ML20055C250

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Summary of CRGR Meeting 167 Held on 890809.Committee Recommended Against Forwarding Proposed Amend to Commission on Basis That Safety Benefit Not Demonstrated as Necessary to Ensure Adequate Protection,As Staff Claimed
ML20055C250
Person / Time
Issue date: 08/23/1989
From: Jordan E
Committee To Review Generic Requirements
To: Taylor J
NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
References
FRN-59FR979, REF-GTECI-070, REF-GTECI-094, REF-GTECI-NI, RTR-NUREG-CR-4692, RTR-NUREG-CR-5186 AC93-1-025, AC93-1-25, NUDOCS 8909140258
Download: ML20055C250 (25)
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{{#Wiki_filter:' dw c ![:- %jo, UNITED STATES g NUCLEAR REGULATORY COMMISSION - wAsmwoTow. o c. roess - %....,/ August 23, 1989 MEMORANDUM FOR: James M. Taylor Acting Executive Director for Operations s FROM: Edward L. Jordan, Chairman Committee to Review Generic Requirements

SUBJECT:

MINUTES OF CRGR MEETING NUMBER 167 The Committee to Review Generic Requirements (CRGR) met on Wednesday, i August 9,1989 from 1:00-5:00 p.m. A list of attendees for-this meeting is attached (Enclosure-1). The following items were addressed at the meeting:

1..

L. Shao (RES), R. Bosnak (RES), and W. Norris (RES) presented for CRGR review a proposed amendment to 10 CFR Part 50.55a, Codes and Standards.- This amendment incorporates by reference subsection IWE of the ASME Code, and addresses inservice. inspection requirements for_ metallic containment liners. The Committee recommended against forwarding the proposed amendment to the Commission on the basis that the safety benefit was not demonstrated as necessary to ensure adequate protection, as the staff had claimed. This matter is discussed in Enclosure 2. 2. L. Shao (RES), R. Bosnak (RES), and S. Aggarwal (RES) presented for CRGR { review a proposed Regulatory Guide DG-1002, " Isolation Devices." The Committee-recommended against forwarding the proposed regulatory guide to { the Commission on the basis that an analysis is needed since some of the i criteria included in the guide would constitute a backfit for some j licensees. This matter 13 discussed in Enclosure 3. l 3. W. Minners (RES), R. Baer (RES), and F. Cherny (RES) presen+.ed for CRGR' ~ review proposed resolutions for Generic Issue 70, " Power Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, " Additional Low-Temperature Overpressure Protection for Light Water Reactors." The Committee did not complete their review of these items and will' continue their review at the next scheduled meeting. This matter is discussed in. i In accordance with the ED0's July 18, 1983 directive concerning " Feedback and Closure of CRGR Reviews," a written response is required from the cognizant office to report agreement or disagreement with the CRGR recommendations in these minutes. The response, which is required within five working days after receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with CRGR recommendations, to the EDO for decisionmaking. di.JmeA t

Questions concerning these meeting minutes should be referred to Jim Conran (492-9855) W Signed ky E.L Jordan Edward L. Jordan, Chairman Committee to Review Generic Requirements x

Enclosures:

As stated cc w/ enclosures: Commission (5) SECY J. Lieberman P. Norry M. Malsch Regional Administrators CRGR Members y ~= Distribution: (w/o enc.) 2 Central File PDR (NRC/CRGR) S. Treby W. Little M. Lesar P. Kadambi (w/ enc ) CRGR CF (w/ enc.) CRGR SF (w/ enc.) M.. Taylor (w/ enc ) L. Shao (w/ enc.) R. Bosna.,(w/ enc.) W. Minners (w/ enc.) W. Norris (w/ enc.) S. Aggargal (w/ enc.) R. Baer (w/ enc.) E. Jordan (w/ enc.-) J. Heltemes (w/ enc.) J. Conran (w/ enc.) C. Sakenas (w/ enc.) , f /)r .~ 0FC

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4. g- =i IRcDon Fue Note and Retum .I g_ Per cieerense per comerneson 1 ire penessearl Per Corvoegen Process Res& -l DlRMlete For Your information See Me benunedt Isweettaste Stanoture b_^~^^-- Justik anwege _j This previous Central File material can now be i made publicly available. l' i I C C (llH O A/L Y) ha n kcchhj e PDie L St. I ,t 00 NOT use this form as a RECORD cf approvals, concurrences, disposals, eteerences, and similar act6ons l-FR000: (Nome, org. symbol, Agency / Post) Room No.- -Bldg. 'I t L/ $ 0 (/ pn.n, no, "'I" ggAl. g 41 (Rev. 7-76) FNR (41 ck sos-11. sos @c.s. coo ier.o toi. 47 am l 1 1 J q 045 f i o

I Materip Related to CRGR Meeting No.167 Tu Be Made Publiclv Available 1. Memo dated August 23, 1989 for J. Taylor from E. Jordan,

Subject:

Minutes of CRGd Meeting Number 167, including 3 enclosures which were not previously released

a., a sumary of discussions of a proposed amendment to 10 CFR Part 50.55a regarding subsection lWE of the ASME Code, I

including 1 attachment.

b., a summary of discussions of a proposed Regulatory Guide DG-1002 regarding isolation devices, including 1 attachment.

i f

c., a summary of discussions of a proposed resolution of l

GI-70 regarding PORY and Block Valve Reliability and GI 94 L regarding additional LTOP for LWR's including attachments. I 2. Memo dated June 13, 1990 for E. Jordan from E. Beckjord forwarding review materials on a proposed amendment to 10 CFR 50.55a. 3. Memo dated July 16, 1990 for E. Jordan from E. Beckjord forwarding review additional materials on a proposed amendment to 10 CFR 50.55a. 4. Memo dated July 21, 1990 for E. Jorden from E. Beckjord forwarding review materials on a proposed Regulatory Guide DG 1002. 5. Memo dated July 11, 1990 for E. Jordan from E. Beckjord forwarding review materials on a proposed resolution for GI-70 and 94. l i SENT TO PDR ON 7/30/90 l i 1 i m,- -~ wn, ,,..,,....-n.

