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==Dear Mr. Lieb:== | ==Dear Mr. Lieb:== | ||
By application dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), FirstEnergy Nuclear Operating Company (FE NOC) submitted a report in accordance with Section 50.46(a)(3) to Title 10 of the Code of Federal Regulations (10 CFR). This report describes an error identified in the emergency core cooling system evaluation model, and an estimate of the effect of the error on the predicted peak cladding temperature (PCT). Correction for this error caused an increase in the PCT predicted for the facility. Most notably, thermal conductivity degradation causes the predicted PCT for loss of coolant accidents that initiate at middle of life or end of life core conditions to increase significantly. The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff evaluated the report and determined that additional information is required to evaluate whether the report satisfies the reporting requirements of 10 CFR 50.46(a)(3). The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Phil Lashley of your staff on February 25, 2015, it was agreed that FENOC would provide a response within 30 days from the date of this letter. The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. | By application dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), FirstEnergy Nuclear Operating Company (FE NOC) submitted a report in accordance with Section 50.46(a)(3) to Title 10 of the Code of Federal Regulations (10 CFR). This report describes an error identified in the emergency core cooling system evaluation model, and an estimate of the effect of the error on the predicted peak cladding temperature (PCT). Correction for this error caused an increase in the PCT predicted for the facility. Most notably, thermal conductivity degradation causes the predicted PCT for loss of coolant accidents that initiate at middle of life or end of life core conditions to increase significantly. The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff evaluated the report and determined that additional information is required to evaluate whether the report satisfies the reporting requirements of 10 CFR 50.46(a)(3). The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Phil Lashley of your staff on February 25, 2015, it was agreed that FENOC would provide a response within 30 days from the date of this letter. The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. | ||
R. Lieb | R. Lieb If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315. Docket No. 50-346 | ||
==Enclosure:== | ==Enclosure:== | ||
Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, R/A Eva A. Brown, Senior Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION EVALUATION MODEL ERRORS FIRSTENERGY NUCLEAR OPERATING COMPANY (FENOC) DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 1. The report dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), states that: FENOC is currently performing a [loss-of-coolant accident] LOCA reanalysis, which is scheduled to be completed by July 31, 2015. AREVA recommends FENOC perform a reanalysis with the revised [evaluation model] EM that uses a COPERNIC2 based [thermal conductivity degradation] TCD uncertainty increase to the TAC03 and GDTACO inputs at [middle of life] MOL. Although the COPERNIC code has been approved by the U.S. Nuclear Regulatory Commission (NRC), as documented in BAW-10231 P-A, the staff does not consider the application of COPERNIC-based uncertainty values to TACO-based fuel performance methods, for application within the NRC-approved licensing topical report A, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, Volume I -Larger Break," to be accordant with NRC-approved methodology. However, the change in fuel temperature uncertainty discussed above has not been submitted to the NRC staff for generic review and approval. In light of the fact that the proposed TACO and GDTACO fuel temperature uncertainty values have not been previously reviewed and approved by the NRC, explain how FENOC will ensure that the corrected emergency core cooling system (ECCS) evaluation is performed using an acceptable evaluation model, pursuant to Section 50.46(a)(1)(i) to Title 10 to the Code of Federal Regulations (10 CFR). In the December 19, 2014, report, FENOC indicated that the TCD-related model changes will be incorporated into a version of the NRC-approved licensing topical report BAW-10192P-A. This model change will significantly change the predicted emergency core cooling performance. Regarding the evaluation of ECCS performance, 10 CFR 50.46(a)(1 )(i) states, in part, that ECCS cooling performance "must be calculated for a number of postulated coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." It was not indicated whether the implementation of the TCD-related changes will affect Enclosure the predicted emergency core cooling performance for the spectrum of break sizes, locations, and other properties, such that the existing, most limiting MOL LOCA event analyzed would become the most severe hypothetical LOCA, when reanalyzed to correct for TCD. Since the December 19, 2014, report indicates that a limited reanalysis of the peak cladding temperature case at MOL conditions will be performed, explain how this analysis will address the 10 CFR 50.46(a)(1 )(i) requirement, regarding assurance that the most severe hypothetical LOCAs are calculated. 