Text

Request for Authorization to Continue Using a Risk-Informed Inservice Inspection Alternative to the ASME Boiler and Pressure Vessel Code Section XI Requirements for Class 1 and 2 Piping
	ML13191A448
Person / Time
Site:	Salem
Issue date:	07/09/2013
From:	Duke P; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LR-N13-0123
	Download: ML13191A448 (46)
	v • d • e

{{#Wiki_filter:PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 JUL 0 9 2013 NuckarlLC 1 0 CFR 50.55a LR-N 1 3-0 1 23 U. S. N uclear Reg ulatory Comm ission ATTN: Document Control Desk Washi ngton, DC 20555-0001 Salem Nuclear Generating Station, Unit 2 Renewed Facility Operating License No. DPR-75 N RC Docket No. 50-3 1 1 S u bject: Request for Authorizatio n to Conti nue Using a Risk-I nformed l nservice I nspection Alternative to the ASME Boiler aod Pressure Vessel Code Section XI Req u i rements for Class 1 and 2 Piping

References:

1. USNRC Letter dated October 1 , 2003, "Salem N uclear Generating Statio n, Unit Nos. 1 and 2 - Risk- I nformed l nservice I nspection Prog ram" (TAC N os. MB7537 and MB7538)

2. USNRC Letter dated April 1 8, 20 1 3, "Salem Nuclear Generati ng Station, Unit No. 2 - Safety Evaluation Of Relief Req uest No. S2-1 4R-123 Regard i ng The Fou rth 1 0-Year lnservice I nspection I nterval (TAC NO. ME8847) (ML13088A219)

I n accordance with 1 0 CFR 50.55a(a)(3), "Codes and standards," PSEG N uclear LLC ( PSEG), hereby req uests N RC approval of proposed Relief Req uest S2-14R-1 31 for Salem Generating Station Unit 2. The p roposed relief will allow Salem Unit 2 to continue to utilize the N RC approved Salem Unit 2 Alternate Risk I nformed l nservice I nspection (RI-ISI) program as an a lternative to the 2004 Edition, ASME Section XI inspection req u i rements for specific Class 1 and Class 2 p i ping welds, in accordance with 1 0 CFR 50.55a(a)(3)( i) by alternatively provid i ng an acceptable level of q ual ity and safety. The RI-ISI p rogram was developed i n accordance with Electric Power Research I nstitute (EPRI) Topical Report (TR) 1 1 2657 Revision 8-A, "Revised Risk I nformed l nservice I n spection Eval uation Proced u re," December 1 999, and was previously approved for use at Salem ( Reference 1 ) on October 1 , 2003. Attachm ent 1 contains the Salem Unit 2 Relief Req uest S2-14R- 1 3 1 , which provides j ustification that the use of the RI-ISI prog ram provides an acceptable level of q ual ity and safety. Attachment 2 contains the i nspection location selection comparison of ASME Section XI Code and EPRI TR- 1 1 2657 by Risk Category. Attachment 3 is a summa ry of the Regu latory Guide 1 .200, Revision 1 , "Approach for Determ in i ng the Tech nical Adeq uacy of Probabilistic Risk Assessment Results for Risk-I nformed Activities," eval uation performed on Revision 4.3 of the

L R- N 1 3-0123 1 0 CFR 50.55a Page 2 PRA model and the impact of the identified gaps on technical adequacy of the Salem PRA model to support the Salem RI-I S I request. Relief is requested for the Fourth Ten-Year lnservice Inspection Interval of the Salem Unit 2 lnservice Inspection Program, currently scheduled to begin on November 27, 201 3, and scheduled to end on November 27, 2023. PSEG is requesting approval by July 9, 20 1 4. There are no regulatory commitments contained in this letter. Should you have any questions concerning this matter, p lease contact M r. Brian Thomas at 856-339-2022. Sincerely, e.IL-{ Manager - Licensing PSEG Nuclear LLC Attachments: 1 . Relief Request S2-14R- 1 31

2. Inspection Location Selection Comparison ASME Section X I Code and EPRI TR-1 12657 by Risk Category

3. Salem PRA Summary cc: W. Dean, Administrator, Region I, N RC N RC Senior Resident Inspector, Salem J. Hughey, Project Manager, Salem, USN RC P. Mul l igan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator T. Cachaza, Salem Commitment Tracking Coordinator LR- N 1 3-0 1 23 Salem Nuclear Generating Station, Unit 2 Renewed Facility Operating License No. D P R-75 N RC Docket N o. 50-3 1 1 Request Number - S2-14R- 1 3 1 Proposed Alternative I n Accordance with 1 0 CFR 50. 55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected System: Various ASME Code Class 1 and 2 Systems Code Class: ASME Code Cl ass 1 and 2 Component

Description:

ASME Code Class 1 and 2 Piping Welds Components Affected: Weld Weld Description Code Item Number Numbers Category ASME Code Class 1 Piping 8-F Various 85.40, 85.70 Welds AS M E Code Class 1 Piping 89 . 1 1 ' 89.2 1 ' 89.3 1 ' Various 8-J Welds 89. 32, 89.40 ASME Code Class 2 Piping C5. 1 1 , C5.2 1 , C5.30, Various C-F- 1 Welds C5 .41 AS M E Code Class 2 Piping Various C-F-2 C5.5 1 , C5.6 1 , C5.81 Welds

2. Applicable Code Edition and Addenda As documented in the NRC approval of relief request S2- 1 4 R- 1 23 dated April 1 8, 201 3 (Reference 8.7), the code edition for the fourth interval for Salem Unit 2 wil l be American Society of Mechanical Engineers (ASM E) Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice I nspection of Nuclear Power Plant Com ponents," 2004 Edition.

3. Applicable Code Requirement The following Code requirements are para p h rased from the 2004 Edition of ASME Section XI:

AS M E Section X I 2004 Edition IWB-24 1 2, I nspection Program B, requires examinations in each examination category be com pleted during each inspection interval. ASME Section X I 2004 Edition IWB-2500 Examination and Pressure Test Requirements (a) Page 1

SWIIQ!I!l't *il!!!l*tt li r* I!RIIUMitlk!lilillrfH:!JEiioo11'1lildii'i Atl iP iii/UI'AifWII\1i;IT n.-ar...-.-........ -...-- Attachment 1 LR-N 1 3-0 1 2 3 Salem Unit 2 l nservice I nspection Program Relief Request S2-14R- 1 3 1 1 0 CFR 50. 55a Com ponents shall be examined and tested as specified in Table IWB-2500- 1 . The method of examination for the com ponents and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWB-2500- 1 except where alternate examination methods are used that meet the requirements of IWA-2240. Applicable category welds in table IWB-2500-1 are B-F (Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles) and B-J (Pressure Retaining Welds in Piping). 1 00% of Category 8- F welds and 25% of Category B-J weld s for the ASM E Code, Class 1, non exe m pt piping shall be selected for volumetric and/or surface examination based on existing stress analyses and cumulative usage factors. AS M E Section XI 2004 Edition IWC-24 1 2, I nspection Program B, requires examinations in each examination category be com pleted during each inspection interval in accordance with Table IWC-24 1 2- 1 . Applicable category welds in table IWC-2500- 1 are C-F- 1 (Pressure Retaining Welds in Austenitic Stainless Steel or Hig h Alloy Piping) and C-F-2 (Pressure Retaining Welds in Carbon or Low Allo y Steel Piping). For Category C-F- 1 welds in Class 2 piping, the welds selected for examination shall includ e 7 . 5%, but not less than 28 welds, of a l l dissimilar metal, austenitic stainless steel or hig h alloy welds not exempted by IWC- 1 220. (Some welds not exempted by IWC-1 220 a re not required to be nondestructivel y examined per Examination Category C-F- 1 . These welds, h owever, shall be included i n the total weld count to which the 7.5% sam pling rate is applied.) The examinations shall be distributed as fol lows: (a) the examinations shall be distributed among the Class 2 systems prorated, to the deg ree practicable, on the number of nonexem pt dissimilar metal, austenitic stainless steel, or hig h a l loy welds in each system (i.e., if a system contains 30% of the nonexempt welds, then 30% of the n ondestructive examinations required by Examination Category C-F- 1 should be performed on that system); (b) within a system, the examinations shall be distributed among terminal ends, dissimilar metal welds, and structural discontinuities p rorated, to the deg ree practicable, on the number of nonexempt terminal ends, dissimilar metal welds, and structural discontinuities in that system; and (c) within each system, examinations shall be distributed between line sizes prorated to the degree practicable. For Category C-F-2 welds in Class 2 piping the welds selected for examination shall include 7 . 5%, but not less than 28 welds, of all carbon and low alloy steel welds not exempted by IWC- 1 220. (Some welds not exem pted by IWC-1 220 are not required to be nondestructivel y examined per Examination Category C-F-2 . These welds, however, shall be included in the total weld count to which the 7.5% sam pling rate is applied). The examinations shall be distributed as follows: (a) the examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the num ber of n onexempt carbon and low alloy steel welds in each system (i.e. , if a system contains 30% of the nonexe m pt welds, th en 30% of the Page 2 LR-N 1 3-0123 Salem Unit 2 l nservice I nspection Program Relief Request S2-14R-13 1 1 0 CFR 50. 55a nondestructive examinations required by Examination Category C-F-2 should be performed on that system); ( b) within a system, the examinations shall be distributed among terminal ends and structural discontinuities prorated, to the deg ree practicable, o n the number of nonexempt terminal ends and structural discontinuities in that system; and (c) within each system, examinations shall be distributed between line sizes prorated to the degree practicable.