ATTENDANCE LIST FOR CRGR MEETING NO. 167 August 9, 1989 CRGR MEMBERS E. Jordan J. Sniezek L. Reyes G. Arlotto J. Goldberg B. Morris (for D. Ross) NRC STAFF J. Conran C. Sakenas

0. Allison L. Shao R. Bosnak W. Norris A. Murphy G. Millman J. Costello C. Y. Cheng K. Wichman R. Hermann S. Aggarwal J. Joyce B. Hayes M. Vagins S. Newberry C. Doutt W. Minners R. Baer F. Cherny R. Kirkwood i

G. Mazetis E. Throm i M. Lopez-Otin 1

I 1 1 to the Minutes of CRGR Meeting No. 167 Proposed Amendment to 10 CFR Part 50.55a, Codes and Standards August 9, 1989 i TOPIC L. Shao (RES), R. Bosnak (RES), and W. Norris (RES) presented for CRGR review a proposed amendment to 10 CFR Part 50.55a, Codes and Standards. This amend-ment incorporates by reference the 1989 edition of Subsection IWE, " Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water-Cooled Power Plants," of Section XI of the ASME B&PV Code. A copy of the slides used } by the staff to guide their presentation and the discussions with the Committee at this meeting is attached to this enclosure. BACKGROUND 1 The packages submitted by the staff for CRGR review of this matter were transmitted by memoranda dated June 13, 1989 and July 6, 1989. E. S. Beckiord to E. L. Jordan. The packages included the following: 1. the proposed rule 2. summary of CRGR review items 3. Regulatory Analysis CONCLUSIONS / RECOMMENDATIONS As a result of their review of this matter, including discussions with the 4 staff at this meeting, the Committee recommended against forwarding the pro-posed rule to the Commission on the basis of being needed to ensure adequate protection. The staff was unable to demonstrate that a sufficient safety problem existed or that existing requirements (Appendix J) were insufficient to ensure that licensees would address the issue. The consensus of the Committee was that this endorsement would go well beyond any present inspection requirements, especially in visual inspection of liner welds. l Aithough the Committee supported the coatings inspection, it was noted that the very real problem of corrosion, especially of inaccessible surfaces, was not addressed although a number of real instances have occurred. If additional guidance is needed in this area, the Committee recommended that the staff develop guidance in this area, outside the ASME' Code context, if necessary. Before returning this package to the Committee, the safety benefit resulting from this action should be demonstrated, or the action should be revised to only address the safety issue. It was recognized by the Committee that this is the first time a consensus standard has failed to be incorporated based on backfit considerations.

slide 1 PROPOSED AMENDMENT TO 10 CFR PART 50.550 CODES AND STAN DAR DS FOR N UCLEAR POWER PLANTS Task Leader: Wallace E. Norris Office of Nuclear Regulatory Research 4 CRGR Meeting of August 9, 1989 N ...-n..

l slide 2 1 i i PROPOSED RULE i INCORPORATES BY REFERENCE THE 1989 EDITION 1 OF SUBSECTION

IWE,

" REQUIREMENTS FOR CLASS i MC AND M ETALLIC LINERS OF CLASS CC COMPONENTS OF LIGHT-WATER-COOLED POWER PLANTS", OF SECTION XI, OF THE ASME B&PV CODE

  • ~

slide 3 SCOPE o CLASS MC (METAL CONTAINMENTS) PRESSURE-RETAINING COMPONENTS (e.g., CONTAINMENT WELDS, PENETRATION WELDS, AIRLOCK WELDS, SEALS AND GASKETS) o CLASS CC (CONCRETE CONTAINMENTS) PRESSURE-RETAINING COMPONENTS (e.g., METALLIC SHELL AND PENETRATION-LINER WELDS)

slide 4 REVIEW AND STAFF APPROVAL O THERE ARE TWO PACKAGES WHICH ARE TO BE COMBINED o THE MAIN PACKAGE WOULD INCORPORATE BY REFERENCE THE 1986 EDITION WITH ADDENDA 1 THROUGH THE 1987 ADDENDA o THE SECOND PACKAGE WOULD INCORPORATE i BY REFERENCE THE 1988 ADDENDA i o THESE TWO PACKAGES COMBINED ARE THE 1989 EDITION o THE 1989 EDITION IS IDENTICAL TO THE 1986 EDITION AS MODIFIED BY THE 1986 THROUGH THE 1988 ADDENDA

slide 5 STATEMENT OF THE PROBLEM o CURRENTLY THERE IS NO CONSISTENT ISI OF CONTAINM ENTS o SECTION III (CONSTRUCTION CODE) DOES NOT REQUIRE CORROSION ALLOWANCES o STUDIES SUCH AS TIRGALEX AND NPAR ASSIGN CONTAINMENTS THE HIGHEST PRIORITY IN RISK STUDIES o EROSION OF THE METAL DRYWELL SHELL AT ONE PLANT WAS FOUND TO BE OCCURRING AT THE RATE OF 20 MILS / YEAR. o AT ANOTHER PLANT, TORUS SHELL WALL THICKNESS WAS AT OR BELOW MINIMUM SPECIFIED WALL THICKNESS. o THREE OTHER PLANTS WILL NEED TO CLEAN AND RECOAT THEIR TORI.

slide 6 4 OBJECTIVES l o ENSURE THE CONTA!NMENT MAINTAINS PF' ESSURE RETAINING INTEGRITY o ESTABLISH NRC POSITION ON A GENERIC BASIS SUBSECTION.lWE WOULD SATISFY, IN PART. o GENERAL DESIGN CRITERION 1 (Quality Assurance Program) o GENERAL DESIGN CRITERlON 16 (Containment Design Conditions) o GENERAL DESIGN CRITERION 53 (Appropriate Surveillance Program) .... ~~.T~~-.

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h 9 slide 7 OBJ ECTIVES (SLIDE 2) i 4 SU BSECTION IWE WOULD SATISFY, IN

PART, i

o TECHNICAL SPECIFICATIONS SECTION

4. 6.1. 7 (Visual inspection of containment during Type A test) i o TECHNICAL SPECIFICATIONS SECTION 4.6.1.7.3 (Visual inspection of accessible interior and exterior su rfaces) a 4

L e y

sifde 8 OBJ ECTIVES (SLIDE 3) SUBSECTION IWE WOULD

SATISFY, IN
PART, o THE GENERAL INSPECTION REQUIRED BY APPENDIX J o APPENDIX B OF PART 50 (Documented Quality Assurance Prog ra m)

slide 9 DESCRIPTION OF THE ACTIVITY REQUIRED BY LICENSEE o DEVELOPMENT OF AN INSERVICE INSPECTION (ISI) PLAN o PERIODIC UPDATES TO ISI PLAN o PERIODIC INSERVICE INSPECTIONS IN CONFORMANCE ~ W!TH THE ISI PLAN

INDUSTRY PARTICIPATION IN SU BSECTION IWE DEVELOPMENT ASME BOILER & PRESSURE VESSEL COMMITTEE e w 4' SUBCOMMITTE ON INSERVICE INSPECTION e SG CONTAINMENT

slide 10A ASME BOILER PRESSURE VESSEL CODE SINCE 1972 THE NRC HAS BEEN ENDORSING THE CODE o SECTION 111 - RULES FOR CONSTRUCTION OF NPP COMPONENTS - CLASS

1. CLASS 2, AND CLASS 3 COMPONENTS o SECTION XI -RULES FOR ISI OF NPP COMPONENTS

- SUBSECTION IWA: GENERAL REQUIREMENTS - SUBSECTION IW8: CLASS 1 COMPONENTS - SUBSECTION IWC: CLASS 2 COMPONENTS - SUBSECTION IWD: CLASS 3 COMPONENTS - SUBSECTION IWE: CLASS MC COMPONENTS - SUBSECTION IWL: CLASS CC COMPONENTS (RECENTLY PASSED BY THE SUBCOMMITTEE) L.-_____ R - - ~ '

') } l slide 11 PROPOSED MODIFICATION AND LIMITATION o MODIFICATION LICENSEES WILL BE REQUIRED TO IMPLEMENT SUBSECTION IWE DURING THE FIRST PERIOD OF THE NEXT INSPECTION INTERVAL WHICH INCORPORATES SUBSECTION IWE o LIMITATION - EDITIONS NO EARLIER THAN THE 1989 EDITION SHALL BE USED.