2. Technical Specification (TS) 5.6.5.b, "Core Operating Limits Report," requires that the: analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in BAW-10179P-A, 'Safety Criteria and Methodology for Acceptable Cycle Reload Analyses.' The TS also permits the use of additional methods not described in BAW-10179P-A, provided such methods are reviewed and approved by the NRC, and identified in the applicable core operating limits report. As the application of COPERNIC-based fuel temperature uncertainties to TAC03 and GDTACO evaluation models is not consistent with NRG-approved fuel performance methodology, address how the proposed TCD correction is consistent with Section 9.2.3 of BAW-10179P-A and the associated TS requirement. | Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, R/A Eva A. Brown, Senior Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION EVALUATION MODEL ERRORS FIRSTENERGY NUCLEAR OPERATING COMPANY (FENOC) DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 1. The report dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), states that: FENOC is currently performing a [loss-of-coolant accident] LOCA reanalysis, which is scheduled to be completed by July 31, 2015. AREVA recommends FENOC perform a reanalysis with the revised [evaluation model] EM that uses a COPERNIC2 based [thermal conductivity degradation] TCD uncertainty increase to the TAC03 and GDTACO inputs at [middle of life] MOL. Although the COPERNIC code has been approved by the U.S. Nuclear Regulatory Commission (NRC), as documented in BAW-10231 P-A, the staff does not consider the application of COPERNIC-based uncertainty values to TACO-based fuel performance methods, for application within the NRC-approved licensing topical report A, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, Volume I -Larger Break," to be accordant with NRC-approved methodology. However, the change in fuel temperature uncertainty discussed above has not been submitted to the NRC staff for generic review and approval. In light of the fact that the proposed TACO and GDTACO fuel temperature uncertainty values have not been previously reviewed and approved by the NRC, explain how FENOC will ensure that the corrected emergency core cooling system (ECCS) evaluation is performed using an acceptable evaluation model, pursuant to Section 50.46(a)(1)(i) to Title 10 to the Code of Federal Regulations (10 CFR). In the December 19, 2014, report, FENOC indicated that the TCD-related model changes will be incorporated into a version of the NRC-approved licensing topical report BAW-10192P-A. This model change will significantly change the predicted emergency core cooling performance. Regarding the evaluation of ECCS performance, 10 CFR 50.46(a)(1 )(i) states, in part, that ECCS cooling performance "must be calculated for a number of postulated coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." It was not indicated whether the implementation of the TCD-related changes will affect Enclosure the predicted emergency core cooling performance for the spectrum of break sizes, locations, and other properties, such that the existing, most limiting MOL LOCA event analyzed would become the most severe hypothetical LOCA, when reanalyzed to correct for TCD. Since the December 19, 2014, report indicates that a limited reanalysis of the peak cladding temperature case at MOL conditions will be performed, explain how this analysis will address the 10 CFR 50.46(a)(1 )(i) requirement, regarding assurance that the most severe hypothetical LOCAs are calculated. 2. Technical Specification (TS) 5.6.5.b, "Core Operating Limits Report," requires that the: analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in BAW-10179P-A, 'Safety Criteria and Methodology for Acceptable Cycle Reload Analyses.' The TS also permits the use of additional methods not described in BAW-10179P-A, provided such methods are reviewed and approved by the NRC, and identified in the applicable core operating limits report. As the application of COPERNIC-based fuel temperature uncertainties to TAC03 and GDTACO evaluation models is not consistent with NRG-approved fuel performance methodology, address how the proposed TCD correction is consistent with Section 9.2.3 of BAW-10179P-A and the associated TS requirement. | ||
ML 15062A651 OFFICE LPLll-2/PM LPLI 11-2/LA NAME EBrown SRohrer DATE 3/17/15 3/17/15 *via-email SN PB/BC JDean 3/3/15}} | |||
}} | |||
Revision as of 13:02, 14 June 2018
| ML15062A651 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/17/2015 |
| From: | Brown E A Plant Licensing Branch III |
| To: | Lieb R A FirstEnergy Nuclear Operating Co |
| Eva Brown, NRR/DORL | |
| References | |
| L-14-403, TAC MF5578 | |
| Download: ML15062A651 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Raymond A. Lieb Site Vice-President FirstEnergy Nuclear Operating Company Mail Stop A-DB-3080 5501 North State, Route 2 Oak Harbor, OH 44081-0097 March 17, 2015
SUBJECT:
DAVIS-BESSE NUCLEAR POWER PLANT, UNIT NO. 1 -REQUEST FOR ADDITIONAL INFORMATION CONCERNING REPORT OF ERRORS IN EVALUATION MODEL (TAC NO. MF5578)(L-14-403)
Dear Mr. Lieb:
By application dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), FirstEnergy Nuclear Operating Company (FE NOC) submitted a report in accordance with Section 50.46(a)(3) to Title 10 of the Code of Federal Regulations (10 CFR). This report describes an error identified in the emergency core cooling system evaluation model, and an estimate of the effect of the error on the predicted peak cladding temperature (PCT). Correction for this error caused an increase in the PCT predicted for the facility. Most notably, thermal conductivity degradation causes the predicted PCT for loss of coolant accidents that initiate at middle of life or end of life core conditions to increase significantly. The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff evaluated the report and determined that additional information is required to evaluate whether the report satisfies the reporting requirements of 10 CFR 50.46(a)(3). The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Phil Lashley of your staff on February 25, 2015, it was agreed that FENOC would provide a response within 30 days from the date of this letter. The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.
