ML13269A323
| ML13269A323 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/05/2013 |
| From: | Nicholas Difrancesco Plant Licensing Branch III |
| To: | Pacilio M Exelon Generation Co |
| Nicholas DiFrancesco, NRR/DORL 415-1115 | |
| References | |
| TAC MF0500 | |
| Download: ML13269A323 (4) | |
Text
Mr. Michael J. Pacilio Senior Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 5, 2013 Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)
Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
LASALLE COUNTY STATION, UNIT 1 AND 2-REVIEW CLOSEOUT OF SURVEILLANCE CAPSULE REPORT AND NON-CONSERVATIVE PRESSURE-TEMPERATURE LIMIT CURVE TECHNICAL SPECIFICATION (TAC NO. MF0500)
Dear Mr. Pacilio:
Appendix H to Part 50 of Title 10, Code of Federal Regulations (1 0 CFR Part 50, Appendix H),
"Reactor Vessel Material Surveillance Program Requirements" (the rule), provides the U.S Nuclear Regulatory Commission (NRC) surveillance and testing requirements for ferritic components of the reactor vessels (RVs) of operating U.S. light-water reactors. The rule requires licensed owners of U.S. light-water reactor facilities to install a number of surveillance capsules within the cavities of their RVs, and to remove capsules and test the capsule materials in accordance with the withdrawal schedule and testing requirements of the American Society for Testing and Materials (ASTM) Standard Practice E-185-82. Paragraph IV.A of the rule requires the RV material surveillance capsule test results to be the subject of a summary technical report that is required to be submitted to the NRC within one year of the capsule withdrawal date. Paragraph IV.B of the rule requires that these reports shall include all data specified by ASTM Standard Practice E-185-82, and the results of all fracture toughness tests conducted on the surveillance capsule materials in both the unirradiated and irradiated condition. Paragraph IV.C of the rule requires the licensee provide the NRC with expected date for submittal of the revised Technical Specifications if the licensee finds that the current Technical Specifications are non-conservative.
By letter dated January 1 0, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13011A005), Exelon Generation Company, LLC (the licensee) submitted a letter to the NRC concerning the identification of non-conservative pressure-temperature limits (P!f limits) for LaSalle County Station, Unit 1 (LaSalle, Unit 1 ), when the information from the 120° surveillance capsule was analyzed in October 2012. The capsule material was reported as having received a fluence of 3.66 x 1017 n/cm2 (E > 1.0 MeV) when it was removed from service on February 12, 2010. The capsule data was submitted in a report to the NRC as BWRVIP-250-NP on November 18, 2011 (ADAMS Accession No. ML11326A290),
21 months after removal. The expected date for submittal of the revised P!f limits technical specification (TS) was provided to the NRC staff by letter dated January 10, 2013, 35 months after removal of the surveillance capsule. The licensee's evaluation of the capsules surveillance data and review of P!f limits was not completed within one year of capsule withdrawal contrary to 10 CFR Part 50, Appendix H reporting requirements. The staff has discussed the matter with LaSalle's resident inspectors for follow-up as part of the Reactor Oversight Process (ROP).
The surveillance program for LaSalle, Unit 1, utilizes the plant's own surveillance capsules as the basis for monitoring the embrittlement trends for the RV beltline materials. The capsule is also part of the BWRVIP-86-A, Integrated Surveillance Program (ISP). The surveillance weld materials were fabricated from submerged arc weld heat No. 1 P3571, Linde 1092 flux, Lot 3958, material. The surveillance plate materials were made from RV plate heat No. C6345-1.
Following review of the capsule surveillance data, the licensee completed an operability determination and established administrative controls for Technical Specification 3.4.11,
"[Reactor Coolant System] RCS Pressure and Temperature (PIT) Limits," since the existing limits were found to be non-conservative. The operability determination used a method of evaluation different from NRC-approved topical reports (TRs) or prior plant-specific approvals.
The licensee used methods described in General Electric-Hitachi (GEH) licensing report TR NED0-33178-A, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves" (ADAMS Accession No. ML092370487), to assess the operability of LaSalle, Unit 1, P-T limits. A limitation and condition of the GEH P-T method requires that fluence must be calculated with a single, NRC-approved, fluence method.
The licensee used an alternative method to calculate fluence which used two different fluence models: one fluence calculation using methods described in BWRVIP-114NP-A, "RAMA Fluence Methodology Theory Manual" (ADAMS Accession No. ML092650376) and second using methods described in NED0-32983-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations" (ADAMS Accession No. ML072480121 ). The NRC staff is concerned that the utilization of two fluence models impacts the biases and uncertainties in the individually approved fluence models. These fluence models were not developed or approved for this intended purpose and the staff is unaware of qualification demonstrating that fluence models can be combined in this manner. Furthermore, such approaches are not addressed by the guidance contained in NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (ADAMS Accession No. ML010890301 ). As such, this letter is not an NRC staff endorsement of the method used to perform the licensee's operability determination.
In the January 10, 2013, letter, the licensee made a regulatory commitment to submit a license amendment request (LAR) by December 31, 2013, to correct the non-conservative TSs for the PIT limits. The current fluence for LaSalle, Unit 1, is between 22-23 effective full power years (EFPY). The current TS PIT limits are based upon a fluence up to 32 EFPY. The licensee's operability determination estimate that the current PIT limit curve remains valid until 27 EFPY at current licensed power levels. Therefore, the staff determined that follow-up actions including review under the ROP and the licensee's regulatory commitment to submit a LAR are appropriate actions to address staff concerns and to implement permanent corrective actions.
Please contact me at {301) 415-1115, if you have any questions.
Docket No. 50-373 cc w/encl: ListServ Sincerely, Nicholas J. DiFrancesco, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ML13269A323 OFFICE LPL3-21PM LPL3-21LA DEIEVIB IBC DSSISRXBIBC LPL3-21BC LPL3-21PM NAME NDiFrancesco SRohrer SRosenberg CJackson TTate NDifrancesco DATE 9130/13 10101113 10102113 10108113 1215113 121512013