ML12164A786

Response to Request for Additional Information Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Cooling System

ML12164A786
Person / Time
Site:	Kewaunee
Issue date:	06/08/2012
From:	Price J Dominion, Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
12-352, LIC/CDS/R2
Download: ML12164A786 (69)

Text

Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 D ommnmon PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 June 8, 2012 ATTN: Document Control Desk Serial No.12-352 U. S. Nuclear Regulatory Commission LIC/CDS/R2 11555 Rockville Pike Docket No.: 50-305 Rockville, MD 20852-2738 License No.: DPR-43 DOMINION ENERGY KEWAUNEE. INC.

KEWAUNEE POWER STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 By letter dated February 16, 2012 (Reference 1), the U.S. Nuclear Regulatory Commission (NRC) issued a letter to Dominion Energy Kewaunee, Inc. (DEK) requesting information pursuant to 10 CFR 50.54(f). Specifically, the letter required DEK to provide information regarding the effect of fuel pellet thermal conductivity degradation (TCD) on peak cladding temperature (PCT) calculated with the Westinghouse Electric Company (Westinghouse) realistic emergency core cooling system (ECCS) evaluation model (EM) employed for the Large Break Loss of Coolant Accident (LBLOCA) analysis for Kewaunee Power Station (KPS). The letter required that DEK provide information on KPS regarding the effect of a potentially significant error, as defined in 10 CFR 50.46(a)(3)(i), associated with TCD and the estimated effect on PCT. The purpose of the request was to ensure continued compliance with the PCT acceptance criterion promulgated in 10 CFR 50.46(b)(1).

By letter dated March 15, 2012 (Reference 2), DEK submitted a response to the NRC's information request related to the estimated effect on peak cladding temperature resulting from thermal conductivity degradation in the Westinghouse furnished realistic emergency core cooling evaluation. DEK also stated that this response served as a 30-day report in accordance with requirements of 10 CFR 50.46.

In the course of their review of DEK's response, the NRC staff determined that additional information is necessary to complete its review. In an e-mail message dated March 26, 2012 (Reference 3) the NRC staff transmitted draft questions. The questions were discussed and finalized during telephone conferences between the NRC staff and DEK on March 30, 2012 and on May 21, 2012. The purpose of this letter is to provide the DEK responses to the NRC staff questions.

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION THAT IS BEING WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390. UPON SEPARATION OF ATTACHMENT 2, THIS PAGE IS DECONTROLLED.

Serial No.12-352 Response to 10 CFR 50.46 RAIs Page 2 of 4 provides the DEK responses to NRC questions 3, 7, and 8. Attachment 2 provides responses to the remaining NRC questions and contains information proprietary to Westinghouse Electric Company LLC. Attachment 3 provides a non-proprietary version of the information contained in Attachment 2. Attachment 4 provides 10 CFR 50.46 report pages that support the response to NRC question 5. is supported by a Westinghouse Application for Withholding Proprietary Information for Public Disclosure and the accompanying Affidavit signed by Westinghouse, the owner of the information, which is provided in Attachment 5. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of Attachment 2 or the supporting Westinghouse affidavit should reference letter CAW-12-3495 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066. A redacted (non-proprietary) version of Attachment 2 has been included as Attachment 3 for public disclosure.

If you have questions or require additional information, please feel free to contact Mr.

Craig Sly at 804-273-2784.

Very truly yours, J Al Price VICKI L.HULL, Vi President - Nuclear*Engineering I Notary Public Commonwealm of Virginia r*

COMMONWEALTH OF VIRGINIA My-Commcision140542 Expires May'31.2014 COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and State aforesaid, today by J. Alan Price, who is Vice President Nuclear Engineering of Dominion Energy Kewaunee, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this day of r.- ,2012.

My Commission expires:LAIAj i/2,/4 31, Notary Public

Serial No.12-352 Response to 10 CFR 50.46 RAIs Page 3 of 4 Attachments:

1. NRC Questions 3, 7, and 8 and DEK Responses

2. NRC Questions 1, 2, 4-6, and 9 and DEK Responses (Westinghouse Proprietary)

3. NRC Questions 1, 21, 4-6, and 9 and DEK Response (Non-Proprietary)

4. 10 CFR 50.46 Reporting Pages that Support the Response to NRC Question 5

5. Westinghouse Affidavit

References:

1. Letter from Michele G. Evans (NRC) to David A. Heacock (DEK), "Kewaunee' Power Station - Information Request Pursuant to 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Core Cooling System Evaluation (TAC No. M99899)," dated February 16, 2012. [ADAMS Accession No. ML120410195]

2. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Response to Information Request Pursuant to 10 CFR50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Core Cooling System (ECCS)

Evaluation and 30-Day Report of ECCS Model Changes Pursuant to the Requirements of 10 CFR 50.46," dated March 15, 2012. [ADAMS Accession No. ML12079A287]

3. E-mail from Karl Feintuch (NRC) to Craig Sly (DEK), "

Subject:

Most Recent Update to ME8460 RAI's - clarification call - continuation - will do record tomorrow," dated May 21, 2012.

Commitments made by this letter:. None.

Serial No.12-352 Response to 10 CFR 50.46 RAIs Page 4 of 4 cc: Director, Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 13-H116M 11555 Rockville Pike Rockville, MD 20852-2738 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 1 of 11 ATTACHMENT I NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 NRC QUESTIONS 3,7. and 8 AND DEK RESPONSES (11 Pages including this cover page)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 2 of 11

3. (ME8460-RAII-SRXB-Parks-003-2012-06-08)

Fully explain all peaking factor adjustments and provide the rationalefor each adjustment.

Response to (ME8460-RAII-SRXB-Parks-003-2012-06-08)

The Kewaunee Power Station (KPS) LBLOCA evaluation for fuel pellet thermal conductivity degradation (TCD) included two sets of peaking factor adjustments:

• Reduction in full-power FAH (hot rod average power, including uncertainties) from 1.8 (analysis of record value) to 1.7 (current full-power limit in the core operating limits report (COLR), with a corresponding reduction in hot assembly average power (including uncertainties); and,

, Peaking factor burndown with increasing rod burnup for FAH, FQ Steady-State, and FQ Transient.

Explanation of the Reduction in Full-Power FAH The maximum FAH peaking factor considered in the LBLOCA analysis of record was 1.8. This exceeded the allowable operating value documented in the core operating limits report (COLR). A reduction of the FAH peaking factor to the plant operation limit of 1.7 (COLR limit) was readily available margin because the plant could not operate at the analyzed FAH peaking factor of 1.8. 'This peaking factor margin was pursued at the beginning of the TCD LOCA analysis to establish a lower baseline PCT from which to assess the potential impacts of TCD. The basis for this evaluation method is described in the Westinghouse letter to the NRC dated March 7, 2012 (Reference 1).

The core is designed and required by Technical Specifications to be operated within the COLR FAH limit of 1.7. Therefore, adjusting the FAH peaking factor to the COLR limit in the LBLOCA TCD evaluation reflects a reduction in the conservatism of the FAH design input value. The FAH adjustment from 1.8 to 1.7 more closely reflects current plant design and operational limits. The adjusted FAH input value bounds the reload core design and reload core operation.

Explanation of Peaking Factor Burndown Limits The effect of peaking factors is integral to the burnup of the fuel due to the decrease in maximum peaking that can be achieved by an assembly with increasing burnup. The evaluation of fuel pellet TCD considered continuous functions of peaking factors versus burnup, starting at the values for beginning-of-life (BOL). This approach was used to explicitly and appropriately consider the peaking factor limits as a function of burnup. In the as-approved evaluation model, the hot assembly and hot rod are assumed to be at BOL. The LBLOCA TCD evaluation explicitly accounted for hot rod and hot assembly peaking factor burndown and considered the full life of the fuel operation.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 3 of 11 The peaking factor burndown limits used in the LBLOCA TCD evaluation are provided in Table 3-1. Dominion reviewed the nuclear core design data for KPS fuel cycles 30, 31 (current at the time of the TCD evaluation), and 32 (current at the time of,this RAI response) and developed FAH and FQ Steady-State at 100% full power equilibrium conditions for all cycle burnups. FQ Transient was evaluated for Cycle 32 to confirm the relationship between FQ Transient and FQ Steady-State. The peaking factor burndown limits input to the TCD evaluation (Table 3-1) wereselected to be bounding with respect to the three cycles of design data.