4. Reason for Request I n accordance with the provisions of 1 0 CFR 50.55a, "Codes and Standards," paragraph 1 0 CFR 50.55a(a)(3)(i), PSEG Nuclear requests relief from the requirement of ASME Code Section X I , Sub-article IW8-2500 and IWC-2500, Tables IW8-2500-1 and IWC-2500- 1 , Examination Categories 8-F, 8-J, C-F-1 and C-F-2, "Pressure Retaining Welds in Piping" welds.

AS M E Section XI Examination Categories 8-F, 8-J, C- F- 1 , and C-F-2 currently contain the requirements for exa mination of piping components by means of nondestructive examination ( N D E). The previously approved Risk-I nformed lnservice Inspection (RI I S I) program ( Reference 8 . 1 ) wil l be substituted for Class 1 and Class 2 piping ( Examination Categories 8-F, 8-J, C-F- 1 , C-F-2) in accordance with 1 0 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of:quality and safety. Other non-re lated portions of the ASM E Section XI Code wil l be unaffected . I n general, a risk-informed program replaces the number and locations of nondestructive examination (N DE) inspections based on ASME Code, Section XI requirements with the number and locations of these inspections based on the risk-informed guidelines. These processes result in a program consistent with the concept that, by focusing inspections on the most safety-significant welds, the number of inspections can be reduced while at the same time maintaining protection of public health and safety.

5. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50. 55a(a)(3)(i), NRC approval of the Salem Unit 2 Alternate RI-ISI program as an alternative to the 2004 Edition, ASM E Section X I inspection requirements for Class 1, Examination Category 8-F and 8-J, and Class 2, Examination Category C F- 1 and C-F-2 piping welds is requested.

The Salem Unit 2 RI-ISI Program has been developed in accordance with the EPRI methodology contained in E PRI TR- 1 12657, "Risk- I nformed l nservice I nspection Evaluation Procedure" (Reference 8.2). The RI-ISI methodology was approved for use at Salem 2 for the Third Ten-year l nservice I nspection I nterval and is applicable for the Fourth Ten-year l nservice Inspection I nterval. The Salem Unit 2 specific RI-ISI program is sum marized in Table 1 (Attachment 2). The RI-ISI program has been updated Page 3 L R- N 1 3-0 1 23 Salem Unit 2 l nservice I nspection Program Relief Request S2-14 R- 1 3 1 1 0 CFR 50. 55a consistent with the intent of N E I-04-05 ( Reference 8.3) and continues to meet E PRI TR-1 1 2657 and Regulatory Guide 1 . 1 74 risk acceptance criteria. PSEG wil l continue to im plement the Rl-ISI Program in accordance with AS M E Code Case N-578- 1 , "Risk-I nformed Requirements for Class 1 , 2, and 3 Piping, Method B, Section XI, Division 1 ." The ultrasonic examination volume to be used based on degradation mechanism and com ponent configuration will be the examination figures specified in Section 4 of EPRI TR- 1 1 2657 . The ultrasonic examination procedures, equipment, and personnel used to detect and size flaws in piping welds will be qualified by p e rformance demonstration in accordance with ASM E Section X I Appendix V I I I, "Performance Demonstration for Ultrasonic Examination Systems." The volumetric sca nning wil l be in both the axial and circumferential directions to detect flaws in these o rientations. As part of the Rl-181 living p rogram update, the delta risk assessment was re-evaluated and was determined to continue to meet the delta risk acceptance criteria of EPRI TR-1 1 2657. This update is based on the most recent Salem PRA, which has been peer reviewed to Regulatory Guide 1 .200, Rev 1 and updated according ly. The PRA has been determined to be adequate for this application as described in Attach ment 3. The majority of the Rl-ISI inspection locations for the fourth interval are identical to those in the third interval a nd the class 1 percentage of welds to be examined for the fourth interval has remained effectively constant compared with third interval at approximately 8.5%. The RI-ISI being a living program, changes in the inspection locations are due to:

Plant modifications im pacting the total number of RI-ISI-piping welds
RI-ISI piping welds required to be included in other augmented inspection programs (i.e. Code Case N-770- 1 )
PRA updates im pacting the consequence category
Degradation mechanism assig nment changes (due to im proved or updated information) impacting the failure potential category
Improved or updated information regarding radiation exposure, access, inspectability or other factors Specific changes are documented in Attachment 2.

Pursuant to 1 OCFR50 .55a(a)(3)(i), relief is requested on the basis that the proposed alternative to continue using a RI-ISI Program would provide an acceptable level of qua lity and safety.

6. Duration of Proposed Alternative Relief is requested for the Salem Unit 2 Fourth Ten-Year Inspection I nterva l of the l nservice Inspection Program, currently scheduled to begin on November 27, 20 1 3, and scheduled to end on November 27, 2023.

Page 4 LR-N 1 3-0 1 23 Salem Unit 2 l nservice I nspection Program Relief Request S2-14R- 1 3 1 1 0 C F R 50. 55a

7. Precedent The N RC previously a p p roved the Salem Unit 2 Alternate Risk-I nformed lnservice Inspection Program in Reference 8. 1 .

Salem considers both the plant and ind ustry operating experience and updates the RI ISI program d u ring the re-evaluation p rocess fol lowing each inspection period per our com mitment in section 4 of our original relief request (Reference 8 . 5)

8. Reference

8. 1 USNRC Letter d ated October 1, 2003, "Salem Nuclear Generating Station, Unit Nos. 1 and 2- Risk- I nformed lnservice Inspection Program" (TAC Nos. MB7537 and MB7538) ( M L032390034) 8.2 EPRI TR- 1 1 2657, Electric Power Research I nstitute Report for Alternative Req uirements of Risk-I nformed lnservice Inspection Evaluation Proced u re, EPRI, Polo Alto, CA: 1 999, Rev B-A.

8.3 N E I-04-05, "Living Program Guidance to Maintain Risk-I nformed l nservice Inspection Programs for Nuclear Plant Piping Systems", dated April 2004. 8.4 Request for Additional I nformation Related to Byron Station, Units 1 and 2, Req uest for relief 1 3R-02, TAC Nos. M D3855 and M D3856, dated Aug ust 8, 2007. ( ML072140023) 8.5 PSEG Letter dated January 2 1 , 2003, "Request for Authorization to Use a Risk I nformed lnservice I nspection Alternative to the AS ME Boiler and Pressu re Vessel Code Section XI Requirements for Class 1 and 2 Piping," Salem Generating Station Unit Nos. 1 and 2, Docket Nos. 50-272 and 50-3 1 1 . (M L030300 1 1 6) 8.6 PSEG Letter dated J u ly 1 , 2003, "Response to NRC Req uest for Additional I nformation Regarding Risk- I nform ed l nservice I nspection Submittal," Salem Generating Station Unit Nos. 1 and 2, Docket Nos. 50-272 and 50-3 1 1 . (ML031950120)

8. 7 US N RC Letter dated April 1 8, 201 3, "Salem N uclear Generating Station, Unit No.

2 - Safety Evaluation Of Relief Request No. S2- 1 4 R- 1 23 Regarding The Fourth 1 0-Year lnservice Inspection I nterval (TAC NO. M E8847)," ( M L 1 3088A2 1 9) Page 5

Attachment 2 L R-N Category Rank Rank BF 5 Medium BF 6 ee 4 Medium es 2 High es 5 Medium eve 2 High eve 4 Medium eve 5 eve 6 eve 7 Low MS 6 Re 2 High Re 2 High Re 2 High Re 2 High Re 2 High Re 4 Medium Re 6 Re 6 Low RHR 2 High RHR 5 RHR 6 SJ 2 High SJ 5 Medium SJ 5 Medium SJ 7 Low sw 4 Medium 1 3-0 1 23 Salem Nuclear Generating Station, Unit 2 Renewed Facility Operating License No. DPR-75 N RC Docket Nos. 50-3 1 1 Table 1 : I nspection Location Selection Com parison ASM E Section X I Code and EPRI TR- 1 1 2657 by Risk Category Ri sk Consequence Failure Potential Code 15r Approved RI-ISI* Interval New RI-ISIInterval System* OMs* Rank Category Weld Count RI-ISI Other Weld Count RI-ISI Other AF 5 Medium Medium n Medium C-F-2 24 4 24 3 a. Medium TASes, TI Medium e-F-2 20 3 31 4 b. Low Medium None Low e-F-2 48 0 51 0 c. High None Low e-F-2 0 0 17 2 d. High Eesee Medium e-F-1 21 6 21 6 es 4 Medium High None Low e-F-1 80 8 75 8 e. Medium IGSee, Eesee Medium e-F-1 8 1 8 1 es 6 Low Medium None Low e-F-1 40 0 40 0 eve 2 High High TASeS, TI Medium B-J 5 2 5 4 f. High n Medium B-J 2 0 4 1 g. High None Low B-J, e-F-1 98 10 99 10 h. Medium Medium n Medium B-J 27 3 27 3 eve 5 Medium Medium Eesee Medium e-F-1 12 1 0 0 I. Low Medium None Low B-J, e-F-1 476 0 178 0 j. Low None Low B-J 0 0 42 0 k. Low Medium None Low e-F-2 249 0 237 0 I. High TASeS, TI, PWSee Medium B-F 1 1 0 0 m. High TASeS, TT Medium B-J 14 3 19 5 n. High TI, PWSee Medium B-F 1 1 0 0 o. High n Medium B-J 9 2 3 1 p. High PWSee Medium B-F 12 5 0 0 q. High None Low B-F, B-J 221 29 227 43 r. Low Medium None Low B-J 9 0 0 0 s. Low IGSee Medium e-F-1 6 0 6 0 Re 6 Low Low Eesee Medium e-F-1 1 0 1 0 Re 7 Low Low None Low e-F-1 74 .0 85 0 t. High TASeS Medium B-J, e-F-1 18 5 18 5 RHR ' 2 High High Eesee Medium e-F-1 3 1 3 1 RHR 4 Medium High None Low B-J, e-F-1 222 23 228 23 u. Medium Medium IGSee Medium B-J 8 1' 8 1 RHR 5 Medium Medium Eesee Medium e-F-1 2 0 2 1 v. Low Medium None Low B-J, e-F-1 130 0 130 0 RHR 7 Low Low None Low e-F-1 20 0 20 0 SJ 2 High High TASeS, TI Medium B-J 12 3 12 3 SJ 2 High High n Medium B-J 13 3 19 5 w. High Eesee Medium B-J 8 4 8 4 SJ 4 Medium High No ne Low B-J, e-F-1 272 38 271 28 x. Medium TI, IGSee Medium B-J 14 1 2 1 y. Medium IGSee Medium B-J 23 3 23 3 SJ 6 Low Medium None Low B-J, e-F-1 622 0 579 0 Z, SJ 6 Low Low TI, IGSee Medium B-J 0 0 12 0 a a. Low None Low B-J 115 0 165 0 bb. High None Low e-F-1 64 7 64 7 3004 168 2764 173

