LR-N11-0211, Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report

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Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report
ML11207A067
Person / Time
Site: Salem  PSEG icon.png
Issue date: 07/18/2011
From: Fricker C
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N11-0211
Download: ML11207A067 (17)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PSEG NuclearLLC JUL 1 8 Z011 10 CFR 50.46 LR-N 11-0211 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Salem Nuclear Generating Station Units I and 2 Facility Operating License Nos. DPR-70 and 75 NRC Docket Nos. 50-272 and 50-311

Subject:

SALEM LOSS OF COOLANT ACCIDENT PEAK CLADDING TEMPERATURE MARGIN TRACKING. - ANNUAL REPORT

REFERENCE:

1) Westinghouse Letter LTR-LIS-11-57, "Salem Units l and 2 10 CFR 50.46 Annual Notification and Reporting for 2010," January 27, 2011.
2) PSEG Letter LR-N10-0250, "Salem Nuclear Generating Station Units 1 and 2 Facility Operating License DPR-70 and 75 NRC Docket Nos. 50-272 and 50-311, Salem Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report,"

July 20, 2010.

In accordance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," paragraph (a)(3)(ii), PSEG Nuclear, LLC (PSEG) is required to submit an annual report of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for Salem Units 1 and 2.

For the reporting period of July 2010 to June 2011, there have been various issues identified via Reference 1; however, no changes to the PCT rack-ups from 2010 are required. The PCT rack-ups are being sent for completeness only. The previous Peak Cladding Temperature (PCT) report PSEG Nuclear filed with the NRC for Salem was dated July 20,2010 (Reference 2). , "Peak Cladding Temperature Rack-Up Sheets," provides updated information regarding the PCT for the limiting small break and large break loss-of-coolant accident (LOCA) evaluations for Salem Units I and 2. , "Assessment Notes," contains a detailed description for each of these previous changes or errors reported.

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JUL 1 8 2011 Document Control Desk LR-N1 1-0211 There are no commitments made in this letter. Ifyou have any questions regarding this letter, please contact Enrique Villar at (856) 339-5456.

Sincerely C,,arll

  • .ricker Site Vice President - Salem cc: Mr. W. Dean, Administrator - Region I U. S. Nuclear Regulatory Commission USNRC Senior Resident Inspector - Salem (X24)

Mr. P. Mulligan, Manager IV Commitment Coordinator - Salem Commitment Coordinator - Corporate

Attachment 1 LR-Nl1-0211 Peak Cladding Temperature Rack-Up Sheets SALEM UNITS 1 AND 2 Docket Nos. 50-272 and 50-311 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments

Attachment 1 LR-Nl1-0211 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 4/27/11 CURRENT OPERATING CYCLE: 21 ANALYSIS OF RECORD (AOR)

Evaluation Model: NOTRUMP Calculation: Westinghouse PSE-93-568, March 1993 Fuel: RFA17x17 Limiting Fuel Type: RFA 17x17 Heat Flux Hot Channel Factor (FQ) = 2.4 Nuclear Enthalpy Rise Hot Channel Factor (FAH) = 1.65 Steam Generator Tube Plugging = 10%

Limiting Break Size: 2 inches Break Location: Cold Leg Limiting Single Failure: loss of one train of ECCS flow Reference Peak Cladding Temperature (PCT) PCT = 1580°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated October 29, 1993 (See Note 1) APCT = -13 0 F 10 CFR 50.46 report dated July 27, 1994 (See Note 2) APCT = -161F 10 CFR 50.46 report dated December 8, 1994 (See Note 3) APCT = +109 0 F 10 CFR 50.46 report dated January 18, 1995 (See Note 4) APCT = 0°F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) APCT = 0°F 10 CFR 50.46 report dated August 2, 1996 (See Note 6) APCT = -80 F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) APCT = 00 F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) APCT = 0°F 10 CFR 50.46 report dated April 27, 1999 (See Note 9) APCT = 0°F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) APCT = +100 F 10 CFR 50.46 report dated September 21, 2000 (See Note APCT = +27 0 F 11) 10 CFR 50.46 report dated August 27, 2001 (See Note 12) APCT = 0°F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = 00 F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = 0°F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) APCT = +40°F 10 CFR 50.46 report dated July 28, 2005 (See Note 16) APCT = 0°F 10CFR 50.46 report dated July 28, 2006 (See Notel7) APCT = 0°F 10CFR 50.46 report dated July 25, 2007 (See Note 18) APCT = 0°F 10CFR 50.46 report dated July 22, 2008 (See Note 19) APCT = 0°F 10CFR 50.46 report dated July 20, 2009 (See Note 20) APCT = 0°F 10CFR 50.46 report dated July 20, 2010 (See Note 21) APCT = 0°F 1