F I i 4 -- R d i M A wRd E E

== m o a 2 e = ait O as m g d b to a k WM d 8 5 5t E_ \\ 5 ' s d 8 E y G W ~ E ,, m = 4 b t g 8 t ,_a o o t

slide 12 L 1 POTENTIAL IMPACT ON' RADIOLOGICAL EXPOSURE OF EMPLOYEES o IMPLEMENTATION OF THIS PROPOSED RULE IS NOT EXPECTED TO SIGNIFICANTLY INCREASE THE OCCUPATIONAL EXPOSURE ASSOCIATED WITH THE ISI EXAMINATIONS o $600 PER. REACTOR OR $75,000 FOR THE REACTOR. POPULATION 4 ARE -VIEWED AS Nil WHEN COMPARED TO COST OF IMPLEMENTATION i- .. ::::: : ~, T.:^.~ ~ ~ ~::::2 7 :: ::.. :.~. L:-- -. -......,-,-

slide 13 BACKFIT STATEMENT - 50.109(a)(4)(ii) THE NRC STAFF HAS DETERMINED THAT SUBSECTION IWE PROVIDES MINIMUM REQUIREMENTS FOR THE CONTAINMENT INSPECTIONS, AND, REPRESENTS RESPONSIBLE APPLICATION OF ENGINEERING JUDGEMENT TO ENSURE ADEQUATE PROTECTION OF THE PUBLIC HEALTH AND SACETY. -.h me. mWe e-4- es-e w 4 e e w 6-@MuWim spil. e e W I-. 6.-.e & Nene .e. Ae -e.-3 y e

1 slide 14 i 4 h I l COSTS o ALL PLANTS ARE AFFECTED 1 o THERE IS A ONE-TIME COST TO A FACILITY OF APPROXIMATELY $241 K TO DEVELOP THE ISI PLAN o TO COMPLY WITH THE ISI REQUIREMENTS THEN, THE COST TO FACILITY OVER A 30 YEAR -PERIOD (1988 DOLLARS) WILL BE ROUGHLY $478K 2 :=:

slide 15 SAFETY IMPACT OF CHANGES IN FACILITY OPERATION SUBSECTION IWE EXAMINATIONS WOULD BE PERFORMED DURING PLANNED SHUTDOWNS AND WOULD NOT IMPACT FACILITY OPERATION _...m.

~ slide 16 OTHER IMPACTS OF IMPLEMENTATION SUBSECTION IWE AUGMENTS THE GENERAL INSPECTION CALLED FOR IN APPENDIX J. IT PROVIDES A CONSISTENT SET OF RULES WITH APPROPRIATE EXAMINATION DETAILS FOR -CONTAINMENT STRUCTURES i i ~.<,- -,, w

&

w-a v '-~ - * - " - ' ^ * - * - - " " ~ * ~ ~ " - * - * ~ * - *

= slide 17 POTENTIAL IMPACT OF DIFFERENCES IN FACILITY TYPE o APPLIES TO. ALL PLANTS o RELIEF REQUESTS EXPECTED TO BE FEW IN NUMBER o OLDER PLANT DESIGNS MAY NOT BE ABLE TO COMPLY WITH ACCEPTABILITY PROVISIONS ~

slide 18 SINCE o RECENT OPERATING EXPERIENCE lNDICATE DEGRADATION OF CONTAINMENTS o TOTAL ABSENCE OF ANY SPECIFIC ISI -PROGRAM i CONCL.USION-o RULE SHOULD BE IMPLEMENTED AS.SOON AS PRACTICABLE

. to the Minutes of CRGR Meetina No. 167 Proposed Regulate'y Guide DG-1002, " Isolation Devices" ~ August 9, 1989 TOPIC L. Shao (RES), R. Bosnak (RES) and S. Aggarwal (RES) presented for CRGP. review a proposed Regulatory Guide DG-1002, " Isolation Devices." Copies of tte slides used by the staff to guide their presentation and the discussions with the Committee at this meeting are attached to this enclosure. BACKGROUND The documents submitted for CRGR review in this matter were transmitted by memorandum dated July 21, 1989 E. S. Beckjord to E. L. Jordan anc included a copy of the draft regulatory guide and responses to the CRGR Chsrter require-ments. CONCLUSIONS / RECOMMENDATIONS As a result of their review of this matter, including discussions with the staff at this meeting, the Committee recommended against forwarding the draft regulatory guide to the Commission on the basis that a backfit analysis was not performed. The Committee stated that although these positions may refle:t staff practice, there is no previously documented staff position which addresses all of the criteria listed in the regulatory guide, and an analysis is needed which shows that there would be a substantial benefit from implementation, since this would be a backfit for some licensees, i

1 l 4 i 3 resen<:o: ion

o CRGR l

Augus 9, ' 989 j j Requ o:ory Guice JG '002 so c: ion Jevices By Sa:is, <. Acgarwa L Eec:rica & Veclanica Engineering

3ronc, Jivision o" Encineerinc 0" ice o' Nuc ear Regu o:ory Researc,

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I l i i i i. i so a: ion Jev. ices Recu a:ory Basis Porograph (h) of Section 50.550 invokes IEEE Std 279-1971, which requires THAT TRANSMISSION OF SIGNALS FROM PROTECTION SYSTEM EQUIPMENT BE THROUGH IS01.AT10N i DEVICES THAT MEET ALL THE REQUIREMENTS OF THE PROTECTION SYSTEM, l $00Ug92

--._....._.-......_.... ~ L i I ) 4 WHEN IT IS NECESSARY l To transmit a singal from (a) safety system to a safety system interface or (b) safety system to non-safety system The signal must be transmitted through an isolation device, which is capable of withstanding credible failures, sooug93

i i l These credible failures include: i .l SHORT CIRCulTS OPEN CIRCUITS GROUNDS V0LTAGE SURGES APPLICATION OF MAXIMUM CREDIBLE FAULTS If an isolation device fails, its failure must not have an adverse effect on the safety system. sooug94 l

.t.7 Control and Protection System Inter-action.2 .t.7.1 Classification of Equ.ipment. Any equipment that is used for both protective and control functions shall be classified as part of the protection. system and shall meet all the L requirements of this document. L 4.7.2 Isolation Devices. The transmission of signals from-protection system equipment I for control system use shall be through isola-tion devices which shall be classified as part of-i -the protection system and shall meet all the requirements of this document. No credible failure at the-output of an isolation -device shall prevent the associated protection system channel from meeting the mi.nimum perform-ance requirenc "c specified .in the design bases. Examples A creF 'e failures. include short circuits, oper. grounds, and-the appli- .ca m cation of tl e 3 naxi' tum credible ac. or de potential. A Laure in an isolation device is 1 evaluated:in the same manner as a failure of other equipment in the protection system. 4.7.3 Single Random Failure. . W here a single random failure can cause a control system action that results in a generating station condition requiring protective action and can also prevent proper action of a pro-i tection system channel designed to protect ( against the conditio nm the w

  • l

1. l i CURRENT LICENSING BASIS for the past 10 years the provisions contained j in this Proposed Regualtory Guide have been uniformly applied for all licensing evaluations. These existing technical positions predate CRGR. Issuance of this guide iscdesirable .(1) to document an existing technical position (2) to maintain uniform application 500U995

l l Il0NS

ystem
-

the rrupting devices acuated i he rrent are considered to be the isolation devices (except

d fuses).

