ML20245B827
| ML20245B827 | |
| Person / Time | |
|---|---|
| Issue date: | 04/10/1987 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20244D847 | List:
|
| References | |
| FOIA-88-198 NUDOCS 8904260302 | |
| Download: ML20245B827 (61) | |
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April 10, 1987 MEMORANDUM TO: Eric Beckjord, Director Office of Nuclear Regulatory Research FROM: Harold R. Denton, Director Office of Nuclear. Reactor Regulation
SUBJECT:
PROPOSALS FOR ENHANCING BWR MARK I CONTAINMENTS With the NRC reorganization effective on April 12 the responsibility for managing the generic issues of severe accidents and any specific proposals for BWR Mark I containments passes to you in th6 Office of Nuclear Regulatory Research (RES). Enclosed is the final draft of the options paper we have -
prepared for the Commission for your possible use; two of the principal managers involved in its preparation, Dr. R. W. Houston and L. G. Hulman are transferring to RES as well.
We have beert concerned about the performance of Mark I containments for some time and have tried to develop an effective way to improve them 50 that, '
should a core melt accident occur, there would be reasonable assurance that its consequences would be substantially mitigated. We have developed the approach spelled out in the enclosed draft to achieve that purpose. The proposed modifications are of reasonable cost and are justified herein by safety arguments and by backfit analysis. In a~ddition, with the shif t to rulemaking as the preferred approach, we believe that OGC concurrence is obtainable.
In the closing days of our preparations we took the additional step of meeting with key scientists from the RES program (on February 3, 1987) and from the industry program (on March 27,1987). In these meetings we reviewed fif teen questions related to the Mark I issue in order to establish the efficacy and justification for different strategies of improvement. The range of technical opinion expressed in the two meetings was wide. Da the one hand the industry spokesmen indicated a fairly substantial confidence in Mark I performance, and perhaps most notable, a confidence that hot core debris would not melt through the drywell shell. In the meeting with RES scientists, on the other hand, drywell meltthrough was viewed as a virtual certainty, and little hope was held out for any means of improving containment performance. In view of this broad ;
difference in views we are reluctant to submit our proposals to the Commission i at this time. We are more inclined to accept the more pessimistic views of the l RES scientists, but we believe that the combined remedies proposed in the enclosure would be effective in giving reasonable assurance of small releases from a BWR/ Mark I plant even in the event of a core melt. I recommend that you allocate the necessary resources to evaluate the efficacy of these containment
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8904260302 890419 PDR FOIA JOHNSON 88-198 PDR ,
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i.;;rovement proposals as soo possible so that-the Individua, Plant Examination (IPE) program car, e adjusted accordingly.-
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Harold R. Denton, Director l Office of Nuclear Reactor Regulation l l
Enclosure:
As stated ,
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t FOR: The Commissioners FROM: Victor Stello, Jr.
Executive Director for Operations
SUBJECT:
PROPOSED SEVERE ACCIDENT MITIGATION REQUIREMENTS FOR THE GE MARX I CONTAINMENTS j PURPOSE: To respond to (1) the Comission concerns raised during November 3,1986 Comission briefing, (2) a memorandum (M861103) dated November 17, 1986 from Samuel J. Chilk to Victor Stello, Jr., (3) a memorandum, dated December 19, 1986, from Chairman Lando W. Zech. Jr. to Victor Stello, Jr., and (4) a memorandum dated February 9,1987, from Samuel J. Chilk to Victor Stello, Jr. supplementing tne Chairman's memorandum dated December 19, 1986.
BACKGROUND: On June 16, 1986, the staff met with the industry representatives and proposed a five element program to -
enhance the BWR containment severe accident mit'gation
. capability. The objective of the proposed program is to gain a reasonable assurance that BWR Mark I containments have substantial capability to mitigate the consequences of severe accidents. The BWR owner's group voluntarily had the Industry Degraded Core Rulemaking (IDCOR) group to evaluate the staff proposals. The industry report was issued in August 1986. Parallel to IDCOR generic evaluations, the Vennont Yankee Power Corporation and the i Boston Edison Company pursued plant specific containment enhancement programs at Vermont Yankee Nuclear Power Station and Pilgrim Nuclear Station respectively. The BWR owner's group effort was subsequently transferred to Nuclear Utilities Management and Resources Comittee (NUMARC) for generic application to all light water reactors, while the plant specific activities of Vermont Yankee and Pilgrim are continuing.
On November 3,1986, the staff briefed the Commission on proposed Merk I containment generic requirements. During that briefing, the staff was requested to orovide for ACRS and CRGR reviews of the staff proposals (ating into consideration other staff programs, inc1L:!! 1g NUREG-1150),
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.4 provide that information to the Commission for consideration, and consult with the Corrission prior to going ahead with a generic letter..
In a memorandum (M861103) from Secretary Samuel J. Chilk to Victor Stello, Jr., the staff was requested to provide a rationale for not following rulemaking and to provide core nelt frequency estimates for the plants identified in the staff's November 3,1986 Connission briefing slides.
In a memorandum frem Chairman Lando W. Zech, Jr. to vi>, Ar Stello, Jr., dated December 19, 1986, suppiamented by a memorandum, 6atec February 9,1987, from Samuel J. Chilk to Victor Stello, Jr., the staff was requested to consider pros and cons of a dual approach consisting of 1) a package !
supporting a generic letter for these issues where the staff is confident of a significant reduction in risk, and 2) a rulemaking for all BWR Mark I containment safety ._ j measures, including those contained in the generic letter.
The staff was requested to consider the following
- additional items in its response:
Resolving the que: tion of BWR - Mark I containment
" fixes" within the confines of rulemaking.
-
- Folding the issue of Mark I containment " fixes" within the Severe Accident Policy implementation program where containment vulnerabilities are systematically addressed.
Issuing a proposed generic letter addressing the BWR Mark I containment issues (generic letter to be forwarded to the Commission for teview).
DISCUSSION: Following the November 3,1986, staff briefing to the Commission on the staff proposals for Kark I containments severe accident mitigation requirements which lend themselves to generic implementation, the staff was requested to consider the following items:
provide for ACRS and CRGR reviews;
' account for other staff programs including NUREG-115C; provide core melt frequency estinates; consider pros and cons of (1) adcressing in a generic letter those issues for which the staff is confident J
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of a sigt:ificant reduction of accident consequences and (2) addressing all Mark I " fixes" within the confines of rulemaking; i
fold the issues of Mark I containtnent " fixes' into W the Severe Accident Policy implementation program; i issue a proposed rule; and issue a proposed generic' letter addressing all Mark I
" fixes" and forward it to the Comission for review.
Following is the staff response:
ACRS Review The staff presented the proposed BWR Mark I containment requirements to the ACRS Subcommittee on Containment ._
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Requirements on December 9,1986, and to the full Comittee on December 12, 1986. Subsequently a letter dated December 17, 1986 (y, the ACRS Enclosure 1) toChairman Chairman wrote I.ando W. Zech, Jr. sumarizing its views - that (1) the staff should issue the proposed generic letter for public coments, and (2) the results of industry studies should be considered along with public coments prior to final resolution of the issues. In addition, the ACRS requested additional infomation in specific technical areas of concern. The staff has included information in its regulatory analyses (Enclosure 4) supporting the proposed
" fixes" ~and has addressed the ACRS request for specific technical information in Enclosure 1. The staff will respond to the ACRS questions in a separate action.
CRGR Review On December 22, 1986, the staff briefed the CRGR on the Mark I containment improvement proposals (see transcript of CRGR meeting No. 104 minutes).
NUREG-1150 Results In February, 1987 the draft of NUREG-1150 was issued for 3 comment. The NUREG-1150 results show a core damage l frequency for a BWR Mark I reference plant (Peach Bottom) to be 8 x 10~6/RY. When compared to core damage frequency estimates in Enclosure 4 (Table 1) for BWR Mark I plants, the NUREG-1150 core dcmage frequency value does not appear to be representative of BWR plants with Mark I cor.tainments.
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The Reactor Safety Study (1975) results indicated a conditional containment failure probability (CCFP) for a reference BWR plant with Mark I containment (Peach Bettom) was about 90%. In the November 1984, IDCOR Technical Sumary Report, Nuclear Power Plant Pesponse to Severe )
Accidents, the CCFP for Peach Bottom containment was. I inoicated~as 20%. More recently (September-1986), the Vemont Yankee Containment Study provided an estimate that CCFP for Vemont Yankee containment was about 71. The NUREG-1150 results for BWR Mark I reference plant (Peach Bottom) show that containment performance is extremely poor (CCFP = 0.1 - 0.9), and appears to be driven upward by the dominant station blackout sequences. -(See regulatory analysis in Enclosure 4.) Our review of the NUREG-1150 results indicates that large source of uncertainties exists in predicting post-accident containment response and determining containment failures which could result in large early offsite releases. The results of NUREG-1150 -
provide additional justification- for staff proposals for BWR containment " fixes", starting with Mark I containments which show the greatest vulnerability to failures resulting in large offsite releases.
Data of Core Melt Frequencies The core melt frecuency estimates have been included in the i staff's regulatory analyses in Enclosure 4.
I Pros and Cons of Generic 1.etter and Rulemaking The staff's proposal to pursue implementation via a generic letter was based on the judgment that the generic letter approach might be more timely than rulemaking, and that the required improvements could be achieved with n.;nimum costs to NRC and the industry.
However, past experience has shown that the generic l
1etters are most cost-and-time-effective if the staff proposals have a nexus to the existico regulations.
Rulemaking, on the other hand, is substantially more time-and-effort-intensive, but has the benefit of more effective
- enforcement of implementation. During earlier discussions with the industry, it was the staff's impression that the industry was supportive of the proposed
- The rule requirements, being the Commission requirements, would be more solemn and therefore more effective.
