ML082971068
| ML082971068 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  (DPR-026, DPR-064) |
| Issue date: | 11/20/2008 |
| From: | Boska J Plant Licensing Branch 1 |
| To: | Entergy Nuclear Operations |
| Boska J, NRR, 301-415-2901 | |
| References | |
| TAC MD9196, TAC MD9197 | |
| Download: ML082971068 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST FOR VESSEL WELD INSPECTION EXTENSION (TAC NOS. MD9196 AND MD9197)
Dear Sir or Madam:
By letter dated July 8, 2008, Entergy Nuclear Operations, Inc. (Entergy), the licensee for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), submitted requests for relief from American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, inservice inspection (lSI) requirements regarding examination of certain reactor pressure vessel welds at IP2 and IP3. Instead, an alternative lSI interval in accordance with Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 2008 (ADAMS Accession No. ML082820046), was proposed. The Nuclear Regulatory Commission (NRC) had approved this topical report.
The NRC staff is reviewing the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). On November 14, 2008, the Entergy staff indicated that a response to the RAI would be provided by approximately December 23, 2008.
Please contact me at (301) 415-2901 if you have any questions on this issue.
Sincerely,
~P~sk~rojectManager a~:~t ILiCenSing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosure:
RAI cc w/encl: Distribution via Listserv
REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST TO EXTEND VESSEL WELD INSPECTIONS ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 By letter dated July 8, 2008, Entergy Nuclear Operations, Inc. (Entergy), the licensee for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), submitted requests for relief from American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, inservice inspection (lSI) requirements regarding examination of certain reactor pressure vessel welds at IP2 and IP3. Instead, an alternative lSI interval of 20 years in accordance with Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 2008 (ADAMS Accession No. ML082820046), was proposed. The Nuclear Regulatory Commission (NRC) had approved this topical report.
The NRC staff is reviewing the submittal and has determined that additional information is needed to complete its review. The specific questions are listed below:
The estimated bounding IP2 and IP3 through-wall cracking frequency (TWCF) is about an order of magnitude above the screening criteria change in TWCF for Westinghouse plants that was reported in WCAP 16168-NP-A, Revision 2, Table 4-1. This indicates that IP's design and operating characteristics are not bounded by the bounding evaluation relied on by the NRC staff to endorse increasing the inspection intervals for reactor vessel welds from 10 to 20 years.
- 1.
Please provide a plant-specific estimate of the change in risk associated with the requested interval extension for IP2 and IP3. This change in risk estimate may use the methods endorsed by the NRC staff in approving the WCAP (which were developed as part of the technical basis for the pressurized thermal shock (PTS) rulemaking).
- 2.
Please summarize the principal steps in the methodology used in the WCAP and confirm that the Entergy methodology is consistent with the WCAP or justify any differences. The summary should briefly describe the major steps in the methodology and should demonstrate that Entergy appropriately applied the WCAP methodology in its calculations.
- 3.
Please describe which results from the PTS technical basis results were used as input to IP2 and IP3's change in risk calculations, and which inputs were developed as IP2 and IP3 specific inputs.
- 4.
If IP2 and IP3 used the PTS sequence frequency and binning from one of the reference plants in the WCAP, please justify that these results are bounding compared to the IP specific values, or that any differences between IP2 or IP3 and the reference plant are not expected to appreciably increase IP's estimated change in risk if IP2 or IP3 specific frequency and binning were used. A methodology to compare plant-specific PTS Enclosure
- 2 characteristics to the characteristics of a reference plant is described in Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants" (ML042880482).
- 5.
Section 5 of Relief Request (RR)-76 for IP2 proposed, "To bound IP2 the Westinghouse pilot plant was reevaluated at a value of EFPY (Condition B) that is well beyond 60 EFPY." Here, EFPY stands for effective full-power years, and Condition B stands for the extended embrittlement level with a mean TWCF around 1x10-6 as defined in NUREG 1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)." The approach described is a significant extension of the analyses previously accepted by the NRC staff when WCAP-16168 was approved. This extension could bring the results much closer to the risk-informed decision-making criteria in Regulatory (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." If Entergy continues to pursue the alternative, please reevaluate the potential sources of uncertainty in the calculation of TWCF and l1TWCF (as defined in WCAP 16168, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection [lSI]
Interval,") to provide confidence that the results which are proposed to bound IP2 and IP3 are consistent with RG 1.174.
