W3F1-2009-0042, Response to Request for Additional Information for License Amendment Revise Departure from Nucleate Boiling Ratio (DNBR) Safety Limit

Response to Request for Additional Information for License Amendment Revise Departure from Nucleate Boiling Ratio (DNBR) Safety Limit

ML092680063
Person / Time
Site:	Waterford
Issue date:	09/22/2009
From:	Christian K Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2009-0042
Download: ML092680063 (21)

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6496

'00ýEntergy Fax 504-739-6698 kchrisl@entergy.com Kenny J. Christian Nuclear Safety Assurance Director Waterford 3 W3F1-2009-0042 September 22, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Response to Request for Additional Information for the License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit Waterford Steam Electric Station, Unit 3 (Waterford 3)

Docket No. 50-382 License No. NPF-38

REFERENCES:

1.

W3F1-2009-0021, Waterford, Unit 3 - License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit, 6/3/09 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091560027).

2. Waterford Steam Electric Station, Unit 3, RAI, Re: License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit (TAC No. ME1424), 8/11/09 (ADAMS Accession No. ML092310748).

Dear Sir or Madam:

In Reference 1, Entergy Operations, Inc. (Entergy).proposed a change to the Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specifications (TS). In particular, the change would modify the TS 2.1.1.1 Departure from Nucleate Boiling Ratio (DNBR) safety limit based upon the Combustion Engineering 16 x 16 Next Generation Fuel design and the associated DNB correlations.

W3F1-2009-0042 Page 2 During the submittal review process, the Nuclear Regulatory Commission (NRC) determined that a Request for Additional Information (RAI) was required to complete the review of the Entergy request (Reference 2).

Thre response to the RAI is included in Attachment 1 to this letter.

This letter contains one, new commitment. This commitment is summarized in .

If you have any questions or require additional information, please contact Robert J.

Murillo at (504)-739-6715.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 22, 2009.

Sincery, Attach nts:

1. Response to Request for Additional Information

2. List of Regulatory Commitments

W3F1-2009-0042 Page 3 cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004

Attachment 1 W3F1-2009-0042 Response to Request for Additional Information to W3F1-2009-0042 Page 1 of 15 RESPONSE TO REQUEST FOR ADDITIONAL. INFORMATION In letter W3F1-2009-0021 (Reference 1), Entergy requested an amendment to the Waterford Steam Electric Station Unit 3 (Waterford 3) Technical Specification (TS) 2.1.1.1, Departure from Nucleate Boiling (DNB) ratio safety limit based upon the Combustion Engineering 16 x 16 Next Generation Fuel (NGF) design and the associated DNB correlations. During the submittal review process, the Nuclear Regulatory Commission (NRC) determined that additional information was required to complete the review of the Entergy request (Reference 2).

Arkansas Nuclear One (ANO) submitted a similar amendment request (Reference 3) and obtained a similar Request for Additional Information (RAI) (Reference 4). The Waterford 3 RAI response follows the format and content of the ANO RAI response (Reference 5).

1. Address compliance with the conditions stated in the NRC safety evaluation reports approving the topical reports that allow application of the ABB-NV and WSSV-T critical heat flux correlations to the Next Generation Fuel design for the Waterford 3 core. The following topical reports should be considered when they are applicable:

WCAP-16523-P-A; CENPD-387-P-A; CENPD-1 61-P-A; CEN-356(V)-P-A; CEN-139(A)-P; and WCAP-16500-P A (Rev. 00).

Several of the topical reports (TRs) listed in RAI Question #1 are currently listed in Waterford 3 TS 6.9.1".11 (Core Operating Limits Report (COLR)) and the Cycle 16 COLR as approved methodologies. When Waterford 3 requested the Technical Specification (TS) to be changed to include these methodologies, the limitations and conditions listed in the safety evaluation reports (SERs) were addressed at that time.

The responses provided at the time of the various TS change requests are repeated here. Some of the responses have been updated based upon revised or supplement submittals or additional information sent to the NRC.

WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes (Reference 23)

In letter W3F1-2007-0037 (Reference 7), Entergy requested changes to Waterford 3 TS. The proposed changes would revise TS 6.9.1.11.1 which would add new analytical methods to support the implementation of NGF. One of these new methods is WCAP-16523-P-A. In the letter W3F1-2007-0037 submittal, the safety evaluation (SE) Conditions and Limitations within each of the licensed TRs being added to the COLR were identified and discussed. The NRC Staff reviewed the disposition of each SE Condition and Limitation and found that Waterford 3 to W3F1 -2009-0042 Page 2 of 15 adequately addressed each oneof them for this TR (Reference 9). For the ease of review the Limitations and Conditions'and how Waterford 3 meets them for this TR are repeated below.

