L-12-076, Response to Nuclear Regulatory Commission Information Request Pursuant to 10 CFR 50.54(f) and Associated 10 CFR 50.46 Report for the Evaluation of Fuel Pellet Thermal Conductivity Degradation
| ML12079A111 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/16/2012 |
| From: | Harden P FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-12-076, TAC M99899 | |
| Download: ML12079A111 (20) | |
Text
FENOC FirstEnergy Nuclear Operating Company Paul A. Harden Site Vice President March 16, 2012 L-12-076 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852
SUBJECT:
Beaver Valley Power Station,Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 724-682-5234 Fax: 724-643-8069 10 CFR 50.54(f) 10 CFR 50.46 Response to Nuclear Regulatory Commission Information Request Pursuant to 10 CFR 50.54(f) and Associated 10 CFR 50.46 Report for the Evaluation of Fuel Pellet Thermal Conductivity Degradation (TAC No. M99899)
On February 16, 2012, the Nuclear Regulatory Commission (NRC) issued an information request pursuant to 10 CFR 50.54(f) to FirstEnergy Nuclear Operating Company (FENOC). The letter requested information regarding the effect of a potentially significant error, as defined in 10 CFR 50.46(a)(3)(i), associated with thermal conductivity degradation (TCD) on peak cladding temperature (PCT) in the Westinghouse Electric Company LLC (Westinghouse) furnished realistic emergency core cooling system evaluation models for the Beaver Valley Power Station, Unit Nos. 1 and 2. Attachment 1 contains the requested information. In support of FENOC's response, Westinghouse submitted directly to the NRC (Reference 1) a description of the methodology and assumptions used to determine the estimated PCT impact due to TCD.
As required by 10 CFR 50.46(a)(3)(ii), Attachment 2 provides the report for the evaluation of fuel pellet TCD in the large break loss-of-coolant accident analysis and the proposed schedule for providing a reanalysis.
The regulatory commitment associated with this correspondence is identified in. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Supervisor - Fleet Licensing, at (330) 315-6808.
Beaver Valley Power Station, Unit Nos. 1 and 2 L-12-076 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on March 16, 2012.
Attachments:
1 Response to Nuclear Regulatory Commission Information Request Pursuant to 10 CFR 50.54(f) 2 10 CFR 50.46 Report for the Evaluation of Fuel Pellet Thermal Conductivity Degradation 3 Regulatory Commitment List
Reference:
- 1. L TR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
cc: NRC Region I Administrator NRC Resident Inspector NRR Director NRR Project Manager Director BRP/DEP Site BRP/DEP Representative L-12-076 Response to Nuclear Regulatory Commission Information Request Pursuant to 10 CFR 50.54(f)
Page 1 of 7 On February 16, 2012, the Nuclear Regulatory Commission (NRC) issued an information request pursuant to 10 CFR 50.54(f) to FirstEnergy Nuclear Operating Company (FENOC). The letter requested information regarding the effeciof a potentially significant error, as defined in 10 CFR 50.46(a)(3)(i), associated with thermal conductivity degradation (TCD) on peak cladding temperature (PCT) in the Westinghouse Electric Company LLC (Westinghouse) furnished realistic emergency core cooling system evaluation models for the Beaver Valley Power Station, Unit Nos. 1 and 2. The NRC staff requests are presented below in bold type, followed by the FENOC response.
- 1.
An estimation of the effect of the thermal conductivity degradation error on the peak fuel cladding temperature calculation for the ECCS evaluations at BVPS-1 and 2.
Beaver Valley Power Station, Unit No.1 (BVPS-1) Large, Break Loss-of-Coolant Accident (LBLOCA) Analysis Estimate Results Consistent with the Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology, the most limiting PCT from each evaluation was taken as the representative PCT. Considering only the beginning-of-life peaking factor reductions, the PCT was 1858°F. Considering TCD and burndown in addition to the beginning-of-life peaking factor reductions, the PCT was 2014°F. The limiting integrated PCT case, considering all design input changes and TCD and burndown, was 1834°F, which is less than the 2200°F acceptance criterion. These are described in greater detail in the FENOC response to the second NRC information request as Case A, Case B, and Case C, respectively.
The estimated effect of TCD and burndown is the difference between the calculated PCT from Case A and Case B, or +156°F. Given the current analysis of record (AOR)
PCT of 2161°F, and an existing +2°F PCT rackup item, the estimated effect of the design input changes for 10 CFR 50.46 reporting purposes is the difference between the current rackup PCT (2163°F) and the new rackup PCT (1834°F), plus the estimated effect of TCD and burndown, or -485°F.
