ML20205D089

Summary of ACRS 317th Meeting on 860911-13 in Washington, DC

ML20205D089
Person / Time
Issue date:	03/25/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2460, NUDOCS 8703300331
Download: ML20205D089 (191)

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TABLE OF CONTENTS h P[ pt.)? ljl j.. Ej N ij h MINUTES OF THE g,;,

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317TH ACRS MEETING i t.;

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I SEPTEMBER 11-13, 1986 WASHINGTON, D.C.

I.

C ha i ma n ' s Re p o rt ( 0 pe n )..........................................

1 II.

Babcock & Wilcox Nuclear Power Plant Long-Term Safety (0 pen)......

1 III.

Decay Beat Removal Subcommittee Report (0 pen).....................

8 IV.

Emergency Core Cooling systems (0 pen)............................

10 V.

Foreign Operating Experience (0 pen)..............................

14 VI.

Improved Light Water Reactors (0 pen).............................

22 VII.

Meeting with the NRC Commissioners (0 pen)........................

27 VIII. Seismic Oualification of Safety-Related Equipment in Operating Nuclear Plants (0 pen)............................................

30 IX.

NRC Incident Investigation Procedures (0 pen).....................

32 X.

Seismic Margins Program (0 pen)...................................

33 XI.

Executive Sessions (0 pen)........................................ 33 A. Subcommittee Assignments......................................

33 1.

ACRS Workload and Pesource Assignments...................

33 2.

Peacto r Ope rati on s.......................................

34 3.

ACRS Officers for Calendar Year 1986.....................

34 B. Reports, Letters and Memoranda................................

35 1.

ACRS Suggestions for an NRC Long Range Plan..............

35 2.

ACRS Comments on the Resolution of USI A-46 (Seismic Oualification of Equipment in Operating Plants)..........

35 3.

ACRS Corrents on the Proposed Revision to the ECCS Rule -

10 C FR 50. 46 a d A pp e nd i x K..............................

3 5 4.

Proposed Resolution of Generic Issue 124, " Auxiliary Feedwater Systen (AFWS)

Reliability".....................

35 5.

ACRS Comments on Degraded Piping Research................

36 DESIGNATED ORIGINAL 8703300331 870325 PDR ACRS

[ty' 2460 PDR Certified By i

4 a

ii a

C.

F u tu re A g e n d a.................................................

36 1.

Future Agenda............................................

36 2.

Future Subcommittee Activities...........................

36 D.

Seabrook Nuclear Power Plant..................................

36 E.

ACRS Meeti ng Dates for CY-1987................................ 36 8

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4

iii TABLE OF CONTENTS APPENDICES TO MINUTES OF THE 317TH ACRS MFETING SEPTEMBER 11-13, 1986 Appendix I Attendees................................................

A-1 Appendix II Future Agenda............................................

A-6 Appendix III ACRS Subcommittee Meetings...............................

A-8 Appendix IV NRC Presentation on B&W Plant Reassessment Program.......

A-14 Appendix V BWOG Presentation - Safety and Performance Improvement P ro g ra m.................................................. A-19 Appendix VI Category C Transient Ccncl usions......................... A-40 Appendix VII NRC Staff Presentation en Generic Issue No.

124..........

A-58 Appenoix VIII Revision of the ECCS Rule - 10 CFR.50.46 and Appendix K..

A-70 Appendix IX Regulatory Analysis - Revision of ECCS Rule.............. A-81 Appendix X Response to ACRS Comments on Revision of ECCS Rule.......

A-88 Appendix XI NRC/RES Proposed Methodology for Measuring Thermal-Hydraulic Code Uncertainty...............................

A-95 Appendix XII Chronology cf the Chernobyl Accident.....................

A-100 Apper. dix XIII Peport on the IAEA Meeting on the Chernobyl Accident.....

A-104 Appendix XIV Proposed Containment Performance Design Objective........

A-136 Appendix XV Additional Documents Provided for ACRS' Use..............

A-144

9 Fadtril Regist:r / Vol. 51, No.165 / Tuesday, August'28,7986 TNeitices /

30451.

_7 7

. si Patent App!! cation 804.039 Method for Procedure for the Determination of, "MNoece 86-441 *t'b"'D-5 "*f N

Machini: - Hofes in Composite Surface Emissivities: filed Deceraber

'-;~

Intent To Grant,a'n Excluelve Patent -

Materialdi icd December 3,1985.

3,1985.

Patent App!: cation 805.012: Quasi-Patent Application 846,429: Ice Detector;< Ucense ContainerInr Glass Formation filed March 31,1986.

Methud u..I Apparatus; filed Patent Application 840.825: Laser N ',. AoENCY:Nati6nal Aeronautics and December 5.1985.

Ranging and Video Display System:

Space Admwatration.

Patent App:icttion 815.099:

filed March 18,1986.

S.__ Acno#c Notice ofIntent to Grant an.

Neighborhood Comparison Operator:

Patent Application 846.430: Braille '

Exclusive Patent unnaa filed Dec: mber 31,1985.

Reading System; filed March 31,1986.

i Patent Application 815,103:

Patent Application 840,812: Semi ' suesstAnV:NAS$herebYgives notice of Programmcble Pipelined Image Interpenetrating Polymer Networks of intent to grant toErnest W.Mllen; Processo, filed December 31,1985.,

High Temperature Polymer Systems:

Seaford, Virginia, a limited, exclusive, Patent Application 815.105: Convoler; filed March 18,1986.

i royalty-bearing, revocable license to filed December 31,1985.

Patent Application 840,900: Oxygen

practice the invention as described in Patent Application 809,975:High Band Diffusion Barrier Coating: filed March.

U.S. Patent No. 4,586,140 for a "Afreraft.

Cap III-IV Tunneling Junction for 18,1986.

p.

+. Ilitmeter," which issued on April 29, Silicon Multijunction Solar Cells; filed Patent Application 834,978:.

IC *o the Admfnistrator of the December 17,1985.

Poly (carbonate-tmides); filed February Patent Application 805.011:

27,1986-

' National Aeronautics and Sp~ ace -

Administration on behalf of the United Reconfigurable Work Station for' a Patent Application 838,655: Process for States of America.The proposed Video Display Unit and Keyboard:

Crosslinking and Extending exclustve license will be for a limited filed December 5,1985.

Conjugated Diene-Containing '

number ofyears and will contain -

Patent Application 798,*13: Liquid Polymers; filed March 11.1986.

. appropriate terms and conditions to be Hydrogen Polygeneration System and Patent Application 838.654: Process for negotiated in accordance with the ~

Process; filed November 15,1985.

Cross-Unidng Methylen@n'aining NASA Patentlicensing Regulations '14 h

Patent Application 751,644: Personnel Aromatic Polymers with Ionizing Emergency Carrier Vehicle; filed July Radiation: filed March 11,1986.

CFRPart1245 Subpart,2.NASAwill negotiate the final terms and conditions 3,1985.

Patent Application 846,428: Liquid m-Patent Application 790,597: Tool and.

Seeding Atomizer: filed March 31, and grant the exclusive license unless, Process for Explosive Joining of 1986.

within 60 days of the date of the Notice, Tubes: filed October 23,1985.

Patent Application 846.439:Swashplate the Director of Patent Licensing receives Patent Application 775,989: Acoustic Control System: filed March 31.1986.

pen $e to grant, together Radiation Stress Measurement; filed Patent Application 846,437: Dual Mode

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September 13,1985.

Leser Velocimeter: filed March 31, s

iH iew Patent Application 806,572:

1986.

all written responses to the Notice and Aminophenoxycyclotriphosphazene Patent Application 831,193: Method and ; then recommend to the Associate Cured Epoxy Resins and the Apparatus for Measuring Distance; General Counsel (Intellectual Property)

Composites Laminates. Adhesives filed February 20,1986.

_ wheth.er to grant the exclusive license.

and structures thereof; filed November Patent Application 852,468: Variable -

oAva: Comments to this notice must be 21,1985.

Energy High Flux, Ground-State received by October 27,1986.

Patent Application 823,712: Airborne Atomic Oxygen Source: filed April 10 Aconses: National Aeronautics and Tracking Sun Photometer Apparatus

. 1986.

Space Administration, Cooe GP.

and System: filed January 29,1985.

Patent Application 855,982: Oxygen Washington, DC 20646.

Patent Application 838,648: Floating Chemisorption Cryogenic Refrigerator:

- Act.

Emitter Solar Cell Junction Transistor; filed April 24,1986.

m mmen maAmW filed March 11,1986.

Patent Application 834.977: Oxidation Mr. John G. Mannixd202) 453-243a Patent Application 802,769: Method of

. Protection Coatings for Polymers; filed Deted: August 14.1988. g, Measuring Field Funneling and Range February 27,1986.

Edward plFrankle,

[

Straggling in Semiconductor Charge-Patent Application 855,983: Lightning Depu(yCenemlCouns'el "

l Collecting Junctions; filed November Discharge Protection Rod; filed April FR d so-te17e Fued 8 25 86:545 am]

27,1985.

24 Patent Application 831,371:Deployable~

Paten' 1986.t Application 83'2,296: Heat.

'"/** **** " " -

~

Geodesic Truss Structure; filed Treatment for Superalloy; filed February 20,1986.

Patent Application 831,372: Inductive-February 24,1986.

UCLEAR REGULATORY Patent Application 855,879: Polyether.

Energy for Rapid Strain Gauge Polyester Graft Copolymer; filed April' COMMISSION

Attachment:

filed February 20,1986.

24,1986.

Patent Application 829,042: Ultransonic Patent Application 838,649: Active

'. Advlaory Committee on Reactor Depth Gauge for Liquids Under High Control of Boundary Layer Transistor Safeguards; Meeting Agenda Pressure; filed February 13,1986.

and Turbulence; filed March 11,1986.

Patent Application 831,377: Adjustable Patent Application 765.991: Planar In accordance with the purpose sections 29 and 182b. of the Atonu,s of Mount for Electro-Optic Transducers Oscillatory Stirring Apparatus; filed c

in an Evacuted Cryogenic System:

August 15,1985.

Energy Act (42 U.S.C. 2039,2232b), the filed February 20,1986.

Advisory Committee on Reactor Patent Application BGl.196: Flat-Panel,.

Dated: August 15,1986.

Safeguards will hold a meeting on Full-Color Electroluminescent Display: Edward A.Frankle, September 11-13,1986, in Room 1046, filed December 3,1985.

DeputyGenem1 Counsel 1717 H Street, NW, Washington, DC.

(

Patent Application 804,040

[TR Doc. 86-19177 File:18-25-86; 8:45 am]

Notice of this meeting was published in Measurement Apparatus and

~ the Federal Register on August 19,1988.

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Federal Register / Vol. 51 No.185 / Tuesday. Au;;ust 28, 1986 / Notices

/

rsday. September 11. Isac regarding plans for review of the long-portions of the meeting when a k

30 A.M445 ARrReport ofACRS term safety of B&W nuclear plants.

transcript is being kept, and questions s.nairman (Open}-%e ACRS Chairman.

P rtions of this session may be closed may be asked only by members of the will report briefly regarding items of as necessary to discuss Proprietary Committee, its consultants, and Staff.

current interest to the Committee.

Information related to B&W nuclear Persons destnng to make oral 8:45A M-11:45 AMrImprovedlight, plants.

statements should notify the ACRS WaterReacton (Open}-The members 2:1 PM415PM Long-Range Executive Director as far in advance as fesC Planning (Open}-%e members of the practicable so that appropriate m miHet w i828s Pmpond ACRS ~ arrangements can be made to allow the ACRS comm n s en ti r, mments and recommendations necessary time during the meeting for b

is tha NRC r.e8ardin8 proposed che eteristics ofimproved light-water regarding the preparation of a long-such statements. Use of still, motion range plan for NRC activities.

picture and television cameras during 5:15 AR4mPXiACRS this meeting may be limited to selected 11:45AR-12MPXiPhtrire ACRS -

Subcommittee Actirities (Open)--%e Activities [Open)-%e members will

$ortions of the meeting as determined discuss anticipated ACRS subcommittee. members win bear and discuss reporta y the Chairman. Information regarding mrstings and items proposed for fulf of designated ACRS subcommittees the time to be set aside for this purpose

)

Committee consideration.ne schedule regarding safety-related matters.

may be obtained by a prepaid telephone for ACRS full Committee meetings for including the NRC incident investigation call to the ACRS Executive Director, CY 1987 wiH also be discussed.

pmgram activities of the NRC Office of R.F. Fraley, prior to the meeting. In view Inspection and Enforcement, and of the possibility that the schedule for j#PE. pf,cs.

evaluation ofseismic margins wilk ACRS meetings may be adjusted by the M =e NRg I

(Open)-%e members wiH discuss the.

NW n M

m-Chairman as necessary to facunate sa presentation ofits report dated August Satuidoy,Septemberit Jagg -

conduct of the meeting, persons I*

i8

"' d s oul ck whh ee 12.1986 (Revised 8/15/86) on the aA.y_ympxipreparadon of

$CRS Ex proposed NRCpolicy statement on

. ACASReports (Open/ Closed)-De ve D standrrdization of nuclear power plants. members will discuss reposed ACRS rescheduling would resultin major MPR-MPXJMeeting with reports and memoran to the NRC inconvenience.

NRC Commissioners (Open}--

~

regarding items considered during this I have determined in accordance with Presentation and discussion of ACRS meeting. In addition. proposed ACRS subsection 10(d) Pub. I.92-463 that it is report dated August 12.1986 (Revised 8/' ' comments on seismic qualification of necessary to close portions of this 15/86) on the proposed NRC safety-related equipment in nuclear meeting as noted above to discuss udardization Policy Statement.

power plants and use of aptitude testing Proprietary Information [5 U.S.C.

15PX-mP.MJKmergency Core in the selection of nr. clear pows plant 552b(c)(4)] applicable to the facilities

.dingSystems (Open)-%e members 2 personnel will be discussed.

being discussed and information the will hiar presentations and discuss Portions of this session will be closed release of which would represent a.

Proposed changes in NRC regulatory as ascessary to discuss Proprietary clearly unwarranted invasion of requirements for emergency core cooling Information applicable to the matters Personal privacy [5 U.S.C. 552b(c)(6)}.

systems. Representatives of the NRC being discussed.

Further information regarding topics St:R will participate in this discussion..

1:Mpm.4mpmaACRS to be ea-- A. whether the meeting mPR-mPR:PrunarySysisar Subco runittee Activities (Open/

has been canceled or rescheduled, the

-l Integrity (Open}-%e members win a. Closed)-De members will hear and Chairman's mling on requests fw the bztr cad discuss the report ofits -

discuss reports ofits subcommittees on oppwtunity to present oral statements subcommittee regardag researd management and conduct of ACRS and the time allotted can be obtained by tetivities related to the integrity of the ^' - activities. including the prioritization

• prepaid telephone call to the ACRS printry coolant systems in nuclear

' and allocation of ACRS resources and Executive Director Mr. Raymond F.

powIr plants.

Priday, September 12.15Er ~

the non-ACRS activities ofindividual -

Fraley (telephone 202/634-3265).

"{,be,,"

between 8:15 a.m. and 5so p.m.

, ns of this session will be closed Dated: August 21.1988.

M AM-M AX *DscoyHeal.

as necessary to discuss information: the John C. Hoyle.

Removal (Open}-He members wiB release of which would represent a hear and discuss a Subcommittee report clearly unwarranted invasion of Adrisery. Committee Management OffYear.

regirding activities related to resolution personal privacy.

[FR Doc, as-19283 Filed 6-2He; etes aml af Unresolved Generic issue 124. -

22p.m.4mPXJMiscellaneous Auxili:ry Feedwater System Reliability.

(Open/ Closed}-De member wid MImbers of the NRC Staff will complete discussion of matters noted Participate as appropriate.-

above.

(Docket No. 50-4101 PSAX-Ju A.Malnternational Portions of this session willbe closed. Niagara Mohawk Power Corpa Nine Oberating Erperience (Open}-Briefing. as necessary to discuss Proprietary Mlle Point Nuclear Station, Unit 2; by m2mber of the U.S. Team regarding Information applicable to the matter Environmental Assessment and ths sequences which contributad to the being discussed.

Finding of No Significant impact Chernobyl Nuclear PowerPlant Procedures for the conduct of and -

cccid:nt...

participation in ACRS meetings were ne U.S. Nuclear R'egulatory l

JMS AR-122 Noon and12P.AL-published in the Federal Register on Commission (the Commission)is

• f5 P.Ma Babcock and WIIcox I.ight.

October 2.1965 (50 FR 191). In considering issuance of exemptions from

!(-

terReactorSofety(Open/ Closed}-. accordance with these procedures. oral certain requirements of 10 CFR Part 50

' msmbers will hear and rNenas a or written statements ir ay be presented to the Niagara Mohawk Power

,.csentation by representatives of the.. by members of the public. recordings Corporation (the applicant) for the Nina Babcock and WUcox Company.,

will be perm!tted only during those.

Mile Point Nuclear Station. Unit 2 i

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'c, UNITED STATES NUCLEAR REGULATORY COMMISSION E"*,

'j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS E

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WASHINGTON, D. C. 20555

%,...,+f Revised:

September 4, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 317TH ACRS MEETING SEPTEMBER 11-13, 1986 WASHINGTON, D. C.,

Thursday, September 11, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 - 8:45 A.M.

Report of ACRS Chairman (0 pen) 1.1) Opening Statement (DAW) 1.2) Items of current interest (DAW /RFF) 2)

8:45 - 11:30 A.M.

ImprovedLight-WaterReactors_(0 pen)

BREAK: 10:00-10:15) 2.1) Discuss proposed ACRS comments /recomen-dations regarding the characteristics of improved LWRs (00/RKM) 3)

11:30 - 12:00 Noon Future ACRS Activities (0 pen) 3.1) Anticipated ACRS Subcomittee activities (MWL) 3.2) Proposed ACRS activities (DAW /RFF) 3.3) Proposed ACRS meeting dates for CY 1987 (DAW,etal/RFF) 12:00 - 1:00 P.M.'

LUNCH 4}

1:00 - 1:30 P.M.

Primary System Integri_ty (0 pen) 4.1) Report of ACRS Subcomittee on f.etal Components regarding research activities related to nuclear power plant primary system integrity (PGS/EGI)

Discussion regarding Meeting with Commissioners 5) 1:30 -

1:50 P.M.

Topen) 5.1) Discuss presentation regarding ACRS report dated August 12, 1986 (Revised, August 15, 1986) on Proposed NRC Standardization Policy j

J Statement (CJW/HA)

J Meeting with NRC Comissioners (Room 1130-H)

/

tr; 2:00 - 3:30 P.M.

(0 pen) 6.1) Presentation and discussion regarding ACRS report on Proposed NRC Standardization Policy Statement dated August 12, 1986 (Revised, August 15,1986) 3:30 - 3:45 P.M.

BREAK p ') 'I l?

f

4 317 h ACRS Meeting Agenda,

7) 3:45 - 5:45 P.M.