Figure 3-1 compares FAH peaking factor limits (with and without uncertainty) to the cycle design data without uncertainty. The FAH was obtained with all-rods-out (ARO) and with D-Bank at the rod insertion limits (RILs). Figure 3-2 compares the FQ Steady-State peaking factor limit to the cycle design data without uncertainty at ARO conditions.

Figure 3-2 also shows the FQ Transient limit. The FQ Transient limit is reduced by an amount less than or equal to the reduction in the FQ Steady-State limit. The relationship between transient and steady-state FQ was evaluated for Cycle 32. FQ Transient decreases with local burnup by a larger amount than the decrease in FQ Steady-State at higher fuel burnups. The FQ Transient and the FQ Steady-State values are bounded by their respective limit curves. Thus, the FQ Steady-State limit and the FQ Transient limit are verified.

Figure 3-1 and Figure 3-2 show that the peaking factor limits used in the LBLOCA TCD analysis are conservative with respect to cycle design data. In addition, Dominion confirmed that the hot assembly average power burndown assumption in the LBLOCA TCD evaluation was bounding. As stated in DEK's letter dated March 15, 2012 (Reference 2), the peaking factor burndown limits in Table 3-1 were incorporated into the reload core design process (via the Reload Safety Analysis Checklist) and will continue to be verified for future KPS core designs to ensure that the LBLOCA TCD PCT estimate remains conservative.

Table 3-1 Peaking Factor Burndown Limits in the Evaluation of TCD Rod Burnup (MWD/ MTU) FQ Steady-State FQ Transient(') FAH (1)(2) 0 2.100 2.500 1.700 30,000 2.000 2.500 1.700 50,000 1.625 2.125 1.406 62,000 1.400 2.125 1.300 (1) Includes uncertainties (2) Hot assembly average power follows the same burndown, since it is a function of FAH.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 4 of 11 Figure 3-1: Kewaunee Cycles 30, 31 and 32 FAH vs Rod Burnup 1.800 I

1.600 1.400 wasfM OvW El

~7~TW~

.1 _________

-~~1 ---------------

1.200 Eu 1.000 EU a

0-800 *1 4-__-

0.600 L ...........

0.400 0.200 0.000 l I F -I-0 10000 20000 30000 40000 50000 60000 70000 MWd/MTU

. FDHARO w/o Uncertainty U FDHDin w/o Uncertainty = Fdh w/o Uncertainty -- "--Fdhwith Uncertainty Figure 3-2: Kewaunee Cycles 30, 31 and 32 ARO FQ Steady-State vs Local Burnup 3.000 2.500 2.000

-1.500 1.000 0.500 0.000 0 10000 20000 30000 40000 50000 60000 70000 MWd/MTU

• FQARO w/o Uncertainty -*- Fq(ss) w/o Uncertainty - Fq(Tr) with Uncertainty

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 5 of 11

References:

1. Letter from J. A. Gresham (Westinghouse) to USNRC Document Control Desk, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f)

Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," LTR-NRC-12-27, dated March 7, 2012.

2. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Response to Information Request Pursuant to 10 CFR 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse-Furnished Realistic Emergency Core Cooling System (ECCS)

Evaluation and 30-Day Report of ECCS Model Changes Pursuant to the Requirements of 10 CFR 50.46," dated March 15, 2012.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 6 of 11

7. (ME8460-RAII-SRXB-Parks-007-2012-06-08)

Please explain how the changed design values will be verified during operation of the plant, i.e. TS limits, Surveillances, etc. Also, explain what compensatory actions will be taken if a value is found to be outside of the limits assumed in the analysis.

Response to (ME8460-RAII-SRXB-Parks-007-2012-06-08)

The following Large Break LOCA (LBLOCA) design input changes with respect to plant operation were considered in the thermal conductivity degradation (TCD) evaluation for KPS:

• Reduction in full-power FAH (hot rod average power, including uncertainties) from 1.8 (analysis of record value) to 1.7 (current COLR limit); and,

" Corresponding reduction in hot assembly average power (including uncertainties).

Since the above change applies the full-power FAH of 1.7 from the COLR, no plant program changes were required. Exceeding the COLR limit of 1.7 would continue to require action as specified in the KPS Technical Specifications.

The following LBLOCA analysis assumptions were modified to offset the effects of fuel pellet TCD. (Note: The rod burnup = 0 MWD/MTU, beginning of life (BOL) FQ Steady-State Limit and FQ Transient Limit are unchanged from the current KPS LBLOCA analysis).

• Maximum Nuclear Enthalpy Rise Peaking Factor, FAH

• Maximum Relative Power in Hot Assembly, PBAR

• Maximum Steady-State Depletion Factor, FQ Steady-State

• Maximum Transient Total Peaking Factor, FQ Transient The modified limits from Reference 1 are repeated below in Table 7-1.

Table 7-1 Peaking Factor Burndown Limits in the Evaluation of TCD Rod Burnup (MWD/ MTU) FQ Steady-State 'FQ Transient(1 ) FAH (1)(2) 0 2.100 2.500 1.700 30,000 2.000 2.500 1.700 50,000 1.625 2.125 1.406 62,000 1.400 2.125 1.300 (1) Includes uncertainties (2) Hot assembly average power (PBAR) follows the same burndown, since it is a function of FAH.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 7 of 11 Verification of the LBLOCA design input changes and the LBLOCA limits specified in Table 7-1 is performed as part of the normal reload core design process for each reload core. Steady state FQ, hot assembly average power, and FAH limits are verified by comparing predicted values from full power conditions against the burnup dependent limits shown in Table 7-1. FQ Transient is verified by comparing power distributions produced during normal operation and from operational transients against the applicable limits specified in Table 7-1. Predicted power distributions used in the reload, analyses are based on core models developed using the NRC-approved nuclear design methodology described in Reference 2.

It was confirmed by cycle design analyses (see response to Question 3) that the peaking factor values with the least margin occur prior to 30,000 MWD/MTU. This margin is illustrated in Figures 3-1 and 3-2 for FAH and FQ Steady-State, respectively. The limiting peaking factors occur at a burnup before the effect of TCD becomes important, so the total amount of peaking factor margin is unaffected by these revised peaking factor limits.

Each reload cycle, the peaking factor limits are analytically confirmed and verified. If the analytical verification produces unacceptable results, then the core is either redesigned or re-analysis of the LBLOCA analysis is performed with revised input peaking factor assumptions. The acceptability of analysis results is based on confirming that the reactor core is operating as designed.

Power distribution surveillances are performed to verify the total peaking factor FQ. KPS Technical Specification (TS) 3.2.1 contains limits on FQ to preclude exceeding the initial linear power generation rate assumed in the LOCA accident analysis.. The FQ limit is verified by performing periodic measurements during the cycle in accordance with the Surveillance Requirements of TS 3.2.1. For relaxed power distribution control operation, FQ is approximated by a steady-state FQ and a transient FQ. The steady-state FQ is an approximation of FQ when the reactor is at the steady-state power at which the power distribution measurement was taken. The transient FQ accounts for power distribution transients that could be encountered during normal operation. Measured FQ is compared against the TS FQ steady-state and FQ transient operational limits. Required actions for the measured FQ exceeding either the steady-state or the transient operational limit are defined in TS and include actions to reduce reactor power level, axial flux difference (AFD) limits and reactor protection system (RPS) setpoints, depending upon the type (either steady-state or transient) and magnitude of the difference between measured FQ and operational limits.

