L-10-333, Units, 1 and 2, Response to Request for Additional Information Related to Risk-Informed Inservice Inspection Requests RI-ISI-1 and RI-ISI-2

From kanterella
(Redirected from ML103540097)
Jump to navigation Jump to search

Units, 1 and 2, Response to Request for Additional Information Related to Risk-Informed Inservice Inspection Requests RI-ISI-1 and RI-ISI-2
ML103540097
Person / Time
Site: Beaver Valley
Issue date: 12/14/2010
From: Harden P
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-10-333, TAC ME4104, TAC ME4105
Download: ML103540097 (16)
v • d • e


Text

FENOC Beaver Valley Power Station 1__%

P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Paul A. Harden 724-682-5234 Site Vice President Fax: 724-643-8069 December 14, 2010 L-1 0-333 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 Response to Request for Additional Information Related to Risk-Informed Inservice Inspection Requests RI-ISI-1 and RI-ISI-2 (TAC Nos. ME4104 and ME4105)

By correspondence dated June 10, 2010*(Accession No. ML101650649), FirstEnergy Nuclear Operating Company (FENOC) requested approval for continued use of the existing Beaver Valley Power Station risk-informed inservice inspection programs, with updates, via 10 CFR 50.55a Requests RI-ISI-1 and RI-ISI-2.

By letter dated October 27, 2010 (Accession No. ML102980099), the Nuclear Regulatory Commission (NRC) staff requested additional information to complete its review of Requests RI-ISI-1 and RI-ISI-2. The attachment provides responses to the NRC staff's comments and questions, as clarified during a teleconference between FENOC and NRC staff on October 20, 2010.

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 A. Harden

Attachment:

Response to Request for Additional Information Related to Risk-Informed Inservice Inspection Requests RI-ISI-1 and RI-ISI-2 cc:

NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment L-1 0-333 Response to Request for Additional Information Related to Risk-Informed Inservice Inspection Requests RI-ISI-1 and RI-ISI-2 Page 1 of 15 By letter dated October 27, 2010, the Nuclear Regulatory Commission (NRC) staff requested additional information related to requests seeking renewal of risk-informed inservice inspection programs, with updates, for the Beaver Valley Power Station (BVPS).

The FirstEnergy Nuclear Operating Company (FENOC) responses for BVPS Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2) are provided below. The NRC staff's questions are presented in bold type, followed by FENOC's responses.

Probabilistic Risk Assessment Licensing Branch Review:

1. In Enclosure C, "Probabilistic Risk Assessment [PRA] Technical Adequacy for Risk-Informed Inservice Inspection [RI-ISI]," of the submittal, the licensee stated that the, [sic] "risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plant." Describe how you plan to maintain a living RI-ISI program for BVPS-1 and 2 during the fourth and third 10-year intervals, respectively, which can be affected by changes in plant design or operations.

Response

Per WCAP-14572, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision 1-NP-A, Section 4.5.2, it's required to re-evaluate RI-ISI programs on a 40-month period basis.

As such, the RI-ISI programs at BVPS are maintained as living programs by performing re-evaluations at the end of the first and second periods of the fourth ten-year inspection interval for BVPS-1, and at the end of the first and second periods of the third ten-year inspection interval for BVPS-2. These re-evaluations follow the guidance provided within WCAP-14572, in addition to the guidance provided within NEI 04-05, "Living Program Guidance To Maintain Risk Informed Inservice Inspection Programs For Nuclear Plant Piping Systems."

As part of these periodic re-evaluations, the following areas are specifically reviewed for changes that may potentially impact the RI-ISI programs.

Examination results Piping failures Probabilistic Risk Assessment (PRA) updates Plant design changes

  • Changes in postulated conditions (including those resulting from changes in plant operations)

Attachment Letter L-1 0-333 Page 2 of 15 Changes having an impact on the RI-ISI programs are incorporated by inserting the new information into the appropriate level of the respective analyses. Based on the significance of the identified changes, it may not be necessary to perform the entire risk-informed examination selection process.

2. Were the risk calculations performed according to all the guidelines provided on page 213, Section 4.4.2 of WCAP-14572, Revision 1-NP-A? If not, provide a description and justification of any deviation. In addition, provide a statement that verifies that the risk from the revised program continues to remain lower when compared to the last deterministic ASME Code, Section Xl inspection program.