EmergencyCoreCoolingSystems(0 pen) 7.1) Report of ACR5 Subcommittee regarding proposed changes in 10 CFR Part 50.46, Acceptance Criteria for ECCS for Light-Water Reactors and 10 CFR Part 50, Appendix K, ECCS Evaluation Models (CYM/PAB) 7.2) Meeting with NRC Staff 8) 5:45 - 6:45 P.M.

Seismic Oualification of Safety-Related Equipment in Operating Nuclear Plants (0 pen) 8.1) Discuss proposed ACRS report to NRC (CJW, et al/RKM) 8.2) Discussion with representatives of the NRC Staff

317th ACRS Meeting Agenda.

Friday, September 12, 1986, Room 1046, 1717 H Street, NW, Washington, D.C..

9) 8:30 - 9:30 A.M.

Decay Heat Removal (0 pen) 9.1)

Report of ACRS Subcommittee on Decay Heat Removal regarding resolution of USI A-124, Auxiliary Feedwater Systems Reliability (DAW /PAB) 10) 9:30 - 11:45 A.M.

ForeignOperatingExperience(0 pen)

(BREAK: 10:30-10:45) 10.1) Briefing by representative of U.S. Team regarding the Chernobyl Nuclear Power Plant accident

11) 11:45 - 12:45 P.M.

Babcock & Wilcox Nuclear Power Plant Long-Tem Safety (0 pen / Closed) 11.1) Opening remarks by Chairman, ACRS Subcom-mittee on B&W Reactors (CJW/RKM) 11.2) Presentation by representatives of the B&W Owners Group regarding evaluation of the long-tem safety of B&W reactors (Note:

Portions of this session will be closed as required to discuss Proprietary Infomation applicable to this matter.)

12:45 - 1:45 P.M.

LUNCH

11) 1:45 - 3:00 P.M.

Babcock & Wilcox Nuclear Power Plant Long-Tem Safety (0 pen / Closed) 11.3) Continue discussion noted above (Note:

Portions of this session will be closed as required to discuss Proprietary Infomation applicable to this matter.)

3:00 - 3:15 P.M.

BREAK

12) 3:15 - 3:30 P.M.

ACRS Subcomittee Activities (0 pen)

Report of ACR5 Subcommittee regarding NRC 12.1) incident investigation procedures (HWL/GRQ)

13) 3:30 - 4:15 P.M.

Seismic Margins Program (0 pen) 13.1)

Discuss results of ACRS August 6, 1986 sub-comittee meeting on seismic margins in the design of nuclear power plants (D0/RPS) 4:15 - 4:30 P.M.

BREAK i

317th ACRS Meeting Agenda,.

14) 4:30 - 6:30 P.M.

Long-Range Planning (0 pen) 14.1) Discuss proposed ACRS coments and recom-mendations regarding preparation of a long-renge plan for NRC activities (MWC/RKM) 4 i

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317th ACRS Meeting Agenda -

0 Saturday, September 13, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

e

15) 8:30 '--12:15 A.M.

ACRS Reports to NRC (0 pen / Closed)

(BREAK: 10:00-10:15) 15.1)

Discuss proposed ACRS reports on:

15.1-1) 8:30-9:30:

Seismic Qualification of Equipment in Operating Nuclear Plants (CJW/RKM) 15.1-2) 9:30-10:00: B&W Long-Tem Safety Review (tentative) (CJW/RKM) 10:00-10:15: BREAK 15.1-3) 10:15-10:45: Auxiliary Feedwater System Reliability (tentative)

(DAW /PAB) 10:45-11:15: Primary S Integrity (tentative) (ystem 15.1-4)

PGS/EGI) 15.1-5) 11:15-11:45:

ECCS Requirements TCYM/PAB) 15.1-6) 11:45-12:15: Long-Range Plan (MWC/RKM)

(

12:15 - 1:15 P.M.

LUNCH

(

16) 1:15 - 2:00 P.M.

ACRS Subcomittee Activities (0 pen) 16.1) 1:15-1:45: Report of ACRS Subcommittee regarding Phase I of the NRC Maintenance Program Plan (CYM/HA) 16.2) 1:45-2:00: Report of ACRS Subcomittee regarding activities of the NRC Office of Inspection and Enforcement (CYM/PAB)

17) 2:00 - 3:15 P.M.

ACRS Subcommittee Activities (0 pen / Closed) 17.1) 2:00-3:00: Report of Subcommittee meetings regarding ACRS Management / Planning (July 9, and September 10,1986) and ACRS Procedures and Administration (August 6, 1986)

(DAW /RFF)

(Note:

Portions of this session will be closed as required to discuss informatien the release of which would represent a clearly unwarranted invasion of personal privacy.)

17.2) 3:00-3:15: Nomination of ACRS Officers -

ACRS Chaiman appoint Nominating Panel for ACRS officers for CY 1987/88 (DAW /RFF)

18) 3:15 - 3:30 P.M.

Miscellaneous (0 pen / Closed) 18.1)

Complete discussion of items considered during this meeting.

(

>4

.L n ' ~l ; 4 ;-u 3

MINUTES OF THE fi j;;;;

317TH ACRS MEETING SEPTEMBER 11-13, 1986 The 317th. meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, U.k., Washington, D.C., was convened by Chairman D. A. Ward at 8:30 a.m., Thursday, September 11, 1986.

[ Note:

For a list of attendees, see Appendix I.

W. Kerr, G. A. Reed, and H. Etherington did not attend the meeting.

D. Okrent and F.,1. Remick did not attend on Saturday, September 13.]

Chairman D. A. Ward noted the existence of the published agenda for the meeting, and identified the items to be discussed.

He noted that the meeting was being held in conformance with the Federal Advisory Committee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively.

He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC's Public Document Room at 1717 H Street, N.W., Washington, D.C.

[ Note:

Copies of the Transcript taken at this meeting are also available for purchase from ACE-Federal Reporters, Inc.,

444 North Capitol

Street, Washington,D.C.20001.]

I.

Chairman's Report (0 pen)

[ Note:

R. F. Fraley was the Designated Federal Official for this portion ofthemeeting.]

Chairman D. A. Ward indicated that Kenneth M. Carr was sworn in for a five-year term as a member of the Nuclear Regulatory Commission on August 14, 1986.

He also mentioned the fact that ACRS member G. A. Reed is now recovering from a heart attack and will be unavailable to the Committee for at least two months.

II. Babcock & Wilcox Nuclear P'ower Plant Long-Term Safety (0 pen)

[ Note:

R. K. Ma,ior was the Designated Federal Official for this portion of the meeting.]

C.

J.

Wylie reminded the Comittee that the NRC made a decision in January 1986 to reassess the Babcock & Wilcox (B&W) plant design after several incidents occurred at B&P reactor plants.

At the urging of the NRC, the B&W Owners Group (BWOG) asemed a leadership role in that reassessment.

In March 1986, the NRC prepared a proposed NRC versico of the reassessment program and forwarded that program to the BW0G.

In June, the NRC comented on BWOG's plans taking account of the NRC input.

The BWOG Trip Reduction and Transient Response Improvements Program was discussed with the C&W Subcomnittee during a June 25, 1986 Subcommittee meetina.

The Subcomittee raised several substantive concerns which led the BWOG to postpone its meeting with the full Committee scheduled for July to the ACRS meeting in September.

During the July ACRS meeting, the

317TH ACRS MEETIMG 2

Committee discussed the BWOG plan as it was then structured, and recorded its concerns in a July 16, 1986 report.

On August 14, 1986, a Staff response to the ACRS letter agreed that a broader based program was needed and stated that the Staff was working with the BWOG toward that effort.

P. C. Jones, NRC, indicated that the Staff has been in contact with the BWOG at various working level meetings regarding 'several major topics.

ii. W. Carbon asked what reassessment of B&W plants means to the NRC Staff today.

R.

C. Jones indicated that the BWOG is to examine operating experience at B&W plants over the last several years to identify sensi-tive systems, kir.ds of problems that have occurred, kinds of behavior exhibited by B&W plants, and look in detail at specific problem systems or sensitive systems.

This broad-based review of the systems should cover their performance, the approoriateness of their design require-ments, and whether some changes ought to be made.

He presented an overview of the BWOG Safety and Performance Improvement Program (see Appendix IV). He indicated that the operating experience review involves HUREG and EDRI reports as well as LERs.

The BWOG plans to interview poerators and maintenance personnel to further identify concerns with d ant behavior. The BWOG has contracted with NPR Associates to study the sensitivity of B&W plants.

The BWOG will also conduct an analytical study to examine the response of B&W plants to trips and upsets and compare the behavior to that of other PWR designs.

M. W. Carbon asked whether this study will look at normal, abnormal, or severe transients.

R. C. Jones indicated that the primary fccus of the program is normal transients, such as instrumentation, or instrumentation and control system (ICS) generated transients, failures of the ICS power supply, loss of feedwater transients, and transients such as stuck-open main steam safety valves.

Also analyzed will be the demands upon the operators during these events.

They will review operating procedures as part of their system reviews to include the ICS/NNI main feedwater system, emergency feedwater system, auxiliary feedwater system, secondary relief system, and the instrument air system. With the exception of the instru-ment air system, all of these systems have caused complicated transient response in B&W plants.

C. flichelson asked if they are doing some systems interactions studies as they relate to the main feedwater system in particular.

R. C. Jones indicated that if transients have cccurred they will be identified through the LERs.

R. C. Jones indicated that the final part of the RUOG program is a recommendatico tracking system and implementation process.

Out of their pregram will come a series o# recommendations which they intend to track and monitor as these items are addressed by individual utilities.

He indicated that the NRC Staff believes that the BWOG program is generally on target, but the Staff needs further discussion in the area of the main feedwater system review and its scope.

The Staff is also closely moni-toring the area of human factors issues and the tracking system and its implementation.

317TH ACPS PEETING 3

R. C. Jones indicated that the prinary effort of the NRC Staff is to review the BW0G results.

The Staff will examine demands on operating i

personnel / procedures, see whether existing B&W PRAs reflect operating experience, and also look at the thermal-hydraulic response and sensi-tivity of B&W plants as they relate to the overall safety of the plant.

The Staff plans an initial SER in December 1986 with open items with supplements, as appropriate.

Completion of the program is scheduled for June 1987.

C. Michelson noted that the thermal-hydraulic response the Staff is interested in probably refers to nornal transient responses. He asked if the Staff plans to look at accident responses.

R. C. Jones indicated that the Staff may examine the area of steam generator tube rupture.

J. C. Ebersole asked if the Staff intends to examine the nonambiguous vessel level gauge at Arkansas Nuclear One, as well as the primary blowdown system being installed at Davis-Besse.

R.

C.

Jones indicated that level instrumentation as well as ICS instrumentation activities are under NUREG-0737 and will not be part of this program.

While the Staff is trying to make the program as broad based as possible, it does not intend to cover the Davis-Besse primary blowdown system because the Staff does not intend to look at new alternate decay heat removal concepts under this program.

D. Okrent expressed interest in the reliability of a reactor trip if the turbine generator does not trip.

R.

C. Jones indicated that, to the best of his knowledge, the turbines always trip.

One area the BWOG is to examine is an alternate relief system designed for the steam generator.

Continued operation of the turbine is one of the ways to try to minimize operation of the main steam isolation valves.- This is a way to keep the steam generator pressure down.

B&W plants, following a reactor trip, actuate the main steam safety valves unlike other PWRs, and there has been difficulty with those valves.

The BWOG is looking at a two-phased program which will improve the reliability and operability of those valves and, at the same time, minimize the challenge to the valves in the first place.

He noted one najor area of concern is overcooling transients with steam generator overfill. B&W plants appear to be more sensitive than other PWRs in that area.

One has to worry about possible pressurized thermal shock conse-quences.

H. B. Tucker, Duke Power Company (Chairman of BW0G Executive Committee),

explained that the BWOG Safety and Performance Improvement Program constitutes an action plan in response to the January 24, 1986 letter from V.

Stello expressing concern about B&W-designed plants.

G.

R.

Skillman, GPU Nuclear, admitted that the BV0G failed to establish recog-nition of the safety orientation of the program when it met with the ACRS Subcommittee. He mentioned three observations and recommendations in the ACRS letter to V. Stello on July 16, 1986.

These were that some B&W plants operate better than others, B&W plants respond differently from other PWRs, and apparently little attention is being given to decay heat renoval in B&W plants.

He identified four basic issues from the ACPS letter:

1) data base lessons, 2) program safety orientation, 3) safety significance of the coce-through steam generator sensitivity, and 4) energy production / removal imbalance--decay heat remeval (see Appendix V).

I

E 317TH ACRS liEETING 4

R. Skillman indicated that the BWOG Safety and performance Improvement Program is a full assessment of fundamental safety and operating issues at B&W plants and includes 13 major tasks.

The program consists of an independent sensitivity study of basic thermal-hydraulic plant charac-teristics.

It includes a detailed review of selected key systems.

He contended that the need for additional decay heat removal and capability is considered unnecessary.

The BWOG recognizes that the imbalance between heat production and removal is a key to understanding B&h units.

R. T. Glaviano, BWOG, indicated that the process of zeroing in on cperat-ing experience has meant a review of about 220 transient assessment /

performance repcrts.

To accomodate that review, the BWOG developed transient classification guidelines to judge the relative complexity of transients.

The central definitions of categories A, B, and C are that Category A events are events where the preferred or expected response of the plant is seen; Category B events are those where the preferred or expected range is exceeded (these events are of concern since they may be precursor events; and Category C events are those where abnormal response was clearly indicated.

J. C. Ebersole asked if Category B events include challenges to safety systems such as the auxiliary feedwater system.

R. T. Glaviano agreed that they did.

D. A. Ward expressed concern regarding the BWGG's ability to differentiate between complexity and seriousness. iie hypothesized an accident that was extremely serious but had a rather simple cause.

R. T. Glaviano played down the sophistication of the classification system as just a convenient tool for putting transients into relative terns.

He noted that the data shcw that the more complex the event the greater the challenge to the operating crew and to the plant systems.

D. W. Moeller asked at what point the BWOG ties the safety significance of the transients to the risk to the health and safety of the public.

R. T. Glaviano again stressed that the BWOG is using this method just to develop basic conclusions and recomendations to improve plant response.

R. T. Glaviano indicated that there were 10 Category C events defined.

There are four iistinct patterns as to what contributed to these events.

Two of the events occurred as a result of loss of offsite power.

Follow-ing the loss cf offsite power, the emergency feed system filled at an excessive rate to the natural circulation control set point.

When the generators reached that set point either the feed system shut down or the pumps kept running but the flow stopped. The B&W plant has a tendency to reheat and repressurize as a result of the onset of high pressure inj,.c-tion.

For both of these events a reheat repressurization follouir rapid refill of the steam generators resulted in the lifting of the Puc.

A second pattern was excessive steam flow through secondary plant valves, whether these were bypass or atmospheric dump valves.

These events generally terd to exhibit excessive steam flow which causes a aouirement for feedwater to maintain inventory. Feed flow is also increar a and the result is an excessive cooling event.

The third type is due to ICS/fiNI power problems.

There is a combination of excessive feed flow and steam flow to the steam flow control valves and emergency flow centrol valves

317TH ACRS liEETIrlG 5

in a partially cpen position.

These events resulted in excessive cool-ing.

A fourth category of events are associated with actuations of safety systems such as the EFIC and steam feed rupture control system.

R. T. Glaviano stated that preliminary conclusions from the review of transients indicate that the events are generally the result of excessive primary to secondary heat transfer with the steam flow through failed open valves, and the feed flow to steam generator inventory.

These events require additional operator attention to achieve plant control in balancing energy removal with energy production and controlling reactor coolant inventory and pressure due to excessive feedback from the secondary side of the piant.

The key to mitigating these events is to l

assure that steam flow and feed flow can be controlled from the control room.

This allows balancing of the production terms with the removal terms reducing feedback to the reactor coolant system.

It requires that plant control be maintained when ICS/flNI power is lost.

Corrective actions are planned in the area of design, maintenance, human interface, and operating experience (see Appendix VI).

F. J. Remick noted that in the case of Davis-Besse the Steam Feed Rupture Control System actuated and isolated the steam generator, then the valve stuck closed.

Would this be a

corrective action to assign to plant maintenance?

R. T. Glaviano indicated that it would be either design or maintenance, one of the two. The point is the ability to control the main steam flow path and the main feed flow path.

G. R. Skillman spoke about the sensitivity of BAW plants, noting that the tuning of the ICS, as well as some other items, has caused a great increase in the reliability and overall safety of the plants.

It is a plant-specific process.

C. Michelson commented that any system that is so arranged that it takes a fine tuning operation to make it function effectively does not sound like a well-designed system to begin with, since it can't stand small drifts and operator maladjustments.

G. R.

Skillman indicated that the ICS is a system that has a design function and needs to stay within its design limits.

D. Okrent irdicated that he thought it certainly valuable to review pricr experience and to see what corrective actions are strongly suggested, llevertheless, he wondered if the BWOG is considering the case of a prior transient and one additional failure occurring.

He asked if the BWOG is reviewing the B&W plants for their ability to recover from multiple failures other than those that have occurred.

G. R. Skillman indicated that one of the objectives of the sensitivity study is to look beyond the data to other kinds of situations as to quantify the relative ser.sitivity of B&W plants with respect to other PWR designs.

D. Okrent insisted that there were certain aspects of the B&W plant which leave it vulnerable to transients which include certain combinations of failures.

G. R. Skillman asked D. Okrent to be more specific.

D. Okrent noted that if there is less water in the steam generator one would have less time to recover fron an event that was not feeding water to the steam generator.

The Instrumentation and Control System in a B&W plant has already been

317TH ACRS FEETING 6

shown to cause multiple instances of incorrect control information to appear simultaneously and to create complex transients.

D. Okrent indicated that he is looking for a different philosophy driving the BWGG studies, one that considers multiple failure analysis rather than trying j

to prevent a single failure.

He suggested that that effort seems to be absent from the BUOG studies, and he noted that it would not be difficult to at least pid out vulnerable points in the plant.

C. Michelson questioned the BWOG effort regarding systems interactions.

E. Swanson, BWOG, indicated that he was associated with specific systen reviews in the BWOG program, such as the air system, emergency feedwater, steam system, ICS/NNI, and main feedwater system reviews.

He noted the ACRS' concern regarding removal of decay heat and indicated that the BWOG program is concerned with decay heat removal after shutdown.

He ex-i plained that the transients the BWOG has seen frcm experience data from power operation have varied.

The objective of the BWOG work regarding decay heat removal is to concentrate on those transients and balance decay heat removal.

He noted that decay heat is renoved from B&W plants in the same way as it is in other PWRs using the steam generators.

P&W plants have the same degree of defense-in-depth as other PWRs.

Even the core melt frequency for B&W plants does not differ much from that for other PURs.

C. Michelson pointed out that the coupling between the reactor core and the steam generator is rather unicue on a B&W plant.

