ML20214A452

Minutes of ACRS 320th Meeting on 861211-13 in Washington,Dc. List of Attendees & Viewgraphs Encl

ML20214A452
Person / Time
Issue date:	05/15/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2481, NUDOCS 8705190461
Download: ML20214A452 (224)

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TABLE OF CONTENTS -

MINUTES OF THE 320TH ACRS MEETING

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DECEMBER 11-13, 1986 I. Chairman's Report (0 pen)....................................... 1 II. Implications of the Chernobyl Nuclear Plant Accident (0 pen)..... 2 III. Reactor Operations (0 pen)...................................... 8 A. Drywell Shell Corrosion at Oyster Creek.................... 8 B. Loss of All Component Cooling Water at Byron-2. . . . .. . . . .. .. 9 C. High Pressure Injection Nozzle External Surface Damage Due to Boric Acid Corrosion at AN0-1........................... 9 D. Design Problems in Plants Operating and Under Construction. 10 E. Loss of Of fs i te Power a t Pil grim. . . . . . . . . . . . . . . . . . . . . . . . . . . 11 F. Failure of Main Feedwater Pipe at Surry-2.................. 11 G. Loss of Low Pressure Service Water at Oconee-2............. 13 IV. Containment Performance (0 pen)................................. 14 V. Reactivation of Nuclear Power Plants (0 pen).................... 20 VI. Pressurized Thermal Shock (0 pen)............................... 23 VII. Meeting with the NRC Commissioners (0 pen)...................... 27 A. Effectiveness of the Programs Which Address Generic and Unresolved Safety Issues............................... 27 B. Committee Views on Advanced and Standardized Nuclehr Power Plants..................................................... 29 VIII. Resolution of ACRS Corinents on Shearon Parris Nuclear Power Plant (0 pen)................................................... 33 IX. Report of TVA Management (0 pen)................................ 34 X. Radwaste Maragement and Disposal (0cen)........................ 34 XI. Hanford Reactor Temporarily Closed (0 pen)...................... 36 XII. Nuclear Plant Security (Closed).........;...................... 37 bc71cgnQ CRIGINAI.

8705190461 070515 PDR ACRS ACRS-2481 PDR C c .,,o, ., .c ,3 37 [Mr ..

. r it XIII. Executive Sessions ,

A. Sucommi ttee Assignments (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

1. Spent Fuel Storage......................................... 37 B. Reports, Letters, and Memoranda (0 pen).................... 37

1. Proposed BWR Mark I Containment Requirements for Severe Accidents...................................... 37

2. ACRS Report on Proposed Policy Statement on Deferred P1 ants................................................ 37

3. Application of GDC-4 to Component Supports............ 38 .

4. ACRS Coments on the NRC Staff Review of DOE's Final Environmental Assessments of HLW Repository Sites..... 38 9

5. ACRS Action on Proposed Regulatory Guide XXXX, " Format and Contents of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," Final Guide Draft, Dated June 1986........ 38

6. ACRS Action on the Proposed Revision 3 to Regulatory Guide 1.63........................................... 38

7. Policy for Nuclear Power Plant License Renewal....... 38

8. Annual Report on the Reactor Safety Research Program. 38

9. ACRS Coments on the implications of the Accident at Chernobyl, Unit 4.................................... 39

10. Improved LWRs........................................ 39 C. Generic Issues (0 pen).................................... 39

1. Additional Safety Requirement in the Federal Republic of Germany........................................... 39

0. Future Agenda (0 pen)..................................... 39

1. Future Agenda........................................ 39

2. Future Subcommittee Activities....................... 39 E. Election of Officers (C1osed)............................ 39 F. Reappointrent of New Members (Closed)................... 39

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G. Appointment of New Members (Closed)..................... 39 ,

H. Retirement of ACRS Member Emeritus (0 pen)............... 40 I. Waste Management and Disposal (0 pen)..................... 40 J. Wingspread International Conference on Reactor Safety..... 40 Supplement - Nuclear Plant Security Supplement - XIII. Executive Sessions E., F., and G.

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TABLE OF CONTENTS '

4 APPEFOICES TO MINUTES OF THE 320TH ACRS MEETING l DECEMBER 11-13, 19P5 i

f Appendix I. Attendees............................................... A-1 l Appendix II. Future Agenda........................................... A-5 4

j Appendix _III. ACRS Subcommittee Meetings.............................. A-7 i Appendix IV. Chernobyl Evaluation.................................... A-12 l

h Appendix V. Recent Significant Events............................... A-22

Appendix VI. Oconee Loss of Low Pressure Service Water.. . ... ... . .. . . . A-57 1

Appendix VII. Proposed BWR Severe Accident Containment Requirements... A-62 ,

! Appendix VIII. Severe Accident Containment Issues...................... A-78 i

i Appendix IX. Pol icy Sta tement on Deferred Pl ants . . . . . . . . . . . . . . . . . . . . . A-79 1

Appendix X. Proposed Final PTS Regulatory Guide..................... A-89 Appendix XI. NRC Staff Response on Shearon Harris Application........ A-98 l Appendix XII. TVA Organizational Issues Presentation Slides............ A-131 l Appendix XIII. Nuclear Waste Management - Low-level Waste, Alternatives

. to Shallow Land Burial................................... A-139 j Appendix XIV. Additional Documents Provided for ACRS' Use.............. A-170 l

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~g UNITED STATES

!" o NUCLEAR REGULATORY COMMISSION

$ ,E ADVidORY COMMITTEE ON REACTOR SAFEGUARDS ,

0,, WASHINGTON. D. C. 20658 "'

Revision 1: December 11, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 320TH ACRS MEETING DECEMBER 11-13, 1986 WASHINGTON, D. C.

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

1) 8:30 - 8:40 A.M. ReportofACRSChairman(0 pen) 1.1) Opening Statement (DAW) 1.2) Items of current interest (DAW /RFF)

2) 8:40 - 9: A.M.

Election of ACRS Officers for CY 1987 (Closed) 2.1) Nominations and discussion regarding:

. ACRS Chainnan

. ACRS Vice-Chairman

. Member-at-Large, ACRS Planning Subcomittee 2.2) ACRS members vote

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9:50 A.M. Preparation for Meeting with NRC Comissioners (0 pen)

TA -------

3.1) Status of ACRS review of the ectiveness of NRC Staff programs whi ddress generic and unresolved safety sues (CPS /SD)

TAB---------- . Standardized Nuc1 Plants - Briefino of ssioners d discussion regarding'ACRS obsery regarding I"lREG-1225, Imple ation 0 Policy on Sta rdization of Nuc erPlants(see 5 report dated 10/15/86)(C

4) 10:00 - 11:30 A.M. Meeting with NRC Comissioners (Room 1130-H)

(0 pen) 4.1) Briefings / discussion of items noted above 11:30 - 11:45 A.M. BREAK

5) 11:45 - 12:30 P.M. .!mprovedLightWaterReactors(0 pen) 5.1) Discuss proposed ACR5 comments /recomenda TAB 5 -------- tions regarding the characteristics of im-provedlightwaterreactors(00/RKM) l 12:30 - 1:30 P.M. LUNCH

{ 6) 1:30 - 2:45 P.M. ImprovedLightWaterReactors(0 pen) 6.1) Continue discussion of proposed ACRS 4

i 320th ACRS Meeting Agenda . , ,

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cannents/ recommendations regarding the characteristicsofimprovedLWRs(D0/RKM) f 2:45 - 3:00 P.M. BREAK 7)'3:00- 5:00 P.M. Reactor Operations (0 pen / Closed) 4 7.1) Report of ACRS Subcommittee chairman regard-ing recent reactor operating incidents and events (JCE/HA)

! 7.2) Meeting with representatives of NRC Staff l (Note: Portions of this session will be closed as 1 necessary to discuss Proprietary Information

- applicable to the matter being discussed.)

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10) 5:00 - 6:9( P.M. Pressurized Thermal Shock (0 pen) i 10.1) Report of ACR5 Subcommittee regarding TAB 10 ------ proposed NRC Regulatory Guide on i Pressurized Thermal Shock of Reactor

) PressureVessels(PGS/EGI) '

Meeting with representatives of the NRC 10.2)

Staff

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320th ACRS Meeting Agenda -3' -

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

9) 8:30 - 9: A.M. Nuclear Plant Security (0 pen / Closed) 9.1) Briefing / discussion with Robert F. Burnett, TAB 9 ------ Director, Division of Safeguards, NMSS.

regarding security provisions at nuclear facilities (Note: Portions of this session will be closed as required to discuss information regarding detailed security arrangements at nuclear facilities.)

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8) 9:30 - 10:30 A.M. Reactivation of Nuclear Power Plants (0 pen) 8.1) Reporc of ACR5 Subcomittee regarding TAB 8 ------ proposed NRC requirements, etc. regarding reactivation of deferred or terminated nuclear power plants (D0/FJR/RPS) 8.2) Meeting with representatives of the NRC Staff 10:30 - 10:45 A.M. BREAK

11) 10:45 - 12: P.M. Implications of the Chernobyl Nuclear Plant Accident (0 pen)

( 11.1) Report of ACRS Subcommittee regarding im-TAB 11 ----- plications of the Chernobyl accident regarding the safety and regulation of nuclear facilities in the United States (00/FJR/RPS) 11.2) Meeting with representatives of the NRC Staff g) 50 .

12: M - 1: E P.M. LUNCH 50 5s~

12) 1: % - 3:45 P.M. ContainmentPerformance(0 pen) 12.1) Report of ACR5 Subcomittee regarding TAB 12 ------ proposed NRC generic letter regarding the capability of pressure-suppression type containments to contain serious reactor accidents (JCM/MDH) 12.2) Meeting with representatives of the NRC Staff 55" f 3:4# - 4:Or/ P.M. BREAK 5

13) 4:07 - 5:30 P.M. Improved Light Water Reactors (0 pen) 13.1) Continue discussion regarding proposed ACRS coments/ recommendations regarding improvedLWRs(00/RKM) k -

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. 320th ACRS Meeting Agenda - . .

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14) 5:30 - 6:15 P.M. Radwaste Management and Disposal (0 pen)

TAB 14 ----- 11.1) ReportofACR5 Subcommittee (DWM/OSM)

Dec. 4-5, 1986 meeting on radwaste management and disposal regarding:

. Environmental Assessment of high-level radwaste repositories

. Rulemaking to conform 10 CFR Part 60 to EPA Standards

, . Implementation of amendments to the Low-level Radwaste Policy Act of 1985

. Alternatives to shallow land burial of radwaste 14.2) Meetings with representatives of the NRC f Staff, as appropriate 3

NRC Safety Research Program (0 pen) i

15) 6:15 - 6:3{P.M. 15.1) Discuss plans for ACRS 1987 report to the j TAB 15 ------ U.S. Congress regarding the NRC safety

! researchprogram(CPS /SD) l i

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. 320th ACRS Meeting Agenda -

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

15) 8:30 - 8:45 A.M. FutureActivities(0 pen)

TAB ----------16.1) Anticipated ACRS subcommittee activities (MWL)

INSERT HANDOUT IN TAB--------16.2) Proposed items for consideration by the full Committee (DAW /RFF)

17) 8:45 - 12:30 P.M. Preparation of ACRS Reports to NRC (0 pen / Closed) 17.1) Discuss proposed ACR5 reports to NRC (10:15-10:30- regarding:

BREAK) 17.1-1) 8:30-9:30: Improved LWRs (D0/RKM) 17.1-2) 9:30-10:15: Containment Performace(JCM/MDH) 10:15-10:30: BREAK 17.1-3) 10:30-11:15: Implications of theChernobylAccident(00/RPS) 17.1-4) 11:15-11:45: Pressurized Thermal Shock (PGS/EGI) 17.1-5) 11:45-12:15: Reactivation of I

Deferred or Terminated Plants (00/RPS) 17.1-6) 12:15-12:30: Radwaste mgt. and disposal (DWM/OSM)

(Note: Portions of this session will be closed as required to consider Proprietary Infonnation applicabletothematterbeingdiscussed.)

12:30 - 1:30 P.M. LUNCH

18) 1:30 - 2:00 P.M. NRC Regulatory Guides (0 pen) 18.1) Report regarding R.G. 1.63. Electrical TAB 18 ------ Penetration Assemblies in the Containments ofNuclearPowerPlants(CPS /CJW/SD)

19) 2:00 - 3:30 P.M. ACRS Subcomittee Activities (0 pen) 19.1) Reports of ACR5 Subcomittees and Subcomittee Chairmen regarding: -

TAB----------------19.1-1) h Cv-e: Ju: Shearon Harris st.'t4-/(rp~ Nuclear Power Plant - Resolution of ACR5 coments in its OL report.-

M ed 1/16/84 .m fyg_ g,. / , TAB ---------------- 19.1-2) 2:30400: Spent Fuel Storage -

o Report of 11/21/86 Subcomittee

. / meeting regarding storage facili-

' ties for interin storage of spent Nucleai power plant fuel TAB---------------19.1-3) 3:00-3:15: Regional Activities -

320th ACRS Meeting Agenda . 6- - -

Report of 12/2/86 meeting with NRC Region III Office (FJR/PAB) -

TAB ---------------- 19.1-4) E L 3E: TVA Management -

Report of 11/21/86 briefing

, /G ' regarding TVA management problems (CJW/RPS)

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• i7. - Fed:ral Register / Vol. 51. No. 230 / Mondry. Dec:mber 1,'3ses 'l Nedess 9"? h .

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I have determined in accordance with present, may exchange prehmmary _. will approve the'appiacallem& W fhes.

g subsection 10(d) Pub. I 92-463 that it is viewe regarding matters to be -

based upon all the informee- seedebie

- j' necessary to close portions of this -

considered during the balance of the i to it. that theiartensamme af enlisend 1 D rneeting information theas noted of which above to discuss meetmg.

trading privileges puesuant to such *

t release would The %W_mittee will then hear applications are consistent with the represent a clearly unwarranted presentations by and hold discussions - '

maintenance of fair and orderly in&dets invasion of personal privacy [5 U.S.C. with representations of the NPC Staff, and the protectan afinvestore ,[- j 552b(c)(6)). Information that involves its consultants. !DCOR representatives. .

detailed security provisions for nuclear and other interested persons regarding F r the Ccmmission, by itne Division of power plants [5 U.S.C. 552b(c)(1)l and this review. ~ ,

Market Regulation, pursuant to delegated . ,

information that involves Propnetary Further information regarding topia ,suthorny. ,- - . q ; $ .j g, Information [5 U.S.C. 552(c)(4)) to be discussed. whether the meetmg "pth e o,goes, ' .-

u. J, applicable to the facility being has been~ cancelled or rescheduled, the .~-

%y discussed. Chairman's ruling on requests for the Further information regarding topics

. ,j opportunity to present oral statements i

"~- .

to be discussed. whether the meeting and the time allotted therefor can be - suam coot smw .

has been cancelled or rescheduled, the obtained by a prepaid telephone call to '

Chairman's ruling on requests for the the cognaant ACRS staff member. Mr. *

• opportuni'y to present oral etatements Dean Houston (telephone 202/634 3287) Setf. Regulatory Cy, .;nns; '7 ~ 7, and the tiwe allotted can be obtained by between 8:15 am and 5:00 pa Persons Applications for Unfisted Trading ,.

, a prepaid telephone call to the ACRS planning to attend this meeting are Privileges and of Op'portunity for Executive Director, hit. Raymond F. urged to contact the above named Hearing; Cincinnati Stock Exchange, -

Fraley (telephone 202/634-3265), individual one or two days before the Inc. -

' . t between 8.15 a.m. and 5:00 p.m. scheduled meeting to be advised of any 4 #

Da ted. November 25.1988 changes in schedale, etc, which maY November 24. isee. "

have occurred. ~"

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. Ad?is co mit:eeMancsementofficer. Det*d Nov'ab*r 25. *B' e above nand nadonal securities

. - Thomas G. WN exchange has filed applications with the (FR Doc. 86-26942 Fded 11-28-a6 e.45 am)

Securities and Exchange Commissfon

• e u nea coot r m m Assistant Executive Directorfor Techazal Acrivaies. pursuant to sect!on 12(f)(1)(B) of the

[rR Doc. 86 28o43 Filed 11 d-ae; a:45 am) Securities Exchange Act of1934 and

, . Advisory Committee on Reactor s u m.a coce ti m e w Rule 12f-1 thereunder. for unlistcJ y Safeguards Subcommittee on Severe trading privileges in the'fobwing stock:

. Accidents; Meeting '

Coca Cola Enwrpnses, lac.

~

, The ACRS Subcommittee on Severe M O Comnen Stock $1Ao Pur Value (Pile No. 7-Accidents will hold a meeting on edit) ..% . .

December 19.1986. Room 1048.1717 H Self. Regulatory Organizations; Street NW Washington. DC. %fs security is listed and registered Applications for Uniisted Trading on one or more other national securities The entire meeting will be open to Privileges and of Opportunity for Pu ca end ce exchange and is reported in the

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Hearing; Boston Stock Exchange,Inc. consolidated transaction reporting shall be as follows: November 24.1986. system. ,

Fnday. December 19.1936-d:30 a.m. The above named national securities exchange has filed applications with the su or D ines untilthe concluswn ofbusiness Securities and Exchange Commission written data, views and arguments The Subcommittee will discuss the concerning the above-referenced '

NRR Implementation Plan for Severe pursuant to section 12(f)(1)(B) of the Securities Exchange Act of 1934 and applications. Persons desiring to make Acctdent Policy Statement regarding Rule 12f-1 thereunder, for unlisted written comments shoufd file these -

Ir.iividual Plant Examinations (IPE) for trading privileges in the following stocle copies thereof with the Secretary of the Existing Plants.

Coca Cola Enterprises. Inc. Securities and Exchange Commieston.

Oral atatements may be presented by members of the public with the Common Stock $100 Par Value (File No. 7 Washington. DC 20540. FoDowing thi concurrence of the Subcommittee I g gg ,

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Chairman written statements willbe This security is listed and registered 11 'PP ifit fmds accepted and made available to the ,d upon au 'the info tion availabIe on one or more other national securities to it, that the extensions of tmitsted Committee. Recordings will be permitted exchange and is reported in the -

only during those pcrtions of the consolidated transaction reporting trading privileges pursuant to such meeting when a transcript is being kept. system. applications are consistent with the and questions may be asked only by Interested persons are invited to maintenance of fair and orderly markets members of the Subcommittee,its submit on or before December 16.1980 and the protection of trmstors.

consultants. and Staff. Persons desiring written data, views and arguments For the Comnusdoa. by the Dwisies el to make oral statements abould notify concerning the above tcleranced Market Regulation. pursuant to delessaed the ACRS staff member named below as applications. Persons desiring to make authority. .

far in advance as is practicable so that written comments should file three * *

(

• appropriate arrangements can be made. copies thereof with the Secretary of the. Jonathas G. Rats. -

• During the initial portion of the Securities and Exchang* Commission. Secretary. .

y Washington. DC 20549. Following this opportunity for hearing, the Commission ' (Mt Doc. 06-26646 FJ!ed 11-as-ea; t4s am) 9 meedng, the Subcornmittee, along with any ofits consultants who may be su,e coo, ,,% ,

9

43258 .

Feder:1 Regist:r / Vol. 51. Na. 230 / Mondry. December 1.1988 / Notices Deted. November 25.190s-

~,.

&mee G. McCrelese. Portions of this session will be closedDiscuss plans for ACRS 1987 Report to as required to discuss Proprietary Assistant Executive Directorfor Technica/Information applicable to the project the U.S. Congress regardin

Activitses.

Safety Research Program. g the NRC

( lFR Doc. s6-289s1 Filed 11-26 48; e-46 aml coag 7,, , ,

being discussed.

4:15p.m.415p.m:ImprovedLight will discuss prop (osed ACRSA30 Satutday. December 13. nos commentsWoterReactors Open)-

am.445 am.:FutumActivities and recommendations regarding the (Open)-The members will discuss Advisory Committee on Reactor anticipated ACRS subcommitfee Safeguards; Meeting characteristics ofimproved light water reactors. -

activities as appropriate and items In accordance with the pumoses of proposed for consideration by the full Friday, December 12.1980

• Comm t sections 29 and 182b. of the Atomic ,

Energy Act (42 U.S.C.2039,2232b), the AJOa.m.4 JOa m. NuclearPlant -

Advisory Committee on Reactor Security (Open/ Closed)-The members ACRS Reports to NRC(Open/ Closed)-

Safeguards will hold a meeting on w!!! hear a report from the Director. The members will discuss proposed NRC Division of Security. NMSS. an'd ACRS reports to the NRC regarding December 11-13.1988. in Room 1048. '

discuss provisions for security of.

1717 H Street NW Washington. DC. matters considered during this meeting.

Notice of this meeting was published in nuclear power plants.. .

Portions of this session will be closed the Federal Register on November 20, Portions of this session wiu be closed as required to discuss Proprietary 1988, as required to discuss mformation Information ap;i: cable to the matter regarding detailed security being discussed.

%ursday. December 11.1988 ""

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&J0am.440a.m.: Report ofACRS d Activities (Open)-The members will Chairman (Open)-The ACRSChairman will discuss a pr(oposed NRC hear and discuss reports of ACRS will report briefly regarding items of RegulatoThermalShock subcommittees and subcommittee OpenHT I- Guide on Pressurized Thermal Shock of a reg current interest to the Committee. Reactor Pressure Vessels. ' d\ns of speci c d

a 40 am.-R09am.:Flection ofACRS Representatives of the NRC Staff will resolution of ACRS recommendations d q i rd d i f th '

ons a 10. .m.- JO availability of candidates and will select the Chernob IAccident(Open)-The m licationsof heato Harris uclear Ian , proposed Committee 3987 officers for Calendar Year members w Ihear and discuss reports gvis nfR R to C id 1 p

.gfI portion of the meeting will beabout the implications of the Chernobyl nuclear plant accident regarding the Nuclear Power Plants, and storage of clos j to discuss information thesafety releaseand regulation of nuclear power nuclear power plant spent fuel.

of which would represent a clearly plants 8n the United States Procedures for the' conduct of and anwarranted invasion of personal ...

f Jop.m. asp.m.: Containment participation in ACRS meetings were p vacy. published in the Federal Register on

. Performance (Open)-The members will October 20.1986 (51 FR 37241). In hear reports and discuss a proposed

.sma.m.-e50 am.:Preparationfor accordance with these procedures, oral Afeeling with N/IC Commissioners NRC generic letter regarding the performance of dynamic reactor or written statements may be presented (Open)-%e members will discuss containment types to contain severe by members of the public, recordings the status of ACRS activit.es and nuclear power plant accidents, will be permitted only during those observations regarding the effectiveness' J:15p.m.-5:15p.m.:Im portions of the meeting when a of NRC Staff programs which address Water Reactors (Open) provedt.ight transcript is being kept, and questions The members may be asked oniyby members of the generic and unresolved safety issues. will continue discussion of proposed and NUREG-1225. Implementation of Committee, its consultants, and staff.

NRC Policy on Standardizauon of ACRS comments and recommendations Persons desiring to make ors! -

regarding the characteristics of Nuclear Power Plants. improved Ifght water reactors. statements should notify the ACRS 10ma.m.-!!:Jo am.: Ateeting with 5:15p.m.-dWp.m.: Radioactive Executive Director as far in advance as NRC Commissioners (OpenHue WasteStanagementandDisposal practicable so that appropriate {

(Open)-ne members will hear and anangements can be made to allow the i members will address and discuss  ?

th topics noted above with the NRC discuss the report ofits subcommittee necessary time during the meeting for Commissioners.- regarding items related to the such statements. Use of still, motion 11:45 am.-l.mp.m/ Reactivation of , management and disposal of radioactive this meeting picture and totelevision may be limited selected cameras dur NuclearPower Plants (Open)--The wastes including the NRC Staff review members will hear reports and discuss portions of the rneeting as determined of the environmental assessment of proposed NRC requirements regarding waste disposal sites nominated by the by the Chairman. Information regarding l the reactivation of deferred or DOP. rulemaking to conform to CFR the time to be set aside for this purpose terminated nuclear power planta.. Part 60 to the EPA Standard for high. may be obtained by a prepaid telephone

. 2Wp.m4Wp.m. Reactor , .. level waste repositories, implementation. call to the ACRS Executive Director. R.

Opemtions (Open/ClosedFue ,, a ,, F. Fraley, prior to the meeting. In view of of the law level Radioactive Weste the possibility that the schedule for .

members will haar reports of and '2, Policy Amendments Act of1985.and

., ACRS meetings may be adjusted by the discuss recent transienta and incidents alternatives to shallow land burial of Chairman as necessary to facilitate the which have occurred at nuclear radioactive westes. Representatives of the NRC Staff willparticipate as conduct of the meeting. persons .

facilities. Representatives of the NRC l

. planning to attend should check with the Staff will participate in this sesolon to . appropriate. .

. l omp.m.430p.m.: ACRS ACRS Executive Director if such gs degree considered approp'riate . .

i s g { Subcommittee Activities (Open)-- . .

. rescheduling would result in major Inconvemence. -

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MINUTES OF THE j T' -

320TH ACRS MEETING l

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DECEMBER 11-13, 1986 The 320th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H. Street, N.W., Washington, D.C., was convened by Chairman D. A. Ward at 8:30 a.m., Thursday, December 11, 1986.

[ Note: For a list of attendees, see Appendix 1.

G. A. Reed did not att'end the meeting.]

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 Reports. Inc., 444 North Capitol Street, Washington, D.C. 20001.]

I. Chairman's Peport (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

Chairman D. A. Ward reported that two versions of the Wingspread meeting

'sumary have been distributed to members, a short version which will become a public document and a proprietary, long version to be issued as the

, official record of the meeting. He requested that members comment on both drafts as soon as possible so that the two versions can be released.

Chairman Ward indicated that a formal written response was received by the ACRS from the General Counsel's office regarding the issue of beckfitting.

It essentially states that recuests by the Staff of a licensee for a systems interactions study do not fall Under the backfit rule.

C. Michelson proposed the case where a licensee wishes to make changes to its nuclear plant and the NRC Staff requests additional changes to maintain the current level of safety. D. A. Ward indicated that this would also not fall under the backfit rule.

Chairman Ward indicated a response from Chairman Zech regarding the move of the ACRS to Bethesda. It states that, since the option on the secord White ,

Flint building has been signed, the Connission wishes to proceed as orig-inally planned. This will mean an August-September 1987 move of the ACRS to the Phillips Building in Bethesda, Maryland.

Chairman Ward indicated that the Comission has approved the restart of the Davis-Besse Nuclear Power Station, Unit 1. The licensee is currently .

perforning hot functional testing. He mentioned the rupture of the 18-inch l feedwater pump suction line at the Surry Powir Station, Unit 2, where one i worker died of burns. He noted that arrangements have been made for -

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320TH ACRS MINUTES 2 -

P. G. Shewmon to accompany the NRC investigation team'to the site. He also -

noted that the rupture was a 360-degree failure. Chairman Ward mentioned receiving feedback from the Commissioners on the selection of a new member for the ACRS. He mentioned a retirement dinner for H. Etherington.

J. C. Mark requested that H. Etherington attend some ACRS meetings when his expertise will be of importance to the subject being discussed.

II. Implications of the Chernobyl Nuclear Plant Accident (0 pen) l

[ Note: R. P. Savio was the Designated Federal Official for this portion of themeeting.]

0. Okrent indicated that a meeting of the Safety Technology, Philosophy, and Criteria Subcommittee was held on November 5, 1986 and on December 10, 1986 to discuss the implications of the Chernobyl accident. He presented a draft letter as one possible approach the Committee might take in respond-ing to the formal request for coment from the Comissioners.

F. J. Remick, Subcommittee member, indicated that the urgency of the ACRS letter has been deferred because the Commission is waiting for the Staff's Chernobyl fact-finding report, which will probably not be available to the Commission this month. B. Sheron, NRC, discussed the status of the fact-finding report. He explained that a draft of the report was prepared prior to the IAEA meeting in Vienna to aid the U.S. delegation. The final factual report will reference and draw from the substantial data provided by the Soviets, as well as b Safety Advisory Group (INSAG)y whichreports done advises by the International the Director Fuclear General of IAEA.

No new information has been received by the Staff since the Vienna meeting (see Appendix IV). He explained that NRC's responsibility regardina this nine-chapter document will be the introduction to the remrt, the suunary, the chapter on accident scenarios, and the chapter on source term and atmospheric dispersion and transporc. A meeting was held to review the status of the report on November 18, 1986. There were a number of cnn-cerns, primarily from DOE, about the FEMA / EPA chapters. It was agreed that DOE should meet with FEMA and EPA to resolve their concerns. A meeting to finalize the report is scheduled for December 19, 1986. A Comission meeting on the subject "Chernobyl Implications Assessment" is scheduled for the week of January 19, 1987. The Staff would need an ACRS letter by January 16.

B. Sheron explained that the approach of the implications assessment report was to identify candidate issues that derive directly from the Chernobyl accident. The Staff intends to assess the candidate issues by identifying current regulatory practico followed by the !!RC and determine whether this regulatory practice is edequate or needs to be strenothened in some arcas.

B. Sheron presented e list of the issues in the report and pointed out changes made since the Comittee last saw the list. He indicated that the reactivity accident section was rewritten, based en new information, to provide more clarity and background as to current NRC practices. The accident management section now discusses operator procedures and their relationship to severe accidents. The auxilfary feedwater system is now considered an engineered safety feature. On multiple unit protection, the Staff was corcerned regarding a radiation hazard if the operators had to leave the control room of an undamaged unit ih order to shut it down. The Staff concluded that there was not a hazard associated with shutting down 1

320TH ACRS HINUTES 3 -

O an undamaged unit and keeping it shut down. The emergency planning section -

was rewritten to be consistent with coments received and is more consis- '

tent with the Staff presentation to the Comittee in November 1986. The Staff has added a new section on graphite-moderated reactors which is .

primarily an assessment of how Chernobyl affects the Fort St. Vrain nuclear generating station.

B. Sheron indicated that the Staff's overall conclusions have not changed since November. The Staff's assessment shows that there is nothing that would prompt the Staff or the Comission to take imediate regulatory

action by means of an order, bulletin, or 50.54(f) letter. The Staff I

concludes that U.S. nuclear power plants are protected against Cherrebyl-like events primarily by the nature of their nuclear design. The

! Staff will take the Chernobyl experience into account in reinforcing some of the aspects of requirements already existing or being developed. There are some areas where the Staff believes some further study or research is probably warranted. Research in areas such as reactivity accidents at low power or shutdown will form the foundation for subsequent consideration of action. The Staff will need to take into account all of the events that

~

occurred at Chernobyl as a general background to make sure that they are considered when the Staff formulates rules, regulations, or makes decisions on sa fe ty. The Staff has specific conclusions regarding the area of operations:

Administrative controls in the regulatory process in the U.S. are generally adequate to assure a safe operating envelope for the plants.

The Staff must reinforce attention to operations factors and manageri-al matters such as consideration of a high-level onsite nuclear safety manager, an individual who would have no other duties.

B. Sheron noted that the shift technical advisor performs some of the functions of an onsite safety ranager but his role is really an advisor to the operators. C. P. Siess noted that the shift technical advisor only functions after an accident. S. Wright, NRC, agreed that the shift techni-cal advisor cannot fulfill all of the functions of this high-level safety 4

individual that the Staff is considerino. P. G. Shewmon noted that there is an onsite safety review committee which should serve this function.

S. Wright indicated that all of the individuals on this comittee have competing duties. Some have production activities and others are associ-ated with naintenance. M. W. Carbon asked if the Staff's proposal is

! similar to the Japanese concept of a safety officer. S. Wright indicated i that he was not familiar with the Japanese concept.

J. C. Ebersole suggested that there is a need for a "redbook " a detailed listing of restrictions that prohibit operators from defeating automatic functions. He speculated that such a book might have prevented Chernobyl.

O. Sheron noted that there are technical specifications which prohibit taking certain systers out of service. J. C. Ebersole thought that was a very low key approach. R. Sheron agreed that' the "redbook" idea might be useful to consider as a longtino item.

B. Sheron indicated that in the area of desfgn the Staff noted that U.S.

nuclear plants do not have the positive void coefficient that the Chernobyl i

320TH ACRS MINUTES 4 reactor had. There are some positive moderator coefficients early , the -

fuel cycle for some U.S. nuclear plants, but once one starts burning up the fuel these coefficients go negative. The Staff should review the ri.s from vulnerable sequences using more sophisticated tools for reconfirmatio.

(PRA and deterministic). J. C. Ebersole asked if this will mean a revisit-ing of the BWR ATWS. B. Sheran indicated that the Staff is not considering a complete reexamination of the ATWS. J. C. Ebersole indicated that he was talkino only about the particular BWR ATWS where a reactivity excursion occurs on the turbine trip. J. C. Mark indicated that he was particularly impressed with the graphite rod followers on the control rods which provide some positive reactivity insertion before the negative effect of the insertion of the control rods.

B. Sheron indicated that, when the Staff was investigating accidents at low power and shutdown, it found a case where technical specifications being proposed for Wolf Creek would have resulted in consequences that were in excess of the original design basis. This was during the analysis of a rod withdrawal accident at startup, which is the worst case. D. A. Ward agreed that technical specifications can be violated. He speculated that at Chernobyl several of their equivalent " technical specifications" were violated. He asked how the NRC can have such high confidence that techni-cal specifications would never be violated in U.S. plants. B. Sheron admitted that no one can absolutely guarantee that an operator will never violate a technical specification. One cannot regulate the attitudes of people; what the Staff can do is attempt to provide the atmosphere and the incentives to make sure that the cperators are sensitized to recognize the risks of violating their technical specifications. D. Okrent referred to l J. C. Ebersole's question on the BWR ATWS reactivity accident. He wondered i

if the Staff had investipated a set of circumstances that might make the first portion of that ATWS event significantly worse. A case in point might be if the water came into the core colder than expected. With the same reactivity there would be more voids to collapse, causing greater reactivity insertion in the first phase when one gets a main steam iso-lation valve closure. J. C. Ebersole added a coincident turbine trip with loss of feedwater. B. Sheron indicated that the Staff has told the BWR owners that they are not allowed to operate with an inlet temperature below some number to compensate for the loss of feedwater heating.

B. Sheron indicated that with regard to multiple unit protection the Staff has examined a number of action items deriving from the TMI accident regarding control room habitability under severe accident conditions.

F. J. Renick asked, if an accident occurred at a multtunit site in the U.S.

with one unit involved, would the other units on the site be habitable?

B. Sheron's conclusion was that the control rooms would have a hich confi-dence of being habitable, althcugh they were not spccifically designed for a 10 CFR 100 release. L. Sofer, NRC, indicated that an SRP revision the Staff is proposing will address the use of BWR suppression pools as part of the cleanup systems. The Staff does not expect to address the question of control room habitability under severe acciderts until the spring of 1987.

B. Sheron noted that the Staff has concluded that there should not be sharinq of systems on a multiple unit site that are part of a shutdown capah'lity to prevent accidents on one unit affecting a second unit.

J. C. Ebersole wondered whether that was an iniplied cendemnation of sharing in general, which is a design to save the installition of extra systens in

320TH ACRS MINUTES 5 1

l a unit. B. Sheron qualified his previous statement by indicating that the -

plant operators ought to be sure that an accident at one unit will not take out the second unit. In severe accidents there are obvious benefits to sharing. W. Kerr suggested that sharing is not the problem one should be ,

discussing; it is inadequate equipment. The Comittee discussed the

possibility of uncontrolled interconnection between the control room j habitability system in an undamaged unit and an eouivalent habitability 1 system in the control room of a damaged unit. B. Sheron indicated that the i

Staff intends to make sure that the operators in the control room of the undamaged unit are protected from the damaged unit. He speculated that it

. was not likely that there would be a harsh environment to prevent the i , operators in the undamaged unit from taking manual actions in the unaffect-i ed unit. B. Sheron indicated that a number of things have been done by the

! Staff regarding the possibility of firefighters being asked to fight fires in the presence of radiation. The Staff intends to provide adequate provi-i sions at nuclear plants to cope with fires in a radiation environment Including access to proper protective equipment such as self-contained breathing equipment..

t B. Sheron indicated that the Staff now believes that if there had been a

! large dry containment around Chernobyl it might have withstood the explo-

! sive force and prevented a radioactive release. This fact has just reem-

! phasized the need for the Staff to look at containment perfonnance in its I evaluation of severe accidents. The Staff will be looking at filtered

venting of some plant containments as a mitigation strategy to perhaps reduce vulnerabilities. The Staff examined what the Soviets did in the way

of emergency plans following Chernobyl. He noted that they distributed KI

( ' the population. It was revealed at Vienna that the Soviets quickly l as essed their civil defense plan and found it to be inadequate since it was not specific enough to be considered an evacuation plan- for that j nucle

• plant. He also took note of the fact that the Chernobyl accident j produce.' a rather unique source term because of the burning of about 25 i percent f the graphite moderator in the core and resultant heat which i lofted ma ' rials into the plume, which went into the upper atmosphere.

! J. C. Ebers 1e wondered if the accident would have been worse if they had had an explo ion but no fire. He noted that the fire was a mechanical l

mechanism to i #t the plume; if there had not been a fire the radioactive

! material might . $ve settled around the plant producing a worse accident.

B. Sheron indicat. i that, after examining what happened at Chernobyl, the Staff has concludec that the 10-mile plume emergency planning zone (EPZ) in

! the U.S. continues o be adequate as well as' the 50-mile EPZ for the ingestion pathway. h noted that there are a number of efforts ongoing regarding EPZs and the e adequacy, and the possibility of charging them I based on new source tern, information. There is ongoing work in this area i in the Office of Researc and some new regulations are being drafted.

NUREG-1150 will also facto. into this subject, and any' conclusions will need to take account of the L 1rnobyl accident. "

D. Okrent noted that the Staff h 11, of course, have to take account of the fact that the dispersion of the .'dioactive cloud would have been quite a bit different and much worse if he short-lived radionuclides had been 1 released at ground level. D. W. A $11er pointed out that the Chernobyl -

i accident really shewed that the rov ment of the contamiration from the j radioactive material could extend over ide areas.

• This points to the need I

i '

1 l

_ ~ .

j.  !

I 320TH ACRS MINUTES 6 1  ; c, ,

. i i for ' compatibility -in the protective action guides ' (PAGs) for adjacent --

4-countries. He thought one of the lessons to be learned was that the NRC ought to make a comparison of the PAGs in Canada and the U.S. He suggested 1 that the. confusion that. could be. caused by different PAGs for Canada 'and i the U.S. (in the : event of an accident along the Canadian border where  !

imaterial moved between the two ' countries) could quite likely lead rapidly l' to chaos. Another problem .is the- State-controlled distribution of KI. The ,

fact that each individual State separately makes a decision whether to . '

administer KI and on what action level to administer KI also invites chaos. 1 One interesting fact from Chernobyl was that ingestion did not prove to be  ;

- pf any consequence as far as acute effects are concerned, but the accident i clearly pointed put that it- is a rather important consideration. He also l

[ noted that the military played a major role in essisting the evacuation and other protective measures at Chernobyl. - He wondered whether the contingen- '

cy of the use of the U.S. military would be useful and he suogested incor- ,

poration of D0D into emergency preparedness for nuclear facilities. S.  ;

! Schwartz, NRC, thought that comparisons of PAGs between different countries i did not really fit into this post-Chernobyl effort. It will take Federal

! effort and high-level government discussions to propose the use of the same

! PAGs. The Federal government. does have a policy on KI and it is to allow.

2 the States to make their own decision with respect to stockpiling and predistribution of KI. The Federal government did a cost-benefit analysis ,

on stockpiling of KI and the decision was not to make it a Federal require-

, ment and to leave the decision at the State level. D. W. Moeller asked if

! the Staff has a table showing each State's action level for distribution of f KI. He thought a table of KI policies would be useful. S. Schwartz agreed that the whole question of distribution, stockpiling, and shelf life with

regard to KI is a concern. D. W. Moeller suggested that if the NRC chooses not to study this cuestion it might be worthwhile asking the National Council on Radiation Protection (NRC provides supporting funds) to indepen-  !

j dently assess, evaluate, and make some recommendations in this area. S. '

Schwartz thought it an excellent svogestion. F. J. Remick pointed out l l that, while the NRC does not have a policy requiring the stockpiling of KI  ;

for the public, the NRC does have a policy as far as stockpiling and i

{ distribution of KI for nuclear plant workers. S. Schwartz admitted that i this was so. He also noted that the need for military transport in the r

] event of a nuclear accident is tied in through the Federal Radiological Emergency Planning Program. Military transportation and personnel are ,

available at the Federal level.  !

