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. 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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 -------