ML14190A211
| ML14190A211 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/08/2014 |
| From: | Jennivine Rankin Plant Licensing Branch IV |
| To: | Edwards M, Hansher B Omaha Public Power District |
| References | |
| MF3412 | |
| Download: ML14190A211 (8) | |
Text
NRR-PMDAPEm Resource From: Rankin, Jennivine Sent: Tuesday, July 08, 2014 4:06 PM To: medwards@oppd.com; BHansher@oppd.com
Subject:
Request for Additional Information - Reactor Vessel Internal Component Aging Management Program (MF3412)
Attachments: RAI final.docx Mr. Edwards and Mr. Hansher, By letter dated September 27, 2012, (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML12276A005) Omaha Public Power District (OPPD) submitted an aging management program (AMP) for the reactor vessel internals at Fort Calhoun Station, Unit 1 (FCS). The MRP-227-A report, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, and its supporting reports were used as technical bases for developing FCSs AMP.
The NRC staff has reviewed the information provided in the September 27, 2012, letter and determined that additional information is required in order to complete its review.
A draft request for additional information (RAI) was transmitted on May 27, 2014, and it was determined a clarification call was not necessary. Please respond to RAIs 4, 6, 7, 8, and 9 within 45 days of this email.
Please respond the remaining RAIs by January 30, 2015. The NRC staff understands that the responses to the remaining RAIs are dependent on an industry-wide effort and the submittal dates for the responses may change. Please contact the project manager if additional time is necessary.
Please treat this email as formal transmittal of the RAIs.
- Thanks, Jennie Jennie Rankin, Project Manager Palo Verde Nuclear Generating Station Fort Calhoun Station Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1
Hearing Identifier: NRR_PMDA Email Number: 1416 Mail Envelope Properties (Jennivine.Rankin@nrc.gov20140708160500)
Subject:
Request for Additional Information - Reactor Vessel Internal Component Aging Management Program (MF3412)
Sent Date: 7/8/2014 4:05:43 PM Received Date: 7/8/2014 4:05:00 PM From: Rankin, Jennivine Created By: Jennivine.Rankin@nrc.gov Recipients:
"medwards@oppd.com" <medwards@oppd.com>
Tracking Status: None "BHansher@oppd.com" <BHansher@oppd.com>
Tracking Status: None Post Office:
Files Size Date & Time MESSAGE 1519 7/8/2014 4:05:00 PM RAI final.docx 33649 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
REQUEST FOR ADDITIONAL INFORMATION AGING MANAGEMENT PROGRAM FOR THE REACTOR VESSEL INTERNALS FORT CALHOUN STATION, UNIT 1 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285
RAI 1
Historically, the following materials used in the pressurized water reactor (PWR) reactor vessel internals (RVI) components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the U.S. Nuclear Regulatory Commission (NRC) staff requests that the licensee provide a list of any additional RVI components (not listed in MRP-227-A and MRP-191 Revision 0) that are manufactured from the following materials. If any of these materials are identified as an additional RVI component at FCS, provide information on the type of aging effect that was detected, and the type of AMP implemented on these components.
(1) Nickel base alloysInconel 600; Weld MetalsAlloy 82 and 182 and Alloy X-750.
(2) Stainless steel type 347 material (excluding baffle-former bolts).
(3) Precipitation hardened (PH) stainless steel materials17-4 and 15-5.
(4) Type 431 stainless steel material.
(5) Alloy A-286, ASTM A 453 Grade 660, Condition A or B.
RAI 2
Related to MRP-2013-025, MRP-227-A Applicability Template Guidelines, (ADAMS Accession No. ML13322A454), the staff has identified two additional questions that all CE and Westinghouse design plants referencing topical report MRP-227-A must answer to close Applicant/Licensee Action Item (AI) 1 related to plant-specific applicability of the topical report. If the answer to either or both questions is yes, then further evaluation will be necessary to demonstrate the applicability of MRP-227-A to FCS. The staff therefore requests the following information:
- 1. Do the FCS RVI have 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? In particular, the staff is interested in plant-specific information on the extent of cold work on its RVI components. The licensee can apply Option 1 or Option 2, as addressed in Appendix A of the report. If Option 2 is applicable to FCS, the licensee should list plant-specific RVI components that have been exposed to cold work equal to or greater than 20%. Plant-specific information related to this issue as addressed in Option 2 in Appendix A, should be provided.
- 2. Has FCS ever utilized atypical 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? The following guidelines provided
by modification/rework package (MRP) should be followed. The licensee is requested to use the MRP document dated October 14, 2013, MRP-2013-025, and it can apply Option 1 or Option 2, as addressed in Appendix B of the report.
