L-10-025, 10 CFR 503.55a Requests for Alternative Non-Destructive Examination Requirements for ASME Class 1 and Class 2 Piping Components

10 CFR 503.55a Requests for Alternative Non-Destructive Examination Requirements for ASME Class 1 and Class 2 Piping Components

ML101650649
Person / Time
Site:	Beaver Valley
Issue date:	06/10/2010
From:	Harden P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-10-025
Download: ML101650649 (40)

Text

FENOC SNrtP.O.

Beaver Valley Power Station Box 4 FirstEnergyNuclear Operating Company Shippingport,PA 15077 PaulA. Harden 724-682-5234 Site Vice President Fax: 724-643-8069 June 10, 2010 L-10-025 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 10 CFR 50.55a Requests for Alternative Non-Destructive Examination Requirements for ASME Class 1 and Class 2 Pipinq Components Pursuant to 10 CFR 50.55a(a)(3), FirstEnergy Nuclear Operating Company (FENOC) is requesting Nuclear Regulatory Commission (NRC) approval for continued use of the existing Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2),

risk-informed inservice inspection (RI-ISI) program, with updates, relevant to certain non-destructive examination (NDE) requirements associated with American Society of Mechanical Engineers (ASME) Class 1 and Class 2 piping components.

Proposed alternative RI-ISI-1, included as Enclosure A, would be implemented during the BVPS-1 fourth ISI interval. FENOC is requesting approval of alternative RI-ISI-1 by February 1, 2011 to support the scope freeze milestone for the BVPS-1 April 2012 refueling outage.

Proposed alternative RI-ISI-2, included as Enclosure B, would be implemented during the BVPS-2 third ISI interval. FENOC is requesting approval of alternative RI-ISI-2 by July 1,2011 to support the scope freeze milestone for the BVPS-2 September 2012 refueling outage.

Pursuant to Regulatory Guide 1.200, "An Approach for Determining the Technical.

Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Revision 1, a summary of the BVPS-1 and BVPS-2 probabilistic risk assessment model's capability for use in RI-ISI program activities and initiatives, is provided as Enclosure C.

AO, 7 k-Idi-

Beaver Valley Power Station, Unit Nos. 1 and 2 Letter L-10-025 Page 2 of 2 There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at (330) 761-6071.

Sincerely, Paul .Hre

Enclosures:

A. Beaver Valley Power Station Unit No. 1, 10 CFR 50.55a Request RI-ISI-1, Revision 0 B. Beaver Valley Power Station Unit No. 2, 10 CFR 50.55a Request RI-ISI-2, Revision 0 C. FirstEnergy Nuclear Operating Company, Beaver Valley Power Station Unit Nos. 1 and 2, Probabilistic Risk Assessment Technical Adequacy for Risk-Informed Inservice Inspection cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Beaver Valley Power Station Unit No. 1 10 CFR 50.55a Request RI-ISI-1, Revision 0 Page 1 of 5 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected ASME Code Class 1 and 2 piping welds as listed in Table 1

2. Applicable Code Edition and Addenda

ASME Code Section XI, 2001 Edition, 2003 Addenda

3. Applicable Code Requirements ASME Code Section XI, 2001 Edition, 2003 Addenda, Inservice Inspection (ISI) requirements for pressure retaining piping welds IWB-2500, Examination and Pressure Test Requirements Table IWB-2500-1, Examination Categories Class 1 Piping Welds Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles Category B-J, Pressure Retaining Welds in Piping IWC-2500, Examination and Pressure Test Requirements Table IWC-2500-1, Examination Categories Class 2 Piping Welds Category C-F-i, Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping Category C-F-2, Pressure Retaining Welds in Carbon or Low Alloy Steel Piping

4. Reason for Request

On April 9, 2004, Nuclear Regulatory Commission (NRC) staff approved FENOC's ASME Code Section XI Class 1 and Class 2 Risk-Informed Inservice Inspection (RI-ISI)

Program for Beaver Valley Power Station Unit No. 1 (BVPS-1), third ISI interval.

In its approval, NRC staff concluded the RI-ISI program is consistent with WCAP.-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, and is an acceptable alternative to the requirements of ASME Code , Section Xl, for inservice inspection of ASME Class 1 and 2 piping, examination categories B-F, B-J, C-F-i, and C-F-2.

RI-ISI-1 Page 2 of 5 On May 1, 2006, NRC staff approved the Pressurized Water Reactor (PWR) Owners Group Topical Report WCAP-14572, Revision 1-NP-A, Supplement 2. The safety evaluation report [Reference 1] states:

"The NRC staff concludes that the proposed RI-ISI program as described in the approved WCAP-14572, and WCAP-14572, Sup. 2, as clarified and revised by the June 22, 2005, supplemental letter, will provide an acceptable level of quality and safety with regard to the number of inspections, locations of inspections, and methods of inspection."

Consistent with the RI-ISI methodology documented in WCAP-14572, including its supplements, [References 2, 3, 4], new information has been incorporated into the RI-ISI analysis as part of the "living" RI-ISI program. The new information includes changes to the BVPS-1 Probabilistic Risk Assessment (PRA) model, revised segments and failure probabilities for some segments based on industry and plant experience and plant modifications, revised consequences based on lessons-learned, and updated test intervals for certain segments and overlays of pressurizer alloy 82/182 welds in the reactor coolant system.

The changes described above required re-performing the risk evaluation. The revised results were reviewed by the RI-ISI expert panel. Compared to the third interval BVPS-1 ISI program, 26 low safety significant (LSS) segments were reclassified as high safety significant (HSS), and 24 HSS segments were reclassified as LSS; 3 quantitative HSS segments were re-categorized by the expert panel as LSS based on the "with operator action consequences" guidance within WCAP-14572, Supplement 2. The expert panel concluded the remaining segment classifications shall remain as-is.

The change in risk evaluation was performed again to compare the original Section XI program with the revised fourth interval RI-ISI program for BVPS-1. Five reactor coolant system segments and one safety injection system segment (six total segments) are to be added to the BVPS-1 RI-ISI program to meet the change in risk criteria discussed in WCAP-14572, page 214. These six additional examinations are VT-2 visual exams.

No examinations were added for defense-in-depth considerations, which is the same as in the previously approved third interval RI-ISI program.

The proposed RI-ISI program, with updates, provides a 77 percent reduction in required examinations. This directly results in reduced outage scope, decreased individual and cumulative occupational radiation exposure, and shortened outage durations. As such, FENOC requests that the BVPS-1 RI-ISI program, with updates, be approved for continued use during the fourth ISI interval.

5. Proposed Alternative and Basis for Use ASME Section XI categories B-F, B-J, C-F-I, and C-F-2 contain the requirements for examining Class 1 and 2 piping components via non-destructive examination (NDE).

The proposed alternative [continued use of the BVPS-1 RI-ISI program, with updates] is limited to ASME Class 1 and 2 piping components, including piping currently exempt from NDE requirements. The proposed alternative will be substituted for the ASME

RI-ISI-1 Page 3 of 5 Section XI category B-F, B-J, C-F-I, and C-F-2 examination requirements. The applicable aspects of ASME Section XI Code not affected by the proposed alternative will be retained.

The basis of the alternative risk-informed inservice inspection program's methodology is fully described in the NRC-endorsed WCAP-14572 and its supplements.

Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative to the ASME Code Section XI examination requirements will continue to provide an acceptable level of quality and safety.

Comparisons of the ASME Section XI inspection program, the third interval RI-ISI program, and the proposed fourth interval RI-ISI program are presented in Table 1.

6. Duration of Proposed Alternative The proposed alternative shall be implemented during the BVPS-1 fourth ten-year ISI interval and will remain effective until the end of the interval on March 31, 2018.

7. Precedents NRC letter to FENOC, April 9, 2004,

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) - Risk-Informed Inservice Inspection (RI-ISI) Program.

[ADAMS Accession Number ML040780805]

NRC letter to Tennessee Valley Authority, April 30, 2007,

Subject:

Sequoyah Nuclear Plant, Units 1 and 2- Risk-Informed Inservice Inspection Program for the Third 10-Year Intervals.

[ADAMS Accession Number ML071070248]

8. References

1. Nuclear Regulatory Commission (NRC), "Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report WCAP-1 4572, Revision 1-NP-A, Supplement 2, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report Clarifications' Pressurized Water Reactor (PWR) Owners Group Project No. 694," May 1, 2006.

[ADAMS Accession No. ML061160035]

2. Westinghouse Electric, WCAP-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, February 1999.

3. Westinghouse Electric, WCAP-14572, Supplement 1, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection," Revision 1-NP-A, February 1999.

4. Westinghouse Electric, WCAP-14572, Supplement 2, "Pressurized Water Reactor Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report Clarifications," Revision 1-NP-A, September 2006.

