ML19282C659
| ML19282C659 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck, Ginna, Yankee Rowe |
| Issue date: | 03/22/1979 |
| From: | Gossick L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Pell C SENATE |
| References | |
| RTR-WASH-1400 NUDOCS 7903300504 | |
| Download: ML19282C659 (2) | |
Text
b UfJITED STATES a
O'
't NUCLEAR REGULATORY COMMISSION j'
WASHlf!GTOrd, D. C. 20555 MAR 2 21979 The Honorable Claiborne Pell United States Senate Washington, D.C.
20510 -
Dear Senator Pell:
By note of March 5, 1979, you requested our views on a letter from Ms. Dorothy Sherman. Ms. Sherman, in her support for a moratorium on new construction permits, cited recent Commission action related to the Reactor Safety Study ("Rasmussen Report") and a Union of Concerned Scientists' claim about defects in nuclear power plants.
I am pleased to provide information on both of these subjects.
In light of the questions Ms. Sherman raised concerning reactor safety, I believe it is important to place in proper perspective the Reactor Safety Study (WASH-1400). A primary goal of WASH-1400, as established in 1972, was to obtain a " quantitative evaluation of the risk from the operation of a nuclear plant." The Safety Study was, in effect, a " measurement," made by analyzing two typical plant designs, of the effectiveness of an existing system of nuclear regulation.
The regulatory system depends on having nuclear plants sited, designed, constructed, and operated on the basis of conservative application of sound and accepted engineering principles, on requirements for multiple and redundant safety systems, and on a set of regulatory requirements that are updated to reflect operating experience.
The designers, builders, and operators of these plants are required to have effective quality assurance programs and their work is subjected to a continuing licensing and inspection process by the NRC.
The re-sults of the licensing and inspection process are, in turn, subject to independent review by the Advisory Committee on Reactor Safeguards and often to examination in public hearings.
What has now been concluded is basically that the " measurement" of the results of our regulatory system, as reflected in the overall risk estimates of the Reactor Safety Study, is much less precise than had been asserted. A specially establisiied Review Group was not able to conclude that the overall risk estimates were higher or lower than reported in the Rasmussen Report, although they speculated on pos-sible factors in both directions, but only that they thought the error bounds on those estimates were substantially larger than had been re-ported. On that account, they recommended to us that the overall risk 79033005o9
The lionorable Claiborne Pell 2
estimates of the Rasmussen Report should be used with great caution --
"should not be used uncritically" were their words -- in the regula-tory process or for public policy purposes. We have accepted and are implementing that recommendation, as well as the other findings and recommendations of the Review Group.
The health and safety regulatory system employed by the Commission, much of which evolved long before the Reactor Safety Study was car-ried out, is unchanged in its basic principles today.
It does not depend on the ability to make precise quantitative estimates of overall risk -- although that ability would be highly useful and should be developed.
We believe this regulatory system has served us well.
It is an ex-ceptionally rigorous system, and appropriately so in view of the technology we regulate.
It is our job as regulators to make sure that there is no undue risk from licensed facilities and, while one must acknowledge strongly held views to the contrary, over 4G0 reactor-years of experience to-date give us reason to believe that we are on the right track.
With the above clarification, I trust that accompanying enclosures provide the information you ha" requested.
Sincerely, JSim <a) T. / - a I,"
r Af Ve -. W. ca_;a
Enclosures:
1.
Memorandum from Samuel J. Chilk to Lee V. Gossick, subject:
" Staff Actions Regarding Risk Assessment Review Group Report," January 18, 1979.
2.
Discussion of Concerns Raised by the Union of Concerned Scientists.
'Cs Y.
UNITED STATES ff "'4 NUCLEAR REGULATORY COMMISSION' y
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- wAssincron, o. c. rosss g
" n-E January 18, 1979 A
Y OFFICE OF 'niE srcarrAny c
MEMORAliDUM FOR:
Lde V. Gossick Executive Director for Operat 4
i' FROM:
Samuel J. Chilk
_ Secretary of the Comissio (/
SUBJECT:
STAFF ACTIONS REGARDING RIJK AL ESSMENT REVIEW GROUP REPORT
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Attached is a pol' icy statement issued by the Comission on January '18,1979.
In addition, the Comission has provided the
'following instructions for the staff.
1.
Sendcopiesofthe'RiskAssessmentReviewGroupReport(NUREG/CR-0400) and of the January 18, 1979 Comission policy statement to all known domestic and international recipients of the RSS.
In the future, copies of the RSS Executive Sumary and the complete RSS will be distri.
buted only when accompanied by a copy of the Review Group's report and a copy of this statement.
2.
Quantitative risk assessment techniques and resul'ts'can be used in the licensing process if proper consideration is given to the results of the Review Group.
The, staff should use the following procedures re-garding the use of quantitative risk assessment techniques and.results pending development of further guidance:
In comparisons of risks from nuclear power plants with a.
other risks, the overall risk assessment results of the RSS (i.e., curves or tables of the probability of occurrence of various consequences) shall not be used -
without. an indicaticn of the wide range of uncertainty,
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asso'ciated with those estimates.
Any such use should note the difficulty of placing high confidence on '
estimates that are well below the values set by experience.
-2 b.' Quantitative risk assessment techniques ray be used to estimate the relative importance of potential nuclear power plant accident sequences.or other features where sufficient
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similarity exists so that the comparisons are not invalidated by lack of an adequate data base.
Such techniques should not be used to estimate absolute values of probabilities of failure
, of subsystems unless an adequate data base exists, and ;it is '
possible either to quantify the uncertainties or to support a conservative analysis.
T[ie quantitative estimates of event probabilitics in the
.c.
RSS should not be used as the principal basis for any regu-latory decision.
However, these estimates may be used for
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relative comparisons of alternative designs or requirements
~ provided that explicit considerations are given to the criti-cisms of those estimates as set forth in the Report of the Risk Assessment Review Group.
d.