B. Sheron explained that Staff has recognized the unfoueness of the Chernobyl accident in the fact that it was not a meltdown but the addition

, of enough energy' deposition in the fuel to get mechanical dispersion. The t

Staff recognizes that there are other ways besides a reactivity excursion, j such ' as' steam explosions, for the occurrence of rechanical dispersion

mechanisms for light water reactors. The Staff also notes that there was t l chemical strippin of fission products from the fuel particle surface (via j j

oxidation to U38*

0 I

B. Sheron noted that there are a number of gr'aphite moderated reactors in i the U.S. either in operation or contemplated. The Fort St. Vrain HTGR in

~

Colorado is the only operating cornrercial graphite noderated gas-enoled reactor in the U.S. He indicated that it is the yespersibility of DOE to evaluate its N-Reactor in Hanford, Washington. He noted that DOE is

320TH ACRS MINUTES 7 l

sponsoring research programs on graphite reactors that might be applicable - -

, to DOE reactors. The Staff is attempting to use the data from that work as  ;

best it can, but the Staff has not specifically looked at any implications '

( ,

of Chernobyl for the DOE reactors. He indicated that the Staff can find no ,

I direct association at Fort St. Vrain with design weaknesses that contribut-ed to the Chernobyl accident. He noted that Fort St. Vrain uses inert i

helium as a coolant, and does not have an oxide core--but a ceramic core 4

with the fuel particles encased in silicon carbide. Silicon carbide is one

! of the major barriers to radioactive release, and there are also pyrolitic

!- graphite layers. Since Fort St. Vrain does not use water, the emission of

) ' '

steam is not a factor. D. Okrent noted that Fort St. Vrain has a steam

generator and a gross " rupture of the s, team generator could result in water i ingress into the core, pressurization, as well as hydrogen generation. B.

Sheron indicated that the Staff fully reccgnizes that reactivity can be l added by virtue of water. ingress under these circumstances, but he noted L that Fort St. Vrain is an under-moderated reactor and Chernobyl was an 1 over-moderated reactor. D. Okrent noted that Fort St. Vrain happens not to have a containment. 'B. Sheron stated that that was a subject of debate some years ago. Fort St. Vrain has a pressurized concrete vessel which some consider to perfom the containment function. There is a steel liner inside the vessel and access to the vessel itself through double penetra-tion closures which are constantly monitored. The Staff looked at the potential for fire in the core which would necessitate the failure of both l! of the penetrations so that air could enter and circulate from the bottom to the top of the core. In the unlikely failure of both penetrations, the Fort St. Vrain plan operators indicate that they could flood the lower  :

compartment to sea" .f the lower penetration and stop any air from enter-

ing the core. This would starve any fire of air. The Staff does not see i any particular concerns regarding severe accident phenomena for Fort St. .

j Vrain, but they do see the desirability of developing a PRA for Fort St.  !

. Vrain. Such a PRA does not now exist. The Staff is also interested in i j graphite themal stress experiments which would be helpful in understanding  :

j the performance of the graphite blocks under the high thermal gradients in L these cores.

I I

F. J. Remick indicated that several questions came up during the Subcommit-tee meeting. One raised by H. W. Lewis involved whether surveillance tests

. are perhaps conducted too frequently in the U.S. He wondered if the Staff  ;

1 has considered that. F. J. Remick asked whether the Staff has considered plant personnel beyond licensed operators in the event of an accident. i B. Sheron indicated that these are two areas that the Staff would certainly want to consider.

J. C. Mark pointed out significant differences between graphite reactors in the U.S. and the USSR. He noted that a crucial difference between the Chernobyl reactor and the Hanford N-Reactor is the fact that if water is voided from the N-Reactor the activity decreases, whereas at Chernobyl

. there is a substantial reactivity insertion. H. W. Lewis thought that the lessons of Chernobyl are really concerned with management and not hardware.

D. Okrent asked members to submit additional' points for inclusion ina 3 possible letter on Chernobyl. F. J. Remick indicated that his only concent ,

for a letter would be for the ACRS to sharpen what the Staff has already done, but avoid introducing new areas for consideration at this stage.  !

.-~~,,y ,-..---. - - , - - -,,--.-n , ,- nm,. , -. ..,,m. - , - - - , - _ . ,, ,,,w-,--. - - , , . , - - - - - ~ . -e,-v ,,,,-.--,--,,.e,-

e O 320TH ACRS MINUTES 8 l

1 III. Reactor Operations (0 pen) '

[ Note: H. Alderman was the Designated Federal Official for this portion of themeeting.)

A. Drywell Shell Corrosion at Ovster Creek J. Donohew, NRC Project Manager for Oyster Creek, explained that the Staff became aware of the wastage of the drywell shell just below the downcomers on November 20, 1986. The Staff was concerned that the corrosinn of the shell would cause ,the thickness to .be less than ndtessary 'for the drywell to be sound for structural 1 pads, including the design basis accident. The integrity of the containment was at issue. He explained that the wastage was below the downcomers in two bays where water leaked through a bellows seal between the outside shell and the biological shield when the refueling cavity was flooded for refueling. l.eakage collected in the drainage channel below the downcomers. The licensee theorizes that the cause of the corrosion was moisture entrained in sand which is present in a small cavity around the circumference at the floor of the drywell (see Appendix V).

The Comittee discussed the general arrangement of the drywell shell.

J. Donohew indicated that the licensee has made measurements in the bays beneath the 10 downcomers and found significant wastage in 2-3 bays. Two bays have minor wastage and five bays are not involved.

The licensee intends to determine an average thickness and what additional corrosion will be allowed. The licensee would then come back to the Staff to indicate whether he will restart without patching the shell. P. G. Shewmon expressed interest in the chemical elec-trolysis aspects of the problem. J. Donohew indicated that the wastage has only been found where the sand is located. C. Michelson suggested that the bonding of insulation to the metal might be a factor. If there is a good band and it has not cracked it might protect the metal. E. Jordan noted that there is a red lead coating where the sand is located. J. Donohew indicated that there is red -

lead in about half of the area all the way around but that the red lead paint was not a major factor.

J. Donohew indicated that this problem has generic implications for Park I containments and the Staff has surveyed Mark I containment shell corrosion at the Dresden-2 and -3 units, Monticello, Nine Mile Point-1, Pilgrim, and Quad Cities-1 and -2. It was found that only the Oyster Creek design has water draining through a sand cavity. The sand cavity appears to be the cause of the problem of wastage of the drywell shell. The sand was wet and damp for probably the entire operating lifetime of the Oyster Creek plant. The licensee said that there were large amounts of water that went through that area during construction. D. Okrent expressed interest in a report by the Staff regarding broader implications on containment integrity associated with this drywell shell corrosion problem. E. Jordan promised a report on Oyster Creek in tuo months regarding other actions the Staff may be taking. ,

l 320TH ACRS MINUTES .9 B. L_oss of All Component Coolino Water at Byron-2 '

R. Woodruff, IE, indicated that all component cooling water was lost for twelve minutes on November 20,.1986 at the Byron Station, Unit 2, ,

during preparation for initial criticality. In .the course of this event, the licensee discovered that a single failure could cause loss of all component cooling water and thereby jeopardize for a short

' period of time the residual heat removal system. The event was caused.

by the lifting of a safety valve on the component cooling water side'

of an excess letdown heat exchanger. The safety valve stuck open.

The component cooling water pump continued to pump water out of the system into the floor dr'ain until _it tripped on low water level in the

~

surge tank. A second pump started automatically and pumped until it i- also stopped on low level. There was no jeopardy to the system since the plant had not operated. The licensee manually isolated that portion of the system which was responsible for the leak and refilled '

, the component cooling water system.

R. Woodruff noted that there are at least six known paths, two for the excess letdown heat exchangers and four for the pumps, for a single failure to drain component cooling water to the floor drains. Since there is but a single pipe in the portion of the system that failed, there would be loss of component cooling water which would require manual action by the operator. The Staff intends to issue an informa-tion notice on this topic and has discussed the matter with NRR. NRR 3

is reviewing the conformance of the system to regulatory recuirements.

j J. Rosenthal, IE, pointed out that this plant was constructed by

.Sargent & Lundy and Sargent & Lundy plants typically have sharing of component cooling water systems and rather intricate piping arrange-

ments.. There are five pumps in parallel to. multiple headers, a number i of crossties, and many isolation valves. By proper closing of the.

!' isolation valves, the system looks like a two-train system with the isolation valves not open and two pumps running. There would be two pumps on standby plus a swing-in standby pump. J. C. Ebersole raised ouestions about the effect on the reactor coolant pump seals during

blowdown. He asked what would happen if the- seals start to leak.

i J. Rosenthal indicated that the seal cooling comes from a ' charging system and from a component cooling vater system. The charging system in turn depends on the component cooling water system, such that given a failure of the component cooling water system you fail the two barrier coolers. J. C. Ebersole thought it clearly a deficient

design. He asked if a fix is contemplated, and requested that_ there be follow-up on this matter.

C. High Pressure Injection Nozzle External Surface Damage Due to Roric -

Acid Corrosion at ANO-1 H. Bailey, IE, indicated that the Staff became aware of high pressure

injection nozzle extensive wastage due to boric acid corrosion at i ANO-1 on October 22, 1986. The plant was' in cold shutdown performing surveillance of the high pressure injection nczzle thennal sleeves.

Hastage was revealed when the insulation was removed from the "A" nozzle. The mechanism of leakage was apparently coolant leaking out of an isolation valve from the bonnet under' the metallic insulation-down the stainless steel line. The valve leakage kept the nozzle i

l 320TH ACRS MINUTES 10 l wetted and that resulted in accelerated corrosion. He indicated that -

the Staff has seen a lot of boric-acid-induced wastage over the years.

Most of it, up to this point, has been on threaded fasteners. The threaded fastener, which is made of a low alloy carbon steel, is susceptible to boric acid corrosion. In 1982 IE issued a bulletin

• which discussed boric acid corrosion on threaded fasteners. In 1983 the ASME Code,Section XI, was revised to require more extensive examination of systems containing boric acid, as a result of boric acid corrosion. P. G. Shewmon asked what sort of repair is contem-plated by the licensee. H. Bailey indicated that the licensee has

. ground out the corrosion and rebuilt the section- by rewelding. They have verified the effectiven'ess of the weld through UT examination but have not heat-treated it. -

D. Okrent expressed interest in the generic implications of this problem for the future. E. Jordan indicated that this is a case where the Staff issued an information notice to advise industry and then used code changes to effect other actions later. D. Okrent expressed skepticism that an information notice advising industry of the issue will be dealt with at all plants in the proper manner. D. Okrent asked if the backfit rule in any way inhibits the Staff frem going beyond an information notice. E. Jordan indicated that the backfit rule certainly causes the Staff to consider carefully the costs and benefits of the need for a new regulatory requirement or a specific requirement of licensees. J. Rosenthal added that IE bulletins and/or generic letters are backed up by regulatory analysis and cost-benefit studies. Information notices are issued based on the Staff's judgment of the generic significance of an event. D. Okrent asked whether the CRGR is involved in examining the effectiveness of the results of the issuance of information notices. E. Jordan indicated the CRGR is not required to examine information notices. CRGR does not nonnally review them, but does receive copies. D. Okrent took notice of the fact that neither NRR nor IE has made a decision about whether to issue an IE bulletin regarding the Oyster Creek, Mark I containment problem. He expressed concern that the Staff may be afraid to close the loop because of the chilling effect of cost-benefit requirements associated with the new backfit rule. He thought this might, in effect, be degrading safety. E. Jordan mentioned an 0I review of this issue, including interviews with members of the NRC Staff as well as CRGR members about a year ago. He offered to provide the report that was issued to the Comittee.

D. Design Problems in Plants Doerating and Under Corstruction J. Rosenthal noted the fact that many of the events being reported to the Committee are the result of design deficiencies, many in older plants. As a result, the Staff has been collecting event reports since August 1986 to form a design deficiency data base. About 50 design problems and 4 possible design problems were identified from 50.75 and regional daily reports. Many of the design problems are of an electrical nature. This is not surprising since a large percentage of the parts in a nuclear plant will be electrical. It is also not surprising that most of the design problems occur at plants construct-ed by the Bechte'. Corporation since Bechtel tras the greatest share of the business. It was surprising that Sargent & Lundy, which has very l

l

320TH ACRS MINUTES 11 l

4 few plants, appears to be involved with many design problems. Even- --

l tually this data base will become a supplement to the performance I indicator program, and eventually will become part of the corporate ,

data network. He noted . that some of the oldest nuclear plants will '

, continue to define what the Staff calls design problems, which account for. about five percent of all licensee event reports.

E. l.oss of Offsite Power at Pilgrim J. Rosenthal indicated that the Pilgrim Nuclear Power Station sus-tained a loss of offsite power on November 19, 1986, when all three -

offsite power lines were lost during a snowstorm. The failure was probably caused by bad weather in the area which consisted of a-combination of high winds, snow, and salt buildup. The plant is located on Cape Cod near the ocean and has had recurring salt-spray problems in the past. There has been salt-spray buildup on switchyard insulators which encouraged the licensee to install a demineralized water spray system to spray down the insulators. However, at this time they do not have a routine preventive maintenance program for spraying the insulators. J. Rosenthal discussed the actual event which resulted in the loss of all three offsite power lines for three hours. Note was taken of the fact that both diesels were parallel to the switchyard network. When power was restored to the switchyard, the diesels were run up to load using the grid effectively as a big resis tor. This is a common practice. J. C. Ebersole thought the use of an unstable grid to load both diesels a potential generic problem.

The diesels do not have the time to protect themselves against a loss of grid, since the breakers are not fast enough. _ He thought the practice of having the two diesels tied to the grid faulty.

J. Rosenthal indicated that General Electric is doing studies for Pilgrim in association with Unresolved Safety Issue (USI) A-44, and another study associated with severe accident considerations. Al-

though there is no regulatory requirement at this time, the licensee is considering adding a third diesel.

F. Failure of Main Feedwater Pipe at Surry-2 1

,. J. Rosenthai explained that an Augmented Inspection Team (AIT) has been dispatched to Surry to investigate the recent December 9,1986 event. Preliminary information suggests an 18-inch main feedwater pipe break on the suction side of the feedwater pump "A" which failed catastrophically. The Staff now attributes that break to a com-bination of wall-thinning on that pipe and a system pressure tran-sient. Eight workers were injured and two have died so far.

J. Rosenthal indicated that the Surry-2 plant was operatin 100 percent power with the main steam isolation valve (MSIV) en g atsteam generator "C" closed. The slamming shut of this MSIV appears to be the initiator of the event, effectively causing a void distribution.

Redistribution in the steam generator, which appears through the steam generator level instruments as low level or Inw-low level, tripped the reactor. All of the control rods entered the core. J. Rosenthal j described the flow in the feedwater system. A pressure transient in ~

l the main feedwater system resulted in the break of the suction piping.

i i

l-

320TH ACRS MINUTES 12 The 360-degree circumferential pipe break occurred in the vicinity of ..

the weld near a tee or an elbow. It was apparently the pipe that broke rather than the weld or the fitting. That pipe is upstream of an isolation valve and the result was loss of both feedwater pumps.

P. G. Shewmon asked if there is anything that would have caused cavitation.~ J. Rosenthal stated that the isolation valve is on the suction side of the pump and, with the changed cross section of pipe in that area, there would be local pressure that could be causing cavitation and erosion. The pressure pulse can be induced by closing of the MSIV or the turbine stop valves. At that point there are two check valves between the steam generator and the feedwater pumps.

. J. C. Ebersole indicated that this was the first instance in his experience where check valves had to function against a catastrophic pressurization upstream of them. This is the most violent manuever that they could have seen. J. Rosenthal indicated that the AIT will be examining the check valves on the affected steam generators, as well as check valves on the unaffected steam generators, in an attempt to understand what the check valve did. Those check valves were certainly forced to shut. The Staff is extremely interested. in gaining a better understanding of the operations that occurred in the 1

transient. .

J. Rosenthal indicated that the Staff now attributes the cause of the piping failure to erosion of the pipe wail and a pressure transient which overloaded the thinned wall section. He did not know whether the thinning or the pressure transient was the larger contributor to 4 the event, and he was not sure which came first. He did note that the MSIV closure was the initial event, since the reactor and the feedwater pump tripped. J. C. Ebersole agreed that the largest loed would come from the main steam feedwater valve closure. J. Rosenthal agreed but noted that the Staff first thought that the feedwater regulator valve was a contributor. The AIT will look into this matter since Surry has had a history in the past of problems with it.

D. Okrent asked if the Staff knows why the MSIV closed. Mr. Rosenthal noted that earlier in the day Surry was having problems with the air pressure that provides the motive force to those valves. The air supply pressure to those valves, which was supposed to be 90 psi, was low at 75 psi.

J. Rosenthal indicated that a 6-12 ft. section of the pipe broke and deflected about 60 degrees, coming to rest against other feedwater suction piping, a case of pipe whip. A piece of the pipe about 2-3 ft. in size separated and came to rest on a cable tray. That piece of pipe has been marked and will receive further examination. He ex-

. plained that the pipe is nominally 1-in. wall thickness. Measurements i

about the middle of that 2-3 ft. section of metal showed a wall thickness of about 1-in., with a 61-mil measurement at the edge.

J. C. Ebersole suggested that this sounds like a mixture 'of wall thinning and water hammer. C. Michelson- noted, however, that it is important to recognize that there was a considerable amount of pipe between the regulating valve and where the break occurred, and there should have been a lot of damping between where the break occurred and

.- - - . , = - -

i 320TH ACRS MINUTES 13 where the. valves snapped shut. Considerable energy absorption must -

have occurred.

, C. Michelson asked why the Staff assigned an AIT to this incident instead of an Incident Investigation Team (IIT). E. Jordan indicated

• that in terms of plant safety the plant did not shut down properly.

This is a balance-of-plant issue and not thought to be safety signifi-cant at this time. He noted that the Staff could still upgrade its response to an IIT if wanranted. D. Okrent asked if the Staff can confirm whether the broad-scope leak-before-break rule would have ruled out this incident. E. Jordan indicated that that should be in the Staff's discussion two months from now. E. Jordan also mentioned that carbon steel pipe is not normally pipe subject to the leak-before-break concept. He noted that the Pipe Study Group is going to the site.

J. Rosenthal explained that heat was removed by the condenser and, since the preferred path was lost, the plant was cooled by use of the atmospheric dump valves, a sgow, long process. Radionuclide concen-trations were about 1 x 10- microcuries per milliliter. He noted that the Staff is also looking at where the steam and water went, since halon and C07 systems were actuated. Since there was halon and C0 in the switchgear and cable tunnel areas, the licensee ventilated th$areabeforeallowingpersonnelaccess. The Staff also intends to look at systems interactions questions but does not have any answers at this time. C. Michelson was told that there were no spurious operations of the fire suppression equipment. H. Etherington ex-pressed concern regarding the current status of the concept of leak-before-break criteria in light of this high-temperature steam pipe break that couldn't happen. R. Hernan iterated that this system is not one of the systems the Staff would have considered covered under the broad scope rule. It was also not one of the systems considered and was ruled out during the Beaver Valley exemption recuest.

G. Loss of Low-Pressure Service Water at Oconee-2 H. Bailey reminded the Committee of his discussion at the October ACRS meeting of an event at Oconee-2 where, during load shed tests, the licensee expected to have a siphon flow to their low-pressure service water pumps after shedding some condenser water circulating pumps (see Appendix VI). They lost the siphon flow after an hour. Gravity flow through the condensers was to supply water to the low-pressure service water pump suctions. This piping did not stay flooded because of drought conditions and subsequent air leakage in the discharge flange of one of these pumps. He explaired that the low-pressure service water pump supplied many safety-grade loads in the plant and, as a result of this event, the Staff decided that when such a system relies on a siphon to operate, tl.e system must surely be airtight to prevent air-in leakage.

H. Bailey explained that durino a meeting with the Staff on Octo-ber 14,1986 the licensee convinced the Staff that it had rectified the problem by repairing the flanges en all 12 ccmponent cooling water pumps to prevent air-in leakage during times of low lake level. The

320TH ACRS MINUTES 14 fix is a . skirt around the leaky flange so that- the flange is rede-signed to prevent air-in leakage. -

pump is lost, the licensee will have a siphon capability.Even Neverthe- if the co lose the component cooling water pumps to begin with blackout.

bility to maintain the siphen.This is a moot point as long as they have show All three units were started back up a few days under after the October 14 meeting and an information notice is preparation to alert all licensees of this event.

J. C. Ebersole asked if the licensee contemplates an emergency suction system where they will take suction directly out of the river.

H. Bailey indicated that they have not proposed such an emergency 1

suction system although they may have. it under consideration.

IV. Containment Performance (0 pen)

[ Note:

of the meeting.]M. D. Houston was the Designated Federal Official for this portion R. Bernero, NRC, indicated that he is proposing, in the form of a generic letter, in some severe accident issue modifications to the Mark I containments the U.S.

Other boiling water reactor pressure suppression containments lend themselves requirements. to the same treatment albeit it with slightly different ing PWR containments.The same can be said for other containments as well includ-He encouraged others to use this as a model of action for application to other containments as well as the Mark I. The Mark I has been selected from among other containments because of its perceived accident. high probability of containment failure followino a severe containments). It involves the largest population of licensed plants (24 8,1985 said that present nuclear power plants are saf one knows that there are not any outliers' or significant vulnerabilities to severe accidents that might be plant-specific in character. The Commis-mitigation for outliers. and the concept of defense-in-depth in dealing tified (see Appendix VII),Special consideration of containment performance was iden-R. Bernero indicated that the search for outliers is really an identiff-cation quantified. of a significant vulnerability which does not necessarily have to be The Severe Accident Policy Statement niques to identify outliers and to backfit When cal as appropriate.alrea a techni-issue goes rulemaking beyond the current regulatory requirements, generic is preferred.

generic letters are also acceptable. Alternatives such as bulletins, orders, or R. Bernero explained that General Design Criterion 16, "Containtrent De-two key design criteria dealing with containments. sign," and Gen severe accident or core melt is beyond the design basis.He indicated that a l

is an essentially leak-tight barrier against the uncontrolled release ofThe containment

320TH ACRS MINUTES 15 radioactivity for as long as postulated design basis accident conditions -

require. Criterion 16 goes on to define the conditions as those that result from degradation but not total failure of emergency core cooling function. The design basis accident postulates certain degraded condi-

, tions, or single failures, but it does not enter the realm of no cooling at

. all where core melt prevails. He spoke of the sloppiness of language that has grown about the issue of containment performance. " Containment fail-ure" cannot be defined as " loss of leak-tightness," because a successfully i~ vented, filtered pressure relief is not a failure. " Failure" ought to be defined as "a large release, or relatively large release for which the containment did not provide substantiel mitigation." For a given plant, I

one has a family of accident sequences, each with its own probability. If one takes a weighted average of those conditional containment failure probabilities what results is a fraction of core melts that can be expected to give a large release. This is the overall success or failure rate of containment. J. C. Mark askcd whether a large release would be defined in terms of the number of curies released, or the number of rems at the plant boundary. R. Bernero indicated one could distinguish a laroe release in a qualitative way to be anything notably in excess of the nnble gas activity released from a filtered, vented release. He admitted that noble gas activity alone would still be a nontrivial release. For purposes of discussion, he defined a "relatively unmitigated release" as "one that involves relatively large quantities of aerosols."

R. Bernero discussed containment conditional failure probabi'ities from U.S. BWR plant specific PRA studies. He explained that the 90-percent failure probabilit Study (WASH-1400) y indicated because byBWR in the H. Denton the dominant derives from the Reactor releases Safety were not fil-tered releases. He indicated that, for the Limerick containment, it is assumed that the containment always bursts in the analysis, but roughly half the tine the burst was a filtered release yielding a 50-percent value for conditional probability. In the case of the GESSAR-II design, the containment bursts in the wetwell all of the time but the unfiltered outccme is only a tiny fraction, or about 1 percent. Therefore, the ccnditional failure prcbability ranges from 90 percent to 1 percent, a minimum on a Mark III containment because the Mark III has a drywell that is vastly stronger than the wetwell. D. Okrent indicated that he recalled an open question concerning the probability of failure of the GESSAR vessel pedestal and the loss of the integrity of the drywell. This will result in a partial pool bypass because some of the fission products may have gotten into the pool, a nontrivial fraction would be available to get through by another path. R. Bernero contended that pedestal collapse was a low-probability event and not significant. D. Okrent contended that there is a large uncertainty in that conclusion if most core melts were to lead to failure of the pedestal and consecuent failure of the drywell. R. Bernero ,

indicated that he expected the results from NUREG-1150, which is not yet '

available, to show containment conditional failure probabilities for Mark Is in the range of 10-90 percent, as previously discussed.

R. Bernero indicated that the Mark I containment is so vulnerable because of its small free volume. He explained that, during the early phases of a core melt, gaseous and volatile activity is driven off as the feel disinte-grates and the relief valves conduct these volatile materials, the gases, cesium, iodine, and possibly cesium iodide, into the suppression pool. The

320TH ACRS MINUTES 16 noble gases are not scrubbed since they are not soluble.. A large-scale --

core melt has to dissolve through the control rod drive assemblies which are in the bottom of the BWR vessel and then flow down onto the floor in the lower drywell area. This area is open to the surrounding containment..

The drywell wall is very close to the lower area and the downcomer vents

• that take the steam away are nearby. As a result of a large-scale core melt, the molten debris has the ability to directly attack the wall.- The-resultant vigorous core concrete interaction will generate significant i hydrogen, CO , and entrained aerosols, besides what is generated in the coredegrada[ ion.

4 R. Bernero .noted that the BWR has nore water . systems, is more adaptable, and is a more flexible system. The upper part of the core is. nomally

. blanketed in a two-phase mixture. Because of these factors, the. likelihood

=

of core melt is probably somewhat lower for a BWR than for a PWR. He, nevertheless, pointed out that the BWR has the vulnerabilities associated with core concrete interaction if there is a large-scale core melt on the e floor. C. J. Wylie pointed out that you have sprays in 'a BWR capable of pouring water on the core debris. R. Bernero admitted that the safety-

. grade redundant spray systems are good; nevertheless, he noted that,-

depending upon the amount of inert gas generated and the average pressure and temperature which will drive the pressure, one can reach overpres-surization ' of the Mark I containment. One is likely to reach a bursting-pressure rather quickly because there is one-eighth the' volume for the Mark I containment than in a large, dry PWR containment.

R. Bernero indicated that the generic letter adopts a five-element approach to the improvement of the Mark I containment. He proposes embracing i control sure reliefof hy(drogen, venting) inchanging order tocontrol of the sprays failure, avoid catastrophic in the drywell, controlpresof .

core debris, and procedures and training for this emergency response. The containment improvement strategy involves preventing hydrogen combustion by inerting and reducing the drywell spray flow rate. The strategy calls for cutting the number of nozzles to permit the use of smaller backup pumps, and to produce a controlled geometry at about one-tenth flow. Now that one has the lower flow, one has to have two backup power supplies. One could be conventional AC buses, and the other should have the capability of operating during a station blackout. A typical plant could probably respond by using existing connections or slight adaptations. An entirely new system should not be necessary.

'~

D. Okrent asked how sure the Staff is that adding water into the drywell would not lead to adverse interactions between water and core debris, i , R. Bernero thought that the primary issue would be one of steam explosion, a vigorous or violent reaction from molten core material dropping into a pool of water. He suggested that several factors indicate that this is not

a significant situation in a BWR. In the BWR the large-scale core melt was designed such that the bulk of the corium would reach the lower drywell area. It is necessarily more dilute and at a somewhat lower temperature 4

than in a PWR. It is also falling into a very' limited pool of water. The i

geometry in the lower drywell area will limit the water to about one foot, and in some plants as little as seven inches. D. Okrent expressed concern regarding the steam explosion that breaches the drywell. He wondered if 4

the Staff had done a bounding calculatien to show'that it does nct matter i

a

- . - , - - - - - . - - - , . - , , , - . , - , - ..-, - - - - - -.- -,-,-c., -,

.' 320TH ACRS MINUTES 17 if a steam explosion occurs. He indicated that it was his impression that --

steam explosions of modest size can occur with a fairly high degree of likelihood in the presence of water and the molten corium. R. Bernero did not deny that there will be a vigorous exchange or that the water-corium interaction would be hannful to the containment. In any case, he explained '

that, if one were to have a truly bypassing failure due to debris injection

, by a steam explosion in the lower cavity, the material released would not only have to miss the hardware but would have to go through the area.

Merely perforating this area is not a significant failure. One will get noble gases in the reactor building, but considerable filtration and attenuation will have occurred. There is not a free path to the environ-ment outside the plant.

R. Bernero indicated that the rest of the strategy is to provide the drywell spray such that there is wetwell pressure relief to the stack. The argument is basically that this is a procedure of last resort since one does not really want to release the gas unless unavoidable. The pressure

. relief, or vent, to get the pool scrubbing at an elevated release requires a hard vent able to stand the pressure one chooses to vent to the stack.

The generic letter says that the licensee should select the pressure somewhere between design pressure and 50 percent overpressure--no more.

The Staff does not want to count on a margin of more than 50 percent.

F. J. Remick asked if the material to be released would be going to the standby gas treatment system. R. Bernero indicated that the standby gas treatment would not withstand the pressure. The Staff does not expect any cleanup beyond the pool scrubbing and delay in the torus.

R. Bernero indicated that the Staff is concerned about pool bypass. He noted that after the volatile activity is released, then lanthanum oxide and aerosols that are generated by the core concrete interaction will be released. The Staff wants to make sure that these products are scrubbed in the suppression pool prior to venting. This is one of the reasons why the presence of a continuous spray in the drywell is so important. It provides a second scrubbing mechanism.

R. Bernero discussed the conditions for the reouirements in the containment improvement strategy. The requirements are intended to be an optimized use of existing eouipment. Added equipment need not meet the quelity or design standards of safety-related equipment, but modifications to, or near, equipment or to systems which are already safety related should not compro-nise the quality of such equipment or systems. Pe explained that contain-ments which are reinforced concrete, such as a BWR Mark II, a large dry PWR containment, or a drywell structure for a BWR Mark I, have inherent capa-bilities that probably make them equal or better than other systems. The Mark I containment has a reinforced concrete biological shield, but the liner inside the biological shield and the independent coupled torus, present a vulnerability. The approach in the generic letter does not provide a mechanism to deal with that apparent vulnerability since there is no obvious mechanism available by optimal use of existing equipment short of literally redesigning the Mark I containment'.

D. Okrent noted that, given the implementation of the generic improvements for the Mark I containments, the Staff may wrtte off any need to look again at the containment seismic capability. He suggested that the Staff is also

j 320TH ACRS MINUTES 18  !

I writing off any further look with regard to other -external events that --

l might be of interest. He noted that there has not been, to date, a good  !

evaluation of Mark Is from the seismic point of view.

. C. Michelson asked about the decisionmaking point for the reactor operator once this equipment is installed. R. Bernero indicated that, during an incident, the operator would follow the symptoms in the reactor coolant system and containment, including the containment pressure and temperature.

The operator's procedures would tell him when to start using the auxiliary spray and the pressure for venting (as low as design pressure or as high as one and one-half times design pressure).. Up to that point, the operator is following his conventional mitigating steps. The operators would not be permitted to vent the containment during an event (noble gas release) without proper authority, e.g., the techr.ical support center or higher authority. D. Okrent suggested the addition of a chilled filter to cut the noble gas activity from the vented release.

R. Bernero indicated that on July 16, 1986 the Commission responded to a hearing question from Congressman Markey about the PWR Mark I containment.

H. Denton had cited a 90-percent chance of failure of the containment in the event of a core meltdown. In effect, this was asking about the ac-ceptability of a 90-percent fraction of core melts resulting in a large release. The Comission indicated that the likelihood of core melt should be very low and there should be substantial assurance that the containment t

will mitigate the consequences. It is not merely a question of low risk, but defense-in-depth assurance of combined protection by prevention and mitigation. The Comittee discussed the cost-effectiveness of making a more extensive modification, such as a chilled filter suggested by D. Okrent. R. Bernero insisted that this modification could not be justi-I fied on a cost-benefit basis. R. Bernero indicated that the proposed generic letter is scheduled for CRGR review about December 22. He indicat-ed an interest in any constructive criticism the ACRS might wish to make at this time. D. Okrent nentioned a technical paper presented at the last American Nuclear Society meeting by Chu and Stoyanov about a new scenario for intersystem LOCAs in BWRs. R. Bernero indicated that he was not familiar with that report but would look into it.

T. Pickens, NSP, representing the B&W Owners Group presented the NUMARC-approved severe accident containment issues approach (see Appendix VIII).

He indicated that NUMARC and the BWR Owners Group have taken on a parallel progran to look at the Mark I issues. He confirmed that utilities do not lack confidence in the capability of the Mark I containment to meet the challenges of a severe accident. They agree that if new information or issues are brought forward in the Severe Accident Policy an appropriate action should be considered. T. Pickens indicated that NUMARC does not see any rew information being brought foniard but a refinement of some old information available at the time that the judgment was made that these plants were safe. The overall objective of the NUMARC procram will be to evaluate containment integrity and potential improvements to minimize offsite releases for severe accident conditions beyond the design basis accident within an appropriate cost-benefit goal. The program will identi-fy challenges to containment such as hydrocen generation, overpressure, or core debris attack and identify initiators for each challenge frcm existing analyses. It has been suggested by some NUMARC ' members that NUMARC put

o .

320TH ACRS MINUTES ,19 l

. forward a comprehensive document identifying all chal-lenges to the Mark I .-

containment, even those being examined under IDCOR and other industry efforts, and how NUMARC believes they can be addressed. The plan is to use IDCOR analysis and any plant-specific PRAs. Concerns of the people at the

• various national laboratories will also be incorporated. The next step will be to assess the abilities of these plants to meet those challenges, identify the vulnerabilities, and propose alternatives to address the vulnerabilities. The alternatives might turn out to be identical 'to those of R. Bernero. A preliminary look at the information has identified the 3

drywell sprays as critical, as well as containment venting, but NUMARC has not identified any fixes. Alternatives will be evaluated using the acci-dent sequences, the apparent benefits, and the risk reduction against .the cost. The BWR Owners Group plans to present the results of its report to the NUMARC Working Group so that NUMARC can identify whether there are initiatives that the industry wishes to sponsor.

T. Pickens indicated that IDCOR produced a number of 20 percent for condi-tional containment failure probability for the Mark I containment. Some IDCOR members believe it to be even lower, as low as 5 percent. Regarding 4 NUREG-1150 and- the 90-percent value postulated in July 1986, the industry took action at that point. Two key parameters that will have a great deal to do with the conditional failure probability that is finally reported in i NUREG-1150 are the ultimate pressure capability of the containment and how the corium acts. There is EPRI work going on in the area of corium phe-nomenology and an Owners Group-sponsored program looking at the ultimate pressure capability of the Mark I containment which is currently reported at 135 psi. Such a- pressure capability will drastically alter the poten-

, tial for failure, reportedly 90 percent in NUREG-1150. It might decrease to the range of 10-30 percent, or much lower.

T. Pickens noted that the individual plant evaluation program is not currently structured to address the seismic capability of the Mark I

containment. D. Okrent was curious that the BWR Owners Group badn't, on l

its own initiative in the past, tried to evaluate the seismic capability of the Mark I containment. D. Pickens iterated that the Owners Group is very i comfortable with the seismic capability of the Mark I containment. There is a Seismic Owners Group that is addressing the entire seismicity issue.

Utilities are participating in that effort. D. Okrent wondered if the Owners Group in its evaluations will allow for situations such as the degradation of the drywell liner that has recently been discovered at Oyster Creek. The Comittee is concerned about degradation that might i significantly affect the containment failure pressure. T. Pickens assured

i. the Committee that, if there were indications that pointed to degradation

[ as a generic problem, it would certainly be incorporated in the BWR Owners Group study.

T. Pickens indicated that the Owners Group expects to issue a final report on the containment issues by March 31, 1987. D. Okrent expressed corcern

that the Owners Group might use a computer code that the Staff ~ has not reviewed and which might precipitate another six-month delay.

! J. C. Ebersole reminded the Comittee that a proposition has been made to

take a source of water and put it into a spray system to spray the contain-ment. He thought that, if there were a supply of water, the first priority ought to be to put the water on the core. R. Bernero agreed and he noted l

. i - _ _ _ _ _ _ . _

=~ .. - -

+

1

' I 320TH ACRS MIfiUTES 20 .

4 that the procedures at Vemont Yankee give the operator the priority --

determination with the first choice to put water on the core and, if for some reason they are unable to do so, to put the water in the containment.

D. A. Ward suggested that the Committee draft a letter for submission to ,

the Commission. J. C. Mark thought the letter might say that R. Bernero should press ahead with hi: proposal and the Commission ought to wait for the results of the NUMARC proposition.

V. Reactivation of Nuclear Power Plants (0 pen)

[ Note: R. P. Savio was the Designated Federal Official for this portion of e themeeting.]

T. Michaels, NRC, discussed the definitions of " deferred," "teminated,"

and " cancelled" plants. He explained that a deferred plant is a plant where the licensee fully expects to reactivate construction but for one reason or another has ceased construction or reduced the construction effort to a maintenance level. The licensee maintains his construction permit. A licensee does not plan to reactivate construction on a termi-3 nated plant. His construction permit is still in effect. He may have 4

requested withdrawal of the construction permit, but the permit has not yet l been withdrawn. A cancelled plant does not have a valid construction

• permit (see Appendix IX). The Policy Statement on Deferred Plants talks' j about a deferred plant and a terminated plant, but does not deal with a j cancelled plant. Cancelled plants are discussed in Commission paper l l SECY-86-359. In answer to a question by D. W. Moeller, T. Michaels indi-  !

cated that there are 4 deferred plants, 12 teminated plants, and 26 cancelled plants at this time.

) T. Michaels discussed directives on deferred / cancelled plants and the i Policy Statement considerations. He explained that a Staff tesk force was

set up to address the directives. Considerations of maintenance, preserva-i tion, and documentation requirements for deferred plants, applicability of new regulatory Staff positions for deferred plants being reactivated, and

. procedures for reactivating deferred plants are addressed in the Policy Statement itself. Identification of renulatory improverents and research initiatives are in Enclosure 3 of SECY-86-359 and population and status of deferred and terminated plants is in Enclosure 4 of. the Commission paper.

The Policy Statement principally focuses on procedures developed for' deferral and reactivation of deferred plants and does not cover reactiva-tion of cancelled plants. D. W. Moeller asked about reactivation of a terminated plant. T. Michaels indicated that, if a licensee has any inkling that it might reactivate a terminated plant, it should follow the

~

same procedures from the standpoint of maintenance and preservation as a deferred plant.

. T. Michaels indicated that one of the more important aspects of the Policy

Statement is the identification of regulations and guides associated with maintenance, preservation, and documentation. The applicable regulvtions

and guides which apply to both plants under construction and on deterred

. status have been identified in the Policy Statement. The licensee may choose to modify existing commitments during extended construction delays by developing a quality assurance plan comensurate with the expected activities and length of delay. The licensee is ' expected to discuss with i

~

320TH ACRS MINUTES 21 ,

the appropriate regional office and headquarters the expected construction --

delay period and the quality assurance program to be implemented during the deferral. Modified quality assurance programs are inspected and reviewed in accordance with appropriate inspection procedures.

T. Michaels explained that plant-specific backfits of new Staff positions will be considered in accordance with the Backfit Rule 10 CFR 50.109, which requires that the Staff justify, by value/ impact analysis, any backfits after October 21, 1985. F. J. Remick was not exactly sure about the application of the Backfit Rule. T. Michaels noted that there is more than one cut-in time in the Backfit Rule and, if a construction pemit came in before October 21, 1985, it would be based on a date six months before the operating license was docketed. A deferred plant would, in a sense, have the same protection as en operating plant and also the same requirements.

D. Okrent asked what would be the case if a plant is deferred for a dozen years under this proposed policy. T. Michaels indicated that the Staff does not see any difference because the plant would be preserved. Any new designs if they were promulgated on an operating plant would also be applied to a deferred plant. R. Hernan noted that there is a defined termination or expiration date for each construction pemit. He asked T. Michaels if the new construction permit renewal process would now apply to these deferred plants. T. Michaels indicated that it would. D. Okrent was not sure that backfitting considerations should be identical for a plant that is 50 percent complete and then is delayed for ten years as for a plant that has been completed on time and is running. T. Michaels explained that, if a plant was licensed at one point in time and 10 years elapsed, any backfits would have been applied with the Backfit Rule. He noted that in the case D. Okrent cited the design has been approved at one point in time by having its operating license docketed; even thcugh it is not completely approved, in any case, it has been accepted and any further plant-specific backfits would be applied by the Backfit Rule. D. Okrent asked if the plant has been reviewed by the Staff. T. Michaels indicated it does not mean that it has been reviewed but has an adequate submission with an FSAR. D. Okrent asked if there are any deferred plants that do not have an OL docketed. . T. Michaels indicated that all of the deferred plants have docketed OLs.

T. Michaels indicated that generic backfits will be implemented either through rulemaking or generic issue resolution. When an issue is resolved from a generic standpoint and the resolution is applied to a particular class of plants, if the plant was of that particular class the backfit would have to be implemented. D. Okrent was still skeptical that a de-ferred plant maybe 10 years from operation would be treated as a plant that was already in operation. T. Michaels indicated that a deferred plant is a plant that is similar to a plant that is under construction. R. Hernan added that the only thing that is different is that the rate of ccnstruc-tion is much slower and can be nonexistent. T. Michaels indicated that provisions of other policy statements applicable to plants under construc- j tion will be implemented. He noted that in the Severe Accident Policy '

,< Statement (NUREG-1070) it states that an integrated, systematic approach must be taken to examine each nuclear pcwer plant now operating or under construction for possible risk contributors that might be plant specific.

