutonium charges, inventories, and discharges for
four reactors are shown in Table 3.1. Results for a pressurized
water reactor (PWR) fueled with uranium and for a PWR fueled with
plutonium are both presented3.
Similar results are presented for two LMFBR conceptual designs:
the 1000 MWe AI Reference Oxide LMFBR4 and the GE 1000 MWe LMFBR5.
The AI LMFBR is considered here because it has been used as the
basis for many literature characterizations of the LMFBR. A design
such as the GE LMFBR is probably closer to that which will be seen
in the first generation of large LMFBR's.
The plutonium which is used in the fuel for the LMFBR's and
for the plutonium-fueled PWR shown in Table 3.1 is plutonium obtained
from uranium-fueled light water reactors. 3»4>5 This would be the
situation for the first several years of a plutonium-fueled reactor
industry because of the available plutonium stocks. By 1985 there
may be as much as $1.7 billion worth of plutonium available from
reprocessing of uranium fuels.6
The fuel reloading schemes for the various reactors in Table
3.1 differ significantly as can be seen from the table. Therefore, the
most important factors for comparison are the maximum plutonium
inventories and the average amounts of plutonium charged and discharged
per year. The larger amounts associated with the LMFBR, as opposed
to the uranium-fueled PWR, are clearly shown. However, just as there
should be serious concern over the 3000 kg of plutonium present in
an LMFBR and the 1500 kg shipped to and from the plant each year,
there must also be appropriate concern over the 500 kg of plutonium
present in a uranium-fueled LWR and the 250 kg shipped from the LWR
plant each year. As stated previously, the plutonium problem
already exists with LWR's and is simply aggravated by the LMFBR.
Note also that the reduction in the amount of plutonium involved
in a plutonium-fueled LWR relative to an LMFBR is not nearly as
dramatic as for the comparison of a uranium-fueled LWR and an LMFBR.
15
-------
1000
•i'.ihlr 3.1
Rcvu-tvr Charges, Discharges, and Inventories of Plutonium
N
X,
Reactor
FWR3 ""•--,,.
(U-Fueled)
PWR3
(Fueled with
Pu fron
0-Fueled PWR)
AI Ref. Oxide Core and
I^FBR4 Axial Blanket
(Fuele.; with
P1-1 fror. Radial Blanket
U-Fueled PWR)
Total
GE JjMFBR^ Core ami. .-»'"
(Fueled with Axial.-Blanket
Pu frcxti
U-Fueled BWR) ,-tedial Blanket
,.--'
Total
F-jol Ave.
fraction Resi- Pu Mtixinvun Ave. Amount of
replaced dence dis- Pu Pu Pu (Kg) ^.--^Ave. Bumuf-
per time chargecl * Charged* Inventory Discharged Charged'' of Core Fue'
.-V.i-oe- (days) (Kg) (Kg) (Ka) per year per"year (MWd/^rT)
'. : li.'O J56 - ••!. 256,--''' — .. ,:,.• .
. 3 " 1200 -142 -!'!0 ,2J:'' 4.5 " '' • -,•.•.':
1. 2 540 1270 ~'i,380 2740 1716 18f5 80,000
.2= -0. ' 223 - \... 56C 3£2 _-_•-
,,•-'"' . 1493 138C . 330C " x 2018 LSb'.
.46 796 1304 1094 ' 2713 1304 "'---._ 1094 100,000
.29 •' 1260 157 356 157
1461 1094 3069 . 1461 1094
* Refers to Pu charged or discharged at an actual refueling. Refueling recurred
annually for the PWR (D-fueled) and GE LMTER reactors so that these rartjers agree
with the average annual anounts for these reactors. Refueling was not annual
for the PWR (Pu-fueled) and AI LMFBR reactors.
-------
Table 3.1
1000 MWe Reactor Charges, Discharges, and Inventories of Plutonium
Reactor
PWR3
(U-Fueled)
PWR3
(Fueled with
Pu from
U-Fueled PWR)
A3. Ref. Ojd.de
LMFBR4
(Fueled with
Pu f ran
U-Fueled PWR)
GE I/4FBR5
(Fueled with
Pu f ran
U-Fueled BWR)
Fuel
fraction
replaced
per
charge
1/3
1/3
Core and
Axial Blanket 1/2
Radial Blanket .28
Total
Core and
Axial Blanket .46
Radial Blanket .29
Total
Ave.
Resi-
dence
time
(days)
1100
1200
540
970
796
'1260.
Pu
dis-
charged*
(Kg)
256
442
1270
223
1493
1304 '
157
1461
Maximum
Pu Pu
Charged * Inventory
(Kg) (Kg)
512
800 2042
1380 2740
560
1380 • 3300
1094 ' 2713
' 356
1094 3069
Ave. Amount of
Pu (Kg)
Discharged Charged
per year per year
256
403 730
1716 1865
302
2018 1865
1304 1094
157
1461 1094
Ave. Burnup-
• of Core Fuel
(MW3/MT)
33,000
33,000
80,000
100,000
* Refers to Pu charged or discharged at an actual refueling. Refueling recurred
annually for the PWR (O-fueled) and GE LMFBR reactors so' that these numbers agree
with the average annual amounts for these reactors. Refueling was not annual
for the PWR (Pu-fueled) and AI IMFBR reactors.
-------
Page Intentionally Blank
-------
3.2 Isotopic Composition of Plutonium
Table 3.2 gives "information on the isotopic composition of
Plutonium in fuels discharged from specific reactor types 3,4,5,7,8,9
and also shows the estimated average composition of plutonium available
for recycle.
The table shows that discharged fuel with the lowest percentage
of fissile plutonium (239pu ancj 24lpu) is that from piutoniurn-fueled
LWR's. This plutonium is much lower in fissile content than that
from either uranium-fueled LWR's or LMFBR's. Next lowest in
fissile content is plutonium from uranium-fueled LWR's. The
plutonium with the highest fissile content (and thus most easily used
in constructing a nuclear explosive) comes from the LMFBR and
in particular from the radial blanket.
It is important to remember, however, that (aside from possible
diversion) high fissile isotopic content is extremely desirable for
reactor fuels. Higher fissile content means better utilization, or
more complete "burning" of the plutonium and it also means an
overall smaller plutonium inventory for reactors designed to use high
fissile plutonium.
Isotopic composition also has a strong effect on the quantity and
character of the radiation from plutonium. A good description of
the contributions made by the various plutonium isotopes to the
activity of interest in fuel manufacturing is provided in Reference 7.
He discusses the neutron doses from spontaneous fission and from (a,n)
reactions with light nuclei. Also the importance of the gamma
activity from the 236pu chain is noted. However, for environmental
considerations, the alpha and beta activity of the various plutonium
isotopes is probably the overriding concern. (All of the plutonium
isotopes of interest are alpha emitters except 241pU) a ^e^a
emitter.) Moreover, the potential for biological damage from reactor
fuels is related not only to the plutonium content but also, to some
extent, to the presence of other higher actinides.
3.3 The Higher Actinides
Although it is theoretically possible to produce elements all the
way up through the highest in the actinide series by successive
neutron absorptions in a reactor, only a few of the higher actinides
are produced in sufficient quantities to be of interest as potential
sources of danger to the environment. Figure 3.1 shows the isotopes
of interest and the principal means of producing them in reactors.
This figure is an elaboration of a figure from Reference 10.
Most of these isotopes will be produced in significant quantities in
both LWR's and LMFBR1s. Calculations and some measurements on the
quantities and activities of the various actinide nuclides
present in reactor fuel during and after exposure have been made.3,4,10
Table 3.3 shows average amounts and activities of several nuclides
of interest that would be discharged each year from various reactor
17
-------
Talkie 3.2
in Disclvirged
1. Uranium-Fueled Reactors
00
'"' "feactor
Bumxy (JVH./HT)
^•'^
•••"'Pu
-^PU
•" * °Pu
' --'Pu
-~;Pu
*Estirated
rWR3 Yankee Ro\je7 BWR5 BWR8 !>resden-I7
33,000 23,9pO 13,000 38,900 20,000 27,500 23,000 3R,400
"X. 1.8 1.00 - 1.92 2. IS ! — 1.0 * /'X .30 1.7!
5&.7 ! 67.7 63.3 56.4 58. '> 57. 2/' : 63.4 53.?
! / !
24.2 X, 13. S 1<>.2 21.9 25." ,,25.7 24.8 2P . 8
11.4 X 10_? ;!_7 n_g -._-,__. •.^f u_6 ' 8]32 :0.-
3.9 :.5i 3.SS 5.77 j \.2 /' 4.5 2.73 ".Pr-
II. Estimated Average Conpssition ot ,pu bailable :or Recycle
Year 1975 '- . 1980 1985
:-!"Pu 1.0 . \. .-.'" - -5 1.7
;- -! 9Pa 64 , " 5g 54
;:';Pu 22 ,"'' 24 '' -,% 25
14 'Pu 10 ..'-'' 11 ''•-. 12
..:.-2irPu 3 ^
III. Plutoni'Jm-rueled Reactors
Reactor
Burnup
Pu
'-'
2H
2U
21*
(MHd/MT)
Source
3Pu
5Pu
°PU
'PU
2Pu
PW1 3
— — ^ ' — •
core 5.
x- ' i axial
33,000, ' •
U-Fueleci PWR
•' 2.7
39.3
25.6
17.3
0- J. i-
blanket
.9
61.5
26.0
7.2
4.5
AI Ref. Oxide IJ4FBR4
radial
blanket
-Core: 80,000-
U-Pueled PWR
.02
97.6
2.33
.04
' 1
core & I core &
blankets ! axial
averaged
.8
66.8
22.5
6.2
3.8
blanket
67.5
24.5
5.2
2.8
GE LMFW D
'"•.^
""•-„ radial
'blanket
-Core: KIO,000-
U-Fueled PV*R
X
X.
94.9
4.9
.2
core &
blankets
averaged
'-••7,0.5
22..4
4.6
2.5
-------
Table 3.2
Isotopic Composition of Plutonium in Discharged Fuels (wt %)
I. Uranium-Fueled Reactors
Reactor '
Bumup (MMd/MT)
238pu
239pu
2"°Pu
2i)1Pu
a^pu
PWR3
33,000
1.8
58.7
24.2
11.4
3.9
Yankee Rowe-
23,900 33,000 38,900
1.00 1.92 2.15
67.7 63.3 56.4
18.8 19.2 21.9
10.0 11.7- 13.8
2.51 " ' 3.88 5.77
BWR5
20,000
58.9
25.7
12.2
3.2
8
BWR
27,500
1.0 * -
57.2
25.7
11.6
4.5
Dresden-I7
23,000 38,400
.80 1.71"
63.4 53.3
24.8 28.8
8.32 10.3
2.73 5.85
*Estimated
II. Estimated Average Composition of Pu Available for Recycle
Year
23.8pu
239pu
2"°PU
2<*lpu
2*2PU
1975
1.0
64
. 22
10
3
1980
1.5
58 " '
24
11
5
1985
1.7
54
25
12
7
III. Plutonium-Fueled Reactors
Reactor
Burnup (MWfl/MTj
Pu Source
238pu
239pu
2-tOpu
^Pu
a^pu
PWR3
33,000
U-Fueled PWR
2.7
39.3
25.6
17.3
15.1
core &
axial
blanket
.9
61.5
26.0
7.2
4.5
AI Ref . Oxide IWFBR4
radial
blanket
-Gore: 80,000-
U-Fueled PWR
.02
97.6
2.33
.04
core &
blankets
averaged
.8
66.8
22.5
6.2
3.8
core S
axial
blanket •
67.5
24.5
5.2
2.8
GE EHFBR
radial
blanket
-Core: 100,000-
U-Fueled PWR
94.9
4.9
.2
core &
blankets
averaged
70.5
22.4
' 4.6
2.5
-------
Page Intentionally Blank
-------
Table 3.3
Average Annual Anvounts and Activities of Selected
Actinides Discharged from Reactors
Isotope
23SU
236U
238U
237Np
239NP
236Pu
238PU
239Pu
2"°Pu
2UPu
2"2Pu
Pu
"'Am
2*2i%n
"2arn
2J(3Am
Am
2"2Cm
2"3Cm
| 2""Cm
I
-••On
Subtotal
Ttotal
U-Fueled PWR3
Curies
Kg Curies after 90d
231 .50 .50
.129 8.20 8.20
2.69x10" 8.97 8.97
13.5 9.51 9.74
2.22 5.17xl08; 489
l.SSxlO-5 9.86 . 9.31
4.63 7.83x10" 8.06x10"
149 9. 14x10 3 9.28xl03
61.4 1.35x10" 1.35x10"
29.1 3.31xlOs 3.29xl06
9.86 38.6 38.6
— 3.39xlOs
.79 2.58xl03 3.80x10'
1.3xlO-s 130 •• 130
2.7xlO-3 2.17xl06 130
2.54 489 489
~ . — 4.55x10*
.35 1.16xl06 7.94xl05
4.0x10-' 175 174
. .87 7.06x10" 7.00x10"
— — 8.64xl05
~ - 5.60xlOa 4.26xl06
— 1.09x10° 4.26xl06
Pu-Fueled PWR3
Curies
Kg Curies after 90d
53 .11 .11
25.7 1.63 1.63
2.67x10" 8.88 8.88
3.55 2.50 2.55
2.09 4.88x10' 5.78x10*
2.24xlO-5 11.9 11.2
10.7 l.SOxlO5 1.92xl05
158 9.70xl03 9.84x10'
103 2.28x10" 2.29x10"
69.9 7.98xl06 7.87xlOG
60.9 238 .238
8.09xl06
3.47 1.13x10" 1.42x10"
.069 668 665
.011 9. 02x10 6 665
30.2 5.78X1Q-3 5.78xl03
-- 2.13x10"
2.29 7.58xlO$ 5.27xl06
4.4xlO"2 2. 05x10 3 2.045C103
' 19.7 1.60xl06 1.58xlO£
~ — 6.85xl06
— 5.19x10' l.SOxlO7
l.llxlO9 1.50x10'
IMFBR4
Curies
Kg Curies after 90d
33.4 .07. .07
.88 .06 .06
2.07x10" 6.87 6.87
2.92 2.06 2.10
5.02 1.17xl09 1.16x10'
1.84xlO-5 9.81 9.27
15.6 2.64xlOs 2.66xl05
1.35xl03 8.28x10" 8.31x10"
454 l.OOxlO5 l.OOxlO5
124 1.42xl07 1.40xl07
77 299 299
— — 1.44x10'
10.8 3.51x10" 4.05x10"
.21 2.04x10' 2.04x10'
4.1xlO~3 3.34xl06 2.04x10'
6.05 1.16xl03 1.16x10'
— 4.57x10"
.52 1.74xlOe 1.20xl06
.02 911 906
.36 2.92x10" 2.89x10"
1.23xlOs
1.19xl09 1.57xl07
— 2.38xl09 1.57x10'
-------
Page Intentionally Blank
-------
7.,.,]o 3.3
' or; •"•-•«. ant •, -irifi Af'.i vi U <-'S of Sclfcted
/\c< V if:'; Discharged fron Rryictorn
isotope
2 3 5U
i3«u
-39u
- :'~NO
235ND
I' :36Pu
'"Pu
1 :3'Pu
s*"Pu
2V1PU
:"2Pu
Pu
^ U-Fueled FWR3
Cui" ies
Kg '\ Curies after 90d
.'31 "'•-. .50 .SO
129 8. 20. 8.20
2.69x10" 8.97 8.97
13.5 9.51 9.74
2.22 5.17x10' 489
1.85x10-' 9.86 9.31
4.63 7.P3xlO" 8.06x10"
149 ' 9.14xl03 9.28x10'
61.4 1.35x10" 1.35x10"
29.1 3.31x10' 3.29xl06
9.86 38.6 38.6
3.39xl06
:"'Air, 1 .79 2.58x10' 3.80xl03
2<2afcr 1.3xlO-3 130 . 130
: ;"2am J2.7xlO-! 2.17xl06 130
; 2 * 3&T1
Am
2.54 489 489
— 4.55xl03
2"20n .35 1.16x10' 7.94xl05
! 2i'3Qtl 4.0xlO-3 175 174
'"'•Qn .87 7.06x10" 7.00x10"
Cm ' — — 8. 64x10 5
. S-Jstotal
5.60x10' 4.26xl06
->otal — 1. 09x10 e 4. 26x10 6
/
l^j-Fvieled PW
C\ji'ies
Kg Curies after 90d
r>3 .11 .11
25.7 1.63 1.63
2.67x10" 8.88 8.88
3.55 , 2.50 2.S5
2.09 4.88x10" r-.78xl' ''
2.2-xlO-5 11.9 11.2
'•• n0.7 l.BOxlO5 1.92xl05
1 8 q.70xl03 9.84xl03
^03 ' \28xlOu 2.29x10"
,/ 69.9 '?.98xlO£ "7.87xin6
60.9 ?38: 238
— x, 8.Q9xl06
3.47 1.13x10" • .1. 42x10"
.069 668 6fe
.011 ^.02xl06 665 '
30.2 •-;. 78x10 3 5.78x10^.
2.13x10"
2.29 7.58x10' 0.27x10*
4.4xlO"2 2.05xl03 2.04xl"3
19.7 1.60x10' 1.58x10*.
6.85xl06
5.19x10' 1.50xl07
1.11x10' 1.50x10'
, LMFBR4
Curies
Kg Curies after 90d
33.4 .07 .07
.88 .06 .06
2.07x10" 6.87 6.87
2.92 2.06 2.10
5.02 1.17xl09 1.16x10'
1.84xlO-5 9.81 9.27
15.6 2.64xl05 2.66xl05
1.35xl03 8.28x10" 8.31x10"
454 • l.OOxlO5 l.OOxlO5
124 1.42xl07 1.40xl07
77 299 299
1.44xl07
10.8 ' 3.51x10" 4.05x10"
.21 2.04xl03 2.04xl03 !
4. 1x10" 3 3.34xl06 2. 04x10 3
6.05 1.16xl03 1.16xl03
4,57x10"
:,52 1.74xl06 1.20x10*
.02., 911 906
.36 *'\ 2.92x10" 2.89x10"
1.23xl06
1.19x10' 1.57xl07
2.38x10' 1.57xl07
19
-------
ro
O
238
236^
I
,?{T, = 22 hr)
236
'Np
237
Np-
238
Np
3(T, = 6.75d)
236
237T
2.35d)
239
'Np
B(T, = 23.5 min)
Z38f,__ 239y
(n,2n) (n,Y)
?, = 26 min)
4.98 hr)
Figure 3.1 Formation. Schars for Iirpoirtarvt Actinides
-------
types. Also the activities are shown after a ,90 day cooling period.
Note that the LMFBR will indeed have the most heavy metal activity
at discharge by about a factor of two over either the uranium-fueled or
piutoniurn-fueled PWR's, with the major activity being that of the beta
decay of 23^Np to 239Pu in all fuels considered. The situation changes
significantly after 90 days of cooling.
The relative amounts of total actinide activities after 90
days of cooling for the LMFBR, piutoniurn-fueled PWR, and uranium-
fueled PWR respectively are : 1, 0.96, and 0.27. The similar ratios
for total plutonium activities are 1, 0.56, and 0.24. The percentages
of total actinide activity which are due to plutonium alone are res-
pectively 92%, 54%, and 80% for the three types of discharged fuels.
The last comparison above is significant. It means that a
reclamation of the plutonium from the uranium-fueled PWR or from
LMFBR discharges would result in separation of most of the actinide
activity. This is not true with the piutoniurn-fueled PWR discharges.
About seven megacuries of americium and curium isotopes would have
to be handled for the piutoniurn-fueled PWR discharged fuel after 90
days, while the uranium-fueled PWR and the LMFBR would each have
about one megacurie of americium and curium isotopes to handle.
The beta emitters among the higher actinides (indicated in^Figure
3.1) all have half-lives of the order of days or less except
Therefore, the activities shown at 90 days in Table 3.3 are mostly
alpha activities. The total plutonium activity in 90 day cooled LMFBR
fuel is 1.44 x 107 curies, but 1.40 x 107 curies of this is the beta
activity from 241pu. Thus about 4.5 x 10b curies of plutonium alpha
activity are present compared to 1.23 x 10^ curies of curium alpha
activity.
The alpha activity of ^41/\m w-ji] continue to build as the
decays. At three years after discharge, there will be about 8.8 x
curies of 24lAm alpha activity. At 30 years after discharge this
activity would be about 3.5 x 10^ curies.
The gamma radiation from the various isotopes and their daughters
is mostly low energy, with less than ~\% of the photons exceeding 400
KeV7. There are a few exceptions such as the 2.6 MeV gamma ray
from 208T1 , a daughter of 236pu. This particular exception would
probably only be important in uranium-fueled reactors.
The neutron production from spent reactor fuels can be signifi-
cant as mentioned previously, 7,10 both from spontaneous fission and
from (a,n) reactions in light elements. Table 3.4 summarizes neutron
production estimates for several nuclides and compounds. The pluto-
nium and americium results are from Reference 7 except where indicated.
The curium results are from Reference' 10.
Tables 3.3 and 3.4 together show that the neutron production from
curium isotopes in discharged fuel will be greater than the neutron
21
-------
Table 3.4
Neutron and Alpha Particle Yields
From Selected Actinides and Their Compounds
n/(g-sec) of Heavy Isotope
Chemical
Form
Pu
Pu02
*Pu02
Pu02
Pu02
Pu02
Pu
Am02
*Am02
Cm203
Cm02
Cm203
Cm02
Isotope
236Pu
238Pu
238Pu
239Pu
240Pu
242Pu
244Pu
241 An,
241 Am
242Cm
242Cm
244Cm
244Cm
Spontaneous
Fission
3.7xl04
2.62xl03
2.4 xlO3
.03
1.02xl03
1.7xl03
B.lxlC3
—
—
2. 30x1 O7
2. 30x1 O7
1.19xl07
1.19xl07
(o,n)
—
1.4xl04
2x1 04
45
1.6xl04
2.7
—
2.6xl03
4x1 03
2. OOxl O7
2.67xl07
4. 29x1 O5
5.72xl05
Total
3.7xl04
1.66xl04
2.2xl04
45
1.7X104
1.7xl03
5.1x103
2.6xl03
4xT03
4,3xl07
4. 97x1 O7
1.23xl07
1.25x10
o/(g-sec) of
Heavy Isotope
1.97xl013
6.47X1011
6.47X1011
2. 27x1 O9
8. 38x1 O9
1.44xl08
6. 54x1 O5
1.20X1011
1.20X1011
1.23xl014
1.23xl014
3. 08x1 O12
3.08xl012
*A11 neutron yields for plutonium and americium are from Reference 7 except these two, which,
along with the curium neutron yields, are from Reference 10.
-------
production from the plutonium despite the iwo to three orders of
mangitude difference in masses of the two < lements which are present
in the fuel.
Moreover, the alpha activity of the curium isotopes will be
greater than that of the plutonium, although the total plutonium
activity is higher due to beta decay of *-4Tpu
It should be mentioned here that, because of the high toxicity
and very long half-lives of the transuranics, and the unique waste
disposal problems created by their presence, work is proceeding to
develop the capability of recycling the actinides along with the
plutonium. This would eliminate the need for handling the trans-
uranics as waste, but would increase the concentration of these
undesirable nuclides in the recycled reactor fuel material.
REFERENCES (Section 3)
1. Plutonium Handbook, A Guide to the Technology, ed. by 0. J. Wick,
Vol. II, Gordon and Breach, New York, 1967.
2. R. Romethsch, "Implementation of International Safeguards - Back-
ground and Future," Trans. Am. Nucl. Soc.. TJ5., 989 (1972).
3. M. J. Bell, "Heavy Element Composition of Spent Power Reactor
Fuels," ORNL-TM-2897, May 1970.
4. "Aqueous Processing of LMFBR Fuels: Technical Assessment and
Experimental Program Definition," ORNL-4436, June 1970.
5. ''Conceptual Plant Design, System Descriptions, and Costs for
a 1000 MWe Sodium Cooled Fast Reactor," GEAP-5678, Dec. 1968.
6. L. C. Schmid, "A Review of Plutonium Utilization in Thermal
Reactors," Nucl. Tech., IjS, 78 (May 1973).
7. R. C. Smith, L. G. Gaust, L. W. Brackenbush, "Plutonium Fuel
Technology Part II: Radiation Exposure from Plutonium in LWR
Fuel Manufacture," Nucl. Tech., 18, 97 (May 1973).
8. "Current Status and Future Technical and Economic Potential of
Light Water Reactors," WASH-1082, January 1968, p. 5-9.
9. D. E. Deonigi, "The Value of Plutonium Recycle in Thermal
Reactors," Nucl. Tech., IjJ, 80 (May 1973).
10. h. S. Bailey, R. N. Evatt, G. L. Gyorey, and C. P. Ruiz,
"Neutron Shielding Problems in the Shipping of High Burnup
Thermal Reactor Fuels," Nucl. Tech., 17, 217 (March 1973).
23
-------
raye iiimnuuiiciiiy Di
-------
4. FISSION PRODUCT GENERATION
4.1 LMFBR Fission Product Generation
Fission product production rates were calculated for two represen-
tative 1000 MWe LMFBR's. Extensive data was available for the AI
Reference Oxide DesignJ but the target bijrnup for this design was
only 80,000 MWd/MT. Since frequently LMFBR; comparisons have been
made for a target burnup of 100,000 MWd/MT, similar results are
reported for a GE 1000 MWe design2 that assumes this burnup. Total
fission product generation should be abou$ equal for the two designs,
except for minor differences such as assumed load factors, different
fractions of power and Pu/U fission ratios in the various core and
blanket regions, and different residence times for the fuel.
A summary of results is presented in Table 4.1. (Although tritium
is a fission product, it is discussed separately in Section 5.1,
and is not included in Section 4). Table 4.1 provides total fuel
discharged annually from each reactor (in metric tons, MT), total
fission product activity discharged with the fuel per year and fission
product power in the discharged fuel, for various cooling times
after reactor shutdown.
The conditions (exposure, specific power, power distribution, length
of time in the core, etc.) for both the AI and the GE designs are
given in Table 4.2.
In Reference 1, values are reported for fission product activities
for the AI design for all fission products which were not negligible
30 days after reactor shutdown. The Reference 1 calculations for the
core were repeated for the most important of these nuclides
(i.e. those which still contributed significantly at 150 days after
shutdown) using fission yields from Reference 3. An energy yield of
215 MeV/fission* was used, which leads to 2.90xl016 fissions/sec MW,
instead of the 203 MeV/fission (3.07x1016 fission/sec MW) used in
Reference 1. For the less important nuclides activity values from
Reference 1 for the core were used; also Reference 1 values were used
for all activities in the axial and radial blankets. (Those nuclide
activities which were calculated are marked with an asterisk in Table 4.3)
It was assumed that 87% of the fissions in the core occurred in
239pu and 13% occurred in "8yt (Activities of several nuclides were
checked using the same input as Reference 1 to assure agreement with
the methods of Reference 1).
*This value is higher than for LWR's primarily because of a higher
value for the kinetic energy of fission products from plutonium
fission than from uranium fission and high gamma energies from
neutron absorption by steel in LMFBR's.
25
-------
Table 4.1
Fuel Mass and Fission Product Activity Discharged Annually
From an LMFBR and Fission Product Power of Discharged Fuel
I. FUEL MASS DISCHARGED
AI Design
Discharge from Core (MT/yr) 8.517
Discharge from Axial Blanket
(MT/yr)
4.948
Discharge from Radial Blanket
(MT/yr) 10.07
II. FISSION PRODUCT ACTIVITY DISCHARGED
GE Design
6.169
6.912
4.869
AI Design
Core
Axial Blanket
Radial Blanket
Total
GE Design
Core Only
Activity Discharged/Yr (Ci/yr)
Coo
30d
3. 77x1 O8
O.lOxlO8
0.44xl08
4. 30x1 O8
3. 08x1 O8
90d
2. 08x1 O8
O.OSxlO8
0.23xl08
2. 36x1 O8
1.67xl08
ing Time
150d
1.40xl08
0.03xl08
O.lSxlO8
1.59xl08
1.14xlQ8
300d
7.81xl07
O.lSxlO7
0.60xl07
8. 59x1 O7
6.10xl07
30y
2. 96x1 O6
0.06X106
0.37xl06
3. 38x1 O6
2. 70x1 O6
III. FISSION PRODUCT POWER
:ission Product Power of Fuel Discharged/Year
Cooling Time (Megawatts)*
AI Design
30d
ts 1.89
90d
1.02
150d
0.71
300d
0.38
30y
0.010
*Equal to the product of MW/MT and MT discharged/year.
26
-------
Table 4.2
Operating Conditions for Two 1000 MWe Designs
AI Reference GE Follow-on
Oxide Design ' Design ^
Average core exposure (MWd/MT 80,000 100,000
Core specific power (MW/MT(U+Pu)) 175 157
Average irradiation time
(equivalent full power days) 458 638
Average chronological (residence)
time in core (days) 540 796
Load factor 0.85 0.80
Fraction of power at mid-burnup,
equilibrium fuel cycle (%)
Core 87.8 87.6
Axial Blanket 1.6 7.6
Radial Blanket 10.6 4.8
27
-------
Table 4.3
Fission Product Activity of Core Discharge Fuel from AI 1000 MWe Reference Oxide
Design (80,000 MWtJ/W Exposure), as a Function of Cooling Time
Activity
V
ro'
Cooling Time
Fission Product 0
*8-Kr •-,. 1.542x10"*
8rRba , 7. 699x10 3
**9Sr .. 2.190xl06
*a:Sr-t--:*Y ' 1.810xl05
*91Y 3.164xlO<;
*95Zr 5.457xl61'
*9a^b 7.093x10"
*a"Nb 5.436x10°
99M~H-9emTc 13.18x10°
* ; c jRu-H ' - 3raRh IS.OSxlO6
*iC6Ru+106Rh 6.372xl06
iioni^a 4. 303x10 3
;::Aga ' 3. 051x10 5
iuAg 4.770xl05
'.uny^a 3. 221x1 0:
'- ; "•'""Cd 9.011xl02
;19lmSn -» ^'47.13
:rnSn ^1.341xlO; X"
173mSn 1. 955x1 0>'X
125Sn 1. 145x1.0 5
* ! 2 5Sb >ii25xlO"*
125mTe / 1.615X101*
1?6Sb /' 9.635xlOu
s .
127Sb ^/x 5.199xl05
127roTeF1'27Te 9. 664x10 5
129mTe 7. 727x10 5
129Te 1.851xl06
129I 0.122
*131Z 4.791xl06
131mXe 4.260xlO!+
132^^^1-1?! 11.468X106
30ri
1.535x10'*
2.522x10*
1.455x10'"
1. 813x10 ';
2.238x10'"
3.973x10'"
.. 5.466x10'"
5,065x10'
8.039x10"
8.933xlOf
6.022X106
4.043X103
5.262xl02
2.985x10"*
3.212X102 ..-""
5.271X102
43.45
"" 1. 341x10 :
1.653x10 3
1.578x10**
6.453x10-
1.691x10''
1.833X103
2.209xl03
2. 976x10 5
4. 213x10 b
2.702xl05
0.124
0.374xl06
1.426x10"
1.936x10"
90d
1.519x10"
2.720x10-'
6.420xlOs
1.806x10'
1.103x10''
2.106x10'
2.898x10"
3.470x10*
0.00273
3.142x10°
5.378xlOfe
3. 4 39x10 3
4. -459x10-
'" 1.162xlO?"
3.183xlO?
2.116xl07
36.74
1.341x10'
1.1 81x10 3
1.889xl02
6.198x10"
1.766x10"
69.71
0.039
2. 012x10 5
1.237x10 ''
7.944x10'*
0.125
2155.810
5.488xl02
0.055
150d
1.503x10"
29.28
2.835xlOs
1.799xlOc
0. 544x1 Of'
1. 116x10'
1.536x10'
2.079xlOi:,'1'"
^'
1.105x10"
4.804xl06
2. 909x10 3
3. 788x10 7
0.455 .
3.155x10'
80.29
31.078 - .; -
1.34Lxl02
8.502xl02
2.267
5.945x10"
1.757x10"
6.471
1.37 5x10 -'
3.646x10"
2.343x10"
0.126
12.43
16.90
-
300d 30 yr
1.464x10" 2.240xl03
3.700x10"
1.781x10- ,_.,'" 8.820x10"
0.092x10*"'
0^228xlOf-
^'" 3. 142x10 :
0.475x10'
f
O.OSlxlO6
3.622x10^ 0.004
1.950x10"-
2.%0xin:
3.095xlO: 73.02
11.12
19.90
1.340xlO: 1.020xl02
*3..600xl02
5.356x10^- 32.52
1.590x10- \ 9.163
5.000 4MJ.9
5.125x10"
1. 650x10 3
1. 095x10 ?
0.126 0.126
-------
"Sable 4.3
Fission Product Activity of Core Discharge Fuel from AI 1000 MKe Reference Oxide
Design (80,000 W8./KC Exposure), as a Function of Cooling Time
Activity (Ci/KT(U+Pu))
ro
00
Cooling Time
Fission Product
A85KJ-,
86Rba
*89gr
*90Sr+90y
*91Y
*9SZr
AgsiriNk*
*95Nb
99M>f99mrc
*103Ru+103mRh'
*106Ru+106Rh
1 1 orn^a
nOAg3
ulAg
U3mcda
llSnca
usngn
1211>ten
I23rasn
l"Sn
*125Sb
1 2 Slttjig
126Sb
127Sb
127mTe+127Te
'lasmjg
l29Te
129Ix
*131j
131Ii^e
132Te+132j
0
1.542X1014
7. 699x10 3
2.190xl06
1.810xl05
3.164xl06
5.457xl06
7.093X101*
5.436xl06
la.lSxlO6^
15.05X105' '
6. 372x10 5
4.303xl03
3. 051x10 5
4. 770x10 5
3.221xl02
9.011xl02
47.13:
1.341xl02
1.955xl03
1.145xl05
6.525X101*
1.615X101*
9. 635x10"*
5. 199x10 5
9. 664x10 5
7. 727x10 5
1.851xl06
0.122
4.791xl06
4.260x10"*
11.468xl06
30d
1.535x10"*
2.522xl03
1.455xl06
1. 813x10 5
2.238xl06
3.973xl06
5.466X101*
5.065xl06
8. 039x10 3
8.933xl06
6.022xl05
4.043xl03
5.262xl02
2.985x10"*
3.212xl02
5.271xl02
43.45
1.341xl02
• " 1.653xl03
1.578x10"*
6. 453x10"*
1.691x10"*
1.833xl03
2.209xl03
. 2.976xl05
4. 213x10 5
2.702xl05
0.124
0.374xl06
1.426x10"*
1.936x10"*
90d
1.519X101*
2.720xl02
6. 4 20x10 5
1. 806x10 5
1.103xl06
2.106xlOs
2.898x10"*.
3.470xl06
0.00273
3.142xl06
5.378xl05
3.439xl03
4.459xl02
1.162xl02
3.183xl02
2.116xl02
36.74
1. 341x1 02
1.181xl03
1.889xl02
6.198x10"*
1.766x10"*
69.71
0.039
2. 012x10 5
1. 237x10 b
7.944x10"*
0.125
2155.810
5.488xl02
0.055
15 Od
1.503x10"*
29.28
2. 835x10 5
1.799xl05
0.544xl06
1.116xl06
1.536x10"*
2.079xl05
1.105xlOs
4.804xl06
2. 909x10 3
3.788xl02
0.455
3.155xl02
80.29
31.078
1.341xl02
8.502xl02
2. '267
5.945x10"*
1.757x10"*
6.471
1. 375x10 5
3.646X101*
2.343X104
0.126
12.43
16.90
300d
1.464x10"*
3.700x10"*
1. 781x10 5
0.092xl06
0.228xl,0s
3.142xl03
0.475xl06
O.OSlxlO6
3.622xl06
1. 950x10 3
2.960xl02
3.095xl02
11.12
19.90
1.340xl02
3.600xl02
5.356x10"*
1.590x10"*
5.000
5. 125x10**
1.650xl03
1. 095x10 3
0.126
30 yr
2.240xl03
8.820x10"*
0.004
73.02
1.020xl02
32.52
9.163
4.119
0.126
-------
Page Intentionally Blank
-------
Table 4.3
(continued-page 2)
Fission Product Activity of Core Discharge Fuel from AJ 1000-MWe Reference,.Oxide
Design (80,000 M!*3/MT Expsoure), as a Function of Cooling Time
Activity (Ci/MS (tH-Pu))
Cooling Time
Fission Product
*133Xe
13"Csa
136Cs
* 137Cs+137ntea- "
1"°Ba+1"clLa
*itice
l^pj.
"i^Cef1""^
:"7Nd
*1"7Pra
item^a
*151Sm
15"Eua
*1S5EU
156Eu
160Tba
lei-n-,
162Gd+162mTb
Total Activity
0
7.867xl06
7.425x10"
3.448xl05
4.903X105
12.242xlOs
( 6. 978x10 6
6.027xlOR
6.337xlOs
2.673xl06
7.673xl05
' 1.672xl05
1.058x10"
2.390xl03
5.221x10"
2.872xl05
3.221x10"
4.478x10"
2.286x10"
(11.70xl07)
30d
0.186xl06
7.226x10"
6.971x10"
4.894xl05
2.541xl06
3.701xl06
1.748xl05
5.886xl06
4.100xl05
7. 743x10 5
. 1. 020x10 5
1.061x10"
2. 380x10 3
5.162x10"
7.330x10"
2.418x10"
2. 201x10 3
2.154x10"
4.421xl07
90d
69.68
6.839x10"
2.843xl03
4.875xl05
9.871x10"
l.OSlxlO5
8.407x10"
5.186xlOs
9. 730x10 3
7.454xl05
3.779x10"
1.059x10"
2.362xl03
5.046x10"
4.581xl03
1.360x10"
5.299
1.927x10"
2.441xi07
150d
0.026
6.461x10"
1.162xl02
4. 857x10 5
3.835xl03
0.287xl06
4.034xl03
4.394xl06
2.286xl02
7. 138x10 5
1.407x10"'
1.058x10"
2. 352x10 3
4.932x10"
2.862xl02
7. 623x10 3
0.013
1.714x10"
1.647xl07
300d
5.400X101*
0.045
4.811xlOs
1.050
0.012xl06
5.100
3.048xl06
0.015
6. 404x10 5
2. 000x10 3
1.055x10"
2. 300x10 3
4.659x10"
0.350
2.000xl03
1.200x10"
9.169xl06
30 yr
2.985
2.463xl05
2.87,4xl02
, 0.849x10"
6.518xl02
«.165xl02
3.472xl05
* These nuclide activities were calculated during the present investigation. Others were obtained from Inference 1.
a. Activation products, produced by neutron activation of a fission product.
-------
Page Intentionally Blank
-------
Table 4.3
(continued-page 2)
Fission Product Activity of Core Discharge Fuel from AJ 1000 MWe Reference Oxide
Design (80,000 MWd^fT Expsoure), as a Function of Cooling Time
Activity (Ci/MT (LH-Pu))
Fission Product
*133Xe
1:"*Csa
136Cs
*137Cs+1-
,-•'•""'*
0
7.867x3.0s
7.425X101*
3. 44 8x1 0s
4. 903x1 O5
12. 242x1.0 G
6.978x10^
6. 027x1 O6
6. 337x1 06
2.673xl06
7. 67 3x10 5
1.672xl05
1.058xlOtl
2. 390x10 3
5.221xlOii
2.872xl05
3.221X1.01*
4.478xlOli
2.286X101*
30d
0.186xlOl-
7. 226x10'*
6.971xlOu
4. 894x10 5
2. 541x10 6
3.701xl06
1. 748x10 r
5. 886x10 f-
4. 100x10 5
7.743xl05
1. 020x10 6
1. 061x1 0L
2.380x10-
5.162X101-
7. 330x3.0 ^
2. 418x10 u
2. 201x1.0 3
2.154X1014
Cc
90d
69.68
6.939X101*
2. 843x10 3
4. 875x10 5
9.871XJ.014
1.031X106
8.407X104
s.isexioe '
9.730xl03
7. 454x10 5
3.779xlOV
1.059X1014
2.362x1.0°
5.046X101*
4. 581x10 3
1.360X1014
5.299
1.927X101*
»ling Tint;
150d
0.026
6.461xlOn
1.162X102
4.857x10
3.835xl03
0. 287x10"
4.034xl03
4.394X.106
2.286xl02
7. 138x10 5
1.407x10''
1.058X1014
2. 352x10 3
4. 9 32X10 •*
2.862xl.02
7. 623x10 3
0.013
1.714X1014
300d
/
5. 400x1 014
0.045
4.811xlOE
1.050
0.012xl06
5. LOO
3.048xi.0':
0.015
6.404x10-
2.000x1?-
1.055x10-
2.300K103
4.659xlOu
r\ 3 =; r
"•-,, 2. OOOxl O3
'--x
1.200x34':
30_yr
2.985
2.463xl05
2.874xl02
0.849X1014
6. 518x1 Oz
8.165xl02
Total. Activity
(11.70xl07)
4.421xl0
7
2.441xl07
1. 647x10 7
9.169xlOe
3.472xl05
* These nuclide activities were calculatjsd daring the present investigate on. Others were obtained fro?r> Reference 1.
a. Activation products, produced by neutron activation of a fission prodxict.
-------
The yields and half lives of the fission products (from Reference 3)
are given in Appendix B. Also the decay schemes of the nuclides
whose activities were calculated are illustrated in Appendix B.
Results for the activities of the important radionuclides are
listed in Tables 4.3, 4.4, and 4.5 for the tore, axial blanket, and
radial blanket of the AI 1000 MWe Reference Oxide Design. Results
are listed as curies per metric ton of metal (U + Pu). Also listed and
noted in the tables are a number of activation products which result
from neutron activation of fission products. Core activities for
the GE 1000 MWe Follow-On Design are listed in Table 4.6. Totals
in Table 4.6 for those nuclides not specifically calculated are based
on results from Reference 1. In Tables 4.3 - 4.6 the totals at zero
cooling time are only the totals for those nuclides shown (hence, they
are shown in parentheses). By 30 days these are the only nuclides
which are not negligible so that the totals from 30 days on are correct.
In Table 4.7 are listed the gamma and beta energy production rates
as a function of cooling time, also listed per MT of metal (U + Pu).
Noble gases and iodine have special significance since they can be
released to the cover gas. These fission product sources are listed
separately in Table 4.8. Saturated activities in a 1000 MWe reactor
are listed, except for 85|
-------
Table 4.4 .
Fission Product Activity of Axial Blanket Discharge from AI 1000 MWe
Reference Oxide Design, as a Function of Cooling Time
Activity [Ci/MT(U+Pu)]
Cooling Time
Fission Product
85KJ-
86Rba
69Sr
90Sr+90y
91y
95zr
95mNb
9%b
99Mo+99mTc
103RUfl03mRh
106Riif106Rh
lioiifcga
110Aga
niAg
nsn^a*
llSniQj-
119mSn
I2imsn
123Jtlsn
125Sn
125Sb
125TOre
'"Sb
127Sb
127n»re+127Te
I29mpe
129Te
129T
Ulj
mmxe
0
7.878xl02
72.64.
1.294xl05
7. 944x10 3
1. 766x10 5
2.834xl05
5.668xl03
2.626xl05
6.480xl05
6.612xl05
1.608xl05
8.162
7.378xl02
1.162X101*
0.580
95.40
3.306
1.001
38.730
4. 090x10 3
9.730xl02
2.947xl02
1.455xl03
1.918X101*
2.097X101*
2.872x10"*
7.236X101*
0.003
1.861xlOs
1. 644x10 3
30d
,7.840xl02
23.80
8.70x10^
7.897xl03
1.247xl05
2. 059x10 5
4.364xl03
2.522xl05
3.949xl02
3. 911x10 5
1. 519x10 5
7.519
0.973
7.283xl02
0.578
58.66
3.042
1.001
32.779
4.478xl02
9.824xl02
3.221xl02
2.758xl02
91.82
4.865xl03
1.568X101*
I'.OOlxlO1*
0.003
1.445X101*
5.970xl02
90d
7. 755x10 2
2.569
3.911X101*
7. 859x10 3
6. 140x10"*
1. 086x10 5
2. 305x10 3
1. 757x10 5
1. 338x10-"*
1. 371x10 5
1. 356x10 5
6.376
0.829
2.843
0.573
22.29
2.569
1.001
23.521
5.365
9.446xl02
3.495xl02
9.919
0.002
3. 533x10 3
4.610xl03
2. 957x10 3
0.003
82.466
23.521
150d
7.670xl02
0.277
1.757x10"*
7.840xl03
3.023X101*
5. 724x10**
1.219xl03
1. 058x10 5
4.799X101*
1. 211x10 5
5.413
0.704
0. Oil
0.568
8.473
2.182
1.001
16.814
0.064
9.097xl02
3.561xl02
0.398
2.409xl03
1.360xl03
8.691xl02
0.003
0.470
0.721
300d 30 yr
7.260xl02 1.143xl02
2. 500x10 3
7. 600x10 3 3. 779x10 3
5. 325x10 3
1.175x10"*
2.800xl02
2.600x10"*
4. 500x10 2
8.780x10"*
3.500
0.450
0.550 0.131
0.815
1.475
1.000 0.763
,7.900
7.85xl02 0.456
3.500xl02 0.189
0.042 0.042 .
8.900xl02
66.00
40.00
0.003 0.003
-------
Page Intentionally Blank
-------
Table 4.4
Fission Pi.o.iix.i. Activity of Axial Blaitet D.i Behsj.yi Iix-iti AT 1000 14*.-
ftefwenee Oxide Design, as a Function of Coo'J ing Tirrse
Activity [Ci/W.1 (i_H-Pu) ]
CO
Cooling Time
Fission Product
FSKr
?6Rba
S9Sr
^Sr+^Y
aiy
"5Zr
95«^b
^5Nb
99JfcH.99mrc
lOBR^iOJmph
106Ru+106Rh
iiaiifcga
110Aga
luAg
1 1 3lT^£|^
1 1 r)IT\Q3
-^msn
1 ; lmSn
123lnSn
125Sn
125Sb
125IIVpe
126Sb
127Sb
127"»tef127Te
izsrope
129Te
129j
131j
I31%e
0
7.878x10-'
72.64
1. 294x10 s
'--.7.944X103
1. 766x1 Ol
2.834x10'-
5.668x10'
-.«:^xio5
6.480x10-"
6. 612x10 5
1. 608x10 5
8.162
7.378xl02
1.162X101*
0.580
95.40
J . JUD
1.001
38.730
4. 090x10 3
9.730xl02
2.947xl02
1.455xl03
LgiSxlO1*
2.097X101*
2.872X101*
7.236x10"
0.003
1. 861x10 5
1. 644x10 3
30d
. 7.840x10-
23.80
8.70xlO:<
7. 897x10 !
1. 247x10 :'
2. 059x10 :
4.364x10-'
2.522x10-
3.949x10'
3.911xl05 ...
1.519x10-
7.519
0.973
7.283xlO?
0.578
58.66
3.G-.2
1.001
32.77,9
4.478xl02
'"' 9.824xl02
3.221xl02
2.758xl02
91.82
4. 865x10 3
1.568X101*
l.OOlxlO1*
0.003
1.445X101*
5.970xl02
90d
7.755xl02
2.569
3.911x10"
7. 859x10 3
6.140x10"
1.086x10"
2. 305x10 3
1.757x10"
1.338x10""
1. 371x10 5
--... 1. 356x10 b
6.376
0.829
2.843 •''•-.
0.573
,..'••22.29
2 . 569
1.001
23.521
5.365
9.446xl02
3. 495x10 :
9.919
0.002
3. 533x10 3
4. 610x10 3
2.957x10"
0.003
82.466
23 . 521
150d
7.670xl02
0.277
1.757x10"
7.840x10-
3.023x10
5. 7 24x10 '^
1. 219x10 :
1.058x10"
4.799x10-
1.211x10-
5.413
0.704
0.011
0.568
• 8.473
2.182
1.001
16.814
0.064
9.097xl02
3.561xl02
0.398
2. 409x10 3
1. 360x10 3
8.691xl02
0.003
0.470
0.721
300d 30 yr
7.260xlO; .,<1 1.143xlO?
..-•'
2.500x10'
7.600x10' 3.779xl0.3
5.325x10
1.175x10-
2.800x10-'
2 , „ „ „•.-. , -
4.500x10-
8.780x10-
3 . 500
0.450
0.550 0.131
0.815
1.475
1.000 0.763
7.900
7,85xlO: 0.456
3.508x10-" 0.189
0.042 '-, 0.042
8.900xl02
66.00
40.00
0.003 0.003
-------
u>
Table 4.4
(continued-page 2}
Fission Product Activity of Axial Blanket Discharge from AI 1000
Reference Oxide Design, ad a Function of Cooling Time
Activity [CiAQ? (U+Pu) J
Cooling Time
Fission Product
132Tef132l
133Xe
I3^csa
136Cs
137cs+137I%a
1"°Ba+I40La
^Ce
lf*3Pr
i^Cefi^^Pr
147Nd
^'Pm
148i%na
151Sm
^W
ISSEu
156Eu
160Tba
lei-n,
162Gd+162ItvEb
0
5.318xl03 .
3.278xl05
3.75Ctod02
6. 981x10 3
1.495x10'*
6. 083x10 s
3.070xl05
2,806x10"
2= 598x10 5
1. 672x10 5
3.731xlOk
1. 861x1 03
4.704xl02
19.27
3 .324x10 3
6.660xl03
56.58
7. 916x10 2
2.418xl02
3M
8.908xl02
7.65bcin2
3. 646x10 2
1. 407x10 3
1.492X101*
1.282xl05
1. 625x10 5
6.849xlOu
2. 399x10 5
?. 569x1 O4
3. 807x10"*
1. 134x10 3
4.723xl02
19.17
3.221jd,03
1. 710x10 3
42o41.
^tf) i^ '' Y
ii.. /J&fet.f'!-
90d
0.003
- 2 '.862
3.457xl02
57.528
1.486X101*
4. 978x10 3
4.496X101*
3. 287x10 3
2. 059x10 5
6 „ 064x1 O2
3.675X1014
4.204xl02
4.723xl02
19.08
3. 023x10 3
l,067xl.02
23.80
0,094
2,040xlQ2
150d
0.001
3.268xl02
2.343
1.481X101*
1.927xl02
1.247xlOu
1. 578x1 O2
1. 785x10 5
14.264
3.523X101*
1.559xl02
4.714xl02
18.89
2. 834x10 3
6.679
13.41
2.258X10-14
1.816xl02
300d 30 yr
2.800xl02 0.015
1.425X101* 7.472xl03
0.070
4.500xl02
0.11C
1. 780x10 5
3.100x10'* 13.98
12.55
4.680xl02 3.722xl02
18.50 5.252
2.35ftKd03 0.034
0.004
0.280
1.33
-------
Table 4.5
Fission Product Activity of Radial Blanket Discharge from Al 1000 MWa Reference
Oxide Design, as a Function of Cooling Time
Activity (Ci/MT (U+Pu) -
Cooling Time
Fission Product
85Kr
86Rba
89Sr
90Sr+90y
91y
95Zr
95mNb
95Nb
saffcH-sgrore
103Ru+103Rh
lOSRu+ioeRh .
i 1 omAga
u°Ag
lnAg
113rnCda
iiarta
119mSn
121ItISn
I23msn
125Sn '
125Sb
lasnvpe
126Sb
127Sb
127mre+127Te
!29nTe
i29Te
129j
131j
0
2. 4 28x10 3
3.363xl02
2.465xl05
2. 371x10"*
3.372xl05
6.102xl05
1.219X101* .