Attachment I LR-Nl1-0211 Peak Cladding Temperature Rack-Up Sheets NET PCT PCT = 1729°F B. CURRENT LOCA MODEL ASSESSMENTS Treatment of Vessel Average Temperature Uncertainty (See APCT = 0°F Note 22)

Pellet Crack and Dish Volume Calculation (See Note 23) APCT = 0°F General Code Maintenance (NOTRUMP) (See Note 24) APCT = 0°F Total PCT change from current assessments Z APCT = 0°F Cumulative PCT change from current assessments 7 1APCTI = 0°F NET PCT PCT = 1729 0 F 2

Attachment 1 LR-Nl1-0211 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE: 4/27/11 CURRENT OPERATING CYCLE: 21 ANALYSIS OF RECORD (AOR)

Evaluation Model: BASH Calculation: Westinghouse 93-PSE-G-0080, September 1993 Fuel: RFA17x17 Limiting Fuel Type: RFA 17x17 Heat Flux Hot Channel Factor (FQ) = 2.4 Nuclear Enthalpy Rise Hot Channel Factor (FAH) = 1.65 Steam Generator Tube Plugging = 10%

Limiting Break Size: Cd = 0.4 Break Location: Cold leg Limiting Single Failure: Loss of one train of ECCS flow Reference Peak Cladding Temperature (PCT) PCT = 1978°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated January 18, 1995 (See Note 4) APCT = +36 0 F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) APCT = 0°F 10 CFR 50.46 report dated August 2, 1996 (See Note 6) APCT = 0°F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) APCT = +15 0 F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) APCT = 0°F 10 CFR 50.46 report dated April 27, 1999 (See Note 9) APCT = 0°F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) APCT = +12 0 F 10 CFR 50.46 report dated September 21, 2000 (See Note APCT = +90 F 11) 10 CFR 50.46 report dated August 27, 2001 (See Note 12) APCT = +60 F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = +201F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = +70 F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) APCT = +50 F 10 CFR 50.46 report dated July 28, 2005 (See Note 16) APCT = 0 OF 10 CFR 50.46 report dated July 28, 2006 (See Note 17) APCT = -50 IF 10 CFR50.46 report dated July 25, 2007 (See Note 18) APCT = +40 F 10CFR 50.46 report dated July 22, 2008 (See Note 19) APCT = 00 F 10CFR 50.46 report dated July 20, 2009 (See Note 20) APCT = 0°F 10CFR 50.46 report dated July 20, 2010 (See Note 21) APCT = 0°F NET PCT PCT = 2042°F 3

Attachment 1 LR-Nl1-0211 Peak Cladding Temperature Rack-Up Sheets B. CURRENT LOCA MODEL ASSESSMENTS Treatment of Vessel Average Temperature Uncertainty (See APCT = 0°F Note 22)

Pellet Crack and Dish Volume Calculation (See Note 23) APCT = 0°F General Code Maintenance (BASH) (See Note 24) APCT = 0°F Total PCT change from current assessments 7- APCT = 0°F Cumulative PCT change from current assessments 7-I APCT = 0°F NET PCT PCT = 2042°F 4

