ML19323H500
| ML19323H500 | |
| Person / Time | |
|---|---|
| Issue date: | 04/17/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1729, NUDOCS 8006130055 | |
| Download: ML19323H500 (15) | |
Text
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ISSUE DATE: April 17,1980 MINUTES OF THE ACRS B&W WATER REACTORS SUBCOMMITTEE
/ M 6 - / '72 9 MEETING WASHINGTON, D. C.
April 8,1980 The B&W Water Reactors Subcomittee of the ACRS met on April 8,1980 at 1717 H St., N.W., Washington, D.C.
The main purpose of the meeting was to continue the Subcomittee's review of the sensitivity of B&W reactor systems to feedwater transient s.
Notice of the meeting was published in the Federal Register on March 24, 1980.
Copies of the notice, meeting attendees list, and meeting schedule are in-cluded as Attachments 1, 2, and 3, respectively.
No written statements nor request for time to make oral comments were received from members of the public.
EXECUTIVE SESSION Mr. H. Etherington, Subcomittee Chairman, convened the meeting at 8:30 A.M.
and introduced the ACRS members and consultants (Attachment 2) who were present. The meeting was conducted in accordance with the Federal Advisory Comittee Act and the government in the Sunshine Act. Mr. Peter Tam was the Designated Federal Employee. Mr. Etherington indicated, in his opening state-ment, that Harold Denton of NRR has recommended that construction of B&W plants not be halted.
(Memo, Denton to Comission, Jan. 22,1980). The Sub-committee would review this recomendation and document its findings in an ACRS letter.
Members and consultants made no comment at this point and the meeting proceeded as scheduled. The first presentation was made by an ACRS task force fonned to study the B&W NSSS sensitivity question. Mr. Etherington noted that a presentation in a Subcomittee meeting by the ACRS staff is " unusual."
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i l PRESENTATION BY THE ACRS TASK FORCE The ACRS task force has documented its findings in an ' official use only' draft report, distributed to the Subcommittee a few days before the meeting.
Revised pages of the report were provided during the meeting and the task force indicated that the report will be finalized after ACRS comments are received.
The task force report, entitled " Review and Evaluation of the Babcock and Wilcox Nuclear Steam Supply System", was rummarized as follows:
Mr. Ed Abbott described the B&W steam generator design. There are three control modes: the boiler following mode (primary system re-sponse changes according to steam demand), the turbine following mode (steam flow corresponds to primary system heat rate) and the optimized mode (integrated master control which uses a composite system of controls for the turbine, steam generator and reactor subsystems). B&W OTSG uses the last mode.
Mr. G. Young addressed the B&W OTSG dryout time question. The water inventory in a B&W steam generator is much less than that in a W or CE steam generator. Therefore, a loss of feedwater results in relatively I
fast dryout of the B&W OTSG. Mr. Young stated that dryout, however, is a problem only if the result is a safety issue. The Crystal River-3 and Rancho Seco incidences have shown that core damage did not occur as a result of dryout. The only way steam generator dryout can occur at a B&W plant now under construction and designed to the post-TMI standards is for multiple failures to the safety-grade emergency feed-watersystem(EFW). The loss of the EFW system at TMI-2 can be attributed to non-safety grade system interaction.
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. Mr. J. Stampelos addressed the integrated control system (ICS). He said that B&W performed a failure mode and effects analysis (FEMA) of the ICS.
The report of the analysis has been evaluated by ORNL. Mr. Stampelos stated that the ICS is an asset to plant operators, based on operating history, but ICS interface with the balance-of-plant needs improvement.
Mr. J. Bickel made a presentation on dynamic response characteristics of the B&W NSSS. He stated that:
The ability to change load in a B&W NSSS with its feed forward control mode is at least twice as fast as NSSS designs using UTSGs (CE and Westinghouse). This capability is the result of the ability to effectively eliminate the effects of the moderator temperature re-activity defect via proper regulation of feedwater flow and enthalpy.
B&W NSSSs have a characteristically more sluggish pressurizer pressure Yrcs and level response to T,yg change due to their inherently small y
ratios and operating pressures, and temperatures.
B&W NSSSs (with their OTSGs), however, are subject to more adverse i
T,yg changes for the most limiting feedwater and steam flow transients than are plants using UTSGs.
The need for safety-grade anticipatory reactor trip on turbine trip and loss of feedwater is clearly evident in view of the larger re-duction in heat removal capability which can be experienced by an OTSG.
