r
&
EPA-520/3-75-003
PRELIMINARY ENVIRONMENTAL
ANALYSIS OF A GENERIC FUEL
REPROCESSING FACILITY
U.S. ENVIRONMENTAL PROTECTION AfiKNICY
Office of Radiation Programs
i
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EPA Review Notice
This report has been reviewed by the EPA and
approved for publication. Approval does not
signify that the contents necessarily reflect the
views and policies of the EPA, nor does mention
of trade names or commercial products consti-
tute endorsement, or recommendation for use.
Copies are available on written request, as
supply permits, from:
U.S. Environmental Protection Agency
Office of Radiation Programs
. Washington, D.C. 20460
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EPA-520/3-75-003
OFFICE OF RADIATION PROGRAMS
ENVIRONMENTAL PROTECTION AGENCY
TASK ORDER N0» 68-01=1121
PRELIMINARY
ENVIRONMENTAL ANALYSIS OF A GENERIC FUEL
REPROCESSING FACILITY
Ho Cooperstein, R. C. Erdmann, R0 R, Fullwood
SAI Services
May 1974
SCIENCE APPLICATIONS, LA JOLLA, CALIFORNIA
ALBUQUERQUE ° ANN ARBOR ° ARLINGTON <> BOSTON o CHICAGO ° HUNTSVILLE o LOS AN3ELES
PALO ALTO o ROCKVILLE ° SUNNYVALE ° TUCSON
1651 Old Meadow Road. Suite 620. McLean, Virginia 22101
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FOREWORD
The Office of Radiation Programs carries out a national program
designed to evaluate the exposure of man to ionizing and nonionizing
radiation, and to promote the development of controls necessary to
protect the public health and safety and assure environmental quality.
Office of Radiation Programs technical reports allow comprehensive
and rapid publishing of the results of intramural and contract projects.
The reports are distributed to groups who have known interests in this
type of information such as the Nuclear Regulatory Commission, the Energy
Research § Development Administration, and State radiation control agencies.
These reports are also provided to the National Technical Information Service
in order that they may be readily available to the scientific community and
to the public.
Comments on this analysis as well as any new information would be
welcomed; they may be sent to the Director, Technology Assessment Division
(AW-559)s Office of Radiation Programs, U.S. Environmental Protection Agency,
Washington, D.C. 20460.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs
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PREFACE
power has become a principal option to satisfy the
national need for a clean, safe and reliable energy supply.,
As a result, the generation of light-water-cooled nuclear
power reactors, using enriched uranium fuel, is experiencing
rapid growth<> This increase in nuclear power reactors will
require similar growth in the associated aspects of the fuel
cycle such as mining and milling of uranium ore, production
of nuclear fuel material, manufacture of fuel elements,
shipping, reprocessing of spent fuel elements and waste
disposal activities. To date, the controlled and accidental
releases of relatively small amounts of radioactivity from
©per&feixig nuclear power and reprocessing plants have been
maintained well below specified limits„ However, these
operations may impinge to a greater extent on the environ-
ment as a result of their anticipated growth.
Projections of the Civilian Nuclear Power Program indicate
that the nuclear economy will expand to about 153 gigawatts
by 1980 and to about 735 gigawatts by the year 2000. Economic
analyses by the AEC and by commercial investors have con-
cluded that generation of electric power by nuclear plants
requires reprocessing of spent fuel to recover residual
fissile materials for re-use in new, fuel elements. Approx-
imate total fuel reprocessing rates in metric tons per year
could reach 3500 in 1980 and increase to 15,000 in the year
2000, The total radioactivity due to beta emitters in the
accumulated wastes will increase from 210 megacuries (1970)
to 18,800 megacuries in 1980 and to 209,000 megacuries in
2000.
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The number of reprocessing plants that will be required
is a function of the individual plant designs assumed and
the amount of spent fuel to be processed„ It is assumed,
for this study, that one aqueous separations plant will be
required for each 1500 metric tons of LWR fuel- A plant
of this size should be capable of processing fuel from about
50 power reactors since each typical 1000 MWe LWR will dis-
charge approximately 30 metric tons of fuel each year.
Reprocessing involves destroying the integrity of the spent
fuel elements to permit dissolution and separation of the
fuel from its metal cladding prior to chemical separation
of the useful fissile materials from waste products by some
adaptation of the Purex solvent extraction process, Destrue-
tion of the integrity of the fuel elements which had been
maintained through the cycle in the reactor represents the
main source of radioactivity from the nuclear power industry
which could potentially enter the environment.
The Environmental Protection Agency, whose charter is to
assure protection of the environment by systematic abatement
and control of pollution, sponsored & program through its
Office o:f Radiation Program:: to perform an environmental analysis
study of a generic fuel reprocessing facility in order to
project what effects accidents, in such a facility, of
potential environmental risk significance, may have on the
public health and safety <>
This report discusses the probabilities and consequences of
hypothetical but credible accidents that could occur in the
operation of a generic LWR fuel reprocessing plant which
could have potential environmental impacts.
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In preparing this report technical data was obtained from numerous sources,
nevertheless as might be expected for an analysis of this type, "hard data"
were not available in most cases and the authors were required to assume
"best judgment values." The limitations which this type of approach places
on these data and analysis should be recognized. However, it is our feeling
that the methodology and approach used in this report are applicable to
environmental analyses at fuel reprocessing plants, and that the information
presented despite its limitations is the best available at this time.
ill
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SUMMARY
A generic reprocessing plant for light-water-reactor spent
fuel utilizing the Purex process has been synthesized from
a review of existing plants and those under construction0
This model is used to develop a quantitative description
of the probability of occurrence of a spectrum of poten~
tially credible accidents and resultant radioactive releases
during its operation. The results can be used to determine
the potential impact on the environment„
It is concluded from evaluation of the generic desifa, sys~
terns and components that the most probable off=site release
pathway that could endanger the public health and safety
would be via airborne releases through the ventilation
system in the event of accidents.
The postulated credible accidents considered include explo-
sions during various unit operations involving different
sources of radioactivity, fires, critieality, loss of cool-
ing to the high level liquid waste facility and accidents
which could derive from the occurrence of natural events
such as earthquakes, tornadoes and floods.
Accidents which might occur during normal operation were
emphasized over those that might occur during shutdown or
neptunium batch processing conditions» In addition, no
releases during decommissioning or start-up operations were
analyzed.
If, during an accident large releases of noble gases or
tritium were noted, these were not assessed because they
IV
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would normally be released in reprocessing spent fuelo The
installation of noble gas collectors, to minimize krypton
releases for example, were examined briefly<> While such
systems exist in the laboratory, production-scale systems
are presently not available. Therefore the risk tradeoff
between continual release of such gases and possible acci-
dental release of stored quantities of such gases, after
some period of radioactive decay, was not conducted»
Hypothetical accident probabilities are estimated by fault
tree analysis of the model plant's safety and confinement
systems during operation„ The expected responses to hypo=
thesiged operating transients and postulated accidents,
including natural events, are evaluated- Realistic assump-
tions based upon existing process technology and experience
are used in the evaluations to determine the consequent
radioactive releases „'
If a processing modification is incorporated into the
reprocessing cycle it will be a relatively simple matter to
estimate the likelihood of releases from such a change for
comparison with those documented herein. Thus, a measure
of the level of acceptability of a processing change from a
release or safety viewpoint is evident from this work-
In such a generic analysis, it is not possible to make a de-
tailed account of the operational and design data which
would be pertinent to an in=depth safety analysis„ We have
attempted, where possible, to utilize existing failure data
supplementing this with estimates of how such an incident
during processing might take place- Our considerations
included types of instrumentation used, human behavior and
v
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designs Thus, our results may be suspect in that errors
in design, etc,, may have been overlooked„ Barring these
anomalies, the accuracy of our results may vary within a
factor of 10 or less from the true situation.
The ventilation system (scrubbers, HEPA filters) played a
very prominent role in decreasing the consequences of many
of the accidents releasing nonvolatile species.. If such
releases are determined to be excessive, one should consider
installation of additional release mitigating equipment as
a positive means of removing hazardous airborne substances«,
Since the spent fuel to be reprocessed has substantially.
less decay heat compared to when it is housed in a reactor,
the concerns of decay heat removal at a reprocessing plant,
while real are not severe„ Interruptions of power for
effecting the latter, can be more readily tolerated. Even
so, auxiliary power generation equipment is available to
provide plant power during emergencies <>
Based upon fault tree analysis and consequence calculations,
consequence/likelihood plots are drawn for selected isotopes
including ruthenium, iodine, plutonium, and other representa=
tive actinides and fission products„
By selecting the dominant accidents from these plots and by
applying a simplified meteorological case, the dose in rems
as a function of distance for a number of pertinent isotopes
are plotted utilizing likelihood as a parameter. These data
permit quantification of accident impact on the environment
for a generic reprocessing plant»
VI
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It is recognized by EPA that this report pyesents only an initial
analysis of the potential environmental impact of accidents at a
generic LWR fuel reprocessing facility. There remain further analyses
which could he completed, using this technique, for both a generic
facility and for any specific facility design sited at a particular
location. Extensions of the current study to derive population dose
estimates and predictions of the health effects which could result
from these exposures are examples of two such possible additional
studies.
VII
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TABLE OF CONSENTS
PAGE
PREFACE ....... . . ........... ................. , ............
FOREWORD ................................... .............. i
SUMMARY ............................ ................... ... iv
I o INTRODUCTION .................................... 1
II o DESCRIPTION OF THE GENERIC FUEL REPROCESSING
PLANT AND CONTROLS ........DO.................... 7
1 o Plant Description ....... 0 o 0 0 0 .. 0 . ... 0 . 0 . 0 . 0 . 7
2a Comparison of Potential Hazards in a
Reprocessing Plant and a Nuclear
Plant o oooooc.o.o.o.oooo..oo'.o.oo.oooo..ll
3c Process Description and Radioactive
oooooooooooooooooooooooooooo
4 o Radioactivity Confinement. . 0 0 . 0 0 ............ 28
So Administrative Controls. ........0... •.<..•.... 29
60 Auxiliary Plant Systems and Controls . . . .. <> . . 33
su Ventilation and Of£°Gas System. . . . . . . . « = 33
. bo Water SUPplVo OOO.... .000.0.0000000.00..0 34
S^ & t£
Co On=Site Electrical Power 0 . . . o . . . „ . 0 . . <> o . 37
do Compressed Air Supply. . «, . o 0 .... ..... » . . » 38
e o Steam Supply .... •....<>,,. ........o.oooooo.o 38
IIIo FAULT TREE CONSTRUCTION OF ACCIDENT
SEQUENCES. . . . . o „ o . . . . . o . . . 0 . o o .... .......... „ . o . 40
1 o Saclc^round ..DO...... .0.0.00.. ........ .oo...o 40
2. Fault Tree Symbolic Language. .. o ......... o .. 42
3. Construction of the Fault Tree for
a Nuclear Fuel Reprocessing Plant. .. . . „ . . o .. 44
IV. ACCIDENT ANALYSIS ............................... 57
1 . General ...o.o.ooooooooo.....o..o»o.....o.oo. 57
2. History of Accidents in Reprocessing. ....... 57
3. Postulated Accidents. . . . .............. ...... 59
a. Criticality in a Process Vessel. ........ 60
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TABLE OF CONTENTS (Cont'd)
PAGE
IV.. 3= Postulated Accidents (Confd)
"* O ' £ ^JL Cv O OO O- tt Q O O OOOOOOOOOOOOOOOOOD OQO OOOOOOOOO OO O 4u
1 o Solvent Fires...0........................ 66
2. Ion-Exchange Resin Fires................. 67
c o Explosion...... ........O.......D............. 68
d. Fuel Receiving and Storage Accidents........ 75
e. Waste Storage Accidents„ ......'..„........... 77
f o Natural Phenomena Events................,.<>. ... „.. 79
V. CONSEQUENCES OF ACCIDENTS............................ 82
1 o HAW Concentrator Explosion, ........................ 93
20 Solvent Fire in Plutonium Extraction Cycle...... 87
3. Solvent Fire in the Codeeontamination Cycl©00oo« 39
4. Explosion in the LAW Concentrator................ 92
5. Ion Exchange Resin Fire......................... 95
6» Nuclear Criticality Incident.................... 98
7 o Explosion in the HAF Tank........................ 99
8. Waste Calciner Explosion........................ 101
9. Fuel Receiving and Storage Accident............. 101
RISK ASSESSMENT. ..............................V. ..... JOS
1. Release Likelihood Spectra...................... 105
2. Dose Quantification, .v.-. 0 0. .•......./?. ...o. .<>. .a 0.. us
2>0 Sit@ R@ls.tsQ Ev@nt@.....ooaao................... X24
REF1S1HCES............ ........... .... o.o..... o....... 128
APPENDIX A - Summary Table describing Bagic
Operations, Process Functions and
Chemical Reactions in the Generic
Spent Fuel Reprocessing Plant.......... A=l
APPENDIX B - Fault Trees Used in Risk Assesment,
» o o o o o
APPENDIX C - Descriptions of Accidents Experienced
in the Nuclear Energy Field and Chem-
ical Industry Relating to Anticipated
Credible Events at a Fuel Reprocessing
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lo INTRODUCTION
As of June 1, 1973, there were 33 operating reactors, 56
being built and 80 additional reactors on order. Based on
nuclear power projections of 300,000 MWe by 1985, additional
fuel reprocessing capability is needed to provide for the
recovery of fissile material remaining in spent fuel elements <,
The two existing fuel reprocessing plants, Nuclear Fuel
Services and the Midwest Fuel Recovery Plant, have combined
capacities for reprocessing the spent fuel discharges from
the equivalent of about 35 1000 MWe light water cooled power
reactors (1050 MTU/year). The Barnwell Nuclear Fuel Plant,
under construction, will have a capacity for reprocessing
an additional 1500 MTU/year of spent LWR fuel bringing the
total annual reprocessing capacity to only 2550 MTU/year„
This will be insufficient capacity for the industry within
the balance of this decade thereby requiring the construction
of added capability to ensure smooth, economical and timely
operation ©f the nuclear fuel cycle for the anticipated
energy requirements.,
In the operation of a nuclear power reactor, the buildup of
fission products and the depletion of fissile material
(U=235 and Pu) requires that for maximum utilization of the
fu©l, the reactor operator must periodically replace about
ona-third of the fuel elements and redistribute the remaining^
partially spent fuel elements throughout the reactor core.
The discharged spent fuel elements still contain between
one-third and one-fourth of the U-235 in the fuel prior to
irradiation and part of the fissile Pu that was produced
from U=238o The fuel reprocessing plant recovers the
unused fissile material so that it can be recycled in reactor
reload fuel* Reprocessing also permits separation and removal
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of tbe'-fission products from the fissile material for con-
version into an. acceptable form for long term isolation
from the biosphere .
The sequential process of transforming uranium ore into
usable fuel for nuclear power reactors and the operations
to recover unused values of uranium as well as the plutonium
and other desirable isotopes produced during irradiation in
2
the reactors constitute the "nuclear fuel cycle" „ These
operations are generally performed at separate installations
in various parts of the country, depending for the most part
upon the economics of transportation 0 The specific components
comprising the LWR supporting fuel cycle are shown in Figure
1-1, page 3, and include the following;
a) Mining uranium ore
b) Milling and refining ore to produce uranium
concentrates (U0)
c). Production of uranium hexafluoride (UP>.) from
uranium concentrates to provide feed for isotopic
enrichment
d) Isotopic enrichment of uranium hexafluoride to
attain reactor enrichment requirements using the
gaseous diffusion process
e) Fabrication of nuclear reactor fuel including;
converting UFg to uranium dioxide (U0_) , pelletizing,
encapsulating in rods and assembling fuel elements
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Mining
(a)
ore
Milling
(b)
U3°8
^
UF- Production
6 (0
UFfi (natural)
Fresh Fuel to IWR ^
Irradiated Fuel from LWR
[V
P
•1
Fuel
Fabrication
(e)
Recycle
(<1% U-2
I
Reprocessing
UF6 Enriched
^ (2-4% U-235)
-J . ...
i •
Enrichment
(d)
(\
I/
Radioactive
Waste
Management
(g)
FIGURE 1-1
Nuclear Fuel Cycle - Light Water Reactors
Uranium Dioxide Fueled - No Plutonium Recycle
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f) ^©processing irradiat©d £u©l aad convcag'feiag
to UFg for recycle through the gaseous diffusion
plant for re-enrichment
g) Radioactive waste management of high level and
other than high level wastes, including long-term
storage of wastes
h) Transportation activities associated with moving
materials to and from each of the above operations.,
Uranium milling and refining (benefieiation of raw uranium
ore into UjO^) is usually done a©ar the mines to avoid th©
coast penalty involved in shipping the comparatively low
value ore over long distances . Uranium or© is r©fin©dl into
U30g for shipment in drums to a "conversion" plant wn©re it
is converted into uranium hexafluoride (UF,) „ Uranium in feh®
o
UFg gaseous state is required as feed material for subsequent
"enrichment" in the gaseous diffusion proc@®s0
Uraaium h®3sa£luorid© -• , although rdfisicid as t© total
uranium content, still contain© 1©©© than 1% of th© fis^
eionable isotope U° 2.35 after "conversion". To be suitable
for fabrication into fuel elements for modern power reaetoso,
it must be "enriched" to approximately 23 to 5% U-235
^©pending upon the specific design requirements of individual
reactors )0 Thus, UF- is fed into enrichment plants where
o
the U-235 isotope is upgraded to the required content „
After enrichment, the gaseous UFg is converted into a
metallic oxide (UO2) for fabrication into fuel elements
4
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which, together with control rods, structure and moderating
i, form the nuclear reactor core,,
In the reactor, fuel elements initiate and sustain the
controlled fissioning "chain reaction" which produces vast
quantities of heat necessary to generate electricity via
the steara~powered turbine generators« On the average, one
pound of slightly enriched uranium produces approximately
3<,2 million kilowatt hours (KWH) of electricity,. This
compares with about 202 million pounds of coal required to
generate the equivalent electricity»
During irradiation in the reactor, various fission products
are created which tend to lower reactivity over time through
absorption of neutrons., Excessive accumulation of fission
products and burn-up of U-235 impair the chain reaction and
eventually shut down the reactor., Irradiated fuel, there-
fore, must be replaced periodically after being only partially
consumed. As the fuel is' irradiated in the reactor core,
a
part of the fertile uranium isotope U-238 is converted to
Plutonium, a portion of which undergoes fission thereby
contributing to the reactor's heat output„ The remaining
Plutonium and other fission products stay intact and become
potential byproducts and waste residues»
Economic considerations favor the recovery of the fissile
material remaining in the spent fuel elements„ The net
value of this residual fissile material, after allowing for
the costs associated with reprocessing, waste disposal and
related transportation services, amounts to about $50,000
per metric ton of irradiated fuel. Thus, during each year
a 1500 MTU capacity reprocessing plant is operated at full
capacity, it will reclaim fissile material having a net
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worth of about $75,000,000. Moreover, by recovery of fissile
material, such a plant will conserve natural resources equiv-
alent to about a million and one-half tons of uranium ore
each year .
Reprocessing accomplishes the objectives of;
a» reclaiming the unused uranium and plutonium for
subsequent recycling into replacement fuel
bo extracting valuable isotopes such as neptunium
and separating the waste fission products from the
reusable fissile material
GO concentrating th@ fission products and associated
irradiated wastes to permit safer, less complicated
handling for permanent storage and more economical
storage that will result in a minimal impact on
the environmento
This study is concerned with determining the probability
of accidents associated with a generic 11ght=water reactor
irradiated fuel reprocessing plant and the consequences of
these accidents to the environment„ This analysis will
permit a quantitative risk comparison with other parts of
the nuclear fuel cycle and to other risks accepted by
society. The study is limited to developing the accident
risk envelope for a generic reprocessing plant that could
impact on the environment« Fault tree ^analyses are used
for the accident probability predictions. Only those trees
needed in the development of the risk envelope are evaluated
although a complete set of trees is presented to serve as
an aid for future work in risk assessments of fuel repro-
cessing facilities.
6
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II. DESCRIPTION OF THE GENERIC FUEL REPROCESSING
PLANT AMD CONTROLS
Public health and safety is a principal concern in reprocessing
spent irradiated fuel to a similar extent as for nuclear power
reactorso In fact, the reprocessing plant may be a greater
4
source of radioactivity in effluents than a reactor „ Thus,
adequate safety margins are included in the design of repro-
cessing plants to prevent accidents and to assure that accept^
able protection systems will function reliably to mitigate ,the
consequences of accidents, if they should occur, because of
multiple system failures or noncompliance with procedures
provided to prevent accidents
The function of a fuel reprocessing plant is to recover the
residual fuel materials, uranium and plutonium, in a form
suitable for reuse and to isolate radioactive wastes for
storage and ultimate permanent disposalo The basic elements
of reprocessing are illustrated in Figure II-l,. page 8°
This simplified flow diagram is generally applicable to any
of the process techniques which have been applied for spent
fuel reprocessing. The generic reprocessing technique for
this study is the Purex process, a solvent extraction process,
which has been in use for two decades in this country and
g
is in uae in other countries where spent fuels are processed „
1» Plant Description
The following assumptions are made regarding the generic plant:
o the facility would be sited in conformance with AEC
siting criteria as expressed in Part 100, Title 10
of the U,So Code of Federal Regulations (10 CFR 100) .
o sufficient water supply would be available for plant
operation
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Spent Fuels
From Reactors
Fuel Receiving -
Feed Preparation
Chemical
Separation
V
Uranium
Plutonium
Products
s>-Shipment
Waste
Treatment
FIGURE II-l
Fuel Reprocessing - Generic Flow Diagram
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o it would have a processing rate of 5 metric tons
of heavy metal (uranium and plutonium) per day of
©pent fuel from light-water power reactors ,
o the major facilities on the site would bes
.(
1. fuel receiving and storage facility
2o main processing building housing the repro-
cessing, product storage and waste solidifica-
tion equipment
3o radioactive-area-ventilation-air filtration
and discharge system
-So high level radioactive liquid waste stprage
5= offices ""
60 warehouse and shops
7o steam-generating plant
Sc cooling towers
9o a retention basin
10. product conversion facilities
Ml processing equipment and systems for processing irradiated
fuel elements, except for waste tank facilities, will be
housed in the process, fuel receiving, and storage station
building by the nature of the radiochemical operation„ Because
of th
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FIGURE II-2
ARC PROCESS BUILDING
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of its larger capacity, which is used for the purpose of
©stimating the fractional releases of radioactive materials
for the accidents considered. This choice is based on
the belief that future plants will be designed for such a
capacity to effect low unit processing costs. High capacity,
small volume equipment will be used to minimize plant inventory
of both reactor fuel and process reagents thereby ensuring a
greater degree of overall safety and economy„ Future plants
may have different characteristics from those used in this
study; however, it is expected that the derived quantities
of radionuclides that could be released in potential credible
accidents will remain unchanged or decrease as a result of
advancing technology
20 Comparison of Potential Hazards in a Reprocessing
Plant and a Nuclear Power Plant
Performance criteria for engineered safety in reprocessing
plants are based upon those proposed for nuclear reactors
Q
(Appendix A, 10 CFR 50? Code of Federal Regulations )
although the function and design of reprocessing plants
are significantly different. Potential hazards in a
separations facility differ considerably from those antici-
pated in power reactors du© to the specific conditions
found in the reprocessing planto Examples of these dif=
ferences follow;
o The potential for a nuclear critieality is very low in a
reprocessing plant albeit fissionable isotopes are present
in quantity and are separated and purified in the course
of operationso The use of soluble and fixed poisons,
favorable geometry, concentration control and mass control
mitigates the possibility of nuclear critieality where a
nuclear chain reaction could take place.
11
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o Fuel reprocessing operations are generally of such a
nature that the rate of approach of critical parameters
(concentrations temperature, acidity, etc.) to prescribed
upper limits will be relatively slow compared to power
reactorso Although reliable instrumentation and control
provisions are required, very rapid response is not
necessaryo The effect of exceeding prescribed operating
limits does not usually present an immediate hazard.
o The cladding, which serves as the primary barrisr fee
fission product escape from fuel in a reactor, must b@
breached in the reprocessing operation to recover th©
fissile materials. The potential hazard from having
mobile radioactive materials in plant process systems?
however, is relatively low sine© the systems do not
contain large amounts of stored energy (like power
reactor primary coolant systems) which could provide
the driving force of radioactivity dispersal.
o In some instances, plant process streams will be corroaiv®,
System failure due to corrosion with subsequent radio~
nuclide releases may not be severe for the streams are
doubly contained for leak tightness and operate at low
pressures.
o Flammable and/or chemically reactive material® are used
in fuel reprocessing operationso Well developed t<§eh=
nologies, however, are used for assuring that potential
hazards from use of such materials are appropriately
controlled.
o The potential hazards from loss of plant cooling capa=
bility are low due to the lower stored energy levels of
the limited fissile inventory in process which has
12
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already undergone extended decay times. Emergency power
supply availability will enhance safety assurance by
providing positive off-gas release control and continuity
of process condition surveillance and safe shutdown
conditions. Continuous cooling is not critical although
necessary for high level wastes which are stored on an
interim basis. Cooling is used to avoid boiling in the
high level liquid wastes and/or overheating of the con-
tainment vessels. Overheating could compromise the
integrity of the storage tanks resulting in uncontrolled
releases of waste fission products to the environment.
o Fuel reprocessing operations are performed at low
temperatures = limited to normal boiling points at
atmospheric pressure - and at low pressures = limited
mabatsaospheric pressure maintained in the process
o Fission product wastes produced during irradiation of
power reactor fuel are separated in the reprocessing
plast and large quantities of these materials must be
stewed and controlled. Process operations are performed
to £©due@ the volumef) of solutions of such wastes to more
manageable volumes which will not be released to local
water courses. Alternatively, the wastes will be con-
verted to low mobility forms (solids) for safe on=site
retention until transferred to authorized permanent
disposal facilities,
o Off-gas treatment processes and controlled effluent
releases are required to assure that gaseous products
which are not amenable to immobilization are not released
from the plant at rates exceeding prescribed limits.