The power distribution surveillances are also performed to verify the Nuclear Enthalpy Rise Hot Channel Factor FAH. TS 3.2.2 contains limits on FAH to preclude exceeding the initial FAH value assumed in the LOCA accident analysis. The FAH limit is verified by performance of periodic measurements during the cycle in accordance with the Surveillance Requirements of TS 3.2.2. Required actions for the measured FAH exceeding the FAH limit are defined in TS and include actions to reduce reactor power level and RPS setpoints.

S

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 8 of 11 Reactivity and power distribution measurements are performed periodically during the cycle as required by TS 3.1.2 (Core Reactivity), 3.2.1 (Heat Flux Hot Channel Factor FQ) and 3.2.2 (Nuclear Enthalpy Rise Hot Channel Factor FAH) to verify that core reactivity and peaking factors are within their respective design limits. Measured power distributions and core reactivity are also compared against predicted (predicted by the core design model) power distributions and core reactivity. These comparisons, when coupled with startup physics testing results following refueling, are used to verify the core design model 'and demonstrate the core is operating as designed. This confirmation provides confidence in the predictive capability of the core design model used to verify LBLOCA accident analysis input assumptions and the model's ability to predict core performance. If the core is determined to not be operating as designed, an evaluation would be performed to assess analysis margins, understand the reasons for the deviation, and make appropriate adjustments on a case-by-case basis to plant operations or setpoints to ensure operation within LBLOCA analysis limits.

In summary, hot assembly average power, transient FQ, FAH, and steady-state FQ limits are analytically confirmed for each reload core as part of the reload design process. If peaking factor assumptions in the LBLOCA analysis are exceeded, then either the reload core is redesigned, or the LBLOCA analysis is revised with new peaking factor assumptions. The acceptability of the peaking factors assumptions used in the LBLOCA analysis is confirmed by verifying the core is operating as designed.

This is accomplished through a startup physics test program following refueling and through periodic power distribution and reactivity measurements performed throughout the cycle. Additionally the power distribution surveillances, performed in accordance with TS 3.2.1 and TS 3.2.2, require that'if an FQ or FAH limit is exceeded compensatory actions (including reductions in reactor power level, AFD limits and RPS setpoints) be taken depending upon the type and magnitude of the discrepancy. These surveillances also provide confirmation that the core is operating as designed., Therefore, it is not necessary to modify the Technical Specification Surveillance Requirements to account for the burn down in the peaking factor limits. The reload design confirms that the predicted cycle will not violate any of the peaking factor limits and performance of TS surveillances serves to confirm the core is behaving as predicted, including confirmation of pre-burndown peaking factor limits.

Verification of the peaking factor burndown limits used in the LBLOCA TCD assessment will be performed as part of the KPS cycle design process. This design verification process is consistent with the confirmation of other KPS safety analysis parameters (e.g., Doppler temperature coefficient, differential rod worth, rod worth used in shutdown margin, and FAH fuel rod census) that are not confirmed by direct plant surveillance.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 9 of 11

References:

1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Response to Information Request Pursuant to 10 CFR 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse-Furnished Realistic Emergency Core Cooling System (ECCS) Evaluation and 30-Day Report of ECCS Model Changes Pursuant to the Requirements of 10 CFR 50.46," dated March 15, 2012.

2. Topical Report DOM-NAF-5, Revision 0.2-A, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," dated January 2011.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 10 of 11

8. (ME8460-RAII-SRXB-Parks-008-2012-06-08)

Page 3 of Attachment 2 to Serial 12-100 states that "Dominionand its vendor, Westinghouse Electric Company, LLC, utilize processes which ensure that LOCA analysis input values conservatively Pound the as-operatedplant values for those parameters." Please explain these processes.

Response to (ME8460-RAII-SRXB-Parks-008-2012-06-08)

Dominion and its vendor, Westinghouse Electric Company LLC, ensure the LOCA analysis input values conservatively bound the as-operated plant values for the relevant parameters via the fuel reload process. The purpose of the fuel reload process is to evaluate plant changes resulting from the loading of different or new fuel into the core.

The fuel reload process for KPS is described in Reference 1 (Westinghouse fuel reload process) and Reference 2 (Dominion fuel reload process). Both the Westinghouse and Dominion fuel reload processes are approved for application to KPS. The evaluations performed for the reload support a licensing approach under the regulations of 10 CFR 50.59. Safety Analyses generally analyze the relevant parameters in a bounding direction compared to the expected operational values. The generic fuel reload evaluation approach relies upon the bounding approach in which safety analyses are performed to accommodate the plant changes resulting from different or new fuel in the core without requiring new safety analyses.

As part of the reload evaluation process, the LOCA analyst generates a list of important parameters to the LBLOCA analysis, and which show a fuel reload dependency. The LOCA analyst then identifies the values of those parameters supported by the LBLOCA licensing basis analyses and evaluations. The parameters are confirmed to support the reload core design or are evaluated with respect to the LBLOCA analysis.

Separate from the fuel reload process, plant changes which may impact the LBLOCA analysis are identified to Westinghouse as needed, and 10 CFR 50.46 evaluations are performed as necessary. During the reload process, a summary of plant changes that have occurred since the previous cycle and changes planned for the upcoming cycle is provided by Dominion to Westinghouse. Westinghouse reviews these changes to ensure the non-reload related parameters analyzed in the LBLOCA analysis, and therefore the LBLOCA analysis, remain applicable. For example, the percentage of plugged steam generator tubes is one such non-reload related parameter reviewed as part of the reload analysis to ensure that the LBLOCA analysis remains applicable.

'Dominion also has system and component surveillance and testing processes (e.g.,

safety injection system flow testing) that apply to the LOCA analysis. The plant surveillance and testing processes are used to confirm the LOCA analysis remains applicable and ensure the LOCA analysis input values conservatively bound the as-operated plant values.

Serial No.12-352 Attachment 1 Response to Questions 3, 7, and 8 Page 11 of 11

References:

1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

2. Topical Report, DOM-NAF-5, Rev. 0.2-A, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to Kewaunee Power Station (KPS)," January 2011.

Serial No.12-352 Attachment 3 Westinghouse letter LTR-LIS-12-234 NP Page 1 of 23 ATTACHMENT 3 NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 NRC QUESTIONS 1. 2. 4-6. and 9 AND DEK RESPONSES (23 pages including this cover page)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

1. (ME8460-RAII-SRXB-Parks-001-2012-05-23)

This item is deleted.

2. (ME8460-RAII-SRXB-Parks-002-2012-06-08)

Justify the evaluation of reduced peaking factors at beginning-of-life conditions to obtain analytic margin to offset the TCD effect. Show that peaking factor reductionsaffect PCT in a manner that is substantiallyindependent of fuel burnup.

Response to (ME8460-RAII-SRXB-Parks-002-2012-06-08)

The maximum FAH peaking factor of 1.8 considered in the Large Break LOCA (LBLOCA) analysis exceeded the allowable operating value of 1.7 documented in the core operating limits report (COLR). As such, a reduction of this peaking factor to the plant operation limit was readily available margin to offset the impact of fuel pellet thermal conductivity degradation (TCD) since the plant. could not operate at the analyzed power peaking.

The effect of peaking factors is integral to the burnup of the fuel, due to the decrease in maximum peaking that can be achieved by an assembly with increasing burnup. The analysis of record (AOR) was first rebaselined to the reduced maximum FAH peaking factor utilized in the evaluation. The evaluation of fuel pellet TCD then considered continuous functions of peaking factors versus burnup, starting at the reduced FAH value and analyzed FQ values for beginning-of-life (BOL). This approach was used to explicitly and appropriately consider the peaking factor limits as a function of burnup, rather than treating the peaking factor reductions independently of burnup.