5.819xl05
1.330xl06
. 1.351xl05
4.742xl05
77.36
5.120xl03
3.316X101*
6.017
1.568xl02
8.643
5.885
1.190xl02
1.077X101*
3. 344x10 3
l.llSxlO3
2. 125x10 3
5.507X101*
6.253X101*
6.546X101*
1.625xl05
0.010
3.939X105
30d
2.409xl03
I.lt)5xl02
1. 653x10 5
2.362X101*
2. 371x10 5
4.430xl05
9.403xl03
5. 526x10 5
8.105xl02
S.OOlxlO5
4.411xl05
71.22
9.257
2.078xl03
5.989
96.35
7.954
5.876
l.OOlxlO2
1.171xl03
3.363xl03
1.171xl03
4.034xl02
2.626xl02
1.719X101*
3.571X101*
2.286x10"
0.011
3.061X101*
90d
2.390xl03
11.90
• 7.415X101*
2. 343x10"
1.171xl05
2.333xl05
4.959xl03
3. 826x10 5
2.749x10""
2. 796x10 5
4. 005x10 5
60.45
7.859
8.105
5.942
36.74
6.735
5.866
71.98
14.07
3.231xl03
1.228xl03
14.73
0.006
1.159x10"
1.049x10"
6.735xl03
0.011
1.748xl02
150d
2.362xl03
1.285
3. 335x10"
2.343x10"
5.772x10"
' 1. 237x10 5
2.617xl03
2. 295x10 5
5.063x10" •
3.571xl05
51.29
6.669
0.032
5.894
13.98 •
5.696
5.857
51.14
0.169
3. 098x10 3 •
,1.237xl03
0.727
7.907xl03
3. 089x10 3
1.984xl03
0.011
0.992
300d 30yr
2.300xl03 3.505xl02
. 4. 285x10 3
2.290x10" 1.126x10"
g.isoxio3
2.500x10"
5. 100x1 O2
5.500x10"
9.500xl02
2.675x10''
34.00
.4.300
5.795 . 1.360
1.150
3.750
5.800 .4.478
21.90
2.800xl03 1.549
1.200xl03 0.643
0.300 0.208
2.500xl03
1.45xl02
1.300xl02
0.011 0.011
-------
Page Intentionally Blank
-------
CO
CO
S5Kr
86Rba
89Sr
106RlH-106Rh
1IO
Ag
U5ntd
119nlSn
12lmSn
123mSn
125Sn
125Sb
126Sb
127Sb
Te
129Te
129j
Table 4.5
1 .iv • a . I (.<. ' '•• '/•'<:' •'•. ' •,.' '
Outvie J»c-,':'i ';;/!,
0
2.428xl03
~ 3.363xl02
* -,2.465x10 5
2.371x10"
3.372x10^
6. 102x1 O5
1.219x10"
5.819xl05
1.330xl06
1.351xl06
4. 742x10 5
77.36
5. 120x10 3
3.316x10"
6.017
1.568xl02
8.643 _,-•"'
5.885 r,-
1.190xlO?
l."677xlO"
3. 344x10 3
l.llSxlO3
2. 125x10 3
5.507x10"
6.253x10"
6.546x10"
1.625xl05
0.010
3. 939x10 5
V i'' K ;• - !
X d FV.i • '
jQd
2. 409x10 (
1.105x10-
1.653x10''
2.362x10"
2. 371x10 b
4.430xl05
9. 403x1 O3
5.526xl05
8.105xl02
8. 001x10 s
4.411xl05
71.22
9.257
2.078xl03'
5.9S9
-96.35
7.954
5. -876
l.OOlxlO2
1. 171x10 3
3.363xl03
1. 171x10 3
4.034xl02
2.626xl02
1.719x10"
3.571x10"
2.286x10"
0.011
3.061x10"
,1 P' -,-.;;. f,,.,',
' i-,: ' ..,.; 1ii.lt
••*. i ivi'i_i' s :
Cooliny
„„,
2 . 390x10 '
11.90
7.415x10"
2.343xin'!
1.171X101
2.333x10'
4. 959x10'
3. 826x10 J'
2.749x10-"
2.796x10 5
4. 00 5x10 5
•60.45
7.859
8.105
5.942
-"36.74
6.735
5.866
71.98
14.07
3. 231x10 3
1. 228x10 3
14.73
0.006
1.159x10"
1.049x10"
6.735xl03
0.011
1.748X102
. M Ui'.O MV>
L,'-,.^.../^!
Tine
^"
2.362xlO!
1.285
3.335x10"
2.343x10"
5.772x10"
1.237x10'
2.617x10"
-2.295x10"
5.063x10"
3. 571x10 5
51.29
6 .669
0.032' •..
5.894
13.98
5.696
5.857
51.14
0.169
3. 098x10 3
1. 237x10 3
0.727
7. 907x1-0 3
3. 089x10 3
1. 984x10 3
0.011
0.992
R«=>f««nr^
SOrsd / 3f>:/.
s*
2.300x101 "' 3.505xl02
4. 285x10 3
2.290x10" 1.126x10"
9.1 50x10 3
2. 500x1 01'
5. 100x1 0?
5.500x10"
9.500xlO?
2.675x10"
34.00
4.300
5.795 1.360.
'"'--,. 1-150
'' 3-.750
5.800 4.478
21.90
2.800xl03 1.549
1. 200x10 3 0.643
0.300 0.208
2.500xl03
1.45xl02
1.300X102
0.011 0.011
-------
Table 4.5
(cxmtinued-page 2)
Fission Product Activity of Radial Blanket Discharge from AI 1000 MWs Reference
Oxide Design as a Function of Cooling Time
Activity (Ci/M1 (U+Pu))
fv. Cooling Time
Fission .Prddtjqt
13imXe
132Te4.132j
133XP .
134Csa
136Cs
137Cs+137mBa
^OBa+i'^La
141Ce
mspj-
^Ce^^Pr
U7Nd
1!*7Pm
iien^a ,^'""""'
151Sm -
1 5W
155Eu
156Eu
160-rfca
161Tb
162Gd+152mTb
0
x._ 3 • V D Qjd 0
i Tvsvi n^
7.047X10K,
2. 645x10 3 -,
1. 880x10 u
4.770X101*
1.254xl06
6.310xl05
5. 913x10 5
6.320xl05
3.164xlt5""
^,^-l":620xl05
7. 784x10 3
1.370xlOJ
1.275xl02
1.219x10'*
1.833X101*
5.129xl02
2. 513x10 2
1. 096x10 3
30d
1.313xl03
1.880xl03
1.625X101*
2. 579x10 3
Xj,J^/xiO
4.770X101*
2. 626x10 5
3.335xlQ5 "
1, 445x10 5
5. 800x10 5
4.865xl04
1.030xl05
4.742xl03
1. 379x10 ?
1.275xl02
l.lSlxlO1*
4.695xl03
3.845xl02
1. 237x10 2
1. 035x10 3
90d
51.20
0.005
6.083
2. 437x10 3
1.549xl02
4.-751X101*
1.020X101*
9.248X104
6. 924x10 3
5. 007x10 5
1. 152x10 2
9.919X101*
1. 757x10 3
1.370xl03
1.266xl02
l.llSxlO1*
2.938xl02
2.163xl02
0.299
9.238xl02
150d
1.578
O.t)02
2. 305x10 3
6.320
4.734x10"*
3.958xl02
2.560X101*
3.325xl02
4 . 326xlOD
,.-- 27.11
9.541x10"
6. 54 6x10 2
1. 370x10 3 „
1.256xl02
1.049X104
18.326
1.209xl02
0.001
8.237X102
,,30dd 30yr
2.000xl03 0.1TT
4.700X101* 2.370xlOu
0.180
1. 000x10 3
0.140
3.975x10-
,0.900
B.Zj^x"101* 37.88
49.00
1. 360x10 3 1. 086x10 3
1.220xl02 34.85
8. 900x10 3 0.125
'-\
0*922
27.50
'S._
e.iooxio2
Total Activity
(11.23xl07)
4.361xl06
2.327xl06
1.515x105
5.943xl05
3.648X101*
-------
Table 4.5
(continued-page 2)
Fission Product Activity of Radial Blanket Discharge from AI 1000 M5fe Reference
Oxide Design as a Function of Cooling Time
Activity (Ci/MT (a+Pu))
Cooling Time
Fission Product
13""xe
132Tef132j
133Xe
13"Csa
136CS
137Cs+137mBa
i^Ba+^La
111 'Ce
i^pr
1""Ce+1'"'Pr
147Nd
ll(7Pm
l^IT^a
151Sra
i5"Eua
155Eu
iSSEu
leo-j^a
161Tb
162Gd+162mIt)
0
3.760xl03
l.llSxlO6
7.047xl05
2.645xl03
1. 880x10"
4.770X101* '
1. 254x10 6
6. 310x10 5
5. 913x10 5
6. 320x10 5
3. 164x10 5-
1.020xl05
7.784xl03
1.370xl03
1.275xl02
1.219xl04
1.833X101*
5.129xl02
2.513xl02
1.096xl03
30d
1. 313x10 3
•1. 880x10 3
1.625X101*
2. 579x10 3
3. 797x10 3
4.770x10''
2. 626x10 5
3. 335x10 5
1.445xl05
S.SOOxlO5
4.865X101* '
1.030xl05
4. 742x10 3
1. 379x10 3
1.275xl02
l.lSlxlO1*
4.695xl03
3.845xl02
1.237xl02
1. 035x10 3
90d
51. 20
0.005
6.083
2. 4 37x10 3
1.549xl02
. 4.751X101*
1.020X104
9.248X101*
6. 924x10 3 ,
5. 007x10 5
1.152xl02
9. 919x10"*
1. 757x10 3
1. 370x10 3
1.266xl02
' l.llSxlO1*
,2.938xl02
2.163xl02
0.299
9.238xl02
150d
1.578
0.002
2. 305x10 3 •
6.320
4.734X101*
3.958xl02
2.560X101*
3.325xl02
4.326xl05
27.11.
9.541X101*
6.546xl02
1.370xl03
1.256xl02
1.049X101*
18.326
1.20.9xl02
0.001
8.237xl02
300d
2.000xl03
4.700X101*
0.180
l.OOOxlO3
0.140
3.975xl05
0.900
8. 200x10^
49.00
1.360xl03
1.220xl02
8.900xl03
0.022
27.50
6-lOOxlO2
30yr
0.107
2.370X101*
37.88
1.086xl03
34.85
0.125
Total Activity
(11.23xl07)
4.361xl06
2.327xl06
1.515x105
5.943xl05
3.648x10'*
-------
Page Intentionally Blank
-------
4.6
:> oil Prodbat Activity of Cure Discharge Fuel
GE 1000 Mfe Follow-on Design
r>f i'-<
CO
Activity(Ci/Kr(U+Pu))
Cooling Time
Fission Product
"Kr
89Sr
90SJ.+90Y
91y
95Zr
95IT^b
95Nb
loaRu+.ioai^,
106^+106^
125Sb
127Sb
IBlj
133Xe
137Cs+1371n%a
141Cr
i^Ce+i^pr
147Pm
»5iaa
155Eu
Total for calculated
nuclides
Total for uncalculated
nuclides
Itotal
0
1.915xlOu
2.322X106
2. 260x10 5
3.357x10^
5.795xl06
7. 533x1 Ou
5.787xl06
15.95xl06
7.396xl06
7.925X101*
5. 510x10 5
5.077x10-
8.336xl06
6. 120x10 5
7.395xl06 .
7.238xl06
9.339xl05
1.324X101*
6.421X101*
30d
1.906xl01<
1.542x10°
2. 263x10 "
2.374xlOc'
4.219xl06
5.805X101*
5.387x10
9.199xl06
6.990xl06
7.829x10"
2.341xl03
0.396xl06
0.197xl06
6. 109x10 5
3.922xl06
6. 723x10 6
9. 387x10 5
1. 327x10"
6.348X101*
4.296xl07
0. 697x10 7
4.99xl07
90d
1.886X1014
0.680xl06
2. 254x10 5
1.171xl06
2.236xlOt;
3.078X1014
3.687xl06
3. 330x1 O6
6.244X106
7. 517x1 014
0.041
2284.590
73.84
6. 086x10 5
1.093xl06
5.809xl06
.9.031xl05
1.325xl04
6.205x10"
2.619xl07
. 085x1 n7
2.70xl07
150d
1. 866x10"
0.300xl06
2.245X105
0.577xlO&
1.185xl06
1.631x10"
2.209xl06
1.072xl06
5.576X106
7.210x10"
13.17
0.028
6. 06 3x10 5
0.304xl06
5.019xl06
8. 648x10 5
1.323x10"
6.066x10"
1.812xl07
0. 037x1 O7
1.85xl07
300d
1. 817x10"
0.039xl06
2. 223x10 5
0.099xl06
0.242xl06
0.334x1-0"
0.505xl06
0.086xl06
4. 204x10 6
6.496x10"
6. 006x10 5
1.246x10"
3.482xl06
7. 758x10 5
1.319x10"
5.730x10"
0.973xl07
o.nifixio7
0.99X107
30yt
0.277x10"
1. 104x10 5
0.008
39.44
3. 075x10 5
. 3.482xl02
1.062x10"
0.100x10"
4. 37x10 5
.01
4. 4x10 5
-------
Table 4.7
Energy Generation Rate from Fission Product Decay for the
AI 1000 MWe Reference Oxide Design as a Function of Cooling Time
Specific Power [Watts/Mr (U+Pu)]
Cooling Tirae
Core
Garam Decay
Beta Decay
Axial Blanket
Gamma Decay
Beta Decay
Radial Blanket
Gamma Decay
Beta Decay
30d
9.56xlOt;
lO.-OQxlO-1*
4. 07x10 3
4. 07x10 3
8. 83x10 3
9. 37x10 3
90d
4.39xl04
6.21X104*
1. 77x10 3
2. 29x10 3
3. 94x10 3
5.60xl03
150d
2. 57x10 k
4. 87x10 **
1. 00x10 3
1.69xl03
2. 27x10 3
4. 29x10 3
300d
1.12x10^
2.81x10^
.43xl03
. 91x10 3
0. 98x10 3
2. 45x10 3
3C^r___
. 49x10 3
.59x10 3
15.
21.
48.
66.
36
-------
Table 4.8
Total Activity of Noble Gas and Iodine Nuclides During
Operation,of a 1000 MWe LMFBR
Radio-
nuclide
83nfe
85%r
85KT
87RC
88Kr
89Kr
Ui^e
ISS^Xe
133Xe
I35nfe
135Xe
I37xe
I38xe
1291
1311
1321
133J
1341
135J
Saturated
Activity, Ci
7.016xlOG
1.301x107
2.4xl05*
2.250x107
2.759x107
3.585x107
4.822x105
3.785xlO«
1.328x108
3.459x107
1.416x108
1.138x108
7.828xl07
1.7*
8.036x107
1.050x103
1.327x108
1.392x108
1.217x108
Half-life
1.86h
4.4h
10.76y
76rn
2.79h
3.18m
11.96d
2.26d
5.27d -
15.7m
9.16h
3.82m
14.2m
1.6xl07y
8.065d
2.284h
20. 8h
52.3m
6.7h
Accumulated
Yields (Fast Fission)
239Pu
(*)
0.350
0.642
0.142
1.108
•1.368
1.653
0.025
0.19S
6.824
1.902
7.447
5.785
3.709
0.922
4.196
5.366
6.817
7.186
6.290
2380
(%)
0.412
0.811
0.173
1.416
1.677
3.010
0.022
0.181
6.471
0.852
5.748
• 5.951
5.908
0.653
3.662
5.300
6.471
6.553
5.673
*Approximate values in the reactor (1000 Mtfe GE design) at shutdown for refueling.
37
-------
Page Intentionally Blank
-------
\ Table 4.8
Total Actiyity of Noble Gas and Iodine NuclideS During
Pperation^of a 1000 MWe LMFBR
Radio -
nuclide
aanfe
asi-icr
85RT
87JCT
88Kr
89Rr
ui^xe
m*xe
133Xe
I35m,ce
135X6
I37xe
138X6
129J
1311
1321
1331,
mi
A 3 51
Saturated
Activity, Ci
7.016xl06
1.301xl07
2.4xl05*
2.250xl07
2. 759x10?
3.585xl07
4.822x105
3.785>J.O«
1.328x108 /'
3.459x107
1.416x108
1. 3,38x10"
7<828xl07
/'" 1.7*
8.086x10'
1.050x10
1.327x10
1.392x10:
1.217x10"
Half-life
1.86h
4.4h
'• 10.76y /"'
\ 76m / '
V.79*1
^18m
11.96d
2.26d,
5.27d
15.7m \
9.16h
3.82m \
14.2m \
1.6xlO'y
8.065ci
2.284h
20. 8h
52.3m
6.7h
Accuraiil-ated
Yields (Fast
239pu
0.350
0.642
0.142
1.108
1.368
1.653
0.025
0.195
6.824
1.902
7.447
5.785
3.709
'0.922
4.196
5 1-366
\
6.8\7
7.186\
6.290 \
Fission)
2380
0.412
0.811
0.173
1.416
1.677
3.010
0.022
0.181
6.471
0.852
5.748
5.951
5.908
0.653
3.662
5.300
6.471
6.553
5.673
*Approxiirate valiiss in the reactor (1000 MWfe GE design) at shutdown for refueling.
37
-------
4.2 Comparison with LWR Fission Product Generation
The fission products generated by a reference LWR have been
estimatec by ORNL staff in their study of siting fcr fuel reproce;. nc
plants and waste management facilities.4 The reference LWR is a
pressurized water type fueled with Zircaloy-clad UG2 (3.3% "5^
operating at an average power level of 30 MW/MTU and achieving a
fuel exposure of 33,000 MWd/MTU. The Diablo Canyon Nuclear Power
Plant Reactor served as a prototype for the reference design.
Values for LWR fission product inventories after cooling times
of 90 and 150 days, as taken from Table 3.9 of the ORNL study,4 are
presented in Table 4.9. Comparison of these values with the corres-
ponding values for an LMFBR core discharge (see Tables 4.3 and 4.6)
reveals that the fission product inventories per metric ton (U + Pu)
are much lower for the LWR than for the LMFBR, as expected from the
large difference in MWd exposure per MT.
The total fuel charge in the PWR is 88.6 MTU, of which one tnird
(or 29.5 MTU) is discharged each year. The total fission product
activity associated with the fuel shipped annually is given in
Table 4.10 as a function of cooling time. On comparing these values
with the corresponding AI LMFBR values of Table 4.1 (i.e. 2.36x108
Ci/yr shipped after a 90-day cooling period and 1.59x10^ Ci/yr
after a 150-day cooling period), it is evident that the 1000 MWe LMFBR
will ship annually a greater quantity of activity from the plant
site than will the 1000 MWe PWR, assuming equal cooling times.
This result does not change by including also the actinide activity
(see Table 3.3) and the cladding activity (Table 5.13). (A comparison
between the GE design and the PWR cannot be made without a calculation
of the blanket discharge activities.)
The higher overall fission product activity in the discharged
LMFBR fuel results primarily from the shorter residence time of LMFBR
fuel in the reactor, i.e., 540 days for the AI LMFBR design vs. 3
years for the PWR. Considering freshly discharged LWR and LMFBR
fuels, the average time elapsed since a particular fuel atom
fissioned is roughly 550 days for the LWR compared to only 270
days for the LMFBR. The longer average decay time prior to reactor
shut-down for the PWR far outweighs the effects of differences between
the two reactor types in1fuel exposure, thermal efficiency, energy
release per fission, and isotopic fission product yields. It is
of interest, however, to compare the isotopic yields of several
important fission products. Table 4.11 gives yields of specific
nuclides from thermal fission of 235y and 239pu and from fast fission
Of 238y ancj 239pU- These are cumulative yields. The table indicates
significantly lower production rates for "^Sr and &$Kr and a higher
production rate for i31i in the LMFBR.
38
-------
Table 4.9
LWR Fission Product Activities as a Function of Cooling Time
Curies/Metric Ton Discharged Fuel
Nuclide
3H
85 Kr
86Rb
89Sr
90Sr
90y
91Y
93Zr
95Zr
95mNb
95Nb
99Tc
103Ru
103mRh
106Ru
110mAg
"0Ag
used
119Sn
123mSn
124Sb
125sn
90 days
6.98E 02
1.13E 04
1.72E 01
2.16E 05
7.69E 04
7.69E 04
3.22E 05
1.88E 00
5.24E 05
1.11E 04
8.69E 05
1.42E 01
2.55E 05
2.55E 05
4.59E 05
3.08E 02
4.01E 01
1.17E 02
1.29E 01
5. HE 02
1.73E 02
1.67E 01
150 days
6.92E 02
1.12E 04
1.85E 00
9.60E 04
7.66E 04
7.66E 04
1.59E 05
1.88E 00
2.76E 05
5.86E 03
5.18E 05
1.42E 01
8.91E 04
8.91E 04
4.10E 05
2.61E 02
3.40E 01
4.43E 01
1.09E 01
3.66E 02
8.63E 01
2.00E 01
39
-------
Table 4.9
(Continued - 2)
L!*IR Fission Product Activities as a Function of Cooling Time
Curies/Metric Ton Discharged Fuel
Nuclide
125Sb
1250^
127^
127Tfi
129me
129Te
131j
131mXe
134Cs
136Cs
137Cs
137mBa
140Ba
140La
141Ce
143pr
144Ce
144Pr
147Nd
147P.
148mPm
148 Pm
90 days
8.48E 03
3.32E 03
9.04E 03
8.94E 03
2.27E 04
1.46E 04
3.81E 02
1.06E 02
2.25E 05
5.10E 02
1.07E 05
9.99E 04
i . 1 1 E 04
1 . 28E 04
2.05E 05
1.44E 04
8.92E 05
8.92E 05
2.16E 03
1.04E 05
1.06E 03
8.82E 01
150 days
8.13E 03
3.28E 03
6.18E 03
6. HE 03
6.69E 03
4.29E 03
2.17E 00
3.27E 00
2.13E 05
2.08E 01
1.06E 05
9.96E 04
4.30E 02
4.95E 02
5.67E 04
6.94E 02
7.70E 05
7.70E 05
5.10E 01
9.94E 04
3.92E 02
3.15E 01
40
-------
Table 4.9
(Continued - 3)
L.WR Fission Product Activities as a Function of Cooling Time
Curies/Metric Ton Discharged Fuel
..Bolide 90 days 150 days
l3}Sm 1.15E 03 1.15E 03
u 1.16E 01 1.15E 01
153Gd 2.66E 01 2.24E 01
"i
-------
., , ~ Table 4.10
Fission Product Activity Transported Annually
Fran a 1000 MWe PWR
Cooling Time, days Activity Transported, Ci/yr
90 1.83 x 108
150 1.30 x 108
42
-------
Table 4.11
Yields of Selected Fission Products from Thermal and Fast Fission
Thermal Fission
Nuciide Yields (%) from
35mKr
^SKT
131!
3iSr
13~/cs
I33Xe
235u
1.332
.285
2.774
5.935
6.228
6.766
239pu
.598
.144
3.889
2.121
6.534
6.838
Fast Fission
Yields (%) from
239PU
.642
.142
4.196
2.089
6.625
6.824
238D
.811
.173
3.662
3.282
5.952
6.471
43
-------
REFERENCES (Section 4)
1. Staff, Chemical Technology Division, ORNL, "Aqueous Processing
Of LMFBR Fuels: Technical Assessment and Experimental Program
Definition," ORNL-4436 (June 1970).
2. "Task II Report, Conceptual Plant Design, System Descriptions,
and Costs for a 1000 MWe Sodium-Cooled Fast Reactor, GEAP-5678
(December 1968).
3. M. E. Meek and B. F. Rider, "Compilation of Fission Product
Yields, Vallecitos Nuclear Center - 1972, "NEDO-12154
(January 1972).
4. ORNL-4451 "Siting of Fuel Reprocessing Plants and Waste
Management Facilities," July 1971, Staff of ORNL.
44
-------
5. OTHER SOURCES*
5.1 Tritium and Its Transport
Tritium produced in an LMFBR comes from two principal sources -
ternary fissions in the fuel and n,t reactions in boron control rods.
Lithium contamination in the fuel might lead to another important
source, and Lithium contamination in the sodium is a minor source.
An estimate of tritium production rates in thermal reactors was
given by Peterson, Martin, Weaver, and Harward, 1*2 who also presented
results for tritium production rates from fast fission in a Pu-fueled
fast reactor. More recently Sehgal and Rempert reported calculated
tritium production rates for EBR-II and FFTF. Kabele reported
tritium calculations for FFTF,4 but the references other than the ANS
summary were preliminary and not available to the public at the time
of this investigation.5 Limited information on the details of Refer-
ence 4 were obtained from Westinghouse personnel.^ Data on tritium
production and transport throughout the EBR-II reactor system have
been reported.'7'" Data on tritium transport through fuel cladding in
fast reactors have been reported;8,9 da-ta on tritium transport through
control rod cladding have been obtainedlO but have not been publicly
reported.
5.1.1 Summary
A summary of estimated annual tritium production rates in a 1000
MWe LMFBR is given in Table 5.1. Results could easily be off by a
factor of two.
Most of the tritium may enter the primary system. It is known
that nearly all of the tritium produced in ternary fission enters the
primary sodium. Some unpublished experimental results indicate that
only a fraction (i.e. ^ 30%) of the tritium in the control rods enters
the sodium. Until firm data is presented to show this, however, it
should be assumed that all the tritium enters the sodium.
EBR-II experience indicates that ^ 0.2% of the tritium that enters
the sodium escapes to the atmospher and ^ 0.01% escapes to the condenser
water (EBR-II has a complete steam cycle). It is unclear to what ex-
tent these percentages will apply to an LMFBR power reactor, but they
represent the best indication currently available.
*References are indicated at the end of each subsection of Section 5,
unlike the procedure used in other sections.
45
-------
Page Intentionally Blank
-------
5. OTHER SOURCES*
5.1 Tri \.urn and Its Transport
Trvrim produced in an LMFBR comes from two principal sources -
te c j "".scions in the fuel and n,t reactions in boron control rods.
Litt, ^m contamination in the fuel might lead to another important
scu-za, and Lithium contamination in the sodium is a minor source.
An estimate of tritium production rates in thermal reactors was
give by Peterson, Martin, Weaver, and Harward, T»2 who also presented
resets for tritium production rates from fast/fission in a Pu-fueled
fas: reactor. More recently Sehgal and Rempert reported calculated
trit'jm production rates for EBR-II and FFTF.3 Kabele reported
tritium calculations for FFTF,^ but the references other than the ANS
sumrrary were preliminary and not available to the public at the time
of this investigation:^ Limited informat on on tbe details of Refer-
ence 4 were obtained from Westinghouse personnel." Data on tritium
production and transport throughout the E3R-II reactor system have
been reported.'*8 Data On tritium transport through fuel cladding in
fast reactors have been reported;8'»9 data on tritium transport through
control rod cladding have been obtained^ but have not been publicly
reported.
5.1.1 Summary
- summary of estimated annual tritium production rates in a 1000
MWe u.'IFBR is given in Table 5.1. Results could easily be off by a
fact:."- of two.
•lost of the tritium may enter the primary system. It is known
the" lear'iy all of the tritium produced in ternary fission enters the
prif.-ic.-y sodium. Some unpublished experimental results indicate that
only a fraction (i.e. ^ 30%) of the tritium in the control rods enters
the sodium. Until firm data is presented to show this, however, it
should be assumed that all the tritium enters the sodium.
EBR-II/experience indicates that ^ 0.2% of the tritium that enters
-one sodium/escapes to the atmospher and ^ 0.01% escapes to the condenser
water (EBR-II has a complete steam cycle). It is unclear to what ex-
tent these percentages will apply to an LMFBR power reactor, but they
represent the best indication currently available.
References are indicated at the end of each subsection of Section 5,
^unlike the procedure used in other sections.
45
-------
Table 5.1
Estimated Tritium Production Rates in a 1000 MWe LMFBR
Source Annual Activity Production Rate
Ternary Fission 20,000 Ci/yr
B«C control rods (shim and safety), 7,000
10B(n,t)2a
Lithium produced in control rods 2,500
Li(n,nt)a
Lithium contamination in fuel (20 ppm Ii in fuel)
6Li(n,t)a , 4,000
Lithium contamination in sodium (5ppm Li in Na)
6Li(n,t)a 100
TOTAL * 30,000 Ci/yr
Extrapolation of the above EBR-II leak rates to the LMFBR would
indicate leakage rates for a 1000 MWe LMFBR of the order of 60 Ci/yr
to the atmosphere and 3 Ci/yr to the condenser water. This value for
leakage to the condenser water compares to liquid effluent tritium
rates of 100 Ci/yr for a BWR and 600 Ci/yr for a PWR reported in the
AEC draft statement on the proposed Appendix I to 10 CFR 50.24
5.1.2 Sources
5.1.2.1 Ternary Fission
Tritium production rate from ternary .fission in a 1000 MWe LMFBR
(2500 MWt) is estimated to be 20,000 curies/year. This value could
be as much as a factor of two lower than t'hfc true value, however,
since the tritium production rate from fast'fission of 239pu is so
poorly known. AEC funding to establish this value more precisely
has been terminated.
The production rate of 20,000 Ci/yr is based on the following
parameters: Tritium yield from 239pu fast fission/^ 2 x 10"'* t/f
(Reference 11) Tritium yield from 238l) fission is assumed to be the
same as from 239pu fast fission. N = number of tritons produced/year,
46
-------
anc is roughly equal to:
(2 x ID"4 IKI?^-) (2.9 x 1016 Sr|i~) (2500 MW) (.
(.8 x 3652*^
3.7 x 1023
(tritium) = 12.4 yr.
\ = 1.77 x 10"9 sec "]
Neglecting decay during the year, the annual activity production rate
from ternary fission is:
Activity = NA %20,000 Ci/yr
3.7 x 10'U dis/sec/Ci
The tritium yield data in Reference 11 consist of three points,
as follows:
Neutron Energy Tritium yield/fission
425 +_ 45 keV (1.9 +_ 0.9)x 10"4
483 + 52 keV (2.3 + 1.0)x 10"4
540 + 55 keV (1.1 + 0.4)x 10"4
These values are preliminary and so are not reported iq a public
Document. The tritium yields for thermal fission of 239Pu are not
well known. Hence it is difficult to compare behavior of tritium
yields of 239pu wjth tritium yields from "5y as a function of
fission energy. Such a comparison would be useful since both the
thermal and fast fission yields for tritium from 235u are fairly well
established.
Four measurements of the thermal fission yield for 235y reported
since 1960 all lie between 0.8 x 10~4 and 1 x 10~4 tritons/fission,
with the latest value of Dudey, Fluss, and Malewicki'2 being
0.85 +_ 0.09 x 10~4. Dudey, et al. report values for fast fission
(i.e. between 200 and 800 keV) between 1.5 and 3.0 x 10'4, with an
average near 2.0 x 10~4 which is nearly constant over the 200 to
800 keV energy rangeJ2 Hence the tritium yield from fast fission
of 235y -js about 2.5 times the yield from thermal fission.
Unfortunately there are no reliable thermal fission tritium
yields for 239pu. Horrocks and White^3 report preliminary values
ranging from 1.8 to 5.0 x 10-4 tritons/fission. An independent
47
-------
estimate of tritons/fission can be inferred from two '. intermediate
.*esu~its which nave been reported—alphas/fission and alphas/triton
;>om fission. The number of alphas/fission for thermal fissions of
?39pu is ^2 x 10~3.'2 The number of alphas/triton is M>J4,15
Yin's gives a value of 3 x 10"^ critons/fission. Dudey reports a
theoretical prediction of 2.3 x 10~4 tritons/fission, even though he
$ays the theory is inadequate.'6
On comparing the preliminary values for tritium yield in fast
"ission of 239pu from Reference 11 with the above range of thermal
fission yields, it is noted that the increase in yield with energy
observed for 235y fission may not apply for 239pu fission.
The tritium yield for 239pu assumed by Sehgal and Rempert^ to
calculate tritium production in FFTF was 1.8 x 10-4, a value which
they estimated as the thermal fission yield for 239pu. They report
an annual tritium production rate of 1670 Ci/yr for ternary fission,
assuming 300 MW(th) operation at a load factor of 0.7.
For comparison the annual tritium production rate from ternary
'•ission in a 1000 MWe light water reactor (for 34% thermal efficiency
and 0.8 load factor) can be calculated to be M5,000 Ci/yr. This
value is based on the following assumptions: 55% of the fissions
jccur in 235y, 41% in 239pUj and 4% in 238U; the tritons/fission in
are tne same as 239pu. thermal fission yields are used; and the
number of tritons per thermal fission in 235y and 239pu fission are
0.85 x 10~4 and 2 x 10~4, respectively. This compares to value of
18,700 Ci/yr for a 1000 MWe light water reactor reported in Reference ls
in which a higher thermal fission yield for 23bU (1.3 x 10"4) was used.
5.1.2.2 Boron Carbide Control Rods
It is likely that 846 will be used for shim control in the
LMFBR. Boron carbide is being used for the early demonstration plants
(e.g. PFR, Phenix, FFTF, U. S. Demonstration Plant). Tantalum
was selected as the shim control material for all five 1000 MWe
follow-on conceptual designs 17 (i.e. (£, W, CE, AI, and B & W),
while 846, was chosen for,safety rods for some of them. Since
operating and planned reactors use 646, however, it is prudent at
this time to assume that 640 will continue to be used for our present
purpose of predicting tritium production.
The principal reaction accounting for tritium production in boron
, arbide control rods of an LMFBR is the '^B(n,t)2« reaction. This
reaction has a threshold at about 1 Me1 and has a cross section
averaged over'the fission spectrum of ^30 millibarns. A second
reaction, 'Li(n5nt)a , contributes some tritium; 7Li is produced from
the '^B(n,o )'Li reaction which is the neutron absorption reaction
that leads to the use of 840 as neutron control material. The
threshold for the 7Li(n,nt)a reaction is 2.8 MeV. A third reaction
that produces tritium is ^'B(n,t)^Be, but this reaction has such a
high threshold (9.6 MeV) that it contributes little to the total
tritium production rate.
48
-------
No reported tritium production ra;es from control rods were
"ound for a large power LMFBR that uses B^C shim control. Sehgal
ind Rempert have reported calculated tritium production rates from
:,*C control rods in EBR II and FFTF. Kabele reports total predicted
;ritium production rates for FFTF. These calculations cannot be
jSeG directly for a large power reactor, however, because the tritium
production rate depends upon the fuel cycle adopted and the amount
of B^C needed in the core for shim control for the particular reactor
aesign.
For this reason, an example calculation of tritium production in
a 1000 MWe LMFBR has been made based on control requirements and
results reported for the GE 1000 MWe follow-on design.'8 The basis
for this design is:
12 month refueling interval. The required shim reactivity
' was 4.76$ (1.80% 6k/k).
16$ worth of safety control (including the backup control
system which is located in the axial blanket during operation) .
"ne GE design assumed tantalum rods for shim control and 640 rods for
safety control. Reference 18 provides a neutron balance at mid-cycle
"or an equilibrium fuel cycle, which provides tantalum absorption
Aates in the core and axial blanket and boron absorption rates in
:ne axial blanket. The fraction of neutrons absorbed by tantalum in
the core is 0.00875, which is one half of the shim control requirement
of 0.0180. Hence, the nertron balance is reported for the true mid-
cycle case, i.e. with half of the shim control rods withdrawn from
the core. Half the shim rods represents the average amount of control
in the core during the entire equilibrium fuel cycle.
For the example calculation, the tantalum absorptions were replaced
with the required number of boron absorptions to provide the same
shim control in order to simulate boron-carbide shim control rods.
Reference 18 gives (a) the total neutron flux, (b) the core average
"lux spectrum, (c) the core adjoint flux, $*, (d) the fraction of
fissions in the core, and (e) the Ta capture-to-fission ratio in the
core at mid-burnup. The tore fission rate can be calculated to
give a total reactor powei of 2500 MW(tf)- A sixteen group cross
section set is also available from General Electric which contains
natural boron and tantalum capture cross sections, both self-shielded
for a control pin array equivalent to a 2cinch rod diameter. These
data permit calculation of the number of B atoms in the core
required for shim control by forcing
The neutron spectrum changes caused by replacing the tantalum with
boron were ignored.
For reference, the intermediate results are:
49
-------
A capture in boron and a capture in Ta have almost equivalent effects
on reactivity. Only 1% more absorptions were required in B than
Ta to provide equal reactivity control.
1 q q
Core average fission rate = 1.7 x 10'° fissions/cm0 sec
From the neutron balance, core control-rod captures (.mid-cycle) = 0.00875
core fission 0.299
Tantalum absorption rate in core = 4.9 x 10^ absorptions/cm3 sec
Boron absorptions in core for same Ak == 4.9 x 10 absorptions/cm sec
16 ]n 8 _-,
X ^.('"BH-j - 1.57 x 10"b sec '
10B concentration: N(10B) = 3.1 x 1019 ^^-
Volume of core: V = 3.66 x 106 cm3
cm-
With these values it was possible to calculate the reaction rate
for tritium production in the core from the;reaction '%(n,t) 2a.
This reaction rate is:
CO
N(10B)vf o(E) .
-------
ence, the triton production rate from B in the core was:
'°
N('8)V= f a(E).
- 3.3.x TO15 tritons/sec
th
"his results in an annual tritium activity production rate from B
.n tie core, A(shim, core), of:
A(shim.core) = 4000 Ci/yr
There are two more contributions to the tritium activity from B--(l)
reactions in the part of the shim rods in the axial blanket and (2)
reactions in the safety and shutdown rods which are present in the
axial blanket. The tritium production from these two sources can
be estimated by using again the neutron balance data from the GE
design and by assuming that the ratio of '^B(n,a )'Li reactions and the
'^B(n,t)2a reactions is the same in the axial blanket and the core.
This assumption is not exact because the spectrum is softer in the
axial blanket so that the assumption leads to a small overestimate
of the tritium production.
From the GE neutron balance, the ratio of tantalum capture in
the shim rods in the axial blanket to that in the core is 0.00300/0.00875.
"."he ratio of boron captures in the safety rods in the axial blanket
:o the tantalum absorptions in the core is 0.00364/0.00875 and this
-atio would be the same (within 1%) if BqC had been used for shim
control. Hence the tritium production rate, if the shim rods were
34C, would be
A(shim, ax. blanket) = A(shim,core)
0.00300
0.00875
= 1400 Ci/yr
A(Safety-shutdown, ax. blanket) = A(shim,core)
0.00364
0.00875
= 1700 Ci/yr
Hence the total tritium production rate in the control rods is:
7000 Ci/yr
10r
This tritium production rate from B capture is significantly
lower than the values which would be extrapolated directly from
References 3 and 4. Both Sehgal-Rempert and Kabele calculate higher
tritium production rates in FFTF from |0B capture than from
ternary fission, a result opposite from that shown here for a power
reactor operating with a typical fuel cycle. Sehgal and Rempert
report an annual tritium production rate for FFTF of 3980 Ci/yr from
i0B, based on 300 MW(th) operation at 0.7 load factor.3 Direct extra-
polation to a 2500 MW(th) LMFBR at 0.8 load factor gives a value of
38,000 Ci/yr.
51
-------
Kabele reports a total of 40 Ci/day generation for tritium
production rate in FFTF.^ It is known from conversations with
Westinghouse personnel that ^ 80% of this is from boron capture in
the B^C control rods. Assuming Kabele used 400 MW(th) (the rated
power level for FFTF) and 0.7 load factor, a direct extrapolation
to a 1000 MWe LMFBR would give 64,000 Ci/yr. Hence, Kabele's number
is significantly higher than Sehgal and Rempert's, and the tritium
production rate from ^OB in a power reactor is grossly over-
estimated by extrapolating either calculation to a power reactor.
This overestimate is probably due to the relatively large amounts of
boron needed for control purposes in a test reactor.
Tritium can also be produced by the Li(n,t)a reaction in a
boron carbide control rod. Lithium-7 builds up in a control rod
since it is the product of the '^B(n,a )'Li reaction that provides
the control.
18
Again, the GE neutron balance provides a means of estimating
the tritium production from this source. During one year of operation,
the number of fissions that occur in the core is:
(1.7 x 1013 fl^S10nS) (3.66 x 106 cm3) (.864 x 105 ~) (0.8 x 365
cm -sec
27
1.6 x 10
From the neutron balance, the number of B(n,a ) Li reactions in the
core is:
(1.6X1027) = 5x1025
The ratio of the integrals / o(E) 4>(E)dE for the Li(n,t)a reaction
and the ^B(n,t)2 a reaction is:
dE - 3>0xlo-ll ' '
7' 20
where the Li(n,t)a cross section was obtained from BNL-325 '
Since the number of B atoms in the core at mid-cycle is l.lx 10 .
at which time the tritium production rate from '^B(n,t)2 ain the core
is 4000 Ci/yr, the tritium production rate from ^Li(n,nt)a in the
core after one year, after refueling, is
25
4000 x °26 x.8 '= 1500 Ci/yr
52
-------
"ne /a".ue varies both as the Li concentration builds up and as the
r-ocs are withdrawn during the fuel cycle. Also the 640 shim rods
ourn out and must be replaced periodically.
~:ie source from Li{n,nt)ct reactions in the shim and safety
rocs in the axial blanket was estimated to be 1000 Ci/yr after one
year of operation. This value would increase as 'Li is built up
in -:ne safety rods, which would not require replacement as often as
the shim rods. The rate of increase in the safety rods would be
^ 500 Ci/yr per year, assuming the neutron balance for the GE 1000
MWe design,18 which assumes a particular amount of B-C safety control
in the axial blanket.
Adding the 1500 and 1000 Ci/yr values gives a total of 2500
Ci/yr for the tritium production rate from the 'Li(n,nt)a reaction
as listed in Table 5.1.
For comparison tritium production rates in boron in a light
water reactor can be estimated from Reference 1. For a 1000 MWe
?w'R (2940 MW(th)), the estimated annual tritium production in the
chemical shim for the equilibrium fuel cycle is ^700 Ci from the
^°B(n,t)2a reaction and ^1250 Ci from the 7|_i(n,nt)a reaction.
"ritlum production rates in BWR control rods are much higher, but
the tritium does not escape from the control rods.
5.1.2.3 Lithium Contamination
Lithium is present as an impurity both in reactor fuel and in
the sodium coolant. Neutron capture by lithium-6 leads to tritium
production through the reaction °Li(n,t)a .
Although fresh fuel is expected to have less than 1 ppm 2i
lithium, reprocessed fuel may contain as high as 20 ppm lithium.
Kabele included this source (20 ppm) in his estimates for FFTF.^
Extrapolating Kabele's calculation to a 1000 MWe LMFBR results in
an estimated 4000 Ci/yr tritium production rate from 20 ppm of
lithium in the fuel.
Lithium content in the FFTF sodiun is specified to be less than
5 ppm. The tritium production rate from this lithium is <100
This value can be estimated from the above 4000 Ci/yr source from
lithium in the fuel since the sodium and fuel volumes in the core
are comparable, the lithium mass concentration is 5 ppm in sodium
instead of 20 ppm in fuel, and the sodium density is a factor of
10 lower than the fuel density.
53
-------
5,;.3 Transport of Tritium in an LMFBR System I
5.i.3.1 Escape into Sodium System
Escape from Fuel Pins j /«
Tritium produced in the fuel from ternary fission and from
lithium in the fuel diffuses through the steel cladding into the
sodium. Roy, Rubin, and Wozadlo" report experimental results of
irradiated mixed-oxide fuel pins with austenitic stainless steel
cladding which show that less than 1% of the tritium produced is
retained in the fuel pin. Hence nearly all tritium gets into the
sodium and little is available for release during fuel repro-
cessing.
Additional data on tritium leakage from fuel is available
for EBR-II driver fuel.''8 The data is of less interest since
the driver fuel is metallic uranium. Assuming 2 x 10~4 tritons
produced/fission, EBR-II staff report that nearly 100% of the
tritium diffuses out of the driver pins at average fuel tempera-
tures greater than 1000°F and that 80% escapes when the average
fuel temperature is 800°F.8
This situation is different from the case of light water
reactors. Little of the tritium diffuses through the zirconium
cladding, so most of the tritium is retained in the fuel pin in
light water reactors. The difference may be caused by the dif-
ference in cladding temperature more than by the difference in
cladding material. The cladding in an LMFBR operates at ^ 400°F
higher than that in a light water reactor.
Control Rods
No published data wen found on diffusion of tritium from control
rods in an LMFBR. It has tjenerally been assumed by LMFBR designers
that tritium produced in B^C rods would diffuse through the cladding
s'nce the cladding is steel at high temperature, as is the fuel
cladding.
Some data have been received directly from the EBR-II staff,
however, that differs from the above assumption. The EBR-II experi-
ence is that all of the tritium produced in 646 clad in steel stays
within the rod. It was assumed at EBR-II that this resulted from
irradiation at lower 840 temperature (1100°F) and higher 846 densities
(2,5 gm/cm^) than planned for power reactors. Later, these 640 rods
were heated to 1500°F for 120 hours and still no loss of tritium oc-
curred. However, EBR-II staff learned that HEDL experimenters who
irradiated 640 control assemblies in EBR-II at 1600°F centerline
temperatures and 2.1 gm/cm^ density—similar co power reactor condi-
t-'ons--found that 70 to 75% of the tritium produced was retained in
the cladding. Unfortunately, so much of the experimental work done
by HEDL for FFTF is unavailable to the public that details on this
question are unavailable; it is still not clear what fracton of the
tritium produced in 646 rods will escape to the sodium.
54
-------
Vhli question may be rendered academic if a different design
approach is taken. Boron,carbide cont ol rods may oe designed to vent
ie nenum produced from B(n,a )'Li eactions to the coolant in order
to avoid the large pressures resulting from helium production. In
tnat case, the tritium produced would also escape to the sodium.
5.1.3.2 Transport in the Sodium and Steam Systems
Tritium is removed in cold traps. The reaction that removes
tritium is unclear. Two possible mechanisms are deposition of
sodium hydride and exchange of 'H and tritiun in the steel mesh in
t.'ie trap.
Some tritium, however, can escape elsewhere—to the cover gas
f^om which it could leak to the reactor building and the environment,
tnrough the primary system boundaries (piping and vessels) to the
:-aactor building and the environment, and to the secondary sodium
across the tube walls of the intermediate heat exchanger.
Host of the tritium which finds its way to the secondary sodium
- £ trapped in the secondary-sodium cold trap. Some, however, escapes
"rough the secondary-system boundaries to the environment, and some
enters the steam system through the steam generators and superheaters.
Proof that tritium is removed in the cold traps comes from
operation represented by Figure 5.1 which is reproduced from Reference 7.
Tht measured tritium concentrations in the primary sodium with and
without cold trap operation clearly indicate the effectiveness of the
cc'Id trap in removing tritium. Also at EBR-II, during one month when
tnc secondary-system cold trap was not being operated, tritium
concentrations in water samples taken from the steam system were
5 to 8 times higher than for normal operation with the secondary cold
traps in service.8
Most of the tritium which reaches the steam system would be
expected to get into the environment eventually, the modes and rates
of .eakage will depend on how the steam system is designed and operated.
EBR-II has reported both tritium production rates and distribu-
•;•:/: of the tritium in the EBR-II compiex. »°>3 The results of
References 7, 8, and 3 are summarized in Table 5.2. All losses to the
-eactor building should appear in the air to the stack since the
reactor building atmosphere is continually exhausted to the stack.
The Known tritium losses to the environment during operation are
^ 1 Ci/yr. This is about 0.2% of the tritium that enters the primary
sodium. About 10% of the tritium at EBR-II remains in the fuel.°
55
-------
b 220
Q 180
COLD TRAP OUT
OF SERVICE -
-NEW COLD TRAP
INSTALLED
I 1 1
COLD TRAP
REMOVED
i 1 i
20 24
32 36
MW DAYS OF OPERATION (XIO~3)
Figure 5.1 Tritium in Primary Sodium (EBR-II)
56
-------
Table 5.2
Summary of EBR-II Tritium Data
Reference 7 Reference 8
Reference 3
Production Rate
Ternary Fission
Boron
Total
100%
Neglected
420 Ci/yr
190 Ci/yr
510
700
Known Loss Rates
In air to stack 0.9 Ci/yr
In condenser water 0.05 Ci/yr
Tritium Distribution
Percent distribution from
Ref. 7 (boron control not
included)
Fuel 20-25% ^ 10%
Primary sodium 4%
Primary cold trap 68-70%
Outside fuel and
primary system 3-7%
Loss from power
plant <1.5%
Absolute activities of typical
sampling data (Ref. 8)
Primary sodium 15.5 Ci
Primary argon 0.0015 Ci
Secondary sodium 0.083 Ci
Secondary argon 0.00018 Ci
Steam system 0.00058 Ci
a. Based on 50 MW(th) operation
b. Based on 62.5 MW(th) at 0.7 load factor
c. Based on 2.0 x 10~4 tritons/fission in 235y
d. Based on 0.8 x 10'4 tritons/fission in 235|j
57
-------
The latest public documentation of EBR-II tritium concentration
was found in Reference 22 in which the following results were reported:
Table 5.3
Tritium Concentrations in EBR-II
Region
Primary Sodium
Secondary Sodium
Turbine Condensates
Concentration
4.
6.
6.
9.
2.
72
13
2
3
7
2
9
X
X
X
X
X
10
10
10
10
10
-2
_9
(-
-3
"3
-3
yCi
/gm
yCi/gm
yCi/gm
yCi
yCi
/gm
/gm
pCi/ml
pCi
/ml
April
May,
April
May,
June,
May,
June,
, 1971
1971
, 1971
1971
1971
1971
1971
Secondary
Cold Trap
Date of Sample Operating?
No
No
No
No
Yes
No
Yes
These values are consistent with the results from Reference 8 quoted
in Table 5.2.
Additional data is now being reported in the non-pub!ic ANL
reactor progress reports. For example starting in September, 1972
(Reference 23) tritium in the steam system is being reported regularly.
The value in Reference 23 is ^ 10 pci/gm with the secondary cold
trap operating, which is similar to the result in Table 5.3. The
tritium activity reported in Reference 23 in the primary argon cover
gas is 16 pci/cm3. This result is about a factor of 10 lower than the
concentration that would be calculated from the primary cover gas
result from Reference 8 data in Table 5.2 above.
The source terms in the EBR-II analysis of Reference 7 include
only ternary fission, and assume 0.8 x 10"^ tritons per fission,^'7
since boron control was not used at that time. Sehgal and Rempert
calculate more than 2 1/2 times as much tritium from boron control
rods as from ternary fission^. However, the tritium in the boron
control rods may not contribute to the source in proportion to
its production rate since, according to Reference 10, the tritium
produced in the B4C rods does not escape from the rods in EBR-II.
An extrapolation of tritium leakage from a 1000 MWe LMFBR can
be made by assuming that the same fraction.of tritium that enters
the primary sodium escapes to the environment as in EBR-II (i.e.
^0.2% to the atmosphere and 0.01% to the condenser water). Based
on ^30,000 Ci/yr entering the sodium in a 1000 MWe LMFBR
58
-------
(see Table 5.1), the amount leaving the reactor through the stack
would be ^ 60 Ci/yr. The amount leaving in the condenser water
would be ^ 3 Ci/yr. Such an extrapolation may be misleading for
several reasons. Leakage of argon to the reactor building is
excessive in EBR-II (See Section 8.2) and much higher than could
be tolerated in an oxide-fuel power plant; hence the stack leakage
extrapolation may be too high. Also the steam generators, a principal
component in the pathway to the condenser, contain double wall
piping which may not be used in a power reactor; hence the fraction
of leakage to the condenser could be higher in a power reactor not
using the double wall design. Also, since EBR-II is a pot design
extrapolation to a loop-type design may not be valid because of
different bulk sodium and wall temperatures, etc.