L-12-076 Page 2 of 7 Beaver Valley Power Station, Unit No.2 (BVPS-2) LBLOCA Analysis Estimate Results The evaluation method described in the Westinghouse letter to the NRC dated March 7, 2012 (Reference 1) for the calculation of the margin PCT was used to assess the impact of the reduction in maximum peaking factors on the BVPS-2 LBLOCA analysis. Specifically, the reference transient was rerun at beginning-of-life conditions with the reduced peaking factors, and a PCT benefit was assigned based on plant-specific run results.
The results of the evaluation of fuel pellet thermal conductivity degradation and peaking factor burndown as well as the design input changes are presented in Table 1.
Table l' Fuel Pellet TCD and Plant Peaking Factor Margin Evaluation Results Evaluation Slowdown Reflood 1 Reflood 2 Composite (OF)
(OF)
(OF)
(OF)
Analysis of Record PCT 1772 1860 1976 1976
.l\\PCT from TCD (including 0
25 10 10 burndown)
.l\\PCT from Peaking Factor
-100
-105
-190
-190 Margins Considering design input changes and TCD and burndown, the PCT was 183rF, which is less than the 2200°F acceptance criterion.
- 2.
A description of the methodology and assumptions used to determine the estimates. This description shall include consideration of experimental data relevant to thermal conductivity degradation and specific information regarding any computer code mod~1 changes which were necessary to address these data.
BVPS-1 Response The BVPS-1 LBLOCA analysis considered nuclear peaking factors in excess of those permitted by the Core Operating Limits Report (COLR). To separate this effect from the effects of TCD and other design input changes, an evaluation was performed (referred to as Case A) to consider the reduction in peaking factors from beginning-of-life, that are more representative of actual plant operation, yet still conservatively bounding. An evaluation was performed to then assess the effect of TCD and peaking factor burndown (referred to as Case B) given the updated beginning-of-life peaking factors.
An evaluation was performed (referred to as Case C) that incorporated additional design input changes and included explicit modeling of TCD and burndown. A schematic of the evaluations is shown in Figure 1.
(+2)r:lF B.
TCD+
Della TeD 156°F
- c.
TCD+
Figure 1: Results of the BVPS-1 Evaluations The evaluation of peaking factor margin with respect to plant operation considered the following input parameter changes to the LBLOCA analysis (Case A):
reduction in transient FQ [heat flux hot channel factor], including uncertainties, from 2.52 to 2.4.
reduction in steady-state FQ, without including uncertainties, from 2.2 to 1.8 reduction in FboH [enthalpy rise hot channel factor], including uncertainties, from 1.75 to 1.62 corresponding reduction in hot assembly average power, including uncertainties For the purposes of estimating the effect of TCD and peaking factor burndown, the following input changes were considered (Case B) in addition to those used in Case A:
fu'el rod design data with Performance Analysis and Design (PAD) 4.0 + TCD code peaking factor burndown shown in Table 2 L-12-076 Page 4 of 7 Burnup (MWD/MTU) 0 30,000 60,000 62,000 a e ea mg ac ors T bl 2 P k"
F t FQ Transient (1) 2.4 2.4 1.9 1.9 (1)
Includes uncertamtles.
v ersus R dB 0
urnup FQ Steady-State FLlH (1},\\:.!)
1.8 1.62 1.8 1.62 1.4 1.30 1.4 1.30 (2)
Hot assembly average power follows the same burndown, since it is a function of FllH.
Notes: FQ = heat flux hot channel factor FllH = enthalpy rise hot channel factor The integrated TCD evaluation, including all design input changes and explicit modeling of fuel TCD and peaking factor burndown, considered the following additional input parameter changes to the LBLOCA analysis (Case C) in addition to those considered for Cases A and B:
reduction in upper bound steam generator tube plugging from 22 percent to 5 percent increase in the conservatively low assumed containment pressure boundary condition (Figure 2) to address the decrease in steam generator tube plugging and existing margin L-12-076 Page 5 of 7 AO~ tf{lck preSlHtfe flevl\\>'Ii!Q fHlH;:;k pr'~~:!;!lr!!
