ML081490282

Submittal of Approved Topical Report DOM-NAF-5, Revision 0.0-A, Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)

ML081490282
Person / Time
Site:	Kewaunee
Issue date:	05/20/2008
From:	Hartz L Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
08-0253
Download: ML081490282 (186)

Text

Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060

-a

-V IV; Dominiow Nay 20, 2008 U. S. Nuclear Regulatory Commission Serial No. 08-0253 Attention: Document Control Desk NL&OS/CDS: RO Washington, DC 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION SUBMITTAL OF APPROVED TOPICAL REPORT DOM-NAF-5. REVISION 0.0-A.

"APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS)"

In a letter dated August 16, 2006, as supplemented by letters dated December 6, 2006, April 16, 2007, May 4, 2007, June 12, 2007, and July 23, 2007 (references 1-6),

Dominion Energy Kewaunee, Inc. (DEK) requested approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)." The NRC approved this topical report by letter dated August 30, 2007 (reference 7).

Although not specified in the August 30, 2007 letter from NRC, DEK has prepared an accepted version of DOM-NAF-5 and is submitting it for information purposes consistent with the typical NRC criteria for topical reports. The accepted version of this topical report is DOM-NAF-5, Revision 0.0-A, and a complete copy is provided as an attachment to this letter.

Should you have any questions or require additional information, please contact Mr.

Craig Sly at 804-273-2784.

Very truly yours, Leslie N. Hartz Vice President - Nuclear Support Services

Attachment:

DOM-NAF-5, Revision 0.0-A, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," dated March 2008.

Serial No. 08-0253 Submittal of DOM-NAF-5 Page 2 of 3

References:

1. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated August 16, 2006 (ADAMS Accession Number ML062370351).

2. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated December 6, 2006 (ADAMS Accession Number ML0063410177).

3. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated April 16, 2007.

4. Letter from G. T. Bischof (DEK) to NRC, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-D/WRB-1 at Kewaunee Power Station," dated May 4, 2007.

5. Letter from G. T. Bischof (DEK) to NRC, "Response to NRC Questions Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated June 12, 2007.

6. Letter from W. R. Matthews (DEK) to NRC, "Response to NRC Questions Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated July 23, 2007.

7. Letter from P. D. Milano to D. A Christian, "Kewaunee Power Station - Safety Evaluation for Topical Report DOM-NAF-5 (TAC NO. MD2829)," dated August 30, 2007(ADAMS Accession Number ML072290373).

Commitments made in this letter: None

Serial No. 08-0253 Submittal of DOM-NAF-5 Page 3 of 3 cc: Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Ms. M. H. Chernoff Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-1H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 08-0253 Attachment Attachment DOM-NAF-5, REVISION 0.0-A "APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS)," DATED MARCH 2008.

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

DOM-NAF-5-0.0-A Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)

Nuclear Analysis and Fuel Department Dominion Richmond, Virginia March 2008 Prepared by:

Todd R. Flowers Reviewed by:

John R. Harrell Recommended for proval:

Suervisor, Fuel Project Engineering Approved:

K. L. asore Director, Nuclear Analysis and Fuel Approve Chýa ir n-in-Faciliti Safety Review Committee

Topical Report DOM-NAF-5, Rev. 0.0-A includes:

" NRC Safety Evaluation Report, dated August 30, 2007 (11 pages)

" Topical Report DOM-NAF-5-A, including the Classification / Disclaimer & Abstract (35 pages)

- submitted to the NRC in a letter dated August 16, 2006 (Serial Number 06-578)

Topical Report DOM-NAF-5-A is supplemented by:

e Attachment A (40 pages)

- Submitted to the NRC in a letter dated December 6, 2006 (Serial Number 06-578A)

• Attachment B (49 pages)

- Submitted to the NRC in a letter dated April 16, 2007 (Serial Number 06-578B)

• Attachment C (23 pages)

- Submitted to the NRC in a letter dated May 4, 2007 (Serial Number 06-578C)

ATTACHMENTS:

1) NRC Request for Additional Information on DOM-NAF-5 and Dominion Responses, dated June 12, 2007 (11 pages)

2) NRC Request for Additional Information on DOM-NAF-5 and Dominion Responses, dated July 23, 2007 (9 pages)

.. UNITED STATES NUCLEAR REGULATORY COMMISSION 0 ,WASHINGTON, D.C. 20555-0001 August 30, 2007 SERIAL #

4Kc'o SEP - 72007 Mr. David A. Christian President and Chief Nuclear Officer Innsbrook Technical Center NUCLEAR LICENSING 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

KEWAUNEE POWER STATION - SAFETY EVALUATION FOR TOPICAL REPORT DOM-NAF-5 (TAC NO. MD2829)

Dear Mr. Christian:

On August 16, 2006, as supplemented on December 6, 2006, April 16; May 4, and June 12, 2007, Dominion Energy Kewaunee submitted Topical Report (TR) DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis methods to Kewaunee Power Station (KPS)."

The Nuclear Regulatory Commission (NRC) staff has found that DOM-NAF-5 is acceptable for referencing in licensing applications for KPS to the extent specified and under the limitations delineated in the TR and in the enclosed safety evaluation (SE). The SE defines the basis for acceptance of the TR.

The NRC staffs acceptance applies only to material provided in the subject TR. The staff does not intend to repeat its review of the acceptable material described in the TR. License amendment requests that deviate from this TR will be subject to a plant-specific review in accordance with applicable review standards. If future changes to the NRC's regulatory requirements affect the acceptability of this TR, Dominion Energy Kewauee will be expected to revise the TR appropriately, or justify its continued applicability for subsequent referencing.

Sincerely, Patrick D. Milano, Senior Project Manager Plant Licensing Branch Il1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosure:

Safety Evaluation cc w/encls: See next page

Kewaunee Power Station cc:

Resident Inspectors Office Ms. Lillian M. Cuoco, Esq.

U.S. Nuclear Regulatory Commission Senior Counsel N490 Hwy 42 Dominion Resources Services, Inc.

Kewaunee, WI 54216-9510 Millstone Power Station Building 475, 5th Floor Ms. Leslie N. Hartz Rope Ferry Road Dominion Energy Kewaunee, Inc. Waterford, CT 06385 Kewaunee Power Station N 490 Highway 42 Kewaunee, WI 54216 Mr. Chris L. Funderburk Director, Nuclear Licensing and Operations Support Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 Mr. Thomas L. Breene Dominon Energy Kewaunee, Inc.

Kewaunee Power Station N490 Highway 42 Kewaunee, WI 54216

oUNITED 0 STATES "0 NUCLEAR REGULATORY COMMISSION

.* WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO FACILITY OPERATING LICENSE NO. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION DOCKET NO. 50-305

1.0 INTRODUCTION

By application dated August 16, 2006 (Reference 1), as supplemented on December 6, 2006, April 16, May 4, and June 12, 2007 (Agencywide Documents Access Management System (ADAMS) Accession Nos. ML070120088, ML063410177, ML071060392, ML071270780, and ML071630521, respectively), Dominion Energy Kewaunee, Inc. (the licensee) requested approval of Topical Report DOM-NAF-5, "Application Of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)." The report describes the various in-scope design and analysis methodologies and documents the assessments of the applicability of these methodologies to KPS. The approval by the Nuclear Regulatory Commission (NRC) would permit the licensee to subsequently request an amendment to the Technical Specifications (TSs) to apply the Dominion Energy (Dominion) nuclear core design and safety analysis methods to the KPS design and licensing analyses.

The nuclear core design methods addressed by the report include the Reload Nuclear Design Methodology, Relaxed Power Distribution Control (RPDC) Methodology, and the Studsvik Core Management System (CMS) Reactor Physics Methods. The safety analysis methods covered by the report include the Vepco Reactor System Transient Analyses using the RETRAN Computer Code, Statistical Departure from Nucleate Boiling ration (DNBR) Evaluation Methodology, and the Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code.

Attachments A (Reference 2) and B (Reference 3) to DOM-NAF-5 provide supplemental material documenting the applicability of Studsvik CMS Reactor Physics methods and Dominion's RETRAN methods to KPS.

2.0 REGULATORY EVALUATION

The NRC staff used the following requirements and guidance documents in evaluating the licensee's amendment request:

Section 50.34, "Contents of Applications; Technical Information," of Part 50 of Title 10 of the Code of FederalRegulations (10 CFR) requires that Safety Analysis Reports analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

As part of the core reload process, licensees perform reload safety evaluations to ensure that their safety analyses remain bounding for the design -cycle. To confirm that the analyses remain bounding, they confirm that the inputs to the safety analyses are conservative with respect to the current design cycle. These inputs are checked using analytical models, and if key safety analysis parameters are not bounded, further analysis of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

3.0 TECHNICAL EVALUATION

OF ANALYTICAL METHODS AND APPLICABILITY The core reload design and safety analysis process is currently performed by the KPS fuel supplier (Westinghouse), whereas other Dominion nuclear plants rely on the Dominion nuclear core design and safety analysis methods. In TR DOM-NAF-5, the licensee has proposed to apply the Dominion nuclear core design and safety analysis methods to KPS, although the KPS fuel supplier will continue to license the fuel design, perform fuel rod design analysis for reload fuel performance assessment and perform certain specific safety analyses, such as small-break and large-break loss-of-coolant accident (LOCA) analyses. The Dominion methods detailed herein are to be applied to KPS in a manner consistent with the conditions and limitations of this safety evaluation report (SER), other relevant NRC SERs and the relevant Dominion TRs.

3.1 Reload Nuclear Design Methodology The reload nuclear design methodology in Dominion TR VEP-FRD-42, Revision 2.0-A, "Reload Nuclear Design Methodology" (Reference 6), consists of the analytical models, methods, reload design and reload safety analysis, and an overview of analyzed accidents. It is an iterative process that involves the determination of a core loading pattern that fulfills cycle energy requirements and the demonstration that the plant with the reload core satisfies the constraints of the plant design basis and safety analysis limits.

The reload safety evaluation uses a bounding analysis concept in which key analysis parameters with limiting directions are identified such that, if all key analysis parameters are conservatively bounded, a reference safety analysis is applicable and no further analysis is necessary. If any values are not bounded, further analysis of the transient or accident in question is performed, the applicable safety analyses are revised, or changes are made in the operating requirements to satisfy applicable safety analysis criteria. The safety analysis process typically consists of steady state nuclear calculations used to derive the core physics related key analysis parameters as well as a dynamic accident analysis that utilizes these parameters to determine the accident result.

The Dominion nuclear design methodology and the current KPS reload design methodology are similar and share a common basis in Westinghouse TR WCAP-9272, "Westinghouse Reload Safety Evaluation" (Reference 7). Specific differences in nuclear steam supply system (NSSS),

reactor protection system (RPS), and fuel features between KPS and other Dominion Westinghouse units are capable of being reflected via modeling inputs in VEP-FRD-42 analytical methods, without changing the methodology. Implementation of this TR at KPS will be done in a manner consistent with the conditions and limitations identified in DOM-NAF-5.

The staff finds the reload nuclear design methodology applicable to KPS as detailed in DOM-NAF-5.

3.2 Relaxed Power Distribution Control (RPDC) Methodoloqy The RPDC methodology, VEP-NE-1 (Reference 8). is a Dominion method for axial power distribution control with a variable axial flux difference (delta-I) band that provides an increasing delta-I band with decreasing power in order to maintain approximately constant analysis margin at all power levels. RPDC provides several operational benefits, such as increased ability to return to power after a trip, reduced control rod motion to compensate for delta-I band restrictions, and reduced reactor coolant system (RCS) boration and dilution requirements.

The RPDC analysis determines acceptable delta-I bands to maintain design bases margin. The process consists of: the generation of power shapes that bound the delta-I range; the selection of delta-I bands such that all bands satisfy the core operating limits report (COLR) height dependent hot channel factor, FQ(Z), limit with verification that the proposed delta-I bands satisfy LOCA FQ and loss of flow accident (LOFA) thermal-hydraulic evaluations; the examination of limiting Condition II events; the verification that over-power delta-temperature (OPAT) and over-temperature delta-T (OTAT) limits are conservative; and N(Z) functions are formulated to support the implementation of FQ TSs surveillance.

Similarity between the Dominion RPDC and Westinghouse, Combustion Engineering (CE), and Exxon-relaxed axial power distribution control methodologies is noted in DOM-NAF-5 and the VEP-NE-1 SER. DOM-NAF-5 also notes that the cooldown transient assumption differs between the RPDC methodology (20 OF) and the Westinghouse relaxed axial offset (RAOC) methodology currently used for KPS (30 OF); the larger value will be used unless a KPS-specific analysis demonstrates that a plant trip will occur before a 30 °F cooldown.

The VEP-NE-1 SER states that the RPDC approach is an acceptable methodology for use with reload cores similar to those of Surry Power Station (SPS) and North Anna Power Station (NAPS) because: approved methodologies are used for the analyses supporting RPDC; justification of uncertainties is provided; and the impact of cycle specific variations on the delta-I power domain, OPAT and OTAT trip setpoints, and other safety analyses are evaluated on a reload basis. In light of the similarity of the NSSS, RPS, and fuel design at KPS, SPS, and NAPS, the NRC staff finds that the RPDC approach is an acceptable methodology for use at KPS as documented in DOM-NAF-5.

3.3 Studsvik Core Management System Reactor Physics Methods The Studsvik CMS reactor physics code package, detailed in TR DOM-NAF-1, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," (Reference 9), consists of CASMO-4, SIMULATE-3, and CMS-LINK. CASMO-4 is a multi-group two-dimensional transport theory code for depletion and branch calculations for a single assembly that is used to generate the lattice physics parameters, including cross sections, nuclide concentrations, pin power distributions and other nuclear data, which are used as inputs to SIMULATE-3. SIMULATE-3 is a two-group, three-dimensional modified coarse-mesh nodal diffusion theory code with coupled thermal-hydraulic and Doppler feedback. CMS-LINK is a linking code that processes CASMO-4 card image files into a binary formatted nuclear data library for use by SIMULATE-3.

Dominion uses the Studsvik CMS package for startup physics testing, RPDC, and licensing applications, including core reload design, core operation, and key core parameters for reload safety analyses.

The Studsvik CMS benchmarking data provided in DOM-NAF-1 was based on the 15x15 and 17x17 fuel designs used at SPS and NAPS respectively, while KPS currently uses 14x14 fuel.

In addition, DOM-NAF-1 SER limits the use of DOM-NAF-1, prohibiting its application to "significantly different or new fuel designs." Since this restriction is not clearly defined and in light of the absence of benchmarking data to 14x14 fuel, the KPS CMS models have been validated by comparison to benchmarks from both higher order Monte Carlo neutron transport calculations and reactor measurements from 10 cycles of operation spanning transitions in fuel enrichment, fuel density, spacer grid design, fuel vendor, core operating conditions and burnable poison design. These benchmarking results, as provided in Attachment A of DOM-NAF-5, are consistent with the staff approved methodology described in DOM-NAF-1.

Thus, the NRC staff finds that the Studsvik CMS methodology as detailed in DOM-NAF-1 and DOM-NAF-5 is applicable to KPS.

3.4 Reactor System Transient Analyses using RETRAN Dominion uses RETRAN to perform transient thermal-hydraulic analyses of the NSSS for best-estimate (e.g. training simulator validation) and licensing applications (e.g. reload core safety analysis), as detailed in TR VEP-FRD-41, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code" (Reference 10). RETRAN calculates general system parameters as a function of time and boundary conditions for input into more detailed calculations of departure from nucleate boiling (DNB) or other thermal and fuel performance margins.

The licensee performed transient analyses to confirm the adherence of reload core design limits to the bounds established by the reference analysis of record parameter values, as well as to verify that the core is acceptable from a safety operational point of view.

Transient analyses form an integral part of evaluations performed to verify the acceptability of a reload core design from the standpoints of safety and operational flexibility. The reload process consists of design initialization, design of the core loading pattern, and detailed characterization of the core loading pattern by the nuclear designer. The latter process determines the values of core physics related key analysis parameters. These key parameters are provided to the safety analyst, who uses them in conjunction with current plant operating configurations and limits to evaluate the impact of the core reload on plant safety.

The Dominion KPS RETRAN models have been validated by selecting representative transient events and comparing the results of the KPS RETRAN models to the vendor RETRAN model that was used to perform the current USAR analyses. This approach is similar to the one taken in VEP-FRD-41. The results of this analysis, as provided in Attachment B of DOM-NAF-5, show that the Dominion KPS RETRAN model compares favorably to the vendor RETRAN model for the selected transients, and the differences can be understood based on differences in noding, inputs, or other modeling assumptions. The NRC staff finds that the RETRAN methodology as detailed in DOM-NAF-5 and VEP-FRD-41 is applicable to KPS.

In performing this evaluation, it is necessary to ensure that those key parameters that influence accident response are maintained within the bounds or "limits" established by the parameter values used in the reference analysis (i.e. the currently applicable licensing calculation). The reference analysis (and the associated parameter limits) may be updated from time to time in support of a core reload or to evaluate the impact of some other plant parameter change.

In the case where a parameter is outside a previously defined limit, an evaluation of the impact of the change on the results for the appropriate transients is performed. This evaluation may be based on known sensitivities to changes in the various parameters in cases where a parameter change is small or the influence on the accident results is weak. For cases where larger parameter variations occur, or for parameters that have a strong influence on accident results, explicit reanalysis of the affected transients is required and performed. Past analytical experience has allowed the correlation of the various accidents with those parameters that have a significant impact on them.

If a reanalysis is performed, the results are compared to the appropriate analysis acceptance criteria. The reload evaluation process is complete if the acceptance criteria are met, and internal documentation of the reload evaluation is provided for the appropriate Dominion safety review. If the analysis acceptance criteria are not met, more detailed analyses and/or TS changes may be required to meet the acceptance criteria.

3.5 Statistical DNBR Evaluation Methodologv 3.5.1 Analytical Methods Topical Report DOM-NAF-5 details the events and analyses that will use the statistical DNBR evaluation methodology as well as those events that will use the deterministic models.

Topical Report VEP-NE-2, "Statistical DNBR Evaluation Methodology" (Reference 11),

describes Dominion's methodology for statistically treating several of the important uncertainties in the DNBR analysis. The statistical DNBR evaluation methodology is used to determine a plant-specific and fuel-specific statistical DNBR limit. This limit DNBR combines the core heat flux (CHF) correlation uncertainty with DNBR sensitivities to uncertainties in key DNBR analysis input parameters. The statistical combination of some of these uncertainties permits a more realistic combination of the independent uncertainties and, thus, provides a more realistic evaluation of DNBR margin. The statistical DNBR evaluation methodology allows thermal-hydraulic evaluations to be performed using nominal operating conditions as opposed to deterministic initial conditions (nominal conditions plus evaluated uncertainty).

The statistical DNBR evaluation methodology is typically applied to all Condition I and II DNB events (except rod withdrawal from subcritical, RWSC), and to the LOFA analysis, the locked rotor accident and the single rod cluster control assembly withdrawal at power (SRWAP). The events modeled statistically (see Table 3.5.1 of Reference 1) are limited by the statistical design limits (SDLs) evaluated in the implementation of the statistical DNBR evaluation methodology for KPS, dated May 4, 2007 (Reference 4). In addition, there are events that will be evaluated with deterministic models. These events will be initiated from bounding operating conditions (nominal value), with appropriate uncertainty added to these nominal values. The events

modeled deterministically are limited by the deterministic design limits (DDLs) stated in DOM-NAF-2 (Reference 12).

In its May 4, 2007, letter (Reference 4), Dominion submitted its KPS-specific statistical DNBR methodology analysis. The May 4, 2007, report supports the application of the NRC-approved TR stated above. In this plant-specific report, Dominion provided the technical basis and documentation necessary to evaluate the plant specific application of the VEP-NE-2-A methods to KPS. In its specific application analysis, Dominion used the VIPRE code with the Westinghouse WRB-1 CHF correlation for the thermal-hydraulic analysis of the Westinghouse 14x14 (422V+) fuel assemblies at KPS. The same report also provides documentation that the core safety limits and protection functions, such as the OTAT, OPAT, do not require revision as a consequence of this implementation.

3.5.2 Uncertainty Analysis Consistent with VEP-NE-2-A, (Reference 11), various plant parameters were selected as the statistically treated parameters in the implementation analysis. The magnitudes and functional forms of the uncertainties for the statistically treated parameters were derived in a rigorous analysis of plant hardware, measurement and calibration procedures, and have been summarized in Table 3.2-1 of the May 4, 2007, submittal (Reference 4).

The uncertainties for core thermal power, vessel flow rate, pressurizer pressure and inlet temperature were quantified using all sensor, rack, and other components of a total uncertainty and combined in a manner consistent with their relative dependence or independence.

Westinghouse quantified these uncertainties for Kewaunee's transition to Westinghouse's 14x14 (422V+) fuel. Total uncertainties were quantified at the 2 sigma level, corresponding to two-sided 95% probability.

The two-sided, 95/95 tolerance interval (95% probability, 95% confidence) for the measurement uncertainty of the nuclear enthalpy rise factor, FAH, is 3.5%. Conservatively, the measured FAH uncertainty was defined as a normal distribution with a 4% tolerance interval for consistency with previous applications.

The magnitude and distribution of uncertainty on the enthalpy rise hot channel factor, FAH. was quantified as a normal probability distribution with a magnitude of 3.0%. The statistical DNBR evaluation methodology (Reference 4) treats the FAH uncertainty as a uniform probability distribution.

3.5.3 Verification of Nominal Set-points Condition 1 of the NRC's SER for VEP-NE-2-A (Reference 11) requires that the nominal statepoints be shown to provide a bounding DNBR standard deviation for any set of conditions to which the methodology may be applied.

Consequently, in the May 4, 2007, submittal (Reference 4), the licensee provided analysis to demonstrate that Sto,, (the total DNBR standard deviation) as calculated in the TR is maximized for any conceivable set of conditions at which the core may approach the SDL. To this end, the licensee performed a regression analysis using as dependent variable the un-randomized

DNBR standard deviations at each nominal statepoint (i.e. the raw MDNBR results obtained from a Monte Carlo simulation). The nominal statepoint pressures, inlet temperatures, powers and flow rates are used as the independent variable. The licensee stated that an evaluation of all the data, linear fits, and regression coefficients indicated that there were no discernible trends in the database. Consequently, the licensee concluded that the total standard deviation had been maximized for any conceivable set of conditions at which the core may approach the SDL, and that the selected nominal statepoints provide a bounding standard deviation for any set of conditions to which the methodology may potentially be applied. The NRC staff finds the licensee's results and conclusion acceptable.

3.6 Reactor Core Thermal-Hydraulics using VIPRE-D Computer Code The VIPRE (Versatile Internals and Components Program for Reactors) computer code is a reactor core thermal-hydraulics code developed by Battelle Pacific Northwest Laboratories.

VIPRE is used to accurately calculate reactor coolant conditions to assure that the DNBR design limit is maintained.

The reactor core thermal-hydraulics code VIPRE-D, as described in TR DOM-NAF-2, is a Dominion-modified version of the VIPRE-01, MOD-02.1, which has been adapted to accommodate the various fuel designs used at Dominion nuclear power stations by incorporating vendor proprietary CHF correlations. The input and output has also been customized to incorporate it into the Dominion thermal hydraulic methodology.

VIPRE-D was approved by the NRC staff for pressurized-water reactor (PWR) licensing calculations up to the CHF using approved CHF correlations in accordance with the conditions and limitations listed in the SERs of DOM-NAF-2 and Electric Power Research Institute Report, NP-2511-CCM. In addition, VIPRE-D must be applied in a manner consistent with plant-specific and fuel-specific application conditions and limitations outlined in DOM-NAF-5. The NRC staff finds that the VIPRE-D thermal-hydraulics analysis methodology is applicable to KPS.

4.0 CONCLUSION

The NRC staff has reviewed Dominion's submittals and supporting documentation and finds the proposed use of Dominion nuclear core design and safety analysis methods at KPS to be acceptable. As such, TR DOM-NAF-5 is acceptable for use in licensing applications at KPS.

Based on the considerations discussed above, the NRC staff has concluded that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 REFERENCES

1. Dominion Energy Kewaunee, Inc. (DEK) letter, Gerald T. Bischof to NRC, request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," August 16, 2006.

2. DEK letter, Gerald T. Bischof to NRC, "Attachment A to Dominion Energy Kewaunee, Inc., Kewaunee Power Station Request for Approval of Topical Report DOM-NAF-5, "Application. of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," December 6, 2006.

3. DEK letter, Gerald T. Bischof to NRC, "Attachment B to Dominion Energy Kewaunee, Inc., Kewaunee Power Station Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," April 16, 2007.

4. DEK letter, Gerald T. Bischof to NRC, "Dominion Energy Kewaunee, Inc., Kewaunee Power Station Request for Approval of Topical Report DOM-NAF-5, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-DIWRB-1 at Kewaunee Power Station (KPS)," May 4, 2007.

5. Dominion letter, Leslie N. Hartz to NRC, "Virginia Electric and Power Company Responses to NRC questions regarding Kewuanee request for approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," June 12, 2007.

6. Dominion Topical Report VEP-FRD-42 Revision 2.0-A, "Reload Nuclear Design Methodology," August 2003.

7. Westinghouse Topical Report WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology," (Proprietary), March 1978.

8. Dominion Topical Report VEP-NE-1, Rev. 0.1-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications," August 2003.

9. Dominion Topical Report DOM-NAF-I, Rev. 0.0-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," June 2003.

10., Dominion Topical Report VEP-FRD-41, Rev. 0.1-A, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code," June 2004.

11. Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology,"

June 1987.

12. Dominion Topical Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," September 2004.

Principal Contributor: A. Attard Date: August 30, 2007

DOM-NAF-5-0.0-A Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)

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DOM-NAF-5, Rev. 0.0-A Classification/Disclaimer The data, information, analytical techniques, and conclusions in this report have been prepared solely for use by Dominion (the Company), and they may not be appropriate for use in situations other than those for which they are specifically prepared. The Company therefore makes no claim or warranty whatsoever, expressed or implied, as to their accuracy, usefulness, or applicability. In particular, THE COMPANY MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OR TRADE, with respect to this report or any of the data, information, analytical techniques, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written approval of the Company. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall the Company be liable, under any legal theory whatsoever (whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report.

Abstract DOM-NAF-5 is a Dominion topical report that documents justification for application of Dominion nuclear core design and safety analysis methods to Kewaunee Power Station (KPS). This report:

a) Describes Dominion nuclear core design and safety analysis methods, and b) Documents assessments of the applicability of Dominion nuclear core design and safety analysis methods to KPS.

Based on the applicability assessments of the Dominion nuclear core design and safety analysis methods described herein, the Dominion methods were determined to be applicable to KPS, and can be employed in design and licensing analyses for KPS. The applicability of certain methods (i.e., VEP-FRD-41, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code"; and DOM-NAF-1, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations") require further demonstration through detailed validation analyses that supplement DOM-NAF-5 (see Attachments A and B). A License Amendment Request (LAR), including a plant-specific and fuel-specific application analysis to define a Departure from Nucleate Boiling Ratio (DNBR) Statistical Design Limit (SDL),

is required to support implementation of these methods at KPS. The LAR will request addition of Dominion topical reports, through DOM-NAF-5, as reference methodologies in the KPS Core Operating Limits Report (COLR) and in Technical Specification (TS) 6.9.a.4. Other conforming Technical Specification changes are to be incorporated into the LAR, as needed to reflect use of Dominion methods.

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DOM-NAF-5, Rev. 0.0-A TABLE OF CONTENTS Page Classification Disclaimer 2.....................................

2 Abstract ............... ....................... . . . ................................................................................................... 2 Table of Contents .................. ............................................................................. 3 List of Tables/List of Attachments ...................................................................................................... 4 1.0 Introdu ction ........................................................................................................................................ 5 2.0 Dominion Nuclear Core Design and Safety Analysis Methodologies ................................................ 7 2.1 Dominion Methods to be Applied to KPS................................................................................... 7 2.2 Applicability Assessment Methodology ..................................................................................... 8 3.0 Applicability Assessments ......................................................................................................... 9 3.1 Applicability Assessment of Reload Nuclear Design Methods - VEP-FRD-42, "Reload Nuclear D esign Methodology". .................................................................................................................. 9 3.2 Applicability Assessment of Relaxed Power Distribution Control Methods - VEP-NE-1, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications" 12 3.3 Applicability Assessment of Core Management System Methods - DOM-NAF-1,. ..... "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Pow er Stations". ......................................................................................................... 15 3.4 Applicability Assessment of RETRAN Methods - VEP-FRD-41, "Vepco... Reactor System Transient Analyses Using the RETRAN Computer Code ..................................................................... 18 3.5 Applicability Assessment of Statistical DNBR Evaluation Methods - VEP-NE-2, .."Statistical DNBR Evaluation Methodology". ........................................................................................................... 23 3.6 Applicability Assessment of VIPRE-D Methods - DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code". ................................................................................... 26 4.0 Conclusions and Implementation ................................................................................................... 29 4.1 Conclusions ................................................................................................................................ 29 4.2 Steps for DOM-NAF-5 Implementation ................................................................................... 30 5 .0 Referen ces ...................................................................................................................................... 31 Page 3 of 35

DOM-NAF-5, Rev. 0.0-A LIST OF TABLES Table 3.3. 1: Fuel Assembly And Component Design Parameters.......................................... 16 Table 3.4.1: KPS-Specific Evaluation of the USNRC Generic RETRAN Code Restrictions and Limitations .21 Table 3.5. 1: USAR Transients Analyzed with VIPRE-D/WRB- 1 for KiPS ................................. 24 Page 4 of 35

DOM-NAF-5, Rev. 0.0-A 1.0 Introduction Kewaunee Power Station (KPS) became part of the Dominion nuclear fleet following Dominion's acquisition of KPS in July 2005. In addition to KPS, the Dominion nuclear fleet presently includes Surry Power Station (SPS),

Millstone Power Station (MPS), and North Anna Power Station (NAPS). Dominion nuclear core design and safety analysis methods were developed for application to.the original Dominion nuclear power stations (SPS and NAPS) in the 1980's. Over the years, these analysis methods have been successfully applied in numerous analytical, operational, and regulatory support activities.

