ML14031A120
| ML14031A120 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 01/23/2014 |
| From: | Mark D. Sartain Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14031A120 (14) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 23, 2014 10 CFR 50.90 U. S. Nuclear Regulatory Commission Serial No. 13-145A ATTN: Document Control Desk NL&OS/GDM R2 Washington, D. C. 20555 Docket Nos. 50-280/281 50-338/339 License Nos. DPR-32/37 NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST (LAR)
ADDITION OF AN ANALYTICAL METHODOLOGY TO THE NORTH ANNA AND SURRY CORE OPERATING LIMITS REPORTS (COLRS) AND AN INCREASE TO THE SURRY MINIMUM TEMPERATURE FOR CRITICALITY RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION By letter dated June 26, 2013 (Serial No.13-145), Virginia Electric and Power Company (Dominion) requested amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, and the Technical Specifications (TS) to Facility Operating License Numbers DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. The proposed LAR requests approval of the following items: 1) generic application of Appendix D, "Qualification of the ABB-NV and WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D Computer Code," to Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," 2) the plant specific application of Appendix D to DOM-NAF-2-A to North Anna and Surry Power Stations (in accordance with Section 2.1 of DOM-NAF A), and 3) an increase in the Surry Power Station TS Minimum Temperature for Criticality. The plant specific application of Appendix D to DOM-NAF-2-A requires the inclusion of the appendix to the TS list of references for determining core operating limits (i.e., the TS list of COLR references).
In an email dated December 24, 2013, the NRC provided a request for additional information (RAI) in support of their technical review. Dominion's response to the RAI is provided in the attachment to this letter.
The information provided in this letter does not affect the conclusion of the significant hazards consideration or the environmental assessment discussion contained in the Dominion letter dated June 26, 2013.
Serial No.13-145A Docket Nos. 50-280/281 and 50-338/339 Page 2 of 3 If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.
Sincerely, Mark D. Sartain Vice President - Nuclear Engineering and Development COMMONWEALTH OF VIRGINIA ))
COUNTY OF HENRICO )
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President of Nuclear Engineering and Development, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this ,j :-day of 2014.
My Commission Expires: 3/ 2C Notary Public Commonwealth of Virginia 140542 My Commission Expires May 31, 2014 Notary Public Commitments made in this letter: None
Attachment:
Response to NRC Request for Additional Information
Serial No.13-145A Docket Nos. 50-280/281 and 50-338/339 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station Ms. K. R. Cotton Gross NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Blvd.
Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health James Madison Building - 7 th Floor 109 Governor Street Room 730 Richmond, Virginia 23219
Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION Virginia Electric and Power Company (Dominion)
Surry Power Station Units I and 2 North Anna Power Station Units 1 and 2
Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION By letter dated June 26, 2013, Virginia Electric and Power Company (Dominion),
requested amendments to the Renewed Facility Operating License (FOL) Numbers NPF-4 and NPF-7 for North Anna Power Station (North Anna) Units 1 and 2, respectively, and the Technical Specifications (TS) to FOL Numbers DPR-32 and DPR-37 for Surry Power Station (Surry) Units 1 and 2, respectively. The proposed license amendment request (LAR) requested the approval of the following items:
(1) generic application of Appendix D, "Qualification of the ABB-NV and WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D Computer Code," to Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," (2) the plant specific application of Appendix D to DOM-NAF-2-A to North Anna and Surry (in accordance with Section 2.1 of DOM-NAF-2-A), and (3) an increase in the Surry TS minimum temperature for criticality.
The NRC technical review staff has reviewed the LAR and determined that additional information is needed to complete its evaluation. The NRC questions and the associated Dominion responses are provided below.
RAI-I Attachment 1. Sections 1. 2.3. and 5.1.2 (a) Describe the impact of increasingthe Minimum Temperature for Criticalityfor Surry Units on maintainingthe shutdown margin during the startup of the units.