Deferred plants will have to have this plant ~ specific vulnerability analy-sis; however, the Backfit Rule will be used to ' decide which identified l

l l

L_ __ _ _

. 320TH ACRS MINUTES 22

• i

plant vulnerabilities need to be . implemented. H. Etherington aske'd if ..

there. is a step change in NRC surveillance for deferred and terminated j'

plants. T. Michaels agreed that a deferred plant goes through lesser inspection requirements than an operating plant or a plant under construc-

tion. Although a plant is terminated it still has to be . inspected semi-

• L annually until withdrawal of the construction permit. R..Hernan added that L whether terminated or deferred the same standards and inspection frequen-cies would apply. C. Michelson asked if a . terminated plant still has full-QA control. . R. Hernan explained that the teminated plant still has an

{; active construction permit. The big step change comes when . the plant is cancelled. At that point- the construction permit is no longer valid and ,

the controls go away. D. Okrent asked if there is any requirement on the

level of engineering staff that a _ utility must maintain for a plant in a j deferred state. T.' Michaels indicated that there is no requirement but he noted, as an example, that there were 50 man-years expended on the mainte-e nance and. preservation of WPPS-1. D. Okrent expressed . concern that a '

! licensee of a deferred plant would not keep up with the current state of  ;

j the industry and not be knowledgeable of current events. T. Michaels ex--

i plained that before a licensee comes in for reactivation they must list all j of the outstanding generic issues that have been prcmulgated on that plant '

4 and all other outstanding action items. D. Okrent asked if the utility has to respond to IE information notices. R. Hernan indicated that the Staff i i requires a response from the utility to any bulletins and engineering ,

! letters that are issued. C. Michelson asked if the NRC reexamines the . i

! competency of the utility at the time the utility expresses a desire to

! restart a project. T. Michaels thought it did and indicated that proce-1 dures for reactivating a plant are stated in the Policy Statement.

T. Michaels explained that during the development of the ' Policy Statement i

the task force was requested to identify any areas where the regulatory -

3 framework could be improved. It was found that the present regulations are j not sufficiently prescriptive with regard to withdrawal of a construction i permit. Nothing is said in the present regulations on what happens if the l licensee asks to have a plant terminated and a construction permit expires-t before the NRC can process the cancellation. The Policy Statement address-

. es this uncertainty by requesting that the licensee provide notice to the j Staff sufficiently far in advance of the expiration of the construction l

permit to permit the Staff to determine appropriate tems and conditions.

i Prior to formal cancellation of the permit by the Commission the permit j holder is expected to comply with the . Connission's rulings, including taking affirmative steps to extend construction permits which may expire <

F before Commission action and the termination request can be completed. He suggested that the Staff may need some additional internal guidance on i specific inforration needed for termination of a construction permit, such I as criteria for site stabilization and other aspect 3 of site termination.

) Also necessary may be detailed cuidance for inspection of deferred plants '

! prior to reactivation.

1 T. Michaels indicated that the Staff met with the Atomic Industrial Forur j (AIF) on February 19, 1986 and requested its -views on the necessity of a i policy statement for deferred and cancelled plants. The AIF provided a j letter on March 31, 1986 on the reactivation of construction projects and i

agreed that a policy statement was necessary, that it should apply to l plants with construction permits, that plants with withdrawn construction I

320TH ACRS MINUTES' 23 i permits should.be hardled on a case-by-case basis, that the Backfit Pule -

should be used to implement new requirements in effect for plants with construction permits, and that the current preservation requirements were adequate as long as records are maintained. Their review and comments on

• the final draft of the Policy Statement was submitted to the Staff .on j July 2, 1986.

j T. Michaels indicated that the Policy Statement does not have any new

safety requirements or address any new. safety issues. It essentially

1. consolidates existing requirements' and clarifies how the Backfit Rule will be applied when a deferred plant is reactivated. It has been coordinated with the industfy through the AIF. Additional coordination is expected  ;

c since the Policy Statement will be issued for public comment. C. J. Wylie

asked if the Staff would do a complete review of the construction organi-zation when a plant is being reactivated. He asked if the Staff would do a thorough and complete inspection of all equipment installed prior to

'- putting the plant on hold to . assure that it has not deteriorated during that period. .T. Michaels indicated that there is a semiannual inspection i performed by the regions of the preservation and maintenance program at the

[ sites. C. J. Wylie expressed concern that equipoent put into the plants j may have been exposed to weather and changing conditions. Unless the equipment is disassembled and inspected the condition of the equipment would still be questionable. .T. Michaels indicated that there are criteria for inspecting that equipment, including appropriate procedures. He added tha t, in addition to the ongoing maintenance and preservation program, there are startup procedures that the licensee would have to conduct before pre-operational testing takes place.

F. J. Remick indicated tnat since the Staff still mandates that the pro-

posed policy statement be issued for public coment the Subccmittee i recommends that there is not a need for a Comittee letter at this time.

i D. Okrent indicated that he had a differing view and expected to present a

draft letter on Saturday morning. A Comittee letter (report) on this 1

subject was approved later in the meeting.

1 .

] VI. Pressurized Toermal Shock (0 pen) 4

[ Note: E. G. Igne was the Designated Federal Official for this portion of j themeeting.]

P. G. Shewmon reminded the Cemittee that a pressurized thermal shock (PTS)

! will take place only if three simultaneous occurrences or events take  :

I place. The first is a lenc-term overcooling; the duration can be from one-half hour up to a few hours. During this time, the system pressure has '

to be maintained at or above operating pressure. Finally, you have a '

i- vessel which, due to irradiation, has had a large increment in its original

ductile brittle transition temperature. The siruitaneous occurrence of

these problems can lead to a very severe accident. To counter the scenar-

. ios, the Staff had the Oak Ridge National Laboratory (ORNL) calculate the

! PTS probability due to all different kinds of transients. The result was a PTS Rule with trip points or screening limits. All licensees were told that if the projected coming within the screening limits before the end of

life of the plant they would have to submit li report to the NRC Staff to l explain their strategy to justify continued operation of the plant to the I

s .- .

320TH ACRS M8NUTES 24 4

end of its life. The report would have to be submitted three years before --

the screening limit was to be exceeded. The PTS Rule has been promulgated but a regulatory guide defining how these analyses would have to be= done has only recently been out for public coment. He indicated that the

• purpose of the' discussion at this meeting was to hear about changes the Staff has made as a result of the public coments on the regulatory guide.

. He noted D. Okrent's particular concern about the substantial uncertainty in the first cf the three reports issued by ORNL..

i R. Woods, _ NRC, indicated that the PTS Rule was promulgated ~ on July 23, f- 1985. The Rule requires extensive analyses to be submitted three years i

before the screening limit defined in the Rule is reached. The Staff I

promised detailed guidance as to how to do those analyses in the form of a

.- regulatory guide which was published for coment in January 1986. Coments .

~g were provided X (see Appendix by)the

. TheACRS StaffSubcommittee on MetalasComponents took those coments well as ACRS in February full 19 Committee comments in a letter of March 18, 1986 and incorporated them in  ;

. the proposed Guide which currently considers public coments and recent NRC

. Staff and CRGR coments. P. G. Shewmon noted that about a year ago all i licensees were asked to indicate whether they projected reaching.the screening limits before the end of life. . Some utilities have changed and l reconfigured their cores as a result of that request. .

l R. Woods indicated that one of the provisions of the -Rule was that each

! licensee was instructed to tell the Staff exactly what its RT values

! are. Those reviews are nearing completion.- According to the Nensees, i with the flux reductions to which they have comitted, no licensee is i indicating it will reach the screening limits before the end of the present license. The Staff has six reviews under way at: the Brookhaven National .

Laboratory (BNL) for plants that have proposed 'large flux reducticns. . BNL r is checking the results. The NRC Staff does not yet know whether it can

! agree that the utilities will achieve, the degree of flux reduction for

! which they are taking credit. The Staff did not think it necessary to send

, analyses for nine additional plants that have made a lesser comitment to flux reduction for review by BNL. The Staff intends to follow and verify, 1 l as we proceed through the years, all of the flux reductions to which the

!_ utilities have comitted. The utilities will have to submit periodic

! capsule surveillance reports and redo the Appendix G limits based upon

, irradiation. P. G. Shewmon asked what was the largest flux reduction i proposed. R. Woods indicated that H. B. Robinson proposed a reduction in .

i the flux factor of 10. H. Etherington asked if the licensees are submit-i ting revised calculations based upon the new Regulatory Guide 1.99. R.

Woods indicated that this is an entirely different subiect. For purposes

( of the screening limits, the PTS Rule says that the licensee must' use a ,

j correlation that is already built into the PTS Rule. H. Etherington noted

that it is really based somewhat on Regulatory Guide 1.99, as they both

! take account of the nickel in vessel welds. R. Woods evolained that you do l get a slightly different result. He did note that there will . be some

further discussions about whether the Rule or the screenino limits should j be changed as the new correlation comes out~ in Regulatory Guide 1.99.

, P. G. Shewmon noted that if there isn't a licensee that l' umps up against i screening limits the correlation in the PTS Rule will never be used.

l R. Woods agreed that this is so by the end of the present licerse for all 3

the plants.  ?!evertheless, the Staff expects that almost 'every licensee i

[

l

I

'320TH ACRS MINUTES 25 e

i will ask for a. license extension and this regulatory guide will be criti- - - -

cally. important to the licensing process. D. W. Moeller noted that the bottom line is that the ' flux reductions are to be verified by the Staff.

He asked how much uncertainty. there is in these flux reductions. R. Woods

.. indicated 5-10 percent. It is a calculation initially that will have to be verified by future capsules that come out of the core,

! R. Woods briefly discussed the public coments that resulted in no recom- '

l* mended change. . He explained that some thought that the PTS Rule was not l adequately conservative. The Staff decided that this was not relevant to 1 i the regulatory guide since the guide merely instructs on how' to do the 4

analyses once one triggers the requirempts of the Rule.- D. Okrent noted

.. an inconsistency regarding the 5 x 10-

! regulatory guide as contrasted with the 10,ger perreactor reactoryear year number value .for in the the i probability of a large release. He noted that studies done by the Staff's-contractors suggest that if one gets through-wall cracks the odds are more

. than half that the vessel will fail. He suggested that if this number is calculated in the regulatory guide it would exceed the suggested limit of

, the Comission in the Safety Goal. W. Minners, NRC, indicated that perhaps

! this should be pointed out to the Comission. R. Woods noted that the PTS 1 Rule leans more heavily toward prevention even though the Staff did do some i preliminary consequence calculations. He agreed with D. Okrent that if one j.

gets a through-wall crack one is a factor of two from having a core melt.

However, he did not believe that one would be within a factor of two of the

] probability for a large release.

! D. Okrent asked about the integrity of the containment in the event of a j catastrophic vessel rupture. R. Woods indicated that a PTS vessel failure

increases the probability of core melt by a factor of two; but there is a very small conditional probability of early failure of the containment l given the failure of the vessel. R. Woods mentioned a paper by Barrett and i

Throm at a conference about a year ago. The paper looked at the likelihood

! of containment failure given catastrophic vessel failure. D. Okrent asked

, if the Staff is backing that study. R. Woods indicated that the Staff is i not issuing it as Comission policy since _there are members on the Staff in j e different areas that disagree with the study. They are looking more at the i methods that would be applicable to ice condenser plants. D. Okrent asked l for a copy of the Barrett and Throm reference.

i.

! Several public coments suggested that the Regulatory Guide was either i premature, incomplete, or not needed, thermal-hydraulic code ;pecifications

! were too loose, and nodalization schemes and air estimation tachniques were l not well specified. The Staff thought the Guide cught to have a fair

! amount of flexibility because one cannot be sure of the method that would I be most applicable many years from now. The Staff, therefore, purposely left a great deal of flexibility in the Regulatory Guide. Regarding the

. fact the some thought that the Guide was not needed because enough flux l reduction actions have been taken to avoid exceeding the limits du_ ring the i present license, some comenters asked why licensees could not do something l other than flux reduction. The Staff acreed that they could, but the only i ,

way to evaluate the safety benefit for other corrective action would be to

! 'do an analysis as described in the Regulatory Guide. He indicated that

some individuals thought the DBA approach should be allowed. The Staff ~

j started the PTS effort years ago in that vein, b0t couldn't find the DBA 8

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320TH ACRS MINUTES 26 that covered all sequences. If one had some sort of-DBA it could lead to .-

falsely . evaluating corrective actions. One needs a PTS that. takes all of the types of events into account in order to properly judge what corrective actions should be applied. The Staff disagreed with this comment. ,

R. Woods discussed public coments that resulted in changes in the-Regula-

. tory Guide. Consideration of condensation outside the pressurizer is now

required by the Guide and pressure control is now listed as -a potential corrective action. R. Woods indicated that it was pointed out, through Staff coment, that the Regulatory Guide might be taken to overemphasize mitigation. As a result, the chapters regarding vessel failure modes, containment performance, and early and late fatalities were deleted from *:

4 the final version and relegated to material by reference.

P. G. Shewmon suggested that' if the operators knew they had an overcoolin'c '

t event and could reduce the pressure over the tens of minutes there would U not be a concern about PTS. He asked if the Staff has factored this

)- element of prevention into their plans as far as providing training or different instrumentation to tell the operators they have an overpres-l

!' surization event instead of an overcooling event. R. Woods indicated that l the Staff has not required any new instrumentation. The Staff will rely on improved training of the operators and procedures. .

D.' Okrent indicated that the level of research effort in the area of the probability of containment failure given a catastrophic. vessel failure has decreased over time. While - the first study that was done concluded that i

the probability of vessel failure given a through-wall crack was quite high, the PTS Rule now states that the Staff is relying en the fact that the containment can withstand a vessel failure. He indicated that he was i not a believer. H. Etherington pointed out that PTS failures are low-i energy failures. The water would be relatively cold; it would not be anything like the energy available on a postulated rupture under operating 4 pressure and temperature. W. Minners noted that the conscouences of a

vessel rupture such as a containment failure and the source term and its i distribution in the environment were all taken out of the Regulatory Guide.

This is not to say that a discussion of consequences is not there. The Staff has tried to deemphasize that aspect and discourage people from using

. consequences as part of their argument. The Staff thinks it knows with reasonable certainty what these effects would be and can begin to predict what would happen in the environment. Nevertheless, the Staff's approach is to have the licensees show the Staff that their vessel will not crack.

R. Woods indicated that it was pointed out at the CRGR meeting that the PTS Rule requires vessel failure prevention but does not provide for a risk

analysis. As a result of NRC lawyers' coments, the Guide now says that

, such analyses might be submitted on appeal to the Commission but are beyond i the scope of the Regulatory Guide.

R. Woods indicated that the Regulatory Guide has not been changed signif-icantly since the last time the Comittee saw it. It has been out for .

i public cogrent and received broad acceptance. The PTS Rule will not be l

, complete until the Guide is out. Analysis cannot be required until a year after the Guide is published and the Staff-is requesting that the ACRS l provide a letter recomending issuance of the final Regulatory Guide.

)

l l

t 320TH ACRS MINUTES- VII. Meeting with the NRC Comissioners (0 pen) - --

[ Note: Those Comissinners present were L. W. Zech, Jr. Chairman, T. M.

Roberts, F. M. Bernthal, and K. M. Carr.]

Comissioner Zech indicated that the subjects to be discussed at this periodic meeting with the Advisory Comittee on Reactor Safeguards are the status of Comittee activities regarding consideration of the effectiveness

. of programs which address generic and unresolved safety issues, and Comit-tee coments on work in areas of advanced and standardized reactor plants.

i A. Effectiveness of Programs Which Address Generic and Unresolved Safety Issues

, C. P. Siess explained that the Comittee has been involved in the review of unresolved and generic safety issues on a rather selective basis. A number of the major issues such as USI A-45 and USI A-46 4 have received a great deal of attention by the full Committee and its subcommittees. The full Comittee reviewed the Staff's - plan and criteria for prioritizing generic issues and. it collected batches of i these generic issues, assigned them to subcomittees, and reviewed the basis for the Staff's priority selections. In a number of cases the-Comittee has disa. greed with the Staff and, after some discussion, decided what the ACRS thought to be the proper priority level, t

C. P. Siess indicated receiving some statistics from the Staff recent-ly on generic issues. The document stated that there are 709 generic items, two-thirds of which have been resolved in one way or another.

There are 228 that remain to be resolved at this time. Of those, 154 j have been assigned priorities, 12 have been made unresolved safety

{ issues, and 73 have been classified as high or medium priority. This

! will mean that the 73 will get worked by the Staff. The remaining 69 generic items have been prioritized as low or dropped, which, with current resources, means nothing will be done about them in the near future. He indicated that there are 79 generic items which have not yet been assigned priorities and have not yet been reviewed by the Committee. If the ratio holds as in the past, about half of those will be assigned high or medium priorities. This will leave about 128 generic items that still remain to be resolved, t

C. P. Siess explained that sometimes resolution of a generic item means that nothing has to be done; other times it means that something

, should be done in the future but nothing sheuld be backfitted.

, Sometimes resolution means a process of action .that might take as ruch as five years. The Comittee's review of the resolution of generic

, items has been considerably more ' selective than its review of the prioritizations. Of the 449 generic issues that have been resolved,

, the ACRS has looked at only the major items. He noted that one 4 measure of effectiveness of the program which. addresses generic and unresolved safety issues is the extent *to which the resolution of generic items has increased safety. He thought the Comittee, unfor-tunately, did not really have sufficient infonnation to assess effec- )

tiveness on that basis. -

4

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320TH.ACRS MINUTES 28

' D. Okrent expressed concern with the progress' on the Backfit ' and - -

Severe Accident Policies and he suggested a connection between them and generic items. He suggested that, at some point, the Comis-sioners will have to seriously address the reason why major countries

• with light water reactor technology, like the Federal Republic of Germany (FRG), France, Switzerland, Finland, and others in Western l Europe, have gone well beyond what the U.S. is requiring of its current plants both with regard to prevention as well as mitigation.

D. Okrent stated that there are numeious specific examples where generic backfits in this country are unlikely to pass the cost-benefit test of $1000 per man-rem. There is the ouestion of whether such backfits may have some significant effect on plant risk and safety.'

Comissioner Bernthal noted that there is definitely a trend since the Chernobyl accident. But he wondered if, before the Chernobyl event, j- any of those countries "antioned issued broad generic requirements

[: both preventive and mitigative that have never been applied in the U.S. He indicated that he was aware of a couple of areas such as station blackout where the U.S. trails other countries. D. Okrent explained that the record is clear that the two major LWR countries in Western Europe, France and - the FRG, before Chernobyl, were taking significant additional steps to prevent core damage and significant steps to mitigate severe accidents. . He . indicated that he was talking primarily about large, dry PWRs although the FRG has a BWR as well.

Switzerland and Sweden took some very strong measures before Cher-nobyl. Comissioner Bernthal asked D. Okrent to be specific about the generic requirements in those countries. He indicated that he did know about the bunkered independent decay heat removal systems in the FRG and the French generic requirement for some kind of filtered vent.

D. Okrent noted that.there are a variety of things that have been done in France and he referred the Comissioner to the NRC Staff's Paluel report. He mentioned that the FRG has noved to the N+2 concept which

gives them greater redundancy in all of their systems. They had moved to even out the contribution for risk from any scenario before Cher-nobyl. The French had decided on a filtered vented containment before Chernobyl. W. Kerr thought the meeting going a hit off the programmed agenda, and he suggested that this topic could be discussed.in depth at a separate meeting with the Comission. C. Michelson pointed out-

. that the Europeans have a significantly different approach than that

, taken in the U.S. with regard to the question of sabotage and security protection of nuclear plants. Comissioner Bernthal acknowledged that

! there is greater protection against sabotage, especially attack from

the cutside. Comissioner Zech pointed out that he did not believe

! that such an approach to sabotage is generic over all of Europe.

! Comissioner Bernthal asked if the Comittee had any thoughts on how the Backfit Rule combined with the Comission's Safety Goal might impact the resolution of unresolved safety issues. He wondered if the i

Backfit Rule would do what it is intended to do, or whether the process will be too cumbersome. C. P. 'Siess noted that, from his observations on the operations of the CRGR, he did not think the process too cumbersome in relation to the time spent on a' technical resolution of these items. Whether the Tracess leads to a backfit is a decision that will be made on the bases 'of the cost-benefit and ,

I 9

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t 320TH ACRS MINUTES . 29 regulatory analysis. It is certainly relevant to the issue of decay . .

heat removal reliability. The CRGR process would only be a problem if it caused a delay in implementation. Comissioner Bernthal asked if the Committee had any opinion on whether the CRGR ought to continue now that the Backfit Rule is in place. C. P. Siess did not.think the Committee had addressed that issue. He noted that the Generic Items-Subcommittee has been trying to follow the CRGR activities. It was his impression that the CRGR has made the regulatory Staff more aware that they must make a case for what they want to do. The Backfit Rule has had the same effect at the . individual reviewer level. Commission-er Roberts noted that that was certainly the intent. Comissioner Bernthal thought that in the generic requirements area the CRGR still had a place and performs a valid function. H. W. Lewis and some other- ,

ACRS members thought the CRGR had done a good job on balance.

Comissioner Bernthal indicated that the Chairman is interested in whether the Committee is prepared to discuss any specific _ unresolved

safety issues, such as station blackout, on.which the Comission plans a broader discussion. He asked the Committee's opinion regarding the

. shutdown decay. heat removal requirement since the Staff may be about .

to issue some recommendations in that area. D. A. Ward explained that l the Committee has been reviewing the shutdown decay heat removal issue all along and plans a meeting next month on the status of the issue.

Nevertheless, the Committee was not prepared to discuss the subject at this time and does not have a consensus opinion. D. Okrent indicated that the Staff group, and their contractors, who are doing studies on shutdown decay heat removal are making estimates ,of the contribution of a loss of the ability to achieve shutdown heat. removal. Some of these estimates are larger than 10-4 per reactor year. This value covers a considerable number of the plants they have studied. Yet drafts of NUREG-1150 and Severe Accident Policy inpligations from the rebaselining work talk of core melt figures like 10 . He expressed.

concern at the discrepancy in the figures since NUREG-1150 will be one of the documents used for making policy decisions. He noted that the seismic problem has not been incorporated into either of those figures and there are a variety of other factors that have not entered into  ;

either of the estimates. Chairman Zech thought this a good subject to consider for a future meeting.

I In addressing C. P. Siess' cuestion on effectiveness, Chairman Zech indicated that he was asking the ACRS for its opinion as to whether the generic and unresolved safety issues are making a contribution to increased nuclear plant safety. Comissioner Bernthal indicated his interest in some Comittee opinions, separately or collectively. on the Safety Goal as promulgated by the Comission. I B. Comittee Views on Advanced and Standardized Nuclear Power Plants  !

C. J. Wylie stated that the Committee was prepared to discuss the '

review of draft NUREG-1225 Implementation of the NRC Policy on Nuclear Plant Standardization. The issue of advanced nuclear power, i plants is being handled by a separate subcomittee and the Comittee l was not prepared to talk about that today. He indicated that the l ACRS, in its coments on draft NUREG-1225, Was in general agreement .

i i

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320TH ACRS MINUTES 30 .

with the draft NUREG. The Comittee did, however, make several .

observations.

C. J. Wylie explained that in the first paragraph of the Committee's

• c letter it states that the Comittee believes that the Commission would benefit frem and should seek public cement on the design certifica-tion rulenaking options. The Comittee recomended that the criteria and threshold for standing and interest for participation in the legislative or adjudicatory rulemaking hearings should be clearly stated in NUREG-1225. F. J. Remick noted that the idea of using a rulemaking to certify standard plants has been the thought for some years. However, he thought that one thing that was truly missing, if the Comission wished to seek public coment, was to define who will participate in that rulemaking. With the site in mind, people living within 50 miles of a plant should be deemed to have an interest and be allowed to participate. Unfortunately when one is talking about a standardized plant, there is, in general, no site in mind. Since we are dealing with a rulemaking for design certification before a site has been selected, the Comission does have different alternatives.

Some of these would be hearings of different formats which would allow the public to participate as a courtesy. The whole point is that when the Comission is soliciting coments en this standardized plant policy it should let the pubite know, in general, who will be allowed to participate in the rulemaking process. W. Parler did not see any particular problem with public participation since the statute spells out who may participate in rulemakings or adjudications. He assumed that anyone wishing to coment in a rulemaking on a stan-dardized design would have his coments reviewed. He did not believe that there would be any particular problem either regarding an adjudi-catory type hearing having to do with the standardization policy.

Comissioner Bernthal thought the only departure would be that one were carrying out a rulemaking with the adjudicatory process more or less fonnally for a class of plant that presumably no lenger would be litigated on plant-specific issues as opposed to site-specific issues and could be readily licensed anywhere in the U.S. except for site-specific issues. He did not believe the process much different from that for the nuclear waste repository proceeding. F. J. Remick explained that the precedence has been basically that, in the case of nuclear plants in the past, the interest is defined as applying to anyone within 50 miles if they meet other thresholds. In this case, one has not defined the site. Other than that he did not see it as a legal question. Comissioner Fernthal agreed that the 50-mile prece-dent would seem inapplicable to this particular situation, but he indicated that he never was under the impression that it would apply anyway. W. Parler certainly agreed that the rules of the game should be defined at the outset and interested individuals ought to be given adequate notice. Chairman Zech thought this should be, and certainly will be, considered in the process.

C. J. Wylie indicated that the Comittee<did not consider the scope and level of detail of information required for design certification adequately defined in the draft NUREG. He suggested that it should be made clear in the Policy Statement that an applicant for a final design approval should be prepared to supply such other-information as l

\

i

1

! '320TH ACRS MINUTES 31 .

is customarily required by the NRC Staff to perfom'a Final Safety .- ,

Analysis Report (FSAR) review. The expansion and clarification of '

information requirements for an " essentially complete design" should

. have input from the principal cognizant NRC Staff reviewers and

+ varicus industry organizations experienced in such matters. Chairman Zech noted that the Staff appears to have different views on this

matter. C. J. Wylie indicated that basically they have provided a- ,

list of categories of information which should be provided and have .

, not gone into great detail as to definino what is in those categories. l

. Commissioner Bernthal asked for specific examples of the detailed ,

information that needs to be specified. C..J. Wylie noted that the  !

location of the standardized plant in a utility's grid has an effect i on the electrical supplies as well as offsite power. Plant systems 4

have to be tailored to those site-specific and grid-specific consid-

erations. C. Michelson suggested that small pipe drawings are neces-1 sary to assess potential pipe breaks. It is unclear from the stan-dardization document. One cannot do a good fire analysis without 4

i

?' electrical layouts also'. H. W. Lewis suggested that nothing is ever j built from complete plans, and you do have to define.what you mean by i " essentially complete." This will be very difficult because one is i talking about plants that presumably will be ordered 10 years after one accepts the final design document. Technology and components will change. He thought it foolish to freeze a design before construction starts.

J. C. Ebersole cited the enormous importance of some of the intricate

[

detail largely due to the absence of dedicated shutdown heat removal

systems. These details would influence the shutdown function.

i C. Michelson added that the point being made is that if one had a

!' hardened, dedicated decay heat removal system one might be able to relax the constraints on fire protection and pipe breaks. One would i need less understanding of the details erd still be able to pass j judgment on the acceptability. Comissioner Pernthal' indicated that it was his impression that the final design - certification would 1 involve a fairly complete set of drawings of the plant design covering everything but the site-specific items. C. J. Wylie wondered whether one needed to have that much detail in a standard design. In the case of the routing of cable, he suggested that one did not need all the cable-routing information but just the criteria by which one is going to route that cable. C. P. Siess didn't think the Staff would approve a standard design if they did not have ercugh information. He sug-i gested that the Staff has never taken an FSAR and icsued an approval

on the basis of it. There are usually numerous amendments and changes i to the original. W. Kerr asked if the Staff has defined the level of

detail as being that that .they would expect to see in an.FSAR at the operating Itcense stage. C. J. Wylie thought that it vas. Chairman Zech cautioned regarding specifying the completeness of the design down to nameplate data. He thought that it should be done as close to being specific as possible yet provide some judgment for site-specific considerations and other processes that' would lend themselves to j increased safety. He agreed that the level of detail is a very j important issue but thought that it would ultimately become a judgment i

call. J. C. Ebersole thought one should get approval at an initial 3 high level of detail and then have agreemen'ts on a tight level of i

320TH ACRS MINUTES 32 follow-up to obtain the remaining detail needed to finish the design. --

D. A. Ward disagreed with J. C. Ebersole that the safety robustness of a plant is dependent on a particular level of design detail. He suggested that J. C. Ebersole's contention was that plants of the present designs shculd not or cannot be standardized. He indicated that he was not sure of that fact. J. C. Ebersole stated that that

, was his thought; he couldn't envision a standardized plant, no less an advance LKR configuration, without due regard to such matters as dedicated shutdown heat removal. C. P. Siess did not share some of these concerns since the Staff will not issue a license for a standard plant but for a standard design. No one will be able to build such a plant and get an operating license without further review. He shared C. J. Wylie's concern that applicants submitting what they think an adequate FSAR may find that it is not sufficient for the standard design review.

D. Okrent suggested that he could not see that a very detailed design submission would preclude new technological improvements. Comission-er Bernthal noted that it was his impression that the design certifi-cation process would go a significant step beyond what is required in the FSAR. He noted that this understanding was not reflected in the Comittee's letter and he thought that possibly might be the reason for the ACRS concerns. He thought that it should be made clear that an FDA will require considerable additional input in order to generate a design certification.

D. A. Ward indicated that he was not as optinistic that one would have the best of both worlds as far as updating or improving a standard plant design. One should recognize the compromise one makes when there is a commitment to a concept or practice of standard plants.

One must admit that one is giving up the opportunity to make continu-ous or yearly improvements in safety systems as the technology per-mits. With the standard plant concept, one is comitting to making technical irprovements in large steps and, unless one is willing to

. give up the benefit of continuous improvements, a standardization policy is not the proper approach. Comissioner Berrthal noted that the Staff is comitted (Backfit Rule) not to require changes unless there is a discovery. M. W. Carbon shared D. A. Ward's views that unless one is willing to be standardized for some extended period of time without change it is not a standardized plant.

Chairman Zech indicated that the whole purpose of standardization is to produce a design in which one has the confidence that it has been proven as a reliable and safe design. The value of standardization is the gaining of reliability. The current state of affairs involves custer-built plants all over the U.S. with custom-butit regulations for these plants ard this is part of the problem. C. P. Siess cited 4

the experience of the French in the fact that their standard designs are backfitted on a consistent, systematic basis; when they have saved up enough improvenents, they go in and 'make them all at one time.

Comissioner Bernthal said that he believed the ACRS did not concur last summer with the Staff's view that standardized plants should be built on preapproved, predesignated sites and he wondered what the basis for that distinction was. F. J. Remick' indicated that the ACRS i

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320TH ACRS MINUTES 33 was not against using standard plants on preapprove, sites, . but --

i thought that the wording seemed to limit it only fnr tha, case. The

Comittee thought it was too restrictive. Comissioner Car. expressed i his opinion that the more detail that is supplied during t e review -

the more likely for everyone to agree on the licensing at th end of the process. He encouraged as detailed a review as possible.

VIII. Resolution of ACRS Coments on Shearon Harris Nuclear Power Plant (Opt. ')

[ Note

R. K. Major was the Designated Federal Official for this portion l

ofthemeeting.]

!; F. J. Remick indicated that the ACRS OL letter (dated January'16,1984) on the Shearon Harris Nuclear Power Plant contained a number of recom-mendations to the Staff and the Applicant, Carolina Power & Light (CP&L).

On December 10, 1986 L. S. Rubenstein, Director for PWR -Project Direc-torate No. 2, Division of PWR Licensing-A, responded to the ACRS letter. .

The Staff plans to go to the Comission during the first week of January l 1987 to request an operating license for the Shearon Harris plant.

3 F. J. Pemick explained that there were seven points in the ACRS letter of January 16, 1984. The ACRS asked to be kept informed regarding the operating experience of Westinghouse D-4 steam generators relative to-i tube degradation. He indicated that a letter from Westinghouse Electric 4

Corporation to CP&L discussed the inspection of one steam generator and i the finding of no evidence of degradation. He noted that a vibration test was done. As evidence of an improvement-in the Licensee's nuclear opera-l tion, the Staff included the results of the _last three SALP reports on

Shearon Harris. He noted that the September 25, 1986 Shearon Harris SALP l report covered 14 functional areas. The report stated that activities

, during this period were conducted in a very professional manner. Major i strengths were identified in five areas which were rated Category 1. No 1 areas were rated Category 3, and ratings improved in three areas. He l

, noted that the Staff considers that these reports provide substantial-i' written evidence of CP&L's improved performance, as discussed in the ACRS letter.

} The Staff response from L. Rubenstein indicated that a June 6, 1984 memo-i randum to the ACRS provided an initial response to the Comittee's con-

{ cerns regarding allegations contained in Mr. Eddleman's letter to the i ACRS. A December 10, 1986 letter provides additional information to f complement the June 6,1984 memorandum and provides the Staff's final j response to Mr. Eddleman's allegations.

i i

F. J. Remick indicated that the ACRS, in its January 16, 1984 letter,-

{ recomended that specific attention be given to assurance' of adeouate -

j - seismic capability of the emergency AC power supplies, DC power supplies, l and small components, such as actuators and instrument lines that are j important to the eccomplishment of safe shutdown and decay heat removal.

i Subsecuent to the recomendation made in the ACRS letter, an ensite audit j of the seismic oualification of selected safety-related electrical and l mechanical components was performed as reported in Supplements 3 and 4 to j the SER. The audit consisted of verification of installation, as well as a review of the cualification documents. Althoudh the items audited were i

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320TH ACRS MINUTES 34 l

\

l not identical .to those recomended by the ACRS, they did include relay -

cabinets, switchgear, control room cabinets, solid state protection system, etc. In all the items reviewed there was evidence of adequate seismic capability. R. Hernan indicated that ACRS o,uestions for the systems mentioned in the January 1984 letter will be picked up during the resolu- '

tion of USI A-46 and other generic issues.

F. J. Remick indicated that the Staff believes that the results of its review of the chilled water system, reouested by the ACRS, is in Supplement 4 to the SER. He noted that the Subcommittee on Air Systems held a de-tailed discussion of chilled water systems during which C. Michelson asked how the Staff reviews such systems. At the conclusion of the meeting C. Michelson reriained concerned about the novel arrangement at Shearon Harris. F. J. Mick noted that the Committee asked for a detailed review but the Staff review was not detailed. C. Michelson noted that chilled water systems are a generic issue that is not being handled well by the NRC Staff. Nevertheless, the question in the ACRS letter should not hold up the OL recommendation. He recommended that the Comittee accept the Staff's explanation and treat the issue of chilled water systems generical-ly. ,

l F. J. Remick noted that, because of the nonoptimum orientation of the turbine relative to vital components at Shearon Harris, the ACRS recom-mended in its 1984 letter that a structured test program for evaluating overspeed protection of the turbine he prepared and submitted to the NRC '

Staff for review and approval before full power operation. He indicated that CP&L has submitted copies of its startup, preoperational, and mainte-nance test procedures used to demonstrate and maintain proper calibration of the main turbine mechanical and electrical overspeed trips. The Staff has reviewed these procedures and has concluded that the structured test program for evaluating turbine overspeed protection is acceptable.

F. J. Remick recomended that the ACRS not maintain any objections to the Shearon Harris OL application. The Committee concurred in his recommenda-tion.

IX. Report of TVA Manacement (0 pen)

[ Note: R. P. Savio was the Designated Federal Official for this portion of the meeting.]

C. J. Wylie indicated that he recently attended, as Chairman of the TVA Organizational Issues Subcommittee, a briefing of the NRC Staff by TVA management. He indicated that the situation is ouite disturbing and, because time did not permit any further discussion, he distributed TVA's presentation slides (see Appendix XII).

X. Radwaste Management and Disposal (0 pen)

[0. S. Merrill was the Designated Federal Official for this portio'n of the meeting.] -

D. k', Moeller indicated that a meeting of the Waste Manacement Subccrmittee was held on December 4-5, 1986 to review several nuclear waste management topics. He indicated that the first topic' discussed during the

w .

320TH ACRS MINUTES 35 Subcommittee meeting was the status of NRC's work on-alternative disposal -

methods for shallow land burial. He indicated that Sections 10 CFR 61.50, 61.51, 61.52, and 61.53 are directly applicable to below-ground vaults, above-ground vaults, earth-mounted concrete bunkers, and augered hole / shaft disposal. A mined cavity as an alternate disposal method is sufficiently different from other near-surface disposal options that it is not covered under these sections of Part 61, but is licensable under Part 61.23 on a case-by-case basis.

D. W. Moeller explained that EG&G, Idaho at the Idaho National Engineering Laboratory has been perfoming studies on the safety assessment of alterna-tives to shallow land burial under an NRC contract. The studies discuss the functions of four design components and weigh the relative importance of each by the use of weighting factors. The four components were cover, structure, fill, and container. He noted that the Subecmmittee was not very impressed with the contractor's work. The study concluded that the cover was the most important design component, although D. W. Moeller thought that the container might more appropriately be considered most important. The studies ignored the characteristics of the site and didn't consider cracks in the vault.

D. W. Moeller discussed the current status of State efforts to meet the mandates of the Low Level Radioactive Waste Policy Amendments Act with regard to providing for the disposal of low-level radioactive waste. He noted that many State compacts have banned shallow land burial as a waste disposal option; the NRC Staff believes that this approach is based on misunderstandings about shallow land burial. There is talk of giving incentives to induce a State to accept low-level waste by the taxing of radioactive waste generators and using the money raised for disposal site operations. Host States have been determined for the waste compacts and the compacts are meeting periodically.

D. W. Moeller indicated that SECY-86-92 was issued March 20, 1986 and the final rule was published in the Federal Recister on July 30,1986, as a rulemaking to conform Part 60 to the EPA Standard. .It contains proposed amendments to eliminate inconsistencies between Part 60 and the EPA Stan-dard for High-Level Waste Geologic Repositories. He noted that there has been close cooperation between the EPA and the NRC. Comments on the proposed rule have been on the substance of the EPA Standard rather than the NRC's proposed adoption of it. EPA does not require monitoring of the

• ite after the high-level waste repository is closed because it could ompromise the engineered barriers to intrusion. The EPA wants any8 intru-fon, with a calculated probability of occurrence of less than 10 to be

1. copped. F. J. Remick noted the controversy around the terminology "rea-sonable assurance" used by the NRC Staff. EPA wishes to use the teminolo-gy " reasonable expectation.'

D. W.' Moeller explained that the subject of assessing compliance with the EPA Standard has been under study by a group at the Sandia National Labora-tories. A draft of their report, Assessino Compliance with EPA Hich-level Weste Standard: An Overview, (NUREG/CP.-4510), has been undergoing a review 1 by the fRSS/ Division of Waste Management Staff since its receipt in Febru-ary 1986. It is anticipated that the ACRS will provide comments to the Staff on this document prior to its final publication. D. W. Moeller

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I 320TH ACRS MINUTES 36

' indicated that. he would like D. Okrent or a suitable ACRS consultant -to --

i examine the report because of its use of the probabilistic approach. He

. noted that the EPA Standard is also probabilistically based and that the report contains development and screening of scenarios, evaluation of ,

consequence assessment, sensitivity and uncertainty analyses, and regulato-

, ry compliance assessment. D. Okrent indicated that a preliminary look at i the document indicates that the methodology review in the document does not make robust judgments. H. W. Lewis wondered why the EPA picked 10,000 years as the period of time for the high-level waste to remain isolated from the environment. D. W. Moeller suggested that it is normally con-sidered the period of time between ice ages,.

D. W. Moeller indicated that the NRC has reviewed DOE's final environmental assessments and the ED0 is planning to send the Staff's coments to the

. Comissioners. Final NRC comments are due-by December 16. The Staff has l' limited itself to major issues such as seismology, tectonics, . hydrology, and geology. ACRS consultant. M. Trifunac has agreed to review the NRC

, work. He noted that the NRC has played up the uncertainties in scenarios an8 models. He also noted that Sweden has gone to an international review of the performance of its high-level waste activities. D. W. Moeller noted

, that the NRC will ask for an independent international review of the EPA Final Environmental Assessments to be done by NEA, OCED, and the IAEA. He l thought that the Staff's primary objective should be to direct its atten-tion to changes that the Department of Ener Characterization Plans (yet to be prepared)gy can incorporate rather than seekinginto the Site additional L changes to the existing Final Environmental Assessments. He noted that the Subcomittee is comfortable with the Staff's review except for the treat-ment of seismicity and tectonics. D. Okrent asked if the subject of

climatology over the 10,000-year period of isolation for the waste was discussed by the Staff. D. W. Moeller indicated that flooding and hydrolo-gy were extensively studied in this context.

l D. W. Moeller indicated that the Staff identified several problem areas in its environmental assessments. These dealt with rock bursts within the

[' basalt medium 'and brine migration to a heat source within the salt medium. .