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" fixes", and would welcome the generic letter approach, since it would minimize implementation costs. Subsequent course of. events has shown that the industry interest in an early resolution has substantially diminished and several industry representatives may be unwilling to support the staff proposals to the extent that would make the generic l 1etter approach successful. In view of this change in the i industry climate and in the absence of a nexus between the staff proposals and existing regulations, and considering the fact that a rule is more effective to implement, the staff has come to the conclusion that rulemaking is the preferred course of action.
, Separating the proposed fixes into those that can be pursued via a generic letter, and those for which rulemaking is preferred, will,have the disadvantages of both approaches. The staff proposal for Mark I " fixes" consists of an integrated approach to enhancing the
- containments' ability to mitigate the consequences of -
severs accidents at successive stages of the course of a
- severo accident. Only when all proposed " fixes" are implemented, can the stated objectives be met.
Based on the above considerations, the staff believes that an integrated rulemaking approach to containment enhancements is the most cost-effective approach.
The staff has developed a draft of a proposed rule and a
. proposed generic letter, which are enclosed for Comission review (Enclosures 2 and 3). The supporting regulatory analyses are presented in Enclosure 4 In rulemaking approach rule based on Enclosure,the staff will 2 infomation and develop su a proposed infomation in generic letter (Enclosure 3) andpplemented regulatory by analysis (Enclosure 4). Requisite information in Enclosures 3 and 4 could be incorporated as a statement of considerations.
Folding Mark I Containment " fixes" Into the Severe Accident Policy Implementation Program The Commf:sion's severe accident policy for operating reactors provided the following guidance:
" Operating nuclear power plants require no further ,
regulatory action---unless" significant new safety i information arises-- ". ---careful assessment shall l l
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be made---whether vulnerability is plant or site specific or of generic importance." "The most cost-effective options for reducing this vulnerability shall be identified and decisions shall be reached consistent with the cost-effectiveness criteria of the Comission's backfit policy as to which option or set of options (if any) are justifiable and required to be implemented."
The analyses presented in NUREG-1150 have identified no new vulnerabilities for BWR containments. The vulnerabilities are generic to containment types and confirm previous conclusions relative to Mark I containments. The staff's regulatory analyses in Enclosure 4 show that generic
" fixes" can be justified consistent with the cost-effectiveness criteria of the Commission's backfit policy. _
In the regulatory analyses (Enclosure 4) the staff has concluded that, because the BWR containments lend themselves to a set of generic " fixes", searching for these " fixes" in Individual Plant Evaluations (IPEs) would be unnecessary. It is recognized that, in time, IPEs performed as originally conceived may identify risk outliers (e.g., plant specific station blackout vulnerabilities) that could be " fixed" to lower the likelihood of severe accidents, thereby also lowering the risk Specifically NUREG-1150 results show that BWR core damage frequencies primarily arise'from station blackout (SBO) sequences *. It is thought that reducing the SB0 sequences' likelihood by generic " fixes" or under IPE program could decrease BWR core damage frequencies to a point where BWRs would not be risk outliers. Without prejudging for such future findings, the staff believes that the uncertainties on containment conditional failure probabilities will still remain unacceptably high, and defense-in-depth capability of containments can not be assured even after the 580 sequence probabilities have been reduced. The staff believes that the uncertainties l inherent in risk reduction efforts are best dealt with when tempered by reasonable engineering and policy
- For example, the SB0 sequences in BWRs could result in a loss of automatic depressurization syster (ADS) leading to l potential high pressure core-velt ejection to containment. j Energetically released core material would disperse !
throughout the containment, directly heat the containment walls, and result in almost certain containment ,
failure (see Enclosure 4 for more details). )
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i judgement regarding the essential purpose of containment-systems. The staff, therefore, concluded that the proposed generic requirements need not await IPEs of containment performance.
Proposed Rule _,
1 The staff considered two options for rulemaking. Option I would be addition of general design criteria-to existing criteria addressing performance requirements for containments under cevere accident conditions. Option 2 would be a separate rule similar. to the Comission's ATWS Rule._ The Option 1.was determined to be inappropriate because 10 CFR 50, Appendix A, which specifies General Design Criteria, deals with postulated design basis accidents, single failures, arid anticipated operational occurrences. Because containment performance requirements
. are being considered for accidents beyond the design basis ~
of the plants, a separate rule similar to the ATWS rule (10
. CFR 50.62) would be more appropriate. Enclosure 2 is a conceptual draft of such a proposed rule.
Proposed Generic Letters Enclosure 3 is a proposed generic letter which has.been revised since the Comission briefing of November 3,1986.
The changes consist of additional " fixes" based on the staff meetings with the research community ir.volved in the severe accident evaluations on February 3,1987, and the NUMRC Comittee on Containment Study on March 27, 1987.-
In view of significant differences between the research comunity and the industry, the staff has incorporated only those changes which are considered to be worthwhile.
In the generic letter approach, pursuant to 10 CFR 50.54(f),
. the licensees would be required to furnish, by a specific date, proposed schedules for completing each of the require-ments identified in the generic letter. Upon receipt of acceptable schedules, the staff would issue Orders, confirming the agreed schedules.
RECOMMENDATIONS:
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The staff recommends that the Comission approve the rulemaking approach to the proposed Mark I containment improvements. The staff will, upon Comission approval, expeditiously develop a proposed rule for publication in the Federal Register. The proposed rule would be based on the approach in Enclosure 2 and would be supplemented by i
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8' l information-in the generic letter (Enclosure 3)'and the regulatory. analysis (Enclosure 4). The supplementary information could be in the form of a statement of considerations. ,
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I Victor Stello, Jr. f Executive Director for Operations '
Enclosures:
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- 1. Letter from ACRS to Chairman Lando W. Zech, Jr. and the Staff {'
Response to ACRS Questions
- 2. Rulemaking Approach -
- 3. Sample Generic Letter
. 4 Regulatory Analysis ._
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Enclosuro 1
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^ ADVISORY COMMITTEE ON RE ACTOR sAFEGUARC
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December 17, 1986 The Honorable Lando W. Zech, Jr. j j
Chairman U. S. Nuclear Regulatory Comission Washington, DC 20555
Dear Chairman Zech:
SUBJECT:
ACRS COMMENTS ON PROPOSED BWR MARK I CONTAINMENT REQUIREMENTS FOR SEVERE ACCIDENTS During our 320th meeting, Decerber 11-13, 1986, the members of the ,
Advisory Cecrnittee on Reactor Safeguards discussed proposed new require-ments for Boiling Water Reactor (BWR) Mark I containments with regard to their capability to withstand severe accidents. During our review, we had the benefit of discussions with representatives of the NRC Staff and i BWR Owners' Group (BWROG), as well as th'e benefit of the documents referenced. Discussions were also held with the Staff during our 315th ,
meeting, July 10-12, 1986, and information was developed during the Containment Requirements Subcommittee meetings on September 23 and December 9,1986.
The Staff's proposed requirements for BWR Mark I containments were l presented in a draft generic letter with an attachment evaluating i containment performance during severe accidents. This evaluation is l admittedly open to quotion ir regard to conditional containment failure l
prot, abilities, and it lacks detailed technical . justification for the se.
1ection of certain procedures or parameters, e.g., a reduction of 905 in the drywell spray flow rates,. We recomend that the Staff strengthen, to the extent feasible, the detailed technical analysis to support the proposed requirements.
We also believe that additional information is needed in several areas, j
including the following:
i e an estimate of the contribution to core melt frequency and to l containment failure from significant external events, including l' seismic events greater than the Safe Shutdown Earthquake ,
e a tabulation of information concerning significant differences among the family of Mark I containments sufficient to ascertain l that the proposed generic requirements would not necessitate l special exceptions and/or additions l i
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The Henorable Lando W. Zech, Jr. December 17, 1986 I
e an examination of possible adverse effects of pool bypass es a
' result of transient thermal stresses and their possible effect on s drywell connections to downcomers
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e an analysis of the proposed Emergency procedure Guidelines (for example, circumstances under which emergency sprays may be initi-i sted or reinitiated) which considers the effects of venting and spraying operations on the continuing availability of the vapor suppression processes and other engineered safety features which may continue to be needed or highly desirable Until sufficient information has been developed on matters such as these, we are unwilling to agree with the proposed position in the draft generic letter: "Given the implementation of the generic improvements of Mark I containments, there is no need for an Individual Plant Eval-
, uation (!p[) for containment performance." -
Nevertheless, the Staff should issue the proposed generic letter for public coment. The BWROG containment studies and the Staff's evalua-tion of the Emergency procedure Guidelines are expected to be issued during the proposed public comens period. The results of these studies should be considered along with public ,cocinents as part of the final-resolution of this issue. We would like' to consider this matter at a future meeting when the actions noted above are completed.
Sincerely, I
.j David A. Ward Chaiman
References:
1., Memorandum from R. Bernero for R. Fraley,
Subject:
proposed Generic Letter on Improvements for BWR Mark 1 Containments, dated December 3, 1986.
- 2. SWR Owners' Group, " Emergency Procedure Guidelines," OEI Document 8390-4, Draf t Revision 4AF, dated August 14, 1986.
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t ENCLOSURE I THE STAFF RESPONSE TO ACRS REOUEST FOR ADDITIONAL INFORMATION Recuest An estimate of the contribution to core melt frequency and to containment failure from significant external events, including seismic events greater than the Safe Shutdown Earthquake.
Response
The external events which are contributors to core melt frequencies have significantly lower probability of occurrence than the internal event probabilities used in the staff's regulatory analyses. SECY 86-162 describes '~
the staff's plans for dealing with external events in context of severe accident policy.
Request A tabulation of information concerning significant differences among the family of Mark I containments sufficient to ascertain that the proposed generic requirements would not necessitate special exceptions and/or additions.