DOM-NAF-5 is a Dominion topical report that documents justification for application of Dominion nuclear core design and safety analysis methods to KPS. This report:

a) Describes Dominion nuclear core design and safety analysis methods, and b) Documents assessments of the applicability of the Dominion nuclear core design and safety analysis methods to KPS.

Section 2.0 of the report identifies the analysis methods that are in the scope of application considered for this report. The following methods are outside the scope of application considered in this report:

a) Containment response and containment integrity analysis methods b) Radiological analysis methods c) Fuel rod design and analysis methods (Note: transient fuel rod thermal response for specific transient events is in scope. The transient fuel rod thermal response is to be calculated using the approved RETRAN hot-spot model as described in Reference 1. With this exception, the responsibility for fuel rod design calculations is to reside with the fuel vendor using the approved methods described in the Core Operating Limits Report.)

d) Small break, and large break loss of coolant accident (LOCA) analysis methods e) Control Rod Ejection analysis methods The reload design and safety analysis process performed by the current KPS fuel supplier (Westinghouse) is essentially the same process as the Dominion process, but it is performed with approved Westinghouse design and analysis methods. The current KPS Core Operating Limits Report (COLR) references the approved Westinghouse design and analysis methods for KPS. These Westinghouse design and analysis methods will remain applicable to KPS, and Dominion intends to retain the Westinghouse methods in the COLR after approval of DOM-NAF-5 to facilitate orderly transition to Dominion analyses.

Section 3.0 of this report describes the various in-scope design and analysis methodologies, and documents assessments of the applicability of those methodologies to KPS. Section 4.0 presents the conclusions derived from the methods applicability assessments.

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DOM-NAF-5, Rev. 0.0-A As described herein, Dominion nuclear core design and safety analysis methods have been determined to be applicable to KPS, and can be employed in design and licensing analyses for KPS. The applicability of certain methods (e.g., VEP-FRD-41, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code"; and DOM-NAF-l, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations") requires further demonstration through detailed validation analyses that supplement DOM-NAF-5 (see Attachments A and B). A License Amendment Request (LAR), including a plant-specific and fuel-specific application analysis to define a Departure from Nucleate Boiling Ratio (DNBR) Statistical Design Limit (SDL), is required to support implementation of these methods at KPS. The LAR will request addition of Dominion Topical Reports, through DOM-NAF-5, as reference methodologies in the KPS Core Operating Limits Report (COLR) and in Technical Specification (TS) 6.9.a.4.

Other conforming Technical Specification changes are to be incorporated into the LAR, as needed to reflect use of Dominion methods.

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DOM-NAF-5, Rev. 0.0-A 2.0 Dominion Nuclear Core Design and Safety Analysis Methodologies 2.1 Dominion Methods to be Applied to KPS Dominion currently applies its nuclear core design and safety analysis methods to its nuclear power stations, while the fuel vendor is responsible for fuel design analyses and reload fuel performance assessments. Dominion has performed reload design and safety analyses for approximately 65 reload cores at Surry and North Anna using both vendor and Dominion-developed tools. Dominion will apply the Dominion nuclear core design and safety analysis methods to KPS in the same manner it applies these methods to the other plants in the fleet. The KPS fuel vendor will retain responsibility for licensing the fuel design, for performing fuel rod design analysis, and for reload fuel performance assessment. For KPS, the fuel vendor will also perform certain specific safety analyses, e.g. small break and large break LOCA analyses.

Dominion has established a process for control and maintenance of its NRC-approved nuclear core design and safety analysis methodologies. Section 2.3 of Topical Report VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology," (Reference 6) refers to this process. This process was further defined in responses to Requests for Additional Information (RAIs) on Dominion's Reload Nuclear Design Methodology (Reference 23). The NRC reviewed the Dominion analysis methods control and maintenance process and found it acceptable, as discussed in their Safety Evaluation Report (SER) for Dominion's Reload Nuclear Design Methodology (Reference 7). The Dominion analysis methods applied to KPS are to be controlled and maintained using these approved processes.

The Dominion nuclear core design methods within the scope of this report are:

a) VEP-FRD-42 (current version: VEP-FRD-42, Rev. 2.1-A), "Reload Nuclear Design Methodology" (Reference 6) b) VEP -NE- 1 (current version: VEP-NE- 1, Rev. 0.1-A), "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications" (Reference 8) c) DOM-NAF- I (current version: DOM-NAF-1, Rev. 0.0-P-A), "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations" (Reference 9)

The Dominion safety analysis methods within the scope of this report are:

d) VEP-FRD-41 (current version: VEP-FRD-41, Rev. 0.1-A), "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code" (Reference 1) e) VEP-NE-2 (current version: VEP-NE-2-A), "Statistical DNBR Evaluation Methodology" (Reference 3) f) DOM-NAF-2 (current version: DOM-NAF-2), "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code" (Reference 2)

Each of the above methods is assessed for applicability to KPS in Section 3.0 using the Applicability Assessment Methodology described below in Section 2.2. Throughout the remainder of this report, each of these reports is cited without reference to the revision. Section 5.0, References, cites the current report versions as of the date of this report (July 2006).

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DOM-NAF-5, Rev. 0.0-A 2.2 Applicability Assessment Methodology Dominion analysis methods are to be applied to KPS in a manner consistent with the conditions and limitations described in the Dominion topical reports and in applicable NRC Safety Evaluation Reports (SERs). Any differences from the methods described in the Dominion Topical Reports that are required for application of the methods to KPS are identified and addressed through the Applicability Assessment Methodology described herein.

The following systematic evaluation process is applied herein to assess the application of candidate methodologies to KPS:

a) Each method is described, including its purpose, key features, and dependencies. Descriptions include:

- Key phenomena/conditions predicted by the method

- General calculation approach or assumptions

- Types of reactor conditions for which the method is used b) Conditions and limitations associated with each method are identified, including:

-Regulatory limitations in NRC Safety Evaluation Reports (SER)

- Physical limitations (e.g., plant systems, plant features & conditions)

- Limitations in Dominion Topical Reports (e.g., specific modeling approaches or inherent assumptions) c) Each method is assessed with respect to the identified Conditions and Limitations. The assessment effort ranges from written evaluations, to validation and benchmark analyses and detailed comparisons of results.

d) The results of the applicability assessment are documented for each method. Results from some of the more involved assessment efforts are presented in Attachments A and B.

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DOM-NAF-5, Rev. 0.0-A 3.0 Applicability Assessments 3.1 Applicability Assessment of Reload Nuclear Design Methods - VEP-FRD-42, "Reload Nuclear Design Methodology" 3.1.1 Description of Methodology The Dominion reload nuclear design methodology includes calculational and process elements that are employed in the design and evaluation of reload nuclear cores. The major activities of the methodology are:

1) determination and fulfillment of cycle energy requirements; 2) determination of a core loading pattern; and

3) a reload safety evaluation that confirms acceptable behavior for the reload core under predicted design basis accident conditions.

/

The Dominion reload nuclear design methods, as documented in VEP-FRD-42 (Reference 6), consist of the following elements:

a) Analytical Models (e.g., Studsvik Core Management System (CMS) Models, VEPCO RETRAN Models, Core Thermal-Hydraulics VIPRE-D Models) b) Analytical Methods (e.g., Core Depletions, Core Reactivity Parameters and Coefficients, Core Reactivity Control, Safety Analysis, Statistical DNB) c) Reload Design Process (e.g., Core Loading Pattern Design & Optimization, Key Parameter Treatment in Nuclear Design Analyses) d) Reload Safety Evaluation Process e) Nuclear Design Report, Operator Curves & Core Follow Data The Dominion methodology for designing a reload core is an iterative process. The process involves determining a core loading pattern which provides the required total cycle energy and then demonstrating through analysis or evaluation that the plant will continue to meet all applicable safety criteria after considering the changes associated with the reload core. Should the characteristics of the proposed loading pattern cause any safety analysis criteria to be exceeded, the loading pattern is revised.

Reload safety evaluation and analysis criteria are established using a bounding analysis concept. This approach employs a list of key analysis parameters with the limiting direction of each parameter identified.

This allows reload core characteristics to be compared with the parameter values assumed in the reference analyses for various transients and accidents. For a proposed core reload design, if all key analysis parameters are conservatively bounded, then the reference safety analysis applies, and no further analysis is necessary. If one or more key analysis parameters are not bounded, then further analysis or evaluation of the transient or accident in question is performed. Occasionally, the applicable safety analyses are revised, or changes are made in the operating requirements (e,g., Technical Specifications or Core Operating Limits Report (COLR) changes) to ensure that plant operation will satisfy the applicable safety analysis criteria for the proposed loading pattem.

Topical Report WCAP-9272, "Westinghouse Reload Safety Evaluation," dated March 1978 (Reference 10) describes the Westinghouse methodology used for reload safety evaluation. WCAP-9272 forms the basis for Dominion's reload methodology as described in Topical Report VEP-FRD-42. The Westinghouse methodology defines the specific key parameters for use in accident analyses and provides limiting directions for consideration in reload evaluations.

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I DOM-NAF-5, Rev. 0.0-A The reload core design is evaluated by comparing the reload core parameters against the assumptions in the current safety analyses. Safety analysis (accident analysis) is the study of nuclear reactor behavior under accident conditions. The accident analyses consider all relevant aspects of the plant and core including the operating procedures and limits on controllable planit.parameters and the engineered safety, shutdown, and containment systems.

There are two stages in the typical safety analysis process. First, steady state nuclear calculations are performed for the core conditions assumed in the accident analysis. The nuclear parameters derived from these calculations are called the core physics related key analysis parameters and serve as input to the second stage. The second stage is the actual dynamic 'accident analysis, which yields the accident results that are applicable for these key analysis parameter values. The accident analyses are transient calculations that usually model the core nuclear kinetics and those parts of the plant systems that have a significant impact on the events under consideration.

The Kewaunee Updated Safety Analysis Report (USAR) documents acceptable plant safety via detailed results of accident analyses performed with the bounding values of key analysis parameters. Plant safety is demonstrated if accident analysis results meet the applicable acceptance criteria. The reload core design is evaluated by comparing the core physics related key analysis parameters against the assumptions in the current safety analyses. The reload evaluation process is complete if the acceptance criteria delineated in the USAR are satisfied with the reload core implemented. If an accident reanalysis is necessary, more detailed analysis methods and/or Technical Specifi.*cations changes may be required to meet the acceptance criteria. Such changes are to be processed in accordance with the applicable regulatory processes.

In summary, the overall reload evaluation process includes the following steps:

a) Determine bounding key analysis parameters, which constitute the current limits for reload cores.

b) Perform (or confirm) accident analysis using the bounding key analysis parameters and conservative assumptions.

c) Establish a proposed core loading pattemn that provides the required total cycle energy.

d)- Determine, for the proposed core loading pattern, the value for each key analysis parameter.

e) Compare key reload analysis parameters to current limits.

f) Evaluate whether an accident reanalysis is needed based on the effect the reload key analysis parameters may have on the accident.

g) Performn reanalysis of specific affected accidents, change operating limits, or revise the loading pattern, as necessary.

Key attributes of the reload nuclear design methodology include the following elements:

a) An analysis framework in which safety analyses establish acceptable limit values for reload core key analysis parameters, while nuclear and fuiel design codes confirm each core's margin to these limits b) The use of bounding key parameter values in reference safety analyses c) Recurrent validation of nuclear design analytical predictions through comparison with reload core measurement data d) Representation of key fuel features via detailed inputs in core design and safety analysis models e) Fuel modeling using approved critical heat flux correlations demonstrated to be applicable and within the range of qualification and identified in the plant COLR section of the TS 3.1.2 Conditions and Limitations Page 10 of 35

DOM-NAF-5, Rev. 0.0-A There are specific and inherent conditions and limitations that are associated with application of the methods documented in VEP-FRD-42 to KPS:

a) Regulatory limitations in NRC Safety Evaluation Reports (SER)

i. Inherent limitation for use at North Anna and Surry Power Stations ii. Prior to its use for fuel types other than Westinghouse and Framatome ANP Advanced Mark-BW fuel, confirm that the impact of the fuel design and its specific features can be accurately modeled with the VEPCO nuclear design and safety analysis codes and methods.

Should the changes necessary to accommodate another fuel product require changes to the reload methodology of Topical Report VEP-FRD-42, these proposed changes are required to be submitted for prior NRC review and approval.

b) Physical limitations (e.g., plant systems, plant features & conditions)

None identified (The methods of VEP-FRD-42 are not dependent on such physical limitations) c) Limitations in Dominion Topical Reports (e.g., specific modeling approaches or inherent assumptions)

None identified 3.1.3 Assessment The Dominion reload nuclear design methods and the current KPS reload nuclear design methods are both based on the Westinghouse Topical Report WCAP-9272. Thus, the Dominion and KPS reload nuclear design methods are similar since they have a common basis in the Westinghouse reload safety evaluation methods.

KPS and the other Dominion Westinghouse nuclear units use Westinghouse designs for nuclear steam supply system (NSSS) and reactor'protection system (RPS). KPS and the other Dominion Westinghouse units have many design and operating similarities. The specific differences in NSSS, RPS and fuel features for KPS are all capable of being reflected via modeling inputs in the analytical methods of VEP-FRD-42. These differences do not impact the execution of the key VEP-FRD-42 methodology elements for the design and evaluation of reload cores.

The reload safety evaluation and analysis process for Dominion and KPS uses a bounding analysis concept.

The method that is used for Dominion and KPS employs a list of key analysis parameters and limiting directions of the key analysis parameters for various transients and accidents.

The key analysis parameters and the limiting directions of those key parameters for the various transients and accidents were evaluated. The key analysis parameters and their limiting direction for KPS are the same as the key analysis parameters and limiting direction for the other Dominion Westinghouse units. The design basis accidents and transients in the safety analyses were also evaluated. The design basis transients and accidents for KPS are similar to the design basis transients and accidents for the other Dominion Westinghouse units. Based on the key analysis parameters assessment, and the design basis transients and accidents that are evaluated in the reload safety evaluation process, the Dominion reload safety evaluation and analysis methods have been determined to be applicable to KPS.

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DOM-NAF-5, Rev. 0.0-A 3.1.4 Summary Dominion reload nuclear design methods documented in VEP-FRD-42 are concluded to be applicable to KPS and can be applied to KPS licensing analysis for reload core design and reload safety evaluation.

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DOM-NAF-5, Rev. 0.0-A 3.2 Applicability Assessment of Relaxed Power Distribution Control Methods - VEP-NE-1, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications" 3.2.1 Description of Methodology The relaxed power distribution control (RPDC) method is the Dominion method for axial power distribution control. RPDC involves a variable axial flux difference (delta-I) band power distribution control strategy that uses a widened full power delta-I band, and provides for an increasing delta-I band with decreasing power.

The widened delta-I band is based on maintaining an approximately constant analysis margin to the design bases limits at all power levels.

RPDC provides several benefits to plant operation, described as follows. The ability to return to power after a trip, particularly at end-of-cycle (EOC), is enhanced. Control rod motion necessary to compensate for delta-I band restrictions is reduced to only that motion needed to maintain operation within a much wider band. The reactor coolant system boration and dilution requirements are decreased due, in part, to the reduced control rod motion. There is generally an enhancement of operational flexibility. The RPDC methodology allows the Dominion units to operate with additional flexibility while at the same time ensuring that the design bases limits are met with an appropriate margin.

RPDC also involves the formulation of Technical Specification surveillance and COLR limits for Total Peaking Factor (FQ). The FQ surveillance uses the measured core axial position-dependent FQ (FQ(Z))

augmented by a non-equilibrium operation multiplier (N(Z)) in order to verify compliance with the peaking factor limits. This FQ surveillance is a required element of the RPDC method.

The objective of the RPDC analysis is to determine acceptable delta-I bands that maintain margin to all the applicable design bases criteria and at the same time provide enhanced delta-I operating margin. Because the RPDC delta-I band is an analysis output quantity rather than a fixed input, power shapes that adequately bound the potential delta-I range must be generated. These power shapes must include the effect of combinations of the key parameters such as bumup, control rod position, xenon distribution, and power level Dominion has developed the methodology to generate the large number of power shapes required for RPDC analyses.

After the power shapes have been created, proposed delta-I bands are chosen such that all shapes within the delta-I bands satisfy the COLR FQ(Z) limit. For the normal operation analysis, power levels spanning the 50% to 100% range are investigated to establish the RPDC delta-I limits. Further verification of the proposed delta-I bands is performed via two different limiting shape evaluations, one based on Loss of Coolant Accident (LOCA) FQ considerations and the other based on a Loss of Flow Accident (LOFA) thermal/hydraulic evaluation. Delta-I bands can be narrowed to satisfy the requirements of these evaluations, if necessary.

Condition II or Abnormal Operation events, which may be the result of system malfunctions or operator errors, are also analyzed for RPDC. This RPDC analysis examines the more limiting of these Condition II events and verifies on a cycle-to-cycle basis that the Over-Power Delta-T (OPAT) and the Over-Temperature Delta-T (OTAT) setpoints are conservative. The OPAT and OTAT setpoints were designed primarily to provide transient and steady state protection against fuel centerline melt and DNB, respectively.

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DOM-NAF-5, Rev. 0.0-A The RPDC methodology takes advantage of the large amount of margin to the design bases limits available at reduced power levels and provides delta-I limits at all power levels that are less restrictive than under Constant Axial Offset Control (CAOC) operation. The RPDC methodology may be summarized as follows:

a) A full range of normal-operation power 'shapes is obtained by combining the key parameters upon which each shape is dependent: xenon distribution, core burnup, boron concentration, core power level and control rod position. Reasonable increments spanning the entire range of values are considered for each of these parameters. A xenon "free oscillation" method is used to create the many and varied axial xenon distributions required for this analysis.

b) Proposed delta-I bands are selected such that the COLR FQ(Z) limit is met for all power shapes within the proposed bands. These power shapes are then analyzed to determine which shapes result in an approach to the LOFA limits.

c) Final normal operation delta-I bands are established such that the LOCA and the LOFA limits are satisfied.

d) Conditions that yield shapes within the final normal operation delta-I bands are used as initial conditions for the bounding Condition II accident simulations.

e) The resultant transient shapes are analyzed and the OPAT and OTAT trip function/setpoints are verified to ensure that margin to fuel design limits is maintained.

f) N(Z) functions (non-equilibrium power distribution multiplier) are formulated based on the maximum composite calculated Condition I FQ(Z) x P (i.e., local FQ times total core thermal power) to support the implementation of FQ Technical Specifications surveillance.

All neutronic calculations are performed with NRC-approved codes and methods. All DNBR calculations are performed using NRC-approved thermal-hydraulic code(s), correlation(s), and methods.

3.2.2 Conditions and Limitations There are no specific conditions or limitations specified in the RPDC SER (Reference 8) for use of this methodology. Commonality amongst the Dominion, Westinghouse, CE, and Exxon versions of this methodology is noted throughout the RPDC SER. The following comments from the RPDC SER are relevant to the assessment of use of the RPDC method for KPS:

a) Approved methods were used for analyses supporting RPDC.

b) Justification was provided for the uncertainties assigned.

c) Impact of cycle specific variations on the delta-I power domain, OPAT and OTAT trip setpoints, and other safety analyses will be evaluated on a reload basis.

d) RPDC is an acceptable methodology for application to reload cores that are similar to those of the Surny (SPS) and North Anna (NAPS) reactors.

The following methodology items must be evaluated to determine if the values stated are applicable to KPS cores:

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.-I DOM-NAF-5, Rev. 0.0-A e) The appropriate plant-specific value for the maximum calculated temperature reduction during an EOC cooldown transient is to be determined for Kewaunee. A cooldown of 20'F was shown to be bounding for North Anna.

f) A conservative relationship is established for the Technical Specification/COLR FQ surveillance such that a 1% increase in FQ is mitigated by a 1% narrowing of the delta-I bands or a 1%

reduction in core power (with commensurate reductions in trip setpoints).

g) A dilution time of 15 minutes is assumed after the control rods pass the insertion limit for the boron dilution event.

h) The stated combined calculation uncertainty (FQU) for predicted FQ is 1.0815.

3.2.3 Assessment Conditions a-c cited in Section 3.2.2 are met for use of RPDC at KPS because the same methods, uncertainty parameters, and analyses will be employed as are currently employed for North Anna RPDC analyses. The specific KPS uncertainty for calculated FQ (FQU, conditions b and h) will be developed as part of KPS validation analysis for the CMS methods (see applicability assessment for DOM-NAF-1 methodology in Section 3.3). Condition d is satisfied due to the many similarities between KPS, SPS, and NAPS. All three stations use Westinghouse designs for the nuclear steam supply system (NSSS) and reactor protection system (RPS). In addition, KPS reload cores are similar to those at SPS because 14x14 fuel and 15x15 fuel have nearly identical fuel pin design and pin pitch (see comparison in Table 3.3.1). The applicability of the RPDC methodology to KPS is also supported by the fact that the Westinghouse Relaxed Axial Offset Control (RAOC) method is currently used by KPS. There are a number of similarities between Dominion's RPDC method and the Westinghouse RAOC power distribution control method (Reference 11),

including:

a) Technical Specifications/COLR Delta I bands (limits) and variation of delta-I bands versus power level b) FQ surveillance requirements and FQ limits are in the Technical Specifications and/or COLR c) Normal (Condition I) and abnormal (Condition II) events are considered d) Skewed axial xenon distributions are used to generate Condition I axial power shape variations e) Combinations of key variables are used (control rod insertion, core bumup, and power level) f) Cycle specific evaluations are performed for delta-I bands vs. power level g) Cycle specific evaluations are performed for OPAT, OTAT setpoint verification h) RAOC non-equilibrium multiplier W(Z) is analogous to the RPDC N(Z) function With regard to the applicability of specific values in the RPDC methodology to KPS (conditions e-h), the cooldown transient assumption of 20'F (condition e) is smaller than the 30'F value in the Westinghouse method currently used for KPS. A cooldown limit of 30'F shall be used unless a KPS-specific analysis can demonstrate that a plant trip will occur prior to reaching 30'F. KPS Technical Specification 3.10.b requires a reduction of 1% in the delta-I bands (axial flux distribution-AFD bands) or 1% in power (with a commensurate reduction in setpoints) for each 1% violation of the FQ limit when evaluated including the W(Z) factor (condition f). This is the same relationship specified for Dominion's RPDC method (Reference 8). The at-power 15 minute dilution time prior to operator action (condition g) value is the same as that stated in KPS USAR Section 14.1.4 (CVCS malfunction event description), and the same as that currently used for KPS. A KPS-specific value for FQU (condition h) will be developed using CMS methodology benchmark results. It is expected that the value 1.0815 will be supported by those results.

3.2.4 Summary Page 15 of 35

DOM-NAF-5, Rev. 0.0-A The Dominion RPDC method is determined to be applicable to KPS and can be applied to KPS licensing analysis for nuclear core design and reload safety evaluation. Specific values assumed in the Dominion RPDC methodology, except for the maximum assumed cooldown temperature, are appropriate for KPS application or will be determined following CMS core design methods validation analyses (FQU). The License Amendment Request (LAR) to add DOM-NAF-5-A to Section 6.9.a.4 of the KPS Technical Specifications will include Technical Specification changes necessary for conformance with the RPDC methodology.

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DOM-NAF-5, Rev. 0.0-A 3.3 Applicability Assessment of Core Management System Methods - DOM-NAF- 1, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations" 3.3.1 Description of Methodology The Dominion reactor physics methods include the Studsvik Core Management System (CMS) core modeling code package. The primary computer codes in the CMS package are CASMO-4 and SIMULATE-3. The CASMO-4 computer code is the fuel assembly lattice code. CASMO-4 is a multi-group, two-dimensional transport theory code used for depletion and branch calculations for a single fuel assembly.

The SIMULATE-3 code is a two-group, 3-dimensional nodal code based on the modified coarse mesh (nodal) diffusion theory calculation technique, coupled with thermal hydraulic and Doppler feedback. The general CMS calculation approach is to model the fuel assembly using the CASMO two-dimensional lattice physics code, and then to construct the three-dimensional SIMULATE reactor core model using lattice physics cross section data.

CMS reactor physics codes are used to model the core physics characteristics of the reload core including depletion/isotopic effects, reactivity, reactivity coefficients, power distribution, and shutdown margin.

Dominion uses CMS reactor physics models in licensing applications, including calculations for core reload design, core operation, and key core parameters for reload safety analyses. CMS models are applied in the analyses for relaxed power distribution control, for startup physics testing (including control rod worth determination using the boron dilution and rod swap measurement techniques), and to provide physics constants for measurement of core power distributions.

CMS models are used to analyze the reactor core in all modes of operation including refueling shutdown, cold shutdown, 0% to 100% reactor power, and conditions associated with design basis transients. CMS models are applied over the entire fuel cycle from beginning to end of cycle.

3.3.2 Conditions and Limitations a) The DOM-NAF- 1 title and several statements in its SER refer to use of CMS for North Anna (NAPS) and Surry (SPS) Power Stations.

b) Benchmarking data was provided for 15x15 (SPS) and 17x17 (NAPS) fuel designs, while the KPS fuel design is 14x14.

c) In Section 5.0, "Conditions and Limitations," the SER lists two conditions that would require further validation and NRC approval:

i. Use of mixed oxide fuel ii. Introduction of significantly different or new fuel designs 3.3.3 Assessment Referring to the Conditions and Limitations listed in Section 3.3.2:

Condition a is not a technical limitation. Condition b is not a technical limitation provided the 14x14 design is not "significantly different" than the 15x15 and/or 17x17 designs. For condition c, part iKPS does not use mixed oxide fuel. The balance of this assessment will focus on condition c, part ii.

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DOM-NAF-5, Rev. 0.0-A The KPS fuel assembly lattice is a 14x14 fuel lattice and the current KPS fuel design is the Westinghouse 422 V+ fuel design. The KPS fuel lattice and fuel design are not significantly different from the SPS (15x15) and NAPS (17x17) fuel designs. Table 3.3.1 demonstrates the similarities of the 14x14 and 15x15 designs.

Table 3.3.1 FUEL ASSEMBLY AND COMPONENT DESIGN PARMETERS Component KPS SPS Fuel Assembly Array 14 x 14 15 x 15 Pitch (Assy) 7.803 in. 8.466 in.

Pitch (Rod) 0.556 in. 0.563 in.

No. guide tubes 16 20 No. instrument tubes 1 1 No. spacer grids 7 7 Fuel Rods Fuel Pellet Diameter 0.3659 in. 0.3659 in.

Fuel Pellet Material Sintered U0 2 Sintered U0 2 Fuel/Clad Diametric Gap 0.0075 in. 0.0075 in.

Fuel Cladding O.D. 0.422 in. 0.422 in.

I.D. 0.3734 in. 0.3734 in.

Material ZIRLO ZIRLO Spacers (Top & Bottom/Mid)

Material Inconel/ZIRLO Inconel/ZIRLO Guide Tube O.D. (above dashpot) 0.526 in. 0.533 in.

I.D. (above dashpot) 0.492 in. 0.499 in.

Material ZIRLO ZIRLO The most significant difference between the 14x14 and 15x15 designs for core physics calculations is the slight asymmetry of the 14x14 design (off-center instrument thimble). Because the SER restriction related to "significantly different" or "new fuel designs" is not clearly defined, any difference between the KPS fuel design and those specifically addressed in DOM-NAF-1 (Reference 9) could cause the 14x14 fuel to be categorized as a "new fuel design." Although the actual differences are minor, additional validation information will be provided to support application of DOM-NAF- 1 methods to KPS.

Using the methods and processes delineated in DOM-NAF-1, the accuracy of the CMS models will be demonstrated through comparisons with reactor measurements and through comparisons with higher order Monte Carlo neutron transport calculations. This demonstration will be consistent with the assessment performed for the Dominion Surry and North Anna units as described in DOM-NAF- 1. Where applicable, nuclear reliability factors (NIRFs) will be determined for the key reactor physics parameters. The KPS NRFs are expected to be similar to those established for NAPS and SPS. The capability of the CMS models to support the KPS Startup Physics Test Program will also be demonstrated through the reactor measurement comparisons. It is expected that the CMS models are fully compatible with the KPS reactor test program.

As part of the development of the KPS model validation data, Dominion will compare CMS and Monte Carlo code calculations of reactivity worth for soluble boron, control rods, burnable absorber, moderator Page 18 of 35

DOM-NAF-5, Rev. 0.0-A temperature defect, and fuel temperature. Dominion will compare SIMULATE predictions to reactor measured data and use statistical methods where applicable to determine CMS model uncertainties.

Dominion will also use the CMS and Monte Carlo code calculations in combination with normalized reactor flux map reaction rate comparisons to determine appropriate power distribution peaking factor reliability factors.

Dominion will use data from multiple KPS operating cycles to benchmark the CMS models. These cycles cover core design changes including transitions in fuel enrichment, fuel density, spacer grid design and material, fuel vendor, core operating conditions (full-power average moderator temperature and rated thermal power), and burnable poison material and design. Dominion will use critical boron concentration and critical rod position, startup physics testing data, measured power distributions, and operational transient data in the CMS model benchmarking. The agreement between the measured and calculated values will be used to validate the application of CMS to KPS' Dominion will demonstrate that the CMS reactor physics models in conjunction with the indicated nuclear reliability factors, adequately represent the operating characteristics of KPS.

3.3.4 Summary Results of the applicability assessment for the DOM-NAF-1 reactor physics methods to KPS will be provided in a supplement to DOM-NAF-5 to be included as Attachment A.

