NL-04-0973, Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2

Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2

ML042320393
Person / Time
Site:	Vogtle
Issue date:	08/13/2004
From:	Gasser J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-0973, WCAP-14416-NP-A, Rev 1
Download: ML042320393 (74)

Text

Jeffrey T. Gasser Southern Nuclear Vfice President Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7721 Fax 205.992.0403 S

SOUTHERN August 13, 2004 COMPANY Docket Nos.:

50-424 NL-04-0973 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise the Vogtle Electric Generating Plant (VEGP)

Technical Specifications (TS) to reflect updated spent fuel rack criticality analyses for Units 1 and 2.

The VEGP current licensing basis permits credit for soluble boron in the spent fuel rack criticality analyses. The current analyses are based on the methodology described in Westinghouse topical report WCAP-14416-NP-A, Revision 1, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology." Subsequent to the issuance of this report, Westinghouse notified licensees of potential non-conservatisms in the axial shape bias and the reactivity equivalencing techniques used in the analyses. Margin in the current analyses was used to demonstrate that the current Technical Specifications continue to provide their intended level of protection.

To address the non-conservatisms and to recover the margin, SNC has chosen to reanalyze the criticality analyses for the VEGP Unit 1 and Unit 2 spent fuel racks.

Westinghouse performed the analyses using methods that address the non-conservatisms previously identified. The methodology is analogous to that described in WCAP-14416-NP-A. The application of the methods used in the revised analyses has been approved by the NRC for R. E. Ginna (Amendment 79 to Facility Operating License DPR-1 8, dated December 7, 2000), Diablo Canyon Power Plant (Amendment 154 to Facility Operating Licenses DPR-80 and DPR-82, dated September 25, 2002), and Millstone Power Station Unit 2 (Amendment 274 to Facility Operating License DPR-65, dated April 1, 2003).

The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The storage configurations and interface requirements in the current Technical Specifications were retained in the revised analyses. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP continues to remain valid. The

U. S. Nuclear Regulatory Commission NL-04-0973 Page 2 analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).

SNC proposes to revise TS 3.7.18, 4.3.1.1, and 4.3.1.2 to reflect the results of the revised analyses described above. In addition, changes to the Bases for TS 3.7.17 and TS 3.7.18 are included to reflect the results of the revised analyses described above.

With the issuance of the revised Technical Specification pages for Amendments 130 and 109 for Units 1 and 2, respectively, the page number for Figure 5.5.6-1 changed from Page 5.5-22 to 5.5-23. However, SNC failed to update Page vi of the Table of Contents to reflect the correct page number in the license amendment request that resulted in Amendments 130 and 109. The page is being corrected to reflect the current page number for Figure 5.5.6-1. There is no change to the figure itself and the conclusions of the safety evaluation report (SER) for Amendments 130 and 109 for Unit Is and 2, respectively, are not affected. This change is considered to be administrative in nature. provides a description and justification for the proposed change. Enclosure 2 contains the 10 CFR 50.92 evaluation and the justification for the categorical exclusion from performing an environmental assessment. Enclosure 3 provides the marked-up Technical Specifications and Bases pages. Enclosure 4 provides the clean-typed Technical Specifications and Bases pages. Enclosures 5 and 6 contain the spent fuel pool criticality analysis reports for Units I and 2, respectively.

SNC requests the NRC review and approve this amendment request no later than July 31, 2005. This amendment will be implemented 60 days after approval.

In accordance with the requirements of 10 CFR 50.91, a copy of this letter and all applicable enclosures will be sent to the designated State official of the Environmental Protection Division of the Georgia Department of Natural Resources.

(Affirmation and signature are on the following page.)

U. S. Nuclear Regulatory Commission NL-04-0973 Page 3 Mr. J. T. Gasser states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company, and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments. If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY Jeffrey T. Gasser

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2004.

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Enclosures:

1. Description of and Justification for Proposed Changes

2. 10 CFR 50.92 Significant Hazards Evaluation and Environmental Assessment

3. Marked-up Technical Specifications and Bases Pages

4. Clean-Typed Technical Specifications and Bases Pages S. Unit 1 Spent Fuel Pool Criticality Analysis Report

6. Unit 2 Spent Fuel Pool Criticality Analysis Report cc:

Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. W. F. Kitchens, General Manager - Plant Vogtle Mr. K. R. Holmes, Performance Analysis Supervisor - Plant Vogtle RType: CVC7000 U. S. Nuclear Regulatorv Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Vogtle Mr. G. J. McCoy, Senior Resident Inspector - Vogtle State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES Description of Proposed Change The VEGP current licensing basis permits credit for soluble boron in the spent fuel rack criticality analyses. The current analyses are based on the methodology described in Westinghouse topical report WCAP-14416-NP-A, Revision 1, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology." Subsequent to the issuance of this report, Westinghouse notified licensees of potential non-conservatisms in the axial shape bias and the reactivity equivalencing techniques used in the analyses. Margin in the current analyses was used to demonstrate that the current Technical Specifications continue to provide their intended level of protection.

To address the non-conservatisms and to recover the margin, SNC has chosen to reanalyze the criticality analyses for the VEGP Unit 1 and Unit 2 spent fuel racks. Westinghouse performed the analyses using methods that address the non-conservatisms previously identified. The methodology is analogous to that described in WCAP-14416-NP-A. The application of the methods used in the revised analyses has been approved by the NRC for R. E. Ginna (Amendment 79 to Facility Operating License DPR-18, dated December 7, 2000), Diablo Canyon Power Plant (Amendment 154 to Facility Operating Licenses DPR-80 and DPR-82, dated September 25, 2002), and Millstone Power Station Unit 2 (Amendment 274 to Facility Operating License DPR-65, dated April 1, 2003).

The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The storage configurations and interface requirements in the current Technical Specifications were retained in the revised analyses. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP continues to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).

SNC proposes to revise TS 3.7.18, 4.3.1.1, and 4.3.1.2 to reflect the results of the revised analyses described above. In addition, changes to the Bases for TS 3.7.17 and TS 3.7.18 are included to reflect the results of the revised analyses described above.

With the issuance of the revised Technical Specification pages for Amendments 130 and 109 for Units 1 and 2, respectively, the page number for Figure 5.5.6-1 changed from Page 5.5-22 to 5.5-

23. However, SNC failed to update Page vi of the Table of Contents to reflect the correct page number in the license amendment request that resulted in Amendments 130 and 109. The page is being corrected to reflect the current page number for Figure 5.5.6-1. There is no change to the figure itself and the conclusions of the safety evaluation report (SER) for Amendments 130 and 109 for Units 1 and 2, respectively, are not affected. This change is considered to be administrative in nature.

El-l

Enclosure I Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES Basis for Proposed Change Summary of Technical Specification and Bases Changes Associated with Rack Criticality Analyses Current Description of Change Table of Contents Figure 3.7.18-1 Figure 3.7.18-2 Bases 3.7.17 Bases 3.7.18 TS 4.3.1.1 TS 4.3.1.2 Figure 4.3.1-1 Figure 4.3.1-2 Figure 4.3.1-3 Figure 4.3.1-4 Figure 4.3.1-5 Figure 4.3.1-6 Figure 4.3.1-7 Figure 4.3.1-8 Figure 4.3.1-9 Updated to reflect changes in Figures Replaced with revised results for Unit 1 Replaced with revised results for Unit 2 Updated description of methods, analyses, and boron concentration requirements Updated description of enrichment requirements for the various storage configurations Update boron concentration requirements, removed reference Kcz requirement, add reference to minimum Integral Fuel Burnable Absorber (IFBA) requirements for Unit 1 Update boron concentration requirements, removed reference Koo requirement, add reference to minimum Integral Fuel Burnable Absorber (IFBA) requirements for Unit 2 Removed Updated results and changed to Figure 4.3.1-8 Updated results and changed to Figure 4.3.1-10 No change to information but changed to Figure 4.3.1-1 No change to information but changed to Figure 4.3.1-2 No change to information but changed to Figure 4.3.1-3 No change to information but changed to Figure 4.3.1-4 No change to information but changed to Figure 4.3.1-5 No change to information but changed to Figure 4.3.1-6 New Figure 4.3.1-7 added New Figure 4.3.1-9 added Current VEGP Licensing Basis SNC has previously submitted license amendment requests to credit boron in the spent fuel pools.

These requests were granted in Technical Specification Amendments 99/77 on February 20, 1998, and Amendments 102/80 on June 29, 1998. Amendments 99/77 were the first application of boron credit at VEGP. Amendments 102/80 were the results of revised analyses for Unit I in support of re-racking the Unit 1 spent fuel pool.

The credit for boron was originally based on the methodology described in Westinghouse topical report WCAP-14416-NP-A, Revision 1, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology." This topical was approved by the NRC on October 25, 1996, and issued by Westinghouse in November 1996.

The methodology provided for limited credit for soluble boron in the spent fuel pool to maintain Keff < 0.95. The criteria set forth in this topical report were that Keff remains less than unity E1-2 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units I and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES with zero soluble boron and that Keff remains less than or equal to 0.95 with credit for soluble boron with a 95% probability at a 95% confidence level (95/95). Fuel enrichments up to 5 weight percent (5 w/o) U-235 were considered in the analyses. In some cases, it was necessary to credit burnup or integral fuel burnable absorbers (IFBA) in spent fuel rack geometry, or to maintain a reference Keff below some specified value in reactor geometry, in order to ensure that the spent fuel pool Keff remained less than unity with zero soluble boron. Several storage configurations, as well as interfaces between these configurations, were considered.

Issues With Current Licensing Basis Methodology Westinghouse issued two Nuclear Safety Advisory Letters (NSAL) reporting potential non-conservatisms in the methodology described in WCAP-14416-NP-A. NSAL-99-003 discussed potential non-conservatisms in the calculated IFBA requirements using the reference Keff technique for reactivity equivalencing. NSAL-00-0 15 discussed potential non-conservatisms in the axial burnup shape reactivity bias. In addition, the NRC issued Regulatory Information Summary RIS 01-012 to notify licensees of the potential for non-conservatisms in spent fuel pool criticality analyses if reactivity equivalencing is used.

An evaluation of these issues was performed for VEGP. Margin in the current analyses was used to demonstrate that the current Technical Specifications continue to provide their intended level of protection.

Description of Revised Analyses SNC decided to update the VEGP spent fuel rack criticality analyses utilizing methods that address the issues described above. The goals of the reanalyses were to show that the acceptance criteria set forth in WCAP-14416-NP-A continue to be met and that the currently permissible storage configurations described in the Technical Specifications continue to be acceptable. The results of the analyses provided updated soluble boron, burnup credit, and IFBA credit requirements. No physical plant changes are being made, i.e., no changes to the spent fuel pools or racks, heat loads, supporting systems, etc. Only the criticality analyses are being updated.

The analysis methodology employs: (1) SCALE-PC, a personal computer version of the SCALE-4.3 code system, with the updated SCALE4.3 version of the 44 group ENDF/B-V neutron cross section library, and (2) the two-dimensional integral transport code DIT with an ENDF/B-VI neutron cross section library.

SCALE-PC was used for calculations involving infinite arrays for the "2-out-of-4", "3-out-of-4",

"All-Cell", and "3x3" fuel assembly storage configurations. In addition, it was employed in a full pool representation of the storage racks to evaluate soluble boron worth and postulated accidents.

SCALE-PC, used in both the benchmarking and the fuel assembly storage configurations, includes the control module CSAS25 and the following functional modules: BONAMI, NITAWL-II, and KENO V.a.

