Text

DOM-NAF-2, Rev. 0.2-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code.
	ML102390419
Person / Time
Site:	Millstone, Kewaunee, Surry, North Anna
Issue date:	08/31/2010
From:	Brackmann R; Dominion
To:	; Office of Nuclear Reactor Regulation
References
	10-486, FOIA/PA-2011-0115 DOM-NAF-2, Rev 0.2-A
	Download: ML102390419 (255)
	v • d • e

Serial No.10-486 Docket Nos. 50-305/336/423/338/339/280/281 ATTACHMENT 2 DOM-NAF-2-NP, REVISION 0.2-A, REACTOR CORE THERMAL-HYDRAULICS USING THE VIPRE-D COMPUTER CODE DOMINION ENERGY KEWAUNEE, INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS 1 AND 2

'N

0 0

0 S

S 0

.0Dominion Fleet Report

DOM-NAF-2,

Rev. 0.2-NP-A 0* (with Appendixes A, B & C) 0 0

Reactor Core 0 Thermal-Hydraulics

Using the VIPRE-D

0 Computer Code 0

0* NuclearAnalysis and Fuel 0 Nuclear Engineering 0

0 August 2010 0

0

S 0

0 S DOM-NAF-2, Rev. 0.2-A 0 Including Appendixes A, B and C 0

0 0

0 0 Reactor Core Thermal-Hydraulics Using the 0 VIPRE-D Computer Code 0

S NUCLEAR ANALYSIS AND FUEL DEPARTMENT S DOMINION RICHMOND, VIRGINIA 0 August 2010 0

Prepared by:

S Robert S. Brackmann 0

S Reviewed by:

S Brian J. Vitiello S 1~

0 S

S Recommended for Approval:

S 0 D. L. Horn S Supervisor, Nuclear Safety Analysis Models and Methods Appro v 0

0 Dir. .oBNasehore Director, Nuclear Analysis and Fuel S

0 Fleet Report DOM-NAF-2, Rev. 0.2-A includes:

" Safety Evaluation by NRC, dated April 4, 2006 (DOM-NAF-2, including Appendixes A & B)

" Corrections to Safety Evaluation by NRC, dated June 23, 20060 S

" Safety Evaluation by NRC, dated April 22, 2009 (Appendix C)

" Safety Evaluation by NRC, dated June 21, 2010 (Appendix BSupplement)0

" Classification /Disclaimer 0

" Abstract 0

" Acknowledgements

" Fleet Report DOM-NAF 2 0

" Appendix A to Fleet Report DOM-NAF-2 0

" Appendix B to Fleet Report DOM-NAF-2 0

0

" Appendix C to Fleet Report DOM-NAF-2 0

Flet Re. 02-Aincude:0Rpor DO-NA-2 ATTACHMENTS (page counts do not include cover pages):

0 1: Request For Additional

aft Information, atdEalaio pil4Set 1, Questions and y206(OMNF-,

RC Answers nluin (31 pages).

ppnixs 0 2: Information Regarding *vauaio aCrrcton LYNXT b NCdaedJue Error Supporting t Sfey the Request for Approval of Fleet0 3,200 Report DOM-NAF-2*(5 pages).0 dte aftyEvlutin Ari 2, 00 (ppndx y RC)0 3: Supplemental Information for Appendix B - Removal of Mixing Vane Grid Spacing0 Restriction (43 pages).0

0 0

UNITED STATES 0 NUCLEAR REGULATORY COMMISSION 0 WASHINGTON, D.C. 20555-0001 0 April 4, 2006 0 SERIAL # C -3/15 Mr. David A. Christian Senior Vice President MC'D APR 11 2006 and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center NUCLEAR LICENSING 0 5000 Dominion Boulevard Glen Allen, VA 23060-6711 0

SUBJECT:

MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 (MILLSTONE 2 AND 3),

S NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 (NORTH ANNA 1 AND S 2), AND SURRY POWER STATION, UNIT NOS. 1 AND 2 (SURRY 1 AND 2) -

APPROVAL OF DOMINION'S FLEET REPORT DOM-NAF-2, "REACTOR 0 CORE THERMAL-HYDRAULICS USING THE VIPRE-D COMPUTER CODE" (TAC NOS. MC4571, MC4572, MC4573, MC4574, MC4575, AND MC4576) 0 0

Dear Mr. Christian:

0 By letter dated September 30, 2004, as supplemented by letters dated January 13, June 30, and September 8, 2005, Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power S Company (the licensees), requested approval for the generic application of Fleet Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code." The NRC staff has defined the term 'fleet report' as a report that can be used by the licensees' 0 nuclear facilities.

In their submittal, the licensees stated that they are using the COBRA IlIc/MIT computer code 0 to perform thermal hydraulic analyses. However, due to the need for enhanced core thermal-hydraulic capabilities, the licensees requested to use VIPRE-D to analyze multiple fuel types.

0 The licensees developed VIPRE-D to fit the needs of the licensees nuclear plants and fuel products.

S Although, the September 30, 2004, submittal identified the docket number for each of the licensees' plants, the Nuclear Regulatory Commission (NRC) staff was requested to approve of this fleet report on a generic basis. The licensees stated that plant-specific applications to implement this fleet report, including applicable appendixes, would be submitted to the NRC staff for review and approval under separate correspondence.

The enclosed Safety Evaluation (SE) documents the basis for the NRC staff's conclusion's that Fleet Report DOM-NAF-2, was found to be acceptable for the licensees' nuclear facilities. The SE defines the basis for acceptance of the report.

In accordance with the guidance provided on the NRC website, the NRC requests that the licensees publish an accepted version of this fleet report within 3 months of receipt of this letter.

The accepted version shall incorporate this letter and the enclosed SE between the title page and the abstract. It must be well indexed such that information is readily located. Also, it must

0 0

D. Christian contain, in appendices, historical review information, such as questions and accepted responses, and original report pages that were replaced. The accepted version shall include an

"-A" (designated accepted) following the report identification symbol.

0 If the NRC's criteria or regulations change such that its conclusions as to the acceptability of the fleet report are invalidated, then the licensees will be expected to revise and resubmit its respective documentation, or submit justification for the continued applicability of the topical report without revision of the respective documentation.

Sincerely, 0

Christopher I. Grimes, Director Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket Nos. 50-336, 50-423, 50-338, 50-339, 50-280, and 50-281

Enclosure:

Safety Evaluation cc w/encl: See next page 0

0 0

S S

S 0 Virginia Electric and Power Company cc:

S Ms. Lillian M. Cuoco, Esq. Office of the Attorney General Senior Counsel Commonwealth of Virginia 0 Dominion Resources Services, Inc. 900 East Main Street Building 475, 5th Floor Richmond, Virginia 23219 Rope Ferry Road Waterford, Connecticut 06385 Mr. Chris L. Funderburk, Director 0 Nuclear Licensing & Operations Support Mr. Donald E. Jernigan Innsbrook Technical Center Site Vice President Dominion Resources Services, Inc.

0 Surry Power Station 5000 Dominion Blvd.

Virginia Electric and Power Company Glen Allen, Virginia 23060-6711 0 5570 Hog Island Road S Surry, Virginia 23883-0315 Mr. Jack M. Davis Site Vice President S Senior Resident Inspector North Anna Power Station S Surry Power Station Virginia Electric and Power Company U. S. Nuclear Regulatory Commission Post Office Box 402 0 5850 Hog Island Road Mineral, Virginia 23117-0402 Surry, Virginia 23883 0 Mr. C. Lee Lintecum Chairman County Administrator 0 Board of Supervisors of Surry County Louisa County S Surry County Courthouse Post Office Box 160 0 Surry, Virginia 23683 Louisa, Virginia 23093 S Dr. W. T. Lough Old Dominion Electric Cooperative Virginia State Corporation Commission 4201 Dominion Blvd 0 Division of Energy Regulation Glen Allen, Virginia 23060 0 Post Office Box 1197 Richmond, Virginia 23218 Senior Resident Inspector S North Anna Power Station Dr. Robert B. Stroube, MD, MPH U.S. Nuclear Regulatory Commission 0 State Health Commissioner 1024 Haley Drive Office of the Commissioner Mineral, Virginia 23"117 Virginia Department of Health 0 Post Office Box 2448 Richmond, Virginia 23218

SREGj UNITED STATES NUCLEAR REGULATORY COMMISSION1 I.- WASHINGTON, D.C. 20555-0001

?S XS SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO FLEET REPORT DOM-NAF-2 MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-336, 50-423, 50-338, 50-339, 50-280, AND 50-281

1.0 INTRODUCTION

By letter dated September 30, 2004 (Reference 1), as supplemented by letters dated January 13 (Reference 2), June 30 (Reference 13), and September 8, 2005 (Reference 14),

Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power Company (the licensees),

submitted a request for Nuclear Regulatory Commission (NRC) staff approval for the application of Fleet Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D 0 Computer Code," Appendix A, "Qualification of the Framatome Advanced Nuclear Power (F-ANP) BWU Critical Heat Flux (CHF) Correlations," and Appendix B "Qualification of the Westinghouse WRB-1 CHF Correlations in the Dominion VIPRE-D Computer Code." Appendix A includes the VIPRE-D code and correlation departure from nucleate boiling ratio (DNBR) design limits, and Appendix B provides an evaluation of DNBR for the Westinghouse WRB-1 CHF correlations that are applicable to the Westinghouse 15xl 5 optimi;zed fuel assembly (OFA) fuel bundle.

0

2.0 REGULATORY EVALUATION

5 Title 10 of the Code of FederalRegulations, Part 50, Section 50.90, requires licensees to submit an application to the NRC whenever they desire to amend the license.

The VIPRE-01 computer code is a core thermal hydraulics computer program developed by the Electric Power Research Institute (EPRI) and approved generically by the NRC staff for the purpose of evaluating departure from nucleate boiling (DNB) for pressurized water reactor (PWR) systems. Since this generic approval did not include specific applications of VIPRE-01 to any particular fuel design, NRC staff review and approval is necessary in order to apply this methodology to a specific fuel design. Therefore, this review addresses the specific application of VIPRE-01 by the licensees to the Framatome and Westinghouse fuel types in the licensees' nuclear steam supply systems (NSSS).

S VIPRE-D is the licensees' version of VIPRE-01, which has been enhanced by the addition of several vendor-specific CHF correlations. The licensees intend to utilize the VIPRE-D computer code to assess the DNBR for the Framatome BWU-N, BWU-Z, and BMU-ZM CHF fuel correlations. Additionally, the licensees intend to apply the VIPRE-D code to assess the S

S S

S

0 O

Westinghouse WBR-1 CHF correlation for the 15x15 OFA fuel design. The licensees have

previously used the COBRA IIIc/MIT computer code (Reference 3) to perform the thermal

hydraulic analyses and is submitting this fleet report to replace COBRA IlIc/MIT computer code O with the VIPRE-D computer program along with the new CHF correlations for the various Framatome and Westinghouse fuel designs. The NRC staff's technical evaluation of the O VIPRE-D code and the new CHF fuel correlations is given below.

3.0 TECHNICAL EVALUATION

In order to evaluate DNB in the licensees' NSSS for the Framatome anid Westinghouse fuel

types, the NRC staff reviewed the application of the VIPRE-D code along with the various pertinent code correlations and models, fuel-specific CHF correlations, and DNBR design limits.

The VIPRE-D code is a modified version of the VIPRE-01 code which is a finite volume subchannel thermal hydraulics code with the specific capability to model a three-dimensional core and other component geometries. With the appropriate boundary conditions from a systems code such as RETRAN, VIPRE-01 computes the flow, void, pressure, and temperature

distribution of the fluid through the core to ultimately compute the minimum DNB for steady state and transient conditions. The VIPRE-01 code also contains a fuel rod model that O computes the radial and axial temperature distribution that is coupled to the cladding surface heat transfer coefficient correlations and CHF correlations that are particular to a given fuel rod and bundle design with the objective of determinating DNB following a non-loss-of-coolant

accident (LOCA) transient event.

In order to compute the single and two-phase flow conditions that develop during transients undergoing a potential DNB, various two-phase flow models for handling subcooled and bulk

boiling are available for use in the code, as well as convective heat transfer correlations for O single and two-phase flow conditions. Correlations are also included in the code to deal with turbulent mixing, axial and cross-flow resistance, and form loss coefficients. As such, the NRC staff's review consisted of reviewing the CHF correlations and the various fluid flow and heat transfer options in the code to assure the correlations and models were validated over the range of conditions for those transients for which DNB is to be evaluated.

It is also noted that the licensees did not modify any of the phenomenological models or correlations in VIPRE-01. The licensees only added the new CHF correlations (Reference 1, Appendix A and Reference 2, Appendix B) to accommodate the DNBR assessments of the O Framatome and Westinghouse fuel types. No other changes were made to VIPRE-01 in O constructing the new VIPRE-D code.

O 3.1 Code Usage O

The licensees indicated it plans to use the VIPRE-D code for the following applications.

O (1) Perform an analysis of 14x14, 15x15, and 17x17 fuel in PWR reactors.

(2) Perform an analysis of DNBR for statistical and deterministic transients in the Updated Final O Safety Analysis Report (UFSAR), as identified in Table 1, below. Additicnal DNBR transients O that are plant specific may be analyzed in a plant-specific application that would be submitted to the NRC staff for review and approval.

0 O

0

(3) Perform steady state and transient DNB evaluations.

(4) Develop reactor core safety limits or core thermal limit lines (CTL).

(5) Provide the basis for reactor protection setpoints.

(6) Establish or verify 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 addendum or appendix to the original VIPRE-D document.

3.2 Code Applications The licensees intend to implement Fleet Report DOM-NAF-2 (VIPRE-D) in its plant-specific S applications through the following methods.

(1) Changes to the technical specifications (TSs) to add Fleet Report DOM-NAF-2 and 0

Appendices A and B to the plant Core Operating Limit Report for that particular plant.

0 (2) Changes to the Statistical Design Limit(s) for the relevant code and .correlation(s).

0 (3) Any TS changes related to over temperature delta T (OTAT), over power delta T (OPAT),

enthalpy rise factor (FAH) or other reactor protection function, as well as, revised reactor core 0 safety limits. 0 (4) Changes to the list of UFSAR transients for which the code and correlations apply, as 0 shown in Table 1. 0 Table 1: UFSAR Transients Analyzed with VIPRE-D 0 1 Accidental depressurization of the main steam system 0 2 Accidental depressurization of the reactor cooling system 3 Excessive heat removal due to feedwater system malfunction S 4 Excessive load increase 51 Inadvertent operation of emeraencv core coolinca system durina o)ower operation 0 Locked reactor coolant pump rotor or shaft break Loss of external electrical load and/or turbine trip S Loss of forced reactor coolant flow 9 Loss\of normal feedwater 10 Major rupture of a main feedwater pipe S 11 Rod cluster control assembly misalignment/dropped rod/bank S 12 Rod cluster control assembly bank withdrawal at power 1 Rod cluster control assembly bank withdrawal from subcritical 0

14 Rupture of a main steam pipe 0 1 Single rod cluster control assembly withdrawal at full power S 1 Startup of an inactive reactor coolant loop 1 Uncontrolled boron dilution S

3.3 Compliance with the VIPRE-01 Safety Evaluation Report (SER)

In order to meet the NRC staff's requirements listed in the VIPRE-01 SER (References 4 and 5), the licensees will apply the VIPRE-D code for PWR licensing applications under the

following conditions:

(1) The application of VIPRE-D is limited to PWR licensing calculations with heat transfer regimes up to CHF. VIPRE-D cannot be used for post-CHF calculations or for

o boiling-water-reactor calculations.

(2) VIPRE-D analyses will use only those DNB correlations reviewed and approved by the NRC

staff in this SER. These correlations include the Framatome BWU-N, BWU-Z, and BMU-ZM

CHF and the Westinghouse WRB-1 fuel CHF correlations.

(3) The Framatome BWU CHF correlations, which have been specifically developed for use with the Framatome Advanced Mark-BW fuel, were used in the 12-channel model. There are

three BWU CHF correlations that constitute the licensing basis for the Framatome Advanced Mark-BW fuel assembly. These correlations use the same basic equation, but are fit to

different databases (References 6 and 7). VIPRE-D applies different BWU correlations at

different axial levels, according to the following guidelines:

- BWU-N, which is only applicable in the presence of non-mixing vane grids (MVG), is

I used from the beginning of the heated length to the leading edge of the first structural MVG (Reference 6).

- BWU-Z, which is the enhanced mixing vane correlation, is used from the leading edge

of the first structural MVG to the leading edge of the second structural MVG (Reference S 6).

- BWU-ZM, which is just BWU-Z with a multiplicative enhancement factor and is applicable in the presence of mid-span mixing grids (MSMGs), is used from the leading

edge of the second structural MVG to the leading edge of the last structural MVG

(Reference 7).

- For the uppermost span, in which the end of heated length occurs less than one grid

span beyond the last MVG, the BWU-Z correlation is used with a grid spacing equal to

the effective grid spacing (the distance from the last grid to the eld of the heated length)

(Reference 6).

(4) As required by the NRC staff in Reference 4, the following model options were reviewed and justified by the licensees for use in the DNB evaluation of the Framatome fuels.

I - Radial Nodalization: The licensees utilize 1/8th core symmetry and the model is applicable to the 14xl 4, 15xl 5, and 17xl 7 fuel arrays. These guidelines are consistent with the previously approved COBRA models (Reference 3). Benchmark calculations

with the Framatome LYNXT code (References 8 and 9) verified this modeling approach.

- Axial Nodalization: Node size is limited to a maximum of 6 inches.

- Fuel Rod Model: The licensees will use the dummy fuel rod model which requires the surface heat flux as input, computed by the RETRAN code. RETRAN accounts for the fuel conduction, gap conductance, and associated delayed energy transport effects. 0 This approach is consistent with previously approved licensees' methodologies (Reference 10). Also, the analysis assumes that 97.4 percent of the reactor power is generated in the fuel while 2.6 percent is generated in the coolant, consistent with the previously approved COBRA modeling techniques.

- Power Distribution: A chopped cosine axial power shape is typically used. The power distribution is modeled to limit the cross flow and mixing in the hct channel since the peak F AH is also applied to the thimble and hot cell. This results in a conservative calculation of DNBR. Also since the data is limited with respect to top peaked axial profiles, the licensees utilize the Tong F-factor to correct for non-uniform axial power shapes, which has been previously approved by the NRC staff. The licensees also performed benchmark comparisons between VIPRE-D/BWU and LYNXT/BWU and VIPRE-D/WRB-1 with COBRA/WRB-1 using symmetric and non-symmetric axial power shapes that show no dependency on the shape of the power distribution.

0

- Turbulent Mixing: The turbulent mixing factor is 0.0 as opposed to the VIPRE Manual recommended value of 0.8. This produces a conservative calculation since momentum mixing is precluded with this assumption. The turbulent mixing for single-phase fluid in single channels is set to 0.038 (range 0.0 to 0.1). This is the default model approved in the original generic VIPRE SER. For flow paths connected to lumped channels, turbulent mixing is set to zero for conservatism.

- Axial Hydraulic Losses and Cross-Flow Resistance: For axial cross flow, the McAdams correlation is used to approximate the Colebrook smooth pipe formulation for single-phase axial friction. Lateral resistance is computed by the Idle Chik empirical correlation (Reference 10) for bundle circular tubes in a vertical column.

- Form Loss Coefficients: These are obtained from the vendor for the particular fuel bundle designs. VIPRE-D properly places the losses at the top of the cell, or at the boundaries between the cells where the grids are located. Varying the location of the grid resistance upward or downward showed an insignificant change in DNBR (much less than the 5 percent uncertainty associated with thermal-hydraulic codes in this 0 application).

- Two-Phase Flow and Heat Transfer Correlations: The licensees will use the following O models to compute CHF for the specific fuel types: EPRI Subcooled Void Model, EPRI Bulk Boiling Void Model, and the EPRI Two-Phase Friction Multiplier. No hot wall friction correlation is used. Results of the comparisons of VIPRE-D with LYNXT justify this choice of correlations and models since this combination produced the lowest standard deviation in DNBR with a value of 0.89 percent. The slip model is not to be employed and cannot be used. The Dittus-Boelter single-phase heat-transfer correlation is also used. O

-Engineering Factors: The licensees include the following factors which adversely affect DNBR: Local Heat Flux Hot Channel Factor, Engineering Enthalpy-Rise Hot Channel 0 Factor, Stack Height Reduction, and Inlet Flow Reduction. These factors are fuel 0

0 0

0

product dependent.

- CHF Correlations: See the "Correlations and DNBR Limits" Section 3.5 below.

- For transient analysis, appropriate time steps are selected to ensure numerical stability and accuracy. The Courant number, which is based on flow velocity, time step and axial node size, is set to be greater than one in VIPRE-D transient calculations whenever a subcooled void model is used.

3.4 Benchmarks VIPRE-D benchmark calculations were performed with the Framatome LYNXT code and the 12-channel model created by Framatome to model the North Anna Power Station, Unit Nos. 1 and 2, cores containing Framatome Advanced Mark-BW fuel assemblies. This benchmark uses 173 state points obtained from the UFSAR Chapter 15 events including the reactor core safety limits, axial offset envelopes (AO's), rod withdrawal at power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss of flow accident (LOFA), and locked rotor accident (LOCROT) events to compare the performance of VIPRE-D and LYNXT. These various limits and events 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 (reactor core safety limits); (b) AOs; (c) elevated hot rod power (misaligned rod); and (d) LOFA and LOCROT. The 173 state points cover the full range of conditions and axial offsets in the North Anna UFSAR Chapter 15 evaluations except for main steamline break (MSLB), which is discussed in Section 5.2 of Reference 1. These results were specifically selected to challenge the three BWU CHF correlations.

This benchmark study showed an average deviation between VIPRE-D and LYNXT of less than 0.14 percent in DNBR, with a maximum deviation of 2.2 percent. These results are well within the uncertainty typically associated with thermal-hydraulic codes, which has been quantified to 0 be 5 percent (References 9 and 12), and these results justify the model selections in Section 4 0 of Reference 1. The close comparison of VIPRE-D to LYNXT over the full range of conditions expected for UFSAR transients justifies the applications of VIPRE-D to the transients identified 0 in Table 1, above. The range of conditions for the benchmarks is given below in Table 2.

S Table 2: Range of VIPRE-D / LYNXT 173 Benchmark State points S

S VARIABLE RANGE 0 Pressure [psial 1860 to 2400 Power [percent of 2942.2 MWt] 66 to 135 S

Inlet Temperature [OF] 506.6 to 626.2 S Flow [percent of Minimum Measured Flow] 64 to 100 S FAH 1.49 to 1.945 S Axial Offset [percent] -48.7 to 57.9 S

S The 12-channel model discussed in Section 5.1 of Reference 1 does not allow the modeling of the peaking and inlet boundary conditions in the fuel assemblies adjacent to the hot assembly, S which is necessary for the analysis of some accidents, such as MSLB. Consequently, a S

S S

0 14-channel model was created to more accurately simulate the behavior of the core during an 0 MSLB event.

0 The VIPRE-D 14-channel model for a North Anna core containing Framatome Advanced Mark-BW fuel assemblies consists of 14 channels (10 subchannels and 4 lumped channels) and 16 rods. The two additional channels provide adequate detail of the flow field in the vicinity of the hot assembly and allow for the modeling of the peaking and inlet boundary conditions in the fuel assemblies adjacent to the hot assembly.

In order to verify the accuracy, the licensees compared the results from the VIPRE-D 14-channel model to the results from the Framatome LYNXT model for high flow (with offsite power) and low flow (without offsite power) MSLB evaluations. The results obtained showed a maximum deviation of 2.12 percent in DNBR. These results demonstrated that VIPRE-D provides results similar to those of other approved codes accepted for analysis of an MSLB event, provided the model has sufficient detail surrounding the hot assembly, such as the 14-channel model described in Reference 1.

0 In addition, the results of the 14-channel model comparison with the DNBR results of the 173 state points obtained with the VIPRE-D 12-channel model showed that there was essentially no difference between the 12-channel and the 14-channel models (the average deviation in DNBR was 0.03 percent), which indicates that VIPRE-D models were created following the methodology discussed in Section 4 of Reference 1 and are acceptable.

3.5 Correlations and DNBR Limits The BWU-Z, BWU-ZM and BWU-N correlations have been qualified with the licensees' VIPRE-D computer code. Table 3 summarizes the DNBR design limits for VIPRE-D/BWU-Z, VIPRE-D/BWU-ZM and VIPRE-D/BWU-N that yield a 95 percent non-DNB probability at a 95 percent confidence level. Table 3 summarizes the applicability and the ranges of validity for all three CHF correlations.

0 Table 3: VIPRE-D DNBR Limits for BWU-Z, BWU-ZM and BWU-N VIPRE-D/BWU-Z DNBR limit below 700 psia 1.59 DNBR limit 700 - 2,400 psia 1.20 VIPRE-D/BWU-ZM DNBR limit below 594 psia 1.59 DNBR limit above 594 psia 1.18 VIPRE-D/BWU-N DNBR limit below 1200 psia 1.39 DNBR limit above 1200 psia 1.22 0

0 0

0

S S

0 These correlations are to be used over the following thermal hydraulic conditions:

0 Table 4: Range of validity for BWU-Z, BWU-ZM and BWU-N 0 BWU-Z BWU-ZM BWU-N S

S Pressure [psial 400 to 2,465 400 to 2,465 788 to 2,616 S Mass Velocity 2

0.36 to 3.55 0.47 to 3.55 0.25 to 3.83

[iMlbm/hr-ft ]

Thermodynamic Less than 0.74 Less than 0.68 Less than 0.70 0 Quality at CHF S Applicability Mixing Vane Grids Mid-Span Mixing Grids Non-Mixing Vane Grids 0

The WRB-1 correlation is applicable to the Westinghouse 15x15 OFA fuel assemblies at Surry 0 Power Station, Unit Nos. 1 and 2. The DNBR limit was found to be 1.17 and was the same as the limits computed using the previously approved methodologies of the licensees (COBRA of Reference 11) and Westinghouse (THINC and VIPRE-01). The range of applicability of the 0 WRB-1 correlation is summarized below in Table 5.

0 Table 5: Range of VIPRE-D / WRB-1 Benchmark State points VARIABLE RANGE Pressure [psia] 1440 to 2490 S Mass Velocity [Mlbm/hr-ft 2] 0.9 to 3.7 0 Thermodynamic Quality at CHF *0.30 Local Heat Flux [Mbtu/hr-ft 2] _*1.00 0

Mixing Vane Grid [in] > 13.0 By letter dated January 13, 2005, the licensees imposed the following additional restrictions on 0

the use of the VIPRE-D/WRB-1 correlation.

S (1) VIPRE-D/WRB-1 will not be used when the local heat flux exceeds 1.0 Mbtu/hr-ft 2, and 0

(2) VIPRE-D/WRB-1 will not be used for fuel with less than a 13-inch mixing vane grid spacing.

0 The licensees imposed these restrictions as a result of the constraints the NRC staff placed on the use of Reference 11, in its letter dated July 25, 1989.

The previously approved W-3 correlation will be used when conditions fall outside the range of the WRB-1 correlation. Specifically, the W-3 correlation will be applied to the lower portion of the fuel assemblies in the RWSC event because of the bottom peaked axial power profile assumed and the MSLB event because of the low pressures encountered. The W-3 will use a limit of 1.3 for the rod withdrawl event. For the MSLB, the limit of 1.45 will be used for pressures 500 to 100 psia and the limit of 1.3 will be used for pressures above 1000 psia.

0 Benchmarking of the VIPRE-D code with the results of the COBRA code for the events listed in Table 1 above (except the MSLB event) showed an average deviation of less than 0.6 percent in DNBR with a maximum deviation of 3.75 percent. This is within the uncertainty for thermal hydraulic codes used to perform analyses of this nature. For the MSLB, the comparison with COBRA using the W-3 correlation, showed the maximum deviation was 1.5 percent.

The licensees utilized a One-Sided Tolerance theory for the VIPRE-D fuel correlation DNBR design limits given above. This theory allows the licensees to calculate a DNBR limit such that values equal to the design limit avoids DNB with a 95-percent probability at a 95-percent confidence level. All of the statistical techniques utilized in the design limit determinations assumed that the original data distribution is normal. As such the licensees verified that the overall measured-to-predicted CHF ratios were also normally distributed evaluated through the use of a "D" normality test.

0 Following the review of References 1 and 2, Requests for Additional Information (RAIs) were sent to the licensees requesting supplemental information regarding the review of the VIRPE-D code model options and usage, the statistical evaluation of the DNBR design limits specific to each fuel type, and the benchmarking evaluations. The RAI responses are documented in Reference 13 and the staff found these responses to be acceptable.

Lastly, an error was uncovered by Framatome in their LYNXT computer code, the results of which, were used by the licensees to qualify portions of the licensees' VIPRE-D code. The licensees' assessment of the impact of the error, reported to the NRC staff in Reference 14, shows that the error does not affect the LYNXT/BWU code or correlation limits. Furthermore, the maximum change in any numerical value reported in Reference 1, Section 5, regarding benchmark DNBR calculations between LYNXT and VIPRE-D, was found to be 0.02 percent.

Appendix B of Reference 2 is not affected by this error. The NRC staff agrees that the impact of the error has a negligible effect on the calculated differences between the VIPRE-D and LYNXT DNBR benchmarking calculations.

4.0 CONCLUSION

The NRC staff finds the proposed use of the VIPRE-D code to evaluate DNBR for selected PWR transients is acceptable. Furthermore, the NRC staff finds the modifications to VIPRE-D to evaluate the Framatome BWU fuel using the BWU-Z, BWU-ZM, and BWU-N CHF correlations as well as the Westinghouse 15xl 5 OFA fuel using the WRB-1 correlation to also be acceptable. The VIPRE-D fuel design limits are also found to be acceptable by the NRC staff for the Framatome and Westinghouse fuel types listed herein. The use of the licensees' VIPRE-D code is limited to only these CHF correlations. The VIPRE-D code can be used subject to the models and options specified in DOM-NAF-2, Rev. 0, Sections 4.0 through and including, Section 4.12 (Reference 1). Evaluation of the Framatome fuel using the BWU-Z, BWU-ZM, and BWU-N CHF correlations is subject to the DNBR limits and ranges given in Section A.5 of DOM-NAF-2, Rev. 0 (Reference 1). Use of the VIPRE-D code is also approved for evaluating the Westinghouse 15x15 OFA fuel using the WRB-1 CHF correlation subject to the DNBR limits and evaluation ranges given in Tables B.8-1 and B.8-2 of DOM-NAF-2, Rev.

0.0 Appendix B (Reference 2). The WRB-1 correlation is limited by the following restrictions:

(1) VIPRE-D/WRB-1 will not be used when the local heat flux exceeds 1.0 MBTU/hr-ft2 , and (2) VIPRE-D/WRB-1 will not be used for fuel with less than a 13-inch mixing vane grid spacing, as discussed in Reference 2 Section B.3. The W-3 correlation will also be used when the S

S S

S

conditions fall outside the range of the WRB-1 correlation as discussed in Section B.3, last paragraph of Reference 2. The VIPRE-D code is further restricted for application to those transients listed in Table 2.1-1 of DOM-NAF-2, Rev. 0.0 (Reference 1) and the uses and

applications listed in Section 2.1 entitled "VIPRE-D Application."

5.0 REFERENCES

1. Letter from Leslie N. Hartz (Dominion Nuclear Connecticut, Inc.) to the USNRC, "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station

Units 2 and 3, Request for Approval of Topical Report DOM-NAF-2 Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code including Appendix A-Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D

Computer Code," dated September 30, 2004.

2. Letter from Eugene S. Grecheck (Dominion Nuclear Connecticut, Inc.) to the USNRC, "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station

Units 2 and 3, Request for Approval of Appendix B of Topical Report DOM-NAF-2

Qualification of the Westinghouse WRB-1 CHF Correlations in the Dominion VIPRE-D Computer Code," dated January 13, 2005.

3. Topical Report, VEP-FRD-33-A, "VEPCO Reactor Core Thermal-Hydraulic Analysis

Using the COBRA IIIc/MIT Computer Code," F. W. Sliz and K. L. Basehore, October

1983.

0

4. 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, dated May 1, 1986.

5. Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 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)," dated October 30, 1993.

6. Topical Report, BAW-10199P-A, Addendum 1, "The BWU Critical Heat Flux Correlations," Framatome Cogema Fuels, August 1996, dated December 6, 2000.

0

7. Topical Report, BAW-10199P-A, Addendum 2, "Application of the BWU-Z CHF Correlation to the Mark-BW17 Fuel Design with Mid-Span Mixing Grids," Framatome

Cogema Fuels, dated September 5, 2002 0

0 S

0 0

0 S

0 0

0

8. Letter from L. N. Hartz (Dominion) to USNRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Changes and Exemption Request, Use of Framatome ANP Advanced Mark-BW Fuel," dated March 28, 2002.

0

9. Technical Report, BAW-1 0156-A, Revision 1, "LYNXT, Core Transient Thermal-Hydraulic Program," Framatome ANP, August 1993.

10. Technical Report, AEC-TR-6630, "Handbook of Hydraulic Resistance, Coefficients of Local Resistance and of Friction," I. E. Idel'Chik, 1960.

0

11. Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," R. C. Anderson and N. P. Wolhope, July 1990.

12. Technical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," R. C.

Anderson, June 1987.

13. Letter from Eugene S. Grecheck (Dominion Nuclear Connecticut, Inc.) to the USNRC, "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, and Millstone Power Station Units 2 and 3 Request for Additional Information on Topical Report DOM-NAF-2:

Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code Including Appendices A and B," dated June 30, 2005.

0

14. Letter from Leslie N. Hartz (Dominion Nuclear Connecticut, Inc.) to the USNRC, "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Information Regarding a LYNXT Error Supporting the Request for 0 Approval of Topical Report DOM-NAF-2, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code Including Appendix A - Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," dated September 8, 2005.

0 Principal Contributor: L. Ward Date: April 4, 2006 0

0 0

0

S S UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 0 June 23, 2006 0 SERIAL #

Mr. David A. Christian Senior Vice President RECO JUL 5 2006 S and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center NUCLEAR LICENSING 0 5000 Dominion Boulevard Glen Allen, VA 23060-6711 0

0

SUBJECT:

MILLSTONE POWER STATION, UNIT NOS. 2 AND 3, NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, AND SURRY POWER STATION, UNIT NOS. 1 0 AND 2 - CORRECTION TO DOMINION'S FLEET REPORT DOM-NAF-2, "REACTOR CORE THERMAL-HYDRAULICS USING THE VIPRE-D 0 COMPUTER CODE" 0

Dear Mr. Christian:

0 On April 4, 2006, the Nuclear Regulatory Commission (NRC) issued Fleet Report DOM-NAF-2, S "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code." Subsequently, in its 0 letter dated June 1, 2006, you informed the NRC that the safety evaluation (SE) for Fleet S Report DOM-NAF-2 contained several editorial errors, including language that could unnecessarily restrict the use of the fleet report to specific fuel vendors. The corrected pages 0 for this SE are enclosed with this letter. The revisions to the SE are identified by lines in the 0 margin.

0 Sincerely, 0

.i/l, .l . .. , ,. .

StepVn R. Monar'que, Projdt Manager S Plant Licensing Branch I1-1 0 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 0

Docket Nos. 50-336, 50-423, 50-338 50-339, 50-280, and 50-281 S

Enclosure:

0 Safety Evaluation cc w/encl: See next page

0 S

UNITED STATES S

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 S

S S

S SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION S

RELATING TO FLEET REPORT DOM-NAF-2 MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 0

S NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 S SURRY POWER STATION. UNIT NOS. 1 AND 2 S S

DOCKET NOS. 50-336, 50-423, 50-338, 50-339, 50-280, AND 50-281 S

S

1.0 INTRODUCTION

S By letter dated September 30, 2004 (Reference 1), as supplemented by letters dated S January 13 (Reference 2), June 30 (Reference 13), and September 8, 2005 (Reference 14), S Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power Company (the licensees),

submitted a request for Nuclear Regulatory Commission (NRC) staff approval for the S

application of Fleet Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D S Computer Code," Appendix A, "Qualification of the Framatome Advanced Nuclear Power (F-ANP) BWU Critical Heat Flux (CHF) Correlations," and Appendix B "Qualification of the 0

Westinghouse WRB-1 CHF Correlations in the Dominion VIPRE-D Computer Code." Appendix S A includes the VIPRE-D code and correlation departure from nucleate boiling ratio (DNBR) S design limits, and Appendix B provides an evaluation of DNBR for the Westinghouse WRB-1 CHF correlations that are applicable to the Westinghouse 15xl 5 optimized fuel assembly (OFA) S fuel bundle. S

2.0 REGULATORY EVALUATION

S S

Title 10 of the Code of FederalRegulations, Part 50, Section 50.90, requires licensees to S submit an application to the NRC whenever they desire to amend the license.

0 The VIPRE-01 computer code is a core thermal hydraulics computer program developed by the S Electric Power Research Institute (EPRI) and approved generically by the NRC staff for the purpose of evaluating departure from nucleate boiling (DNB) for pressurized water reactor S

(PWR) systems. Since this generic approval did not include specific applications of VIPRE-01 S to any particular fuel design, NRC staff review and approval is necessary in order to apply this S methodology to a specific fuel design. Therefore, this review addresses the specific application of VIPRE-01 by the licensees to the NRC staff-approved pressurized water reactors (PWR) fuel S types in the licensees' nuclear steam supply systems (NSSS). S S

S Corrected by letter dated June 23, 2006 S S

S S

S

VIPRE-D is the licensees' version of VIPRE-01, which has been enhanced by the addition of several vendor-specific CHF correlations. The licensees intend to utilize the VIPRE-D computer

code to assess the DNBR for the Framatome BWU-N, BWU-Z, and BMU-ZM CHF fuel

correlations. Additionally, the licensees intend to apply the VIPRE-D code to assess the

Westinghouse WBR-1 CHF correlation for the 15x15 OFA fuel design. The licensees have

lpreviously used the COBRA IlIc/MIT computer code (Reference 3) to perform the thermal hydraulic analyses and is submitting this fleet report to replace COBRA IIIc/MIT computer code

with the VIPRE-D computer program along with the new CHF correlations for the NRC

staff-approved PWR fuel designs. The NRC staffs technical evaluation of the VIPRE-D code and the new CHF fuel correlations is given below.

3.0 TECHNICAL EVALUATION

In order to evaluate DNB in the licensees' NSSS for the NRC staff-approved PWR fuel types,

the NRC staff reviewed the application of the VIPRE-D code along with the various pertinent

code correlations and models, fuel-specific CHF correlations, and DNBR design limits.

The VIPRE-D code is a modified version of the VIPRE-01 code which is a finite volume

subchannel thermal hydraulics code with the specific capability to model a three-dimensional core and other component geometries. With the appropriate boundary conditions from a systems code such as RETRAN, VIPRE-01 computes the flow, void, pressure, and temperature

distribution of the fluid through the core to ultimately compute the minimum DNB for steady

state and transient conditions. The VIPRE-01 code also contains a fuel rod model that computes the radial and axial temperature distribution that is coupled to the cladding surface heat transfer coefficient correlations and CHF correlations that are particular to a given fuel rod

and bundle design with the objective of determinating DNB following a non-loss-of-coolant

accident (LOCA) transient event.

0 In order to compute the single and two-phase flow conditions that develop during transients

undergoing a potential DNB, various two-phase flow models for handling subcooled and bulk boiling are available for use in the code, as well as convective heat transfer correlations for single and two-phase flow conditions. Correlations are also included in the code to deal with turbulent mixing, axial and cross-flow resistance, and form loss coefficients. As such, the NRC staff's review consisted of reviewing the CHF correlations and the various fluid flow and heat transfer options in the code to assure the correlations and models were validated over the range of conditions for those transients for which DNB is to be evaluated.

It is also noted that the licensees did not modify any of the phenomenological models or 0 correlations in VIPRE-01. The licensees only added the new CHF correlations (Reference 1,

Appendix A and Reference 2, Appendix B) to accommodate the DNBR assessments of the NRC staff-approved PWR fuel types. No other changes were made to VIPRE-01 in constructing the new VIPRE-D code.

3.1 Code Usage

The licensees indicated it plans to use the VIPRE-D code for the following applications.

C

Corrected by letter dated June 23, 2006 S

S S

0 0

(1) Perform an analysis of 14x14, 15x15, and 17x17 fuel in PWR reactors.

(2) Perform an analysis of DNBR for statistical and deterministic transients in the Updated Final Safety Analysis Report (UFSAR), as identified in Table 1, below. Additional DNBR transients that are plant specific may be analyzed in a plant-specific application that would be submitted to the NRC staff for review and approval.

3.3 Compliance with the VIPRE-01 Safety Evaluation Report (SER)

In order to meet the NRC staff s requirements listed in the VIPRE-01 SER (References 4 and 5), the licensees will apply the VIPRE-D code for PWR licensing applications under the following conditions:

S (1) The application of VIPRE-D is limited to PWR licensing calculations with heat transfer regimes up to CHF. VIPRE-D cannot be used for post-CHF calculations or for boiling-water-reactor calculations. S S

(2) VIPRE-D analyses will use only those DNB correlations reviewed and approved by the NRC staff in this SER. These correlations include the Framatome BWU-N, BWU-Z, and BMU-ZM CHF and the Westinghouse WRB-1 fuel CHF correlations.

S (3) The Framatome BWU CHF correlations, which have been specifically developed for use with the Framatome Advanced Mark-BW fuel, were used in the 12-channel model. There are S three BWU CHF correlations that constitute the licensing basis for the Framatome Advanced Mark-BW fuel assembly. These correlations use the same basic equation, but are fit to different databases (References 6 and 7). VIPRE-D applies different BWU correlations at 0 different axial levels, according to the following guidelines: S

- BWU-N, which is only applicable in the presence of non-mixing vane grids (MVG), is used from the beginning of the heated length to the leading edge of the first structural MVG (Reference 6).

- BWU-Z, which is the enhanced mixing vane correlation, is used from the leading edge 0 of the first structural MVG to the leading edge of the second structural MVG (Reference 6).

- BWU-ZM, which is just BWU-Z with a multiplicative enhancement factor and is applicable in the presence of mid-span mixing grids (MSMGs), is used from the leading 0 edge of the second structural MVG to the leading edge of the last structural MVG (Reference 7).

S

- For the uppermost span, in which the end of heated length occurs less than one grid span beyond the last MVG, the BWU-Z correlation is used with a grid spacing equal to the effective grid spacing (the distance from the last grid to the end of the heated length) S (Reference 6).

C S

S Corrected by letter dated June 23, 2006 S

S S

S

(4) As required by the NRC staff in Reference 4, the following model options were reviewed and justified by the licensees for use in the DNB evaluation of the NRC staff-approved PWR fuels.

S- Radial Nodalization: The licensees utilize 1/8th core symmetry and the model is applicable to the 14x14, 15x15, and 17x17 fuel arrays. These guidelines are consistent

with the previously approved COBRA models (Reference 3). Benchmark calculations

with the Framatome LYNXT code (References 8 and 9) verified this modeling approach.

Axial Nodalization: Node size is limited to a maximum of 6 inches.

0*-

- Fuel Rod Model: The licensees will use the dummy fuel rod model which requires the surface heat flux as input, computed by the RETRAN code. RETRAN accounts for the

fuel conduction, gap conductance, and associated delayed energy transport effects.

This approach is consistent with previously approved licensees' methodologies (Reference 3). Also, the analysis assumes that 97.4 percent of the reactor power is

generated in the fuel while 2.6 percent is generated in the coolant, consistent with the

previously approved COBRA modeling techniques.

- Power Distribution: A chopped cosine axial power shape is typically used. The power

distribution is modeled to limit the cross flow and mixing in the hot channel since the peak F AH is also applied to the thimble and hot cell. This results in a conservative calculation of DNBR. Also since the data is limited with respect to top peaked axial 0profiles, the licensees utilize the Tong F-factor to correct for non-uniform axial power

shapes, which has been previously approved by the NRC staff. The licensees also

performed benchmark comparisons between VIPRE-D/BWU and LYNXT/BWU and VIPRE-D/WRB-1 with COBRA/WRB-1 using symmetric and non-symmetric axial power

shapes that show no dependency on the shape of the power distribution.

0

- Turbulent Mixing: The turbulent mixing factor is 0.0 as opposed to the VIPRE Manual recommended value of 0.8. This produces a conservative calculation since momentum mixing is precluded with this assumption. The turbulent mixing for single-phase fluid in

single channels is set to 0.038 (range 0.0 to 0.1). This is the default model approved in the original generic VIPRE SER. For flow paths connected to lumped channels, 0turbulent mixing is set to zero for conservatism.

0

- Axial Hydraulic Losses and Cross-Flow Resistance: For axial cross flow, the McAdams correlation is used to approximate the Colebrook smooth pipe formulation for

single-phase axial friction. Lateral resistance is computed by the Idle Chik empirical correlation (Reference 10) for bundle circular tubes in a vertical column.

0- Form Loss Coefficients: These are obtained from the vendor for the particular fuel bundle designs. VIPRE-D properly places the losses at the top of the cell, or at the boundaries between the cells where the grids are located. Varying the location of the

grid resistance upward or downward showed an insignificant change in DNBR (much

less than the 5 percent uncertainty associated with thermal-hydraulic codes in this application).

C

Corrected by letter dated June 23, 2006 0

0 0

0 S

0

- Two-Phase Flow and Heat Transfer Correlations: The licensees will use the following models to compute CHF for the specific fuel types: EPRI Subcooled Void Model, EPRI 0

Bulk Boiling Void Model, and the EPRI Two-Phase Friction Multiplier. No hot wall friction 0 correlation is used. Results of the comparisons of VIPRE-D with LYNXT justify this S choice of correlations and models since this combination produced the lowest standard deviation in DNBR with a value of 0.89 percent. The slip model is not to be employed S and cannot be used. The Dittus-Boelter single-phase heat-transfer correlation is also 0 used. S

-Engineering Factors: The licensees include the following factors which adversely affect S DNBR: Local Heat Flux Hot Channel Factor, Engineering Enthalpy-Rise Hot Channel Factor, Stack Height Reduction, and Inlet Flow Reduction. These factors are fuel S

S These correlations are to be used over the following thermal hydraulic conditions: S Table 4: Range of validity for BWU-Z, BWU-ZM and BWU-N 0

S BWU-Z BWU-ZM BWU-N S S

Pressure [psial 400 to 2,465 400 to 2,465 788 to 2,616 Mass Velocity 0.36 to 3.55 0.47 to 3.55 0.25 to 3.83 S

[Mlbm/hr-ft ]

2 S Thermodynamic Less than 0.74 Less than 0.68 Less than 0.70 S Quality at CHF 0

Applicability Mixing Vane Grids Mid-Span Mixing Grids Non-Mixing Vane Grids S S

The WRB-1 correlation is applicable to the Westinghouse 15x15 OFA fuel assemblies at Surry S Power Station, Unit Nos. 1 and 2. The DNBR limit was found to be 1.17 and was the same as S the limits computed using the previously approved methodologies of the licensees (COBRA of Reference 11) and Westinghouse (THINC and VIPRE-01). The range of applicability of the S

WRB-1 correlation is summarized below in Table 5. S Table 5: Range of VIPRE-D / WRB-1 Benchmark State points S

S VARIABLE RANGE S Pressure [psia] 1440 to 2490 S Mass Velocity [Mlbm/hr-ft 2] 0.9 to 3.7 Thermodynamic Quality at CHF S

_.0.30 Local Heat Flux [Mbtu/hr-ft 2]

S

!51.00 Mixing Vane Grid [in] > 13.0 S

S By letter dated January 13, 2005, the licensees imposed the following additional restrictions on S the use of the VIPRE-D/WRB-1 correlation. S S

Corrected by letter dated June 23, 2006 S S

S S

(1) VIPRE-D/WRB-1 will not be used when the local heat flux exceeds 1.0 Mbtu/hr-ft 2, and (2) VIPRE-D/WRB-1 will not be used for fuel with less than a 13-inch mixing vane grid spacing.

9

The licensees imposed these restrictions as a result of the constraints the NRC staff placed on the use of Reference 11, in its letter dated July 25, 1989.

