DCL-10-098, Response to NRC Letter Dated July 15, 2010, Request for Additional Information (Set 10) for License Renewal Application
| ML102360040 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/12/2010 |
| From: | Becker J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-10-098 | |
| Download: ML102360040 (10) | |
Text
Pacific Gas and Electric Company James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P. 0. Box 56 Avila Beach, CA 93424 805.545.3462 August 12, 2010 internal: 691.3462 Fax: 805.545.6445 PG&E Letter DCL-10-098 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Letter dated July 15, 2010, Request for Additional Information (Set 10) for the Diablo Canyon License Renewal Application
Dear Commissioners and Staff:
By letter dated November 23, 2009, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA), and Applicant's Environmental Report -
Operating License Renewal Stage.
By letter dated July 15, 2010, the NRC staff requested additional information needed to continue their review of the DCPP LRA.
PG&E's response to the request for additional information is included in. LRA Amendment 9, resulting from the responses, is included in showing the changed pages with line-in/line-out annotations.
PG&E makes no regulatory commitments (as defined in NEI 99-04) in this letter.
If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4160.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon 9 Palo Verde
- San Onofre
- South Texas Project
- Wolf Creek HKZ1V4
Document Control Desk August 12, 2010 Page 2 PG&E Letter DCL-10-098 I declare under penalty of perjury that the foregoing is true and correct.
Executed on August 12, 2010.
Sincerely, James R. Becker Site Vice President pns/50327747 Enclosures cc:
Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial B. Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, Office of Nuclear Reactor Regulation A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway ° Comanche Peak
- Diablo Canyon
- Palo Verde
- San Onofre
- South Texas Project e Wolf Creek PG&E Letter DCL-10-098 Page 1 of 5 PG&E Response to NRC Letter dated July 15, 2010 Request for Additional Information (Set 10) for the Diablo Canyon License Renewal Application RAI2.4.4-1 Based on the staffs review of Section 2.4.4, "Turbine Building," and Table 2.4-4 of the license renewal application (LRA), it is not clear if the roof and roofing membranes have been included within the scope of license renewal and subject to an aging management review (AMR). If the components are not included due to oversight, please discuss whether these components are within the scope of license renewal and subject to an AMR. If they are excluded from the scope of license renewal, please provide the basis for their exclusion.
PG&E Response to RAI 2.4.4-1 By letter dated June 18, 2010, PGE submitted license renewal application (LRA)
Amendment 1, which revised the turbine building screening and aging management reports and updated LRA Section 2.4.4 and Table 2.4-4, "Turbine Building," to add additional details regarding the turbine building. The response explains that the turbine building roof structure and materials are within the scope of license renewal but the design does not have a membrane.
PG&E Letter DCL-10-098 Page 2 of 5 RAI 2.5-1 NEI 95-10, "Industry Guideline for Implementing the Requirements, of 10 CFR Part 54 -
The License Renewal Rule, " Appendix B indicates that elements, resistance temperature detectors (RTDs), sensors, thermocouples and transducers should be included in the list of component or commodity groups subject to an AMR if a pressure boundary is applicable. LRA Section 2.1.3.1 states that instrument and control (I&C) components with mechanical functions such as flow elements, flow indicators, flow orifices, and sight gauges were evaluated in their respective mechanical systems.
However, it is not clear if I&C components such as RTDs, thermocouples and transducers that may have a pressure boundary function or other "mechanical function" have been included within the scope of license renewal. Please explain how such components were evaluated. Provide examples or references to such components and indicate where in the LRA these components have been listed as component types subject to an AMR, if required in accordance with 10 CFR 54.4(a).
PG&E Response to 2.5-1 License'renewal application Section 2.1.3.1, "Mechanical System Scoping Methodology," indicates that instrument and control components with mechanical functions such as flow elements, flow indicators, flow orifices, and sight gauges are evaluated in their respective mechanical systems. Resistance temperature detectors (RTDs), sensors, thermocouples, transducers, and various elements at Diablo Canyon Power Plant do not have a pressure boundary since they are not in-line components.
Thermowells or mounting brackets that may provide a pressure boundary for the sensing devices are evaluated as part of the piping system. Therefore, an aging management program for RTDs, sensors, thermocouples, transducers, and various elements is not required in accordance with 10 CFR 54.4(a).
