ML13115A279

10 CFR 50.46 30-day Report and Annual Report for Changes to the Emergency Core Cooling System Performance Analysis

ML13115A279
Person / Time
Site:	Calvert Cliffs
Issue date:	04/22/2013
From:	John Stanley Calvert Cliffs 3 Nuclear Project, EDF Group, Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13115A279 (11)

Text

Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 CENG.

a joint venture of Constellation jmeDF O ergy, VO*%

CALVERT CLIFFS NUCLEAR POWER PLANT April 22, 2013 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318 10 CFR 50.46 30-day Report and Annual Report for Changes to the Emergency Core Cooling System Performance Analysis

REFERENCES:

(a) Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP),

dated February 18, 2011, Amendment re: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (b) Letter from Ms. N. S. Morgan (NRC) to Mr. G. H. Gellrich (CCNPP),

dated December 19, 2012, Safety Evaluation of the Realistic Large-Break Loss-of-Coolant Accident Summary Report (c) Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC),

dated January 19, 2012, 10 CFR 50.46 30-day Report for Changes to the Emergency Core Cooling System Performance Analysis (d) Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC),

dated April 30, 2012, 10 CFR 50.46 30-day Report and Annual Report for Changes to the Emergency Core Cooling System Performance Analysis This letter is submitted pursuant to 10 CFR 50.46(a)(3)(ii) to provide notification of a significant change to the peak cladding temperature analysis result for the large break loss-of-coolant accident (LB LOCA) and the small break loss-of-coolant accident (SB LOCA) analyses. Because the effect on the peak cladding temperature of the changes is greater than 50'F from the temperature calculated for the limiting transient using the last acceptable model, the analysis changes qualify as significant as defined in 10 CFR 50.46(a)(3)(i) and, consequently, are provided in Attachment (1).

The analyses for the LB LOCA Emergency Core Cooling System performance have been re-performed for Unit 2, Cycle 20. The analysis was performed using the latest Nuclear Regulatory Commission

Document Control Desk April 22, 2013 Page 2 accepted version of the AREVA evaluation model for pressurized water reactors to Calvert Cliffs (References a and b). The new analyses explicitly model the AREVA fuel used in Unit 2, Cycle 20 in the spring of 2013.

The analysis for the SB LOCA Emergency Core Cooling System performance has been assessed for Unit 2, Cycle 20. The SB LOCA analysis performed for Unit 2, Cycle 19 is applicable to Unit 2, Cycle 20. The analysis applied the latest Nuclear Regulatory Commission accepted version of the AREVA evaluation model for pressurized water reactors to Calvert Cliffs (Reference a). A 10 CFR 50.46 30-day letter (Reference c) was previously submitted to document a greater than 50'F change in peak clad temperature for the SB LOCA analysis for Unit 2 due to an error.

The new LB LOCA analysis constitutes the new licensing basis for Unit 2 as of March 22, 2013, when Unit 2 entered Mode 4. The SB LOCA assessment for Unit 2, Cycle 20 maintains the current licensing basis for Unit 2. Attachment (1) contains the results of the change in the peak cladding temperature based on the change to the LB LOCA evaluation model used for the analysis for Unit 2.

In addition, AREVA has reported changes to, or errors in, the acceptable evaluation models for calendar year 2012. There are no changes in the peak clad temperature for the SB LOCA reported for 2012. This annual report is contained in Attachment (2). These changes were previously reported for Unit I in Reference (d).

Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.

Vanefs%.Stan¶'

Manager-Engineering Services JJS/PSF/bjd Attachments: (1) 10 CFR 50.46- 30 Day Report (2) 10 CFR 50.46 - Annual Report cc: N. S. Morgan, NRC Resident Inspector, NRC W. M. Dean, NRC S. Gray, DNR

ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT Calvert Cliffs Nuclear Power Plant, LLC April 22, 2013

ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT INTRODUCTION This letter is submitted pursuant to 10 CFR 50.46(a)(3)(ii) to provide notification of a significant change to the peak cladding temperature analysis result for the Unit 2 large break loss-of-coolant accident (LB LOCA) analyses. Because the effect on the peak cladding temperature of the changes is greater than 50'F due to a previously reported error combined with a change in the LB LOCA methodology, the analyses changes qualify as significant as defined in 10 CFR 50.46(a)(3)(i) and, consequently, are provided below.

