ML20245F164
| ML20245F164 | |
| Person / Time | |
|---|---|
| Issue date: | 01/26/1989 |
| From: | Jordan E Committee To Review Generic Requirements |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| NUDOCS 8902130177 | |
| Download: ML20245F164 (49) | |
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NUCLEAR REGULATORY COMMISSION N PDR' t
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MEMORANDUM FOR:
Victor Stello, Jr.
j-
-Executive Director for Operations FROM:
Edward L. Jordan, Chairman Committee to Review Generic Requirements
SUBJECT:
MINUTES OF CRGR MEETING NUMBER 155 The Committee to Review Generic Requirements (CRGR) met on Wednesday, January 11.1988 f rom 11: 00 a.m. - 5:15 p.m.
A list of attendees for this meeting is enclosed (Enclosure 1).
The following items were addressed at the meeting:
1.
The Committee completed at this meeting their review (begun at Meeting No.
155) of proposed modifications to Mark I containments, and related safety enhancements (e.g., accident management procedures), aimed at improving significantly the severe accident capability of Mark I cot:tainments.
The Committee recommended in favor of imposing the proposed severe accident related backfits, subject to satisfactory resolution of several residual-issues identified in their review and incorporation of a number of specific changes to the wording of the package (all changes to be coordinated with the CRGR staff).
This matter is discussed in Enclosure 2, 2.
S. Crockett (0GC) and J. Wilson (RES) presented for CRGR review the pro-posed final rule, 10 CFR Part 52, on standardization of advanced reactor designs, passive LWR designs, and evolutionary LWR aesigns.
The Committee recommended in favor of sending the proposed rule forward for final con-sideration by the Commission, subject to several modifications (to be coordinated with CRGR staff).
This matter is discussed in Enclosure 3.
In accordance with the ED0's July 18, 1983 directive concerning " Feedback and Closure on CRGR reviews," a written response is required from the cognizant office to report agreement or disagreement with the CRGR recommendations in these minutes.
The response, which is required within five working days after receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is. disagreement with CRGR recommendations, to the EDO for decisionmaking.
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e Questions concerning these meeting minutes should be referred to Jim Conran (492-9855).
Originni Frid Bvi E. L hrden Edward L. Jordan, Chairman Committee to Review Generic Requirements
Enclosures:
As stated cc/w enclosures:
Commission (5)
SECY Office Directors Regional. Administrators CRGR Members Distribution: (w/o encl)
Central File PDR (NRC/CRGR
- 5. Treby W. Little M. Lesar P. Kadambi (w/ enc.)
W. Houston (w/ enc.)
A. Thadani (w/ enc.)
- 5. Crockett (w/ enc.)
J. Wilson (w/ enc.)
CRGR SF (w/ enc.)
M. Taylor (w/ enc.)
E. Jordan (w/ enc.)
l J. Heltemes (w/ enc.)
J. Conran (w/ enc.)
3CXSakenas'(w/ enc:)
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- SEE PREVIOUS CONCURRENCE OFC
- *CRGR: AEOD :DD D
- C:CR OD -
1
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, ____..___.q NAME :JConran
- C ltemes rdan I
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.........................................................--............a l
DATE :1/19/89: jr :1/4j)89
- 1/23/89 0FFICIAL RECORD COPY i
_____________j
. - _ = _ - _ - _ _ _ _ _ _ _ _ _
- i.
L Attendance List CRGR Meeting No. 155 CRGR Members E. Jordan J. Sniezek
- 8. Sheron (for D. Ross)
R. Bernero l'
'J. Goldberg C. Paperiello NRC Staff C. J. Heltemes C. Sakenas J. Conran E..Beckjord W. Houston t
B. Beckner l
A. Thadani
}
J. Ridgeley l
J. Kudrick l
L. Soffer T. Collins M. Thadani T. Cox l
A. Serkiz M. Taylor
-J. Scinto l
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_-_.___m_.___
_.____-_.m__..___J
r l to the Minutes of CRGR Meeting No. 155 Proposed (Severe Accident) Enhancements to Mark I Containments TOPIC E. Beckjord (RES), T. Murley (NRR), W. liouston (RES), B. Beckner (RES), and A. Thadani (NRR) presented for CRGR review a proposed package of modifications to Mark I containments, and related safety enhancements (e.g., accident manage-ment procedures), aimed at improving significantly the severe accident cap-ability of the Mark I Containments.
(The Committee began their review of this matter at Meeting No. 152.) Copies of the briefing slides used by the staff to guide their presentations and the discussions at this meeting are attached (Attachment 1).
BACKGROUND 1.
Subsequent to Meeting No. 152, the staff revised the original proposed Mark I package in response to CRGR comments.
The revised material was transmitted to CRGR by memorandum dated January 4, 1989, E.S. Beckjord to E.L. Jordan; the revised material was comprised of the following:
a.
Draft Commission Paper, " Mark I Containment Performance Improvement Program," and (revised) enclosures as follows:
- i. - Regulatory Analysis ii. - Draft Proposed Rule iii. Appendix A - Backfit Analysis 2.
The January 4 version of the package was further revised on the basis of EDO office comments.
Due to time constraints, that latest version of the Ma.'k I package was not transmitted formally to CRGR, but was provided directly to Committee members at Meeting No.155.
That material is attached to these minutes for completeness of record (Attachment 2); it includes the following:
a.
Draft Commission Paper dated January 11, 1989 and (further revised) to the Commission Paper, " Draft Proposed Rule."
b.
Draft Plant-Specific Order.
CONCLUSIONS / RECOMMENDATIONS As a result of their review of the propcsed Mark I upgrades, including the discussions with the staff at Meeting No. 152 and this meeting, the Committee recommended in favor of imposing the proposed Mark I improvements, subject to resolution of the following comments / recommendations (to be coordinated with the CRGR staff):
t l ' -.
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, -1.z The Committee recommended expedited rulemaking as the method for imple-l menting approved Mark I upgrades.
First, this is consistent with existing Commission guidance on preferred approach for addressing new severe-accident requirements.
Second, the concern was stated in discussions at-this meeting that litigation of the issues involved in the proposed Mark I upgrades (in hearings that could be requested under'the plant specific Orders approach recommended by the staff).could delay implementation of the recommended upgrades significantly longer _than expedited rulemaking.
The
' staff.did not agree that such litigative risks are an overriding concern, but did agree to further highlight that concern to the attention of the Commission in the final package.
2.
Instrumentation requirements in this package (e.g., at the top of p.2 of the Draft Rule) should be sharpened and more fully defined-to better assure' that the improved severe accident coping functions intended can actually be successfully carried out by plant operators.
3.
The extensive cost-benefit treatment of the separate elements of the over-all package of Mark I fixes containe in this package (e.g., in the discus--
sion of Alternatives ii., iii., iv. and v. in the Regulatory Analysis) detracts from the case the staff is trying to make for the synergistic,.
integrated set of modifications finally recommended (i.e., Alternative vi.).
The package should be revised to give greater emphasis to the staff's objective of providing defense-in-depth protection (i.e., both preventive and mitigative measures) against the dominant severe accident sequences for Mark I containments, so that even if a (low probability) core melt occurs, there is reasonable expectation (i.e., comparable to that for most PWRs) that the containment will be able to mitigate the consequences.
4.
A major weakness of the current package is the discrepancy (a factor of three-or-so) between the staff's estimate of licensee costs to implement the recommended backfits and the actual costs incurred by one-licensee in
' implementing (voluntarily) a number of those same Mark I upgrades.
The staff should resolve that discrepancy, and revise the package to better; explain the apparent difference.
5.
The Committee questioned seriously the feasibility or practicality of accelerating ATWS and Station Blackout (SBO) rule implementation, as recommended in the current package.