            *Acronyms defined:

AF- Auxiliary Feedwater System BF- Steam Generator Feedwater system CS - Containment Spray System CC- Component Cooling System CVC- Chemical and Volume Control System OMs - Damage Mechanisms ECSCC - External Chloride Stress Corrosion Cracking 1 LR- N 1 3-0 1 23 IGSCC- lntergranular Stress Corrosion Cracking MS - Main Steam System PWSCC- Primary Water Stress Corrosion Cracking RC- Reactor Coolant System RHR- Residual Heat Removal System RI-ISI- Risk Informed lnservice Inspection SJ - Safety Injection System SW- Service Water System TASCS- Thermal Stratification TT- Thermal Transients Notes (all table references below refer to th e tables in EPRI TR-112657)

a. Reduced selections by one; meets requirements

b. 11 welds added as a result of Design Changes (i.e. Steam Generator replacement, piping replacement due to Flow Accelerated Corrosion)

c. 3 welds added as a result of Design Changes.

d. CC system scope is newly added due t o change in weld exemption.

e. 5 CS welds moved to RHR.

f. 2 extra selections to maintain Class 1 percentage of welds to be examined consistent with third interval at approximately 8.5%

g. 2 branch connection welds moved from RC to eve.

h. 1 branch connection weld moved from RC to eve.

L All 12 small-bore socket welds removed from CVC scope, exemption IWA-1220(a)(1) applies, line is sealing water to Reactor Coolant Pum ps, previously assigned to High Pressure Safety Injection.

j. 2 welds added as a result of Design Change and then 42 welds moved to RC 7 as a result of consequence change. 258 small-bore socket welds removed from CVC scope, exem ption IWA-1220(a)(1) applies, line is sealing water to Reactor Coolant Pum ps, previously assigned to High Pressure Safety Injection.

k. 42 welds moved from RC 6 as a result of consequence change I. 20 welds removed (welds are integral part of flanged relief valve) and 8-welds added as a result of Design Change.

m. 1 weld moved to RC 2 with DM- TT per Tables 5-2 and 5-5 ,

n. 1 weld added per Table 5-1 (no Change No. identified). 4 welds added as a result of Design Change and Table 5-1

o. 1 weld moved to RC 2 with DM- TT per Table 5-2 of EPRI TR-112657

p. 1 weld moved from RC 2 with DM- TASCS, TT, PWSCC per Tables 5-2 and 5-5 of EPRI TR-112657. 1 weld moved from RC 2 with DM- TT, PWSCC per Table 5-2 of EPRI TR-112657.

Six welds moved to SJ RC 2 and 2 moved to eve RC 2

q. 6 welds moved to RC 4 per Table 5-2 of EPRI TR-112657; 2 welds m oved to RC 4 per Table 5-5; 4 welds deleted per Table 5-5

r. 9 welds added per Design Changes; 8 welds added from RC 2 with DM of PWSCC; 4 welds deleted per Table 5-5; 1 Weld moved to CVC; 4 welds moved to SJ; 2 welds moved to RHR; 2 welds added due to category change from B-M-1 valve body welds (reducer to valve) to R A piping welds; 2 welds moved to RC 6 as a result of consequence change

s. 2 welds added from RC 4, then 11 welds moved to RC 7 as a result of consequence change

t. 11 welds moved from RC 6 as a result of consequence change

u. 5 CS welds moved to RHR, 2 RCS RC 4 welds m oved to RHR RC 4 and 1 RHR RC 4 moved to SJ RC 4

v. 1 weld selection added during 3rd 10-year Interval living program update to meet requirement

w. 6 welds moved from RCS RC 2 to SJ RC 2

x. 6 welds moved to RC 6 as a result of consequence change. 4 welds moved from RCS RC 4 to SJ RC 4 and 1 weld moved from RHR

y. 12 welds moved to RC 6 as a result of consequence change 2

LR-N13-0 1 23

z. 6 welds moved from RC 4 as a result of consequence change. 51 welds moved from RC 7 as a result of consequence change and 101 welds moved to RC 7 as a result of consequence change. A new weld was added to scope.

aa. 12 welds moved from RC 5 as a result of consequence change bb. 51 welds moved to RC 6 as a result of category change. 101 welds moved from RC 6 as a result of consequence change 3 L R-N 1 3-0 1 23 Salem Nuclear Generating Station, Unit 2 Renewed Facility Operating License No. DPR-75 N RC Docket No. 50-311 Salem PRA Summary The Salem PRA has been updated several times to maintain current with the plant desig n and operation and to support peer review. Revision 3 to the m odel was released as a d raft in November 2001 in p re paration for the Westing house Owners Group (WOG) peer review. Documentation for Revision 3 was fina lized in June 2002. This is the version of the model that was used for the original RI-I S I submittal. The PWR Owners Group conducted a peer review in accordance with Regulatory Guide 1.200, Rev 1, "Approach for Determining the Technical Adequacy of Proba bilistic Risk Assessment Results for Risk- Informed Activities," in November of 2008. A final report of that peer review was issued in March of 20 1 0. PSEG Nuclear has made changes to the model post peer review and maintains the attached Table A 1 of "Identified Gaps to Capability Category I I of the ASME Standard" which discusses sig nificance of the gap and is assessed for each application. The latest PRA model used for this evaluation is Revision 4 . 3, which is adequate to support this application based on a review of the gaps and their significance . The original RI-ISI evaluation concluded external events are not likely t o im pact the consequence ranking. This position is further supported by Section 2 of EPRI Report 1 02 1 46 7, "Nondestructive Evaluation: P robabilistic Risk Assessment Technical Adequacy Guidance for Risk- I nformed lnservice Inspection Programs" which concludes that quantification of these events will not change the conclusions derived from the RI I S I process. As a result, there is no need to further consider these events. 1 L R- N 1 3-0 1 23 Table A1: Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Finding Finding Description Supporting Resolution Requirements IE-A1 -01 A loss of an AC bus may not result in a reactor trip, but may result IE-A1 Minimal impact, referenced event is in a forced shutdown due to technical specifications. If the lost bounded by the modeled reactor trip bus happens to be the operating bus for equipment, systems will initiator. I nitiator impact is a qualitative be challenged. Loss of an AC bus is generally modeled in most consideration. PRAs. This F&O is characterized as a finding based on the lack of sufficient documentation to allow verification of SR. Include events for loss of 4Kv bus if they require a forced shutdown consistent with most industry PRAs. IE-A3-01 Historical events appear to lead to somewhat more complex IE-A3 Response: Table 3-2 indicates this was situations than the assigned grouping would indicate. The plant- binned as a trip with loss of feedwater, specific history indicates that on 1 2/31/01 an event occurred consistent with the classification scheme resulting in Sl. The categorization of initiating events does not employed for Salem (Spurious Sl Tp). = account for this or the case of ESFAS actuation . Negligible impact on technical adequacy This F&O i s characterized a s a finding based o n the lack of of the PRA model. sufficient documentation to allow verification of S R. L__. __ - fonsider : ategorizing this event aan SF a.ctuation (QR9). re-c 2 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements I E-A3a-01 Although sited as an input source, there appears to be no I E-A3a Comparisons were made to industry data documentatio n supporting a comparison of initiating events with and to other plants. regard to plants of similar design. The documentation indicates that "past probabilistic risk assessments" were used for source This is judged to be a documentation and experience. However, there is not documentation of such a consideration only and does not affect the comparison. It also does not identify that any examination was technical adequacy of the PRA model. made for Salem-like designs. A com parison to similar designs can potentially identify those design-specific events than may have Ul}ique consequences which may not be defined in more generic sources. It also provides an industry basis for selection. This F&O is characterized as a finding based on the lack of sufficient documentation to allow verification of SR. Utilize available industry summary documentation to define generally appropriate initiating event l ist for specific design. IE-A4-01 The requirement is to address each system, including support IE-A4 systems to assess potential for initiating events. The analysis only Loss of charging not i ncluded based on addresses support systems and does not address the impact of screening criterion, m ust cause automatic other operating systems with regard to events resulting in a plant or manual trip AND frontline systems are upset and subsequent trip signal. For charging this has the significantly affected. No significant potential to im pact both the initiator and response models such impact on frontline systems expected. that consequential failures could be possible. This F&O is characterizing as a finding based on the lack of This is judged to be a documentation sufficient documentation to allow verification of S R. consideration only and does not affect the Add evaluations for frontline operating systems that in particular technical adequacy of the PRA model. are part of the PRA response model. - - - - - - - - - - 3 LR- N 1 3-0123 Applicable Finding Finding Description Supporting Resolution Requirements I E-AS-01 SA PRA Initiating Events Notebook, SA-PRA-001 , Revision 0, IE-AS Other than at-power events were Section 2 . 1 .2 describes the review of Salem Generating Station evaluated. Experience and Trip Review. No mention is made of consideration of events that occurred at conditions other than at- This is judged to be a documentation power operation. Also, events resulting in controlled shutdown consideration only and does not affect the were excluded on the basis that they present only m ild challenges technical adequacy of the PRA model. rather than being determined to be not applicable to at-power operation. Failure to consider non-power events and controlled shutdown events could result in exclusion of valid initiating events. Provide an expl icit discussion of the review of non-power events. Improve the justification for exclusion of controlled shutdown events to add ress applicability to at-power operation or to provide a quantitative justification for exclusion. IE-A6-01 SA PRA Initiating Events Notebook, SA-PRA-001 , Revision 0, IE-A6 This is judged to be a documentation Section 2 .1 .2 does not indicate that plant operations, consideration only and does not affect the maintenance, engineering, and safety analysis personnel were technical adequacy of the PRA model. interviewed or included in the review process for the initiating events notebook to determine if potential initiating events have been overlooked. Documentation was not available to show that the Category I I/I I I requirement was satisfied. The initiating event analysis should document a reasonably complete identification of initiating events. Docum ent the required interviews. - - - - - -- - --- --- - - 4