3. (ME8460-RAII-SRXB-Parks-003-2012-06-08)

Fully explain all peaking factor adjustments and provide the rationalefor each adjustment.

Response to (ME8460-RAII-SRXB-Parks-003-2012-06-08)

Dominion Responsibility

4. (ME8460-RAII-SRXB-Parks-004-2012-05-23)

This item is deleted.

NP-2

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

5. (ME8460-RAII-SRXB-Parks-005-2012-06-08)

Your submittal referenced a March 7, 2012 letter sent by Westinghouse Electric Company to the NRC.

a) The final paragraphon Page 2 of 9 refers to small differences in fuel characteristics that were claimed to be compared. The paragraphalso discusses confirmatory evaluations concluding that other operatingcharacteristicswere acceptable.Provide the results of this comparison for Kewaunee, including the relevant conclusions and the technical basis supporting those conclusions. Forany conclusion that differences in a particularfuel or operating characteristicare offset by other conservatisms, list those conservatisms and provide a quantitative estimate of each conservatism, as well as a brief descriptionof the rigorassociatedwith that estimate.

Response to (ME8460-RAII-SRXB-Parks-005-2012-06-08), Part a)

The key fuel parameters used for fuel temperature analyses were compared to a TCD analysis of a representative rod type.

]ac The specifics of the comparison for Kewaunee are as follows:

NP-3

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Table 5-1: Comparison of Plant C and Kewaunee PAD Data Parameters a,c a,c NP-4

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

5. (ME8460-RAII-SRXB-Parks-005-2012-06-08)

Your submittal referenced a March 7, 2012 letter sent by Westinghouse Electric Company to the NRC.

b) Please provide the values for the coefficients used in the PAD 4.0+TCD U02 thermal conductivity equation.

Response to (ME8460-RAII-SRXB-Parks-005-2012-06-08), Part b)

The functional form used to model TCD 1a,c is as follows:[

a,c NP-5

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

5. (ME8460-RAII-SRXB-Parks-005-2012-06-08)

Your submittal referenced a March 7, 2012 letter sent by Westinghouse Electric Company to the NRC.

c) Please explain any errorcorrections,code improvements, and miscellaneouscode cleanup between the WCOBRA/TRAC and HOTSPOT code versions used in the TCD evaluations and those used in the plant's AOR.

d) What is the thermal conductivity model impact of code version changes in HOTSPOT?

Response to (ME8460-RAII-SRXB-Parks-005-2012-06-08), Parts c) and d)

The error corrections, code improvements, and miscellaneous code cleanup between the WCOBRA/TRAC and HOTSPOT code versions used in the analysis-of-record versus the evaluation of fuel pellet thermal conductivity degradation (TCD) are described in the 10 CFR 50.46 reporting pages enclosed with this response (Attachment 4)..

The addition of a fuel conductivity model appropriate for the TCD evaluations was incorporated into WCOBRA/TRAC and HOTSPOT as discussed in LTR-NRC-12-27 (Reference 1). The error corrections and code improvements referenced in the prior paragraph do not impact the thermal conductivity model. It is more appropriate to estimate the effect of TCD using code versions with these changes because the impact of TCD on the PCT may be affected by the corrections in the updated code versions (for example the fuel relocation model correction in HOTSPOT).

Reference

1. LTR-NRC-12-27, "Westinghouse Input Supporting'Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.

NP-6

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

5. (ME8460-RAII-SRXB-Parks-005-2012-06-08)

Your submittal referenced a March 7, 2012 letter sent by Westinghouse Electric Company to the NRC.

e) Explain the differences between the HOTSPOT and PAD thermal conductivity models and the impact of those differences. Provide graphs or other quantified descriptionsthat aid in explanation.

Response to (ME8460-RAII-SRXB-Parks-005-2012-06-08), Part e)

For the fuel thermal conductivity degradation (TCD) evaluation, PAD 4.0 TCD was used to generate the initial maximum fuel average temperature input into WCOBRA/TRAC and HOTSPOT. The PAD 4.0 TCD fuel thermal conductivity equation, for fuel at a nominal density of 95% theoretical density is given in LTR-NRC-12-27 (Reference 1) with the coefficients provided in response to Part b) of this request for additional information (RAI) and repeated below.

a,c For the TCD evaluation, WCOBRA/TRAC and HOTSPOT used a fuel thermal conductivity model based on [

]ac For fuel at a nominal density of 95% theoretical density, the model in WCOBRA/TRAC and HOTSPOT is given in LTR-NRC-12-27 (Reference 1) and repeated below.

[

ac The functional form and units between the two models are different. For ease of comparison, the degradation terms (f(Bu) in both equations) are compared in Figure 5-1 NP-7

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment at burnups of 20, 40 and 65 GWD/MTU. As seen from Figure 5-1, a,c Figures 5-2 through 5-5 compare the overall fuel thermal conductivity models at burnups of 0, 20, 40 and 65 GWD/MTU, respectively. Also included in the figures is a comparison with the FRAPCON 3.4 thermal conductivity model (Reference 2). As seen from the figures, a,c For a given maximum fuel average temperature and burnup, the differences between the PAD 4.0 TCD and WCOBRA/TRAC and HOTSPOT fuel thermal conductivity models a,c References

1. LTR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.

2. NUREG/CR-7022, Vol. 1 / PNNL-1 9418, Vol.1, "FRAPCON-3.4: A Computer Code for the Calculation of Steady-State Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup," March 2011.

3. WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 - 5 (Revision 1), "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.

(Proprietary)

4. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.

(Proprietary)

NP-8

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Figure 5-1: Fuel Thermal Conductivity Degradation Model Comparison a,c Figure 5-2: Fuel Thermal Conductivity Model Comparisons - 0 GWD/MTU -

a,c NP-9

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Figure 5-3: Fuel Thermal Conductivity Model Comparisons - 20 GWDIMTU a,c Figure 5-4: Fuel Thermal Conductivity Model Comparisons -40 GWD/MTU a,c Figure 5-5: Fuel Thermal Conductivity Model Comparisons - 65 GWDIMTU a,c NP-10

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

6. (ME8460-RAII-SRXB-Parks-006-2012-06-08)

Pleaseprovide additionaldetails concerning the steady-state ASTRUM/CQD initializationprocess. In particular,please explain what fuel characteristicsare adjusted within the ASTRUM/CQD models to obtain convergence and agreement among HOTSPOT, WCOBRA/TRAC, and PAD4.OTCD.

Response to (ME8460-RAII-SRXB-Parks-006-2012-06-08)

The following parameters in WCOBRA/TRAC are used to determine steady-state convergence, as discussed in Section 20-5 of WCAP-1 2945-P-A (Reference 1) and Section 12-4-1 of WCAP-1 6009-P-A (Reference 2).

[

]a,c NP-11

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment a,c NP-12

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment a,c References

1. WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 - 5 (Revision 1), "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.

(Proprietary)

2. WCAP-1 6009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.

(Proprietary)

NP-13

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Table 6-1: Initial Gap Thickness and Average Fuel Temperature Comparison for Sample 17x17 Plant ac Table 6-2: Initial Gap Thickness and Average Fuel Temperature Comparison for Sample 15x15 Plant aC NP-14

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Table 6-3: HOTSPOT and WCOBRA/TRAC Steady-State Gap Heat Transfer Coefficient and Average Fuel Temperature Comparison for Sample 17x17 Plant a7C Table 6-4: HOTSPOT and WCOBRAITRAC Steady-State Gap Heat Transfer Coefficient and Average Fuel Temperature Comparison for Sample 15x15 Plant a,

NP-15

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Figure 6-1: WCOBRAITRAC and HOTSPOT Cladding Temperature Comparison for 17x17 Plant NP-16

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Figure 6-2: WCOBRA/TRAC and HOTSPOT Cladding Temperature Comparison for 15x15 Plant

,a,c NP-17

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

7. (ME8460-RAII-SRXB-Parks-007-2012-06-08)

Please explain how the changed design values will be verified during operation of the plant, Le. TS limits, Surveillances, etc. Also, explain what compensatory actions will be taken if a value is found to be outside of the limits assumed in the analysis.