The RSS consequence model shall not be used as the basis for licensing decisions regardirig individual nuclear power plant sites until significant refinements and sensitivity' tests are accomplished.
However, the consequence model may be used for relative comparisons provided that such estima es' are not the primary basis for such reviews and provided that
~ explicit consideration is given to the criticisms of the various elements of that model as set forth in the Report-of the Risk Assessment Review Group.
The staff shall prepare and submit by June 30, 1979, detailed pro-cedures to ensure the proper and effective use of risk assessment theory, cethods, data development 'and statistical analyses by the staff.
Pending review by the Commission of these detaile'd procedures and the bases and rationale supporting them, the Office Directors will obtain the advice of the ED0's Regulatory Requirements Review Comittee should questions
.arise regarding the implementation of the above instructions.
3.
The staff shall review the extent to which past and.pending 11-censing or other regulatory actions, including Comission ACRS and li-censing board actions and statements, have relied on the risk assessment models and risk estimates of the RSS.
The Comission will examine the results of this review to detemine whether the degree of reliance.,
identified was and continues to be justified and to decide whether regulatory modifications are appropriate.
9
3
-4.
The staff shall give special attentien to'those activities identified by the Review Group as bei.ng especially amenable'to' risk ~
assessment, i.e., deali.ng with_ generic safety issues, formulating new r,egulatory requirements, assessing and re-validating existing regulatory requirements, evaluating new designs, and fonnulating reactor safet;y.
research and inspection priorities
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5.
The staff shall prepare a review of current. HRC practices and pro-cedures in two areas of particular concern to the Review Group:
the peer review process for risk assessment devchpments.
a.
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and b.
the coordination among the research and probabilistic analysis staff and the licensing and regulatory staff, in order to promote the effective use of these techniques.
The Comission will make whatever changes are necessary to assure that effective pear review and interoffice coordination are integral features of NRC'.s risk assessment program.
6.
The staff shall examine the significance of the tei:hnical issues raised by the Raview Group and the appropriate courses of action for dealing with the n.
These issues include questions abou; statistical methods, data base quality and availability, consequence modeling, human factor considerations, earthquakes, fires, and conmon cause failures.
Thb Comission will address what (.hanges should be proposed in the approved FY 79 and proposed FY 80 research program to improve the data base, including that on human behavior.
As an addditional action, the staff shall undertake a review of statistical methods and human factor considerations used in risk assessment.
Attachment:
As' stated cc:
Chairman Hendrie Comissioner Gilinsky r
'Comissioner Kennedy Comissioner Bradford
' Comissioner Ahearne
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-Jimes 1,. Kelley, 0GC
- Kennefh Pedersen, OPE Joseph J. Fouchard, OPA Carlton C. Kamerer, OCA
. January 18, 1979
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~ NRC STATEMENT ON RISK ASSESSMENT AND
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THE REACTOR SAFETY STUDY REPORT (WASH-1400) f= =.-
IN LIGHT OF THE RISK ASSESSMENT REVIEW GROUP REPORT
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. The Risk Assessment R'eview Group, :hartered' by the NRC,in July,1977
.to " provide advice and information to the Cormission on the final report of the Reactor Safety Study, WASH-1400," and related matters, _lf
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.subnitted its report to the Comission on September 7,1978. The Review-Group, chaired by Professor Harold Lewis of the University of California.
at Santa B'arbara, 2] was formed in response 'to letters fr6= Congrassrian s
Udall,. Chairman of the House Comittee on Interior and.. Insular Affairs, c'xpressing misgivings hbout the Reactor Safety Study (WASH-1400), and in particular about the " Executive Sumary" published with the Main Report.
It was expected that the Review Group's report would " assist the Comission in e~stablishing policy regarding the use of risk assessment in the.
regulat'ory process" and that it would " clarify the achievements and
' limitations of the Reactor Safety Study."
In August,1972, the Chairman of the Atomic Energy Comission informed the Chaiman of the Joint Comittee on Atomic Energy that the Atomic Energy Comission had undertaken an in-house study "to provide a
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basis for submitting recomendations to the Congress regarding the extension or modification of the Price-Anderson Act." A draft version of the s.tudy report was circulated for coment in April,1974.
On L
October 30, 1975, the Nuclear Regulatory Comission 3/ announced that-the final report had been completed.
Criticism of the document following C-relhase centered on the mathod of treating peer cements on th'e draft-
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report as well as on the substance of the report. Thh NRC' press release' accompanyf ig publication of WASH-1400 praised the report, describing it as a "rea,istic assessment.... providing an objective and meaningful estimate of the' present risks associated with the operation of present day light water reactors in the United States," gave several comparisons to show that the risk from nuclear power was much less than from other man-made activities, and included a statement that "the final report is a soundly based and impressive work....
Its overall conclusion is that the risk attached to the operation of nuclear power plants is very low compared with other natural and man-made risks." _4f In view of the importance attached to the Reactor Safety Study, within and outside the Commission, both prospectively and after it was made public, the Comission has reexamined it's views regarding the Study in light of the Review Group's critique.
9 e?.* L &
h[ #g Uhile praislng the study's general methodology and recognizing.its Th t
9entribution to assessing the risks of nuclear power, the Review Group
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was critical of the Executive.Sumary, the procedure forlowed in producing the final report and the calculations in the body of the report.
Among the major failings of the study, the Review Group cited-ille Executive Summary: The Review Group concluded that "the Executive Summary of the RSS is a poor description of the contents of the report, should not be portrayed as"such, and.
f has lent itself to misuse in the" discussion' of reactor risks."
The Review Group indicated the Executive Summary-does not,
adeqtfately indicate the full extent of the consequences of
. reactor accidents and does~not.sufficiently emphasize the uncertainties involved in the calculations of their probability.