For comparison, the AEC has calculated tritium leak rates in the
liquid effluent for a BWR and PWR.24 For a 1100 MWe BWR, the
annual tritium release rate in the liquid effluent was predicted to
be 110 Ci/yr (or 100 Ci/yr for a 1000 MWe plant). For a 870 MWe
PWR the corresponding release rate was 500 Ci/yr (or 600 Ci/yr for
a 1000 MWe plant). The AEC report did not consider tritium in
gaseous effluents.
The EBR-II experience, therefore, indi£ate,s that although the
jritium production rate in an LMFBR is higher than in a LWR, the
amount that finds its way to the liquid effluent is smaller in an
:.MFBR.
REFERENCES (Section 5.1)
1. H. T. Peterson, J. E. Martin, C. L. Weaver, and E. D. Harward,
"Tritium Contamination from Increasing Utilization of Nuclear
Energy Sources," Proceedings pf the International Conference_on
Environmental Contamination of Radioactive Materials. SM-117/78,
IAEA, Vienna, 1969.
2. C. L. Weaver, E. D. Harward, and H. T. Peterson, "Tritium in the
Environment from Nuclear Power Plants," Public Health Reports, 84,
363, 1969.
3. 5. R. Sehgal and R. H. Rempert, "Tritium Production in Fast
Reactors Containing B4C," Trans. Am. Nucl. Soc.. 14^ 779, 1971.
4. r. 3. Kabele, "Tritium Distribution in FFTF," Trans. Am. Nucl. Soc.,
15, 79, 1972.
5. xeply to Aug. 18, 1972 request to H. Bullinger for HEDL-TM-7219,
Report on Tritium Distribution in FFTF", by T. J. Kabele.
6. IfJestinghouse staff, personal communication, August, 1972.
59
-------
7. E. R. Ebersole, W. R. Vroman, and J. R. Krsul, "Tritium in trie
EBR-II Reactor Complex," Trans. Am. Nucl. Soc. 14, 321, 1971 „
8. EBR-II Staff, Revision of Data in Reference 7, supplied to
participants in an LMFBR workshop sponsored by ANL, AUA, and AEC,
December 11-15, 1972
9. C. -P. Wozadlo, B. F. Rubin, P. Roy, "Tritium Analysis of Fas,: Flux
Irradiated Mixed-Oxide Fuel Pins," Trans. Am. Nucl. Soc., 1_54 200,
1972.
10. EBR-II Staff, personal comnunication, December, 1972.
11. Reactor Development Progress Report, NAL-RDP-2, p. 7.35,
February 1972.
12. N. D. Dudey, M. J. Fluss, and R. L. Malewicki, "Tritium and
Alpha Particle Yields in Fast and Thermal Neutron Fission of
Uranium-235," Phys. Rev., C6(6), 2252 (1972).
13. D. L. Horrocks and E. B. White, "Tritium Yield in the Thermal
Fission of U-233", Nucl. Phys., A151, 65 (1970).
14. P. Cavallini, et al., "Application of New Fast Chemical Separation,
to the Determination of Charge Distribution in Low-Energy Fission",
Proceedings of the Second Symposium on the Physics and Chemistry
of Fission, IAEA, Vienna, 1969.
15. T. Krogulski, et al., "Emission of Ligh Nuclei in Thermal Neutron
Fission of Pu-239", Nuclear Physics, AT.'8^ 219 (1969).
16. N. D. Dudey, personal communication, July 24, 1972.
17. Five papers on 1000 MWe Follow-on Designs of GF., W_, Al, CES and
B & W Proceedings of the International Conference on Sodium
Technology and Large Fast Reactor Design, ANL-7520, Part T, November,
1968.
18. "Task II Report, Conceptual Plant Design, Systems Descriptions,
and Costs for a 1000 MWe Sodium-Cooled Fast Reactor," GEAP-5678,
pages 91-101, December, 1968.
19. D. C. Irving, "Evaluation of Neutron Cross Sections for Boron-10",
ORNL-TM-1972, October, 1967.
20. BNL-325, Second Edition, Supplement No. 2, "Neutron Cross Sections,
Vol. I, Z = 1 to 20," J. R. Stehn, M. D. Goldberg, B. A. Magurno,
and R. Wiener-Chasman--Editors, May, 1964.
21. 6. R. Taylor, WARD, personal communication, August, 1972.
22. Reactor Development Progress Report, ANL-7845, July, 1971.
23. Reactor Development Progress Report, ANL-RDP-9, September, 1972.
60
-------
24. Staff, Directorate of Regulatory Standards, USAEC, "Draft
Invironmental Statement Concerning Proposed Rule Making Action:
Numerical Guides for Design Objectives ;nd Limiting Conditions
for Operation to Meet the Criterion 'As Low as Practicable1
for Radioactive Material in Light-Water-Cooled Nuclear Power
Reactor Effluents," (Proposed Appendix j to 10 CFR50),
January, 1973.
5.2 Activated Corrosion Products
Sodium slowly corrodes metallic surfaces in and near the cere.
These metallic surfaces are radioactive as a result of neutron
activation. Radioactive corrosion products inay remain in solution in
the sodium or may be deposited on other surfaces in the primary system,
such as the reactor vessel, piping, intermediate heat exchangers,
pumps, and cold traps.
Both corrosion rates and deposition rates are influenced by surface
and sodium temperature, flow velocity, and oxide concentration in the
sodium. Corrosion and deposition rates and the dependence of these
rates on the above parameters vary for different metals. Although
experimental information on corrosion and deformation rates has
been obtained by General Electric, 1>2 by Hanford Engineering
Development Laboratory^ (HEDL), and in the United Kingdom4 (as discussed
in References 5 and 6), analysis of the distribution of corrosion
product activity throughout the primary sodium system is still quite
uncertain. In the analysis of activated corrosion product distribution,
the primary sources of uncertainty are assoc ated with corrosion
rates and deposition patterns; uncertainties involving reaction rates
and amounts of target nuclides present are less important.
The best analyses available are two analyses* for FFTF. The most
recent and complete analysis was performed by HEDL,5 and a slightly
earlier analysis was made hy General Elactric.^ The corrosion data of
GE and other data and methods used by HEDL were applied here to
extrapolate corrosion product activity and distribution to a 1000 MWe
LMFBR. Additional data on corrosion product reactions appear in AI's
STP-1 code7 and in an ORNL reference8 which quotes data from their
reference library.
The principal corrosion products, the reactions that produce them,
and their half lives, are given in Table 5.4.
* The principal purpose of the two reports was to estimate dose at
primary system components for maintenance purposes. However,
production and distribution of the activation products were intermediate
steps in both reports.
61
-------
Table 5.4
Activation Reactions in Stainless Steel
Nuclide Reaction Tl/2
60Co 59Co(n,Y)60Co 5.24yr
60Ni(n,p)60Co
58Co 58Ni(n,p)58Co 71 d
54Mn 54Fe(n,p)54Mn 313 d
59Fe 58Fe(n,Y)59Fe 45 d
55Fe* 54Fe(n,Y)55Fe 2.4 yr
51Cr 50Cr(n,Y)51Cr 28 d
54Fe(n,a)51Cr
182Ta 181Ta(n,Y)182Ta 115 d
The following additional reactions are possible, but they contribute
little to corrosion product activity:
58Ni(n,Y)59Ni 52Cr(n,2n)51Cr
62Ni(n,Y)63Ni 55Mn(n,2n)54Mn
59Co(n,2n)58Co
The above reactions have cross sections that are too low to make
them of interest except possibly the reactions that produce ^Ni and
63Ni since their half lives are so Long, (8;'xx 1CT yr and 92 yr
respectively); these two products are included in the section on
Cladding Activation (Section 5.3.2). >
*This corrosion product was not included in the HEDL calculation because
bbFe decays only by electron capture, giving up a maximum of 0.22
MeV energy as internal bremsstrahlung, apd this energy is of little
consequence to primary system maintenance.
62
-------
5.2.1 Estimated Corrosion Product Activity In 1000 MWe LMFBR
Estimates of the corrosion rate and activity of each activated
corrosion product in the primary system of a 1000 MWe LMFBR are
presented here. It is important to emphasize that great uncertainty
exists in the estimates. The methods used here show how to estimate
corrosion product activity, and provide an order-of-magnitude
result. The large differences between the GE and the HEDL cal-
culations for FFTF are then discussed in order to indicate the
degree of uncertainty in the calculations.
Corrosion product inventories in the primary system are given
for 30 years of reactor operation. 58Q0j5lcr559pe ancjl82ja achieve
equilibrium during the first year. ^Mn takes somewhat longer,
and °°Co and 55pe still longer. The data used to estimate these
inventories are listed in Table 5.5. Estimated inventories are given
in Table 5.6.
Table 5.5
Data for Corrosion Product Calculation
Region Corrosion Surface Average Neutron
Rate Area Flux ?
(mils/yr) (ft^) (n/cm sec)
Core 0.13* 17,000 7 x 1015
Upper axial blanket 0.3 8,000 3 x 1015
Gas Plenum (above core) 0.3 17,000 0.5 x 10
Radial blanket 0.1 10,000 -2 x 1015
(upper half)
*Based on corrosion vs. temperature curve in Reference 6 and
an inlet temperature of 800°F and outlet temperature of 1100°F.
63
-------
Page Intentionally Blank
-------
5.2.1 Estimated Corrosion Product Activity in 1000 MWe LMFBR
Estimates of the corrosion rate,and'activity of each activated
corrosion.product in the primary"system of a tQOO Mtye LMFBR are
presented here. It is important to emphasize that great uncertainty
exists in the estimates. The methods used here show how to estimate
corrosion product activity, and proyice an order-of-magnitude
result. The large differences between the GE and the HEDL cal-
culations for, FFTF are then discussed in order to indicate the
degree of uncertainty in the calculations.
Corrosion product inventories in the primary system are given
for 30 years of reactor operation. 58Q0,5lcr,59pe and!82ja Achieve
equilibrium during the first year. 5^Mn'takes somewhat longer,
and °°Co and 55pe st>n longer. The data used to estimate these
inventories are listed in Table 5.5. Estimated inventories are given
in Table 5.6. ; . , ... , ,-;^
Table 5v5:.v:s -'q.fii
Data for Corrosion Product Calculation
Region
Core
Upper axial blanket
Corrosion
Rate
(mils/yr)
0.13*
0.3 J
Gas Plenum (above core) 0.3 ,
Radial blanket 0.1
Cupper half)
Surface Average Neutron
Area Flux 0
(ft?)
17,000
g:»000
17,000
10,000
(n/cm sec)
7 x 10
3 xllO
15
15
0.5 x 10
2 x 1015
15
*Based on corrosion vs. temperature curve in Reference 6 and
an inlet temperature of 8000F and outlet temperature of 1100°F.
63
\
-------
Table 5.6
Estimates of Activated Corrosion Products in the Primary
System of the 1000 MWe LMFBR After 30 Years Operation
Contribution to the Primary
System Activity
Formation Axial Gas Radial Total Primary
Isotope Reaction Core Blanket Plenum Blanket System Activity
60rrt
Co
58Co
54Mn
55Fe
59Fe
51 Cr
82Ta
(n,y)
(n,p)
(n,p)
(n,p)
(M.Y)
(n,Y)
(n.Y)
(n,o)
(H.Y)
(CD
1400b
1000
20,000C
16,000d
26,000
300
2500
200
800
(CD
9300b
2800
2400
f
500
2800
3200
(CD (CD
6600b 2200b
300 400
300
f f
300
1800
2400
(Ci)
20,000
23,000
19,000
>26,000
1,000
7,000
6,000e
a. All values based on stoichiometric corrosion, assuming 316 stainless
steel (see Table 5.9 for composition).
b. Based on only 0.02% by weight cobalt in stainless steel. (HEDL
assumed 0.02%, GE6 assumed 0.1%).
c. This value was calculated using <|>(E) and a(E) (see page 55). For
comparison the value calculated from of Table 5.5 and aof Table
5.7 was 21,000 Ci.
d. This value was calculated using (E) ami o(E) (see page 66). For
comparison the value calculated from <{>OF Table 5.5 and o of Table
5.7 was 15,000 Ci.
e. The assumption of stoichiometric corrosion is believed by HEDL to
be particularly poor here. (HEDL assumed only 1% of stoichiometric
corrosion).
-------
Cross sections were unavailable for the soft spectra in these reg~;
It Is expected that the 55re generation from the n9v reaction
would be higher outside than inside the core. It was decided not
to pursue this calculation further, however, because of the low
importance of the 5^pe isotope since it is neither a3 or y
emitter (see footnote, Table 5.4)
The corrosion rates are based on a curve presented in the GE
report^ for a sodium flow rate of 15 to 28 ft/sec and an oxygen
concentration of 2.6 +_ 1.5 ppm. The 0.13 mils/yr corresponds to an
average corrosion rate over the core for an inlet of 800op and an
outlet cf 1100°F. The 0.3 mils/yr corresponds to 1100°F sodium.
The lower value for the radial blanket accounts for both a lower
sodium temperature and a lower flow rate.
The HEDL calculation assumed much lower corrosion rates even though
the same inlet and outlet temperatures were assumed. Two cases were
reported by HEDL--one for an oxygen content of 5 ppm and one for an
oxygen content less than 2 ppm. Even for the high oxygen case, the
corrosion rates were a factor of 2 to 3 lower than the GE values.
Values used by HEDL were: core-C.055 mils/yr; axial reflector-0.13
.T.ils/yr; gas plenum-0.095 mils/yr; and radial reflector-0.025 mils/yr.
For the case of <2 ppm oxygen, HEDL assumed a further reduction in
corrosion rate of a factor of four, i.e. 0.014 mils/yr was used for
the core. The source of such a large discrepancy in corrosion rate was
not discussed. GE claims that its curve is based on mass transfer
data from GE and UK. GE refers to the experimental results of
Srehm of HEDL, indicating agreement in some areas and disagreement 3
in others. HEDl's analysis references only a later paper by Brehm.
The GE analysis" also used lower corrosion r< tes than in Table 5.5 but
their calculation assumed a 700°F inlet and a 1050°F outlet.
Stoichiometric corrosion rates were assumed by GE . Significant
deviation from Stoichiometric corrosion was assumed by HEDL,5 however,
cased on specific Information about corrosion of particular elements.
for example, cobalt in stainless steel is assumed by HEDL to corrode
at about 20% of the rate of the stainless steel in high oxygen sodium,
and manganese is assumed to corrode at twice the rate of stainless
steel. Tantalum is assumed to corrode at only 1% of the rate of
stainless steel, an assumption which HEDL says is backed up by the
fact that no 182ja is observed in EBR-II and tantalum corrosion
"ates are low in other stainless-sodium systems. However, 182ja was
observed on the primary pump walls at EBR-II, even though EBR-II
itaff suggested that the source of the '82ya was the cladding of an
antimony neutron source instead of the stainless steel structure.
3my iron and chromium are assumed to be released at Stoichiometric
rates by HEDL. The GE analysis did not include a calculation of
cantalum activation.
The surface area of the core is tha; calculated for the 1000 MWe
65
-------
g
GE follow-on design. The upper axial blanket has half the area in
the GE design, and it was assumed that the plenum area is equal to
the core area. The sodium in the lower axial blanket has too low a
temperature to contribute to corrosion.
The fluxes were based on both the GE 1000 MWe follow-on design , r
report9 and on the HEDL FFTF report.5 The average flux of 7.1 x 10
sec in the core is given in Reference 9. For FFTF the core average
flux is 4,2 x 10'5; the value is smaller primarily because of the
relatively smc.ller size of the core. Average fluxes in the rest of
the FFTF are: axial reflector--! .5 x 10'5; gas plenum-0.3 x 1015: and
radial reflector-0.9 x 10^5. it was assumed that the average fluxes
in the corresponding zones in a 1000 MWe reactor would be in the same
ratio to the core flux in both the 1000 MWe LMFBR and FFTF. Since
cross sections for n,Y reactions are high in the gas plenum region,
the activation rates are sensitive to the correct flux in this region.
Also the flux extrapolation from FFTF to a 1000 MWe LMFBR in the gas
plenum region may introduce a large error.
As discussed in Section 5.2.3, the actual energy dependent
neutron flyx was used in the calculations of the 5°Ni(n,p) 58Co and
the 54Fe(n,p) 5fyln reactions in the core. The flux spectrum was
combined with the energy dependent n,p cross sections obtained from
BNL-325.10 This procedure offered an independent check of the HEDL
method (and of the HEDL flux averaged cross section), which was useful
due to the important contributions from these two reactions in the
core. As indicated in Table 5.6 (footnotes c and d), nearly identical
results were obtained using the actual <(>(£) and a(E) as were obtained
using (total) and o .
Flux averaged activation cross sections were taken directly from
the HEDL report,5 as follows:
66
-------
Table 5.7
Corrosion Product
Neutron Cross Sections (Barns)
Zone
Reaction
5-Co(n,Y)60Co
5°-(n,p)60Co
58Mn,p)58Co
54-e(n,P)54Mn
38F£(n,Y)59Fe
54^(n,Y)55Fe
D°:r(n,Y)51Cr
54Fe(n,a)51Cr
i£T^n,Y)182Ta
Core
.122
.0006
.014a
.009a
.012
.012b
.035
.0007
1.12
Axial Reflection
in Blanket
1.73
.001
.004
.003
.042
.086
—
9.18
Gas
Plenum
3.47
.001
.001
.001
.070
—
.150
—
19.7
Radial
Reflector
or Blanket
1.95
Not Determined
.003
.002
Not Determined
—
Not Determined
—
Not Determined
Energy averaged cross sections for a typical LMFBR core spectrum
were also given in References 7 and 8. These cross sections, together
with values for the less important reactions, are given in Table 5.8
for comparison with the HEDL values^ used for the present study.
As described above, the actual cross sections as a function of
5°
energy were used for the 5°Ni(n,p)58co and the
reactions in the core and the results are compared to results
based on these flux averaged cross sections in Table 5.6.
"his cross section for Fe(n,v) Fe comes from Reference 8.
67
-------
Table 5.8
57 8
Conparison of HEDL, AI, and ORNL Cross Sections
Averaged Over an LMFBR Core Energy Spectrum
Cross Section (barns)
Reaction
59Co(n,Y)60Co
60Ni(n,p)60Co
58Ni(n,p)58Co
5JtFe(n,p)5£*Mn
58Fe(n,Y)59Fe
51*Fe(n,Y)55Fe
50Cr(n,Y)51Cr
181Ta(n,Y)182rEa
58Ni(n,Y)59Ni
62Ni(n,Y)63Ni
52Cr(n,2n)51Cr
55Mn(n,2n)51+Mn
59Co(n,2n)58Co
5ttFe(n,a)51Cr
HEDL
0.122
0.0006
0.014
0.009
0.012
0.035
1.12
AI
0.136
0.0004
0.016
0.010
0.017
0.027
2 x 10~5
3 x iO~5
3 x 10~5
0.0007
ORNL
0.012*
0.0003
0.017
0.011
0.008
0.012
0.012
0.007
0.005
1 x 10~5
6 x 10~5
4 x 10~5
0.0006
*Tnis value, as quoted in the OFNL reference, appears to be a factor of 10
too low.
68
-------
The composition and isotopic abundances assumed for the stainless
steel are given in the following table:
Table 5.9
316 Stainless Steel Composition and Isotopic Abundance
Isotopic
Element Weight Percent I sotope Abundance
Fe 65% 54Fe 0.0582
Ni 12 58Fe 0.0033
Cr 18 58Ni 0.6788
Mn 2 50Ni 0.2623
Co 0.02 50Cr 0.0431
Ta 0.01 59Co 1.000
Mo 2 181Ta 1.000
The 0.02% cobalt content was the value assumed by HEDL in Reference 5.
Jhis value might vary significantly from one reactor to another, and the
-'°Co activity is dependent on (nearly directly proportional to) this
number. For example, the GE analysis^ used 0.1% cobalt. RRD standards
specify <.05% cobalt in fuel cladding materials.
5,2.2 Distribution of Corrosion Products in the Primary System
The corrosion products are deposited on surfaces throughout the
v'imary system. Graphical data on deposition rates are given in the GE
analysis." From this data GE estimated where the activation products
were deposited. HEDL quotes some general qualitative experimental
"esultsi^ for example, manganese preferentially deposits in cold
parts of the system whereas cobalt is deposited in the hot parts of the
system. The uncertainty concerning deposition location is illustrated
by comparing, in Table 5.10, the final results of GE and HEDL for
deposition in FFTF components. No GE results were presented for
5lCr,5yFe, and 182Ta.
As discussed in the GE reports, it is unclear whether deposition
can continue at the rate indicated by current experiments after a thick
eeposit has built up on the surfaces. Perhaps the cold traps which will
be removed periodically will trap more of the metal corrosion products than
Indicated in Table 5.10 since the surfaces there will be periodically
renewed. Some metallic products will remain dissolved in the sodium
aid will be part of the coolant activity when the plant is decomissioned.
It is likely, however, that most of the corrosion products will plate
out onto system components or be removed by the cold traps.
69
-------
Table 5.10
Fraction of Nuclides Deposited in Primary System Components
HEDL5 and GE6 Results
Component Nuclide
60r
Co
HEDL GE
.31 —
.249 —
.12 —
.04 —
58,,
Co
HEDL GE
.51
.20
.08
.03
54,.
Mn
HEDL GE
.06 --
.12 —
.08 —
.01 —
51. 59, 182T
Cr, Fe, Ta
HEDL
1.08
.109
.030
.017
Vessel
Hot leg piping,
Vessel-to-pump
Pump
Hot leg piping
Pump-to-IHX
T,,Y Top Half
1MA Bottom Half
Cold leg piping
Cold Trap
Primary tank, inlet —
Bottom shield
5.2.3 Calculational Method
The method used by HEDL for calculation of activated corrosion
product inventories in the primary system was adopted, and the method
is outlined here.
o
The activity (in curies/cm of steel) for a given isotope, j, is:
jC £ \\ •» J. """ • I*
.258
.02
.003
—
_ _ —
!280
.145
.300
.034
.127
.168
.01
.003
.114 ,y
.280 '^
.145 .25
.300 .03
.034 ---
.127 —
.105
.333
.250
.035
.059
.218
.684
.046
.005
—
Y o i T — VI ~\ &
j = 3.7 x 1010M 0.J* 0 - e )
e
(1)
70
-------
where f = weight percent of the element, e.
f. = abundance of reactant nuclide, i.
p = Steel density.
N = Avogadro's number.
M = atomic weight of element, e.
a. = activation cross section (average over energy spectrum) for
1 nuclide i which results in production of nuclide j.
4, = neutron flux (average over space, integrated over energy).
X- = decay constant of radionucl ide, j.
\j
t = duration of reactor operation.
The fuel is replaced after it has been in the core for some
residence time, TR, a time which is short compared to the 30 year life
of the plant. Hence, an average value of corrosion product activity
over the residence time is needed. Assuming X = 0 for fresh fuel, the
average value of X is:
-X,t /r'R e j'R
- e J)dt/y dt = K(l +-—T —4 ) (2)
V^ XJ TR
where K is defined by Equation 1.
In a 1000 MWe LMFBR, it was assumed that the fuel residence time,
TR, was two years in the core (and axial blanket and gas plenum) and
three years in the radial blanket. (Some of the core fuel and the outer
part of the radial blanket would probably have longer fuel cycle
intervals.) In the core, therefore, half of the fuel would be
replaced each year.
An alternate way to derive Equation 2 is to obtain an average
activity over a one year period. For example, assuming two-batch loading
and a refueling interval of one year (TR/2 = 1 year), the average activity
during a typical year is:
Tn 12 TR
-Mt 1 / -A,t
XJ=
— / / """i0 ' / ~AiL
27 0 - e J )dt + 2 Jr „ (1 - e J )dt
TR/2
Jfdt
0
71
-------
which is equivalent to Equation 2.
The activity transported to the primary system, Q(curies), is
obtained from the corrosion rate C and the corrosion product activity
X", as follows:
dQ, __ ,
— J-.'X . CA - A.Q.
dt 1 J J - :
For Qj = 0 at t = 0
Q, *
J j - (i . e
+
3.7 x 10lUM_ TT U A,- T
-X.tx
J )
e Aj J 'R j.'R
where t = total reactor operating time (assumed 30 years for results in
this report)
C = corrosion rate of stainless steel (cm/sec)
o
A = corrosion surface area (cm )
For the work reported by HEDL, a further parameter would be
needed to account for the deviation from the stoichiometric rate of
the corrosion rate of the specific element in stainless steel being
condiered.
For the Ni(n,p) Co and the Fe(n,p) Mn reactions in the core,
Equation 1 was modified so that °j $ was replaced by the integral
4> (E)dE. The cross sections were obtained from BNL-325.10 The flux
was obtained by the method described in Section 5,1.2.2 for the tritium
production rate calculations, i.e. 4> (E) below 2.2 MeV was obtained from
the multi group flux in Reference 9 and (E) above 2.2 MeV was assumed
to be the properly normalized fission spectrum, with the energy de-
pendence:
^ e -E/1.41 MeV.
I QO
5.2.4 Corrosion of Ta from Tantalun Control Rods
Tantalum is being considered as a material for shim control. One
of the disadvantages of the use of Ta is its high activation rate to
72
-------
182T.
i &.
Because of the high activation rate of tantalum, shim rods mace
z~ this material will probably be clad in stainless steel. The GE
1000 MWe follow-on des gn, which includes twelve tantalum shim rods,
specifies that the tantalum would be in large (2 tc 3 inch diameter)
solid rods clad with 5 mil stainless steel. Hence direct corrosion by
sodium would not occur unless preceded by cladding failure.
Little information was found on corrosion rates of tantalum
cy sodium. The HEDL r'eport says that 182ja was found only in small
quantities in LAMPRE, a Los Alamos plutonium fueled, sodium cooled
reactor, which had Ta in the core, and HEDL claims that the corrosion
rate of Ta in low oxygen sodium is known to be quite low.
The total activation rate of tantalum shim rods can be estimated
from tne GE 1000 MWe follow-on report.9 Based on their neutron balance,
tne tantalum midcycle capture rate is 2.5 x 10'° captures/sec, which
corresponds to an equilibrium activity of 7 x 10' Ci. Since the
-alf-"Kfe is 115 days, this activity would remain in the environment
long after the rods were removed from the reactor.
5.2.5 Activated Jorrosion Product Experience at Operating Sodium-
Cooled Reactors
5.2.5.1 Summary
A review article by Zwetzig contains information about corrosion
products in sodium or faK cooled reactors. Table 5.11 summarizes
the corrosion products observed, in the manner which he used, with
adcitions as referencec .
73
-------
Table 5.11
Corrosion Products in Sodium-Cooled Reactors
(Other than Tritium)
EBR-II12'13
Neutron spectrum i fast
Typical outlet temperature 900°F
54 124
Corrosion and Activation Mn, Sb,
products in primary Ag, Sn,
-, , 113mT 117m_
coolant In, Sn
Corrosion products on \ Mn, Co,
18?
primary system surfaces i Ta
I
Corrosion anu activation 'Mn.^^Co
products in cold traps Zn, Sb
14
Rapsodie
fast
54Mn,58Co,
6°CO
65Zn j
i
SRE11,15,16
thermal
1000°F
51Cr,54Mn,
59Fe,60Co
54 59
60r 51 .
Co, Cr
HPNF11
thermal
950°F
54M 60r
Mn, Co
S8ER11
thermal
1300°F
56Mn,60Co
54Mn,59Fe
58Co,6°Co
-------
t,:.5.2 EBR-II
At EBR-II, tne fcllowing products in the primary sodium are
-:requantly monitored: 54Mn, 110mAg, 117mSn, and 112Sn - 113tT1In.
Of these only 54^n comes from corrosion of stainless steel; the
*es~ are peculiar to EBR-II.
I,i December, 1970, the primary pump of EBR-II was replaced and
;ua".itative information on activation products was obtained. Detailed
results were reported in July, 197.1.12 Activation products on the
pump prior to steam cleaning included 22^3, 54^ 60^0, anc| 182Ta-
Also present was the fission product '^'Cs. In steam cleaning the
aump, the following were removed: all of the 22^3j 44-57% of the
°4Mn, 42-92% of the 60Co, and * 65% of the 137 Cs. None of the l8^Ta
was removed. On the uncleaned surface the 182ja activity was less
than, 0.1 of the 54wn activity. On the cleaned surface the activation
products remaining were 54Mn, ^Co, anj 182Ta, listed in decreasing
order of activity. Reference 12 speculated that the source of
tantalum was the cladding of the antimony neutron source used in EBR-I1.
58 59 51
It is noted that the presence of Co, Fe, and Cr was not
indicated in either Reference 10 or in any other of the ANL reactor
development program progress reports.
137
Activation products and Cs were also reported to be present
D"; the inside surface of the reactor tank at the cover-gas-tank
interface.12 Both 54^n anci 60c0 were present. It was postulated that
'•^'Cs vaporized from the sodium and redeposited on the tank wall,
bu; it was not known how the 54^n and 60co got there.
Activated corrosion products were identified in the EBR-II cold
trap from gamma scans.13 These included 54Mn, and 60co. Also
accivation products 22Na, 134Cs, ^In, and '24$b were observed in the
cold trap.
5.2.5.3 Rapsodie
In Rapsodie, the corrosion products Co, Co, and Mn were
observed on the primary pump after three years of operation at 24 MW(th)
{530 equivalent full power days)J2 The ^4Mn activity distribution
.ji"cng the axis of the primary pump is shown in the report. Both
3fv;n and $&Ca were observed on the pipes of the primary system. The
^•v.n was distributed far ly uniformly along the pipes. The 54wn
s^"*ace activity on tne old leg piping (400°C) is 2 to 5 times higher
;nar, on the hot leg pipi g (500°C). Values of 54Mn surface activity
/a-,ec between 0."; and 1 u C
5.2.5.4 SRE
Corroston product contamination was oiserved on the piping walls
of SRE.13 The radionuclides 54Mn, 60Co, and 59Fe were identified.
The activity levels of these nuclides were roughly equal to
75
-------
2
activity level on the piping, ^0.01 y Ci/cm , at shutdown on
July 26, 1959.
As described in Section 7.2.4.4, Table 7.9, corrosion-product
elements were concentrated in the primary cold trap at the SREJ6
However, concentration ratios were not large, ranging from ^ 10 to
\ 100 for Fe, Ni, Cr, and Mn.
5.2.5.5 SEFOR
No data on radioactive transport of corrosion products was
reported for SEFOR. However, a report on cold trap experience at
SEFOR^7 did show some corrosion-product elements in the^ 200 Ibs
of sodium oxide removed from the primary cold traps. The following
concentrations of impurities were listed for the SEFOR cold traps
(Table 5.12).
Table 5.12
Weight Percent of Impurities in SEFOR Cold Traps
Element ppm
Cu 200
Fe 50
Cr 20
Ni 6
C-(carbonate) 240
C 130
The copper was an unexplained surprise. The cold traps were
not radioactive because the traps reported in Reference 17 were removed
and analyzed prior to power operation.
REFERENCES (Section 5.2)
1. G. P. Wozadlo and C. N. Spalaris, "Corrosion of Stainless Steel
and Deposition of Particulates in Flowing Sodium Systems," GEAP-
13544, September, 1965.
2. P. Roy and M. F. Gebhardt, "Corrosion and Mass Transport of Stainless
Steels in Sodium Systems," GEAP-13548, September, 1969.
3. W. F. Brehm, et al., 'Radioactive Material Transport in Flowing
Sodium Systems", in Cc.rrosion by Liquid Metals, Draley and Weeks,
editor, pp 97-113, PTl num Press, New York, 1970.
76
-------
4. A. W. Thorley and C. Tyzack, "The Corrosion Behavior of Steels
ana Nickel Alloys in High Temperature Sodium." Proceedings of
Symposium on Alkali Metal Coolants, IAEA, Vienna, Austria, November,
1966.
5. I. u. Kabe'ie, W. F. Brehm, D. R. Marr, "Activated Corrosion Product
Raaiation Levels Near FFTF Reactor and1 Closed Loop Primary System
Components," HDL-TME 72-71, May, 1972. (Also identical results
presented in Trans. An.. Nucl. Soc.. 15, No. 1, June, 1973.
5. G. P. Wozadlo, C. E. Boardman, and M. L. Weiss, "Calculated
Radioactivity of the FFTF Primary Sodium System Due to Mass Transfer,'
GEAP-13671, August, 1971.
7. G. B. Zwetzig and R. F. Rose, "Interim Description of a Computer
Code (STP-1) for Estimating the Distribution of Fission and
Corrosion Product Radioactivity," AI-AEC-12847, June, 1969.
8. Staff, Chemical Technology Division, ORNL, "Aqueous Processing
of LMFBR Fuels" Technical Assessment and Experimental Program
Definition," ORNL-4436, June 1970.
9. "Task-II Report, Conceptual Plant Design, System Descriptions,
and Costs for a 1000 MWe Sodium-Cooled Fast Reactor," GEAP-5678,
December, 1968. ''»
t ' ''
"10. BNL-325, Second Edition, Supplement No.2, "Neutron Cross Sections,
Vol. 1, Z = 21 to 40," J. R. Stehn, M. D. Goldberg, B. A. Magurno,
and R. Wiener-Chasman--Editors, May, 1964.
"il. G. B. Zwetzig, "Survey of Fission and Corrosion-Product Activity in
Sodium-or NaK-Cooled Reactors," AI-AEC-MEMO-12790, February, 1969.
12. Reactor Development Program Progress Report, ANL-7845,,p. 1.15,
July, 1971.
13. Reactor Development Program Progress Report, ANL-RDP-7, p. 1.10,
July, 1972.
14. R. de Fremont, "Observations on the Behavior of Radioactive
Products in Rapsodie," DRNR/STRD, 71.1146, 1971.
15. R. S. Hart, "Distribution of Fission Product Contamination in the
SRE," NAA-SR-6890, March, 1962.
16. A. I. Hansen, "The Effects of Long-Term Operation on SRE Sodium
Systems Components," NAA-SR-11396, August, 1965.
7. A. D. Gadeken and M. C. Plummer, "SEFOR Cold-Trap Experience,"
GtAP-10548, April ,1972.
77
-------
- -i Activation Products
:;uo,l Sodium Activation
SoGium has the disadvantage that it is activated by neutrons. The
principal activation product is ^Na, formed by the absorption of a
neutron by ^Ha. A second activation product is 22Na also formed from
in an n,2n reaction. Although the 2^Na activity is far greater
than the Na activity during reactor operation, Na decays with a
14.7 hour half life while 22Na has a 2.6 year half life. Therefore in
considering the long term environmental effects of storage of sodium after
:ts use in an LMFBR, or the effects from dispersion of sodium by a
sodium fire, the long half-life isotope 22Na -js more important than
24Na. The 22Na activity becomes greater than the 2^Na activity about
ten days after reactor operation ceases.
5.3.1.1 Sodium- 24
Sodium-24 in the Primary System
24
Calculation of the activity of Na is straightforward since it
does not result from a threshold reaction. A typical calculation
is provided for the General Electric 1000 MWe conceptual design described
in Reference 1. The equilibrium activity produced in the primary
coolant is 2 x 10' Ci . It would have been of interest to compare
shls result with a result extrapolated from EBR-II, but Na activity in
the primary system is not reported in the ANL Reactor Development
Program Reports.
Sodium-24 Activity in the Secondary System
2* Sodium-24 can enter the secondary system in two ways—by leakage of
Na from the primary to the secondary system through small leaks in the
•intermediate heat exchanger and, in a pot-type LMFBR design, by direct
activation of the secondary sodium. It should be noted here that
leakage of sodium from the primary to the secondary system would normally
be minimized by controlling the primary system pressure lower than
the secondary so that leakage would be in the other direction.
Activation of secondary sodium in pot-type designs will be made
u.all by shielding the secondary sodium loop from neutrons. No
estimate of secondary sodium activation was available from conceptual
sesign reports, and both FFTF and the Demonstration Plant will be loop
resigns instead of pot designs.
24
EBR-II is a pot-typ. design and some secondary Na activity is
reported. Early values reported in the ANL Reactor Development Progress
Reports were corrected for an earlier calibration error in Reference 7.
During 1972, the 2^Na activity in the E! R-II secondary system varied
from 8.6 to 38nCi/gm, with an average value of ^20 nCi/gm. The 2a
secondary sodium inventory is 6 x 10' gm Hence the total secondary Na
activity is of the order of 1 Ci.
78
-------
5.3.1.2. Sodium- 2 2
23 22
Soaium-22 results from the threshold reaction, Na(n,2n) Na, the
threshold for the reaction being 12.5 MeV. On decay a 1.28 MeV
gamma and two 0.5 MeV gammas are released from positron annihilation.
22
Two estimates of Na activity produced in a 1000 MWe LMFBR were
compared. The first estimate is an extrapolation from measured data
in EBR-Ii. The second is a result quoted by General Electric in their
1000 MWe follow-cr, report. These two estimates disagreed so greatly
a third value was calculated for this report in order to try to under-
stand the possible source of the disagreement. (A third comparison
was possible, based on extrapolation from a calculation for SEFOR.
However, the methods used for the SEFOR calculation were likely the
same as for the GE 1000 MWe calculation so that the extrapolated
result is not necessarily an independent calculation.)
EBR-II Extrapolation
Sodium-22 activity in the primary system of EBR-II has been
measured. Knowing the "Na activity in EBR-II, the power level, the
operating time, and the load factor, one can extrapolate approximately
to the activity produced in a 1000 MWe LMFBR.
The Na activity in November, 1972, in EBR-II was 60 nCi per
gram of sodium in the primary system. ^ The primary system contains
90,000 gallons of sodium. The accumulated exposure at this time was
60,309 MWd.2 During the year prior to November, 1972, the exposure
was 10,446 MWd, 3 and the exposure had been fairly constant for the
last four years. The power level has been 62.5 MW(th)* since 1970,
before which it was 50 MW(tn). Assuming that reactor operation has
always been the same as the year prior to November, 1972, the load
factor for EBR-II would be 0.46 and the chronological time for operation
would be 5.75 years at 62.5 MWth.
The total activity, A, in November, 1972, was 60 nCi/gm in
90,000 gallons of sodium, or 17 Ci . This activity and the equilibrium
activity, A , are related by
CD . ,
A = fA^ (1 - e "X22t)
f = load factor
22
= Na decay constant
"•ince completing this investigation, it was learned that since about 1972
tne actual power level has been about 9% below 62.5 MW(th), or approximately
57MW(th), even though 62.5 MW is still quoted as the "nominal" power.
:*o modifications were made to the EBR-II calculated results given in
this report to account for this lower power level.
79
-------
t = chronological operation time
Hence A = 47 Ci .
77
Next assume that the Na equilibrium activity is proportional to
the power level. The geometry and sodium volume fractions in EBR-I1 and
a power reactor are different, but still the extrapolation is expected
to be a reasonable approximation. Also the fission spectrum for the
235y fue-| Of EBR-II is sufficiently similar to the fission spectrum
for the Pu fuel in a power reactor, even above the 12.5 MeV threshold
of the "Na(n,2n)"Na reaction, that the different fuel does not
introduce a large uncertainty in the extrapolation. One can extra-
polate the activity after long operation in a 1000 MWe LMFBR (2500
MWth, 0.8 load factor) to:
A= (47 Ci) (.8)
" 1500 Ci
77
It is of interest to note that the Na activity in EBR-II is
manner consistent with the equations utilized here.
lovember, 1971, the "Na activity was 54 ^-, and the
increasing in a
For example in November
gm
ra^io of 60 nCi/gm to 54 is about equal to the ratio of (1 - e
-5.75X
22) /
(1 - e~4-75A22), where 5.75 yr and 4.75 yr are the chronological opera-
ting times for November, 1971 and November, 1972 respectively.
(A22 = -266 yr-"1)
GE 1000 MWe Follow-on Value
The result for the 1000 MWe follow-on jdesign reported by General
Electric is:' t /
1 , ' \
A = 9600 Ci , :. . ]
SEFOR Extrapolation
77
The calculated "Na equilibrium activity for a load factor of
unity is 65 Ci4. The SEFOR power level was 20 MWth. Extrapolation to
a 1000 MWe LMFBR at .8 load factor gives
A = 6500 Ci
This value is closer to the 1000 MWe follow-on result than the
EBR-II extrapolation, but both are based on General Electric calculations
which may have used the same "Na(n,2n) cross sections.
Present Calculation
77
Since the above results varied so widely, the "Na activity was
calculated from the fission spectrum and the ^^A(n,2r\) cross section.
80
-------
The activity was estimated from the equation:
where the subscript k refers to the region—core and blanket. The flux
was obtained from the fission spectrum, normalized to the first group
of a multi group flux distribution. Details of the first-group flux
calculation are given near the end of this section. The fission spectrum
was represented by
*(E) = /FTe~E/T
where T = 1.41 MeV for 239Pu. The cross section for the 23Na(n,2n)22Na
reaction was obtained from BNL-3255 (which used the 1963 data of Picard
and Williamson), although a large margin of error was possible in
reading the cross section as a function of energy.
pp
The calculated equilibrium "Na activities generated in the core and
blanket were
A(core) = 2.4 x 103 Ci
A(blanket) = 0.3 x 103 Ci
22
she activity of Na produced outside the blanket was not calculated.
It is probably small since the neutron leakage is small from the
blankets, and in calculating the blanket flux (see next section for
details) no leakage was allowed. There may be leakage of uncoil ided
high energy neutrons into the sodium pool , however, which might provide
a "Na source that is not negligible. In SEFOB, which was a small
^eactor with thin reflectors, the calculated 22|^a production outside
the reflectors was ^25% of the total.
Based on the above results, the total Na activity from a 1000
MWe LMFBR is estimated to be:
A % 3000 (1 - e"A22t) curies
One further estimate is possible, based on the recalculation of
the integral
f a(E) * (E)dE
H2.5 MeV
81
-------
A value of 8.9 x 10" barns MeV was obtained for this integral
(see next section). For SEFOR the calculated yalu* for this integral
was reported4 as 2.3 x 10 . Since the SEFOR 22na production rate,
together with the extrapolation of 6500 Ci for the 1000 »te LMFBR
from the SEFOR numbers, wls based on the SEFOR value for this integral,
one would be justified in renormalizing the SEFQft extrapolation to
our calculated integral. This renormalization gives:
A = 6500 x 8'9 x 10". = 2500 Ci
2.3 x 10"b
which is close to both the above estimated result of 3000 Ci and the
result of 1500 Ci extrapolated from EBR-II.
Further Details of the 22Na Activity Calculation
The sodium-22 activity is given by
N
23,b
12.5 12.5
MeV MeV .'
where ^3 c = total # of atoms of Na-23 in the core
N23 b = total # °f atoms on Na-23 in tue blanket
. (E) = space averaged, energy dependeit flux in the blanket
°n,2n = n,2n reaction cross section
pp 1
\22 = decay constant for "Na = .266 yr"1
t = chronological time since initiation of reactor operation
f = load factor
4>(E) was obtained as follows: The high energy neutron flux is
proportional to the fission spectrum, if»(E), where
where T = 1.41 MeV for Pu-239 fission
Hence 4>(E) = oup (E) = a/E~e~^-^ at high energy.
The integrals in Eq'n (1) were rewritten, following the method
of reference 4, to obtain:
82
-------
f ,.-, ,r> ... f , . =r -E/1.4I.- 0 Q in-6 /bcrn-neutvons
i o o It; 9 (tjut = a I a 9 \E)/E e uE = 8.9 x 10 a( y~^ 7
j n,^n J n,^n Cnr - sec
12.5 12.5
MeV MeV
(2)
The constant a was determined as follows. In terms of group 1 of
a iTiultl group flux structure,
^ was estimated for a group structure for which E , = 3.7
MeV. The flux in the core -)core = total fission/sec in the core
i
and the subscript 1,c refers to group 1 of the core.
For the blanket the equation for cj>i ^ was:
/. 4 X""* r o. \ u. n D £~JL~
where the sources were both fission in the blanket and group 1
leakage from the core, and no leakage is assumed from the blanket.
The calculated fluxes were:
*]jC = 0.9 x 1014
^-, b = 0.8 x 1013
00
The integral in Eq'n (3) is f ^(E)dE = 0.23
3.7 MeV
83
-------
Hence the proportionality constants, a, are:
a(core) = 4 x 1014
a(blanket) = 3 x 1013
The sodium-22 activities generated in the GE 1000 MWe Follow-on core
and blanket are:
^ (N23 ) (9 x 10'V)f(l - e'^ZZ*)
A(core) = ^J'c £
3.7 x 1010 dis/sec Ci
= 2.4 x 103 (1 - 3 22 ) Ci
based on the following numerical values:
?
.. „ , In4 atom-cm
Noo _ = 3.1 x 10
barn
from: core volume = 3.7 x 10^ cm3
sodium volume fraction =0.37
sodium density = 0.85
9x10"° a = 3 6 x IP9 ba£n-neutrons
J n I \J **" *J * \J A I \J T
c cttr - sec
f = 0.8
A(blanket) = .26 x 103 (1 - e ~X22t) Ci
A(total) = 2.7 x 103 (1 - e "A22 ) Ci
84
-------
REFERENCES (Section 5.3.1)
1. "Task 1 Report of 1000 MWe LMFBR Follow-on Work," GEAP-5618,
p. 252, June, 1968.
2. Reactor Development Program Progress Report, ANL-RDP-11, November,
1972 (Limited-distribution report).
3. Reactor Development Program Progress Report, ANL-7887 November, 1971.
4. A. B. Reynolds and D. F. Sudborough, "SEFOR Shielding Nuclear
Design Calculations," GECR-5199, p. 7-4 to 7-6, March, 1967.
5. J. R. Stehn, M. D. Goldberg, B. A. Magurno, and R. Wiener-Chasman,
"Neutron Cross Sections," BNL-325, Second Edition, Supplement No. 2,
May, 1964.
6. "Task II Report, Conceptual Plant Design, Systems Descriptions, and
Cost for a 1000 MWe Sodium-Cooled Fast Reactor," GEAP-5678,
December, 1968.
7. Reactor Development Program Progress Report, ANL-RDP-9, September,
1972 (limited distribution report).
5.3.2 Cladding Activation
The stainless steel structure in and near the core of an LMFBR
becomes activated. The only part of this activation considered in this
report is the fuel cladding activation, which is to be shipped from
the site with the fuel.
V ;
For a fuel assembly design that has s^eeV hexagonal cans enclosing
the entire assembly (as in present FFTF and demonstration plant design),
the steel can would contain from 50 to 100% as much steel as the cladding.
The specific activity of this steel would be equal to that of the
cladding. Hence the total steel activity shipped from the site would be
nearly a factor of two higher than the cladding activity alone.
The cladding activity for the AI 1000 MWe Reference Oxide Design
was calculated by ORNL, and these values are reproduced in Table 5.13.
These values were checked for order of magnitude by using cross
sections and fluxes from Section 5.2 (Activated Corrosion Products) and
approximate cladding volumes. The results agreed within factors of
2 to 3, which was considered acceptable for the purpose of this review.
The fuel mass discharged annually from the AI design was 8.517 MT
(U + Pu) (see Table 4.1). The cladding activity discharged annually
with the fuel is therefore 8.517 times the totals in Table 5.13.
These results are given in Table 5.14.
-------
Table 5.13
CladdiJig Activity of Spent Uff ;3R
Core Fuel as a Function of Cooling Time
Activity [Ci/MT(U + Pu)]
Isotope
5'cr
51*Mn
sSfe
59Fe
58Co
60Co
5*Ni
63m
Total
Cooling Time
0
6.17x10"*
l.SOxlO5
7.60X101*
1.03x10"
4. 10x10 5
1. 35x10 3
2.1
67.8
7. 09x10 s
30d
2.92x10**
1. 40x10 5
7.43X101*
6.52X103
3. 07x10 5
1. 33x10 3
2.1
67.7
5. 58x10 5
90d
6.55xl03
1.22xl05
7.12X101*
2. 59x10 3
1.71xl05
1. 31x10 3
2.1
67.6
3.75xl05.
150d
1. 47x10 3
1. 07x10 5
e.sixio1*
1. 03x10 3
9. 55x10 "*
1. 28x10 3
2.1
67.6
2. 74x10 5
300d
35
7.50x10"*
6.00x10"*
l.OOxlO2
2. 26x1 01*
1.20xl03
2.1
67.5
1 .59xl05
30yr
25.5
25.9
2.1
54.1
108
Half-life
28 d
313 d
2.4 yr
45 d
71 d
5.2 yr
SxlO1* yr
92 yr
Table 5.14
Cladding /Activity Discharged Annually Fran
1000 MHe LMFH
Activity (Ci)
Cooling Time
0
6.0xl06
30d
4.8xl06
90d
3.2xl06
150d
2.3xl06
300d
1.3xl06
30 yr
900
86
-------
REFERENCES (Section 5. .2)
L Staff, Chemical Technology Division, ORNL, "Aqueous Processing
of LMFBR Fuels: Technical Assessment and Experimental Program
Definition," ORNL-4436 (June, 1970).
5.3.3 Activation Products 39Ar, 41Ar. and 23Ne
5.3.3.1 Argon-39
Argon-39 has a 269 year half life and undergoes 3~ decay (0.59 MeV
maximum 6, no v), It is produced from 39K -jn ^ne sodium coolant from
the reaction: 39|<(n,p) 39Ar. Although no reported observations of 39Ar
were found from operating fast reactors, Reference 1 estimates that
0.13 Ci/day of 39Ar would be produced in the 975 MWe Clinch River Breeder
Reactor (CRBR) if the potassium impurity concentration in the coolant
were 1000 ppm. For reference, some potassium concentrations have been
- 160 ppm in EBR-II (e.g. Ref. 2, and Table A.28 of this report)
and ^ 300 ppm in SEFOR3. Our independent check on 39Ar activity agrees
with the Reference 1 value to within 50%, whi( h is within the accuracy
;:f the (n,p) reaction cross section available''. Extrapolating from
•c:ie CRBR value, and assuming 300 ppm potassium impurity in the coolant,
gives a 39Ar activity production rate of ^ 30 Ci/yr for a 1000 MWe LMFBR.
5 3.3.2 Argon-41
This radionuclide has a 1.83-hr half-life and undergoes negatron
cecay to 4 K. 4'Ar can be produced in the LMFBR by two mechanisms:
" 1K(r,,p)41Ar and 40Ar(n,y)4lAr. The EBR-II staff suspects that the
•"irst reaction is the principal production source in EBR-II.5 EBR-II
cover gas typically contains about 1.5 uCi of 4'Ar per liter of cover
gas. Rapsodie was reported^ to have a cover gas 4'Ar content of
200 yCi/* and BR-5 a cover gas 41Ar content3 of 100 yd/A .
5.3.3.3 Neon-23
This radionuclide has a 37.6 sec half-life and undergoes negatron
decay to 23[\|a. 23^ -js produced by the following reaction
Na(n,p) Ne
8 23
SEFOR was reported to have a specific Ne cover gas activity of f
.:£,000 yCi/£ when operating at 10MW. With a cover gas volume of 0.56 x 10
cc, the total cover gas activity was 16.4 Ci from 23^e> yne 23^
cccivity in the SEFOR core sodium was estimated to be 5700 Ci at 10 MW.
"Ga these numbers, the average time for disengagement of the 23Ne atoms
rrorn the sodium was determined to be 5.5 min.
fi po
Rapsodie was reported to have a cover gas Ne content of 10,000
jCi/£ . Both the primary coolant and tne cover gas of the #2 primary
pump were analyzed for 23^e at BR-5.7 Tie sodium contained 500,000
and the pump cover jas contained 700,000
37
-------
REFERENCES (Section 5.3.3)
1. I. A. Nemzek, transmittal of Feb. 21, 1974, to universities, of
WARD document, "Assessment of the Demonstration Plant Design
Decisions", January, 1974.