4lJ'-r--~------------------"""I 21J to +--\\.-...-'--"---'---,-...,.I,-,--'--"--'---r...,..J---+-"~~r-'----'-~-+=T""'"'"-----'---+--i!
o Figure 2: BVPS-1 Comparison of AOR-assumed Containment Pressure Boundary Condition with that of the Design Input Changes and TCD Evaluation (Case C)
The evaluation method discussed in Reference 1 was used to determine the estimated effect of the fuel pellet TCD and peaking factor burndown. A specific adaptation of the approach, as described above, was applied for the BVPS-1 evaluation to address the difference between the as-analyzed peaking factors and those permitted by the COLR.
L-12-076 Page 6 of 7 For both Case B and Case C, which explicitly considered TCD and burndown, a total of 24 WCOBRAfTRAC executions were performed. The uncertainty attributes of these executions were taken from among the most limiting cases from the original 124-run ASTRUM analysis. The evaluation considered an adequate range of burnup such that the effects of TCD and related burnup effects were captured. HOTSPOT executions were performed for each WCOBRAfTRAC case to consider the effect of local uncertainties. Additionally, IFBA (Integral Fuel Burnable Absorber) fuel was considered for Case C.
For Case A, to consider the effect of the beginning-of-life peaking factor margins, a total of 12 WCOBRAfTRAC executions were performed. Again, the uncertainty attributes were taken from among the most limiting cases from the original 124-run ASTRUM analysis.
BVPS-2 Response The evaluation of design input changes with respect to plant operation considered the following input parameter changes to the LBLOCA analysis:
reduction in transient Fa, including uncertainties, from 2.52 to 2.4 reduction in steady-state Fa, without uncertainties, from 2.1 to 1.8 reduction in F~H, including uncertainties, from 1.75 to 1.62 corresponding reduction in hot assembly average power, including uncertainties The evaluation of fuel pellet thermal conductivity degradation considered the following input parameter changes to the LBLOCA analysis:
fuel rod design data with PAD 4.0 + TCD code peaking factor burndown shown in Table 3 Table 3-Peaking Factors Versus Rod Burnup Burnup FQ Transient (1)
FQ Steady-State FAH (1},(:.!)
(MWD/MTU) 0 2.4 1.8 1.62 30,000 2.4 1.8 1.62 60,000 1.9 1.4
.1.30 62,000 1.9 1.4 1.30 (1)
Includes uncertainties.
(2)
Hot assembly average power follows the same burndown, since it is a function of FLlH.
Notes: FQ =*heat flux hot channel factor FLlH = enthalpy rise hot channel factor The evaluation method described in Reference 1 for the calculation of the margin PCT was used to assess the impact of the reduction in maximum peaking factors on the L-12-076 Page 7 of 7 BVPS-2 LBLOCA analysis. Specifically, the reference transient was rerun at beginning-of-life conditions with the reduced peaking factors, and a PCT benefit was assigned based on plant-specific run results.
The evaluation method discussed in Reference 1 was used to determine the estimated effect of fuel pellet TCD and peaking factor burndown.
References
- 1.
L TR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
L-12-076 10 CFR 50.46 Report for the Evaluation of Fuel Pellet Thermal Conductivity Degradation Page 1 of 10 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN FOR THE BEAVER VALLEY POWER STATION, UNIT NO.1 (BVPS-1)
(Non-Discretionary Change)
Background
The Nuclear Regulatory Commission (NRC) approved 2004 Westinghouse Realistic Large Break LOCA [Ioss-of-coolant accident] Evaluation Model Using ASTRUM
[Automated Statistical Treatment ofUncertainty Method] (Reference 1) is based on the PAD 4.0 fuel performance code (Reference 2). PAD 4.0 was licensed without explicitly considering fuel pellet thermal conductivity degradation (TCD) with burnup. Explicit modeling of fuel pellet TCD in the fuel performance code leads to changes in the fuel rod design parameters beyond beginning-of-life, which are input to the large break LOCA (LBLOCA) analysis. The effects of explicitly modeling fuel pellet TCD on the BVPS-1 LBLOCA analysis have been considered due to a request from the NRC staff.
Westinghouse Electric Company LLC (Westinghouse) considers the modeling of fuel pellet TCD a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451 (Reference 3).
Fuel performance data that accounts for fuel pellet TCD (using an unlicensed model) was used as input to the BVPS-1 evaluation. The new PAD fuel performance data was generated with a representative model that includes explicit modeling of fuel pellet TCD.
Therefore the evaluations performed consider the fuel pellet TCD effects cited in NRC Information Notice 2011-21 (Reference 4).