Attachment I LR-N11-0211 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 2 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 4/27/11 CURRENT OPERATING CYCLE: 19 ANALYSIS OF RECORD (AOR)

Evaluation Model: NOTRUMP Calculation: Westinghouse (PSE-04-131), December 2004 Fuel: RFA 17 x 17 Limiting Fuel Type: RFA 17x17 Heat Flux Hot Channel Factor (FQ) = 2.5 Nuclear Enthalpy Rise Hot Channel Factor (FAH) = 1.65 Steam Generator Tube Plugging = 10%

Limiting Break Size: 3 inches Break Location: Cold Leg Single Failure: loss of one train ECCS flow Reference Peak Cladding Temperature (PCT) PCT = 9871F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10CFR 50.46 report dated July 22, 2008 (See Note 19) APCT = 0°F 10CFR 50.46 report dated July 20, 2009 (See Note 20) APCT = 0°F 10CFR 50.46 report dated July 20, 2010 (See Note 21) APCT = 0°F NET PCT PCT = 9870 F B. CURRENT LOCA MODEL ASSESSMENTS Treatment of Vessel Average Temperature Uncertainty (See APCT = 0°F Note 22)

Pellet Crack and Dish Volume Calculation (See Note 23) APCT = 0°F General Code Maintenance (NOTRUMP) (See Note 24) APCT = 00 F Total PCT change from current assessments 7- APCT = 00 F Cumulative PCT change from current assessments 7-I APCT = 0°F NET PCT PCT = 987 0F 5

Attachment 1 LR-Nll-0211 Peak Claddi ng Temperature Rack-Up Sheets PLANT NAME: Salem Unit 2 ECCS EVALUATION MODEL: Larae Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE: 4/27/11 CURRENT OPERATING CYCLE: 19 ANALYSIS OF RECORD (AOR)

Evaluation Model: BASH Calculation: Westinghouse 93-PSE-G-0080, September 1993 Fuel: RFA 17 x 17 Limiting Fuel Type: RFA 17x17 Heat Flux Hot Channel Factor (FQ) = 2.4 Nuclear Enthalpy Rise Hot Channel Factor (FAH) = 1.65 Steam Generator Tube Plugging = 25%

Limiting Break Size: Cd = 0.4 Break Location: Cold Leg Limiting Single Failure: loss of one train ECCS flow Reference Peak Cladding Temperature (PCT) PCT = 19781F MARGIN ALLOCATION A. -PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated January 18, 1995 (See Note 4) APCT = +36 0 F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) APCT = 0°F 10 CFR 50.46 report dated August 2, 1996 (See Note 6) APCT = 0°F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) APCT = +1 50 F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) APCT = 0°F 10 CFR 50.46 report dated April 27, 1999 (See Note 9) APCT = +24 0 F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) APCT = -12'F 10 CFR 50.46 report dated September 21, 2000 (See Note APCT = +90 F 11) 10 CFR 50.46 report dated August 27, 2001 (See Note 12) APCT = +60 F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = +201F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = +7°F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) APCT = -45 0 F 10 CFR 50.46 report dated July 28, 2005 (See Note 16) APCT = 0°F 10 CFR 50.46 report dated July 28, 2006 (See Note 17) APCT = 0°F 10 CFR 50.46 report dated July 28, 2007 (See Note 18) APCT = +40 F 10CFR 50.46 report dated July 22, 2008 (See Note 19) APCT = -41 OF 10CFR 50.46 report dated July 20, 2009 (See Note 20) APCT = 0°F 10CFR 50.46 report dated July 20, 2010 (See Note 21) APCT = 0°F NET PCT PCT = 2001°F 6

Attachment 1 LR-Nll-0211 Peak Cladding Temperature Rack-Up Sheets B. CURRENT LOCA MODEL ASSESSMENTS Treatment of Vessel Average Temperature Uncertainty (See APCT = 0°F Note 22)

