Text

Exemption for Use of M5 Advanced Zirconium Fuel Cladding and Approval Request for Application of Supporting Analysis Methods
	ML15282A036
Person / Time
Site:	Surry
Issue date:	09/30/2015
From:	Mark D. Sartain; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	15-438
	Download: ML15282A036 (31)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 September 30, 2015 10 CFR 50.12 10 CFR 51.22(c)(9)

U.S. Nuclear Regulatory Commission Serial No.15-438 Attention: Document Control Desk NL&OS/GDM R0 Washington, DC 20555 Docket Nos. 50-280/281 License Nos. DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS I AND 2 EXEMPTION FOR USE OF M5 ADVANCED ZIRCONIUM FUEL CLADDING AND APPROVAL REQUEST FOR APPLICATION OF SUPPORTING ANALYSIS METHODS Pursuant to 10 CFR 50.12, Virginia Electric and Power Company (Dominion) requests an exemption from certain requirements of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and Appendix K to 10 CFR 50, "ECCS Evaluation Models," related to the use of up to eight lead test assemblies (LTAs) containing fuel rods fabricated with MS advanced zirconium cladding at Surry Power Station (SPS) Units 1 and 2. While the current SPS Technical Specifications (TS) permit the use of LTAs in non-limiting core locations, the requested exemption is required since the types of cladding material specified in the cited regulations for use in light water reactors do not currently include MS advanced zirconium cladding material. Dominion also requests NRC review and approval to extend the application of the Dominion and AREVA analytical methods to the evaluation of the AREVA AGORA-5A-I LTAs to demonstrate compliance with the SPS TS.

The request for exemption from the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 is provided in Attachment 1. Also, as required by 10 CFR50.22(c)(9),

Dominion has determined that the proposed exemption does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is also provided in Attachment 1.

A discussion of the analysis methods used to demonstrate compliance with the SPS TS requirement that LTAs must be loaded in non-limiting core locations, as well as the applicability of the spent fuel pool criticality analysis of record, is provided in . NRC approval is requested for application of specific portions of the analytical methods to the AREVA AGORA-5A-I LTAs.

The proposed exemption, method of determining non-limiting core locations, and applicability of the spent fuel pool criticality analysis of record have been reviewed and approved by the station Facility Safety Review Committee. The LTAs are currently

Serial No.15-438 Docket Nos. 50-280/281 Page 2 of 3 scheduled to be placed in the SPS Unit 1 core during the fall 2016 refueling outage (RFO). To facilitate the receipt and storage of the LTAs onsite in advance of core insertion during the 2016 RFO, NRC approval of the applicability of specific Dominion and AREVA reload methods and the spent fuel pool criticality analysis of record to the LTAs is requested by July 30, 2016. NRC approval of the proposed exemption is requested by September 30, 2016.

Should you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Respectfully, Mark D. Sartain Vice President - Nuclear Engineering Commitments made in this correspondence: None Attachments:

1. Exemption Request for Use of M5 Advanced Zirconium Fuel Cladding

2. Request for NRC Approval to Apply Analysis Methodologies for Evaluation of AGORA-5A-I Lead Test Assemblies

Serial No.15-438 Docket Nos. 50-280/281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 Ms. K. R. Cotton Gross NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 Dr. V. Sreenivas NRC Project Manager - North Anna U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station

Serial Number 15-438 Docket Nos. 50-280/28 1 Attachment I EXEMPTION REQUEST FOR USE OF M5 ADVANCED ZIRCONIUM FUEL CLADDING Virginia Electric and Power Company (Dominion)

Surry Power Station Units I and 2

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1 Exemption Request for Use of M5 Advanced Zirconium Fuel Cladding Surry Power Station Units 1 and 2 1.0 Exemption Summary Description Pursuant to 10 CFR 50.12, Virginia Electric and Power Company (Dominion) requests an exemption from certain requirements of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and Appendix K of 10 CFR 50, "ECCS Evaluation Models," to facilitate the use of up to eight AREVA AGORA lead test assemblies (LTAs) containing fuel rods fabricated with M5 cladding material at Surry Power Station (SPS). The requested exemption is specific to the types of cladding material specified in these regulations for use in light water reactors.

As written, 10 CFR 50.46 and 10 CFR 50, Appendix K presume the use of zircaloy or ZIRLTM uelrodclading Threfrein order to use M5 fuel rod cladding, a limited exemption to these regulations is needed. The AGORA LTAs also use a different advanced zirconium alloy, Q12TM, for the guide tubes and instrument tubes; however, an exemption is not required for the use of Q12 TM since the cited 10 CFR regulations apply specifically to fuel rod cladding.

The eight LTAs are currently planned to be used in SPS Unit I Cycle 28. The LTAs will be inserted in non-limiting core locations during the next Unit 1 refueling outage currently scheduled for fall 2016. However, the exemption request is applicable to both SPS units to maximize loading flexibility for the LTAs.

2.0 Detailed Description of Proposed Exemptions 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," states, "Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLOTM cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section." As 10 CFR 50.46 specifically refers to fuel with zircaloy or ZIRLO TM cladding, the use of fuel with MS cladding would, in effect, result in noncompliance with this section of the Code.

Also, paragraph I.A.5 of Appendix K to 10 CFR Part 50 states that the rates of energy release, hydrogen generation, and cladding oxidation from the metal-water reaction shall be calculated using the Baker-Just equation. The Baker-Just equation presumes the use of zircaloy clad fuel. Therefore, use of fuel with MS cladding would result in noncompliance with this section of the Code as well.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1

3.0 Background

The LTAs are manufactured by AREVA, Inc. (AREVA) and are identified as the AGORA-5A-I design. While a full reload batch of this fuel design has not been operated in a reload core to date, the AGORA-5A-I design represents an improved evolution of the AFA 3GTM design that is widely used in European reactors. In addition, AGORA-5A-I LTAs are currently being operated in the Ringhals Unit 2 reactor in Sweden. AREVA currently supplies nuclear fuel to numerous US reactors including Dominion's Millstone Power Station Unit 2.

The AGORA-5A-I design is very similar to the current SPS fuel assemblies. The primary difference between these assemblies is the use of M5 cladding for the AGORA-5A-I LTAs. Use of M5 as a cladding material would not be in compliance with 10 CFR 50.46 and Appendix K of 10 CFR Part 50 requirements, which has created the need for this exemption request.

SPS Technical Specification (TS) 5.2.1, "Fuel Assemblies," specifically allows the Use of a limited number of LTAs loaded in non-limiting core locations, thus no TS changes are required for the use of the AGORA-5A-I LTAs. Additional descriptions of the AGORA-5A-l assembly and how the LTAs are evaluated to be non-limiting is provided in Attachment 2.

