ML14205A438
| ML14205A438 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 07/23/2014 |
| From: | Cadogan J Arizona Public Service Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 102-06908-JJC/TNW | |
| Download: ML14205A438 (7) | |
Text
John 3. Cadogan Qaps Vice President, Nuclear Engineering Palo Verde Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072 102-06908-JJC/TNW Mail Station 7605 July 23, 2014 Tel 623 393 5403 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos. STN 50-528, 529, and 530 Response to Request for Additional Information (RAI) -
Pressurized Water Reactor (PWR) Internals Aging Management Program Plan Arizona Public Service Company (APS) provided to the Nuclear Regulatory Commission (NRC) its Pressurized Water Reactor (PWR) Internals Aging Management Program Plan for the Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2 and 3, by letter number 102-06599, dated September 28, 2012
[Agencywide Documents Access & Management System (ADAMS) Accession No. ML122780119)].
In e-mail dated March 24, 2014, the NRC staff indicated that the APS submittal had been reviewed and that additional information was needed to complete the review.
The NRC initially requested that the information be provided by May 21, 2014.
During subsequent communications, APS was granted an extension until July 25, 2014. The Enclosure to this letter contains the requested information.
There are no commitments being made in this letter. Should you need further information regarding this response, please contact Thomas N. Weber, Department Leader, Nuclear Regulatory Affairs, at (623) 393-5764.
Sincerely,
&(
JJC/TNW/hsc
Enclosure:
Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan A member of the STARS (Strategic Teaming and Resource Sharing) Alliance 4 cc Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde 0 Wolf Creek
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan Page 2 cc: M. L. Dapas NRC Region IV Regional Administrator B. K. Singal NRC NRR Project Manager for PVNGS S. R. Sandal NRC NRR Project Manager M. M. Watford NRC NRR Project Manager M. A. Brown NRC Senior Resident Inspector for PVNGS (electronic)
ENCLOSURE Response to Request for Additional Information (RAI)
Pressurized Water Reactor (PWR) Internals Aging Management Program Plan
Enclosure Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan
Background
By letter number 102-06599, dated September 28, 2012 [Agencywide Documents Access Management System (ADAMS) Accession No. ML122780119], Arizona Public Service Company (APS, the licensee) submitted its Pressurized Water Reactor (PWR)
Internals Aging Management Program Plan, which credits the implementation of Materials Reliability Program (MRP)-227-A, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, for Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2 and 3. The MRP-227-A report and supporting reports were used as the technical bases for developing the PVNGS aging management program (AMP). The U.S.
Nuclear Regulatory Commission (NRC) staff reviewed the MRP-227-A report and issued a final safety evaluation (SE) on December 16, 2011 (ADAMS Accession No. ML11308A770). Based on the review of the APS AMP Plan, the NRC staff provided an e-mail dated March 24, 2014, that documented specific areas that required additional information to complete its review. The NRC request for additional information is stated first, followed by the APS response.
NRC Reauest for Additional Information In Section 5.1 of the submittal, the licensee addresses Action Item 1 of the NRC staff's SE for MRP-227-A by stating:
The assumptions regarding plant design and operating history made in MRP-191 are appropriate for PVNGS. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, PVNGS Units 1, 2, and 3 are bounded by the assumption in MRP-191.
As discussed in Section 1.8.4.1 of this document, operations at PVNGS conform to the assumptions in Section 2.4 of MRP-227-A.
" PVNGS Units 1, 2, and 3 historic core management practices meet the requirements of MRP-227-A
- PVNGS Units 1, 2, and 3 operate as base load units
- No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (C-E or Westinghouse)
The NRC staff has determined that additional information, as discussed in References 1 and 2 below, should be provided to verify the applicability of MRP-227-A for each unit at PVNGS. The two specific generic issues that need to be addressed are summarized as follows:
- 1. Do the Reactor Vessel Internals (RVIs) for the units at PVNGS have any non-weld or bolting austenitic stainless steel components with 20% cold work or greater, and if so do the affected components have operating stresses greater than 30 ksi? If so, perform a plant-specific evaluation to determine the aging management requirements for the affected components.
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Enclosure Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan APS Response Plant-specific drawings, procurement specifications, ASTM and ASME standards, and related information were retrieved and reviewed to determine components and locations that may have been cold-worked during fabrication. In particular, stainless steel RVIs were reviewed to determine if cold working could be present due to fabrication, or if materials could have been ordered as strain hardened. The review also considered whether there may be evidence of other conditions, such as replacement parts, which may have led to higher strength, cold-worked austenitic stainless steels having been placed in service.
No core support component materials were found to be in the _>20% cold-worked condition. However, one set of socket head cap screws, which attach shims to the core barrel for alignment of the upper guide structure, were found with an unidentified amount of cold working on the head to shank fillet of the screws. The amount of cold working is not known, but given bolt/screw manufacturing practices, the cold working of the head to shank region could contain greater than 20% cold working. Also, since the cap screws are made from high strength fastener material, the service stress is likely greater than 30 ksi due to bolt preload. It is, therefore, assumed that these cap screws are cold-worked to greater than the 2 0 % threshold and have operating stresses greater than 30 ksi.