1. The WSSV correlationmust be used in conjunction with the VIPRE code since the correlationwas developed based on VIPRE and the associatedVIPRE input specifications. Other uses of the WSSV correlationshould reference this TR and be based on appropriatebenchmarking with VIPRE.

This condition is not applicable to Waterford 3 as the WSSV correlation with VIPRE will not be used at this time.

2. The WSSV-T correlationmust be used in conjunction with the TORC code since the correlation constants were developed based on TORC and the associated TORC input specifications. The correlationsmay also be used in the CETOP-D code in support of reload design calculationsbenchmarked by TORC.

The WSSV-T correlation is used in conjunction with TORC and CETOP-D codes in support of reload design calculations.

3. The WSSV and WSSV-T correlationsmust also be used with the optimized Tong Fc shape factor for non-mixing and side-supported mixing vane grids to correct for non-uniform axial power shapes.

The optimized Tong Fc shape factor was utilized for non-mixing and side-supported mixing vane grids in the Waterford 3 NGF Thermal Hydraulic (TH) implementation analyses.

to W3F1 -2009-0042 Page 3 of 15

4. The range of applicabilityfor both the WSSV and the WSSV-T correlationsare:

Parameter Units Range Pressure -, psia 1,495 to 2,450 Local coolant quality -- <0.34 Local mass velocity 106 0.90 to 3.46 Ibm/2 hr-ft Matrix heated hydraulic diameter, inches 0.4635 to I Dhm 0.5334 Heated hydraulic diameter ratio, -- 0.679 to 1.00 Dhm/Dh Heated length, HL inches 48* to 150 Grid spacing inches 10.28 to 18.86

• Set as minimum' HL value, applied at all elevations below 48 inches.

The WSSV-T correlation was applied according to Section 6.2 of WCAP-1 6523-P within the above range of applicability in the Waterford 3 NGF TH implementation analyses.

CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuels (Reference 22)

In letter W3F1 -2007-0037 (Reference 7), Entergy requested changes to Waterford 3 TS. The proposed changes would revise TS 6.9.1.11.1 which would add new analytical methods to support the implementation of NGF. One of these new methods is CENPD-387-P-A. In the letter W3F1-2007-0037 submittal, the SE Conditions and Limitations within each of the licensed TRs being added to the COLR were identified and discussed. The NRC Staff reviewed the disposition of each SE Condition and' Limitation and found that Waterford 3 adequately addressed each one of them for this TR (Reference 9). 'For the ease of review, the Limitations and Conditions and how Waterford 3 meets them for this TR are repeated below.

The following conditions are satisfied when applying the ABB-NV correlation for non-mixing vane grid spans for CE 16 x 16 Standard and NGF assemblies:

1. The ABB-NV and ABB-TV correlationsindicate a minimum DNBR limit of 1.13 will provide a 95 percent probability with 95 percent confidence of not experiencing CHF on a rod showing the limiting,value.

to W3F1 -2009-0042 Page 4 of 15 The ABB-NV correlation is applied for non-mixing vane grid spans for CE 16 x 16 standard and NGF assemblies. The minimum DNBR correlation limit of 1.13 is used. The WSSV-T correlation is applied for the mixing vane grid spans of the NGF fuel as described in Section 6.2 of WCAP-1 6523-P instead of the ABB-TV correlation.

2. The ABB-NV and ABB-TV correlationsmust be used in conjunction with the TORC code since the correlationswere developed on the basis of the TORC and the associated TORC input specifications. The correlationsmay also be used in the CETOP-D code in support of reload design calculations.

The ABB-NV correlation for non-mixing vane grid spans for CE 16 x 16 standard and NGF assemblies is used in conjunction with both TORC and CETOP-D codes.

3. The ABB-NV and ABB-TV correlationsmust also be used with the ABB-CE optimized Fc shape factor to correctfor non-uniform axialpower shapes.

The ABB-NV correlation will be used with the ABB-CE optimized Fc shape factor to correct for non-uniform axial power shapes.