R. Lieb If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315. Docket No. 50-346
Enclosure:
Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, R/A Eva A. Brown, Senior Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION EVALUATION MODEL ERRORS FIRSTENERGY NUCLEAR OPERATING COMPANY (FENOC) DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 1. The report dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14353A228), states that: FENOC is currently performing a [loss-of-coolant accident] LOCA reanalysis, which is scheduled to be completed by July 31, 2015. AREVA recommends FENOC perform a reanalysis with the revised [evaluation model] EM that uses a COPERNIC2 based [thermal conductivity degradation] TCD uncertainty increase to the TAC03 and GDTACO inputs at [middle of life] MOL. Although the COPERNIC code has been approved by the U.S. Nuclear Regulatory Commission (NRC), as documented in BAW-10231 P-A, the staff does not consider the application of COPERNIC-based uncertainty values to TACO-based fuel performance methods, for application within the NRC-approved licensing topical report A, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, Volume I -Larger Break," to be accordant with NRC-approved methodology. However, the change in fuel temperature uncertainty discussed above has not been submitted to the NRC staff for generic review and approval. In light of the fact that the proposed TACO and GDTACO fuel temperature uncertainty values have not been previously reviewed and approved by the NRC, explain how FENOC will ensure that the corrected emergency core cooling system (ECCS) evaluation is performed using an acceptable evaluation model, pursuant to Section 50.46(a)(1)(i) to Title 10 to the Code of Federal Regulations (10 CFR). In the December 19, 2014, report, FENOC indicated that the TCD-related model changes will be incorporated into a version of the NRC-approved licensing topical report BAW-10192P-A. This model change will significantly change the predicted emergency core cooling performance. Regarding the evaluation of ECCS performance, 10 CFR 50.46(a)(1 )(i) states, in part, that ECCS cooling performance "must be calculated for a number of postulated coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." It was not indicated whether the implementation of the TCD-related changes will affect Enclosure the predicted emergency core cooling performance for the spectrum of break sizes, locations, and other properties, such that the existing, most limiting MOL LOCA event analyzed would become the most severe hypothetical LOCA, when reanalyzed to correct for TCD. Since the December 19, 2014, report indicates that a limited reanalysis of the peak cladding temperature case at MOL conditions will be performed, explain how this analysis will address the 10 CFR 50.46(a)(1 )(i) requirement, regarding assurance that the most severe hypothetical LOCAs are calculated. 2. Technical Specification (TS) 5.6.5.b, "Core Operating Limits Report," requires that the: analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in BAW-10179P-A, 'Safety Criteria and Methodology for Acceptable Cycle Reload Analyses.' The TS also permits the use of additional methods not described in BAW-10179P-A, provided such methods are reviewed and approved by the NRC, and identified in the applicable core operating limits report. As the application of COPERNIC-based fuel temperature uncertainties to TAC03 and GDTACO evaluation models is not consistent with NRG-approved fuel performance methodology, address how the proposed TCD correction is consistent with Section 9.2.3 of BAW-10179P-A and the associated TS requirement.
ML 15062A651 OFFICE LPLll-2/PM LPLI 11-2/LA NAME EBrown SRohrer DATE 3/17/15 3/17/15 *via-email SN PB/BC JDean 3/3/15