L-PI-15-061, Submittal of 2014 10 CFR 50.46 LOCA Annual Report

Submittal of 2014 10 CFR 50.46 LOCA Annual Report

ML15181A080
Person / Time
Site:	Prairie Island
Issue date:	06/30/2015
From:	Davison K Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-15-061
Download: ML15181A080 (13)

Text

Xcel Energy"'

JUN 3 o 2015 L-PI-15-061 10 CFR 50.46 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 2014 10 CFR 50.46 LOCA Annual Report Pursuant to 10 CFR 50.46, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM") submits the 2014 Annual Report of changes and errors to the Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2 Large Break Loss of Coolant Accident (LBLOCA) and Small Break Loss of Coolant Accident (SBLOCA) analysis. contains the "Non-Plant Specific LOCA Errors and Changes" and summarizes the changes made to both the large break LOCA (LBLOCA) and small break LOCA (SBLOCA) analyses. No changes to the peak cladding temperature (PCT) for Prairie Island Units 1 and 2 occurred since the last annual report for the LBLOCA analysis.

The SBLOCA and LBLOCA peak clad temperature (PCT) assessment sheets for Unit 1 and Unit 2 are included in Enclosure 2. The limiting LOCA analysis PCT for PINGP Unit 1 and Unit 2, with consideration of all1 0 CFR 50.46 assessments, remains the LBLOCA analysis as summarized in Enclosure 2.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

~P~iL~

Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East

• Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Regional Administrator, Region Ill Page 2 Enclosures (2) cc: Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC

ENCLOSURE 1 Non-Plant Specific LOCA Errors and Changes 5 pages follow

Attachment to LTR-LIS-15-54 February 19,2015 Page 1 of 10 GENERAL CODE MAINTENANCE

Background

Various changes have been made to enhance the usability of codes and to streamline future analyses.

Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 ofWCAP-13451.

Affe'cted Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

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Attachment to LTR-LIS-15-54 February 19, 2015 Page 2 of10 ERRORS IN DECAY GROUP UNCERTAINTY FACTORS

Background

En*ors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. The uncertainty factors for 239 Pu were applied to 238 U, and those for 238 U were applied to 239 Pu. This error causes an over-prediction of the uncertainty in decay power from 239 Pu and an under-prediction of the uncertainty in decay power from 238U. Further, the decay group uncertainty factor for Decay Group 6 of 235 U was erroneously coded as 2.5% instead of 2.25%. Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) analysis results. The resolution of these issues represents a closely-related group ofNon-Discretionary Changes in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The issues described above are judged to have either no effect or a negligible effect on the LBLOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F for Prairie Island Units 1 and 2.

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Attachment to LTR-LIS-15-54 February 19, 2015 Page 3 of 10 FUEL ROD GAP CONDUCTANCE ERROR

Background

An error was identified in the fuel rod gap conductance model in the NOTRUMP computer code (reactor coolant system response model). The error is associated with the use of an incorrect temperature in the calculation of the cladding emissivity term. This error corresponds to a Non-Discretionary Change as described in Section4.1.2 ofWCM-13451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The estimated effect was determined based on a combination of engineering judgment of the phenomena and physics of a small break LOCA and sensitivity calculations perfonned with the advanced plant version of NOTRUMP. It was concluded that this error has a negligible effect on small break LOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

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Attachment to LTR-LIS-15-54 February 19, 2015 Page 4 of 10 RADIATION HEAT TRANSFER MODEL ERROR

Background

Two errors were discovered in the calculation of the radiation heat transfer coefficient within the fuel rod model of the NOTRUMP computer code (reactor coolant system response model). First, existing logic did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated. These erroneous calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1°F. Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine. These errors represent a closely related group ofNon-Discretionary problems in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The estimated effect was determined based on a combination of engineering judgment of the phenomena and physics of a small break LOCA and sensitivity calculations performed with the advanced plant version of NOTRUMP. It was concluded that this error has a negligible effect on small break LOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

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Attaclu11ent to LTR-LIS-15-54 February 19,2015 Page 5 of 10 SBLOCTA PRE-DNB CLADDING SURFACE HEAT TRANSFER COEFFICIENT CALCULATION

Background

Two errors were discovered in the pre-departure from nucleate boiling (pre-DNB) cladding surface heat transfer coefficient calculation in the SBLOCTA code (cladding heat-up calculations). The first error is a result of inconsistent time units (hours vs. seconds) in the parameters used for the calculation of the Reynolds and Prandtl numbers, and the second error relates to an incorrect diameter used to develop the area tenn in the cladding surface heat flux calculation. Bofu of these issues impact the calculation of fue pre-DNB convective heat transfer coefficient, representing a closely related group of Non-Discretionary Changes to the Evaluation Model as described in Section 4.1.2 ofWCAP-13451.