4. Range of applicabilityfor the ABB-NV and ABB-TV correlations:

Parameter ABB-NV Range ABB-TV Range Pressure (psia) 1750 to 2415 1500 to 2415 Local mass velocity (Mlbm/hr-ft2) 0. 8 to 3.16 0.9 to 3.40 Local quality -0.14 to 0.22 -0.10 to 0.225 Heated length, inlet to CHF location (in) to 150 _48 _ 48 to 136.7 Grid Spacing (in) 8 to 18.86 8 to 18.86 Heated hydraulic diameter ratio, 0.679 to 1.08 0. 679 to 1.000 Dhm/Dh The specified range of applicability forthe designated parameters is used when applying the ABB-NV correlation.

5. The ABB-NV and ABB-TV correlationswill be implemented in the reload analysis in the exact manner described in Section 7.1 of Topical Report CENPD-387-P, Revision 00-P.

to W3F1 -2009-0042 Page 5 of 15 The ABB-NV correlation is applied according to Section 7.1 of CENPD-387-P-A for non-mixing vane grid spans for CE 16 x 16 Standard and NGF assemblies.

The WSSV-T correlation is applied for the mixing vane grid spans of the NGF fuel as described in Section 6.2 of WCAP-1 6523-P instead of the ABB-TV correlation.

6. Technology transfer will be accomplishedonly through the process describedin Reference 5 which includes ABB-CE performing an independent benchmarking calculation for comparison to the licensee generatedresults to verify that the new CHF correlationsare properly applied for the first application by the licensee.

There is no technology transfer between Westinghouse and Entergy at this time.

CENPD-1 61-P-A, TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core (Reference 6)

The constraints listed in the SER for this topical were superseded by the review of CENPD-206-P-A (Reference 8). Westinghouse has declared that the information showing how the constraints for the proprietary computer code TORC (CENPD-206-P-A) are being met is proprietary. Entergy has provided this information previously in support of the Extended Power Uprate (EPU). In this case, in lieu of repeating the proprietary information and requesting the NRC to withhold this information from the public, a reference to the previously submitted response is provided. Letter W3F1-2003-0074 (Reference 10) Appendix 1 Table 15 provides the requested information.

The extended power uprate NRC safety evaluation (Reference 11) Section 2.8.3 states the following:

The evaluation methodology that is applied for the EPU thermal-hydraulic design is unchanged from the current methodology. Steady-state departure from nucleate (DNB) ratio (DNBR) analyses for EPU core have been performed using the TORC code (References 47 and 48), the CE critical heat flux correlation (References 49 and 50), and the CETOP code (Reference 51).

The information in these references have'been reviewed as part of this effort and found to still be true and accurate.

to W3F1 -2009-0042 Page 6 of 15 CEN-356(V)-P-A, Modified Statistical Combination of Uncertainties (Reference 20)

The CEN-356(V)-P-A modified SCU (MSCU) methodology was originally NRC -

approved for use at the Palo Verde Nuclear Generation Station Unit 1 (Reference 12). Waterford 3 informed the NRC of the CEN-356 use in letter W3F1-92-0053 (Reference 17) for the Cycle 6 Reload Analysis Report.

The NRC approval to use this methodology was part of a change to the Waterford 3 TS 6.9.1.11 (Reference 13). This specification lists the analytical methods used to determine the core operating limits report. A review of this topical demonstrated that the NRC did not list any Limitations or Conditions for the use of this methodology.

The NRC Core Operating Limit Report (COLR) safety evaluation (Reference 13) states the following regarding the COLR:

The report provides the values of cycle-specific parameter limits that are applicable for the current fuel cycle. Furthermore, this specification requires that the NRC-approved methodologies be, used in establishing the values of these limits for the relevant specifications and that the values be consistent with all applicable limits of the safety analysis. The approved methodologies are the following: ... (c) CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties."

CEN-356(V)-P-A methodology is augmented by WCAP-1 6500-P Supplement 1-P Revision 1-P (Reference 26) which is currently under NRC review (refer to Attachment 2 commitment).

CEN-139(A)-P, Statistical Combination of Uncertainties Combination of System Parameter Uncertainties in Thermal Margin Analyses for Arkansas Nuclear One Unit 2 (Reference 21).

Letter W3P86-1686 (Reference 14) originally submitted the use of a statistical combination of uncertainties (SCU) as part of the Waterford 3 Cycle 2 Reload Analysis Report (RAR). W3P86-1686 Section 6.1 stated that the SCU methodology presented in CEN-283(S)-P (Reference 28) was applied with Waterford 3 specific data using the calculational factors listed in W3P86-1686 Table 6-1 and other uncertainty factors at the 95/95 confidence/probability level to define a design limit of 1.26 on CE-1 minimum DNBR.