RI-ISI-1 Page 4 of 5 Table 1 BVPS-1 STRUCTURAL ELEMENT SELECTION RESULTS AND COMPARISON TO ASME SECTION XI 1989 EDITION REQUIREMENTS Safety ASME Total ASME XI Third Interval RI-ISI Fourth Interval RI-ISla System High Safety Degradation Significant Mechanism(s) Class Code Weld Count Program Segments Exam [Welds requiring Examinations (Qty. of HSS in Category Volumetric (Vol)

Augmented and Surface (Sur)]

Program / Total Vol and Sur only Vol and Sur SES Matrix Number of SES Matrix Number of Qty. of Segments Sur Sur only Region Exam Region Exam in Aug. Program) Locations Locations BD 0 (0 / 27c) FAC/TF Class 2 N/A 0 0 0 0 3 0 3 0 1 B-J 25 287 7 64 1,2,3,4 0 1,2,3,4 3e CH 28(0/0) TF/VF, TF Class 303 17 18 8 + 189b + 2 _ _ 8+17e Class 2 C-F-1 317 Cl 0(0/0) FAC/TF, TF Class 2 N/A 0 0 0 0 4 0 4 0 DV 0(0/0) TF Class 1 B-J 0 106 0 27 4 0 4 0 FW 0 (0 / 27z) FAC/TF Class 2 C-F-2 62 0 14 0 3 0 3 0 HY 0(0/0) TF Class 2 N/A 0 0 0 0 4 0 4 0 MS 8 (8 / 48*) FAC/TF Class 2 C-F-2 106 0 23 0 1, 3 8 1,3,4 8 QS 5(0/0) TFVF Class 2 C-F-1 157 50 12 4 2,3,4 3 2,3,4 19 RC 23(0/0) SCC/TF, Class 1 B-F 18 0 18 0 2,4 7 2,4 23+5 Class 1 B-J 207 181 55 53 13+ 2a SCC/TF/VF/SS RH 19(0/0) TF, TFNF Class 1 B-J 26 0 6 0 2, 3, 4 2 2,3,4 2 Class 2 C-F-I 177 0 14 0 15+_2 b 15 + 2e RS 10(0/0) TF, VF Class 2 C-F-1 84 14 7 2 2,_4 10 2,_4 10 Class 1 B-J 193 108 43 31 1,2, 4 11 +4%+ -1 2,4 12+ 3e SI 37(0/0) TF Class 2 C-F-1 826 147 70 16 16+1l_ 18 + 5 1 SS 0(0/0) TF Class 1 N/A 0 0 0 0 4 0 4 0 Class 2 N/A 0 0 0 0 0 0 FAC/TF, TF, 38 NDE + 43 NDE +

SCC/TF, Class 1 469 682 129 175 3 VISUAL 12 VISUAL TOTAL 130 (8 / 102) SCC/TFNF/SS, TF/VF, VF Class 2 1729 514 157 40 55 NDE + 72 NDE +

28 VISUAL 24 VISUAL Total 2198 1196 286 215 93 NDE + 115 NDE +

31 VISUAL 36 VISUAL

RI-ISI-1 Page 5 of 5 Table 1 BVPS-1 STRUCTURAL ELEMENT SELECTION RESULTS AND COMPARISON TO ASME SECTION XI 1989 EDITION REQUIREMENTS Summary: ASME Section XI selected a total of 501 welds while the proposed RI-ISI program selects a total of 115 welds (plus 36 visual exams), which results in a 77% reduction.

Degradation Mechanisms:

FAC - Flow-Assisted Corrosion SCC - Stress Corrosion Cracking SS - Striping/Stratification TF - Thermal Fatigue VF - Vibratory Fatigue "X/X" indicates combination of mechanisms.

Systems:

BD - Steam Generator Blowdown System CH - Chemical and Volume Control System Cl - Containment Isolation System DV- Reactor Plant Drains and Vents Systems FW - Steam Generator Feedwater System HY - Hydrogen Control System MS - Main Steam System QS - Quench Spray System RC - Reactor Coolant System RH - Residual Heat Removal System RS - Recirculation Spray System SI - Safety Injection System SS - Sampling System Notes for Table 1

a. System pressure test requirements and VT-2 visual examinations shall continue in all ASME Code Class systems.

b. VT-2 visual examination at one location within segment.

c. Augmented programs for erosion-corrosion and high energy line break continue.

d. Examinations added for change in risk considerations (total of six segments- five RC and one SI).

e. VT-2 visual examination for entire segment.

Beaver Valley Power Station Unit No. 2 10 CFR 50.55a Request RI-ISI-2, Revision 0 Page 1 of 5 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected ASME Code Class 1 and 2 piping welds as listed in Table 1

2. Applicable Code Edition and Addenda

ASME Code Section XI, 2001 Edition, 2003 Addenda

3. Applicable Code Requirements ASME Code Section XI, 2001 Edition, 2003 Addenda, Inservice Inspection (ISI) requirements for pressure retaining piping welds IWB-2500, Examination and Pressure Test Requirements Table IWB-2500-1, Examination Categories Class 1 Piping Welds Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles Category B-J, Pressure Retaining Welds in Piping IWC-2500, Examination and Pressure Test Requirements Table IWC-2500-1, Examination Categories Class 2 Piping Welds Category C-F-I, Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping Category C-F-2, Pressure Retaining Welds in Carbon or Low Alloy Steel Piping

4. Reason for Request

On April 9, 2004, Nuclear Regulatory Commission (NRC) staff approved FENOC's ASME Code Section XI Class 1 and Class 2 Risk-Informed Inservice Inspection (RI-ISI)

Program for Beaver Valley Power Station Unit No. 2 (BVPS-2), second ISI interval.

In its approval, NRC staff concluded the RI-ISI program is consistent with WCAP-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, and is an acceptable alternative to the requirements of ASME Code ,Section XI, for inservice inspection of ASME Class 1 and 2 piping, examination categories B-F, B-J, C-F-i, and C-F-2.

RI-ISI-2 Page 2 of 5 On May 1, 2006, NRC staff approved the Pressurized Water Reactor (PWR) Owners Group Topical Report WCAP-14572, Revision 1-NP-A, Supplement 2. The safety evaluation report states:

"The NRC staff concludes that the proposed RI-ISI program as described in the approved WCAP-14572, and WCAP-14572, Sup. 2, as clarified and revised by the June 22, 2005, supplemental letter, will provide an acceptable level of quality and safety with regard to the number of inspections, locations of inspections, and methods of inspection."

Consistent with the RI-ISI methodology documented in WCAP-14572, including its supplements, [References 2, 3, 4], new information has been incorporated into the RI-ISI analysis as part of the "living" RI-ISI program. The new information includes changes to the BVPS-2 Probabilistic Risk Assessment (PRA) model, revised segments and failure probabilities for some segments based on industry and plant experience and plant modifications, revised consequences based on lessons-learned, and updated test intervals for certain segments and overlays of pressurizer alloy 82/182 welds in the reactor coolant system.

The changes described above required re-performing the risk evaluation. The revised results were reviewed by the RI-ISI expert panel. Compared to the second interval BVPS-2 ISI program, 48 low safety significant (LSS) segments were reclassified as high safety significant (HSS), and 6 HSS segments were reclassified as LSS: 7 quantitative HSS segments were re-categorized by the expert panel as LSS based on the "with operator action consequences" guidance within WCAP-14572, Supplement 2. The expert panel concluded the remaining segment classifications shall remain as-is.

The change in risk evaluation was performed again to compare the original Section XI program with the revised third interval RI-ISI program for BVPS-2. Three reactor coolant system segments and six safety injection system segment (nine total segments) are to be added to the BVPS-2 RI-ISI program to meet the change in risk criteria discussed in WCAP-14572, page 214. These nine additional examinations are VT-2 visual exams.

No examinations were added for defense-in-depth considerations, which is the same as in the previously approved second interval RI-ISI program.

The proposed RI-ISI program, with updates, provides a 78 percent reduction in required examinations. This directly results in reduced outage scope, decreased individual and cumulative occupational radiation exposure, and shortened outage durations. As such, FENOC requests that the BVPS-2 RI-ISI program, with updates, be approved for continued use during the third ISI interval.

5. Proposed Alternative and Basis for Use ASME Section XI categories B-F, B-J, C-F-i, and C-F-2 contain the requirements for examining Class 1 and 2 piping components via non-destructive examination (NDE).

The proposed alternative [continued use of the BVPS-2 RI-ISI program, with updates] is limited to ASME Class 1 and 2 piping components, including piping currently exempt from NDE requirements. The proposed alternative will be substituted for the ASME Section XI category B-F, B-J, C-F-I, and C-F-2 examination requirements.

RI-ISI-2 Page 3 of 5 The applicable aspects of ASME Section XI Code not affected by the proposed alternative will be retained.

The basis of the alternative risk-informed inservice inspection program's methodology is fully described in the NRC-endorsed WCAP-14572 and its supplements.

Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative to the ASME Code Section X1 examination requirements will continue to provide an acceptable level of quality and safety.

Comparisons of the ASME Section XI inspection program, the second interval RI-ISI program, and the proposed third interval RI-ISI program are presented in Table 1.

6. Duration of Proposed Alternative The proposed alternative shall be implemented during the BVPS-2 third ten-year ISI interval and will remain effective until the end of the interval on August 28, 2018.

7. Precedents NRC letter to FENOC, April 9, 2004,

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) - Risk-Informed Inservice Inspection (RI-ISI) Program.

[ADAMS Accession Number ML040780805]

NRC letter to Tennessee Valley Authority, April 30, 2007,

Subject:

Sequoyah Nuclear Plant, Units 1 and 2 - Risk-Informed Inservice Inspection Program for the Third 10-Year Intervals.

[ADAMS Accession Number ML071070248]

8. References

1. Nuclear Regulatory Commission (NRC), "Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report WCAP-14572, Revision 1-NP-A, Supplement 2, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report Clarifications' Pressurized Water Reactor (PWR) Owners Group Project No. 694," May 1, 2006.

[ADAMS Accession No. ML061160035]

2. Westinghouse Electric, WCAP-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, February 1999.

3. Westinghouse Electric, WCAP-14572, Supplement 1, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection," Revision 1-NP-A, February 1999.

4. Westinghouse Electric, WCAP-14572, Supplement 2, "Pressurized Water Reactor Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report Clarifications," Revision 1-NP-A, September 2006.