13
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The Purex process basically involves solutions and as
such makes the recovery of tritium especially difficult „
3» Process Description
The overall process function is to recover the contained
uranium and plutonium from spent fuel assemblies „ The Purex
solvent extraction process will be used to separate the
latter from fission products „ To accomplish the overall
function, several processes must be used . They are described
in the following and are illustrated by the simplified block
diagram in Figure 11=3, page 15, which is the particular
flow diagram for the BNFP facility0 For comparative purposes,
the block diagrams for the MFRP and NFS processes are illus-
trated in Figures 11=4 and 11=5, page® 15 and 17 . Principal
similarities and differences among the commercial plants and
the assumed model are listed in Tables XI~1 and 11=2, page®
18 and 19 c
o Fuel Receipt and Storage „ Irradiated fuel assemblies
arrive at the reprocessing plant in shielded casks „
These are monitored for outside contamination, cleaned,
removed from the carriers and submerged in a pool of
water for unloading the fuel assemblies o The cask is
opened, the fuel assemblies removed and placed in storage
canisters o The canisters are moved to a fuel storage
pool where they are held until the fuel is scheduled for
reprocessing,,
o Mechanical Disassembly^ When scheduled for reprocessing,
the fuel assembly is remotely transferred to a mechanical
facility where it is sheared, as rods or as fuel bundles
into short lengths to expose the fuel to the dissolvent »
,
14
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» SQLIQOF11CATOOKI
1 FACILITY
FE0EKAIL BEFO80TOC3V
ATMOSPHERE
OFF-CAS
TREATfiflEMT
MICH-LEVEL CHASTE
STORAGE TANKS
MITRIC AOO>
RECOVERY
PRODUCT
a- u T
I PLUTONIUM
g PRODUCT
• FACILITY
in
CHOPPING
DISSOLUTION
SOLVEWT
EXTRACTION
PARTITION
STRIPPING
EXTRACTION
CVCO.E
IPu
WITRATE
Pu
PURIFICATION
CYCLE
REACTOR
FUEL
RECEIVING
STORAGE
HULLS TO
INTERIM
STORAGE
(BURIAL)
OJ
EXTRACTION
CYCLE
U
SILICA GEL
PURIFICATION
(U
WOTRATE
r-
i
i
>J UF-FACILITY }
1
I.™ I
PRODUCT
FIGURE II-3
0
Simplified Flow Diagram of BNFP Process
-------
oo
HULLS
SHEARING AND
COBE LEACHING
SOLVENT EXTRACTION
ANIOW 6HCHANOI
PROTECTED
STORAGE Of
PROCESS
WASTES
USEFUL PQODUCTS PACKAGED
AND SHIPPED TO FUEL
MANUFACTURERS, ETC.
9GANIUW CONCINfBATlOW
AWB CALCINATION
UBAMIUM FLUORINATION
AND PUBLICATION
DRY
WASTE
WASTE STOBAOE
a-
MFRP
General Electrlc's proposed Aqua-
fluar process makes.use of aqueous
and fluoride volatility fuel recovery
tachnolopics. The process uses
well-demonstrated fuel recovery
techniques Including fuel shearing
and leaching, solvent extract Ion,
union encliango, calcination, fluor-
liuillnn, and uranium Miwflur/rlrln
dlBtlllatlon. Aiiloii enchanue lo
used in the recovery and purifica-
tion of plutonium and neptunium.
Uranium will be converted to
the volatile uranium heulluorlde
(Ur0) and purified to make It
Dutiable for toll enrichment. Tito
hLgh level liquid waoteo will to
converted to a olid form and stored
temporarily on-oite in a watertUled
concrete storage baain.
16
-------
FIGURE II-5
NUCLEAR FUEL SERVICES, INC., SPENT FUEL PROCESS1
!FUEL
STORAGE
/ MECHANICAL
I DISASSEMBLY
( NITRIC ACID
V DISSOLUTION
SOLVENT
EXTRACTION
PROCESS DESCRIPTION
Spent nuclear furl is transported from tlit* rcai lor
site to (he ri'priHVfeiuK plant in shielded r.i-.lt> The
rusk* are unloaded .underwater nnd the fuel assem-
blies are stored prior to reprocessing.
The fuel is transferred from the SIOI.IKI- pool to a
mechanical processing cell where end-fitting hard-
ware is removed and the fuel is sheared into small
pieces. The pieces an- collected in a canister, which
is placed in a dissolvrr, where the fuel values and
fission products nre dissolved in nitric acid. (The
insoluble cladding materials— hulls — are monitored
to establish completeness of dissolution of fuel values
and buried in the plant1* radioactive waste burial
ground, i The disiuilvcr solution is subjected to solv-
ent extraction with a mixture of tnbutyl phosphate
and a diluent to separate the majority of the fission
products from the contained uranium and plu-
tonium values. The pluloniurn and uranium are
then separated by solvent extraction techniques
and the resulting plutonium and uranium streams
are subjected to.solvent extraction purification to
remove the remaining fission products.
The uranium product stream is concentrated by
evaporation and subjected to a final decontamina-
tion with silica gel. resulting in a uranyl nitrate solu-
tion product. The plulonium product stream is sub-
jected to an ion exchange treatment to rlTecl both
the concentration and further dttCOntnmiruiUOfl ot
the plulonium—yielding the final plulonium nitrate
solution product.
Liquid waste i.s subjected to evaporation, neutral-
ized and stored in underground liquid waste tanks.
FISSION PRODUCTS URANIUM & PLUTONIUM
f S<
SOLVENT
V EXTRACTION
URANIUM
PLUTONIUM
17
-------
TABLE II-l
PRINCIPAL SIMILARITIES OF THE MODEL SEPARATIONS FACILITY
COMPARED TO OTHER COMMERCIAL REPROCESSING FACILITIES
Model
Comparison Item
BNFP
NFS
MFR1
Fuel Unloading and
Storage
Headend Process
Stored Fuel Criti-
cality Safety
Fuel Chopping
Fuel Dissolution
Material
Fuel Dissolution
Technique
Fuel Dissolution
Equipment
Solvent Cleanup
Final Exhaust
Filters
Underwater
Chop-Leach
Spacing
Mechanical Shear
Nitric Acid
Semicontinuous
Baskets in
Dissolvers
Alternate Contact
with Sodium
Carbonate and
Nitric Acid
Solutions
Roughing and
Underwater
Chop-Leach
•«f
Spacing
Mechanical Shear
Nitric Acid
Semicontinuous
Baskets in
Dissolvers
Alternate .Contact
with Sodium
Carbonate and
Nitric Acid
Solutions
Roughing and
Underwater
Chop-Leach
Spacing
Mechanical Shear
Nitric Acid
Batch
Baskets in
Dissolvers
Alternate Contact
with Sodium
Carbqnate and
Nitric"Acid
Solutions
Deep Fiberglass
and HEPA
Underwater
Chop-Leach
Spacing
Mechanical Shear
Nitric Acid
Semicontinuous
(No Similarity)
Leaching trough
Alternate Contact
with Sodium
Carbonate and
Nitric Acid
Solutions
(No Similarity)
Sand Filter
-------
TABLE II-2
PRINCIPAL.DIFFERENCES OF THE MODEL SEPARATIONS FACILITY
COMPARED TO OTHER COMMERCIAL REPROCESSING PLANTS
Comparison Item
Model
BNFP
Separations
Facility
NFS
MFRP
^Location
Design Capacity
'Shear/, ng
Criticality Control
During Dissolution
Fuel Dissolution
Technique
Fuel Dissolution
Equipment
HA Contactor
Partitioning
Interiir High-Level
waste storage Form
Iodine Removal from
Process Off-gas
Process Vent Filters
*
Final Exhaust Filters
Peed Clarification
Tritium Disposal
Uranium Product Form
Liquid Effluent
Compliance with
10 CFR 100
5 MTU/day
Entire.fuel
elements including
end fittings
Soluble poison
Semicontinuous
Baskets in
Dissolvers
Centrifugal
Contactor
lor. Exchange
Acidic solution
(1-5 molar)
Mercuric Nitrate
Iodine Scrubbers
plus Iodine Silver
Zeolite Adsorption
Bed?
Roughing and HEPA
Roughing and HEPA
Centrifuge
As vapor up stack
Hexafluoride
Noncontamina ted
South Carolina
5 MTU/day
Entire fuel elements
including end
fittings
Soluble t-oison
Semicontinuous
Baskets in
Dissolvers
Centrifugal
Contactor
Electropulse column
Acidic solution
(1-5 molar)
Mercuric Nitrate
Iodine Scrubbers
plus Iodine Silver
Zeolite Adsorption
Beds
Roughing and HEPA
Roughing and HEPA
Centrifuge
As vapor up stack
Nitrate solution
Noncontaminated
New York
3 MTU/day
End' fittings may be
removed before
cheering
Geometric
limitationo
Batch
Baskets in
Ditt solvers
Pulse Column
Solvent extraction
with chemical
valence adjustmen"
Acidic solution
Mercuric nitrate
scrubbers plus sil
ver zeolite
adsorption beds.
Multiple HEPA
Deep Bed Fiberolappsand Filter
plus HEl'A
None
Aa water to creek
Hexafluoride
Contaminated
Illinois
1 MTU/day
Pins removed from
fuel elements before
ohoaring pine only
Gaomatric
liraitationo
iomicontinuouo
Vibratory Lo&chor
Tray
Pulse Column
Ion exchange
Calcined solid
Sodium Hydroxide
Scrubbers. Heated
Silver Zeolite
Packed Fiberglass
Filter
None
As vapor up stack
Hexafluoriae
Noncontaminated
19
-------
The fuel segments fall into ©r are fed into & dissolver
vesselo
Fuel Dissolution,, The segmented fuel containing the
Plutonium, neptunium and fission products formed during
irradiation, as well as unspent uranium is dissolved out
of the cladding hulls with nitric acid to form the feed
for subsequent remote liquid-liquid solvent extraction
steps,, After dissolution, the undissolved cladding
hulls, made of zirconium, zirconium alloys or stainless
steel, are separated from the solution, ringed, monitored
for residual fissile material and transferred to ® pro-
tected interim vault storage arefu Gases., generated during
the dissolutions are channeled to an off-gas treatment
system,, This system contains decontaminating units such
as condensers, scrubbers, chemical traps for iodine
removal and particulate filters which remove
radioactive gases and particulates other than inert gass
e^g0, Kr-85 and tritium, to level® below allowable
release limits before being exhausted to the atmosphere.
Nitrogen oxides formed during the dissolutions are also
retained by the ventilation treatment system to minimise
their release to the atmosphere.
o Solvent Extraction. The chemically adjusted aqu@ous
feed solution is then subjected to a Purex-type extrac-
tion » It is contacted countercurrently in a centrifugal
contactor with an organic solution of tributyl phosphate
(TBP) dissolved in normal paraffin hydrocarbon diluent
(dodecane)o The organic solution preferentially extracts
the tetravalent plutoniuxn and hexavalent uranium, leaving
about 95% of the fission products in the aqueous solution.,
The organic solution from the extraction passes through
20
-------
a scrubbing column where it is washed with additional
nitric acid solution„ This step removes about 96% of
the residual extracted fission products from the organic
product solutiono The wash solution is recycled back
to the centrifugal contactor. The aqueous solution
leaving the codecontamination cycle contains about 99.8%
of th© fission products from the initially dissolved
solutiono It is routed to a high level waste treatment
system where it is concentrated for protected interim
liquid waste storage and/or calcined to an immobile
solid form, loaded into specially designed containers
sad transferred to protected storage facilities for
ultimate authorized dff=site disposal„ A simplified
block diagram illustrating the solvent extraction cycle
is shown in Figure 11=6, page 22 o
o Product Separation, Anion exchange or electrochemical
reduction is used to partition plutonium and uranium
into separate streams following the codecontamination
step» In the former operation, solution from the solvent
extraction system concentrator is fed through a cooler
to a series of semicontinuous ion exchange contactors,
in an ion=exchange cell, where the plutonium is sorbed
on the regin as anionic Pu(N03)~= and subsequently
removed as a nitrate solution for concentration and
loadout= Alternatively, the organic solution from the
codecontamination step can be passed through a partitioning
column where tetravalent plutonium is electrochemically
reduced to the less extractable trivalent stateo The
reduced plutonium is then stripped into an aqueous nitric
acid solution containing hydrazine as a holding reductant.
The organic uranium containing solution is then stripped
into acidified water„ Electrochemical reduction and
solvent extractions eliminate the need for chemical
21
-------
eo
to
Scrub Solution
(Feed Solution
IT, Pt8B Fission'
Products
Organic
Extractant
Scrub
Stages
Extraction
ss
Us, Pu Products
in Organic Phase
Waste Solution
-?.-*- Salting Agents and
Fission Products
A
Back Extractant Solution
-*=~ Organic To Treatment
Back-
Extraction
Stages
-&~US Pu Products in
Aqueous Phase
FIGURE II°(S
SOLVENT EXTRACTION CYCLE
-------
additions for valence adjustment and the use of ion-
exchange resin columns„ The quantities of waste to be
subsequently handled and disposed of, are also reduced.
In this analysis, the former unit operation is reviewed
for its accident potential„
o Uranium Purification and Recovery, The aqueous uranium
strip solution is concentrated and its acidity is
adjusted prior to resubjeeting it to another partitioning
cycle and filtration through a silica gel bed for final
concentration loadout as uranyl nitrate solution., The
operations remove additional residual fission products
and plutonium further ensuring that the uranium solution
meets product specifications for reuse in the fuel cycle0
Th© nitrate solution is ultimately shipped as such or
converted to UF, in an associated facility prior to ship-
ment for reuse in the fuel cycle.
o Plutonium Purification and Recovery. Plutonium in the
aqueous stream may be cyclically converted to anionic
Pu(HO3)g' and sorbed on strong base anion exchange resin
while the associated uranium and fission products are
washed out in the raffinate. The sorbed plutonium from
th© last purification cycle is eluted as nitrate solution,
concentrated and stored pending conversion to plutonium
dioxide for use in plutonium recycle or breeder reaction
fuel elements. Alternatively, the aqueous plutonium
stream leaving the partitioning column may be reoxidized
to the extractable tetravalent state and subjected to
additional extraction-scrubbing sequences as described
above to further decontaminate the product solution from
fission products. The final plutonium nitrate solution,
after concentration, is analyzed and stored in geometrically
favorable tanks until it is converted in the Plutonium
23
-------
Product Facility to the solid oxide form for storage
and/or off-site shipment„
Organic Solvent Systems. The organic solvent waste
streams from the decontamination and partition cycles
are washed successively with dilute aqueous solutions
of sodium carbonate, followed by nitric acid and
neutralized by sodium carbonate to remove organic
degradation products by extraction or precipitation0
Precipitated solids are removed by filtration. A®
required, fresh TBP or dilutent is added to maintain the
TBP concentration and the total solvent inventory.
Liquid Waste Treating and Storage„ The aqueous raffinat©
streams from the plutonium and uranium cycles are
reprocessed for residual fissile material content by
extraction into TBP organic solution which is recycled
back to the decontamination cycle for recovery. Th©
aqueous raffinate, essentially depleted of radioactive
materials, is then concentrated in the low level waste
concentratoro The radioactive waste streams from all
the solvent extraction cycles are concentrated in th©
high=level or low-level waste concentrators to recover
nitric acid and water for reuse in the process while
reducing the waste volumes to be stored in appropriate
waste facilities; miscellaneous waste streams containing
salts, minimal fission products and no appreciable fissile
material are acidified and concentrated. The concentrator
bottoms are stored in appropriate waste tanks for ultimate
disposal at a federal repository and the condensed
heads are vaporized to the stack.
24
-------
o grocesii Off°Gas Treatment. In order to maintain near-
atmospheric internal working pressures, with very few
exceptions, all of the process equipment pieces -
vessels, extractors, condensers, etc. - are vented to
on© or more process vent systems. The vent gases are
treated by scrubbing with circulating mercuric nitrate
solution to remove radioactive iodine, then treated in
an absorber to convert nitrogen oxides to nitric acid
suitable for recycling. The dissolver off-gas and
v©©s©l off~ga& streams are combined and passed through
s ©©eond decontamination train which includes an iodine
scrubber unit, iodine adsorber beds and a series of high
efficiency filter banks before being released to the
stack. Thus, the relatively small amounts of radio=
activity including most of the remaining radioiodine are
removed from the vent gases prior to discharge, via a
©tack, to the atmosphere <, The vent gases will probably
certain most of the tritium as a result of operations
in the acid recovery system. Virtually all of the
tritium from the original fuel follows the water phase,
a® tritiated water, to the waste system and thence to
th© process condensates derived during evaporation-acid
r©eov©ry operations. In early process designs, this
material was discharged to the environment following
©veporative and chemical treatments, for removal of
entrained non-volatile activity. Current design
approaches recycle process condensates in the process,
as process water, with evaporation of a small-volume
remainder of the condensate to discharge tritium to the
atmosphere via the off-gas system and the stack.
25
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Th@3 gaseous wastes from a reprocessing plant include
a ventilation air, which is indirectly exposed to the
process, besides- the process off-gases which are
directly exposed to the process and which contain some
volatile radionuclides . Both systems are shown by
Figure 11=7, page 2.7 „ In this block diagram, the
process building is- shown schematically as the large
block on the left,, The building operating spaces are
divided into four zones of ventilation, control, with
corresponding increasing levels of potential eontamina?
tion by radioactivity, and increasing degrees of access
control requirements. As plotted on the diagram, th©:
building ventilation system is engineered to maintain
pressure differentials between zones such that the air
flow is always toward the zone of greater potential
contamination. The Zone 1 spaces, shown at a slightly
positive air pressure, include offices., lunchrooms, etc.,
and no special radiological control measures are required.
Examples of Zone 2 spaces include parts of the analytical
laboratories. The potential for radioactive contamination
is low, but controlled access is required. Zone 3 spaces
include plutonium product loadout spaces, other space©
of the analytical laboratories, etc. Access is iuper~
vised. Zone 4 spaces are those which are expected to
be routinely contaminated, such as process cells. The
negative pressure in Zone 4 is typically an inch of
water. . .
A relatively large volume of air continuously
through the process building zones, through an efficient
filtration system, and to the atmosphere via a stack.
26
-------
Zone
Access
Contamination
Air Pressure
Supply Air
To Atmosphere
Stack
1
Unlimited
None
i
,
2 3
Controlled Supervised
Definite
Low Potential _ A . . ,
Potential
|
,
_^_
4
Planned
Contami nated
J
>
Treatment
System
*. .
Process ~~
Off-Gas
Ventilation Air
Air Flow
FIGTOE 11=7
OFF-GAS AND VENTILATION WASTES
-------
With the exception of incoming.'fuel shipments and the
transfer of solid scrap to burial, a^i radioactive material
handled in the operation of the plant is located in an
exclusion area» All processing equipment and systems for
processing irradiated fuel elements, except for waste tank
facilities, may be housed in the process and fuel receiving
and storage station building as exemplified in the isometric
drawing, Figure XI-=2, page xo° A summary of the process
.functions and chemical reactions involved in reprocessing
is given in Appendix A0
The .general concept of radioactivity eenfin©m
-------
o Treatment to remove radioactive material from
fluids or gases discharged to the environment so
that established limits are not exceeded,,
o Burial to confine certain non-mobile radioactive
solid wastes within the site boundaries„
The methods used, depending upon the mobility, quantity,
type and intensity of the associated radioactivity for the
uait operation involved, are shown in Table 11=3, page 30.
Accidental radioactive releases,, as measured by an increase
in off°iiite radiation level, could follow three pathways?
release to liquid effluents, release to ground water and
release to the atmosphereo The confinement and ventilation
systems in fuel reprocessing plants remove particulates of
n©n~volatiles dispersed under accidental conditions and
liquid releases to such an extent that off-site environ-
mental risks are dominated by airborne releases of volatile
and semi-volatile materials only. Therefore, the airborne
release pathway was the major one considered for evaluations
of the off°site environmental risks from potential acci-
dents in the plant operationso Secondary emphasis was
jplae©d ©a liquid releases to the ground. A radioactive
material flow diagram for a reprocessing plant, as exempli-
fied fey 183FP, is shown schematically in Figure 11=8, page 31,
So Administrative Controls
The operation of a radiochemical separations plant is in
most respects like the operation of any large chemical
plant except for the complications introduced by the radio-
active nature of the process materials <, Working with radio-
active materials necessitates adherence to extensive
government regulations for their control. These are found
29
-------
TABLE 11=3
CONFINEMENT
a. Pool Water
b. Fuel Elements-Undamaged
c. Fuel Elements-Damaged
Process
d. Fluids
e. Solids
Product
f. Uranium (Storage £ Shipping)
g. Plutonium £ Neptunium
0 h. Plutonium & Neptunium (Storage £ Shipping)
Waste-
i. High Level Liquid
Intermediate Liquid
k. Low Level Liquid with Tritium
1. Other Low Level Liquid
m. Solids (Hulls £ Equipment) in cell
n. Hulls during Transport
o. Contaminated Equipment £ Solid
Radioactive Wastes during Transport
Conlmenu-m MiMhqd Used
Warners 'r'*cnt
0 j.
l(l)
xd)
X
X
X
X
X
X
X
x •
X
X
X
X
X
X
X
* (1) Number of barriers will be dependent upon the radiation level of.the nuite.rial. If the
* radiation level is high, shielding will be provided during transport.
X
X
X
X
30
-------
BNFP RADIOACTIVE MATERIAL FLOW
LJ>
H1
SPENT
FUEL
ATMOSPHERE
SITE
STACK
PLANT
OFF-GAS
CLEANUP
SYSTEMS
SEPARATION &
PURIFICATION
SYSTEMS
J HIGH-LEVEL
IQUID WASTE
TREATMENT
HIGH-LEVEL
LIQUID WASTE
I STORAGE
SOLIDIFICATION
PRODUCT
STORAGE
LIQUID TREAT-
MENT SYSTEMS
INTERMEDIATE
LEVEL WASTE
STORAGE
SOLID WASTE
STORAGE
FEDERAL REPOSITORY
RETURN
TO FUE
CYCLE
FIGURE II-8
-------
under Code of Federal Regulations, Title 10, Chapter I0
These?,regulations serve as minimum limits for operational
The principles upon which critieality safety and radio-
active contamination controls are based, are implemented
in plant design and plant operation„ In addition, however,
administrative responsibilities are assigned to specific
individuals or groups, at the plant, for plant functions.
to assure that the reprocessing plant is operated and
maintained under the full rang® ©f normal and potential
accident conditions without risk t© public health and
The daily functioning of the fuel reprocessing plant is
governed by an on-site organization which is. self-sufficient
in regard to assuring public, plant personnel and facility
safety on a day-to-day basis. Functional components in
this., organisation include engineering, produetion, safety
and analytical services, employee and community relations
and financial groups. These licensed personnel have
specific qualifications for overseeing eritieality and
radiation safety, accountability for special nuclear materials,
plant operation, plant maintenance, plant assistance and plant
S^ S^ f S^ * & S^
services which include decontamination and waste disposal
operations» The on~site staff assure that all ssfety-
related activities are performed in accordance with ©gtabli©h©d
procedureso Reviews and audits of plant safety are performed
at appropriate intervals both on an interna-l and independent
basis for compliance with prescribed requirements,
A criticality-safety control committee appointed by manage-
ment establishes the limits on th© operating variables that
have a bearing on criticality safety. This eommitt©©,
32
-------
composed of representatives from all the functioning groups,
reviews proposed changes in equipment or in operating
procedure. The committee's approval is required before
any change is implemented,
Radiation survey inside the plant as well as in the site
environs within a radius of up to 50 miles verifies the
effectiveness of contamination control.
Through a system of checks and balances among/the functional
component© in carrying out their daily operations of the
plant; the maintenance of performance records, the institu-
tion of training programs and testing of personnel to assure
their ability to discharge safety-=related responsibilities
and the performance of audits at regular intervals, plant
safety under all normal and abnormal operation conditions
as well as full compliance with license and regulatory
requirements are further assured„
60 Auxiliary Plant Systems and Controls
a° Ventilation and Off°Ga@ System
In th@ generic plant0 only gaseous and solid radioactive
wast©© will eventually leave the site,, No radioactive
liquid effluents will be exposed to the environment„
The primary effluent that could have an impact on the
environment during operation of the facility is the gaseous
effluent from the stacko This was discussed briefly in an
earlier section and in more detail in the following„
33
-------
The gaseous effluent from the main stack consists of
building ventilation air, vaporized process condensate
and the off-gases from the dissolver and vessel off=gas
treatment system. The building ventilation exhaust air
is routed through at least two stages of high efficiency
(HEPA) filters prior to discharge to the atmosphere.
Excess process condensate is decontaminated by evaporation
and condensationo The decontaminated water is recycled
to the process or alternatively may be revalorised and-1-
discharged to the ,atmosphere-via the 100~meter- tall stack.
The dissolver off-gases (DOG)- are routed through a treat-
ment train consisting of a condenser, knock~out pot, iodin©
scrubber and an NO., absorber o The HO,., absorber is- -designed
to recover 70 percent of the NO as 45 percent nitric acid»
*"*
The treated dissolver off-gas, which still contains residual
amounts of NO and iodine, is further treated through the
vessel off=gas treatment system (VOG) which' also collect®
gases vented from various other process or storage vessels.
The VOG treatment system consists of a condenser, knock-
out pot, iodine scrubber, pre~filter, iodine1 absorber and
a two-stage high efficiency (HEPA) filter. The stack is
equipped with samplers, monitors and alarms which identify
the amounts of radioactivity in the effluent. Th© system
is illustrated in Figure II-9. The expected radioactive
release from use of the treatment systems shown in Figure
11=9 in processing high-exposure spent fuel (average annual
fuel exposure of 32,000 MWD/MTU at 40 MW/MTU) cooled 180
days from reactor discharge, and processed at the rat© of
1500 MTU/year is listed in Table 11=4, page 360
b. Water Supply
Water must be provided for process streams, for makeup
water for closed loop cooling systems, for cooling towers
34
-------
M&IM PROCESS
SYSTEM
TREATCS3EWT SYSTEM F'33? MCM-COWDEWSABLE EFFLUEWTS
FUEL
(SHEAR & ~0_J|_CO
IDISSOLVER g n
DISSOLVER
DUST
SCREEN
'11 f^
—
WSSCLVCR OFF-GAS
CONDENSER AND
KNOCK-OUT POT
I
NO. 1 IODINE
SCRUBBER
—
N0a
ABSORBER
_i ' 1 Si si
! 1 1.
r —
I i
1 PR
1
L
rr
•J Wi
ST<
tfAIN
OCESS
•y
i_
{ II (2) VESSEL OFF-GAS N ^^ 1 I
I OFF — GAS IVWUCK UUt fU 1 1 I1
LII -A " i r~ v
II Wt CONDENSER AND (ALTERNATE) ^ AI TFRMATF inniMF & AITFRWATF
KNOCK-OUT POT " ABSOHSER FILTERS ^
« {
1 TREATMENT SYSTEM FOR CONDENSABLE EFFLUENTS
i _, ; (^
1 ^ ^ ^ NOTE 3
®MAW II i AMJ R ^r*in 'I v \
H " CONCENTRATOR CCNOCI^CR j » CON(:ENTRATOR ^ FRACTIONATOR i^ CONOCh-CR — ^ VAPOniZCR |^
1 • V " ' •
11 ' ~JL Jtt. _I?