Should Dominion decide to use the AGORA-5A-I fuel design for full batches, Dominion would submit a License Amendment Request and Core Operating Limits Report (CO LR) changes for NRC review and approval.

4.0 Basis for Exemption Request Per 10 CER 50.12, the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not present an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.12(a)(2) are present. The requested exemption to allow the use of MS advanced zirconium alloy rather than zircaloy or ZIRLO TM for fuel cladding material for LTAs at SPS Units 1 and 2 satisfies these requirements as described below.

1. The requested exemption is authorized by law.

This exemption would allow the use of MS advanced zirconium alloy, instead of zircaloy or ZIRLOTM, for fuel rod cladding in up to eight LTAs at SPS Units 1 and 2. As stated above, 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50. Therefore, the exemption is authorized by law.

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Serial Number 15-438 Docket Nos. 50-280/28 1 Attachment 1

2. The requested exemption does not present an undue risk to the public health and safety.

The underlying purpose of 10 CFR 50.46 is to ensure nuclear power facilities have adequate acceptance criteria for their EGOS. In the NRC approved Topical Report BAW-10227(P)(A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5R) in PWR Reactor Fuel," dated June 18, 2003, Framatome ANP demonstrated the effectiveness of the ECCS will not be affected by a change from zircaloy fuel rod cladding to M5 fuel rod cladding.

The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5 clad fuel.

Appendix K, paragraph I.A.5, of 10 CFR 50 ensures cladding oxidation and hydrogen generation are appropriately limited during a loss-of-coolant accident (LOCA), and conservatively accounted for in the EGGS evaluation model.

Appendix K requires the Baker-Just equation be used in the EGGS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation. In the approved Topical Report BAW-10227(P)(A), Revision 1, Framatome ANP demonstrated that the Baker-Just model is conservative in the evaluated post-LOCA scenarios with respect to the use of the MS advanced alloy as a fuel rod cladding material; therefore, the amount of energy release, cladding oxidation, and hydrogen generated in an M5-clad core during a LOCA will remain within the station design basis.

Based on the above, no new accident precursors are created by the use of MS fuel cladding for up to eight LTAs at SPS Units 1 and 2; thus, the probability of postulated accidents is not increased. Also, the consequences of postulated accidents are not increased. Therefore, there is no undue risk to public health and safety.

3. The requested exemption will not endangier the common defense and security.

The M5 fuel rod cladding is similar in design to the current cladding material used at SPS Units 1 and 2. This change in cladding material will not result in any changes to the security aspects associated with the control of special nuclear material. The change in cladding material is unrelated to other security issues.

Therefore, the common defense and security are not impacted by this exemption.

4. Special circumstances are present which necessitate the request of an exemption to the regulations of 10 GER 50.46 and 10 CFR 50 Appendix K.

Special circumstances, in accordance with 10 GER 50.12, paragraph (a)(2)(ii),

are present whenever application of the regulation in the particular circumstances would not serve the underlying purpose of the rule, or is not necessary to achieve Page 3 of 8

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1 the underlying purpose of the rule.

The underlying purpose of 10 CFR 50.46 is to ensure nuclear power facilities have adequately demonstrated the cooling performance of their EGCS. As discussed above, Topical Report BAW-10227(P)(A) concluded the M5 fuel rod cladding does not alter the effectiveness of the ECCS and also demonstrated the EGGS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5 fuel rod cladding.

As currently written, the criteria of 10 CFR 50.46 are not applicable to M5, even though analysis shows applying the zircaloy or ZIRLO TM criteria to M5 fuel yields acceptable results. Application of the regulation in this instance is not necessary to achieve the underlying purpose of the rule; therefore, special circumstances exist.

The underlying purpose of 10 CFR 50, Appendix K, paragraph I.A.5 is to ensure that cladding oxidation and hydrogen generation are appropriately limited during a LOGA and conservatively accounted for in the EGOS evaluation model.

Specifically, Appendix K requires the Baker-Just equation be used in the EGGS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation. Topical Report BAW-10227(P)(A), Revision 1, demonstrated that the Baker-Just model is conservative in the evaluated post-LOCA scenarios with respect to the use of the MS advanced alloy as a fuel rod cladding material. Therefore, the amount of energy release, cladding oxidation, and hydrogen generation for an MS advance alloy reactor core during a LOCA would remain within the station design basis.

As currently written, the criteria of 10 GFR S0, Appendix K, paragraph I.A.5 are not applicable to MS, even though analysis shows applying the zircaloy or ZIRLO TM criteria to MS fuel yields acceptable results. Application of the regulation in this instance is not necessary to achieve the underlying purpose of the rule; therefore, special circumstances exist.

Based on the above, the underlying purpose of 10 GFR 50.46 and 10 CFR S0, Appendix K will continue to be satisfied for the planned operation with eight LTAs with MS fuel rod cladding. Therefore, required special circumstances exist, which necessitate the exemption request.

In conclusion, issuance of an exemption from the specified regulations for the use of M5 fuel rod cladding material for up to eight LTAs in SPS Units 1 and 2 reactors will not compromise the safe operation of the reactors. Because the underlying purpose of the applicable NRC regulations has been preserved, it is concluded that the proposed exemptions do not present an undue risk to the public health and safety and are consistent with the common defense and security.

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Serial Number 15-438 Docket Nos. 50-280/28 1 Attachment 1 4.1 Precedents Similar exemptions have been issued for other licensed reactors, including, but not limited to: Crystal River, Unit No. 3 [ML032380538], North Anna, Units No. I and 2

[ML032590881], Arkansas Nuclear One, Unit No. 1 [ML051790417], Calvert Cliffs, Unit No. 2 [ML030640137], Calvert Cliffs, Units No. 1 and 2 [ML062260123], Calvert Cliffs, Unit No. 1 [ML073200694], Shearon Harris, Unit No. 1 [ML12025A162], and H. B. Robinson, Unit No. 2 [ML11297A103].

5.0 No Significant Hazards Consideration Determination Virginia Electric and Power Company (Dominion) is requesting an exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, to insert up to eight AGORA-5A-I Lead Test Assemblies (LTAs) containing fuel rods fabricated with M5 cladding material at Surry Power Station (SPS) Units I and 2. While current SPS Technical Specifications permit the use of LTAs in non-limiting core locations, the requested exemption is required since the types of cladding material discussed in the cited regulations for use in light water reactors do not include M5 cladding material.

The NRC has provided standards for determining whether a significant hazards consideration exists in 10 CFR 50.92(c). A determination that a proposed exemption involves no significant hazards consideration may be made if operation of the facility in accordance with the proposed exemption 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 a margin of safety. Dominion has evaluated if a significant hazards consideration (SHC) is involved with the proposed exemption request. A discussion of these standards as they relate to this exemption request is provided below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accidentpreviously evaluated?

