LR-N07-0181, Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.
| ML072150246 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/25/2007 |
| From: | Braun R C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N07-0181 | |
| Download: ML072150246 (18) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236
© PSEG Nuclear LLC JUL 2 6W LR-N07-0181 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 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 10 CFR 50.46
Subject:
Reference Annual Report of the Emergency Core Cooling System Evaluation Model Changes end Errors required by 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors." PSEG Letter LR-N06-0331, "Salem Nuclear Generating Station Unit Nos. 1 and 2 Facility Operating License DPR-70 and DPR-75 Docket Nos. 50-272 and 50-311, Annual Report for the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46," dated July 28, 2006 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 submitting the annual report of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for Salem Units 1 and 2.The last Peak Cladding Temperature (PCT) report PSEG Nuclear filed with the Nuclear Regulatory Commission (NRC) for Salem was dated July 28, 2006 (Reference 1).Enclosure 1, "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 1 and 2.A0DZ 95-2168 REV. 7/99 JUL S 5 200, Document Control Desk Page 2 LR-N07-0181 Enclosure 2, "Assessment Notes," contains a detailed description for each change or error reported.If you have any questions concerning this report, please contact E. H. Villar at (856) 339 -5456.Sincerely, Robert C. Braun Site Vice President
-Salem Enclosures (2)cc: Mr. Samuel Collins, Administrator
-Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Attn: Mr. R. Ennis, Licensing Project Manager -Salem Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector
-Salem (X24)Mr. P. Mulligan, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625 Enclosure 1 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 Peak Cladding Temperature Rack-Up Sheets Peak Cladding Temperature Rack-Up Sheets PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: Salem Unit 1 Small Break Loss of Coolant Accident (SBLOCA)5/18/2007 19 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 (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 = -16 0 F 10 CFR 50.46 report dated December 8, 1994 (See Note 3) APCT = +109°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 = -8 0 F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) APCT = 0°F 10 CFR 50.46 report dated June 10, 1998 (See Note 8) APCT = 0 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 = +10°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 0 F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = 0°F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = 0 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 10 CFR 50.46 report dated July 28, 2006 (See Note17) APCT = 0-F NET PCT PCT = 1729 0 F 1 B. CURRENT LOCA MODEL ASSESSMENTS NOTRUMP-EM Refined Break Spectrum (See Note 20)Error in IMP Vessel Nozzle Collections (See Note 22)General Code Maintenance (NOTRUMP) (See Note 18)APCT = 0 0 F APCT = 0°F APCT = 0°F I NET PCT PCT = 1729 0 F 2 Peak Cladding Temperature Rack-Up Sheet PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: Salem Unit 1 Large Break Loss of Coolant Accident (LBLOCA)5/18/2007 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 = 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 = +9 0 F 11)10 CFR 50.46 report dated August 27, 2001 (See Note 12) APCT = +6 0 F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = +20°F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = +7 0 F 10 CFR 50.46 report dated July 29, 2004 (See Note 15) APCT = +5 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 = -50 OF NET PCT PCT = 2038°F 3 B. CURRENT LOCA MODEL ASSESSMENTS BASH Minimum and Maximum Time Step Sizes (See Note APCT = 0°F 19)Rebaseline of Limiting LOCBART Calculation (see Note 21) APCT = -8 0 F LOCBART Pellet Volumetric Heat Generation Rate (see Note APCT = 12 0 F 21)LOCBART Oxide to Metal Ratio (see Note 21) APCT = 0°F General Code Maintenance (BASH Code) (Note 18) APCT = 0°F NET PCT PCT = 2042°F 4 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: Salem Unit 2 Small Break Loss of Coolant Accident (SBLOCA)5/18/2007 16 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 (FQ) = 2.4 Nuclear Enthalpy Rise Hot Channel Factor (FAH) = 1.65 Steam Generator Tube Plugging = 25%Limiting Break Size: 2 inches Break Location:
Cold Leg Single Failure: Loss of one train ECCS flow Reference Peak Cladding Temperature (PCT)PCT= 15801F 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 = -16 0 F 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 = -8 0 F 10 CFR 50.46 report dated July 11, 1997 (See Note 7) APCT = 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 = +10°F 10 CFR 50.46 report dated October 18, 1999 (See Note 10) APCT = 0 0 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 = 0°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 10 CFR 50.46 report dated July 28, 2006 (See Note 17) APCT = 0°F NET PCT PCT = 1729 0 F 5 B. CURRENT LOCA MODEL ASSESSMENTS NOTRUMP-EM Refined Break Spectrum (See Note 20)Error in IMP Vessel Nozzle Collections (See Note 22)General Code Maintenance (NOTRUMP) (See Note 18)APCT = O°F APCT = OF APCT = O°F I NET PCT PCT = 1 729 0 F 6 Peak Cladding Temperature Rack-Up Sheets PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: Salem Unit 2 Large Break Loss of Coolant Accident (LBLOCA)5/18/2006 16 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= 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°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 = +24 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 = +9 0 F 11)10 CFR 50.46 report dated August 27, 2001 (See Note 12) APCT = +6 0 F 10 CFR 50.46 report dated August 27, 2002 (See Note 13) APCT = +20°F 10 CFR 50.46 report dated August 08, 2003 (See Note 14) APCT = +7 0 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, 2005 (See Note 17) APCT = 0°F NET PCT PCT = 20381F 7 B. CURRENT LOCA MODEL ASSESSMENTS BASH Minimum and Maximum Time Step Sizes (See Note APCT = 0 0 F 19)Rebaseline of Limiting LOCBART Calculation (See Note 21) APCT = -8°F LOCBART Pellet Volumetric Heat Generation Rate (See APCT = 12 0 F Note 21)LOCBART Oxide to Metal Ratio (see Note 21) APCT = 0°F General Code Maintenance (BASH Code) (See Note 18) APCT = 0 0 F NET PCT PCT = 2042°F 8 Enclosure 2 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 Assessment Notes 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 -13 0 F. The first assessment entailed a +150 0 F penalty that resulted from explicitly modeling safety injection into the broken loop in the NOTRUMP model. The second assessment entailed a -150OF benefit that resulted from the implementation of an improved condensation model. The third assessment entailed a -13 0 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 if 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 -6 0 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 0 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°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 -6 0 F that resulted from corrections made to the LUCIFER code.1
- 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°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 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 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 0 F.2
- 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 I 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+14 0 F PCT assessment due to an error in the feedwater line volume calculation and a+ 13 0 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 0 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 0 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.