B n
vO . r "^rii-n«i ,NOTE 3
— , ... ., . — — — i... „ -i .1——.— ^ ui_i»i-n«u n g M
r
QUID
IVSTE
JRAGE
I ] i FfiFwn
" x FILTER
' ^NOTE 2
> 5LD5.
ENTlLATlOM
NOTE 2
f
IMTERNAL PROCESS STREAM
NOTEIS i
I. 7 HIS STREAM IS CONTINUOUSLY SAMPLED
FCJ3 RADIOACTIVE CONTENT.
2. 1HE5E STREAMS ARE SAMPLED FOR
RADIOACTIVE CONTENT AS REQUIRED.
3. THESE STREAMS ARE CONTINUOUSLY
SAMPLED & MONITORED WITH PROVISION
FOR DIVERSION 3 RECYCLE IF REQ*D.
FIGURE II-9
SEPARATIONS FACILITY EFFLUENT TREATMENT SYSTEM7
-------
TABLE II -4
ESTIMATED AVERAGE RADIOACTIVE F.FFLUEtiTS
(curies/sec) - Separations Plant Main Stock
Stream
NO. •
°H-3
°Kr-85
Sr-89
°Sr-90
Y-90
Y-91
»r-9S
Mb-9S
Ru-103
Ru-lOS
"1-129
"t-131
«CO-134
"Co-137
Co- 141
Co-144
Pm-147
U-234
U-235
U-238
»Pu-2J8
Ofu-239
opu-240,
°Pu-241
Pu-242
°Aia-24l
AB-242
°CB-Z42
Co- 2 4 3
"Co-244
1
3.6E-3
4.3E-1
2.0E-S
2.3E-6
2.3E-S
4.2E-6
. 7.4E-6
1.42-3
2.22-S
l.JE-3
1.4B-0
1.1B-3
S.OB-6
3.2E-6
1.4B-6 •
2.2E-5
4.0E-6
4.2E-11
4.2E-13
9.9E-12
l.OE-7
9.0E-9
1.6E-8
4.2B-8
8.62-11
0.7E-9
1.7B-10
l.OE-S
S.7E-10
1.1E-7
2
«
—
l.OE-5
1.2E-5
U2E-5
3.1E-5
3.8E-S
7.1E-5
1.1E-5
6.7B-S
--
—
2.6E-3
1.6E-S
7.3E-5
1.1E-4
2.0E-S
2.4E-10
2.4E-12
5. BE- 11
2.8E-7
2.5E-B
4.6E-9
1.2E-S
2.4E-10
4.3E-S
8.4E-10
5.1E-S
3.4B-9
5.4B-7
3
—
—
2.0E-S
2.3E-5
2.1E-S
4.2E-S
7.4B.-S
1.4E-4
2.2E-i
1.3E-4
—
--
9.0E-3
3.2E-3
LIE- 5
2.2E-4
l.OE-5
2.2E-12
2.2E-14
3.4E-13
1.1E-0
9.9E-10
1.8E-?
4.3E-7
9.3E-12
0.7E-0
1.7E-9
l.OE-5
6.7E-9
1.1E-S
4
1.4E-2
~
4.1
4.8
4.9
.8.7
IS. 4
29
4.6
28
0.1E-D
3. JK-7
11
4.7
3.0
48
a. 3
4.4E-11
4.4E-9
1.1B-7
2.2E-3
2.0E-4
3'.6E-4
9.UE-2
1.9E-6
1.4E-2
2.7B-4
1.7
1.1B-3
.. 1.8E-1
3
—
1.1E-3
1.3B-3
1.3E-3
2.-3B-3
4.2B-3
7.9B-3
i.je-3
7.4B-I
«.1B>0
S.JK-7
2.9B-3
1.02-3
o.ie-4
1.3B-3
2.2S-3
2.212-12
2.2E-10
S.4B-9
3.6E-9
3.02-6
9.1E-6
2.3B-3
4.8B-0
4.3E-5
8.42-0
3.1S-4
3.42-7
5.1E-9
6
3.6E-3
4.3E-1
6.3E-9
7.6E-9
T.SE^
1.4C-0
2.4B-0
4.&E-S
7.2E-9
4.3B-0
l.4r.-to
\.tr,-i
1 . 7K - a
l.Ot-B
4.7B-9
7.JB-0
1.3B-8
4.4B-14
4.4E-18
1.4R-14
7.9E-11
7.3E-12
1.3E-11
3.3B-9
ft.ee-i4
2. 92-11
S. 98-13
J.JE-9
2.12-12
3.4E-10
7
1.4E-2
"
1.12-9
1.32-9
1.32-9
2.3E-9
4.02-9
7.6E-9
1.2E-9
7.JB-9
1 , IK-9
1. IR-Q
t.nf.-v
1.7B-9
7.8E-10
1.22-0
2 . 2E-9
2.2E-14
2.2E-16
3. 42-19
3.32-11
4.8B-12
8.72-12
2.22-9
4.6E-14
4.32-12
8.42-14
4.12-10
3.42-13
S. 42-11
Total
To
Stack
1.8E-2
4.32-1
7.CE-9
0.0E-9
B.OS-5
.1.
-------
and for the fuel storage pools. In addition, water must be
provided as backup for the plant fire water system, emergency
cooling for the HLLW storage facilities, for the emergency
utility area, and for the 76°F cooling loop.
In the generic plant? it is assumed that water will be available
from deep wells on the site- (At BNFP, water is obtained
from three deep wells. Normal cooling is by circulation
through the cooling tower but the wells can supply straight- ,
through cooling emptying into a pond which serves as an .
emergency reservoir. The BNFP pond is a 15-acre pond having ;
a capacity of 60 million gallons of water.) Recirculation„ •
for several months, with such a pond is possible by means of ;
an emergency diesel-powered pump. Redundant pumps are ;
connected to separate emergency busses with automatic start !
features as a back-up safety measure. i
GO On°Site Electrical Power :
The loss of on-site electrical power, if sustained, could
lead to an unsafe plant condition. To minimize this
possibility, consideration is given to a high reliability
©cure© of off-site power. Should off-site power fail,
standby diesel powered generators will come on and assume
priority loads. These and day-tank fuel supplies are
located in structures designed to meet earthquake and
C /C *7
tornado criteria ' ' • An uninterruptable source of ac
power is used to supply power to the process control
equipment and to provide control for starting the diesel
generators. A 24V dc battery supply provides a highly
reliable supply of power for monitors and some control
functions„
37
-------
do Compressed Air Supply
Oil free compressed air is required for process control,
instrumentation and starting the emergency diesel powered
generatorso It is used for air lifts, air pulsers for the
extraction column operations^ for air circulation as in the
high-level liquid waste storage tanks, and for air purging
of radiolytic hydrogen concentrations;, generated during
plant operation, to prevent formation of potentially
explosive concentrations.,
The generic plant fs provided with two air compressors t©
provide © doubly redundant supply of air« The©® compressors)
are provided with pressure relief valves as a precaution
in the event of failure of the pressure cut-off switch„
Accumulator tanks located at critical locations about the
plant are equipped with reverse flow check valves in their
feed lines so compressor failure or pipe failure will not
necessarily result in immediate pressure failureo .As an
example, the emergency diesels have accumulator tanks
capable of 5 starts without resupply,,
e» Steam Supply
Steam is used for process transfer jets, process heating,
space heating, steam turbines„ decontamination, deaerator
heating and stripping, and yard steam tracing»
In geaeral, it is not a critical service except for two
aspects? the high-level liquid waste transfer and the
control of steam pressure (100 psig) and hence the temperature
(270°F)o To inhibit the possibility of a "red-oil" explosion
which requires approaching a threshold temperature a® one of
38
-------
the conditions, the steam pressure is limited to 40 psig in
the solvent extraction operations and subsequent processing
of the effluent streams containing organic solvents„ Redun-
dant pressure regulators are used to maintain the maximum
allowable process temperatures during the various unit
The generic plant, having a large quantity of high-level
liquid waste, has an emergency steam generator for emergency
liquid waste transfer which is designed to earthquake
acceleration criteria and is housed in an enclosure built
to tornado criteria»
39
-------
IIIo FAULT TREE CONSTRUCTION OF ACCIDENT SEQUENCES
lo Background
Quantitative safety analysis has been developing rapidly
due to the requirements of the space program and to the
growth in size and numbers of nuclear power reactors» The
techniques which have developed in these disciplines are
used in this study for the analysis of hypothetical nuclear
fuel reprocessing accidents and therefore a .brief review
of the development of reactor safety analysis ,will be pre-
sented,.
The accident potential in nuclear power reactors has been
recognized for some time,, Safety analysis of nuclear power
reactors has been approached by designing them for protec°
tion against the maximum credible accident (MCA)* Later,
in a desire for more realism, the design basis accident(s)
(3QBA) was defined and mitigated against „ Currently there
is a desire to present accidents on a numerical risk basis
so that comparisons can be made with risks already accepted
by societyo Early work in probabilistic reactor accident
12 13
assessment was- done by Mulvihill , Garriek et al , Farm-
1 A T K
er and Otway and Erdmann . All of these authors esscept
Farmer used fault tree analysis for the calculation of
accident probabilities„ This work is actively continuing
under AEC sponsorship. To date, the most complete d©scrip=
tions of reactor risk are provided by Otway and Srdmaxm
and Starr who place these risks in perspective by compar=
isons with other social risks» The present work provides
a similar safety analysis for a generic nuclear fuel repro-
cessing planto
-------
The central problem of probabilistic safety analysis is
the representation of a complex system such as a fuel
reprocessing plant in a form suitable for safety analysis.
A plant is represented by construction, plumbing, electri-
cal drawings, process flow charts, etc.; these must be
synthesized into a unified description of the plant acci-
dent spectrum which is suitable for probabilistic analy-
sis 0 This synthesis can be done through the use of fault
treeso •
Engineering for safety is not a new concept. Common ex-
amples are pressure relief valves on a boiler, or the
safety on a loaded firearm. In both of these examples
two simultaneous events must occur for an accident to
result. Early safety engineering was usually done after
the fact of an accident to prevent its recurrence. And in
fact, the beginning of probabilistic safety was concerned
with analyzing after=the=fact missile misfirings. Nuclear
safety engineering has introduced a new aspect; namely,
the calculation of the occurrence probability of accidents
that in many cases have never occurred, or have never
occurred with the safety system being analyzed.
Normally, reasoning proceeds from cause to effect; this
process, when applied to safety, is called failure mode
and effects analysis. This method proceeds by a series
of "what if" statements to the final undesired event but
generates many final results that are not of particular
interest.
Fault tree analysis begins with a final result that is
significant and proceeds back through the system identify-
ing causes. In a certain sense, time is going backwards
41
-------
in the logic of developing the tree., in this manner,
maay extraneous paths that would be generated by a fail-
ure-mode and effects analysis are eliminated and only
those paths that lead to the top event are generated <,
The logical structure dictated by the form of the fault
tree prescribes the manner in which the probabilities
must be combined to yield both the probability of the top
node (final hazard event), and the probability of individ-
ual event sequences.
The overall accuracy of event sequence prediction, there-
fore, depends directly upon both the availability and
quality of the basic probability data
2o Fault Tree Symbolic Language
Fault tree analysis is basically a two=§tate Boolean
logic and as such uses the operation of logical addition
often referred to as logical ™orw ("union")„
Truth Table
ABC
1 01
A •!• B = C Oil
0 00
111
Thus if A or B have a true input, th© output is true,, In
the case of fault trees, true corresponds to failure» If
either of the inputs A or B are in the failed state the
gate output is in the failed state„
42
-------
Another logical operation is logical multiplication and
is referred to as logical "and" ("intersection").
•B = C
Truth Table
ABC
1
0
0
1
0
1
0
1
0
0
0
1
Thus if A and B are true, the output is true and two
safety systems must simultaneously fail to get system
failureo In the earlier example of a loaded firearm, the
pulling of the trigger, and the safety being off must
both occur for the weapon to fire,,
Other logical operations have been defined and are used
by safety analystso In this report, care has been used
to avoid other operations because; they usually can be
represented by combinations of "and" and "or" gates, for
reasons of clarity and because these operations are the
only ones that can be directly treated in existing fault
tree computer programs.
The accident events are represented by a square which de-
scribes the event logically represented by a gate,,
A circle is used to represent an event for which probabil«
ity-of-occurrence data is available. Hence it represents
a terminal event requiring no further development.
43
-------
A diamond is used to represent an event which is not fur°
ther developed because it is not believed to be of signifi-
cance or adequate information is unavailable.
The triangle is used as a linking symbol„
LINK IS-.. . . /—A LINK OUT
A triangle, always to th© right ©£ an aao£Qs>g
t© th© gjubroutia© calculation
3o Fault Tree Constructien•feg a Kuelear
Fuel Reprocessing Plant
A chemical plant is a very complex entity in both its arch=
itecture and processing operationso Correspondingly a fault
tree modeling the safety analysis of the plant is also highly
complexo To systematize and maintain control over the comple*
tion of the safety modeling a procedure for fault tree con=
struction known as Leak Path Analysis was used. Briefly.
-------
the method enumerates the intersection of barrier failures
encountered in tracing all paths from radioactive sources
to the environmento The union of these Leak Paths forms the
top event - the uncontrolled release of radiation (URR) to
the environment. This very large equation is both unwieldy
and contains redundancies but when factored into most com-
pact form (terms appearing only once) or as close thereto
as can be achieved, it is suitable for conversion to a fault
tree of the plant. This is performed by the replacement of
intersection operations by the fault tree symbol AND and
similarly union by OR.
The top of a fault tree constructed in this manner is shown
in figure III-lo The modes of release are air and liquid
pathways. Solid pathways were not considered-
The release by liquid pathways was developed in a series of
fault trees for qualitative evaluation of critical paths
but based on historical data from high level liquid waste
18
storage was not further analys<
mdiaxaond=off" on the fault tree.
18
storage was not further analysed and the event is now
Similarly, an investigation of risks under neptunium proces-
sing failed to identify consequences as severe as those which
could occur during normal processing. Since neptunium pro-
i.cessing is performed about once a year, the probability is
correspondingly reduced and the risks are not further con-
sidered.
J .
An investigation of URR during shutdown did not reveal risks
comparable to normal operation and was therefore excluded
from detailed analysis.
45
-------
Figure
i
Generic Reprocessing Plants Top of Tree
46
-------
Release by air pathways under normal plant operation is
divided into normal and reverse air flow because reverse
air flow allows particulate from cells such as the PPC to
impact the environment without any high efficiency filter-
ing = In the investigation it was found that reverse air
flow cannot occur by equipment failure alone. There is
one exception, however, the air pulser can in principle
result in a higher pressure in the process cells than out-
side the building. Considering the reverse flow ventilation
damper, the redundant protection for maintaining normal air
flow,, the interlock on the pulser and the volume of building
and site of air flow, this mechanism was determined to be
ineffective and was not pursued. Fault trees were constructed
for the reverse air flow diamonded events shown in Figure
111=1 but are not included because they were not numerically
evaluated and void in the risk assessment.
A tornado is a natural event that can result in reverse air
flow from its depressurization and this is included in the
tornado analysis. Because of the uncertainty of failure
modes under tornado conditions, little value is derived from
the reverse flow fault trees.
The fault tree development continues with the linking tree
_o
NFR (figure III-2). The failure probability of 10 /yr or
10= /yr for single and double HEPA filters respectively are
from references 19 and 20. This is the probability of the
filter having a particulate transmission greater than design
specification. Clogging of the filters was not considered
to be a failure mode for present purposes.
The event URR in RFC is developed in figure III-3. The re-
leases .from this are further divided into those due to
47
-------
00
Figure III-2
Generic Reprocessing Piamts Ventilation System Fault Tree
-------
LIST OF EQUIPMENT
vo
10-t-lOI
lO-t-cm
io-c-ras
a-o-ooo
lo-s-tso
10-E-IOO
01-a-COI
OI-B-OOO
(0-a-ioo
B-a-is3
I9-CM30
C3-ci-ooia/a
oo-D-coic/o
CO-D-COI
es-T-flta
T-
-------
solvent fire, aqueous solutions and zirconium hull fires.
The aqueous solution spill is developed to material failure
from chemical, physical and natural phenomena. Statement
of the use of water (aqueous) seals for confining the gase-
ous products within the plumbing boundary is expressed
explicitly even though the probability of occurrence is
one for about 1 psig overpressure0
In the design of the fault trees for the reprocessing plant,
it was decided to present the generic tree for the occur-
rences in that cello This is further supplemented by the
list of equipment present in the cello In order to evalu-
ate the accident probability in that cell, repeated appli-
cations of the tree to encompass all the equipment given in
the list must be performedo This reduces the repetitive
complexity of the trees and helps to maintain a better
perspective over the details.,
The data base used in evaluating figure XII-3 and th© other
trees included in this report comes from many source®» Xfe
should first be stated that there -are no evaluated data
bases for nuclear fuel reprocessing plantsg therefore fail-
ure rate data from similar equipment in similar environments
must be usedo
Probably the best source of component data would be from
chemical plants using solutions of similar acidity„• Acci-
dents are reported to the Manufacturing Chemists Associa-fe
21
tion but there is no reporting of component failure and
there are no compilations of this data»
50
-------
Some component failures and all accidents are reported to
the USAEC but there is no systematic compilation of this
data» The body of this data is for nuclear reactors,, an
example of which is a recent analysis of pipe rupture
occurrenceso Using data reported in 1972, it is found
that the log mean pipe failure rate for PWR and BWRs is
Io6 x l(T5/yr~ft.
Similar data is available in the General Electric Pipe
22
Rupture Study series„ The report GEAP~10207~25 contains
pipe break data for both nuclear and fossil power plantsi.
23
Anyakora et al have published instrument failure rate
data on chemical plants in Great Britain in three environ-
mental categorieso Their work is summarized in a compila-
O A
tion by Powers and Tompkinso Additional British data is
25
contained in the useful text by Bourne and Green and the
O £
article by Bourne„
Data on U.S. instruments can be found in MIL-HDBK-217A
28
and in the recent IEEE survey of industrial plants. Data
on the reliability of fire prevention systems may be found
29
in the paper by Miller.
General collections embracing electrical, instrumentation
and plumbing failure rates are found in the publications by
Garrick et al and Atomics International memo0 A very
useful recent nuclear plant reliability evaluation is that
32
due to Erdmann et al.
The probabilities for natural phenomena are quite uncertain
and subject to the siting of the generic plant. A proba-
bility of 10° /yr is estimated for exceeding the design
51
-------
basis earthquake by a factor of 2 in acceleration„ This
is based upon an unpublished study prepared in connection
with an environmental impact analysis for a plant in
Richland, Washington„ The probability of 10 /yr for a
design basis tornado is from an unpublished study of a
plant in Oklahoma„
Using these data, the fault tree of 111=3 can be evaluated„
The linking tree steam explosion needed for the bottom
left hand event is presented in figure 111=40 This steam
explosion is presented as the intersection of pressure
buildup„ the failure of pressure relief devices and operator
failureo In some portions of the plant where aqueous seals
are used; a damaging explosion is not possibleo The seal
will blow and some of the entrapped radioactivity will
constitute a small release within that cello
Figure XXX-5, a red oil explosion shows the elements that
must be present to result in such a reactions the presence
of heavy ions, excess acidity, organic solvent, excessive
temperature and the failure of the operator to correct the
upset condition. These events are developed into subsidiary
trees as requiredo
Figure XIX-6 shows the fault tree for a critieality scci=
dento This is developed in a general way and some care must
be exercised in applying it to assure that all th@ failure
modes are indeed possible „. The '-valws £®ilur©§ shmm wiss-ii
originally intended to be developed into subsidiary trees
to include possible common mode problems„ Because these
were not found, chemical plant failure rate values were
usedo The failure rate for instrument power used in figure
52
-------
Coot.
1
r«/*\
p»f* \
S7x
V/e
Figure III-4
Generic Reprocessing Plants Faullt Tree for Steam Explosion Accident.
-------
U1
Fault
Figure III-5
for Red Oil Easplosiom
'O
-------
•O!
" Figure III-6.
Generic Reprocessing Plants
Fault Tree for Criticality Accident
in Process Cell
<=>C3osO«Sff
-------
£11=6 was taken from the Final Safety Analysis Report for
the uninterruptable instrument power system in the Monti-
cello power plant. This also agrees with other data on
do GO system powero
This completes the failure probability analysis that began
with figure III-3o The analysis continues by reference to
figure 111=2 and picking up another process cello The
complete collection of fault trees used in this report are
presented in Appendix- B«
56
-------
IV, • ACCIDENT ANALYSIS
1. General
A fuel reprocessing plant represents a substantial poten-
tial release source for radioactivity because of the
presence of fissile material and fission products, the use
of organic solvent® and the handling of quantities of
radioactivity in aqueous solutions« Because of the redun=
dant safety systems designed into the plant and the inherent
nature of the process,, however, it is believed that most of
the potential accidents have a low probability of occurrence»
Anticipated operational accidents are considered to be
situations in which only one independent failure, human or
equipment, is involved. Environmental radioactive releases
from such events would be expected to be less than those
allowable by 10 CFS 20 although the technical specifications
on releases may be exceeded <> A& &pper limit accident is
considered to be a situation resulting from multiple opera-
tor errors, from multiple equipment malfunctions or from
stresses imposed by natural phenomena which may have notice-
able potential environmental consequences.
0
2« History of Accidents in Reprocessing
Most of the accidents that have occurred during irradiated
(
fuel reprocessing operations took place at AEC installations
when the industry was in its infancy,, These experiences, in
general, might not correspond with anticipated commercial
plant experienceo The Nuclear Fuel Services plant at West
Valley, New York is the only commercial light water irradiated
57
-------
fuel reprocessing plant which has operated in this country.
It had 6 years of operation before it suspended operation for
§
modification to allow higher throughput . Reprocessing
experience that has been applied and operated on a produc-
tion basis in the U»So employing essentially the same
process technologies, however, is greater than 100 plant-
years o
A literature survey was made to accumulate a data base of
accideats which were categorized into types of accidents
for the nuclear energy and related chemical processing in-
dustries,, The survey ©howed that they do happen in spite
of safety precautions that have been taken. The data baa©
provided guidance in the evaluation of the safety features
assumed for the generic reprocessing plant too0 Specific
descriptions of the accident circumstances from the survey
are presented in Appendix C.
The types of accidents which have occurred or ar© probable
during reprocessing operations,, according to the litera-
ture, are categorized as followss
a) Criticality accidents
b) Fires
o
c) Explosions
d) Fuel receiving and storage accidents
e) Waste storage accidents
f) Natural phenomena events.
These are discussed*in the following sectiono
58
-------
3. Postulated Accidents; Bases for Selection
in the Hazards Analysis
Accident situations selected for analysis were hypothesized
from a review of reprocessing experiences, the design of
the generic reprocessing plant and the unit operations
involved in the process« The accidents selected are be-
lieved to have the most severe consequences in terms of
potential exposure to the environment. These events? of
low probability, are credible only if one assumes simultan-
eous failure of engineered safety features and where per=
tinente, compromise of administrative procedures established
as safety barriers„ The types of incidents considered are
those most likely to result in the dispersion of radio-
activity beyond the primary confinement, These includes
o Nuclear criticality - a solution criticality in a
process vessel in the Remote Process Cell
o Fires involving solvent and process equipment? a
fire in leached zirconium alloy cladding was also
consideredo The consequences of such an incident,
however, wer© found to have a negligible poten-
tial environmental impact
o Explosions involving different types of radioactiv=
ity sources
o Fuel receiving and storage maloperation
o Loss of cooling to the high level waste facility
o Earthquake and tornado consequences on the model
reprocessing plant's stmactures and systems con-
taining the normal inventory of design basis fuel
were also reviewed and evaluated.
59
-------
The potential accidents are discussed in the following
_in terms of malfunctions or errors required for the acci-
dent to occur, the probabilities of their occurring, the
protective measures available and the of f=site consequen-
ces that could result from such accidental releases „
a» Criticality Accidents
In the 30 years of the nuclear age, representing 432 plants
years, there have been 30 criticality accidents in the
"3 /I O C • • • . ' •
UoS* • ' o Most of these occurred in experimental facili°
ties? some of the early accidents are attributable to th®
limited understanding of nuclear energy at the time,,
There have been no criticality accidents since 1968 „ Seven
of the accidents which occurred were in solution systems
which could exist in a fuel reprocessing plant* The
balance of the incidents involved conditions that would
not be encountered in a reprocessing facility. A summary
of the relevant accidents is presented in Table IV=1,
page 61° The tabulated incidents occurred during reproces-
sing operations but not at a production facility „ However,
they provide anticipated general characteristics for crit-
icality accidents which could occur in a processing plant "s
solution systems. These are summarized as follows:
1. The number of fissions in such an event would b®
than 10
2» Th@ accident would have to be sustained
minutes
fissions
minutes to produce a fission yield approaching 10
60
-------
TABLE IV=1
Solution Criticality Accidents3"9"35
DATE
June 16,
1953
December
30,1958
October
16,1959
January
25,1361
a\
April 7,
1962
July 24,
1964
Jan. 30,
1968
LOCATION
Y-12
Processing Plant,
Oak Ridge, Tenn.
LASL,
New Mexico Pu
Processing Plant
Chemical Proces-
sing Plant
Idaho Reactor
Chemical Proces-
sing Plant
Idaho Reactor
Testing Area
Hanford Works,
Rich land, Wash.
The Wood River
Junction, R.I.
scrap recovery
facility
Y-12 Processing
Plant - Oak
Ridge, Tenn.
ACTIVE
MATERIAL
2.5 kg235U
U02(N03>2 in
56 liters H_0
2
3.27 kg Pu
Pu00(NO_)_ in
£. J A
"» 1£8 liters
v JL v U A Jk WCJ» 0
34 o 5 kg235U
^800 liters
SI-O-UO-CNO,),
<& &• *3 £
8 kg 235U
UO.(NO-) in
40 liters
HO
<&
1.55 kg
Pu
2.64 kg
235n ,•«
U in
2 3 ' 2
3.3 Kg U-233
UO_(NO.,)0 in
£. 32
1 A 1 £.**» fr • -. A. ^*. **•
TOTAL
GEOMETRY FISSIONS
Cylinder 1.3 x 1018
concrete
reflected
below
Cylinder 1.5 x 1017
water
reflected
below
19
Cylinder ^4 x 10
concrete
reflected
below
Cylinder 6 x 1017
Cylinder 8 x 1017
unref lected
Cylinder 1.3 x 1017
unref lected
Sphere 1.1 x 1016
water
reflected
PHYSICAL
CAUSE DAMAGE
Wash water None
added to
U02(N03)2
solution . .