The previously approved W-3 correlation will be used when conditions fall outside the range of the WRB-1 correlation. Specifically, the W-3 correlation will be applied to the lower portion of

the fuel assemblies in the RWSC event because of the bottom peaked axial power profile assumed and the MSLB event because of the low pressures encountered. The W-3 will use a

limit of 1.3 for the rod withdrawl event. For the MSLB, the limit of 1.45 will be used for pressures 500 to 1000 psia and the limit of 1.3 will be used for pressures above 1000 psia.

Benchmarking of the VIPRE-D code with the results of the COBRA code for the events listed in Table 1 above (except the MSLB event) showed an average deviation of less than 0.6 percent

in DNBR with a maximum deviation of 3.75 percent. This is within the uncertainty for thermal

hydraulic codes used to perform analyses of this nature. For the MSLB, the comparison with COBRA using the W-3 correlation, showed the maximum deviation was 1.5 percent.

D The licensees utilized a One-Sided Tolerance theory for the VIPRE-D fuel correlation DNBR

design limits given above. This theory allows the licensees to calculate a DNBR limit such that values equal to the design limit avoids DNB with a 95-percent probability at a 95-percent

confidence level. All of the statistical techniques utilized in the design limit determinations

assumed that the original data distribution is normal. As such the licensees verified that the overall measured-to-predicted CHF ratios were also normally distributed evaluated through the use of a "D" normality test.

9

Following the review of References 1 and 2, Requests for Additional Information (RAIs) were sent to the licensees requesting supplemental information regarding the review of the VIRPE-D

code model options and usage, the statistical evaluation of the DNBR design limits specific to

each fuel type, and the benchmarking evaluations. The RAI responses are documented in Reference 13 and the staff found these responses to be acceptable.

9Lastly, an error was uncovered by Framatome in their LYNXT computer code, the results of

which, were used by the licensees to qualify portions of the licensees' VIPRE-D code. The licensees' assessment of the impact of the error, reported to the NRC staff in Reference 14, 9 shows that the error does not affect the LYNXT/BWU code or correlation limits. Furthermore, the maximum change in any numerical value reported in Reference 1, Section 5, regarding benchmark DNBR calculations between LYNXT and VIPRE-D, was found to be 0.02 percent.

Appendix B of Reference 2 is not affected by this error. The NRC staff agrees that the impact

t of the error has a negligible effect on the calculated differences between the VIPRE-D and

LYNXT DNBR benchmarking calculations.

4.0 CONCLUSION

The NRC staff finds the proposed use of the VIPRE-D code to evaluate DNBR for selected

PWR transients is acceptable. Furthermore, the NRC staff finds the modifications to VIPRE-D 0

Corrected by letter dated June 23, 2006 9

0 0

S to evaluate the Framatome BWU fuel using the BWU-Z, BWU-ZM, and BWU-N CHF correlations as well as the Westinghouse 15x15 OFA fuel using the WRB-1 correlation to also 0

be acceptable. The VIPRE-D fuel design limits are also found to be acceptable by the NRC staff for NRC staff-approved PWR fuel types. The use of the licensees' VIPRE-D code is 0 limited to only these CHF correlations. The VIPRE-D code can be used subject to the models 0 and options specified in DOM-NAF-2, Rev. 0, Sections 4.0 through and including, Section 4.12 (Reference 1). Evaluation of the Framatome fuel using the BWU-Z, BWU-ZM, and BWU-N 0

CHF correlations is subject to the DNBR limits and ranges given in Section A.5 of DOM-NAF-2, 0 Rev. 0 (Reference 1). Use of the VIPRE-D code is also approved for evaluating the 0 Westinghouse 15x1 5 OFA fuel using the WRB-1 CHF correlation subject to the DNBR limits and evaluation ranges given in Tables B.8-1 and B.8-2 of DOM-NAF-2, Rev. 0.0 Appendix B S (Reference 2). The WRB-1 correlation is limited by the following restrictions: (1) VIPRE- 0 D/WRB-1 will not be used when the local heat flux exceeds 1.0 MBTU/hr-ft 2, and (2) VIPRE-D/WRB-1 will not be used for fuel with less than a 13-inch mixing vane grid spacing, 0

as discussed in Reference 2 Section B.3. The W-3 correlation will also be used when the 0 0

0 0

0 Corrected by letter dated June 23, 2006 0 0

0

S 0 UNITED STATES SERIAL # W *O 0 0 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 0 April 22, 2009 REC'D APR 28 2009 0 C,*

0 NUCLEAR LICENSING Mr. David A. Christian 0 President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center S 5000 Dominion Boulevard Glen Allen, VA 23060-6711 0

S

SUBJECT:

KEWAUNEE POWER STATION, MILLSTONE POWER STATION, UNITS 2 AND 3, NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, AND SURRY S POWER STATION, UNIT NOS. 1 AND 2 - APPENDIX C TO DOMINION FLEET S REPORT DOM-NAF-2, "QUALIFICATION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION IN THE DOMINION VIPRE-D COMPUTER CODE" (TAC S NOS. MD8703, MD8704, MD8705, MD8706, MD8707, MD8708, MD8709)

Dear Mr. Christian:

0 By letter dated April 4, 2008, Dominion Energy Kewaunee, Inc., Dominion Nuclear Connecticut, 0 Inc., and Virginia Electric and Power Company (Dominion), submitted an application to use Appendix C to Dominion Fleet Report DOM-NAF-2, "Qualification of the Westinghouse WRB-2M 0 CHF [Critical Heat Flux] Correlation in the Dominion VIPRE-D Computer Code" to the U.S.

S Nuclear Regulatory Commission (NRC) for Kewaunee Power Station, Millstone Power Station, Units 2, and 3, North Anna Power Station, Unit Nos. 1 and 2, and Surry Power Station, Unit Nos. 1 and 2, respectively. The purpose of this report was to justify the use of the previously 0 approved WRB-2M CHF Correlation in the previously approved Dominion VIPRE-D Computer Code. The proposed change would allow Dominion to use the WRB-2M CHF Correlation in S VIPRE-D when performing thermal-hydraulic analysis on 17x17 Robust Fuel Assembly fuel.

0 On the basis of its review, the NRC staff finds the licensee's request acceptable. The enclosed 0 safety evaluation documents the findings. Please contact me at (301) 415-1864, if you have any questions on this matter.

0 Sincerely, 0

0 0

Donna N. Wright, Pr Manager 0 Plant Licensing Branch I1-1 6 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation S

Docket Nos. 50-305, 50-336, 50-423, 0

50-338, 50-339, 50-280, and 50-281

Enclosure:

Safety Evaluation cc: Distribution via Listserv

S

,R REG",

UNITED STATES 0 *NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 0

0 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 0 RELATED TO APPENDIX C TO DOMINION FLEET REPORT DOM-NAF-2, 0

"QUALIFICATION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION IN THE DOMINION VIPPRE-D COMPUTER CODE" DOMINION ENERGY KEWAUNEE, INC., DOMINION NUCLEAR CONNECTICUT, INC.,

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION, MILLSTONE POWER STATION, UNITS 2 AND 3, NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 SURRY POWER STATION, UNIT NOS. 1 AND 2 0 DOCKET NOS. 50-305, 50-336/423, 50-338/339, AND 50-280/281

1.0 INTRODUCTION

By letter dated April 4, 2008 (Agencywide Documents Access and Management System 0 (ADAMS) Accession No. ML080980229), Dominion Energy Kewaunee, Inc., Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power Company (Dominion), submitted Appendix C to Dominion Fleet Report DOM-NAF-2, Rev 0.0, and "Qualification of the Westinghouse WRB-2M CHF [Critical Heat Flux] Correlation in the Dominion VIPRE-D Computer Code" (Reference

1) to the Nuclear Regulatory Commission (NRC) for Kewaunee Power Station, Millstone Power Station, Units 2 and 3, North Anna Power Station, Unit Nos. 1 and 2, and Surry Power Station, Unit Nos. 1 and 2, respectively. The purpose of this report was to justify the use of the previously approved WRB-2M CHF Correlation (Reference 2) in the previously approved Dominion VIPRE-D Computer Code (Reference 3). The proposed change would allow Dominion to use the WRB-2M CHF Correlation in VIPRE-D when performing thermal-hydraulic analysis on 17xl 7 Robust Fuel Assembly (RFA) fuel.

2.0 REGULATORY EVALUATION

0 0

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.34, "Contents of construction permit and operating license applications; technical information," requires that Safety Analysis Reports be submitted that analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the 5 consequences of accidents. As part of the core reload design process, licensees are responsible for reload safety evaluations to ensure that their safety analyses remain bounding 0

Enclosure 0

S 0

0

S 0

for the design cycle. To confirm that the analyses remain bounding, licensees confirm those key inputs to the safety analyses (such as the CHF) are conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a re-analysis or a re-0evaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

The NRC staffs review was based on the evaluation of the technical merit of the submittal and

compliance with any applicable regulations associated with the review of topical reports.

3.0 TECHNICAL EVALUATION

0 3.1 BackQround Information

Boiling crisis occurs when the boiling water flowing past a fuel rod transitions from nucleate

boiling to film boiling. This transition decreases the heat transfer rate at the fuel rod surface, forcing the fuel rod surface temperature to dramatically increase in order to maintain the same total heat transfer. This large increase in fuel rod surface temperature may lead to fuel damage.

The heat flux which causes this transition from nucleate boiling to film boiling is known as CHF.

To prevent possible fuel damage, boiling crisis is prevented via correlations used to predict the CHF. For normal reactor operations, thermal-hydraulic analysis is used to demonstrate that the

I peak heat flux in the core will remain below the CHF.

In pressurized-water reactors (PWR), CHF is primarily a local phenomenon caused by bubbles

which crowd the surface of the fuel rod. If the bubbles prevent the cooling water from reaching

the surface of the fuel rod, the flow can transition from nucleate boiling to film boiling. This form of CHF happens very quickly and is known as departure from nucleate boiling (DNB). Keeping

with common practice, DNB and CHF are used interchangeably. Many parameters can impact

DNB such as: flow pattern, bubble size and population, bubble layer thickness, wall superheat

and flow memory, flow instability, local pressure, local enthalpy, mass velocity, inlet conditions, heated length, rod bundle shape, grid spacers, and others (Reference 4). Due to the complex Snature of the phenomenon of DNB, the functional form of DNB correlations are generally empirical and are often based solely on experimental observations of the relationship between the measured DNB and the measured DNB parameters.

To prevent DNB, the departure from nucleate boiling ratio (DNBR) is used. DNBR is the ratio of the CHF at a location along the fuel rod divided by the current heat flux at that same location 5under the same flow conditions. To ensure an accurate prediction of DNBR, CHF experiments

are performed in which the heat flux in prototypical fuel assemblies increases to the point that CHF is reached. The flow conditions of the experiment are measured and those same flow conditions are input into a specific thermal-hydraulic computer code with a specific CHF correlation. The measured CHF value from the test is compared with the predicated CHF value from the thermal-hydraulic computer code and an analysis is performed to determine if the computer code with the specific CHF correlation can accurately predict the CHF behavior of the

fuel assembly.

0 0

0

The WRB-2M CHF correlation along with Dominion VIPRE-D thermal-hydraulic computer code were used to predict the CHF behavior of Modified 17x17 Vantage 5H fuel with or without modified intermediate flow mixer (MIFM) grids (with MIFM grids the fuel is referred to as 17x17 RFA fuel).

3.2 Critical Heat Flux Test Progjram 0

Data for development of WRB-2M was obtained at the Columbia University Heat Transfer Research Facility. This facility consisted of an instrumented high pressure loop that could supply water at pressures up to 2500 psia, flow rates up to 650 (gallons per minute), and inlet temperatures up to 650 *F. The power supply was capable of producing 12.5 megawatts of direct current.

Four types of test sections were analyzed to evaluate the combinations of different grid spacing and the effects of a central control rod guide thimble. The test sections were designed to have a chopped cosine axial power distribution and a non-uniform radial power distribution so that the highest power rods and peak heat flux locations were in the middle of the bundle and were prototypical of modified 17x17 Vantage 5H and 5H/IFM fuel.

CHF tests were performed by maintaining a constant test section outlet pressure, inlet temperature, and mass flux. Total power to the test section was then increased in small increments until a sudden temperature increase occurred in one or more of the thermocouples positioned on the heater rods. This temperature excursion indicated that DNB had occurred.

When the temperature excursion occurred, power to the test section was reduced and preparation for the next test was begun.

3.3 Use of WRB-2M CHF Correlation with VIPRE-D 0

The WRB-2M CHF correlation is based on local conditions within the fuel bundles. The WRB-2M CHF correlation has been previously reviewed and approved by the NRC staff 0 (Reference 2) and no further review is intended in this evaluation.

0 To evaluate local conditions within the Modified Vantage 5H test bundles, Dominion used the VIPRE-D code. VIPRE-D is a modified version of the Electric Power Research Institute's VIPRE-01 computer code. The VIPRE-D computer code has already been previously and approved by the NRC staff (Reference 3) and no further review is intended in this evaluation.

The NRC staff focused their review efforts on verifying that the WRB-2M correlation when used in the VIPRE-D computer code provided a conservative predicted CHF value with no bias or trends in the prediction of CHF.

The NRC staff reviewed the intended range of the correlation (Table 1) and finds that the behavior of the correlation is consistent over that range. The NRC staff reviewed the trend analysis and finds that there are no trends in the WRB-2M CHF correlation prediction of CHF as a function of any of the thermal-hydraulic variables (pressure, mass flow, and quality), as 0 consistent with the WRB-2M approved topical report.

0 0

S 0

0 O 0 3.4 Statistical Evaluation of DNBR Part of the correlation procedure involves the reduction of the calculated CHF to account for

non-uniform axial flux shapes. This is accomplished by the use of the Tong F-factor

(Reference 4). This factor takes into account that the CHF is affected by a bubble that later separates the main stream in a fluid channel from the superheated liquid near the heated

surface. The bubble later is affected by axial distribution of the upstream heating so that for the O same total power input, a peaked heat flux location will have a lower CHF than if the axial heating rate had been uniform. Tong derived the F-factor from theoretical considerations with empirical constants that were determined from test data. Use of the F-factor permits O development of CHF correlations that are independent of the axial flux shape. The F-factor must also be applied when correlations are used to predict CHF for nuclear reactor safety analysis.

O WRB-2M CHF correlation was developed from test data taken from two types of grid structures, with and without the MIFM grids. Of the 241 CHF tests performed, 143 tests contained MIFMs.

In the other 98 tests, the MIFMs were omitted. Similar to the original WRB-2M Topical Report, O Dominion performed statistical analyses which determined both data sets were random samples from the same population and therefore could be correlated together. Both data sets were first determined to be random samples of normal distributions in accordance with Regulatory

Guide 5.22, "Assessment of the Assumption of Normality (Employing Individual Observed O5 Values)," (Reference 5). The statistical variances of the two populations of data taken from similar test sections were then compared using the F-distribution test. The F-test demonstrated 0 that both data sets were from the same total population and could therefore be correlated

together. The tests with and without MIFM grids are of the same population since the mixing vanes in the structural support grids were modified to be of similar shape and size to those in the MIFMs. The only effective difference between the test assemblies is the grid spacing and

the presence of thimble tubes. Both of which are accounted for in the data correlation.

During plant operation, the ratio of heat fluxes between the CHF and the actual heat flux, which 0 is the DNBR, provides a method for describing the safety margin to fuel damage. One component in this margin is the minimum DNBR limit for acceptance of reactor core thermal/hydraulic calculations. A separate DNBR limit is calculated for each CHF correlation O based on the scatter in predicted test results. The NRC staff has accepted DNBR limits that O ensure a 95 percent probability that CHF will not occur with a confidence of 95 percent for the hottest pins of the reactor core (Reference 6). A DNBR limit of 1.14 was derived for the VIPRE-D computer code using the WRB-2M CHF correlation to meet this criterion.

0 O The DNBR limit which meets the 95/95 acceptance criterion was determined using Owen's one-sided tolerance limit method (Reference 7). The general equation for Owen's method is as O follows:

O O 95/95 DNBR limit = 1

- - K 95 /95 07 SP O

O O

0 O

O

0 0

Where, M is the test population mean of measured to predicted CHF ratios.

P 0 0

oa is the effective standard deviation of all the M/P data.

K 95/95 is a tolerance multiplier which provides the 95/95 probability/confidence limit. The constant K 95 / 95 is a function of the effective degrees of freedom in the test series.

0 Consistent with previous CHF correlations, the standard deviation, o-, shall be calculated from combining the variance within the test series and the variance among the test series. The effective degrees of freedom shall also be calculated in a similar manner. 0 Considering the evaluation above, the NRC staff finds that with a DNBR limit of 1.14, the VIPRE-D computer code using the WRB-2M CHF correlation will conservatively predict the CHF behavior of the fuel designs described herein.

3.5 Conditions and Limitations Based on the forgoing considerations, the NRC staff concludes that the use of the of VIPRE-D computer code with the WRB-2M CHF correlation with a DNBR limit of 1.14 is acceptable for plant safety analyses provided that the following conditions are met:

1. Because WRB-2M CHF correlation was developed from test assemblies designed to simulate Modified 17x1 7 Vantage 5H fuel with or without modified intermediate flow mixer grids, the correlation may only be used to perform evaluations for fuel of that type without further justification.

2. The WRB-2M CHF correlation shall not be applied outside its range of applicability defined by the original WRB-2M topical report and repeated in Table 1 of this evaluation.

3. The WRB-2M CHF correlation shall be used with a DNBR limit of 1.14 with the Dominion VIPRE-D computer code. WRB-2M is dependent on calculated local fluid properties that shall be only be calculated by a computer code approved by the NRC staff for that purpose, such as the VIPRE-D computer code.

4. The WRB-2M CHF correlation can be used for PWR plant analyses of steady state and reactor transients other than loss of coolant accidents. The WRB-2M CHF correlation shall not be used for loss of coolant accident analysis before additional justification is provided to the NRC staff which demonstrates that the applicable regulations are met and the computer code used to calculate local fuel element thermal/hydraulic properties has been approved for that purpose.

0 The NRC staff will require licensees referencing this topical report in licensing applications to document how these conditions are met.

0 0

S 0

S4,0 CONCLUSION 0 When implemented as stated, the NRC staff has reasonable assurance that the use of the

WRB-2M CHF correlation with Dominion's VIPRE-D computer code, as documented in

Reference 1, is acceptable in calculating the CHF for the specified fuel types. The NRC staff has reviewed the qualification of the WRB-2M CHF correlation in the VIPRE-D computer code,

and finds the method applicable only when implemented in accordance with the conditions and

limitations described in Section 3.5 of this safety evaluation. The NRC staff does not intend to review the associated topical report when referenced in license applications.

If the NRC's criteria or regulations change so that its conclusions about the acceptability of the thermal-hydraulic methods or statistical analyses are invalidated, the licensee(s) referencing the report (Reference 1) will be expected to revise and resubmit its respective documentation, or

submit justification for the continued effective applicability of the methodologies without revision of the respective documentation.

Table 1: WRB-2M Applicability Range 0v Parameter WNG-1 Applicability Range 0

Pressure (psia) 1405 to 2425 0

0 Local mass velocity (Mlbm/hr-ft 2 ) 0.97 to 3.1 Local quality (fraction) -0.1 to 0.29 Heat length (ft) <514 0

Grid spacing (inches) 10 to 20.6 S

Equivalent hydraulic diameter (inches) 0.37 to 0.46 Equivalent heated diameter (inches) 0.46 to 0.54 0

5.0 REFERENCES

S 0 1. Letter from Gerald T. Bischoff to U.S. NRC, Serial Number 08-0174 "Request for Approval of Appendix C of Fleet Report DOM-NAF-2 Qualification of the Westinghouse WRB-2M CHF 0 Correlation in the Dominion VIPRE-D Computer Code" dated April 4, 2008 (ADAMS 0 Accession No. ML080980229 (Publically Available)).

0 2. WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids", April 1999 (ADAMS S Accession No. ML081610106 (Non-Publically Available)).

S 3. Letter from Gerald T. Bischoff to U.S. NRC, Serial Number 06-773 "Approved Topical Report 0 DOM-NAF-2, Rev 0.0-A Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code Including Appendixes A and B" dated September 14, 2006 (ADAMS Accession S No. ML062650184 (Publically Available)).

Hsee footnotein Appendix C, page C-22, regarding typo in Table 1

4. L.S. Tong, "Boiling Crisis and Critical Heat Flux," Temporary Instruction Inspection Documentation-25687, 1972.

5. U.S. Atomic Energy Commission Regulatory Guide 5.22, "Assessment of the Assumption of Normality (Employing Individual Observed Values)," April 1974 (ADAMS Accession No. ML003739999 (Publically Available)).

6. U.S. Nuclear Regulatory Commission Standard Review Plan, Section 4.4, "Thermal and Hydraulic Design, Rev 2" NUREG-800, March 2007.

7. D.B. Owen, "Factors for One-Sided Tolerance Limits and for Variables Sampling Plans",

SC-R-607, Sandia Report, March 1963.

Principal Contributors: A. Attard, NRR J. Kaizer, NRR Date of Issuance: April 22, 2009 0

0 S

S 0

0 0

S 0

0 S

S S

O . %.tLRNIC4 UNITED STATES

-o0 S NUCLEAR REGULATORY COMMISSION WASHINGTON, M.C. 20555-0001 SRAL 0.

U"- JUN 2 3 ?0111

O June 21, 2010 NUCLEAR LICENSING O Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear O Innsbrook Technical Center O 5000 Dominion Boulevard Glen Allen, VA 23060-6711 O

SUBJECT:

KEWAUNEE POWER STATION, MILLSTONE POWER STATION UNITS 2 AND 3, NORTH ANNA POWER STATION UNITS 1 AND 2, AND SURRY O POWER STATION UNITS 1 AND 2 - REMOVAL OF MIXING VANE GRID O SPACING RESTRICTION IN APPENDIX B TO FLEET REPORT DOM-NAF-2-A

(TAC NOS. ME2321 - ME2327)

Dear Mr. Heacock:

O By letter dated August 28, 2009, Mr. J. Alan Price of Dominion Resource Services, requested O Nuclear Regulatory Commission (NRC) approval of a change to the NRC-approved Topical

Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code." The change involves removing the mixing vane grid spacing restriction in Appendix B to

DOM-NAF-2-A. Mr. Price supplemented the August 28, 2009, letter with letters dated O November 20, 2009, and April 16, 2010.

The NRC staff has reviewed the proposed change conveyed by the cited submittals and found O the proposed change acceptable. Details of the NRC staffs review are set forth in the enclosed Supplemental Safety Evaluation for DOM-NAF-2-A.

O Sincerely,

.7 O*S. Tam, Senior Project Manager O Plant Licensing Branch Il1-1 Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation O

Docket Nos. 50-305, 50-336, 50-423, O 50-338, 50-339, 50-280, and 50-281 O

Enclosure:

As stated O cc w/encl: Distribution via ListServ O

O O

9p 0

O

0 0

40PAREG UNITED STATES 4,,1o,, NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205550001 0

00 e- SUPPLEMENTAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 0 DOMINION RESOURCE SERVICES, INC., TOPICAL REPORT DOM-NAF-2-A KEWAUNEE POWER STATION 0

MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS 1 AND 2 0

DOCKET NOS. 50-305. 50-336, 50-423, 50-338, 50-3.39, 50-280, AND 50-281

1.0 INTRODUCTION

0 By letter dated August 28, 2009 (Reference 1), as supplemented by letters dated November 20, 2019i and April 16, 2010 (References 8 and 9), Dominion Resources Services, Inc. requested the Nuclear Regulatory Commission (NRC) staff to approve a change to the NRC-approved Fleet Topical Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code." This topical report, and its Appendices A, B, and C were approved by NRC0 via Safety Evaluations dated April 4, 2006 (ADAMS Accession No. ML060790496), June 23, 0 2006 (ADAMS Accession No. ML061740212), and April 22, 2009.(ADAMS Accession No. ML091030639). The approved version of Fleet Report DOM-NAF-2-A, Rev. 0.0-A (Reference 3), including Appendices A and B, was submitted by Dominion by letter dated 0 September 13, 2006 (ADAMS Accession No. ML062650184).

Dominion requested approval to remove the mixing vane grid spacing restriction in Appendix B,

'Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," of Fleet Topical Report DOM-NAF-2-A.

Dominion provided the above-cited submittals to justify the removal of the restriction from Appendix B of DOM-NAF-2-A. Appendix B to DOM-NAF-2-A currently includes a restriction associated with mixing vane grid spacing. Specifically, the subject restriction states that "VIPRE-D/WRB-1 will not be used for fuel with less than, 13" mixing vane grid spacing." The purpose of this restriction was to exclude the use of the VIPRE-D/WRB-1 code/correlation with Westinghouse Intermediate Flow Mixing (IFM) vane grids fuel types. The restriction was originally placed on WRB-1 in the safety evaluation for the COBRAIWRB1 Topical Report VEP-NE-3-A, ."Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code,"

July 1,990 (Reference 5). When the VIPRE-D fleet report was developed, the restriction from the COBRA safety evaluation was included in the WRB-1 Qualification in Appendix B of DOM-NAF-2-A. However, Westinghouse has developed several fuel products that incorporate IFMs (e.g., 15x1 5 VANTAGE+ and 15x1 5 Upgrade fuel products). Consequently, for Dominion 0 Enclosure 0

0 0

O~

S

*to be able to use VIPRE-D/WRB-1 to perform departures from nucleate boiling (DNB)

11 calculations for these fuel types, the grid spacing restriction must be rescinded.

2.0 REGULATORY EVALUATION

Title 10 of the Code of FederalRegulations Part 50, Section 34, "Contents of applications;

technical information," requires that safety analysis reports be submitted that 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 design process, licensees (or Vendors) performwreload safety evaluations to ensure that their

safety analyses remain bounding for the design cycle. To confirm that the analyses remain bounding, licensees confirm those key inputs to the safety analyses such as the critical heat flux 0(CHF) are conservative with respect to the current design cycle. If key safety analysis 0parameters are not bounded, a re-analysis or a re-evaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

Appendix B to DOM-NAF-2-A describes Dominion's methodology for implementing the WRB-1 CHF correlation in its sub-channel code, VIPRE-D. The NRC staff has previously reviewed and approved this correlation (see Reference 3). The NRC staffs review will be based on the

evaluation of technical merit and compliance with any applicable regulations of the revisions.

3.0 TECHNICAL EVALUATION

3.1 Background Information

Dominion is requesting the removal of the gdd spacing restriction note above imposed by

Appendix B of the VIPRE-D Fleet Report DOM-NAF-2-A. The purpose of this restriction was to exclude the use of the VIPRE-DIWRB-1 code/correlation with Westinghouse IFM vane grids fuel

types. The restriction was originally placed on WRB-1 in the issuance of the safety evaluation

for the COBRA/WRB-1 topical report, VEP-NE-3-A. When the VIPRE-D fleet topical report was developed, the restriction from the COBRA safety evaluation was included in the WRB-1

Qualification in Appendix B of DOM-NAF-2-A (Reference 3). However, Westinghouse has

developed several fuel products that incorporate IFMs (e.g. 15x15 VANTAGE+ and 15x15 Upgrade). Consequently, for Dominion to be able to use VIPRE-D/WRB-1 to perform DNB

calculations for these fuel types, the grid spacing restriction must be rescinded.

0 3.2 Technical Justification for Removing the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A Appendix B of DOM-NAF-2-A documents Dominion's qualification of the WRB-1 correlation with

the VIPRE-D code. This qualification was performed against a subset of the data from the

Columbia-EPRI CHF database for Westinghouse "R" grid 17x1 7 and 15x1 5 fuel (Reference 4).

This is the same subset of the Columbia-EPRI CHF database used by Dominion in the

qualification of the WRB-1 correlation with the COBRA, code (Reference 5). Appendix B

) summarizes the data evaluations that were performed to qualify the VIPRE-D/WRB-1 code/correlation pair and to develop the corresponding departure from nucleate boiling ratio (DNBR) design limits for the correlation. In addition, Appendix B provides the range of

application for operating conditions.

0 0

Westinghouse conducted confirmatory CHF testing in the time period December 1998 to January 1999 to revalidate the WRB-1 applicability to the 15xi5 Vantage+ fuel designs which included IFM grids. The CHF test section consisted of a 4x4 typical cell with geometry prototypical of 15x15 Vantage+ fuel design. Details of the test section, axial layout of the grids, assembly instrumentation, and axial and radial power distributions, are provided in Reference 2.

Westinghouse demonstrated in Reference 6 that the quantity of test data was comparable with other Westinghouse test bundles and that the test conditions were representative of the WRB-1 database. Westinghouse evaluated the test data for normality, and compared the test data to the WRB-1 database using the F- and T-statistical tests. The results indicated that the test data produced a 95/95 DNBR limit less than 1.17 (the accepted WRB-! design limit) and a mean M/P (measured to predicted CHF) greater than 1.0, which shows that these results are conservative.

Westinghouse showed similar results using VIPRE and THING computer codes. Westinghouse concluded that the CHF test data can be conservatively considered as part of the WRB-1 database and the WRB-1 DNBR design limit of 1.17 can be conservatively applied to 15X15, Vantage+ fuel. The test results were discussed during a March 17, 1999, meeting between the NRC, Westinghouse, and New York Power Authority (NYPA). Westinghouse documented this meeting in a letter to the NRC dated March 29, 1999 (Reference 6). Further, the NRC staff reviewed data and documentation of the tests performed by Westinghouse, leading to approval of an amendment for Indian Point Unit 3 (Reference 7).

Dominion replicated the Westinghouse evaluation of the CHF test data using the VIPRE-D Code with the WRB-1 correlation to develop a 95/95 DNBR limit for the test data. One-sided tolerance theory (Reference 8)was used by Dominion for the calculation of the test data DNBR design limit. This theory allows the calculation of a DNBR limit so that, for a DNBR equal to the design limit, DNB will be avoided with 95 percent probability at a 95 percent confidence level. A comparison of the WRB-1 M/P results between the Westinghouse VIPRE-W and Dominion VIPRE-D Code indicated excellent agreement.

Because all the statistical techniques used in Table 1 of Dominion's November 20, 2009, submittal (Reference 8) assume that the original data distribution is normal, it is necessary to verify that the overall distribution for the M/P ratio is a normal distribution. To evaluate whether the distribution is normal, the D' normality test was applied. This D' value is within the range of acceptability for the number of data points for conducting a 95/95DNBR limit. Thus, Dominion concluded that the M/P distribution for the test data is indeed normal. Based on the results listed in Table 1, the 95/95 DNBR limit for the test data can be calculated using the typical DNBR relation which is presented in Reference 1.

In addition, Dominion also performed an F-Test and a T-Test. The F-Test is performed to determine if the variances of two variables, the WRB-i variance and the variance of the test data, are equal. Thus, the NRC staff agrees that the result of this analysis showed that the two variances were indeed equal, and the additional IFM data collected for the 15x15 and 17x17 fuel designs can be included in the WRB-1 data base (Reference 8).

The T-test was used to compare the means of populations with equal variances. The NRC staff requested the licensee to provide additional information in order to complete the review. The questions, conveyed in an e-mail dated March 9, 2010 (ADAMS Accession No. ML100700236),

requested Dominion to clarify statistical calculations and define the expected operational range of the correlation. Dominion responded by Reference 9, providing a sample calculation of its S

0 0

0

w statistical method and providing a specific table of ranges for the correlation. The NRC staff 0 found Dominion's response in Reference 9 acceptable.

0

4.0 CONCLUSION

The NRC staff has reviewed the request by Dominion to remove a mixing vane grid spacing restriction contained in Appendix B of Fleet Topical Report DOM-NAF-2-A. Based on its review, t* the NRC staff concludes that the removal of the restriction is acceptable.

1

5.0 REFERENCES

1. Letter from J. Alan Price (Dominion Resources Services, Inc.) to NRC, proposing to

revise Appendix B to Fleet Report DOM-NAF-2-A to remove a restriction associated with mixing vane grid spacing, August 28, 2009 (ADAMS Accession No.

ML092430338).

2. Letter from E. S. Grecheck (Dominion) to NRC, requesting clarification of the NRC safety

evaluation for Dominion Fleet Report DOM-NAF-2, June 1, 2006 (ADAMS Accession

No. ML061530114).

3. Dominion Fleet Report, DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the

) VIPRE-D Computer Code," R. M. Bilbao y Leon, September 2006 (ADAMS Accession No. ML062650184).

4. Technical Report EPRI NP-2609, "Parametric Study of CHF Data, Volume 3, Part 1; Critical Heat Flux Data," C. F. Fighetti & D. G. Reddy, September 1982.

5. Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," R. C.Anderson, July 1990 (ADAMS Legacy Library Accession

No. 90092600.97).

0

6. Letter from H. A. Sepp (Westinghouse) to NRC. "Notification of FCEP Application for DNB Testing and Revalidation of WRB1- Application to the 15x15 Vantage+ Fuel Design

(Proprietary)," March 29, 1999 (ADAMS Legacy Library Accession No. 9904050141)

7. Letter from G. F. Wunder (NRC) to J. Knubel (NYPA), "Indian Point Nuclear Generation

Unit No. 3 - Issuance of Amendment No. 191," September 2, 1999 (ADAMS Accession

No. ML003780850).

8. Letter from J. Alan Rice (Dominion Resources Services, Inc.) to NRC, supplementing the August 20, 2009 letter regarding revising Fleet Topical Report DOM-NAF-2-A, November 20, 2009 (ADAMS Accession No. ML093310330).

0

9. Letter from J. Alan Price (Dominion Resources Services, Inc.) to NRC, providing addition information in response to an NRC Request for Additional Information, April 16, 2010 (ADAMS Accession No. ML101100608).

0 Principal Contributor. Anthony Attard, NRR

Date: June 21,2010 0

0 0i 0

0 CLASSIFICATION/DISCLAIMER 0 S

The data, information, analytical techniques, and conclusions in this report have been prepared 5 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 S FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE S 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, 5 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.

S S

ABSTRACT S

As part of a continuing effort to improve core thermal-hydraulics methods, Dominion (Virginia S Electric and Power Company) is updating its capability for performing nuclear reactor analyses in support of its nuclear power stations. VIPRE is a core thermal-hydraulics computer code currently in wide use throughout the nuclear industry. VIPRE-D is the Dominion version of VIPRE, which has been enhanced by the addition of several vendor specific CHF correlations. Dominion has validated VIPRE-D with extensive code benchmark calculations, and the accuracy of VIPRE-D has been demonstrated through comparisons with other NRC-approved methodologies. VIPRE-D has been shown to meet or exceed the same standards for accuracy as methodologies currently being used by Dominion.

R S

S S

S DOM-NAF-2, Rev. 0.2-A2 S

0

S 0

S ACKNOWLEDGEMENTS We would like to express our immense gratitude for the labors that Dr. Rosa M. Bilbao y Le6n, also recognized as Madame VIPRE, had in development and implementation of the 0 VIPRE-D methodology at Dominion. She worked tirelessly to get every part of the 0 methodology to its current state today and lay the ground work for the future S development.

The original author [Dr. Rosa M. Bilbao y Le6n] would like to express her huge appreciation to Mr.

0 Kurt F. Flaig for his insight and considerable technical assistance in the development of the VIPRE-D methodology. In addition, she would like to thank Mr. Dana M. Knee, Ms. Jennifer L.

0 Meszaros and Mr. Sean M. Blair for their significant contributions in the development of the methodology documented herein. The author would also like to acknowledge all the members of 0 the Dominion Nuclear Safety Analysis group for their help and support through the long process involved in the production of this Fleet Report. Finally, the advice provided by Mr. Noval A. Smith, 0 Dr. Ross C. Anderson, Mr. Kerry L. Basehore, Mr. David A. Farnsworth and Mr. John H. Jones (from AREVA), and Mr. Yixing Sung (from Westinghouse) was very appreciated.

S 0

0 0

0 DOM-NAF-2, Rev. 0.2-A 3

S TABLE OF CONTENTS S

CLASSIFICATION/DISCLAIMER 2 ABSTRACT 2 ACKNOWLEDGEMENTS 3 TABLE OF CONTENTS 4 LIST OF TABLES 5 LIST OF FIGURES 5 ACRONYMS AND ABBREVIATIONS 6

1.0 INTRODUCTION

7 2.0 TOPICAL METHODOLOGY 9 2.1 VIPRE-D APPLICATION 9 2.2 COMPLIANCE WITH VIPRE-01 SER 11 3.0 CODE DESCRIPTION 12 3.1 VIPRE-01 12 3.2 VIPRE-D 13 4.0 VIPRE-D MODELING 14 4.1 RADIAL NODALIZATION 15 4.2 AXIAL NODALIZATION 16 4.3 FUEL ROD MODELING 16 4.4 POWER DISTRIBUTION 17 4.5 TURBULENT MIXING 18 4.6 AXIAL HYDRAULIC LOSSES AND CROSSFLOW RESISTANCE 19 4.7 FORM LOSS COEFFICIENTS 20 4.8 TWO-PHASE FLOW AND HEAT TRANSFER CORRELATIONS 20 4.9 CRITICAL HEAT FLUX CORRELATIONS 21 4.10 ENGINEERING FACTORS 22 4.11 BOUNDARY CONDITIONS 23 4.12 RUN CONTROL PARAMETERS 23 5.0 QUALIFICATION OF THE VIPRE-D SUBCHANNEL MODEL 24 5.1 STEADY STATE APPLICATION 24 5.2 MAIN STEAM LINE BREAK APPLICATION 28 5.3 TRANSIENT APPLICATION 30 5.4 SENSITIVITY STUDIES 33

6.0 CONCLUSION

S 36

7.0 REFERENCES

38 R

S S

S DOM-NAF-2, Rev. 0.2-A 4 S

S

S S

0 S APPENDIXES S

0 APPENDIX A: QUALIFICATION OF THE F-ANP BWU CHF CORRELATIONS IN THE DOMINION VIPRE-D COMPUTER CODE APPENDIX B: QUALIFICATION OF THE WESTINGHOUSE WRB-1 CHF CORRELATION IN 0 THE DOMINION VIPRE-D COMPUTER CODE APPENDIX C: QUALIFICATION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION IN S THE DOMINION VIPRE-D COMPUTER CODE 0

0 0 LIST OF TABLES 0 Table 2.1-1: UFSAR Transients Analyzed with VIPRE-D 10 0 Table 4.0-1: VIPRE-D Modeling Summary 15 S Table 4.4-1: Typical Radial Peaking Factors for a 1 / 8 th Core Model of 157 17x17 Fuel Assemblies Modeled with 12 Channels and 14 Rods 18 0 Table 5.1-1: Range of VIPRE-D / LYNXT 173 Benchmark Statepoints 25 0 Table 5.3-1: Summary of VIPRE-D Sample Transients 31 0 Table 5.4-1: Statistical Analysis of the MDNBR Results for the Four Sets of Two-Phase Models 34 0

0 0 LIST OF FIGURES 0 Figure 5.1-1: Typical North Anna VIPRE-D 12-Channel Model for F-ANP AMBW Fuel Assemblies 26 0 Figure 5.1-2: VIPRE-D vs. LYNXT for the 173 Statepoints 27 Figure 5.2-1: Typical VIPRE-D 14-Channel Model for North Anna Cores with F-ANP AMBW 0 Fuel 29 0 Figure 5.3-1: VIPRE-D RWAP Transient Sample Calculation Results 32 Figure 5.3-2: VIPRE-D LOFA Transient Sample Calculation Results 32 Figure 5.4-1: VIPRE-D vs. LYNXT for the Four Sets of Two-Phase Models 35 0

DOM-NAF-2, Rev. 0.2-A 5

ACRONYMS AND ABBREVIATIONS S

AMBW Advanced Mark-BW AO Axial Offset BWR Boiling Water Reactor CHF Critical Heat Flux CTL Core Thermal Limit DNB Departure from Nucleate Boiling 0 DNBR Departure from Nucleate Boiling Ratio EPRI Electric Power Research Institute F-ANP Framatome Advanced Nuclear Power FLC Form Loss Coefficients FTM Turbulent Momentum Factor LOCROT Locked Rotor Accident 0 LOFA Loss of Flow Accident MDNBR Minimum Departure from Nucleate Boiling Ratio S MSLB Main Steam Line Break MSMG Mid-Span Mixing Grid MVG Mixing Vane Grid NMVG Non-Mixing Vane Grid NAPS North Anna Power Station PWR Pressurized Water Reactor RWAP Rod Withdrawal At Power RWSC Rod Withdrawal from Subcritical SER Safety Evaluation Report UFSAR Updated Final Safety Analysis Report USNRC US Nuclear Regulatory Commission VIPRE Versatile Internals and Components Programs for Reactors - EPRI R

0 S

S S

DOM-NAF-2, Rev. 0.2-A 6 S

S

0 S

1.0 INTRODUCTION

0 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 surface temperature. DNB is a concern in reactor design because of the possibility of fuel rod failure resulting from the increased rod surface temperature.

0

In order to preclude potential DNB related fuel damage, a design basis is established and is 0 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. DNBRs

greater than the design limit indicate the existence of thermal 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.

0 0 Dominion (Virginia Power) has used the COBRA IIIc/MIT computer code (Reference 8) to perform the thermal-hydraulic analyses discussed above. COBRA is licensed to evaluate the thermal 5i margin for North Anna Power Station (NAPS) and Surry Power Station cores containing Westinghouse fuel. However, Dominion's nuclear assets and fuel products require enhanced core 5thermal-hydraulic capabilities. As a consequence, Dominion has decided to implement a new

thermal-hydraulic analysis computer program to analyze multiple fuel types.

VIPRE-D is the Dominion version of the computer code VIPRE (Versatile Internals and Components Program for Reactors - EPRI), developed for EPRI (Electric Power Research Institute) by Battelle

Pacific Northwest Laboratories in order to perform detailed thermal-hydraulic analyses to predict

CHF and DNBR of reactor cores (References 1 through 5). VIPRE-01 has been approved by the

U.S. Nuclear Regulatory Commission (USNRC) (References 6 and 7). VIPRE-D, which is based

t upon VIPRE-01, MOD-02.1, was developed by Dominion to fit the specific needs of Dominion's nuclear plants and fuel products by adding vendor specific CHF correlations and customizing its input and output. Dominion, however, has not made any modifications to the NRC-approved constitutive models and algorithms in VIPRE-01.

S

This report describes Dominion's use of the VIPRE-D code, including modeling and qualification

for Pressurized Water Reactors (PWR) thermal-hydraulic design. This report demonstrates that

the VIPRE-D methodology is appropriate for PWR licensing applications.

R S

0 S

DOM-NAF-2, Rev. 0.2-A7 S

S

0 S

0 This report is organized into six sections. Section 2 provides a description of VIPRE-D 0 methodology and intended applications, including a discussion on VIPRE-D compliance with the 0 VIPRE-01 Safety Evaluation Report (SER). Section 3 describes the VIPRE-D code and its 0 capabilities. Section 4 describes the VIPRE-D modeling of PWR cores and fuel rods. Section 5 provides VIPRE-D benchmark calculations against other subchannel codes for PWR DNB 0

analyses, such as Framatome ANP (F-ANP) LYNXT (Reference 14). Conclusions and references 0 are presented in succeeding sections. The topical allows for a series of appendixes, each one 0 containing the verification and qualification of additional CHF correlations with the VIPRE-D code. 0 0

0 0

0 S

0 0

DOM-NAF-2, Rev. 0.2-A 8 0

0

2.0 TOPICAL METHODOLOGY

2.1 VIPRE-D APPLICATION The intended VIPRE-D applications are consistent with the Dominion COBRA applications for

PWRs using USNRC approved methodologies (Reference 8). The VIPRE-D applications include

DNB analyses to define PWR core safety limits that provide the basis for reactor protection

setpoints, and to perform DNBR calculations in reactor transients. While VIPRE-D is able to

- model Boiling Water Reactors (BWR), its BWR features and capabilities are not discussed for

qualification in this report. Furthermore, the rod conduction model present in VIPRE-D will not be used. All VIPRE-D models will employ the dummy rod model.

Dominion plans to use the VIPRE-D code for:

0

1) Analysis of 14x14, 15x15 and 17x17 fuel in PWR reactors.

0

2) Analysis of DNBR for statistical and deterministic transients in the Updated Final Safety Analysis Report (UFSAR), as identified in Table 2.1-1. Additional DNBR transients that are plant specific may be analyzed in a plant specific application

that would be submitted to the USNRC for review and approval.

3) Steady state and transient DNB evaluations.

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

5) Providing the basis for reactor protection setpoints.

0

6) 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 this document.

0

Plant specific applications of VIPRE-D would include:

0

1) Technical Specifications change request to add DOM-NAF-2 and Appendixes to

the plant's COLR list.

2) Statistical Design Limit(s) for the relevant code/correlation(s) 0 0

0 DOM-NAF-2, Rev. 0.2-A9 0

0 0

3) Any technical specification changes related to OTAT, OPAT, FAI or other reactor 0

protection function, as well as revised Reactor Core Safety Limits.

0

4) List of UFSAR transients for which the code/correlations apply (see Table 2.1-1).

0 Table 2.1-1: UFSAR Transients Analyzed with VIPRE-D 0 0

1 Accidental depressurization of the main steam system 0 2 Accidental depressurization of the reactor cooling system 0 3 Excessive heat removal due to feedwater system malfunction 4 Excessive load increase 0 Inadvertent operation of emergency core cooling system during power 0 operation 0 6 Locked reactor coolant pump rotor or shaft break 0

7 Loss of external electrical load and/or turbine trip 8 Loss of forced reactor coolant flow S

9 Loss of normal feedwater S 10 Major rupture of a main feedwater pipe S

11 Rod cluster control assembly misalignment / Dropped rod/bank 12 Rod cluster control assembly bank withdrawal at power 0

13 Rod cluster control assembly bank withdrawal from subcritical 0

14 Rupture of a main steam pipe S 15 Single rod cluster control assembly withdrawal at full power 16 Startup of an inactive reactor coolant loop 0

17 Uncontrolled boron dilution S 0

S 0

S S

0 S

0 0

0 S

DOM-NAF-2, Rev. 0.2-A 10

0

2.2 COMPLIANCE WITH VIPRE-01 SER

In order to meet the USNRC's requirements listed in the VIPRE-01 SER (References 6 and 7),

Dominion will apply the VIPRE-D code for PWR licensing applications under the following conditions:

1) The application of VIPRE-D is limited to PWR licensing calculations with heat transfer regime up to CHF. VIPRE-D will not be used for post-CHF calculations or

for BWR calculations.

2) VIPRE-D analyses will only use DNB correlations that have been reviewed and

approved by the USNRC. The VIPRE-D DNBR calculations will be within the

USNRC approved parameter ranges of the DNB correlations, including fuel assembly geometry and grid spacers. The correlation DNBR design limits will be derived or verified using fluid conditions predicted by the VIPRE-D code. Each

DNB correlation will be qualified or verified in appendixes to this report.

0

3) This report provides the necessary documentation to describe the intended uses of

VIPRE-D for PWR licensing applications. The report 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.

Dominion only applies models and correlations already existing in VIPRE-01 and

previously approved by the USNRC (Section 4).

4) For transient analysis, appropriate time steps are selected to ensure numerical stability and accuracy. The Courant number, which is based on flow velocity, time

step and axial node size, is set to be greater than one in VIPRE-D transient calculations whenever a subcooled void model is used.

5) VIPRE-D is maintained within Dominion's 10CFR50, Appendix B Quality Assurance program.

0 S O-A-,Re.02A1 0

0

0 3.0 CODE DESCRIPTION 3.1 VIPRE-01 VIPRE is a computer code developed for EPRI by Battelle Pacific Northwest Laboratories in order to perform detailed thermal-hydraulic analyses of reactor cores (References 1 through 5). 0 VIPRE-01, MOD-02 was previously approved by the USNRC (References 6 and 7). The code 0 errors reported and verified since the release of VIPRE-01, MOD-02, as well as some documentation changes and other minor enhancements, were incorporated into version VIPRE-01, MOD-02.1 of the code, which was released in May 2001. These changes did not alter the basic models, equations and algorithms in the code, and it was verified that all significant differences between the results of the VIPRE Standard Testcases for MOD-02.1 and MOD-02 were accounted for and were the result of error corrections.

