EMVIRQM1-JTAL PROTECTION AGE-ICY
WASHINGTON, D. C. 20460
November 1974
Comments by the Environnental Protection Agency'
on
Reactor Safety Study
AN ASSESSMENT OF ACCIDENT RISKS IN U.S. CCMERCIAL
NUCLEAR POWER PLANTS
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^-\>K4 V UNITED STATES ENVIRONMENTAL PROTECTION AGENCY
**'«< w.uti''1'*" WASH ING! ON. D C 2CMGO
27NOV 1974
Mr. Saul Levinc
Project Staff Director
Reactor Safety Study
U.S. Atomic Energy Commission
Washington, D.C. 20545
Dear Mr. Levine:
The Environmental Protection Agency's comments from the initial
phase of its review of WASH-1400 ("An Assessment of Accident Risks in
U.S. Commercial Nuclear Power Plants") arc transmitted with this letter.
Because the assessment reported in WASH-1400 is expected to he
a principal step toward establishing the accident risk associated with
nuclear power plants, we are reviewing it in two phases. The first phase
is represented by the enclosed preliminary comments based on a. two-month
review effort. The second phase will include an in-depth review of
selected aspects of the study with technical assistance being provided
to EPA through a contract with Intermountain Technologies, Inc. This
second phase should be concluded by May 1975, at which time our intent
is to issue a final report detailing all of our comments. During this
period of continuing review we hope to maintain a close liaison with the
Atomic Energy Commission so that our final report will reflect an up-to-
date awareness of any resolution attained regarding comments by EPA or
others on the draft report.
We have reviewed the work plan for our continuing effort with members
of your staff as well as others in the technical community. We are also
including it as a part of our review comments so that others may be
cognizant of our planned efforts.
Our initial review indicates that the Reactor Safety Study provides
an innovative forward step in risk assessment of nuclear power reactors.
The general methodology and approach utilized in determining risk levels
developed in the Reactor Safety Study appear to provide a meaningful
basis for obtaining useful assessments of accident risks at nuclear power
plants. Certainly, significant improvements in obtaining and utilising
nuclear plant operating data could considerably narrow the uncertainty
range of risk estimates. We do, however, believe that certain aspects of
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the report require modification and information additions. In particular,
the consequence modeling assumptions appear to underestimate the health
effects resulting from the accident sequences associated with the larger
releases of radioactivity. It is uncertain what the impact of this
apparent underestimation may be on the resultant risk assessment.
Although the report does not make an absolute judgment on nuclear
power plant accident risk acceptability, the comparative risk approach
highlighted in the summary and the main volume of the study will certainly
imply an acceptability judgment to the average reader. EPA recognizes
that the comparative risk approach is a first step in addressing this
question, but by itself is misleading. However, studies in progress by
EPA and others indicate that judgments on "risk acceptability" are
extremely complex, with comparative risk evaluations representing only
one of numerous inputs which must be considered.
We are interested in the plans for application of this methodology
to other reactor systems and other components of the nuclear fuel cycle.
Certainly, we would recommend that studies of this type should be con-
sidered by the applicable AEC successor and that their intent in these
areas be publicly stated.
We would be pleased to discuss our comments with you if they require
any clarification.
Sincerely yours,
U—c_
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs (AW-558)
Enclosure
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Table of Contents
Page
Introduction and Conclusions 1
Assessment of Accident Risks... 5
Calculation of Reactor Accident Consequences 6
Accident Sequences, Reactor Meltdown
Processes and Radioactivity Releases 11
Definition of Failure Data and Patliways 16
Design Adequacy 19
Summary Report, General Ccmrients 20
Additional Garments 21
Attachment - Contract with Intermountain Technologies,
Inc. - Continuing WASH-1400 Review Tasks
A. Failure Mode Patiis Selected for Review 34
E. Critical Radiological Source Terra Parameters
Selected for Review 34
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AMD CONCLUSIONS
iteviow Perspective
The Environmental Protection Agency has completed a preliminary
review of the draft report "Reactor Safety Study - An Assessment of
Accident Risks in U.S. Commercial Nuclear Power Plants," V7ASH-1400,
prepared by the Atomic Energy Commission. Our review process will
continue through April 1975 at which time we will issue a final set of
Garments. During this period of continuing review, we hope to maintain a
close liaison with those responsible for the Study so our final Garments
will reflect an up-to-date awareness of any resolution attained regarding
comments by EPA or others on the draft report.
EPA's review of the Study cannot be considered as exhaustive in that
many of the calculational details and data base have not been checked.
Our focus has rather been one of emphasizing a review of major
assumptions, concepts, methodology and approach. Although EPA's
resources are limited when compared with those utilized in the
development of the Study, it was deemed necessary to do as comprehensive
a review as possible due to the many significant implications the Study
l\as with regard to areas of EPA responsibility. EPA's primary concerns
deal with the health and safety of the public and the protection of the
environment froir the consequences of accident releases. In this respect,
we are involved in the planning for mitigation of these potential
cons©[uences, including the development of appropriate Federal guidance,
and in assuring that the public risks incurred are societally acceptable.
Within this context we attempt to maintain cognizance of accident
analysis activities so that we can be continually aware of !x>th the
probability of accidents and the consecjuences for such accidental
releases. Due to the importance we attach to this subject and the broad
range of subject matter considered in the Study, we telieve it is
imperative that it receive a thorough and critical review by the general
technical community and the public. !7e realize that rruch of this review
that we suggest is already underway or planned. However, we feel that
corti>Tents developed on the Study should be referenced in the final version
of the Study and copies of these reviews should be pviblicly available.
In continuing its review, EPA has contracted with Intermountain
Techno logics, Inc. to assist us in the evaluation of the range of
applicability of the various analytical models and assumptions utilized
in the assessment. The preliminary work plan for this effort is
presented in an attachment to our comments. If, in the initial stages
this detailed review of the selected failure mode paths or critical
source term parameters indicates a general agreement with the Study1 s
evaluation, that portion of the investigation will be terminated and
other failure mode paths or source term parameters may be substituted in
this work plan.
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Review Format
Following this Introduction and Conclusions section, our review takes
up individual groups of volumes of the TJASH-1400 document by first
presenting general comments and then specific comments. This sequence
begins with the main volume of the Study and continues with Appendix VI
(environmental consequences); Appendices Vf VII, and VTII (accident
secjuence, meltdown processes, and radioactivity releases); Appendices I,
II, III, and IV, (definitions of failure pathways), Appendix X (design
adequacy) and the sumary volume, in that order. The last section of our
review presents Additional Comments in order of tho Study volumes
theroselves. These latter comments were not felt to be of the same level
of significance as those referred to in the previous sections of our
revie//.
Main Coirrrents and Conclusions
EPA has made a broad spectrum of specific cements on the Study,
realizing that they have varying degrees of impact on final results.
However, as the document is bound to be used as a reference for many
follow-on studies and analyses, we feel it is desirable to make it as
complete and accurate as possible in all its facets. EPA's main corrients
and conclusions, although of a preliminary nature, are as follows:
1. The Study is innovative in both its concept and methodology and
provides an innovative forward step in risk assessment of nuclear
power reactors. In this respect, the AEC is to be commended. The
general methodologies and rationale developed in thp Study to
determine risk levels appear to provide a meaningful basis for
obtaining useful assessments of accident risks of nuclear power
plants.
?. Appendix VI (environmental consequences) received particular
attention in our review due to its ^xsrtiiience to rPA concerns. This
appendix was found to be quite weak in a number of respects and not
up to the general thoroughness tliat appears to pemeate many other
sections. Our preliminary review indicates, for example, that if the
rocor.nendations of the 3HIR Report are followed, the consequences
estimated in the Study may be low, in certain cases, by factors of 2
to 5. In addition, the evacuation model assuned for the reference
case consequence calculation also appears somewhat overly optir.iistic.
.Vised on the information presentel in the Study, this could increase
consequences by at most a factor of 2 to 4 (i.e., no evacuation).
Therefore, the canbination of these factors could result in an
underestimate, by about an order of magnitude, of the consequences
Ni "associated with the "high" release accident sequences. Since the
liigh release accident sequences are significant, but not dominating,
contributors to the overall risk assessment, the resultant assessed
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risk magnitude would be increased but by a lesser factor. It is
suggested that appropriate modifications should be made or tlie
rationale for utilizing other assumptions should be provided.
Furthermore, the description of certain critical portions of the
overall calculational process should be significantly expanded to
permit a clear understanding of the relationships between the
radioactive material releases, its dispersion, population
distributions, and the resulting health effects.
