DCL-10-167, Response to NRC Letter Dated December 20, 2010, Request for Additional Information (Set 36) for the Diablo Canyon License Renewal Application

Response to NRC Letter Dated December 20, 2010, Request for Additional Information (Set 36) for the Diablo Canyon License Renewal Application

ML110130425
Person / Time
Site:	Diablo Canyon
Issue date:	01/12/2011
From:	Becker J Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-10-167
Download: ML110130425 (41)

Text

Pacific Gas and Electric Company James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601

p. O. Box 56 Avila Beach, CA 93424 805.545.3462 Internal: 691.3462 January 12, 2011 Fax: 805.545.6445 PG&E Letter DCL-1 0-167 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 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 December 20, 2010, Request for Additional Information (Set 36) 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 December 20,2010, 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 Enclosure 1.

LRA Amendment 31 resulting from the responses is included in Enclosure 2 *showing the changed pages with line-in/line-out annotations. Enclosure 3 contains a list of all reactor coolant pressure boundary valves and their design codes.

If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4160.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 12, 2011 .

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway . Comanche Peak. Diablo Canyon. Palo Verde . San Onofre . South Texas Project. Wolf Creek

Document Control Desk PG&E Letter DCL-1 0-167 January 12, 2011 Page 2 tlg/50363556 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, Licensing Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway. Comanche Peak. Diablo Canyon. Palo Verde. San ,Onofre

• South Texas Project. Wolf Creek

Enclosure 1 PG&E Letter DCL-10-167 Page 1 of 17 PG&E Response to NRC Letter dated December 20, 2010 Request for Additional Information (Set 36) for the Diablo Canyon License Renewal Application RAI 4.1-6

Background:

License renewal application (LRA) Section 4.3.2.6, "Absence of a TLAA for Reactor Coolant System Boundary Valves," provides the applicant's basis for its conclusion that the current licensing basis (CLB) for the safety-related valves in the reactor coolant pressure boundary (RCPB) valves does not include any analyses that need to be identified as Time-limited Aging Analyses (TLAA) for the LRA under the TLAA identification criteria in 10 CFR 54.3. Final Safety Analysis Report (FSAR) Table 5.2-9 provides the list of applicable RCPB valves. FSAR Table 5.2-2 identifies that the applicable design codes and standards for the reactor coolant pressure boundary valves are: (1) "USAS B16.5," (2) MSS-SP-66; (3) "ASME III 68," or (4) "ASME III 74."

Additionally, the review of the CLB indicates that some of the RCPB values may have been designed to one or more of the following additional code and standards not currently reflected in FSAR Table 5.2-2: (1) ANSI B31.7 [several editions listed]; (2)

ASME Boiler and Pressure Vessel Code,Section III, 1966 Edition; (3) ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition, inclusive of 1973 Addenda; (4) for Target Rock Head Vent Valves, ASME III, Class II, 1977 Edition; (5) draft ASME Pump and Valve Code for Nuclear Power Plants, 1968 Edition; and (6) ASME Code Section III, 1986 Edition.

FSAR Table 5.2-2 identifies that the design code for the reactor coolant system safety valves is the 1965 Edition of ASME Code Section, III, Article 9, and that the design code for the reactor coolant system relief valves is USAS B16.5 (edition not specified).

Issue and Requests:

The information and basis in LRA Section 4.3.2.6 does not give the staff a sufficient basis for verifying that there is no need for any TLAAs to be identified in the LRA for the RPCB valves based on the following observations:

Issue 1:

FSAR Table 5.2-9 identifies the valves that are applicable to the RCPB design. The table does not identify which specific design code was used for the design, design analysis, procurement, and fabrication of each RCPB valve that was listed in the table.

In addition, FSAR Table 5.2-2 only identifies the codes and standards that are applicable to the RCPB valves based on a commodity grouping, not on an individual RCPB valve basis. In addition, DCPP has not provided the staff with access to the specific design specifications that were used for the design stress analyses of the RCPB valves that are listed in FSAR Table 5.2-9. Thus, the staff is unable to verify

Enclosure 1 PG&E Letter DCL-10-167 Page 2 of 17 (based on the current information) whether the design code for a given RCPB valve required a time-dependent fatigue analysis based on its design code and its nominal valve size.

Request 1:

Clarify whether FSAR Table 5.2-9 provides a comprehensive list of all Class 1 or Class A valves in the RCPB. For each Class 1 or Class A valve in the RCPB, identify which design code or standard was designated in the owner's design specification for the valve's design stress analysis. For each valve: (1) identify whether the code used for the valve's design analysis included a cycle dependent cumulative usage factor (CUF) analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cool down cycles), or a maximum allowable stress reduction analysis, and if so, (2) summarize the criteria in the code that would call for a given valve to be included within the scope of the code's fatigue analysis criteria.

Issue 2:

LRA Section 4.3.2.6 states that ASME III, Article 9, did not require a time dependent analysis. FSAR Table 5.2-2 appears to appropriately indicate that the reactor coolant safety valves were procured to 1965 Edition of ASME Code Section III, Article 9. Staff review of this code has determined that the code is only applicable to the design of vessel components, and that ASME III, Article 9, is limited only to the application of the low-pressure overpressure protection (LTOP) system setpoints associated with these valves. Article 9 in this code clearly identifies that the remaining design rules and aspects for the valves are to be done in accordance with other applicable standards or codes. Thus the applicant has not provided a clear basis on which code or standard was used to perform the design stress analysis for these safety valves or whether the designated Code or Standard required either a CUF or It type explicit fatigue analysis or an implicit fatigue analysis (i.e., maximum allowable stress range reduction analysis, as might be required by ANSI B31.1 or B31.7).

Request 2:

Clarify which design code was used for the design stress analysis of the 6-inch nominal size reactor coolant system safety valves. Identify whether the specific code required the valve to be within the scope of a cycle dependent CUF analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Issue 3:

FSAR Table 5.2-2 indicates that some of the Class 1 or Class A valves in the RCPB were procured to ASME Code Section III, 1968 Edition. However, there appears to be an inconsistency in the design basis information in that table because the design requirements in the 1968 Edition of the ASME Code Section III, Subarticle NB, appear to be limited only to vessel components and do not appear to be applicable to Class 1 or Class A valves in the RCPB. Thus, it is not evident how some of the valves in the RCPB could have been procured to ASME Code Section III 1968 Edition or which

Enclosure 1 PG&E Letter DCL-10-167 Page 3 of 17 design code was used for the stress analysis of the valves and whether the code or standard for the stress analysis required a cycle dependent CUF analysis. It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Request 3: Clarify whether, consistent with the information in FSAR Table 5.2-2, any Class 1 or Class A valves in the RCPB have been designed to the design requirements (including design stress requirements and cyclical fatigue analysis requirements) in the 1968 Edition of the ASME Code Section III. If so, justify the basis for using a vessel-related Code for the design, fabrication, analysis, and procurement of a given Class 1 or Class A valve in the RCPB, and clarify, with an explanation and justification, whether or not the cyclical metal fatigue analysis in Section N-415 of the Code would have been required to have applied as part of the stress analysis for the valves procured to this ASME Code Section III edition.

Issue 4:

The staff has determined that the some of the small bore Class 1 or Class A valves in the RCPB have been designed, fabricated, analyzed, and procured to a 1968 Draft ASME Code for Pumps and Valves for Nuclear Power Code and that Sections 452 and 454 of this Code include applicable time-dependent cyclic or fatigue assessment criteria for pumps and valves designed and procured to this code. Specifically, Section 454 of the Code has a It parameter metal fatigue analysis (cycling loading analysis) that is similar to the type of CUF analysis that is required for ASME Code Class 1 or Class A components in ASME Section III Article NB-3200 requirements or N-415 requirements for older versions of ASME Section III. The staff has verified that Section 142 of this Code identifies that the fatigue analysis requirements in Section 452 and 454 would need to be performed only if the inlet nozzle size for the Class 1 pump or valve was greater than 4 inches diameter nominal pipe size. However, Section 410 of the Code qualifies this somewhat by stating the Code's Chapter 4 procedures and analyses (including those in Sections 452 and 454) would need to be performed for small bore pumps or valves (i.e., for those pump or valves with inlet nozzles less than or equal to 4 inches in nominal pipe size) if the owner's design specification for a given small bore pump or valve specified this need, as determined by the owner. Thus, there could be circumstances where a small-bore pump or valve could be within the scope the Code's fatigue assessment criteria (Section 452) and cyclical loading assessment criteria (Code Section 454).

Request 4:

Clarify the review and steps that DCPP took to confirm whether or not the owner's design specification for a small bore Class 1 or Class A valve designed to the 1968 Draft ASME Code for Pumps and Valves for Nuclear Power Code had designated the valve for analysis pursuant to the Code's It fatigue analysis criteria. Identify all small bore Class 1 or Class A valves that were designed to the 1968 Draft ASME Code for Pumps and Valves for Nuclear Power Code and were permitted to be exempted from the It analysis based on the exemption criteria in Section 410 of this Code based and

Enclosure 1 PG&E Letter DCL-10-167 Page 4 of 17 their nominal valve inlet size. In addition, identify all small bore Class 1 or Class A valves (if any) that were designed to this draft Code for which the owner had gone beyond the small bore fatigue exemption criteria in Section 410 of the Code and had specifically designated the time-dependent It analysis to be performed in the owner's design specification for a given small bore Class 1 or Class A valve.