VIPRE-01 uses the subchannel analysis concept where a reactor core is divided into a number of flow channels that communicate laterally by crossflow and turbulent mixing. Conservation equations of mass, axial and lateral momentum, and energy are solved for the fluid enthalpy, axial flow rate, crossflow, and momentum pressure drop. A detailed description of the VIPRE-01 subchannel equations can be found in Reference 1. The VIPRE-01 flow field is assumed to be incompressible and homogeneous. It is assumed that any lateral flow is directed by the gap through which it flows, and it loses its sense of direction after leaving the region. Since crossflow S is assumed to exist only between two adjacent channels, no external lateral boundary conditions are required.

The VIPRE-01 heat transfer model is capable of solving the conduction equation for the temperature distribution within the fuel rods and provides the heat source term for the fluid energy 0 equation. The full boiling curve can be incorporated into the heat transfer model, from single-phase convection through nucleate boiling to the DNB point, and from transition boiling to the film boiling regime. A detailed description of the VIPRE-01 heat transfer model can be found in Reference 1.

VIPRE-01 offers two numerical solution options: the upflow solution, which is similar to the one in COBRA-IIIC; and the recirculation solution scheme adapted from COBRA-WC. Both solution schemes iteratively solve the same finite difference equations and use the same model and 0 correlations for heat transfer, wall friction, fluid state and two-phase flow. The difference between them is in the numerical method used to obtain the flow and pressure fields. Both solution schemes yield essentially the same results (Reference 4, Section 7.3). However, the recirculation solution scheme is applicable to core conditions having flow reversal and recirculation. Either solution scheme can be used for PWR analysis.

0 0

0 DOM-NAF-2, Rev. 0.2-A 12 0

0

S O VIPRE-01 modeling of a PWR core is based on the one-pass modeling approach (Reference 1),

O in which hot channels (subchannels with the highest enthalpy rise) and their adjacent region are modeled in detail, while the remainder of the core is modeled simultaneously on a relatively coarse mesh. A reactor core can be modeled in a small number of channels while still maintaining sufficient detail and accuracy around the hot channels. A one-pass model contains lumped channels that comprise total flow area and heated and wetted perimeters of the individual subchannels. The lumped channel gives uniform conditions over the entire flow area of the

channel. Some input parameters of the lateral momentum equation in the VIPRE-01 code are O adjusted in order to obtain the correct crossflow for the lumped channel. The VIPRE-01 one-pass modeling has been approved by the USNRC (References 6 and 7).

3.2 VIPRE-D S

VIPRE-D, which is based upon VIPRE-01, MOD-02.1, was developed by Dominion to fit the specific needs of Dominion's nuclear plants and fuel products by adding vendor specific CHF O correlations and customizing its input and output. Dominion, however, has not made any

modifications to the NRC-approved constitutive models and algorithms in VIPRE-01 and the O computational philosophy of VIPRE-D remains unchanged from VIPRE-01. Therefore, the O VIPRE-01 qualification is fully applicable to VIPRE-D.

In addition to minor formatting changes and corrections to reported code errors, the main O enhancement made to VIPRE-01, MOD-02.1 to obtain VIPRE-D is the addition of several vendor O proprietary CHF correlations. Additional customizations were made in VIPRE-D's input and output O to integrate it seamlessly into Dominion's thermal hydraulic methodologies. Additional CHF correlations may be added to the code in the future. Each one of these DNB correlations will be O qualified or verified in its own appendix to this report, and submitted to USNRC for review and approval, prior to licensing use.

The VIPRE-D coding changes do not alter the fundamental computational method and solution

scheme of the VIPRE-01 code. It has been demonstrated by running the VIPRE Standard

I Testcases that the additions and modifications made to create VIPRE-D have been correctly implemented into the code and have not affected in any way the original internal models and O algorithms in the code. VIPRE-D has been developed and is maintained in accordance with Dominion's 10CFR50 Appendix B Quality Assurance program. The VIPRE-01 User's Manual (Reference 2) is fully applicable to VIPRE-D, but it has been augmented with an in-house User's O Manual that clarifies the selection of VIPRE-D specific CHF correlations and enhancements.

R S

0 S

S 0

S ODOM-NAF-2, Rev. 0.2-A 13 0

S

0 4.0 VIPRE-D MODELING The goal of Section 4 of this report is to comply with Condition 3 of the VIPRE-01 SER S (References 6 and 7), which requires that each organization using VIPRE-01 provide justification for specific modeling assumptions, including the choice of two-phase flow models and correlations, heat transfer correlations and turbulent mixing models. As such, the methodology and guidelines used to create the VIPRE-D model for a typical Dominion reference plant core are described in this section.

As discussed in Section 3 of this report, no substantive modifications have been made to VIPRE-D. All the models selected and discussed herein were previously approved in VIPRE-01.

The modeling choices described below (Table 4.0-1), which are not plant specific, were developed in a manner consistent with the USNRC approved Dominion COBRA models (Reference 8) and with standard industry practice. These modeling choices will be used for the qualification or verification of all the CHF correlations included in the Appendixes to this report, unless otherwise specified in the particular Appendix.

0 Sections 4.1 (radial nodalization), 4.2 (axial nodalization), 4.4 (power distribution), 4.7 (form loss coefficients), 4.9 (CHF correlations), 4.10 (engineering factors) and 4.11 (boundary conditions) describe modeling areas that are fuel or accident dependent and would have to be determined based on the particular core and the type of analysis to be performed. The remaining sections, listed in Table 4.0-1, describe modeling choices that are independent of the fuel type.

Section 5.0 of this report describes a specific example applying these guidelines to a North Anna Power Station core containing F-ANP Advanced Mark-BW (AMBW) fuel assemblies. Extensive code benchmark calculations have confirmed that the VIPRE-D models specified in sections 4.1 through 4.12 in this report produce essentially the same results as equivalent F-ANP LYNXT models (Reference 13).

R S

S S

0 S

S S

S DOM-NAF-2, Rev. 0.2-A 14 S

S

S 0

S Table 4.0-1: VIPRE-D Modeling Summary 0

VIPRE-01 MODEL DOMINION SELECTION SECTION S

Fuel Rod Modeling "Dummy" rod model Section 4.3 S

Turbulent Mixing No momentum mixing Section 4.5 S ABETA fuel dependent Axial Friction Losses McAdams Correlation Section 4.6 9

0 Crossflow Resistance Idel'Chik Correlation Section 4.6 9 Two-Phase Flow EPRI Correlations Sections 4.8 & 5.4 0

S Heat Transfer Dittus-Boelter Correlation Section 4.8 Run Control Parameters Default Options with Section 4.12 0 Courant > 1 for transients 0

S 4.1 RADIAL NODALIZATION While the techniques used in formulating the hydraulic representation of a typical core are S applicable in general to all PWRs, the specifics of the model change with the type of fuel present 0 in the particular core and the type of analysis being performed. In general it is assumed that the core presents 1 /8th symmetry, and thus it is only necessary to model 1 /8th of the core. It is also assumed that the hot assembly is located at the center of the core, and therefore, the 1 /8th core 0 model will contain 1 /8th of the hot assembly. The adequate number of channels to model a given 0 core must allow simulating the entire core, while having a detailed subchannel model surrounding the hot channels. A set of subchannels surrounding the hot channels (i.e., hot thimble cell and 0 hot typical cell) is sufficient to provide adequate solution detail of the flow field in the vicinity of the hot subchannels (Reference 2). If the model is used for the analysis of main steam line break S (MSLB) events, it is also necessary to account for the core inlet enthalpy maldistribution when defining the number of channels. This modeling methodology is applicable to 14x14, 15x15 and S 17x17 PWR fuel.

These modeling guidelines are consistent with the USNRC approved Dominion COBRA models (Reference 8) and with standard industry practice. The adequacy of using a 1/8th core model and the above modeling guidelines has been verified through benchmark calculations with the F-ANP LYNXT code (References 13 and 14), and will be discussed in Section 5.0.

DOM-NAF-2, Rev. 0.2-A 15

0 4.2 AXIAL NODALIZATION 0 The finite differences methods used in VIPRE-D require that sufficient axial nodes be provided to S resolve the details of the flow field and the axial power profiles. Dominion models use an axial nodalization scheme that places all the mixing and non-mixing vane grids at the upper edges of the axial nodes for better numerical convergence, while preserving the actual grid spacing. This is important because VIPRE-D applies the pressure loss associated with a node at the top edge of the node. Therefore, it is important to create a nodal distribution that ensures that the axial locations where the pressure losses are applied match the actual axial locations for each spacer grid.

0 VIPRE-D allows a PWR core to be modeled with variable axial nodal length. VIPRE-D offers a great deal of control and flexibility by allowing the user to define both the geometry and the axial power shapes with as much detail as needed in the critical areas of the model and with not so much detail in less critical areas. Dominion models use typical node lengths of 2 inches. A maximum node length of 6 inches will be used in the models. Selection of a very small node length is not reasonable since an excessive number of nodes will add significantly to the run time of the problem and the memory required to store the results without actually improving the precision. The subchannel model qualification in Section 5.0 demonstrates the acceptability of 5 using maximum node lengths of 6 inches with axial node lengths of about 2 inches in the MDNBR region.

The length of the axial nodes should also be taken into account when running transient problems in order to satisfy the Courant number limit (The Courant number is defined as the axial velocity u times the numerical approximation of the time derivative - uAt/AX ). The VIPRE-D solution methods are generally fully implicit and have no time step size limitations for numerical stability.

However, solution instability could occur in transient calculations using a subcooled void model that was developed based on steady state data, such as the EPRI subcooled void model. In these cases, and to avoid numerical instabilities, appropriate time step sizes and axial node sizes are selected in transient heat flux and DNBR calculations to ensure that the Courant number is greater than one.

0 These modeling guidelines are consistent with the USNRC approved Dominion COBRA models (Reference 8) and with standard industry practice. Dominion VIPRE-D axial nodalizations are created according to these guidelines.

0 4.3 FUEL ROD MODELING S A typical VIPRE-D model defines the number of rods appropriate for the number of channels S selected in the radial nodalization (Section 4.1), normally in accordance with the type of fuel 0 present in the core, and uses the "dummy" rod model to represent them. In the dummy rod model 0

0 DOM-NAF-2, Rev. 0.2-A 160 0

0

S 0

there is no calculation of the heat transfer and the temperature distribution within the fuel rod, and

the surface heat flux for each rod is specified as an input parameter. Unheated rods, such as instrument tubes and guide tubes, do not need to be modeled as rods. They are taken into account when calculating the flow area, the wetted and heated perimeters, and the crossflow gaps in the appropriate channels, but they are not modeled as separate entities. Dominion does Snot plan to use the conduction model present in the code.

The VIPRE-D model accounts for a fraction of the core power being generated directly in the coolant due to gamma heating and neutron absorption. For the safety analysis, it is assumed that 97.4% of the reactor power is generated within the fuel rods, and the remaining 2.6% is generated directly in the coolant. VIPRE-D fuel rod modeling and the treatment of the gamma heating is

consistent with the USNRC approved Dominion COBRA production models (Reference 8).

4.4 POWER DISTRIBUTION

In the VIPRE-D model, an axial power profile is entered to specify the power generated by each 0axial node relative to the average. A radial power factor that determines the rod power relative to 0the average core power is assigned to each rod.

0 DNBR calculations are typically performed with reference axial power shapes. For example, the typical reference axial power shape used in establishing core thermal limits is a chopped cosine shape with a peak-to-average value of 1.55. This reference power shape is supplemented by

other axial shapes skewed to the bottom or to the top of the core to determine the reduction of trip setpoints on excessive axial power imbalance. Dominion's VIPRE-D model interpolates in the

axial power table using the spline fit option, as opposed to the default linear interpolation option.

The spline fit option was added to VIPRE-01, MOD02.1 and provides a slightly smoother axial

power profile integration. A sensitivity analysis of the impact of this option was performed by Dominion, and virtually identical MDNBR results were obtained with both options.

The radial power distribution is specified by assigning to each dummy rod a radial power factor that specifies the rod power relative to the average core power. The power distributions provide a

gradual power gradient with the highest peaking around the hot channels (i.e., hot thimble cell

and hot typical cell) to reduce the benefit of crossflow into the hot channel. The VIPRE-D models

i apply the peak FAH to a rod in the hot thimble cell and the hot typical cell. This radial modeling results in a conservative evaluation of DNBR in the hot channel and hot pin, since the mixing effects in the center of the core are significantly reduced. A typical radial power distribution for a

1/8th core model of 157 17x17 fuel assemblies, adjusted for a 1.587 maximum peaking factor, is

D described in Table 4.4-1.

R 0

0 0

0 I*DOM-NAF-2, Rev. 0.2-A 17 0

0

0 S

0 0

Table 4.4-1: Typical Radial Peaking Factors for a 1 /8th Core Model of 157 17x17 Fuel 0 Assemblies Modeled with 12 Channels and 14 Rods Statistical 0

Rod Relative Powerofrd Number MaiuFH Number Nubr fiiof ~1.8 rods Maximum FAH 0 Ni 1.587 0 1 1.0 0.5 1.587 2 0.99748 0.5 1.583 0 3 0.993699 0.5 1.577 S 4 0.994959 1 1.579 0 5 0.986767 0.5 1.566 6 0.988658 1 1.569 0

7 0.996219 1 1.581 0 8 0.988028 0.25 1.568 9 0.986767 0.5 1.566 0 10 0.991178 0.5 1.573 0 11 0.983617 0.5 1.561 0 12 0.980466 0.125 1.556 13 0.982987 26.125 1.560 S

i=14 i=13 Z NI.- ZY FmaxA H fi Ni 14 i=1 5148 0.99639 N 14 0 S

4.5 TURBULENT MIXING 0 The VIPRE-01 turbulent mixing model accounts for the exchange of energy and momentum S between adjacent subchannels due to turbulence. This is not a turbulence model, but an attempt 0 to empirically account for the effect of turbulent mixing. The following inputs are needed to setup 0 this model:

S

Turbulent Momentum Factor (FTM), which can range from 0.0 to 1.0, measures how efficiently the turbulent crossflow mixes momentum. The VIPRE-01 User's Manual (Reference 2) recommends a value of 0.8 for FTM and explains that VIPRE is not very 0 sensitive to the value of FTM. In Dominion models FTM has been conservatively set to S 0.0, which indicates that the turbulent crossflow mixes enthalpy only and not momentum.

This modeling approach is consistent with USNRC approved Dominion COBRA models (Reference 8).

0 0

The model for turbulent mixing chosen for single phase mixing describes the mixing as 0 w'=A x S x G, where A is an empirical mixing coefficient (the variable ABETA in VIPRE-D) S entered by the user, S is the rod-to-rod gap width (ft), and G is the average mass velocity 0 0

0 S

DOM-NAF-2, Rev. 0.2-A 18 0

S 0

in the channels linked by a given gap (Ibm/ft 2-s). This coefficient ABETA, which depends

I on the particular fuel type and can range from 0.0 to 0.1, is typically set to 0.038. The two-phase turbulent mixing is computed in the same way as the single phase. This is the default model in the code and it is consistent with USNRC approved Dominion COBRA

models (Reference 8).

Since turbulent mixing is a subchannel phenomenon, the value of the turbulent mixing coefficient needs to be corrected for lumped channels to reflect the effect of lumping 0 together many rod-to-rod gaps. The value of ABETA for the flow path between a subchannel and a lumped channel is defined as:

0 Subchannel CentroidDistance A

ABETumped = ABETAubchannel x [4.5.1]

LumpedChannelCentroid Distance S

The impact of correcting the value of the turbulent mixing coefficient for the flow paths

connecting to lumped channels has been quantified with a sensitivity analysis, which demonstrated that both approaches yield essentially the same results. This methodology is 4consistent with standard industry practice. In larger lumped regions, on the order of a bundle or larger, turbulent mixing tends to be smeared out by the effect of averaging on both flow and enthalpy. As a consequence, the turbulent mixing coefficients for the flow paths between lumped channels are set to zero (Reference 4, Section 7.2).

9

4.6 AXIAL HYDRAULIC LOSSES AND CROSSFLOW RESISTANCE Axial friction losses are calculated with the McAdams correlation, which has been shown to provide an excellent approximation to the Colebrook smooth pipe formulation for single phase axial friction factor for the range 3.104 < Re < 2.106 (Reference 11). This is the same correlation used in the USNRC approved Dominion COBRA models (Reference 8).

F= MAX (0.184

Re-0 2 + 0.0 [turbulent], 64.0

Re-10 + 0.0 [laminar])

0* [4.6.1]

0Lateral resistance for a subchannel is calculated in both the turbulent and laminar regions with a

l Blasius-type function of the gap Reynolds number, where the coefficient A is calculated using the Idel'Chik empirical correlation for a bundle of circular tubes in vertical columns (Reference 12, p.332).

KG = A Relateralo02 [4.6.2]

-00 5

where A is defined as:

~FuelRodOD A = 1.52 SubchannelPitch _1 ] . [4.6.3]

0 0

5 DOM-NAF-2, Rev. 0.2-A 19 0

0

0 In order to correctly calculate the effective crossflow resistance for the lumped channels, the subchannel crossflow resistance is multiplied by the ratio of the lumped channel centroid distance and the subchannel centroid distance. This treatment is consistent with the USNRC SER for VIPRE-01 (Reference 6). 0 0

4.7 FORM LOSS COEFFICIENTS The local form loss coefficients (FLC) associated with a given fuel assembly type are obtained by 0 the vendor from full-scale hydraulic tests of the fuel assemblies. These form losses are specified for each fuel component (non-mixing grids, mixing grids, mid-span mixing grids, etc.) and for each type of subchannel (unit cell, corner cell, etc). Thus, VIPRE-D allows the definition of different FLCs for different channels and at different axial locations.

In the VIPRE-D models, the FLCs are axially placed at the upper edges of the axial nodes immediately below the corresponding component (mixing vane grids, mid-span mixing vane grids, 0 etc). VIPRE-D places the pressure loss associated with a node at the top edge of the node, thus 0 applying the pressure losses at the actual axial locations for each spacing grid. The impact of slightly varying (upward and downward) the axial location where the FLCs are applied was studied with a sensitivity analysis, which showed an insignificant change in DNBR.

4.8 TWO-PHASE FLOW AND HEAT TRANSFER CORRELATIONS O VIPRE-01 has a number of empirical correlations available to simulate two-phase flow effects (Reference 1). These correlations can be grouped in three major categories: 1) two-phase friction multipliers; 2) subcooled void correlations; and 3) bulk boiling void correlations. In Reference 4, a sensitivity study was performed to assess the differences in the performance of the various 0 correlations and, although significant differences were not found, the EPRI models were defined 0 as the default models for VIPRE-01. The USNRC, in Reference 6, concluded that the EPRI void O models and EPRI correlation for two-phase friction are acceptable for licensing calculations.

Dominion performed yet another sensitivity study to verify that this set of two-phase flow correlations provided results approximate to results already approved by USNRC for the F-ANP AMBW fuel product. The set of two-phase flow correlations listed below was shown to provide the O closest comparison to the USNRC approved F-ANP LYNXT code for F-ANP AMBW fuel products 0 (Section 5.4) and was deemed to be the most suitable for Dominion applications. Dominion will apply this set of two-phase flow correlations for all applications unless future fuel types necessitate the use of a different set. In those cases, the selection of two-phase flow correlations will be described and justified in the appendix where the CHF correlations associated to that particular fuel type are qualified or verified.

0 0

0 DOM-NAF-2, Rev. 0.2-A 20 O 0

0

The selections are:

0

Subcooled Void Model: EPRI

Bulk Boiling Void Model: EPRI 0 Two-Phase Friction Multiplier: EPRI

Hot Wall Friction Correlation : NONE

VIPRE-D also requires the user to select the heat transfer correlations that describe the boiling curve. These selections (except the Single Phase Forced Convection Correlation), however, are

D only applied to the heat transfer solution if the conduction model is used. Since Dominion VIPRE-D models described herein use the "dummy" rod model (Section 4.3), the conduction model is ignored.

0

The Single Phase Forced Convection is modeled with the standard Dittus-Boelter correlation,

which is commonly used for this type of configuration (Reference 2).

ShDB = 0.023 .Re," 8 .Pr 0 4 . k [4.8.1]

De 0

where Re, is the Reynolds number for the liquid, Pr is the Prandtl number, k is the thermal conductivity of the fluid (Btu/s-ft-°F) and De is the hydraulic diameter in ft. This selection is

consistent with the USNRC approved Dominion COBRA models (Reference 8) and with standard

industry practice.

0

4.9 CRITICAL HEAT FLUX CORRELATIONS

VIPRE-D currently includes several CHF correlations applicable to various F-ANP and

Westinghouse fuel types. Dominion intends to add appendixes to the present report qualifying

various CHF correlations for fuel products to be used within the Dominion nuclear units. This

modular approach will allow simple submittals of additional CHF correlations for new fuel types in the future. The critical heat flux correlation to be used for a particular fuel type will be qualified in one of the appendixes and will have been approved by the USNRC for use with such fuel product.

The VIPRE-D CHF correlations will be used within the USNRC approved parameter ranges of the

CHF correlations, including fuel assembly geometry and grid spacers. The DNBR design limits

applied to each CHF correlation will be derived or verified using fluid conditions predicted by the

VIPRE-D code.

0 R

0 0

0 ODOM-NAF-2, Rev. 0.2-A 21 0

0

0 0

0 4.10 ENGINEERING FACTORS Variations in the fuel fabrication and core flow adverse to DNB margin are also considered in the 0 VIPRE-D models. Typical VIPRE-D models account for engineering hot channel factors for both enthalpy-rise and heat flux, as well as for inlet flow maldistribution. These engineering factors are fuel product dependent.

Local Heat Flux Engineering Hot Channel Factor, F E:

FeE accounts for pellet-to-pellet variations in enrichment, density and burnable absorber plus the effects of pellet-to-clad eccentricity and variations in the clad outer diameter. Used in the evaluation of the maximum linear heat generation rate, FeE has been determined to have negligible effect on DNB, and it is not used for most fuel types. FaE will be applied according to fuel vendor approved methodologies.

Engineering Enthalpy-Rise Hot Channel Factor, FAHE:

FAHE accounts for variations in the fuel enrichment, density, rod dimensions and pin pitch that affect the heat generation rate along the flow channel. Uncertainties in these variables are determined from sampling of manufacturing data. For deterministic analyses, FAHE is incorporated in the model as a multiplier to the energy input to the hot channel without affecting the surface heat flux. In statistical DNBR methods, FAHE is statistically convoluted into the DNBR design limit.

Stack Height Reduction:

Active fuel stack height varies during reactor operation due to the combined effects of fuel densification, swelling and thermal expansion. However, the treatment of this phenomenon is vendor specific and fuel specific. VIPRE-D models comply with the treatment specified by the fuel vendor.

0 Inlet Flow Reduction:

Core inlet flow maldistribution accounts for non-uniform flow distribution into each fuel assembly at the core inlet. Consistent with the USNRC approved Dominion COBRA methodology for PWR applications (Reference 8), a 5% flow reduction (maldistribution) to the hot assembly is applied in VIPRE-D models.

0 0

DOM-NAF-2, Rev. 0.2-A 22 0 0

0

4.11 BOUNDARY CONDITIONS

The VIPRE-D models require the following parameters as the input or the boundaries for calculations:

Core inlet temperature or enthalpy

. Core average power

System pressure

Core inlet flow rate

Core power distributions 0The core inlet temperature and inlet flow may be uniform or non-uniform, depending on the core

conditions being analyzed. The core power defines the thermal energy entering the fluid through

the fuel rods. The system pressure is assumed to be uniform throughout the VIPRE-D model. The core inlet flow conservatively excludes flow through bypass leakage, such as through the guide tubes.

0The core boundary conditions for VIPRE-D transient calculations can be obtained from system 0computer codes and neutronic codes. For example, the system code provides time-dependent reactor coolant system pressure, core average power, core flow rate and core inlet temperature for transient DNBR calculations. The neutronic codes provide core power distributions and

nuclear peaking factors such as FAH.

0 4.12 RUN CONTROL PARAMETERS The run control parameters determine the maximum and minimum number of iterations to be performed to find a solution, as well as the convergence limits and the damping factors used.

After a careful review, these values have been set to the defaults provided by the code (Reference 2). In a few occasions, when convergence problems have been reported by the code, 0the damping factors and/or the convergence limits have been adjusted in the models to allow the

code to converge. These convergence problems do not necessarily mean bad results or false convergence, just some numerical instability. Indeed, in most occasions, the results obtained by the code with the adjusted convergence limits or damping factors are nearly identical to the non-

converging results (Reference 6, Section 2.1).

The VIPRE-01 solution methods are generally fully implicit and have no time step size limitations 0 for numerical stability. However, solution instability could occur in transient calculations using a

t subcooled void model that was developed based on steady state data, such as the EPRI subcooled void model. In these cases, and to avoid numerical instabilities, appropriate time step sizes and axial node sizes are selected in transient heat flux and DNBR calculations to ensure that the Courant number is greater than one. This modeling guideline is consistent with VIPRE-01 SER Restriction #4 (see Reference 6 and Section 2.2 herein).

0 0

DOM-NAF-2, Rev. 0.2-A 23 0

0

0 0

5.0 QUALIFICATION OF THE VIPRE-D SUBCHANNEL MODEL The analyses shown in this section demonstrate that Dominion VIPRE-D models created using the 0 selections and modeling guidelines described in Section 4 of this report provide close comparison to other USNRC approved subchannel codes. This section is provided as an example to demonstrate in sufficient detail the validity of the methodology discussed herein, and it is not meant to be linked to a specific plant or fuel product.

5.1 STEADY STATE APPLICATION Dominion created a 12-channel model for F-ANP AMBW fuel at North Anna Power Station in accordance with the methodology described in Section 4 of this report. This VIPRE-D model of 0 the 1 /8th North Anna core consists of 12 channels (10 subchannels and 2 lumped channels) and 0 14 rods, as shown in Figure 5.1-1. The axial nodalization used in this model has been customized 0 for F-ANP AMBW fuel assemblies and contains 87 non-uniform axial nodes with typical node 0 lengths of 2 inches and a maximum node length of 6 inches. The reference axial power profile 0 (1.55 chopped cosine) was defined as an axial power profile table with 37 points. All other axial power shapes are defined as axial power profile tables with 32 points.

The AMBW fuel assembly consists of 264 fuel rods with an outside diameter of 0.374 inches 0 arranged in a 17x1 7 matrix with a pin pitch of 0.496 inches. The AMBW fuel contains several 0 advanced design features, such as mixing vane grids (MVG) and mid-span mixing grids (MSMG) in the upper two thirds of the heated length (Reference 13). The local FLCs used in this VIPRE-D 0 12-channel model were developed by F-ANP from full-scale hydraulic tests. 0 The F-ANP BWU CHF correlations, which have been specifically developed for use with the 0 AMBW fuel, were used in the 12-channel model. There are three BWU CHF correlations that 0 constitute the licensing basis for the F-ANP AMBW fuel assembly. These correlations use the 0 same basic equation, but are fit to different databases (References 9 and 10). VIPRE-D applies different BWU correlations at different axial levels, according to the following guidelines:

BWU-N, which is only applicable in the presence of non-mixing vane grids (NMVG), 0 is used from the beginning of the heated length to the leading edge of the first 0 structural MVG (Reference 9). 0

BWU-Z, which is the enhanced mixing vane correlation, is used from the leading 0 edge of the first structural MVG to the leading edge of the second structural MVG 0 (Reference 9). 0

" BWU-ZM, which is just BWU-Z with a multiplicative enhancement factor and is 0 applicable in the presence of MSMGs, is used from the leading edge of the second structural MVG to the leading edge of the last structural MVG (Reference 10). 0 0

0 0

0 DOM-NAF-2, Rev. 0.2-A 24 0 0

0

0 S

For the uppermost span, in which the end of heated length occurs less than one grid S

span beyond the last MVG, the BWU-Z correlation is used with a grid spacing equal S to the effective grid spacing (the distance from the last grid to the end of heated 0 length) (Reference 9).

VIPRE-D benchmark calculations were performed with the F-ANP LYNXT code and the 0 12-channel model created by F-ANP to model North Anna Power Station cores containing AMBW fuel assemblies. This benchmark uses 173 statepoints obtained from the UFSAR Chapter 15 0

events including the reactor core safety limits, axial offset envelopes (AOs), rod withdrawal at 0 power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss of flow 0 accident (LOFA), and locked rotor accident (LOCROT) events to compare the performance of VIPRE-D and LYNXT. These various limits and events provide sensitivity of DNB performance to the following: (a) power level (including the impact of the part-power multiplier on the allowable 0 hot rod power FAH), pressure and temperature (reactor core safety limits); (b) axial power shapes (AOs); (c) elevated hot rod power (misaligned rod); and (d) low flow (LOFA and LOCROT). The 9 173 statepoints cover the full range of conditions and axial offsets in the North Anna UFSAR 0 Chapter 15 evaluations (except for MSLB that is discussed in Section 5.2), and were specifically 0 selected to challenge the three BWU CHF correlations (Table 5.1-1).

0 This benchmark study showed an average deviation between VIPRE-D and LYNXT of less than 0.14% in DNBR, with a maximum deviation of 2.2%. These results are well within the uncertainty typically associated with thermal-hydraulic codes, which has been quantified to be 5% (Reference 0 15), and justify the model selections in Section 4. Figure 5.1-2 shows graphically the performance 0 of VIPRE-D versus LYNXT for the 173 statepoints. The close comparison of VIPRE-D to LYNXT 0 over the full range of conditions expected for UFSAR transients justifies the applications of VIPRE-D to the transients identified in Table 2.1-1 (MSLB will be discussed in Section 5.2).

S Table 5.1-1: Range of VIPRE-D / LYNXT 173 Benchmark Statepoints S

VARIABLE RANGE 0

Pressure [psia] 1860 to 2400 S

Power [% of 2942.2 MWt] 66 to 135 S

Inlet Temperature [OF] 506.6 to 626.2 Flow [% of Minimum Measured Flow] 64 to 100 FAH 1.49 to 1.945 Axial Offset [%] -48.7 to 57.9 DOM-NAF-2, Rev. 0.2-A 25

0 S

0 0

Figure 5.1-1. Typical North Anna VIPRE-D 12-Channel Model for F-ANP AMBW Fuel Assemblies 0

0 Guide Tube 0

Instrument Tube 0 0

Rod 13 0

Channel 11 0 Remainder 0 of hot assembly 0

0 0

9 0

0 0

0 Hot Assembly 0

0 Rod 14 0

Channel 12 0 Remainder of 0 1/ 8 th core 0 0

0 0

0 DOM-NAF-2, Rev. 0.2-A 26 0

0 9 ~~~~~~Figure 5.1-2: VIPRE-D vs. LYNXTfoth17 aepis z 1.20 1.15 1.10 1.05 I,-

x z

, 1.00 0

a-0.95 0.90 0.85 0.80 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 STATEPOINT

0 0

5.2 MAIN STEAM LINE BREAK APPLICATION S The 12-channel model discussed in Section 5.1 does not allow the modeling of the peaking and 0 inlet boundary conditions in the fuel assemblies adjacent to the hot assembly, which is necessary for the analysis of some accidents, such as MSLB. Consequently, a 14-channel model was created to more accurately simulate the behavior of the core during a MSLB event.

The VIPRE-D 14-channel model for a North Anna core containing F-ANP AMBW fuel assemblies consists of 14 channels (10 subchannels and 4 lumped channels) and 16 rods as shown in Figure 5.2-1. The two additional channels provide adequate solution detail of the flow field in the vicinity of the hot assembly and allow the modeling of the peaking and inlet boundary conditions in the fuel assemblies adjacent to the hot assembly.

S The 14-channel model defines the inlet temperature for each one of the 14 channels. In addition, the inlet flow fraction is also specified for each of 14 channels. This modeling choice is of key importance for MSLB events, since the inlet temperature may change for each channel and it is then necessary to adjust the flow fraction to obtain the appropriate values of core inlet flow rate and channel flow rate.

S The results from the VIPRE-D 14-channel model were compared to the results from a F-ANP LYNXT model for high flow (with offsite power) and low flow (without offsite power) MSLB evaluations. The results obtained show a maximum deviation of 2.12% in DNBR. These results demonstrate that VIPRE-D can analyze a MSLB event, provided the model has sufficient detail surrounding the hot assembly, such as the 14-channel model described here.

S In addition, the accuracy of the 14-channel model was demonstrated through comparison with the DNBR results of the 173 statepoints obtained with the VIPRE-D 12-channel model. As discussed in Section 5.1, this set of statepoints is consistent with the list of intended applications of the VIPRE-D code discussed in Section 2.1. (Table 2.1-1). This comparison shows that there is essentially no difference between the 12-channel and the 14-channel models (the average deviation in DNBR is 0.03%), which indicates that VIPRE-D models created following the methodology discussed in Section 4 of this report are adequate.

R 0

S 9

S S

S 0

DOM-NAF-2, Rev. 0.2-A 28 0 S

S

S 0

0 0 Figure 5.2-1. Typical VIPRE-D 14-Channel Model for North Anna Cores with F-ANP AMBW Fuel S

0 Instrument Tube Guide Tube 0

0 Rod 13 S Channel 11 Remainder of the hot 0 Symmetry line assembly 0

0 Guide Tube 0

0 9

0 0

0 12 0

S Hot Assembly 9

0 Rod 15 Channel 13 Rod 16 S 1/2 Bundle Channel 14 0 Remainder of core S 1 / 8 th S

DOM-NAF-2, Rev. 0.2-A 29

0 0

5.3 TRANSIENT APPLICATION VIPRE-D has the capability to perform transient calculations by using boundary conditions 0 obtained from a reactor systems code or a neutronic code. The reactor systems code provides O time-dependent forcing functions for pressure, core average power, core flow rate and core inlet temperature and the neutronics code provides core power distributions and nuclear peaking factors.

VIPRE-D transient capability was tested by performing several sample transient calculations, two O of which are described in this report. These two transient calculations were only intended to be 0 samples designed to exercise the transient capabilities of the VIPRE-D code and a typical 0 VIPRE-D model created according to the guidelines discussed in Section 4. In both cases, the behavior of the VIPRE-D results was successfully compared to the behavior to the COBRA analysis of record in the UFSAR. In addition, the VIPRE-D transient results were benchmarked against the steady state analysis of the most limiting statepoint in the transient. Two statepoints were selected in each case, the statepoint with the highest value of the power to mass flow ratio, and the 0 limiting statepoint determined in the transient calculation (if different). O 0

As discussed in Section 4.12, a numerical instability could occur in transient calculations using a subcooled void model that was developed based on steady state data, such as the EPRI model.

For that reason, in order to avoid numerical instabilities, the time steps used for these transient simulations were selected to ensure that the Courant number is greater than one.

The damping factors and the convergence limits were set to the defaults provided by the code (Section 4.12). In a few occasions, when convergence problems were reported by the code, the damping factors and/or the convergence limits were adjusted in the models to allow the code to converge. These convergence problems do not necessarily mean bad results or false convergence, just some numerical instability. Indeed, in most occasions, the results obtained by the code with the adjusted convergence limits or damping factors were nearly identical to the non-converging results. 0 0

The first sample transient selected to verify the capabilities of the VIPRE-D code and the 12-channel model was the RWAP accident. Forcing functions for the RWAP transient were 0 obtained from a NAPS UFSAR case (Dominion COBRA analysis of record for Westinghouse fuel). The length of the transient was 4.0 seconds, with a 0.05-second timestep. VIPRE-D results show similar behavior to the COBRA analysis of record in the UFSAR, but the MDNBR results are 0 different because the analyses use different fuel types and CHF correlations (see Figure 5.3.1).

Comparison with the results of the steady state calculation of the limiting statepoint show MDNBR 0 values that are essentially the same as the results obtained in the transient.

0 0

DOM-NAF-2, Rev. 0.2-A 30 a 0

0

0 0

0 0 The second sample transient selected to perform this verification was the LOFA. Forcing 0 functions for the LOFA transient were obtained from the NAPS UFSAR. In particular, COBRA forcing functions were obtained for a F-ANP uprated core tripping on reactor coolant pump 0 undervoltage. The length of the transient was 20.4 seconds, with a 0.1 -second timestep. COBRA 0 analysis of record and VIPRE-D calculations exhibited similar behavior, but the MDNBR results 0 are different because the analyses use different fuel types and CHF correlations (see Figure S 5.3.2). Comparison with the results of the steady state calculation of the limiting statepoint show MDNBR values that are essentially the same as the results obtained in the transient.

0 0 The transient analyses demonstrate that VIPRE-D is capable of performing stable transient calculations and the results obtained are adequate. Table 5.3-1 summarizes the results of the 0

transient analysis.

0 0 Table 5.3-1: Summary of VIPRE-D Sample Transients 0 RWAP Sample Transient POWER FLOW TEMPERATURE PRESSURE DNBR 0 [MBtu/hr-ft 2] 2

[Mlbm/hr-ft ] [OF] [psia]

0 INITIAL CNITION 0.20578 2.469 553.7 2250.0 CONDITION 2.847 0

LIMITING 2.597 [transient]

0 CONDITION 0.22290 2.467 553.9 2286.5 2.597 [transie]

[2.75 s] [2.75 s]2.598 [steady state]

0 IIII 0 LOFA Sample Transient 0 POWER FLOW TEMPERATURE PRESSURE DNBR

[Mgtu/hr-ft 2] [Mlbm/hr-ft 2] [OF] [psia]

0 INITIAL CNITION 0.20578 2.469 553.7 S CONDITION 2250.0 2.847 0 LIMITING 1.820 [transient]

CONDITION 0.19726 1.649 552.9 2360.7 1.820 [transie]

0 [9.41.796 [steady state 0

0 0

0 DOM-NAF-2, Rev. 0.2-A 31

Figure 5.3-1: VIPRE-D RWAP Transient Sample Calculation Results 2.9 0

0 2.85 0 0

2.8 0

2.75 0

w z

0 2.7 0

0 2.65 0 0

2.6 0 0

2.55 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 0

TIME (s]

0 Figure 5.3-2: VIPRE-D LOFA Transient Sample Calculation Results 0

14 0

0 12 0

0 10 0 0

8 0

S 0

4. 0 0

2 0 0

0 10 15 20 25 0

TIME [s]

DOM-NAF-2, Rev. 0.2-A 32

0 S

5.4 SENSITIVITY STUDIES 0t VIPRE-D has a number of empirical correlations available to simulate two-phase flow effects (Reference 1). These correlations can be grouped in three major categories: 1) two-phase friction

multipliers; 2) subcooled void correlations; and 3) bulk boiling void correlations. In Reference 4 (Section 3.0), a sensitivity study was performed to assess the differences in the performance of the various correlations and, although significant differences were not found, the EPRI models were chosen as the default models for VIPRE-01. The USNRC staff reviewed these sensitivity

I studies and concluded in the SER for VIPRE-01 MOD-01 (Reference 6) that the EPRI void 0models and the EPRI correlations for two-phase friction are acceptable for licensing calculations.

0 51 Dominion performed another sensitivity study to determine the set of two-phase flow correlations

most suitable for Dominion models. This sensitivity analysis provides justification for Dominion's modeling assumptions as discussed in Section 4.8, thus fulfilling condition (3) of the SER for

I VIPRE-01 MOD-01 (Reference 6). A detailed analysis of the available correlations was 0performed, including the modeling assumptions used in deriving the various correlations and four 0sets of correlations were chosen. The selected sets use together only those correlations that have consistent or complementary bases and take advantage of previous industry experience and vendor recommendations. The four cases studied were:

. Case 1 (EEE)

Subcooled Void Model: EPRI

Bulk Boiling Void Model: EPRI

t Two-Phase Friction Multiplier: EPRI 0

. Case 2 (LSE)

Subcooled Void Model: LEVY Bulk Boiling Void Model: SMITH Two-Phase Friction Multiplier: EPRI 0

. Case 3 (LHH)

I Subcooled Void Model: LEVY

Bulk Boiling Void Model: HOMOGENEOUS

Two-Phase Friction Multiplier: HOMOGENEOUS Case 4 (LSH)

Subcooled Void Model: LEVY

I Bulk Boiling Void Model: SMITH

Two-Phase Friction Multiplier: HOMOGENEOUS 0

0 0

DOM-NAF-2, Rev. 0.2-A 33 0

0

0 0

The 173 statepoints and the typical 12-channel model described in Section 5.1 were executed by 0 VIPRE-D using the four sets of two-phase models and correlations. The results were compared to 0 the results of the USNRC approved code F-ANP LYNXT. Table 5.4-1 lists the average and 0 maximum percent deviations in DNBR between the codes and Figure 5.4-1 shows the same results graphically. The set of EPRI correlations (option EEE), which is the default in the code, was then 0

selected for VIPRE-D models as discussed in Section 4.8.

0 0

Table 5.4-1: Statistical Analysis of the MDNBR Results for the Four Sets of Two-Phase Models 0 0

% DEVIATION IN DNBR 0 LYNXT - VIPRE 0

LYNXT 0 EEE LSE LHH LSH 0 AVERAGE 0.14 1.87 3.21 1.00 0

0 STANDARD 0.89 1.26 1.48 1.28 DEVIATION 0 0

0 0

DOM-NAF-2, Rev. 0.2-A 34

0 Z Figure 5.4-1: VIPRE-D vs. LYNXT for the Four Sets of Two-Phase Models CD 1.2000

!C p~

1.1500 1.1000 I- 1.0500 z

EEE

.J oLSE i3, 1.0000 o LHH X LSH 0.9500 0.9000 0.8500 0.8000 0C1 0 20 40 60 80 100 120 140 160 180 STATEPOINT

)

6.0 CONCLUSION

S S The VIPRE-01 code has been approved by the USNRC and is widely used throughout the nuclear 0 industry for PWR safety analyses. VIPRE-D is the Dominion version of VIPRE-01. Other than the addition of vendor proprietary CHF correlations and minor input/output customizations, VIPRE-D is equivalent to VIPRE-01 as Dominion has preserved all the USNRC approved constitutive models and algorithms in the code. Dominion has shown VIPRE-D compliance with the requirements of the USNRC SERs regarding VIPRE-01 code applications. Dominion has validated VIPRE-D with extensive code benchmark calculations using the modeling methods 0 outlined in this report, and the accuracy of the VIPRE-D models has been demonstrated through 0 comparisons with other NRC-approved methodologies. VIPRE-D has been shown to meet or exceed the same standards for accuracy as other methodologies currently being used by Dominion and approved by the USNRC.

VIPRE-D includes several CHF correlations applicable to various F-ANP and Westinghouse fuel types, and the qualification of each one of them will be documented in the appendixes to this 0 report. The critical heat flux correlation to be used for a particular fuel type will be documented and qualified in one of the appendixes and will have been approved by the USNRC for use with 5 such fuel product prior to use by Dominion. The VIPRE-D CHF correlations will be used within the USNRC approved parameter ranges of the CHF correlations, including fuel assembly geometry and grid spacers. The DNBR design limits applied to each CHF correlation will be derived or verified using fluid conditions predicted by the VIPRE-D code.

S With the modeling methods outlined in this report, and in conjunction with the appropriate CHF correlation and DNBR design limits qualified in the appendixes to this report, Dominion plans to use the VIPRE-D code for:

S

1) Analysis of 14x14, 15x15 and 17x17 fuel in PWR reactors.

S

2) Analysis of DNBR for statistical and deterministic transients in the Updated Final Safety Analysis Report (UFSAR), as identified in Table 2.1-1. Additional DNBR transients that are plant specific may be analyzed in a plant specific application that would be submitted to the USNRC for review and approval.

3) Steady state and transient DNB evaluations.

S

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

R S

S S

DOM-NAF-2, Rev. 0.2-A 360 S

S

C 0

0 5) Providing the basis for reactor protection setpoints.

0

6) 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 this document.

0 0

DOM-NAF-2, Rev. 0.2-A 37

7.0 REFERENCES

0

1. Technical Report, EPRI NP-2511-CCM-A Volume 1, Revision 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 1: Mathematical Modeling," February 2001.

2. Technical Report, EPRI NP-251 1-CCM Volume 2, Revision 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 2: User's Manual," February 2001.

3. Technical Report, EPRI NP-251 1-CCM-A Volume 3, Revision 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 3: Programmer's Manual," February 2001.

4. Technical Report, EPRI NP-251 1-CCM-A Volume 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 4: Applications," April 1987.

5. Technical Report, EPRI NP-251 1-CCM Volume 5, 'VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 5: Guidelines," March 1988.

0

6. 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.

0

7. Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 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.

8. Topical Report, VEP-FRD-33-A, "VEPCO Reactor Core Thermal-Hydraulic Analysis Using the COBRA IlIc/MIT Computer Code," F. W. Sliz and K. L. Basehore, October 1983.

S

9. Technical Report, BAW-10199P-A, "The BWU Critical Heat Flux Correlations," Framatome Cogema Fuels, August 1996, including Addendum 1, December 2000.

10. Technical Report, BAW-10199P-A, Addendum 2, "Application of the BWU-Z CHF Correlation to the Mark-BW17 Fuel Design with Mid-Span Mixing Grids," Framatome Cogema Fuels, June 2002.

S

11. Technical Report, "Nuclear Power Plant Engineering," James H. Rust, Haralson Publishing Company, Atlanta, Georgia, 1979.

R S

S DOM-NAF-2, Rev. 0.2-A 38 0 0

0

S 0

S

12. Technical Report, AEC-TR-6630, "Handbook of Hydraulic Resistance, Coefficients of Local 0

Resistance and of Friction," I. E. Idel'Chik, 1960.

13. Letter from L. N. Hartz (Dominion) to USNRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Changes and Exemption 0 Request, Use of Framatome ANP Advanced Mark-BW Fuel," Serial No.02-167, March 28, 2002.

S 14. Technical Report, BAW-10156-A, Revision 1, "LYNXT, Core Transient Thermal-Hydraulic 0 Program," Framatome ANP, August 1993.

S 15. Technical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," R. C. Anderson, June 1987.

S S

0 0

0 S

0 0

0 S

S 0

DOM-NAF-2, Rev. 0.2-A 39

0 DOM-NAF-2, Rev. 0.2-A 0

APPENDIX A 0

0 0

Qualification of the F-ANP BWU CHF Correlations 0 in the Dominion VIPRE-D Computer Code 0 0

0 0

NUCLEAR ANALYSIS AND FUEL DEPARTMENT 0 DOMINION 0 RICHMOND, VIRGINIA 0 0

0 0

Approved by NRC Safety Evaluation 0 dated April 4, 2006 and corrected June 23, 2006 0 0

0 0

0 S

0

S

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.

0 0

ABSTRACT

This appendix documents Dominion's qualification of the Framatome-ANP (F-ANP)

BWU-N, BWU-Z and BWU-ZM correlations with the VIPRE-D code. This qualification

was performed against the same CHF experimental database used by F-ANP to develop and license the correlations. This appendix summarizes the data evaluations that were performed to qualify the VIPRE-D/BWU code/correlation pair, and to develop the corresponding DNBR design limits for each correlation.