_g-With proper anticipatory reactor trip operation, the B&4 OTSG can be operated safely. B&W should develop a better trip that would permit the ICS to function properly. The current anticipatory FW trip is only good for the interim; for the future, a sustained-loss-of-heat-sink trip should be developed.
DISCUSSIONS WITH THE NRC STAFF Mr. Novak indicated that Mr. Denton of NRR has sent a memorandum to the Commission to recomrend that construction of B&W plants should not be halted.
The Staff has requested that the ACRS review this recommendation and document its concurrence (or lack of) in a letter to the Commission.
In addition, since draft NUREG-066,7 (to be described below) was published prior to the meeting, the Staff would also like the ACRS to comment on the report.
Mr. Etherington indicated that since members and consultants have just received the report in the mail (date of report, April 2,'1980), the Subcommittee has no way of doing a fair review in this meeting. Such review would take place in a future meeting.
1.
NUREG-0667 - Mr. R. Tedesco Mr. Tedesco is chairman of a Staff task force formed to provide an assessment of the apparent sensitivity of the B&W designed plants to transients and the consequences of malfunctions and failures of the ICS and non-nuclear instru-mentation (NNI).
The task force, consisting of about a dozen individuals, spent about two weeks studying the issues and documented its findings in NUREG-0667, " Transient Response of Babcock and Wilcox-Designed Reactors." General findings are:
B&W designed plants are more responsive to secondary side perturbations than other light water reactors.
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. The once-through steam generator design is basically sound; however, it requires a highly interactive and responsive control system (i.e., the integrated control system).
A high degree of overall plant interaction is inherent in the integrated control system and the once-through steam generator.
Based on the design features and the faster response of B&W plants during transients and upset conditions, the operators may be required to take more rapid action and have a better understanding of instrument re-sponse than operators on plants having other designs.
Implementation of the 22 specific requirements would be incorporated into TMI-2 action plan. Details of the implementation, to be presented in Section 7 of the report, were not finalized yet. The 22 requirements fall into four action areas, and are described in detail in the report:
AUXILIARY FEEDWATER SYSTEM 1.
Classify AFW system as engineered safety feature, Where upgrade may not be feasible, consideration would be given to the addition of a 1
dedicated AFW system (i.e., a separate train).
2.
Automatically initiated and controlled by engineered safety features which are independent of the ICS, NNI, and other non-safety systems.
3.
Installation of a diverse-drive AFW pump for Davis-Bessee.
4.
Reevaluate and modify such that system is capable of differentiating be-tween steam line break and overcooling and undercooling transients.
INSTRUMENTATION AND CONTROL 5.
Improve reliability of instrumentation and plant. control, separate and channelize power buses and signal paths for NNI and associated control systems.
i
' Prompt followup actions should be taken on:
BAW-1564 (ICS reliability analysis)
NSAC-3/INP0-1 recomendations (evaluation of CR-3 incident)
IE Bulletin 79-27 (loss of NNI power supplies) 6.
Establish prompt implementation of select data set of principal plant parameters for operator (safety grade).
7.
Increased usage of incore thermocouples.
8.
Provide a safety-grade containment high radiation signal to initiate containment vent and purge isolation.
DESIGN AND OPERATIONAL MATTERS 9.
Plant operating and control functions should be modified to maintain pressurizer level on scale and pressure above HPI actuation setpoint.
- 10. Perform sensitivity studies of possible modifications which reduce the response of the OTSG to feedwater flow perturbations.
(Consider active and passive measures).
- 11. Modifications should be made, to the extent feasible, to reduce or eliminate manual immediate actions for emergency procedures.
- 12. Provide a qualified I&C technician on duty with each shift.
- 13. Operator training on CR-3 event as well as plant-specific loss-of-NNI/ICS analysis and procedures.
- 14. B&W develop generic guidelines for loss of-NNI/ICS.
- 15. Mandatory one-week simulator training for operators as part of re-qualification program.
- 17. Staff review alternative solution to PORV unreliability/ safety system challenge rate concerns.
- 18. Expeditious completion of Crystal River 3 IREP study.
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. 19. Staff develop plant performance criteria for anticipated transients for all light water reactors.
- 20. Continue studies of need to trip RCp during small break loca.
(Con-ducted jointly by industry and NRC).
- 22. Staff analysis to determine significance and cause of LER's due to licensed personnel error being higher for B&W plants than other PWR's.