The DIT code is used for simulation of in-reactor fuel assembly depletion. KENO V.a was used in the calculation of biases and uncertainties.

E1-3 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES Models were made for each storage configuration as well as for the entire pools. Each configuration was modeled as an infinitely repeating pattern in the X-Y plane. A water reflector was modeled above and below the spent fuel storage cells to account for axial reactivity effects.

The KENO model for the entire pool modeled each individual storage rack module.

KENO was used to calculate Keff for the various storage configurations considered as well as for the entire pool. For demonstrating that Keff remains below unity for zero soluble boron, SNC chose to apply an acceptance criterion of 0.995. For the storage configurations considered, the target value of Keff for these calculations was selected to be less than 0.995 by an amount sufficient to cover the magnitude of the analytical biases and uncertainties. KENO was also used to determine burnup and IFBA versus enrichment requirements to meet the target Keff for configurations that credit burnup or IFBA.

Calculations were performed for the entire pool with various fuel storage configurations to demonstrate that the Keff for the entire pool remains below 0.995 with zero soluble boron.

KENO was used to determine the soluble boron requirements for non-accident and accident conditions to ensure that Keff remains less than or equal to 0.95.

The details of the modeling and analyses are provided in Section 3.0 of the reports in Enclosures 5 and 6.

Licensing Precedence The application of the above-described methodology for spent fuel rack criticality analyses has been approved by the NRC for the following plants:

1. R. E. Ginna Nuclear Power Plant, Amendment 79 to Facility Operating License DPR-18, dated December 7, 2000,

2. Diablo Canyon Power Plant, Amendment 154 to Facility Operating Licenses DPR-80 and DPR-82, dated September 25, 2002, and

3. Millstone Power Station Unit 2, Amendment 274 to Facility Operating License DPR-65, dated April 1, 2003.

Summary of Results The key results are:

1. The soluble boron concentration requirement to maintain Keff less than or equal to 0.95 is 511 ppm for the Unit 1 pool and 394 ppm for the Unit 2 pool for non-accident conditions.

2. The soluble boron concentration requirement to mitigate postulated accidents is 851 ppm for the Unit 1 pool and 1098 ppm for the Unit 2 pool.

E14 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units I and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES

3. Westinghouse 17 x 17 fuel assemblies with nominal enrichments no greater than 3.556 w/o U-235 may be stored in all storage cell locations of the Unit 1 pool. Fuel assemblies with initial nominal enrichments greater than 3.556 w/o U-235 must satisfy a minimum burnup requirement as shown in TS Figure 3.7.18-1 or satisfy a minimum IFBA requirement as shown in TS Figure 4.3.1-7.

4. Westinghouse 17 x 17 fuel assemblies with nominal enrichments no greater than 5.0 w/o U-235 may be stored in a 3-out-of-4 checkerboard arrangement with empty cells in the Unit 1 pool. There is no minimum burnup requirement for this configuration.

5. Westinghouse 17 x 17 fuel assemblies with nominal enrichments no greater than 1.73 w/o U-235 may be stored in all storage cell locations of the Unit 2 pool. Fuel assemblies with initial nominal enrichments greater than 1.73 w/o U-235 must satisfy a minimum burnup requirement as shown in TS Figure 3.7.18-2.

6. Westinghouse 17 x 17 fuel assemblies with nominal enrichments no greater than 2.40 w/o U-235 may be stored in a 3-out-of-4 checkerboard arrangement with empty cells in the Unit 2 pool. Fuel assemblies with initial nominal enrichment greater than 2.40 w/o U-235 must satisfy a minimum burnup requirement as shown in TS Figure 4.3.1-8.

7. Westinghouse 17 x 17 fuel assemblies with nominal enrichments no greater than 5.0 w/o U-235 may be stored in a 2-out-of-4 checkerboard arrangement with empty cells in the Unit 2 pool. There is no minimum burnup requirement for this configuration.

8. Westinghouse 17 x 17 fuel assemblies may be stored in the Unit 2 pool in a 3 x 3 array.

The center assembly must have an initial enrichment no greater than 3.2 w/o U-235 or satisfy a minimum IFBA requirement for higher enrichments as shown in TS Figure 4.3.1-9. The surrounding fuel assemblies must have an initial nominal enrichment no greater than 1.39 w/o U-235 or satisfy a minimum burnup and decay time requirement for higher initial enrichments as shown in TS Figure 4.3.1-10.

9. Keff for the entire pool was demonstrated to be less than unity for the various storage configurations considered.

10. The allowed storage configurations and interface requirements in the current Technical Specifications continue to remain valid.

A detailed discussion of the results of the analyses is provided in Section 3.0 of the reports in Enclosures 5 and 6. A summary of the results is provided in Section 4.0 of the reports in Enclosures 5 and 6.

Boron Dilution Event A spent fuel pool dilution evaluation was presented in support of Amendments 99 and 77 for Units 1 and 2, respectively. The Staff evaluation is described in the Safety Evaluation Report (SER) for these amendments. The dilution evaluation addressed a dilution from the minimum boron concentration requirement of 2000 ppm (TS 3.7.17) to 600 ppm. The evaluation concluded E1-5 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 DESCRIPTION OF AND JUSTIFICATION FOR PROPOSED CHANGES that a dilution event would be detected by alarms and plant personnel and will be terminated prior to reaching 600 ppm. For non-accident conditions, the boron concentration requirement to maintain Keff less than or equal to 0.95 is 511 ppm for the Unit 1 pool and 394 ppm for the Unit 2 pool. Therefore, the conclusion that the dilution event will be detected and mitigated prior to exceeding Keff of 0.95 remains valid and no new dilution evaluation was required.

Conclusions Based on the results of the analyses discussed above and in Enclosures 5 and 6, the following conclusions can be made:

1. Keff remains below unity for the various storage configurations considered with zero soluble boron with a 95% probability at a 95% confidence level (95/95) thereby meeting the requirements of Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50),

Appendix A, General Design Criteria (GDC) 62.

2. Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).

3.

The current storage configurations in TS 3.7.18 and TS 4.3.1 continue to remain unchanged.

4. The current interfaces between storage configurations in TS 3.7.18 and TS 4.3.1 continue to remain unchanged.

5. The revised burnup and IFBA versus enrichment limits ensure that the first conclusion continues to be met.

6. The current minimum spent fuel pool boron concentration limit in TS 3.7.17 and the current dilution event analysis continue to ensure that any credible dilution event could be terminated before reaching a boron concentration corresponding to Keff greater than 0.95.

EI-6 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION AND ENVIRONMENTAL ASSESSMENT Description of Proposed Change The VEGP current licensing basis permits credit for soluble boron in the spent fuel rack criticality analyses. The current analyses are based on the methodology described in Westinghouse topical report WCAP-14416-NP-A, Revision 1, "Westinghouse Spent Fuel Rack Criticality Analysis Methodology." Subsequent to the issuance of this report, Westinghouse notified licensees of potential non-conservatisms in the axial shape bias and the reactivity equivalencing techniques used in the analyses. Margin in the current analyses was used to demonstrate that the current Technical Specifications continue to provide their intended level of protection.

To address the non-conservatisms and to recover the margin, SNC has chosen to reanalyze the criticality analyses for the VEGP Unit 1 and Unit 2 spent fuel racks. Westinghouse performed the analyses using methods that address the non-conservatisms previously identified. The methodology is analogous to that described in WCAP-14416-NP-A. The application of the methods used in the revised analyses has been approved by the NRC for R. E. Ginna (Amendment 79 to Facility Operating License DPR-18, dated December 7, 2000), Diablo Canyon Power Plant (Amendment 154 to Facility Operating Licenses DPR-80 and DPR-82, dated September 25, 2002), and Millstone Power Station Unit 2 (Amendment 274 to Facility Operating License DPR-65, dated April 1, 2003).

The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The storage configurations and interface requirements in the current Technical Specifications were retained in the revised analyses. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP continues to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).

SNC proposes to revise TS 3.7.18, 4.3.1.1, and 4.3.1.2 to reflect the results of the revised analyses described above. In addition, changes to the Bases for TS 3.7.17 and TS 3.7.18 are included to reflect the results of the revised analyses described above.

With the issuance of the revised Technical Specification pages for Amendments 130 and 109 for Units I and 2, respectively, the page number for Figure 5.5.6-1 changed from Page 5.5-22 to 5.5-

23. However, SNC failed to update Page vi of the Table of Contents to reflect the correct page number in the license amendment request that resulted in Amendments 130 and 109. The page is being corrected to reflect the current page number for Figure 5.5.6-1. There is no change to the figure itself and the conclusions of the safety evaluation report (SER) for Amendments 130 and 109 for Units 1 and 2, respectively, are not affected. This change is considered to be administrative in nature.

10 CFR 50.92 Significant Hazards Evaluation In 10 CFR 50.92(c), the Nuclear Regulatory Commission (NRC) provides the following standards to be used in determining the existence of a significant hazards consideration:

E2-1 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION AND ENVIRONMENTAL ASSESSMENT

.a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22, or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in the margin of safety.

Southern Nuclear Operating Company (SNC) has reviewed the proposed amendment request and determined that its adoption does not involve a significant hazards consideration based upon the following discussion:

I. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

SNC has chosen to reanalyze the criticality analyses for the VEGP Unit 1 and Unit 2 spent fuel racks. Westinghouse performed the revised analyses using methods that address the non-conservatisms previously identified in the current analyses. The methodologies used for the revised analyses have been previously approved for use by the NRC.

The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The storage configurations and interface requirements in the current Technical Specifications were retained in the revised analyses. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP continues to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).

Core design procedures ensure that new fuel can be stored in one or more of the allowed storage configurations. Administrative controls during fuel fabrication ensure that the fuel is fabricated accordingly to ensure proper loading of fuel in the fuel assemblies.

Administrative controls used to load fuel assemblies into the spent fuel pool ensure that fuel assemblies are stored in compliance with the allowed storage configurations. Fuel handling is performed under many administrative controls and physical limitations.

These controls provide reasonable assurance that a criticality accident, fuel fabrication error, or fuel handling accident will not occur.

The change to the page number of Figure 5.5.6-1 on Page vi of the Table of Contents is administrative in nature.

E2-2 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units I and 2 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION AND ENVIRONMENTAL ASSESSMENT Therefore, based on the conclusions of the above analysis, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?

The types of accidents previously evaluated include fuel fabrication errors, criticality accidents, and fuel handling accidents. The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. No new or other kind of accident can be postulated as a result of the revised analyses.

The change to the page number of Figure 5.5.6-1 on Page vi of the Table of Contents is administrative in nature.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant decrease in the margin of safety?

The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP was shown to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95%

confidence level (95/95).

The change to the page number of Figure 5.5.6-1 on Page vi of the Table of Contents is administrative in nature.

Therefore, the proposed change does not involve a significant decrease in the margin of safety.

Environmental Assessment Southern Nuclear has evaluated the proposed changes and determined the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in the individual or cumulative occupational exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and an environmental assessment of the proposed changes is not required.

E2-3 Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 MARKED-UP TECHNICAL SPECIFICATION AND BASES PAGES

TABLE OF CONTENTS LIST OF FIGURES 2.1.1-1 3.4.16-1 3.7.18-1 3.7.18-2 Reactor Core Safety Limits....................

2.0-2 Reactor Coolant Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power with the Reactor Coolant Specific Activity > 1 pCi/gram Dose Equivalent 1-131..................