0 R 0

0 0

SDOM-NAF-2, Rev. 0.2-A, APPENDIX A A-2 0

S

0 0

TABLE OF CONTENTS 0 0

CLASSIFICATION/DISCLAIMER A-2 0 ABSTRACT A-2 0 TABLE OF CONTENTS A-3 0 LIST OF TABLES A-4 LIST OF FIGURES A-4 0

ACRONYMS AND ABBREVIATIONS A-5 0 A.1 PURPOSE A-6 0 A.2 DESCRIPTION OF THE F-ANP CHF CORRELATIONS A-6 0 A.3 DESCRIPTION OF CHF EXPERIMENTAL TESTS A-8 0 A.3.1 BWU-Z CORRELATION A-8 A.3.2 BWU-ZM CORRELATION 0

A-9 A.3.3 BWU-N CORRELATION A-9 0

A.4 VIPRE-D RESULTS AND COMPARISON TO LYNXT/LYNX2 A-11 0 A.4.1 VIPRE-D/BWU-Z RESULTS A-12 0 A.4.2 VIPRE-D/BWU-ZM RESULTS A-23 0 A.4.3 VIPRE-D/BWU-N RESULTS A-33 0 A.5 CONCLUSIONS A-43 A.6 REFERENCES A-44 0

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-3 0

LIST OF TABLES

Table A.3.1-1: BWU-Z CHF Experimental Database A-8

Table A.3.2-1: BWU-ZM CHF Experimental Database A-9

Table A.3.3-1: BWU-N CHF Experimental Database A-10 Table A.4.1-1: VIPRE-D/BWU-Z M/P Ratio Results A-13 Table A.4.1-2: VIPRE-D/BWU-Z DNBR Design Limit A-14

Table A.4.1-3: VIPRE-D/BWU-Z DNBR Limits for Pressure Groups A-15 O Table A.4.2-1: VIPRE-D/BWU-ZM M/P Ratio Results A-23 O Table A.4.2-2: VIPRE-D/BWU-ZM DNBR Design Limit A-24

Table A.4.3-1: VIPRE-D/BWU-N M/P Ratio Results A-33

Table A.4.3-2: VIPRE-D/BWU-N DNBR Design Limit A-34 Table A.4.3-3: VIPRE-D/BWU-N DNBR Limits for Pressure Groups A-35 Table A.5-1: VIPRE-D DNBR Limits for BWU-Z, BWU-ZM and BWU-N A-43 O Table A.5-2: Range of validity for BWU-Z, BWU-ZM and BWU-N A-43 O

LIST OF FIGURES Figure A.4. 1-1: Measured vs. Predicted CHF for BWU-Z A-16 O Figure A.4.1-2: M/P vs. Pressure for BWU-Z A-17

Figure A.4.1-3: M/P vs. Quality for BWU-Z A-18

Figure A.4.1-4: M/P vs. Mass Velocity for BWU-Z A-19

Figure A.4.1-5: DNBR vs. Pressure for BWU-Z A-20 Figure A.4.1-6: DNBR vs. Quality for BWU-Z A-21 Figure A.4.1-7: DNBR vs. Mass Velocity for BWU-Z A-22

Figure A.4.2-1: Measured vs. Predicted CHF for BWU-ZM A-26 O Figure A.4.2-2: M/P vs. Pressure for BWU-ZM A-27

Figure A.4.2-3: M/P vs. Quality for BWU-ZM A-28

Figure A.4.2-4: M/P vs. Mass Velocity for BWU-ZM A-29

Figure A.4.2-5: DNBR vs. Pressure for BWU-ZM A-30 Figure A.4.2-6: DNBR vs. Quality for BWU-ZM A-31 Figure A.4.2-7: DNBR vs. Mass Velocity for BWU-ZM A-32

Figure A.4.3-1: Measured vs. Predicted CHF for BWU-N A-36 O Figure A.4.3-2: M/P vs. Pressure for BWU-N A-37 O Figure A.4.3-3: M/P vs. Quality for BWU-N A-38 O Figure A.4.3-4: M/P vs. Mass Velocity for BWU-N A-39 O Figure A.4.3-5: DNBR vs. Pressure for BWU-N A-40 Figure A.4.3-6: DNBR vs. Quality for BWU-N A-41 Figure A.4.3-7: DNBR vs. Mass Velocity for BWU-N A-42 0

SDOM-NAF-2, Rev. 0.2-A, APPENDIX A A-4 0

0 0

S 0

ACRONYMS AND ABBREVIATIONS 0 0

0 0

ARC Alliance Research Center 0 CHF Critical Heat Flux 0 DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio 0

F-ANP Framatome Advanced Nuclear Power 0

HTRF Heat Transfer Research Facility at Columbia University 0 M/P Ratio of Measured-to-Predicted CHF S MSMG Mid-Span Mixing Grid 0 MVG Mixing Vane Grid 0 NMVG Non-Mixing Vane Grid P/M Ratio of Predicted-to-Measured CHF (equivalent to DNBR) 0 PWR Pressurized Water Reactor S USNRC US Nuclear Regulatory Commission 0 0

0 0

0 S

0 0

S 0

0 0

0!

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-5 0 0

S S

S

A.1 PURPOSE Dominion has purchased fuel assemblies from Framatome ANP (F-ANP) for use at North Anna Power Station, Units 1 and 2. These new fuel assemblies are designated as 0Advanced Mark-BW fuel and are a one-for-one replacement for the resident fuel product,

which is the North Anna Improved Fuel with ZIRLO components and PERFORMANCE+

! debris resistant features (a Westinghouse fuel product). The thermal-hydraulic analysis of the F-ANP fuel product requires the use of the F-ANP BWU CHF correlations (References S All and A2).

0To be licensed for use, a critical heat flux (CHF) correlation must be tested against

experimental data that span the anticipated range of conditions over which the correlation

will be applied. Furthermore, the population statistics of the database must be used to

establish a departure from nucleate boiling ratio (DNBR) design limit such that the probability of avoiding departure from nucleate boiling (DNB) will be at least 95% at a 95%

confidence level.

S

This appendix documents Dominion's qualification of the BWU-N, BWU-Z and BWU-ZM

correlations with the VIPRE-D code. This qualification was performed against the same

CHF experimental database used by F-ANP to develop and license the correlations. This appendix summarizes the data evaluations that were performed to qualify the VIPRE-D/BWU code/correlation pair, and to develop the corresponding DNBR design limits for each correlation.

A.2 DESCRIPTION OF THE F-ANP CHF CORRELATIONS In pressurized water reactor (PWR) cores, the energy generated inside the fuel pellets leaves the fuel rods at their surface in the form of heat flux, which is removed by the reactor

coolant system flow. The normal heat transfer regime in this configuration is nucleate boiling, which is very efficient. However, as the capacity of the coolant to accept heat from

the fuel rod surface degrades, a continuous layer of steam (a film) starts to blanket the tube. This heat transfer regime, termed film boiling, is less efficient than nucleate boiling and can result in significant increases of the fuel rod temperature for the same heat flux.

1 Since the increase in temperature may lead to the failure of the fuel rod cladding, PWRs Sare designed to operate in the nucleate boiling regime and protection against operation in

film boiling must be provided.

S

The heat flux at which the steam film starts to form is called CHF or the point of DNB. For design purposes, the DNBR is used as an indicator of the margin to DNB. The DNBR is

the ratio of the predicted CHF to the actual local heat flux under a given set of conditions.

R 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-6 0

0

0 Thus, DNBR is a measure of the thermal margin to film boiling and its associated high temperatures. The greater the DNBR value (above 1.0), the greater the thermal margin.

The CHF cannot be predicted from first principles, so it is empirically correlated as a function of the local thermal-hydraulic conditions, the geometry, and the power distribution measured in the experiments. Since a CHF correlation is an analytical fit to experimental data, it has an associated uncertainty, which is quantified in a DNBR design limit. A calculated DNBR value greater than this design limit provides assurance that there is at least a 95% probability at the 95% confidence level that a departure from nucleate boiling will not occur.

F-ANP has developed and uses the B&W-2, the BWC and the BWCMV CHF correlations.

The first two of these correlations apply to fuel assemblies with non-mixing vane spacer grids of inconel or zircaloy. The BWCMV correlation applies to fuel assemblies with mixing vane grids. These correlations are limited to applications in a high flow regime, but modern applications require the use of a correlation in the middle and low flow regimes. Using the response surface model and sequential optimization techniques, F-ANP developed a universal local conditions CHF correlation form. This correlation form, designated BWU, was modified and applied to three different fuel design types over the wider required ranges in Reference Al. This reference describes the CHF tests that provided the bases for the new correlations, analyzes the performance of the correlation for each fuel type, and provides limits and guidelines for its application.

The F-ANP BWU CHF correlations are defined in Reference Al as:

S FMsM FLS*.0.j

= FS [A.2.1]

QCHF 7

where QCHF is the critical heat flux in Btu/hr-ft2 , FMSM is a dimensionless performance factor dependent on the grid arrangement of the assembly and defined in References Al and A2, FLS is a dimensionless length spacing factor, FTong is the dimensionless non-uniform flux shape factor (Tong factor) and Qunif is the uniform heat flux in Btu/hr-ft2 . The specific formulations for each one of these components, as well as the corresponding constants are F-ANP proprietary and can be found in References Al and A2.

References Al and A2 discuss the application of the BWU correlation form to three different grid types:

0

BWU-N, which is only applicable in the presence of non-mixing vane grids (NMVG).

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-7 S

0

" BWU-Z, which is the enhanced mixing vane correlation, is applicable to the DNB analysis of the fuel assembly in the mixing region.

" BWU-ZM, which is BWU-Z with a multiplicative enhancement factor, is applicable in the presence of mid-span mixing grids (MSMGs).

A.3 DESCRIPTION OF CHF EXPERIMENTAL TESTS A.3.1 BWU-Z CORRELATION F-ANP developed the BWU-Z correlation to be used for fuel designs with mixing spacer grids based on the experimental data obtained at the Heat Transfer Research Facility of Columbia University (HTRF) and with the Mark BW17 spacer grid designs. The HTRF is a ten-megawatt electric facility capable of testing full length (up to 14 ft heated length) rod arrays in up to a 6-by-6 matrix. HTRF testing conditions cover the full range of PWR operating conditions with pressures up to 2,500 psia, mass velocities up to 3.5 Mlbm/hr-ft 2 and inlet temperatures approaching saturation. Seven series of tests were used to develop the BWU-Z CHF correlation (References Al and A4). These same tests were also used by Dominion to qualify the VIPRE-D/BWU-Z code correlation pair. Seven full assembly models were created for VIPRE-D to model these experimental test sections. Table A.3.1-1 summarizes the seven series of tests in the BWU-Z CHF experimental database.

Table A.3.1-1: BWU-Z CHF Experimental Database PIN OD PI D// HEATED GRID AXIAL HEAT GUIDE TUBE HAE GRD NUMBER TYPE MATRIX LENGTH SPACING OFMTES TEST FLUX SHAPE OD OF TESTS

[ichs]

[inches] [inches] (inches]

BW 12.0 Unit Cell 5x 5 1.55 Symmetric 0.374 / - 143.4 20.5 99 BW 13.1 Unit Cell 5x 5 1.55 Symmetric 0.374/ - 143.4 20.5 94 BW 14.1 Guide Tube 5x 5 1.55 Symmetric 0.374 / 0.482 143.4 20.5 76 BW 15.1 Cold Unit 5x 5 1.55 Symmetric 0.374/- 143.4 20.5 92 BW 16.0 Cold Row 5x 5 1.55 Symmetric 0.374 / - 143.4 20.5 48 BW 19.0 Guide Tube 5x 5 1.55 Symmetric 0.374 / 0.482 143.4 20.5 94 BW 20.0 Unit Cell 5 x5 1.55 Symmetric 0.374 / - 143.4 20.5 48 DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-8

0 0

A.3.2 BWU-ZM CORRELATION 0 0

F-ANP developed the BWU-ZM correlation to be used for fuel designs with MSMGs 0 based on the experimental data obtained at the HTRF and with the Mark BW17 spacer 0 grid designs. Three series of tests were used to validate the BWU-ZM CHF correlation 0 (References A2 and A4). These same tests were also used by Dominion to qualify the VIPRE-D/BWU-ZM code correlation pair. Three full assembly models were created for 0 VIPRE-D to model these experimental test sections. Table A.3.2-1 summarizes the three 0 series of tests in the BWU-ZM CHF experimental database. 0 0

Table A.3.2-1: BWU-ZM CHF Experimental Database S

0 0

S 0

0 0

0 A.3.3 BWU-N CORRELATION C

F-ANP developed the BWU-N correlation for fuel designs with NMVGs based on the S experimental data obtained at the heat transfer facility at the Alliance Research Center (ARC) with the Mark C and Mark BZ non-mixing spacer grid designs. This experimental 0

facility was similar in capacity to HTRF, but has since been decommissioned. Seven 0 Mark C tests and 3 Mark BZ tests were used to develop the correlation (References Al 0

and A3). These same tests were also used by Dominion to qualify the VIPRE-D/BWU-N code correlation pair. Ten full assembly models were created for VIPRE-D to model S

these experimental test sections. Table A.3.3-1 summarizes the ten series of tests in the S BWU-N CHF experimental database. 0 0

0 0

0 C

0 0

S DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-9 0D

Table A.3.3-1: BWU-N CHF Experimental Database PIN OD /

HEATED S TEST TYPE MATRIX AXIAL HEAT GUIDE TUBE LENGTH GRID SPACING NUMBER FLUX SHAPE OD OF TESTS

[ichs] [inches] [inches]

0 [inches]

S C-3 Unit Cell 3 x 3a 1.0 Uniform 0.379/- 72.0 21.0 107 0 C-6 Unit Cell 5x5 1.0 Uniform 0.3797/ - 144.0 21.0 130 S

C-7 Guide Tube 5x5 1.0 Uniform 0.379 / 0.465 144.0 21.0 122 0 1.662 Cosine b C-8 Unit Cell 5x 5 0.379/- 144.0 155 0 Symmetric 1.662 Cosine 0 C-9 Guide Tube 5x 5 Symmetric 0.379 / 0.465 144.0 b 85 0 C-11 Unit Cell 5x 5 1.595 Sine 0.379/- 144.0 b 34 S Symmetric 1.595 Sine 0 C-12 Guide Tube 5x 5 Symmetric Symmetric 0.379 / 0.465 144.0 b 133 0 B-15 Guide Tubet 5 5 1.68 Cosine 0.430 / 0.554 144.0 21.1 47 Symmetric S

B-16 Unit Cell 5x 5 16Coie 0.430 /- 144.0 21.1 131 0 Symmetric Intersection 1.68 Cosine 0 B-17 Cell 5 x5 Symmetric 0.430/- 144.0 21.1 157 0

S S

6 S

0 0

0 S

a Bundle C-3 has a heated strip in each of the four walls (1.381" x 72.0").

b Grid centerline distances from the end of the heated length are 15.66", 37.66", 59.41", 80.91",

102.16", 123.16", 143.53".

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-10

S S

0 A.4 VIPRE-D RESULTS AND COMPARISON TO LYNXT/LYNX2 5 References A3 and A4 describe the mathematical model for each separate test section S by providing the bundle and cell geometry, the rod radial peaking values, the rod axial flux shapes, the types, axial locations and form losses associated to the spacer grids, as 5 well as the thermocouple locations. References Al and A2 provide the data for each CHF observation within a test, including power, flow, inlet temperature, pressure and CHF location (rod and axial location).

S Each test section was modeled for analysis with the VIPRE-D thermal-hydraulic computer code as a full assembly model following the modeling methodology discussed in Section 4 in the main body of this report. For each set of bundle data, VIPRE-D produces the local thermal-hydraulic conditions (mass velocity, thermodynamic quality, heat flux, etc) at every axial node along the heated length of the test section. The ratio of measured-to-predicted CHF (M/P) is the variable that is normally used to evaluate the thermal-hydraulic performance of a code/correlation pair. The measured CHF is the local S heat flux at a given location, while the predicted CHF is calculated by the code using the 5 CHF correlation of interest (BWU-Z, BWU-ZM or BWU-N). The ratio of these two values provides the M/P ratio, which is the inverse of the DNB ratio. M/P ratios are frequently used to validate CHF correlations instead of DNB ratios, because their distribution is usually a normal distribution, which simplifies their manipulation and statistical analysis.

The axial location, the hot rod and the hot channel that are used to perform the M/P comparison are important. For each test, the M/P ratio must be evaluated at the axial location where burnout was observed experimentally, as listed in References A3 and A4.

The axial nodalization for the various VIPRE-D models was developed taking into account the actual test location of the thermocouples, as well as the locations of the various spacer grids. The criteria used to select the hot channel and hot rod are supported by engineering judgment and use the information regarding burnout location provided by References A3 and A4. In general, when burnout was observed 0 experimentally in a hot rod, a hot rod and a central (hot) channel were selected to 5 perform the comparison. When the burnout was observed experimentally on a cold rod, a hot rod was still selected because it was considered unphysical to observe burnout in a cold rod earlier than in a hot rod (experimentally, even though a cold rod was reported to experience burnout first, the reality was that several rods saw burnout almost simultaneously and the limitations of the instrumentation and a desire to minimize damage to the test cell, caused the discrepancy). In this case, however, an external channel (cold) was selected to be the hot channel.

S In addition to comparing to the experimental results, the results obtained by VIPRE-D when modeling the Mark BW, Mark C and Mark BZ experiments were benchmarked against the results obtained by F-ANP with the LYNXT/LYNX2 codes (References Al R

S DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-li 5~

S S

0 and A2). This comparison was just a sanity check to verify that there are no suspect

datapoints and that the statepoint conditions were correctly input to the code.

Some of the tests analyzed were discarded prior to their incorporation into the

VIPRE-D/BWU database. Two criteria were used to justify data deletions.

1) If the M/P ratio obtained for a given data point was greater than 3.5 standard deviations from the average, the data point was eliminated. This criterion is

consistent with the methodology used by F-ANP in Reference Al.

2) If any of the local conditions (pressure, mass velocity or thermodynamic quality) was outside the range of applicability of the correlation as given in References Al and

A2, the data point was eliminated. This criterion is also consistent with the

methodology used by F-ANP in Reference Al.

Overall, 23 data points were excluded from the BWU-Z database (F-ANP discarded 21 data points in Reference Al), and 11 were excluded from the BWU-N database (F-ANP eliminated 8 data points in Reference Al). No data points were eliminated from the

BWU-ZM database. The reason the VIPRE-D/BWU database is slightly smaller than the

LYNXT/BWU database is that the local conditions predicted by VIPRE-D for a few test

! data were just barely outside the range of validity of the BWU correlations as given in Reference Al.

This section summarizes the VIPRE-D results and the associated significant statistics. In addition, this section shows a comparison to the results obtained by F-ANP with the

LYNXT/LYNX2 codes as reported in References Al and A2. This section also shows the 0l variation of the M/P ratio with each independent variable to assess if there are any

biases in the data. Finally, it provides the VIPRE-D overall statistics for the seven

BWU-Z tests, the three BWU-ZM tests and the ten BWU-N tests, and generates the

DNBR design limits for the various BWU CHF correlations with VIPRE-D.

00 A.4.1 VIPRE-D/BWU-Z RESULTS

The BWU-Z correlation was developed by F-ANP correlating the CHF experimental results obtained in tests BW 12.0, BW 13.1, BW 14.1, BW 15.1, BW 16.0, BW 19.0 and BW 20.0. Dominion used those same experimental data to develop the VIPRE-D/BWU-Z

DNBR limit. Table A.4.1-1 summarizes the relevant statistics for each test, and

D calculates the aggregate statistics for the entire set of data.

One-sided tolerance theory (Reference A5) is used for the calculation of the VIPRE-

D/BWU-Z DNBR design limit. This theory allows us to calculate a DNBR limit so that, for a DNBR equal to the design limit, DNB will be avoided with 95% probability at a 95%

confidence level.

R 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-i12 0

Table A.4. 1-1: VIPRE-D/BWU-Z M/P Ratio Results NUMBER MIP RATIO M/P RATIO M/P RATIO M/P RATIO OF TESTS AVERAGE STDEV MAX MIN BW 12.0 99 1.0230 0.0848 1.1683 0.7812 BW 13.1 94 0.9907 0.0900 1.1609 0.7669 BW 14.1 76 0.9869 0.0951 1.1538 0.7261 BW 15.1 92 1.0086 0.0917 1.2974 0.7717 BW 16.0 48 0.9475 0.0716 1.0840 0.6980 BW 19.0 94 0.9833 0.0893 1.1693 0.7833 BW 20.0 25 1.0108 0.0971 1.1642 0.8342 BWU-Z 528 0.9950 0.0907 1.2974 0.6980 Because all the statistical techniques used below assume that the original data distribution is normal, it is necessary to verify that the overall distribution for the M/P 0 ratios is a normal distribution. To evaluate if the distribution is normal, the D' normality test was applied (Reference A6). A value of D' equal to 3,430.23 was obtained for the 0 VIPRE-D/BWU-Z database. This D' value is within the range of acceptability for 528 data points with a 95% confidence level (3,387.6 to 3,449.4)c. Thus, it is concluded that the S M/P distribution for the VIPRE-D/BWU-Z database is indeed normal. S Based on the results listed in Table A.4.1-1, the deterministic DNBR design limit can be calculated as:

S 1.0 S DNBRL = [A.4.1.1]

M/P-KNC,P .M/P 0 0

where M/P = average measured-to-predicted CHF ratio 0 GyM/P = standard deviation of the measured-to-predicted CHF ratios of the database 0 KN,C,p = one-sided tolerance factor based on N degrees of freedom, C S confidence level, and P portion of the population protected. This number is taken from Table 1.4.4 of Reference A5. S 0

c From Table 5 in Reference A6 D' Lower Limit (528) [P = 0.025] = 3,310 + (8 / 20) x (3,504 - 3,310) = 3,387.6 D' Upper Limit (528) [P = 0.975] = 3,371 + (8 / 20) x (3,567 - 3,371) = 3,449.4 DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-13

S Then, the DNBR design limit for the VIPRE-D and the BWU-Z correlation can be calculated as described in Table A.4.1-2:

0 Table A.4.1-2: VIPRE-D/BWU-Z DNBR Design Limit 0

VIPRE-D/BWU-Z 0

Number of data n 528 0

Degrees of freedom N =.n 14 513 0

Average M/P M/P 0.9950 0

0 Standard Deviation 0 YM/P 0.0907 Corrected Standard 0 Deviation N= 1MP [(n -1)/N] ' 0.0919 0 Owen Factor K(513,0.95,0.95) 1.7607 0

0 BWU-Z Design limit DNBRL = 1/ (0.9950 - 1.7607

0.0919) 1.2002 S

Figures A.4. 1-1 through A.4.1-4 display the performance of the M/P ratio, and its 0 distributions as a function of the pressure, mass velocity and quality. The objective of 0 these plots is to show that there are no biases in the M/P ratio distribution, and that the performance of the BWU-Z correlation is independent of the three variables of interest.

The plots show a mostly uniform scatter of the data and no obvious trends or slopes.

0 These plots also show that all the tests in the BWU-Z database are within 3.5 standard 0 deviations from the average. Figures A.4.1-5 through A.4.1-7 display the performance of the P/M ratio (i.e. the DNBR) against the major independent variables for the BWU-Z 0 database. These plots also include a DNBR design limit line at 1.20. It can be seen that 0 only 19 data points (3.6% of the database) are above the DNBR design limit, and that 0 these data in excess of the limit are distributed over the variable ranges tested.

In Reference Al, the USNRC argued that the performance of the BWU-Z correlation 0

might be deficient at the extremely low end of the pressure range. For that reason, F-ANP developed individual DNBR design limits for each low pressure group in the database. This approach allows users to use the BWU-Z correlation at low pressures but imposes a higher DNBR limit to ensure that the correlation is used conservatively. Table A.4.1-3 summarizes the VIPRE-D/BWU-Z DNBR limits calculated for the different pressure groups and compares them with the BWU-Z DNBR design limits obtained by F-ANP in Reference Al.

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-14

0 0

Table A.4.1-3: VIPRE-D/BWU-Z DNBR Limits for Pressure Groups 0

400 psia 700 psia 1000 psia 1500 - 2400 psia 0

1.0623 0.9883 0

AVERAGE M/P 0.8504 1.0452 0

STDEV 0.0121 0.0879 0.0787 0.0883 0

DATA 4 20 40 464 9

K(N,0.95,0.95) 6.882 2.396 2.125 1.768 VIPRE-D 0 DNBR 1.304 DNBR LIMIT 1.198 1.117 1.202 S LYNXTDNBR 1.590 1.199 1.125 1.193 0 LIMIT 0 0

Dominion will take the VIPRE-D/BWU-Z DNBR limit to be 1.20 for pressures greater 0 than or equal to 700 psia, and 1.59 at pressures lower than 700 psia. Since the VIPRE-D/BWU-Z database at 400 psia only has four datapoints, Dominion has used the S F-ANP more conservative DNBR limit of 1.59.

0 0

0 S

0 0

S 0

0 S

A-15 DOM-NAF-2, Rev. 0.2-A, APPENDIX A 0

ee....O9@@@.eO*@@SOO*S*O@@@**Oe@**@9ee@@@@e 0

0 Figure A.4.1-1: Measured vs. Predicted CHF for BWU-Z z

"-n 1.20 CD 0 1.10 1.00 m

z 0.90 X

" 0.80 S0.70 0.60 L-v 0.50

____. 4 .1:~--~-~* St.. .- -___ __

.4 (LD0.40 a.

0.30 0.20 0.10 0.00 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 1.10 1.20 Measured CHF (MBtu/hr-ft**2)

0 Figure A.4.1-2: M/P vs. Pressure for BWU-Z 0

"11 1.50 CD 5

1.40 1.30 m

z X

1.20 1.10 C.

1.00 0.90 0.80 0.70 0.60 0.50 0 500 1000 1500 2000 2500 3000

-4 Pressure (psia)

0 Figure A.4.1-3: M/P vs. Quality for BWU-Z T

1.50 CD 1.40 1.30 m

z 0

1.20 1.10 C-1.00 0.90 0.80 0.70 0.60 0.50

-0.20 -0.10 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0>

Quality

0 Figure A.4.1-4: M/P vs. Mass Velocity for BWU-Z z0 1.50 1.40 1.30 m

z X

1.20 1.10 a.

1.00 0.90 0.80 0.70 0.60 0.50 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 Mass Velocity (Mlbm/hr-ft**2)

Figure A.4.1-5: DNBR vs. Pressure for BWU-Z z

,-n 2.00 CD 1.90 0

O 1.80 1.70 1.60 m

z 1.50 X

1.40 1.30 1.20 1.10 z

1.00 IL 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0 500 1000 1500 2000 2500 3000 0

,0 Pressure [psia]

0 Figure A.4.1-6: DNBR vs. Quality for BWU-Z 1-1" 2.00 CD 1.90 p0 1.80 1.70

-u 1.60 m

z 1.50 X

1.40 1.30 -

1.20 1.10 z

0 1.00 -

II 0.90 -

0.80 -

0.70 -

0.60 -

0.50 0.40 -

0.30 -

0.20 -

0.10 0.00 -

-0.20 -0.10 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 Quality

0 Figure A.4.1-7: DNBR vs. Mass Velocity for BWU-Z 2.00 z

1.90 m 1.80 1.70 1.60 z 1.50 X

1.40 1.30 1.20 1.10 z

1.00 a.

0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 Mass Velocity (Mlbm/hr-ft**2)

0 0

0 A.4.2 VIPRE-D/BWU-ZM RESULTS 0

The BWU-ZM correlation was developed by F-ANP correlating the CHF experimental results obtained in tests BW 12.0, BW 13.1, BW 14.1, BW 15.1, BW 16.0, BW 19.0 and 0 BW 20.0. F-ANP used the experimental data obtained in tests BW 18.0, BW 18.1 and S BW 43.0 to determine FMSM and to calculate the DNBR limit for the BWU-ZM correlation (Reference A2). 0 S

Dominion has used those same experimental data to determine the VIPRE-D/BWU-ZM 0 DNBR limit. Table A.4.2-1 summarizes the relevant statistics for each test, and 0 calculates the aggregate statistics for the entire set of data.

S One-sided tolerance theory (Reference A5) is used for the calculation of the VIPRE-0 D/BWU-ZM DNBR design limit. This theory allows us to calculate a DNBR limit so that, 0 for a DNBR equal to the design limit, DNB will be avoided with 95% probability at a 95% 0 confidence level. 0 S

Table A.4.2-1: VIPRE-D/BWU-ZM M/P Ratio Results 0

0 NUMBER M/P RATIO M/P RATIO M/P RATIO M/P RATIO OF TESTS AVERAGE STDEV MAX MIN 0 0

BW 18.0 18 0.9931 0.1136 1.1467 0.8334 BW 18.1 58 1.0322 0.0945 1.2299 0.8142 0

BW 43.0 72 1.0041 0.0715 1.1747 0.7793 S

BWU-ZM 148 1.0138 0.0875 1.2299 0.7793 0 0

Because all the statistical techniques used below assume that the original data 0 distribution is normal, it is necessary to verify that the overall distribution for the M/P 0 ratios is a normal distribution. To evaluate if the distribution is normal, the D' normality S test was applied (Reference A6). A value of D' equal to 510.55, was obtained for the VIPRE-D/BWU-ZM database. This D' value is within the range of acceptability for 148 0

data points with a 95% confidence level (497.82 to 5 1 5 .04 )d. Thus, it is concluded that 0 the M/P distribution for BWU-ZM is indeed normal. S 0

Based on the results listed in table A.4.2-1, the deterministic DNBR design limit can be 0 calculated as: 0 d From Table 5 in Reference A6 0 D' Lower Limit (148) [P = 0.025] = 456.9 + (8 / 20) x (559.2 - 456.9) = 497.82 D' Upper Limit (148) [P = 0.975] = 473.2 + (8 / 20) x (577.8 - 473.2) = 515.04 DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-23 0

S 9

1.0 DNBRL= [A.4.2.1]

9 M/P-KN,C,P .'M/P 0 where M/P = average measured to predicted CHF ratio 0 CTM/P standard deviation of the measured to predicted CHF ratios of the 0 database 9 KN,C,P = one-sided tolerance factor based on N degrees of freedom, C confidence level, and P portion of the population protected. This number is taken 0 from Table 1.4.3 of Reference A5.

0 Then, the DNBR design limit for the VIPRE-D and the BWU-ZM correlation can be 0 calculated as described in Table A.4.2-2:

0 Table A.4.2-2: VIPRE-D/BWU-ZM DNBR Design Limit 0

0 VIPRE-DIBWU-ZM 0

FNumber of data n 148 0 Degrees of freedom N =n -1 147 Average M/P M/P 1.0138 0 Standard Deviation CVM/P 0.0875 Owen Factor K(147,0.95,0.95) 1.872 0 BWU-ZM Design limit DNBRL = 1 / (1.0138 - 1.872 - 0.0875) 1.1765 9

Figures A.4.2-1 through A.4.2-4 display the performance of the M/P ratio, and its distributions as a function of the pressure, mass velocity and quality. The objective of these plots is to show that there are no biases in the M/P ratio distribution, and that the performance of the BWU-ZM correlation is independent of the three variables of interest.

The plots show a mostly uniform scatter of the data and no obvious trends or slopes.

Figures A.4.2-5 through A.4.2-7 display the performance of the P/M ratio (i.e. the DNBR) against the major independent variables for the BWU-ZM database. These plots also include a DNBR design limit line at 1.18. It can be seen that only 4 data points (2.7% of the database) are above the DNBR design limit, and that these data in excess of the limit are distributed over the variable ranges tested.

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-24

0 9

0 For the BWU-ZM database, no individual DNBR design limits were calculated for the low 0 pressure data. However, in order to extend the validity of the BWU-ZM CHF correlation 0 over the same range as the BWU-Z CHF correlation, the VIPRE-D/BWU-ZM DNBR 0 design limit at pressures less than 594 psia was set to 1.59 (The same as for BWU-Z at low pressures). The DNBR design limit for VIPRE-D/BWU-ZM for pressures equal to or 0

greater than 594 psia is 1.18. S 0

0 0

9 0

0 0

S 0

0 S

0 0

S 0

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-25 0

o Figure A.4.2-1: Measured vs. Predicted CHF for BWU-ZM 1.20 z 1.10 -

0I 1.00 0.90 i 0.80 E0.70 "0.60 L-0.50

.2 2 0.40 0.30 0.20 0.10 0.00 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 1.10 1.20 Measured CHF (MBtu/hr-ft**2)

0 Figure A.4.2-2: M/P vs. Pressure for BWU-ZM 0

"1-1.50 CD 0

1.40 1.30 m

z X

1.20 1.10

- 1.00 0.90 0.80 0.70 0.60 0.50 0 500 1000 1500 2000 2500 3000

-4 Pressure [psia]

0 Figure A.4.2-3: M/P vs. Quality for BWU-ZM z

"11 NG 1.50 CD o 1.40 "3 1.30

> 1.20 _,____

1.10

0.90*

0.80 - - -*---

0.70 0.60 0.50

-0.10 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 SQuality co

Figure A.4.2-4: MIP vs. Mass Velocity for BWU-ZM 1.50 1.40 1.30 1.20 1.10 a.

1.00 0.90 0.80 0.70 0.60 0.50 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 Mass Velocity (Mlbm/hr-ft**2)

0 Figure A.4.2-5: DNBR vs. Pressure for BWU-ZM 2.00 1.90 z 1.80 Xn 1.70 1.60 1.50 1.40 1.30 1.20 6 w

1.10 z

1.00 II 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 500 1000 1500 2000 2500 3000 00 C)

Pressure [psia]

Figure A.4.2-6: DNBR vs. Quality for BWU-ZM z

-13 2.00 C0 0

1.90 5

1.80 p

CD 1.70 1.60 m

z 1.50 9

X>

1.40 1.30 1.20 1.10 -

z 1.00 a.

0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00

-0.10 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80

1' Quality 910 Ole I** Ilb o **** 010000 0 0 0900 0004000 000000,*0*00 00*0

0o Figure A.4.2-7: DNBR vs. Mass Velocity for BWU-ZM Z

T 2.00 CD 1.90 1.80 1.70

-U 1.60 m

z 1.50 0

1.40 1.30 1.20 1.10 z

1.00 a.

0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 Mass Velocity (Mlbm/hr-ft**2)

A.4.3 VIPRE-D/BWU-N RESULTS The BWU-N correlation was developed by F-ANP correlating the CHF experimental results obtained in the ARC tests C-3, C-6, C-7, C-8, C-9, C-1 1, C-1 2, B-15, B-1 6 and B-

17. Dominion has used those same experimental data to determine the VIPRE-D/BWU-N DNBR limit. Table A.4.3-1 summarizes the relevant statistics for each test, and calculates the aggregate statistics for the entire set of data.

Table A.4.3-1: VIPRE-D/BWU-N M/P Ratio Results 0

S NUMBER M/P RATIO M/P RATIO M/P RATIO M/P RATIO OF TESTS AVERAGE STDEV MAX MIN 0 0

C-3 107 1.0655 0.1128 1.3251 0.7501 0 C-6 128 0.9445 0.1188 1.2966 0.6635 0 C-7 120 0.9757 0.0942 1.1553 0.6707 S C-8 155 1.0076 0.0816 1.2127 0.7396 0 C-9 85 1.0373 0.0605 1.1681 0.8934 0 C-ll 34 0.9986 0.0862 1.1389 0.8041 0

C-12 133 1.0083 0.0881 1.2003 0.7346 0

B-15 47 0.9806 0.0971 1.1263 0.7438 0

B-16 129 1.0052 0.1219 1.2627 0.6985 B-17 152 0.9988 0.1004 1.3507 0.8002 0 BWU-N 1090 1.0018 0.1038 1.3507 0.6635 0 0

One-sided tolerance theory (Reference A5) is used for the calculation of the 0 VIPRE-D/BWU-N DNBR design limit. This theory allows us to calculate a DNBR limit so that, for a DNBR equal to the design limit, DNB will be avoided with 95% probability at a 0

95% confidence level. 0 Because all the statistical techniques used below assume that the original data 0

distribution is normal, it is necessary to verify that the overall distribution for the M/P 0 ratios is a normal distribution. To evaluate if the distribution is normal, the D' normality test was applied (Reference A6). A value of D' equal to 9,963.21 was obtained for the S

VIPRE-D/BWU-N database. This D' value is not within the range of acceptability for 0 1090 data points with a 95% confidence level (10,082.0 to 10,210.60)e. Since the value S of D' is less than the lower critical value, the BWU-N distribution has greater kurtosis 0

e From Table 5 in Reference A6 S D' Lower Limit (1090) [P = 0.025] = 9,530 + (40 / 50) x (10,220 - 9,530) = 10,082.0 D' Upper Limit (1090) [P = 0.975] = 9,653 + (40 / 50) x (10,350 - 9,653) = 10,210.6 0 S

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-33

S 0

0 than a normal distribution. Therefore, the one-sided theory is conservative for VIPRE-D/BWU-N. This behavior was also observed by F-ANP in Reference Al.

0 Based on the results listed in Table A.4.3-1, the DNBR limit can be calculated as:

1.0 DNBRL = [A.4.3.1]

0 M/P - KN,C,P .M/P 0

0 where M/P = average measured to predicted ratio 0

CTM/P = standard deviation of the measured to predicted ratios of the database 0

KN,C,P = one-sided tolerance factor based on N degrees of freedom, C 0 confidence level, and P portion of the population protected. This number is taken 0 from Table 1.4.4 of Reference A5.

0 0 Then, the DNBR design limit for the VIPRE-D/BWU-N code/correlation pair can be 0 calculated as described in Table A.4.3-2:

0 Table A.4.3-2: VIPRE-D/BWU-N DNBR Design Limit 0

VIPRE-D/BWU-N 0

0 Number of data n 1090 Degrees of freedom N n 14 1075 0 Average M/P M/P 1.0018 0 Standard Deviation aTM/P 0.1038 0 Corrected Standard Deviation NMIP" [(n-i)! N ] 0.1045 0

Owen Factor K(1075,0.95,0.95) 1.7239 0

S BWU-N Design limit DNBRL = 1 / (1.0018 - 1.7239

0.1045) 1.2170 0

Figures A.4.3-1 through A.4.3-4 display the performance of the M/P ratio, and its distributions as a function of the pressure, mass velocity and quality. The objective of these plots is to show that there are no biases in the M/P ratio distribution, and that the performance of the BWU-N correlation is independent of the three variables of interest.

The plots show a mostly uniform scatter of the data and no obvious trends or slopes.

Figures A.4.3-5 through A.4.3-7 display the performance of the P/M ratio (i.e. the DNBR)

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-34

0 0

0 against the major independent variables for the BWU-N database. These plots also 0 include a DNBR design limit line at 1.22. It can be seen that only 65 data points are 0 above the DNBR design limit, and that these data in excess of the limit are distributed 0 over the variable ranges tested.

0 In Reference Al, the USNRC argued that the performance of the BWU-N correlation 0 might be deficient at the extremely low end of the pressure range. For that reason, 0 F-ANP developed individual DNBR design limits for each low pressure group in the 0 database. This approach allows users to use the BWU-N correlation at low pressures 0 but imposes a higher DNBR limit to ensure that the correlation is used conservatively.

0 Table A.4.3-3 summarizes the VIPRE-D/BWU-N DNBR limits calculated for the different pressure groups and compares them with the DNBR design limits obtained by F-ANP in 0

Reference Al. 0 0

Table A.4.3-3: VIPRE-D/BWU-N DNBR Limits for Pressure Groups 0 0

800 psia 1200 psia 1500 - 2616 psia 0 AVERAGE MIP 1.0019 1.0598 1.0007 0 STDEV 0.1186 0.0865 0.1036 0 N, # DATA 20 20 1050 0 K(N,0.95,0.95) 2.396 2.396 1.7249 0 VI DNBRPRE-D 0 1.393 1.173 1.217 DNBR LIMIT LDNBR YNX2 0

DNBR LIMIT 1.387 1.290 1.207 0 0

Dominion will take the VIPRE-D/BWU-N DNBR limit to be 1.22 for pressures equal to or 0 greater than 1200 psia, and 1.39 at pressures less than 1200 psia. 0 0

0 0

9 0

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-35 0

Figure A.4.3-1: Measured vs. Predicted CHF for BWU-N 1.60 CD 5

1.40 m

z 1.20 1.00 0.80 L-I

. 0.60 a.

0.40 0.20 0.00 0.00 0.20 0.40 0.60 0.80 1.00 1.20 1.40 1.60 Measured CHF (MBtu/hr-ft**2)

0 Figure A.4.3-2: M/P vs. Pressure for BWU-N 0

1.50 CD 5

1.40 1.30 m

z 0

1.20 1.10 0.

1.00 0.90 0.80 0.70 0.60 0.50 0 500 1000 1500 2000 2500 3000 4>

Pressure [psia]

3 Figure A.4.3-3: M/P vs. Quality for BWU-N 1.500 m 1.400 N,

XD 1.300 1.200 1.100 CL 1.000 0.900 0.800 0.700 0.600 0.500

-0.60 -0.40 -0.20 0.00 0.20 0.40 0.60 0.80 Quality

Figure A.4.3-4: M/P vs. Mass Velocity for BWU-N 1.50 1.40 1.30 1.20 1.10 --

a-1.00 0.90 0.80 0.70 0.60 0.50 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50 Mass Velocity (Mlbm/hr-ft**2)

OeOO@@@OOO@@OeO@@OOeOOOO.@........@...OO.O@

0 Figure A.4.3-5: DNBR vs. Pressure for BWU-N 2.00 z

1.90 1.80 1.70 1.60 m

z 1.50 X

1.40 1.30 1.20 1.10 z

II 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00

>0 500 1000 1500 2000 2500 3000 0 Pressure [psia]

U 0 Figure A.4.3-6: DNBR vs. Quality for BWU-N 2.00 CD 1.90 1.80 1.70 1.60 m

z 1.50 X

1.40 1.30 -- E 1.20 1.10 -

z 1.00 -

0-0.90 -

0.80 -

0.70 -

0.60 -

0.50 -

0.40 0.30 0.20 0.10 0.00

-0.60 -0.40 -0.20 0.00 0.20 0.40 0.60 0.80 Quality

Figure A.4.3-7: DNBR vs. Mass Velocity for BWU-N 2.00 1.90 m 1.80 X>

1.70 1.60 z 1.50 1.40 1.30 1.20 -'

1.10 z

1.00 a.

0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 -

0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50 Mass Velocity (Mlbm/hr-ft**2)

0 0

A.5 CONCLUSIONS 0 0

The BWU-Z, BWU-ZM and BWU-N correlations have been qualified with Dominion's 0

VIPRE-D computer code. Table A.5-1 summarizes the DNBR design limits for VIPRE-D/BWU-Z, VIPRE-D/BWU-ZM and VIPRE-D/BWU-N that yield a 95% non-DNB 0

probability at a 95% confidence level. 0 0

Table A.5-2 summarizes the applicability and the ranges of validity for all three CHF S correlations, which are the same as those reported by F-ANP in References Al and A2.

Table A.5-1: VIPRE-D DNBR Limits for BWU-Z, BWU-ZM and BWU-N S

VIPRE-D/BWU-Z 0

DNBR limit below 700 psia 1.59 0

DNBR limit 700 - 2,400 psia 1.20 VIPRE-D/BWU-ZM 0 DNBR limit below 594 psia 1.59 0 DNBR limit at or above 594 psia 1.18 0

VIPRE-D/BWU-N 0 DNBR limit below 1200 psia 1.39 0 DNBR limit at or above 1200 psia 1.22 0

0 Table A.5-2: Range of validity for BWU-Z, BWU-ZM and BWU-N 0 0

BWU-Z BWU-ZM BWU-N Pressure Psiae

[psia]

400 to 2,465 400 to 2,465 788 to 2,616 0 Mass Velocity 0.36 to 3.55 0.47 to 3.55 0.25 to 3.83 0

[Ml bmlh r-ft2]

Thermodynamic Less than 0.74 Less than 0.68 Less than 0.70 Quality at CHF Mid-Span Mixing Non-Mixing Vane Mixing Vane Grids GidS Grids Applicability Grids Grids DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-43

S S

0 0 A.6 REFERENCES Al. Technical Report, BAW-10199P-A, "The BWU Critical Heat Flux Correlations,"

Framatome Cogema Fuels, August 1996, including Addendum 1, December 2000.

S 0 A2. Technical Report, BAW-10199P-A, Addendum 2, "Application of the BWU-Z CHF 0 Correlation to the Mark-BW17 Fuel Design with Mid-Span Mixing Grids,"

0 Framatome Cogema Fuels, June 2002.

0 A3. Technical Report, BAW-10143P-A, "BWC Correlation of Critical Heat Flux,"

S Babcock & Wilcox, April 1985.

S A4. Technical Report, BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design," Framatome Technologies, January 1996.

0 A5. Technical Report, "Tables for Normal Tolerance Limits, Sampling Plans, and S Screening," R. E. Odeh and D. B. Owen, 1980.

0 0 A6. Technical Report, "Assessment of the Assumption of Normality (employing individual observed values)," American National Standards Institute, ANSI 0

N15.15.1974.

0 0

S 0

0 0

0 DOM-NAF-2, Rev. 0.2-A, APPENDIX A A-44

0 0

0 DOM-NAF-2, Rev. 0.2-A APPENDIX B 0

0 0

S Qualification of the Westinghouse WRB-1 CHF 0 S

Correlations in the Dominion VIPRE-D 0 Computer Code S NUCLEAR ANALYSIS AND FUEL DEPARTMENT DOMINION RICHMOND, VIRGINIA 0

0 S

0 Approved by NRC Safety Evaluation S dated April 4, 2006 and corrected June 23, 2006 S S

and Supplemental Approval by NRC Safety Evaluation S dated June 21, 2010 S 0

0 S

S S

S 0

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 0contract, tort, warranty, or strict or absolute liability), for any property damage, mental or 0physical 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.

0 0

S

ABSTRACT This Appendix documents Dominion's qualification of the Westinghouse WRB-1 CHF Scorrelation with the VIPRE-D code. This qualification was performed against the same CHF

experimental database used by Dominion to qualify the COBRA/VRB-1 code/correlation

pair. This Appendix summarizes the data evaluations that were performed to qualify the

VIPRE-DNVRB-1 code/correlation pair and to develop the corresponding DNBR design limit for the correlation.

S R S

S 0

S S

t*DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-2 S

S

0 TABLE OF CONTENTS C LA SS IFIC A T IO N/D IS C LA IME R ....................................................................................................... B-2 0

A BST RA C T ......................................................................................................................................... B -2 TA BLE O F C O NTE NTS ..................................................................................................................... B-3 LIST O F TA B LE S ............................................................................................................................... B-4 LIST O F F IGURE S ............................................................................................................................. B-4 0

A C R O NY MS A ND A BBR EV IATIO NS ............................................................................................... B-5 B .1 P UR P O S E ................................................................................................................................. B-6 B .2 A P P LIC A BILITY ........................................................................................................................ B-6 B.3 DESCRIPTION OF THE WESTINGHOUSE WRB-1 CHF CORRELATION ........................... B-7 0 B.4 DESCRIPTION OF VIPRE-DIWRB-1 DATABASE .................................................................. B-8 0 B.5 VIPRE-D/WRB-1 TEST ASSEMBLIES ..................................................................................... B-9 B .5 .1 4x4 G EO METRY T ESTS .......................................................................................................... B -9 B .5 .2 5x5 G EO METRY T ESTS ........................................................................................................ B-10 0 B.6 VIPRE-D RESULTS AND COMPARISON TO COBRA ......................................................... B-1i1 B.7 BENCHMARK OF THE VIPRE-D/WRB-1 SUBCHANNEL MODEL ...................................... B-22 0 B.7.1 STEA DY STATE A PPLIC ATIO N ...................................................................................... B-22 B.7.2 MAIN STEAM LINE BREAK APPLICATION .................................................................... B-26 B .7.3 T RA NS IENT A PP LIC A TIO N ............................................................................................ B-26 B .8 C O NC LUS IO NS ...................................................................................................................... B-2 8 B .9 R E F E R E NC E S ........................................................................................................................ B-2 9 R

0 0

0 S

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-3 0 0

0

S

LIST OF TABLES

Table B.5.1-1: 4x4 VIPRE-D/WRB-1 Experimental Database B-9 Table B.5.2-1: 5x5 VIPRE-DNIRB-1 Experimental Database B-1 0 Table B.6-1 VIPRE-DNWRB-1 M/P Ratio Results B-12

Table B.6-2: VIPRE-DNVRB-1 DNBR Design Limit B-1 3

Table B.7.1-1: Range of VIPRE-D / COBRA 164 Benchmark Statepoints B-23 Table B.8-1: DNBR Limits for WRB-1 B-28 Table B.8-2: Range of Validity for VIPRE-D/WRB-1 B-28 0

0 0

LIST OF FIGURES Figure B.6-1: Measured vs. Predicted CHF for VIPRE-D/WRB-1 Database B-15 Figure B.6-2: M/P vs. Pressure for VIPRE-D/WRB-1 Database B-16 Figure B.6-3: M/P vs. Mass Velocity for VIPRE-DNIRB-1 Database B-17 Figure B.6-4: M/P vs. Quality for VIPRE-D/WRB-1 Database B-18 Figure B.6-5: DNBR vs. Pressure for VIPRE-DNVRB-1 Database B-19 Figure B.6-6: DNBR vs. Mass Velocity for VIPRE-DNWRB-1 Database B-20 Figure B.6-7: DNBR vs. Quality for VIPRE-D/WRB-1 Database B-21 Figure B.7.1-1: Typical Surry VIPRE-D 19-Channel Model for Westinghouse 15x15 SIF Fuel Assemblies B-24 Figure B.7.1-2: VIPRE-DIWRB-1 vs. COBRANWRB-1 for the 164 Analyzed Statepoints B-25 Figure B.7.3-1: VIPRE-D FWMAL Transient Sample Calculation Results B-27

Figure B.7.3-2: VIPRE-D LOCROT Transient Sample Calculation Results B-27 R

0 S

0 0

S 0

0 S

5DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-4 S

0

S 0

0 ACRONYMS AND ABBREVIATIONS 0 0

AO's Axial Offset Envelope 0

CHF Critical Heat Flux 0 DNB Departure from Nucleate Boiling 0 DNBR Departure from Nucleate Boiling Ratio FLC Form Loss Coefficient 0

FWMAL Feedwater Malfunction Transient 0 HTRF Heat Transfer Research Facility at Columbia University 0 LOCROT LOFA Locked Rotor Accident Loss of Flow Accident S

M/P Ratio of Measured-to-Predicted CHF 0 MSLB Main Steam Line Break 0 MVG Mixing Vane Grid 0 NMVG Non-Mixing Vane Grid P/M Ratio of Predicted-to-Measured CHF (equivalent to DNBR)

S PWR Pressurized Water Reactor 0 RWAP Rod Withdrawal at Power 0 RWSC Rod Withdrawal from Subcritical SIF Surry Improved Fuel 0

SPS Surry Power Station S USNRC US Nuclear Regulatory Commission 0 0

0 S

S 0

0 0

0 S

S S

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-5 0

B.1 PURPOSE

Dominion currently uses Westinghouse 15x1 5 OFA fuel assemblies at Surry Power Station, Units 1 and 2. This fuel product as implemented at Surry is also known as Surry Improved Fuel (SIF). The thermal-hydraulic analysis of this Westinghouse fuel product requires the use of the Westinghouse

WRB-1 CHF Correlation (References B1 and 83). In fact, Westinghouse WRB-1 CHF correlation has

been approved by the USNRC for use with Westinghouse 15x1 5 and 17x1 7 "R" grid type fuel, and

with Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel products (Reference B13).