Response

Yes. The change-in-risk calculations were performed in accordance with the guidelines provided within WCAP-14572, Revision 1-NP-A, Section 4.4.2.

Additionally, the risk from implementation of these programs is expected to slightly decrease when compared to that estimated from the last deterministic American Society of Mechanical Engineers (ASME) Code Section Xl inspection programs. This decrease is expected since examinations are optimizing (performed on risk-significant components) versus completing randomly selected examinations.

3. Regulatory Guide 1.200, Rev. 2 includes supporting requirements describing acceptable fire and seismic PRA analyses. Describe how fire and seismic initiating events are accounted for in the proposed RI-ISI program for BVPS-1 and 2.

Response

As discussed during an October 20, 2010 teleconference, FENOC is providing justification why fire and seismic are insignificant contributors to the PRA results.

Per WCAP-14572, Revision 1-NP-A, Section 3.6.2, if a PRA model does not exist for an external event, the expert panel considers the segment function in mitigating the consequences of the external event, as well as the likelihood of the event in the determination of whether a segment is high or low safety significant (that is, the expert panel considers fire and seismic events from a deterministic perspective).

As such, fire and seismic events were considered in the risk metric data, as well as deterministically, by the expert panel for the BVPS-1 and BVPS-2 RI-ISI programs.

EPRI Topical Report 1021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection

Attachment Letter L-10-333 Page 3 of 15 Programs," Section 2.2, notes the following with respect to internal fire and seismic events.

Regarding internal fire, the potential contribution of piping failure to internal fire risk is insignificant as the failure probability of piping is insignificant when compared to the failure probabilities of other systems, structures and components such as power supplies, pumps and valves.

Regarding seismic events, well engineered piping systems are considered seismically robust. Individual Plant Examination for External Events (IPEEE), nuclear industry, and NRC studies have shown piping systems to have seismic fragility capacities greater than the screening values typically used in seismic assessment and are considered not likely to fail during a seismic event. Inservice inspection is not considered in establishing the fragilities of such systems, structures and components.

Meeting defense-in-depth and safety margin principles provides additional assurance that these conclusions remain valid. Inservice inspection is an integral part of defense-in-depth, and the RI-ISI process maintains the basic intent of ISI, thus providing reasonable assurance of an ongoing substantive assessment of piping condition.

Additionally, there are no changes to any design basis events as a result of implementing inservice inspection programs; therefore, safety margins are maintained.

Piping and Non-Destructive Examination Review:

4. Identify the PRA model number used for the current RI-ISI program.

Response

The PRA model numbers used for the current RI-ISI programs are BV1 REV4 for BVPS-1, and BV2REV4 for BVPS-2.

5. Provide values for the changes in core damage frequency and changes in large early release frequency for BVPS-1 and 2 compared to ASME Code,Section XI ISI program values, and to the third ISI interval RI-ISI program values.

Response

As clarified during an October 20, 2010 teleconference, the third ISI interval applies to BVPS-2; and, FENOC is providing values for the BVPS-1 fourth ISI interval.

Changes in core damage frequency (CDF) and large early release frequency (LERF),

with and without operator action, for the ASME Section XI program and the RI-ISI programs for both the current and prior intervals, are provided in Table 1 for BVPS-1, and Table 2 for BVPS-2. The following notes are applicable to both tables.

Attachment Letter L-10-333 Page 4 of 15 The ASME Section Xl results are based on calculations from the current intervals for BVPS-1 and BVPS-2. Therefore, a direct comparison of the previous interval RI-ISI results for BVPS-1 and BVPS-2 to the current interval ASME Section Xl results is not meaningful since they are based on different inputs. However, the previous interval RI-ISI results (after conversion to point estimate values), along with the current interval RI-ISI results and the current ASME Section Xl results, are provided in Tables 1 and 2.

In the RI-ISI program submittals for the BVPS-1 third interval and BVPS-2 second interval (previous intervals), the change-in-risk results were presented based on mean values. The change-in-risk results for the current BVPS-1 and BVPS-2 intervals are presented based on point estimate values. The change-in-risk results from the previous BVPS-1 and BVPS-2 intervals were converted to point estimate values to allow a meaningful comparison between the current and prior BVPS-1 and BVPS-2 intervals, and the current interval ASME Section Xl results.