' As a result, the reader may be left with a misplaced confidence in
'the validity of the risk estimates and a more favorable impression of reactor risks in comparison with other risks than warranted by thestudy.S/
The peer Review process:
The Review Group Report criticized the. RSS staff response.. p.ointing..<ut that in s. cme cas,es cogent comtents from critics either were not acknowledged or were evaded and that, in general, the record of response to valid criticisni, The Report points out
, as weaker than it should have been.
that the lack of clarity of WASH-1400 itself led-to major diffi -.
w rF culty in tracing a line of thought through the study and crippled many efforts to accomplish responsible peer reviews.-
Accident Probabilities:
The Review Group was unab.le to deter--
mine whether the absolute probabilities of accident seqiiences in WASH-1400 are high or low, but believes that the error bounds on those estimates are, in general, greatly understated.
This, the Report said, is true in part because there is in many cases an inadequate data base, in part because of an inability to quantify comon cause failures, and in part because of some.
questionable methodological and statistical procedures.
The Reyiew Group also criticized, in some cases severely. various of the calculation,al techniques in the Study as well as its lack of clarity.
.The R'eview Group cited the following as major achievements of the st'udy':
" WASH-1400 was a substantial advance over previous attempts to
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estimate the risks of the nuclear option.
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159
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" WASH-l'400 was largely successful in at least three ways; in making the study of reactor safety more rational, in *
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. establishing the topology of rany accident sequences, and in delineating procedures through which quantitative estimates of the risk'can be derived for those sequences for which a
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data base exists.
'i'Despite its shortcomings, d4SH-1400 provides at this-time.
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the most complete single picture of accident probabilities associated with nuclear reactorse The fault-tree / event-tree approach coupled with an adequate data base is the best available -
J tool with which to quantify these probabilities..,
" WASH-1400 made clear the importance to reactor safety dis-cussions of accident consequences. ther than early fatalities."
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The Commission _ accepts these findings and takes the following a.ctions:
Executive Summary: The Commission withdraws any explicit or
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implicit past endorsement of the Executive Summary.
The Peer Review. Process 0 aThe Comissi.on. agrees. that. the_g'y....
peer review-process followed in publishing RASH-1400 was inadequate and that proper peer review is fundamental to The Commission will take making sound, technical decisions.
+EE whatever corrective action is necessary to assure that effective peer review is an integral feature of the NRC's
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risk assessment program.-
Accident Probabilities:
The Commission accepts the Review Group Report's conclusion that absolute values of the risks presented by WASH-1400 should not be used uncritically either in the regulatory process or for public policy purposes and.
has taken and will continue to take steps to assure that any such use in the past will be corrected as appropriate.
In particular, in light of the Review Group conclusion ~s on accident probabilities, the Commission does not regard as reliable the Reactor Safety Study's numerical estimate of the overalT risk of reactor accident.
. Communication with the Congress and the Public:
Commission correspondence and statements involving RASH-1400 are being reviewed and corrective action as necessary will be taken.
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With respedt to the component parts of the Study, the Comission expects the staff to make use of them as appropriate, that is, where the data
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base is adequate and analytical techniques permit. Taking due account of the reservations expressed in the Review Group Report ar.d in its presentation to the Co;miission, the Conaission supports the extended use of probabilistic risk assessment in regulatory decisionmaki'ng.
Thi Comission has provided additional detailed instructions to the.fiRC staff concerning continued use of risk assessment techniques and results.
. in response to specific criticisms raised by the Risk' Assessment Review s
Group.
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NOTES
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Its charter reads:. The Review Group will provide advice and
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information tc the Commission regarding the final report of the Reactor Safety Study, WASH-1400, and the peer corr.ments on
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the Study, advice end; recommendations on. developments in the-
- field of risk assessment methodology and on future courses of
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action which should be taken to improve this methodology and its application. This advice and infomation will, assist the Commission in establishing policy regarding the use of risk s
assessment in the regulatory process, in improving the base for the use of such assessments.
It.will also clarify the achieve-ments and limitations of the Reactor Safety Study."
..c 2f.
The other members were Dr. Robert J. Budnitz (Lawrence Berkeley Laboratory, University of California)
Dr. Herbert J. C. Kouts (Brookhaven National Laboratory), Dr. Walter Lcewenstein (Electric Power Research Institute), Dr. William Rowe (Environ-mental Protection Agency),' Dr. Frank von Hippel (Prim.eton University) and Dr. Fredrik Zachariasen (California Institute ~
ofTechnology).
Dr. Budnitz is presently on leave from the University of California and is serving (since August 1978).as Deputy Director of the NRC's Office ~of Nuclear Regulatory ~
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Research..
- =mm
- The Nuclear Regulatory Comission was established on January 19, 3/
1975 to carry out the regulatory functions of the Atomic Energy
==-
Commission, which was abolished on that date.
The press release at the time of publication said that the report 4]
is "the culmination of the most comprehensive risk assessment of nuclear power plants made to date.
The objectives of the The overall study were to make a realistic assessment....
conclusion...is that the risks attached to the opcration of present day nuclear power plants are very low compared to other Nuclear power plants are about natural and man-made risks....
10,000 times less likely to produce fatal accidents than man-Non-nuclear accidents. involving made non-nuclear activities....
comparable large dollar value damage are about 1,000 times The chance more 1,i,kely than nuclear power plant accidents....
that a person living in the general vicinity of a nuclear power plant will be fatally injured in a reactor accident is one in five billion per year....
In the event of an unlikely reactor accident with a probability of one in a million per reactor per year, latent health effects except for thyroid nodules would be such a small percentage of the nomal incident rates that they would be difficult to detect...."
m 8
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.srs thifi The tufC Chairman was quoted as saying, "The Comission believes that the Reactor Safety Study Report provides.
an objective and meaningful estimate of the public risks associated with 'the operation of present day light water reactors in' the. United States....
The.
final report is a soundly based and impressive work....
Its overall conclusion is that th risk attached to the operation of nuclear power plants is very low compared s
with other natural. and man-made risks." The press release went on to sai that more than 1800 pages of comments were received from a broad spectrum.of people and all'were carefully considered in preparing the final report. -.