Issue 5:

The staff has determined that FSAR Table 5.2-2 indicates that USAS B16.5 is designated as an appropriate design code for specific small bore and large Class 1 or Class A valves in the RCPB. However, the staff has noted that the scope of USAS B16.5 only is limiting to the following valve design and quality activities: (1) pressure-temperature ratings; (2) size and methods for designated openings; (3) markings; (4) minimum requirements for valve material selection; (5) valve dimensions; (6) valve tolerances; and (7) valve hydrostatic test criteria. The staff has noted that the scope of USAS B16.5 does not appear to include design stress analysis criteria for Class 1 or Class A valves in the RCPB. Thus, for a given Class 1 or Class A valve procured to the USAS B16.5 pressure-temperature rating criteria, it is not evident which design codes (if any) were used to perform the design stress analyses for the specific valve, and if applicable, whether the code used for the design stress analysis required either a cycle dependent CUF analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Request 5:

For each Class 1 or Class A valve that was procured to USAS B16.5 pressure-temperature rating criteria, identify the code or standard (if any) that was used to perform the design stress analysis for the procured valve, and if applicable, clarify whether the design code or standard used for the stress analysis of the valve required the valve to be analyzed in accordance with either an applicable cycle-dependent CUF analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Issue 6:

The staff has determined that some of the Class 1 or Class A valves in the RCPB piping subsystems were designed to either ANSI B31.1 or B31. 7 design. LRA Section 4.3.5 identifies that the implicit fatigue analyses (i.e., maximum allowable stress reduction analyses) for piping, piping components, and piping elements in these subsystems are analyses that meet the definition of a TLAA in 10 CFR 54.3. The staff has determined that the scope of components in piping systems designed to ANSI B31.1 code criteria includes applicable valves in the systems. Thus, it is not evident to the staff why Class 1 or Class A valves in portions of the RCPB designed to ANSI B31.1 or B31. 7 criteria would not be within the scope of the ANSI B31.1 or B31.7 stress analysis criteria or the implicit fatigue analysis criteria in these codes.

Enclosure 1 PG&E Letter DCL-10-167 Page 5 of 17 Request 6:

Identify all Class 1 or Class A valves in the RCPB that were designed to ANSI B31.1 stress analysis criteria and all Class 1 or Class A valves in the RCPB that were designed of ANSI B31.7 stress analysis criteria. For those Class 1 or Class A valves procured to these design codes, clarify, with a justified explanation, on whether the implicit fatigue analysis in these Codes are applicable to any Class 1 or Class A valves that are procured to these design code criteria, and if so, justify whether or not the implicit fatigue analyses performed on the subsystems containing the valves need to be identified as a TLAA for the DCPP LRA.

Issue 7:

The staff has determined that the applicant has indicated that, based on its current review of the CLB, there were some small bore Class 1 or Class A valves (less than or equal to 4 inches nominal size) in the RCPB where the applicant could not determine which the design code or standard was used for the design, analysis, fabrication, and procurement of the valves, but where the applicant indicated there would not be any associated fatigue-related TLAAs based on their size. Presumably, these valves are valves in the RCPB and Safety and Seismic Class 1 valves. Thus, pursuant to 10 CFR Part 50, Appendix B, Criterion III, Design Control, the NRC would have required that these valves be within the scope of appropriate design standards (or Codes). Thus, it is not evident to the staff why these valves would not have been required to be within the scope of applicable design codes or standards, including those governing the stress analyses for such valves. Thus, without further clarification, the staff cannot determine whether these valves were procured to appropriate design codes or standards, and if so, whether the given code or standard for a valve in this category would have required the valve to be analyzed in accordance with either a cycle-dependent CUF analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Request 7:

Identify the design codes or standards that were used for the design of these valves in order to comply with the provision in 10 CFR Part 50, Appendix B, Criterion III, Design Control, that states that the design measures shall include: "provisions to assure that appropriate design standards are specified and included in design documents ..." For each valve in this category, identify whether the design code or standard used (if any) for the design stress analysis of the valve required either a cycle-dependent CUF analysis, It analysis (similar to CUF except the analysis only considers cyclical stresses imposed to heat/cooldown cycles), or a maximum allowable stress reduction analysis.

Justify your basis for concluding that the CLB does not include any fatigue related analyses in the CLB that meet the definition of a TLAA in 10 CFR 54.3.

Enclosure 1 PG&E Letter DCL-10-167 Page 6 of 17 PG&E Response to RAI 4.1-6 Response to Request 1:

DCPPs licenses pre-date the establishment of the Class 1 or Class A designations per ASME Section III for reactor coolant pressure boundary (RCPB) valves. The DCPP RCPB valves were based on classification as Safety Class 1 per the 1970 draft of ANSI N-18.2. The definition of Safety Class 1 per ANSI N-18.2 applies to Reactor Coolant System components whose failure could cause a loss of reactor coolant inventory in excess of that which can be made up with normal reactor coolant makeup and prevents an orderly reactor shutdown.

FSAR Table 5.2-9 lists valves between major components in the main RCPB process lines. Table 1 in Enclosure 3 provides a comprehensive list of the codes used in the valve design for all currently existing RCPB valves.

The DCPP valves were purchased using USAS B16.5 for pressure rating, except for the 14-inch valves that were rated per MSS SP-66. Many small bore valves (<4 inches) have been replaced with valves designed to 1974 edition and later editions of ASME Section III, Class 1 as allowed by ASME Section XI Repair/Replacement rules and reconciled to the design code of record. None of the codes include a cycle dependent cumulative usage factor analysis, It analysis, requirement for a valve of the size for which it is identified or maximum allowable stress reduction analysis.

Response to Request 2:

The reactor coolant safety valves were procured to the 1968 Edition of ASME III, Article 9. The 1968 Edition of ASME III, Article 9 is the basis for capacity certification and application of ASME "NV" Code Symbol Stamping on the safety relief valves for overpressure protection.

In accordance with the original DCPP specification, the valves were designed to the 1968 Edition of ASME Section III, Section VIII and USAS B16.5. ASME Section III, Article 9 and ASME Section VIII were used for overpressure protection of the vessels with the use of a safety relief valve, and USAS B16.5 is used for inlet outlet flange rating. These codes do not require a fatigue analysis based on a cyclic operation for safety relief valves or maximum allowable stress reduction analysis.

Response to Request 3:

FSAR Table 5.2-2 refers to valves in the reactor coolant pressure boundary (RCPB) that were designed to ASME Code Section III, 1968 Edition. In these cases, the ASME Code Section III, 1968 Edition is the 1968 Draft ASME Code for Pumps and Valves. No RCPB valves at DCPP used a vessel-related Code for the design, fabrication, analysis, and procurement.

Enclosure 1 PG&E Letter DCL-10-167 Page 7 of 17 Response to Request 4:

Table 1 in Enclosure 3 identifies the valves within the reactor coolant pressure boundary. Valves purchased to the 1968 Draft ASME Code for Pumps and Valves apply to valves with inlet piping connections of 4-inches nominal pipe size and smaller and may be designed by any method that has been demonstrated to be satisfactory for the specified design. All valves at DCPP identified as being designed to the 1968 Draft ASME Code for Pumps and Valves are 4 inches or less. These valve models were shown to meet the USAS B16.5 design rating and this standard does not require a cycle dependent CUF analysis requirement.

PG&E reviewed the original design specification for these valves and discussed them with Westinghouse who owner accepted the valves during the construction. The review determined that no RCS pressure boundary valve specification designated the time-dependent It analysis.

Response to Request 5:

For valves that were procured to USAS B16.5, USAS B16.5 constitutes the complete design basis for which valves were analyzed in the current licensing basis. As stated in LRA Section 4.3.2.6, valves purchased to USAS B16.5 did not impose the requirement for cyclic analysis or maximum allowable stress reduction analysis.

Response to Request 6:

DCPP reactor coolant pressure boundary piping systems are designed and analyzed to USAS B31.1 and B31.7, which includes pipe, flanges, bolting, gaskets, valves, relief devices, fittings and the pressure containing parts of other piping components. USAS B31.1 and B31.7 cite the codes and standards that are acceptable for valve design in B31.1 piping, which include ASME Codes, USAS B16.5, and MSS SP-66. These design codes and standards do not require the use of a maximum allowable stress reduction factor. The codes and standards that apply to the individual valves are identified in Enclosure 3. Therefore, valves at DCPP designed to ANSI B31.1 or B31.7 do not use an implicit fatigue analysis (i.e., a maximum allowable stress reduction factor) or other time dependent analysis.

Response to Request 7:

Table 1 in Enclosure 3 identifies all reactor coolant pressure boundary (RCPB) valves and their design codes. The governing codes and standards applicable for the design of RCPB valves are:

USAS B16.5 MSS SP-66 1968 Draft ASME Code for Pumps and Valves for Nuclear Power ASME Section III, 1974 Edition and later for replacement valves These codes and standards do not require a cycle dependent CUF analysis, It analysis, for a valve of the size for which it is identified, or maximum allowable stress reduction analysis. The specifications were reviewed and determined not to contain any

Enclosure 1 PG&E Letter DCL-10-167 Page 8 of 17 requirement for a cycle dependent CUF analysis, It analysis, for a valve of the size for which it is identified, or maximum allowable stress reduction analysis. Because no time dependent aging analyses were performed or required for the DCPP RCPB valves, 10 CFR 54.3(a) Criterion 2 is not satisfied and there are no Time-Limited Aging Analyses that support the design of the DCPP RCPB valves.

Enclosure 1 PG&E Letter DCL-10-167 Page 9 of 17 RAI 4.1-7 The applicant includes its TLAA for the reactor vessel internal (RVI) core support structure components in LRA Section 4.3.3. The applicant stated that the CUF analysis for the baffle-former bolts originally calculated a CUF value less than the design limit of 1.0. The applicant also stated that the adequacy of the baffle-former bolt is an industry issue and the design analyses and evaluations may not currently be sufficient to support their initial safety determination. The applicant stated that the baffle-former bolt analyses will be addressed by participation in industry level initiatives.