3.4.16 Vogtle Unit I Burnup Credit Requirements for All Cell Storage.........

3.7.1 Vogtle Unit 2 Bumup Credit Requirements for All Cell Storage 3.7.1 Vogtle Unit 1 Bumup Credit Requirements for 1%

-i

- At A!t

% n a 6-4 8-

[A q 1 1"-,

7-Ia J-UI ord 1 atworape.........................................................................

.U 4.3.1 Vogtle Unit 2 Bumup Credit Requirements for 3-out-of-4 Storage....................

4.0 4.3.1 Vogtle Unit 2 Bumup Credit Requirements for 3x3 Storage.

...................................................................... 4.0 4.3.1 Vogtle Units I and 2 Empty Cell Checkerboard Peripheral E

fjj Storage Configurations.........................

4.0 4.3.1 Vogtle Unit 2 3x3 Checkerboard Storage Configuration.4.0 4.3.1-Vogtle Units 1 and 2 Interface Requirements Nj (All Cell to Checkerboard Storage).4.0 4.3.1 Vogtle Unit 2 Interface Requirements (Checkerboard Storage Interface).4.0 4.3.1 Vogtle Unit 2 Interface Requirements

\\W (3x3 Checkerboard to All Cell Storage).4.0 4.3.1 Vogtle Unit 2 Interface Requirements Insert (3x3 to Empty Cell Checkerboard Storage).4.0

/

Schedule of Lift-Off Testing for Two

[3Int Containments at a Site...............

5.5

-7

-8 Assemblies forl

-9

-10

-11

-12

-13

-14

/

I l Note: Items 4.3.1.1 through 4.3.1.10 will be reordered in sequence and pages renumbered l

I accordingly.

s:\\vogTS&B\\TS-TOC.d0c Vogtle Units 1 and 2 vi Amendment No. 112 (Unit 1)

Amendment No.

90 (Unit 2)

INSERTS TO TABLE OF CONTENTS INSERT A Figure 4.3.1-7 INSERT B Figure 4.3.1-9 Vogtle Unit 1 IFBA Credit Requirements for All Cell Storage Vogtle Unit 2 IFBA Credit Requirements for Center Assembly for 3x3 Storage

Fuel Assembly Storage in the Fuel Storage Pool 3.7.18 Figure 3.7.18-1 Vogtle Units 1 and 2 Vogtle Unit 1 Bumup Credit Requirements for All Cell Storage 3.7.18-3 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

INSERT "A" FIGURE 3.7.18-1 9,000 8.000 7.000 7,000 5

6,000 1 5,n00 C.

c E

X 4,000 8

3,000 2,000 1,000 0

_---T----

_ _ _ _ _ _ _ _ _ _ _ _ I __ _ _

=-_

=

=

=

=

=

=

=

=

=

X X

_ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _- Z _

fi = = = = = = _ = _ = _ = ===

= _r

= _

_ = _ = _ _ _ = = _ _ _ _== = f = = _

=

=

=

=

=

=

X

=

=

=

=

=

=

== S b

=

. = = = = = = = = = = = = ===

a

==

_ __ CE 'TA 3U}-----E-g E---

. _ _ _ _ _ _ _ _ _ _ _ _ _ 7 _ _ __

_ _ _ _ _ _ _ = = = _ _ _ Z:_ = =-= _

= = = = = = = = = =

,-=

= = _ = _ _

7 =-_

= = = = = _ = = = _ = r = = === =

._ _ ____ __ _ __ /- _ _ ____ _ l

_ _ _ _ _ _ _ _ _ _ f _ {!AC :CE'TS BLE 2 _ _

=

=

=

=

=

=

=

=

=

I T

1 1

1

=

=

._ _ _ _ _ - - - E - r X E t

_ _ _ _ _ _ _ X _ _ _ _ W+1 T r g _ _ =- _ I I I I I v _ _ _

1 1 1 1 1 3.0 3.5 4.0 4.5 5.0 Initial 235U Enrichment inominal w/o)

Fuel Assembly Storage in the Fuel Storage Pool 3.7.18 500 4500C 40000 35 0

30000 E 25000 co u, 20 0

/

15000 10000 5 0 0

/

/

/

/

/

.~~~~

IA I IT I I f 111ll§1 REPLACE WITH INSERT "B"

_=

7 I

A r-I l

I II I.~i I I

~~~/

I _I

{l(111

)

A i i iT IIII4In IIJIIIIl f1111 _

_S I§fv

~

lIIIIv I I lIIII T

A I

I~--

ACAPTABLE I

VA-11 Z _Ai I

I I I 1

E

,I

)

I IA--_

EK I -1 1:1:

1:

I:- Ufl j

,n

/

/

I,

/

1.5 4V.U

n. ? 4 U 2

-E c

I'p.

i naI w

/.u P

,Inii U-235 Enrichme&i (nominal w/o) f f -

j1-f I

"-'5j U.u Figure 3.7.18-2 Vogtle Units 1 and 2 Vogtle Unit 2 Burnup Credit Requirements for All Cell Storage 3.7.18-4 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

INSERT "B" FIGURE 3.7.18-2 AC e m 4*J,wW I

II III I~ I I~1111111 I

rMiac I1 iii ii 1111II:1II II 1111.............

I I000I1 I I I I.I I

I Sr 30,000 I

c. 25,000 E

E 20,000 0

Ui.

15,000 I

I I

I I

I I

I III I

I I

I I

I I

I I

I I

I I I I

I I

I I

Ir I I

I I

I I

I I

E 11........

...J -

- - - - - - - -1 I I 1 I1 I II1I 11I1I I11 1I 1 1 1 I1 1 V I...........................

.. I 1.......... 1 § I I 1 1.1.1. 1.

ABLE 111111111 11

!!! M M I[

Iiiiii I I Ii :1 I

I 10,000 1 11 1 111111 1 A Li I I I I I I I I ill 1

1 11 11 1 1 1 5,OD I I 1 I

OfII I

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial 236U Enrichment (norirnal wfo)

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality (Unit 1) 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

I

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

Keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described In Section 4.3 of the FSAR.

C.

Keff < 0.95 when fully flooded with water borated tom/ ppm, -

which Includes an allowance for uncertainties as described in Section 4.3 of the FSAR; I

l INSERT A'A Vs

d.

New or partially spent fuel assemblies with a combination of bumup and initial nominal enrichment in the maccer table bumu doain ofFigures 3.7.18-1 odhaving aA maximum rofcrcn cc fuool zicccm bly K le thaItt eqen t6or ctIt

1. 31 i, 6 may be aiiowea unrestnctea storage In the unit 1 tuel storage pool.

e.

New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 1 fuel storage pool In a 3-out-of-4 checkerboard storage configuration as shown In Figure 4.3.1 Interfaces between storage configurations in the Unit 1 uel INSERT "B' storage pool shall be in compliance with Figure 4.3.1 "An

\\

assemblies are new or parbially spent fuel assemblies with a

\\ combination of bumup and initial nominal enrichment in the akaccepale bumu domains of Figure 3.7.18-1, or whichfibayce a l ~

Fexmu eofrcee fuel assembly Kit les-s thanRorequal to 1.4181 at l"i assemblies are assemables with initial enrichments up to a maximum of 5.0 weight percent U-235.

(continued)

Vogtle Units 1 and 2 4.0-2 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

f.

A nominal 10.25 inch center to center pitch in the Unit 1 high density fuel storage racks.

I (Unit 2) 4.3.1.2 The spent fuel storage racks are designed and shall be maintained with:

I

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

Kff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR.

c.

Keff < 0.95 when fully flooded with water borated to

ppm, which includes an allowance for uncertainties as described in I

Section 4.3 of the FSAR; ElI

d.

New or partially spent fuel assemblies with a combination of bumup and initial nominal enrichment in the "acceptable burnup domain" of Figure 3.7.18-2 may be allowed unrestricted storage in the Unit 2 fuel storage pool.

e.

New or partially spen uel assemblies with a combination of bumup and initial nomi al enrichment in the "acceptable burnup domain" of Figure 4.3.1 may be stored in the Unit 2 fuel storage pool in a 3-out-of-4 checkerboard storage configuration as shown in Figure 4.3.1-i.

I1 I, decay time, New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 2 fuel storage pool in a 2-out-of-4 checkerboard storage nfiguration as shown in Figure 4.3.1 E.

Ne r partially spent fuel assemblies with a combination of burnu and initial nominal enrichment in the "acceptable bumup domain" of Figure 4.3.1 may be stored in the Unit 2 fuel (continued) 10 Vogtle Units 1 and 2 4.0-3 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (c I INSERT "C" ontinued) storage pool as "low enrichment" fuel assemblies in the 3x3 /

checkerboard storage configuration as shown in Figure 4.3.1-.

New or partially spent fuel assemblies with initial nominal enrichments less than or equal to 3.20 weiQht percent U-235 or

,m Fee

^ fue asebl 6lss than or l

equa to1.1Q a 68Flmay be ~stored -in the Unit 2 fuel storage/'

pool as "high enrichment" fuel assemblies in the 3x3 checkerboard storage configuration as shown in Figure 4.3.1 Er Interfaces betwee storage configurations in the Unit 2 fug storage pool shall b in compliance with Figures 4.3.1 4.3 and 4.3.1a. "A" assemblies are new or partially l speel assemblies with a combination of burnup and initia nominal enrichment in the "acceptable bumup domain" of Figure 3.7.18-2. "B" assemblies are new or partially spent fuel assemblies with a combination of bumup and initial nominal

' X\\enrichment in the "acceptable bumup domain" of Figure 4.3.1-

0. "C" assemblies are assemblies with initial enrichments up to a maximum of 5.0 weight percent U-235. "L" assemblies are new oipartially spent fuel assemblies with a combination of bumupland initial nominal enrichment in the "acceptable burnup domain" of Figure 4.3.1A. "H" assemblies are new or partially I

onspent fuel assemblies/ith initial nominal enrichments less than I, decay time,

f.

A nominal 10.58-inch center to center pitch in the north-south direction and a nominal 10.4-inch center to center pitch in the east-west direction in the Unit 2 high density fuel storage racks.

I (continued)

Vogtle Units I and 2 4.0-4 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

INSERTS FOR TS 4.3.1 INSERT A

... satisfying a minimum Integral Fuel Burnable Absorber (IFBA) requirement as shown in Figure 4.3.1-7...

INSERT B

... satisfy a minimum IFBA requirement as shown in Figure 4.3.1-7.

INSERT C

... which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher initial enrichments...

INSERT D

... which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher initial enrichments.

Design res 4.0 (This figure ha een deleted.

gu re 4.3.1-1 ogtle Unit 1 Buup Credit Requirents for 3-out-of Strage Vogtle Units and 2 4 -6 Amendment No. 9 (Unit 1)

AmendmentN 77 (Unit 2)

Design Features 4.0

.3.(

I M

I 11 IV I}

I I I I I

11 I

I 1

Ii^ I

/

/

~REPLACE WITH INSERT "A"

____7 25000 lN 1 lII I

1 1

1 Yi I I I Iy I 7_____

/

'll.

I Tl It: T-l I I0 2WO

_ I r11 l IJ 1 :1 IYI 1 y I01 I

i 000_ FHX ~H1 i IF-7

/

~~

ACCE3-

-llhI

0gE 7 _-

Gf~~~~~~~~~~

I

_III,

/11 IZI mf 500011,11111 011 E

___r

/.1 oo

,011 1/ 1 1 1 1 II Z 1 t1 IA 1111 T T 1 1 1 TZ I I I 1

} 7 A 1,f I7II

___r_

1.57 2. T~

3.01ll~

4.

.0Li

,,f

~

~

~

~

Iita _ncrnnp

_min lll1 11{

1 PAU Figure 4.3.1-0 Vogtle Unit 2 Bumup Credit Requirements for 3-out-of-4 M L -J Storage Vogtle Units 1 and 2 4.0-7 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

INSERT "A" FIGURE 4.3.1-8 25,000 20,000 i.

t 15,000 E

.0D g 10,000 5L 5,000 ACEPTASL NAICCEPTAE Y

I I I I

/

II111I I I..

I 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial 235U Enrichment (norrinal wfo)

Design Features 4.0 Jr ~

~

I I I I I I I I VI II I I I I T/ I I I I I I fTr 11I TfT II An rI I

Xl Jr

~

~~~~~~~~~~~~~~

Y I

-A IIV 2

f S-

_ttI REILACE 1YITH INSR NYI_

)

f-00 If

/

IIIII1111 II R I L

r 9 ^^2 __ _I II f I I

J 1 t 1

__A g~~~~~~~~~~~~~~~~~0 orr_.___flf IlIl11111tHd F

/

_ _Z z

flt I

I It1 AI I

I A

f 0

/

~I If If I t 1

0 111 lll f

J

= -,

1} I I IA I

I 1 II I1 111 f

IIY 1

I I III_

_I_

J l §~~~I 1X1 1t-l lK I _

JII I

if

/1-I I I I I IA I

I T

1.

