ML19224C501
| ML19224C501 | |
| Person / Time | |
|---|---|
| Issue date: | 03/24/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1611, NUDOCS 7907020465 | |
| Download: ML19224C501 (38) | |
Text
,
e_
,r j, m.
g
/f //
p.....
9' 4
. ;}
1 pp/zr//v DATE ISSUED:/n in i
3/24/79
.C/ L -
xL.;i iblyi MIfiUTES OF THE ACRS SUBC0i"ilTTEE MEETING ON ANTICIPATED TRAtiSIEriTS WITHOUT SCPH1 JANUARY 31, 1979 WASHINGTON, D.C.
On January 31, 1979, the ACRS Anticipated Transients Without Scram (ATWS)
Subcommittee held a meeting in Washington, D.C., to continue discussion of the NRC Staff position on ATWS as stated in Volume 3 of NUREG-0460,
" Anticipated Transients Without Scram for Light Water Reactors." The notice of the meeting appeared in the Federal Register, on Friday, January 19, 1979.
No request to submit oral or written statements were received from members of the public, and none were made at the meeting.
Attachment A is the meeting agenda. The attendees list is Attachment B.
A tentative meeting schedule is Attachment C to the minutes. Attachment 9 contains selected slides and handouts used at the meeting. A complete set of slides and handouts is attached to the office copy of these minutes.
OPEN SESSION (8:30 am - 4:20 pm) INTRODUCTION Dr. Kerr, Subcommittee Chairman, called the meeting to order at 8:30 a.m.
The Chairman explained the purpose of the meeting and the rules and procedures for conducting the meeting, pointing out that Dr. Thomas G.
McCreless was the Designated Federal Employee in attendance.
Dr. Kerr introduced Dr. Roger Mattson, Director of the Division of Systems Safety, to begin the day's presentations.
NRC STAFF _ RESPONSE TO ATWS SU3 COMMITTEE QUESTIONS - R. MATTSON, A. THADANI Dr. Mattson began the day's presentation by summarizing the NRC activities since the NRC Staff presentation at the January Committee meeting.
The following points were noted by Dr. Mattson:
790702 0 e//5 272 049
g,,
ATWS 1/ 31/ 79 s
(1) The fiRC Regulatory Requirements Review Committee completed its deliberatio.s on Volume 3 of NUREG-0460 and summarized its conclusions in a January 18, 1979 letter from Ed' Case to Lee Gossick (Attachment U-l).
(2) NRC will issue a request to the LWR vendors for generic analyses as indicated in Volume 3 of NUREG-0460. These requests are scheduled to issued in early February.
(3) At the Atomic Safety and Licensing Board meeting on Monticello, the Staff stated that Alternative 3 of Volume 3 of NUREG-0460 would be implemented for the Monticello plant.
In the interim the licensee will adopt procures that involve upgrading of noerator training to deal with an ATW5.
(4) The NRC Staff has issued a letter to the 11 operating BWRs without recirculation pump trip advising them that they must install RPT on a timely basis (within 2 years) utilizing either the Monticello or modified Hatch RPT designs.
In the time prior to installation, the licensee must update operator training for ATWS as noted above.
There were several Subcommittee questions concerning details of the NRC letter on implementation of BWR RPT.
Copies of the letter were provided to the Subcommittee and are attached to these minutes (Attachment D-2).
Mr. Thadani (NRC Staff) provided responses to four Subcommittee questions noted below.
272 050 Question 1:
" Outline ar.y significant differences that exist between analysis (and results thereof) of the course of the various postulated ATWS events carripd out by the NRC Staff and by the vendors."
Mr. Thadant said there was
ATWS 1/31/79 1
a two part response to this question.
The first part of the response was a discussion of the analytical methods used by the vendors, and the second part of the response was a discussion of expected plant behavior using the analytical models. Mr. Thadant said that there are no significant differences in the results between NRC and PWR vendor analyses. Regarding the BWRs, NRC believes there is no significant differences between GE and NRC ATWS analysis methods. NRC is interested in obtaining confirmatory data foi chat part of the model which describes two-phase flow through PWR pressure relief valves.
In lieu of these data, they have recommended a conservative model for vendor use.
Mr. Thadani discussed the results of the ATWS analyses using event-tree diagrams to illustrate the outcome (Attachments D-3&4).
The PWR event-tree begins with an ATWS, followed by automatic actuation circuitry function, auxiliary feedwater function, power operated relief valves, and finally long-term cooling. Mr. Thadani noted that in an ATWS event involving loss-of-heat sink a considerable amount of auxiliary feedwater is necessary for successful recovery.
The analyses reviewed by NRC assumes some amount of auxiliary feedwater. Mr. Thadani replied that this is
~
assumed to be the case if the sional for auxiliary feed-waster is derived from the reactor protection system.
The Staff intends to require that the auxiliary feedwater system is derived from the reactor protection system. The Staff intends to require that the auxiliary feedwater system actuation function be independent of the scram signal logic.
Mr. Thadani said that, assuming half of the auxiliary feedwater system functions, Westinghouse reactors will not see exces-sively high pressures. All PWR plants require operation of 272 051
ATWS 1/31/79 the power operated relief valves (PORV) to avoid.
high pressures that might disable long-term cooling capability, (however W plants can lose one PORV without adverse consequences).
It was noted that loss of long-term cooling capability can lead to core melt.
However because of the time involved, this is unlikely.
Dr. Mark asked if the differences be-tween the NRC and vendor event-tree analyses are the probability of the events,- not the outcome.
NRC replied in the affirmative.
In response to a question from Dr. Lee concerning the possibility of exceeding Service Level C pressure limits for Alternatives 2&3 in NUREG-0460, Dr. Mattson replied that Alternatives 2&3 contain the same hardware requirements; however, Alternative 3 requires a demonstration that the relieving capacity assumed to exist under Alternative 2 does indeed exist.