The applicant is currently an active member of the Electric Power Research Institute's Materials Reliability Program (EPRI MRP) and for license renewal, the applicant has committed to participating in the EPRI MRP activities to ensure the structural integrity of Westinghouse designed RVI components, including Westinghouse designed baffle bolts and former bolts. Specifically, the staff confirmed that the applicant's FSAR Supplement Table Commitment No. 22, Part B, states the following:

B. For Reactor Vessel Internals:

(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.

The current industry-wide program for Westinghouse designed facilities is defined in MRP-227 and that the industry-wide initiatives include appropriate measures to perform ultrasonic testing (UT) inspections of Westinghouse baffle and former bolts for evidence of either stress induced or fatigue induced cracking. The staff noted that this is consistent with Standard Review Plan License Renewal (SRP-LR) Sections 3.1.2.2.15 and 3.1.2.2.17.

Issue:

The staff noted that the applicant is taking the position that the CUF calculation for the baffle bolts no longer serves a safety basis and the CUF analysis for the bolts does not need to be identified as a TLAA because the analysis did not meet 10 CFR 54.3, Criterion 4. The staff noted that the CUF calculation for the baffle and former bolts was required to meet the 1968 Edition of the ASME Code Section III, Article NG CUF calculation requirements for core support components in the reactor vessel. The staff further noted that the fact that the EPRI MRP is currently investigating industry initiatives to inspect for cracking in these components, does not invalidate the applicant's CLB or design basis that required a CUF calculation for these components.

Enclosure 1 PG&E Letter DCL-10-167 Page 10 of 17 Request:

Explain why the performance of the required design CUF calculation for the baffle bolts does not satisfy 10 CFR 54.3, Criterion 4, the analysis was used in a safety basis decision. Justify why the CUF analysis for the baffle bolts does not need to be identified as TLAA for the LRA.

PG&E's Response to RAI 4.1-7:

The cumulative usage factor calculation for the baffle and former bolts is considered a Time-Limited Aging Analyses. The Reactor Vessel Internals Aging Management Program (as described in LRA Table A4-1, Commitment 22) will ensure that fatigue is adequately managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii). See amended LRA Sections A3.2.2, 4.3.3, and Table 3.1.2-1 in .

Enclosure 1 PG&E Letter DCL-10-167 Page 11 of 17 RAI B2.1.21-1 (follow-up)

Background:

Generic Aging Lessons Learned (GALL) aging management program (AMP) XI.M.37 program element "acceptance criteria" states, in part:

The acceptance criteria will be technically justified to provide an adequate margin of safety to ensure that the integrity of the reactor coolant system pressure boundary is maintained. The acceptance criteria will include allowances for factors such as instrument uncertainty, uncertainties in wear scar geometry, and other potential inaccuracies, as applicable, to the inspection methodology chosen for use in the program. Acceptance criteria different from those previously documented in NRC acceptance letters for the applicant's response to Bulletin 88-09 and amendments thereto should be justified.

In LRA Section B2.1.21, "Flux Thimble Tube Inspection" Program, the applicant states that its program is an existing program that is consistent with the recommended program element criteria in GALL AMP XI.M.37, "Flux Thimble Tube Inspection." By letter dated July 14, 2010, the staff issued request for additional information (RAI)

B2.1.21-1, and requested that the applicant clarify its basis for the AMP's through-wall wear acceptance criterion and clarify how sources of instrument measurement and wear scar uncertainties and inaccuracies are accounted for in the AMP, as recommended in both GALL AMP XI.M37 and NRC Bulletin 88-09. In its response dated August 12, 2010, the applicant stated that the AMP's current through-wall wear acceptance criterion basis was established in the February 1991 revised inspection procedure for the AMP, which set the acceptance criterion at 68% of the nominal thimble tube wall thickness.

However, the applicant also explained that the updated procedure eliminated the application of the applicant's prior 10% uncertainty adjustment on the nondestructive examination (NDE) estimate, as was made based on the applicant's steps to confirm the accuracy of the program's NDE testing methods in during Unit 1 refueling outage (RO) 1 R4 and the applicant's review of the generic Westinghouse acceptance criteria bases in Westinghouse Proprietary Class 2 Report WCAP-12866, which was issued in January 1991.

The applicant's response letter of August 12, 2010, indicates that the applicant eliminated application or accounting for any source of measurement uncertainty and wear rate estimation uncertainty in the program elements for the AMP. However, Westinghouse Class 2 Proprietary Report review of WCAP-12866 does include an appropriate allowance for the wall thickness acceptance criterion that is recommended in the generic report, and this appears to satisfy the need to account for appropriate uncertainties generic flux thimble tube program report, as recommended in NRC Bulletin 88-09 and in GALL AMP XI.M37.

Issue: The staff has determined that the current DCPP Flux Thimble Tube Program does not include any uncertainty allowances in the program, even though the applicant

Enclosure 1 PG&E Letter DCL-10-167 Page 12 of 17 has set acceptance criterion for the AMP to a value that is more conservative than that recommended for these types of programs in the Westinghouse report. This does not appear to conform to recommendation in either NRC Bulletin 88-09 or in the "monitoring and trending" program element of GALL AMP XI.M37 which state that these types of programs should include appropriate allowances for instrument measurement and wear scar uncertainties. In addition, DCPP's elimination of appropriate instrument measurement and wear scar uncertainties from the scope of the AMP may be non-conservative when taken in light of the flux thimble tube wear data for Unit 2 thimble tube L13, as obtained from eddy current inspections of the tube during Unit 2 ROs 2R11, 2 R12, and 2R13, and in light of the fact that this tube leaked within 4 months of returning to power operations out of RO 2R13. Specifically, the wear data obtained from the inspections of Unit 2 tube L13 indicate that the wear in the tube might have been occurring at an increasingly non-linear rate. Thus, the staff finds that the applicant's decision to eliminate appropriate instrument measurement uncertainties and wear scar uncertainties for the scope of the AMP is out of conformance of the recommendations of the applicable NRC bulletin and GALL AMP, and may not be conservative relative to relevant thimble tube operating experience for the facility.

Request: In light of the respective operating experience for Unit 2 thimble tube 2L13, justify the basis for not including an appropriate margin term to account for NDE measurement and wear scar uncertainties in either the wear projection basis for the AMP or accounting for them in the acceptance criterion for the AMP, and provide a basis for not identifying this as an appropriate exception to the "acceptance criteria" program element in GALL AMP XI.M37, "Flux Thimble Tube Inspection."

PG&E's Response to B2.1.21-1 (follow-up)

PG&E has maintained a trending program for the flux thimble tubes. This trending includes comparisons of PG&E's wear projection methodology to the methodology used in WCAP-12866 with DCPP site specific wear data. It also contains comparisons between values predicted by the calculations for the following cycle eddy current inspection and the actual Non Destructive Examination (NDE) measurements made in that Refueling Outage (RFO). These comparisons substantiate that the PG&E projection methodology is conservative compared to the WCAP criteria of 80 percent.

DCPP contributed thimble tubes from the 1R3 RFO along with copies of the in-situ eddy current exam results to Westinghouse as part of the WCAP-12866 program, which were measured both mechanically and by in shop eddy current testing. These results led to the development of eddy current testing standards for the whole industry.

WCAP-12866 states in the Abstract and the Conclusions that "Comparison of eddy current measurements of the thimble wear scars to those measured by mechanical means showed that eddy current identifies the scar as being as deep, or deeper than it actually is. This result indicates that it is not necessary to add any uncertainty to the eddy current indications." In addition WCAP-12866 states: "Based on thimble segment

Enclosure 1 PG&E Letter DCL-10-167 Page 13 of 17 collapse tests it was concluded that the thimbles have a high residual strength even when subject to wall loss on the order of 90% percent. The thimbles will retain their functional and structural integrity with up to 85% wall loss for all plant operating modes."

They go on to state that, "For conservatism, a wall loss of 80% should now be used to determine when thimble action is required (i.e. repositioned, replaced, etc.)."

PG&E considers that the acceptance criteria of 68 percent includes an extra 17.5 percent margin compared to the 80 percent limit recommended in the WCAP.

PG&E considers the 17.5 percent margin, in addition to the enhanced acceptance criteria, to be an acceptable margin term to account for NDE measurement and wear scar uncertainties applied to the calculations of predicted wear scar growth. See the following response to RAI B2.1.21-2 (follow-up).

Therefore, PG&E believes that the acceptance criteria are comprehensive and conservative.

Enclosure 1 PG&E Letter DCL-10-167 Page 14 of 17 RAI B2.1.21-2 (follow-up)

Background:

The GALL AMP XI.M37 program element, "monitoring and trending," states, "[t]he wall thickness measurements will be trended and wear rates will be calculated. Examination frequency will be based upon wear predictions that have been technically justified as providing conservative estimates of flux thimble tube wear." The GALL AMP recommends that the "interval between inspections" should be established "such that no flux thimble tube is predicted to incur wear that exceeds the established acceptance criteria before the next inspection."

The "operating experience" (OE) program element for the Flux Thimble Tube Program discussed the impacts of a leak that occurred in thimble tube L13 of DCPP Unit 2 in 2006. This leak occurred at normal operating pressure with no prior warning or expectation, and occurred within four months of returning to power operations out of Unit 2 RO 2R13 and repositioning corrective actions that were implemented on that tube during the RO. In its August 12, 2010, response to RAI B2.1.21-3, the applicant added a "License Renewal Commitment" to preclude repositioning a tube more than once (without capping or replacing).

Issue:

The OE and related observations on plant-specific wear rate projections do not conform to, or meet the intent of, the GALL AMP "monitoring and trending" program element. As noted in RAI B2.1.21-2, issued by letter dated July 14, 2010, the "incremental wear" and "cumulative wear" projection methods as implemented in the applicant's AMP do not provide conservative wear projection because they do not account for possible accelerating wear nor do they account for uncertainty in the method of wear projection.