RI-ISI-2 Page 4 of 5 Table 1 BVPS-2 STRUCTURAL ELEMENT SELECTION RESULTS AND COMPARISON TO ASME SECTION XI 1989 EDITION REQUIREMENTS Degradation Safety ASME Total ASME XI Second Interval RI-ISI Third Interval RI-ISla System High Safety Significant Mechanism(s) Class Code Weld Count Program Segments Exam [Welds requiring Examinations (Qty. of HSS in Category Volumetric (Vol)

Augmented and Surface (Sur)]

Program / Total Vol and Sur only Vol and Sur SES Matrix Number of SES Matrix Number of Qty. of Segments Sur Sur only Region . Exam Region Exam in Aug. Program) Locations Locations BDG 3 (3 / 24c) FAC/TF Class 2 N/A 0 0 0 0 3 0 1, 3 3_+_3_

CHS 34(0/0) TF/VF, TF Class 1 B-J 4 369 3 57 2,3,4 0 1,2,3,4 0 Class 2 C-F-1 343 315 26 27 19 + 14 b 19 + 15 Cl 4(0/0) TF/SCC, TF Class 2 N/A 0 0 0 0 4 0 2,4 4 DAS 0(0/0) TF Class 1 B-J 0 36 0 24 4 0 4 0 FWA 0 (0 / 57c) FAC/TF Class 2 C-F-2 56 0 9 0 3 0 3 0 GNS 0(0/0) TF Class 2 N/A 0 0 0 0 4 0 4 0 HCS 0(0/0) TF Class 2 N/A 0 0 0 0 4 0 4 0 MSS 15 (9/44c) FAC/TF, TF Class 2 C-F-2 136 3 17 0 1,3 8 1,2,3,4 12+ 3 g QSS 15(0 /0) TF, VF Class 2 C-F-1 200 0 16 0 1,2, 4 15+40 2,4 17727+49 RCS 30(6/ 7e) SCC/TF, Class 1 B-F 18 0 18 0 2,4 26 + 2d 1,2,4 30+3 SCC/TF/VF/SS, Class 1 B-J 217 350 57 136 TF RHS 1 (0/0) TF/SCC, TF Class 1 B-J 22 6 7 2 2,4 1 2,4 1 Class 2 C-F-1 283 0 23 0 0 0 RSS 9(0/0) TF Class 2 C-F-1 199 0 16 0 4 0 2,4 9 SIS 47(0/0) TF Class 1 B-J 222 157 43 14 2,4 0 2,4 12+3 Class 2 C-F-1 934 200 71 17 19 + 5) 21 + 2'+ 1 4g

+ d SSR 0(0/0) TF Class 1 N/A 0 0 0 0 4 0 4 0 Class 2 N/A 0 0 0 0 0 0 FAC/TF, TF, 27 NDE + 43 NDE +

SCCITF, Class 1 483 918 128 233 2 VISUAL 9 VISUAL TOTAL 158 (18 / 132) SCC/TF/VF/SS, TF/ VF, VF 61 NDE + 83 NDE +

Class 2 2151 518 181 44 25 VISUAL 43 VISUAL Total 2634 1436 309 277 88 NL- + 1265 NUD +

27 VISUAL 52 VISUAL A I A fl ____________ -

RI-ISI-2 Page 5 of 5 Table 1 BVPS-2 STRUCTURAL ELEMENT SELECTION RESULTS AND COMPARISON TO ASME SECTION XI 1989 EDITION REQUIREMENTS Summary: Prior ASME Section XI selects a total of 586 welds while the proposed RI-ISI program selects a total of 126 welds (plus 52 visual exams), which results in a 78% reduction.

Degradation Mechanisms:

FAC - Flow-Assisted Corrosion SCC - Stress Corrosion Cracking SS - Striping/Stratification TF - Thermal Fatigue VF - Vibratory Fatigue "X/X" indicates combination of mechanisms.

Systems:

BDG - Steam Generator Blowdown System CHS - Chemical and Volume Control System Cl - Containment Isolation System DAS - Reactor Plant Drains and Vents Systems FWA - Steam Generator Feedwater System GNS - Gaseous Nitrogen System HCS - Hydrogen Control System MSS - Main Steam System QSS - Quench Spray System RCS - Reactor Coolant System RHS - Residual Heat Removal System RSS - Recirculation Spray System SIS - Safety Injection System SSR - Sampling System Notes for Table 1

a. System pressure test requirements and VT-2 visual examinations shall continue in all ASME Code Class systems.

b. VT-2 examination at one location within segment.

c. Augmented program for erosion-corrosion continues.

d. Examinations added for change in risk considerations (total of nine segments - three RCS and six SIS).

e. Augmented program for alloy 82/182 welds continue.

f. VT-2 examination on socket welded portion of segment.

g. VT-2 visual examination for entire segment.

FirstEnergy Nuclear Operating Company Beaver Valley Power Station Unit Nos. 1 and 2 Probabilistic Risk Assessment Technical Adequacy for Risk-Informed Inservice Inspection Page 1 of 28 Summary Statement of Beaver Valley Power Station Unit No. 1 (BVPS-1) and Beaver Valley Power Station Unit No. 2 (BVPS-2) Probabilistic Risk Assessment Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Introduction FirstEnergy Nuclear Operating Company (FENOC) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the probabilistic risk assessment (PRA) models for all operating FENOC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach, as it applies to the BVPS-1 and BVPS-2 PRA models.

PRA Maintenance and Update The FENOC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the FENOC risk management program, which consists of a governing procedure and subordinate implementation procedures. These procedures delineate the responsibilities and guidelines for updating the full-power internal events PRA models at all operating FENOC nuclear generation sites and delineate the responsibilities and guidelines for use of the PRA models in applications.

The overall FENOC risk management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files.

To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

• Design changes and procedure changes are reviewed for their impact on the PRA model.

" New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

" Unavailability due to maintenance is captured, and the impact on core damage frequency (CDF) is trended.

• Plant-specific initiating event frequencies, failure rates, and unavailability due to maintenance is updated approximately every three years.

In addition to these activities, FENOC risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

• The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full-power, internal events PRA models for FENOC nuclear generation sites.

Page 2 of 28 Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65 (a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur every three years, although longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant. FENOC performed a regularly scheduled update to the BVPS-1 PRA model in 2006 and BVPS-2 PRA model in 2007, and is currently in the process of updating the BVPS-1 PRA model.

PRA Self Assessment and Peer Review Several assessments of technical capability have been made, and continue to be planned, for the BVPS-1 and BVPS-2 PRA models. These assessments are as follows:

" An independent PRA peer review [Reference 1] was conducted in 2002 under the auspices of the Westinghouse Owners Group (WOG), following issuance of the industry PRA peer review process guidance [Reference 2]. This peer review included an assessment of the PRA model maintenance and update process.

" During 2005, the BVPS-1 and BVPS-2 PRA model results were evaluated in the WOG PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process. Results of this cross-comparison are presented in WCAP-16464-NP [Reference 3]. Notably, after allowing for plant-specific features, there are no MSPI cross-comparison outliers for BVPS-1 or BVPS-2.

" In 2007, a gap analysis [Reference 4] was performed against the ASME PRA Standard [Reference 5] and Regulatory Guide 1.200, Revision 1 [Reference 6].

• Follow-up peer review [Reference 7] of the human reliability analysis (HRA) element, following the industry follow-on PRA peer review process [Reference 8],

was performed in 2007 to evaluate the change in HRA methodology since the 2002 WOG Peer Review.

• As part of the transition to NFPA-805, an independent PRA peer review

[Reference 9] was conducted in January 2009 of the fire PRA model under the auspices of the Pressurized Water Reactor Owners Group (PWROG), following the industry PRA peer review process [Reference 10]. This peer review included an assessment of the PRA model maintenance and update process for both BVPS-1 and BVPS-2.

A summary of the disposition of the 2002 industry PRA peer review facts and observations for the BVPS-1 and BVPS-2 models are documented within FENOC's Corrective Action Program.

The resolutions were reviewed and documented in the 2007 gap analysis report, Table A-3

[Reference 4].

A gap analysis for BVPS-2 [Reference 4] and HRA follow-up peer review for the 2006 BVPS-1 and 2007 BVPS-2 PRA models [Reference 7] was performed. These evaluations were performed against the ASME PRA Standard [Reference 5] and Regulatory Guide 1.200, Revision 1 [Reference 6]. The gap analysis identified 67 supporting requirements with potential gaps to Capability Category II of the Standard. The HRA follow-up review identified 10

Page 3 of 28 supporting requirements that did not meet Capability Category II requirements primarily in the analysis for pre-initiator human actions.

The gap analysis [Reference 4] documented 55 facts and observations that were written against the 67 supporting requirements with potential gaps to Capability Category II. Of these 55 facts and observations, 48 were considered to be documentation issues. Of the remaining seven facts and observations that were considered PRA modeling issues, five were categorized as Capability Category I, which was deemed an acceptable categorization for this application. The last two were categorized as not meeting the supporting requirements. However, these were written against the internal flooding PRA analysis, which was not used directly to support the development of the RI-ISI program, but instead was used as one source for identifying potential flooding sources and possible equipment affected by flooding.

In the HRA follow-on peer review [Reference 7], five facts and observations were written against the ten supporting requirements with potential gaps to Capability Category II. Of these five, all were considered to be documentation issues.

Table1 lists the status and importance to the Risk-Informed Inservice Inspection of these 60 facts and observations that had potential gaps in meeting Capability Category II of the ASME PRA Standard. The gaps pertaining to the internal flooding analysis, fire analysis, HRA and large early release frequency (LERF) analysis will be addressed during the model update process that is ongoing. Specifically, the analysis updates for LERF and pre-initiator human actions will be integrated into the BVPS-1 PRA model update; the analysis updates for flooding and fire will be integrated into the BVPS-1 Level 1 internal events model update; and the analysis updates for flooding, LERF and pre-initiator human actions will be integrated into the BVPS-2 PRA model update. The analysis updates for fire will be integrated into the BVPS-2 level 1 internal events model update that is scheduled to occur in 2011. The other remaining gaps will be reviewed for consideration during the PRA model update process, but are judged to have low impact on the PRA model and its ability to support a full range of PRA applications.

The remaining gaps are documented, so they can be tracked and accounted for in applications where appropriate.

General Conclusion Regarding PRA Capability The BVPS-1 and BVPS-2 PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions. As specific risk-informed PRA applications are performed, remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

Assessment of PRA Capability Needed for Risk-Informed Inservice Inspection In the risk-informed inservice inspection (RI-ISI) program at BVPS-1 and BVPS-2, the PWROG RI-ISI methodology [Reference 11] is used to define alternative inservice inspection requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking and delta risk evaluation steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three processes in the PWROG methodology.

Page 4 of 28

" In the PWROG methodology two sets of consequences are developed based on the operators taking no action to isolate or mitigate the piping failure and based on the operators being perfect in taking action to isolate or mitigate the piping failure, ifthere is a credible operator action. Based on this, four risk evaluation workbooks are created for core damage frequency (CDF) and large early release frequency (LERF). If the risk metrics from any of these four risk evaluation workbooks are quantitatively high safety significant (HSS), the segment is identified as quantitatively HSS.