Attachm ent 3 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements IE-A?-01 SA PRA Initiating Events Notebook, SA-PRA-001 , Revision 0, IE-A? Section 2 . 1 .2 does not indicate that a review of plant-specific or Plant and industry operating experience industry operating experience was performed for the purpose of was reviewed. identifying initiating event precursors. Failure to consider precursor events and controlled shutdown This is judged to be a documentation events could result in exclusion of valid initiating events. consideration only and does not affect the The model owner stated that precursors were considered during technical adequacy of the PRA model. the review of plant operating experience. However, because this is not documented, the SR cannot be considered met. This should be explicitly stated in the Initiating Events Notebook. I E-83-01 Initiating events are not grouped with less severe events without IE-83, AS-A5 assuming the worst potential effects. For example, the potential Initiating events should be grouped for a spurious Sl actuation is grouped in the general transient reasonably, as PRA should be realistic category with events such as reactor trip and considered to be no and not conservative. Spurious Sl will worse than the reactor trip . However, unmitigated spurious S l generally be recovered and the event will events can challenge a PORV resulting in a consequential LOCA. be a transient. If S l is not reset prior to Spurious Sl events should not be grouped with general reactor PORV operation, what results is a trips. Also, the loss of AC power bus (F) is said to result in a transient with improved reliability of feed-degraded loss of condensate/feedwater performance. However, it and-bleed cooling (already initiated). S l is placed in the PCS available category. This presents a problem can still b e reset and PORV closed . If when developing the conditional failure PCS in response to the difficulty is experienced in closing PORV, event. block valve can be closed. Regard ing the This F&O is characterized as a finding based on the lack of loss of AC bus, this does not result in sufficient documentation to allow verification of SR. even a trip so it would be quite conservative to bin such events as trips Separate out events on basis of unique impacts to the response sequence. with loss of PCS. I Minimal impact on the ability to assess significance of proposed application. i 5 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements I E-C1 b-01 The loss of SW initiating event fault tree S1 R4.Caf (gate IE-TSW) IE-C1 b, IE-C6 This is judged to be a documentation was reviewed and the logic appears to capture the appropriate consideration only and does not affect the combinations of equipment failures that contribute to the initiator. technical adequacy of the PRA model. However, the documentation of the development of the initiator fault trees appears to be lacking. Section 3.3 of the Salem SA-PRA-001 , Revision 0 notebook does not provide much detail of how the i nitiators modeled 'as-fault' trees are developed. It refers to the system m odel notebook. For the loss of SW initiator, the SW model notebook SA-PRA-005. 1 3, Revision 0 was reviewed and there was no discussion of the development of the loss of SW initiator fault tree. For the loss of CC initiator fault tree, Section 4.2 of notebook SA-PRA-005. 1 2, Revision 0 provides a good description of how that initiator fault tree is developed. This F&O is characterized as a finding based on the lack of sufficient documentation to allow verification of S R I E-C1 b. Document how the loss of SW initiating event fault tree is developed. Likewise for other system initiators, as needed. Include a discussion of the recoveries credited in the initiator. This should also be done for the other initiators that are fault trees. IE-C3-01 The initiators that are fault trees, such as loss of SW and loss of IE-C3 This is a minor conservative modeling CC, the initiator frequency is not based on reactor year. For issue and would not affect the ability to example, under gate I E-TSW, basic event SWS-PIP-RP-TBH D R assess the impact of an application . has a m ission time of 8760 hours. This F&O is characterized as a finding because it does not m eet the SR. Use reactor year when quantifying the initiator frequencies. --*- - - - - 6 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements AS-A?-01 Accident Sequences and Event Tree Development Notebook, SA- AS-A? Sequences TTS04 and TIS05 differ only PRA-002, Revision 0 delineates the possible accident sequences in whether containment is isolated, which for each modeled initiating event. However, some sequences are is of concern only in level 2, not for CDF. not explicitly modeled in the single-top fault tree (e.g., TT Level 2 analysis does address sequences S04 and S05 are combined into a single fault tree containment status. gate). No documentation was fou nd to describe the basis of these combinations. Not excluded based on very low Subsuming non-minimal sequences in the single-top fault tree frequency. model could result in loss of risk insights or masking of importance in non-standard configurations. This is judged to be a documentation Provide a description of the process used to combine non-minimal consideration only and does not affect the sequences with their bounding equ ivalent sequence in both the technical adequacy of the PRA model. Accident Sequences and Event Tree Notebook or in the Quantification Notebook. Discuss how it is ensured that risk insights are not impacted by the subsuming of sequences. Provide a more complete basis for not modeling sequences judged to have "very low frequencies" such that a reviewer can evaluate the basis for the exclusion. AS-A?-02 The VS ISLOCA sequence with no piping failure is assumed to be AS-A7 Inventory loss from the postulated term inated with operator isolation of the suction path using the ISLOCA would not be expected to flood pump suction isolation MOVs . However, isolation cannot be more than the lower levels of the auxiliary accomplished until primary pressure is reduced. The potential for building. RH-4 valves which could be flooding of adjacent areas by water lost through the RHR pump used for isolation are located a floor seals and/or RHR heat exchangers prior to isolation does not above the postulated break. Flooding appear to have been evaluated. analysis add resses plant response and demonstrates that the plant can be safely Flooding of adjacent areas could im pact additional equipment shut down without the potentially affected affecting the ability to achieve a safe, stable condition. components, Evaluate the potential volume of water which can be released prior to isolation of the VS sequence with no piping failure to This is judged to be a documentation determ ine if add itional mitigation equipment could be affected. consideration only and does not affect the technical adequacy of the PRA model. 7 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements AS-A8-01 Accident Sequences and Event Tree Development Notebook, SA- AS-A8, AS-86 Mission times vice recovery of offsite PRA-002, Revision 0 and the associated CAFTA event trees power are addressed in current PRA define the end state of each sequence as success or core model . Documentation issue. damage. However, the SBO sequences S08, S 1 1 , S1 4, and S1 7 are assumed to be successful based on offsite power recovery. Operator action to restore mitigating systems after power recovery is not addressed . In addition, given the fact that power recovery is only credible out to 4 hours, 20 hours of m itigating system operation and the potential failures of that equipment over a significant portion of the 24 hour mission time is not being addressed. This failure to address recovery of mitigating systems following power recovery does not ensure a safe, stable end state has been reached for some SBO sequences. There is also concern that the application of offsite power recovery is included twice in the modeling of the SBO event. Recovery is credited in the application of a diesel mission time of 6 hours and again through the appl ication of offsite power recovery top event RBU. Recovery of offsite power does not guarantee restoration of mitigating systems needed to establish a safe stable condition in the plant. I n some plant models, operator action to restore required mitigating systems following power recovery has been shown to be significant. In addition, mitigating system operation over a significant portion of the 24 mission time is not being addressed. Extend the event tree models to address restoration and operation of required safety functions following offsite power covery. Potential events to incl ude are decay heat removal and primary inventory makeup. - L___ 8 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements AS-A1 0-01 Systems and operator actions required to m eet each key safety Varying seal leak rates are now explicitly function are discussed in general terms in the Accident addressed. Sequences and Event Tree Development Notebook, SA-PRA-002, Revision 0 Sections 3 through 9. Operator actions and diverse systems to satisfy top events are included in the fault tree but are grouped under common top events in the accident sequence model (e.g., core decay heat removal includes AFS, operator action to depressurize, and condensate under a common top event). However, the m odeling of offsite power recovery in the SBO event tree does not explicitly model the differences in recovery times or plant response associated with different RCP seal leakage rates. Instead, a single lumped recovery event is modeled . The lumping of RCP seal leakage rate with offsite power recovery under the RBU top event does not provide sufficient detail to determine d ifferences in requirements for mitigation systems and operator responses. For example, RCP seal l eakage of 21 gpm per pump may proceed l ike a general transient and only require secondary side cooling whereas larger seal leakage rates may also require primary makeup for success. Provide explicit event tree branches for each RCP seal leakage rate the event timing and mitigation requirements for different leakage rates can be shown to be the same. This will ensure that significant d ifferences in mitigation requirements and event timing are captured . . .- * -

9 L R- N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements AS-C2-01 Documentation does not clearly address the procedural guidance, Operator actions and related procedural operator actions and interfaces of the plant event trees with plant guidance are discussed in detail in HRA damage states. notebook. Plant event trees, success I The current documentation does not include sufficient detail to failure paths are discussed in accident allow correlation of operator actions required to m itigate the sequence and success criteria notebooks.