,ed for y Position C.1: t'he 1 'or the effects. pting devices actuated by fault and then ered acceptable isolation devices ts. 7ty-related to non-safety-related tcontrol circuits, with the exception or pedu6auon and fuses. (This does not apply to

) that would

'uses in electrical power feeder ie safety system. scoug97 scoug96

s l= .afety 8ystem l Non-safety System-1 4 o-I eC I l.0LATION u i DEVICE t MCP T j. p j l o' l,67 'u l c I l l l MCP i 2 Pleure 1

  • MCF test'should be applied to l

l the power supply input if the operating voltage is less than the MCF. fety.yste i.efe,v.ystem. DMelon I i . DMalon 2 o' I c o l = n MW P I T u I P t I u o j. 'o I I I l Pigure t (MCF is applied to both sides) MCF = Maximum Credible Fault 1 T = Test equipment (oscilloscope, chart recorder) sensitive enough to detect any perturbation, i

i l l i Safety System i Non-safety System t I l .O -l oC I ISOLATION u-DEVICE t T P l u I .P-SWC 4 i t I O l 'C ~ i i { I I t Figure 3 (- SWC = Surge Withstand Capability voltage t T = Test measurement equipment ~ k l .l 1 ~, w w-e + w-r, n.-w--

F s r (3) For a safety to a safety system interface: i in addition to. Position 2, the credible failures should be applied to the input terminals.- l The isolation Device should not allow any energy pass-through from the output to the input or input to the output that will have-a detrimental effect.on the safety system (opposite from where the fault is applied). sooug98

~ O i q (4) When using relays as isolation devices: .i Qualification testing should include an 4 oscillatory voltage surge test - to assure that voltage surges in the non-safety system are. effectively blocked. IEEE Std 472/1974/ ANSI C37.90a - 1974-Surge Withstand Capability Tests - Acceptable. scoug99 - - - - ~

L Y i (5) Physical arrangement of components in an electric isolation device should be designed to prevent, in the event-of failure, shattered parts (such as solder splatter) from adversely' affecting the safety system. a sooug910

6 L IMPLEMENTATION 1. cps issued after issue date of this guide f 2. OLs docketed 6m after issue date of this guide 3. Operating plants that add or modify a safety system design requiring the use.of isolation device saoug911

i - g l l to the Minutes of CRGR Meetina No. 167 Proposed (Combined) Resolution for GI-70 and GI-94 August 9, 1989-TOPIC R. Baer (RES)-and F. Cherny (RES). presented for CRGR review the propcsed (combined) resolution for Generic Issue 70 (" Power-Operated Relief Valve and Block Valve Reliability") and Generic Issue 94 (" Additional Low-Temperature Overpressure Protection for Light Water Reactors"). Copies of-the slides used by the staff to guide their presentation and the discussions with the Commit-tee at this meeting are enclosed (Attachments 1 and 2). BACKGROUND The documents submitted for CRGR review in this matter were transmitted by memorandum dated July 11, 1989,. E. S. Beckjord to E. L. Jordan; the review package included the following documents:

1. - Background Information for CRGR Review of GI-70 and GI-94 Resolutions (submitted in accordance with Section IV.B of the CRGR Charter) r
2. - Proposed Generic Letter (undated), "NRC Position on the Resolution of Generic Issue 70...and Generic Issue 94..."

3. Enclos 3 - Draft NUREG-1316 dated June 1989, " Evaluation of Power-Operated Relief Valve-and Block Valve Reliability in PWR Nuclear Power Plants" t

4. - NUREG/CR-4692 dated October 1987, " Operating Experience Review of Failures of Power Operated Relief Valves and Block Valves in Nuclear Power Plants"
5. - NUREG/CR-4999 dated March 1988, " Estimation of Risk -

Reduction from Improved PORV Reliability in PWRs"

6. - Proposed Revision 2 (undated) to Standard Review Plan (SRP) Section 3.2.2, " System Quality Group Classifi-cation"
7.

-Proposed' Revision 3 (undated) to SRP Section-5.2.2, " Overpressure Protection"

8. - Proposed Revision 3 (undated) to SRP Section 5.4.7,

" Residual Heat Removal System" 9. Cnclosure 9 - Safety Issues Management System (SIMS) data for GI-70, dated July 29, 1989

2 1

10. 0 - Draft NUREG-1326 dated March 1989,-" Regulatory Analysis for the Resolution of Generic Issue 94:

Additional i Low-Temperature Overpressure Protection for Light-Water Reactors"

11. 1 - NUREG/CR-5186 dated November 1988,." Low Temperature Overpressure Systems for Pressurized Water Reactors"
12. 2 - SIMS data for GI-94, dated _ July 29, 1988 l

l CONCLUSIONS / RECOMMENDATIONS The Committee did not complete their review of these issues at this meeting; the review will be continued at the next scheduled CRGR meeting. The staff will. revise the package (in particular the proposed generic letter) to reflect CRGR comments / suggestions from the discussions at this meeting, and will provide the revisions to the Committee prior to the next meeting on these

topics, i

1 l

PRESENTATION TO CRGR AUGUST 9, 1989 GENERIC ISSUE 70 PORV AND BLOCK VALVE RELIABILITY RESPONSIBLE ORGANIZATION: EIB/DSIR/RES SENIOR TASK MANAGER: R. KIRKWOOD ElB PRESENTATION BY:. F. CHERNY SECTION LEADER EiB-f SLIDE 1 to Enclosure 4