Calvert Cliffs request to amend the Unit 2 renewed operating licenses to transition from Westinghouse to AREVA-designed fuel was approved by the Nuclear Regulatory Commission (NRC) in Reference 1. As part of that transition, the Emergency Core Cooling System (ECCS) performance for both the LB LOCA and the SB LOCA were re-analyzed. The NRC acceptance of the LB LOCA analysis permitted a single cycle of operation for Unit 2. A revised realistic LB LOCA methodology that included fuel burned in multiple cycles and removed the statistical treatment of the decay heat multiplier was approved by the NRC in Reference 2. This approved methodology was used to re-perform the LB LOCA analysis for Unit 2, Cycle 20.

REFERENCE ANALYSES LB LOCA The Unit 2 LB LOCA ECCS performance analysis was performed with the AREVA evaluation method approved in Reference 2. The analysis includes twice-burned AREVA fuel and restricted the decay heat multiplier to a single value of 1.0, rather than ranging the multiplier.

The analysis resulted in an absolute change in peak clad temperature from the prior analysis of record of 50'F. The analysis used the S-RELAP5 code version that contains an error in the treatment of the Sleisher-Rouse correlation. That error changes the peak clad temperature by 8'F, and was previously reported (Reference 3). The combined difference from the last acceptable evaluation model is greater than 50'F. A comparison of peak clad temperature results is provided in Table 1.

Table 1, AREVA LB LOCA Peak Clad Temperature Analysis Results Item PCT, -F New Analysis of Record (twice-burned AREVA fuel) 1,620 Old Analysis of Record (first-burned AREVA fuel) 1,670 Table 2 provides the results of the new LB LOCA analysis demonstrating conformance with the acceptance criteria of 10 CFR 50.46(b).

Table 2, LB LOCA versus Acceptance Criteria Parameter Criterion Result Peak Cladding Temperature, 'F 2200 1620 Maximum Cladding Oxidation, % <17 2.460 Maximum Core-Wide Cladding Oxidation, % <1 0.0111 Coolable Geometry Yes Yes I

ATTACHMENT (1) 10 CFR 50.46 - 30 DAY REPORT

SUMMARY

The new LB LOCA analysis constitutes the new licensing basis analyses (analyses-of-record) for Calvert Cliffs Unit 2. It is used as the reference analysis to evaluate the impact on peak cladding temperature of changes to or errors in the AREVA LB LOCA evaluation model and its application to Calvert Cliffs.

REFERENCES

1. Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP), dated February 18, 2011, Amendment re: Transition from Westinghouse Nuclear-Fuel to AREVA Nuclear Fuel

2. Letter from Ms. N. S. Morgan (NRC) to Mr. G. H. Gelirich (CCNPP), dated December 19, 2012, Safety Evaluation of the Realistic Large-Break Loss-of-Coolant Accident Summary Report

3. Letter from Mr. J. J. Stanley (CCNPP) to Document Control Desk (NRC), dated April 30, 2012, 10 CFR 50.46 30-day Report and Annual Report for Changes to the Emergency Core Cooling System Performance Analysis 2

ATTACHMENT (2) 10 CFR 50.46 - ANNUAL REPORT Calvert Cliffs Nuclear Power Plant, LLC April 22, 2013

FAB13-00140-002 Page 2 Calvert Cliffs Unit 1 and Unit 2 NUCLEAR PLANTs - 10CFR50.46 ANNUAL REPORT -

EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES In accordance with the annual reporting requirements of 10CFR50.46 (a)(3)(ii), the following is a summary of the limiting design basis accident (loss-of-coolant accident) analysis results established using the current Calvert Cliffs ECCS evaluation model from January 2012 to end of March 2013.

Table 1 Calvert Cliffs RLBLOCA PCT for ANP-2834-000 Report Analysis PCT (OF) Delta PCT (OF) Year Notes Licensing Basis - Analysis of 1670 2009 September 2009.

Record Updated Licensing Basis 1678 December 2011 Annual PCT at the end of 2011 Report Liquid fallback into surrounding 6 assemblies 2012 Attachment I Cathcart-Pawel Uncertainty Implementation in RLBLOCA 0 2012 Attachment 2 Applications Updated Licensing Basis 1678 PCT Net Change 0

FAB13-00140-002 Page 3 Table 2 Calvert Cliffs RLBLOCA PCT for ANP-3043-001 Report Analysis POT (°F) Delta PCT (OF) Year Notes Licensing Basis - Analysis 1620 2011 December 2011.

of Record S-RELAP5 Sleicher-Rouse +8 2011 Attachment 3 correlation Cathcart-Pawel Uncertainty Implementation in 0 2012 Attachment 2 RLBLOCA Applications Updated Licensing Basis 1628 0 PCT Net Change 8 Table 3 Calvert Cliffs Small Break LOCA PCT Analysis PCT (OF) Delta POT Year Notes (OF)

Licensing Basis AnalysisgofaRecord- 1626 2009 September 2009.