Neither the OE discussion (for L13 event in 2006) nor the applicant's response to RAI B2.1.21-3 identified the apparent cause (aging mechanism) of the degradation in Unit 2 thimble tube L13, or explained why the leak occurred so soon after returning to power operations, even after indicating repositioning (corrective action) of the tube during RO 2R13.

The wear history of several flux thimble tubes, including Unit 2 thimble tube L13, indicates that the wear in the tubes may be occurring at an increasingly accelerated wear rate, and in other instances, repositioning of the tubes appears to have moderated the wear rate increase. The applicant has not addressed whether cracking could have been a main contributing factor in the rapid-time failure of Unit 2 thimble tube L13. Thus, multiple repositioning of tube L13 may not be the only feasible explanation for the rapid failure in the tube, and the staff is concerned that either rapidly progressing wear, rapidly propagating cracking, or rapidly propagating wear coupled to cracking may have been the main contributing factor for the leak in Unit 2 thimble tube L13 during Unit 2 operating cycle 14.

Enclosure 1 PG&E Letter DCL-10-167 Page 15 of 17 Request:

1. Identify the quality activities that DCPP takes to identify and confirm the apparent cause of age-related degradation that is detected in a DCPP flux thimble tube, and identify all age-related degradation effects and mechanisms (including any cracking and its mechanisms, if applicable) that have been detected in the DCPP flux thimble tubes to date.

2. Describe how the trending of thimble tube wear rates accounts for the possibility of a non-linear or accelerating wear rate.

3. Identify all aging effects and mechanisms that contributed to the degradation in Unit 2 flux thimble tube L13 over time (Le., as detected during ROs 2R11, 2R12, and 2R13) and discuss the failure analysis activities that were performed at the site or were contracted out to confirm the apparent cause of the degradation that had occurred in the tube and the rapid progression of the degradation mechanism that lead to the relative rapid leak in 2006 (i.e., the leak occurred within four months of returning to power).

4. Provide your basis for concluding that the "monitoring and trending" activities, "acceptance criteria" and "corrective action" criteria for the Flux Thimble Tube Program will be capable of detecting degradation in a flux thimble prior to the occurrence of a through-wall failure.

5. If aging effects other than wear were determined to have occurred in tube L13 or any other thimble tube, describe how these other aging effects will be managed by the Flux Thimble Tube Program.

PG&E's Response to B2.1.21-2 (follow-up)

1) The quality activities that DCPP takes to identify and confirm the apparent cause of flux thimble tube (FTT) degradation include the eddy current examinations and reports, the Surveillance Test Procedure, STP R-22, "Thimble Tube Inspection Program", and the DCPP Corrective Action Program. PG&E has performed 100 percent eddy current testing in every outage since 1R3/2R3. The only degradation mechanism that has been observed is wear scars caused by flow induced vibration. In 2R14, FTT L-13 was inspected by eddy current and showed a through wall flaw but no evidence of cracking.

Cracking has not been detected in any thimble tube eddy current exam.

PG&E performed an evaluation after the failure of Unit 2 FTT L-13. The evaluation determined that the cause was flow induced vibration magnified by multiple repositionings which caused destabilization of the tube when moving the 15-inch chrome band below the core plate. This moved the hardened chrome band below the wear locations of the fuel bottom nozzle and the core plate. In 2R14, FTT L-13

Enclosure 1 PG&E Letter DCL-10-167 Page 16 of 17 was inspected by eddy current and showed a through wall flaw at the location of the fuel bottom nozzle.

2) The possibility of non-linear or accelerating wear rates was addressed by STP R-22 FTT enhanced acceptance criteria. This includes consideration of the corrective actions taken to address the through wall flaw in Unit 2 FTT L-13 to cap or replace any tube with:

a. Any wear scar with a growth rate greater than 25 percent per year,

b. Any two wear scars greater than 40 percent,

c. and to limit the total repositioning of any tube to a maximum of 6 inches.

Any of these three new STP R-22 FTT acceptance criteria would have caused Unit 2 FTT L-13 to be capped during 2R13. These criteria have resulted in more frequent capping or replacing of thimble tubes.

As discussed in PG&E Letter DCL-10-096, dated August 12, 2010, PG&E committed to limiting repositioning of any tube to one time. This will be included in the STP R-22, acceptance criteria before 2R16 in May 2011. There are currently no tubes in either unit that have been repositioned more than once.

Since 1R14 all replacement tubes have an extended 12-ft chrome band which covers the area from approximately 8 inches above the fuel bottom nozzle to below the entry into the bottom of the reactor vessel. This has been shown to be effective in the eddy current inspections performed in the subsequent outages.

There has been no detectable wear found in any of the thimble tubes within the chrome hardened band.

3) PG&E has performed 100 percent eddy current testing in every outage since 1R3/2R3. The only degradation mechanism which has been observed is wear scars caused by flow induced vibration. In 2R14, FTT L-13 was inspected by eddy current and showed a through wall flaw but no evidence of cracking.

The flow induced vibration that causes the formation of wear scars is caused by reactor coolant system flow which enters the vessel, is turned by the downcomer, then turns up through the lower internals, through the coreplate, and up past the fuel assemblies. A small portion of the flow goes up through the guide structure with the thimble tubes. Flow past all locations in the core is different.

PG&E attempted to recover the piece of the Unit 2 L-13 FTT from the bottom nozzle location during 2R14. One piece that was examined felt like it had an anomaly. This was captured and packaged up for shipment to Westinghouse for a destructive analysis. The piece had several wear scars on it but none were through wall. The wear scars conformed to the scars Westinghouse had seen during development of the WCAP-12866. Their determination was that the event was caused by flow

Enclosure 1 PG&E Letter DCL-10-167 Page 17 of 17 induced vibration of the thimble tube against the lower internals, core plate or bottom nozzle. This was similar to other failures they had analyzed previously.

The history of Unit 2 L-13 thimble tube is that it was repositioned and capped in 2R3 after the first eddy current exam. It was replaced in 2R10 with a tube having a 15-inch hardened chrome band centered around the fuel bottom nozzle. In 2R13, L-13 was repositioned for the second time. By repositioning the L-13 tube 5 inches twice, the hardened chrome band had been moved below the core plate. The three wear scars within a 10-inch section of tube, located approximately 5 ft below the core plate, allowed that section of the tube to vibrate more than normal and the bare tube at the fuel bottom nozzle wore until the tube failed.

4) PG&E has maintained a trending program for the FTTs. This trending includes comparisons of PG&E's wear projection methodology to the methodology used in WCAP-12866 with DCPP site specific wear data. It also contains comparisons between values predicted by the calculations for the following cycle eddy current examination and the actual NDE measurements made in that refueling outage (RFO). These comparisons substantiate that the PG&E methodology is conservative compared to the WCAP criteria of 80 percent.

DCPP contributed thimble tubes from the 1R3 RFO along with copies of the in-situ eddy current examination results to Westinghouse as part of the WCAP-12866 program, which were measured both mechanically and by in shop eddy current testing. These results led to the development of eddy current testing standards for the whole industry.

Therefore, the monitoring and trending activities, acceptance criteria, and the corrective actions for the FTT program will be capable of detecting degradation in a FTT prior to the occurrence of a through wall failure.

5) PG&E has performed 100 percent eddy current testing in every outage since 1R3/2R3. The only degradation mechanism which has been observed is wear scars caused by flow induced vibration. In 2R14, FTT L-13 was inspected by eddy current and showed a through wall flaw but no evidence of cracking.

Therefore, PG&E believes that the acceptance criteria are comprehensive and conservative.

PG&E Letter DCL-10-167 Page 1 of 4 LRA Amendment 31 LRA Section RAI Section A3.2.2 4.1-7 Section 4.3.3 4.1-7 Table 3.1.2-1 4.1-7 Appendix A PG&E Letter DCL-10-167 Final Safety Analysis Report Supplement Page 2 of 4 A3.2.2 Fatigue Analyses of the Reactor Pressure Vessel Internals The reactor internal components are not ASME code components. The reactor internals were designed and built prior to the implementation of Subsection NG of the ASME Boiler and Pressure Vessel Code,Section III, for reactor vessel internals. Therefore, no plant-specific ASME Code stress report was written during the initial design. However, these components were designed to meet the intent of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code with addenda through the Winter 1971. The qualification of the reactor vessel internals was first performed on a generic basis. Some internal components were subsequently analyzed on a plant-specific basis to account for plant modifications.

The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section A2.1 monitors fatigue design transients for the period of extended operation and provides reasonable assurance that the fatigue in the reactor vessel internals will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

The Reactor Vessel Internals Aging Management program will monitor the integrity of the baffle and former bolts and provides reasonable assurance that the fatigue in the baffle and former bolts will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

Section 4 PG&E Letter DCL-10-167 Time-Limited Aging Analysis Page 3 of 4 4.3.3 Fatigue Analyses of the Reactor Pressure Vessel Internals Baffle-Former Bolts The fatigue usage factor of the baffle-former bolts was originally shown to be less than 1.0 based on evaluation of test data which demonstrated acceptable performance for a set of bolt displacements. However the The adequacy of baffle-former bolts is an industry issue and the design analyses and evaluations are not sufficient to support the safety determination.

Therefore, the design of the baffle bolts is not supported by a TLAA, in accordance with 10 CFR 54.3(a), Criterion 4. Ttheir extended operation is addressed by participation in industry level initiatives as described below.

Flow Induced Vibration in the Reactor Vessel Internals FSAR Section 3.9.1 and the original SER for DCPP discuss the design and vibration test programs for the reactor vessel internals performed as part of preoperational and startup testing. The dynamic behavior of reactor internals has been studied using experimental data obtained from prototype plants along with results of model tests and static and dynamic tests.

Indian Point Nuclear Generating Unit 2 was the prototype for the DCPP Unit 1 internals verification program. Trojan Nuclear Plant data provide additional internals verification for Unit 2 (Unit 1 lower internals are similar to Indian Point Unit 2; Unit 2 lower internals are similar to Trojan). The tests indicated that no unexpected large vibration amplitudes existed in the internal structure during operation.