PG&E Letter DCL-10-098 Page 3 of 5 DCPP RAI 3.1.2.1-1 The Generic Aging Lessons Learned (GALL) Report item IV.B2-21 recommends the Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program to manage loss of fracture toughness due to thermal aging and neutron irradiation embrittlement for the CASS lower internal assembly (lower support casting and lower support plate columns) exposed to reactor coolant (>250'C) and neutron flux. In addition, GALL Report item IV.B2-37 recommends the Thermal Aging and Neutron Irradiation Embrittlement of CASS Program to manage loss of fracture toughness due to thermal aging and neutron irradiation embrittlement for the CASS upper internals assembly, upper support columns exposed to reactor coolant
(>2500C) and neutron flux.
Diablo Canyon Power Plant (DCPP) LRA Table 3.1.2-1, citing GALL Report items IV.B2-21 and IV.B2-37, indicates that the loss of fracture toughness of CASS Reactor Vessel Internal (RVI) lower core support structure (core support casing (U1)) and RVI upper core support structure (upper support colimns) exposed to reactor coolant is managed by the Water Chemistry Program and the Final Safety Analysis Report (FSAR) Supplement commitment to (1) participate in industry RVI aging programs, (2) implement applicable results, and (3) submit for NRC approval >24 months before the extended period an RVI inspection plan, based on industry recommendation. The LRA uses Note E, along with a plant-specific note that states "Consistent with the GALL Report for material, environment and aging effect, but a different aging management program is credited or the GALL Report identifies a plant specific aging management program.
Clarify why the Water Chemistry Program with the commitment, which is different from the recommendation of the GALL Report, is adequate to manage the aging effect of the CASS components.
PG&E Response to RAI 3.1.2.1-1 In license renewal application (LRA) Table A4-1, "License Renewal Commitments,"
PG&E commits to the following activities for managing the aging effects of reactor vessel internals (RVI) components: (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, PG&E will submit an inspection plan for reactor internals to the NRC for review and approval.
PG&E Letter DCL-10-098 Page 4 of 5 EPRI's Material Reliability Program (MRP) has been in the process of developing guidance on managing RVIs, including cast austenitic stainless steel (CASS) components susceptible to thermal aging and neutron irradiation embrittlement. PG&E Pressurized Water Reactor (PWR) Vessel Internals Program will rely on the results of the ongoing MRP initiative to develop a comprehensive aging management program (AMP) for PWR reactor internals. As indicated in the EPRI report, MRP-227, the screening process includes the evaluations of aging effect of loss of fracture toughness in the CASS RVIs components as a result of thermal aging embrittlement and neutron irradiation embrittlement.
Implementation of the PWR Vessel Internals Program will provide reasonable assurance that aging effects will be managed such that components within the scope of this program will continue to perform their intended function(s) during the period of extended operation and is therefore an acceptable substitution for NUREG-1 801, AMP XI.M1 3, "Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)."
LRA Table 3.1.2-1, "Reactor Vessel, Internals, and Reactor Coolant System -
Summary of Aging Management Evaluation - Reactor Vessel and Internals," has been revised to remove water chemistry as part of the AMP and to indicate that the loss of fracture toughness of CASS reactor vessel internal (RVI) lower core support structure (core support casting (Ul)) and RVI upper core support structure (upper support columns) exposed to reactor coolant is managed by the Final Safety Analysis Report supplement commitment to: (1) participate in industry RVI aging programs; (2) implement applicable results; and (3) submit for NRC approval greater than 24 months before the extended period an RVI inspection plan, based on industry recommendation.
See revised LRA Tables 3.1.1 and 3.1.2-1 in Enclosure 2.
PG&E Letter DCL-10-098 Page 5 of 5 RAI 3.5.2.2.1.7-1 DCPP LRA Section 3.5.2.2.1.7 addresses the further evaluation of cracking due to stress corrosion cracking (SSC) of stainless steel penetration sleeves, penetration bellows, and dissimilar metal welds, as not applicable to DCPP.