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DOM-NAF-5, Rev. 0.0-A 3.4 Applicability Assessment of RETRAN Methods - VEP-FRD-41, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code" 3.4.1 Description of Methodology 3.4.1.1 Background Dominion has developed the capability to perform system transient analyses. This capability, coupled with core thermal/hydraulic analysis capability, encompasses the non-LOCA licensing analyses required for the Condition 1,11, 11m, and IV transients and accidents addressed in the Updated Safety Analysis Report (USAR). In addition, the capability for performing best-estimate analyses for plant operational support applications has also been developed.

The purposes of having transient and accident safety analysis capability are to: 1) maintain in-house cognizance and expertise in the system transient analysis area; 2) support plant operation; and 3) provide a basis for the reload core safety analysis and licensing process. The principal analysis tool is the RETRAN computer code (Reference 12), which determines the time-dependent (transient) thermal-hydraulic response of a Nuclear Steam Supply System (NSSS). The RETRAN computer code calculates: 1) general system parameters as a function of time; and 2) boundary conditions for input into more detailed calculations of Departure from Nucleate Boiling (DNB) or other thermal and fuel performance margins. The theory and numerical algorithms, the programming details, and the user's input information for the RETRAN computer code have been documented by its developers, Energy Incorporated (EI) and the Electric Power Research Institute (EPRI), in Volumes I through IV of Reference 12. Volume IV of Reference 12 provides the results of the extensive verification and qualification of the RETRAN code. The verification activity consisted of qualification of the code by comparison of code results with separate effects experiments, with systems effects tests, and with integrated system responses based on actual plant data or USAR results.

In conjunction with both an analysis tool and system models, the development of a non-LOCA licensing analysis capability requires conservative analysis assumptions and input data. For licensing calculations, the major Dominion analysis assumptions are consistent with those documented in the units' USAR. If a change in analysis assumptions is required by a plant modification, core reload, or a related change, the change will be assessed via the 10 CFR 50.59 process. Depending on the results of that assessment, either the analysis is submitted to the NRC for approval or a normal update of the appropriate section of the USAR is prepared.

3.4.1.2 Licensing Applications Dominion's system transient analysis capability is intended for both best-estimate (e.g. training simulator validation) and licensing applications (e.g. core reload analysis). Since core reloads are the most common and expected reason for accident reanalysis, Dominion's system transient methodology is discussed in that context.

Transient analyses form an integral part of evaluations performed to verify the acceptability of a reload core design from the standpoints of safety, economics, and operational flexibility. The reload process consists of design initialization, design of the core loading pattem, and detailed characterization of the core loading pattern by the nuclear designer. The latter process determines the values of core physics related key analysis parameters. These key parameters are provided to the safety analyst who uses them in conjunction Page 20 of 35

DOM-NAF-5, Rev. 0.0-A with current plant operating configurations and limits to evaluate the impact of the core reload on plant safety.

hin performing this evaluation, it is necessary to ensure that those key parameters that influence accident response are maintained within the bounds or "limits" established by the parameter values used in the reference analysis (i.e. the currently applicable licensing calculation). The reference analysis (and the associated parameter limits) may be updated from time to time in support of a core reload or to evaluate the impact of some other plant parameter change.

For cases where a parameter is outside of these previously defined limits, an evaluation of the impact of the change on the results for the appropriate transients must be made. This evaluation may be based on known sensitivities to changes in the various parameters in cases where a parameter change is small or the influence on the accident results is weak. For cases where larger parameter variations occur, or for parameters that have a strong influence on accident results, explicit reanalysis of the affected transients is required and performed. Past analytical experience has allowed the correlation of the various accidents with those parameters that have a significant impact on them.

If a reanalysis is performed, the results are compared to the appropriate analysis acceptance criteria. The reload evaluation process is complete if the acceptance criteria are met, and internal documentation of the reload evaluation is provided for the appropriate Dominion safety review. If the analysis acceptance criteria are not met, more detailed analyses and/or Technical Specifications changes may be required to meet the acceptance criteria. Analysis changes are evaluated in accordance with the requirements of 10 CFR 50.59.

3.4.1.3 System Model Application The production of a conservative, reliable safety analysis of a given anticipated or postulated transient is accomplished by combining a system transient model with appropriate transient-specific input. A system transient model is designed to provide an accurate representation of the reactor plant and those associated systems and components that significantly affect the course of the transient. Transient-specific input ensures that the dynamic response of the system to the postulated abnormality is predicted in a conservative manner, and includes: a) initial conditions; b) core reactivity parameters such as Doppler and moderator temperature coefficients, and control rod insertion and reactivity characteristics; and c) assumptions concerning overall systems performance. Important system performance assumptions include the availability of certain system components (such as pressurizer spray or relief valves) and control and protection system characteristics (setpoints, instrument errors, and delay times).

RETRAN affords the modeling flexibility to develop an infinite number of representations for a given nuclear plant. At Dominion, several standard plant models are assembled and maintained for performance of the entire spectrum of system transient analyses. RETRAN makes use of an input structure that allows modification of the base deck input for specific cases by use of override cards. Thus, specific transient cases may be executed without altering the base plant models.

The base models are designed to provide a basic system description comprised of those parameters that would not ordinarily change from cycle to cycle. Thus, such parameters as system volumes and flow areas, characteristics of various relief and safety valves, and primary coolant pump characteristics form part of the base models.

Dominion's RETRAN Topical Report VEP -FRD-4 1, (Reference 1) describes Dominion's history with the use and application of RETRAN, dating from the early days of code development. The report also provides Page 21. of 35

DOM-NAF-5, Rev. 0.0-A a detailed description of the three-loop base models that have been developed for Dominion's Surry and North Anna plants.

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DOM-NAF-5, Rev. 0.0-A 3.4.1.4 Evolution of the Kewaunee Models A KPS RETRAN-02 model was developed in the mid-1980's. Prior to development of the KPS RETRAN-02 input, the DYNODE-P code was used to address plant operation, licensing, and design change issues, and to support the reload core design and safety evaluation. The KPS RETRAN-02 model was developed as the RETRAN code was becoming an industry standard for transient safety analysis.

In 2000, the RETRAN-02 input model was converted to a RETRAN-3D input model so it could be used to support steam generators replacement (SGR) and the need to reanalyze the design basis transients for the SGR project. As part of the conversion to RETRAN-3D, a detailed steam generator model was incorporated to replace the RETRAN-02 model single -node steam generator model. The RETRAN-3D model was also upgraded to include the replacement steam generator geometry and associated plant changes.

To support changing from the DYNODE-P code to the RETRAN-3D code for safety analysis and plant support, the RETRAN-3D model was qualified by benchmarking the RETRAN-3D KPS model results to DYNODE results for several USAR Chapter 14 transients.

In the fall of 2005, the KPS RETRAN-3D model was used to address an issue regarding primary- and secondary-side thermal-hydraulic response to a feedwater line break in conjunction with a reactor coolant pump seal failure, and to determine whether the core would remain covered. As part of this project, the RETRAN-3D model was modified to reflect a power uprate from 1683 to 1772 MWt, including the reactor protection system (RPS) and the engineered safety features (ESF).

After the acquisition of KPS by Dominion, the KPS RETRAN-3D model was modified to conform to the Dominion RETRAN Topical Report for the RETRAN-02 models of the Surry and North Anna plants, and overlay decks were prepared to be consistent with the Surry and North Anna overlay decks.

The KPS RETRAN model changes and conversion from RETRAN-3D to RETRAN-02 were performed based on the following governing principles:

a) The KPS RETRAN model is consistent with the NRC-approved methods used in VEP-FRD-41 (Reference 1). There are a few minor differences:

i The KPS model explicitly models the safety injection accumulators.

ii The KPS model has separate volumes for the steam generator inlet and outlet plenums, where the Reference 1 models lump these volumes with the hot leg and pump suction leg respectively.

iii. The KPS model treats the pressurizer spray line with explicit volumes and junctions. In the Reference 1 models, the spray is modeled as a pair of fill junctions with boundary conditions controlled by the control system model.

iv. The KPS models have an explicit volume representation of the main feedwater piping from the point where the auxiliary feedwater piping ties in. In the Reference 1 models, the time to purge residual hot main feedwater from the feedwater lines following auxiliary feedwater initiation is modeled using control blocks.

b) The KPS RETRAN model does not violate any restrictions and limitations of use presented in the RETRAN-02 Safety Evaluation Report (SER) (Reference 13).

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DOM-NAF-5, Rev. 0.0-A c) The model is as consistent as possible with the input parameters and methods used for the KPS USAR Chapter 14 transient analysis.

d) Wherever possible, the KPS RETRAN model is consistent with the other Dominion RETRAN models (similar nodalization and node numbers). Adopting a consistent noding and numbering scheme facilitates use of the models by analysts who are familiar with the Surry and North Anna plant models. As noted in (a) above, there are a few minor differences.

3.4.2 Conditions and Limitations Appendix 7 of Reference 1 provides a detailed discussion of the conformance of Dominion's Surry and North Anna RETRAN models to the restrictions, limitations and conditions of use imposed by NRC staff in the generic RETRAN code Safety Evaluation Reports (SERs) issued for the RETRAN code Topical Report. Based on the principles cited above for development of the KPS model, the Reference 1 assessment is applicable to the KPS model. Table 3.4.1 lists the KPS-specific evaluations of the NRC RETRAN code restrictions and limitations for which further explanation was warranted for application of VEP-FRD-41 to KPS.

Table 3.4.1 KPS-Specific Evaluation of the USNRC Generic RETRAN Code Restrictions &

Limitations RETRAN02 Mod 002 VEP-FRD-41 Evaluation Kewaunee Disposition Restrictions Dominion does not propose to a) Conservative usage of Rod ejection performed apply VEP-NFE-2 methods to a-D kinetics must be with point kinetics per KPS at this time. Rod ejection demonstrated Dominion Topical analyses will continue to be Report VEP-NFE-2. done with approved Westinghouse methods.

Dominion will apply hot pin model with metal-water Model only used for rod reaction to KPS analyses of e) Metal-water reaction ejection hot pin model. rod withdrawal from subcritical will have to be justified for Justification/ and locked reactor coolant sphae analysesinDmioTpca specific oificednalysebenchmarking provided pump rotor events.

in Dominion Topical Report VEP-NFE-2. Justification/benchmarking is provided in Dominion Topical Report VEP-NFE-2.

RETRAN02 No differences between VEP-MOD003/004 Restrictions FRD-41 and KPS models have been identified.

RETRAN02 No differences between VEP-MOD005.0 Restrictions FRD-41 and KPS models have been identified.

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DOM-NAF-5, Rev. 0.0-A 3.4.3 Assessment RETRAN is approved for application to KPS with the Westinghouse 422 V+ fuel design and is used for the current KPS safety analyses of record (References 15, 16, 17 and 18). RETRAN was also approved for application to KPS prior to the recent implementation of the Westinghouse fuel design (Reference 19).

KPS and other Dominion units (Surry, North Anna and Millstone Unit 3) use Westinghouse designs for nuclear steam supply system (NSSS) and reactor protection system (RPS). SPS, NAPS and KPS have many design and operating similarities.

The design basis transients and accidents for KPS are similar to the design basis transients and accidents for the other Dominion Westinghouse units. The range of KPS transients and accidents is bounded by those transients and accidents addressed in Reference 1.

The reactor thermal hydraulic conditions of the design basis transients and accidents are similar between KPS and the other Dominion units, and are within the qualification and capability of the RETRAN code, as demonstrated in References 15, 16, 17, 18 and 19.

Validation of the Dominion KPS RETRAN model involves comparison of RETRAN calculations to the KPS analysis of record for selected transients using the following strategy:

a) Identify unique classes of events (RCS heatup, RCS cooldown/depressurization, reactivity excursion, loss of RCS flow, and loss of secondary heat sink).

b) Select transients that represent the range of transient responses generated by these events.

c) Perform demonstration analyses of selected events to validate the capability to model key phenomena.

d) Verify that applicability assessment criteria are met:

i. Key phenomena are appropriately modeled and predicted ii Predicted results are technically sound and are in reasonable agreement with the KPS USAR analyses of record (or differences are understood and assessed as acceptable) iii. General trends in key parameters are consistent with USAR analyses of record 3.4.4 Summary Dominion's RETRAN methods (Reference 1) are determined to be applicable to KPS and can be applied to KPS licensing analysis for reload core design and safety analysis. The applicability of these methods is to be further demonstrated in a supplement to DOM-NAF-5 (to be included as Attachment B) by providing:

a) A base model noding diagram and region descriptions b) Results of benchmarking comparisons to the analyses of record for selected transients Page 25 of 35

DOM-NAF-5, Rev. 0.0-A 3.5 Applicability Assessment of Statistical DNBR Evaluation Methods - VEP-NE-2, "Statistical DNBR Evaluation Methodology" 3.5.1 Description of Methodology Topical Report VEP-NE-2, "Statistical DNBR Evaluation Methodology," (Reference 3) describes Dominion's methodology for statistically treating several of the important uncertainties in Departure from Nucleate Boiling Ratio (DNBR) analysis. Previously, these uncertainties were treated in a conservative deterministic fashion, with each parameter assumed to be simultaneously and continuously at the worst point in its uncertainty range with respect to DNBR. The Statistical DNBR Evaluation Methodology is used to determine a plant-specific and fuel-specific statistical DNBR limit. This limit combines the correlation uncertainty with the DNBR sensitivities to uncertainties in key DNBR analysis'input parameters. The statistical combination of some of these uncertainties permits a more realistic combination of the independent uncertainties and thus provides a more realistic evaluation of DNBR margin. Even though the statistical DNBR limit (the Statistical Design Limit or SDL) is larger than the deterministic DNBR limit (the Deterministic Design Limit or DDL), its use is advantageous. The Statistical DNBR Evaluation Methodology allows thermal hydraulic evaluations to be performed using nominal operating conditions as opposed to deterministic initial conditions (nominal conditions plus evaluated uncertainty).

In the performance of in-house DNB thermal-hydraulic evaluations, design limits and safety analysis limits are used to define the available retained DNBR margin for each application. The difference between the safety analysis (self-imposed) limit and the design limit is the available retained margin. For deterministic DNB analyses, the DDL is set equal to the applicable code/correlation limit (see Section 3.6). For statistical DNB analyses, the design DNBR limit is set equal to the applicable statistical design limit (SDL).

The Statistical DNBR Evaluation Methodology will be applied to all Condition I and HIDNB events (except Rod Withdrawal from Subcritical, RWSC), and to the Loss of Flow analysis, the Locked Rotor Accident and the Single Rod Cluster Control Assembly Withdrawal at Power, SRWAP. The events modeled statistically (see Table 3.5.1) are limited by the statistical design limits (SDLs) evaluated in the implementation of the Statistical DNBR Evaluation Methodology for KPS, which will be submitted for NRC review and approval.

In addition, there are events that will be evaluated with deterministic models. These events will be initiated from bounding operating conditions considering the nominal value and the appropriate uncertainty value, and require the application of the bypass flow, F? HN (measurement component) and F? HE (engineering uncertainties component), etc. The events modeled deterministically are limited by the deterministic design limits (DDLs) stated in DOM-NAF-2 (Reference 20).

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DOM-NAF-5, Rev. 0.0-A Table 3.5.1 USAR Transients Analyzed with VLPRE-D/WRB-1 for KIPS ACCIDENT KPS USAR APPLICATION SECTION Rod cluster control assembly bank withdrawal from DET-DNB subcritical 14.1.1 Rod cluster control assembly bank withdrawal at power 14.1.2 STAT-DNB Rod cluster control assembly misalignment / Dropped 14.1.3 STAT-DNB rod/bank Uncontrolled boron dilution 14.1.4 Non-DNB Full and partial loss of forced reactor coolant flow 14.1.8 STAT-DNB Startup of an inactive reactor coolant loop 14.1.5 Non-DNB Loss of external electrical load and/or turbine trip 14.1.9 STAT-DNB Loss of normal feedwater 14.1.10 Non-DNB Loss of offsite power 14.1.12 Non-DNB Excessive heat removal due to feedwater system 14.1.6 STAT-DNB malfunction Excessive load increase 14.1.7 STAT-DNB Rupture of a main steam pipe 14.2.5 DET-DNB Locked reactor coolant pump rotor or shaft break 14.1.8 STAT-DNB 3.5.2 Conditions and Limitations Topical Report VEP-NE-2 was reviewed and generically approved by the NRC in May 1987. The fuel-specific and plant-specific implementation of the VEP-NE-2 methodology must be submitted to the NRC for review and approval. Therefore, a plant-specific implementation of the Statistical DNBR Evaluation Methodology for Kewaunee Power Station will be submitted.

The NRC SER for VEP-NE-2 listed the following conditions that must be met by any plant-specific implementation of this generic methodology:

a) The selection and justification of the Nominal Statepoints used to perform the plant-specific implementation must be included in the submittal.

b) The justification of the distribution, mean and standard deviation for all the statistically treated parameters must be included in the submittal.

c) The justification of the value of model uncertainty must be included in the plant-specific submittal.

d) For the relevant critical heat flux (CHF) correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submittal, unless there is an approved Topical Report documenting them.

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DOM-NAF-5, Rev. 0.0-A 3.5.3 Assessment A DNBR evaluation method involving the statistical treatment of uncertainties is currently approved for application to KPS with the Westinghouse 422 V+ fuel design using the Westinghouse WRB- 1 correlation (References 15, 17 and 22). This DNB evaluation method (called the Revised Thermal Design Procedure, RTDP) is used for the current safety analyses of record for KPS. The RTDP is similar to Dominion's Statistical DNBR Evaluation Methodology.

The plant-specific, fuel-specific implementation of the Dominion Statistical DNBR Evaluation Methodology to KPS cores will be submitted for NRC review and approval. This submittal will provide the specific justification for Conditions (a), (b) and (c) cited in Section 3.5.2 above. Although no specific justification is provided herein for (a), (b) and (c), the application of the Westinghouse RTDP methodology clearly demonstrates that these conditions can be justified for the application of the Dominion Statistical DNBR Evaluation Methodology. The implementation of this methodology to KPS cores will result in a Statistical Design Limit (SDL) that is plant-specific and fuel-type specific. Since Appendix B to Topical Report DOM-NAF-2 has been approved by NRC (Qualification of the VIPRE-D/WRB-1 code/correlation pair), condition (d) has been met for the Westinghouse 422V+ fuel. Should Dominion elect to load in the KPS core a fuel /

product that uses a CHF correlation not previously qualified with the VIPRE-D computer code, a new submittal would be provided for the plant-specific and fuel-specific application of the Statistical DNBR Evaluation Methodology.

3.5.4 Summary Statistical DNB evaluation methods are determined to be applicable to KPS and can be applied to KPS licensing analysis for reload core design and safety analysis. A plant-specific and fuel-specific application for Kewaunee Power Station cores containing Westinghouse 422 V+ fuel assemblies will be submitted to the NRC for review and approval.

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DOM-NAF-5, Rev. 0.0-A 3.6 Applicability Assessment of VIPRE-D Methods - DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code" 3.6.1 Description of Methodology The basic objective of core thermal-hydraulic analysis is the accurate calculation of reactor coolant conditions to verify that the fuel assemblies constituting the reactor core can safely meet the limitations imposed by departure from nucleate boiling (DNB) considerations. DNB, which could occur on the heating surface of the fuel rod, is characterized by a sudden decrease in the heat transfer coefficient with a corresponding increase in the fuel rod surface temperature. DNB is a concern in reactor design because of the possibility of fuel rod failure resulting from the increased fuel rod surface temperature. In order to preclude potential DNB-related fuel damage, a design basis is established and is expressed in terms of a minimum departure from nucleate boiling ratio (MDNBR). The departure from nucleate boiling ratio (DNBR) is the ratio of the predicted heat flux at which DNB occurs (i.e. the critical heat flux, CHF) and the local heat flux of the fuel rod. By imposing a DNBR design limit, adequate heat transfer between the fuel cladding and the reactor coolant is assured. If the MDNBR is greater than the design limit, there is adequate thermal hydraulic margin within the reactor core. Thus, the purpose of core thermal-hydraulic DNB analysis is the accurate calculation of DNBR in order to assess and quantify core thermal margin.

VIPRE-D is the Dominion version of the computer code VIPRE (Versatile Internals and Components Program for Reactors), developed for EPRI (Electric Power Research Institute) by Battelle Pacific Northwest Laboratories in order to perform detailed thermal-hydraulic analyses to predict CHIF and DNBR of reactor cores. VIPRE-D, which is based upon VIPRE-0 1, MOD-02. 1, was adapted by Dominion for the specific analysis needs of the various Dominion nuclear power stations and their different fuel designs. The main enhancement made to VIPRE-01, MOD-02.1 to obtain VIPRE-D is the addition of several vendor proprietary CHF correlations. Additional customnizations were made in VIPRE-D's input and output to integrate it into Dominion's thermal hydraulic methodologies.

Topical Report DOM-NAF-2, "Reactor Thermal Hydraulics using the VIPRE-D Computer Code,"

(Reference 2) describes Dominion's use of the VIPRE-D computer code and the justification of all input, default parameters, and the specific modeling choices selected by Dominion. DOM-NAF-2 demonstrates that the VIPRE-D core thermal-hydraulics methodology is appropriate for pressurized water reactor (PWR).

licensing applications. In addition, the various appendices to Topical Report DOM-NAF-2, document the qualification of several CHF correlations with the Dominion VIPRE-D computer code, as well as their associated code/correlation deterministic design limits. Topical Report DOM-NAF-2, including several appendices, received generic NRC approval (Reference 21) and, as such, the VIPRE-D core thermal hydraulics methodology can be used for any of Dominion's nuclear facilities.

3.6.2 Conditions and Limitations The conditions and limitations associated with the implementation of Topical Report DOM-NAF-2 can be split into three groups. The first group of conditions and limitations is related to the general use of the VIPRE code, and its maintenance, and were imposed by the NRC in the SER for VIPRE-01 (References 24 and 25).

1.A. The application of VIPRE-D is limited to PWR licensing calculations modeling heat transfer regimes up to CHF. VIPRE-D will not be used for post-CHF calculations or for BWR calculations.

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DOM-NAF-5, Rev. 0.0-A 1..B. VIPRE-D analyses will only use DNB correlations that have been reviewed and approved by the NRC. The VIPRE-D DNBR calculations will be within the NRC-approved parameter ranges of the DNB correlations, including fuel assembly geometry and grid spacers. The correlation DNBR design limits associated with these approved CHF correlations will be derived or verified using fluid conditions predicted by the VIPRE-D code. Each CHF correlation used will be qualified or verified in the appropriate appendixes to Topical Report DOM-NAF-2.

1.C. Any plant-specific, fuel-specific application of the DOM-NAF-2 methodology will strictly use the modeling choices approved in Topical Report DOM-NAF-2, which describes the intended uses of VIPRE-D for PWR licensing applications, and provides justification for Dominion's specific modeling assumptions, including the choice of two-phase flow models and correlations, heat transfer correlations and turbulent mixing models.

1.D. The Courant number, which is based on flow velocity, time step, and axial node size, will be set at greater than 1.0 in VIPRE-D transient calculations whenever a subcooled void model is used to ensure numerical stability and accuracy.

L.E. VIPRE-D is maintained within Dominion's I0CFR50 Appendix B Quality Assurance program.

A second set of conditions and limitations is related to Dominion planned uses and applications for VIPRE-D, and were imposed by the NRC in the SER for DOM-NAF-2 (Reference 21). According to this group of conditions and limitations, VIPRE-D can be used for:

2.A. Analysis of 14x14, 15xl 5 and 17x17 fuel in PWR reactors.

2.B. Analysis of DNBR for statistical and deterministic transients in the Updated Safety Analysis Report (USAR). Additional DNBR transients that are plant-specific may be analyzed in a plant specific application.

2.C. Steady state and transient DNB evaluations.

2.D. Development of reactor core safety limits (also known as core thermal limit lines, CTL).

2.E. Providing the basis for reactor protection setpoints.

2.F. Establishing or verifying the deterministic code/correlation DNBR design limits of the various DNB correlations in the code. Each one of these DNBR limits would be documented in an appendix to the original DOM-NAF-2 Topical Report.

The third and final set of conditions and limitations is related to the plant-specific and fuel-specific application of the VIPRE-D methodology:

3.A. Changes to the Technical Specifications (TS) to add Topical Report DOM-NAF-2 and applicable approved appendixes to the plant Core Operating Limit Report.

3.B. A plant-specific and fuel-specific Statistical Design Limit(s) for the relevant code/correlation pairs, to be used in statistical evaluations, which is evaluated following the Statistical DNBR Evaluation Methodology (see Section 3.5).

3.C. Any TS changes related to Over-Temperature Delta-T (OTAT), Over-Power Delta-T (OPAT), or other reactor protection function, as well as revised reactor core safety limits.

3.D. Changes to the list of USAR transients for which the code/correlations and limits apply.

3.6.3 Assessment KPS is a standard 2-loop PWR that uses 14x14 fuel. This is one of the approved applications of the DOM-NAF-2 methodology according to conditions L.A and 2.A.

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DOM-NAF-5, Rev. 0.0-A Conditions 1.B and 2.F are met by the use of the WRB-1/W-3 CHF correlations (which are NRC-approved), with the design limits and ranges of applicability listed in Reference 20. The WRB-1 correlation has been qualified with Dominion's VIPRE-D computer code in Appendix B to Topical report DOM-NAF-2 (Reference 20). A DNBR design limit of 1.17 was obtained for VIPRE-D/WRB- 1 that yields a 95% non-DNB probability at a 95% confidence level. The range of validity for VIPRE-D/WRB-1 is also listed in Reference 20. The Westinghouse W-3 correlation will be used when the local conditions fall outside the range of applicability of the WRB-1 correlation. Specifically, the W-3 correlation will be applied to the lower portion of the fuel assemblies (below the first mixing vane grid) and in low-pressure events, such as main steam line break (MSLB). The DNBR design limit for W-3 is 1.45 for pressures between 500 to 1000 psia and 1.3 for pressures above 1000 psia. This application was specifically approved in Reference 21.

The application of DOM-NAF-2 is approved for all NRC-approved PWR fuel types (Reference 21). Should Dominion elect to load in the KPS core a fuel product that uses a C-F correlation not previously qualified with VIPRE-D, a submittal would be made to the NRC in accordance with DOM-NAF-2 to qualify the new CHF correlation with the VIPRE-D computer code and provide the associated code/correlation deterministic design limit.

Dominion has developed VIPRE-D models for Kewaunee cores containing Westinghouse 422 V+ fuel.

These models use all the modeling inputs approved in Topical Report DOM-NAF-2, including two-phase flow models and correlations, heat transfer correlations and turbulent mixing models, thus meeting conditions 1.C and 1.D. Should Dominion elect to load in the KPS core a different fuel product, Dominion would develop new VIPRE-D models for Kewaunee cores containing the new fuel product, and these models would strictly follow all the modeling guidelines specified in Topical report DOM-NAF-2, thus meeting conditions L.C and 1.D.

These models are used to evaluate the DNB-related design basis transients and accidents (Table 3.5.1). The KPS transients and accidents are a subset of the ones listed in Table 2.1-1 of Topical report DOM-NAF-2.

The reactor thermal hydraulic conditions of the design basis transients and accidents are similar between KPS and the other Dominion units and within the qualification and capability of the VIPRE-D code.

Therefore, conditions 2.B, 2.C, 2.D and 2.E are met.

These models are also used to evaluate the plant-specific and fuel-specific Statistical Design Limit (SDL) within the context of the Statistical DNBR Evaluation Methodology, which will be submitted for NRC review and approval (see Section 3.5). These models will also be used to verify the Kewaunee setpoint functions, core thermal limit lines and USAR statepoint and transient analyses. Any changes to, the Kewaunee reactor protection system setpoints and core thermal limit lines will be evaluated per the provisions of 10CFR50.59.

Conditions 3.A, 3.B, 3.C, and 3.D are thus met for application of DOM-NAF-2 methods to KPS.

3.6.4 Summary The DOM-NAF-2 core thermal hydraulic analysis methodology, including the applicable appendices, can be used for the thermal hydraulic evaluation of Kewaunee power station cores containing NRC-approved PWR fuel (currently Westinghouse 422 V+ fuel). The methods therein are determined to be applicable to KPS and can be applied to KPS licensing analysis for reload core design and safety analysis.

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DOM-NAF-5, Rev. 0.0-A 4.0 Conclusions and Implementation 4.1 Conclusions Dominion nuclear core design and safety analysis methods were assessed for applic ability to KPS. The Dominion reload nuclear design methods, as documented in the Dominion Topical Reports below, were determined to be applicable to KPS, and can be employed in the licensing design and evaluation of reload cores for KPS. The bases for this conclusion are provided in the Section 3.0 methodology applicability assessments.

• VEP-FRD-42, "Reload Nuclear Design Methodology" (Reference 6)

• VEP-NE- 1, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications" (Reference 8)

" DOM-NAF-1, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations" (Reference 9)

• VEP-FRD-41, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code" (Reference 1)

" VEP-NE-2, "Statistical DNBR Evaluation Methodology" (Reference 3)

• DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code" (Reference 2)

The applicability of the RETRAN and CMS methods to KPS will be further demonstrated through detailed validation analyses that will be documented in a supplement to DOM-NAF-5 (see Sections 4.2.3 and 4.2.4). In addition, the plant-specific and fuel-specific application analysis to define the DNBR Statistical Design Limit (SDL) will be completed to support the applicability of the Statistical DNBR methods to KPS (see Section 4.2.2).

KPS and other Dominion units (Surry, North Anna and Millstone Unit 3) use Westinghouse designs for nuclear steam supply system (NSSS) and reactor protection system (RPS). KPS has many design and operating similarities with the other Dominion Westinghouse units. KPS plant-specific considerations and features were evaluated and the differences from the methods as described in the Dominion Topical Report required for the application to KPS were identified in Section 3.0. The identified differences do not affect the conclusions on applicability of the methods to KPS.