The licensing basis does not describe any time limited effects for a licensed operating period associated with flow-induced vibration. Therefore there are no TLAAs, in accordance with 10 CFR 54.3(a) Criteria 2 and 3.

Participation in Industry Programs for Reactor Vessel Internals PG&E will (1) participate in industry programs for investigating and managing the aging effects on the reactor vessel 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 prior to entering the period of extended operation, PG&E will submit an inspection plan to the NRC for review and approval.

Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)

The design basis number of transients will be managed for the period of extended operation by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary program, which is summarized in Sections 4.3.1 and B3.1. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. The continued implementation provides reasonable assurance that fatigue in the reactor vessel internals will be managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(iii).

The integrity of the baffle and former bolts will be managed by the Reactor Vessel Internals Aging Management program, which DCPP committed to implement in LRA Table A4-1, Commitment 22. The implementation of the program provides assurance that fatigue in the baffle and former bolts will be managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(iii).

Section 3.1 PG&E Letter DCL-10-167 AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Page 4 of 4 Table 3.1.2-1 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -

Reactor Vessel and Internals Aging Effect Component Intended Aging Management NUREG-1801 Table 1 Material Environment Requiring Notes Type Function Program Vol. 2 Item Item Management RVI Baffle &

TLAA, evaluated in Former Stainless Reactor Coolant Cumulative SS accordance with 10 IV.B2-31 3.1.1.05 A, 4 Assembly Steel (Ext) Fatigue Damage CFR 54.21(c)

4. The commitment to implement the RVI inspection plan will manage fatigue of the Baffle-Former Bolts in the Reactor Vessel Internals Baffle and Former Assembly

Enclosure 3 PG&E Letter DCL-10-167 Page 1 of 18 Table 1 Reactor Coolant Pressure Boundary Valves and Design Codes Valve Component Component Design Code Size Number Description (in.)

UNIT 1 RCS-1-513 PZR Spray Line Drain To RCDT-1st Off ASME III-83 w/84A 0.75 RCS-1-514 PZR Spray Line Drain To RCDT-2nd Off ASME III-83 w/84A 0.75 RCS-1-515 PCV-455B Drain To RCDT-1st Off USAS B31.1-67 / B16.5-68 0.75 RCS-1-516 PCV-455B Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 0.75 RCS-1-517 PCV-455B Vent-2nd Off ASME III-89 0.75 RCS-1-518 PCV-455B Vent-1st Off ASME III-89 0.75 RCS-1-519 PCV-455B Drain To RCDT-1st Off USAS B31.1-67 / B16.5-68 0.75 RCS-1-520 PCV-455B Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 0.75 RCS-1-521 PCV-455A Vent Vlv-2nd Off ASME III-89 0.75 RCS-1-522 PCV-455A Vent Vlv-1st Off ASME III-89 0.75 RCS-1-523 PCV-455A Drain To RCDT-1st Off USAS B31.1-67 / B16.5-68 0.75 RCS-1-524 PCV-455A Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 0.75 ASME III-80 w/82A or RCS-1-665 Isol To Vacuum Degas Sys - 1st Off Later 2 ASME III-80 w/82A or RCS-1-666 Isol To Vacuum Degas Sys - 2nd Off Later 2 ASME III-80 w/82A or RCS-1-731 RCS Vacuum Refill Connection From Pressurizer, Later 1 ASME III-80 w/82A or RCS-1-732 RCS Vacuum Refill Connection From Pressurizer, Later 1 RCS-1-744 RVLIS Manifold Isolation Valve ASME III-01 w/03A 1 RCS-1-8000A Motor Stop Upstream Of PZR PWR RV PCV-474 USAS B31.1-67 / B16.5-68 3 RCS-1-8000B Motor Stop Upstream Of PZR PWR RV PCV-455C USAS B31.1-67 / B16.5-68 3 RCS-1-8000C Motor Stop Upstream Of PZR PWR RV PCV-456 USAS B31.1-67 / B16.5-68 3

Enclosure 3 PG&E Letter DCL-10-167 Page 2 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-1-8033A PCV-455B Dwnstrm Stop USAS B31.1-67 / B16.5-68 4 RCS-1-8033B PCV-455B Upstream Stop USAS B31.1-67 / B16.5-68 4 RCS-1-8033C PCV-455A Dwnstrm Stop USAS B31.1-67 / B16.5-68 4 RCS-1-8033D PCV-455A Upstream Stop USAS B31.1-67 / B16.5-68 4 RCS-1-8050 PCV-455A Bypass ASME III-80 or Later 0.75 RCS-1-8051 PCV-455B Bypass ASME III-80 or Later 0.75 Iso Vlv For PZR RV Loop Seal Drain Header (2nd RCS-1-8052 Off) ASME III-86 0.75 High Leg Iso Vlv To PZR LT-459 And PTs 455 And RCS-1-8053A 474 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8053B High Leg Iso Vlv To PZR LT-460, PT-456 & LT-406 ASME III-86 0.75 RCS-1-8053C High Leg Iso Vlv To PZR LTS 461 462 & PTs 457 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8054A Low Leg Iso Vlv To PZR LT-459 & PTs 455 And 474 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8054B Low Leg Iso Vlv To PZR LT-460, PT-456 & LT-406 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8054C Low Leg Iso Vlv To PZR LTS 461 462 PTs 457 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8056 Vent On PZR PWR RV Header USAS B31.1-67 / B16.5-68 0.75 RCS-1-8057A Loop 1-1 Drain To RCDT - 1st Off USAS B31.1-67 / B16.5-68 2 RCS-1-8057B Loop 1-2 Drain To RCDT - 1st Off USAS B31.1-67 / B16.5-68 2 RCS-1-8057C Loop 1-3 Drain To RCDT-1st Off USAS B31.1-67 / B16.5-68 2 RCS-1-8057D Loop 1-4 Drain To RCDT-1st Off & RVRLIS Conn USAS B31.1-67 / B16.5-68 2 RCS-1-8058A Loop 1-1 Drain To RCDT - 2nd Off USAS B31.1-67 / B16.5-68 2 RCS-1-8058B Loop 1-2 Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 2 RCS-1-8058C Loop 1-3 Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 2 RCS-1-8058D Loop 1-4 Drain To RCDT-2nd Off USAS B31.1-67 / B16.5-68 2 RCS-1-8059A FT-416, Loop 1-1 Low Pressure Iso ASME III-86 0.75 RCS-1-8059B FT-426, Loop 1-2 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8059C FT-436, Loop 1-3 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 3 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-1-8059D FT-446, Loop 1-4 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8060A FT-415, Loop 1-1 Low Pressure Iso ASME III-86 0.75 RCS-1-8060B FT-425, Loop 1-2 Low Pressure Iso ASME III-86 0.75 RCS-1-8060C FT-435, Loop 1-3 Low Pressure Iso ASME III-86 0.75 RCS-1-8060D FT-445, Loop 1-4 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8061A FT-414, Loop 1-1 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8061B FT-424, Loop 1-2 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8061C FT-434, Loop 1-3 Low Pressure Iso ASME III-86 0.75 RCS-1-8061D FT-444, Loop 1-4 Low Pressure Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8062A High Pressure Leg Iso Vlv For FT's 414, 415 & 416 ASME III-86 0.75 RCS-1-8062B High Pressure Leg Iso For FT's 424, 425 & 426 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8062C High Pressure Leg Iso Vlv For FT's 434, 435 & 436 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8062D High Pressure Leg Iso For FT's 444, 445 & 446 USAS B31.1-67 / B16.5-68 0.75 RCS-1-8064A Drain Vlv On Loop Seal Upstream of PZR RV-8010A ASME III-86 0.75 RCS-1-8064B Drain Vlv On Loop Seal Upstream of PZR RV-8010B ASME III-86 0.75 Drain Vlv On Loop Seal Upstream of PZR RV-RCS-1-8064C 8010C ASME III-86 0.75 RCS-1-8066 RVRLIS LI-400A,B,C Conn On Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 RCS-1-8070 Reactor Vessel Head Vent ASME III-01 w/03A 1 RCS-1-8071A PT-406 RHR Suction From Loop 1-4 Press Iso ASME III-86 0.75 RCS-1-8071B PI-476 RHR Suction From Loop 1-4 Press Iso USAS B31.1-67 / B16.5-68 0.75 RCS-1-8076 Loop 1-2 Letdown Iso Vlv To CVCS USAS B31.1-67 / B16.5-68 3 RCS-1-8077 Loop 1-2 Excess Letdown Iso Vlv To CVCS USAS B31.1-67 / B16.5-68 1 RCS-1-8086 Loop 1-1 Hot Leg To NSSS System USAS B31.1-67 / B16.5-68 0.75 RCS-1-8090 RCS Loop 1-4 Hot Leg To NSSS USAS B31.1-67 / B16.5-68 0.75 RCS-1-8093 Iso Vlv For PZR RV Loop Seal Drain Header ASME III-86 0.75 RCS-1-9365A PZR Steam Sample To NSSS Iso USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 4 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-1-9365B PZR Liquid Sample To NSSS Iso ASME III-86 0.75 RCS-1-PCV-455A PZR Spray Control Valve (From Loop 1 Cold Leg) ASME III-80 or Later 4 RCS-1-PCV-455B PZR Spray Control Valve (From Loop 2 Cold Leg) ASME III-80 or Later 4 RCS-1-PCV-455C PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-1-PCV-456 PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-1-PCV-474 PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-1-RV-8010A PZR Safety Valves ASME III-68, Article 9 6 RCS-1-RV-8010B PZR Safety Valves ASME III-68, Article 9 6 RCS-1-RV-8010C PZR Safety Valves ASME III-68, Article 9 6 CVCS-1-283 RCP 1-1 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-294 RCP 1-2 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-303 RCP 1-3 Seal Inj. Water Line Drain To The RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-308 RCP 1-4 Seal Inj. Water Line Drain To The RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-8145 Pressurizer Aux. Spray After Regen. Hx USAS B31.1-67 / B16.5-68 2 CVCS-1-8148 Pressurizer Aux. Spray Bypass USAS B31.1-67 / B16.5-68 2 CVCS-1-8166 Excess Letdown To Hx 1-1 USAS B31.1-67 / B16.5-68 1 CVCS-1-8167 Excess Letdown To Hx 1-1 USAS B31.1-67 / B16.5-68 1 CVCS-1-8364A RCP 1-1 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-8364B RCP 1-2 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-8364C RCP 1-3 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-8364D RCP 1-4 Seal Inj. Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 ASME III-80 w/82A or CVCS-1-8367A RCP 1-1 Seal Inj. Water Inlet Check Later 2 CVCS-1-8367B RCP 1-2 Seal Inj. Water Inlet Check ASME III-80 w/82A or 2

Enclosure 3 PG&E Letter DCL-10-167 Page 5 of 18 Valve Component Component Design Code Size Number Description (in.)

