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{{#Wiki_filter: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
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 17, 2015 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


==SUBJECT:==
==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)  
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:==
==Dear Mr. Lieb:==
By application dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14353A228), 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.
By application dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14353A228), 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 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 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.
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  
R. Lieb                                         If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315.
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 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. ML14353A228), states that: FENOC is currently performing a [loss-of-coolant accident]
Request for Additional Information cc w/encl: Distribution via Listserv
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.
REQUEST FOR ADDITIONAL INFORMATION EVALUATION MODEL ERRORS FIRSTENERGY NUCLEAR OPERATING COMPANY (FENOC)
However, the change in fuel temperature uncertainty discussed above has not been submitted to the NRC staff for generic review and approval.
DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346
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.
: 1. The report dated December 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14353A228), states that:
This model change will significantly change the predicted emergency core cooling performance.
FENOC is currently performing a [loss-of-coolant accident] LOCA reanalysis, which is scheduled to be completed by July 31, 2015.
Regarding the evaluation of ECCS performance, 10 CFR 50.46(a)(1  
AREVA recommends FENOC perform a reanalysis with the revised
)(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
[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.
)(i) requirement, regarding assurance that the most severe hypothetical LOCAs are calculated.
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 BAW-10192P-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.
: 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.'
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).
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.
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.
R. Lieb If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315.
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 loss-of-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
Docket No. 50-346


==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 highest-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 DISTRIBUTION:
ML15062A651              *via-email                    NRR-088 OFFICE     LPLll-2/PM       LPLI 11-2/LA   SN PB/BC          LPLI 11-2/BC NAME       EBrown           SRohrer         JDean            TTate DATE         3/17/15         3/17/15       3/3/15             3/17/15}}
Sincerely, IRA/ 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 PUBLIC LPLlll-2 Reading File RidsAcrsAcnw_MailCTR Resource RidsNrrDorllpll 1-2 Resource RidsNrrDssSrxb Resource RidsNrrDorl Resource RidsNrrPMDavisBesse Resource RidsOgcMailCenter Resource RidsNrrLASRohrer Resource JWhitman RidsNrrDssSnpb ADAMS Accession No. ML15062A651 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 OFFICIAL RECORD COPY BP arks JRowley RidsRg3MailCenter Resource NRR-088 LPLI 11-2/BC TT ate 3/17/15}}

Latest revision as of 14:37, 31 October 2019

Request for Additional Information Concerning Report of Errors in Evaluation Model (TAC No. MF5578) (L-14-403)
ML15062A651
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 03/17/2015
From: Ellen Brown
Plant Licensing Branch III
To: Lieb R
FirstEnergy Nuclear Operating Co
Eva Brown, NRR/DORL
References
L-14-403, TAC MF5578
Download: ML15062A651 (5)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 17, 2015 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

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. ML14353A228), 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.

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 Docket No. 50-346

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

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. ML14353A228), 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 BAW-10192P-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 loss-of-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 highest-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.

ML15062A651 *via-email NRR-088 OFFICE LPLll-2/PM LPLI 11-2/LA SN PB/BC LPLI 11-2/BC NAME EBrown SRohrer JDean TTate DATE 3/17/15 3/17/15 3/3/15 3/17/15

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