DCL-14-074, ASME Section XI Inservice Inspection Program Request for Alternative RPV-U1-Extension to Allow Use of Alternate Reactor Reactor Inspection Interval Requirements

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ASME Section XI Inservice Inspection Program Request for Alternative RPV-U1-Extension to Allow Use of Alternate Reactor Reactor Inspection Interval Requirements
ML14230A618
Person / Time
Site: Diablo Canyon Pacific Gas & Electric icon.png
Issue date: 08/18/2014
From: Allen B
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
PG&E Letter DCL-14-074
Download: ML14230A618 (11)
v • d • e


Text

Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.4888 August18,2014 Internal: 691.4888 Fax: 805.545.6445 PG&E Letter DCL-14-074 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 ASME Section XI lnservice Inspection Program Request for Alternative RPV-U 1-Extension to Allow Use of Alternate Reactor Inspection Interval Requirements

References:

1. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition with Addenda through 2003.
2. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207).
3. PWROG Letter OG-1 0-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," July 12,2010 (ADAMS Accession Number ML11153A033).

Dear Commissioners and Staff:

In accordance with Title 10 of the Code of Federal Regulations (1 0 CFR),

Part 50.55a, "Codes and Standards," paragraph (a)(3)(i), Pacific Gas and Electric Company (PG&E) proposes an alternative to the requirements of the ASME B&PV Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," (Reference 1) as applicable to Diablo Canyon Power Plant (DCPP),

Unit 1.

PG&E's Request for Alternative (RFA) RPV-U1-Extension pertains to the requirements of ASME Code,Section XI, Paragraph IWB-2412, "Inspection Program 8," requiring volumetric examination of essentially 100 percent of reactor pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk PG&E Letter DCL-14-074 August18,2014 Page 2 PG&E proposes to extend the DCPP Unit 1 reactor vessel inspection interval from 10 years to 20 years in accordance with WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval" (Reference 2).

The proposed interval extension is a deviation from the latest revised implementation plan, OG-1 0-238 (Reference 3). The impact to the intent of the implementation plan in OG-1 0-238 is considered by PG&E to be minimal since a majority of the reactor volumetric inspections scheduled for the DCPP Unit 1 third lSI interval were completed to the extent feasible in the recent Unit 1 eighteenth refueling outage ( 1R 18), including approximately 84 percent of the linear footage of beltline region welds. The results of the 1R 18 beltline inspections performed are essentially identical to those of the previous examinations with only one indication in the beltline region detected. This indication was found to be acceptable per the applicable ASME code. Unresolved inspection robot control issues in 1R 18 led to the deferral of the remainder of the examinations to Unit 1 nineteenth refueling outage (1 R 19), the last outage in the third inspection interval for DCPP Unit 1.

PG&E has concluded that the proposed alternative provides an acceptable level of quality and safety, in accordance with 10 CFR 50.55a(a)(3)(i). The requisite supporting information and basis for use are provided in the enclosed RFA RPV-U1-Extension for DCPP, Unit 1.

PG&E requests approval of the RFA by August 1, 2015, to support 1R 19 planning.

This RFA is similar to recent alternatives granted for Duke Energy's McGuire Nuclear Station, Unit 2, by letter dated September 6, 2012 (ADAMS Accession No. ML12249A175), and Dominion's Surry Power Station, Units 1 and 2, by letter dated April30, 2013 (ADAMS Accession No. ML13106A140).

