ML20205C795
| ML20205C795 | |
| Person / Time | |
|---|---|
| Issue date: | 01/21/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| REF-GTECI-A-17, REF-GTECI-SY, TASK-A-17, TASK-OR ACRS-2484A, NUDOCS 8703300262 | |
| Download: ML20205C795 (52) | |
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Meeting Minutes on the January 7, 1987 b ;",' s . ' 4D Meeting of the ACPS Subcomittee on Safety Philosophy, Technology, and Criteria
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The ACRS Subcommittee on Safety Philosophy, Technology, and Criteria met on January 7, 1987 and 1717 H Street, N.W., Washington, DC. The purpose of-this meeting was to: (1) continue the review of the NRC Staff proopsed plan for implementing the Safety Goal Policy, (P.) continue the review of the NRC Staff's proposed resolution to USI A-17, " Systems Interactions in Nuclear Power Plants," and (3) to review the status of the NRC Staff work on the steam generator overfill issues. The Subcommittee discussions begin at 2:00 p.m. and were concluded at 7:20 p.m. All of the discussions were held in open sessier.
The attendees at this meeting were as follows:
ACRS NRC Staff D. Okrent, Subcommittee Chairman M. Taylor, E00 F. Remick, Member N. Anderson, NRR M. Carbon, Member D. Thacher, NRR C. Siess, Member K. Shaukat,flRR J. C. Mark, Member J. Martin, RES D. Ward, Member B. Sheron, NRR J. Ebersole, Member C. Michelson, Member C. Wylie, Member R. Savio, Staff Highlights
- 1. M. Taylor discussed the NRC Staff's prccress in developing an implementaticn plan for the August 1986 Commission Safety Goal Policy. The August 1986 policy statement (See Attachmert A): (1) established two qualitative goals, (2) established two quantitative objectives, (3) stated a Commission intent to pursue a course of action "that has as its objective providing reasonable assurance, DESIGNATED ORIGINAL B703300262 870121 -
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l SPTC Meeting Minutes January 7, 1987 1 i
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while giving appropriate consideration to the uncertainties in-volved, that a severe core damage accident will not occur at a U.S.
nuclear power plant, and (4) proposed the use of general perfor- l mance guideline specifying that the frequency of a large release be less than 10-6/ reactor year. The policy statement also expresses a Commission make a best effort to ensure that the cuantitative techniques used in decision making take uncertainty into account.
(The features of the policy statement are described in more detail on pages 1-5 of Attachment B and in the policy statenent which is included as Attachment A).
The NRC Staff has developed a proposed implementation plan which they believe incorporates the Commission policy statement guidance.
The main feature of the Staff's implementation are displayed in the
" Integrated Safety Goal Matrix" which is attached as page 6 of Attachment B. The matrix is similar to what was proposed by the NRC Staff in 1986 prior to the adoption of the current Safety Goal Policy. The NRC Staff is proposing that the policy be used for about a year as one of the decision element in the resolution of Generic Issues, and plant specific requirements, the implementa-tion, and the NRC;s Policy Statement on Severe Accidents, in environmental statements, and in allocating NRC resources. The experience gained in the trial application would be used to access the adequacy of the implementation plan. (The NRC Staff subse-ouently made this proposal to the Ccemission on January 8. The Comission appeared to be reluctant to let them start the trial implementation without additional Commission guidance. The Commis-sion is currently developing additional guidance for the Staff and expect to complete this process in 1-2 conths.)
- 2. The proposed Safety Goal Policy implementation plan was discussed at some length. The highlights of this part of the discussion were:
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SPTC Meeting Minutes January 7, 1987 (a) A general performance guideline specifying that the frequency of a large release is used in the Safety Goal Policy Statement and in the proposed implementation plan. The definition of a large release is controversial. The NRC Staff proposal defines a large release a's one which cause one or more prompt fatalities at the site boundary. Commission Asselstine, in the additional views which he attached to the Safety Goal Policy Statement, defines a large release as one which would result in a whole body dose of 5 rem to a individual at the site boundary. The guidelines of 10 CFR 100 have been proposed by others. The NRC Staff proposal may allow releases as large as a SST-1.
(b) The NRC Staff believes that the core nelt frequency / cost-benefit guideline relationship expressed in the " Integrated Safety Goal Matrix" (see page 6, Attachment B) implements the
" reasonable assurance... that a severe core damage accident will not occur" statement in the Safety Goal Policy Statement.
This was also controversial. Opinions were expressed that, given what PRA has told us core melt frequency is true, another core melt in the U.S. would not be unlikely in the lifetime of the currently operating reactors.
(c) The NRC Staff proposal for implementing the Safety Goal Policy Statement will utilize the results of NUREG-1150 in making quantitative measures of risk. Questions were raised as to the adequacy of this approach. The six plants analyzed in NUREG-1150 may well not be very representative of U.S. operat-ing plants.
(d) It was suggested that the NRC Staff review their cost-benefit guidelines and methodology in light of the Chernobyl experi-ence. Loss of unique societial resources are not specifically
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SPTC Meeting Minutes d- January 7, 1987 considered in the NRC cost-benefit methodology. It was suggested that this be reconsidered.
(e) The NRC Staff's " matrix" includes averted on-site costs (AOSC) in'the cost-benefit rule. The subject of including AOSC continues to be controversial.
(f) The NRC Staff, in the proposed implementation guidance, suggested that a factor of 3 for a reduction in core melt frequency and a factor of 10 reduction in the frequency of a large release be considered as " substantial". These factors, by the NRC Staff's arguments, correspond to the difference between "best estimates" and "95% confidence" estimates in the calculations done for NUREG-1150. These factors (3 and 10) would be used as general but not absolute guidelines in judging whether or not a proposed reactor modification was worthwhile. There was some feeling that these factor were too high and would in practice end up being used as thresholds rather than guidelines.
- 3. The NRC Staff has integrated the work being carried out on the steam generator overfill issues under a single Generic Issue (GI-135, " Integrated Steam Generator and Steam Line Overfill Issues). The proposed work plan is summarized on page 7 of Attach-ment B and the proposed Task Action Plan is included as Attachment C. The Staff will be obtaining a contractor for the technical assistance work in the near future. The proposed work will not resolve the issue of steam line water hammer. Preliminary work indicates that this is beyond the scope of the resources allotted to this Generic Issue. The NRC Staff reovested Subcommittee on the adequacy of the proposed Task Action Plan.
.P ed SPTC Meeting Minutes January 7, 1987 4 D. Thacher summarized the progress that the NRC Staff has made on developing a resolution for USI A-17, " Systems Interactions for Nuclear Power Plants." The ACRS last commented on this subject in a May 13, 1986 report, recommending extensive changes in the propose NRC Staff resolution. A summary of the NRC Staff changes is shown an page 8 of Attachment B. The NRC Staff, in general, did not accommodate the ACRS recommendations in their revised, but intends to address the ACRS concerns under a new Generic Issues.
ORNL is currently developing a scope and work plan for this new work.
NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, DC 20001, (202) 347-3700.
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ATTACHMENT A
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[7590-01] j NUCLEAR REGULATORY COMMISSION .
- 10 CFR PART 50 l
Safety Goals for the Operations of Nuclear Power Plants; Policy Statement AGENCY: Nuclear Regulatory Commission.
ACTION: Policy Statement.
SUMMARY
- This policy statement focuses on the risks to the public from nuclear power plant operation. Its objective is to establish goals that broadly define an acceptable level of radiological risk. In developing the policy statement, the NRC sponsored two public workshops during 1981, obtained
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public comments and held four public meetings during 1982, conducted a 2-year evaluation during 1983 to 1985, and received the views of its Advisory Committee on, Reactor Safeguards.
The Commission has established two qualitative safety goals which are supported by two quantitative objectives. These two supporting objectives are based on the principle that nuclear risks should not be o significant addition to other societal risks. The Commission wants to make clear that no death .
attributable to nuclear power plant operation will ever be " acceptable" in the sense that the Commission would regard it as a routine or permissib*ie event.
TheCommijsionisdiscussingacceptablerisks,notacceptabledeaths.
o The cualitative safety coals are as follows:
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Individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health.