Mfected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect These errors have been corrected in the SBLOCTA code. Because this condition occurred prior to DNB, it was judged that these errors had no direct impact on fue cladding heat-up related to fue core uncovery period. A series of validation tests were performed and confirmed that these errors have a negligible effect on SBLOCA analysis results, leading to an estimated Peale Cladding Temperature (PCT) impact of 0°F.

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ENCLOSURE 2 LOCA Peak Clad Temperature Summary (Rack-Up Sheets)

Prairie Island Nuclear Generating Plant (includes plant specific changes and non-zero non-plant specific changes) 4 pages follow

Attachment to LTR-LIS-15-54 February 19, 2015 Page 6 of 10 Westinghouse LOCA Peale Clad Temperature Summary for ASTRUM Best Estimate Large Brealc Plant Name: Prairie Island Unit 1 Utility Name: Xcel Energy, lnc Revision Date: 2/5/2015 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP(%): 10 Notes:

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. Evaluation of fuel Pellet Thennal Conductivity Degradation and Peaking 227 2 (a)

Factor Bum down 2 . Revised Heat Transfer Multiplier Distributions -2 3 3

• Error in Burst Strain Application 25 4 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2014 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2015

• It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References I . WCAP-17783-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Prairie Island Units I and 2 with Replacement Steam Generators Using ASTRUM Methodology," June 2013.

2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown," September 20, 2012.

3 . LTR-LIS-13-366, Revision I, "Prairie Island Units I and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions," August 2013.

4 . LTR-LIS-14-50, "Prairie Island Units I and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction," January 2014.

Notes:

(a) This evaluation credits peaking factor bumdown, see Reference 2.

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Attachment to LTR-LIS-15-54 February 19, 2015 Page 7 of 10 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 1 Utility Name: Xeel Energy, Inc Revision Date: 2/5/2015 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 10 Notes: Zirlo (14X14), Framatome RSG Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 959 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2014 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 959

• It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References I . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units 1 and 2 PCT Summaries," February 2008.

Notes:

None Page 11

Attachment to LTR-LIS-15-54 February 19, 2015 Page 8 of 10 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/5/2015 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP(%): 10 Notes:

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. Evaluation of Fuel PelletThennal Conductivity Degradation and Peaking 227 2 (a), (b)

Factor Bumdown 2 . Revised Heat Transfer Multiplier Distributions -2 3 3 . Error in Burst Strain Application 25 4 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2014 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2015

• It is recommended that the licensee detennine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References I , WCAP-17783-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Prairie Island Units I and 2 with Replacement Steam Generators Using ASTRUM Methodology," June 2013.

2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notification and Reporting for. Fuel Pellet Thennal Conductivity Degradation and Peaking Factor Bumdown," September 20, 2012.

3 . LTR-LIS-13-366, Revision I, "Prairie Island Units 1 and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions," August 2013.

4 . LTR-LIS-14-50, "Prairie Island Units I and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction," January 2014.

Notes:

(a) This evaluation credits peaking factor burndown, see Reference 2.

(b) The reporting text and line item originally identified for Unit I in Reference 2 is applicable to Unit 2 with RSGs.

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Attachment to LTR-LIS-15-54 February 19, 2015

,

• Page 9 of 10 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/512015 Analysis Information EM: NOTRUMP Analysis Date: 1121/2008 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%):. 10 Notes: Zirlo (14X14), AREVA RSG Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 959 1,2 a PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2014 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT== 959

• It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References 1 . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units 1 and 2 PCT Summaries," February 2008.

2 . LTR-LIS-13-274, "Prairie Island Units 1 and 2, I 0 CFR 50.46 Summary Sheets for the Evaluation to Support the Unit 2 Installation of AREVA Model 56119 Replacement Steam Generators (RSGs)," June 2013.

Notes:

(a) The Unit l AOR is applicable to Unit 2 with the RSGs installed.

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