The package should be revised to more clearly indicate (in accordance with discussions with the staff at this meeting) that (a) implementation of approved Mark I upgrades must be carefully coordinated with those ongoing rule implementation eff. orts, but (b) no acceleration of licensees' actions is intended.
1 L
.The Committee also recommended-the following specific modifications to the 6.
current package:
-a.
Draft Rule, p.2, paragraph 2.b.:
Delete the first sentence, and revise the remaining. wording, if necessary, to make clearer the staff's intent-(i.e., provision must '
-be made.to assure the capability to vent at design pressure; but venting at low pressure is not precluded, and no special provision-need be made to preclude inadvertent venting at ' low pressure).
b.
Draft Rule, p.2, paragraph 2.d.:
Delete entirely. f(Also delete. corresponding paragraph IV.A.1.d..at-
- p. 7 of the Draft-Order.)
c.
Draft Rule, p.2,' paragraph 2.e.:
- Change "should" to "shall."
d.
Draft Rule, p.3,~ paragraph 2.g.:
Change to read ".. alarmed and indicating in the control room..," and make clear that the requirement for a. radiation monitor in this para-graph could be' met with an approved poet-TMI stack radiation monitor,.
if already installed.
e.
Draft Rule, p.3, paragraph 3.:
The EQ requirements in subitem 1.) should be deleted if (as indicated in the discussions at this meeting) no. additional or more stringent.
qualification of the subject cabling, beyond that provided by compliance with 10CFR50.49 requirements, would be required for assured operation prior to vessel failure.
f.
Draft Rule, p.3, paragraph 4.:
Change "should' to "shall" in the second sentence, and delete the last sentence.
g.
Draft Rule, p.4, first paragraph:
Change ". 30 days.." to ".4 60 days.." in the first sentence, and add to the end of the second sentence "..and licensees shall certify to' NRC that they have met the rule."
i
7 a.
t 4-h.
Draft Rule, p.4, paragraph 2. a. :
Change to read "..by use of an alternate AC (AAC) source, as defined in Section 3.3.5 of Reg. Guide 1.155,.."
l i.
Draft Order, p.6, third sentence of the first full paragraph:
l l
Delete the words ".. raising the vent valve operability pressure and/or.."
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j Draft Order, p.8, last sentence of paragraph 2. :
Change "should" to "shall."
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4
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1 BRIEFING ON BWR MARK 1 CONTAINMENT PERFORMANCE BEFORE CRGR JANUARY 11, 1989 WAYfE H0JSTON WILLIAM J. 3ECKNER OFFICE OF NUCLEAR REGULATORY RESEARCH U. S. NUCLEAR REGULATORY COMISSION A 4+a c h m e n Y l
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BACKGROUND
. JUNE 1986, STAFF PROPOSED 5 ELEMENT PROGRAM FOR 1%RK 1 CONTAINTB6 PERFORMANCE ENHANCEMENT l
JUNE-JULY 1987, TWO LICENSEES INFORMED THE STAFF 0F THElR INTEtHION T0 ITNESTIGAE CONTAlf4 MENT AND SAFETY ENHANCEMENT JULY 7,1987, STAFF BRIEFED C0mlSS10N ON A PLAN FOR CLOSURE OF SEVERE ACCIDENT ISSUES DECEMBER 1987, " MARK I CONTAlf#E R PERFORMANCE PROGRAM PLAN",
3 (SECY-87-297)
FEBRUARY 1988, WORKSHOP ON MARK I ISSUES MAY 1988, " INTEGRATION PLAN FOR CLOSURE OF SEVERE ACCIDENT ISSUES",
(SECY-88-1!47)
JULY 1988, " STATUS OF MARK I CONTAINENT PERFORMANCE EVALUATION",
DECEMBER 6,1988, ACRS SUBC0mlTTEE ON CONTAlffENT SYSTB4S DECBGER 114, 1988, CRGR REVIEW DECEMBER 15, 1988, ACRS FULL COMITTEE 2
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SUfEARY 0 S"AFF 1
RECOTBm T: ONS fDLiML l
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ACCELERATE IWLEENTATION OF STATION BLACK 0UT RULE
( ATWS IWLEENTATION TO BE ESSENTI ALLY COMPLETE BY 1/89)
REQUIRE ALTERNATE WATER SUPPLY FOR DRYtELL SPRAY / VESSEL l
]
INJECTION WITH PUTING CAPABILITY INDEPENDENT OF NORF%L AND EERGENCY AC 1
REQUIRE HARDENED VENTING CAPABILITY FROM WET \\ ELL (ABLE
-TO WITHSTAND SEVERE ACCIDENT PRESSURES).
ISOLATION VALVES TO BE REMOTELY OPERABLE INDEPENDENT OF NORMAL AND EERGENCY AC.
1 i
REQUIRE ENHANCED ADS RELIABILITY. ADDITIONAL POWER AND/0R NITROGEN SUPPLY AND CABLE RELIABILITY REQUIRE iMPLBETATION OF IWROVED EPG'S (REV. 4 0F BiE0G) l 3
l.
CHANGES TO CPI PACKAGE FOR l%RK ls PROPOSED IMPLEMENTATION THROUGH PLANT-SPECIFIC BACKFITS C0ORDINATED RE0J1RBENTS OF THE SB0 RULE WITH PROPOSED CPI BACKUP POWER NEEDS INCREASE ESTIMATED AVERAGE REMAIN 1f4G PLANT LIFE FROM 20 YEARS To 25 YEARS PERFORMED COST-BENEFIT SENSITIVITY FOR INCRBENTAL ADDIT 10f4 0F ADS AND BACKUP WATER SUPPLY (SECTION 4.1.7)
REVISED COST-BENEFIT ANALYSIS TO If4CLUDE EFFECT OF PROPOSED MITIGATION ENHANCBENTS ON SUE ATWS SCENARIOS 4
4 COST-BENEFIT RESULTS (MAN-REM AVERTED PER MILLION DOLLARS)***
LOW RISK PLAT 4T (TV=10-5)
LOW INDUSTRY COSTS
- 1,970 HIGH INDUSTRY COSTS **
500 BIGRRISKPLANT (TW=10-4)
LOW INDUSTRY COSTS
- 29,600 HIGH INDUSTRY COSTS **
4,570 LOW INDUSTRY COST IS ESTIMATED TO BE $48 MILLION HIGH INDUSTRY COST IS ESTIMATED TO BE $176 MILLION INCLUDES AVERTED ON-SITE COST OF CLEANUP, REPAIR AND REPLACEMENT POWER i
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COST BENEFIT FOR INCREMENTAL ADDITION 0:
ADS AND BACKUP aARR SUPPLY UiAN-RBi AVERTED PER MILLION DOLLARS)
HIGH INST.
ASSUPPTION AVERTED QST g)31 BASE CASE' 33 630 190 AVERAGE SB0 PROB."
55 1,050 310 AVERAGE SB0 PROB.
93 1,780 530 l
HIGH LINER ELT PROB."*
AVERAGE SB0 PROB.
135 2,580 760 HIGH LINER ELT PROB.
ATWS MITIGATION""
HIGH SB0 PROB."*"
165 3,150 930 HIGH SB0 PROB.
280 5,344 1,580 HIGH lifer ELT PROB.
'HIGH SB0 PROB.
322 6,145 1,816 HIGH LlfER ELT PROB.
A W S MITIGATION HIGH SB0 PROB.
64 1,221 361 HIGH LINER ELT PROB.
AW MITIGATION LOW POPULATION
-6 BASED ON SB0 FREQUENCY OF 6x10 /RY AND A C0t4DIT10NAL LINER E LT PROBABILITY OF 0.5 GIVEN A CORE ELT.
BASED ON A SB0 FREQUENCY OF 1x10-5/RY WHICH IS Afl AVERAGE FOR PLANTS.