0To be licensed for use, a critical heat flux (CHF) correlation must be tested against experimental data

that span the anticipated range of conditions over which the correlation will be applied. Furthermore,

the population statistics of the database must be used to establish a departure from nucleate boiling ratio (DNBR) design limit such that the probability of avoiding departure from nucleate boiling (DNB) will be at least 95% at a 95% confidence level.

This Appendix documents Dominion's qualification of the WRB-1 correlation with the VIPRE-D code.

1 This qualification was performed against a subset of the data from the Columbia-EPRI CHF database for Westinghouse "R" grid 17x17 and 15x15 fuel (Reference 82). This is the same subset

of the Columbia-EPRI CHF database used by Dominion in the qualification of the WRB-1 correlation

with the COBRA code (Reference B3). This Appendix summarizes the data evaluations that were

performed to qualify the VIPRE-DNVRB-1 code/correlation pair, and to develop the corresponding DNBR design limits for the correlation.

B.2 APPLICABILITY S Dominion intends to use the VIPRE-DMNRB-1 code/correlation pair for the analysis of Westinghouse 15x15 and 17x17 "R" grid type fuel, and Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel

1 products in PWR reactors. When evaluating these types of fuels outside of the range of validity of the

WRB-1 CHF correlation, Dominion intends to use the VIPRE-D/W-3 code/correlation pair. W-3 is one of the CHF correlations contained in the USNRC approved generic version of VIPRE-01 (References

B7 and 88).

0

The intended VIPRE-D/WRB-1 applications discussed in this Appendix are consistent with the generic intended applications listed in the main body of this report (Section 2.0). Also, more specifically, Dominion intends to use VIPRE-D/WRB-1 to analyze the transients delineated in Table

2.1-1 in Section 2.0 of the main body of this report.

0 The qualification of the WRB-1 CHF correlation with the VIPRE-D code has been performed following the modeling guidelines described in Section 4.0 of this report. In addition, extensive code

benchmark calculations have confirmed that the VIPRE-D models specified in sections 4.1 through

4.12 in the main body of this report produce essentially the same results as equivalent Dominion

COBRA models. Some of these benchmarks are described in section B.7 of this Appendix.

This Appendix is submitted to the USNRC for review and approval in order to meet the USNRC's

requirement #2 listed in the VIPRE-01 SER, as outlined in Section 2.2 in the main body of this report.

0 S

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-6 0

0

S 0

B.3 DESCRIPTION OF THE WESTINGHOUSE WRB-1 CHF 0 CORRELATION S In pressurized water reactor (PWR) cores, the energy generated inside the fuel pellets leaves the fuel rods at their surface in the form of heat flux, which is removed by the reactor coolant system flow. The normal heat transfer regime in this configuration is nucleate boiling, which is very efficient.

However, as the capacity of the coolant to accept heat from the fuel rod surface degrades, a continuous layer of steam (a film) starts to blanket the tube. This heat transfer regime, termed film boiling, is less efficient than nucleate boiling and can result in significant increases of the fuel rod temperature for the same heat flux. Since the increase in temperature may lead to the failure of the fuel rod cladding, PWRs are designed to operate in the nucleate boiling regime and protection against operation in film boiling must be provided.

The heat flux at which the steam film starts to form is called CHF or the point of DNB. For design purposes, the DNBR is used as an indicator of the margin to DNB. The DNBR is the ratio of the predicted CHF to the actual local heat flux under a given set of conditions. Thus, DNBR is a measure of the thermal margin to film boiling and its associated high temperatures. The greater the DNBR value (above 1.0), the greater the thermal margin.

The CHF cannot be predicted from first principles, so it is empirically correlated as a function of the local thermal-hydraulic conditions, the geometry, and the power distribution measured in the experiments. Since a CHF correlation is an analytical fit to experimental data, it has an associated 0 uncertainty, which is quantified in a DNBR design limit. A calculated DNBR value greater than this design limit provides assurance that there is at least a 95% probability at the 95% confidence level that a departure from nucleate boiling will not occur.

The Westinghouse WRB-1 CHF correlation is defined in Reference B1 as: 5 0

QCHF = PF + A] + B3 (G1 0 c) B 4 (GL°C) XLOC [B.3.1] 0 106 106 106 where QCHF is the critical heat flux in Btu/hr-ft2 , PF is a dimensionless performance factor dependent S on the outer diameter of the rods and defined in Reference B1, GLOC is the local mass velocity in Mlbm/ft2-hr, and XLOC is the local quality. The specific formulations for each one of these components, as well as the corresponding constants, are Westinghouse proprietary and can be found in Reference B1. Reference B1 discusses the application of the WRB-1 correlation form to the "L"and "R" grid fuel assembly designs. Westinghouse WRB-1 CHF correlation has been approved S by the USNRC for use with Westinghouse 15x15 and 17x17 "R" grid type fuel, and with 5 Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel products (Reference B1). Its intended range of application for operating conditions is as follows (Reference B3):

1440 < Pressure < 2490 psia 5 2

0.9 < Mass Flux < 3.7 Mlbm/hr-ft Local Quality *0.30 R

S DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-7 5 S

S

S O In response to concerns raised by the NRC in Reference B3, Dominion will impose one additional restriction on the intended range of applicationa:

O

VIPRE-D/WRB-1 will not be used when the local heat flux exceeds 1.0 Mbtu/hr-ft2 .

The W-3 correlation is used when conditions are outside the range of the WRB-1 DNB correlation.

Specifically, the W-3 correlation is applied to the lower portion of the fuel assemblies in the rod withdrawal from subcritical event because of the bottom peaked axial power distribution assumed, and in the steam line break event because of the low pressures involved. The W-3

correlation with a correlation limit of 1.30 is used below the fuel assembly first mixing vane grid for the rod withdrawal from subcritical event. For the steam line break event, the W-3 correlation is used with a correlation limit of 1.45 in the pressure range of 500 to 1000 psia and 1.30 for pressures

above 1000 psia (Reference B131). The Westinghouse W-3 CHF correlation is described on page 10 O* in Reference B10.

B.4 DESCRIPTION OF VIPRE-D/WRB-1 DATABASE The WRB-1 CHF correlation was developed from a large body of rod bundle CHF data obtained at the Columbia University Heat Transfer Research Facility (HTRF) using full-scale, electrically heated

rod bundle test sections (Reference B2).

The Dominion qualification of WRB-1 in VIPRE-D was performed against a subset of the data from the Columbia-EPRI CHF database for Westinghouse "R" grid 17x17 and 15x15 fuel (Reference B2).

5Dominion analyzed 19 test series out of the 22 series used to develop the correlation; in particular, Dominion did not consider the three series of "L"grid tests and as a consequence no "L" grid data

were included in the test population. The 19 tests represent the same subset of the Columbia-EPRI

CHF database used by Dominion in the qualification of the WRB-1 correlation with the COBRA code (Reference B3).

O Two criteria were used to justify data deletions:

1) The first was consistency with the practice of the test sponsor. Certain points were excluded from the COBRAiWRB-1 database because they had been excluded from the THINCIWRB-1 O database in Reference BI. Most excluded data were deleted under this condition. These points O were also excluded from the COBRANWRB-1 database.

2) The second exclusion criterion was consistency of the input data in References B1 and B2.

0I Although some differences were expected, data points that differed by more than ten standard deviations were excluded as being probable typographical errors in Reference B2.

With the exception of the "L" grid data, and the 25 data points that were thrown out under the second

criterion, the VIPRE-D/WRB-1 database is the same as the one used in Reference B1 to qualify

THINCNVRB-1 for "R" grid fuel. This is also the same database used by Westinghouse to qualify

VIPRE-01NVRB-1 (Reference B6). The same 945 statepoints used in Reference B3 by Dominion in O ,dh a The 13" mixing vane grid spacing restriction was removed via NRC SER dated June 21, 2010.

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-8

S S

the qualification of the WRB-1 correlation with the COBRA code were used in this calculation. Since no "L" grid data were included in the test population, Dominion does not intend to apply the WRB-1 S

correlation to Westinghouse 15x1 5 standard fuel. S S

B.5 VIPRE-DIWRB-1 Test Assemblies B.5.1 4x4 Geometry Tests Twelve of the nineteen tests used by Dominion to qualify the VIPRE-DINRB-1 code/correlation pair have a 4x4 geometry. These 4x4 test bundles have essentially a 15x1 5 subchannel geometry S (Reference B3, page 13). Table B.5.1-1 provides a summary of the key information about each test.

S Table B.5.1-1: 4x4 VIPRE-DN/RB-1 Experimental Database S

PIN 00 / HEATED GRID NUMBER OF TESTS S TEST MATRIX AXIALSHAPE FLUX HEAT GUIDE ODTUBE LENGTH SPACING IN VIPRE-DNWRB-1 S

[nhs S[inches]

[inches]

[inches] DATABASE S

124 4x4 Non-Uniform 0.422/- 96 20 32 S S

125 4x4 Non-Uniform 0.422/- 96 20 33 S

127 4 x4 Non-Uniform 0.422 / - 96 Non-Uniform 36 S S

131 4x4 Non-Uniform 0.422/- 168 26 32 S

132 4x4 Non-Uniform 0.422/- 168 20 36 S 133 4x4 Non-Uniform 0.422/- 168 13 35 S

S 134 4x4 Non-Uniform 0.422/- 168 32 38 S 140 4x4 Non-Uniform 0.422/- 96 32 30 S

S 148 4x4 Non-Uniform 0.422/- 168 26 70 S

153 4x4 Uniform 0.422/- 168 26 40 S S

146 4x4 Non-Uniform 0.422 / 0.545 168 26 37 S

139 4x4 Non-Uniform 0.422 / 0.545 168 32 37 S S

S S

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-9 S

S S

S 0

B.5.2 5x5 Geometry Tests Seven of the nineteen tests used by Dominion to qualify the VIPRE-DIWRB-1 code/correlation pair 0 have a 5x5 geometry. These 5x5 test bundles have the same subchannel geometry as the current 0 Westinghouse 17x17 "R"grid fuel. Table B.5.2-1 provides a summary of key information about each 0 test.

0 Table B.5.2-1: 5x5 VIPRE-D/WRB-1 Experimental Database 0 PIN 0D/ HEATED GRID NUMBER OF TESTS 0 TEST MATRIX AXIAL HEAT FLUX GUIDE ODTUBE LENGTH SPACING IN VIPRE-DANRB-1 FLUX SHAPE SHAE [e ([inches] [inches]

[inches] DATABASE 0

161 5x5 Uniform 0.374/- 168 22 71 0

156 5x5 Uniform 0.374/- 168 26 70 0

0 160 5x5 Uniform 0.374/- 96 22 65 0 157 5x5 Uniform 0.374/- 96 26 76 0

S 164 5x5 Non-Uniform 0.374/- 168 22 74 0 162 5x5 Non-Uniform 0.374 / 0.485 168 22 70 0 158 5x5 Uniform 0.374 / 0.482 96 26 63 0

0 0

S DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-10

0 B.6 VIPRE-D RESULTS AND COMPARISON TO COBRA S Reference B2 describes the mathematical model for each separate test section by providing the bundle and cell geometry, the rod radial peaking values, the rod axial flux shapes, the types, axial locations and form losses associated to the spacer grids, as well as the thermocouple locations.

Reference B2 provides the data for each CHF observation within a test, including power, flow, inlet temperature, pressure and CHF axial location.

Each test section was modeled for analysis with the VIPRE-D thermal-hydraulic computer code as a full assembly model following the modeling methodology discussed in Section 4 in the main body of this report. For each set of bundle data, VIPRE-D produces the local thermal-hydraulic conditions (mass velocity, thermodynamic quality, heat flux, etc.) at every axial node along the heated length of the test section. The ratio of measured-to-predicted CHF (M/P) is the variable that is normally used to evaluate the thermal-hydraulic performance of a code/correlation pair. The measured CHF is the local heat flux at a given location, while the predicted CHF is calculated by the code using the WRB-1 CHF correlation. The ratio of these two values provides the M/P ratio, which is the inverse of the DNB ratio. M/P ratios are frequently used to validate CHF correlations instead of DNB ratios, because their distribution is usually a normal distribution, which simplifies their manipulation and statistical analysis.

In addition to comparing to the experimental results, the results obtained by VIPRE-D when modeling the experiments were benchmarked against the results obtained with the COBRA code in the USNRC approved COBRA topical (Reference B3). This comparison was just a sanity check to verify that there are no suspect datapoints and that the statepoint conditions were correctly input to the code.

S This section summarizes the VIPRE-D results and the associated significant statistics. In addition, this section shows a comparison to the results obtained with the COBRA code as reported in Reference B3. This section also shows the variation of the M/P ratio with each independent variable S to demonstrate that there are no biases in the data. Finally, it provides the VIPRE-D overall statistics for the nineteen WRB-1 tests and generates the DNBR design limit for the WRB-1 CHF correlation with VIPRE-D.

The WRB-1 correlation was developed by Westinghouse by correlating the CHF experimental results obtained in the tests as described in Reference BI. Westinghouse also used these test data to calculate a DNBR design limit of 1.17 for the WRB-1 correlation (References B1 and B6). Dominion used a subset of this experimental data, as described in section B.4, to develop the VIPRE-DIVRB-1 5 DNBR limit. Table B.6-1 summarizes the relevant statistics for each test, and calculates the aggregate statistics for the entire set of data.

One-sided tolerance theory (Reference B4) is used for the calculation of the VIPRE-DNVRB-1 DNBR design limit. This theory allows us to calculate a DNBR limit so that, for a DNBR equal to the design 0 limit, DNB will be avoided with 95% probability at a 95% confidence level.

R S

0 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-li1 S

S

S S

0 0

Table B.6-1: VIPRE-DNVRB-1 M/P Ratio Results 0

NUMBER M/P RATIO M/P RATIO MIP RATIO M/P RATIO TEST OF TEST AVERAGE STDEV MAX TESTS MIN S TS124 32 0.9984 0.0510 1.083 0.851 TS125 33 0.9352 0.0533 1.041 0.802 0 TS127 36 1.0209 0.0877 1.307 0.897 TS131 32 1.0383 0.0827 1.188 0.799 0

TS132 36 1.0382 0.1006 1.185 0.804 TS133 35 0.9473 0.0713 1.093 0.786 0 TS134 38 1.0321 0.0891 1.236 0.864 TS140 30 1.0206 0.0714 S 1.154 0.803 TS148 70 1.0138 0.0815 1.170 0.766 9 TS153 40 0.9220 0.0592 1.013 0.760 S TS146 37 0.9942 0.0516 1.086 0.899 S TS139 37 0.9597 0.0834 1.122 0.787 S 4x4 456 0.9941 0.0847 1.307 0.760 TS161 71 1.0012 0.0624 1.170 0.833 S TS156 70 1.0132 0.0780 1.163 0.812 9 TS160 65 1.0238 0.0812 1.171 0.763 S TS157 76 1.0222 0.0769 1.223 0.800 0 TS164 74 1.0468 0.0869 1.271 0.841 TS162 70 0.9825 0.0712 1.156 0.845 TS158 63 1.0168 0.0848 1.223 0.823 0

5x5 489 1.0154 0.0794 1.271 0.763 VIPRE-D/WRB-1 945 1.0051 0.0827 1.307 0.760 0 COB RA/WRB-1

( R B3) 945 1.0010 0.0838 (Reference 133) 1.287 0.745 0

S Because all the statistical techniques used below assume that the original data distribution is normal, it is necessary to verify that the overall distribution for the M/P ratios is a normal distribution. To evaluate if the distribution is normal, the D' normality test was applied (Reference B5). A value of D' equal to 8,160.9 was obtained for the VIPRE-D/VRB-1 database. This D' value is within the range of acceptability for 945 data points with a 95% confidence level (8,134.0 to 8 ,2 4 5 .4 )b. Thus, it is concluded that the M/P distribution for the VIPRE-DNVRB-1 database is indeed normal.

b From Table 5 in Reference B5 D' Lower Limit (945) [P = 0.025] = 7,558 + (45 / 50) x (8,198 - 7,558) = 8,134.0 D' Upper Limit (945) [P = 0.975] = 7,664 + (45 / 50) x (8,310 - 7,664) = 8,245.4 DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-12

S 0

S S

Based on the results listed in Table B.6-1, the deterministic DNBR design limit can be calculated as:

S

= 1.0 [B.6.1]

S DNBJRL M/P-KN,C,p oM/P S where

= average measured-to-predicted CHF ratio 0

M/P GM/P = standard deviation of the measured-to-predicted CHF ratios of the database 0

KN,C,P = one-sided tolerance factor based on N degrees of freedom, C confidence level, S

and P portion of the population protected. This number is taken from Table 1.4.4 S

in Reference B4. S 0

Normally, the number of degrees of freedom would be the total number of data minus one. However, because Westinghouse used these experimental data to correlate the 12 constants that appear in 0

the WRB-1 correlation, the total number of degrees of freedom must be corrected to account for this. S In addition, the standard deviation of the database needs to be corrected accordingly to account for S this reduced number of degrees of freedom:

S N = n 12 S

= G-M/P.

=YN [ (n -1) / N] [B.6.2]

0 S

Then, the DNBR design limit for the VIPRE-D and the WRB-1 correlation can be calculated as S described in Table B.6-2:

S Table B.6-2: VIPRE-D/WRB-1 DNBR Design Limit 0 S

VIPRE-D/WRB-1 S S

Number of data n 945 S

Degrees of freedom N = n 12 932 S S

Average M/P M/P 1.005 S

Standard Deviation GM/P 0.083 S Corrected Standard S Deviation = O'M/p" [(n-i)! N] 0.084 S Owens Factor K(N,0.95,0.95) 1.730 0 WRB-1 Design limit DNBRL = 1 / (1.005 - 1.730- 0.084) 1.163 S

S S

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-13 0 S

S

S S

With a large database such as this, with 945 statepoints, correcting for the number of constants in the WRB-1 correlation has no significant effect, though technically it is more conservative to make the correction. Either way, the calculated DNBR limit results in a value of 1.17.

Figures B.6-1 through B.6-4 display the performance of the M/P ratio and its distributions as a function of the pressure, mass velocity and quality. These plots show that there are no biases in the M/P ratio distribution, and that the performance of the WRB-1 CHF correlation is independent of the three variables of interest. The plots show a mostly uniform scatter of the data and no obvious trends or slopes. These plots also show that all the tests in the WRB-1 database are within 3.6 standard deviations from the average. Figures B.6-5 through B.6-7 display the performance of the P/M ratio (i.e. the DNBR) against the major independent variables for the WRB-1 database. These plots also S include a DNBR design limit line at 1.17. It can be seen that only 35 data points (3.70% of the database) are above the DNBR design limit, and that these data in excess of the limit are distributed over the entire range of the relevant variables.

S 0

9 S

S S

9 S

0 0

S S

0 S

S DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-14 0

Figure B.6-1: Measured vs. Predicted CHF for VIPRE-DIWRB-1 Database 1.300 1.200 1.100 1.000 0.900 C.,'

1~ 0.800 0.700 U-I 0.600 U

a) 0.500 4-U 4) 0~ 0.400 0.300 0.200 0.100 0.000 0.000 0.100 0.200 0.300 0.400 0.500 0.600 0.700 0.800 0.900 1.000 1.100 1.200 1.300 Measured CHF (MBtu/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-15

Figure B.6-2: M/P vs. Pressure for VIPRE-DIWRB-1 Database 1.400 1.300 1.200 1.100

a. 1.000 0.900 0.800 0.700 0.600 0 500 1000 1500 2000 2500 3000 Pressure (psia)

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-16

Figure B.6-3: M/P vs. Mass Velocity for VIPRE-DIWRB-1 Database 1.400 1.300 1.200

(, 110o ,r 0.900 *~4 1

7.

0.800 4 I-0.700 0.600 0.000 0.500 1.000 1.500 2.000 2.500 3.000 3.500 4.000 Mass Velocity (Mlbm/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-17 0 0 1100000000 0i 000 i s 1&00 0 000 0*0 0 00 090 0 0

Figure B.6-4: M/P vs. Quality for VIPRE-DIWRB-1 Database 1.400 1.300 1.200 1.100 CL n

1.000 0.900 0.800 0.700 0.600 0 T 0 1

-0.300

_0.300

-0.250

-0.200

-0.150

-0.100

-0.050 0.000 0.050 0.100 0 h 0.150 0.200 0.250 0.300 0.350 0.400 Quality DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-18

Figure B.6-5: DNBR vs. Pressure for VIPRE-DIWRB-1 Database 1.400 1.300 -_- -- _

/IPRE-DIWRB-1 Design Limit=1.17 1.200 -- -

1.100 - _.... _ _ _ _ . .

01.000 --- -- _

0.900

_ _ _ .. -

0.800 '-

0.700 0.600 .....

0 500 1000 1500 2000 2500 3000 Pressure (psia)

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-19 O00000.00000000S0000 *0000*0000*000000000**00

Figure B.6-6: DNBR vs. Mass Velocity for VIPRE-DIWRB-1 Database 1.400 1.300 1.200 1.100

. 1.000 z

0.900 0.800 0.700 0.600 4 0.000 0.500 1.000 1.500 2.000 2.500 3.000 3.500 4.000 Mass Velocity (Mlbm/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-20

Figure B.6-7: DNBR vs. Quality for VIPRE-DIWRB-1 Database 1.400 1.300

O 1.200 VIPRE-DIWRB-1 Design Limit=1.17 *

1. 100Ko _ o o_

1.1000 - -- -- -- - . * - *4 - .. . -88. .

t , - -

,% ** *4.

4 4 *j°* ~

01.000 0.800 °.* * #* $#

8.t 8 0.900 ..... ..

0.800 8 ~

0.800 0.600

-0.200 -0.150 -0.100 -0.050 0.000 0.050 0.100 0.150 0.200 0.250 0.300 QUALITY DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-21

B.7 BENCHMARK OF THE VIPRE-DIWRB-1 SUBCHANNEL MODEL In Section 5 of the main body of this report, the Dominion VIPRE-D models created using the selections and modeling guidelines described in Section 4 in the main body of this report provided close comparison to the Framatome ANP LYNXT code, which is a USNRC approved subchannel code. This section in Appendix B demonstrates that the Dominion VIPRE-D models created using the selections and modeling guidelines described in Section 4 in the main body of the report provide close comparison to Dominion's COBRA code, which is also a USNRC approved subchannel code.

This benchmark is provided as an example to demonstrate in sufficient detail the validity of the methodology discussed in the body of this report, and it is not meant to be linked to a specific plant or fuel product.

B.7.1 STEADY STATE APPLICATION Dominion created a 19-channel model for Westinghouse 15x15 SIF fuel at SPS in accordance with the methodology described in Section 4 of this report. This VIPRE-D model of the 1/8th Surry core consists of 19 channels (15 subchannels and 4 lumped channels) and 20 rods, as shown in Figure B.7.1-1. The axial nodalization used in this model has been customized for Westinghouse 15x15 SIF fuel assemblies and contains 73 non-uniform axial nodes with typical node lengths of 2 inches and a maximum node length of less than 6 inches. The reference axial power profile (1.55 chopped cosine) was defined by the default function provided by the VIPRE-D code.

The Westinghouse SIF fuel assembly consists of 204 fuel rods with an outside diameter of 0.422 inches arranged in a 15x15 matrix with a pin pitch of 0.563 inches. The Westinghouse SIF fuel contains several advanced design features, such as mixing vane grids (MVG). The local FLCs used in this VIPRE-D 19-channel model were provided by Westinghouse from full-scale hydraulic tests.

VIPRE-D benchmark calculations were performed against the Dominion COBRA code and the COBRA 19-channel model created by Dominion to model SPS cores containing Westinghouse 15x15 SIF fuel assemblies. This benchmark uses 164 state points obtained from the UFSAR Chapter 14 events including the reactor core safety limits, axial offset envelopes (AO's), rod withdrawal at power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss of flow accident (LOFA), and locked rotor accident (LOCROT) events to compare the performance of VIPRE-D and COBRA. These various limits and events 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 (reactor core safety limits); (b) axial power shapes (AOs);

(c) elevated hot rod power (misaligned rod); and (d) low flow (LOFA and LOCROT). The 164 statepoints cover the full range of conditions and axial offsets in the Surry UFSAR Chapter 14 evaluations (except for MSLB that is discussed in Section B.7.2), and were specifically selected to challenge both the WRB-1 and W-3 CHF correlations (Table B.7.1-1).

This benchmark study showed an average deviation between VIPRE-D and COBRA of less than 0.6% in DNBR, with a maximum deviation of 3.75%. These results are well within the uncertainty typically associated with thermal-hydraulic codes, which has been quantified to be 5%

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-22

0 0

(Reference B9), and justify the model selections in Section 4. Figure B.7.1-2 shows graphically the 0

performance of VIPRE-D versus COBRA for the 164 statepoints. The close comparison of VIPRE-D 0 to COBRA over the full range of conditions expected for UFSAR transients justifies the applications of VIPRE-D to the transients identified in Table 2.1-1 in the main body of this report (MSLB will be 0 discussed in Section B.7.2).

0 0

Table B.7.1-1: Range of VIPRE-D / COBRA 164 Benchmark Statepoints 0 0

VARIABLE RANGE 0

Pressure [psia] 1800 to 2483.2 0

Power [%of 2546 MWt] 53.4 to 144.5 0

Inlet Temperature [OF] 505.1 to 631.7 9

Flow [% of Minimum Measured Flow] 66.8 to 100 0

FAH 1.56 to 2.106 0

Axial Offset [%] -76.6 to 32.2 0

0 0

0 S

S 0

S S

0 S

S 0

0 0

0 S

S 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-23

S S

Figure B.7.1-1. Typical Surry VIPRE-D 19-Channel Model for Westinghouse 15x15 SIF Fuel Assemblies 0

0 0

0 Rod 17 0 Channel 16 Remainder of PLB 0

0 S

S S

0 0

S 0

Power Limiting Bundle DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-24

Figure B.7.1-2:VIPRE-DIWRB-1 vs. COBRAIWRB-1 for the 164 Analyzed Statepoints 4.000 3.500 3.000 m

Z o 2.500 LU a-2.000 1.500 1.000 1.000 1.500 2.000 2.500 3.000 3.500 4.000 COBRA/WRB-1 MDNBR DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-25 0100100 00000 *0010 6 *4b*** 40 09 0*000 0*04000090 0 0

B.7.2 MAIN STEAM LINE BREAK APPLICATION The VIPRE-D 19-channel model discussed in section B.7.1 was also used to simulate the behavior of the core during a MSLB event, as it allows the modeling of the peaking and inlet boundary conditions in the fuel assemblies adjacent to the hot assembly. The two most limiting cases from a recent reload were evaluated with the VIPRE-D code, and their results compared to the COBRA results. The results obtained show a maximum deviation of 1.5% in DNBR. These results demonstrate that VIPRE-D can analyze a MSLB event, provided the model has sufficient detail surrounding the hot assembly, such as the 19-channel model described here. It is important to note that both MSLB statepoints evaluated occurred at pressures below 1000 psia, and therefore the MDNBR was evaluated with the W-3 CHF correlation, and the appropriate correlation limit was 1.45 (Reference 86).

B.7.3 TRANSIENT APPLICATION As demonstrated in Section 5.3 in the main body if this report, VIPRE-D has the capability to perform transient calculations by using boundary conditions obtained from a reactor systems code or a neutronic code. The reactor systems code provides time-dependent forcing functions for pressure, core average power, core flow rate and core inlet temperature and the neutronics code provides core power distributions and nuclear peaking factors.

VIPRE-D/WRB-1 transient capability was tested by performing two sample transient calculations.

These two transient calculations were only intended to be samples designed to exercise the transient capabilities of the VIPRE-D code and a typical VIPRE-D model created according to the guidelines discussed in Section 4 in the main body of this report. In both cases, the behavior of the VIPRE-D results was successfully compared to the behavior of the COBRA analysis of record in the UFSAR.

The first sample transient selected to perform this verification was the Feedwater Malfunction Transient (FWMAL). Forcing functions for the FWMAL transient were obtained from the SPS UFSAR. The length of the transient was 195 seconds, with a 0.5-second time step. COBRA analysis of record and VIPRE-D calculations exhibited similar behavior, and the MDNBR results show a maximum deviation of less that 0.4% (see Figure B.7.3-1).

The second sample transient selected to perform this verification was the Locked Rotor Transient (LOCROT). Forcing functions for the LOCROT transient were obtained from the SPS UFSAR. The length of the transient was 9.5 seconds, with a 0.025-second time step. COBRA analysis of record and VIPRE-D calculations exhibited similar behavior, and the MDNBR results show a maximum deviation of less that 1.6% (see Figure B.7.3-2).

The transient analyses demonstrate that VIPRE-DNVRB-1 is capable of performing stable transient calculations and the results obtained are essentially the same as the COBRA/WRB-1 results documented in the SPS UFSAR.

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-26

0 0

0 Figure B.7.3-1 VIPRE-D FWMAL Transient Sample Calculation Results 0

0 2.8 S

0 2.6 S

2.4 2.2 0

w COBRA S ca z 2

-VIPRE-D 0 1.8 S 1.6 0

1.4 S

1.2 9 1

S 0 20 40 60 80 100 120 140 160 180 200 S Time [s] 0 S

Figure B.7.3-2: VIPRE-D LOCROT Transient Sample Calculation Results 4

S S

3.5 S

0 S

3 0

COBRA S

ca z - VI PRE-D S 2.5 S S

S 2 S S

S 1.5 S 0 1 2 3 4 5 6 7 8 9 10 Time [s] S S

S DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-27 S

S

B.8 CONCLUSIONS The WRB-1 correlation has been qualified with Dominion's VIPRE-D computer code. Table B.8-1 summarizes the DNBR design limits for VIPRE-DNWRB-1 that yields a 95% non-DNB probability at a 95% confidence level. The limit of 1.17 from VIPRE-D is the same limit as found with three other, approved code packages: COBRA (Reference B3), THINC (Reference B1), and Westinghouse's version of VIPRE-01 (Reference B6). Westinghouse WRB-1 CHF correlation has been approved by the USNRC for use with Westinghouse 15x15 and 17x17 "R" grid type fuel, and with Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel products.

Table B.8-1: DNBR Limits for WRB-1 Table B.8-2 summarizes the applicability and the ranges of validity for VIPRE-D/WRB-1, which are the same as those on page 2 of the Dominion COBRA SER in Reference B3 and augmented by NRC SER dated June 21, 2010 that removed the 13" mixing vane grid spacing restriction.

Table B.8-2: Range of Validity for VIPRE-D/WRB-1 Pressure

[psial 1,440 to 2,490 Mass Velocity 0.9 to 3.7

[Mlbm/hr-ft2 ]

Thermodynamic <0.30 Quality at CHF Local Heat Flux 2

[Mbtu/hr- ft ] *1.0 Finally, extensive code benchmark calculations have confirmed that the VIPRE-DNVRB-1 models created using the modeling guidelines specified in Section 4 in the main body of this report produce essentially the same results as USNRC approved equivalent Dominion COBRAIWRB-1 models.

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-28

B.9 REFERENCES BI. Technical Report, 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.

B2. Technical Report, EPRI NP-2609, "Parametric Study of CHF Data, Volume 3, Part 1; Critical Heat Flux Data," C. F. Fighetti, & D.G. Reddy, September 1982.

B3. Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," R.C. Anderson & N. P. Wolfhope, July 1990.

B4. Technical Report, "Tables for Normal Tolerance Limits, Sampling Plans, and Screening," R. E.

Odeh and D. B. Owen, 1980.

B5. Technical Report, "Assessment of the Assumption of Normality (employing individual observed values)," American National Standards Institute, ANSI N15.15.1974.

B6. Technical Report, WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," Y. X. Sung, P. Schueren, and A. Meliksetian, October 1999.

B7. Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), "Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

B8. Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 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.

B9. Topical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," R. C. Anderson, June 1987.

B10. Technical Report, "Boiling Crisis and Critical Heat Flux," TID-25887, 1972.

B1 1. Letter from A.C. Thadani (NRC) to W. J. Johnson (Westinghouse), "Acceptance for Referencing of Licensing Topical Report, WCAP-9226-P, WCAP-9227-NP, Reactor Core Response to Excessive Secondary Steam Releases," 1989.

DOM-NAF-2, Rev. 0.2-A, APPENDIX B B-29

0 0

0 0 DOM-NAF-2, Rev. 0.2-A I APPENDIX C 0

0 0

0 0 Qualification of the Westinghouse WRB-2M CHF 0 Correlation in the Dominion VIPRE-D Computer Code 0

S NUCLEAR ANALYSIS AND FUEL DEPARTMENT DOMINION 0 RICHMOND, VIRGINIA 0

0 Approved by NRC Safety Evaluation 0 dated April 22, 2009 0

0 0

S 9,

S

0 0

0 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.

0 ABSTRACT This appendix documents Dominion's qualification of the Westinghouse WRB-2M correlation with the VIPRE-D code. This qualification was performed against the same CHF experimental database used by Westinghouse to develop and license the correlation. This appendix summarizes the data evaluations that were performed to qualify the VIPRE-D/WRB-2M code/correlation pair, and to develop the corresponding DNBR design limit.

R 0

0 0

0 DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-2 0

0

TABLE OF CONTENTS S

CLASSIFICATION/DISCLAIMER ....................................................................... C-2 A B S T R A C T ............................................................................................ ..... . C -2 TABLE OF CONTENTS .................................................................................. C-3 0

LIS T O F T A B LE S ......................................................................................... C -4 LIS T O F F IG UR E S ....................................................................................... C -4 9

ACRONYMS AND ABBREVIATIONS ................................................................ C-5 C .1 P UR P OS E ............................................................................................ C-6 S C .2 A P P LIC A B ILIT Y .................................................................................... C -6 0.3 C DESCRIPTION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION.... C-7

1 C.4 DESCRIPTION OF THE VIPRE-D/WRB-2M DATABASE AND TEST A S S E MB LIE S ..................................................................................... C -8 C.5 VIPRE-D/WRB-2M RESULTS AND COMPARISON TO VIPRE-01 .................. C-9 S C .6 C O NC LU S IO NS .................................................................................... C-22 C .7 R E F E R E NC E S ..................................................................................... C -23 R

0 S

0 0

S S

S 0

S S

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-3 S

0 0

LIST OF TABLES 0 Table C .4-1: Sum m ary of O HF Tests ................................................................. C-9 0 Table C.5-1: Sum m ary of VIPRE-D Results ......................................................... C-10 0

Table C.5-2: Statistical Analysis of W RB-2M Design Limit ..................................... C-11 Table C.5-3: Range of Validity for W RB-2M ....................................................... C-11 0

Table C.5-4: M/P CHF Performance by Independent Variable Grouping of the WRB-2M 0 Database at 95% Confidence Level ................................................... C-13 0 Table 0 .6-1: DN BR Lim its for W RB-2M .............................................................. C-22 0 Table 0.6-2: Range of Validity for VIPRE-DIW RB-2M ........................................... C-22 0 9

0 LIST OF FIGURES 0 Figure C.5-1: Measured vs. Predicted CHF for VIPRE-D/WRB-2M Database ............. C-14 Figure C.5-2: M/P vs. Pressure for VIPRE-D/VRB-2M Database ............................ C-15 0

Figure C.5-3: M/P vs. Mass Velocity for VIPRE-D/WRB-2M Database ...................... C-16 9 Figure C.5-4: M/P vs. Quality for VIPRE-DNVRB-2M Database ............................... C-17 0 Figure C.5-5: DNBR vs. Pressure for VIPRE-DA/RB-2M Database ......................... C-18 9 Figure C.5-6: DNBR vs. Mass Velocity for VIPRE-D/WRB-2M Database ................... C-19 0 Figure C.5-7: DNBR vs. Quality for VIPRE-DNVRB-2M Database ............................ C-20 Figure C.5-8:

0 VIPRE-D/WRB-2M Probability Density Function ................................ C-21 0

0 0

0 9

0 0

9 DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-4 0

S S

0 S

0 ACRONYMS AND ABBREVIATIONS 0

0 0

0 CHF Critical Heat Flux S DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio S FLC Form Loss Coefficient HTRF Heat Transfer Research Facility at Columbia University 9 IFM Intermediate Flow Mixer 0 LPD Low Pressure Drop M/P Ratio of Measured-to-Predicted CHF 0 MIFM Modified Intermediate Flow Mixer 9 MPS Millstone Power Station MVG Mixing Vane Grid NMVG Non-Mixing Vane Grid 0 P/M Ratio of Predicted-to-Measured CHF (equivalent to DNBR)

PWR Pressurized Water Reactor 0 RFA Robust Fuel Assembly 0 USNRC US Nuclear Regulatory Commission 0

S 0

0 0

S 0

0 0

S 0

S DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-5

0 S

0 C.1 PURPOSE S Dominion currently uses the Westinghouse 17x17 Robust Fuel Assembly (RFA) fuel product at 0 Millstone Power Station (MPS), Unit 3. The thermal-hydraulic analysis of this Westinghouse fuel product requires the use of the Westinghouse WRB-2M CHF correlation (Reference Cl). In fact, the Westinghouse WRB-2M CHF correlation has been approved by the USNRC for use with the 17x17 RFA fuel design with or without the Intermediate Flow Mixer (IFM) grids (Reference Cl).

To be licensed for use, a critical heat flux (CHF) correlation must be tested against experimental 0 data that span the anticipated range of conditions over which the correlation will be applied. S Furthermore, the population statistics of the database must be used to establish a departure from nucleate boiling ratio (DNBR) design limit such that the probability of avoiding departure from nucleate boiling (DNB) will be at least 95% at a 95% confidence level.

This addendum documents Dominion's qualification of the WRB-2M correlation with the 0 VIPRE-D code. This qualification was performed against the data from the Columbia University Heat Transfer Research facility (HTRF) for the Modified Vantage 5H and Modified Vantage 9 5H/IFM fuel types (Reference Cl). This is the same set of the Columbia-EPRI CHF database used by Westinghouse in the qualification of the WRB-2M correlation with the VIPRE-01 code (Reference Cl). This addendum summarizes the data evaluations that were performed to qualify the VIPRE-DNWRB-2M code/correlation pair, and to develop the corresponding DNBR S design limits for the correlation. 5 S

C.2 APPLICABILITY S Dominion intends to use the VIPRE-D/WRB-2M code/correlation for Westinghouse 17x17 RFA S fuel products, with or without modified intermediate flow mixers (MIFM) in a PWR reactor. S When evaluating this type of fuel outside of the range of validity of the WRB-2M CHF 5 correlation, Dominion intends to use the VIPRE-D/W-3 code/correlation pair. W-3 is one of the CHF correlations contained in the USNRC approved generic version of VIPRE-01 (References C3 and C4), and it has already been approved for use with the VIPRE-D code (Reference C5).

The intended VIPRE-D/WRB-2M applications discussed in this addendum are consistent with S the generic intended applications listed in the main body of this report (Section 2.0 in Reference 5 C5). Also, more specifically, Dominion intends to use VIPRE-DINRB-2M to analyze the transients delineated in Table 2.1-1 in Section 2.0 of the main body of this report (Reference C5). The qualification of the WRB-2M correlation with the VIPRE-D code has been performed following the modeling guidelines described in Section 4 of this report (Reference C5).

This Addendum is submitted to the USNRC for review and approval in order to meet the S USNRC's requirement #2 listed in the VIPRE-01 SER, as outlined in Section 2.2 in the main 5 body of this report (Reference C5). 5 S

S DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-60 S

S 0

0

C.3 DESCRIPTION OF THE WESTINGHOUSE WRB-2M CHF

CORRELATION In pressurized water reactor (PWR) cores, the energy generated inside the fuel pellets leaves

the fuel rods at their surface in the form of heat flux, which is removed by the reactor coolant

system flow. The normal heat transfer regime in this configuration is nucleate boiling, which is very efficient. However, as the capacity of the coolant to accept heat from the fuel rod surface degrades, a continuous layer of steam (a film) starts to blanket the tube. This heat transfer regime, termed film boiling, is less efficient than nucleate boiling and can result in significant increases of the fuel rod temperature for the same heat flux. Since the increase in temperature 0may lead to the failure of the fuel rod cladding, PWRs are designed to operate in the nucleate 0 boiling regime and protection against operation in film boiling must be provided.

0

The heat flux at which the steam film starts to form is called CHF or the point of DNB. For

design purposes, the DNBR is used as an indicator of the margin to DNB. The DNBR is the ratio of the predicted CHF to the actual local heat flux under a given set of conditions. Thus,

DNBR is a measure of the thermal margin to film boiling and its associated high temperatures.

The greater the DNBR value (above 1.0), the greater the thermal margin.

0

The CHF cannot be predicted from first principles, so it is empirically correlated as a function of

the local thermal-hydraulic conditions, the geometry, and the power distribution measured in the experiments. Since a CHF correlation is an analytical fit to experimental data, it has an associated uncertainty, which is quantified in a DNBR design limit. A calculated DNBR value 0greater than this design limit provides assurance that there is at least a 95% probability at the

95% confidence level that a departure from nucleate boiling will not occur.

0 0Correlations to predict the occurrence of CHF have undergone evolutions as nuclear fuel

designs have changed. Westinghouse developed the WRB-1 CHF correlation for the prediction of DNB for Westinghouse fuel assemblies with mixing vane grids (MVGs). Subsequently, Westinghouse developed the WRB-2 CHF correlation for the prediction of DNB in

Westinghouse fuel assemblies with MVGs and intermediate flow mixing grids (IFMs). More recently, Westinghouse has modified their nuclear fuel design to reduce fuel rod mechanical wear and to further improve thermal/hydraulic performance.

0

The new fuel design includes modified low pressure drop (LPD) mixing vane grids and modified intermediate flow mixing grids (MIFMs). This new fuel design is called the modified Vantage 5H and Modified Vantage 5H/IFM depending on whether the MIFMs have been included. (When

the design includes the MIFM grids, it has also been referred to as the Robust Fuel Assembly

(RFA)). CHF tests with the modified grids were conducted at the Columbia HTRF with and

without control rod guide thimbles and with and without MIFM grids. Although the new data was successfully correlated by Westinghouse using the WRB-2 CHF correlation, a better correlation, 0the WRB-2M CHF correlation (a modification of the WRB-2 CHF correlation), was obtained by 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-7 0

0 incorporation of a multiplier 'M' (Reference Cl). The sensitivity of WRB-2M to other parameters, such as various power shapes is very similar to WRB-2. Thus, WRB-2M is applicable for 17x17 fuel with 0.374 inch OD rods, and Modified LPD grids with or without MIFM's. The range of applicable parameters is given in Table C.5-3. The WRB-2M correlation has been approved by the NRC (Reference Cl).

i The WRB-2M DNB correlation was developed from test bundles simulating the RFA fuel design with only a cosine axial power shape. As part of a scoping study for new grid designs, DNB tests were performed with the uniform axial power shape at the Columbia University test loop.

Although the grid designs were not the same, the mixing vanes of the test bundles were similar to the RFA fuel design. When compared to the data from those tests, the WRB-2M measured- S to-predicted (M/P) CHF average ratio was lower than 1.0. No significant trend in M/P was O observed with respect to key parameters such as local flow rate, local equilibrium quality, and O pressure. Based on this comparison with the test results, Westinghouse decided to adjust the DNB predictions for the WRB-2M DNB correlation. The adjustment factor does not constitute a change in the methodology as described in the licensing basis. The NRC staff has reviewed the adjustment factor and its consequences and found it acceptable (Reference C6).

The W-3 correlation is used when conditions are outside the range of the WRB-2M DNB 0 correlation. Specifically, the W-3 correlation is applied to the lower portion of the fuel assemblies in the rod withdrawal from subcritical event because of the bottom peaked axial power distribution assumed, and in the steam line break event because of the low pressures involved. The W-3 correlation with a correlation limit of 1.30 is used below the fuel assembly first mixing vane grid for the rod withdrawal from subcritical event. For the steam line break event, the W-3 correlation is used with a correlation limit of 1.45 in the pressure range of 500 to 1000 0 psia and 1.30 for pressures above 1000 psia (Reference C8). The Westinghouse W-3 CHF S correlation is described on page 10 in Reference C7. O i

C.4 DESCRIPTION OF THE VIPRE-DIWRB-2M DATABASE AND TEST ASSEMBLIES

The WRB-2M CHF correlation was developed from CHF data obtained at the Columbia 0 University HTRF using full-scale, electrically heated rod bundle test sections (Reference Cl). 5 The Dominion qualification of WRB-2M in VIPRE-D was performed against the same test data 5 from the Columbia-EPRI CHF database for Westinghouse 17x17 fuel. Dominion used the CHF experimental data used by Westinghouse to develop the WRB-2M correlation. No data point was deleted or excluded.

0 The HTRF test assemblies had a 5x5 geometry, thus the test assemblies used by Dominion to O qualify the VIPRE-DNVRB-2M code/correlation pair have a 5x5 geometry. These 5x5 test 5 bundles have essentially a 17x17 subchannel geometry (Reference Cl). Table C.4-1 provides a summary of the key information about each test.

R 0

DOM-NAF-2, Rev. 0.2-A, APPENDIX C 0-8 0

0 Table C.4-1: Summary of CHF Tests 0 PIN OD I AXIAL HEAT GUIDE TUBE HEATE MIFM NUMBER FLUX SHAPE OD LEnGh grids? OF TESTS

[inches] [inches]

S A-1 5x5 Non-Uniform 0.374/- 168 Yes 66 0

A-2 5x5 Non-Uniform 0.374 168 Yes 77 0 0.474 0 A-3 5x5 Non-Uniform 0.374/

0.474 168 No 70 0

0 A-4 5x5 Non-Uniform 0.374/- 168 No 28 0

0 0 C.5 VIPRE-DIWRB-2M RESULTS AND COMPARISON TO S VIPRE-01 0 Reference C1 describes the mathematical model for each separate test section by providing the 0 bundle and cell geometry, the rod radial peaking values, the rod axial flux shapes, the types, axial locations and form losses associated to the spacer grids, as well as the thermocouple locations. Reference C1 also provides the data for each CHF observation within a test, 0 including power, flow, inlet temperature, pressure and CHF axial location.

0 0 Each test section was modeled for analysis with the VIPRE-D thermal-hydraulic computer code as a full assembly model following the modeling methodology discussed in Section 4 in the main body of this report. For each set of bundle data, VIPRE-D produces the local thermal-hydraulic 0 conditions (mass velocity, thermodynamic quality, heat flux, etc.) at every axial node along the 0 heated length of the test section. The ratio of measured-to-predicted CHF (M/P) is the variable that is normally used to evaluate the thermal-hydraulic performance of a code/correlation pair.