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation as discussed in Reference 5 was performed to assess the peak cladding temperature (PCT) effect of TCD and peaking factor burndown with other considerations of burnup on the BVPS-1 LBLOCA analysis and concluded that the estimated PCT impact is 156°F for 10 CFR 50.46 reporting purposes. The peaking factor burndown included in the evaluation is provided in Table 4. FirstEnergy Nuclear Operating Company and its vendor, Westinghouse, utilize processes that ensure the LOCA analysis input values conservatively bound the as-operated plant values for those parameters.
L-12-076 Page 2 of 10 Table 4: Peaking Factors Versus Rod Burnup Burnup FQ Transient (1)
FQ Steady-State FAH (1),(2)
(MWD/MTU) 0 2.4 1.8 1.62 30,000 2.4 1.8 1.62 60,000 1.9 1.4 1.30 62,000 1.9 1.4 1.30 (1)
Includes uncertainties.
(2)
Hot assembly average power follows the same burndown, since it is a function of F.t.H.
Notes: FQ = heat flux hot channel factor F.t.H = enthalpy rise hot channel factor References
- 1.
WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
January 2005.
- 2.
WCAP-15063-P-A with Errata, Revision 1, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000.
- 3.
WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992.
- 4.
NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011.
- 5.
L TR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
L-12-076 Page 3 of 10 EVALUATION OF DESIGN INPUT CHANGES WITH RESPECT TO THE BEAVER VALLEY POWER STATION, UNIT NO.1 (BVPS-1) PLANT OPERATION
Background
To demonstrate compliance with the 10 CFR 50.46(b)(1) acceptance criterion concerning peak cladding temperature (PCT) when explicitly considering fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown in the BVPS-1 large break loss-of-coolant accident (LBLOCA) analysis of record (AOR), design input values were revised to more closely represent current plant operation. These input changes are not changes to the approved 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM). The updated inputs for BVPS-1 include:
reduction in transient FQ [heat flux hot channel factor] to COLR [Core Operating Limits Report] value reduction in steady-state FQ reduction in Fb.H [enthalpy rise hot channel factor] to COLR value reduction in hot assembly average power reduction in upper bound steam generator tube plugging increase in the assumed containment pressure boundary condition FirstEnergy Nuclear Operating Company and its vendor, Westinghouse Electric Company LLC, utilize processes that ensure the LOCAanalysis input values conservatively bound the as-operated plant values for those parameters.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation as discussed in Reference 1 was performed to estimate an overall PCT change due to the changes in design input parameters. The evaluation concluded that the estimated PCT impact of these analysis input changes is
-485°F for 10 CFR 50.46 reporting purposes.
Reference
- 1.
LTR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
L-12-076 Page 4 of 10 THE BEAVER VALLEY POWER STATION, UNIT NO.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE (PCT)
SUMMARY
Description Clad Temp Reference Notes (OF)
LICENSING BASIS Analysis of Record PCT 2161 1
A. Prior ECCS Model Assessments
- 1. PAD Data Evaluation 2
2 B. Planned Plant Modification Evaluations
- 1. PBOT/PMID Evaluation 0
2
- 2. Design Input Changes with Respect to Plant
-485 3,4 Operation C. 2012 ECCS Model Assessments
- 1. Evaluation of Pellet Thermal Conductivity 156 3,4 Degradation and Peaking Factor Burndown D. Other
- 1. None 0
LICENSING BASIS PCT + PCT ASSESSMENTS 1834
References:
. 1. WCAP-17052-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Beaver Valley Unit 1 Nuclear Plant Using the ASTRUM Methodology," March 2009.
(a)
(a)
- 2. L TR-LlS-1 0-427, "10 CFR 50.46 Report for the Large Break LOCA PBOT/PMID Evaluation and for the Large Break LOCA PAD Data Evaluation for Beaver Valley Unit 1 (DLW) ASTRUM Implementation," August 2010.
Notes:
- 3. LTR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
- 4. FENOC-12-37, "Transmittal of Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Including Design Input Changes,"
March 12,2012.
(a) These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown and design input changes.
L-12-076 Page 5 of 10 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN FOR THE BEAVER VALLEY POWER STATION, UNIT NO.2 (BVPS-2)
(Non-Discretionary Change)
Background
The Nuclear Regulatory Commission (NRC) approved 1996 Westinghouse Best Estimate Large Break LOCA [Ioss-of-coolant accident] Evaluation Model (Reference 1) is based on the PAD 3.4 fuel performance code (Reference 2). Upon NRC approval of PAD 4.0 (Reference 3), its usage was extended to the 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model, as reported to the NRC in Reference 4.