Option 1:
FCS complies with the MRP-227-A assumptions regarding core loading/core design. Neutron fluence and heat generation rates are concluded to be Option A or Option B.
Option A: acceptable based on the following assessment to the limiting MRP guidance threshold values.
Option B: unacceptable based on an assessment to the limiting MRP guidance threshold values.
If Option A as addressed under Option 1 is applicable, the following plant-specific values should be submitted: (a) active fuel to fuel alignment plate distance; (b) average core power density; and, (c) heat generation figure of merit.
If Option B under Option 1 is applicable to FCS, the licensee should justify the usage of its fuel management program.
Option 2:
FCS does not comply with the MRP-227-A assumptions regarding core loading/core design. The licensee should provide a technical justification for the application of MRP-227-A criterion to FCS.
RAI 3
Action Item 2 in the staffs SE for the MRP-227-A requires the licensee to confirm that Table 4-4 of the MRP-191, Revision 0, Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs, report includes all of RVI components for a licensees Combustion Engineering-designed (CE-designed) reactor facility, or else to identify the missing components and propose any necessary modifications to the program defined in the MRP-227-A report for CE-designed reactors.
RAI 4
Chapter 7 of the MRP-227-A report addresses how a licensee will evaluate and disposition relevant plant-specific or generic operating experience (OE) that is applicable to RVI components at its PWR facility. The NRC staff requests that the licensee identify any and all generic and plant-specific OE that is applicable to the design of the RVI components at Fort Calhoun, including but not limited to OE that is applicable the following components at FCS:
Panel to former bolts, core barrel bolting, thermal shields (including positioning pins),
fuel alignment pins, guide lug inserts and bolts, guide lugs, core support barrel girth
welds, in-core instrumentation flux thimble tubes, core barrel at the prior thermal shield bracket attachment areas, RV flow skirt, and, include the RVI components addressed in Appendix B of the September 27, 2012 submittal.
RAI 5
As discussed in Section 3.3.7 of Revision 1 to the Safety Evaluation for MRP-227, AI 7 requires that the licensees of Westinghouse reactors develop plant-specific analyses to be applied for their facilities to demonstrate that lower support column cast austenitic stainless steel (CASS) bodies will maintain their function during the extended period of operation. MRP-227-A Table 3-2 (Final disposition of CE internals) classifies CASS lower support columns as Primary Components based on susceptibility to irradiation embrittlement (IE) and irradiation assisted stress corrosion cracking (IASCC) and thermal embrittlement (TE). After further review of the existing literature data for the threshold limits for IE and TE of CASS materials, the staff developed a new position on these limits. The bases for the staffs new threshold limits are described in Attachment 1 of this document.
To enable evaluation of the susceptibility of the lower support columns to TE and IE, the staff requests that the licensee should provide the following information:
(a) provide the neutron exposures for these columns and assess their susceptibility to IE and TE consistent with Attachment 1, or provide a functionality evaluation of the lower support columns considering the effects of IE and TE, (b) provide the ferrite content for each lower support column and, (c) provide the casting method for the column (static or centrifugal), if known.
RAI 6
The Fort Calhoun Updated Safety Analysis Report (USAR) includes Chapter 15, which summarizes the applicants aging management programs, time-limited aging analyses, and license renewal commitments that will be implemented to manage or analyze the effects of aging during the period of extended operation. This USAR section includes Commitment Nos. 16, 17, and 18 relative to recommended inspections, evaluations, or analyses for RVI components at Fort Calhoun. However, these commitments were established well before the industrys development of the augmented inspection and evaluation guidelines for CE-designed plants in MRP-227-A. Thus, the staff seeks information on how implementation of the new MRP-based program relates to these commitments. Specifically, clarify and justify how implementation of the MRP-227-A based program for Fort Calhoun supersedes, augments, adjusts, replaces or fulfills the criteria for RVI components in Commitment Nos. 16, 17, and 18 of USAR Chapter 15.
RAI 7
For three of the Primary inspection category components applicable to FCS, MRP-227-A permits a demonstration of fatigue life versus a time-limited aging analysis (TLAA) instead of inspection. These components are the Core Support Barrel Assembly - Lower Flange Weld the Lower Support Structure - Core Support Plate, and, Upper Internals Assembly--Fuel Alignment Plate. Details of these plant-specific fatigue evaluations were not provided in the RVI Program Description.