3. Although the Study indicates that no absolute judgment on
nuclear power plant acceptability is intended, the comparative risk
ajjproach highlighted in the surmary nay well imply an accepta) dlity
judgment to the average reader. It should be further pointed out in
the report that the comparative risk approach is only a first step in
addressing this question and by itself can be misleading. It can be
noted that studies in progress by EPA, National Science Foundation,
and others, indicate that judgments on "risk acceptability" are
extremely complex, with comparative risk evaluations representing
only one of numerous inputs which must be considered.
4. As can be expected with such a voluminous report, a number of
apparent inconsistencies, format difficulties, and cases of
insufficient supporting information were encountered. Particularly
in Appendix II there were inconsistencies in identification of
components and levels of detail in the various fault trees and system
descriptions. There were also problems with the lack of a readily
accessible glossary of abbreviations and with inadequate cross-
referencing among appendices. It is suggested that the Study be
subjected to the necessary editing to eliminate abbreviations
v/herever possible, glossaries bo added for those abbreviations used
(e.g., foldout in Appendix I) and the cross-referencing between
appendices be improved. The formats employed in Appendices III and
VII are worthy of consideration for use in all appendices.
5. There is sorne concern relative to a lack of certainty as to what
tha follow-on actions in this program area will be. This is
intensified by the recent reorganization of the AKC and its functions
and lack of definition as to where this effort will be picked up and
continued. We would expect that some follow-on effort should te
directed toward additional verification that the design, operational,
or other variations among the 100 nuclear plants to which the Study
is applied, do not significantly affect the overall risk calculated
by the Study. Of major interest would be consideration of other
plant designs such as Westingliouse 2 and 4 loop reactor coolant
systems, and BV7R fterk II and III containment designs. Other items
such as the use of hydrogen recombiners and differing modes of
containment spray injection should also be considered for
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examination. It is realized that in many of these cases what may
appear as a significant difference on the system or component level
may not significantly change the overall risks but some documentation
of this should be presented in order to show that to be the case. A
further concern relative to continuing effort in this area but not
related to this specific Study is the application of this methodology
to other reactor systems and other components of the nuclear fuel
cycle. Certainly we would recormend that these studies should be
considered by the applicable AEC successor and that their intent in
tliese areas be publicly stated.
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ASSESSMENT OF ACCIDENT RISKS IN U.S. COMMERCIAL HUCLEAR POWER PLANTS
MAIN VOLUME
General Comments
The main volume presents a well written introduction to and summary of
various analyses presented in the supporting appendices; tlierefore,
Garments on the material within this volume are generally covered
elsewhere in this review. The discussion iji Section 5.3 pertaining to the
process of assessing release category probabilities was especially
informative. The assignment to a category release probability of a 10%
contribution from adjacent categories certainly adds a significant elerient
of conservatism to the resulting probability values. The Study also
attributes additional conservatism to the Monte Carlo process used to
assess failure rate median values. However, the degree of conservatism
attached to the Monte Carlo process throughout the Study, relative to its
ability to compensate for wide ranges in available input data, may be
somewhat misleading, especially if the log normalised data are "processed"
through a series of "and" or "or" gates. It would appear that, in such
cases, the Monte Carlo process would be expected to yielC a point estimate
similar to that attained through a straight additive or multiplicative
process of input value median value failure rates, with an associate:!
error factor. Although the Monte Carlo process is statistically correct,
a further explanation of this process indicating the differences between
it and the point estimate approach should be presented with regard to the
evaluation of associated error factors, liowever, it should be noted that
statistical techniques such as this, although appropriate analytical
rnethodology, can never conclusively shew that all critical pathways to an
accident occurrence have been considered.
Chapter 6 of the main document presents a comparison of the nuclear
accident risks to other societal risks. Although the Study does not make
an absolute judgment on nuclear power plant accident risk acceptability,
the comparative riak approach certainly implies an acceptability judgment
to the average reader. EPA recognizes that the comparative risk approach
is a first step in addressing this question; however, studies in progress
by the EPA and others indicate that judgments on "risk acceptability" are
extremely complex, with comparative risk evaluations representing only one
of numerous inputs which must be considered.
Specific Comments
The question of applicability of the Study results to all current
commercial water reactors is very pertinent. The discussion on page 27
appears to be the only place in trie entire report that the question is
considered, and only a brief general assessment is attempted. It is
generally recognized that there are certain design differences in Babcock
& Wilcox and Combustion Engineering plants as well as the Wtestinghouse
plants of four-loop design (more camon than the three-loop system
selected). Similarly, the BWR containment design, in particular, has
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undergone two inajor changes (the Mark II and Jlark III containi. tents) since
the reference design Peach Bottom plant, which would te exacted to at
least change the details of the contairuncnt response analyses. It would
appear that the Study could benefit significantly by recognizing these
design differences and presenting the necessary argu.ients wld.ch support
the thesis that these design and response differences at the system design
level do not have a major effect on the overall ris.V assessr;ient. Further
discussion appears warranted and any continuing analyses by the AIXT to
further verify tliis conclusion should be included.
On page 45, the safety irprova
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dispersion, the population distributions, and the resulting health
effects. Obviously, many refinements to the various calculation models
are available. Those which were assessed and found negligible in effect
for the purposes of this report should be discussed to give a better
appreciation of the range of applicability of the calculation model use!.
Specific Comments
The reasons given in Section 6.2 of Appendix VI for selection of
critical radioisotopes do not support the omission of plutonium-241. Data
presented by the USAEC in the draft WASH-1327, "Generic Environmental
Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in LWR," on
pages 1-13, I (A)-2, and 11-20, indicate that for exposure by inhalation
of plutoniura within a few years after its production in the LWR uraniun
fuel cycle, Pu-241 contributes more to the dose than Pu-239. Similarly,
the data referenced by the Study (ICRP-II) and the plutoniura isotopic mix
of WASH-1327 indicate that Pu-241 contributes more to the dose than Pu-239
in the majority of the organ doses considered in Table VI-15 of Appendix
VI, the proportion of their contributions depending upon the solubility of
the plutonium aerosol.
The discussion of meteorological models and assunptions should be
expanded to discuss the expected calculational differences incurred with
the Study's use of a simplified model as opposed to the more conventional
but complex models in general use. For example, it is inferred frorn the
discussion on page 16 of Appendix VI that much of the meteorological
frequency information is taken from greater heights than the release
heights predicted in this Study. Since wind velocity generally increases
with altitude, using such information will tend to decrease the estimated
downwind dose levels. The acute health effects, therefore, could be
underestimated. Furthermore, the uniform distribution in the crosswind
direction used in the atmospheric model, as described in Section 6.4 of
Appendix VT, is also likely to produce an underestimate of acute health
effects, since the sector averaged dose estimates should be lower than
actual peak doses. Finally, without consideration for wind meander, the
constant angular widths chosen appear to broaden the plume more tlian would
be expected (Ref. Figure A.2, page 408, of "Meteorology and Atomic Energy
- 1968"), again contributing to lower peak doses and thus fewer acute
health effects. It is judged unlikely, however, that any underestimate of
acute health effects resulting from the treatment of meteorological
information and the dispersion model is greater than a factor of two.
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The model selected to account for the effect of evacuation on the
calculation of medical consequences is described on page °.l of tiiis
appendix. An jJ?A report/ "LVacuation Risks - An Evaluation," I3?/\-!520/6-
74-002, is referenced in support of certain assitiiptions use; in the
evacuation model. Altliough a number of parametric calculations relating
to the evacuation i.olel assumptions are presented in Table VT-21
(including a no_ evacuation case study), we believe the H-\RP case
evacuation model to be overly optimistic.
'x'he .'-PA report on evacuation risks was priiiiarily -lire'rtec at an
assessment of the risk of death and injury, and the costs associated with
past evacuations of population groups. Tho data and information utilize.i
in tnis study were obtained by contacting persons and organisations
involved with previous evacuations precipitated by nat.iral or uin-nade
causes. Factors which were hypothesised to influence the tine required
for the liistoric data base evacuations included: (1) ti ic lapse before
onset of incident, (2) availability of evacuation plans, (3) tine of day,
(4) waather conditions, (5) population size, (6) area sirle fra : tlie tine
associated with acca.plishTient of the evacuation/ ix) quantitative
evaluation of this paraneter was r-iade. A correlation of evacuation tines
with population density, however, was perforiiTed assuming in 'eponlence froin
otiier parameters. A trend showing an increase in tiuie required for
evacuation was inaiuated as population density j.ecrease!. In applying
this conclusion to evacuations which my result frai potential nuclear
accidents, an clement of caution needs to be exercised. It shoal 1 Lie
rei.ienbered tliat the data on l~dstaric evacuations generally include
situations applying to small areas or in the case of larger areas, when
there is a lengthy forewarning tine. I lore significant is the fact that
evacuation travel distances were almost always short aiy" safe destination
points were generally obvious.