Dominion analysis methods will be applied to KPS consistent with the conditions and limitations described in the Dominion Topical Reports and in applicable NRC Safety Evaluation Reports (SER). The conditions and limitations for each method were addressed in Section 3.0. The conditions and limitations will be met when the method is applied to KPS.

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DOM-NAF-5, Rev. 0.0-A 4.2 Steps for DOM NAF-5 Implementation The following steps are necessary to fulfill the applicability assessment for Dominion nuclear core design and safety analysis methods to KPS. Accomplishing these steps enables DOM-NAF-5 to be cited as the methodology reference in the KPS Technical Specification 6.9.a.4 and COLR.

4.2.1 Submit a license amendment request (LAR) to add DOM-NAF-5-A to Section 6.9.a.4 of the KPS Technical Specifications. Other conforming Technical Specification changes are to be incorporated into the LAR, as needed, to reflect use of Dominion methods.

4.2.2 Submit, as part of the step 4.2.1 LAR, a plant-specific and fuel-specific application analysis to define a DNBR Statistical Design Limit (SDL) per the provisions of VEP-NE-2 and DOM-NAF-2. The scope of the analysis is defined in Section 3.5 and 3.6 of this report.

4.2.3 Submit, as Attachment A of this report, detailed validation analyses for application of CMS methods (DOM-NAF-1). The scope of the analysis is defined in Section 3.3.

4.2.4 Submit, as Attachment B of this report, detailed validation analyses for application of RETRAN methods (VEP-FRD-41). The scope of the analysis is defined in Section 3.4.

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DOM-NAF-5, Rev. 0.0-A 5.0 References

1. Topical Report VEP-FRD-41, Rev. 0.1-A, "Vepco Reactor System Transient Analyses Using the RETRAN Computer Code," June 2004.

2. Topical Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code,"

September 2004.

3. Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," June 1987.

4. Deleted.

5. Deleted.

6. Topical Report VEP-FRD-42, Rev. 2. 1-A, "Reload Nuclear Design Methodology," August 2003.

7. Letter from Scott Moore (NRC) to D. A. Christian (VEPCO), "Acceptance of Topical Report VEP-FRD-42, Revision 2, Reload Nuclear Design Methodology, North Anna and Surry Power Stations, Units 1 and 2,"

June 11, 2003.

8. Topical Report VEP-NE- 1, Rev. 0. I-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications," August 2003.

9. Topical Report DOM-NAF-1, Rev. 0.0-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," June 2003.

10. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (Proprietary), March 1978.

11. WCAP 10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control-FQ Surveillance Technical Specification," February 1994.

12. EPRI NP-1850-CCM-A, "RETRAN-02, A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, Revision 6," December 1996.

13. Letter from C. 0. Thomas (NRC) to T. W. Schnatz (UGRA), "Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, "RETRAN-A Program for One Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems," and EPRI NP-1850-CCM, "RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," September 4, 1984.

14. Deleted.

15. Letter from J. G. Lamb (NRC) to T. Coutu (NMC) transmitting the NRC SER for Amendment No. 167 to the Operating License, revising TSs for use of Westinghouse 422 VANTAGE + nuclear fuel with PERFORMANCE + features, April 4, 2003.

16. Letter from J.G. Lamb (NRC) to T. Coutu (NMC) transmitting the NRC SER for Amendment No. 168 to the Operating License, approving the Measurement Uncertainty Recapture (MUR) Power Uprate (1.4%),

Letter No. K-03-094, July 8, 2003.

17. Letter from J.G. Lamb (NRC) to T. Coutu (NMC) transmitting the NRC SER for Amendment No. 172 to the Operating License, approving the 6% Stretch Power Uprate, Letter No. K-035, February 27, 2004.

Page 34 of 35

DOM-NAF-5, Rev. 0.0-A

18. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis," (Proprietary), April 1999.

19. Letter from G. Lamb (NRC) to M.E. Reddemann (NMC), transmitting the NRC SER approving WPSRSEM-NP Revision 3, "Kewaunee Nuclear Power Plant - Review for Kewaunee Reload Safety Evaluation Methods Topical Report," Letter No. K 112, September 10, 2001.

(

20. DOM-NAF-2, Appendix B, "Qualification of the Westinghouse WRB- 1 CIF Correlation in the Dominion VIPRE-D Computer Code," December 2004.

21. Letter from C.I. Grimes (NRC) to D. A. Christian (VEPCO) transmitting approval of Topical Report DOM-NAF-2, 'Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," April 4, 2006 and letter from S. R. Monarque (NRC) to D. A. Christian (VEPCO) transmitting corrected pages for the NRC SER, June 23, 2006.

22. WCAP-8762-P-A, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," F. E. Motley, et al., July 1984.

23. Letter from E. S. Grecheck, Virginia Electric and Power Company, to USNRC, "Response to Request for Additional Information, Dominion's Reload Nuclear Design Methodology Topical Report," December 2, 2002.

24. Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), "Acceptance for Referencing of Licensing Topical Report, EPRI NP-251 1-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

25. Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-0l Maintenance Group), "Acceptance for Referencing of the Modified Licensing Topical Report, EPRI NP-251 1-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores,' (TAC No. M79498)," October 30, 1993.

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DOM-NAF-5-Rev. 0.0-A DOM-NAF-5-0.0-A Attachment A CMS Benchmarking Information Prepared by:

Christopher J. Wells Robert A. Hall Walter A. Peterson

,Oý Applo;ra W. hlei e SupeMsor, IUOCuclear Core Design 1

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 2 of 40 Table of Contents TABLE OF CONTENTS ............................................................................................................ 2 LIST OF TABLES ............................................................................................................................ 3 LIST OF FIGURES .......................................................................................................................... 4 SECTION 1 -INTRODUCTION AND

SUMMARY

................................................................ 5 1.1 INTRO D UCT IO N .......................................................................................................................... 5 1.2 S UM MARY .................................................................................................................................. 6 SECTION 2- STATISTICAL METHODS ............................................................................... 8 2.1 NULL HYPOTHESIS TESTS FOR NORMALITY ............................................................................. 8 2.2 DETERMINING TOLERANCE LIMITS ASSUMING NORMALITY ................................................ 8 2.3 DETERMINING NON-PARAMETRIC TOLERANCE LIMITS ......................................................... 9 SECTION 3- HIGHER ORDER CODE BENCHMARKING ............................................... 10 3.1 CASMO BENCHMARKING TO HIGHER ORDER CALCULATIONS ..................... 10 3.2 SIMULATE BENCHMARKING To HIGHER ORDER CALCULATIONS ................................. 12 SECTION 4- SIMULATE BENCHMARKING TO MEASURED DATA ........................... 15 4.1 CRITICAL BORON CONCENTRATION .................................................................................. 16 4.2 INTEG RAL ROD W ORTH .......................................................................................................... 19 4.3 PEAK DIFFERENTIAL ROD WORTH ................................................................................... 21 4.4 ISOTHERMAL TEMPERATURE COEFFICIENT ..................................................................... 23 4.5 DIFFERENTIAL BORON WORTH ........................................................................................ 25 4.6 DOPPLER COEFFICIENTS AND DEFECTS ........................................................................... 28 4.7 REACTION RATE COMPARISONS ........................................................................................ 29 4.8 DELAYED NEUTRON AND PROMPT NEUTRON LIFETIME DATA .......................................... 33 SECTION 5 - NORMAL OPERATION POWER TRANSIENTS ........................................ 34 SECTION 6- CONCLUSIONS ............................................................................................ 38 SECTION 7- REFERENCES ................................................................................................ 40 2

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 3 of 40 List of Tables Table Title Page 1 Kewaunee Nuclear Reliability Factors 6 2 CASMO-4 Reactivity Benchmarking Versus Monte Carlo Codes 11 3 CASMO-4 W-prime and Pin-to-box Ratio Comparisons 14 4 Critical Boron Tolerance Limits and Nuclear Uncertainty Factors 18 5 Integral Rod Worth Tolerance Limits and Nuclear Uncertainty Factors 20 6 Peak Differential Rod Worth Tolerance Limits and NUFs 22 7 Isothermal Temperature Coefficient Tolerance Limits and NUFs 24 8 Differential Boron Worth Tolerance Limits and Nuclear Uncertainty Factors 27 9 Integral and 32 Node Reaction Rate Tolerance Limits and NUFs 30 10 Comparison of North Anna/Surry and Kewaunee Nuclear Reliability Factors 38 3

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 4 of 40 List of Figures Figure Title Page 1 Critical Boron Comparison 18 2 Integral Rod Worth Comparison 20 3 Peak Differential Rod Worth Comparison 22 4 Isothermal Temperature Coefficient Comparison 24 5 Reactivity Computer Bias Influence 26 6 Differential Boron Worth Comparison 27 7 Partial Power & HFP Flux Map Integrated Reaction Rate Comparisons 31 8 Partial Power & HFP Flux Map 32 Axial Node Reaction Rate Comparisons 32 9 KPS C27 November 2005 Transient - Power and D-bank Changes 35 10 KPS C27 February 2006 Transient - Power and D-bank Changes 35 11 KPS C27 November 2005 Transient - Axial Flux Difference 36 12 KPS C27 February 2006 Transient- Axial Flux Difference 36 13 KPS C27 November 2005 Transient - Critical Boron Concentration 37 14 KPS C27 February 2006 Transient - Critical Boron Concentration 37 4

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 5 of 40 SECTION 1 - INTRODUCTION and

SUMMARY

1.1 Introduction The NRC approved topical report DOM-NAF-1,Rev. 0.0-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations" (Reference 1) details the Dominion methodology for validating the Studsvik/CMS core modeling code package for use in reactor physics calculations. The CMS system consists of CASMO-4, CMS-LINK, and SIMULATE-3 (described in detail in DOM-NAF-1). In DOM-NAF-1 the accuracy of the CMS system was demonstrated through comparisons with measurements from over 60 cycles of operation at the Surry and North Anna Nuclear Power Stations.

The benchmarking methodology, as described fully in DOM-NAF-1, consists of the following broad steps:

• Higher order codes are used to identify any biases in CMS model key parameters (such as control rod worth, burnable poison worth, fuel temperature (Doppler) defect and soluble boron worth). Most bias corrections are applied prior to assembling the SIMULATE cross-section library (Reference 1). Control rod corrections are applied using a grey factor in SIMULATE. Higher order codes are also. used to verify the accuracy of peak-to-average pin power calculations in CASMO and SIMULATE.

• SIMULATE-3 predictions are compared with measured data for the key physics parameters and normal operation power transients are modeled to test the integrated CMS models in a dynamic manner.

" Where applicable, Nuclear Uncertainty Factors (NUF) and Nuclear Reliability Factors (NRF) are derived for key physics parameters using statistical methods. For parameters that cannot be directly evaluated using statistical techniques the basis for arriving at conservative estimates for the NUF and NRF is presented.

5

.DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 6 of 40 1.2 Summary This report (DOM-NAF-5, Rev.0.0 Attachment A) documents the application of the DOM-NAF-1 methodology to Kewaunee Power Station. Ten cycles (cycles 18-27) of Kewaunee have been modeled with CMS for the benchmark effort. These ten cycles of operation span a wide range of design and operating history, including transitions in fuel enrichment, fuel density, spacer grid design and material, fuel vendor, core operating conditions (full-power average moderator temperature and rated thermal power) and burnable poison material and design.

The Kewaunee CMS models have been validated by a set of benchmarks to both higher order calculations and ten cycles of measured data. Based on these benchmarks, the following set of nuclear reliability factors (NRF) were determined to account for model predictive bias and uncertainty.

Table 1 Kewaunee Nuclear Reliability Factors NRF Parameter Upper Lower Integral Control Rod Bank Worth (Individual banks) 1.10 0.90 Integral Control Rod Bank Worth (Total of all banks) 1.10 0.90 Differential Control Rod Bank Worth 1.20 0.80 Critical Boron Concentration +70 ppm -70 ppm Differential Boron Worth 1.05 0.95 Isothermal and Moderator Temperature Coefficient +2 pcm/°F -2 pcm/°F Doppler Temperature Coefficient 1.10 0.90 Doppler Power Coefficient 1.10 0.90 Effective Delayed Neutron Fraction 1.05 0.95 Prompt Neutron Lifetime 1.05 0.95 FAH 1.04 N/A FQ 1.05 N/A 6

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 7 of 40 In addition to the Nuclear Reliability Factors, two Kewaunee normal operation power transients have also been modeled to demonstrate how well the CMS system can predicted simultaneous changes in multiple parameters such as core power, moderator temperature, fuel temperature, and control rod bank position. For both modeled transients SIMULATE provided accurate predictions of the timing and magnitude of overall reactivity (critical boron) and core axial power (axial offset).

The following sections present the details of the CMS benchmarking for Kewaunee Power Station. Since the basic methodology is the same as described in DOM-NAF-1, the results of the Kewaunee benchmark are presented directly while relying heavily on Reference 1 for background and supporting information.

7

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 8 of 40 SECTION 2- STATISTICAL METHODS The following sub-sections review the statistical methods used to construct tolerance intervals for the predicted versus measured plant data. The full discussion of these methods can be found in Reference 1.

2.1 Null Hypothesis Tests for Normality Since no single test for normality is particularly strong, the standard for accepting the hypothesis of normality for any sample data set is such that at least two out of three tests must indicate that the hypothesis of normality is not rejected. The normality tests employed for the Kewaunee predicted versus measured data are the same as used in Reference 1: Shapiro-Wilk, Kolmogorov-Smirnov, Kuiper, and D'Agostino (D') tests.

Note that not all tests apply to all sample sizes. All statistical normality tests are conducted at the 0.05 level of significance.

2.2 DeterminingTolerance Limits Assuming Normality From DOM-NAF-1-Rev.0.0-P-A, the one-sided tolerance limit (TL) for data that is assumed to be normally distributed is given by:

TL =Xm + (K* a) where Xm and a are the mean and standard deviation of the sample data set. The multiplier K is chosen such that 95% of the population is less than the value of TL, applied in a conservative direction, with a 95% confidence level. Several references (References 2, 3) describe how to obtain the multiplier K. The K values in this report were calculated using the following formula:

8

DOM-NAF-5-0.O-P-A Attachment A CMS Benchmarking Information pg 9 of 40 z +. P-ab a

2 a=1-2(n -1) 2 .Zy2 b=z pPn--

The values of z refer to the inverse of the cumulative normal distribution function. P refers to the population that is less than the TL (95% for our case), while y refers to the desired confidence level (also 95% in this case). Both zp(O.95) and Zy(O.95) evaluate to

-1.645.

2.3 DeterminingNon-parametricTolerance Limits If the assumption of normality for a certain data set cannot be supported then non-parametric methods must be used to determine the tolerance limits. The non-parametric ranking method of Somerville (Reference 4) is used to determine the mth value of a sorted list that bounds 95% of the population with 95% confidence. This method of computing distribution-free tolerance limits is also referenced in USNRC Regulatory Guide 1.126 (Reference 3).

9

DOM-NAF-5-0.0-P-A Attachment A CNIS Benchmarking Information pg 10 of 40 SECTION 3 - HIGHER ORDER CODE BENCHMARKING 3.1 CASMO Benchmarking To Higher OrderCalculations As part of the development of the Kewaunee models, Dominion has performed a comparison of CASMO and Monte Carlo code (MCNP-4B" and KENO-V.a) calculations of reactivity worth for soluble boron, gadolinia (gad) loading, AIC (silver-indium-cadmium) control rods, temperature defect, and Doppler defect. These comparisons identify any significant biases so that corrections may be applied prior to assembling the SIMULATE cross-section library (Reference 1). Control rod corrections are applied using a grey factor in SIMULATE.

Table 2 shows the results of the CASMO reactivity benchmarking. Statistical uncertainty associated with each Monte Carlo calculation was limited to a range of 0.00004 to 0.00034 AK (one standard deviation). The data represents a range of fuel enrichments from 3.0 to 5.0 w/o U-235, soluble boron concentration from 0 to 2000 ppm, and temperature from 100 to 547 'F. Doppler comparisons are for enrichments of 3.0 and 5.0 w/o U-235 (burned and new fuel) over a fuel temperature range of 300 to 900 K. Statistically significant biases are apparent for Doppler defect, control rod worth, and soluble boron worth. These results are consistent with results for both the 15xl 5 and 1 7x1 7 fuel from Reference 1. No statistically significant bias is apparent for gadolinia worth.

10

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 11 of 40 Table 2 CASMO-4 Reactivity Benchmarking Versus Monte Carlo Codes Parameter Mean Std. Deviation Number of

(% difference) (% difference) Observations AIC Control Ros1.70 0.6 24 Rods Soluble Boron -1.23 0.4 30 Worth Gadolinia 0.40 0.4 36 Doppler Defect -10.8 0.8 3 Note: % Difference is 100 x (CASMO WORTH - MONTE CARLO WORTH) / (MONTE CARLO WORTH) 11

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 12 of 40 3.2 SIMULATE Benchmarking To Higher OrderCalculations The method of comparing CASMO/SIMULATE versus Monte Carlo code calculations of pin-to-box ratios and flux thimble instrument reaction rate ratios used in combination with normalized flux map reaction rate comparisons to determine appropriate peaking factor (FAH and FQ) uncertainty factors is fully described in DOM-NAF-1.

In order to estimate the W-prime (normalized ratio of assembly power to flux thimble instrument reaction rate) and pin-to-box uncertainty for the Kewaunee models, five 2x2 assembly models and one 5x5 model have been constructed using both SIMULATE and MCNP. The 2x2 cases include combinations of:

1) Two fuel enrichments (4.0 & 5.0 w/o)

2) Two soluble boron concentrations (0 & 2000 ppm)

3) The maximum gadolinia loading used to date at KPS (20 pins @ 8 w/o Gd 20 3)

4) Control Rod Insertion

5) An approximation of once-burned fuel (4w/o U235 at 20 GWD/MTU)

The 14x14 Kewaunee assemblies are inherently asymmetric, with an off-center instrument thimble and asymmetric gadolinia loadings. In light of this, the 5x5 model is intended to test the ability of CASMO/SIMULATE to predict the effect of large flux gradients (approximately a factor of 2 diagonally across the target assemblies) on the W-prime for asymmetric assemblies on opposite sides of the core.

The set of cases modeled (5 2x2 cases and 1 5x5 case) produce large power gradients but do not encompass the full range of burnable poison loadings, fuel enrichments, and fuel burnup that can occur in face-neighbors. However, large power gradients across the fuel assembly faces produce large pin-to-box uncertainties, resulting in a more bounding analysis.

CASMO/SIMULATE uses a gamma smoothing technique to account for redistribution of fission energy released as gamma radiation. This method redistributes approximately 12

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 13 of 40 7% of the assembly power, effectively flattening the intra-assembly pin power distribution. The MCNP flux tallies do not account for gamma smoothing directly, therefore the MCNP pin-to-box ratios have been adjusted to include this effect. Both the CASMO/SIMULATE and Monte Carlo pin powers include the effects of gamma smoothing and are compared on an equal basis.

Results of the SIMULATE / MCNP model comparisons are presented in Table 3. All W-prime data was determined to be normal and all pin-to-box ratio data was determined to be non-normal using the methods described in Section 2. The one-sided tolerance limit for pin-to-box is -1.9% for all pins. These results are bounded by the 2% pin-to-box uncertainty determined in DOM-NAF-1. It is important to note that the KPS results could be improved by eliminatingnon-limiting low power pins (such as rodded assembly and gadolinia pins) as was done in DOM-NAF-1.

Including the MCNP uncertainty, the W-prime tolerance interval is 1.7%*. If the MCNP uncertainty is excluded via root sum square from the overall uncertainty then the W-prime tolerance interval is recalculated to be 1.5%, which is the same as the DOM-NAF-1 value. It is important to note that the results of the 5x5 case clearly demonstrate that the CMS models are able to accurately account for the effects of offset instrument thimble and asymmetric burnable poison loadings when constructing W-prime values.

For the 5x5 "core", MCNP predicted 4.4% difference in W-primes between cross core symmetric partners while SIMULATE predicted a difference of 4.6%.

The RSS (root sum square) combination of W-prime and pin-to-box uncertainty for use in determining measured peaking factors is 2.6% (1.026 multiplier) using the conservative W-prime tolerance and 2.5% (1.025 multiplier) using the W-prime tolerance with the MCNP uncertainty removed. These values are in line with the 3.0%

and 2.5% RSS combination of W-prime values obtained for the 15x1 5 / 17x1 7 fuel in Reference 1.

• A conservatively low estimate of the MCNP statistical uncertainty for W-prime is 0.4% based on MCNP case output.

13

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 14 of 40 Table 3 CASMO-4 W-prime and Pin-to-box Ratio Comparisons Sample Mean (%) Std. Dev. Tolerance Parameter Size (%) Normal Limit Pin-to-Box* 7876 0.06 1.26 No -1.9%

W-prime (including 44 0.01 0.82 Yes 1.7%

MCNP Uncertainty)

W-prime (Excluding 44 0.01 0.72 Yes 1.5%

MCNP Uncertainty)

Note: Difference is ((SIMULATE- MCNP) / SIMULATE) x 100%

• Both the Monte Carlo and CASMO/SIMULATE models are adjusted to include the effects of gamma smearing.

14

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 15 of 40 SECTION 4- SIMULATE BENCHMARKING TO MEASURED DATA The following sections present the results of comparisons of SIMULATE-3 predictions with measurements from Kewaunee Power Station. The calculations were performed using full core, 32 axial node, 2x2 X-Y mesh per assembly geometry.

All comparisons of SIMULATE predictions with measured data will by nature represent a combination of SIMULATE bias, SIMULATE uncertainty, measurement bias, and measurement uncertainty. These comparisons will be used to derive appropriate uncertainty factors for SIMULATE predictions. In cases where the comparison data lead to unrealistically high estimates for SIMULATE uncertainty, attempts to quantify and account for measurement bias and uncertainty will be made. Statistical methods are discussed in Section 2. Both the nuclear uncertainty factors and nuclear reliability factors for each parameter are presented in the following sections. In the statistics presented, the sign convention used is such that a positive value indicates over-prediction of the magnitude of a parameter by SIMULATE, and a negative value indicates under-prediction by SIMULATE. Percent differences are determined by the following: (SIM-Measured)/SIM x 100%.

The following sections are summaries of the benchmarking effort for each parameter, with the tolerance limits and Nuclear Uncertainty Factors presented directly, followed by the determination of Nuclear Reliability Factors. More discussion on the general effect of measurement bias and uncertainty on the total observed bias and uncertainty can be found in Reference 1.

15

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 16 of 40 4.1 CriticalBoron Concentration Table 4 presents the statistical benchmark data for critical boron concentration. Figure 1 presents the difference between SIMULATE and measured boron concentrations in histogram format. Since Kewaunee currently recycles its soluble boron, only boron measurements at EOC (< 50 ppm) and measurements for which there is a corresponding isotopic analysis of the RCS boron are compared against SIMULATE.

Based on Figure 1, the critical boron differences do not seem to come from a normal distribution. In fact, the boron data appears to be grouped into several separate distributions. From examining the underlying data (not presented), several observations are made about the critical boron comparisons:

1) Before Cycle 27, there is a consistent, tightly distributed bias centered around-28 ppm. This bias persisted through changes in fuel design, transition from discrete burnable poison to gadolinia, and transition from annual to 18 month operating cycles.

This is the "middle" distribution seen in the Figure 1 histogram.

2) During Cycle 27 it appears that the negative bias disappeared. The Cycle 27 boron differences are centered close to 0 ppm and most observations fall within +/-10 ppm.

3) There are several data points that are very negative, grouped about the-50 ppm point. These samples are from cycles before 27. The number of samples in this subset is too small to determine if these points are outliers or are part of distinct third distribution.

From the NUF values alone, the NRF for critical boron could be set to some conservatively bounding value, say, +/-60 ppm. Although +/-60 ppm is supported by the benchmark data, there appears to be some unknown driver that is biasing the boron comparisons. This bias does not appear originate from CMS because in Cycle 27 the bias driver appears to have disappeared. Until the cause of the bias can be positively 16

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 17 of 40 identified or enough data is collected to determine that the driver has disappeared, the NRF for critical boron comparisons is conservatively set to +/-70 ppm.

17

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 18 of 40 Table 4 Critical Boron Tolerance Limits and Nuclear Uncertainty Factors (Nonparametric Distribution Assumed)

Mean Std. Number Sorted Upper Lower Min. NUF Max. NUF Dev. of Obs. List Tolerance Tolerance Value Value (ppm) (ppm) mth Limit Limit (ppm) (ppm) value (ppm) (ppm)

-20.1 17.3 93 2 11 -55 -11 55 Figure 1 Critical Boron Comparison SIMULATE Minus Measured 12-10- r3 End of Cycle Critical Boron n3 BI01B Atom Ratio Adjusted Critical Boron 0

.0 0

51 -46 211-41 -36 -31 -26 -21 -16 -11 -6 -1 5 10 PPM 18

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 19 of 40 4.2 IntegralRod Worth Table 5 presents the statistical benchmark data for integral rod worth. Figure 2 presents the percent difference between SIMULATE and measured integral rod worths in histogram format.

A nuclear reliability factor of +/-10% (multiplier range 0.90 to 1.10) is chosen for integral rod worth calculations as it bounds the NUF and is therefore conservative for SIMULATE integral rod worth calculations. Since the uncertainty for the total rod worth cannot be larger than the uncertainty for individual banks the NRF for total bank worth is also set to +/-10%.

19

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 20 of 40 Table 5 Integral Rod Worth Tolerance Limits and Nuclear Uncertainty Factors (Normality Assumed)

Mean Std. Number Std. Dev. Upper Lower Min. NUF Max. NUF

(%) Dev. of Obs. Multiplier Tolerance Tolerance Multiplier Multiplier

(%) K Limit(%) Limit (%)

-0.3 4.6 61 2.01 8.9 -9.5 0.91 1.095 Figure 2 Integral Rod Worth Comparison SIMULATE Minus Measured 1.

1:

Ii 0

0) 0l

-7 -5 -3 -1 1 3 5 7 9 11 13 15

% Difference 20

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 21 of 40 4.3 Peak DifferentialRod Worth Table 6 presents the statistical benchmark data for peak differential rod worth. Figure 3 presents the percent difference between SIMULATE and measured peak differential rod worths in histogram format.

The NRF multiplier range for Peak Differential Rod worth is set to bound the NUF at 0.80 to 1.20.

21

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 22 of 40 Table 6 Peak Differential Rod Worth Tolerance Limits and Nuclear Uncertainty Factors (Normality Assumed)

Mean Std. Number Std. Dev. Upper Lower Min. NUF Max. NUF

(%) Dev. of Obs. Multiplier Tolerance Tolerance Multiplier Multiplier

(%) K Limit (%) Limit (%)

0.1 6.6 9 2.99 19.8 -19.6 0.8 1.2 Figure 3 Peak Differential Rod Worth Comparison u)

.0

,0

-14 -10 -6 -2 10

% Difference 22

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 23 of 40 4.4 Isothermal Temperature Coefficient Table 7 presents the statistical benchmark data for the isothermal temperature coefficient. Figure 4 presents the difference between SIMULATE and measured isothermal temperature coefficients in histogram format. From the histogram, it appears one point is not grouped with the others. If this data point (a Cycle 20 predicted-measured comparison) is removed from the sample pool and sample statistics are computed, the point lies over 6a from the mean. Also, from documentation of the measurement, it appears that there may have been difficulty in obtaining the measured ITC. Therefore, the ITC difference data point for Cycle 20 is assumed to be an outlier and has been removed from the tolerance interval dataset.

A nuclear reliability factor of +/-2 pcm/°F conservatively bounds the NUF of -0.96 / +0.84 pcm/°F and is therefore considered appropriate.

23

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 24 of 40 Table 7 Isothermal Temperature Coefficient Tolerance Limits and Nuclear Uncertainty Factors (Normality Assumed, Cycle 20 Outlier Removed)

Mean Std. Num. Std. Dev. Upper Lower Min. NUF Max. NUF Dev. of Multiplier Tolerance Tolerance Value Value (pcm/°F) (pcm/°F) Obs. K Limit Limit (pcm/IF) (pcm/°F)

(pcml° F) (pcm/° F) 0.06 0.30 9 2.99 0.96 -0.84 -0.96 0.84 Figure 4 Isothermal Temperature Coefficient Comparison 3f I

I El Isothermal Temperature Coefficient I I

0 0

I- 7f

-1.75 -1.375 -1.125 -0,875 -0.625 -0.375 -0.125 0.125 0.375 Difference 24

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 25 of 40 4.5 Differential Boron Worth Use of measured data to determine differential boron worth (DBW) uncertainty can lead to unrealistically large uncertainty estimates. As discussed in DOM-NAF-1, reactivity computer bias can be a large portion of the DBW measurement uncertainty. Figure 5 clearly demonstrates the correlation between DBW error and RW error for 8 measurements at Kewaunee. The correlation coefficient of 81 % indicates a strong relationship between DBW error and RW error, which except for the reactivity computer are independent quantities. Also, based on the best estimate and standard error of the slope of the least squares fit line, the true slope of the fit line is most certainly greater than zero. Therefore, the data shows that reactivity computer bias is contributing significant uncertainty to the DBW measurements.

Figure 6 presents the difference between SIMULATE and measured differential boron worth in histogram format. Using the same methodology described in DOM-NAF-1, tolerance limits and NUF values for DBW have been calculated with the uncertainty due to reactivity computer measurements removed (Table 8). Even without reactivity computer bias, the derived NUF range is unreasonable.