2. Reactor Development Progress Report, ANL-7845, July, 1971. |
3. G. Billuris, General Electric Breeder Reactor Department, ^
personal communication of a 1967 analysis. f
*"t
4. J. R. Stehn, M. D. Goldberg, B. A. Magurno, and R. Wiener - " '>
Chasman, "Neutron Cross Sections", BNL-325, Second Edition, «
Supplement No. 2, May, 1964 J
!7
5. EBR-II Staff, personal communication, Dec. 1972. I,
6. CEA-R-3626 October, 1968.
7. V. V. Orlov, M. S. Pinkliasik, N. N. Aristarkhov, I. A. Efinov,
A. V. Karpov, M. P. Nikulin, "Some Problems of Safe Operation of
the BR-5 Plant," Paper Va-7, Proceedings of the Int. Conf. on the
Safety of Fast Reactors. Aix-en-Provence, Sept. 19-22, 1967.
8. J. J. Regimbal, W. P. Kunkel, and R. S. Gilbert, "Measurement of
Noble Gas Transport Dynamics in SEFOR Sodium," Trans. Am. Nucl. Soc. 14, "I
No. 2, 773-774. ~ <4
-------
5.3.4 Miscellaneous Activation
5.3.4.1 From Fission Products
A number of radioactive nuclides are produced from activation (n,Y
reactions) of fission products. In this report, these radionuclides
are included under "fission products." They are listed in Tables 4.3-4.6,
and indicated by the footnote "a." The "activation fission products"
of importance are: 86Rb, 110mAg, 110Ag, 113mCd, 134Cs, 148mPm, 154Eu,
and 150Tb.
5.3.4.2 From Impurities in Sodium Systems
A number of miscellaneous activation products are reported from
sodiurn-cooled-reactor operating experience. These are frequently peculiar
to the particular reactor system and are generally low in activity level.
Such activation products observed in EBR-II include In, Sb, Sb,
113sn, llSmSn, 117msn, l^Ag, and 210p0 (see Appendix A Table A25). The
^'°Po comes from the bismuth (210Bi) in the tin-bismuth seal in the
EBR-II cover. It is presumed that the 125Sb, 113Sn, 113mlh, and 117mSn
come from activation of the tin. It is unclear whether the ^'^Ag results
from activation of a fission product or activation of a silver impurity
in the sodium. Also the source of ^24Sb is unclear—perhaps from activa-
tion of ^23$b in the sodium. The 6$In presumably comes from activation
of the ^4Zn present in the EBR-II sodium, although, like Sb, zinc is not
listed as a trace metal in the sodium in EBR-II reports.
In Rapsodie, Po was also observed. Like EBR-II, Rapsodie has
a tin-bismuth cover seal. Also 65zn wa. observed in Rapsodie-
In SRE, Sb was observed. The ' Sb was reported to come from
activation of 124Sn. '10Ag was observed in the cold trap of SRE, but
the source is unclear.
In SEFOR, Ag and Sb were both observed in the sodium (see
Appendix A, Table A22). S nee the only other fission product observed
in SEFOR was 86Rb, and sin
-------
5.4 Tramp Fuel
/
Tramp fuel is the term used to describe fuel material presentxon the
outer ^surface of fuel pins as a contaminant from the process of/fuel
fabrication. When this tramp fuel is exposed to neutrons in the core
it becomes a source for direct introduction of fission products into
the primary coolant. • ' /
/
• \ . f'
The exact amount of tramp fyel .present in any core will be a
function of many things but primarily a function of the fuel element
fabrication pVocess.
Estimates of the amounts of tramp fuel in SEF()R and in EBR-II2'3
have been made. \ Neither of these reactors utili/e fuels typical of
those which will\be found in large LMFBR's. Tlj£ EBR-II fuel is metal
rather than oxid'e\and is therefore fabricated/In a significantly
different way from\oxide fuels. The SEFOR fuel is indeed oxide fuel,
but is of significantly different diameter/than will be used in future
LMFBR's. \ /
\ /
Despite the norpratotypic fuel in EfiR-II and SEFOR, the tramp fuel
information from these reactors will be used to estimate tramp fuel
inventories in a 1000 MWe LMFBR. The basic assumption for extrapolation
from EBR-II and SEFOR data is that/the mass of tramp fuel per unit
length of fuel pin in the core will be similar for the large LMFBR.
5.4.1 SEFOR
The total tramp fuel inventory in SEFOR has been estimated as 0.2 mg
of fissile material, or about 1 mg of heavy metal fuel atoms. Based
on a total pin length estimated of 1740 ft., the inventory per foot
of fuel pin in the core/is 6 x 1Q~4 mg/ft.
5.4.2 EBR-II
9 9^
E. R. Ebersole/has estimated the U in tramp fuel in EBR-II
to be 2 mg based ,dn the normal tramp background of 133Xe and '35Xe
observed in the/cover gas. 6. S. Brunson3 claims an inventory of
roughly 7 mg of unclad "5U in the core. Since the EBR-II fuel is
enriched to about 50% 235u, a range of 4, to 14 mg of fuel is indicated
for the amount of tramp fuel in EBR-II. \0n the basis of 6910 ft. of
fuel pins >n the core5, a corresponding range of 5.79 x 10~4 to 2.03 x
10"3 mg/ft' can be determined.
5.4.3 .Rapsodie
/
HO direct data are available for tramp fuel in Rapsodie; however,
a tramp fuel inventory for Rapsodie which is in general agreement with
that of EBR-II can be inferred from other published information as
follows: Differences in fast fluxes, cover gas volumes, fissile fractions
0f core fuel, and total length of fuel pins in the EBR-II and Rapsodie
cores tend to cancel. Therefore, ignoring cold trapping of precursors,
^ N \ / '
V, '•-, """ 90
-------
5.4 Tramp Fuel
Tramp fuel is the term used to describe fuel material present on the
outer surface of fuel pins as a contaminant from the process of fuel
fabrication. When this tramp fuel is exposed to neutrons in the core
it becomes a source for direct introduction of fission products into
the primary coolant. •
The exact amount of tramp fuel present in any core will be a
function of many things but primarily a function of the fuel element
fabrication process.
1 03
Estimates of the amounts of tramp fuel in SEFOR and in EBR-II '
have been made. Neither of these reactors utilize fuels typical of
those which will be found in large LMFBR's. The EBR~II(fuel is metal
rather than oxide and is therefore fabricated ir> a significantly
different way from oxide fuels. The SEFOR fuel is indeed oxide fuel,
but is of significantly different diameter than will be used in future
LMFBR's.
Despite the nonprototypic fuel in EBR-II and SEFOR, the tramp fuel
information from these reactors will be used to estimate tramp fuel
inventories in a 1000 MWe LMFBR. The basic assumption for extrapolation
from EBR-II and SEFOR data is that the mass of tramp fuel per unit
length of fuel pin in the core will be similar for the large LMFBR.
5.4.1 SEFOR
The total tramp fuel inventory in SEFOR has been estimated as 0.2 mg
of fissile material, or about 1 mg of heavy metal fuel atoms. Based
on a total pin length estimate^ of 1740 ft., the inventory per foot
of fuel pin in the core is 6 x 10-4 mg/ft.
5.4.2 EBR-II
?
E. R. Ebersole has estimated^ the • U in tramp fuel in EBR-II
to be 2 mg based on the normal tramp background of xe and
observed in the cover gas. G. S. Brunsorr claims an inventory of
roughly 7 mg of unclad "by ^ £ne core> since the EBR-II fuel is
enriched to about 50% 235|j, a range of 4 to 14 mg of fuel is indicated
for the amount of tramp fuel in EBR-II. On the basis of 6910 ft. of
fuel pins in the core^, a corresponding range of 5.79 x 10~4 to 2.03 x
10~3 mg/ft can be determined.
5.4.3 Rapsodie
No direct data are available for tramp fuel in Rapsodie; however,
a tramp fuel inventory for Rapsodie which is in general agreement with
that of EBR-II can be inferred from other published information as
follows: Differences in fast fluxes, cover gas volumes, fissile fractions
of core fuel, and total length of fuel pins in the EBR-II and Rapsodie
cores tend to cancel. Therefore, ignoring cold trapping of precursors,
90
-------
Page Intentionally Blank
-------
the cover gas specific activities due to tramp fuel should be similar.
Indeed this is. found to be true. The '^xe activity in the EBR-II
cover gas at saturation from tramp fuel2»3 -js about 3 x 10~3 uCi/cc.
The measured saturation activity of 135xe in the RAPSODIE cover gas
right after initial startup6 was 1 x 10~2 yCi/cc.
5.4.4 Extrapolation to 1000 MWe LMFBR
The information on tramp fuel inventories presented above suggests
the use of a value of 10-3 mg of heavy metal fuel atoms per foot of
fuel pin in the core. Large LMFBR's will contain tens of thousands of
fuel pins having total lengths of tens of miles. For an average linear
power density of 9 kW/ft, the total core fuel length is ^ 250,000 ft.
Hence, a total inventory of 0.25 gm of tramp fuel is estimated.
The coolant fission product inventory due to this load of tramp fuel
can be estimated using the data of Reference 7. If no deposition or
other removal mechanism is assumed for fission products which enter
the primary sodium, about 300 curies of fission product activity could
be present in the sodium for the equilibrium fuel cycle. Most of this
activity would be from short-lived isotopes. The long-lived isotopes
would eventually contribute an activity of a few tens of curies if
the same primary sodium were utilized throughout a normal plant life.
The corresponding upper limit activities for the higher actinides
would be 20 to 30 curies for the equilibrium fuel cycle with a buildup
to a few curies of long-lived actinides over the plant life. It is
important to emphasize that these estimates take no credit for cold-
trapping, plating out, or other removal mechanisms for the radioactive
nuclides.
The activities discussed above are quite small in magnitude
compared to that from activation of primary sodium, impurities in the
sodium, and corrosion products in the socium.
REFERENCES (Section 5.4)
1. J. J. Regimbal, R. S. Gilbert, W. P. Kunkel, R. A. Meyer, and C. E.
Russell, "Fuel Failure Detection Capability at SEFOR," Trans. Amer.
Nucl. Soc., 14, 69 (1971)
2. R. R. Smith, et al., "Effects of Driver-Fuel Cladding Defects on
the Operation of EBR-II," ANL-7787, Feb. 1972.
3. G. S. Brunson, "On-Line. Noble Gas Fission Product Monitoring
in EBR-II," Nucl. Tech., 10, 33 (Jan. 1971).
4. Massoud T. Simnad, Fuel Element Experience in Nuclear Power
Reactors, Gordon and Breach, New York, 1971, p. 596.
5. G. S. Brunson, R. M. Fryer, and R. V. Strain, "Post-Shutdown
Surges in Cover Gas Activity in Experimental Breeder Reactor II
(EBR-II)" Nucl. Tech. 13, 6 (Jan. 1P72).
91
-------
6. M. Chapelet, et al., "Experimental Study of the RAPSODIE Pro-
tections," CEA-R-3626, Oct. 1968. (In French).
7. "Aqueous Processing of LMFBR Fuels: Technical Assessment and
Experimental Program Definition," ORNL-4436, June, 1970.
6. TRANSPORT OF FISSION PRODUCTS FROM FAILED FUEL
6.1 Introduction
The task of predicting activity releases to the primary coolant
and cover gas from the fuel of an operating LMFBR is extremely
complicated. At this time there is no consensus of opinion on almost
every major question that arises in answering questions like the following:
What fraction of the noble fission gases are released from a mixed
oxide pellet operating under specific conditions? What are the release
times for the gases from the pellet? What pin failure rate will be
observed? How can the failures be characterized, e.g. size, location,
time of occurrence, etc. For large cladding failures, how much of
each of the various radionuclides will be leached from the exposed
fuel by the flowing sodium?
These and other similar questions need to be answered before realistic
release predictions can be made. This section contains partial
answers to all of the questions listed above, and more. The fuel
testing which will accompany operation of the FFTF should provide better
answers. There is a considerable amount of art (engineering judgement)
involved in the prediction of fuel performance today, due to limited
experience.
Radioactivity releases from intact non-vented fuel in normally
operating LMFBR's will be limited to tritium. Any non-vented fuel
which is not intact will be categorized here as failed fuel; this
includes either "leaky" fuel or fuel exhibiting gross cladding failure.
These two types of failures will, of course, have significantly
different consequences. The distinction between small holes in
cladding and relatively large failures has been made with varying
degrees of consistency in available literature on fuel failures,
thus making it difficult to interpret reported failure rates.
Release from vented-to-coolant fuel will be treated separately,
although much information from venting tests (e.g. holdup times, etc.)
can be important in defining release rates from non-vented fuel.
The problem of activity release from fuel pins will be handled in
two main parts: release of activity from the fuel proper, i.e. from
the pellet, powder, etc.; and release of activity from the pin itself
to the primary coolant. Each of these main parts of the problem will
involve several subproblems and phenomena.
92
-------
6.2 Brief Background Description of Irradiation Experience Relating
to Fission Product Release"
The FFTF, the LMFBR Demonstration plant, and probably the first
generation of large ( ^1000 Mwe)LMFBR's will use mixed oxide fuels
which are about 20% Pu02 by weight and 80% UO?. These fuel pins
will probably consist of mixed oxide pellets and helium bond gas in
an austenitic stainless steel cladding. Later reactors may employ
so-called advanced fuels (carbides and nitrides instead of oxides),
but the choice of oxide fuel for early plants is governed by the
much more extensive experience and testing that has been achieved
with oxide fuels, as opposed to carbide or nitride. However, even
though there is considerable information available on oxide fuels,
much uncertainty still exists about oxide performance under the
extreme operating conditions forecast for LMFBR fuels.
Extremely high burnups ( ^100,000 MWd/MT)* and fast fluences
( <\,3 x I023n/cm2) are the goals for LMFBR fuels. The fuel pins
would undergo these irradiations while operating at peak linear
powers of the order of 18 kw/ft with corresponding fuel surface and
centerline temperatures of 1800°F and 4900°F, respectively, in flowing
sodium with velocities of the order of 25 ft/sec and maximum outlet
temperatures of 1300°F.1>2
The conditions described above have not been simultaneously
achieved for fuel tests to date. Moreover, even while falling
short of prototypic test conditions, problems have been encountered
in oxide fuel testing which have not been completely solved and which
may prevent achieving original performance goals.3
The main reason that oxide fuel problems have not been solved is
the lack of appropriate test facilities. The fast fluxes of sufficient
magnitude needed to achieve ^3 x lO^n/cnr have not been available.
EBR-II and DFR require over three years to accumulate desired fast
fluences. The FFTF will provide higher fast fluxes (achieving the
desired fluence in about 1.5 years) for testing fuels, but its own
driver fuel will be operating at slightly lower specific power and
lower coolant, cladding, and fuel temperatures than proposed LMFBR's.
Still, many of the limitations indicated today in fuel performance
should be better quantified by FFTF programs. (Design burnup -- i.e.
>_ 100,000 Mwd/MT -- has been readily achieved in low power fast test
reactors like EBR-II and DFR and in thermal test reactors by enriching
the uranium portion of the U02 - Pu02 fuel. However, the combination
of design fast neutron fluence at the cladding and fuel burnup cannot
be readily achieved.)
* Experimental fuel exposures appear in the literature both as atom
percent burnup and as megawatt days per metric ton. The conversion
factor is almost exactly: 1 atom % = 10,000 MWd/MT.
93
-------
Fuel swelling and release of fission gases with resulting
cladding stress due both to fuel-cladding mechanical interaction and
to high fission gas pressures were early recognized as problems.4
Satisfactory mitigation of the effects of these two phenomena can
probably be achieved by (1) incorporating sufficient voidage within
the fuel proper to accommodate solid fission products and any un-
released fission gases,5 and (2) by providing sufficient plenum
volume (or venting) to handle gases released from the fuel. However,
knowledge of the amounts of gas released from the fuel but remaining
within the pin is important in terms of potential releases.
Two unanticipated problems with the stainless steel cladding were
discovered: the significant swelling of stainless steel on irradiation
to high fast fluences ,6 and irradiation induced creep.7>8 Austenitic
stainless steels irradiated to fast fluences of the order of
7 x 1 O22n/cin2 have been observed to exhibit volume increases of as
much as 7%.6 Techniques such as heat treating prior to irradiation
may reduce such swelling significantly;^ however, even with the data
on control thimble swelling of EBR-II, extrapolation by factors of
two on fluence must be made.
Creep rates of 304 stainless steel in a fast flux of only 2
5 x 1012n/(cm2sec) and a simultaneous thermal flux of 6 x 10^n/(cm sec)
have been found to be a factor of 2 to 5 higher than those of an
unirradiated specimen.6 Experiments on 316 stainless irradiated
in EBR-II to fast fluences approaching 7 x lQ22n/cm2 have shown creep
rates increasing about tenfold while linear creep strain decreases
fourfold.'^ Again extrapolations to fluences of 3 x l()23n/cnr are
needed.
More recently, Foster, et al. have reported measurements on
solution annealed and cold-worked types 304 and 316 stainless steel.
The irradiation creep rates are linear with stress and essentially
independent of temperature at low fluence levels:. At high fluence
levels there are limited data which indicate that irradiation creep
increases as swelling becomes significant. The relationship between
swelling and irradiation creep is of great technological significance
to the design of fuel rods and assemblies. In particular, the
beneficial role of irradiation creep in relieving the stresses created
by differential swelling is becoming more clearly understood, and
is being applied to core design. 67
Because of the sensitivity of predicted total plastic strain to
various levels of saturation of cladding swelling (10-15%) and fuel
swelling (20-35%), additional prototypic test data at high fluence
and burnup is necessary to determine the combined effects of fuel
swelling, fission gas pressures, cladding swelling, irradiation-induced
cladding creep, and perhaps other, as yet undiscovered phenomena.
Indeed, the ultimate limiting phenomena for fuel performance may be the
deterioration of the cladding due to fuel-cladding interactions.'2
94
-------
The uncertainties indicated above make it difficult to predict
failure rates and activity releases for LMFBR fuel pins operating at
or near design conditions. Indeed, the ability of the reactor rad-
waste system to handle released activities from failed fuel will
ultimately determine the quantity and quality of failed pins that
can be accommodated in the core at one time.
The statistics of LMFBR oxide fuel pin irradiations have recently
taken a sizable leap. As shown in Table 6.1, the number of pins ir-
radiated or undergoing irradiation in the world's fast reactors was
-v 110,000 as of January 1974. This total is up from ^ 23,000 as a
result of the startup of the Russian BN-350 and the French Phenix
and replacement of fuel in existing fast reactors. Because the
first three prototype LMFBR power stations plan to replace fuel at
rather frequent intervals*, continued rapid growth in oxide pin
operational statistics is expected.
These three early power stations will provide the first fast
neutron irradiation experience on large multipin fuel assemblies
at LMFBR prototypic temperatures and flow rates, and much more
realistic fluence-to-burnup ratios, than have heretofore been
possible.
6.3 Tritium Release from Fuel Pins
Tritium produced in the fuel from ternary fission or from
lithium impurities in the fuel can diffuse through the stainless
steel cladding of intact pins. Results of fast flux irradiation
of stainless-steel-clad mixed-oxide fuel 13 indicate that more
than 99% of the tritium produced in the fuel ,of .large operating
LMFBR's will be released to the primary coolan|. Tritium produced
in the fuel could introduce about 24,000 Ci/yr into the pYimary
coolant system, as discussed in Section 5.1 of this report.
6.4 Release Fractions for Noble Gases from Oxide Fuels
Release fractions of 7 to 25% were observed in the initial fast-
flux irradiation of PuO? - U02 fuel in EBR-II even though it was a
very short irradiation.'4
At about the same time information on the Russian BR-5 plutonium
oxide fuel became available,T5 indicating release fractions of 40-54%
for burnups of less than 3% and about 60% for burnups greater than
3.5%.
*BN-350 reportedly will have one-fifth of its fuel replaced every 54
days; Phenix will have one-sixth replaced at 60-day intervals, and
the PFR (Prototype Fast Reactor) in Great Britain will have one-sixth
replaced at 49-day intervals.
95
-------
Table 6.1
Oxide Fuel Pins Irradiated in LMFBR's
USSR
USA
SEFOR
EBR-II
UK
DRF
DEBENELUX
Rapsodie
DRF
Other
TOTAL
Completed On-Going
BR-5
BR-10
BOR-60
BN-350
France
DFR
Rapsodi e-Core I
Rapsodi e-Fortissi mo
Phenix
* 2,490
* 8,000
* 8,000
41
4,305
^13,600
M.520
3,400
38,200
^3,700
23,002
648
^800
^800
73
48
^38,800
^1,000
^200
60
Total
61,600
44,650
-^2,450
^•1,000
181
^150
•^71,100 ^110,000
96
-------
The first data from Rapsodie, which uses mixed-oxide sintered
fuel pellets in 316 SS cladding, showed gas releases of up to
60% for pins operated at peak powers of 12.2 kW/ft to burnups of
20,000 MWd/MT.
One of the largest sets of experimental data is that from the
General Electric F2 Series. The series consisted of 21 encapsulated
fuel pins in an experimental program designed to "investigate such
parameters as fuel density, compaction process, stoichiometry,
diametral and axial gap, cladding material, and cladding wall
thickness."'7 Results of the irradiations have been published as
data became available. 5,17-20 yne resu]ts are summarized in
Table 6.2. The fission gas releases are seen to approach 100% at high
burnups and are not strongly a function of initial fuel density in this
range of burnup.
Table 6.3 shows irradiation results of twelve other mixed-oxide
fuel elements irradiated in EBR-II under Argonne National Laboratory's
fuel element performance program. The first group of four ANL
elements were some of the first unencapsulated elements to achieve
large burnups in EBR-II. The NUMEC^' elements were designed to give
a comparison of fabrication processes. The last group of four
ANL elements were designed to assess various void deployment techniques
in accommodating fuel swelling. The information in Table 6.3 is
mostly from Reference 22. The data indicate that fission gas release
approaches 90% at 10 at% burnup (* 100,000 MWd/MT.)
23
Other irradiations such as the PNL tests could be added to
these examples, but the trend is clear. On the basis of data presented
here plus other irradiations in EBR-II or DFR, Lambert, et al.,
have concluded that, for fuel operating in the linear power range
expected for LMFBR's (9 to 18 kW/ft), fission gas release increases
with burnup in the manner shown in Table 6.4, regardless of initial
smear density or form of fuel.22
The experimental information presented thus far in this section
provides a good basis for estimating fission gas release from fuel in
LMFBR pins. Moreover, the proliferation^ Of theoretical models to
explain the data will hopefully aid in extrapolation of data to
other conditions through an understanding of the phenomena involved.
In order to estimate activity release from failed pins to the
coolant from the above information, the burnup of the failed pins
must be known. It is a reasonable assumption that failure rates will
also increase with burnup. This has certainly been borne out by
experience.25,26 if m0st of the pins do achieve the goal burnups,
then the highest rate of failures would occur at burnups approaching
100,000 MWd/MT. This argument leads to the assumption that the
"average" failed pin contains fuel which has released at least 75%
of its total inventory of noble fission gases and will continue to
release noble gases at about the same rate as they are produced.
97
-------
Table 6.2
General Electric F2 Series of Fuel ?in Irradiations
Pin
F2A
F2B
F2E
F2F
F2N
F2P
F2Q
F2S
F20
F2W
F2Y
F2Z
F2C
F2H
F2R
F2T
F20
F2D
F2G
F2V
F2X
Fuel
Density
Sneared
(% Theoret-
ical)
94.0
94.6
94.4
94.8
92.6
94.5
94.9
96.1
87.0
83.8
83.9
87.2
93.1
94.9
95.2
94.6
94.0
94.3
93.0
86.7
83.9
Fuel
Density
(% Theoret-
ical)
95.6
95.2
95.3
96.1
96.2
95.6
96.2
97.1
89.5
NA
NA
89.2
94.6
95.3
95.3
96.1
95.8
96.4
95.6
89.7
m.
Max
Linear
Bcwer
(kW/ft)
15.6
16.0
16.3
16.3
'17.7
9.7
16.3
17.4
15.8
13.6
13.3
15.7
16.1
16.7
16.1
16.9
16.1
16.5
16.1
15.7
15.3
Peak
Burnup
(at. %)
5.19
5.26
5.35
5.33
5.42
3.26
5.25
5.49
5.61
5.08
4.91
5.50
7.20
7.30
7.17
7.07
7.39
12.8
12.7
13.1
11.9
%
Fission
Gas
Release
49
48
MA
50
50
32
47
55
63
58
56
58
64
67.5
60.8
59.2
NA
NA
•vlOO
NA
i-lOO
98
-------
Table 6.3
ANL Irradiations in EBR-II
Pin
ANL-012
ANLr-007
ANL-021
ANL-026
NUMEC B-2
NUMEC C-l
NUMEC C-ll
NUMEC C-15
ANL SOPC-1
ANL SOPC-3
ANL SOPC-5
ANL SOVG-17
Fuel
Density
Smeared
(% Theoret-
ical)
78.5
78.6
80.2
82.5
88.8
88.4
88.3
82.7
80.0
82.7
77.1
80.0
Fuel
Density
(% Theoret-
ical)
81.1
79.0
82.8
85.2
90.6
91.1
89.8
84.4
82.2
85.3
86.0
NA
Ma>.
Lirioar
Power
(kW/ft)
16.0
15.7
16.4
17.0
15.0
14,7
14.0
12.9
14.5
14.8
12.1
13.7
Peak
Burnup
(at. %)
2.91
4.73
4.7
4.7
9.8
10,9
10.9
10.6
3.6
3.5
3.2
3.5
%
Fission
Gas
Release
69.1
82.1
85.6
m
89.8
95.7
90.8
68.0
69.0
62.6
56.4
59.5
99
-------
Table 6.4
Percent of Fission Gas Released vs Fuel Burnup
Fuel Burnup Fission Gas
(MWd/MT) ' Released (%)
<2000 ^30
30,000 ^50
50,000 VO
100,000 >85
100
-------
6.5 Fuel Failure Rates
As discussed previously, failures can be of two types: small
leaks and gross cladding failures. No data is available which gives
failure rates of either type for fuel pins tested under prototypic
1,000 MWe LMFBR conditions. For this reason, failure rates and failure
types will be assumed for example calculations in this report,
using available experience as a guideline, but with emphasis on the
fact that realistic rates must await future experience. Certainly,
the failure rates which will be tolerated will depend on the coolant
and cover gas cleanup systems.
Some overall feel for fuel reliability can be obtained from
numbers presented by Bernath and Wolfe2' in April, 1971. In
summarizing world-wide fast reactor experience at that time, they pointed
out that less than 2% of all rods tested had failed. Ignoring
the BR-5 failures, " which resulted when pins with very small fission
gas plena were pushed well past their design limit, less than 0.5%
have failed. This picture gets even better if DRF failures"* caused
by cover-gas entrainment in the coolant are ignored. However, it
is important to note that different fuel tests have different goals
(conditions at specific burnup, power, etc.) and are not all aimed
at achieving high burnups. For example, the first twelve pins of
the General Electric F Series tests were removed after burnups
between about 35,000 and 60,000 MWd/MT.19
The only extensive oxide driver fuel experiences have been
obtained from BR-5 and Rapsodie, together with limited experience at
SEFOR. The BR-5 experience,25 as indicated above, was intentionally
oriented toward run-to-failure with the goal of learning the effects
of operating with leaking fuel elements. Also BR-5 used straight
in its first core, not a mixed oxide.
Experience with the 24 MW core at Rapsodie has been encouraging.
Through February 22, 1970, when the reactor was shut down for
modification, no operation-limiting failures had been observed in
Rapsodie test pins in nearly three years of operation. 29 However,
the reactor core was known before shutdown to contain at least one
failed fuel pin with direct contact between the sodium coolant and
the fuel. '^' Twenty of the driver fuel subassemblies had obtained
a burnuo of 50,000 MWd/MT compared to a design value of 30,000
MWd/MT. 32 However, the gross cladding failure did not occur in any
of these assemblies. It was found in a pin which had achieved a
burnup of 40,000 MWd/MT and a fast fluence of 4 x 1022n/cm2 at a linear
power of only 6.3 kW/ft.^3 It was concluded from instrument records
that the rupture had occured in January, 1969, when the pin had
experienced a burnup of only 24,000 MWd/MT. Also there were indications
that the pin had started leaking at 16,000 MWd/MT.
It is important to note that this -is only one pin out of 2300, and
no gross cladding failures were observed in the higher burnup pins
which had operated at twice the power ( ^12 kW/ft). However, several
leakers have been observed during Rapsodie operation, including at
101
-------
34
least three during the first year of operation. Conservative overall
failure rates of 0.5% with 10% of these gross cladding failures
could be inferred from the very limited data.
Reports of fuel failure experience with the Rapsodie "Fortissimo"
(40 MW) core are not available. This experience should be followed
closely since the failure rate may be significantly higher for the
new core.35
Experience at SEFOR was even better than at Rapsodie, but this
experience is not as significant with regard to fuel operation. No
failures of any type were detected in the first two years of
operation.36 Perhaps this was to be expected because of the much
thicker cladding in SEFOR, combined with much smaller power densities
and fast fluxes as compared to Rapsodie.3' Indeed SEFOR was designed
to check physics characteristics, especially the Doppler effect,
rather than to serve as a fuel irradiation facility.
In conclusion, the available experience with fuel failure
rates can only give very general guidelines. For purposes of
creating a source term for the primary coolant in this report, a
failure rate of 1% will be assumed, 10% of which are gross cladding
failures (i.e. overall 0.1% gross cladding failure).
6.6 Leakage of Fission Products from Failed Fuels - Gaseous and Solid
An important factor in determining noble gas activity in the
primary coolant due to failed fuel is the rate at which gases can
escape through the pin defect. The degree of enhancement of
concentrations of long-lived vs. short-lived isotopes (or the corres-
ponding daughters of such isotopes) in cover gas samples is an
indication of release rates to the coolant and transport times in
the coolant to the cover gas.
6.6.1 Escape Rates from Plenum to Sodium
Mixed-oxide pins irradiated to burnups of MO,000 MWd/MT in
EBR-II were punctured at different axial locations to measure escape
rates.3° The results indicate that the fuel and insulator columns
provide significant resistance to gas releases. The times required
to release 50% of the noble gases varied from less than a minute
to about 20 minutes, depending on the puncture location. Punctures
at the top of the fuel, at the top insulator, and in the plenum all
released over 80% of the gas within about two minutes. One bottom
insulator puncture released 80% of the gas in four minutes. From
this data, an average release time of about four minutes seems
reasonable. Two limiting factors about the experiment were that the
test was run out of pile and that the sizes of the punctures were
not given.
39
Carelli and Coffield calculated internal pin pressures as a .
function of time after failure for punctures of areas 10~° in2 and 10
102
-------
9
in . The larger hole gave almost complete gas release for various
puncture locations in a matter of seconds. The smaller hole held up
the gas release to about three minutes, consistent with the
measurements of Reference 38.
Other experimental measurements of release times have given quite
different results. Studies by Gregoire, Novak, and Murata^O on two
mixed oxide pins in naturally convecting NaK indicate much larger
delay times. The pins failed while under irradiations in the General
Electric Test Reactor (GETR) at a burnup of about 18,000 MWd/MT. The
one pin on which plenum depressurization results were given required
about four hours for total gas escape, with the peak escape rates
within the first 45 minutes after failure. The pin was continually
moved to lower power regions during the time immediately after
failure, so the results are not exactly applicable to full power
operation either. Moreover, although many cracks were observed in
the failed region (14 inches below the plenum), no areas were given.
A more recent GE test showed even longer release times, but
is not really applicable to sitations of interest here because the
pins had only undergone 3900 MWd/MT irradiation at the time of
defection. This test is more applicable to determination of in-fuel
diffusion rates.
At this point one is left with a wide range of five minutes to
one hour for possible delay times for release of plenum gas from the
fuel pin.
6.6.2 Transit Time from Failure to Cover Gas
Delay times for transit to the cover gas from the defected pin
may show less uncertainty. These times will, of course, depend on
reactor and vessel dimensions, coolant flow rates, etc.
A number of measurements of the transit time between fuel
and cover gas have been made. Measurements in SEFOR36 on disengagement
times for ^He gas of 5.5 min. from core to cover gas were assumed to
be the same as for fission gas bubbles.36 Studies in Rapsodie42
(the Tempete tests) were made on disengagement times for noble gases
released from the lower region of Rapsodie pins. The results
indicated that "once the clad barrier is passed, the transfer of
xenon and krypton from clad failures from the pin to the reactor
cover gas will be fast and practically complete". The Tempete test
results indicated disengagement times of the order of a few minutes,
thus agreeing with SEFOR results. This also agrees with previous
out-of-pile results from Atena^3 and with observed cladding failures
at Rapsodie. An averave assumed disengagement time of 5 to 10 minutes
could be inferred from the SEFOR and Rapsodie data.
103
-------
-------
p \
in . The\larger hole gave almost complete gas release for various
puncture locations in a matter of seconds. The smaller hole held up
the gas release to about three minutes, consistent with the /
measurements of Reference 38.
Other experimental measurements of release times have given quite
different results. Studies by Gregoire, Novak, and Murata^u on two
mixed oxide pins in naturally convecting NaK indicate much larger
delay times. The pins failed while under irradiations in the General
Electric Test Reactor (GETR) at a burnup of about 18,000 MWd/MT. The
one pin on which plenum depressurization results were given required
about four hours for total gas escape, with the peak escape rates
within the first 45 minutes after failure. The pin was continually
moved to lower power regions during the time immediately after
failure, so the result^ are not exactly applicable to full power
operation either. Moreover, although many cracks were observed in
the failed region (14 inches below the plenum), no areas were given.
41
A more recent GE test showed even longer release times, but
is not really applicable to sitations of interest here because the
pins had only undergone 3900-.MWd/.MT irradiation at the time of
defection. This test is more applicable to determination of in-fuel
diffusion rates. •/
At this point one is left w|th a wide range of five minutes to
one hour for possible delay times for release of plenum gas from the
fuel pin.
6.6.2 Transit Time from Failure to Cover Gas
Delay times for transit to the cover gas from the defected pin
may show less uncertainty. These times will, of course, depend on
reactor and vessel dimensions, coolant flow rates, etc.
A number of measurements of the transit time between fuel
and cover gas have been made. Measurements in SEFOR36 on disengagement
times for "Ne gas Of 5.5 m-jnm from core to dpyer gas were assumed to
be the same as for fission gas bubbles.3° studies in Rapsodie^
(the Tempete tests) were made on disengagement times for noble gases
released from the lower region of Rapsodie pins. \ The results
indicated that "once the clad barrier is passed, the transfer of
xenon and/krypton from clad failures from the pin to the reactor
cover ga,£ will be fast and practically complete". The Tempete test
results/indicated disengagement times of the order of a few minutes,
thus agreeing with SEfOR results. This also agrees with previous
out-pf-pile results f t om Atena^3 and with observed cladding failures
at-Rapsodie. An average assumed disengagement time of 5 to 10 minutes
could be inferred from the SEFOR and Rapsodie data.
, \ / • ' . : . 103
\ >- i'
-------
6.6.3 Time for Diffusion out of the Fuel
From the previous discussion, a delay time of five minutes to
one hour seems reasonable for gas which has already been released
from the fuel proper but is still inside a leaking pin. Another delay
of the order of five to ten minutes might be seen in transporting
noble fission gases to the cover gas. However, these total delay
times of 10 to 70 minutes may well be insignificant compared to the
time required for fission products, whether rioble gas, semi-volatile,
or solid, to diffuse out of the main body of the fuel proper.
IT
A recent GE test cited previously indicates diffusion times
for noble gases of the order of many hours. Perhaps the best data
for operating gins is that obtained from Gulf General Atomic experi-
ments at ORNL- From this data a diffusion time of the order of
twelve hours is indicated for noble gases. Often the assumption is
made that other elements diffuse much more slowly in reactor fuel;
however, the results of Davies, Long and Stanaway**5 indicate migration
times for iodine, tellurium, and cesium comparable to the noble gases.
6.6.4 Diffusion Direction
The question of diffusion times leads logically to the question
of diffusion direction. Elements which tend to diffuse toward the
center of the pin, i.e. up the temperature gradient, may contribute
significantly less to coolant contamination from gross cladding
failures than those elements which diffuse down the temperature
gradient. Note, however, that some elements may have high concentra-
tions at both the central void and at the outer surface, with minimal
in between. A fair amount of migration behavior has been determined.
In mixed oxide fuels operating at sufficient linear powers
(> 11 kW/ft), plutonium will preferentially migrate to the central
void,46 apparently by preferential vapor transport mechanisms. This
phenomenon is desirable in the sense that it reduces the amount of
plutonium that can be leaked into the coolant. The plutonium migration
is undesirable because of its effect on the Doppler Coefficient
(small effect) and on the maximum allowable power rating (bigger effect).
48
Duncan, et al., have observed fission product migration in
the F Series of fuel pin tests. Volatile elements such as I and Cs
migrate to the cladding. Some elements, such as the noble metals Mo,
Ru, Rh, Tc and Pd, form metallic ingots in the center void of high
burnup fuels. Other elements, such as Ba, Zr, and Sr, in the form
of oxides, produce nonmetallic deposits on the walls of the central void.
These nonmetallic deposits all contain some Pu and U oxides.
49
Lambert, et. al., also observed outward radial migration of
Cs along with axial migration. They also found the noble metals
mentioned above in the form of metallic ingots in the central void.
In addition they detected Zr. Nb, and Ce by gamma spectroscopy which
indicated the presence of 103Ru, 106Rh, 95Zr, 95Nb, and 141Ce.
104
-------
50
Johnson, Steidl, and Crouthamel observed similar behavior for
as was given in References 48 and 49. Reference 50 also
indicated a 13/cs as was given in References 48 and 49. Reference 50
also indicated a !37cs increase near the central void, with a minimum
13'Cs concentration at an intermediate radius. Essentially uniform
concentration of Ce and the other rare earths was observed, while Mo
increased markedly in concentration in the outward radial direction.
The outward radial migration of cesium is unfortunate in two
respects. First of all, cesium oxide is apparently important in
cladding corrosion51 (along with Mo02). Moreover, when the cladding
fails the Cs then has access to the sodium coolant. The same
undesirable access to sodium is achieved by I in its migration outward.
6.6.5 Experimental Data on Transport of Specific Fission Products
to the Sodium or NaK Coolant
The migration behavior described above is generally borne out
by experience. General Electric has run a series of tests on
irradiated defected fuel to determine leakage of fission products from
the fuel .40,41 ,52,53,54 The tests varied greatly in important aspects
such as burnup of fuels tested (3900 to 41,000 MWd/MT at failure),
environment (in-and-out-of-pile) , coolant, coolant velocity, and size
of defects (0.0007 in2 to 0.050 in2).
Fuel pin B9A was irradiated to only 8700 MWd/MT before being
intentionally defected by three 31 mil holes (each of area 0.00075^
lined up within a one inch notch but in the cladding. The only
really non-prototypic condition was the low sodium velocity. Almost
03
no 103RU or 9$lr was released to the coolant (as opposed to previous
tests). However, almost all of the 137cs left the fuel. Less than
1% of the fuel was lost from the pin. In general leakage of both
fuel and non-gaseous fission products from pinhole defects is extremely
slow relative to leakage from the large defects in experiments reviewed
below.
Rods B9D-1 and B9D-2 ' were intentionally defected under
irradiation at a burm
-------
Table 6.5
Leaching Results from Grossly Defected Oxide Pin (B3C) in NaK
,40
Isotope
Concentrations
in NaK
(atoms/gm NaK)
NaK
**
ugn/in2
Plateout
Concentrations
(atoms/in2)
i*e.
137CS
'""Ce
106Ru
103Ru
iMgb
95zr
91Y
23^
u
2.4xl015
2.1xl017
7.7xl010
4.2xlOn
9.0xl013
1.3xlOn
3.2xl012
<8.9xl0ltf
* .54xlO~3
*1.3
1.23xl013
1.42X1015
.96xl015
2.9xl011+
G.lxlO15
1.5xl0llt
3.6x10^
<9.7xl013
** 12
106
-------
An intentionally defected section of fuel pin F2U which had been
irradiated to 41,000 MWd/MT was exposed to 1150°F flowing sodium for
18 hours.53 The defect had an area of about .050 in2, versus ^.002
in2 for B9A. There was substantial loss of selected measured
fission products (shown in Table 6.6) but less than 1% loss of fuel,
with uranium preferentially escaping compared to plutonium (presumably
because of Pu migration toward the center of the pin). Leached
fuel deposits showed U/Pu ratios of from 1643/1 to 42/1, compared to
an original ratio of 4/1.
25
The BR-5 experience is generally consistent with the above
test results. The first indication of gross cladding failure was
detection of 137cs in coolant samples. The most important results
for determining defective fuel source terms is that the escape
fractions of 13/cs, 136cs, and 133xe, were an order of magnitude
higher than the escape fractions of '31i, 95zrj 95|\|b, 140ea, 140La,
and 135xe. Also, the '311 and '36(;s activities were appreciable
only with very large leaks. In addition to the major radionuclides
discussed above, small amounts of ^^Ce, l^Pr, and IQ^Ru were found
in the BR-5 primary coolant.
6.6.6 Theoretical Models for Fission Product Transport
Theoretical models to describe the migration of the various
chemical species and thus their availability for transfer through
failed cladding to the coolant, have been available since the late
1950's. A simple diffusion model concerned strictly with fission
product concentrations within isothermal particles or grains was
developed by Booth.55 Neglect of the temperature gradient present in
operating fuel renders the Booth model inadequate, however. Others56*57
have used Booth's release equations and have included the effect
of fuel surface temperature gradients. TheseS6,57 models have been
widely used. Yuill, et. al.,5° have derived the relationship between
release fraction and temperature gradient directly from equations
for diffusion in the fuel controlled by the Gibbs free-energy gradient.
Models which incorporate both concentration and temperature effects
have been moderately successful, but release estimates used for
the example calculation in the present report will be based on
experimental work.
6.7 Vented Fuel
Vented fuel is of interest for two reasons. First it may be
desirable to use vented fuel, either for safety reasons or simply
to maintain cladding integrity to the high burnup goals of LMFBR's.
In addition, the experimental work from vented fuels gives added
information on release rates from the fuel and from the pin which
can be used in analysis of defective non-vented fuel.
A good discussion of vented fuel elements is given by Keilholtz
and Battle.59 Little new work on vented fuel has been reported
since their paper. One possible venting design not discussed in
107
-------
Table 6.6
Loss of Fission Products from Grossly Defected Oxide
53
Fuel in Flowing Sodium
Nuclide % Loss From
Fuel
137Cs 66
144Ce 32
106Ru 85
90Sr 29
147Pm 32
95Nb-Zr <1
125Sb «1
U.Pu <1
108
-------
Reference 59 was a modified "diving bell" concept with separate
inlet and outlet capillaries that is mucy shorter than the original
GE "diving bell" design.
rn
The main arguments for vented fuel elements have been ones of
long term performance and safety: fewer cladding failures because
of reduced fission gas pressure buildup, better transient behavior
with respect to coolant voiding because of reduced amounts of gases
available, and reduction of failure propagation that might be caused
by blanketing of neighboring pins by gas expelled from failed
non-vented pins.
If the dominant mechanism for fuel failure turns out to be a
combination of fuel and cladding swelling, mechanical property change
of the cladding under irradiation, and/or fuel-cladding interactions,
the first argument for vented fuel gi\en in the previous paragraph
may not carry much weight. Thus, the main arguments may be ones
of safety, e.g. no sudden gas bursts to add reactivity, or propagate
failures, or possibly transport fuel.
One disadvantage of using vented fuel (other than the obvious
complications of increased shielding plus larger cover gas and
sodium cleanup system) is the possible transport of non-gaseous
fission products from the fuel to the coolant. This could happen
by diffusion of volatile elements (such as Cs) or by release of
gaseous precursors which subsequently decay to solids. (See the
notes on DFR experience at the end of this section).
One additional disadvantage may be a long term buildup of
certain radionuclides in the primary sodium, notably 134^$, produced
by neutron activation of 133c$, a daughter of ^3/e.
Whatever the final judgement is on use of vented fuel, the
testing of vented fuel has produced valuable data on release fractions
of various radionuclides.
The work of O'Neill, et. al. on mixed oxide fuel exposed to
16,500 MWd/MT in a thermal flux gave the following major results:
1. About 44% of noble fission gases were released from the
the fuel proper, i.e. were available for venting.
2. Effective gas delay time was 5 days.
24 134
3. Dominant sodium activity after decay of Na was Cs,
although this would be less of a problem for fast fluxes.
4. Release fractions of all isotopes (including fissile) except for
the noble gases were extremely small - in the range of
10-10 to 10-6.
109
-------
Table 6.7 gives measured release fractions. It is a reproduction
of Table 5.1 from Reference 61. Note that all of the long-lived
85 ancj 41/\ wn-jch vary by as much as a factor of 100
depending on the sampling location. Typical values for the Xe
isotopes are on the order of 10^ dpm/cc. A table of possible
radionuclides accounting for measured activities in the NaK is
given in Reference 64. Amounts of U and Pu in the primary circuit
are small (less than 0.5 gm and 20 mg. respectively.) Coolant ^4
activity during normal operation is about 100 times that due to Na,
with '3°Cs the major nuclide of interest.
The data presented above for DFR are so different from expected
vented oxide fuel behavior (due mainly to the effectively direct
exposure of so much of the DFR fuel surface to the primary coolant)
that it can contribute little to estimating activities from either
vented or failed mixed-oxide fuels.
6.8 Example Calculations of Releases of a Few Selected Radionuclides
imp
"tb
to tne Primary Coolant ana the Cover Gas System
110
-------
Table 6.7
Isotopic Release Fractions from
GE Vented Fuel Test
Radionuclide
Noble
Gases :
85Kr
131mxe
133Xe
133mXe
135Xe
Half -Life
10.76 yr
11.96 d
5.27 d
2.26 d
9.16 h
Release
Fraction
0.44
0.30
0.27
0.05
0.0003
Solids:
103
131
Ru
137Cs
Other (Sr, Y, Zr,
Ba, Ce)
4x10
1x10
-10
-8
IxlO'6
10"8 to 10"6
111
-------
From discussions and data already presented in this section, it is
clear that predictions of releases from the fuel are dependent on
a large number of complicated factors, the biggest one of which is:
how much activity can the cleanup systems and containment handle
and still satisfy federal guidelines and regulations. This factor
will obviously be the governing one. However, example calculations
will be made here using several assumptions (some based on available
information and some arbitrary) of possible releases from failed
fuel in an operating 1000 MWe LMFBR.
Assumptions will include the following:
1% failed fuel
90% of the failed pins are leakers
10% of the failed pins have gross cladding failures
75% of the fission gas is released from the fuel proper, i.e.
pellets, of the failed pins.
For the gross failures, the following percentages of long-lived
elements are assumed to escape:
Fuel 1%
Br, I, Cs 15%
Te, Ru, Tc, Mo 5%
All Others 1%
The choice of escape fractions was based on relative volatility of
the elements (or their oxides) and release data from References 40
and 53. Escape fractions based on only the volatilities and
Reference 40 data would have suggested lower values for the group of
Te, Ru, Tc, and Mo and for the group of "all others." On the other
hand, Reference 53 suggests much higher release percentages.
Neither test was prototypic so a compromise was made. Limited BR-5
data available generally support the choices for long-lived Cs
and "all others." Plutonium fractions in the released fuel will be
assumed equal to 0.1 of the original fraction based on observed
preferential leaching of the uranium, due partially to plutonium
migration.
At Westinghouse's Advanced Reactors Division, the following
escape rates are assumed:°' , ^ v ,n5.
•» ~ I • D X I U A,
Noble Gases f a!.5xl05X
112
-------
1 _ e-1.5 x 106X
Halogens f = 0.2 -j 5 x iQ6x
Alkali Metals (Cs) f = 1
All Others f = 0.01
where f is the fraction of fission products produced in a defected
fuel pin which escapes to the coolant.
Table 6.8 gives calculated equilibrium cover gas activities
for several noble gases assuming various total delay times between
production and entry into the cover ges. The table is simply a
modification of Table 6.2 from Refererce 61 to account for 1%
failures and 75% gas releases plus slightly different nuclear data
available today. The table reflects the same cover gas purge rates
as in Reference 51 with the corresponding limitation on long-lived
radionuclide activity. Note that the difference in total cover gas
activity between no delay and a 15 hour delay is only about a factor
of 4. Thus, the results are not too critically dependent on a good
knowledge of migration or diffusion rates within the fuel pellets.
Table 6.9 gives calculated annual contamination of the primary
system from important long-lived nuclides. The numbers are based
on assumptions stated above plus calculated activities from the GE
1000 MWe LMFBR as given in Table 4.6. The activities shown in
Table 6.9 are important because much of the^se will collect in the
cold traps and be shipped from the reactor or else will deposit
on the colder surfaces of the primary system.
85
Note also that the Kr which is removed by,the cover gas
cleanup system may also be shipped away eventually, just as all the
activity which remains inside the fuel pin will be. The amount of
85Kr removed by the gaseous radwaste system under the conditions
described here would be about 1900 Ci/yr.
\
The last activity source to be discussed here is fuel leached
from elements experiencing gross cladding failures. An assumed
fuel loss of less than 1% from such pins is consistent with almost
all of the references cited above.40»4',52,53,54 Combined with the ''
assumption of 0.1% gross cladding failures and the assumption that the
Pu fraction in leached fuel is only 0.1 of the original fraction,
burdens of fuel contamination of the primary circuit can be estimated.
The annual leaching rate thus calculated is 130 gm per year of
metal fuel atoms, of which about 2.5 gm would be plutonium. This
would represent about 20 Ci of plutonium activity, most of which would
be beta activity of 241Pu.