Cilice the evacuations called for in the larger conafiquenco accidents
appear to involve evacuation areas of a few hundred square :iiles, tlie
application of cvaciaa-tion tii'se requirements fraa the J'PA evaluation stuc'y
to areas of this size is questionable.
Although tiie .\^peiidi:: VT discussion states that a: ou':c^nic^st li. it for
evacuation of 20 idles was assumed, it is not clear how tl'iis evacuated
population segr.ient is treated in terms of actual do.se received.
Similarly, in tlie assessment of property dauiage on piigo ;"3, t'ie
effectiveness of tlie 10 rem calculated yearly dose as a basir, for
tei.porary evacuation is unclear since the e;^pecte.l dose rate as a function
of time is not indicated.
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Tiie assumption stated on page 63 regarding a first year projected
of 10 ren as the criteria for deternining the decision ho evacuate ay b;
unwarranted. A suggested value of 10 to 20 re a is cited, but the
reference, although relevant, dees not contain sue]: a suggestion. In it,
the recommendation is made that for small population groups, the use of
evacuation as a protective action be considered if the anticipated
exposure during 30 days might exceed a whole body .lose of 2 rod or a
thyroid dose of 10 rad. The reference suggests that un lor less favorable
circmnstances evacuation wight not be considered as d protective action
unless larger exposures were anticipated. The reference ~:oes not support
the implication that 5 reia per year or less is acceptable because it is
below the occupational dose limit, nor does it suggest 10 re.i or any other
projected dose as a criterion for decontamination.
The evaluation of health effects appears to rtx'juire significant
Modification and information additions. Our preliminary indications , as tlioy
state, riay be too liigh or too low. Although the value of :-:ero could not te
excluded by the data, use of this number was rejected on a nunber of
cogent bases (pages 2 and 88 of tlie I3LIR Report). Also, on pages 34 and
43, tlie somtic excess deatli risk used (50-165 deaths/106 Kian-ren) only
reflects tlie range of absolute risk estimates fron t?ie D'JIIl Import.
Consideration should be given to tlie estimates wliich, in cor.iljination witli
tlie ^ibsolute risk estiuiates lead to tlie DEIR Cauidttee "riost li]:ely"
nur^txirs (150-200 deatlis/106 iian-rera). Using H]IR "inost ll'-.ely" nu..>ix^rs
for a lifetime plateau instead of a 30 year plateau for cr-cpression of
effects, the EPA estimate of 200 excess deaths/10" J.t-m-re i has been
developed (page 167-174 HUIR).
TlTtJ I3JIR CoMuiittee genetic effects estirates refcrra.' to on page 35
*ray be construed as uieaning that 10" rian-rera would pra';ic.:e 10-100 uorainant
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diseases and 1-100 congenital anomalies in the first gem-oration after
exposure. Hovjever, tliis is 1/5 of tlie total impact e:cpecto1. An
additional 40-400 ucminant diseases and 4-400 congenital anafvilies shoul'i
be attributed to future generations (page 53 RCIR) .
The reference for the numbers in rads used as criteria for estimating
acute effects is not presented. The average dose wliioh will cause
fatalities to 501 of the people so exposed in 60 days is given by
Lushbaugh, Ca>as, Udwards, and Andrews (Secc 17 in A7: -7)1' C 10410, 19'i4)
as about 235 rads. These autliors estir.iates for the lose \/liich will cause
fatalities to 10% of the people so exposeu, Lt\Q, is o£ t;ie order of 75- bse
range of 40-140 rads given in .'CRP Report $29. Therefor';, the state. iei ir-
on page 37 tiiat there is little chance of death fra.i i.ian-rera of exposure is presenttxt, v/liich also appears to
be a irisinterpretation of the DEIR Report. The LLIR Co udntee esti'iate."1-
that tlie average mutant persisted ii\ tiie population for five generations
not "...in tlie first and also in all generations..." Therefore, the total
increments shown in Table VT-14 should be five ti; es greater.
Sii-ilarly, tlie quotation frori the .l'.IH Report (page- ;jl) appearing on
pages 52-53 is truncated to an extent that, in our opinion, a
'oisiaterpretation of tie BilIR Report results. The paragraph quote."
continues: "r>y extrapolation, it can }JG estii.iate.7. tliat 'd\<.± nunjyir of
deatlis per 0.17 reiu per year in tlie entire U.S. population ;<«y rringe
rouglily from 3,000 to 15,000 with tlie i.ost likoly val:io falling in til;;
range of 5000 to 7000 (or 3500 per 0.1 reisi a"f otlior
doatli and radiation sic/aiess are discus.s&j only ]jrif%fly OM mge 4G. For
e-car.ple, bei.porary aspen .iia in tlie :vale has been o]js^-irve ' following
exposures as low as 12.5 R. The personal trauc a of Ixdng u;ial.)le to
reproduce or of it being reca.TOeixIeu that no attonpc. Ix1. ;:a.;c to conceive a
child for some extended period after exposure is not negligible, at least
for normal "peacetime" operations. Furtlieriiore, tlie disruption of the
honeostasis of tlie finely tuned endocrine system, wliilo possibly ai 'tenable
to hon.nne replacertient therapy, does not necessarily represent
insignificant individual trauma or financial burden. Tlmrefore, a
significant expansion of this presented discussion appears warranted.
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SCQUHTCHS, WJC2GR MKLTDOW-I PROCESSES Al 7)
VTrY reU3AST;S Appendices V, VII and VIII
General Ga fluents
These appendices, which follow various accident sequences through the
meltdown, [process and associated releases of radioactivity, represent a
significant effort to quantify the consecruences of reactor meltdown
accidents. It is recognized that to present a meaningful discussion of
the i.iany accident sequences evaluated, to relate these sequences to the
timing and physical processes associated with a reactor meltdown, arid to
prelict the resulting radioactivity releases via several containment
failure mechanisms is a formidable task both technically air,1
documentarily.
Of these three appendices, Appendix V (possiMy tecause it pulls
together much of the infomatiori presented in Appendices VII and VIII)
appears to require soi;>e additional effort to resolve apparent
inconsistencies and to supply additional information on accident sequences
other than the large LOCA. Furthermore, the Study should higlilight and
e/qjand the sensitivity analyses on UCCS functionability and the evaluation
and significance of the various containment failure node prol^abilities.
Our comments on both Appendices VII and VIII are dealt with in the
specific cam'lent section which follows.
Specific Ccffitients
Appendix V
One problei in revievdng Appendix V involves apparent inconsistencies
between the various tables which relate accident sequences to release
categories. Tor example, on pages 21 and 24 (Tables V-'J and V-'l), it is
not clear how these; lists were compiled. Both tables do not include sone
of the dcrninant large I£CA sec^uences fra'i Table V-6 (e.g. AF—*• arkl ATD-a
in category 1, and AF-6 in category 3) but do include sequences v^iich are
not considered dariinant (e.g., 7CDGI- a ~). Basel on fie liscussion (]>igo
140) , it is also not obvious why sequence ACDGI '-a is clas.sifiel as
release category 1 instead of category 3. Furthermore, in corinaring the
probabilities givc^n in Talkie V-f- with tlie relative contain!.tent failure
i.iode proljaljilities listed in table 2 page 124 of the attanhaent, certain
sequences listed as "other large JjOCA accidrait sequencer," ap^-^ar to bo
significaiit contrilvators to a release category' proLaidlity (e.g., catc->jor^'
2, AJjT - a, 3x10-19? .V,-6 , 4 x 1Q-H). If tbese contril •utions"aro
nur.erically correct, the sum a:\»arinj at the bottan of the table must
only represent the sum of the listed dominant sequence. proValilitios.
Since tlie large irCA'r, do not da.iinat'i the pro}.alilitio- of Table> v-]/,,
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information similar to that presented in Table 2 of the attachment, page
124, applicable to small LOGAs and transients would appear pertinent for
inclusion in this appendix.