The Staff called on DOE to use greater knowledge in formulation of its Site

Characterization Plans. He noted that there were scathing coments on the Hanford site and he wondered whether they would be considered in the Site Characterization. It was indicated that a Comittee report on the Environ-mental Assessments would be appropriate and one would be put on the table on Saturday. However, D. W. Moeller said he thought it not appropriate to go into specific details regarding the Staff's coments, but to make the

, letter of a general character. '

I XI. Hanford Reactor Temporarily Closed (0 pen) l l

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

!' H. W. Lewis explained that he participated in a recent review of the Hanford N-reactor. The Department of Enercy (DOE) had commissioned a six-member panel of independent consultants to review the safety of the

N-reactor. He indicated that he was one of* two of the six-member panel that suggested that the N-reactor be ^1osed imme'diately and permanently.

320TH ACRS MINUTES 37 Three of the panel members recomended that the N-reactor be allowed to --

opera'te for an additional two years after the implementation of certain safety modifications. He noted that there is an upcoming refueling outage for the N-reactor and in a December 12 news conference DOE announced the shutdown of the reactor for an additional six months to make all of the modifications recomended by the panel. DOE is doing two PRA studies on the Hanford N-reactor and preliminary reports will be available in April 1987.

XII. Nuclear Plant Security (Closed)

[ Note: R. F. Fraley was the Designated Federal Official for th'is portion ,

ofthemeeting.]

R. F. Burnett, Director, Division of Safeguards, NMSS, briefed the ACRS full Committee on security provisions at nuclear facilities. The text of this discussion is contained in a Supplement.

XIII. Executive Sessions

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

A. Subcommittee Assignments (0 pen)

1. Spent Fuel Storage During the report of the ACRS Subcomittee on Spent Fuel Storage, C. P. Siess, Subcommittee Chairman, reported on and recommended that the ACRS need not be involved in the review of- licensing ISFSIs or reracking and consolidation of fuel rods at existing nuclear power plants. The ACRS should, however, be involved in the Monitored Retrievable Storage (MRS) facility review. The Comittee expects to take up the proposed changes to' 10 CFR Part 72 with regard to the storage of spent fuel rods and other high-level radioactive wastes in onsite er offsite ISFSIs, including MRS-type facilities after public coments have been reviewed by the NRC (expected by the April or May 1987 ACRS meeting).

B. Reports, letters, and Memoranda (0 pen)

1. Proposed BWR Mark I Containment Recuirements for Severe Accidents The Committee prepared a report to the Comissioners regarding the proposed new recuirements for BWR Mark I containments in regard to severe accidents.

2. ACRS Report on Proposed Policy Statement on Deferred Plants l l

The Comittee prepared a report to the Corrmissioners nn the i NRC Staff's proposal for a Comission policy statement on ' i deferred plants (SECY-86-359). .

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320TH ACRS MINUTES 38

3. Application of GDC-4 to Component Supports --

The Committee prepared a memorandum to the EDO reaarding the NRC Staff's interpretation of GDC-4 as it applies to the '

design of supports for components in existing plants taking into account elimination of the dynamic effects of pipe l ruptures. '

4 ACRS Coments on the NRC Staff Review of DOE's Final Environmental Assessments of HLW Repository Sites The Comittee prepared a letter to the ED0 concurring in the NRC Staff approach for review of DOE's Final Environ-mental Assessments of HLW repository sites.

5. ACRS Action on Proposed Regulatory Guide XXXX, " Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," Final Guide Draft, Dated June 1986 The Committee prepared a memorandum to the EDO which recom-mends issuance of the final Regulatory Guide XXXX, " Format and Content of Plant-Specific Pressurized Thermal Shock

, Safety Analysis Reports for Pressurized Water Reactors."

6. ACRS Action on the Proposed Revision 3 to Reculatory Guide 1.63 The Coninittee prepared a letter to the E00 concurring in the regulatory position of proposed Revision 3 of Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants."

i, 7. Policy for Nuclear Power Plant License Renewal The Committee prepared a memorandum to the.ED0 expressing Committee interest in participation in the Development of a Policy for Nuclear Power Plant License Renewal beyond the 40-year lifetime of current licenses.

8. Annual Report on the Reactor Safety Research Program The Committee prepared a letter to the President of the Senate and the Speaker of the Pouse proposing alternate means to fulfill its statutory requirement that the ACRS provide an annual report to the Congress on the NPC Safety Research Prograr. Instead of writing detailed annual reports on the kRC Safety Research Program, the ACRS pro-posed to write more focused reports on fewer important issues. -

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9. .ACRS Coments on the Implications of the Accident at - . .

. Chernobyl, Unit 4 The Comittee discussed a draft report to the Comissioners on its consideration of lessons learned from the accident at

' Chernobyl, Unit 4, and the' implications for - U.S. nuclear power plants. Time did not pennit full consideration of this report, and it will be scheduled for further discussion during the 321st ACRS meeting (January 1987).

10. Improved LWRs .

.The Committee members continued their discussion of a proposed ACRS report to NRC regarding the characteristics of improved LWRs. Time did not permit completion of this report.- Additional time to complete this report will be scheduled during the January (321st) ACRS meeting.

C. Generic Issues (0 pen)

1. Addi tional Safety Reouirement- in the Federal Republic of Germany .

D. Okrent noted that the Federal Republic of Germany has recently expressed concern regarding the ' ability of their large dry containments to withstand a hydrogen burn. They are considering a recuirement for their containments to provide the capability to handle. hydrogen generation during accidents. H. Alderman, ACRS Staff Engineer, was assigned ~

to investigate the matter. He will be assisted by Moni De, ACRS Fellow.

D. Future Agenda (0 pen)

1. Future Agenda The Committee agreed on tentative agenda itects for the 321st ACRS meeting, January 8-10,1986(seeAppendixII).

2. Future Subcomittee Activities A schedule of future subcomittee activities was distributed to members (see Appendix III).

E. Election of Officers (Closed) f Contained in Supplement.

! Reappointment of New Members (Closed)

Contained in Supplement.

G. Accointment of New Menbers (Closed)

Contained in Supplement.

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320TH ACRS MINUTES- 40 H. Retirement of ACRS Member Emeritus ._

H. Etherington completed his last ACRS meeting as Member Emeri-tus. -He will be available to consult in areas of his expertise such as mechanical and metallurgical engineering and is available to attend ACRS meetings when specifically requested to partici-pate regarding specific issues.

I. Waste Management and Disposal During D. W. Moe11er's report of the December 4-5, 1986 meeting of the Subcommittee on Waste Management, a short discussion of a NUREG report prepared by Sandia National Laboratories took place.

The Comittee decided that this report, Assessing Compliance with .

the EPA High-level Waste Standard: An UTerview (NUREG/CR-4510),

should be reviewed by ACRS consultant W. E. Kastenberg before the Comittee takes a position regarding its merits. The NUREG report has been sent to Dr. W. E. Kastenberg, UCLA, for his review.

J. Wingspread International Conference on Reactor Safety .

D. W. Moeller has provided a long and a short summary of th'e highlights of the Wingspread International Conference on Reactor Safety held on October 20-22, 1986 in Racine, Wisconsin. The ACRS Executive Director, R. F. Fraley, requested that any member comments be forwarded to T. G. McCreless, ACRS Assistant Execu-tive Director, as soon as possible for incorporation into the meeting summaries so they can be promulgated for public release (short version) and distributed to meeting participants as a -

detailed record of the meeting (long revision).

The 320th ACRS I4eeting was adjourned at 2:30 p.m., Saturday, December 13, 1986.

e 9

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APPEilDICES TO MIllVTES OF THE 320TH ACRS MEETIllG DECEMBER 11-13, 1986 06RS-M9/

O O

I APPENDIX I

)

NRC' ATTENDEES ATTENDEES 9 320TH ACRS MEETING DECEMBER 11-13, 1986 Thursday, December 11, 1986 0FFICE OF NUCLEAR REACTOR REGULATION R. W. Hernan D. Scaletti D. D. Lynch, Jr.

J. Donohew 0FFICE OF INSPECTION AND ENFORCEMENT A. E. Rosenthal R. W. Woodruff E. L. Jordan H. A. Bailey E. Weiss O

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A-I

I PUBLIC ATTENDEES

~320TH ACRS MEETING DECEMBER 11-13, 1986 Thursday, December 11, 1986 John Trotter, NUS Corporation A. Wyche, SERCH Licensing-Bechtel D. Airozo, McGraw Hill J. Verdiles, Framatone J. Berga, EPRI D. Weaver, McGraw Hill N. S. Grifford, General Electric B. Jordan, McGraw-Hill T. Imai, JEPIC S. Reines, JEPIC S. Naleagawa, Kansai B. Jordan, McGraw Hill R. Turner, IEAL s

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l NRC ATTENDEES l v

! 320TH ACRS MEETING l l DECEMBER 11-13, 1986

i j Friday, December 12, 1986 i 0FFICE OF NUCLEAR REACTOR REGULATION i i

R. W. Hernan  !

l T. S. Michaels  !

i S. E. Bryan i j E. Butcher i -H. Richings  !

j

. B. Buckley l h

0FFICE OF INTERNATIONAL PROGRAMS ,

4  :

! H. Faulkner l l

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b PUBLIC ATTENDEES

! 320TH ACRS MEETING DECEMBER 11-13, 1986 b

Friday, December 12, 1986 K. Arn, SERCH Licensing P. F. Riehm, KMC, Inc.

M. G. Gagliardi, CP&L/Ebasco Services S. B. Presgrove, NUS Corp.

C. Chester,.IEAL H. M.Fontecilla, VA Power

W. H. Bamford, Westinghouse B. J. Snyder, Brown Bovers, Nollear .

! M. Beaumont, Westinghouse R. Borsum, Babcock & Wilcox L. I. Loflin, Carolina Power & Light R. Sweeney, NHY/Ebasco L. Connor, DSA K. Unnerstall, Neuman & Holtzinger Colette Power, Embassy of Japan -

T. Imaz, JEPIC i D. Tibbits, Carolina Power & Light 4

D. C. McCarthy, Carolina Power & Light i s W. Washington, Darolina Power & Light i R. L. Turner, RLT M. Webb, Carolina Power & Light l' T. Pickens, NSP S. Murphy, NIRS

^

O. Batum, Georgia P'ower Company A. B. Cutter, California Power & Light J. F. Lang, EPRI

L. S. Gifford, General Electric 1

V. S. Boyer, PECO W. Koch, Small Newspaper Group W. H. rasin, Duke Power Company S. B. Maia, TEPC0 A. Taucher, Chubu EPC j

F. Warner, The (Pottstown, Pa.) Mercury '

B. Jordan, McGraw Hill l D. Weaver, McGraw Hill J

J. Kerdiles, Frametone F. Warner, The Mercury J

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APPENDIX II FUTURE AGENDA X APPENDIX A I

FUTURE AGENDA JANUARY ACRS MEETING NRC Regulatory Guides --

Regulatory Guide 1.35, Revision 3, Inservice Inspection Deferred in-of Ungrouted Tencons in Prestressed Concrete Containments definitely Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containment 1

Radiation Damage -- ACRS comments requested regarding Deferred to d

Regulatory Guide 1.99, Revision 2, Radiation Damage Feb./ March Meeting with EDO -- Briefing / discussion regarding NRC Staff 1 hr reorganization l BWR Pipe Crack Guidance -- ACRS comments requested regarding Deferred to proposed incorporation of public comments into NUREG-0313, February Revision 3 regarding monitoring, repair, etc. of BWR pipe cracking (medium priority)

Safety Goal Policy Im)1ementation -- Status report / briefing li hrs regarding proposed NR: implementation plan

\ Systems Interactions -- ACRS response requested regarding i hr proposed resolution of ACRS May 13, 1986 consnents regarding proposed resolution of USI A-17, Systems Interactions in Nuclear Plants Review of Advanced BWR -- Briefing / discussion regarding 2 hrs ACR5 participation in review of this advanced standard plant  ;

Improved Light Water Reactors -- Complete report to NRC As needed to regarding characteristics of improved LWRs complete i Meeting with the Director, NRR -- Discuss items of mutual I hr interest j Bypassing of the Suppression Pool in Mark I and Mark II 3/4 hr  !

Containments -- Briefing regarding failure of SRV discharge line in the airspace of Mark I and Mark II containment utwell Safety Research Proc ram-- Preparation of ACRS " report" to the As needed the proposed NRC safety research U.S.

program Congress regarc ing(tentative) for FY 1988/89 l Generic Issue 61 -- C. Michelson recommendation on the i hr prioritization of Issue 61, "SRV Line Break Inside the BWR Wetwell Airspace of Mark I and Park II Containments" l I

Station Blackout -- ACRS consnents requested regarding Defermd to l proposed NRC rule on station blackout March / April i

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320TH ACTIONS AND AGREEMENTS i

i j Standard Plant Improvements -- Proposed report by Deferred until i

J. C. Ebersole (319th ACR5 meeting) report on - 7

' improved LWRs is " complete" General Desis n Criterion 4 ' Environmental and Missile Deferred to l Design -- ACES coments regarding proposed changes in GDC-4 Feb./ March regarding design of pipe whip restraints in nuclear plants Safety of B&W Water Reactors -- ACRS comments requested Deferred to regarding NRC Staff review of BWOG evaluation of the March long-term safety of B&W PWRs Seabrook Nuclear Station -- ACRS coments requested regarding Deferred to the basis for reduced emergency planning distances February 1

Implementation of Severe Accident Policy -- ACRS coments Deferred to '

requested regarding proposed NRC Staff plan for implementation Feb./ March of the NRC Severe Accident Policy Statement Performance Indicators for Operating Reactors -- Status report Deferred to by NRC Staff (SECY 86-327 received October 29,1986) April Spent Fuel Storage Facilities -- ACRS coments regarding the Deferred to -

storage of spent fuel rods and other high-level radioactive April /May wastes in onsite or offsite ISFSIs, including MRS-type facilities j Topics for Februa'ry Meeting with NRC Comissioners -- Discuss I hr (tenta-proposed topics for February meeting with NRC Comissioners tive)

Review of TVA Plant Restarts -- Review proposed TVA corporate Deferred to "get-well" plan February 1

i ACRS Subcomittee Activities Instrumentation and Control Systems -- Report regarding i hr

. meeting on December 18, 1986 to consider the effects of adverse environmental conditions on the perfomance j of solid state components in nuclear plants L i

{ Extreme External Phenomena -- Diablo Canyon long-tem i hr  !

j seismic program Regional Activities -- Report of December 2, 1986 meeting i hr

, at Region III TVA Management -- Report of briefing on November 21, 1986 & hr

! Severe Accident Policy - Report of meeting on December 19, .

l 1986 regarding NRC 5taff proposed implementation plan  !

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l gg DEC 131986 APPENDIX III fL ACRS SUBCOMMITTEE MEETINGS ACRS SUBC0tiftITTEE ttEETINGS Instrumentation and Control Systems, December 18, 1986, 1717 H Street, NW, Washington, DC (El-Zeftawy), 8:30 A.M., Room 1046. The Subcommittee will discuss the effect of adverse conditions such as high temperature on

. solid-state components in nuclear power plants. Attendance by the follow-ing is anticipated, and reservations have been made at the hotels indicated for the night of December 17:

Mr. Ebersole DAYS INN Dr. Moeller LOMBARDY Dr. Kerr LOMBARDY Mr. Wylie DAYS INN Mr. Michelson DAYS INN Dr. Lipinski NONE Severe Accidents, December 19, 1986, 1717 H Street, NW, Washington, DC (Fouston), 8:30 A.M., Room 1046. The Subcommittee will discuss the NRR Implementation Plans for the Severe Accident Policy Statement regarding Individual Plant Examinations (IPE) for Existing Plants. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of December 18:

Dr. Kerr LOMBARDY Dr. Catton DUPONT PLAZA Dr. Carbon STATE PLAZA Dr. Corradini ANTHONY Dr. Mark LOMBARDY Mr. Davis HOLIDAY TNN Dr. Shewmon NONE Dr. Lee ANTHONY Safety Research Program (Closed), January 7, 1987, 1717 H Street, NW, b Washington,DC (Duraiswamy), 8:30 A.M., Room 1167. The Subcommittee will discuss the following and gather information for use by the ACRS in its preparation of the annual report to the Congress on the NRC Safety Research Program and budget for FY 1988: (1) changes made to the FY 1988 NRC Safety Research Program and Budget since the December 10, 1986 meeting; (2) final OMB Mark and the impact of the OMB-proposed reductions on the continuing and proposed research contracts, and (3) Draft 0 of the ACRS report to the Congress. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of January 6:

Dr. Siess ANTHONY Dr. Remick* NONE Dr. Carbon

• STATE PLAZA Dr. Shewmon* NONE Dr. Kerr* LOMBARDY Mr. Ward

• NONE Dr. Mark LOMBAP.DY Mr. Wylie* NONE Mr. Michelson* DAYS INN Dr. Moeller LOMBARDY Denotesparttime--sharedwithGEReactors(ABWR) meeting.

O m A-7

4 Joint General Electric Reactors (ABWR)/ Safety Philosophy, Technology, and l

Criteria, January 7,1987,1717H5treet,NW, Washington,DC(Major /5avfo),

9:00 A.M., Room 1046. The GE Reactors Subcommittee will begin its review of the ABWR. This will be a preliminary session to explore the status of

this project and to be briefed on efforts regarding a licensing basis

agreement between GE and the NRC. A current description of the ABWR is sought as well as schedules from GE and the Staff. The SPTC Subcommittee will review the status of the NRC Staff's work on the Safety Goal Policy i and on USI A-17. " Systems Interactions on Nuclear Power Plants." Attend-

! ance by the following is anticipated, and reservations have been made at the hotels indicated fnr the night of January 6:

Dr. Okrent ANTHONY Dr. Remick* NONE Dr. Carbon

• STATE PLAZA Dr. Shewmon* NONE

. Mr. Ebersole DAYS INN Mr. Werd* NONE l Dr. Kerr* LOMBARDY Mr. Wylie* DAYS INN Dr. Lewis (tent.) HYATT i Mr. Michelson* DAYS INN J

• Denotes part time -- shared with Safety Research meeting.

Planning Subcommittee (tenta'.ive) (Closed), January 7, 1987, 1717 H Street,

! NW, Washington, DC (Fraley), Time and Room to be determined. The Subcom- '

, mittee will discuss categorization and prioritization of ACRS activities. '

Attendance by the following is anticipated:

ACRS Chairman Member-at-Large ACRS Vice-Chairman R. Fraley

, 321st ACRS Meeting, January 8-10, 1987, Washington, DC, Room 1046.

Regulatory Policies and Practices, January 14, 1987, 1717 H Street, NW, Washington, DC (Quittschreiber), 8:30 A.M., Room 1046. The Subcommittee 1

will begin its current review of the nuclear plant regulatory process.

Attendance by the following is anticipated, and reservations have been made

, at the hotels indicated for the night of January 13:

Dr. Lewis HYATT Dr. Siess ANTHONY Dr. Kerr LOMBARDY Mr. Ward NONE Dr. Remick NONE Mr. Wylie DAYS INN l

l i

l Metal Components, January 15 and/or 16, 1987, 1717 H Street, NW, i Washington, DC (Igne), 8:30 A.M., Room 1046. The Subcommittee will review:

(1) hear a status mport of the Whipjet program (application of broad scope GDC-4 criteria) as applied to lead plant Beaver Valley Unit 2; and (2) review public comments on NUREG-0313, Revision 2 (long range fix for BWR-IGSCCproblems),(3) Reg. Guide 1.99,Rev.2,and(4)otherrelated matters. Attendance by the following is anticipated:

Dr. Shewmon Dr. Bush Mr. Michelson Dr. Kassner Mr. Ward Mr. Odette Mr. Bender Mr. Rodabaugh Standardization of Nuclear Facilities, January 21, 1987, 1717 H Street, NW, Washington, DC, (Alderman), 8:30 A.M., Room 1046. The Subcomittee will review the NRC evaluation of Chapter I ("Overall Requirements") of the EPRI Advanced Light Water Reactor Program. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of January 20:

Mr. Wylie DAYS INN Mr. Michelson DAYS INN Mr. Ebersole DAYS INN Mr. Reed NONE Structural Engineering, January 21 and 22, 1987, at the AMFAC Hotel, 2910 YaleBlvd.,SE, Albuquerque,NM(Igne),8:00A.M. The Subcommittee will review containment integrity and Category I structures, programs, and test facilities. Attendance by the following is anticipated, and reservations have been made at the AMFAC Hotel (telephone #505/843-7000) for the nights of January 20 and 21:

Dr. Siess Dr. Mark Dr. Carbon Pr. Bender Decay Heat Removal Systems January 22, 1987, 1717 H Street, NW, Washing-ton, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of the NRR Resolution Position for USI A-45. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of January 21:

S Mr. Ward NONE Mr. Wylie DAYS INN Mr. Ebersole DAYS INN Dr. Catton DUPONT PLAZA Mr. Michelson DAYS INN Mr. Davis HOLIDAY INN Mr. Reed NONE O

A4

i Joint' Occupational and Environmental Protection Systems / Severe Accidents /

5eabrook, January 29,1987,1717 H 5treet, NW, Washington, DC -

(Igne/ Houston / Major), 8:30 A.M., Room 1046. The Subcomm1ttees will review

. Brookhaven National Laboratory's Draft report of the Seabrook Emergency i Planning Sensitivity Study. Attendance by the Following is anticipated, and reservations have been made at the hotels indicated for the night of l January 28:

4 Dr. Moeller LOMBARDY Dr. Remick NONE Dr. Kerr (P.M. only)LOMBARDY Dr. Siess ANTHONY Dr. Mark LOMBARDY Corradini NONE

Naval Reactors (Closed), January 30, 1987, National Center #2 Building, I ' Crystal City, VA (Boehnert), 8

30 A.M. The Subcommittee will review the j Naval Reactor Moored Training Ship Project. Attendance by the following is ,

j anticipated, and reservations have been made at the hotels indicated for

the night of January 29

!

1 Dr. Kerr LOMBARDY Dr. Remick NONE

Dr. Lewis HYATT Mr. Ward NONE l'

Advanced Reactor Designs, February 4, 1987, 1717 H Street, NW, Washincton, DC adva(El-Zeftawy), 8:30 A.M., regarding Room 1046. The Subcommittee will review DDE

, nced non-LWR designs the use of proven technology and stan-i dardization. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of February 3:

1 Dr. Carbon STATE PLAZA Dr. Okrent ANTHONY

! Dr. Mark LOMBARDY Dr. Remick NONE 4

Mr. Michelson DAYS INN Dr. Siess ANTHONY 1 Dr. Kerr LOMBARDY l i Waste Panagement, February 12-13, 1987, 1717 H Street, NW, Washington, DC l

(Merrill), 8:30 A.M., Room 1046. The Subcommittee will review several

' pertinent nuclear waste management topics, which are to be detemined during an agenda planning session with the NMSS and RES Staffs on January <

21, 1987. Lodging will be announced later. Attendance by the following is

! anticipated:

T j Dr. Moeller Mr. Reed

. Dr. Carbon Dr. Remick i Dr. Kerr Dr. Shewmon .

Dr. Mark I I

) ,

i i

O 4-10 l

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_ , ,,_ _ _,_.-__ -.,,.. ._ ,_ __-.._ ,_ ._ _ ~ _.._____ --_,. _ ,_--,-- _ ,,.. __ - ~

Human Factors, February 18, 1987, 1717 H Street, NW, Washington,~ DC -

( Aldennan) . The Subcommittee will review " Safety Conscience" cnncept at utilities. Lodging will be announced later. . Attendance by the following is anticipated:

Dr. Remick Mr. Michelson Mr. Ebersole Mr. Ward Dr. Kerr Mr. Wylie AC/DC Power Systems Reliability, Date to be determined (February / March).

Washincton, DC (El-Zeftawy). The Subcommittee will review the proposed Station Blackout rule. Attendance by the following is anticipated:

Dr. Kerr Dr. Lewis Mr. Ebersole Mr. Wylie Regional and I&E Programs, Date to be detemined (mid-March), Washington, DC (Boehnert). The Subcommittee will continue its review of the activities of the Office of Inspection and Enforcement. Attendance by the following is anticipated:

Dr. Remick 'Mr.~ Reed Mr. Michelson Mr. Ward Dr. Moeller Mr. Wylie Thermal Hydraulic Phenomena, Date to be determined (2-day meeting, March /-

April), INEL, Idaho Falls, ID. The Subcomittee will review: (1) the FinalECCSRuleandassociateddocumentation,and(2)TICactivitiesat INEL. Attendance by the following is anticipated:

Mr. Michelson Dr. Catton Mr. Ebersole Dr.:Schrock Dr. Kerr Dr. Sullivan Mr. Reed Dr. Tien Mr. Ward Seabrook Unit 1, Date to be detemined (winter), Washington, DC (Major).

The Subconnittee will review the application for a full power operating license for Seabrook Unit 1. Attendance by the following .is anticipated:

Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson Regional and I&E Programs, Date to be detemined (May), Region IV, Arling-ton, TX. The Subcommittee will continue its review of the activities under the control of the Region IV Office. Attendance by the following is anti-cipated:

Dr. Remick Mr. Reed Mr. Michelson Nr. Ward Dr. Moeller Mr. Wylie A-tt

NRR STAFF PRESENTATION TO THE  !

D ACRS APPEt4 DIX IV CHERT 10BYL EVALUATION

\

4 i

SUBJECT:

CHERNOBYL EVALUATION

! DATE: 12/12/ee PRESENTER: BRIAN SHERON

'O PRESENTER'S TITLE / BRANCH /DIV: DEPUTY DIRECTOR DIVISION OF SAFETY REVIEW & OVERSIGHT NRR PRESENTER'S NRC TEL. NO.: 492-73o3 i

SUBCOMMITTEE:

l FOR INTERNAL ACRS USE ONLY A -l;L- l

4s O

4 STATUS OF FACT-FINDING REPORT i

DRAFT REPORT PREPARED PRIOR TO VIENNA MEETING TO AID U.S.

DELEGATION INFORMATION PROVIDED BY SOVIETS AT VIENNA MEETING WAS

! SUBSTANTIAL AND WILL PROVIDE PRIMARY BASIS FOR FINAL FACTUAL i REPORT iO

FINAL FACTUAL REPORT WILL REFERENCE AND DRAW FROM SOVIET

! PLUS INSAG REPORTS AS APPRUPRIATE t

NO NEW INFORMATION RECEIVED SINCE VIENNA 1

I i 1

O A-/3

l i

,V O STATUS LHAPTER TITLF LEAD AGENCY I INTRODUCTION NRC 2

SUMMARY

NRC 3 PLANT DESIGN DESCRIPTION DOE 4 SAFETY ANALYSIS EPRI 5 t.CCIDENI SCENARIO NRC 6 ROLE OF OPERATING PERSONNEL INPO 7 SOURCE TERti AND ATMUSPHERIC DISPERSION AND TRANSPORT NRC 8 EMERGENCY PREPAREDNESS FEMA 9 ENVIRONMENTAL CONSEQUENCES EPA SCHEDULE ALL DRAFT CHAPTERS PREPARED AND.SENT TO NRC FRELIMINARY DRAFI FACTUAL REPORT ISSUED TO PARTICIPANTS 11/17 MEETING TO REVIEW REPORT WAS HELD 11/18 EXPECT TO ISSUE FINAL DRAFT FACTUAL REPORT BY 12/17 MEETING TO FINALIZE THE REPORI IS SCHEDULED FOR 12/19 EXPECT FINAL TEXT TO BE COMPLETED BY END OF MONTH i

,O A-It l

l i

i O

CHERN0BYL IMPLICATIONS ASSESSMENT APPROACH: IDENTIFY CANDIDATE ISSUES

!' ASSESSJ DRAW CONCLUSIONS, RECOMMENDATIONS (ACTION, DROP, OR FURTHER STUDY: NEW PROPOSED ACTIONS VIA j GENERIC ISSUES PROCESS)

SEEK BROAD, OVERALL CONCLUSIONS AS MAIN FOCUS (DETAILED 1

ISSUE ANALYSES ARE SUBSIDIARY)

. ERIEF REPORT (NOT MUCH OVER 100 PP.), WITH

SUMMARY

l (PERHAPS S RP.)

l REPORT: OVERALL ASSESSMENT; INCLUDE ASSESSMENT OF EACH ISSUE (STATEMENT OF ISSUE, PRESENT PRACTICE, WORK IN j PROGRESS, ASSESSMENT, CONCLUSIONS AND RECOMMENDATIONS).

A-6 i 1

1 1

- - . - --,----,nn---c, ,,,.,.,,..__,,nn, c-n_-,-,, ,-,.. --,,,,,,,,,,,._,,,,n-__~.,,m,-._,,.- -

-n-.-,--,--.~--,., ,~ww._n- -w,.,- ,

.A ISSUES I. OPERATIONS (ADMINISTRATIVE CONTROLS) l 1.1 ADMINISTRATIVE CONTROLS TO ASSURE THAT PROCEDURES ARE FOLLOWED, AND PROCEDURE ADEQUACY.

! I.2 APPROVAL OF TESTS-AND OTHER UNUSUAL OPERATIONS

I.3 BYPASSING SAFETY SYSTEMS 1.4 AVAILABILITY OF ENGINEERED SAFETY FEATURES I.5 OPERATOR ATTITUDES TOWARD SAFETY I6 MANAGEMENT SYSTEMS 1.7 ACCIDENT MANAGEMENT II. DESIGN II.1 REACTIVITY ACCIDENTS 11.2 ACCIDENTS AT LOW POWER AND WHEN SHUT DOWN -

i II.3 MULTIPLE UNIT PROTECTION II 4 FIREF

! 111. CONTAINMENT 111.1 DEYOND DBA CAPABILITIES III.2 FILTERED VENTING IV. EMERGENCY PLANNING IV.1 EPZ SIZE IV.2 MEDICAL SERVICES IV.3 INGESTION PATHWAY MEASURES  ;

IV.4 DECONTAMINATION AND RELOCATION V. SEVERE ACCIDENT PHENOMENA V.1 SOURCE TERMS 1 V.2 STEAM EXPLOSIONS V.3 COMBUSTIBLE GAS j VI. GRAPHITE MODERATED REACTORS A-A

OVERALL CONCLUSIONS (INTERIM)

O

1. NO IMMEDIATE REGULATORY ACTION NEEDED:

! U.S. REACTORS PROTECTED AGAINST CHERNOBYL-LIKE EVENTS BY NUCLEAR DESIGN, SHUTDOWN MARGIN, CONTAINMENT, FABRIC 0F HUMAN ACTION SAFEGUARDS.

2. TAKE CHERNOBYL EXPERIENCE INTO ACCOUNT.IN REINFORCING SOME ASPECTS OF REQUIREMENTS ALREADY EXISTING OR BEING j

! DEVELOPED. (HUMAN ACTION CONTROLS, EMERGENCY PLANNING, l CONTAINMENT PERFORMANCE). ,

I

() 3. STUDIES, RESEARCH IN SOME SPECIFIC CHERNOBYL-LESSON AREAS, AS FOUPDATION FOR SUBSEQUENT CONSIDERATION OF ACTION.

(REACTIVITY ACCIDENTS, ACCIDENTS AT LOW POWER OR SHUT DOWN, SOUPCE TERM CHARACTERISTICS).

4. CHERNOBYL EXPEPIENCE WILL REMAIN A CONTINUING PART OF THE BACKGROUND INFORMATION TAKEN INTO ACCOUNT IN A VARIETY OF REACTOR' SAFETY AREAS -- EVEN' APART FROM ANY ACTIONS THAT MAY BE TAKEN IN DIRECT RESPONSE TO IT.

1 s

j A-/7 I

~

i SPELIFIC AREA CONCLUSIONS t I

1

1. OPERATIONS:  ;

ADMINISTRATIVE CONTROLS BASED ON EXISTING REGULATORY r PROVISIONS ARE GENERALLY ADEQUATE TO ASSURE A SAFE f

OPERATING ENVELOPE. (PROCEDURAL ADEQUACY AND i a

COMPLIANCE, APPROVAL OF TESTS AND UNUSUAL OPERATIONS, 1

i BYPASSING OF SAFETY SYSTEMS, ESF AVAILABILITY, OPERATOR  !

l ATTITUDES TOWARD SAFETY, MANAGEMENT SYSTEMS).

I i 1 1 1

REINFORCE ATTENTION TO HUMAN FACTORS:

l i

i  !

CONSIDER A HIGH-LEVEL ON-SITE NUCLEAR SAFETY

! MANAGER, WITH NO OTHER DUTIES.

• l i i

i CONSIDER A PROGRAM OF PREPAREDNESS FOR ACCIDENT MANAGEMENT. (TRAINING AND PROCEDURES FOR COPING

{ WITH SEVERE CORE DAMAGE'AND FOR MANAGEMENT OF l

! L CONTAINMENT). l l

1 REVIEW ADMINISTRATIVE CONTROLS TO STRENGTHEN

} PROCESS OF TECHNICAL REVIEWS AND APPROVAL OF  ;

I 3 i CHANGES, TESTS, AND EXPERIMENTS.

j l

REVIEW ESF SAFETY SYSTEM STATUS DISPLAYS AND AVAILABILITY FOR POTENTIAL WORTHWHILE IMPROVEMENTS. f i .

\ A-16  ;

i l

l . - - - - - . . - . - - , - - - - , , , - , . - . - - - - - , . , - . , . , - - - . , - - - , - - - - - - . - - - - - - - - - - .

L I i j

l l:.1 REACTIVITY ACCIDENTS: POSITIVE VOID COEFFICIENT I

ABSENT. 'BUT SHOULD REVIEW RISKS FROM VULNERABLE I SEQUENCES (PRA AND DETERMINISTIC). i J II.2 ACCIDENTS AT LOW POWER'AND SHUT DOWN: REVIEW, IN VIEW  ; 4 0F PRESUMPTION SOMETIMES THAT FULL POWER OPERATION q i LIMITS. (RECOGNIZE ESF STATUS.) l l. j 11.3 MULTIPLE UNIT PROTECTION: l ASSESS ADEQUACY WITH " REALISTIC" SOURCE TERMS (VS. I PRESENT ASSESSMENT PER TID-14844). I i , i. FOR FUTURE PLANTS, DO NOT SHARE SYSTEMS THAT ARE l j PART OF SHUTDOWN CAPABILITY. l t

 !                                            II.4 FIRES: FIRE FIGHTING WITH RADIATION PRESENT                                     ASSURE t
 !                                                 ADEQUATE PROVISIONS.

l - I i l l l i I i

!                                                                                                                                                                 l
}

l I i h-11 l'

    --..__ ._-.,_ _~. _ _ _ ,.._ ._--_-. ~ . -                              _ _ . ~ . . - . _                - _ . _ _ . . ,            _ . _ - - . - , _ . - _ _

i III. CONTAINMENT / VENTING: TAKE CHERNOBYL INTO ACCOUNT IN CURRENT

   }                                         EVALUATION OF:

CONTAINMENT PERFORMANCE IN SEVERE ACCIDENTS. FILTERED VENTING AS A SEVERE ACCIDENT MITIGATION STRATEGY. ! IV. EMERGENCY PLANNING: l j 20-MILE PLUME EPZ CONTINUES TO BE VIEWED AS ADECUATE. (INCLUDES CONCEPT OF PROTECTIVE ACTION OUTSIDE IT IF ! NECESSARY). O

• STUDY CHERNOBYL ACCIDENT IN COMBINATION WITH NRC SOURCE

! TERM WORK FOR LESSONS FOR BASIS OF RELOCATION, I DECONTAMINATION, INGESTION PATHWAY. (WITH FEMA.) V. SEVERE ACCIDENT PHENOMENA: RECOGNIZE MECHANICAL DISPERSAL MECHANISMS. CHEMICAL STRIPPING OF FPS FROM FUEL PARTICLE SURFACE

!                                                                    (VIA OXIDATION TO U        3g0 )

i l ' l 3

N VI. GRAPHITE MODERATED REACTORS (FORT ST. VRAIN HTGR, MODULAR HTGR CONCEPT): hu vantCT ASSOCI ATI0H WITH DESIGN WEAKNESSES THAT CONTRIBUTED TO CHERNOBYL ACCIDENT. (HTGRS HAVE HELIUM COOLANT, CERAMIC CORE, NEGATIVE OVERALL REACTIVITY COEFFICIENT, DIVERSE ALTERNATE SHUTDOWN AND COOLING SYSTEMS). NO NEW CONCERNS RE HTGR SEVERE ACCIDENT PHENOMENA, BUT REINFORCED DESIRABILITY OF: TARGETED PHA O - GRAPHITE THERMAL STRESS EXPERIMENTS O A-21

s

,  .                                                                        APPENDIX Y                   l RECENT SIGNIFICANT EVENTS            I AGENDA FOR ACRS MEETING Friday December 11, 1986 3:00 p.m.