It should be emphasized here that the failure rates and release
fractions assumed in calculating all types of release from LMFBR
fuel were, to a certain extent, arbitrary and may be proven high by
113
-------
Table 6.8
Example Equilibrium LMFBR Cover-Gas Activity from Failed Fuel for
Various Delay Times After Birth for Gaseous Radionuclides
\ (1% fuel failure, 75% release) /
\ Curies /
Nuclide
89Kr
137Xe
138Xe
135mKr
87Kr
83mKr
00
Kr
85mKr
135Xe
1 O wMH v A
xe
133Xe
131mxe
85Kr
"Half-Ltfe"
V
3.18m \
3.82m
14.2m
15.7m
76m
1 .86h
2.79h
4.4h
9.16H
2.26d
5.27d
li;96d
/ 10.76y
0
2. 69x1 O5
\ 8.54xl05
/5.87xl05
2:§9xl05
1.69X105
5. 26x1 O4
c
2.07x10
9. 76x1 O4
1.06xl06
2. 79x1 O4
1.14xl06
3. 30x1 O3
20.2
10m
3. 08x1 O4
1.36X105
3. 60x1 O5
1.66X105
1.55xl05
4.91xl04
c
1.99x10°
\ 9. 50x1 O4
Xosxio6
2. 79x1 O4
1.14X1I)6
3.30xlO\
20.2
1 h
.m/
',*£&
3.14xl04
1.84xl04
9. 90x1 O4
3. 64x1 O4
c
1.62x10
8. 32x1 O4
9.80xl05
2. 74x1 O4
1.14xl06
3. 30x1 O3
\20.2
5h
.256
.471
8. 44x1 O3
789
A
6.19x10*
4,41xl04
7. 24x1 O5
2.61xl04
l.lOxlO6
3. 30x1 O3
20.2
15h
59.2
206
•3
5.17x10
8. 99x1 O3
3.14xl05
2.31xl04
1.05xl06
3.17xl03
20.2
Totaf
4.73xl0
3.39xl0
1.97xl06 1.40xl06
\
114
-------
Table 6.8
Example Equilibrium LMFBR Cover-Gas Activity from Failed Fuel for
Various Delay Times After Birth for Gaseous Radionuclides
(1% fuel failure, 75% release)
Nuclide "Half-Life"
Curies
10m
Ih
5h
15h
Kr
137X
138Xe
135mKr
87Kr
83mKr
88Kr
85mKr
135Xe
133mx
133Xe
131mxe
85Kr
3.18m
3.82m
14.2m
15.7m
76m
1.86h
2.79h
4.4h
9.16h
2.26d
5.27d
11.96d
10.76y
2.69xlOJ
8. 54x1 O5
.5.87x!05
2.59xl05
1.69xl05
5. 26x1 O4
2. 07x1 O5
9.76xl04
1.06xl06
2.79X104
l.HxlO6
3. 30x1 O3
20.2
3.08x10^
1.36xl05
3.60xl05
1.66xl05
1.55xl05
4.9'lxlO4
1.99xl05
9. 50x1 O4
l.OSxlQ6
2. 79x1 O4
1.14xl06
3.30x10^
20.2
.269
12.8
3.14xl04
1.84xl04
9. 90x1 O4
3. 64x1 O4
1.62xl05
8. 32x1 O4
9.80xl05
2. 74x1 O4
1.14xl06
3. 30x1 O3
20.2
.256
.471
8.44xl03
789
6.19xl04
4.41xl04
7.24xl05
2.61xl04
l.lOxlO6
3. 30x1 O3
20.2
59.2
206
5.17xl03
8. 99x1 O3
3.14X105
2.31xl04
1.05xl06
3.17xl03
20.2
Total
4.73xlOc
3.39x1O6 2.58x1O6 1.97xl06 1.40xl06
114
-------
Page Intentionally Blank
-------
Nuclide
85Kr
90Sr-Y
106Ru-Rh
125Sb
125mTe
137Cs-Ba
151Sm
i55Eu
2^Pu (B)
Pu (a)
Pu
Table 6.9
Calculated Annual Activities of Long-Lived
Radionuclides Entering Primary
Sodium from Failed Fuel
Annual Release
(CD
1900
30
4600
150
1100
900
100
20
0.6
2.5 grams
Activity Present One
Year After 30 Year
Operation Period
(Ci)
25,000
600
6000
40
10
350
30,000
800
500
50
4
60
400
20
75 grams
*Activation product of 133Cs (daughter of 133Xe); therefore, activities
depend on irradiation history of failed pins, i.e., time of failure.
115
-------
a significant factor in future years. However, the numbers used
were reasonably consistent with experience and do serve as a conve-
nient basis for calculating releases. If, for example, overall fuel
failure rates are 0.1% instead of 1%, the gas release estimates
may be scaled down by a factor of ten, etc.
Returning to the question of fuel leaching from pins having
gross cladding failures, this leaching may indeed be the ultimate
limiting factor on fuel performance. Little information is available
on the magnitude of fuel contamination which can be tolerated in
the primary circuit. Certainly BR-5 operated with plutonium con-
tamination from failed elements," and DFR has found no problem with
having about 0.5 gm of fuel in the primary coolant;^^ however, only
20 mg of this was plutonium. Perhaps the best estimate of how much
fuel leaching will be tolerated can be found by looking at the
question of tramp fuel (Section 5.4 of this report). Tramp fuel
inventories per unit fuel pin length consistent with those seen in
SEFOR, EBR-II, and Rapsodie would indicate total tramp fuel
inventories of 0.5 gm in a large LMFBR, of which about 0.1 gm would
be Pu. To restrict fuel leaching to this order of magnitude per
year would require a reduction by more than a factor of 200 of the
leached fuel mass calculated above; however, the observed pre-
ferential leaching of U over Pu can result in annual Pu leach
rates calculated above that need only be reduced by a factor of 20
to approach the magnitude of Pu inventories in tramp fuel.
116
-------
REFERENCES (Section 6)
1. "Conceptual Plant Design, System Descriptions, and Costs for a
1000 MWe Sodium-Cooled Fast Reactor-Task II Report," GEAP-5678,
p. 18 (December 1(J68).
2. "1000 MWe LMFBR Follow-on Study," BAW-1328, vol. 5, p. 1-10
(January 1969).
3. E. E. Kintner, "LMFBR Fuel Design - Why Can't We Do Better,"
Proceedings of the Conference on Fast.Reactor Fuel Element Technology
New Orleans, La., April 13-15, 1971, p.l, Am. Nucl.'Soc. (1971).
4. H. Lawton, et. al., "The Irradiation Behavior of Pu-Bearing
Ceramic Fuel," London Symposium on Fast Breeder Reactors,
British Nucl. Energy Soc., London (1966).
5. R. N. Duncan, D. A. Cantley, K. J. Perry, and R. C. Nelson,
"Fuel Swelling - Fast Reactor Mixed Oxide Fuels," Proceedings of
the Conference on Fast Reactor Fuel Element Technology, New
Orleans, La., April 13-15, 1971, p. 291, Am. Nucl. Soc. (1971).
6. C. Cawthorne and E. J. Fulton, "Voids in Irradiated Stainless
Steels," Nature, 216, 576 (1967).
7. E. E. Bloom and J. R. Weir, "In-Reactor and Postirradiation
Creep-Rupture Properties of Type-304 Stainless Steel," Trans.Am.
Nucl. Soc., 10, 131 (1967).
8. R. M. Willard, "Design Criteria for 304 and 316 Austenitic
Stainless-Steel Cladding for FBR Fuel Element," Trans. Am. Nucl.
Soc., TO., 486 (1967).
9. S. Oldberg, D. Sandusky, P. E. Bohaboy, F. A. Comprelli,
"Analysis of Swelling of Austenitic Stainless Steels in Fast
Reactors," Trans. Am. Nucl. Soc., 12., 588 (1969).
10. A. J. Lovell, J. J. Holmes, "Creep Rupture Properties of Type 316
Stainless Steel after High-Temperature Irradiation," Trans. Am.
Nucl. Soc., 14, 153 (1971).
11. A. Boltax, T. P. Soffa, A. Biancheria, "Sensitivity of Fuel Pin
Behavior to Void Swelling and Irradiation Creep of Stainless
Steels," Trans. Am. Nucl. Soc., 14, 631 (1971).
12. Proceedings of the Conference on Fast Reactor Fuel Element Tech-
nology. New Orleans, La., April 13-15. 1971, Session III,
pp. 393-458, Am. Nucl. Soc. (1971).
13. G. P. Wozadlo, B. F. Rubin, P. Roy, "Tritium Analysis of Fast
Flux Irradiated Mixed-Oxide Fuel Pins," Trans. Am. Nucl. Soc., 1_5,
, 200 (1972).
117
-------
14. S. A. Rabin, R. W. Darmitzel, W. W. Kendall, "Short-Term Fast-
Flux (EBR-II) Irradiation of Pu02 U02 Fuel Pins," Trans.
Am. Nucl. Soc., 9, 41 (1966).
15. M. T. Simnad, Fuel Element Experience in Nuclear Power Reactors,
p. 530, Gordon and Breach, New York (1971)."
16. F. Anselin, R. G. Mas, J. P. Mustelier, "Irradiation Behavior
of Plutonium Mixed-Oxide Driver Fuel of Rapsodie," Trans. Am.
Nucl. Soc. 11, 514 (1968).
17. R. C. Nelson, B. F. Rubin, W. W. Kendall, and W. E. Bailey,
"Performance of Mixed-Oxide Fuel Pins Irradiated in a Fast
Reactor to 50,000 MWd/T," Trans. Am. Nucl. Soc., 1_0, 460, (1967).
18. W. E. Bailey, C. N. Spalaris, D. W. Sandusky, and E. L. Zebroski,
"Effects of Temperature and Burnup on Fission Gas Release in
Mixed Oxide Fuel," Ceramic Nuclear Fuels - International Symposium,
May 3-8, 1969, Nuclear Div. of American Ceramic Society (1969).
19. C. N. Craig, R. R. Asamoto, R. N. Duncan, "Fast Reactor
(PuU)02 Fuel Pin Irradiations in EBR-II to 75,000 MWd/Te,"
Trans. Am. Nucl. Soc., 1_2, 566 (1969).
20. C. N. Craig, W. K. Appleby, K. J. Perry, R. N. Duncan, W. E.
Bailey, and C. N-. Spalaris, "Steady-State Performance of Pu02-U02
Fast Reactor Fuel," Proceedings of the Conference on Fast
Reactor Fuel Element Technology. New Orleans, La.. April 13-15,
lojf
1971, p. 555, Am. Nucl. Soc. (1971).
21. "Fabrication of 20 wt% Pu02-U02 Fast Breeder Fuels for Irradiation
Testing in EBR-II," NUMEC-3524-74 (June 1970).
22. J. D. B. Lambert, L. A. Niemark, W. F. Murphy, and C. E. Dickerman,
"Performance of Mixed-Oxide Fuel Elements - ANL Experience,"
Proceedings of the Conference on Fast Reactor Fuel Element
Technology, New Orleans, La., April 13-15, T971, p. 517, Amer.
Nucl. SocT (1971).
23. J. E. Hanson, "Experimental Description and Hazards Evaluation
for the Pacific Northwest Laboratory Mixed Oxide (U02-Pu02)
Irradiations in EBR-II, Task A. Subtask I. Irradiations,"
BNWL-650 (July 1968).
24. Reference 12, Sessions II and III, pp. 137-493.
25. V. V. Orlov, et. al., "Some Problems of Safe Operation of the
BR-5 Plant," Int. Conf. on the Safety of Fast Reactors, Aix en
Provence, September, 1967, paper no. Va-7, Commissiat a 1'Energie
Atomique (1967).
118
-------
26. N. J. Olson, G. C. McClellan, J. E. Flinn, D. G. Franklin, S. C.
Miller, "Analysis of Mark-IA Run-to-Failure Experiments in
EBR-II," Trans. Am. Nucl. Soc.. 15, 748 (1972).
27. L. Bernath and W. B. I'olfe, "Fuel Design for the AI LMFBR
Demonstration Plant," Proc. of the Conf. on Fast Reactor Fuel
Element Technology, Nr*v Orleans, La., April 13-15", 1971, p. 47, Am.
Nucl. Soc. (1971
52
28. J. Graham, Fast Reactor Safety, p. 228, Academic Press, New York
(1971).
29. G. Vendryes, "A General Survey of the French Fast Reactor
Program," Trans. Am. Nucl. Soc., 13, 104 (1970).
30. R. DeFremont, "Observations on the Behavior of Radioactive
Products in Rapsodie," DRNR/STRS.71.1146, 1971.
31. D. P. Roux and J. Max, "Detection and Location of a Fuel Pin
Failure in Rapsodie Using Noise Analysis," Trans. Am. Nucl. Soc.,
1_4, 304 (1971).
32. C. Moranville, "Fuel Development for French Fast Reactors,"
Trans. Am. Nucl. Soc., V3, 104 (1970).
33. P. Bussy, "Observations on a Rupture of a Rapsodie Fuel Pin,"
International Meeting on Fast Fuel and Fuel Elements, Karlsruhe
(1970).
34. G. Gajac, J. L. Ratier, L. Reynes, M. A. Valantin, "Rapsodie's
First Year of Operation," CEA-CONF-1247 (1969).
35. Personal Communication, EBR-II Staff, December, 1972.
36. J. J. Regimbal, W. P. Kunkel, and R. S. Gilbert, "Measurement
of Noble Gas Transport Dynamics in SEFOR Sodium," Trans. Am.
Nucl. Soc., 14, 773 (1971).
37. Reference 15, pp. 587, 596.
38. Reference 21, p. 566.
39. M. D. Carelli and R. D. Coffield, Jr., "Fission Gas Ejection
Characteristics and their Effect on Adjacent Fuel Pins in an
LMFBR," Proc. of the Conf. on Fast Reactor Fuel Element
Technology, New Orleans, La., April 13-15, 1971, p. 617,
Am. Nucl. Soc. (1971).
40. K. E. Gregoire, P. E. Novak, and R. E. Murata, "Failed Fuel
Performance in Naturally Convecting Liquid Metal Coolant,"
GEAP-13620 (June 1970).
119
-------
41. J. J. Regimbal, R. S. Gilbert, and P. E. Bohaboy, "Radionuclide
Release from Intentionally Defected Fuel in the B9D In-Pile
Sodium Loop," Trans. Am. Nucl. Soc., 1_5, 197 (1972).
42. R. DeFremont, "Tempete Test: In-Pile Experiment on the Transfer
of Fission Gases," DRNR/STRS.71.1147, (1971).
43. "Diffusion des produits de fission," Proc. of the Conf. of the
French Society for Radiation Protection, Saday, November 4-6,
1969, paper no. 27.
44. A. W. Longest, N. Baldwin, J. A. Conlin, R. B. Fitts, and
J. R. Lindgren, "Fission Gas Release Measurement from Fast Breeder
(U,Pu)02 Fuel," Trans. Am. Nucl. Soc.. H, 604 (1970).
45. D. Davies, G. Long, and W. P. Stanaway, "The Emission of Volatile
Fission Products from Uranium Dioxide," UK Report AERE-R4342 (1963).
46. R. 0. Meyer, E. M. Butler, and D. R. O'Boyle, "Actinide
Redistribution in Mixed-Oxide Fuels Irradiated in a Fast Flux,"
Trans. Am. Nucl. Soc.. 15, 216 (1972)
47. W. T. Sha, P. R. Huebotter, and R. K. Lo, "The Effect of
Plutonium Migration on Allowable Power Rating and Doppler
Broadening," Trans. Am. Nucl. Soc., 1_4, 183 (1971).
48. Reference 5, pp. 294-296.
49. Reference 22, pp. 542-545.
50. C. E. Johnson, D. V. Steidl, and C. E. Crouthamel, "Distribution
of Gaseous Fission Products in Irradiated Mixed Oxide Fuels,"
Proc. of Conf. on Fast Reactor Fuel Element Technology,
New Orleans, La., April 13-15, 1971, p. 603, Am. Nucl. Soc. (1971).
51. C. E. Johnson, I. Johnson, and C. E. Crouthamel, "Fuel-Cladding
Chemical Interactions in U02-20 wt% Pu02 Fast Reactor Fuel Clad
with Stainless Steel," Proceedings of the Conf. on Fast^Reactor
Fuel Element Technology. New Orleans, La., April 13-15, 1971.
p. 393, Am. Nucl. 5007(1971).
52. R. E. Murata, C. N. Craig, H. C. Pfefferlen, and P. E. Novak,
"Effect of Stoichiometry on the Behavior of Mixed-Oxide Fuel
during Extended Operation in Failed Pins," GEAP-13730
(July 1971).
53. D. E. Plumlee and P. W. Novak, "Measured Loss of Fission Products
from High-Burnup Mixed-Oxide Fuel in a Miniature Pumped Sodium
Loop," GEAP-13731 (August 1971).
54. P. E, Bohaboy, C. N. Craig, and G. L. Stimmell, "Performance of
Failed Plutonium-Uranium Oxide Fuel Elements in Flowing Sodium,"
Trans. Am. Nucl. Soc., 1_5, 196 (1972).
120
-------
55. A. H. Booth and G. T. Rymer, "Determination of the Diffusion
Constant of Fission Xenon in U02 Crystals and Sintered Compacts,"
AECL-692 (August 1958).
56. G. VI. Parker, et. al., "Prompt Release of Fission Products from
Zircaloy Clad U02 Fuels," ORNL-4228 (April, 1968).
57. D. L. Morrison, et. al., "An Evaluation of the Applicability
of Existing Data to the Analytical Description of a Nuclear
Reactor Accident," BMI-1810 (July 1, 1967).
58. F. W. A. Yuill, V. F. Baston, and J. H. McFadden, "An Analytical
Model Describing the Behavior of Fission Products in Operating Fuel
Pins," IN-1467 (June 1971).
59. G. W. Keilholtz and G. C. Battle, Jr., "Fission Product Release
and Transport in Liquid Metal Fast Breeder Reactors," ORNL-NSIC-37,
section 2.4 (March 1969).
60. A. Gerosa and M. Martini, "Venting Device for Sodium-Cooled
Fast Ceramic Reactor Fuel Elements," Trans. Am. Mud. Soc., 11,
508 (1968).
61. G. L. O'Neill, J. Duffy, D. B. Sherer, J. C. Gilbertson, and
F. W. Knight, "A Technical and Economic Evaluation of Vented
Fuel for Sodium-Cooled Fast Ceramic Reactors," GEAP-4770
(May 1965).
62. G. L. O'Neill, J. H. Davies and J. L. Johnson, "Demonstration
of Fission-Gas Venting from Fast Oxide Reactor Fuel Elements,"
Trans. Am. Nucl. Soc., 7_, 92 (1964).
63. Reference 16, p. 538 ff.
64. J. L. Phillips, "Full Power Operation of the Dounreay Fast
Reactor," Proc. of National Topical Meeting on Fast Reactor
Technology, Detroit. April 26-28, 1965, ANL-100. p. 7 (1965).
65. J. Kirk, "Radioactive Maintenance on DFR," Trans. Am. Nucl. Soc.,
1_3, 789 (1970).
66. J. P. Foster, et. al., Analysis of Irradiation-Induced Creep
of Stainless Steel in Fast Spectrum Reactors,
Proceedings of the BNES, London, 1973.
67. Letter from G. W. Hardigg, General Manager, Westinghouse Advanced
Reactors Division, to E. D. Harward, Director, Technology
Assessment Division, EPA, Comments on Draft Report, February 28, 1974.
121
-------
Page Intentionally Blank
122
-------
7. FISSION PRODUCTS IN SODIUM SYSTEMS
Fission products may enter the primary sodium from the fuel,
either through failures in the cladding, or from the purposeful venting
of fission product gases to the sodium. The extent of fission product
release from fuel is discussed in Section 6.
In that section it was noted that fission product gases (e.g. Xe
and Kr) escape from both fuel failures and vented fuel. Iodine is
volatile at fuel temperatures and some escapes; some cesium and
rubidium enter the sodium by the decay of(xepon and krypton
precursors before these gases escape from' the sodium. The major source
of fission products such as ^37CSj 90sr, 140ea, 95zr, ancj 141ce
is the leaching of fission products from the fuel in large cladding
ruptures. The release of fission product tritium was discussed in
Section 5.1.3; it can be assumed that all tritium from ternary fission
enters the sodium.
Much of the interest in cleanup of solid fission products is
generated by analyses of reactor accidents in which relatively large
amounts of fuel might melt and come directly into contact with sodium.
Although this report is concerned with normal operation, some solid
fission products enter the sodium even in normal operation; therefore,
results are presented for the behavior of solid fission products in this
report.
In the first part of this section, a review of fission product
behavior in sodium is presented, including operating experience for
sodium-cooled reactors. In the second part, the role of the cold
traps in sodium purification is discussed. This second part will
include discussions of (a) cold trap operation, (b) cold trap
experimental results, and (c) cold trap operating experience for
sodium-cooled reactors.
7.1 Fission Product Behavior in Sodium
Fission products entering the sodium generally experience one
of the following fates: (a) escape to the cover gas, (b) deposit
on system surfaces, (c) removal by a cold trap, or (d) they remain in
solution in the sodium. The thermodynamics of vaporization (hence,
of process a above) is reviewed by Castleman and Tang.'
A review of fission product behavior in sodium is given by
Castleman.2 His article forms the basis of much of the review here,
although it has been augmented with additional material. Of particular
interest is a loop experiment reported by Plumlee and Novak^ in
which fission products and fuel were leached from a purposely defected
fuel pin by flowing sodium, and the fate of the fission products in the
loop was determined.. (This experiment was also reviewed in Section 6.) In
particular the relative retention in the sodium versus deposition
on the cold steel surfaces of the system was measured for '^'Cs
^alkali metal), 90Sr (alkaline-earth metal), 144Ce and 147Pm (rare earths),
123
-------
the results of which are discussed in the sections below. Experiments
by Sarour>5 also provide information on fission product behavior
in sodium. Following the review of each type of fission product,
operating from sodium-cooled reactors is reported.
7.1.1 Behavior of Each Fission-Product Type
7.1.1.1 Noble Gases
The noble gases of principal interest are xenon and krypton.
Noble gases escape into sodium from leaking fuel (see Section 6).
Effectively all of the noble gases then escape from the coolant to
the cover gas within a few minutes after entering the sodium (See
Reference 6,7 for example).
The time delay before escape, however, allows some decay of
xenon to cesium and krypton to rubidium in the sodium. For example,
138Cs results from decay of '38Xe. With the vented fuel of Dounreay
the on-line activity of the NaK coolant is 100 times the 24^|a
activity, with !38Cs reported as being the major contributor.8 Some
'38Xe (half-life 17 min.) may escape from failed fuel, but the
half-life of '38Cs is only 32 min.; hence, it would not be an
important long-term environmental source in the sodium. Also some
13'Cs is produced from decay of !3?Xe in the sodium; however the
short half-life of '3'Xe (4.7 min.) prevents much of it from escaping
from failed fuel. Although significant 135Xe (9.2 hr. half-life)
escapes from the fuel, its daughter ^3^Cs is relatively stable
(2 x 1()6 y). Perhaps some 88Rb is produced in the sodium from decay
of 88Kr (2.77 hr. half-life) which leaks from failed fuel, but the
half-life is short (18 min.) so that it is not a long-term environmental
problem. The daughters of 8'Kr and 85Kr are stable; the resulting
87Rb and 85Rb could activate to 88Rb and 86Rb, with half-lives of 18 min.
and 18.7 days respectively.
Vented fuel elements can be designed to delay fission gas transfer
to the sodium and thereby substantially reduce the entrance of the
short-lived noble gases into the sodium, as described in Section 6.7.
This was not the case for the vented fuel in Dounreay, however,
since the NaK was, in effect, able to contact much of the fuel directly
in Dounreay. As noted above, in Dounreay ^38Cs, which is the
daughter of ^38Xe, is the main source of activity in tbe coolant
during on-line operation,8 although the half-life of '38Xe is only
17 min.
Saroul reports experiments in the Pirana and Aetna facilities
in which fission gases from molten irradiated uranium were allowed to
enter first sodium and then an argon cover gas. He reported
significant retention of noble gases by the sodium. However, un-
certainties concerning whether equilibrium was reached and the meaning
of material balances led Castleman2 to emphasize other noble gas
solubility experiments to argue that noble gas retention in the sodium
should be negligible.
124
-------
7.1.1.2 Iodine
Iodine which enters the sodium reacts with sodium to form Nal.
Sodium iodide remains in solution in the liquid, with only small amounts
being vaporized into the cover gas, still as Nal. Extensive data
on Nal volatility have been reported by Castleman and Tang (e.g.
Reference 1) and by Pollock, Silberberg, and Koontz (e.g.
Reference 9). In Reference 9, relative volatility data for NaT
are reported in terms of a distribution coefficient, KJ, defined as
the ratio of the mole fraction of solute in the vapor to the mole
fraction of solute in the liquid.
Sodium iodide does not generally react chemically with other
fission products. Some reactioanwith cesium to form Csl is possible;
but Castleman, Tang, and Mackay'u showed experimentally that, for the
low concentrations and for the sodium temperatures involved in reactors,
Csl readily decomposes to Nal and Cs. This lack of reaction between
Cs and Nal was also confirmed by Cooper, Grundy, and TaylorJl
It has been observed (for example in the Pirana experiments ) that
in stagnant sodium a large fraction of the iodine is concentrated*
near the gas-liquid phase boundary. In other Pirana experiments,
when argon was bubbled through the sodium the iodine was distributed
homogeneously in the sodium.
Fission product isotopes of another halogen, bromine, are of
sufficiently short half-life not to be of environmental concern.
7.1.1.3 Alkali Metals
Cesium and rubidium are alkali metals, with cesium being the more
important for environmental considerations.
Plumlee and Novak found that cesium is retained in sodium far
more than any other fission product, which might be expected since
sodium itself is an alkali metal. They report that in their loop
experiment, 50% of the '"cs which was leached from the fuel remained
in the sodium and 50% plated out on the colder loop surfaces.
This finding is somewhat consistent with EBR-II experience (Section
7.1.2.2) and BR-5 experience (Section 7.1.2.3). It is also consistent
with the results of Clifford,34 which is described in Section 7.2.3.1
on cold trapping of cesium.
Cesium is present in both sodium liquid and sodium vapor as
elemental metallic cesium. Cesium reacts little with other fission
products. Cesium will react with carbon if presentJ2
Cesium is highly volatile in sodium. Experimental results by
Pollock, Silberberg and Koontz^ and theoretical and experimental
results by Castleman and Tangl indicated high volatility of cesium
relative to sodium (far higher, for example, than Nal). Clough and
confirmed these high volatilities. They found, further, that
125
-------
the volatility was decreased significantly either by adding graphite
or charcoal to the sodium or the gaseous phase. Cesium apparently
both reacts with graphite and is adsorbed on graphite surfaces.
The concentration of cesium does exhibit some inhomogeneity at
a sodium liquid-gas interface, with higher concentrations beinq found
near the surface than below the liquid level. On draining steel
vessels which contained cesium dissolved in sodium, Saroul found that
appreciable cesium remained at the vessel surface and a significant
amount had penetrated the vessel wall to a 3 to 4 p depth4 ,5
Although less work has been reported on rubidium, its properties
are similar to those of cesium. For example lastleman reports
thermodynamic properties for rubidium which indicate that it is also
highly volatile relative to sodium.2
7.1.1.4 Alkaline-Earth Metals
Alkaline-earth fission products include strontium and barium.
Little of these materials enter the primary sodium from fuel failures.
However some 89Sr, 90Sr, and '40Ba has been observed in the sodium
from failed fuel in operating sodium-cooled reactors.
3 90
Plumlee and Novak reported that, of the Sr that entered the
sodium in their loop experiment, only 0.024% remained in the sodium.
Hence nearly all of the 90Sr presumably plated out on the system walls.
The alkaline earth metals have low volatility in sodium.
Castleman reports that their chemical state in sodium is not well
established; they probably interact with dissolved oxygen in sodium.
but the nature of the oxygen compounds in sodium is not well known.2
Clough reports experimental values for strontium volatility in
sodium that are lower than expected for elemental Sr, indicating that
some relatively nonvolatile oxygen species has been formed." Later
Clough and Wade again suggest that barium and strontium are present
in sodium as BaO and SrO.'2
4 i40 140
Saroul reported in the Pirana experiments that Ba - La
tended to concentrate near the liquid sodium-argon gas boundary in
stagnant sodium.
Saroul also reported4 that most (83%) of the 140Ba - 140La
released from the uranium into the sodium deposited on the stainless
steel walls of the sodium vessel upon removal of the sodium at 250°C.
7.1.1.5 Rare Earths
Rare earth fission products include cerium, lanthanum and pro-
methium. Little cerium and promethium ente,r .the sodium1 from fuel •,..
failures although Plumlee and Novak3 found significant amounts of Ce
and !47Pm leached in their experiment. Lanthanum-140 is produced by
decay of the alkaline earth 140Ba and is found with 140Ba.
126 ; >
-------
Plumlee and Novak report <0.0024% retention of Pm in sodium;
hence nearly all of these two fission products plated out on the
system walls. Saroul also showed that most of tbe cerium and lanthanum
is transported to the wa-lls of sodium systems.4'5 The rare earths
are relatively nonvolatile.
7.1.1.6 Transition Metals
95Zr 95 .
Transition metals include among fission products.
Information on their behavior in sodium was not found. Such small
amounts of these two isotopes were leached in the Plumlee-Novak
experiment3 that relative retention in sodium and deposition on system
surfaces could not be measured.
7.1.1.7 Noble Metals
Noble metal fission products include palladium, rhodium, and
ruthenium. Little information on their behavior m sodium was found.
Plumlee and Novak report that a large amount of '00Ru was leached
from the fuel in their experiment and less than 0.023% was retained
in the sodium.3 Presumably, this means that most plated out on the
system walls.
7.1.1.8 Tritium
The fission product tritium, and its behavior in sodium, are
discussed in Section 5.1.
7.1.2 Operating Experience with FissionProducts in Sodium (or NaK)
Cooled Reactors (Excluding Experience with Cold Traps)
7.1,2.1 Summary
14
Zwetzig has reported a summary of fission-product operating
experience in the coolant systems of sodium or NaK cooled reactors.
Table 7.1 is patterned after Zwetzig's summary; it includes his
results plus additional results as referenced in Table 7.1.
7.1.2.2 EBR-II
Activities of various radionuclides in the primary system of
EBR-II during 1971 (the last year they were publicly reported) are
listed in Table A25 of Appendix A. The principal fission products
observed are ^37Cs and13'I. Also observed was the activation product
^34cs which results from activation of the fission product '33Cs.
In July, 1971, the fission and activation products on the pump
walls of the primary pump were reported.^ 137cs was found on the
pump walls; 65% of the '3'Cs was removed by cleaning the surface.
Also Cs was reported in the walls of the primary tank at
the argon cover gas level. This was believed to have resulted from
vaporization and subsequent recondensation of '37Gs.
127
-------
Table 7.1
Fission Products Observed in the Primary System of
Sodium and NaK Cooled Reactors (other than tritium)
Neutron Spectrum
Coolant
In Primary Coolant
)n Primary Piping
)r Pump
fn Cold Trap
Fermi14
fast
Na
140Ba-La,137Cs,
89Sr,13V
141Ce, 144Ce, 133I,
103Ru, 95Zr-Nb
v
BR-517
fast
Na
144Ce, 141Ce, 144Pr,
140Ba-La, 137Cs
136Cs106Ru,
95Zr-Nb, 90Sr, 131l'
137CS, 136Cs, 131I5 133I
135I, 95Zr-Nb, 140Ba-La
EBR-II15
fast
Na
137- 131T
OS $ i
137Cs
137Cs, 134CS
8
Dounreay
fast
NaK
141Ce, 144Ce, 132Te
131I, 103Ru, 106Ru,
132I,'137Cs. 95Zr-Nb,
140Ba-La, 138Cs
140Ba-La
137Cs
ro
oo
-------
Table 7.1 (Continued)
TO
Rapsodie
SEFOR
(See Appendix A,
Table A21)
SRE14'20
S8ER
14
Neutron Spectrum
fast
fast
thermal
thermal
Coolant
Na
Na
NaK
NaK
In Primary Coolant
137
Cs
ro
vo
86
Rb
141
Ce,
103
Ru
137. 131T 132To T
Cs, I, Te-I
137Cs, 83Sr, 90Sr, 95Zr-Nb,
140Ba-La, 144Ce, 106Ru
On Primary Piping
or Pump
141ro 137rc 131 T
Ce, Ls, I,
R_ ,
bd-i_a ,
95Zr-Nb, 90Sr-Y,
91 „
89Sr, 90Sr, 95Zr-Nb,
144Ce, 137Cs, 106Ru
89Sr, 90Sr, 95Zr-Nb,
103RU 144Ce, 106Ru-Rh
140Ba-La, 141Ce
In Cold Trap
137C5, 106Ru, U4Ce-Pr
-------
An interesting results concerns Cs segregation in the primary
sodium system of EBR-II at low temperature.'° After a reactor
shutdown on November 15, 1970, the primary pumps were turned off and
the sodium was cooled to 350°F on November 17. Sampling of '37Cs and
22fta continued during and after this time. The 22^ activity in the
sodium remained constant. The ^7QS activity, however, steadily
decreased from 11 nCi/gm Na to 4 nCi/gm in one month. This decrease
can be seen in Table A25 of Appendix A, as reported in Reference 16.
It was supposed that the 137c$ segregated from the bulk of the
quiescent sodium and concentrated at the sodium-metal and sodium-gas
interfaces in the primary tank. After the sodium was reheated and
operation again started, the '37Cs activity returned to its original
value, as can be seen from later results in Table A25, Appendix A.
7.1.2.3 BR-5
The following results for BR-5 operation were obtained from
Reference 17.
The USSR sodium-cooled Pu02~fueled 5MW(th) BR-5 fast reactor was
operated for 8 years from 1959 to 1967. At the end of the first stage
of operation (1962-1964) there were 63 assemblies with Pu02 fuel with
5.0 - 6.5% fuel burnup in the core. Between 1964 and 1967, the reactor
was operated with PuC fuel, to 2.4% burnup. Integrated power for the
8 years was 4100 MW days.
During the eight years of operation there was no situation
endangering the integrity of the reactor because of sodium leakage
from the heat-transfer system. No sodium leakage occurred at pipe
welds. Isolated leakages did occur in liquid metal fittings, through
the level metering devices, in the heat exchanger equipment, and
through a fault in the drainage piping of the primary circuit. Four
of 65 valves were replaced due to sodium leaks.
A unique feature of BR-5 operation was long-term operation with
an excessive number of fuel failures. Before completion of operation
with the Pu02 core, 17 of the 63 fuel assemblies contained failed fuel.
The concentrations of fission product activities in the sodium and in
the primary system walls at the end of the first stage of operation is
given in Table 7.2.
Before 2% burnup, the residual activity in the sodium was due
only to 22Na. At 3% burnup, 137Cs was detected ( * 20% of the 22^
activity). At 5% burnup, 137Cs was 200 times its activity at 3%
burnup, and other fission products were found in the primary sodium
(see Table 7.1).
7.1.2.4 Dounreay
Dounreay fuel is U-Mo alloy and is vented to the NaK coolant.
During operation the fission product activity is ^ 100 times the 24Na
'activity, the dominant isotope being 138cs ('3oQS activity = 0.6 Ci/gm
NaK).8
130
-------
Table 7.2
Fission Product Activity in BR-5 During
First Stage of Operation (1962-64)
Isotope Activity
Primary Sodium 1311 0.8 mCi/liter
137Cs 7 mCi/liter
QC
yDZr-Nb 0.3 mCi/liter
140Ba-La 2 mCi/liter
Walls of Primary System 131 j 7Q mC1/cn]2
137Cs 74 mCi/cm2
95Zr-Nb 55 mCi/cm2
140Ba-La 19 mCi/cm2
Table 7.3
Gamma Activity of Fission Products in DFR Coolant,
6 Days After Sampling
?s>- Activity (nCi/gm NaK)
0.7
1.2
6.8
0.7
7.3
2.3
8.8
Energy (MeV)
0.14
0.22
0.364
0.5
0.67
0.76
1.6
Possible Isol
143Ce, 144Ce
132Te
131I
103Ru, 106Ru
132T 137-
I , Cs
95Zr-Nb
140Ba-La
131
-------
The fission product gamma activities 6 days after sampling
in the DFR coolant are given in Table 7.3.8
7.1.2.5 Rdpsodle
After three years of operation, including 500 equivalent days at
24 MW, Rapsod'te was shut down for conversion to 40 MW operation. At
that time a study of fission oroducts in the sodium was made and the
results are summarized here.'° Operation had proceeded with one
failed fuel pin, with direct contact of sodium coolant and the U02-PuOp
fuel. \
Cs and Ba were the main fission products deposited on
primary system pipesV
141
Fis5Jon P^ucts deposited on the P^WAfy PumP included: Ce,
137Cs, 131I, 13Zi. 140B^La, 95Zr-Nb,and 90Sr-Y. The axial
distribution of ^'Cs was plotted in Reference 18 for the primary pump.
137
The Cs level in the primary sodium rose steadily to 0.05wCi/gm
Na at the end of the 500 effective days of operation at 24 MW.
The primary pump was decontaminated by alternate washing in
water and dilute nitric and phos,p}»bric acids. A 90% decontamination
factor was obtained for 13'Cs, wjtich was not considered adequate for
future work. A sample steel bolt from the pump was washed with alcohol
with a resulting 99% decontamination factor, but it was considered
too dangerous to use alcohol/for the,entire pump.
7.1.2.6 SRE
During Run 14 of SRE, from July 12-26, 1959, fuel element cladding
failures occured in 14/of the 43 elements. The total accumulated
irradiation through Ran 14 was 2426 MWd. The fuel in SRE was
metallic uranium, bonded with NaK, and clad in stainless steel. gn
The fate of fissiop products from these failures is well documented
and is reviewed here in some detail. Unfortunately some uncertainty
exists on its direct applicability to an LMFBR system since 7 to
70 Ibs. of carbon were also in the system.
Prior to Run 14, small amounts of fission products were found
in the primary sodium. The fission product levels detected prior to
Run 14 are/given in Table 7.4, which is reproduced from Reference 20.
/ \
After Run 14 the fission product levels rose to the values listed
in Table 7.5, again reproduced from Reference 20. It is interesting
to note7 in this table that the variation in fraction of isotope
released to the primary sodium was only a factor of 10 between the
lowes't ('03Ru) and the highest (137cs) isotope.
/ In Table 7.6 are listed primary sodium levels for three sampling
, dates — at the end of Run 14,3 months later, and one year later.
132
-------
The fission product gamma activities 6 days after sampling
in the DFR coolant are given in Table 7.3.8
7.1.2.5 Rapsodie
After three years of operation, including 500 equivalent days at
24 MWS Rapsodie was shut down for conversion to 40 MW operation. At
that time a study of fission products in the sodium was made and the
results are summarized here.'" Operation had proceeded with one
failed fuel pin, with direct contact of sodium coolant and the U02-PuOo
fuel.
Cs and Ba were the main fission products deposited on
primary system pipes.
141
Fission products deposited on the primary pump included: Ce,
137Cs, 1J1I, 13^I, l40Ba-La, 95Zr-Nb,and 90Sr-Y. The axial
distribution of ^37Cs was plotted in Reference 18 for the primary pump.
137
The Cs level in the primary sodium rose steadily to 0.05wCi/gm
Na at the end of the 500 effective days of operation at 24 MW.
The primary pump was decontaminated by alternate washing in
water and dilute nitric and phosphoric acids. A 90% decontamination
factor was obtained for '37Cs, which was not considered adequate for
future work. A sample steel bolt from the pump was washed with alcohol
with a resulting 99% decontamination factor, but it was considered
too dangerous to use alcohol for the entire pump.
7.1.2.6 SRE
During Run 14 of SRE, from July 12-26, 1959, fuel element cladding
failures occured in 14 of the 43 elements. The total accumulated
irradiation through Run 14 was 2426 MWd. The fuel in SRE was
metallic uranium, bonded with NaK, and clad in stainless steel. ?Q
The fate of fission products from these failures is well documented"
and is reviewed here in some detail. Unfortunately some uncertainty
exists on its direct applicability to an LMFBR system since 7 to
70 Ibs. of carbon were also in the system.
Prior to Run 14, small amounts of fission products were found
in the primary sodium. The fission product levels detected prior to
Run 14 are given in Table 7.4, which is reproduced from Reference 20.
After Run 14 the fission product levels rose to the values listed
in Table 7.5, again reproduced from Reference 20. It is interesting
to note in this table that the variation in fraction of isotope
released to the primary sodium was only a factor of 10 between the
lowest (^03RU) anc| the highest (137cs) isotope.
In Table 7.6 are listed primary sodium levels for three sampling
dates -- at the end of Run 14, 3 months later, and one year later.
132
-------
Page Intentionally Blank
-------
Table 7.4
Typical Radioactivity Levels of SHE
Primary Sodium Prior to Run 14 20
Sample tto.
Sanplrng Location
Date of Sample Ranoval
Date of Last Reactor Scram
95Ito-Zr
137CS
R-24
Material Evaluation
Facility
10/2/58
9/25/58
5. 9xl
-------
Table 7.5
Initial Fission Product Analysis of SRE
Primary Sodium After Run 14
20
Isotope
137CS
"-as
89 Sr
<>°sr
131!
i^Ce
l^Ce
llt0Ba-La.,:-
,->5Zr-Nb
103BU
Primary Coolant Activity
(pCi/gm Na)a
1.26
0.02
20.0
0. 97
0.74 ...,-•"'""'
4.38 ,.-.-"•""""
^•5.18'"'
"""" 1. 65
13.9
0.95
Total Coolant Inventory
(curies')3
2.77X101
-4X10-1
4.44xl02
. 2.14X101
i:63xlOi
9.65X101 "'
1.41xl02 ""-•-„.
3.63X101
3.06xI02
2.09X101
Ttotal Reactor Inventory
(curies)^ ,---"""
8.70X10.3
2x10 2C
i.soxins
8. 15x10 3
1.68X101*
1. 27x10 5
1.69xl05
""•-'-.,5. 61x10"
5.53x13^
7.52xlOu "--,
-^SST"—
3.1?xlO~3
2xlO-3c
2.78x10"-
2.63xlO"?
0. 97x10- 3
0. 76x10- 3
0.67xlO-3
0. 65x10- 3
0.-5X10-3
0.28xlO~3
(a) As of July 26, 1959
(b) Multiply values in this column by - 3 to adjust fraction released to average values for those fuel elements
which suffered cladding failures (14 of 43 elements failed).
(c) From neutron capture in 133Cs; estimated.
-------
Table 7.5
Initial Fission Product Analysis of SRE
Primary Sodium After Run 14
20
Isotope
137Cs
13 ''Cs
89 Sr
9°Sr
131!
l«Ce
i^Ce
lOOBa-La
9 5Zr-Nb
103RU
Primary - Coolant Activity
(uCi/gm Na)a
1.26
0.02
20.0
0.97
0.74
4.38
5.18
1.65 •
13.9
0.95
Total Coolant Inventory
(curies) a
2.77X101
~4xlO-J
4.44xl02
2. 14x10 l
1.63X101
9.65X101
1.41xl02
3.63X101
3.06xl02
2.09X101
Total Reactor Inventory
(curies) a
8.70xl03
2xl02C
1.60xl05
8. 15x10 3
1.68X101*
1.27xl05
1.69xl05
5.61X101*
5.53xl05
7.52X101*
Fraction of Inventory
Released13
,3.18xlO"3
2x10-3°
2.78xlO-3
2.63xlO"3
0.97xlO"3
0.76x10-3
0.67xlO-3 -
0.65xlO-3
0.55xlO-3
0.28xlO"3
(a) As of July 26, 1959
ft>) Multiply values in this column by -3 to adjust fraction released to average values for those fuel elements
which suffered cladding failures (14 of 43 elements failed).
(c) From neutron capture in I33cs; estimated.
-------
Page Intentionally Blank
-------
Table 7.6
Fission Product Analyses of SEE Primary Sodium
As a Function of Time After Run 14 20
Primary Ccol.mt Acliivity (yCi/g Na)
Sanple Date
Time After Run 14
137Cs
13^CS
89Sr
90Sr
13l!
^Ce
i^Ce
11+0Ba-La
95Zr-Nb
103Ru
7/26/59
0
1.26
0.02
20.0
0.97
0.74
4.38
5.18
1.65
13.9
0.95
10/31/59
97 days
0.45
0.006
0.25
0.060
0.00012
0.000088
0.00031
—
0.0067
0.0045
7/26/60
1 year
0.028
Undetectable
Not analyzed
Not analyzed
Undetectable
Undetectable
Undetectable
Undetectable
Undetectable
Undetectable
Ratio
EQct. 31 (actual)"
Oct. 31 (decay
only) _
0.36
0.3
0.043
0.062
0.63
0.00016
0.00008
0.0013
0.024
135
-------
Also listed are the ratio of the Oct. 31 results to the values which
would result if radioactive decay were the only loss mechanism for
the isotope. The fact that the ratios are below unity indicates
that other mechanism such as deposition on primary system walls or
deposition in the cold trap are effectively removing fission.products
from the sodium. Discussion of cold trap purification of the SRE
system is given later in this section (see Section 7.2.414). By
July 26, 1960 (one year later), the 13TI, 140Ba-Las and 141Ce had
decayed to the extent that they were undetectable. However the
decreases in '44Ce, '03Ru, and 137Cs were attributed to other removal
mechanisms, such as cold trap cleanup. A strontium analysis was not
made in the final sample.
Analysis of fission product activity on primary pipe samples
was also made. Residual sodium on samples of pipe walls was
removed by methanol and water. Next the pipe was subjected to a
series of etches by hydrochloric acid. Sodium, methanol wash, and
HC1 etch solutions were analyzed. An example analysis of the etch
solution at the surface is given in Table 7.7. In addition.
Reference 20 shows a graph of the activities of ^Ir-Hb, '^Ce, and
137Cs as a function of depth into the pipe wall, to a depth of 0.2 mils.
7.1.2.7 SEFOR
oc
The only fission product reported in the SEFOR sodium was Rb
(see Appendix A, Table A22), which is actually an activation product
of 85Rb which results from decay of 85|
-------
Table 7.7
Example of SEE Primary Pipe Wall Fission-Product
Contamination from HC£ Etch at Pipe Surf ace 20
Isotope Contamination Level*
(pCi/cm2)
15.2
90Sr 0.78
95Zr-Nb 2.7
i^Ce 2.1
137Cs 0.022
*Corrected for radioactive decay since 7/26/59 for
coraparative purposes.
137
-------
Other nonradioactive impurities removed by cold traps include carbon
and hydrogen.
Cold traps are also effective in removing some fission products
from the sodium. Removal occurs even when the fission product
concentration is lower than the saturated value for the material at
the cold trap temperature. For some materials, such as cesium and
sodium iodide, the concentration in sodium at a metal surface is
higher than the concentration in the bulk liquid, and adsorption or
some other transfer mechanism occurs at the surface to remove the
material from solution.
A brief review of cold trap operation and experience is given by
Hinze."
In a cold trap a bleed stream from the main sodium system (i.e.
the primary or secondary system) is cooled, and precipitation of the
impurity (e.g. Na20) occurs. A large surface.on which the impurities
are collected is present in the trap, frequently in the form of
stainless steel mesh. The collection process includes one or more of
the following processes and operations: crystal formation and re-
tention on metal surfaces, filtration, and settling. After leaving
the cold trap (or the steel mesh part), the sodium is reheated and
returned to the main system. Initial reheating is generally done in
an economizer (usually, but not necessarily, external to the cold trap)
in which the exiting stream is heated by cooling the incoming bleed
stream.
22
Hinze describes early designs and experience for the cold traps
for the Submarine Intermediate Reactor (SIR), Fermi, EBR-II, Sodium
Reactor Experiment (SRE), Hallam (HNPF), and Dounreay, and also
reports some USSR experience. Both the Fermi and the EBR-II primary
cold traps were 500-gallon traps containing stainless steel mesh.
Only one trap was in each primary system. The traps were run until
they "plugged" with oxide, i.e. until the pressure drop across the
trap due to oxide deposition increased such that insufficient flow
could be maintained. In some cases (e.g., the primary trap in Fermi
and the secondary trap in EBR-II) the trap plugged early during the
purification of the sodium. Then a new trap was installed which has
not yet required replacing. The primary cold trap in EBR-II lasted
until June, 1968, when it was replaced.
SEFOR had two primary cold traps, only one of which was operated
at any one time. Each trap had to be replaced after one year (^1500
hours of operation each) due to plugging.23 Total ^0 collected was
^200 Ib. Excessive oxide buildup had resulted from zero power
operation with the vessel cover removed, when the argon in the
refueling cell and over the sodium was contaminated from excessive
leakage of nitrogen (and oxygen impurity) from adjacent cells.
The first Fermi primary trap was also removed prior to power
operation. The trap was examined by Westinghouse; however, the
138
-------
24 '
report * is not available to the public and no final public report was
issued. The report was available to Hinze, however, because he
reports in Reference 22 that the trap was found to contain 50 Ib. of
oxygen, 1 Ib. of carbon, and lesser amounts of hydrogen, nitrogen,
and metallic impurities.
The above experience indicates that it is difficult to predict
how long a cold trap will last in an LMFBR power plant. Hence, it is
difficult at this stage to estimate how often cold traps, with their
accompanying charge of fission and activation products, will be
shipped away from the reactor for decontamination or storage in the
environment.
Brief descriptions of cold traps appear in the 1000 MWe follow-on
reports. For example Reference 25 (the GE design) shows four primary
system cold traps operating in parallel and six secondary system cold
traps. Each trap (in both systems) has a 20 cu ft volume (150 gallons)
and a maximum oxide capacity of 560 Ibs. The traps are constructed
of 304 stainless-steel, each with a 35 inch high and SS^nch^tHameter
bed of stainless-steel packing. The traps are cooled by forced
convection of the cell atmosphere nitrogen. The economizers are not
"built-in", but are separate from the cold trap.
7.2.2 Cold Trap Decontamination Terminology
It is useful to review some of the theory and definitions
concerning cold traps in order to appreciate the data on fission product
removal reported in the literature. Different reports quote a
variety of measured or design quantities but no summary of the
relations between these quantities was found. Hence, this background
is provided in this section.
A number of reports on removal of fission products by cold
traps report values for a surface deposition constant K. This constant,
with units of length, is defined as:
2
., _ grams deposited/cm of deposition surface area
- 3
grams/cm concentration in sodium
This constant is generally found to be inversely proportional to the
absolute temperature of the sodium. Experimental information for cesium
and iodine is reviewed below in Section 7.2.3.
Another parameter that is used in cold trap technology is a
"decontamination factor," D. For a particular nuclear species,
this factor is defined as: Concentration in coolant
D =Decontamination factor = without cold traps operating
Concentration in coolant
with traps operating
* Personal communication, P. Cohen, Westinghouse, December 21, 1972.
139
-------
Another related term In use is a "Concentration ratio", C, defined
for a particular nuclear species as: 3
Grams In trap/cm sodium in trap
C = Concentration ratio = ConcentratTon in remainder of the
system
The factors D and C are related as follows:
Let X = concentration in the system without a trap
Y = concentration in the system outside the cold traps, with cold
traps operating.
Vs = system volume
V = cold trap volume
If traps are operating, the concentration in a trap is CY. From
conservation of total production of the particular nuclide with or
without trap operation,
XVs = YVS + CYVT
The decontamination factor, D,, is related to X and Y by: D = X/Y.
Therefore
VT
D = 1 + C «i
s
Values of concentration ratio found at SRE are given in Section 7.2.4.4.
Cold trap efficiency, e is frequently reported, where e is the
efficiency for precipitation, defined as:
e = entering concentration - exit concentration
~ entering concentration - saturation concentration at the
minimum cold trap temperature
None of the parameters listed above are related to rate of
deposition in a cold trap. The experimental work found on cold trap
deposition did not provide information on deposition rate. Atomics
International does, however , provide for a rate calculation of cold
trap deposition in their STP-1 fission product transport code.26,27
Shown below is an equation similar to one proposed by Atomics
International to describe the removal rate of material i from the
sodium in a cold trap:
i ., "i
- ai l + _L_ | ' ~ c |
PZ V /
140
(D
-------
where N 7 = atoms of nuclide i in coolant (atoms)
c
•3
V = volume of coolant in total primary (or secondary) system (cm )
e1' = cold trap efficiency for removal of nuclide i
o
F = flow rate in the trap (cm /sec)
o1 = solubility of nuclide i in,the trap at the minimum cold
trap temperature (atoms/cm ^
2.
P = surface area per unit length of traps (cm )
= deposition rate parameter of the trap /atoms deposited/sec/cm
(atoms deposited/sec/cm \
atoms/cm /
Z = length of trap (cm)
Only depletion is considered in the above equation (no source term is
included).
N,
The factor (--'— ° !) is a measure of the excess of the concentra-
tion above saturation, and therefore the first term of Equation (1)
represents removal by precipitation. This term is assumed to be
positive or zero in the AI codes.
The second term of Equation (1) represents deposition by
adsorption. The following derivation of the second term is useful to
provide an understanding of it:
Let N (t)/V = Concentration of an impurity entering the trap
c (equal to the average concentration in the total
system, (atoms/cm^)
N (z,t)/V = concentration at height z in the trap
^ \*
The change in N as a function of height is:
N(z,t) p
C ic r
Vc F
Integrating from z = 0 to Z, and letting N (Z,t) = N (t) at z = 0, gives
Nc(Z,t) = Nc(t) e-
Next we consider the time dependence of N _(t). The transit time of
the sodium in the cold trap is PZ/F. Hence,
141
-------
dNc(t) Nc(Z,t) - Nc(t)
—ar~ ="
which is the second term of Equation (1).