No. The proposed exemption would allow Dominion to insert up to eight AGORA-5A-I LTAs with MS cladding at SPS.

The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, to permit the use of the MS fuel cladding material in the SPS cores does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structure, or components, or their roles in accident prevention or mitigation. MS fuel cladding material is not an accident initiator. The response of the fuel to an accident is analyzed using conservative techniques, and the results are compared to approved acceptance criteria.

Reload specific analyses conducted by Dominion and the fuel vendor demonstrate the design limits of the fuel cladding are met.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1 Station operation and analysis will continue to be in compliance with NRC regulations. In the NRC approved Topical Report BAW-10227(P)(A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," dated June 18, 2003, Framatome ANP demonstrated the effectiveness of the emergency core cooling system (ECCS) performance will not be affected by a change from zircaloy fuel rod cladding to M5 fuel rod cladding. The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5 fuel rod cladding. Thus, the plant will continue to meet applicable design criteria and safety analysis acceptance criteria.

Consequently, permitting the insertion of up to eight LTAs with M5 fuel rod cladding in the SPS cores does not affect the probability of an accident or transient or the consequences thereof. Therefore, the proposed exemption does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, does not impact the plant configuration or system performance. The proposed exemption does not modify any interfaces with existing equipment, change the equipment's function, or change the method of operating the equipment.

Use of the M5 fuel cladding material in the SPS cores does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structure, or components, or their roles in accident prevention or mitigation. In the NRC approved Topical Report BAW-10227(P)(A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," dated June 18, 2003, Framatome ANP demonstrated ECCS performance will not be affected by a change from zircaloy fuel rod cladding to M5 fuel rod cladding.

The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with MS fuel rod cladding. Using approved methods, Dominion and AREVA will demonstrate on a cycle specific basis that the LTAs perform within the fuel design limits.

The proposed exemption assures there is adequate margin available to meet safety analysis criteria and does not introduce any failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident. Therefore, the proposed exemption does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1

3. Does the proposed amendment involve a significant reduction in a margin of safety?

No. The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, does not impact the plant configuration or system performance and use of the M5 cladding material in the SPS cores does not adversely affect any fission product barrier. Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed exemption does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, and the dose analysis acceptance criteria are not affected. The proposed exemption does not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that respond to safely shutdown the plant and maintain the plant in a safe shutdown condition.

In the NRC approved Topical Report BAW-10227(P)*A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5 ) in PWR Reactor Fuel,"

dated June 18, 2003, Framatome ANP demonstrated the effectiveness of the ECCS will not be affected by a change from zircaloy fuel rod cladding to M5 fuel rod cladding. The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5 fuel rod cladding. Using approved methods, Dominion and AREVA will demonstrate on a cycle specific basis that the LTAs perform within the fuel design limits.

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

Based on the above, Dominion concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

6.0 Environmental Consideration Dominion has reviewed the proposed exemption for environmental considerations. The proposed exemption does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed exemption meets the eligibility criterion for categorical exclusion from an environmental assessment as set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemption.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 1 7.0 Conclusions 10 CFR 50.46 and 10 CFR 50, Appendix K (specifically the use of the Baker-Just equation), only apply to the use of fuel rods clad with zircaloy or ZIRLO TM .

10 CFR 50.46 and 10 CFR 50, Appendix K, do not apply to the proposed use of LTAs with M5 cladding since the composition of the cladding in these fuel rods differs from that of zircaloy or ZIRLO TM .

In order to insert up to eight AGORA-5A-I LTAs at SPS, an exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, is requested. As required by 10 CFR 50.12, the requested exemption is authorized by law, does not present undue risk to public health and safety, and is consistent with common defense and security. Approval of this exemption request does not violate the underlying purpose of the rule, and special circumstances exist to justify the approval of an exemption from the subject requirements.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 REQUEST FOR NRC APPROVAL TO APPLY ANALYSIS METHODOLOGIES FOR EVALUATION OF AGORA-5A-I LEAD TEST ASSEMBLIES Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 Request for NRC Approval to Apply Analysis Methodologies for Evaluation of AGORA-5A-I Lead Test Assemblies Surrv Power Station Units I and 2 1.0 Summary Description Dominion requests NRC approval for use of specific methodologies outside of the restrictions delineated in their respective NRC safety evaluations to evaluate AGORA-5A-I Lead Test Assemblies (LTAs).

Specifically, Dominion plans to insert up to eight LTAs fabricated by AREVA into one of the units at Surry Power Station (SPS). The approval is requested to be applicable to both SPS units to maximize loading flexibility for the LTAs. The intent of the fuel assembly irradiation program is to demonstrate the use of the AREVA AGORA-5A-l design for the SPS units.

The proposed methodology changes would allow Dominion to utilize NRC-approved methods for the limited purpose of validating the placement of AGORA-5A-I LTAs in non-limiting locations in a SPS reload core. Use of these methods would confirm the safety analyses remain bounding by confirming key inputs to the safety analyses remain conservative with respect to the cycle design.

Dominion is also requesting NRC approval for use of AREVA fuel evaluation methods at SPS for evaluation of the AGORAR-5A-I LTAs and for application of the existing spent fuel pool criticality analysis to the storage of the LTAs.

At this time, Dominion is not seeking to alter the SPS TS or Core Operating Limits Report (COLR). Should Dominion decide to pursue the use of the AGORA-5A-l fuel design for full batches, Dominion would submit the necessary TS changes for NRC review and approval.

2.0 Detailed Description 2.1 Introduction As the nuclear industry has pursued longer operating cycles with increased fuel discharge burnup and fuel duty, the performance requirements for nuclear fuel cladding have become more demanding. M5 is an advanced fuel rod cladding composed primarily of zirconium and niobium. It provides significant improvements in corrosion resistance, hydrogen pickup, axial growth, and diametral creep relative to zircaloy.

Currently, the only 15x15 fuel assembly design approved for use in the US nuclear industry with M5 cladding is the AREVA HTP design. This design, while mechanically robust, does not offer the same thermal hydraulic margins associated with the current SPS fuel product, the Westinghouse 15x15 Upgrade design. AREVA has developed Page 1 of 18

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 another design, AGORA-5A-I, to combine the benefits of the M5 cladding with thermal hydraulic performance comparable to the current SPS fuel product.

The AGORA-5A-I design is an evolutionary Westinghouse 15x15 (W15) array design that has been developed by AREVA and used in Europe. The AGORA-5A-I design is very similar to fuel types currently utilized by Dominion at its power stations, including the Westinghouse 15x15 Upgrade fuel type used at SPS. The performance of the AREVA assemblies under design basis conditions is expected to be comparable to the performance of the current fuel design. A description of the AGORA-5A-I design is provided in Section 2.2 of this attachment.