Agitator None
created
critical geom-
etry
Solution None
siphoned from
safe to unsafe
geometry
Solution pumped None
from safe to
unsafe geometry
Concentrated None
solution
incorrectly
siphoned
Concentrated None
solution poured
into unsafe
geometry. Addi=
tional moderation
in tank
Solution surged None
from safe to
unsafe geometry
-------
3.o The rate of energy release would be too low to be
explosive, i.e.-, no shock front generation would
be anticipated
4o The event would be associated with a change from
normal procedures
5o The environmental impact would be very small?
total property loss^would be less than $'70;jjOOO as
indicated from the incidents which occurred.
Criticality incidents have typically resulted in initial
18 ' • '
bursts of less than 10 fissions followed in some instan=
ces by subsequent bursts of less than 3 x 10 fissions
per secondo Little or no damage resulted to the confine-
ment equipment from the criticality events„
In this analysis, a criticality is assumed to occur in th©
Remote Process Cello The assumptions concerning the radio-
active releaseiare discussed in section V on consequences <>
Accidental criticality in fuel receiving and storage opera=
tions is unlikely because the areas where these operations
are performed are designed to be subcritical with unirradi*
ated fuel of 5% enrichment. Light water reactor fuel is
normally enriched to less than 4%0 After a burnup of
30?000 MWD/tonne the enrichment is reduced to less than
3%o Fission products generated in the irradiation also
contribute neutron poisons in the elements thereby further
reducing their fuel wortho
Criticality in the Remote Process Cell could accidentally
occur by overfilling a dissolver through a failure to
62
-------
switch the diverter chute. Depending on geometry, this
would have to occur in conjunction with a reduction in
neutron poison and failure of monitoring instruments.
In the multi-batch dissolution process if the filter (40
mesh) fails or is eliminated it is possible to transfer
Plutonium fines to the dissolver transfer tank and to the
accountability tanko Accidental transfer of fines to the
HAF feed tank is minimized or eliminated by centrifugal
clarification of the feed after it leaves the accountabil-
ity tanko
Some possibility of criticality exists in the centrifuge.
This is made unlikely by an interlock on excessive current
to prevent further operation if it is loaded with more than
1.5 kg of insoluble fines.
Following centrifugal contacting and concentration of uran-
ium and plutonium, criticality is inhibited by controlling
the concentrations of fissile material in the solutions.
Criticality could result from the formation of an excessive
fissile material concentration due to out-of-specification
process control coincident with the failure of the concen-
tration monitors. A© an example, the dilute aqueous plu~
tonium stream (IB?) is continuously monitored for plutonium
concentration. The density is also checked by the analysis
of grab samples. In addition to these monitors, neutron
monitors are mounted on the HS column (scrubber) and the IB
electropulse column (product separation) to warn of exces-
sive plutonium concentration.
A criticality accident could occur in this part of the pro-
cos s from a failure of process control that results in
63
-------
higher-than-=normal fissile material concentrations in
§£>lution concurrently with multiple monitoring failures or
it could result from administrative error by processing
higher enrichment fuel under specifications normally used
for lower uranium enrichment.
Criticality in product loadout is also possible„ It would
be of higher consequence in the plutonium loadout area than
in the uranium loadout area., Criticality is avoided by
density control, neutron monitoring? use of equipment hav~
ing favorable geometryf and fixed neutron poisons. The
formation of plutonium polymers is controlled by keeping
the solution greater than 1 M acido The plutonium product
is stored in product tanks of favorable geometry designed
to earthquake and tornado criteria both as to integrity
and positiono
Criticality in product loadout could result from failure
of concentration control coincident with monitor failureo
It could result from plutonium polymer formation coincident
with monitor and acid control failureo It could result from
flooding of the plutonium product cell or greater than
design basis earthquake or tornadoo
bo Fires
Three postulated incidents involving dispersal of radioac°
tive material through the agency of fise have been.analyzed.
One involves a contaminated solvent, assumed to contain a
substantial loading of iodine along with plutonium and
fission products. Another involves a solvent fire in the
plutonium extraction cycle. The third fire evaluated is
-------
assumed to occur with the ion-exchange resin during product
purification 0
A fire involving leached hulls (chopped cladding containing
residual fuel material) was initially considered too» Ex-
perience in six years of shearing and handling the leached
zirconium hulls indicated a very low probability of a major
fire in this material „ More than 95S of the radioactivity
associated with the leached hulls is induced radioactivity
and is an integral part of the metal itself. In order for
this radioactivity to escape, the metal itself must burn,
potentially producing volatile particulates„ The remainder
of the radioactivity associated with the hulls is fission
product and transuranic species in the form of refractory
oxidesc All of this radioactivity which remains with the
hulls failed to be removed from the hulls during several
hours of vigorous boiling in nitric acid and subsequent
washingo In order for a hull fire to occur, there would
have to be source of ignition of sufficient energy to
initiate an active fire in the zirconium hulls which would
either go undiscovered, or not respond to firefighting
efforts using dry chemical extinguishing agents. There have
been instances under which leached zirconium hulls have
glowed but at no time has this ever affected more than a
small fraction of the hulls nor has it ever resulted in
anything more active than a transient glowing of the few
hulls involved. The probability of such an event going
unnoticed is small also, as there is an operator present at
the viewing window of the hull inspection and canning sta-
tion whenever they are being handled. In comparison to
the hypothetical solvent fire, a zirconium fire results in
lower potential off-site releases and the heat release in
65
-------
such an incident is estimated to be approximately 5S that
for the postulated solvent fire „
1. Solvent Fires
Codecontamination is the operation which removes most of
the fission products and other undesirable impurities from
solutions of uranium and plutonium without separating the
uranium and plutonium components, Partitioning is the op©r°
ation in which the uranium is separated from the plutonium
(neptunium) o Solvent extraction cycles, employing 30
volume percent tributyl phosphate (TBP) in a normal paraf=
finic hydrocarbon (dodecane) as the water immiscible sol-
vent, is used for these operations in conjunction with var=
ious chemical adjustments. Because of the moderate flash
point of the organic solvent, 70°C, there is a potential
for the occurrence of solvent fires during these processing
steps due to upsets or system leaks« Operating temperatures
are held below the 70°C flash point by temperature controls
and flow rates are monitored to avoid spills and to main«=
tain the desired compositions in all feed and discharge
streams of the equipment used in these processing steps.
This applies to the use of anion exchange resin columns or
electrochemical reductions for the partitioning steps.
A solvent fire could result from a failure of temperature
control which would allow the flash point to be reached.
Loss of temperature control could be caused by failure of
the temperature sensor, temperature control servo or missing
valve failure. Leaks or spills due to process control up-
sets or pipe rupture under earthquake conditions could
result in a solvent fire if failure of sump level sensors to
66
-------
solvent accumulations on the cell floors occurred.
2» Ion Exchange Resin Fires
Ion exchange resin columns are used to partition plutonium,
uranium and/or neptunium into separate streams while pro-
viding for additional fis'sion product decontamination fol~
lowing the initial codecontamination step,. Potential auto-
thermal resin-nitrate reaction in these ion exchange
columns can be prevented in this processing step by limit-
ing the temperature to less than 135°C and by avoiding con-
tact of the material with oxidising agents.
A resin fire could result from failure of process control
to limit the acidity as monitored by specific gravity meas-
urement of the feed streams. Monitoring is done continu-
ously and alarmed on high density indication. Failure of
such alarms and monitors to indicate a potentially hazar-
dous condition could lead to a resin fire.
Failure of temperature control to maintain the resin temp-
erature to less than 135°C could also result in a resin-
nitrate reaction. Temperature regulation is performed by
sensors with servo controls,, Thus, anything that can upset
this control, such as failure of the sensor or the servo-
mechanism, could lead to excessive heating and an incident.
Low pressure steam is commonly used for heating purposes in
reprocessing plants. Under upset conditions such that the
process came into temperature equilibrium with the steam,
a temperature of 135°C could be attained which could lead to
the onset of a resin-nitrate reaction.
67
-------
If the '..resin b©ds were highly loaded with plutoaiusi-, ?&dio«
active heating could augment the temperature of the bed,,
This could occur in the plutonium purification and recovery
operation carried out in the plutonium product cell. Fail-
ure to remove the product from loaded resins in the event
of a plant shutdown would also be conducive to initiating
such an incident-
safeguards, i0e';, the use of heavy bar grids to
inhibit th© ©sspulsion of resin from the eolisnns and prossur®
relief instrumentation,, are incorporated in th© ©q^ipmgnt
design to minimize th® consequences of such an event0
Advancing reprocessing technology may preclude th© n@©d for
utilizing ion exchange resins in separations operations.
occurrence of such an accident was therefore excluded from
consideration in this analysis.
Postulated incidents involving th© dispersion of radioactiv
ity by explosions in process equipment are analyzed. Ex-
plosions considered included th© following s
o an explosion in the high aqueous feed tank (HAF)
o an explosion in the high aqueous ^aste concentra~
tor (HAW concentrator)
o an explosion in the low aqueous waste concentra-
tor (LAW Concentrator)
o an explosion in the silver seolite iodine adsorber
o an explosion in the waste calciner
o an explosion in the plutonium product calciner,,
An explosion in the high aqueous feed tank could
ably be caused by ignition of an explosive mixture of radio*
lytieally generated hydrogen in the air ebov© th© liquid .
68
-------
An explosion in the HAW or LAW concentrator could conceiv-
ably be caused by ignition of an explosive mixture of
radiolytically generated hydrogen in the air above the
liquid in the evaporator or a "red-oil" explosion. An
explosion in the silver zeolite iodine adsorber could re-
sult from the formation of silver azide due to the presence
of excess ammoniacal material in the off-gases, A waste
calciner explosion could also conceivably occur from the
excessive presence of hydrogen and/or "red-oil" in the
equipment used in this unit operation<>
An explosion in the plutonium product calciner could
conceivably result from ignition of an explosive mixture
of hydrogen which may be used in the process of decomposing
the intermediate plutonium oxalate product or from an
accelerated decomposition of moist oxalate crystalline
materialo
These circumstances are normally precluded from developing
by means of both design and operational safety features;
multiple failures of these protective systems could con-
ceivably lead to the incidents noted.
"Red-oil" is a material that can be formed from a heavy
metal nitrate, e.g., uranium, and/or nitric acid solutions
mixed with tributyl phosphate solvent at temperatures ex-
ceeding 1350C37'38«, The exact nature of the reaction of
tributyl phosphate (TBP) with hot concentrated solutions
of heavy metal nitrates and nitric acid has not been deter-
mined. However, under optimum conditions, the reaction
becomes explosive and oxides of nitrogen are evolved. Al-
though Puress-process conditions do not approach those
-------
giving rise to such a reaction, safeguards are provided in
the concentration of uranium, plutonium and nitric acid
solutions to prevent their accidental occurrence. One
method of avoiding an explosion is to keep the evaporation
temperature below 135°C, the minimum temperature at which
the reaction occurs« Another method is to remove the TBP
from the aqueous stream prior to evaporation by steam
stripping the aqueous stream.
In order for a "red-oil" explosion to occur, several inde-
pendent instrument control failures and administrative
failures would have to precede the occurrence. The code-
contamination column would have to be out of control dump-
ing solvent with the high aqueous waste (HAW) stream for
a number of hours without being noticed or cbrrected? the
controller for the reduction of 150 psig steam (l81°C)
to, less than 40 psig steam (131°C) would have to malfunc-
tion; the relief valve which restricts the low pressure
steam would have to fail to operate? the pressure control-
ler on the steam to the evaporator heating coils would have
to malfunction, causing steam pressure to rise above 40
psig, and not be noticed for an extended period of time,
and the evaporator bottoms product would have to be sub-
stantially overconcentrated while all the normal indicators
of this condition either malfunction or are ignored. Given
these conditions, an explosion involving complexes of a
heavy metal, TBP and nitric acid-is possible.
Radiolysis of aqueous solutions results in the production
of hydrogen and oxygen. Thus, all process streams would be
expected to evolve hydrogen. It is estimated that 3 ft
of hydrogen (STP) per 10 BTU of fission product heat in
70
-------
acid waste and 10 ft /10 BTU in alkaline waste is formed
by radiolysis „ The combustion threshold of hydrogen
in air occurs at 4% concentration. To avoid reaching
this hydrogen concentration, dilution of the off-gases
with continuously flowing air is used in the processing
operationso The HAF storage tank, the HAW and LAW concen-
trators and the high and intermediate level liquid waste
systems have the highest potential for such an explosion
due to hydrogen gas accumulation as a result of failure
of the air purge system„ Such a failure would constitute
a common mode failure to all the systems.
To reduce the likelihood of air flow failure, the plant is
designed with redundant air flow features which have been
discussed under special safety systems. These systems
include redundant ventilation blowers with a spare blower
in reserve, emergency electrical power, air reservoirs and
spare air compressors.
A hydrogen explosion might occur as a result of failures
in the process ventilation system such as two DOG/VOG
blowers failing together or their power failing, filter
blockages, ventilation control failure such that air flow
balancing dampers are closed and failure of emergency
compressed air which could result in a hydrogen explosion
in the HLLW -or ILLW storage tanks if the waste is not
stored under self-boiling conditions.
The process off-gas system, composed of the Dissolver Off-
Gas (DOG) and Vessel Off-Gas (VOG) systems removes iodine,
particulate radioactive contaminants and nitrogen oxides
which are volatilized during the fuel segment dissolution
and subsequent processing steps. These off-gases are then
combined with the ventilation air, filtered through two
stages of HEPA off-gas filters and discharged through the
stack*
71
-------
Silver zeolite sorbents are incorporated in the subsystem
scrubbing trains to further restrict volatile radioiodine
releases to the environment. ^
An explosion in the waste calciner facility can be postu-
lated on essentially the same basis as the high level
waste concentrator explosion for the calciner would be
:fed from the HAW concentrator. The calciner, however,
operates at a much higher temperature than the concentra-
tor (>450°C) . This would tend to in'crease the quantity
of ruthenium that could be volatilized during an accident
by about a factor of 10 over the amount that could be
volatilized from a HAW concentrator explosion . Aside
from this variation and the fact that the equipment em-
ployed in this operation, having a small holdup which
would limit the probability of such an incident, the
results of a waste calciner explosion should be essentially
the same as for the HAW concentrator explosion analysis.
An explosion in the plutonium product concentrator can be
postulated on essentially the same basis as the high level
waste concentrator explosion. The hydrogen generation
rate from a plutonium product solution, however, is much
lower than from a high level waste solution. Hoover and
44 '
Ingalls quote a hydrogen generation rate for plutonium
as 0.05-0.06 ml/day/gram Pu-239 at the probable nitric
acid concentration range anticipated in the evaporator
concentrate. Increasing this generation rate by an order
of magnitude, to reflect the higher specific activity of
design basis plutonium, and assuming an equilibrium quan-
tity of process solution of 10 liters at 200 grams per
liter of plutonium, the rate of hydrogen generation in
the evaporator would be about 50 ml per hour. At a volume
free space in the evaporator of 38 liters, at least 30
hours would be required to reach the minimum lower explo-
sive limit of hydrogen content, disregarding the flow of
72
-------
instrument air to the probes. (This in itself would be
sufficient to keep all hydrogen generated at a concentra-
tion below the lower explosive limit for the equipment
has a small holdup capacity.)
Although the plutonium product concentration explosion
is highly improbable, it has been evaluated as an upper
limit accident for the inventory of the plutonium in the
concentrator is large = 2,000 grams or 23,000 total curies
at the time of the postulated explosion . Using the same
assumptions as were employed to analyze the high activity
waste evaporator explosion, the radioactivity release
is calculated to be 105 x 10°2 curies.
A criticality and/or explosion accident during conversion
of plutonium nitrate to oxide powder is also considered
to be highly improbable. In this operation, plutonium
nitrate in solution from anion exchange or evaporation is
precipitated as the oxalate. This product is filtered
and washed. The wet oxalate crystals are dried at 400°C
for a fixed period and then calcined at 750°C also for a
fixed period, in a slow stream of air or hydrogen. To
ensure criticality control, fissile concentration controls
and favorable geometry with fixed poisons are employed in
the operations. Batch sizes of solids processed are always
maintained subcritical. The product powder is screened,
sampled, weighed and sealed in metal containers for subse-
quent shipment or storage.
The governing radioactive material that could be released
to the environment due to such an accident would be in a
73
-------
particulate form., Decomposition of plutonium oxalate to
oxide results in particles having an indicated Mass Median
45
Diameter of 8-12 microns . Oxide particles were found
to be. 26 to 68% of the size of their precursor. These
measurements were made under laboratory conditions? plant
oxide produced under accident conditions will probably be
coarser o Mishima et al, report fractional releases up to
IS of the source when heating either the oxalate or par-
A iC
tially oxidized oxalate in are upsweep of air „ Their
finely divided, free flowing powder was composed a£
spheres with a Mass Median Diameter of 32 microns. „ As
much as Oo9% of the plutonium used in the source was made
airborne during a l=hour period at temperatures up to
1000°C and air velocities up to 100 cm/sec
For the postulated accident , it is assumed that in the
startup of the calcining furnace, hydrogen is introduced
before the air is displaced with nitrogen, violating
the procedure and resulting in a hydrogen-air explosion o
The oxalate batch size being calcined is assumed to con°
tain the equivalent of 3 kg of Pu» Overpressurization of
the furnace will not be enough to destroy the furnace „
Consequently, the explosion would be directed toward the
ends of the furnace ° The powder is assumed to be entrained
at a windspeed of 20 mpho At this condition, up to 15§ of
the solids would be expected to be airborne into the venti~
lation exhaust system. The airborne powder particles0
size is so large that the filtering efficiency of the three
stages of HEPA filters in series would be almost 100% . A
filter release factor of 8 x 10 is assumed
74
-------
If the furnace does explode, at most the surrounding glove
box might rupture and powder would be ejected into the
roosio This would create a temporary internal contamina-
tion problem in the building- However, the size of the
particles would be too large to cause a sizeable external
release of radioactivity., It is assumed a 1% release to
the exhaust ventilation system could occur„ The anticipa-
ted stack releases for the cases discussed are as follows
for a Pu mixture having a specific activity of 14 „!
Ci/g190
lo Furnace intact, direct release to vent systems
(3000 g Pu)(OolS)(8 x 10~9) = 3»6 x 10=6g Pu = Sol x
10~5 Ci
2o Furnace explodes with release to room;
(3000g Pu)(0.01)(8 x 10=9) = 204 x 10~7g Pu s 3o4 x
10°6 Ci
The off-site release from such incidents will have insig-
nificant environmental impact even under the worst disper-
sion conditions "for the EEPh filters will reduce these values
by a factor of £10 0
d» Fy©l R©G©iving and Storage Area Accident
An accident in the fuel receiving and storage area result-
ing in the release of radioactivity that could have an
environmental impaet is a credible event. The consequences
of an uncontrolled release in this plant area could be
serious although the probability of such an occurrence
would be small. Consideration of such an event was made in
this analysis„
75
-------
Based on regulatory standards and requirements for package
design/ quality assurance and handling and monitoring
procedures, the probability of a cask being breached is
,47
low
A hypothetical incident which may bound a variety of events
as to the nature and the magnitude of a release in the
fuel receiving and storage.area is assumed for this eval-
uation and is described as followss
In shipment, it is assumed that the spent fuel
cask loses its heat removal capability„ Th©
spent fuel rods self~heat due to fission pro=
duct decay heating to a temperature approaching
1225°Co This causes cladding failure and re-
lease of a large fraction of the more volatile
fission products to the hermetically se&led? dry
cask cavityo After receipt of the shipping cask
at the reprocessing plant; it is transferred to
the cask unloading pool in the fuel receiving
and storage area where it is opened» On open-
ing the cask; mobile radioactive species are
expelled from the cask cavity as a stream of
bub&les which rise to the pool surface. Seat® of
the fission products released in the cask cavity
will plate out on the cask's internal surface?
some will remain in the pool water. The balance
of the fission products - primarily volatile
species -will be airborne within the building
area and are assumed to pass through the vessel
off-gas system., Of the fission products in the
spent fuel cask inventory, it is assumed that
76
-------
all of the-noble gases, tritium, halogens,
cesium,, strontium and ruthenium in the breached
elements are released to the unloading pool» '
For this evaluation, the airborne release of
the aoble gases and tritium is neglected since
they are accepted as normal releases to the
atmosphere. Ninety percent of the halogens
(essentially iodine)„ ruthenium and cesium and
all of the strontium expelled from the cask
are conservatively assumed to remain in the pool
water* The balance are airborne and pass through
the vessel off°gas system0
Table IV=3 lists the anticipated release data for the
radioactive species involved.
®° Leakage of Fission Products From High Level
Waste Tanks
High level wastes will be stored on an interim basis in
high integrity tanks as solutions, with the radiolytic
heat removed by heat transfer to a cooling water system.
As a safety device for the eventuality of failure of the
cooling system water supply, provision is included for
venting the storage tanks to the atmosphere via an off-gas
system designed for a total decontamination factor, boil-
ing waste to atmosphere, of at least 1 x 10100 '-3*'40'41
Because of the defenses in depth which would be operative -
high integrity design of the cooling water supply system,
tank~saucer-vault containment design, a failure and/or
77
-------
TABLE
,51)
Anticipated Releases ° Fuel Receiving
a XaciclQ'nt
Group Isotopes Ci/Tonne
Source/4 . 5 tonne
Shipment
Fraction
Released
Ru
Ru~103
Ru-106
1-129
1-131
6.1X10
3.6xlO
1.6
=2
5.4(101
27o4(105)
16o2(10
7,2
"2)
OFF*
Cs-134
Cs-137
1.2xl0
10.8(lCr)
5.4(105)
Other Fission Products
78
-------
accident in this part of the facility which could have an
environmental impact is not expected. Tank storage of
high level wastes has been accomplished safely over the
25 years since they began to be generated, despite the
fact that some leaks have occurred (see Appendix C)„ Ex-
tensive measurements at the leak locations show that the
material released has remained in the vicinity of the
tankso It is anticipated that developing technology will
incorporate additional capability to detect and contain
any leakage that might occuro
fo Natural Phenomena Incidents
The reprocessing plant is designed, evaluated and construc-
ted to criteria and guidelines accepted as adequate to
provide reasonable assurance that the plant could be opera-
ted without undue risk to the health and safety of the
53
public o These criteria include requirements that the
structures, systems and components important to safety, be
designed to withstand the effects of natural phenomena
(earthquakes and tornadoes)„ The design bases that satisfy
the natural phenomena criteria reflects
o Appropriate consideration of the most severe of the
natural phenomena that has historically been report-
ed for the site and the surrounding area
o Appropriate combinations of the effects of normal
and accident conditions with the effects of natural
phenomena
o The importance of the safety functions performed„
Because (1) operating and accident stresses are lower, (2)
process parameters cannot change as rapidly, and (3) conse-
quences of any failure are generally less severe for fuel
79
-------
reprocessing operations, no plant components have perform-
ance requirements that are critical to safety assurance in
the same sense as power reactor safety system response or
cooling continuity requirements„ Protection against uncon-
trolled release of radioactive materials is assured by
maintaining the mechanical and structural integrity of
relatively passive confinement and off-gas treatment facil-
ities and of certain monitoring and control system^ compon-
ents o No violent expulsion of process materials would be
anticipated in the event of a natural phenomenon incident
as system components are not highly stressed during opera-
tion. Operating, loads are accounted for. in design and
construction because system components and interconnections
are generally of small size (over 90S of the process piping
is less than 2 inches in diameter) and relatively thick -
walled as a result of corrosion allowances provided. The
availability of multiple confinement barriers further
assures that radioactive materials escaping from the pri-
mary process systems or from effluent collection, treatment
or disposal facilities are monitored and controlled so a®
5—7
not to result in uncontrolled releases to the environment
An earthquake may negate some or all of the confinement of
individual systems but will have little, if any, direct
dispersive mechanismso
The potential off-site exposures from such an event would
be much less than one percent of the accident exposure
guideline
The plant structure will also protect the radioactive inven-
tory from the effect of a design basis tornado„ The Class
1 Ventilation System which ventilates the cells and the
80
-------
emergency cooling system for potentially self-boiling
solutions will maintain the process in a safe shutdown
conditiono The inventories of the plant areas which have
a potential for release during or subsequent to a design
basis tornado would be negligiblec Although the tornado
has great potential for releasing radioactivity, it has
great potential for dispersing the release thereby miti<=
gating the consequences„ The maximum potential off-site
releases and exposures would be well within the guide=
lines
81
-------
Consequences of Occident®
the purpose of demonstrating the degree of inherent
safety of the generic reprocessing plant and its confinement
systems, the credible consequences of operational upsets
and of stresses which might be imposed by the design basis
natural phenomena were selected in evaluating the postulated
accidentso The accidents examined are those believed to
have the most severe consequences in terms of potential
exposure to the environments These accidents are events of
low probability which are credible only if one assume© simul
taneous failure of engineered safety feature© and where
pertinent, administrative procedures established as safety
barriers are bypassed.
For each accident probability sequence, there is, at that
point in the reprocessing operations, a corresponding
consequence of radioactivity dispersal beyond the primary
confinement o These source term rel@&@<$ valu©s have b@©n
realistic
in Soefeios 2ZI
The types of accidents considered w©re discussed in Section
IVo The cases examined are listed in Table V-l. Some of
these accidents could appear in more than one location on a
fault tree or on several fault trees. The physical
tions used in the analyses are generally based upon the
design of the Barnwell Etoclear Fuel '
B2
-------
TABLE V°l
Postulated Reprocessing Plant Accidents Examined
HAW Concentrator Explosion
Solvent Fire in the Plutonium Extraction Cycle
£
Solvent Fire in the Co-Decontamination Cycle
Explosion in the LAW Concentrator
Ion Exchange Resin Fire*
Nuclear Criticality Incident
Explosion in the HAF Tank
Waste Calciner Explosion
Fuel Receiving and Storage Aceident
*» Accident Examined Was In The Neptunium Recovery Cycle
Syscideat releases might be initiated in any of the number
of process cells in the reprocessing plant. Airborne releases
would obviously have to pass through the plant's ventilation
system prior to escaping to the environment. In the realis=
tic calculations the following measured filter efficiencies
were used in developing the source termss'
First HEPA filter removes 99=9%
Second HEPA filter removes 99,, OS
Third HEPA filter removes 94«,0%
Fourth HEPA filter removes 83,0%
The accidents listed in Table V=l are considered in detail in
the following s
1. HAW Concentrator Explosion
The assumptions and results for this accident as calculated
in the Barnwell SAR are given in Table V-2«
-------
TABLE V=2
SAR .Source Calculation HAW Concentrator Explosion (Barnwell)
Group*
Ru:
Zr-Nb:
Is
OFPT :
Pu:
;
Isotope
Ru-103
Ru-106
Zr-95
Nb-95
.1^129
1-131
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce-141
Ce-144
Pm-147
Am-241
Am-242
.Cm-242 -
Cm-243
Cm-244
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ci/Tonne y SF
l.C
1.2xlOJ
06.1*105
, - 1,0'
3.5xl05
6.5xl05
,:8.3xlO'
3.6xl,0~2
1.6
/
1.0
9.0xl04
8.-4xl04
8.4xl04
1.9x105
2.4xlO|
'1.2X105'
7.9xl04
8.8xl05
1.4x10 5
250.