Much like the North Anna and Surry DBW NUFs presented in DOM-NAF-1, the Kewaunee DBW NUFs do not seem consistent with other evidence we have about boron worth. The maximum HZP BOC critical boron concentration difference between SIMULATE and measured is -55 ppm. Based on the DBW NUF range however, it is expected we would see boron differences due to boron worth error alone in the range of 98 ppm (7% x 1400 ppm) to 276 ppm (12% x 2300 ppm). The fact that the worst observed HZP boron difference (-3%) does not fall within this range indicates that the true DBW uncertainty is significantly lower than the DBW statistics suggest.

Therefore, based primarily on the evidence of critical boron concentration difference data, a NUF and NRF of +/-5% (multiplier range of 1.05 to 0.95) is considered to be sufficiently conservative for SIMULATE differential boron worth predictions.

25

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 26 of 40 Figure 5 Reactivity Computer Bias Influence DBW % Difference Versus Rod Worth % Difference 15.0 10.0 4 DBW vs RW Difference 5.0 - Linear (DBW vs RW Difference) -___

5.0 j Q4

.: 0.0 ly =1.402d M -5.0 -

-10.0-*

-15.0

-10.0 -8.0 -6.0 -4.0 -2.0 0.0 2.0 4.0 6.0 8.0 Control Rod Worth % Difference 26

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 27 of 40 Table 8 Differential Boron Worth Tolerance Limits and Nuclear Uncertainty Factors (Normality Assumed, Reactivity Computer Influence on Variance Removed)

Mean Std. Number Std. Dev. Upper Lower Min. NUF Max. NUF

(%) Dev. of Obs. Multiplier Tolerance Tolerance Multiplier Multiplier

(%) K Limit(%) Limit(%)

-2.5 3.0 8 3.14 6.92 -11.92 0.93 1.12 Figure 6 Differential Boron Worth Comparison 3-2.5-Cl Differential Boron Worth 2-C-

0 U)

.0 1- A 0.5- 7

-8 -4 0 4 8 12

% Difference 27

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 28 of 40 4.6 Doppler Coefficients and Defects As discussed in DOM-NAF-1, direct determination of the NRF for Doppler feedback is very difficult. The value of 1.10 for the Doppler Temperature Coefficient and Doppler Power Coefficient that was proposed for North Anna and Surry models in DOM-NAF-1 is proposed for Kewaunee CMS predictions. There are three indications that support the use of a +/-1_0% NRF (multiplier range 1.10 to 0.90)

1) Benchmarking of CASMO Doppler Temperature Defects to Monte Carlo methods enables a best-estimate correction to be performed that effectively eliminates the theoretical CASMO Doppler bias (as determined using higher order calculations) from the SIMULATE model.

2) The Kewaunee fuel pellet diameter is almost identical to the Surry fuel pellet diameter. The Doppler feedback is therefore expected to be just as well predicted for Kewaunee as it currently is for Surry. DOM-NAF-1 presented data for Surry that supported the use of a +/-10% NRF.

3) The Kewaunee Cycle 27 February 2006 operational transient modeled in Section 5 includes an undamped xenon oscillation. The good agreement of the measured and predicted axial offset oscillation magnitude in the modeled operational transient demonstrates that the xenon-Doppler balance is reasonable.

28

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 29 of 40 4.7 Reaction Rate Comparisons Table 9 presents the statistical benchmark data for the reaction rate comparisons.

Figure 7 presents the percent difference between SIMULATE and measured integral reaction rates in histogram format. Figure 8 presents the percent difference between SIMULATE and measured 32 node reaction rates in histogram format.

Table 9 presents the combined tolerance limits (the pin-to-box tolerance limits root-sum-squared with the reaction rate tolerance limits) and NUFs for integral and nodal reaction rates. Since under-prediction of reaction rates is the only concern from a safety analysis standpoint, only the lower tolerance limit and corresponding NUF is presented.

Based on the Table 9 values, a NRF for FAH of 1.04 is conservative and an NRF for FQ of 1.05 is conservative.

29

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 30 of 40 Table 9 Integral and 32 Node Reaction Rate Tolerance Limits and Nuclear Uncertainty Factors (Nonparametric Distributions Assumed)

Data Mean Std. Number One Sided Pin-to-box Combined NUF Type (%) Dev. of Obs. Reaction Rate Tolerance Tolerance Multiplier

(%) Tolerance Limit  % Limit Integral 0.02 2.08 838 -3.03 -2.0 -3.63 1.036 32 Node -0.12 2.72 23333 -4.20 -2.0 -4.65 1.047 Note: The observations in this table are from a combination of partial power and HFP maps.

30

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 31 of 40 Figure 7 Partial Power & HFP Flux Map Integrated Reaction Rate Comparison 200 180- 03Kewaunee Integrated RR 160 140- 1 S120-0) 100 0on "

8 -7 -6 -5 3 -2 -1 0 1 2 3 4 5 6 7 8

% Difference 31

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 32 of 40 Figure 8 Partial Power & HFP Flux Map 32 Axial Node Reaction Rate Comparisons 3800-3600-3400-3200-3000-2800-2600-M Kewaunee RXR 2400-2 2200-0 2000 1800-100-0 16:00-24 20 16 12 8 -6 -4 -2 0 2 4 6 8 10 12 14 16 18 20 22 24

% Difference 32

DOM-NAF-5-0.0-P-A Attachment A CIMS Benchmarking Information pg 33 of 40 4.8 Delayed Neutron and PromptNeutron Lifetime Data The same effective delayed neutron and prompt neutron fraction NRFs developed in DOM-NAF-1 are also assumed for Kewaunee. A NRF of +/-5% (multiplier range of 1.05 to 0.95) is considered conservative for both of these values. The Tuttle delayed neutron set was used for all CASMO modeling of Kewaunee fuel.

33

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 34 of 40 SECTION 5- NORMAL OPERATION POWER TRANSIENTS Two Cycle 27 power maneuvers were modeled with SIMULATE. These maneuvers are useful for demonstrating the ability of the CMS model to accurately predict core behavior involving combinations of large power changes, control rod movements, temperature variations, xenon worth changes, and boron changes. Figures 9 and 10 show the time dependent power and D-bank positions for the 11/2005 and 2/2006 events, respectively.

Results of the modeling (Figures 11 and 12) showthat SIMULATE axial offset changes closely match the pattern of the measured ex-core axial offsets. The magnitude of the changes are reasonably well predicted (generally within 2-3% delta-I). Some of the differences may be due to the dependence of the ex-core detectors on a few peripheral fuel assemblies that are less axially sensitive to D-bank insertions than average.

Critical boron predictions show a modest initial bias of 10-20 ppm and follow the measured changes within approximately 20 ppm. The total boron change is -600 ppm for the 11/05 event and -200 ppm for the 12/06 event. No adjustment for B-10 depletion effects have been made. Based on the consistency of the boron bias before and after each event, there does not appear to be a significant change in B-10/B-1I ratio during the event. Results for A/O and boron are shown in Figures 13 and 14.

34

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 35 of 40 Figure 9 KPS C27 November 2005 Transient Power and D-bank Changes 100 - - ' '- - 210

- Power 90 1901 90 ~~~~D-Bank - - - - - - -

40 -. 170 30 a - - - 0 3, 10---_j- -- -- 30 0-- -10 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 Time (hours)

Figure 10 KPS C27 February 2006 Transient Power and D-bank Changes 110 -J---- - - - 230 10 - -0 70 1 150 0 lb 0.0 o 0 70, --- , .130

60. 0 4130 o 0 50 -2110 ,

0 10 20 30 40 50 60 70 80 90 100 Time (hours) 35

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 36 of 40 Figure 11 KPS C27 November 2005 Transient Axial Flux Difference 4.0 2.0 00.0 X -2.0 x

o20

-4.0

-6.0

-8.0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 Time (hours)

Figure 12 KPS C27 Febuary 2006 Transient Axial Flux Difference 6.0 4.0 2.0 S 0.0 U

S

-2.0 0

x

-4.0

-6.0

-8.0

-10.0

-12.0 10 20 30 40 50 60 70 80 90 100 Time (hours) 36

DOM-NAF-5-0.0-P-A Attachment A CMS Benchmarking Information pg 37 of 40 Figure 13 KPS C27 November 2005 Transient Critical Boron Concentration 1500/- 1 - 1 1 1400-I- -SIMULATE 1300" & Measured a-0.

C 1200 -- -----

.2 C

0 U

1100-C 0

U C 1000*

0 0

U 900 0 U

600----

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 Time (hours)

Figure 14 KPS C27 Febuary 2006 Transient Critical Boron Concentration CL

.2 0

0 0

M 0 10 20 30 40 50 60 70 80 90 100 Time (hours) 37

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 38 of 40 SECTION 6- CONCLUSIONS Using the methods and processes delineated in DOM-NAF-1, the accuracy of the CMS models has been demonstrated through comparisons with reactor measurements and through comparisons with higher order Monte Carlo neutron transport calculations. As Table 10 demonstrates, the Kewaunee NRF values are consistent with those derived from the assessment performed for the Dominion Surry and North Anna units.

Table 10 Comparison of North Anna I Surry and Kewaunee Nuclear Reliability Factors North Anna I Surry Kewaunee Parameter NRF NRF Upper Lower Upper Lower Integral Control Rod Bank Worth 1.10 0.90 1.10 0.90 (Individual banks)

Integral Control Rod Bank Worth 1.10 0.90 1.10 0.90 (Total of all banks)

Differential Control Rod Bank 1.15 0.80 1.20 0.80 Worth Critical Boron Concentration +50 ppm -50 ppm +70 ppm -70 ppm Differential Boron Worth 1.05 0.95 1.05 0.95 Isothermal and Moderator +2 pcm/F -2pcm-F Temperature Coefficient pcm/°F_+

Doppler Temperature Coefficient 1.10 0.90 1.10 0.90 Doppler Power Coefficient 1.10 0.90 1.10 0.90 Effective Delayed Neutron 1.05 0.95 1.05 0.95 Fraction Prompt Neutron Lifetime 1.05 0.95 1.05 0.95 FAH 1.04 N/A 1.04 N/A FQ 1.05 N/A 1.05 N/A 38

DOM-NAF-5-0.0-P-A Attachment A CMVS Benchmarking Information pg 39 of 40 The largest NRF difference between North Anna/Surry and Kewaunee is for the critical boron concentration parameter. Although the critical boron tolerance interval would support a NRF of +15/-60 ppm at Kewaunee, both the shape of the difference distribution and the limited number of samples (93 samples for Kewaunee versus over 1000 for North Anna and Surry) introduce uncertainty as to how well the Kewaunee CMS models predicted critical boron.\A+/-70 ppm NRF is chosen because it is sufficiently wide for conservative safety analysis but not so wide as to mask potential plant operation issues. The critical boron NRF can be adjusted in the future as more boron samples are collected and/or a cause for the bias between measured and predicted values is determined.

The Dominion Kewaunee CMS models, including appropriately determined NRF values, have been demonstrated to be appropriate for use in reload applications such as core reload design, core follow, and calculation of key core parameters for reload safety analysis at Kewaunee Power Station. The robust model development process, in conjunction with code and model quality assurance practices, as described fully in DOM-NAF-1 and applied specifically to Kewaunee Power Station in this document, provide assurance that future changes to core, fuel and burnable poison designs will be modeled with accuracy and appropriate conservatism.

39

DOM-NAF-5-0.0-P-A Attachment A CIVIS Benchmarking Information pg 40 of 40 SECTION 7- REFERENCES

1. R. Hall, R. Kepler, J. Miller, C. Wells, W. Peterson, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," Dominion, DOM-NAF-1 -Rev.0.0-P-A, June 2003.

2. M.G. Natrella, "Experimental Statistics," National Bureau of Standards Handbook 91, August 1963.

3. "An Acceptance Model and Related Statistical Methods for the Analysis of Fuel Densification," U.S.N.R.C. Regulatory Guide 1.126, Rev. 1, March 1978.

4. P. N. Somerville, "Tables for Obtaining Non-Parametric Tolerance Limits," Ann.

Math. Stat., Vol 29, 1958.

40

DOM-NAF-5-o.O-A.

Attachment B RETRAN Benchmarking Information Prepared by:

John C. Lautzenheiser Mark C. Handrick Approved:

Cary Larue-Supervisor, Nuclear Safety Analysis

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg 2 of 49 Table of Contents TABLE OF CONTENTS ......................................................................................................... 2 1.0 Introduction and Summ ary ......................................................................................... 3 1.1 Introduction ...................................... ................ 3 1.2 Sum mary ............................................................................................................ 3 2.0 KPS RETRAN Model....... ...................................... 4 3.0 Method of Analysis ...................................................... ...... ......... 8 4.0 Demonstration Analysis Results ....................................... ................ 8 4.1 Loss of Load ............................................................ ........... 9 4.2 Locked Rotor .......................................................... ......... 15 4.3 Loss of N orm al Feedw ater ..................................................................................... 24 4.4 M ain Steam Line Break ......................................................................................... 30 4.5 Control Rod Bank Withdrawal at Power .............................................................. 37 4 .6 L oss of F low ............................................................................................................. 44 5.0 C onclu sions .................................................................................................................... 49 6 .0 Referen ces ....................................................................................................................... 49

DOM-NAF-5-0.0-A Attachment13 RETRAN Benchmarking Information pg 3 of 49 1.0 Introduction and Summary 1.1 Introduction Topical report VEP-FRD-41, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," (Reference 1) details the Dominion methodology for Nuclear Steam Supply System (NSSS) non-LOCA transient analyses. This methodology encompasses the non-LOCA licensing analyses required for the Condition I, II, III, and IV transients and accidents addressed in the Updated Safety Analysis Report (USAR). The VEP-FRD-41 methods are also used in support of reload core analysis. In addition, this capability is used to perform best-estimate analyses for plant operational support applications. The material herein supplements the applicability assessment of RETRAN methods for Kewaunee Power Station (KPS) that is presented in Section 3.4 of DOM-NAF-5, demonstrating that the VEP-FRD-41 methods are acceptable for the stated applications.

1.2 Summary This report provides a description of the RETRAN base model for KPS and results of demonstration analyses using this model. The KPS model was developed in accordance with the methods in VEP-FRD-41, with certain noding changes noted below. This assessment reaffirms the conclusion in Section 3.4 of DOM-NAF-5, that the Dominion RETRAN methods, as documented in topical report VEP-FRD-41, are applicable to KPS and can be applied to KPS licensing analysis for reload core design and safety analysis. Dominion analyses of KPS will employ the modeling in VEP-FRD-41, as augmented with the noding changes listed below.

Thus, VEP-FRD-41, as augmented, is the Dominion methodology for analyses of non-LOCA NSSS transients for KPS.

The KPS RETRAN base model contains the following alterations in noding with respect to the modeling that is documented in VEP-FRD-41.

a) The KPS model explicitly models the safety injection (SI) accumulators.

b) The KPS model has separate volumes for the steam generator inlet and outlet plenums.

c) The KPS model includes cooling paths between downcomer and upper head (Main Steam Line Break overlay).

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg 4 of 49 2.0 KPS RETRAN Model A KPS RETRAN-02 Base Model and associated model overlays are developed using Dominion analysis methods described in the Dominion RETRAN topical report (Reference 1). The Dominion analysis methods are applied consistent with the conditions and limitations described in the Dominion topical report and in the applicable NRC Safety Evaluation Reports (SERs).

-J The KPS Base Model noding diagram is shown on Figure 2-1. Volume numbers are circled, junctions are represented by arrows, and the heat conductors are shaded. This model simulates both reactor coolant system (RCS) loops and has a single-node steam generator (SG) secondary side, consistent with Dominion methodology. The SG primary nodalization includes 10 steam generator tube volumes and conductors. There is a multi- node SG secondary overlay that can be added to the Base Model for sensitivity studies although none of the analysis results presented herein utilize this overlay.

In addition to the base KPS model, an overlay deck is used to create a split reactor vessel model to use when analyzing Main Steam Line Break (MSLB) events, consistent with Dominion methodology. This overlay adds volumes to create a second, parallel flow path through the active core from the lower plenum to the upper plenum such that RCS loop temperature asymmetries can be represented. This overlay also includes flow paths between the downcomer and the upper head to model the small amount of cooling flow to the upper head. These flowpaths may also be added for other events when flashing in the upper head is expected to occur. A noding diagram of the split reactor vessel is shown on Figure 2-2. This figure shows the hot leg volumes (101, 201) and cold leg volumes (116, 216) so the reactor vessel can be seen in context of the RCS interface. Otherwise, the noding for all other regions of the model are unchanged from Figure 2-1.

The base KPS model noding is virtually identical to the Surry (SPS) and North Anna (NAPS) models with the exception of some minor noding differences listed as follows, which are updated from the original list provided in DOM-NAF-5 Section 3.4.1.4.

a) The KPS model explicitly models the SI accumulators.

b) The KPS model has separate volumes for the SG inlet and outlet plenums.

c) The KPS model includes cooling paths between downcomer and upper head (MSLB overlay).

The SI accumulators are part of the KIPS model because injection from the accumulators is more likely to occur during a MSLB cooldown event for a two-loop plant. The cooling paths are included in the MSLB overlay to appropriately model the effects of flashing in the head, as noted above. The use of separate volumes for the inlet and outlet should have little effect on transient

DOM-NAF-5-0.0 -A' Attachment B RETRAN Benchmarking Information pg 5 of 49 response since the fluid temperature in these volumes is generally the same as the connecting RCS piping.

The Dominion models, including the KPS model, have some differences compared to the vendor RETRAN model that was used to perform the current USAR analyses. Table 2-1 and the subsequent text discussion provides an overview of these differences. Additional details concerning differences between the Dominion KPS and USAR RETRAN models are discussed in the demonstration analyses in Section 4.

A description of the Dominion RETRAN methodology is provided in Reference 1, where specific model details are discussed in Sections 4 and 5 of that reference.

Table 2-1 Dominion USAR RETRAN Model Comparison Parameter Dominion USAR Noding:

Reactor Vessel Single flow path - 3 axial nodes for Split (two parallel flow path) - 4 active core (special split core axial nodes for active core.

overlay for MSLB only) Increased nodalization in other vessel regions.

Steam Generator Single node secondary. Five axial Multi-node secondary. Four axial levels (10 nodes) for' SG tubes levels for SG tubes (primary and primary side. Local Conditions Heat secondary).

Transfer model available for loss of heat sink events.

Reactivity Model Doppler Feedback Doppler temperature coefficient that Doppler-only power coefficient is a function ofTFUEL. and a Doppler temperature coefficient effect driven by moderator temperature.

Moderator Feedback Moderator temperature coefficient Moderator density coefficient Decay Heat ANS-5.1 1979 Standard ANS-5.1 1979 Standard U-235 with 1500 day burn. Equilibrium decay heat, Q = 190 MeV/fission. Bounds additional 2o uncertainty Bounds additional 2o uncertainty Other:

Heat Transfer Option Forced Forced + Free Convection HT I _Map Gap Expansion Model Used for HZP events only. Used for all events.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 6 of 49 MFW

• xrw

÷*AFW Figure 2-1. KPS Base Model Nodalization Diagram

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 7 of49 351 352 J16 Figure 2-2. KPS Split Vessel Nodalization

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information,' pg. 8 of 49 3.0 Method of Analysis As discussed in Section 3.4.3 of DOM-NAF-5, validation of the Dominion KPS RETRAN method involves comparison of RETRAN analyses to the KPS USAR analysis of record (AOR) for select events. These events represent a broad variation in behavior (e.g. RCS heatup, RCS cooldownrdepressurization, reactivity excursion, loss of heat sink, etc.), and demonstrate the ability to appropriately model key phenomena for a range of transient responses. The transients selected for comparison with their corresponding KPS USAR section are provided in Table 3-1. For each transient, an analysis is performed using the Dominion KPS RETRAN model and analysis methods. Initial conditions are established to be consistent with the input used in USAR analyses.

Table 3-1 Transients Analyzed for USAR Comparison Transient KPS USAR Section Control Rod Withdrawal at Power 14.1.2 Loss of Flow 14.1.8 Locked Rotor 14.1.8 Loss of Load/Turbine Trip 14.1.9 Loss of Normal Feedwater 14.1.10 Main Steam Line Break 14.2.5

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 9 of 49 4.0 Demonstration Analysis Results A summary for each transient comparison is presented in the following sections. Included in each section is an input summary identifying key inputs and assumptions along with differences from USAR assumptions. A comparison of the results for key parameters is provided with an explanation of key differences between the Dominion and USAR cases.

4.1 Loss of Load The Loss of Load/Turbine Trip (LOL) event is defined as a complete loss-of-steam load and turbine trip from full power without a direct reactor trip, resulting in a primary fluid temperature rise and a corresponding pressure increase in the primary system. This transient results in degraded steam generator heat transfer, reactor coolant heatup and pressure increase following a manual turbine trip.

The LOL transient scenario presented here was developed to analyze primary RCS overpressurization. It is initiated by decreasing both the steam flow and feedwater flow to zero immediately after a manual turbine trip. The input summary is provided in Table 4.1-1.

Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.1-1 LOL Input Summary Parameter Value Notes Initial Conditions NSSS Power (MW) 1815.6 Includes 2% uncertainty and pump heat RCS Flow (gpm/loop) 89,000 Thermal Design Vessel TAvG (F) 579 Includes +6 F uncertainty RCS Pressure (psia) 2200 Includes -50 psia uncertainty (delays trip)

Pressurizer Level (%) 53 Includes +5 % uncertainty SG Level (%) 44 SG Pressure (psia) 836.8 USAR = 797.98 Assumptions/Configuration Reactor trip - only Hi Pzr Pressure is active Automatic rod control - Not credited Pressurizer sprays, PORVs - Not credited Main steam dumps, SG PORV - Not credited AFW flow - Not credited SG tube plugging (%) 10 Max value Reactivity Parameters Doppler Temp. Coefficient (pcm/F) -2.5, Most USAR uses least negative Doppler-only negative power coefficient, and a least negative DTC (driven by moderator temp).

Moderator Temp. Coefficient (pcn/F) 0.0 USAR uses 0.0 Ak/(grn/cc) for MDC

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 10 of 49 Results - LOL Pressure in the RCS increases during a LOL due to degraded heat transfer in the steam generator and is alleviated only when the pressurizer safety valves (PSV) open. The pressurizer pressure response is shown on Figure 4.1-1 and the peak RCS pressure values are listed in Table 4.1-1. Pressure for the Dominion case increases slightly faster initially but peaks at about the same time and same value as the USAR data.

Figure 4.1-2 shows the power response which is nearly constant until a reactor trip on high pressurizer pressure occurs. The Dominion case trips slightly earlier than the USAR data because of the higher RCS pressurization rate.

The core average temperature is shown on Figures 4.1-3. The Dominion and USAR temperature are virtually identical until after the reactor trip and peak pressure occurs, at which time they diverge somewhat but trend together. Because of the RCS. heat up and coolant expansion, there is a liquid insurge to the pressurizer as shown by the pressurizer liquid volume increase on Figure 4.1-4. The pressurizer liquid volume increases faster for the Dominion case early in the event, yielding the slightly faster pressure increase discussed earlier. By the time the peak pressure is reached at approximately 11 seconds, the liquid volumes for both cases compare well. The difference after that time is consistent with the temperature response.

The steam generator pressure is shown on Figure 4.1-5. The steam generator heat transfer degradation is strongly related to the secondary pressure increase (saturation temperature increase) since the dominant secondary heat transfer mode is boiling heat transfer. The Dominion case starts from a higher initial pressure, which is the result of model initialization to match the design heat transfer surface area for the single node steam generator, and peaks at a higher pressure than the USAR case.

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rkingInforination pg. 11 of 49 Table 4.1-2 LOL Primary Overpressure Results Parameter DOM USAR Sequence of Events:

Reactor Trip (sec) 8.4 8.9 (High Pzr Pres)

Peak RCS Pressure (psia) 2697 2697 Peak MSS Pressure (psia) 1192 1182 Summary - LOL The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the LOL event. The RCS temperatures agree well early in the event and although they diverge somewhat later in the event, this does not occur until after the RCS pressure peak occurs and pressure relief begins. There are small differences in pressurization rates early in the event; however, the peak RCS pressure values are the same for both cases. In addition, the peak SG pressure is slightly higher for the Dominion case. There is adequate margin to the RCS pressure acceptance criterion of 2750 psia.

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 12 of 49 Figure 4.1-1 LOL - Pressurizer Pressure 2600-2500' 2400 2300 2200' 0 2 4 6 8 10 12 14 16 18 20 Time (seconds)

Figure 4.1-2 LOL - Reactor Power 1.20 1.00 0.80 00.60 0

0.40 0.20 0.00 0 2 4 6 8 10 12 14 16 18 20 Time (seconds)

DOM-NAF-5-00-A Attachment B RETRAN Benchma rking Information pg. 13 of 49 Figure 4.1-3 LOL - RCS Average Temperature 590 0D a)

C-E g- 580 0 2 4 6 8 10 12 14 16 18 20 Time (seconds)

Figure 4.1-4 LOL - Pressurizer Liquid Volume 0)

E

. 580 560 0 4 6 8 10 12 14 16 18 20 Time (seconds)

DOM-NAF-5-0.0-A AttachmentB RETRAN Benchma rking Information pg. 14 of 49 Figure 4.1-5 LOL - Steam Generator A Pressure 1300, 1200' 1100 L? 1000 900 800 0 2 4 6 8 10 12 14 16 18 20 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 15 of 49 4.2 Locked Rotor The Locked Rotor / Shaft Break (LR) event is defined' as an instantaneous seizure of a Reactor Coolant Pump (RCP) rotor, rapidly reducing flow in the affected reactor coolant loop leading to a reactor trip on a low-flow signal from the Reactor Protection System. The event creates a rapid expansion of the reactor coolant and reduced heat transfer in the steam generators, causing an insurge to the pressurizer and pressure increase throughout the reactor coolant system (RCS).

The LR transient scenario presented here was developed to analyze primary RCS overpressurization. It is initiated by setting one RCP speed to zero as the system is operating at full power. The reactor coolant low loop flow reactor trip is credited, with a setpoint of 86.5% of the initial flow. The input summary is provided in Table 4.2-1. Most of the input parameters are the same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

Table 4.2-1 LR Input Summary Parameter Value Notes Initial Conditions NSSS Power (MW) 1815.6 Includes pump heat and 2% uncertainty RCS Flow (gpm/loop) 89,000 Thermal Design Flow Vessel TAvG (F) 579 Includes +6 F uncertainty Initial Fuel Average Temperature (F) 1332 RCS Pressure (psia) 2300 Includes +50 psia uncertainty Pressurizer Level (%) 48 Nominal SG Level (%) 44 Nominal SG Pressure (psia) 836.8 (Note 1)

Assumptions/Configuration Reactor trip - Only Low RCS Loop Flow is credited Automatic rod control - Not credited Pressurizer sprays, PORVs - Not credited Main steam dumps, SG PORV - Not credited AFW flow - Not credited SG tube plugging (%) 10 Max value RCP/motor moment of inertia (lbm/flC) 72,000 90% of nominal Reactivity Parameters Doppler Temp. Coefficient (pcm/F) -1.2, Least USAR uses most negative Doppler-only Negative power coefficient, and a least negative DTC (driven by moderator temp). /

Moderator Temp. Coefficient (pcm/F) 0.0 USAR uses 0.0 Ak/(gm/cc) for Moderator Density Coefficient 1 - SG pressure adjusted for minimal change to the SG tube surface areas by steady-state initialization

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg& 16 of 49 Results - LR RCS Overpressure Case Pressure in the RCS increases during a LR event due to degraded heat transfer in the steam generator and is alleviated only when the pressurizer safety valves (PSV) open. The pressurizer pressure response is shown on Figure 4.2-1 and the peak RCS pressure values are listed in Table 4.2-1. Pressure for the Dominion case increases slightly faster initially but peaks at about the same time and approximately the same value as the USAR data.

Figure 4.2-2 shows the pressure response in the reactor vessel lower plenum, which compares well to the USAR data, particularly until the point of peak pressure.

Figures 4.2-3 and 4.2-4 show the total core inlet volumetric flow and faulted loop volumetric flow (fraction of nominal), respectively. The predicted volumetric core inlet volumetric flow rate decreases more rapidly than the USAR data. This behavior is also present for the faulted loop volumetric flow, where the loop flow reverses earlier than the USAR data. This is conservative behavior in comparison to the USAR data. Each analysis assumes 90% of nominal RCP inertia to limit the coastdown.

Figure 4.2-5 shows the core thermal power response, which matches the USAR analysis well, except for small differences prior to the reactor trip. The USAR Doppler feedback model is a function of (1) Doppler-only power coefficient (DPC), and (2) Doppler temperature coefficient (DTC). The DTC modeled in the USAR analysis is actually a function of moderator temperature rather than fuel temperature. The Dominion reactivity model uses a Doppler Temperature Coefficient, dependent only on changes in fuel temperature, which provides the prompt feedback component. The Dominion DTC model is described in Section 5.13 of Reference 1. The Dominion model predicts core power to decrease prior to reactor trip, which is expected due to negative Doppler feedback as the fuel temperature increases.

The computed core average heat flux shown in Figure 4.2-6 compares well with the USAR data. The small difference in the core heat flux response during the first second of the transient is probably due to the difference in modeling of the core heat transfer coefficients (HTC). In the USAR analysis, the core HTCs are held fixed at their initial values. In the Dominion model, the forced convection HTCs are allowed to decrease with the decaying RCS flow rate, effectively reducing the core heat flux during the first second of the event.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 17 of 49 The faulted loop hot leg and cold leg temperatures are shown on Figure 4.2-7. The Dominion and USAR temperature are virtually identical until time when the peak pressure occurs, at which time they diverge somewhat but trend together.

A summary of the LR transient analysis comparison is provided in Table 4.2-2.