Mr. Thadani rt f ewed the ATWS event-tree for BWRs (AttachmentD-4).
This event-tree begins with an ATWS followed by recirculation trip pump, poison injection, actuation of high pressure coolant injection (or core spray), followed by long-term cooling.
Mr. Thadani noted that both NRC and GE agree that if an ATWS occurs and RPT fails, then core melt is likely.
NRC believes that rapid poison injection is necessary to assure enough high pressure makeup water and to maintain containment integrity, which NRC believes is necessary.
It was Mr. Thadani's opinion that one of the central contentions between NRC and GE is the assurance of the necessary high-pressure make up water.
272 052
.:,. u...
... a.
ATWS 1/31/79 At Dr. Mattson's suggestion, Mr. Strcupe of General Electric discussed the central points concerning their difference of opinion with the f4RC on the ATWS transient scenario.
General
~
Electric believes that during the course of an ATWS, the operator will take the appropriate action to assure a source of high pressure make-up water.
GE calculates that approximately 20% of the rods will see CHF, but that does not imply fuel failure.
As a ;esult of further discussion, Mr. Stroupe noted that mother major point of disagreement between GE and the f1RC is the belief on GE's part-that suppression 7oo1 integrity is not required to prevent core melt, although GE believes suooression pool integrity would be maintained thr'aughout the event.
There was an extensive question period.
Dr. Kerr asked if the operator would quickly recognize an ATWS event given the possibility of contradictory instrumen-tation signals and/or failure of some instrumentation.
Dr. Kerr also questioned whether prompt operator action to initiate standby liquid control would be taken, given the cost of clean up for spurious poison injection. Mr. Epler also questioned the clean up costs estimated by GE for poison injection.
Mr. Stroupe noted that CE's high cost.estimato ($60 million/ event) will depend on such parameters as the plant rad-waste system design, and whether or not the standby liquid control system is automated. Dr. Mattson noted that flRC is considering allowing GE to install time delay actuation circuitry on the standby liquid control system to minimize the possibility of inadvertant actuation. Mr. Stroupe observed that GE believes installation of recirculation pump trip and possibly alternate rod injection with upgraded operator training is sufficient to n,n r
u._
ATWS 1/ 31/ 79 address the ATFC concern for BWRs. Mr. Thadani noted that assuming no sMigle failures in an 80 - 36 gpm capacity standby liquid control system, the suppression pool temperature would stabilize around 200 F, proviced the system is actuated within 2 minutes.
Dr. Lee requested information on the reliability of the BWR rod position indicators.
Mr. Stroupe said he would provide this information at the next ATWS Subconnittee meeting.
Dr. Mattson raised the question of reliance upon instrumentation that may not function in an ATWS environment.
He said such instrumentation failure could give the operator the problem of deciding whether the instrumentation has failed, or whether a serious accident has occurred.
The Subcommittee and NRC agreed that Question 3 (" Discuss the most probable accident scenarios for linking ATWS to a core melt.") was covered in the extensive discussion of Question 1 above. The Subcommittee proceeded to discussion of Question 2.
Question 2:
" Explain the reasoning that leads the Staff to conclude that the prescriptions of Alternative 4 will lead to lower risk than those of Alternative 3 for new reactors.
How much lower does the Staff exp.at the risk will be for each vendor's reactor?" Mr. Thadani noted that Alternative 3 is a mix of mitigation and prevention while Alternative 4 incorporates high mitigation capability for an ATWS. Mr. Thadani noted that fnr Westinghouse there is no real difference between the two Alternatives.
For B&W and CE reactors Alternative 4 mitigates all ATWS events; Alternative 3 provides protection for most ATWS events but may result 272 054
...........u.:.......
.c. c...
ATWS 1/ 31/ 79 in exceeding Service Level C p" essure limits by a small amount for some components.
NRC estimates that for new B&W and CE reactors, there is a factor of 40 difference in protection between the Alternatives.
In response to questions from Dr. Lipinski, Mr. Thadani noted that the principal factor account-ing for these differences is the degree of mitigation capability available for each of the two Alternatives.
For BWRs, most alt 4S's are mitigated by both Alternatives.
Since Alternative 4 provides more reliable mitigating systems than Alternative 3, there is a factor of 10 in the unreliability associated with use of Alternative 4 as compared to Alternative 3.
Dr. Lipinski requested documentation of the probability numbers quoted in Appendix F (Risk Ar.alysis) of Volume 3 of NUREG-C460.
Dr. Kerr noted that Dr. Okrent had requested the same information at the January Cornittee meeting.
Dr. Mattson said that he would provide this information to the Subcommittee prior to its next meeting (March 2,1979).
There was extensive Subconmittee discussion regarding the variation of tLe unreliability estimates presented in Appendix F of Volume 3.
This discussion led to a question from Dr. Lee regarding tha unreliability of the mechanical portion versus the electrical portion of the scram system.
Dr. Lee was concerned with the fact that the prevention aspect of the ATWS alternatives focuses on the electrical portion of the scram system.
He felt there was a need to discuss the mechanical unreliability of the scram system as well.
Dr. Mattson said that it is the NRC's best engineering judgment that the mechanical portion of the scram system is more reliable than the electrical portion, out since this reliability can not be quantified na credit has been given for it.
Regarding 272 055
ATWS 1/ 31/ 79 the event-tree scenarios described earlier, Dr. Lee asked whether the single failure consideration or unfavorable moderator temperature coefficient contributed most to peak pressure, flRC replied that the moderator temperature coefficient makes the greater contribution.
Question 4:
" Discuss the procedures. for performing analyses of plant mitigating systems as described in Alternatives 3 and 4, and for jriging the acceptability of the fixed-up system." Mr. Thadani reviewed the procedures for performing the plant analyses and detailed the systems and components to be analyzed. He noted that in some cases, a small runHr of components may exceed Service Level C stress limits, and it may be necessary to use inelasti, analysis. fiRC is insisting that the mitigating sys ems be ioliable, separate, and diverse.