Information on the deposition rate parameter K.was not found.
Perhaps it is dependent on conditions in the cold trap which vary with
time, thus making Ktime dependent.
It is noted further that the parameter rein Equation (1) is a
deposition rate (i. e. per sec) whereas the deposition constant K
is an equilibrium-type value and not related to a rate. Although
many of the experiments which report K are made in flowing sodium,
it remains unclear how the two constants sand K are related -- a re-
lation that is necessary before using Equation (1).
7.2.3 Experiments on.Cold Trapping of Particular Radionuclides
7.2.3.1 Cesium
28
Cooper and Taylor of Westinghouse studied cesium sorption from
sodium by the following surfaces: polished 304 stainless-steel, as-
received 304 stainless-steel, polished nickel, single-crystal
aluminum oxide, and oxidized zirconium. Cesium concentrations from
<0.1 apm (atom parts per million) to 46 apm were studied.
It was concluded that cesium was sorbed by Van der Waal forces
as opposed to chemisorption. Numerical results showed the sorbed
cesium surface concentration (atoms/cm?) to be inversely proportional
to temperature and directly proportional to cesium concentration in
the sodium (atoms/cm3).
Later the same experimenters ran experiments on cesium trapping
by 304 stainless steel to study the effect of N«20 on deposition
rate.29 Cesium was cold trapped from almost oxygen-free sodium and
from sodium containing oxygen. Initial cesium concentrations were
0.13 and 0.059 apm. Their results are reported as fraction of the
initial cesium removed by cold trapping, and this fraction varied from
0.18 to 0.52.
Among their conclusions were:
1. Cesium is removed from flowing sodium by reversible physical
adsorption on metal surfaces in the absence of precipitated Na20, or
by adsorption on both metal and Na20 surfaces in the case of
precipitation.
142
-------
2. Precipitation of ^0 increased the Cs fraction removed
from the sodium. Values are given for atoms of cesium deposited
per cm2 of surface and per gram of sodium for various conditions
(particularly oxygen concentrations) and for various flow rates
(flow rate had little effect).
3. Adequate LMFBR Cs traps can be designed based on adsorption
on clean metal surfaces. Precipitation of f^Q in this trap would
increase the capacity.
pc
Zwetzig, Guon, and Silberberg of Atomics International showed
relative trapping levels by stainless steel for cesium concentrations
of 65 ppm and three different oxygen concentrations (5, 55, and 105
ppm), as a function of temperature. The deposition levels increased
with increasing Na20 concentration, and the iog of the deposition
level was inversely proportional to temperature. Deposition occurred
at temperatures above which Na20 had not precipitated, indicating
that adsorption occurs directly on metal instead of on Na20. Later
studies of deposition of cesium on 304 stainless steel in the range
of cesium concentrations of 0.7 to 6 ppm and oxygen concentrations
from 10 to 25 ppm were reported by Guon of AI.-^O Among the conclusions
were:
1. Cesium deposition requires the presence of a third constituent.
2. Cesium deposition and dissolution kinetics are rapid with
no apparent hysteresis.
3. A deposition constant, K, (defined in Section 7.2.2) can
be used to express the partition of cesium between the sodium solution
and stainless-steel wall, in agreement with earlier results from
Westinghouse^S for different cesium concentrations.
4. Surface treatments of stainless steel prior to sodium
loading can result in increased cesium deposition by a factory of
10 and possibly 100. The surface treatment referred to concerned
the temperature history of the surface prior to deposition; the
report shows a relation between K and surface temperature.
31
Further studies by Guon showed further distribution coefficients
(called a "partition parameter" in Reference 31) for cesium, barium,
and manganese on stainless-steel surfaces.
Recently Colburn of Westinghouse has presented two papers
summarizing work there on cold trapping of cesium and iodine.32,33
In the first32 he reports distribution coefficients for both 137Cs
and '3'I. Further conclusions presented in the paper were:
1. Large Cs deposits observed were not due to physically adsorbed
metallic Cs but, rather, are part of a nonmetallic precipitate.
143
-------
2. The distribution coefficient for Cs and I at cold-trap
temperature is strongly influenced by non-metallic contaminants in
sodium.
33
In the second work Colburn studied mechanisms for cesium and
iodine deposition in sodium on stainless steel which had been previously
exposed to hydrogen or oxygen. Examination showed that the deposition
behavior was dominated by interactions with the nonmetallic
contaminants, i.e. hydrogen or oxygen. Tests showed that hydrogen
was more effective than oxygen. Colburn suggests that the importance
of the surface impurities and possible differences in impurity
concentrations between experiments could have led to earlier dis-
crepancies in cesium surface distribution coefficients, K, reported
in the literature. He reports experimental values for "phase
distribution coefficients," D, at 250°F of 8.5 x 105 for Cs and
2.27 x 106 for I, where
n = atoms of Cs (orl)/gram hydrogen in the deposit
" atoms of Cs (orl)/gram sodium in bulk solution
He suggests that the intentional addition of hydrogen to the sodium
may enhance the ability to cold trap cesium and iodine (while
simultaneously enhancing tritium removal by isotopic substitution
in the hydride precipitate).
34
Clifford showed that some cesium could be removed by cold
trapping, although most of the cesium remained in the sodium in
his experiments. In two loops which operated for 2300 to 2500
hours, equilibrium was believed to have been achieved with the following
cesium distribution: one third of the ^/QS deposited in the cold
trap, one half remained in the sodium, and the remainder was distributed
around the system on stainless steel surfaces. Adding 100 ppm
oxygen to the sodium had little effect on the amount of trapped
although the cause could have been that the oxygen was absorbed
elsewhere in the system than the cold trap. The hot leg of the loops
were operated at 500°C, the cold leg at 300°C, and the cold trap in
the range from 110°C to 175°C. A total of 3 to 4 mCi of 137Cs was in
each trap but no data was provided concerning loop or trap sodium
inventories.
7.2.3.2 Iodine
Cold trapping of iodine fission products appears to be effective.
Two reports on iodine deposition by Colburn32,33 a]so gave results
for cesium; hence they were discussed in the previous section on
cesium.
35
Cooper, Grundy, and Taylor reported experimental values of
the distribution coefficient, K, for Nal in sodium. They found that
log K is inversely Proportional to the sodium absolute temperature, as
was the case with cesium. This relationship held both for low iodine
concentrations ( ^10~6 to 10~9 apm) and for high concentrations (0.05 apm),
144
-------
although the distribution coefficients for the high concentration were
about a factor of five larger than those at low concentrations. In
all cases, more than 90% of the iodine was removed by cold trapping
at 250°F. They also conclude that 99% of iodine may be cold trapped
in high oxygen/hydrogen systems or by the addition of sufficient
natural iodine to increase the concentration beyond the Nal solu-
bility limit at the cold trap temperature.
7.2.3.3 Strontium, Barium, and Zirconium
or
Clifford reported some experience with strontium in cold traps.
He reported that strontium deposited on the stainless-steel and
zirconium surfaces of a cold trap at 300° to 500°C, with the
strontium collection at 300°C being an order of magnitude higher
than at 500°C. Slightly more deposition occurred on stainless steel
than on zirconium.
At BR-5, barium and zirconium were collected in the cold trap,
but much less effectively than iodine and cesiumJ'
7.2.3.4 Tritium
Cold trapping of tritium (a fission product as well as an
activation product) was discussed in Section 5.1.3.2.
7.2.4 Operating Experience on Cold Trapping of Fission Products at
Sodium-Cooled Reactors
7.2.4.1 Summary
Experience at each reactor for which data are available is reported
in Section 7.2.4.
Fission products which have been observed in cold traps are
listed in Tables 7.1 and 7.8.
Table 7.8
Fission Products Observed in Primary Cold Traps
of Sodium or NaK Cooled Reactors
EBR-II38 137CS, 134CS, I
BR-517 137CS> 136CS> 131^ 133^ 135,f
O
Dounreay 137Cs
SRE39 137Cs> 106Ru>
145
-------
7.2.4.2 EBR-II
Despite experimental work reviewed in this report and results from
other reactors that show the success of 137Cs removal by cold traps,
EBR-II personnel maintain that the primary system cold trap does not
reduce 137Cs satisfactorily at EBR-II.37 This result is shown by
observing the reported '37Cs activity levels in the primary sodium,
as shown in Table A25 of Appendix A. In 1971 the level built up
to 20 nCi/gm Na from failed fuel, and stayed there.
Experience with iodine (and tritium—see Section 5.1.3.2) at
EBR-II was different. The cold trap does remove ^1 so that the
levels are generally below 0.1 nCi/gm in the primary sodium (see
Table A25. Appendix A). Also EBR-II personnel can observe increases
in the '3'I levels in the cover gas when the primary cold trap is
cut off.37
A primary system cold trap was removed from operation from EBR-II
in 1965. Unfortunately the contents of this cold trap were never
analyzed; the trap still sits in a field near EBR-II.
Recently limited data have been reported concerning the EBR-II
primary cold trap.3° A gamma spectral scan of the trap during
shutdown period^in 1972 identified radiation from 2^a> 54wn9
65Zn, 124Sb, l£DSb, IMcs, and 137Cs. The ratio of the l6/Cs to Na
activities was ^18. The same ratio in the primary sodium was ^0.36.
Hence, the '37Cs concentration ratio in the cold trap for this
measurement was ~50. This is far below the value reported by SRE
(see Table 7.9 below).
The dose rate from the cold trap during the 1972 shutdown was
90% higher than the value during a shutdown one year before, in
1970-71. The measurements in the previous shutdown are reported in
Reference 15. The dose rate 2 in. from the surface was 290 mR/hr
at 132 days after the 1972 shutdown, compared to 153 mR/hr at 132 days
after the 1970-71 shutdown.
7.2.4.3 BR-5
The BR-5 cold trap was reported to trap 131IS 137Css and 136Cs.17
More than 90% of the I and Cs activity was trapped. The cold trap
also collected zirconium and barium, but much less efficiently than
I and Cs.
The Xe and the Xe activities in the cover gas were reduced
by factors of two and three, respectively, when the cold trap was
operating, due to trapping of the precursors 135I and '33I.
7.2.4.4 SRE
Extensive data are available from SRE cold trapping experience
because the cold traps were used to clean up the sodium system after
146
-------
Table 7.9
Comparison of Impurity Levels in SRE Cold Trap to those in Sodium
Coolant 22»39
Impuri ty
Carbon
137Cs
125Sb
Fe
Si
Mn
Pb
Cr
Ni
In Cold Trap
144-1550 p. p.m.
4.0xl02pCi/g
4.3 yCi/g
200- >500 p. p.m.
200- >500 p. p.m.
50- 500 p. p.m.
5 >500 p. p.m.
5- >500 p. p.m.
10- 300 p. p.m.
In Coolant
18-60 p. p.m.
1.5xlO"2yCi/g.
O.SxlO"2 yCi/g.
50 p. p.m.
50 p. p.m.
<5 p. p.m.
10 p. p.m.
5 p. p.m.
5 p. p.m.
Concentration Ratii
2-80
2.7xl04
7.2xl02
4- >10
4- >10
10- >100
0.5- >50
1- >100
2- 60
147
-------
22 39
extensive fuel cladding failure. ' There was a large amount of
carbon in the system, however, which leads to uncertainty in applying
the results directly to a cold trap system without carbon. Hansen39
provides arguments that oxide impurity in the sodium was responsible
for the greater retention of 137cs instead of the carbon impurity.
The most interesting results are the concentration ratios, which
are reported in Table 7.9. The large concentration ratio for
T37cs is particularly noted. The total 137r,s trapped was ^10 Ci.
In addition to those shown in Table 7.9 (only one of which is a fission
product), the fission products 106Ru, ^Ke-Pr, and ''°Ag were also
observed in the cold trap.
7.2.4.5 SEFOR and Fermi
23
Although reports on oxide removal by cold traps in SEFOR and
Fermi 22 are available (as discussed in Section 7.2.1), no results
were found on fission product removal by cold traps at these facilities.
REFERENCES (Section 7)
1. A. W. Castleman, Jr., and I. N. Tang, "Fission Product Vaporization
from Sodium Systems," Proceedings of the International Conference
on Sodium Technology and Large Fast Reacto"r~"Design, ANL-7520, I.
noogy
7-9,
540, November 7-9, 1968.
2. A. W. Castleman, Jr., "LMFBR Safety, I. Fission-Product Behavior
in Sodium," Nuclear Safety, II, 379 (September - October 1970).
3. D. E. Plumlee and P. E. Novak, "Measured Lb,ss of Selected Fission
Products from High-Burnup Mixed-Oxide Fue,l in a Miniature Pumped
Sodium Loop," GEAP-13731, August 1971.
4. J. Saroul, "Investigation on the Behavior of Fission Products in
Sodium and Argon--Pirana Experiments," Proceed i ngs of the ^
International Conference on the Safety of Fast Reactors," "A"ix-en~
Provence, France, Session Vb-1 , September 19-22, 1967.
5. J. Saroul, "Out of Pile Studies and Diffusion of the Contamination
in Liquid Sodium and Fast Reactor Containment, New Experimental
Program," Proceedings of the International Congress on the Diffusion
of Fission Products, Sac! ay, France. November 4-6. 196?.
6. R. deFremont, "TEMPE'TE Test: In Pile Experiment on the Transfer
of Fission Gases. Comparison to Out of Pile Tests," DRNR/STRS,
71.1147 (France), 1971.
7. J. J. Regimbal, W. P. Kunkel , and R. S. Gilbert, "Measurement of
Noble Gas Transport Dynamics in SEFOR Sodium," Trans. Am. Nucl .
Soc., 14, 773, (1971).
148
-------
8. J. L. Phillips, "Full Power Operation of the Dounreay Fast
Reactor," Proceedings of the National Topical Meeting on Fast
Reactor Technology, ANS-100, p. 23, April 26-28. 1965.
9. B. D. Pollock, M. Silberberg, and R. L. Koontz, "Vaporization of
Fission Products from Sodium," Proceedings of the International
Conference on Sodium Technology ancTLarge Fast Reactor Design,"
ogy
7-T,
ANL-7520, 1, 549, November 7-9, 1968.
10. A. W. Castleman, Jr., I. N. Tang, and MacKay, "Fission Product
Behavior in Sodium Systems," Proceedings of the Symposium on
Alkali Metal Coolants, IAEA, Vienna, 1966, p. ~~^
11. M. H. Cooper, B. R. Grundy, and G. R. Taylor, "Behavior of
Iodine in Sodium Systems," Trans. Am. Nucl . Soc. , 1_5, 232 (1972).
12. W. S. dough and S. W. Wade, "Caesium Behavior in Liquid Sodium —
The Effect of Carbon," Proceedings of the Eleventh AEC Air
Cleaning Conference. 1, 393, September, 1970.
13. W. S. Clough, "The Behavior of Barium and Strontium Fission
Products in Liquid Sodium," Proceedings of the International
Congresson the Diffusion of Fission Products, Saclay, France,
November 4 - 6, 1969.
14. G. B. Zwetzig, "Survey of Fission - and Corrosion - Product
Activity in Sodium - or NaK-Cooled Reactors," AI-AEC-MEMO-12790
(February 1969).
15. Reactor Development Program Progress Report, ANL-7845,
Section 1 (July 1971).
16. Reactor Development Program Progress Report, ANL-7776, 1 - 24
(January 1971).
17. V. V. Orlov, M. S. Pinkhasik, N. N. Aristarkhov, I. A. Efimov,
A. V. Karpov, M. P. Nikulin, "Some Problems of Safe Operation of
the BR-5 Plant," Proceedings of the International Conference
on the Safety of Fast Reactors, Aix-en-Provence, France, Session
Va-7, September 19 - 22, 1967.
18. R. deFremont, "Observations on the Behavior of Radioactive
Products on Rapsodie," DRNR/STRS, 71.1146, 1971.
19. R. deFremont, "Decontamination Experience on Rapsodie," DRNR/STRS,
71.1145, 1971.
20. R. S. Hart, "Distribution of Fission Product Contamination in the
SRE," NAA-SR-6890 (March 1962).
21. R. L. Eichelberger, "A Recommended Expression for the Solubility
of Oxygen in Liquid Sodium," Trans. Am. Nucl . Soc., 12, 613 (1969).
i - - -
149
-------
22. R. B. Hinze, "Cold Trap Performance Limitations (A State-of-the-Art
Review)," Chem. Eng. Prog., Symposium Series, 66, No. 104, 94 (1970),
23. A. D. Gadeken and M. C. Plummer, "SEFOR Cold-Trap Experience,"
GEAP-10548 (April 1972).
24. 0. Herb, "Examination of the Enrico Fermi Sodium Cold Trap,"
WCAP-4321 (November 1965) (Limited distribution report).
25. "Task II Report, Conceptual Plant Design, System Descriptions,
and Costs for a 1000 MWe Sodium-Cooled Fast Reactor," GEAP-
5678, p. 206 (December 1968).
26. G. B. Zwetzig, J. Guon, and M. Silberberg, "The Distribution of
Fission Products in LMFBR Sodium Systems/1 Proceedings of the
International Conference on Sodium Technology and Large Fast
Reactor Designs, ANL-7520, I. 527, November 7 - 9, 1968.
27. G. B. Zwetzig and R. F. Rose, "Interim Description of a Computer
Code (STP-1) for Estimating the Distribution of Fission and
Corrosion Product Radioactivity," AI-AEC-12847 (June 1969).
28. M. H. Cooper and G. R. Taylor, "Cesium Sorption from Liquid Sodium,
Trans. Am. Nucl. Soc., n_, 525 (1968).
29. M. H. Cooper and G. R. Taylor, "Cesium Cold Trapping in a Forced-
Convection Na System," Trans. Am. Nucl. Soc., 1_2^, 611 (1969).
30. J. Guon, "Studies of Cesium Deposition in a Sodium/Stainless-
Steel System," Trans. Am. Nucl. Soc., 1_2, 612 (1969).
31. J. Guon, "Effect of Surface/Liquid Partition on the Analysis of
Impurities in a Sodium System," Trans. Am. Nucl. Soc., 14,
625 (1971).
32. R. P. Colburn, "Nature of Cs and I Deposits in Sodium Systems,"
Trans. Am. Nucl. Soc.. 1_4, 626 (1971).
33. R. P. Colburn, "Fission Product Removal from Sodium by Hydride
Slagging," Trans. Am. Nucl. Soc., TJ5, 235 (1972).
34. J. C. Clifford, Advanced Plutonium Fuels Program, Second Annual
Report, LA-3993, September, 1968.
35. M. H. Cooper, B. R. Grundy, and G. R. Taylor, "Behavior of Iodine
in Sodium Systems," Trans. Am. Nucl. Soc.. 15, 232 (1972).
36. J. C. Clifford, "Behavior of Fission Products in Sodium,"
Proceedings of Symposium on Alkali Metal Coolants, I.A.E.A.,
Vienna, 759 (1967).
150
-------
37. EBR-II personnel, personal communication, December, 1972.
38. Reactor Development Program Progress Report, ANL-RDP-7, p. 110
(July 1971).
39. A. I. Hansen, "The Effects of Long-Term Operation on SRE Sodium
System Components," NAA-SR-11396 (August 1965).
8. GASEOUS RADWASTE MANAGEMENT
The proposed gaseous radwaste systems of FFTF and EBR-II together
with some review of present systems for EBR-II, Fermi, SEFOR, Rapsodie,
and Dounreay are discussed in this section. Also a comparison is
made with LWR gaseous radwaste systems.
The quantities of gaseous activity that will be released from
FFTF are expected to be trivial. This results from two factors:
(a) a sophisticated gaseous radwaste system will be used on FFTF, and
(b) there is virtually no liquid coolant leakage from an LMFBR from
which noble gases can escape to the environment. Although sophisticated
gaseous radwaste systems have not been used heretofore on fast reactors,
such systems will probably be used on future LMFBR power plants.
Nevertheless, results are presented here for 85|
-------
reactor, being designed and constructed at Hanford under the management
of the Westinghouse Hanford Company for the USAEC Division of
Reactor Development and Technology. The purpose of FFTF is to provide
experimental data in support of the LMFBR program in a number of
areas, including: fast neutron effects on fuels and materials; fast
reactor fuel performance; and system and component performances. In
keeping with this purpose, the design wil. allow reactor operation with
continuous noble gas release to the primary system from up to 1%
of the fuel pins. Also, four special sodium-cooled closed loops will
permit testing of vented or defective fuel. FFTF is designed to
release practically zero quantities of radionuclides to the environment.
This "near zero release" operation will bi achieved primarily by
means of high-integrity sealing of the primary sodium systems and
through the use of two gas processing systems, namely, the Radio-
active Argon Processing System (RAPS) and the Cell Atmosphere
Processing Systems (CAPS). Salient details of these features are
discussed below.
8.1.1. Primary Sodium System Seals
The FFTF reactor will operate with a maximum outlet temperature
of the coolant of approximately 1050°F. At this temperature, the vapor
pressure of sodium is only 0.018 atmospheres absolute. In order to
prevent inleakage of air into the reactor, it is necessary to
pressurize the reactor with an inert gas. At FFTF argon has been
chosen for this service; the reactor cover gas pressure is nominally
10 inches WG or approximately 1.025 atmospheres absolute. The closed
loops require an argon cover gas pressure of about 55 psig in order to
prevent sodium pump cavitation. Since the argon cover gas lies on
top of the sodium in the reactor and closed loops, certain gaseous
fission products, primarily the Kr and Xe isotopes, which escape
from defective or vented fuel pins can disengage from the sodium and
collect in the argon cover gas. Consequently, it is important;to
reduce the leakage of this potentially contaminated cover gas into the
reactor building.
To accomplish this, gas buffered seals are used in the reactor
head and in the closed loops. Each buffered seal consists of two seals
in series, with positive argon buffer gas pressure (e.g., 2 psig in the
reactor head seal) maintained in the annular space between the two
seals. Since all seals leak to some extent, there is some argon
buffer gas continuously leaking into the reactor and into the reactor
building from the inter-seal spaces. Therefore, it is necessary to vent
argon cover gas from the primary system, at a rate which depends on the
amount of leakage, in order to maintain the cover gas pressure in the
proper range. The flow rate of argon from the reactor cover gas region
is expected to be about 4 standard cubic feet per minute ( scfm). The
closed-loop coyer gas will contribute an additional 0.02 scfm. This
flow of contaminated argon cover gas goes to the Radioactive Argon
Processing System (RAPS), where its activity is substantially reduced.
The relatively "clean" argon leaving the RAPS is recycled for use as
coven- gas or for pressurization of the buffered seals.
152
-------
The buffer gas which leaks into the rpactor building from the
interseal spaces should present no significant radiation hazard,
as it has practically the same specific activity as the effluent from
RAPS. Although the references''^ report no values for the volumetric
leak-rates into the reactor building, it is reasonable to assume a
value equal to the total leak-rate from the inter-seal spaces into
the reactor and the closed loops, i.e., about 4 scfm. The specific
activity of the buffer gas is estimated in Reference 1 to be 10~5 Ci/ml,
most of which is ^Kr. Assuming this activity and the leak rate of 4
scfm, the rate at which activity leaks into the reactor building and
thence to the environment via the reactor building ventilation system
is:
4 ft3 v 28317 ml v 10"5 yCi v 1440 min „ , ',, ,n-3 Ci
x x x - '•DJ x IU
min ft ml day
Thus, the annual discharge of activity to the environment stemming
from leakage of buffer gas into the reactor building is only about 0.5 Ci
8.1.2 Radioactive Argon Processing System (RAPS)
This system (RAPS) is designed to receive the contaminated argon
cover gas from the reactor and the four closed loops and to process it
on a continuous basis. The system is designed to process an inflow
of about 700,000 Ci of noble gases per day,2 yielding a purified
argon effluent having a specific noble gas activity of 10~°yCi/ml or
less, which corresponds to a maximum allowable specific activity of
1 MPC* for Kr-85.
The basic flow-sheet of RAPS is shown in Figure 8.1 Contaminated
argon, vented by pressure controllers from the various cover gas
regions, is piped to a surge tank, from which it is metered into a
processing loop consisting of four cryogenic charcoal delay beds,
four heat exchangers for removal of decay heat, a fractionation column,,
a gas circulator, a surge gas storage tan<, and various control
elements.
The delay beds are quite effective in holding up xenon (and
iodine, if any exists in the cover gas), less so for krypton. Table
8.1 summarizes the delay times under design conditions.^ For these
delay times, virtually all radioactive xenon and most of the short-lived
krypton is eliminated in the decay beds.
Argon leaving the last decay bed is recooled to -280°F before
being injected into the fractionation column. The stable xenon
isotopes and the krypton isotopes are concentrated in a pool of liquid
argon in the bottom of the fractionation column by the refluxing action
* From 10CFR20, Appendix B, Table 1 (maximum permissible average concen-
tration in restricted areas to persons of age 18 or more).
153
-------
-125»F
Figure 8.1
154
-------
Contaminated Gas
25 PSIG
Surge Tank
Compressor
-175°F
X Charcoal
Delay 1*12
Beds Cooling
-125°F
Purge Gas
Storage Tank
Gas Return
To Reactor
IN2 Cooling
VXAAr—Electrical Heater
Fractionation
Column
Figure 8.1 Rfdioactive Argon Processing System
154
-------
Page Intentionally Blank
-------
Table 8.1
Radioactive Argon Processing System Delay Times
Xenon Delay, Days
Krypton Delay, Days
Delay Bed
No. 1
9
0.27
Delay Bed
No. 2
45
0.78
Delay Bed
No. 3
42
0.76
Delay Bed
No. 4
40
0.73
-------
of the column. The fractionation column is expected to remove 99.9%
of the xenon and krypton isotopes from the gas stream. The purified
argon gaseous effluent from the column is expected to have a specific
noble gas activity of 10~5 yCi/ml or less. The purified argon is
recycled back to the buffered seals.
When it becomes desirable to remove the accumulation of noble
gas nuclides in the bottom of the column, the column is isolated and
its contents are gasified and transferred to an ambient-temperature
tank for long term storage. Under design operating conditions (1%
defective fuel), the annual accumulation of Kr-85 in the fractionation
column will be about 300 Ci. Other noble gas nuclides will be
present in only trace quantities.
If it is- assumed that leakage from RAPS is negligible, there is
virtually no release of radioactivity from this system to the
environment.
8.1.3 Cell Atmosphere Processing System (CAPS)
The primary sodium equipment cells are provided with virtually
inert atmospheres of nitrogen with approximately one percent oxygen.
The cells are sealed and the atmospheres are maintained by feed-and-
bleed pressure controls. Effluents from these cells are processed by
the Cell Atmosphere Processing System (CAPS) before release to the
environment.
The basic flow-sheet for CAPS is shown in Figure 8.2. Gas vented
from the inert atmosphere cells is pumped into a surge tank, from which
it is metered into the processing equipment, consisting of a disiccant
unit, two cryogenic charcoal delay beds9 two liquid nitrogen cooled
heat exchangers for removal of decay Seat, a gas circulator,, a'nd^vartous
control elements. The effluent from CAPS is mixed with air passing
through the FFTF heating and ventilation system and exhausted to the
environment.
Although the final design of CAPS has not yet been made, an
estimate of the delay times associated with the charcoal delay beds is
53 days for xenon and 2 days for krypton at a flow rate of 25 scfm
and a temperature of -100°F.3 CAPS should be able to process between
0 and 50 scfm of contaminated inert gas, depending on the demand.
The normal release of activity from CAPS is virtually zero9 since
there should be no release of activity from the primary system under
normal conditions.
8.2 EBR-II Gaseous Radwaste Systems
8.2.1 Present Operation
EBR-II is used to test fuel for the LMFBR development program.
The driver fuel is metallic U-235. Test pins are made of potential
156
-------
Potentially Contamnated
Gases from Inert Cells
Compressor
Kbrmal
Exhaust
to H & V
System
I—I
Surge Tank
Gas
Dryer
Molecular
Sieve
Emergency Circulation
Gas
0 ' Circulator
-100°P
IN2 Cooling
Charcoal
Delay
Beds
IN2 Cooling
Figure 8.2 Cell Atmosphere Processing System
157
-------
Page Intentionally Blank
-------
Potant-ially
fcxsn Inert Celie
Ccntpsri
Bfchsust
tsoH & V
fiysrtatn
Figwr* 8.2 Cell
157
-------
LMFBR fuels such as oxides, carbides, and nitrides, with oxide test
pins predominating.
At present EBR-II cannot operate with failed test pins. When
oxide pins fail, fission product gas is rapidly released. Leakage from
the cover gas to the reactor building is sufficiently high that
EBR-II must be shut down when a test pin fails and remain shut down .,
until the failed pin is located and removed.
EBR-II can operate with failed driven fuel, however. Failed
metallic fuel releases fission product gas at such a slow rate that
the present cover gas cleanup system can reduce the activity from
failed driver fuel adequately.
The present gas radwaste system is designed: (a) to operate
during normal reactor operation and (b) to purge the cover gas when
a failure occurs in a test pin.
8.2.1.1 Normal Operation
During normal operation the escape rate of the cover gas from the
reactor tank is -v/lOOO ml/min. Of this M30 ml/min passes through
the various monitoring systems and then is discharged to the atmosphere
through the stack. The remainder (i.e. -\.900 ml/min) leaks to the
reactor building. The air in the reactor building is continually being
purged, with the exhaust being discharged to the atmosphere through
the stack. Hence all 1000 ml/min of cover gas eventually is discharged
directly to the atmosphere via the stack.
8.2.1.2 Fast Gas Purge System
In the event of a test fuel pin failure the reactor is shut down,
and the Fast Gas Purge System is put into operation. This system
removes the cover gas and eventually sends it to the atmosphere
through the stack.
The flow rate to the system can be varied up to 3 standard cubic
feet per minute (scfm). The purged argon is replaced with fresh argon
while monitoring the cover gas at slightly above atmospheric pressure.
The activity in the cover gas can be returned to a tolerable level
in 3 to 4 hours.
In the Fast Gas Purge System, the first step is to remove sodium
vapor in a vapor trap. An aerosol trap filters out particles of size
greater than 5u . This is followed by a gas sampling and monitoring
station. Finally there is a variable speed pump and a flowmeter. The
gas is then sent out of the containment to the suspect exhaust stack
and to the atmosphere.
8.2.2 Proposed Gas Radwaste System
4
A system has been proposed for use at EBR-II which would allow
158
-------
operation with failed test fuel. The proposed system is described
here because of its educational value. It is an example of a system
that has been extensively analyzed and one for which the analysis is
available. If it is implemented, it will serve as a useful demonstra-
tion that operation with failed oxide fuel is feasible, or at least it
will identify problems involved with such operation.
8.2.2.1 Criteria
The first step toward designing a system for operation with failed
test fuel was to determine the required design criteria.5 Ultimately
this meant specifying the flow rate to the proposed cover gas cleanup
system and the activity of the gas to be processed by the system.
The design criteria were:
• Operation with 12 defective oxide fuel pins at a linear power
density of 16 kw/ft.
• Detection of a new test pin failure by a step release of
133YQ
Xe.
• Activity in the reactor building below the maximum permissible
concentration as defined by 10 CFR 20.
• Gas release to the environment from the stack to be below
the maximum permissible concentration at ground level as
defined by 10 CFR 20.
The number of defective fuel pins and linear power density were
based on proposed fuel failure test requirements by the General Electric
Co. Calculations were made of fission gas release rates from
defective oxide pins to determine the rate at which activity of each
isotope would be added to the cover gas.5 The Booth diffusion model
was used for these calculations.
Detection of failed oxide pins by xenon tagging has been successfully
demonstrated in EBR-II, and test fuel pins are now being tagged. In
order for the xenon tagging method to work, the level of xenon
isotopes in the cover gas must be kept low. The fact that a new
failure has occurred is indicated by a rise in '33Xe activity. It
was determined that a 25% rise in '^3Xe activity due to a new fuel
pin failure was sufficient for detection. A pin failure is expected
to increase the cover gas activity by 0.25uCi/ml. Therefore, the second
design criteria meant that the cover gas '33Xe activity from 12 failed
fuel pins must be held to <1.0 yCi/ml.
133
the required cover gas purge Fate was
In order to meet the 1 yCi/ml Xe activity from 12 failed pins,
;eterminea to be TO scfm.
159
-------
For the resulting activity levels in the cover gas, the present
leakage rate of ^1000 ml/min from the cover gas to the reactor
building is too great. In order to reach 10 CFR 20 MFC levels, the
leakage rate must be reduced by a factor of 100, to 10 ml/min. It
is anticipated that this can be done by replacement of seals known
to be principal sources of the current high leakage.
8.2.2.2 Cover Gas Cleanup System
The 10 scfm of cover gas purged must be treated in order to
remove radioactive krypton and xenon isotopes from it. The method
selected for the proposed EBR-II system is the use of charcoal adsorption
beds. This method was selected over other possibilities (e.g.
cryogenic distillation, permselective membranes, and selective absorption
in liquid fluorocarbons) on the basis of relative costs, relative
effectiveness, complexity, possible material problems, and space
requirements.
Before passing through the charcoal delay:beds, the gas passes
through an aerosol removal system. This system will remove sodium
liquid entrainment and fine particles. Aerosol traps will be
followed by high efficiency filters, but the specific design for neither
has yet been selected. A gas flowmeter will also be in the aerosol
removal system. There will be two redundant modules, each containing
a trap, filter, and flowmeter, and each module will have sufficient
capacity to perform the entire aerosol removal function independently.
Two conceptual charcoal delay-bed systems have been designed from
which a final selection will be made. The simpler design is a 7-day-
delay system.
8.2.2.3 Seven-Day-Delay System
The 7-day-delay system is shown in Figure 8.3. Flow capacity
through the system is 10 scfm, although lower flow rates (and longer
delay times) would be used if fewer than 12 fuel pins had failed. For
seven days flow is directed through one delay bed, e.g. Tl. The
argon cover gas is cooled in cooler HX1 to -136°F. At this low
temperature, the xenon and krypton in the cover gas are adsorbed on the
charcoal. Decay heat from Xe and Kr in the delay bed causes the
temperature to rise to -64°F as the argon passes through the bed.
The argon is then filtered (F3), reheated to 80°F, and returned to the
reactor. The delay bed provides a seven-day holdup of xenon and a
seven-hour holdup of krypton.
After seven days of service, the cover gas flow is switched to the
second delay bed, T2, and the first bed, Tl, is regenerated, i.e.
the xenon and krypton isotopes are removed. Regeneration is accomplished
by heating the bed to 400°F (at which temperature the adsorbed xenon
and krypton are released from the charcoal) and backflushing the bed
with a small flow of cover gas diverted from the outlet of the
operating delay bed, T2. This hot gas from Tl, which now contains the
160
-------
CTl
From 700°F
Reactor
To
Reactor
FM2 V4
J
Aerosol Removal
Module
Figure 8.3 Seven-Day-Delay Cover Gas
Cleanup System for EBR-II
C3
V20
-------
Page Intentionally Blank
-------
Systan for EBR-II
C3
V20
-------
xenon and krypton from Tl, is cooled in HX1 and is compressed into
bottles. The volume of gas bottled each week is 162 standard cubic
feet. The bottles are shipped off site for further processing or
storage. After regeneration, the gas in the bed is recirculated
through the blower and cooled until the bed returns to operating
temperature (-100°F).
Another small bleed stream is sent to a Xe-tag cold trap, T3. The
xenon is held up in this trap for about one hour. In the event of a
fuel element failure, the trap will collect and retain the xenon tag
sample for later analysis.
This delay bed system removes nearly all of the xenon activity
from the cover gas stream before it is returned to the reactor. Since
the krypton is held a shorter time on the beds, some of it returns to
the reactor. The fraction returned for each krypton isotope is
listed in Table 8.2. Also shown in Table 8.2 are the calculated
activities and decay heat rates in the delay bed in service for both
xenon and krypton for operation with 12 failed fuel rods.
8.2.2.4 24-Hour Delay System
The 24-hour-delay-system is shown in Figure 8.4. Three delay
beds are used on a 24 hour cycle each.
The most fundamental change in this system compared to the 7-day
system is the addition of the secondary delay beds T4 and T6. After
regeneration of one of the 24-hour-delay beds, the argon is cooled
and sent to the secondary delay beds T6 and T4. The delay bed T4
operates at room temperature and provides a 50-day holdup of xenon
isotopes. The outlet gas from T4 is cooled to -60°F and flows through
the krypton retention bed, T6. The regeneration flow rate is
approximately 9 scfh and approximately 6 hours is required, so that
approximately 54 standard cubic feet of gas is used for regeneration.
The outlet from T6 is either sent to the stack or recirculated back
through the primary delay beds and to the reactor. The krypton holdup
time in T6 is 7 days. Hence all of the krypton isotopes will decay in
T6 except 85Kr.
Once each week the krypton retention bed, T6, is regenerated, and
the effluent is compressed into bottles, for storage or shipment off-
site. During regeneration, T6 is heated to 300°F. Regeneration of
this bed is accomplished during regeneration of a primary delay bed;
hence the same flow rate (9 scfh) is used. Regeneration requires only
2.1 hr., hence only 18.9 standard cubic feet each week must be bottled.
8.3 Gaseous Radwaste Experience in Other Operating Fast Reactors
8.3.1 Fermi
Tables of cover gas data for Fermi are given in Appendix A of this
i
162
-------
01
CO
TABLE 8.2
Xenon and Krypton Conditions in Delay Beds
Isotope Half Life
83raKr 1.86 hr
D Oirtpf-y- A A Hf~
JVL *± • *i ILL-
85 Kr 10.76 hr
87 Kr 76 min
88 Kr 2.8 hr
89 Kr 3.2 min
90 Kr 33 sec
Fraction Returned Activity
To Reactor On Bed
(Ci)
0.209 19.8
0.517 32.9
1.00 0.0032
0.102 83.5
0.355 55.5
^L0~6 10.2
S >7ri
/Vt- J • ^ / d
135mXe 15.6 min
135 Xe 9.2 hr
137 Xe 3.8 min
138 Xe 14 min
139 Xe 41 sec
llf0 Xe 13.7 sec
,C 0.87
>0 15.1
^0 384
.o-O 17 . 3
^0 92 .-4
O 92.0
M) 41.1
^0 0.45
M) 0.05
0.00
0.07
1.53
0.18
1.07
2.75
1.60
0.02
0.00
Total decay heat =
7.22 tc
15.70 Btu/hr
-------
From 700°F
Reactor
10 SOPM, 80°F
Aerosol Removal
Module
AAAAAAAA
Heater
Delay Bed Module
V18 FM4
Figure 8.4 24-Hour-Delay Cov-
er Gas Cleanup
System for EBR-II
VI9 FM5
Suspect Exhaust
1b
Bottle
C4
-CZ1
FM7
-------
report. Initially there was a problem of achieving tight sealing
of the cover gas system,^ but there was no real problem in keeping
the cover gas clean as long as the primary sodium temperature was
kept below 600°F. The waste-gas system was quite oversized for the
associated systems. In fact, Bruzzi et a!.7 calculated that the
Fermi waste-gas system could adequately handle the large activities
which would result if the original fuel were replaced by vented-to-
coolant fuel, i.e. it could handle perhaps 100 times more activity than
expected in normal operation.
8.3.2 SEFOR
Performance of the gaseous radwaste system is only partially
indicated in the tables of cover gas activity in SEFOR in Appendix A
of this report. Tho main reasons that t!;e cover gas showed so little
activity were that the pins were not pusled to excessive performance g
limits, i.e. little leakage, and that tr
-------
8.4.1 LWR Gaseous Releases
Two recent studies provide excellent summaries of radioactive
gas emission from LWR's, the first an ORNL study'^ and the second
a comprehensive USAEC Regulatory study.^ yne QR^L and the
Regulatory studies both assumed 0.25% failed fuel for the calculated
fission-gas releases. Although this assumption leads to generally
higher estimates for activity releases to the environment than is
warranted by actual LWR operating experience, these results are
used here for the purpose of comparison with LMFBR results (for
which 1% failed fuel has been assumed).
8.4.1.1 ORNL Study
A comprehensive survey of LWR gaseous waste systems was presented
by ORNL staff members at the 12th AEC Air Cleaning Conference.'2
This survey involved a detailed study of roughly 100 LWR plants
based upon information contained in docketed documents such as the
Preliminary Safety Analysis, the Final Safety Analysis, the
Applicant's Environmental Report, and the Amendments thereto, as
well as information obtained by direct questioning of the applicants,
reactor vendors, and architect engineers.
As a result of this study, it was determined that those
radionuclides which are normally available for escape in gaseous
form include the noble fission product gases (Kr and Xe), the
fission product halogens (Br and I), certain activation products
such as '6^5 13^5 19Q5 ancj 41/\rj ancj tritium, which may originate
either from ternary fission or activation. Experience has shown
that the noble gases and the iodines contribute virtually all
of the radiologically significant gaseous activity released from
LWR's of current design.
The sources of emission can be divided into two major categories:
(1) inadvertent leaks from tanks, piping, valves, etc. which allow
gaseous activity to escape without being processed by the radgas
system, and (2) operational releases in which fluid is deliberately
withdrawn from the cooling system of the reactor. The latter
category would include steam generator blowdown, exhaust from the
condenser air ejectors and releases from various system degassing
operations.
The ORNL study presents tables of "typical" gaseous releases
from PWR's and BWR's, identifying sources as well as isotopes.
Table 8.3 is a summary of releases from a typical 1000 MWe PWR,
based on numbers taken from the ORNL study. It is seen that a 1000
MWe PWR with 0.25% defective fuel will "typically" release roughly
2,500 Ci annually of noble gas radionuclides, most of which consists
of 133xe with a 5.77-day half-life. More important, because of its
10.7-year half-life, is ^^Kr, which has an estimated annual release
of about 800 Ci.
166
-------
Table 8.3
Typical Annual Gaseous Release from a 1000 MWe PWR Operating
With 0.25? Defective Fuel (based on Reference 12)
Release Rate, Ci/yr
Coolant
cr>
Kr-83m
85m
-85
-87
-88
-89
Xe-131m
-133m
-133
-135m
-135
-137
-138
I -131
-133
Concentration
y Ci/ml
3.865E-02
2.076E-01
1.219E-01
1.125E-01
3.604E-01
8.546E-03
1.518E-01
3.724E-01
2.775E 01
2.393E-02
6.003E-01
1.756E-02
8.317E-02
6.166E-01
6.845E-01
Auxi 1 iary
Bui Iding
1.068E 00
5.737E 00
3.368E 00
3.110E 00
9.960E 00
?.362E.-m
4.196E 00
1.029E 01
7.669E 02
5.614E-01
1.659E 01
4.852E-01
2.298E 00
8.519E-02
9.458E-02
Containment
Purge
2.643E-03
3.359E-02
6.751E 00
5.240E-03
3.708E-02
1 .665F-05
1.573E 00
7.428E-01
1.292E 02
2.302E-04
2.033E-01
4.109E-05
8.657E-04
4.375E-01
5.283E-02
Primary
Degasifi cation
G.O
0.0
9.252E 01
0.0
0.0
0.0
7.729E 00
2.625E-04
5.277E 01
0.0
0.0
0.0
0.0
0.0
0.0
Shim Bleed
Degasif ication
0.0
0.0
7.194E 02
0.0
0.0
0.0
2.239E 01
1.623E-04
7.481E 01
0.0
0.0
0.0
0.0
0.0
0.0
Steam
Generator
1.079E 00
5.796E 00
3.403E 00
3.142E 00
1..006E 01
2.386E-01
4.239.E 00
T.040E 01
7.747E 02
6.682E-01
1.676E 01
4.902E-01
2.322E 00
9.907E-01
7.078E-01
Total
2 ,150E 00
1 .157E 01
8.254E 02
6.257E 00
2.006E 01
4.748E-01
4.013E 01
Z.143E 01
1.798E 03
1.330E 00
3.355E 01
9.754E-01
4.621E 00
1 .513E 00
8.552E-01
-------
It should be stressed that most of the activity releases in
the above "typical" PWR stem from leaks of reactor coolant in the
Reactor Building, Auxiliary Building, and steam generators. Gaseous
activity from these sources are, for the most part, vented directly
to the environment. Over 90% of the 133Xe, and virtually all of the
other short half-life isotopes, escape to the environment in this
fashion. Consequently, the effect of adding a cryogenic "cleanup"
system to the tail-end of the radgas system would be to reduce the
8%r emission rate substantially (by about 80-90%), but to diminish
the emission of the short half-life isotopes only marginally (since
these isotopes come principally from points outside the radgas system).
The net effect of such a cryogenic system would be to reduce off-site
radiation exposure rates by marginal amounts. In view of these
considerations, it does not seem practical for PWR's to incorporate
cryogenic units into e>isting radgas systems until sources stemming
from coolant leaks can be reduced to insignificant levels.
8.4.1.2 USAEC Regulatory Study
Another recent and very comprehensive study of the radioactive
liquid and gaseous releases from LWR's was performed by the USAEC
Directorate of Regulatory Standards.13 in this report, a number of
alternative gaseous radwaste systems wore considered and evaluated
for both PWR's and BWR's. Results for six PWR radgas systems are
presented in Table 8.4 based on 0.25% cefective fuel). Annual
releases were estimated for each system. The total annual releases
for the six systems are summarized in Table 8.5. (A more detailed
presentation of the releases from each system, indicating sources for
for each radionuclide, is found in Reference 13.)
For all radgas systems represented in these two tables, the
annual emission of noble gas radionuclides ranged from 1300 Ci to
170,000 Ci, with ^33Xe (5.77 day) accounting for the larger part
of the released activity. The more important (radiologically)
85Kr annual releases ranged from 5 Ci to 800 Ci. The upper limit
represents releases from PWR radgas systems of current design. (Note
the excellent agreement between this value and the corresponding
value reported in the ORNL study).
Results for gaseous releases from a BWR were similar to those for
a PWR (except for the total noble gas -elease with little radwaste
equipment), and therefore are not presented here in detail. For
example, based on the Regulatory Report,'3 the annual emission of
noble gases from a 1000 MWe BWR for 0.25% defective fuel would
range from 2300 Ci to 2 x 10^ Ci, and the 85Kr annual releases would
range from 1 Ci to 600 Ci.
8.4.2 Comparison of L'/IR and LMFBR Radioactive Gas Releases
In Section 8.1.1 it was estimated that the FFTF would discharge
about 0.5 Ci/yr of °^Kr to the environment. Normalizing this value
to a 1000 MWe (2500 MWt) unit, the annual release of 85Kr from a
168
-------
UD
TABLE 6.4
Summary of Variables for PWR Gaseous Radwaste Treatment Systems
Xe
I
Kr
Primary system gases
Secondary system
qases
Reactor containment
purge
Auxiliary building
ventilation
Turbine building
ventilation
PWR Gas Case No .
1 2
Degree of Removal
low nigh
low medium
low low
kqmpment Units, Functions, and
None 60-day decay storage
tanks, HEPA filter
None None
i*jne Charcoal kidney
adsorber for
iodine
None None
None None
3
high
medium
low
Flow Paths*
60-day decay on
charcoal bed,
HEPA filter-
Charcoal adsorber
for iodine, HEPA
filter, clean steam
for gland seal,
blowdown tank
vented to con-
denser
Charcoal kidney
adsorber for
iodine
None
None
A
high
low
60-day decay on
charcoa ' bed ,
HEPA filter
Charcoal adsorber
for iodine, HEPA
filter, clean steam
for gland seal,
blowiown tank
vented to con-
denser
Charcoal kidney
adsorber for
iodine , charcoal
adsorber for
iodine, HEPA filter
Charcoal adsorber
for iodine, HEPA
filter
Charcoal adsorber
for iodine, HEPA
filter
5
high
high
high
. Reoontoiner, 60-day
decay storage tanks ,
selective a.i.s--jrt:ican
Charcoal adsorber
for iodine, HEPA
filter, clear, steam
for gland seal,
blowdown tank
vented to con-
denser
Charcoal kidney
adsorber for
iodine, charcoal
adsorber for
iodine, HEPA filter
Charcoal adsorber
for iodine, HEPA
filter
Charooal adsorber
for iodine, HEPA
filter
6
high
high
high
Cover gas recycle
Charcoal adsorber
for iodine, HEPA
filter, clean steam
for gland seal,
blowdown tank
vented to con-
denser
Charcoal kidney
adsorber for
iodine, charcoal
^^orber for
iodine, HEPA filter
Charcoal adsorber
for iodine, HEPA
filter
Charcoal adsorber
for iodine, HEPA j
filter
gases to 56-meter roof vent unless stack is indicated.
-------
o
Table 8.5
Estimated Annual Releases (Ci/yr of Radioactive Gaseous Effluents from 1000 MWe PWR
with 0.25% Defective Fuel (Based on Reference 13)
Nuclide
Kr-83m
85m
85
87
88
89
Xe-131m
133m
133
135m
135
137
138
Total Noble
Gas
1-131
133
System
1
210
1,100
800
620
2,000
31
920
2,100
160,000
120
3,400
70
420
170,000
1.8
0.7
System
2
3
19
800
10
33
1
35
36
2,900
2
55
2
7
3,900
1.2
0.6
System
3
3
19
800
10
33
1
35
36
2,900
2
55
2
7
3,900
0.3
0.2
System
4
3
19
800
10
33
1
35
36
2,900
2
55
2
7
3,900
0.04
0.03
System
5
3
19
26
10
33
1
18
36
2,900
2
55
2
7
3,000
0.04
0.03
System
6
3
17
5
10
33
1
5
20
1,200
2
48
2
7
1,300
0.04
0.03
-------
LMFBR operating with 1% defective fuel and an FFTF-type
radwaste system would be only about 3 Ci. Unlike the LWR, there
should be almost no release of short-lived fission gas to the
environment from an LMFBR. Confirmation of these low release
rates, of course, must await actual operating experience since
these are only estimates at this time.
39
The A production rate for a 1000 MWe LMFBR with 30 ppm potassium
impurity in the coolant was estimated in Section 5.3.3.1 to be ^30
Ci/year. All of this would eventually leak to the environment
regardless of whether argon or helium is used as the cover gas
(unless argon is deliberately separated from helium in a purification
system for a helium cover gas). This radioactive source is not
present in a LWR.
These values compare favorably with gaseous activity releases
from light water reactors. As indicated in Section 8.4.1 for an
assumed 0.25% defective fuel, even the most sophisticated PWR
radgas system'^ would release 1300 Ci of noble gases annually,
including about 5 Ci of 85«rj whereas "typical" PWR annual
radioactive gaseous releases amount to roughly 2500 Ci of noble gases,
including about 800 Ci of 85 Kr. jne ms^ sophisticated BWR radwaste
system would release only 1 Ci of &$}\r annually.'^
REFERENCES (Section 8)
1. Fast Flux Test Facility Design Safety Assessment HEDL-TME 72-92,
July 1972, HEDL Section 2.3.
2. C. J. Foley, "Fission Gas Control at FFTF," Proceedings of the
12th AEC Air Cleaning Conference, Oak Ridge, August, 1972.
3. FFTF Environmental Statement, WASH-1510, USAEC, May 1972,
pp. IV-28 to IV-32.
4. A. Panek, C. McPheeters, P. Nelson, "Cover Gas Cleanup System
for EBR-II Conceptual Design," ANL report, October 27, 1972.
5. ANL Memo, J. F. Koenig to R. C. Matlock, "Considerations for
Cover Gas Cleanup System," Sept. 13, 1972.