0 The measured CHF is the local heat flux at a given location, while the predicted CHF is calculated by the code using the WRB-2M CHF correlation. The ratio of these two values 0 provides the M/P ratio, which is the inverse of the DNB ratio. M/P ratios are frequently used to validate CHF correlations instead of DNB ratios, because their distribution is usually a normal 0 distribution, which simplifies their manipulation and statistical analysis.

0 This section summarizes the VIPRE-D results and the associated significant statistics. This 0 section also shows the variation of the M/P ratio with each independent variable to demonstrate that there are no biases in the data. Finally, it provides the VIPRE-D overall statistics for the WRB-2M tests and generates the DNBR design limit for the WRB-2M CHF correlation with VIPRE-D.

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-9

S 0

0 The WRB-2M correlation was developed by Westinghouse by correlating the CHF experimental results obtained in the tests as described in Reference C1. Westinghouse also used these test data to calculate a DNBR design limit of 1.14 for the WRB-2M correlation (Reference C1).

Dominion used these experimental data, as described in section C.4, to develop the VIPRE-DIWRB-2M DNBR limit. Table C.5-1 summarizes the relevant statistics for each test, and calculates the aggregate statistics for the entire set of data.

Table C.5-1 : Summary of VIPRE-D Results 0

Test Number of M/P Ratio M/P Ratio M/P Ratio MIP Ratio Tests Average STDEV Max Min S

A-1 66 1.0178 0.0789 1.2114 0.8287 0

A-2 77 0.9834 0.0538 1.1017 0.8614 0 A-3 A-4 70 28 1.0144 0.9731 0.0559 0.0490 1.1444 1.1002 0.8954 0.8688 0

Thimble 147 0.9982 0.0568 1.1444 0.8614 0 Typical 94 1.0045 0.0740 1.2114 0.8287 0 With MIFM 143 0.9993 0.0685 1.2114 0.8287 Without MIFM 98 1.0026 0.0570 1.1444 0.8688 0 All Results 241 1.0006 0.0640 1.2114 0.8287 0 0

One-sided tolerance theory (Reference C2) is used for the calculation of the VIPRE-DNVRB-2M 0 DNBR design limit. This theory allows the calculation of a DNBR limit so that, for a DNBR equal 0 to the design limit, DNB will be avoided with 95% probability at a 95% confidence level. 0 First, it is necessary to verify that the overall distribution for the M/P ratios is a normal 0

distribution, because all the statistical techniques used below assume that the original data 0 distribution is normal. To evaluate if the distribution is normal, the D' normality test was applied. 0 A value of D' equal to 1047.04 was obtained for the VIPRE-D/WRB-2M database. This D' value 0 is within the range of acceptability for 241 data points with a 95% confidence level (1038.60 to 0 1066.75)[1]. Thus, it is concluded that the M/P distribution for the VIPRE-D/IWRB-2M database is indeed normal. Based on the results listed in Table C.5-1, the deterministic DNBR design limit 0

can be calculated as:

0 0

DBRL.0 [C.5.1]

0 M /P - KN,C,t,

C"M/I1 0 where 0 M/P = average measured to predicted CHF ratio c'M/P = standard deviation of the measured to predicted CHF ratios of the database 0

KN,C,P = one-sided tolerance factor based on N degrees of freedom, C confidence level, 0 and P portion of the population protected. This number can be obtained from 0 Table 1.4.4 of Reference C2.

0 1 From Table 5 in Reference C9 D' Lower Limit (241) [P = 0.025] = 1038.60 0

D' Upper Limit (241) [P = 0.975] = 1066.75 0 0

0 DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-10 0 0

S 0

S Normally, the number of degrees of freedom would be the total number of data minus one.

However, because Westinghouse used these experimental data to correlate the 6 constants 0 that appear in the WRB-2M correlation, the total number of degrees of freedom must be 0 corrected to account for this. In addition, the standard deviation of the database needs to be corrected accordingly to account for this reduced number of degrees of freedom:

0 N =n-1-6 [C.5.2]

0 (3N c'M/P"[ (n -1) / N]

0 Then, the DNBR design limit for the VIPRE-D and the WRB-2M correlation can be calculated as 0 shown in Table C.5-2.

0 Table C.5-2: Statistical Analysis of WRB-2M Design Limit 0

Number of data n 241 0

Degrees of freedom N = n 6 234 0

Average M/P M/P 1.0006 Standard Deviation TaM/P 0.0640 0

Corrected Standard Deviation CFN = GM/P [(n -1) / N] 0.0648 S Owens Factor K(N,0.95,0.95) 1.8170 0 WRB-2M Design limit DNBRL = 1 / (M/P - K(N,0.95,0.95)

CYN) 1.1327 0

0 Even though this is not a large database, correcting for the number of constants in the WRB-2M 0 correlation has no significant effect, and it is more conservative to make the correction. The calculated DNBR limit results in a value of 1.14. This is the same number reported by S Westinghouse in Reference C1 and has been approved by the NRC.

0 Table C.5-3 summarizes the ranges of validity for the VIPRE-D/WRB-2M correlation. These 0 ranges, are identical to those submitted by Westinghouse and already approved by the NRC (Reference Cl).

Table C.5-3: Range of Validity for WRB-2M S VIPRE-D Pressure [psia] 1495 to 2425 2

Mass Velocity [Mlbm/hr-ft ] 0.97 to 3.1 Thermodynamic Quality at CHF -0.1 to 0.29 Figures C.5-1 through C.5-4 display the performance of the M/P ratio, and its distributions as a function of the pressure, mass velocity and quality. The objective of these plots is to show that DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-11

0 S

S there are no biases in the M/P ratio distribution, and that the performance of the WRB-2M S correlation is independent of the three independent variables of interest. The plots show a mostly uniform scatter of the data and no obvious trends or slopes. These plots also show that S

all the tests in the WRB-2M database are within 3.5 standard deviations from the average.

S Figures C.5-5 through C.5-7 display the performance of the P/M ratio (i.e., the DNBR) against S the major independent variables for the WRB-2M database. These plots also include the DNBR S design limit line. It can be seen that only six data points (2.49% of the database) are above the S DNBR design limit, and that these data in excess of the limit are distributed over the variable S ranges tested.

S A more formal determination of the lack of bias of the average M/P ratio can be done using the S analysis of variance test (ANOVA) shown in Table C.5-4. ANOVA tests are normally applied to S highly controlled situations, but they can be somewhat useful in CHF testing and correlation. S However, the ANOVA test cannot be used as the sole measure of the performance of a CHF S correlation, but it would indicate an extremely bad mismatch (with a very large F statistic). The S variables analyzed were pressure, quality, mass velocity and test cell type. The ANOVA results for VIPRE-DIWRB-2M slightly exceed the critical values of F for pressure and quality, but other S

comparisons prove the hypothesis that all the groups belong to the same distribution; i.e., that S there is no bias of the results regarding the analyzed variables. Furthermore, when looking at S the figures in this section, there does not appear to be any trend or bias in the data. Therefore, S it can be concluded that the WRB-2M M/P ratio database is independent of the pressure, S quality, and mass velocity.

S S

0 S

S 0

S S

S DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-12 S

S S

Table C.5-4: M/P CHF Performance by Independent Variable Grouping of the WRB-2M Database at 95% Confidence Level S GrouingNumber AveragelStandardlMaximuml Minimum of Data MIP lý Deviation M/P MIP 0

Analysis by Pressures Below 1575 psia 29 0.9831 0.0610 1.1153 0.8746 1575 - 1850 psia 67 0.9967 0.0666 1.1327 0.8614 S 1850 - 2250 psia 74 1.0210 0.0680 1.2114 0.9058 0 Above 2250 psia 71 0.9903 0.0534 1.1002 0.8287 0 Fdistribution = 4.0812 Fcriti.a(3,237) = 2.6427 Analysis by Qualities Below 5% 42 1.0107 0.0521 1.1193 0.8688 5% to 10% 71 1.0126 0.0581 1.1302 0.8862 0 10% to 15% 66 1.0051 0.0633 1.1744 0.8794 0 15% to 20% 46 0.9718 0.0628 1.1444 0.8287 Above 20% 16 0.9858 0.0961 1.2114 0.8336 0 Fdistribution = 3-6650 Fcriti.i(4,236) = 2.4099 Analysis by Mass Velocities Below 1.25 Mlbm/hr-ft2 32 0.9915 0.0645 1.1444 0.8688 1.25 - 1.75 Mlbm/hr-ft2 75 1.0055 0.0642 1.1302 0 1.75 - 2.25 Mlbm/hr-ft2 74 1.0023 0.0602 1.1744 0.8614 0.8287 0 2.25 - 2.75 Mlbm/hr-ft 2 51 0.9911 0.0637 1.1264 0.8336 0 2.75 - 3.25Mlbm/hr-ft' 9 1.0323 0.0879 1.2114 0.9063 Fdistribution = 1.1176 Friti. 1(4,236) = 2.4099 0 Analysis by Geometry Type With MIFM Grids 143 0.9993 0.0685 1.2114 0.8287 0 Without MIFM Grids 98 1.0026 0.0571 1.1444 0.8688 Fdistrbution = 0.1580 Fcriti.(1,239) = 3.8807 0 Analysis by Thimble vs. Typical 0 Thimble 147 0.9982 0.0568 1.1444 0.8614 0 Typical 94 1.0045 0.0740 1.2114 0.8287 Fdistribution = 0.5546 Fcriticai(1,239) = 3.8807 All Data WRB-2M S All Data 241 1.0006 10.0640 1.2114 0.8287 S

0 0

0 S

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-13

Figure C.5-1: Measured vs. Predicted CHF for VIPRE-DIWRB-2M Database 1.000 0.900 0.800 eq 0.700 "I-0.600 10 0.500

=-.

C.)

0.400 0.300 0.200 0.100 0.000 0.000 0.100 0.200 0.300 0.400 0.500 0.600 0.700 0.800 0.900 1.000 Measured CHF (MBtu/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-14 OOOOOOOOOOOOOOOOOO 0OOOOOOOOOOOO0 OOOOOOOOOOOOOO*0 *04

Figure C.5-2: M/P vs. Pressure for VIPREIWRB-2M Database 1.3 1.2 1.1 0.9 0.8 0.7 0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 Pressure (psia)

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-15

Figure C.5-3: M/P vs. Mass Velocity for VIPRE-DIWRB-2M Database 1.3 1.2 1.1

,it

- 11 0.9 0.8 0.7 0.000 0.500 1.000 1.500 2.000 2.500 3.000 3.500 Mass Velocity (Mlbm/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-16 1*

Is 0 41

0 0 90 0 0 9 0

0 9 40

0 0 0 0 0 0 0 0 0 0 0 10 0 1 10 0

Database Figure C.5-4: M/P vs. Quality for VIPRE-DIWRB-2M 1.3 1.2 1.1

_ t. . .

0.9 0.8 0.8 0.7 0.350 0.100 0.150 0.200 0.250 0.300

-0.150 -0.100 -0.050 0.000 0.050 Quality DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-1 7

Figure C.5-5: DNBR vs. Pressure for VIPRE-D/WRB-2M Database 1.3 1.2 1.1 C-II 1

z 0

0.9 0.8 0.7 0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 Pressure (psia)

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-1 8 400004 01000 0*000 4000 0 000 004419004010040*040 *0 04

Figure C.5-6: DNBR vs. Mass Velocity for VIPRE-DIWRB-2M Database 1.3 1.2 1.1 II m 1 z

a 0.9 0.8 0.7 0.000 0.500 1.000 1.500 2.000 2.500 3.000 3.500 Mass Velocity (Mlbm/hr-ft2)

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-1 9

Figure C.5-7: DNBR vs. Quality for VIPRE-DIWRB-2M Database 1.3 1.2 11 CL 1

z 0.9 0.8 0.7 -

-0.150 -0.100 -0.050 0.000 0.050 0.100 0.150 0.200 0.250 0.300 0.350 Quality DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-20

S S

S The 241 data points of the VIPRE-DNVRB-2M M/P distribution calculated by Dominion were S

used to create the empirical probability density function. These data points were distributed S among 21 equal bins that covered the entire range of M/P in the VIPRE D/WRB-2M distribution, 0 and the frequency of data in each bin was determined. The resulting empirical probability density functions for the VIPRE-D/WRB-2M distribution were then compared with the probability S density function of a normal distribution of mean 1.0006 and standard deviation 0.0640, which is S the mean and standard deviation for the VIPRE-D/WRB-2M distribution calculated in Section C.5 above. Figure C.5-8 displays the resulting empirical probability density function for the 0

VIPRE-DNVRB-2M M/P distribution, and compares it with the probability density function of the normal distribution of mean 1.0006 and standard deviation 0.0640.

0 S

0 0 Figure C.5-8: VIPRE-DIWRB-2M Probability Density Function 0 1800U% -r 1

0 16.00%

S 0

14.00%

1' \

H 12.00% t 0

>, 10.00%

0 K

C o" 8.00%

0 LL 6.00%

0 0

0 4.00%

0.00%

+

2.00% t C) 0 (N C(N 0

0 0 (N n

0 0

U) 0 0N Ný fl 0

0ý H

U) o U) 00 U

0 0o Uo 0 0

04 (N

0 0

In

,:0 04 CC 0) n) O5 o) 0 0 0 (N (N!

0 o a 0 M/P 0 IE2 VIPRE-D/WRB-2M Data - NORMAL(1.0006, 0.0640)1 0

0 S

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-21

0 0

0 C.6 CONCLUSIONS 0 The WRB-2M correlation has been qualified with Dominion's VIPRE-D computer code. Table C.6-1 summarizes the DNBR design limits for VIPRE-D/WRB-2M that yields a 95% non-DNB 0 probability at a 95% confidence level. The limit of 1.14 from VIPRE-D is the same limit as found with Westinghouse's version of VIPRE (Reference Cl). The Westinghouse WRB-2M CHF 0

correlation has been approved by the USNRC for use with Westinghouse 17x17 with 0.374 inch 0 OD rods, and Modified LPD grids with or without MIFM's in a PWR reactor.

0 0

Table C.6-1: DNBR Limits for WRB-2M 0

S 0

Table C.6-2 summarizes the applicability and the ranges of validity for VIPRE-D/WRB-2M, 0

which are the same as those on page 4-2 of Reference C1. 0 S

0 Table C.6-2: Range of Validity for VIPRE-DIWRB-2M 0

Pressure 1495[2l to 2425

[psia] 0 Mass Velocity 0.97 to 3.1 0

[Mlbm/hr-ft 2] 0 Thermodynamic -0.1 to 0.29 Quality at CHF 0

0 0

2 0 The NRC SER [Serial #09-290] for WRB-2M (Appendix C) typographically listed the beginning of the pressure range as 1405 psia and stated it should match the WCAP-15025-P-A applicability range of 1495 to 2425 psia. Also a header in the applicability table, Table 1, in the NRC SER listed WNG-1, which should be WRB-2M.

The original submittal by Dominion to the NRC [Serial #08-0174] for Appendix C had the beginning of the pressure range was listed as 1440 psia and stated that the applicability range should match those which are on page 4-2 of WCAP-15025-P-A. The WCAP-15025-P-A [page 4-2] pressure applicability range is 1495 to 2425 psia.

DOM-NAF-2, Rev. 0.2-A, APPENDIX C C-22

C.7 REFERENCES S

C1 .Technical Report, WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids," L.D. Smith, et al, April 1999.

C2.Technical Report, "Tables for Normal Tolerance Limits, Sampling Plans and Screening," R.

E. Odeh and D. B. Owen, 1980.

C3. Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), "Acceptance

for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

l C4.Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 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.

5!

S M79498)," October 30, 1993.

C5.Fleet Report, DOM-NAF-2-A, including Appendixes A and B, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," Rosa M. Bilbao y Le6n, August 2006.

(ML062650184) 0

) C6.Letter from D. S. Collins (NRC) to J. A. Gresham (Westinghouse), "Modified WRB-2 Correlation WRB-2M for Predicting Critical Heat Flux in 17X17 Rod Bundles with Modified LPD Mixing Vane Grids," February 3, 2006.

S

l C7.Technical Report, "Boiling Crisis and Critical Heat Flux," TID-25887, 1972.

I C8.Letter from A.C. Thadani (NRC) to W.J. Johnson (Westinghouse), "Acceptance for

I Referencing of Licensing Topical Report, WCAP-9226-P, Reactor Core Response to

Excessive Secondary Steam Releases." 1989.

C9.Technical Report, "Assessment of the Assumption of Normality (employing individual

observed values)," American National Standards Institute, ANSI N15.15.1974.

R 0

0 S

S 0

SDOM-NAF-2, Rev. 0.2-A, APPENDIX C C-23 S

S

0 0

0 I.

Fleet Report DOM-NAF-2, Rev. 0.2-A 0 Reactor Core Thermal-Hydraulics Using the 0 VIPRE-D Computer Code 0

0 Attachment I 0 0

0 Request for Additional Information Set 1 0 Questions and Responses 0

(31 pages) 0 0

0 0

0 DOM-NAF-2, Rev. 0.2-A, Attachment 1

e 1)0.,inion Resou.rces

-,*o Services, Inc..

Itinion lhonl*v.i'd. Ovi' Allcn, VA

!#'° D o m in io n -

VDt600 20

June 30, 2005 United States Nuclear Regulatory Commission Serial No.05-328

Attention: Document Control Desk NL&OS/ETS: R0 Washington, D.C. 20555 Docket Nos. 50-280/281 50-338/339

50-336/423 License Nos. DPR-32/37 NPF-4/7 DPR-65/NPF-49 0

VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

DOMINION NUCLEAR CONNECTICUT, INC. (DNCI

) NORTH ANNA AND SURRY POWER STATIONS UNITS 1 AND 2

) MILLSTONE POWER STATION UNITS 2 AND 3

REQUEST FOR ADDITIONAL INFORMATION ON TOPICAL REPORT DOM-NAF-2:

.REACTOR CORE THERMAL-HYDRAULICS USING THE VIPRE-D COMPUTER CODE

INCLUDING APPENDICES A AND B

]In letters dated September 30, 2004 and January 13, 2005 (Serial Nos.04-606 and

05-020, respectively), Dominion and DNC submitted the Topical Report DOM-NAF-2,

"Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code" and associated Appendices A and B, for NRC review and approval. VIPRE-01 is a core thermal-hydraulics computer code developed by EPRI, approved by the NRC, and currently in

use throughout the nuclear industry. VIPRE-D is the Dominion version of VIPRE-01,

which has been enhanced by the addition of several vendor specific CHF correlations.

In a May 19, 2005 letter, the NRC requested additional information to complete their review of VIPRE-D and the associated appendices.

S

On May 25, 2005 Dominion and NRC held a public meeting on VIPRE-D licensing issues at which Dominion discussed the proposed scope for the RAIs. The NRC staff

agreed that the proposed scope for the responses was acceptable. The attachment to

) this letter provides the detailed responses discussed in the scope at the public meeting, including appropriate references and supporting information. If you have further questions or require additional information, please contact Mr. Thomas Shaub at (804)

273-2763.

0

Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services

Virginia Electric and Power Company

Dominion Nuclear Connecticut, Inc.

S 0

S S

Serial No.05-328 W Docket Nos. 50-280/281/338&339/336/423 0 Page 2 of 3 Commitments made in this letter: None 0 cc: U.S. Nuclear Regulatory Commission 0 Region II 0 Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW 0 Suite 23T85 Atlanta, Georgia 30303 0 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road 0 King of Prussia, Pennsylvania 19406-1415 0 Mr. J. T. Reece (w/o Att.) 0 NRC Senior Resident Inspector 0 North Anna Power Station Mr. N. P. Garrett(w/o Att.)

NRC Senior Resident Inspector Surry Power Station Mr. S. M. Schneider (w/o Aft.)

NRC Senior Resident Inspector Millstone Power Station Mr. V. Nerses 0 NRC Senior Project Manager - Millstone Unit 2 U. S. Nuclear Regulatory Commission One White Flint North 1 11555 Rockville Pike 0 Rockville, MD 20852 Mr. G. F. Wunder 0 NRC Senior Project Manager - Millstone Unit 3 0 U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike 0 Rockville, MD 20852 Mr. R. E. Martin NRC Lead Project Manager - North Anna and Surry 0 U. S. Nuclear Regulatory Commission One White Flint North 0 11555 Rockville Pike 0 Rockville, Maryland 20852 0 O

St 0

V Serial No.05-328 Docket Nos. 50-280/281/33813391336/423 Page 3 of 3

Mr. Stephen R. Monarque

NRC Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North

11555 Rockville Pike

Rockville, Maryland 20852 Mr. J. Honchark

NRC Project Manager- North Anna

U. S. Nuclear Regulatory Commission One White Flint North

11555 Rockville Pike

Rockville, Maryland 20852 S

S S

0 S

S S

S 0

0 S

S

S 0

Serial No.05-328 S Docket Nos. 50-280/281/338/339/336/423 S S

S S

COMMONWEALTH OF VIRGINIA ))

COUNTY OF HENRICO )

S S

The foregoing document was acknowledged before me, in and for the County and S Commonwealth aforesaid, today by Eugene S. Grecheck who is Vice President - 0 Nuclear Support Services, of Virginia Electric and Power Company and Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of these companies, and that the 0 statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this - day of 1,.2005. 0 S

My Commission Expires: ,e&4;c,/,O66 S

0 0

S S

Notary P'ublic S S

S S

0 S

S (SEAL) S 0

S S

0 0

S S

S 0

0 0

0 0 Attachment 0 (Serial No.05-328) 0 S Request for Additional Information S Topical Report DOM-NAF-2 Reactor Core Thermal-Hydraulics Using the 0 VIPRE-D Computer Code 0

S 0

0 0

S S

0 0

0 S

0 0 Virginia Electric and Power Company (Dominion) 0 Dominion Nuclear Connecticut (DNC)

S

6 0

S REQUEST FOR ADDITIONAL INFORMATION ,

PROPOSED TOPICAL REPORT DOM-NAF-2 0 0

VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

DOMINION NUCLEAR CONNECTICUT (DNC)

NRC Question 1 00 Departure from Nucleate Boiling Ratio (DNBR) is sensitive to the turbulent mixing coefficient, Please justify the value of the turbulent mixing coefficient. Stating it is conservative is insufficient. Please show the sensitivity of the DNBR to the turbulent mixing coefficient Dominion/DNC Response: 0 The turbulent mixing coefficient (ABETA) is obtained empirically by the fuel vendor.

Both AREVA and Westinghouse have conducted several subchannel mixing tests in pressurized water loops at Reynolds numbers similar to that of a PWR core under single phase and two-phase flow conditions (References 13, 14, 15) to determine ABETA. It has been determined experimentally that the value of ABETA is a function of grid spacing, and as such, it is dependent on the fuel product design. A turbulent mixing coefficient value of 0.038 has been validated by the NRC for analyzing Westinghouse 17 x 17 and 15 x 15 fuel assemblies with mixing vane grids having a spacer span of 26 inches or less (Reference 15). The Dominion submittal for the AREVA Fuel Transition (Section 4.2.5 in Reference 5) documented a turbulent mixing coefficient of 0.038 for the AREVA Advanced Mark-BW fuel. Therefore, a turbulent mixing coefficient value of 0.038 is applicable to current AREVA and Westinghouse fuel designs, and will be used by Dominion in VIPRE-D models for current fuel products. The turbulent mixing coefficient used by VIPRE-D models for future fuel designs will be provided by the fuel vendor.

As agreed during the May 25, 2005 Dominion-NRC public meeting, the sensitivity of the DNBR to the turbulent mixing coefficient has not been provided herein.

NRC Question 2 Please describe the basis for the use of the drift flux correlationsemployed as part of the Electric Power Research Institute (EPRI) void model. Please describe the flow regime/regimes that the EPRI drift velocity correlation is applicable to and show that this applies to the flow regimes experienced during DNB in the plant analyses. Comparisonsof the drift velocity correlation/correlationsto void data in rod bundles and small pipes would be desirable. This could be done using all other particularinputs and correlation choices (and code corrections) included to show the effect/ability to continue to predict the test data presented in Volume 4 of the VIPRE Manual entitled "Applications"dated 1987. Since Page 1 of 26 5 0

S S

0 0

the EPRI void model appears to employ only the drift velocity correlation applicable to

churn turbulent bubbly flow (for void fractions less than 0.3), please explain and justify why

this correlation is applied to slug and annular flow where Critical Heat Flux (CHF) can

occur. Define the limitations of the drift flux correlations (i.e. pressure range, flow conditions, etc.). What distributionparameteris assumed? Since more voiding may occur

near the walls of the hot rods, how does the distribution parameter account for this

condition in the drift flux modeling? Pleaseexplain.

0 Dominion/DNC Response:

VIPRE-01 has been approved by the NRC (References 10 and 11). VIPRE-D, which is based upon VIPRE-01, MOD-02.1, was developed by Dominion to fit the specific needs of Dominion's nuclear plants and fuel products by adding vendor specific CHF correlations

and customizing its input and output. Dominion, however, has not made any modifications

to the NRC-approved constitutive models and algorithms in VIPRE-01. Therefore, none of the models for bulk void, subcooled void and two-phase friction factor present in VIPRE-01 0* has been modified in any way in VIPRE-D, including the EPRI models.

The EPRI models for bulk void, subcooled void and two-phase friction factor are described in detail in the VIPRE-01 documentation (Section 2.6 of Reference 16 and Section 2.7 of Reference 7), including their underlying assumptions and ranges of

validity. Sections 2.5 and 3.0 of Reference 9 document the VIPRE-01 comparisons

performed by the code developer. VIPRE-01 two-phase friction factor models were evaluated against experimental data from the FRIGG rod bundle test loop. VIPRE-01

void models were evaluated against experimental data from the FRIGG rod bundle test

loop, ANL void test and Martin void measurements at high pressure. The comparisons

show that although all the models available in VIPRE-01 match the experimental data reasonably well, the EPRI set of correlations compare more favorably with the measured data. In consequence, the EPRI models for bulk void, subcooled void and two-

phase friction factor are the default selections in VIPRE-01 for PWR analysis.

Based on the evaluation of the benchmark calculations discussed above, NRC staff

t concluded that the EPRI models for bulk void, subcooled void and two-phase friction are

acceptable for use in licensing calculations (Reference 10).

The benchmark studies mentioned above, as well as the NRC approval of the EPRI

models provided Dominion a good starting point for the selection of void and two-phase

friction multiplier models. In addition, Dominion performed another sensitivity study to

determine the most suitable set of models for Dominion applications. This sensitivity analysis, which is summarized in Section 5.4 of DOM-NAF-2, provides justification for

Dominion's modeling selections, thus fulfilling condition (3) of the SER for VIPRE-01

MOD-01 (Reference 10).

Dominion performed. a detailed analysis of the correlations available in the code, and

four sets of correlations were chosen based on the compatibility of the modeling

assumptions used in deriving the various correlations. The selected sets use together only those correlations that have consistent or complementary bases and take advantage of previous industry experience and vendor recommendations. The four

cases studied were:

Page 2 of 26 0

0

Case 1 (EEE)

Subcooled Void Model: EPRI Bulk Boiling Void Model: EPRI Two-Phase Friction Multiplier: EPRI Case 2 (LSE)

Subcooled Void Model: LEVY Bulk Boiling Void Model: SMITH Two-Phase Friction Multiplier EPRI Case 3 (LHH)

Subcooled Void Model: LEVY Bulk Boiling Void Model: HOMOGENEOUS Two-Phase Friction Multiplier: HOMOGENEOUS Case 4 (LSH)

Subcooled Void Model: LEVY Bulk Boiling Void Model: SMITH Two-Phase Friction Multiplier: HOMOGENEOUS Although other relevant parameters such as void fraction or quality were verified throughout the benchmark evaluation, Dominion's acceptance criterion was based on the DNBR performance of the analyzed cases. The EPRI models for bulk void, subcooled void and two-phase friction factor provided the best DNBR comparison for all the following:

0

VIPRE-D/BWU against the CHF experimental database for the AREVA BWU CHF correlations (DOM-NAF-2 Appendix A).

VIPRE-D/BWU against the LYNXT/BWU code results for a set of representative operating conditions (DOM-NAF-2 Section 5.0)

VIPRE-D/WRB-1 against the CHF experimental database for the Westinghouse WRB-1 CHF correlation (DOM-NAF-2 Appendix B).

VIPRE-D/WRB-1 against the COBRA/WRB-1 code results for a set of representative operating conditions (DOM-NAF-2 Appendix B Section B.7). 9 Consistent with the results of these external and in-house benchmark studies, Dominion has demonstrated that the selection of the EPRI set of void and two-phase friction models is acceptable for application in DOM-NAF-2.

During the May 25, 2005 NRC-Dominion public meeting, Dominion agreed to provide some graphic comparisons to show VIPRE-D's performance for void fraction and quality.

Figures 1 and 2 provide a comparison between the VIPRE-D/WRB-1 and the 0 COBRA/WRB-1 codes. In particular, these figures show the void fraction and quality axial distributions in the hot channel at the time of minimum DNBR (3.0 seconds) for a Locked Rotor Transient event at Surry Power Station. Additional description and results S associated with this transient were included in Section B.7.3 of Appendix B to DOM-NAF-

2. In addition to the VIPRE-D/WRB-1 with the EPRI-EPRI-EPRI set of models, Figures 1 Page 3 of 26 0

0 0

0 and 2 show VIPRE-D/WRB-1 with the LEVI-SMITH-HOMOGENEOUS set of models, which is the set of bulk void, subcooled void and two-phase friction models closest to 0 COBRA/WRB-1 modeling in Reference 12 (LEVY-SMITH-BAROCZY).

0 Figures 1 and 2 show that the different void and two-phase friction models result in comparable void fraction and quality trends for COBRA and VIPRE-D. However, as noted above, DNBR comparisons are the main criterion used to evaluate the performance of VIPRE-D against other thermal-hydraulics codes and models. These comparisons 0 demonstrated the acceptability of the EPRI models for bulk void, subcooled void and two-phase friction factor for use in the Dominion VIPRE-D models.

0 S

0 Figure 1: Locked Rotor Transient Equilibrium Quality Results at Time 3.0 Seconds 0.2~ - -- -- - -------------

0 ° '. 1- **, -- ............... ...

0 0

0 20 4 0

-0.1 .....-.

0 COBRA 5 -V)PRE EEE

/'00 - - - VIPRELSH 0 -02 O

-0.3 O

0 -0.4 0

Axial Height (in) 0 Page 4 of 26

S Figure 2: Locked Rotor Transient Void Fraction Results at Time 3.0 Seconds 0

0.35 0 0.3 0

9 0.25..

iI 0.2-

0 C 0

-COBRA

o U-0.15- -VIPRE EEE 0

-- VIPRE LSH 0.1 0 0.05

,--/- I -- - i 0 i, ,i 1 /./ - I !

S 0

40,' 8 1U o 1io0 40 160

- I -,-----------L----' ...

S

-0.05 Axial Height (In) 0 0

9 0

0 0

Page 5 of 26

S 0

0* 'NRC Question 3 Please explain the basis for choosing the EPRI bulk void model. Please describe the transient test cases employed to determine the values given in Table 5.4-1 and show the

void distributions for the VIPRE and LYNXT codes at the initiation.of DNB in some

example cases. Also provide justification for the choice of the subcooled boiling model as

well as the two-phase friction multiplier. Pleaseshow comparisonsof the VIPRE-D code to data for these models. Please also describe the limitations and identify the ranges of applicabilityof each of these correlations.

Dominion/DNC Response:

The basis for choosing the EPRI bulk void, subcooled void and two-phase friction

multiplier models is provided in the response to Question 2.

The transient test cases employed to determine the values in Table 5.4-1 of

DOM-NAF-2 are the same statepoints that were used in Section 5.1 of DOM-NAF-2 to

benchmark VIPRE-D/BWU against LYNXT/BWU, and were also used to support the

Licensing Amendment Request (LAR) for the Framatome ANP Fuel Transition (Reference 5). These statepoints were obtained from the UFSAR Chapter 15 events

including the reactor core safety limits, axial offset envelopes (AO's), rod withdrawal at

1 power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss

of flow accident (LOFA), and locked rotor accident (LOCROT) events. 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 (reactor core safety limits); (b) axial power shapes (AOs); (c)

elevated hot rod power (misaligned rod); and (d) low flow (LOFA and LOCROT). These statepoints cover the full range of conditions and axial offsets in UFSAR Chapter 15

evaluations (except for MSLB that was evaluated separately), and were specifically

selected to challenge the three BWU CHF correlations.

In the May 25, 2004 NRC-Dominion public meeting, Dominion indicated that LYNXT

models use different (proprietary) void models, and that as a consequence, void

distribution comparisons for VIPRE-D/BWU and LYNXT/BWU would not be expected to match. A comparison of the void and quality distributions for COBRANWRB-1 and VIPRE-D/WRB-1 is provided in the response to Question 2.

0 NRC Question 4

In section 4.12, please explain what "nearly identical" means. Identifying the percent

difference between the two results would be helpful or show the plot of the two DNBR calculations.

Dominion/DNC Response:

In the few occasions in which V1PRE-D did not converge when using the default values of

the convergence criteria and damping factors, and these criteria and/or damping factors

Page 6 of 26 0

were adjusted to ensure numerical convergence, the difference in the DNBR results reported by the code was 0.5% or less.

NRC Question 5 Section 5.3 describes a comparison with the COBRA code but states the Minimum DNBR (MDNBR) results are different because the analyses use different fuel types and CHF correlations. Please provide the latest comparisons between the codes using the same fuel type and CHF correlations. Please show the channel void distribution and quality at several selected times during the events. Show the steamline break, feedline break, and loss-of-flow events.

Dominion Response: 0 0

Section B.7 in Appendix B to Topical Report DOM-NAF-2 includes comparisons of VlPRE-D/WRB-1 with COBRAiWRB-1 for Surry Power Station. These comparisons use the same fuel type (Westinghouse 15x15 SIF) and the same CHF correlation (Westinghouse WRB-1).

The comparisons include:

A statepoint safety analysis evaluation of 164 statepoints obtained from the Surry UFSAR Chapter 14 events, including the reactor core safety limits, axial offset envelopes (AO's), rod withdrawal at power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss of flow accident (LOFA), and locked rotor accident (LOCROT) events.

A Main Steam Line Break statepoint application.

A Feedwater Malfunction Transient (FWMAL).

A Locked Rotor Transient (LOCROT).

Selected examples comparing channel void and quality distributions were provided in the 0 answer to Question 2.

0 NRC Question 6 The qualification document identifies the DNBR limits for the correlations for several pressure groups. Please explain what DNBR limit is applied or how the situation is 0 handled when the range of validity is exceeded for the other parametersidentified in Table A.5.2. Please also define the quality rangefor the correlation.

Dominion/DNC Response:

The qualification of the BWU CHF correlations with the VIPRE-D code was performed consistent with AREVA's approved Topical Report (Reference 4). The evaluations.

performed for the various pressure groups as well as the resulting limits were included in Appendix A for completeness. However, Dominion will only apply the DNBR limits listed in Table A.5-1:

0 Page 7 of 260 0

0

S 0

VlPRE-DIBWU-Z 0

DNBR limit below 700 psia 1.59 DNBR limit 700 - 2,400 psia 1.20 VIPRE-DIBWU-ZM 0 DNBR limit below 594 psia 1.59 DNBR limit at or above 594 psia j 1.18 VIPRE-D/BWU-N 0 DNBR limit below 1200 psia 1.39 S DNBR limit at or above 1200 psia 1.22 0

0 The ranges of validity for the three BWU CHF correlations, including quality, are listed in 0 Table A.5-2:

BWU-Z BWU-ZM BWU-N Pressure 400 to 2,465 400 to 2,465 788 to 2,616

[psia] ---

Mass Velocity

[Mlbm/hr-ft 2] 0.36 to 3.55 0.47 to 3.55 0.25 to 3.83 Thermodynamic Less than 0.74 Less than 0.68 Less than 0.70 Quality at CHF I Applicability Mixing Vane Mid-Span Mixing Non-Mixing Vane Grids Grids Grids The Low Flow Main Steam Line Break (MSLB) statepoint has been identified as the only event that might fall outside of the range of validity of the BWU correlations (for the mass velocity). This issue was already identified in the Framatome Fuel Transition Submittal (Reference 5). The Low Flow MSLB statepoint results in minimum DNBR values of the order of 5.8, which are obviously very far from being limiting. In this case Dominion will use the DNBR limit corresponding to the pressure at which the event takes place.

NRC Question 7 CHFis also sensitive to the axial power distribution. Since the correlationswere developed from data with uniform or symmetric power distributions,please justify applicabilityof the correlations to the asymmetric power distributions that may be limiting in the plant calculations. Explain how the correlationsare applied and describe any correction factors that may be applied to accommodateskewed distributions.

DominionlDNC Response:

This issue is a generic concern with all CHF correlations, since these correlations were developed using a limited number of uniform and non-uniform axial power distributions.

Both the AREVA BWU set of correlations (qualified with VIPRE-D in Appendix A) and Page 8 of 26

0 0

the Westinghouse WRB-1 correlation (qualified in Appendix B) include a correction factor, the F-factor proposed by Tong (Reference 8), that is used to correct for non-uniform axial power shapes. The Tong Factor has been extensively verified by both Westinghouse (Reference 6) and AREVA (Reference 4) with numerous CHF test data, that over the years have included uniform and non-uniform power shapes. The use of this F-factor has been previously approved by the NRC staff.

In addition to the verifications performed by both vendors, Dominion performed a statepoint safety analysis evaluation using symmetric and non-symmetric, positively and negatively skewed axial power shapes to benchmark VIPRE-D/BWU with LYNXT/BWU and VIPRE-D/WRB-1 with COBRAIWRB-1. The results of these evaluations do not show differences in the performance of the CHF correlations dependent on the form or uniformity of the axial power distributions.

NRC Question 8 Please justify applicability of the steady-state DNB correlations to steamline breaks since these events have rapid depressurizations where the steady-state correlations may not be applicable. These transients may also transition through slug and annular flow. As such, please justify the use of the EPRI bulk boiling drift flux model since it only applies to bubbly flow.

Dominion/DNC Response:

The issue of the applicability of steady-state CHF correlations to transient analysis is generic to all thermal hydraulics codes and CHF correlations. Section 6.7 of Reference 9 provides a study to determine the applicability of steady-state CHF correlations to transient analysis for VIPRE-01. The NRC staff review concluded that the studies have shown that the transient CHF for power ramp and flow coastdown transients are higher than the steady-state CHF, and that, except for very rapid depressurization events (LOCA), the use of CHF correlations developed with steady-state CHF data can correctly or conservatively predict the transient CHF when the instantaneous local fluid conditions are used (Reference 10). S Dominion's Main Steam Line Break (MSLB) DNBR evaluations are performed at the limiting statepoint. Dominion does not perform full DNBR transient analysis for this event. For MSLB all the DNB limiting statepoints occur after the pressurizer has drained and the upper head has flashed. Therefore, pressure is changing relatively slowly and the quasi steady-state assumption is appropriate.

The use of the EPRI bulk boiling void correlation was justified in the response to Question 2.

NRC Question 9 Please discuss whether the slip option will be used and if so justify the slip ratio employed in the DNBR calculations.

0 Page 9 of 26 S

0 0

DominionlrDNC Response:

Dominion VIPRE-D models do not use the slip option.

NRC Question 10 Since the conduction model will not be used, please explain how the stored energy in the rod is accounted for. Please explain why use of the dummy rod model is conservative since the conduction model does not include the effects of gap conductance and initial stored energies. Please also describe how the heat flux is calculatedfor use as input to the VIPRE-D code.

DominionIDNC Response::

Conduction models are typically used to perform fuel temperature calculations and to simulate delay of energy transport. Dominion does not plan to use VIPRE-D for these applications. The use of the dummy rod model is consistent with previously approved Dominion methodologies (Reference 12).

The use of the dummy rod model requires the user to provide the fuel rod surface heat flux as one of the operating input conditions. For steady-state statepoint analysis this value is easily calculated based on core thermal power. Fuel rod surface heat flux forcing functions for transient calculations are provided by an NRC-approved transient system code (e.g., RETRAN). Transient system codes account for fuel conduction, gap conductance, and delayed energy transport effects. Therefore, the use of the dummy rod model appropriately includes the relevant effects into the analysis.

Page 10 of 26

S 0

Appendix A to DOM-NAF-2 NRC Question 1 The appropriatestatisticalanalysis of the data, which form Tables A.3.1-1, A.3.2-1, and A.3.3-1 is an analysis of variance of a mixed-effects model.

a) Give the appropriateanalysis of variance tables for these mixed-effects models.

b) Formulate the appropriatestatisticalhypothesis tests to justify the values for MIP and alp used in Eq. A. 1.1 based on the data in Tables A.3. 1-1, A.3.2-1, and A.3.3-1, c) For those cases where individual DNBR design limits were developed for each low pressuregroup, how were the results of the above analysis of variance taken into account?

Dominion/DNC Response:

0 As clarified in our May 25, 2005 meeting with the NRC staff, this question refers to Tables A.4.1-2, A.4.2-2 and A.4.3-2.

a) The qualification of the BWU CHF correlations with the VIPRE-D code was performed consistent with AREVA's approved Topical Report (Reference 4). An analysis of variance (ANOVA) test was performed for each one of the three BWU correlations. This analysis was not included in the original submission because it was not deemed to provide additional substantial information to the qualification.

0 An ANOVA test divides the database in several groupings according to a given variable and then evaluates whether or not the distributions for each one of the groups appear to belong to the overall distribution. If all the groupings belong to the same distribution it can be deduced that the total population does not show a bias with respect to that particular variable Even though it was recognized that ANOVA tests cannot be used as the sole measure of the performance of a CHF correlation, they can be useful to indicate an extremely bad mismatch (very large F statistic).

The variables analyzed were mass velocity, pressure, quality, test cell type and axial flux shape type.

The results of the ANOVA tests for the BWU-Z, BWU-ZM and BWU-N correlations are provided in Tables 1 through 5. The results for the BWU-ZM (Table 3) and BWU-N (Table 4) correlations prove the ANOVA hypothesis: all the groups analyzed belong to the same distribution, i.e. there is no bias of the results regarding the analyzed variables. The results for the BWU-Z correlation (Table 1) show F values slightly above the critical value, but still reasonably small. Following the AREVA approach in Reference 4, an additional ANOVA test was performed excluding the low pressure data from the BWU-Z database. This treatment is consistent with the fact that the low pressure data were taken separately by determining a separate DNBR design limit. Table 2 summarizes the results of this second ANOVA analyses.

While the values of the F statistic do decrease somewhat in most cases, they are still slightly above the critical value of F for the appropriate number of degrees of freedom. An additional ANOVA test was also performed for BWU-N without the low 0 pressure data (Table 5), and the trends observed are the same as those shown in Table 4. For all cases, these are the same trends showed by AREVA in Reference 4.

0 0

Page 11 of 260 0

0

Table 1: MWP CHF Performance by Independent Variable Grouping using the entire BWU-Z database at 95% confidence level 0 Number Average Standard - Maximum Minimum 0 of Data MWP Deviation MWP M/P Analysis by Mass Velocities 0.5 Mlbm/hr-ft 2 9 0.9559 0.1147 1.1083 0.7833 0 1.0 Mlbm/hr-ft 2 90 1.01 20 0.1088 1.2974 0.6980 1.5 Mlbm/hr-ft2 97 0.9934 0.0726 1.1609 0.8321 2.0 Mlbm/hr-ftz 134 0.9721 0.0878 1.1 612 0.7717 2.5 Mlbm/hr-ftz 85 1.0018 0.0908 1.1693 0.7667 3.0 Mlbmrhr-ft 2 71 1.0216 0.0891 1.1481 0.7261 3.5 Mlbm/hr-ft2 42 0.9853 0.0749 1.1148 0.8387 Fdistrbution- = 3.4995 Fcrtical(6,521) 2.1159 .......