Response to (ME8460-RAII-SRXB-Parks-007-2012-06-08)

Dominion Responsibility NP-18

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

8. (ME8460-RAII-SRXB-Parks-008-2012-06-08)

Page 3 of Attachment 2 to Serial 12-100 states that "Dominionand its vendor, Westinghouse Electric Company, LLC, utilize processes which ensure that LOCA analysis input values conservatively bound the as-operatedplant values for those parameters."Please explain these processes.

Response to (ME8460-RAII-SRXB-Parks-008-2012-06-08)

Dominion Responsibility.

NP-19

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

9. (ME8460-RAII-DORL-Paige-009-2012-06-08)

Based on the NRC's review of the March 15, 2012, submittal it appears that the licensee has revised inputs to a method of evaluation as describedin the FSAR (as updated) used in establishingthe design bases or in the safety analyses.

Revision 1 to NEI 96-07, "Guidelinesfor 10 CFR 50.59 Implementation," Section 3.8, "InputParameters,"provides clarifying information concerning whether an input parameteris considered to be an element of a methodology for the purposes of addressingthe applicable requirements found at 10 CFR 50.59, "Changes,Tests, and Experiments." Address whether the methodology permits the licensee to establish how to select the value of an input parameterto yield adequately conservative results and whether the revised value is more conservative than that requiredby the selection method.

Address whether any of the changes (i.e., to the U02 thermal conductivity equation) constitutes a change in the calculationalframework used for evaluating behavioror response of a system, structure or component. Explain whether and how 10 CFR 50.59(c)(4) might apply to such a change.

Response to (ME8460,RAII-DORL-Paige-009-2012-06-08)

Westinghouse currently employs three best estimate Evaluation Model (EM) methodologies for analysis of the large break loss-of-coolant accidents (LBLOCA) in pressurized water reactors (PWRs):

• 1996 Westinghouse Best Estimate LBLOCA Evaluation Model (Code Qualification Document (CQD) EM, Reference 1)

• 1999 Westinghouse Best Estimate LBLOCA Evaluation Model, Application to PWRs (Pressurized Water Reactors) with Upper Plenum Injection (CQD-UPI EM, Reference 2)

• 2004 Westinghouse Realistic LBLOCA Evaluation Model using ASTRUM (Automated Statistical Treatment of Uncertainty Method) (ASTRUM EM, Reference 3)

As described in Reference 7, the best estimate Evaluation Model (EM) methodologies for analysis of the large break loss-of-coolant accidents applicable to Kewaunee Power Station (KPS) is the LBLOCA EM described in References 1 and 2. In application of a Westinghouse best estimate large break LOCA methodology to a plant analysis, Westinghouse works with the licensee to establish several parameter values input to the specific analysis per the Nuclear Regulatory Commission (NRC) - approved evaluation model requirements (including applicability restrictions specified by the NRC in their Safety Evaluation Reports (SERs)). The licensee is permitted to establish the values of these parameters on the basis of plant-specific considerations; as such they are input to the methodology and not part of the methodology, as defined in NEI 96-07 Revision 1 (Reference 6) Section 3.8. The input parameter values may be selected conservatively NP-20

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment in order to support current plant operation, as well as accommodate expected future changes or otherwise at the discretion of the licensee. Table 9-1 summarizes the selected design input changes evaluated in conjunction with the execution of the thermal conductivity degradation (TCD) evaluation(s) performed as described in the Reference 7 submittal, and relevant governing topical report references identifying how these values are to be selected.

In the evaluations of design input changes performed as described in the Reference 7 submittal, the changes to design input values were made to more closely represent current plant operation. Selection of the revised input parameter values was made in accordance with the approved EM. Therefore, the design input changes reflect reduction in the conservatism of these values and are considered an input parameter change and not a change to the methodology, consistent with Reference 6 Section, 3.8.

Westinghouse and its licensees utilize processes which ensure that the LBLOCA analysis input values conservatively bound the as-operated plant values for these parameters.

Fuel pellet TCD and peaking factor burndown' were not explicitly considered in the as-approved Westinghouse best estimate LBLOCA EMs. In order to evaluate the PCT effect of TCD and peaking factor burndown as described in the Reference 7 submittal, evaluation techniques were used that are outside of the as-approved EMs. This was necessary to explicitly consider the fuel performance effects of TCD, and to adequately evaluate the burnup-dependent aspects of the fuel performance changes considering TCD. Specifically, the following aspects of the TCD evaluation(s) were outside of the as-approved best estimate LBLOCA EM:

[

]aC NP-21

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment

[

a,c 10 CFR 50.46 establishes criteria for reporting and for action regarding changes or errors involving methods for loss of coolant analyses. For the evaluation and reporting of PCT impact, the changes to the LBLOCA EM to explicitly consider the fuel performance effects of TCD and to adequately evaluate the burnup-dependent aspects of the fuel performance are governed by 10 CFR 50.46. Consistent with 10 CFR 50.59(c)(4) and Reference 6 Section 4.1.1, the provisions of 10 CFR 50.59 do not apply for the LBLOCA EM changes for evaluations and reporting of PCT impact because the 10 CFR 50.46 regulation establishes more specific criteria for reporting and for action for changes involving methods for loss of coolant accidents.

In summary, in the evaluations of TCD and design input changes as described in the Reference 7 submittal, two types of changes were made:

• Design input values were changed to more closely represent plant operation, or analysis input changes were made to reduce conservatism in as-analyzed values.

The licensee is permitted to establish the value of these parameters on the basis of plant-specific considerations; as such these are changes to the input of the methodology and are not part of the methodology. Therefore, the design input changes reflect reduction in the conservatism of these values and are considered an input parameter change and not a change to the methodology.

" Techniques to appropriately account for the burnup-dependent effects of TCD were used in the evaluation(s) which are outside of the as-approved EMs. These changes to the calculational framework (as defined in 10 CFR 50.46(c)(2)) were required to assess the TCD phenomena which are not explicitly accounted for in the as-approved EMs. The provisions of 10 CFR 50.59 do not apply for the LBLOCA EM changes for evaluations and reporting of PCT impact because the 10 CFR 50.46 regulation establishes more specific criteria for reporting and for action for changes involving methods for loss of coolant accidents.

NP-22

Westinghouse Non-Proprietary Class 3 LTR-LIS-12-324 NP-Attachment Table 9-1: Applicable Evaluation Model Reference(s) for Selection of the Design Input Parameters Modified in TCD Evaluation for Kewaunee Design Input Change Relevant Section(s) of UPI -CQD Topical Report (Reference 2)

Specification of Section 5-2-1 peaking factors Section 5-2-2 Table 5-1

References:

1. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2-5, Revision 1 (Proprietary), Bajorek, S. M. et al., Code QualificationDocument for Best Estimate LOCA Analysis, March 1998.

2. WCAP-14449-P-A, Revision 1 (Proprietary), Dederer, S. I., Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection, October 1999.

3. WCAP-16009-P-A Revision 0 (Proprietary), Frepoli, C., et al., Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of UncertaintyMethod (ASTRUM), January 2005.

4. LTR-NRC-12-27, Letter from J. A. Gresham (Westinghouse) to NRC, Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(o Letter Regarding Nuclear Fuel Thermal Conductivity Degradation,(Proprietary/Non-Proprietary),

March 7, 2012.

5. LTR-NRC-01-6, Letter from H. A. Sepp (Westinghouse) to J. S. Wermiel (NRC), U.S.

Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2000, March 13, 2001.