LRA Section 3.5.2.2.1.7 indicates that this further evaluation is not applicable because DCPP has no in-scope stainless steel penetration sleeves, penetration bellows, or dissimilar metal welds subject to stress corrosion cracking. The FSAR, on Page 3.8-33, indicates that penetration sleeves for containment structures are made of carbon steel and that the flued heads, which appear to be welded to the penetrations sleeves, are made of stainless steel as shown in FSAR Figure 3.8-6. The staff noted that this information of the FSAR potentially contradicts the basis of the applicant's claim that the further evaluation described in LRA Section 3.5.2.2.1.7 is not applicable to DCPP.
Clarify why there are no in-scope stainless steel penetration sleeves, penetration bellows, or dissimilar metal welds subject to SCC, taking into consideration the stainless steel flued heads which are apparently welded to the penetration sleeves.
PG&E Response to RAI 3.5.2.2.1.7-1 Diablo Canyon Power Plant (DCPP) aging evaluation of the flued head dissimilar welds on containment penetrations concludes that the design and plant configuration preclude a corrosive environment from affecting the dissimilar welds. Therefore the welds should not be subject to stress corrosion cracking and do not require an aging management program. Final Safety Analysis Report, Figure 3.8-6 shows the configuration of the penetration sleeves. Inside containment the environment is plant indoor air, which is non-corrosive. Also the welds are protected from spray and drip by a leak chase, which was installed for testing purposes during plant construction. On the auxiliary building side, the flued head welds are isolated by insulation and a fire protection metal cover.
Although the configuration on the auxiliary building side is not hermetically sealed, it does limit the free transfer of air from the auxiliary building environment to the dissimilar metal welds. The dissimilar metal welds are tested during the integrated leak rate test, which was recently conducted in 2008 for Unit 2 and 2009 for Unit 1. DCPP operating experience has not shown any failures of dissimilar metal welds for this unique plant configuration.
PG&E Letter DCL-10-098 Page 1 of 3 LRA Amendment 9 LRA Section RAI Table 3.1.1 RAI 3.1.2.1-1 Table 3.1.2-1 RAI 3.1.2.1-1 PG&E Letter DCL-10-098 Page 2 of 3 Section 3.1 AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.1 Summary of Aging Management Evaluations in Chapter IV of NUREG-1801 for Reactor Vessel, Internals, and Reactor Coolant System (continued)
Item Component Type Aging Effect / Mechanism Aging Management Further Discussion Number Program Evaluation Recommended 3.1.1.80 Cast austenitic Loss of fracture toughness due Thermal Aging and Neutron No Consistent with stainless steel reactor to thermal aging and neutron Irradiation Embrittlement of NUREG-1 801 for material, vessel internals (e.g.,
irradiation embrittlement CASS (B2.1.39) environment, and aging upper internals effect, but a different aging assembly, lower management program WVate internal assembly, Chemistry (B2.1.2) aRd CEA shroud FSAR supplement assemblies, control rod commitment to (1) participate guide tube assembly, in industry RVI aging core support shield programs (2) implement assembly, lower grid applicable results (3) submit assembly) for NRC approval > 24 months before the extended period an RVI inspection plan based on industry recommendations is credited.
PG&E Letter DCL-10-098 Page 3 of 3 Section 3.1 AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-1 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -
Pn tr'trr lt/,, ccd nnd Inf,*rnmIl Component Type Intended Material Environment Aging Effect Aging Management Program NUREG-Table 1 Notes Function Requiring 1801 Vol.
Item Management 2 Item RVI Lower Core SS Stainless Reactor Loss of ate.r Chemistry (12.1.2) and IV.B2-21 3.1.1.80 E, 3 Support Structure Steel Cast Coolant (Ext) fracture FSARsupplement commitment (Core Support Austenitic toughness to (1) participate in industry RVI Casting (Ul))
aging programs (2) implement applicable results (3) submit for NRC approval > 24 months before the extended period an RVI inspection plan based on industry recommendation RVI Upper Core SS Stainless Reactor Loss of
^
-atcr Chemistry (12.1.2) and IV.B2-37 3.1.1.80 E, 3 Support Structure Steel Cast Coolant (Ext) fracture FSAR supplement commitment (Upper Support Austenitic toughness to (1) participate in industry RVI Columns) aging programs (2) implement applicable results (3) submit for NRC approval >.24 months before the extended period an RVI inspection plan based on industry recommendation