-- Societal risks to life and health from nuclear power plant operation should be corrparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risk.
o The following ouantitative objectives are to be used in determining achievement of the above safety goals:
-- The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed.
-- The risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from all other causes.
EFFECTIVEDATE: August 4, 1986. .
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k tions and FOR FURTHER INFORMATION CONTACT: Merrill Taylor, Reg for Operations, Generic Requirements Staff, Offi,ce of the 55. Executive Telephone Director U. S. Nuclear Regulatory Commission, Washington, DC 20S .
i (301/492 4356). .
SUPPLEMENTARY INFORMATION:
t on Safety The following presents the Commission's Final Policy Statemen Goals for the Operation of Nuclear Power Plants:
- 1. INTRODUCTION A. Purpose and Scoce
' Commission on
' In its response' to the recommendations of the President s l tory Commission (NRC) the Accident at Tnree Mile Island, the Nuclear licitRegu policy a statement stated that it was " prepared to move forward d with an expffs in the on safety philosophy and the role of safety-cost tra eo .
.- becisions." This policy statement is the result.
the basic Current regulatory practices are believed blic, is met. toNeverthe-ensure that statutory requirement, adequate protectionbetter of the pu for testing means less,curJentpracticescouldbeimprovedtoprovidea d regulatory requirements.
The the adequacy of and need for current and propose e coherent and Commission believes that such improvement could lead to a
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consistent regulation of nuclear power plants, a more predictable regulatory process, a public understanding of the regulatory criteria that the NRC applies, and public confidence in the safety of operating plants. This state-ment of NRC safety policy expresses the Commission's views on the level of risks to public health and safety that the industry should strive for in its nuclear power plants.
This policy statement focuses on the risks to the public from nuclear power plant operation. These are the risks from release of radioactive materials from the reactor to the environment from normal operations as well as I
from accidents. The Ccemission will refer to these risks as the risks of nuclear power plant operation. The risks from the nuclear fuel cycle are not included in the safety goals.
These fuel cycle risks have been considered in their own right and determined to be quite small. They will continue to receive careful considera-tion. The possible effects of sabotage or diversion of nuclear material are also not presently included in the safety goals. At present there is no basis A.** It is the Comission's on which to provide a measure of risk on these matters.
intention that everything that is needed will be done to keep these types of risks at their present very low level; and it is the Comission's expectation that efforts on this point will continue to be successful. With these ex-ceptions,IitistheCommission'sintentthattherisksfromallthevarious initiating mechanisms be taken into account to the best of the capability of current evaluation techniques. -
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In the evaluation of nuclear power plant operation, the staff considers .
several types of releases. Current NRC practice addresses.the risks to the Before a nuclear power public resulting from operating nuclear power plants. ,
plant is licensed to operate, NRC prepares an environmental impact assessment which includes an evaluation of the radiological impacts of routine operation of the plant and accidents on the population in the region around the plant site.
The assessment undergoes public comment and may be extensively probed in adjudicatory hearings. For all plants licensed to operate, NRC has found that there will be no measurable radiological impact on any member of the public from routine operation of the plant. (
Reference:
NRC staff calculations of radiological impact on humans contained in Final Environmental Statements for '
specific nuclear power plants; e.g., NUREG-0779, NUREG-0812, and NUREG-0854.)
The objective of the Comission's policy statement is to establish goals that broadly define an acceptable level of radiological risk that might be While this imposed on the public as a result of nuclear power plant operation.
policy statenent includes the risks of normal operation, as well as accidents, the Commission believes that because of compliance with Federal Radiation Council (FRC) guidance, (40 CFR 190), and NRC's regulations (10 CFR Part 20 and Appendix I to Part 50), the risks from routine emissions are small compared to the safety goals. Therefore, the Comission believes that these risks need not be routinely analyzed on a case-by-case basis in order to demonstrate confom-ancewith'hesafetygoals.
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w Development of this Statement of_ Safety Policy .
fited In developing the policy staterant, the ComissionNRC-solicited and 4
from the information and suggestions provided il 1-3,by1981 workshop and dis sponsored workshops were held in Palo Alto, California, on Apr 23-24, 1981. The first workshop in Harpers Ferry, West Virginia, on July The second work-addressed general issues involved in developing safety ft goals. goals.
B.oth shop focused on a discussion paper whichdrawn presented from industry,proposed workshops featured discussions among knowledgeable ntedpersons a broad public interest groups, universities, and elsewhere, who represe range of perspectives and disciplines.
for its The NRC Office of Policy Evaluation submitted to the Commis '
Power P1 ants in consideration a Discussion Paper on Safety Goals for Nuclear November 1981 and a . revised safety goal report in July 1982.
tions The Commission also took into consideration the comments licy Statement on " Safety received from the public in response to the proposed Po 1982 (47 FR_ 7023).
Goals for Nuclear Power Plants." published on February d on March 14 17, f Following public coment, a revised Policy Statement was issue f 1983 (48 FR 10772) and a 2-year evaluation period began.
h t resulted l The Commission used the staff report andfinal itsPolicy recommenda from the t
2-year evaluation of saf.ety goals in developing this '
Additionally, the Comission had benefit of further comm ,
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from its Advisory Comittee on Reactor Safeguards (ACRS) and by senior NRC l l
management. .
Based on the results of this information, the Comission has determined that the qualitative safety go'als will remain unchanged from its March 1983 revised policy statement, and the Comission adopts these as its safety goals for the operation of nuclear power plants.
II. QUALITATIVE SAFETY GOALS The Comission has decided to adopt qualitative safety goals that are supported by quantitative health effects objectives for use in the regulatory decisionmaking process.
The Comission's first qualitative safety goal is that the risk from nuclear power plant operation should not be a significant contri-The intent is to butor to a person's . risk of accidental death or injury.
require such a level of safety that individuals living or working near nuclea power plants should be able to go about their daily lives without special Thus, the Comission's concern by virtue of their proximity to these plants.
first safety goal is--
Individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that indivi bear ['no significant additional risk to life and health.
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Even though protection of individual members of the public inherently provides substantial societal protection, the Comission also decided tha 0:
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limit should be placed on the societal risks posed by nuclear power plant ,
operation. The Comission also believes that the risks of n0 clear power plant 6
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operation should be comparable to or less than the risks from other viable means of generating the same quantity of electrical energy. Thus, the Comissica's ,
second safety goal is--
Societal risks to life and health from nuclear power plant operation should be ccmparable to or less than the risks of generating electricity by viable competing technologies and sSculd not be a significant addition ,
to other societal risks.
The broad spectrum of expert opinion on the risks posed by electrical .
generation by coal and the absence of authoritative data make it impractical to calibrate nuclear safety goals by comparing them with' coal risks based on what we know today. However, the Comission has established the quantitative health effects objectives in such a way that nuclear risks are not a significant addition to other societal risks. e Severe core damage accidents can lead to more serious accidents with the potential for life-threatening offsite release of radiation, for evacuation of members of the public, and for contamination of public property. Apart from their health and safety consequences, severe core damage accidents can erede ,
. publiccofifidenceinthesafetyofnuclearpowerandcanleadtofurthe.-
instabilf5yandunpredictabilityfortheindustry. In order to avoid thess -
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- program that has as its objective providing reasonable assurance, whil_e giving s
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'g' t core appropriate consideration to th\ uncertainties involved, that a severe darrge accident wi11 not o: cur at a U.S. nuclear power plants.
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III OUANTITATIhE OBJECTIVES USED TO GAUGE AC OF THE SAFETY GOAL.S A. General Considerations _
The quantitative _ health effects objectives establish NRC guid h ld strive to public protection which nuclear plant designers and operators s ou 4
achieve.
A key element in formulating a qualitctive safety goal whos j tives is to understand ment is measured by quantitative health effects ob ec i by which one judges Ath the strengths and limitat,f ,
ons of the techn qJes
" whether ,the qualitative safety goal has been met.
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andcompleted refinement of ac 2 '. in 01 quantification was taken in the Reactor Safety Study (WASH-i )
1975.
The objectivk of the Study was "to N; try to rea:h some meaning N. !