4.0
4.4xl04
34.0
5.7xl03
• . , o.oi
4.3xl03
3.2xl02
6.3xl02
1.7xl05
3.6
X RF X M£ ° Ci Rclcajc-d
1.0xlO"3 1.05
4.6x10
1.0
-7
4..6x10
-7
4.6x10
-7
a ..os
..1.05
1.05
.1.05
126
640
.17
.31
3.2xlO'3
0.14
.043
.042
.042
.094
.119
.060
.039
.434
.070
2.1xlO~c
.021
1.7xlO"5
.0028
2.0xlO"5
1.5xlO~6
2.'9xlO~6
8.3xlO"4
1.7xlO~8
Release Factor; t
Non-volatile fraction dispersed:
Concentrator volume =600 liters of solution.
Cell volume = 2850 m3; dispersion = 100 mg/m3
Tot.al solution dispersed = 0.285 kg of concentrate
Density of ^concentrate = 1.4 kg/J-
Fraction of concentrate dispersed « 3...3x1.0
Dispersion passing through filter = 0.14 mg/m3
Filter factor » 1.4xlO~3
Non-Volatile Volatile
Fraction Dispersed x Filter Factor + Fraction
lxlO~3
0
1.0
0
Ru
Zr-Nb
I
.OFPT
Pu
Equivaloncc Factor:
3.3xlO~4
3.3::10-4
0
3.3X10"1
3.3xlO-4
1.4x10
1.4x10'
,-3
1.4xI.T
1.4x10'
EF =,600 litorr. of conconLratf:/D67>./MTU - 1.03'MTU
: RF
l.OxlO'3
4.GX10-7
1.0 ..
4.f,y.lO~'
4.6x10
-7
84
-------
Since th@ HAW Concentrator is in the Remote Process Cell
(RFC), there are 2 HEPA filters in series before stack
release„ To carry out a realistic calculation one needs
the nonvolatile release fraction which is (2850 m-^ cell
volume)/(100mg/m3), the denominator being a representative
^ A
density»J This yields a release factor RF = 3«33 x 10
for non=volatileso In addition 100% of the iodine and
0.13 of the ruthenium are volatilized. With these data the
following releases are calculated.
V=3
Volatile Source From HAW Concentrator Explosion
ug Ci/tonne x SP X E£ + Volatile Fraction Volatile Source;
Ci
103 1.2E5 1 1.05 l.OE-3 1.26E2
106 6.1E5 1 1.05 l.OE-3 6.41E2
2-129 3.6E°2 8.3E-2 1.05 1.0 3.14E-3
l.SEO 8.3E-2 1.05 1.0 1.39E-1
is 3 a 1 SJ iO
35
-------
Table V~4
Ru:
Pus
up Isotope
It . . . _ . _i .
Ru-103
Ru-106
Nb
Zr-95
Nb-95
1-129
1-131
'T;
Sr-39
Sr-90
Y-90
Y-91
C:i-134
Cs-137
Ce-141
Ce-144
" Ptt-147
XT.- 2 41
;-.T,-242
C--242
C-.-244
>
»
?u-238
Pu-239
Pu-240
Pu-241
Pu-242
Notes lo2E5-
B8on~Volatile Source From HAW
Ci/Tonne X SF X EF x RF
ioO Io05 3.3E-4
1.2E5
SclES
IoO Io05 3o3E-4
3o5E5
6.5E5 .
8.3E-2 1,05 0
3oSE-2
loSEO
ioO 1.05 3.3E-4
9o OE4
8o4E4
8.4E4
lc9E5
2.4B'5
192E5
7.9B4
8o8E5
1.-4B5
250. OE'6
4 o OEO
4.4E4
5.7E3
OoOl Io05 3o3E-=-4
4 o 3E3
3o2S2
So3E2
lo7E5
3-.6EO
- 1 2S i©5
Concentrati
No Filter
Working
4.1SE1
2.12E2
1.21E2
2.25E2
0
0
3ol2El
2«91E1
2.91E1
6.S3E1 .
8o33El
4 o!7Ei
2.74E1
3o05E2
4.81E1
80S7E-2
1.39E-3
1.52E1
1.98B-0
lo49E0
1.11E-1
2oiSE=l
5<,2>Ei
lo25E=3
One Filter
Working
4»16E=2
2ol2E=l
lo2lE-l
2c25E-l
lcllE-4
5.9.-.B-2
1.25E-6
Two Filters
Working
4016E-4
2P12E-3
l«2lE-3
2..25E-3
0
0
3.12E1
2«91E1
2.91E1
6.S3E1 .
8o33El
4.17E1
2.74E1
3o05E2
4.81E1
80S7E-2
1. 39E=3
1.52E1
iol8E-2
1.98B-0
0
0
3oi2E~2
2o9lE=2
2o9,lE-2
SoS3E=2
8o33E~2
4.17E-2
2o74E=2
3.05E-1
4.81E-2
8.67E-5
lo39E=-6
1.52E-2
loi8E=S
io2>8E=3
0
0
3.12E-4
2.91E-4
2.91E-4
6.S3E-4
8o33E-4
4»17E=4
2o74E=4
3.05E-3.
4 o 8 IE— 4
8o67E-7
ic39E-8
lo52E-4
1.18B-7
1 . 98E-S
l=liE-6
2ol8E-S
5o9 E-4
1.25E-8
-------
Solvent Fire in Pu Extraction Cycle
The Barnwell SAR calculation is given below„
TABLE V=5
SOLVENT FIRS IS T12S Pu SOLVENT EXTRACTION CYCLE
Group
Pu;
Isotope; *
Pu-238
Pu-239
Pu-240
Pu-241.
Pu-242
Ci/Tonne
4.3xl03
3.2x102
6.3x10^
1.7xl05
3.6
X SF
1.0
1.0
1.0
1.0
1.0
X RF. X
IxlO"4
IxlO-'j
ixio";.
IxlO-J
1x10
EF =
0.034
0.034
0.034 '
0.034
0.034
Ci Releas
0.015
0.001
0.002
0.58
1.3x10
Other radioisotopes are considered to be negligible relative to
plutonium.
Equivalence Factor;
• Organic consumed = 14 liters
Processing rate = 0.208 MTU/hr
3AP normal flow =85 A/hr = 409 2-/MTU
EF =
14 9, = 0.034 MTU
40STT/MTU
Release Factor:
Pu dispersed = 1%
Filter release = 1% ,
RF « (.01)(.01) = 1x10
!
This incident would occur in the Plutonium Product Cell (PPC)„
There are 3 HEPA filters in series before stack release. The
calculation yielded the following datas
87
-------
TABLE
Solvent Fire in the Pu Eastra'ctiom Cycle
Group Isotope:* Ci/Tonne X SF X EP
Fraction No Filter
Dispersed Working
Stack Release Curies
One Filter Two Filters Three Filters
Working Working Working
?u-238
Fu-239
oo „ ?u-240
Fu-241
?u-242
4.
3.
6.
1.
3xl03
2xl02
3xl02
7xl05
3.6
•' 1
1
1
1
1
.0
.0
.0
.0
.0
0.
0.
0.
0.
0.
034
034
034
034
034
.01
.01
.01
.01
.01 .
1.46EO* 1.4&B.-3
1.09E-1 1.09E-4
2.14E-1 2.14E-4
5.77E1 5o77E~2
JL o at oJCj *=s* <3) JL o & &j£t "^ V
1.46E-5
lo02>E~S
2..14E-6 •
5.77E-4
Tl *5 OlC0™. Q
JL o o* a»j£» O
8.
6.
lo
3,
7.
76E-7
54E-8
28E-7
46E-5
32E-10
1.4(850 '- loOS %. 10
-------
{Neptunium Recovery Campaign)
Tne Barnwell^ Calculation is given in Table V-7.
TABLE V-7 - SAR Results
Croup0
flu:
Ss-Nbi
It
SET?1?!
Ptti
IscitujM! Ct /Tonne a
MORu 1.2 a 10'
lo°Ru 6.1 a 10°
"2? 3.3 a 10°
00WG» S.3 K 10°
>80| 3.8 K 10-°
•""S l.fi
°°S? 9.0 a 10*
°°SB 8.4 a 10*
oo^ 8.4 „ lo»
°«Y 1.9 a 10°
BOOCo 2.4 a 10°
DO'Co 1.2 a 10°
""Co 7.9 a 10*
'**Ca 8.8 a 10°
•*9PB 1.4 a 10"
8*'Asa 230.
8*8fts 4.0
"""OB 4.4 a 10*
"""Cra 34.0
On 5*7 is 10
800Pu 4.3 « 10°
800Pu 3.2 a 10s '
oe>0P« 6.3 » 10°
8<>8pu 1.7 tt 10°
aoaPu 3.6
DP00 a S7 «
0.01
0.16
0.73
0.01
0.14
0.14
0.04
1.0
0.032
0.001
0.20
1.0
1.0
0.23
0.79
0.96
0.13
0.66
0.90
1.0
1.0
0.43
0.99
1.0
0.01
1.0
1.0
1.0
1.0
1.0
RP a EP ° Ci Released
0.10 0.083
1.69
38
1 a 10"* 0.085
4.1 a 10-"
7.7 a 10-»
1.0 0.083
1.2 a ID''
1.8 a 10~»
1 a I0°* 0.083
1.3 a 10"»
7.2 a 10"»
7.2 a 10~»
3.6 a 10'»
1.7 a 10~»
1.0 a 10"«
0.4 a lo"9
3.0 a 10-'
1.1 a 10"'
2.1 a 10"°
3.4 a 10'°
1.6 a 10'"
2.9 a 10-'
4.8 a 10"'
1 a 10"* 0.083
3.S a 10~*
2,7 a 10"'
3.4 a 10~°
1.3 a 10-8
2.7 a 10"'
o oeher fioiion products and transueaniura olemento.
°°0I7 o Bac-ty Factor; cot roc to for simultaneous accumulation anil decay o£ otored waato
2off one yo&s be for a Np recovery campaign.
Roloooe Factor;
090 filter efeiclancy.
iO o2 contcrainonto in burned organic dispersed.
i6Q o? 3u end 1000 o£ I volotilisoo and passes out tho otack
Dlappraed
0.01
0.01
0.01
0.01
Egulvolenco ractOF* i
t&eo of thn HILC o lose ft2.
Total burned = 100 onllons a 378 liters.
2260 liters/MTU during Np campaign.
ftX o o.Sl dutlnn tip campaign.
o IIAJ> o 4^31 litera/tfTU.
S3?
37H lifcero.
Fraction
Passing
Filter
0.01
.01
0.01
0.01
Volatile
Fraction
0.10
1.0
RP
0.10
1 a 10—
1.0
1 a 10'*
1 a 10-'
4431 Utaso/MTU
0.083 MTU
-------
This incident takes place in the High Intermediate Level
Cell /(HILC) „ There are 2 HEPA filters Before stack release,
The volatile releases are given in Table V=8 and the non-
volatile releases are given in Table V°9.
TABLE V-8
VOLATILE RELEASES
Volatile Stack
Group Isotope Ci/Tonne X DF X SF X -EF Fraction Release Ci
Ru: ' -• .0.01
Ru-103 1.2xlOe 0.16 0.085 0.1 lo63EO*
Ru-106 6..1xl
-------
TABLE V-9
Group Isotope Ci/Tqnne X
Ru:
Zr-Nb:
I:
OFPTs
to
Pu:
Ru-103
Ru-106
Zr-95
Nb-95
1-129
1-131
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce-141
Ce-144
Pm-147
Am-241
Am-242
Cm-242
Cm-243
Cm-244
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
1.2x10;?
6.1xlOb
3.5x10;?
6.5xlOb
3.6x!0"2
1.6
9.0x10^
8.4x10^
8.4x10^
1.9x10;? -
2.4x10;?
1.2xlOT
7.9x10*
8.8x10;?
1.4xlOb
250.
4.0
4.4xl04
34.0
5. 7x10 J
4.3x10^
3.2x10,
6.3x10^
K
1.7x10
3.6
DF** X
0.16
0.73
0.14
0.14
1.0
0.032
0.20
1.0
1.0
0.23
0.79
0.96
0.13
0.66
0.90
1.0
1.0
0.43
0.99
1.0
1.0
1.0
1.0
1.0
1.0
NON-VOLATILE RELEASES
Stack
Fraction No Filter
SF X EF Dispersed Works
0.01 0.085 1.636-1
•01 lo63E-l
3o78EO
0.01 0.085
-01 4017E-1
7«74E-1
0.04 0.085
0 0
0
0.001 0.085
.01 1.53E-2
7ol3E-2
7ol3E=2
"3o7lE-2
lo6lE-l
9c78E-2
8.74E-3
4.94E-1
lo07E-l
2d2E-4
3o4 E-6
Io6 E-2
2o86E-5
4.84E-3
0.01 0.085 3.65E-2
.01 2o72E-3
5o35E-3
1.44EO
3.06E-5
Release Ci
One Filter
Works
lo63E-4
1.63E-4
3o78E-3
4»17E-4
7«74E-4
0
0
lo53E-5
7ol3E-5
7ol3E-5
3o7lE-5
lo6lE-4
9.78E-5
8.74E-6
4o94E-4
1.07E-4
2ol2E-7
3o4 E-9
1,6 E-5
2.86E-8
4o84E-6
3.65E-5
2.72E-6
5.35E-6
1.44E-3
3.06E-8
Two Filters
Works
1..63E-S
1.63E-6
3.78E-5
4.17E-6
7.74E-6
0
0
lo53E-7
7.13E-7
7»13E-7
3.71E-7
lo6lE-6
9.78E-7
8..74E-8
4.94E-6
1.07E-6
2»12E-9
3o4 E-ll
Io6 E-7
2o86E-10
4..84E-8
3o65E-7
2o72E-8
5o35E-8
1.44E-5
3.06E-10
-------
Explosion in the LAW Concentrator
The Barnwell7 calculation is giv©n ia Tabl© V=10,
TABLE V-10 - SAR Results
Ru:
Zr-NL-:
I:
OFPT:
Pu:
Isotope;,
Pu-103
Ru-106
7,r-95
Nb-95
1-129
1-131
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce- 141
Ct-144
Tin- 1 4 7
Am-241
Ani-242
Cm-242
Cm-243
Cm-244
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ci/Tonrc >; SF_ >. riF
0.02 O.OCL
1.2xl05
6.1x105
. ' 0.02 l.lKiO'7
3.5xlOf>
O.SxlO5
0.032 1.0
3.6x10-2
1.6
0.002 l.lxlO'7
9.0x104
B.4xlO"
8.4x104
1.9x10°
2.4xlOr'
J..2xlOJ
7.SxlO?
J.4xl05
'250.
4.0
4.4xl04
34.0
5.7x10-'
4.3xl.03
3. 2xl02
6.3x10^
1.7x10^
3.6
t.f a Cl KcJ-Jili'l
5.5
B.G
43 . 0
3.S
.no. ;t;
.00', 2
3.5
4 .OxlO"3
0.18 •
' • t ..
7.0x.I.n-:'
O.oxin"^
6 • l» '• 1 r' '
1 . !?xl 0"^J
1 . B x 1. 0 _'
9. 4x10" a
6.2x10" ;
6.9x10";
l.lxIO"4
t.i 1
3.r>xio"^
2 . 7 ^ 10"
4 . 3 X J li " ^
1 ' r» v i 0 "*
1.1x10^
— t
C.OxiO 5
Nil
OOFFT = Other fisaion products ./P->
Filter factor - 1.4 •/. 10'-5
EliMn-nt
Ku
Ji-Mb
A
Ol-T'i1
I- 1.
Kipji.v.ilc n-'C I •."•'.')i
Concent, i •••
'' er - T- *
i-'r;>.ct.ic:n
L'iKj)'.'i .''-t-c: x
%>io^
8x10
i!;;i?:^
:..- VO fc/5M
RTU
[•'i 1 1 < t Vo 1 ••: i J<:
r.ici.oi - LL'^l'-L.1™1. r —
1.4:-:: ()"3 .Oil1. f.f'M .
1.4xl(r3 , 0 i:;;'''1
i.4:.-io-lj ;• :•;'!;- 7
i.'i>:IO'J '•• i.)>:'"
-
- 4:.' R/M-M- (,;o.,- i. •».-/•:!• 1^ ff-l» .,
n-^
-------
The LAW concentrator is in the High Intermediate Level Cell
(HILC) and there are 2 HEPA filters prior to stack release.
Table V-ll lists the volatile releases and Table V-12
lists the non-volatile releases.'
TABLE V~ll
VOLATILE RELEASES
Volatile Stack
Isotope Ci/Tonne X SF : X EF Fraction Release Ci
Ru; . 0.02 3.5 .001
Ru-103 1.2xl05 8.4EO
Ru-106 6.1xl05 4o27El
/
*s 0.032 3.5 1,0
1-129 3,6xlO~2 4c03E=3
1-131 1.6 1.79E-1
93
-------
NON-VOLATILE RELEASES
Stack Release Ci
Group*
Ru:
Zr-Nb:
Is
OFPT:
Pus
Isotopes Ci/Tonne X SF X
.Ru-103
Ru-106
Zr-95
Nb-95
1-129-
1-131
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Fraction
EF Dispersed
i.2xio;
6.1x10
3.5x10;
6.5x10'
3.6x10
1.6.
-2
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce-141
Ce-144
Pm-147
Am-241
Am- 24 2
Cm- 24 2
Cm-243
Cm-244
9.0xlOJ|
8.4x10*
8.4x10^
1.9x10^
2.4x10^
1.2x10^
7.9x10*
8.8x10^
1.4x10
250.
4.0
4.4x10^
34.0
5.7xlOJ
4.3x10;
3. 2x1 Of
6.3x10;
1.7x10'
3.6
0.02 3.5 8oOE=5*
0.02 3.5 8oOE-5
0.032 3.5 0
0.002 3.5 8.0E=5
9xlO~ 3.5 8.0E-5
No Filter
Works
6o72E-l
3»42EO
lo96EO
3.64EO
0
0
5,02E=2
1 = 06E-.1
1.34E-1
S»72E=2
4«,42E=2
2.46E-2
1.08E-3
8.06E-5
One Filter
Works
6=72E~4
3o42E-3
1.96E-3
3.64E-3
0
0
5o02E~5
4.7 E-5
4.7 E-5
1.34E-4
6o72E-5
4o42E-5
4.83E-4
7.84E-5
1.40E-7
2.46E-5
lo90E-8
3 o19E-6
1.08E-6
8.06E-8
8.06E-8
4o28E=5
9o07E-10
Two Filters
Works
6.72E-6
3.42E-5
1.9'6E-5
3o64E-5
0
0
5o02E-7
4.7 E-7
4o7 E-7
loOSE-S
S.72E-7
4.42E-7
7o84E-7
1.40E-9
2.24E-11
2o46E-7
1.90E-10
3ol9E-8
1.08E-8
8o06E-10
8.06E-10
4»28E-7
9«07E-12
~5 = 8.0
-------
5. Ion Exchange Resin Fire (Neptunium Recovery Campaign)
The Barnwell7 Calculation is given in Table V-13.
TABLE V-13
Isotope
Source Activity,
Ci/tonnc
SAR Results
SF
RF
EH =
Activity Released,
Ci
Ku-103
Ru-106
Zr-95
Nb-95
1-129
M3I
Sr-89
Sr-90
Cs-134
Cs-137
lia-137m
Cc-144
Cm-242
Cm-244
Np-238
Pu-23H
l'u-239
Pti-140
l'u-241
1.24 x IQ5
7.22 x K)5
3.58x 10s
6.82 x 10.5
3.55 x 10'2
1.6
8.2 x 10-
8.2 x 104
1.79 x 105
l.25x 10-s
l.lOx 105
5.6. x K)5
4.0 x 104
4.9 x 103
6.38
4.37 x 1C)3
3.2 x 102
'6.3 x 102
1.6 x I05
3.0xlO-6 5
3.0x10-6 5
6.6x10-6 i
6.6 xlO'6 1
3.1 x 10"7
3.1 x 10'7
l.Ox ID'8
.Ox lO'8
.Ox ID'8
.Ox 10-8
.Ox JO'8
.Ox lO-8
•1.0x10-8
l.Ox lO-8
0.95
5.0 x lO'4
5.0 x lO'4
5. Ox lO'4
5.0 x lO'4
.1 x lO'2
.IxlO'2 .
.Ox lO-3
.0 x 10'3
.50
.50
.Ox JO'3
.Ox ID'3
.Ox 10-3
.Ox 103
.Ox 10-3
.Ox 10-3
.Ox lO-3
.Ox JO"3
.Ox ID'3
.Ox 10-3
.0 x 1 0-3
.Ox 10-3
.Ox JO'3
5
5
5
5
5
5
5
5
5
5
5 .
5
5
5
5
5
5
5
5
9.49 x lO'2
0.552
1.18x lO'2
2.25 x lO'2
2.75 x 10-8
1.24x 10'6
4.10 x lO-6
4.10x 10-6
8.95 x ID'6
6.25 x 10-6
S.SOx 10-r'
2.80x ID'5
2.00 x ID'6
2.45 x 10'7
3.03x 10'2
1.09x 10'2.
8.00x 10'4
1.5«x ID'3
0.40 ....:...
Non-Volatile Fraction Vulalilo
Passing Filter
Ru
Zr-Nb
I
Sr, Cc, Cs,
Ba, Cm
Pu
Np
0
9.0 x 10'4
l.Ox 10--
0
1.0x10-3
l.Ox JO'3
l.Ox ID'3
• i • ii i » •
.05
0
.50
0
0
0
Jill
5.1 x 10'2
l.Ox ID'3
.50
1.0 x 10-3
l.Ox 10-3
l.Ox ID'3
Equivalence Factor (EF)
EF=(5MTU/day)(1 day) = 5MTU
95
-------
The accident occurs in the Plutonium Product Cell (PPC)
where there are three HEPA's in series before stack release„
10% of the Ru and 100% of the I are volatilized during
the burning. The results are given below for the volatile
and nonvolatile releases. These results are included for
completeness. They are not included in risk assessment
considerations since it is anticipated that this unit opera-
tion will not be used in the reprocessing plants under
cons ider ation.
TABLE V-14
VOLATILE SOURCE
Source Activity
Isotope 'Ci/tonne
X
SF
Ru-103
Ru-106
1-129
1-131
1.24xl05
7.22xl05
3.55xlO~2
1.6
3.0xlO"6
3.0xlO~6
3.1x!0"7
3. IxlO-7
X EF
5
5
5
5
Volatile
Fraction
.05
.05
5-}-
a J
.5
Stack
Release Ci
5.42E-1
2.75E=©
1.24B-6
* 9.3E-2 - §>o3 at 10
+ Reference 5
96
-------
TABLE V-15 NON-VOLATILE SOURCE
vo
Isotope
Ru-103
Ru-106
2r-95
Nb-95
1-129
1-131
Sr-89
Sr-90
Cs-134
Cs-137
Ba-137m
Ce-144
Cm-242
Cm-244
Np-238
Pu-238
Pu-239
Pu-240
Pu-241
Source Activity
Ci/tonne
1.24x10;
7.22x10;
3.58x10;
6.82x10'
3.55x10"
1.6
4
8.2x10
8.2x10
1.79x10;
1.25x10;
1.10x10"
5.6x10^
4.0x10^
4.9xl03
6.38
4.37xlQ3
3.2x10
5
1.6xl05
SF
3.0x10-6
3.0x10-6
6.6xlO~6
6.6x10-6
3.1x10"^
3. 1x10
1.0xlO~g
1.0xlO~p
1.0x10 j!
1.0x10 £
1.0xlO_g
1.0xlO~
1.0x10 Q
1.0x10 B
0.95
5.0xlO~J
5.0x10"*
5 . OxlO~
5.0xlO~4
EF
5
5
5
5
5
5
5
5
5
5
5
5
5
5
5
5
5
5
5
Fraction
Dispersed
.45
.45
.5
.5
0
0
.5
.5
.5
.5
.5
.5
. 5
.5
.5
.5
.5
.5
.5
No Filter
Works
One Filter
Works
80 37E-16
4..87EO
5 o 91EO
lo!3EO
0
0
2»05E-3
2o05E-3
4o48E-3
3ol3E-3
2«75E-3
lo40E-2
loO E-3
lol8E-4
1,52E1
5o47EO
4oOE-l
7o88E-l
2.0E2
8.37E-4
4.87E=3
5.91E-3
1.13E-2
0
0
2»05E-6
2o05E-6
4.48E-6
3.13E-6
2.75E=6
lo40E-5
1.0 E=6
1.18E-7
lo52E-2
5o47E-3
4.0E-4
7o88E=-4
2.0E-1
Stack Release Ci
Two Filters Three Filte:
Work
Work
8.37E-6
4.87E-5
5.91E-5
l»13E-4
0
0
2.05E-8
2.05E-8
4o48E-8
3ol3E-8
2o75E-8
1.40E-7
1.0 E-8
ld8E-9
1.52E-4
5.47E-5
4.0E-6
7o88E-6
2,OE-3
5o02E-7
2.92E-S
3.54E-6
6..78E-6
0
0
lo23E-9
1.23E-9
2.69E-9
1 „ 88E-9
1,65E-S)
8.40E-9
SoO E-10
7o08E-ll
9ol2E-S
3.28E-S
2.40E-7
4o37E-7
lo20E-41
8o37E-l = 8o37
10
=1
-------
6° Nuclear Criticality Incident
The Barnwell results for a criticality incident in the
Remote Process Cell (RFC) are given in Table V-16«
TABLE V-16 /
SAR Results
NUCLEAR CRITICAIilTY INCIDENT (1010 Fissions)
Activity Released,*
Isotope Ci@ t--0
1-131(8.05d) 0.75
1-132(2.4h) 3.30
•1-133(20.5h) 18.0
1-134(52.5m) 450.
1-135(6.68hr) . 48.0
Xe-135m(15m) 395.
Xe-138(17m) 1050.
Kr-87(1.3h) . 112.
Kr-83m(1.86h) 13.5
Kr-88(2.8h) 69.5
Kr-85m(4.4h) 18.5
Xe-135(9.2h) 36.4
Xe-133m(2.3d) 0.20
Xe-133(5.27d) 2 ..70
Xe-131ra(12.0d) 60OxlO 2 -
Kr-85(10.4y) 2.0xlO"3
*The iodine is assumed to be volatile. Of the amount
generated, 1% is assumed to be released to the vent
system, the rest remaining in solution.