From the steam generator out into the secondary, the situation is not much different from other PWRs.

There certainly is considerable differ-ence in how decay heat gets from the core to the steam generator.

E. Swanson contended that the difference between B&W FWRs and other PWRs is a matter of time.

C. Michelson noted that it is also the thermal-hydraulic a rrangement that is significantly different.

E. Swanson claimed at least three lines of defense, three lines or levels of defense for removal of decay heat within/without the steam generators.

The first line of defense is the main feedwater system, the second line is the emergency feedwater, and the third line of defense is bleed-and-feed capability.

He contended that bleed-and-feed is very powerful for B&W plants because of the very high-head high-flow pumps that are able to provide flow throuah the PORY or through the pressurizer safety valves.

l The Committee discussed the fact that at Davis-Besse there is enough pump net positive suction head to open the safeties but one cannot get flow into that head. As a result, Toledo Edison has to use the PORV.

D. A. Ward expressed concern that the thermal hydraulics of the bleed-and-feed process are not adequately understood and satisfactorily repre-sented in the emergency operating procedures that are in place at BtM plants.

E. Swanson admitted that the operators at Davis-Besse were a bit reluctant to engage bleed-and-feed. All plants have taken some action to better define procedures so that the operators will definitely engage bieed-and-feed when necessary.

C. Michelson wondered whether thermal-hydraulic understanding through calculations and computer codes is adequate.

E. Swanson insisted that the calculations that were made show that the process does work.

217TH ACRS MEETING 7

E.

Swanson attempted to address the ACRS concern regarding the main feedwater system.

He indicated that the emphasis of the BWOG program is to keep the main feedwater system on line.

The BWOG is looking at what degrades system performance and wishes to improve the reliability of the system and, in particular, the main feedwater pumps.

This is important to reduce emergency feedwater challenges which occur even when the main feedwater system is on line. The BWOG wants a snooth post-trip response.

It wants to control the ability to keep one pump on line if the other pump trips which is very closely tied to the tuning of the ICS and the main feedwater pump turbine and controls.

This is an important part of the BWOG program.

Another thing being done in the program is to try to minimize unnecessary feedwater pump trips, ano this is done by correcting the controls on the main feedwater system, and the ICS. Mention was made of the interaction between the turbine-driven pumps and steam pressure control.

If one does not control steam pressure after the trip the head output of the pumps is affected.

J. C. Ebersole commented that electric-motor-driven pumps, while costing more to run, are much more reliable than the turbine pumps and would remain steady in this situation.

J. H. Taylor, B&W, indicated that an extensive study of feed pump trips has been done by B&W.

He offered to provide a copy to the Committee.

J. C. Ebersole expressed interest in the report.

E. Swanson indicated that all B&W plants have initiation of emergency feedwater independent of the ICS power supplies.

He explained that the turbine-driven emergency feedwater pump reliability suffers because of the short startup time placed as a requirement on the design.

If B&W owners can lengthen the startup time to allow the pumps' speed to be increased more gradually from the low-speed stop to the high-speed stop it will improve or reduce the number of trips cn overfrequency or over-speed.

The ' BWOG will probably make such a recommendation to the NRC Staff.

He also noted that maintenance practices probably acccunt for a reasonable amount of unreliability of the pumps in the system.

Another way of increasing the reliability of the system is to reduce the number of challenges to the system.

G. R. Skillman indicated that the BWOG expcts, in its Safety and Ferfor-mance Improvement Pregram, to develop about 500 recommendations from the 13 main tasks.

These will be available by the secord ouarter of 1987.

It is the intent of the BWOG to dispose of all the recorrendations formally and to include che NRC Staff in the disposition of the recommendations.

He noted that the implementation phase of appropriate changes will have to fit with the various cutage schedules and other key activities and commitments of the utilities.

He expected that the implementation phase will continue for several years.

P. G. Shewmon noted that the Oconee units did not have any Category C events. He asked if the BWOG has identified tow things are being done differently at the Oconee plants than at other B&W plants, and how this might be factcred into the program.

H. B. Tucker indicated that, while it is true that no Category C or abnormal response events have occurred at the Oconee plants, there are several precursor events or Category B events that were

317TH ACRS MEETING 8

of significance at Oconee.

The position of the BWOG is to take the collective response of all the B&W plants to learn the lessons that there are available and apply them to all of the B&W plants.

C. Michelson asked how the BWOG was addressing the ouestion of steam generator overfill in this review.

G. R. Skillman indicated that it will be addressed partially through the main feedwater system review and, to a large degree, during work on the secondary plant relief systen, instrument air, and others.

C. J. Wylie asked if the BWOG program, as presented, resolves the Commit-tee's previous concerns as expressed in its letter of July 16, 1986.

C. Michelson thought it premature to endorse the BWOG program withcut caveats.

He remained skeptical about a lack of emphasis on systems interactions in the BWOG program.

D. W. Moeller thought that the BWOG has taken the ACRS concerns in the July 16 letter seriously and addressed them. He wondered, however, if the BWOG had studied all the PRAs for B&W plants and produced findings.

G. R. Skillman indicated that the BWOG has structured its PRA review around the Class 3 PRA at Crystal River and the Class 1 review at Oconee.

The BWOG is in the process of taking the findings from those two PRAs and applying them to Davis-Besse, Rancho Seco, TMI-1, and Arkansas Nuclear One.

D. F. Moeller thcught it useful for the Committee to hear periodic reports as progress is made in the BWOG program.

J. C. Ebersoie expressed concern regarding the control of the bleed-and-feed process for B&W plants.

P. G. Shewmon mentioned cycling tests done at Davis-Besse.

C. Michelson noted that they did not cover the full spectrum of fluH conditions.

M. W. Carbon expressed unease regarding the spectrum of accident scenarios examined by the BWOG progran.

He thought that they shculd cor. sider accidents which haven't happened yet, and they are only looking at historical experience.

He thought that a key question is if the EWOG is doing a systems inter-actions study of the type that will reveel these new pcssibilities.

No one present was aware of any B&W systems interactions studies that have been published.

Chairman Ward noted that there did not appear to be any interest on the Committee to write another letter at this time.

He did express the interest of the Committee in hearing more about the program in the future.

III. Decay Heat Removal Subcommittee Report (0 pen)

[ Note:

P.

A.

Boehnert was the Designated Federal Official for this portion of the meeting.1 D. A. Ward indicated that the Subccamittee on Decay Heat Renoval Systens held a meeting to review the NRR resolution effort for Generic Issue 124

" Auxiliary Feedwater System Reliability."

He explained that, prior to 1975, auxiliary feedwater systens were built to conform to good engineer-ing practice.

There were no specific AEC or NRC requirements.

After 1975 a requirement was developed that new auxiliary feedwater systens be safety grade, appropriately seismic resistant, subject to 0A require-ments, and tolerant of single failures.

After the Three Mile Island

r 317TH ACRS MEETIfiG-9 accident, considerable reevaluation and rethinking of the importance of auxiliary-feedwater systems led to a new requirement after July 1981 for new. plants.

They must show that their auxiliary 4feedwater system should have an unavailability upon demand less than 10 This requirement was placed upon applicants for operating licenses rather than for con-struction permits and was placed in the Standard Review Plan. Twenty to twenty-five units have been licensed under this provision, which means that 75 or 80 ' plants do not-necessarily meet this requirement.

This concern was focused in Generic Issue 124.

The focus is on 7 plants in which there is a particular concern about unreliability of the systems.

All are single units which have two-train -auxiliary feedwater systems.

While a number of other units have two-train auxiliary feedwater systems, they are on multi-unit sites and credit is given to ' certain cross-connections between the units. The Cemittee reviewed the Staff program and in December 1985 wrote a letter which pointed out that the ACRS thought the Staff's resolution for this generic issue was not adequate,.

the schedule not prompt enough, and.the plan not well enough focused.

The Comittee also; complained that operating experience with regard to the performance of auxiliary feedwater systems had not been adequately analyzed for useful information. The Subcommittee cencluded on September

' 9,1986, after. hearing presentations by NRR and AE0D, that the work on analyzing data on actual operating experience is proceeding better than in the last year.

The Subcomittee does have some concerns - with the approach being taken to review in some detail the systems at each of these seven plants and to recomend specific fixes in hardware or proce-dures.

D. A. Ward scored the lack of development of objective guidelines for making judgments as to what fixes are needed.

A more objective basis is needed for deciding whether systems are adequate.

He noted that MRR will propose plant-specific hardware-oriented fixes for the original seven plants.

J. C. Ebersole posed _a scenario in which loss of main and auxiliary feedwater is assumed at the outset.

Relief is accomplished through the PORVs. At some point the flow path shuts down to the point where removal of decay heat can cnly be effected' by relief aided by heat transport through the secondary systen.

But loss of the secondary systen is assumed at the beginning of the event and there is now a lack of options for the reduced relief through the primary system.

He asked the NRC Staff how they would deal with such a scenario, fSubsequent to the neeting the Staff agreed to discuss this issue at a future Subcomittee neeting on the resolution of USI A-45.]

S. S. Diab, NRR, presented the status of a modified resolution approach to Generic Issue 124 (see Appendix VII).

He indicated that the modified resolution approach basically consists of short-term, concentrated reliability reviews for each of the seven plants followed by a findings report.

The review effort will berefit from ongoing tasks like the Rancho Seco restart effort and the BWOG design reassessment effort.

The HRR team, the Auxiliary Feedwater Systen Revis.w Team, will also benefit from the IE program for Safety System Functional Inspections, and

317TH ACRS MEET 9!G 10 licensee auxiliary feedwater systems reliability analyses that are being done, as well as relevant industry efforts.

S. S. Diab discussed the status of each one of the seven plants in the modified resolution approach, as far as their auxiliary feedwater system reliability studies are concerned.

[The seven plants are ANO-1 and

-2, Rancho Seco, Crystal River, Prairie Island-1 and

-2, and Fort Calhoun.]

l'e indicated that the Staff is also reviewing the BWOG design reassess-ment. This assessment is attempting to improve the reliability of main i

feedwater systens, to improve the reliability of auxiliary feedwater l

systems, and to limit the challenges to the auxiliary feedwater system 3

(reduce scrams).

D. Okrent asked if the Staff's review for Generic Issue A-124 will be sufficiently detailed to get into dependencies of not only support systems but support systems of support systems.

S. Diab dis-cussed the composition of a Staff auxiliary feedwater system review team, as well as the review's scope, which he indicated would cover support systems such as power supplies, compressed air or nitrogen systems, lubrication, and cooling.

The Staff will review all of the post-TMI modifications, paying particular attention to finding common-mcde vul-nerabilities of the design or the arrangement of the equipment.

The reviews will cover operator recovery, control room adequacy for indi-cation control and recovery, ease of 1.0CA recovery, and alternate decay heat removal means.

C. Michelson asked if the Staff will include a fire analysis as well as pipe-break analysis in the feedwater area.

S. Diab indicated that the Staff will be looking at the functional reliability, as well as the environment.

C. Michelson asked if the Staff will look at fire protection around the auxiliary feedwater turbine and the prcbabil-ity of inadvertent actuation as a factor in reliability.

S. Diab indi-cated that the Staff plans to do that.

W. Minners indicated that the Staff does not wish to promise that they will do a fcnral quantitative i

reilability analysis on each plant.

If that analysis is available and uses plant-specific data

+'le Staff will utilize it.

But, based upon schedule and resources, the Staff does not plan to do a formal cuantita-l tive, detailed reliability analysis for each of the seven plants.

C. Michelson expressed interest in the team reports as they are generat-ed.

W. Minners indicated that the Staff could provide them.

D. A. Ward still remained concerned regarding the continually changing definition of

" reasonably assured reliability" or " sufficient reliability."

IV. Emergency Core Cooling Systems (0 pen)

[ Note:

P.

A.

Boehnert was the Designated Federal Official for this portion of the teeting.]

W. Beckner, NRC, indicated that the existing errergency core cooling system (ECCS) Rule 10 CFR 50-45 was based upon knowledge available in 1975. Even at that time, the Staff knew that parts of Appendix K to the Rule were very conservative, particularly the calculation of decay heat.

Conservatism was left in the rule to cover the overall uncertainty and the understanding of the overall performance of ECC systems at that tire

M7TH ACRS MEETING 11 (see Appendix VIII).

Additional features not specifically called out in Appendix K to the Rule have been typically treated very conservatively, either by the Staff's conservative requirements or by the licersee, or applicant, proposing simplified or conservative models.

The Staff believes that the current overall ECCS evaluation model is very conserva-tive and that calculated temperatures during a LOCA, using Appendix K models, are much higher than would be expected in reality.

The Staff now believes that distortions created by the use of these artificial conservatisms in Appendix K may adversely affect the overall safety of plant design and operation.

W. Beckner explained that the existing ECCS rule is specific and pre-scriptive.

All - errors must be immediately reported to the NRC and a reanalysis is required if there is a significant error, even when no regulatory or safety threshold is surpassed.

Errors of 20 degrees in peak cladding temperature generate a reanalysis even if the temperature does not exceed 2200 F.

W. Beckner indicated that there is broad support within the NRC Staff, the ACRS, and industry with no indication of any outside opposition to revising this rule.

There is also broad support for the basic approach which is based upon SECY-83-472 (Realistic Evaluation Model with an Uncertainty Evaluation).

This is an approach originally proposed by the General Electric Company and used in SAFER.

W. Beckner described the proposed rule revision.

He noted that calcula-tions would still be required of ECCS performance.

However, the analyt-ical technique would attempt to describe the behavior of the reactor systems realistically as defined by comparisons to applicable experi-mental data.

Uncertainty of the calculation would be accounted for so that there would be a high prcbability that the 2200 F criteria would not be exceeded.

Existing evaluation models using Appendix K would be grandfathered.

The grandfathering is indefinite if desired by the licensee.

Error reporting would be modified to make a reevaluation commensurate with the effect of the change or error, whereas the current rule gives the Staff an option to simply shut down a plant because of an error or exceedence of the 2200 F criterion.

The proposed rule allows a utility to de-rate the plant rather than totally shut it down if some i

l problem occurs. All errors and all changes made to the evaluation models should be reported at least annually.

If the change is significant there would be a report required within 30 days along with a schedule for the reanalysis.

D. W. Moeller noted that these requirements give the im-pression that each individual utility is doing a separate, totally independent analysis of their specific plant as contrasted to vendors doing a generic calculation.

W. Beckner indicated that in reality the vendor typically holds the model and makes changes to it, but the licen-see is submitting his model with the vendor acting as his agent.

W. Beckner indicated that the Dougall-Rohsenow correlation is removed from Appendix K and another heat transfer correlation reference is

- ~ _ _.

317TH ACRS MEETIliG 17 updated.

The Staff has found that the Dougall-Rohsenow correlation can be nonconservative in certain areas.. As a result, the Staff intends to monitor -evaluation models as approved and if there is a significant reduction in conservatism, defined as more than 50 F, the Staff would start to look at such things as Dougall-Rohsenow and pronounce it unacceptable in any area where it is not conservative.. The Staff has made changes in documentation requirements to remove the 20*F definition of a significant change, as well as putting into Appendix K explicit reporting requirements for errors.

C.tiichelson noted that the old requirement of 20 F now becomes 50 F even for the old Appendix K' users.

W. Beckner indicated that the ECCS rule revision package contains three pieces:

Revised Rule Regulatory guide Compendium of ECCS research (Sumary Report by the Office of Re-search of research performed over the past 10 years)

I W. Beckner indicated that the rule itself is cereral and does not refer-l ence either the Regulatory Guide or the compendium. The Regulatory Guide

~

expands upon the Rule by giving guidance on what is acceptable to meet the Rule in practice; if the applicant or licensee follows the guidance of the Regulatory Guide, the NRC will not mount a challenge. The appli-cant, or licensee, can propose another method of meeting the rule.

The Rule pemits the use of the latest technology, specifically best-estiriate calculations, combined with uncertainty evaluations based on data comparisons.

The use of such realistic calculations shculd lead to more unierstandable regulations and, hopefully, will be of benefit to safety.

The Regulatory Guide proposes acceptable models and data related to Appendix K, but allows flexibility for the Staff to accept industry initiatives.

It defines requirements for estimating overall code l

uncertainty at the 95 percent probability limit.

W. Beckner indicated that the proposed Rule is still being studied by the CRGR and the Staff hopes to send the proposed Rule to the ECO and to the Cornission in September.

The Rule should be issued with a three-month comment period with a final Rule available sometime in !!ovember 1987.

l W. Beckner mentioned the Staff's regulatory analysis of the effect of the rule change, notinp that there is the potential for large cost savings by the industry (see Appendix IX).

Westinghouse plants may be upgraded in power by about 5 percent as a result of this rule change.

There is a potential cost savings simply by increasing the flexibility of the fuel cycle and the way plants operate.

Because of the more realistic analy-sis, plants can go to higher peakinp factors, alleviate overly-tight diesel generator start times, which will increase diesel reliability, and 1

increase the overall safety of the plant.

The rule charge may also alleviate PTS concerns, i

i

~..

..--.-.-.._.-,.__m.,___,_.-,,m..__...-_-.-..

e 117TH ACRS PEETIflG 13 W. Beckner discussed some of the comments made at the ACRS subcommittee meeting in August 1986 (see Appendix IX).

He explained that the Staff finds it hard to justify on a safety basis phasing out Appendix K and forcing all licensees to develop a realistic calculation.

C. Michelson asked if a five-to-ten year phaseout would be feasible?

W. Beckner indicated that the Staff cannot justify that based on a safety benefit.

He noted the ACRS' concern regarding more guidance on uncertainty and he irdicated that three potential areas of guidance have been considered.

These involve the clarification of high probability, the providine of general principles to be used in realistic best-estimate models, and broad general principles on uncertainty methodology..He defined the objectives of the compendium of ECCS research as supporting the rulemaking and consistent with the Regulatory Guide.

He acknowledged that the compendium needs better organization to make it more readable.

Regarding the impact of the backfit rule on the NRR review of new evalua-tion models, he indicated that, according to the NRC Office of the General Counsel, as long as the issues are confined to the adenuacy of the evaluation model and whether it meets the criteria, there should be no backfit. The backfit rule could be invoked only if the Staff raises a different issue.

L.

Shotkin, NRC, discussed the llRC Research (RES) Staff's proposed methcdology for measuring thermal-hydraulic code uncertainty (see Appen-dix XI).

The RES Staff proposes to examine the uncertainty of the Staff's own best-estimate codes.

The study will not be incorporated directly in the proposed Rule, nor in the Regulatory Guide.

It will be addressed in the compendium of the ECCS research and has very little relationship to the Rule.

L. Shotkin explaineo that RES is proposing a comprehensive methodology for looking at the uncertainty in the calculat-ed peak clad temperature.

This methodology will consist of code versus data comparisons in a systematic examination of code models and correla-tions.