Dominion will operate these LTAs in accordance with the current SPS TS, specifically TS 5.2.1, which allows the use of a limited number of LTAs in non-limiting locations of the core. For these LTAs, Dominion will demonstrate existing SPS UFSAR analysis limits, which are currently evaluated based on the existing Westinghouse fuel design, are met.

The SPS COLR contains multiple methods approved for use in evaluating the reload performance of the SPS reload cores and ensuring that all applicable operating limits are met. Dominion's reload design and analysis methods have been approved for a broad array of fuel types, including Westinghouse and AREVA (with M5 cladding) assemblies, and are not specific to a single fuel assembly design. Dominion has experience successfully applying these methods to reload evaluations. For cores operating with AGORA-5A-l LTAs, Dominion will demonstrate the results of the existing UFSAR safety analyses remain set by the resident fuel design and are bounding for the reload cores containing the AREVA assemblies by applying a peaking factor reduction to the LTAs.

Dominion will evaluate the reload core containing the LTAs with these approved methods to ensure the LTAs are in non-limiting core locations as required by the current TS. The thermo-mechanical performance of the LTAs will be evaluated using approved AREVA methods to ensure they remain within the design and safety limits.

Within the subset of approved Dominion analytical methods that Dominion desires to use for the LTAs, certain portions of the approved methods will be used outside of the restrictions associated with their approval in order to evaluate the LTAs. The AREVA fuel evaluation methods are not currently listed in the SPS TS. The spent fuel pool criticality analysis does not explicitly account for the differences between the resident and AGORA fuel design.

Therefore, Dominion is requesting NRC approval for the application of Dominion core thermal-hydraulic and reload design methods (methods that are not specifically approved for application to the AGORA fuel product) for the limited purpose of validating non-limiting performance. Dominion is also requesting NRC approval for use Page 2 of 18

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 of AREVA fuel evaluation methods at SPS for evaluation of the AGORA-5A-I LTAs and for application of the existing spent fuel pool analysis to storage of the LTAs.

2.2 AGORA-5A-I Design Description The features of the ARE VA AGO RA-5A-I design are:

0.424 inch OD, M5 clad fuel rods in a 15x15 array

0.555 inch OD, QI2 TM MONOBLOC TM guide tubes and instrument tube

Standard reconstitutable upper end fitting

TRA*PPER TM debris resistant lower end fitting

A bottom Alloy 718 HMP TM spacer grid, 5 intermediate AFA 3GTM mixing vane spacer grids, 3 mixing vane mid-span mixer grids (MSMG) , and a top M5 HTP TM grid

Gadolinia burnable absorber The AGORA-5A-I design is an evolution of a European W15 design, the AFA 3GTM design, which has a lengthy history of excellent operation in European reactors. The original design used the 0.422 inch M5 clad rods with AFA 3GTM mixing grids, AFA 3GTM MSMGs, and M5 MONOBLOC TM guide tubes, and used the TRAPPER TM lower end fitting. The AGORA-5A design added a standard, quick-disconnect reconstitutable upper end fitting, replaced the lowermost AFA 3G TM grid with an Alloy 718 HMPTM, and increased the rod diameter to 0.424 inch. Improvements were made to the AGORA-5A design that changed the guide tube and instrument tube material from MS to Q12 TM and replaced the top AFA 3GTM grid with an MS HTP TM . Vaned mid-span mixing grids were an option on the AFA 3GTM and AGORA-5A designs and were supplied for reloads at one European Pressurized Water Reactor (PWR). The AGOA -A-ldesign incorporates the improvements made to the AGORA-5A design and includes anti-hangup tabs on the sideplates on the AFA 3GTM grids between each rod position instead of at every other rod position. The grid internal strips are also modified to provide support to the guiding vanes, and their bottom edges are chamfered to compensate for any pressure drop changes from the additional anti-hangup tabs.

The AFA 3GTM design has been used in reload cores since 2000. The AGORA-5A design has been used in reload cores since 2007. Several AGORA-5A-I LTAs are currently being irradiated in Europe.

For SPS, several minor changes were made to the AGORA-5A4I design:

The end caps have been changed to match the configuration of the end caps used in the US:

o The upper end cap change makes the rod compatible with the post irradiation inspection tooling, and o The lower end cap change facilitates rod insertion into the fuel assembly during manufacturing.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2

The plenum spring was modified to preclude the possibility of inserting it incorrectly because the manufacturing process is slightly different in Europe.

All of the identified features except the AFA 3GTM grids, MSMGs, and Q12 TM have been previously used in reloads in the US in other AREVA PWR fuel designs. The AFA 3GTM spacer has been used in Europe for several reloads at two W15 type plants. The MSMG has also been used in multiple reloads. Q12 TM is currently being used in LTA programs at both a US and European reactor.

The Q12TM material is a quaternary Zirconium, Niobium, Tin, and Iron alloy developed for PWR fuel assembly structural components (e.g., guide tubes, instrument tubes, etc.). The alloy increases the creep resistance with respect to an M5 tube, and maintains the corrosion behavior within the Zircaloy-4 experience. The increased creep resistance results in the assembly being less susceptible to fuel assembly bow due to compressive loads. AREVA has multiple lead assembly programs underway in several fuel array sizes -that use the Q12TM material. The post irradiation examinations from these lead programs show that the observed properties of the Q1 2 TM components are well behaved and within the design methods. The irradiation growth has been positive (i.e., it has grown), but small. The variability in growth has also been small when compared with the Zircaloy-4 and M5 experience.

The AGORA-5A-I assemblies are designed for full compatibility with the SPS mechanical interfaces, including the core internals, control and insert components, the resident fuel, and shipping and handling tools. Compatibility of AREVA supplied fuel with resident Westinghouse fuel and core components, as well as Westinghouse designed core internals, has been demonstrated in the past through successful reload transition experiences at other stations, including Dominion's North Anna Power Station.

The LTAs are very comparable to the current fuel design in use at SPS. AREVA has experience with design, manufacture, and analysis of fuel designs in use at US reactors today and has used this expertise in the design of the AGORA -5A-I assemblies. Thus, the performance of the AGORA-5A-I LTAs under design basis conditions is expected to be comparable to the performance of the current fuel design. The properties of M5 fuel rod cladding are well understood by AREVA, and Dominion has experience with M5cladding from AREVA assemblies at the North Anna Power Station.

2.3 Post Irradiation Examination (PIE)

The AGORA-5A-I fuel design is an evolution of a European W15 design, the AFA 3GTM design, which has a lengthy history of excellent operation in European reactors, but has not been used in US reactors in the past. Dominion plans to conduct post irradiation examinations of the fuel in cooperation with AREVA throughout the LTA program.