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

If you have any questions regarding the information enclosed, or other lSI program activities, please contact Mr. Tom Baldwin at (805) 545-4720.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk PG&E Letter DCL-14-074 August 18, 2014 Page 3 Sincerely,

-.5~ 5. AtL.-

Barry S. Allen Site Vice President rntt/4231/50621143 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Thomas R. Hipschman, NRC Senior Resident Inspector Gonzalo L. Perez, Branch Chief, California Department of Public Health Balwant K. Singal, NRC Project Manager State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Enclosure PG&E Letter DCL-14-074 Proposed Alternative for Diablo Canyon Power Plant Unit 1 In Accordance with 10 CFR 50.55a(a)(3)(i)

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

1. ASME Code Component(s) Affected
2. Applicable Code Edition and Addenda
3. Applicable Code Requirement
4. Reason for Request
5. Proposed Alternative and Basis for Use
6. Duration of Proposed Alternative
7. Precedents
8. References 1

Enclosure PG&E Letter DCL-14-074 Proposed Alternative for Diablo Canyon Power Plant Unit 1 In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Diablo Canyon Power Plant (DCPP) Unit 1 reactor vessel (RV); specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (8PV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IW8-2500 and Table IW8-2500-1 of the ASME BPV, Code Section XI.

Category 8-A welds are defined as "Pressure Retaining Welds in Reactor Vessel."

Category 8-D welds are defined as "Full Penetration Welded Nozzles in Vessels."

Examination Category Item No. Description 8-A 81.11 Circumferential Shell Welds 8-A 81.12 Longitudinal Shell Welds B~A 81 .21 Circumferential Head Welds 8-A 81.22 Meridional Head Welds 8-A 81.30 Shell-to-Flange Weld 8-D 83.90 Nozzle-to-Vessel Welds (Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME 8PV Code,Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

The DCPP Unit 1 third 10-year inservice inspection (lSI) interval is based on the ASME Code Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"

2001 Edition through 2003 Addenda (Reference 1). ASME Code Section XI, 2001 Edition without Addenda applies to ultrasonic examinations performed per ASME Code Section XI, Appendix VIII requirements. The applicable ASME Section XI Code for the fourth 10-year lSI interval for DCPP will be the 2007 Edition with 2008 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program 8, requires volumetric examination of essentially 100 percent of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The DCPP Unit 1 third 10-year lSI interval is scheduled to end on May 6, 2015.

2

Enclosure PG&E Letter DCL-14-07 4

4. Reason for Request

An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100 percent of reactor vessel pressure-retaining Examination Category B-A and 8-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use Pacific* Gas and Electric Company (PG&E) proposes not to perform the ASME Code required volumetric examination of the DCPP Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and 8-D welds for the third lSI, which was scheduled for 2014. PG&E will perform the third ASME Code required volumetric examination of the DCPP Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and 8-D welds in the fourth lSI interval in 2025 plus or minus one refuelling outage. The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2). The impact to the implementation plan in OG-1 0-238 would increase the number of inspections in 2025 from two to three and decrease the number of inspections in 2015 from six to five. Based on Figures 3 and 4 of OG-1 0-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval" (Reference 4). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for DCPP Unit 1 were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for DCPP Unit 1 are bounded by the results of the Westinghouse pilot plant qualifies DCPP Unit 1 for an lSI interval extension. Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of DCPP Unit 1.

Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

3

Enclosure PG&E Letter DCL-14-074 Table 1: Critical Parameters for the Application of Bounding Analysis for DCPP Unit 1 Additional Pilot Plant Plant-Specific Parameter Evaluation Basis Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization No (PTS) Transients in the NRC PTS (Reference 5) Study (Reference 6)

Risk Study are Applicable Through-Wall Cracking Frequency 1. 76E-08 Events per 5.56E-09 Events per No (TWCF) year (Reference 4) year (Calculated per Reference 4)

Frequency and Severity of Design 7 heatup/cooldown Bounded by 7 No Basis Transients cycles per year heatup/cooldown (Reference 4) cycles per year Cladding Layers (Single/Multiple) Single Layer No Single Layer (Reference 4) 4

Enclosure PG&E Letter DCL-14-074 Table 2 below provides a summary of the latest reactor vessel inspection for DCPP Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the DCPP Unit 1 RV.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for DCPP Unit 1 The 2005 lSI was conducted in accordance with the ASME Code,Section XI and Section V, 1989 Edition, with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Examinations of Category B-A and B-D welds were performed to ASME Section XI, Appendix VIII, 1995 Inspection methodology:

Edition with 1996 Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future ISis will be performed to ASME Section XI, Appendix VIII requirements.