Letter W3P86-3331 (Reference 15) submitted the Waterford 3 Cycle 2 TS change requests. This change request included the DNBR safety limit to incorporate the SCU methodology.

Waterford 3 Cycle 2 Technical Specification Amendment 12 NRC Safety Evaluation Report (SER) (Reference 16) Section 3.1 approval states the following:

to W3F1 -2009-0042 Page 7 of 15 ForCycle 2, the uncertainties associated with fuel manufacturing variations, as well as certain thermal-hydraulic uncertainties, will be combined using the methodology associated with the Statistical Combination of Uncertainties (SCU).

These methods have been previously used on other Combustion Engineering (CE) plants and have been approved by the NRC.

CEN-343 (Reference 29) and CEN-338 (Reference 30) were written for the Waterford 3 specific SCU methodology. CEN-338 explicitly states that the Waterford 3 analysis includes penalties imposed by the NRC in their review of these methods from CEN-1 39 (Reference 18 and 19).

CEN-356(V)-P-A is the current approved methodology which extends the previous Statistical Combination of Uncertainty Topicals. CEN-356(V)-P-A methodology is augmented by WCAP-1 6500-P Supplement 1-P Revision 1-P (Reference 26) which is currently under NRC review (refer to Attachment 2 commitment).

WCAP-16500-P-A, Revision 0-P-A, CE 16 x 16 Next Generation Fuel Core Reference Report (Reference 25).

In letter W3F1-2007-0037 (Reference 7), Entergy requested changes to Waterford 3 TS. 'The proposed changes would revise TS 6.9.1.11.1 which would add new analytical methods to support the implementation of NGF. One of these new methods is WCAP-16500-P-A. In the letter W3F1-2007-0037 submittal, the SE Conditions and Limitations within each of the licensed TRs being added to the COLR were identified and discussed. The NRC Staff reviewed the disposition of each SE Condition and Limitation and found that Waterford 3 adequately addressed each one of them for this TR (Reference 9). For the ease of review, the Limitations and Conditions and how Waterford 3 meets them for this TR are repeated below.

1. Using approved methods, the licensee must ensure that all of the design criteria specified in TR WCAP-16500-P are satisfied on a cycle-specific basis (SE Section 3.3.1).

As part of the reload methodology, all of the new design criteria specified for CE 16 x 16 NGF per WCAP-1 6500-P, Table 1-1 will become part of the reload analysis basis. Using approved models and methods, the reload analysis, which is reviewed per the requirements of 10 CFR 50.59, will check/confirm that these design criteria are met.

2. Fuel assembly component design and configuration(e.g., type and distribution of spacergrids and IFM grids) are limited to the five designs described in TR WCAP-16500-P and in response to RAI No. 2 (SE Section 3.2).

The Waterford 3 NGF assembly is consistent with the Plant B design defined in Figure 1-1 of WCAP-16500-P and for the Plant B design documented in the

Attachment.1 to W3F1 -2009-0042 Page 8 of 15 response to RAI No. 2 of Westinghouse letter LTR-NRC-06-66.

3. The reference fuel assembly design,. CE 16 x 16 NGF, its fuel mechanical design

-methodology and design criteria,are approved up to a peak rod average burnup of 62 GWd/MTU. A fuel burnup limit may exist, either explicitly or implicitly, in other portions of a plant's licensing,basis. The NRC staff's approval of this,,

topical report allows the CE 16 x 16 NGF assembly to reach a rod average burnup of 62 GWd/MTU. However, a lice nse amendment request, specifically addressing each plant's licensing basis including radiologicalconsequences, is requiredprior to extending burnup beyond current levels. Further,the NRC staff's SE for Optimized ZIRLOTM (Addendum I to TR WCAP-12610-P-A and TR CENPD-404-P-A) specified a 60 MWd/kgU burnup limit and this limitation must be revised prior to extending the peak rod average burnup for the NGF design (SE Section 3.4).

The current Waterford 3 licensing basis restricts peak rod average burnup to 60 MWd/kgU. Entergy is not proposing a change to this limit.

4. Licensees shall demonstrate the accuracy of their growth predictionsbased upon measured data and this validation shall be ahead of the burnups achieved by batch implementation. The growth model validation (e.g., measured versus predicted) should be documented in a letter(s) to the NRC (SE Section 3.2. 1).