While subject to change, the current expectations for PIE for the AGORA-5A-I LTAs are as follows:

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 Post Cycle 1:

Visual inspection o Including specific checks for assembly growth, bow, and twist and rod growth

Peripheral rod inspection of oxide thickness Post Cycle 2:

Visual inspection o Including specific checks for assembly growth, bow, and twist and rod growth

Peripheral rod inspection of oxide thickness Post Cycle 3:

Visual inspection o Including specific checks for assembly growth, bow, and twist and rod growth

Oxide thickness measurements

Crud scraping/sampling

Grid-to-rod fretting wear examinations including rod diameter checks 2.4 Definition of Non-Limiting The SPS TS allow for a limited number of LTAs to be inserted in non-limiting locations in the SPS cores. For the AGORA-5A-I LTAs, non-limiting refers to ensuring the AGORA-5A-I LTAs are bounded by the resident fuel (Westinghouse 15x15 Upgrade) with respect to SPS UFSAR Chapter 14 non-Loss of Coolant Accident (LOCA) and LOCA performance.

The Dominion method for demonstrating compliance with the SPS TS 5.2.1 requirement of non-limiting locations for non-LOCA transients is: 1) the extension of applicability of existing, approved core and thermal-hydraulic design methods to the AGORA-5A-I LTAs, and 2) demonstration that an applied constraint on allowable assembly-wise peaking factor (FdH) for the LTAs will ensure the performance of the AGORA-5A-I LTAs is less limiting than the resident fuel product. Maintaining approximately 5% FdH margin between the AGORA-5A-I LTAs and the resident fuel product is sufficient to ensure non-limiting performance for the LTAs. NRC approval is requested for the application of Dominion core thermal-hydraulic and reload design methods (methods that are not specifically approved for application to the AGORA* fuel product) for the specific purpose of validating non-limiting performance for the AGORA -5A-I LTAs.

The Dominion method for demonstrating compliance with the SPS TS requirement of non-limiting locations for LOCA is the maintenance of FdH margin between the AGORA-5A-I LTAs and the resident fuel product in reload core designs to ensure a reduction in total peaking factor (Fq). Maintaining approximately 5% FdH margin between the AGORA-5A-I LTAs and the resident fuel product is sufficient to ensure non-limiting performance for the AGORA-5A-I LTAs.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 2.5 Description of the Use of Approved Methods Outside of the Constraints of the NRC Safety Evaluation Reports The current SPS TS Section 6.2.C lists the following approved methods for use in the analysis of reload cores:

VEP-FRD-42, "Reload Nuclear Design Methodology"

WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)"

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code"

WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code"

WCAP-1 2610-P-A, "VANTAGE+ Fuel Assembly Report"

WCAP-1 2610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO"

VEP-N E-2-A, "Statistical DN BR Evaluation Methodology"

VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code"

DOM-NAF-2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," August 2010 and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code"

WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function" To ensure the LTAs are located in non-limiting locations, as required by the current TS 5.2.1, Dominion is seeking approval to utilize the approved reload design method, VEP-FRD-42, and the approved thermal-hydraulic evaluation method, DOM-NAF-2, outside of specific restrictions associated with their respective approvals. The non-limiting locations (and associated FdH margin described in Section 2.4), will be used to ensure the safety analyses performed in accordance with approved COLR methods remain applicable.

For the application of these methods to demonstrate that the LTAs are placed in non-limiting locations for the cycle specific core design, Dominion will utilize these approved methods outside of their respective Safety Evaluation Reports (SER):

VEP-FRD-42-A, "Reload Nuclear Design Methodology." The approved reload nuclear design method is used to demonstrate that the cycle-specific core design remains within the bounds of the ap,.[licable safety analyses for SPS. This includes cycles containing the AGORA -5A-I LTAs.

o The SER for VEP-FRD-42 explicitly states that its use is limited to Westinghouse fuel and AREVA's Advanced Mark-BW fuel product.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 Therefore, application to the AGORA-5A-I fuel design is conservatively deemed to be outside of the NRC SER, and NRC approval is requested for application to the evaluation of LTAs only.

DOM-NAF-2(P)-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code." The approved VIPRE-D code and Critical Heat Flux (CHF) correlations are used to demonstrate that DNB margin is present in the cycle-specific core design. For cycles containing the AGORA-5A-I LTAs, the VIPRE-D/WRB-1 code/correlation pair will be used to demonstrate that maintaining approximately 5% FdH margin for the AGORA-5A-l LTAs ensures non-limiting performance for the AGORA design, including the effects of flow redistribution.

o As part of the application of DOM-NAF-2 for the specific use of the LTAs in non-limiting core locations, Dominion plans to apply the WRB-1 CHF correlation. The WRB-1 correlation is currently approved only for application to Westinghouse fuel.

For evaluations of the thermo-mechanical performance of the LTAs, which are not associated with the definition of non-limiting, Dominion is seeking approval to utilize the following NRC-approved AREVA methods that are not included in the current SPS TS:

1) EMF-92-1 16, "Generic Mechanical Design Criteria for PWR Fuel Designs,"

2) BAW-10240, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," and 3) BAW-10084, "Program to Determine In-Reactor Performance of BWFC Fuel Cladding Creep Collapse." These approved AREVA methods will be utilized to demonstrate the mechanical performance of the AGORA-5A-I LTAs.

EMF-92-116(P)-A, "Generic Mechanical Design Criteria for PWR Fuel Designs" and BAW-10240(P)-A, "Incorporation of M5TM Properties in Framatome ANP Approved Methods." These methods are used for evaluating the mechanical design of the AGORA-5A-I LTAs with M5 clad fuel rods.

o Both methods were approved based on use of the RODEX2 code, but the approved COPERNIC TM code (BAW-10231(P)-A, "COPERNIC TM Fuel Rod Design Computer Code") is proposed for the analysis of the AGORA-5A-I LTAs instead. COPERNIC TM results include the effects of Thermal Conductivity Degradation (TCD), which would have required a separate disposition if RODEX2 were used.

BAW-1 0084(P)-A, "Program to Determine In-Reactor Performance of BWFC Fuel Cladding Creep Collapse." This method refers to the CROV computer code for calculation of the cladding creep collapse behavior.

o This method has not been approved for use at SPS.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 Dominion will not alter the existing spent fuel pool analysis (Reference 10). Per this evaluation, fuel stored in Region 2 of the pool is allowed to have enrichment of 4.3 weight percent and there are no restrictions on storage pattern or credit for soluble boron. The use of this analysis with respect to the AGORA-5A-I LTAs has not been approved at SPS.