Clarify whether the current licensing basis (CLB) or current design basis for these components includes either a cumulative usage factor analysis, implicit 7000 cycle maximum allowable stress range reduction analysis, fatigue flaw growth analysis or cycle-based flaw tolerance analysis, and if so, whether the applicable cycle-based TLAA was previously found to be acceptable in accordance with the TLAA acceptance criterion in either Title 10 of the Code of Federal Regulations (10 CFR) 54.21(c)(1)(i) or (ii). Otherwise, justify why your program does not propose inspections of the components using the MRP-227-A recommended inspection bases if either the CLB does not include an applicable fatigue-based or cycle-based TLAA for each of these components, or if the TLAAs were not found to be acceptable previously in NUREG-1782, Safety Evaluation Report Related to the License Renewal of the Fort Calhoun Station, Unit 1, in accordance with either 10 CFR 54.21(c)(1)(i) or (ii).
RAI 8
Table C-1 on page C-5, the licensee stated that the inspection requirements for deep beams in lower support structure are not required for FCS. Since FCS has full-height shroud plates, the inspection criteria should be applicable to FCS. Please confirm.
RAI 9
In Section 4.4.2 in MRP-227-A report, the MRP stated that for all CE units, the American Society of Mechanical Engineers (ASME) Code,Section XI inspection criteria applies to all of the following RVI components binned in Existing inspection category(a) guide lugs and lug inserts and bolts; and (b) fuel alignment plates. However, in Table C-3 in the September 27, 2012 submittal, the licensee stated that the AMP for the aforementioned RVI components is not applicable to FCS, and this position is inconsistent with the MRP-227-A. Please provide an explanation for this inconsistency. : Threshold Limits for TE and IE in CASS
ATTACHMENT 1 Threshold limits developed by the staff for thermal (TE) and irradiation embrittlement (IE) in cast austenitic stainless steel (CASS).
(MRP-227-Action Item 7 in the staffs safety evaluation (SE) for the MRP-227-A)
Background
On March 12, 2014, the staff held a conference call with MRP-EPRI members and conveyed the staffs position (addressed below) related to the threshold limits for thermal (TE) and neutron irradiation embrittlement (IE) in cast austenitic stainless steel (CASS).
Based on the lower bound estimate of fracture toughness of austenitic stainless steel materials as a function of fluence from NUREG/CR-7027, Degradation of LWR Core Internal Materials Due to Neutron Irradiation, the NRR staff proposes that the fluence screening criteria in the Grimes letter dated May 19, 2000, (ADAMS Accession No. ML003717179) for all CF-3 and CF-8 materials be increased to 0.45 dpa (displacements per atom (dpa) = 3x1020 n/cm2 (E > 1MeV)), as described in the following tables. Furthermore, the NRR staff proposes that CF-3 and CF-8 materials in an RVI component can be treated as a wrought stainless steel when the ferrite levels are low enough such that significant TE is not expected to occur; with the current state of knowledge, the staff considers the maximum ferrite level is 15%. In MRP-175 and MRP-227, wrought stainless steel requires aging management when the neutron fluence > 1.5 dpa.
Therefore, the staff recommends that the current Grimes letter be revised to read as follows the CF-3 and CF-8 materials can be treated as a wrought stainless steel for aging management when the ferrite content is less than or equal to 20% and its exposure to neutron fluence is 0.45 dpa. For CF-3 and CF-8 materials that are exposed to a fluence value > 0.45 dpa and
< 1.5 dpa, aging management is required when the ferrite content exceeds 15%. Above 1.5 dpa, all stainless steel RVI components are considered susceptible to loss of fracture toughness and require aging management.
Proposed NRC position on Screening of CASS Reactor Vessel Internal Components The NRC staff requires licensees/applicants to consider embrittlement from both TE and IE.
The staff would consider screening for susceptibility to embrittlement based on fluence and ferrite content (measured or calculated from Hulls equation and summarized in Tables A, B, and C shown below) to be conservative.
Table A Screening for Components with < 0.45 dpa neutron exposure Casting Molybdenum (wt. %) Susceptibility Delta ferrite %
Method TE > 14%
static High 2.0-3.0% No 14%
(CF-8M) TE > 20%
centrifugal No 20%
TE > 20%
static Low 0.5% max No 20%
(CF-3 and CF-8) centrifugal No All Table B Screening for Components with 0.45 dpa neutron exposure 1.5 dpa Casting Molybdenum (wt. %) Susceptibility Delta ferrite %
static High 2.0-3.0% No 10%
centrifugal No 15%
static Low 0.5% max No 15%
(CF-3 and CF-8) centrifugal No All Table C Screening for Components with > 1.5 dpa neutron exposure Casting Molybdenum (wt. %) Susceptibility Delta ferrite %
static High 2.0-3.0% IE 10%
centrifugal IE 15%
static Low 0.5% max IE 15%
(CF-3 and CF-8) centrifugal IE All