6. E. L. Alexanderson, C. E. Branyon, and W. R. Olson, Proceedings
of the ANS National Topic Meeting on Fast Reactor Technology,
ANS-100, p. 41 (196:^17~~
7. L. Bruzzi, G. Gondoni, P. S. Lindsey, and R. E. Mueller, "Plant
Design Problems Associated with the Application of Vented Fuel
to a Fast Breeder Reactor," Proceedings of the International
Conference on Sodium Technology on Larae Ifast Reactor Design,"
ANL-7520 II, p. 154 (1968). x7"^
171
-------
8. J. J. Regimbal, R. S. Gilbert, W. P. Kunkel, R. A. Meyer, and
C. E. Russell, "Fuel Failure Detection Capability at SEFOR,"
Trans. Am. Nucl. Soc.. 14, 69 (1971).
9. G. Gajac, J. L. Ratier, N. Reboul, L. Reynes, and M. A. Valantin,
"Rapsodie's First Year of Operation," Proceedings of the
International Conference on Sodium Technology and Fast Reactor
Design," ANL-7520 II, p. 52 (1968). '
10. G. Gajac, "Experience de Fonctionnement de Rapsodie en ce qui
Concerne la Fiabil'te des Composants," Proceedings of the
Internation Conference on the Engineering of Fast Reactors
for Safe and Reliable Operation, Karlsruhe, October 9-13, 1972.
11. J. L. Phillips, "Full Power Operation of the Dounreay Fast
Reactor, Proceedings of the ANS National Topical Meeting on
Fast Reactor Technology, ANS-100, p. 1 (1965).
12. Binford, F. T.; Hamrick, T. P.; Parker, G. W.; Row, T. H.,
"Analysis of Power Reactor Gaseous Waste Systems," Proceedings
of the 12th AEC Air Cleaning Conference, Oak Ridge, Tenn.,
August, 1972.
13. Directorate of Regulatory Standards, USAEC, "Draft Environmental
Statement Concerning Proposed Rule Making Action: Numerical
Guides for Design Objectives and Limiting Conditions for Operation
to Meet the Criterion 'As Low as Practicable1 for Radioactive
Material in Light-Water-Cooled Nuclear Power Reactor Effluents,"
January, 1973.
9. LIQUID AND SOLID RADWASTE MANAGEMENT AT EBR-II
The only information collected on Mquid and solid radwaste from
an operating fast reactor was for EBR-II J»2 This is a test reactor
with elaborate hot cell facilities that would not normally be present
at an LMFBR power plant. For example, irradiated fuel test pins
are routinely dismantled in the hot cells, and irradiated cladding
and other materials must be stored.
In Section 9.2, it is shown that the high level solid waste
stored at the EBR-II -'5 of the order of 106 Ci/yr, while the
intermediate solid waste is ^3000 Ci/yr. For comparison, for a 1000
MWe light water reactor the solid waste activity is 5000 to 10,000
Ci/yr.2
9.1 Liquid Radwaste System
Suspect liquid waste from EBR-II is liquid waste that contains
radioactive material, generally in water solutions. Approximately
100,000 gallons per year of suspect waste is processed. No estimate
of the activity in this waste is available. After processing, this
172
-------
waste is added to non-radioactive industrial waste which is sent
to a leach pit, where it either evaporates or settles into the lava
below. Typical sources of this suspect waste include decontamination
of equipment, solutions from chemical laboratories, and emergency
showers.
In the early days of EBR-II the su >pect waste was pumped directly
to the leach pit. Later this method was changed to an evaporation
process, carried out several times each week. This process is shown
schematically in Figure 9.1(A). The present system provides a
decontamination factor of 10? to 10)3.
4
A decontamination factor of 10 is now desired at EBR-II.
Therefore, the liquid waste system will be modified to add a settling
tank for solids before the liquid enters the evaporators and to add
additional equipment after evaporation. These additions are shown
schematically in Figure 9.1(B).
The solid sludge from the evaporators is stored in 55 gallon
drums, which are encased in concrete for shielding. This sludge is
eventually processed as solid radwaste.
9.2 Solid Radwaste Management ,
Solid waste from EBR-II is classified as low, intermediate, or
high level waste. The concern here is with solid waste other
than the irradiated fuel itself, which is shipped offsite for
reprocessing. About 90% of the solid waste volume is low level;
nearly all of the activity is high level. In Table 9.1 are listed the
activity concentration ranges for the different levels, together
with typical annual volumes and the 1971 activity totals for each
level. The activity levels are values after 15 days of storage.
The intermediate wastes are placed into 1 cu ft shielded
containers and sent, together with the low level waste, to the
National Reactor Testing Station (NRTS) Burial Ground. Typical
sources of low level wastes are dry decontamination and filters.
Intermediate wastes come mostly from analytical cells and chemical
facilities.
High level wastes are stored in a 7 acre storage facility at
EBR-II. About half of this capacity had been used through 1972. The
wastes are placed in 1 ft. or 2 ft. high containers, these are then
placed in a 6 ft. high can, and the can is inserted in a 12 ft.
deep, 16 inch diameter hole in the ground, or on grids at 6 ft.
intervals. Finally the 6 ft. above the can is filled with gravel
shielding.
Any solid waste that contains plutonium or other transuranium
elements is wrapped several times in plastic bags, placed in 55
gallon drums and stored outdoors on an asphalt pad at an interim sto-
rage facility at NRTS. On the order of 3 mg of plutonium was stored
in 1971.
173
-------
Liquid
Waste
Liquid Industrial
Liquid
Condenser
f Vapor
Evaporator
Sludge
1
Filter
| Liquic
Condenser
f Vapor
Evaporator
-~ to Leacn PC
I
Solids
" Stored in i>5
Hastes
inds
Gal lor
(A) Present System
Liquid
Industrial
Liquid
Waste H"~
Settling
Tank
k • >
Liquid
Conds
T -qu'd
3iser
Vapor
Sludge
Filter
Liquid
Condenser
1 Vapor
$ f Wastes
Conaaiser t . )iri
^ Vapor
Mist Liquid , Liquid
Separator Centrifuge Ion Exchange
'
Particles
> 10 v
Solids
(B) Madified System
Figure 9.1 EBR-II Liquid Radwaste Systan
-------
Table 9.1
EER-II Solid Waste Management
Annual Production ^ 5 x 107 Ci
Shipped from site with fuel ^ 90% of production
Processed at site ^ 5 x 106 Ci
1971
Range Volume Activity
Level (Ci/ft3) (ft3) (Ci) Disposal Site
Low 1 x 10-5 to .1 13,000 3 x 102 NTRS Burial Ground
Intermediate .1 to 5 1,000 3 x 103 NTRS Burial Ground
High > 5 150 1.5 x 106 EBR-II Site High Level
Storage Facility
Pu & Other Stored in 55 gallon
Transuranium drums, wrapped in
Materials plastic, on asphalt
pad above ground at
NTRS (interim storage)
175
-------
REFERENCES (Section 9)
1. M. Jackson and W. Grady, EBR-II Liquid Radwaste System, Personal
Communication, Fast Reactor fcbrksbop, EHR-II Site, Idaho Falls,
December 1972.
2. P. Stoddart, EBR-II Solid Radwaste Management, Personal Oonmunication,
Fast Reactor Workshop, EBR-II Site, Idaho Falls, December 1972.
3. Directorate of Regulatory Standards, U.S.A.E.C., "Draft Environmental
Statement Concerning Proposed Rule Making Action: Numerical Guides
for Design Objectives and Limiting (Conditions for Operation to Meet
the Criterion 'As Low as Practicable1 for Radioactive Material in
Light-Water-Cooled Nuclear Power Reactor Effluents," January 1973.
176
-------
APPENDIX A:
Environmental Operating Data for Fermi, SEFOR, and EBR-II
Environmental radiation data for Fermi was reported monthly to
the AEC, and data for SEFOR was reported quarterly. These reports
are available in the AEC's public documents room in Washington, D. C.
Similar data for EBR-II are reported in the ANL monthly Reactor
Development Program Progress Reports. Data for the EBR-II are
tabulated only through 1971 because the AEC did not make the 1972
and 1973 progress reports (the ANL-RDP series) available to the
general public while this review was in progress. Since completion
of this review, these progress reports have been made available to the
public.
The data are summarized in the following tables. Much of the
data is quoted as total y, 3, ora activity. These data are of
limited usefulness since specific isotopes are not identified.
It is useful, however, to observe the type of environmental data
that must be reported to the AEC by the reactor operations. It is
also of limited use to observe that apparently no significant
differences in radiation in the environment surrounding Fermi
and SEFOR were observed over the background present before plant
operation. The integrated power for SEFOR is very low since it was
an experimental reactor which operated at low load factor and at a
power level of 20MW(th). Total exposure date (MWd) were not reported
for SEFOR.
LIST OF TABLES
Fermi:
A 1 Integrated Power and List of Reports
A 2 Liquid Waste Discharge
A 3 Gaseous Waste Discharge
A 4 Environmental Surveys -- Airborne Dust
A 5 Environmental Surveys -- Precipitation
A 6 Environmental Surveys -- Surface Water
A 7 Environmental Surveys -- Drinking Water
A 8 Environmental Surveys -- Milk
A 9 Environmental Surveys -- Fish
A10 Environmental Surveys -- Gamma Radiation
Page
179
180
181
182
183
184
185
186
187
188
177
-------
All Activity of Liquid and Gaseous Samples 189
A12 Primary Sodium Composition 193
A13 Primary Sodium Activity 194
A14 Uranium in the Sodium in the Transfer Tank System 195
SEFOR:
AT5 Off-Site Radioactivity Release in Gaseous Waste -- Noble Gases 196
and Activation Products
A16 Off-Site Radioactivity Release in Gaseous Waste —- Halogens 197
and Particulates
A17 Off-Site Radioactivity Release in Liquid Waste — Fission 198
Products and Activation Products
A18 Off-Site Radioactivity Release in Liquid Waste -- Tritium 199
and Carbon-14
A19 Environmental Sampling of Vegetation, Soil, and Water 200
A20 Environmental Filrr) Monitoring 202
A21 Primary Sodium Composition 203
A22 Primary Sodium Radioactivity 204
A23 Cover Gas Activity 205
EBR-II:
A24 Integrated Power and List of Reports (1971) 206
A25 Radionuclide Activity in Primary Sodium 207
A26 Radionucl de Activity in Secondary Sodium 215
A27 Gamma Activity in Cover Gas Due to Tramp Source 217
A28 Primary Sodium Composition — Trace Metals 218
A29 Primary Sodium Composition -- Non-metals 220
A30 Secondary Sodium Composition -- Trace Metals 221
A3! Secondary Sodium Composition -- Non-metals 223
178
-------
Report t & Date
Table Al
Integrated Power and List of Reports (FERMI)
Operation Period
Integrated Power
EF-101, January 1972
EF-100, December 1971
EF-99, November 1971
EF-98, October 1971
EF-97, September 1971
EF-96, August 1971
EF-95, July 1971
EF-94, June 1971
EF-93, May 1971
EF-92, April 1971
EF-91, March 1971
EF-90, February 1971
EF-89, January 1971
EF-88, December 1970
EF-87, November 1970
January 1972
July 16, 1970-January 31, 1972
December 1971
July 16, 1970-Docember 31, 1971
November 1971
July 16, 1970-Nwember 1971
October 1971
July 16, 1970-October 1971
September 1971
July 16, 1970-£eptember 1971
August 1971
July 16, 1970-August 1971 '
July 1971
July 16, 1970-July 1971
June 1971
July 16, 1970-June 1971
May 1971
July 16, 1970-May 31, 1971
April 1971
July 16, 1970-April 1971
March 1971
July 16, 1970-Htorch 1971
February 1971
July 16, 1970-February 1971
January 1971
July 16, 1970-January 1971
December 1970
July 16, 1970-December 1970
November 1970
July 16, 1970-November 1970
0
5941
0
5941
1491
5856
0
4365
0
4365
0
4365
0
4365
1029
4365
19
3336
155
3317
0
3162
142
3162
427
3020
145
2593
1381
2448
179
-------
EF-101, January 1972
EF-100, December 1971
EF-99, November 1971
EF-98, October 1971
EF-97, September 1971
EF-96, August 1971
EF-95, July 1971
EF-94, June 1971
EF-93, toy, 1971
EF-92, April 1971
EF-91, March 1971
EF-90, February 1971
EF-89, January 1971
EF-88, December 1970
EF-87, November 1970
Table A2
Liqu id Waste Discharge (FEBMI)
Total amount o>: discharge
(gallons)
0
6467
0
6937
0
7068
0
0
6967
0
0
6967
0
6967
1810
Total activity
mCi
0
4.02
0
0.72
0
1.1
0
0
1.376
0
0
4.62
0
(a)
(a)
(a)
(a) (b)
(a)
9.14
4.23
(a)
(a)
(a) All effluents released to the environment after dilution with the
circulating pump discharge were: below MPC.
(b) This value was -eported in EF-93 as 1376 mCi. By letter of Oct. 18,
1972, fron W. C. Morison, Fermi Assistant Plant Superintendent, to
A. B. Reynolds, it was learned that the correct number was 1.376,
not 1376 as reported.
180
-------
Table A3
Gaseous Waste Discliarge (FERMI)
The following paragraph appeared in every monthly report:
"Approximately 1x1O9 cubic feet of gaseous effluent were released
through the plant stack. The concentration of particulates and halogens
with half-lives greater than 8 days was less than 0.143 MPC at the waste
gas stack outlet. The concentration of all other isotopes was less than
100 MPC at the gas stack outlet. These levels meet the requirements of
the Technical Specifications."
181
-------
Erviromental Survoys Airhorn Dunt (FERMI)
Iteactor Area*
Report 1
.and Date
EF-101
Jan., 1972
EF-10C
Dec., 1971
EF-99
Nov., 1971
EF-98
Oct., 1»71
EF-97
apt., 1971
EF-%
Aug. , 1971
EF-95
July, 1971
EF-94
June, 1971
EF-93
May, 1971
EF-92
April, 1971
EF-91
(torch, 1971
EF-90
Feb., 1971
EF-89
Jan., 1971
EF-88 ,'
Dec., 1970/
/
KF-87/
Nov/.', 1970
/
Period
A (Aug. 5 - Sept. 2, 1 )71)
B )$ept. 2 - Sept. 30, 1971)
1%7\
No reports on environmental
\
A (June 10 -iJuly 8, 1971)
B (July 8 - Aug. 5, 1971)
1967
A (April 15 - May',13, 1971)
B (May 13 - June 10, 1971)
1967 \
March 18 - April 15, 1^71
V
No reports on environmental
A (Jan. 21 - Feb. 18, 1971)
B (Feb. 18 - March 18, 1971)
1967
No reports on environmental
/
No reports. on environmental
/ .
No reports op environmental
/
/
/
/
A (Sep1 . 3 - Oct. . 1 , 1970)
B (On . 1 - net . .">, 1970)
19*7
A (Juru' ] 1 - July 'I, I'l/ID
B (July '1 - Auq. (., 1 '!'/())
C (Aug. (> - Sept. 3, 1970)
No. of Gross a
Sanples uCi/ocx.O-15
39 2.18
40 1.70
2.67
surveys were received during
40 1.97 •/'
40 1.87 ./
2.67 /
40 2 . 3/
40 K77
/' 2.67
40 / 1.77
/ 2.67
surveys' were received during
•"•. 40 1.87
40 1.48
2.67
surveys were received during
surveys were Ireceivef1 during
surveys were received during
40 1.50
40 1.39
.'.hi
411 1 .70
40 i.v:,
40 2.07
Gross B
33.7
14.2 ;
7.5 /
this period.
62.7
58.8
7.5
53.1
60.3
7.5
29.5
7.5
this period.
10.5
14.1
7.5
this period.
th :; period.
th • s period.
\ 1.2.8
"'•. 1 2 . 9
\''
'•'•\
12.4 \
10. B \
Gross a Gross B
uCi/cad.0-15 uCd/ccxlO-111
1.84 33.6
1.75 13.5
3.31 7.7
1.85 60.3
1.55 57.9
3.31 7.7
2.25 56.8
1.93 .65.5
3.31 7.7
1.77 30.5
3.31 7.7
1.76 10.6
1.63 15.5
3.31 7.7
•
1.59 12.4
1.70 13.0
3.31 7.7
I.B'i (,|.4
l.Wi 44. '1
L.97 40.2
1967
2.67
3.31
**niere are five utopling stations around t^hn rnact-or .irea and f ivo st.ation away from the reactor area where
saitples are continuously collected and removed weekly, collectsd samples during the periods were analyzed
and averaged, and compared to the 1967 yearly average results (background group).
\
xj-
182
-------
Table A
-------
Page Intentionally Blank
-------
Table AS
Environrental Surveys - Precipitation (FERMI)
Reactor Area*
Background
Report 1
and Date
EF-101
Jan., 1972
No. of
Period . Sanples
A
B
(Aug. 5 -
(Sept. 2
- Sept. 2, 1971) 9
- Sept. 30, 1971) 10
1967
EF-100
Dec., 1971
EF-99
NOV., 1971
EF-98
Oct., 1971
No
A
B
reports
,
(June 10
(July 8 -
on environmental surveys were
- July 8, 1971) 10
- Aug. 5, 1971) 10
1967
EF-97
Sept. 1971
A
B
(April 15 - May 13, 1971) 10
(May 13 -
- June 10, 1971) 10
1967
EF-96
Aug., 1971
EF-95
July, 1971
EF-94
June, 1971
March 18 -
April 15, 1971 10
1967
ND
A
B
reports
(Jan. 21
(Feb. 18
on environniental surveys were
- Feb. 18, 1971) 10
- March 18, 1971) 10
1967
EF-93
May, 1971
EF-92
April, 1971
EF-91
March, 1971
EF-90
Feb., 1971
EF-89
Jan., 1971
EF-88
Dec., 1970
No
ND
No
A
B
reports
reports
reports
(Sept. 3
(Oct. 1 •
on environmental surveys were
on environmental surveys were
on environmental surveys were
- Oct. 1, 1970) 10
- Oct. 29, 1970) 9
1967
EF-87
Nov., 1970
A
B
C
(June 11
(July 9 •
(Aug. 6 •
- July 9, 1970)
- Aug. 6, 1970)
- Sept. 3, 1970)
Gross a Gross 6 Gross a
nCi/sq. mile nCi/sq. mile nCi/sq. mile
0.
0.
0.
151
113
205
received during
0.
0.
0.
0.
0.
0.
0.
0.
151
083
205
117
202
205
090
205
received during
0.
0.
0.
164
106
205
received during
received during
received during
0.090
0.
057
0.206
12
5
5
this
20
10
5
17
33
5
15
5
this
7
11
5
this
this
this
7
6
5
.2 0.173
.03 0.156
.1 0.337
period.
.7 0.219
.7 0.138
.1 0.337
.96 0.130
.3 0.177
.1 0.337
.5 0.140
.1 0.337
period.
.82 0.174
.2 0.197
.1 0.337
period.
period.
period.
•
.80 0.108
.99 0.067
.1 0.337
Gross
nCi/sq. i
10.4
4.85
6.04
25.1
11.8
6.04
18.08
28.3
6.04
12.7
6.04
6.86
10.3
6.04
7.57
8.76
6.04
*Two groups of samples wore, collected frcrn five locations around the reactor site and five locations away frcro
the reactor area to indicate background. The samples wore analyzed for gross a, 6 and -(—activity, and compared
to the 1967 yearly average results.
183
-------
Page Intentionally Blank
-------
9
10
0.151
0.113
0.205
12 -' 0.173.,-'
' 3 0.156
5 i .0.337
10.4
4.85
6.04
>:r.-i r ,rimcr---i: Surveys - Pi -i-i .tior. (FFKMI)
Reac'-or Area* Background
Report * No. of Gross a Grc s [3 Gross a Gross fi
arid Data Period Sanples m'i/sq. mil.- nCi/s . mile nCi/sq. mi,le mCi/aq. mile
EF-101 \ A (Aug. 5 - Sept. 2, 1971)
Jan., 1972 \ B (Sept. 2 - Sept. 30, 1971)
1967
EF-100 No reports on environmental surveys werp icoeived during this [.leriod. /
Dec., 1971 ;, /'
EF-99
Nov., 1971
EF-98 A (June 10 - July 8, 1971)
Oct., 1971 B (July 8 -..Aug. 5, 1971)
19C.7 ~'.
EF-97 A (April 15 - May 13, 1971)
Sept. 1971 B (May 13 - June\10, 1971)
1967 . \
EF-% March IS - April 15,-. 1971
Aug., 1571 1967 :
EF-95 No reports on environmental s
July, 1971
EF-94 A (Jan. 21 - Feb. 18, 1971)
June, 1971 B (Feb. 18 - March 18, 1971)
1967
EF-93 No reports on environmental surveys were received during this period.
May, 19*1
EF-92 Nc reports on environmental surveys were received during this period.
April, 1971
EF-91 No reports on environmental surveys were received during this period.
March, 1971
EF-90
Feb., 1971
EF-89
Jan., 1971
EF-88 A 'Sept. '; - (x:t. 1, 1970) 10
Dec., 1970 B (Oct. 1 -'IT*. 29, i970j ')
EF-87 'A (June 11 - July '), 1970)
»>•/., 1970 / ft fJuly 9 - Auq. >., 1970)
r~. (Ajfj. f. - Sept. -, 1970)
*"••!<:/'• f<, i: -. -' :; urn Jen •...•'•"• • -v;l i. •';(/.*:' f r-x-- : i1'- ]<>>M( if i--, ir^'inil f.lv- i".n-lcn -;i(.- ,nu! I i vn Isx'ntiour; away frcni
f b/ rf:\<~''ir .\r<-,\ t^ i nr: I'Mt/: b.r-K'n'o'jM'i. 'Die -:.an'-lr-s •"•[' injly/.cx! for qrns:: M, : ,uuf v-,u't i vi (-A-, anci ccsnnaret!
to tlic 1%7 y.-.irlv .iv..|.»; rc.HM I.K.
10
10
10
10
10
>ys were
10/
:10
0.151
0.083
' 0.205
0.117
0.202
0.205
0.090
0.205
received during
0.164
0.106
0.205
2( '.7
10.7
r.l
1 '.96
31.3
5.1
15.5
5.1
this period.
7.82
11.2
5.1
0.219
0.138
0.337
0.130
0.177
0.337
0.140
0.337
0.174
0.197
0.337
25.1
11.8
6.04
, 18.08
28.3
6.04
12.7
6.04
6.86
10.3
6.04
0.090
0.0'i7
n..>or
: 7.-'i. 0.108
',(,.-)') C.OG7
-------
T.iblc.' A*
Cn'/ironmental Surveys - Surfa " Wator (FERMI)
Report *
and Date
EF-101
Jan., 1972
EF-100
Dec., 1971
EF-99
Nov., 1971
EF-98
Oct., 1971
EF-97
Sept., 1971
EF-96
Aug., 1971
EF-95
July, 1971
EF-94
June, 1971
EF-93
May, 1971
EF-92
April, 1971
EF-91
March, 1971
EF-90
Feb., 1971
EF-89
Jan., 1971
EF-88
Dec., 1970
EF-87
Nov., 1970
*T\JO qroups
groas , act
No. of
Period Samples*
A (Aug. 5 - Sept. 2, 1971) 16
B (Sept. 2 - Sept. 30, 1971) 16
1967
No reports on environmental surveys were
A (June 10 - July 8, 1971) 12
B (July 8 - Auq. 5, 1971) 12
1967
A (April 15 - May 13, 1971) 12
B (May 13 - June 10, 1971) 12
1967
March 18 - April 15, 1971 12
1967
No reports on environmental s- rveys were
A (Jan. 21 - Feb. 18, 1971) 12
B (Feb. 18 - March 18, 1971) 12
1967
No reports on environmental surveys were
No reports on environmental surveys were
No reports on environmental surveys were
•
A (Sept. 3 - Oct. 1, 1970) 12
B (Oct. 1 - Oct. 29, rncj 12
Swan Creek
xlO~' vQ/nH
6.73
5.92
8.68
received during
8.22
7.13
8.68
10.2
9.74
8.68
6.35
. 8.68
received dur JKJ
14.5
8.49
8.68
rece ived dur ing
received dur .ng
received during
9.14
8.59
Lake Erie
(Intake)
xlO-' yCi/fel
5.68
6.29
5.hO
this period.
6.97
5.78
5.60
6.47
8 . 50
5.60
8.87
5.60
this period.
15.2
7.12
5.60
this period.
this period.
this period.
4.58
5.21 ,
of sotrples were oollected frtm four or three locations (one sample a week
ivity. The results were compared to 1967 yearly averaged results.
Reactor Reactor
Channel (Outlet)
5.17 4.91
5.58 5.45
5.79 Not Collected
6.77 —
6.13 —
5. "79 —
6.53 --
10.8 —
5.79 —
7.00 —
5.79 —
6.70 —
7.08
5.79 —
6.80 —
5.42 —
from each) and analyzed foi
184
-------
Table A7
Envirormental Surveys - Drinking Water (FERMI)
Report #
EF-101
EF-100
EF-99
EF-98
EF-97
_ EF-96
CO
in
EF-95
EF-94
EF-93
EF-92
EF-91
EF-90
EF-89
EF-88
EF-87
Period
A
B
1967
No of Monroe '
Samples* xlO~' tiCi/i
32
32
No reports on
No reports on
A
B
1967
A
B
1967
3/18/71
4/15/71
1967
32
32
32
32
32
No reports on
A
B
1967
32
32
No reports on
No reports on
No reports on
No reports on
No reports on
A 32
B
1957
32
No reports on
4.18
4.64
3.26
environmental
environmental
5^58
4.60
3.26
5.14
5.06
3.26
4.16
3.26
environmental
11.8
7.15
3.26
environmental
environmental
environmental
environmental
environmantal
4.32
4.02
3.26
environmental
Flat ftock Dundee Toledo
nl xlO"9 >iCi/ml xlO-9 uCJ/inl xlO~9 jjCi/ml
6.83
6.41
3.73
surveys were
surveys were
7.49
7.68
3.73
6.04
. 7.27
3.73
6.26
3.73
surveys were
12.6
10.6
3.73
surveys were
surveys were
surveys were
surveys were
surveys were
6.54
5.83
3.73
surveys were
6.17
7.14
3.17
received during
received during
7.37
6.99
3.17
6.69
5.37
3.17
5.83
3.17
received during
13.9
11.4
3.17
received during
received during
received during
received during
received during
7.20
6.24
3.17
received during
4.20
3.65
2.51
this period.
this period.
5.60
4.88
2.51
4.34
3.89
2.51
6.05
2.51
this period.
9.36
7.69
2.51
this period.
this period.
this period.
this period.
this period.
4.14
3.53
2.51
this period.
Detroit
xlO"9 yCi/ml
3.
3.
2.
4.
4.
2.
4.
4.
2.
3.
2.
8.
4.
2.
4.
3.
2.
86
87
64
24
16
64
98
64
64
16
64
25
22
64
01
66
64
Allen Park
XlO-9 pCi/ml
3.76
4.27
Not Collected
5.25
4.07 '''
Not Collected
4.30
4.96
Not Collected
4.32
Not Collected
9.89
6.15
Not Collected
3.82
•$ 3.77
Not Collected
Ann Arbor
xlO-9 pCi/ml
3.99
3.76
2.23
5.47
4.76
2.23
6.38
4.89
2.23
5.18 v
2.23
9.67
38.7
2.23
4.49
4.22
2.23
Colchester
xlO"~! yCi/ml'
6.47
3.54
Not CollectaJ
-
8.24
5.24
- Not Collected
5.57
4.94
Not Collected
4.66
Not Collected
19.8
7.93
Not Collected
3.29
9.18
Not CollectaJ
*TVo groups each containing thirty-two samples (daily composite samples collected weekly) from neighboring cities were analyzed for gross 6
activity, and compared to the 1967 results.
-------
Page Intentionally Blank
-------
Table A7
Enviroraicntnl .Surveys - Drinkim Water (FKFNI)
Report *
EF-101
EF-100
ET-99
EF-98
•'•- ' :-."„
EF-97
: _, ",.-"•
•_, r EF-%
OO
tn
-•;• EF-95
/ EF-94
EF-93
EF-92
EF-91
EF-90
EF-89
EF-88
Period
A
B ..
1967
No of Manroe Flat Rock Dundee Toledo
Sanples* xlO~' pCi/nd xlO~' pCi/ml xlO-' uCi/ml xlCr' MCi/ml
32
32
No reports on
No reports on
A
B
1967
A
B
1967
3/18/71
4/15/71
1967
32
32
32
32
32
No reports on
A
B
1967
32
32
No reports on
No reports on
No reports on
No reports on
No reports on
A
B
1967
32
32
4.18
4.64
3.26
environmental
environmental
5.58
4.60
3.26
5.14
5.06
3.26
4.16
3.26
environmental
11.8
7.15
3.26
environmental
environmental
environmental
environmental
environmental
4.32
4.02
3.26
6.83
6.41
3.73
surveys were
surveys were
7.49
7.68
3.73
6.04
7.27
3.73
6.26
3.73
surveys were
-LZ.O
10.6
3.73
surveys were
surveys were
surveys were
surveys were
surveys were
6.54
5.83
3.73
6.17
7.14
3.17
received during
received during
7.37
6.99
3.17
6.69
5.37
3.17
5.83
3.17
received during
13. :•
11.4
3.17
received during
received during
received during
received durifig
received during
7.20
6.24
3.17
4.20
3.65
2.51
this period.
this period.
5.60
4.88
2.51
4.34
3.89
2.51
6.05
2.51
this period.
>.3fc
7.69
2.51
this period.
this period.
this period.
this period.
this period.
4.14
3.53
2.51
Detroit
xlO"' yCi/ift
3
3
2
4
4
2
4
4
2
3
2
-
4
2
4
3
2
.86
.87
.64
.24
.16
.64
.98
.64
.64
.16
.64
--
.22
.64
.01
.66
.64
Allen Park Ann Arbor
1 xlO~' pCi/ml xlO-' pCi/ml
3.76 3.99
4.27 3.76 ,
Not Collected 2.23
5.25 5.47
4.07 4.76
Not Collected 2.23
4 . 3C i. . jo
4.96 4.89
Not Collected 2.23
4.32 5.18
Not Collected 2.23
3.8s 9.t
6.15 38.7
Not Collected 2.23
3.82 4.49
3.77 4.22
Not Collected 2.23
Colchester
xlO-' MCi/ral
b.47
3.54
Not Collects!
8.24
S.24
Not Collected
b.b7
4.94
Not Collected
4.66
Not Collected
19.8
7.93
Not Collected
3.29
9.18
Not Collected
EF-87
No reports on environmantal surveys were received during this period.
"Two groups eacii containing thirty-two samples (daily conpogite arables collected weekly) from neighboring cities were analyzed for gross &
activity, and oonpared to the 1967 results.
-------
Table AS
l-'nvironmental Purveys - Milk (F 3
-------
Table A9
Environmental Surveys — Pish (FEEME)
EF-101: Twelve sample; of Lake Erie fish (six from the reactor area
and six from the Buffalo area) were taken and analyzed for
gross 3 and Y activity during period A. No significant change
in the results was noted when compared to previously analyzed
samples.
EF-94: Twelve samples of Lake Erie fish (six from the reactor area and
six from the Buffalo area) were taken and analyzed for gross
B and y activity during period B. No significant change in the
results were noted when compared to previously analyzed samples.
137
-------
Table A10
Environmental Surveys --- Gamma Radiation (Fermi)
Two groups (A»B) each containing ten gamma radiation exposure
analyses were conducted (ten locations of on , four week exposure)
by the usage of environmental film packets. No gamma exposures
above normal were reported.
The same analyses were done for each period and the same
results were reported.
183
-------
Table .All
Activity of Liquid and Gaseous Sanples (FERMI)
Liquid
Report f
and date
reported
EF-101
Jan. 1972
EF-UO
Dec. 1971
EF-99
Nov. 1971
EF-98
Oct. 1971
Date
Taken
Jan. 19
Jan. 19
Dec. 7
Dec. 10
Dec. 14
Dec. 14
Dec. 14
Dec. 14
Dec. 20
Dec. 20
Nov. 18
Nov. 18
Oct. IS
Oct. 15
Oct. 18
Location
Demineralized water
Potable Water
—
MK-15 Liquid Waste Tank
MK-15 Liquid Waste Tank
Cut-up Pool before Ion Exchange
Cut-up Pool after Ion Exchange
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Demineralized Water
Potable Water
Demineralized Water
Potable Water
Demineralized Water
Potable Water
Waste Liquid Tank MK-15
Gross e
activity
(highest
concentra-
tion)
jjCi/cm3
1.22xlO~8
1.22xlO~8
3.1xlO~5
1.7x10-"
6.8xlO~8
l.SxlO"8
2.0xlO~7
3.9xlO~8
1.2xlO~8
4.5xlO~8
4.76xlO~8
1.22xlO"8
1.95xlO~8
5.86xHT8
2.73xlO"5
Gaseous
No. of
Samples
4
4
3
1
4
5
4
1
1
5
4
5
1
1
4
4
4
1
Location
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Reactor Cover Gas
Waste Gas Storage Tank
No: 2
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Reactor Cover Gas .
Waste Gas Storage Tank
No: 2
Waste Gas Storage Tank
No: 1
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Reactor Cover Gas
Waste Gas Storage Tank .
Waste Gas Discharge Line
No: 1
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
FARB Transfer Tank
Gross 8
activity
(highest
concentra-
tion)
yCi/cm3
3.5xlO~8
1.6xlO~5
3.5xlO~5
3.1xlO~5
3.5xlO-8
1.4xlO"6
4.2xlO~3
2.2x10-"
4.1X10-1*
2.75xlO"3
4.55xlO-6
1.82X10"1
1.32X10-1*
3. 19x1 0"5
1.24xlO-6
3.16xlCT8
5.39xlO~5
5.94xlO~7 '
00
-o
-------
Page Intentionally Blank
-------
Table All
Activity of Liquid and Gaseous Sarqoles (FERMI)
CO
V
Report.^ Date
and dateT~ '-^_ Taken
reported "--.^
""•-..,
~"~"--~,^
""•"
EF-101 Jan. 19
Jan. 1972
1 Jan. 19
i
1 EF- X)0 Dec. 7
Dec. 1971 ,_
Dec. 10
Dec. 14
Dec. 14
j Dec. 14
j ' , .--
'' , Dec. 14
j •••• Dec. 20
Dec. 20
Er'-99 Itov. 18
Nov. 1971 _X'"'
„-••"'
,-^
^-Nsv. 18
,s"
EF-98 Oct. 15
Oct. 1971
Oct. 15
Oct. 18
Liquid
Location
Demineralized water
"*•--,..
""--,.,
Potable Water
"~~-'--,.^
MK-15 Liquid Waste Tank
MK-15 Liquid Waste Tank
Cut-up Pool before Ion Exchange
Cut-up Pool after Ion Exqharige
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Demineraiizea .vaner
Potable Water
Detnineralizea Watei
Potable Water
Demineralized Water
Potable Water
Waste Liquid Tank MK-15
Gross 8
activity
(highest
concentra-
tion)
vCi/an3
1.22xlO"8
1.22x10""
3.1xfe^
1.7X10-1* '"\
6.8x10"'
1.8x10"'
2. 0x10" 7
3. 9x10" "
^.2xlO"f
4.5xlO"8
4.76xlO"8
1.22xlO"e
1.95xlO-8
5.86xlO-8
2.73xlO"5
No. of
Samples
4
4
.-•• '3
1
4
5
'••-._.
""'M-,
1 "' --..
1
5
4
5
1
1
4
4
4
1
Gaseous
Location
.-•••- Primary Shield Tank
Auxiliary Fuel Storage
Facility-
Reactor Cover Gas
Waste Gas Storage Tank
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Reactor Cover Gas
Waste Gas Storage Tank
Mo: 2
'Waste Gas Storage Tank
'{*>: 1
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Reactor Cover Gas
Waste Gas Storage Tank
Waste Gas Discharge Line
No: 1
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
FARE Transfer Tank
X
Gross i
activity
(highest
concentra-
tion)
uCi/on3
3. 5x10- f
l.SxlO"7
3.5x10-"
3.1X10"5
3.5x10-"
1.4xlO"£
4.2x10"-
2.2x10"- I
4.1x10-"
2.75xlO"3
4.55xlQ"6
1. 82x10" :
1. 32x10-"
3. 19x1 0~-
1.24xlO-6
3.16xlO"s
5.39x10"^
5.94x10-
-------
Table AIL
Activity of I.itjuid and Gaseous Samples (FERMI)
(continued)
page 2
Liquid
Report «
\and date
reported
"'*-.
: EF-97
Sept. 1971
EF-96
Aug. 1971
EF-95
July 1971
EF-94
June 1971
^.'•'
EF-93
May 1971
Date
Taken
^v
*""\,
Sept. *"r>^
Sept. 13
Sort . 1 3
Sept. 13
Sept. 20
---<-. 21
Aug. 18
Aug. 24
Aug. 24
July 19
July 19
June 7 ^x""
S*tuns 1
.X*" June 7
June 7
June 18
June 18
May 6
May 19
May 19
Location
Cut-up Pool before Ion Exchange
"-•-,..
Cut;-i?> after Ion Exchange
Decay 'PSQ! bff™re. Ion Exch-tikje
Decay Pool after Ion Exchange
Demineralized Wafer,,^
~ ' . ': ' '•'"-t ~ -.-.
Waste Liquid Tank MK-15
Demineralized Water
Potable Water
Detiineralized Water
Potable J^ater
.„-''""'
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Cut-up Pool, before Ion Exchange
Cut-up Pool after Ion Exchange
Demineralized Water
Potable Water
Liquid Waste Tank MK-15
Damineralized Water
Potable Water
clross K
activity
(highest
concentra-
tion)
..ci/™'
9.62x10""
. 3.54x10""
1.66x10-
Gasooa'; y
No. of
Samples
5
5
?•"''"" -
1.22x10-" \f--°"
1.22x10—
K 34x10- X
4 '.'08x10-'
2. 1x10" '
1.22xlO~s-.,
Below detectable
limits
4
4
3
"" , 5
Below uetectable < 5
liitdts
2.24x10""
1.68x10'"
S.SxlO"7
1.22xlO"8
l.,34xlO-8
5.81xlO-8
5. 18x10" 5
2.6xlO-8
7.32x10-°
5
4
5
6
3
3
5
4
4
Location
_.X'
/"'"
^Auxiliary Fuel 2toracje
,.,-"" Facility
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
Auxiliary Puel Storage
Facility
Primary Shielrf Tank
"'* „ Reactor Cover Gas
Auxiliary Fuel Storage
Primary Shield r
O
-------
Table All
Activity of Liquid and Gaseous Samples (FERMI)
(continued)
page 2
Liquid
Report #
and date
reported
EF-97
Sept. 1971
EF-96
Aug. 1971
EF-95
July 1971
EF-94
June 1971
EF-93
Kay 1971
Date
Taken
Sept. 13
Sept. 13
Sept. 13
Sept. 13
Sept. 20
Sept. 21
Aug. 18
Aug. 24
Aug. 24
July 19
July 19
June 7
June 7
June 7
June 7
June 18
June 18
May 6
May 19
May 19
Location
Cut-up Pool before Ion Exchange
Cut-up after Ion Exchange
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Dsnineralized Water
Potable Water
Waste Liquid Tank MK-15
Demineralized Water
Potable Water
DOTineralized Water
Potable Water
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Cut-up Pool before Ion Exchange
Cut-up Pool after Ion Exchange
Denineralized Water
Potable Water
Liquid Waste Tank MK-15
Demineralized Water
Potable Water
Gross B
activity
(highest
concentra-
tion)
jiCi/cm3
9.62xlO~8
3.54xlO~8
1.66xlCT7
1.22xlO~8
1. 22xlO-8
1.34xlO~8
4.08xlO~5
2.1xlO"8
1.22xlO~8
Below detectable
limits
Below detectable
limits
2.24xlO~7
1.68xlCT7 •
5.5xlO~7
1.22xlO~8
- 1.34xlO~8
S.SlxlQ-8
5. 18x10" 5
2. 6xlO~ 8
7.32xlO-8
Gaseous
No. of
Samples
5
5
5
4 '
4 .
3
5
5
• 5 •
4
5
6
3
3
5
4
4
Location
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Primary Shield Tank
Reactor Cover Gas
Waste Gas Tank tfo: 1
Waste Gas Tank No: 2
Reactor Cover Gas
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Gross 6
activity
(highest
concentra-
tion)
jjCi/cm3
1.37xlO"6.
7.49xl(T7
9.8X10-1*
1.55xlO"5
2.83xlO~8
5.84xlO~5
1.6xlO~6
4.9xlO~7
1. 2xlO-3
1.32xlO~6
2. 15x1 0"1*
1.8xlO-: •
1.28x10,-"
9. 94x1 O"5
.
9.94xlO~5
1.97xlO~7
1.16xlO~6
to
O
-------
Page Intentionally Blank
-------
Table AH
Activity of Liquid and Gaseous Samples (FERMI)
(continued)
page 3
Liquid
Report #
and date
reported
EF-92
April
EF-91
torch 1971
.Date
Taken
April 16
April 16
March 1
March 1
March 2
* ' March 2
March 2
March 2
March 22
March 22
Location
Denineralized Water
Potable Water
Demineralized Water
Potable Water
Cut-up Pool before Ion Exchange
Cut-up Pool after Ion Exchange
Decay Pool before Ion Exchange
Decay Pool after Ion Exchange
Potable Water
Demineralized Water
.Gross 6
activity
(highest
concentra-
tion)
.yCi/cm3
1.2xlO-8
1.2xl(T8
1.2xlO"8
3.5xlO~8
9.9xlO~8
3.3xlO~8
l.VxlCT7
8.6xl(T8
1.2xlO~8
1.2xlO~8
' EF-90
Feb. 1971 •
EF-89
Jan. 1971
Jan. 13
Jan. 13
Jan. 20
Potable Water
Desidneralized Water
Condenser out-Fall
' 2.44xlO~8
1.22X1CT8
i.esxicr7
Gaseous
•No. of
Samples
5
5
5
4
3
2
1
4
4
• 3
1
1
Z
3
4
4
1
1
Location .
Reactor Cover Gas
Primary Shield Tank
Auxiliary Fule Storage
Facility
Reactor Cover Gas
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Containment Building
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Waste Gas Storage Tank
No: 1
Transfer Tank
Containment Building
Reactor Cover Gas
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Machinery Dome
Waste Gas Tank No: 1
Gross B
activity
(highest
concentra-
tion)
jjCi/cm3
9.1X10"1*
2.3xl(T7
l.BxlQ-6
5.6xlO~5
3.7xlO~8
. 1.3xlO~6
»
8.7xl
-------
Page Intentionally Blank
-------
Table All
Activity of Liquid and Gaseous Samples (FERMI)
(continued)
page 3
x\ Liquid
Report » x Date Location
and date " Taken
reported
EF-92 . April 16 Deinineralized Water
Aprl April 16 Potable -Water
EF-91 .March 1 Danineralized Water
; 'vil-h "" "• March 1 Potable Water
March 2 Cut-up Pool before Ion Exchange
March 2 Cut-up Pool after Ion Exchange
March 2 Decay Pool before Ion Exchange
i
March 2 Decay Pool after Ion Exchange
March 22 Potable Water
March 22 Denineralized Water
EF-90
Feb. 197]
EF-89 Jan. 13 Potable Water
Jan. 1971
Jan. 13 Deraineralized Water .
Jan. 20 Condenser out-Fall
Gross i<
activity
(highest
concentra-
tion)
uCi/cm3
1.2x10-"
1.2x10""
1.2x10-" .-
9.9x10-
•-•". 3.3x10-"
1.7x10-
8.6x10--
1.2x10""
1.2x10- :
2.44xlO-s
1.22xlO-a
1.68xlQ-7
No. of
Samples
5
5
5
3
2
1
4
3
1
I
2
3
4
4
1
1
Gaseous
location
Reactor Cover 'las
,..---•'" Primary Shield Tank
Auxiliary Fule "tcraae
Facility
Reactor Cover 3as
Priniary Shield 7arJ<
Auxiliary Fuel Stcraae
Facility
Containnient Building
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Waste Gas Storaoe Tank
ND: 1
Transfer Tank
Containment Building
Reactor Cover Gas
Primary Shield Tank
Auxiliary Fuel Storage
Facility
Machinery Dare
Waste Gas Tank lie: 1
Gross f.
activity
(highest
jjpijcentra-
uCi/cm3
9.1x10"-
2.3x10-
1.5x10-
i
5.6xlC--'
3.7x10-'
1.3X1Q-"
8.7x10"^
1 'V.-1 A— '
- 1.2x10--
1.6x10"-
2.0x10-'
1.1*10-'
1. 26x10- ':
, 1. 57x10- :
9.19xlO-;
5.05x10-'
1.37x10-'
8.69x10-'
-------
Table All
Activity of Liquid and Gaseous Samples
-------
Table All
Activity of Liquid and Gaseous Samples (FERMI)
(continued)
page 4
Liquid
Report # Date
and date Taken
reported
EF-88 Dec. 7
Dec. 1970 Dec- ?
Dec. 7
Dec. 7
Dec. 9
Dec. 15
Dec. 15
EF-87
Nov. 1970
Location Gross g
activity
(highest
concentra-
tion)
vCi/cm3
Decay Pool before Ion Exchange 1.7xlO~6
Decay Pool after Ion Exchange 1.4xlO~7
Cut-up Pool before Ion Exchange 7.4xlO~7
Cut-up Pool after Ion Exchange • 7.4xlO~7
Waste Liquid Tank MK-15 3.4X10"1*
Demineralized Water 2.6xlO~8
Potable Water 2.8xlQ-8
4
Gaseous
No. of
Samples
5
5
4
4
1
5
4
4
4
1
1
1
1
1
Location
Primary Shield Tank
Auxiliary Fuel Storage
Tank
Reactor Cover Gas
Waste Gas Tank No: 1
Containment Building
Primary Shield Tank
Reactor Cover Gas
Auxiliary Fuel Storage
Facility
Waste Gas System FARE
Waste Gas Tank No: 2
Waste Gas Tank No: 1
Waste Gas Tank No: 2
Waste Gas Tank No: 1
Gross 8
activity
(highest
concentra-
tion)
yCi/on3
1.6xl.CT6
1.6xlO-6
2.9xlO-3
3.8xlCT5
1.7xlO~6
2.6xlCTtt
8.1xlO-2
2.8xlO~6
3.2xlQ-7
6.4xlO-5
4.6xlO-5
2.5xlO~3
2.7x10-"
-------
Page Intentionally Blank
-------
Table A12
Primary Sodiun Composition (FERMI)
Chemical Analysis
Report #
Sample #
Date
taken
(ppn)
Cl
(ppn)
Cr
(ppn)
(ppn)
Fe
(ppn)
Non-hydroxide
(ppn)
0
(ppn)
r-9J Purity variation of Primary sodiur.. test was concerned with the effect that temperature and radiation would have on the shield
plug graphite for releasing impurities into the primary sodiun. Comparison of analyses of a sodiun sample taken prior to the
power demonstration program and one taken immediately following the 200 m(t) operation showed negligible changes in impurity
levels. Oxygen level remained below 10 ppn,- carbon level vas less than 35 ppn; and the indicated hydrogen level was less than
3 ppm. There was no apparent correlation between power operation and impurity level. Operation of the cold trap during power
operation apparently maintained the low level impurities in the sodiun.
10
oo
EF-92
Coil #58
Coil #59
April 5, 1971
April 12, 1971
0.7
0.3
not published
EF-88
Coil #54
Oct. 22, 1970
33
0.078
Hydroxide H = 2.30 ppm
0.268
0.66
0.138
-------
Table A13
Prinary Sodium Activity (FERMI)
Fission Products
activity
Report # ( yCi/cc of Na)
EF-101 2.5x10"'
EF-100
EF-99
EF-98
EF-97
EF-96
EF-95
EF-94
EF-93
EF-92
EF-91 1.8xlO~2
EF-90
EF-89
EF-88
EF-87
24
Na activity
(mCi/cc of -Na)
***
13
13
2.3xlO"5
2.3x10"5
2.5xlO~5
2.5xlO~2
12.23
0.865
2.68
0.5
1.0
1.6
13
Date
Reported
January 1972
December 1971
November 1971
October 1971
September 1971
August 1971
July 1971
June 1971
May 1971
April 1971
March 1971
February 1971
January 1971
December 1970
November 1970
Secondary System: No radiation levels greater than instrument
background were detected at the surface of the
steam generators which indicates that the ?A
radioactivity was well below 2.5xlO~^yCi of Na
per cc of sodium as required by the Technical
Specifications.
***Purpose of the sample taken is an investigation to determine the cause
of Na24 found in the containment building atmosphere in December, 1971.
The sample is being analyzed to establish base data on Na^2 and other
constituents.
194
-------
Table A14
Uranium in the Sodium in the Transfer Tank System (Fermi)
A sodium sample taken from the electromagnetic pump line of
the FARB cold trap room transfer tank system on May 27, 1970 was
analyzed in two parts for December, 1970 (EF-88) as follows:
Sample A Sample B
OOQ
U 11.8+0.7 ppb 2.9+0.4 ppb
235
"DU 1.9+0.5 ppb <0.5 ppb
195
-------
Table A15
Off-Site Radioactivity Release in Gaseous Waste
ttoble Gas and Activation Products (SEPOR)
x—
vo
en
Report 1 \
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
6th Quarterly
5th Quarterly
4th Quarterly
3rd Quarterly
2nd Quarterly
>slf Quarterly
Total Activity Total
Released Volute of
\ (Noble Gases Gas
Xand activation Released
products) (ft3)
\(Ci)
\ . •
8.99xl6\-x 261,000
6.38x10" \ 127,000
\
•
1.37 ''155,700
7.1x10"' ] ?" son
390,000
125,280
153,380
108,650
265,200
234,350
Time MPC
Average used
Release (pCi/ml)
Rate
(uCi/sec)
1.14x10-' 2xlQ~f
.81x10"' 2x10—
.174 2xlO"r
9.0x10" 2x1 0""1
%>,v Maximum Radioactivity Jlec
^"~L. "
Lorig-rlived Jaong-lived
Gross X^ Sr Gross 6
_,.--' '^-^
x-uio- ^o-;=
Licensed Percent Maximum License; 1 Percent
Limit of Annual Hourly Average Limit of Hourly
for Annual Limit Release Rate for Hourly Limit
Average (»iCi/«ec) Average
(wCi/sec) (uC,i/sec)
800 1.4x10-" 544 ,.,,-'"' 3400 16
800 IxlO-1- ^
-------
Table A15
Off-Site Radioactivity Release in Gaseous Waste--
Ifcble Gas and Activation Products (SEFOR)
Report f Total Activity
Released
(Ifoble Gases
and activation
products)
(Ci)
10th Quarterly 8.99xlO~3
9th Quarterly 6.38xlO~3
8th Quarterly 1.37
7th Quarterly 7.1xlO~2
' Total
Volume of
Gas
Released
(ft3)
261,000
127,000
155,700
]24,500
Time
Average
Release
Rate
(liCi/sec) .
1.14xlO~3
.SlxlO"3
.174
9.0xlO~3
MPC1^
used
(pCi/rel)
2xlO~8
2xlO-8
2xlO~8 '
2xlO-8
Licensed Percent ' Maximum
Limit of Annual Hourly Average
for Annual Limit Release Rate
Average (pCi/sec)
(yCi/sec)
800 1.4X10-1* 544
800 IxlO-1* 5.1xlO~2
800 2.2xlQ-2 18
800 l.LxlO-3 .19
Licensed Percent
Limit of Hourly
for Hourly Limit
Average
(pCi/sec)
3400 16
3400 l.SxlO-3
3400 .53
3400 5.6xlO~3
Maximum Radioactivity Measured
6th Quarterly • .