5]..
professor !.ewis..in reporting -to the Comission..said that the Executive Sumary was not a sumary of the' report.
He concluded it was written as a public statement that reactors were safe compared to other risks to which the public is exposed and he stated it should not have been attached to the report and described as a part of it.
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DISCUSSION OF COI!CERtS RELATED TO WASH-1400 RAISED BY THE UtlION OF CONCERiiED SCIENTISTS
DISCUSSION OF CONCERNS RELATED TO WASH-1400 RAISED BY THE UNION OF CONCERNED SCIENTISTS The staff's use of the RSS was commented on in a Union of Concerned Scientists (UCS) pr'ess release issued on January 26, 1979.
The first two paragraphs of that release stated:
"The government may have to require shutdown of at least 16 of the country's nuclear power stations, the Union of Concerned Scientists said today.
The continued operation of these 16 plants, despite officially acknowicJged safety defects, has been allowed based on theoretical risk estimates from the so-called Rasmussen Report.
(See Attachment 1 and 2 for plant names and locations.) In an unprecendented act, the government repudiated the Rasmussen Report last week." of the UCS press release If sted 12 operating plants that the NRC Staff had previously identified as " Plants Requiring an Alternate or Dedicated Shutdown System" in a table following page 4 of our memorandum to the Commission of July 6,1979, concerning the UCS Petition for Reconsideration of May 2,1978. As discussed in our memorandum to the Commission, this Staff finding was based on our ongoing fire protection reviews of operating plants. of the UCS press release listed 6 operating plants in which " defective equipment was discovered as a result of the Union of Concerned Scientists' Petition to the NRC of November 1977."
These
. pir ats had been previously identified in various NRC St.ff memoranda to the Commission on the UCS petition as plants in which electrical equipment in safety systems had been replaced or requalified because of the inability of licensees to demonstrate, on the basis of available information, that the equipment <as environmentally qualified for appropriate accident conditions, wo of the plants listed in UCS's Attachment 2 were also listed in r
Attachmsnt 1.
The combination of these two lists apparently provided the basis for tnt number of plants specified in the title and lead of the UCS press release.
In another attachment to the press release, UCS listed and discussed "Three serious safety hazards which the NRC has not acted upon because of past NRC reliance on the Rasmussen Report." These three items were " Safety system electrical cables will fail in fire (discussed above); Safety system equipment cannot withstand accident it is designed to control" (also discussed above); and " Catastrophic accidents for which there is no protection required."
. Based on the foregoing, it appear, that the principal UCS concerns related to the use of WASH-1400 that formed the basis for its press release of January 26 are fire protection measures in operating p* ants, performance of electrical equipment in safety systems of operating UCS concerns plants, and class 9 accidents in operating plants.
over staff use of WASH-1400 in response to the UCS petition concerning fire protection and electrical connectors were also excressed in its letter to the Commissioners of October 16, 1978.
UCS concerns over the role of WASH-1400.r defining Class 9 accidents were also expressed in its letter to the Commissioners of November 1,1978.
The fire protection, electrical equipment and class 9 issues are discussed in more detail in the enclosures to this report.
FIRE PROTECTION FIASURES IN OPERATING PLANTS NRC regulations require, in General Design Criterion 3 (10 CFR Part 50 Appendix A), that structures, systems and component important to safe *y of nuclear power plants be designed and Mcated to minimize, consistent with other safety requircments. the probability and effects of fires.
Prior to the fire at the Browns Ferry Station that occurred on March 22, 1975, NRC staff requirements for fire protection were minimal.
Since that time, considerable staff effort has been devoted to developing comprehensive guidance to assure that nuclear plants have adequate fire protection systems. This effort has produced the guidance set forth in Section 9.5.1 of the NRC Standards Review Plan and in Regulatory Guide 1.20.
These review procedures are used to evaluate fire protection systems for all facilities.
And, although they contain considerably more stringent requirements than were imposed prior to 1975, they are being backfitted on all existing operating plants.
During this backfitting period, the safety of the nuclear plants during the time interval taken to implement these requirements has been questioned.
In fact, this question has been of paramount impo-tance and the focus of staff attention since the Browns Ferry Fire. This question was raised during Congressional Hearings in September 1975 and was most recently resurrected by the Union of Concerned Scientists (UCS) in the January 26, 1979, press release announcement. The basis for the most recent questioning is the premise that the reliance of plant safety was based mainly on the results of the Reactor Safety Study (WASH-1400). This premise is incorrect.
i e
. The basis for the staff conclusion that these plants can continue to operate while these modifications are being made does not depend on t'.e conclusions of the WASH-1400_ study.
No reliance has been placed by the staff on quantitative analyses derived from WASH-1400 in the licensing reviews of fire protection systems for operatir.g plants.
The NRC's Special Review Group Report on the Browns Ferry Fire (NUREG-0050),
and WASH-1400 Appendix XI indicate the limited applicability of the Reactor Safety Study calculations on fire risks. These reports state that rather straightforward measures can be used to improve fire protection and fire fighting capability and that these improvements would significantly reduce the likelihood of a core melt accident caused by fire. As will be discussed further, the staff's conclusions on plant safety is primarily dependent on these measures and not on quantitative assessments of risk caused by fire.
The staff's current judgment that the probability of occurrence of serious fires is low in operating plants is based upon the plant conditions in 1975, the subsequent positive and effective actions taken to improve the fire protection programs at nuclear plants since that time and the plant conditions observed during our site visits.
This judgment
is independent of and corroborates with similar judgments arrived at previously by the Browns Ferry Special Review Group and Reactor Safety Study.M This staff conclusion on continued plant operation pending the implemen-tation of additional fire protection measures was expressed in a memorandum of December 15, 1977, to the Commission on the UCS petition which presented the overall staff conclusion and recommendations concerning the petition. Page 33 of that memorandum includes the following statements:
"6.
Basis for Continuation of Plant Operation and Licensing The staff has previously indicated its basis for the continued operation of licensed plants pending completion of the full im[lementation of our current fire protection guidance in the November 9,1977 report (pp. 8-9) and the November 22, 1977 report (pp. 12-18).