There was discussion of the usefulness of diversity from the aspect of operator action defeating diversity, as has happened during periodic plant testing. Questicas from Drs. Kerr and Saunders and Mr. Epler led the Staff to state that there is indeed always the possibility of the diverse systems being defeated by operator action, and the concept of defense in depth is relied on.for this situation.
Dr. Lipinski noted some criteria in 3 flVREG report and said that merely citing criteria does not necessarily insure reliability just because diversity is obtained.
Mr. Ditto agreed, stating that he believed diversity in and of itself is not a cure-all.
2[2056
ATWS 1/31/79
(
EPRI PRESENTATION ON ATHS - G. S. LELLOUCHE Dr. Lellouche discussed flRC comments made concerning the EPRI ATWS report.
He said that many of the comments were incorrect (Attachments D-5-8).
Mr. Lellouche went on to characterize the t1RC risk analysis in Volume 3 as inscrutable, and said that Volume 3 cannot be used for decision-making by the Commission.
ATOMIC INDUSTRIAL FORUM PRESENTATION ON ATWS - R. NEWTON, L. LONG, W. LARSON Mr. Newton, Acting Chairman of the AIF ATWS Committee, said t'
' AIF would offer comments on Volume 3 of fiUREG-0460.
Mr. Newton introduced Mr. Louis Long tn present ATWS history and general comments on Volume 3.
Mr. Long reviewed the history of ATWS, focusing on the continuing evolvement of NRC ATWS requirements (Attachment D-9).
He offered the following general comments:
(1) ATWS requirements have continually escalated over the years; (2) AIF is discouraged that Alternative 4 (mitigation only) has been retained as a Staff option; ( 3) AIF is also discouraged that value-impact appears to be abandoned in the Staff analysis; (4) AIF is encouraged that NRC has factored engineering judgment and cost-benefit analysis into the resolution of ATWS; (5) AIF believes that consideration of the conditional core-melt probability, given an ATWS event, would sustantially reduce the " societal
. risk" probability given in Appendix F of Volume 3.
Dr. Kerr asked Mr. Long how the utilities as a body concluded that ATWS does not pose an unacceptable risk to society.
After considerable dis-cussion, it was determined that the industry conclusion was based primarily on risk analysis along with engineering judgment. Mr. flewton said the industry" believes the probability numbers calculated by NRC are unduely conservative, and the true risk from ATW."
l-2 orders of magnitude less than the Staff believes, without attempt. 3 to factor in the benefit of operator action.
c/2 057 r-
ATWS 1/31/79 Mr. Larson gave a presentation on ATWS versus core melt. He focused on the premise that industry believes ATWS does not equal core melt; thus, resolution of ATWS is not based ca societal benefit, but is a licensing issue. Mr. Larson said that high pressure in a PWR will not impair valve functionability for long-term core cool-down, and thus cora melt will be prevented.
The proposed AIF. ATWS fix for PWRs is as follows: operating PWRs-no modifications.
For future PWR designs, AIF stated that an alternate system (undefined) should be provided to eliminate ATWS as a licensing concern.
In the case of BWRs, AIF believes that an appropriate ATHS fix for' BWRs currently operating and under construction is installation of RPT, and prompt operator action.
AIF also said that ARI (alternate rod injection), would be implemented on the BWR shutdown system. Mr. Larson said that Alternatives 3 and 4 would lead to future regulatory instability; however, in response to a question from Dr. Kerr, Mr. Larson said that if regulatory stability is maintained, Alternative 3 may be acceptable.
There was additional discussion of the NRC Staff requirement in Alternative 3 to demonstrate by analysis the 71tigation capability of the plants. The industry is opposed to this re' irement.
Prior to going into open executive session, Dr. Kerr announced that the next ATWS Subcommittee Meeting would be held on Friday, March 2, 1979.
OPEN EXECUTIVE SESSION Dr. Kerr surveyed the Subcommittee members and consultants to determine if there was any additional information they would like to see provided based on the day's discussion.
Dr. Lipinski indicated that he would like to see the reference material for Appendix F, Volume 3 that Dr. Mattson committed to providing to the Subcommittee.
He also requested a brief discussion on the difference in logic systems for the modified Hatch and Monticello RPT designs approved by the NRC Staff. Mr. Ray suggested the AIF previde a formal statement as to the reasons for the unanimous 27q OGS
f
~ _.
ATWS 1/ 31/ 79 consensus among the utilities that ATWS is not a safety issue.
Dr. Lee questioned specifics on the utilities operator training program concern-ing recognition of ATWS.
cr lee also requested information on the basis for such systems as ARI (alternate rod injection) on BWRs.
The meeting was adjourned at 4:30 p.m.
NOTE: Additional meeting details can be obtained from a transcript of the meeting which is available at the NRC Public Document Room, 1717 H St., N.W., Washington, D.C., or can be obtained from Ace-Federal Reporters, 444 North Capitol St., N.W., Washington, D.C.
272 059
s NN Anvrsoar couxnm oM rEAcTon san.
GUARDS SUBCOMMITTIE ON ANTK3 PATIO At th.e conclus!on of the Executive TRAN54fN15 WITHot/T SCRAM (ATW5)
Session, the SubcommJttee 3-111 hear presentations by and hold di.scussions m eng
'with representativs of the NRC Staff, various Industrim. and their consul.