BVPS-1 Systems:

o BD - Steam Generator Blowdown System o

CH - Chemical and Volume Control System o

Cl - Containment Isolation System o

DV - Reactor Plant Drains and Vents Systems o

FW - Steam Generator Feedwater System o

HY - Hydrogen Control System o

MS - Main Steam System o

QS - Quench Spray System o

RC - Reactor Coolant System o

RH - Residual Heat Removal System o

RS - Recirculation Spray System o

SI - Safety Injection System o SS - Sampling System BVPS-2 Systems:

o BDG - Steam Generator Blowdown System o

CHS - Chemical and Volume Control System o

CI - Containment Isolation System o

DAS - Reactor Plant Drains and Vents Systems o

FWA - Steam Generator Feedwater System o

GNS - Gaseous Nitrogen System o

HCS - Hydrogen Control System o

MSS - Main Steam System o

QSS - Quench Spray System o

RCS - Reactor Coolant System o

RHS - Residual Heat Removal System o

RSS - Recirculation Spray System o

SIS - Safety Injection System o SSR - Sampling System

Attachment Letter L-1 0-333 Page 5 of 15 Table 1 Comparison of CDF and LERF for Delta Risk Evaluation [BVPS-1]

Case Third Interval Fourth Interval ASME

  • Systems RI-ISI RI-ISI Section Xl I CDF

[Without Operator Action]

  • BD 1.91E-11 1.20E-14 1.20E-14
  • CH 2 1.32E-08 1.41E-09 1.49E-09 C1 1.15E-11 1.79E-11 1.79E-11
  • DV 0.OOE+00 0.OOE+00 0.OOE+00
  • FW 3.04E-12 2.69E-15 2.62E-15 sHY 0.OOE+00 0.OOE+00 0.OOE+00 NMS 7.17E-13 1.08E-13 1.10E-13
  • QS 9.41E-11 1.38E-11 1.27E-11
  • RC 2 2.19E-09 1.20E-09 1.20E-09
  • RH 1.29E-12 7.89E-13 7.86E-13 I RS 7.64E-12 1.88E-13 4.14E-13 SI 2 1.07E-08 2.66E-09 2.78E-09
  • SS 3.89E-12 4.11E-13 4.11E-13
  • Total 2.63E-08 5.29E-09 5.50E-09 CDF

[With Operator Action]

  • BD 8.12E-14 1.20E-14 1.20E-14 eCH 2.85E-09 3.33E-11 7.12E-11

.C1 1.15E-11 1.66E-13 1.66E-13

'DV 0.OOE+00 0.OOE+00 0.OOE+00

  • FW 3.OOE-14 2.66E-15 2.59E-15
  • HY 0.OOE+00 0.OOE+00 0.OOE+00 eMS 7.37E-13 1.09E-13 1.11E-13 eQS 4.15E-12 1.66E-14 1.66E-14
  • RC 2 2.17E-09 1.19E-09 1.19E-09
  • RH 1.29E-12 7.89E-13 7.86E-13
  • RS 4.93E-12 7.38E-16 3.45E-14
  • SI 2 4.OOE-09 1.84E-09 1.84E-09
  • SS 1.11E-16 1.22E-14 1.22E-14
  • Total 9.05E-09 3.06E-09 3.10E-09

Attachment Letter L-10-333 Page 6 of 15 Table 1 Comparison of CDF and LERF for Delta Risk Evaluation [BVPS-1]

Case Third Interval Fourth ASME

  • Systems RI-ISI Interval RI-ISI Section Xl I LERF

[Without Operator Action]

BD 1.41E-15 6.26E-17 6.26E-17 eCH 8.33E-11 1.40E-13 1.48E-13

.C1 3.68E-13 1.86E-14 1.86E-14 IDV O.OOE+00 O.OOE+00 O.OOE+00 FW 4.47E-13 9.91E-19 6.28E-19 sHY O.OOE+00 6.69E-15 6.69E-15 IVMS 3.29E-14 2.87E-16 3.08E-16 0QS 3.55E-13 4.90E-14 4.52E-14 RC 1.74E-11 3.04E-14 3.04E-14 RH 1.03E-14 4.62E-15 4.62E-15