This basis includes,
- 1) the actions taken as a result of the I&E inspections and subsequent follow-up actions by licensees; 2) the conclusions of the Brons Ferry Fire Special Review Group Report (NUREG-0050) that the probability of fires EIt should be noted that the analysis presented in WASH-1400 showed that the potential for a significant release from a severe fire was about 2G". of that calculated from all other causes analyzed. This analysis could be interpreted to infer that no further actions were needed for fire protection. This clearly was not the approach recommended by the Special Review Group or adopted by the staff after the Browns Ferry Fire.
_4 of a large and disruptive nature of the magnitude of the Browns Ferry fire is small and that 'there is no need to restrict operation of nuclear power plants for publi.
safety', and 3) improvements made since that time by licensees in fire presention measures and fire brigade capability and training that have been noted in the plants visited to date and are expected to exist in the remaining plants, which further reduce the probability and consequences of fire."
A similar staff conclusion was expressed in a July 6,1978, staf f memorandum to the Ccmmission on the UCS petition for reconsideration.
Page 45! of that memorandum includes the following statements:
"For those plants not yet evaluated, and those plants for which the staff has required enhancement of the fire protection system, the staff believes that the probability of occurrence of severe damaging fires is acceptably low for the interim period until staff evaluations and licensee enhancements are completed.
This conclusion is based upon the information discussed by the Browns Ferry Fire Special Review Group in NUREG-0050 and upon
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This page immediately precedes the Table identifying the 12 plants requiring an alternate or dedicated shutdov;n system which was included as an attachment to the UCS press release.
the additional defense-in-depth protection provided by the staff's overall fire pr' tection upgrading program which provides (1) controls ovedr ignition sources, combustibles and access to the areas, (2) physical separation and use of flame retardants to delay or prevent propagation, and (3) fire detection, fire suppression and trained fire brigades to effect prompt manual suppression of fires."
The Commission's Memorandum and Order on the UCS Petition of April 13, 1978, on pages 36 through 40 discusses the Commission's conclusion concerning the need for immediate action with respect to fire protection of operating plants and the bases for its conclusion that no immediate action is necessary.-
The bases discussed in this Order are generally consistent with those advanced by the staff discussed previously, but additional details concerning the view of the Browns Ferry Fire Special Review Group are provided by a quotation from its Report (NUREG-0050).
Within the quotation is a paragraph describing the results of the RSS probabilistic assess-ment of the risks of fires.
It is this particular paragraph that apparently provides the basis for the UCS statement in its press release that "the Commission adopted this Reactor Study finding...."
as a basis for allowing these plants to continue to operate.
-3/
The UCS press announcement quotes only 1 paragraph of this multi-page discussion, and omits the footnote applicable to the specific portion it did quote.
As previously mentioned, since 1975, there has been a significant improvement at each operating plant in the fire prevention program, and in the capability of plant equipment to detect and extinguish fires promptly. These improvements have provided the primary basis for the staff conclusion that a severe fire in operating plants is not likely. To provide a better understanding of these improvements, we review the accomplishments achieved since the Browns Ferry fire:
(a) As a result of the Office of Inspection and Enforcement's special bulletins to all licensees of operating power reactors on March 24, 1975, and April 3,1975, directing controls over ignition sources, a review of procedures for alternate shutdown and cooling methods, and a review of flammability of materials used in floor and wall penetration seals, some of the changes and improvements at operating plants are:
(1) modifications of plant administrative procedures for work permits to assure consideration of the safety significance of electrical cables and piping in the work area; (2) incorporation of the control of combustible materials into plant administrative procedures; (3) improved plant administrative procedures for the control of ignition sources:
. (4) development of new procedures and guidelines covering the use of water on electrical cable;S/
(5) study and development of procedures for a variety of means to provide decay heat removal; (6) addition, upgrading and repair of cable penetration fire stops; and (7) addition of fire suppression equipment.
(b) As a result of the special inspections by the NRC Office of Inspection and Enforcement completed at all operating power reactors in April and May 1975 covering the installation of fire stops on electrical cables and penetration seals, any inspection findings which reflected noncompliance with then current NRC requirements resulted in prompt corrective action by licensees. Follow-up I&E inspections have confirmed that licensees implemented the required corrective actions and that administrative control procedures were in place.
(c) More detailed procedures for inspection of fire prevention and protection measures have been incorporated in the NRC Operating Reactor Inspection Program.
Since September 9, 1975, the Office of Inspection and Enforcement has been conducting detailed annual inspections of licensees' fire protection 4/ This improvement alone tends to preclude the development of a fire in plant areas in the proportions of the Browns Ferry fire.
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programs as one of the routine I&E inspection modules.
In addition, a plant tour is conducted quarterly, during which the inspector looks for conditions that might contribute to fires. These inspections include review of fire insurance inspection reports.
The results of_ these. inspections show that the licensees' fire protection programs have improved and that the licensees have an increased appreciation for the need for effective fire prevention measures and improved fire-fighting capability.
Some of the more significant improvements include:
(1) The licensees have completed sealing the safety-related cable penetrations and have instituted controls to ensure that penetrations are not subsequently degraded.
(2) The licensees have surveyed their plants, identified the possible fire hazards, and are in the process of eliminating s
or reducing the identified hazard.
In cases where the hazard could not be eliminated, additional automatic fire protection systems have been or will be installed.
In other cases in which fire protection of safe shutdown systems was uncertain, modifications to provide alternate shutdown methods have been or will be installed or additional fire protection measures taken.
Periodic tests of fire protection systems are now being performed.
(3) Fire brigades have been established or enlarged and formally organized with duties and defined responsibilities.
Formal training and requalification programs including perisdic drills, have been or wiis be implemented.
(4) Formal agreements with local fire departments have been established and joint participation in some fire drills is now taking place.