5th Quarterly
4th Quarterly
3rd Quarterly
2nd Quarterly
1st Quarterly
390,000
125,280
153,380
108,650
265,200
234,350
Long-lived
Gross a
(uCi/ml)
-------
Page Intentionally Blank
-------
Table A16
Off-Site Radioactivity Release in Gaseous Waste-
Halogens and Particulates (SEFOR)
Report #
Total
Activity
Released
(Halogens
arid Parti-
culates)
(Ci)
Total
Volume -.
of Gas
Released
(ft3)
Tiine
Average
Release
Rate
(pCi/sec)
MPC
Licensed
Limit
for Annual
Average
(nCi/sec)
Percent
of Annual
Limit
Maximum
Hourly Average
Release Rate
(jiCi/sec)
Licensed
Limit
for Hourly
Average
(pCi/sec)
Percent of
Hourly
Limit
' 10th Quarterly <7.39xlO~5 261,000 <9.4xlO~6 IxlO"10 5.6xlO~3
.17
<.34
5.6xlO~2
9th Quarterly <3.6xlO~
127,000
. 6xlO-9
5.6xlO-3
<9.4x10-
5.6x10'
-2
.7x1O'3
8th Quarterly <4.4xlO~7 155,700 <5.6xlO~8 IxlQ-10 5.6xlO-3
<1.0xlO"
<9.4xlO
-7
5.6x10
1-2
<1.7xlO-3
7th Quarterly 3.5xlQ-7
124,500
4.4xlO-9
1x10
i-lO
5.6xlO~3
.79xlO~3
9.4x10
1-7
5.6xlO~2
1.7xlO~3
6th - 1st Quarterlies: None Observed
(a) Halogens and particulates with half-lives > 8 days.
(b) Based on the possible presence of 131I.
-------
Page Intentionally Blank
-------
Table A16
Off-Site Radioactivity Release in Gaseous Waste-
Halogens and Particulates (SEPOR)
Report # """""--
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
•--..Total
Activity
Released ~~~
(Halogens
and Parti-
culates) (a)
(Ci)
<7.39xlO~5
<3.6xlQ-s
..4.4x10-
3.5xlO~7
Total
Vfolune
--- — ...of Gas
Released
(ft1) ~""~-
261,000
127,000
^5,700, -' '
..,'•124,500
Time
Average
Release
Rate
9.4x10-
•-4.6MO-"
• - 5 . 6xi 0 "
4.4xlO~':
MFC
used to)
(nCi/ml)
1x10- ;:
1x10- ;r
ixlQ---
bclO-::
Licensed
Limit
for Annual
Average
(uCi/sec)
"sTfia^,^
5. 6x10- ^
,.^0-'
5. 6x10- ?
Percent
of Annual
Limit
.17
-•IxlO-'1'---,.
•-1.0x10-''
.79x10"
Hpurly Average
Release Rate
<.34
•;9.4xlO-^
•9.4xlO-7 - . .
9.4x10-'
Licensed Percent si
Limit Hourly
for Hourly Limit
d.Ci/gec)
5.6x10-- 6
5.6x10-- -1.7x10-
,„=••«*"- -;xl°-
5.6x10"- 1.7xlO~?
6th - 1st Quarter1i$s: None Ctoserved
(a) Halogens and particulates with half-lives > 8 days.
(b) Based on the possible presence of l 3'I.
-------
Table A17
Off-Site Radioactivity Release in Liquid
Vfeste—Fission Products and activation Products (SEPOR)11
oo
R4»rt # Total activity of
X^ Fission Products and
"""X^ Activation Products
*\, Released (Ci)
\,
10th Quarterly 1/28x10" 5
9th Quarterly < 3. 3x1 0"' : """>• ,
8th Quarterly 4.6xlQ~7
7th Quarterly <7.2xlO"?
Total volume Total volume Volume average MFC Percent ,-• mxijnum concentra-
of liquid of dillution concentration at used of limit' tion released, averaged
waste discharged water discharge point (uCi/ml) .^ftl over not more than
(gallons) (gallons) (pCi/W.) ,,-''' 24 hours (uCi/ml)
5.29x10' 5.0x10'- 6.8x10.-" >.'6xlO-'-(b> 0.72 9. 0x10"''
1.15x10" 5.0x10' -1.9x10" 3.0x10" (b) ^0.06 3.0x10'
12x10 62x10" 1.9x10"" 1.0x10"" <1.9 <3xlO~"
4H::8xlO- 50x10 ,3.8x10"" l.OxlO"7 <3.8 jxxu '
•-•.I „:••-, ,.h " Maximum radioactivity level measured
'xv _, "" (Fission products and activation products) Volume discharged
-.•-"" , ., . -• (gallons)
-» (uCi/ml) A (,,ci..'r-J.;
6th Quarterly ,,.>""
5th Quarterly
4 th>«Quarter ly
,f"f 3rd Quarterly
2nd Quarterly
1st Quarterly
1x10 ; 1.6xlO~; (Identified as 'tritium * 1773
and C-14 . No v emitters
observed above lxlO-'v"''uCi/ml)
-1x10- ' 8x10"" . 7774
1x10" 3xlO~; , 7399
1x10-' --1x10"' v, 4154
:lxlO-" '^IxlO"1' 6718
^1x10"' • lxlO~9 3750
(a) All liquids are released to a tile field. Measured concentrations refer to values at the point of discharge intc
(b) Na2? identified as aamK emitter.
the tile field.
-------
Table A17
Off-Site Radioactivity Release in Liquid
Waste—Fission Products and Activation Products (SEFOR)'
Report #
Total activity of Total volume
Fission Products and of liquid
Total volume Volume average MPC Percent
of dillution concentration at used of limit
Activation Products waste discharged water discharge point (yCi/ml) (%)
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly '
.-
6th Quarterly
5th Quarterly
4th Quarterly
3rd Quarterly
2nd Quarterly
1st Quarterly
Released '(Ci) (gallons)
1.28xlO~5 5.29xl03
<3.3xlO~6 l.lSxlO1*
4. 6xlO~ 7 12x10 3
<7.2xlO~7 4.8xl03
a yCVml)
-------
Page Intentionally Blank
-------
Table A18
Off-Site Radioactivity Release in Liquid Wastes-
Tritium and Carbon-14 (SEFOR)
Report #
, Total Curie
Activity PeLeased
(Ci)
Tritium
Volume Average
Concentration at
Discharge Point (a)
(pCi/ml)
(b)
Percent of Limit %
Total Curie
Activity Released
(Ci)
Carbon-14
Volute Average
Concentration at
Discharge Point (a)
(pCi/ml)
(c)
Percent of Limit %
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
6th Quarterly
5th Quarterly
4th Quarterly
3rd Quarterly
2nd Quarterly
1st Quarterly
1.35xlO-2- . 7.1xlO-5
4.4xlO-2 2.3x10-"
9.4xlO~2 4.0x10-"
8.0x10-' 4.2x10-*
Total activity of tritium and carbon-14 is
Not reported.
Not reported. "
Not reported.
Not reported
Not reported.
2.4 1.3xlO-6 6.9xlO-9 8.6x10-"
7.7 4xlO-3-iestimated
13.4 3. 5x10-" -estimated
1.4 3.0xlO-"-estimated
1.6 x 10-5 yci/ml.
Not reported.
Not reported. •
Not reported.
Not reported.
Not reported.
(a) All liquids are released to a tile field. Measured concentrations refer to values
at the point of discharge into the tile field.
Volume average concentration
at discharge
Total activity released
Total volume of dilution water
(b) MPC used 3 x 10-3 yCi/ml for one week breathing (soluble in water)
(c) MPC used 8 x 10~" yCi/ml for one week breathing (soluble in water)
-------
Page Intentionally Blank
-------
vo
T^ble A18
Off-Site Radioactivity Rolf vise in Litpiid Wasto?;-
Tritiun ami Girt«n-M
NX--,. '
Tritium Carbon-14 ,'
Volume Average Volume Average
Total Curie Cdncenbration at Total Curie Concentration at
Activity Released Discharge .Point (a) (b) Activity Released Discharge Point (a) (c)
Report # (ci) (yCi/ml) .... Percent of Limit % (ci) - (nCi/ml) Percent of Limit %
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
6th Quarterly
5th Quarterly
4th Quarterly
3rd Quarterly
2nd Quarterly
1st Quarterly
-(a) All liquids are
at the point of
1.35xlO-2
4.4x10-'
9.4xlO-2
S.OxlO-3
Tt>tal activity of
Not reported.
Not reported.
Tfot reported.
.f&t reported
-s>
Not reported.,,-"
released to a tile
discharge into the
Vblune average concentration
7.1xlO-5 - ,,. 2.4 1.3x10-' ,,,-" 6.9x10-' 8.6x10-"
2 . 3x10-" "\ 7 . 7 4xlO->-«stiia«te
-------
TSble A19
Environmental San^ling
of Radioactivity in Vegetation, Soil, and Water (SEPOR)
f
to
o
o
V\ Vegetation 'a)
Report *\^
All Quarterlies
10th Quarterly
9th Quarterly
8th Quarterly
7th .Quarterly
'
6th Quarterly
5th Quarterly ..-•''
. .-"-
Month
~"l*echeck level
pre-bperational
average ^""^
August, 1971 ~"v""-,.
'^optcri-T. ' 1971 ' """"•!
October, 197 1
.,,,. --
June, 1971
July, 1971
February, 1971 .
March, 1971
April, 1971
November, 1970 i;,= '
December, 1970
January , 1-971 ,,.-••""" , . -;
Auguat2-f'*1970
September, 1970
.,- -' October, 1970
May, 1970
June, 1970
July, 1970
i activity i ac.ti v i. t y
pCi/gm-ash
50
13
16
•V, 18
'*K-,. '
15
•15
• 15
32.6
30
20.7
16
23
19
<15
16.6
<15
15
17
18 '
1820
987
-549
861
862
11.08
1151 '--,., •
917
.. ^
1657
1640
1332
1509
1463
964
1471
1395
762
949
1317
Soil Y
-------
Table A19
Environmental Sampling
of Radioactivity in Vegetation, Soil, and Water (SEFOR)
r\»
o
o
Vegetation
Report #
All Quarterlies
10th Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
6th Quarterly
5th Quarterly
Month
Recheck level
pre-operational
average
August, 1971
September, 1971
October, 1971
May, 1971
June, 1971
July, 1971
February, 1971
March, 1971
April, 1971
November, 1970
December, 1970
January, 1971
August, 1970
September, 1970
October, 1970
May, 1970
June, 1970
July, 1970
a activity S activity
pCi/gm-ash.
50
13
16
18
<15
<15
<15
<15
32.6
30
20.7
16
23
19
<15
16.6
<15
15 -
17 •
18
1820
987
549
861
862
1108
1151
917
1497
1657
1640
1332
1509
1463
964
1471
1395
762
949
1317
Soil
a activity
32
25
<15
22
19
24.3
22
23
26
<15
20
26
15
18
<15
15.4
20
16
15
22
(b)
8 activity
i
45
34
27
19
44
25
27
22
29
55
23.3
21
23
15
48
43
22
35
29
26
Vfeter^
a activity 6 activity
3xlO~8
<2xlO-9
d.0~8
-------
Page Intentionally Blank
-------
Table .A19
Environmental Sampling
of Radioactivity in Vegetation, Soil, and Water (SEPOR)
(continued - page 2)
Report #
4th Quarterly
>, V
3rd Quarterly
2nd Quarterly
1st Quarterly
Month
February, 1970
March, 1970 .
April, 1970
November, 1969
December, 1969
January, 1970
August,. 1969
September,' 1969
October, 1969
May, 1969
June, 1969
July, 1969
Vegetation
a activity g activity
pCi/gm-ash
18 1007
15 2154*
15 762
<15 951
<15 966
18 705
<15 , 941
20 1003
<15 1055
18 1026
<15 966
<15 941
(b) (b)
Soil Water
a activity g activity
pCi/gm
16
15
16
22
21
31
22
<15
31
22
22
21
22
27
35
24
32
38
28
25
32
29
17
28
ti activity 6 activity
uCi/ml
,i*io-8
-------
Page Intentionally Blank
-------
Table A19
Enviroimental Sampling
of Radioactivity In Vegetation, Soil, and Water (SIPOR)
(continued - page 2)
ro
o
Report f
4th Quarterly
3rd Quarterly
2nd Quarterly
1st Quarterly
Nbnth
February, 1970
March, 1970
April, 1970
November, 1969
December, 1969
January, 1970
August, 1969
Septsriber, 1969
October, 1969
May, 1969
June, 1969
July, 1969
Vegetation (a)
a activity 8 activity
pCi/gm-ash
18 1007
15 2154*
15 762
<15 951
-15 966
18 705
'15 941
20 1003
<15 1055
18 1026
<15 966
•15 941
a activity 6 activity
pCi/gm
16
15
16
22
21
31..,,
22
<15
31
22
22
21
22
27
35
24
32
38
28 """--.,..
25
32
29
17
2S
Itetar*'
a activity 6 activity
yCi/ml
-------
Table A20
Environmental Film Monitoring (SEFOR)
Number of Stations: 17
Total Films Analyzed during Each Quarter: 51
Report t and
Report Period
10th Quarterly
Aug. 1, 1971 thru
Oct. 31, 1971
9th Quarterly
May 1, 1971 thru
July 31, 1971
8th Quarterly
Feb. I, 1971 thru
April 30, 1971
7th Quarterly
Nov. 1, 1970 thru
Jan. 31, 1971
6th Quarterly
Aug. 1, 1970 thru
Oct. 31, 1970
5th Quarterly
May 1, 1970 thru
July 30, 1970
4th Quarterly
Feb. 1, 1970 thru
April 30, 1970
3rd Quarterly
Nov. 1, 1969 thru
Jan. 31, 1970
2nd Quarterly
Aug. 1, 1969 thru
Oct. 31, 1969
1st Quarterly
May 1, 1969 thru
July 31, 1969
Maximum Radiation
Level Reported
17 millirad/quarter
16 millirad/quarter
12 millirad/quarter
0 millirad/month
0 ndllirad/month
0 millirad/month
0 ndllirad/month
0 millirad/itonth
4 millirad/month
4 millirad/month
Maximum Radiation
Level Reported during
Pre-Operational Survey
(millirad/month)
8
202
-------
Table A21
Privrary Sodium Composition (SEFOR)
Concentration (ppm) .
Report #
10th Quarterly
(8/1/71-10/31/71)
9th Quarterly
(5/1/71-7/31/71)
8th Quarterly
(2/1/71-4/30/71)
7th Quarterly
(11/1/70-1/31/71)
IS3
0
CO
Report #
10th Quarterly
(8/1/71-10/31/71)
9th Quarterly
(5/1/71-7/31/71)
8th Quarterly
(2/1/71-4/30/71)
7th Quarterly
(11/1/70-1/31/71)
Date
Taken Al Ag B Ba Be Bi C Ca Co Cr Cu Fe Li
8/16/71 20 <.2 <20 <6 <.6 <.2 25 6 <.6 <.6 .2 8
6/16/71 100 2 <10 <5 <.l <.l 22 6 6 4 <3 4 <10
3/17/71 <2 2 <12 <12 13 5 6 6 <20
12/18/70 < 6 <.6 <-6 <6 <.6 24 6 <6 < 6 <.6 4
Primary Sodium Composition (SEFOR)
Concentration (ppm)
(part 2 - continued)
Date
Taken Mn MD Nb Hi P Pb Sb Si Sn Ti U235 U238 V Zn • Zr
8/16/71 <.2 <.2 <60 <6 <60 5 <.6 <.2 .2(a) 7 (a) <.2 <6
6/16/71 .6 <.3 <30 <1 ' <100 <1 <30 2 < 3 < 1 .2(a^ 5(a) <.3 <1
3/17/71 <12 20 < 2 • (4±2) (a) (200+50) (a) <12
12/18/70 <.6 <.6 <6 <.6 12 <.6 <.6 <.6 <6
Mg
.6
6
6
<6
< Represents lower limit of detection for instrument used.
(a) ppb
-------
Page Intentionally Blank
-------
Table A21
Primary Sodium Oonpoaition (SEPQR)
Oonoantration (ppn)
Report *
10th Quarterly
(8/1/71-10/31/71)
' 9th Uuai tcrly
(5/1/71-7/31/71)
.f^f 8th Quarterly
". .X (2/1/71-4/30/71)
7th Quarterly
(11/1/70-1/31/71)
/ ro
o
to
Report 1
10th Quarterly
9th Quarterly
(5/V71-7/31/71)
8th Quarterly ...»••'
(2/1/71-4/30/71)"
7th Quarterly
(11/1/70-1/31/71)
Date
Taken Al Ag
8/16/71 20 <.2
6/16/71 100 2
3/17/71 <2
12/18, 70 < 6 <.6
Date
Taken Mn Mo Nb .
8/16/71 <.2
12/18/70 <.6 <.6
B BaBeBiGCaOoCrCuFe^j,
<20 <6 <.6 <.2 25 6 <.6 <.6 .2 8^
<10 <5 <.l <.l 22 66 4 <3 4 <10
2 <12 <12 13 5 6 6 <20
<.6 <6 <.6 24 6 <6 < 6 <.6 4
Primary Sodium Cfcrrjposition (SEPOR)
' (part 2 - dontinued)
Ni P Pb Sb si . . Sn Ti U2 ' * ' viio V Zn Zr
<6 <60 5 <.6 <.2 .2(a) 7
-------
ro
CD
Table A22
Prdroary Scxiiun Radioactivity (SEFOR)
dptv/16g sanple
Report #
10th Quarterly
(8/1/71-10/31/71)
9th Quarterly
(5/1/71-7/31/71)
8th Quarterly
(2/1/71-4/30/71)
7th Quarterly
(11/1/70-1/31/71)
Date Ag110 Co60 I131, I133
Taken
8/16/71 3.1x10" (a)
6/16/71 4.4x10"
3/17/71 4.9x10" (b)
12/18/70 2. 0x10" 4.4xl02 (b)
Na22 Na?" Rb86 Sb12"
1.2xl06 None
11 vi n^ <4vi n ^
• O^A-L.L' ^-lA^LU
1.9xl06 1.1x10"
4.1xl05 vLxlO11 1.6xl05 4. 8x10 3
(a) No evidence of I131 or i133 above detection limits was observed.
(b) .I131 activity was below the limits of detection.
-------
Table A23 .
Cover Gas Activity (SEFOR)
Report i
IQth Quarterly
9th Quarterly
8th Quarterly
7th Quarterly
6th Quarterly
5th Quarterly
4th Quarterly
The cover gas monitor was in service during the quarter and indicated no anomalous
gas activity. Cover gas samples were obtained at monthly intervals before and
after the Subprompt Test Series and each Sufaearprompt Transient Test. No signifi-
cant increase in the concentration of the f i ssion products in the cover gas was
observed. Examination of this data continues to indicate good correlation with
cover gas samples obtained since December, 1970.
The cover gas monitor was in service during the quarter and indicated no anomalous
fission gas activity. Ten cover gas samples were obtained between May 6, 1971
and May 8, 1971; six in June, and one in July to quantitatively measure the
isotppic constituents. These samples consistent of routine monthly cover gas
analysis, special experiments to further refine sampling and identification techni-
ques, and pre- and post-FKED transient samples. No significant increase in the
concentration of the fission products in the cover gas was observed. Preliminary
examination of these data indicate good correlation with other cover gas sanples
obtained since December, 1970.
The cover gas monitor was in service during the quarter, and indicated no anomalous
fission gas activity. Cover gas samples were obtained to quantitatively measure
the isotopic constituents. These samples consisted of routine monthly cover gas
analyses, special experiments to further refine sampling and identification techni-
ques, and pre- and post-FKED transient samples. These results indicate no fission
gas levels other than those normally anticipated from tramp uranium and/or pin
hole cladding penetrations.
The cover gas monitor was in service continuously during the quarter. Cover gas
samples were obtained to quantitatively measure the isotopic constituents. Two
samples were obtained in November, 1970, five during January, 1971. A special
series of measurements were also conducted in December to further refine the
techniques for sampling the low level fission gas (Xe and Kr) conponents. These
results indicated no fission gas levels other than those normally anticipated
from tramp uranium and/or pin hole cladding penetration.
•The cover gas monitor operated during the quarter and data were obtained as a
function of reactor power level. These results were supplemented with a cover
gas sampling technique which used a 50 ml charcoal filter and a millipore filter
in addition to the 400 ml cylinder gas sample which was used on previous gas
samplings. With the edition of these filters the sensitivity of the sampling
technique was increased sufficiently to observe a background level of Ssnon radio-
activity. The magnitude of jtenon corresponded to that anticipated from the low
level "tramp" uranium. On-line spectrometric studies were conducted also which
revealed the presence of neon-23, a short lived (T
from fast neutron activation of Na2S.
RHAPSODIE.
1/2
38 sec) isotope resulting
This isotope has been observed at EBR-II and
Reactor cover gas activity was monitored and samples were taken for spectrographic
analysis. The A-41 activity was near the levels anticipated. No other isotopic
activity was observed.
Reactor cover gas activity was monitored and samples were taken for spectrographic
analysis. The A-41 activity was near the anticipated level for short term operation
at low power levels.
205
-------
ro
o
en
Report No. and Date
Table A24
Integrated Power*and List of Reports (1971) (EBR-II)
Operated Period of Time
ANL-7776, January 1971
ANL-7783, February 1971
ANL-7798, March 1971
ANL-7825, April-May 1971
ANIr-7833, June 1971
ANL-7845, July 1971
AN1>7854, August 1971
ANL-7861, September 1971
ANL-7872, October 1971
ANL-7887, November 1971
ANL-7900, December 1971
March 2, 1971-March 15, 1971
November 1963-March 15, 1971
March 15, 1971-May 15, 1971
November 1963-May 15, 1971
May 15, 1971-July 15, 1971
November 1963-July 15, 1971
July-15, 1971-September 15, 1971
November 1963-September 15, 1971
September 15, 1971-November 15, 1971
November 1963-November 15, 1971
Integrated Power
Not Reported
Not Reported
505
42321
1467
43788
Not Reported
1752
45540
Not Reported
1969
47509
Not Reported
2354
49863
Not Reported
* See footnote on page 7? concerning EBR-II power level.
-------
Table A25
Radionuclide Activity in Primary Sodium (EBR-II)
Report No. and
Date
ANL-7776
January 1971
ro
o
--4
Sampling
Data
9/30/70
10/6/70
10/13/70
10/28/70
11/2/70
11/9/70
11/13/70
11/18/70
11/23/70
11/30/70
12/4/70
12/9/70
12/14/70
12/30/70
1/4/71
22Na
(nCi/q)
48
49
51
51
50
53
53
51
52
54
51
51
50
51
51
foci/?)
11
n
11
12
12
13
11
10
7
8
6
6
3
4
4
-------
A25
Page 2 {continued)
r>o
O
co
Report No. and
Date
ANL-7798
March 1971
Sampling
Date
2/11/71
2/16/71
2/L8/71
2/19/71
2/22/71
2/24/71
3/2/71
(pCi/g)
None detected
None detected
None detected
None detected
None detected
28
12
137Cs
(nCi/g)
5
12
12
10
13
12
21°PO
15.7
-------
Table A25
Page 3 (continued)
ro
o
Report No. and
Date
ANLr-7825
April-May, 1971
Sampling-
Date
3/5/71
3/10/71
3/12/71
3/15/71
3/26/71
4/5/71
4/12/71
4/16/71
4/21/71
4/27/71
4/27/71
4A6/71
(pCi/g)
25
67
150
140
86
47 .
79
140
150
143
130
137Cs 5^Mn
(nCi/g) (pCi/g)
11
12
13
13
13
14
14
18
18
19
18
None
(nCi/g) (nCi/g) (rCi/g) (rCi/g)
n KIA c: in n no IAS
detected
4/23/71
50.6
0.651
5.46
4.81 .506
-------
Table A25
Page 4 (continued)
Report No. and Sampling 131j 137Cs
Date Date (pGi/g) (nCi/g)
ANL-7833 4/12/71 79 14
June1971 4/16/71 140 18
4/21/71 150 18
4/27/71 136 18
^ 5/5/71 110 17
5/14/71 58 17
5/19/71 25 16
5/24/71 66 17
5/28/71 72 17
: 6/1/71 147
6/4/71 97
6/7/71 310 17
6/8/71 463 ~21
-------
ISJ
Table A25
Page 5 (continued)
Report No.
and Date
ANL-7845
July 1971
Sampling 3H
Date (nCi/g)
5/5/71
5/14/71
5/19/71
5/24/71
5/28/71
6/1/71
6/4/71
6/7/71
6/8/71
6/11/71
6/14/71
6/18/71
6/23/71
6/28/71
4/29/71 42
5/24/71 63
5/13/71
6/18/71
7/2/71
(pCi/g)
110
58
25
66
72
147
97
31
463
294
196
113
82
68
foci/q)
17
17
16
17
17
18
17
20
22
23
21
22
21
23
(PCi/g)
-1 * 3mln
(nCi/q)
117Sn
(nCi/g)
125Sb
(nCi/g)
28.9 .651
72.3 .578
61.4 .687
5.96
5.78
6.47
4.34
3.61
5.06
.542
.108
.867
-------
ro
Table A25
Page 6 (continued)
Report No.
and Date
September
1971
Sampling 3H 22^ I3lj
Date (iCi/g) (nCi/g) (pCl/g)
1/6/70
2/5/70
3/19/70
6/16/70
7/31/70
8/21/70
9/21/70
11/24/70
1/8/71
2/16/71
4/23/71
5/13/71
6/18/71
7/2/71
7/9/71 51 42
7/14/71 52 50
7/16/71 34
7/16/71 30
7/19/71 51 49
(rCi/g) (pCVg)
28
59
24
53
84
19
49
None Detected
None Detected
None Detected
44
25
63
53
19
21
20
(rCi/g)
.50
.41
.47
.53
.59
.59
.47
.69
.59
.53
.56
.57
.50
.59
H3Sn_ll3mIn
4.4
5.0
5.6
3.8
7.2
6.6
4.7
7.2
5.7
4.4
4.7
5.2
5.0
5.6
117Sn
(nCVg)
4.1
4.1
5.6
5.3
4.4
4.4
2.2
3.8
0.38
0.07
4.2
3.8
3.0
4.4
125Sb
(nCi/g)
.59
.66
.23
.10
.50 i
.17
.17
.20
None Detected
.13
.44
.48
.094
.75
continued on next page.
-------
Table A25
Page 7 (continued)
ro
CO
Report No.
and Date
ANL-7861
Septentoer
1971
Sampling 3H
Date (nCi/g)
7/23/71
7/28/71
8/3/71
8/6/71
8/19/71
8/20/71 20
8/27/71
22Na
(nCi/g)
53
53
53
51
53
(PCi/g)
63
76
59
49
41
137Cs
(nCi/g)
20
20
20
21
20
(pCi/g) (nCi/g>
(nCVg)
117Sn
(rCi/g)
18
.59
6.0
4.7
.20
-------
Table A25
Page 8 (cx>ntinued)
Report No.
and Date
ANIr-7887
November
1971
Sampling 3H
Date (nCi/g)
9/1/71
9/8/71
9/13/71
9/24/71
9/29/71 46
10/4/71
10/6/71
10/18/71
10/20/71
10/21/71
10/27/71
10/29/71
(nCi/g)
51
52
51
51
51
53
53
53
54
131!
(pCi/g)
43
53
45
65
63
35
47
66
65
(nCi/g)
21
21
21
20
20
22
22
21
21
(pCVg) (rCi/g)
28
.59
i 13mm
(rCi/g)
6.1
117Sn
(rCi/g)
4.2
125Sb
(nCi/g)
.35
.59
5.8
3.6
.28
-------
Table A26
EBR-II Radionuclide Activity
in Secondary Sodium (1971)
Concentration (nCi/g)
Report No. and Date
ANL-7845
July 1971
ANI>7861
September 1971
ANL-7887
November 1971
Sampling 3H
Date (nCi/g)
4/29/71 6.7
5/24/71 9.2
6/23/71 2.9
7/16/71 0.83
7/16/71 2.4
7/16/71 3.6
8/2/71 0.98
8/2/71 0.73
9/10/71
9/29/71 1.2
10/27/71
11/30/68
4/5/68
5/20/68
6/12/68
9/4/68
12/20/68
3/26/69
4/23/69
5/19/69
6/10/69
7/10/69
8/19/69
9/26/69
10/17/69
1/29/70
3/10/70
6/22/70
2"Na
(nCi/g)
30
41
6.5
31
32
32
30
25
10
6.5
26
27
22
22
24
9.3
11
30
28
continued.
215
-------
Table A26
EBR-II Radionuclide Activity
in Secondary Sodium (1971)
(continued)
Concentration (nCi/g)
Sanpling 3H 21*Na
Report No. and Date Date (nCi/g) (nCi/g)
ANL-7887 7/15/70 24
November.1971 /1Q 33
(oontinuecl)
10/6/70 34
4/21/71 16
6/30/71 30
7/29/71 36
216
-------
Table A27
Gamma Activity in EBR-II Cover Gas
due to Tramp Souxe
(data taken Sept. 7-8, 1971)
Absolute Activity
Repprt No. and Date Nuclide yCi/ml
ANL-7872 133Xe
October 1971
135Xe
85mKr
88Kr
87Kr
138Xe
135mxe
23Ne
4.34 x 10"4
2.05 x 10"3
3.07 x 10"4
5.0 x 10"4
2.48 x 10"4
4.22 x 10"5
2.1 x 10"5
2.5 x 10"3
(a) Normalized to grab-sample determination of 2.05 x 10~ yCi/ml for
135xe during this period. Principal nuclides identified in the
EBR-II coyer gas are: 85mKr, 133Xe, ]35Xe, 85Kr 87Kr 88Kr,
135mxe> T^ ?3^Xe$ 133mXej 137^, 23Ne] 41A/88Rb/138Csf
(ANL-7872, October 1971).
217
-------
Table A28
EBR-II Primary Sodium Composition
(trace metals)
Concentration, ppn
ro
«i
oo
Report No. and Date
ANI/-7776
January 1971
ANL-7798
March 1971
ANL-7825
April-May 1971
AN07845
July 1971
ANL-7861
September 1971
ANL-7887
November 1971
Sanpling Date
10/27/1970
11/24/1970
1/8/1971
2/16/1971
4/23/1971
5/13/1971
7/2/1971
6/18/1971
5/19/1971
7/7/1971
7/14/1971
7/16/1971
8/19/1971
1/4/1971
9/13/1971
10/18/1971
10/20/1971
Ag
0
0
0
0
0
0
0
0
0
0
0
0
.06
.05
.05
.05
.05
.07
.13
.045
.1
.06
.06
.05
Al
<0.6
<0.6
<0.6
<0.6
<0.6
<0.6
<0.6
0.027
<0.6
<0.6
<0.6
Bi
1.9
1.9
0.9
1.3
3
3.2
1.1
1.4
VL.O
2.0
1.5
1.1
1.8
1.6
Ca
<0.02
<0.02
<0.02
<0.02
<0.02
<0.02
<0.01
<0.001
<0.01
<0.01
<0.01
Cd Co
<0.02
<0.02
<0
<0
<0
<0
<0
0.02
<0
0.03
<0
<0
<0
0.08
.02
.02
.02
.02
.02
.002
.002
.02
.02
Cr
<0.02
<0.02
<0.02
<0.02
<0.02
0.02
<0.02
0.063
<0.02
0.07
<0.02
Cu
<0.02
<0.02
<0.02
<0.02
0.03
0.04
0.03
0.02
<0.02
<0.02
<0.02
Fe
0.09
0.1
0.28
0.08
0.23
0.97
0.45
0.55
0.25
0.18
0.48
continued
-------
Continued
Table A28
EBR-II Primary Sodium Composition
{trace metals)
Concentration, ppm
Report tto. and Date
ANL-7776
January 1971
ANL-7798
March 1971
AM>7825
April-May 1971
ANL-7845
July 1971
ANL-7861
Septenter 1971
ANL-7887
Nbvenfcer 1971
Sanpling Date
10/27/1970
11/24/1970
1/8/1971
2/16/1971
4/23/1971
5/13/1971
7/2/1971
6/18/1971
5/19/1971
f/7/1971
7/14/1971
7/16/1971
8/19/1971
1/4/1971
9/13/1971
10/18/1971
10/20/1971
In
<0.06
<0.06
<0,06
<0.06
<0.06
<0.06
<0.06
0.008
<0.06
<0.06
<0.06
K Mg
0
0
0
<0
0
0
0
165
0
0
0
0
.009
.01
.005
.005
.009
.013
.018
.007
.023
.007
.015
Mn
<0.005
0.009
<0.005
<0.005
<0.005
0.009
0.005
0.015
<0.005
<0.005
<0.005
M3
<0.07
<0.07
<0.07
<0.07
<0.07
<0.07
<0.07
<0.01
<0.07
0.07
0.07
Ni
<0.04
<0.04
<0.04
<0.04
<0.04
<0.04 '
<0.04
0.056
<0.04
<0.04
<0.04
Pb
11.3
10.6
2.9
7.11
9.8
5
1.8
8.27
•^•9.0
11.0
9.1
11.0
11.0
12
Sn
24
23
24
28
23
26
25
28
26
25
24
.2
.1
.6
.8
.5
.9
.5
.0
.7
.0
.0
Zn
<0.06
<0.06
<0.06
-------
Table A29
Primary Sodivm Ccrposition
(Non-metals)
Concentration, ppn
Report No. and Date
ANL-7776
January 1971
ANL-7798
March 1971
ANL-7825
April-May 1971
PJ ANL-7845
ro July 1971
ANL-7861
Septenfaer 1971
AM>7887
November 1971
Total Hydride Hydroxide
Sampling Date BC 0 Si H H HN
11/24/1970 1.1 ± 0.2
11/13/1970 <0.1
2/16/1971 2.2 ± 1.0
11/24/1970 0.8 ± 0.06
2/16/1971 1.6 ±0.4
12/23/1971 <0.06
2/18/1971 0.22
4/7/1971 1.3 ± 0.4
4/23/1971 1.8 ± 0.7
2/11/1971 1.6
4/29/1971 0.9
4/29/1971 <0.05
5/14/1971 1.0 ± 0.5 1.6 ± 0.6
6/11/1971 1.1 ± 0.2 1.2 ±0.3
6/28/1971 <0.1
7/16/1971 1.2
8/20/1971 <0.05 3.1 • .. -.
5/13/1971 0.09 0.10
6/11/1971 0.07 0.08
7/19/1971 1.9 ± 0.1 0.8 + 0.2S <0.0 0.10 0.07
8/20/1971 0.16
9/29/1971 <0.05 ' 0.4
10/21/1971 <0.05 0.8
8A9/1971 ' 0.09 ± 0.02
9/13/1971 . 0.16 ± 0.05
9/24/1971 0.07
10/20/1971 1.0 ± 0.5 <0.1
10/21/1971 1/0+0.4
-------
Table A30
EBR-II Secondary Sodium Composition
• (Trace Metals)
Concentration, ppm
Report No. and Date
ANIr-7798
March 1971
ANL-7825 •
April-May 1971
ANL-7845
July 1971
ANL-7861
September 1971
ANL-7887
November 1971
Sampling Date
2/16/1971
2/22/1971
3/10/1971
3/16/1971
4/16/1971
5/13/1971
6/16/1971
7/13/1971
8/10/1971
9/7/1971
10/12/1971
10/21/1971
Ag
0.01
0.048
0.011
0.03
0.06
0.16
0.05
0.08
0.23
0.09
0.02
Al Bi
<0.6 <0:l
0.05 0.019
0.02 0.013
<0.6 <0.1
<0.6 <0.1
<0.6 <0.1
<0.6 <0.1
<0.6 <0.1
<0.06 <0.01
. <0.06 <0.1
<0.06 <0.1
Ca
0.03
0.026
<0.001
0.013
<0.6l
0.04
0.11
0.02..
0.01
<0.01
<0.03
Co
<0.02
<0.002
0.002
<0.02
<0.02
<0.02
<0.02
<0.02
<0.02
<0.02
<0.02
Cr
0.58
0.004
0 ,001.
0.009
0.016
O.OS
0.05
0.03
0.05
0.06
0.02
Cu
<0.02
0.007
0.045
0.015
0.02
0.05
<0.02
0.02
0.04
<0.02
0.04
Fe
3.6
0.11
0.08
0.07
0.08
1.27
0.23
0.21
0.11
0.17
0.20
continued....
-------
Page Intentionally Blank
-------
Table A30
-II Secondary Sodiun Composition
(Trace Metals)
Concentration, ppm
ro
ro
Report No. and Date
ANL-7798 '
March 1971
ANL-7825
April-May 1971
ANL-7845
July 1971
ANL-7861
September 1971
ANL-7887
November 1971
Sanpling Date
2/16/1971
"2/22/1971
3/10/1971
3/16/1971
4/16/1971
5/13/1971
6/16/1971
7/13/1971
8/10/1971
9/7/1971
Ag
0.01
0.048
0.011
0.03
0.06
0.16
0.05
0.08
0.23
0.09
Al Bi
<0.6 <0.1
0.05 0.019
0.02 0.013
<0.6 <0.1
<0.6 <0.1
<0.6 <0.1
<0.6 <0.1
...<:. 6 -C.I
<0.06 <0.01
<0.06 <0.1
Ca
0.03
0.026
<0.001
0.013
<0.01
0.04
0.11
n,02
0.01
<0.01
Co
<0.02
<-0.002
0.002
<0.02
<0.02
<0.02
<0.02
•'?.??
<0.02
<0.02
Cr
0.58
0.004
0.001
0.009
0.016
0.06
0.05
n ^ ->
0.05
0.06
Cu
<0.02
0.007
0.045
0.015
0.02
0.05
<0.02
n.02
0.04
<0.02
Fe
3.6
0.11
0.08
0.07
0.08
1.27
0.23
0.21
0.11
0.17
10/21/1971
0.02
<0.06
<0.03
<0.02
0.02
0.04 0.20
continued
-------
(continned)
Table A30
' EBR-II Secondary Sodium Conposition
(Trace Metals)
Concentration, ppm
ro
ro
ro
Report No. and Date
ANL-7798
March 1971
ANL-7825
April-May 1171
ANL-7845
July 1971
ANI>7861
September 1971
ANL-7887
November 1971
Saitpling Date
2/1S/1971
2/22/1971
3/10/1971
3/16/1971
4/16/1971
5/13/1971
6/16/1971
7/13/1971
8/10/1971
9/7/1971
10/12/1971
10/21/1971
In K
<0.06
<0.01
<0:01
<0.06
<0.06
<0.06
<0.06
<0.06
<0.06
<0.06
158
<0.06
:,g
0.005
0.013
0.007
0.043
0.007
0.038
0.028
0.032
0.038
0.04
0.02
Mn
0.031
0.001
0.002
<0.006
<0.006
0.021
<0.006
0.014
<0.06
<0.006
<0.005
Mo
<0.07
<0.01
<0.0i5
<0.07
<0.07
<0.007
<0.07
<0.07
<0.07
<0.07
0.11
Ni
0.37
0.005
0.008
<0.04
<0.04
0.06
<0.04
<0.04
<0.04
0.08
<0.04
Pb.
0.5
0.67
0.30
0.55
0.20
0.18
0.23
0.77
0.66
0.11
Sn
<0.5
0.06
<0.5
<0.5
<0.5
<0.5
<0.5
<0.5
<0.5
<0.5
-------
A3!
3K-II Secondary Sodiizn Ccnpoeition
(Non-met^.1. -,)
ro
ro
oo
Report No. and Date
ANL-7776
January 1971
ANL-7798
March 1971
ANL-7825
April-May 1971
ANL-7845
July 1971
ANL-7861
September 1971
AW/-7887
NowHtoer 1971
' ^*" iv^j|i Hydride Hydroxide
Sanpling Data B C O 3i H H HN
11A8/1970 0.8 ± 0.4
10/27/1970 <0.06
2/17/1971 ' 3.7-0.5
2/17/1971 1.0 t 0.3
3/3/1S.71 . 1.2 ± 0.2
2/17/1971 0.27
3/3/1971 1.0 t 0.1
4/14/1971 0.4 ! 0.1
4/U/1971 1.4 ±0.4
2/16/1971 1.6
3/15/1971 0.8
4/21/1971 0.9
4/26/1971 <0.05
5/11/1971 1.6 ± 0.2 1.3 ± 0.9
6/7/1971 1.3 ± 0.4 1.0 ± 0.5
6/17 A971 <0.1
5/19A971 0.4
6/24A971 4.4
5/20/1971 <0.05
6A8/1971 <0.05
7/12/1971 <0.05
7/27/1971 1.3
8/11/1971 <0.05
8A2A971 0.5
3/25 A971 ' 0.08 0.10
4/19/1971 , o.07 0.04 '
5/18/1971 0.04 <0. 04
6/29/1971 ' 0.09 0.12 0.03
7/28/1971 0.2 0.06 0.11
9/20A971 2.3
9/21A971 <0.05
10/25/1971 <0.05
10/29/1971 0.7
V26/1971 O.OOi 0.05
9/14/1971 ' 0.12
9/28/1971 0.7 0.04 ± 0.02
10/19/1971 1.2+0.3
-------
APPENDIX B:
Fission Product Data
OOQ
Half lives, decay constants, and fast fission yields for Pu
and 238y are given in Table Bl. The reference is:
M. E. Meek and B. F. Rider, "Compilation of Fission Product
Yields, Vallecitos Nuclear Center, 1972," NEDO-12154
(January 1972).
The results in this reference are output of a computer tabulation at
General Electric which is frequently updated and which weights in
importance the different experimental data assembled.
The decay schemes of the fission products for which the activities
were.-calculated' during the present study are shown in Figure Bl.
These decay schemes ignore nuclides of very short half lives that
occur before the first nuclide shown in each chain in the figure,
i.e. nuclides of such short half life that they do not affect the
results -in Section 4.
224
-------
Table Bl
Half Lives and Fission Yields of Fission Products Listed in Section 4.
NucHde
85lV
85Kr
86Rba
"sr
9°Sr
90y
91Sr
91nv
t\3
, 9 1 v
Zr
SSmNt,
95Nb
99Mo
99mTc
1°3RU
10311^
106Ru
lOSRh
HOltlftg3
HOAga
Hlftg
1 1 SmQja
1 1 5iT£<3
1 1 9n^3n
121mSn
123ItlSn
Sn
125Sb
Accumulated Percent
(Fast Fission)
239pu
0.642
0.142
1.810 x 10~5
1.719
2.089
2.089
2.464
1.503
2.464
4.586
0.060
4.586
5.609
4.936
6.533
6.468
4.517
4.519
7.300 x 10~s
8.250 x 10~6
0.367
4.180 x 10~s
0.006
0.001
0.001
0.043
0.064
0.192
Yield
23 By
0.811
0.173
1.510 x 10"7
3.016
3.282
3.282
4.506
2.748
4.506
5.579
0.073
5.579
6.424
5.653
6.395
6.331
2.835
2,835
2.050 x 10~10 •
2.310 x 10"10
0.103
3.330 x 10~9
0.003
o;ooi
7.860 x 10~6
0.0210
0.068
0.113
Half-Life
Tl/2
4.4h
10.76y
18.66d
50. 8d
28. 9y
64. Oh
9.67h
50.5m
58. 8d
65. 5d
87h
35. Id
66. 6h
6.007h
39. 8d
55m
36 8d
30s
253d
24.6s
7.48d
13. 6y
44. Id •
245d
76y
129d
9.65d
2.73y
Decay Constant
X, sec l
4.375 x 10~5
2.042 x 10~9
4.298 x 10~7
1.579 x 10~7
7.604 x 10~10
3.008 x 10~6
1.991 x 10~5
- 2.287 x 10~"
1.364 x 10~7
1.225 x 10~7
2.213 x.!0~6
2.285 x 10"7
2.890 x 10~6
3.208 x 10~5
2.015 x 10~7
2.100 x. lO'1*
2.180 x'10~8
2.310 x 10~2
3.170 x 10~8
2.817 x 10-2
1.072 x 10~6
1.616 x 10~9
1.819 x 10~7
3.274 x 10"8
2.891 x 10-10
6.218 x 10~8
8.312 x 10~7
8.049 x 10~9
-------
Page Intentionally Blank
-------
Table Bl
Half Lives and Fission Yields of Fission Products Listed in Section 4.
Nucfide " ,
85V
85Kr
86Rba
89Sr
9°Sr
90y
9'ny
91y
95Zr
SSitNb
95Nb
99MO
99mrc
103RU
lOSniRh
106Ru
lOSRh
HOity^a
n°Aga
mAg
nSf
119nfen //
12'msn '
125Sn
125Sb
Accumulated Percent
(Fast Fission)
"^
0.642
0.142
1.810 x IQ--
1.719
2.089
2.089
2.464
1-503
2.464
4.586
0.060
4.586
5.609
4.936
6.533
--v-S.468
4.517
, 4.519
7.300 x 10"s
8. 250 x 10~6
0. 367
/ 4.180 x 10~6
,/ 0. 006
0. 001
0. 001
0.043
0.064
0.192
Yield
238u
0.811
0.173
1.510 x 10~7
3.016
3.282
3.282
4.506
2.748
4.506
5.579
0.073
5.579
6.424
5.653
6.395
6.331
2.835
2.835
2.050 x 10" 1 z
2.310 x 10~10
0.103
3.330 x 10~9
0.003
0,001
7.860 x 10~6
0.0210
0.068
0.113
Half- Life
Tl/2
4.4h
10.76y
18.66d
50. 8d
28. 9y
64. Oh
9.67h
50.5m
58. 8d
65. 5d
87h
35. Id
66. 6h
6.007h
39. 8d
55m
368d
30s
253d
24.6s
7.48d
13. 6y
44. Id
245d
76y
129d..
9.65d
2.73y
Decay Constant
>. , sec"1
4.375 x 10~5
2.042 x 10~s
4.298 x 10~7
1.579 x 10~7
7.604 x 10"10
3.008 x 10~6
1.991 x 1C"5
2.287 x 10""
1.364 x 10~7
1.225 x 10""
2.213 x, 10~6
2.285 x 10~7
2.890 x 10" 6
3.208 x 10"5
2.015 x Kf7
2.1.0 x 1C""
2.180 x 10"s
2.310 x 10"2
3.170 x 10" 8
2.817 x iO~2
1.072 x 10~6
Nl. 616 x 10"9
1.819 x 10" 7
3.274--X 10" e
2; 891 x 10- 10
6.218 x 10~8
8.312 x 10~7
8.049 x 10~9
-------
Table Bl
(continued-page 2)
Half Lives and Fission Yields of Fission Products Listed in Section 4.
Nuclide Accumulated Percent
•(Fast Fission)
739 Pu
125n*Te 0:040
126Sb 0.301
l27Sn 0.209
127Sb 0.500
127mj.e 0.086
127Te 0.501
129Sb 0.687
129ntpe 0.335
129Te 0.801
129I 0.922
££ Ulmpe 0-609
en , 51
131Te 3.548
13!I 4.196
AG \j , •_/ ^ 1>
132Te 5.265
132I 5.366
133I 6.817
133mxe 0.195
133Xe 6.824
13"Csa 1.440 x 10"1*
.. 136Cs 0.151
137Cs 6.625
1 37niBa 6.195
11(0Ba 5.142
ll<0La 5.150
^'La 6.094
i^iCe 6.094
ilt3Ce 4.312
Yield
238u
0-024
0.063
0.049
0.098
0. 190
1.067
0.513
0.222
0-573
0.653
0.367
3.361
3.662
0.024
•' 5.298
5.300
6.471
' 0.181
6.471
1.020 x 10~7
'0.011
'5.952
5.563
5.947
5.947
5.447
5.447
4.533
Half -Life
ij-1/2
58d
12. 4d
2.12H
3.8d
109d
9.3h
4.34H
34. Id
69m
1.6 x 107y
30h
25m
8.065d
11. 96d
78h
2.284h
20. 8h
2.26d
: S.27d
2.06h
13. Od
30. 2y
2.551m ' '
12. 8d
40.23h
3.87h
32.53d
33h
Decay Constant
X, sec"1
1.383 x 10~7
6.468 x 10~7
9.080 x 10~5
2.111 x 10~€
7.359 x 10" b
2.070 x 10~5
4.435 x 10~c
2.352 x 10~7
1.674 x 10"14
1.373 x 10"15
6.417 x 10~6
4.620 x 10" 4
9.945 x 10-7
6.706 x 10" 7
2.468 x 10"e
8.428 x 10~5
9.255 x 10"6
3.549 x 10"6
1.522 x 10~6
9.345 » 10~5
6.170 x 10~7
7.276 x ID'10
4.528 x 10"3
6.266 x 10~7
4.785 x 10~6
4.974 x 10"5
2.466 x 10~7
5.833 x 10~6
-------
ro
IN3
Table Bl
(continued-page 3)
Half Lives and Fission Yields of Fission Products Listed in Section 4
Nuclide Accumulated Percent
(Fast Fission)
»•»
ll*3Pr 4.313
l^Ce 3.604
lui'^r 3.609
ll(7Nd 2.022
llt7Pm 2,023
11<8nfena 3.110 x lO"1*
I51Pm 0.819
151Sni. 0.820
151*Eua 4.488 x KT3
155Eu Q.258
156an 0.125
156Eu 0.154
160Tba 0.005
161Ib 0.014
162Gd
16211*0.,
Yield
238D
4.533
4.543
4.543
2.564
2.565
5.39 x 10~8
0.925
0.925
3.300 x 10~6
0.139
0.074
0.07=
4.580 x 10~6
0.002
4.080 x 10~"
2.210 x 10"1*
Half-Life
Tl/2
13.58d
284 .4h
17.3m
11.06d
2.623y
41. 5d
28. 4h
93y
7.8y
5.0y
9.4h
15. 2d
72. 4d
7.0d
10.4m
2.24h
Decay Constant
X, sec"1
5.906 x 10"7
6.769 x 10~7
6.P7G x 10~"
7.252 x 10~7
8.378 x 10~9
1.933 x 10~7
6.778 x 10~6
2.363 x 10" 10
2.817 x 10~5
4.395 x 10~9
2.048 x 10"5
5.277 x 10~7
1.108 x 10~7
1.146 x 10~6
1.111 x 10~3
8.594 x 10~5
aTheae isertop®s are produced more by activation of other fission products than by accumulated yield from fission.
-------
Figure Bl. Fission Product Decay Schenes Used for
Calculations in Section 4.
85
Kr:
21.76%
78.24%
89
Sr:
89Sr
50.8d
89y
90Sr + 90Y.
90Sr
28.9y
-*.90y
64h
91Y:
39%
58.8d
61%
91my
50.5m
•91Zr
97'7%
65.5d
35. Id
1.3%
228
-------
Figure Bl. Fission Product Decay Schemes
Used for Calculations in Section 4.
(continued)
io3Rh
99%
106RU +
106Ru
368d
l°6Pd
30s
9.65d
79%
2.73y
21%
' 12 5mpe
58d
2.12h
83%
3.8d
127Te
9.3h
17%
109d
131
I:
18%
30h
99.4%
131];
8.065d
82%
0.6%
11.96d
229
-------
133Xe.
137SC
Figure Bl. Fission Product Decay Schemes
Used for Calculations in Section 4.
(continued)
133
97.2%
133
Xe
20.8h\ 5.27d
2.8%
2.26d
6.5%
137Cs , ..137^
30.
2yX
93.5%
2.551m
,133Cg
3.87h
32.53d
.141
Pr
284.4d
17.3m 2.1xl015y
/ If7pm;
. 06d
2.623y 1.07xlOny
151Sm:
151pm.
28.4h
.151Sm
93y
•51Eu
155Eu.
155Eu
5y
230
------- |