" A simplified uncertainty analysis is performed to ensure that no low safety significant segments could move into high safety significance when reasonable variations in the pipe failure and conditional CDF/LERF probabilities are considered.

" The PWROG RI-ISI methodology is a risk-informed process and not a risk-based process. The quantitative results from the risk evaluation along with deterministic insights and other input data are presented to an expert panel in an integrated decision making process. The primary focus of the expert panel is to review all pertinent information and determine the final safety-significance category for each of the piping segments. The expert panel is comprised of plant personnel with a wide breadth and depth of experience as specified in WCAP-1 4572 [Reference 11]. Segments that have been determined to be quantitatively HSS are typically categorized as HSS by the expert panel. The focus of the expert panel is to add segments to the higher classification. The BVPS-1 and BVPS-2 expert panel categorized 53 BVPS-1 segments and 59 BVPS-2 segments as HSS that were not quantitatively HSS based on deterministic insights, high failure potential and/or high consequences. Additionally, as part of the integrated decision making process the expert panel considers limitations in the process when categorizing segments as HSS or LSS. This may include PRA model limitations and limitations in modeling the consequences using the PRA model.

The limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174 [Reference 12].

Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:

There are, however, some applicationsthat, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

An example is risk-informed inservice inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodicallyexamined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary. Therefore, the staff review of plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.

In the PWROG RI-ISI process the PRA model is not used as the basis for the risk evaluation, but instead is used as an input to the risk evaluation process. The vast majority of the piping failure consequences are identified as loss of a system or train of a system. The PRA results are then used as an input to the risk evaluation for the relative ranking of the segments. Table 1.3-1 of the ASME PRA Standard1 [Reference 8] identifies the bases for PRA capability 1 Table A-1 of Regulatory Guide 1.200 identifies the NRC staff position as "No objection" to Section 1.3 of the ASME PRA Standard, which contains Table 1.3-1.

Page 5 of 28 categories. The bases for Capability Category I for scope and level of detail attributes of the PRA states:

Resolution and specificity sufficient to identify the relative importance of the contributors at the system or train level including associatedhuman actions.

Based on the above, in general, Capability Category I should be sufficient for PRA quality for a RI-ISI application.

In addition to the above, it is noted that segments and their associated welds determined to be low risk significant are not eliminated from the ISI program on the basis of risk information. For example, the risk significance of a segment may be determined by the expert panel to be low safety significant, resulting in it not being a candidate for inspection. However, it remains in the program and, if in the future the assessment of its ranking changes (either by damage mechanism, PRA risk, or deterministic insight), then it can again become a candidate for inspection. If it is discovered during the RI-ISI update process that a segment is now susceptible to flow-accelerated corrosion (FAC), inter-granular stress corrosion cracking (IGSCC), or microbiological induced cracking (MIC), it is addressed in an augmented program where it is monitored for those special damage mechanisms. That occurs no matter what the risk ranking of the segment or weld is determined to be.

Conclusion Regarding PRA Capability for Risk-Informed ISI The BVPS-1 and BVPS-2 PRA models continue to be suitable for use in the RI-ISI application.

This conclusion is based on:

• the PRA maintenance and update processes in place,

• the PRA technical capability evaluations that have been performed and are being planned, and

• the RI-ISI process considerations, as noted above, that demonstrate the relatively limited reliance of the process on PRA capability.

References

1. Westinghouse Electric Company, "Beaver Valley Power Station PRA Peer Review Report, Final Report," December 2002.

2. Nuclear Energy Institute, NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision A3, March 2000.

3. Westinghouse Electric, WCAP-16464-NP, "Westinghouse Owner's Group Mitigating Systems Performance Index Cross Comparison," Revision 0, August 2005.

4. "Summary of Beaver Valley Power Station Unit 2 PRA Regulatory Guide 1.200 App. B /

ASME PRA Standard 'Gap' Assessment," attached to LTR-RAM-I-08-016, January 2008.

5. American Society of Mechanical Engineers, ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," New York, New York, December 2005.

6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007.

Page 6 of 28

7. "Focused Peer Review of the Human Reliability Analysis Against the ASME PRA Standard Requirements for the Beaver Valley Power Station Probabilistic Risk Assessment," attached to LTR-RAM-II-08-006, March 2008.

8. Nuclear Energy Institute, NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard (Internal Events)," Revision 1 (Draft), November 2007.

9. "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Beaver Valley Unit 1 Fire Probabilistic Risk Assessment," LTR-RAM-II-09-006, April 2009.

10. Nuclear Energy Institute, NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Draft Version E, November 2008.

11. Westinghouse Electric, WCAP-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, February 1999, including:

Supplement 1, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection," Revision 1-NP-A, February 1999.

Supplement 2, "Pressurized Water Reactor Owners Group Application of Risk Informed Methods to Piping Inservice Inspection Topical Report Clarifications," Revision 1-NP-A, September 2006.

12. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.

Page 7 of 28 Table 1 - Status of Open Gaps to Capability Category IIof the ASME PRA Standard2 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection IE-A6-01 There is no documentation of interviews of plant IE-A6 Open - Plan is to document None. Capability Category I is met and personnel (for example: operations, interviews of plant personnel that- appropriate for this application.

maintenance, engineering, safety analysis) to determined if potential initiating determine if potential initiating events have events have been overlooked. This gap is a documentation been overlooked. This is required to meet consideration only.

Capability Category II.

IE-C9-01 Plant-specific information used in the IE-C9 Open - Will document the None. This gap is a documentation assessment and quantification of recovery assessment and quantification of consideration only.

actions included in the support system initiating any recovery actions assumed in event analysis is not included in the support the support system initiating event system notebooks. Analysis of the recovery analysis. If no recovery actions actions should be consistent with the applicable are used or modified, also note requirements in the human reliability analysis. that in the documentation.

IE-C10-01 There is no comparison of the initiating event IE-C10 Partially Resolved - A comparison None. This gap is a documentation analysis with generic data sources or was made between the BVPS consideration only.

explanation of differences to provide a initiating event data and the WOG reasonableness check of the results. initiating event database and NUREG/CR-5750 values. Any outliers have justification provided.

SC-A5-01 This supporting requirement requires that for SC-A5 Open - Additional evaluations or None. Capability Category I is met and sequences in which stable plant conditions modeling by using an appropriate appropriate for this application.

would not be achieved by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using the technique will be performed for modeled plant equipment and human actions, sequences in which stable plant This gap is a documentation perform additional evaluation or modeling by conditions would not be achieved consideration only.

using an appropriate technique. by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using the modeled plant equipment and human The makeup to refueling water storage tank actions.

(MU) top event for medium loss of coolant accident (LOCA) and small LOCA/general For top event MU, document that transient uses refueling water storage tank the plant conditions reach (RWST) makeup as part of the success path acceptable stable values and that when recirculation has failed. While a mission using the analyzed RWST makeup time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is assumed, the plant is not at flow rate would not result in 2 The gap analysis is conducted independently of RI-ISI and is based on comparing the PRA model against the supporting requirements of ASME PRA standard at Capability Category II. Many of the identified gaps are not applicable to RI-ISI since in general Capability Category I is sufficient. For completeness, all current gaps that do not meet at least Capability Category II are identified in Table 1.

Page 8 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection a safe stable state because another action is containment flooding issues that required for long term success. The RWST would impact any equipment or refill results in additional water to the instrumentation important for containment which eventually will result in the mitigating the accident. Use the design basis flooding level being exceeded and containment water level and the potential for subsequent loss of volume SAMG CA-5 for guidance instrumentation and control. The impact of on what equipment and continued RWST makeup and injection into instrumentation could become containment needs to be discussed in relation submerged based on the RWST to the achievement of a safe stable state where makeup flow rate.

no additional operator actions are required.

A similar situation exists for steam generator tube rupture (SGTR) and interfacing system LOCA (ISLOCA) where RWST refill is being used to maintain core cooling, but the justification for mission time of only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not apparent given that the plant is not in a safe stable state by traditional definitions.

SC-A5-02 The success criteria for top event makeup to SC-A5 Open - Will provide justification for None. Capability Category I is met and RWST given leakage through secondary (WM) 400 gpm success criteria for top appropriate for this application.

for the SGTR states that 400 gallons per minute event WM to maintain HHSI for (gpm) makeup to the RWST is sufficient to RCS inventory control at full RCS This was judged not to impact PRA maintain high head safety injection (HHSI) for pressure despite leakage through results and is not required to meet reactor coolant system (RCS) inventory control a ruptured SG tube. Also noting SC-A5. This is expected to be a at full RCS pressure, despite leakage through a that by the time makeup is clarification of the use of these success ruptured steam generator (SG) tube. required the RCS would not be at criteria and is therefore assigned a full RCS pressure due to the Level C.

The maximum RCS inventory loss through a breach in the SG tube.

single SGTR is approximately 600 gpm if the This gap is a documentation primary side is at normal operating pressure consideration only.

and the secondary side of the SG is not depressurized. This is in excess of the 400 gpm makeup and therefore appears to invalidate the success criteria as stated. Also, if continued HHSI at full system pressure is required, SG overfill is likely to occur and the SG will be depressurized and the leakage through the ruptured tube will even be higher.

Page 9 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection SC-C1-02 The ASME PRA standard for SC-Cl requires SC-Cl, Open - Will gather all (system) None. This gap is a documentation that success criteria be documented in a SY-Ci success criteria in the success consideration only.

manner that facilitates applications, upgrades, criteria Notebook to facilitate future and peer reviews. The current state of the usage.

BVPS PRA success criteria is that the accident sequence success criteria are gathered in the success criteria notebook, but other success criteria are scattered throughout the PRA.

Examples include the SW success criteria and ISLOCA success criteria for BVPS-1. It is recommended that FENOC consider gathering all success criteria in the success criteria notebook to facilitate future usage.