Response
See Table 2 of Enclosure 4.
Recuest An examination of possible adverse effects of pool bypass as a result of transient thermal stresses and their possible effect on drywell connections to downcomers.
Response
The staff has long recognized the potential for overpressurization of the containment if a wetwell-to-drywell vacuum breaker were to stick open during a loss of coolant accident. As a result, the staff has imposed stringent operability requirements on these vacuum breakers. The issue and its resolution are discussed in NUREG-0474, dated July 1978. The specific requirements are stated in Appendix A. " Steam Bypass for Kark I, II and III Containments," to Standard Review Plan (SRP) 6.2.1.1.C, " Pressure-Suppression Type BWR Containments." In general, the SRP states that the plant not be operated unless the vacuum breakers (VBs) are indicated in their closed position, that VB valve and limit switch operability be demonstrated monthly, j and leak tightness of the VBs be demonstrated at refueling outages. These i I
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2 actions are to be assured by inclusion in the Technical Specifications governing plant peration. The staff additionally has taken the position that either VBs be redundant with single closed position indication, or their closed position indication be redundant. Compliance with the staff position as currently stated in SRP 6.2.1.1.C reduces the credible open failure of a VB between the wetwell and the drywell to a passive failure (i.e., disk or body rupture).
Recuest An analysis of the proposed Emergency Procedure Guidelines (for example, circumstances under which emergency sprays may be initiated or reinitiated) which considers the effects of venting and spraying operations en the continuing availability of the vapor suppression processes and other engineered safety features khich may continue to be needed or highly desirable. ,
Respons_e _
The BWR Owner's Group is evaluating the impact of the proposed " fixes" on the Emergency Procedures Guidelines (EPGs). The staff will review the Owner's Group proposals to ascertain the effect of venting cnd spraying on the continuing availability of the vapor suppression processes.
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o Enclosure 2 CONCEPTUAL RUELMAKING APPROACH A. PREAMBLE SEVERE ACCIDENT PERFORMANCE CRITERIA Severe accidents beyond the design bases of light water reactor facilities should be realistically appraised to ensure that both of the following severe accident performance criteria are satisfied.
(a) Severe Core Damage Frecuency - The likelihood of events causing severe
- camage to the reactor core should be very low. -
(b) Mitication of Releases - Containment and other features of the plant shall be such that, given the occurrence of one of the spectrum of the more likely severe accidents, there is reasonable assurance that it will not result in the release of a large amount of radioactive materials from the plant.
The requirements demonstrating compliance with criterion (a) should be developed in the future regulations.
The requirements for enhancing containment performance to meet criterion (b) should be developed for each containment type for both boiling water reactors and pressurized water reactors.
For boiling water reactors, criterion (b) should be applied to Mark I, Mark !!, and Mark 111 containments taking advantage of the salient features of each containment type. The rule proposed here deals with Mark I containments only. Future rules will address the requirements for Mark II and Mark III containments.
B. PROPOSE 0 RULE 50.64 SEVERE ACCIDENT REQUIREMENTS FOP MARK I CONTAINMENTS (a) All licensees with boiling water reactors and Mark I containments shall develop procedures considering those actions necessary in a severe accident to mitigate their consequences. Modifications of existing procedures will assure that the actions in severe accidents are not pre-empted by other procedures. ,
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(b) Each boiling water reactor with a Mark I containment shall perform plant specific analyses to the extent that may be necessary to identify modifications to existing plant systems and/or emergency operating procedures to achieve the following performance objectives:
(1) Hydrogen Control - Provide assurance that deinerting of an initially inerted contaiment cannot occur following a station blackout event.
(ii) Drywell Spray - Provide assurance that backup water supplies, pumping capability, and spray features are available during a station blackout event that would provide a well distributed spray pattern in the drywell and at a total flow rate sufficient to .
(iii) Pressure Relief - Provide assurance of the capability to open and reclose, during a station blackout event, a vent path from the wetwell vapor space. The vent line should be capable of .
handling a water vapor flow equivalent to li decay heat at the
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highest pressure selected to initiate venting.
(iv) Core Debris Coolability - Provide assurance that torus room will be able to hold all torus water in the event of torus failure caused by core debris attack. Additionally, fire suppression system sprays should be (a) capable of scrubbing airborne radioactivity in the torus room, and (b) pH control should be provided in the torus room to prevent reevolution of radiciodines in the water retained in the torus room. Finally, torus room ventilation space fire protection spray enhancements should be provided if plant specific cost-benefit analyses support such improvements.
(c) The equipment changes required herein shall be installed curing the first refueling outage which begins nine (9) months after the effective date of this letter. The procedures and training required shall be implemented on a schedule reviewed and approved by the NRC.
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l Enclosure 3 TO ALL BOILING WATER REACTOR (BWR) LICENSEFS AND APPLICANTS FCP POILING WATER REACTOR LICENSES Gentlemen:
SUBJECT:
PROPOSED BWR SEVERE ACCIDENT CONTAINMENT REQUIREMENTS (GENERIC LETTER 87- )
Severe accidents dominate the risk to the public associated with nuclear power j plant accidents. A fundamental objective of the Somission's Severe Accident !
Policy is to take all reasonable steps to reduce the chances of occurrence of I a severe accident and to mitigate the consequences of such an accident should -
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- one occur. The Reactor Safety Study report issued in 1975 found that for most* l BWRs the probabilities of accidents resulting in core melt were low, but the containment performance following a severe accident was poor. Subsequent actions resulting from the TMI Action Plan have led to several plant modifica-tions and required improvements in plant procedures to further reduce the like-lihood of severe accidents. In December 1980, an industry initiative on severe ,
accidents resulted in the formation of the Industry Degraded Core Rulemaking 1 (IDCOR) group to address the concerns related to core damaging accidents. The IDCOR effort has . led to industry methodology for Individual Plant Evaluations (IPEs) to searen for the risk outliers and to address system reliability and containment performance on a plant specific basis. The staff has concluded, bewever, that for BWR containments, a set of g(neric requirements has been idatified that moots the need to await plant specific analyses of containment performance and will lead to speedier implementation than would be possible via the IPEs. Severe accident analyses have indicated several areas for improvement in BWR containments which should be promptly pursued. This Generic Letter deals with Mark I containments only. Subsequent Generic Letters will address Mark II c.id Mark III containments. Following areas of improvements or adequacy of present requirements are identified for Mark I containments.
- 1. Hydrocen Control Present inerting requirements imposed by 10 CFR Part 50.44 and the technical specifications are adequate. The licensees should, however, assure that deinerting(of sequences suchthe as inerted containments will not occur in severe accident station blackout).
- 2. Containment Spray All BWRs with Mark I containment shall previde at least two backup water supply systems for the containment drywell spray, ore of which shall be functional ciuring station blackout. Water to the spray system from the:.e
- Those BWRs for which core melt frequencies have been calculated.
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1 backup supplies shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during a severe accident scenario.
In addition, the spray nozzles shall be adjusted so that an evenly distributed spray pattern will be developed in the drywell whether water is supplied by the primary source or either of the backup sources. A ficw rate on the order of 1/10 of the present flow rate is considered typical, the licensee shall select the flow based on an analysis of plant specific parameters.
- 3. Pressure Relief The licensee shall select a pressure between design pressure and li times design pressure at which to open an exhaust path from the wetwell vapor space. This line should be capable of handlitig water vapor flow ecuivalent to 1% decay heat at the vent pressure selected without significant chance of rupture before the desired release point. The line shall be equipped ~
with isolation valves which can be opened and reclosed by remote manual operation er by simple procedures which can be implemented during severe accident scenarios including station blackout. '
- 4. Core Debris Control The licensee shall ensure that the water in t'h'e suppression pool in the event of torus failure due to core debris attack is held within the confines of the torus room and the corner rooms and cannot flow cut to other parts of the plant. Further, fire suppression systems should be provided in the ventilation spaces above the torus room that could become flow path for airborne radioactivity that would be released due to core debris attack and failure of the downcomers or torus. Lastly, stainless steel wire baskets of trisodium phospnate (TSP) should be provided in the torus room to assure that in the event of the torus failure, the water pH in the torus room will be such as to minimize the likelihood of the re-evolution of iodine.
- 5. Procedures and Training The licensee shall implement emergency operating procedures and other '
procedures based on all significant elements appropriate to its plant of Emergency Procedure Guidelines, Revision 4.
- 6. ADS Enhancements The staff has determined that uncertainties about continued operability of the Automatic Depressurization System (ADS), during protracted station blackout events, could lead to core trelt ejection from the vessel at high reactor pressure and alm i t certain containment failure. Because there 1
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1.0 BACKGROUND
1 A fundamental objective of the Comission's Severe Accident Policy of Augut 8, 1985 is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to assure substantial capability to mitigate the
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consequences of such an accident should one occur. The commission also called for a balancing of accident prevention and mitigation, and special consideration of containment performance in searches for risk outliers.
Enhancements to the performance of containments in severe accidents should increase assurance of mitigation of severe accident consequences as required by the Comission fundamental policy of defense-in-depth. The Reactor Safety Study report issu'ed in 1975 found that for most* BWRs the probabilities of accidents resulting in core melt were low, but the containment following a severe accident could be severely challenged and tended to offset the benefits-of low BWR core melt probabilities. Subsequent actions resulting from the TMI Action Plan have led to several plant modifications and required improvements in plant procedures to further reduce the likelihood of severe accidents. Other post TMI actions have also involved containment enhancements, particularly in the areas of isolation dependability and hydrogen control.
In concert with the Comission's policy to further reduce the chances of occurrence of severe accidents and to mitigate their consequences, an industry initiative is underway to develop a methodology for Individual Plant Evaluation (IPE) directed to search for risk outliers. The resulting approach will be
- Those BWRs for which core melt frequencies have been calculated.