0 Analysis by Pressures Below 1250 psia 64 1.0437 0.0936 1.2974 0.8342 1250 - 1649 psia 111 0.9789 0.0881 1.1683 0.7770 1650 - 1949 psia 108 0.9722 0.0916 1.1693 0.7667 0 1950 - 2249 psia 116 1.0089 0.0914 1.1481 0.6980 Above 2250 psia 129 0.9914 0.0796 1.1576 0.7669 0 Fcfstrlbution = 8.3625 Fcritcai.4,523) = 2.3889 0 Analysis by Qualities Below 5% 34 0.9634 0.0794 1.0840 0.7261 5% to 10% 84 1.0415 0.0740 1.1693 0.8306 10% to 15% 127 0.9886 0.0884 1.1576 0.7717 15% to 20% 122 0.9750 0.0774 1.1479 0.8126 20% to 25% 90 0.9755 0.0913 1.1411 0.6980 25% to 30% 29 0.9969 0.1039 1.1609 0.7669 0 30% to40%

Above 40%

18 24 1.0343 1.0553 0.1006 0.1198 1.1642 1.2974 0.8484 0.8342 0 Fstr-.boi.= 7.9433 F.~1~1(7,520) 2.0271 0 Analysis by Test Cell Type 0 Unit Cell 218 1.0077 0.0894 1.1683 0.7669 0 Guide Tube . 170 0.9849 0.0917 1.1693 0.7261 0 Other 140 0.9877 0.0899 1.2974 0.6980 Fdistributon- = 3.6760 Fcrqbcai(2,525) = 3.0128 All Data BWU-Z All Data 528 0.9950 0.0907 1.2974 0.6980 Page 12 of 26

0 0

0 Table 2: M/P CHF Performance by Independent Variable Grouping using the BWU-Z database without the low pressure data at 95% confidence level 0 Number Average Standard Maximum Minimum 0 of Data M/P Deviation M/P MIP 0

nalysis by Mass Velocities 1.25 Mlbm/h r-f 70 0.9909 0.1025 1.1457 0.6980 0 1.5 Mlbm/hr-ft 2 86 0.9876 0.0700 1.1576 0.8321 0 2.0 Mlbm/hr-ft 2 119 0.9626 0.0853 1.1492 0.7717 2.5 Mlbm/hr-ft2 77 0.9994 0.0945 1.1693 0.7667 0 3.0 Mlbm/hr-fte 70 1.0201 0.0889 1.1481 0.7261 3.5 Mlbm/hr-ftz 42 0.9853 0.0749 1.1148 0.8387 0

_Fdistributjon = 4.2356 Fcriai(5,458) 2.2337 0 0

Analysis by Pressures 1250- 1649"psia 111 0.9789 0.0881 1.1683 0.7770 0 1650- 1949 psia 108 0.9722 0.0916 1.1693 0.7667 1950 - 2249 psia 116 1.0089 0.0914 1.1481 0.6980 0 Above 2250 psia 129 0.9914 0.0796 1.1576 0.7669 0

Fdistribution = 3.8436 Fcbul 3,460) = 2.6243 0

Analysis by Qualities -- 0

< 10% 116. 1.0186 0.0839 1.1693 0.7261 10%-15% 122 0.9859 0.0887 1.1576 0.7717 0

15%-20% 115 0.9737 0.0787 1.1479 0.8126 0

> 20% 111 0.9746 0.0951 1.1411 0.6980 0 Fdistribution = 6.7560 Fcuc 1(3,460) = 2.6243 0

Analysis by Test Cell Type .... 0 Unit Cell 188 1.0024 0.0880 1.1683 0.7669 Guide Tube 150 0.9810 0.0931 1.1693 0.7261 0 Other 126 0.9761 0.0805 :1.1457 0.6980 0 Fdistributilon = 4.1660 Fcrifcal(2,461 = 3.0153 0

BWU-Z without low pressure data 0 All Data 464 0.9883 0.0883 1.1693 0.6980 0

0 0

0 Page 13 of 26

Table 3: M/P CHF Performance by Independent Variable Grouping using the entire BWU-ZM database at 95% confidence level Number Average Standard Maximum Minimum of Data MIP Deviation M/P M/P Analysis by Mass Velocities 0 <1.25 Mlbm/hr-ft 2 36 1.0338 0.0712 1.2110 0.8828 0 1.5 Mlbrn/hr-fte 32 1.0256 0.1039 1.2299 0.8142 2.0 Mlbm/hr-fte 41 0.9859 0.0890 1.1467 0.8255 2.5 Mlbm/hr-ft 2 18 0.9918 0.0843 1.1607 0.7793 S 3.0 Mlbm/hr-ft2 10 1.0480 0.0809 1.1699 0.9612 3.5 Mlbm/hr-ftr 11- 1.0228 0.0676 1.1324 0.9195 0 Fditrbution = 1.9439 Fcriicaj(5,142) = 2.2779 Analysis by Pressures_'

S < 1000 psia 11 0.9758 0.0545 1.0749 0.8832 1000- 1500 psia 15 1.0202 0.0840 1.2110 0.9125 0 1500 - 2000 psia 53 1.0249 0.0982 1.1844 0.8142

2000 psia 69 1.0099 0.0833 1.2299 0.7793 Fdistribution = 1.0528 Fcrifical(3,144) = 2.6674 S

Analysis by Qualities 0 < 10% 11 1.0004 0.0622 1.1249 0.9195 S 10%-15% 39 1.0111 0.1023 1.1699 0.7793 15%-20% 35 1.0082 0.0737 1.1680 0.8783

> 20% 63 1.0209 0.0897 1.2299 0.8496 9 F'cstributjon = 0.2794 Fcnvcai(3,144) = 2.6674 0 Analysis by Test Cell Type -

0 Unit Cell 76 1.0230 0.1000 1.2299 0.8142 Guide Tube 72 1.0041 0.0715 1.1747 0.7793 Fditibuton = 1.7240 Fcrica1(1, 146) = 3.9059 0 All Data BWU-ZM All Data 148 1.0138 0.0875 1.2299 0.7793 0

0 0

0 Page 14 of 26

0 0

0 Table 4: M/P CHF Performance by Independent Variable Grouping 0

using the entire BWU-N database at 95% confidence level 0 Number Average Standard Maximum Minimum 0 of Data MIP Deviation M/P M/P Analysis by Mass Velocities 0

0.5 Mlbm/hr-ft 2 147 1.0052 0.1469 1.3507 0.6707 1.0 Mlbm/hr-W 172 0.9870 0.1331 1.3251 0.6635 0 1.5 Mlbrn/hr-ftV 194 1.0018 0.0985 1.2983 0.7262 2.0 Mlbm/hr-ft2 208 1.0086 0.0758 1.2541 0.7560 0 2.5 Mlbm/hr-ft2 149 0.9980 0.0808 1.2061 0.7747 0 3.0 Mlbm/hr-ft 124 1.0050 0.0767 1.1845 0.8154 0 3.5 Mlbm/hr-fte 96 1.0102 0.0882 1.2826 0.8002 FdIstribufon 0.9171 Fcriici(6 ,1083) = 2.1069 0 Analysis by Pressures 0 Below 1250 psia 40 1.0309 0.1066 1.2569 0.6985 1250 - 1649 psia 192 1.0092 0.1110 1.3251 0.7396 0 1650 - 1949 psia 198 0.9900 0.0999 1.3507 0.7372 0 1950 - 2249 psia 446 1.0021 0.1002 1.2868 0.6635 Above 2250 psia 214 1.0002 0.1069 1.3030 0.6930 0

Fdisrjbution = 1.6829 Fcrt'ca (4,1085) = 2.3801 _

0 Analysis by Qualities; 0 Below 5% 222 1.0129 0.0788 1.1834 0.7867 5% to 10% 192 0.9993 0.0758 1.1845 0.8041 0 10% to 15% 234 1.0051 0.0885 1.2787 0.7372 0 15% to 20% 158 1.0110 0.1029 1.3030 0.7747 0 20% to 25% 106 0.9766 0.1393 1.3251 0.6746 25% to 30% 45 0.9821 0.1371 1.2983 0.6635 30% to 40% 56 0.9915 0.1341 1.2917 0.7264 Above 40% 77 1.0008 0.1533 1.3507 0.6707 Fdlstrlbuton = 1.8027 Fcritai(7,1082) = 2.0180 Analysis by Test Cell Type Unit Cell 553 1.0031 0.1140 1.3251 0.6635 Guide Tube 385 1.0012 0.0889 1.2003 0.6707 Intersection 152 0.9988 0.1004 1.3507 0.8002 SFdisufiuon = 0.1133 Fc,. 1 (2,1087) = 3.0040 Analysis by Axial Flux Shape Type Uniform 355 0.9915 0.1200 1.3251 0.6635 0 Non Uniform - 568 1.0069 0.0968 1.3507 0.6985 0 Symmetric 0 Non Uniform - 167 1.0063 0.0875 1.2003 0.7346 Asymmetric I 0 Fdistribution = 2.6068 Fcriicai(2,1087) = 3.0040 All Data BWU-N 0

All Data 1090 1.0018 0.1038 1.3507 0.6635 0 0

0 0

0 Page 15 of 26 0 S

0

Table 5: M/P CHF Performance by Independent Variable Grouping using the BWU-N database without the low pressure data at 95% confidence level Number Average Standard- Maximum1 Minimum Grouping of Data M/P Deviation M/P M/P Analysis by Mass Velocities 0 1.25 Mlbrn/hr-ft 304 0.9945 0.1387 1.3507 0.6635 0 1.5 Mlbm/hr-ftl 188 1.0001 0.0995 1.2983 0.7262 0 2.0 Mlbm/hr-ft 2 202 1.0074 0.0765 1.2541 0.7560 2.5 Mlbrn/hr-fte 143 0.9961 0.0814 1.2061 0.7747 9 3.0 MlbnVhr-fte 117 1.0040 0.0772 1.1845 0.8154 3.5 Mlbm/hr-ft' 96 1.0102 0.0882 1.2826 0.8002 Fdistribution = 0.6302 Fcriicai(5,1'044) = 2.2227 0 Analysis by Pressures S 1250 - 1649 psia 192 1.0092 0.1110 1.3251 0.7396 1650 - 1949 psia 198 0.9900 0.0999 1.3507 0.7372 1950 - 2249 psia 446 1.0021 0.1002 1.2868 0.6635 Above 2250 psia 214 1.0002 0.1069 1.3030 0.6930 Fdistribution = 1.1604 Fc,iticat(3,1046) = 2.6134 Analysis by Qualities 0 <5% 222 1.0129 0.0788 1.1834 0.7867 9 5%- 10% 186 0.9986 0.0764 1.1845 0.8041 0 10%- 15%

15%-20%

227 151 1.0041 1.0076 0.0891 0.1027 1.2787 1.3030 0.7372 0.7747 20%-25% 102 0.9734 0.1409 1.3251 0.6746

> 25% 162 0.9924 0.1430 1 1.3507 0.6635

.Fdistribution = 2.4577 Fcrticai(5,1044) = 2.2227 0 Analysis by Test Cell Type S Unit Cell 537 1.0033 0.1141 1.3251 0.6635 0 Guide Tube 385 1.0012 0.0889 1.2003 0.6707 Intersection 128 0.9884 0.0979 1.3507 0.8002 F

Fdistribution = 1.0784 Fcribcal(2,1047) = 3.0043 Analysis by Axial Flux Shape Type Uniform 355 0.9915 0.1200 1.3251 0.6635 Non Uniform- 528 1.0051 0.0959 1.3507 0.7045 0 Symmetric ......

Non Uniform - 167 1.0063 0.0875 1.2003 0.7346 Asymmetnic I Fdistribution = 2.1303 Fcritcal(2,1047) = 3.0043 BWU-N without low pressure data All Data 1050 1.0007 1 0.1036 1.3507 0.6635 S

S 0

0 Page 16 of 26 S

0

0 0

0 b) As clarified in our May 25, 2005 meeting with the NRC staff, this question refers to Eq. A.4.1.1 and Tables A.4.1-1, A.4.2-1 and A.4.3-1. The validity of this equation is based on two assumptions: 1) the average M/P is 1.0 and 2) the 0 M/P distribution is normal. These two assumptions were demonstrated in Appendix A: S MIP DISTRIBUTION CORRELATION AVERAGE MWP = 1.0 NORMAL NORMAL BWU-Z Table A.4.1-1 Page A-13 0 BWU-ZM Table A.4.2-1 Page A-23 Page A-33 9 BWU-N Table A.4.3-1 (Hypernornnal distribution, 0 also seen by AREVA in Reference 4)

S S

c) Different DNBR limits were developed for low pressure groups for the BWU-Z S

and BWU-N correlations following the guidance in Reference 4. The plots (not S the ANOVA tests displayed in Tables 1 and 4 respectively) seemed to indicate a poorer performance of the correlation at low pressures, and for that reason a S separate DNBR limit was calculated at low pressures. In those cases, a new S ANOVA analysis was performed excluding the low-pressure data (Table 2 for BWU-Z and Table 5 for BWU-N). These ANOVA results were similar to the S previous results. All Dominion results were similar to AREVA results in S Reference 4. S S

NRC Question 2 S As in Appendix B, you state that the plots show that there are no biases in the MIP ratio 0 distributions, and that the performance of the CHF correlations is independent of the S three variables of interest. The plots show a mostly uniform scatter of the data and no obvious trends or slopes. The plots again suggest but do not demonstrate that the 0 claims made in those sentences are true. Please give the appropriate statistical 0 analysis, that demonstratesthe truth of the claim.

S Dominion/DNC Response: 0 As discussed in the response to Question 1 for Appendix A, an ANOVA analysis was 0 performed for each correlation to formally demonstrate that the performance of the S BWU CHF correlations is not biased by the three independent variables present in the correlations (mass velocity, pressure and quality). Please refer to the response to 0 Question 1 for analyses demonstrating no biases in the M/P distributions. 0 S

0 0

S Page 17 of 26 0 0

0 0

0 NRC Question 3 0

Please show the empirical probability density functions for the MIP values used in the 0 analyses together with the estimate of the 95-percent fractile for each correlation.

0 Dominion/DNC Response:

0 The probability density functions for the BWU-Z, BWU-ZM and BWU-N correlations are shown 0 in Figures 3, 4 and 5.

0 The 528 datapoints of the VIPRE-D/BWU-Z M/P distribution summarized in Section A.4.1 of DOM-NAF-2 Appendix A were used to create the empirical probability density function. These datapoints were distributed among 24 equal bins that covered the entire range of M/P in the 0 VIPRE-D/BWU-Z distribution, and the frequency of data in each bin was determined. The resulting empirical probability density function for the VIPRE-D/BWU-Z M/P distribution was 0 then compared with the probability density function of a normal distribution of mean 0.995 and standard deviation 0.0907, which are the mean and standard deviation calculated for the 0 VIPRE-D/BWU-Z MWP distribution in Section A.4.1 of DOM-NAF-2 Appendix A. Figure 3 also displays the obtained 95% fractile (1.13) for the data and the VIPRE-D/BWU-Z DNBR limit 0 obtained in Section A.4.1 of DOM-NAF-2 Appendix A (1.20).

S 0

Figure 3: VIPRE-D/BWU-Z Probability Density Function 0

S 14.00%

S 12.00%

0 0 10.00%

S 8.00%

0 6.00%

IL 0 4.00%

2.00%

0.00%

0.65 0.68 0.71 0.74 0.77 0.80 0.83 0.86 0.89 0.92 0.95 0.98 1.01 1.04 1.07 1.10 1.13 1.16 1.19 1.22 1.25 1,28 1.31 1.34 W/P I= VIPRE-D/BWU-Z CHF DATA "NORMAL(0.995, 0.0907)1 Page 18 of 26

0 S

S The 148 datapoints of the VIPRE-D/BWU-ZM M/P distribution summarized in Section A.4.2 of DOM-NAF-2 Appendix A were used to create the empirical probability density function. These datapoints were distributed among 27 equal bins that covered the entire 0 range of M/P in the VIPRE-D/BWU-ZM distribution, and the frequency of data in each bin was determined. The resulting empirical probability density function for the VIPRE-D/BWU-ZM M/P distribution was then compared with the probability density function of a normal distribution of mean 1.0138 and standard deviation 0.0875, which are the mean and 0

standard deviation calculated for the VIPRE-D/BWU-ZM M/P distribution in Section A.4.2 of 0 DOM-NAF-2 Appendix A. Figure 4 also displays the obtained 95% fractile (1.16) for the data and the VIPRE-D/BWU-ZM DNBR limit obtained in Section A.4.2 of DOM-NAF-2 Appendix A (1.18).

S Figure 4: VIPRE-D/BWU-ZM Probability Density Function S 12.00% -I i

oFRACTILE

__%F 116 I16 S

S 10.00% ---- '- *..*.I -

S

--'-I--"

8.00%

0 I VIPRE-D/BWU-ZM FOE I DNBR LIM,,1= 1.18 S V

Z w

0 6.00% 0 Lu UL 0 4.00%

S r _ _

0 2.00%

075 07 0 0 0 951 101 1 121 1 125. 127 0.00%-

0.75 0.77 0.79 0.81 0.83 0.85 0.87 0.89 0.91 0.93 0.95 0.97 0.99 1.01 1.03 1.05 1.07 1.09 1.11 1.13 1.15 1.17 1.19 1.21 1.23 1.25 1.27 MIP I VIPRE-D/BWU-ZM CH-F DATA - NORMAL(1.0138, 0.0875) 1 Page 19 of 26

S S

The 1090 datapoints of the VIPRE-D/BWU-N WP distribution summarized in Section A.4.3 of DOM-NAF-2 Appendix A were used to create the empirical probability density

function. These datapoints were distributed among 26 equal bins that covered the.entire

range of M/P in the VIPRE-D/BWU-N distribution, and the frequency of data in each bin was determined. The resulting empirical probability density function for the VIPRE-

D/BWU-N M/P distribution was then compared with the probability density function of a

normal distribution of mean 1.0018 and standard deviation 0.1038, which are the mean and standard deviation calculated for the VIPRE-D/BWU-N M/P distribution in Section

' A.4.3 of DOM-NAF-2 Appendix A. Figure 5 also displays the obtained 95% fractile (1.16)

for the data and the VIPRE-D/BWU-N DNBR limit obtained in Section A.4.3 of DOM-NAF-

2 Appendix A (1.22).

0 Figure 5: VIPRE-DIBWU-N Probability Density Function 14.00% 951/%FOACTI*LE- 1.16 12 .0 0% ..................... . ........... ..... ....

0 ~10.00% I 8.00% .... ...... ..

__ I ~ IIPF -DI~

600% I U.

0 _ VP*-i 4.O 2.00% ,

o.oo%4 0.64 0.67 0.70 0.73 0.76 0.79 082 0,85 0.88 0.91 0.94 0.97 1.00 1.03 1.06 1.09 1.12 1.15 1.18 1.21 1.24 1.27 1.30 1.33 1.36 1.39 M/P E IVIPRE-D/BWU-N CHF DATA NORMAL(1.01 8, 0.1038)

P 0

0 0

Page 20 of 26 0

0

Appendix B VIPRE-D NRC Question 1 The appropriatestatisticalanalysis of the data that form Table B.6-1 is an analysis of 49 varianceof a mixed-effects model.

a) Give the appropriateanalysis of variance table for this mixed-effects model.

b) Formulate the appropriatestatisticalhypothesis tests to justify the values for MIP and om used in Eq. B.6.1.

Dominlon/DNC Response:

a) The qualification of the WRB-1 CHF correlations with the VIPRE-D code was performed consistent with Dominion's approved Topical Report for COBRA/WRB-1 (Reference 12) and Westinghouse's approved Topical Report (Reference 6). An analysis of variance (ANOVA) test was performed but was not included in the original submission because it was not deemed to provide additional substantial information to the qualification. As mentioned in the response to Question 1 on Appendix A, it was recognized that ANOVA tests cannot be used as the sole measure of the performance of a CHF correlation, but they can be useful to indicate an extremely bad mismatch. The variables 0 analyzed were mass velocity, pressure, quality and test cell type.

The ANOVA results for VIPRE-D/WRB-1 (Table 6) exceed the critical values of F for some comparisons, but other comparisons prove the hypothesis that all the groups belong to the same distribution, i.e. that there is no bias of the results regarding the analyzed variables. These are the same trends discussed in the NRC's Safety Evaluation of WRB-1 (Reference 6).

0 0

6 0

0 0

Page 21 of 260 0

0 0

Table 6: M/P CHF Performance by Independent Variable Grouping using the entire WRB-1 database at 95% confidence level Number Average Standard Maximum Minimum of Data MIP Deviation M/P M/P

-"__ _ Analysis by Mass Velocities Below 1.25 58 0.9849 0.0753 1.1158 0.8005 Mlbm/hr-ft 2 ____

1.25- 1.75 131 1.0125 0.0970 1.3069 0.7632 Mib rn/hr-ft 2 0 1.75-M2.25 247 0.9930 0.0914 1.2711 0.7600 0 Mlbm/hr-ft 2 ____ ___________ _______

0 2.25-2.75 203 1.0129 0.0688 1.1715 0.8276 Mlbm/~hr-fte___ _____

0 2.75-3.25 159 1.0074 0.0755 1.1858 0.8337 0 Mlbm/hr-ft 2 ___

Above 3.25 147 1.0137 0.0787 1.2356 0.8017 Fdistrniufion= 2.6870 Fcica,(5,939) = 2.2236 0 Analysis by Pressures 0 Below 1575 psia 227 1.0201 0.0851 1.3069 0.8017 1575 - 1850 psia 179 0.9878 0.0861 1.2711 0.7632 1850 - 2250 psia 277 1.0076 0.0774 1.1873 0.7600 0 Above 2250 psia 262 1.0014 0.0814 1.2356 0.8005 0 Fcristribubon = 5.4330 Fcrivcal(3,941) = 2.6144 1 0 _jiIiII I Analysis b Qualities I I 0 Below 5% 262 1.0000 0.0762 1.3069 0.8017 0 5% to 10%

10% to 15%

199 247 1.0119 1.0090 0.0778 1.2034 0.8226 0.0810 1.2711 0.8005 0 15% to 20% 169 0.9927 0.0942 1.2312 0.7600 0 20% to 25% 68 1.0222 0.0922 1.2234 0.7662 2.4209 Fcrfica(4,940) = 2.3814 0 Fdistribution =

Analysis by Geometry Type 0 4x4 5x5 456 489 0.9941 1.0154 0.0847 0.0794 1.3069 1.2711 0.7600 0.7632 0 __Fdistntbum.on = 15.9428 Fcridcai(1,943) = 3.8513 Analysis by Thimble vs. Typical 0 Thimble 207 0.9910 0.0771 1.2229 0.7873 Typical 738 1.0091 0.0838 1.3069 0.7600 Fdstributjon =7.83 72 Fcriucai(1,943) = 3.8513 0 All Data WRB-1 All Data 945 1.0051 0.0827 1.3069 0.7600 Page 22 of 26 0

0 S

S d) The validity of equation B.6.1 is based on two assumptions: 1) the average M/P is 1.0 and 2) the M/P distribution is normal. These two assumptions were demonstrated in Appendix B: 0 0

CORRELATION AVERAGE M/P =1.0 M/P DISTRIBUTION NORMAL WRB-1 Table 1.6-1 Page B-12 0 0

NRC Question 2 S S

On page B-14 you state, "These plots show that there are no biases in the MIP ratio distributions,and that the performance of the WRB- 1 CHF correlationis independent of S

the three variables of interest. The plots show a mostly uniform scatterof the data and S no obvious trends or slopes." The plots suggest but do not demonstrate that the claims S made in those sentences are true. Please give the appropriatestatisticalanalysis that demonstrates the truth of the claim.

Dominion/DNC Response:

As discussed in the response to Question 1 for Appendix B, an ANOVA analysis was 0 performed to formally demonstrate that the performance of the WRB-1 CHF correlation is not biased by the three independent variables present in the correlation (mass S

velocity, pressure and quality). Please refer to the response to Question 1 for an analysis demonstrating no biases in the M/P distribution.

NRC Question 3 0

Please show the empirical probability density function for the MIP values used in the analyses together with the estimate of the 95-percent fractile.

S Dominion/DNC Response: 0 The probability density function for the WRB-1 correlation is shown in Figure 6.

0 The 945 datapoints of the VIPRE-D/WRB-1 M/P distribution summarized in Section B.6 of DOM-NAF-2 Appendix B were used to create the empirical probability density function. S These datapoints were distributed among 31 equal bins that covered the entire range of 0 MWP in the VIPRE-DIWRB-1 distribution, and the frequency of data in each bin was S determined. The resulting empirical probability density function for the VIPRE-D/WRB-1 M/P distribution was then compared with the probability density function of a normal S distribution of mean 1.0051 and standard deviation 0.0827, which are the mean and S standard deviation calculated for the VIPRE-D/WRB-1 M/P distribution in Section B.6 of S DOM-NAF-2 Appendix B. Figure 6 also displays the obtained 95% fractile (1.14) for the data and the VIPRE-D/WRB-1 DNBR limit obtained in Section B.6 of DOM-NAF-2 S Appendix B (1.17). S Page 23 of 26 0 S

S

Figure 6: VIPRE-DIWRB-1 Probability Density Function 19) 95% F ACTILE= 1-14 0.1 0.08 ýFI *0.08

( I VIPRE O/WRB--1 0 ,~~~I... N~m...

.. T- 1.17 S z LU 0.08

DNBR LIMIT=

0.06 0

S 0.04 0.04 0

S 0.02 0.02 S

S 0 0.740.760.780.800.82 0.840.860.880.90 0.92 09D4 08898 O 1.00 1.02 1,04 1.06 1.08 1.10 1,12 1.14 1.16 1.18 1.20 1.22 1.24 1.28 1.28 1830 1.32 1.34 S WP S IEa=VIPRE-D/WRB-1 CHF DATA NORMAL(1,0051,0.0827)

S S

S Page 24 of 26 S

S

0 0

References S

1. Letter from S. R. Monarque (NRC) to D. A. Christian (Dominion), "North Anna and Surry Power Stations, Units 1 and 2 and Millstone Power Station, Units 2 and 3 -

Request for Additional Information Rearding Topical Report DOM-NAF-2, 'Reactor Core Thermal Hydraulics Using the VIPRE-D Computer Code' (TAC Nos. MC4571 through MC4576)", Serial No.05-328, May 19, 2005.

2. Letter from L. N. Hartz (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station 0 Units 2 and 3, Request for Approval of Topical Report DOM-NAF-2, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code Including Appendix A, Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," Serial No.04-606, dated September 30, 2004.

3. Letter from L. N. Hartz (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc. (DNC)

North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Request for Approval of Appendix B of Topical Report DOM-NAF-2, Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," Serial No.05-020, dated January 13, 2005.

4. Topical Report, BAW-10199P-A, "The BWU Critical Heat Flux Correlations," B&W 0 Fuel Company, August 1996, including Addendum 2, "Application of the BWU-Z CHF Correlation to the Mark-BW17 Fuel Design with Mid-Span Mixing Grids,"

Framatome Cogema Fuels, November 2000.

5. Letter from L. N. Hartz (Dominion) to US NRC Document Control Desk "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Changes and Exemption Request, Use of Framatome ANP Advanced Mark-BW Fuel," Serial No.02-167, dated March 28, 2002. (Proprietary version).

6. Topical Report, WCAP-8762-P-A (Proprietary) and WCAP-8763-A (Non-proprietary),

"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids", Westinghouse Electric Corporation, July 1984.

7. Technical Report, EPRI NP-2511-CCM Volume 2, Revision 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 2: User's Manual," February 2001.

8. Technical Report, TID-25887, "Boiling Crisis and Critical Heat Flux," L. S. Tong, AEC Critical Review Series, 1972. 0

9. Technical Report, EPRI NP-2511-CCM-A Volume 4, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. Volume 4: Applications," April 1987.

10. Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), 0 "Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM,

'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

Page 25 of 26 0

0 S

0 0

0 11.Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 Maintenance Group),

0 "Acceptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor 0 Cores,' (TAC No. M79498)," October 30, 1993.

12.Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," R. C. Anderson and N. P. Wolfhope, November 0 1986.

13. Topical Report, WCAP-7667-P-A (Proprietary) and WCAP-7755-A (Non-proprietary),

0 "Interchannel Thermal Mixing with Mixing Vane Grids," F. F. Cadek, Westinghouse Electric Corporation, January 1975.

0 14.Topical Report, WCAP-7941-P-A (Proprietary) and WCAP-7959-A (Non-proprietary),

0 "Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid," F. F. Cadek et al., Westinghouse Electric Corporation, January 1975.

0

15. Topical Report, WCAP-8298-P-A (Proprietary) and WCAP-8299-A (Non-proprietary) 0 "The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," F.

E. Motley, et al., Westinghouse Electric Corporation, January 1975.

0 16.Technical Report, EPRI NP-2511-CCM-A Volume 1, Revision 4, "VIPRE-01: A 0 Thermal-Hydraulic Code for Reactor Cores. Volume 1: Mathematical Modeling,"

February 2001.

0 0

0 Page 26 of 26

0 0

S S

S 0

0 S

S Fleet Report DOM-NAF-2, Rev. 0.2-A I.

Reactor Core Thermal-Hydraulics Using the S VIPRE-D Computer Code S

S Attachment 2 S S

S Information Regarding a LYNXT Error Supporting the S Request for Approval of Fleet Report DOM-NAF-2 S 0

S (5 pages) S S

S S

S DOM-NAF-2, Rev. 0.2-A, Attachment 2 S

S S

I00

I~~~~omnion Resources Scrvices, Inc.

Do ino

September 8, 2005

United States Nuclear Regulatory Commission Serial No. 05-020A 0 Attention: Document Control Desk NL&OS/ETS: RO Washington, D.C. 20555 Docket Nos. 50-280/281 50-338/339

50-336/423 License Nos. DPR-32/37 NPF-4/7 DPR-65/NPF-49

VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

NORTH ANNA AND SURRY POWER STATIONS UNITS 1 AND 2

MILLSTONE POWER STATION UNITS 2 AND 3,

INFORMATION REGARDING A LYNXT ERROR SUPPORTING THE REQUEST FOR APPROVAL OF TOPICAL REPORT DOM-NAF-2

REACTOR CORE THERMAL-HYDRAULICS USING THE VIPRE-D COMPUTER CODE

INCLUDING APPENDIX A - QUALIFICATION OF THE F-ANP BWU CHF

CORRELATIONS IN THE DOMINION VIPRE-D COMPUTER CODE

In a September 28, 2004 letter (Serial No.04-406), Dominion/DNC submitted Topical

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

Code," and Appendix A to theTopical Report DOM-NAF-2, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," for NRC review and

approval. Dominion/DNC was recently notified by Framatome ANP of an error identified in

the LYNXT code, which was used to benchmark the VIPRE-D code. Although the impact of this error is considered negligible and should not affect the NRC's review of VIPRE-D, Dominion/DNC is providing a description of the error and the impact for your information.

The attachment to this letter summarizes the error and the impact on the DNB benchmark

analyses.

0Although the docket number is identified for each Dominion/DNC unit, Dominion/DNC is

requesting the approval of the generic application of this topical report. Plant specific

applications of this topical report, including applicable appendixes, will be submitted to the NRC for review and approval, in accordance with Section 2.1 of DOM-NAF-2.

If you have further questions or require additional information, please contact Mr. Thomas

Shaub at (804) 273-2763.

Very truly yours,

Leslie N. Hartz Vice President - Nuclear Engineering Virginia Electric and Power Company

Dominion Nuclear Connecticut, Inc.

S S

0 Serial No. 05-020A 0 Docket Nos. 50-280/281/338/339/336/423 Page 2 of 2 Attachment Commitments made in this letter: None 0 cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW 0 Suite 23T85 Atlanta, Georgia 30303 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406-1415 Mr. J. T. Reece (w/o Att.)

NRC Senior Resident Inspector North Anna Power Station 0 Mr. N. P. Garrett w/o Att.)

NRC Senior Resident Inspector Surry Power Station 0 Mr. S. M.Schneider (w/o Att.)

NRC Senior Resident Inspector Millstone Power Station 05 Mr. V. Nerses NRC Senior Project Manager - Millstone Unit 2 U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. G. F. Wunder NRC Senior Project Manager - Millstone Unit 3 U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Stephen R. Monarque NRC Project Manager - Surry and North Anna U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike 0 Rockville, Maryland 20852 0

0 0

0

S S

0 0

0 S

0 0

0 Attachment 1 0 Topical Report DOM-NAF-2 0

05-020A REACTOR CORE THERMAL-HYDRAULICS USING THE VIPRE-D COMPUTER CODE INFORMATION REGARDING A LYNXT ERROR 0

0 0 Virginia Electric and Power Company (Dominion)

,Dominion Nuclear Connecticut (DNC) 0 0

0 0

0

Characterization of the LYNXT Error During the development of a new version of the LYNXT computer code, Framatome ANP detected an error that affects the LYNXT DNB predictions for previous code versions. It was discovered that two different surface fits to the ASME steam tables were used to calculate the values for the thermodynamic quality for a given local coolant condition within the code. The difference in the two values is based on two different water property subroutines used to generate the saturated liquid enthalpy (hf) and the latent heat of vaporization (hfq), which are needed to calculate the thermodynamic quality given the local coolant enthalpy. One thermodynamic quality definition was being used in the flow field calculations and the second definition was being used in the calculation of the DNBR. Although the two definitions are very close (maximum observed differences less than 0.2 percent in hf and hfg) there is an impact on DNBR predictions. Neither surface fit is incorrect by itself, but the inconsistency of using two different values for the quality was characterized as a code error.

Generic Impact of the LYNXT Error Due to the location of the error in the LYNXT code, there is no impact to the LYNXT/BWU code/correlation limits reported in Framatome ANP's Topical Report BAW-10199 and Addendum 2. The differences between the DNBR results provided by the corrected and the uncorrected versions of the LYNXT code are extremely small, but observable.

Specific Impact to DOM-NAF-2 Topical Report DOM-NAF-2 describes Dominion's use of the VIPRE-D code, including modeling and qualification for Pressurized Water Reactors (PWR) thermal-hydraulic design. The Topical is entirely based on VIPRE-D calculations, and it does not rely on LYNXT results. However, the Topical includes information about VIPRE-D benchmarks to the NRC-approved code, LYNXT, to assist the NRC in the review of the VIPRE-D Topical Report.

For the 173 North Anna statepoints used in Section 5.1 in the Topical Report DOM-NAF-2, the average difference between LYNXT predictions with the error and without the error is less than 0.02%, and the maximum difference is less than 0.15%.

The maximum change to any numerical value reported in Section 5 of the main body of Topical Report DOM-NAF-2 regarding benchmark DNBR calculations between VIPRE-D and LYNXT is 0.02%.

" The comparisons between the corrected LYNXT and VIPRE-D are slightly better than the comparisons between the uncorrected LYNXT and VIPRE-D.

Appendix A to Topical Report DOM-NAF-2 documents Dominion's qualification of the BWU-N, BWU-Z and BWU-ZM correlations with VIPRE-D. Tables A.4.1-3 and A.4.3-3 of Appendix A list the LYNXT/BWU code/correlation limits for information in comparison

to the calculated values for VIPRE-D/BWU. Since there is no impact to the LYNXT/BWU code-correlation limits reported in Framatome ANP's Topical Report BAW-10199 (induding Addendum 2), Appendix A to Topical Report DOM-NAF-2 is not affected by

this error.

Appendix B to Topical Report DOM-NAF-2 documents Dominion's qualification of the

WRB-1 correlation with VIPRE-D. Appendix B to Topical Report DOM-NAF-2 is not

affected by this error.

Conclusion

Based on the above discussion, Dominion has concluded that the LYNXT error has a

negligible impact to the information provided in Topical Report DOM-NAF-2, including Appendixes A and B.

0 0

S 0

0 0

S 0

0 0

S 0

S S

S 0

0 S

Fleet Report DOM-NAF-2, Rev. 0.2-A S Reactor Core Thermal-Hydraulics Using the S VIPRE-D Computer Code S

S S

Attachment 3 S Supplemental Information for Appendix B - S Removal of Mixing Vane Grid Spacing Restriction S S

Initial Submittal (Serial No.09-528)

S Supplemental Information to Initial Submittal (Serial No. 09-528A) S Request for Additional Information - Questions and Responses (Serial No. 09-528B) S S

(43 pages) S S

S S

S DOM-NAF-2, Rev. 0.2-A, Attachment 3 S

S S

0 Dominion Resources Services, Inc.

qOl)00Dominion Boulevard, Glen Allen. VA .131 6; l aom lniOn

',h .\ddrcss: www.doi.Coi August 28, 2009

U.S. Nuclear Regulatory Commission Serial No.09-528

Attention: Document Control Desk NL&OS/GDM R1

Washington, DC 20555 Docket Nos. 50-305 50-336/423

50-338/339

50-280/281

License Nos. DPR-43 DPR-65/NPF-49 NPF-4/7 SDPR-32/37 DOMINION ENERGY KEWAUNEE, INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY

KEWAUNEE POWER STATION

MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS 1 AND 2

SURRY POWER STATION UNITS 1 AND 2

REMOVAL OF MIXING VANE GRID SPACING RESTRICTION IN APPENDIX B TO

FLEET REPORT DOM-NAF EVALUATION OF 4X4 DNB TEST OF 15X15 VANTAGE+ WITH IFMS USING VIPRE-D/WRB-1

In a September 30, 2004 letter (ADAMS Accession No. ML042800118), Dominion

submitted Fleet Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," and Appendix A to DOM-NAF-2, "Qualification of the F-ANP 0 BWU CHF Correlations in the Dominion VIPRE-D Computer Code," for NRC review and

approval. Pursuant to discussions with the NRC during an August 4, 2004 public

meeting (ADAMS Accession No. ML042520317), Dominion submitted Appendix B to the

Fleet Report, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," by letter dated January 13, 2005 (ADAMS

Accession No. ML050180257) for NRC review and approval. Appendix B documented

the qualification of the Westinghouse WRB-1 CHF Correlation with the VIPRE-D code

and the code/correlation departure from nucleate boiling ratio (DNBR) design limits.

The NRC approved Fleet Report DOM-NAF-2, including Appendices A and B, in the

Safety Evaluation Report (SER) included in their letter dated April 4, 2006 (ADAMS

Accession No. ML060790496). Dominion provided comments on the NRC SER in a

letter dated June 1, 2006 (ADAMS Accession No. ML061530114), and the NRC issued a revised SER that addressed Dominion's comments in a letter dated

June 23, 2006 (ADAMS Accession No. ML061740212). The approved version of Fleet

Report DOM-NAF-2-A, Rev. 0.0-A, was submitted to the NRC for information in a letter

0 0

Serial No.09-528 Docket Nos. 50-305/3361423/338133912801281 Page 2 of 4 dated September 13, 2006 (ADAMS Accession No. ML062650184). Appendix C to Fleet Report DOM-NAF-2, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," was submitted to the NRC in letter dated April 4, 2008 (Serial No. 08-0174), and subsequently approved by the NRC in letter dated April 22, 2009. The approved version of Fleet Report DOM-NAF-2, Rev. 0.1-A, with the inclusion of Appendix C was submitted to the NRC for information in a letter dated August 4, 2009 (Serial No.09-479).

Appendix B to Fleet Report DOM-NAF-2-A currently includes a restriction associated 0 with mixing vane grid spacing. Specifically, the subject restriction states that "VIPRE-D/WRB-1 will not be used for fuel with less than 13" mixing vane grid spacing." The purpose of this restriction was to exclude the use of the VIPRE-D/NRB-1 code/correlation with Westinghouse Intermediate Flow Mixing Vane grids (IFM) fuel types. The restriction was originally placed on WRB-1 in the SER for the COBRA/NRB-1 topical report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," July 1990. When the VIPRE-D fleet report was developed, the restriction from the COBRA SER was included in the WRB-1 Qualification in Appendix B of DOM-NAF-2-A. However, Westinghouse has developed several fuel products that incorporate IFMs (e.g. 15x15 VANTAGE+ and 15x15 Upgrade 0 fuel products); consequently, for Dominion to be able to use VIPRE-D/WRB-1 to perform DNB calculations for these fuel types the grid spacing restriction must be rescinded.

Therefore, Dominion hereby requests NRC approval to remove the mixing vane grid spacing restriction contained in Appendix B of Fleet Report DOM-NAF-2-A. The technical justification for this request is provided in the attachment.

0 If you have any further questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Sincerely, 0

cIresrident - Nuclear Engineering Dominion Energy Kewaunee, Inc.

Dominion Nuclear Connecticut, Inc.

Virginia Electric and Power Company Attachment Commitments made in this letter: None 0

0 0

0

Serial No.09-528 Docket Nos. 50-305/336/423/338/339/280/281

Page 3 of 4

cc: U.S. Nuclear Regulatory Commission

Region I

475 Allendale Road King of Prussia, Pennsylvania 19406-1415

U.S. Nuclear Regulatory Commission

Region II Sam Nunn Atlanta Federal Center

61 Forsyth Street, SW

Suite 23T85

Atlanta, Georgia 30303 0U.S. Nuclear Regulatory Commission

Region III

2443 Warrenville Road

Suite 210 Lisle, Illinois 60532-4352

Mr. P. S. Tam

NRC Senior Project Manager- Kewaunee Power Station U. S. Nuclear Regulatory Commission

- One White Flint North

11555 Rockville Pike

Mail Stop 8 H4A

Rockville, Maryland 20852-2738

Ms. C. J. Sanders

NRC Project Manager - Millstone Power Station Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North

11555 Rockville Pike Mail Stop 8 11A

Rockville, Maryland 20852-2738

Ms. K. R. Cotton

NRC Project Manager

U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 16 E15

11555 Rockville Pike

Rockville, MD 20852-2738 0

0 0

0

Serial No.09-528 Docket Nos. 50-305133614231338133912801281 Page 4 of 4 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station 0 0

NRC Senior Resident Inspector Millstone Power Station 0 NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station 0

0 0

0

Serial No.09-528 Docket Nos. 50-30513361423/338/33912801281 ATTACHMENT Technical Justification for the Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A 0 DOMINION ENERGY KEWAUNEE, INC.

0 DOMINION NUCLEAR CONNECTICUT, INC.

0 VIRGINIA ELECTRIC AND POWER COMPANY 0 KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 0 NORTH ANNA POWER STATION UNITS 1 AND 2 0 SURRY POWER STATION UNITS I AND 2 0

0 0

Serial No.09-528 Docket Nos. 50-30513361423/338133912801281 Attachment Technical Justification for the Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A S

Introduction In a September 30, 2004 letter (Reference 1), Virginia Electric and Power Company (Dominion) submitted Fleet Report DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," and Appendix A to the Fleet Report DOM-NAF-2, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," for NRC review and approval. Pursuant to discussions with the NRC during an 0 August 4, 2004 public meeting (Reference 2), Dominion submitted Appendix B to the Fleet Report, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," by letter dated January 13, 2005 (Reference 3) for NRC review and approval. Appendix B documented the qualification of the 0 Westinghouse WRB-1 CHF Correlation with the VIPRE-D code and the code/correlation departure from nucleate boiling ratio (DNBR) design limits. The NRC approved DOM-NAF-2 including Appendices A and B and provided their associated Safety Evaluation Report (SER) in Reference 4. Dominion provided comments on the NRC SER in Reference 5, and the NRC issued a revised SER that addressed Dominion's comments in Reference 6. The approved version of Fleet Report DOM-NAF-2-A, Rev. 0.0-A was submitted to the NRC for information in a letter dated September 13, 2006 (Reference 7). Appendix C to Fleet Report DOM-NAF-2, "Qualification of the 0 Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code,"

was submitted to the NRC in letter dated April 4, 2008 (Reference 17) and subsequently approved by the NRC in letter dated April 22, 2009 (Reference 15). The approved version of Fleet Report DOM-NAF-2, Rev. 0.1-A, with the inclusion of Appendix C was submitted to the NRC for information in a letter dated August 4, 2009 (Reference 16). 0 Dominion herein requests the removal of the mixing vane grid spacing restriction contained in Appendix B of Fleet Report DOM-NAF-2-A. The restriction requested for removal is as follows: "VIPRE-D/WRB-1 will not be used for fuel with less than 13" mixing vane grid spacing." The basis for this request is provided below. 0

Background

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 S 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 surface temperature. DNB is a concern in reactor design because of the possibility of fuel rod failure resulting from the increased rod surface temperature.

0 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 0

Page 1 of 6 0

0

Serial No.09-528 Docket Nos. 50-305133614231338133912801281

Attachment

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.

DNBRs greater than the design limit indicate the existence of thermal margin within the

reactor core. Thus, the purpose of core thermal-hydraulic DNB analysis is the accurate

calculation of DNBR to assess and quantify core thermal margin.

The computer code VIPRE (Versatile Internals and Components Program for Reactors -

EPRI) was developed for the Electric Power Research Institute (EPRI) by Battelle

Pacific Northwest Laboratories to perform detailed thermal-hydraulic analyses to predict

CHF and DNBR of reactor cores. Topical Report VIPRE-01 was approved by the U.S.

Nuclear Regulatory Commission (USNRC) in References 8 and 9 for referencing in licensing applications. VIPRE-D is the Dominion version of the VIPRE computer code

based upon VIPRE-01, MOD-02.1. VIPRE-D was developed to fit the specific needs of

Dominion's nuclear plants and fuel products by adding vendor specific CHF correlations and customizing its input and output. However, Dominion has not made any modifications to the NRC-approved constitutive models and algorithms contained in

~VIPRE-01.

Dominion Fleet Report DOM-NAF-2-A describes Dominion's use of the VIPRE-D code, including modeling and qualification for Pressurized Water Reactors (PWR) thermal-

hydraulic design. The Fleet Report demonstrates that the VIPRE-D methodology is

appropriate for PWR licensing applications.

Appendix B of DOM-NAF-2-A documents Dominion's qualification of the WRB-1 correlation with the VIPRE-D code. This qualification was performed against a subset

of the data from the Columbia-EPRI CHF database for Westinghouse "R" grid 17x17

and 15x15 fuel (Reference 10). This is the same subset of the Columbia-EPRI CHF database used by Dominion in the qualification of the WRB-1 correlation with the COBRA code (Reference 11). Appendix B summarizes the data evaluations that were

performed to qualify the VIPRE-D/WRB-1 code/correlation pair and to develop the

corresponding DNBR design limits for the correlation. In addition, Appendix B provides

the range of application for operating conditions:

1440

Pressure

2490 psia 2 o 0.9

Mass Flux _3.7 Mlbm/hr-ft

Local Quality

0.30 0 and imposed two additional restrictions on the intended range of application as follows:

a VIPRE-D/WRB-1 will not be used when the local heat flux exceeds 1.0 Mbtu/hr-ft2 ,

and

° VIPRE-DIWRB-1 will not be used for fuel with less than 13" mixing vane grid

spacing.

0 S

S

Page 2 of 6 0

0

Serial No.09-528 Docket Nos. 50-305133614231338/33912801281 S Attachment Discussion Dominion is requesting the removal of the grid spacing restriction note above imposed by Appendix B of the VIPRE-D Fleet Report DOM-NAF-2-A. The purpose of this restriction was to exclude the use of the VIPRE-DNVRB-1 code/correlation with Westinghouse Intermediate Flow Mixing Vane grids (IFM) fuel types. The restriction was originally placed on WRB-1 in the issuance of the SER for the COBRANVRB-1 topical report, VEP-NE-3-A (Reference 11). When the VIPRE-D fleet report was developed, the restriction from the COBRA SER was included in the WRB-1 Qualification in Appendix B of DOM-NAF-2-A (Reference 12). However, Westinghouse S has developed several fuel products that incorporate IFMs (e.g. 15x15 VANTAGE+ and 15x15 Upgrade). Consequently, for Dominion to be able to use VIPRE-D/WRB-1 to perform DNB calculations for these fuel types the grid spacing restriction must be rescinded. 0 Westinghouse conducted confirmatory DNB testing on the 15x15 fuel design (with Intermediate Flow Mixers) in December 1998 / January 1999. The measured and predicted critical heat flux for the range of the experimental data were used to statistically determine the 95/95% DNBR limit. The test data yielded a limiting DNBR value of 1.114. The results of this testing demonstrated that the use of the WRB-1 correlation for the 15x15 VANTAGE+ fuel is conservative and confirmed its applicability for this fuel type. It was therefore concluded that the DNB tests verified that application of the WRB-1 correlation to the 15X15 VANTAGE+ fuel was appropriate. The DNB test results were discussed during a March 17, 1999 meeting between the NRC, Westinghouse, and New York Power Authority (NYPA). Westinghouse documented this meeting in a letter to the NRC dated March 29, 1999 (Reference 13). Further, the NRC staff reviewed data and documentation of the DNB tests performed by Westinghouse in an amendment request for Indian Point Unit 3 (Reference 14) and determined that, because the new test data yielded a DNBR lower than the bounding limit of 1.17, the WRB-1 correlation is applicable to the 15X15 VANTAGE+ fuel. 0 Dominion has performed a similar evaluation of the Westinghouse CHF test data. The 5 approved VIPRE-D methods with WRB-1 (Reference 12) were used to yield a limiting DNBR value of 1.112. Additionally, Dominion's evaluations obtained essentially identical results to those Westinghouse provided to the NRC in Reference 13. The standard deviation determined using VIPRE-D/WRB-1 was within 0.004 of the Westinghouse results with THINC and VIPRE-W. (It is noted that the Westinghouse data and results transmitted in Reference 13 are proprietary to Westinghouse and are referred to herein by reference only.)

0 With essentially identical results, Dominion concludes, as did Westinghouse, that the DNB test data can be conservatively considered part of the WRB-1 database population. Therefore, the existing VIPRE-D/WRB-1 qualification in Reference 12 can be conservatively applied for IFM fuel types with respect to the VIPRE-D/WRB-1 limit DNBR of 1.17 (Appendix B, Reference 12).

0 0

Page 3 of 65 0

0

0 Serial No.09-528 Docket Nos. 50-305133614231338/33912801281 0 Attachment Conclusion 0 Dominion requests NRC review and approval to remove the following restriction from Fleet Report DOM-NAF-2-A, Appendix B (Reference 12):

0 "VIPRE-D/WRB-1 will not be used for fuel with less than 13" mixing vane grid 0 spacing." (page B-8 of Appendix B) 0 and the mixing vane grid spacing row from the range of validity table for 0 VIPRE-DNVRB-1 as follows:

0 Range of Validity for VIPRE-DIWRB-1 (pageB-28 of Appendix B)

Pressure 0 [psia]

1,440 to 2,490 0 Mass Velocity 0.9 to 3.7

[Mlbm/hr-ft 2] 0.9_to_3.7 0 Thermodynamic *0.30 0 Quality at CHF 0 Local Heat Flux

[Mbtu/hr- ft2 ] <1.0 0 Mixing Vane Grid .3.

Sp~> 113._

0 With the removal of the grid spacing restriction, Dominion will have approval to use the 0 VIPRE-D/WRB-1 code/correlation pair to analyze Westinghouse fuel products with IFMs 0 (e.g., 15x15 VANTAGE+ and 15x15 Upgrade). Test data provided by Westinghouse demonstrates that WRB-1 can be conservatively used for IFM fuel types 0 (Reference 13). Dominion has evaluated the same test data and has come to the same 0 conclusions using its approved in-house DNB methods, i.e., VIPRE-D. Thus, Dominion 0 has concluded that the VIPRE-D/WRB-1 limit of 1.17 and associated statistics can be conservatively applied to the analysis of Westinghouse fuel products with IFMs.

0 0 References 0

1. Letter from L. N. Hartz (Dominion) to Document Control Desk (USNRC), "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

0 (DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Request for Approval of Topical Report DOM-NAF-2, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code Including Appendix A, Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," Serial No.04-606, dated September 30, 2004, ADAMS Accession No. ML042800118.

Page 4 of 6

Serial No.09-528 Docket Nos. 50-305133614231338/33912801281 Attachment

2. Meeting Notes (S. R. Monarque, USNRC), "Summary of August 4, 2004, Meeting with VEPCO on Topical Report DOM-NAF-2 (TAC NOS. MC3904, MC3905, MC3906, AND MC3907)," September 3, 2004, ADAMS Accession No. ML042520317.

3. Letter from E. S. Grecheck (Dominion) to Document Control Desk (NRC), 'Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc.

(DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Request for Approval of Appendix B of Topical Report DOM-NAF-2, Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," Serial No.05-020, dated January 13,2005, ADAMS Accession No. ML050180257.