Room 1046 H-Street Washington, D.C. RECENT SIGNIFICANT EVENTS Presenter / Office i Date Plant Event , Telephone Page 11/20 Oyster Creek Drywell Shell Corrosion J. Donohew, NRR 29421 3 11/20 BJron:2 Loss of Both Component Cooling Water Pumps R. Woodruff, IE ' 27205 7

      's p/22   ANO-1                High Pressure Injection Nozzle External Surface H. Bailey, IE 29006 3

Damage Due to Boric Acid Corrosion 11/19 Pilgrim Loss of Offsite Power J. Rosenthal, IE (( 24193 8/1-10/14 ------- Design Problems at Plants R. Woodruff, IE /A Operating and Under Constr 27205 12/9 Surry 2 Failure of Main Feedwater J. Rosenthal, IE /3 Pipe 24193

                                                                                                     ~

10/1 Oconee 2 Loss of Low Pressure Service J. Rosenthal, IE Water (UPDATE) 24193 i O A -*

OTHER EVENTS PRESENTED TO J. Ebersole

    )                       Meeting on December 9, 1986 j                                   (Tuesday) 8:30 a.m.                                      ~

Room P-422 Phillips Building Bethesda, Md RECENT SIGNIFICANT EVENTS 1 Presenter / Office Date Plant Event Telephone Page 10/23 Indian Pt. 2 Auxiliary Feedwater Problems R. Woodruff, IE 27205

                                                                            /7 10/4  Clinton         Fuel Handling Problems        M. Wegner, IE 24511
                                                                           /7 11/5  Nine Mile Pt. 2 Problems During Initial       M. Wegner, IE       31 Fuel load                     24511 10/28 McGuire 1&2     Both Trains of ECCS           A. Dromerick, IE    13 Inoperable                    24784
    'a/22 Point Beach     Scram with Complications      T. Colburn, NRR     1N 29787 11/3  Pilgrim         Stanby Gas Treatment System   R. Auluck, NRR      37 Design Deficiencies           29476 10/21 Trojan          Potential EQ Deficiency of    T. Chan, NRR         30 Safety System Components      27136 10/21 Trojan          Low 'emperature Overpressure  T. Chan, NRR        3/

Protection 27136 8/19 WNP 2 No Analysis of Potential .l. Bradfute, NRR p Flooding Due of Fire 29414 12/4 Hatch 1 & 2 AIT for Leak from Spent E. Weiss, IE Fuel Pool 29005 3  ! C\ Q 1 A -23 _2

                                                                      ..        .__ N

DECEMBER 1, 1986 ( OYSTER CREEK - DRYWELL SHELL CORROSION

                                                                                         ~

NOVEMBER 20, 1986 (J. DON 0 HEW, NRR) - PROBLEM: SIGNIFICANT REDUCTION OF DRYWELL SHELL DUTSIDE THICKNESS, AROUND CIRCUMFERENCE AT FLOOR OF DRYWELL CAUSE: 1.ICENSEE THEORIZES CAUSE IS CORROSION DUE TO MOISTURE ENTRAINED IN SAND IN SMALL CAVITY AROUND CIRCUMFERENCE. SIGNIFICANCE: DEGRADED SHELL MAKES CONTAINMENT INTEGRITY UNCERTAIN. DISCUSSION: WATER LEAKED THROUGH BELLOWS SEAL BETWEEN OUTSIDE SHELL AND BIOLOGICAL SHIELD WHEN REFUELING CAVITY FLOODED FOR REFUELING.  ; LEAKAGE COLLECTED FROM DRAINAGE CHANNEL BELOW DOWNCOMERS. WATER QUALITY INDICATED THIS WAS REACTOR GRADE WATER. SEAL BETWEEN SHELL AND CONCRETE FIXED IN CYCLE 11R OUTAGE. WATER DRAINED DOWN OUTSIDE OF SHELL TO CAVITY, FILLED WITH SAND, AND OUT DRAINAGE CHANNEL. SAND USED IN CAVITY IS SAME AS SAND USED IN MAKING CONCRETE AT OYSTER CREEK j UT USED TO MEASURE SHELL THICKNESS. MEASUREMENTS RANGE FROM 0.38" l TO 1.09". AVERAGE READING IS 0.92". NOMINAL WALL THICKNESS IS 1.15 " , FOLLOWUP: MEETING WITH GPU NUCLEAR ON DECEMBER 1, 1986 IN PROGRESS. OYSTER CREEK RESTART FROM CYCLE 11R OUTAGE DELAYED. LICENSEE MAY REMOVE ONE OR TWO PLUGS FROM SHELL IN CORROSION AREAS TO INVESTIGATE OUTSIDE OF SHELL. PROBLEM HAS GENERIC IMPLICATIONS FOR MARK I CONTAINMENTS. i t A-:tt  ! 3

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GPU NUCLEAR CORPORATION . GENERAL ARRANGEMENT {,, ' OYSTER CREEK NUCLEAR GENERATING STATION REACTOR BUILDING SECTION C-C l UPDATED I FINAL SAFETY ANALYSIS REM)RT REY.1. 2/86 l FIGURE 6.2-1E l A-W y

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St'PVEY OF PLANTS IN BWDI/ DEL /NRR MARK I CONTAINMENT SHELL CORROSION

                      ~

ULTRASONIC (UT) PLANT- DESIGN MEASUREMENTS DRESDEN ?/3 WATER DOES NOT DRAIN WITHIN DPYWELL l THP0 UGH SAND CAVITY SPECIFICATIONS MONTICELLO WATER DOES NOT DRAIN WITHIN DRYWELL THROUGH SAND CAVITY SPECIFICATIONS NINE MILE POINT 1 WATEP.DOES NOT DRAIN NONE THROUGH SAND CAVITY O OYSTEP CREEK WATER DOES DRAIN WASTAGE OF l THB00GH SAND CAVITY DRYWELL SHELL  ; 1 l PILGRIM WATER DOES NOT DRAIN NONE l THROUGH SAND CAVITY OUAD CITIES 1/2 FATER DOES NOT DRAIN WITHIN DRYWELL THROUGH SAND CAVITY SPECIFICATIONS A -21 6 . i

l

                                                                            .           I BYRDN 2 - LOSS OF ALL COMPDNENT COOLING WATER (CCW)                     l NOVEMBER 20, 1986 - R. WOODRUFF, IE                             l PROBLEM:                                                                          j

• DURING PREPARATION FDR INITIAL CRITICALITY, ALL CCW WAS LOST FDF. 12 MINUTES CAUSE:
• THE SAFETY VALVE ON THE CCW SIDE OF AN EXCESS LETDDWN M AT EXCHANGER (ELHX) LIFTED AND STUCK DPEN SIGNIFICANCE:
• SEVERAL SINGLE ACTIVE FAILURES COULD CAUSE TEMPORARY LDSS OF ENGINEERED SAFETY FEATURES (ESFs)

DISCUSSION:

• ON 11/20, ONE OF FIVE CCW PUMPS WAS STDPPED AND ANOTHER WAS STARTED ,

A PRESSURE SURGE CAUSED THE SAFETY VALVE ON DNE OF TWO ELHXs TO OPEN THE SAFETY VALVE STUCK AND CCW WAS PUMPED TO A FLDOR DRAIN UNTIL THE OPERATING PUMP TRIPPED DN LOW WATER LEVEL IN THE ( SURGE TANK THE BACKUP CCW PUMP STARTED ON LOW PRESSURE IN THE PUMP DISCHARGE HEADER AND PUMPED UNTIL IT TRIPPED THE LICENSEE MANUALLY ISOLATED THAT PART OF THE CCW SYSTEM SERVING COMPONENTS WHICH ARE NOT ESFs AND REFILLED THE CCW SYSTEM FOLLOWUPr

• THE SYSTEM CONFORMS TO THE FSAR
• IE IS PREPARING AN INFORMATION NOTICE l
• NRR IS REVIEWING THE CONFORMANCE OF THE CCW SYSTEM TD REGULATORY REQUIREMENTS i

O l A -* , i 7  ;

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2 ANO-1, HIGH PRESSURE INJECTION (HPI) NOZZLE EXTERNAL SURFACE DAMAGE DUE TO BORIC ACID CORROSION OCTOBER 22, 1986 (H. BAILEY, IE) i PROBLEM: HPl NOZZLE HAD EXTENSIVE WASTAGE DUE TO BORIC ACID CORROSION CAUSE: . EQUIPMENT FAILURE AND INADEQUATE SURVEILLANCE SIGNIFICANCE: DEGRADATION OF RCS PRESSURE BOUNDARY. 4 CIRCUMSTANCES: , PLANT.lN COLD SHUTDOWN PERFORMING SURVEILLANCE OF HPl NOZZLE THERMAL SLEEVES. INSULATION REMOVED FROM THE "A" NOZZLE REVEALED EXTENSIVE WASTAGE. l HPI ISOLATION VALVE, DIRECTLY ABOVE HAD PREVIOUSLY BEEN LEAKING (.09 GPM) FROM BONNET FOR SEVERAL MONTHS. THE VALVE LEAKAGE KEPT THE NOZZLE WETTED AND THAT RESULTED IN ACCELERATED CORROSION. 4 DISCUSSION: O

• lE BULLETIN 82-02 DISCUSSES THREADED FASTENER WASTAGE DUE TO BORIC ACID.

ASME CODE, SECTION X1, WAS REVISED.IN 1983 TO PROVIDE MORE I RESTRICTIVE VISUAL INSPECTIONS OF SYSTEMS CONTAINING BORATED WATER. FOLLOWUP: GTHER THREE HP1 NOZZLES WERE INSPECTED AND SHOWED NO EVIDENCE

OF VALVE LEAKAGE OR ANY DAMAGE.

! THE DAMAGED HPl NO22LE REPAIRED BY GRINDING OUT AND REBUILDING i BY WELDING. AN IE INFORMATION NOTICE IS UNDER PREPARATION. l A-30 l

O O O~ Figure 1: ANO-1 HPI Line/ Nozzle Configuration i (,w W7 l(Il lJ mer I m S. L g 7 RCS Cold Leg Piping SS Line i

                         --                 SS Safe End
             ~ 8 Ft. 4--       _

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                                                   /m l _  _          ,

d NnZZle-to-Cold leg Weld inconel Weld Carbon Steet Nozzle (Damaged area shown darkened) N

PILGRIM - LOSS OF OFFSITE POWER NOVEMBER 19, 1986, (J. ROSENTHAL, IE) i PROBLEM: DURING SNOW STORM ALL THREE OFFSITE POWER LINES ARE LOST I CAUSE: COMBINATION OF HIGH WINDS, SNOW AND SALT BUILD-UP ARE PROBABLE CAUSE SIGNIFICANCE: SALT BUILD-UP HAS CAUSED RECURRING PROBLEMS AT THIS SITE DISCUSSION: OFFSITE POWER LOST AT ABOUT 0805 SHUTDOWN COOLING LOST OPEN TELEPHONE LINK ESTABLISHED BETWEEN HQ, REGION AND SITE T SHUTDOWN COOLING RESTORED AT ABOUT 0920 PLANT HAS BEEN SHUT DOWN SINCE APRIL (RCS TEMP 114*F) HEAVY WET SNOWFALL OF ABOUT 4 INCHES WITH WINDS OF 45 TO 50 ,. MPH WITH GUSTS TO 60 MPH SITE LOCATED ON CAPE COD BAY BOTH DIESEL GENERATORS START AND LOAD 20,000 GALS OF DlESEL FUEL AVAILABLE, WHICH IS ADEQUATE FOR MORE THAN 100 HOURS 345 KV LINE FROM NORTH RESTORED TO YARD BUT NOT PLANT AT 1103 SAFETY BUSES RERAIN ON DIESEL GENERATORS UNTIL LOAD DISPATCHER TROUBLESHOOTS YARD OFFSITE POWER RESTORED TO PLANT AT ABOUT 1130 DIESEL GENERATORS PARALLELED TO GRID PRIOR TO SHUTDOWN-

DG SHUTDOWN AT 1238 FOLLOWUP

LIKELlHOOD OF OCCURRENCE WARRANTS ATTENTION MUST BE INTEGRATED WITH BLACKOUT RULE A -3.L

                                                                                         //

DESIGN PROBLEMS AT PLANTS OPERATING AND UNDER CONSTRUCTION AUGUST 1 THRU OCTOBER 14, 1986 - R. WOODRUFF, IE PROBLEM: 50 DESIGN PROBLEMS AND 4 POSSIBLE DESIGN PROBLEMS WERE IDENTIFIED FROM 50.72 AND REGIONAL DAILY REPORTS DISCUSSION: DESIGN PROBLEMS AS OPPOSED TO FAILURES TO CONSTRUCT, OPERATE, TEST, MAINTAIN, AND MODIFIY THE PLANT AS REQUIRED BY DESIGN

• SYSTEM PROBLEMS 12 ELECTRICAL (6 DIESEL GENERATOR, 3 BUS, AND OTHERS) 8 CONTAINMENT 7 ECCS '

5 HEATING, VENTILATING, AND AIR CONDITIONING

• PLANT PROBLEMS 6 COMANCHE PEAK 6 PERRY 4 PALISADES
• NSSS PROBLEMS 1

23 GE 21 WESTINGHOUSE

• AE PROBLEMS 16 BECHTEL (PALISADES, HOPE CREEK, SUSQUEHANNA, GRAND GULF, AND OTHERS) 9 SARGENT & LUNDY (FORT ST VRAIN, ZION, AND OTHERS) 6 GIBBS & HILL (COMMANCHE PEAK) 6 GILBERT (PERRY)

I THESE EVENTS HAVE BEEN GIVEN THE SAME SCRUTINY BY NRR, AEOD, IE, AND THE REGIONS AS OTHER EVENTS (OPERATING, MAINTENANCE, i ETC) FOLLOWUP: ' MAINTAIN THE DATA BASE ON AN INTERIM BASIS UNTIL THE CORPORATE DATA NETWORK IS AVAILABLE l A-33 4

                                                                                                      /;L

s %,/ SURRY 2 - FAILURE OF MAIN FEEDWATER PIPE DECEMBER 9, 1996 - J. ROSENTHAL PROBLEM:

• AN 18-INCH MAIN FEEDWATER PIPE ON THE SUCTION SIDE OF FEEDWATER PUMP "A" FAILED CATASTROPHICALLY PROBABLE CAUSE:
• PIPE WALL THINNING AND SYSTEM PRESSURE TRANSIENT SIGNIFICANCE:
• 5 WORKERS INJURED DISCUSSION:
• AT 2:40 PM ON 12/9/86, THE REACTOR WAS OPERATING AT 100X DF FULL POWER
• MAIN STEAM ISOLATION VALVE "C" CLOSED ,
• REACTOR TRIPPED ON LOW LOW LEVEL IN STEAM GENERATDR "C"
• PRESSURE TRANSIENT IN MAIN FEEDWATER SYSTEM RESULTED IN BREAK OF SUCTION PIPING TO FEEDWATER PUMP "A"
• OPERATORS CLOSED VALVES IN ALL LINES SUPPLYING HIGH ENERGY FLUIDS TO THE INUNDATED AREA
• STEAM GENERATOR LEVELS WERE MAINTAINED WITH AUXILIARY FEEDWATER AND STEAM WAS DUMPED TO ATMOSPHERE
• AT 2 AM, THE RESIDUAL HEAT REMOVAL SYSTEM WAS PUT IN SERVICE
• UNIT REACHED COLD SHUTDOWN DURING THE MORNING OF 12/10/86 FOLLOWUP:
• THE LICENSEE DECLARED AN ALERT
• THE REGION DISPATCHED AN AUGMENTED INSPECTION TEAM
• THE LICENSEE SHUT DOWN UNIT 1 FOR EXAMINATION
• PRELIMINARY INFORMATION ATTRIBUTES CAUSE OF PIPING FAILURE TO EROSION OF PIPE WALL AND A PRESSURE TRANSIENT WHICH OVERLOADED THE THINNED WALL SECTIION O A-M

                                                                                            /:5

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' INDIAN POINT 2 - AUXILIARY FEEDWATER PROBLEMS OCTOBER 20 & 23, 1986 - R. WOODRUFF, IE i i PROBLEM: i i

• THREE DEFECTIVE CDMPONENTS CAUSED LDSS OF SDME AUXILIARY l FEEDWATER (AFW) CAPABILITY (10/20) j
• THE GOVERNOR VALVE ON THE AFW TURBINE-DRIVEN PUMP WAS

;                       IMPROPERLY SET (10/23)

CAUSEs  ;

• IMPROPER NAINTENANCE (10/20 & 23) I
• AGING AND WEAR LIKELY (10/20)

SIGNIFICANCE: l

• MARGINAL MAINENANCE DISCUSSION: l

$

• ON 10/20, THE REACTOR WAS AT 100% OF FULL POWER l

    '\

l

• DURING SURVEILLANCE, A LODSE WIRE IN THE REACTOR PROTECTIDN  !

SYSTEM WAS DISTURBED l

,

• A SPURIOUS SIGNAL WAS GENERATED INDICATING TPAT SAFETY INJECTION HAD OCCURRED
• THE REACTOR SCRAMMED IN RESPONSE TO THE SIGNAL l

BOTH MOTOR-DRIVEN AFW PUMPS STARTED BUT DNE TRIPPED DN SPURIOUS OVERCURRENT BECAUSE ON A DEFECTIVE CARD

• TURBINE-DRIVEN AFW PUMP STARTED AND RAN ALTHOUGH THE STEAM SUPPLY RELIEF LIFTED BECAUSE THE PRESSURE CDNTROL VALVE WAS DEFECTIVE

$

• THE LOOSE WIRE WAS TIGHTENED
• THE OVERCURRENT PROTECTION CARD WAS REPLACED AND THE SETPDIN1 WAS INCREASED BY 10%
• THE SOVERNOR VALVE WAS RESET FROM A CORRECT VALUE OF 0% TO AN INCORRECT VALUE OF 20%
• THE UNIT WAS RESTARTED
• AT 38% POWER, MAIN FW WAS LDST AND THE REACTOR WAS f1ANUALLY s SCRAMMED

                                                              } ~ hh 17

l 1 t

• THE TURB1NE-DRIVEN AFM TRIPPED DN DVERSPEED
• THE GOVERNOR VALVE WAS RESET TD 0%

FOLLOWUP:

• IE DBTAINED AN INTERPRETATION OF REGULATORY REQUIREMENTS a

'l f i 9

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                                                                                                                             /8

CLINTON - FUEL HANDLING PROBLEMS OCTOBER 1986 - (MARY WEGNER, IE) PROBLEM: WRONG ENRICHMENT BUNDLES HAVE BEEN PLACED IN CORE LOCATIONS, l MISSED SUB-CRITICALITY VERIFICATIONS, AND FUEL TRANSFER CARRIAGE FAILURE, HAVE OCCURRED IN THE PERIOD OF DCTOBER 4-8, i 1986, i CAUSE: PERSONNEL ERROR SIGNIFICANCE: FAILURE TO FOLLOW FUEL LOADING PROCEDURES SIGNIFIES AN INATTENTION TO A MOST IMPORTANT OPERATION CIRCUMSTANCES: O OCTOBER 4 - TWO MEDIUM ENRICHMENT BUNDLES WERE PLACED IN HIGH ENRICHMENT BUNDLE LOCATIONS FUEL LOADING SPECIFICATIONS ARE FOR ENRICHMENT TYPE . (HIGH, MEDIUM, NONE) RATHER THAN BUNDLE NUMBER OCTOBER 7 - THE FUEL TRANSPORT CARRIAGE FAILED TO STOP AT l THE REQUIRED POSITIGN DURING TRANSPORT OF TWO NEW FUEL

BUNDLES l

                        - THE CARRI AGE TRAVELED FOUR FEET BEYOND ITS UPPER STOP      l i

l

                         - MINOR MECHANICAL DAMAGE RENDERED THE CARRIAGE:lN0PER-ABLE UNTIL OCTOBER 9 i                WHILE THE FUEL TRANSFER CARRIAGE WAS Ih0PERABLE, THE LICENSEE DEVI ATED FROM THE SPIRAL LOADING PATTERN OCTOBER 7 - THE LICENSEE FAILED TO CONDUCT A SUBCRITICALITY TEST FOLLOWING THE LOADING OF BUNDLE 248                              !
                        - BUNDLES 249 AND 250 WERE LOADED BEFORE DISCOVERY OF THE OMISSION
                       - BUNDLES 249 AND 250 0FF-LOADED AND TEST COMPLETED            l w

I l /9 L J

CLINTON - FUEL HANDLING PROBLEMS OCTOBER 1986 - (MARY WEGNER, IE) ,' OCTOBER 8 - THE LICENSEE IDENTIFIED A MEDIUM ENRICHMENT BUNDLE ABOUT TO BE PLACED IN A LOCATION FOR A HIGH ENRICHMENT BUNDLE NO ADDITIONAL PROBLEMS REPORTED SINCE BUNDLE 261 FOLLOW-UP: FUEL HANDLING CONCERNS DISCUSSED WITH THE REGION O O A-Al 2.0 _m _m_r._ . . _ . . - _ _ _ , _ _ _ _ _ - . , - , _ _ - . . , _ . . . , , , _ . . , . . _ .- . - . . . . - , . _ . - , . - . _ . . - - . _ , _ , . ,

l i p- p . q ' c _. , 55 . . 51- . 5 , - 47 p , *! , 1 s 43 - M~ . 35 - . 9 Fr S&S - . 31 " +rg 6z. 1, - i?>%# M G fC r,s Y((;gQav 37 ~ Sb s 2@ ' 13 - ~fD,, "t,fg rFM

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                                                                                                                                     .                       .; eV e' .                 ,

4 = SAM Detector Location A = Alternate Source Location . 1 = High a fuel asses 61y (8C18219-4GDZ-103M-150) l 2 = Medium e fuel asserbly (SC181M-44DZ-100M-150) A4h - e in.3-pn-te . 2f

N1tlE MILE POINT 2 FUEL HANDLING PROBLEMS NOVEMBER, 1986 (M. S. WEGNER, IE) RPS ACTUATIONS, TECHNICAL SPECIFICATION (TS) VIOLATIONS, AND OTHER PROBLEMS DURING FUEL LOADING CAUSE: PERSONNEL ERRORS AND EQUlPMENT FAILURE SIGNIFICANCE: DEGRADED RPS DURING FUEL LOAD, LOSS OF RAD DETECTION, UNNECESSARY CHALLENGES TO RPS SEVERAL VIOLATIONS OF TECHNICAL SPECIFICATIONS DISCUSSION: ON 11/05 INTERMEDIATE RANGE MONITOR (IRM) "D" SPIKED CAUS SIGNAL - APPARENT CAUSE WAS BUMPING BY FUEL BUNDLE B

                - SCRAM DISCHARGE VOLUME HIGH LEVEL SCRAM OCCURRED WHEN

• FAILED TO FOLLOW PROCEDURE FOR RESETTING SCRAM ON 11/07 SOURCE RANGE MONITOR (SRM) "C" TRIP FUNCTIONS WERE BYPASSED TO PERFORM FUNCTIONAL CHECKS AND LEFT IN THIS CONDITIO bq HOURS.

                - WITH ALL TRIP FUNCTIONS OF SRM "C" INOPERABLE, 19 FUEL BUNDLES WERE LOADED IN THE QUADRANT, CONTRARY TO TS REQUIREMENTS ON .11/08, THE RAD MONITOR FOR THE REACTOR BUILDING BELOW THE i

REFUELlHG FLOOR WAS REMOVED FROM SERVICE FOR MAINTENANCE WHILE FUEL LOADING WAS IN PROGRESS; CONTRARY TO TS REQUIREMENTS,

• THE STANDBY GAS TREATMENT SYSTEM WAS NOT PLACED IN SERVICE ON 11/09, TWO APRM-H1 TRIPS ON CHANNEL "C" WERE RECEIVED.

               - THE FIRST WAS ATTRIBUTED TO WELDING IN THE AREA 0F THE LP
               - AFTER THE SECOND OCCURRED WITH NO WELDING IN PROGRESS, A FAULTY CARD IN THE LPRM CIRCulTRY WAS FOUND.

DURING FUEL LOADING, R0D BLOCKS WERE BYPASSED CONTRARY TO TS REQUIREMENTS FOLLOWUP: REGION IS FOLLOWING OPERATIONS CLOSELY AND DISCUSSING EVEN WITH LICENSEE ENFORCEMENT CONFERENCE ON TS VIOLATIONS TO BE SCHEDULED IN EARLY DECEMBER LICENSEE IS TAKING CORRECTIVE ACTIONS, (E.G. DISCIPLINARY ACTION) w n

MCGUIRE 1 8 2 - BOTH TRAINS OF ECCS INOPERABLE OCTOBER 28, 1986 - (A. DROMERICK, IE) () FROBLEH: LICENSEE CHANGES TO MOV TORQUE SWITCH SETTING FROM THOSE SET AT FACTORY MADE VALVE POST ACCIDENT OPERABILITY QUESTIONABLE CAUSE: INSTALLATION ERROR, LICENSEE MADE CHANGES TO MOV TORQUE SWITCH SETTINGS WITHOUT CONSULTING VALVE OPERATOR VENDOR SIGNIFICANCE: POTENTIAL COMMON MODE FAILURE OF A SIGNIFICANT NUMBER OF SAFETY-RELATED VALVES DISCUSSION: WHILE LICENSEE PERFORMING IEB 85-03 EVALUATIONS, IT WAS DETERMINED THAT TORQUE SWITCH SETTINGS MAY NOT BE CONSERVATIVE ON ROTORK MOV LICENSEE CONCLUDED THAT TWO CHARGING HEADER ISOLATION VALVES AT EACH UNIT MIGHT NOT CLOSE WHEN REQUIRED LICENSEE LOCKED VALVES CLOSED AND SHUT DOWN BOTH UNITS LICENSEE HAS VERIFIED THAT VALVES AT MCGUIRE AND OTHER DUKE

'()

• PLANTS HAVE CONSERVATIVE TOROUE. SWITCH.SETTlNGS REGION II AND IE ISSUED AN INFORMATION NOTICE TO ALL CP AND OL PLANTS (THIS IS SECOND IN ISSUED AS A RESULT OF INFOR-MATION OBTAINED BY BULLETIN ACTIONS).

NPRDS DATA INDICATES THE FOLLOWING APPROXIMATE DISTRIBUTION FOR VALVES WITH ROTORK OPERATORS UNIT NUMBER OF VALVES MCGUIRE 1 & 2 275 EACH CATAWBA 1 & 2 175 EACH OCONEE 1, 2, & 3 15 EACH 3 ADDITIONAL 30-40 EACH 15 ADDITIONAL 10 EACH ROTORK INFORMATION INDICATES THAT 29 PLANTS OTHER THAN DUKE PLANTS HAVE VALVES WITH THEIR OPERATORS FOLLOW-UP: REGION 11 FOLLOWING LICENSEE ACTIVITIES IE IS REVIEWING BULLETIN RESPONSES, ( ) k11

                                                                                 ;t3

P01hT EEACH 1 - LOSS OF RFD INSTRUMENT BUS REACTOR SCRAM NOVEKEER 17, 1986 - (T COLEURN, NRR) PROBLEM: LOSS OF FED INSTRUMEhT BUS SIGNALED 1 OUT OF 2 LOGIC FOR FEED FLOW STEAM FLOW MISMATCH RESULTING IN REACTOR TRIP CAUSE: SHCRTED DIODE IN INVERTER SUPPLYING RED INSTRUMEhT EUS CAUSES 4 BLOWN OUTPUT FUSES AND LOSS OF BUS

.      SIGNIFICANCE:
      ~*

! BRIEF TEMPORARY LOSS OF ONE OF THREE AUXILIARY FEECKATER SOUFCES (OhE AIR-0PERATED AFW FUMP DISCHARGE VALVE CONTROLLER IS POWERED FROM RED INSTRUMENT BUS) ONE SOURCE RANGE NUCLEAR INSTRUMENT DETECTOR FAILED LOW ONE lhTERMEDIATE RANGE DETECTOR APPAPEFTLY FAILED LOW DISCUSSION: POWER RESTORED TO INSTRUMENT BUS VIA ALTERNATE INVERTER IN 4 MINUTES SUBSEQUENT CHECKS OF FAILED IR DETECTOR INDICATED INCORRECT COMPENSATlhG VOLTAGE SET FOR LOW POWER, NOT A DETECTOR FAILURE

ANY LOSS OF AN INSTRUMENT BUS WILL RESULT IN A REACTOR TRIP (4 FEED FLOW STEAM FLOW MISMATCH CHANNELS, 2 PER LOOP, EACH PChERED FR0h SEPARATE INSTRUMENT EUS)

LOSS OF RfD OR BLUE INSTRUMENT EUS ALSO CAUSES LOSS OF PCKER TO A OR B AIR-0PERATED AFW PUMP DISCHARGE VALVE CONTROLLER, TbhEINE DRIVEN AFW PUMP UNAFFECTED REACTOR CRITICAL - 4 HOURS AFTER EVENT FOLLOWUP: RESIDENT INSPECTOR FOLLOWING LICENSEE ACTIONS l i w M - . . .

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OTHER EVENT OF INTEREST ' ' PILGRIM - STANDBY GAS TREATMENT (SBGT) SYSTEM DESIGN y DEFICIENCIES _' NOVEMEEP 3, 1986 - (R. AULUCK, NRR) PROBLEM: SINGLE ACTIVE FAILURES WHICH DIMINISH THE ABILITY OF i SEGT SYSTEM TO PERFORM ITS SAFETY FUNCTION CAUSE: DESIGN DEFICIENCY - REDUNDANCY NOT PROVIDED I SIGNIFICANCE: POTENTIAL FOR RELEASE EXCEEDING 10 CFR P GIVEN A SINGLE FAILURE IN STANDBY GAS TREATMENT; .i ,

'                        DISCUSSION:                                                              '      '

i ON AUGUST 27, 1986, THE LICENSEE (BECO) CONCLUDED THAT SlHCLE i ACTIVE FAILURE OF SBGT DELUGE SYSTEM DURING POS e LOCA OR FUEL HANDLING ACCIDENT COULD RESULT IN OFFS EXCEEDING PART 100 LIMITS, j

                               - INADE00          ATE SBGT SYSTEM / DELUGE SYSTEM INTERACT THE ORIGINAL DESIGN,
                               - DELUGE SYSTEM PROVIDES COOLING AND FIRE PROTECTION SPRAY IN,'

i RESPONSE TO HIGH TEMPERATURES IN SBGT CHARC0AL FILTERS, ' j .

                               - INITI ATION OF THE DELUGE SYSTEM DUE TO FAILURE O l

COMPONENTS WILL RESULT IN SOAKING OF CHARC0AL BE RENDERING THAT TRAIN INCAPABLE 0F FURTHER IODINE FILTRATION; SYSTEM CANNOT MEET DESIGN BASIS j ON OCTOBER 30, 1986, THE LICENSEE (BECO) INFORMED NRC ABOUT. l TWO ADDITIONAL DESIGN DEFICIENCIES } - FAILURE OF THE CROSS 0VER DAMPER RESULTING IN NO GASES THROUGH THE STANDEY TRAIN.

- SINGLE CONTROL FAILURE RESULTING IN A HEATER FAILURE AND '

! A41 l i _ . _ _ _ _ - _ - - -- - - - - ---~~~ ~" ~ 21

2-DISCUSSION: - CONTINUED DAMPER FAILURE '

                          - C-ROSS TIE DANPER PROVIDES THE REQUIRED FL0h' PATH THROUGH                                        .

EITHER FILTER TRAIN WHEN THAT TRAIN IS IN STANDEY.

                          - PARTIAL OR FULL CLOSURE OF THE DAMPER COULD LEAD TO ELEVATED TEMPERATURES OR FIRE IN THE STANDBY TRAIN,                                                                         '
                          - PARTIAL OR FULL CLOSURE WILL INTERRUPT FLOW THROUGH BOTH TRAIN THUS SYSTEM CANNOT MEET DESIGN BASIS,
        ~

HEATER FAILURE

                         - FOR CHARC0AL FILTERS TO NAINTAIN THElR EFFICIENCY FOR FILTERING OUT RADIO 10 DINES THE HUMIDITY OF THE AIR STREAM MUST BE CONTROLLED FOR BOTH PP.1 MARY AND STANDBY TRAINS.                                                           ,
                         - IN THE PRESENT DESIGN, THE HEATERS DO NOT HEAT THE AIR PASSING THROUGH THE STANDBY TRAIN ONCE FAN IS DE-ENERGIZED: THUS, SYSTEM CANNOT MEET DESIGN BASIS                                                                           ;
                         - FAILURE OF THE HEATER AC POWER OR TEMPERATURE SWITCH ON THE.

ACTIVE TRAIN COULD ALSO IMPACT OPERABILITY FOLLOW-UP: LICENSEE WILL RE-EVALUATE DESIGN BASIS

        ,                DESIGN MODIFICATIONS WILL BE COMPLETED AS NECESSARY E

t 1 A- 41

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TROJAN - POTENTIAL E0 DEFICIENCY OF SAFETY SYSTEM COMPONENTS OCTOBER 21, 1986 - (T. CHAN, NRR) PP0BLEM: AllXILIARY FEEDWATER (AFW) PUMPS, EMERGENCY DIESEL GENERATOP (EDG) CONTROLS AND PEMOTE SHUTDOWN PANEL MAY NOT BE QUALIFIED FOR A HARSH ENVIRONMENT CAllSE: IMPROPER ASSl!MPTIONS IN ANALYSES SIGNIFICANCE: POTENTIAL FAllllRE OF SYSTEM / COMPONENTS REQUIRED FOR MITIGATION OF A STEAM OR FEEDWATER LINE BPEAK IN THE TURBINE BUILDINS, DISCllSSION: OCTOBER 1985 - LICENSEE'S INTERNAL EQ REVIEW REVEALS POTENTIAL OVEPPRESSURE OF ROOMS ENCLOSING SAFETY RELATED E0llIPMENT DUE TO STEAM /FEEDWATER LINE BREAK IN TURBINE BllILDING, AND ERRORS IN ORIGINAL HELB ANALYSIS JUNE 1986 - SUBSEQUENT INVESTIGATIONS FOUND ROOMS NOT' SEALED FOR HARSH ENVIRONMENTS AS DESIGNED AND ASSUMED IN HELB ANALYSIS - (INGRESS THROUGH VENTTLATION OPENINGS) LICENSEE INITIATES NEW ANALYSIS TO RECONCILE OVERPRESSilRE AND TEMPERATURE CONCERNS 1978 BECHTEL ANALYSIS AND 1986 LICENSEE /IMPELL ANALYSES UTILIZE DIFFEPENT ASSUMPTIONS, PESilLTING IN DISPARITING ENVIFONMENTAL CONCLUSIONS AS AN INTERIM MEASURE, LICENSEE REMOVED LARGE PORTIONS OF THE TURBINE 'BilILDING SIDING TO ALLEVIATE DIFFERENTIAL PRESSURE AND TEMPERATURE CONCERNS UNTIL DIFFERENCE IN ANALYSES ARE RESOLVED EQUIPMENT IS OPERABLE: JUSTIFICATION FOR CONTINUED OPERATION HAS BEEN WRITTEN FOLLOW-UP LICENSEE PERFORMING ADDITIONAL ANALYSES LICENSEE HAS TEMPOPAPILY SEALED ROOMS AND WILL REVIEW REVISED HIGH-ENERGY LINE BREAK ANALYSIS LICENSEE TO DETERMINE ROOT CAUSE OF DEFICIENCY IN ANALYSIS ASSUMPTIONS REGION V TO F0LLOW-UP DN CORRECTIVE ACTIONS A-SI I 1 -- . 30 ._

l TROJAN - LOW TEMPERATURE OVEP PRESSUPE PROTECTION (LTOP) l DCTOBER 21, 1986 - (T. CHAN, NRR) PP0BLEM: POWER OPERATED RELIEF VALVE STR0KE TIME EXCEEDS LTOP SAFETY ANALYSIS ASSUMPTION CAUSE: FAILURE TO TRANSLATE ANALYSIS ASSUMPTIONS INTO TEST REQUIREMENTS SIGNIFICANCE INADEQUACY OF RCS OVERPRESSURE PROTECTION BELOW 290*F WITH THE POTENTIAL FOR EXCEEDING 10 CFR 50 APPENDIX G LIMITS DISCUSSION: AMENDMENT 78 STATED ACCEPTABILITY OF THE OVEPPROTECTION MITIGATION SYSTEM BASED ON PORV OPENING STROKE TIME OF 0.28 SEC. OCTOBER 21, 1986 - LICENSEE DISCOVERED THAT LTOP ANALYSIS ASSUMPTION FOR PORV STROKE TIME HAD NOT BEEN MET

          - ACTUAL STROKE TIMElS APPROX 1MATELY' B~ SECONDS TECHNICAL SPECIFICATION APPLICABLE ONLY AT OR BELOW 290*F LICENSEE CIIRRENTLY IS NOT IN AP ACTION STATEMENT LICENSEE HAS DECLARED SYSTEM IN0PEPABLE PORV SilRVEILLANCE TESTING REQUIREMENTS BASED ON ASME CODE FOLLOW-UP:

LICENSEE IS PERFORMING NEW ANAYSIS TO REANALYZE STROKE TIME AND SETTING RE0i!IDEMENTS REGION V FOLLOWING LICENSEE'S ACTION A -Ch

            ~                     _.        ___      .__      ._                              . . . . _ _ - .

i WNP NO ANALYSIS OF POTENTIAL FLOODING DUE TO FIRE I MAIN BREAKS-  ! AUGUST 19, 198G --JOHN 0. BRADFUTE, NRR f PROBLEM: . NO FLOODING ANALYSIS CONDUCTED FOR WET-SPRINKLER SYSTEM ADDED  : TO CONTROL ROOM COMPLEX ' j

• NO FLOODING ANALYSES EVER CONDUCTED FOR FIVE OTHER AREAS OUTSIDE REACTOR BUILDING AND CONTAINING SAFETY RELATED

!                            EQUIPMENT CAUSE:

j 0VERSIGHT DURING DESIGN OF CONTROL ROOM COMPLEX SPRINKLER SYSTEM BURNS AND R0E FAILED TO COMPLETE ANALYSES IN 1983 - (BURNS AND ROE COMPLETED FLOODING ANALYSES FOR REACTOR BUILDING) SIGNIFICANCE: FLOOD IN CONTROL ROOM FROM' ADJOINING AREASTOUED TREVENT PLANT O\ SHUTDOWN FROM CONTROL ROOM

                            - NO DRAINS
                            - CABLING UNDER FLOOR l                            - PROCEDURES CHANGED TO EVACUATE CONTROL R00M IN EVENT OF FLOODING l
                            - REMOTE SHUTDOWN PANEL CABLING INDEPENDENT OF CONTROL ROOM CABLING
                            - GAFE SHUTDOWN FROM REMOTE-SHUTDOWN PANEL IS ASSU. RED PRELIMINARY EVALUATIONS OF OTHER AREAS INDICATE FLOODING C 4CL          0     -

E PU L C SN SK DISCUSSION: AREAS AFFECTED:

- CONTROL ROOM (IN RAD WASTE BUILDING)

                           - CABLE SPREADING ROOM (IN RAD WASTE BUILDING)

A -33 , 32

l DECEMBER 8, 1986 l HATCH 1 AND 2 - AIT FOR LEAK FROM SPENT FUEL POOL DECEMBER 4, 1986, (E. WEISS, IE) PROBLEM: 141,000 GALLONS LEAKED FROM SPENT FUEL POOL TRANSFER CANAL CAUSESi 4 AIR SUPPLY INADVERTENTLY SHUT TO INFLATABLE SEALS IN CANAL DRAINS ON LEAK DETECTOR INADVERTENTLY LEFT OPEN SIGNIFICANCE: LEAK NOT IDENTIFIED FOR HOURS (AIR SUPPLY SHUT 2200 CST DEC 2) LEAKAGE PATH TO ENVIRONMENT AND CAUSE NOT IMMEDIATELY APPARENT i IF FUEL BUNDLE HAD BEEN IN TRANSIT, POTENTIAL FOR UNC0VERY EXISTS i DISCUSS 10N t UNIT 1 AT 100% POWER THROUGHOUT EVENTJ UNIT 2 SHUTDOWN TRANSFER CANAL SEAL IS INFLATED DURING REFUELING INFLATABLE SEAL IS USED IN GAP BETWEEN REACTOR BUILDINGS FOR SEISMIC CONSIDERATIONS DOUBLE INFLATABLE SEAL ON GATE IS IN PLACE DURING REACTOR OPERAT

SEAL LEAK DETECTION DRAIN VALVES (F238 AND F239) LEFT OPEN PRIOR TO EVENT - SEAL LEAK DETECTION ALARM DID NOT WORK PROBLEM WITH PRESSURE REGULATOR; AIR VALVE THROTTLED AIR VALVE MOVED TO CLOSED POSITION WHILE RESTORING FROM CLEARANCE 1430 CST DEC 3, THIRD LOW LEVEL ALARM IN FUEL POOL 2200 CST DEC 3, COULD NOT OPEN DOOR TO NITROGEN ROOM, WATER POURING INTO CABLE TRAYS, LEAK FOUND LEVEL DOWN ABOUT 5 FEET ALARM ON SPENT FUEL POOL LEVEL WORKED 4

17,000 GAL TO RAD WASTE 40,000 GAL CONTAINED BETWEEN REACTOR BUILDINGS 84,000 GAL TO STORM DRAIN AND SWAMP 1,26 X MPC CS-134, CS-137, ZN-65, C0-60, MN-54 DIKES BUILT FROM OUTFALL TO SWAMP RIVER RISING BECAUSE OF RECENT RAINFALL TANKER TRUCKS USED TO REMOVE WATER CLEANUP 0F WATER BY RECIRCING THRU DEMINS TO TANK TRUCK NT 33

i HATCH - CONTINUED 2 O FOLLOWbP: i HATCH AIT SCOPE IS TO DETERMINE  ! ROOT CAUSES OF EVENT AND FAILURE OF ALARM DETERMINE AMOUNT OF SPILL AND RELEASE LICENSEE RESPONSE TO IEB 84-03 AND IN 84-93 POTENTIAL RISK IN DRAINING BOTH POOLS POTENTIAL FOR LOSS OF SECONDARY CONTAINMENT  !

• LICENSEE'S KNOWLEDGE OF PRECURSOR EVENTS i PRELIMINARY OBSERVATIONS BY AIT RELATE T0:

ADMIN CONTROLS FOR PRESSURE REGULATOR AND LACK OF! DEFICIENCY REPORT CLOSING OF AIR SUPPLY VALVE WITHOUT PROCEDURE PROCEDURE TO CAllBRATE LEAK DETECTION ' PROCEDURE TO CHECK AIR PRESSURE DID NOT INCLUD CANAL SEALS

                -                                                                                                              i

• AIR SUPPLY TO SEALS NOT REDUNDANT IE CONSIDERING INFORMATION NOTICE IF CONFIGURATION UNIQUE i

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O A-54 35

APPENDIX VI OCONEE LOSS OF LOW PRESSURE SERVICE WATE OCONEE - LOSS OF LOW PRESSURE SERVICE WATER (LPSW) OCTOBER 1, 1986 (HENRY BAILEY, IE)  ! l l FOLLOWUP (12/11/86): REPAIRED FLANGE ON ALL 12 CCW PUMPS TO PREVENT AIR INLEAKAGE WITH LOW LAKE LEVEL. CCW AND LPSW SIPHON FLOW TESTED. STANDBY SHUTDOWN FACILITY DESIGN REVIEWED SEISMIC FRAGILlTY ANALYSIS FOR CCW AND OTHER BOP SYSTEMS REVIEWED. EMERGENCY POWER FOR CCW PUMPS EVALUATED. MEETING WITH NRC ON OCTOBER 14, 1986 ALL 3 UNITS RESTARTED

                                                                      ~

IE INFORMATlON NOTlCE UNDER PREPARATION. C; I l ) I A-57

SSINS No.: 6835 IN 86-99 UNITED STATES NUCLEAR REGULATORY ColmISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 December 8, 1986 l IE INFORMATION NOTICE NO. 86-99: DEGRADATION 0F STEEL CONTAINMENTS ! Addressees: i All nuclear power reactor facilities holding an operating license or a con- . struction permit.

Purpose:

This notice is to provide recipients with current information of a potentially significant safety problem regarding the degradation of a steel containment resulting from corrosion. It is expected that recipients will review this ! information for applicability to their facilities and consider actions, as i appropriate, to promptly recognize or prevent a similar problem from occurring. l However, suggestions contained in this notice do not constitute NRC requirements; ) therefore, no specific action or written response is required. , j Description of Circumstances: . l The Oyster Creek Nuclear Generating Station first discovered water in the gap i between the boiling-water-reactor drywell and the concrete shield in 1980 and l began investigation of the cause in 1983. It appeared that the collection of l water varied from a few drops to 2 gallons per minute, depending on whether i the unit was in operation or an outage for refueling. During the spring and I summer of 1986, the licensee planned work to identify and eliminate this water problem. The bellows at the drywell to cavity seal was repaired and a gasket was replaced, thus stopping the leakage. Since the bellows is located at the - top of the drywell and the region above the bellows is flooded during refueling, l it would explain why leakage was high during refueling and low during operation.