(5) Administrative controls have Deen implementea to limit the use of combustible materials within the plants. The control of ignition sources has also been improved by limiting the use of open flames in the nuclear plant.
When open flames are required, such as welding or burning, trained f're watches are used to monitor the operation and to take any necessary corrective action.
(d)
Quality assurance inspection procedures have steadily improved over those in effect at the time of preoperational QA inspections of Browns Ferry Units 1 and 2.
(e) As a result of implementing the staff's improved guidelines on administrative cpntrols for fire protection, the licensees' fire protection programs now provide the following:
(1) Control use and storage of combustible materials; (2) 1.imit use of ignition source ano provide protective measures where ignition sources are used, including fire watches in critical areas;
4 e
- 10 (3) Establish training programs for fire brigades to include classroom instructions, " hands-on" practice, and periodic drills; (4) Establish fire-fighting procedures to include notification of operators, offsite fire departments, fire brigade and any other required personnel, and procedures that include fire-fighting strategies for fires in specific safety-related areas.
(f) Technical Specifications for fire protection systems have been incorporated into the license of each plant to assure the operability of fire protection systems.
For example, the specifications require the periodic testing, inspection, and surveillance of fire protection systems and equipment (i.e.,
fire protection instrumentation, fire suppression systems, firehose stations, penetration fire barriers) and require compensatory actions when such equipment is deemed inoperable.
These specifications establish the minimum shift strength of the onsite fire brigade for each individual plant. The specifi-cations also require a periodic independent fire protection and los: prevention program inspection and audit that uses either qualified offsite licensee personnel or an outside fire protection firm to assure that the fire protection program is being properly carried out..
All operating plants have been visited by the staff's fire protect'un review teams, and improvements in fire prevention and fire fighting continue to be made as a result of the ongoing evaluations. During these visits, we have also observed an increased awareness on the part of the utility management and plant personnel that fire prevention and fire-fighting readiness are important elements in the fire protection of their plants. Correction of deficiencies in fire prevention are being made on a schedule that is commensurate with the concern expressed by both the licensees and NRC staff. We have alto found that, at every facility reviewed to date, the licensees have established administrative controls that substantially conform to staff guidelines.
In the overall conclusions of our fire protection Safety Evaluation Report (SER's) on individual plants, we have also referenced certain comments made by the Special Review Group report on the Browns Ferry Fire (NUREG-0050).
The Special Review Group concluded that there was no need to restrict operation of nuclear plants based on (1)
WASH-1400 conclusions that fires contribute negligibly to the overall risk of nuclear plant operation; and (2) the Special Review Group's conclusion that, based on its evaluation of events occurring before, during and after the Browns Ferry Fire, the probability of disruptive fires of the magnitude of the Browns Ferry event is low.
The staff's bases for not restricting operation pending completion of our reviews and implementation of all modifications were the recommendations of the Special Review Group as well as the actions
- 12 already being taken at operating plants to further reduce the likelihood of disruptive fires. These other actions include the administrative controls, the fire brigade staffing and training, and the previously discussed technical specifications.
At plants where the staff review found that smaller fires may affect safe shutdown, immediate preventive actions were taken to establish methods to safely shutdown the reactor if such fires were to occur.
The Reactor Safety Study '.<as one element considered by the Special Review Group in making its recommentation; however, the RSS is not a primary element of the staff's basis to allow continued operation pending im ' ementation of all facility fire protection modifications.
The RSS calculations did not consider the effects of actions taken in the plants since 1975 and, therefore, are not indicative of the risk of fire in any presently operating plant.
The UCS press release states that fire tests sponsored by NRC at Sandia Laboratories in 1977 and at Underwriters' Laboratories (UL) in September 1978 showed that plant designs meeting current standards do not provide adequate protection against fire. The " current standards" implied by the UCS statement are Regulatory Guide 1.75 and IEEE-384, which deal with separation of redundant safety systems, and IEEE-383, which specifies cable fire-retardant criteria. With regard to these standards, the staff agrees with the UCS that these standards alone do not provide adequate protection against fires.
Since the Browns Ferry Fire, the staff has taken a position that sole reliance should not be placed on these standards for fire protection of nuclear power plants.
, The staff considers that the Sandia tests confirm the validity of the staff po,ition; namely, that Regulatory Guide 1.75 cable separation criteria and IEEE-383 cable flame-retardant criteria by themselves are not sufficient to protect against exposure to fires and that additional fire protection measures are required.
These additional measures include fire barriers, fire-retardant coating on cable, automatic fire detection and extinguishing systems, backup fire extinguishing capability (fire hoses and portable extinguishers), administrative procedures and controls to minimize fire hazards due to poor housekeeping or to plant maintenance activities, and plant fire brigade staffing and training to assure adequate response to fire emergencies.
This staff position was taken more than a year before the 1977 Sandia cable fire tests.
Thus, the test results confimed the staff position that additional fire protection measures beyond Regulatory Guide 1.75 and IEEE-383 were necessary as a safe and conservative basis for the plant fire protection evaluation program that is now being implemented.
The UL tests referred to by UCS were generic separate effects tests that did not test a specific fire protection configuration in an operating plant.
Vertical cable trays (unbarriered) have been identified during the course of the operating plant review that are grouped in a manner similar to the tested cable tray configuration. Licensees have
proposed various systems of fire protection for vertical cable trays that include the use of fire barriers. The types of barriers proposed to protect redundant divisions of saf, tty-related cables include cable tray covers, cerar,1c wool blankets with tray covers, insulating board
~
material (Marinite), and fire-retardant coatings.
Representative barrier and suppression systems will be, or have been, tested in NRC-sponsored or licensee-sponsored test programs.
The particular barrier configuration chosen for the UL test is that currently recommended by the barrier material manufacturer to protect cable trays. The UL test provided data with which the staff can evaluate such barrier systems.
The UCS press release also includes a memorandum from Mr. Cohn of Gage-Babcock dated September 30, 1977. The press release alleges that Gage-Babcock agrees that fire protection is inadequte in many existing pl an ts.