The Study did not directly clusions about the risk of nuclear accidents."
idents was accept-address the question of what level of risk from nuclear acc x
able. N in Since : the completion of the Reactor Safety Study, further progr l t data have
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developing probabilistic risk assessment and in accumulating i fety re l
led t$p recognition that it is feasible to begin to use quant ;
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h 4 3 111 present in the methods and the. gaps in the data base--essential element ,
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- objectives should be. viewed as aiming points or numerical benchmarks of p In particular, because o' the present limitations in the state of J formance. ,
the art of quantitatively estimating risks, the quantitative health effects objectives are not a substitute for existing re'gulations.
h The Commission recognizes the importance of mitigating the; consequence a core-melt accident and continues to emphasize features such as containmen '
siting in less populated ar as, and emergency planning as integral parts o defense-in-depth concept associated with its accident prevention and mitig l
philosophy.
B. Quantitative Risk Objectives
'V The Commission wants to make clear at the beginning of this section that ble" no death attributable to nuclear power plant operation will ever be "accepta ge. sense ,that the Comission would regard it as a routine In any We are discussing acceptable risks, not acceptable deaths.
event.
fatal accident, a course of conduct posing an acceptable risk at one moment This is true whether one results in an unacceptable death moments later.
Each speaks of driving, swimming, flying or generating electricity from coal.
of these activities poses a calculable risk to society and to individuals.
Some of th se who accept the risk (or are part of a society that ', accep We intend that no such accidents will occur, but the .
do not survive it.
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possibility cannot be entirely eliminated. Furthermore, individual and societal risks from nuclear. power plants are generally estimated to be considerably less than the risk that society is now exposed to from each of the other activities mentioned above.
C. Health Effects - Promet and latent Cancer Mortality Risks The Commission has decided to adopt the following two health effects as the quantitative objectives concerning mortality risks to be used in determining achievement of the qualitative safety goals--
- The risk to an aver' age individual in the vicinity of a nuclear power plant of prompt fatalities that might result fror., eactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are
- generally exposed.
- The risk to the population in the area near a nuclear power plant of a
cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from all other causes.
The Commission believes that this ratio of 0.1 percent appropriately
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reflects both of the qualitative goals--to provide that individuals and society .
bear no significant additional risk. However, this does not necessarily mean e
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's, that an additional risk that exceeds 0.1 percent would by itself constitute a j i
- significant additional risk. The 0.1 percent ratio to other risks is low enough to support an expectation that people living or working near nuclear power plants would have no special concern due to the plant's proximity.
The average individual in the vicinity of the plant is defined as the average individual biologically (in terms of age and other risk factors) and locationally who resides within a mile from the plant site boundary. This means that the average individual is found by accumulating the estimated individual risks and dividing by the number of individuals residing in the vicinity of the plant.
In applying the objective for individual risk of prompt fatality, the Commission has defined the vicinity as the area within 1 mile of the nuclear power plant site boundary, since calculations of the consequences of major reactor accidents suggest that individuals within a mile of the plant site boundary would generally be subject to the greatest risk of prompt death attHbutable to radiological causes. If there are no individuals residing within a mile of the plant boundary, an individual should, for evaluation purpeses, be assumed to reside 1 mile from the site boundary.
In applying the objective for cancer fatalities as a population guideline for individuals in the area near the plant, the Commission has defined the populatic generally considered subject to significant risk as the population .
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t within 10 miles of the plant site. The bulk of significant exposures of the population to radiation would be concentrated within this distance, and thus this is the appropriate population for comparison with cancer fatality risks from all other causes. This objective would ensure that the estimated increase
' in the risk of delayed cancer fatalities from alk potential radiation releases at a typical plant would be no more than a small fraction of the year-to-year normal variation in the expected cancer deaths from nonnuclear causes. More-over, the prompt fatality objective for protecting individuals generally That is, if the provides even greater protection to the population as a whole.
quantitative objective for prompt fatality is met for individuals in the immediate vicinity of the plant, the estimated risk of delayed cancer fatality to persons within 10 miles of the plant and beyond would generally be much lower than the quantitative objective for cancer fatality. Thus, compliance with the prcmpt fatality objective applied to individuals close to the plant would generally mean that the aggregate estimated societal risk would be a number of times lower than it would be if compliance with just the objective The distance for averaging applied to the population as a whole were involved.
the cancer fatality risk was taken as 50 miles in the 1983 policy statement.
The change to 10 miles could be viewed to provide additional protection to
-individuals in the vicinity of the plant, although analyses indicate that this It also pro-objective for cancer fatality will not be the controlling one.
vides more representativ'e societal protection, since the risk to the people
- beyond 10! miles will be less than the risk to the people within 10 miles.
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IV. TREATMENT OF UNCERTAINTIES l l
l The Commission is aware that uncertainties are not caused by use of quanti-tative methodology in decisionmaking but are merely highlighted through use of the quantification process. Confidence in the us2 of probabilistic and risk assessment techniques has steadily improved since the time these were used in the Reactor Safety Study. In fact, through use of quantitative techniques, .
important uncertainties have been and continue to be brought into better focus and may even be reduced compared to those that would remain with a sole reliance on deterministic decisionmaking. To the extent practicable, the Commission intends to ensure that the quantitative techniques used for regulatory decision-making take into account the potential uncertainties that exist so that an estimate can be made on the confidence level to be ascribed to the quantitative results.
s The Cormtission has adopted the use of mean estimates for purpof,es of imple-menting the quantitative objectives of this safety goal policy (i.e., the mortality risk objectives). Use of the mean estimates comports with the customary practices for cost-benefit analyses and it is the correct usage for purposes of the mortality risk comparisons. Use of mean estimates does not however resolve the need to quantify (to the extent reasonatle) and understand those important uncertainties involved in the reactor accident risk predictions.
Anumbehhfuncertainties(e.g., thermal-hydraulicassumptionsandthe phenomenology of core-melt progression, fission product release and transport, and containment loads and performance) arise because of a direct lack of severe - ,
accident experience or knowledge of accident phenomenology along with data related to probability distributions.
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t In such a situation, it is necessary that proper attention but also be give i t only to the range of unce'rtainty surrounding For probabilistic this reason, est ma e to the phenomenology that most influences the uncertainties. t i ties most l l
sensitivity studies should be performed to determine those uncer a nl The results of sensitivity of '
important to the probabilistic estimates. i tion studies should be displayed showing, for example, i the range of var that dominate together with the underlying science or engineering assumpt ons Depending on the decision needs, the probabilistic resu this variation. f deterministic should also be reasonably balanced and supported through use o In this way, judgments can be made by the decisionmak arguments. This is a ,
degree of confidence to be given to these estimates onservatism and assu key part of the process o'f determining the degree of regulatory c This defense-in-depth approach that may be warranted for particular decisions. d safety.
is expected to continue to ensure the protection of public health V. GUIDELINES FOR REGULATORY IMPLEMENTATION l ding use The Commission approves use of the qualitative safety goal, decisionmaking of the quantitative health effects objectives in the regulatory fl The Commission recognizes that the safety goal can provi process. decisions regarding tool by which the adequacy of regulations or regulatory Likewise, the safety goals could be changes to.the regulations can be judged. h existing of benefi{ in the much more difficult task i h pastof andassessing current .
whe plants, designed, constructed and " operated ft goalto comply policy. w t regulations, conform adequately with the intent of the sa e y e
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However, in order to do this, the staff will require specific guidelines to use as a basis for determining whether a level of safety ascribed to a plant is consistent with the safety goal policy. As a separate matter, the Commission intends to review and approve guidance to the staff regarding such determina-tions. It is currently envisioned that this guidance would address matters such as plant performance guidelines, indicators for operational performance, and guidelines for conduct of cost-benefit analyses. This guidance would be derived from additional studies conducted by the staff and resulting in recom-mendations to the Commission. The guidance would be based on the following general performance guideline which is proposed by the Commission for further staff examination--
Consistent with the traditional defense-in-depth approach and the accident mitigation philosophy requiring reliable performance of containment systems, the overall mean frequency of a large release of radioactive .
materials to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactar operation.