3_

3t 4

~

v 11 111 lr

.0 4.5 5.

f ~

~

~

~

~~~n,10000m 1111na 11/l l

w/o)U ItI

/ :frv ^

111A1 7 11111 Figure 4.3.1 41 Vogtle Unit 2 Bumup Credit Requirements foj 3x3 Storage I Peripheral Assemblies for 1

Vogtle Units 1 and 2 4.0-8 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

INSERT "B" FIGURE 4.3.1-10 55,000.................................

1-1.1 50,000 45,000 111111111 III I II 11111111 III 111 TM 40,000 II 0.

E

.0 E

U.

35,000 30,000 25,000 20,000 01O ACCEPTABLE if I

Y1 A

I A.

09 1

100 O11 XO oO'YhO'l UNACC PTABLE MY lo 4%09 91 VI I Decay Time

-0 years 5 years 10years

-15 years

-2 years 15,000 10,000 5,000 0

T 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial 236U Enrichment (nominal w/o)

Design Features 4.0 3-out-of4 Checkerboard Storage (Units 1 and 2) 2-out-of-4 Checkerboard Storage (Unit 2)

L Empty Storage Cell 0

Fuel Assembly in Storage Cell Figure 4.3.1g

-Vogtle Units 1 and 2 Empty Cell Checkerboard Storage Configurations Vogtle Units 1 and 2 4.0-9 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

Design Features 4.0 3x3 Checkerboard Storage Low Enrichment Fuel Assembly in Storage Cell [m] High Enrichment Fuel Assembly in Storage Cell Fi Figure 4.3.1 Vogtle Unit 2 3x3 Checkerboard Storage Configuration Vogtle Units 1 and 2 4.0-10 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

Design Features 4.0 Interface A

A A

A A

A A

A A

A A

A A

A A

A A

A Empty B

Empty A

A A

B I B B

A I A A I Empty B

Empty A

A l A Note:

A-AIlCell Enrichment B Out-Of-4 Enrichment Empty - Empty Cell Boundary Between All Cell Storage and 3-out-of4 Storage (Units 1 and 2)

I A

A A

A A

A A

A A

A A

A 4

I 4

.I Interface A

A A

A A

A Note:

A - All Cel Enrichment B -

3-Out-OfW4 Enrichment C Out-Of-4 Enrichment Empty - Empty Cell

-r -

w e-a mu -

I Empty B

Empty A

A A

C Empty B

A A

A Empty C

EmptyL A

jA A I

Boundary Between AB Cell Storage and 2-out-of4 Storage (Unit 2)

Note:

1. A row of empty cells can be used at the Interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty celL Figure 4.3.1-0 Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage)

El Vogtle Units 1 and 2 4.0-11 Amendment No. 102 (Unit 1)

Amendment No. 80 (Unit 2)

Design Features 4.0 Interface B

Empty B

Empty B

Empty B

B I B B

B B

B Empty B

Empty B

Empty Empty C

Empty B

B B

C Empty C

Empty B

Empty Empty C

Empty B

B B

mm Note:

B Out-OfM4 Enrichment C Out-Of-4 Enrichment Empty - Empty Cefl Boundary Between 2-out-of-4 Storage and 3-out-of-4 Storage Empty B

Empty B

B B

B B

B B

Empty B

Interface Note:

B Out-Of-4 Enrichment C Out-Of-4 Enrichment Empty - Empty Cell Empty B

Empty B

B B

b mU mUm

 m

 mum c

Empty C

Empty B Empty Empty C Empty B

B B

C Empty C

Empby B

Empt I

Boundary Between 2-out-of-4 Storage and 3-out-of-4 Storage Note:

1. A row of empty cells can be used at the Interface to separate the configurations.

2. It is acceptable to replace an assembly with an empty cell.

Figure 4.3.1-7 4}

Vogtle Unit 2 Interface Requirements (Checkerboard Storage Interface)

Vogtle Units 1 and 2 4.0-12 Amendment No.

Amendment No.

102 (Unit 1) 80 (Unit 2)

Design Features 4.0 Interface A

A A

A A

A A

A A

A A

A L

L Ll L

A A

L 1L L

L A

A L

Hl L

L A

A L

L L

L A

A I

Note:

A - Al Cell Enrichment L = Low Enrichment of 3x3 Checkerboard H - High Enrichment of 3x3 Checkerboard Note:

1. A row of empty cells can be used at the Interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty cell.

Figure 4.3.1 Vogtle Unit 2 Interface Requirements (3x3 Checkerboard to All Cell Storage)

Vogtle Units 1 and 2 4.0-13 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

Design Features 4.0 B

B B

B B

B*

Empty B

Empty B

Empty B

Interface L

L L

L B

B Note:

B Ot-Of4 Enrichment L -Low Enrichment of 3x3 Storage H - High Enrichment of 3x3 Storage Empty - Empty Cell

_in wi A--

I L

L L

EmptyI B L

H L IL B

B L

L L

L Empty B

I Boundary Between 3x3 Storage and 3-out-of-4 Storage Interface

• 41 C

Empty C

Empty C

Empty Empty B

Empty B

Empty C

L _L**

L L**

B Empty L

L L

L Empty C

L H

L l7*

B Empty L

L L

L EmptyC I

Note:

B Out-O4 Enrichment L - Low Enrichment of 3x3 Storage H1-High Enrichment of 3x3 Storage C Out-Of4 Enrichment Empty - Empty Cell Boundary Between 3x3 Storage and 2-out-of-4 Storage Note:

1. A row of empty cells can be used at the interface to separate the configurations

2. It Is acceptable to replace an assembly with an empty cell.

3. For the 3-out-of-4 configuration, the row beyond the Low enrichment can swap empty and and B assemblies, however the next outer row must change the Indicated assembly (*) to an empty cell.

4. For the 2-out-of-4 configuration, the row beyond the Low enrichment can swap empty and B assemblies, however the next outer row of empty and C assemblies must also swap locations.

5. If empty cells are In Indicated locations (**), then the face adjacent B assemblies can be C assemblies.

Figure 4.3.1; Vogtle Unit 2 Interface Requirements (3x3 to Empty Cell I

I t

Checkerboard Storage)

Vogtle Units 1 and 2 4.0-14 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

NEW FIGURE ADDED 140 120 100 E

i) 80 60

-ACCEPTABLEF-_

_____-=_

L--

.~~~~~

UN

__gcf ACCETABLE IFBA

~ 1.0X

-1.5X

-~2.OX 40 20 0

3.0 3.5 4.0 4.5 5.0 Initial 2 5U Enrichment (nominal w/o)

Figure 4.3. 1-7 Vogtle Unit I IFBA Credit Requirements for All Cell Storage

NEW FIGURE ADDED 100 90 80 70 E

00 0I U-E z

60 50 40 30 -

20 -

10 -_

0-3.0 Figure 4.3.1-9

-<IFBA

- iox]

- -tj j]

\\_-_-2-7--

-2c

_UNACCEPTABLE

==

=Z__

V 3.5 4.0 4.5 5.0 Initial Enrichment [nominal wlo]

Vogtle Unit 2 IFBA Credit Requirements for Center Assembly for 3x3 Storage

Fuel Storage Pool Boron Concentration B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Fuel Storage Pool Boron Concentration BASES BACKGROL)ND Fuel assemblies are stored in high density racks. The Unit 1 spent fuel storage racks contain storage locations for 1476 fuel assemblies, and the Unit 2 spent fuel storage racks contain storage locations for 2098 fuel assemblies. The Unit 1 racks use boral as a neutron absorber in a flux trap design. The Unit 2 racks contain Boraflex, however, no credit Is taken for Boraflex.

Westinghouse 17x17 fuel assemblies with initial enrichments of up to and including 5.0 weight percent U-235 can be stored in any location in the Unit 1 or Unit 2 fuel storage pool provided the fuel burnup-enrichment combinations are within the limits that are specified in Figures 3.7.18-1 (Unit 1) or 3.7.18-2 (Unit 2) of the Technical Specifications. Fuel assemblies that do not meet the

-enrichment combination of Figures 3.7.18-1 or 3.7.18-2 may bed in the storagepools of Units 1 or 2 in accordance ith checkerbo orage configurations described in Figures

4.

iho ouse Shnt Fuel Ra4.k I

Or;t;^-lihz Illntl^^etel iAn~.,ri,..p4 ivMIAU'AD-4AA49Z-MDR QDip I I INSERT -

at_ r bIItlIIIY IVIUttlUUUIUUV.

UlGlSIIUUd 111 VbfOr 1'1-1 MU lsr as.. rev.

-1111--

Al

_O I-I-y*"-

-t"I 'v,-1 1r -- Inn itri -

'-yin.1 iln.rma~

s xr n

_B_

l_

r_

_wT

_rv bencthmarking, spent fue rask criticalit' calculations methodeolog, reactivity equivalencing methodology, accident methodology, and soluble boron credit mothodolofKG'.

The Westinghouse Spent Fuel Rack Criticality Methodology ensures that the multiplication factor, Keg, of the fuel and spent fuel storage racks is less than or equal to 0.95 as recommended by ANSI 57.2-1983 (Reference 3) and NRC guidance (References 1, 2 and 6). The codes, methods, and techniques contained in the methodology are used to satisfy this criterion on Kff.

The methodology of the NITAIL II, XSDRNPM S, and KENO Va codes is used to establish the bias and bias uncertainty. PHOENIX P. a nuclear design code used primarily for core reactor physics calculations is used to simulate spent fuel storage rack geometries.

I INSERT -

(continued)

Vogtle Units 1 and 2 B 3.7.17-1 Rev. 2-7198

Fuel Storage Pool Boron Concentration B 3.7.17 BASES BACKGROUND (continued)

Reference 4 describes how credit for fuel storage pool soluble boron is used under normal storage configuration conditions. The storage configuration is defined using Kff calculations to ensure that the Kff will be less than 1.0 with no soluble boron under normal storage conditions including oerances 51 and uncerlai nxlU.

oluble boron credit is then used to aintain K less than or equal to

95. IThe Unit 1 pool requires W ppm and the Unit 2 poo` requires ppm to maintain Kef less than or equal to 0.95 for all allowed combinations of storage configurations, enrichments, and burnups.L The analyses assumed 19.9% of the boron atoms nave atomic weight 10 (B-10). IThe effects of B 40 depletion on theo boron conentration for maintaining K. -t ! 09 aFe neglgible. The treatment of reactivity equivalncing uncortainties, a; well as the calculation of postulated accidents crediting coluble boron is described in WCAP 14416 NP A, Rev. 1.

I INSERT M(

This methodology was used to evaluate the storage of fuel with initial enrichments up to and Including 5.0 weight percent U-235 in the Vogtle fuel storage pools. The resulting enrichment, and burnup limits for the Unit 1 and Unit 2 pools, respectively, are shown in Figures 3.7.18-1 and 3.7.18-2. Checkerboard storage configurations are defined to allow storage of fuel that is not within the acceptable burnup domain of Figures 3.7.18-1 and 3.7.18-2.

These storage requirements are shown in Figures 4.3.1 through 4.3a-.

A boron concentration of 2000 ppm assures that no BLJdibble dilution event will result in a Kff of > 0.95.

APPLICABLE SAFETY ANALYSES I MARSt ;uor ASAMARio MOol anrnIM nt AMMAIttiont Mir not roRuIt1.

lR an incrffaS9 in.K'.M xamples of such accidents are the drop of-a fuel aceombly on top Of a rcFk, and the drop of a fuel asoembly betwee;n rack modulo, Or between Prak ofnduloc and the pool wall.

From a criticality standpoint, a dropped assembly accident occurs when a fuel assembly in its most reactive condition is dropped onto the storage racks. The rack structure from a criticality standpoint is not excessively deformed. Previous accident analysis with unborated water showed that the dropped assembly which comes to rest horizontally on top of the rack has sufficient water separating it from the (continued)

Vogtle Units 1 and 2 B 3.7.17-2 Rev. 2-7/98

Fuel Storage Pool Boron Concentration B 3.7.17 BASES APPLICABLE SAFETY ANALYSES (continued)

I net-"

P J-I nErt3F active fuel height of stored assemblies to preclude neutronic interaction. For the borated water condition, the interaction is even less since the water contains boron, an additional thermal neutron absorber.

HOLeu eor therso acidente ean be poctulated foFr each etriago configuration which could incrra6c reactivity beyond the analyzed conditiln. The WMret poctulated accident would be a ehange in pool temperature to outside the range of temperatures assumed in the criticality analyses (60F to 1 850F). The serond accident would be dropping a fuel assembly into aR already loaded cell. The irW would be the misloading of a fuel assembly into a cel! for which the ecstr~irtn oR location, enRichmeRt, or burnup are Rot caticfied.