The ACRS Subcommittee on Antie!*
tants, pertinent to this review. The pated Transients without Scram SubcommJttee may then caucus to de-(ATWS) will hold an opening meeting termine whether the matters ident!.
on January 31, 1979, in Room 1046.
fled in the initial session have been 1717 H Street N.W Washington, DC adequately covered and whether the 20555 to continue its discussion of Vol project La ready for review by the full 3 NUREG-0460, " Anticipated Tran-Committee.
sients without Scram for Lfght. Water Purther informattoo regardfr4 Reactors."
topics to be discuse,ed, whether the In accordance with the procedures meeting has been canceHed or resched-outlined in the Ferno. Rtc1sm on uled. the Chairman's ruling on re-October 4,1978. (43 FR 4b926), oral or quests for the opportunity to present
- Titten statements may be presented oral statements and the time aUotted by members of the pubuc. recordings therefoe can be obtained by a prepajd w1U be permitted only during those telephoae call to the Designated Fed.
portions of the ;necting schen a tran.
eral Dnployee for this meeting, Dr.
script is be!ng kept, and questions may Thornas O. McCreless, (telephone 202/
be asked only by members of the Sut>
634-3267) between 8:15 a.m. and 5:00 committee, its consultants, and Staff.
p.m EST.
Persons desiring to make oral state.
Background Information concerning ments should notify the Designated items to be considered at this meeting Federal Employee as far in advance as can be found in documents on fue and practicable so that apprepnate ar.
available for public inspection at the rangements can be m*de to allow the NRC Public Document Room.1717 H necessary time during the meeting for Street, N.W Washington, DC 20555.
such statements.
Dated: January 9,1979.
'Ite agenda for subject meeting shall be as follows: Wednesday, Icns.
Joitw C. HortA ary 31,1979. 8:30 a.m. until the conclu.
Advisory Committee ston of business.
Mnnement WW.
The Subcomm!ttee may meet in Ex.
[FR Doc. 79-2238 Pued 1-18-79; 10:30 sm!
ceutive Session, with any of its consul.
tants who may be present, to explore and exchange their preliminary opin.
lons regarding Instters which should be considered during the meeting and to formulate a report and recommen.
dations to the full Committee.
FEct1At REGt5Ttt. VOL. 44, NO.14-fuDAY, JANUART 19,19T9 gracHmfNT A 272 060 r~
ACRS SUBCOMMITTEE MEETING ON ANTICIPATED TRANSIENTS WITHOUT SCRAM JANUARY 31, 1979 WASHINGTON, D.C.
ATTENDEES LIST ACRS NRC NORTHEAST UTILITIES W. Kerr, Chairman R. Woods, D0R C. Mark, Member T. Novak, DSS R. L. McGuinness J. Ray, Member D. Thatcher, DSS S. Ditto, Consultant R. Tedesco, DSS JCP&L E. Epler, Consultant A. Thadani, DSS J. Lee, Consultant C. P. Speis, DPM K. R. Goddard W. Lipinski, Consultant J. Norberg,0SD S. Saunders, Consultant M. Stolzenberg, RES DUKE POWER T. McCreless, Staff
- R. Mattson, DSS P. Boehnert, Staff P. M. Abraham WESTINGHOUSE W. H. Rasin
- Designated Federal Employee S. Miranda TOLCYO ELECTRIC EBASCO SERVICES, INC B. Steitler H. Hamada M. P. Horrell 0FFSHORE POWER SYSTEMS BALTIMORE GAS &
B&W R. M. Randall ELECTRIC CO A. McBride PSE&G R. C. L. Olson F. McPhafter E. Oelkers C. W. Veprek CE_
Bu TVA NORTHERN STATES POWER n an D. Wilson J. A. Gonyeau WASHINGTON PUBLIC POWER SUPPLY SYSTEM STONE & WEBSTER R. G. Cockrell D. R. Jaquette A. D. C. K1mmins P. V. Holton C. F. Grochmal EPRI PASNY WISCONSIN ELECTRIC POWER C. P. Chen R. A. Newton SOUTHERN CO. SERVICES GENERAL ELECTRIC E. M. Page L. B. Long H. C. Pfefferlen AIF NRC - OCM-Ahearne E. P. Stroupe F. Stetson V. Harding, ELD BOSTON EDIS0N GEORGIA POWE_R
_R W. W. Larson W. E. Burns 272 061 ATTACHMENT B
A'IWS WORKING GROUP MEETI?G JANUARY 13, 1979 RASHINGTON, D.C.
'IENTATIVE SCHEDULE T PRESENTATIOBE 8:30 a.m.
I.
INTROULCTION W. KERR - WORKING GROUP CHAIRMAN 8:45 a.m.
II.
WORKING GROUP DISCUSSION WITH IRC STAFF ON A'IWS 12:00 noon III.
LUNCH IV.
EPRI A'IWS PRESENTATION 1:00 p.m.
G. S. LELIDUCHE V.
AIF PRESENTATION ON A'IWS 2:00 p.m.
A.
INTRODUCTION R. NPEON - ACTING CHAIRMAN - A'IWS CCF'J4ITTEE B.
'IESTIMONY L. LOtG W. IARSON VI. WORKING GROUP DISCUSSION 3:30 p.m.
4: 30 p.m.
VII. ADJOURN n17AdilhHNT C 272 062 c
1/31/79 ATWS DISCUSSION 'IDPICS WITH NRC STAFF 1.
Outline any significant differences that exist between analysis (and results thereof) of the course of the various postulated ATWS events carried out by the NRC Staff and by vendors.
2.
Explain the reasoning that leads the Staff to conclude that the prescriptions of Alternative 4 will lead to lower risk than those of Alternative 3 for new reactors.
How much lower does the Staff expect the risk will be for each vendor's reactor?
3.
Discuss the most propable accident scenarios for linking A'fWS to a core melt.
4.
Discuss the procedures for performing analyses of plant mitigating systems as described in Alternatives 3 and 4 and for judging the acceptability of the fixed-up system.