!             To determine if the water in the gap had caused damaged to the steel contain-                                                    ,

i ment, the licensee measured the wall thickness, using an ultrasonic testing ('UT) i technique at two elevations. The 51-ft level near the drywell seal wa: sound, . j but there appeared to be loss of metal on the gap side at the 11-ft 3-in. level j immediately above the concrete floor. In this area, the gap is packed with i sand and contains five equally spaced drain pipes (see attached Figure 1). A 1 total of 143 measurements were made at this level and 60 indicated a reduction in thickness of more than 1/4 in, from.the drawing thickness of 1.154 in. These readings were found throughout seven of'the ten downcomer bays. The - , licensee plans to cut the steel containment and remove about 12 samples to confirm and evaluate the corrosion damage. 8612050463 A-58

IN 86-99 . December 8, 1986 Page 2 of 1 : b Q = The licensee plans to remove a section of the drain pipe to perform a visual examination of the outside of the drywell. Wipe samples will be taken from

• several areas and a chemical analysis will be performed. Sand samples will be taken adjacent to the core holes and will be analyzed for chemicals, bac-teria, and water composition. Some channels are being cut in the concrete floor that is inside the drywell to provide access for further UT examination of the containment-sand interface.

Discussion: The purpose of the sand is to act as a cushion and allow expansion of the drywell during operation. The steel containment is in contact with sand in those areas where corrosion has been detected. The containment material is ASTM A-212 Grade B carbon steel plate. The licensee stated that the outside surface was protected with a red lead coating from above the drywell down to about the 10-ft. level, which means that the interface between the lead paint and the unprotected steel was in contact with wet sand. Red lead protects steel by providing a stable and impenetrable surface, but the steel is sacri-ficial with respect to the lead in dilute, acidic water conditions. It is possible that condensation during initial construction, moisture pickup through the drain line during operation, and the leaking bellows wetted the sand, thereby causing corrosion of the containment steel plates. During con-struction, water was seen running down the outside of containment into the sand. The five drain lines, as well as other penetrations in the concrete

                                                                                  ~

shield, are open during operation and would allow moist air 'to ente'r"anTriie'"~-

 \                  up the gap and later cool and condense as water. Water also was able to enter the gap through the holes in the bellows during refueling until repairs were made.

A related matter is discussed in IE Information Notice 86-35, " Fire in Com-pressible Material at Dresden Unit 3," where a large amount of water was used to extinguish the slowly burning fire between the drywell and the concrete shield. Oyster Creek uses different filler material. The NRC is continuing to obtain and evaluate pertinent information. If spe-cific actions are required, an additional notification will be made. No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regier,a1 Administrator of the appropriate regional office or this office. v +- ! dward L.' Division %j0ordan, f Emergency DirUtor 7 Preparedness J and Engineering Response Office of Inspection and Enforcement , Technical

Contact:

Paul Cortland, IE (301) 492-4175 g i Attachments: l 1. Fig ne 1, Sketch of Possible Degraded Area f P. LW of Recently Issued IE Information Notices

Attachment 1 IN 86-99 - December 8,*1986 Page" 1 of 1 Sketch of Possible Degraded Areas 3 IN. -- ANNUL AR ' SPACE T C' f os 1 WITH FILLER MATERIAL ,a* od . DRYWELL f4 8a 0 Cf4

                                    ~

o' o o' 1.154 IN. DESIGN THICKNESS oaod#S

                                                                                                                                                                                               ,#,F,d,,*

• o

                                                                                                                                                                                             *O'o*ae'e ELEVATION 11 FT.-                                                                                                                                         o i
                                                                                                                                                                 ![4                .

c,'* ELEVAT20N 10 FT. 3 IN. '

                                                                                                                              .,:                   e e* mee e
                                      * ,"                          es, e e ,e e e * *e,. ..

• 0, d

                                            . CONCRETE
                                                                     , f,, e                       n
                                                                                                                                              , o, b O FLOOR                   ,                        ,4 4                     oo                                                                     DOWNCOMER
                                         **er                   ,

SAND g,6 00 oeTO TORUS C oa s CD,# ay e O Og D f* A q % OD e CONCRETE a#e

• 8 aa* dA6 AREA OF u a o o , , ,a3, a e a ,p ,p ,1,NVESTIGATION 4 DRAINAGE CHANNEL
• O # aC
• 4 A WITH 4 IN. DIA. o O*O 4 8 j

o SAND FILLED TAAIN I' a o '4 4

                                        , PIPE (EVERY OTHER d                                                                                                         4 DRYWELL VENT) a #

d 4fgea e - l l l l Figure 1 bY

e - Attac hent 2 IN 06-99 Seceaner 8, 1986 L157 0F RfCENTLY 1550tD = If INF00 MAT 10h NCTICE5

                                                                                                                                                                                            ]

i wr.a o.n . j . ate .r , he*. ice ho Sutiect issue issued to l 86-21 Recognition of American 12/4/86 All power reactor Sup 1 Society Of Mechanical facilities holding Engineers Accreditation an OL or CP Progeas For h 5 tamp holders

                                                                                                                                                                                    ~

86-98 Cffsite Nesical Services 12/2/86 All power reactor facilities holding en DL er CP 86-97 feeegency Connanicatices 11/28/06 All power reactor systee facilities holding an OL or CP and fuel facilitlea 86 96 Neat faceanger Fauling Can 11/20/06 All power reactor Cause Inaceouste Ope +atility facilities holding Of Seevice Water systems an OL or CP 86 95 Leak testing lodine 125 11/14/86 All met licensees Seales Soufces in Limi. Inc. authorized te use lea;ing Cevices and tone List, Inc. imaging Mineral Analyzers devices 86-94 Hilti Contrate lapansion 11/6/86 All power reactor Ancho* Bolts facilities holding en CL er CP 86-93 ]E8 85-03 Evaluation Of 11/3/86 All power reacter Mctor-Operators leentifies facilities holding leproper forgve $ witch an OL or CP 5ettings 86-82 Failures Of Scree Dischaege 11/4/86 A11 power reactor Rev. 1 Volume bent And Drain Valves facilities holding ' an 0L or CP 86 92 Pressu'irer Safety Yalve 11/4/86 All PWR fact 11 ties Reitatility holding an CL or CP

0.

• sperating ifcense CP
• Co*stru ction Perett UNITED STATES

   ' NUCLEAR REGULATORY COMMISSION                                                                                                                         ,oT. .M ,1, WASHINGTON. D.C. 20555                                                                                                                                    **=ac teasu D C 3GRest? he G47 OFFICIAL BUSINESS PEN ALTY FOR PAIVATE USE. 4300 a
                                                                                                                                  ~

120555003911 1 1CDICY1FB11A1 . US NRC ACRS EXECUTIVE DIRECTOR H-1016 WASHINGTON %SC 7D555 1 i A4/ l _ - - _ , - , ,- , - _ - _ - _ __- n _ , _ , - - - - , ,_ . - _ . - - - . _ _ - , , , , - _ -

, L, 's s.

     .;                                                                                                                       APPENDIX VII PROPOSED BWR SEVERE ACCIDENT CONTAINf1ENT REQUIREtiENTS                             l I

i l t 4 PROPOSED BWR l SEVERE ACCIDENT  ; I - CONTAINMENT REQUIREMENTS i i R. M. BERNER0 l l i i i l DECEMBER 9, 1986 4 1 ( i I k'b 0 . i

_. n a -.- h i I l l l l l l l A-t a

I i - 4 l NRC SEVERE ACCIDENT POLICY l e AUGUST S, 1985 e PRESENT REACTORS ARE SAFE ENOUGH, BUT... 4 l e SEARCH FOR OUTLIERS e CONSIDER EALANCE OF PREVENTION AND MITIGATION j - SPECIAL CONSIDERATION OF CONTAINf!ENT PERFORi%NCE O 1 l l . I i J i i 0 A -4 +

     .:                                                                                                                                                                                     -2 t

1 {o THE SEARCH FOR OUTLIERS i i  ! e SEARCH FOR SIGNIFICANT VULNERABILITY i FIND OUTLIERS NOT NECESSARILY GUANTIFY INLIERS 1 l 1 i e INDIVIDUAL PLANT EXAf;If1ATION l UNLESS ALREADY DONE l - IDENTIFY OUTLIERS BACKFIT AS APPROPRIATE . 1  ; l0 i e WHERE TECHNICAL ISSUE G0ES BEYOND CURRENT REGULATORY

REGUIREMENTS l -

GENERIC RULEMAKING PREFERRED ALSO USE BULLETINS, ORDERS OR gel 4ERIC LETTERS  : l I i i { jO ! . A-w t - _ _ _ . . _ . . _ _ _ _ . _ _ . _ _ _ _ _ _ _ . - - _ . , _ _ . _ _ _ _ . _ _ _ . . _ _ _ . . . _ . . _ _ _ _ , _ . _ _ _ _ _ _ . _ _ _ _ _ -

                                                                                                               3 GDC 16:

CRITERION 16 - CONTAINMENT DESIGN. AN ESSENTIALLY LEAK-TIGHT BARRIER AGAINST THE UNCONTROLLED RELEASE OF RADI0 ACTIVITY TO THE ENVIRONMEi4T AND TO ASSURE THAT THE C0i1TAliiiiEllT DESIGN CONDITIONS IMPORTANT TO SAFETY ARE NOT EXCEEDED FOR AS LONG AS POSTULATED ACCIDENT C0iiDIT10iiS REQUIRE." 1 GDC 50: O CRITERION 50 - CONTAINMENT DESIGN BASIS.

                                                                                       --AS REQUIRED BY SECTION 50.44, ENERGY FROH METAL-WATER AND OTHER CHEMICAL REACTIONS TifAT MAY RESULT FROM T;EGRADATION BUT NOT TOTAL FAILURE OF EMERGENCY CORE COOLING FUNCTIONING, (2) THE LIMITED EXPERIENCE A!iD EXPERIMEi3TAL DATA AVAILABLE FOR DEFINING ACCIDENT PHENOMENA AND CONTAltiMENT RESPONSES, AtlD (3) THE CONSERVATISM 0F THE CALCULATIONAL H0 DEL AND INPUT PARAMETERS."

A-44

i

                                                                                                                                                                                     ... M.x-TABLE 1 - U.S. 8WR PLANT-SPECIFIC PRA STUDIES
                                                                                                                                                                                    '.; 44g 1

J REPORT CORE / REACTOR CORE-DAMAGE EVENTS MEDIAN, CONTA100ENT ! PROGRAM YEAR CONTAlteENT POWER (InfT) FREQUENCY PRA CONSIDERED MEAN OR CONDITIONAL PLAMI MPME REPORT ESTINATE POINT FAILURE I ESTIMATE PROBABILITY i. 3x10

                                                                                                                                        -5       Internal /      Median          Not evaluated Peach                      RSS         WASH-1400       1975      BWR-4/MK I       3293 Botton                                                                                                    External 4x10
                                                                                                                                        -5       Internal         Mean           0.2 Peach                      IDCOR       Tech Summary 1984          SWR-4/MK I      3293                                      ,
                                      -Botton                             Task 21                                                         5

,' Peach IPE IPE 1986 BWR-4/MK I 3293 2x10 Internal Mean Not evaluated Bottom

                                                                                                                                        ~4

! Millstone IREP NUREC/CR 1983 BWR-3/MK I 1727 3x10 Internal Median Not evaluated l 3085 4 .! Millstone NUSCO Millstone 1 1986 BWR-3/MK I 1727 5x10 Internal Mean Not evolueted PSS

                                                                                                                                        ~4 i                                   Braun. Ferry IREP                      NUREC/CR        1982       BWR-4/MK I      3293          2x10          Int M             Pold          MmW 2801                                                                                 Estteete 3x10
                                                                                                                                        -5       Internal          Mean          0.07 Verment                    VYCSS       VYCSS           1986       WWR-4/MK I      1593

> iankee

'                               bg Ble Rock                   Conseners   Big Rock        1981        BWR-1/ Dry     158           1x10
                                                                                                                                        -3        Interne1/        Mean            0.25 g      Fotnt                              Point PRA                                                       3 External 3i                         EG86/SNL    EG&G-EA-        1982        BWR-1/ Dry     158           1x10           Internet/        Mean            0.25

{ kRock int 5533 Rev. 1 Externe 1 . 7x10

                                                                                                                                        -5        Internal /       Mean          1.0 Limerick                   PEPCO      Limerick PRA 1981           BWR-4/MK II    3293
 ;                                                                                                                                                External Limerick                  ONL         IIUREG/CR-   -

1983 gWR-4/NK II 3293 1x10'4 Internal / Mean 1.0 j 3028 External Shorehme LILCO Shoreham PRA , 1983 gwit-4/MK 11 2436 5x10-5 Internal Pelat Not evalented i Estimate i Shorehan ONL IIUREC/CR- 1985 BWR-4/MK II 2436 1x10 4 Internal Point last evalented 4050 Estlante Shoreham IPE shorehen IPE 1986 SWR-4/MK II 2436 8x10 5 Internal Mean Met evaluated

                                                                                                                                         ~7 Sisquehenne                IPE         IPE            1986        BWR-4/MK !!    3293           2x10           Internal        Mean           Itetevaluate(

Grand Gulf RSSMAP NUREG/CR- 1981 OWR-6/let III 3833 4x10-5 Internet Itedien Isot evolunted 1659 Grand Gulf Istet Tech Summary 1984 BWR-6/MK III 3833 8x10

                                                                                                                                         -6        Internet        Moon           Itet evolueted j                                                                       Task 21               -

j CESSAR GE GESSAR 11 PRA BWR-6/MK III 3579 4x10 Internal / Moen last evaluated External

 }

l

    .<                                                                                             7 i

l KEY RESULTS FOR BWR CONTAINMENTS i l l l e. REACTOR SAFETY STUDY - PEACH EDTTOM 90% EARLY RELEASE e IDCOR - PEACH BOTTOM 20% EARLY RELEASE e VERMONT YANKEE - 7% EARLY RELEASE i O

e NUREG-1150 ,

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s , 1 d

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k,'$,5 j e i 1 - ,[ 1 g o' gj. g [ ( PRIMARY gygygg , DRYWELL l j_ Nr%PTk dj [ : , i = f

               $ PEDESTAL+ C AVITY W ' M                         '
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TYPICAL MARK 1 CONTAINMENT DESIGN: l ll FIGURE 1 'O^ - A -41

  - -___.-_         -                    - _ . . - .             . - . _ - . _ -                                       _-                  ~    _ _ _ _ - -

10 BWR C0i4TAINMENT ISSUES - IMRK I , e SMALL OLUME MORE RAPID DVERPRESSURE ESPECIALLY VULNERABLE TO HYDROGEN BURN e SIMLL DRYWELL FLOOR LOWER HEAD AREA CLOSE TO DRYrr.LL WA1.L l POTENTIAL FOR DIRECT I)EERIS ATTACK DIRECT RADIATION AND CONVECTIDN HEATING O

e LIMITED PASSIVE CAPABILITY BUT QPIl0ES FOR ACTIVE RESPONSE e 5-ELEhENT APPROACH HYDROGEN CONTROL SPRAY IN DRYWELL PRESSURE RELIEF DEBRIS CONTROL i -

PROCEDURES AND TRAINING i i t

O A-70  ;

i  ! I l

  .                                                                11 CONTAINMENT IMPROVEMENT STRATEGY e PREVENT HYDROGEN COMBUSTION BY INERTING                         ,

e REDUCE DRYWELL SPRAY FLOW RATE PERMITS ALTERNATE SUPPLIES TO PRODUCE SPRAY EXTENDS WATER SUPPLIES e PROVIDE RELIABLE BACKUP SUPPLIES FOR DRYWELL SPRAY i PROVIDES SHALLOW POOL OF WATER ON DRYWELL FLOOR DIRECT SPRAY C00LIliG 0F ANY CORE DEBRIS LEAVING LOWER HEAD AREA SPRAY SCRUBBING OF DRYWELL VOLUf'I DIRECT COOLING OF WALLS , e WETWELL PRESSURE RELIEF TO STACK POOL SCRUBBING - ELEVATED RELEASE e DEBRIS CONFINEMEllT f a TRAINED OPERATORS i O A-7/

r " n [g }"" 13 l u;i ;M3mi* mM m m n:g a:: jSiO E c i'::::;:: 3 mas um3-t g;4

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                                                            =: :~~~
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                                                                                                                                                                                                      = = -- - _

FIGURE 2 _ l- _ g -. 1 s EXPECTED WHOLE BODY RADIATION DOSE (REM)- =- i a N

                                                                                       =,

i -i 4 i i i d l e 5 5 i 2 i 4  ;  ? M M= FROM RELEASE OF 1005 N0BLE GASES ns = ee

                                                                          =      -

s ee g. =

                                                                                      -               =
                                                                                                              = =
                                                                                                                    =.

1: :  :  : (I HOUR DECAYED AND 5 HOURS DURATION OF

                                                                                                                                                                                                                                                                                                                                                                                                         -+

j .- . RELEASE) FROM 3412 MWt LWR VS. DISTANCE ei

                                                                                                                                                       -  -                                                                                                                                                                                                            - - =
                                                                                                                                                                                                                                                                                                                                                                       - s;

_ _ :j {_ _ - a . . . NOTES: - - E E I I E I E E E E I lillll';;;;;;

;                                                                                                                               1. Graphs assume one hour holdup and decay prior ll l ' l l l l l l l l l to release. Greater delay in release can produce l                    1000. 3!m M : 7                                                                        1
                                                                                                                  ! lower doses (e.g., as much as a factor of about                                                                                                                                                                                                                                                   nio
!               :           .H iin = s Q 30 at one alle for 12 hours of inreactor holdup                                                                                                                                                                                                                                                                                                                         -

i l g 7 e m e compared to one hour). i: _ _=- 5-_

                                                                                                                                                                                                                                                                                                                                                                                                       =
                                                                                                                                                                                                                                                                                                                                                                                                                             )
                                                                                                                                                                                                                                                                                                                                                                                     -s                                      ;

d +

                            =~                                                                                                                                                                                                                                                                                                           98-5                                  :

j w H M; 2.' Dose estimates are based upon MCCS computer " ! id of .- JU lls codecalculationsusingrevised(relativeto m-1 i E 2 1

                                                                         ==

1 5 _ 3,

                                                                                         .--. CRAC and CRAC2) meteorological sempling models. - - =i;_ _.                                                                                                                                                                                                       =

a s-i - a === _ g-i I5 =y~

3. The likelihood of exceedinq the =i.

estimated 95 percentile dose ; s less ~ ~

=.

l E! l than 55 given release of 1005 of Ai noble gases as'shecified above. g

                                                                                                                                                                                                                                                  '                                         '"                                                                                                                        i      !
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I I I W j E" 'D R H 'E ! ! ! l l l E l I I l l ' 9 HH' H H'44EEEEEEEEEEEEEE [ ' f $$ N )$$ 23! [i OM i = 2 -

                                                                                                                                    =          =

TOM Nyhdh 10 METER $= l g; - a, 6 _'l ._ . _ _ ~_ ) g _ -_ - _  : - g - WW g h * =a F E t n ,- d  ; $ 4 1 n e s  : u -

                                                                                                                                                                                           =

m 3 .. . . , 2 -

                                                                                                                                                                                                                                        , ESTIMATED 9'5' PERCENTILE y           ;

si 8 g 4 ~: ? 9I? I--0 N.  : .:  : 4 i j - ELEVATE) Ri1 EAST 100 METERS i -}.'g iis {=-E E .iW gg p - ;=_ ..- _J Q  : :  :

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                                                                                                                           = =                 == n
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                                                                                       ' ELEVATED                                                         RELEA$d                                  100              METERS"-            EE          i-
                                                                                                                                                                                                                                                                                                                                   ~;" g ;                                                 _ __

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                                                                                                                                                                                                                                                                                                                     '!! '                         ii = Hi 5 iE

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y,/- w:n:: ::: ::: :::  ;-: :::  :: :: - - '-- - - - - w< a l } jjj jj ll"lll jj! lll" D! STANCE FROM RELEASE P0 INT l ";

. n;n;iis i um nen s us ;;(MILES) i is i s i >.-[

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                                                                                                                                       ~*                              5 ~6                                                   7~a                                                   9                    e                 it

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16 O CONDITIONS-QUALITY AND DESIGN STANDARDS SINCE THESE REGUIREMENTS ARE INTENDED TO BE AN OPTIMIZED USE OF EXISTING EQUIPMENT IT IS EXPECTED THAT ADDED EQUIPMENT, 0F ITSELF, NEED NOT MEET THE QUALITY OR DESIGN STANDARDS OF SAFETY RELATED EQUIPMENT. NEVERTHELESS, MODIFICATIONS TO OR NEAR EQUIPMENT OR SYSTEMS WHICH ARE ALREADY SAFETY RELATED SHALL NOT COMPROMISE THE QUALITY OF SUCH EQUIPMENT OR SYSTEMS. IMPLEMENTAT1011 THE EQUIPhENT CHANGES REQUIRED HEREIN SHALL BE INSTALLED DURING THE FIRST REFUELING OUTAGE WHICH EEGINS NINE (9) MONTHS AFTER THE EFFECTIVE DATE OF THIS LETTER. THE PROCEDURES AND TRAINING REQUIRED SHALL BE IllPLEMENTED ON A SCHEDULE REVIEWED AND APPROVED BY THE NRC. GIVEN THE liiPLEMENTATION OF THE GENERIC IMPROVEf1ENTS OF MARK I CONTAINMENTS THERE IS NO HEED FOR AN INDIVIDUAL PLANT EVALUATION (IPE) FOR CONTAINMENT PERFORMAllCE. THIS DOES NOT REMOVE THE NEED FOR AN IPE WHICH COVERS THE SYSTEM RELIABILITY OR CORE MELT FREQUENCY PORTION OF THE SEVERE ACCIDENT QUESTION.

A d 6 ,s - 17 e j SEVERE ACCIDENT POLICY STATEfENT 1 e OPERATING NUCLEAR POWER PLANTS REQUIRE NO FURTHER REQULATORY' { !; ACTION TO DEAL WITH SEVERE ACCIDENT ISSUES UNLESS SIGNIFICANT  ! l NEW SAFETY INFORMATION ARISES TO QUESTION WHETHER THERE IS ADEQUATE ASSURANCE OF NO UllDUE RISK TO PUBLIC HEALTH AND ] hl SAFETY. f e IN THE LATTER EVENT, A CAREFUL ASSESSMENT SHALL BE MADE OF THE l SEVERE ACCIDENT VULNERABILITY POSED BY THE ISSUE AllD WHETHER l l THIS VULNERABILITY IS PLANT OR SITE SPECIFIC OR OF GENERIC IMPORTANCE.

O  !

j e THE MOST COST-EFFECTIVE OPTIONS FOR REDUCING THIS VULNERABILITY l I SHALL BE IDENTIFIED AND A DECISION SHALL BE REACHED CONSISTENT WITH THE COST-EFFECTIVENESS CRITERIA 0F THE COMulSS10N'S ) BACKFIT POLICY AS TO WHICH OPTION OR SET OF OPTIONS (IF ANY)  : ARE JUSTIFIABLE AND REQUIRED TO BE IMPLEllENTED. ( l - ! , i e IN THOSE INSTANCES WHERE THE TECHNICAL ISSUE G0ES BEYOND CURRENT REGULATORY REQUIREMENTS, GENERIC RULEMAKING WILL BE THE PREFERRED l SOLUTION. IN OTHER CASES, THE ISSUE SHOULD BE DISPOSED OF  ! l l THROUGH THE CONVENTIONAL PRACTICE OF ISSUING BULLETINS AND ORDERS OR GENERIC LETTERS WHERE MODIFICATIONS ARE JUSTIFIED THROUGH BACKFIT POLICY, OR THROUGH PLANT-SPECIFIC DECISION-MAKING ALONG THE LINES OF THE INTEGRATED SAFETY ASSESSMENT l PROGRAM (ISAP) CONCEPTION. A.79 l l

,..' . 16 l 00ffilSS10N RESPONSE TO A HEARING QUESTION JULY 16, 1986 4 ? ! QUESTION I IS A 90 PERCENT CHANCE OF FAILURE IN THE EVENT OF A CORE MELTDOWil All ACCEPTABLE FAILURE RATE 7 . ANSWER THE NRC HOLDS THE POSITION THAT THE LIKELIHOOD OF CORE MELT ! ACCIDENTS IN ANY PLAfiT SHOULD BE VERY LOW AND, IN ADDITION, THAT THERE SHOULD BE SUBSTANTIAL ASSURANCE THAT THE CONTAINMENT i WILL MITIGATE THE CONSEQUENCES OF A CORE MELT SHOULD ONE OCCUR i

IN ORDER TO ENSURE LOW RISK TO THE PUBLIC. IT IS NOT MERELY A QUEST 10fl 0F HAVING LOW RISK BUT OF HAVING AS WELL THE DEFENSE-4 IN-DEPTH ASSURAf1CE OF C0lBINED PROTECTION BY PREVENTION AND filTIGAT10N...

i A-fr i

a 18A ' 10 - TABLE 3 COST-BENEFIT ANALYSIS COST: $0.7-2.2H BENEFIT:(1) F'CM CCFP CCFP AVERTED AVERTED BEFORE AFTER LOSS /YR LOSS PRES. VALUE BASE I CALCULATION 1x10'4/yr 0.5 0.05 $4x105 /yr $3M/$12M

                                                                              -5/yr LOWER FCM                                       1x10                        0. 5                      0.05                   $4x104 /yr                        $0.3M/$1.2M l

r LESS CHANGE I ~4 IN CONTAINHENT 1x10 /yr 0.5 0.1 $4x105 /yr $3M/$12M j BETTER , CONTAINMENT

                                                                              ~4 TO START                                         1x10                        0.2                       0.05                   $2x105 /yr                        $2M/$6M
                         OPTIMISTIC"                                                                                                                                                                                     i
                                                                              -5 f                        CALCULATION                                      1x10                        0.2                       0.05                  $2x104 /yr                         40.2M/$0.6M i

, " PESSIMISTIC" CALCULATION 3x10'4 0.9 0.1 $2x105 /yr $16M/$60M F (1) FCM = Frequency of Core Melt CCFP = Conditional Containment Failure Probability i' AVERTED LOSS PRESENT VALUE expressed as A/B where A is the averted loss - per year times 8 (roughly equivalent to discount at 12X/yr rate) and B is/ -// the averted loss per year times 30 (no discount).

                                                                                                                                                                                                                         /

i

      .--.,,.-,_,-----c       -. - - - . . . - - - - - , . - - - - - - ,                     - - - -       - - - - . . - - - -      . - - - - - . . . - - - , , - - - , , - - - , . ,           . - - - .-,, - - , - - -

      *'                                                                                                                      19 O.

• PROPOSED ACTION j e DEC. 9 a 12, 1986 ACRS REVIEW l

l e DEC. 19, 1986 CRGR REVIEW e JANUARY 1987 REVIEW 0F ACRS AND CRGR REACTION WITH l C0fEISSION i e FEB 1,1987, PUBLISH PROPOSED GENERIC LETTER FOR C0ffiENT .i e MAY 1987 ISSUE FINAL GENERIC LETTER l f l j SIMILAR LETTERS ON MARK II AND MARK III T0' FOLLOW I , l i i l t

                                                                                           #7/                                   -

i _ . _ . _ _ . _ __ _ _ _ _.___.__ _ _ _ _ _ ._.-_ _ __ _ __ _ _ _ _ _ _ _. _ _ . ~ . _

APPENDIX VIII SEVERE ACCIDENT C N W NMENT ISSUES BWR OWNERS' GROUP j SEVERE ACCIDENT CONTAINMENT ISSUES NUMARC APPROVED SEVERE ACCIDENT CONTAINMENT ISSUES APPROACH q 1. OBJECTIVE i EVALUATE CONTAINMENT INTEGRITY. IF APPROPRIATE, ASSESS

POTENTIAL IMPROVEMENTS TO MINIMIZE OFFSITE RELEASES FOR SEVERE ACCIDENT CONDITIONS (BEYOND DBA) WITHIN AN APPROPRIATE COST / BENEFIT GOAL.

2. IDENTIFY CHALLENGES TO CONTAINMENT

• 1.)

                                                                               ~

H2 GENERATION i

)

2.) OVERPRESSURE 3.) TEMPERATURE 4.) CORE DEBRIS ATTACK , 5.) FISSION PRODUCT CONTROL I ) 6.) HUMAN ACTIONS  ! 14 DIRECT CONTA!NMENT HEATING . . . _ _ . .. . . ,

q. .

3. IDENTIFY INITIATORS TO EACH CHALLENGE FROM EXISTING ANALYSES ,

PICK KEY EVENTS / INITIATORS (MOST SEVERE - LESS SEVERE) SEQUENCE THAT PRODUCES MOST SEVERE CHALLENGE l

4. ASSESS PLANTS' ABILITIES TO MEET CHALLENGES

5. ASSESS PLANT VULNERABILITIES

6. PROPOSE ALTERNATIVES TO ADDRESS VULNERABILITIES l

7. EVALUATE ALTERNATIVES i 8. REACH DECISIONS O

• OTHER ISSUES TO BE ADDED g OG37/12.12

      - a              - - . . -               .     . - _ -                        --              . - - - . - - - -

a- -~ a . . - . . - - - __&--m_ _. l 4 S O APPENDIX IX POLICY STATEMENT ON DEFERRED PLANTS POLICY STATEMENT ON O DEFERRED PLANTS O A-11

4 i DEFINITIONS 4 , DEFERRED PLANT

                               - CP IS IN EFFECT
                               - CONSTRUCTION CEASED OR REDUCED TO MAINTENANCE LEVEL i
                               - LICENSEE HAS NOT ANNOUNCED TERMINATION O                        TERMINATED PLANT                                            ,

- CP IS IN EFFECT i - LICENSEE HAS ARNOUNCED THAT CONSTRUCTION HAS BEEN TERMINATED PERMANENTLY I CANCELLED PLANT l - A PLANT WITHOUT A VALID CP

l 1 ' ([$3) ! A-80  :

I l i

DIRECTIVES ON DEFERRED / CANCELLED PLANTS i

• PORTIONS OF THE 1985 POLICY AND PLANNING GUIDANCE. ITEM IV.B, PLANNING GUIDANCE 2
• REQUEST FOR PROCEDURES TO REACTIVATE A PROJECT AFTER CONSTRUCTION AND LICENSING HAVE STOPPED
• COMMISSIONER ZECH'S REQUEST
• SEVERE ACCIDENT AND STANDARDIZATION POLICY STATEMENTS i

i O

                                                                       /         -lf /

4

r , POLICY STATEMENT CONSIDERATIONS 1

                           ,

• MAINTENANCE, PRESERVATION AND DOCUMENTATION

                   )        i    REQUIREMENTS FOR DEFERRED PLANTS
             }V
              /.
                         /

• APPLICABILITY OF NEW REGULATORY STAFF POSITIONS h )! FOR DEFERRED PLANTS BEING REACTIVATED v
• PROCEDURES FOR REACTIVATING DEFERRED PLANTS
• IDENTIFICATION OF REGULATORY IMPROVEMENTS AND l 8 ) RESEARCH INITIATIVES cg>
• POPULATION AND STATUS OF DEFERRED'AND TERMINATED PLANTS b

A-82 i

7_..._.__ i t Le i i i l  : I } i SCOPE l i  ; I i i' I t  ! a 4 -

• PROCEDURES DEVELOPED FOR DEFERRAL AND REACTIVATION l 0F DEFERRED PLANTS .

!

• DOES NOT COVER REACTIVATION OF CANCELLED PLANTS  :

{ t l -t i  ; i i i, ! l i l l O 4 k 13

                                                                                                                                                                                           -l

i r . ELEMENTS OF POLICY l I 1 l MAINTENANCE, PRESERVATION AND DOCUMENTATION REQUIREMENTS IDENTIFIED t

                                           - APPLICABLE REGULATIONS - 10 CFR PARTS-21, 50.55, 50.71, 50 APPENDICES A&B                                                                                   l

l

                                           - APPLICABLE GUIDES - R.G. 1.28, 1.37, 1.38, 1.58, 1.88                                                l AND 1.116 2
                                           - INSPECTION PROCEDURES AS APPROPRIATE l

l 4 4 O A-74 1

  ,-        ,---,-v---         . - , - . .       , , - - -                   - - - , - . _ _ . .      .---.,.           -
                                                                                                                          --r..-,, - - . .. . - .

1 0 ELEMENTS OF POLICY (CONTINUED)

• APPLICABILITY OF NEW REGULATIONS DURING DEFERRAL

                    - PLANT-SPECIFIC BACKFITS OF NEW STAFF-POSITIONS WILL BE CONSIDERED IN ACCORDANCE WITH BACKFIT RULE -

10 fr" DART 50.109

                    - GENERIC BACKFITS WILL BE IMPLEMENTED EITHER THROUGH RULEMAKING OR GENERIC ISSUE RESOLUTION i
                    - PROVISIONS OF OTHER POLICY STATEMENTS APPLICABLE

. TO PLANTS UNDER CONSTRUCTION WILL BE IMPLEMENTED  ; a

O A-Kr

O PROPOSED REGULATORY IMPROVEMENTS 4

• REGULATORY CHANGES MAY BE NEEDED T0:

            - ESTABLISH THAT CP REMAINS IN EFFECT (EVEN IF IT EXPIRES) UNTIL NRC WITHDRAWS CP

• INTERNAL NRC GUIDANCE ON SPECIFIC INFORMATION NEEDED FOR TERMINATION OF CP
• DETAILED GUIDANCE FOR INSPECTION OF DEFERRED PLANTS PRIOR TO REACTIVATION f

O ~ A-74

O AIF PARTICIPATION

• STAFF MET WITH AIF, FEBRUARY 19, 1986
• AIF PROVIDED LETTER, MARCH 31, 1986 ON " REACTIVATION OF CONSTRUCTION PROJECTS"

]

                     - AGREES THAT POLICY STATEMENT IS NECESSARY       ,

i

                     - POLICY SHOULD APPLY TO PLANTS WITH cps
                     - PLANTS WITH WITHDRAWN cps SHOULD BE HANDLED ON A CASE-BY-CASE BASIS
                     - 10 CFR PART 50.109, BACKFIT RULE, SHOULD BE USED TO IMPLEMENT NEW REQUIREMENTS IN EFFECT FOR PLANTS WITH cps                                                     i l
                     - CURRENT PRESERVATION REQUIREMENTS ADEQUATE AS LONG AS

~ RECORDS MAINTAINED.

• REVIEW AND COMMENTS ON FINAL DRAFT, JULY 2, 1986 i A -If

  . - -          .          .                     .     .       _-       -_   a

   . -.-...._ ---... _ ...~._ - - -.. - - -.                          , -  _-. . .  .  . - . . - _ . - - - - .

I i. i i L- != !t 1 l l l i u l ! POLICY OVERVIEW l i i  !

.t

]- t i i i  ; i

                                             - NO-SAFETY IMPLICATIONS                                             l i

i ! - ESSENTIALLY CONSOLIDATES EXISTING REQUIREMENTS E i 4

                                             - COORDINATED WITH INDUSTRY                                        .;

l l 1 1 1 l l 1 l I I 1 1 4

    ~

1 i i i@ , A-f i i

    ~ - - . - - . . . _ . . . . . . - . .                       - _ . - . .-. -                              ,         .     . - . - .      -   - - - .         . . . .        . . . . . .

l

                                                                                                                                        . APPENDIX X                                   '   '

PROPOSED-. FINAL PTS REGULATORY GUIDE a , , I 1 t r f t . PROPOSED FINAL PTS REGULATORY GUIDE l 1 ACRS REVIEW MEETING . 1 i  ! ! DECEMBER 12, 1986 l t I g ! r

i i i I

i i i I L i . i I 4 l J !' t l I I , 4 A -77 t

c

                                                        ~        ~

[ ) i

BACKGROUND PTS RULE PROMULGATED 7/23/85 (10CFR50,61)

REQUIRES EXTENSIVE ANALYSES 3 YRS, BEFORE RT SC GUIDE PROMISED IN RULE

                     - PUBLISHED FOR PUBLIC COMMENT 1/17/86 3
                     - A,CRS METAL SUBCOMPONENTS SUBCOMMITTEE MET TO PROVIDE f

j COMMENTS ON 02/28/86 l( ) - ACRS FULL COMMITTEE PROVIDED COMMENTS BY 03/18/86 LETTER i - PROPOSED FINAL GUIDE NOW BEFORE YOU ! -CONSIDERS PUBLIC COMMENTS 4

                              -CONSIDERS RECENT NRC-STAFF AND CRGR COMMENTS                                !

l l' . l l . A -to

3-i l PUBLIC COMMENTS I r i SIX LETTERS, ONE STAFF MEMO RECEIVED 2 FROM UTILITIES (VEPC0, YANKEE ATOMIC ELECTRIC) l 3 FROM WESTINGHOUSE - 2 FROM PNL (OUR CONTRACTOR) y 1 FROM ACRS CONSULTANT (DR, CATTON) MEMO FROM DR. D. ROSS, RES ' { 4 DETAILED COMMENTS AND RESPONSES! ATTACHED TO 10/07/86 J. FUNCHES~TO R. FRAGY MEMO i i I A-9/ 4 r p.-eu w- w- ---n.,,----n-,--rw--- ,-e- , - r,- ,-,,---~n,--w- -.,n,,,,, ,,,,,,,wa,w. v--n.e,- ~,w-- -- ,.,-,---,n n ,, ,w , e r,w- m a -,-, we-,.w w m-m,,--

14-PUBLIC COMMENTS THAT RESULTED IN NO RECOMMENDED CHANGE PTS RULE NOT ADEQUATELY CONSERVATIVE NOT RELEVANT TO REGULATORY GUIDE GUIDE IS PREMATURE AND INCOMPLETE, NO ONE NEEDS NOW, WAIT

                        - (CRITERIA AND METHODS EVOLVING)
                        - (T-H CODE SPECIFICATIONS ARE TOO LOOSE)
                        - (NODALIZATION SCHEMES AND ERROR ESTIMATION TECHNIQUES ARE NOT SPECIFIED)
                        - GUIDE SHOULD ALLOW FLEX EILITY                                                                                      f
 ,                      - DO NEED NOW - SEVERAL PLANTS ~COULD EXCCED RT SC DBA APPROACH SHOULD BE ALLOWED NO DBA COVERS ALL SEQUENCES
                        - NEED PRA TO EVALUATE FIXES ONLY DETERMINISTIC FRACTURE MECHANICS (F.M.) ANALYSES SHOULD BE REQUIRED "WHERE NO CRACK IS POSSIBLE"
                       - CRACK ALWAYS "POSSIBLE" NO DETERMINISTIC F.M. ANALYSES SHOULD BE REQUIRED (ONLY PROBABILISTIC F.M. NEEDED)

NEED DETERMINISTIC F.M. TO DETERMINE PRA PARAMETERS i A-1A.

   - - - , .                      -                  . . - - ,       ,_.--,,---._----.-r-,-    - - ,   ,-                     . , . - , , ,

v PUBLIC COMMENTS THAT RESULTED IN CHANGES 1 CONDENSATION OUTSIDE PRESSURIZER

                - NOW " REQUIRED" BY GUIDE PRESSURE CONTROL
                - NOW LISTED AS POTENTIAL CORRECTIVE ACTION CIRCUMFERENTIAL FLAWS
                - 300*F RT PTS NOW STATED TO BE BOUNDING, NOT CALCULATED EFPY, FLUENCE, AND RT PTS
                - NO LONGER SUGGEST USING EFPY TO CHARACTERIZE EXPOSURE CERTAIN MATERIALS ARE NOW OK TO INCORPORATE BY 

REFERENCE:

i

                - PORTIONS OF UNCERTAINTY ANALYSIS
                - PROJECTED FLUENCE CALCULATIONS
                - T. H. CODE BENCHMARKING

! - OPERATOR GUIDANCE, TRAINING I O A-13

I i.

I

!- l i 3 STAFF ~ COMMENTS THAT RESULTED IN CHANGES . l l GUIDE OVER-EMPHASIZED MITIGATION ! - STAFF WISHES TO EMPHASIZE PREVENTION

                                      - DELETED CHAPTERS REGARDING:

l j VESSEL. FAILURE MODES' CONTAINMENT PERFORMANCE .i PERSON-REM  !

                                          .                                                                                                t

) , 4 ' i 1 GUIDE PRESENTED TO CRGR (AND TRANSMITTED TO ACRS). . I l PROVIDED THAT MATERIAL BY REFERENCE i  ! . I I i f A. e 1 1 I l l I I A k * !! il ..