98
-------
7.
oson
This accident is assumed to occur in the High Level Cell
(HLCJ. The consequences of this accident have been evaluated
assuming that there is only one HEPA filter before stack release
The release data are given in Tables V-17 and V-18.
Note: It is more likely that for present and future plant
designs that at least two filters will be in line
between the HLC and the stack.
TABLE V-17
VOLATILE SOURCE
Split Equivalence Volatile Stack Release
Group Isotope Ci/Tonne Factor Factor Fraction Ci
Rus
Ru-103 1.2xl05
Ru-106 6.1xl05
1.0
.001
2.44E3
1-129
1-131
3.GxlO
1.6
0.1
1.0
1.4E-2
6.4E-1
^4.812 ^ 4.8 X 10'
99
-------
TABLE V-18 NON-VOLATILE SOURCE
Group
Split Equivalence
Isotope Ci/Tonne Factor Factor
Fraction
Dispersed
Stack Release Ci
No Filter
Works
One Filter
Works
Ru:
Ru-103
Ru-10-6
1.2x10;
6.1x10-
1.0
2.83EO
1»44E1
2o83E=3
ZR-N'b:
I:
Zr-95
:;b-95
1-129
1-131
3.5x10^
6.5xlOb
3.6xlO~2
1.6
1.0
0.5
5»9E-6
8.26EO
1.53E1-
0
0
8o26E-3
1.53E-2
OFPT:
1.0
5.9E-6
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce-141
Ce-144
Pra-147
Att-241
Am-242
Cm-242
Cra-243
Cm-244
9.0x10
8.4x10
8.4x10]
1.9x10;
2.4x10;
1.2x10'
7.9x10^
8.8x10;
4
250.
4.0
4.4x10
34.0 .
5.7x10'
4
2=12EO
1=98EO
lo98EO
4o48EO
5.66EO
2..83EO
lo86EO
2.08E1
3.31EO
5.90E-3
9o45E-5
1.04EO
8.03E--S
lo35E=l
2.12E-3
l»98E-3
lo98E-3
4o48E-3
5oS6E-3
2»83E-3
1.86E-3
2»08E-2
3.31E-3
5.90E-6
9.45E-8
1.04E-3
1.35E-
Pvi:
Pu-238
Pu-239
Fu-240
Pu-241
Pu-242
4.3x10!
3.2x10!
6.3x10;
1.7x10'
3.6
1.0
5o9E=S
7.55E-3
lo49E-2
l»02E-4
4«01E=3
3£ 10
-6
-------
8. Waste Calciner Explosion
The calciner would be fed from the High Aqueous Waste (HAW)
concentrator. It would probably be located in the same
cell area, which in the case of Barnwell, is the Remote
Process Cell (RFC). The calciner operates at several
hundred degrees Celsius and this would raise the amount
of Ru volatilized during an accident by about a factor of
10 over the HAW concentrator explosion results. Aside
from this variation, the results of a waste calciner ex-
plosion would be essentially what they were in the HAW
concentrator explosion analysis.
§. Fuel Receiving and Storage Accident
Incidents that release radioactivity in the receiving and
storage area are varied. A scenario which may bound many
similar incidents in the nature and magnitude of the release
was chosen for evaluation.
While being shipped the spent fuel cask has lost its heat
removal capability. The fuel rods fail the clad and release
a large fraction of the more volatile fission products.
When the cask is opened in the receiving and storage area
it will be submerged in water. The cask interior may be
dry and thus, much of the fission activity leaves the cask
in a large stream of bubbles that rise to the surface. Some
of the fission products will have plated out on the cask
interior walls and some will remain in the pool water. The
released products will either enter the Fuel Receiving and
Storage area (FRS) atmosphere or they may enter the DOG/VOG
system.
ior
-------
Table V~19 lists some expected release data for Ru and I0
The tritium and noble gas releases are neglected as being
part of the normally accepted release„ The material quan-
tities used were taken from the data for rail shipments
given in Table V-20.
TABLE V-19
Expected Release Data50'51
Group Isotope; Ci/Tonne
Source/4 . 5 tonn®
Shipment
Building
Fraction
Released
Ci
Released
Rus
Is
Ru-103
Ru-106
1-129
1-131
Cs-134
Cs-137
1.2x10 1(
6.1xlb3
3.6xlO"2
1.6
2.4x10;?
1.2xl03
OFF = Other
5=4Kl0
5
27o4x10'
Produets
16o2x10
7o2
10o8x10;
5 o 4x10'
TABLE V~;
~2
10%
13
Volatile Isotope Activity in Spent Fuel Shipments
1.62E-2
7'o 2 E=l
1>08E4
Isotope
Kr-85
H-3
1-131
Kr-85
H-3
1-131
Kr-85
H-3
1-131
Type of ' MTU per
Shipment Shipment
Truck 0.45-
Truck
Truck
Rail 4.5
Rail
Rail
Water 18
Water
.Water
Total Activity
per Shipment
4350
187
0.64
43,500
1,870
6.4
174,000
7,480
25. G
% Released
To Plenum
17
1
2.3
17
1
2.3
17
1
2.3
Activity
In Plenum
740
1.87
0.015
7400
IB. 7
0.15
29,600
75
0.5D
102
-------
The probability of cask heat transfer failure during ship-
-1
—3 =4
ment was estimated at 10 to 10 per shipment. With
approximately 250 shipments per year, this became 2,5 x 10
to 2.5 x 10= per year likelihood. We chose 10" per yi
for a 4.5 tonne delivery of failed fuel.
Prior to the cask being opened it was assumed that 100%
of the I, 10% of the Ru and IS of the Cs was volatile. The
amount that rises through the pool after the cask is opened
under water was assumed as 10% of what was initially volatile
for the Ru and I and 0.1% for the Cs.
The released gases subsequently pass through an iodine
scrubber (VOG) and then through 2 HEPA filters. It was
assumed that 7% of the Ru, 0.01% of the I and 0.1% of the Cs
would pass through the iodine scrubber. It was also assumed
that the passage through the iodine scrubber would cause
the volatiles to become nonvolatile. Hence the HEPA fil-
==2 =3
t®£ failus© probabilities are taken as 10 asad 10 per demand,
as previously noted. Table V-21 presents the results. Noble
gases were considered as normal releases.
103
-------
TABLE V°21
FUEL RECEIVING AND STORAGE ACCIDENT
O
in Cask toto in
Ci in a 4,5 tonne Atmosphere AtmoSoafter AmtoAfter AmtoAfter
Isotope Shipment Before Release Cask is Opened Iodine Scrubber 1 Filter
Amto After
2 Filter
Rna 103
Raa 106
I 129
I 131
Cs 134
Cs 137
So
2o
lo
7o
lo
So
4xl05
74xl06
62xlO"3
2
08xlOS
4xl05
MO)*
( = 10)
•(loO)
(loO)
(.01)
(.01)
5o4E4 (ol)6
2o74E5 (ol)
1.62E-1 (ol)
7o2EO (ol)
1.08E4 (oOOl)
5o4E3 (.001)
5o4E3
2.74E4
1 o 62E-2
7o2E-l
loOSEl
504EO
(.07)*
(o07)
(10~4)
(10~4J
(10~3)
do"3)
3 =
lo
1.
7c
lo
So
78E2
92E3
62E-S
2E-5 •
08E-2
4E-3
3
1
1
7
1
5
o78E-l
o92EO
«62E-6
o 2E"~9
o08E-5
o4E-6
3«78E-3
lo92E-2
lo62E-6
7o2E-5
lo08E~7
5o4E=8
Fraction available fojr release
-------
VI« Risk Assessment
1. Release Likelihood Spectra
To quantify the risk from a generic reprocessing plant re-
quires the synthesis of accident likelihoods and their
consequences. In Section 3, a number of fault trees keyed
to each process cell were constructed. While these fault
trees are generic in nature they do indicate the complica-
tions that are required in order to have an accident and
also indicate the probable likelihood of such an event. In
Section 5, the consequences from each of these accident
sequences identified in the fault trees are evaluated. In
this section, the results of Sections 3 and 5 are combined
to produce a spectrum of release likelihood curves for a
variety of isotopes.
Since there are 10 to 15 accidents that were considered,
an alphabetical code is employed in plotting up the data for
accident identification. This alphabetical code is given
in Table VI-1. Also shown in Table VI-1 are the number of
HEPA filters normally found between the process cell of
interest and the final exhaust stack. Both the process
cell and the fault trees that were utilized to evaluate
the likelihood of the incident occurring in the cell are
given in the table as well as the probability of occurrence
of the incident.
Figures VI-1 through VI=5 are plots of the data for the
classes of isotopes considered. Iodine and ruthenium are
105
-------
TABLE VI-1
ACCIDENTS CONSIDERED IN PRESENT ANALYSIS
ACCIDENT LOCATION
A. HAW Cone Explosion
B. Solvent Fire Pu Extraction
Cycle
Co Solvent Fire Codecontamination
Cycle
Do LAW Cone Explosion
Eo Ion Exchange Resin Fire
(Mot Plotted)
Fo Criticality Incident
Go HAF Tank Explosion
Ho Waste Calciner Explosion
Io Fuel Receiving and Storage
Accident FRSS
NO, OF HEPA
TREES
LIKELIHOOD/YEAR
RPC
PPC
HILC
ILC
PPC
Various
(RPC Typical)
HLC
(RPC)
FILTERS
2
3
2
2
3
2
1*
2
RPC,
PPC,
HILC
ILC,
PPC,
RPC,
HLC,
RPC ,
S, RO
SF
, SF
S, RO
IER
CP
H
S, RO, H
lO'5
ID"6
io-6
ID'4
ID'4
ID"5
ID'5
ID'8
FRS
10
-1
* The analysis presented in this report assumes that there is only one HEPA filter between HLC
and stack. However, it is more likely that for present and future plant designs that at least
two filters will be in line between HLC and the stack.
-------
(UJ
VI .
<
(UJ •
—a
UJ
02
.
Si
U'
l-l
H,
LEGEND
A0 HAW Cone Explosion
Bo Solvent Fins in Pu Extraction Cycle
Co Solvent Fire Co-decontamination Cycle
Do^ LAW Cone Explosion
FTCriticality Incident
Go HAFTank Explosion _
Ho Waste Calciner Explosion
L Fuel Receiving and Storage Accident
!=>13U
^
9=>129j
D,
10~5
10
,-3
10-2 10°'
i FREQUENCYAR
Figure VI =1 IODINE RELEASES ANTC1PATED FQILTHE^ 107
rJU: ACC If? E MTS AMAl'^Ef"
-------
o
IAJ
to
<
UJ
=J
UJ
fi£
103,
6=>
Ru
Ru
10"
10-
FREQUENCY/W
Vh2 RUTHENIUM RELEASES ANTIC9PATED FO8
THE HYPOTHETICAL ACCIDENTS ANALYZED
-------
A,'
D3
«n
H, ?
1Q°*|U1 C241
2 => 242.
Am
Xm
A 3=?243Cm
4=>244
§ 3? 8=>238No
^ n^ • NP
FREQUENCY/YR
Figure Vl=3 NON°VOLATILE ACTINIDE RELEASES ANTICIPATED
i FOR THE HYPOTHETICAL ACCIDENTS 'ANALYZED"
-------
• I
(LSJ
on
vls>
Co Solvent Fire Co-decontatnlndtiv» 1.-.V'
Do LAW Cone Explosion
Eo Ion Exchange Resin F?ro(Not Plotted)
F0 Criticalify Incident
Go HAF Tank Explosion
Ho Wast® Celciner Explosion
'L Fuel Receiving and Storage Accident
8=>238py
9 => 239
0 => 240'
. 1 => 241
Pu
2 => 242
Pu
Pu
Pu
10 1 Filter Foils on Demand
10 All Filters Foil on Demand
_JL
•FREQUENCV/VR
Figure Vh4 PLLTTONIMUM RELEASES ANTICIPATED FOR
THE HYPOTHETICAL ACCIDENTS ANALYZED
-------
OS.
LEGEND
Ao HAW Cone Explosion
Bo -Solvent Fire in Pu Extraction Cycle
Co Solvent Fire Co-decontamination Cycle
D. LAW Cone Explosion
Fo Critical ity Incident
Go HAF Tank Explosion
Ho Waste Calciner Explosion_
L Fuel Receiving and Storage Accident
*4 N=> 95"
£.0
u
.4 - H
10,
. FREQUENCY/ YR
f
Figure Vh5 OTHER NONVOLAmFFTsslOiNl PRODUCT RELEASES ANTTCIPATE~D
FOR THE .HYPOTHETICAL ACCIDENTS ANALYZED
111
-------
plotted separately because of their expected importance in
the analysis. The other three curves identify the actinide
releases, the plutonium releases and the remaining fission
products which are .all nonvolatile . Only 1^ is plotted for
the criticality incident in th©
To obtain the abscissa values shorn in the figur@© it
necessary to calculate or obtain from the literature a
probability of HEPA filter failure. We have found that a
probability of failure of 10= per demand for the first
filter is realistic. If more than one, filter occurs in
the series then we have assumed that an additional factor
of ten is needed for the likelihood of complete filter bank
failure. That is for two or three HSPA filters in series
=3
the probability of failure would be 10 per demand. In
the case of volatile releases the filters are not assumed
to work at all. Hence, in the case of iodine and partially
in the case of ruthenium the filters do not work because
of the volatile nature of the release,
The likelihood values in the figures are then the products
of the numbers in Table VI-1 with the probability of filter
failure. For example in Figure VI~3, the probability of a
HAF tank explosion occurring is 10° /year. When the filter
O A 1
is working this appears as G. for Am and as Gc for
242 _c •*•
Cm at the 10 vertical line. When the filter is failed
these two points are shifted to the 10~ /year line, as
seen in the figure, in which the probability of filter
failure of 10 /demand has been factored. Since these
curves are plotted on log/log paper the points that appear
furthest from the origin are those of most significance,
For example in the iodine curve, Figure VI-1, the points
G^ and D^ dominate the spectrum. These figures therefore
graphically indicate thos® accident© of most significance „
112
-------
2. Dos© Quantification
We have selected some of the data given in Figures VI-1
through VI-5 and have calculated the likelihood of receiving
a certain dose at a given distance from the generic reproces-
sing plant site. To do this we utilized the following sim-
plified meteorological model . For a puff release following
an accident we have
D = QoKo(x/Q) (6.1)
where
Q is the source strength in curies
K is a dose conversion factor in rem meters /curie second
D is the received dose in rem
and where
(X/Q) = bu
-------
TABLE VI-2
Distance From
Source
[Meters)
100
500
1,000
5,000
10,000
50,000
100,000
Category
-x
. 6.
30 =
55.
220.
400.
1500.
2700.
D Meteorology
°z
5.
19.
32.
90.
114.
310.
420.
(X/Q) 3
[sec/meter 3
h=0,y=0
5.32E-3
2.79E-S
9.0 E-5
8.0SE-S
3c50E-S
3.43E-7
1.41E=7
•to
(X/Q) -
[sec/meter ]
h=100 meters,y=0
2.76E-10
6.74E-7
4o29E-6
2.38E-6
3o25E-7
1.37E-7
'5.32E-3 = 5.32 k
-------
TABLE VI-3
Isotopic Dose Conversion Factors
(1)
Group
Rus
Zr-Nb;
Is
U1
Isotope
Ru-103
Ru-106
Zr-95
Nb-95
1-129
1-131
1-131
K factors
frem m3
[Ci sec
15.9
238
47.6
15o9
8.5 (10")
1.6 (Ifl3)
3.8 (102)
K factors
Group
OFPT;
Isotope
Pui
Sr-89
Sr-90
Y-90
Y-91
Cs-134
Cs-137
Ce-141
Ce-144
Pm-147
Am-241
Am-242
Cm-242
Cm-243
Cm-244
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
JGi secj
47.6
238
47.6
119
95
9.5
238
15.9 .
9.52(10*)
\
U
1.26(10D)
3.81(10^)
3.81(10^)
3.81(10,)
3.81(10^)
3.8K105)
(1) Conversion factors for dose to lung from inhalation of insoluble particles
except for iodine for which the organ of reference is the thyroid.
(2) Ingestion of milk by infant.
(3) Inhalation by adult.
-------
CO
s
UJ
UJ
o
Q
I: Fuel Receiving and
DISTANCE (METERS)
Figure Vl-6a RUTHENIUM DOSE AT DISTANCE FO8 A CERTAIN LIKELIHOOD
-------
10
o
Q
10
=12
LEGEND
Is Fuel Receiving and Storage Accident1
3=> 103
Ru
6 => 106
Ru
P= 10=/YR
103
10'
DISTANCE (METERS)
10'
Figure Vh6b RUTHENIUM DOSE AT DISTANCE FOR A LIKELIHOOD OF 10
117
-------
103
UJ
io
10
10
103
LEGEND
1 Fuel Receiving and Storage Accident
«"> 106^%
.
1
104 ^'i-W'-f:-'
DISTANCE (METERS)
10
Figur© Vh^s RUTHENIUM DOSE AT DISTANCE FOT A LIKELIHOOD OF
-------
UJ
e
UJ
10
10
10"3/YR
10=1/YR
LEGEND
Fuel Receiving and Storage Accident
7=>137Cs
DISTANCE (METERS)
Figure VI-7 CESBUM DOSE AT DjSTANCE FOR SEVERAL LIKELIHOODS
119
-------
Id*
5 16
UJ
u»
o
10
10
6
10*
1(f5/YR
10"6/YR
.LEGEND
G; High Acfivify Feed TANK Explosion
H: Waste Colcinsr Explosion
DISTANCE (METERS)
Figure Vh8 LUNG DOSE FROM PLUTONIUM VS DISTANCE
FOR SEVERAL LIKELIHOODS
120
-------
LU
CO
o
Q
10°
10
LEGEND
G: HAF Tank Explosion
G4 => 244
Cm
10=7/YR
10 VYR
.DISTANCE (METERS)
Figure Vl=9 CURIUM DOSE AT DISTANCE FOR SEVERAL LIKELIHOODS
121
-------
10...
10
=0
10
CO
5
UJ
UJ
8
£ 10
10
-IB
10
10 /YRiMILK INGEST ION
10 /YR«M1LK INGESTION
1P°5/YR INHALATION
10°4/YR INHALATION
1 -10=1/YR MILK INGESTION
LEGEND
D; LAW Cone„ Expl osfon
Gs HAF Tank Explosion
° Fuel Receiving and Storage Accident
10=1/YR INHALATION
n -- it
DISTANCE (METERS)
Figure VhlOa THYROID DOSE FROM IODINE-129 VS DISTANCE
-------
id2
LM
e
LU
«/1
o
o
LEGEND
LAV/Cone. Explosion
HAF Tank Explosion
Fuel Receiving and Storage Accident
© 10 /YR MILK INGESTION
10"V.YR MILK INGESTION
10"y YR INHALATION
lO^/YR INHALATION
I 0.1/YR MILK INGESTION
I Ool/YR INHALATION
)s 10* 10s 1Q6
DISTANCE (METERS)
-------
By inspecting Figures VI-1 through VI-5 those points that
provided the larger releases with the higher probabilities
were used as source terms to generate the doses at distance
given in Figures VI=6 through VI-10, The isotopes plotted
were ruthenium 103 & 106, cesium 134 & 137, plutonium 238 &
241, curium 242 & 244, and iodine 129 & 131. Any of the
other isotopes could be plotted in a similar manner.,
The distance where the largest dose occurs is approximately
5,000 meters from the plant and this appears to be common
for each of the isotopes examined. Category D meteorology
is perhaps that which occurs with the highest frequency,
A more pessimistic, less likely meteorology could, of course,
lead to a series of different dose/distance curves.
Inspection of the results presented in this chapter indicate
that incidents involving the high activity feed tank and the
fuel receiving and storage cask, dominate much of the re-
lease data for non=plutonium releases, Plutonium releases
are most evident in ion exchange resin fires and in the
high activity waste concentrator and waste calciner explos=
ions. Other incidents, such as low activity waste concentra-
tion explosions, contribute to a lesser extent,
3. Site Related Events
It is difficult in a generic study to utilize actual site
related data in the accident analysis. We have assumed that
the generic plant would be built to withstand forseeable
124
-------
site related events„ Listed in Table VI-4 are several severe
phenomena whose occurrence might damage a portion of the
plant. While we cannot do a specific failure analysis for
these initiating events„ the following statements appear
appropriate,,
Were a severe tornado to strike a reprocessing plant it
might initiate the following failuresi
©. loss of off site electric power ;-:':
bo filter failure , |.>
Co missile penetration of a portion of the.building
do stack structural failure, either partial or complete
eo ; possible loss of storage pool watero .
There appears to be practically no way in which a tornado can
cause the process cell walls to failo The loss of electric
power or storage pool water are expected to be temporary in
natureo The possible release of excess radioactivity due to
filter failure or stack collapse will be counteracted by the
extensive turbulence and dispersion caused by the tornado
itselfo Hence little radiological risk is expected from tor-
nado induced events«
An earthquake can cause the following malfunctions:
a. possible structural failure
bo loss of offsite electric power
c. internal piping failures
do stack collapse
e« loss of pool watero
125
-------
TABLE VI-4
Selected Natural Event Data
Event
Tprnado
19
Frequency of Occurrence ~"
6dO""Vyro
Earthquake •
Intensity IX
10°5/yr.
Meteorites
57
io"9/yr.
Airplane Impact
58
io"s/yr.
-------
The incident of importance here is the piping failure possi-
bility. This event could cause releases to occur in many of
the cells simultaneously. Building collapse is not expected
to occur even during severe earth tremors. The releases
would probably be liquid in nature and consequently would
not contribute significantly to airborne releases0 Moreover
^c
the frequency of 10 /year is about the same as the process
cell incidents. Hence the earthquake induced releases
Hill not impact the upper limit process cell accidents to
any noticeable extent.
Airplane or meteorite impact with subsequent fire would cause
local process cell failure„ However these likelihoods are
smaller than cell initiated events and are therefore not
significant,, •:
127
-------
REFERENCES
lo "Environmental Statement Related to Construction and
Operation of Barnwell Nuclear Fuel Plant", U.oS. AEC
Docket No. 50=332, January, 1974.
2. "Environmental Survey of The Nuclear Fuel Cycle", U.S.
AEC, Fuels and Materials, Directorate of Licensing,
November, 1972.
3o Allied-Gulf Nuclear Services, Inc., "Barnwell Nuclear
Fuel Plant Environmental Report", U.S. AEC Docket No.
50°332, October, 1971.
4. J. A. McBride, "Reprocessing, Transportation and Waste
Management", presented at the Atomic Industrial Forum
workshop on "The End of the Fuel Cycle", Chicago, Illi-
nois, August 22=25, 1971.
5o "Nuclear Fuel Services, Inc., Safety Analysis Report, Reprocessing
Plant, West Valley, New York, 1973."
60 General Electric Co., NEDO°10178, "Final Safety Analy-
sis Report - Midwest Fuel Recovery Plant", UoS0 AEC
Docket No. 50-268, December, 1970»
7o Allied-Gulf Nuclear Services, Inc., "Safety Analysis
Report - Barnwell Nuclear Fuel Plant", U.So AEC Docket
No. 50=332, October, 1973.
80 Eo Ro Irish and W. H. Reas, "The Purex Process = A
Solvent Extraction Reprocessing Method for Irradiated
Uranium", HW-49483A, April 8, 1957.
9. Code of Federal Regulations? Title 10, Office of Fed-
eral Register, National Archives and Records Service,
General Services Administration, Washington, B.C., 1971.
10. Ho C-o Rathvon, "Reprocessing Techniques and Their
Licensing and Regulatory Framework", presented at the
Atomic Industrial Forum workshop on "The End of the
Cycle", Chicago, Illinois, August 22=25, 1971o
11. "Siting of Fuel Reprocessing Plants and Waste Management
Facilities", ORNL-4451, Oak Ridge National Laboratory,
Oak Ridge, Tenn«, July 1970, p. 4-5.
128
-------
(Confd)
12 o R. Jo Mulvihill, "A Probabilistic Methodology for the
Safety Analysis of Nuclear Power Reactors", PRC-R-657,
Planning Research Corporation, February 1966.
13o Bo Jo Garrick, W0 Co Gekler, Lo Goldfisher, Ro Ho Kar-
cher, Bo Shimizer and Jo Ho Wilson, "Reliability Anal-
ysis of Nuclear Power Plant Protective Systems", HN-
190, Holmes and Marver, May, 1967»
14o P. Ro Farmer, "Siting Criteria - A New Approach in
Containment and Siting of Nuclear Power Reactors",
IAEA Report STI/PUB/154, Symposium Proceedings, pp0
303-329, Vienna, Austria, 1967»
15o Ho Jo Otway and R« Co Erdmann, "Reactor Siting and De-
sigsi From a Risk Viewpoint", Nuclear Engineering and
Design, 13,. 365, (1970) o
16o Co Starr, "Benefit-Cost Relationships in Socio-fechni-
cal Systems", IAEA-SM-146/47, Symposium on Environmen-
tal &apoets of Nuclear Power Stations, U»N0 Headquar-
ters, R.Y., August 14, 1970» _
17 o E. Ko f \allwood and S» C= Erdmann, "On th® Uss® of Leak
PatJo ^sialyois in fault Tree Construction for Fait loac-
toe Sivfsty0, Fast Reactor Safety Bteeting, Beverly Hills,
California, April 2-4, 1974 .
IQo Uo Lo E,@sia
-------
22
23
24
26<
27,
28,
2S
30
31
am68o
Liquid Metal Engineering Center, "Failure Data Handbook
for Nuclear Power Facilities00, LMEC=Memo=69°7, August 15,
1969o
-R0 Co "Esdsaann, Do Okrent, P0 Godbout and K» &0 Solomon,
"Fault Tree Analysis of Reactor Safety Systems with Appli-
cation to the Residual Heat Removal System of -a BWRM , paper
V-47 in Conference on Mathematical Models and Computational
Techniques for Analysis of Nuclear Systems", USAEC-CONFo
730414-P2, Ann Arbor, Michigan, AprS.1 9°ll,1973o
130
-------
REFERENCES (Cont'd)
33. Northern States Power Co., "Final Safety Analysis Report,
Monticello Nuclear Generating Plant", USAEC Docket No.
50-263, 1969.
34, "Operational Accidents and Radiation Exposure Experience
Within the USAEC 1943-1970", U.S. AEC WASH-1192, 1971o
35. J0 A. McBride et al, "Safety Aspects of U. S. Fuel
Reprocessing", Third International Conference on the
Peaceful Uses of Atomic Energy, Paper A/Conf« 28/P/278
(1964).