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1 2 i applied on a plant-specific basis. The initial IPE trials have been made by industry, and have encompassed internal accident initiators and systems reliabilities to the point of estimating core melt probabilities. Source terms, containment performance and offsite risks have not been considered, but have been discussed as future extensions of initial IPEs.
With respect to BWRs with Mark I type containments, the staff has reviewed these initial IPEs, historical probabilistic risk assessments and the plans for completing the search for individual plant outliers. The review has indicated that sufficient bases exist to enhance accident consequence mitigation by backfitting in five areas of containment performance improvements as discussed in the subsequent sections. That is, by requiring improvements in five areas,
, no further evaluations of accident mitigation for BWR Mark I type containments --
are considered necessary.
The staff identified containment enhancements lend themselves to generic imple-mentation and have the potential to significantly reduce the consequences of several severe accident sequences including station blackout and some ATW5 se-quences. In the Policy Statement the Commission stated that the rulemaking route would generally be a preferred route to implement future severe accident related actions. However, rulemaking is extremely time consuming. The Commis-sion's statement regarding operating reactors recognized the time element and the continued severe accident risk to public health and safety, and provided other options to dispose of the issues through the conventional practice of issuing Bulletins and Orders or Generic Letters.
The design and sizing of containment are required to assure that the containments are essentially leak-tight barriers against the uncontrolled release of radio-activity to the environments and to assure that containment design conditions important to safety are not exceeded for as long as
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3 postulated accident conditions require. The containments should accommodate with sufficient margins, the pressures and temperatures resulting from any loss-of-coolant-accident (LOCA).
Although a postulated design basis LOCA is not expected to produce more than a few percent fuel f ailures, an accident radiological " source term" used in calculating offsite dose consequences is representative of a substantial core melt accident (10 CFR 100). Even for this source term, containments are designed such that calculated offsite doses are unlikely to result in an early or major latent health hazard if the containments were to maintain their low leakage ~apability". .
What is at issue is the capability of containments to perform a mitigating --
safety function as long as practicable during very low probability severe accidents, where the stress on containment may significantly exceed that of a design basis LOCA and the consequences of containment failure may be very significarst. The structural integrity of BWR containments is seriously challenged for accidents with high energy release to the containment because in spite of the positive pressure suppression feature, they are relatively small and the likelihood of their failure in a severe accident is perceived to be higher than necessary.
Overall plant core melt probabilities for BWRs with Mark I, II and III contain-ments have been estimated to range from one in a thousand per reactor year to two in ten'million per reactor year for BWR designs evaluated by the NRC and the industry. Many of these estimates have not fully included assessments of l the benefits of post-TMI backfits, operator respenses, or the increases in core melt probabilities arising from factors not considered in plant specific analyses such as earthquakes, floods and fires. Contemporary analyses break down such probabilities into classes and subclasses of accidents. The sum of "Part 100 specifies "the expected demonstrable leak rate from the containment",
a value which is made part of each licensee's Technical Specifications.
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i the core melt probabilities for all classes and subclasses of accidents is considered to be the overall core melt probability. For BWRs with Mark I containments, 10COR2,2 has proposed the following five classes of events for core melt accidents: )
loss of core cooling with containment at low pressure and failure after core merit;
- loss of core cooling with containment failure befc e core melt;
- loss of core cooling with containment failure soon after core melt due to high ontainment pressure at the time of core melt;
- loss of core cooling with containment failure before core melt due to failure to depressurize; and -
- containment bypas2. , _ ,
Our review of the core melt probability estimates to date generally indicates that they are low. The BWR core melt frequencies of past evaluations are summarized in Table I. Given a core melt, the estimates of likelihood of Mark I, II and III containment failures have been high relative to other containment types. In all of these past evaluations, little or no credit has been given to containment features which can be used with relatively modest upgrading to prevent or mitigate accidents.
The Reactor Safety Study (WASH-1400, NUREG-75/14) indicates a conditional containment failure probability for the BWR Mark I containment reference plant (Peach Bottom) of about 90% (inferred from Table 5-3, page 81). That is, given a core melt in a BWR with a Mark I containment (Peach Bottom) there is about 90% chance of containment failure. In the November 1984, IDCOR Technical Sumrtary Report, Nuclear Power Plant Response to Severe Accidents, the estimate
'or Peach Bottom was about 20% (inferred from Table 10-1, page 10-6). More 1 Individual Plant Evaluation - Peach Bottom Atomic Station, May 1986.
2IDCOR Technical Report 85-3-B1, BWR Accident Sequence - Individual Plant Evaluation Methodology, April 1986.
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recently, the Vermont Yankee Containment Study provided an estimate that vermont Yankee, a somewhat smaller BWR with Mark I containment, has a condi-tional containment failure probability of about 7%. In all chose Mark I containment failure estimates the challenges come from a spectrum of accidents
-including ATWS, station blackout, and other transients. The principal causes of failure are overpressure and direct attack of the drywell. In February, 1987 the draft of NUREG-1150 was issued for comments. The NUREG-1150 results show that for BWR Mark I reference plant (Peach Bottom) the containment performance is extremely poor (CCFP = 0.1 - 0.9), and appears to be driven upward by the dominant station blackout sequence. Figures 1 and 2 show an early CCFP distribution for Peach Bottom presented in Draft NUREG-1150. The dominance of station blackout sequences appears to drive the containment failure frequencies upward.
The accidents of interest span a spectrum of sequences and will have a probability' distribution unique to each plant. Nevertheless, because of the uncertainties in calculating the dominant accident sequences, it is prudent to consider each principal type as the cause of large scale core melt and containment challenge.
For a BWR Mark II containment (Limerick), Brookhaven National Laboratory (BNL)
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estimated about 50% likelihood of containment failure leading to a large release, givea a core melt (inferred from BNL 33835; April 1984). Only IDCOR and GESSAR II evaluations considered containment venting.
For most accidents considered, the core is postulated to melt, interact with steam, water, and the structural features in the vessel and coolant system, seit through the vessel, and attack the concrete and structural features of the lower containment. Depending on the sequence of events, the containment has the potential to fail either before or af ter vessel melt through. For the remainder of the accidents postulated, the containment would be bypassed, allowing radioactivity a direct path to portions of a plant not designed to l contain the releases, but with some capability to attenuate radioactivity. l I
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BWR containments respond to heatup of the fuel in the vessel directly or indirectly. The direct transfer of energy is through pipe breaks, through blowdown into the suppression pool or by the aerosols generated when the core melts through the vessel. Indirectly, radiant heat is transferred through the vessel and piping. The blowdown or depressurization process, and '.5e use of the relatively large quantity of suppression pool water as a heat sink and fission product scrubbing device, act in combination with the structural capability of the containment (including penetrations) to mitigate the high temperatures, pressures and subsequent radioactivity releases. Core melt scenarios have been identified, however, which can produce conditions that could lead to containment failures, and release 9f fission products to the environment without the benefit of the suppression pool scrubbing. There is
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strong evidence that BWR containments are capable of withstanding substantially higher stresses than those for which they have been explicitly designed and this potential containment strength can be drawn upon to demonstrate additional protection to the public at low to modest cost. The longer a containment can be expected to hold, the greater the likelihood that failure can be avoided. If failure were to occur, however, reductions in the radioactivity released would be achieved due to delay in failure. Actions that can be taken to prevent a j catastrophic failure of containment before the fission products are adequately attenuated include such items as operator actions to vent the wetwell space l above the suppression pool, and providing reliable spray capability.
l In a core melt accident with temperatures in excess of 5000 degrees F, fission products are released from the fuel in three general groups. The noble gases and the more volatile species of fission products are released from the fuel relatively early in a core melt accident. Later, the less volatile species are released as the fuel melts down into the vessel and combines with the in-vessel structural materials. Finally, af ter melting through the vessel, refractory materials may be released during interactions of core debris with concrete on the floor of the containment.
The amount of radioactivity that could be released to the environment in core melt or degraded core accidents has been the subject of considerable analysis l
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i for a number of years. Present estimates (NUREG-0956) for MK I and III BWRs j indicate that substantial quantities of important fission products can be {
released in a core melt accident and these analyses provide clues that suggest that releases can be reduced by a number of actions to enhance .ontainment i performance.
Within the core of a contemporary BWR with MK I, II, or III containment at full power there are over five billion curies of radioactivity. Severe accident releases to the environment have been estimated to exceed 40% of such important fission products as iodine and cesium (releases of over 300 million curies of iodine and over seven million curies of cesium fer a 3454 MWt reactor).
2.0 NEEDS AND STRATEGY FOR CONTAINMENT IMPROVEMENT Consideration of the insights dnwn from previous analyses suggests that no single simple feature can be added to a BWR pressure suppression containment to provide substantial assurance that it will successfully sitigate the consequences of a large scale core' melt should one occur. Rather, one must conceive of some integral approach which deals with the principal concerns.
Consider new only the Mark I containments, 24 of which are now found in licensed U.S. reactors. This analysis and development of requirements will deal first with Mark I containments because they constitute about 2/3 of the BWR population. Subsequent analyses will deal with Mark II and III containments.
Compared to many other U.S. reactor containments, the Mark I containment j
(Figure 3) has a small volume relative to the size of the reactor it contains, l
With a free volume of less than 300,000 cubic feet, the drywell wall is very i I
close to the reactor and to the lower head area where melten core material would most lika:1y fall from the reactor vessel. Even with a relatively higt design pressure, typically about 50-60 psig, the small volume makes the Mark I f containment more vulne'ible to overpressure failure, given a comparable core 1 melt event. Any strategy to enhance Mark I containment performance must certainly consider preventing hydrogen combustion, cooling the non-condensibles l
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in containment, and as a last resort, venting serious overpressure through some 3 I
available path, whGre the consequence of venting is known and is preferred to the potential of uncontrolled release due to containment loss. The enhancement strategy must also take account of any variations in systems and components characteristics of the operating U. S. Mark I containments (Table 2).