The BVPS-2 large break LOCA (LBLOCA) analysis utilized fuel rod design input from PAD 4.0. PAD 3.4 and PAD 4.0 were licensed without explicitly considering fuel pellet thermal conductivity degradation (TCD) with burnup. Explicit modeling of fuel pellet TCD in the fuel performance code leads to changes in the fuel rod design parameters beyond beginning-of-Iife, which are input to the LBLOCA analysis. The effects of explicitly modeling fuel pellet TCD on the BVPS-2 LBLOCA analysis (Reference 5) have been considered due to a request from the NRC staff. Westinghouse Electric Company LLC (Westinghouse) considers the modeling of fuel pellet TCD a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451 (Reference 6).
Fuel performance data that accounts for fuel pellet TCD (using an unlicensed model) was used as input to the BVPS-2 evaluation. The new PAD fuel performance data was generated with a representative model that includes explicit modeling of fuel pellet TCD.
Therefore the evaluations performed consider the fuel pellet TCD effects cited in NRC Information Notice 2011-21 (Reference 7).
Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect A quantitative evaluation as discussed in Reference 8 was performed to assess the PCT effect of TCD and peaking factor burndown with other considerations of burnup on the BVPS-2 LBLOCA analysis and concluded that the estimated PCT impact is O°F for Blowdown, 25°F for Reflood 1, and 10°F for Reflood 2 for 10 CFR 50.46 reporting purposes. The peaking factor burndown included in the evaluation is provided in Table 5. FirstEnergy Nuclear Operating Company and its vendor, Westinghouse, utilize processes that ensure the LOCA analysis input values conservatively bound the as-operated plant values for those parameters.
L-12-076 Page 6 of 10 a e. ea mg ac ors T bl 5 P k"
F t V ersus R dB 0
urnup Burnup FQ Transient (1)
FQ Steady-State FAH(l),(")
(MWD/MTU) 0 2.4 1.8 1.62 30,000 2.4 1.8 1.62 60,000 1.9 1.4 1.30 62,000 1.9 1.4 1.30 (1)
Includes uncertainties.
(2)
Hot assembly average power follows the same burndown, since it is a function of Fb.H.
Notes: FQ = heat flux hot channel factor Fb.H = enthalpy rise hot channel factor References
- 1.
WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.
- 2.
WCAP-1 0851-P-A, "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," August 1988.
- 3.
WCAP-15063-P-A with Errata, Revision 1, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000
- 4.
LTR-NRC-01-6, "10 CFR 50.46 Annual Notification and Reporting for 2000,"
March 13, 2001.
- 5.
WCAP-15900, Revision 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Beaver Valley Unit 2 Nuclear Plant," December 2002.
- 6.
WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992.
- 7.
NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011.
- 8.
L TR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
L-12-076 Page 7 of 10 EVALUATION OF DESIGN INPUT CHANGES WITH RESPECT TO THE BEAVER VALLEY POWER STATION, UNIT NO.2 (BVPS-2) PLANT OPERATION
Background
To demonstrate compliance with the 10 CFR 50.46(b)(1) acceptance criterion concerning peak cladding temperature (PCT) when explicitly considering fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown in the BVPS-2 large break loss-of-coolant accident (LBLOCA) analysis of record (AOR), design input values were revised to more closely represent current plant operation. These input changes are not changes to the approved 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model. The updated inputs for BVPS-2 include:
reduction in transient FQ to COLR value reduction in steady-state FQ reduction in FflH to COLR value reduction in hot assembly average power FirstEnergy Nuclear Operating Company and its vendor, Westinghouse Electric Company LLC, utilize processes that ensure the LOCA analysis input values conservatively bound the as-operated plant values for those parameters.
Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect A quantitative evaluation as discussed in Reference 1 was performed to estimate an overall PCT change due to the changes in design input parameters. The evaluation concluded that the estimated PCT impact of these analysis input changes is -100°F for Blowdown, -105°F for Reflood 1, and -190°F for Reflood 2 for 10 CFR 50.46 reporting purposes.
Reference
- 1.