The following four factors will be addresssed by the NRC/RES uncertainty methodology, and taken together these four factors ccrstitute the code applicability to analyze a given scenario in a given full-scale vendor geometry:

1) code medeling capability, 2) quantitative measure of code uncertainty, 3) detection of compensating errors, and 4) scalability of the calculated peak cladding temperature. The uncertainty methodolcgy will first provide a quantitative estimate te go to full scale and make an estimate of the uncertainty.

At the same tire there will be a backup methodology that looks at all of the reasons why the results that the Staff is getting might be wrong.

He indicated that the nethodology will be reviewed in the beginning of October 1986.

C. Michelson indicated that the ACRS is cbliged to provide the Staff with an endorsement of the Rule as far as stipulating that it is ready to issue for public comment.

He indicated that the Subcommittee did ont have a problem with this endersement.

F. J. Remick applauded the rule

revision, indicating that it was long overdue.

D. Okrent asked C. Michelson how ACRS consultants received the Staff presentation at the Subcommittee meeting.

C. Michelson indicated that all had minor detailed

317TH ACr SEC

> COOLDOWN -->

FluIo

-> OUTSURGE -> STM. BussLE HEAT TRANSFER CONTRACTION EXPANSION LIouIo i

FLASHING AT0G STM. GEN.

RCS RCS RCS

! y STABILITY PRESS TEMP INVENTORY PRESS 3

];

P_ARAMETER g

3 g

INVENTORY l

A A

CONTRIBUTING h

h bSTEAM OR y

OD.H.

@U/Lo 9PORV/Sv MITIGATING MFEED ORCPS MHPI QSPRAY

_FACIO_RS bHPI HEATERS

O

SUMMARY

OF FINDINGS 0

CATEGORIZATION OF 1980 - 1985 EVENTS 0

TRANSIENT DISTRIBUTION 0

CAUSES & CONTRIBUTING FACTORS O

DATA

SUMMARY

i i

l I

O se i

3

't i

SUMMARY

OF FINDINGS CATEGORIZATION OF 1980 - 1985 EVENTS CATEGORY NUMBER PERCENTAGE i

A 78 35%

)

B 134 60%

l C

10 5%

1 O

A -33

O TRANSIENT DISTRIBUTION j

SIG PLANT CAT. C.

CAT. B TOTAL ANO-1 1

6 7

• TMI-1 SHUTDOWN CR-3 3

3 6

DURING THIS i

DB 2

8, 10 PERIOD i

OC-1 0

8 8

OC-2 0

3 3

OC-3 0

2 2

RS 4

5 9

10 35 45 l

0 1

i l

l I

l O

A -3+

~

0

i CATEGORY VENTS O

^T T"E BSwoo o 1"a u"1Ts

]

UNIT 118_0 1_981 19_82 1983 1984 1985 4

ANO-1 LOOP 4

CR-3 ICS/NNI PWR LOOP EFIC ACT POWER 1

D-B SFRCS ACT FWP TRIP

[MSSV]

SFRCS ACT

{

OC.1 OC.2 0C.3 i

RS TBV/ADV NNI POWER (1) FW HTR RV 4,)

(2) ICS/NNI PWR TMI-1 g

i i

TOTAL 2

2 0

0 2

4

#. TRIPS 35 47 42 44 20 34 l FREQUENCY %

6%

4%

0%

0%

10%

12%

\\

c

({])

SUMMARY

OF FINDINGS COMPLEX TRANSIENT DATA

SUMMARY

0 0

FREQUENCY OF OCCURRENCE IS 5%

0 TREND SHOWS 3 FOLD INCREASE FOR

'84 '85 OVER

'80 '83 EXPERIENCE O

CAUSES EXCESSIVE HEAT REMOVAL - OVERSTEAM & OVERFEED o

BY EFW FOLLOWING O

LOSS OF 0FFSITE POWER ICS/NNI POWER LOSS TBV/ADV/RV FAILED OPEN SPURIOUS EFIC/SFRCS ACTUATION INADEQUATE HEAT REMOVAL - LOSS OF STEAM o

GENERATOR (s) AS HEAT SINK SPURIOUS / IMPROPER SFRCS/EFIC ACTUATION l

0 SECONDARY PLANT RESPONSE IS REFLECTED INTO THE PRIMARY PLANT 1

_.,.i_____....

PRELIMINARY CATEGORY 'C' REVIEW CONCLUSIONS BASED ON A REVIEW OF CATEGORY 'C' TRANSIENTS 4

1.

EVENTS ARE GENERALLY THE RESULT OF EXCESSIVE PR TO SECONDARY HEAT TRANSFER 0

STEAM FLOW THROUGH FAILED OPEN VALVES 0

FEED FLOW TO MAINTAIN SG LEVEL 2.

CATEGORY 'C' TRANSIENTS REQUIRE ADDITIONAL OPERATOR ATTENTION TO ACHIEVE PLANT CONTROL i

O BALANCE ENERGY PRODUCTION AND REMOVAL O

CONTROL OF RCS INVENTORY & PRESSURE DUE TO FEEDBACK FROM THE SECONDARY PLANT 8

3.

KEY TO MITIGATION IS ENSURING CONTROL OF STEAM FLOW i

& FEED FLOW FROM THE CONTROL ROOM O'

BALANCING ENERGY TERMS REDUCES FEEDBACK INTO THE RCS 0

MAINTAIN PLANT CONTROL WHEN ICS/NNI POWER IS LOST l

4.

CORRECTIVE ACTIONS REQUIRED IN THE FOLLOWING AREAS O

PLANT DESIGN l

0 PLANT MAINTENANCE O

HUMAN INTERFACE i

0 LEARN FROM OPERATING EXPERIENCE l

+37 l

O PRELIMINARY CORRECTIVE ACTIONS AREAS BASED ON A REVIEW OF CATEGORY 'C' TRANSIENTS 1.

PLANT DESIGN 0

ENHANCE CONTROL OF STEAM FLOW & FEED FLOW FROM THE CONTROL ROOM UNDER POST-TRIP CONDITIONS 0

REDUCE SFRCS & EFIC ACTUATIONS AND IMPROVE RESPONSE ONCE ACTUATED 2.

PLANT MAINTENANCE O

IMPROVE MAINTENANCE PROGRAM WITH SPECIAL EMPHASIS ON STEAM FLOW & FEED FLOW COMPONENTS, CONTROLS AND THEIR MOTIVE POWER j

3.

HUMAN INTERFACE O

IDENTIFICATIONOFOPERABLECbHTROLS&

l INSTRUMENTATION ON LOSS OF ICS/NNI POWER 1

l 0

VERIFICATION OF AT0G STABILITY PARAMETERS 0

CONTROL OF PLANT PER AT0G GUIDANCE

4. OPERATING EXPERIENCE l

0 IMPROVE PROCESS TO IDENTIFY & ELIMINATE RECURRING PROBLEMS AT EACH PLANT i

0 IMPROVE PROCESS TO LEARN FROM B&WOG COLLECTIVE EXPERIENCE 9 _ggf' i

e=-----

--Tt---w--w--*

• • ~w-*-~~-*

--.--,-,e--

O CATEGORY 'C' EVENT REVIEW CATEGORY 'C' EVENT REVIEW IS ONE ELEMENT OF THE SPI PROGRAM DATA ENCOMPASSES 6 FULL YEARS OF B&WOG OPERATING EXPERIENCE REVIEW CONCLUDES 4 MAIN ITEMS:

O EVENTS ARE GENERALLY THE RESULT OF EXCESSIVE PRIMARY TO SECONDARY HEAT TRANSFER i

0 CATEGORY 'C' TRANSIENTS REQUIRE ADDITIONAL O

OPERATOR ATTENTION TO ACHIEVE PLANT CONTROL O

KEY TO MITIGATION IS ENSURING CONTROL OF FEED FLOW & STEAM FLOW FROM THE CONTROL ROOM I

0 CORRECTIVE ACTION IS REQUIRED IN THE FOLLOWING l

AREAS:

1 PLANT DESIGN PLANT MAINTENANCE HUMAN INTERFACE LEARN FROM OPERATING EXPERIENCE i

APPENDIX VI CATEGORY C TRANSIENT CONCLUSIONS

O CareGoRv C conc'USIoNS AND THEIR RELATIONSHIP TO THE SPI PROGRAM TASKS THE SPIP PROGRAM BUILDS ON KNOWN AND UNDERST00D' OPERATING EXPERIENCE.

FOCUS ON THE CATEGORY C CONCLUSION:

i 0

EVENTS ARE GENERALLY THE RESULT OF EXCESSIVE PRIMARY i

TO SECONDARY HEAT TRANSFER 1154 EFFORT - MAIN FEEDWATER SYSTEM REVIEW EFW/AFW REVIEW i

ICS/NNI SYSTEM REVIEW

)

INSTRUMENT AIR REVIEW SECONDARY PLANT RELAY SYSTEM REVIEW 0

CATEGORY C TRANSIENTS REQUIRE ADDITIONAL OPERATOR ATTENTION TO ACHIEVE PLANT CONTROL O

svsreas "Ev'sw <^Bove)

OPERATIONS / MAINTENANCE PERSONNEL INTERVIEWS RANCHO SECO - NUREG 1195 SYSTEMS 0

KEY TO MITIGATION IS ENSURING CONTROL OF FEED FLOW AND STEAM FLOW FROM THE CONTROL ROOM i

SAME AS AB0VE 0

CORRECTIVE ACTION IS REQUIRED IN PLANT DESIGN, PLANT MAINTENANCE, HUMAN INTERFACE, LESSONS LEARNED OPERATIONS / MAINTENANCE PERSONNEL INTERVIEWS OPERATING EXPERIENCE REVIEW PAST RECOMMENDATIONS / INDUSTRY REVIEW 1

L

B&W OWNERS GROUP

([)

SAFETY & PERFORMANCE IMPROVEMENT PROGRAM 1.

SENSITIVITY STUDY 2.

ICS/NNI EVALUATION i

3.

OPERATING EXPERIENCE REVIEW (PAST 6 YRS) 4.

EFW SYSTEM REVIEW 5.

MFW SYSTEM REVIEW 6.

SEC.

PLANT RELIEF SYSTEM REVIEW 7.

INSTRUMENT AIR SYSTEM REVIEW O

8.

RISK ASSESSMENT 9.

EMERGENCY OPERATING PROCEDURES REVIEW 10.

OPS /MAINT PERSONNEL INTERVIEWS 11.

REVIEW 0F PAST RECOMMENDATIONS 12.

DAVIS BESSE TASK FORCE (NUREG-1154)

E.G.

- NEW ROOT CAUSE PROCESS

- MOV WORKSHOPS

- MFW RELIABILITY

(])

- STEAM TRAP REVIEW Ab~Y/

13.

RANCHO SEC0 RRG ACTIONS (NUREG-1195)

O

O B&WOG/0TSG PLANT SENSITIVITY STUDY - DVERVIEW o

THE STUDY'S OBJECTIVE IS TO QUANTIFY THE RELATIVE SENSITIVITY OF B&W NSSS PLANTS WITH RESPECT TO OTHER PWR DESIGNS

- PROVIDES A LOGICAL, SYSTEMATIC BASIS FOR COMPARISON

- QUANTIFIES " SENSITIVITY" IN MEASURABLE TERMS

- CORRELATES WITH ESTABLISHED SAFETY LIMITS AND PERFORMANCE REQUIREMENTS

- IS BROAD IN SCOPE TO FULLY ASSESS DIFFER-ENCES o

MPR IS PERFORMING CALCULATIONS AND ANALYSES OF THE O

THERMODYNAMIC RESPONSE OF THE THREE NSSS PWR DESIGNS TO UPSETS IN FW FLOW, STEAM FLOW, AND 1

REACTIVITY

- USES SPECIFIC PLANTS, CHOSEN TO REFLECT DIFFERENCES WITHIN VENDOR LINES i

- IMPORTANT B&W PLANT DIFFERENCES ARE BOUNDED

- ASSESS THE RESPONSE OF CONTROL AND PROTEC-l TIVE SYSTEMS TO UPSETS, FAILURES

- ADDRESSES FAILURES REFLECTED IN OPERATING EXPERIENCE

- LOOKS AT THE NEED FOR OPERATOR ACTION AND CONSEQUENCES OF FAILURE'TO PERFORM

- LOOKS AT PLANTS AT POWER O

^~**

O O

O

.i QUANTIFICATION OF SENSITIVITY l

OPERATIONAL j

INDICES SAFETY PARAMETERS LIMITS / CRITERIA l

o SECONDARY DESIGN o STEAM LINE FLOODING

{

PRESSURE o SG OVERFILL 1

o MARGIN o PRIMARY DESIGN o SG DRYOUT PRESSURE, TEMPERATURE o SAFETY VALVE CHALLENGE o SATURATION MARGIN o PORY CHALLENGES o TIME o Kw/FT LIMIT o STEAM / FEED ISOLATION o LOSS OF PRESSURIZER i

LEVEL ll 0 FREQUENCY o MINIMUM DNBR o RPS TRIP LIMITS 0 PTS LIMITS o SAFETY INJECTION LIMITS j

o HEATUP/COOLDOWN RATE LIMITS

• Q

~

I

)

i j

O ANALvSIS NATRIx ceARTIAL LIST 1NG)

(SPECIFIC ANALYTICAL ACTIVITIES)

SIGNIFICANT DISTURBANCES, WITH CORRECTIVE ACTION A.

TURBINE TRIP

- WITH REACTOR RUNBACK

- WITH REACTOR TRIP i

- WITH STEAM FLOW UPSET B.

LOSS OF ONE FEEDWATER PUMP

- WITH SLOW RUNBACK

- WITH FAST RUNBACK C.

LOSS OF ALL FEEDWATER PUMPS i

- NORMAL EFW

- DELAY EFW

- NO EFW O

'- excessive erw D.

CONTROL SYSTEM UPSETS

!l

- VARYING FEED SYSTEM CONFIGURATIONS AND INITIAL CONDITIONS E.

LOSS OF COOLANT FLOW I

- LOSS OF ONE PUMP

- LOSS OF ALL PUMPS e

THESE CALCULATIONS WILL BE REVIEWED TO DETERMINE IF, WHEN, AND HOW OFTEN THESE OPERATIONAL LIMITS l

ARE REACHED.

l e

AS A RESULT, THE RELATIVE SENSITIVITY OF THE THREE DESIGNS WILL BE QUANTIFIED.

l k ~~h f i

--,_6

o i

B&WOG SENSITIVITY STUDY -

SUMMARY

o THE STUDY WILL CLEARLY IDENTIFY WHERE THE B&W PLANT RESPONSE IS DIFFERENT, BOTH POSITIVELY AND NEGA-TIVELY, FROM THE OTHER DESIGN.

I

- WHERE THERE IS DIFFERENCE IN MARGIN OR DIFFERENCE.

IN TIME TO RESPOND.

Oo THE STUDY WILL PROVIDE AN ENGINEERING BASIS FROM WHICH TO ASSESS THE SAFETY SIGNIFICANCE OF THESE DIFFERENCES.

- SAFETY SIGNIFICANCE IN TERMS OF ESTABLISHED i

LIMITS

~

o THE STUDY WILL PRODUCE RECOMMENDATIONS FOR CHANGES TO CORRECT AREAS WITH ADVERSE DIFFERENCES, TO BE EVALUATED IN LIGHT OF THEIR SIGNIFICANCE.

a>-s g

n a

a e-.-

l I

O l

l l

i l

i I

l DECAY HEAT REMOVAL B&W PLANTS O

AN OVERVIEW s

5

a_.

2.-

+

u e

2

.a O

PURPOSE OF PRESENTATION PROVIDE A GENERAL OVERVIEW 0F DHR ON B&W PLANTS O

SHOW THE SPIP RELATIONSHIP TO DHR O

I t

9 A-+ 7 4

--~---,-,-,r------

__,.,__,,_,w____n,,,,,na,,,,,,,

(

O l

l l

DECAY HEAT REMOVAL REQUIRED o

AFTER SHUTDOWN FROM POWER o

NOR' MAL O

o ABNORMAL o

DURING REFUELING 4

,,...mm------,,e

_w,,,_,-,

O j

IN A BROAD SENSE:

o DECAY HEAT REMOVAL IS ACCOMPLISHED IN B&W PLANTS AS IN ANY PWR I

o DEFENSE IN DEPTH IS PROVIDED IN B&W PLANTS TO THE SAME DEGREE OR MORE THAN IN OTHER PWRs o

PRA RISK EVALUATIONS INDICATE THAT THE CORE DAMAGE i

FREQUENCY FOR INTERNAL EVENTS IS THE SAME FOR B&W AS FOR OTHER PWRs f

l 1

l l

l

DECAY HEAT REMOVAL USING STEAM GENERATOR WITHOUT STEAM GENERATOR FIRST LINE OF SECOND LINE OF THIRD LINE OF DEFENSE DEFENSE DEFENSE MAIN FEEDWATER EMERGENCY FEEDWATER FEED AND BLEED B&W PLANTS - GENERAL:

CONDENSATE HPI AND PORV OR q

PRESSURIZER SAFETY lq I

STEAM RELIEF STEAM RELIEF DAVIS-BESSE:

T e MSSV e MSSV MAKEUP AND PORV e TBV/ADV ICS/NNI EFIC, SFRCS CONTROLS OR EQUIVALENT CONTROLS RC PUMPS (FORCED FLOW) e RC PUMPS OR' s NATURAL CIRCULATION B0XES INDICATE ITEMS INCLUDED IN SPIP e

9 8

O SPIP REVIEW OF-SYSTEMS IMPORTANT TO DECAY HEAT REMOVAL COVERS ALL ASPECTS o

DESIGN o

OPERATIONS o

TESTING / INSPECTIONS O

o aaratsa^ ace

____________________________o__________________________

OBJECTIVE - INCREASE LEVEL OF OVERALL PLANT SAFETY BY:

o BETTER SYSTEM AND COMPONENT RELIABILITY o

BETTER OPERATION FOR NORMAL AND ABNORMAL CIRCUMSTANCES o

REDUCED PLANT / SYSTEM TRIP FREQUENCY l

O A-r/

I

()

DECAY HEAT REMOVAL - FIRST LINE OF DEFENSE MAIN FEEDWATER/ICS CONTROLS 4

BACKGROUND INFO OF INTEREST TRANSIENT (1983 & PRIOR DATA) i o

LOFW 0.9/YR o

EXCESS F/W 0.2/YR o

ICS POWER 5x10-2/YR EQUIPMENT' PERFORMANCE (APPROXIMATE) o OTHER PWR MFP TURBINE 3.3 PUMP TRIPS /YR o

B&W MFP TURBINE 2.3 PUMP TRIPS /YR o

NNI SENSOR 0.6 Rx TRIPS /YR SPIP REVIEW OBJECTIVES IMPROVE SYSTEM AND MFW PUMP RELIABILITY O

KEEP'MFW ON LINE REDUCE EFW CHALLENGES SM0OTH POST TRIP RESPONSE OBSERVATIONS TO DATE MFW/ICS REQUIRE DILIGENCE TO MAINTAIN TUNING AND CONTROL o

SPEED CONTROL AND RESPONSE UNNEEDED MFW PUMP PROTECTION TRIPS o

LOW SUCTION PRESSURE ICS SENSOR INPUT FAILURES NNI POWER SUPPLY MIDSCALE FAILURES

({])

(AFFECTS CONTROL COMPONENT POSITIONS)

$.Eb t

---,-,-nnn-m,

O DECAY HEAT REMOVAL - SECOND LINE OF DEFENSE EMERGENCY FEEDWATER/ CONTROLS BACKGROUND INFO OF INTEREST ALL B&W PLANTS - INITIATION AND CONTROL TO BE o

INDEPENDENT OF ICS AND ICS POWER SUPPLIES i

SPECIFIC PLANT ACTIONS UNDERWAY, E.G.

o RANCHO SECO - EFIC INSTALLATION o

o DAVIS-BESSE - SFRCS MODS i

- MOTOR DRIVEN EFW PUMP

- TURBINE DRIVEN PUMP STEAM 1

o RESPONSE TO SSFI REVIEWS OF EFW o

MOV PROGRAMS j

SPIP - REVIEW GOALS i

o REDUCE SFRCS AND EFIC INITIATIONS AND IMPROVE i

PERFORMANCE ONCE INITIATED.