272 063 r--
WASHINGTON, D. C. 20$55
- Q
- g
.< /
JAN 181979 MEMORANDUM FOR: Lee V. Gossick, Executive Director for Operations FROM:
Edson G. Case, Chairman Regulatory Requirements Review Committee
SUBJECT:
SUMMARY
OF REGULATORY REQUIREMENTS REVIEW COMMITTEE MEETING NO. 81, JANUARY 2,1979 The Committee continued its review of ATWS, particularly Volume 3 to NUREG 0460 dated December 1978, and reached the following conclusions and recommendations:
a.
The Comittee concurred in the overall approach used for developing staff recorrendations for resolving the ATWS issue as set forth in NUREG 0460, Volume 3.
In particular, the Committee agreed with the use of engineering judgment as the primary basis for reaching decisions on the ATWS issue, with quantitative risk assessment used in a supportive role.
b.
The Committee recomended that the following general requirements for ATWS protection of LWR's be established by a notice and comment rulemaking proceeding:
(1)
Early Operating Plants s
Provide as a minimum the modifications of Alternative 2 of NUREG 0460, Volume 3.
Because of the unique characteristics of these older and generally smaller facilities, plant unique analyses and reviews should be performed to determine the needed additional ATWS prevention or mitigation features that are cost effective for these plants in the context of the overall safety of these facilities.
(2)
Other Operating Plants And Plants With Construction Permits Issued Prior To January 1 a 1978 a.
Provide the modifications of Alternative 3 of NUREG 0460, Volume 3, for all other operating plants with OL's issued on or before the effective date of the rule A797tHM5/if D-/
272 064
s JAN 181979 Lee V. Gossick.
t (estimated to be January 1, 1980). A reasonable amount of time to accomplish the modifications should be allowed so as to minimize down time.
It appears that 2 years should be sufficient.
b.
Provide the modifications of Alternative 3 of NUREG 0460, Volume 3, for plants currently under construction with CP's issued prior to January 1, 1978. A reasonable amount of time to accomplish the modifications should be allowed.
It appears that 2 years after the effective date of the rule, or before issuance of an operating license, whichever is later, should be sufficient.
(3)
Plants With Construction Permits Issued On Or After January 1,1978, New Plants, And Standard Plant Design Approvals Provide the modifications of Alternative 4 of a.
NUREG 0460, Volume 3, for plants with CP's issued on or after January 1,1978, but before the effective date of the rule.
Based on the premise that the effective date of the rule will be no later than January 1,1980, the modifications should be completed before these plants are issued an OL.
With regard to those plants with CP's issued on or after January 1, 1978, but which reference a standard or duplicate or which replicate a base plant design approved prior to that date, the Committee was evenly divided as to whether such plants should be required to provide modifications of Alternative 3 or Alternative 4 of NUREG 0460, Volume 3.
The Committee recommended that the Director, Office of Nuclear Reactor Regulation decide this question.
b.
Provide the modifications of Alternative 4 of NUREG 0460, Volume 3, for new plants with CP's issued on or after the effective date of the rule.
The modifications should be completed before these plants are issued an OL.
272 065 7.
JAN 181979 Lee V. Gossick.
4 c.
Stan3ard plant Design Approvals provide the modifications of Alternative 4 of NUREG 0460, Volume 3, by amendment of all currently effective preliminary reference design approvals.
Based on the premise that the effective date of the rule will be no later than January 1,1980, the modifications of all plants that reference these amended preliminary design approvals should be completed Lefore these plants are issued an OL.
provide the modifications of Alternative 4 of NUREG 0460, Volume 3, for all new preliminary reference design approvals, for all final reference design approvals that can be referenced in construction permit applications or that are associated with preliminary reference design approvals amended to provide the modifications of Alternative 4, for all preliminary duplicate plant design approvals and final duplicate plant design approvals that can be referenced in construction permit applications, and for all manufacturing licenses issued before the effective date of the rule.
Based on the premise that the effective date of the rule will be no later than January 1,1980, the modifications of all plants that reference these standard plant design approvals should be completed before these plants are issued an OL.
provide the modifications of Alternative 4 for preliminary reference design approvals and final reference design approvals that can be referenced in constructica permit applications, preliminary duplicate plant design approvals and final duplicate plant designs that can be referenced in construction permit applications, and manufacturing licenses issued on or after the effective date of the rule.
The nadifications of all plants that reference these standard plant design approvals should be completed before these plants are issued an OL.
/ Zdson G. Casa *, Chairman Legulatory Requirements Review Committee cc: See Next page t'
-272 066 r-
e WASHINGTO" 0. C. 20555 3
f, '
, /
JANUARY 6 1973 MEMORANDUM FOR:
T. A. Ippolito, Chief, Operating Reactors Branch #3, 00R FROM:
V. L. Rooney, Project Manager, Operating Reactors Branch
- 3, DOR
SUBJECT:
SUMMARY
OF MEETING HELD ON JANUARY 9,1979 TO DISCUSS ATWS RECIRCULATION PUMP TRIP On January 9,1979, representatives of BWR licensees which do not have recirculation pump trips installed met with the NRC staff.
Mr. Harold R. Denton, Director of the Office of Nuclear Reactor Regulation, reviewed the history of the ATWS issue, and described the present ATWS status. The general agreement that recirculation pump trip (RPT) significantly limits the consequences of an ATWS event was noted.
It was pointed out that, although licensees have delayed RPT installation awaiting firmer NRC requirements, NRC has a firm and carefully considered pcsition that RPT is required.
It was further pointed out that RPT is a minimum requirement in that regardless of the resolution of the renaining ATWS considerations, RPT will be required although it mav not be a " total" or sufficient modification.
Letters requesting RPT installation were distributed to licensees and discussed with them. The letters, together with enclosures:
1.
Describe the bases for requiring an RPT at this time, 2.
Describe two alternative ways to provide an acceptable RPT (referred to as the Monticello fix and the modified Hatch fix),
3.
Request that within 90 days, licensees provide an RPT implementation schedule and that installation be accomplished within two years, 4.