accident sequences to the H RA or the interface between the Level 2 notebook describes interfaces. accident sequences and plant damage states carried forward to This information has been provided in the the Level 2 analysis. accident sequence, success criteria and Expand the event trees to include important operator actions as L2 notebooks. separate top events or provide a table which describes operator actions included u nder each existing top event. Provide a - description of the procedural guidance used in m itigation of each accident sequence or group of accident sequences. Document the interfaces between the event tree end points and plant damage states in the Accident Sequence notebook or through a specific reference to the appropriate section of the Level 2 notebook. SC-A 1 -01 The ASM E standard defines core damage as "uncovery and Documentation clarified regarding core heatup of the reactor core to the point at which prolonged damage. oxidation and severe fuel damage involving a large section of the core is anticipated." In the Salem PRA Success Criteria Notebook, SA-PRA-003, a "big picture" definition as described in the ASME PRA standard appears to m issing. In the Salem PRA, core damage is defined as maintaining core temperature below 1 200 degrees F which deals with heatup but not uncovery. The big picture definition of core damage is incomplete in that it defines core heatup but not uncovery. Include core uncovery in the definition of core damage. 10

Attachm ent 3 LR- N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements SC-A2-01 lin the Salem PRA core cooling was defined as successful if core Documentation clarified regarding core exit temperatures do not exceed 1 200 degrees F. This represents damage the temperature below which no core damage is expected to occur and the core exit thermocouple temperature at which the operators transfer to severe accident guidelines. The 1 200 degrees F core temperature success criteria were interpreted to be the core hottest node tem perature (TCRHOT) in MAAP. However, in the T/H notebook a peak cladding tem perature of 1 800 degrees F was referenced. The MAAP code used 1 800 degrees as TCRHOT. Also, there is no mention of core collapsed l iquid level. The temperature defined for core damage in the success criteria notebook was not the temperature used for TCRHOT in the MAAP code. Reconcile the definition of core damage between the T/H calculations and the success criteria notebook. SC-84-01 IMAAP Thermal-Hydraulic Calculations Notebook (SA-PRA-007, Revision 1 ) Sections 1 .2 and 1 .3 provide a discussion of the Documentation u pdated.

              **codes available and the advantages associated with using MAAP, respectively. However, MAAP is used in establishing large LOCA success criteria, although the code is not su itable for analysis of this plant upset. A d iscussion of code limitations needs to be documented.

Use of a non-applicable code could result in incorrect success criteria. -* - Base the success criteria for large LOCA on an appropriate T/H code. Provide a general discussion of known T/H code lim itations. 11 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements SC-85-01 A check of the reasonableness and acceptabil ity of the success SC-85 This is judged to be a documentation criteria results is not documented. consideration only and does not affect the Comparing success criteria results with those of similar plants or technical adequacy of the PRA model. performed using other plant-specific codes provides greater assurance that the results are correct. Document a check of the reasonableness and acceptability of the success criteria results. Supporting requ irement SC-85 provides example methods. Note that the PWROG PSA database identifies success criteria for its constituent plants and may be a helpful resource. SC-C3-02 Sources of uncertainty are addressed in a draft evaluation using SC-C3, AS-C3, frhis is judged to be a documentation guidance from draft EPRI report, "Treatment of Parameter and HR-D6, H R-G9, f.'onsideration only and does not affect the Model Uncertainty for Probabilistic Risk Assessments." H R-13, DA-E3, QU- echnical adequacy of the PRA model. An appropriate characterization of uncertainty is required to E1 , QU-E3, QU-E4, support risk-informed decision making. QU-F4, LE-F2, LE-Apply the EPRI guidance, once final ized, to identify the sources of G4 uncertainty in the analysis. SY-A4-01 System walkdown documentation not included in the system SY-A4 This is judged to be a documentation notebook documentation. consideration only and does not affect the A review of system notebooks and available documentation does technical adequacy of the PRA model. not include system walkdown information. A draft document containing photos and docu mentation of insights from a system walkdown was provided to the peer review but is not finalized . Finalize the provided notebook. SY-A6-01 Missing boundary definitions for system models. I SY-A6, SY-C2 This is judged to be a documentation The system notebooks do not clearly define the boundaries. The consideration only and does not affect the training documentation is not adjusted to be specific to the PRA technical adequacy of the PRA model. model. Additionally some systems, such as ac power, do not include discussion of modeled events. The diesel generator and the fuel oil transfer system are not addressed explicitly. Develop PRA specific illustrations and expand documentation to clearly describe the system boundaries to ensure that no components are doljble counted or m issed. 12

Attach m ent 3 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements SY-A8-01 Review of notebooks and data notebook did not provide a source SY-A8 This is judged to be a documentation for i nclusion or exclusion of failure modes based on data consideration only and does not affect the boundaries. technical adequacy of the PRA model. No documentation of component boundaries. Expand the data discussion to provide component definitions. SY-A1 0-01 Some systems do not incl ude expected fail u re modes and SY-A1 0 This is judged to be a docu mentation although this may be correct, there is no documentation as to how consideration only and does not affect the the data boundaries encompass the expected failures. technical adequacy of the PRA model . One example is the diesel generator model does not include the diesel generator day tank and instrumentation. The response to inquiries was that these components are part of the diesel skid package. This is usually separate modeling to capture m iscalibrations. Define what is included within the diesel generator "box" or expand the model. SY-A1 2-01 Review of system models identified some m issing component SY-A1 2 Components are included in the PRA failure modes. model either explicitly or as part of a Required components are not always addressed in the model. super-component. This is judged to be a For example, the diesel generator day tank and fuel oil check documentation consideration only and valves are not included . Additionally, restart of some components does not affect the technical adequacy of (such as dampers having to re-open for CAV) are absent in the the PRA model. model. Define boundaries to show incorporation of failure modes by other events or expand model. SY-A1 3-01 The modeling excludes some required component failures without SY-A1 3 This is judged to be a documentation

              *ustification.                                                                        consideration only and does not affect the Some failure modes listed for inclusion i n the SR are not found or                   technical adequacy of the PRA model .

are excluded from the m odel. This includes the*transfer closed/plugging failure modes for valves and the absence of some check valves and/or tanks. J ustify the exclusion of any failure mode or model the failure mode. --* -* - - - 13

Attachm ent 3 LR-N 1 3-0 1 23 Applicable Finding I Finding Description I Supporting Resolution Requirements SY-A1 6-01 IThe SWS fault tree includes recovery via alignment of the header ISY-A1 6 Action can be taken from the control room crosstie. HFE SWS-XHE-FO-OVER2 is used for this recovery and will be taken within 1 0 minutes or action in all cases, even LOSP. However, the timing used in the less based on responses to control room H RA for this action is based on room heatup fol lowing a loss of alarms, which should be adequate. CAV, not on the more restrictive timing req uired for recovery of Minimal impact to appl ication. cooling to a d iesel following LOSP. jApplication of the HFE for recovery of SW via the header crosstie in the incorrect context may result in underestimating the importance of the HFE and associated equipment required for the recovery. Create a variation of the SWS-XH E-FO-OVER2 HFE accounting for differences in tim ing during LOSP conditions where cooling to a diesel generator is required. SY-A1 9-01 !System notebooks do not include discussions on potential jsY-A1 9 , SY-A20 ,This is judged to be a docu mentation adverse operating conditions that could impact operation. consideration only and does not affect the No documentation of any potential for loss of desired system technical adequacy of the PRA model. function, e.g., excessive heat loads, excessive electrical loads, excessive hum idity, etc.

             'Add brief d iscussion.

SY-A21 -01 T he current type code does not provide consistent nomenclature ISY-A21 , QU-A2b Same data are used for different types of

                                                                                                    !failu res when d ata are lacking and a for same failure data.

1The SR indicates that the nomenclature should use the same surrogate data set is required (e.g. diesel identifier for the same failure mode. The type code changes by air compressors). Minimal impact to system although the data is from the same source. application . . For data sources from the same reference the same type code should be used. Using type codes by system may obscure the state of knowledge information. _ * . 14 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements SY-83-01 The review of the system model and documentation identified Additional battery charger CCF terms are cases the selection of CCF combinations are not com plete and now included . those selected are not the most limiting. Exam ple of incorrect usage is found for the de chargers. Combinations of 3, 4 and 5 of six chargers are not included in current model. Additionally, CCF for two of two on same bus is m odeled but cross train is not addressed (A & 8, A & C, B & C) which are more significant. Review and revise as appropriate the selection of CCF combinations and model all possible combinations of CCF. SY-85-01 Documentation i nd icated that the heated water circulating system SY-85 This is judged to be a documentation was requ ired. consideration only and does not affect the Documentation for several system notebooks (AFW, CVCS and technical adequacy of the PRA model. RWST) indicated that the heated water circulating system was required to prevent freezing, but was not modeled. Model the heated water circulating system or justify the reason for not modeling. SY-86-01 No documentation provided related to analysis of support system Support system requirements are requirements. analyzed, modeled, documented. There appears to be no analysis of support system requirements concurrent with their definition in the system notebooks. Perform the required engineering analysis. SY-81 1 -01 Some AFW signals (SI, LOSP) are not defined and no justification SY-81 1 This is judged to be a documentation for exclusion is provided - consideration only and does not affect the The SR states that actuation signals m usfbe considered or technical adequacy of the PRA model.

               *ustification provided. The AFW start signals are not completely modeled and justifications for exclusion are not provided .