Response
Shutdown margin (SDM) is defined in Surry TS as the amount by which the reactor core would be subcritical at hot shutdown (average temperature -> 547°F) conditions. By definition, changing the minimum temperature for criticality (from 522 0 F to 5380 F) does not technically affect SDM. The impact of the change is to the analytical calculation of SDM. As specified in internal analysis guidance documents for Dominion's Reload Safety Analysis Checklist (RSAC) process, the calculation of SDM conservatively addresses parameters which may produce conditions outside of the definition, such as instrument uncertainty and minimum temperature for criticality. An increase in the minimum temperature for criticality reduces the temperature range to be bounded in Dominion's RSAC process, resulting in an increase in the minimum available shutdown margin for two otherwise identical cores. The process of unit startup is unaffected by the increase in the minimum temperature for criticality. The increased minimum temperature for criticality TS limit will be reflected in the applicable station procedures.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment (b) It is stated that "the increased Minimum Temperature for Criticalitywill continue to be verified against the assumptions in the safety analyses on a reload basis and does not impact the NRC approved analytical methods used to determine the core operating limits such as the MTC. Describe the processes/proceduresin place to achieve this objective.
Response
The minimum temperature for criticality is an input to various RSAC calculations. The RSAC calculations are performed each cycle per Dominion's NRC-approved Reload Nuclear Design Methodology (Reference 1.1). The proposed value will be incorporated into the appropriate Dominion internal analysis guidance documents that include the minimum temperature for criticality, such as Moderator Temperature Coefficient (MTC) calculations. Increasing Surry's minimum temperature for criticality will result in required updates to affected Dominion internal analysis guidance documents to reflect the new minimum temperature for criticality.
(c) Explain how the increase in TS Minimum Temperature for Criticality coupled with introduction of the W-3 alternate CHF correlations at Surry units will provide (1) increased flexibility in loading pattern development as well as improved margins, and (2) improved predictive capabilities in determining the thermal-hydraulic performance at Surry, as claimed in Section 2.3 of Attachment I of the LAR.
Response
The improved predictive capabilities of the W-3 Alternate CHF correlations provide Dominion additional thermal margin as shown in the response to RAls 3(b) and 4 below.
This additional margin removes the reload core design constraint imposed by the Main Steam Line Break (MSLB) and Rod Withdrawal from Subcritical (RWSC) accident analyses, and reload axial power shape departure from nucleate boiling (DNB) analysis.
The increase in the minimum temperature for criticality results in more margin to the most positive MTC limit, which allows consideration of loading patterns with slightly higher soluble boron. Together these changes provide flexibility to consider more optimized loading patterns.
References 1.1. Topical Report, VEP-FRD-42-A, Rev. 2.1-A, "Reload Nuclear Design Methodology," August 2003.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment RAI-2 Attachment I Section 4.2 Pleaseprovide details of how the retainedDNBR margin is used to offset generic DNBR penalty for rod bow. Explain how the rod bow penalty is quantifiedin the analysis.
Response
Retained DNB ratio (DNBR) margin is quantified as the percent difference between the Safety Analysis Limit (SAL) and Deterministic Design Limit (DDL) as discussed in , Sections 4.2 and 4.3 of Reference 2.1 for North Anna and Surry, respectively. The DDLs are developed using the population statistics of the appropriate CHF correlation test databases and are set such that the probability of avoiding DNB will be at least 95% at a 95% confidence level (Attachment 6, Section D.6 of Reference 2.1 for ABB-NV and WLOP). Therefore, the DDLs are fixed and any changes to their values would require NRC review and approval. Dominion evaluates the Updated Final Safety Analysis Report (UFSAR) Chapter 14 and 15 safety analyses for Surry and North Anna, respectively, against the self-imposed SAL, which is greater than the DDL. The difference between the DDL and SAL provides DNB margin to offset the effect of rod bow and other DNB penalties, which are most effectively addressed on a generic basis.