An increase in the temperature of the water passing through the stored fuel assemblies causes a decrease in water density which results in an addition of negative reactivity for flux trap design racks such as the Unit 1 racks. However, since Boraflex is not considered to be present for the Unit 2 racks and the fuel storage pool water has a high concentration of boron, a density decrease causes a positive reactivity addition.theactivitk alects of a temherature range from 320 F to 2400 F were evaluat The increase in reactivity due to the-increase in temperature is bounded by the misload accident, for the

_l Unit 2 racks4 The increase in reactivity due to the decrease int temperature below 50° F is bounded by the misplacement of a fuel assembly between the rack and pool walls for the Unit 1 racks...,

For the accideRt ef dropping a fuel assembly irte an alr!Feady leaded cell, the upward axial leakage of that cell will be reduced, hoWeF or, the ^everall efect oR the rack reactivity will be insignificant. Thic is because the total axial leakage in both the upWard and downwrr d diFeroctic for the eRtire fuel arrvay i Worth about 0.003 Ak. Thus, minimizing the upward only leakage of just a in;gle cell w 11 net cauoe aRy significaRt ncreae in r6_eactivity.

Furthermore, the neutronic coupling between the dropped ascembly and the already leaded assombly will be lew due to several inches of assembly nozzle ctrueture which would separate the active fuel Fegienc. Therefore, this accident weuld be bounded by the micload accident.

(continued)

Vogtle Units 1 and 2 B 3.7.17-3 Rev. 2-7/98

Fuel Storage Pool Boron Concentration B 3.7.17 BASES APPLICABLE SAFETY ANALYSES (continued)

Tho fuol assembly misloading accidont inRolvoe placomont of a fuel assombly iR a lroca!o for WhIch it dooe Rot meot tho roguirom^onts for onrichmont or burnuP, including tho placomont of an assombly in a location that ic roquired to be loft ompty. Tho rosult of tho misloading is to add positivo roactity, incroasing.., toward 0.05. A fourth acidont was evaluated for tho n I 1 fo Btorage raolcc coRtaining boral. Tho fourth accident wa:

the mipSfacomont of a fuol assembly botwoon tho rFack and pool wall. This was tho limiting accidont for tho Unit 1 racss. Tho (continued)

Vogtle Units 1 and 2 B 3.7.17-4 Rev. 2-7/98

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

APPLICABLE mximum required additional boron to compensate for this SAFETY ANALYSES event Qs 1260 ppm for Unit 2, and 800 Ppm for Unit 1 which (continued) i; well bolow the limit of 2000 ppm.

JNERG The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO The fuel storage pool boron concentration is required to be

> 2000 ppm. The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in reference 5. The amount of soluble boron required to offset each of the above postulated accidents was evaluated for all of the proposed storage configurations. That evaluation established the amount of soluble boron necessary to ensure that Kgff will be maintained less than or equal to 0.95 should pool temperature exceed the assumed range or a fuel assembly misload occur. The amount of soluble boron necessary to mitigate these events was 851 pmfor Unit 1 determined to be 11 60 PPFm 9fo Unit 2 and 800 viam forUni 11 The and 1098 ppm for specified minimum boron concentration of 2000 ppm assures that Unit 19 the concentration will remain above these values. In addition, the boron concentration is consistent with the boron dilution evaluation that demonstrated that any credible dilution event could be terminated prior to reaching the boron concentration for a Kff of

> 0.95. These values are ppm for Unit l and ppm for Unit 2.

APPLICABILITY This LCO applies whenever fuel assemblies 'are stored in the spent fuel storage pool.

ACTIONS A.1. A.2.1. and A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident In progress. This is most (continued)

I Vogtle Units 1 and 2 B 3.7.17 Rev. 3-1 0/01

INSERTS TO B 3.7.17 INSERT A The acceptable fuel assembly storage configurations are based on NRC-approved acceptance criteria for crediting soluble boron as described in the NRC's safety evaluation report in WCAP-14416-P-A (Reference 4).

INSERT B The analysis methodology employs: (1) SCALE-PC, a personal computer version of the SCALE-4.3 code system, with the updated SCALE4.3 version of the 44 group ENDF/B-V neutron cross section library, and (2) the two-dimensional integral transport code DIT with an ENDF/B-VI neutron cross section library.

SCALE-PC was used for calculations involving infinite arrays for the "2-out-of-4", "3-out-of-4", "All-Cell", and "3x3" fuel assembly storage configurations. In addition, it was employed in a full pool representation of the storage racks to evaluate soluble boron worth and postulated accidents.

SCALE-PC, used in both the benchmarking and the fuel assembly storage configurations, includes the control module CSAS25 and the following functional modules: BONAMI, NITAWL-II, and KENO V.a.

The DIT code is used for simulation of in-reactor fuel assembly depletion. KENO V.a was used in the calculation of biases and uncertainties.

INSERT C However, to account for the effects of variations in the natural abundance of B-1O, the calculated boron concentrations, as well as the concentrations for accidents, were adjusted to correspond to a B-1I fraction of 19.7%.

INSERT D The soluble boron concentration, in units of ppm, required to maintain Ktff less than or equal to 0.95 under accident conditions is determined by first surveying all possible events which increase the Kff value of the spent fuel pool. The accident event which produced the largest increase in spent fuel pool K~ff value is employed to determine the required soluble boron concentration necessary to mitigate this and all less severe accident events. The list of accident cases considered includes:

• Dropped fresh fuel assembly on top of the storage racks,

• Misloaded fresh fuel assembly into an incorrect storage rack location,

• Misloaded fresh fuel assembly outside of the storage racks, I of 2

• Reduction in rack module-to-module water gap due to seismic event,

• Spent fuel pool temperature outside the normal range of 50 OF to 185 OF.

INSERT E Several fuel mishandling events were simulated with KENO V.a to assess the possible increase in the Keff value of the spent fuel pools. The fuel mishandling events all assumed that a fresh Westinghouse OFA fuel assembly enriched to 5.0 W/O 235U (and no burnable poisons) was misloaded into the described area of the spent fuel pool. These cases were simulated with the KENO V.a model for the entire spent fuel pool.

For Unit 1, the fuel mishandling event which produced the largest increase in spent fuel pool KYff value is the misloading of a fresh fuel assembly between a "3-out-of4" fuel assembly storage configuration and the pool wall. The additional soluble boron concentration necessary to mitigate this and all less severe accident events is 340 ppm.

For Unit 2, the fuel mishandling event which produced the largest increase in spent fuel pool Keff value is the misloading of a fresh fuel assembly in an incorrect storage rack location for the "2-out-of-4" configuration. The additional soluble boron concentration necessary to mitigate this and all less severe accident events is 704 ppm.

For the accident due to a seismic event, the gap between rack modules was reduced to zero. For both Units I and 2, the reactivity increase is an order of magnitude less than that for the fuel mishandling events.

INSERT F This bounds the temperature range assumed in the criticality analyses (50 0F to 185 OF).

INSERT G Including the effects of accidents, the maximum required boron concentration to maintain Keff < 0.95 is 851 ppm for Unit 1 and 1098 ppm for Unit 2 which is well below the limit of 2000 ppm.

2 of 2

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Fuel Assembly Storage in the Fuel Storage Pool BASES BACKGROUND The Unit 1 spent fuel storage racks contain storage locations for 1476 fuel assemblies, and the Unit 2 spent fuel storage racks contain storage locations for 2098 fuel assemblies.

I Westinghouse 17X17 fuel assemblies with an enrichment of up to and including 5.0 weight percent U-235 can be stored in the I

acceptable storage configurations that are specified in Fiqures 3.7.18-1 (Unit 1) 3.7.18-2 (Unit 2) and 4.3.1 throu h

_1

. The acceptable fuel a43embly vtOrage locatiGne re based on the Westinghousc Spent Fuel Rack Criticality 3i~~-

Methodology, described in WCAP 14116 NP A, Rcv. 1 Ireferene 14. Additional background discussion can be found in B3.7.17.

-~

_3.556.56 Westinghouse 7x17 fuel assemblies ith nominal enrichments no greater thank w/o 235U may be ored in all storage cell locations of the Unit I pool. Fuel ass mblies with initial nominal enrichment greater than m w/o235U must satis a

minimum burnu requirement as shown in Figure 3.7.18-1.

INS conduton smb:les bavie t Korod i

.ll ate cold feactoF co conditions mnay also be stored in all ceils of the nt1fe 6toBFeS§ras II Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 5.0 w/o235U may be stored in a 3-out-of-4 checkerboard arrangement with empty cells in the Unit I pool.

There are no minimum burnup requirements for this configuration.

I I

I Move to ** on Page B3.7.18-3 Westinghouse 17x17 fuel assemblies with nominal enrichments no greater that 5.0 w/o235U may be stored in a 2-out-of-4 checkerboard arrangement with empty cells in the Unit 2 pool.

There are no minimum bumuD requirements for this configuration. 1 (continued)

Vogtle Units I and 2 B 3.7.18-1 Rev. 2-7/98

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18

[3~\\

BASES BACKGROUND (continued)

Westinghouse 7x17 fuel assemblies with nominal enrichments no greater than k w/o235U may be stored in all storage cell locations of the Unit 2 pool. Fuel assemblies with initial nominal enrichment greater than w/o235U must satisfy a minimum burnup requirement a shown in Figure 3.7.18-2.

(continued)

Vogtle Units 1 and 2 B 3.7.18-2 Rev. 2-7/98

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 BASES BACKGROUND (continued)

E-Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 2.40 w/o235U may be stored in a 3-out-of-4 checkerboard arrangement with empty cells in the Unit 2 pool.

Fuel assemblies with initial nominal enrichment greater than 2.40 w/o235U must satisfy a minimum burnup requirement as shown in Figure 1 IN Westinghouse a

7x17 fuel assembliesmatio stored in the Unit42 pool in a 3x3 array. The center assembl nust have an initial enrichment no areater than 3.20 w/o 235ul Ateornatively, the eenteF Gf the 3x3 array may be leaded with any assembly which mooets a mfiaxfimu~m Infinite multiplication factor (K.) value ef 1.410 at 6°F.

ne mnethod of aehleyvin this value of Ki is-be the use of IFBAs.lThe surrounding fuel assemblies must have an initial nominal enrichment no greater than m