CONDITIONAL Lil4ER MELT PROBABILITY OF 1.0 GIVEN A CORE ELT.
ASSU T S ATWS MITIGATION OF 42 MAfi-REM PER RY.
-5 BASED ON A SB0 FREQUENCY OF 3.5x10 /RY.
L 6
CD-ORDINATION W1TH SB0 RULE (10 CFR 50.63)
COI'PLIANCE WITH 10 CFR 50.63 BY:
ALTERNATE AC (AAC) POWER SOURCE DOES I40T NEED ADDITIONlt POWER SOURCE JE ACC SATISFIES POWER NEEDS FOR PROPOSED i%RK 1 ENHANCEMENTS COPING ANALYSIS MUST PROVIDE ADDITIONAL POER SOURCE TO SATISFY PDER I4EEDS FOR PROPOSED !%RK 1 Ei1HANCEENTS l
1 7
a_______.
PROPOSED liPLEENTATION HAVE EXAMllED TWO OPT 10!1S:
RULEMAKli4G PLANT SPECIFIC BACKFITS (ORDERS)
PREPARE ENVIR0t4Et1TAL ASSESSENT OF VENTING SCHEDULE LICENSEES TO SUBMIT PLANS Af4D ANTICIPATED SCHEDULE WITHitt 60 DAYS OF BACKFIT REQUEST ItPLEENTATION TO BE C0FPLETED WITHlf4 30 MONTHS OF BACKFIT REQUEST l
L_ _ _---_ _-
.l CONCLUSIONS Ai4D REC 0t1ENDAT10NS PROPOSED ENHANC&ENTS PR0/IDED SUBSTANTIAL INCREASE IN-DVERALL PROTECTION OF PUBLIC HEALTri AND SAFETY PROPOSED ENHANCEMENTS ARE GETERALLY COST BENEFICIAL PROPOSE TO IMPLETEllT VIA ORDERS CONTINUE CONFIRMATORY RESEARCH ON PHEN 0iENA RELEVANT TO IN-VESSEL AND EX-VESSEL ACCIDENT PROGRESSION, THE EFFECT OF WATER ON THE PROBABILITY OF LINER MELT-THROUGH, AND ASSOCIATED SOURCE TERMS l
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.s55 BWRs Dominant Accident Sequences ATWS Station Blackout SB0 Long Tem Heat Removal Cont. Failure Before Core Melt TW-Others-Decay Wat
. Key Containment Failure Moaes Overpressure Liner Melt-thru.
(Uncertain)
Proposed Fixes Reduce Core Melt Nequency (TW, SB0,, Decay Heat)
Cont. Failure Probability (Reduces Potential for Key. Failure Modes)
Source Term (Prevides Water) i 4
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4 REVISED PAGES - REVISIONS MARKED January 11, 1989 For:
The Commissioners From:
Victor Stello, Jr.
Executive Director for Operations
Subject:
MARK 1 CONTAINMENT PERFORMANCE IMPROVEMENT PROGRAM
Purpose:
To present staff recommendations on Mark I containment performance improvements and other safety enhancements.
Category:
This paper covers a major policy question.
Summary:
As noted in the Integration Plan for Closure of Severe Accident issues (SECY 88-147) and in interim reports to the Commission (SECY 87-297 and SECY 88-206), the staff has undertaken a program to determine what actions, if any, should be taken to reduce the vulnerability of containments to severe accident challenges. The containment performance improvement effort is one main eiement of the integrated approaci, to closure of severe accident issues. Staff efforts have focused initially on BWR plants with a Mark I containment. The staff has now completed its assessment of generic severe accident challenges and failure modes as well as potential improvements for plants with the Mark I containment. The review of Mark II, Mark III, and other
- ontainment types are the subject of parallel but separate trogram efforts, as discussed in SECY 88-147.
P.obabilistic Risk Assessment (PRA) studies have been performed for a number of BWRs with Mark I containments.
These studies indicate that BWR Mark I risks are dominated by Loss of Decay Heat Removal, Station Blackout (SBO), and Anticipated Transient Without Scram (ATWS) sequences.
Although these studies do not show the BWR Mark I plants to be risk outliers as a class relative to other plant designs, they do suggest that the Mark I containment integrity could be challenged by a large scale core melt accident, principally due to its smaller size. However, estimates of containment failure likel? hood under such conditions are based on analysis of corrplex accident conditions, where there remains a broad band of uncertainty.
Contact:
W. Beckner, RES 492-3975 L. Soffer, RES 492-3916 l
A +Ya c k
- nT To E,, w l
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The staff has concluded that the optimum way to reduce overall risk in BWR Mark I plants is to pursue a balanced approach' utilizing accident prevention and mitigation.
Based on our assessment including the atove described balanced approach, the staff recommends five specific improvements for Mark I containment plants: 1)animproved hardened vent' capability, 2) improved reactor pressure vessel (RPV) depressurization system reliability, 3) an alternate water supply'to the reactor vessei and drywell sprays, 4)' extended emergency procedures and training and
- 5) accelerated implementation of the existing ATWS and SB0-rules. These improvements, when fully implemented, will substantially enhance the safety of Mark I plants, including improvement to containment performance. The staff has evaluated them and found them to be cost effective. The staff proposes.that orders be issued-to all licensees with Mark I containments to implement these improvements.
Background:
The Reactor Safety Study _{ WASH-1400)_'found that, for the Peach Bottom BWR Mark I nuclear plant, even though the core melt probability was relatively low, the containment could be severely challenged if a large core melt occurred.
Based on this conclusion and reinforced by the anticipation confirmed) in the draft of similar findings (subsequently (NUREG-1150, February 1987)
Reactor Risk Reference Document a five_ element program was proposed in June 1986 to enhance the performance of the BF!R Mark I containment. Elements of this proposal included 1) hydrogen control, 2) containment drywell spray, 3) containment venting, 4) core di:bris control, and 5) emergency procedures and training.
After the initial proposal, the staff held two separate meetings in early 1987 with researchers representing NRC contractors and industry. There was a wide range of views expressed regarding accident phenomenology as well as the efficacy of the various improvements.
In view of the lack of technical consensus on the effectiveness of the proposed improvements, the staff decided to undertake additional efforts.
In July 1987, the Staff informed the Commission of its intention to examine the Mark I issue in the context of an integrated approach to the closure of severe accident issues.
On December 18, 1987, the staff issued a plan (SECY 87-297) for resolving generic severe accident containment performance issues for Mark I and other containment types. As part of the plan, a workshop was held on February 24-26, 1988 to discuss a number of issues associated with Mark I containment challenges, failure modes and potential containment improve-ments with researchers, industry representatives and interested members of the public. A major topic at the workshop was the phenomena associated with containment shelI melttbrough
t?
l4, m.
3 as discussed.in Enclosure 6.
The Integration Plan for Closure,of Severe Accident-Issues, (SECY 88-147) character-izes the containment: performance. improvement effort as'being one of the main elements of the integrated approach to-l closure of severe' accident issues. Other main' elements include a) Individual Plant Examinations (IPEs), b) improved plant operations, c) the severe accident research program, d) examination of external events, and e) a program on 1
accident management. The containment performance improvement program is related to the IPE effort, and is considered complementary to it,: since this effort is primarily focused on the potential generic vulnerabilities of specific containment classes, whereas the IPE effort is focused on plant unique vulnerabilities.
A Commission paper (SECY 88-206) dated' July 15, 1988-provided a status report on the staff's efforts regarding the Mark I containment. This paper reaffirmed that the risk from BWR Mark Is is' low. Nevertheless, the staff proposed a program intended to further reduce overall risk in BWR Mark I plants by pursuing a balanced approach
-involving accident prevention and mitigation. A number of safety enhancements were identified which appeared attractive.
in terms of their potential risk. reduction capability as well as implementation costs.