During the discussion, Mr. J. Taylor of B&W stated that a lot of people have the perception that the B&W pressurizer level goes off scale all the time.
He said that it does not; of 350 reactor trips B&W examined, there were indi-cations of only' eighteen occurrences of off-scale behavior.
Mr. C. Domeck of Toledo Edison commented on NUREG-0667. He commended the Staff effort in producing the report but pointed out that some of the requirements may overlap with others. He suggested that there should be active owner parti-cipation in the writing of Section 7, " Implementation of Recomendations."
Mr. J. Taylor of B&W commended the Staff for its NUREG-0667 and the ACRS task force for its report. He indicated he would like a copy of the ACRS report.
B&W has made a number of specific recommendations :.o its utility customers to The improve plant performance in light of the Crystal River-3 incident.
utilities 'are currently evaluating these recomendations for plant specific applicability. B&W plants under construction have many of the features that NUREG-0667 recommends. He said that there should be some criteria for the acceptability of certain transient response, and B&W is suggesting, as a first-cut, the following statements. He called these " transient success statements."
1.
RCS pressure should remain above HPI setpoint.
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. 2.
RCS pressure should remain below safety valve setpoint.
3.
RCS temperature decreases at rates within technical specification limits.
4.
Primary coolant be contained within the primary system and pressurizer quench tank.
S.
Pressurizer level remains on scale.
6.
OTSG 1evel remains on scale.
B&W would also like to participate in the writing of Section 7 of NUREG-0667 Regarding the auxiliary feedwater system, B&W reconinends a reliability-oriented upgrading as opposed to just safety-oriented upgrade.
2.
RELAP-4 SIMULATION OF A HYPOTHETICAL OVERC00 LING TRANSIENT - W. Jensen Mr. Jensen performed a number of calculations using the RELAP-4 code.
The analysis is designed to be a best estimate and is plotted for comparisons with similar analyses by B&W. B&W used a detailed steam generator model (MAXI-TRAP) and a less detailed model (MINI-TRAP). The purpose of the RELAP calculation is to provide an audit of B&W's TRAP computer code which has not been approved by the NRC.
The assumed sequence of events was a reactor trip and turbine trip with the failure of main feedwater to throttle back to maintain the shutdown level of 32 inches.
Feedwater was assumed to continue to flow causing shrinkage of the primary system. Letdown was assumed to isolate and the makeup flow of one charging pump was assumed. The increase in steam pressure following the turbine trip caused the turbine bypass and the steam relief valves to open for a few seconds. The RELAP results show more cooling than B&W, but with less loss of reactor system pressure and pressurizer level than B&W's Maxi-TRAP V
-g-model. No voids were formed in the primary system over the 100 second inter-val of the R LAP analysis.
Mr. Ebersole said that there is a lot of " unreality about the mechanical evolution" in Mr. Jensen's study. Mr. Catton and Mr. Zudans said that plants should be better instrumented to provide more complete transient response dra;
)
mere computer study is not a satisfactory substitute for good data.
3.
ANL SENSITIVITY STUDY ON B&W OTSG - B. Siegel The objectives of this study is to determine the sensitivity of the cooling
)
dynamics to perturbations in the secondary system, to detennine effects of proposed applicant's modifications to reduce sensitivity of the coupling of primar to secondary systems, and' to determine the effects of the use of secondary systems to generate steam (at Midland only). Task 1, to be completed by August,1980, is a parametric study of the effectiveness of proposed modi-fications on transients (including the effects of location of AFW injection, MFW and AFW runback flow rates, time of initiation of runback, and 0TSG water i
level ). Task 2, to be done by July 1981, is an assessment of the change in sensitivity of the primary-to-secondary coupling due to the use of a tertiary heatexchanger(atMidlandonly).
All work is to start soon.
4.
INTEGRATED RELIABILITY EVALUATION PROGRAM (IREP) - J. Murphy The Staff has briefed the Subcommittee on' this program in its Jan. 8,1980 meeting. Mr. Murphy described new developments since then. The Integrated Re-liability Evaluation Program is continuing. General results identified are:
System interactions are significant (especially auxiliary cooling systems, DC and AC power sources).
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. Likelihood-of core damage and of high release categories may exceed those predicted in WASH-1400. However, analytical methods and data may differ and the analyses are not easily compared.
Eventually, IREP will be extended to study six plants:
Indian Point, Zion, Oconee, Calvert Cliffs, Browns Ferry and Dresden. Also, plant owners may be asked to participate in the study.