  • RS 8.90E-13 1.66E-13 6.68E-13 S

S12 5.41E-11 1.83E-11 1.88E-11

  • SS 2.77E-14 2.42E-17 2.42E-17
  • Total 1.57E-10 1.87E-11 1.97E-11 LERF

[With Operator Action]

BD 1.35E-15 6.25E-17 6.25E-17 C

OH 2 7.75E-12 6.12E-13 9.64E-13

'Cl 3.68E-13 1.76E-14 1.76E-14

.DV 0.OOE+00 O.OOE+00 O.OOE+00 FW 5.57E-16 8.62E-19 5.04E-19 sHY O.OOE+00 6.69E-15 6.69E-15 IMS 3.37E-14 2.88E-16 3.09E-16 eQS 2.66E-14 1.10E-16 1.10E-16 RC 1.73E-11 3.01E-14 3.01E-14 RH 1.03E-14 4.62E-15 4.62E-15 RS 8.75E-13 2.04E-15 3.36E-13 SI 2

3.16E-11 2.95E-13 3.98E-13

  • SS 8.33E-19 2.81E-19 2.81E-19
  • Total 5.80E-11 9.69E-13 1.76E-12 Notes:
1. Based on inputs from the fourth interval
2. Identifies a dominant system per the definition of dominant systems in WCAP-14572, Revision 1-NP-A, Section 4.4.2.

Attachment Letter L-1 0-333 Page 7 of 15 Table 2 Comparison of CDF and LERF for Delta Risk Evaluation [BVPS-2]

Case Second Third Interval ASME 0 Systems Interval RI-ISI RI-ISI Section Xl I CDF

[Without Operator Action]

BDG 8.04E-10 2.38E-11 2.59E-11 CHS 3.19E-08 5.21E-10 5.41E-10 oCl 1.18E-10 4.20E-15 3.73E-12 DAS 0.OOE+00 0.OOE+00 0.0OE+00 FWA 2.07E-13 9.86E-14 9.36E-14 GNS 1.26E-15 1.40E-13 1.40E-13 HCS 0.OOE+00 0.OOE+00 0.OOE+00 MSS 3.47E-12 3.33E-12 9.51E-12 QSS 3.08E-11 1.06E-09 1.20E-09 RCS 2 1.21E-08 1.17E-08 1.17E-08 RHS 2.71E-13 1.43E-13 1.43E-13 RSS 1.71E-11 5.67E-11 6.65E-11 SIS 2 4.13E-08 3.99E-09 5.53E-09 SSR 6.70E-11 4.01E-13 4.01E-13

  • Total 8.63E-08 1.73E-08 1.91E-08 CDF

[With Operator Action]

BDG 8.02E-10 2.38E-11 2.58E-11 CHS 7.66E-11 9.36E-12 3.82E-12

  • CI 1.18E-10 4.20E-15 3.73E-12 DAS 0.OOE+00 0.OOE+00 0.OOE+00 FWA 4.15E-14 1.22E-14 7.13E-15 GNS 1.26E-15 1.40E-13 1.40E-13 HCS 0.OOE+00 0.OOE+00 0.00E+00 MSS 3.76E-12 4.30E-12 1.24E-11 QSS 1.38E-12 4.56E-10 5.89E-10 RCS 2 1.21E-08 1.17E-08 1.17E-08 RHS 2.71E-13 1.43E-13 1.43E-13 RSS 1.53E-11 3.57E-12 4.18E-12 SIS2 6.81 E-09 1.75E-09 1.75E-09
  • SSR 1.59E-12 1.47E-14 1.47E-14
  • Total 1.99E-08 1.39E-08 1.41E-08

Attachment Letter L-10-333 Page 8 of 15 Table 2 Comparison of CDF and LERF for Delta Risk Evaluation [BVPS-2]

Case Second Third Interval ASME 0 Systems Interval RI-ISI RI-ISI Section XI 1 LERF

[Without Operator Action]