Following that meeting the Commission requested additional information-via a staff requirements memorandum dated August 1,.1988. Responses to these questions are included-as Enclosure 1.
Disassion:
Probabilistic Risk Assessment (PRA) studies for BWRs indicate that accidents initiated by transients rather than toss-Of-Coolant-Accidents (LOCAs) dominate the total core damage f.equency estimates. The principal accident sequences for BWRs consist'of Long-term Loss of Decay Heat-Removal (TW),StationBlackout(SBO),andAnticipated Transient Without Scram (ATWS). WASH-1400. indicated that TW is the dominant core damage accident sequence for Peach Bottom. Draft HUREG-1150, huever, indicated that the dominant' contribution to core olt frequency at Peach Bottom is due.to Station Blackout, and estimated that TW has been greatly reduced at Peach Bottom by implementation of containment venting procedures with the assumption that said venting actions can be successfully accomplished. For those plants in which TW has been e iminated as the dominant contributor, the residual risk is largely due to ATWS and SB0 sequences. These studies also indicate that the estimated likelihood of core damaging accidents for existing Mark I plants is predicted to vary widely over two orders of magnitude or more. The primary containment challenges and potential failure modes for BWR Mark I containments are shown in Enclosure 2.
e 4
The staff has examined potential Mark I containment'and.
plant' improvements in the following six areas: '(1) hydrogen control, (2) alternate water supply for reactor vessel-
. injection or containment drywell sprays (3) containment pressure relief capability (venting!, (4) enhanced RPV-depressurization system reliability, (5) core debric controls, and (6) procedures and trrining. Each of these was evaluated to determine their potential benefits in terms of reducing the (1) core melt frequency (2) containment failure probability, and (3) offsite consequences.
Hydrogen' Control:
Although BWR Mark' Is are required to be operated with an inerted containment atmosphere, plant Technical Specif t-cations permit de-inerting to commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to plant shutdown, and do not require inerting to'be completed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after plant-startup, in order to permit plant personnel access.
In the event of a severe accident, such as a long-term station blackout, a concern was expressed that loss of control of the valves and containment leakage could eventually lead to containment-de-inerting.
Two potential improvements with regard to hydrogen' control were evaluated. These were: (1) elimination of the two 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> de-inerted periods and (2) providing a backup supply of nitrogen. Since the probability of a severe-accident occurring during either of the two 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> de-inerted' periods is small compared to the probability _of accident occurrence during normal operations, eliminating this time of de-inerting would not significantly reduce risk.
During a severe accident, reactor pressure is anticipated to increase, releasing steam and non-condensable gases into the containment. This will increase containment pressure, preventing ingress of air. Therefore, the containment atmosphere would not become de-inerted for an extended period of time. Since offsite supplies of nitrogen could readily be obtained during this period, an onsite backup j.
supply of nitrogen would not significantly reduce risk.
Therefore, the staff concludes that additional Mark I improvements to control hydi e n beyond the existing hydrogen control rule and the prucedures in Revision 4 of the Emergency Procedure Guidelines would have no significant l
benefit and are not warranted.
1.
Alternate Water Supply for Drywell Spray / Vessel Injection An important proposed improvement would be to employ a backup or alternate supply of water and a pumping capability that is independent of normal and emergency AC
4{
.. l m;
power.
By connecting this source to-the low pressure L residual heat removal (RHR) system as well as to the-existing drywell sprays, water.could be delivered eith" into the reactor vessel or to the drywell, by use of an appropriate valving arrangement.
An alternate source of water injection into the' reactor vessel would. greatly reduce the likelihood of' core melt due to station blackout or ' u s'of long-term decay heat removal, as well as provide significant accident management capability.
Water for.the drywell sprays would also provide significant mitigative. capability to' cool core debris, to cool the containment liner to delay or prevent failure', and'to scrub air borne particulate fission products from the atmosphere.
A review of some BWR Mark I facilities indicates that most plants have one or more diesel driven pumps-which could be used to provide an alternate water supply. The flow' rate' using this backup water' system may be significantly less than the design flow rate for the drywell sprays. The potential benefits of modifying the spray. headers to assure a spray were compared to having the water run out of the spray nozzles.. Fission product removal in the small crowded volume in which the sprays would be effective was judged to be small compared to the benefit of.having a.
water pool on top of the'corium. Therefore, modifications to.the spray nozzles are not considered warranted.
Containment pressure Relief Capability (Venting):
Venting of-the containment is currently inc~luded.in BWR emergency operating procedures. The vent path external to existing containment penetrations typically consists of a ductwork system which has a low design pressure of only a few psi. Venting under high pressure severe accident conditions would fail the ductwork, release the containment atmosphere into the reactor building, and potentially contaminate or damage equipment needed for accident recovery.
In addition, with the existing hardware and procedures at some plants, it may not be possible to open or to close the vent valves for some severe accident scenarios. The staff has concluded that venting, if properly implemented, can have a significant benefit on plant risk. However, venting via a sheet metal ductwork path, as currently implemented at some Mark I plants, is likely to greatly hamper or complicate post-accident recovery activities, and is therefore viewed by the staff
{
as yieloing reduced improvements in safety. The capability to vent has long been recognized as important in reducing risk from operation of BWR Mark I facilities for loss of long term decay heat removal events. Controlled venting can prevent the failure of ECCS pumps from inadequate NPSH l
6 and re-closure of the ADS valves. The staff agrees with this view as long as the potential downsides of using the existing hardware are corrected.
A hard pipe vent capable of withstanding the anticipated severe accident pressure loadings would eliminate these disadvantages. The vent isolation valves should also be remotely operable from the control room and should be provided with a power supply independent of normal or emergency AC power. Other changes, such as raising the vent valve operability pressure and/or raising the RCIC turbine back pressure trip setpoint, may also be desirable and should be considered as part of the IPE. 1his capa-bility, in conjunction with proper operating procedures and other improvements discussed in this paper, would result in greatly reducing the probability of core melt due to the TW and SB0 sequences.
Given a core melt accident, venting of the wetwell would provide a scrubbed venting path to reduce releases of particulate fission products to the environment. Venting has been estimated to reduce the likelihood of late containment over-pressure failure and to reduce offsite consequences for severe accident scenarios in which the containment shell does not fail for other reasons.
Failure of the shell due to corium attack (shell meltthrough) would reduce the benefits from venting in that it would release fission products directly into the reactor building.
Inadvertent venting could result in the release of normai coolant radioactivity to the environment even when core degradation is averted or vessel integrity maintained.
Measures to reduce the probability of inadvertent venting, such as a rupture disk, should be considered in the vent design.
Enhanced Reactor Pressure Vessel (RPV) Depressurization I
System Reliability:
i The Automatic Depressurization System (ADS) consists of I
relief valves which can be operated to depressurize the reactor coolant system. Actuation of the ADS valves I
requires DC power.
In an extended station blackout after -
station batteries have been depleted, the ADS would not be available and the reactor would re-pressurize. With enhanced RPV depressurization system reliability, l
depressurization of the reactor coolant system would have a greater degree of assurance. Together with a low pressure alternate :;ource of water injection into the reactor vessel, the major benefit of enhanced RPV depressurizotion l
reliability would be to provide an additional source of I
core cooling which could significantly reduce the i
I likelihood of high pressure severe accidents, such as from the short-term station blackout.
)
i
'4 7
Another important benefit is in.the area of accident
. mitigation. Reduced reactor pressure would greatly reduce-the possibility of core debris being expelled under' high pressure,-given a core melt and failure of the reactor pressure vessel. Enhanced RPV depressurization system l
reliability would also delay containm'ent failure and reduce the quantity and type of fission prod ets ultimately
~
released to the environment.