After the Crystal River-3 incident, the Staff has decided that it should "de-emphasize quantitative risk assessment, but should emphasize diverse appli-cability of accident sequence analysis."
Mr. Taylor pointed out that the IREP should produce a " scrutable report with sufficient documentation to allow good peer review."
DISCUSSION WITH UTILITIES 1.
TVA ( D. Terrill)
TVA has responded to Mr. Denton's 50-54 letter last year (as reported to the Subcomittee in its Jan. 8,1980 meeting). TVA concurs with Mr. Denton's Jan. 22 letter which recommends to the Comission against halting of construction of B&W plants.
2.
WPPSS (A. Hosler)
Mr. Hosler presented a handout which provides a list of items WPPSS has comitted to work on as a result of Mr. Denton's 50-54 letter. WPPSS has briefed the Subcomittee in its last meeting on these items (see minutes, Jan.8,1980). The current list describes the status of implementation of these comitments.
3.
CONSUMERS POWER COMPANY - M. Salerno Mr. Salerno said that Midland construction is about 60% complete. He e
. stated that the Staff should have been supplied sufficient information to make a decision with regard to construction stoppage. Although the
' sensitivity' ' issue is not closed, he urged that it be pursued during the normal licensing process such that there will be no more delay in licensing reviews.
EXECUTIVE SESSION Mr. Etherington commended the ACRS task force for its efforts in writing the report.
lne Subcommittee decided that all items on the schedule should be presented
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to the full ACRS, which would meet during the same week.
The Subcommittee agreed to recommend that the ACRS write a letter to the Commission on the Committee's concurrence with Mr. Denton's Jan. 22,1980 memorandum, " Determination Whether B&W - Designed Plants Presently Under Construction Be Allowed to Continue."
No future meeting date was picked but subsequent to adjournment, Mr. Etherington picked April 29, 1980 for the Subcommittee to meet again to review NUREG-0667.
(Whereupon, the meeting was adjourned at 3:30 P.M.)
A complete transcript of the meeting is on file at the NRC Public Document Room at 1717 H St., N.W., Washington, D.C. or International Verbatim Reporters, Inc., Suite 107, 449 S. Capitol St. S.
W., Washington, D.C. 20002,202/484-3550 e
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ATTACHMENT 4 List of documents received before and during the meeting. Those that were in the public domain have been distributed in the meeting.
(Members and consultants have received a copy of everything.
For this reason, copies of these documents are not attached to these minutes; one copy of each, however, has been filed in the ACRS office).
1.
R. Tedesco, "B&W Transient Response Task Force Presentation Before ACRS Subcommittee on B&W Water Reactors."
2.
W. Jensen, " RECAP-4 Investigation of S&W Transients."
3.
B. Siegel, "ANL Sensitivity Study on B&W OTSG."
4.
J. Murphy, " Integrated Reliability Evaluation Program (IREP)."
5.
A. Hosler, " Recommendations for WNP-1/4."
6.
Tam to Subcommittee, " Status Report for the April 8,1980 ACRS Sub-committee Meeting on B&W Water Reactors."
7.
Draft " Transient Response of Babock & Wilcox - Designed Reactors",
NUREG-0667, April 2, 1980.
8.
Draft, " Review and Evaluating of the Babcock and Wilcox Nuclear Steam Supply System" (by the ACRS Staff task force).
l
Vcl. 45. No. 58 / M:ndry. M:rch 24. 1980 / Nstic:s Fed:ral Regist
[9102 opd:ted as required.no meeting Babcock and Wilcox their consultanta af s trac ss:b(cl.Covernment to the and other interested persons. -
schedule will also continue to be distributed through the current mailing sunshins Act.
in addition,it may be necessavy for service, the Federal Register and the AutLonty To Close Meetins This the Subcommittee to hold one or more Public Document Room, as before: no detzrminstion was made by the Committee Macasemsne otBeer pursuant to provtst closed sessions for the purpose of change is anticipated in this distribution.
exploring matters involving proprietaQ De recording will operate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a
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information.Ihave determined in day.Because the Commission schedule t
a nO was delegated the authority to make such accordance with Subsection 10(d) of th is subject to late changes, those who are determinstions by the Director. NSF. on Federal Advisory Committee Act(Pub.
July eagra.
- l. 92,-463). that, should such sessions b planning to attend a meeting should re-verify the status of the meeting M. Mecca Winkler, aquind. it is necessary to close these whenever possible.
g sessions to protect proprietary )(4).