The SALs for DNB analyses may be changed without prior NRC review and approval provided the changes meet the criteria established in Reference 2.2.
The method for calculating the rod bow penalty and using retained DNBR margin to offset the rod bow penalty is consistent with the North Anna and Surry licensing basis.
Dominion calculates the generic DNBR penalty for rod bow using the NRC approved methodology described in WCAP-8691, Revision 1 (References 2.3 and 2.4). Specific discussion can be found on the rod bow methodology and its application to retained DNBR margin in the North Anna UFSAR, Sections 4.4.1.1 and 4.4.2.3.4.2 and the Surry UFSAR, Section 3.4.3.5.
References 2.1. Letter from David A. Heacock (Dominion) to Document Control Desk (NRC),
"Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Proposed License Amendment Request (LAR),
Addition of an Analytical Methodology to the North Anna and Surry Core Operating Limits Reports (COLRS) and an Increase to the Surry Minimum Temperature for Criticality," Serial No.13-145, ADAMS Accession No. ML13179A014, June 26, 2013.
2.2. Technical Report, NEI 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation," Nuclear Energy Institute, November 2000.
2.3. Topical Report, WCAP-8691-P-A, Revision 1, "Fuel Rod Bow Evaluation," July 1979.
2.4. Letter from C. 0. Thomas (NRC) to E. P. Rahe, Jr. (Westinghouse), "Acceptance for Referencing of Licensing Topical Report WCAP-8691(P)ANCAP-8692(NP),"
December 29, 1982.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment RAI-3 Attachment 1, Sections 4.2 and 4.3 It is stated that WLOP CHF correlation is to be applied when the conditions occur outside of the range of applicabilityof the primary CHF correlation,say, for low pressure or low flow conditions. Please provide response to the questions below:
(a) Provide details of the correlation(s) currently used to analyze low pressure or low flow transients at Surry and North Anna Units.
Response
As described in Attachment 1, Section 2.1 of Reference 3.1, the WLOP CHF correlation is a replacement for the W-3 CHF correlation at low pressure and low flow operating conditions. The WLOP CHF correlation will replace the W-3 correlation as applied to the MSLB event because of the lower pressures encountered.
Dominion currently uses the W-3 correlation when determining the thermal-hydraulic performance of Westinghouse fuel products within North Anna and Surry's cores when the conditions occur outside of the range of applicability of the primary CHF correlation, such as at low pressure and low flow conditions. W-3 is one of the CHF correlations contained in the USNRC approved generic version of VIPRE-01 (References 3.2 and 3.3). Appendix B, Section B.2 and Appendix C, Section C.2 of DOM-NAF-2-A (Reference 3.4) describe the approved application of W-3 with the WRB-1 and WRB-2M CHF correlations in VIPRE-D, respectively. DOM-NAF-2-A, Appendix B has been incorporated into the Surry Core Operating Limits Report (COLR) list of approved methodologies and DOM-NAF-2-A, Appendix C has been incorporated into the North Anna COLR list of approved methodologies.
(b) What is the impact/advantage on DNBR and thermal performance margins when Surry and North Anna transition to WLOP CHF correlation to analyze low pressure or low flow transients.
Response
The advantage to DNBR and thermal performance margins with a North Anna and Surry transition to the WLOP CHF Correlation can be illustrated by the increase in retained DNBR margin. Retained DNBR margin is discussed in detail in the response to RAI 2. Currently, the W-3 CHF Correlation provides 9.7% retained DNBR margin at and above 1000 psia and 9.9% retained DNBR margin below 1000 psia for both Surry and North Anna (Attachment 4, Section 4.3 of References 3.5 and 3.6). The resulting available retained DNBR margin with use of WLOP in VIPRE-D will yield 12.8% retained DNBR margin for Surry (Attachment 1, Section 4.3 of Reference 3.1) and 21.2%
retained DNBR margin for North Anna (Attachment 1, Section 4.2 of Reference 3.1).