L TR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
L-12-076 Page 8 of 10 THE BEAVER VALLEY POWER STATION, UNIT NO.2 LARGE BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE (PCT) COMPOSITE
SUMMARY
Description Clad Temp Reference Notes CF)
LICENSING BASIS Analysis of Record PCT 1976 PCT ASSESSMENTS (Delta PCT)
A. Prior ECCS Model Assessments
- 1. MONTECF Version 2.4 0
2
- 2. Revised Blowdown Heatup Uncertainty Distribution 5
3
- 3. HOTSPOT Fuel Relocation Error 40 6
B. Planned Plant Modification Evaluations
- 1. RAOC Evaluation 0
2
- 2. Accumulator Pressure Range Evaluation
-4 4
- 3. Containment Heat Sink Changes 0
5
- 4. Design Input Changes with Respect to Plant Operation
-190 7,8 C. 2012 ECCS Model Assessments
- 1. Evaluation of Pellet Thermal Conductivity 10 7,8 Degradation and Peaking Factor Burndown D. Other 0
References:
- 1. WCAP-15900, Revision 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Beaver Valley Unit 2 Nuclear Plant," December 2002.
- 2. FENOC-04-143, "BELOCA Evaluation For Changes in Containment Data," August 2004.
- 3. FENOC-05-48, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.
- 4. L TR-LlS-06-605, "Beaver Valley Units 1 & 2 (DLW/DMW) Accumulator Pressure Range Change," October 2006.
- 5. L TR-LlS-06-632, "Transmittal of Revised PCT Sheets for Beaver Valley Units 1 and 2 Best-Estimate Large Break LOCA and Appendix K Small Break LOCA," October 2006.
- 6. LTR-LlS-07-380, "10 CFR 50.46 Reporting Text for HOTSPOT Fuel Relocation Error and Revised PCT Rackup Sheets for Beaver Valley Units 1 and 2," June 2007.
(a)
(b)
(b)
L-12-076 Page 9 of 10 Notes:
- 7. LTR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
- 8. FENOC-12-37, "Transmittal of Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Including Design Input Changes,"
March 12, 2012.
(a) Accumulator Pressure Range Evaluation of 625 psia to 700 psia.
(b) These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown and design input changes.
L-12-076 Page 10 of 10 THE BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 LARGE BREAK LOSS-OF-COOLANT ACCIDENT REANALYSIS SCHEDULE The estimated impact on the Beaver Valley Power Station (BVPS), Unit Nos. 1 and 2 Large Break LOCA (LBLOCA) Evaluation Model (EM) from fuel pellet thermal conductivity degradation (TCD) results in a significant change in peak cladding temperature (PCT), as defined in 10 CFR 50.46(a)(3)(i). 10 CFR 50.46(a)(3)(ii) requires the licensee to provide a report within 30 days, including a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46. FirstEnergy Nuclear Operating Company (FENOC) has reviewed the information provided by Westinghouse Electric Company LLC (Westinghouse) and determined that the adjusted LBLOCA PCT values and the manner in which they were derived continue to comply with the requirements of 10 CFR 50.46. FENOC has evaluated the requirement for reanalysis specified in 10 CFR 50.46(a)(3)(ii) and hereby proposes a schedule for reanalysis.
On or before December 15,2016, FENOC will submit to the Nuclear Regulatory Commission (NRC) for review and approval LBLOCA analyses that apply NRC-approved methods that include the effects of fuel pellet TCD. The date for the analyses submittal is projected on the following milestones needed to perform a revised licensing basis LBLOCA analysis with an NRC-approved emergency core cooling system EM that explicitly accounts for TCD:
- 1) Submittal by Westinghouse, to the NRC for review and approval, of revised fuel performance and LBLOCA EM methodologies that include the effects of TCD.
- 2) Prior NRC approval of a fuel performance analysis methodology that includes the effects of TCD. The new NRC-approved methodology would replace the current licensing basis methodology for the Beaver Valley Power Station, Unit No.1 that is described in WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005 and for the Beaver Valley Power Station, Unit No.2 that is described in WCAP-12945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.
- 3) Prior NRC approval of a LBLOCA EM that includes the effects of TCD and accommodates the ongoing 10 CFR 50.46(c) rulemaking process. The new methodology would replace the current licensing basis methodology, WCAP-16009-P-A for Beaver Valley Power Station, Unit No.1 and WCAP-12945-P-A for Beaver Valley Power Station, Unit No.2.
. This information satisfies the reporting requirements of 10 CFR 50.46(a)(3)(ii).
L-12-076 Regulatory Commitment List Page 1 of 1 The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for the Beaver Valley Power Station, Unit Nos. 1 and 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not Regulatory Commitments. Please notify Mr. Phil H. Lashley, Supervisor - Fleet Licensing, at (330) 315-6808 of any questions regarding this document or associated Regulatory Commitments.
Regulatorv Commitment Due Date
- 1. Submit to the NRC for review and approval LBLOCA On or before analyses that apply NRC-approved methods that include December 15, 2016.
the effects of fuel pellet TCD.