Later ASME III-80 w/82A or CVCS-1-8367C RCP 1-3 Seal Inj. Water Inlet Check Later 2 CVCS-1-8367D RCP 1-4 Seal Inj. Water Inlet Check USAS B31.1-67 / B16.5-68 2 ASME III-80 w/82A or CVCS-1-8372A RCP 1-1 Seal Inj. Water Inlet Check Later 2 ASME III-80 w/82A or CVCS-1-8372B RCP 1-2 Seal Inj. Water Inlet Check Later 2 ASME III-80 w/82A or CVCS-1-8372C RCP 1-3 Seal Inj. Water Inlet Check Later 2 CVCS-1-8372D RCP 1-4 Seal Inj. Water Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-1-8377 Pressurizer Aux. Spray Line Check USAS B31.1-67 / B16.5-68 2 CVCS-1-8378A Charging Line To Loop 3 Cold Leg Check USAS B31.1-67 / B16.5-68 3 CVCS-1-8378B Charging Line To Loop 4 Check USAS B31.1-67 / B16.5-68 3 CVCS-1-8379A Charging Line To Loop 3 Cold Leg Check USAS B31.1-67 / B16.5-68 3 CVCS-1-8379B Charging Line To Loop 4 Check USAS B31.1-67 / B16.5-68 3 CVCS-1-93 Pressurizer Aux. Spray Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-1-94 Pressurizer Aux. Spray Line Drain To RCDT ASME III-86 0.75 CVCS-1-LCV-459 Regen. Hx Letdown Inlet USAS B31.1-67 / B16.5-68 3 CVCS-1-LCV-460 Regen. Hx Letdown Inlet USAS B31.1-67 / B16.5-68 3 SI-1-100 Vent Stop RHR PP 1-1 Loop 2 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-102 Vent Stop RHR PP 1-2 Loop 3 Cold Leg (1st Off) ASME III-83 W/84A 0.75 SI-1-104 Vent Stop RHR PP 1-2 Loop 4 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-106 Drain Stop RHR PP 1-1 Loop 1 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-108 Drain Stop RHR PP 1-1 Loop 2 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-110 Drain Stop RHR PP 1-2 Loop 3 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-112 Drain Stop RHR PP 1-2 Loop 4 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 6 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-1-169 Drain Stop To RCDT (RCS Loop 1 Hot Leg) (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-301A Accum Tk #1 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-1-301B Accum Tk #2 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-1-301C Accum Tk #3 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-1-301D Accum Tk #4 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-1-309 Accum Fill From Charging Injection (1st Off) ASME III-86 0.75 SI-1-311 Test Conn Stop on Accum Fill (1st Off) ASME III-86 0.75 SI-1-313A Hot Leg - Loop 1 Test Isol (1st Off) ASME III-86 0.75 SI-1-313B Hot Leg - Loop 2 Test Isol (1st Off) ASME III-86 0.75 SI-1-313C Hot Leg - Loop 3 Test Isol (1st Off) ASME III-86 0.75 SI-1-313D Hot Leg - Loop 4 Test Isol (1st Off) ASME III-86 0.75 SI-1-40 (1st Off) Vent Stop SIS To RCS Loop 1 Hot Leg ASME III-83 W/84A 0.75 ASME III-80 w/82A or SI-1-42 (1st Off) Vent Stop SIS To RCS Loop 2 Hot Leg Later 0.75 SI-1-48 1st Off Vent SI Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-50 1st Off Vent SI Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-51 2nd Off Vent SI Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-52 1st Off Vent SI Loop 2 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-56 1st Off Vent SI Loop 3 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-58 1st Off Vent SI Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-59 2nd Off Vent SI Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-1-74 Vent Stop On SI Line (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-78 Drain Stop SI Loop 1 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-80 Drain Stop SI Loop 4 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-82 Drain Stop SI Loop 3 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-84 Drain Stop SI Loop 2 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-86 Drain Stop SI PPS Loop 1 Hot Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 7 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-1-88 Drain Stop SI PPS Loop 2 Hot Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 Adjusting Vlv For CCP Runout RCS Loop 1 Cold ASME III-80 w/82A or SI-1-8810A Leg Later 1.5 Adjusting Vlv For CCP Runout RCS Loop 2 Cold ASME III-80 w/82A or SI-1-8810B Leg Later 1.5 Adjusting Vlv For CCP Runout RCS Loop 3 Cold ASME III-80 w/82A or SI-1-8810C Leg Later 1.5 Adjusting Vlv For CCP Runout RCS Loop 4 Cold ASME III-80 w/82A or SI-1-8810D Leg Later 1.5 SI-1-8818A Check Vlv RHR Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-1-8818B Check Vlv RHR Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-1-8818C Check Vlv RHR Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-1-8818D Check Vlv RHR Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-1-8819A Check Vlv SIS Loop 1 Cold Leg ASME III-86 2 SI-1-8819B Check Vlv SIS Loop 2 Cold Leg ASME III-86 2 SI-1-8819C Check Vlv SIS Loop 3 Cold Leg ASME III-86 2 SI-1-8819D Check Vlv SIS Loop 4 Cold Leg ASME III-86 2 SI-1-8820 Charging Injection Hdr Check Vlv USAS B31.1-67 / B16.5-68 3 SI-1-8900A Check Vlv RCS Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 1.5 SI-1-8900B Check Vlv RCS Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 1.5 SI-1-8900C Check Vlv RCS Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 1.5 SI-1-8900D Check Vlv RCS Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 1.5 ASME III-80 w/82A or SI-1-8902A Iso Vlv FE-924 RCS Loop 1 Cold Leg (Lo Leg) Later 0.75 ASME III-80 w/82A or SI-1-8902B Iso Vlv FE-925 RCS Loop 2 Cold Leg (Lo Leg) Later 0.75 ASME III-80 w/82A or SI-1-8902C Iso Vlv FE-926 RCS Loop 3 Cold Leg (Lo Leg) Later 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 8 of 18 Valve Component Component Design Code Size Number Description (in.)