Table 4.2-2 LR RCS Overpressure Results Parameter DOM USAR Sequence of Events:

Reactor Trip on Low RCS Flow (sec) 0.80 0.80 Peak RCS Pressure (sec) 4.2 4.5 Peak RCS Pressure (psia) 2681 2683 Summary - LR RCS Overpressure Case The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the LR event. There are small differences in RCS coastdown flow and core power response during the early portion of the transient. However these differences occur prior to the time of peak RCS pressure (at approximately 4 seconds), and are relatively insignificant for this transient. The peak RCS pressure values are essentially the same for both cases. The predicted peak RCS pressure for the Dominion model (2681 psia) is just slightly below the USAR peak pressure (2683 psia). In each case, there is adequate margin to the RCS peak pressure acceptance criterion of 2750 psia.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmna rking Information pg. 18 of 49 Figure 4.2-1 LR - Pressurizer Pressure N

0 a,

0~

0 N

0 0 2 4 6 8 10 12 Time (seconds)

Figure 4.2-2 LR - Lower Plenum Pressure E

ID 9

0 2 4 10 12 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 19 of 49 Figure 4.2-3 LR - Core Inlet Volumetric Flow 300-LL.

B 250-0 S200-2 4 6 8 10 12 Time (seconds)

Figure 4.2-4 LR - Faulted Loop Volumetric Flow 0

E 0

_j LL 0 2 4 6 8 10 12 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 20 of 49 Figure 4.2-5 LR - Core Power 0

2-0) 0 0 2 4 6 8 10 12 Time (seconds)

Figure 4.2-6 LR - Core Heat FLux 1.2 1.0 0 O:B S0.6 8

S0.4 0.2 0.0 0 2 4 6 8 10 12 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 21 of 49 Figure 4.2-7 LR - Faulted Loop Temperatures 630 610 CU S590 (D

0.

0 0

--' 570 a,

L_ 550 530 0 2 4 6 8 10 12 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 22 of 49 LR Peak Cladding Temperature The Locked Rotor event is also analyzed to demonstrate that a coolable core geometry is maintained. Acceptance criteria for this analyses are met by showing that the peak cladding temperature (PCT) remains below 2700 'F, and that the oxidation level is below 16.0 percent by weight. A hot spot evaluation is performed to calculate the peak cladding temperature and oxidation level. The Dominion Hot Spot model is described in Topical Report VEP-NFE-2-A, "VEPCO Evaluation of the Control Rod Ejection Transient."

(Reference 2) The Dominion Hot Spot model was used to evaluate the Kewaunee PCT and oxidation level for the LR event.

The Dominion hot spot model is used to predict the thermal- hydraulic response of the fuel for a hypothetical core hot spot during a transient. The hot spot model describes a one- foot segment of a single fuel rod assumed to be at the location of the peak core power location during a transient. The hot spot model uses boundary conditions from the LR system transient analysis to define inlet flow and core average power conditions. The hot spot model uses Kewaunee-specific values for fuel dimensions, fuel material properties, fluid volume, and junction flow areas.

The hot spot model is run to 0.001 seconds and a restart file is saved. Upon restart, the fuel/cladding gap conductance (thermal conductivity) is modified to simulate gap closure by setting the gap heat transfer coefficient to 10,000 Btu/ft2 -hro-F for a gap conductance of 3.125 Btulft-hr-°F. The hot spot model input summary is provided in Table 4.2-3. Most of the input parameters are the same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

Table 4.2-3 Hot Spot Model Input Summary Parameter Value Notes Computer Code Used RETRAN-02 USAR uses FACTRAN Initial Conditions Ratio of Initial to Nominal Power 1.02 RCS Thermal Design Flow (gpm/loop) 89,000 Hot Spot Peaking Factor 2.50 Assumptions/Configuration Pre-DNB Film Heat Transfer Coefficient Thorn USAR uses Dittus-Boelter or Jens-Lottes Time of DNB (sec) 0.001 Post DNB Film Boiling Heat Transfer Bishop-Coefficient Sandberg-Tong Fuel Pin Model Post DNB Gap Heat Transfer Coefficient 10,000

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 23 of 49 (Bt/hr-ft2--F)

Gap Thermal Expansion Model activated? Yes Zircaloy-Water Reaction activated? Yes LR Peak Cladding Temperature Results The peak cladding inner surface temperature obtained from the Dominion Kewaunee hot spot model is 1633 OF. The maximum zircaloy-water reaction depth into the cladding is 2.04E-06 feet, which corresponds to approximately 0.10% by weight based on the nominal cladding thickness of 2.025E-03 feet. A summary of the LR Peak Cladding Temperature Hot Spot analysis comparison is provided in Table 4.2-4.

Table 4.2-4 LR Hot Spot Results Parameter DOM USAR Peak Cladding Temperature 1633 OF 1900 OF Maximum Zr-water reaction (w/o) 0.10 0.61 The Dominion peak cladding temperature and maximum oxidation values are less than the USAR values, however both cases demonstrate considerable margin to the acceptance criterion of 2700 OF and 16.0% by weight, respectively.

The difference in zirconium-water reaction results is understood by examination of the Baker-Just parabolic rate equation, which shows that the zirconium-water reaction becomes significant above cladding temperatures of 1800 OF. Since the Dominion hot spot model results do not predict peak cladding temperatures of this magnitude, it is expected that the corresponding zirconium-water reaction would be less than the USAR analysis. In the RETRAN-02 code, the rate of change in the reacting metal-oxide interface is proportional to exp(-41200/T) where T is the absolute clad temperature (OR).

The ratio of exp[-41200/(1633+460)] / exp[-41200/(1900+460)] = 0.1. This value corresponds approximately to the oxidation ratio of the Dominion vs. USAR result.

DOM -NAF-5-0.0-A Attachmnent B RETRAN Benchmarking Information pg. 24 of 49 4.3 Loss of Normal Feedwater The Loss of Normal Feedwater (LONF) event causes a reduction in heat removal from the primary side to the secondary system. Following a reactor trip, heat transfer to the steam generators continues to degrade resulting in an increase in RCS fluid temperature and a corresponding insurge of fluid into the pressurizer. There is the possibility of RCS pressure exceeding allowable values or the pressurizer becoming filled and discharging water through the relief valves. The event is mitigated when Auxiliary Feedwater (AFW) flow is initiated and adequate primary to secondary side heat removal is restored. This analysis shows that the AFW system is able to remove core decay heat, pump heat and stored energy such that there is no loss of water from the RCS and pressure limits are not exceeded. The LONF input summary is provided in Table 4.3-1. Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.3-1 LONF Input Summary Parameter Value Notes Initial Conditions NSSS Power (MW) 1790.62 Includes 0.6% unc. and pump heat RCS Flow (gpm/loop) 89,000 Thermal Design Vessel TAvO (F) 579 HFP nominal + 6 F RCS Pressure (psia) 2300 Nominal + 50 psi Pressurizer Level (%) 53 Nominal + 5%

SG Pressure (psia) 863.4 USAR= 829.0 SG Level (%) 51 Nominal + 7%

Assumptions/Configuration Low-Low Level Reactor Trip Setpoint 0% Percent of narrow range span Pressurizer: sprays, heaters, PORVs - assumed operable AFW Temperature (F) 120 max value AFW Pump configuration - one motor-driven pump per SG Auxiliary feedwater flow rate (gpm) - variable as function of SG press.

AFW Delay after Low-Low-SG level (sec) 800 Max delay Local Conditions Heat Transfer model active SG secondary side USAR= multi-node SG Steam Generator MSSV blowdown (%) 3 USAR= blowdown not modeled Reactivity Parameters Doppler Temp. Coefficient (pcm/F) -2.5, Most USAR uses most negative Doppler-negative only power coefficient, and a least negative DTC (driven by moderator temp).

Moderator Temp. Coefficient (pcu/F)

Moderator_____________

Temp. Coeficen TModerator 0.0 USAR uses 0.0 Ak/(gm/cc) for Density Coefficient

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 25 of 49 Results - LONF The results for the LONF comparison analysis are presented in Table 4.3-2 and Figures 4.3-1 through 4.3-6. The loss of feedwater flow to the steam generators (SG) results in a reduction in SG level until a reactor trip occurs on Low-Low SG level. Normalized power is shown on Figure 4.3-1. The power response is similar for the Dominion and USAR cases, .except that the trip for the Dominion case occurs about 10 seconds later due to slightly higher initial fluid mass in the Dominion SG secondary side. (Higher initial mass has been demonstrated in a sensitivity case to yield conservative results due to delayed reactor trip).

The reduction in SG level results in degraded heat transfer from the primary to secondary systems and an increase in RCS temperature, plotted on Figure 4.3-2. The heatup prior to reactor trip is more pronounced for the Dominion model due in part to the delay in reactor trip. After the reactor trip occurs, the RCS cools somewhat until the loss of SG level and related heat removal is no longer able to remove decay and residual heat. The temperature then increases until AFW flow is actuated and adequate heat removal is restored.

The effect of the temperature change is reflected in the fluid density and associated pressurizer level change, as seen on Figure 4.3-3. The initial pressurizer insurge is higher for the Dominion case, which receives a later reactor trip signal as noted earlier. The maximum pressurizer level, which occurs after the reactor trip, is higher for the USAR case.

This appears to be due primarily to higher pressurizer spray flow rates and more conservative decay heat assumptions for the USAR cases. Note, both the Dominion and USAR methods use decay heat profiles that conservatively bound the values for the 1979 ANS-5.1 decay heat model plus 2-sigma uncertainty; however, the USAR method uses a higher value for the decay heat multiplier. Also, the Dominion methodology assumes a conservative value for pressurizer spray flow, although the USAR model appears to apply additional conservatism.

Next, pressurizer pressure responds to the level insurge as shown on Figure 4.3-4. The initial pressure increase for the Dominion case is sufficiently high to cause the pressurizer Power Operated Relief Valves (PORV) to open. The subsequent Dominion pressure response is below the USAR profile but eventually increases above the USAR values during the second insurge. Overall, the pressure response for the USAR case is somewhat flatter due to the higher pressurizer spray flow, which tends to suppress the pressure increases.

Also, the delay in reactor trip causes the initial pressure increase for the Dominion case to be more severe.

DOM-NAF-5-O.O-A Attachment B RETRAN Benchmarking Information pg. 26 of 49 The secondary side response for SG mass and pressure is shown on Figures 4.3-5 and 4.3-6, respectively. Other than the differences in initial pressure and fluid mass, the responses for the Dominion case and USAR case are similar. Note, the Dominion case models Main Steam Safety Valve (MSSV) blowdown, resulting in slightly lower steady state SG pressures and a more clearly defined valve opening and closing response. Finally, as noted in the plot for SG mass, once AFW flow is initiated, the SG level gradually increases and adequate heat removal is eventually restored.

Table 4.3-2 LONF Results Parameter DOM USAR Sequence of Events:

Loss of Feedwater (see) 20.0 20.0 Reactor Trip (sec) 65.4 54.5 (Low-Low SG Level)

Peak RCS Pressure (psia) 2433 2341 Peak PZP Liquid Volume (ft3) 842 925 Summary - LONF The Dominion analysis provides results that are similar to the USAR analysis for the LONF event. Differences in the maximum pressurizer level are explained primarily by differences in pressurizer spray assumptions and decay heat modeling. Both analyses are conservative and demonstrate adequate margin to pressurizer overfill acceptance criteria.

DOM-NAF-5-0,0-A Attachment B RETRAN Benchmarking Information pg. 27 of 49 Figure 4.3-1 LONF - Core Power 1.2' 1.0 S

o 0.8.

0 t

T 0.6.

Z 0.4' z

0.2' I 10 100 Time (seconds) 1000 10000 Figure 4.3-2 LONF - RCS Average Temperature 0)

I-10 100 1000 10000 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 28 of 49 Figure 4.3-3 LONF - Pressurizer Liquid Volume 1200 a3 E=

0

.5 10 100 1000 10000 Time (seconds)

Figure 4.3-4 LONF - Pressurizer Pressure 2500 9D CL 1 10 100 1000 10000 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 29 of 49 Figure 4.3-5 LONF - Steam Generator Mass no 10 100 1000 10000 Time (seconds)

Figure 4.3-6 LONF - Steam Generator A 1200 1100 S1000 900 800 1 10 100 1000 10000 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 30 of 49 4.4 Main Steam Line Break The Main Steam Line Break (MSLB) event is a rupture in the main steam piping resulting in a rapid depressurization of the SG secondary and corresponding cooldown of the primary.

The temperature reduction results in an insertion of positive reactivity with the potential for core power increase and DNBR violation.

The MSLB transient scenario presented here is modeled as an instantaneous, double-ended break at the nozzle of one steam generator from hot shutdown conditions with offsite power available. The input summary is provided in Table 4.4-1. Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.4-1 MSLB Input Summary Parameter Value Notes Initial Conditions Core power (MW) 1772.9E-9 HZP; USAR - 1.0% RTP Pump power (MW) 8 Nominal RCS Flow (gpm/loop) 89,000 Thermal Design Vessel TAvO (F) 547 HZP nominal RCS Pressure (psia) 2250 Nominal Pressurizer Level (%) 21 HZP nominal SG Level (%) 44 Nominal Assumptions/Configuration Heat transfer option Forced HT Map USAR = Forced + Free (note 1) Convection HT Map Manual Reactor Trip - Assumed at time 0 sec Main feedwater flow (% HFP value) 100 initiated at time 0 sec Auxiliary feedwater flow rate (gpm) 600 initiated at time 0 sec; per SG SG tube plugging (%) 0 Minimum value Reactivity Parameters Boron Reactivity (pcm/ppm) -8.0 USAR= reactivity f(boron concentration).

Doppler Reactivity Feedback Doppler Power USAR - Doppler power defect defect, DTC plus DTC model disabled Moderator Feedback Moderator Moderator density feedback density feedback Decay heat multiplier 1.0 USAR= 1.E-20 1 - Dominion method maximizes heat transfer coefficients for the faulted SG secondary side.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 31 of 49 Results - MSLB with Offsite Power Available The forcing function for the MSLB transient is the break flow. The main steam flow from the faulted and unfaulted SGs is plotted on Figure 4.4-1. Flow from the unfaulted SG stops at approximately 11 seconds due to Main Steam Isolation Valve (MSIV) closure. Flow from the faulted SG continues and there is good agreement between the Dominion 'and USAR cases for the first 85 seconds. After that time, the break flow predictions from the faulted steam generator begin to diverge slightly. This is primarily due to the differences in the predicted core heat flux response as discussed below.

The SG pressure response (Figure 4.4-2) matches well initially with the USAR data until the MSIV closes. After the MSIV closure, the unfaulted SG pressure increases and the USAR SG pressure remains higher than the Dominion model, most likely due to difference in SG modeling.

The core heat flux response is shown on Figure 4.4-3. The Dominion case core heat flux increases to a higher value compared to the USAR case. This results primarily from differences in the amount and timing of boron reaching the core. The Dominion boron transport method conservatively models the various system delays associating with purging of fluid from the SI lines and transport of borated water from the Refueling Water Storage Tank (RWST) to the core. The initial fluid in the SI piping is assumed to be at a boron concentration of zero ppm. This affects the timing at which boron reaches the core from the RWST and begins to suppress power. In addition, the USAR analysis RCS pressure decreases more quickly than the Dominion pressure. While this difference is relatively small, it allows for greater injection of accumulator flow for the USAR cases, which is further reflected in the power response as well as the RCS temperature and pressure response. Note that since accumulators are included in the KPS RETRAN model, the Dominion boron transport model described in Reference 1 was modified slightly to include the accumulators, however the accumulators are not subject to the system delays associated with the purge time for the Safety Injection (SI) system piping.

The total core reactivity is initially similar to the USAR data as shown in Figure 4.4-4.

After 60 seconds, the USAR core reactivity becomes negative, reflecting the sudden increase in core boron concentration. After approximately 100 seconds the negative reactivity for the Dominion model increases as the core boron concentration increases.

However, this occurs after the point of peak heat flux.

The core average boron concentration response is shown in Figure 4.4-5. As can be seen, boron starts to reach the core prior to 60 seconds for both the USAR case and the Dominion

DOM -NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 32 of 49 case. In the USAR case, a higher accumulator flow rate results in a larger increase in core boron concentration. As seen on Figure 4.4-3 for the USAR analysis, the effect on core heat flux is immediate. Once accumulator flow ceases, core boron concentration continues to increase slowly as a result of continued SI flow from the RWST. In the Dominion model, the accumulator flow is less than the USAR case during the first 100 seconds of the transient. Due to the SI piping purge delay times in the Dominion model, borated water from the RWST does not enter the core until approximately 140 seconds. The higher core heat flux predicted by the Dominion case is mainly attributable to the later timing of boron injection.

The pressurizer pressure response is shown in Figure 4.4-6. The pressure initially decreases at a rate comparable to the USAR result. At approximately 20 seconds, the upper head begins to flash and the depressurization rate is decreased. The timing of the upper head flashing and the following depressurization is a contributor to when the accumulators activate. After the accumulator flow stops, the RCS pressure starts to rise slowly.

The reactor vessel inlet temperature response (Figure 4.4-7) shows that the initial cooldown matches well for both the USAR and Dominion cases. After approximately 150 seconds, the temperature differences are attributed to the different core heat flux response. However, this occurs well after the point of peak heat flux, as core power is steadily decreasing.

The sequence of events is compared to the USAR in Table 4.4-2. USAR values are taken from Kewaunee USAR Table 14.2.5-1.

Table 4.4-2 MSLB with Offsite Power Results Parameter DOM USAR Sequence of Events:

Initiate Break 0.0 0.01 Unfaulted SG High-High Steam flow setpoint reached 0.79 0.71 Faulted SG Low-Low Steam Pressure Signal 1.02 1.44 Unfaulted SG Low-Low Steam Pressure Signal 1.64 2.01 Safety Injection Actuation Signal 2.79. 2.72 Steam line Isolation (MSIV Closure) occurs 10.29 10.22.

Peak Heat Flux occurs 100.0 56.5 Feedwater Isolation occurs 86.72 87.82 Peak Heat Flux (fraction of nominal) 0.318 0.288

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 33 of 49 Figure 4.4-1 MSLB w/ Offsite Power - Steam Flow 4500 4000-G,33500-3000 -

2500 0

.2 2000 2

0 1500 E

1000-5OO 500 - '9 Unfaulted SG Isolation 01 0 60 120 180 240 300 Time (seconds)

Figure 4.4-2 MSLB wI Offsite Power - Steam Generator Pressure 800 2

8 600 0

E400 0 120 180 240 300 Time (seconds)

DOM-NAF-5-O.O-A Attachment B RETRAN Benchma rking Information pg. 34 of 49 Figure 4.4-3 MSLB w/ Offsite Power - Core Heat Flux z

0 4-

-0.2

1)

U-0 60 120 180 240 300 Time (seconds)

Figure 4.4-4 MSLB w/ Offsite Power - Core Reactivity 0

60 120 180 240 300 Time (seconds)

DOM oNAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 35 of 49 Figure 4.4-5 MSLB w/ Offsite Power - Core Average Boron Concentration 600 500 E

400 g 300 0

2o0 100 0 60 120 180 240 300 Time (seconds)

Figure 4.4-6 MSLB wI Offsite Power - Pressurizer Pressure 60 120 180 240 300 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 36 of 49 Figure 4.4-7 MSLB wI Offsite Power - Reactor Vessel Inlet Temperature LL 500-E

- 400-5 300 60 120 180 240 300 Time (seconds)

Summary - MSLB This section presents a comparison of a RETRAN-02 Main Steam Line Break transient calculation with the Kewaunee model using the Dominion RETRAN transient analysis methods (Reference 1) compared to the USAR results. The key observations from these comparisons are that:

1) The peak power and heat flux reached with the Dominion methods is higher than the USAR result.

2) The effect of boron is significant in these transients because it can determine the timing and the magnitude of the transient peak power. The core average boron concentration resulting from SI and accumulator flow is different between the USAR and the Dominion model. The amount of flow from the accumulators greatly affects the amount of boron in the system and core power.

DOM -NAF-5-0.0-Ak Attachment B RETRAN Benchmarking Information pg. 37 of 49 4.5 Control Rod Bank Withdrawal at Power The Control Rod Bank Withdrawal at Power (RWAP) event is defined as the inadvertent addition of core reactivity caused by the withdrawal of rod control cluster assembly (RCCA) banks when the core is above no load conditions. The RCCA bank withdrawal results in positive reactivity insertion, a subsequent increase in core nuclear power, and a corresponding rise in the core heat flux. The RWAP event described here is terminated by the Reactor Protection System on a high neutron flux trip or the overtemperature AT trip (OTAT), consistent with the USAR analyses.

The RWAP event is simulated by modeling a constant rate of reactivity insertion starting at time zero and continuing until a reactor trip occurs. The Dominion analysis involves two different reactivity insertion rates, 3 pcm/sec and 100 pcm/sec that match the reactivity insertion rates described in the USAR. Both cases assume that the reactor is initially operating at 100% power, with minimum core reactivity feedback. Most of the input parameters are he same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

Table 4.5-1 RWAP Input Summary Parameter Value Notes Initial Conditions NSSS Power (MW) 1780 Nominal plus pump heat RCS Flow (gpm/loop) 93,000 Minimum Measured Flow Vessel TAVG (F) 573 Nominal RCS Pressure (psia) 2250 Nominal Pressurizer Level (%) 48 Nominal SG Level (%) 44 Nominal Assumptions/Configuration Reactor trip - High neutron flux or OTAT Automatic rod control - Not credited Pressurizer. level control - Not credited Pressurizer heaters - Not credited Pressurizer sprays, PORVs - Active SG tube plugging (%) 10 Max value Reactivity Parameters Doppler Temp. Coefficient (pcm/F) -1.2 Dominion least negative DTC model.

USAR uses least negative Doppler-only power coefficient, and a least negative DTC (driven by moderator temperature).

Moderator Temp. Coefficient (pcm/F) 0.0 USAR uses a value of 0.0 pcm/°F for MTC Moderator Density Coefficient N/A USAR uses 0.0 Ak/(gmlcc)

Fuel Heat Conduction Model Initial Fuel Average Temperature Minimum USAR targets a minimum value

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 38 of 49 Results - RWAP 3 pcm/sec Case Figure 4.5-1 shows the core power response, which is slowly changing until a reactor trip occurs. The core power rate of increase for the Dominion case is somewhat greater than the USAR data. The Dominion case trips on high neutron flux at about 41 seconds, while the USAR case trips on an OTAT signal at about 45 seconds. It is noted that the USAR core power response very nearly reaches the 118% setpoint for the high flux trip. The difference in reactor trip mechanisms between the Dominion and USAR cases is reasonable, considering the breakpoint for switching between OTAT and high flux occurs at approximately 4 pcm/sec, as shown in USAR Figure 14.1.2-3. The pressure response also affects the OTAT setpoint (setpoint will be lower in the USAR case, due to the USAR pressure response discussed below).

The pressurizer pressure response is shown in Figure 4.5-2. For the Dominion case, the pressure rises faster than the USAR result. For the first 5 seconds the results are similar.

However, the USAR result shows a flat pressure response from about 5 to almost 40 seconds maintaining the pressure at about 2255 psia. In the Dominion case, the pressure steadily increases rising above the USAR value and continuing until the reactor trips.

The RCS pressure response is determined by the modeling assumptions, especially pressurizer spray flow. As noted previously in Section 4.3, the Dominion method uses a conservative value for pressurizer spray flow rate; however, it appears that the USAR model adds additional conservatism, which suppresses the pressure increase associated with the RWAP event.

Figure 4.5-3 shows the RCS Loop A average temperature. There is good agreement between the USAR analysis and the Dominion model, as the peak temperatures are approximately 585 'F for the USAR, and 584 'F for the Dominion model. The time of peak temperature is related to the time of eactor trip as shown in Figure 4.5-1. The sequence of events for the 3 pcrmsec RWAP transient is compared to the USAR in Table 4.5-2.

Table 4.5-2 RWAP 3 pem/sec Time Sequence of Events Event Time (seconds)

DOM USAR Reactivity Insertion at 3 pcrr/sec 0.0 0.0 Reactor Trip Signal Initiated 41.27* 45.28**

• Trip on high neutronflux

• • Trip on OTAT

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 39 of 49 Results - RWAP 100 pcm/sec Case Figure 4.5-4 shows the core power response which rises rapidly until a reactor trip on high flux occurs. The Dominion case trips on a high neutron flux signal of 118% at about 1.8 seconds, compared to about 2.03 seconds for the USAR case (each includes a 0.65 second delay). The Dominion case includes decay heat while the USAR analysis neglects decay heat. The effect of decay heat modeling differences is seen post-trip, where the USAR case power drops to the delayed neutron stable period following shutdown, while the Dominion case follows a decay heat curve defined by the ANS-5.1 1979 single isotope (U-235) model.

The 100 pcm/sec transient is a fast transient and the time period before the reactor trip is so brief that the any differences in fuel pin heat transfer modeling assumptions have little impact on Doppler reactivity feedback.

The pressurizer pressure response is shown in Figure 4.5-5. As is the case with the RWAP analysis for a 3 pcm/sec reactivity insertion rate, the Dominion model predicts a higher pressurizer pressure response. This is mainly due to the differences in modeling of the pressurizer spray.

Figure 4.5-6 shows the RCS Loop A average temperature. There is good agreement between the USAR analysis and the Dominion model, as the peak temperatures are approximately 576 'F for both models. The sequence of events for the 100 pcm/sec RWAP transient is compared to the USAR in Table 4.5-3.

Table 4.5-3 RWAP 100 pcmnsec Time Sequence of Events Event Time (seconds)

DOM USAR Reactivity Insertion at 3 pcm/sec 0.0 0.0 Reactor Trip Signal Initiated 1.78* 2.03*

Trip on high neutronflux

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 40 of 49 Figure 4.5-1 RWAP 3 pcm/sec - Core Power 0'

0 10 20 30 40 50 60 70 80 90 100 Time (seconds)

'-I Figure 4.5-2 RWAP 3 pcm/sec - Pressurizer Pressure

.9 0~

0 10 20 30 40 50 60 70 80 90 100 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 41 of 49 Figure 4.5-3 RWAP 3 pcm/sec - RCS Average Temperature 5805

[ 575.

W I-570' 565*

560, 10 20 30 40 50 60 70 80 90 100 Time (seconds)

Figure 4.5-4 RWAP 100 pcm/sec - Core Power 1.

0.8-

_ 0.6-z 2 3 4 5 6 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 42 of 49 Figure 4.5-5 RWAP 100 pcm/sec - Pressurizer Pressure a

E 0.*

a, a) 0.

1 2 3 4 7 Time (seconds)

Figure 4.5-6 RWAP 100 pcrn/sec - RCS Average Temperature 575 E 570 4,

U M 565

.0 1 2 3 4 5 6 7 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 43 of 49 Summary - RWAP The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the RWAP event. There are small differences in core power response during the early portion of the transient. For lower reactivity insertion rates, the fuel pin heat transfer modeling differences can affect the time to reactor trip; however, the peak core powers and peak average coolant temperatures are in close agreement. The USAR spray flow rate appears to include additional conservatism, resulting in a less severe pressurizer pressure response compared to the Dominion analysis.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 44 of 49 4.6 Loss of Flow The Loss of Flow (LOF) event is the loss of one or two reactor coolant pumps (RCP) and an associated coastdown of reactor flow. The reduction in core flow results in a sudden increase in coolant temperature and increased probability of violating a DNBR limit.

The LOF transient scenario presented here is a complete loss of flow resulting from the trip of both RCPs. The input summary is provided in Table 4:6-1. Where differences from USAR inputs exist they are indicated in the Notes column.

Table 4.6-1 LOF Input Summary Parameter Value Notes Initial Conditions NSSS Power (MW) 1780 Nominal plus pump heat RCS Flow (gpm/loop) 93,000 Minimum measured flow Vessel TAVG (F) 573 Nominal RCS Pressure (psia) 2250 Nominal Pressurizer Level (%) 48 Nominal SG Level (%) 44 Nominal SG Pressure (psia) 804.5 Assumptions/Configuration Pump inertia (lbm-fl2) 72,000 Reactor trip - Low RCS flow is active Automatic rod control - Not credited Pressurizer sprays, PORVs - Not credited Main steam dumps, SG PORV - Not credited SG tube plugging (%) 10 Max value Reactivity Parameters Doppler Temp. Coef (pcm/F) -1.2 USAR uses most negative Doppler-only power coefficient, and a least negative DTC (driven by moderator temp).

Moderator Temp. Coef 0 Results - LOF RCS flow coasts down following LOF as shown on Figure 4.6-1. The Dominion results compare well with the USAR data demonstrating good agreement between the pump model and friction losses. Since there is minimal reactivity feedback, the core power remains nearly constant prior to the reactor trip on low RCS flow, and the Dominion and USAR power response compares well as seen on Figure 4.6-2. The resulting core average heat flux is provided on Figure 4.6-3. The Dominion case provides a higher, more conservative value

DOM-NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 45 of 49 than the USAR case. This is primarily due to the fact that the Dominion model is initialized at a higher fuel temperature with higher stored energy. The USAR assumes a higher gap conductance and U0 2 thermal conductivity resulting in lower initial fuel temperature and stored energy. A sensitivity case is also shown on Figure 4.6-3 that modifies the inputs to more closely match the USAR case initial conditions and assumptions, including the higher gap heat transfer conductance and U0 2 thermal conductivity multiplier. In this case, the heat flux closely tracks the USAR value.

The RCS loop temperature response is shown on Figure 4.6-4. Again, the higher stored energy results in higher loop temperatures for the Dominion case compared to the USAR response. This is also reflected in the pressurizer pressure response shown on Figure 4.6-5 where the RCS pressure for the Dominion case peaks above the USAR case and remains higher for the duration of the event. The effect of pressurizer liquid flashing after the pressure peak can be seen by a reduction in the rate of pressure decrease.

Table 4.6-2 LOF Results Parameter DOM USAR Sequence of Events:

Reactor Trip (sec) 2.61 2.57 (Low RCS flow)

Summary - LOF The Dominion Kewaunee LOF analysis provides RCS flow response that is very similar to the USAR results, demonstrating close agreement between the pump model and friction losses. The timing for reactor trip and the core power response is also in close agreement.