? L i HISTORICAL

SUMMARY

PRIORIT'IZED AS MEDIUM IN MAY 1984 TASK ACTION PLAN APPROVED: (APRll 1985) PRESENTATION TO ACRS S/C IN APRIL 1985 ~ STAFF COMMISSION PAPER ISSUED (FEBRUARY 1986)- 'RE: WHETHER PORVs Sil0VLD BE SAFETY GRADE PRESENTATION TO ACRS S/C IN JANUARY 1989 PRESENTATION TO ACRS. FULL COMMITTEE IN FEBRUARY 1989 e e SLIDE 2 i l

l i l WHAT 00 TYPICAL PORVs LOOK LIKE? 4 m 1 e s SLIDE 3

o 0 % t. Ow(, a p. ag,g g gio 1 ADJU$1 tNG $CRt W ? B E L LOWS 19 3 8tLLOW5 FLANGE Q 8 8t LLOW5 Pa570N 6 Ot$C $PRING (( 18 \\ 5 CAGE Q 7 DR AIN 8 GUIDE 'i 4 9 LIVER l 10 LO wtR $P:N DLE I t1 Pilot B Ast g. ir V U 12 PILOT Ot$C ./' 13 PILOT St AT BUSHING 14 PLUNGER $PRING 15 SOL E NOID 15 16 SOLINCIDCOVER i P N BRACKET (T=T) fr=w, f-te switch Asstu8tw o o 20 UPPE R $PINDLE 21 VALVEBODY r. ?? VALVE OtSK e b v ',(x oI a to j o e OUTLET ~' 3 I )l d e 'N ( =- c h 7' i i -ppp' 8 7 m -g g ._g M ! I -d it .i >j%_ F/ = 1 INLLI a s PILOT-0PERATEDRELIEFNALVE (CourtesyofDresserIndustries)' SLIDE L1

e omNL-owG 8 7 3690 t to i r 1 00NN(T-7 CAGE WITH St AT s 3 CAG(SPACEM 4 GL AND 8 0L.LOWE R 5 guide SU$HING k ) 6 LE AKOFF CONNECTION \\ 7 OPER ATOR A$$tMBLY \\ B P ACKING ,1 9 P ACKING GLAND ~~ 10 PLUG 11 Situ j 12 VALVE BODY i g e Vs \\sc' c e h K! 7 0' s Y mIMMIh atow '/ t i s s s AIR-0PERATED (SPRING-LOADED) RELIEF VALVE (Courtesy of.' Copes-Vulcan) SLIDE 5 l

l i i l WHERE/HOW ARE PORVs AND BLOCK VALVES INSTALLED ON PWRS IN THE V.S.? t 9 e 9 e k SLIDE 6

-{ N VN y mm ~ 3n .YALVES A 3" NDLT VALVES A VA b 6" YALYES 6" 6" 6" l 8 l b jN 4.. 12" PFSSSUMZER - i rm ){ i ~ ~ - - -. -~ C00lANT T.%T y-g;;g n;,,7;,x,7; .) th0PS PRESSURIZER SAFETY AND REUEF VALVES SL1DE 7

i i GI-70 AREAS OF EVALUATION l r i NUREG-1316 - ADDRESSES THE FOLLOWING TOPICS: o HISTORICAL USE OF PORVs o SAFETY FUNCTIONS THAT ARE PERFORMED BY PORVs o OPERATING EXPERIENCE (ORNL) NUREG/CR-4692 o PRA RISK REDUCTION STUDY (BNL) NUREG/CR-4999 + o CURRENT PORV AND BLOCK VALVE CONSTRUCTION REQUIREMENTS o RECOMMENDED CHANGES o SUPPORTING REGULATORY ANALYSIS SLIDE 8-

i. 1 HISTORICAL' USE OF PORVs l 'o AUTOMATIC ACTUATION AT PRESSURE LOWER THAN SAFETY VALVE ACTUATION PRES'SURE TO AVOID SAFETY ~ VALVE ACTUATION i] DURING ROUTINE PLANT PRESSURE TRANSIENTS i o NO CREDIT TAKEN IN-ASME CODE OVERPRr.SSURE PROTECTION REPORT OR FSAR FOR PORV PRESSURE RELIEF CAPABILITY FOR TRANSIENTS DURING NORMAL PLANT OPERATION of INADVERTENT OPENING OR STUCK OPEN PORV IS A CHAPTER 15 FSAR ANALYZED EVENT -'" ANTICIPATED OPERATIONAL OCCURRENCE" l 0 PRE-1979 LICENSED PLANTS-D0 N0T HAVE SAFETY GRADE PORVs. THAT IS, VALVE OPERATORS AND THEIR ELECTRICAL-CONTROL. SYSTEMS WERE NORMALLY DESIGNED TO NONSAFETY-RELATED-STANDARDS. HOWEVER, IHE PRESSURE-RETAINING ELEMENTS OF-PORVs ARE WITHIN THE RCPB AND WERE-CONSTRUCTED TO THE SAME CODES AND STANDARDS AS'THOSE REQUIRED FOR SIMILAR SAFETY-RELATED RCPB COMPONENTS, SLIDE 9

SAFETY FUNCTIONS THAT ARE PERFORMED BY PORVs-F 4 o PORVs ORIGINALLY PROVIDED FOR PLANT OPERATIONAL FLEXIBILITY, o ROLE OF PORVs HAS CHANGED OVER A PERIOD OF TIME, PORVs ARE NOW RELIED UPON BY MANY WESTINGHOUSE, B8W, AND CE PLANTS WITH PORVs TO PERFORM ONE OR MORE OF THESE SAFETY-RELATED FUNCTIONS: i 1. MITIGATING A DESIGN BASIS STEAM GENERATOR TUBE RUPTURE ACCIDENT, 2. LOW-TEMPERATURE OVERPRESSURE PROTECTION OF THE: REACTOR VESSEL DURING STARTUP AND SHUTDOWN, 3. PLANT C00LDOWN IN COMPLIANCE WITH BTP RSB-5-1, OR. l 4. REACTOR COOLANT SYSTEM VENTING,- o PORVs ALSO PROVIDE SAFETY-RELATED FUNCTIONS FOR EVENTS BEYOND THE D$~ SIGN BASIS, SUCH AS FEED AND BLEED AND l ATWS MITIGATION, l SLIDE 10 l

OPERATING EXPERIENCE (0RNL) NUREG/CR-4692 o NUREG/CR-4692, "0PERATING EXPERIENCE REVIEW 0F FAILURES 0F POWER OPERATED REllEF VALVES AND BLOCK VALVES lN NUCLEAR POWER PLANTS," WAS PREPARED BY ORNL IN SUPPORT OF THE RESOLUTION OF_GI-70. o REPORT REVIEWED EVENTS REPORTED FROM 1971 TO MID-1986, EACHEVENTWASJUDGEDASTOTHESEVERITYOFITHEDEGREE o l OF FAILURE AS FOLLOWS: 1. " DEGRADED" (BUT OPERABLE), THE COMPONENT OPERATED AT LESS THAN ITS SPECIFIED PERFORMANCE LEVEL,.AND 2. " FAILED," THE COMPONENT WAS COMPLETELY UNABLE TO PERFORM ITS FUNCTION, o FAILURE SEVERITY OF PORV EVENTS:; DEGRADED FAILED TOTAL PORV-MECHANICAL 77