Provide justification for exclusion of the AFW signals or model these signals. 15

Attachm ent 3 LR- N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements SY-81 2-01 ISome run times for components do not reflect the actual required Mission times changed in current m odel m ission time. Several components reflect 8 hour run times (DG and control room fans as examples) when the required mission time is continued operation 24 hours. The design generator and turbine driven pump run time of 6 hours is not sufficient to address the total run time of 24 hours. Justify m ission times or revise the m ission tim es to the required value. H R-82-01 I DO NOT screen activities that could simultaneously have an Section 4.3.3.1 of the H RA Notebook i mpact on m ultiple trains of a redundant system or diverse which allows screening of actions that systems (HR-A3). could simultaneously have an im pact on m ultiple trains of a redundant system or diverse systems is in violation of this. This requirement is not met. Change the documentation to reflect that the activities are being screened because they are either not in the PRA model or do not impact any success criteria. No screening performed. Documentation clarified . 16 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements HR-C3-01 There is no documentation showing that miscalibration as a mode H R-C3, SY-81 6 Can be considered to be modeled by of failure of initiation of standby systems was considered. An common-cause failure events. No impact example of this is that there is no HFE for miscalibration of bus expected to application. undervoltage bus, RPS relays, etc. There is no documentation showing that miscalibration as a mode of failure of initiation of standby systems was considered. An example of this is that there is no HFE for miscalibration of bus undervoltage bus, RPS relays, etc. Consider analyzing the miscalibration of standby systems. H R-F2-01 Complete the definition of the H FEs by specifying: This information is available. The cited (a) Accident sequence specific timing of cues, and time calculations were reviewed and found to window for successful completion provide appropriate basis for the related (b) Accident sequence specific procedural guidance (e.g., operator actions. No impact to AOPs, and EOPs) application. (c) The availability of cues and other indications for detection and evaluation errors. (d) The complexity of the response. (Task analysis is not required.) The accident sequence specific timing of time window for successful completion for CCS-XH E-FO-ISOLT is not based on a calculation that addresses leakage. The calculation S-CC-MDC-2 1 1 1 is for loss of Service Water and does not address leakage of the Component Cooling Water System. The time window should account for leakage that would drain the CCW system and make it inoperable. This is the limiting time since the CCW system will continue to cool with the leak until the surge tank is drained. This is only one example of a timing window error. Review all HRA analysis to verify that the time window in the analysis is based on an applicable calculation. This review needs to be documented 17 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements DA-A1 a-01 No discussion of component boundary definition is provided in DA-A1 a, DA-C1 This is judged to be a documentation either the data or systems analysis. Boundaries for unavailability consideration only and does not affect the events are not established. technical adequacy of the PRA model. Boundary definitions help assure that failures are attributed to the correct component and that calculated failure rates and u navailability values are appropriate. Some component boundaries are d iscussed in the notes to Appendix A, "Generic (Industry) Failure Data" of the Data Notebook. Note 32 states to "Assume that CCW/RHR HX failure rates apply to TDAFW Pump Bearing and governor jacket coolers", however unless the Salem TDAFW pum p has unique features that require this to be modeled separately, cooling to the TDAFW pump is usually included in the component boundary to the pump. Define the component boundary for each component consistent with the failure data source. Establish boundaries for unavailabil ity events consistent with definitions in the systems analysis. DA-A2-01 Mean values for failure rates appear in the model; however no Data distribution information is uncertainty distributions could be found in the basic event documented. database. Failure rates used in the model are not exact and uncertainty distributions are needed to help bound the analysis. Include data distributions in the database in the model. DA-C1 -01 Generic u navailability data is used for some SSCs without DA-C1 This is judged to be a documentation demonstrating that the data is consistent with the test and consideration only and does not affect the maintenance philosophies for the subject plant. technical adequacy of the PRA model. Generic unavailability data may not be appl icable to Salem if its T&M approach is d ifferent. --- Review and state that any generic unavailability data used is applicable to the Salem model. 18