SC-C2-01 No discussion of the limitations of the modular SC-C2 Open - Will add a discussion of None. This gap is a documentation accident analysis program (MAAP) code for MAAP limitations (similar to the consideration only.

success criteria are provided in the success EPRI assessment for MAAP 3) to criteria notebook. Two known limitations are be documented or referenced in the use of MAAP for early phase large LOCAs the success criteria notebook. Also and the use of MAAP for steam generator reference the MAAP users guide dryout assessments without benchmarking to for additional info.

design basis codes (for example, bleed and feed initiation). It was observed in the success criteria notebook that MAAP runs were made to justify only one accumulator (but that two of two intact accumulators appear to have been actually used as stated to be used in section 3.1 of the notebook). It is recommended that a discussion of MAAP limitations (similar to that provided in the EPRI assessment for MAAP 3) be documented or referenced in the success criteria notebook.

SY-A14-01 The draft revision 4 system notebooks for SY-A12, Open - Will add a discussion for None. This gap is a documentation auxiliary feedwater, service water, component SY-A14, the excluded failure modes and consideration only.

cooling water secondary side, component SY-C1 contributors to system cooling water primary side, and main feedwater unavailability and unreliability.

were reviewed. Discuss failure modes and However, it is unlikely that these contributors to system unavailability and contributors will significantly unreliability that are excluded from the systems impact PRA results.

analysis. However, a SY-A14 criterion does not appear to have been applied consistently

Page 10 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection throughout the analysis. The only exceptions found where the SY-A14 criteria are explicitly met is in the CCS notebook, section 14, c, assumption 2, and the AFW notebook section 14, c, assumption 3. In some instances, such as the CCP notebook section 14, c, assumption 1, there was no explanation given for why the contributor was not modeled.

SY-C1-01 In providing the response to peer review DA-09, .SY-C1 Open - Will add brief summary of None. This gap is a documentation which deals with providing documentation of the the CCF group selections, possibly consideration only.

common cause failure (CCF) groupings, as part of the system notebook, FENOC noted that the systems analysis section 15 "Common Cause".

overview and guidance notebook provides the process used to identify CCF groupings. The response further suggests details of the common cause groups that were retained in the PRA system models and presented in appendix C of the BVPS-2 PRA system notebooks, under the common cause sections of the risk management software program (RISKMAN)

System notebook files are adequately documented and can be found by knowledgeable personnel.

The reviewer agrees that one can review Appendix C of the systems notebooks and see what the CCF groupings are and how the CCF probabilities were generated. The reviewer also agrees that high level guidance is provided in the systems analysis overview and guidance notebook. However, it appears a link between the two documents is missing.

For example, the guidance states "When identical, non-diverse, and active components are used to provide redundancy, they should be considered for assignment to common cause groups, one group for each identical redundant component." When the systems notebook appendix C is reviewed, the components

Page 11 of 28 Facts and Description of Gap Observations IIRequirementý Supportingt JCurrent Status or Comment Importance to Risk-Informed Inservice Inspection contained in the CCF group are clearly identified, but there is no documentation that states that those components are "identical, and/or non-diverse" or used to provide redundancy.

Further examination of other sections of the system notebooks (such as section 3 "System Success Criteria", or section 6 "Operating Features", would lead a reviewer to find this type of information. But this documentation is not always intuitively obvious and makes peer review difficult at times.

SY-C1-02 The BVPS-2 system notebooks have no SY-C1 Partially Resolved -- In the process None. This gap is a documentation indication of system engineering reviews, of documenting system notebooks consideration only.

These reviews help ensure that systems are reviews by system engineering.

modeled in accordance with day-to-day plant operations and additionally expand the PRA knowledge of the system engineers.

HR-B1-01 This is a carry-over from the HR-2 peer review. HR-B1, Open - Will calculate specific None. Capability Category I is met and HR-D2 misalignment error of omission appropriate for this application. Refer to A generic error of omission term from the failure probabilities for important the section "Assessment of PRA Pikard, Lowe, and Garrick, Inc. (PLG) database systems using the EPRI human Capability Needed for Risk-Informed (ZHEO1A) was used for all misalignment reliability analysis (HRA) Inservice Inspection."

human error probabilities without regard for calculator.

procedural or operational failure barriers such It is not expected that the BVPS specific as independent verification, peer checks, misalignment values will be significantly walkdowns, etc. However, plant-specific data different from the generic values used.

was used for test and maintenance frequencies.

Therefore, the overall misalignment errors were a hybrid of generic and plant-specific data. This was used for systems which are important to CDF (for example, auxiliary feedwater and safety injection).

HR-D3-01 While the discussion in the system notebooks HR-D3 Open - Will confirm and document None. Capability Category I is met and (auxiliary feedwater, quench spray, and that the procedure quality is appropriate for this application.

recirculation spray notebooks were reviewed) sufficient to support the crew references the procedures, no documentation response within the times This gap is a documentation of quality of those procedures or administrative assigned in the PRA evaluation, consideration only.

controls was found.

Page 12 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection HR-I1-01 The BVPS-2 system and data notebooks have HR-I1, HR-12 Partially Resolved - Complete the None. This gap is a documentation been updated and exist in draft form, but there update of the PRA analysis and consideration only.

is no record of formal review and approval, system notebooks with formal Furthermore, only a subset of the total PRA review and approval.

notebooks have been updated for this revision of the PRA.

HR-12-01 The BV human reliability analysis does HR-12 Open - Will document the process None. This gap is a documentation document a process to perform a systematic used to perform a systematic consideration only.

search for dependent human actions credited search for dependent human on individual sequences. It is clear from the actions credited on individual human action identifier sheets documented in sequences.

the BVPS-2 human reliability analysis notebook that such an evaluation has been performed, but there is no evidence of the process documented in the human reliability analysis notebook.

To be consistent with current human reliability analysis methods, there must be a systematic process to identify, assess and adjust dependencies between multiple human errors in the same sequence, including those in the initiating events.

HR-12-02 There is no evidence in the human reliability HR-12 Open - During the recent extended None. This gap is a documentation analysis or success criteria notebooks that an power uprate evaluation, plant consideration only.

operator review of the human reliability analysis operations did review the operator has been performed. actions and timings. There are reports to document these reviews

[See Note 21. Furthermore, several operator action scenarios were evaluated using the plant simulator. The results of the review of operator actions will be incorporated into the human reliability analysis notebook or success criteria notebook.

HR-13-01 The human reliability analysis notebook IE-D3, Open - Will document all of the None. This gap is a documentation sporadically discusses assumptions and AS-C3, human reliability analysis consideration only.

uncertainties. Per the clarification to Regulatory SC-Cl, assumptions and uncertainties into Guide 1.200, Revision 1, there is an increased SC-C3, a new "Assumptions and

Page 13 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection importance in the industry to identify HR-l1, HR-13, Uncertainties" section in the assumptions and uncertainties in the PRA IF-F3, human reliability analysis model. In reviewing the human reliability QU-F4, notebook.

analysis notebook, it is difficult to locate the LE-F3, assumptions and uncertainties. LE-G4 Also, the quantification notebook lists an evaluation of the model uncertainties; however, a more comprehensive set of assumptions and uncertainties will be documented.

DA-C4-01 A clear basis for the identification of events as DA-C4 Partially Resolved - The None. This gap is a documentation failures is not included in the data analysis methodology for a clear basis to consideration only.

notebook. This basis could be used to identify events as failures was distinguish between those degraded states for interpreted from NUREG/CR-2823 which a failure, as modeled in the PRA, would and is documented in a draft copy have occurred during the mission and those for of the data analysis notebook.

which a failure would not have occurred (for This provides the basis for example, slow pick-up to rated speed). guidance to distinguish between degraded states for which a failure It could not be determined from the data is modeled in the PRA.

analysis notebook if any failures were screened out or if the maintenance rule maintenance preventable functional failures are used as the data source.

DA-C5-01 There is no listing or description in the data DA-C5 Open - Will document a listing or None. This gap is a documentation analysis notebook of repeated component description of repeated component consideration only.

failures that were counted as a single failure. failures that were counted as a single failure in the data analysis Repeated component failures occurring within a notebook.

short time interval should be counted as a single failure if there is a single, repetitive problem that causes the failures. In addition only one demand should be counted.

DA-C8-01 Plant records should be used and documented DA-C8 Open - Will use plant records to None. Capability Category I is met and to determine the time that components are determine and document the time appropriate for this application.

configured in their standby status. This is that components are configured in required to change DA-C8 from Capability their standby status. This gap is a documentation Category I to Il1. consideration only.

Page 14 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection DA-C10-01 Decompose failure modes into sub-elements DA-C1 0 Open - If the component failure None.. Capability Category I is met and and count demands and failures individually in mode is decomposed into sub- appropriate for this application. Refer to the sub-elements. elements that are fully tested, will the Section "Assessment of PRA review test procedures to ensure Capability Needed for Risk-Informed that tests that exercise specific Inservice Inspection."

sub-elements are used for their evaluation. There are only a limited number of component failure modes that are decomposed into sub-elements, no significant impact is expected.

IF-Ala-01 It is not clear from the documentation that a IF-Ala Open - The internal flooding None. Capability Category I is met and comprehensive assessment has been analysis gaps will be addressed appropriate for this application.

conducted to finalize the combined rooms during the internal flooding model including propagation, barriers, etc. The update that is scheduled for 2010. In addition, the internal flooding PRA internal flooding assessment is based on large This flooding analysis update will analysis was not used directly to support flood areas but there is no description of the then be integrated into the level 1 the development of the RI-ISI program, process used to define those areas with respect internal events model update. but instead was used as one source for to flood propagation and barriers, identifying potential flooding sources and possible equipment affected by flooding.

This gap is a documentation consideration only.