Request that in the interim, i.e., until RPT installation is made, licensees train o should one occur.perators to recognize and respond to an ATWS event This would provide significant protection from those ATWS events which occur at low power levels where the rise in the vessel pressure and the containment temperature is limited to acceptable values by manual recirculation pump trip and actuation of the existing standby liquid control system.
If the operator 272 067 M 7
Meeting Summary.
were to promptly (in a few seconds) trip the recirculation pumps to assure that the short term rise in vessel pressure is not excessive, protection will also be provided for those ATWS events where the common mode failure occurs in either the electrical portion of the scram system or in some portions of the drive, system, and 5.
Include Volume 3 of NUREG-0160 which provides the latest ATWS recommendations and information on cost and other impacts.
It was pointed out that the Regulatory Requirements Review Committee has completed their review and concurred with the staff approach described in Volume 3 of NUREG-0460 in so far as it applies to the BWR class of plants under discussion.
After a caucus the licensees indicated that they had no question at that time with respect to the technical position described in the letter.
A list of attendees is enclosed as Attachment A, and a sample letter is enclosed as Attachment B.
V. L. Rooney, roject fanager Operating Reactors Branch #3 Divisien of Operating Reactors
Enclosure:
1.
List of Attendees 2.
Sample letter 272 068 Y
ATTACHMENT A LIST OF ATTENDEES Names Organization J. Hannon NRC P. Pol k NRC R. Bevan NRC D. Skovholt NRC P. Collins NRC H. Daniels NRC D. Tondi NRC J. Beard NRC G. Lainas NRC J. Wetmore NRC R. Campbell NRC H. Denton NRC V. Stello NRC R. Mattson NRC T. Ippolito NRC P. O'Connor NRC B. Grimes NRC G. Klinger NRC P. Boehnert NRC A. Thadani NRC D. Bunch NRC R. Woods NRC P. Shemanski NRC R. Cooley NRC J. Scinto NRC D. Eisenhut NRC E. Case NRC J. Co f fman NRC V. Rooney NRC B. Mayn '
NRC B. D' Angelo Niagara Mohawk R. Huston Consumers Power D. Bixel Consumers Power B. McGuinness Northeast Utilities C. Ondash Boston Edison W. Larson Boston Edison W. Armstrong Boston Edison L. DelGeorge Commonwealth Edison C. Reed Commonwealth Edison 272 069 r
ATTACHMENT A (Cont'd)
LIST OF ATTENDEES Name Organization R. Groce Yankee Atomic R. Shimshak Dairyland Power W. Nowicki Dairyland Power J. Parkyn Dairyland Power S. Rafferty Dairyland Power D. Gridley General Electric R. Brugge General Electric R. Milos Nuclear Energy Services V. Latorre Lawrence Livermore Laboratory 272 070 E94-
,ye
UNITED STATES f
- g NUCLEAR REGULATORY COMMISSION g-g g,,
y.
- y WASHINGTON, D. C. 20555 a
N %.
,o JANUARY 8 N
i Docket No. 50-245 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Conpany Post Office Box 270 Hartford, Connecticut 06101 Dear Mr. Counsil.
Attached for your infomation is a copy of NUREG Oa60, Volume 3 which details our current view related to ATWS.
In this supplement a variety of options are considered regarding ATWS. We intend to select one of the ATWS options in the near future and to pursue it to adoption.
However, it is important to note that all of the options under serious consideration by the NRC staff (options #2, 3, and 4 in Volume 3 of NUREG 0460) regarding resolution of the ATWS issue for BWRs require installation of an RPT. While you have committed to install a RPT on your facility, Millstone Unit 1, you have not yet begun to take steps towarc such installation, on the grounds that you were awaiting fimer requirements by NRC. The NRC staff now has a firm position that RPT is recuired for your facility. Therefore, we see no bases for any further delay in implementing an RPT for your facility. The RPT designs discussed in this letter are compatible with ATWS requirements.
.c To expedite your installation of an approved RPT, the staff is providing a modified description (Appendix A, attached) of design requirements which provide sone additional flexibility over those previously provided (May, 1978), but which the staff has found
~
acceptable for RPT systems to be installed in the near future.
~
For all operating plants, the Monticello RPT design described in NEDO 25016 and summarized in Appendix B has been accepted by the staff as meeting.the Appendix A criteria. Sections of NEDO 25016 related to ARI should be ignored as that system is not addressed by this letter.
Some operating plants have already installed the "BWR/4" or " Hatch" RPT, and the staff also accepts that design as meeting the Appendix A criteria provided the changes specified in Appendix B, or equivalent changes, are incorporated.
272 071 r"
Mr. W. G. Counsil Both the Monticello design and the modified "BWR/4" or " Hatch" design
~
utilize generator field breakers which have been modified so that they are provided with two trip coils. One coil for each breaker is actuated only by reactor pressure and water level sensors in RPT division A, and the other coil is actuated by pressure and level sensors in RPT division B, thereby providing redundancy of power supplies available to the overall system and increasing trip reliability.
Either the Monticello or modified '8WR/4" or " Hatch" design, would be an acceptable RPT design provided diverse final trip relays of a different type are used, or obtained from a different manufacturer than the primary scram relays used in the RPS.
The staff has not reviewed the specific design of the time delay circuitry recently proposed fcr the Monticello RPT design for low-level initiated pump trips. We agree that time delays on the order of 10 seconds are desirable to avoid making the consequences of a postulated LOCA more severe, and we agree that such delays of around 10 seconds have insignifi-cant effect on ATWS consequences (for low-level initiated ATWS pump trips only). Therefore, we find incorporation of such circuitry on either RPT design discussed above to be acceptable,.provided:
1.
The time delay is realized only for low-level initiated pump trips;
- and, 2.
The circuitry is incorporated in such a way that it does not signiff-cantly affect the overall reliability of the RPT; that is, that no single failure in the timing circuit (s) can cause failure of the pump trip to occur. This could be accomplished, for example, by use of a
. separate, independent timing (delay) circuit with each low-level sensor, or equivalent.