Meetings willbe at1717 H Street l
Committee Management Coontinator.
f Information. See 5 U.S.C. 552b(c t
Further information regarding topics unless otherwise indicated.
' p,,,g,
Further details of meetings are mu.
to be discussed, whether the meeting available from NRC staff during regular has been cancelled or rescheduled the work hours at (202) 634-1410. At the end Chairman's ruhng on requests for the of the 80 day trialperiod. consideration opportunity to present oral statements will be given to extending this service to juCl. EAR REGULATORY and the time allotted therefor can bean W Wme] nuder.
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- OMMISSION e% a pMpaM te@m ard 1s.1mo.
Advisory Committee on Reactor j[, p,'t t ephone
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t Safecu:rds, Subcommittee on p
kbcock and Wilcox Water Reacto e; 202/634-1413) between 8:15 a.m. and office of the Secretory.
5:00 p.'n EST.
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'I Meeting ns ACRS Subcommittee on Babcock Deted:Marchis.tsso.
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and Wilcox Water Reactors willhold aW C Hoyk.
1 mesting on April 8,1980 in Room 1048 g,7 g,j,,,, y gg (DodetIsos. 50-325.Ed-3241 1717 H St NW., Weahington, DC 20555 Caroline Power & Ugttt Co.t Issuance to complete its review of the NRC Staff
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Study to determine whether of Amendmenta to Facility Operating l
construction should be halted on certain Advisory Committee on Reactor ne U.S. Nuclear Regula'ory Uconsee D & W plants because of sensitivity of.
Safeguarda. Ad Hoc Subcommittee on Commission (the Commission) has the once-through steam generator Three Mlle Island Unit 2 Acebent
. Issued Amendment Nos. 28 and 50 to (OTSC) to feedwater transients.
Notice of this 'neeting was published Action Plan; Meeting; Chant,e Facility Operating Ucense Nos. DPR-71 I
%e April 1-2.1980meetingof the and DPR-82 issued to Caro!!na Power L
'[ with th d
ACRS Ad Hoc Subcommittee on%reeLight Company (the licensee) which g
Mile Island. Unit 2 Accident Action Plan revised the Technical Speci c.ations for in the ederal R ro outlin Octobst 1.1979. (44 ER 56408), oral or willbe held in Room P-118.Phillips operation of the Brunswick Steam Building 7920 Norfolk Avenue, Electric Plant. Units Nos.1 and 2 (the written statements may be presented by b
f th bli rding wiu Bethesda, MD. starting at 8.30 a.m. each facility). located in Brunswick County, I
he rmitted ont# d"^"8 ose portions North Carolina.%e amendments are day.Noticeof thismeetingwas pubushed on March 17 (45 FR 17007) and effective as of the date ofissuance.
ept, an g stio may be do y all other items pertaining to the meeting ne amendments revise the Techn!
by msmbers of the Subcommittee,its remain the same as pubushed.
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g g 2) cons'ultants, and Staff. Persons desiring Dated:Mardis.1sso.
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to make oral statements should notify.
provide for systematic implementation r
the Dtsignated Federal Employee as far John C. Hoyle' of instrumentation modif2 cations, and (3) in advance as praticable so that Advisory Committee Management Officer.
eliminate the requirement for removing C
appropriate arrangements can be madepen,,m3w me M*3 the SRM shorting links" during core r
to cllow the necessary tims during the same cooe WM alterations with control rods withdrawn.
I mesting for such statements.
' He appucations for amendments
/
The agenda for subject meeting shall Announcementby Automatic comply with the standards and d
bs as follows:
Telephone Answeringm nquirements of the Atomic Energy Act 1
of1954, as amended (the Act), and the r
Tucsday. Aprila.1980 Annecv. Nuclear Regulatory Commission's rules and regulations.%e s
Commission has made appropriate f'
m on. UntMe Conclusion of Commission..
During the next 80 days, the NRC will findings as requiredby the Act and the s
U ##"'88 test the effectiveness of an automatic Commission's rules and regulations in 10 h
ne Subcommittee may meet in Telephone Answering Service as an CFR Chapter L which are set forth in the Executive Session,with any ofits additional method of providing current
!! cense amendments. Prior public notice a
consultants who may be present to information to the pub 11c concerning the of the amendments was not required I:
explore and exchange their preliminary scheduling of Commission meetings.ne since the amendments do not involve a opinions regarding matters which should telephone number to callis (202) 634-significanthazards consideration.