The core heat flux is higher for the Dominion case due to differences in initial fuel temperature and stored energy. This is also reflected in higher primary side pressure and temperature values. The higher heat flux is conservative for DNBR acceptance criteria.

DOM-NAF-5-0.0-A Attachment B RETRAN Benchrnarking Information pg. 46 of 49 Figure 4.6-1 LOF - Total Core Inlet Flow 460 400 340 S 280 220 160 100 0 2 4 6 8 10 12 Time (seconds)

Figure 4.6-2 LOF - Nuclear Power 1.20 1.00 S0.80' 0.60' 0

a)

S0.40' 0 2 4 6 8 10 12 Time (seconds)

DOM -NAF-5-0,0-A Attachment B RETRAN Benchma rking Information pg. 47 of 49 Figure 4.6-3 LOF - Core Average Heat Flux 2 4 6 1 8 10 12 Time (seoonds)

Figure 4.6-4 LOF - RCS Loop Temperature 640 620 OM

ýDOM HOT LEG 600 U-USAA Rý

. 580 12a-E 60 540 nN n, 1Fr.

r9n 0 2 4 6 8 10 12 Time (seconds)

DOM -NAF-5-0.0-A Attachment B RETRAN Benchmarking Information pg. 48 of 49 Figure 4.6-5 LOF - Pressurizer Pressure 2360 a) 0 2 4 6 10 12 Time (seconds)

DOM-NAF-5-0.0-A Attachment B RETRAN Benchma rking Information pg. 49 of 49 5.0 Conclusions This report presents demonstration transient analyses performed with the KPS RETRAN model developed in accordance with VEP-FRD-41. These analysis results are compared with current Kewaunee USAR results. The following conclusions are drawn based on these analyses.

1) This report demonstrates that the Dominion RETRAN-02 model and analysis methods can predict the response of transient events with results that compare well to USAR results.

2) Where there are differences between the Dominion results and the USAR results, they are understood based on differences in noding, inputs, or other modeling assumptions.

3) The Dominion Kewaunee RETRAN-02 model is consistent with current Dominion methods (Reference 1). These methods have been applied extensively for Surry and North Anna licensing, engineering and plant support analyses.

4) The RETRAN comparison analyses satisfy the DOM-NAF-5 applicability assessment criteria and provide further validation of the conclusion that Dominion's RETRAN analysis methods are applicable to-Kewaunee and can be applied to Kewaunee licensing analysis for reload core design and safety analysis.

6.0 References

1) Topical Report, VEP -FRD-4 1, Rev. 0.1-A, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," June 2004.

2) Topical Report, VEP-NFE-2-A, "VEPCO Evaluation of the Control Rod Ejection Transient," December 1984.

IMPLEMENTATION OF THE DOMINION STATISTICAL DNBR EVALUATION METHODOLOGY WITH VIPRE-D/WRB-1 AT KEWAUNEE POWER STATION (KPS)

Prepared By: 4/Dat7 Date Reviewed By. 419/200, Date Approved By:

Date Supervisor. Nuclear Safety Analysis

TABLE OF CONTENTS TABLE OF CONTENTS .......................................................................................................... 2

1. INTRODUCTION ...................................................................................................................... 3

2. BACKGROUND .............................................................................................................. 4

3. IMPLEMENTATION OF THE STATISTICAL DNBR EVALUATION M ETHODOLOGY .................................................................................... 6 3.1. METHODOLOGY REVIEW .................................... 6 3.2. UNCERTAINTY ANALYSIS...................... 7 3.3. CUF CORRELATIONS .......................... ...

3.4. MODEL UNCERTAINTY TERM ................................................... 10 3.5 CODE UNCEKRAINwY.................................... 10 3.6. MONTE CARLO CALCULATIONS ........................................................ 11 3.7. FULL CORE DNB PROBAmILITY SUMMATION.

3.8. VERIFICATION OF NOMINAL STATEPOINTS ....

3.9. SCOPE OF APPLICABILITY ................ . ................. .. 1 310.

SUMMARY

OF ANALYSIS ................ ..... ... -... .......... £ F

4. APPLICATION OF VIPRE-D/WRB-1/W-3 TO KPS ............................... .................... ... 20 4.1. VIPRE-D/WRB-1 STATISTICAL DFI.SGN LIMIT (SDL) FOR KEWAUNEE ................ 20 4.2. VERIFICATION OF EXISTING REACTOR CORE SAFETY LIMITS, PROTECTION SETPOINTS AND CHAPTER 14 EVENTS . . ....................... 21

5. CONCLUSION S ..................................................................................................................... 22

6. REFEREN CES ....................................................................................................................... 23 2

1. Introduction DOM-NAF-5 (Reference 7) was submitted to the NRC, in August 2006, to document application of Dominion nuclear core design and safety analysis methods to Kewaunee Power Station (KPS). This report provides the plant specific application for KPS cores containing Westinghouse 422V+ fuel assemblies, in accordance with Section 3.5 of DOM-NAF-5. Specifically, this report supports the application of U.S. Nuclear Regulatory Commission (USNRC) approved Dominion Topical Report VEP-NE-2-A. "Statistical DNBR Evaluation Methodology" (Reference 1) to KPS. It provides the technical basis and documentation required by the USNRC to evaluate the plant specific application of VEP-NE-2-A methods to KPS. This application employs the VIPRE-D code with the Westinghouse WRB-1 Critical Heat Flux (CHF). correlation (VIPRE-D)WRB-1) for the thermal-hydraulic analysis of Westinghouse 422V+ fuel assemblies at KPS. In particular, Dominion requests the review and approval of the Statistical Design Umit (SDL) documented herein as per 10 CFR 50.59(c)(2Xvii) it constitutes a Design Basis Limit for Fission Products Barrier (DBLFPB).

The approval of a subsequent license amendment request (LAR) to add DOM-NAF-5 (Reference 7) to Section 6.9.aA of the KPS Technical Specifications, will provide Dominion with the ability to use the VIPRE-D/WRB-1 code/correlation set to perform licensing calculations for Westinghouse 422V+ fuel in KPS cores, using the Determinislic Design Limit (DDL) qualified in DOM-NAF-5 (which includes references to DOM-NAF-2-A) and the SDL documented herein. The intended applications are described in DOM-NAF-5.

3

2. Background Kewaunee Power Station (KPS) became part of the Dominion nuclear fleet following Dominion's acquisition of KPS in July 2005. Currently, departure from nuclear boiling (DNB) analyses required to support the use of Westinghouse 422V+ fuel at KPS are performed using the Westinghouse Revised Thermal Design Procedure (RTDP) methodology which was approved for Kewaunee's fuel transition to Westinghouse's 422V+

fuel (Reference 5).

This report documents the plant specific application of the USNRC approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," (Reference 1) for KPS cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-DIWRB-1 code/correlation in accordance with Section 3.5 of DOM-NAF-5.

In 1985, Virginia Power (Dominion) submitted to the USNRC Topical Report VEP-NE-2-A describing a proposed methodology for the statistical treatment of key uncertainties in core thermal-hydraulic DNBR analysis. The methodology provided DNBR margin through the use of statistical rather than deterministic uncertainty treatment. The methodology was reviewed and approved by the USNRC in May 1987, and the SER provided by the USNRC listed the following conditions for its use (Reference 8):

1) The selection and justification of the Nominal Statepoints used to perform the plant specific implementation must be Included in the submittal (Sections 3.6 and 3.8).

2) Justification of the distribution, mean and standard deviation for all the statistically treated parameters must be included in the submittal (Section 32).

3) Justification of the value of model uncertainty must be included in the plant specific submittal (Section 3.4).

4) For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be Included In the submission, unless there is an approved Topical Report documenting these (such as Reference 2).

This report provides the technical basis for Implementation of the Dominion Statistical DNBR Evaluation Methodology for Westinghouse 422V+ fuel at Kewaunee with VIPRE-DIWRB-1, as well as the SDL obtained by this implementation. This report also documents that the existing Reactor Core Safety Limits and protection functions (OTAT, OPAT, FAI, etc) do not require revision as a consequence of this implementation. The list of.

USAR transients for which the code/correlations will be applied is also included herein.

4

Section 3 of this report summarizes the implementation of the Dominion Statistical DNBR Evaluation Methodology to Westinghouse 422V+ fuel at Kewaunee Power Station with the VIPRE-D/WRB-1 code/correlatiori. Section 4 provides the necessary information for the plant specific application of the VIPRE-DNVRB-1 code/correlation to Kewaunee, Including the SDL. The verification of the existing Reactor Core Safety Limits, Protection Setpoints and KPS USAR Chapter 14 events with the above DNBR limits Is documented in Section 4-5

3. Implementation of the Statistical DNBR Evaluation Methodology 3.1. Methodology Review The Statistical DNBR Evaluation Methodology (Reference 1) is employed herein to determine a statistical DNBR limit for KPS. This new limit combines the correlation uncertainty with the DNBR sensitivities to uncertainties in key DNBR analysis input parameters. Even though the new DNBR limit (the Statistical Design Limit or SDL) is larger than the deterministic code/correlation design limit (DDL), its use is advantageous as the Statistical DNBR Evaluation Methodology permits the use of nominal values for operating initial conditions instead of requiring the application of evaluated uncertainties to the initial' conditions for statepoint and transient analysis.

The SDL is developed by means of a Monte Carlo analysis. The variation of actual operating conditions about nominal statepoints due to parameter measurement and other key DNB uncertainties is modeled through the use of a random number generator. Two thousand random statepoints are generated for each nominal statepoint The random statepoints are then supplied to the thermal-hydraulics code VIPRE-D, which calculates the minimum DNBR (MDNBR) for each statepoint. Each MDNBR Is randomized by a code/correlation uncertainty factor as described in Reference 1 using the upper 95%

confidence limit on the VIPRE-DAWRB-1 measured-to-predicted (M/P) CHF ratio standard deviation (Reference 2). The standard deviation of the resultant randomized DNBR distribution is increased by a small sample correction factor to obtain a 95% upper confidence limit, and is then combined Root-Sum-Square with code and model uncertainties to obtain a total DNBR standard deviation (stW). The SDL is then calculated as:

SDL = 1 + 1.645*so [Eq. 3.1]

in which the 1.645 multiplier is the z-value for the one-sided 95% probability of a normal distfibution. This SDL thus provides peak fuel rod DNB protection at greater than 95/95.

As an additional criterion, the SDL is tested to determine the full core DNB probability when the peak pin reaches the SDL. This process is performed by summing the DNB probability of each rod in the core, using a bounding rod census curve and the DNB sensitivity to rod power. If necessary, the SDL is increased to reduce the full core DNB probability to 0.1% or.

less.

6

3.2. Uncertainty Analysis This section is included herein to satisfy Condition 2 in the SER (Reference 8) of VEP-NE-2-A (Reference 1).

Consistent with VEP-NE-2-A, inlet temperature, pressurizer pressure, core thermal power, reactor vessel flow rate, core bypass flow, the nuclear. enthalpy rise factor and the engineering enthalpy rise factor were selected as the statistically treated parameters in the implementation analysis. The magnitudes and functional forms of the uncertainties for the statistically treated parameters were derived in a rigorous analysis of plant hardware and measurement/calibration procedures, and have been summarized in Table 3.2-1.

The uncertainties for core thermal power, vessel flow rate, pressurizer pressure and inlet temperature were quantified using all sensor, rack, and other components of a total uncertainty and combined in a manner consistent with their relative dependence or Independence (Reference 6). Westinghouse quantified these uncertainties (Reference 6) for Kewaunee's transition to Westinghouse's 422V+ fuel (Reference 5). Total uncertainties were quantified at the 2(r level, corresponding to two-sided 95% probability. The standard deviations a Were obtained by dividing the total uncertainty by 1.96, which is the z-value for the two-sided 95% probability of a normal distribution.

The magnitude and distribution of uncertainty for pressurizer pressure (system pressure) per the pressurizer pressure control system were quantified. The pressurizer pressure uncertainty is a normal, two-sided, 95% probability distribution with a magnitude of +/-35.1 psi (Reference 6) and a standard deviation (a)of 17.9 psi.

The magnitude and distribution of uncertainty on the average temperature (Tavg) per the Tavg rod control system were quantified. The average temperature uncertainty is a normal, two-sided, 95% probability distribution with a magnitude of +/-4.9'F (Reference 6) and a standard deviation (a) of 2.50F.

The core power uncertainty is defined as a normal, two-sided, 95% probability distribution with a magnitude of 1.7% (Reference 6) and a standard deviation of 0.9%.

The reactor coolant system (RSC) flow uncertainty is defined as a normal, two-sided, 95%

probability distribution with a magnitude of 2.7% (Reference 6) and a. standard deviation of 1.4%.

The two-sided, 95195 tolerance interval (95%- probability, 95% confidence) for the measurement uncertainty of the nuclear enthalpy rise factor, FmN, is 3.5%. Conservatively, the measured Fm" uncertainty was defined as a normal distribution with a 4% tolerance interval for consistency with previous applications.

The magnitude and distribution of uncertainty on the engineering hot channel factor, FmE, was quantified as a normal probability distribution with a magnitude of +/- 3.0%. The 7

Statistical DNBR Evaluation Methodology (Reference 1) treats the FpE uncertainty as a uniform probability distribution.

The total core bypass flow consists of separate flow paths through the thimble tubes, direct leakage to the outlet nozzle, baffle joint leakage flow, upper head spray flow and core-baffle gap flow. These five components were each quantified based on the current Kewaunee core configuration, their uncertainties conservatvely modeled and the flows and uncertainties totaled. The Monte Carlo analysis ultimately used a best estimate bypass flow of 5.1% with an uncertainty of 1.9%. The.analysis assumed that the probability was uniformly distributed.

8

Table 3.2-1: Kewaunee Parameter Uncertainties (Reference 6)

PARAMETER NOMINAL STANDARD UNCERTAINTY DISTRIBUTION VALUE DEVIATION Pressure 2250 17.9 psi 35.1 psi at 2o Normal

[psia Temperature 541.2 2.5°F 4.9°F at 2cy Normal FF] .,

Power [MW] 1,772 0.9% 1.7% at 2a Normal Flow [gpm] 186,000, 1.4% 2.7% at 2, Normal FM N 1.635 2.0% 4.0% at 2y Normal FmE 1.0 N/A 3.0% Uniform Bypass [%] 5.1 N/A 1.9% Uniform 9

3.3. CHF Correlations The WRB-1/W-3 CHF correlations are used for the calculation of DNBRs in Westinghouse 422V+ fuel assemblies. Only WRB-1 is applicable to the operating conditions for which the Statistical DNBR Evaluation Methodology applies. Table 3.3-1 presents the Design Limit correlation data for VIPRE-D/WRB-1. The W-3 correlation is only used below the first mixing grid or when the operating conditions are outside of the range of validity of the WRB-1 CHF correlation, such as the main steam-line break evaluation, where there are reduced temperature and pressure. The W-3 CHF correlation is always used deterministically.

Table 3.3-1: CHF CodelCorrelatlon Data (Reference 2)

WRB-1 Average MIP .1.005 S(MIP) 0.083 n 945 KW 1.03963 Kx S(MIP) 0.08629 3.4. Model Uncertainty Term This section is included herein to satisfy Condition 3 in the SER (Reference 8) of the Statistical DNBR Evaluation Methodology Topical Report (Reference 1).

The VIPRE-D 20-channel production model for Kewaunee was used in the development of the VIPRE-DNVRB-1 SDL for Kewaunee. Since this is the production model that Dominion intends to use for all Kewaunee evaluations once the Topical Report DOM-NAF-5 is approved, and the VIPRE-D code becomes the approved code for the determination of the core operating limits in the Core Operating Limits Report (COLR), there Is no additional uncertainty associated with the use of this model. In summary, it is concluded that no correction for model uncertainty is necessary, and the model uncertainty term is set to zero for the calculation of the total DNBR standard deviation.

3.5 Code Unceirtainty The code uncertainty accounts for any differences between Dominion's VlPRE-D and Westinghouse's THINC codes, with which the WRB-1 CHF data were correlated, and any Kis a sample size correction factor that gives a one-sided 95% upper confidence limit on the estimated standard deviation of a given population. It can be calculated .as:

K=" 2(nn-1) 10

effect due to the modeling of a full core with a correlation based upon bundle test data.

These uncertainties are clearly independent of the correlation, the model, and parameter induced uncertainies. The code uncertainty was quantified at 5%; consistent with the factors specified for other thermal/hydraulic codes in Reference 1. The basis for this uncertainty is described in detail by USNRC staff in Reference 8. In Reference 8, the USNRC Staff refers to the 5% uncertainty as being a 2a value. The 5% code uncertainty is certainly conservative in light of the excellent V1PRE-DNIPRE-W and VIPRE-D/CHF data comparisons. Howeyer, the 5% uncertainty serves as a conservative factor that may be shown to be wholly or partially unnecessary at a later time. A one-sided 95% confidence level on the code uncertainty is then 3.04% (=5.0%/1.645). The use of the 1.645 divisor (the one-sided 95% tolerance interval multiplier) is conservative since the USNRC Staff considers the 5% uncertainty to be a 2c value.

3.6. Monte Carlo Calculations In order to perform the Monte Carlo analysis, nine Nominal Statepoints covering the full range of normal operation and anticipated transient conditions were selected. These statepoints must span the range of conditions over which the statistical methodology will be applied. Two statepoints were selected at each of the four Reactor Core Safety Umit (RCSL) pressures (2426, 2250, 2000, and 1800 psia). For each of the RCSLs, a high power, 120%, and low power, near the intercept of the DNBR limit line with the vessel exit boiling line, were chosen. In order to apply the rrmethodology to low flow events, a low flow statepoint is also included (Statepoint Z). The inlet temperature used for each statepoint is calculated by determining the inlet temperature that would result In the desired MDNBR (1.24) for each statepoint. The selected Nominal Statepoints are listed in Tables 3.6-1.

Table 3.6-1: Nominal Statepoints for Westinghouse 422V+ Fuel at Kewaunee with VIPRED-/WRB-1 STATE PRESSURIZER INLET POINT PRESSURE TEMPERATURE POWER FLOW MDNBR Ppsial [%] [%] FHrN 1 2425.0 572.1 120% 100% 1.635 1.243 2 2250.0 562.5 120% 100% 1.635 1.244 3 2000.0 548.3 120% 100% 1.635 1.244 4 1800.0 536.2 120% 100% 1.635 1.244 5 2000.0 580.9 100% 100% 1.635 1.243 6 1800.0 563.2 104% 100% 1.635 1.244 7 2250.0 606.6 90% 100% 1.684 1.241 8 2425.0 623.6 85% 100% 1.709 1.242 9 1 2250.0 541.2 100% 63.5% 1.635 1.247 The Monte Carlo analysis itself consisted of 2000 calculations performed around each of the nine Nominal Statepoints. As described in Section 3.1, the DNBR standard deviation at each Nominal Statepoint was augmented by the code/correlation uncertainty, the small "The part-power multiplier described in the Kewaunee Core Operating Limits Report (COLR) is used for less than 100% power statepolnts.

11

sample correction factor, and the code uncertainty to obtain a total DNBR standard deviation.

The Total STot, is obtained using the Root-Sum-Square method according to Equation 3.2:

STOTAL =

2 SDNBR {21.0+ { n 1-1 1 0 12

'.+

+{

2 22

+F2 +F2 [Equation 3.21 where

• SDWR is the standard deviation for the Randomized DNBR distribution.

• The factor { !1_1.0t is the uncertainty in the standard deviation of the 2,000 Monte Carlo simulations, and provides a 95% upper confidence limit on the standard deviation.

" YJ-. is the uncertainty in the mean of the correlation. N is the number of number of degrees of freedom in the original correlation database.

" F, is the code uncertainty, that has been defined as 5% (2a value), i.e. 5.0%/11.645

=3.04% (1y value). See Section 3.1.5 in Reference 1.

• FMis the model uncertainty, which is 0.0 in our caseas we are running the Monte Cado simulation with the production model (see Section 3.1.4 In Reference 8).

Note that this equation differs slightly from the equation listed In Reference 1. It has an additional factor applied to the Randomized DNBR SDNBR, the / factor to correct for the uncertainty in the mean of the correlation. This factor has been used in previous implementations of the Statistical DNBR Evaluation Methodology, such as Reference 3 and Reference 4.

The limiting peak fuel rod SDL was calculated to be 1.24 for VIPRE-DNWRB-1. The Monte Carlo Statepoint analysis is summarized in Table 3.6-2.

12

Table 3.6-2: Peak Pin SDL Results for Kewaunee 422V+ with VIPRE-D/WRB-1 STATEPOINT Randomized Total DNB Pin Peak DNB SDNBR3 STOTAL SDLgWt,5 1 0.1309 0.1398 1.230 2 0.1338 0.1428 1.235 3 0.1339 0.1429 1.235 4 0.1298 0.1386 1.228 5 0.1306 0.1395 1.229 6 0.1342 0.1432 1.236 7 0.1320 0.1409 1.232 8 0.1299 0.1388 1.228 9 0.1310 0.1398 1.230 3.7. Full Core DNB Probability Summation After the development of the peak pin 95/95 DNBR limit, the data statistics are used to determine the number of rods expected in DNB. The DNB sensitivity to rod power is estimated as a(DNBR)/ a(I/FVh). The specific values of a(DNBR).a(1I/Fh), denoted 0. are listed in Table 3.7-1.

To ensure that the calculations are conservative, a one-sided -tolerance limit of f3is used:

in which:

• [3 is the one-sided tolerance limit on P

• t(oyv) is the T-statistic with significance level a and v degrees of freedom. For 2,000 observations at a 0.05 level of significance t(O.05,2000) = 1.645.

• se(13) is the standard error of 6-The variable 1/FAh is the most statistically significant independent variable in the linear regression model, yielding RW values larger than 99%. The value of the stafistic parameter F of I/FAh was the largest for all statepoints. which indicates that the variable 1/FAJh accounts for the largest amount of the variation in the DNBR.

Table 3.7-1: D(DNBR)/Y (l/Fi~h) Estimation for WRB-1 STATEPOINT se() -80 1 4.05387 0.00914 4.038a4 99.5%

2 4.18659 0.00861 4.17243 99.6%

3 4,31365 0.00878 4.29921 99,6%

4 4.33980 0.00951 4.32416 99.5%

5 4.16476. 0.00962 4.14894 99.5%

6 4.14116 0.00951 4.12551 99.5%

7 4.19940 0.00997 4.18301 99.5%

8 4.03524 0.01006 4.01869 99.5%

9 4.41089 0.00936 4.39549 99.5%

A representative fuel rod census curve used for the determination of the SDL is listed in Table 3.7-2. The full core DNB probability summation will be reevaluated on a reload basis 13

to verify the applicability of the fuel rod census (F &-N versus % of core with F*nN greater than or equal to a given FAH limit) used in the implementation analysis. The DNB probability summation for V1PRE-DIWRB-1 is summarized in Table 3.7-3.

Table 3.7-2: Representative Fuel Rod Census for a Maximum Peaking Factor FAh = 1.635 MAXIMUM % OF FUEL RODS IN CORE WITH FAh>>L to:

0.0 1.6350 0.1 1.6330 0.2 1.6300 0.3 1.6270 0.4 1.6230 0.5 1.6194 0.6 1.61 74 0.7 1.6154 0.8 1.6124 0.9 1.6104 1.0 1.6074 1.5 1.5977 2.0 1.5896 2.5 1,5805 3.0 1.5725 4.0 1.5594 5.0 1.5514 6.0 1.5453 7.0 1.5412 8.0 1.5362 9.0 1.5315 10.0 1.5271 20.0 1.4676 30.0 1.4213 40.0 1.3709 PEAK 1.635 14

Table 3.7-3: Full Core DNB Probability Summation for Kewaunee 422V+ with VIPRE-D/WRB-1

% of Rods in Full Core STATEPOINT STOTAL DNB SDL9* 9 1 0.1398 0.0992 1.232 2 0.1428 0.0991 1.236 3 0.1429 0.0983 1.234 4 0.1386 0.0983 1.224 5 0.1395 0.0992 1.229 6 0.1432 0.0990 1.238 7 0.1409 0.0984 1.232 8 0.1388 0.0996 1.230 9 0.1398 0.0988 1.225 15

3.8. Verification of Nominal Statepoints Condition 1 of the USNRC's safety evaluation report for Reference 1 (Reference 8) requires that the Nominal Statepoints be shown to provide a bounding DNBR standard deviation for any set of conditions to which the methodology may potentially be applied.

It is therefore necessary to demonstrate that st as calculated herein Is maximized for any conceivable set of conditions at which the core may approach the SDL To do so, a regression analysis is performed using as dependent variable the unrandomized DNBR standard deviations at each Nominal Statepoint (i.e. the raw MDNBR results obtained from the Monte Carlo simulation). The Nominal Statepoint pressures, inlet temperatures, powers and flow rates are used as the independent variable. If no dear trend appears in the plot it can be concluded that the standard deviation has been maximized. If a clear trend is displayed, the regression function is determined. This regression equation is evaluated to determine the values of the independent variable for which the standard deviation would be maximized, and it is verified that the Nominal Statepoints selected bound those conditions.

In addition, the residuals of the regression are plotted again against all the independent variables, and it is verified that no trends are discernible.

Table 3.8-1 shows the RW coefficients obtained for the verification of the nominal statepoints. The largest linear curve fit RW coefficient is 9.33%, thus confirming that there is not dependence.

An evaluation of all the data, linear fits, and RW coefficients indicates that there are no discernible trends in the database. Therefore, it was concluded that STOTAL had been maximized for any conceivable set of conditions at which the core may approach the SDL and that the selected Nominal Statepoints provide a bounding standard deviation for any set of conditions to which the methodology may potentially be applied. Figure 3.8-1 displays a sample regression plot for WRB-1 and clearly shows the trends discussed above.

16

Table 3.8-1: W2 Coefficients for the Verification of the Nominal Statepoints for Kewaunee 422V+ with VIPRE-D/WRB-1 R*z Linear Regression Pressure 6.97%

Temperature 4.93%

Flow Rate 8.33%

Power 9.33%

z E

02 0.0000 4-0.15 0.17 0.19 0.21 0.23 0.25 0.27 Power (MBTUIhr-ftA2) 1.- SIGMA DNBR -Unear (SIGMA DNBR)

Figure 3.8-1: Variation of the Unrandomized Standard Deviation with Power for the WRB-1 CHF Correlation 17

3.9. Scope of Applicability This section is included herein to satisfy Condition 4 in the SER (Reference 8) of VEP-NE-2-A (Reference 1).

The Statistical DNBR Evaluation Methodology may be applied to all Condition I and II DNB events (except Rod Withdrawal from Subcritical (RWFS) which is initiated from zero power), and to the Loss of Flow analysis and the Locked Rotor Accident. The accidents to which the methodology is applicable are listed In Table. 3.9-1. This table corresponds to Table 3.5.1 in Reference 7. The range of application is consistent with previous applications of Dominion Statistical DNBR Evaluation Methodology for other Dominion PWR applications (North Anna and Surry). This methodology will not be applied to accidents that are initiated from zero power where the parameter uncertainties are higher.

The Statistical DNBR Evaluation Methodology provides analytical margin by permitting

-transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, FbHN (measurement component) and hot channel uncertainties. These uncertainties are convoluted statistically into the DNBR limit.

Table 3.9-1: USAR Transients Analyzed with VIPRE-D/WRB-1iW-3 for Kewaunee (Reference 7, Table 3.5.1)

KPS USAR ACCIDENT APPLICATION SECTION Rod cluster control assembly bank withdrawal from subcritical 14.1.1 DET-DNB Rod cluster control assembly bank withdrawal at STAT-DNB power 14.1.2 Rod cluster control assembly misalignment / STAT-DNB Dropped rod/bank 14.1.3._______

Uncontrolled boron dilution 14.1.4 Non-DNB Full and partial loss of forced reactor coolant flow 14.1.8 STAT-DNB Startup of an inactive reactor coolant loop 14.1.5 Non-DNB Loss of external electrical load and/or turbine trip 14.1.9 STAT-DNB Loss of normal feedwater 14.1.10 Non-DNB Loss of offsite power 14.1.12 Non-DNB Excessive heat removal due to feedwater system 14.1.6 STAT-DNB malfunction Excessive load increase 14.1.7 STAT-DNB Rupture of a main steam pipe 14.2.5 DET-DNB Locked reactor coolant pump rotor or shaft break 14.1.8 STAT-DNB 18

3.10. Summary of Analysis The steps of-the SDL derivation analysis may be summarized as follows:

In accordance with the Statistical DNBR Evaluation Methodology, 2,000 random statepoints are generated about each nominal statepoint and VIPRE-D is then executed to obtain MDNBRs. The standard deviation for the distribution of 2,000 MDNBRs is referred to as the unrandomized standard deviation. At the limiting Nominal Statepoint (W), the standard deviation of the randomized DNBR distributions, which is the unrandomized corrected for CHF correlation uncertainty, was found to be 0.1342. This value was then combined Root Sum Square with code and model uncertainty standard deviations to obtain a total DNBR standard deviation of 0.1432, as listed in Table 3.6-2.

The use of 0.1432 in Equation 3.1 yields a peak pin DNBR limit of 1.236 with at least 95% probability at a 95% confidence level. The total DNBR standard deviation was then used to obtain 99.9% DNB protection in the full core of 1.238. Therefore the VIPRE-DJWRB-1 SDL for Westinghouse 422V+ fuel is set to 1.24.

19

4. Application of VIPRE-DIWRB-1/W-3 to KPS VIPRE-DJWRB-1 together with the Statistical DNBR Evaluation Methodology will be applied to all Condition I and II DNB events (except Rod Withdrawal from Subcritical, RWFS), and to the Complete Loss of Flow event and the Locked Rotor Accident The Statistical DNBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, FaN (measurement component) and FM E uncertainties. These uncertainties are convoluted statistically into the DNBR limit.