36. Nuclear Fuel Services, Inc., "Safety Analysis Report, NFS
Reprocessing Plant, West Valley, NoY.1?, UoS. AEC Docket
No. 50=201, Volume II, 1973.
37-, G. Mo Nichols, "Decomposition of the Tributyl Phos-
phate - Nitrate Complexes", USAEC Report DP-526, E.I.
DuPont de Nemours and Co., Savannah River Laboratory,
November, 1960.
3i» T. J. Colvin, Jr., G. M. Nicols and T. H. Siddall,
"TNX Evaporator Incident January 13, 1953", UoS.AEC
Report DP-25, Eolo DuPont de Nemours and Co., Savannah
River Laboratory, May 15, 1953.
39. E. R. Irish, "Separations Plant Silver Reactor Inci-
dent", HW-57048, September 15, 1958.
40. L. T. Lakey and J. R. Bower, "ICPP Waste Calcining
Facility.Safety Analysis Report" IDO-14620, December 1,
1963.
41. G. E. Lohse and R. E. Commander, "Initial Operation of
the Idaho Waste Calcining Facility and Radioactive
Feed", Proceedings of the Symposium on the Solidifica-
tion and Long-Term Storage of Highly Radioactive
Wastes, Richland, Wash., Feb. 14-18, 1966, rj.fi. AW
CONF-660208, 1966..
42. Ko J. Schneider, editor, "Waste Solidification Program,
Process Technology-Pot, Spray and Phosphate Glass Solidi-
fication Processes", BNWL-1073, August 16, 1969.
131
-------
REFERENCES (Cont'd)
43o A = Go Blasewitze editor,, "Fixation of Radioactive Resi-
dues™, BNWL-1074,. Quarterly Progress Report, February-
April 1969.
44o Do A= Hoover and W. B. Ingalls, "Study of Polyethylene
Bottles as Containers for Plutonium Nitrate™, Proceedings
of the 2nd International Conference on Packaging and
Transportation- of Radioactive Materials, Union Carbide/
U o S * AEC CONF- 6 810 0 31 (196 8) .
43,, Pa E'o. Potter, "Studies of the Sintering Behavior of
Plutonium Oxide% AERE^R~4729, Metallurgical Division^
AERE, Harwell, England, September, 1964„
46o Jo Mishisaa et al, "Plutonium Release'Studies", XIXo Release
from Pu Bearing Powders", BNWL-786, July, 1968o
47 o "Environmental Survey of Transportation of Radioactive
Materials to and from Nuclear Plants", u. So AEC Report,
December, 1972.
48o Go W. Parker,: J0 W» Martin and Go E. Creek, "Fission
Product Release from Reactor Grade UO0 by Oxidation,
Diffusion and Melting", ORNL-CF-60=12-140
49o G. Wo Parker et al, "Fuel Element Decomposition Pro-
ducts", TID-7627, Seventh AEC.Air Cleaning Conference,
October 10-12, 1961o
50. Jo Ho Goode and V0 Co A» Vaughen, "Experiments on the
Behavior of Tritium During Head-Sad Processing of
Irradiated Reactor Fuels", ORNL-TM-2793, February, 1970,
51. "The Safety of Nuclear Power Reactors and Related Facili-
ties", UoSo AEC Report WASH-1250, July, 1973o
52. J0 Eo Mendel and J0 Lo McElroy, "Waste Solidification
Program, Volume 10, Evaluation of Solidified Waste
Products", BNWL-1666, July, 1972,
132
-------
MiF KKEMCES (Cont' d)
53. code of Federal Regulations: Title 10, Office of Fed-
eral Register, National Archives and Rocords Service,
General Services Administration, Washington, D.C.,
Appendix A of 10 CFR Part 50 "General Design Criteria
of Nuclear Power Plants"? Appendix B to 10 CFR 50,
"Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants"; Appendix A to 10 CFR
Part 100, "Seismic and Geologic Criteria for Nuclear
Po%j®r Plants".
54. Letter from Mr. G. E. Kley, Special Assistant to the
Director, USAEC Division of Operational Safety„ October
4, 1973 on safety-related incidents in nuclear reproces-
sing and fabrication facilities for 1971 and 1972.
55. D. H. Slade, editor, "Meteorology and Atomic Energy",
USAEC, DRDT 1968.
56. "Environmental Analysis of the Uranium Fuel Cycle",
EPA-520/9-73-003-D, U.S. Environmental Protection Agency,
Office of Radiation Programs, October, 1973.
57. "Probability and consequences of Airplane Crashes into
!S?*la^-,Areas"' K° Solomon' •*• C. Erdmann, D. Okrent.
UCLA 0(1974).
58. "Estimate of the Hazards to a Nuclear Reactor from the
133
-------
APPK.'IDIX A
Summary Table Describing the Basic Operations, Process
Functions and Chemical Reactions in the Generic Spent Fuel
Reprocessing Plant (Modeled after BNFP).
A-l
-------
Cask receiving
and handling
Fui' 1 sto rage
o
and tranr.fer
FUNCTION' AND PRINCIPAL
CHKMICAL REACTIONS
Receipt and preparation of
shipping cask for unload-
ing
Storage of fuel elements
until dissolution
Preparation of fuel for
dissolution
DESCRIPTION
Cask and carrier will be monitored for outside contamination
and washed to remove outside dirt. The cask will be
removed and the condition of fuel and coolant determined
by temperature, pressure, and coolant radioactivity
measurements. The cask •.•.'ill be vented to the vessel off-
qas "system and the primary coolant replaced, if necessary.
The cask will be placed ir. the cask unloading pocl •.-.'here
the lid will be rer.cvod ^r.d fuel elements unloaded re-
r.'ctely under water shield. Empty casks will be decor.tam-
ir.atcci, r.cr.itcrod, ar.J returned to customer.
fuel elerer.t identity •••.•ill cc confirmed and the elements
placed in storage canisters ir. the storage pool. Pcol
vater will be circulated through heat exchangers, inorganic
an.: rauioajtive ccr.t.-'.-;:-..i.-.t.-5. r.icm-:-r.t= :-:i^i LV 'remotely
transferred from the- -jr-ol tr the feed mechaniEm c: shear.
Fuel elements will L-o me ch.^r.i rally chcrpcd ;ntc .-m.ill
segments, execs ing cxidc- fuel inside- the el one :•.;.:• to dis-
solution wh: le ^;it.-=id.' jl.": J.:ir. 3 istair.lesr i; t t.i-1 or
-------
PROCESS STEP
Dissolution and
feed preparatior
Dissolution
Solids handling
and waste
Co-decontamina-
tion and parti-
tion cycle
FUNCTION A1JD PRINCIPAL
CHEMICAL REACTIONS
Conversion of the fu
to a liquid solution
3L-0
4H2O -»• 2NO
21! .O -»• 2NC:
FISSION PRODUCTS «• x KN'Oj-*
F.P. (NOj) + yHjO+Z NO
X
PuO2 + 4HN03 - > P
2H2O
NO + 2HN'Oj
Disposal of ur.ciissolved
cladding hulls
Separation of the plutonium
and uranium from the bulk
of the fission products
and partitioning of—the
plutonium from the uranium
DESCRIPTION
The chopped fuel elements will be contacted with hot,
concentrated nitric acid which .will convert ur.i.-.iun,
plutonium, and irost of the fission products to soluble
nitrate salts. Undissolved cladding (hulls) will rcr.im
in cissolver basket. Gases generated during dissolution
will be channeled to off-gas treatment system. Nitrate
salt solution will be transferred to tanks for sampling
measurement and final acid adjustment.
The cladding hulls will be rinsed and transferred by
shielded trailer to a burial ground. Intermittently,
or in case of abnormalities during dissolution, hatches
of hulls will be checked for complete dissolution of
plutonium and uranium.
Adjusted aqueous feed solution and tributyl phosphate (TBP)
diluted in a normal paraffin hydrocarbon will be mixed
counter-currently in a bank of centrifugal contactors.
The organic solution, which preferentially extracts the
-------
PROCESS STEP
Extraction
Reduction and
Partitioning
Strim
Second uranium .
cvclo
FUNCTION AND PRINCIPAL
CHEMICAL REACTIONS
+ 2NO3" + 2TBP
3)Z-2TBP
Pu*" + 4NO3~ •«• 2TBP nTCKH2S>
Pu(NO3),,: 2TBP
""* + 2NO, + 2TBP
n"C|2K?6>Pu02(NO,)a- 273P
aq
aa
aq
2e
aq
U+" + 2PU*" * 2lijO
aq aq
2Pu
*}
2HN02
N,O
UOZ (NO,) j -2T3P + II20 + 2H
UOj** + 2HNO, + H2O*2TBF
Further decontamination of
uranium from fission
products
DESCRIPTION
nitrate complexes of tetravalent plutonium and hexavalent .
uranium, will exit from the centrifugal contactor and pass
through a pulsed scrub column where an aqueous nitric
acid solution will remove extracted fission products
from the organic stream. The organic stream will pass
through a partitioning column where plutoniun will be re-
duced to the inextractable trivalent state and stripped
into another aqueous nitric acid stream containing hy-
drazine. The organic stream will pass through another
colurr.r. where the urar.iur. will be stripped into acidified
water. (Alternatively, anion exchange would be used
for partitioning plutonium and uranium into separ=-
ate streams)
Nitric acid will be added to the aqueous strip stream
containing the uranium, and the uranyi nitrate complex
will again be preferentially extracted by another TEP
solution in a pulsed column. Before leaving the column.
-------
PROCESS STEe
Ul
Uraniur. silica
gel, product
storage
Second and t
third plutoniurn
cycle, storage,
and shipping
Oxidation
Extraction
Reduction and
Stripping
Fl'NVTION AND PRINCIPAL
CHEMICAL REACTIONS
Final decontamination
and disposition of uranium
Final decontamination
and disposition of plu-
tonium
2NO
•§• NO
4NO," + 2TB?
Pu(NO,)„ • 2TBP
j_ i „ ~ ,. n i, * '
aq
aq
2e
aq
2H.O
DESCRIPTION
the organic stream will bo scrubbed successively with
strong and dilute nitric acid solutions which rer.ovo
extracted ruthenium and :irconiun-nicbium, respectivoiv.
Uranium will be stripped from the organic strea- in
another column, using acidified water, and this solution
will be subsequently concentrated by evaporation.
Concentrated uranium solution will be passed through silica
gel beds to remove traces of zirconium-niobium. Uranyl
nitrate product solution will be analyzed and stored in
tanks until shipment..
Plutonium in aqueous stream leaving partitioning column
will be reoxidized to the extractable tetravalent state,
which will be preferentially extracted into the T3P
organic stream in a pulsed column. In the same column,
the organic stream will be scrubbed successively with
strong and dilute nitric acid solutions, which will remove
extracted ruthenium and zirconium-niobium, respectively.
The organic stream will pass through a strip column where
plutonium will be reduced to inextractable trivalent
state, which will transfer to the aqueous stream of dilute
nitric acid and hydrazine. The extraction-stripping
-------
PROCESS STKP
Scrubbing
>
I
No. 1 solvent
systom
Carloriat«? wash
FUNCTION AND PRINCIPAL
CHEMICAL REACTIONS
2Pu
+lt
2H,0
aq
TBP + UOz"1"+ + 2ND,'
U02(N0})
2T3P
Removal of degradation
products from solvent
Na^CO, + 2(C.H,-); H?0, - -
2(CJ'.,) , Na P0b + H2CQ,
2RCI!:N'O:.----^2RC:i = NOOH +
Na.CO -—• 2?.C':i = NOON a
«- J
H2CO,
DESCRIPTION
sequence will be repeated in the third plutonium cycle
for further decontamination. A TB? scrub stream will
remove residual uranium fro.Ti the plutonium aqueous streair.
as it leaves the last strip column. Plutonium concentration
will be acccr.plished by nair.tainir.g a high ratio of .
organic to aqueous flow.in the strip columns. Final plu-
tonium nitrate solution will be .washed with an organic
streat?, of r.orival paraffin hydrocarbon (diluent for TBP) to
remove traces of T3F and phosphate. Product solution
will be analyzed and stored in tanks until shipment. Solvent
streams leaving plutonium cycles will pass through a strip
calu.T.n to rer.ove residual. inextractablp species of uranium
and plutonium and will be recycled to the co-decontamination
cycle.
Organic solvent stream fror. co-deccr.tar.ir.aticn ar.d partition
cycle will be Washed successively with dilute ftrv-ecus
solutions of sodium carbonate, nitiic acid,and sodium
carbonate (or sodium hydroxide) to remove organic
degradation products iry sxtr.iction or precipitation;
precipitated solids will be ro.roveo. by a filter. Fresh
TBP or diluent (normal paraffin hydrocarbon) will be added,
-------
PROCESS ST£?
N'o. 2 solvent
system
Liquid waste
treatir.a and
storage
Acid reduction
FUNCTION AND PRINCIPAL
CHEMICAL REACTIONS
Removal of degradation
products from solvent
Disposal of liquid
waste streams with
minimum residual waste
volume for storage.
C12"z2Ol« + 18'2 HNO3 *
12C02 + 14.9 NO + 3.3 N'O2 +
20.1 H2O
DESCRIPTION
as required, to' maintain proper TDP concentration or
total solvent inventory.
Organic solvent strearr. ffor. second uranium cycle -will be
treated similarly to No. 1 system, except the secc.-.d
alkaline wash will be emitted.
The highly radioactive waste stream from the co-decor.taT.ina-
ticn cycle will be concentrated by evaporation; acidity of •
the concentrated bottoms will be reduced to permit lonq-
term storage in stainless steel tanks by reacting with a
sugar solution; overheads will be fed to the low-activity
evaporator for further decontamination. Xost of the re-
maining nitric acid waste streams containing low levels
of fission products, uranium, and plutonium will be con-
centrated in the low-activity waste evaporator; cor.cer.tr.'.tsd
bottoms will be recycled to the co-decontamination cycle:
overheads will be condensed and fed to the acid recovery
system. Miscellaneous waste streams, containing salts,
low levels of f-ission products and no appreciable uranium
or plutonium, will be acidified and concentrated in tho
general-purpose evaporator; bottoms will be stored; over-
heads will be monitored for radioactivity content and then
-------
PROCESS STEP
Nitric acid
recovery and
storage
>
cc
Off-gas treating
Iodine Scrub
FUNCTION AND PRINCIPAL
CHEMICAL REACTIONS
Recovery of nitric acid
and reduction of nitrogen
oxides release to the en-
virons.
Removal of radioactive and
other pollutants from
gaseous effluents
"Hg(NO,)j + 4I~- - -3>
Hgl,.
°Hg(NO,)
°'Hg/I2 mole ratio >, 4
DESCRIPTION
discharged.
Overheads from LAW evaporator will contain most of the
tritium (as tritiated water) and some undestrpyed nitric
acid from the process; they will be condensed and fed
to the fractionator which concentrates nitric acid. Re-
covered acid will be stored and used in make-up of
various acid streams; overheads containing tritiated
water will be monitored for radioactivity (other than
tritium) and released to the stack.
Off-gas from dissolver will pass through a scrubber where
radioactive iodine will be removed by contact with dilute
aqueous solution of nitric acid and ir.ercurous/mercuric
nitrate; it will subsequently pass through an acid ab-
sorber where nitrogen oxides will be removed. Cissolver
off-gas and vessel off-gas streams will be combined, passed
through another mercurous/mercuric nitrate scrubber, an
iodine adsorber bed, and a high-efficiency filter
before release to the stack.
-------
Appendix Bs Fault Trees Used in Risk Assessment
Page Description Abbreviation
B-l Top of the Reprocessing Plant Fault Tree
B-2 URR Normal Airflow NFR
B-3 URR in Fuel Receiving and Storage FRS
B-4 URR in Remote Process Cell RPC
B-5 URR in High-Level Cell HLC
B-6 URR in High Intermediate-Level Cell HILC
B-7 URR in Intermediate Level Cell ILC
B-8 URR in Plutonium Process Cell PPC
B-9 URR in Plutonium Loadout Cell PLC
B-10 URR from High-Level Liquid Waste Tank LPS^WTV.WTS
B-ll Failure of Heat Transfer from HLLW HT
B-l2 Failure to Institute Emergency Reflood ER
Dissolver Off-Gas System DOG
Vessel Off-Gas System VOG
B-13 Cooling Water Failure WS
Dissolver and Vessel Off-Gas Common System-VD
B-14 URR from Resin Reaction RRR
B-15 Solvent Fire in Partition and Purification-SF
B-16 Red Oil Explosion RO
B-17 Steam Explosion S
B-18 Criticality Accident in Process Cell CP
B-19 Controller Fails Unsafe C
Failure of Personnel Intervention PI
On-Site Power Failure OSP
B-20 Acid Fraction Overhead AFO
Pump Failure P
-------
Top of the Reprocessing Plant Fault Tree
D-l
-------
URR Normal Airflow
-------
td
I
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URR in Fuel Receiving and Storage
-------
IST OF EQUIPMENT
W
19-C-101
io-c-isa
19-C-IOS
lo-a-ioo
lO-O-lOO
10-o-ioo
41-a-aoi
41-I-OQO
13-tt-IM
19 -a-199
10-0-130
C3-P-OOIC/D
OO-n-COIC/0
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CG-7-CX.
ID -7- 101
IS-T-IOQ
in-f-ioD
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18-T-IOT
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i
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MOO eOBCHaTOATOQ QU08ILOQ
MAO coocnaTDflToa
-------
EQUIPMENT LIST
to
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19-0-
19-0-
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• • •
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IT 0 104 OJ3OT.WO reoOUCT IOCS POT
21 o lot MA ft to matt
81 0 S40 MAF COaSTAaT LffVQL Htt^O POT
II 0 IBS tOO HCAO POT
41 0 04oo UNIRAL ninpooi coacanTQAToR »nn
4> 0 412 HUMP COLLECTION TAUR
490 44T Q. P. rCCO REAO POT
0 tea OIS90LVCO ACIl) 9UROI TAH«
o aiT HULL nnsf sunot TANK
0 013 HUU. RINSC CONSTANT LEVXL HIM) POT
0 M6 MUU. "'NSC MIX FCCDKBO POT
01 0 099 MQ I OtS90LVCA ACIO HEAO POT
01 D 6SO »0»- • •'
0 Ml HO. J " • "
I 0 QOX HO 4 - • • •
I 0 <>! OlUOLVZD MILL Ria£Z DIAL KXAO POT
VALVE POTKO SAMPLE PCBMT
00 0 001 -Ol ACCOUOTAOIUTT TCSt
CO 0 COS -02 KO.I FCEO AOjUOTagtJT TABB
000 018 -18 HAa CIRCE TAttK
ODD Oil -IS LAD COaCHnTRATOa FOM TABU
oo o 019 -in Kit ptopoa conciciTKiToa POO) nan
CO 0 OO -10 Htm 8ADPUE TAHR
•0 0 02B -20 OUUP COLLECTIOH TAW
00 0 OK it LABO CHECK TANK
000 BO 99 80D LINE
CO 0 Otl -Ql OI1SOLVER FLU9H ACCUOULATOa
00 D 002 08 BO.8 FCCD aOJUgTtlffNT TANK
ooo oro -TO noioissoi. EB TRANSFER lann
• 00 0 071 71 KO 8 "
000 078 72KO.S-
COT079 -7S 00.4 '
CO 0 074 -74 FEED OUROE TAUO
URR in High-Level Cell
c «io aap c»XL£CTioa TANK RZFLUI
J E 4^1 Ot>i- fllO TANK REFLUX COMOXN9CR
-------
EQUIPMENT LIST
•> CU.UMU
tO* {OL*MM
w« •<*« POT
oi..w» •((& *oi
.f^* OuAMQ »KO ratjfl KIAO «»
;•*;• *»t^
" :X.i:iCM tww DtUMTIO
:>-<^» i»Ka KfM POT
x.to y* us « o POT
o«-| S'*tot*o*, (AM
tO., IV* SflTftl r«() TMM
• .im *•.«*» »»*«
c«* *ac*>i3CNT cvu ezcwrta
.-o «!:-.-. I »|40 POT
• •« «i?»;.i •<*» POT
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tOUCUT >CAO P
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vol..
URRin High Intermediate-Level Cell
-------
EQUIPMENT LIST
«»-;-«oc »«i •aatessfv
V* C-tAC MC »(UC1Q**TO3
4J-CO VX«<»' ft*)C« ll«*»"* *
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4IM-4J C«»"6»S
J4-I-1K RtCCMtMC *CiD COACH
4l-(-*Ot "Mill »OiVtliI OuMSIM
4»-« 4» NO I lOC'Wt $CB«M*«
49-B-4M 00 I <00>xf KRuOCin
DO-C-OiO OwPLCn **L«t *OT
CD-0-0" i*«»Ll«* **t-vi POT
OC-O-eiT SAMPLE" •**.»( ^T
CO-O-OJT i*u^i.(* **tvt *c*
URR in Intermediate Level Cell
-------
CO
I
CO
COUIPttCNT LIST
». Of «
-------
a
i
vo
TOO TZO
TOZ-7S7
7os-?eoA/a
PLUTONIUM
soa-soi
OS3A.-SOI
ssfl-sos
0 SSA-SO4
38A-303
0 3 6 A- JOG
S6A-307
S8A-3O3
360-3000 THRU f
300-3084 THRU f
360-3 IO a THOUr
ISO-Ill & THRUF
036D-3I2 A THRUF
0360-319 A THRU?
036D-3I4A THRU?
0360-3134 TMHUF
3SO-3IG
36D-3I7
36G-3OO
36GJOB
36G-3IO
•6G-3M
n!6G-3!2
0 36G-3I3
DS6G-SI4
036G-3I5
366-316
366-31?
H36G-3IO
75B-7SOA/S
0 7SB-7SI
79K-7SO
73H-7SI
79K-7S2
7SK-7S3
75K-751
0 T5K-7S5
D 7SK-7S6
0 7SK-787
7&K-7SB
75K-739
75K-7SO
0 73 K -761
0 73K-762
OTSK-78S
75K-764
TSK-76S
0 7SR-766
0 78K-7S7
75K-760A/0
75 K -76 9
000-OIOS
80D-BI06
eoo-eior
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0 800-8109
0 800-6110
D 800-0111
0 OOD-OII2
C a EHPANSioa TOia
C a CSCHAN6ER
c a cnaiLaTicn p\ia?s
NITRATE srcaaeE
PUHP 6LO»S DOT 47 1
P-JMP 6LOVE DOT^Z
VALVE CLOVE 0 OS *• 1
VALVE GLOVE oon^a
SAdPLE 6LCVE OCX *T 1
SAUPLE OLOVE BOS &*
PM LCADOUT GtOVE DOB
UAiaTEOANCE 6LOVE COB
pLj-oniuw nrroATE STORAGE TOKOS
CLUT5Wb« IglTDATE STORACZ TARKJ
PLUTOBIU3 W7«ATE STORAaE TABUS
pLuTcamei HITRATE STORASE TANKS
PLJTONIUU NITDATE STODAGE TANKS
PLUTCMlUa NI'DATE STORAGE TANKS
PLUTCftllua dlTRATC STORAGE TANKS
PLOTCMIUH QTOATE STORAGE TANKS
SOLDER h.TOUTE SAMPLE TANK
1C i-Tq NITRATE dCASjRING TANK
P.ttS TRANSFER PUMP
f.a.s. ra&usFiR ?uur
PBS TRANSFER PUtyP
PM3 TQAHSfiR PUMP
fKLO TRANSFER PUt3>
PBS TRANSFER PUMP
P.tJ.S. TRANSFER futit
P. N.S. TRANSFER PUMP
PNLO TRAH87ER PUMP
SUMP PUUP PB CELL J(f\
SUMP Piap PH CELL &l
TANK VAULT COOLING UNIT
TANK VAULT COOLING UNIT
PNLO GLOVE BOS INLET FILTER
PNLO GLOVE BOX EXH FILTER
PUMP GLOVE eon INLET FILTER
t-UUP GLOVE BOBJEKH. FILTER
PUMP GLOVE BOX INLET FILTER
PUMP GLOVE BOS INLET FILTER
PUMP GLOVE BOX EKH. FILTER
PUMP GLOVE BOX INLET FILTER
VALVE GLOVE BOX INLET FILTER
VALVE GLOVE BOX EXH. FILTER
VALVE GLOVE BOX INLET FILTER
VALVE GLOVE BOX INLET FILTER
VALVE GLOVE BOX EXH FILTER
VALVE' OLOVE 80X INLET FILTER
SAMPLE GLOVE BOX EXH. fILTER
SAMPLE GLOVE BOX INLET FILTER
SAUPLE 6LOVE BOX EXH. FILTER
SAMPLE GLOVE SOX INLET FILTER
TANK VAULT RCCIRCUL. FILTER
EMERGENCY INLET FILTER
SAMPLE VALVE POT
SAMPLE VALVE POT
SAMPLE VALVE POT
SAMPLE VALVE POT
SAMPL E VALVE POT
SAMPLE VALVE POT
SAMPLE VALVE POT
SAMPLE VALVE POT
+, J® L7<2 SPliLt
URR in Plutonixim Loadout Cell
-------
DO
I
URR from High-Level Liquid Waste Tank
-------
J-"
M
Failure of Heat Transfer from HLLW
-------
'oaej&ya
w
I
VY«
"'X,
Failure to Institute Emergency Reflood
-------
to
M
U)
COOi I
'»yVA
N.O.
-------
URR from Resin Reaction
SES
r \
RCffiw
r\
/so »*«««?«
'ra
>
-------
w
I
S-1
ui
Id.
Solvent Fire in Partition and Purification
-------
(x)
Red Oil Explosion
-------
W
M
-J
Steam Explosion
-------
f \
r \
00
I
!-•
00
&ccidlest in Process 'C®11 °
-------
Controller Fails Unsafe
Vpai(,UK?t3/
V y
/O
Failure of Personnel Intervention
On-Site Power Failure
B=19
-------
Acid Fraction Overhead
Pump Failure
B-20
-------
APPENDIX C
Descriptions of Accidents Experienced in the Nuclear
Energy Field and Chemical Industry Relating to Anticipated
Credible Events at a Fuel Reprocessing Plant.
C-l
-------
This Appendix reviews the types of accidents which have
occurred in AEC operational activities relating to opera-
tions performed at a nuclear fuels reprocessing plant.
The information is drawn from AEC field office reports
34
as described in USAEC Report WASH-1192 and supplemental
information supplied by the USAEC Division of Operational
54
Safety. There is a tendency to emphasize exposures and
contamination from criticality incidents in discussions
concerning radiation hazards because of the AEC"s involve-
ment. These occurrences, however, are in the minority in
the overall picture of potential environmental impact.
By far, the majority of potential incidents which could
occur during activities at a fuel reprocessing plant,
that might have an environmental impact are comparable,
in kind, to those occurring in the chemical industry.