Analysis of Record Updated Licensing Basis PCT at the 1695 December 2011 Annual Report end of 2011 Updated Licensing 1695 0 Basis PCT Net Change 0 There are no changes in the PCT reports for Small Break LOCA to be reported in 2012 annual 50.46 reporting.

FAB13-00140-002 Page 4 Attachment I Liquid fallback into surrounding 6 assemblies Backqound The issue was discovered when reviewing the results for a W 3-loop plant sensitivity study. The limiting case revealed a non-conservatism in the core exit modeling relative to the form loss coefficient (FLC) between the upper plenum and the central core region. Liquid communication from the central core region, which represents six assemblies surrounding the hot channel, to the hot channel resulted in steam cooling and a reduction in the hot rod PCT at elevations associated with the peaked axial power.

This Condition Report is related to CR 2011-1688, where in all plant cases, a high reverse form loss coefficient is applied to the HC and central core to UP junctions at the beginning of the core reflooding phase. In CR 2011-1688, cases were rerun that had liquid down flow into the hot channel and potentially affected the AOR PCT limit. Whereas CR 2011-1688 involved rerunning cases that had liquid downflow into the hot channel, CR 2012-2301 involves rerunning cases that had liquid downflow into the central core. That is, CR 2012-2301 only changes the criteria for rerunning cases with the high reverse form loss coefficients.

Results The PCT impact for Calvert Cliffs Units 1 and 2 RLBLOCA is 0 F for the Analysis of Record of ANP- 2384P-000, There is no impact on ANP-3043P-001 as the analysis included this fix, therefore no PCT reporting is required.

Cathcart-Pawel Uncertainty Implementation in RLBLOCA Applications Backqound

References:

FAB13-00140-002 Page 5 Ref. 1. J.V. Cathcart and R.E. Pawel, Zirconium Metal-Water Oxidation Kinetics: IV. Reaction Rate Studies, ORNL/NUREG-17, August 1977.

In realistic large break loss of coolant accident (RLBLOCA) analyses, energy released through the oxidation of cladding is calculated from the Cathcart-Pawel correlation for oxide growth (Reference 1, Equation A2). The correlation has the form:

52/2 = A

• exp(-Q/R*/I-)

Where A and Q are experimentally determined constants and R and T are the gas constant and temperature, respectively. The uncertainty parameter for the A value is given in terms of the natural logarithm: In(A). The value of In(A) follows a normal distribution and the value of A follows a log-normal distribution. RLBLOCA applications implement the Cathcart-Pawel uncertainty using a log-normal function for the uncertainty multiplier, B, applied to a constant, A.

The equation to determine the uncertainty multiplier, B, was determined to be incorrect.

Results The incorrect equation still has a log-normal distribution like the corrected equation for the uncertainty multiplier, B. In addition, the range of sampled values for B falls within the range expected for the corrected equation for the uncertainty multiplier, B. The estimated impact of this change on the LBLOCA analyses for the Calvert Cliffs analyses is +0°F on the calculated peak cladding temperature.

S-RELAP5 Code Programming of Sleicher-Rouse correlation

Background

Sleicher-Rouse is one of the correlations used to define the heat transfer between the fuel and coolant. This correlation is applicable to both Large and Small Break analyses performed with the S-RELAP5 computer code.

During development of a BWR LOCA methodology based on S-RELAP5, the behavior of the Sleicher-Rouse correlation relative to other single-phase vapor heat transfer correlations was reviewed and it was questioned whether Sleicher-Rouse correlation was correct. It was also discovered that another industry code uses the Sleicher-Rouse correlation, but the form for the correlation is different than that used in S-RELAP5 implementation of Sleicher-Rouse. The concern is related to the form of the equation for calculating the exponent of the temperature ratio correction term.

The S-RELAP5 form is n = -loglO(Tw/Tg)1/4 + 0.3 The alternate form used in another industry code is n = -[IoglO(Tw/Tg)]1/4 + 0.3

FAB13-00140-002 Page 6 The alternate form appears to be more consistent with other heat transfer correlations and expected physical trends.

Results Preliminary assessments of the potential impact of using the alternate Sleicher-Rouse correlation form were performed. A development version of S-RELAP5 was prepared with the alternate Sleicher-Rouse form and several code validation and plant sample problems were repeated. Additional analyses were performed using a different heat transfer correlation for single-phase vapor heat transfer. The assessments included analyses for both RLBLOCA and SBLOCA.