In ord :r to increase reliability of the RPV depressuriza$jon system, assurance l
. of electrical power beyond the requirements of existing regulations may be necessary as discussed later in this paper.
In addition, performance of the cables needs to be l.
reviewed for temperature capability during a severe accident.
Core Debris' Controls:
Core debris controls, !n the form of curbs in the crywell and/or curbs or weir walls in the torus room'under the wetwell have been proposed in the past to prevent containment shell meltthrough and to retain sufficient water to permit fission product scrubbing.
However, as noted in SECY 88+206, the technical feasibility for such controls has not been established, and the design and installation costs as well as the occupational exposure during installation could be significant. The staff intends to pursue research programs to evaluate the need for and feasibility of core debris controls. There is a growing consensus that water in the containment (from an alternate supply to the drywell sprays) may help mitigate risk either by. fission product scrubbing or by preventing or delaying shell melt by core debris.' Research is continuing in order to confirm and help quantify these initial conclusions.
A discussion of Mark I shell melt phenomena and the current state of knowledge is included in Enclosure 6.
Emergency Procedures and Training:
A major element of the Mark I containment performance -
improvement evaluation involves emergency procedures and training. Current emergency operating procedures (EOPs) are symptom-based procedures that originated from require-ments of TMI Task Action Plan item I.C.1.
Plant-specific lemented based on generic Emergency E0Ps are generally imp (EPGs) developed by the BWR Owners Procedure Guidelines Group. As part of the balanced approach to examining potential BWR Mark I plant improvements, both the generic EPGs and the plant-specific implementation of E0Ps and training have been examined.
l l.
l
8 MC has recently reviewed and approved Revision 4 of the BWR Owners Group EPGs (General Electric Topical Report NED0-31331, BWR Cwner's Group " Emergency Procedure Guidelines, Revision 4," March 1987). Revision 4 to the BWR Owners Group EPG is a significant improvement over earlier versions in that they continue to be based on symptoms, they have been simplified, and all open items from previous versions have been closed. The BWR EPGs extend well beyond the design bases and include many actions appropriate for severe accident management.
The improvement to EPGs is only as good as the plant-specific E0P implementation and the training that operators receive on use of the improved procedures. A recent staff safety evaluation report (Ltr. Thadani to Grace, " Safety Evaluation of 'BWR Owners' Group - Emergency Procedure Guidelines, Revision 4,' NED0-31331, March 1987," dated September 12,1988) encouraged licensees to implement Revision 4 of the EPGs and reiterated the need for proper implementation and training of operators.
Implementation of the guidelines has been voluntary, but is strongly recommended in the SER.
Impact of Existing Requirements:
As part of the balanced approach, for completeness, and to provide a more accurate picture of Mark I plant risk, the staff nas also evaluated the impact on Mark I risk of several recent rules that have been imposed on light water reactors - the Station Blackout Rule and the ATWS Rule. As discussed earlier, PRAs typically indicate that Mark I reactor risks are dominated by TW, SB0 and ATWS sequences.
Upon implementation of these two rules at all Mark I plants, risk from SB0 and ATWS sequences would be expected to be reduced to a low level. The response to Question #2 in Enclosure 1 provides a discussion of expected rO:k reductions from changes to Mark I plants as a result of these rules.
J Assuring the operability of the proposed improvements under severe ccident conditions, including an extended period of station blackout, may require assurance of electrical power beyond the requirements of the recent Station Blackout
]
(SBO) rule, 10 CFR 50.63. The proposed improvements have I
been coordinated with the requirements of this rule in order not to cause an undue proliferation of power supplies, which could be counter-productive to safety. The staff proposes that licensees intending to implement the i
SB0 rule by use of an alternate AC (AAC) source, need not j
provide additional electric power supplies for the proposed j
Mark I improvements, provided that the capacity of the AAC l
is sufficient for the requirements of both the SB0 rule and j
the improvements proposed here. Further details are given in Enclosures 7 and 8.
t
9 Benefit of Improvements:
The improvements that the staff is recommending) include:(2? impro (1)-an improved hardened venting capability RPV depressurization systems reliability, (3) an l.
alternative ~ water supply to the reactor vessel and drywell c
sprays, and (4) emergency procedures and training.
Accelerated implementation of the existing station blackout and ATWS_ rules is also planned. These improvements are unchan ed from those indicatec in the interim report (SECY 88-206 to the Comission.
A major benefit of these improvements.is that they can provide a reduction in core melt frequency of about a factor of five.to ten. With the proposed enhancements, the coremeltfrequegcywouldbeexpectedtobereducedto-about I to 2x10 per reactor-year.
It should be noted that these. estimates apply to internal events only.
~
For plants with a high TW probability, a large. fraction of the reduction in core melt frequency is attributable to improved venting which, by allowing the removal of long-term decay heat from the containment, greatly reduces the likelihood of core melt 'from the TW sequence. Another reduction in core melt frequency from station blackout is attributable to the enhancements taken~ together.
In the-event of station blackout, enhanced RPV depressurization l.
s reliability would permit depressurization of the reactor, availability of a low pressure backup source of water injection into the vessel would permit core cooling, while venting would allow decay heat removal from the con-tainment.
It is important to note that under these circum-stances, venting would prevent core damage and not result in releases of fission. products 'of any. significance.
Accident mitigation benefits are also considered to be significant. Mitigation of fission product releases would be realized for all accident sequences, including ATWS.
Venting would be effective in preventing containment failure arising from slow over-pressurization. Venting via the suppression pool would provide significant scrubbing of non-noble gas fission products by about a factor of 10 to 100 if no containment shell failure occurs. Water in the-
?
drywell may be effective in preventing or at least delaying l
failure of the shell by molten core debris. Finally, even l
if shell failure should occur, the presence of a water L
layer atop the core debris combined with the drywell spray.
l-would reduce any source term releases to the environment by l-a factor judged to range from 2 to 10.
Because of the combination of reduced core melt likelihood, reduced fission product releases due to mitigation, and possible reduction or elimination of a significant contain-ment failure mode, the staff concludes that the overall f
y
- 79 c
+
_ o m-W o-'
10
-1~
risk reduction of.the proposed improvements.is'in-excess of:
one order-of magnitude.
The benefits of the proposed enhancements in terms of their reduction in offsite risk can be calculated in terms of.
_ person-rem..: Depending upon'the. probability of__ core. melt-due to the TW sequence the estimated' reduction-in risk,-
expressed in. person-rem, for the proposed enhancements.
- ranged. from about-145 person-rem per reactor-year.to about
'1330 person-rem per reactor-year, for plantg having a probability of core melt due to TW of 1x10- per reactor--
year and 1x10-4 per reactor-year, respectively. Of this total value, the risk reduction produced by. lowering' the likelihood of core melt due to. station blackout and raitigation.of ATWS accounts for a reduction'of about 33 te
+
x 210 person-rem per reactor-year.
For plants whose probability of core' melt due to the TW sequence.is highj (about 10- per reactor-year), the bulk of the risk rediuction'can be attributed to the large reduction in.the TW sequence brought about by.. improved venting. Additional details are provided in Enclosure 4.
Finally,'as'noted earlier, the recommended improvements' form a package in the sense that'they complement one another 'in prevention or mitigation. _ This results in the maximum risk reduction when all Lare taken together.
Summary of Costs of Improvements:
Cost estimates were made of the proposed. improvements.
These are given in Enclosure 3 which provides a summary for all improvements that include's'high and low.' estimates ~
ranging from $3.1 to $1.6 million dollars.. For purposes of the. regulatory analysis included in
Enclosure:
4, a'best estimate cost of $2.0M has been used.
Estimates of cost as high as $7.3M were obtained based on actual costs of-similar improvements at an existing Mark I. plant. Actual costs at many plants may be less since, as shown in
~, some plants already have many features of the r
l l-proposed improvements.