w/o235U or satisfy a minimum burnup requirement for highe nitial enrichments as shown In Figure

\\

and decay time APPLICABLE SAFETY ANALYSIS Most fuel storage poolaccident conditions will not result In an Increase in K^...I xmp of such; accidents arc the drop of a fuel assembly en to ofarakad the drop of a fuel assembly between Fack( modules er. bet-WeeR rack moedules and the poolt w However, accidents can be postulated for each storage configuration which could Increase reactivity beyond the analyzed condition. A discussion of these accidents is contained in B 3.7.17.

The configuration of fuel assemblies In the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

I LCO The restrictions on the placement of fuel assemblies within the fuel storage pool ensure the Kff of the fuel storage pool will always remain < 0.95, assuming the pool to be flooded with borated water.

The combination of initial enrichment and burnup are specified In Figures 3.7.18-1 and 3.7.18-2 for all cell storage in the Unit 1 and Unit 2 pools, respectively. Other acceptable enrichment burnup l enrichment - IFBA l and checkerboard combinations are described In Figure 4.3.1-1 through 4.3.1 (continued)

Vogtle Units 1 and 2 B 3.7.18-3 Rev. 3-1 0/01

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 BASES (continued)

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored In the fuel storage pool is not in accordance with the acceptable combination of initial enrichment, burnup, and storage configurations, the immediate action Is to Initiate action to make the necessary fuel assembly movement(s) to bring the configuration Into compliance with Figures 3.7.18-1 (Unit 1),

3.7.18-2 (Unit 2), or Specification 4.3.1.111 (Unit 1) or 4.3.1.2 (Unit 2).

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action Is independent of reactor operation. Therefore inability to move fuel assemblies Is not sufficient reason to require a reactor shutdown.

I I NOTE DELETION [

SURVEILLANCE REQUIREMENTS SR 3.7.18.1 This SR verifies by administrative means that the Initial enrichment and burnup of the fuel assembly is within the acceptable burnup domain of Figures 3.7.18-1 (Unit 1) or 3.7.18-2 (Unit 2). For fuel assemblies In the unacceptable range of Figures 3.7.18-1 and 3.7.18-2, performance of this SR will also ensure compliance with Specification 4.3.1.1 (Unit 1) or 4.3.1.2 (Unit 2).

I Fuel assembly movement will be In accordance with preapproved plans that are consistent with the specified fuel enrichment, burnup, and storage configurations. These plans are administratively verified prior to fuel movement. Each assembly Is verified by visual Inspection to be in accordance with the preapproved plan prior to storage In the fuel storage pool.

Storage commences following unlatching of the fuel assembly in the fuel storage pool.

(continued)

Vogtle Units 1 and 2 B 3.7.18-4 Rev. 2-7/98

INSERTS TO B 3.7.18 INSERT A The acceptable fuel assembly storage configurations are based on NRC-approved acceptance criteria for crediting soluble boron as described in the NRC's safety evaluation report in WCAP-14416-P-A (Reference 1).

INSERT B

... or a minimum Integral Fuel Burnable Absorber (IFBA) requirement as shown in Figure 4.3.1-7.

INSERT C

... or satisfy a minimum IFBA requirement for higher initial enrichments as shown in Figure 4.3.1-9.

Vogtle Electric Generating Plant Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2 TYPED REVISED TECHNICAL SPECIFICATION AND BASES PAGES

TABLE OF CONTENTS LIST OF FIGURES 2.1.1-1 Reactor Core Safety Limits.................................................. 2.0-2 3.4.16-1 Reactor Coolant Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power with the Reactor Coolant Specific Activity> 1 pCi/gram Dose Equivalent 1-131

.3.4.16-4 3.7.18-1 Vogtle Unit 1 Burnup Credit Requirements for All Cell Storage 3.7.18-3 3.7.18-2 Vogtle Unit 2 Bumup Credit Requirements for All Cell Storage 3.7.18-4 4.3.1-1 Vogtle Units I and 2 Empty Cell Checkerboard Storage Configurations 4.0-6 4.3.1-2 Vogtle Unit 2 3x3 Checkerboard Storage Configuration.

.4.0-7 4.3.1-3 Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage)

.4.0-8 4.3.1-4 Vogtle Unit 2 Interface Requirements (Checkerboard Storage Interface)..

4.0-9 4.3.1-5 Vogtle Unit 2 Interface Requirements (3x3 Checkerboard to All Cell Storage)

.4.0-10 4.3.1-6 Vogtle Unit 2 Interface Requirements (3x3 to Empty Cell Checkerboard Storage)

.4.0-11 4.3.1-7 Vogtle Unit 1 IFBA Credit Requirements for All Cell Storage 4.0-12 4.3.1-8 Vogtle Unit 2 Burnup Credit Requirements for 3-out-of-4 Storage

,4.0-13 4.3.1-9 Vogtle Unit 2 IFBA Credit Requirements for Center Assembly for 3x3 Storage.................

4.0-14 4.3.1-10 Vogtle Unit 2 Burnup Credit Requirements for Peripheral Assemblies for 3x3 Storage...

4.0-15 5.5.6-1 Schedule of Lift-Off Testing for Two Containments at a Site..............

5.5-23 Vogtle Units 1 and 2 vi Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Fuel Assembly Storage In the Fuel Storage Pool 3.7.18 9,000 8,000 7,000 6,000 5,000 4,000 c

E.

E

.0IE V

a0 ox I,

3,000 2,000 1,000 0

3.0 3.5 4.0 4.5 Initial 235U Enrichment (nominal w/o) 5.0 Figure 3.7.18-1 Vogtle Units 1 and 2 Vogtle Unit 1 Burnup Credit Requirements for All Cell Storage 3.7.18-3 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Fuel Assembly Storage in the Fuel Storage Pool 3.7.18 45,000 1 1 1 1 11 1 1 1 11 1 11 1 1 1 1 1 1 40,000 1 1 1 1 11 1 1 1 1 1 1 1 1 1 1 111 1 1 1 I T T T I I TTT1 I 1

T I I 3 5, 0 0 11 I

Vl I I 30,000-

-ACCEPTABL 25,000 6

E 2

Z.

.C

• ina 12 20,000 1 _ _ _ _

I I

I I I TFF I I F l I I I I I

,ooo _

1111111111I _Z

• EE 10,000 5,000

,.1 I H III I

I I I IIIII I I I

LA I

II I I I I I I I I I I I I X~~lAIIITI.

111111 o I 1 / 1 ! 1 1 I 1 1 1 111 1

1 11 1

11 1 1 I

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial235U Enrichment (nominal wlo)

Figure 3.7.18-2 Vogtle Unit 2 Burnup Credit Requirements for All Cell Storage Vogtle Units 1 and 2 3.7.18-4 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticalitv (Unit 1) 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

Kiff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described In Section 4.3 of the FSAR.

c.

Kit < 0.95 when fully flooded with water borated to 511 ppm, which Includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d.

New or partially spent fuel assemblies with a combination of bumup and Initial nominal enrichment in the acceptable bumup domains of Figures 3.7.18-1 or satisfying a minimum Integral Fuel Burnable Absorber (IFBA) requirement as shown In Figure 4.3.1-7 may be allowed unrestricted storage in the Unit 1 fuel storage pool.

e.

New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored In the Unit 1 fuel storage pool in a 3-out-of-4 checkerboard storage configuration as shown In Figure 4.3.1-1.

Interfaces between storage configurations In the Unit 1 fuel storage pool shall be In compliance with Figure 4.3.1-3. OA" assemblies are new or partially spent fuel assemblies with a combination of burnup and Initial nominal enrichment in the "acceptable bumup domaln" of Figure 3.7.18-1, or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-7. ABE assemblies are assemblies with Initial enrichments up to a maximum of 5.0 weight percent U-235.

(continued)

Vogtle Units 1 and 2 4.0-2 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

f.

A nominal 10.25 inch center to center pitch in the Unit 1 high density fuel storage racks.

(Unit 2) 4.3.1.2 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

Kff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR.

c.

Kff

• 0.95 when fully flooded with water borated to 394 ppm, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d.

New or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the "acceptable burnup domains of Figure 3.7.18-2 may be allowed unrestricted storage in the Unit 2 fuel storage pool.

e.

New or partially spent fuel assemblies with a combination of bumup and initial nominal enrichment in the "acceptable bumup domain" of Figure 4.3.1-8 may be stored in the Unit 2 fuel storage pool In a 3-out-of-4 checkerboard storage configuration as shown In Figure 4.3.1-1.

New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 2 fuel storage pool in a 2-out-of4 checkerboard storage configuration as shown In Figure 4.3.1-1.

New or partially spent fuel assemblies with a combination of bumup, decay time, and Initial nominal enrichment In the acceptable bumup domains of Figure 4.3.1-10 may be stored (continued)

Vogtle Units 1 and 2 4.0-3 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) in the Unit 2 fuel storage pool as glow enrichment" fuel assemblies in the 3x3 checkerboard storage configuration as shown in Figure 4.3.1-2. New or partially spent fuel assemblies with initial nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher initial enrichments may be stored in the Unit 2 fuel storage pool as "high enrichment" fuel assemblies in the 3x3 checkerboard storage configuration as shown in Figure 4.3.1-2.

Interfaces between storage configurations in the Unit 2 fuel storage pool shall be in compliance with Figures 4.3.1-3,4.3.1-4, 4.3.1-5, and 4.3.1-6. 'A' assemblies are new or partially spent fuel assemblies with a combination of bumup and initial nominal enrichment in the "acceptable bumup domain" of Figure 3.7.18-

2. "B' assemblies are new or partially spent fuel assemblies with a combination of bumup and initial nominal enrichment in the acceptable bumup domains of Figure 4.3.1-8. 'C' assemblies are assemblies with initial enrichments up to a maximum of 5.0 weight percent U-235. "L" assemblies are new or partially spent fuel assemblies with a combination of bumup, decay time, and initial nominal enrichment in the "acceptable bumup domains of Figure 4.3.1-10. "H assemblies are new or partially spent fuel assemblies with initial nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher Initial enrichments.

f.

A nominal 10.58-inch center to center pitch in the north-south direction and a nominal 10.4-inch center to center pitch In the east-west direction in the Unit 2 high density fuel storage racks.

(continued)

Vogtle Units 1 and 2 4.0-4 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Design Features 4.0 0000 EJOLIJO 0000 LJ0EIJ0 0000 0EIJ0EIJ 001z11z1 DOEJO 3-out-of4 Checkerboard Storage (Units 1 and 2) 2-out-of4 Checkerboard Storage (Unit 2)

[

Empty Storage Cell 0

Fuel Assembly in Storage Cell Figure 4.3.1-1 Vogtle Units 1 and 2 Empty Cell Checkerboard Storage Configurations I

Vogtle Units 1 and 2 4.0-6 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 3x3 Checkerboard Storage Low Enrichment Fuel E

Assembly in Storage Cell Figure 4.3.1-2 Vogt Vogtle Units 1 and 2 EL High Enrichment Fuel Assembly in Storage Cell le Unit 2 3x3 Checkerboard Storage Configuration I

4.0-7 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 A

A A

A A

A A

A A

A A

A Interface A

A A

A A

A Empty B

Empty A

A A

B B

B A