24 101 PORV CONTROL 30 61 91 PORV DESIGN -

6 _Q 6 113 '85 -198 0 23% OF THE PORV MECHANICAL EVENTS AND 67% OF THE'PORV CONTROL EVENTS WERE-FAILURES, i SLIDE 11 w

PRA RISK REDUCTION STUDY (BNL NUREG/CR-4999 1 o NUREG/CR-4999, " ESTIMATION OF RISK' REDUCTION FROM IMPROVED PORV RELIABILITY IN PWRs," WAS. PREPARED BY BNL IN SUPPORT OF THE RESOLUTION OF GI-70, o USED EXISTING PRAS (INDIAN PolNT 3 AND OCONEE 3) TO PERFORM AN ANALYSIS OF RISK REDUCTION FOR IMPROVING PORV AND BLOCK VALVE RELIABILITY, o-CORE MELT FREQUENCIES ATTRIBUTABLE TO PORV OR BLOCK l VALVE FAILURES'WERE FOUND TO BE RELATIVELYz INSIGNIFICANT AND TO REPRESENT ONLY A VERY SMALL FRACTION 0F TOTAL CORE MELT FREQUENCY ATTRIBUTABLE.TO INTERNAL PLANT EVENTS, .o NRC STAFF BELIEVES BNL RESULTS-UNDERESTIMATES SAFETY BENEFIT THAT WOULD BE ACHIEVED BY IMPROVING'PORY AND BLOCK VALVE RELIABILITY, o HOWEVER, BNL FAULT TREES ARE DOMINATED BY OPERATOR ERROR CONSIDERATIONS, AND IT DOES NOT: APPEAR THAT RESULTS WOULD l HAVE CHANGED A GREAT DEAL EVEN IF HIGHER PORV FAILURE L RATES WERE USED, o CONSIDERATION OF FEED AND BLEED (NOT'WITHIN' THE SCOPE OF THE BNL STUDY - EVALUATED.AS PART-0F'USI A-45) INDICATES-A MUCH GREATER SAFETY IMPORTANCE OF PORVs AND BLOCK VALVES, SLIDE 12

l SAFETY BENEFITS (FEED AND BLEED) o EFFECT OF FEED AND BLEED UPON PROBABILITY OF CORE' MELT P(CM) EXAMINED IN NUREG/CR-5230 AS.A SENSITIVITY ISSUE AS A PART'0F THE RESOLUTION OF USI A-45 i 0 P(CM) FOR INTERNAL EVENTS ONLY, WITH AND WITHOUT FEED-AND BLEED CALCULATED. REPORT INDICATES FEED AND BLEED REDUCES ESTIMATED P(CM) BY SIGNIFICANT AMOUNT, ORDER OF 25 TO 907, (4.8E-5 TO 1.14E-3 PER REACTOR YEAR)' o CORE MELT PROBABILITY WITH AND WITHOUT FEED AND BLEED FROM THE CASE. STUDIES IN USI'A-45'(INTERNAL EVENTS ONLY) WITH REC 0VERY P(CM)PER REACTOR P(CM)PER REACTOR AP(CM) PER PLANT YEAR WITHOUT FEED YEAR WITH FEED AND BLEED AND BLEED REACTOR YEAR A 1.87E-4 1.39E-4 4.8E-5 B 1.00E-4 7.1E-5 2.9E-5 C 4.8E-5 1.4E-5 3.4E-5 D 1.23E-3 8.8E-5 1.14E-3 o DECISION TO FEED AND BLEED MUST BE MADE EARLY IN AN ACCIDENT TO BE SUCCESSFUL SLIDE 13 s

CURRENT PORV AND BLOCK VALVE CONSTRUCTION REQUIREMENTS CODES AND STANDARDS - PRESSURE ~ RETAINING PORTIONS OF o 1 PORVs AND BLOCK VALVES ARE CONSTRUCTED TO SAME CODES AND STANDARDS AS OTHER RCPB COMPONENTS IN CONFORMANCE WITH-10 CFR 50,55A.. i PORVs CURRENTLY CONSTRUCTED TO ASME-SECTION 111,. CLASS 1, PRIOR TO 1971 PORVs WERELCONSTRUCTED TO EARLIER CODES AND STANDARDS, SUCH AS DRAFT ASME CODE FOR. PUMPS AND VALVES,:USAS B31.1.0,-ETC, a o SEISMIC DESIGN - NO UNIFORM APPLICATION OF SEISMIC DESIGN REQUIREMENTS, SINCE 1972 RG 1,29 HAS SPECIFIED THAT RCPB SHOULD BE SEISMIC CATEGORY 1. UNLESS SPECIFICALLY REQUESTED BY CUSTOMER, PORVs WERE -NOT NORMALLY QUALIFIED-T0 SEISMIC CATEGORY 1 1 REQUIREMENTS, o QUALITY ASSURANCE - PORVs GENERALLY CONSTRUCTED TO QA PROGRAM IN COMPLIANCE WITH 10 CFR 50, APPENDIX B, SINCE APPR0XIMATELY 1971, PRIOR TO 1971 PORVs WERE CONSTRUCTED TO MANUFACTURER'S l QA PROGRAM, - o CONTROL SYSTEMS - NOT SUPPLIED BY VALVE MANUFACTURER BUT-BY NSSS SUPPLIER OR A-E, PRIOR-T0 TMI-2 ACCIDENT, PORVs AND BLOCK VALVE CONTROL-SYSTEMS NOT QUALIFIED TO L STANDARDS, SUCH AS IEEE-382, i SLIDE 14

PROVISION OF PROPOSED GENERIC LETTER o FOR PWR PLANTS CURRENTLY UNDER CONSTRUCTION, WHEN PORVs AND ASSOCIATED BLOCK VALVES ARE USED FOR ANY OF THE SAFETY-FURCTIONS DISCUSSED ABOVE, THESE COMPONENTS SHOULD BE CLASSIFIED-AS SAFETY RELATED AND A MINIMUM 0F TWO PORVS AND BLOCK VALVES INSTALLED, o FOR OPERATING PWR PLANTS, A NUMBER OF IMPROVEMENTS (SHORT OF UPGRADING TO FULLY SAFETY-GRADE HARDWARE) CAN INCREASE-RELIABILITY OF PORVs AND BLOCK VALVES AND PROVIDE ASSURANCE THEY WILL FUNCTION AS REQUIRED, THESE IMPROVEMENTS ARE: 1, INCLUDE PORVs AND BLOCK VALVES WITHIN THE SCOPE OF A LICENSEE'S OPERATIONAL QUALITY ASSURANCE PROGRAM IN COMPLIANCE WITH 10 CFR 50, APPENDIX B. 2. PROVIDE A MAINTENANCE / REFURBISHMENT PROGRAM FOR PORVs AND BLOCK VALVES BASED ON DETAILED WRITTEN PROCEDURES AND IMPLEMENTED BY TRAINED PLANT i MAINTENANCE PERSONNEL, 3. INCLUDE PORVs, VALVES IN PORV. CONTROL SYSTEMS, AND BLOCK VALVES WITHIN THE-SCOPE OF THE ASME SECTION XI, " INSERVICE TESTING PROGRAM." 4. FOR MOST OPERATING PLANTS, MODIFY THE LIMITING CONDITIONS OF OPERATION OF PORVS AND BLOCK VALVES IN THE TECHNICAL SPECIFICATIONS FOR MODES 1, 2,'AND 3,.THE_ INTENT FS T0 (A) ENSURE THAT PLANTS'THAT RUN WITH THE BLOCK VALVES CLOSED (E G., DUE TO LEAKING PORVs) MAINTAIN ELECTRICAL POWER TO THE BLOCK VALVES S0 THEY CAN BE READILY OPENED FROM THE CONTROL ROOM UPON-DEMAND, AND (B) NOT PERMIT UNLIMITED PLANT OPERATION WITH PORVs AND BLOCK VALVES IN0PERABLE FOR REASONS OTHER THAN SEAT LEAKAGE, 5. USE, TO THE EXTENT POSSIBLE, MORE RELIABLE PORV AND PORV BLOCK VALVE DESIGNS THAT ARE RESISTANT TO FAILURE. SLIDE 15 4