IMPROVE RELIABILITY FOR DECAY HEAT REMOVAL o

OBSERVATIONS TO DATE TURBINE DRIVEN EFW PUMP RELIABILITY o

INCREASE START TIME MORE CONSISTENT TESTING MAY BE DESIRABLE o

MAINTENANCE PRACTICES ARE IMPORTANT o

o UNNECESSARY INITIATIONS O

A -ss 9

A

DECAY HEAT REMOVAL - FIRST/SECOND LINE OF DEFENSE STEAM PRESSURE AND RELEASE CONTROL l

BACKGROUND INFORMATION OF INTEREST o

SAFETY VALVE FAILURES - STUCK OPEN

~2 x 10-3 (95% CONFIDENCE) o OPERATOR ACTIONS TO SOLIDLY RESEAT MSSVs ARE NEEDED o

FAILURES OF ICS/NNI POWER CAUSING OPEN BYPASS CONTROL VALVE FAILURES o

ONE PLANT - BEING CORRECTED l

SPIP PROGRAM REVIEW GOAL $

O_

o aeouce cua'teNGeS To SarsrieS o

SM0OTHER PRESSURE CONTROL PRESSURIZER LEVEL i

MFP/EFWP RESPONSE l

o REDUCE OPERATOR ACTIONS l

OBSERVATIONS TO DATE o

SAFETY VALVE PERFORMANCE IS GOOD BUT NOT OPTIMAL i

EARLY LIFTS BLOWDOWN GREATER THAN ORIGINAL DESIGN l

o STEAM PRESSURE CONTROL UNEVENESS INTERACTION BETWEEN TBVs AND MSSVs O

erw A -54

-...,w m

3--

wy-

l

SUMMARY

o THREE LEVELS OF DEFENSE FOR DECAY HEAT REMOVAL ARE i

PROVIDED FOR THE SPECTRUM OF NORMAL AND ABNORMAL CONDITIONS i

o THE OTSG CHARACTERISTICS DEMAND CAREFUL CONTROL OF i

NORMAL SYSTEMS o

NORMAL SYSTEMS PROVIDE THE PREFERRED METHOD FOR HEAT REMOVAL (FIRST LINE OF DEFENSE) AND ARE BEING ADDRESSED MAIN FEEDWATER STEAM i

CONTROLS o

IMPROVEMENTS IN EMERGENCY FEEDWATER AND CONTROLS ARE BEING ADDRESSED (THE SECOND LINE OF DEFENSE) o FEED AND BLEED COOLING (THIRD LINE OF DEFENSE) IS NOW MORE THAN ADEQUATE AND REQUIRES NO FURTHER ACTION *

• DAVIS-BESSE MAY ELECT ADDITIONAL CHANGES 0

/-6

--,,,_,_-_,----,m,.

-.m.,_

O CLOSING COMMENTS EXPECT TO DEVELOP APPROXIMATELY 500 SAFETY AND o

PERFORMANCE IMPROVEMENT REC 0144ENDATIONS FROM 13 MAIN TASK EFFORTS BY 2Q 87.

i o

INTEND TO DISPOSITION ALL. RECOMMENDATIONS FORMAL i

IMPLEMENTATION OF SEVERAL RECOMMENDATIONS HAS o

ALREADY RESULTED IN BENEFIT (TRIP AVOIDANCE).

EXPECT THAT ISSUE / CONCERN DISTILLATION AND o

PRIORITIZATION ACTIVITIES WILL IDENTIFY AREAS MOST SAFETY-SIGNIFICANT-APPROPRIATE FOR CHANGE, AND FOR RECOMMENDATIONS AFFECTING THOSE AREAS TO BE ACTED UPON FIRST.

EXPECT THE IMPLEMENTATION PHASE TO BE CONTINUING o

FOR SEVERAL YEARS.

/ -SA

..x -

()

ACRS OBSERVATIONS AND RECOMMENDATIONS 1.

THE PROGRAM APPEARS TO CONCENTRATE ENTIRELY ON DESIGN; IT SHOULD INCLUDE THE EFFECT OF THE OPERATING ORGANIZATION ON SYSTEM PERFORMANCE.

THE B&W SYSTEMS WERE DESIGNED TO~ RESPOND DIFFERENTLY.

THE PROGRAM SHOULD DETERMINE WHETHER THESE CHARACTERISTICS ARE GOOD,

BAD, OR INDIFFERENT FROM A SAFETY

~

PERSPECTIVE.

3.

T00 LITTLE ATTENTION APPEARS TO BE GIVEN TO DHR.

IF DHR IS ADEQUATE PUBLIC SAFETY IS ENSURED.

O A-57

f APPENDIX VII NRC STAFF PRESENTATION ON

' GENERIC ISSUE N0.124 NRR STAFF..PRESbNTATION TO THE ACRS f

SUBJECT:

.AUXILIkRYFEEDWATERSYSTEMRELIABILITY, GENERIC I,SSUE NO. 124 DATE:

SEPTEMBER 12, 1986 j

PRESENTER: SAMMY S. DIAB c

PRESENTER'S TITLE / BRANCH /DIV: TASK MANAGER, REACTOR SAFETY ISSUES BRANCH / DIVISION OF SAFETY REVIEW & OVERSIGHT PRESENTER'S NRC TEL. NO.:

492-4083 i

FULL COMMITTEE:

A-M O

4 SEPTEMBER 12, 1986 PRESENTATION TO THE ACRS AUXILIARY FEEDWATER SYSTEM (AFWS) RELIABILITY GENERIC ISSUE NO.

124

~

THIS IS A STATUS PROGRESS REPORT MODIFIED RESOLUTION APPROACH MODIF'IED RESOLUTION PLAN PROGRAM l

SCHEDULE FOR RESOLUTION I

i

(~~

THE MODIFIED RESOLUTION APPROACH SHORT TERM CONCENTRATED RELIABILITY REVIEW FOLLOWED BY AN AFWS REVIEW FINDINGS REPORT FOR EACH OF THE SEVEN PLANTS REVIEW CARRIED OUT BY ONRR MULTIDESCIPLINE AFWS REVIEW TEAM REVIEW EFFORT WILL BENEFIT FROM RANCHO SECO RESTART EFFORT B&W OG DESIGN REASSESSMENT EFFORT THE DIE SSFI PROGRAM THE LICENSEE'S AFWS RELIABILITY ANALYSES RELEVANT INDUSTRY EFFORTS v

0 A-i o i

O WHY NOT STICK WITH "0LD" PROPOSED RESOLUTION?

NEGATIVE COMMENTS FROM NRR REVIEW 0F BACKFIT ANALYSIS i

1 e

LENGTHY PROCESS (ISSUANCE OF GL, LICENSEES TO CONDUCT AFWS ras, STAFF REVIEW 0F ras FOLLOWED BY ISSUANCE OF i

SERs) i 4

!lO 1

l i

I

/t -d /

O l

MODIFIED RLSOLUTION PLAN TWO TIER APPROACH THE SEVEN PLANTS:

LOW ESTIMATED PRE-TMI AFWS RELIABILITIES.

TWO PUMP AFWSs.

THE REST OF PWRs:

ACCEPTABLE AFWS RELIABILITIES.

MOSTLY THREE PUMP AFWSs.

OUTCOME OF THE SEVEN PLANT REVIEW WILL DICTATE ACTION REGARDING REST OF PWRs.

O p.g n l

O THE SEVEN PLANTS ANO-1

• SSFI CONDUCTED BY OIE

• B&W OG FOR DESIGN REASSESSEMENT

• AFWS REVIEW TEAM

' RANCHO SECO

• RESTART EFFORT:

EXTENSIVE MULTIDECIPLINE STAFF REVIEW 0F AFWS

- DESIGN MODS

- PROCEDURES a TRAINING

- SUPPORT SYSTEMS

- INDICATION 8 CONTROL

• B&W OG FOR DESIGN REASSESSEMENT

• AWFS REVIEW TEAM O

g g3 e

CRYSTAL RIVER B&W OG FOR DESIGN REASSESSMENT LICENSEE'S AFWS RELIABILITY ANALYSIS AFWS REVIEW TEAM PRAIRIE ISLAND 182 LICENSEE'S AFWS RELIABILITY ANALYSIS AFWS REVIEW TEAM ANO-2 LICENSEE'S AFWS RELIABILITY ANALYSIS O-DEDICATED BLEED a FEED ARRANGEMENT AFWS REVIEW TEAM FT.CAHOUN AFWS REVIEW TEAM i

i l

) ff G

l l

BaW OG (INCLUDES R. SECO, C. RIVER a ANO-1)

• B&W DESIGN REASSESSMENT

- IMPROVE RELIABILITY OF MFhS j

- IMPROVE RELIABILITY OF AFWS

- LIMIT CHALLENGES TO AFWS j:

• CURRENTLY UNDERWAY i

• WILL BE PRESENTED TO THE FULL COMMITTEE TODAY l

l l

REVIEW PROGRAM ApyS REVIEW TEAM A FIVE PERSON TEAM PLUS A TEAM LEADER REVIEW MATERIAL REQUESTED AND ENROUTE TEAM STARTS OPERATION MONDAY, SEPTEMBER 15 REVIEW SCOPE AREAS TO BE C0VERED ARE SHOWN ON TABLE 1 AFW AND SUPPORT SYSTEMS AUDIT SELECTED PREVENTIVE AND CORRECTIVE MAINTENANCE ACTIVITIES DURING LAST 12 M0S.

AUDIT SELECTED SURVEILLANCE TESTING PROCEDURES AND POST MAINTENANCE TESTING POST TMI MODS COMMON MODE VULNERABILITIES OPERATOR REC 0VERY a WALK-THROUGHS CONTROL ROOM ADEQUACY FOR INDICATION, CONTROL & REC 0VERY EASE OF LOCAL RECOVERY, INDICATION & CONTROL ALTERNATE DECAY HEAT REMOVAL MEANS k ~bb

O' TABLE 1.

Ar,EAS COVERED BY STAFF REVIEW 1.

PalDs A g fDg g RVs (FEED EED)

>huPPLIES, LUBRICATION,C00LIN 2.

FSAR

SYSTEMS, SYSTEM DESCRIPTION 4.

TECHNICAL SPECIFICATION 5.

IaC LOGIC DIAGRAMS 6.

SERS SINCE 1980 s

7.

MAINTE$ANCE(PREVENTIVEPROGRAMS,CORRECTIVEACTIVITIES) p LAST 12 MOS v

8.

SURVEILLANCE TESTING (FEW PROCEDURES AND RESULTS), POST MAINTENANCE TESTING, LERs, NPRDS, SOEs, R0s, A0s DURING THE LAST 12 MOS 9.

EMERGENCY OPERATING PROCEDURES (LOSS OF HEAT SINK) 10.

FaBLEED PROCEDURES (SEE 9 AB0VE), TRAL'ING (YES/NO)

N O

A-47 e

O

)

REVIEW PROGRAM (CONT.)

POSSIBLE REVIEW OUTCOMES

• THE REVIEW TEAM IS REASONABLY ASSURED THAT THE AFWS IS ADEQUATE AND SUFFICIENTLY RELIABLE

• PROVIDED CERTAIN PLANT-SPECIFIC MEASURES WERE TAKEN THE AFWS IS ADEQUATE AND SUFFICIENTLY RELIABLE,

• WITHOUT SUBSTANTIAL CHANGES TO THE AFWS' DESIGN MAINTENANCE, TESTING AND/0R OPERATION, THE REVIEW TEAM CAN NOT BE REASONABLY ASSURED THAT THIS SAFETY SYSTEM IS SUFFICIENTLY RELIABLE IMPLEMENTATION OF FINDINGS i

l

• ANY REVIEW TEAM FINDINGS WILL BE DISCUSSED WITH LICENSEE

• ANY PLANT BACKFITS WILL BE HANDLED BY NORMAL PROCEDURES.

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SCHEDULE REVIEW TEAM VISIT REPS.BI P. ISLAND 182 SEP. 29, 1986 Nov. 3, 1986 ANO-2 NOV. 17, 1986 DEC. 22, 1986 FT. CALHOUN JAN. 5, 1987 F,EB. 9, 1987 AN0-1 NOTE (1)

NOTE (2)

R. SEC0 NOTE (1)

NOTE (2)

C. RIVER NOTE (1)

NOTE (2)

NOTE (1) THE NEED FOR AND SCHEDULE OF VISITS TO THESE PLANTS WILL BE DETERMINED IN LIGHT OF THE ONG0ING STAFF WORK ON THESE PLANTS.

NOTE (2) REVIEW TEAM REPORTS FOR THESE PLANTS WILL BE COORDINATED WITH ONG0ING STAFF EFFORTS.

9 4

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APPENDIX VIII

~

REVISION OF THE ECCS RUI.E 10 CFR 50.46 AND APPENDIX K wSx "x

0o 2S5 EM$

Po ACRS FEVIEW & Ff0 FUSED ouz Z O a:t "i=-

% S *-

FEVISICfi & TE ECCS FRE

S bx "ti 10 CFR 50.f6 #0 APF90lX K j

2 317* ACRS FETING SEFTEE ER 11, 1986 O

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H l-WILLI #1 EE02ER i

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PACKGPOLNMIFf1NT RLE 50.r,6(A) EQJIFES TFAT REPLENCY CORE COOLING SYSTEMS "...E DESIGED SlDi TliAT ITS CALCILATED C00LI!E EFF0FFANCE C0EOFPS TO TE CRITERIA SET FORTH IN PARAGRAFH (B)

...AffB0lX K, ECCS EVALUATIQi PGELS, SETS FORTH ERTAIN EQJIPED MO ACEPTABLE FEATLFES T EVAtt!ATION PEELS."

50.t6(B) CRITERIA IfEtt0E: (1) CALQLATED PEAK CLADDING TEPFERATLEE LIMIT T 2200*F (2) FAXUiM CLADDING OXIDATION OF 17% OF THIOKSS, (3) MAXIf1M HYDROEN GEERATIQi LESS TFAN 1% T HYPOTETICAL #0lNT FPm TOTAL ETAL EACTION, (4) C00UBLE GEDETRY FAINTAIED, #0 (5) LCNG TEFM COOLING PFD/IDED.

APE 10lX K CONTAltG SECIFIC FEAllFES TilAT EVALUATION P00ELS (EPS) MJST CONTAIN (E.G.,

SOURCES OF EAT TO EE INQifR IN TE CALQLATIONS), SPECIFIC FEATURES EMS MJST EXCLl0E k

(E.G., STEAM COOLING OtLY FOR LOW PEFLOOD PATES), SPECIFICALLY EQJIRED P00ELS (E.G.,

F000Y EEAK FLOW), M0 P00ELS WilOf AE ACCEPTABLE, BUT NOT EQJIRED (E.G., SPECIFIC EAT TRANSFER CCFSELATIONS).

EXISTING PJ.E HAS PPD /IDED SUCESSRL EGLATION #0 WILL BE "GPJM)FAllEED" lNTIL If00STRY FEASES IN EW FILE: IGEWR, IT IS LNIGE IN TM) ASRCIS:

1.

PESCRIPTI\\E t00ELS 2.

PESTRICTI\\E PER)RTING OF ERRORS, EVEN WEN NO EGLATORY OR SAFETY TliES110LD IS SURPASSED IT IS APPF0PRIATE, AFTER A ECADE OF ECCS RESEARQi, TO REVISE TE RAE TO pal (E IT PDE GESISTENT WITH OTIER NRC EGULATIDS.

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a tPf, DE, liOUSTRY AfD F0EIGI ESEARCii PERFORWD SINE 1975 PfWIDES A TEGNICAL BASIS milch PEff!ITS P0RE EALISTIC TEATENT OF ECCS ANALYSES.

~

1973CatilSSIONPOSITIONSTATErfNI:

"TE rDtilSSION EXPECTS THAT BOTH C0ftiMMffAL A?O PRIVATE PiGTWiS WILL b DILIGNil.Y, A?O EXPECTS TO CONSIDER PfDFILY TlE EW KNOWLEDGE AS IT BECOMES AVAILABLE, AM) TO CONSIDER SUCH CHANGES IN TESE EedLATIONS AS TEY APPEAR k

APPROPRIATE IN TE LIGHT OF ALL INFORMATION THEN AVAILABLE".

BILAD SUPPORT FOR REVISING TlE RULE BY NRC STAFF, ACRS, AND INDUSTRY; NO Q

It01 CATION OF OPPOSITION FF@l PUBLIC OR INTERVENORS AT THIS TIE.

BROAD SUPPORT FOR CURRENT RULE REVISION APPROACH BASED ON SECY-83-472-(I. E., EALISTIC EM WITH LNCERTAlt(IY EVALUATION).

PESENT APPENDIX K EQUIREENTS RESULT IN CONSIDERABLE DIVERSION OF INDUSTRY JE) AND EGULATORY ESOURCES TO SAFETY ANALYSIS TilAT AE IN FACT IRRELEVANT, TAKING ATTENTION AWAY FROM FDE IFFORTANT ISSUES 9

e

PROPOSED RULE EVISION 50.46(A)(1)

(1)

... ANALYTICAL TEGINIWE REALISTICALLY DESCRIBES TE BEHAVIOR OF TE EACTOR SYSTEM..."

"C0WARIS0NS TO APPLICABLE EXPERIENTAL DATA..."

"lECERTAINTY PUST BE ACCOUNTED FOR S0 THAT...lllERE IS A HIGH LEVEL OF PROBABILITY TilAT TiiE CRITERIA [ PARAGRAPH B1 WOULD NOT BE EXCEEDED."

(II) " ALTERNATIVELY, AN ECCS EVALUATION MODEL MAY BE DEVELOPED IN COEORMANCE WITli... APPENDIX g

K..."