4. Letter from C. I. Grimes (USNRC) to D. A. Christian (Dominion), "Millstone Power Station Unit Nos. 2 and 3 (Millstone 2 and 3), North Anna Power Station Unit Nos. 1 and 2 (North Anna 1 and 2) and Surry Power Station Unit Nos. 1 and 2 (Surry 1 and

2) - Approval of Dominion's Fleet Report DOM-NAF-2, 'Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, (TAC Nos. MC4571, MC4572, MC4573, MC4574, MC4575 and MC4576)", Serial No.06-319, April 4, 2006, ADAMS Accession No. ML060790496.

5. Letter from E.S. Grecheck (Dominion) to NRC Document Control Desk, "Dominion Nuclear Connecticut, Inc. (DNC), Virginia Electric and Power Company (Dominion),

Millstone Power Station Unit Nos. 2 and 3, North Anna Power Station Unit Nos. 1 and 2, and Surry Power Station Unit Nos. 1 and 2, Requested Clarification of the NRC SER for Dominion Fleet Report DOM-NAF-2, (TAC Nos. MC4571, MC4572, MC4573, MC4574, MC4575 and MC4576)", Serial No.06-319, June 1, 2006, ADAMS Accession No. ML061530114.

6. Letter from S. R. Monarque (USNRC) to D. A. Christian (Dominion), "Millstone Power Station Unit Nos. 2 and 3, North Anna Power Station Unit Nos. 1 and 2, and Surry Power Station Unit Nos. 1 and 2 - Correction to Dominion's Fleet Report DOM-NAF-2, 'Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code'," Serial No.06-560, June 23, 2006, ADAMS Accession No. ML061740212.

7. Letter from G. T. Bischof (Dominion) to NRC Document Control Desk (USNRC),

"Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, Inc. (DNC), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Approved Topical Report DOM-NAF-2, Rev. 0.0-A, 'Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code including Appendixes A and B'," Serial No.06-773, September 13, 2006, ADAMS Accession No. ML062650184.

8. Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee),

"Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, 0

'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

P 0

0 Page 5 of 6 0

0

0 Serial No.09-528 0 Docket Nos. 50-30513361423/338133912801281 O Attachment

9. Letter from A. C. Thadani (USNRC) to Y. Y. Yung (VIPRE-01 Maintenance Group),

O "Acceptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor

Cores,' (TAC No. M79498)," October 30, 1993.

10. Technical Report, EPRI NP-2609, "Parametric Study of CHF Data, Volume 3, Part 1;

Critical Heat Flux Data," C. F. Fighetti, & D.G. Reddy, September 1982.

11. Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the O Virginia Power COBRA Code," R. C. Anderson, July 1990.

12. Fleet Report, DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics using the VIPRE-D

Computer Code,' R. M. Bilbao y Le6n, August 2006.

13. Letter from H. A. Sepp (Westinghouse) to Document Control Desk (USNRC) (NSD-

NRC-99-5828), "Notification of FCEP Application for DNB Testing and Revalidation of WRB-1 Applicability to the 15x15 VANTAGE+ Fuel Design, (Proprietary)," March 29, 1999.

14. Letter from G. F. Wunder (USNRC/NRR) to J. Knubel (NYPA), "Indian Point Nuclear Generation Unit No. 3 - Issuance of Amendment RE: Removal of Footnote from

Technical Specifications (TAC NO. MA5193)," September 2, 1999, ADAMS O Accession No. ML003780850.

15. Letter from D. N. Wright (NRC) to D. A. Christian (Dominion), "Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Unit Nos.

1 and 2, and Surry Power Station Unit Nos. 1 and 2 - Appendix C to Dominion Fleet O Report DOM-NAF-2, ,,Qualification of the Westinghouse WRB-2M CHF Correlation O in the Dominion VIPRE-D Computer Code" (TAC Nos. MD8703, MD8704, MD8705,

MD8706, MD8707, MD8708, MD8709)," Serial No.09-290, April 22, 2009.

16.Letter from J. A. Price (Dominion) to NRC Document Control Desk (USNRC),

"Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut,

Inc. (DNC), Dominion Energy Kewaunee, Inc. (DEK), Surry and North Anna Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Kewaunee Power Station, Approved Topical Report DOM-NAF-2, Rev. 0.1-A," Serial No.09-479,

August 4, 2009.

17. Letter from G. T. Bischof (Dominion) to Document Control Desk (NRC), "Dominion Energy Kewaunee, Inc. (DEK), Dominion Nuclear Connecticut, Inc. (DNC), Virginia O* Electric and Power Company (Dominion), Kewaunee Power Station, Millstone Power O Station Units 2 and 3, North Anna and Surry Power Stations Units 1 and 2, Request

for Approval of Appendix C of Fleet Report DOMNAF-2, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer O* Code," Serial No. 08-0174, dated April 4, 2008.

0 0

0

l Page 6of 6 0

0

0 Dominion Resources Services, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 D Web Address: www.dom.com PROPRIETARY-Withhold under 10 CFR 2.390 November 20, 2009 0

U.S. Nuclear Regulatory Commission Serial No. 09-528A Attention: Document Control Desk NL&OS/GDM RO Washington, DC 20555 Docket Nos. 50-305 50-336/423 50-338/339 50-280/281 License Nos. DPR-43 DPR-65/NPF-49 NPF-4/7 DPR-32/37 S

DOMINION ENERGY KEWAUNEE, INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS I AND 2 SURRY POWER STATION UNITS I AND 2 REMOVAL OF MIXING VANE GRID SPACING RESTRICTION IN APPENDIX B TO FLEET REPORT DOM-NAF SUPPLEMENTAL INFORMATION 0

In an August 28, 2009 letter (Serial No.09-528), Dominion requested NRC approval to remove a mixing vane grid spacing restriction contained in Appendix B of Fleet Report DOM-NAF-2-A, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code." Specifically, the subject restriction states that "VIPRE-D/WRB-1 will not be used for fuel with less than 13" mixing vane grid spacing."

Dominion provided technical justification for removing the restriction in the subject letter; however, we are providing supplemental information herein to further support our request and to facilitate NRC review. The supplemental information is provided in . contains information proprietary to Westinghouse Electric Company, LLC.

Therefore, this information is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations. The affidavit is included in Westinghouse authorization letter LTR-CAW-09-2701, "Application for Withholding 0

Attachment I contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation this page is decontrolled.

0 0

S Serial No. 09-528A Docket Nos. 50-3051336142313381339/2801281 lit Page 2 of 4 0 Proprietary Information from Public Disclosure," which also includes a Proprietary

Information Notice and a Copyright Notice. The Westinghouse authorization letter is

provided in Attachment 3. Correspondence with respect to the copyright or proprietary

aspects of the Westinghouse information noted above or the supporting Westinghouse affidavit should reference the authorization letter and should be addressed to

J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse

- Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355. A

redacted, non-proprietary version of the Westinghouse information is provided in Attachment 2.

If you have any questions or require additional information, please contact

Mr. Gary D. Miller at (804) 273-2771.

0 Sincerely,

J. la riceL

Vi eeresident - Nuclear Engineering

Dominion Energy Kewaunee, Inc.

Dominion Nuclear Connecticut, Inc.

Virginia Electric and Power Company

Attachments:

1. Supplemental Information (PROPRIETARY), Removal of the Grid Spacing

Restriction in Appendix B to Fleet Report DOM-NAF-2-A 0

2. Supplemental Information (NON-PROPRIETARY), Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A

3. Westinghouse Electric Company LLC Authorization Letter LTR-CAW-09-2701,

"Application for Withholding Proprietary Information from Public Disclosure," dated

November 10, 2009

Commitments made in this letter: None 0

0 0

S 0

0 0

0 Serial No. 09-528A Docket Nos. 50-305133614231338133912801281 Page 3 of 4 cc: U.S. Nuclear Regulatory Commission 0 Region I0 475 Allendale Road King of Prussia, Pennsylvania 19406-1415 U.S. Nuclear Regulatory Commission Region It Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 0 Atlanta, Georgia 30303 U.S. Nuclear Regulatory Commission 0 Region III 2443 Warrenville Road Suite 210 Lisle, Illinois 60532-4352 Mr. P. S. Tam NRC Senior Project Manager- Kewaunee Power Station U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8 H4A Rockville, Maryland 20852-2738 Ms. C. J. Sanders NRC Project Manager- Millstone Power Station Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8 BIA Rockville, Maryland 20852-2738 Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission 5 One White Flint North Mail Stop 16 E15 11555 Rockville Pike Rockville, MD 20852-2738 S

0 0

0 S

S S

S

Serial No. 09-528A Docket Nos. 50-305133614231338/33912801281

Page 4 of 4

Dr. V. Sreenivas

NRC Project Manager

U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A

11555 Rockville Pike

Rockville, MD 20852-2738 NRC Senior Resident Inspector

Kewaunee Power Station

NRC Senior Resident Inspector Millstone Power Station

NRC Senior Resident Inspector

North Anna Power Station NRC Senior Resident Inspector

Surry Power Station 0

0 S

0 0

0 Serial No. 09-528A Docket Nos. 50-30513361423/3381339/280/281 0

0 0

0 ATTACHMENT 2 0 0

0 0

Supplemental Information (NON-PROPRIETARY) 0 Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A 0 0

0 0

S 0

DOMINION ENERGY KEWAUNEE, INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS I AND 2 SURRY POWER STATION UNITS 1 AND 2 Page I of 8

O Non-Proprietary Serial No. 09-528A

Docket Nos. 50-305/336/423/338/339/280/281 Attachment 2 0

Supplemental Information in Support of the Removal of the. Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A

) Background

The:WRB-1 CHF cbrrelatibn was developed from a large body of rod bundle'CHF data obtained

at the Columbia University Heat Transfer Research Facility (HTRF) using full-scale, electrically heated rod bundle test sections (Reference 1). Westinghouse used this database in Reference 2 to qualify the THINC/WRB-1 code/correlation for "R" grid fuel. This is. also the same database used by Westinghouse to qualify the VIPRE-O1/WRB-1 code/correlation (Reference 3).

0Dominion used a subset of this experimental data, as described in Reference 4, for the

qualification of the WRB-1 correlation with the VIPRE-D code. Appendix B of Reference 4 Simposed an additional restriction on the intended range of application as follows:

"VIPRE-D/WRB-1 will not be used for fuel with less than 13"' mixing vane grid spacing."

- Dominion in a letter dated August 28, 2009M (Reference 5) requested NRC approval to remove the mixing vane grid spacing, restriction contained in Appendix B of Fleet Report DOM-NAF-2-A.

Additional technical justification is provided herein for the WRB-1 application to Westinghouse 15xl 5 fuel designs containing IFM grids.

0

Technical -Justification

Westinghouse conducted confirmatory critical heat flux (CHF) testing in December 1998 /

January 1999 to revalidate the: WRB-1 applicability to the 15,x15 Vantage+ fuel design which

included intermediate flow mixer (IFM) grids. The CHF test section consisted of a 4x4 typical

cell with geometry prototypical of 15x15 Vantage+ fuel design. Reference 6 provides details of

the test section, axial layout of the grids and instrumentation, and axial and radial power

distributions. Westinghouse demonstrated in Reference 6 that the quantity of test data was

comparable with other Westinghouse test bundles and that the test conditions were 0representative of the WRB-1 database. Westinghouse evaluated the test data for normality,

and compared the test data to the WRB-1 database using the F- and T-statistical tests. The

results indicated that the test. data produced a 95/95 DNBR limit less than 1.17 (the accepted

WRB-1 design limit) and a mean M/P (measured to predicted CHF) greater than 1.0, which 0shows that these results are conservative. Westinghouse showed similar results using VIPRE

and THINC. Westinghouse concluded that the CHF test data can be conservatively considered

as part of the WRB-1 database and the WRB-1 DNBR design limit of 1.17 can be conservatively

Page 2 of 8 0

0 0

0

0 Non-Proprietary Serial No. 09-528A 0 Docket Nos. 50-305/336/423/338/339/280/281 Attachment 2 applied to 15x15 Vantage+ fuel. The test results were discussed during a March 17, 1999 meeting between the NRC, Westinghouse, and New York Power Authority (NYPA).

Westinghouse documented this meeting in a letter to the NRC dated March 29, 1999 (Reference 6). Further, the NRC staff reviewed data and documentation of the tests performed by Westinghouse in an amendment request for Indian Point Unit 3 (Reference 7).

Dominion replicated the Westinghouse evaluation of the CHF test data using the VIPRE-D code with the WRB-1 correlation to develop a 95/95 DNBR limit for the test data.-One-sided tolerance theory (Reference 8) is used for the calculation of the test data DNBR design limit. This theory allows the calculation of a DNBR limit so that, for a DNBR equal to the design limit, DNB Will be avoided with 95% probability at a 95% confidence level. A comparison of the WRB-1 M/P results between Westinghouse VIPRE-W and Dominion VIPRE-D codes is shown in Table 1.

0 Because. all the statistical techniques used in Table 1 assume that the original data distribution is. normal, it is necessary to verify that the overall distribution for the M/P ratio is a normal distribution. To evaluate whether the distribution is normal, the D' normality test was applied (Reference 9). A value of D' equal to [ ]ab~c was obtained for the. test data. This D' value is 0 within the range of acceptability for [. ]ab,c data points with a 95% confidence level [

]"b.cl Thus, it is concluded that the M/P distribution for the test data is indeed normal.

0 Based on the results listed in Table. 1, the 95/95 DNBR limit for the test data can be calculated as:

DNBR._ 1.0 (Equation 1)

M/P - KNCP * (qao1) where:

M/P = average measured-to-predicted CHF ratio 0

M-p = standard deviation of the measured-to-predicted CHF ratios of the database 0

0 0

1 From Table 5 in Reference 1 [a*1

]a~c D'Lower Limit [ ]a,b~c D' Upper Limit [ ]ab~c 01 0

Page 3Of 8 0 0

0 0

0

0 0

0 Non-Proprietary Serial No. 09-528A 0 Docket Nos. 50-305/336/423/338/339/280/281 Attachment 2 S

K*N.c,p = one-sided tolerance factor based on N degrees of freedom, C confidence 0 level, and P portion of the population protected. This number is taken from S Table 1.4.4 in Reference 8.

0 Then, the DNBR design limit for test data with the VIPRE-D can be calculated as described in 0 Table 1:

0 Table 1: WRB-1 DNBR Design Limit for the Westinghouse Co nfirmatory Test Data 0 VIPRE-D VIPRE-W 0

0 Number of data . [ ]a,b'c [ aAb~c Degrees, of N n-1 "'C 0 freedom Average M/P M/P .]

a,,C ] a,b,c 0

Standard ] ab,c abC 0 Deviation <aMp [ [

0 Corrected 0 Standard aN =MIP[ (n -1)/ N ] ].bc .

Deviation 0 Owens Factor K(N,0.95,0.95) ]a,b,c --

0 0 95/95 DNBR Limit DNBRL 1.112 1.108 0

0 The F-Test Two Sample for Variances (Reference 10) was considered to determine whether the 0 variances for the WRB-1 database and the test data are equal. The hypothesis is tested by S comparing the T statistic ([ ]abc, at a 0.05 level of significance), and Fcwý,1(932, [ ]a"b.0.05 ) =

0 0 T 2, _ 0.00692 = [ ]a,bc 0 3 kfrtdata [a cb, c (Equation 2) 0 0 Therefore, it can be concluded that the variances are the same (i.e., equal), since the T statistic is less than Fcrdcal (i.e. both the WRB-I database and the additional test data have the same variability). Thus, the test data can be conservatively considered as part of the WRB-1 database, 0

which is also the same conclusion Westinghouse reported in Reference 6 from performing the F-Test.

0 Page 4 of 8

0 Non-Proprietary 0 Serial No. 09-528A 0 Docket Nos. 50-305/336/423/338/339/280/281 Attachmnent 2 0 0

The T-Test (Reference 10) was used to compare the means of the two populations with equal 0 variance. The T-Test was employed to test the hypothesis that IP1 - P2 = 0.0. The T-test produced all interval, [ ]3*"*, 0 thus the hypothesis that the means are jequal is rejected and is confirmed by the low probability 0

[a.bc of equal means. The T-Test for the test data failed, but was found to be 0 conservative since the mean of the test data [ ]r,b. is greater than the mean of the VI PRE-D/WRB-i code/correlation (1.0051). Westinghouse also failed the T-Test for equal means in 0 Reference 6, but found it conservative with respect to the VVRB-1 database. 0 0

Figure 1 displays the performance of the M/P ratio. This plot also includes the VIPRE-DIWRB-1 DNBR design limit line at 1.17. It can be seen that only 1 data point is outside the 0 VIPRE-DAWRB-i DNBR-design limit. Figures 2 through 4 display the performance of the M/P ratio S and its distributions as a function of the pressure, mass velocity and quality. The figures are 0 effectively the same as the Westinghouse results (Reference 6). 0 a,b,c 0

0 0

S 0

0 0

S 0

0 0

0 Figure 1: Measured vs. Predicted CHF 0 0

0 Page 5 of8

Non-Proprietary Serial No. 09-528A 0 Docket Nos. 50-305/336/423/338/339/280/281 0 Attachment 2 S

0 a,b,c 0

0 0

0 Figure 2: M/P vs. Pressure abc 0

S 0

0 0

S 0

Figure 3: M/P vs. Quality Page 6 of 8

S Non-Proprietary S Serial No. 09-528A 0 Docket Nos. 50-305/336/423/338/339/280/281 Attachment 2 S

0 3,b,c 0 0

0 0

0 S

0 0

Figure 4: M/P vs. Mass Velocity 0 S

Conclusions 0

The VIPRE-DBARB-1 analysis of the Westinghouse confirmatory CHF test data results in a

[ b]c" M/P ratio with a [ ]3,b.C standard deviation. Table 2 gives the 95/95 DNBR limit for 0 the test data, which is compared to the Westinghouse results from Reference 6. 0 Table 2:95/95 DNBR Umit for the WRB-1 Test Data I VIPRE-D Westinghouse 1.114 (THINC) 0 95/95DNBR Limit 1.112 1.108 (VIPRE-W)

Finally, the conclusions in Reference 5 (Serial No.09-528) that are supplemented herein, support the use of VIPRE-DM/RB-1 with a 1.17 DNBR Design ULmit for WRB-1 application to Westinghouse 15x15 fuel designs containing IFM grids.

Page 7 oF'8

S Non-Proprietary Serial No. 09-528A

Docket Nos. 50-305/336/423/338/339/280/281

Attachment 2 0

References.

1. Technical Report, EPRI NP-2609, "Parametric Study of CHF Data, Volume 3, Part 1; Critical Heat Flux Data," C. F. Fighetti, & D.G. Reddy,.September 1982.

2.. Technical Report, 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.

3. Technical Report, WCAP-14565-P-A, 'VIPRE-01 Modelihg and Qualification for

Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," Y. X. Sung,

P. Schueren, and A. Meliksetian, October 1999.

4. Fleet Report, DOM-NAF-2, Rev. 0.1-A "Reactor Core Thermal-Hydraulics using the

VIPRE-D Computer Code," R. S. Brackmann, July 2009.

5. Letter from J. A. Price (Dominion) to Document Control Desk (USNRC),. "Removal of

Mixing Vane Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF *Evaluation of 4x4 DNB Test of 15x15 Vantage+ with, IFMs Using VIPRE-D/WRB-1,"

August 28, 2009, Serial No.09-528.

6. Letter from H. A. Sepp (Westinghouse) to Document Control Desk (USNRC) (NSD-NRC-99-5828), "Notification of FCEP Application for DNB Testing and Revalidation of WR3-1 Applicability to the 15x15 Vantage+ Fuel Design, (Proprietary)," March 29, 1999.

7. Letter from G. F. Wunder (USNRC/NRR) to J. Knubel (NYPA), "Indian Point Nuclear Generation Unit NO.3-Issuance. of Amendment RE: Removal of Footnote from Technical

Specifications (TAC NO. MA5193)," September 2, 1999, ADAMS Accession No.

ML00378085

8. Technical Report, "Tables for Normal Tolerance Limits, Sampling Plans, and Screening,"

R. E. Odeh and D. B. Owen, 1980.

9. Technical Report, "Assessment of the Assumption of Normality (employing individual

observed values)," American National Standards Institute, ANSI N15.15.1974.

10. Technical Report, WAPD-TM-1292,. "Statistics for Nuclear Engineers and Scientists; Part 1: Basic Statistical Inference," February 1981, W. J. Beggs, Contract No. DE-ACl 1-

~76PN00014.

Q ~Page.8of 8 0

0

0 Serial No. 09-528A 0

Docket Nos. 50-30513361423/338/339/280/281 0 0

0 0

ATTACHMENT 3 0

S S

Westinghouse Electric Company LLC Authorization Letter LTR-CAW-09-2701, "Application for Withholding Proprietary Information from Public Disclosure,"

dated November 10, 2009 0

0 0

DOMINION ENERGY KEWAUNEE, INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS I AND 2

O Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Conunission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (412) 374-3846 Washington, DC 20555-0001 e-mail: greshraja@westinghousm.com Our ref: CAW-09-2701 November 10, 2009 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Supplemental Information in Support of the Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2 (Proprietary) 0 The proprietary information for which withholding is being requested in the above-referenced report is 0 further identified in Affidavit CAW-09-2701 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. Thie affidavit, which accompanies this letter, sets forth the basis 0 on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's S regulations.

S Accordingly, this letter authorizes the utilization of the accompanying affidavit by Dominion Generation.

0 Correspondence with respect to the proprietary aspects of the application for withholding or the 0 Westinghouse affidavit should reference this letter, CAW-09-2701, and shoUld be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company S LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

0 Very truly y uS, 0

0 Manager 1AIGres ham.

0 Regulatory Compliance and Plant Licensing 0 Enclosures 0

0 S

S 0

CAW-09-2701O 0 0

AFFIDAVIT COMMONWEALTHI OF PENNSYLVANIA:

ss 0

COUNTY OF ALLEGHENY: S 0

Before me, the undersigned authority, personally appeared J.A. Gresham, who, being by me duly 0 sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of S Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this 0 Affidavit are true and correct to the best of his lknowledge, information, and belief:

J. A. Gresham, Manager Regulatory Complimnce and Plant Licensing Sworn to and subscribed before me this 10d' day of November, 2009 Notary Public O 0

0 COMMONWEALTH OF PENNSYLVANIA NOTARIAL SEAL Renee Glampole. Notary Public Penn Township, Westmoreland County My Commlislon Expires September.25. 2013 O S

S S

0 S

S S

0 S

2 CAW-09-2701 0 (1) I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse 0 Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

0 (2) Iam making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse "Application for 0 Withholding" accompanying this Affidavit.

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

0 (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, 0 the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

0 (i) The information sought to be withheld from public disclosure is owned and has been held 00 in confidence by Westinghouse.

0 (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute 5 Westinghouse policy and provide the rational basis required.

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

0 advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, S structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

S 0

0

O O

3 CAW-09-2701 (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

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

S 0

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. 0 (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

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

0 There are sound policy reasons behind the Westinghouse system which include the following: 0 O

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. O (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.,

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

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. S 0

0 S

S

4 CAW-09-2701 0

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the 0competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a 0competitive advantage.

0 (iii) The information is being transmitted to the Commission in confidence and, under the 0provisions of 10 CFR Section 2.390, it is to be received in confidence by the

Commission.

(iv) The information sought to be protected is not available in public sources or available 0information has not been previously employed in the same original manner or method to best of our knowledge and belief.

0*the (v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in, "Supplemental Information in Support of the Removal of the

Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2 (Dominion Letter

09-528A)," dated November, 10, 2009 (Proprietary), for submittal to the Commission, being transmitted by Dominion Generation and Application for Withholding Proprietary

Information from Public Disclosure to the Document Control Desk. The proprietary 0information as submitted by Westinghouse for Dominion Generation is expected to be 00used in support of the removal of the grid spacing restriction.

0*This information is part of that which will enable Westinghouse to:

0(a) 0 Assist customer to obtain a license change.

Further this information has substantial commercial value as follows:

0 (a) Westinghouse can use this information to further enhance their licensing position

with their competitors.

0 0

S 0

5 CAW-09-2701 (b) The information requested to be withheld reveals the distinguishing aspects of 0 Westinghouse fuel designs and CHF correlations.

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

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

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

Further the deponent sayeth not.

0 0

0 S

0 0

S 0

S S

0 S

S S

S 0

0 S

S 0

0 0

0

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

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the

protection of proprietary information so submitted to the NRC, the information which is proprietary in the

proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the

brackets in the proprietary versions having been deleted). The justification for claiming the information

so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the

types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

0 0

0

COPYRIGHT NOTICE S

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its 0internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public 0disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright

protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in

order to have one copy available for public viewing in the appropriate docket files in the public document

room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include 0*the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

0 0

0

Dominion Resources Services, Inc. Dominion....

I 0(11 )ominion Boulevard, Glen Allen. VA 2,1ini, h \ddress: www.dorn.corn 0

April 16, 2010 0

U.S. Nuclear Regulatory Commission Serial No. 09-528B Attention: Document Control Desk NL&OS/GDM R1 Washington, DC 20555 Docket Nos. 50-305 50-336/423 50-338/339 50-280/281 0 License Nos. DPR-43 DPR-65/NPF-49 NPF-4/7 DPR-32/37 DOMINION ENERGY KEWAUNEE. INC.

DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY KEWAUNEE POWER STATION MILLSTONE POWER STATION UNITS 2 AND 3 NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS 1 AND 2 REMOVAL OF MIXING VANE GRID SPACING RESTRICTION IN APPENDIX B TO FLEET REPORT DOM-NAF-2-A RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION 0

In an August 28, 2009 letter (Serial No.09-528), Dominion Energy Kewaunee, Inc.,

Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power Company (collectively "Dominion") requested NRC approval to remove a mixing vane grid spacing restriction contained in Appendix B of Fleet Report DOM-NAF-2-A, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code."

Specifically, the subject restriction states that "VIPRE-D/WRB-1 will not be used for fuel 5 with less than 13" mixing vane grid spacing." Dominion provided technical justification for removing the restriction in the August 28, 2009 letter. In addition, Dominion provided supplemental information in a subsequent letter dated November 20, 2009 (Serial No.

09-528A.)

In an e-mail dated March 9, 2010, the NRC provided a draft request for additional information (RAI) associated with Dominion's submittals. A conference call was held on March 16, 2010 to discuss the questions contained in the RAI, and at the conclusion of the call, Dominion stated that it would formally respond to the RAI on the docket.

Dominion's response is provided in the attachment. S 0

S 0

S S

W Serial No. 09-528B

Docket Nos. 50-305133614231338133912801281

Page 2 of 3 0

If you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

D Sincerely, 0

J. Al 'n ce Vic Pr sident- Nuclear Engineering Dof n Energy Kewaunee, Inc.

Dominion Nuclear Connecticut, Inc.

Virginia Electric and Power Company

Attachment:

Response to Request for Additional Information, Removal of the Grid Spacing

Restriction in Appendix B to Fleet Report DOM-NAF-2-A Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission

! Region I

475 Allendale Road King of Prussia, Pennsylvania 19406-1415

U.S. Nuclear Regulatory Commission

t Region II 245 Peachtree Center Avenue, NE Suite 1200 0* Atlanta, Georgia 30303-1257

U.S. Nuclear Regulatory Commission

Region III 2443 Warrenville Road

Suite 210

Lisle, Illinois 60532-4352

Mr. P. S. Tam

NRC Senior Project Manager - Kewaunee Power Station

U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike

Mail Stop 8 H4A

Rockville, Maryland 20852-2738 0

0 0

0

S Serial No. 09-528B Docket Nos. 50-305/33614231338133912801281 Page 3 of 3 Ms. C. J. Sanders NRC Project Manager - Millstone Power Station Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8 B1A Rockville, Maryland 20852-2738 Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A 11555 Rockville Pike S Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station S

NRC Senior Resident Inspector Millstone Power Station NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station S

S S

S 0

Serial No. 09-528B Docket Nos. 50-3051336/423/338/3391280/281 ATTACHMENT Response to Request for Additional Information Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A 0 DOMINION ENERGY KEWAUNEE, INC.

0 DOMINION NUCLEAR CONNECTICUT, INC.

VIRGINIA ELECTRIC AND POWER COMPANY 0 KEWAUNEE POWER STATION 0 MILLSTONE POWER STATION UNITS 2 AND 3 0 NORTH ANNA POWER STATION UNITS I AND 2 SURRY POWER STATION UNITS I AND 2 0

0 0

Serial No. 09-528B Docket Nos. 50-305/336/423/338/339/280/281 Attachment Response to Request for Additional Information

Removal of the Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF-2-A 0

Introduction By letter dated August 28, 2009 (Reference 1), Dominion Energy Kewaunee, Inc.,

Dominion Nuclear Connecticut, Inc., and Virginia Electric and Power Company O (collectively "Dominion") requested NRC approval to remove a mixing vane grid spacing restriction contained in Appendix B of Fleet Report DOM-NAF-2-A (Reference 2),

"Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code." Specifically, the subject restriction states that "VIPRE-D/WRB-1 will not be used for fuel with less than 13" mixing vane grid spacing." Dominion provided 0 the technical justification for removing the restriction in the subject letter (Reference 1).

Dominion provided supplemental information to further support the request and to 0 facilitate NRC review in a November 20, 2009 letter (Reference 3). The NRC accepted the Dominion request for review via Reference 4.

The NRC staff forwarded by email a draft Request for Additional Information (RAI) to complete the review of the proposed change that consisted of five questions 0 (Reference 5). A teleconference was held on March 16, 2010 between the NRC and

Dominion staff to discuss the draft questions. The questions were finalized and the aforementioned email became a formal RAI (Reference 5). Dominion's response to the RAI questions is provided herein.

Background

Appendix B of DOM-NAF-2-A documents Dominion's qualification of the WRB-1 O correlation with the VIPRE-D code. Table B.8-1 lists the departure from nucleate boiling 0 ratio (DNBR) design limit of 1.17 for VIPRE-D/NRB-1, which is the value that yields a

95% non-DNB probability at a 95% confidence level (95/95 limit). Appendix B also includes a restriction that states "VIPRE-D/VVRB-1 will not be used for fuel with less than 13" mixing vane grid spacing." The purpose of this restriction was to preclude the use of the VIPRE-D/WRB-1 code/correlation with Westinghouse fuel with Intermediate

Flow Mixing Vane grids (IFM), which have a nominal 13" grid spacing. The restriction O was originally placed on WRB-1 in the issuance of the Safety Evaluation Report (SER) 0 for the COBRANVRB-1 topical report, VEP-NE-3-A (Reference 6). When the VIPRE-D fleet report was developed, the restriction from the COBRA SER was included in the WRB-1 qualification in Appendix B of DOM-NAF-2-A (Reference 2).

0 Pagel1 of 8 0 0

0

- Serial No. 09-528B

Docket Nos. 50-3051336/42313381339/280/281 Attachment

In Reference 7 the NRC issued a license amendment for the Indian Point Nuclear

) Generating Unit No. 3 (IP3) to accommodate the transition from VANTAGE 5 fuel without IFMs to VANTAGE+ with IFMs for Cycle 10. In Section 2.3 of the safety evaluation for that amendment, the NRC staff stated:

"The Westinghouse rod bundle criticalheat flux (CHF) correlation, WRB-1,

predicts critical heat flux in rod bundles based on subchannel local fluid conditions. This correlation was initially approved for the standard 14X14, 15X15 and the 17X17 standard Westinghouse fuel (WCAP-8762).

Evolution of the standard I 7X1 7 and 15X15 fuel have been developed by

Westinghouse and their behavior simulated by using an NRC-approved

scaling technique. This scaling technique was validated for all four of the different 17X1 7 fuel types, but not for the 15X15 OFA and the VANTAGE+

(w/IFMs) fuel. No testing was conducted to verify that the scaling

technique applied to the 15X15 standard fuel; however, cycle 10 analyses

has shown that there is substantial departure from nucleate boiling ratio (DNBR) margin. Consequently, until such time as fuel tests are conducted

on the 15X15 VANTAGE+ (w/IFMs) to validate the scaling technique and

the applicability of the WRB-1 correlation, [it] is acceptable for the

upcoming cycle 10 only. Also, DNB analyses must be submitted to the staff for review and approvalprior to cycle 11."

Westinghouse conducted confirmatory DNB testing on the 15XI5 VANTAGE+ fuel

design with IFMs in December 1998 and January 1999. The DNB test results were

discussed with the NRC during a March 17, 1999 meeting. Westinghouse documented

this meeting in a letter to the NRC dated March 29, 1999 (Reference 8). The measured and predicted critical heat flux for the range of the experimental data were used to 0statistically determine the 95/95% DNBR limit. The test data yielded a limiting DNBR

value of 1.114, which is less than the WRB-1 design limit of 1.17. The results of this

testing demonstrated that the use of the WRB-1 correlation for the 15X15 VANTAGE+

fuel with IFMs was conservative and confirmed its applicability for this fuel type.

Westinghouse concluded that the DNB tests verified the application of the WRB-1 correlation with a 95/95 correlation limit of 1.17 to the 15X15 VANTAGE+ fuel.

Subsequently, in Reference 9 the NRC issued a license amendment to remove a

footnote in the IP3 TSs stating "Current DNB analysis contains adequate margin for

Cycle 10. Prior to achieving criticality in Cycle 11, the DNB analysis must be reviewed

and approved by the NRC staff." In the associated SER, the NRC staff stated:

"The NRC staff reviewed the conditions of Amendment 175 safety

evaluation and the presentation of the DNB test results documented in a

letter from Westinghouse dated March 29, 1999. The restriction in the footnote of TS Section 3.1 was intended to ensure that adequate DNB margin would exist on cycle-specific basis until the fuel vendor 0

0

I.

Serial No. 09-528B Docket Nos. 50-305/3,3614231338133912801281 0 Attachment demonstrated the applicability of the WRB-1 correlation to the 15X15 VANTAGE+ fuel design. The staff reviewed the licensee's amendment 0 request and the documentation of the DNB tests performed by Westinghouse and determined that, because the new test data yielded a DNBR lower than the bounding limit of 1.17, the WRB-1 correlation is applicable to the 15X15 VANTAGE+ fuel; therefore, use of the WRB-1 correlationis acceptable for 15X15 VANTAGE+ fuel beyond Cycle 10 and removal of the footnote is acceptable. The DNB testing obviates the need for using the scaling technique; therefore, any question as to the acceptabilityof this technique is no longer relevant."

Thus, the NRC approved the application of the Westinghouse WRB-1 correlation with 95/95 correlation limit of 1.17 to the 15X1 5 VANTAGE+ fuel with IFIVls.

Westinghouse submitted Reference 10 to the NRC, which described evaluations performed for the 15x15 Upgrade fuel design. This submittal served as Westinghouse notification to the NRC, as required by the SER on Westinghouse Fuel Criteria Evaluation Process (FCOEP), that the NRC-approved process in WCAP-12488-A (Reference 11) was used for the validation of the WRB-1 DNB correlation applicability to the 15X15 Upgrade fuel design. The Westinghouse FCEP specifies the guidelines relevant to determine correlation applicability to a new grid design. In Reference 10, Westinghouse presented the results of the evaluation process and concluded that the WRB-1 DNB correlation with 95/95 correlation limit of 1.17 and the associated correlation ranges are acceptable for application to the 15X15 Upgrade fuel assembly design.

Dominion employed a two-step process to confirm two necessary methodology 0 elements for analysis of the 15x15 Upgrade design. First, it was necessary to confirm the applicability of the qualification of the VIPRE-D/WRB-1 qualification as documented in Appendix B of DOM-NAF-2-A to the Westinghouse 15X15 Upgrade fuel product. The second element was to support removal of the 13" grid spacing restriction. As a first step, Dominion replicated the Westinghouse evaluation of the confirmatory CHF test data and provided the results of this evaluation to the NRC in References 1 and 3.

Reference 1 summarized the Dominion results in narrative terms. Reference 3 provided 0 the detailed numerical and technical description of the Westinghouse proprietary data and data analysis. No significant differences were found between the Westinghouse and VIPRE-D results. The application of VIPRE-D/WRB-1 with 95/95 DNBR limit of 1.17 was demonstrated to be conservative for grid spacings of 13" for 15X15 fuel (i.e.,

15X15 fuel with IFMs). The second step of the process applies the conclusions of the Westinghouse FCEP evaluation for the 15X15 Upgrade in Reference 10. Therefore, Dominion has concluded that the VIPRE-D/WRB-1 design limit of 1.17 and associated statistics can be conservatively applied to the safety analysis of the Westinghouse 15X15 Upgrade fuel product with IFMs.

Page 3 of 8 P

01 0t 0

Serial No. 09-528B

Docket Nos. 50-305133614231338133912801281 Attachment

It is with respect to the first step described in the above paragraph that the NRC staff

provided an RAI to complete the review of the proposed change in References 1 and 3.

Dominion's response to t:he RAI questions is provided below.

0

NRC Question 1 11/20/09 supplement, page 4 of 8, Equation 2, reference is made to the T-statistic test.

It appears that this should be the F-Statistic test. The same applies in the sentence

above the equation, "comparingT statistic"should be "comparingF-statistic."

0 Dominion Response

Dominion understands that the NRC staff requires clarification of the terminology used

in Equation 2 and associated paragraphs on page 4 of 8 of Reference 3. Dominion

utilized the F-Test to determine whether the variances for the WRB..1 database and the test data are equal as discussed on page 4 of 8 of the supplemental material dated

November 20, 2009. However, in Equation 2, and in the preceding and following

sentences, reference is made to the "T statistic". The "T" used here identifies the

specific calculation of the test statistic for use in the F-Test and did not refer to the use of the T-Test as is discussed on page 5 of the supplement. In each instance on page 4

where "T" or "T statistic" is used, substitute "F" or "F statistic", as shown below.

S

"The hypothesis is tested by comparing the F statistic (2 nd paragraph)

"F = 2RB1 2test data (Equation 2)

... since the F statistic is less than ( 3 rd paragraph) 0 0S NRC Question 2

11/20/09 supplement, Page 4 of 8, Table 1, row 5.

0

(a) Is sigma calculatedusing n or N = n-1 in the denominator?

(b) What is the standard deviation corrected for?.

(c) The correction factor (n-1)/N are identically 1 (one). What is accomplishedby

taking the square root of 1?

0 Dominion Response

The calculation of the DNBR limit as described in Table 1 of Reference 3 for Test 100

was arranged in a form to be consistent with the previously reviewed calculation of the DNBR design limit for VIPRE-D/WRB-1 code/correlation as described in Table B.6-2 in Appendix B of DOM-NAIF-2-A (Reference 2).

P 4 of 8 Page

S Serial No. 09-528B Docket Nos. 50-305133614231338/33912801281 0 Attachment 0

The qualification of the VIPRE-D/WRB-1 code/correlation in Appendix B of DOM-NAF-2-A utilized 945 data points from 19 test series (n). Normally, the number of degrees of freedom (N) would be the total number of data minus one (n-1). However, because Westinghouse used these experimental data to correlate the 12 constants that appear in the WRB-1 correlation, the total number of degrees of freedom was corrected to account for this. Thus, the standard deviation of the database in Appendix B (OM/p) was corrected accordingly to account for this reduced number of degrees of freedom.

Equation B.6.2 of Reference 2 N = n 12 CYN = C'M/P* [ (n -1)/N

/N The DNBR design limit for the test data from Test 100 with VIPRE-DNVRB-1 code/correlation is calculated as described in Table 1 of Reference 3. The calculations in Table 1 are identical to those described in Table B.6-2 in Appendix B of DOM-NAF A. In Table 1, the Standard Deviation (sigma) of the test data is calculated using n-i.

The number of degrees of freedom is the total number of data minus one. No additional 0 correction to the degrees of freedom is necessary to account for the correlation of the 12 constants. Hence, the correction factor, (n-1)/N, is identically 1 (one).

The specific responses to the questions are then:

(a) The standard deviation of the sample population, sigma or O'M/p, is determined using n-1 in the denominator.

(b) The standard deviation of the database may be corrected to account for a reduced number of degrees of freedom. The number olf degrees of freedom is the total number of data minus one. No additional correction to the degrees of freedom is necessary to account for the correlation of the 12 constants. Thus, for Test 100, the number of degrees of fireedom, N, is equal to n-i.

0 (c) The Corrected Standard Deviation, aN, is determined by multiplying the Standard Deviation of the sample, aM/p, by the correction factor, which is the ratio of (n-1))N. The value of this correction factor is identically one.

Thus, the Corrected Standard Deviation is equal to the Standard Deviation of the sample. The purpose of presenting the calculation of the DNBR design limit for Test 100 as described in Table 1 of Reference 3 is to provide a calculation comparable in all ways to the established method of Table B.6-2 in Appendix B of DOM-NAF-2-A.

P 0

0 0

Page 5 of 8 5D 0

0

Serial No. 09-528B

Docket Nos. 50-3051336142313381339/2801281

Attachment

NRC Question 3 11/20/09 supplement, Page 6 of 8, Figure 2. There is only one data point for low quality

(< -0.1) and only one poirit for high quality (>0.2).

Dominion Response

The qualification of the VIPRE-D/WRB-1 code/correlation in Appendix B of DOM-NAF-2-A utilized 945 data points from 19 test series. Westinghouse conducted confirmatory

DNB testing on the 15),15 fuel design (with Intermediate Flow Mixers) in December

1998 and January 1999. Only one test series was run (Test 100). It is to be expected

that the range for the local conditions for Test 100 would be smaller and a subset of those in the 19 test series.

NRC Question 4 0 8/28/09 submittal, page 2 of 6, the last sentence of the 4th paragraph provides the

ranges of operation for the VIPRE-D code. Comparing these ranges with the plots

provided in attachment 1' of supplemental information for DOM-NAF-2-A, one notes that discrepancy exists between the tabulated data and the plotted data for both the Mass Flux and Local Quality. The upper end range of the Mass flux is less (-2.9), versus the

tabulated value of 3.7. Similarly, the upper end of the quality is less than .2, while the

tabulateddata is .3. Pleaseprovide a table containing the operationalranges of the fuel

in question.

0 Dominion Response

Appendix B of DOM-NAF-2-A (Reference 2) documented the qualification of the

Westinghouse WRB-1 CHF Correlation with the VIPRE-D code and the code/correlation departure from nucleate boiling ratio (DNBR) design limits. Appendix B provided the

range of application for the VIPRE-DNVRB-1 code/correlation set:

1440 < Pressure < 2490 psia 0.9 : Mass Flux < 3.7 Mlbm/hr-ft2

Local Quality < 0.30

Local Heat Flux < 1.0 Mbtu/hr-ft 2 The qualification of the VIPRE-DN/RB-1 code/correlation in Appendix B utilized 945

data points from 19 test series. Westinghouse conducted confirmatory DNB testing on

the 15X15 fuel design (with Intermediate Flow Mixers) in December 1998 and January

1999. Only one test series was run (Test 100). It is to be expected that the range for the local conditions for Test 100 would be smaller and a subset of those for the 19 test

series. The F-Test was used to demonstrate that the variances for the WRB-1 database

and the Test 100 data were equal. Thus, the test data can be conservatively considered 0

0

Page 6 of 8 0

0

S Serial No. 09-528B Docket Nos. 50-305133614231338133912801281 Attachment as part of the WRB-1 database, which is also the same conclusion Westinghouse reported in Reference 8 from performing the F-Test. Therefore, the range of applicability for VIPRE-DNVRB-1 will continue to be that provided in DOM-NAF-2-A, Appendix B (Reference 2), as listed above.

NRC Question 5 11/20/09 supplement, Table 2 on page 7 of 8 of attachment 1, shows the DNBR limit for VIPRE-D to be 1.112 while that for VIPRE-W is 1.108. The VIPRE-D value is more conservative. But the VIPRE-D/WRB-1 has a DNBR of 1.17, which is more conservative than 1.112. Which DNBR value does Dominion intend to use in the reload analysis?

Dominion Response The F-Test was utilized to determine whether the variances for the WRB-1 database and the test data are equal. It was determined that the variances are the same (i.e.,

both the WRB-1 database and the additional test data have the same variability). Thus, it is concluded that the test data can be conservatively considered as part of the WRB-1 database. This conclusion was reached by Westinghouse in Reference 8.

The T-Test was utilized to compare the means of the two populations with equal variance. The T-Test for the test data failed, but was found to be conservative since the mean of the test data was greater than the mean of the VIPRE-D/WRB-1 code/correlation of DOIM-NAF-2-A, Appendix B. The Westinghouse evaluation also involved a failed T-Test for equal means in Reference 8, but found it conservative with respect to the WRB-1 database.

Thus, Dominion has concluded that the DNBR design limit of 1.17 for the VIPRE-D/WRB-1 codelcorrelation as established in DOM-NAF-2-A, Appendix B (Reference 2) is conservative with respect to Test 100. Further, Dominion concludes that a DNBR design limit of 1.17 for the VIPRE-D/WRB-1 code/correlation can be utilized in the applicatioi to Westinghouse 15X15 Upgrade fuel design containing IFM grids. Dominion will use this value in safety analysis to support reloads with the Westinghouse 15X15 Upgrade fuel.

REFERENCES

1. Letter from J. Alan Price (Dominion) to NRC, "Dominion Energy Kewaunee, Inc., S Dominion Nuclear Connecticut, Inc., Virginia Electric And Power Company, Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Removal of Mixing Vane Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF Evaluation of 4x4 DNB Test of 15x15 Vantage+ with IFMs Using VIPRE-DANRB-1," dated August 28, 2009; Serial No.09-528. [ADAMS Accession No. ML092430338]

P S

Page 7 of 8 S

S

0 Serial No. 09-528B

Docket Nos. 50-30513361423/338133912801281

Attachment

2. Fleet Report, DOM-NAF-2, Rev. 0.1-A, "Reactor Core Thermal-Hydraulics using the

VIPRE-D Computer Code," July 2009. [ADAMS Accession No. ML092190894]

3. Letter from J. Alan Price (Dominion) to NRC, "Dominion Energy Kewaunee, Inc.,

Dominion Nuclear Connecticut, Inc., Virginia Electric And Power Company,

Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Removal of Mixing Vane Grid Spacing Restriction in Appendix B to Fleet Report DOM-NAF Supplemental

Information," dated November 20, 2009; Serial No. 09-528A. [Non-Proprietary

ADAMS Accession No. ML093310330]

4. Email from Peter Tam (NRC) to Gary D. Miller & Craig D. Sly (Dominion), "Dominion

-Mixing Vane Grid Spacing Restriction submittals (TAC ME2321, 2, 3, 4, 5, 6,

and 7)," dated December 8, 2009. [ADAMS Accession No. ML093430415]

5. Email from Peter Tarn (NRC), "Dominion Fleet Report DOM-NAF Draft RAI (TAC

ME2321 thru ME2327)," March 9, 2010. [ADAMS Accession No. ML100700236]

6. Topical Report, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," July 1990.

0

7. Letter from G. F. Wunder (NRC/NRR) to J. Knubel (NYPA), "Issuance of Amendment for Indian Point Nuclear Generating Unit No. 3 (TAC No. M97482),"

July 15, 1997. [ADAMS Accession No. ML003779946]

8. Letter from H. A. Sepp (Westinghouse) to Document Control Desk (NRC) (NSD-

NRC-99-5828), "Notification of FCEP Application for DNB Testing and Revalidation of WRB-1 Applicability to the 15x15 VANTAGE+ Fuel Design, (Proprietary),"

March 29, 1999.

0

9. Letter from G. F. Wunder (NRC/NRR) to J. Knubel (NYPA), "Indian Point Nuclear Generation Unit No. 3 - Issuance of Amendment RE: Removal of Footnote from

Technical Specifications (TAC NO. MA5193)," September 2, 1999. [ADAMS 4 Accession No. ML003780850]

10. Letter from Westinghouse (J. A. Gresham) to J. S. Wermiel (NRC), LTR-NRC-04-8, "Fuel Criterion Evaiuation Process (FCEP) Notification of the 15x15 Upgrade

Design," February 6, 2004.

0

11. WCAP-12488-P-A, 'WNestinghouse Fuel Criteria Evaluation Process," October 1994.

P

Page 8 of 8 0

v t e Letter Sequence Request
TAC:ME4600, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers 16 April 2010 Results Approval... Other: ML102560185
MONTHYEAR2010-08-31 [Table View]

ML102390419

Text

SUBJECT:

Dear Mr. Christian:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

5.0 REFERENCES

SUBJECT:

Dear Mr. Christian:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

SUBJECT:

Dear Mr. Christian:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

5.0 REFERENCES

SUBJECT:

Dear Mr. Heacock:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

5.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

Background

Subject:

Attachment:

Background

Navigation menu

Search