ML13172A009

Nd 2, Closure Evaluation for Report Pursuant to 10 CFR 50.46(a)(3) Concerning Significant Emergency Core Cooling System Evaluation Model Error Related to Nuclear Fuel Thermal Conductivity Degradation

ML13172A009
Person / Time
Site:	Surry
Issue date:	07/01/2013
From:	Cotton K Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Cotton K
References
TAC MF0599, TAC MF0600
Download: ML13172A009 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 1, 2013 Mr. David A Heacock President and Chief Nuclear Officer Virginia Electric and Power Company lnnsbrook Technical Center 5000 Dominion Blvd.

Glenn Allen, VA 23060

SUBJECT:

SURRY POWER STATION, UNIT NOS., 1 AND 2- CLOSURE EVALUATION FOR REPORT PURSUANT TO 10 CFR 50.46(a)(3) CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERROR RELATED TO NUCLEAR FUEL THERMAL CONDUCTIVITY DEGRADATION (TAC NOS. MF0599 AND MF0600)

Dear Mr. Heacock:

The Virginia Electric and Power Company (Dominion), the licensee for Surry Power Station, Unit Nos. 1 and 2 (SPS), submitted a report pursuant to Title 10 of the Federal Code of Regulations (1 0 CFR) 50.46(a)(3), describing the estimated effect of thermal conductivity degradation on the predicted peak cladding temperature associated with the emergency core cooling system (ECCS) evaluation model for large-break loss-of-coolant accidents in their letter dated July 10, 2012.

The U.S. Nuclear Regulatory Commission (NRC) has completed the review. Upon evaluating the report, the NRC staff has determined that the SPS satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR),

for the realistic ECCS evaluations revision of 10 CFR 50.46 (53 FR 35996).

A copy of the Closure Evaluation is enclosed.

If you have any questions regarding this matter, I may be reached at (301) 415-1438 or via e-mail at karen.cotton@nrc.gov.

Sincerely,

~~

Karen Cotton, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-280 and 50-281

Enclosure:

As stated cc w/encl: Distribution via Listserv

CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2

• REPORT DESCRIBING THE NATURE OF AND ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE OF A SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERROR DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated July 10, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12199A061), Virginia Electric and Power Company (Dominion),

submitted a report describing a significant error identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the error on the predicted peak

• cladding temperature (PCT) for Surry Power Station (SPS), Unit Nos. 1 and 2. This report was submitted pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50, Section 46 (1 0 CFR 50.46), paragraph (a)(3).

The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR), for the realistic ECCS evaluations revision of 10 CFR 50.46 (53 FR 35996). The staff review is discussed in the following sections of this closure evaluation.

2.0 REGULATORY EVALUATION

2.1 Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCSs for light water nuclear power reactors are promulgated at 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change to, or error in, an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature different by more than 50 degrees Fahrenheit (°F) from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 °F.

Enclosure

For each change to or error discovered in an acceptable evaluation model or in the application of such a model, paragraph (a)(3)(ii) to 10 CFR 50.46 requires the affected licensee to report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually. If the change or error is significant, the licensee is required to provide this report within 30 days and include with the report a proposed schedule for providing a re-analysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.

2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the statement of considerations published on September 16, 1988, in the FR for the best estimate loss-of-coolant-accident (LOCA) revision of 10 CFR 50.46 (53 FR 35996):

[Paragraph (a)(3) of Section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although re-analysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications mC!de subsequent to NRC review of the evaluation model. ..

Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model. .. More timely reporting (30 days) is required for significant errors or changes ...

the final rule revision also allows the NRC to determine the schedule for re-analysis based on the importance to safety relative to other applicant or licensee requirements.

The NRC staff considered the discussion in the Federal Register in its evaluation of the error report submitted by the licensee.

3.0 TECHNICAL EVALUATION

The report submitted by the licensee described the effects of an error in the ECCS evaluation model associated with the degradation of thermal conductivity in nuclear fuel. This issue is discussed in NRC Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," and its potential effects in realistic emergency core cooling system evaluation models are described in IN 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation [TCD]."

Based on the nature of the reported error, and on the magnitude of its effect on the PCT calculation, the NRC staff determined that a detailed technical review is necessary .. Based on the regulatory evaluation discussed above, the staff's review was performed to ensure that the NRC staff agrees with the licensee's assessment of the significance of the error, and to enable the staff to verify that the evaluation model, as a whole, remains adequate. Finally, the NRC staff's review also establishes that the licensee's proposed schedule for re-analysis is acceptable in light of the safety significance of the reported error.