The staff response to Mr. Cohn's concern is discussed at length on pages 5,13 and 14 in the staff response to the Commission dated July 6,1978, on the subject of the UCS petition for reconsidera-tion. A brief summary of the staff response follows:
(a) At a meeting with the staff on October 20, 1977, subsequent to the September 30 memorandum, Mr. Cohn agreed that no changes in the NRC fire protection guidelines were necessary.
- 15 (b)
In a letter dated July 1,1978, Mr. Cohn reiterated his position on fire safety in nuclear power plants as follows:
"It is my belief, based on my knowledge and experience of conditions in the plants I have either personally visited or discussed in depth with our engineers, that
~-
sufficient precautions have been taken and that operation can ccntinue in the interim period during which additional measures are implemented to fully meet NRC guidelines."
The Nuclear Regulatory Commission staff reiterates its previous conclusion that the fire protection features i:: operating plants are adequate to permit operation during the interim period until certain additional fire protection features are installed.
This conclusion, previously explained, is primarily based on the many improvements in fire protection systems already accomplished in operating nuclear plants as a result of staff review which began immediately following the 1975 Browns Ferry Fire and does not rely on WASH-1400 results or calculations.
Performance of Electrical Eauipment in Safety Systems of Goerating Plants NRC regulations require, in General Design Criterion 4
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(10 CFR Part 50 Appendix A), that structures, systems and components important to safety of nuclear power plants be designed to accommodate the effects of the environmental conditions associated with postulated accidents.
The staff review procedures and acceptance criteria for the environmental qualification of safety equipment were first developed on a case-by-case basis in the late 1960s and are now contained in the Standard Review Plan issued in 1975 and in several national standards and Regulatory Guides, principally Regulatory Guide 1.89, also issued in 1975.
These review procedures and acceptance criteria are used in the review of all new CP and OL applications.
Prior to 1975, earlier versions of related national standards were used for CP and OL reviews. Thus, the plants now in operation have been reviewed against detailed acceptance criteria that have changed (in fact, grown more stringent) with time. All plants, however, must meet the same overall requirement of General Design Criterion 3 of having safety equioment qualified for service in an appropriate accident environment.
In November 1977, the UCS filed a petition with the NRC regarding the effects of fires and the validity of environmental qualifi-cations of a certain type of electrical connector used in safety systems.
. Act',ns by the NRC in connection with the petition identified a num;er of plants which had insufficiently cualified electrical connectors or insufficient documentation of environmental qualifica-tions. T'1e questionable connectt. s have been either replaced or requalified in all operating olants.
In the course of acting on the UCS petition, the staff identified several other tyoes of electrical connections that were used in safety systems and also had questionable qualifications.
These too were either replaced or requalified by the operators of the plants in which the equipment existed.
The UCS and the staff have also raised the question of whether this experience with the special class of electrical equicment, namely electrical connections, is indicative of a general inadequacy of environmental qualifications of electrical equipment in safety systems of operating nuclear power plants.
The staff, on its own accord and in response to the Commission's April 1978 Memorandum and Order on the UCS Petition, has ongoing activities in the inspection, licensing and research areas to confirm its present judoment that electrical eauipment in safety systems of operating nuclear power plants, previously considered to be acceptably qualified, remains acceptable in light of today's knowledge.
These actions are described in a number of staff filings with the Commission on the U:. Petition and are publicly available.
Two useful summaries are NUD.EG-0413 published in February 1978 that describes the evolution of environmental qualification criteria, including the IEEE-323,1971 standard criticized by the staff as referenced by the UCS, and NUREG-0458 published in May 1978 that contains a short term
, safety assessment of the environmental qualifications of safety-related electrical equipment 'n eleven of the oldest ;perating reactors.
The NRC sta/f has not relied upon the Reactor Safety Study for its conclusion that plant operation could safely continue pending resolution of questions about the continued acceptability of the earlier qualification of electrical qualification of electrical eauip-ment used in safety systems of operating plants.
The basis for the staff conclusion on continued operation is described on pages 34-36 of Appendix B of a staff memorandum to the Commission dated December 15, 1978, in connection with the UCS Petition (later published as NUREG-0413).
In reachina a conclusion on the environmental qualification aspects of the petition, the staff stated that:
"In reaching the judgment that no immediate action is required on operating reactors, the staff, as discussed elsewhere in this report, considered the following:
1.
Nuclear power plants include provisions, such as redundancy and diversity, to cone with equipment failures without affecting the oublic health and safety.
2.
Operating exoerience indicates that electrical equipment has performed adequately under both normal operating environmental conditions and on the few occasions where severe environmental conditions have existed.
. 3.
Even the older operating reactors used conservative design and construction practices and m.ny improve-ments have been made in the area of ereironmental qualification.
4.
A preliminary audit of the environmental qualification of electrical connectors and penetrations in ooerating reactors has indicated that there is ceasonable assurance that this equipment would perform its safety function under accident conditions even though complete documentation is not readily available in all cases.
It is the staff's belief that these findings would be essentially the same for other safety-related equipment.
5.
The likelihood that essential safety-related eouipment or other non-safety equipment would not perform the necessary safety function prior to failure due to environmental reasons coupled with the likelihood of a major accident requiring the performance of this equip-ment is very low.
6 The regulations have included reauirements for environ-mental qualification and a comprehensive quality assurance program since 1971.
The requirement for environmental qualification was included in initial versions of these regulations in the mid-1960s.
The NRC compliance effort by the Office of Inspection and Enforcement has emphasized review of environmental qualification test results for safety systems in its routine inspection program."
One part of the six reasons relied on by the staff is that "the likelihood of a major accident requiring the performance of this equipment is very low." This statement was auoted in the UCS press release and apnarently provides the only basis for the UCS concern that the Rasmussen Report was relied on for the staff conclusion that no immediate action is required for operating plants because of environmental qualification concerns.
_S.
. e.