For all other methods used to evaluate the reload core, the non-limiting location of the AGORA assemblies is used to disposition their irradiation against the existing analyses of record. No method changes were used to perform the dispositions.

3.0 Technic:al Evaluation Dominion will develop core designs for SPS wherein the LTAs are placed in core locations where the peaking factors are limited under normal operating conditions. The current resident fuel will be allowed to operate within the FdH limit defined in the SPS COLR. Dominion will utilize approved design methods to ensure the LTAs maintain approximately 5% FdH margin to the most limiting resident fuel throughout the cycle.

With regards to methods being used outside of the NRC SERs, Dominion and AREVA have evaluated the use of these methods for compatibility with the AGORA-5A-I LTAs.

In all cases, it was determined the methods would result in conservative evaluations of the LTAs. The use of the LTAs in non-limiting locations ensures that margin, including the effects of core flow redistribution, is preserved to the existing SPS limits. Prior to the use of AGORA-5A-I assemblies for a reload batch, or unrestricted use, Dominion will submit a License Amendment Request to the NRC.

VEP-FRD-42-A The VEP-FRD-42 NRC SER explicitly identifies that the approved method is applicable to Westinghouse fuel and AREVA's Advanced Mark-BW fuel product. Dominion proposes to apply VEP-FRD-42 to the analysis of reload cores containing AGORA-5A-I LTAs in non-limiting locations. The VEP-FRD-42 SER states that the method may be applied to other fuel designs if Dominion is able to provide confirmation that the fuel design of interest and its specific features can be accurately modeled with the approved nuclear design and safety analysis codes and methods. The AGORA-5A-I fuel design is neutronically, thermal hydraulically, and mechanically similar to the resident Westinghouse fuel, including: the base design (WI5), cladding and structural materials, and pellet composition. Dominion has utilized the VEP-FRD-42 reload method for the design and operation of multiple Surry, North Anna, and Kewaunee cycles without issue; thus, demonstrating the robustness of the method. Furthermore, previous North Anna cycles with M5 cladding material were effectively analyzed using the VEP-FRD-42 method. Thus, the AGORA-5A-I LTAs are within the Dominion operational experience for application of VEP-FRD-42, and Dominion is able to accurately model the AGORA-5A-I assembly design on a cycle-specific basis without any modification to the existing codes and methods.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 The use of the VEP-FRD-42 method for evaluation of reload core designs was approved with the following conditions. The responses following each condition demonstrate that the implementation of the proposed change for SPS will remain in compliance with requirements identified in the Safety Evaluation (SE) associated with the topical report.

1. Prior to use of the VEP-FRD-42 method for fuel types other than Westinghouse and Framatome ANP Advanced Mark-BW fuel, VEPCO [Dominion] must confirm that the impact of the fuel design and its specific features can be accurately modeled with the nuclear design and safety analysis codes and methods describedin the VEP-FRD-42 topical report.

The AGORA-5A-I assembly design is very similar, neutronically, thermal hydraulically, and mechanically to the current fuel products modeled with the nuclear design and safety analysis codes and methods described in the VEP-FRD-42 topical report. Dominion has demonstrated that modeling the LTAs is possible with the existing codes and methods without changes to the reload methodology.

2. When transitioning fuel products, VEPCO [Dominion] must submit a license amendment request to add the applicable and approved thermal hydraulic method references to the COLR TS section. These methods must have NRC review and approvalprior to being listed in the COLR section of the TS.

Dominion is not transitioning fuel products at this time. Prior to using AGORA-5A-I fuel design for full batches, Dominion would submit a License Amendment Request for NRC review and approval.

3. Use of the core design methods for North Anna and Surry shall be in accordance with the restrictionsand limitations of the approved VEPCO [Dominion] methods.

No changes to the approved core design methods are required for modeling the AGORA -5A-I LTAs. Dominion shall continue to utilize its approved core design methods in accordance with the restrictions and limitations of the associated NRC SER.

Therefore, Dominion requests approval for the use of the VEP-FRD-42 reload design method to SPS cores containing the AGORA-5A-I LTAs in non-limiting locations.

DOM-NAF-2(P)-A The DOM-NAF-2 SER allows the use of multiple CHF correlations for the thermal-hydraulic analysis of multiple fuel types. WRB-1 is an approved CHF correlation for use with the resident Westinghouse fuel (15x15 Upgrade). Dominion proposes to use the VIPRE-D/WRB-1 code/correlation pair for the specific purpose of validating non-limiting Page 9 of 18

Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 performance for the AGORA-5A-I LTAs. Dominion has applied the WRB-1 correlation to experimental test data provided by AREVA to demonstrate the applicability of the VIPRE-D/WRB-1 code/correlation pair to the LTAs. AREVA performed CHF tests at the Columbia Heat Transfer Research Facility and supplied the data from the AFA TM tests to Dominion for use in a validation study.

Dominion was able to model the tests using the VIPRE-D code and compare the CHF results from application of the WRB-1 correlation to AREVA's experimental data. Based on this evaluation, Dominion is able to conservatively calculate DNB ratios for the AGORA-5A-I fuel under the current design limit of 1.17 with the following constraint:

A lower mass velocity limit of 1.40 Mlbm/hr-ft 2 is required to ensure conservative prediction of the AGORA-5A-I fuel thermal performance (instead of the current lower mass velocity limit of 0.90 Mlbm/hr-ft). This limit is only applicable to evaluations of the AGORA-5A-I LTAs and does not change the approved mass velocity range in DOM-NAF-2. Dominion will check this limit on a reload basis for the LTAs.

1.5 . . . ... ...... ... .......... ..... ... . . . .... . . .. . . ..

1.4

CU47.1 U CU48.1 1.3 ... ... . .. ............. ... .....t 1.2 lii

54 1t 1.1 ...

iF 0.

4 U.

1 P 4 m 4 - 4

'4 I D vO 0.9 ---------------

a.I 0.8 0.7 0.6 ...

0.5 1 1.5 2 2.5 3 3.5 Mass Velocity (Mlbm/hr-ft 2 )

Figure 1: MeasuredlPredicted TM ratios for the application of WRB-1 to data from two of the AFA CHF tests.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 As shown in Figure 1, the M/P ratios are well aligned with the AREVA data set. There are six test points that fall below the VIPRE-D/WRB-1 code/correlation design limit of 1.17 (-0.85 in M/P space as shown in Figure 1). All six of these points are grouped at low mass velocities (less than 1.40 Mlbm/hr-ft'). The data grouped in this low mass velocity regime also, as a set, appears offset relative to the sets at higher mass velocities. To ensure that application of the VIPRE-D/WRB-1 code/correlation pair remained conservative, Dominion imposed the limit on mass velocity described above and will demonstrate the limit is met for each reload cycle containing the LTAs.