Should molten core material (corium) reach the drywell floor, the direct attack I of the drywell becomes a serious concern. Since the corium is likely to be of low viscosity, it can easily reach the nearby wall of the drywell. There it will attack the steel wall of the drywell between the vents or attack one of the large steel vent passages leading down to the wetwell. The steel shell of the drywell is typically backed by a 1-2 inch construction gap filled with a _
plastic spacer and then by a very thick, reinforced concrete biological shield.
Most analyses do not attempt to treat attack through the shell and shield mechanistically because of the complexity of the path, but it is apparent that this path to the reactor building and the ambient is not an open one, especially if some means are available to reduce the vigor of the attack by the hot corium or mitigate the consequences of such an event.
The presence of hot corium on the floor of the drywell raises the potential of other challenges. The corium can be expected to react with the concrete floor, thus generating large quantities of non-condensible gas (including hydrogen) as well as voluminous aerosols carrying non-volatile health threatening radionu-clides such as lanthanum and plutonium. In addition, radiative and convective heat transfer can directly attack the steel shell and its penetrations. Any strategy to enhance the performance of Mark I containments must seek some means to cool or quench the core debris. That strategy should also include means to cool the drywell wall to prevent overheating.
Overpressure failure of the Mark I containment may be averted even in a large scale fuel melt if debris cooling and quenching limit the amount and the temperature of the non-condensible gases in containment. Nevertheless, it is possible that pressure and temperature can build up to levels which could cause containment penetration (seal) failures or catastrophic rupture of the
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containment. It is desirable to have available a procedure of last resort whereby the threatening overpressure can be relieved from the wetwell vapor space so that all gases released from containment will have passed through the water in the suppression pool, and thus will have been scrubbed of most non gaseous fission products. The pressure at which such relief should be taken into account must account for the ultimate strength of containment, the reliability of the valves used for venting and backpressure effects on SRV operation. In addition, consideration should also be given to the material vented from the containment. At a minimum it will contain water vapor, nitrogen, unburned hydrogen, and (depending on the stage of the accident) fission product noble gases (principally Xe 133). If the path through the downcomers and the pool water is bypassed, perhaps through a vacuum breaker line, the effluent could contain other fission products as well. .-
The core debris of concern includes not only the corium which melts through the pressure vessel, but the large amount of aerosols which may be released and captured by the water in the suppression pool. Using Three Mile Island experience as a guide, the suppression pool water might absorb radioactive material on the order of 0.01 to 0.1 Ci/ml (0.4-4 gigaBq/el). That water, almost 1 million gallons of it, would be so radioactive that it would be desirable to see to it that it stays in the torus, or at least in the torus room and immediately adjacent spaces should the torus fail. If some molten corium does pass down through one or more of the eight drywell-to-torus vent ducts, then it would most likely cause torus rupture. In that event the water, if retained, would be available to quench the corium. The fission products released to the torus room atmosphere in such a scenario would still be substantial, however, and would include the radioiodines re-evolved from water.
The fission products would be driven through existing ventilation spaces connected to the torus room to the environment. The use of trisodium phosphate (TSP) in wire baskets in the torus room would prevent the reevolution of radioiodine by assuring a suitable pH level in the torus water. Lastly, the use of conventional fire suppression systems (sprays) in the ventilation spaces above the torus room is necessary to substantially reduce airborne fission products that could escape to the environment.
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Finally, it is evident from all previous studies that the Mark I containment should not be treated as a simple passive body. To be effective in mitigating core melt accidents, its features must be used by trained on-site personnel who are prepared to deal with these extreme events using the equipment at hand.
Thus, the containment improvement strategy should include procedures and training for such accident management.
3.0 MARK I MITIGATION FEATURES Af ter considering the technical factors identified in the preceding, a 5 element strategy stands out as an effective solution to improve Mark I containment performance to the point that there would be reasonable assurance
. that Mark I containments can substantially mitigate the consequences of a ~
large-scale core melt accident. The five elements are:
- 1. Hydrogen Control i
- 3. Pressure Relief
- 4. Core Debris Control
- 5. Procedures and Training 3.1 Hydrocen Control Under the present requirements of 10 CFR Part 50.44, all Mark I containments are required to have their containments inerted (with nitrogen gas) during operation. Allowance is made for a period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the beginning and at the end of the power operations cycle to operate with air in the containment to enable operators to inspect equipment in the containment for leaks etc. From time to time small, unidentified system leaks will start inside the containment during power operation. Although operators sometimes i
% ; ~ ;. -c.: x c8.m .u u.mm:. <.:w. . 1.w ..~ - wu w.u . ... w: . au w 11 question whether the 10 CFR 50.44 based technical specifications permit deinerting and entry during a cycle to investigate such leakage, they have used the 20-hour deinerted periods permitted in the technical specifications to investigate and to the extent practicable repair such leaks. Data recently presented in the Vermont Yankee Containment Study indicate that non-inerted operation at that plant amounted to 1.1% of power operations in a period of 14 years. Such a low fraction indicates a very low risk from hydrogen even now as long as inerting system power scurces operate as intended The impact of the station blackout sequences for inerting system operation must be considered in any evaluation of hydrogen, control for inerted containments. Taking Vermont Yankee's experience as representative, the Mark i strategy here should reaffirm the existing controls for hydrogen.
3.2 Containment Spray All Mark I containments except Oyster Creek and Nine Mile Point 1, are equipped with a dual header drywell spray system. The two spray headers are rings located well up in the cylindrical part of the drywell with branches holding spray nozzles pointing down at an angle. The headers are fed through each division of the RHR system with spray operation as an alternate mode of RHR operation. Due to the characteristically large size of RHR pumps (3,000 10,000 gpm) the drywell spray has a very high flow rate. Precautions are usually included in operating procedures to avoid excessive use of this powerful spray system. Oy, ster Creek and Nine Mile Point I have separate dedicated spray systems. See Table 2 for a summary of key features of the 24 plants with Mark I containments.
Most plants have other systems already connected to the spray header feed lines outside of containment. They include such systems as RHR Service Water, Condensate, and in some cases bolted blind flanges which are removed to install lines for periodic containment integrated leak rate tests. Thus, it is easy for a plant to provide one or more backup supplies for the drywell spray, even in the event of a station blackout, because of the availability of fire main systems with independent pumping capability. But the available backups are all
"i m/ b ' "LJU.K JiA.&,:.C & MW.; M..M.e", n .,n e.;Ch nO h.b. W7 ".' ~ ,1 :.h6 T'T "Y T 12 smallejrsystems,ontheordero 10%d the size of the RHR. If they were used they would probably not be able to develop sufficient header pressure for even spray flow distribution in the drywell. If there is a high assurance of drywell spray during severe accidents, even in station blackouts, a number of benefits accrue. First, the walls and penetrations of the drywell are cooled to reduce the threat of heat induced failure. Second, the drywell floor is I kept flooded to provide a quenching pool for molten corium if it melts through the reactor vessel. Third, the continuing spray cools any corium which begins to travel over the open floor toward the wall of the drywell or its vents. I Fourth, the spray is available to begin washout of aerosol particles even before they pass to the suppression pool; this is, another filtering and condensing mechanism which will reinforce defense-in-depth if some flow were to
, bypass the suppression pool. ~
Thus, the Mark I strategy is to replace all sprr.y nozzles with smaller sizes (about a tenfold reduction) and to provide at least two backup water supply systems (including one for station blackout) which can be turned on by remote manual operation or by simple procedures for connection and startup.
3.3 Pressure Relief Currently available structural analyses for Mark I containments show ultimate failure at about twice the design pressure, usually failing at the knuckle between the upper cylindrical and lower sections of the drywell. However, these analyses have not taken into account the mechanical backing which may be provided by the biological shield surrounding the drywell. The ultimate strength of the Mark I may be quite a bit higher.
Other factors may control the selection of a pressure limit. The vent valves already on containment are tested or qualified to levels about enual to design pressure and may not be reliable at pressure far above that. In addition, such high back pressures would reclose SRVs, possibly' exacerbating the accident.
I
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O 13 The size and the durability of the vent path involves questions of accident scenario. Assuming that pressure considerations lead one to select a relief at a level on the order of design pressure, some alternatives become apparent.
First, this vent need not be the large steam escape path desired for an ATWS scenario; for the ATWS the operator would use one or more main steam lines to the turbine bypass. With ATWS set aside, only a decay heat level vapor flow vent capacity is needed to prevent containment failure due to the pressure rise. Since the containments are already designed to absorb initial sensible heat and the high, early decay heat within the design pressure, a flow equivalent to 1% of rated power at the venting pressure is tufficient. One percent power is equivalent to the decay heat generation rate after 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. For Mark I containment, this translates to six to eight inches vent .-
diameter, based on venting at about 60 psig. Figure (4) was obtained from the August 1986 IDCOR study submitted to BWROG, and gives the estimated vent diameter as a function of power level.
The figure can be used to determine venting capacity for ATWS. For a 2800 MWe reactor ATWS power levels can range from 30 to 40% of or B40 to 1120 MWt. A vent diameter of 40 inches or so may be needed to manage ATWS by venting the containment.
One has the choice cf designing a special purpose vent for this purpose leading directly to the stack or to use existing vent valves and ducts. The staff knows of no plant which already has a high pressure (60-90 psig) vent path in place. Given the highly undesirable effects of the potential vent path !
rupturing inside the plant, the Mark I containment venting strategy is to provide a burst-resistant path with reliable valves, capable of remote manual opening and reclosing even in station blackout, to vent steam equal to 1% of rated power to the plant stack or a high point vent. The use of stack or other I high point release will assure a substantial reduction of radiation doses due l to post-accident venting. Figure (5) shows expected whole body dose as a function of distance for unmitigated release due to direct breach of the containment following a core melt. Also shown are the whole body dose as a function of distance for ground level and elevated release of noble gases for
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doses for elevated releases.