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment North Anna and Surry have different SALs for the WLOP CHF correlation due to the DNBR results for each site's specific MSLB transient analysis. The DNBR results from the site-specific transient evaluation set an upper bound on the allowed SAL.
Dominion's response to RAI 4 provides a sample minimum DNBR (MDNBR) comparison between W-3 and WLOP for a representative MSLB event.
References 3.1. Letter from David A. Heacock (Dominion) to Document Control Desk (NRC),
"Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Proposed License Amendment Request (LAR),
Addition of an Analytical Methodology to the North Anna and Surry Core Operating Limits Reports (COLRS) and an Increase to the Surry Minimum Temperature for Criticality," Serial No.13-145, ADAMS Accession No. ML042800118, dated June 26, 2013.
3.2. 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," dated May 1, 1986.
3.3. 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-2511-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores,' (TAC No. M79498)," dated October 30, 1993.
3.4. Fleet Report, DOM-NAF-2-A, Rev. 0.2-P-A, including Appendixes A, B and C, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," ADAMS Accession No. ML102390419, August 2010.
3.5. Letter from J. Alan Price (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request Relocation of Core Operating Limits to the Core Operating Limits Report (COLR) and Addition of COLR References," Serial No.09-581, ADAMS Accession No. ML092960616, October 16, 2009.
3.6. Letter from J. Alan Price (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR,"
Serial No.10-404, ADAMS Accession No. ML102020165, July 19, 2010.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment RAI-4 Attachment 1. Sections 4.2 and 4.3 Please show with typical calculations in support of your statement that there are no changes to the OTAT, OPA T, FAH, reactor protection system, or the Reactor Core Safety Limits (RCSLs) due to the implementation of the W-3 alternate CHF correlations at North Anna and Surry units.
Response
The OTAT, OPAT and f(AI) reactor protection functions are established to ensure bounding protection for the core thermal limits. The response below demonstrates that the ABB-NV and WLOP CHF correlations provide improved calculated thermal performance over the W-3 CHF correlation, and that the current reactor protection setpoints would continue to provide bounding protection if analyzed with the alternate CHF correlations. The RCSLs are unaffected by the W-3 Alternate CHF correlations as the RCSLs fall within the NRC-approved range of applicability of the primary CHF correlation (e.g. WRB-1 or WRB-2M).
Sections 4.2 and 4.3 of Reference 4.1 describe how the W-3 Alternate CHF correlations will be applied at North Anna and Surry and the impact of the W-3 Alternate correlations on the Reactor Protection System (RPS) and UFSAR transients. The ABB-NV CHF correlation is to be applied when the limiting location of DNBR occurs below the first mixing vane grid (MVG) and the WLOP CHF correlation is to be applied when the MDNBR condition occurs outside of the range of applicability of the primary CHF correlation. Specifically, the ABB-NV CHF correlation is employed in the presence of a highly bottom skewed power shape (when the limiting MDNBR occurs below the first MVG, e.g., RWSC) and the WLOP CHF correlation is employed at low pressure and low flow conditions (e.g., MSLB).
Dominion performed statepoint calculations across the licensing basis range of the W-3 CHF correlation at North Anna and Surry to quantify the DNB margins generated at each statepoint with use of the W-3 Alternate CHF correlations. The current licensing basis range of W-3 at North Anna and Surry includes the RWSC event (ABB-NV limiting), MSLB event (WLOP limiting), and the statepoints used in the development of reactor protection (RP) setpoints which result in a MDNBR below the first MVG (ABB-NV limiting). Table 4-1 lists the MDNBR results for a selection of North Anna and Surry UFSAR and RP statepoints evaluated to quantify the thermal-hydraulic performance margin provided with use of the alternate CHF correlations. The difference between the VIPRE-D result and the DDL is provided in Table 4-1 as the margin to the DDL:
/MoNBR - DL)
Margin to the DDL [%] =, DDL ) 100 The appropriate DDLs for comparison are contained within the table footnotes. The results of the calculations demonstrate that the minimum DNBR values are equal to or Page 6 of 10
Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment greater than the appropriate North Anna or Surry DNBR DDLs and SALs for the RCSLs, the OTAT, OPAT, and f(AI) reactor trip setpoints, as well as the applicable UFSAR events (i.e., RWSC and MSLB). Thus, the current RPS setpoints and RCSLs would continue to provide bounding protection of the core thermal limits when analyzed with the ABB-NV and WLOP CHF correlations in lieu of the W-3 CHF correlation.