l I l SUPPORTING REGULATORY ANALYSIS STAFF ESTIMATES THAT OUTAGE AVOIDANCE COSTS BASED'ON INDUSTRY DATA REPORTED BY EPRI, WOULD FAR EXCEED COST'0F IMPLEMENTING !TEMS 1, 2, 3, AND 4. SPECIFICALLY, PRESENT VALUE ASSOCIATED WITH IMPROVEMENTS TO PORVs AND BLOCK VALVES FOR ITEMS 1,-2, 3, AND 4 IDENTIFIED AB0VE ARE ESTIMATED TO BE $127,200 FOR A PLANT WITH TW0 PORVs AND TWO BLOCK VALVES, ' RESENT VALUE 0F OUTAGE AVOIDANCE COST IS ESTIMATED-T0 BE $2 541,000, .0VERALL COST BENEFIT IS ESTIMATED TO RESULT IN A SAVINGS OF $2,413,800 PER REACTOR, s 1 SLIDE 16

a i ACRS LETTER ACRS CONCURRED IN PROPOSED RESOLUTION OF GI-70 AND 94, PROVIDEDTHEFOLLOWINGTHREECLARIFICATIONSARESPECIFICALLY. ADDEDTOTHEPLANTLTECHNICAL'SPECIFICATIONACTIONSTATEMENTS 1. W}lEN ONE OR MORE BLOCK VALVES ARE CLOSED BECAUSE OF EXCESSIVE PORV SEAT LEAKAGE, POWER SHOULD BE MAINTAINED TO BLOCK VALVE T0 ENSURE QUICK RE0PENING CAPABILITY FROM CONTROL' ROOM, THIS ITEM WAS DISCUSSED IN NUREG-1316, BUT WAS NOT INCLUDED IN MODIFIED TS DUE TO A MANAGEMENT DECISION BY NRR/0TSB, RES STAFF' ACTION: DID NOT COMPLY WITH ACRS COMMENT IN ORDER TO OBTAIN NRR CONCURRENCE, HOWEVER, THE'RES STAFF PREFERS THE REVISION TO THE PLANT TS ACTION STATEMENT AS STATED IN THE ACRS LETTER OF 2/16/89, 2, MODIFIED TS SURVEILLANCE REQUIREMENTS - RCS'SHOULD BE IN-HOT SHUTDOWN-RATHER THAN COLD SHUTDOWN WHEN PERFORMING OPERABILITY TEST ON BLOCK VALVES OR PORVs, RES STAFF ACTION: ?EVISED-MODIFIED TS TO COMPLY WITH ACRS COMMENT. SLIDE 17 -l =

t a ACRS LETTER (CONTINUED) Y 3. MODIFIED TS SURVEILLANCE REQUIREMENTS SHOULD INCLUDE SOLEN 0ID AIR CONTROL VALVES AND CHECK VALVES ON AIR ACCUMULATORS IN PORV CONTROL SYSTEMS, IST REQUIREMENTS IN REGULATORY ANALYSIS HAD INCLUDED VALVES IN PORY CONTROL SYSTEMS, ACRS COMMENT WAS A CLARIFICATION, RES STAFF ACTION: REVISED MODIFIED TS TO COMPLY WITH ACRS

COMMENT, i

t SLIDE 18 P

l t A-5 Generic issue 70 Enclosure A to Gener_ic Letter 89-XX = Attachment A-1 Modified Standird Technical Specifications for Combustion Engineering and Westinghouse Plants s REACTOR COOLANT SYSTEM i i 3f4.4.4 PELIEF VALVES ) LIMITING CONDITION FOR OPERATION The following is to be used when two PORVs are provided: 3.4.4 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2. and 3. ACTION: g g g g g g .g O Ob b}O Yd)Y i With one or both PORVs inoperable, because of a. xcessive seat leikage, within 1 hour either restore the POR (s) to OPERABLE statusorclosetheassociatedblockvalve(s]inHOTS least HOT STANDBY within the next 6 hours ancotherwise, be in at the following 6 hours, within i b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, ~ With boti. PORVs ino etable due to causes other than excessive seat c. leakage, within 1 hour either restore at least one PORY to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours, and in HOT SHUTDOWN within the following 6 hours. d. With one or both block valves inoperable, within I hour restore the block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. within the next hour if both block valves are inoperable, restore 1 otherwise be in at least HOT S'ANDBY within the next 6 ho HOT SHUTOOWN in the following 6 hours, d The provisions of Specification 3.0.4 are not applicable. e. \\

i l l r Presentation To The i Committee to Review Generic Requirements j August 9,1989 I i Generic leeue 94 l " Additional Low-Temperature Overpresse Protection for LWRs" 1 Edward D. Throm l Senior Task Manager i Reactor and Plant Safety issues Branch Division of Safety lasue Resolution, RES Mall Stop NL/S-314 x-23911 t i l Slide 1' to Enclosure 4

1 l LTOP Background Iri1979, Multi-Plant Action B-04 resulted in the imposition of new requirements for procedures and equipment to reduce the potential for cold j overpressure events based on the recommendations from GSI A-26. I j Current staff requirements are in: I SR'P 5.2.2 - Overpressure Protection I BTP RSB 5-2 to SRP 5.2.2 l Tech Spec for Overpressure Protection Basis for LTOP: l GDC 15 of Appendix A to 10CFR Part 50 GDC 31 of Appendix A to 10CFR Part 50 Appendix G to 10CFR Part 50 Anticipated Operational Occurrences Slide 2 l

l Regulatory Alternatives Addressed i 1. Base Case - No Action, t 2. Assure Both OMS

  • trains operable when water solid.