3.4 Core Debris control The strategy identified in sections 3.2 and 3.3 should provide cooling cf molten corium should it come out of the reactor vessel. If molten corium does reach the drywell wall, the combination of a spray-cooled interior and a heavily backed extarior make drywell to reactor building failure unlikely. The vents on the other hand are a remaining possible debris travel path and the torus room is pneumatically open to the reactor building. Therefore, the Mark I strategy is to ensure that, if the torus fails, the water in the torus will ~~
be retained in the torus room and the corner rooms, and will quench any corium which might reach there and Ifmit the spread of damage by intensely radioactive material. In addition, the fire suppression sprays should be provided 'sn the ventilation spaces to scrub airborne radioactivity, and trisodium phor.phate baskets be placed in the torus room to assure pH control to prevent reevolution of radiciodines from torus water.
- 3.5 Procedures and Training The Emergency Procedure Guidelines, Rev. 4, now under review, have the scope and content to satisfy the needs identified in Section 2.0. The Mark I strategy then is to require that all licensees adopt all principal elements of EPG Rev. 4, and revise or modify as riecessary to reflect the changes occasioned by 3.1 to 3.4 above.
4.0 FORMULATION OF REQUIREMENTS Based on the preceding analusis the following requirements should be met by any BWR with Mark I containment.
4.1 Hydrogen Control Present requirements imposed by 10 CFR 50.44 and the Technical Specifications shall be adhered to. The licensees should, however, assure that deinerting will not occur in severe accident sequences (such as station blackout).
~
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15 4.2 Containment Spray l All BWRs with Mark I containment shall provide at least two backup water supply systems for the enntainment drywell spray, one of which shall be functional during station blackout. Water to the spray system from these backup supplies shall be available by remote manual operation or by simple procedures for ,
connection and startup which can be implemented during a severe accident scenario.
In addition, the spray nozzles shall be adjusted so that an evoly distributed spray pattern will be developed in the drywell whether water is supplied by the primary source or either of the backup sources. A flow rate on the order of 1/10 of the present flow rate is considered typical, the licensee shall select the flow based on an analysis of plant specific parameters.
4.3 Pressure Relief The licensee shall select a pressure between design pressure and 1 times design pressure at which to open an exhaust path from the wetwell vapor space to the highest vent point (stack or pipel available. This line should be capable of handling water vapor flow equivalent to 1% decay heat at the vant pressure selected without significant chance of rupture before the desired release point. The line shall be equipped with isolation valves which can be opened and reclosed by remote manual operation or by simple procedures which can be implemented during severe accident scenarios including station blackout.
4.4 Core Debris Control The licensee shall ensure that the water in the suppression pool in the event of torus failure is held within the confines of the torus room and the corner rooms and cannot flow out to other parts of the plant. In addition, the fire suppression sprays should be provided in the ventilation spaces to scrub airborne radioactivity, an'd trisodium phosphate baskets be placed in the torus room to assure pH control to prevent reevolution of radiciodines from torus water.
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! 4.5 Procedures and Trainino i
The licensee shall implement emergency operating procedures and other procedures baseo on all significant elements of Emergency Procedure Guidelines, Revision 4.
4.6 ADS Enhancement High pressure melt ejection of core debris has been postulated to cause direct containment heating and failure of the containment due to overpressurization.
The molten core materials may be ejected from the reactor pressure vessel at high pressure in the form of fine aerosols which directly heat the containment
. and cause a large pressure pulse. This pressure pulse threatens containment ~~
integrity at the time of reactor pressure vessel failure. This mode of containment failure can be essentially eliminated by enhancement of the Automatic Depressurization System (ADS). The ADS enhancement has the additional benefit of reducing the core melt frequency.
The ADS normally serves as a backup to the high pressure coolant injection system. It performs the function of rapid vessel depressurization when high pressure injection systems are inoperable or are unable to maintain adequate water inventory. The ADS control operates solenoid-actuated air (or nitrogen) valves that allow the gas to open safety / relief valves and rapidly depressurize the reactor vessel. The ADS receives electrical power from the plant direct current system and is automatically or manually actuated.
Actuation of the ADS is dependent on an adequate air or nitrogen supply and control power to the solenoids which control the air or nitrogen supply.
NUREG-0737 item II.K.3.28 addressed the adequacy of the air supply and compliance with II.K.3.28 is sufficient to assure a reliable air or nitrogen supply. In a protracted loss of alternating current power, the batteries providing power to the solenoids may be depleted rendering ADS inoperaDie.
Alternative dedicated backup power (batteries or small generator driven power supply) for the ADS could substantially increase the availability of the ADS
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17 for station blackout events and avoid containment failure due to direct heating. Also, the cabling for the ADS may need to be thermally insulated or shielded te assure operability in the hot drywell for the protracted station blackout event.
Because no suitable modifications have been found which can assure containment integrity in direct heating environment, enhancing the availability of ADS to prevent direct heating must be considered.
4.7 Quality and Desien Standards Since these requirements are intended to be an optimized use of existing
~~
. equipment it is expected that added equipment, of itself, need not meet the quality or design standards of safety related equipment. Nevertheless, j modifications to or near equipment or systems which are already safety related shall not' compromise the quality of such equipment or systems.
4.8 . Implementation The equipment changes required herein shall be installed during the first refueling outage which begins nine (9) months after the effective date of this letter. The procedures and training required shall be implemented on a schedule reviewed and approved by the NRC. Given the implementation of the generic improvements of Mark I containments there is no need for an Individual Plant Evaluation (IPE) for containment performance. This does not remove the need for an IPE which covers the system reliability or core melt frequency l portion of the severe accident question.
5.0 JUSTIFICATION FOR REQUIREMENTS There are three possible bases for justification of Mark I containment improvements. They are: l (1) the improvements are needed for safety;
G:.::d d.%;4 3 v 'MF.tdi.? ? . ' G ,'~ IE%BL C GJT<!Nh:e.WE,',WW JrLNG 7" 18 (2) the competing risks due to the proposed improvements are not significant; and (3) the improvements are justified backfits by cost-benefit analysis.
Examination of these bases shows that they support the containment improvements.
5.1 Needed For Safety The present General Design criteria (GDC) set requirements for containment performance. GDC 16 - Containment Design says, ",an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment
- and to assure that the containment design conditions important to safety are '-
not exceeded for as lonc as postulated accident conditions recuire." (Emphasis added). It is clear from the long application of this GDC to many designs that
" postulated accident conditions" are design basis accident conditions, not severe accident conditions. In a similar way GDC,,5D Containment Design Basis says, "[the containment can accommodate with sufficient margin) the pressure and temperature resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as the energy in the steam generators and as required by $50.44 energy from metal water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the cal:ulational model and input parameters." The words " degradation but not total failure of emergency core cooling" clearly limit the application of this GDC to design basis accidents. Thus, consideration of both these GDC indicates that any mandate for change to protect against core melt accidents lies outside the requirements of the existing regulations.
I
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The Commission spoke to the need in the Severe Accident Policy Statement of August 8, 1985:
o Operating nuclear power plants reovire no further regulatory action to deal with severe accident issues unless significant new safety information arises to question whether there is adequate assurance of no undue risk to public health and safety.
o In the latter event, a careful assessment shall be made of the severe accident vulnerability posed by the issue and whether this vulnerability is plant or site specific or r,f generic importance.
o The most cost-effective options for reducing this vulnerability shall be identified and a decision shall be reached consistent with the cost-effectiveness criteria of the Commission's backfit policy as to which option or set of options (if any) are justifiable and required to be implemented.
o In those instances where the technical issue goes beyond current regulatory requirements, generic rulemaking will be the preferred solution. In other cases, the issue should be disposed of through the conventional practice of issuing Bulletins and Orders or Generic Letters where modifications are justified through backfit policy, or through plant-specific decisionmaking along the lines of the Integrated Safety Assessment Program (ISAP) conception.
From these passages it is clear that the Commission intends to deal with severe accident issues if there is a question whether there is adequate assurance of no undue risk to public health and safety. As noted in Section 1.0, the BWR Mark I containment performance in the face of core melt inferred from the Reactor Safety Study was that it had a 90% chance of failure. It was expected l
that more refined analyses of risk available now would show a much lower level of severe accident risks. In many ways that expectation has been satisfied but with the Mark I containment the later results have not been so encouraging.
l
G J D D T ! 5 K m.% ETG 2 ?.'..:: %2i art:.:,X M? . " nXL . ' , d.:. . V.4 WCWb &C. ' M 20 Again as noted in Section 1.0, the IDCOR Technical Summtey Report presents an inferred containment failure rate of about 20%. The analysis done for the smaller Vermont Yankee plant yielded a 7% estimate. Considering the continuing debate on uncertainties in these estimates, it is fair to say that the early failure rate for Mark I containment lies in the range of 90% to 10%. Latest results of the staff analyses for Peach Bottom in support of NUREG-1150 indicate that the highest density of probability of early containment failure occurs at containment failure of 90%.
In discussing a plant with Mark 1 containment in a Congressional hearing on July 16, 1986, the Commission responded as follows to the question:
Question Is a 90 percent chance of failura in the event of a core meltdown an acceptable failure rate?
Answer The NRC holds the position that the likelihood of cere melt accidents in any plant should be very low and, in addition, that there should be substantial assurance that the containment will mitigate the consequences of a core melt should one occur in order to ensure low risk to the public.
It is not merely a question of having low risk but of having as well the defense-in-depth assurance of combined protection by prevention and mitigation...