We would like to emphasize at this point that inclusion in a release
category probability estimate of a 10% contribution from adjacent release
categories adds considerable conservatism to certain summed release
category probabilities; however, an attempt should be made to correct and
clarify the interpretation of these summary tables.
In the discussion of the smoothing of release category probabilities,
p. 50, it is not clear how the smoothing technique necessarily swamps any
common mode failure contribution. The presentation would also be
clarified if even just an illustration were included which would show the
bar chart in Figure V-l reversed in relation to the severity categories.
Considering the interest attached to the ECCS functionability, the
discussion on pages 52-55 is especially pertinent. This sensitivity
analysis discussion might be considerably improved by not only relating
the ECP contributions to overall release category probabilities but also
to the "large IDCA" contribution. This latter Relationship would show, a
larger percentage contribution. For example, given a large LOCA (A)
Followed by ECF (E), one accident sequence would be AC- e with the same
consequences as AD->-e (category 7). The probability would be AE- e = (1 x
10-4) (10-2) (~ 1) = 10-6 assuming the high end of the ECF failure rate.
Although this and other sequences would have a moderate influence on large
LDCA release categories, the limited impact on the overall release
category probability would be highlighted. Since the ECF failure
probabilities are of general interest, it would appear appropriate to
identify the rationale for assuming the failure occurrence range utilized
(10-2- io~5). Considerable confusion is also caused by not including ECF
sequences in Table V-l6 while including such sequences in Table V-6.
With regard to the BWR transient tree quantification on page 68, it is
not clear from this discussion, in conjunction with Table V-19, which
transients were slow enough such that credit for reserve shutdown can be
taken .
In attachment 1, Table 2, certain sequences are shown with
"containment rupture - vessel steam explosion" failure mode probabilities
of zero which are nevertheless estimated as 0.01 in Table V-6. Since
similar tables are not included for S^ and 82 initiating events, the
relationship between the various containment failure mode probabilities
shown in Table V-7 and V-8 cannot be determined (e.g., the relationship
between S2 C-d and S2 C-a ).
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Appendix VII
The information contained in Appendix VII is well presente I,
sufficiently documental, and based on our preliminary review, presents a
reasonable appraisal of the extent of radioactivity releases.
In discussing the neltdown release ccnponont for ar:a.lino earths ont;
noble metals (p.p. 11; 13), the probable values selecte-1 appear somwhat
low if consistency with the selection basis of other released coruxinents
(e.g., lialogen, alkali metals) is to 3:e maintained. In fact, the text, in
discussing the alkaline eartlis, indicates a release rarvjo of 2-20% and
suggests that the probable value should lie in the upper :"iortion of this
range, yet selects 103 as a irost probable value.
On page C-l, last sentence, it is not clear wliat is referred to by
"...the IDCA's postulated; i.e., successful JTJC and recovery," since the
Study is concerned with many LOGAs in which successful IT/' and recovery
are not assumed.
In outlining the accident sequence and core response on page C-2, the
basis for the 100% rod failure at a maximum clad temperature of 2.').00 " F
should be stated since tliis failure value appears to lie quite conservative
in view of vendor calculations (eg. SuafCY, Final Safety Analysis Report).
Also, in tlie discussion of six critical points involved with the
evaluation of the LOCA pronpt release fission product source-term, the
terra "release coefficient" (escape fraction) should be made consistent
with nomenclature used elsewhere in the report.
In Appendix K, p. K-19, it is not clear if the text is implying that a
potential important pathway for release of fission products, between the
containinent shell and cofferdam, was not considered in the Study.
Appendix VTII
The discussion under "Linitations" on page 3 of Appendix VIII contains
a disclaimer regarding the potential non-applicability of "these studies"
(presumably core meltdown studies) to other B-TRs and nWRs. As mentioned
previously under the specific comuients on the main volume of the Study,
the discussion on this topic should be expanded.
In discussing the basic assumptions for the analysis of degraded.
accident behavior (p. 7), the basis for ar.sii.iing that "core melting would
take place witliout significant metal-water reaction and that there would
be no possibility of steam e:
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14
depressor izat ion and the time of accumulator discharge ap;x>ar a factor of
2-3 too short compared to the results in the paper, "Corparison of
TlTeErnal-ifydraulic Response of LOFT and a Large PUTR to LOTA Conditions,"
authored by P. Davis and J. nuctone, presented at tho topical meeting on
water reactor safety. CONF-730304, March 1973.
On page 12, the dismissal of the potential for a large energy release
from a stean explosion between the molten core am water laden gravel
s to te contradicted by tlie Arriico Incident describe "> on page B-2.
In describing containment response (p-13) , the asrnmption of IPTIS
cavitation at the tine of containinent failure seems pessioistic, if
I-tegulatory Guide 1 is followed. The guide states, "TiTor jenny core cooling
and containment heat rerioval systens should be designer"1 "TO t'lat rvVxjuato
ncd; ,ior;itive suction head O'IPSII) is provided to systen pirns assuriung
rutri.ium expected temperatures of pur pel fluids an . no increase in
coiitaiment pressure frcr. that present prior to postulate1. IO1A."
secjucince rliscussion, v.^iich incl\:>l^s CTp\'3 an
failures (p. 17) , describes tlie molten core vessel '^net-ration an.'"!
interaction vdtli the \jater in the reactor cavity an1, tho CSRR vsiter. Our
UJTtlorstantling is tliat tlie GSRS water should not bo csr^ete-1. in tlie reactor
cavity except for ;x;ssi3jly a small amount of lea]:agc;. If tiiis
inter[)retation is correct, all sources for reactor cavity vafear need
further clarification. Similarly, on page 21 it appears that CSI5 is
assu; TQ I not to deliver water to the reactor cavity while the opposite is
true for GSRS.
The assessment of containment failure tiode pro] -oi dli ties includes tlie
probability of contai:irnent failure resulting frora a steavt eiq^losion
estii.iated as P - 10-2 (+l,-2). Since \ie are not aware of any discussion
which indicates tliat this proljability is sequence dependent, tlie
probabilities associated with certain sequences in Table V-1G are not
understood ^e.g., SjD-a, J x 10-* while S^D -e/ t> x irH>aii1 nc i S£ G-a
v;hicli sliould te at least 2 x 10-8 base" on S2 C-a of 2 :•; in-6 ) . The
latter exai.ple nay be elir-unated h«cause containMent ovorprossurG failure
occurs before initiation of core meltdown. If tiiis or other sequences are
logical exceptions to the contairuMent failure prolxibility associate! with
vessel steam explosion, the exceptions sl'iould te discussed.
Tlie assessments of containrrent failure probabilities from hydrogen
ca^ustion or overpressurization lx>tli are strongly dependent on the
assumed noritial clistr dilution of containment failure pressures of about 100
psia \.dth a 15 psi standard deviation and containi lent nelt-tlirougVi tine
(for which tlie meaning of tlie skewed distribution, m+10,-5) hours, is
not clear). Given the information in figures 4 through 9, general
correlation with containment failure mode probabilities listed in table 2,
attaclinient 1, Appendix V could be observe:!. However, such was not the
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case for sequences AIIF-6 and AD-6 . It would be helpful if the text
provided an example of such a calculation, which would define the various
probabilities listed in the text.
Additional clarifying remarks would, seem appropriate at several points
within Appendi;; A, which discusses the thernal evaluation model. On page
A-16, the assumptions used for the core teuperature distribution an"!
vessel water inventory following blowdown should be state:! and justifiel.
Itegarding the fission product release fraction equation, the Ivisir;
referred to should be specifically referenced. Siiilavly, tlie reasoning
for assur.iirig no change in steal a properties due to hydrogen mixing should
be presented. Under the heading "Convective Heat Transfer" tlie values
chosen for Tw and hB should be discusse" since it would appear that hfi
should var^'' ^itli O , ^T, and pressure.
On page A-3J, a question arises as to whether vensol failure can ocoir
L>y fracture due to tliermal stress occurring wiien tan ix^.ten core contacts
tlie lower vessel nead.
It is not clear on Page A-JG if the continued addition of water on top
of tlie core melt could cause a steam explosion similar to the East Geman
incident described on page D-3.
The containinent failure mode evaluation presento."1. in Aptxytli:: E
considers several factors whicli could affect the ultJj'Pt^ oontainrr-xi-i;
strength. I;\artho.r discussion or clarification of tlio po':.-"vi':ial
significance of these factors on 'the anstrxd 100 ± 15 psia failure
pressure appears warranted. Since the assumed failure pressure coul-1
alter the containr.ient failure mode prolialoilities for several accident
secjuenccs, an indication of tlie sensitivity of the roloarse category
probabilities to a cliange in tlie assit-iecl containment fair .ire
should 1x2 provided.