A A

Empty B

Empty A

A A

Note:

A-AllCeD Enrichment B Out-Of-4 Enrichment Empty - Empty Cell I

Boundary Between All Cell Storage and 3-out-of-4 Storage (Units 1 and 2)

Interface A

A A

A A

A A

A A

A A

A A

A A

A A

A W Empty B

Empty A

A A

-_-A C

Empty B

A A

A Empty C

Empty A

A A

a Note:

A-ACell Enrichment B Out-Of-4 Enrichment C Out-Of-4 Enrichment Empty - Empty Cell Boundary Between All Cell Storage and 2-out-of-4 Storage (Unit 2)

Note:

1. A row of empty cells can be used at the Interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty cell.

Figure 4.3.1-3 Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage)

I Vogtle Units 1 and 2 4.0-8 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 B

Empty B

IEmpty B

Empty B

B B

B B

B Interface B

Empty B

Empty B3 Empty Empty C

Empty B

B B

C Empty C

Empty B

Empty EEmpty C

Empty B

B B

m Note:

B Ont-Of-4 Enrichment C Out-Of4 Enrichment Empty - Empty Cell I

Boundary Between 2-out-of-4 Storage and 3-out-of-4 Storage Interface Empty B

Empty B

B B

F-B B

B B

Empty B

3-Empty I B Empty.

B B

B C

Empty C

'Empty B

Empty Empty C

Empty B

l B B I C_

Empty C

Empty B

Empt I

Note:

B Out-Of-4 Enrichment C Out-Of-4 Enrichment Empty - Empty Cel Boundary Between 2-out-of4 Storage and 3-out-of4 Storage Note:

1. A row of empty cells can be used at the Interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty cell.

Figure 4.3.1-4 Vogtle Units 1 and 2 Vogtle Unit 2 Interface Requirements (Checkerboard Storage Interface) 4.0-9 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 A

A A

A A

A A

A A

A A

A

.~

Note:

A - All Cell Enrichment L -Low Enrichment of 3x3 Checkerboard H - High Enrichment of 3x3 Checkerboard Interface L

L L

L A

A Mm ---

w j

-I-'-

L L

L L

A A

L H

L I L A

A L

L.

L I L A

A I

Note:

1. A row of empty cells can be used at the interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty cell.

Figure 4.3.1-5 Vogtle Unit 2 Interface Requirements (3x3 Checkerboard to All Cell Storage)

I Vogbe Units 1 and 2 4.0-10 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 Interface B

B B

B B

B*

Empty B

Empty B

Empty B

L L

L L

B B

L L

L L

EmptyB L

H L

L B

B L

L L

L I EmptyB I

Note:

B Out-Of-4 Enrichment L - Low Enrichment of 3x3 Storage H - High Enrichment of 3x3 Storage Empty -Empty Cell Boundary Between 303 Storage and 3-out-of4 Storage Interface 411 C

EmptyC Empty C

Empty Empty B

Empty B

Empty C

L L*

L L**

B Empty L

L L

L Empty C

L H

1 L

L*

B Empty L

L L

L Empty C

I Note:

B Out-Of-4 Enrichment L -Low Enrichment of 3x3 Storage H -High Enrichment of 3x3 Storage C - 2l-ut-Of4 Enrichment Empty - Empty Cell Boundary Between 3x3 Storage and 2-out-of4 Storage Note:

1. A row of empty cells can be used at the interface to separate the configurations.

2. It Is acceptable to replace an assembly with an empty cell.

3. For the 3-out-of-4 configuration, the row beyond the Low enrichment can swap empty and and B assemblies, however the next outer row must change the Indicated assembly (*) to an empty cell.

4. For the 2-out-of-4 configuration, the row beyond the Low enrichment can swap empty and B assemblies, however the next outer row of empty and C assemblies must also swap locations.

5. If empty cells are In Indicated locations (**), then the face adjacent B assemblies can be C assemblies.

Figure 4.3.1-6 Vogtle Units 1 and 2 Vogge Unit 2 Interface Requirements (3x3 to Empty Cell Checkerboard Storage)

I 4.0-11 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 140 120 100 4b c0 80 60 ACEPTABLE-

--'UACCEPTALE_

7 IFBA 1.OX

-1.5X

2. OX 40 20 0

3.0 3.5 4.0 4.5 5.0 Initial 2M U Enrichment (nominal w/o)

Figure 4.3.1-7 Vogtle Units 1 and 2 Vogtle Unit 1 IFBA Credit Requirements for All Cell Storage 4.0-12 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 25,000-;t!

20,000-15,000 _ACC

_E ALI_

10,000-.

7-_

-IUNA CCEPTABLE CF_

0 ac n.E 4,S 2S U-

/

II-A l I

III V

_- It i

1 1 11 IT I Il 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 InitialI'U Enrichment (nominal wlo)

Figure 4.3.1-8 Vogtle Unit 2 Burnup Credit Requirements for 3-out-of-4 Storage Vogtle Units 1 and 2 4.0-13 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 100 90 80 70 2t

.0 0

E f

C E-Ez 60 50 40 ACCEPTABLE---

_Z ACEPTABLE IFBA

-1 OX I-2.0X 30 20 10 0

3.0 3.5 4.0 4.5 5.0 Initial Enrichment [nominal w/ol Figure 4.3.1-9 Vogtle Units I and 2 Vogtle Unit 2 IFBA Credit Requirements for Center Assembly for 3x3 Storage 4.0-14 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Design Features 4.0 55,000 50,000 45,000 40,000 E0 6) 6)

E)

U-35,000 30,000 25,000 20,000 I 1011 A v

10 I 1ACCEPTABLE II Zs I II

'00, OW I

IIA I

A I

I eZAI 1UNACCEPTABLE 07 I I Decay Time

-0 years

-5 years

-10 years

-15 years

-- 20 years 15,000 10,000 5,000 0

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial 235 U Enrichment (nominal wo)

Figure 4.3.1-10 Vogtle Unit 2 Bumup Credit Requirements for Peripheral Assemblies for 3x3 Storage Vogtle Units 1 and 2 4.0-15 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Fuel Storage Pool Boron Concentration B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Fuel Storage Pool Boron Concentration BASES BACKGROUND Fuel assemblies are stored in high density racks. The Unit 1 spent fuel storage racks contain storage locations for 1476 fuel assemblies, and the Unit 2 spent fuel storage racks contain storage locations for 2098 fuel assemblies. The Unit 1 racks use boral as a neutron absorber in a flux trap design. The Unit 2 racks contain Boraflex, however, no credit is taken for Boraflex.

Westinghouse 17x17 fuel assemblies with initial enrichments of up to and including 5.0 weight percent U-235 can be stored In any location in the Unit-1 or Unit 2 fuel storage pool provided the fuel bumup-enrichment combinations are within the limits that are specified in Figures 3.7.18-1 (Unit 1) or 3.7.18-2 (Unit 2) of the Technical Specifications. Fuel assemblies that do not meet the burnup-enrichment combination of Figures 3.7.18-1 or 3.7.18-2 may be stored in the storage pools of Units 1 or 2 in accordance with checkerboard storage configurations described in Figures 4.3.1-1 through 4.3.1-10. The acceptable fuel assembly storage configurations are based on NRC-approved acceptance criteria for crediting soluble boron as described in the NRC's safety

-evaluation report in WCAP-14416-P-A (Reference 4).

The Westinghouse Spent Fuel Rack Criticality Methodology ensures that the multiplication factor, Kff, of the fuel and spent fuel storage racks Is less than or equal to 0.95 as recommended by ANSI 57.2-1983 (Reference 3) and NRC guidance (References 1, 2 and 6). The codes, methods, and techniques contained in the methodology are used to satisfy this criterion on Kff.

The analysis methodology employs: (1) SCALE-PC, a personal computer version of the SCALE-4.3 code system, with the updated SCALE-4.3 version of the 44 group ENDF/B-V neutron cross section library, and (2) the two-dimensional integral transport code DIT with an ENDF/B-VI neutron cross section library.

SCALE-PC was used for calculations involving infinite arrays for the 2-out-of-4", "3-out-of-4", "AII-Ceir, and "3x3" fuel assembly storage configurations. In addition, it was employed in a full pool representation of the storage racks to evaluate soluble boron worth and postulated accidents.

SCALE-PC, used in both the benchmarking and the fuel assembly storage configurations, includes the control module CSAS25 and the following functional modules: BONAMI, NITAWL-I1, and KENO V.a.

(continued)

Vogtle Units 1 and 2 B 3.7.17-1

Fuel Storage Pool Boron Concentration B 3.7.17 BASES BACKGROUND

• (continued)

The DIT code is used for simulation of in-reactor fuel assembly depletion. KENO V.a was used in the calculation of biases and uncertainties.

Reference 4 describes how credit for fuel storage pool soluble boron is used under normal storage configuration conditions. The storage configuration is defined using Keff calculations to ensure that the K.,( will be less than 1.0 with no soluble boron under normal storage conditions including tolerances and uncertainties. Soluble boron credit is then used to maintain Keff less than or equal to 0.95. The analyses assumed 19.9% of the boron atoms have atomic weight 10 (B-10). However, to account for the effects of variations in the natural abundance of B-10, the calculated boron concentrations, as well as the concentrations for accidents, were adjusted to correspond to a B-10 fraction of 19.7%.

The Unit 1 pool requires 511 ppm and the Unit 2 pool requires 394 ppm to maintain Keff less than or equal to 0.95 for all allowed combinations of storage configurations, enrichments, and burnups.

This methodology was used to evaluate the storage of fuel with initial enrichments up to and including 5.0 weight percent U-235 in the Vogtle fuel storage pools. The resulting enrichment, and burnup limits for the Unit 1 and Unit 2 pools, respectively, are shown in Figures 3.7.18-1 and 3.7.18-2. Checkerboard storage configurations are defined to allow storage of fuel that is not within the acceptable burnup domain of Figures 3.7.18-1 and 3.7.18-2.

These storage requirements are shown in Figures 4.3.1-1 through 4.3.1-10. A boron concentration of 2000 ppm assures that no credible dilution event will result in a K0f of > 0.95.

APPLICABLE SAFETY ANALYSES The soluble boron concentration, in units of ppm, required to maintain KeH less than or equal to 0.95 under accident conditions is determined by first surveying all possible events which increase the Kff value of the spent fuel pool. The accident event which produced the largest increase in spent fuel pool Ktff value Is employed to determine the required soluble boron concentration necessary to mitigate this and all less severe accident events. The list of accident cases considered includes:

Dropped fresh fuel assembly on top of the storage racks, Misloaded fresh fuel assembly into an incorrect storage rack

location, Misloaded fresh fuel assembly outside of the storage racks, (continued)

Vogtle Units 1 and 2 B 3.7.17-2

Fuel Storage Pool Boron Concentration B 3.7.17 BASES APPLICABLE Reduction in rack module-to-module water gap due to seismic SAFETY ANALYSES

event, (continued)

Spent fuel pool temperature outside the normal range of 50 OF to 185 "F.

From a criticality standpoint, a dropped assembly accident occurs when a fuel assembly in Its most reactive condition is dropped onto the storage racks. The rack structure from a criticality standpoint Is not excessively deformed. Previous accident analysis with unborated water showed that the dropped assembly which comes to rest horizontally on top of the rack has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction. For the borated water condition, the interaction is even less since the water contains boron, an additional thermal neutron absorber.

Several fuel mishandling events were simulated with KENO V.a to assess the possible increase in the Keff value of the spent fuel pools.

