LR-N13-0236, Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report

Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report

ML13289A301
Person / Time
Site:	Salem
Issue date:	10/11/2013
From:	Jamila Perry Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N13-0236
Download: ML13289A301 (19)

Text

PSEG Nuclear LLC P,O, Box 236; Hancocks Bridge, NJ 08038-0236 Nurim)' LLC OCT 11 2013 10 CFR 50.46 LR-N13-0236 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Salem Nuclear Generating Station Units 1 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

References:

1) Westinghouse letter LTR-LlS-13-1 04, "Salem Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2012," March 1, 2013.

2) PSEG letter LR-N12-0328, "Salem Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - 30 Day Report," October 19, 2012.

In accordance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nucl\3ar Power Reactors," paragraph (a)(3)(li), PSEG Nuclear 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 this reporting period of October 2012 to October 2013, there have been various issues identified via Reference 1; however, no significant changes to the' PCT rack-ups from 2012 are required. The previous Peak Cladding Temperature (PCT) report PSEG Nuclear filed with the NRC for Salem was dated October 19,2012 (Reference 2).

In Reference 2 for Salem Units 1 & 2, PSEG Nuclear committed to the NRC a Large Break Loss of Coolant (LBLOCA) analysis that applies NRC approved methods that include the effect of fuel Thermal Conductivity Degradation (TCD) and accommodates rulemaking associated with the proposed 10 CFR 50.46c.

OCT 1-1 2013 Document Control Desk LR-N13-0236 Page 2 There are no Commitments contained in this letter.

If you have any questions regarding this letter, please contact Chris Dahms at (856) 339-5456.

Sincerely, 1d-F.P~

John Perry _U Site Vice President - Salem Attachments (2): : Peak Cladding Temperature Rack-Up Sheets : Assessment Notes

Document Control Desk LR-N13-0236 Page 3 cc: Mr. W. Dean, Administrator, Region I, NRC Mr. J. Hughey, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. Cachaza, Salem Commitment Tracking Coordinator

LR-N13-0236 Attachment 1 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

LR-N13-0236 Attachment 1 Peak Cladding Temperature Rack~Up Sheets PLANT NAME: Salem Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 9/18/13 CURRENT OPERATING CYCLE: 23 ANALYSIS OF RECORD (AOR)

Evaluation Model: NOTRUMP.

Calculation: Westinghouse PSE-93-568, March 1993 Fuel: RFA 17 x 17 Limiting Fuel Type: RFA 17x17 Heat Flux Hot Channel Factor (Fa) 2.4=

Nuclear Enthalpy Rise Hot Channel Factor (Fi.\H) = 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)

MARGIN ALLOCATION A. P RIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated October 29, 1993 (See Note 1) llPCT = -13°F 10 CFR 50.46 report dated July 27, 1994 (See Note 2) llPCT = -16°F 10 CFR 50.46 report dated December 8, 1994 (See Note 3) llPCT = +109°F 10 CFR 50.46 report dated January 18, 1995 (See Note 4) llPCT = O°F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) llPCT = O°F 10 CFR 50.46 report dated August 2, 1996 (See Note 6) llPCT = -8°F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) llPCT = O°F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) llPCT = O°F 10 CFR 50.46 report dated April 27, 1999 (See Note 9) llPCT = O°F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) llPCT = +10°F 10 CFR 50.46 report dated September 21, 2000 (See Note llPCT = +2rF 11 )

10 CFR 50.46 report dated August 27,2001 (See Note 12) llPCT = O°F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) .L\PCT = O°F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) .L\PCT = O°F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) .L\PCT = +40°F 10 CFR 50.46 report dated July 28, 2005 (See Note 16) .L\PCT = O°F 10 CFR 50.46 report dated July 28, 2006 (See Note17) .L\PCT = O°F 10 CFR 50.46 report dated July 25, 2007 (See Note 18) .L\PCT = O°F 10 CFR 50,46 report dated July 22, 2008 (See Note 19) .L\PCT = O°F 10 CFR 50.46 report dated July 20, 2009 (See Note 20) .L\PCT = O°F 10 CFR 50.46 report dated July 20, 2010 (See Note 21) .L\PCT = O°F 10 CFR 50,46 report dated July 18, 2011 (See Note 22) .L\PCT = O°F 10 CFR 50.46 report dated July 16, 2012 (See Note 23) .L\PCT = O°F NET peT