The AGORA-5A-I LTAs have an equivalent heated hydraulic diameter, for the thimble cell only, of ,-0.60 inches. This is larger than the maximum of the WRB-1 applicable range (0.58 inches) as described in WCAP-8762, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids." For the AGORA-5A-I LTAs, Dominion utilized AREVA test results *erformed with a heated hydraulic diameter of -0.61 inches (greater than the AGORA -5A-I LTAs) to validate use of the VIPRE-D/W\RB-1 code/correlation pair. Dominion utilized the AFA TM experimental results to demonstrate the WRB-1 CHF correlation yielded a conservative assessment of the test data and, therefore, is acceptable for use with the AGORA-5A-I LTAs.

The AGORA-5A-I LTA rod outer diameter is 0.424" instead of the 0.422" outer diameter in the approved WRB-1 topical. This difference in outer diameter measure (0.002") is considered too small to have a significant impact on CHF performance. This difference is on the same order as the manufacturing tolerance on the AGORA rod diameter, and the rod pitch to rod diameter ratio of the as-built design is within 0.5% of CHF test data. Thus, the difference between the rod diameter of the AGORA-5A-I design and that used during generation of the AFATM test data is not considered significant enough to perturb the relationship between local conditions and CHF predictions defined through the WRB-1 CHF correlation. Therefore, CHF tests performed with a 0.422" outer diameter are viewed as applicable for rods with a 0.424" outer diameter.

An assessment of the thermal performance of the AGORA-5A-I LTAs with the VIPRE-D/WRB-1 code/correlation pair demonstrated that the described FdH margin of approximately 5% is sufficient to ensure non-limiting performance for the AGORA -5A-I fuel design. This assessment included modeling a single AGORA-5A-I assembly in a core of Westinghouse 15x1 5 Upgrade fuel to account for the flow redistribution between dissimilar assemblies.

The use of the DOM-NAF-2 method for evaluation of Departure from Nucleate Boiling Ratio (DNBR) for PWR transients was approved with the following conditions. The responses following each condition demonstrate that the implementation of the proposed change for SPS will remain in compliance with the requirements identified in the SE associated with the topical report.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2

1. The use of the VIPRE-D code is limited to only the CHF correlationsapprovedin the topical report.

Dominion proposes the use of WRB-1 for the evaluation of AGORA-5A-I LTAs, a correlation which is approved by the NRC for use with VIPRE-D per Appendix B of DOM-NAF-2. In addition, Dominion proposes the use of the W-3 alternate correlations, as approved in Appendix D of DOM-NAF-2, in place of the W-3 correlation for the evaluation of the AGORA-5A-I LTAs: the ABB-NV correlation for the non-mixing vane region below the first mixing vane grid, and the WLOP correlation for low pressure transients (e.g., Main Steam Line Break event). The AGORA-5A-I design parameters are within the approved ranges for use of the ABB-NV and WLOP correlations.

2. The VIPRE-D code can be used subject to the models and options specified in Sections 4.0 - 4.12 of DOM-NAF-2.

Dominion continues to utilize the approved VIPRE-D code in compliance with the models and options specified in Section 4.0 - 4.12 of DOM-NAF-2. Use of these models and options is ensured through training and procedures.

3. Use of the VIPRE-D code with WRB-1 is explicitly approved only for use with Westinghouse fuel.

This condition constitutes an exception to the use of the method and requires NRC approval per this request.

4. The WRB-1 correlation is limited to conditions where the local heat flux is less than 1.0 MBTU/hr-ft2 .

Dominion ensures the local heat flux remains less than 1.0 MBTU/hr-ft 2 when using the VIPRE-D code for reload specific evaluations.

5. The W-3 correlation will be used when the conditions fall outside the range of the WRB-1I correlation.

As documented in Appendix D of DOM-NAF-2, Dominion has received NRC approval for the use of the W-3 alternate correlations, ABB-NV and WLOP.

Dominion will apply the W-3 alternate correlations outside of the range of applicability for the WRB-1 correlation within the constraints of the DOM-NAF-2 SER.

6. The VIPRE-D code is restricted for application to the transients listed in DOM-NAF-2 (Table 2.1-1) and the uses and applications listed in Section 2.1 of DOM-NAF-2.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 Approval for the use of DOM-NAF-2 at SPS was provided by the NRC in a letter dated October 19, 2010 (Reference 8). As part of that approval, the list of transients for which DOM-NAF-2 constitutes an acceptable method of analysis was slightly changed relative to Table 2.1-1 and Section 2.1 of DOM-NAF-2.

Analysis of the AGORA-5A-l LTAs is consistent with the NRC approval for use of DOM-NAF-2 at SPS. The AGORA-5A-I LTAs are manufactured with a 15x15 design and will be located in a PWR. VIPRE-D will be used on a reload specific basis to demonstrate DNBR limits are met for the approved statistical and deterministic transients and for steady state and transient DNB evaluations.

Core thermal limit lines, reactor protection setpoints, and DNBR design limits are not being altered for the implementation of AGORA-5A--l LTAs.

Based on the conservative results obtained relative to AREVA experimentally obtained CHF data, Dominion requests approval to use the VIPRE-D code and WRB-1, ABB-NV, and WLOP correlation pairs for the specific purpose of validating non-limiting performance for the AGORA-5A-I LTAs.

EMF-92-116(P)(A), BA W-1O240(P)(A), BA W-10231(P)(A), and BA W-10084(P)(A)

NRC approval for both EMF-92-116 and BAW-10240 was based on AREVA submittals utilizing the RODEX2 code. Dominion proposes to use the methods and limits described in both topical reports in conjunction with the COPERNICTM code instead of RODEX2 for the analysis of AGORA-5A-I LTAs in non-limiting locations. AREVA will be performing the fuel mechanical design evaluations for the AGORA-5A-I LTAs. The AREVA topical reports (EMF-92-116 and BAW-10240) are approved for use with fuel clad in M5, the same cladding used for the AGORA -5A-I LTAs. The COPERNIC TM fuel performance code was reviewed and approved by the NRC for use in the evaluation of AREVA fuel designs with M5 cladding in Topical Report BAW-1 0231 (P)(A).

The use of AREVA approved methods for M5 cladding was approved in Topical Report BAW-10240 with the following conditions. The responses following each condition demonstrate the implementation of the proposed change for SPS will remain in compliance with requirements identified in the SE associated with the topical report for the AGORA-5A-I LTAs.

1. The corrosion limit, as predicted by the best-estimate model will remain below 100 microns for all locations of the fuel.