==
Conclusions:==
Many of the proposed enhancements would require plant E
backfits. The staft has examined these iii light of the backfit rule, 10 CFR 50.109.
Section (a) 3 of that i
regulation indicates that the Commission shall require backfittire,, only when "there is a substantial increase in the overall protection of the public health and safety" and "that the direct and indirect costs... are justified in view of this increased protection".
In reaching a conclusion with respect to the first test indicated above, the staff considered the effect of the proposed enhancements upon reductions in core melt frequency and improved containment performance. A major n
n
~
11' 1
benefit of these enhancements is in their ability to reduce the likelihood of core melt. Core melt frequencies'for BWR
.i Mark I plants prior to.any of'the enhancements considpred W
would be expected to range from about 1x10- to 2x10- per reactor year.. With the combined ~ enhancements, core melt' frequency. would be reduced by about a factor of five to ten..Thus, the proposed enhancements clearly offer.a substantial reduction in core melt frequency.. The core melt frequency reductions do not give credit for existing
. venting capability assumed in NUREG-1150 since the current' venting capability at plants has significant uncertainty regardirg its overall effectiveness.
The im es ;ed ability to cool core debris and to' remove' excess st from the containment by venting, given the
~
c occurrence of an accident, is also expected to reduce ther likelihood of containment failure, although this is not as readily quantifiable because'of the uncertainty in core-melt progression and shell'meltthrough phenomenology which is discussed in Enclosure 6.
In addition, the ability to scrub particulate f bsion products by use of venting through the suppression pool and.by the use of a water layer atop any core debris also adds significant mitigative capability.
Since the proposed enhancements would be expected to reduce the likelihood of core melt by about.a factor of five'to r
tre, and provide significant additional accident mitigation capoility as well, the staff concludes that the proposed enhancements do provide a substantial increase in the overall protection of.the public health-and safety.
With regard to the second or cost-benefit test required by-the backfit rule,.the discussion given earlier has shown that the costs of the enhancements are estimated to ranm-from 1.6 to 3.1 million dollars per plant, although sic
.r improvements at an existing Mark I plant may have cost about 7.3 million dollars. Both the estimated cost and the cost associated with an existing Mark I plant were used in the cost-benefit analysis. Based on the survey results for nine Mark I plants, the staff believes that many plants i
have some of these improvements already in place. Since L
the estimated benefits ranged from 3.6 to 33 million l--
dollars per reactor based Lpon 1000 dollars per person-rem and an average remaining plant life of 25 years for Mark I 1
plants, the staff concludes that the proposed enhancements-are generally cost beneficial.
For the reasons stated above, the staff concludes that backfit of these proposed enhancements is warranted for all Mark I plants.
Options:
1.
Take no action.
Pro:
No further resources W uld be l
1
]
12 required. Con: This option would result in a situation I
where a number of enhancements to safety that the staff believes to be cost effective would not be implemented and closure of severe accident issues would not be obtained for Mark I plants.
2.
Issue a generic letter. Pro: This option would be the quickest way to inform licensees of the staff's views and would require the least resources. Con: The generic letter can inform industry of the staff's finding, but can only request, not require, licensees to make chcoges to their facilities.
3.
Issue an order. Pro: This option could ce accom-plished quickly and provide a regulatory requirement to implement the improvements. Con: This option could result in requests for hearings from both licensees and interveners contesting the orders. A draft proposed order is included as Enclosure 7.
4 Initiate Rulemaking. Pro: This option would provide a regulatory basis for requiring the improvements.
It is generally preferable to impose generic require-ments by rule. Con: This option would require some staff resources and cause a delay in implementing the proposed improvements. A draft proposed rule is attached as Enclosure 8.
Recommendations:
The proposed improvements could be implemented as a regulatory requirement either by use of orders or through rulemaking. Of these two viable options, although the staff considers that it is generally preferable to impose generic requirements by rule, the improvements could be carried out more quickly via orders and for this reason the staff recommends that orders be issued to require the improvements. The staff would also prepare an Environmental l
Assessment of venting of the containment using the improved hardware and procedures.
Coordination:
0GC has no legal objections. The ACRS has reviewed these recommendations and will provide their comments separately.
I Victor Stello, Jr.
Executive Director for Operations
Enclosures:
- 1. Response to Commission Questions
- 2. Mark 1 Challenges and Relative Likelihood of Failure Modes
- 3. Summary of Costs
- 4. Regulatory Analysis
- 5. Results of Survey of Mark I Plants
- 6. Mark I Liner Melt Status
- 7. Draft Proposed Order Et moco ermrwmm FMo
GNr.ksar3')
J In the Matter of
)
)
)
License No.
)
Docket No.
)
)
ORDER MODIFYING LICENSE I.
(Name of Licensee), (Licensee) is the holder of Operating License No.
issued by the Nuclear Regulatory commission (NRC/ Commission) on The license authorizes the licensee to operate (Name of Facility). The facility is a Boiling Water Reactor (BWR) located at the Licensee's site in which utilizes a Mark I containment.
II.
Probabilistic Risk Assessment (PRA) studies have been performed for a number of Boiling Water Reactors with Mark I containments. These studies indicate that, although the risk from the BWR Mark I is low, containment integrity could be challenged if a large scale core melt accident were to occur, principally due to the smaller size of the containment. The studies which have been performed indicate that BWR Mark I plant risks are dominated by 1.oss of Decay Heat Removal (TW), Station Blackout (S80) and Anticipated Transient WithoutScram(ATWS) sequences. The staff has concluded that the optimum way to reduce overall risk in BWR Mark I plants is to pursue a balanced approach utilizing accident prevention and mitigation.
Based on this assessment, l
potential improvements have been identified in the following specific areas
-2 which, when implemented, will substantially enhance the overall safety of Mark I plants:
(1) Containment Pressure Relief Capability (Venting).
The capability to vent has long been recognized as important in reducing risk from operation of BWR Mark I facilities for loss of long term decay heat removal events.
However, the vent path external to most of the existing containment penetrations typically consists of a ductwork system which has a low design pressure of only a few pounds per square inch (psi.) Venting under high pressure severe accident conditions would fail-the ductwork, release the containment atmosphere into the reactor building, and potentially contaminate or damage equipment needed for accident recovery.
/'arthermore, with the existing hardware and procedures at some plants, it may not be possible to open or close the vent valves for sone severe accident scenarios. A har.d pipe vent capable of withstanding the anticipated severe accident pressure loadings would eliminate or minimize the consequences of these disadvantages.
Other cnanges, in conjunction with proper operating procedures, would result in greatly reducing the probability of core melt due to the Loss of Decay of Heat Removal (TW) and Station Blackout (5B0) sequences.
i 1
p
'4 '.
-. (2) -Reactor Pressure Vessel Depressurization System Reliability.
The reactor pressure vessel (RPV) depressurization system consists of relief valves which can be operated to depressurize the reactor coolant sy stem. Actuation of these valves requires DC p,ower.
In an extended station blackout af ter station batteries have been depleted, the RPV.
depressurization system valves would not be available and the reactor would re-pressurize. With enhanced RPV depressurization system reliability, depressurization of the reactor coolant system would have a greater degree of assurance. Together with a low pressure alternate source of water injection into the reactor vessel, the major benefit of enhanced RPV depressurization reliability would be to provide an additional source of core cooling which could significantly reduce the likelihood of high pressure severe accidents, such as from the short-term station blackout.
Another important benefit is in the area of accident mitigation.
Reduced reactor pressure would greatly reduce the possibility of core debris being expelled under high pressure, given a core melt and failure of the reactor pressure vessel,.
Use of the RPV depressurization would also delay containment failure and reduce the quantity and type of fission products ultimately released to the environment.
In order to i erease reliability of the RPY depressurization system assurance of electrical power beyond the requirements of existing regulations may be necessary.