Implementation as soon as possible of an RPT in accordance with the attached design criteria will provide an increased level of safety over the lifetime of the plant and should be installed as pronptly as is reasonable.
s 272 072 r
Mr. W. G. Counsil '
The staff has given careful consideration to the concern expressed by some licensees that RPT design requirements may change in the future.
We have concluded that the design criteria outlined in this letter
~
(Appendix A) are, for operating plants, equivalent to those enclosed with the May, 1978 letters to all BWR licensees, and we intend to effect no changes to those criteria in the future.
We believe that RPT design, procurement, and installation can be accomplished within a two year period without requiring additional outage time beyond refueling outages.
We have given consideration to steps that can be taken at present, in order to reduce the risk from ATWS events during the interim period before recirculation pump trip circuitry and any other necessary plant modifications are completed. We have detemined that many of the following steps are practicable and appropriate for your facility for this interim period. We therefore, request that you inform us within 50 days that you have done the following:
1.
Developed emergency procedures to enable operators to recognize an ATWS event, including consideration of scram indicators, rod positio.. indicators, flux monitors, vessel level and pressure indicators, relief valve and isolation valve indicators, and containment temperature, pressure, and radiation indicators.
2.
Train operators to take actions in the event of an ATWS including consideration of manually tripping the recirculation pumps and scramming the reactor by using the manual scram buttons, changing individual rod scram switches to the scram position, stripping the feeder breakers on the reactor protection system power
" distribution buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the RHR in the pool cooling mode to reduce the severity of the containment conditions.
Early operator action as described above would provide significant protection from those ATWS events which occur at low ocwer levels where the rise in the vessel pressure and the containment temperature is limited to acceptable values by manual recirculation pump trip and actuation of the existing standby liquid control system.
If the operator were to promptly (in a few seconds) trip the recirculation pumps to assure that the short tem rise in vessel pressure is not excessive, protection will also be provided for those ATWS events where the common mode failure occurs in either the electrical portion of the scram system or in some portions of the drive system.
272 073 r
Mr. W. G. Counsil Within 90 days inform us of your schedule for implementation of your commitment to install an RPT system for your plant. Such system should conform to the acceptable systems described in this letter and your schedule should be consistent with the staff's overall objective of assuring that an acceptable RPT system is installed at your facility within two years.
Sincerely,
/
~
Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
NUREG 0460, Volume 3 2.
Appendices A and B cc w/ enclosure No. 2:
see next page e
272 074
Northeast Nuclear Energy Company.
CC William H. Cuddy, Esquire Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103
~
Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N. W.
Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut 06385 Mr. Janes R. Himmelwright Post Office Box 270 Hartford, Connecticut 06101 Nuclear Regulatory Commission, Region I Office of Inscection and Enforce:ent ATTN: John T. Shediosky 631 Park Avenue
-c King of Prussia, Pennsylvania 19406 r
272 075 7.--
APPENDIX A CRITERIA FOR HIGH PRESSURE-LOW LEVEL INITI ATED RECIRCULATION PUMP TRIP _(RPT TO BE INSTALLED IN OPERATING BWRs BEFORE NOVEMBER 1, 1979*
A.
General Functional Recuirement The RPT system shall automatically initiate the appropriate action whenever the conditions monitored by the system reach a preset level.
B.
Independence and Intecrit1 The RPT system and components shall be independent and separate from components and/or systems that initiate anticipated transient (s) being analyzed and diverse from the normal scram system to minimize the probability of disabling the operation of the mitigating system.
Diversity can be achieved by incorporating as many of the following methods as is practicable:
1.
Use of RPT final trip relays from different manufacturers (required).
2.
Use of energized versus de-energized trip status.
3.
Use of AC versus DC power sources.
It shall be demonstrated that the function of the RPT system and components will not be disabled as a consequence of events being analyzed.
Diversity of the RPT pressure and level sensing devices (including relays used in such sensing devices) from similar or identical devices used on the RPS is not required, since failure of those devices on both the RPT and the RPS is not likely to cause an ATWS due to the presence of other diverse trips on the RPS (high flux, valve position, etc.).
- The NRC staff has reviewed the Monticello RPT design and the " Hatch" RPT design, and finds that they meet these criteria (provided the changes specified in the cover letter are made to the " Hatch" design). Plant specific reviews will be conducted only as necessary to ascertain that the plant design is the same as, or equivalent to, one of the approved designs.
272 076
C, Eouiement Oualification The RPT system equipment and components shall be tested to verify that the system will provide, on a continuing basis, its functional capability under conditions relevant to postulated ATWS events, in-cluding extremes of conditions (as applicable) relating to environ-ment, which are expected to occur in the lifetime of a plant.
-D.
Periodic Surveillance and Preventative Maintenance Testing and Calibration Periodic surveillance and preventative maintenance tests and calibra-tion requirements shall be identified to provide continuine assurance that the RPT system, including sensors and actuated equipment, is capable of functioning as designed and that system accuracy and per-formance have not deteriorated with time and usage. These recuirements shall be particularly directed toward the detection of those failures or degradation of accuracy and performance which would not otherwise be likely to be detected during the course of normal operations.
Integrated system testing shall also be performed to verify overall system performance.
E.
Quality Assurance A quality assurance program in conformance with the requirements of 10 CFR 50 Appendix B shall be applied to the RPT system design and equipment.
F.
Administrative Controls
.c Administrative controls shall be established to control the access to all set point adjustments, calibration and test points.
G.
Information Readout The RPT system shall be designed to provide the operator with accurate, complete and timely information regarding its status. For those functions, including operations, test or maintenance, and calibration, which require direct operator interaction, human engineering factors such as information displays (e.g., display formats, layout and con-trols)~ and functional controls (e.g., methods, location and identifi-cation) shall be included in the design.