BDG 4.70E-12 2.39E-14 3.16E-14 CHS 8.30E-11 1.35E-14 1.93E-14

.CI 2.65E-13 1.66E-14 4.10E-14 DAS O.OOE+00 O.OOE+00 O.OOE+00 FWA 2.07E-14 5.27E-15 5.22E-15 GNS O.OOE+00 6.92E-19 6.92E-19 HCS O.OOE+00 O.OOE+00 0.OE+00 MSS 4.77E-14 1.67E-14 1.87E-14 QSS 1.08E-13 8.83E-13 9.12E-13 RCS 9.49E-11 1.16E-13 1.23E-13 RHS 2.09E-15 1.54E-18 1.54E-18 RSS 5.90E-14 5.19E-13 1.07E-11 SiS 2

1.19E-10 1.14E-11 3.95E-11

  • SSR 3.04E-13 1.36E-16 1.36E-16
  • Total 3.02E-10 1.30E-11 5.13E-11 LERF

[With Operator Action]

BDG 4.50E-12 1.79E-14 2.60E-14 C OHS 3.56E-13 2.42E-15 6.51E-15 eCI 2.65E-13 1.66E-14 4.10E-14 DAS O.OOE+00 O.OOE+00 O.OOE+00 FWA 2.19E-15 2.01E-16 1.57E-16

  • GNS 0.OOE+00 6.92E-19 6.92E-19 HCS O.OOE+00 O.OOE+00 O.OOE+00 MSS 2 8.89E-14 3.63E-14 5.07E-14 QSS 1.77E-14 2.05E-14 3.89E-14 RCS 2 9.49E-11 1.16E-13 1.23E-13 RHS 2.09E-15 1.54E-18 1.54E-18 RSS 5.25E-14 7.01E-17 7.98E-17 SIS 2 6.OOE-11 4.18E-14 1.85E-13 SSR 5.24E-14 2.44E-18 2.44E-18 Total 1.60E-10 2.52E-13 4.71E-13 Notes:
1. Based on inputs from the third interval
2. Identifies a dominant system per the definition of dominant systems in WCAP-14572, Revision 1-NP-A, Section 4.4.2.

Attachment Letter L-1 0-333 Page 9 of 15

6. For the 24 pipe segments of BVPS-1 and the 6 pipe segments of BVPS-2 that were moved from the High Safety Significant category to the Low Safety Significant category, as a result of the expert panel evaluation, identify these pipe segments, provide the Risk Reduction Worth value, and explain why each segment's safety significance changed.

Response

As clarified during an October 20, 2010 teleconference, the pipe segments that moved from high safety significant (HSS) to low safety significant (LSS) for the current ISI intervals, including their basis for change, are provided in Table 3 for BVPS-1, and Table 4 for BVPS-2. As depicted within Table 3, all risk reduction worth (RRW) values are less than 1.001; additional explanation is not required.

During development of this response, it was identified that seven BVPS-2 pipe segments changed from HSS to LSS, not six as stated in 10 CFR 50.55a Request RI-ISI-2, submitted by FENOC to the NRC on June 10, 2010. Additionally, it was also identified that 50 LSS pipe segments changed to HSS, not 48 as stated in the same BVPS-2 submittal. These compilation errors are considered editorial in nature, do not affect any supporting analyses or actual quantity of pipe segments changing categories, were documented within FENOC's correction action program on November 18, 2010, and communicated to NRC staff on November 24, 2010.

No similar compilation errors were identified within 10 CFR 50.55a Request RI-ISI-1 for BVPS-1.

Table 3 Segments Changing from HSS in the Third Interval to LSS in the Fourth Interval [BVPS-1]

Segment ID Basis for Making Segment LSS CH-134 All fourth interval RRWs are less than 1.001 RS-007 All fourth interval RRWs are less than 1.001 RS-008 All fourth interval RRWs are less than 1.001 RS-009 All fourth interval RRWs are less than 1.001 RS-010 All fourth interval RRWs are less than 1.001 RS-030 All fourth interval RRWs are less than 1.001 RS-031 All fourth interval RRWs are less than 1.001 RS-032 All fourth interval RRWs are less than 1.001 RS-033 All fourth interval RRWs are less than 1.001 SI-032 All fourth interval RRWs are less than 1.001 SI-033 All fourth interval RRWs are less than 1.001

Attachment Letter L-10-333 Page 10 of 15 Table 3 Segments Changing from HSS in the Third Interval to LSS in the Fourth Interval [BVPS-1]