The cap screws are used for alignment of the upper guide structure assembly and, therefore, have no core support function. The screws are visually examined during removal/insertion of the upper guide structure assembly to identify any deformed, degraded, broken, or missing cap screws. Additionally, the upper guide structure physically captures/contains these components during operation; so, in the event of cracking, they would not become loose parts. Therefore, no additional examinations of these components or others are required because of cold-working.
NRC Reauest for Additional Information
- 2. Have the units at PVNGS ever utilized atypical fuel design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representative for that plant, including power changes/uprates? If so, describe how the differences were reconciled with the assumptions of MRP-227-A or provide a plant-specific aging management program for affected components as appropriate.
APS Response The PVNGS core designs and fuel management have been compared to the assumptions of MRP-227-A and the screening criteria of MRP 2013-025.
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Enclosure Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan MRP-227-A fluence estimates assumed a 60 year plant operation, with 30 years of "high leakage" cores followed by 30 years of "low leakage" cores. PVNGS operating experience shows approximately 1.5 years (one cycle) of high leakage cores, followed by greater than 26 years of low leakage core operation. This shows significant margin to the assumed fluence in MRP-227-A. PVNGS satisfies each of the MRP-227-A assumptions except for the average core power density. Thus, further screening in accordance with MRP 2013-025 was required.
PVNGS satisfies each of the three boundary screening criteria in MRP 2013-025, as summarized below for each boundary limitation.
MRP 2013-025 Radial Boundary Limitation (components laterally surrounding the core)
Plant-specific applicability of MRP-227-A in the radial direction, with no further evaluation required, is demonstrated by satisfying the following limits for Combustion Engineering (CE) plants:
Heat generation figure of merit (F) being less than or equal to 68 Watts/cm 3 and average core power density being less than 110 Watts/cm 3 .
PVNGS satisfies each of these limits. Specifically, PVNGS Unit 3, Cycle 19, is representative of current and planned future core designs at PVNGS and is considered low leakage. Unit 3, Cycle 19, has a maximum figure of merit value (F) of 58.59 Watts/cm 3 . The current rated power level at PVNGS results in an average core power density of 100.23 Watts/cm 3 . Each of the radial boundary limits are, therefore, satisfied.
MRP 2013-025 Upper Axial Boundary Limitation (components above the core)
Plant-specific applicability of MRP-227-A in the upper axial direction, with no further evaluation required, is demonstrated by satisfying the following limits for CE plants:
Active fuel to fuel alignment plate distance being greater than 12.4 inches and the average core power density being less than 110 Watts/cm 3 .
PVNGS satisfies these limits. PVNGS has an additional design feature not considered in MRP-227-A. PVNGS has control element assembly (CEA) shroud tubes that extend below the fuel alignment plate. The bottom of these tubes are considered the lowest point of the upper internals instead of the fuel alignment plate. The distance from the active fuel to the CEA shroud tubes at PVNGS is greater than 19 inches. The distance from the active fuel to the bottom of the fuel alignment plate is greater than 29 inches. This distance from the fuel to the upper internals results in a decrease in fluence to the upper internals compared to that assumed in MRP-227-A.
As stated previously, the current rated power level at PVNGS results in an average core power density of 100.23 Watts/cm 3 . Each of the upper axial boundary limits are, therefore, satisfied.
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Enclosure Response to Request for Additional Information (RAI) - Pressurized Water Reactor (PWR) Internals Aging Management Program Plan MRP 2013-025 Lower Axial Boundary Limitation (components below the core)
Plant-specific applicability of MRP-227-A in the lower axial direction, with no further evaluation required, is demonstrated by satisfying the MRP-227-A, Section 2.4 assumptions. The MRP-227-A, Section 2.4 assumptions are stated first, followed by a description of how the assumptions are addressed at PVNGS.
The assumptions from MRP-227-A, Section 2.4 are as follows:
- i. 30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripherallocations) followed by implementation of a low leakage fuel management strategy for the remaining 30 years of operation; PVNGS operating experience shows approximately 1.5 years of high leakage cores (one cycle), followed by greater than 26 years of low leakage core operation. This shows significant margin to the assumed fluence in MRP-227-A. This assumption is, therefore satisified.
ii. base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendaror load demand schedule.
PVNGS has not operated in a load follow capacity, therefore, this assumption is satisifed.
iii. no design changes beyond those identified in general industry guidance or recommended by the original vendors.
PVNGS has not implemented any fuel design changes beyond those identified in general industry guidance or those recommended by the fuel vendor. PVNGS has, therefore satisified each of the assumptions of MRP-227-A, Section 2.4. As a result, no further evaluation of fluence in the lower axial direction is necessary.
In conclusion, while APS has performed numerous core reload designs, including two NRC staff approved power uprates over the course of the PVNGS operating history, there have been no significant changes in the areas of atypical core designs or fuel management that would invalidate the assumptions or results of MRP-227-A and the screening criteria in MRP 2013-025.
References
- 1. Meeting Summary EPRI-Westinghouse January 22-23, 2013, dated February 21, 2013 (ADAMS Accession No. ML13042A048)
- 2. Summary of February 25, 2013 Telecom with EPRI and Westinghouse Electric Company, dated March 15, 2013 (ADAMS Accession No. ML13067A262) 4