Table 4-1: Sample Statepoints Used for the Development of North Anna and Surry DNBR Limits for the W-3 Alternate CHF Correlations Power W-3 ABB-NVIWLOP Case psure [% of Rated Margin to Mar
[psial Thermal Power] DDL1 [%] MDNBR DDL" to ABB-NV - RP1 2400 100 1.44 10.77 1.85 62.28 ABB-NV- RP2 2250 100 1.44 10.77 1.84 61.40 ABB-NV- RP3 2000 100 1.44 10.77 1.80 57.89 ABB-NV - RP4 1860 100 1.44 10.77 1.77 55.26 ABB-NV - RP5 2250 90 1.44 10.77 1.88 64.91 ABB-NV - RP6 2250 118 1.44 10.77 1.76 54.39 ABB-NV - RWSC 2320 35.4 2.01 54.62 2.34 105.26 WLOP - MSLB 869.8 23.7 2.34 61.38 2.01 64.75
- 1. VIPRE-D/V-3 DDL (2:1000 psia): 1.30; VIPRE-DAW-3 DDL (<1000 psia): 1.45 (References 4.2 and 4.3)
- 2. VIPRE-D/ABB-NV DDL: 1.14; VIPRE-DNVLOP DDL: 1.22 (Reference 4.1)
References 4.1. Letter from David A. Heacock (Dominion) to Document Control Desk (NRC),
"Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Proposed License Amendment Request (LAR),
Addition of an Analytical Methodology to the North Anna and Surry Core Operating Limits Reports (COLRS) and an Increase to the Surry Minimum Temperature for Criticality," Serial No.13-145, ADAMS Accession No. ML042800118, dated June 26, 2013.
4.2. Letter from J. Alan Price (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request Relocation of Core Operating Limits to the Core Operating Limits Report (COLR) and Addition of COLR References," Serial No.09-581, ADAMS Accession No. ML092960616, October 16, 2009.
4.3. Letter from J. Alan Price (Dominion) to Document Control Desk (NRC), "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR,"
Serial No.10-404, ADAMS Accession No. ML102020165, July 19, 2010.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment
RAI-5
Limitations and Conditions Number 3 for the Safety Evaluation Report for WCAP-14565-P-A stipulates that Selection of the appropriate DNB correlation, DNBR limit, engineering hot channel factors for enthalpy rise, and other fuel dependent parameters will be justified for each application of each correlation on a plant specific basis.
This is also listed as Number 7 of Section 7.0 Conclusions of Topical Report WCAP- 14565-P-A.
Please describe how the Limitations and Conditions number 3 is justified/implemented in connection with the use of ABB-NV and WLOP correlationsin North Anna and Surry units.
Response
Dominion performs DNBR analysis of the North Anna and Surry Units in accordance with the approved thermal-hydraulic methodology outlined in DOM-NAF-2-A (Reference 5.1). Consistent with the requirements of DOM-NAF-2-A, qualification of the ABB-NV and WLOP CHF correlations has been performed and the results have been documented in Appendix D to DOM-NAF-2-A (Reference 5.2, Attachment 6).
Furthermore, use of the ABB-NV and WLOP CHF correlations in licensing DNBR calculations for the North Anna and Surry Units is governed by DOM-NAF-2-A.