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b2 considered during the meeting.
ne Commission has determined that 1498.
a At the conclusion of the Executive ne schedule of Commission meetinge the issuance of the amendments willnot s
Session, the Subcommittee will hear wiB be recorded daily on or before 3 PM, result in any significant environmental presentations by and hold discussions of the date preceding the meeting and with npresentatives of the NRC Staff, D**D "D
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April 10, 1980 S'.!' NARY OF THE APRIL 8,1980 MEETING OF THE B&W REACTOR SUBCOMMITTEE Purpose of Meeting:
To continue the Subcomittee's review of the sensitivity of B&W reactor systems to feedwater transients.
Attendees:
ACRS NRC H. Etherington, Chairman D. R. Quick, Reg. II W. M. Mathis J. Murphy, PAS J. Ebersole A. Bournin, NRR J. Ray Bruce Wilson, NRR S. Lawroski F. Rowsome, PAS
- 1. Catton, Consultant T. M. Novak, DSS T. Theofanous, Consultant B. L. Siegel, DSS Z. Zudans, Consultant Z. P. Spets, DPM T. McCreless Darl Hood, OPM G. Young, Fellow Hal Ornstein, AE00 J. Stamepelos, Fellow Eward Blackwood, IE E. Abbott, Fellow V. Panciera, NRR W. Kastenberg, Fellow S. Israel, NRR P. Tam, DFE G. Zech, NRR J. Bickel, Fellow Nina DiPaolo, IVRI TVA B&W L. A. Haack C. W. Connell Dennis L. Terrill D. H. Roy D. W. Wilson D. A. Womack J. H. Taylor OTHERS R. R. Steinke Bruce Karrasch N. S. Porter, WDPSS Roland L. Reed A.G. Hosler, WPPSS John S. Shively M. J. Salerno, CP Co.
DsJ. Short R. M. Ham, CP Co.
R. O. Vosburgh David Long, WSPR Al Ahamkhani Bob Leyse, NSAC F. R. Miller, TE Co.
OTHERS T. D. Murray, Toledo Edison Chuck Domeck, Toledo Edison H. Filacchions, SAI Mary Gust, Shaw Pittman A. A. Garcia' SAI Ken Wilson, Duke Power N. M. Cole, MPR S. Corsanico, IVRI T
[ '
ADVISORY COMMITTEE ON REACTORS SAFEGUARDS 8&W WATER REACTORS SU8 COMMITTEE b
.TCNTATIVE MEETING SCHEDULE APRIL 8,1980 h.
APPROXIMATE TIME 8:30 a - 8:45 a EXECUTIVE SESSION _
yChairman'sOpeningStatement(Mr.H.Etherington)
- Review the Meeting Schedule Comments by the Subcomittee i
DISCUSSION WITH NRC STAFF 8;45 a - 9:00 a A.
Introduction (Mr. T. Novak) 8.
Staff Analysis of Sensitivity of B&W Plants to Feedwater Transients New developments since the January Subcommittee 9:00 a
-9:40 a 1.
Meeting.(W.Jensen) 2.
Progress Report: ANL Plant Sensitivity 9:40 a - 10:30 a Program (B. Siegel) 10:30 a - 10:40 a
- * * * * * * * * * * * * * * * * * * * *
- BREAK 3.
Pertinent Res61ts from the Integrated 10:40 a - 11:30 a i
Reliability Eyaluation Program Report by B&W Reactor Response Vul'erability 11:30 a - 12:30 p n
4.
Task Force (2-week study of the Crystal River-3 event) 12:30 p - 1:30 p
- LUNCH ANALYSIS DISCUSSION WITH UTILITIES BUILDING 88W PLANTS:
OF B&W REACTOR SENSITIVITY TO FEEDWATER TRANSIENTS 1:30 p - 2:15 p 1.
WPPSS 2:15 p - 3:00 p 2.
Consumers Power Company 3:00 p - 3:45 p 3.
TVA (Mr. D. Terrill)
- BREAK **********************
3:45 p - 4:00 p 4:00 p - 5:00 p DISCUSSION WITH ACRS STAFF (T. McCreless. ACRS Fellows)
( Results of Analysis Petformed by the ACRS Staff 5:00 p - 5:30 p EXECUTIVE SESSION Discuss recomendations to be made in an ACRS letter to the Comissio l
_