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11TINITIQM OF F/ilTAJKS DATA ATJD PATHWAYS
(Appendices I, II, III, A' 3D IV)
General Ca"T.ients
Our review of Appcnlices I and II has, for tho r.ost part, T-vxin linito ,
to questions regarding the treatment of (1) specific fail' ro pathways
which are not acknowledged in these appendices or (2) the rationale? for
disrdssal of otlier failure pathways or their relations] i.r > to assigned
contoiiji.Tent failure r,o:le probabilities. Review7 of certain sections of
Appendix II is presently anticipated in our continuing review of the
Study.
In our limited reviev; of Appendix IV, the role of f 10 "GOTO; ion node
failure" in the overall risk assesanent process is difficult to assess.
Some i ethods, quantification tecliniques, causes and results are discusse i
in Appendix IV, but the material necessary to properly un Hern tan; ' the
totc'il role and significance of ca;ition i-iode failures an-1 to .-letennine that
a reasonal)le degree of caipleteness has Ijeon develop:)! a.-T^ars to be
sproa.1 through A[j|.Deii.iices I, II, III and V. Many sunoazy statements in
tiie earlier and later sections of the report assert the sicjnificance of
contiion i-iode failures witlwut quantification or reference when, in fact,
the netjdod i material is in other appen.iices. Furtlier cross-referencing to
such analyses and cjuantifications Which support tl"ie assertions of fiis
appendi>c could resolve tliese conccims.
Specific Cortments
In the LOCA functional evenL tree development (p. 13 footnote) , it
appears that the containment building purge system }iar, a probability of
failure which is not acJinowledged. Similarly, the possibility of
containment overj^ressure failure prior to core nelt \-1iich is treate"! j.n
appendices, is not included in the discussion on page 31.
In Lho aeveloptnent of tlie HJR snail LOCA event tree, 82, p-l^» it
•would seeia possil^le tiiat vessel melt-througli coul.1 occur while the prii iciriJly reduceJ,
allov/iiig [x>ssil)le accumulator water injection onto the nolten core,
potentially causing contaira;ient rupture fraa a steaii c^r^losion. It could
not Ix; determined if tlie containment failiore mode prol>abilities for 82
LOCA's considered this possiliility.
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17
With regard to F.JR reactor vessel rupture, p. 141, it is not clear ITOW
the polar crane presents an effective missile barrier for the entire upper
portion of the containment.
On page 145, the reason for not considering rupturo of stean generator
tubes and subsequent overpressurization of the secondary system with
potential for rupture outside the containment, should be stated.
Siioilarly, on page 159 it is not obvious vjhy the situation of autcnatic
trip failure occurring with loss of electric power sequenco ^xis eliminate* 1
fron consideration.
Appendix II, Vol. 2_
The discussion of the electrical power system, off site emnon node
failures (p. 33), does not specifically indicate whether earthquakes have
Jjecn considered in assessing covinon inodo failures, errx^-ially for power
sul ^systems which are not specifically designed for cart1'q-'Ake response
(diosel fuel system). Sinilarly, in the text on oa-j^ 3">, a discussion of
how the failure analysis of tl'ie diesel generator synten accounts for the
failure nodes discussef. would be an important acVition to this so-::.-ion.
The evaluation of the reactor protection synton l''J"") • "'iFriimos th< .
inpact of a prossrorizcr vapor space npt'ore on the PPf (sig-vil djiitiation)
failiir,: probability. Since it is posniJ'lf; for the ^jv ;_>r; r-.r.urizc?r lov^l
signal not to function due to frothing, the effect of nr •') a fai""nr.'5
sliould )« addressed. On page 155, under CGir, faj.l,ir^ . o "os, it aujx-Mrr;
that tho p/^fueling ^atcr Storage Tan]-; (R'.TT), "siir-tio/ lir-.r plagg^',11
should also ITO listal under sLiglo failure rosul:::mj ii-i • i:vT/fii."1.a]jility of
HITS? v.ater.
The Conscx_iuencc Li iting Control Systc', (C"/^) description o-i pago 174
is confusing since it ap}Xjars frcn this '"isc^.ssion tlia'.. t!^- ojxirator -ay
not be able to switch frai CHIS to CT-"£> vntil tho .^ont.-it^-vv.t ;)rosmiro
falls to -0.5 psig (s";ch a pressurre iRy not occ ir in tr . to p^ri'iit
successful switch of these systeros) .
The.' results of analysis of systm interfaces un^or La.T Pressure
Injection Gysten (LPIS) in."icates, on page 250, that I'O'.vnLajr^/
onavailal'ility of \reiter at tiie sbirt of TPIS JTUJ^J op^TP.ti.n'i is not
considered a failure, *mle inavailaLility of v^itor to th.-> cr^r- is
considiirod a failure for tlic saioe reason (Aprxavii:-: I, p. 102). A
clarification of tliis situation appears warrante."1. Also, in listing tv;
single failure-failure ; odes for JjPIS \p. 2C4), consllov-i'-.ion slioul"1. TO
given to pipe ruptures betv;ecin J3 and eitier V4 an" V5 (figi.ire 11-53) an-1
to J^JST puip suction drain plug.
In tho escariiination of potential faults for the Lov/ JT\s:;ure
l
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13
ruptures between PI, P2 and J3 will not cause system failure. The flow
would split betwaen the path to the cold leg and the rupture and,
depending on relative flow resistances, the delivery of water to the cold
leg may not achieve the necessary 300 gpm.
Appendix II_ - Volume 3_
In the evaluation of the BWR electric power system, an assumption is
rnade on page 27 that all emergency buses are available immediately prior
to a UDCA based on the Technical Specification requirement that the
reactor be shutdown if an emergency bus is not available. However, it
appears there should be a finite probability that all emergency buses are
not available which should bo dependent on the failure probability of the
failure detection system. Also in discussing offsite power common mode
failures, the omission of earthquake as an initiating event should be
addressed.
Discussion on page 134 relates failure of vacuum breaker valves in the
open position to the defeat of the vapor suppression function. It would
seem appropriate that the assumptions regarding the two or more valve
failures required should be justified with calculations or referenced to
pertinent information. Similarly, the assumption on page 243 that rupture
of branch piping of 2-inch diameter or less will not significantly affect
core spray injection system operation at the time of a LOCA or during
injection requires justification.
Appendix III
This appendix on failure data is well written, well organized and
appears to be appropriately integrated into the Study. In our continuing
review, reflected in the attached work plan, EPA does intend to perform a
selective review of the failure rate data base.
Appendix IV
Our review of Appendix IV, to date, has been limited, especially with
respect to the analysis and quantifications applied in the Study;
therefore, our comments at this time are very general in nature. The
concept and influence of the "cannon mode failure" appear to need
additional development. A distinction should be provided between the
causative effects of certain control mode failures in initiating accident
situations vs the influence effect of common mode failures resulting
during an established accident sequence (e.g., the influence of a check
valve slam and subsequent water hammer damage as a cornTDn mode initiator
vs the consequence of such an event occurring during an accident in
progress).
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1'J
The completeness of tlie consideration given to potential sources of
caiinon mode failures also appears to require sane expansion. Although,
tlirough searching in other appendices, it is evident that particular tynes
and areas of ccrttron node failures are considered, sources, such as the
requirement for pump inlet subcooling for emergency coolant recirculating
systems, pump bearing lubrication systens, instrunent and cornponent
service water and air, heat tracing for nlateout prevention, control root-i
and cable tray fires, drain plugging of storage water systa.is, etc, ought
to be discussed and treated in Appendix IV to support the claims and
assertions developed.
Tinally, the treatment of the method of screening for relevant comnon
mode sources discussed on pages 18 through 39 appears crelible but is
unsupported and in need of tabular listing of (or reference to) the
"numerous" types of sources considered in order to demonstrate tliat the
inetliod is indeed comprehensive in identifying all conceivable canponent,
system, and operation vulnerabilities to common mode failures. The
examples cited are useful but create questions of "what else," "how r.-any,"
and. what does a complete list look like.
DESIGN ADEQUACY
(Appendix X)
General Comments
This appendix does not appear to be tied in with the rest of the
Study. There is no mention of how the results of this Study were utilize1.
in the risk assessiients. Since Appendix X indicates that a significant
number of tlie systens exariined were either not properly qualified, not
properly analyzed, or didn't meet current standards, it would see;a very
important that these deficiencies be readily traceable to tie quantitative
risks, or that they be shown to be peculiar to the plant: analyze t.