Pellet Crack and Dish Volume Calculation (See Note 23) APCT = 0°F General Code Maintenance (BASH) (See Note 24) APCT = 0°F Total PCT change from current assessments Y_APCT = 0°F Cumulative PCT change from current assessments TZ_

APCT = 0°F NET PCT PCT = 2001OF 7

Attachment 2 LR-Nl1-0211 Assessment Notes SALEM UNITS 1 AND 2 Docket Nos. 50-272 and 50-311 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System

.Evaluation Model Changes and Errors Assessments

Attachment 2 LR-Nl1-0211 Assessment Notes

1. Prior Loss-of-Coolant Accident (LOCA) Model Assessment The 10 CFR 50.46 report dated October 29, 1993, implemented the current Analysis of Record for the SBLOCA evaluation model (PCT = 1580 0 F), in support of the Fuel Upgrade / Margin Recovery Program. However, three PCT assessments were also included, resulting in a PCT benefit of -1 30 F. The first assessment entailed a +1 50°F penalty that resulted from explicitly modeling safety injection into the broken loop in the NOTRUMP model. The second assessment entailed a -1 50°F benefit that resulted from the implementation of an improved condensation model. The third assessment entailed a -1 30 F benefit that resulted from the correction of drift flux flow regime errors.
2. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 27, 1994, reported an assessment to the SBLOCA model, which resulted in a -16 0 F PCT benefit. This PCT benefit was a result of corrections made to the reactor vessel and steam generator geometric and mass calculations in the VESCAL subroutine in the LUCIFER code.
3. Prior LOCA Model Assessment The 10 CFR 50.46 report dated December 8, 1994, reported evaluations for the SBLOCA model due to three errors, for a penalty of +109 0 F. The first assessment entailed a +85 0 F PCT penalty that was a result of correcting nodalization and overall fluid conservation errors in the SBLOCTA code and implementing a revised transient fuel rod internal pressure model. The second assessment entailed a -60 F PCT benefit that was a result of error corrections made to the boiling heat transfer regime correlations in NOTRUMP. The third assessment entailed a

+30°F PCT penalty as a result of errors affecting the steam line isolation logic in the SBLOCA evaluation model.

4. Prior LOCA Model Assessment The 10 CFR 50.46 report dated January 18, 1995, reported no changes in the SBLOCA model, which caused the PCT to remain unchanged. The current Analysis of Record for the LBLOCA evaluation model (PCT = 1978 0 F) was implemented in support of the Fuel Upgrade / Margin Recovery Program. However, three PCT assessments were also included, resulting in a PCT penalty of +36 0 F. The first assessment entailed a +94 0 F PCT penalty that resulted from the absence of Intermediate Flow Mixers (IFMs) in the core. The second assessment was a PCT benefit of -52 0 F that resulted from four changes to the LOCBART code; including modifications made to convert the LOCBART code from a Cray to a Unix platform, corrections made to the rod heat-up code, the addition of a new model used to determine zircaloy cladding burst behavior above 1742 0 F, and the implementation of a revised burst strain limit model for the rod heat-up codes. The third assessment entailed a PCT benefit of -60 F that resulted from corrections made to the LUCIFER code.

1

Attachment 2 LR-Nl1-0211 Assessment Notes

5. Prior LOCA Model Assessment The 10 CFR 50.46 report dated December 7, 1995, reported no changes in the SBLOCA and LBLOCA models for both Salem Units 1 and 2, which caused the PCTs to remain unchanged.
6. Prior LOCA Model Assessment The 10 CFR 50.46 report dated August 2, 1996, reported no changes in the LBLOCA model, which caused the PCT to remain unchanged. The SBLOCA model was assessed an -8 0 F PCT benefit as a result of three assessments. The first assessment was a +20°F PCT penalty due to an error in the specific enthalpy equation in NOTRUMP. The second assessment was a

+10°F PCT penalty due to an error in the Fuel Rod Initialization algorithm of the SBLOCTA code, as well as several changes in the fuel rod creep and strain model. The third assessment was a -38 0 F PCT benefit as a result of an error in the relative loop seal elevation of the crossover leg.

7. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 11, 1997, reported no changes in the SBLOCA model, which caused the PCT to remain unchanged. The LBLOCA model was assessed a +15 0 F PCT penalty as a result of translating the fluid conditions used for subchannel analysis of the fuel rods from one computer code (SATAN) to another computer code (LOCTA).
8. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 10, 1998, reported no changes in the SBLOCA and LBLOCA models for both Salem Units 1 and 2, which caused the PCTs to remain unchanged.
9. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 27, 1999, reported no changes in the Salem Unit 1 SBLOCA and LBLOCA models, which caused the PCTs to remain unchanged. However, unit-and cycle-specific PCT assessments were applied to Salem Unit 2. For the Salem Unit 2 LBLOCA evaluation model, a partial re-analysis was performed that incorporated the effects of Intermediate Flow Mixers (IFMs), features of the Robust Fuel Assembly (RFA), and other model updates. The cumulative impact of these PCT changes resulted in an increase in the Salem Unit 2 LBLOCA PCT of +24°F.

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Attachment 2 LR-N 11-0211 Assessment Notes

10. Prior LOCA Model Assessment The 10 CFR 50.46 report dated October 18, 1999, reported evaluations for the SBLOCA and LBLOCA models for both Salem Units due to three errors. The first error resulted from the use of incorrect geometric data related to the accumulator lines and the pressurizer surge line. The second error was discovered in the length-averaging logic for heat transfer coefficient calculations in the LOCBART code. The third error was found in the Baker-Just metal-water reaction calculation in the LOCBART code. These errors were assessed together on a plant-specific basis and resulted in a -12 0 F PCT benefit for LBLOCA and no change (0°F) in the PCT for SBLOCA for both Salem Units. Thus, the Salem Unit 2 SBLOCA PCT remained

.unchanged, while the Salem Unit 2 LBLOCA PCT decreased by -12 0 F. In addition to the assessment above, further unit- and cycle-specific PCT assessments were applied to Salem Unit 1. For the Salem Unit 1 SBLOCA evaluation model, a generic PCT penalty of +10°F was assessed due to the impact of fully enriched annular pellets. For the Salem Unit 1 LBLOCA evaluation model, a partial re-analysis was performed that incorporated the effects of the Robust Fuel Assembly (RFA) features, Intermediate Flow Mixers (IFMs), and other model updates. In addition, a generic transition core PCT penalty was assessed to account for the effects of mixed fuel types (RFA and V5H) in the core. The cumulative impact of all of these PCT changes resulted in an increase in the Salem Unit 1 LBLOCA PCT of +12 0 F.

11. Prior LOCA Model Assessment The 10 CFR 50.46 report dated September 21, 2000, reported evaluations for SBLOCA model changes, which resulted in a +27 0 F PCT increase. This increase consisted of a +140 F PCT assessment due to an error in the feedwater line volume calculation and a +130 F PCT assessment due to the discovery of several closely related errors dealing with mixture level tracking and region depletion errors in NOTRUMP. The LBLOCA model was assessed a +90 F PCT penalty as a result of an error in the LOCBART vapor film flow regime heat transfer correlation.
12. Prior LOCA Model Assessment The 10 CFR 50.46 report dated August 27, 2001, reported no changes in the SBLOCA model, which caused the PCT to remain unchanged. The LBLOCA model was assessed a +60 F PCT penalty as a result of using non-conservative cladding surface emissivity values in LOCBART.
13. Prior LOCA Model Assessment The 10 CFR 50.46 report dated August 27, 2002, reported no changes in the SBLOCA model, which caused the PCT to remain unchanged. The LBLOCA model was assessed a +20 0 F PCT penalty as a result of using a non-conservative assumption for accumulator water temperature.
14. Prior LOCA Model Assessment The 10 CFR 50.46 report dated August 8, 2003, reported no changes in the SBLOCA model, which caused the PCT to remain unchanged. A partial re-analysis was performed for the LBLOCA transient using the latest BASH-EM code version that incorporated the "LOCBART transient extension method," that ensured adequate termination of the fuel rod cladding temperature and oxidation transients predicted by LOCBART. This partial re-analysis allowed 3