Segment ID Basis for Making Segment LSS SI-034 All fourth interval RRWs are less than 1.001 SI-035 All fourth interval RRWs are less than 1.001 SI-036B All fourth interval RRWs are less than 1.001 SI-037B All fourth interval RRWs are less than 1.001 SI-069 All fourth interval RRWs are less than 1.001 SI-070A All fourth interval RRWs are less than 1.001 SI-072A All fourth interval RRWs are less than 1.001 SI-074A All fourth interval RRWs are less than 1.001 SI-076A All fourth interval RRWs are less than 1.001 SI-077A All fourth interval RRWs are less than 1.001 SI-078A All fourth interval RRWs are less than 1.001 SI-079A All fourth interval RRWs are less than 1.001 SI-086A All fourth interval RRWs are less than 1.001 Notes:

LSS = low safety significant RRW = risk reduction worth All RRWs include the RRWs for CDF without operator action, CDF with operator action, LERF without operator action, and LERF with operator action (both with and without uncertainty)

Attachment Letter L-10-333 Page 11 of 15 Table 4 Segments That Changed from HSS in the Second Interval to LSS in the Third Interval [BVPS-2]

Second Interval Third Interval Basis RRWs RRWs Segment Without With Without With ID Case Uncert.

Uncert.

Uncert.

Uncert.

CDF 1.001 1.003 1.000 1.000 PRA values (CCDP and CLERP, with and without operator action) w/o decreased. Failure probability increased slightly.

CF 1.001 1.005 1.000 1.000 CDF w

Third interval RRWs are less than 1.001, with the exception of the LERF 1.000 1.002 1.000 1.000 with operator action LERF RRWs, which are low.

w/o LERF 1.000 1.002 1.001 1.002 w

CDF 1.000 1.000 1.000 1.000 PRA values (CCDF and CLERF, with operator action) decreased.

w/o CDF Third interval RRWs are less than 1.001, with the exception of the w

1.002 1.005 1.000 1.000 with operator action with uncertainty LERF RRW, which is low.

LERF 1.000 1.000 1.000 1.000 w0o LERF 1.003 1.008 1.000 1.002 W

CDF 1.001 1.003 1.000 1.000 Third interval RRWs are less than 1.001.

w/o CDF 1.002 1.006 1.000 1.000 w

SIS-059B LERF 1.003 1.007 1.000 1.000 w/o LERF 1.003 1.008 1.000 1.000

.W

Attachment Letter L-10-333 Page 12 of 15 Table 4 Segments That Changed from HSS in the Second Interval to LSS in the Third Interval [BVPS-2]

Second Interval Third Interval Basis RRWs RRWs Segment Without With Without With ID Case Uncert.

Uncert.

Uncert.

Uncert.

CDF 1.002 1.004 1.000 1.000 PRA values (CCDP and CLERP, with and without operator action) w/o decreased.

CDF 1.002 1.005 1.000 1.000 Third interval RRWs are less than 1.001, with the exception of the SIS-061B with operator action with uncertainty LERF RRW, which is low.

LERF 1.001 1.003 1.000 1.000 LERF with operator action with uncertainty RRW is less than 1.001, w/o LERF but rounds to 1.001.

LERF 1.003 1.009 1.000 1.001 w

CDF 1.004 1.008 1.000 1.000 Failure probability decreased.

w/o CDF 1PRA values (CCDF/CCDP and CLERF/CLERP, with and without w

operator action) decreased with the exception of the CCDP portion LERF for LERF without operator action, which increased slightly. The SIS-064A w/o 1.000 1.000 decrease in failure probability was greater than the increase in CCDP portion for LERF without operator action.

LERF 1.000 1.000 1.000 1.001 Third interval RRWs are less than 1.001, with the exception of the w

with operator action with uncertainty LERF RRW, which is low.

LERF with operator action with uncertainty RRW is less than 1.001, but rounds to 1.001.

CDF 1.000 1.000 1.000 1.000 Previously classified HSS to be consistent with SIS-064A. Revised w/o failure probability on SIS-064A made SIS-064A risk metrics CDF 1.000 1.000 1.000 1.000 consistent with SIS-065A.

w SIS-065A LERF Third interval RRWs are less than 1.001, with the exception of the w/o0 1.000 1.000 1.000 1.000 with operator action with uncertainty LERF RRW, which is low.