LR-N13-0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets B. CURRENT LOCA MODEL ASSESSMENTS NOTRUMP-EM evaluation of pellet Thermal Conductivity bPCT = O°F Degradation (see Note 25)

General Code Maintenance (NOTRUMP) (See Note 26) bPCT = O°F Total PCT change from current assessments L LlPCT = O°F Cumulative PCT change from current assessments L !LlPCT! = O°F NET PCT

LR-N13-0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE: 9/18/13 CURRENT OPERATING CYCLE: 23 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 (Fo) = 2.4 Nuclear Enthalpy Rise Hot Channel Factor (Fi.1H) = 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)

MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated January 18, 1995 (See Note 4) LlPCT = +36°F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) ilPCT = O°F 10 CFR 50.46 report dated August 2, 1996 '(See Note 6} LlPCT = O°F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) LlPCT = +15°F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) LlPCT = O°F 10 CFR 50.46 report dated April 27, 1999 (See Note 9) LlPCT = O°F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) LlPCT = +12°F 10 CFR 50.46 report dated September 21, 2000 (See Note LlPCT = +9°F 11) 10 CFR 50.46 report dated August 27, 2001 (See Note 12) LlPCT = +6°F 10 CFR 50.46 report dated August 27,2002 (See Note 13) LlPCT = +20°F 10 CFR 50.46 report dated August 08,2003 (See Note 14) ilPCT = + 7°F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) LlPCT = +5°F 10 CFR 50.46 report dated July 28, 2005 (See Note 16) LlPCT = 0 of 10 CFR 50.46 report dated July 28, 2006 (See Note 17) ilPCT = -50 of 10 CFR 50.46 report dated July 25, 2007 (See Note 18) ilPCT = +4°F 10 CFR 50.46 report dated July 22, 2008 (See Note 19) LlPCT = O°F 10 CFR 50.46 report dated July 20, 2009 (See Note 20) ilPCT = O°F 10 CFR 50.46 report dated July 20, 2010 (See Note 21) ilPCT = O°F 10 CFR 50.46 report dated July 18, 2011 (See Note 22) LlPCT = O°F 10 CFR 50.46 report dated July 16, 2012 (See Note 23) LlPCT = O°F 10 CFR 50.46 report dated October 19,2012 (See Note 24) LlPCT = +87°F NET PCT

LR-N13~0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (BASH) (See Note 26) .6PCT = OaF Total PCT change from current assessments L LlPCT = OaF Cumulative PCT change from current assessments L I.6PCTI = OaF NET PCT

--~------~- ~-----~----~-~-------~-------~- ~---- ~~ ---.---.--~.-.~--~------- ..-.. ----

LR-N13-0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 2 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 9/18/13 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD (AOR)

Evaluation Model: NOTRUMP Calculation: Westinghouse (PSE-04-131), December 2004 Fuel: RFA 17 x 17 Limiting fuel Type:_RFA 17!<JL ___~____ _ _________ _

Heat Flux Hot Channel Factor (Fa) = 2.5 Nuclear Enthalpy Rise Hot Channel Factor (F"H) = 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)

MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated July 22, 2008 (See Note 19) i1PCT = oaF 10 CFR 50.46 report dated July 20, 2009 (See Note 20) i1PCT = oaF 10 CFR 50.46 report dated July 20, 2010 (See Note 21) . i1PCT = oaF 10 CFR 50.46 report dated July 18, 2011 (See Note 22) i1PCT = OaF 10 CFR 50.46 report dated July 16, 2012 (See Note 23) i1PCT = oaF NET PCT B. CURRENT LOCA MODEL ASSESSMENTS NOTRUMP-EM evaluation of pellet Thermal Conductivity i1PCT = oaF*

Degradation (see Note 25) .