If we 'are debating in the range of 90% to 10% failure probability, even with the likelihood that it is closer to the lower figure, that is hardly
" substantial assurance that the containment will mitigate the consequences of a core melt should one occur." There is no quantitative synonym for substantial ,
assurance but it is a defensible proposition if the range of debate can be shifted down to something more like 10% to 1%. The Mark I strategy developed in the preceding sections is nut quantified but it does provide significant I
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. 21 changes for the better in each of the areas of greatest uncertainty and I 1
significance for Mark I performance in core melt.
And so it can be argued that these containment enhancements are needed for safety, to ensure low risk to the public by establishing substantial assurance that the containment will mitigate the consequences of a core melt should one occur.
5.2 Competina Risks The competing risks of proposed actions were cdnsidered and conclusions reached are summarized as follows:
' Hydrocen Control Since no new requirements have been identified to maintain hydrogen ignition control on Mark I containments during severe accidents, there are no competing risks introduced.
Containment Spray Enhancement The purpose of this discussion is to examine the competing factors affecting risk which should be considered relative to the installation of enhanced containment sprays in SWRs with Mark I containment. For purpose of the discussion it is assumed that the containment spray systems are modified to provide two backup water supplies with at least one of these water supplies designed to be functional during station blackout. In addition the spray nozzles are assumed to be sized such that an evenly distributed spray pattern will be developed in the drywell whether water is supplied by the primary source (i.e., the RHR system) or either of the backup sources. A flow rate on the order of 1/10 of the present flow rate is considered typical. (Need an explanation of why the 1/10 flow rate is 0.K. , e.g. does not create a problem !
for original spray needs.)
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The modification to the spray system contributes to a decrease in risk based on l the following considerations:
- 1) Increased probability of spray availability decreases containment failure probability as follows:
a) Reduction in containment pressure decreases the likelihood of failure by overpressurization.
b) Sprays flood the drywell floor providing a pool for quenching molten corium thereby delaying or preventing. containment melt through.
L c) Continuous spray cooling retards corium flow reducing likelihood of direct attack on drywell wall or vents.
d) Cooling of containment walls and penetrations reduces the likelihood I of heat induced failure of containment (e.g. transient thermal stress failure of drywell opening and downcomer interface).
- 2) The likelihood of core melt should decrease due to the addition of spray system backup water supplies, one of which is functional during station blackout. This is because the spray equipment interfaces directly with the LPCI systems and thus the additional water supply would also be available for core cooling.
- 3) Accident consequer.ces are decreased due to:
a) Scrubbing of aerosols by sprays, b) Delaying time to failure due to overpressure and core debris attack.
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I The modifications may adversely affect risk as follows:
a) Actuation of drywell spray may degrade electrical equipment such as motors for containment isolation valves or containment cooling fans.
Use of these systems for recovery may be hampered.
b) The possibility of ex-vessel steam explosions may be incre' ased due to greater availability of a pool of water on the drywell floor.
c) The addition of backup equipment for the spray system may result in new interfaces between safety grade an,d non-safety grade systems.
These new interfaces could lead to system interactions which degrade
- the reliability of the safety systems.
d) The availability of a reliable spray system may influence en operator to incorrectly divert core cooling flow to containsint cooling, particularly in situations where instrumentation is unreliable. This
. could increase the chances of a core melt.
It should be noted that the adverse impacts of containment spray are much less certain than the potential benefits. For example, studies of steam explosions, although mostly performed for in-vessel conditions, have concluded that events of significant energetics are unlikely. It is generally believed that ex-vessel considerations would not change these conclusions since the energetics in ex-vessel interaction will be even lower. The chance of an operator incorrectly diverting flow from the core to the containment can be minimized through training and procedures. Likewise equipment interface problems between safety and non-safety equipment can be addressed through careful design considerations. Degradation of drywell equipment is not considered to be significant since the accident environment itself for which the electrical equipment is qualified is likely to be at least equally hostile.
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24 On the other hand, the pressure suppression, containment cooling, and fission product scrubbing benefits of spray are to be expected and the availability of an additional water supply for core cooling is a clear advantage.
Pressure Relief When one considers venting as a means of controlling the course of a severe transient, one must also consider the risk tradeoffs. (It should be noted that venting is provided for in the regulations under 10 CFR 50.44, and in BWR emergency procedure guidelines and emergency procedures.) The risks associated with venting must first consider whether the core has melted. If venting is considered prior to core damage, little or no fission products beyond a fraction of the noble gas inventory would be released. But then, one must *-
include in the risk assessment, the risks from causing a core melt due to venting. For example, the venting process could damage equipment that could provide the necessary core cooling later on in the transient. During the venting process failure of the line from containment to the stack could produce a radioactive steam environment in the rooms containing important equipment.
This could cause failure of the equipment as well as prevent repair of damaged equipm3nt. Such an event could severely impact the ability of the operators to control the course of the accident, or to implement post-accident recovery.
Assuming that the core has melted prior to the decision to vent, the most obvious negative aspect is venting after core damage when it is unnecessary.
In other words, the containment would survive the transient without venting.
In this situation there would be an unnecessary release of radioac' > ty to the atmosphere. The radioactivity released, however, would be virtually only noble gases since suppression pool st, rubbing would most likely occur. Additionally, one must consider the possibility of being unable to close the vent valves after the necessary venting is completed. This would result in a loss of !
containment function for the remainder of the transient.
Another risk is associated with the potential of containment collapse. Venting during the transient can result in the loss of a significant fraction of the
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, 25 mass of noncondensibles within the containment. Upon closure of the vent valves, the containment atmosphere will eventually cool, if containment sprays continue to operate, until a negative pressure is established in the containment. Due to the loss of noncondensibles, the resulting reversal of pressure differential could cause containment failure. For free standing steel containments, however, vacuum breakers are provided to prevent this type of failure. Therefore, containment failure would also require failure of the vacuum breakers.
There are also risks of not venting which should be considered. The most serious consequences of not venting are the complete loss or ECCS as a result of containment failure due to overpressure, and the uncontrolled release of 1
large quantities of fission products. Such a containment failure could be below the suppression pool wster level, thereby losing the pool as a source of core cooling and fission product scrubbing. It could also fail violently, causing failure of vital equipment in and around the containment. Excessive There are also risks of not venting which should be considered. The most serious consequences of not venting are the complete loss of ECCS as a result of containment failure due to overpressure, and the uncontrolled release of large quantities of fission products. Such a containment failure could be below the suppression pool water level, thereby losing the pool as a source of core cooling and fission product scrubbing. It could also fail violently, causing failure of vital equipment in and around the containment. Excessive containment pressure could also cause loss of primary system pressure relief capability provided by the automatic depressurization system (A05). Since the ADS valves are air operated, high containment pressure could result in an inability of the air operators to open the valves. High containment pressure could also effect the operability of a wide range of equipment and instrumentation inside containment. Exceeding the equipment environmental profile could cause failure of the components to perform their function.
Therefore, the risks of not venting far outweigh the risks of venting.
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26 Core Debris 1
The requirement to assure that the torus water is retained in the torus room, can pose some risk of loss of equipment in the corner rooms due to flooding.
For plants where the corner room equipments are elevated at least two feet above the ficor level, there will be no flooding risk. For plants where the equipment is at the floor level, the risk due to loss of any equipment in the corner room due to flooding will be small compared to the risk due to escape of highly contaminated torus water outside the torus room following a core melt accident. Similarly, if containment is breached the use of fire sprays may inhibit accident management and cleanup operatichs. However, the potential mitigation benefits are considered to substantially outweigh the disadvantages. ._
5.3 Costs and Benefits 1
The estimated costs of proposed action would vary substantially depending upon specific designs of plants and ease with which performance enhancements could be incorporated. IDCOR* has presented approximate ranges of costs. The estimated costs do not reflect any unique engineering difficulties or time
~
available for modifications to be incorporated in plant maintenance outages.
The cost considerations included in the IDCOR study include the following:
o Hardware o Installation o Test and Maintenance o Plant Unavailability o ALARA (Exposure Costs) o Costs of Procedure Changes and Training and Impact of Proposed l Backfits
- Evaluation of BWR Accident Mitigation Capability Relative to Proposed NRC Changes, August 1986.
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The cost of drywell sprays using fire pumps available at all plants was estimated to range from 0.6 million dollars to 1.1 million dollars. The cost of venting using vents of 6 to 18 inch ducts ranged from 0.1 to 1.1 million dollars. The cost of installing a short debris barrier in the torus room, the use of trisodium phosphate in the torus room and fire spray enhancement in the reactor building was estimated by the staff to be 2.5 million dollars. No additional cost is expected for Mark I containment hydrogen control. No significant new costs are- expected for implementing the emergency procedures I guidelines that the industry is pursuing as a result of TMI actions. Based on the above estimates it wou,1d appear that the cost of the proposed initiative should range from 3.2 to 4.7 million dollars per reactor.
~
To estimate the benefits of containment improvements one must estimate the averted loss. The terms needed to estimate it are:
o FCM, Frequency of Core Melt o CCFP, Conditional Containment Failure Probability o Loss, the monetized cost of a large release.
Taking Frequency of Core Melt first, the containment improvements include operats' training and procedures for handling the containment. Because of the close interaction of systems in a BWR improved procedures will undoubtedly have an effect of reducing FCM. For reference, in the Reactor Safety Study, about two-thirds of all core melts were caused by the failure of containment due to overheating which failure then caused the loss of core cooling. For simplicity in the calculation here the reduction of FCM will not be included, thus apparently underestimating the value of the containment features. It is reasonable to do this because the reduction in FCM will come principally from the training and procedures whose costs are not included in the preceding section since the BWR owners are already committed to adopting most of them.