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- 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 +1 5 0 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 0 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 I LBLOCA model was assessed a +5 0 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 0 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 0 F.16. Prior LOCA Model Assessment The 10 CFR 50.46 report dated July 28, 2005, reported a 0°F 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 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 0°F increase in the PCT of the SBLOCA analysis due to a SBLOCA evaluation 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 code maintenance (NOTRUMP).
The report also included a 0°F increase in the PCT of the LBLOCA analysis due to the LBLOCA model assessment.
The model assessment for LBLOCA included replacing previously transmitted 4
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. General Code Maintenance (BASH / NOTRUMP)Various changes in code input and output format have been made to enhance usability and help preclude errors in analyses.
This includes both input changes (e.g., more relevant input variables defined and more common input values used as defaults) and input diagnostics designed to preclude unreasonable values from being used, as well as various changes to code output which have no effect on calculated results. In addition, various updates were made to eliminate inactive coding, improve active coding, and enhance commenting, both for enhanced usability and to facilitate code debugging when necessary.
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 0 OF.19. BASH Minimum and Maximum Time Step Sizes A review of some recent BASH-EM sensitivity calculations led to a recommendation to reduce the minimum and maximum time step sizes in BASH during reflood. These changes are being recommended for generic application and have been evaluated for impact on existing analysis results. These changes represent a closely-related group of Non-discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3351.Sensitivity calculations using BASH and SMUUTH show that reducing the minimum and maximum time step sizes in BASH during reflood results in either a negligible change or a modest increase in the integral flooding rate for most cases, leading to an estimated impact of 0 OF for 10CFR50.46 reporting purposes.20. NOTRUMP-EM Refined Break Spectrum The Nuclear Regulatory Commission (NRC) questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM). The NRC was concemed that the resolution of the break spectrum used in the NOTRUMP EM (1.5, 2, 3, 4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10CFR50.46.
That is, the plant could be SBLOCA limited with regard to overall LOCA results. Based on the reanalysis performed for Salem Unit 2 as part of the upcoming steam generator replacement, Westinghouse determined that a specific evaluation was not necessary for Salem Unit 1. Thus, for both Salem Units, the estimated PCT impact is 0 OF.21. LOCBART Version 37.0 Issues The LOCBART code has been modified to correct an inverted term in the calculation of the pellet volumetric heat generation rate. This change affects the steady-state and transient heat generation for all three rods and could result in either an increase or decrease in peak cladding temperature for a given calculation.
This represents a Non-discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
A rebaseline calculation was performed for Salem Units 1 and 2 to determine the limiting LOCBART calculation PCT prior to the error correction, resulting in an estimated PCT benefit of 5 8 OF. The LOCBART Pellet Volumetric Heat Generation Rate impact assessment was then estimated to be 12 OF using the difference between the PCTs from LOCBART calculations with and without the error correction.
The net effect is a PCT increase of 4 OF.An option has been added to the LOCBART code to convert the user-specified zirconium-oxide thickness to equivalent cladding reacted. This adjustment is made during problem initialization, and the cladding outside diameter is modified accordingly.
This change represents a Discretionary Change that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451.
This change is expected to produce a minimal effect of the limiting peak cladding temperature, leading to an estimated effect of 0 OF.22. Errors in IMP Vessel Nozzle Collections Some minor errors were discovered in the reactor vessel nozzle data collections that potentially affect the vessel inlet and outlet nozzle fluid volume, metal mass and surface area. The corrected values have been evaluated for impact on current licensing-basis analysis results and will be incorporated into the plant-specific input databases 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.
These errors are considered minor and would be expected to produce a negligible effect on large and small break PCT results for 10CFR50.46 reporting purposes.
The estimated PCT impact is 0 OF.6