The growth data presented in Figure 2-15 of WCAP-1 6500-P Supplement 1-P is ahead of the projected exposure for the first cycle implementation of NGF. The fluence for the data is approximately 7 x 1021 nvt, which corresponds to an assembly average burnup of 39 MWD/kgU. The projected end of first cycle assembly average burnup is approximately 27 MWD/kgU. As indicated in the responses to RAIs 1a and 1b in Westinghouse letter LTR-NRC-06-66, additional growth data will be obtained from future Lead Test Assembly (LTA) exams ahead of the exposure .achieved by batch implementation. This data has been sent by Westinghouse as it is obtained and is being tracked by the Westinghouse processes.

5. To compensate for NRC staff concerns related to the digitalsetpoints process, an interim margin penalty of 6 percent must be applied to the final addressable constants (e.g., BERRI* 1.06, [(1+EPOL2)*1.06- 1.0]) calculated following the 1/64 hypercube setpoints process (Response No. 6 of Reference 6). Removal of this interim margin penalty will be consideredafter the digital setpoints methods have been formalized, documented (e.g., revision to TR WCAP-16500-P), and approved by the NRC (SE Section 3.7).

For the first cycle of Waterford 3 that contains a batch of NGF (Cycle 16), the analysis that calculates the uncertainty addressable constants for the Core Operating Limit Supervisory System (COLSS) on-line monitoring system and the to W3F1 -2009-0042 Page 9 of 15 Core Protection Calculator (CPC) System will not account for the NGF design and Critical Heat Flux (CHF) correlations. Therefore, the resultant DNB uncertainty addressable constants will not credit the DNB margin gain due to NGF, will not require application of the interim 6% margin penalty and will not require use of the 1/64 hypercube setpoints process.

Full DNB margin credit for NGF will begin with the next cycle (Cycle 17) where the NGF CETOP-D model with the WSSV-T and ABB-NV CHF correlations will be used in the COLSS and CPC uncertainty analyses. The Modified Statistical, Combination of Uncertainties (MSCU) analysis performed each cycle, as described in CEN-356(V)-P-A, will automatically calculate appropriate DNB uncertainty addressable constants for COLSS and CPC reflecting the DNB margin impact of NGF. The 1/64 hypercube setpoints process as well as other process steps described in response to RAI 6 of WCAP-1 6500-P Supplement 1-P will be utilized in this analysis. WCAP-16500-P Supplement 1-P Revision 1-P (Reference 26) updates the response to RAI 6. In addition, the 6% interim margin penalty will be applied to the resultant addressable constants until its removal has been approved by the NRC.

6. Licensees are requiredto demonstrate that during transition cores, DNB margin gains associated with the NGF design offset (1) any impacts of flow starvation due to increasedpressure drop and (2) uncertainty associatedwith predicting local flow characteristics. Further,licensees must detail the analyticalmethods and results of their transition core LOCA and non-LOCA analyses (SE Sections 3.7 and 3.10).

First time engineering implementation analyses have been performed for transition and full cores of NGF. The analytical methods are defined in WCAP-16500-P for NGF implementation. For the transition cycle the COLSS on-line monitoring system and the CPC system will continue to utilize the current models and the CE-1 CHF correlation. For the first cycle of Waterford 3 that contains a batch of NGF (Cycle 16), the potential DNB margin gain, after accounting for the flow redistribution, is expected to be 12%. This margin gain is sufficient to compensate for any negative impact of the mixed core of NGF and standard fuel on transient and setpoint analyses. For non-LOCA analysis of the transition core NRC approved analytical methods are applied and results are based on a CETOP-D model for standard fuel so there is no transition core impact on transients. Full DNB margin credit for NGF will begin with the next cycle (Cycle

17) where the WSSV-T and ABB-NV DNB correlations will be used in the NGF CETOP-D model. The transition core LOCA evaluations for ECCS Performance including the implementation of CE 16 x 16 NGF assemblies are being finalized and will be submitted to NRC for review.

ANO response to this condition contained additional relevant information (Reference 5).

to W3F1-2009-0042 Page 10 of 15

'7. Implementation of CE 16 x 16 NGF assemblies necessitatere-analysis of the plant-specific LOCA analyses. Licensees are required to submit a license amendment containing the revised LOCA analyses for NRC review. Upon approval, the revised LOCA analyses constitute the analysis-of-recordand baseline for which future changes will be measuredagainstin accordance with 10 CFR 50.46(a)(3) (SE Section 3.7).