ASME III-80 w/82A or SI-1-8902D Iso Vlv FE-927 RCS Loop 4 Cold Leg (Low Leg) Later 0.75 ASME III-80 w/82A or SI-1-8903A Iso Vlv FE-924 RCD Loop 1 Cold Leg (Hi Leg) Later 0.75 ASME III-80 w/82A or SI-1-8903B Iso Vlv FE-925 RCD Loop 2 Cold Leg (Hi Leg) Later 0.75 ASME III-80 w/82A or SI-1-8903C Iso Vlv FE-926 RCS Loop 3 Cold Leg (Hi Leg) Later 0.75 ASME III-80 w/82A or SI-1-8903D Iso Valve FE-927 RCS Loop 4 Cold Leg (High Leg) Later 0.75 SI-1-8905A Check Vlv SI PP 1-1 Disch To Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-1-8905B Check Vlv SI PP 1-1 Disch To Loop 2 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-1-8905C Check Vlv SI PP 1-2 Disch To Loop 3 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-1-8905D Check Vlv SI PP 1-2 Disch To Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-1-8948A Check Vlv RCS Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-1-8948B Check Vlv RCS Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-1-8948C Check Vlv RCS Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-1-8948D Check Vlv RCS Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-1-8949A Check Vlv - SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-1-8949B Check Vlv - SIS To RCS Loop 2 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-1-8949C Check Vlv - SIS To RCS Loop 3 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-1-8949D Check Vlv - SIS To RCS Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-1-8956A Check Vlv Accum 1-1 Disch USAS B31.1-67 / B16.5-68 10 SI-1-8956B Check Vlv Accum 1-2 Disch USAS B31.1-67 / B16.5-68 10 SI-1-8956C Check Vlv Accum 1-3 Disch USAS B31.1-67 / B16.5-68 10 SI-1-8956D Check Vlv Accum 1-4 Disch USAS B31.1-67 / B16.5-68 10 SI-1-90 Drain Stop SI PPS Loop 2 Hot Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 SI-1-92 Drain Stop SI PPS Loop 4 Hot Leg (1st Off) ASME III-86 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 9 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-1-98 Vent Stop RHR PP 1-1 Loop 1 Cold Leg (1st Off) USAS B31.1-67 / B16.5-68 0.75 RHR-1-2 RHR PPS Suction From RCS Loop 4 HL Drain Iso USAS B31.1-67 / B16.5-68 0.75 RHR-1-3 RHR PPS Suction From RCS Loop 4 HL Drain Iso USAS B31.1-67 / B16.5-68 0.75 RHR PPS Suction From RCS Loop 4 HL Drn to RHR-1-4 RCDT USAS B31.1-67 / B16.5-68 0.75 RHR-1-6 Valve 8702 Upstream Line Vent-Second Off USAS B31.1-67 / B16.5-68 0.75 RHR-1-7 RHR PPS Suction From RCS Loop 4 HL Test Conn. ASME III-83 W/84A 0.75 USAS B31.1-67 / MSS SP-RHR-1-8701 Reactor Coolant Loop 4 Outlet To RHR System 66 14 USAS B31.1-67 / MSS SP-RHR-1-8702 Reactor Coolant Loop 4 Outlet To RHR System 66 14 Pressure Balancing Valve For Double Disc. Mov-RHR-1-8710A 8701 ASME III-83 w/84A 0.75 Pressure Balancing Valve For Double Disc. Mov-RHR-1-8710B 8702 USAS B31.1-67 / B16.5-68 0.75 RHR-1-8740A RHR Disch. To RCS Hot Leg 1 Check Valve USAS B31.1-67 / B16.5-68 8 RHR-1-8740B RHR Disch. To RCS Hot Leg 2 Check Valve USAS B31.1-67 / B16.5-68 8 RHR-1-935 Valve 8702 Upstream Line Vent-First Off USAS B31.1-67 / B16.5-68 0.75 RHR-1-936 Valve 8702 Downstream Line Vent-First Off USAS B31.1-67 / B16.5-68 0.75 UNIT 2 RCS-2-513 PZR Spray Line-First Off Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 RCS-2-514 PZR Spray Line-Second Off Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 RCS-2-515 PCV 455B Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-516 PCV 455B Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-517 PCV 455B Vent-Second Off Valve ASME III-89 0.75 RCS-2-518 PCV 455B Vent-First Off Valve ASME III-89 0.75 RCS-2-519 PCV 455B Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 10 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-2-520 PCV 455B Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-521 PCV 455A Vent Valve-Second Off ASME III-89 0.75 RCS-2-522 PCV 455A Vent Valve-First Off ASME III-89 0.75 RCS-2-523 PCV 455A Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-524 PCV 455A Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-570 Loop 4 RHR Suction Drain (2nd Off) USAS B31.1-67 / B16.5-68 0.75 RCS-2-579 Loop 4 RHR Suction Drain (1st Off) USAS B31.1-67 / B16.5-68 0.75 ASME III-80 w/82A or RCS-2-684 Loop 2-1 CL High Point Vent, 1st Off Isolation Valve Later 2 Loop 2-1 CL High Point Vent, 2nd Off Isolation ASME III-80 w/82A or RCS-2-685 Valve Later 2 ASME III-80 w/82A or RCS-2-687 RCS Vacuum Refill Connection From Pressurizer, Later 1 ASME III-80 w/82A or RCS-2-688 RCS Vacuum Refill Connection From Pressurizer, Later 1 RCS-2-744 Iso Vlv RCS Vent System ASME III-01 w/03A 1 RCS-2-8000A Motor Stop Upstream Of PZR PWR RV PCV 474 USAS B31.1-67 / B16.5-68 3 RCS-2-8000B Motor Stop Upstream Of PZR PWR RV PCV 455C USAS B31.1-67 / B16.5-68 3 RCS-2-8000C Motor Stop Upstream Of PZR PWR RV PCV 456 USAS B31.1-67 / B16.5-68 3 RCS-2-8033A PCV 455B Downstream Stop Valve USAS B31.1-67 / B16.5-68 4 RCS-2-8033B PCV 455B Upstream Stop Valve USAS B31.1-67 / B16.5-68 4 RCS-2-8033C PCV 455A Downstream Stop Valve USAS B31.1-67 / B16.5-68 4 RCS-2-8033D PCV 455A Upstream Stop Valve USAS B31.1-67 / B16.5-68 4 RCS-2-8050 PCV 455A Bypass Valve ASME III-80 or Later 0.75 RCS-2-8051 PCV 455B Bypass Valve ASME III-80 or Later 0.75 RCS-2-8052 Iso Valve For PZR RV Loop Seal Drain Header ASME III-86 0.75 RCS-2-8053A High Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 11 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-2-8053B High Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75 RCS-2-8053C High Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75 RCS-2-8054A Low Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75 RCS-2-8054B Low Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75 RCS-2-8054C Low Leg Iso Valve To PZR Instruments USAS B31.1-67 / B16.5-68 0.75 RCS-2-8056 Vent On PZR PWR RV Header USAS B31.1-67 / B16.5-68 0.75 RCS-2-8057A Loop 2-1 Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 2 RCS-2-8057B Loop 2-2 Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 2 RCS-2-8057C Loop 2-3 Drain To RCDT-First Off Valve USAS B31.1-67 / B16.5-68 2 ASME III-80 w/82A or RCS-2-8057D Loop 2-4 Drain To RCDT-1st Off & RVRLIS Later 2 RCS-2-8058A Loop 2-1 Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 2 RCS-2-8058B Loop 2-2 Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 2 RCS-2-8058C Loop 2-3 Drain To RCDT-Second Off Valve USAS B31.1-67 / B16.5-68 2 ASME III-80 w/82A or RCS-2-8058D Loop 2-4 Drain To RCDT-Second Off Valve Later 2 RCS-2-8059A FT-416, Loop 2-1 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8059B FT-426, Loop 2-2 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8059C FT 436, Loop 2-3 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8059D FT 446, Loop 2-4 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8060A FT-415, Loop 2-1 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8060B FT-425, Loop 2-2 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8060C FT-435, Loop 2-3 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8060D FT 445, Loop 2-4 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8061A FT-414, Loop 2-1 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8061B FT-424, Loop 2-2 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-8061C FT 434, Loop 2-3 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 12 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-2-8061D FT 444, Loop 2-4 Low Pressure Iso Valve USAS B31.1-67 / B16.5-68 0.75 High Pressure Leg Iso Valve For FT's 414, 415 &

RCS-2-8062A 416 USAS B31.1-67 / B16.5-68 0.75 High Pressure Leg Iso Valve For FTs 424, 425 &

RCS-2-8062B 426 USAS B31.1-67 / B16.5-68 0.75 High Pressure Leg Iso Valve For FTs 434, 435 &

RCS-2-8062C 436 USAS B31.1-67 / B16.5-68 0.75 High Pressure Leg Iso Valve For FTs 444, 445 &

RCS-2-8062D 446 USAS B31.1-67 / B16.5-68 0.75 RCS-2-8064A Drain Valve Loop Seal Upstream of PZR RV-8010A USAS B31.1-67 / B16.5-68 0.75 RCS-2-8064B Drain Valve Loop Seal Upstream of PZR RV-8010B USAS B31.1-67 / B16.5-68 0.75 RCS-2-8064C Drain Valve Loop Seal Upstream of PZR RV-8010C USAS B31.1-67 / B16.5-68 0.75 RCS-2-8066 Loop 2-4 RVRLIS Standpipe On Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 RCS-2-8070 Reactor Vessel Head Vent Vlv ASME III-01 w/03A 0.75 PT 406 RHR Suction From Loop 2-4 Pressure Iso RCS-2-8071A Vlv USAS B31.1-67 / B16.5-68 0.75 RCS-2-8071B PI 476 RHR Suction From Loop 2-4 Pressure Iso Vlv ASME III-86 0.75 RCS-2-8076 Loop 2-2 Letdown Iso Valve To CVCS USAS B31.1-67 / B16.5-68 3 RCS-2-8077 Loop 2-2 Excess Letdown Iso Valve To CVCS ASME III-80 w/82A 1 RCS-2-8086 Loop 2-1 Hot Leg To NSSS System ASME III-86 0.75 RCS-2-8090 RCS Loop 2-4 Hot Leg To NSSS ASME III-86 0.75 RCS-2-8093 Isol Vlv for PZR Loop Seal Drain ASME III-86 0.75 RCS-2-9365A PZR Steam Sample To NSSS Iso Valve USAS B31.1-67 / B16.5-68 0.75 RCS-2-9365B PZR Liquid Sample To NSSS Iso Valve ASME III-86 0.75 RCS-2-PCV-455A PZR Spray Control Valve (From Loop 1 Cold Leg) ASME III-80 or Later 4 RCS-2-PCV-455B PZR Spray Control Valve (From Loop 2 Cold Leg) ASME III-80 or Later 4

Enclosure 3 PG&E Letter DCL-10-167 Page 13 of 18 Valve Component Component Design Code Size Number Description (in.)

RCS-2-PCV-455C PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-2-PCV-456 PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-2-PCV-474 PZR Power Operated Relief Valve USAS B31.1-67 / B16.5-68 3 RCS-2-RV-8010A PZR Safety Valves ASME III-68, Article 9 6 RCS-2-RV-8010B PZR Safety Valves ASME III-68, Article 9 6 RCS-2-RV-8010C PZR Safety Valves ASME III-68, Article 9 6 CVCS-2-283 RCP 2-1 Seal Inj Water Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-294 RCP 2-2 Seal Inj Water Line Drain To RCDT ASME III-86 0.75 CVCS-2-303 RCP 2-3 Seal Inj Water Line Drain To The RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-308 RCP 2-4 Seal Inj Water Line Drain To The RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-8145 Pressurizer Aux Spray After Regen Hx USAS B31.1-67 / B16.5-68 2 CVCS-2-8148 Pressurizer Aux. Spray Bypass USAS B31.1-67 / B16.5-68 2 CVCS-2-8166 Excess Letdown To Hx 2-1 USAS B31.1-67 / B16.5-68 1 CVCS-2-8167 Excess Letdown To Hx 2-1 USAS B31.1-67 / B16.5-68 1 CVCS-2-8364A RCP 2-1 Seal Inj Wtr Line Drn To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-8364B RCP 2-2 Seal Inj Wtr Line Drn To RCDT ASME III-86 0.75 CVCS-2-8364C RCP 2-3 Seal Inj Wtr Line Drn To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-8364D RCP 2-4 Seal Inj Wtr Line Drn To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-8367A RCP 2-1 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8367B RCP 2-2 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8367C RCP 2-3 Seal Inj Wtr Inlet Check ASME III-86 2 CVCS-2-8367D RCP 2-4 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8372A RCP 2-1 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8372B RCP 2-2 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8372C RCP 2-3 Seal Inj Wtr Inlet Check ASME III-86 2

Enclosure 3 PG&E Letter DCL-10-167 Page 14 of 18 Valve Component Component Design Code Size Number Description (in.)