- To provide adequate protection of the public health and safety, current NRC regulations require conservatism in design, construction, testing, operation and maintenance of nuclear power plants. A defense-in-depth approach has been mandated in order to prevent accidents from happening and to mitigate their Furthermore, consequences. Siting in less populated areas is emphasized.
emergency Yesponse capabilities are mandated to provide additional defense-in-depth protection to the surrounding population. n e
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These safety goals and these implementation guidelines are not meant ,
as a substitute for NRC's regulations and do not relieve nuclear power !
plant permittees and licensees from complying with regulations. Nor are the safety goals and these implementation guidelines in and of themselves meant to serve as a sole basis for licensing decisions. However, if pursuant to these guidelines, information is developed that is applicable to a particular licensing decision, it may be considered as one factor in the licensing decision. !
The additional views of Comissioner Asselstine and the separate views of Comissioner Bernthal are attached.
Dated at Washington, D.C., this 3 day of h 1986.
. For the Nuclear Regulatory Comission
% w.24 %. ! U Lando W. Zech, Jr.
Chaiman T' .
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edditional Views by Commissioner Asselstine on the Safety Goal Policy Statement The comercial nuclear power industry started rather slowly and cautiously in the early 1960's. By the late 1960's and early 1970's the growth of the New orders were coming in for regulatory industry reached a feverish pace.
The result was the designs of the plants out-review on almost a weekly basis.
A's paced operational experience and the development of safety standards.
experience was gained in operational characteristics and in safety reviews, safety standards were developed or modified with a general trend toward stricter requirements. Thus, in the early 1970's, the industry demanded to In this Safety Goal Policy Statement, the know "how safe is safe enough."
Much credit Comission is reaching a first attempt at answering the question.
should go to Chairman Palladino's efforts over the past 5 years to develop this policy statement. I approve this policy statement but believe it needs to go further. There are four additional aspects which should have been addressed by the policy statement.
Containment Performance I
First, I believe the Comission should have developed a policy on the relative emphasis to be given to accident prevention and accident mitigation.
Such guidance is necessary to ensure that the principle of defense-in-d T The Comission's Advisory Comittee on Reactor Safeguards has maintained. '
As a step in that direction, I repeatedly urged the Comission to do so.
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offered for Comission consideration the following containment performance criterion: -
In order to assure a proper balance between accident prevention and accident mitigation, the mean frequency of containment failure in the event of a severe core damage accident should be less than 1 in 100 severe core damage accidents.
l Since the Chernobyl accident, the nuclear industry has been trying to distance 1t self from the Chernobyl accident on the basis of the expected per-Unfortunately, formance of the containments around the U.S. power reactors.
the industry and the Comission are unwilling to comit to a level of perform-ance for the containments.
The argument has been made that we do not know how to develop contain-ment performance criteria (accident mitigation) because core meltdown phenomena and containment response thereto are very complex and involve substantial uncertainties. On the other hand, to measure how close a plan.t comes to the quantitative guidelines contained in this policy statement and to perform analyses required by the Comission's backfit rule, one must perform just those kinds of analyses. I find these positions inconsistent.
The other argument against a containment performance criterion is that However, a containment such a standard would overspecify the safety goal.
performance objective is an element of ensuring that the principle of defense-in-depth is maintained. Since we cannot rule out core meltdown accidents in the foresjeable future, given the current level of safety, I believe it un' ise not to establish an expectation on the perfomance of the final barrier to a e
s .
substantial release of radioactive materials to the environment, given a core meltdown. .
General performance Guideline !
l While I have previously supported an objective of reducing the risks to an as low as reasonably achievable level, the general performance guideline articu-lated in this policy (i.e., "...the overall mean frequency of a large release of radioactive materials to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation.") is a suitable compromise. I believe it is an objective that is consistent with the recom-mendations of the Commission's chief safety officer and our Director of Research, and past urgings of the Advisory Committee on Reactor Safeguards. Unfortunately, the Commission stopped short of adopting this guideline as a performance objective in the policy statement, but I am encouraged that the Commission is willing at least to examine the possibility of adopting it. Achieving such a standard coupled with the containment performance objective given above would go a long way toward ensuring that the operating reactors successfully complete their useful lives and that the nuclear option remains a viable component of the nation's energy mix.
In addition to preferring adoption of this standard now, I also believe the.Consnission needs to define a "large release" of radioactive materials. I would have defined it as "a release that would result in a whole body dose of S rem to an individual located at the site boundary." This would be consistent * ,
.with the EPA's emergency planning Protective Action Guidelines and with the level proposed by the NRC staff for defining an Extraordinary Nuclear Occurrence
, , , - - - - . , . . - - - , . , . , - ~ - . . - - , . . - , . . . - , . - . ~ . , - _ , . . - , , . . . . --_-,-,...---nn . - - , - -- _ - - - - . - - - , , , - - , - - , ,
s 4-under the price-Anderson act. In adopting such a definition, the Comission ,
would be saying that its objective is to ensure that there is no more than a I
1 in 1,000,000 chance per year that the public would have to be evacuated from the vicinity of a nuclear reactor and that the waiver of defenses provisions of the Price-Anderson Act would be invoked. I believe this to be an appropriate objective in ensuring that there is no undue risk to the public health and safety associated with nuclear power.
Cost-Benefit Analyses I believe it is long over'due for the Commission to decide the appropriate way to conduct cost-benefit analyses. The Comission's own regulatierts require these analyses, which play a substantial role in the decisionmaking on whether to improve safety. Yet, the Commission continues to postpone addressing'this fundamental issue. .
~
Future Reactors In my view, this safety goal policy statement has been developed with a steacy eye on the apparent level of safety already achieved by most of operating reactors. That level has been arrived at by a piecemeal approach to designing, constructing and upgrading of the plants over the years as experi-ence was' gained with the plants and as the results of required research became available. Given the performance pf the current generation of plants, I believe a safety goal for these plants is not good enough for the future. This policy
. statement should have had a separate goal that would require substantially
s - 5- i better plants for the next generation. To argue that the level of safety achieved by plant designs that are over 10 years old is good enough for the next generation is to have little faith in the ingenuity of engineers and in the potential for nuclear technology. I would have required the next generation of plants to be substantially safer than the currently operating plants.
s e
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Separate Views of Commissioner Bernthal on Safety Goals Policy I do not disapprove of what has.been said in this policy statement, but too much remains unsaid. The public is understandably desirous of reassurance since Chernobyl; the NRC staf'f needs clear guidance to carry out its responsi-bilities to assure public health and safety; the nuclear industry needs to plan for the future. All want and deserve to see clear, unambiguous, practical safety objectives that provide the Commission's answer to the question, "How safe is safe enough?" at U.S. nuclear power plants. The question remains unanswered.
It is unrealistic for the Commission to expect that society, for the foreseeable future, will judge nuclear power by the same standard as it does all other risks. The issue today is not so much calculated risk; the issue is public acceptance and, consistent with the intent of Congress, preservation of i the nuclear option.
' In these early decades of nuclear power, TMI-style incidents must be
. rendered so rare that we would expect to recount such an event only to our grandchildren. For today's population of reactors, that implies a probability -
for severe core damage of 10~4 per reactor year; for the longer term, it implies'sjmethingbetter. I see this as a straightforward policy conclusion If the thatever)newspapereditorinthecountryunderstandsonlytoowell. .
Commission fails to set (and realize) this objective, then the nuclear option
.will cease to be credible before the end of the century. In other words, if b
r- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- -
TMI-style events were to occur with 10-15 year regularity, public acceptance of nuclear power would almost certainly fail, h And while the Comission's primary charge is to protect public health and safety, it is al'so the clear intent of Congress that the Comission, if possible, regulate in a way that preserves rather than jeopardizes the nuclear option. -
So, for example, if the Comission were to find 100 percent confidence in some impervious containment design, but ignored what was inside the containment, the primary mandate would be satisfied, but in all likelihood, the second would note Consistent with the Comission's long-standing defense-in-depth philosophy, both core-melt and containment performance criteria should therefore be clearly stated parts of the Comission's safety goals.