IF-A3-01 There is no evidence in the internal flooding IF-A3 Open - The internal flooding None. The internal flooding PRA notebook that it represents the current as-built, analysis gaps will be addressed analysis was not used directly to support as-operated plant. Revision 4 documentation in during the internal flooding model the development of the RI-ISI program, another document may include the information update that is scheduled for 2010. but instead was used as one source for to show that the internal flooding assessment is This flooding analysis update will identifying potential flooding sources and current, but it is not in this notebook. IF-A3-01 then be integrated into the level 1 possible equipment affected by flooding.

was written as a B level fact and observation to internal events model update.

provide documentation that the internal flooding This gap is a documentation assessment still represents the as-built as- consideration only.

operated plant in 2007. This probably also applies to other PRA elements from the ASME PRA standard (for example, system analysis, success criteria, human reliability analysis, etc.)

and should be addressed generically for the BVPS PRA. This would facilitate future reviews and development of PRA applications.

Page 15 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection IF-Bl-01 The ASME PRA standard states "for each flood IF-B1 Open - The internal flooding None. The internal flooding PRA area, identify the potential sources of flooding." analysis gaps will be addressed analysis was not used directly to support Section C3.1 identifies flood sources in each during the internal flooding model the development of the RI-ISI program, area but clear documentation of each source in update that is scheduled for 2010. but instead was used as one source for an area is lacking. The standard expects a This flooding analysis update will identifying potential flooding sources and more systematic approach for identifying then be integrated into the level 1 possible equipment affected by flooding.

potential flood sources and then later screening internal events model update.

them. The internal flooding assessment here This gap is a documentation includes initial screening without written consideration only.

justification. It is suggested that a complete discussion of potential sources be documented and used as the basis for screening potential sources.

IF-B1-02 Section C3.1 states that major flood sources IF-B1 Open - The internal flooding None. The internal flooding PRA were reviewed to identify potential flood analysis gaps will be addressed analysis was not used directly to support locations. The ASME standard suggests that during the internal flooding model the development of the RI-ISI program, first you identify flooding areas then identify all update that is scheduled for 2010. but instead was used as one source for flooding sources in that area. This method This flooding analysis update will identifying potential flooding sources and used for BVPS may have led to overlooking then be integrated into the level 1 possible equipment affected by flooding.

other sources of flooding within each area. internal events model update.

This gap is a documentation consideration only.

IF-B2-01 B-2 of the PRA standard requires "For each IF-B2, IF-B3 Open - The internal flooding None. The internal flooding PRA source of flooding, identify the flooding analysis gaps will be addressed analysis was not used directly to support mechanisms that would result in a fluid release during the internal flooding model the development of the RI-ISI program, including failure models, human-induced update that is scheduled for 2010. but instead was used as one source for mechanisms, and other events resulting in a This flooding analysis update will identifying potential flooding sources and release into the flood area." In addition, B-3 then be integrated into the level 1 possible equipment affected by flooding.

requires "For each source and its identified internal events model update.

failure mechanism, identify the characteristic of This gap is a documentation release and the capacity of the source." consideration only.

Section C3.1 of the internal flooding notebook does not provide enough detail to judge whether this requirement is met. One example is that although a few human error induced floods (for example, testing or maintenance errors) were considered, there is no evidence of a systematic assessment of potential test and maintenance errors.

Page 16 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection IF-C2b-01 Section C3.1 does not have enough detail to IF-C2b Open - The internal flooding None. The internal flooding PRA show that the capacity of the drains and the analysis gaps will be addressed analysis was not used directly to support amount of water retained by the sumps, berms, during the internal flooding model the development of the RI-ISI program, dikes, and curbs was estimated. The reviewer update that is scheduled for 2010. but instead was used as one source for notes that it is likely that this was performed but This flooding analysis update will identifying potential flooding sources and there is no record of the assessment. The then be integrated into the level 1 possible equipment affected by flooding.

capacity of drains and the amount of water internal events model update.

retained by sumps, etc. should be documented This gap is a documentation in the internal flooding notebook. consideration only.

IF-C3-01 The PRA standard states "for each SSCs IF-C3, IF-C3a Open - The internal flooding None. The internal flooding PRA identified in IF-C2c identify the susceptibility of analysis gaps will be addressed analysis was not used directly to support each SSC in the flood area to flood-induced during the internal flooding model the development of the RI-ISI program, failure mechanism". Also, C3a states, "to update that is scheduled for 2010. but instead was used as one source for determine susceptibility of SSC to flood-induced This flooding analysis update will identifying potential flooding sources and failure mechanism, take credit for the operability then be integrated into the level 1 possible equipment affected by flooding.

of SSC identified in IF-C2c with respect to internal events model update.

internal flood impact only if supported by an This gap is a documentation appropriate combination of: 1) test or consideration only.

operational data, 2) engineering analysis, and

3) expert judgment." It is likely that flood-induced failure mechanisms were considered in the internal flooding assessment but are not identified in the internal flooding notebook.

Section C3.1 does not provide enough detail on the impact of the flood on SSCs.

IF-C3b-01 IF-C3b requires that all potential mechanisms IF-C3b Open - The internal flooding None. The internal flooding PRA that can create interconnections between analysis gaps will be addressed analysis was not used directly to support flooding areas be considered for Capability during the internal flooding model the development of the RI-ISI program, Category 11 (CCII) and that barrier unavailability update that is scheduled for 2010. but instead was used as one source for also be considered for Capability Category III This flooding analysis update will identifying potential flooding sources and (CCIII). There is no evidence in appendix C of then be integrated into the level 1 possible equipment affected by flooding.

the initiating events notebook that any internal events model update.

mechanism other than open obvious pathways This gap is a documentation (for example, vents in doors, tunnels, etc.) were consideration only.

considered. This may be just a documentation issue for CCII.

Also, the RI-ISI program did a comprehensive assessment of flooding potential for various break locations. A comparison should be

Page 17 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection performed between the RI-ISI flooding assessment and the PRA internal flooding assessment to ensure consistency.

Note that upgrading to CCIII requires the additional consideration of barrier unavailability, for example due to maintenance activities or maintenance unavailability.

IF-C3c-01 Develop engineering calculations for all flooding IF-C3c Open - The internal flooding None. The internal flooding PRA scenarios, not just the worst case scenarios, analysis gaps will be addressed analysis was not used directly to support This is likely just a documentation issue, but during the internal flooding model the development of the RI-ISI program, since it is missing from the internal flooding update that is scheduled for 2010. but instead was used as one source for notebook, IF-C3c is not met. This flooding analysis update will identifying potential flooding sources and then be integrated into the level 1 possible equipment affected by flooding.

internal events model update.

This gap is a documentation consideration only.

IF-C4-01 The operator actions credited in the internal IF-C4 Open - The internal flooding None. The internal flooding PRA flooding assessment are based on detailed analysis gaps will be addressed analysis was not used directly to support HRA assessments for two operator actions. during the internal flooding model the development of the RI-ISI program, Cues, procedures, etc. are detailed in the HRA update that is scheduled for 2010. but instead was used as one source for assessment. It is not clear if these actions are This flooding analysis update will identifying potential flooding sources and also applied to scenarios other than those used then be integrated into the level 1 possible equipment affected by flooding.

to quantify the human error probability in the internal events model update.

HRA notebook. In addition, there are a number This gap is a documentation of other instances in which the operators are consideration only.

assumed to be highly reliable. There is also no indication that these are validated by operator interviews. Cleaner documentation of the operator actions that are credited (as well as those not credited), and their basis, should be completed to assist in future reviews and for risk applications in which the performance of operators is important. Also a clear linkage between the internal flooding and HRA notebooks should be documented for the basis of the important HRA input and some of the operator actions to screen scenarios are based on highly reliable operator actions.

Page 18 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection IF-C4-02 IF-C4 requires the development of flood IF-C4 Open - The internal flooding None. The internal flooding PRA scenarios by examining the equipment and analysis gaps will be addressed analysis was not used directly to support relevant plant features in the flood area and during the internal flooding model the development of the RI-ISI program, area in potential propagation paths, taking update that is scheduled for 2010. but instead was used as one source for credit for appropriate flood mitigation systems This flooding analysis update will identifying potential flooding sources and or operator actions, and identifying susceptible then be integrated into the possible equipment affected by flooding.

SSCs. No flood scenarios are developed in the subsequent level 1 internal events internal flooding notebook. model update. This gap is a documentation consideration only.

IF-C5-01 The screening methodology documented in IF-C5, Open - The internal flooding None. The internal flooding PRA Section C3.1 does not follow the systematic IF-C5a, analysis gaps will be addressed analysis was not used directly to support methodology described in the standard. For the IF-C7, IF-D7 during the internal flooding model the development of the RI-ISI program, internal flooding assessment, the screening is update that is scheduled for 2010. but instead was used as one source for performed at the source and location level and, This flooding analysis update will identifying potential flooding sources and in some cases, without adequate basis as then be integrated into the possible equipment affected by flooding.

discussed in IF-B1-01. The method used in the subsequent level 1 internal events internal flooding assessment may be technically model update. This gap is a documentation adequate, if the basis is better documented, consideration only.

even though it does not meet the standard supporting requirements for C-5, C5a and C7.

IF-D1-01 The FENOC response to DE-06 from the IF-DI Open - The internal flooding None. The internal flooding PRA owners group peer review is incomplete. The analysis gaps will be addressed analysis was not used directly to support fact and observation is concerned about the during the internal flooding model the development of the RI-ISI program, vintage of the data used to estimate pipe break update that is scheduled for 2010. but instead was used as one source for frequencies and the FENOC response talks This flooding analysis update will identifying potential flooding sources and about walkdowns. then be integrated into the possible equipment affected by flooding.

subsequent level 1 internal events model update. This gap is a documentation consideration only.

IF-D1-02 The internal flooding assessment does not rely IF-D1 Open - The internal flooding None. The internal flooding PRA on grouping of initiating events, sources, analysis gaps will be addressed analysis was not used directly to support locations, etc. The screening methodology during the internal flooding model the development of the RI-ISI program, discussed in the internal flooding notebook and update that is scheduled for 2010. but instead was used as one source for assessed under the IF-C-xx supporting This flooding analysis update will identifying potential flooding sources and requirements methodology resulted in only a then be integrated into the possible equipment affected by flooding.

handful of flooding events to be considered. subsequent level 1 internal events These were individually assessed in the overall model update. This gap is a documentation PRA quantification using RISKMAN. The consideration only.

methodology used may be technically adequate in spite of not meeting the ASME standard

Page 19 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection supporting requirements for grouping if it can be justified that only a handful of events are important.