Attachm ent 3 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements DA-C2-01 Plant-specific data is only collected for MSPI components. The DA-C2 Plant-specific data is collected and PRA procedure requires plant specific data to be collected for maintained by the station for components components with a RAW >2 or an F-V greater than 0.005. and systems such as those tracked for MSPI components are only a subset of the risk-significant MSPI. The PRA was u pdated with this components. plant specific d ata. This is believed to be Expand collection of plant-specific data to all modeled appropriate. Documentation issue. components or justify why the generic data is applicable. DA-C4-01 Documentation describing the process of evaluating maintenance Plant data were developed from existing records for failures could not be identified. All failures m ust be plant programs (e.g. MSPI, maintenance reviewed for applicability to the PRA model. rule, etc.) which is also an acceptable approach. Failure rates are dependent on an accurate failure count. Document the process of evaluating maintenance records for failures, ensuring failures are reviewed for applicability to the PRA m odel in accordance with S R-DA-C4. DA-C5-01 Documentation describing the process for counting component This information is available in the failures could not be identified. process descriptions for relevant plant programs. Failure rates are dependent on an accurate failure count. Counting repeated failures occurring within a short interval could skew the importance of sse. Document the process for counting component failures, consistent with S R DA-C5. The draft data procedure provided did not discuss counting of repeated failures in a short interval. 19 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements DA-C6-01 Documentation describing the process of evaluating the num ber of DA-C6 This is judged to be a documentation plant specific demands for standby components could not be consideration only and does not affect the identified. Standby components were identified in Table 1 of the technical adequacy of the PRA model. Data Analysis Notebook and plant specific demands for some of these components were listed in Appendix B, however the basis for these num bers of demands was not provided . The draft data procedure states that plant specific data should be estim ated by actual counts of hours or demands from logs or counters, use of surveillance procedures to estimate the frequency of demands and run times, or estimates based upon input from the System E ngineer. Failure rates are dependent on an accurate demand count or component im portance could be skewed. Standby components were identified in Table 1 of the Data Analysis Notebook and plant specific demands for some of these components were l isted in Appendix B, however the basis for the num ber of demands was not provided. The draft data procedure states that plant specific data should be estimated by actual counts of hours or demands from logs or counters, use of surveillance procedures to estimate the frequency of demands and run times, or estimates based upon i nput from the System Engineer. Issue the data collection guidance document and document/justify the basis for the demands used . --- -- 20 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements DA-C7-0 1 Documentation describing the process of collecting the num ber of surveillance tests and planned maintenance activities on plant Plant data were developed from existing components could not be identified . In Appendix C for example plant programs (e.g. MSPI, maintenance CCS MOVs in test and Maintenance were described . The source rule, etc.) which is an acceptable of the data was listed as Salem 3.2 PRA, however no specific approach. breakdown of the surveillance tests included was provided. The draft data procedure identifies surveillance tests as a source of data. Maintenance and testing unavailability are dependent on an accurate review of test and maintenance procedures. Document the source of maintenance and testing activities. DA-C9-01 Documentation describing the process of estimating the Plant data were developed from existing operational time of standby components from testing was plant programs (e.g. MSPI, maintenance identified in draft procedure. Standby components were identified rule, etc.) which is an acceptable in Table 1 of the Data Analysis Notebook and operational times for approach. some of these components were listed in the Data Analysis Notebook, however the source of the data was not provided . Failure rates are dependent on an accurate run times. Document the source of data for the actual run times of standby components. DA-C1 0-01 Compare the initiator frequencies used in the Salem model with DA-C1 0 This is judged to be a documentation other generic data sources. consideration only and does not affect the This F&O is characterized as a suggestion because the IE technical adequacy of the PRA model. notebook does include a comparison with N U REG/CR-5750. It is recommended for com pleteness to check how the Salem set of initiators com pares with other data sources. 21 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements DA-C1 1 -01 Maintenance and testing u navailability were identified in the Plant data were developed from existing I model; however no specific surveillance tests were discussed in plant programs (e.g. MSPI, maintenance ! the Data Analysis Notebook. MSPI/Maintenance Rule sources rule, etc.) which is an acceptable were identified. Docum ent the specific surveillances or plant approach. maintenance contributing to the unavailability of plant components. Document the process for counting these durations in a data procedure. DA-C1 1 A- Documentation describing the process of using m aintenance and Plant data were developed from existing 01 testing durations to determine plant specific durations was plant programs (e.g. MSPI, maintenance identified in a draft document. rule, etc.) which is an acceptable approach. Component availability depends on an accurate count of m aintenance unavailability. Document the process for counting m aintenance unavailability in a data procedure. DA-C1 2-01 There was no specific documentation or guidance document Plant d ata were developed from existing provided that d iscusses how maintenance was treated for shared plant programs (e.g. MSPI, maintenance systems. rule, etc.) which is an acceptable approach. Com ponent availability depends on an accurate count of maintenance unavailability including shared systems. While a table of critical hours was provided and the Maintenance Unavailability Table provided in Appendix C appears to address these hours there was no specific documentation or guidance document provided that discusses how maintenance was treated for shared systems. 22 LR-N 1 3-0 1 23 Applicable Finding Finding Descripticn Supporting Resolution Requirements DA-C1 3-01 Coincident u navailability for service water pumps was modeled as With the exception of service water and a shown in Appendix C of the Data Analysis Notebook, however, no handful of other systems, concurrent overall guidance document could be found to ensure all systems unavailability of m ultiple components ** were reviewed for coincident unavailability. would require a prompt shutdown. This condition is remote and is not modeled. Component availabil ity depends on an accurate count of maintenance unavailability. Document the review of coincident unavailability in plant systems. DA-D3-01 Several items listed in Table A-1 do not contain any reference Bounds have been provided. information for either error factor or basic input parameters from I which an error factor can be derived. Provide information related to the bounds of the failure rates. 23 L R-N 1 3-0 1 23 Applicable Finding Finding Description Resolution Supporting Requirements OA-04-01 No documentation exists related to the comparison between the OA-04 This is judged to be a documentation generic value and the plant-specific update value to ensure consideration only and does not affect the accurate and m eaningful im plementation of Bayes approach. The technical adequacy of the PRA model. documentation only indicates that data came from NUREG/CR-6928 and that MSPI data was used to perform the update. It then references Appendix B which is only Table B-1 . The table provides limited information related to the update and does not provide any comparisons of results or discussions with regard to applicability of results. The documentation only indicates that data came from N UREG/CR-6928 and that MSPI data was used to perform the update. It then references Appendix B which is only Table B-1 . The table provides l imited information related to the update and does not provide any comparisons of results or discussions with regard to applicability of results. Perform comparisons of results with regard to initial ranges of possible generic values and confirmation that the updated results are within the expected range. Also confirmation that plant data, due to relatively small generic alpha factors is not biasing the updated value. OA-06-01 No documentation is present that provides any comparisons OA-06 This is judged to be a documentation between data sources. consideration only and does not affect the Perform the evaluation. technical adequacy of the PRA model. OA-E2-01 A draft document was provided that documented how to establish OA-E2, OA-C6 This is judged to be a documentation component boundaries, how to establish failure probabilities, OA-C7 consideration only and does not affect the sources of generic data, etc. This procedure needs to be OA-C8 technical adequacy of the PRA model. formalized . OA-C9 The draft document discussing how to perform data analyses needs to be finalized to ensure qual ity. Provide procedure on how to perform data malysis. 24 LR-N 1 3-0 1 23 Applicable Finding Finding Descriptign Supporting Resolution Requirements I F-A4-01 Appendix A of the SA-PRA-01 2, Revision 0 contains a summary This information was available; location of of the walkdown performed for the internal flooding analysis. information is at most a documentation However, it does not contain the details of the walkdown notes issue. such as spatial information, plant design features, mitigating equ ipment such as drains, sumps, doors, wall penetrations, etc. This F&O is characterized as a finding because there was insufficient documentation available to verify the SR. Include walkdown sheets in with the documentation that includes the observations of the walkdowns. IF-B 1 a-01 This SR requires consideration flood sources for multi-unit sites. This was done. AB-0848 scenario is an The internal flooding notebook does not contain documentation exam ple. that Unit 2 flood sources could or could not impact Unit 1 and vice versa. This F&O is characterized as a finding because there was insufficient documentation available to verify the SR. Assess whether Unit 2 flood sources can im pact Unit 1 and vice versa . I F-C1 -01 Propagation paths are not documented for each flood area. See Appendix E of the Internal Flooding report. Very low risk areas were not The requirement specifies that the propagation paths should be addressed using the same level of detail identified . as for higher risk areas. (Note 1) Document propagation paths for each flood area. - - 25 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements IF-C2-01 Plant design features that have the ability to term inate or contain Plant design feature information is the flood propagation are not documented for all defined flood provided for those areas which could not areas. be shown to be unimportant. The information contained in Appendix A does not provide Information was not gathered if no documentation information for each flood area. important floods were identified in the area. Document the required information for each flood area. IF-C2b-01 The documentation does not provide spatial information for See Appendix E of the Internal Flooding components. report. Very low risk areas were not addressed using the same level of detail This is required information for flood areas. as for higher risk areas. (Note 1) Document the required information for each flood area. I F-C2C-01 The propagation paths and spatial information is not provided for See Appendix E of the Internal Flooding SSCs contained in flood areas. The evaluation limits the report. Very low risk areas were not propagation paths to only those found to be of highest frequency. addressed using the same level of detail Spatial information is not provided for components listed in as for higher risk areas. Appendix D with respect to potential flood sources. (Note 1) Document the required_i_nformation fo_l"_each flood area. - -- I 26 LR- N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements IF-C3a-01 Appendix D of the PRA I nternal Flood Evaluation states that "For This was our informed judgment based on spray scenarios, however, walkdown observations revealed that empirical observation. Experience shows Air-Operated Valves (AOVs) and Motor-Operated Valves (MOVs) that water spray does not generally were of a robust design that would exclude them from being prevent AOVs and MOVs from operating . susceptible to water damage. Hence, these components were not Therefore the assumption is believed to automatically failed (PRA event equal to T RUE) for quantification be appropriate for best-estimate PRA of the CCDP." This is not an adequate basis for determining the work. No impact expected to application. susceptibility of these components to flood-induced failure mechanisms per this S R. I I mproperly screening SSCs from flood-induced failure could lead to underestimating the risk associated with a flood sequence. Per the SR, take credit for the operability of SSCs identified in I F-C2c with respect to internal flooding impacts only if supported by an appropriate combination of: (a) test or operational data (b) engineering analysis (c) expert judgment. IF-C3b-01 Propagation was not performed for initial screening. This was done for any flood which could contribute to CDF. This information was The propagation paths for systems during the initial quantification not developed for all areas. If, for were not defined or util ized to perform the flood area screening. instance, no source within or external to This can result in screening sequences that could be im portant. an area could impact equipment in that area or in other areas which the area Identify propagation paths for each flood area. would drain to, then it was not necessary to develop detailed propagation information. 27 L R-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements IF-C4-01 Flood scenarios were screened without development of flood rate, Detailed assessments were provided for source, and operator actions. Detailed assessments were only those floods which could not be shown by i provided for selected h igh-frequency floods. screening to be negligible risk I contributors. Guidance does not ind icate Improperly screening flood scenarios could lead to that detailed information must be underestimating the risks associated with i nternal floods. gathered for locations once they are shown not to contribute to flood risk. Provide a more thorough development of all flood scenarios IF-C4a-01 Documentation of m ulti-unit scenarios could not be identified . This was considered; see for exam ple flood AB084B For completeness, the potential for m ulti-unit scenarios needs to be addressed. Address the potential for multi-unit internal flood scenarios. QU-A2b-01 Parametric uncertainty is not performed on the quantification Uncertainty information has now been results. In addition, it is not clear that the same type code is used provided. for multiple events based upon the same underlying data. For Category I I , the "state-of-knowledge" correlation m ust be accounted for in determining the m ean. Since the u ncertainty characterization for basic events is not carried into the CAFTA database and no Monte Carlo techniques are used to generate the mean CDF, this S R is only met at Category I . Incorporate the uncertainty bounds i nto the CAFTA database to allow generation of a CDF mean accounting for the "state-of-knowledge" correlation. This may also requ ire revision of the type code applications to ensure all basic events relying on the same underlying date> are correctly correlated. ---- 28 LR- N 1 3-01 23 Applicable Finding Finding Description -- - Supporting Resolution Requirements QU-A4-01 Recovery events NRAC-12H, NRAC-OSP, and N REDG-4H are Recovery file was reviewed . included in the S 1 R4REC.CAF file, but their application is not discussed in the Accident Sequences and Event Tree notebook or in the A.C Power System Notebook. The model owner stated that the recovery events in question should not have been used in the latest revision. However, it appears that their inclusion did not significantly affect the results. Review the recovery file to ensure only those events intended to I be applied are included. Provision of a l isting of all recovery events and their intended application in the Quantification Notebook could facilitate this review for future model revisions. QU-83-01 Either applies a truncation limit satisfying the criteria of "final Truncation evaluation has been updated. change is less than 5%" for both CDF and LERF or use a lower truncation limit to the LERF quantification to satisfy the criteria. I QU-85-01 Salem Quantification Notebook SA PRA-2008-01 Attachment E QU-85 Convergence validation was updated . documents the convergence analysis performed to set an The circular logic issue is judged to be a appropriate truncation value. The truncation level for both CDF documentation consideration only and and LERF was set at 1 .0E-1 1 . The percentage change between does not affect the technical adequacy of 1 .0E-1 0 and 1 .0E-1 1 was 2 .2% for CDF, but 6 .1 % for LERF. the PRA model . Therefore, this S R was not satisfied for LERF. The supporting requirement applies the same criteria for convergence to both CDF and LERF. The criteria were satisfied for CDF, but not LERF. Document the overall philosophy and method for breaking circular logic in the Quantification notebook and provide sufficient documentation in the system notebooks to provide assurance that unnecessary conservatisms or non-conservatisms are not introduced. - L__ 29 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resolution Requirements QU-89-01 Split fractions and undeveloped events are included in the model . QU-89 This is judged to be a documentation Examples include main feedwater availabil ity for ATWS (MFI- consideration only and does not affect the UNAVAI LABLE) and some Unit 2 systems credited for recovery of technical adequacy of the PRA model. Unit 1 CAV failure (G2SW22). The derivation of the values for these events is not documented. The derivation of split fractions and undeveloped events is not documented sufficiently to allow identification of shared events and results i nterpretation based on individual events subsumed into the split fraction. Document the derivation of any split fractions and undeveloped events used in the model sufficiently to allow results interpretation and to provide assurance that the im pact of any shared components is appropriately considered. QU-D1 b-01 There is no discussion in the quantification notebook that indicates QU-D 1 b This is judged to be a documentation a review of the results was performed for the purpose of consideration only and does not affect the assessing modeling and operational consistency. Also, since the technical adequacy of the PRA model. sequences were not quantified, it is difficult to perform this verification. This F&O is characterized as a finding because there was insufficient documentation available to verify the SR. Review the results for modeling consistency (e.g ., event sequence models consistency with systems models and success criteria) and operational consistency (e.g., plant configuration, procedures, and plant-specific and i ndustry experience) and include in the quantification notebook. 30 LR-N 1 3-0 1 23 I Applicable Finding Finding Description Supporting Resolution Requirements QU-03-01 This is a Capability Category I because there is no documentation QU-03 This is judged to be a documentation to i ndicate that the Salem results were com pared to the results of consideration only and does not affect the a similar plant. technical adequacy of the PRA model. No documentation was provided showing this requirement was met. Provide a comparison of initiating event contributions and significant basic event im portances between Salem and similar plants based on information available in the PWROG PRA Com parison Database. QU-04-01 There is no documentation indicating that a sampling of non- QU-04 This is judged to be a documentation significant accident cutsets or sequences were reviewed to consideration only and does not affect the I determine they are reasonable and have physical m eaning. technical adequacy of the PRA model. Quantification Notebook Section 2 only requires a review of the top 1 00 cutsets. Review of a sam pling of non-significant cutsets can also reveal logic problems or recovery rules which are not being applied correctly. Include a requirement for review of a sampling of non-significant sequences in Section 2 of the Quantification Notebook and in procedures governing the model update process. 31 LR-N 1 3-0 1 23 Applicable Finding Finding Description Supporting Resol ution Requirements QU-F2-01 This requirement was only partially met as described below: QU-F2, QU-F3, QU-F6 This is judged to be a documentation (a) This requirement is met by the system and HRA notebooks . consideration only and does not affect the (b) There is a cutset review process description technical adequacy of the PRA model. (c) There is no description of how the success systems are accounted for. Since a one top tree is used the software already accounts for I this. A statement stating would be satisfactory. The truncation values and how they were determined were documented. The method for applying recovery and how post initiator HFE's are applied was not described. (d) This requirement was met. (e) This requirement was met (f) This requirement was not met since the cutsets per accident sequence were not discussed. (g) This requirement was not met since equipment or human actions that are the key factors in causing the accidents sequences to be are not discussed. (h) This requirement was not met since sensitivities were not documented. (i) This requirement was not met since the uncertainty notebook was not finalized.