Attachment 2 LR-N11-021.1 Assessment Notes several prior PCT "generic evaluation" assessments (Accumulator Line / Pressurizer Surge Line Data Error, LOCBART Spacer Grid Single Phase Heat Transfer Error, LOCBART Zirc-Water Oxidation Error, LOCBART Vapor Film Flow Regime Heat Transfer Error, LOCBART Cladding Emissivity Error, Changes due to RFA Fuel Features, and Non-Conservative Accumulator Water Temperature Evaluation) to be replaced with a plant-specific analytical estimation. In addition, a +150 F PCT penalty was assessed to the LBLOCA model that resulted from corrections to the LOCBART ZIRLO Cladding Specific Heat Model. As a result of this penalty and the partial re-analysis, the LBLOCA PCT increased by +70 F.

15. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 29, 2004, reported a +40)F increase in the PCT of the SBLOCA evaluation model as a result of inconsistency corrections made to the NOTRUMP Bubble Rise and Drift Flux models and burst and blockage and time in life. The Salem Unit 1 LBLOCA model was assessed a +50 F PCT penalty as a result of the correction of discrepancies in the LOCBART Fluid Property Logic. The Salem Unit 2 LBLOCA model was also assessed this +50 F penalty, in addition to the removal of a +50°F Transition Core Penalty that resulted from operating with a mixed core of V5H and RFA fuel types, for a decrease in the PCT of -451F.
16. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 28, 2005, reported a 0 0 Flincrease in the PCT of the SBLOCA evaluation model due to the SBLOCA model assessment. The model assessment for SBLOCA was performed for reactor coolant pump reference conditions and general code maintenance (NOTRUMP). The report also reported a 0°F increase in the PCT of the LBLOCA evaluation model due to the LBLOCA model assessment. The model assessment for LBLOCA was performed for reactor coolant pump reference conditions, LOCBART fluid property logic, steam generator inlet/outlet plenum flow areas, initial containment relative humidity assumption and general code maintenance (BASH).
17. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 28, 2006, reported a 00 F increase in the PCT of the SBLOCA evaluation model due to the SBLOCA model assessment. The model assessment for SBLOCA included replacing previously transmitted pressurizer fluid volumes with nominal cold values, correcting for an error in the lower guide tube assembly weight, corrected modeling of the spilling flows in the RWST draindown calculation and general code maintenance (NOTRUMP). The report also reported a 0°F increase in the PCT of the LBLOCA evaluation model due to the LBLOCA model assessment. The model assessment for LBLOCA included replacing previously transmitted pressurizer fluid volumes with nominal cold values, correcting for an error in the lower guide tube assembly weight, and general code maintenance (BASH). Additionally, the 50OF transition core PCT penalty applied to Salem Unit 1 LBLOCA was removed.
18. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 25, 2007, reported a 00 F increase in the PCT of the SBLOCA evaluation model due to the SBLOCA model assessment. The model assessment 4

Attachment 2 LR-N11-0211 Assessment Notes for SBLOCA included the impact of the SBLOCA break size spectrum, errors in the IMP code vessel nozzle collections, and general code maintenance (NOTRUMP). The report also reported a +40 F increase in the PCT of the LBLOCA evaluation model due to the LBLOCA model assessment. The model assessment for LBLOCA included BASH minimum and maximum time step sizes (0°F), a rebaseline calculation to determine the limiting LOCBART calculated PCT (-8°F), LOCBART code correction for pellet volumetric heat generation rate

(+12 0 F), LOCBART code option to convert user-specified zirconium-oxide thickness to equivalent cladding reacted (0°F), errors in the IMP code vessel nozzle collections (0°F), and general code maintenance (BASH).

19. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 22, 2008, reported a 0°F increase in the PCT of the SBLOCA evaluation model due to the SBLOCA model assessment. The model assessment for SBLOCA included the impact of errors in the reactor vessel lower plenum surface area calculation and general code maintenance (NOTRUMP). A new Small Break LOCA Analysis of Record was implemented for Salem Unit 2 with implementation of the replacement steam generators in Salem 2 Cycle 17. The report also provided a 0°F increase in PCT of the LBLOCA evaluation model for Salem Unit 1 due to the LBLOCA model assessment. The Salem Unit 1 model assessment for LBLOCA included BASH pellet volumetric heat generation rate, error in reactor vessel lower plenum surface area calculations, and general code maintenance (BASH). The Salem Unit 2 model assessment for Large Break LOCA included a net -41OF benefit due to implementation of the replacement steam generators and change in steam generator tube plugging limits from 25% to 10% (-470 F), removal of a rebaseline calculation not applicable to Salem Unit 2 with the new steam generators (+80 F); BASH pellet volumetric heat generation rate correction (00 F); LOCBART pellet volumetric heat generation rate correction (-20 F), and errors in the reactor vessel lower plenum surface area calculation (00 F), and general code (BASH) maintenance (0°F).
20. Prior LOCA Model Assessment The 10CFR50.46 report dated July 20, 2009, reported a 0°F increase in the PCT for the Salem Unit 1 and Salem Unit 2 small and large break LOCA model assessments. Discrepancies were discovered in the use of metal masses from drawings. The updated reactor vessel metal masses and fluid volumes have been evaluated for impact on current licensing basis analysis results and will be incorporated on a forward-fit basis. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

The differences in the reactor vessel metal mass and fluid volume are relatively minor and produce a negligible effect on large and small break LOCA analysis results, leading to a PCT impact of 0°F for 10 CFR 50.46 reporting purposes. General code maintenance (NOTRUMP for SBLOCA and BASH for LBLOCA) resulted in a 0°F PCT increase for Salem Unit 1 and Salem Unit 2.

21. Prior LOCA Model Assessment The 10CFR50.46 report dated July 20, 2010, reported a 0°F increase in the PCT for the Salem Unit 1 and Salem Unit 2 small and large break LOCA model assessments. No discrepancies were identified in the 1 OCFR50.46 LOCA models or methods for this reporting period for Salem Unit 1 and Salem Unit 2.

5

Attachment 2 LR-Nl1-0211 Assessment Notes

22. Treatment of Vessel Average Temperature Uncertainty Historically, the overall vessel average temperature uncertainty calculated by Westinghouse considered only "-" instrument uncertainties, corresponding to the indicated temperature being lower than the actual temperature. The uncertainty was then applied as a "+/-" uncertainty in some LOCA analyses, rather than using specific "+" and "-"uncertainties. This discrepancy has been evaluated for impact on existing Large and Small Break LOCA analysis results, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. The issue was judged to have a negligible impact on existing Large and Small Break LOCA analysis results, leading to an estimated PCT impact of 0°F.
23. Pellet Crack and Dish Volume Calculation Error Two issues were identified related to the normalized pellet crack and dish volumes utilized in the LOCA peak clad temperature (PCT) analyses. These issues were: 1) the incorrect tables of normalized volume versus linear heat generation rate were being used (the table for clad outer diameters of <0.4 inches were using tables for clad outer diameters >0.4 inches and vice versa), and 2) the normalized volume at 18 kw/ft was incorrectly programmed in one of the tables as 1.58 instead of 1.59. This discrepancy has been evaluated for impact on existing Large and Small Break LOCA analysis results, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451. These issues were judged to have a negligible impact on existing Large and Small Break LOCA analysis results, leading to an estimated PCT impact of 0°F.
24. General Code Maintenance (BASH/NOTRUMP)

Various changes have been made to enhance usability and help preclude errors in analyses.

This includes items such as modifying input and 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-1 3451. The nature of these changes leads to an estimated PCT impact of 00 F.

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