O

                                                                          ~     ~

CRGR COMMENTS AND RESULTING CHANGES 1

                               - PTS RULE REQUIRES VESSEL FAILURE PREVENTION ANALYSES, DOES NOT PROVIDE FOR " RISK" ANALYSES WE HAVE NOW DELETED THOSE REFERENCES GUIDE NOW SAYS SUCH ANALYSES A) MIGHT BE SUBMITTED, ON APPEAL, TO COMMISSION, BUT B) ARE BEYOND SCOPE OF REG. GUIDE CORRECTIVE ACTIONS ~MUST BE' EVALUATED CONSIDERING OVERALL                                                 ,

PLANT SAFETY, NOT JUST PTS l t Y FURTHER WORDS ADDED THROUGHOUT TO FURTHER EMPASIZE 4 1 i ) i

O  :

1 A-15

        ,--.-,-,w       -
                                     ,-c+--.v,,         -,----w,e---   ,,   ,          -     -,e-,,  ,r,..-      ,~r-~,--,en.e--,--w,-,-w--

i. l j

i i

SUMMARY

AND RECOMMENDATION

                                                                                                                                                    ]

! l i- GUIDE NOT CHANGED SIGNIFICANTLY. -l  ; t i RECEIVED BROAD ACCEPTANCE . I l i NEEDED TO COMPLETE PTS ISSUE

• TAG '4"N i

i , i ' i

• REQUEST ACRS LETTER RECOMMENDING ISSUANCE i

0F FINAL PTS REGULATORY GUIDE iO I I 2 i I f A-94 l 4 i

                                                                                                               +

4 N i - BACKUP - 1 b i STATUS OF PLANT RT PTS REVIEWS 4 , i 40 SERS ISSUED ~ j 31 - NO FLUX REDUCTION 9 - FLUX REDUCTION COMMITMENT ~ t 1 SER DRAFTED

!                  - NO FLUX REDUCTION 1                                           .

6 REVIEWS UNDERWAY AT BNL - i

                                                     ~

l LARGE REDUCTIONS PROPOSED BNL CHECKING RESULTS

16 ADDITIONAL SERS TO BE ISSUED BY 2/28/87 1 l

j (63 TOTAL) NOTE: FLUX REDUCTIONS TO BE VERIFIED i 1 \ l 1 l i I l A-97

     ... . . - - . ~ . -.                                  c. .- . ,-- . .--..-       -              . . _ .   . . -.- - - . -.~. - _.               ..-. _ _

i, . 1 APPENDIX X , PROPOSED FINAL PTS REGULATORY GUIDE i I , i a  ; i i. i . PROPOSED FINAL PTS REGULATORY GUIDE ACRS REVIEW MEETING , i DECEMBER 12, 1986 i 4  ! i 4 i i i I i 1 i i i i i a i i 1

                                                                                                                                                                               \
                                                                                                             ~

l l

                                                                                                                                                                               )
   ,r.,n.--             , , - , . -,. - . , - , . . - , - ,                n,-,--,-,-           . . - . .                              _.                        . _ . - - -

APPENDIX XI [oan. %, UNITED STATE: NRC STAFF RESPONSE ON NUCLEAR REGULATORY COMMISSION g(V{! g wasacTow.o.c.nossa SHEARON HARRIS APPLICATION s

                   %**...                                           December 10, 1986 Docket No. 50-400                                                                                              ,

MEMORANDUM FOR: Raymond F. Fraley, Executive Director Advisory Comittee on Reactor Safeguards (ACRSI FROM: Lester S. Rubenstein. Director PWR Project Directorate #2 Division of PWR Licensing-A

SUBJECT:

RESPONSE TO ACRS LETTER DATED JANUARY 16, 1984, j IN REGARD TO THE SHEARON HARRIS APPLICATION The ACRS report dated January 16, 1984, on the Shearon Harris Nuclear Power Plant, contained a number of recomendations to the staff and applicant. It also requested the staff to investigate the ellegations described in Mr. Wells Eddleman's letter dated January 13, 1984, and to proviJe a written repnrt to the Comittee. The purpose of this letter is to describe the manner in which , O. the staff and applicant have addressed the above issues. i The specific recomend3tions delineated in the ACRS letter were: 1

1) The ACRS requested that it be kept infomed of the control room habitability evaluation perfomed by the applicant during the operational test of the

       ~

control room emergency air recirculation ~ system.

2) The ACRS wants to be kept informed regarding the operating experience of the Westinghouse D-4 steam generators relative to the tube degradation problem.

       #                3)         The ACRS believes that written evidence of an improvement in the licensee's nuclearoperation(i.e.,thetwoscheduledSystematicAssessmentof Licensee Perfomance (SALP) reports) should be, available prior to full power operation.

4) The ACRS requested the staff to investigate Mr. Eddleman's allegations and provide a written report to the Comittee.

5) The ACRS recomended that, in addition to items already considend, I ' specific attention be given to assurance of adequate seismic capability of the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines that are itportant to the accomplishment of safe shutdown and decay heat removal. The ACRS also

       '                            suggested that specific attention be given to the adequacy of clearances-between adjacent buildings.                                            -

6) The ACRS would like to receive a detailed discussion of the chilled water system in a supplement to the Safety Evaluation Report.

A- 7f _. __ /

1 l sQ Raymond F. Fraley .

7) Because of the nonoptimum orientation of the turbine relative to vital components in this plant, the ACRS recomended that a structured test program for evaluating overspeed protection of the turbine he prepared and submitted to the NRC staff for review and approval before full power operation.

Following is a discussion of each of the items in the order that they appear above:

1) Enclosed is a letter dated December 2, 1985 (Attachment 1), from Carolina Power & Light Company (CP&L) which is being provided for your infomation.

As indicated, the testing shows that the Technical Specification require-ments were met with some margin and that further testing will be conducted when plant conditions permit.

2) Enclosed is infomation from Westinghouse Corporation (Attachment 2) describing the operating experience of Westinghouse D-4 steam generators.

3) In, January 1984, when the ACRS conducted its 06 review of Shearon Harris, the three SALP reports available at that time indicated several areas

       /7             of perfomance needing improvement by CP&L. Since that time, the V              staff has completed three additional SALP evaluations and issued SALP reports on August 21, 1984, January 15, 1986, and September 25, 1986.

The last report covers only Shearon Harris. As noted in item 4) below, this area was also the subject of extensive review by the Shearon Harris ASLB.

         ~

The results of these last three SALP reports were as follows:

         .            Aucust 21, 1984 (2/1/83 - 4/30/84)
         ;            The overall corperate evaluation was that CP&L's perfomance has changed "from being considered as a poor performer during the previous y            SALP period to a significantly improved utility." Ten functional areas were evaluated at Brunswick I and 2 and Robinson 2, nine areas at Shearon Harris. At Brunswick, improvement was noted in 9 out of 10 areas and 4 of those areas were rated Category J. At Robinson, improvement was noted in 7 areas, and also was rated Category 1 in 4 areas. At Shearon Harris, improvement was noted in 4 areas with 3 Category 1 areas. No functional area at any of these sites was rated Category 3.

January 15,1986(5/1/84-10/31/85) The overall corporate assessment indicated that CP&L "contimled to demonstrate a proper concern for nuclear safety." At Brunswick and Robinson, performance in 6 areas showed an improving trend with none declining. At s j Shearon Harris,13 functional areas were assessed with 6 areas Q improving. Only one area (Electrical Equipment and Cables at Harris) was rated Category 3. A-71 t

.,p. Raymond F. Fraley September 25, 1986 (11/1/85 - 7/31/86) (ThisreportcoveredonlyShearonHarrisandcovered14 functional

                      ! areas. The report states that activities during this period "have.been
                   / conducted in a very professional manner." Major strengths were identified in 5 areas, which were rated Category 1. No areas were rated

( Category 3 and ratings improved in 3 areas. The staff considers that these reports, copies of which have been sent e to the ACRS, provide substantial " written evidence" of CPAL's improved

perfornance as discussed in the ACRS letter.

4) In a June 6,1984 memorandum (Attachment 3), the NRC staff provided an initial response to the Corriittee on allegations contained in Mr. Eddleman's letter to the ACRS. The following is provided on those issues that were not resolved in the June 6, 1984 memorandum and

_ provide our final response to Mr. Eddleman's allegations. The following paragraphs are numbered to. correspond to the June 6,1984 memorandum. _; Paradraph 1 o Management of the Brunswick and Robinson Plants was discussed.before the ASLB in conjunction with management capability issues. The Board's (V) interpretation of Brunswick and Robinson management is contained in their partial initial decision dated August 20, 1985, for SHNPP which resolved the mana 235 ff.)gement . capabilityunder See quotation issueParagraph in favor of2 the applicants (22 NRC at 232, below. Paracraph 2 .

                                                                                                                           ~

As indicated in No. I above, the management capability issue (Joint Contention 1) has been resolved by the ASLB in the applicants' favor. This decision is :'ocumented in the partial initial decision dated August 20, 1985, wherein the Board concurs with the following conclusion

      -r               of the NRC staff (22 NRC at 2571:
                                   "The staff concludes CP&L is technically qualified to operate the Harris facility within the purview of the regulations and with due regard for public health and safety. The Region II inspection and enforcement program will be applied to assure that CP&L continues to operate within the regulations and continues to make improvements in the nuclear program."

Paragraph 3 ~ The issue of pipe hangers (Eddleman Conter. tion 41) was raised before and resolved by the ASLB. The Board ruled in favor of the applicants as documented in the December 11, 1985, partial initial decision which p included the following conclusion (22 NRC at 930): '

                                                                                / -/oD k

                 . . . . . _ . = -
       ^%                                                                                                      ,

d V) Raymond F. Fraley \

                                   "The Board finds CP&L and its contractors had a variety of problems in carrying out the pipe support welding activities. This contention may have had merit when it was initially raised, but remedial actions have averted a possible breakdown in quality construction. No uncorrected errors that would affect safe plant operation were identified in the proceeding."

This ASLB decision has been affimed by the Appeal Board in its August 15, 1986 decision. Paragraph 4 Steam Generators and water hammer considerations have been raised before and resolved by the ASLB as Joint Contention VII and Eddleman Contention 45, respectively. Joint Contention VII included four subparts, the first three of which were decided in the applicants' favor by sunrnary disposition

                - and documented in the Board's December 11, 1985 decision as follows (22 NRC at 931):
                                ** Sufficient experimental data, analyses and testing have been
    ~~

performed to provide assurance that vibration in the Shearon Harris steam generators will not be a major problem. The use of all-volatile treatment (AVT) water chemistry is state-of-the-art and provides reasonable assurance of obviating many of the corrosion, cracking and denting problems experienced in earlier systems. The ' loose-parts' detection system for foreign objects in the steam generators has been tested and found to be capable of. detection of any such objects which might affect the integrity of the steam generators." The fourth subpart was also decided in the applicants' favor and documented 7 in the December 11, 1985 decision (22 NRC at 934). The Board reached the following conclusion:

                                    "The Board has evaluated the record before' us and finds that there is no need for multiple-tube failure to be considered in the FSAR.

It has been established that single-tube failures are rsre and that multiple-tube failures are even more unlikely. We conclude that the testing, design modification, water chemistry procedures, loose parts monitoring and inspection and maintenance procedures which have been or will be implemented should make tube failure even more unlikely than it has been historically. The Board..'therefore, finds that analysis of potential single-tube failure is adequate." . V) c

                                                                      }}-/Of i

Y Raymond F. Fraley - I The Board's resolution of Joint Contention VII was affimed by the ASLAB i in their August 15, 1986 decision. Eddleman Contention 45 regarding , water hamer was decided in the applicants' favor by the ASLR via sumary ' disposition. Paragraph 5 The staff considers that the issue of an inactive fault under the containment is resolved as discussed in our June 6,1984 memorandum to the ACRS. Paragraph 6 The issue of the effects of earthquakes as a source of comon mode #ailure i of all the transmission lines to the Shearon Harris site is discussed in our June 6, 1984 memorandum to the ACRS. We consider this issue to be resolved. Paragraph 7

         ~

The staff considers that the issue of earthquake faulting at the Shearon Harris Site is resolved as discussed in our June 6,1984 memorandum to s the ACRS.

Paragraph 8 No contentions regarding the Physical Security Plan were admitted by the ASLB.

Paragraph 9 Theissueofcontrolroomdesign(EddlemanContention1232C(II))was raised before and resolved by the ASLB. The Board granted sumary disposition in favor of the appifcants' as documented in their August 20 1985, decision (22 NRC at 297) with the following conclusion: ! 'The Board, therefore, finds, based on Applicant's arguments which were corroborated by the Staff, that no material fact exists to be litigated in this proceeding concerningJcontrol room configuration, and the contention is, therefore, dismissed." Chapter 18 in Supplement 4 to the Shearon Harris SER also provides updated infomation regarding the staff's control room design review.

5) i Subsequent to the recomendation made in the ACRS letter, an on-site audit l of the seismic qualification of selected safety related electr.ical and

! / mechanical components was performed as reported in Supplements 3 and 4 to

                      / the SER (Section 3.101 The audit consisted of verification of installation 1 as well as a review of the qualification documents. Although.the items audited were not identical to those recomended by the ACRS, they did include relay cabinets, switchgear, control room cabinets, solid state protection systems, etc. In all the items reviewed, there was evidence of adequate seismic capability.
                                                                                                                      /

h

               ~

x s

                                                                                                                                       \

O Raymond F. Fraley - The peak ground acceleration for the SSE at the Shearon Harris plant is O.15g which is substantially lower than earthquake experienced in Chile in 1985. This earthquake was carefully investigated by the Seismic Oualification Utility Group. A substantial body of earthquake experience data exists to conclude that properly supported equipment and components

have inherent ruggedness to resist earthquake loading substantially greater than that caused by 0.15g peak ground motion. The staff audit provides additional assurance that the licensee's program for equipment anchorage is adequate.

4 In regard to the building clearances, this matter was addressed in i Chapter 19 of Supplement 4 to the SER. i i

6) The results of the staff's review of the chilled water system is /

documented in Section 9.2.7 of Supplement 4 to the SER. In addition, the staff and applicant have made numerous presentations to various ACRS J

                             .subcomittees during the past two years to discuss reliability and fracture toughness of the chilled water system. Transcripts and minutes of these meetings can be provided to ACRS. The staff considers this issue to be resolved.

7) Carolina Power & Light submitted copies of their startup, pre-operational and maintenance test procedures used to demonstrate and maintain proper j calibration of the main turbine mechanical and electrical overspeed trips. These letters are enclosed for your infomation (Attachments 4 i

and 5). The staff has reviewed these procedures and has concluded that i the structured test program for evaluating turbine overspeed protection i  ; is acceptable. The staff's evaluation is also enclosed (Attachment 6). 1 Validation of the vendor's roll-up curves is based upon actual turbine i cycrspeed events from full lond at Salem. The staff considers that i action on this issue is responsive to the ACRS letter and is available to further discuss this testing with the ACRS upon request. I hope that the above discussions have been responsive to ACRS recomendations. [ dwM k Lester . Rubenstein Director PWR Project Directorate #2 Division of PWR Licensing-A i

Enclosures:

l As stated 1

O -

! p-ps , I

I - . . _ _ - . . - . _ . _ , , , _ _ _ _ , _ .,,,.._ _ _ _ , l ATTACHMENT 1 f.s " CP&L Carolina Power & Light Company l DEC 0 219c6 SERIAL: NLS-86-444 Mr. Harold R. Denton, Director . Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20$35

                                                                                                                                         ,~

SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 30-400/ LICENSE NO. NPF-53 CONTROL ROOM HABITABILITY IN EMERGENCY RECIRCULATION MODE

Dear Mr. Denton:

Carolina Power & Light Company hereby submits additionalinformation concerning of the Control Room Area Ventilation System (CRAVS)in the emergency testinglation recircu mode. During CP&L's presentation to the ACR5 on January 12,1984, CP&L indicated that the Centrol Room Ventilation System would be tested in the recirculation mode to determine its ability to maintam 1/4 inch watergauge positive pressure as well as verify the systems cooling capability. In the ACR5' letter to the NRC N dated January 16,1984, J. C. Ebersole to N.'J. Palladino, the ACR5 expressed a desire to be kept informed of the results of these tests. , W) . As part of an operations surveillance test, the system's ability to pressurize the Control Room area was tested in November 1986 with the system aligned in the emergency /8 inch recirculation mode with makeup air, in this alignment, the system exceeded the 1 watergauge pressure criteria relative to adjacent areas as identified in the FSAR and

       -             Technical Specifications

• i.e., effectively > 1/4 inPh (differential cressure between the Control Room and outside atmospheric pressure) pressure at a make-up flowrate

       -             (pressurization rate) < 315 cim. Although system balancing has not yet been completed, as such, should not direct! Impact the system's ability to achieve design d).

balancing, As far as t he system cooling capacit is concerned, the capacity of the CRAV5 cooling coils was verified in conjunction wit a preoperational test performed in July of this . year. During this test, the coolers exhibited greater than a 25 percent margm over the cooler vendor specifications. While this is not a direct measure of the resulting Control Room area temperature, it is a good indication that when performed, the temperature maintenance test will be satisf actory. Preparations are underway to perform this temperature maintenance test and it is expected that it will be completed within the

• next few weeks. f the tests performed Even though final integrated system testing has not been completed,lation System wn!!

to date provide enough assurance that this control Room Area Venti function as designed in all modes of operation. If you have any discuss this further, please contact Mr. Steven Chaplin at (919) question 536-6623. Yours very truly. em S. . Zi erman . I nager A l Nuclear 1.icensing Section U(l SDC/bmc (5078SDC) cet Mr. B. C. Buckley (NRC) Mr. G. F. Maxwell (NRC-SHNPP) Dr. J. Nelson Grace (NRC-Ril) h YO '4. 1

                                                                                                                                             )

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                                                                             .-                      '             .-                              ATTACHMENT 2
                                                                                                                                                                                                                     .         CQL-9522 i                                                  .
                        ' ' WeStinghoUSB .                                                                  Power, Systems e imye sysa nim Electric Corporation                                          -
                                                                                  .                                   .                                                                                         sans PinswpPensten152n c355 1
                                                                                                                                                                                                                                                                )

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                                                                                                                                                                                                      . November 25, 1986 I                       Mr. L. I. Loflin, Manager                                                                                                                                                                                                    !
           ;                       HarrisProjectEngineer1'ng 8

Carolina Power & Ligh,t C.ompany P.O.' Box 101 . . 4 '

Ne'w Hill .NC4.27562 j - I g

                                                                                                                  ' CAROLINA POWER & LIGHT COMPAN
                                                                                                            !SHEARONHARRISNUCLEARPOWERP                                                              NT

: Post Modification Field ExDerien e on

                                                                                                    - .:'                                Counterflow Preheater Tube                                   -

Dear Kr. Loflin:

[h i ,. .

i. . j Attached per your reques['is the information on the operating experience of '.

                                .Westinghouse counterflow tiodified steam generators. Ws trust CP&L will find

l. -

this information useful.8 w .. , j :. ! e If there are any quest.ichs.. please feel free to call us. g . ji Very truly youts, WESTINGHOUSE E(ECTRIC CORPORATION Y: ': .

                              . ,s ..~ .

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                                                                                                                                                -                  I    A                            ana er
                                                                                                                                                                / T . Carolina T. Parke.r            f)& L!ght, Project ,

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               .,,.. Attachment' , 1                                     ,
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                           . cci                     L. 1. Leflin, CPt.L' i                                                                                                             .
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                                                   'R. A. Watson, CP&L' i.te                                                                                                                             ll,1 A 8 :                                                                                                                   te
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                                                  .A.,C.               Anderson,                           Eba           c'o                                                   .                         3L, 3A
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L. Rowell,'CP&L Sitt ll,1 A 1 L', 1 A

                            -       **/ . RJ.. EL..' 'L'ums            W1.111s,         d e n , C P&'         Pla'n't    '                   G'entral'Hanager, CP&L Site                         -       IL,1 A ~ *

• 5 L. H. Martin, CP&Yi . 1L, I A-

                                                   .G. L'. Forehand, C'P& , Aite                                                                                                                         1L,1 A
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' . ' POST MODIFICATION FIELD EXPEAIENCE ON l l 1 . CO l j;UdTERFLOWPREHEATERTUBEVIBRATION l i

                                                            -                                                                                                                                                             1 t                     :                                                                                                  .

i l l I Post ahdific.tio,r tube inspection data have bien obtained from three

                     !                                                          plants with count)erflow preheater steam generttors (Models 04, 05
                     !                        .                            'and E) and vibrption data have been acquired gn one plant.

I

                      ;                                                  . In on't unit,!the ;two Model 04 steam generator [ have been' inspected                                                                        )

8 after 3 years of ' post-modification operating a xperience. There were no 1 reportable indigations of tube wear in the *' preheater (20% threshold  ! l l -

                                                                             ' sensitivity). , ' , , .

i Eddy current'. inspection of the three Model E tteam generators in each i of two units lafte'r one cycle of operation res ilt'ed in no reportable wear.' In addition, expanded tubes were inspe ted with a UT method t which has s'nent ,tivity of about one mil wall thickness and can detect

       ,        ,j                                                          'the small wall't inning due to tube expansion! No measurable wear was
                                                                           , detected with th .s inspection snethod.

i' -- Tube' vibration gifta were obtained ori one expadded tube and two unexpanded tubes steam generator, ,

                                                                            ;Theresultsjndic'duringoperationofaModelf        ated small vibration levels tnd ti delta values lessE tha'n .3 g-mils'! This 6 delta value is almost 3 orders of magnitude

• lower than thQaximum premo.dification values . Tube vibration data were"also reyuire'd for an expanded tube in a Uodel 04 steam generator.

                                                                                                                                                  ~

i .. ' The 6 delta yalpe' for this tube was .4 g-eilst With extensive plate

                                                                         . position searchin,g to maximize tube response,la maximum 6 delta value I             !                        .
      !i                                           - -

of approximaiely *,4 was measured in laboratory flow model testing of t , thepost-modificationcondition.s. This value R less than the design i~j . objective value a'nd is approximately a factor of 50 lower than the i;

               -s              .
                                       .. *./ ,:.... maximum prem6difi,. cation values..
                                                                                      ,c: . .             . - -                                                          *
                                                .'/*.-   '.
          .Ij'
                                                                           ' Modified Medg1 0 and 05 steam generators in two other units are r-                                                            . cur'r'         a ntly optrat g in the first fuel c;cle. These units are expected                                                   *
                         -                    '. , . .                         to,pomplete tip , cycle in'spections in 1987.                                                    ,
                                                            .'               '.In sunre.ary', fiel experience to date indicatto no reportable wear en
                                     ..                                        the 8 steam geht tors that have been inspectrd and low vibration
                                -                                            #1evels' on counte flow preheater steam generators. These data suppdrt I                             -
                                                                             .the adequacy'of' he counterflow preheater m'odLfications for the Model                      '

j ,

04. units at,$he,hr,on Harris-1.

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I ATTACHiENT 3 l, JUN 0 IS64 , f d PEPORA'O'JM FCR: Dr. Jesse C. Ebersole, Chaiman i Advisory Comittee or Reactor Safeguards

        .                                                                                                 7 FMf!:                   Thomas P.. l'ovak. Assistant Directnr for Licensing                                               "
                                          - Divisten of Licensirg

SUBJECT:

P.E5FCtl5E TO Mr:. EDDLEMAN'S LETTEP TO ACR5 DATED JA}!L'AP.Y 1, 1984 IN F.ERAF.D TO THE SHEAn.ON HARRIS APPLICATION in your January 16.19M ACRS report on the Shearon Harris Nuclear Pcwer Plart, you recuested the PRC staff to investigate the allegations described in Nr. Wells [. Eddlenan's letter and provide a written report to the Comittee. J. The purpose cf this letter is to provide.an initial respnnse to the Comittee

          -         en Mr. Ed81eman's allegations. Hr. Eddlerin is an intervenor in the engoing
       +"           adninistrative proceedir.es which is part of the review of the operating licens,e f;           application. We have reviewed Mr. Eddlei.an's letter and.it appears to raise p-           concerns 'in the folleming areas related to the Shearon Harris application.                           .

r*

      ~

1. On page 1 and 5 there are allegations concerning management of .the Brunswick and Robinson plants. Testimony is currently beirp prepared on this issi:e 9..

and is scheduled to be filed in mid-August. As you know, the staff nov provides the ACRS with copies of all staff testincey to'be filed in administrative proceedings. This will provide you with the staff position. Me will also provide the ACPS with the ASLB's Inittel Decisien on this "f issue. In regPrd to the pipe crack issue, the Carolina Power & Light (CP&L) Company did inspect the piping of the' Brunswick unit which was

        ~..                   already shutdown but requested a delay of several sienths on the other unit'
        -                . because they had completed in inspection of the piping..which met the r,            ,

inspection criteria. .iust several renths before. L4 2. Page 1 raises an issue as to the cualificatiens of Carolina Ptwer and Light Corpany management to operate the Shearon Harris facility. This riatter is fFy currently before the Atenic Safety and Licensipp Board (ASLB) as Joint

     *                       'Crntention 1 and an evidentiary hearing will be held on this issue. Fe will provide the ACRS with a copy of the staff's testie.oriy en this issue. This will provide you with the staff position. We will provide the ACRS with a ency of the ASLB's Inittel Decision on this metter.

e s .,

                                                                                                      ,.js -l lL                                                                             A-/07                        .

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        .                                                                                                                         ;e 1

3. Page ? questions the aperepriateness of the pipe hangers. Theipplicar?

has instituted a 100* reinspection progran cf the pipe hangers, the .

• resul,ts of this progran uill be reviewea by the NP.C staff. This is now befora the ASi.E as Icdler.an Ccntention 41 and will be resolved by that Licensing Roa-d based upon a public record. We will provide the ACF.S.

the staff's testinony and a copy of the ASLB Initial Decisten on this ratter. 4 Page 3 raises a concern abou+ the "fix" fer the stean rene-ator art water Samer censideratters which are currently before the Licensing Beart' as k. Joint Contention 7 and Eddienan Contention 45 cr.d en evidentiary hearirt t 5; will be held on these issues. .We will provit'e the ACRS with tne staff's te,stinony and a copy of the ASLR's initial Decision en these natters. 1

g. -

y 5. In regard te the aller i- the faults in the rocE;ation on an inective feult under the certaineent. beneath the site were,evalueted and fcund to he

;        f.                                at least.several Milion years old. Feuits' that have been inactive for i'

that lo' rip are not likely to be reactivated during the life of the plant and' are not considered to be hazardo'us features fcapable fault-Apeertix A. 10 CFR Part 100). All rock is broken by joints and/or faults. These i features are taken into account in the engineering design of nuclear

        !                                 plants. The Unit 3M holes are isolated from Units 1&2 by a retaining i
                 .,                      wall and would have en effect on the operation of the Shearon Harris                                            -

H- plant. ' , L. ~ N.

5. '

Page 4 - The allegation is in regard to the effects of eartheurkes.as a N source of comen rede failure of all the transmission liner to the Shearor. i . Harris site. Also, the alleg'ation states that comon mode *eilure o~f the .

           .h                            transnission lines could result from terrorist attecks, huge ice stoms,                                 ~

x - wind sterns or hurricanes end that the applicant only I;bes a probability g for tornado-caused comon sede failure. General Design Criteria (GCC'.17 lf r dres not require transn!ssion lines to be designed to. Class IE requirenerts. i Less of off-site pcwer is assured during accident cer.ditions and power i fron the Class 1E diesel generators is relief upon 'to operate safet,v-related ecuipment to r.itigate the consequerces of such accidents. The effects of environrental conditions on transmission lines such as ice stems, wind, hurricanes and tornados have been evaluated by the applicant and their conclusin~ns found to be accepte.ble to the NRC staff. l!!th regard te terrorist acts, as described abnve, the staff assures the loss r3 off-site power, regardless of the cause, t!uring accident condittees. , ,,-

                                                                          '                                                       >~.

7. ' Page 4 - This allegation is about earthquake Taulting at. the Steron Herris site and the effects of faults specifically en the containrent.zasling Water supply piping, and the spent fuel pool building. The effaqs of

(' earthquakes en the Shearen Harris fartlity have already been evaluated in both the construction phase and operating license review and our findings have been reported in the NRC staff Safety Evaluation Peports. 4--/D I .

                                                                                                                        ~

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       -                                                                                                                 i

8. Fage 4 ~- In recard tn the Security Plen, the ASLB is presently censidering pro"ered cor.tentions or. this retter. We will acvise the ACll$ cs to the outtone c( the ASLR's actions. As t:e stated in Section 13.6 of the t'RC ^

SER dr.ted liove-ber 1933, the staff considers the Security Plans for Shearen Harris tc, En adecuate. l i 9. Hr. Eddleman cuestions the control roon design. The staff has not cor.pleted 1 j its review. When we do, our views will be set forth in a supplenent te cur

    .                    S E F.. Also the ASLB has this rv.tter before it as Edeler2r Contentien 1320.

We will provide the ACRS a copy of the ASLS's Initial Decistor r> this 1 l satter. It is pessible that ser.e of the cer.cerns raiset in Mr. Eddleman's letter ray be 1 l- resolved in a surrury disposition procedure. If this is done, we will provide the ACES pith ll documents subnitted to" the ASLB and with the ASLB Order resolving the,aissues. , In regard,ts.the chilled water sys'.en, the staff will review this systen and p Teport its findings in a supplement tn the SER, as resucsted by the ACRS. p

                                                                                                     ~        -

h' Thomas P. Novak, Assistant Director for Licensing

      .I                                                                     Division of Licensing
     ..j                                ,              .,

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ATTACHMENT 4 O cpat Carolins Power & Light Company DEC 5 1986 SERIAL: NLS-86-452 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20535 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/ LICENSE NO. NPF-53 , TURBINE OVERSPEED PROTECTION

Dear'hjr. Denton:

Carolina Power & Light Company (CP&L) hereby submits additional information concerniniturbine overspeed protection at the Shearon Harris Nuclear Power Plant (SHNPP). This information is submitted in response to a verbal request for additional f information from the NRC reviewer. The fo!!owing additionalinformation on turbine (3)

     %./

overspeed testing is provided to satisfy this request:

                           -     The test of the Turbine Mechanical Overspeed Protection System is conducted in the no load condition during power ascension testing (9103-5-27).                                j
       -                   -     The electrical overspeed protection system was tested as part of preoperational test (1-5015-P-01).                                                                              l
        -                  =     Per Technical Specification 4.3.4.2.b, the electrical and mechanical overspeed                   :

• protection systems are physically tested every 18 months per the following , i surveillance tests: l r Mechanical Overspeed - Operations Surveillance Test 057-1075 Electrical overspeed - Maintenance Surveillance Test MST-1-319

                            -    The Westinghouse analysis of a SHNPP turbine overspeed event assumed a generic rotor design. The generic rotor assumption envelopes the new SHNPP Low Pressure Rotor design.

• The Westinghouse analysis of a SHNPP turbine overspeed event used a computer model which has been shown to be conservative based on actual overspeed events at j the Salem Nuclear Plant. The Westinghouse analysis for Salem predicted an l overspeed of 105.5% which was corroborated by three actual overspeed.pvents from full power which resulted in overspeeds averaging 104.8%.

l( 1

                                                                                                                  ~~

A -tm l l m reyene..n. sneet . P o sei isst

• Rese.p N C U602 ,

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Mr. Harold R. Denton g[g g E,O

• C NLS-86-452/Page 2 If you have any questions on this subject or require additional Information, please contact me.

                                                                                                .Yours ve                ruly, il
                                                                                                 . , .                4L
                                                                                                     .'R. Zimmerma%n 6               Manager Nuclear Licensing Section 3HE/kts (50865DC) cc:      Mr. B. C. Buckley (NRC)

Dr. 3. Nelson Grace (NRC-RII) ' Mr. R. 3. Giardina (NRC-PAPS) i

             .                     Mr. G. F. Maxwell (NRC-SMNFP) b b

e 9 h Y . I l

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ATTACHMENT 5

 -                                                            CP&L carehna Po.e, a uom company                     SERIAL: NL5-86-426                  j O 21 086  ..
                             .                                                                                                            l Mr. Harold R. Denton, Director                                                                                           !

Office of Nuclear Reactor Regulation  ! United States Nuclear Regulatory Commission l Washington, DC 20355 J SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/ LICENSE NO. NPF-53 l TURBINE OVERSPEED TESTING

                                                                                                                                          ]

Dear Mr. Denton:

I Carolina Power & Light Company (CP&L) hereby submits additional information concerning turbine overspeed testing at the Shearon Harris Nuclear Power Plant (SHNPP). This information is submitted in response to a request from the NRC Advisory { Committee on Reactor Safeguards (ACR57 requiring the submittal of a structured test l

    .           program for evaluating turbine overspeed protection prior to full-power operations.                                       !

Attached is a copy of Startup Test Procedure 1-9103-5-27, Turbine Overspeed Trip l Test". This startup test will be conducted during power ascension testing. In addition, the following information concerning the turbine overspeed protection at SHNPP is subrpitted for your review. Three overspeed protection functions are provided on the c

   -            SHNPP turbine:                                                                                        ~

(1) Auxiliary governor control / load drop anticipation - setpoint of 103 percent of rated speed, (2; Mechanical overspeed trip - setpoint of 110 percent of rated speed,

0) Electrical overspeed trip - setpoint of !!O percent of rated speed.

The t.uxiliary governor control function controls turbine speed without a turbine trip In the event of a fullload rejection. When the setpoint is reached, the governor and - interceptor valves close. Then, the interceptor valves cycle open and closed until the

      .         entrained steam / energy is released to the condenser. When speed decreases to rated speed, the interceptor valves remain open, and the governor valves begin controlling '-

speed. Overspeed is limited to the roll-up immediately following auxiliary governor control action. Reducing the setpoint below 103 percent of rated speed would create the risk of spurious auxiliary governor control action during normal turbine speed , oscillations. If the auxiliary Eevernor control logic fails, turbine speed willincrease to  ; approximately 110 percent of rated speed. At this spee'd, either the mechanical or  ; electrical overspeed protection system will operate. Both systems have the same l setpoint (!!0 percent of rated speed) and provide redundant protection against turbine overspeed. Either system may activate first depending on particular instrument / device tolerances. Both the mechanical and electrical overspeed trips cause all of the turbine steam valves (governor, throttle, reheat stop, and interceptor) to close. These turbine sicam valves are very fast-closing valves, and closure of these valves occurs near simultaneously to minimize roll-up. The closing of these valves stops all steam flow through both the high pressure and low pressure turbines. . l bb 0 kN l 4" Fa yette.. tie stree' *

• C E ra 9111
• fle e ;* P. C Pit::

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Office of Nucletr Reactor Regulation ) NLS-86-426 / Page 2 The mechanical and electrical trip setpoint of 110 percent of rated speed is chosen for V two reasons: (1) to prevent the turbine from reaching the design overspeed of 120 percent taking into account expected roll-up following turbine trip, and (2) to prevent a turbine trip af ter auxiliary governor control action due to the normal roll-up following the control action. Reducing the trip setpoint below 110 percent would create a situation in which a turbine trip would occur even though the auxiliary governor control had functioned as designed. This is an undesirable situation. The roll-up of the turbine following a trip at the startup test power level is expected to be I to 2 percent of full power speed. The reduced steam flow results in minimal entrained energy which can be converted to rotational energy af ter steam valve closure. The roll-up following a trip at

                                                                                                 ~

full power has been calculated by Westinghouse based on auxiliary governor control action at 103 percent of rated speec. This results in a maximum turbine speed of 108.4 percent of rated speed. Specific roll-up data has not been supplied for a trip at 110 percent of .ated speed. However, Westinghouse has stated that roll-up following a trip at 100 percent will not result in a turbine speed above the 120 percent design overspeed limit. Based on the overspeed protection system design described above and the satisfactory completion of the attached startup test, turbine overspeed protection at SHNPP will be demonstrated to be adequate. If you have any questions on this subject or require

                    -additional information, please contact me.

Yours very truly, F _

5. . Zi merman

                                                                                             ,ager Nuclear Licensing Section SR2/JDK/kts (5060]DK)

l. Attachment cc: Mr. B. C. Buckley (NRC)

, i" Mr. G. F. Maxwell (NRC-SHNPP) , i Dr J. Nelson Grace (NRC-Ril) l y Mr. R. J. Giardina (NRC-PAPS) l l l 4 4-//3 4

Cs2 PolLN Wo. su-8.1 (Ezy. 6) CP&L SHEARON MARRIs NUCLZAR POWER PW

                                                                                                              '                 )

START-UP MANUAI. INSTRUCTION /PROCEDUEE l l 1 SUM VOLUME VI No. 3-9103-5-27 e. 4- . . TITLE: TVRBINE OVERSPEED TRIP TEST i . ~ '

c

   .*                  EZV.      APPROVED BY       DATE               RIV.            APPROVED BY               DATE .

E l i l a APPROVED BY

                                             .                g          -                         -

7,3-h NAME/ TITLE ' DATE p=m ned" M/hABER-0)iRAT0hS 4-n Vol. VI 3 l t

                                                                                                                                )
                                                                                                                          /7    ;

ADVANCI CHANCI TOR.w (1) Proc. No f.9/ty+77 Rev. O Change # / O h} i2) Tit 1e _'llebbi Car pew s / fers [hl (3) Rea:on f of the Change $nrerribs'+ sI r#@w4) <r/<M 8/ Jd(Wlf/dv (/W.*b l (4) Description of the Change ru d/ ,f74,f0,///Erel//tf(1)- l (5) Additional pages Attached f of Pages 6 (6) Prepared by I M'/R /h Date It 29/4 l Name/Iitle (6a) Verify Tech. Spec. Cross Reference reviewed if procedure is used to satisfy Tech. Spec. Surveillance Requirements. ) f AIIft Data ff 19*ll

• l Preparar ,,

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                                                                                                                                                           ~
                                  ,' [ 7 ) Escosaanded for approval: .                .                                                'T
                               ~

7 o'<m iN .. Data k 29 Si  ;

         )                                   1st Technical Reviewer Sigssture title                                                                            i
 %J Jf /$ &~lfre t . Jure                     we . 4' <. e           Date is-30 8b 2t.d Technical Reviewer Signature Title (8) SATITY REVIEW                                                                                ,,

Two qualified safety reviews are required prior to Final Approval. Attach Wuclear Safety Review Checklist in Accordance with AP-011. .

   ~

(9) A1A1A concurren i applicable Signatur N d Date // *20 - 8 (10) Fire Protec A enee if appilc'able Signature VJW4 - Date//[to[N i (11) QA CONCURREWCE, if appilcable Signature n/a Date (12) FINAL APPROVAL or k Approved by J. I.*st Arm

• S5Y6a ultt/W Date afant/ Title .? ,

s ' \ Remarks s (Form AP-07-6-4) h-M [

052 Proc. No. 1-9103-5-27 Issuet Rev. o List of Effective Pages M Revision 1 - 12 0 4C 2),6,9,li ACcll ti rx RBt O O G

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_;t- _ _ _ _ _ _ __ 052 Proc. No. 1-9103-5-27 Issue: Rev. O

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T_able of Contents Section g 1.0 PURPOSE Ah*D OBJECTIVE 4 2.0 PREREQUISITES AND INITI AL C0h*DITIONS 4

3.0 TRECAUTIONS 5

4.0 REFERENCES

                                       ).O TEST EQ'JIPMENT                                                          $

6.0 TEST INSTRUCTIONS $ 7.0 ACCEPTANCE CRITERIA 9 8.0 TICURES 10 9.0 TABLES 10 10.0 DATA SHEETS 10 11.0 ELECTRICAL LINEUP SHEETS 12 12.0 VALVE LINEUP SHEETS 22

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052 Proc. No. 1-9103-5-27 Issue: Rev. O CN

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1.0 Purpcse and Objective The purpose of this procedure is to verify the on-line operation of the Turbine Mechanical Overspeed Protection Mechanism. 2.0 Prerequisites and initial Conditions n t i al s /Da t e I,n,i 2.1 Initial Criticality and Low Power Physics testing plateau is complete per 1-9100-5-01.

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2.2 Reactor power is less than 202.

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2.3 The Turbine-Cenerator has been at 2 101 Lead (95 - 145 HV) for eight (8) hours or mere. l 2.t All test personnel have been briefed on their responsibility. during the test by the Power Ascension Lead Test Engin(er. .

 \                                         2.5 Ar. operator is standing by the trip lever at the Covernor Pedestal, in contact with the Control Rocc.
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2.6 Comunications have been established. . I - 1 2.7 A frequency counter has been installed at the Covernor Pedestal to allow the operator to see the speed locally. (Spare speed e pickup, terminal Box "A", 781, points 40,41 and 42 CRND)

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1 2.8 The operational status of the unit' will allow the test plant l prerequisites to be established and performance of the test to l occur without violation of Technical Specifications or creation of conditions adverse to safe operation.

                                                                                                             -              l 2.9 The shift Forer.an has grarted permission to perfgra this test by signing in the space provided below.                       t                         ,

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                                                                      '                                        ?            /            1

($nift Foreman) Vol. V2 4 ear 1 1 1 l l f}

052 Proc. No. 1-9103-5-27 Issue: Rev. 0

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3.0 Precautions i 3.1 The Turbine speed should not be allowed to escoed 2010 RPM i (111.61). If the unit has not tripped when this speed is reached, the turbine is to be tripped manually.

                                      .3.2    should bearing vibration levels increase past the alars point of                          1 7 mils while above 1800 rpm, this test should proceed only with the approval of the Manager-Operations.