IF-D5-01 The internal flooding pipe and tank break IF-D5, IF-D5a Open - The internal flooding None. The internal flooding PRA frequencies used in the internal flooding analysis gaps will be addressed analysis was not used directly to support assessment are based on 1988 and 1990 data. during the internal flooding model the development of the RI-ISI program, The prior pipe break frequencies should be update that is scheduled for 2010. but instead was used as one source for updated to reflect more recent experience and This flooding analysis update will identifying potential flooding sources and should include plant-specific experience. In then be integrated into the possible equipment affected by flooding.

estimating pipe break frequencies, it is subsequent level 1 internal events recommended that experience with safety model update. In the PWROG RI-ISI methodology, related vs. balance of plant piping be piping failure probabilities are estimated considered along with active pipe degradation using the Win-SRRA code for each mechanisms. Credit for condition monitoring segment in the scope of the RI-ISI programs should also be applied where program. Failure probability estimates applicable, include appropriate degradation mechanisms including active degradation mechanisms.

Although, Beaver Valley should consider more recent pipe rupture data that considers an increased knowledge of pipe degradation mechanisms and considers plant aging concerns, updating is not expected to significantly impact the flooding results.

IF-D5-02 The initiating event frequency (IEF) for pipe IF-D5 Open - The internal flooding None. The internal flooding PRA breaks is based on a generic 80 percent analysis gaps will be addressed analysis was not used directly to support capacity factor. There are two issues with this during the internal flooding model the development of the RI-ISI program, method: a) current capacity factors are typically update that is scheduled for 2010. but instead was used as one source for greater than 80 percent so that the IEFs are This flooding analysis update will identifying potential flooding sources and slightly lower, and b) the method is inconsistent then be integrated into the possible equipment affected by flooding.

with the method used to calculate other IEFs. It subsequent level 1 internal events is recommended that the calculation for internal model update. In the PWROG RI-ISI methodology, flooding IEF be revised to be consistent with the piping failure probabilities are estimated method used for other IEFs. using the Win-SRRA code for each segment in the scope of the RI-ISI program. Failure probability estimates include appropriate degradation mechanisms including active

Page 20 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection degradation mechanisms.

Although, Beaver Valley should consider current capacity factors and methodology, updating is not expected to significantly impact the flooding results.

IF-El-01 The standard states "for each flood scenario, IF-El Open - The internal flooding None. The internal flooding PRA review the accident sequences for the analysis gaps will be addressed analysis was not used directly to support associated plant-initiating event group to during the internal flooding model the development of the RI-ISI program, confirm applicability of other accident update that is scheduled for 2010. but instead was used as one source for sequences model." A spot check was made to This flooding analysis update will identifying potential flooding sources and provide reasonable confidence that the overall then be integrated into the possible equipment affected by flooding.

results are correct. However, there is no record subsequent level 1 internal events that each scenario was reviewed, model update. This gap is a documentation consideration only.

IF-Fl-01 The internal flooding documentation does not IF-Fl, SY-A4 Open - The internal flooding None. The internal flooding PRA include the results of the walkdowns performed analysis gaps will be addressed analysis was not used directly to support during the original assessment. FENOC's during the internal flooding model the development of the RI-ISI program, response to the owners group peer review DE-4 update that is scheduled for 2010. but instead was used as one source for indicates that the RI-ISI walkdowns are This flooding analysis update will identifying potential flooding sources and documented and cover the issues required for then be integrated into the possible equipment affected by flooding.

an internal flooding walkdown. To facilitate subsequent level 1 internal events future maintenance and reviews of the internal model update. This gap is a documentation flooding assessments, the use of the RI-ISI consideration only.

walkdowns for internal flooding should be documented in the internal flooding notebook and a direct reference to a retrievable copy the RI-ISI walkdowns should also be included.

IF-F1-02 If the current internal flooding methodology is IF-F1 Open - The internal flooding None. The internal flooding PRA retained, a comparison of the current analysis gaps will be addressed analysis was not used directly to support methodology to the ASME standard is during the internal flooding model the development-of the RI-ISI program, recommended to facilitate future reviews, update that is scheduled for 2010. but instead was used as one source for This flooding analysis update will identifying potential flooding sources and then be integrated into the possible equipment affected by flooding.

subsequent level 1 internal events model update. This gap is a documentation consideration only.

IF-F2-01 The documentation of the processes to identify IF-F2 Open - The internal flooding None. The internal flooding PRA flood areas, sources, pathways, scenarios, etc. analysis gaps will be addressed analysis was not used directly to support are not clearly documented. For example, the during the internal flooding model the development of the RI-ISI program,

Page 21 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection rules used to screen out sources and areas are update that is scheduled for 2010. but instead was used as one source for not defined and the bases for eliminating or This flooding analysis update will identifying potential flooding sources and justifying propagation pathways is either not then be integrated into the possible equipment affected by flooding.

clearly defined or not provided at all. subsequent level 1 internal events model update. This gap is a documentation consideration only.

IF-F2-02 The internal flooding notebook states that the IF-F2 Open - The internal flooding None. The internal flooding PRA annual frequency of a flood scenario in location analysis gaps will be addressed analysis was not used directly to support X is Rx = Fi

• fx i

• fsx

• fp,x and the equation during the internal flooding model the development of the RI-ISI program, used to quantify scenarios in which recovery update that is scheduled for 2010. but instead was used as one source for actions can be included is Sx = Rx (Dx + Ix). This flooding analysis update will identifying potential flooding sources and However, the frequency is never quantified then be integrated into the possible equipment affected by flooding.

using these equations. This is confusing for a subsequent level 1 internal events reviewer - what is the purpose of these model update. This gap is a documentation statements if they are not used or if they are consideration only.

used, an explanation is needed. Note 1 defines the variables.

QU-B9-01 Component boundary conditions are not well QU-B9 Open - Will provide a discussion None. This gap is a documentation defined. The data analysis notebook, as well of component boundaries in the consideration only.

as several system notebooks (auxiliary data analysis or system feedwater and service water) were reviewed notebooks. This may be and there is no discussion of component addressed at a higher level by boundary. There are assumptions made identifying typical groupings of regarding system boundaries, but no discussion components.

of component boundaries. As a result, module definitions can not be determined.

QU-D5a-01 The revision 3B quantification notebook and QU-D5a Open - Will identify significant None. Category I is met and appropriate revision 4 initiating events notebook were SSCs and operator actions that for this application.

reviewed. Significant contributors to core contribute to initiating event damage frequency have been identified, but frequencies and event mitigation in This gap is a documentation there is no identification of SSCs and operator the quantification notebook. consideration only.

actions that contribute to initiating event frequencies and event mitigation.

QU-F4-01 The revision 3B quantification notebook, section QU-F4, Open - Will characterize all major None. This gap is a documentation 5 states that the PRA notebooks "include an QU-E4, sources of uncertainty using consideration only.

estimation of the uncertainty introduced by the IE-D3 WCAP-1 63043 or EPRI data used to quantify the PRA model...This TR-1 0096524 guidance.

3 Westinghouse Electric, WCAP 16304-P, "Strategy for Identification and Treatment for Modeling Uncertainties in PSAs Applications to LOCA and LOOP."

4 Electric Power Research Institute (EPRI), Report 1009652, "'Guideline for the Treatment of Uncertainty in Risk-Informed Applications' Technical Basis Document."

Page 22 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection uncertainty estimation does not, however, reflect possible effects on the results from other sources of uncertainty. Such sources may include such things as: optimism or pessimism in definitions of sequence, component, or human action success criteria; limitations in sequence models due to simplifications (for example, not modeling available systems or equipment) made to facilitate quantification; uncertainty in defining human response within the emergency procedures...; degree of completeness in selection of initiating events; assumptions regarding phenomenology or structures, systems, and components (SSC) behavior under accident conditions... While it is difficult to quantify the effects of such sources of uncertainty, it is important to recognize and evaluate them because there may be specific PRA applications where their effects may have a significant influence on the results."

QU-F4 requires that these sources of uncertainty be characterized regardless of the difficulty of the evaluation. By Beaver Valley's own admission, it is important to recognize and evaluate them because there may be specific PRA applications where their effects may have a significant influence on the results.

Furthermore, the documentation provided in chapter 5 of the quantification notebook makes a start at identifying the sources of model uncertainty. PWROG guidance suggests the number of identified sources of uncertainty typically is on the order of 50 items. It is also suggested that BVPS perform a more rigorous search to complete a fairly complete list of sources of uncertainty.

Page 23 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection QU-F4-02 A detailed description of the risk management QU-F4 Open - Will add a brief discussion None. This gap is a documentation program (RISKMAN) quantification process is regarding the limitations of the consideration only.

provided. However, the revision 3B RISKMAN methodology to the quantification notebook does not discuss quantification notebook.

limitations in the methodology.

QU-F6-01 Beaver Valley does list important operator QU-F6 Open - Will define the term None. This gap is a documentation action basic events; however, there is no "significant" and add a discussion consideration only.

documented definition of "significant". The or reference to justify the risk-revision 3B quantification notebook lists top importance rankings for systems accident sequences but provides no definition and basic events.

of whether they are "significant" or not. The only discussion is that there is "no single sequence makes up a large fraction of the CDF".

The quantification notebook states the following definition for important systems: "The system rankings for determining High Importance is based on having an F-V Importance greater than 5.OE-02 or a RAW greater than 10, while the Low Importance is based on having an F-V Importance less than 5.0E-03 and a RAW less than 2. Medium Importance systems are comprised of everything else in between these importance measures." This definition agrees with the Regulatory Guide 1.200 definition for "significant contributors." However, there is no documented justification (no reference to a standard definition, such as R.G. 1.200 or the EPRI PRA Applications Guide).