50.46(A)(2) q "ESTRICTIONS ON REACTOR ORRATION WILL BE I&OSED...IF TE EVALUATIONS OF ECCS COOLING PERFORMANCE SUBMITTED AE NOT CONSISTENT WIT'i PARAGRAPHS (A)(1)(1) AND (II)...AND EQUIRED TO PROTECT PUBLIC HEALTH AND SAFETY."

50.46(A)(3)

... ESTIMATE TlE EFRCT OF ANY 01ANGE TO OR ERROR IN AN ACCEPTABLE EVALUATION MODEL..."

...SIGNIFICANT OlANGE OR ERROR IS DE milch RESULTS IN...TEFFERATUE DIFFERENT BY FDE THAN 50*F..."

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"FOR EAOl GiANGE T0, OR ANY ERROR..cREPORT lliE NARIRE OF TIE 01ANGE OR ERROR AND ITS ESTIMATED EFECT ON TIE LIMITING ECCS ANALYSIS...AT LEAST A?EUALLY..."

"IF THE GIANGE OR ERROR IS SIGNIFICANT...REN RT WITHIN 30 DAYS..."

... REPORT A PROPOSED SOEDULE...T0 SHOW COPFLIANCE..."

... FACILITIES NOT HAVING NRC APPROVED INTEGRATED SCHEDULING SYSTEM, A SOEDULE FOR AOilEVING C0FPLIANCE WILL BE ESTABLISED BY TIE NRC STAFF WITHIN 60 DAYS..."

"ANY OlANGE OR ERROR...D0ES NOT C0 WORM TO CRITERIA SET FORTH IN PARAGRAPH (B)...IS A REPORTABLE EVENT AS DESCRIBED IN 50.55E, 50.72 AND 50.73."

5 APPENDIX K N

I.C.S.B - DOUGALL-R0llSENOW CORRELATION RENVED AND ANOTHER' HEAT TRANSFER CORR UPDATED.

• I.C.5.C

" EVALUATION MDDELS APPIDED AFTER IEFFECTIVE DATE OF RULEl...SHALL NOT USE...[DO

" EVALUATION MODELS... APPROVED PRIOR TO [ EFFECTIVE DATE OF RULE] CONTINUE TO BE A UNTIL...GiANGE... ESULTS IN A SIGNIFICANT REDUCTION IN OVERALL CONSERVATISM IN TlE EVALUATIONPDDEL."

e

...SIGINIFICANT REDUCTION IN THE OVERALL CONSERVATISM...WOULD BE A EDUCTION EAK FUEL CLADDING T&PERATURE OF AT LEAST 50*F FROM THAT WilCH WOULD HAVE BEE ON [ EFFECTIVE DATE OF RULEl..."

II. - 01ANGES IN DOCTEKfATION REQUIREENTS T0:

1.

RBTNE 20*F DEFINITION OF SIGNIFIC.'ST 01ANGE 2.

PROVIDE A COPPLETE LISTING OF COPPUlER PROGRAM ONLY IF REQUESTED BY STAFF.

k i

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mal (E CONSISTENT WITli GiANGES IN 50.46(A).

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ECCS RULE REVISION PACKAGE HAS 3 PARTS:

1.

REVISED RUE 2.

EGULATORY GUIDE 3.

CO WENDIlN OF ECCS RESEARGI 1.

RUE DOES NOT ENTION EITTER TE EG GUIDE OR TIE COWENDILN; TESE AE ONLY REFERENCED IN THE SUPPLEENTARY INFORMATION AND THUS HAVE LITTLE EGAL STATUS.

S 2.

REG GUIDE PROVIDES " GUIDANCE" ON "ACCEPTABE" MODELS/ DATA AND UNCERTAINIY PROCEDURES WHICH, IN PRACTICE, NRC WILL NOT OiALLENGE IF USED.

g 3.

COWENDIlN IS REFEENCED IN REG GUIDE AS REPESENTATIVE OF "A LARGE BODY OF DATA GENERALLY APPLICABE TO B.E. MDDELS"; Blff TEE IS N0 GUARANTEE TllAT ANYTHING ENTIONED IN TIE COWENDIlN IS "ACCEPTABE" IN TIE EG GUIDE SENSE OF TIE TERM "ACCEPTABE".

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1.

REVISED RULE NO SKCIFICITY ON 10DELING EQUIREENTS (0TER lliAN DOUGALL-ROHSENOW)

MORE FLEXIBILITY ON REPORTING EQUIREENTS PERMITS USE OF LATEST TE0f0 LOGY BEST-ESTIIMTE CODES I

CODE UNCERTAINTY BASED ON DATA COPPARISON T

USE OF B. E. CODES SHOULD LEAD TO MDE UNDERSTANDABLE EGULATION CONSISTENT WITli INDUSTRY C0FMITENT TO USE OF SECY-83-472 (B.E. CODE + UNCERTAINTY +

APPENDIX K CORECTIONS)

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2.

EGULATORY GUIDE PROPOSES, FOR PUBLIC COMENTS, ACCEPTABLE MODELS/ DATA RELATED TO APPENDIX K.

DEFINES REQUIREENTS FOR ESTIMATING OVERALL CODE UNCERTAINTY AT TijE 95%

PROBABILITY LIMIT.

FLEXIBILITY TO ACCEPT INDUSTRY INITIATIVES.

h 3.

COWBDIlN OF ECCS RESEARCH w

\\

ROAD-MAP TO DECADE OF RESEARCH ON ECCS SINCE AKRICAN RlYSICAL SOCIETY REf0RT.

SHOWS NRC 11AS PERFORWD SUFFICIENT RESEARCH TO BETTER UNDERSTAND MARGIN OF SAFEIY IN CALCULATED ECCS OPERATION AND TO JUSTIFY THE RULE REVISION.

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SCHEDULE FOR REVISION OF ECCS RULE ACRS EETING TO UPDATE ECCS RULE ACTIVITIES APRIL, AUGUST, SEPTEFEER 1986 NRR, DRR, ELD, RES CONCUR WITH PROPOSED RULE JUE 1986 CRGR E ETING JULY, AUGUST 1986 SEPTEMBER 1986 EDO T

C0FMISSION SEPTEMBER 1986 C0FPENDIlti0F ECCS RESEARCil ISSUED FOR PUBLIC COPE NT OLT0BER 1986 3

NOTICE OF PROPOSED RULEMAKING ISSIED FOR PUBLIC C0FIENT NOVEMBER 1986 6

REGULATORY GUIDE ISSUED FOR PUBLIC COM U TS NOVEFEER 198%

C0ftENTS PERIOD ENDS FEBRUARY 1987 FINAL RULE PUBLISilED NOVEMBER 1987 9

APPENDIX IX REGULATORY ANALYSIS REVISION OF ECCS RULE CE x

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EGULATORY ANALYSIS EVISION OF ECCS RULE 317* ACRS EEilNG sL%

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SUPPLEENTARY INFORMATION N

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REGULATORY ANALYSIS-EF RCT OF RULE 01ANGE CALCULATED EAK CLADDING TDPERATUES (PCT) DURING LARGE BEAK LOCA (INCLUDING WOULD BE REDUCED, At0UNT OF REDUCTION v0JLD BE PLANT SECIFIC AND ALSO DERND ON TlE ACCURACY OF THE CALCULATION. HOWEVER, THE REDUCTION IN CALCULATED PCT VQJLD LIKELY BE LARGE EN00Gi SO Ti!AT LARGE LOCA CONSIDERATIONS VajLD NO LONGER BE LIMITING. OTTER CONSIDERATIONS (E.G., DNB, SBLOCA WOU W LIMIT PLANT OPERATION.

SMALL BREAK LOCA (SBLOCA) MODELS AE GENERALLY MBE REALISTIC TllAN LARGE BRFA E LESS AFFECTED BY THE PROPOSED RULE OiANGE, SBLOCA MAY BECOE LIMITING.

REDUCED CALCULATED LARGE BREAK LOCA PCT COULD RESULT IN:

INCEASED ALLOWED PEAK LOCAL POWER INCREASED TOTAL POWER Q

01ANGES IN EQUIPENT, SURVEILLANCE OR LCO i

T ALL TIE ABOVE OlANGES WOULD LIKELY NOT BE POSSIBLE AT 111E SAE TIE AND MANY OllE h0VLD llAVE TO BE CONSIDERED SU0i AS:

DtB LIMITS PLANT HARDWARE LIMITS OTTER CHAPTER 15 EVENTS l

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EGULATORY ANALYSIS-PLANTS EXKuul TO BEEFIT GE AND WESTItG100SE PLANTS AE COERALLY LIMITED IN OKRATING FLEXIBILITY BY TE ECCS RULE AS EVIDENCED BY:

WRITTEN RESPONSFS FROM GE AND WESTINGHOUSE SIGNIFICANT RESLES INVESTED BY GE AND H IN IlPROVED ECCS CALCULATIONS RECENT BWR APPLICATIONS OF NEW GE SAFER FDDEL TlE GE SAFER MODEL APPROVED UNDER SECY-83-472 PROVIDES BWRS WIT Q

REDUCTION IN O KRATING LIMITATIONS.

IT MAY E MBE DIFFICULT TO REDUCE OKRATING LIMITATIONS ON i

N H PLANTS T1100GH TllE USE OF SECY-83-472 W/0 A RULE CHANGE.

B8W AND CE CLAIM NO BENEFIT FOR TEIR PLANTS.

9

KGl1LATORY ANALYSIS-POTENTIAL COST SAVINGS WESTINGHOUSE CLAIMS MDST PLANTS AE LOCA LIMITED AND COULD BE UP PRESENT VAllE (10% DISCOUNT RAID OF SUOf A POER INCREASE FOR 5% IF LOCA LIMITS REMOVED.

H PLANT ASSlflING A 30 YEAR LIR RANGES FROM $13M TO $1ti7M, WITil AN AVERAGE FOR OPERATINGHPLANTSOF$68M. LGER ASSlfED DISCOUNT RATES OR LONGE TliESE ESTIMATES.

INCREASING ALLOWED LOCAL PEAK POWER (SAE TOTAL POWER) ESULTS I AND MANEINERING CAPABILITIES. TilESE ARE C0FPLICATED SUBJECTS AND SEVERAL LIMITING FACTORS. IDEVER, SAVINGS OF $3-fN ER PLANT PER YEAR MAY BE POSSIBLE.

N GENERIC RELOAD ANALYSES AND FEWER REANALYSES OFFER POTENTIAL SAVINGS.

POTENTIAL HARTHAE 01ANGES, LCO, ETC HAVE NOT BEEN EVALUATED b

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LOCA LIMITS ON WESTING 0 LEE PLANTS CURENT TECHNICAL SKCIFICATIONS AND UPDATED FSAR'S INDICATED TilAT:

15% OF W PLANTS AE t0T RESTRICTED BY APPENDIX K LOCA CRITERIA (F = 2.32; LIMITIN 0

LBLOCA PCT < 2000*F) 41% HAVE MDDERATE OKRATIONAL ESTRICTIONS LIMITATIONS ON LOAD FOLLOWING LIMITATIONS ON STEAM GENERATOR RIBE PLUGGING (Fg = 2.32; LIMITING LBLOCA h

PCT > 2000*F) i 44% HAVE STRONG OPERATIONAL RESTRICTIONS f

LIMITATIONS AS ABOVE INCREASED COE MDNITORING EQUIRENNTS POSSIBLE DIFFICULTIES IN ACHIEVIfE FULL POWER f

(E.G., D. C COOK 2; F = 1,97; LIMITING LELOCA PCT = 2187'F)

[

0 2,2; LIMITING LBLOCA PCT > 2100*F)

(GENERALLY F0 f

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REGULATORY ANALYSIS-SAFETY IWACT OfANGES Mil 01 MAY ESULT FROM PROPOSED RULE COULD RESULT IN POSSIBLE EGA SAFETY IWACTS.

DO NOT EC0ffEND ATTENTING TO OUANTIFY ET IWACT ECAllSE:

EGATIVE ASRCTS EXAMINED EE FOUND TO E SMALL COWARED TO UNCEPTAltRY IN OV Q

i RISK MAft/ OF TIE POSITIVE IWACTS, M1101 WE BELIEVE TO E REAL, AE HIGILY SUBJECTIVE.

t MAJOR RISK IWACT BELIEVED TO RESULT FROM POTENTIAL OiANGES TO PLANT EQlIRTNT M1IOi, MlILE POSSIBLE UNDER TIE PROPOSED RULE, IS NOT T100Gif TO BE A LIKELY ESULT.

TlE PROPOSED RULE MAY:

ALLEVIATE OVERLY TIG1T SETPOINTS; REDUCING NEEDLESS SCRAMS ALLEVIATE OVERLY TIGIT DIESEL GEEPATDR START TIES; INCEASING DIESEL RELIABILITY PERMIT NEUTRON FLUX PROFILES WlIOl REDUCE FLUENCE ON VESSEL AND CORRESP0 RISK POSSIBLY PETHIT POWER INCEASE l

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SIM9RY A RULE EVISION HAS BEEN PROPOSED BASED ON INCREASED [710h1 EDGE OF E GAIED SINCE TIE RULE WAS kRITTUl TIE EFFECT OF TIE RULE IS TO PEDUE titECESSARY PLANT PESTRICTIONS WITil A FOR ECONOMIC BENEFIT WITli NEGLIGIBLE EFFECTS ON SAFETY.

RULE INCORPORATES TE EXISTING LICENSING KT110DS SO AS TO NOT PLAE ANY N

BURDEN ON PLANTS NOT NEEDING OR DESIRING TO MAKE USE OF NEW RULE PROVISIONS, 4x e

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5 APPENDIX X RESPONSE TO ACRS COMMENTS ON m$w REVISION OF ECCS RULE W =l x5" x"O Bud t

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hE RESPONSE TO ACRS COWENTS ON I

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REVISION OF ECCS RULE B ACRS MEETItE 317

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SEPTEMBER 11, 1986 I

I WILLIAM BECKNER 3

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(

ACRS CatENT:

SHOULD APPENDIX K BE AVAILABLE AS AN OPTION FOREVER OR SHOULD IT EVENTUALLY A

RESPONSE

N TEE APKARS TO BE POTENTIAL BENEFITS FROM A REALISTIC CALCULATION AS OPPOSED TO CONSERVATIVE, i

BUT UNREALISTIC APP 90lX K CALCULATIONS. TliUS, IT WOULD BE " NICE" TO EVENTUALLY HAVE ALL T

PLANTS LICENSED lEDER TE NEW ETlID. HOWEVER, IT IS DIFFICULT TO JUSTIFY FORCING ALL LICB1 SEES TO SUBMIT NEW EVALUATION MODELS, PARTICULARLY IN LIGHT OF TliE BACKFIT RULE.

A CASE FOR SAFETY IS HARD TO MAKE BECAUSE OF TE LARGE CONSERVATISM IN APPENDIX K TIE COST WOULD BE HIG1. SUCH AN ACTION WOULD ALSO BE DIFFICULT TO JUSTIFY EVEN FOR NEW PLANTS.

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ACRS C0ftENT:

WHAT GUIDANCE IS TIE REG 11ATORY GUIDE EXPECTED TO PROVIDE? MiAT IS T GUIDE? 10RE GUIDANE SHOULD BE PROVIDED ELATIJ TO TIE UNCERTAINTY EVA IS A KEY PART OF TIE EW ETHODOLOGY.

ESPONSE:

THREE POTENTIAL AREAS OF GUIDANCE HAVE BEEN CONSIDERED:

1.

CLARIFICATION OF "HIGi PROBABILITY"-

O Ch A MMOR OBJECTIVE OF TE REGULATORY GUIDE IS TO CLARIFY TE RULE WOR I

0F PROBABILITY THAT TE CRITERIA WOULD NOT BE EXCEEDED." SRCIFIC EQUIR N

INTENTIONALLY LEFT OUT OF TE RULE S0 AS TO AWID TIE POTENTIAL FOR LITI MATHEMATICS AND STATISTICS. TE GUIDE PROVIDES SRCIFIC CRITERIA AND GUID AS TO M1AT TIE STAFF E(llIES TO EET TIE RULE INTENT.

2.

BEST ESTIMATE 10DELS-TIE GUIDE PROVIDES GEERAL PRINCIPLES TO BE USED IN EALISTIC KDELS.

IN ADDITION, TIE STAFF IS REVIEWING SRCIFIC MODELS TllAT MAY BE LISTED AS " ACCEPTA EALISTIC CALCULATIONS. IlE DEGREE OF SRCIFIC GUIDANCE ON t00ELS THAT WIL INCLUDED IN TIE GUIDE HAS NOT YET BEEN DETERMIED AND PUBLIC 00ffENT IN WILL BE SOLICITED.

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3.

[NCERTAINTY ltTH000 LOGY-THIS SECTION ONLY PROVIDES BROAD PRINCIPLES W/0 DETAILED KTHODOLOGIES.

DECIDED NOT TO PROVIDE DETAILED GUIDANCE IN THIS AREA BECAUSE:

N A.

TIE INDUSTRY IS ALREADY WORKING ON TIEIR.0WN ET10DS AND IS IN MANY AREAS OF TIE NRC STAFF IN THIS AEA.

B.

SINCE THIS IS A NEW AREA, TE STAFF WANTS TO PROVIDE MAXIlVI INCENTIVE AfD FLEXIBILITY FOR TlE INDUSTRY TO DEVELOP AND PROPOSED SKCIFIC ETHODS.

a

ACRS C0ftENT:

M1AT IS TE OBJECTIVE OF TE C0ffEf0IlN OF ECCS RESEARO1? WHAT IS TIE RELATI0f6 HIP OF TE C0ffENDILN OF RESEAR01 TO TlE EGULATORY GUIDE AND M1AT IS TE STATUS OF TE C0ffENDIlN ONCE REFERENCED BY TIE GUIDE? TE TWO D0GIENTS SHOULD BE CONSISTENT.

RESPONSE

TIE OBJECTIVE OF TE C0ffENDIlN OF ECCS RESEARCH IS TO PROVIDE A BROAD " ROAD MA TO lliE EXTENSIVE ECCS RESEAR01 TliAT HAS BEEN PERFORED SINCE TIE ECCS RULE WAS WRITTEN.

IT SUPPORTS TlE RULEMAKING IN TE SENSE THAT IT PROVIDES EVIDEN& THATg TIE TEON0 LOGY NOW EXISTS TO PERFORM REALISTIC CALCULATIONS OF ECCS PERFOR i

TE EGtLATORY GUIDE REFERENCES TIE COPPENDIlN AS A SOURE OF INFORMATION ABOUT T

THIS TEON0 LOGY. HOWEVER, TE REGULATORY. GUIDE LANGUAGE HAS BEEN CAREFULLY WRIlTEN S0 THAT MDDELS OR DATA CONTAIED IN TIE C0ffENDIlN AE NOT, BY VIRTIE OF REFERENCE ALONE, TO BE CONSIDERED " ACCEPTABLE" IN TE REGULATORY SENSE AND THUS NOT SUBJECT TO EVIEW. THIS WAS DONE BECAUSE TIE C0ffENDIlN IS A VERY BROAD AND EXTENSIVE SlPNARY. AS SUCH, A DETAILED REVIEW 0F WHAT RESEARGi IS ACCEPTABLE UNDER WHAT CONDITIONS WOULD NOT BE PRACTICAL.

llE STAFF AGREES TliAT TE TWO DOCLNENTS SHOULD BE CONSISTENT. TllIS HAS BEEN AN OBJECTIVE FROM TIE START AND WE WILL ATTBfT TO WET THIS OBJECTIVE.