.. In short, this pudding lacks a theme. Meaningful assurance to the public; substantive guidance to the NRC staff; the regulatory path to the future for i the industry -- all these should be provided by plainly stating that, consistent l
with the Co.m. ission's " defense-in-depth" philosophy:
- 1) Severe core-damage accidents should not be expected, on average, to occur i
f in the U.S. more than once in 100 years; f) Containment perfomance at nuclear power plants should be such that severe t
accidents with substantial offsite damages are not expected, on average, to occur 1,n the U.S. more than once in 1,000 years; T .
e e
o e m .
- 3) The goal for offsite consequences should be expected to be met after ervative consideration of the uncertainties associated with the estimated frequency of severe core-damage and the estimated mitigation thereof by containment.I The term " substantial offsite damages" would correspond to the Comission's legal definition of " extraordinary nuclear occurrence." " Conservative consideration of associated uncertainties" should offer at least 90 percent confidence (typical good engineering judgment, I would hope) that the offsite release goal is met. .
The broad core-melt and offsite-release goals should be met "for the average power plant"; i.e., for the aggregate of U.S. power plants. The decision to fix or not to fix a specific plant would then depend on achieving
- goal for offsite consequences." As a practical matter, this offsite socie'tal risk objective would (and should) be significantly dependent on site-specific population density.
The absence of such explicit population density considerations in the Comission's 0.1 percent goals for offsite consequences deserves careful thought. Is it reasonable that Zion and Palo Verde, for example, be assigned 1 Interestingly enough, the Comission has adopted proposed goals similar to the above core-melt and containment performance objectives -- without clearJy saying so. Taken together, the Comis 1) 0.1 percent offsi~tepromptfatalitygoals,2) proposedper-reactor-year 10gion's "large -
offsite release" criterion; 3) comitment "to provide reasonable assurance...that a severe core-damage accident will not occur at a U.S. ,
nuclear power plant," though they may be ill-defined, can be read to be
' more stringent than the plainly stated criteria suggested above. .
?.
ATTACHMENT B
..=..
7
)
1 i
ELEM&rrS OF SAFETY GOAL POLICY ESTABLISHED Two OUALITATIVE SAFETY GOALS ESTABLISHED,Two QUANTITATIVE OBJECTIVES FOR GAUGING ,
ACHIEVED OF SAFEW GOALS STATED INTENT TO PURSUE COURSE OF ACTION ON REASONABLE tSSUPANCE THAT SEVERE COPE-DMAGE ACCIDENTS ARE PPEVENTFD-
"WILL NOT OCCUR AT A !!.S. NUCLEAR POWER PLANT."
STATEDTHE HFED TO C0t! SIDER UNCEPTAINTIES IN DECISIONS FOR IfPLEMENTING DEFENSE-IN-DEPm: PROPOSED A GENERAL PERFOPF#'CE GUIDELINE ON THE FREQUENCY OF A LARGE RELEASE
g OUALITATIVE SAFETY GOALS FIRST SAFETY GOAL - INDIVIDUAL MEMBERS OF THE PUBLIC SHCULD BE PROVIDED A LEVEL OF PROTECTION FROM TE CONSEQUENCES OF NUCLEAR PCHER PLANT OPERATION SUCH THAT INDIVIDUALS BEAR NO SIGNIFICAffT ADDITIONAL RISK TO LIFE AND HEALTH.
SECCfD SAFETY GOAL - SOCIETAL RISKS TO LIFE AND HEALTH FROM NUCLEAR POWER PLANT OPEPATION SHOULD BE C0f1PARABLF TO 09 LESS THAM THE RISKS OF GENEPATIllG ELECTRICITY BY VIABLE COMPETING TECHNOLOGIES AND SHOULD NOT BE A SIGNIFICAtiT ADDITION TO OTHEP SOCIETAL P.ISKS.
O OutNTITATIVE HEALTH EFFECTS OBJECTIVES To GAUGE ACHI5VEMEttr OF THE SAFETY GOAL PR@PT FATALITY - THE RISK TO AN AVERAGE INDIVIDUAL IN THE VICINITY OF A NUCLEAR POWER PLANT OF PPCNT FATALITIES THAT MIGHT RESULT FROM REACTOR ACCIDENTS SHOULD NOT EXCEED ONE-TENTH OF ONE PERCENT (0.1%) 0F THE SUM OF PROMPT FATALITY RISKS PISULTS FROM OTHER ACCIDENTS TO WHICH MEMBEPS OF THE U.S. POPULATION ARE GENERALLY EXPOSED.
- !.ATENT CANCER FATALITY - THE RISK TO THE POPULATION IN THE AREA NEAR A NUCLEAR PCWER PLANT OF CANCER FATALITIES THAT MIGHT RESULT FRCH NUCLEAP PWEP PLANT OPERATION FHOULD NOT EXCEED ONE-TENTH OF ONE PERCENT (0.1") CF THE StM OF CAFCER FATALITY RISKS PESULTING FROM ALL OTHER CAUSES.
i 8
. --_- 3
.i REASONABLE ASSitRANCE OBJECTIVE SEVERE CORE-DAMAGE ACCIDENTS CAN LEAD TO MORE SERIOUS (CCIDEWS WITH THE POTENTIAL FOP LIFE-THREATENING OFFSITE RELEASE OF PADIATION, FOR EVALUATION OF MEMBERS OF T E PUBLIC, AND FOP COWAMitETION OF PUBLIC PROPERTY.
APART FROM THEIR HEALTH IND SAFETY CONSEQUENCES, SEVERE CORE DAMAGE ACCIDENTS CAN ERODE PUBLIC C0FFIDENCE IN THE SAFETY OF NUCLEAR POWER AND C/N LEAD TO FURTHER INSTABILITY AND UNPREDICTABILITY FOR THE IPDUSTRY. IN ORDER TO AVOID THESE /JVERSE CONSEQUENCES, THE CrMISSION INTEFOS TO CONTINUE TO PURSUE A , 'e PEGULATORY PROGRAM THAT HAS AS ITS OBJECTIVE PROVIDING REASf*ABLE ASSURANCE, WHILE GIVING APPROPRIATE CONSIDFPATION TO THE UNCERTAINTIES INVOLVED,.THAT A SEVERE COPE-DAMAGE ACCIDENT
- WILL NOT OCCUR AT A U.S. NUCLEAP POWER PLAW.
f
. _ - . . - , , .- - - . . - - - - - , . - . _ _ -..-_.__.,_m , - _ , , . - . , . . ,--.-. _ . -,,.,_ __-, _ , ,,..,,,.._.._,-_.-. ,_,.,,, . . m_. _. - . - _ . - -
s
.....:.____.___. .~
i TPEATMENT OF UNCERTAINTIES TO THE EXTEND PPACTICABLE, THE COPf41SSION INTENDS TO ENSURE THAT THE QUANTITATIVE TECHNIQUES USFD FOR REGULATORY DECISION-MAKING TAKE INTO ACCOUPTT THE POTENTIAL UNCERTAINTIES THAT EXIST SO THAT AN ESTIMATE CAN BE MADE ON THE CONFIDENCE LEVEL TO BE ASCRIBED TO THE QUANTITATIVE RESULTS.
THE Cmi!SSION HAS ADOPTED THE USE OF MEAN ESTIVATES FOR
, . PURPOSES OF. IMPLEMFFING THE QUANTITATIVE OBJECTIVES OF THIS SAFETY GOAL POLICY (I.E., THE MORTALITY RISK OBJECTIVES). USE OF THE MEAN ESTIt%TES COMPORTS WITH THE CUSTOPARY PRACTICES FOR COST-BENEFIT ANALYSES AND IT IS THE CORRECT USAGE FOR PURPOSES OF THE MORTALITY RISK COMPARISONS, l
i r
I I i
, TFTfteATES SAftTY s NG. PATRif '
(sastettlet fee CDeF-8!ELT.
EFFtfTS M etM Ftf-tN57)*
tary Radleactlee teacett Cast lieelth Effects 99.11/RT fil sen/p-r 4 aveeted targe-scale Cere. Delease Frequency
, Promptflatent*** ~ enslte cests)***'
Felt Frequency (Per RT) (Per NTJ'*
d se fer m er,sa mty
< 1e-5 S te . er Itset both sedecesses
> 1 -8, ne.ia effects see's meet e. se,ren (ts.anem-r)
-i. eses -, ne .e.st.ed .