The staff did not rely'on the results of the Reactor Safety Study in reaching its overall conclusion nor this specific conclusion on this issue, nor has it conducted a specific, independent analysis of the probability of a major accident in arriving at this judgment.
Rather, the staff meant by this statement that in deciding whether immediate action should be taken to further reduce risk to the public, the existing level of orotection provided in the facilities to orevent the occurrence or loss-of-coolant accidents was considered.
That is, the staff considered the past experience in commercial power reactor operation that is sufficient to demonstrate that the likelihood of such events is low.
Data developed from similarly designed high pressure piping systems in other industries is in agreement with this experience.
This engineering experience is sufficient to support a judgment by the staff on the low likelihood of occurrence of such an accident, which is, in turn, a part of the overall basis for requiring no imnediate action where the technical data also provides reasonable assurance that the safety-related electrical equipment will provide its intended accident mitigating function.
As initially outlined in the staff memorandum of December 15, 1977, and subsequently on July 6,1978, in Item 11 of Enclosure 1 of another staff memorandum to the Commission concerning the UCS Petition, the scope and timing of staff programs to provide additional confidence that adequate environmental qualification exists for safety equipment in
.... operating plants is based on several factors, including the likelihood of a major accident requiring the performance of this eouinment.
The degree to which this factor has shaped the staff's actions is difficult to quantify, but other licensing actions taken by the staff serve to illustrate the partial reliance placed upon the low probability of an accident.
In cases in which the licensing staff has low confidence that equipment important to safety would function in an accident, plants are required to shutdown and remedy the problem (e.g., see description of staff actions on D. C. Cook Unit 1 in the memorandum to the Commission on November 18, 1977).
Such decisions flow from the staff's view that the low probability of an accident within the design basis does not, by itself, provide a sufficient basis to permit continued operation in the face of significant related questions regarding the safety of a plant.
That is, the low orobability of a severe accident is not considered to be sufficient justification, by itself, to allow continued operation in light of a staff judgment that safety equipment provided to mitigate the consequences of that accident is not likely to function under the accident conditions expected.
In other cases, the staff has judged that the available technical information was sufficient to conclude that the equipment was likely to perform adequately or could be demonstrated to be qualif.ied and that reasonable time should therefore be allowed to complete the demonstration of the qualifications.
,,.. The tia:e frame allowing for confirming or further documenting qualifications was chosen, in part, based upon the generally understood low likelihood of occurrence of an accident that would environmentally challenge this equipment during that time frame.
In continuing to recommend that no immediate action need be taken, the staff relies neither on the Reactor Safety Study nor solely on the low likelihood of a major accident, but rather is guided primarily by the judgment, as discussed in Commission memoranda of December 15, 1977 and May 12, 1978, and in NUREG-0458, that the electrical equipment in safety systems of operating plants will not fail before performing its safety function when exposed to expected accident conditions, and there is ongoing work by the staff to confirm this conclusion for these plants.
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,'y. f Februlry 3,.:.:
I
~E HE E.
Hon. Claiborne deb. Pell 325 Pussell Senate office Blde, Washington, D. C. 20510
Dear Senator:
I am usinr ny perogative as an American citizen to participate in the decision-making processes of my governnent.
Now that the Nuclear Regulatory Commission has withdrawn endorse-ment of the Rasmussen Report, the necessity of a nuclear reappraisal becomes appa re nt; Formerly the Rasmussen Report was used as a criteria for safety.
It has now been proven that the Report was flawed and may understate or overstate the risk of a nuclear catastrophe.
I have received a communication from Representative Hamilton Fish, Jr.
(R. NY) stating that he has introduced legislation addressing nuclear power The Nuclear Energy Reappraisal Act which would halt the is-plants, namely:
suance of construction licenses for new nuclear power plants for 5 years while a thorough study of the entire nuclear fuel cycle is conducted by the Office of Technological Assessment.
I am writing to ask you to cosponsor this Nuclear Reappraisal Act.
The Union of Concerned Scientists says there are two main types of defects in plants - electrical cables that cannot withstand fire and mi ht E
let all safety systems be disabled by a blaze and safety equipment that cannot survive the type of accident it was designed to control.
This information was sent to the Providence Journal via United Press It also listed twelve plants which were risky International from Washint on.
t as having either or both of the sbove named ha zards. They are:
The Hadding Neck Plant in Conn.
Brunswick Nos. 1 & 2, Southport, N.C, Plymouth No.1 Plant at Plynouth, Mass. Ocanee Nos.1, 2 & 3, Seneca, S.C.
enkee Rowe in Eowe, M8ss.
Rancho Seco, Clay STA., Calif.
o r.
.I,
- 5. C.
- 2. 2. C. r k '..
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f r-ree F:ile Irland.ks.
.2 2 at Middle town, Fa. Erowns. c rry :.. ;, :. L ;r.
Irojan at Prescott, Oreton Cinna, Onta ri o,
.. Y.
The number of the Nuclear Enercy P.eappraisal J.ct bill it H.?.
336.
It should be of great concern. to not only we, the Ameri..ans but to all ci tizens of the world particularly in this time of stress, This bill would address the many econonic, environments 1 end moral ca nce rs of nuclear power which are of very great concern to me. Please co-sponsor thic bill and let me know what you are doing to protect us from the dangers of nuclear power.
Sincerely yours, Dorothy Sherman e
"W;regf
)
- ' ~
local PDR Docket No.
Haddem Neck 50-0213 Pilgrim, No. 1 50-0293 Yankee Rowe 50-0029 Brunswick No. l 50-0324 Brunswick No. 2 50-0325 Oconee No.'l 50-0269 Oconee No. 2 50-0270 Oc'onee No. 3 50-0271 Rancho Seco 50-0312 Robinson No. 2 50-0261 Three Mile Island No. 1 50-0289 Three Mile Island NO. 2 50-0320 Trojan 50-0344 D. C. Cook No. 1 50-0315 D. C. Cook No. 2 50-0316 Browns Ferry No. 3 50-0296 Ginna 50-0244