The revised LOCA analyses for ECCS Performance including the implementation of CE 16 x 16 NGF assemblies for full core configuration are being finalized and will be submitted to NRC for review. It should be noted that the introduction of NGF does not impact the post-LOCA long term cooling analysis. Upon approval and implementation of NGF, these revised LOCA analyses will constitute the new analysis-of-record and baseline for Waterford 3.

ANO response to this condition contained additional relevant information (Reference 5).

8. Using approved models and methods, Westinghouse will continue to limit peak local power experienced during Condition I and II events to ensure that fuel temperature remains below melting temperature at all burnups. This evaluation may be both plant and cycle-specific (SE Section 3.3.4).

Peak local power experienced during Condition I and II events will be limited to ensure fuel temperature remains below melting temperature at all burnups in accordance with Waterford 3 Technical Specification 2.1.1.2. This will be confirmed during each reload analysis.

9. The NRC staff's approvalof TR WCAP-16500-P establishes the licensing basis for batch implementation of the CE 16 x 16 NGF assembly design. Licensees wishing to implement this fuel design are required to submit a license amendment request, where applicable, updating their Core Operating Limits Report list of methodologies with the "A"version of this TR.

Letter W3F1 -2007-0037 (Reference 7) requested the COLR update and was approved under Reference 9.

10. The NRC staff's review did not include the LOCA model changes describedin Appendix A of TR WCAP-16500-P. Therefore, a licensee will have to submit a license amendment, if they desire to use The Appendix A LOCA model changes.

Changes to the LOCA model outlined in Appendix A of TR WCAP-1 6500-P were resubmitted to the NRC by Westinghouse under CENPD-1 32, Supplement 4-P-A, Addendum 1-P, and have been approved for use in license amendment to W3F1 -2009-0042 Page 11 of 15 applications as described in the CENPD-1 32, Supplement 4-P-A, Addendum 1-P section.

The conclusion of the SE for CENPD-1 32, Supplement 4-P-A, Addendum 1-P states that Limitations and Conditions 3, 4,.and 5 are appropriate for use when evaluating CE 16 x 16 NGF design fuel assemblies. The first two Limitations and Conditions included in the SE for CENPD-1 32, Supplement 4-P-A, Addendum 1-P are for fuel designs other than CE 16 x 16 NGF. The optional steam cooling model is not being used to support the implementation of CE 16 x 16 NGF assemblies at this time. However, the applicable Limitations and Conditions and the means of satisfying them are included below for future reference.

3. Limitation on the Optional Steam Cooling Heat Transfer Model The result of the grid model enhancement cannotresult in the use of a heat transfer coefficient greaterthan FLECHT. The FLECHT upper-bound heat transfer coefficient, as requiredby the current NRC licensing constraint,is also applied to the spacergrid optional steam cooling model improvement.

The computer code logic for the optional steam cooling heat transfer model in the STRIKIN-II hot rod heatup computer code contains a specific algorithm to insure that the current NRC licensing constraint on the use of

,the FLECHT upper-bound heat transfer coefficient is also applied to the spacer grid steam cooling model improvement calculated in the PARCH steam cooling module. Therefore, this limitation and condition is automatically satisfied when performing the licensing calculations using the version, of the STRIKIN-II computer code containing the approved optional steam cooling heat transfer model.

4. Use of the OptionalSteam Cooling Model If a licensee wants to use the optional steam cooling model, then a license amendment request should be submitted including the analyses performed to determine its applicabilityto the specific fuel design being evaluated, as discussed in CENPD-132Section3.3.1, 3.3.2, and 3.3.3 above. In addition, the licensee should provide the results of the evaluation with and without the optional steam cooling model, in a format similar to the graphical results provided in the reference calculations presentedin the supplemental TR. The PCT, local oxidation, and steam cooling flow rates should be included in the submittal. These comparisons will enable the NRC staff to confirm the acceptabilityof the use of the optional steam cooling model.

to W3F1-2009-0042 Page 12 of 15 If the optional steam cooling model were to be used for ECCS Performance Analyses at some time in the future, then a license amendment request would be submitted including the analyses and comparison graphical results needed to confirm the acceptability of the use of the optional steam cooling model.

5. Use of Flow Blockage and Reynolds Number Limits (Section 3.3.3)

For use of this topicalreport at a specific plant, the flow blockage and Reynolds number limits, as discussed in Section 3.3.3 above, should be confirmed by plant-specific analyses.