In addition, performance of the depressurization system valves needs to be reviewed for temperature capability during a severe accident.
_ a
_ (3) Alternate Water Supply for Drywell Spray / Vessel Injection.
An important proposed improvement would be to employ a backup or alternate supply of water and a pumping capability that is independent of normal and emergency AC power.
By connecting this source to the low pressure residual heat removal (RHR) systems as well as to the existing drywell sprays, water could be delivered either into the reactor vessel or to the drywell, by use of an appropriate valving arrangement.
An alternate source of water injection into the reactor vessel would greatly reduce the likelihood of core melt due to station blackout or loss of long-term decay heat removal, as well as provide significant accident management capability.
Water for the drywell sprays would also provide significant mitigative capability to cool core debris, to cool the containment liner to delay or prevent failure, and to scrub airborne particulate fission products from the atmosphere.
(4) Emergency Procedures and Training.
l A major element of the Mark I containment performance improvement j
evaluation involves emergency procedures and training.
Proper operator actions can preclude milder events from progressing to core damage or core meltdow accidents, and can greatly mitigate the consequences of
e u even severe accidents. Since a variety of unusual conditions can be present particularly for beyond design bases events, emergency training :
and symptom based procedures are ' essential for guiding the operation to those. actions which provide the greatest measure of protection to the public.
NRC has recently reviewed and approved Revision 4 of the BWR Owners Group EPGs (General Electric Topical Report NED0-31331, BWR Owner's Group
" Emergency Procedure Guidelines, Revision 4," March 1987).
Revision 4 to the BWR Dwners Group EPG is a significant improvement over earlier versions in that they continue to be based on symptoms, they have been simplified, and all open items from previous versions have been closed.
The BWR EPGs extend well beyond the design bases and include many actions appropriate for severe accident management.
Since operator actions affect the risk for all severe accident scenarios, implementation of procedures based upon Revision 4 to the EPGs is important for maximizing overall risk reduction.
III.
Improvements in the above mentioned areas can provide a reduction in core melt frequency of about a factor of five to ten. Furthermore, accident mitigation benefits are also considered to be significant. Mitigation of fission product releases would be realized for all acci~ dent sequences, including ATWS.
Venting would be effective in preventing containment failure arising from slow
L, over-pressurization. Venting via the suppression pool would provide m
significant scrubbing of n'on-noble gas fission products by about a factor of 10 to 100 if no containment shell failure occurs. Water in the drywell may be effective in preventing or at least. delaying failure of the shell by molten l
core debris. Finally, even.if shell failure should occur,. the presence of a water layer atop the core debris combin d with the drywell spray would reduce P
any source. term releases to the environment by a factor ' judged to range from 2 to 10. _ In sum, improvements in these areas would result in reduced core melt' L
likelihood, reduced fission product releases due to mitigation, and possible reduction or elimination of a significant containment failure mode, and provide a substantial increase in the overall protection of the public health and safety. [ STAFF: ADDRESS i 50.109]
IV.
Accordingly, in view of the foregoing, and pursuant to sections 103,161b.,1611.,1610. and 182 of the Atomic' Energy Act of 1954, as amended, and the Commission's regulations in 10 C.F.R. I 2.204 and 10 C.F.R. Part 50, it is hereby ordered that the Licensee A.
Within 30 months of the date of this Order:
1.
Provide its BWR Mark I containment with an exhaust line'from the wetwell vapor space to a suitable release point (e.g., plant stack).
The basic design objective shall be to provide sufficient venting capacity to
l^
r 7-prevent long-term overpressure failure of containment. This "hard vent" system shall meet the following criteria:
a); The vent shall be sized such that under conditions of 1) constant heat input at a rate equal to 1% of rated thermal power, and 2)containmentpressureequaltotheventsetpointpressure,the exhaust flow through the vent is sufficient to prevent the containment pressure from increasing, b)
The venting setpoint shall be set at or above containment design pressure.
Capability of RPV depressurization system valves, torus vent valves, or other equipment should not limit the venting setpoint to less than containment design pressure.
c)
The venting capability shall be available during severe accident conditions and for a period up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of a station blackout.
d)
The hardened vent path should include a means to prevent premature or inadvertent actuation.
e)
The vent path up to and including the second containment isolation barrier should be designated safety Class 2.
. f)~
The hard vent path shall be capable of withstanding,'without loss of functional capability, experted venting conditions and the effects of. potential combustion phenomena.
g)
The hardened vent path shall have a radiation monitor, alarmed in control room and functional during extended station blackout.
2.
Examine the reactor pressure vessel (RPV) depressurization system and make modifications to ensure its functional capability during severe accidents and during extended station blackout conditions. As a minimum, the following shall be provided:
- 1) the capability of the RPV depressurization system cables and components to withstand, without loss of functional capability, the environment in the containment during a severe accident prior to vessel failure, and 2) an alternate power supply system capable of opening and maintaining open at least one RPV depressurization system valve for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of station blackout. Any sources of electrical power required to assure the operability of the backup water supply, containment venting system, and RPV depressurization system during an extended station blackout should be coordinated with the requirem!nts of 10 C.F.R[50.63, as follows:
a)
Those licensees who chcose to implement the requirements of 10 C.F.R. 5 50.63 by the use of an alternate AC (AAC) source need not provide any additional power supplies to comply with the provisions of this section, provided that the capacity, capability, and l
^l i
I 9..
duration of the AAC can be shown to meet the requirements of both 10
-C.F.R. 9 50.63 and this section b)
Those licensees who choose'to implement the requirements of 10 C.F.R. 5 50.63 solely by means of a coping analysis must provide additional power supplies of sufficient capacity and reliability to assure the operability of the backup water supply, containment venting system and RPV depressurization systems during an extended station blackout.
[ STAFF:
FIX ABOVE TO BE SPECIFIC FOR THE LICENSEE TO WHOM THIS IS ISSUED]
3.
Provide at least one water supply system for the containment drywell spray which shall be functional during an extended station blackout. An extended station blackout is defined as loss of all normal and emergency
'AC power and loss of DC power due to depletion of station batteries.
Operability of controls and valves during such an event may require an independent source of power such as a dedicated battery set or a means to recharge the station batteries. Water to the spray system from this supply shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during severe accident conditions.
e l
= The water supply system shall also be capable of delivering water to the reactor vessel once the vessel has been depressurized. The mass flow rate shall'be equal to or greater than the boiling rate which would occur under depressurized, saturated conditions with a constant heat input rate equal to 1% of rated thermal power.
All valve realignmr.nts or other actions necessary to realize this capability shall be reasonably achievable during an extended station blackout.
Instrumentation needed to realize these capabilities shall be functional in the expected accident conditions and should, as a minimum, include [tobedetermined).
4.
Implement procedures based on Emergency Procedure Guidelines (EPGs) developed by the BWR Owners' Group.
Initially, Revision 4 to the EPGs 1/
as modified by the staff safety evaluation 2) should be used as the basis for the procedures.
Subsequent revisions to the EPGs as developed by the BWR Owners Group (or equivalent) should be used to update the procedures in a timely fashion.
1/
BWR Owners' Group Emergency Procedure Guidelines, Revision 4, NED0-31331,
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March,'1987.
-2/
Letter from A. Thadani to D. Grace " Safety Evaluatic.n at BWR Owners' Group Emergency Procedure Guidelines" dated 9/12/88.
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B..
Within 60' days, submit to the Director of Nuclear Reactor Regulation-plans for implementation of the above improvements and a schedule for implementation.
The licensee r r any person adversely affected by this10rder may request a hearing'within 30 days of the date of this Order. A request for hearing should be clearly marked as a " Request for Hearing" and shall be addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555, with copies to the Assistant General Counsel for Enforcement at the same address, the Regional Administrator, Region
, and the NRC Resident Inspector, at (Plants affected).