3.3 All work and testing will be performed in accordance with the CP&L Safety Manual. , 4.0 References 4.1 Drawints F/A

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4.2 Miscellaneous . 4.2.* 5tes. Turbines Vol I, Operation and control. Tab 36-9, Itev. 4,

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Wes:inghouse Tile # 16-5005 PO NY001. y 4.2.2 1-9100-5-01, Power Ascension Test Prograr.- Power Escalation 4.2.3 CP&'. Start-up Manual, Chapter 22 1 4.2.4 CP-006, Normal Plant Shutdown from Power Op'eration to Net I Standby 4.2.5 Regulatory cuide 2.68, Appendia A, 4K , 4.2.6 CM-M0164, Mechanical Overspeed Trip Calibration' a 3.0 Test Equipment s 3.1 Trequency Counter - Fluke 1900A or equivalent CP&L ID # Cal Due."Dete 6.0 Test Precedure Initials /Date l 6.1 Prepare fer overspeed Test 6.1.1 Re=ove the cenerator Load in accordance with applicable portions of CP-006 (normal plant shutdown to hot stan,dby). s

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t-6.1.2 Establish coar:unications between operator se the governor 2 pedestal and the operator at the MCB. , i I Vol. V1 5 k-//f l

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prac. 53. 1-9103-9-27 OS2 issuet Rev. 0

• Inittels/Date i

AC 6.1.3 Plitt the RPM indications on the MCS DEN panel and the frequency sounter used in section 5.0. Oll Record any difference in Section 10.1.2. fgg y b Q 29h , 6.2 Oil pressure Test of Trip Mechanisz 6.2.1 Mold speed at 18vs hPM. This portion of the test will be performed at the Governor pedestal of the Turbine. J

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6.2.2 Move the lever marked ' Test' and ' Normal' to the ' Test' position and hold it there for the duration of this oestion. I  ! u c 8/l 6.2.3 Slowly open the valve marked 'Overspeed Test'(iL0j!!): eli pressure gauge FT4ff0,

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i > I . fra M 4/k 1 6 . 2 . t. Record the pressure 3on Data Sheet 10.1 when the ' Reset' lever drops to the ' Trip' pasition for baseline data purposes.

/

                                                                                                                                                       \   }C This pressure has no set value for acceptance, but is                                                     p/f liOTE:

ysportisq[ to speed at the trip point, and so can be used to gg ," verify proper operations during regular operational , ' g gp., l , testing. (Baseline Data) - Still holding the ' Test' lover over, close the hand valve marked

i:

l 6.2.5

                                                  'Overspeed Test' tightly.

i p / i y Mold the ' Reset' lever in the 'Rese.t' position long enough so l 6.2.6 J that it will only return to 'Worsal' position, not ' Trip'.

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052 Proc. No. 1-9103-5-27 Issue: Rev. 0 n l w Initials /Date 6.2.7 Once the ' Reset' is in the 'Woraal' Position, the ' Test' lever may be released.

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6.2.8 Defeat the electrical overspeed protection by turning the

                                                     'Overspeed Trip' ketysvitch (on the trip test panel) to the
                                                     ' Inhibit' position
                                                                                                                             /

6.2.9 Defeat the DEN protection by turning the 'OPC' keysvitch (on the manual DEH panel) to the 'Overspeed Test Permissive' pcsition. 6.3 Press ' RET' on the turbine operator's panel, and enter a speed reference of 2015 RPM (112* speed). ' Hold' will light. Run il / Run.f2 / Run (3 */ Ir 6 . t. Press 'ACCEI. RPM l MIN' button and enter a rate of 50 (fifty) RPM / MIN. Run #1 / Run #2 / Run #3 / 6.5 Notify the operator at the turbine to watch the digital counter

   -                                           carefully, and to trip the unit at 2010 RPM.

• Run il / Run #2 / Run #3 /

e i 6.6 Assign test personnel to watch only the digital speed indicator on the operator's panel. ..

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Run il / Run #2 / Run f3 6.7 Press 'CO'. When the turbine trips ,Ibg the speed from both e indicators on Data Sheet 10.1 Run il / Run #2 _ / Run f3 / l 6.8 1.et the speed decrease to below 1700 RPM and relatch the turbine fro the valve test and latch panel. Run #1 / Run #2 / Run f 3 .'.' / 1 6.9 Press 'REF' and enter a speed of 1700 RPM, Press C0 . N' Run f1 ./ Run (2 / Run #3' /

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052 Petc. No. 1-9103 5-27

                                                                              !ss:st Rev. O D\                                 .

Initisis/Date

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m./ l 6.10 When tirbine speed reaches 1700, press ' Transfer TV' to return ' to Covernor Valve Control. Verify 'TV' extinguishes and 'CV' illuaicates. 1 Run #1 / Run #2 / Run #3 / l

                       ~6.11 Press 'AOCEl. RPM / MIN' and verify the rate is still 50 RPM / MIN.                                                           :

Run fl /_ tun #2 / _ tan #3 / y If the turbine tripped itself in Step $.7 within 1962 to 1998 C/l 6.12 1 EPM (1980 1 11), then repeat steps 6.3 through 6.11 twice f/. sore, . gg I

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6.12.2 If the turbine does not trip between 1962 to 1998 RPM, notify Ac Maintenance to perform CM-M0164, Mechanical Overspeed Trip

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Calibration. )[fd 2)fbid b it*Ptit0 fthtp fr & f f. y

                                                                                                   '/                                                 ki 6.15 On the Emergency Trip System Panel, take the Overspeed Trip i                                                           AC keysvitch to the INSERVICE positioc.                                                                                   sif
                                                                                                                   /                                  LI b v

AG 6.14 On the operator's panel on the MC5, take the CPC keysvitch to the INSEEVICE position. cli A

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052 Prse. 53. 1-9103-s-27 Issunt Rev. 0 7.0 Acceptsece criteria Initials /Date

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Level 1* Wene Level !! 7.1 The mechanical overspeed trip sechanism has successfully tripped three (3) times consecutively at speeds between 1962 and 1998

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                                                                                                                            #ti trx (11c1 2 11). see M 5 u n i                                                             ,         t/s.

101916 Level III Wone Acceptance criteria Approved

             . .-             Power Ascension Lead Test Engineer                                  Date e

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i 052 Proc. No. 1-9103-5-27

                                                                                                    !ssue: Rev. 0 l

S.O Tinuree w/A l 9.0 Tables I I i W/A ' 4 , 10.0 Data sheets i j - 10.1 Turbine Overspeed Trip Test i I e t i I i

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052 Proc. No. 1-9103-5-27 Issue: Rev. O t 10.1 Turbine overspeed Trip Test feitials/Date Cit Pressure.obtained during Step 6.2.4 / (Saseline Data) Mechanical Trips DEN Speed Local Speed (Step 6.73 Acceptance Criteria 7.1) Run il / Run #2 / Run f3 / If unit did not trip on first attempt (tthvMW #/A,Wlful s#MdiftIll4.h}' Cil pressure from Step 6.2.4 (second attempt) / M 18aseline Data) Old (lfl. Recceded by Time Date 10.1.2 Test Cons =ects r i O S O e l e. 4 iP i

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052 Proc. Wo. 1-9103-5-27 Issue: Rev. o l I 11.0 Electrical Lineup N/A 12.0 Valve Lineur N/A

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                                                                     *,ATTACHf1ENT 6
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                          'ks                                 UNITED STATES NUCLE AR REGULATORY COMMISSION                                      .
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dc.cketNo.: 5047 N YE*QRAf DUV FOR: Rart Rucklev. Senior Pro.iect Fanaoer

       .                                          PWP Pro.tect Directorate No. 2 o                                           Division of PWP Licensing-A                             .

N FR0v: Charles E. Rossi. Assistant Director

     .{ ,                                         Divisinn of PWR Licensing - A

(( SURJECT:

                             '                    SWEADDV HARRIS UNIT 1: TUPRINE OVEPSPEED PROTECTinN SYSTE**

j. _ TEST PROCRA!' (SRP 10.21 (TAC Fo. 630951 v

Plant Nar:e:'Shearen Farris Nuclear 8ower Plant. Unit 1

    ,                Utility: Carolina Power anff Licht Company

( Docket No.: 50-400 7 Pro.iect Manager: Bart Buckley ' Review Rranch: Plant Systems

• ranch Review Status: Complete g

h in the January 16, 1994 memorandum to Chatman Pa111dino from J. Ebersole (ACES), 3

                   entitled "ACRS Report on the Shearon Harris Fuclear power Plant, the ACRS recon-mended        "...a structured test program for evaluating overspeed protection of the-v-               turbine be prepared and subnitted to the NPC staff for review and approval before a-full power operation." The ACRS requested that the test procram show

t. 1. The riechanical and electrical overspeed trip setroints are calibrated to j' the proper value (110% of rated speed).

2

2. Given the proper overspeed setpoint. the roll-up to lowing a turbine trip would not exceed the Design Overspeed Limit (120% of rated speed).

The enclosed evaluatien from the Plant Systems Franch (PSB) addresses the Shearen Harris test program for calibrating and verifying the mechanical and electrical turbine overspeed trip setpoints, and the calculatfor.a1 model used to deterrine the acceptability of these setpoints. The' staff finds that the pre-operational, start-up and maintenance test procedures are acceptable, and meet.the criteria set forth by the ACDS for item 1 above. The staff also finds that the esicula-tional model used to detemine the acceptab111tv of the overspeed toip setpoints is conservative, and that the setpoints determined by the model wi13*not result p) i in a turbine roll-up that exceeds the design overseeed limit (120% 6f rated spe ed). Therefore, the staff finds that the intent of item 2. above'is met by V the model, and is an acceptable alternative to actual testino of thit system. 1 l 1C A 1>7

9. Buckley 9- .

i i This completes P58's e' fort on TAC 63Q85. A SALD input is enclose'd as Enclosure

2. -

s To 1, h0irce.tne i Divisier o' Pilo Licensinn - A

Enclosures:

i As stated g cc: T. Novak s L. Rubenstein i W. Minners l' Contact R. 1. riiardina f'8 .- y27900 i 1. - Y 3 ,* . . f . . , <>b

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                                 .          SPEAC.0L PADDIS NllCLEAD PO'.'ER PLA' T                                                                              :

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                 ..                                            PLAUT SYSTEFS BRANCu (TACNo.6?eF5)
     .           In the January 16, leM ACRt menorandur, entitled "ACD.S Depa.rt on the Shearon i

L Harris l'uclear Power Plant", the ACRS recomended "...a structured test emera"

     ;!         -for evaluating overspeed protection of the turbine be prepared and submitted to I           the dC sta#' for review and approval he'ere full power operatier." The Acc5
     ,;;.        requeste(1ha+ the test procrar show:

1. The mechanical and electrical overspeed trip setpoints were calibrated tn O)

   \t.                 theprnrervalue(110%ofratedspeed).                                       ;           -

2. Given the proper overspeed setpoint, the roll-up followine a turbine trio

[y would not exceed the Design Overspeed Linittf120% of rated speedt. E- - Jy letters dated November 21 and Decenber 5,19R6, Carolina Power and Light h,, submitted copies of start-up, pre-op, and maintenance test procedures that N would be used to demonstrate proper calibration o' the mechanical and electrical overspeed trips and to raintain the calibration of the trips over the li'e o' the plants. The staff has reviewed the procedures and f'inds that they setisfy iten 1 above. In attition the November ?! and Decenber 5, loa 6 letters discussed why a structured test program was not needed to meet the second obiective prescribed by the ACRS. The licen'ee s stated that the overspeed trip setpoints }110t of rated speed) were detemined based on a Vestinghouse calculationaltodel which predicted that the turbine roll-up would not exceed the design overspeed 11 nit (120% of rated speedi. The calculational nndel used to determine e overspeed trip setpoints and turbine roll-up following a trip has been verified as beino

 ,              conservative based 1on actual turbire overspeed trip everts at the Salen nuclear power plant.                               -

g- L f *

                                                                                      /

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                        )
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                                                                                                              'l For the Salem events the model preficted turbine roll-ur to 105.5% of ret *d speed followire a trip o' the auxiliary covernor control, with the$turh're a* W' load, The average roll-up for the three Salem events, when the turbine wts e' ful' lear', was 104.f" c' reted speed. The sta" has reviewed the in'orr.etion submitted and concurs with the licensee that the calculattenel model is conse'-

vative. The staff finds that overspeed trip setooints deteminet by this method will not result in turbine rell-up follow'ng a turbire trip that e>ceeds the

          !!                     design overspeed limit. Therefore, the staff finds that the second nh.iec+1ve cf

'f -the test progran recorrended by the ACRS can be met with the calculatter.71 redel, s~ ard finds this approach to be an acceptable alternative to actual testirp of the p.

i. systen.

,4 , f Baser' cr'th'e foregoing discussion, the staff concludes that the licensee's re-

    ,!                           sponse in the ,1anuary 16,1984 ACP.5 report recomendine that a structured test

'I progran for eveluating turbine overspeed protection be in place before 'ull power

t. . . ,

         ,                       operation, is acceptable. Furthemore, the staff concludes that the licensee tes                       ,

i s established acceptable procedures 'or calibrating and verifying the nechanical l p and electrical turbine overspeed trip setpoints..and that the calculatter.a1

              ~

approach for assuring that turbine roll-up following a turbine trip will mot exceed the design overspeed limit is an accertable alternative to actual testir'o of the systen. . } g* - e.' i i

                                                                                                        .' l *! .

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[_-o o - on l NUCLEAR PERFORMANCE PLAN GOALS UNCHANGED 1 i h e CENTRAllZED ORGANIZATION IN-PLACE o STRENGTHENED MANAGEMENT TEAM o RESTORING EMPLOYEE TRUST AND CONFIDENCE . i e IMPROVING CONTROL OF NUCLEAR ACTIVITIES i i - MANAGEMENT AWARENESS g S k - MANAGEMENT SYSTEMS AND CONTROLS l 5 j- -IMPROVED CORRECTIVE ACTION SYSTEM $ a i j i

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O - 1 RESTART APPROACH 0 i CORPORATE NUCLEAR PERFORMANCE PLAN j e Assemble Capable Senior Management Team e Consolidation Of Nuclear Activities Within A Single Organization e Centralized Direction And Control From Headquarters Division e C: ear Responsibility Established For Functional Areas i e Employee Concern Program j

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                  - Permanent Program e increasing Management Awareness
                  - Direct Access To Board
                 - Office Of Inspector General e Management Systems And Control
                 -Centralized Programs
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1 I REPLACE LOANED MANAGERS WITH ADVISORS  : e NEXT LOGICAL STEP IN PROGRESSION OF TVA MANAGEMENT o TVA MANAGER APPOINTED WITH ADVISORS TO CONTINUE TRANSFER OF INDUSTRY EXPERIENCE o ADVISORS PROVIDE CONTINUITY WITH ESTABLISHED PROGRAMS o ADVISORS SERVE AS PERSONAL MANAGEMENT CONSULTANTS

                        - CONTINUOUS INTERACTION BETWEEN ADVISOR AND MANAGER
                        -INCREASED ATTENTION ON MANAGEMENT ISSUES:

ORGANIZATIONAL EFFECTIVENESS STAFFING , PLANNING DIRECTION - CONTROL COMMUNICATION I

                        - RAPID DEVELOPMENT PLAN FOR TVA MANAGERS e ADVISORS OBSERVE MANAGER AND PROVIDE FEEDBACK                      ,

e ADVISORS PROVIDE TECHNICAL INPUT e ADVISORS 1NVESTIGATE PARTICULAR PROBLEM AREAS e PLAN TO ADD ADVISORS WITH UTILITY EXPERIENCE AT UPPER LE AND MORE ADVISORS TO LOWER LEVELS ' k-/3f _ 4

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KEY LOANED MANAGERS REPLACED DY TVA MANAGERS Position TVA Manager Advisor N , 4 Director of Quality Assurance R. C. Parker J. E.Huston Director of Nuclear Engineering R. W. Cantrell J. A. Kirkebo - I Director of Construction W. R. Brown B. R. McCullough Manager Planning and Finance J. L. McAnally J. C. Krummel Browns Ferry Project Engineer J. G. Chapman J. P. Stapleton - Manager of Modifications B. R. Painter C. Turnbow Site Quality Assurance Manager J. E. Law G. G. Turner i

KEY LOANED MANAGERS RETAINED 4 4 Position 4 Manager A Director of Licensing R. L. Gridley Site Director Browns Ferry i H. P. Pomrehn j Site Director Watts Bar G. Toto

• 4 j Director Engineering and Technical Services L.J.Ses

! Manager NMRG R. K. Selberling i

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                        - LEADERSHIP DEVELOPMENT                                     .

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e COMPREllENSIVE VACANCY LISTING -

I e EXTENSIVE ADVERTISING e BEARCll FIRMS , e TARGETED RECRUITMENT 6~ A G '5 k k b PkJ 4 - SELECTED MANAGEMENT POSITIONS

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SUMMARY

OF NRC'S WORK ON ALTERNATIVE DISPOSAL METHODS i TO SHALLOW LAND BURIAL l

 .                        - REVIEW ALTERNATIVES                       I
                          - PURPOSE OF FRN y                      - RESPONSE TO FRN f

g - STATUS & PLANS UER $

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LOW-LEVEL RADI0 ACTIVE WASTE POLICY AMENDMENTS ACT (LLRWPAA) 0F 1985 BY JANUARY 15, 1987 NRC MUST: IDENTIFY ALTERNATIVE' METHODS FOR DISPOSAL (SEC. 8) ESTABLISH PROCEDURES AND DEVELOP TECHNICAL CAPABILITY TO PROCESS LICENSE APPLICATIONS (SEC. 9(A)) BY JANUARY 15, 1988 NRC MUST: t 2h6 IDENTIFY TECHNICAL INFORMATION AN APPLICANT MUST PROVIDE FOR A LICENSE IN ORDER TO PURSUE SUCH METHODS (SEC. 9(B)) D . l i

DRAFT Tl:ClifilCAL POSITION ON ALTERNATIVE DISPOSAL CONCEPTS . (FRN MARCH 6, 1986) A 3-YEAR STUDY CONDUCTED BY NRC STAFF WITH CONTRACTOR ASSISTANCE IDENTIFIED ALTERNATIVE DISPOSAL TkCHNIQUES CURRENTLY.BEING - CONSIDERED AS DISPOSAL OPTIONS; LLRWPAA SEC. 8 REQUIREMENT 10 CFR PART 61 IS APPLICABLE TO ALTERNATIVE DISPOSAL CONCEPTS A 4 . t i N -

D D d~ .. i t i 10 CFR PART 61 SUBPARTSAl&B: PROCEDURAL REQUIREMENTS (APPLIES TO ANY METHOD OF LAND DISPOSAL)

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SUBPART C: PERFORMANCE OBJECTIVES (APPLIES TO ANY METHOD OF LAND DISPOSAL) , i

• TECHNICAL REQUIREMENTS (APPLIES TO NEAR-SUBPART D:

SURFACE DISPOSAL METHODS)

• SUBPARTS E, F, & G: ADMINISTRATIVE REQUIREMENTS (APPLIES TO ANY A METHOD OF LAND DISPOSAL) ,

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SPECIFIC / ILLY EVALUATED TECHNICAL REQUIREMENTS, 10 CFR 61 SUBPART D I SITE SUITABILITY (61.50) SITE DESIGN (61.51) g  ; l d SITE OPERATIONS AND CLOSURE (61.52) i W ENVIRONMENTAL MONITORING (61.53) t i e i ) l

O , o o~ SECTIONS 10 CFR 61.50, 61.51, 61.52, 61.53 ARE i DIRECTLY APPLICABLE BELOWGROUND VAULTS ~ AB0VEGROUND VAULTS EARTH MOUNDED CONCRETE BUNKERS AUGERED HOLES / SHAFT DISPOSAL MINED CAVITY IS SUFFICIENTLY DIFFERENT FROM OTHER NEAR-SURFACE DISPOSAL OPTIONS: HOWEVER, IT IS LICENSABLE UNDER PART 61.23 ON 4 g A CASE BY CASE BASIS s N 1 Y ' i . l l 6N 1 i

t MOUNDED Soll CovtR . J j w. VAULT COVER AND l m

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                                                                         /,i,i T '# '                                   FLOOR NATURAL F                                          ORENotNitRED
                                                               \ ,; ' ..

1 i I CONCEPTUAL CUTAWAY VIEW BELOW GROUND VAULT DISPOSAL j ( - 7

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Low-perissethy heemerene Siepe theer Gravel to Shgle Orah I i i i l l Generic aboveground LLW disposal vautt and foundation cross seetlen showing yavel era 6aste layer, *(.* 9*rmabluty maer488 eloped eaeavallen, ane porleheral eubourtase erein H f6 . . 7

( 40eattv5L W.8788 N .T N G Vgett.760 Top to 3s.,0.C50 CO.C#f,8 OLOCa t

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i M c6a( W e.. .es evers. I EARTH MOUNDED CONCRETE BUNKER 4 ! I l i l

                               ~

A-191 9

                                                         "S-w               ,

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                       .                                                                    I wo'uEoYo*foYo*E2 Ear g' "f
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                                      #'     A Y/ h%Y//-

IN SITU SOIL 2 3 CK 1 _=_ g't wwnmwan CONCEPTUAL ELEVATION VIEW . AUGERED HOLE DISPOSAL \C j' - A-/98 _

                                                                                         /0

O . O O~ . i i . . CONCLUSIONS OF TASK 2, VOL. 2, BELOWGROUND VAULTS L ,-

LICENSABLE UNDER PART 61 D0fS NOT RELY AS HEAVILY AS AB0VEGROUND ON DESIGN FEATUPES AND CONSTRUCTION MATERIALS FOR LONG-TERM PERFORMANCE NO LONG-TERM EXPERIENCE (CURRENTLY USED FOR STORAGE ONLY)

LIKELY TO PROVIDE IMPROVED SHORT-TERM PERFORMANCE j (100 YEARS) - 4 EXPECTED LONG-TERM PERFORMANCE SHOULD AT LEAST EQUAL SLB POSSIBLE HIGH EXPOSURE OF WORKERS - i w . l i s i i

                                                                                                                                                ~

l 1 CONCLUSIONS OF TASK 2, VOL. 3, AB0VEGROUND VAULTS e

              . LICENS6BLE UNDER PART 61                                                                                         .

d NO LONG-TERM EXPERIENCE WITH DEMONSTRATING PERFORMANCE I SERIOUS DIFFICULTY MEETING: ,

- 61.51(A)(1) AVOIDANCE OF LONG-TERM MAINTENANCE . .

j - 61.52(A)(2) 500 YEAR INTRUDER PROTECTION 4 CANNOT RELY ON SECONDARY CONTAINMENT BY GE0 LOGIC MEDIA i LIMITED SHORT-TERM EXPERIENCE ( l 4* DIFFICULTY IN DEMONSTRATING LONG-TERM DURABILITY OF i 4 ENGINEERING MATERIALS REQUIRES WELL DEFINED QUALITY ASSURANCE PROGRAM j l ' POSSIBLE HIGH EXPOSURE OF WORKERS j . i l . lk __ __ __-- -__ __ __ __r - - _ m 1 ___ ____ ____ -- - -

i  ; CONCLUSIONS OF TASK 2, VOL. 4, EARTH MOUNDED CONCRETE BUNKER ) ! LICENSABLE UNDER PART 61 ONLY ALTERNATIVE CURRENTLY BEING USED FOR ACTUAL WASTE DISPOSAL . (FRANCE SINCE 1969) ' l.* l POSSIBLE DIFFICULTY DEMONSTRATING LONG-TERM PERFORMANCE OF CONSTRUCTION MATERIALS . l

• CORPS RECOMENDS SEVERAL IMPROVEMENTS OVER FRENCH SYSTEM TO MITI

INTERNAL PROBLEMS i

4 .,

4 .

                                                                                                                                                                        ~

i

i l H , CONCLUSIONS OF TASK 2, VOL. 5, SHAFT DISPOSAL /AUGERED H0LES LILLf6ABLE UNDER PART 61 . iSHORT-TERM EXPERIENCE (APPROXIMATELY 25 YEARS) , I INITIAL USE HAS BEEN STORAGE

MOST LIKE SHALLOW-LAND BURIAL

! b EASILY CONSTRUCTED AND ADAPTABLE TO LINERS IF DEEMED 4 APPROPRIATE

h '

f EXPECTE LONG-TERM PERFORMANCE WOULD BE EQUAL TO OR GREATER THAN SLB -. . j l l ' i . t

O TD ^O . l THE DRAFT TECHNICAL POSITION SOLICITED ANSWERS TO FOUR QUESTIONS (

SUMMARY

STATEMENT OF EACH QUESTION AND RESPONSES) QUESTION 1: SHOULD ANY ADDITIONAL ALTERNATIVES BE CONSIDERED RESPONSE: FRN IDENTIFIED ALL ALTERNATIVES QUESTION 2: WHAT ADDITIONAL REGULATORY GUIDANCE IS NEEDED RESPONSE: AEQUEST VARIED ! QUESTION 3: 00 YOU SUPPORT THE APPLICATION OF STANDARDIZED DESIGN  ;

STANDARDIZATION WAS SUPPORTED RESPONSE

BEST APPLIED ON A REGIONAL BASIS QUESTION 4: D0Y0bFAVORPRELICENSINGREVIEWSTOEXPEDITETHE '

LICENSING PROCESS - RESPONSE: FAVORABLY SUPPORTS PRELICENSING REVIEWS Obw , i W > Di .

s l i ADDITIONAL WORK ON ALTERNATIVES \ DRAFT STANDARD FORMAT AND CONTENT OF LICENSE APPLICATIONS FOR NEAR-SURFACE DISPOSAL OF RADI0 ACTIVE WASTE (FRN MARCH 14,1986)' l \ STANDARD REVIEW PLAN (INTERNAL DRAFT AUGUST / SEPTEMBER) l REGIONAL WORKSHOPS ~ OUTREACH PROGRAM FOR TECHNICAL ASSISTANCE i CONCENTRATED EFFORT ON SELECTED ALTERNATIVES

          -    BELOWGROUND VAULTS
          -     EMCB i

j . k . Q . I . N i%

D. O o. . i 't ) e i NRC FOCUS ON BELOWGROUND VAULTS AND EMC8 PREFERRED OPTIONS CURRENTLY BEING CONSIDERED - s i i LIMITED RESOURCES ,

' CRITICAL TIME SCHEDULE ,

                                                                                               ~

.t 1

                                                                         ~

1 . l j - i ! 4. l 1

       \

i ! N , i

I 1 , i i . ! LOW-LEVEL RADI0 ACTIVE WASTE POLICY AMENDMENTS ACT (LLRWPAA) 0F 1985 -L  ; BY JANUARY 15, 1987 NRC MUST: i l IDENTIFY ALTERNATIVE METHODS FOR DISPOSAL (SEC. 8) FRN IDENTIFIED ALTERNATIVE DISPOSAL CONCEPTS I

ESTABLISH PROCEDURES AND DEVELOP TECHNICAL CAPABILITY TO

! PROCESS LICENSE APPLICATIONS (SEC. 9(A)) -

                  -            STANDARD FORMAT AND CONTENT GUIDE AND STANDARD REVIEW       -

PLANS MECHANISM PROCESS APPLICATION I BY JANUARY 15, 1988 NRC MUST: IDENTIFY TECHNICAL INFORMATION AN APPLICANT MUST PROVIDE ' i FOR A LICENSE IN ORDER TO PURSUE SUCH METHODS (SEC. 9(B)) j i ! - REGIONAL WORKSHOPS PROVIDE INPUT 4 -

m .

!4

[

O O O~ -

                                                             ~

SAFETY ASSESSMENT OF ALTERNATIVES TO ' SHALLOW LAN D BURIAL OF LOW LEVEL WASTE i OBJECTIVE: 1.) IDENTIFICATION OF DESibN FEATURES IMPORTANT TO SAFETY DF THE ENGIN'EERED ENHANCEMENTS / ALTERNATIVES TO SHALLOW LAND BURIAL j 2.) F4ELIABILITY ANALYSIS OF THE "lMPORTANT" DESIGN FEATURES ( identify failure mechanisms and relevant environmental j c.on dition s, estimate performance lifetimes) l s D e N l -

,                               \

i

4 . l i j , CONTRACTOR: INEL FY 1986 WbRKSCOPE l TASK 1: ALTERNATIVE PtTHODS REVIEW

l!TEPATUPE REVIEW OF THE CONCEPTUAL DESIGNS OF THE FOLLOWING FIVE ENHANCEMENTS / ALTERNATIVES TO SHALLOW LAND BURIAL

MINED CAVITY, ABOVE GRADE ENGINEERED VAULT, BELOW GRADE ENGINEERED VAULT, AUGURED HOLES, AND CONCRETE BUNKERS TASK 2: IDENTIFICATION OF IPPORTAPIT DESIGN COPPONENTS

IDENTIFICATION OF SIGNIFICANT DESIGN FEATURES THAT CAUSE AN ENHANCE E NT/ ALTERNATIVE TO DIFFER FROM SHALLOW LAND BURIAL AND THE RELATIVE ENHANCENNTS AND/OR DETRIMENTS ASSOCIATED WITH THE g IDENTIFIED FEATURES RANKING OF IPPORTANT DESIGN COMPONENTS kTASK3:

R ' \ ELATIVE RANKING OF THE DESIGN FEATURES ACCORDING TO THEIR CONTRIBUTION TO THE OVERALL DESIGN PERFORPMNCE FY 1987 E mKscuit . h !FICATION OF THE BENEFITS / RISKS ASSOCIATED WITH THE DESIGN FEATURES DETERMINED TO BE IPPORTANT TO OVERALL FACILITY PERFORMANCE .g - l 1

s l ! BASIS OF TECHNICAL APPROACH , 1 o IDENTIFICATION OF DESIGN COMPONENTS D5RIVEDFROM'SumARYCHARACTERIZATIONOFLOW-LEVEL i RADIDACTIVE WASTE DISPOSAL TECHNOLOGIES," ROGERS

ASSOC. ENGINEERING, TIM 850lv1-1, JUNE 1985.

o DEFINITION OF PERFORMANCE OBJECTIVES 2 FOLLOWS 10CFR61 " LICENSING ~~ REQUIREMENTS FOR LAND lilSPOSAL OF RADI0 ACTIVE WASTE" SUBPART C - PERFORMANCE OBJECTIVES.

;                                                                      o                ANALYSIS OF ALTERNATIVES BASED ON FAILURE MODE AND EFFECT ANALYSIS (FMEA).

FIGURESOFMERITARECALCULATEDFOR: J

RELATIVE PERFORMANCE OF EACH ALTERNATIVE ,

RELATIVEIMPORTANCEOFEACHCOMPONENTWiTHIN EACH ALTERNATIVE.

9 4

l 1

                                                                                                                                                                                                                                                                              ;fl 1

D _

                                                                                                                                                                                 /-/s7
         , , _ - - - - - - - . , - . _ _ _ . . . , . _ - - - - - - - - - , - - - - - - , - . . - - - - , - - - - - - - - - - - - - - - - - - - . - - - - - - - - - - - -                    - , - - - - - - - - - - ~ - - - - , - - - - - - - ~ - ~ - - - - ~ - -

                                                                                            ---           -    _.       _.   ~ _.

i i FUNCTIONS OF FOUR DESIGN COMPONENTS

                 +                                      ,

COMPONENT ' 7 FUNCTIONS COVER Blo-BARRIER DIVERTS INCIDENT WATER )

             ,                                                          PREVENTS AIRBORNE RELEASE SHIELDS DIRECT RADIATION STRUCTURE ASSURES STABILITY PREVENTS RELEASE TO AIR, WATER SHIELDS DIRECT RADIATION O                  FILL                                           -

ASSURES STABILITY SHIELDS DIRECT RADIATION CONTAINER ASSURES STABILTIY l PREVENTS RELEASE TO AIR, WATER

                                                                                                                                            )

SHIELDS DURING OPERATIONS

                                             # e*           I r    WS

• 1 e

s%  !

           .L
                              ~

A -/&& ZZ ,

  ,  .,....___-.---...-_,......--..--,s..--                     -,       .____. ..-- ._,. .                       _                 _ - - .

l s_- i SETS OF DISTINGUISHING COMPONENTS FOR SELECTED ALTERNATIVE DISPOSAL SYSTEMS Relation - Alternative ' to Grade Baseline Case Enhancement Shallow Land Burial below Cover (SLS) C'ove r Fill Fill Container Selow Ground Vault below Cover (SGV) Cover Structure Structure Fill j Augered Hole ' ' below Cover (AH) None

                                                                   ^ Structure Fill Earth Mounded Concreta         above             Cover                    Cover Bunker Tumulus (EMC3)                            Fill Fill Container Above Ground Vault             above Structure         il Structure (AGV)

Cover 12 Structure Cover

                                                                                          , Container

_ A-H/ p

1 i CONCLUSIONS . l 1.) The cover design component is one of the most important design components regardless - of the presence of other components 2.) Engineered enhancements that improve the b N performance of the cover and/or provide redundancy with the cover can significantly enhance a disposal system f 3.) The above ground vault is unique among the alternatives due to the reliance for system performance resting on one engineered component A-/42 .

r E _ A3 _ 'Y As A _YS S 1.) Identify relevant time scales 2.) Identify benign and detrimental environmental conditions 3.) Develop a data analysis method to accurately estimate performance lifetimes attributable to the "important" environmental conditions 4.) Present results in a form which will support the development of design criteria and p'erformance assessments i O _ A-/63 25

1

     }              .

FAILURE MECHANISMS LEADING TO LOSS OF CONTAINMENT BY STRUC IV: LOSS OF CONTAINMENT OF STRUCTURE Failure Mechanism Contributing Cause Factors

1. Concrete Cracking a. Loss of 5tability -

of Cover (!!) i t

b. External L'oads Dead Load .

Live Load Seismic Mud . i Ponding Espansive Soils

c. Shrinkage Drying Mixture i Stre of Member Restraint '

d. Tempera ture Internal Environment External Environment

                                                  ,                                    Fire

e. Settlement in Plastic Suu

f. Degradation of Concrete Failure of Coatings Freeze / Thaw Chemical Esposure Reactive Aggregates Correston of Reinforcement Electrolysis

g. Foundation Settlement .

(!!-1.e)

2. Seals a. Installation

                "                                                                     Doors & Portals                                    -

Vents Construction Joints Expansion Joints

b. Degradation of Seal Chemistry inside Cheefstry Outside Failure of Coatings

c. Overpressurization Failure of Vent

3. Barrier Penetration a. Corrosion Failure of Coatings

.. Chemtstry Outside ' Chemistry inside Dissimilar Metals ' Motsture

b. Degradation of Failure of Coatings Materials Chemistry Inside Cheatstry Outside Motsture

4. Leaching a. Concrete Cracking (IV-1) '

                                  ~

b. Seals (IV-2)

!( -

c. Inft1tratton (1-2) 4-n, 7 M

                                                                                     .               1 l

I 100 i i _3----_ ' Data available a

                     ---Insufficient data available         Environmental          W atta:k             Benign b

g g _ environment _ g _ 0 1 10 15 100 j Time (yr) ewa

 )

e* e 1 ('. ~ 4-/if Graphical Representation of Expected Component Perfomance Z[

Current Status of Each State in Providing Disposal of Low-Level ) Radioactive Waste . Prepared by Office of State Programs - Current as of December 1,1986 For further infonnation, contact Dr. Stephen N. Salomon, (301) 492-9881

1. Chart entitled, " Interstate Low-Level Radioactive Waste Compacts Consented to by Congress as of December 1,1986." .

2. Compact status map as of December 1, 1986.

eP

3. DOE's Draft Generic Plan for Development of a New LLW Disposal Facility.

4. Discussion of the following:

A. Meeting the Congressional Milestones o Compacts and States Covered by Congressional Milestones , o Milestones and Penalties o Meeting the Congressional Milestone of July 1,1986 Compacts consented to by Congress

               , . ;.            o      Central-Midwest Compact o      Midwest Compact o      Central Compact o      Northeast Compact Compacts requiring Congressional consent o       Appalachian Compact o       Western Compact Unaffiliated States B. Progress in the $1ted Compacts Note:          New infonnation since the last status report dated October 6,1986, is highlighted by a bar in the left column. Also, where possible 1

information on compacts and States is organized following the draft

 ; %d       J DOE Generic Plan for Development of a New LLW Disposal Facility.
                              ~

j d '/(( 27

( 1 INTERSTATE LOW-LEVEL RADIOACTIVE WASTE COMPACTS CONSENTED TO BY CONGRESS AS OF DECEMBER 1, 1986 i

Northwest Rocky Mountain Central Midwest Central-Midwest Southeast Northeast Alaska Colorado Arkansas Indiana Illinois Alabama Connecticut

!               Hawaii              Nevada                        Kansas           Iowa          Kentucky          Florida            New Jersey Idaho             New Mexico                    Louisiana        Michigan                        Georgia            Washington, DC (2)           .

Montana North Dakota (1) Nebraska Minnesota Mississippi

;               Oregon              Wyoming                       Oklahoma        Missouri                         North Carolina                                  i Utah                                                               Ohio                            South Carolina Washington                                                        Wisconsin                        Tennessee Virginia i

t

        .                                       COMPACTS REQUIRING                                                  Unaffiliated States 1

CONGRESSIONAL CONSENT Maine Massachusetts (4) N Appalachian Western (2 options) New Hampshire j h (5.2679;H.R.5338) New York (4) t g Delaware Rhode Island Texas (4) i j Maryland Arizona Arizona Puerto Rico - Pennsylvania California (3) North Dakota (3) Verinont

West Virginia South Dakota i,

! Notes: ! (1) MD membership effective July 1, 1987. Rocky Mountain Board must approve.

(2) Washington, DC has enacted temporary legislation and must petition to join the compact. Northeast '

! Connission must approve. ! (3) CA and ND aust enact legislation to become members of the comoact. l (4) MA, NY and TX are each planning to dispose of its own LLW. . 1 .. j .

                                                                                                          ..           g j       *                                                                                             .

l . . 4

i LGW-LEVEL RADIOACTIVE WASTE COMPACT STATUS

                                                                             .                              DECEMBER 1988                                               .

NORTHWEST MIDWEST UNAFFILIATED STATES FIF##21

                          ,

• WA is host State
• No host State selected
• 21% National LLW(10 Stateel AK = 5% National LLW
• 7% National LLW
• NY to host site - 6% National LLW- SLB banned j *
• SL8
• SLB banned
• MA to host site - 4% National LLW- SLB banned
• VT. NH. ME RI DC. PR each less than 1% National LLW- SLB banned in ME eo

                                                -                                                                                                                     VT NH 1

l I WA MN ' HI%e ej MT 1

• Q ---) l ND K We p f1 NV l l

g - OR I l g g l i

                                                                                                                                                ) MI l
               \

i MA NORTHEAST i  % h 1

                                                                                                                     %                           IN OH                  0                  h

• NJ and CT are party States

!  ! I l a [PA h {Q ~ MD rwJ

• No host State selected J '"l 80 3 '

l J l MO DE

• 5% National LLW

{8 g7 g f gy

• Burial technology to be i

!" / / y DC g determined by host state

                                                                                      /

{e CALIFORNIA

                                                                                                             /                                                            .         APPALACHIAN f /r                          /                                        CENTRAL MIDWEST                 e Introduced in 99th e CA to host site e 9% National LL
                                           ;                           g                 wy        ,,, #

• IL is host State - Congress. 2nd Session, i

i e SLS

• 14% National LLW for consent i I AZ W '

NE il

• SLB banned
• PA is host State
• 11% National LLW CO l WESTERN NV y
• SLB banned
• Not yet introdwesd into KS .

Congress for consent , _

• AZ is host State

                   * <1% National LLW

• SLB NM OK AR F I TN / NC .

ROCKY MOUNTAIN f SC k

• NV current host State LA j AL GA '

'

• CO next host State i
• SC le now host State

                                               * <1% National LLW

• SLB 1 -J
• NC selected by SE Comunission to replace SC as host State

                                                                           '                                                                                          FL           with operating site by 1991       s TEXAS                                                                                        '        *"*'

s - CENTRAL' , , **C,'P8 G 2 2 ~4

• e,
• TX to host site

                                                                                    ,                                          e No host State selected

• New burial technology to be ans uw nesomess see * <1% National LLW e 3% National LLW determined by NC
• SLB restricted
• SLB restricted .

E M P PR h

          . g Note: National LLW wolume for 195 = 2.00 tnillion cubic feet.

Sowee: Office of State Programs. NRC 9 e e

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p. 320TH ACRS MEETING APPENDIX XIV ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Report, Revitalizing Nuclear Safety Research, Committee on Nuclear Safety Research, Consnission on Physical Sciences, Mathematics and Resources, Mational Research Council, 1986

2. lE Information Notice No. 86-99: Degradation of Steel Containments, U.S. NRC, Office of Inspection and Enforcement, December 8,1986

3. Report, M. Sander, R. Loth, Brown Boveri ReaktorGmbtt, Design and Operational Aspects of the Decay Heat Removal Systems of the BBR-Reactor,

4. SECY 86-357, V. Stello, EDO to the Commissioners, NRC Staff Review and Proposed Follow-Up Actions Related to the DOE Final Environmental Assessments for the First High-Level Radioactive Waste Repository, November 28, 1986

6. Report, International Conference-by-Computer September 29 through October 17, 1986, Chernoby1: Where Do We Go From Here?

6. SECY-86-1F, V. Stello, Jr., ED0 to the Comissicners, Status of Staff Actions Regarding TVA- December 9,1986

\

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