Number of past Two complete 10-year ISis have been performed.

inspections:

There was one indication identified in the beltline region during the most recently completed lSI (2005). This subsurface indication is located in the intermediate shell to lower shell circumferential weld (Item 20 in Table 3).

Number of indications The indication is acceptable per Table IWB-351 0-1 of Section XI of the found: ASME Code. This indication is not within the inner 1/1 oth or 1 inch of the reactor vessel thickness; therefore, it is inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

The third lSI was scheduled for 2014. This inspection will be performed in 2025 plus or minus one refuelling outage. The proposed inspection date for DCPP Unit 1 is a deviation from the schedule presented in OG-1 0-238 Proposed inspection (Reference 2). The impact to the implementation plan in OG-1 0-238 would schedule for balance of increase the number of inspections in 2025 from two to three and plant life: decrease the number of inspections in 2015 from six to five. Based on Figures 3 and 4 of OG-1 0-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

5

Enclosure PG&E Letter DCL-14-074 Table 3 summarizes the inputs and outputs for the calculation of TWCF.

Table 3: Details of TWCF Calculation for DCPP Unit 1 at 54 Effective Full-Power Years (EFPY)

Inputs Inter. & Lower Shell Twall [inches] : 8.85 Reactor Coolant System Temperature, T c [°F]: N/A Upper Shell T wall [inches]: 10.97 cu<1l Ni<1l (1) Fluence Material R.G . 1.99 CF(1) [°F] RTNDT(u) 19 No. Region and Component Description [1 0 n/cm 2 ,

Heat No. [wt%] [wt%] Pos.<1l (oF]

E>1.0 MeV] <1l 1 Upper Shell Plate 84105-1 C-2624-1 0.12 0.56 1.1 82.2 28 0.0341 2 Upper Shell Plate 84105-2 C-2624-2 0.12 0.57 1.1 82.4 9 0.0341 3 Upper Shell Plate 841 05-3 C-2608-28 0.14 0.56 1.1 98.2 14 0.0341 4 Intermediate Shell Plate 841 06-1 C-2884-1 0.125 0.53 1.1 85.3 -10 2.02 5 Intermediate Shell Plate 84106-2 C-2854-2 0.120 0.50 1.1 81 .0 -3 2.02 6 Intermediate Shell Plate 841 06-3 C-2793-1 0.086 0.476 1.1 55.2 30 2.02 7 Lower Shell Plate 84107-1 C-3121-1 0.13 0.56 1.1 89.8 15 2.01 8 Lower Shell Plate 84107-2 C-3131-2 0.12 0.56 1.1 82 .2 20 2.01 9 Lower Shell Plate 84107-3 C-3131-1 0.12 0.52 1.1 81.4 -22 2.01 10 Upper Shell Long. Weld 1-442A 27204/12008 0.19 0.97 1.1 215.7 -20 0.0245 11 Upper Shell Long. Weld 1-4428 27204/12008 0.19 0.97 1.1 215.7 -20 0.0149 12 Upper Shell Long . Weld 1-442C 27204/12008 0.19 0.97 1.1 215.7 -20 0.0306 13 Intermediate Shell Long. Weld 2-442A 27204 0.203 1.018 2.1 214.1 -56 1.49 14 Intermediate Shell Long. Weld 2-4428 27204 0.203 1.018 2.1 214.1 -56 1.49 15 Intermediate Shell Long. Weld 2-442C 27204 0.203 1.018 2.1 214.1 -56 0.768 16 Lower Shell Long . Weld 3-442A 27204 0.203 1.018 2.1 214.1 -56 1.19 17 Lower Shell Long . Weld 3-4428 27204 0.203 1.018 2.1 214.1 -56 1.19 18 Lower Shell Long. Weld 3-442C 27204 0.203 1.018 2.1 214.1 -56 2.01 19 Upper To Inter. Shell Circ. Weld 8-442 13253 0.25 0.73 1.1 197.5 -56 0.0341 20 Inter. To Lower Shell Circ. Weld 9-442 21935 0.183 0.704 1.1 172.2 -56 2.01 Outputs Methodology Used to Calculate LlT3o: Regulatory Guide 1.99, Revision 2\~J Controlling Material Fluence RTMAX-XX 19 FF (Fiuence Region No. (From [1 0 Neutron/cm 2 , LlT3o [°F] TWCF9s-xx (oR] Factor)