The fuel mishandling events all assumed that a fresh Westinghouse OFA fuel assembly enriched to 5.0 W/o 235U (and no burnable poisons) was misloaded into the described area of the spent fuel pool. These cases were simulated with the KENO V.a model for the entire spent fuel pool.

For Unit 1, the fuel mishandling event which produced the largest increase in spent fuel pool Kff value is the misloading of a fresh fuel assembly between a "3-out-of-4" fuel assembly storage configuration and the pool wall. The additional soluble boron concentration necessary to mitigate this and all less severe accident events is 340 ppm.

For Unit 2, the fuel mishandling event which produced the largest increase in spent fuel pool Keff value Is the misloading of a fresh fuel assembly in an incorrect storage rack location for the M2-out-of-4" configuration. The additional soluble boron concentration necessary to mitigate this and all less severe accident events is.704 ppm.

For the accident due to a seismic event, the gap between rack modules was reduced to zero. For both Units 1 and 2, the reactivity increase is an order of magnitude less than that for the fuel mishandling events.

An increase in the temperature of the water passing through the stored fuel assemblies causes a decrease In water density which results in an addition of negative reactivity for flux trap design racks such as the (continued)

Vogtle Units 1 and 2 B 3.7.17-3

Fuel Storage Pool Boron Concentration B 3.7.17 BASES APPLICABLE SAFETY ANALYSES (continued)

Unit 1 racks. However, since Boraflex is not considered to be present for the Unit 2 racks and the fuel storage pool water has a high concentration of boron, a density decrease causes a positive reactivity addition. The reactivity effects of a temperature range from 320 F to 2400 F were evaluated. This bounds the temperature range assumed In the criticality analyses (500 F to 1850 F). The Increase in reactivity due to the decrease in temperature below 500 F is bounded by the misplacement of a fuel assembly between the rack and pool walls for the Unit 1 racks. The increase in reactivity due to the increase in temperature is bounded by the misload accident, for the Unit 2 racks.

Including the effects of accidents, the maximum required boron concentration to maintain Keff < 0.95 Is 851 ppm for Unit 1 and 1098 ppm for Unit 2 which is well below the limit of 2000 ppm.

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

I LCO The fuel storage pool boron concentration is required to be

> 2000 ppm. The specified concentration of dissolved boron In the fuel storage pool preserves the assumptions used In the analyses of the potential criticality accident scenarios as described in reference 5. The amount of soluble boron required to offset each of the above postulated accidents was evaluated for all of the proposed storage configurations. That evaluation established the amount of soluble boron necessary to ensure that Keff will be maintained less than or equal to 0.95 should pool temperature exceed the assumed range or a fuel assembly misload occur. The amount of soluble boron necessary to mitigate these events was determined to be 851 ppm for Unit 1 and 1098 ppm for Unit 2. The specified minimum boron concentration of 2000 ppm assures that the concentration will remain above these values. In addition, the boron concentration is consistent with the boron dilution evaluation that demonstrated that any credible dilution event could be terminated prior to reaching the boron concentration for a Kef of

> 0.95. These values are 511 ppm for Unit 1 and 394 ppm for Unit 2.

I I

APPLICABILITY This LCO applies whenever fuel assemblies are-stored in the spent fuel storage pool.

(continued)

Vogtle Units 1 and 2 B 3.7.17-4

Fuel Storage Pool Boron Concentration B 3.7.17 BASES (continued)

ACTIONS A.1. A.2.1. and A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. Immediate action to restore the concentration of boron is also required simultaneously with suspending movement of fuel assemblies. This does not preclude movement of a fuel assembly to a safe position If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the concentration of boron In the fuel storage pool Is within the required limit. As long as this SR Is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time. The gate between the Unit 1 and Unit 2 fuel storage pool Is normally open.

When the gate is open the pools are considered to be connected for the purpose of conducting the surveillance.

REFERENCES

1. USNRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition.

NUREG-0800, June 1987.

2.

USNRC Spent Fuel Storage Facility Design Bases (for Comment)

Proposed Revision 2, 1981. Regulatory Guide 1.13.

3.

ANS, ODesign Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations,"

ANSI/ANS-57.2-1983.

(continued)

Vogtle Units 1 and 2 B 3.7.17-5 I

Fuel Storage Pool Boron Concentration B 3.7.17 BASES REFERENCES (continued)

4.

WCAP-1 4416 NP-A, Rev. 1, Westinghouse Spent Fuel Rack Criticality Analysis Methodology," November 1996.

5.

Vogtle FSAR, Section 4.3.2.

6.

Nuclear Regulatory Commission, Letter to All Power Reactor Licensees from B. K. Grimes, OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

April 14, 1978.

Vogtle Units 1 and 2 B 3.7.17-6 I

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Fuel Assembly Storage In the Fuel Storage Pool BASES BACKGROUND The Unit 1 spent fuel storage racks contain storage locations for 1476 fuel assemblies, and the Unit 2 spent fuel storage racks contain storage locations for 2098 fuel assemblies.

Westinghouse 17X17 fuel assemblies with an enrichment of up to and including 5.0 weight percent U-235 can be stored in the acceptable storage configurations that are specified in Figures 3.7.18-1 (Unit 1), 3.7.18-2 (Unit 2), and 4.3.1-1 through 4.3.1-iO. The acceptable fuel assembly storage configurations are based on NRC-approved acceptance criteria for crediting soluble boron as described In the NRC's safety evaluation report in WCAP-1 4416-P-A (Reference 1). Additional background discussion can be found in B 3.7.17.

Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 3.556 w/o235U may be stored in all storage cell locations of the Unit 1 pool. Fuel assemblies with Initial nominal enrichment greater than 3.556 w/o235U must satisfy a minimum burnup requirement as shown in Figure 3.7.18-1 or a minimum Integral Fuel Burnable Absorber (IFBA) requirement as shown In Figure 4.3.1-7.

Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 5.0 w/o235U may be stored in a 3-out-of-4 checkerboard arrangement with empty cells in the Unit 1 pool.

There are no minimum burnup requirements for this configuration.

Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 1.73 w/o235U may be stored In all storage cell locations of the Unit 2 pool. Fuel assemblies with initial nominal enrichment greater than 1.73 w/o235U must satisfy a minimum burnup requirement as shown in Figure 3.7.18-2.

Westinghouse 17x17 fuel assemblies with nominal enrichments no greater than 2.40 w/o235U may be stored In a 3-out-of-4 checkerboard arrangement with empty cells in the Unit 2 pool.

Fuel assemblies with initial nominal enrichment greater than 2.40 w/o235U must satisfy a minimum burnup requirement as shown in Figure 4.3.1-8.

(continued)

Vogtle Units 1 and 2 B 3.7.1 8-1

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 BASES BACKGROUND (continued)

Westinghouse 17x17 fuel assemblies with nominal enrichments no greater that 5.0 w/o3 5U may be stored in a 2-out-of-4 checkerboard arrangement with empty cells In the Unit 2 pool. There are no minimum burnup requirements for this configuration.

Westinghouse 17x17 fuel assemblies may be stored in the Unit 2 pool In a 3x3 array. The center assembly must have an Initial enrichment no greater than 3.20 w/o235U or satisfy a minimum IFBA requirement for higher initial enrichments as shown In Figure 4.3.1-9. The surrounding fuel assemblies must have an Initial nominal enrichment no greater than 1.39 w/o235U or satisfy a minimum burnup and decay time requirement for higher Initial enrichments as shown in Figure 4.3.1-10.

APPLICABLE Most fuel storage pool accident conditions will not result SAFETY ANALYSIS In an Increase in Kff. However, accidents can be postulated for each storage configuration which could increase reactivity beyond the analyzed condition. A discussion of these accidents Is contained in B 3.7.17.

The configuration of fuel assemblies In the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO The restrictions on the placement of fuel assemblies within the fuel storage pool ensure the Koff of the fuel storage pool will always remain < 0.95, assuming the pool to be flooded with borated water.

The combination of Initial enrichment and burnup are specified In Figures 3.7.18-1 and 3.7.18-2 for all cell storage in the Unit 1 and Unit 2 pools, respectively. Other acceptable enrichment-burnup, enrichment-IFBA, and checkerboard combinations are described in Figures 4.3.1-1 through 4.3.1-10.

APPLICABILITY This LCO applies whenever any fuel assembly Is stored In the fuel storage pool.

(continued)

I1 Vogtle Units 1 and 2 B 3.7.18-2

Fuel Assembly Storage in the Fuel Storage Pool B 3.7.18 BASES (continued)

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0;3 does not apply.

When the configuration of fuel assemblies stored In the fuel storage pool is not In accordance with the acceptable combination of Initial enrichment, burnup, and storage configurations, the immediate action Is to Initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figures 3.7.18-1 (Unit 1),

3.7.18-2 (Unit 2), or Specification 4.3.1.1 (Unit 1) or 4.3.1.2 (Unit 2).

If unable to move Irradiated fuel assemblies while In MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move Irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action Is Independent of reactor operation. Therefore Inability to move fuel assemblies Is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies by administrative means that the Initial enrichment and burnup of the fuel assembly is within the acceptable burnup domain of Figures 3.7.18-1 (Unit 1) or 3.7.18-2 (Unit 2). For fuel assemblies In the unacceptable range of Figures 3.7.18-1 and 3.7.18-2, performance of this SR will also ensure compliance with Specification 4.3.1.1 (Unit 1) or 4.3.1.2 (Unit 2).

Fuel assembly movement will be in accordance with preapproved plans that are consistent with the specified fuel enrichment, burnup, and storage configurations. These plans are administratively verified prior to fuel movement. Each assembly is verified by visual Inspection to be In accordance with the preapproved plan prior to storage In the fuel storage pool.

Storage commences following unlatching of the fuel assembly in the fuel storage pool.

REFERENCES

1.

WCAP-14416-NP-A, Revision 1, 'Westinghouse Spent Fuel Rack Criticality Analysis Methodology," November 1996.

Vogtle Units 1 and 2 B 3.7.18-3-