LERF LERF with operator action with uncertainty RRW is less than 1.001, W

1.000 1.000 1.000 1.001to.001.

Attachment Letter L-10-333 Page 13 of 15 Table 4 Segments That Changed from HSS in the Second Interval to LSS in the Third Interval [BVPS-2]

Second Interval Third Interval Basis RRWs RRWs Segment Without With Without With ID Case Uncert.

Uncert.

Uncert.

Uncert.

CDF 1.005 1.009 1.000 1.000 Third interval RRWs are less than 1.001.

w/o CDF 1.000 1.000 1.000 1.000 SIS-097B w

LERF 1.002 1.006 1.000 1.000 w/o LERF 1.000 1.000 1.000 1.000 W

1_00_0 Notes:

CCDF = conditional core damage frequency CCDP = conditional core damage probability CLERF = conditional large early release frequency CLERP = conditional large early release probability LSS = low safety significant HSS = high safety significant w = with operator action w/o = without operator action

Attachment Letter L-10-333 Page 14 of 15

7. Identify any augmented inspection programs which have been subsumed in the proposed RI-ISI program and discuss the reason(s) for the change.

Response

Per WCAP-14572 Revision 1-NP-A, Addendum 1-A, "Addendum to 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report' to Address Changes to Augmented Inspection Requirements," the BVPS-2 break exclusion region (referred to as the break exclusion zone at BVPS) augmented ISI program has been subsumed into the proposed RI-ISI program for BVPS-2. This break exclusion region augmented ISI program was also subsumed into the previous RI-ISI program for BVPS-2; therefore, there is no change.

8. Describe how the industry initiative Materials Reliability Program-139, Revision 1, inspection guidelines have been implemented in conjunction with the RI-ISI program.

Response

The following weldments are included in the scope of both the MRP-139, Revision 1 inspection guidelines, and the RI-ISI programs:

Five BVPS-1 top-head pressurizer nozzles, consisting of:

o one pressurizer spray nozzle o four pressurizer safety/relief nozzles

" Six BVPS-2 pressurizer nozzles (including the surge)

Six BVPS-2 reactor vessel nozzles Currently, all eleven BVPS-1 and BVPS-2 pressurizer nozzle welds have been remediated using a full structural weld overlay (FSWOL). The six BVPS-2 reactor vessel nozzle welds have not been remediated.

MRP-1 39 categorizes the eleven pressurizer welds as Category B weldments not made of resistant materials, having no known cracks, and having been reinforced by full structural or optimized weld overlays made of primary water stress corrosion cracking (PWSCC) resistant material. For this category, MRP-1 39 requires inspection, including a preservice exam, according to a schedule consistent with the existing ASME Code Examination program or its approved alternative. Per ASME Section Xl, Non-Mandatory Appendix Q, all eleven pressurizer nozzle FSWOL welds received a preservice ultrasonic testing (UT) exam prior to service, and were re-examined during the second refueling outage following their application. The current RI-ISI program scope includes these welds. Examination selection is in accordance with WCAP-14572, Revision 1-NP-A, Section 3.7, and WCAP-14572, Revision 1-NP-A, Supplement 2, Section 2.3.

Attachment Letter L-1 0-333 Page 15 of 15 MRP-1 39 categorizes the three BVPS-2 reactor vessel hot leg nozzle welds as Category D, and the three BVPS-2 reactor vessel cold leg nozzle welds as Category E. Category D welds are required to be examined once every five years; Category E welds, once every six years. The current RI-ISI program scope includes these welds. These welds are selected for examination (Region 1A, locations susceptible to or analyzed as highly susceptible to an active degradation mechanism) per the RI-ISI program, along with a second weld (Region 1 B, other locations on the Region 1 segment) located within the same hot/cold leg loop.

These welds have not yet been examined during the current interval; however, they have been scheduled for examination.

Retrieved from "https://kanterella.com/w/index.php?title=L-10-333,_Units,_1_and_2,_Response_to_Request_for_Additional_Information_Related_to_Risk-Informed_Inservice_Inspection_Requests_RI-ISI-1_and_RI-ISI-2&oldid=5287789"