6. NEI 96-07 Revision 1, Guidelines for 10 CFR 50.59 Implementation, November 2000.

7. Letter from A. J. Jordan (DEK) to Document Control Desk (NRC), "Response to Information Request Pursuant to 10 CFR 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Core Cooling System (ECCS)

Evaluation and 30-Day Report of ECCS Model Changes Pursuant to the Requirements of 10 CFR 50.46," dated March 15, 2012.

NP-23

'Serial No.12-352 Attachment 4 10 CFR 50.46 Reporting Pages ATTACHMENT 4 NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 10 CFR 50.46 REPORTING PAGES THAT SUPPORT THE RESPONSE TO NRC QUESTION 5 (23 pages including this cover page)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Source: NSD-NRC-98-5575 Date: April 8, 1998 TUBE HEATED CONDUCTOR ERROR

Background:

WCOBRAITRAC allows metal structures to be modeled as either a "Heated Conductor" in which axial conduction is calculated, or as an "Unheated Conductor" in which axial conduction is assumed to be relatively unimportant. The geometry of either Conductor can be a "WALL", a "TUBE", or a "ROD". In PWR models, Heated Conductors with a ROD geometry are used for fuel rods only. Other metal structures are modeled using Unheated Conductor types. It has been discovered that no heat is transferred to the inside Channel of a Heated Conductor if it is modeled with a TUBE geometry.

This was determined to be a Non-discretionary change as described in Section 4.1.2 of WCAP- 1345 1.

Affected Evaluation Models SECY UPI WCOBRAJTRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect No SECY UPI or Best Estimate WCOBRA/TRAC calculation models heated conductors as TUBE geometry. This error does not occur for Unheated Conductors using the "TUBE" geometry type. Therefore, no estimated PCT effect is required to be assessed.

1997

Attachment 1 Our ref: LTR-NRC-02-10 March 13, 2002 Oxidation Thickness Index Error For Best Estimate WCOBRAITRAC

Background

A coding error has been identified in the initial outside oxidation thickness array used for fuel rods. The error was an incorrect index for storage of the oxide thickness for each fuel rod. Coding used the rod number index instead of the rod type index. This issue was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect The error was found to have no effect for standard BELOCA analyses that follow the published guidance material for input of this variable. The error also did not affect any test simulations performed to support the licensing of the BE Evaluation Model. Thus, there was found to be no instance of use of erroneous oxidation thickness and there is no PCT impact for this error. The error will be corrected during the next revision of the Best Estimate WCOBRA/TRAC code.

Official recordelectronically approved in EDMS 2000 .Page 8of 16 A BNFL Group company

Attachment 1 Our ref: LTR-NRC-02-10 March 13, 2002 Neutronics Calculation Moderator Density Weighting Factor Error

Background

An error was discovered in WCOBRA/TRAC whereby power used in normalization of moderator density weighting factors was double-accounted for channels with multiple simulated rods. The error biases the average moderator density to be slightly higher, resulting in slightly higher power generation in the hot rod. The error is qualitatively conservative, however, quantitatively insignificant. This issue was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP- 13451.

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect At the beginning of the transient calculation, the difference in weighted density is less than 1% for all plant types. This difference is similar to the density difference between (2250 psia, 586°F) and (2250 psia, 588.8'F) thermodynamic state points. The difference in average moderator density affects the reactivity. The difference in reactivity at the beginning of the transient is negligible. As the transient progresses, with voiding of the core, the strong negative reactivity dominates. Therefore, it was estimated thit the error has 00 F PCT impact on plant calculations. The error will be corrected during the next revision of the Best Estimate WCOBRA/TRAC code.

Official recordelectronically approvedin EDMS 2000 Page 9 of 16 A BNFL Group company

Attachment 1 - Standard Text Page 16 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 1-D MINIMUM FILM BOILING TEMPERATURE MODEL SELECTION ERROR

Background

Section 6-3-6 of WCAP-12945-P-A indicates that the minimum film boiling temperature calculation for 1-D components is calculated as the maximum of the homogeneous nucleation temperature and that predicted by the Iloeje correlation. The comparison of these two correlations is made if a flag (ITMIN) is set greater than zero. Otherwise, the homogeneous nucleation temperature is used. It was found that ITMIN was not initialized, resulting in the Iloeje correlation not being considered. This error has the potential to affect the heat transfer calculations in the steam generator tubes of the STGEN component.

The coding was corrected to be consistent with the description in Section 6-3-6. This coding change was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluatiori Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect The homogeneous nucleation temperature exceeds the minimum film boiling temperature predicted by the Iloeje correlation for pressures less than about 100 psia. Therefore, this error could only potentially have an effect until the system pressure drops below about 100 psia, which typically occurs within 20-30 seconds. Examination of a typical PWR transient indicated-that the transition boiling regime occurs in the steam generator tubes for only a few seconds during blowdown. Given the short period of time in the transition boiling regime, and relatively small difference between the homogeneous nucleation temperature and the Iloeje correlation results during this time period, it is concluded that the effect of the error is small enough to be considered negligible. Therefore, the estimated effect of this error correction is 0°F.

A BNFL Group company

Attachment 1 - Standard Text Page 17 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 1-D CONDENSATION RAMP ERROR Background /

Section 5-3-5 of WCAP-12945-P-A indicates that condensation in specified one-dimensional components is suppressed if the pressure drops significantly below the containment pressure, using Equation 5-95a.

This ramp was erroneously applied to the interfacial heat transfer for superheated liquid, affecting the evaporation process as well as the condensation due to subcooled liquid. The coding has been corrected so that it is applied to condensation conditions only. This coding change was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP- 1345 1.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect Superheated liquid is not expected to be present in the affected components for any significant portion of a large break LOCA. A sensitivity study was performed using a PWR model in which the condensation ramp was applied. It was confirmed that the effect of the error correction on the peak cladding temperature was negligible. Therefore, the estimated effect of this error correction is 00F.

A BNFL Group company

Attachment 1 - Standard Text Page 18 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 CLADDING AXIAL THERMAL EXPANSION ERROR

Background

The cladding axial thermal expansion enters into the calculation of the fuel rod internal pressure, via the time-dependent gas plenum volume (Equation 7-46 of WCAP-12945-P-A). Equation 7-39 shows how the cladding axial thermal expansion over the length of the rod is calculated. Table 7-1 shows that the cladding axial thermal expansion is based on a linear interpolation scheme over a temperature range of 1073-1273'K. The CALL statement for the interpolation subroutine had a typographical error in one of the arguments, such that the axial thermal expansion was evaluated incorrectly. The error was corrected.

This coding change was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect Rod internal pressures vary on the order of several hundred psi prior to burst, primarily as a result of changes in the temperatures of the various gas volumes (plenum, pellet-clad gap, effective porosity, etc.).

Correction of the cladding axial thermal expansion error affects the rod internal pressure transient by only a few psi. This change is considered negligible, and the estimated effect on plant calculations is 0°F.

A BNFL Group company

Attachment 1 - Standard Text Page 19 of 25 Our ref LTR-NRC-03-5 March 7, 2003 ERROR IN TIME AFTER SHUTDOWN FOR NEUTRON CAPTURE TERM

Background

Equation 8-45 of WCAP-12945-P-A shows the neutron capture correction factor specified by the ANSI/ANS. 5.1-1979 standard. The time after shutdown term, t, was incorrectly programmed to use the total calculation time, including the steady state calculation. The coding has been corrected so that it is defined as the time after initiation of the break. This coding change was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP- 13451. (Note that for the SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model, this change affects only the superbounded analysis. The Appendix K analysis is unaffected.)