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ACRS C0 TENT:

TIE C&FENDILM OF ECCS RESEARQi NEEDS W6RK IN SEVERAL AREAS. TIE ORGANIZATION COULD BE DIANGED TO MAE IT MDRE READABLE, llE DEFINITION OF ISSUES AND THE RELATIONSHIP BE M EN ISSUES (OlAPTER 11) Am RESULTS (OiAPTER 7) I WROVED, A2 MAJOR ITBOVEENTS MADE TO TE UNCERTAINTY DISCUSSION. PARTS OF TE DOCLENT AE S0 NEGATIVE S0 AS TO APPEAR NOT TO SUPPORT TIE RULE REVISION.

ESPONSE:

N TIE LARGE SIZE OF THIS REPORT HAS MADE TlilS A WNTENTAL TASK AND THUS DELAYED TIE

{-

COWLETION OF TIE REPORT. THIS DOClMNT IS A ROUGH DRAFT BEING EVIEWED BY TIE ACRSj' i

IN PARALLEL WITH llE ES STAFF REVIEW. E GEERALLY AGREE WITH TIE ACRS C&KNTS AM) ELCOE SKCIFIC INPUT TO ASSIST LE IN TE TASK 0F S20TlilNG SOE VERY ROUGH EDGES. h E ALSO ECOGNIZE TIE IWORTANCE OF A GOOD DOCTENT TO BEURE A Sm0111 RULEMAKING ANi ARE GIVING TlilS EFFORT A HIGi PRIORITY.

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ACRS C&f0 6:

M1AT IS TE IWACT OF TE BACKFIT RULE ON NRR EVIEW 0F EW EM's? IF A LICENSEE APPLIES FOR AN NODelT TO HIS LICENSE RELATIVE TO A EW EM AE BELIEVES THAT TE STAFF IS IWOSING UNDUE CONSERVATISM OR CHANGE TO A PROPOSED MODEL, MAY TE LICENSEE INVOKE TE BACKFIT RULE?

RESNNSE:

ACCORDING TO OGC STAFF, AS LONG AS TE ISSLES AE COWIED TO TIE ADEQUACY OF TIE EM AE WHETER IT EETS TE CRITERIA, TIEE SHOULD BE NO BACKFIT. TIE BACKFIT RULE COULD BE INVOKED ONLY IF TIE STAFF RAISED A DIFFERENT ISSLE.

TlilS CONCERN DIFFERS FROM TlE " INSULATION ISSUE" PROWTING CONCERN OVER TIE ADEQUAC OF StW PLNP DESIGN. IN CASE OF TIE ECCS RULE, TE LICENSEE ALWAYS IIAS TE FALLBACK OF USING APPE E IX K AND HIS EXISTING EM.

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STATUS EVISED ECCS RULE EQUIES OUANTITATIVE WASUE OF LNCERTAINIY (95% PROBABILIT KAK CLAD THPERATUE (PCT)

NRC/ES IS PROPOSING COWRElENSIVE Ell 0D0 LOGY, FOR INDEPENDENT PEER REVIEW IN OCTOBER,1986, QMmSED OF BOTH CODE VS. DATA CO WARISONS AND SYSTEMATIC EXAMINATION OF CODE MODELS AND CORRELATIONS g

I ET1000 LOGY ORIGINALLY DEVELOPED F0 Q

C0HERENT INTEGRATION OF ICAP ASSESSENT RESULTS BASIS FOR ESTABLISHING CODE APPLICABILITY TO ANALYZE B&W PLANTS (PF0 GRAM AT T/

TEONICAL INTEGRATION CENTER)

WKN APPLIED TO ECCS RULE, BOTH ASKCTS COWLBENT EACH OTTER IN A SINGLE C0[ RENT lETHODOLOGY O

e

PROPOSED NRC/ES UN&RTAINTY PETH000 LOGY EOLIIRES, AT START CODE D00KNTATION (MANUAL, USER GUIDE, 0/A DOCTENT, ASSESSENT REFORTS)

ASSESSENT AGAINST DATA IDENTIFICATION OF IEY PROCESSES A?0 PARAETERS FOR GIVEN SCENARIO UNCERTAINIY IN INPUT Ato BOLNDARY QPOITIONS UNCERTAINTY IN FUEL PAR /MTERS UNCERTAINIY IN EXPERIN NTAL DATA FOUR FACIORS ADDRESSED BY NRC/ES UNCERTAINTY KT10DOLOGY N

k 1.

CAPABILITY OF CODE t0DELLING TO ANALYZE AND SCALE UP MAJOR PHENDENA EXPEClED IN TIE ACCIDENT SCENARIO j

QUANTITATIVE EASUE OF CODE BIAS Af0 STAT 0ARD DEVIATION TO WITHIN 95% PROBABILITY, 2.

INCLUDING NODALIZATION SENSITIVITY WIT 111N USER GUIDELIES 3.

DETECTION OF C0ffENSATING ERRORS WilOl WOULD MASK TIE PERCEIVED ACCURACY OF TE}

CALCULATED PCT I

4.

SCALABILITY OF TIE CALCULATED PCT l

TAl(EN TOGET11ER TESE FOUR FACTORS CONSTITUTE TIE CODE APPLICABILITY TO ANALYZE A SCENARIO IN A GIVEN FULL-SCALE VENDOR GEDETRY

}

9 e

PROF 0 SED NRC/RES UNCERTAINW KTHODOLOGY:

DETAILS OF EA01 FACTOR 1.

CODE t0DELING CAPABILITY A.

IDENTIFY AND DEFIE MAJOR PHENDENA EXPECTED, FOR A SCENARIO OR SET OF SCENARIOS B.

ASSESS CODE MODELS TO CALCULATE AND SCALE TllESE PHEfDENA C.

IF MODELLING DEFICIENT, ESTIMATE EFFECT ON ACCURACY OF CALCULATED RESULT D.

IDENTIFY SCENARIOS WlIOi CANNOT BE ADDRESSED BY CODE h>

2.

QUANTITATIVE WASURE OF CODE UNERTAIN1Y (BIAS AND STANDARD DEVIATION)

A.

CALCULATE APCT = KASURED - CALCULATED, FOR SEVERAL TESTS IN DIFFERENT SCALED FACILITIES AND SIMILAR (LOCA) SCENARIOS B.

COMBINE ALL A'S TO CALCULATE CODE BIAS C.

ESTIMATE PROBABILITY DENSITY FUNCTION WITH HIST 0 GRAM 0F FREQUENCY OF A VS A; DERIVE STANDARD DEVIATION FROM CURVE MAT 0ilNG TllIS HISTOGRAM D.

WITlDJT FURTHER ANALYSIS, WOULD ASSifE THIS BIAS AND STANDARD DEVIATION CAN BE EXTRAPOLATED TO FULL-SCALE E.

HOWEVER, WE STILL NEED TO EVALUATE EFFECTS OF C0FFENSATING ERRORS AND SCALABILITY TO CONFIRM EXTRAPOLABILITY O

O O

3.

COIPENSATING ERRORS A.

REPEAT (2) FOR OTER KEY PARAETERS MAKE ENGINEERING JUDGEENT ON ACCURACY RE0JIRED FOR TESE KEY PARAETERS; TEN B.

IDENTIFY FROM (1), MDDELS/ CORRELATIONS LEADING TO COEENSATING ERRORS AND DETER C.

SENSITIVITY AND IIPORTANCE D.

EITER IMPROVE 10DELLING OR ADD EFFECT TO UNCERTAINTY I4, SCALABILITY 4 i PLOT BIAS OF EACH FACILITY VS SCALE AND LOOK FOR SIGNIFICANT SLOPE; AT SAE TIE N

A.

B.

EXAMINE CODE MODELLING/ CORRELATIONS:

DATABASE ADEQUATE YES FOR SCALING NO y

, N0 10 DEL / CORRELATION IIPORTANT TO SCENARIOS YES U

,NO SENSITIVITY OF CORRELATION TO SCALING IS LARGE YES V

V OUANTITATIVE EXTRAPOLABILITY CODE SCALABILITY IN OttSTION TO FULL-SCALE IS RRMITTED UNTIL MODELLING

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-2:00 PM POWER DESCENT DELAYED FOR 9 HOURg 1:19 AM OPERATORS PULL ROOS SY LOAD DISFATCHin SEYONO ALLOWABLELIMITS TO ECCS DISCONNECT 20 NEACM J00 MWrh 1

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1 DENSITY RATIO OF VAPOR TO LIQUID AT PREVAILING WHERE

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9//2/F4 APPENDIX XIII.

O REPORT Ofl THE IAEA MEETING ON THE CHERN0BYL ACCIDENT REPORT ON THE IAEA MEETING ON THE CHERNOBYL ACCIDENT HAROLD R. DENTON THEMIS P. SPEIS BRIAN W. SHERON FRANK J. CONGEL O

PRESENTED TO THE NRC COMMISSIONERS SEPTEMBER 3, 1986 b-/Oh I..-..

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i-KEY DESIGN AND NUMAN FACTOR ASPECTS WHICH CONTRIBUTED TO ACCIDENT SEVERITY HUMAP,' FACTOR ASPECTS TEST PROCEDURES VIOLATED CERTAIN SOVIET SAFETY REGULATIONS, OPERATORS VIOLATED CERTAIN PARTS OF TEST PROCEDURE, 4

SOVIETS BELIEVE 0,PERATORS LOST SENSE OF VIGILANCE TOWARDS

SAFETY, DESIGf' ASPECTS APPARENT SIMPLICITY WITH WHICH SAFETY AND PROTECTION SYSTEMS COULD BE OVERRIDDEN, SLOW CONTROL R0D INSERTION RATE.

POSITIVE VOIE REACTIVITY FEEDBACK, i

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G - rod AC-2 M - steam flow of the drum separator A - power B - total reactivity H - rod AC-3 N - T,,, (T-temperature 1 l

C - pressure in the drum separator K - flow of the main 0-circulation pump P-D - power E - rod AC-1 (automatic control rod)

L - feed water flow 5 - water level in the drum separator Il n

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i Fault in measwoment settien of 123%o-

'nd of AZ 5 automatic reydators AR1, AR2 sper aten Over presswe in drum separatcr irt up Triggermg of f ast-atteg steam-dump system i

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O Lt(VIRONMENTAL RELEASE TERMS

~

IN THE CHERNOBYL ACCIDENT OVERVIEW THE SOVIETS HAVE DEVELOPED ESTIMATES FOR ENVIRONMENTAL RELEASES OF RADIONUCLIDES AS A FUNCTION OF TIME, THE ESTIMATED RELEASE FRACTIONS APPEAR TO BE:

1)

CONSISTENT WITH THE SOVIET CONCEPTION OF THE ACCIDENT SCENARIO, 2)

CONSISTENT WITH MEASUREMENTS MADE IN OTHER COUNTRIES, i

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ESTIMATED RELEASES OF RADIONUCLIDES O

FROM inE AcciDEur unir Or CHERN0BYL NUCLEAR POWER PLANT

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Total release by Way 6,1986.

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SOVIET ESTIMATES OF THE CHERN0BYL SOURCE TERM ESTIMATED RELEASE TO ENVIRONMENT g.

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NOBLE GASES 50 MCI-0THER RADIONUCLIDES 5D_fiCL 100 MCI AVERAGE RELEASE FRACTION EXCLUDING NOBLE GASES 3.5%

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ESTIMATED RELEASE FRACTIONS FOR KEY ELEMENTAL GROUPS NOBLE GASES UP TO 100%

IODINE 20%

TELLURIUM 10%

CESIUM 13%

O-INVOLATILE GROUPS 3 - 4%

PHASES OF RELEASE PHASE 1. APRIL 26 RELEASE ASSOCIATED WITH VERY HIGH FUEL TEMERATURES IN EXCURSI0t HIGH RELEASE OF VOLATILE FISSION PRODUCTS PHASE 2. APRIL 26 - MAY 2 REDUCED RATE OF RELEASE RELEASE PROPORTIONAL TO INITIAL INVENTORY OF FUEL (IMPLYING TRANSPORT AS FUEL PARTICLES)

O

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OQ SOVIET ESTIMATES OF THE CHERNOBYL SOURCE TERM (CONT.)

PHASE 3. MAY 2 - MAY 9 INCREASED RELEASE OF RADIONUCLIDES, PARTICULARLY VOLATILE SPECIES IMPLIED HEATUP 0F CORE PHASE 4. MAY 10 -

RAPID DECREASE IN RELEASE OBSERVED (MECHANISM FOR PRODUCING TO U 038 OX10 ail 0N OF U02 FULE AEPOSOLS)

G O

9 A

i FIRE FIGHTING 3 TEAMS WENT TO SITE IMMEDIATELY i,

l FIF.ES LOCALIZED TO ROOFS BY 2:30 A.M.

FIRES QUENCHED BY 5:00 A.M.

i i

OBJECTIVE WAS TO PREVENT SPREAD TO UNIT #3 i

PROTECT CABLE ROOMS h

OIL TANK ROOMS i

i USED PRIMARILY WATER T0.EXTINGUISti FIRES, FIRES WERE MAINLY ON SURFACE

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SOVIETS IDENTIFIED " FIRE-FIGHTING LESSONS LEARNED" l

LIST OF PROPOSALS GIVEN T0 IAEA FOR CONSIDERATION f

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O ENTOMBMENT OF UNIT 4 A CONCRETE-WALLED BUILDING WILL BE CONSTRUCTED AROUND UNIT 4 TO ENSURE CONFINEMENT OF RADI0 ACTIVITY, AN INNER CONCAETE PARTITION WALL IN THE TURBINE HALL WILL SEPARATE THE THIRD AND FORTH UNITS, A METAL PARTITION WALL WILL SEPARATE UNITS 2 AND 3.

i A PROTECTIVE ROOF WILL COVER THE TURBINE HALL, 1

l THE CENTRAL HALL AND OTHER REACTOR ROOKS WILL BE SEALED OFF.

I CONCRETE WILL BE POURED OVER DEBRIS IN SOME AREAS, 1

I i

h-///

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ENTOMBMENT OF UNIT 4 (CONT)

O CLOSED LOOP AND OPEN LOOP SYSTEMS WERE CONSIDERED.

CLOSED LOOP SYSTEMS ARE MORE DESIREABLE FROM THE VIEWPOINT

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OF PUBLIC PERCEPTION, AN OPEN LOOP SYSTEM WAS SELECTED BECAUSE:

PERMITS SIMPLIFIED CONTROL OF HYDROGEN (DILUTION),

EASIER MONITORING AND MAINTENANCE.

A NEGATIVE PRESSURE DIFFERENTIAL WILL BE MAINTAINED BETWEEN THE EUILDING AND THE ENVIRONhENT, Oh!TS 1 AND 2 ARE EXPECTED TO RESUME OPERAT10N IN 1986, UNIT 3 WILL UNDERGO A THOROUGH SAFETY REVIEW BEFORE RESUMING OPERATION.

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PROPOSED RBMK-1000 MODIFICATIONS (SHORT AND LONGER TERM)

CONTROL RODS PERMANENTLY INSERTED IN THE CORE TO A DEPTH OF l

1.2M.

ABSORBER-TYPE CONTROL RODS ALWAYS PRESENT IN THE CORE WILL BE INCREASED TO 80, TO FURTHER REDUCE THE POSITIVE VOID COEFFICIENT (BY A FACTOR OF 2): TEMPORARY MEASURE UNTIL FUEL ENRICHMENT IS INCREASED TO 2.4% FROM THE PRESENT 2.0%.

ADDITIONAL INDICATORS OF THE CAVITATION OF THE' MCPS WILL BE INSTALLED.

i AUTOMATIC CALCULATION OF REACTIVITY WITH EMERGERCY SHUTDOWN i

SIGNAL WHEN EXCESS REACTIVITY MARGIN < SPECIFIED LEVEL.

0FGANIZAT10NAL STEPS TO REINFORCE TECHNOLOGICAL DISCIPLINE AND TO IMPROVE QUAllTY OF OPERATIONS, EVALUATE ADDITIONAL DIVERSE AND FAST ACTING ABSORBERS SUCH AS LIQUID, GAS, OR SOLID FOR FUTURE USE.

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I PLUME MOVEMENT TOWARD THE WEST AND NORTH (AROUND PRIPYAT)

IST LAY i

TOWARD THE NORTH (THROUGH PPIPYAT) 2ND-3RD DAYS i

I 1

TOWARD THE SOUTH (TOWARD KIEV) l 4TH DAY 4

)

PLUME HEIGHT EXCEEDED 1200M ON APRIL 27 i

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DOSE RATE IN PLUME. 1 R/HR AS MEASURED BY AIRPLANES 5-10 KM FROM $1 j

L PLUME CONTAlkED FISSION AND ACTIVATION PRODUCTS (CS-134 s NP-239 i

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O INDIVIDUAL DOSES 4 0RKERS h'ORKERS - PRIMARILY FIREFIGHTEPS ONE DIED AT 6:00 AM FROM BURNS IST VICTIMS ANOTHER WAS APPARENTLY BURIED IN COLLAPSED SECTIONS OF BUILDING 12 HOURS thTO ACCIDENT - 350 WORKERS CLOSELY EXAMINED FOR BLOOD CHANGES AND OTHER PHYSICAL SYMPTOMS OF HIGH RADIATION EXPOSURE 129 PEOPLE HOSPITALIZED IN SFECIAL CENTER IN MOSCOW O

84 DIAGNCSED AS EXPERIENCING ACUTE RADIATION ILLNESS -

IMMEDIATE SYMPTOMS LE Pl? L N ? ExrcSuppc A';r. FEFE CESEPVED Fort 1 'c 1.5 Po':THE i

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INDIVIDUAL DOSES - WORKERS (CONT.)

TOTAL OF 203 PERSONS HOSPITALIZED WITH ACUTE RADIATION SICKNESS - NO MEMBERS OF THE GENERAL PUBLIC WERE INCLUDED IN THIS GRCUP NO EVIDENCE OF NEUTRON EXPOSURE BASED ON ANALYSIS FOR NA-22 l

SUBSTANTIAL AMOUNTS OF CS AND PU FOUND IN VICTIMS DOSE DISTRIBUTION - ALL 203 HOSPITALIZED YlCTIMS RECEIVED

> 100 REM.