General Code Maintenance (NOTRUMP) (See Note 26) i1PCT = oaF Total PCT change from current assessments L i1PCT :: oaF Cumulative PCT change from current assessments L I i1PCTI = oaF NET PCT

LR-N13-0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: Salem Unit 2 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE: 9/18/13 CURRENT OPERATING CYCLE: 20 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 (Fo)::: 2.4 Nuclear Enthalpy Rise Hot Channel Factor (FL',H)::: 1.65 Steam Generator Tube Plugging::: 25% (Reduced to 10% for RSG)

Limiting Break Size: Cd::: 004 Break Location: Cold Leg limiting Single Failure: loss of one train ECCS flow

, Reference Peak Cladding Temperature (PCT)

MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated January 18, 1995 (See Note 4) LlPCT  ::: +36°F 10 CFR 50.46 report dated December 7, 1995 (See Note 5) LlPCT  ::: OaF 10 CFR 50.46 report dated August 2, 1996 (See Note 6) LlPCT  ::: OaF 10 CFR 50046 report dated July 11, 1997 (See Note 7) LlPCT  ::: +15°F 10 CFR 50.46 report dated June 10, 1998 (See Note S) LlPCT  ::: OaF 10 CFR 50.46 report dated April 27, 1999 (See Note 9) LlPCT  ::: +24°F 10 CFR 50.46 report dated October is, 1999 (See Note 10) LlPCT  ::: *12°F 10 CFR 50.46 report dated September 21, 2000 (See Note ilPCT  ::: +9°F 11 )

10 CFR 50.46 report dated August 27, 2001 (See Note 12) LlPCT  ::: +6°F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) LlPCT  :::' +20°F 10 CFR 50.46 report dated August OS, 2003 (See Note 14) ilPCT  ::: +7°F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) LlPCT  ::: -45°F 10 CFR 50.46 report dated July 2S, 2005 (See Note 16) ilPCT  ::: OaF 10 CFR 50.46 report dated July 2S, 2006 (See Note 17) ilPCT  ::: OaF 10 CFR 50.46 report dated July 2S, 2007 (See Note is) LlPCT  ::: +4°F 10 CFR 50.46 report dated July 22, 200S (See Note 19) ilPCT  ::: -41°F 10 CFR 50.46 report dated July 20, 2009 (See Note 20) ilPCT  ::: OaF 10 CFR 50.46 report dated July 20, 2010 (See Note 21) LlPCT  ::: OaF 10 CFR 50.46 report dated July is, 2011 (See Note 22) LlPCT  ::: OaF 10 CFR 50.46 report dated July 16, 2012 (See Note 23) LlPCT  ::: OaF 10 CFR 50.46 report dated October 19, 2012 (See Note 24) LlPCT  ::: +S7°F NET PCT

LR-N 13-0236 Attachment 1 Peak Cladding Temperature Rack-Up Sheets B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (BASH) (See Note 26) llPCT = OaF Total PCT change from current assessments L llPCT = OaF Cumulative PCT change from current assessments L IllPCTI = OaF NET PCT

LR-N13-0236 Attachment 2 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

LR-N13-0236 Attachment 2 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°F), in support of the Fuel Upgrade / Margin Recovery Program. However, three PCT assessments were also included, resulting in a PCT benefit of -13°F. The first assessment entailed a +150°F penalty that resulted from explicitly modeling safety injection into the broken loop in the NOTRUMP model. The second assessment entailed a -150°F benefit that resulted from the implementation of an improved condensation model. The third assessment entailed a -13°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°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 +1 09°F. The first assessment entailed a +85°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 -6°F PCT benefit that was a result of error corrections made to the boili(lg 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°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°F. The first assessment entailed a +94°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°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°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 -6°F that resulted from corrections made to the LUCIFER code.

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.

LR-N13-0236 Attachment 2 Assessment Notes

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°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

+1 OaF 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°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°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 SBLOCA evaluation model, a generic PCT penalty of +1 OaF was assessed due to the impact of fully enriched annular pellets. 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.

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°F PCT benefit for LBLOCA and no change (O°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°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 +1 OaF was

LR-N 13-0236 Attachment 2 Assessment Notes 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°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°F PCT increase. This increase consisted of a +14°F PCT assessment due to an error in the feedwater line volume calculation and a +13°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 +9°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 +6°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°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 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 +15°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 + 7°F.

LR~N13~0236 Attachment 2 Assessment Notes

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 +5°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 +5°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 ~45°F.

16. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 28, 2005, reported a OaF increase 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 OaF 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 OaF increase in the PCT of the SBLOCA evaluation model due to the SBLOCA model :a.ssessment. The model a$sessment 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 OaF 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 50°F 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 OaF increase in the PCT of the SBLOCA evaluation model due to the SBLOCA model assessment. The model assessment 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 +4°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 (O°F), a rebaseline calculation to determine the limiting LOCBART calculated PCT (~8°F), LOCBART code correction for pellet volumetric heat generation rate

(+12°F), LOCBART code option to convert user"specified zirconium-oxide thickness to

LR~N13~0236 Attachment 2 Assessment Notes equivalent cladding reacted (oaF), errors in the IMP code vessel nozzle collections (oaF), and general code maintenance (BASH).

19. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 22,2008, reported a oaF 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 oaF 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 ~41 OF benefit due to implementation of the replacement steam generators and change in steam generator tube plugging limits from 25% to 10% (-4rF), removal of a rebaseline calculation not applicable to Salem Unit 2 with the new steam generators (+8°F); BASH pellet volumetric heat generation rate correction (O°F); LOCBART pellet volumetric heat generation rate correction (_2°F), and errors in the reactor vessel lower plenum surface area calculation (OOF), and general code (BASH) maintenance (O°F).

20. Prior LOCA Model Assessment The 10CFR50.46 reportdated July 20,2009, reported a OaF increase in the PCT for the Salem Unit 1 and Salem Unit 2 small and large break LOCA mod~1 assessments. Discrepancies were discovered in the use of metal masses from drawings. The updated reactor ve'ssel 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 c1osely-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 OaF for 10 CFR 50.46 reporting purposes. General code maintenance (NOTRUMP for SBLOCA and BASH for LBLOCA) resulted in a OaF 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 OaF 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 10CFR50.46 LOCA models or methods for this reporting period for Salem Unit 1 and Salem Unit 2.

22. Prior LOCA Model Assessment The 1 OCFR50.46 report dated July 18, 2011, reported a OaF increase in the PCT for the Salem Unit 1 and Salem Unit 2 small and large break LOCA model assessments. Discrepancies were discovered and are summarized. Historically, the overall vessel average temperature

LR-N13-0236 Attachment 2 Assessment Notes 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 OaF. 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 kwlft 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-13451. 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 OaF.

23. Prior LOCA Model Assessment The 10CFR50.46 report dated July 16,2012, reported a OaF increase In the PCT for the Salem Unit 1 and Salem Unit 2 small and large break LOCA model assessments. Discrepancies were discovered and are summarized. Two errors were discovered in the calculation of the radiation heat transfer coefficient in the SBLOCTA computer code. First, existing diagnostics did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated. These calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1 degree. Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine. These errors have been corrected in the SBLOCTA code and represent a closely related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of OaF for existing Small Break LOCA analysis results. An error was discovered in the SBLOCTA code that allowed the fuel rod time step to exceed the specified maximum allowable time step. The time step logic has been corrected in the SBLOCTA code.

This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of OaF for existing Small Break LOCA analysis results.

24. Prior LOCA Model Assessment The 10CFR50.46 report dated October 19, 2012, reported a rebaseline +8rF increase in the PCT for the Salem Unit 1 and Salem Unit 2 large break LOCA model assessments.

Evaluations have been completed to estimate the effect of fuel pellet thermal conductivity degradation (TCD) on peak cladding temperature (PCT) for analyses using the 1981 Westinghouse Large-Break Loss of-Coolant Accident Evaluation Model with BASH (BASH-EM) with the LOCBART Transient Extension Method. Note the impact on PCT due to TCD was OaF. These evaluations utilized fuel rod performance input from a version of the PAD code that accounts for pellet TCD and considered the beneficial effects of assembly power and peaking factor burndown resulting from the depletion of fissionable isotopes. This change represents a

LR-N13-0236 Attachment 2 Assessment Notes Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. The estimated effect was determined on a plant-specific basis. The peaking factor burndown used in the evaluation is provided in LTR-LlS-12-512; it is conservative for the current cycle and will be validated as part of the reload design process. PSEG Nuclear and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the corresponding LOCA analysis input parameters conservatively bound the as-operated plant values. The utilization of the LOCBART Transient Extension Method led to an ~stimated rebaseline PCT impact of +S7"F for existing Large Break LOCA analysis results.

25. NOTRUMP-EM EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION An evaluation has been completed to estimate the effect of fuel pellet thermal conductivity degradation (TCD) on peak cladding temperature (PCT) for plants in the Unites States with analyses using the 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (NOTRUMP-EM). This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. Based on phenomena and physics of the SBLOCA transient, in combination with limited sensitivity calculations, it is concluded that TCD has a negligible effect on the limiting cladding temperature transient, leading to an estimated PCT impact of OaF.

26. 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-13451. The nature of these changes leads to an estimated PCT impact of OaF.