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect The neutron capture correction factor is a multiplier slightly larger than unity, which increases with time after shutdown. The error resulted in a longer time after shutdown, which is slightly conservative. The effect of the error correction was estimated by evaluating Equation 8-45 of WCAP- 12945-P-A, using typical analysis values. The results indicated that the G multiplier is reduced by about 0.4% with the correction, which would cause the total decay heat energy to be reduced by about 0.4%. This change is considered negligible, and the estimated effect on plant calculations is 0°F.

,I, A BNFL Group company

Attachment 1 - Standard Text Page 22 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 USER CONVENIENCES IN HOTSPOT

Background

The HOTSPOT code is used to quantify the propagation of local model uncertainties in the Westinghouse best estimate LOCA methodologies. HOTSPOT was updated as part of normal process improvement initiatives. The coding changes included increased array size to support longer transients, running the burst case automatically at the exact burst location, and consolidation of the advanced plant and gap re-opening versions with the production version. These changes were determined to be Discretionary changes in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect Test cases with longer transient run times supported the conclusion that the prior version was adequate.

Therefore, there is no PCT impact for this change.

The automation of the burst node calculation is a process improvement that reduces analyst resource requirements. There is no PCT impact for this change, as the capability previously existed for the analyst to specify the exact burst location in a separate run.

Consolidation of the advanced plant and gap re-opening versions with the production version eliminates the need to maintain single application code versions. There is no PCT effect on design basis analyses.

Based on the above, no licensees are affected by these changes, and 50.46 reporting is not required.

A BNFL Group company

Attachment 1 - Standard Text Page 23 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 POTENTIAL DIVIDE BY ZERO ERROR DURING PUMP ROTATION REVERSAL

Background

While modeling a pump suction leg break, it was discovered that a divide by zero can occur if the pump speed goes to zero during the reversal. Logic was added to branch to the reverse flow coding if the speed' is zero. Cold leg breaks, in which the flow is always forward, are considered in design basis analyses.

Therefore, this coding change was determined to be a Discretionary change in accordance with Section 4.1.1 of WCAP-1345 1.

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect Design basis analyses are performed for the most limiting break location, which is in the cold leg, between the pump and the vessel inlet nozzle. The rotation of the pump is in forward direction for cold leg breaks, such that the rotation is not reversed. Therefore, this error has no effect on PWR large break LOCA design basis analyses, and 50.46 reporting is not required.

A BNFL Group company

Attachment 1 - Standard Text Page 24 of 25-Our ref: LTR-NRC-03-5 March 7, 2003 APPLICATION OF DECAY HEAT UNCERTAINTY TO PROMPT FISSION ENERGY ERROR

Background

WCOBRA/TRAC contains an option to apply a built-in decay heat uncertainty based on the ANSI/ANS 5.1-1979 Standard. Use of this option resulted in the application of the uncertainty to the prompt fission energy in addition to the decay heat energy.

For the SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model, the 1979 decay heat model is used only in the so-called superbounded analysis, which serves as a validation of the analysis with Appendix K required features. The Appendix K analysis determines the licensing basis PCT. The current coding will be retained as an accepted conservatism in the superbounded calculations, and not considered an error.

The built-in decay heat uncertainty option is not used in the current Westinghouse Best Estimate Large Break LOCA Evaluation Models (1996 and 1999 versions). However, it will be used in a future methodology improvement. Therefore, this coding change was determined to be a Discretionary change in accordance with Section 4.1.1 of WCAP- 13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect As'noted above, the built-in decay heat uncertainty option is not used in the affected evaluation models.

Therefore, this error has no effect on PWR large break LOCA design basis analyses, and 50.46 reporting is not required.

A BNFL Group company

Attachment 1 - Standard Text Page 25 of 25 Our ref: LTR-NRC-03-5 March 7, 2003 BYPASS OF ORIFICE ENTRAINMENT MODEL IN DOWNFLOW WITH CHANNEL SPLITTING

Background

Entrainment during downward flow is calculated as described in Section 4-6-4 of WCAP-12945-P-A. An orifice entrainment model is used if the void fraction is greater than 0.8, and if there is an area expansion of greater than five percent in the downflow direction. There was a coding error.that would result in the orifice entrainment model being bypassed if there was channel splitting (one channel above two or more channels below). This error was corrected.

A review of the nodalization used in PWR analyses and test simulations indicated that only the G-2 test predictions were potentially affected by this error. The G-2 test predictions were not used to establish any of the uncertainty distributions used in the methodology. Therefore, this coding change was determined to be a Discretionary change in accordance with Section 4.1.1 of WCAP- 13451.

Affected Evaluation Models SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect As discussed above, the nodalization used in PWR analyses and the test simulations used to.establish code and model uncertainties precluded this error from occurring. Therefore, this error has no effect on PWR large break LOCA design basis analyses, and 50.46 reporting is not required.

A BNFL Group company

Attachment 1: Standard Text Page 14 of 22 Our ref:, LTR-NRC-04-17 March 25, 2004 INPUT ERROR RESULTING IN INCOMPLETE SOLUTION MATRIX

Background

Input parameter MSIM identifies the last cell number in each simultaneous solution group for the 3-D vessel component. A survey of WCOBRA/TRAC input decks identified two plant models and one test simulation model in which the MSIM input value was less than the total number of cells in the vessel.

This resulted in an incomplete solution matrix. An input diagnostic check has been added to prevent future occurrences. This input correction was determined to be a Non-Discretionary change in accordance with Section 4.1.2 of WCAP- 13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect Plant specific calculations were performed to estimate the PCT effect of this error for the two analyses affected. It was confirmed that the fundamental LOCA transient characteristics (e.g., blowdown cooling and reflood turnaround timing and behaviors) were unchanged by the error correction. The reference double-ended guillotine break was used to develop the PCT assessments for each plant.

The test simulation model affected by this error was also corrected, and the transient calculation repeated.

It was found that the error correction had no significant effect on the calculation results, and the prior validation conclusions remain valid.

A BNFL Group company

Attachment 1: Standard Text Page 21 of 22 Our ref: LTR-NRC-041-17 March 25, 2004 IMPLEMENTATION OF AUTOMATED STEADY STATE AND RESTART

Background

Westinghouse has submitted a revised treatment of uncertainties for its Large Break LOCA evaluation models, for NRC review and approval. The Automated Statistical Treatment of Uncertainties Methodology is described in WCAP-16009-P. As part of the implementation; of the revised methodology, enhancements were introduced that help to automateconvergence of the steady state solution to the desired set of conditions, as well as automating the restart process for beginning the LOCA transient.

These changes were determined to be Discretionary changes in accordance with Section 4. 1.1 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect These changes are for forward-fit automation purposes only, and have no effect on existing analyses.

A BNFL Group company

Attachment 1: Standard Text Page 22 of 22 Our ref: LTR-NRC-04-17 March 25, 2004 GENERAL CODE MAINTENANCE (BEST ESTIMATE)

Background

A number of coding changes were made as part of normal code maintenance. These include improvements in user flexibility for non-standard (non-design basis) analyses, and enhancements in the information available via output edits or for plotting purposes. All of these changes are considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-i 3451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect None of these changes affect the results of design basis analyses. Therefore, the estimated effect is zero.

A BNFL Group company

Attachment 1 - Standard Text Our ref: LTR-NRC-05-20 April 11, 2005 REVISED BLOWDOWN HEATUP UNCERTAINTY DISTRIBUTION

Background

Correction of modeling inconsistencies and input errors in the LOFT input decks have resulted in a change in the predicted peak cladding temperature transients. Revised analyses of the LOFT and ORNL tests were performed using the current version of WCOBRA/TRAC. As a result of this re-analysis, revised blowdown heatup heat transfer coefficients were developed and the revised cumulative distribution function (CDF) was programmed into a new version of HOTSPOT. The revised CDF was previously reported to the NRC in LTR-NRC-04-1 1. The overall code uncertainty for blowdown was also recalculated and programmed into a new version of MONTECF. The overall code uncertainty for reflood was not affected. These corrections were determined to be Non-Discretionary changes in accordance with Section 4.1.2 of WCAP-1 3451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect An estimate of the PCT effect of the revised blowdown heatup CDF was performed for the 1996 and 1999 Evaluation Models by calculating the impact on the reference transient for representative 2-, 3- and 4-loop plants. The estimates bound all of the 95th percentile HOTSPOT results. Estimates of the effect of the revised overall code uncertainty for blowdown were made on a plant-specific basis by repeating the MONTECF analysis, for those plants that track the blowdown period.