0) This requirement is not met since there is no discussion of importance.

(k) This requirement is not met because there is not list of mutually excl usive events and there justification. (I) This requirement is not met because there is no discussion of asymmetries in quantitative modeling to provide appl ication users the necessary understanding regarding why such asymmetries are present in the model. (m) This requirement is met since CAFTA and Forte are being used. Both of these pieces of software are industry standards and therefore no further testing is required. Several documentation items called for in this supporting requirement were not available for review. Specific items not included in the documentation were: the process used to account for system successes, accident sequence results, discussion of factors causing accidents to be non-dominant, sensitivity assessments, uncertainty distribution, importance measure results, basis for elimination of mutually exclusive events, asymmetries in the model, and a quantitative definition of significant basic event, significant cutset, and significant accident sequence . Expl'md the documentation to address the items documented in the F&O. 32 L R-N 1 3-01 23 Applicable Finding Finding Description Supporting Resolution Requirements LE-C8a-01 Equipment survivability and human actions under adverse LE-C8a, LE-C8b, No credit is taken for equipment or environments m ust be considered to reach Category I I . LE-C9a, LE-C9b actions under adverse environments. N o documentation provided o r credit taken for equipment or This is judged to be a documentation operators in adverse environment. consideration only and does not affect the Provide d iscussion on environmental conditions and the effects on technical adequacy of the PRA model. operator actions. LE-D1 b-01 Requirements are to address penetrations, hatches and seals. This was evaluated as only a documentation issue. No documentation presented in the containment isolation documentation that the required analysis was performed. Perform analyses for penetrations, hatches and seals. LE-06-01 Consider both failures of isolation and safety systems. The Cl This was evaluated as only a model (SA-PRA-005.07) does not provide sufficient information documentation issue. and does not address potential failures due to air locks or other locations. Perform detail analyses for failures due to air locks and other locations. LE-F1 b-01 Other than verifying that the sum of the three end states (INTACT, LE-F1 b This is judged to be a documentation LATE and LERF) is approximately equal to the core damage consideration only and does not affect the frequency, no checks on the reasonableness of the LERF technical adequ acy of the PRA model. contributors is documented . A review for reasonableness is required to meet the intent of this S R. Review contributors for reasonableness (e.g., to assure excessive conservatisms have not skewed the results, level of plant SpE)Cificity is appropriate for significant contributors, etc.). --------- -- - 33 LR-N 1 3-01 23 Applicable Finding Finding Description Supporting Resolution Requirements LE-F3-01 LERF uncertainties are not characterized consistent with the LE-F3 This is judged to be a documentation requirements in Tables 4.5.8-2(d) and 4.5.8-2(e). consideration only and does not affect the LERF uncertainties m ust be appropriately characterized to meet technical adequacy of the PRA model. the intent of this S R. The NEI 04-1 0 methodology explicitly Characterize the LERF uncertainties consistent with the addresses u ncertainty. requirements in PRA Standard Tables 4.5.8-2(d) and 4.5.8-2(e). LE-GS-01 Limitations in the LERF analysis that would impact applications LE-GS This is judged to be a documentation are not documented. consideration only and does not affect the Limitations in the LERF analysis that would impact applications technical adequacy of the PRA model. m ust be discussed to meet this S R. Document the lim itations in the LERF analysis that would impact applications. LE-G6-01 A definition for significant accident progression sequence is not LE-G6 This is judged to be a docu mentation documented. consideration only and does not affect the A definition for significant accident progression sequence must be technical adequacy of the PRA model. included to m eet this S R. Include in the documentation a definition for significant accident progression sequence. M U-C1 -01 There is no reference to a review of the cumulative impact of Documentation issue. pending changes. Multiple changes to PRA inputs can necessitate the need for a PRA model update/upgrade. Revise the FPIE model procedure to require a review of the cumulative im pact of impending changes 34 L R- N 1 3-0 1 23 Note 1 : During the review of the Unit 1 RI-/SI relief request f.or..the-Unit 1 Fourth 1 0-year interval, the NRC asked the following question which is documented in PSEG 's response dated Apri/ 12, 20 1 1 (ADAMS Accession No. ML111030186) NRC Question 1: "Attachment 3 of the licensee's Jetter dated October 2 1 , 20 1 0, lists the following findings as gaps to the internal flooding requirements to the ASME Standard: IF-C1-01, IF-C2b-01, and /F-C2C-01. The resolution of these gaps state "See Appendix E of the Internal Flooding report. Very low risk areas were not addressed using the same level of detail as for higher risk areas. " A quantitative guideline is not provided which would discern "low risk" from "high risk" areas as is provided in resolution of supporting requirement JF-C3b-01. Since a qualitative approach is assumed to be used for screening, please provide an explanation which shows that meeting IF-C1-01, JF-C2b-01, and IF-C2C-0 1 in entirety would not result in greater pipe segments for high or medium categories in the risk matrix. " Response to Question 1: The i nternal flood analysis d id q u antify a l l postulated scenarios that were not screened on the basis of S u p p o rting Req u i re m ent ( S R) I F-C5 of ASM E/ANS RA-Sb-2005, which is now identified as I FS N-A1 2 per ASM E/ANS RA-Sa-2009a, and also accounted for affected SSCs d ue to i n ter-area pro pagatio n pathways per I F-C3b ( I FS N-A8 p e r ASM EIAN S RA-Sa-2009a). For those q u antified scenarios, a con ditional core d amage probab i l ity (CCDP) was com puted that con servatively considered a l l PRA-modeled SSCs to be damag ed by a flood orig inating or propagating i nto a particular flood a rea, whi ch satisfied the ASM E S R I F-C1 ( I FSN-A 1 per ASM E/ANS RA-Sa-2009). When a n alyzing the vario u s i nternal flood scenarios, credit was not i n itially g iven for area d rainage a n d other m itigating effects i n o rder t o estimate a worst-case C C D P for that scenario. The CCDP was then m u lti p l ied by t h e flood i n itiating freq u e n cy to est i m ate the core damage fre q uency ( C DF). If the CDF for a g i ve n flood scenario was sufficiently low, e . g . , less than 0. 1 % of the nominal i nternal events CDF, then no further refinement was d ee m ed necessary. H owever, if first estimates of the core d am age freq uencies for that com partment proved too pessimistic, the affected a rea of the plant was a n alyzed i n g reater detai l to take i n to account spatial effects, s pecific flooding flow rates , operator actions, d ra i n ag e pathways, etc . , which satisfied the considerations given in SRs I F-C2b a n d I F-C2c ( I FSN-A4 a n d I FSN-A5 per ASM E/ANS RA-Sa-2009, respectively). Therefore, the j ustification for m o re detailed modeling of certai n floodi ng scenarios was aimed at removing some of the conservatism of the m ethodology, whi l e at the same time provi d ing-a real istic basis for not ass u m i ng complete fai l u re of a l l scenario-specific e q u i p m ent due to a cred ible flood ing event. In reviewing the CDF resu lts for the flood scenarios rep o rted in the I nternal Flood Analysis documentation, many of the scena rios were less than a tenth of a percent of the nominal CDF value. Col l ectively, these 35 LR-N 1 3-0 1 23 scenarios contributed a s u m total of less than 0 . 6% of the nominal C D F and were excluded from a n y further detailed a nalysis, which wou l d have i nvoked the m itigative measu res a l l owed per S Rs I F-C2b a n d I F-C2c. Therefore, those i nternal flood scenarios where deta i led modeling was not performed were not "scren_edll fmm q u antification. These scena rios were m erely excluded from any further detailed analysis that would i nvol ve hyd rau l i c modeling of flood areas, which wou l d invo lve deve l o p ment of specific flood ing flow rates and height of water as a fu n ction of time for various flood a reas. There was l ittle benefit to be gained from perfo rming time i ntensive detailed eva luations of scenarios that p roved to be relatively insign ificant in com parison to other i nternal flood sce n a rios. 36}}

LR-N13-0123, Request for Authorization to Continue Using a Risk-Informed Inservice Inspection Alternative to the ASME Boiler and Pressure Vessel Code Section XI Requirements for Class 1 and 2 Piping

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LR-N13-0123, Request for Authorization to Continue Using a Risk-Informed Inservice Inspection Alternative to the ASME Boiler and Pressure Vessel Code Section XI Requirements for Class 1 and 2 Piping

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