ML16089A226
| ML16089A226 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/22/2016 |
| From: | Hafenstine C Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA 16-0025 | |
| Download: ML16089A226 (6) | |
Text
Cynthia R. Hafenstine Manager Regulatory Affairs March 22, 2016 RA 16-0025 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555
Reference:
- 1)
Letter RA 15-0080, dated November 4, 2015, from C. R. Hafenstine, WCNOC to USNRC
Subject:
Gentlemen:
- 2)
Westinghouse Letter L TR-LIS-16-53, dated February 18, 2016, "Wolf Creek 10 CFR 50.46 Annual Notification and Reporting for 2015" Docket No. 50-482:
10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the annual reporting requirement for the Wolf Creek Generating Station (WCGS).
In Reference 1, WCNOC submitted a 30 Day Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes to the Nuclear Regulatory Commission (NRC) to report changes of greater than 50°F in the peak cladding temperature (PCT) from those previously reported for a large break loss-of-coolant accident (LOCA).
Those changes were due to implementation of a new best-estimate large break LOCA methodology known as Automated Statistical Treatment of Uncertainty Method (ASTRUM) that was approved by the NRC for WCGS.
WCNOC has reviewed Reference 2, which addresses 10 CFR 50.46 reporting information pertaining to the ECCS Evaluation Model changes that were implemented by Westinghouse for 2015. The review concludes that with the exception of the above changes that were already addressed in a 30 day ECCS model change report (Reference 1 ), the additional effect of changes to, or errors in, the Evaluation Models on the limiting transient PCT is not significant for 2015. Therefore, changes to the ECCS Evaluation Model are being reported as an annual report.
P.O. Box 411 / Burlington, KS 66839 I Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET
RA 16-0025 Page 2 of 2 Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2015.
Except for the exceptions noted above, these model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rack-up forms.
Attachment II provides PCT rack-up forms for the calculated Large Break LOCA and Small Break LOCA PCT margin allocations in effect for the 2015 WCGS Evaluation Models. The PCT values determined in the Large Break and Small Break LOCA analysis of record, combined with all of the PCT allocations, remain below the 10 CFR 50.46(b)(1) regulatory limit of 2200 °F.
Therefore, WCGS is in compliance with 10 CFR 50.46 requirements and no reanalysis or other action is required.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204 or Bill Muilenburg at 620-364-4186.
CRH/rlt Attachment II Sincerely, Cynthia R. Hafenstine Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss-of-Coolant Accidents (LOCA)
Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms cc:
M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), w/a Senior Resident Inspector (NRC), w/a
Attachment I to RA 16-0025 Page 1 of 2 ASSESSMENT OF CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS FOR LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENTS (LOCA)
GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451 "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting."
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0 °F.
LOWER SUPPORT
- PLATE, CORE
- BARREL, AND VESSEL WALL UNHEATED CONDUCTOR ERRORS
Background
Modeling errors were discovered in the lower support plate, core barrel, and vessel cladding unheated conductors in the Best-Estimate Large Break Loss-of-Coolant Accident (BELBLOCA) analysis-of-record. The modeling errors impacted the volume and surface area of the core barrel, the surface area and thermal resistance of the lower support plate, and the thermal resistance of the vessel wall.
The resolution of these issues represents a closely-related group of Non-Discretionary Changes in the application of the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A qualitative evaluation was completed concluding that the modeled net stored energy and heat transfer rate of the vessel wall, core barrel, and lower support plate unheated conductors were adequate. This error is estimated to have a PCT impact of 0°F.
Attachment I to RA 16-0025 Page 2 of 2 CORE CHANNEL GAP ERROR
Background
A modeling error was discovered in the BELBLOCA analysis-of-record. The modeling error over-represented the flow area between the guide tube and non-guide tube core average channels.
The resolution of this issue represents a Non-Discretionary Change iri the application of the Evaluation Model as described in Section 4.1.2 ofWCAP-13451.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A qualitative evaluation was completed concluding the magnitude of the error is negligible.
The error is estimated to have a PCT impact of 0°F.
Attachment II to RA 16-0025 Page 1 of 2 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORMS Evaluation Model:
Fuel:
Peaking Factor:
SG Tube Plugging:
Power Level:
Limiting Break Size:
LICENSING BASIS
ASTRUM (2004)
RFA-2 FQ=2.50, FdH=1.65 10%
3565 MWth DEG Clad Temp (°F)
Ref.
Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (~PCT)
A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. Containment Fan Cooler Capacity
- 2. Decay Group Uncertainty Factors Errors B. PLANNED PLANT CHANGE EVALUATIONS
- 1. Containment Fan Cooler Capacity
- c. 2015 PERMANENT ECCS MODEL ASSESSMENTS
- 1. None D. OTHER
- 1. None 1900 °F 1
0 2
-10 3
0 2
(a) 0 0
LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1890 °F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES SINCE LAST 30-DAY REPORT (LETTER RA 15-0080)
References:
rj ~PCT! = o °F
- 1.
WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology,"
January 2014.
- 2.
LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.
- 3.
L TR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.
Notes:
(a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.
Attachment II to RA 16-0025 Page 2of2
Evaluation Model:
Fuel:
Peaking Factor:
SG Tube Plugging:
Power Level:
Limiting transient:
LICENSING BASIS 1985 EM with NOTRUMP 17x17 RFA-2 w/IFM FQ=2.50, FdH=1.65 10%
3565 MWth 4-inch Break Clad Temp (°F)
Ref.
Notes Analysis of Record PCT 936 °F 1
MARGIN ALLOCATIONS (APCT)
A.
PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. None 0
B.
PLANNED PLANT CHANGE EVALUATIONS
- 1. Loose Part Evaluation 45 2
(a)
- c.
2015 PERMANENT ECCS MODEL ASSESSMENTS
- 1. None 0
D.
TEMPORARY ECCS MODEL ISSUES
- 1. None 0
E.
OTHER
- 1. None 0
LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 981 °F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES rl APCTI = o °F
References:
- 1. WCAP-16717-P, Rev. 0, "Wolf Creek Generating Station (SAP), MSIV/MFIV Replacement Project, Small Break Loss of Coolant Accident Analysis Engineering Report," January 2007.
- 2. SAP-90-148/NS-OPLS-OPL-1-90-239, 'Wolf Creek Nuclear Operating Corporation, RCS Loose Part Evaluation," April 1990.
Notes:
(a) This penalty will be carried to track the loose part which has not been recovered.
Cynthia R. Hafenstine Manager Regulatory Affairs March 22, 2016 RA 16-0025 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555
Reference:
- 1)
Letter RA 15-0080, dated November 4, 2015, from C. R. Hafenstine, WCNOC to USNRC
Subject:
Gentlemen:
- 2)
Westinghouse Letter L TR-LIS-16-53, dated February 18, 2016, "Wolf Creek 10 CFR 50.46 Annual Notification and Reporting for 2015" Docket No. 50-482:
10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the annual reporting requirement for the Wolf Creek Generating Station (WCGS).
In Reference 1, WCNOC submitted a 30 Day Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes to the Nuclear Regulatory Commission (NRC) to report changes of greater than 50°F in the peak cladding temperature (PCT) from those previously reported for a large break loss-of-coolant accident (LOCA).
Those changes were due to implementation of a new best-estimate large break LOCA methodology known as Automated Statistical Treatment of Uncertainty Method (ASTRUM) that was approved by the NRC for WCGS.
WCNOC has reviewed Reference 2, which addresses 10 CFR 50.46 reporting information pertaining to the ECCS Evaluation Model changes that were implemented by Westinghouse for 2015. The review concludes that with the exception of the above changes that were already addressed in a 30 day ECCS model change report (Reference 1 ), the additional effect of changes to, or errors in, the Evaluation Models on the limiting transient PCT is not significant for 2015. Therefore, changes to the ECCS Evaluation Model are being reported as an annual report.
P.O. Box 411 / Burlington, KS 66839 I Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET
RA 16-0025 Page 2 of 2 Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2015.
Except for the exceptions noted above, these model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rack-up forms.
Attachment II provides PCT rack-up forms for the calculated Large Break LOCA and Small Break LOCA PCT margin allocations in effect for the 2015 WCGS Evaluation Models. The PCT values determined in the Large Break and Small Break LOCA analysis of record, combined with all of the PCT allocations, remain below the 10 CFR 50.46(b)(1) regulatory limit of 2200 °F.