Specific Connents
The section on seisraic loads (p. 47) appears incomplete in that current
design response spectra were not evaluated for the structures and
equipment. On pages b2 and 57, it is stated that the c irrent spectra
would increase seisraic loads (by as much as a factor of 2). It is not.
clear what these increases mean relative to the general seismic
vulnerability of the 100 plants and what risks are associated with the
increases.
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20
SUMMARY REPORT
General Oomnents
The summary document is a relatively well written volute, which
satisfies its intent through a question and answer format. Our cements
on certain quantifications of assessed impact are incorporated into our
review of the Appendix VI volume. Of particular interest was the
discussion comparing the Study predicted consequences with the earlier
WASH-740 evaluation. It would appear that the significance of the four
factors leading to differences in the two studies is substantial. Plume
rise and evacuation and possibly population, as treated in WASH-1400, have
relatively little impact on consequence when compared to the effect of the
differences in assumed release of radioactivity to the environment. The
Study indicates that given a PWR core meltdown event, a chance of only
about one in one hundred exists that the resultant containment failure
mode will be other than melt-through with its relatively insignificant
radioactivity release. A somewhat similar case exists for the BWR
meltdown event where a chance of only one in ten exists that the
containment failure mode will be other than containment isolation failure
in the drywell with, again, a relatively insignificant release of
radioactivity. It would appear appropriate that the discussion of this
variable in the summary document should be expanded.
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Additional Comments
Main Volume
Clarifications
1. Page 122, Table 5.2 - It is not clear why AB, Ad IF, SiB and S2B do
not lead to containrnent failure by overpressure since loss of containment
heat removal should lead to overpressure failure.
2. Page 126, Section 5.3.2.1 - The definition of a large IDCA being
a rupture equivalent to a hole greater than 6 inches in diameter is not
consistent with the definition used by vendors, AEOHegulatory, etc.,
v;hieh is a 0.5 ft2 (9" hole). It is not clear why a different definition
was chosen here.
3. Pago 154, item (1) - The SL-1 accident was a military power
reactor nuclear accident which resulted in 3 fatalities. It appears that
the statement ignores the SL-1 accident.
4. Page 21f>, last sentence in Section 6.4.7 and Figure 6-10 - The
statement that the calculated probability of a dam failure resulting in
10,000 fatalities "...agrees with the extrapolation of the data..." does
not appear justified. A straight line can be drawn through the three
data points (as was done in Figure 6.9), and, if anything, an upward
inflection of the curve is indicated by the data, rather than downward as
drawn, to include the calculated point.
rxiitorial
1. Page 1^6, Section 5.4.4 - The source for the probability of
aircraft impact accidents should be referenced.
2. Page 150, Section 5.4.6 - Near site explosions, which must be
considered for reactor sites, are not mentioned.
3. Page 200, Section G.4.1, 1st sentence - The reference does not
agree with the reference at the end of Table 6.8.
4. Page 204, reference 1 - This reference appears incorrect,
"....forth Atlantic Hurricanes..." since the Galveston hurricane is
apparently included (#1, described on page 200).
5. Page 205 - The average number of tornado fatalities is stated as
11G while the division indicated yields a value of 46.
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22
Appendix !_
Clarifications
1. Page 83, Figure 1-13 - It is not clear why the success path for
containment leakage is chosen as the drywell and the failure path, the
wet well. Wet well leakage should produce the lesser consequences due to
fission product scrubbing in the torus (see Appendix I, page 37).
2. Page 102, 3rd sentence, and page 134, item F. - The basis for
the GSRS failure assumption is not clear since CSRS should eventually
operate.
3. Page 199 - Further justifications of the unanticipated transient
probability of 10"5 per year should be presented.
4. Page 205, Figure 1-23 (also Footnote 1, page 207) - The EPS
failure probability for unanticipated transients (Part C) has been
increased from 4 x 10~7 to 4 x 10~6 to account for the fact that only the
scram system may be effective for reactor shutdown. This reduction does
not agree with the fault tree at the bottom of page 68, Appendix V, which
assigns the failure of RPS to scram a value of 1.3 x 10"5.
Editorial
1. Page 45, 3rd paragraph, 1st sentence - It appears that CR-VSE
should be CR-CSE.
2. Page 77, Figure 1-10 - The IPIS is missing from the ECI segment
of this figure.
3. Page 198, 1st paragraph - The apparent distinction between a
transient which causes a LOCA and a transient which causes a ruptured
reactor coolant system is not clear.
4. Page 222, 4th sentence under RIRS - This sentence is not
complete.
5. Page 233 - These footnotes appear to be used in Table 1-13, but
the heading does not match the heading for Table 1-13.
6. Page 259, item 5 - This item appears out of place in that it is
not a "...design feature provided to keep the likelihood of loss of pool
water small..."
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23
Appendix II, Vol. !_
Editorial
1. Page 11, last paragraph - This disclaimer paragraph scorns to
indicate that if data did not exist for a particular system failure
contribution, it was not considered.
Appendix II, Vol 2_
Clarifications
1. Page 253, item 3 at top of page - There does not appear to be any
basis for the assumption that pipe ruptures of 2 innh diameter or less
will not cause failure of accurtulator injection.
2. Page 385, Introduction - It is not clear if the SICS analysis
also applies to the small break case.
3. Page 490, top of page, and page 529, 1st paragraph - It is not
clear how realignment of the LPR system to the hot legs will prevent an
"undesirably high boron concentration or accumulation of residue and
debris in the core that could result from continuous boiling." LPR system
water injected in the hot legs will enter the upper plenum, run down the
outer (cold) core and core structure region into the lower plenum, and be
available for boiling in the hot central core region.
4. Page 490, top of page - It appears that closure of V10 is also
required to effect the realignment.
5. Page 501, 2nd paragraph - It is not clear why air suction from
the RWST occurs for this failure in view of the discussion under item (2),
page 493.
Appendix II, Vol. _3_
Clarifications
1. Page 92, Section 3 - It is not clear why O (1.3 x 10~5 ) is
considerably different from the RPS unavailability snown in Figure 11-131
(2.47 x 10-6 ).
Editorial
1. Page 359, item 2 - The use of the term "suppression chamber" is
inconsistent with "wetwell" and "vapor suppression system" used elsewhere
in the report.
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24
Appendix III
Editorial
1. Page 187, first line - The bibliography section mentioned here
appears to be missing.
Appendix IV
Clarifications
1. Page 43 - Results of the susceptibility analysis are presented
but no specific reference is given to where the analysis is presented and
the specific fault and event trees to which it was applied.
Editorial
1. Glossaries and definitions are sorely needed for this appendix,
not only to track the latter sections in relation to Appendix II, but to
understand the distinctions between the PWR and BWR treatments.
2. Page 8 - The treatment of ideas at this early stage in the
appendix requires the reference to other unspecified appendices in order
to understand the terms used and messages developed. An introductory
tutorial treatment with a description of the other appendices which
intimately interface with Appendix IV is needed.
3. Page fl-15 - This section could benefit by specific cross
references, examples and limited numerical results to give significance
and meaning to this important portion of the report.
4. Pages 40-41 - The list of "classes of potential common mode
roechanisras" could benefit by a sub-category of items under each major
topic to provide an index of completeness, e.g., where would failure
causes fall for wearout due to exercising a given component, or for
partial or delayed performance due to degradation from lack of service,
or for transient behavior of a component (check valve water hammer).
5. Pages 40-63 - Although Sections 3.3 through 4.0 portray a
reasonable description of the methods applied to the "quantifications" in
the study, the interpretation could be considerably aided by examples
with numerical results or tabulations, such as that of Table IV-4 on
coupling probability.
6. Pages 65 and 87 - These two sections are intended to treat PWRs
and BWRs separately and this should be stated in the introductory
paragraphs.
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25
7. Pages 65-98 - This Section, "Summary of Results," acknowledges
the performance of the "fault analysis" in Appendix II and from those
results identifies selected "sequences in the event tree...chosen
because...(of) some potential susceptibility for common modes" and
develops "impact" conclusions as "insignificant," "minor impact," etc.
The support for and meaning of these conclusions should be identified.
The event sequences selected for the follow-on discussions appear
without comparative discussion to other cases which have been dismissed.