S.1er8 ,er sinet he m ensecteves smyreseits,
< 1o-* -le-5 et -i> es Aest ;
tes't eget ese fuerose (88.ese m-r
> 1o-8. health effects +I895 A8 SCI aselyses Ittely regelred .
fleet hath shjettises' soprove($1.8085-r 1o-3 -led Presemed set to meet health
+IS-)l1885C) effects matti detatted analyses reveel otherwise son't meet one levewee($1.eae#-r eness mo5C) pres-ed set te seet hesith sinet hem shsetttees - smyreve its.sw
>1o-2 ele 05A05C) effects entti detailed .
analyses reveal ettervlee esma t emet ese supe = (e st as 1seet)
- A u values are taken as mean values C* The everall guidellee for the frequency'ef a large and Itfe threatentog release less then Id per Rf meF serve 88 an acceptable sorrogate Ifor health effects analyses) that provides a ht$ degree of asseronte 19et the Canudtstee't safety Goals are achieved. Otherwise. ^ m .. _ ; le atttdent attigetten er preventles may be destreble for added de fense-5.. depth.
- Prompt effects integrated to 1 elle from site benaderyg 1 stent effects to 19 miles.
- 1 AOSC = Averted ensite costs (See IIUPFG/CR-3548, A Itemeest for Vales-luomet Assessuunt]
l.2)1 p-r persen-ree. totegrated to 50 miles ~
s
7 i
d LANL PERFORMS SGTR STUDIES STAFF REVIEWS LANL REVISES AND j FOR BBW, WH & CE PLANTS LANL DRAFT REPORT
- /u/u/H/u/H/H//H/H//// Q ~u/u/u/niu/uu QlSSUES Q y FINAL REPORT l
DEVELOP DEVELOP POSITION STAFF POSITION ON OFFSITE DOSES, LICENSING OUESTIONS ON LICENSEE /EPRI RESPONSES REVIEWS l ON OVERFILL OPER. ACTION, INITIATE REV.
l OVERFILL FROM SGTR ITMil _ Q PROVIDES SGTR ANALYSES Q RESPONSE _ [ T DUE TO SGTR _ /"% TUBE INTEGRITY _ A TO SRP 15 6 3 _ y ,/,7 j unuusu//u/uuassar v///yy///////H///HH/w// y - v 2;gy %,,,r4;gi v sigy J H L DEVELOP POSITION ON STAFF TU8E INTEGRITY INTEGRATE l Gl 66 lACTIONS I
FOR INDUSTRYI LICENSEES REVIEWS INTO G.L. 85-02 A RESPOND _ A RESPONS_E C GI-135 l u uuu///u/ \ _,7yyy, C - g, y ,,,.,, g ASSES RESPONSES AND PREPARE i USS A-3.4.5 SG COMMISSION PAPER
! m TUYE INTEGRITY U niu//u n ,,, \
[ uniuuu/usuuan =
j Gl 67 ISTAFF ACTIONSI 2/87 PUBLIC COMMENTS FINAL RESOLUTION DEVELOP POSITION ON DRAFT NUREG-0844 ON TUBE INTEGRITY suiuuuuu. n/~ ~ .= Q FINAL v NUREG 0844y_O 9 ,, ,
' l RESOLUTH3N INTEGRATE SOME SUS-ISSUES OF GI-67 INTO Gl135 _[T OF GI-135 _
i ' \. /
DEVELOP INTEGRATED W L
{ SGTR STUDY lOFFSITE DOSESI GENERIC SER_ f"gREQUIREMENTSf g cl.135 uinuni.; Ah,ims, = [C,\lOFFSITE
\ DOSESI
,g,g,ggggu, u, -
, ,,,,, v 3,,, -
ISSUE WCAP 11002 TUBE INTEGRITY & STAFF REVIEW INITIATE PLANT OPERATOR ACTION _ WCAP 11002 SPECIFIC ACTIONS
/////s ////u//// "
CREARE/INEL PERFORMS STUDY ON WATER HAMMER & STRUCTURAL INELISSUES STAFF INEL ISSUES STAFF DEVELOP POSITION ON l n ANALYSIS OF MAIN STEAMLINES Q COMMENTS _ Q FINAL REPORT _ Q WATER HAMMER MITIGATION U/uusuiusuuuuuusuuuss= Q DRAFT vuuuuuu/3 REPORT vuuiusur viuuuuuur v 1
I i
i INTEGRATED STEAM GENERATOR ISSUES WORK PLAN
! GI-135 s j .-
1 s.
a4 4 - 4 - L_ _
4 4
ATTACH 4ENT C i
i 1
1 DRAFT TASK ACTION PLAN (January 1987)
INTEGRATED STEAM GENERATOR AND STEAM LINE OVERFILL ISSUES (GENERIC ISSUE 135)
Lead Organization: Division of Safety Reviiw and Oversight (DSRO)
Engineering Issues Branch (EIB)
Task Manager: Syed K. Shaukat, EIB, DSR0 Lead Manager: Robert J. Bosnak, EIB, DSR0 NRR Principal Reviewers: Richard Johnson Engineering Issues Branch, DSR0 E. S. Chelliah Reliability and Risk Assessment Branch, DSR0 E. J. Sullivan Engineering Branch 4 -
Division of PWR Licensing - A C. Y. Cheng Engineering Branch Division of PWR Licensing - B
- Bernard Mann Reactor Systems Branch Division of PWR Licensing - A Chu Liang Reactor Systems Branch Division of PWR Licensing - B Gene Suh Reactor Systems Branch Division of BWR Licensing Applicability: All Reactors Projected Completion Date: June 1988
4 i
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- 1. PROBLEM DESCRIPTION t s
Steam Generator (SG) overfill evints that have occurred in several facilities indicate that the probability of steam generator overfill transients is not small. The> frequency of occurrence and severity of i overfill transients have received staff / industry attention. In addition,
! there have been a lot of overfeed events which can possibly lead to overfill. Therefore there is a need'to investigate these areas. A number of ongoing staff activities are concerned with SG overfill related issue. This issue will coordinate and integrate all these activities. A brief description of each is given in Enclosure 1. The following activities are included:
A. Los Alamos National Laboratory (LANL) Steam Generator Tube Rupture (SGTR) Studies.
' B. Licensing Questions on,SGTR (TMI Review).
C. Generic Issue 66, " Steam Generator Requirements" (Actions for
. Industry). V 1 j D. Generic Issue 67, " Steam Generator Staff Actions".
- E. Westinghouse SGTR Studies.
, F. SteamLineWaterHammerStudies(INEL/CREARE).
2.
PLAN FOR PROBLEM RESOLUTION .
'(
, The overfill programs described in Enclosure 1 are in various stages of resolution. This program will consider the resciution' milestones for each overfill program and provide a schedule for resolution of each in a '
more cohesive manner. An integrated work plan and important' milestone- t schedule is contained in Enclosure 2. Specific tasks for resolution are described in Section 5 under Technical Assistance.
l ,
I 3. BASIS FOR CONTINUED PLANT OPERATION AND LICENSING '~
PENDING COMPLETION OF
,' TASK rimarily The safetytoissue applicable addressed Steam Generators byofthis Task Action pressurized waterPlan reactors(TAP) is p(PWRs) s and the reactor vessel of boiling water reactors (BWRs). Fcr both PWRs and "
BWRs currently licensed for operation, we have concluded that, pending
,s the completion of this TAP, continued operation does not constitute an ?i undue risk to the health and safety of the public for the following g reasons:
I The probability of the design basis accident during normal operation is small and the probability that the accident would occur during 'the short period of time between leak detection and the plant' shutdown is even smaller. ,
A small amount of leakage (e.g., less than the Technical Specification limit) can be tolerated during nonnal operation without exceeding the offsite dosage limits of 10 CFR Part 20.
\ ,
t iQ~s J b 4 Continuous feedback from operating experience and the TAP efforts will be utilized to update interim criteria and requirements.