The computer code logic for the optional steam cooling heat transfer model in the PARCH module of the STRIKIN-II hot rod heatup computer code contains specific computational constraints to print warning and diagnostic output messages to alert the user if the calculation is found to be outside the range of applicability for flow blockage and Reynolds number. Therefore, this Limitation and Condition is automatically satisfied when performing the licensing calculations using the version of the STRIKIN-II computer code containing the approved optional steam cooling heat transfer model.

The NRC Staff reviewed the disposition of each SE condition and limitation for CENPD-132, Supplement 4-P-A, Addendum 1-P and found that Waterford 3 adequately addressed each one of them for this TR (Reference 9).

2. The NRC Staff may need to perform an audit of the Westinghouse calculation, at its Rockville, Maryland, offices. These calculations show how the value of 1.23 was determined from the WSSV-T and ABB-NV values of 1.12 and 1.13. This audit can be arranged for the week of July 13, 2009.

The exact date has to be decided.

For ANO-2, the NRC Staff performed an audit of the Westinghouse calculation, at its Rockville, Maryland, offices. These calculations show how the new safety limit was determined. However, the audit was not done for Waterford 3. Please describe how the safety limit DNBR of 1.24 was obtained, including codes and methods used.

The Modified Statistical Combination of-Uncertainties (MSCU) methodology presented in CEN-356 (Reference 20) was applied in Cycle 17 at the 95/95 confidence/probability level to verify a design limit of 1.24 on both WSSV-T and ABB-NV CHF correlations minimum DNBR (applicable to NGF assemblies).

The DNBR limit of 1.24 on WSSV-T and ABB-NV correlations has been shown to be valid for Cycle 17 due to the calculated Cycle 17 DNBR value being less than the 1.24 Analysis of Record (AOR) value. This MSCU derived limit of 1.24 is in.

Attachment I to W3F1-2009-0042 Page 13 of 15 compliance with Safety Evaluation Report (SER) constraints pertaining to CENPD-387-P-A (References 22) and WCAP-1 6523-P-A (Reference 23). This AOR DNBR limit of 1.24, in addition to other calculational and uncertainty factors, includes the following allowances:

1) NRC-specified allowances for the TORC code uncertainty, as discussed in Reference 19;

2) Rod bow penalty, as discussed below.

The effect of fuel rod bowing on DNBR margin has been incorporated in the safety and setpoint analysis in the manner discussed in CENPD-225-P-A (Reference 24). The penalty used for this analysis, 2.3% on minimum DNBR, is valid for assembly burnups up to 33 GWD/T. This penalty is included in the 1.24 DNBR limit.

For assemblies with burnup greater than 33 GWD/T, sufficient available margin exists to offset rod bow penalties due to the lower radial power peaks in these higher burnup assemblies. Hence, the rod bow penalty based upon CENPD-225-P-A for 33 GWD/T is applicable for all assembly burnups expected for Cycle 17.

The codes and calculation methods are explicitly described in CEN-356(V)-P-A (Reference 20) and WCAP-1 6500-P Supplement 1-P (Reference 26).

REFERENCES

1. W3F1-2009-0021, Waterford, Unit 3 - License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit, 6/3/09 (Agencywide Documents Access-and Management System (ADAMS) Accession No. ML091560027).

2. Waterford Steam Electric Station, Unit 3, RAI, Re: License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit (TAC No. ME1424), 8/11/09 (ADAMS Accession No. ML092310748).

3. 2CAN050901, Arkansas Nuclear One, Unit 2 - License Amendment Request to Revise the Departure From Nucleate Boiling Ratio (DNBR) Safety Limit, 5/13/09 (ADAMS Accession No. ML091340153).

4. Arkansas Nuclear One, Unit 2 - E-mail Request for Additional Information Regarding License Amendment Request, Modify Technical Specification 2.1.1.1, "DNBR," to Revise Departure from Nucleate Boiling Ratio Safety Limit (TAC ME1 328), 6/12/09 (ADAMS Accession No. ML091630573).

5. 2CAN070901, Arkansas Nuclear One, Unit 2, Response to Request for Additional Information for the License Amendment Request to Revise the to W3F1-2009-0042 Page 14 of 15 Departure from Nucleate Boiling Ratio (DNBR) Safety Limit, 7/8/09 (ADAMS Accession No. ML092050637).

6. CENPD-1 61-P-A, TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core, April 1986.

7. W3F1-2007-0037, Waterford Steam Electric Station, Unit 3, License Amendment Request NPF-38-271 to Support Next Generation Fuel, 8/2/07 (ADAMS' Accession No. ML0721180042).