CVCS-2-8372D RCP 2-4 Seal Inj Wtr Inlet Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8377 Pressurizer Aux Spray Line Check USAS B31.1-67 / B16.5-68 2 CVCS-2-8378A Charging Line To Loop 3 Cold Leg Check USAS B31.1-67 / B16.5-68 3 CVCS-2-8378B Charging Line To Loop 4 Check USAS B31.1-67 / B16.5-68 3 CVCS-2-8379A Charging Line To Loop 3 Cold Leg Check USAS B31.1-67 / B16.5-68 3 CVCS-2-8379B Charging Line To Loop 4 Check USAS B31.1-67 / B16.5-68 3 CVCS-2-93 Pressurizer Aux Spray Line Drain To RCDT ASME III-86 0.75 CVCS-2-94 Pressurizer Aux Spray Line Drain To RCDT USAS B31.1-67 / B16.5-68 0.75 CVCS-2-LCV-459 Regen Hx Letdown Inlet USAS B31.1-67 / B16.5-68 3 CVCS-2-LCV-460 Regen Hx Letdown Inlet USAS B31.1-67 / B16.5-68 3 SI-2-100 Vent Stop RHR PP 2-1 Loop 2 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-102 Vent Stop RHR PP 2-2 Loop 3 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-104 Vent Stop RHR PP 2-2 Loop 4 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-106 Drain Stop RHR PP 2-1 Loop 1 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-108 Drain Stop RHR PP 2-1 Loop 2 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-110 Drain Stop RHR PP 2-2 Loop 3 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-112 Drain Stop RHR PP 2-2 Loop 4 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-169 Drain Stop To RCDT RCS Loop 1 Hot Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-301A Accum Tk #1 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-2-301B Accum Tk #2 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-2-301C Accum Tk #3 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-2-301D Accum Tk #4 Leak Test (Reactor Side - 1st Off) ASME III-86 0.75 SI-2-309 Accum Fill From Charging Injection (1st Off) ASME III-86 0.75 SI-2-311 Test Conn Stop on Accum Fill (1st Off) ASME III-86 0.75 SI-2-313A Hot Leg - Loop 1 Test Isol (1st Off) ASME III-86 0.75 SI-2-313B Hot Leg - Loop 2 Test Isol (1st Off) ASME III-86 0.75 SI-2-313C Hot Leg - Loop 3 Test Isol (1st Off) ASME III-86 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 15 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-2-313D Hot Leg - Loop 4 Test Isol (1st Off) ASME III-86 0.75 SI-2-40 1st Off Vent Stop SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-2-48 1st Off Vent Stop SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-2-50 1st Off Vent Stop SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-2-51 2nd Off Vent Stop SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 0.75 SI-2-52 1st Off -Vent SI Loop 2 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-54 1st Off -Vent SI Loop 2 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-55 2nd Off -Vent SI Loop 2 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-60 1st Off -Vent SI Loop 4 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-62 1st Off -Vent SI Loop 4 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-63 2nd Off -Vent SI Loop 4 Hot Leg Before Chk USAS B31.1-67 / B16.5-68 0.75 SI-2-74 Vent Stop On SI Line-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-78 Drain Stop SI Loop 1 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-80 Drain Stop SI Loop 4 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-82 Drain Stop SI Loop 3 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-84 Drain Stop SI Loop 2 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-86 Drain Stop SI PP's Loop 1 Hot Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-88 Drain Stop SI PP's Loop 2 Hot Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-8810A Adjusting Valve For CCP Runout RCS Loop 1 CL ASME III-86 1.5 SI-2-8810B Adjusting Valve For CCP Runout RCS Loop 2 CL ASME III-86 1.5 SI-2-8810C Adjusting Valve For CCP Runout RCS Loop 3 CL ASME III-86 1.5 SI-2-8810D Adjusting Valve For CCP Runout RCS Loop 4 CL ASME III-86 1.5 SI-2-8818A Check Valve RHR Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-2-8818B Check Valve RHR Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-2-8818C Check Valve RHR Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-2-8818D Check Valve RHR Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 6 SI-2-8819A Check Valve SIS Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 2

Enclosure 3 PG&E Letter DCL-10-167 Page 16 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-2-8819B Check Valve SIS Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 2 SI-2-8819C Check Valve SIS Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 2 SI-2-8819D Check Valve SIS Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 2 SI-2-8820 Charging Inj Header Check Vlv USAS B31.1-67 / B16.5-68 3 SI-2-8900A Check Valve RCS Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 1.5 SI-2-8900B Check Valve RCS Loop 2 Cold Leg ASME III-86 1.5 SI-2-8900C Check Valve RCS Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 1.5 SI-2-8900D Check Valve RCS Loop 4 Cold Leg ASME III-86 1.5 Isolating Valve FE-924 RCS Loop 1 Cold Leg (Lo SI-2-8902A Leg) ASME III-86 0.75 Isolating Valve FE-925 RCS Loop 2 Cold Leg (Lo SI-2-8902B Leg) ASME III-86 0.75 Isolating Valve FE-926 RCS Loop 3 Cold Leg (Lo SI-2-8902C Leg) ASME III-86 0.75 Isolating Valve FE-927 RCS Loop 4 Cold Leg (Lo SI-2-8902D Leg) ASME III-86 0.75 Isolating Valve FE-924 RCS Loop 1 Cold Leg (Hi SI-2-8903A Leg) ASME III-86 0.75 Isolating Valve FE-925 RCS Loop 2 Cold Leg (Hi SI-2-8903B Leg) ASME III-86 0.75 Isolating Valve FE-926 RCS Loop 3 Cold Leg (Hi SI-2-8903C Leg) ASME III-86 0.75 Isolating Valve FE-927 RCS Loop 4 Cold Leg (Hi SI-2-8903D Leg) ASME III-86 0.75 SI-2-8905A Check Valve SI PP 2-1 Disch To Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-2-8905B Check Valve SI PP 2-1 Disch To Loop 2 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-2-8905C Check Valve SI PP 2-2 Disch To Loop 3 Hot Leg USAS B31.1-67 / B16.5-68 2 SI-2-8905D Check Valve SI PP 2-2 Disch To Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 2

Enclosure 3 PG&E Letter DCL-10-167 Page 17 of 18 Valve Component Component Design Code Size Number Description (in.)

SI-2-8948A Check Valve RCS Loop 1 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-2-8948B Check Valve RCS Loop 2 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-2-8948C Check Valve RCS Loop 3 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-2-8948D Check Valve RCS Loop 4 Cold Leg USAS B31.1-67 / B16.5-68 10 SI-2-8949A Check Valve - SIS To RCS Loop 1 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-2-8949B Check Valve - SIS To RCS Loop 2 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-2-8949C Check Valve - SIS To RCS Loop 3 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-2-8949D Check Valve - SIS To RCS Loop 4 Hot Leg USAS B31.1-67 / B16.5-68 6 SI-2-8956A Check Valve Accumulator 2-1 Disch USAS B31.1-67 / B16.5-68 10 SI-2-8956B Check Valve Accumulator 2-2 Disch USAS B31.1-67 / B16.5-68 10 SI-2-8956C Check Valve Accumulator 2-3 Disch USAS B31.1-67 / B16.5-68 10 SI-2-8956D Check Valve Accumulator 2-4 Disch USAS B31.1-67 / B16.5-68 10 SI-2-90 Drain Stop SI PP's Loop 3 Hot Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-92 Drain Stop SI PP's Loop 4 Hot Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 SI-2-98 Vent Stop RHR PP 2-1 Loop 1 Cold Leg-1st Off USAS B31.1-67 / B16.5-68 0.75 RHR-2-2 RHR PPS Suction From RCS Loop 4 HL Drain Isol USAS B31.1-67 / B16.5-68 0.75 RHR PPS Suction From RCS Loop 4 HL Drn to RHR-2-4 RCDT USAS B31.1-67 / B16.5-68 0.75 RHR-2-5 RHR PPS Suct From HL 4 Drain ASME III-86 0.75 RHR-2-6 RHR PPS Suct From HL 4 Drain Line Test Conn. ASME III-86 0.75 USAS B31.1-67 / MSS SP-RHR-2-8701 Reactor Coolant Loop 4 Outlet To RHR System 66 14 USAS B31.1-67 / MSS SP-RHR-2-8702 Reactor Coolant Loop 4 Outlet To RHR System 66 14 Pressure Balancing Valve For Double Disc Mov-RHR-2-8710A 8701 USAS B31.1-67 / B16.5-68 0.75 Pressure Balancing Valve For Double Disc Mov-RHR-2-8710B 8702 USAS B31.1-67 / B16.5-68 0.75

Enclosure 3 PG&E Letter DCL-10-167 Page 18 of 18 Valve Component Component Design Code Size Number Description (in.)

RHR-2-8740A RHR Disch To RCS Hot Leg 1 Check Valve USAS B31.1-67 / B16.5-68 8 RHR-2-8740B RHR Disch To RCS Hot Leg 2 Check Valve USAS B31.1-67 / B16.5-68 8 RHR-2-935 Valve 8702 Upstream Line Vent-First Off USAS B31.1-67 / B16.5-68 0.75 RHR-2-936 Valve 8702 Upstream Line Vent-Second Off USAS B31.1-67 / B16.5-68 0.75 RHR-2-937 Valve 8702 Downstream Line Vent-First Off USAS B31.1-67 / B16.5-68 0.75