Following the SGTR event at GINNA in January 1982, the staff proposto to develop steam generator requirements to help reduce tube ruptures. These
,s requirements v,ere issued to licensees through generic letter 85-02 on April 17,
?
1985. This gius further assurance of safety because of the considerations of 4 the following items: I Augmented inservice inspection requirements and preventive tube plugging criteria have been established to provide assurance th6t a great majority of degraded tubes will be identified and removed from service before leaks develop. s Primary to secondary leakage te 3.imit:, and associated surveillance i requirements, have been established.to provide assurance that the occurrence of abnormal tube degradation during operation will be detected
,1 and appropriate corrective action will be taken such that an individual defect will not become unstable under nonnal operating, transient, or accident conditions.
The above mentioned rationale, which constitutes the basis for continued operation of licensed facilf ties, also supports continued licensing of new facilities. Further, to the extent that is practicable, for facilities not yet licensed for operation, " state-of-the-art" design improvements and operating procedures which are expected to decrease the potential for or rate of steam generator tube degradation are required by the staff.
,- 4 NRR. TECHNICAL ORGANIZATIONS INVOLVED This section indicates the responsibilities of each NRR Branch in supporting this task until final disposition of all encompassed issues.
A. Engineering Issues Branch, Division of Safety Review and Oversight, has lead ' responsibility for completion of all issues to their resolution and will provide coordination for scheduling and budgeting activities.
(
' Personpower Estimate: FY-87 0.6 PSY A' y FY-88 0.6 PSY l Reliability and Risk Assessment Branch, Division of Safety Review if.
'( and Oversight will provide review of proposed implementation and assist in resolution of coments.
Personpower Estimate: FY-87 0.1 PSY FY-88 0.1 PSY
.\
4 C. Engineering Branch, Division of PWR Licensing-A, will provide review of proposed implementation and assist in resolution of coments.
Personpower Estinate: FY-87 0.2 PSY ' -
FY-88 0.2 PSY D. Engineering Branch, Division of PWR Licensing-B, will provide review of proposed implementation and assist in resolution of coments.
Personpower Estimate: FY-87 0.2 PSY l FY-88 0.2 PSY E. Reactor Systems Branch, Division of PWR Licensing-A, will provide review of proposea implementation and assist in resolution of coments.
Personpower Estimate: FY-87 0.2 PSY FY-88 0.2 PSY F. Reactor Systems Branch, Division of PWR Licensing-B, will provide reivew of proposed implementation and assist in resolution of coments .
Personpower Estimate: FY-87 0.2 PSY FY-88 0.2 PSY r
G. Reactor; Systems Branch, Division of BWR Licensing, will provide review of proposed implementation and assist in resolution of comments.
Personpower T-tinate: FY-87 0.2 PSY FY-88 0.2 PSY l 5. TECHNICAL ASSISTANCE A. Contractor: To Be Detennined l Furis Rec"iired: $90K FY-87
$90K FY-88 The Technical Assistance Program consists of the following tasks:
Task 1, To assess the impact of water carry over or water in the main steam line following the SG overfill and find ways to mitigate the consequences or develop conclusions on whether or not these consequences of SG overfill are acceptable.
g a,. _--- - - - - - - - - - --
.I .
- S-Task 2. Study the effect of main steam line overfill considering weight of water, the sagging of main steam line and consequences on the attachments, valves, and branch lines. -
- Task 3. Review current ASME Code requirements with respect to licensing needs for eddy current inservice inspection of steam generator tubes. Formulate recommendations for improving code provisions for eddy current testing.
Task 4. Develop regulatory guidance such as SRP revisions, based on the results of the above tasks and conclusions of other activities in this integrated program.
- 6. INTERACTIONS WITH OUTSIDE ORGANIZATIONS The primary purpose for interactions with the following organizations is to exchange information on the experience and research work they are sponsoring:
EPRI, Plant Owners Group, etc.
Interactions with other organizations such as the Electric Power Research Institute (EPRI) and the "ad hoc" organization of plant owners may also be appropriate regarding the safe operation of steam generators or reactor vessels in general and, in particular, the various safety problems associated with the degradation of steem generators or, reactor vessels.
6*
L RESOURCE REQUIREMENTS
SUMMARY
FY-87 FY-88 Contract Collars for technical assistance in thousands 90 90 NRR Personpower in person years (PSY)
EIB,DSR0 0.6 0.6 SORR, DSR0 0.1 0.1 PAEB 0.2 0.2 PARS 0.2 0.2 PBEB 0.2 0.2 PBRS 0.2 0.2
_ . . . . . BWRS 0.2 0.2 TOTAL 1.7 1.7
- 9
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ENCLOSURE 1 A. LANL SGTR Studies SGTR transients have received staff / industry attention because of their frequency of occurrence and severity During some of these incidents it was several hours before the primary-to-secondary leakage was terminated.
Some existing Safety Analysis Reports state that leakage can be stopped within 30 minutes and that overfilling of tube-ruptured steam generator does not occur. Studies have shown that operator action times that have been observed are longer than those assumed in analysis of design-basis SGTRs. This caused reinvestigation of SGTR transients by vendors and by NRC. LANL performed studies and analyses for SGTR events for three typical PWR plants using various SG overfill transients, multiple failure SGTR and control system failures. The results of these analyses have been submitted to NRC in a draft report in September 1985. l The findings in this report are plant specific, but the operator action times were found to be an important parameter for all three plants. It was found that overfill is not a problem if the operators acted in less time than that specified in plant procedures. But the primary-to-secondary a SGTR leakage than 30 minutes caused by(approx.) incident as stated in theisFSAR.
predicted to continue To help much longer mitigate the SGTR, certain recomendations for operator actions were made as follows:
- 1. Reduce hot-leg temperature to 555 K (540*F) using secondary Atmospheric Relief Valves (ARVs).
- 2. After the hot-leg temperature has been reduced to 555 K (540*F),
isolate the Damaged Steam Generator (DSG).
l
!- 3 Prevent primary system overfill by controlling the High Pressure
! Injection (HPI) on pressurizer level increases beyond some predetennined level, e.g. , 2.54 m (100in).
l
- 4. Continue Generators primary)
(ISGs . system Maintaincooldown water levelbyinusing theby the ISG Intact usingSteam the Auxiliary Feedwater (AFW).
- 5. Reduce primary-system pressure by using the Auxiliary Pressurizer Spray (APS) or the Power-operated Relief Valve (PORV) but limit Subcooling Margin (SCM) to a minimum of 13.9 K (25'F).
- 6. Continue cooldown and depressurization down to the Residual Heat Removal System (RHRS) entry conditions.
The above steps were collected from several SGTR guidelines.
~ The draft report from LANL has been reviewed by the staff. Coments have been provided to LANL to finalize the report, and issue the final version.
\
B. LicensingQuestionsonSGTR(TMIReview},
During t'1e review process of TMI-1 Overfill due to SGTR the staff requested additional infonnation on SGTR plant specific guidelines.
Among many questions asked there were some questions on tube integrity and radiological consequences, asking the utility to provide tube rupture analyses for design basis SGTR accidents with and without offsite power.
The utility contracted with EPRI to perform analyses and provide responses to the fiRC questions.
EPRI's report NSAC-101 which incluoes the results of SGTR analyses was
, received by the licensee on October 19, 1986. Licensee responded to NUREG 0737, I.C.1 (Emergency Operating Procedures) SER questions on November 18, 1986. Staff is currently reviewing the licensee response.
C. Generic Issue 66, " Steam Generator Requirements" '~
(Actions for Industry)
The staff recomended actions for industry were issued through a generic letter 85-02 on April 17, 1985. These actions have been found to be effective measures on a plant specific basis for significantly reducing (1) the incidence of tube degradation, (2) the frequency of tube ruptures and corresponding potential for significant non-core melt release, ar.d (3) occupational exposures, and are consistent with good operating and engineering practices.
The responses from licensees have been reviewed. The staff was asked by the Comission to prepare an information paper providing staff's i asse'ssment of the reponses and follow-up efforts.
The staff is in the process of revising the draft NUREG-0844 which is expected to be issued in January 1987. The issuance of NUREG-0844 constitutes the resolution of Generic Issue 66, after which the staff intends to develop position on tube integrity.