Section 2.2 of Reference 5.1 provides the conditions under which Dominion applies the VIPRE-D computer code to meet the USNRC's requirements listed in the VIPRE-01 Safety Evaluation Report (SER) (References 5.3 and 5.4) which satisfy Limitation and Condition Number 3 of the WCAP-14565-P-A, Addendum 2-P-A SER (Reference 5.5).
The specific VIPRE-D conditions which comply with Limitation and Condition Number 3 of Reference 5.5 are provided below:
Condition 2 of Section 2.2 "Compliance with VIPRE-01 SER" of DOM-NAF-2-A states: "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."
Section B.2 of DOM-NAF-2-A, Appendix B and Section C.2 of DOM-NAF-2-A, Appendix C state the range of applicability of the W-3 CHF correlation with use of the WRB-1 and WRB-2M CHF correlations, respectively. The ABB-NV and WLOP CHF correlations are used as alternatives to the W-3 correlation where W-3 is applicable. The applicability of the W-3 Alternate CHF correlations is provided in Section D.2 of DOM-NAF-2-A, Appendix D: "The ABB-NV and WLOP CHF correlations are applicable for use in the thermal-hydraulic evaluation of Page 8 of 10
Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment 14x14 fuel products with a rod outside diameter (OD) of 0.400 or 0.422 inches; 15x15 fuel products with a rod OD of 0.422 inches; 16x16 fuel products with a rod OD of 0.360 or 0.374 inches; and 17x17 fuel products with a rod OD of 0.360 or 0.374 inches (Reference D4)." Therefore, condition 2 of DOM-NAF-2-A, Section 2.2 with the approved CHF correlation applicability ranges discussed above satisfy the "selection of the appropriate DNB correlation" and "DNBR limit" portion of Limitation and Condition Number 3 of the WCAP-14565-P-A, Addendum 2-P-A SER.
Condition 3 of Section 2.2, "Compliance with VIPRE-01 SER" of DOM-NAF-2-A states: "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."
The engineering enthalpy-rise hot channel factor is obtained from the fuel vendor for a given fuel type and is independent of the DNB correlation. The hot channel factor is accounted for in licensing DNBR calculations as described in Section 4.10 of DOM-NAF-2-A (Reference 5.1). The selection of other fuel dependent parameters and their use in licensing DNBR calculations are outlined in Section 4.0 of DOM-NAF-2-A. As noted above, Dominion performs DNBR analysis of the North Anna and Surry Units in accordance with the approved thermal-hydraulic methodology outlined in DOM-NAF-2-A. Therefore, condition 3 of DOM-NAF-2-A, Section 2.2 with the approved VIPRE-D modeling criteria described in Section 4.0 of DOM-NAF-2-A satisfy the selection of the "appropriate engineering hot channel factors for enthalpy rise" and "other fuel dependent parameters" portion of Limitation and Condition Number 3 of the WCAP-14565-P-A, Addendum 2-P-A SER.
References 5.1. Fleet Report, DOM-NAF-2-A, Rev. 0.2-P-A, including Appendixes A, B and C, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," ADAMS Accession No. ML102390419, August 2010.
5.2. Letter from David A. Heacock (Dominion) to Document Control Desk (NRC),
"Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Proposed License Amendment Request (LAR),
Addition of an Analytical Methodology to the North Anna and Surry Core Operating Limits Reports (COLRS) and an Increase to the Surry Minimum Temperature for Criticality," Serial No.13-145, ADAMS Accession No. ML042800118, dated June 26, 2013.
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Serial No. 13-145A Docket Nos. 50-280/281 and 50-338/339 Attachment 5.3. 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.
5.4. 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-2511-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores,' (TAC No. M79498)," October 30, 1993.
5.5. Letter from Ho K. Nieh (NRC) to James A. Gresham (Westinghouse), "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-14565-P, Addendum 2, Revision 0, 'Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP [Westinghouse Low Pressure] for PWR [Pressurized Water Reactor] Low Pressure Applications' (TAC No. MD3184)," ADAMS Accession No.
ML080360381, February 14, 2008.
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