Above) E > 1.0 MeV]

Limiting Axial Weld - AW 18 658.54 2.01 1.190 254.87 2.42E-09 Limiting Plate - PL 7 581 .57 2.01 1.190 106.90 4.22E-13 Circumferential Weld - CW 20 608.66 2.01 1.190 204.99 O.OOE+OO TWCF9s-TOTAL(aAwTWCF9s-Aw+ OpLTWCF9s-PL + OcwTWCF9s-cw) : 5.56E-09 (1) Per Reference 8. The Pos1t1on 2.1 chemistry factor (CF) value for the 1ntermed1ate and lower shell longitudinal welds (heat 27204) incorporates all surveillance capsule data, including sister plant data from Palisades. Per Reference 8, the surveillance data for weld heat 27204 was deemed credible.

(2) Reference 9 6

Enclosure PG&E Letter DCL-14-074

6. Duration of Proposed Alternative This request is applicable to the DCPP Unit 1 lSI Program for the third and fourth 10-year inspection intervals.
7. Precedents
  • "Palo Verde Nuclear Generating Station, Units 1, 2, and 3- Relief Request No.

RR-40, Reactor Vessel Weld Examination Interval Extension (TAC Nos. ME1634, ME1635, and ME1636)," dated February 22, 2010 (ADAMS Accession Number ML100290415).

  • "Safety Evaluation of Relief Requests to Extend the lnservice Inspection Interval for Reactor Vessel Examinations for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. ME1478, ME1479, ME1480 and ME1481)," dated February 22, 2010 (ADAMS Accession Number ML100491550).
  • "Arkansas Nuclear One, Unit 2 - Request for Alternative AN02-ISI-004, to Extend the Third 10-Year lnservice Inspection Interval for Reactor Vessel Weld Examinations (TAC No. ME2508)," dated September 21, 2010 (ADAMS Accession Number ML102450654).
  • "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2)- Relief Request for Extension of the Reactor Vessel lnservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession Number ML101750402).
  • "Three Mile Island Nuclear Station, Unit 1 (TMI-1)- Request to Extend the lnservice Inspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed Alternative Request Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484)," dated September 21, 2010 (ADAMS Accession Number ML102390018).
  • "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April30, 2013 (ADAMS Accession Number ML13106A140).
  • "McGuire Nuclear Station, Unit 2, Relief 10-MN-002 to Extend the lnservice Inspection Interval for Reactor Vessel Category 8-A and 8-D Welds (TAC Nos.

ME7329 and ME 7330)," dated September 6, 2012 (ADAMS Accession Number ML12249A175).

7

Enclosure PG&E Letter DCL-14-074

8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition through 2003 Addenda, American Society of Mechanical Engineers, New York
2. PWROG Letter OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." PA-MSC-0120, July 12, 2010 (ADAMS Accession Number ML11153A033)
3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission, November 2002
4. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207)
5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission, March 2010
6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482)
7. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. 1, dated January 4, 2010, and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010
8. Westinghouse Report WCAP-17315-NP, Revision 0, "Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations," July 2011
9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988 8
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