Therefore, WCGS is in compliance with 10 CFR 50.46 requirements and no reanalysis or other action is required.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204 or Bill Muilenburg at 620-364-4186.
CRH/rlt Attachment II Sincerely, Cynthia R. Hafenstine Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss-of-Coolant Accidents (LOCA)
Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms cc:
M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), w/a Senior Resident Inspector (NRC), w/a
Attachment I to RA 16-0025 Page 1 of 2 ASSESSMENT OF CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS FOR LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENTS (LOCA)
GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451 "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting."
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0 °F.
LOWER SUPPORT
- PLATE, CORE
- BARREL, AND VESSEL WALL UNHEATED CONDUCTOR ERRORS
Background
Modeling errors were discovered in the lower support plate, core barrel, and vessel cladding unheated conductors in the Best-Estimate Large Break Loss-of-Coolant Accident (BELBLOCA) analysis-of-record. The modeling errors impacted the volume and surface area of the core barrel, the surface area and thermal resistance of the lower support plate, and the thermal resistance of the vessel wall.
The resolution of these issues represents a closely-related group of Non-Discretionary Changes in the application of the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A qualitative evaluation was completed concluding that the modeled net stored energy and heat transfer rate of the vessel wall, core barrel, and lower support plate unheated conductors were adequate. This error is estimated to have a PCT impact of 0°F.
Attachment I to RA 16-0025 Page 2 of 2 CORE CHANNEL GAP ERROR
Background
A modeling error was discovered in the BELBLOCA analysis-of-record. The modeling error over-represented the flow area between the guide tube and non-guide tube core average channels.
The resolution of this issue represents a Non-Discretionary Change iri the application of the Evaluation Model as described in Section 4.1.2 ofWCAP-13451.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A qualitative evaluation was completed concluding the magnitude of the error is negligible.
The error is estimated to have a PCT impact of 0°F.
Attachment II to RA 16-0025 Page 1 of 2 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORMS Evaluation Model:
Fuel:
Peaking Factor:
SG Tube Plugging:
Power Level:
Limiting Break Size:
LICENSING BASIS
ASTRUM (2004)
RFA-2 FQ=2.50, FdH=1.65 10%
3565 MWth DEG Clad Temp (°F)
Ref.
Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (~PCT)
A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. Containment Fan Cooler Capacity
- 2. Decay Group Uncertainty Factors Errors B. PLANNED PLANT CHANGE EVALUATIONS
- 1. Containment Fan Cooler Capacity
- c. 2015 PERMANENT ECCS MODEL ASSESSMENTS
- 1. None D. OTHER
- 1. None 1900 °F 1
0 2
-10 3
0 2
(a) 0 0
LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1890 °F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES SINCE LAST 30-DAY REPORT (LETTER RA 15-0080)
References:
rj ~PCT! = o °F
- 1.
WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology,"
January 2014.
- 2.
LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.
- 3.
L TR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.
Notes:
(a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.
Attachment II to RA 16-0025 Page 2of2
Evaluation Model:
Fuel:
Peaking Factor:
SG Tube Plugging:
Power Level:
Limiting transient:
LICENSING BASIS 1985 EM with NOTRUMP 17x17 RFA-2 w/IFM FQ=2.50, FdH=1.65 10%
3565 MWth 4-inch Break Clad Temp (°F)
Ref.
Notes Analysis of Record PCT 936 °F 1
MARGIN ALLOCATIONS (APCT)
A.
PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. None 0
B.
PLANNED PLANT CHANGE EVALUATIONS
- 1. Loose Part Evaluation 45 2
(a)
- c.
2015 PERMANENT ECCS MODEL ASSESSMENTS
- 1. None 0
D.
TEMPORARY ECCS MODEL ISSUES
- 1. None 0
E.
OTHER
- 1. None 0
LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 981 °F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES rl APCTI = o °F
References:
- 1. WCAP-16717-P, Rev. 0, "Wolf Creek Generating Station (SAP), MSIV/MFIV Replacement Project, Small Break Loss of Coolant Accident Analysis Engineering Report," January 2007.
- 2. SAP-90-148/NS-OPLS-OPL-1-90-239, 'Wolf Creek Nuclear Operating Corporation, RCS Loose Part Evaluation," April 1990.
Notes:
(a) This penalty will be carried to track the loose part which has not been recovered.