272 077 r-
3 H.
Maintainability The design shall include measures which enhance maintainability to reduce mean-time-to-repair and to assure the continued availability and reliability of the system for the life of the plant. The system design shall include features which facilitate the recognition, loca-tion, replacement, repair and/or adjustment of malfunctioning equipment and components or modules.
e 272 078 I
Apoendix B Acceptable RPT Designs Monticello RPT Desian The Monticello design simultaneously trips both MG sets "A" and "B" generator field breakers upon receipt of eitner reactor high pressure or icw-low water level control logic input signals. The logic to each breaker is two-out-of-two (pressure) or two-out-of-two (level)
(2/2 or 2/2), i.e., contacts "A" and "C" or contacts "B" and "D" must close to trip the breaker. The Monticello desiga employs diversity, testability, separation and redundancy.
Modified BWR/4 or Hatch RPT Design The modified "BWR/4" or " Hatch" design results in the independent (separate) trip of each of the two recirculation pumps upon receipt of either one reactor high pressure signal or one low-low water level signal. The logic to each MG set "A" and "B" generator field breaker is one-out-of-two (level) or one-out-of-two (pressure) (1/2 or 1/2).
The modified "BWR/4" or " Hatch" design employs diversity, testability, separation, and redundancy.
The modification to the existing " Hatch" design which makes it acceptable is accomplished as follows:
- 1) Add a second trip coil to each recirculation loop's M-G set genera:cr field breaker, as per the identical codification made to Monticello.
- 2) Connect one of the pressure sensors and one of the low level sensors in RPT train A to the old (existing) trip coil in the recirculation loop A M-G set generator field breaker. Connect one of the pressure sensors and one of the low level sensors in RPT train B to the new trip coil in the recirculation loop A M-G set generator field breaker.
- 3) Connect the other pressure sensor and the other low level sensor in RPT train A to the new trip coil in the recirculation loop B M-G set generator field breaker. Connect the other pressure sensor and the other low level sensor in RPT train B to the old (existing) trip coil in the recirculation loop B M-G set generator field breaker.
272 079
_.(
>~
i i.
Omo NN 3
D TN EM H
CATTA des teav rl ea E
pV E
O R
f T
re ei T
wl N
oe E
PR V
E S
W A
R yr W
re P
at ia l w idxe ue AF cny ior tit ati mau ouc tt r uci AAC i
SW D
T E
A E
C L
C I
U A
S F
m
^
o v
9
i
~
co O
N N
Long-Term Cooling High Pressure Core Spray Poison Injection SUCCEED
^
o If FAIL t
ATTACHMENT D-4 e
e
FIGURE II MISSTATEMENTS C0flCERNING EPRI WORK 2 jic t.x, '; s 9
QUESTION 6 STAFF SAYS:
"EPRI'S BEST ESTIMATE FOR THE FUTURE IS 3 PER RY, NUREG 0460 USES 6."
EPRI SAYS:
"EPRI'S BEST ESTIMATE FOR THE FUTURE IS 1.7/RY (0.6 WITH APPROPRIATE BYPASS). IF A FULL CORRECTION FOR POWER LEVEL EFFECT IS MADE THEN THE BEST ESTIMATE IS 1.3/RY (0.4 WITH APPROPRIATE BYPASS)."
272 082 g-5
FIGURE III QUESTION 9B l
1 STAFF SAYS:
rMINADVERTENTOPENINGOFADDITIONALRELIEFVALVES (0.05fYR)ADDSINCIDENTALRISK EPRi SAYS:
,FAILUREp0RESEATANALREADYOPENEDVALVE
,k
/(0.05/ DEMAND) ADDS CONSIDERABLE RISK /,,', :,,.
/[.),
/
d?
0
, L -y,, -..'-'
)
~
.~
Tij 272 083 D-6
.,1 GC 70
FIGURE IV QUESTION 13 WfiY REDUCE 1.1x10-4 TO 3x10-57 STAFF GIVES NO ANSWER, BUT:
PREDICTION OF SCRAM FAILURES 1.1x10-4 d"N## 3x10-5 ACTUAL b
LWR'S 7-8
-e2 0 q, V
NAVAL 11 Uj3 0
272 084 p-7 1-26-79 w
FIGURE V 2
QUESTION 20 r o.. ' ;;" "
n, f, : 4 ~
EFFECT OF BYPASS ON BWR's NOT ANSWERED, ACTUAL EFFECT IS FACTOR OF 3.
i f
i 272 085 0-9 v
1-26-79 i
.~
Y~ ~
EJOLUTION OF ATWS REQUIREP'ENTS s
ATWS-1 l
ATWS-Il PRE WASH -
ATWS-Ill ATWS-IV AinS-V WASH-NUREG 0460 1270 STATUS NUREG NUREG 1270 ISSUE PHASE REPORT 0460 0460 PHASE PHASE vo!. 122 Vol. 3 Time Period 1968-9/73 9/73-12/75 12/75-4/78 4/78-Present 12/78-Present w
involvement vendors vendors vendors Utilities Vendors Min. Utility A/E's N
EPRI EPRI UtilItles N
ATwS Solution Study Only Mitigate Deterministic Mitigation +
Prevention and/or N
Only Mitigation Rellability Mitigation Type of Event Special Study Special Study Special Study DBA Generic Analyses Probabiiity None None Goal 10-7/ year 10-6/ year None Models Best Estimate Conservative Conservative Like LOCA Conservative Estimate Anticipated None 0.1-l.0 1.0 6.0 6.
Transients Parameters None Nominal 99% HTC 99% HTC 95% MTC and IEEE 279 No 99% MTC No No Yes Yes and No Safety Grade No No No Yes Yes and No Stress Limit None-Faulted Emergency Emergency Emergency Emergency Dose Modei None Nominal Modified None Parameters Accident Parameters l
j:
i I
t