Although these discussions improve one's insight to the "controlling"
common mode sequences, tabulations or some form of overall results
presentation should be developed to enable the reader to gain a "feel"
for the relative influence or "impact" of other sequences which could be
important to plants of newer design than those chosen for analysis. The
companion treatment given to the BWRs (page 87), although different in
style, is equally obscure in portraying understanding and confidence that
the treatment of common mode failures is comprehensive and complete.
APPENDIX VI
Clarifications
1. The description of the release and dispersion calculation in
Appendix VT appears sketchy in that there is not a clear description of
the radioactive material release magnitudes as a function of time over
the release durations presented. Thus, any interaction of the airborne
release with the population being evacuated cannot be evaluated. The
description suggests that the fraction of core inventory released is
modeled as a uniform release over the indicated duration of release. An
alternative model could be a distribution of discrete releases as shown
in Figure J-8 of Appendix VII. A clarification of this subject is in
order.
2. The discussion of the consequence calculation and population
distribution patterns of Appendix VI does not describe the model of the
population distributions used for calculation of consequences within 70
miles; i.e., it is not discernible from the information presented whether
the sector population, originally obtained as a function of distance from
the reactor, was averaged over the first 70 miles or averaged over
segments of sectors using differences in the cumulative populations from
Table VT-6, or whether some other distribution model was used.
3. A more careful explanation of the population averaging method on
page 24 would be helpful. In particular, the top 1% sectors are
reflected in the peak case consequence results. The range of populations
averaged into the top 1% would clarify the nature of the top population
category.
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26
4. The application of the plume broadening for meander over
extended periods of time, described in Section 6.43 of Appendix VI, needs
to be specified more clearly. Table VI-2 shows categories PV7R 6 and PW1<
7 having a duration of release of 10 hours and all other categories
having sliorter releases; it is not clear whether categories PWR 6 and PWR
7 are the only categories having "releases that last for many hours,"
i.o. categories to which the broadening was applied, or whether the
broadening was applied to shorter releases as well.
r>. Because this appendix does not present the necessary information
regarding individual organ or whole body doses as a function of release
category and downwind distance, several questions arise as to the
significance of certain omissions from Table VT-16; namely, (1) lung dose
contribution from noble gas inhalation, (2) consideration of Pu-241, Am,
G.i, and U releases, (3) releases of longer lived isotopes, such as 1-129
and II-3, and (4) any possible significant release of activation products.
6. With regard to the evacuation model, clarification is needed of
the manner in which the warning time for evacuation T^ (time between
awarcaiess of impending core melt and leakage for accident type j) was
determined. It is observed that, in Table VI-2, tliis time is constant
for each reactor type and independent of the containment failure mode,
and also that for release category PWR I, awareness of impending core
nelt is immediate at the outset of the accident.
7. Page 36, Section 6.6.3. - This section is based on available
data and is apparently extended for standard man only. The uncertainties
iii the estimates, particularly as they apply to differences in age and
state of health, should be at least underscored- and, if possible,
explored further.
H. Page 37 - The listing of peripheral blood element response
should be compared to data given by Wald (Chapter 23, Haematological
Parameters after Acute Radiation Injury, pp. 253-264 in Hanual on
Radiation Haematology, IAEA Technical Report Series No. 123, 1971).
9. Page 47, Section 6.6.4.4. - Reference and justify assumptions,
particularly the "...slightly increased number of induced mutations." If
a value judgment is to be made, a frame of reference must be established.
in. Although reference is made to the BEIR Report, the discussion
regarding Table VI-13 is misleading. This table, taken from p. 171 of
the BJ-ITR Report, refers to the "...absolute risk for those aged 10 or
more at the time of irradiation..." This is neither the complete estimate
of the BEIR Conmittee nor the only population considered.
11. Page 49, Section 6.6.4.3 - This section does not mention the
rather generalized "increased ill-health" considered in the BEIR Report.
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12. The discussion of thyroid illness on pages 53 and 54 appears to
need considerable clarification. In particular, the apparent treatment
of production of nodules as an illness requiring a surgical process is
not understood. For an estimate of nonfatal malignancies, reference to
the HEIR Report would seem appropriate.
]3. Page 54, first paragraph - The assumptions on incidence of
nodule formation following thyroid exposure discussed on page 54 should
be justified. For example, data in Reference 42 of the subject draft
report suggest that, in a mixture of external and internal radiations,
gainra and beta exposures are equivalent. The HEIR Committee points out
studies evidently showing a species difference in response to beta
irradiation of the thyroid and also points out the problems in some
available human and 1-131 data (a thyroid ablating dose is used) .
Reference 42 does mention thyroid nodularity incidences ranging
from 0.47% to 1.6%, it should be pointed out that the 1.6% incidence was
in a population of 30 to 59 years of age and 0.47% was in a general
population. The 0.36% to 1.7% values in controls in various studies
reflect small numbers in the populations and, perhaps, the regions of the
country from which the populations were derived.
Lilien, et al (AM Lilienfeld, M. L. Levin and I. I. Kessler, Cancer
in the United States, Harvard University Press, 1972) , suggest a thyroid
cancer incidence rate of 4 0/10 6 persons based on state tumor registry
data. Even if the ratio of fatal to occult cancers of 1 to 100 (ABCC
Tech Report 25-68) is used and the incidence of 40/105 thyroid cancers is
considered fatal, the total incidence of thyroid cancer would be 4000/10 6
persons. The relationship between these occult carcinomas and the total
number of nodules has not been established yet, but some nodules are
occult thyroid carcinomas. The nodules, as pointed out in Reference 42,
represent malignant and benign tumors, but also nodular goiter,
Hashimoto's thyroiditis , colloid diseases, local hyperplasia, local
lymphnodes, etc.
14. Table V3T-15 is somewhat misleading in that it apparently refers
only to acute or subacute fatality and to "illness" in which thyroid
should not be included since nodularity is not an "illness." The table
does not include all effects, e.g. effects of pituitary injury or
carcinogenesis, aspermia, etc.
15. References pertaining to in utero acute fatality and acute
soiratic injury are as follows: Evaluation for the Protection of_ the
Public in Radiation Accidents; IAEA Safety Series $ 21, IAEA Geneva
(1967) ; Nokkentved, K. Effect of_ Diagnostic Radiation on the Human
Fetus; Munksgaard, Copehagen (1968) ; Griera, M. L. The Effects of
Radiation on the Fetus ; Lying in; Journal of Reproductive Medicine
1:367-372 (1968) ; Hammer-Jacobsen, E. Therapeutic Abortion on Account of
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X-ray Examination During Pregnancy; Danish "fedical Bulletin. 6:113-122
71959) ; Brent, R.L. and Gorson, R.O. Radiation Exposure in Pregnancy
Current Problems in Radiology Vol. 215 (1972) ; Grahan, S., Levin, M. L. ,
Lilienfeld, A.M., Schuman, L.M, Gibson, R. , Dowd, J.D. , and Hempelmanri,
L. Preconception, Intrauterine, and Postnatal Irradiation as Related to
Leukemia, pp. 347-371 in Epidemiological Approaches to the Study of
Cancer and Other Chronic Diseases National Cancer Institute Jfanograph 19 ,
NCI (1966).
16. There is also no indication that individual organ doses have
been aggregated as "organ-rem" for summation in the estimate of "latent"
cancers and genetic effects. Estimates of some isotopes and the
distribution of organ doses and variations with age can be obtained from
such publications as ICRP-17 (ICRP Publication #17, Protection of the
Patient in Radionuclide Investigations, Perganon Press. 1971).
17. Page 55, Section 6.7.3 - The use of ICRP-2 dose models, while
defining what was done, does not seem adequate in light of advances in
the field of physiology and dosimetry. As pointed out by Eve (I.S. Eve.,
"A Review of the Physiology of the Gastrointestinal Tract in Relation to
Radiation Doses from Radioactive Materials," Health Physics 12:131-161,
1966) residence times and mass of contents for the GI tract used in IG^J3-
2 may be in error by factors of 2 or 3 in various segments and the values
used for the stomach may be in error by a factor of 24 when residency
time for inhaled material is being evaluated.
Dolphin and Eve (G.W. Dolphin and I.S. Eve, "Dosimetry of the Gastro-
intestinal Tract", Health Physics, 12:163-172, 1966) suggest that
differences of the order of a factor of 2 result, when a more
sophisticated GI tract model is used rather than the ICRP-2 model.
TCve also made pertinent comments on the dose to the ovary from GI
tract contents and the insensitivity of rnucosal cells to radiation
exposture at a depth of less than 140 microns.