8. CENPD-206-P-A, TORC Code, Verification and Simplified Modeling Methods, June 1981.

9. Waterford, Unit 3, License Amendment 214, Re: Request to Support Next Generation Fuel; Review and Approval of ECCS Performance Analysis; and Review and Approval of Supplement to ECCS Performance Analysis, 4/15/08 (ADAMS Accession No. ML080880014).

10. W3F1-2003-0074, Waterford, Request for Amendment, Increase Unit's Rated Thermal Power Level From 3441 Megawatts Thermal (MWt) to 3716 MWt, 11/13/03 (ADAMS Accession No. ML040260317).

11. Waterford, Unit 3, License Amendment 199 regarding Extended Power Uprate, 4/15/05 (ADAMS Accession No. ML051030068).

12. Palo Verde, Unit 1, Amendment 24, Revising Several Portions of TS to Incorporate Changes in Support if Cycle 2 Operation, 10/21/87 (ADAMS Accession No. ML021690079).

13. Waterford, Unit 3, License Amendment 102, Amendment Changes Appendix A TSs by Modifying Specifications Having Cycle Specific Parameter Limits by Replacing Values of Those Limits With Reference to a Core Operating Limits Report for Values of Those Limits, 3/1/95 (ADAMS Accession No. ML021780077).

14. W3P86-1686, Waterford 3 Cycle 2 Reload Analysis Report (RAR), 8/29/86.

15. W3P86-3331, Technical Specification Change Requests, 10/1/86.

16. Waterford, Unit 3, Amendment 12 & Safety Evaluation to License NPF-38.

Amendment revises Appendix A Technical Specifications, 1/16/87 (ADAMS Accession No. ML02117502117)

17. W3F1-92-0053, Waterford 3 Cycle 6 Reload Analysis Report (RAR), 4/24/92.

18. Arkansas Nuclear One Unit 2, Amendment No. 24, Authorizing Cycle-2 Operation Subject to Condition in License Which Temporarily Restricts Operation of Facility to Seventy Percent of Licensed Full Power Level of 2815 MWt Pending Completion of Staff Review, 6/19/81 (ADAMS Accession No. ML021490407).

19. Arkansas Nuclear One, Unit 2, Amendment.26, consisting of changes to license in accordance with satisfactory completion of issues previously required limiting to W3F1 -2009-0042 Page 15 of 15 authorized level to 70% of full power rating of 2815 MWT, 7/21/81 (ADAMS Accession No. ML021490394).

20. CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, May 1988.

21. CEN-1 39(A)-P, Statistical Combination of Uncertainties - Combination of System Parameter Uncertainties in Thermal Margin Analyses for Arkansas Nuclear One Unit 2, November 1980.

22. CENPD-387-P-A, Revision 00, ABB Critical Heat Flux Correlations for PWR Fuel, May 2000.

23. WCAP-1 6523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, August 2007.

24. CENPD-225-P-A, Fuel and Poison Rod Bowing, June 1983.

25. WCAP-1 6500-P-A, Revision 0, CE 16 x 16 Next Generation Fuel Core Reference Report, August 2007.

26. WCAP-16500-P, Supplement 1-P, Revision 1-P, Application of CE Methodology for CE 16X1 6 Next Generation Fuel (NGF), October 2008, (ADAMS Accession No. ML083050498).

27. Not Used

28. CEN-283(S)-P, Statistical Combination of Uncertainties, Part 1, June 1984, Part II and Part III, October 1984.

29. CEN-343(C)-P, Statistical Combination of Uncertainties for Waterford 3, October 1986.

30. CEN-338(C)-P, Statistical Combination of Uncertainties, August 1986.

9 Attachment 2 W3F1-2009-0042 List of Regulatory Commitments

Attachment 2 to W3F1 -2009-0042 Page 1 of I List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided-for information purposes and are not considered to be regulatory, commitments.

TYPE (Check One) SCHEDULED ONE-TIME CONTINUING COMPLETION DATE COMMITMENT ACTION COMPLIANCE (If Required)

WCAP-16500 Supplement 1 Revision After the WCAP-1 6500 1 Safety Evaluation Report limitations Supplement 1 Revision or conditions will be evaluated and 1 Safety Evaluation how they are met will be documented X Report is issued and in the implementation package of the prior to use as part of revision to the COLSS and CPC the reload process.

setpoints and the cycle specific COLR.