Overview of ASTRUM The licensee uses the NRC-approved Automated Statistical Treatment of Uncertainty Method (ASTRUM), documented in WCAP-16009-NP-A (ADAMS Accession Nos. ML050910157, ML050910159, and ML050910161 ), to evaluate ECCS performance. ASTRUM relies on an approach based on order statistics, in which a set number of cases with randomly varied initial conditions are analyzed using the WCOBRAITRAC (WC/T) reactor system analysis code. The number of cases is chosen so that the highest predicted PCT within the case set becomes a predictor of the 95/95 upper tolerance limit for the PCT associated with a hypothetical populations of LOCA scenarios. The result is used to show compliance with the 10 CFR 50.46(b)(1) acceptance criterion concerning PCT.

3.1 SUMM~RY OF TECHNICAL INFORMATION IN THE REPORT The licensee's report indicated that the effect of the TCD error was 183 oF for Units 1 and 2.

The nature of the error, and the method used to estimate its effect on the calculated peak fuel cladding temperature, is discussed in greater detail in Attachment 3 to the licensee's July 10, 2012, report. The report also disclosed the estimated effect of a model change to account for a non-uniform pellet radial profile, which had an estimated effect of -13 oF on the predicted peak' cladding temperature.

TCD Error Correction The error in the ECCS evaluation model was caused by the inability of the Westinghouse Improved Fuel Rod Performance and Design (PAD 4.0) fuel performance model to account for the effects of TCD with increasing fuel burnup. This error caused fuel temperature initial conditions to be non-conservatively low for higher burn up fuel rods that were analyzed in the ECCS evaluation. In order to correct for the error, a burnup-dependent term was added to the nuclear fuel thermal conductivity equation, which caused the predicted initial fuel temperatures to compare better with experimental data obtained from the Halden Reactor Projed. The results from the modified PAD (PAD 4.0 + TCD) code were then used to re-initialize the WC/T cases that are performed in execution of ASTRUM.

• Note that the TCD correction also includes a peaking factor burndown effect, which captures a reduction in the core peaking factors that naturally occurs through fuel life. This phenomenon partially offsets the net effect of TCD by lowering the initial stored energy in the fuel.

Estimation of the Effect of TCD in the PCT Calculation In order to estimate the PCT effect of the TCD error correction, the licensee systematically identified a subset of 45 cases within the ASTRUM analysis that had the potential to produce a limiting result, once corrected for TCD. These cases were re-analyzed within WC/T, using the adapted initial conditions described above. The following passage from Section 3.0 of to Dominion's July 10, 2012, letter, describes this process in greater detail:

1 Although comparisons of PAD 4.0 and PAD 4.0 + TCD predictions to Halden Reactor measurements and data are Westinghouse proprietary information, related information and similar comparisons are available from the NRC's FRAPCON computer code in NUREG/CR-7022, "FRAPCON-3.4: Integral Assessment" See in particular Chapter 3 of NUREG/CR-7022.

To estimate the effect of fuel TCD and peaking factor burndown, a total of 45 WCOBRA/TRAC [WC/T] executions were performed. The uncertainty attributes of these executions were taken from among the most limiting cases from the original 124-run ASTRUM analysis. The evaluation considered an adequate range of burnup such that the effects of TCD and related burn up effects were captured ...

The estimated effect of TCD was then taken as the difference between the maximum PCT when considering the effects of fuel TCD and peaking factor burn down and the 95/95 PCT from the Surry ASTRUM analysis with 15x15 Upgrade Fuel.

Reported Results

• Following the correction for TCD and the model change, the current predicted PCT for SPS Units 1 and 2 is 2037 °F. The licensee also stated that Dominion will conduct a re-analysis following approval by the NRC of a revised large-break loss-of-coolant accident (LBLOCA) evaluation model that includes the effects of TCD and accommodates the 10 CFR 50.46(c) rulemaking process.

3.2 Summary of Staff Evaluation In its evaluation, the NRC staff reviewed (1) the approach used to estimate the effects of TCD, (2) the estimated effect of TCD at both units, and (3) the licensee's schedule for re-analysis in consideration of the approach used to estimate the effects of TCD. As discussed in the following paragraphs, the NRC staff determined that the licensee's estimate and proposal for re-analysis are acceptable.

To estimate the effects of TCD, the licensee used a modified uranium thermal conductivity model to account for TCD, andre-executed the most sensitive WC/T cases using inputs from the revised thermal conductivity model. The explicit fuel performance model is described in a March 7, 2012, Westinghouse letter to the Commission (ADAMS Accession No. ML12072A036). A proprietary enclosure to the Westinghouse letter also provides information to show that the modified uranium thermal conductivity model more accurately reflects available high-burnup data, as described in Section 3.1 in this evaluation. The NRC staff also notes that the licensee has examined the ASTRUM run set to identify the WC/T cases with potential to be most significantly affected by TCD, and explicitly re-analyzed these cases with initial conditions that accurately capture the TCD phenomenon.