D. Generic Issue 67, " Steam Generator Staff Actions" _ __. __
To meet the objective of reducing steam generator tube ruptures the staff identified and recomended the following fifteen staff actions:
- 1. Reassessment of Radiological Consequences;
- 2. Reevaluation of SGTR Design Basis;
- 3. Supplemental Tube Inspections;
- 4. Integrity of Steam Generator Tube Sleeves;
- 5. Denting Criteria;
- 7. Steam Generator Overfill;
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- 8. Pressurized Thennal Shock;
- 9. Improved Accident Monitoring;
- 10. Reactor Vessel Inventory Measurement;
- 11. RCP Trip; .
- 12. Control Room Design Review;
- 13. Emergency Operating Procedures;
- 14. Organizational Responses; and
- 15. Reactor Coolant System Pressure Control.
Except for Eddy Current Tests (ECT), all the other staff actions were considered either part of on-goine activities or issues of low priority.
The need and scope of work for ECY is being reevaluated because this is a focal point having a potential reduction in Occupational Radiation Exposure (ORE).
Item 2 above, although rated as low priority, has been a concern expressed by staff and any need for further work will be evaluated after
'LANL, WESTINGHOUSE, and EPRI submit their reports on SGTR analyses.
Item 7 above is divided into: 1) Steam Generator Overfill due to Control System Failures (consi-dered under USI A-47).
- 2) Steam Generator Overfill due to SGTR(willbeconsideredinthis integrated program when position on
' overfill due to SGTR is developed).
E. WEST.INGHOUSE SGTR Studies -
' The frequency of occurrence and severity of SGTR transients, particularly the SGTR event of GINNA (NUREG-0909 and NUREG-0916) prompted Westinghouse to initiate a study on structural integrity of Steam Generator Tubes and .
operator action times to terminate the primary-to-secondary leakage in such events. A W/ CAP Report 10698 and Supplement 1 on offsite doses with some open items (like structural tube integrity) has been issued. The staff has reviewed this report and issued an SER for offsite doses calculations. Another report W/ CAP 11002 containing overfill transients,
- operator action times and structural integrity of tubes has been issued subsequently and is under review by the staff. The SER on this report is expected to be issued by May 1987.
F.
Steam Line Wate Hamer Sf.udies (INEL/CREARE) -- '-
One of the' categories of USI A-47 transients involves the overfilling of a steam generator in PWRs or the vessel in BWRs. It is postulated that the overfill transient could lead to steam line failure caused by dynamic loading of a moving slug of water or condensation resulting in water hammer. One of the tasks under USI A-47 was to assess the potential i
. \
severity of water hammer events in main steam lines during postulated steam generator overfill transients. This task was carried out by INEL.
INEL completed and submitted a draft report on this task in November 1985. It is entitled "The Potential Severity of Water Hammer In PWR Main Steam Lines During Postulated Steam Generator Overfilling Events."
The staff reviewed and provided comments on this draft report. The final report was published in January 1986.
Steam line water hammer study is one of the tasks developed through thermal-hydraulic analyses studied under USI A-47 (Control System Failures). The intent of this investigation was to determine the bounding water ham er scenario and the loads thereby imposed upon the piping system. It was found that the calculation of overpressure and ,
impulse loading cannot be done with simple thermal-hydraulic correlations. A method was introduced that facilitates the determination of the estimated overpressure as a function of both subcooling and void fraction. The overpressures were determined and presented for a range of subcoolings, and mitigating factors were discussed which could possibly reduce the overpressures and impulse loadings.
Since the thermal-hydraulic analyses could not show that a main steam line water hammer would be of little or no consequence, structural analyses were performed to determine whether or not a typical main steam line could withstand the postulated water hammer loads. Again, the intent was to perform a bounding structural analysis, taking into consideration the plasticity of the pipe. The results of this analysis show that the piping system will fail in bending if subjected to the postulated water hammer loads. Real life experience does not show such failure. Therefore, it is logical to assume that there are conservatisms and uncertainties in the assumptions made in the analyses.
Sensitivity studies were then made to learn more about these conservatisms and uncertainties. The effects of variation in subcooling temperatures, slug length and support locations were considered.
The conclusions were that:
l a) We weren't able to predict water hammer occurrence; .
b We couldn't establish parameter thresholds; l c We couldn't demonstrate that pipe will not fail; and d It would take major (Multi-million Dollar) experimental program to resolve water hammer.
l I
, ., \
- o. .
Therefore, the line of action now will be to assess the impact of the situation that can occur in the main steam line following the SG overfill and find the ways to mitigate the consequences or develop conclusions on -
whether or not these consequences of SG overfill are acceptable, using
- bes'. estimate methodology.
e
b
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EHCLOSURE. 2 e
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~~
LANL PE'R'F'O'RMESGTR 5TdlilES SfAFF REVIEWS LANL REVISES AND FOR EtsW. WH b CE PLANTS
,O//HHHHHH//uMH/HH/s r_ Qsuu/uuuu/uu= Q LANL DRAFT REe* ORT QlSSUES Q FINAL REPORT I i . i :l= l DEVELOP DEVEL6P POSITION
._ _ _ _ . . _ STAFF POSITION ON OFHITE DOSES. .
LICENSING QUESTIONS ON LICENSEE /EPRI RESPONSES REVIEWS j ON OVERFILL OPER. ACTION. INITIATE REV.
OVERFILL FROM SGTR ITMil _ Q PROVIDES SGTR ANALYSES ATUBE INTEGRITY _ A TO SRP 15 6 3 _ Q,/,7 ON/H/HHHuunu/Hur V///////////H////HH/HHi' Q) RESPONSE Q f T' DUE y 2igyTO SGTR ' \ ,,,Iajgy
., i i i y sigy ' %,, j DEVELOP POSITION ON 5TNFF TURE INTEGRITY INTEGRATE GI 66 (ACTIONS _ _._ _
INTO FOR INDUSTRY) LICENSEES REVIEWS GI 135
, G L. 85 02 G RESPOND A RESPONSE l /H Hu/uHu// h unne \ Cjyyy,f g C ASSES RESPONSES OSI'~ -315'SG" AND PREPARE
/N COMMISSION PAPER O,TUIE INTEGRITY,,,,,,,,,,,,,, . _ . _
u n u n nistu u u u u -
" C)G167 ISTAFF ACTIONSI . _ . __
2/87 PUBLIC COMMENTS "
FINAL RESOLUTION DEVELOP POSITION ON TUBE INTECPITY ON DRAFT NUREG unusuususuuun y 0844 ,% FINAL _ [C NUREG 0844 _ -
COMPLETE l' RESOLUTION INTEGRATE SOME SUR ISSUIS OF Gl 67 INTO GI-135 . [' T OF GI 135 _
'%/
DEVELOP .. INTEGR ATE d uL W PERFORMS ISSUE WCAP 10698 ON ISSUE SER ON WCAP 10698 ISSUE PLANT SPEC. INTO SGTR STuOY unuiu,=(C, y
"% IOFFSITE DOSES)
,, p\ ,,,,,,,,,,,,,,,,,,,,,:
lOFFSITE DOSESI -- GENERIC SER AREQUIREMENTd. f'ag
,,,,, = \ ,,,,,
Gl.135 ISSUE WCAP 11002 TUDE INTEGRITY b STAFF REVIEW ' INITIATE PLAfJT OPERATOR ACTION WCAP 11002 SPECIFIC ACTIONS nuinunuiu 9 4
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CREARE/INEL PERFdRi455TUOFON _ . . .. .. . . _ . . . . .
WATER HAMMER b STRUCTURAL INELISSUES___ STAFF INEL ISSUES STAFF DEVELOP POSITION ON ANALYSIS OF MAIN STEAMLINES FINAL REPORT WATER flAMMER MITIGATION u/uuuusunnunuusuius- Q DRAFT REPORTv _////uu,, Q COMMENTS _ C ,,,,,,,,,,,,,,_,/"% @
y suuuu,uur INTEGRATED STEAM GENERATOR ISSUES WORK Pi N 1 G!-135 i
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