The proper choice of a typical FCM is dif ficult to make. Current IDCOR and
-5 NRC analyses of Peach Bottom suggest FCM on the order of 1x10 /yr although the NRC results are dominated by station blackout sequences while the 10COR results
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.- ' gg are dominated by ATVS. Examination of the results presented in Table 1 indicates a number of plant specific cases where FCM ranges up to and above 1x10'#/yr. Considering the diversity of systems and the flexibility of
-5 operation in a BWR, a FCM of 1x10 /yr may well be attainable. However, the modelling differences between IDCOR and NRC results at that level and the
~4 results from other plants at the higher level suggest the choice of 1x10 /p as the typical value for cost-benefit analysis purposes.
The cost-benefit equation calls for a quantitative estimate of the CCFP before and after the containment changes. Considering the range of debate on the present state of containments, given previously ,as 90% to 10%, it is reasonable to use 50% as the "before" figure. If substantial improvement is achieved the
. exact value of the "after" figure is not important, but 5% will be used here. ~
The monetized consequences of a large release coming from early containment failure can be large. Consideration should be given to counting health effects above or counting offsite economic consequences as well. Previous work" indicates that large early releases can cause on the order of 10 person-rem offsite exposure. Monetized at $1000 per person-rem, this gives consequences 10 of $10 per event. Studies also show** that offsite economic consequences of large early releases can be up in the tens of billions of dollars or even more.
On the other hand, source term studies centinue and some argue that these high consequences are derived from WASH-1400 vintage source terms. For this cost-benefit calculation the health effects only consequence of 10 person-rem 10
. or $10 will be used. Another question arises in converting annual averted health effects into a present worth value. Since these are human health effects some argue that they should not be discounted in a present worth calculation. If that approach is taken then the averted loss per year is
- NUREG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accidents," Appendix A, September 1982
- NUREG/CR-3673, " Economic Risks of Nuclear Power Reactor Accidents" p.2-13, April, 1984
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a 4 29 multiplied by the remaining years of plant operation. The other approach is to use a discount factor such as would be associated with averted economic losses.
Table 2 has been prepared to illustrate the central estimate for the cost benefit analyses.
Referring to Table 2, the cost is listed as 3.2 to 4.7 million dollars, or in a rounder number, less than $5 million per reactor. The base calculation gives a benefit of $3 million to $12 million, indicating a balance of cost and benefit or a clear justification. The other calculations in Table 2 are a sensitivity analysis to explore the range of outcomes with different assumptions. The lower FCM case uses the frequency of core melt currently being calculated by 10COR and NRC for Peach Bottom. The next case illustrates less improvement in .-
~
containment performance, only a factor of five. The next case assumes that containment performance now is better, the 20% CCFP inferred in the IDCOR report. The " optimistic" case uses the IDCOR values for FCM and present containment performance while the " pessimistic" case assumes a relatively high FCM and CCFP.
Comparison of these estimated benefits to the range of costs indicates that these proposed changes are justified backfits.
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TABLE 3 COST-BENEFIT ANALYSIS COST: $3.2 TO 4.7 MILLION BENEFIT:(1) FCM CCFP CCFP AVERTE0 AVERTE0 BEFORE AFTER LOSS /YR LOSS PRES. VALUE BASE
~4 CALCULATION 1x10 /yr 0. 5 0.05 $4x105 /yr $3.6M/$12M
-5 LOWER FCM 1x10 /yr 0.5 0.05 $4x104 /yr 50.3M/$1.2M l LESS CHANGE
~4 IN CONTAINMENT 1x10 /yr 0. 5 0.1 $4x105 /yr $3.2/$12M BETTER CONTAINMENT
~4 TO START 1x10 0.2 0.05 $2x105 /yr $2M/$6M "0PTIMISTIC"
-5 CALCULATION 1x10 0.2 0.05 $2x104 /yr $0.2M/$0.6M
" PESSIMISTIC"
~4 CALCULATION 3x10 0. 9 0.1 $2x105 /yr $16M/560M (1) FCM = Frequency of Core Melt CCFP = Conditional Containment Failure Probability AVERTE0 LOSS PRESENT VALUE expressed as A/B where A is the averted loss per year times 8 (roughly equivalent to discount at 12%/yr rate) and B is the averted loss per year times 30 (no discount).
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. Safety Evaluation No.: 510lp SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION Rev. No. O PDC PCN System Calc.
Group: No.: Name: No.: Date: 9.- t - g> g Initiator: Dept:
Do'eSd Sa_
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SAFETY EVALUATION CONCLUSIONS:
The proposed change, test or experiment:
1, (x) Does Not ( ) Does increase the probability of occurrence er consequences of an accident or malfunction of equipment important to safety previously evaluated in the F54R.
- 2. (g) Does Not ( ) Does increase the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
- 3. (X) Does Not ( ) Does decrease the margin of safety as defiesd la the basis for any technical specification. I I
BASIS FOR SAFETY EVALUATION CONCLUSIONS:
5n Rfta e h wreAT A.
A-Change Change
( ) Not Recommended Q() Recommended b Date 77 _
SE Performed by j
Exhibit 3.07-A Rev. 3 Sheet 1 of 3
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SAFETY EVALUA?!0N No.: [ ) ' ((h ~" b P2LGRIM NUCLEAR POWER STATf0N Rev. No. ()C A. APPROVAL
- X) This proposed change does not involve a change in the Technical Specifications.
(X) This proposed change, test or experiment does ( ) does not Q()
involve an unreviewed saf ety question as defined in 10CFR, Part 50.59(a)(2).
$) This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR50.59(b).
(X) Conuments: f9C 8b'HA U CA'*k C.v.'l a:4 s.fe & Qe
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(X) Approved () Ilot Approved l l
/ / m 4 4 -B 'I Discipli te Suppcrting Discipline Group Leader /Date fleas S. RE APPROVAL
) Co u nts: MN -
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& i @ Grifulp Leader /6 ate C. ORC REVIEW
() This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with the Directorate of Licensing, alRC prior to implementation. ,
J
() This proposed change does not involve an unreviewed safety question.
ORC Chairman Sate ORC fleeting Number cc:
Exhibit 3.07-m Rev. 3 Sheet 2 of 3 l
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- PIL6 RIM STATION FSAR REvlEW SHEET 4
References :
Safety Evaluation: P 10 (/ Rev. No.: O Support a change Date: '
4-S-e]
List FSAR ' test, diagrams, and indices af fected by this change and
, corresponding FSAR revision. )
Affected FSAR
)
Section Revision to affected FSAR Section is shown en:
Preliminary Final I i, nt 2 P Attachment 3 Attachment 4
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Attachment $ J Attachment 6 i 1 t
PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation) .
Prepared by: .016
/Date: Reviewed by M te: /N7 l Approved by:M W <0 ate: 3 ff& ! I U O
{ 1 FINAL FSAR REV1510N (Prepared following Operational turnover Of related !
systems structures of components for use at PNPS). (1) j
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Prepared by: /Date: Reviewed by: I !
/Date: l l
(1) Attach completed FSAR Change Request Foru (Refer to NOP).
Exhib.~t 3.07-A Rev. 2 Sheet i of 3
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SAFETY EVALUATION WORK SHEET I
Rev. No. O
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A. System Structure Component Failure and Consequence Analyses.
System Structure Component Failure Modes Effects of Failure Comments j MdfP liCAlb h N 'N 1.
(S'em. Er hibcP 3.o7-M 2.
3.
General Reference Material Review cat.CULATIONS REGULATORY FSAR SECTION PNPS TECHNICAL SPECS. DESIGN.. SPECS PROCEDURES GUIDES STANDARDS CODES WY AWw 4 l .
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- 8. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix 6).
Prepared by Bate 3b' I
l NOTE: It is a requirement to include this work sheet vtsh the Safety I Evaluation.
Exhibit 3.07-C Aev. 2 W
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. 1 Attachment A to Safety Evaluation # S tok slut t t'6 t-and PDC 86-56A A. DESCRIPTION OF CHANGE This change describes the preparation activities required to prepare the site yard area for the installation of a new diesel generator facilities:
The preparation activities include the.following:
(a) Excavation for the fuel oil tanks, the halon tank (s)/ neutral resistor foundation, the switchgear foundation, the diesel generator foundation and radiator foundation.
(b) Compact subgrade below diesel generator.
(c) Placement of reinforcement and concrete for the foundations.
(d) Installation of the fuel oil storage tanks and associated piping.
(e) Installation of concrete duct banks for electrical cable.
(f) Installation of grounding for diesel generator, neutral grounding resistor, switchgear and duct bank.
(g) Backfilling the fuel oil tanks with pea gravel.
(h) Structural backfill under diesel generator.
(i) General backfill and compaction.
(j) Placement of reinforcement and concrete for protective slabs above the fuel oil tanks.
(k) Installation of cathodic protection for all underground piping.
(1) Installation of paving, curbs, finishing grading and installation of crushed stone and landscaping stone ground cover.
The final result will be a facility ready to receive the diesel generator mechanical and electrical components.
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'B. Purnose of the Chance The purpose of this modification is to prepare the site yard area located south of the Turbine Building and adjacent to the Relay House for the installation of a new diesel generator set. 'This new diesel generator. set will provide an additional source of back-up electrical power and reduce the probability of a complete station blackout.
C. System. Subsystem. Comeonents Affected The changes implemented under this PDC package have no impact on any existing safety-related buried comodities nor do they modify any existing safety-related electrical / mechanical systems and/or civil structures.
D, Safety Function of Affected Systems Since this PDC affects no existing plant systems it has no effect on any safety functions.
E&F Effect on Safety Functions /and Analysis See Q-above G. Summary This modification describes the preparation of a yard site for the installation of a new diesel generator. Since these activities do not affect any existing plant systems, it has no effect on any safety related system of safety function This modification does not involve an unreviewed safety questior.
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