The NRC staff has reviewed estimating techniques for the same phenomena in the generically approved ASTRUM evaluation model for several other licensing actions. In a recent request for extended power uprate, the licensee addressed a staff request for additional information by identifying a limiting subset of cases to re-execute, and then by completely re'-executing the entire ASTRUM run set. In this investigation, the original, limited set of cases contained the new limiting PCT. Also, s~veral reports submitted pursuant to 50.46 have provided TCD effect estimates by re-executing a more limited subset of the original ASTRUM run set. In the case of the uprate, the NRC staff concluded that the licensee had acceptably accounted for the effects of TCD in its ECCS evaluation; in the case of the 50.46 reports, the NRC staff determined that the estimates provided in the reports satisfied the applicable reporting requirements.

Based on the following considerations: (1) The PAD 4.0 + TCD and revised HOTSPOT fuel performance models generate fuel stored energy initial conditions that in reasonable agreement with available high burnup data, and (2) the licensee has identified the limiting WC/T cases and re-executed them using the revised fuel performance models, the NRC staff concludes that the licensee's estimate of the effects of TCD is acceptable. The NRC staff also notes, as discussed above, that this approach has been applied previously at other licensed facilities and accepted by the staff.

• The estimated effect of TCD at Surry is 183 oF for Units 1 and 2. Recently received explicit estimates of the effects of TCD using the ASTRUM evaluation model have ranged from 73 oF to 384 oF; these estimates fall within that range. The updated PCTs, which also include the effect of the pellet radial power profile change, are 2037 oF for Units 1 and 2, which fall within the regulatory acceptance criterion of 2200 oF. Because the effect of TCD is consistent with other, similar estimates, and because the updated PCTs meet the 10 CFR 50.46(b)(1) acceptance criteria, the staff did not identify any significant issues with the estimates.

In its cover letter, Dominion stated the following:

Before June 15, 2017, Dominion will submit to the NRC for review and approval a LBLOCA analysis that applies NRC-approved methods that include the effects of fuel TCD. The date for the analysis submittal is based on the following milestones, which must be completed in order to perform a revised licensing basis LBLOCA analysis with an NRC-approved ECCS EM that explicitly accounts for TCD:

1. NRC approval of a fuel performance analysis methodology that includes the effects of TCD. The new methodology for developing inputs to the LBLOCA EM would replace the current SPS licensing basis methodology in WCAP-15063-P-A, Revision 1, which is referenced in Section 3.4 of the SPS Updated Final Safety Analysis Report (UFSAR).

2. NRC approval of a LBLOCA EM that includes the effects of TCD and accommodates the ongoing 10 CFR 50.46(c) rulemaking process. The new methodology would replace the current licensing basis analysis methodology in WCAP-16009-P-A, which is referenced in Section 14.5 of the SPS UFSAR.

10 CFR 50.46(a)(3)(ii) states, in part, that the licensee "shall include with the report a proposed schedule for providing a re-analysis or taking other action as may be needed to show compliance with [1 0 CFR] 50.46 requirements." As described in the Regulatory Evaluation, the statements of consideration explain further that "the final rule revision also allows the NRC to determine the schedule for re-analysis based on the importance to safety relative to other applicant or licensee requirements." The NRC staff determines herewith, based on the provided date of prior to June 15, 2017, the licensee will submit to the NRC for review and approval a LBLOCA that applies NRC-approved methods that include the effects of fuel TCD, that the re-analysis requirement is satisfied.

In summary, the NRC staff reviewed the licensees report estimating the effect of TCD on the large break LOCA analyses for SPS, Units 1 and 2. Based on the technical rigor employed by the licensee, which included correcting the TCD error using a model that agrees with available experimental data and explicitly re-evaluating a limiting subset of WC/T cases, the NRC staff

concluded that the TCD estimate was acceptable. The NRC staff did determine that the licensee satisfied the re-analysis requirement set forth in 10 CFR 50.46(a)(3)(ii) providing a re-analysis submittal date of June 15, 2017.

4.0 CONCLUSION

Based on the considerations discussed above, the NRC staff finds that the report* submitted pursuant to 10 CFR 50.46(a)(3), concerning an ECCS evaluation model error pertaining to TCD, satisfies the intent of the 10 CFR 50.46 reporting requirements. The report and supplemental information enabled the staff to ( 1) determine that it agrees with the licensee's assessment of the significance of the error, (2) confirm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b). The licensee did include a proposed schedule for providing a re-analysis, June 15, 2017, the NRC staff did conclude that the licensee satisfied this aspect of the 10 CFR 50.46(a)(3)(ii) reporting requirement.

ML13172A009 NRRIDORULPL2- NRRIDORULPL2- NRRIDORULPL2- NRRIDORULPL2-NRR/DSS/BC 1/BC 1/PM OFFICE 1/PM 1/LA NAME KCotton SFigueroa CJackson RPascarelli KCotton DATE 06/25/13 06/24/13 05/23/13 06/27/13 07/01/13