In addition, there are a few events that will be evaluated with the VIPRE-DMW-3 and deterministic models because they do not meet the applicability requirements of the Statistical DNBR Evaluation Methodology (see Table 3.9-1, DET-DNB events). These events will be initiated from bounding operating conditions considering the nominal value and the appropriate uncertainty value, and require the application of the bypass flow, FMN (measurement component) and FM2 uncertainties. The events modeled deterministically are limited by the deterministic design limit (bDL) stated in DOM-NAF-2-A (Reference 2).

4.1. VIPRE-DJWRB-1 Statistical Design Limit (SDL) for Kewaunee The Statistical Design Limit for Kewaunee cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-D/WRB-1 code was derived in Section 3 of this report. The SDL for VIPRE-DIWRB-1 is determined to be 1.24. The SDL limit provides a peak fuel rod DNB protection with at least 95% probability at a 95% confidence level and a 99.9% DNB protection for the full core. This SDL is plant specific as it already includes the Kewaunee specific uncertainties for the key parameters accounted for in the application of the Statistical DNBR Evaluation Methodology. Therefore, this limit is applicable to the analysis of statistical DNB events of Westinghouse. 422V+ fuel in Kewaunee cores with the VIPRE-DNVRB-1 code.

20

4.2. Verification of Existing Reactor Core Safety Limits, Protection Setpoints and KPS USAR Chapter 14 Events This section is included herein to satisfy Condition 3 in the SER (Reference 8) of VEP-NE-2-A (Reference 1).

To demonstrate that the DNB performance of the Westinghouse 422V+ fuel is acceptable, Dominion performed calculations for full-core configurations of Westinghouse 422V+ fuel.

The calculations were performed using the VIPRE-DNVRB-1 and VIPRE-D/W-3 code/correlation pairs and selected statepoints including: the reactor core safety limits (RCSL), axial offset limits (AO), rod withdrawal from subcritical (RWFS), rod withdrawal at power (RWAP), loss of flow (LOFA), locked rotor events (LOCROT), hot zero power steam line break (MSLB), dropped rod limit line (DRLL), and static rod misalignment (SRM).

These various statepoints provide sensitivity of DNB performance to 'the following: (a) power level (including the impact of the part-power multiplier on the allowable hot rod power FAh), pressure and temperature (RCSL); (b) limiting axial flux shapes at several axial offsets (AO); and (c) low flow (LOFA and LOCROT). The statepoints for the RWFS and MSLB were evaluated with deterministic DNB. methods. The remaining statepoints were evaluated using statistical DNB methods. The evaluation criterion for these analyses is that the minimum DNBR must be equal to or greater than the applicable safety analysis limit (SAL) listed below.

The results of the calculations demonstrate that the minimum DNBR values are equal to or greater than the applicable safety analysis limit for all of the analyses that are performed to address statepoints of the Reactor Core Safety Limits, the OTAT, OPAT and FAI trip setpoints, as well as all the evaluated Chapter 14 events (including the LOFA and LOCROT) with an FAh of 1.635 (COLR limit of 1.7 divided by the measurement uncertainty of 1.04 = 1.635).

VIPRE-DIWRB-1 DDL 1.17 SDL 1-24 SAL 1.31 VIPRE-D/W-3 D-DL (;000 psia) 1.30 DDL (<1000 psia) 1.45 SAL (21000 psla) 1.42 SAL (<1 000 psia) 1.58 21

5. Conclusions This report supports the application of USNRC approved VEP-NE-2-A to KPS. Itprovides the technical basis and documentation required to evaluate the plant specific application of VEP-NE-2-A methods to KPS. This application employs the VIPRE-D code with the Westinghouse WRB-1 Critical Heat Flux (CHF) correlation (VIPRE-D/WRB-1) for the thermal-hydraulic analysis of Westinghouse 422V+ fuel assemblies at KPS. In particular, Dominion requests the review and approval of the Statistical Design Umit (SDL) of.1.24 as documented herein as per 10 CFR 50.59(cX2Xvli) it constitutes a Design Basis Umit for Fission Products Barrier (DBLFPB).

22

6. References

1. Topical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," R. C.

Anderson, June 1987.

2. Fleet Report, DOM-NAF-2, Rev 0.0-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," R. M. Bilbao y Le6n, August 2006.

3. Letter from E. S. Grecheck (VEPCO) to US NRC Document Control Desk, "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 - Proposed Technical Specification Changes Addition of Analytical Methodology to COLR,"

Serial No.05-419, July 5, 2005.

4. Letter from W. L. Stewart (VEPCO) to US NRC Document Control Desk, "Virginia Electric and Power Company Surry Power Station Units I and 2 - Proposed Technical Specification Changes - FAh IncreaselStatistical DNBR Methodology,"

Serial No.91-374, July 8, 1991.

5. Letter from Mark Warner (NMC) to US NRC Document Control Desk, "Kewaunee Nuclear Power Plant License Amendment Request 187 to the Kewaunee Nuclear Power Plant Technical Specifications, Conforming Technical Specification Changes for Use of Westinghouse VANTAGE+ Fuel", Serial No.02-067, July 2, 2002.

6. Technical Report, WCAP-1 5591, Rev. 1 (proprietary), "Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology Kewaunee Nuclear Plant,"

December 2002.

7. Letter from G. T. Bischof (Dominion) to Document Control Desk (USNRC), "Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," Serial No.06-578 dated August 16, 2006.

8. Letter from L. B. Engle (NRC) to W. L. Stewart (Virginia Power), "Statistical DNBR Evaluation Methodology, VEP-NE-2, Surry Power Station, Units No. 1 & No. 2 (Surry-I&2) and North Anna Power Station, Units No. 1 & No. 2 (NA-I&2)," Serial No.87-335 dated May 28, 1987.

23

Topical Report DOM-NAF-5, Rev. 0.0- A Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)

Attachment 1 NRC Request for Additional Information on DOM-NAF-5 and Dominion Responses dated June 12, 2007 11 pages after the cover page

Dominion Energy Kcwaunee, Inc. -¥DO

',000)D~ominion Boulevard. Glen,Allen. VA 23060 J D m n oi June 12, 2007 U. S. Nuclear Regulatory Commission Serial No. 06-578D Attention: Document Control Desk NL&OS/CDS: R4 Washington, DC 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO NRC QUESTIONS REGARDING KEWAUNEE REQUEST FOR APPROVAL OF TOPICAL REPORT DOM-NAF-5. "APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS)"

In a January 31, 2006, public meeting with Nuclear Regulatory Commission (NRC) staff, Dominion Energy Kewaunee, Inc. (DEK) presented a conceptual approach and implementation strategy for application of existing NRC-approved nuclear core design and safety analysis methods to Kewaunee Power Station (KPS) (reference 1). These design and analysis methods are already in use within the remainder of the Dominion fleet. Fundamental to the proposed approach was creation of a composite topical report (DOM-NAF-5) that would document the application of the relevant methodologies to KPS.

On August 16, 2006, DEK submitted Dominion Topical Report DOM-NAF-5 without attachments A and B (reference 2). Attachment A to DOM-NAF-5, containing Core Management Systems benchmark analysis results, was submitted on December 6, 2006 (reference 3). On April 16, 2007, DEK submitted Attachment B to DOM-NAF-5, containing RETRAN benchmark analysis results (reference 4). This submittal, in conjunction with References 2 and 3, provided the complete contents of DOM-NAF-5.

On May 4, 2007, DEK submitted the KPS plant specific application of the NRC approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," for KPS cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-D/WRB-1 code correlation (reference 5).

Subsequently, the (NRC) staff communicated two questions regarding these submittals.

These questions and DEKs responses are provided in Attachment 1.

Should you have any questions, please contact Mr. Craig D. Sly at 804-273-2784.

Very truly yours, G. T. Bischof Vice President - Nuclear Engineering

Serial No. 06-578D Response to NRC Questions Page 2 of 4

References:

1. Summary of Meeting on January 31, 2006, 'To Discuss the Applicability of Dominion Safety and Core Design Methods to Kewaunee Power Station (TAC No. MC 9566),"

(ADAMS Accession Number ML060400098).

2. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated August 16, 2006 (ADAMS Accession Number ML062370351).

3. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated December 6, 2006 (ADAMS Accession Number ML0063410177).

4. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated April 16, 2007.

5, Letter from G. T. Bischof (DEK) to NRC, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-D/WRB--1 at Kewaunee Power Station," dated May 4, 2007.

Attachment:

1. Response to NRC Request for Additional Information Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)."

Commitments made in this letter: None

Serial No. 06-578D Response to NRC Questions Page 3 of 4 cc" Regional Administrator U. S. Nuclear Regulatory Commission Region III 2443 Warrenville Road Suite 210 Lisle, Illinois 60532-4352 Ms. M. H. Chernoff Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8 G9A Washington, D. C. 20555 Mr. S. C. Burton NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 06-578D Response to NRC Questions Page 4 of 4 COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, 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 that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this /,3 vjhday of .*III.,L._..

.2007.

My Commission Expires: e'2.oZ3 J/?608 Notary P Nblic -

NoayPUblic C

Commonwoolth of Vliginio J

yCommlaslon Exipres Aug 31. 2008 (SEAL

Serial No. 06-578D ATTACHMENT 1 Response to NRC Questions Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)"

KEWAUNEE POWER STATION DOMINION. ENERGY KEWAUNEE, INC.

Serial No. 06-578D Attachment 1 Page 1 of 6 Response to NRC Questions Regarding Kewaunee Reauest for Approval of Topical Report DOM-NAF-5. "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)"

On August 16, 2006, Dominion Energy Kewaunee (DEK) submitted Dominion Topical Report DOM-NAF-5 without attachments A and B (reference 2) to the NRC. On December 6, 2006, Attachment A to DOM-NAF-5, containing Core Management Systems benchmark analysis results, was submitted (reference 3). On April 16, 2007, DEK submitted Attachment B to DOM-NAF-5, containing RETRAN benchmark analysis results (reference 4). This submittal, in conjunction with references 2 and 3, provided the complete contents of DOM-NAF-5.

On May 4, 2007, DEK submitted the KPS plant specific application of the NRC approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," for KPS cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-D/WRB-1 code correlation (reference 5).

Subsequently, the (NRC) staff communicated two questions regarding these submittals.

These questions and DEKs responses are provided below.

NRC Question 1 In the August 16, 2006, submittal, a brief explanation (on page 12 of 34), and a summary (on page 14 of 34) is provided for the relaxed power distribution control methodology. However, no analysis results are provided. Pleaseprovide "before"and "after"results of the bounding analyses conducted with the VEP-FRD-42 methodology, demonstrating continued adherenceto the respective limits.

Dominion Response The question above was clarified during a teleconference with NRC staff members on May 15, 2007. During the teleconference it was determined that before and after analysis results are not currently available and providing them was not practicable.

Dominion agreed to provide a comparative analysis showing the Dominion relaxed power distribution (RPDC) methodology, when applied to Kewaunee Power Station, will produce similar results to those provided by the Westinghouse relaxed axial offset control (RAOC) methodology currently in use (WCAP-10216-P, Revision 1A (reference 7)).

Dominion expects that the cycle-specific RPDC analysis to be performed for KPS will support similar delta-I limits to those that are calculated using the Westinghouse RAOC methodology. Delta-I is defined as the difference in power generated in the top and

Serial No. 06-578D Attachment 1 Page 2 of 6 bottom halves of the core (in percent of rated thermal power). The two methodologies are very similar to each other and the physics codes (ANC (Westinghouse) and CMS (Dominion)) used in the main calculations should generate a similar set of axial power shapes. This expectation is based on Dominion's experience with RPDC implementation for the North Anna units and informal comparisons to generic RAOC analysis results.

To further illustrate the similarity of the two methods, Table 1 presents a side-by-side comparison of key elements for both methodologies. This table provides greater detail about items a) through h) listed on page 14 of 34 of the August 16, 2006 submittal (reference 2). Since Table 1 illustrates that the key elements of the RPDC methodology are directly comparable to the elements of the RAOC methodology, it may be concluded that the RPDC results obtained during the KPS reload analysis will be essentially the same.

Serial No. 06-578D Attachment 1 Page 3 of 6 Table 1 Comparison of Key Elements of Dominion Relaxed RPDC and Westinghouse ROAC Methodology Category Element Westinghouse RAOC Dominion RPDC Comparison Technical Specification /

COLR Limits Operating Axial Flux Difference (AFD) limits Axial Flux Difference (AFD) limits Same Limits versus reactor power versus reactor power FQ Surveillance Non- RAOC applies a cycle specific RPDC applies a cycle specific N(z) Essentially the equilibrium W(z) factor to the measured FQ to factor to the FQ limit to account for conditions account for non-equilibrium non-equilibrium operation. same operation.

Condition I Analysis RAOC methodology populates a The RPDC xenon shape library is xenon shape library using a built by reducing Doppler feedback Both xenon reconstruction model. in the base neutronics model and meth The reconstruction model is anmethodologies dependent on several parameters allowing a to oscillation divergent xenon xenon occur. Actual gnrt xa generate axial Xenon whose ranges are determined by distributions are sampled and xenon Distributions xenon transient analysis. stributisaramplend distributions that Parameters for a given xenon saved from this transient, No cover essentially shape are retained only if Delta-I consideration is given during the the same delta-I control can be maintained for transient calculation to whether or space those shapes within a tentative not the shapes are obtainable limit, during normal operation

Serial No. 06-578D Attachment 1 Page 4 of 6 Category Element Westinghouse RAOC Dominion RPDC Comparison A range of power levels between Minimum of three power levels, 50% and 100% power with small Power 100% 500/ and an intermediate enough increments to ensure an Essentially the Levels power are required adequate number of power same distributions are being analyzed (typically 10% power intervals).

Control Rod Range of control rod positions Range of control rod positions from from ARO to power dependent ARO to power dependent Rod Same Rod Insertion Limits (RILs). Insertion Limits (RILs).

Burnups BOL, MOL, EOL BOL, MOL, EOL Same FQ Analysis Each power shape generated for The FQ x Power for each shape is Condition I is analyzed to compared to the LOCA FQ x Power Loss of determine if LOCA constraints are Coolant met or exceeded. For each x K(z) limit at each power level to Essentially the Accident power level the results of this determine which axCial shapes same (LOCA) analysis will indicate a ofi dltaI tentative approach thea LOCA whih tereareestablishing limit, thereby preliminary allowable range of delta-I in which there are delta-I versus power band.

no violations of the LOCA limits.

Normal operation power The entire set of axial power distributions are evaluated relative distributions from the normal Loss of Aow to the assumed limiting normal operation analysis are evaluated Essentially the Accident operation power distribution, against the 1.55 cosine design axial same (LOFA) typically the 155 cosine, sed in power distribution for the LOFA theiaccident1anysins i analysis with the applicable the accident analysis thermal-hydraulic code(s) and correlation(s).

Serial No. 06-578D Attachment 1 Page 5 of 6 Category Element Westinghouse RAOC Dominion RPDC Comparison Condition II Analysis Analyzed Cooldown Accident, Control Rod Cooldown Accident, Control Rod Accidents Withdrawal, Boration / Dilution Withdrawal, Boration / Dilution Se Initial statepoints for Condition II Initial statepoints for Condition II Shape analysis are limited to the analysis are limited to the Condition Selection Condition I axial power I axial power distributions that fit Same distributions that fit within within tentative delta-I bands.

tentative delta-I bands.

Serial No. 06-578D Attachment 1 Page 6 of 6 NRC Question 2 On page 26 of 34, Section 3.6.1 addresses the conditions and limitations associated with VIPRE-D. In section 3.6.2, parts 1.C and 1.D allude to various models being used without- stating when and how these models are used. Please provide additional information regarding the model selection process, e.g., when one particularmodel is chosen, how that choice is made, and who makes the decision.

Dominion Response The model selection process is controlled by NRC approved Dominion Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code" (reference 6). Dominion will develop and use VIPRE-D models for Kewaunee cores that strictly follow the modeling guidelines and usage requirements specified in DOM-NAF-2-A. DOM-NAF-2-A prescribes the specific constitutive models to be selected for use in the VIPRE-D models for two-phase flow models and correlations, heat transfer correlations, and turbulent mixing models. DOM-NAF-2-A allows no flexibility to select different options for these constitutive models. These VIPRE-D models are used to analyze the non-LOCA, DNB-related transients and accidents as listed in DOM-NAF A.

References:

1. Summary of Meeting on January 31, 2006, 'To Discuss the Applicability of Dominion Safety and Core Design Methods to Kewaunee Power Station," (TAC No. MC 9566),

(ADAMS Accession Number ML060400098).

2. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated August 16, 2006 (ADAMS Accession Number ML062370351).

3. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated December 6, 2006 (ADAMS Accession Number ML0063410177).

4. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," dated April 16, 2007.

5. Letter from G. T. Bischof (DEK) to NRC, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-DNWRB-1 at Kewaunee Power Station," dated May 4, 2007.

6. Dominion Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2006.

7. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," February 1994.

Topical Report DOM-NAF-3, Rev. 0.0-A Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)

Attachment 2 NRC Request for Additional Information on DOM-NAF-3 and Dominion Responses dated July 23, 2007 9 pages after the cover page

Dominion Energy Kewaunee, Inc.

July 23, 2007 U. S. Nuclear Regulatory Commission Serial No. 06-578E Attention: Document Control Desk NL&OS/CDS: RO Washington, DC 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO NRC QUESTIONS REGARDING KEWAUNEE REQUEST FOR APPROVAL OF TOPICAL REPORT DOM-NAF-5. "APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS)"

On January 31, 2006, a public meeting was held with the Nuclear Regulatory Commission (NRC) and Dominion Energy Kewaunee, Inc. (DEK) staff. During this meeting, DEK presented an approach and implementation strategy for application of existing NRC-approved nuclear core design and safety analysis methods to Kewaunee Power Station (KPS) (reference 1). These design and analysis methods are already in use within the remainder of the Dominion fleet. Fundamental to the proposed approach was creation of a composite topical report (DOM-NAF-5) that would document the application of the relevant methodologies to KPS.

On August 16, 2006, DEK submitted Dominion Topical Report DOM-NAF-5 without Attachments A and B (reference 2). On December 6, 2006, Attachment A to DOM-NAF-5 was submitted, which contained Core Management Systems benchmark analysis results (reference 3). On April 16, 2007, DEK submitted Attachment B to DOM-NAF-5, containing RETRAN benchmark analysis results (reference 4). This submittal, in conjunction with References 2 and 3, provided the complete contents of DOM-NAF-5.

On May 4, 2007, DEK submitted the KPS plant specific application of the NRC approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," for KPS cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-D/WRB-1 code correlation (referernce 5).

Subsequently, on June 27, 2007 the (NRC) staff communicated four questions regarding the April 16, 2007 letter (reference 4). On June 28, 2007 a telephone discussion was held between members of the NRC and Dominion staff to discuss each question. The DEK responses to the four questions are provided in Attachment 1.

Serial No. 06-578E Response to NRC Questions Page 2 of 3 Should you have any questions, please contact Mr. Craig D. Sly at 804-273-2784.

Very truly yours, William R. Matthews Senior Vice President - Nuclear Operations COMMONWEALTH OF VIRGINIA

)

COUNTY OF HENRICO The foregoing. document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by William R. Matthews, who is Senior Vice President - Nuclear Operations 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 that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this AM!day o 2007.

My Commission Expires: 6?. 9 ~ 3/.3 AooB Notary Public 4 MARGARET S.SENNEIT INotary Pubft3S36 Commonwealth of vkgtnla my commwulof Ep~es Au~g 31. 2W#

Serial No. 06-578E Response to NRC Questions Page 3 of 3

References:

1. Summary of Meeting on January 31, 2006, 'To Discuss the Applicability of Dominion Safety and Core Design Methods to Kewaunee Power Station (TAC No. MC 9566),"

(ADAMS Accession Number ML060400098).

2. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated August 16, 2006 (ADAMS Accession Number ML062370351).

3. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated December 6, 2006 (ADAMS Accession Number ML0063410177).

4. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated April 16, 2007.

5. Letter from G. T. Bischof (DEK) to NRC, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-D/WRB-1 at Kewaunee Power Station," dated May 4, 2007.

Attachment:

1. Response to NRC Request for Additional Information Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)."

Commitments made in this letter: None cc: Regional Administrator U. S. Nuclear Regulatory Commission Region IlI 2443 Warrenville Road Suite 210 Lisle, Illinois 60532-4352 Ms. M. H. Chernoff Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8 G9A Washington, D. C. 20555 NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 06-0578E ATTACHMENT 1 Response to NRC Questions Regarding Kewaunee Request for Approval of Topical Report DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)"

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 06-0578E Attachment 1 Page 1 of 5 Response to NRC Questions Regarding Kewaunee Request-for Approval of Topical Report DOM-NAF-5. "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)"

On August 16, 2006, Dominion Energy Kewaunee (DEK) submitted Dominion Topical Report DOM-NAF-5 without Attachments A and B (reference 2) to the NRC. On December 6, 2006, Attachment A to DOM-NAF-5 was submitted, which contained Core Management Systems benchmark analysis results (reference 3). On April 16, 2007, DEK submitted Attachment B to DOM-NAF-5, containing RETRAN benchmark analysis results (reference 4). This submittal, in conjunction with references 2 and 3, provided the complete contents of DOM-NAF-5.

On May 4, 2007, DEK submitted the KPS plant specific application of the NRC approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," for KPS cores containing Westinghouse 422V+ fuel assemblies with the VIPRE-D/WRB-1 code correlation (reference 5).

Subsequently, on June 27, 2007, the (NRC) staff communicated four questions regarding the April 16, 2007 letter (reference 4). On June 28, 2007, a telephone discussion was held between members of the NRC and Dominion staff to discuss each question and clarify the scope and detail of the information being requested by the NRC staff. Consistent with the results of these communications, the DEK responses to the four questions are provided below.

Note: For the purposes of consistency and clarity in the following responses, the most recently adopted analytical models will be referred to as "Dominion KPS" models. The presently used and proposed-to-be-changed models will be referred to as "current USAR" models.

NRC Question 1 In Appendix B, of the April 16, 2007, submittal, DEK provided their RETRAN calculations as they are applied to Kewaunee. Table 2-1 on page 5 of 49, provides a comparison of the Dominion-USAR models. It appears to the staff that the USAR model is more detailed than the more recently adopted RETRAN model. Please be prepared to discuss each section of Table 2-1 with the intent of demonstrating the reasons for the preferreduse of a less detailed DOM model over the USAR model.

Response

The differences between the Dominion KPS and current USAR RETRAN models, as summarized in Table 2-1 of DOM-NAF-5 Attachment B, are the result of certain model

Serial No. 06-0578E Attachment 1 Page 2 of 5 noding differences. The base model noding used in the Dominion KPS model was selected to ensure consistency with the Dominion RETRAN Topical Report (VEP-FRD-

41) for non-LOCA licensing analyses. Wherever practical, it is a Dominion objective to maintain consistency within the models used among Dominion plants. As such, the Dominion KPS model noding was chosen to be the same as the Surry and North Anna models with the exception of some minor differences that have been described in DOM-NAF-5 Section 3.4.1.4 and Attachment B. The Dominion KPS model is applied consistent with the conditions and limitations described in the NRC Safety Evaluation Report (SER) for VEP-FRD-41.

Dominion uses special modeling methods to address those transients where the base model noding may not be adequate for detailed phenomena prediction. For example, an overlay deck is used to create a split reactor vessel model to use when analyzing Main Steam Line Break (MSLB) events. This overlay adds volumes to create a second, parallel flow path through the active core from the lower plenum to the upper plenum such that RCS loop temperature asymmetries can be represented. In addition, this overlay maximizes the steam generator (SG) tube heat transfer coefficients to ensure conservative SG heat transfer from the primary to secondary. Again, these special modeling methods are consistent with VEP-FRD-41.

It should be noted that the differences in the transient results for the current USAR and Dominion KPS models are generally attributable to differences in modeling and/or initial condition assumptions rather than noding differences. Responses to Questions 2 through 4 provide some specific examples of such differences.

NRC Question 2 On page 20 of 49, the Locked Rotor analysis (Figure4.2-6) indicates that the core heat flux drops in the firstsecond of the transient. Please explain.

Response

The core heat flux drops in the first second of the transient because the core conductor heat transfer coefficients in the Dominion KPS model change during the transient as a result of the decreasing core flow rate. In comparison, the current USAR model incorporates a different core heat transfer treatment that does not involve calculation of time-dependent values of heat transfer coefficients during the transient.

For the Dominion KPS model, the heat transfer coefficients for the core conductors decrease during the first 0.5 seconds of the Locked Rotor transient. This is a result of the decreasing RCS flow rate. The Dominion KPS model initially predicts a single-phase, subcooled forced convection heat transfer regime in the core regions. As core flow decreases, so does the single-phase forced convection heat transfer coefficient.

Serial No. 06-0578E Attachment 1 Page 3 of 5 This causes a corresponding decrease In heat flux. At approximately 0.5 seconds, the heat transfer mode begins to transition from single-phase forced convection to nucleate boiling, based on local fluid conditions. This causes an increase in the magnitude of the heat transfer coefficient in the affected core volumes. As a result, the core heat flux begins to increase in the time frame between approximately 0.5 to 1.0 second. At approximately one second, the Dominion KPS model closely matches the current USAR heat flux response.

This modeling difference between static (current USAR) and dynamic (Dominion KPS) heat transfer coefficients has little impact on the Locked Rotor analysis acceptance criterion of peak RCS pressure. The Dominion KPS model agrees very well with the current USAR RCS pressure response, as shown in DOM-NAF-5, Attachment B, Figure 4.2-2 (page 18).

NRC Question 3 On page 28 of 49, Figure 4.3-4, LONF pressurizer pressure, requires additional explanation.

Response

Pressurizer pressure response demonstrates wider pressure swings (higher initial peak and more pronounced post-reactor trip pressure decrease) for the Dominion KPS model when compared to the current USAR model response. One reason for this difference is the current USAR model dampens pressurizer pressure response by assuming an artificially high pressurizer spray flow value. By suppressing pressurizer pressure, pressurizer PORV actuation is minimized or precluded, reducing the initial pressure increase and the subsequent pressure decrease.

A second factor driving pressurizer pressure is the pressurizer volume insurge, which in turn, is influenced by reactor vessel temperature increase. This initial temperature increase is more pronounced for the Dominion KPS model case due to an approximately 10 second delay in reactor trip, causing increased energy input into the reactor coolant system. The SG low-low level reactor trip is delayed because the Dominion KPS model Is initialized at a higher initial SG water mass when compared to the current USAR model. The current USAR model has less SG water mass due, in part, to the zero-slip assumption for the current USAR multi-node steam generator (MNSG). It is noted that when the Dominion KPS initial SG water mass is established at the current USAR value, the reactor trip occurs at the same time as the current USAR model.

The benchmark criteria for this event are satisfied and the benchmark results are acceptable. The difference in pressurizer pressure response between the Dominion

Serial No. 06-0578E Attachment 1 Page 4 of 5 KPS model and the current USAR model are understood as being the result of differences in initial conditions and assumed control system response.

NRC Question 4 On page 34 of 49, Figures 4.4-3 and 4.4-4, MSLB reactivity and core power indicate significantdifferences between the USAR and DOM calculationsboth of which are done with RETRAN-02. Please explain the different reactivity curves in light of the fact that both are done with the same kinetics model, i.e., point kinetics.

Response

The difference in core reactivity response between the Dominion KPS and current USAR models is due to the difference in boron injection from the emergency core cooling system (ECCS) safety injection and accumulator injection. ECCS flow rate and timing are different between the two models causing core boron concentration and core reactivity differences during the transient. The boron concentration difference can be seen in DOM-NAF-5, Attachment B, Figure 4.4-5 (page 35).

The significant difference in core reactivity response is observed after approximately 130 seconds into the Main Steam Line Break (MSLB) transient. For the first 130 seconds or so, the core reactivity response predicted by the Dominion KPS model matches the current USAR model fairly well. The primary driver for the observed difference in core reactivity after 130 seconds is the core boron concentration, as shown in Figure 4.4-5. In the Dominion KPS model, boron injection from the refueling water storage tank (RWST) by way of the safety injection (SI) system is delayed until all of the fluid from the SI piping Is purged. The initial fluid in the SI piping volumes is assumed to be at zero parts per million (ppm) boron concentration. For the Dominion KPS model, significant boron concentrations do not reach the core until approximately 130 seconds after the start of the transient. The current USAR model predicts a sharper increase in boron concentration in the 55 to 70 second time frame, as the accumulators discharge borated water into the RCS. The current USAR model predicts higher flow rates from the accumulators than the Dominion KPS model.

The model differences in ECCS flow rate and core boron concentration affect the core power and heat flux response. The Dominion KPS model shows a slightly higher core heat flux during the transient as shown on DOM-NAF-5, Attachment B, Figure 4.4-3 (page 34). Higher core heat fluxes are conservative for the MSLB core response transient, as they lead to lower Departure from Nucleate Boiling Ratios (DNBR).

Serial No. 06-0578E Attachment 1 Page 5 of 5

References:

1. Summary of Meeting on January 31, 2006, "To Discuss the Applicability of Dominion Safety and Core Design Methods to Kewaunee Power Station," (TAC No. MC 9566),

(ADAMS Accession Number ML060400098).

2. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),'" dated August 16, 2006 (ADAMS Accession Number ML062370351).

3. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated December 6, 2006 (ADAMS Accession Number ML0063410177).

4. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated April 16, 2007.

5. Letter from G. T. Bischof (DEK) to NRC, "Implementation of the Dominion Statistical DNBR Methodology with VIPRE-DNVRB-1 at Kewaunee Power Station," dated May 4, 2007.

6. Dominion Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2006.

7. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," February 1994.