3(a). No HPSI pumps operable if water solid. 3(b). No RCP restart if water solid. 4. ACl+ removal and use RHR SRVs. 5. Safety Grade OMS. 6. No water solid operations, require pressurizer bubble (N )- 2 OMS - Overpressure Mitigation System + ACI-Auto-Closure Interlock on RHR System i I t Slide 3 ^ l l-. s..... . ~. -.. _. - -. - - - -, _. _.,.. _ _ _ - - _. _ - ~. _ - - - -,. ~... _

Base Caso / Operating History Data from 1980 through 1986 OMS Category PORV RHR SRV PORV+N2 OMS Failures 2 1 0 OMS Challenges 23 7 0 l Plant Years 244 56 56 No. of Plants 40 15 (+12) 8 l Actual Max Pres 1100 700 I 750 4 1 The frequency of LTOP events is about the same as previously observed, before 1980,0.1/R-y. The likelihood of exceeding the Appendix G pressure / temperature limit is about 1 in 10 LTOP events. 4 Slide 4 y---- y y v---yg grww-y-s,y-g,--g,,v,- m w o mia, ytw v erw w e g w-g w,- ,,---vm-p-y,-ryg->"e--swy.m a eev--.--m+gpe%s,

  • -___me.,m___,,m._._.

m m. m _--s_,,.,,, y-__m e -_er_m_.,_m____

i l Base Case Risk Analysis i

1. Pressurization sources are varied l

- Safety injection j - Charging / letdown Imbalance I -Thermal expansion of RCS water i

2. Pressurization is rapid when water-solid l

- 100 psiincrease for: l 25 galinjacted 1 Tave increase of 10F

3. OMS operability Most T/S allow one channel out for 7 days.

Events in which OMS failed occurred simultaneously with other channel out for maintenance. 4 Slide 5 ..,a,.,,w.n.n.--w._ m m v~ ~-#' ~**"""~'"'*'~~~~~~#'

i = l Analysis Methodology l 1. Reanalyze each event in data base for each l alternative: l - la event still possible, for alternative? i -If so, assume OMS failure: - what is maximum pressure achievable? l l 2. Estimate change in OMS unavailability for each alternative. l 3. Assumptions: l - Flow rates, heat-up rates and relief capacities are same as in data base. l - Operator intervention terminates press.urization 3 minutes after start, or other relief path available. - Base case OMS unavailability represents operation with single channel. i l Slide 6

l l i Vessel Through-Wall Crack Probability ) Thermal-hydraulic transient for VISA Code: l Based on operating data i Twater = 120 0F Heatup rate = 25 0F/hr Calculate conditional TWC probability for various l peak pressures (2500,1500,700 pala). Integrate over plant life based on RTndt shift to j determine mean value. CDF = P(e) x P(OMS) x P(pres) x P(TWC at pres) - CDF la Core Damage Frequency - P(e) is frequency of an LTOP transients. - P(OMS) is probability of OMS failure. l - P(pres) is probability of potential peak pressure being reached. - P(TWC at press) is vessel TWC probability at potential peak pressure. 4 Slide 7 i .--a .,,,..-.._,y,. _.,.,,,,,..,y__w. y.m,,,,, ,y .,ygg..,. w7.y,7,-_,._7..v,m,.,.p,

l Consequence Evaluation Operating history indicates that LTOP events occur in Mode 5, therefore containment may be bypassed during a transient, or non-Isolatable. Base case consequences evaluated for a late core melt with containment bypass. Results were scaled to plant specific consequences based on Strip's data (NUREG/CR-2723) to account for plant specific features - population, power level and environmental factors. SST1 - containment bypass (PWR-2) SST2 - failure to isolate (PWR-5). Each plant specifically evaluated for both TWC probability and consequences. ~ I Slide 8 i a . - - - ~. - - -, - - -. -, - - -. -,... - -, - - - -, - - -,,,, -.,.. -..., - -, - - - ~ ~ - -

i Base Case Results l CDF Results: For Westinghouse & Combustion Engineering - Mean CDF is 3.24 x 10-6/R-year For Babcock & Wilcox: - Mean CDF is negligible - Also no observed LTOP transient which l challenged toe OMS. 1 Consequences Results (Industry Total W & CE) - Best Estimate (50% SST1/ 50% SST2) 16,000 person-rem - High Estimate (Scaled SST1) 29,000 person-rem - Low Estimate (10% SST1/ 90% SST2) 5,300 person-rem l 1 l Slide 9 ._.___.._.-____.,_.~..,,..__,,.-,_..,1

) Risk Reduction Estimate Bases Alt. Benefit Plants Affected 2 Improve OMS l Availability All l j 3 Reduce initiating i events All t 4 Provide additional i relief path 14/40 PORV plants l 5 Improve OMS Availability 40 PORV plants 6 Reduce OMS challenges All l 4 Slide 10 L. -. - - - - - -. - -

Resource Estimates " Simple" T/S Change (From NUREG/CR-4627, " Generic Cost Estimates") Staff: Review and approve proposed License Amendment, Preliminary Significance and Hazards Analysis. 1.5 weeks technical + 1 week management & legal j 1 page Federal Register Notice ($600) l Final License Amendment, SER, lasue Amendment. I 2.5 weeks technical + 1 week management & legal l 1 column Federal Register Notice ($200) ) 6 Staff-weeks per Plant for T/S Amendment j $14,200 per Plant for T/S Amendment l Industry: 8 weeks for T/S, technical, management & legal. 1 week for two procedure revisions. 9 Staff-weeks per Plant $17,400 per Plant for T/S Amendment $ 1,900 for Two Procedure Changes l $19,300 per Plant Total Cost plus $ 2,000 per PORV Plant (replacement power cost overIlfetime) Total: $33,500 per RHR Plant, $35,500 per PORV Plant Slide 11

BEST ESTIMATE VALUE/ IMPACT StweeARY FOR GI-94 AoorTIONAL Low-TEMPERATURE OVERPRESSURE PROTECTION FOR LWRS ALT T E FREo. Dose Occuparromat ImposTar IEC V/I Rarzo REoucTrom REoUCTIon Exposant Cosi Cost ($ pea (PtER R-TEAR) (Person-rem) (PEkson-rem) ($1,000s) ($1,000s) PERsom-mEm) 2 2.89E-6 14,500 n/A 1,370 950 160 3(a) 1.07E-6 7,000 m/a 3,630 1,840 780 3(s) 0.21E-6 1,400 n/A 1,290 950 1,600 3(A&s) 1.20E-6 8,400 n/A 4,920 2,790 920 4(a) 0.16E-6 700 m/A 770 650 1,900 4(s) 0.16E-6 700 n/a 4,770 650 7,750 5 1.82E-6 8,200 900 16,000 570 2,000 5(a) 3.00E-6 13,400 900 16,000 570 1,200 6(a) i.24E-6 16,000 23,000 41,450 1,450 2,700 6(s) 1.74E-6 9,300 23,000 41,450 1,450 4,600 ALT: 2 - TECamICAL SPEcgrrearton CuansE 67 PtamTs 3(a) - SI LocEour 3(a) - RCP RESTART 67 PtamTs 3(aSe) - SI & RCP 67 Ptamis 4(a) - ACI REnovat, w/o oxscommEcr cosy 40 PORV PtamTs 4(s) - ACI REmovat, w/ orscommEcr cost 40 PORV PtamTs 5 - SarETT GamoE LTOP STsven 49 PORV PLamTs 5(a) - SEmsrTxTTT STuor roR SarETT GRaoE 40 PORV PtamTs 6(a) - PmesssRrZER BUBBLE, <600 Ps1 67 PLamTs 6(s) - PREssORIZEa BUBBLE, 19% To 2500 psr 67 PLamTs SLIDE 12 - - -}}

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