The revised blowdown heatup heat transfer multipliers have been and will be used for all analyses based on the 2004 ASTRUM Evaluation Model. Therefore, no PCT assessments are necessary for those plants.

A BNFL Group company

Attachment I - Standard Text Our ref: LTR-NRC-05-20 April 11, 2005 IMPLEMENTATION OF ASTRUM CAPABILITY IN HOTSPOT Back-ground The HOTSPOT code was modified to be compatible with the Automated Statistical Treatment of Uncertainty Methodology (ASTRUM, described in WCAP-16009-P-A). An option is used to trigger the ASTRUM HOTSPOT technique (single iteration mode) or the Monte Carlo mode used in the previous Best Estimate Large Break LOCA evaluation models. These changes were considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-1 3451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect None of these changes affect the results of design basis analyses performed with these evaluation models. Therefore, the estimated effect is zero.

A BNFL Group company

Attachment I - Standard Text Our ref: LTR-NRC-05-20 April 11, 2005 GENERAL CODE MAINTENANCE (WC/T)

Background

A number of coding changes were made as part of normal code maintenance.

Examples include correction of debug plots not used in design analyses, and improved consistency between the HOTSPOT nominal PCT (not used in the uncertainty analysis) and WCOBRAITRAC PCT. All of these changes are considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-1 3451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect None of these changes affect the results of design basis analyses. Therefore, the estimated effect is zero.

A BNFL Group company

Attachment I - Standard Text Our ref, LTR-NRC-06-8 March 16,2006 Page 14 of 21 REVISED ITERATION ALGORITHM FOR CALCULATING THE AVERAGE FUEL TEMPERATURE (Discretionary Change)

Background

Under certain conditions, the iteration scheme to calculate an average fuel temperature in HOTSPOT converged slowly, exceeding the maximum iteration count. This led to an average fuel temperature calculation that was inconsistent with the WCOBRAITRAC temperature for calculating the stored energy in the fuel. A revised iteration scheme, based on a combination of a secant method and a parabolic interpolation with a bracketing scheme, was implemented to resolve the non-convergence issue. This change is considered to be a Discretionary change in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The prior inconsistencies between the WCOBRA/TRAC temperature and the HOTSPOT average fuel temperature always resulted in a higher HOTSPOT average fuel temperature. Therefore, a 0'F impact is conservatively assigned for 10 CFR 50.46 reporting purposes.

A BNFL Group company

Attachment 1 - Standard Text Our ref: LTR-NRC-06-8 March 16, 2006 Page 15 of 21 PELLET RADIAL PROFILE OPTION (Discretionary Change)

Backgqround The radial power profile of fuel pellets was previously assumed to be uniform when setting up the conduction network over the fuel pellet in HOTSPOT. However, the accuracy of this approximation decreases for highly burned fuel since the radial power profile tends to increase from the center towards the outside of the fuel pellet at higher burnups. As such, an option was added in HOTSPOT to use a non-uniform radial power profile consistent with the WCOBRA/TRAC code. These changes were considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect This change is for forward-fit purposes only, and has no effect on existing analyses.

A BNFL Group company

Attachment 1 - Standard Text Our ref: LTR-NRC-06-8 March 16, 2006 Page 17 of 21 GENERAL CODE MAINTENANCE (Discretionary Change)

Background

A number of coding changes were made as part of normal code maintenance.

Examples include more descriptive file naming, improved automation in the ASTRUM codes, and improved input diagnostics in the WCOBRA/TRAC code., All of these changes are considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-1 3451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect None of these changes affect the results of design basis analyses. Therefore, the estimated effect is zero degrees.

A BNFL Group company

Attachment Standard Text Page 14 of 15 Our ref: LTR-NRC-07-23 May 15, 2007 GENERAL CODE MAINTENANCE (Discretionary Change)

Background

A number of coding changes were made as part of normal code maintenance.

Examples include additional information in code outputs, improved automation in the ASTRUM codes, increased WCOBRA/TRAC code dimensions, and general code cleanup. All of these changes are considered to be discretionary changes in accordance with Section 4.1.1 of WCAP-1 3451.

Affected Evaluation Model(s) 1996 Westinghouse best estimate large break LOCA evaluation model 1999 Westinghouse best estimate large break LOCA evaluation model, application to PWRs with upper plenum injection 2004 Westinghouse realistic large break LOCA evaluation model using ASTRUM Estimated Effect The nature of these changes leads to an estimated PCT impact of 00 F.

Page 13 of 32 Our ref: LTR-NRC-08-24 May 15, 2008 HOTSPOT FUEL RELOCATION (Non-Discretionary Change)

Background

In the axial node where burst is predicted to occur, a fuel relocation model in HOTSPOT is used to account for the likelihood that additional fuel pellet fragments above that elevation may settle into the burst region. It was discovered that the effect of fuel relocation on local linear heat rate was being calculated, but then cancelled out later in the coding. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect 1996 and 1999 BELOCA EMs analyses were assessed on a plant-specific basis, via the HOTSPOT reanalysis of a representative WCOBRA/TRAC case using the corrected code version at the burst elevation/burst model enabled sub-case. The HOTSPOT 95% probability PCT results were used to establish the plant-specific PCT penalty.

2004 ASTRUM EM analyses were assessed on a plant-specific basis, via the reanalysis of all of the burst cases from the original HOTSPOT calculations using the corrected HOTSPOT code version.

Serial No.12-352 Attachment 5 Application for Withholding Proprietary. Information ATTACHMENT 5 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 WESTINGHOUSE ELECTRIC COMPANY LLC, APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FOR PUBLIC DISCLOSURE AND THE ACCOMPANYING AFFIDAVIT (8 pages including this cover page)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

• Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: WPS-12-35 CAW-12-3495 June 5, 2012 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-LIS-12-324 P-Attachment, "Kewaunee Response to NRC RAIs Related to the Evaluation of Fuel Pellet Thermal Conductivity Degradation" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-12-3495 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Dominion Energy Kewaunee.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should referenceCAW-12-3495, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, AA.Gresham, Manager Regulatory Compliance Enclosures

CAW-12-3495 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 5th day of June 2012 Notary Public COMMONWEALTH OF PENNSYLVANIA NOTARIAL SEAL Renee Giampole, Notary Public Penn Township, Westmoreland County My Commission Expires September 28, 2013

2' CAW-12-3495 (1) *I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse, Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to'be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390,of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in thatconnection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive.

advantage, as follows:

(a) The information reyeals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-12-3495 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b). It is information that is marketable in many ways. The extent to which such information is available to competitors. diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-12-3495 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to. invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-LIS-12-324, P-Attachment, "Kewaunee Response to NRC RAIs Related to the Evaluation of Fuel Pellet Thermal Conductivity Degradation" (Proprietary), for submittal to the Commission, being transmitted by Dominion Energy Kewaunee letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with fuel thermal conductivity degradation, and maybe used only for that purpose.

5 CAW-12-3495 This information is part of that which will enable Westinghouse to:

(a) Assist customers in providing responses to RAls dealing with the 10 CFR 50.46, 30-day report.

Further this information has substantial commercial value as follows:

(a) Provide licensing support with respect to thermal conductivity degradation.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the.information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to.

make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.