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

-------
   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

-------
                                           » 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

-------
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.

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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

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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

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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

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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

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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

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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

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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

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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

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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

-------
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 2.   "Environmental Survey of The Nuclear Fuel Cycle", U.S.
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 3o   Allied-Gulf Nuclear Services, Inc., "Barnwell Nuclear
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 4.   J.  A. McBride, "Reprocessing, Transportation and Waste
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 60   General Electric Co., NEDO°10178,  "Final Safety Analy-
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 7o   Allied-Gulf Nuclear Services, Inc., "Safety Analysis
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                              128

-------
           (Confd)
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15o   Ho Jo Otway and R« Co Erdmann, "Reactor Siting and De-
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16o   Co Starr, "Benefit-Cost Relationships in  Socio-fechni-
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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-
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 730414-P2, Ann  Arbor, Michigan, AprS.1 9°ll,1973o
                       130

-------
REFERENCES (Cont'd)
33.  Northern States Power Co.,  "Final Safety Analysis Report,
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34,  "Operational Accidents and Radiation Exposure Experience
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37-,  G. Mo Nichols, "Decomposition of the Tributyl Phos-
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     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
U)
                                                                                  '•Xw
                                                                                 H\ /*-%*
                                  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
                DO-a-coi
                CG-7-CX.
                ID -7- 101
                IS-T-IOQ
                in-f-ioD
                10-VlOO
                18-T-IOT
                00-T-OOI
                C3-T-C90
                «a-?.an
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                C3-T-000
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tPOtaOSOLVED
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    i
(PSOIOOOLVOQ HUflTan/COSUtQ
MOO eOBCHaTOATOQ QU08ILOQ
MAO coocnaTDflToa
 
-------
                           EQUIPMENT  LIST
to
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                   IT 0 IO1   DO  I  TOO AOJUOTnlCJT WHn
                   I? 0 IO«   00. a  •       ••
                   IT 0 III    0I0001VEQ FLUOH ACCUafttflTOO
                   IT 0 111    fCBO CUQCZ TAOH
                   IT 0 104   OJ3OT.WO reoOUCT IOCS POT
                   21 o lot    MA ft to matt
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                   II 0 IBS   tOO HCAO POT
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                   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 «»
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                                                             URRin  High  Intermediate-Level  Cell

-------
                           EQUIPMENT LIST
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                     OC-O-eiT   SAMPLE" •**.»( ^T
                     CO-O-OJT   i*u^i.(* **tvt *c*
                                                                                          URR   in  Intermediate   Level  Cell

-------
CO
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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
eoo-sioa
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

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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

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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

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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

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          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

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          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

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          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

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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

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                                         •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

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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

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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

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           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

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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

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           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.

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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

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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.

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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

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           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.

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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

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