OF THAT GROUP, 35 PEOPLE EXCEEDED 400 REM UP TO A, MAXIMUM OF 1200-1600 REM i

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areas In t..n 30 km - zone arround ChNPS b

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O EMERGENCY RESPONSE MEASURES TAKEN IMMEDIATELY AFTER ACCIDENT PRIPYAT POPULATION (45,000) ADVISED TO REMAIN INDOOR AND CLOSE WINDOWS ON APRIL 26 OPEN-AIR ACTIVITIES BANNED AT ALL KINDERGARQENS, AND SCHOOLS; IODINE PROPHYLATIC TREATMENT GIVEN THERE EVACUATION OF PRIPYAT BEGAN AT 2 PM ON APRIL 27 AS DOSE RATE WORSENED; COMPLETED BY 5 Ph THE SAME DAY l

4 REMAINING POPULATION (90,000) FROM 30-KM ZONE EVACUATED IN NEXT,FEW DAYS BECAUSE OF CONTINUING CONTAMINATION DUE TO CHANGING PLUME DIRECTION CONSUMPTION OF MILK CONTAINING 1 x 10-7 C1/L OR MORE OF l-131 WAS BANNED ALL CHILDREN FROM 30-KM ZONE WERE SENT TO SUMMER SAN!TORIUM IN THE COUNTRY O

R -/3/

O EMERGENCY REPONSE MEASURES TAKEN (CONT.)

STANDARDS FOR PERMISSIBLE LEVELS OF

~

CONTAMINATION IN FOOD PRODUCTS ISSUED BEGINNING EARLY IN MAY 1986 1240 DOCTORS, 920 NURSES AND SEVERAL THOUSAND SUPPORTING ASSISTANTS MOBILIZED TO PROVIDE MEDICAL CARE OF EVACUEES EACH EVACUEE' EXAMINED; BLOOD TESTS CARRIED IN SOME CASES EXAMINATION AND TESTS OUT; O

REPEATED EVACUEES WHO SHOWED IRREGULARITIES WERE HOSPITAll2ED IN SPECIAL SECTIONS SET-UP AT CENTRAL REG IONAL HOSPITALS LONG-TERM PROGRAMS ARE BEING ESTABLISHED FOR MEDICAL AND BILOGICAL MONITORING OF POPULATION AND PERSONNEL O

A -/3D z3 5

O DECONTAMINATION--OFFSITE BUILDING AND HOUSES ARE BEING DECONTAMINATED BY SPRAYING DECONTAMINATION SOLUTION AFTER WASHING, CONTAMINATED Soll AROUND THE BUILDINGS TURNED OVER OR REMOVED WITH BULL-DOZERS AND TAKEN AWAY TRANSPORT VEHICLES DECONTAMINATED USING

~ SOLUTIONS BY SPRAYING AND STEAM JETS.

LONG iERM DECONTAMINATION' PROCEDURES OF THE ENVIRONMENT ARE BEING RESEARCHED 1-A -/33 3

- - = -.

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DECONTAMINATION -- PLANTSITE

~

THE SITE, THE EJRBINE BUILDING ROOF AND THE SIDES OF THE

, ROADS TREATED WITH RAPID POLYMERIZING SOLUTIONS TO REINFORCE UPPER LAYERS OF S0ll AND PREVENT DUST FORMATION.

THE NUCLEAR POWER PLANT SITE WAS DIVIDED INTO ZONES FOR THE PURPOSES OF DECONTAMINATIONi DECONTAMINATION IN EACH ZONE CARRIED OU) IN THE F'OLLOWING

~-

ORDER:

REMOVALbFDEBRISANDCONTAMINATEDEQUIPMENTFROM S I,TE ;

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REMOVAL OF A SOIL LAYER, 5-10 CM THICK, AND TRANSPORTATION TO REPOSITORIES)

LAYING, WHERE NECESSARY, OF CONCRETE SLABS OR FILLING IN WITH CLEAN SOIL; COVERING OF SLABS AND NON-CONCRETED PARTS OF SITE WITH FILM-FORMING MATERIAL; l

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^y CPS 8/28/86 APPENDIX XIV PROPOSED CONTAINMENT PERFORMANCE DESIGN

/

OBJECTIVE i

PROPOSED CONTAINMENT PERFORMANCE DESIGN OBJECTIVE 1.

INTRODUCTION At the 316th meeting, August 1986, I was directed to prepare a

" straw man" Containment Performance Design Objective for consideration by the ACRS as part cf its recomendations for design criteria for improved reactors.

This has been done, and is offered herewith.

It is based on previous recomendations by the ACRS and is more or less consistent with what the Comission has said in its Policy Statement and what individual Comissioners have proposed.

The remainder of this report is arranged as follows:

2.

Historical Review. An attempt to sumarize previous ACRS recomendations and provisions or comments in the Policy Statement. The first may be incomplete; it begins in 1982.

3.

Containment Performance Design Objective. A statement of the ground rules followed, a statement of the proposed CPD0 itself,-

A comments relating it to previous recomendations and the Policy Statement, and coments on its application and possible significance.

4 Design to Meet the CPDO. A brief list of some of the design changes that might be made or required to meet the CPDO.

5.

Questions. A very brief and incomplete list of what we might need to know before such a CPD0 is adopted.

4 I

2. HISTORICAL REVIEW ACRS RECOMMENDATIONS With no fear of prejudicing what we might say now, I will review briefly below what we have said about Containment Performance Objectives in previous ACRS letters.

4 (1) Letter of 9 June 1982:

Coments on Proposed Policy Statement on Safety Goals for Nuclear Power Plants (NUREG-0880, a Discussion Paper)

"For plants yet to be designed, it may be practical to I

set containment performance standards for accidents leading to large-scale-core melt, but not automatically involving a direct loss of containment integrity by 4 -/N

Containment Performance Design 2

' Objective the frequency of core nelt accidents that are directly i

coupled with an early loss of containment integrity is and must be kept very low, recent studies indicate that

~

it is practical to establish stringent performance requirements on containment capability for other core melt accidents. We believe that additional study is needed before numerical guidelines are set for the containment of future plants."

4 (2) Letter of 15 September 1982: ACRS Coments on the NRC Staff Questions to the Comission Concerning the Policy Statement on Safety Goals for Nuclear Power Plants.

"We believe that, in view of the continuing uncertainties to be expected in the art of PRA and a continuing inability to satisfactorily treat all initiators and other contributors to core melt frequency, and in view of the potentially very large differences in release magnitudes among different core melt accidents, containment performance design objectives are needed and should be developed expeditiously."

(3) Letter.of 15 September 1982: ACRS Report on the Draft Action Plan for Implementing the Comission's Proposed Safety Goals for Nuclear Power Plants.

"We believe that priority should be given to developing containment performance criteria for several reasons, including the following:

There are major uncertainties in the calculation of a.

statistical health effects from very small doses to large numbers of people.

i b.

There are large uncertainties in calculation of accident dose.

Evacuation models, for example, are j

fairly arbitrary and do not reflect the potential I

effects of earthquakes or offsite loss of power on l

the effectiveness of emergency actions, Assumptions concerning land areas dich woula c.

require interdiction and problems in large-scale decontamination require futher study.

d.

Uncertainties in prediction of core melt frequency would be compensated, at least in part, by a containment having a significant potential to mitigate core melt accidents."

(4) Letter of 9 August 1983: ACRS Coments on Proposed Safety Goal Evaluation Plan.

"We observe that the proposed safety goals contain no It is design objective for containment performance.

stated that the evaluation process will include a review [-/

of whether containment performance is to be a specific e

-.--l. -

Containment Performance Design ~

3 Objective design objective. Discussions with the NRC Staff indicate that they have concluded that uncertainties in containment performance are too great to make a f~s f

1 performance objective meaningful at this time.

It is V

strange that the NRC Staff considers the uncertainty in describing the progress of a large scale core melt to be significantly less than the uncertainty in describing containment performance. We continue to believe that containment performance objectives are important as an indication of the need for mitigation, just as the core melt design objective is an indication of the emphasis on accident prevention."

(5) Letter of 17 July 1985: ACRS Coments on Proposed NRC Safety Goal Evaluation Report.

"The NRC Staff has not developed a containment performance. guideline, nor has any serious NRC Staff effort to do so been apparent to the Comittee. The ACRS continues to believe, as it did in its report of June 9, 1982, that the' development of a containment performance guideline warrants high priority, and recommends that the Comission require early NRC Staff attention to this matter as part of maintaining its defense-in-depth principle. Approximate compliance to an appro'priate criterion should be an NRC objective."

(6) ' Letter of 19 March 1986:

ACRS Coments on Proposed Safety 3

Goal Policy.

"In a severe accident, it is the releases from the containment which constitute the risk to the health and safety of the public. Thus, risk cannot be assessed without a judgment on containment performance. We reiterate our recomendations to develop a containment performance objective."

(7) Letter of 15 April 1986: Additional ACRS Coments on Proposed NRC Safety Goal Policy Statement.

"We believe the Comission should adopt certain performance guidelines as one satisfactory means to assure conformance with the safety goal objectives.

these guidelines should be structured so that the principle of defense-in-depth is maintained....

"We propose that the plant performance guidelines be regarded as fully acceptable surrogates for the safety goal objectives.....

"There should be two performance guidelines and consideration should be given to the development of a third.

"The first guideline should be that the chance of a loss of adequate core cooling with consequent severe core damage should be less that 10E-4 per reactor-year for all but a few small reactors.

"The second guideline should relate to containment performance and should be such that the chance of a very large release of radioactiv aterials to the l

Containment Performance Design 4

Objective environment should be less than 10E-6 per reactor-year."

h COMMISSION POLICY STATEMENT d

On 30 July 1986, the Commission issued its " Final Policy Statement on Safety Goals for the Operation of Nuclear Power Plants."

Portions cf that Statement, or of the attached remarks of Comissioners Asselstine and Bernthal, relating to containment performance are cited in the following:

(8) Under the heading V. Guidelines for Regulatory Implementation.

....the staff will require specific guidlines to use as a basis for determining whether a level of safety ascribed to a plant is consistent with the safety goal policy........... The guidance would be based on the following general performance guideline which is proposed by the Comission for further staff examination--

" Consistent with the traditional defense-in-depth approach and the accident mitigation philosophy requiring reliable performance of containment systems, the overall mean frequency of a large release of radioactive material to the environment from a reactor accident should be less than 1 in r

1,000,000 per year of reactor operation."

(9) Comissioner Asselstine, in his additional views, citing accident prevention and mitigation, defense-in-depth, and the ACRS, proposes "the following containment performance criterion:

"In order to assure a proper balance between accident prevention and accident mitigation, the mean frequency of containment failure in the event of a severe core damage accident should be less than 1 in 100 severe core damage accidents."

(10) Comissioner Asselstine further addresses the " general performance guideline" of one in a million probability of a large release, and proposes that this be adopted as a performance objective.

He then defines a "large release" as one "that would result in a whole body dose of 5 rem to an individual located at the site boundary". He points out that this is. consistent with the EPA's emergency planning Protective Action Guidelines and thus would not require evacuation of the public."

l (11) Comissioner Bernthal, in his separate views, proposes the following as being consistent with the Comission's defense-in-depth philosophy:

Ov on average, to occur in the U.S. more than once in 100 "1) Severe core-damage accidents should not be expected, years; "2)

Containment performance at nuclear power plants A -t 39

- /d

Containment Performance Design 5

Objective should be such that severe accidents with substantial offsite damages are not expected, on average, to occur g

in the U.S. more than once in 1000 years; a

"3) The goal for offsite consequences should be expected to be met after conservative consideration of the uncertainties associated with the estimated frequency of severe core-damage and the estimated mitigationthereofbycontainment.[Footnoteomitted]

The term " substantial offsite damages" would correspond to the Commission's legal definition of "extraordinay nuclear occurrence" [5 rem whole body ?].

" Conservative consideration of associated uncertainties" should offer at least 90 percent confidence (typical good engineering judgment, I would hope) that the offsite release goal is met."

3.

CONTAINMENT PERFORMANCE DESIGN OBJECTIVE GROUND RULES This " straw man" Containment Peformance Design Objective (CPD0) is intended to be used in connection with the ACRS recomendations for future plants. Earlier discussion suggests that this would be limited to future standard plant designs for LWRs.

With this

/-

limitation, BWR Mark I and II containments and PWR ice-condenser G

containments would not be included. Some previous discussion indicated also that the proposed ACRS recommendations would not necessarily apply to existing standard plant designs, such as GESSAR II and CESSAR, or to possible reactivated cps or replications of existing plant designs.

However, there is nothing in the proposed CPD0 that would not apply to such plants if they Nor is do not involve Mark I or II or ice-condenser containments.

there anything that would preclude applying the CP00 to future HTGR or LMR designs.

PROPOSED CP00 L

The following is proposed as a CPDO:

The overall mean frequency of a large release of radioactive material to the environment from a reactor accident should be less than 10E-6 per reactor-year.

A "large release" is one that results in a whole body I

dose of 5 rem to an individual located at the site boundary.

This objective is the same as that p oposed by the ACRS in its letter of 15 April 1986 [1.(7) above and included in the Policy O

StatementforconsiderationbytheNRCStaff[1.(8)above].

It l

V orobably is not inconsistent with Commissioner Bernthal's proposal

[1.(11) above], requiring 10E-3 per year for all plants in the U.S. with 90 percent confidence.

The definition of a large release corresponds to that offered by l

A Yo

//

Containment Performance Design 6

ObjectiveCommissionerAsselstine[1.(10)above],andisconsistentwith Commissioner Bernthal's definition of " substantial offsite damage"

[1.(11)above].

-COMMENTS Taken alone, this can be considered a " general performance guideline" [1.(8) above] rather than a containment performance design objective. However, in the ACRS recommendation [1.(7) above], this objective was coupled to a limit on frequency of severe core damage.

Although the Policy Statement does not mention a meaningful limit on frequency of severe core damage, it does relate this performance guideline to defense-in-depth and "the accident mitigation philosophy requiring reliable performance of containment systems.

In any case, it is highly unlikely that the 10E-6 probability can be met either without a containment or by the containment alone.

The definition of a large release allows no credit for evacuation.

In fact, it is intended to define an accident that does not require evacuation. To this extent, it is intended to both protect and assure the public.

If a severe core damage accident should occur--but no evacuation is required--some of the psychological trauma expected from a severe core damage accident might be avoided. This may be idealistic or optimistic in view of the possible bases upon which a decision to evacuate may be made, but it is clearly one reason behind the Commission's proposal.

An important, and perhaps controversial feature of this CPD0 is the calculation of dose to a hypothetical individual at the site This, however, is essential if the CP00 is to be used boundary.

to evaluate standard designs to be ap' proved for use at unknown sites.

Some assumption will have to be made about the meteorology, but this should be tractable since the release point is a function of the scenario and only local meteorology is needed.

I have no idea as to whether this limit on releases will satisfy the health effects safety goals. Most likely, it will be well below the quantitative goals of the Policy Statement.

Such conservatism may be considered desirable or even necessary to offset the uncertainties in estimating the frequency of severe core damage accidents, their course, and the performance of containment systems.

In fact Compliance with this CPD0 will of course require a PRA.

it is the ultimate use of the " bottom line". Since a PRA will be required for all standard plant designs, and probably for any other future designs, this should introduce no more than the usual problems, including what to do about external events such as In this respect, should some thought be earthquakes and floods.

given to external events of such magnitude that' the event itself, absent the presence of the nuclear power plant, would require p

(

evacuation or result in catastrophic consequences?

For the most likely severe-core-damage accidents (SCDAs), the If challenge to the containment is slow overpressure.

unmitigated, this scenario will lead eventually to failure of the

&M/

/p t

Containment Performance D:: sign 7

Objective containment, perhaps by gross rupture but most likely by large leakage.

For this case, the CPD0 will require that the containment either o

(a) be able to resist the overpressure for a long enough time that the eventual release is not "large", or (b) be vented prior to failure in such a manner that the release will not be "large."

For SCDA scenarios leading to very large pressures on the containment at early stages of the accident, it may not be possible to prevent containment failure.

Such scenarios include steam explosion producing a missile, hydrogen detonation on a large. scale, or direct heating.

In such cases the CPD0 must be met by insuring that the frequency of all such accidents is below 10E-6 per reactor-year.

Similarly, accidents involving containment bypass (Event V),

pre-existing leakage, or failure to isolate cannot be mitigated by containment or containment system designs. Again, the CPD0 must be met by keeping the frequency of such accidents, in total, below the 10E-6 criterion.

4. DESIGN TO MEET THE CPD0 Application of the proposed CPD0 could lead to design changes of the following kinds:

,m (v) 1.

Increased pressure capacity of the contairment to prevent or further delay failure by slow overpressure.

2.

Provisions for venting through a filtering medium, and procedures for doing so.

3.

Provisions to reduce the probability of pre-existing leakage, such as continuous leak monitoring to detect gross openings.

4.

Provisions to prevent containment bypass as a result of Event V sequences, if the frequency is not low enough.

5.

Provisions to reduce the probability of hydrogen detonation, if it is not low enough.

It is likely, or at least we can hope, that provisions such as these will be either passive or procedural, and simple enough that complex analyses of their effectiveness will not be required.

5.

QUESTIONS 1.

Most of the discussion above relates chiefly to large dry containments for PWRs. What is not applicable to BWRs with x

Mark III containments? What should be added? How do Bernero's recomendations for BWRs fit in? Would GESSAR II meet the proposed criterion.

f gg

~

/3

Containment Performance Design 8

Objective 2.

Can the proposed CPD0 be achieved?

3.

If so, how much mar in will be provided against the quantitative safety goa s?

For existing designs, will the doses for late containment 4.failure under slow overpressure, or for venting, meet the definition of large releases?

S'.

Could the performance required by the proposed CPD0 be met by requiring certain design features, as Bernero proposes, rather than by the proposed performance criterion?

O

7 1

APPENDIX XV V

ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE 1.

Memorandum, H. Etherington to ACRS members, Graphite-Steam Reaction at Chernobyl, September 9,1986 2.

Announcement No.135, NRC Chairman L. W. Zech, Jr. to all NRC employees, Director, Office of Nuclear Regulatory Research, September 9, 1986 3.

Report, USSR State Committee on the Utilization of Atomic Energy, The Accident at the Chernobyl' Nuclear Power Plant and Its Consequences, Part I.

General Material (Information compiled for the IAEA Experts' Meeting, 25-29 August 1986, Vienna), August 1986 4.

Report, USSR State Comittee on the Utilization of Atomic Energy, The Accident at the Chernobyl' Nuclear Power Plant and Its Consequences, Part II. Annexes 1, 3, 4, 5, 6 (Information compiled for the IAEA Experts' Meeting, 25-29 August 1986, Vienna), August 1986 5.

Report, USSR State Comittee on the Utilization of Atomic Energy.

The Accident at the Chernobyl' Nuclear Power Plant and Its f

Consequences, Part II. Annexes 2, 7 (Information compiled for the y]

IAEA Experts' Meeting, 25-29 August 1986, Vienna), August 1986 6.

Report of the International Task Force on Prevention of Nuclear Tarrorism, a project of the Nuclear Control Institute, June 25, T)86 9