To broaden the statistical base for accident probability
estimates for this study, available relevant accident case
histories from the chemical industry are also included in
21
this Appendix. These case histories, covering the period
1951 through 1972 are voluntarily submitted to the Manu-
facturing Chemists Association for publication in an
endeavor to improve plant safety in the industry. This
compilation, to be sure, is not complete. It does, how-
ever, indicate the types of accidents that have occurred
in the operation of a reprocessing plant or related
facility and was utilized in selecting the hypothetical
accidents considered in this study.
The abbreviations used for the USAEC field offices are as
follows:
AL Albuquerque Operations Office
BH Brookhaven Office
c-2
-------
CH Chicago Operations Office
GJ •Grand Junction Office
HA Hanford Operations Office
ID Idaho Operations Office
LAR Lockland Aircraft Reactors Office
NV Nevada Operations Office
NY New York Operations Office
OR Oak Ridge Operations Office
PNR Pittsburgh Naval Reactors Office
RL Richland Operations Office (Formerly HA)
SAN San Francisco Operations Office
SNR Schenectady Naval Reactors Office
SR Savannah River Operations Office
SNPO-C Space Nuclear Propulsion Office-Cleveland
SNPO-N Space Nuclear Propulsion Office-Nevada
C-3
-------
lo Critical!ty Accidents
a. Oak Ridge, Tenn., June 16, 1958
A nuclear accident occurred in a 55-gallon stainless steel
drum in a processing area in which enriched uranium is re-
covered from various materials by chemical methods in a
complex' of equipment. This recovery process was being re-
modeled, at the time of the accident.
The incident occurred while they were draining material
thought to be water from safe 5-inch storage pipes into
an unsafe drum.
Eight employees were in the vicinity of the drum carrying
out routine plant operations and maintenance. A chemical
operator was participating in the leak testing which inad-
vertently set off the reaction. He was within three to
six feet of the drum, while the other seven employees were
from 15 to 50 feet away.
Using special post hoc methods for determining the neutron
and gamma exposures of the employees involved, it was
estimated that the eight men received? 461 rem,, 428 rem,
413 rem, 341 rem, 298 rem, 86 rem, 86 rem, and 29 rem.
Area contamination was slight, with decontamination costs
amounting to less than $1,000.
18
During this incident 1,3 x 10 fissions occurred.
-------
b. Los Alamos, N. Mex., Dec. 30, 1958
The chemical operator introduced what was believed to be
a dilute plutoniurn solution from one tank into another
known to contain more plutonium in emulsion. Solids
containing plutonium were probably washed from the bot-
tom of the first tank with nitric acid and the resultant
mixture of nitric acid and plutonium-bearing solids was
added to the tank containing the emulsion. A criticality
excursion occurred immediately after starting the motor
to a propeller type stirrer at the bottom of the second
tank.
The operator fell from the low stepladder on which he was
standing and stumbled out of the door into the snow. A
second chemical operator in an adjoining room had seen a
flash, which probably resulted from a short circuit when
the motor to the stirrer started, and went to the man's
assistance. " The accident victim mumbled h'e felt as
though he was burning up. Because of this, it was assumed
that there had been a chemical accident with a probable
acid or plutonium exposure. There was no realization that
a criticality accident had occurred for a number of min-
utes. The quantity of plutonium which actually was pres-
ent in the tank was about ten times more than was supposed
to be there at any time during the procedure.
The employee died 35 hours later from the effects of a
radiation exposure with the whole-body dose calculated to
be 12,000 rem + . '
C-5
-------
Two other employees received radiation exposure of 134
and 53 rem, respectively. Property damage was negligible,
(See TID-5360, Suppl, 2, p0 30? USAEC Serious Accidents
Issue #143, 1=22-59.) s
GO Idaho Falls, Idaho, Oct. 16, 1959
A nuclear incident occurred in a process equipment waste
collection tank when an accidental transfer was made of
about 200 liters of uranyl nitrate solution, containing
235
about 34 kilograms of enriched uranium (91 percent U ),
from critically safe process storage tanks to a geometric-
ally unsafe tank through a line formerly used for waste
transferSo
Limited visual inspections and test indicated that no sig~
nificant property damage or loss resulted beyond the
approximately $60,000 cost to recover contaminated uranium
solution resulting from the incident.
Of the 21 personnel directly involved in this incident,
seven received external exposures to radiation. The ex-
posures were 8, 6, 3,95, 1 = 50, 1-38, 1,17 „ and 1.17 renu
Two individuals also received external exposures to the
skin of 50 rem and 32 rem. No medical treatment was re=
quired for the 21 personnel involved,
do Idaho Falls, Idaho, Jan,. 25, 1961
A nuclear excursion of approximately 6 x 10 fissions
occurred in a first-cycle product evaporator at a chemical
processing plant. The criticality accident resulted when
C-6
-------
"—a^ solution- ot~ en r irctved~\ir^ny1r-n±tra"te:~a'
-------
sump and was drawn into the' transfer tank through a tem-
porary line between this tank and the sump.
When the excursion occurred, radiation and evacuation
alarms sounded. All but three employees left the building
immediately, according to well-prepared and well-rehearsed
evacuation plans. Fortunately, they were not in close
proximity to the involved system nor in a high radiation
field.
The course of the nuclear reaction involved initial erit-
icality (10- fissions)? a subsidence,0 one or more later
peaks? and after approximately one~half hour, a declining
rate of fission, which terminated in a subcritical condi-
tion 37 hours later. The total number of fissions was
approximately 8 x 10 . •
Of the 22 persons in the building at the time, only four
employees, those who were in the room with the system,
were hospitalized for observation„ Three of them were
the system operators, who were in close proximity to the
excursion, and who received estimated radiation doses of
110, 43, and 19 rem. None of them showed symptoms defi-
nitely referable to their radiation exposures. The fourth
was sent to the hospital only because he was in the room
at the time of the incident.
Some fission product activity, airborne via the vent
system and the exhaust stack, was detected in the atmosphere
for a brief period after the accident. The physical damage
amounted to less than $1,000.
C-8
-------
f„ Wood River Junction, R.I., July 24, 1964
Because of startup difficulties an unusually large amount
of highly enriched uranium-contaminated trichloroethane
(TCE) had accumulated at United Nuclear"s Scrap Recovery
Planto The recovery was by mixing the TCE with a sodium
carbonate solution. On the day of the incident the pro-
cess was shifted to an 18 in« dia. by 25 in. deep tank .
to try to catch up with the backlog. The plant evaporator
failed resulting in a plug of uranium nitrate crystals in
the converting line. This plug was dissolved with steam
and the concentrated solution was drained into a. polyethy-
lene bottle, i; This bottle was mistaken for trichlorethane
and the operator poured it into the tank of solution. Cri-
ticality was reached in a burst of 10 fissions creating
a flash of light. 1/5 of the solution was ejected and the
operator knocked to the floor. He ran to the emergency
building 200 yards away but having received about 10,000
rad, died 49 hours later.
Later two men entered the area to drain the solution into
saf® containers. When the stirrer was turned off, the
geometry change resulted in a second criticality of 2-3x
10 fissions and these men received 60 and 100 rads.
g; Oak Ridge, Tenn., Jan. 30, 1968
Unexpected criticality was achieved in a volume of an
233
aqueous solution of a salt of U during a series of rou-
tine critical experiments in progress in a well-shielded
assembly area of a critical experiments facility. The
C-9
-------
criticality-radiation alarm system functioned as designed,
the evacuation of personnel from the building was prompt
and orderly, and the excursion was terminated expeditiously
by a negative coefficient of reactivity and was prevented
from recurring by the action of the safety devices. The
fission yield was 1.1 x 10 . Gamma-ray sensitive personnel
dosimeters read immediately following the excursion showed
no direct exposure greater than 5 mr to any person present.
There was no property damage or loss of fissile materials.
An estimated 100 cm of solution (15 g of U) were spilled
when a rubber-stoppered connection immediately above the
sphere was dislocated.
The purpose of the particular experiment in progress was to
establish the critical concentration of a sphere of the
solution of uranyl nitrate surrounded by a thick water
reflector. In the course of approaching criticality by
incremental additions of solution, a small volume of air
was observed entrapped in a flexible transparent tube.
Supercriticality occurred during an attempt, by remote man-
ipulation of liquid levels, to remove the air.
C-10
-------
2.
Reporting AEC
Field Office
OR
LAR
1960
HA
1962
OR
1963
HA
OR
OR
RL
Multiple circuit breaker fail-
ure led to severe electrical
fire. Property damage $86,000.
No exposures.,
Electrical fire due to severe arc-
dng on the lineside of heater break-
ers o Property damage $30,000. No
exposures.
Fire and explosion in pyrophoric
metal contents of a chemical dis-
solver, off-gas filter, and rela-
ted process equipment., Contamina-
tion spread to cell, canyon and
crane. Cause (s) of the accident
not established. Property damage
$250,443. No exposures.
Fire occurred in ventilation system,
probable cause electrical spark.
Property damage $24,700. No expo-
sures.
Air ventilation equipment failure.
Property damage $10,000. No expo-
sures.
Explosion and fire in cell. Pro-
perty damage $2,900,000. No expo-
sures.
Fire (definite cause undetermined)
originated in building exhaust
system and was confined to labora-
tory hoods and exhause system; smoke
damaged building. Property damage
$43,400. No exposures.
Fire (definite cause undetermined)
in plutonium purification facility.
Pu contamination in immediate
area of fire. Firemen received
slight skin contamination, readily
removed. Costs related directly
to fire $85,400,- decontamination
costs $251,300; overhead related to
direct losses $60,300.
C-ll
-------
Fires (Continued)
Reporting AEG
Field Office
SR
1965
RL
1966
SR
OR
1968
RL
Fire (definite cause undetermined)
occurred around an anion exchange
column in hot canyon. Fire caused
airborne contamination to crane
used for remote maintenance. Water -...
to quench fire damaged electric
motors. Property damage $21 ,.000.
During an aluminum jacket dissolu-
tion in a dissolver, an exothermic
reaction involving ammonia and/or
hydrogen occurred with an electric
heater, through which these gases
were accidentally vented. The
reaction, which continued for 3 hours,
totally destroyed the heater. Prop-
erty damage $7,200. No exposures.
A fire occurred when a drying oven
overheatedo Faulty loading blocked
the thermostat sensing element,
causing it to indicate erroneously
low temperature and call for addition-
al heat. Property damage ($6,000)
was confined to the room of origin.
No exposures.
A fire, of undetermined cause, occurred
in a laboratory. It was confined to
one hood and a section of ductwork
because of the successful operation of
a sprinkler head, a fire damper in the
exhaust system and other fire protec-
tion controls. Property damage $5,5©0.
No exposures.
An electrical short circuit and the
resulting power arc in the main elec~
trical switchgear damaged two breakers
extensively (, when the lights went out
and the building ventilation stopped.
Emergency actions were.taken to pre-
clude any contamination spread. Opera-
tions were curtailed for two and one-
half days while repairs were being
made. Property damage $34,000, No
exposures,
C-1.2
-------
Fires (Continued)
Case History No. - Manufacturing Chemists Association
41
129
141
150
255
341
348
612
643
699
701
1025
1217
1234
1966
1970
"Static SparX Flashes "Empty" Sty--
rene Drum"
Fire Due to Static Spark - Benzene
Toluene Vapors - Flash Fire
Escape of Vapor From Condenser
Ignition of Solvent Vapors -^Employee
Burned
**
Solvent Fire
Electric Mixer Fires Solvent
Waste Solvent Fire
Flash Fire in Exhaust Duct
Toluene - Static Fire
"Boil Over" - Flash Fire
Flammable Solvents - Electric Motor
Kerosene Vapor Flash - Synthesis
Kettle
Zirconium Fines Flash Fire
"Static" Ignition of Flammable Solvent
Solvent-Vapor Flash Fire
C-13
-------
Explosions
Reporting AEG
Field Office
SR
HA
1960
AL
OR
OR
OR
OR
AL
Gasket on head of secondary condenser
in unit failed. Relief valve vented
open due to overpressure. 4-5T of
H2S gas released to atmosphere. $7,000
property damage. No exposures.
Plutonium glovebox explosion. $9,500.
property damage., No exposures.
Autoclave explosion. Property damage
$4?000. No exposures.,
Drybox explosion. Property damage
$1,933. No exposures.
Chemical explosion in innercycle evap-
orator. Property damage $350,000.
No exposures.
An explosion occurred in a digester.
Property damage $10,000-4-. No exposures.
Hydrogen gas explosion occurred in
gas furnace enclosure in metal plant.
Property damage $5,000. One employee
suffered serious injuries.
Explosion occurred in a uranium sinter-
ing furnace located in a foundry. Major
structural damage to furnace and build-
ings o. " Property damage $20,000. No
exposures.
The accidental discharge of radioactive
material into a room occurred as a
result of pressure buildup in a dry°
box. This was due to an inlet solenoid
being locked in the open position and
a venting solenoid being closed due to
a malfunction. The pressure built up
to a point that one of the drybox gloves
blew out, thereby releasing radioactive
particulate material into the room.
Property damage $31,360. Eleven persons
received minor exposures.
C-14
-------
3.
KxploHJona (Continued)
Reporting AEC
Field Office
AL
1963
1964
ID
AL
1965
RL
RL
Pressure buildup in closed caustic
scrubber system forced airborne
radioactive material into room. Prop-
erty damage $4,016. One employee
received 71 rem exposure to bone.
Area contaminated.
Low-level spread of plutonium con-
tamination from gloveboxc Property
damage $25,451. No exposures.
Chemical explosion in metal hood
when methanol vapors reached flash-
point. Two sets of gloves were
shredded by the explosion. Contam-
ination spread in operating area.
Property damage $34,922. Three em-
ployees received slight contamina-
tion.
An explosion occurred in a boiler
during an attempt to relight the
oil-fired burner with a kerosene
torch after the automatic re-igni-
tion system failed to function.
Property damage $75,000. No expo'sures.
Plutonium contamination spread follow-
ing an explosion and fire occurring
in a glovebox when cleaning fluid
ignited. Ten employees left the
building immediately. Prompt show-
ering easily removed all skin con-
tamination. One employee received 10%
of a maximum permissible body burden
(bone) of Pu-239 by inhalation. Con-
tamination did not spread outside the
building. About 90% of the cost
($76,800) was incurred for decontam-
ination .
C-15
-------
•t
•V
*-'•
3o Explosions (Continued)
Reporting AEC
Year Field Office
1965 AL An explosion and fire occurred when
acetone fumes from a "cocoon" used
in a glovebox paint-stripping opera-
tion, contacted a hot muffle furnace
in another part of the glovebox line,,
Plutonium contamination spread to
adjacent rooms and the second floor.
12 employees required skin decontam-
ination? none received internal rad-
iation exposureso Property damage
costs ($23«,253> was for decontamina-
tion of facilities,
CH An explosion resulting from the ig-
nition of a hydrogen-air mixture,
the hydrogen apparently evolved from
nickel~iron batteries, occurred in
the equipment airlock joining a reac-
tor building and a fuel cycle facility.
No radioactive material was involved.
Property damage $22,600.
AL An undetermined small quantity of
Pu-238 was released when a double=
contained vessel, nearly full of
drybox seived material exploded, dis-
persing a quantity of the waste
material into the laboratory. Property
damage $19,100. Mo exposures.
C-16
-------
3. Fires (Continued)
Case History No. - Manufacturing Chemists Association
103 Nitrating Operation Explosion
116 Storage Battery Explosion
128 Nitrogen Peroxide - Cyclohexane
Mix Explosion
131 Nitric Acid - Waste disposal Ex-
plosion
163 Reaction in Solvent Recovery Tank
223 Laboratory Explosion - No Injuries
258 Explosion - Ignited by Vacuum Cleaner
347 Hydrogen dessicator = Drainage Trench
Explosion
569 Runaway Nitration Reaction
576 Hydrogen Compressor - Explosion
611 Oil Vapor Explosion
678 Explosion in Nitrobenzene Recovery
Kettle
679 . Unsafe "Fail Safe" Flame Safety
Device
703 Explosion in Vent Stack - Static Gen-
eration
859 Spilled Four Gallons Solvent on Lab-
oratory Floor - Fire
880 Chemical Fire - Azido Compound
976 Silver Complex Detonation
987 Explosion and Fire - Lead Azide
988 Tank Explosion
1048 Explosion - Silver Oxide
C-17
-------
3» Explosions (Continued)
Case. History Nov - Manufacturing Chemists Association
1068 Gas Explosion - lighting Burner
1097
1105
1310
131L
1496
1499
1554
1733
1957
1958
Explosion - Hydrogen gas vent stack
Americium Solution Shipping Container
Explosion
Hazardous Solvent Causes Explosion
in,a. Plutonium Fuels Laboratory
Glovebox (Furnace)
Flammable Vapor Explosion - Slurry
Mix Tank
Nitration Explosion - Organic Inter-
mediate Mixed with Nitric and Sulfuric
Acids
Drums burst from internal pressure -
Accumulation of Hydrogen within
Tank Rupture - Organic Material
Nitrated with Nitric Acid
Ammoniacal Silver Nitrate Explosion
Laboratory Explosion - Silver Nitrate,
Ammonium Hydroxide, Alcohol Silvering
Solution
Disposal of Deposits of Metal Azides
Mix Tank Explosion
C-18
-------
Fuel Receiving & Storage
Reporting AEC
Field Office
SR
1964
CH
1967
RL
During shipment of irradiated fuel
elements, 30 to 40 gallons of con-
taminated water leaked from the
casko $24,000 cost due to decontam-
ination of area. No exposures„
Broken valve on autoclave, housed
in lead shipping cask, allowed con-
taminated water to seep out of cask
during transit, contaminating con-
tainer and truck floor» No property
damage.
A diesel locomotive collided with a
cask car during coupling operations,
due to the inattention of the die-
sel 's engineer. The cost of $5,124
was for repairing the locomotive?
the cask car was not damaged. .No
exposures.
C-19
-------
Waste and Product Storage
Reporting AEC
Field Office
SR
SR
SR
1960
SR
1961
HA
HA
1963
AL
AL
Leaking compression fitting,
damage. No exposures.
$20,000
Solution leaked from the loosened
flange during maintenance work on
a waste evaporator in hot canyon,
vaporized and contaminated crane.
Property damage $129,324. No exposures,
Loose contaminated particles on the
lid of a waste burial box were
scattered by the wind, contaminating
the ground^ locomotive and spacer
car. Property damage $5,200. No ex-
posures o
Contaminated cooling water discharged
from canyon onto floor. Property
damage $250,000 (due to decontamina-
tion). No ovesexposures.
Uranyl nitrate (1355 Ibs. of depleted
uranium) lost to ground when tank
trailer was overfilled due to misun-
derstanding between regular operators
and their lunch relief. Property
damage $13,000. No exposures.
Approximately 1,089 pounds of depleted
uranium lost to chemical sewer in
plant. Property damage $9,000. No ex-
posures .
Leak in line carrying high-level plu-
tonium solution caused contamination
of building and equipment. Property
damage $8,364. No exposures.
Spill of contaminated nitric acid sol-
ution. Property damage $5,662. No
exposures.
C-20
-------
5-.
Waste and Product Storage (Continued)
Reporting AEC
Year Field Office
1964
CH
ID
1965
SR
AL
SR
SR
SR
Clean water, being used to test two
new waste tanks, was contaminated by
condensation from contaminated vent
line connected to one tank. Water
subsequently drained onto asphalt
surface, contaminating it and drain-
age ditcho Property damage $6,075.
No exposures.
During steam flushing to remove rad-
ioactive contamination from pipelines
to permit tie-in to new lines, leak
developed in hose coupling„ Contam-
inated fluid and steam issuing from
leak were rapidly dispersed by high
wind over approximately 10 acres.,
Majority of $12,884 cost due to
cleanup.
Leaking nitric acid corroded canyon
cell equipment beyond repair. Pro-
perty damage $6,000. No exposures.
While attempting to activate a prod-
uct transfer line, contaminated acid
solution was sprayed out of a flanged
union that had not been tightened.
Three contaminated employees were
readily decontaminated. Property dam-
age $7,557 for decontamination and for
replacing contaminated equipment.
Process water (2400 Ibs) was lost when
it leaked through an unseated sleeve.
Property damage $33,600. No exposures.
Process water (700 Ibs) was lost when
it leaked through an unseated sleeve.
Property damage $9,800. No exposures.
A cooling coil in a vessel developed
a leak and allowed contaminated solu-
tion from a tank to enter the cooling
water system when the coil was pres-
surized. The cost for cleaning the
system and associated work was $19,-500
No exposures.
C-21
-------
Waste and Product Storage (Continued)
Reporting AEC
Year Field Office
1966
RL
RL
RL
1967
AL
An estimated 420 Ibs. of uranium
solution were lost to radioactive
waste through a milling tank over-
flow, caused by the failure of a
normally closed supply line valve.
Property damage $7,200., No exposures-
During the repair of an air circula-
tion valve, approximately 10 gallons
of high°level radioactive waste solu-
tion were spilled onto the floor.
Three employees, wearing protective
clothing, were sprayed with droplets
of the solution, but were readily
decontaminated. Property damage
$19,746.
Less than 5 grams of concentrated
plutonium nitrate solution spilled on
the elevator floor when a product
receiver assembly overturned and the
lid of the inner container came off.
Cost for decontamination was $13,443.
No exposures.
Abandoned storage vessels inside a
stainless steel glovebox were being
flushed with 7-9N nitric acid to re-
cover plutonium nitrate. During this
operation, the air monitor alarmed,
and the odor of nitric oxide fumes was
detected. Shortly thereafter, a puddle
of dark liquid (plutonium nitrate solu-
tion) was seen on the ofloor under the
glovebox. The solution had leaked from
one of the storage vessels into the
glovebox well and thence onto the floor
of the room. 3 contaminated personnel
were readily decontaminated by shower-
ing ? the cost for decontamination of
the storage vessel area was $16,465.
-------
5.
Waste and Product Storage (Continued)
Reporting AEC
Year Field Office
1967
SR
1969
SR
SR
Radioactive liquid waste was stored
in an underground' tank., ' The pipeline
for the concentrate entered the stor-
age tank through a shielded riser/
extending from the top of the tank to
approximately one foot above ground.
Access plugs were sealed with mastic
compounds. The inlet pipe entered
the riser horizontally below ground
and terminated with a valve near the
center of the riser. When the riser
became plugged with concentrate crys-
tals below the inlet line, the liquid
flow reversed and forced its way
through the access plugs. Approximately
13 curies of radioactive liquid waste,
primarily cesium, were released to plant
streams but sampling showed that radio-
activity concentration standards were
not exceeded in streams beyond the
plant boundary. The cost for decontam-
ination of ground in the vicinity of
the tank was $49,179. No exposures.
Approximately 20 millicuries of air-
borne radioactive contamination
(mostly curium-244) were released via
an exhaust stack and were spread by a
northeasterly wind across the roof of
a building and along a line leading
from the main entrance of the building
to a parking lot. The level of-radio-
activity on the roof was 4 x 10 d/m/lOOcm"*
and on vehicles inside the area fence.
1.5 x 104 d/m/100cm2. The highest level
of activity outside the area was approx-
imately 5000 d/m/100cm2. All activity
was contained within the plant's boun-
dary. The cost for decontamination
was $37,506. No radiation exposures.
Acidic waste solution (approximately 8200
pounds), which was being processed for
neptunium-237 and plutonium-238 recovery,
was lost when inadvertently transferred
to an underground waste system due to a
leaking valve in the stream supply to a
transfer jet. Property damage $32,000.
No exposures.
C=23
-------
5. Waste and Product Storage (Continued)
Reporting AEC
Year Field Office
1970 SR A solution containing 20 grains of
curium-244 and americium-243 was
transferred by mistake to the waste
system- Property damage $124,523.
No exposures.
RL Minor cerium-cesium contamination
resulted from a routine change of
a stack filter in the 300 area-
Over 200 employee's shoes were
checked and none were found to be
contaminated. Walkways and roadways
were washed down? no radioactivity
was found in surveys beyond the 300
area. No property damage? no ex-
posures.
-------
5. Waste and Product Storage (Continued)
Case History No. - Manufacturing Chemists Association
254
298
1088
1498
1716
Pressure Build-Up in Pfandler Kettle
Operation
Collapse of 20,000 gallon SoS. Tank
Implosion in Still During Cleaning
Operation
Gross Leakage of Plutonium Nitrate
Solution from Stored Polyethylene
Bottles
Dry Radioactive Waste Unloading
Incident - Localized Radioactive
Dust Escape
C-25
-------
Natural Phenomena
Reporting AEC
Year Field Office
1959
1960
OR
SR
LAR
SR
SR
OR
AL
1963
PNR
1965
CH
CH
AL
Lightning damaged transformer.
Property damage $6,500. No exposures.
Lightning damaged two 750 KVA trans-
formers. Property damage $13,750.
No exposures.
Wind damage to aluminum side wall of
building. Property damage $7,500.
No exposures.
Hurricane damaged water dam. Property
damage $50,000. No exposures.
During an electrical storm, lightning
struck two 200-hp pump motors in an
out-of-door pump pit. $6,000 cost for
rewinding of burned out motors. No
exposures.
During violent storm, severe power
system disturbance caused oil circuit
breaker failure. Property damage
$18,132. No exposures.
High-velocity winds caused circuit
breaker failure in substation, result-
ing in fire readily controlled by fire
extinguisher. Property damage $8,200.
No exposures.
Severe winds during electrical storm
damaged roofs, stacks, ventilation
ducts, trees and fences. Property
damage $9,400. No exposures.
Four transformers were damaged by
lightning. Property damage $35,000.
No exposures.
Lightning caused the destruction of
a breaker and the burning of a cubicle.
Property damage $8,000. No exposures.
Repeated lightning strikes damaged
transformers. Property damage $5,400.
-------
6.
Natural Phenomena
Reporting AEC
Year Field Office
1966
1967
AL
AL
1970
SR
Roofing destroyed by high winds.
Property damage $47,000. No exposures.
A severe wind and hailstorm, with
winds in the range of 80 to 100 miles
per hour and hailstones the size of
oranges, caused extensive roof and
other structural damage to numer-
ous buildings, disrupted utilities,
demolished a warehouse wall, leveled
security fencing and caused severs
vehicle damage. Property damage
$1,872,000. No exposures.
Water supply lines, drainlines and
traps, water-jacketed equipment,
heating and cooling coils, instruments,
gages, and fire sprinkler lines froze
in numerous plant locations during a
period of extremely low temperatures,
unusual and unexpected in the area.
Property damage $38,200. No exposures.
C-27
-------
A©IENCV
Washing?©*). O.C.
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POSTAGE ANS PEiS PAID
U.S.
AGENCY
gPA-335
ClASS BULK
------- |