LE-C2a-01 LE-C2a is assigned a Capability Category I LE-C2a, Open - Will include realistic None. Capability Category I is met and because BVPS 2 does not use operator actions LE-C2b, operator actions as part of the appropriate for this application. Refer to post core damage. This is considered LE-C3, level 2 analysis based on severe the Section "Assessment of PRA conservative treatment of operator actions LE-C6 accident management guidelines, Capability Needed for Risk-Informed following the onset of core damage. To meet emergency operating, procedures, Inservice Inspection."

Capability Category III for this supporting and WCAP-16657-P.

requirement, BVPS-2 level 2 analysis must Any credit for post core damage operator contain realistic operator actions, based on actions would only help to reduce the 5 Westinghouse Electric, WCAP 16657-P, "SAMG Template for Level 2 PRA."

Page 24 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection severe accident management guidelines LERF.

(SAMGs), emergency operating procedures (EOPs), etc. such as WCAP-1 6657-P.

LE-C2b-01 Only recovery of AC power after uncovery of LE-C2b Open - Will discuss post core None. Capability Category I is met and top of active fuel is discussed in the Level 2 damage recoveries that could appropriate for this application.

notebook. It is concluded that not enough time impact LERF such as restoring exists to assign a high success probability. No feedwater to a ruptured steam This gap is a documentation other recoveries are discussed. generator, using WCAP-16657 consideration only.

and WCAP-16341 as a reference.

LE-C9a-01 Level 2 and LERF analysis stopped at LE-C9a, Open - Will justify the lack of None. Capability Category I is met and containment failure and continued operation of LE-C9b credit of equipment survivability appropriate for this application.

equipment and operator actions were not and review NUREG/CR-6595 6 for modeled. Operation of mitigating systems after guidance. This gap is a documentation containment failure is not modeled either. consideration only.

Justify the lack of credit of equipment survivability.

LE-C10-01 SGTR and containment bypass did not take LE-C10 Open - Will credit scrubbing for None. Capability Category I is met and credit for scrubbing. WCAP-16657 suggests SGTR and containment bypass appropriate for this application.

that scrubbing for tube rupture events can be events based on WCAP-16657.

credited by an operator action restart auxiliary The ASME standard recognizes Any credit for SGTR scrubbing would feedwater to the ruptured steam generator. scrubbing during SGTRs as a way only help to reduce the LERF.

to reduce LERF.

LE-D5-01 Beaver Valley Thermal Induced SGTR is based LE-D5 Open - Will update the thermal None. Capability Category I is met and on a 1995 Fauske and Associates report and induced SGTR model to appropriate for this application. Refer to Westinghouse Calculation CN-RRA-02-38. incorporate new methodology from the Section "Assessment of PRA Recent investigations suggest that these results WCAP-1 6341. Capability Needed for Risk-Informed may be too optimistic. A more reasonable Inservice Inspection."

approach may be implementing WCAP-16341, Simplified LERF Model, and characterizing the This is not expected to have a significant uncertainties based on that latest EPRI, impact on LERF.

PWROG, and NRC interactions.

LE-D6-01 The containment isolation analysis for LE-D6 Open - Will revise the BVPS-2 None. Category I is met and appropriate BV2REV3b is based on a sub-atmospheric LERF notebook to account for the for this application.

containment. BVPS-2 has been converted to change to atmospheric atmospheric so this analysis must be revisited, containment. This gap is a documentation BVl REV4 does account for the atmospheric consideration only.

containment conversion in the containment 6 U.S. Nuclear Regulatory Commission, NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events."

Page 25 of 28 9

Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection isolation notebook. The results of a similar assessment for BVPS-2 need to be incorporated in the LERF notebook.

LE-E4-01 The BVPS-2 LERF model is quantified using LE-E4 Open - Will develop database None. This gap is a documentation RISKMAN. Only point-estimates for each top distributions for level 2 split consideration only.

event are used and there are no uncertainty fractions, so that a Monte Carlo estimates or uncertainty propagation. quantification can be used for the LERF uncertainty propagation analysis.

No significant impact to the LERF mean value is expected.

LE-F2-01 The PRA peer review team suggested in L2-02, LE-F2 Open - Will use the LERF None. This gap is a documentation using uncertainty analysis for the LERF top uncertainty propagation analysis to consideration only.

events to ensure that future applications are not identify key sources of uncertainty, affected by use of point estimates. then perform and document sensitivity studies for the This fact and observation was entered into significant LERF contributors.

FENOC's Corrective Action Program as CA 02-09043-26 to track and resolve the No significant impact to the LERF issues. The suggested PRA Peer Review mean value is expected.

Team resolution to this observation was not addressed in the BV2REV3B PRA model update, but will be evaluated sometime later in a future PRA model update.

This update has not yet been completed. At the time, it was a "C" level fact and observation but the PRA standard raises the requirements for PRA quality and this fact and observation is now a "B" level.

LE-G5-01 Limitations of the LERF analysis are identified LE-G5 Open - Will document the None. This gap is a documentation throughout the BVPS-2 Level 2 notebook., limitations of the LERF analysis consideration only.

However, they need to be gathered into a single identified throughout the level 2 location to facilitate future usage. notebook into a single location to facilitate future usage.

HR-PR-001 BVPS does not have a written process for HR-D5, Open - Will explicitly describe the None. This gap is a documentation evaluating dependencies between multiple HR-G7, process used to identify and consideration only.

human error probabilities (HEPs) occurring in a HR-H3, evaluate the dependencies single accident and does not provide a HR-i1, HR-12 between multiple HEPs occurring

Page 26 of 28 p1 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection summary of those that were explicitly evaluated in the same accident sequence.

for dependencies and the associated levels of dependencies and joint HEPs.

HR-PR-002 BVPS does not appear to have evaluated their HR-G6, Open - Will perform and document None. This gap is a documentation HEPs for internal consistency consistent with HR-12 an explicit process for reviewing consideration only.

the requirements of HR-G6 and does not have the HEPs for internal consistency a documented process to do so. with respect to scenario, context, procedures and timing.

HR-PR-003 The method for quantifying pre-initiator HR-D2, Open - Use the EPRI HRA None. This gap is a documentation misalignment errors as described on page 8 of HR-D3, calculator to update and document consideration only.

the "Beaver Valley Power Station Unit 2 PRA HR-D4, the process for identification and Notebook- Human Reliability Analysis," HR-1i, HR-12 quantification of pre-initiator HFEs.

Revision 2, dated 10/01/07, relies on the use of a generic error of omission rate that does not

.reflect any detailed assessment of the human error probabilities. The process also does not consider the quality of plant-specific written procedures, administrative controls or the man-machine interface, and does not include an explicit assessment of the potential for recovery that specifically delineates which procedures and processes influence the potential for identification and recovery. Furthermore, the method for quantifying post-maintenance miscalibrations relies on a single generic error of omission rate.

A complication in reviewing the pre-initiator human failure events (HFEs) was that the HRA notebook does not include a list of the pre-initiator HFEs or their probabilities. The system

• notebooks provide evidence of the search for and identification of misalignments but they do not present a list of such events or their probabilities.

HR-PR-005 The BVPS HRA is documented in the "Beaver HR-13 Open - Will add an "Assumptions" None. This gap is a documentation Valley Power Station Unit 2 PRA Notebook - section to the HRA notebook and consideration only.

Human Reliability Analysis," Revision 2, dated collect all of the high level

Page 27 of 28 Facts and Description of Gap Supporting Current Status or Comment Importance to Risk-Informed Inservice Observations Requirement Inspection 10/01/07. This notebook does not have an assumptions into it. Also, explicit assumptions section to identify and qualitatively characterize the characterize assumptions. A review of this potential impact of the notebook revealed assumptions scattered assumptions and potential impact throughout the text. of alternate assumptions (if any) on the HRA analysis.

HR-PR-007 In general, BVPS excludes virtually all HR-B1 Open - Will develop a table which None. This gap is a documentation miscalibration events based on the assumption lists the individual test, consideration only.

that events related to instrument miscalibrations maintenance and calibration are captured in the equipment failure rate data activities that were reviewed as and the on-line maintenance program precludes potential pre-initiator human common-cause miscalibration by scheduling actions, and list the screening rule work on opposite trains in different weeks. that was applied for each action Post-maintenance misalignments were that was screened from further excluded for normally operating system based consideration.

on the assumption that misalignments on normally operating systems would be quickly detected and corrected. While these rules seem reasonable, they are applied to classes of maintenance and test activities to screen them from further consideration. This is sufficient for Capability Category I but not for Capability Category II.

Page 28 of 28 Notes for Table I

1. The variables used in the above equation [Cell IF-F2-02] for Rx are:

Rx = annual frequency of a flood scenario in location x.

Fi = total annual frequency of the flood of any severity in building i.

fx, i = conditional frequency of the flood occurring in location x of building i, given that the flood has occurred in building i.

fs,x = severity factor; conditional frequency of the flood being of a severity to cause equipment failure.

fp,x = propagation factor; conditional frequency of the flood propagating to the adjacent locations, given that the flood occurred at location x with the severity specified to cause equipment failure (for localized cases, fp,x = 1.0).

The variables used in the above equation [Cell IF-F2-02] for Sx are:

Sx = the annual frequency of the scenario and recovery failure.

Rx = the annual frequency of the scenario before recovery.

Dx = the probability that timely detection of the flood fails. This includes consideration of detection capabilities (alarms, etc.), the likelihood of operator diagnosis, and time available.

Ix = the probability of not isolating or mitigating the flood prior to failing critical systems, given detection of the flood. This includes human actions.

2. Reports are attached to the following FENOC letters:

FENOC letter to NRC, January 25, 2006, Beaver Valley Power Station, Unit No. 1 and No. 2, BV-1 Docket No. 50-334, License No. DPR-66, BV-2 Docket No. 50-412, License No. NPF-73, Additional Information in Support of License Amendment Request Nos. 302 and 173.

[NRC Accession Number ML060330262]

FENOC letter to NRC, February 14, 2006, Beaver Valley Power Station, Unit No. 1 and No. 2, BV-1 Docket No. 50-334, License No. DPR-66, BV-2 Docket No. 50-412, License No. NPF-73, Supplemental Response in Support of License Amendment Request Nos. 302 and 173.

[NRC Accession Number ML060520569]