The lack of information on particiilate aerosol characteristics of the
expected releases used in this section precludes applying the more
accurate Task Group Lung Ttodel or determining the extent of departure
from the simple ICRP-2 model which would be expected. However, the
current biological half-times for the various isotopes could be eroployed.
1". In the evaluation of damage from an accident, the health effects
and dollar costs appear to be considered as mutually exclusive. This
fails to consider the dollar costs of health effects. There is of
course, the obvious cost of lost productivity but it is also noted, for
instance, that thyroid nodules are passed off as being surgically
treatable with no consideration as to the dollar cost of that treatment.
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19. In Section 6.8.4, Non Core Accidents, Table VI-23 appears to
over estimate the consequences by up to three orders of magnitude.
Editorial
1. In Section 6.4.4, for the phrase in parentheses, "vertical
velocity toward the ground," substitute "ratio of the ground
concentration to the integral over time of the adjacent air
concentrations." This substitution will avoid furthering the false
impression that the deposition velocity is indeed the vertical velocity
toward the ground.
2. The PWR 7 category description on page II of Appendix VI needs a
few more words of clarification, since the sprays do not act on the
leakage occurring upward around the containment.
3. In the second paragraph on page 14 of Appendix VI, insert the
word "acute" before the word "illness."
4. On page 14, the sentence "It was found, in particular, that the
wind blew 0.1% of the time toward the 0.1% highest population density
sector" needs clarification. The explanation on page 110 of the main
volume is much clearer.
5. In Section 6.5.1, the reference to the isolated Idaho Falls site
is of questionable interest, since Idaho Falls is not the site of any
commercial nuclear power plant.
6. Table VI-6, on page 28 of Appendix VI, needs correction in that
it shows, for categories 11 and 12, that the cumulative population
decreases as the distance increases from 2 miles to 5 miles.
7. Page 32 - Experience with human radiation effects is not small
and includes much more than Japanese data. The experience with acute
effects is much less.
n. Page 35, Section 6.6.2 - The question of prophylaxis and adverse
effects thereof is an open question. The fact that the treatment may be
worse than the disease in some cases should also be considered.
9. Table VI-II, page 40 indicates up to 5?, mortality at 165 rad
(250 R) and a cutoff around 100 rad (150 R). Uncertainties in population
response suggest that there must be a range around these values and that
effects at lower exposure levels are possible.
10. Page 47, Section 6.6.4.2 - There is some confusion about the
data studied by the BEIR Conmittee. Probably most of the data is on
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relatively acute exposure to low LET radiation, the type most applicable
to the emergency situation studied in the subject report.
11. In Section 6.8 of Appendix VI, the last sentence on page 67
implies that a Monte-Carlo type of determination was employed, as
contrasted to the assertion in the second paragraph on page 3.
12. The title to figure VI-8 on page 76 should be changed since the
thyroid nodules do not include all thyroid consequences to be expected.
Appendix VII
Editorial
1. Page C-2, item 2 - The core, taken as a whole, cannot "heatup"
from sensible heat as stated here.
2. Page C-9 - The "Little Mamu" program should be referenced to
supporting documentation.
3. Page 1-2, equation (3) - Since this equation involves an
integration over time, a distinction in the various time parameters is
required since C is a function of "t".
Appendix VIII
Clari fications
1. Page A-3, 1st paragraph under Fission-Product Release - It
appears that the pin rupture temperature was assumed to be 1500°F in the
BOIL code calculations. This does not correspond to either of the two
temperatures cited in Appendix VII.
2. Page A-12, last sentence under Bottom Flooding - The meaning of
this sentence is not clear, particularly the reference to "these"
flooding rates, and the reasoning that heatup of cores at elevated
temperatures is not prevented.
Editorial
1. Page 6, top of page - Nomenclature problem: The ECR system
described here appears to be the same as the LPRS system used in most of
the rest of the Study documentation.
2. Page 7, last sentence - The starting time for CSIS is iirportant.
The fact that it must operate for a considerable length of time has
nothing to do with start time considerations.
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3. Page 8/ 1st paragraph under Pore Meltdown - It is not clear what
is included in SIS failure (not previously defined).
4. Page 34, Accident Time Scale - A discussion similar to this for
the PWR case would clarify the PWR containment discussion.
5. Page A-l, 1st paragraph under Pore Heatup Calculations - In view
of the application of the core heatup results to other PWRs and BWRs, the
statement that some of the results apply only to the specific designs
considered needs elaboration.
6. Page A-6, equation (A-9) - QMELT apparently should be C^,^.
7. Page E-9 - The pressures in this assessment should be labelled
psig or psia, whichever is appropriate.
Appendix X
Clarifications
1. Page 6, first paragraph - Although the site geology is
described, a description of what the plant is actually built on
is not mentioned, as was done for the BWR on page 7.
2. Page 45, Note (4) - The Bijlaard formulae have not been defined
in the text.
3. Page 94, first paragraph - It is indicated that the IHSIS (LPIS
elsewhere) injects into the PCS hot legs. Figure 11-53 of
Appendix II, Vol. 2, shows injection into the cold legs and the
text associated with the figure also indicates cold leg
injection.
4. Page 94, third paragraph - The discharge pressure of 300 psig
does not appear compatible with the 225-foot head stated on page
275 of Appendix II, Vol. 2.
5. Page 168, item 2 at bottom of page - This item states that the
assumption of a 40" tilt of the MSIV actuator axis is a
conservative assumption since "one expects vertical installation
to be the usual practice." It is not clear why the actual
orientation for the Surry plant was not determined in order to
establish the validity of this conservatism. (Figure 28 shows
an MSIV with about a 40° tilt to the actuator).
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Editorial
1. The nomenclature used for the various reactor systems is not
consistent with the rest of the Study. Examples are:
Apn A, Page 18 - Lew Head Safety Injection System vs Low Pressure
Injection Systems
High Head Safety Injection System vs High Pressure
Injection Systems plus Accumulator Systems
Containment Pvecirculation Spray Systems vs Containment
Spray Recalculation Systems
Core Spray Systems vs Core Spray Injection System
Residual Heat Removal Systems vs Post Accident Heat
itenoval.
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Additional Comments
Summary Report
Editorial
1. Page 2, 1st sentence - The sources for the results in Figures 1, 2, &
3 should be identified, and the figures explained in nore detail (ie, time
period covered, population covered, etc).
2. Page 8, 1st paragraph, last sentence - Depending on schedules and
definitions, this staterrent may be incorrect. Fort St. Vrain (330 Mfe-
I£TGR) should start up this year, and Fulton 1 (1140 We-IffGR) is scheduled
for startup in 1979.
3. Page 26, Section 2.21, 1st paragraph - A more effective qualification
of tlie v-aSH-740 results would be to quote the cover letter transinitting
the Study to the JCAE in March 1959. This letter, presumably written by
the authors of the report, says, in part:
"Pessimistic values, leading to great hazards, were cliosen for the
numerical values of many uncertain factors wliich influence the final
magnitude of the resulting damage. It can therefore be concluded that
tliese theoretical estimates are greater than the damages which would
actually result in the unlikely event of such an accident."
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CONTRACT WITH INTEIMXJNTAIN TECHNOLOGIES, INC.
CONTINUING WASH-1400 REVIEW TASKS
A. Failure Mode Paths Selected for Review
1. BWR-Reacto'r Protection System-Review to determine credit taken
2. BWR-Transient #1 for backup Boron injection under
3. BWR-Transient #2 BWR transients selected following
4. BWR-Transient #3 investigation of BWR Praetor
Protection System.
5, PWR-Electric Power Systems - Independent evaluation of Electric
Power System Availabildty-
6. PWR-High Pressure Injection Review to determine the extent that
System possible troublesome break 1 ocati.ons
7. PWR-Small Break #1 have been accounted for.
8. PWR-Small Break #2
9. PWR-Loss of Power Transient - Review relationships considered
between this accident sequence
and specific containment failure
modes.
10. PWR-Low Pressure Injection Review to determine range of
System applicability of assured failure
11. PWR-Low Pressure Recirculation paths and sensitivity of results
System on accident risk magnitude.
B. Critical Radiological Source Term Parameters Selected for Review
1. Core conditions prior to meltdown calculation
a. vessel residual water
b. blowdown heat-txarsfer
c. blowdown duration
2. Core meltdown calculation
3. Containment response
a. failure pressure
b. containment safeguards and containment pressure response.
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