If a person other than the licensee requests a hearing, that person shall set forth with particularity the manner in which the petitioner'.s interest is adversely affected by this Order and should address the criteria set forth in 10 C.F.R. 5 2.714(d).
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If a hearing.is requested, the Commission will issue an Order designating the time and place of the hearing.
If a hearing is held, the issue to be.
considered shall be whether this Order should be sustained. Upon the failure to answer or request a hearing within the specified time, this Order shall be final without further proceedings.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, Director Office of Nuclear Reactor Regulation Dated this, IIk_ y of da l
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DRAFT PROPOSED RULE 1
SECTION 5'0.XX -_ SEVERE ACCIDENT REQUIREMENTS 70R BOILING-WATER REACTORS HAVING
- MARK 1 CONTAINMENTS a)
Applicability The requirements of this section apply to all boiling water 5
reactors-(BWR) having Mark 1 containments.
b )'
Requirements
- 1.
Backup Water Suppy for Drywell Spray / Core injection Ali BWRs with a Mark 1 containment shall provide at least one water supply system for the. containment drywell spray which shall be functional during an extended station blackout.
Water to the spray system from this ' supply.
l shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during severe accident conditions.
The water supply system shall also be capable of. delivering water to the reactor vessel once the vessel has been depressurized. The mass flow rate shall be equal to or greater than the boiling rate which would occur under depressurized, saturated conditions with a constant heat input rate equal to 1% of rated thermal power. All valve realignments or other actions necessary to realize this capability shall be achievable during an extended station blackout.
i 1
An extended station blackout is defined as loss of all normal and emergency AC power and loss of DC power due to depletion of station batteries. Operability of controls and valves during such an event may require an independent source of power such as a dedicated battery set or i,
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'a means to recharge the station batteries.
L _-___- - -____-_-_-_ __--______ __ __-__________ = ______ _ _ __ _ __________________________________________ ________________
Instrumentation needed to realize this capability shall be functional in l
the expected accident conditions and should, as a minimum, include [to be determined].
2.
Containment Venting l
For BWR plants with a Mark I containment an exhaust line which is capable of withstanding expected venting conditions shall be provided from the wetwell vapor space to a suitable release point (e.g., plant stack). The basic design objective shall be to provide sufficient venting capacity to prevent long-term overpressure failure of containment. This "hard vent" system shall meet the following criteria:
a)
The vent shall be sized such that under conditions of 1) constant heat input at a rate equal to 1% of rated thermal power, and 2) containment pressure equal to the vent setpoint pressure, the exhaust flow through the vent is sufficient to prevent the containment pressure from increasing.
b)
The venting setpoint shall be set at or above containment design pressure.
Capability of RPV depressurization system valves, torus vent valves, or other equipment should not limit the venting setpoint to less than containment design pressure.
c)
The venting capability shall be available during severe accident conditions and for a period up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of a station blackout.
d)
The hardened vent path should include a means to prevent premature or l
inadvertent actuation, e)
The vent path up to and including the second containment isolation barrier should be designated safety Class 2.
f)
The harG vent path shall accommodate effects of potential combustion I
phenomena and remain functional.
2
g)
The hardened vent path shall have a radiation monitor, alarmed in control room and functional during extended station blackout.
3.
[eactorPressureVesselDepressurizationCapability All licensees having BWR Mark I containments shall examine the reactor pressure vessel (RPV) depressuriration system and make modifications to ensure its functional capability during severe accidents and during extended station blackout conditions. As a minimum, the following shall be provided:
- 1) the capability of the RPV depressurizatici. system cables and components to withstand, without loss of functional capability, the environment in the containment during a severe accident prior to vessel failure; and 2) an alternate power supply system capable of opening and maintaining open ? least one RPV depressurization system valve for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of station blackout. Coordination of this requirement with the Station Blackout Rule (10 CFR 50.63) is discussed in Section 4.C.2 below.
4.
Procedures and Training All BWRs with Mark I containments shall implement procedures based on Emergency Procedures Guidelines (EPGs) developed by the BWR Owner's Group.
2 Initially, Revision 4 to the EPG's as modified by the staff safety 3
evaluation should be used as the basis for the procedures. Subsequent revisions to the EPGs as developed by the BWR Owners Group (or equivalent) should be used to update the procedures in a timely fashion.
c)
Implementation 1)
Schedule 2
BWR Owner's Group Emergency Procedure Guidelines, Revision 4, NED0-31331, March, 1987.
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Letter from A. Thadani to D. Grace " Safety Evaluation of BWR Owners Group Emergency Procer,ure Guidelines" dated September 12, 1988.
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3
I All licensees to whom this section applies shall submit their plans and anticipated schedule within 60 days after a final rule is issued which identifies any actions taken and those needed to be taken to comply with the requirements of this section. The requirements of this section shall be fully implemented within 30 months after a final rule is issued.
2)
Co-ordination with requirements of the Station Blackout Rule (10 CFR 50.63)
Any sources of electrical power required to assure the operability of the backup water supply, conta;nment venting system, and RPV depressurization system during an extended station blackout, as required in part (b) above, should be coordinated with the requirements of 10 CFR 50.63, as follows:
a)
Those licensees who choose to implement the requirements of 10 CFR 50.63 by the use of an alternate AC (AAC) source need not provide any additional power supplies to comply with the provisions of this section, provided that the capacity, capability, and duration of the AAC can be shown to meet the requirements of both 10 CFR 50.63 and this section, b)
Those licensees who choose to implement the requirements of 10 CFR 50.63 solely by means of a coping analysis, must provide additional power supplies of sufficient capacity and reliability to assure the operability of the backup water supply, containment venting system and RPV depressurization systems during an extended station blackout.
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t L to;the Minutes of CRGR Meeting No.;155 Draft Final Rule on Standardization and Licensing Reform Topice
. S. Crockett (0GC) and J. Wilson'(RES) presented'for CRGR review'a draft final rule (Part 52) on standardization and licensing reform.
This package reflects
. resolution of public comments.
Background
' The package submitted for review by CRGR in this matter was transmitted by memorandum dated January 6, 1989, S. Crockett to E. Jordan.
The review p:ckage included the redrafted rule.
Conclusions / Recommendations
.As a result'of their review of this matter,. including discussions with the staff at this meeting, the Committee made the following recommendations:
1.
The rule specifies that an early site permit O valid for twenty years.
The rule should allow for shorter permit times:s.1ce many factors can change over a 20 year period.
2.
As presently written (see p. 23 & p. 29), the rule places emphasis on testing a full-size prototype prior to certification.
The rule should be neutral on this issue to permit testing and analysis in lieu of building a ful.1-size prototype, since this can be an acceptable method for certifying a design.
- 3.
The rule (p. 25) requires completion of.a design-specific PRA and inclusion of the PRA in the application for design certification.
A phrase should be added, such as " including an estimate of the uncertainties." PRA has too much uncertainty to compare to the safety goals without an inclusion of the uncertainty analysis.
4.
Under section 52.47, Conter.ts of Applications, delete ix(3) because it is redundant.
The Atomic Energy Act already provides for acquisition of this information.
5.
Under 52.47 (b)(2), the current language for prototype testing includes very specific conditions (i.e., normal, transient, and accident).
This should be revised to read "over a suitable range of conditions" to avoid excessive testing, such as, beyond the design basis.
6.
In section 52.63 (a)(2), clarify the language discussing design certification modifications by either switching the two sentences or modifying the words.
As presently stated the meaning is ambiguous.
4.g y
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.The Committee rer.ommended in favor of forwarding the draft final rule to the Commission.
The staff was requested to forward the Statement of Considerations to the Committee for review.
The Committee will identify. issues but will not meet again on this topic.
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