ML091030639

Appendix C to Dominion Fleet Report DOM-NAF-2, Qualification of the Westinghouse WRB-2M CHF

ML091030639
Person / Time
Site:	Millstone, Kewaunee, Surry, North Anna
Issue date:	04/22/2009
From:	David Wright Plant Licensing Branch II
To:	Christian D Virginia Electric & Power Co (VEPCO)
Wright D, NRR/DORL, 301-415 -1864
References
TAC MD8703, TAC MD8704, TAC MD8705, TAC MD8706, TAC MD8707, TAC MD8708, TAC MD8709, FOIA/PA-2011-0115
Download: ML091030639 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 2009 Mr. David A. Christian President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

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 REPORT DOM-NAF-2, "QUALIFICATION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION IN THE DOMINION VIPRE-D COMPUTER CODE" (TAC NOS. MD8703, MD8704, MD8705, MD8706, MD8707, MD8708, MD8709)

Dear Mr. Christian:

By letter dated April 4, 2008, Dominion Energy Kewaunee, Inc., Dominion Nuclear Connecticut, 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 CHF [Critical Heat Flux] Correlation in the Dominion VIPRE-D Computer Code" to the U.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 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 VIPRE-D when performing thermal-hydraulic analysis on 17x17 Robust Fuel Assembly fuel.

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

04*W*

Donna N. Wrigh~anager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-305, 50-336, 50-423, 50-338, 50-339, 50-280, and 50-281

Enclosure:

Safety Evaluation cc: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO APPENDIX C TO DOMINION FLEET REPORT DOM-NAF-2, "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 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 (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 17x17 Robust Fuel Assembly (RFA) fuel.

2.0 REGULATORY EVALUATION

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 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 Enclosure

- 2 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 evaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

The NRC staff's 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

3.1 Background 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 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 nature 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 under 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.

- 3 The WRB-2M CHF correlation along with Dominion VIPRE-D thermal-hydraulic computer code were used to predict the CH F 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 Program 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 of. 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 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 (Reference 2) and no further review is intended in this evaluation.

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 consistent with the WRB-2M approved topical report.

- 4 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 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 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.

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, 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 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 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 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 based on the scatter in predicted test results. The NRC staff has accepted DNBR limits that 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.

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 follows:

95/95 DNBR limit =

1 M

--K95/95 *cr P

- 5

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

p (J" is the effective standard deviation of all the M/P data.

K9S/9S is a tolerance multiplier which provides the 95/95 probability/confidence limit. The constant K9S/95 is a function of the effective degrees of freedom in the test series.

Consistent with previous CHF correlations, the standard deviation, (J", 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.

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 !\\IRC 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 17x17 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.

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

~ 6

4.0 CONCLUSION

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 Iicensee(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 Parameter WNG*1 Applicability Range Pressure (psia) 1405 to 2425 Local mass velocity (Mlbm/hr-ft£)

0.97 to 3.1 Local quality (fraction)

-0.1 to 0.29 Heat length (ft)

514 Grid spacing (inches) 10 to 20.6 Equivalent hydraulic diameter (inches) 0.37 to 0.46 Equivalent heated diameter (inches) 0.46 to 0.54

5.0 REFERENCES

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 Correlation in the Dominion VIPRE-D Computer Code" dated April 4, 2008 (ADAMS Accession No. ML080980229 (Publically Available)).

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 Accession No. ML081610106 (Non-Publically Available)).

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

- 7

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

April 22, 2009 Mr. David A. Christian President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 SUB"IECT:

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 REPORT DOM-NAF-2, "QUALIFICATION OF THE WESTINGHOUSE WRB-2M CHF CORRELATION IN THE DOMINION VIPRE-D COMPUTER CODE" (TAC NOS. MD8703, MD8704, MD8705, MD8706, MD8707, MD8708, MD8709)

Dear Mr. Christian:

By letter dated April 4, 2008, Dominion Energy Kewaunee, Inc., Dominion Nuclear Connecticut, 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 CHF [Critical Heat Flux] Correlation in the Dominion VIPRE-D Computer Code" to the U.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 approved WRB-2M CHF Correlation in the previously approvedDominion VIPRE-D Computer Code. The proposed change would allow Dominion to use the WRB-2M CHF Correlation in VIPRE-D when performing thermal-hydraulic analysis on 17x17 Robust Fuel Assembly fuel.

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

Sincerely, IRA!

Donna N. Wright, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-305,50-336, 50-423, 50-338,50-339,50-280, and 50-281

Enclosure:

Safety Evaluation cc: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrLAMO'Brien Resource RidsRgn1 MailCenter Resource LPl2*1 RlF RidsNrrPMSurry Resource RidsRgn2MailCenter Resource LPL3-1 RlF RidsNrrPMCSanders Resource RidsRgn3MailCenter Resource LPL1-2 RlF RidsNrrPMKewaunee Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDorl Resource RidsOgcRp Resource RidsNrrDssSnpb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl1-1 Resource JKaizer, NRR RidsNrrDorlLpl2-1 Resource RidsNrrDorlLpl3-1 Resource AAttard, NRR Accession N Ml091 03069 proviOded b dated ADAMs

o.

3

)ymemo OFFICE NRRlLPl2-1/PM NRRlLPl2-1/LA NRRlSNPB/BC DORULPl2-1/BC NAME DWright MO'Brien (SRohrer for)

AMendiola*

MWong DATE 4/21/09 4/20109 04/02109 4122109 OFFICIAL RECORD COPY