- 6.
Table 3 of RR-76 and RR-3-43(1) contains plant-specific information for the TWCFs that you calculated for the IP2 and IP3 reactor pressure vessels (RPVs). To complete the review, the NRC staff requests you:
Explain how the neutron fluence and flux values for the IP2 and IP3 RPVs were obtained, including references to these calculated results where necessary.
Identify source documents for determining the manganese content for the IP2 and IP3 RPV materials and describe how you averaged them when more than one data source is available.
Provide your calculations for obtaining the embrittlement shift, l1T30, for the circumferential weld of the IP2 RPV. The NRC staff has verified your l1T30 values within 1 degree F for the axial weld and the plate of the IP2 RPV, but is not able to conclude that the l1T30 for the circumferential weld is the same as the l1T30 for the plate.
Provide your calculations for obtaining the reference temperature for the axial weld (RTMAX-AW) and the reference temperature for the circumferential weld (RTMAX-CW) of the IP2 RPV. Please note that determination of RTMAX-AW and RTMAX-CW as defined in WCAP-16168 requires consideration of reference temperatures of adjacent plate or forging.
- Section 3.4 in the staff's safety evaluation (SE) ofWCAP-16168 indicated the RTMAX-X and the l1T30 value must be calculated using the latest approved methodology documented in RG 1.99 or other NRC approved methodology. The latest approved methodology is documented in RG 1.99, Revision 2. This RG also specifies a methodology for evaluating reactor vessel surveillance material data. An alternate methodology for calculating l1T30 value is documented in an
- 3 alternative PTS rule, 10 CFR 50.61(a), which was published for public comment in the Federal Register on August 11, 2008. The alternative PTS rule also documents a methodology for evaluating reactor vessel surveillance material data. If the licensee determines RTMAX-X and boT3o values using either the methodologies in RG 1.99, Revision 2 or the alternative PTS rule, the licensee must provide the NRC the analysis of its reactor vessel materials surveillance data.
- 7. During a teleconference between the licensee and the NRC staff on November 5,2008, the licensee indicated that the current applicable ASME Code of record for inservice examination is the 2001 Edition with 2002 Addenda for Indian Point Unit No.2 and the 1989 Edition, no Addenda for Indian Point Unit NO.3.
Considering the unique nature of the RRs, which were intended to apply to the end of the current license, the NRC staff requests you revise the RRs according to the following:
- The edition and addenda of the Code should be the recent year ASME Code that you would use for the fourth interval for other ASME Code applications.
Table 2 of the RR provides the results from prior reactor vessel examinations.
This table indicates that the inspection method for the prior examinations were performed in accordance with RG 1.150. This is an acceptable methodology for prior examinations; however, the staff requires that all future examinations be qualified in accordance with ASME Code,Section XI, Appendix VIII.
Enclosures 1 and 2 to NL-08-096 indicate that the fourth reactor vessel inspection is proposed to be completed by 2032 for Indian Point Unit No.2 and by 2035 for Indian Point Unit No.3. These dates are beyond the current licensing term for each of these units. The NRC can only approve a RR until the end of the licensee's current license term. Therefore, the RR should be modified to identify all inspections that are planned within the current licensing term. If Indian Point Unit Nos. 2 and 3 have their licenses extended in the future, a new RR will be needed for the extended license terms.
ML082820046), was proposed. The Nuclear Regulatory Commission (NRC) had approved this topical report.
The NRC staff is reviewing the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). On November 14, 2008, the Entergy staff indicated that a response to the RAI would be provided by approximately December 23, 2008.
Please contact me at (301) 415-2901 if you have any questions on this issue.
Sincerely,
/RA!
John P. Boska, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosure:
RAI cc w/encl: Distribution via Listserv DISTRIBUTION:
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