The restriction that the corrosion limit, as predicted by the best-estimate model, will remain below 100 microns for all locations of the fuel is implementing the AREVA fuel design processes. This limit is verified for each reload as part of the cycle-specific reload analysis.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2

2. All of the conditions listed in the NRC SEs for all ARE VA methodologies used for M5* fuel analysis will continue to be met, except that the use of M5*~ cladding in addition to Zircaloy-4 cladding is now approved.

Conditions from SEs are incorporated as restrictions in AREVA design procedures and guidelines that control the core reload designs for SPS. This is verified for each reload as part of cycle-specific reload analysis.

3. All ARE VA methodologies will be used only within the range for which M5 data was acceptable and for which the verifications discussed in BA W-10240(P)(A) and BA W- 10227(P)(A) were performed.

Limitations to ensure AREVA methodologies will be used only within the range for which M5 data was acceptable, and for which the verifications discussed in BAW-10240(P)(A) or BAW-10227(P)(A) were performed, are incorporated as restrictions in AREVA design procedures and guidelines that control the core reload designs for SPS. This is verified for each reload as part of cycle-specific reload analysis.

4. The burnup limits for implementation of M5 is 62 GWd/MTU.

This limit, identified in approved methodologies, is contained in SPS core functional requirements and AREVA design processes, and is currently limited to 62 GWd/MTU. The limit is verified for each reload as part of cycle specific reload analysis.

The use of COPERNICTM was approved with the following conditions. The responses following the conditions demonstrate that implementation of the proposed change for SPS will remain in compliance with requirements identified in the SE associated with the topical report.

1. The burnup limits for implementation of M5is 62 GWd/MTU.

This limit, identified in approved methodologies, is contained in SPS core functional requirements and AREVA design processes, and is currently 62 GWd/MTU. The limit is verified for each reload as part of cycle specific reload analysis.

2. Licensees referencing the topical report need to meet 10 CFR 51.52, "Environmentaleffects of transportationof fuel and waste" - Table S-4.

Use of the COPERNIC TM topical report does not affect Dominion's compliance with 10 CFR 51.52. Existing Dominion procedures and processes ensure that transport of fuel to and from the reactor remains in compliance with 10 CFR 51.52 requirements.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 BAW-10084 describes the method used to calculate the creep collapse behavior of Zirconium alloy tubes with the CROV code. The code uses a cladding creep model driven by the pressure difference between the system pressure and the rod internal pressure to determine the ovalization and time to collapse of the fuel rod tube. A gap in the fuel column greater than the maximum potential fuel column irradiation densification is evaluated. The system and rod conditions (e.g., temperature, pressure, flux, etc.) are provided by a COPERNIC TM fuel rod performance evaluation. The results of the CROV evaluation are compared with defined collapse criteria to assure the cladding will not collapse over the design lifetime, thus satisfying the specified acceptable fuel design limit (SAFDL) requirement of NUREG-0800, the Standard Review Plan. The CROV code has been reviewed and generically approved by the NRC and has been used for reload applications of AREVA fuel designs used at B&W and Westinghouse reactors.

Dominion proposes the COPERNIC TM code be used for evaluation of the fuel mechanical design criteria (e.g., rod internal pressure analyses, cladding corrosion, cladding strain, cladding fatigue, and cladding stress) for the AGORA-5A-I LTAs in non-limiting locations. The use of the CROV code to calculate cladding creep collapse is also proposed. Therefore, Dominion requests approval for the use of EMF-92-1 16, BAW-10240, BAW-10231, and BAW-10084 for the evaluation of the AGORA-5A-I LTAs at SPS.

3.1 Spent Fuel Pool Criticality with AGORA L TAs NRC approval is also requested for application of the spent fuel pool criticality analysis of record to storage of the AGORA -5A-l LTAs. The current SPS Spent Fuel Pool analysis, approved by Reference 10, allows for storage of fuel enriched to 4.3 weight percent in U-235. However, this anal~sis was performed for the standard Westinghouse 15x15 fuel design, while the AGORA -5A-I LTAs have the following differences: slightly larger fuel clad outer diameter, slightly higher fuel pellet density, and a slightly larger diameter fuel pellet. The increase in clad diameter for the AGORA-5A-I LTAs is on the same order as the manufacturing tolerance for the analyzed fuel, and the effect with regard to fuel reactivity is neutral or slightly decreasing. The increase in uranium content caused by the higher pellet density and pellet diameter is similar in magnitude to uranium content tolerances documented in the current analysis. These changes increase fuel reactivity.

Dominion has concluded that application of the following restrictions to the eight AGORA-5A-I LTAs more than offsets the reactivity effect of the identified fuel assembly design differences, and ensures the current criticality analysis results remain bounding:

1) U-235 enrichment less than or equal to 4.0 weight percent; and 2) administrative restriction of the LTA storage locations to only Region 2 of the pool. Per the current analysis, fuel stored in Region 2 of the pool is allowed to have enrichment of 4.3 weight percent and there are no restrictions on storage pattern or credit for soluble boron.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The NRC approved use of the Dominion and AREVA methods described herein subject to the conditions set forth in the NRC SEs for the following topical reports:

VE P-F RD-42 (P)-A, DO M-NAF-2 (P)-A, EM F-92-1 16(P)-A, BAW-1 0240(P)-A, BAW-10231(P)-A, and BAW-10084(P)-A. Each of the conditions associated with the topical reports required for NRC approval were specifically addressed for applicability to the AGORA-5A-I LTAs at SPS in Section 3.0. In summary:

-VEP-FRD-42: NRC approval is requested to apply this method to the evaluation of the AGORA-5A-l LTAs in non-limiting core locations at SPS. All other limits, conditions, and requirements are verified herein and on a reload specific basis.

Use of this method supports Dominion's compliance with the SPS requirement for placement of the LTAs in non-limiting core locations.

-DOM-NAF-2: NRC approval is requested to apply this method, and the WRB-1 correlation, to the evaluation of the AGORA-5A-I LTAs for the specific purpose of ensuring non-limiting performance for the LTAs. In addition, NRC approval is requested to apply the WRB-1, ABB-NV, and WLOP CHF correlations, which were originally approved for use with Westinghouse fuel, to the AGORA-5A-I LTAs. All other limits, conditions, and requirements are verified or evaluated herein and on a reload specific basis. Use of this method supports Dominion's compliance with the SPS requirement for placement of the LTAs in non-limiting core locations.

-EMF-92-116, BAW-10240, BAW-10231, and BAW-10084: NRC approval is requested to apply these methods to the evaluation of the AGORA-5A-I LTAs in non-limiting core locations at SPS. In addition, NRC approval is requested to apply the COPERNIC TM code in place of the RODEX2 code for evaluations of the LTAs otherwise performed in conformance with the limits, conditions, and restrictions of EMF-92-116, BAW-10240, BAW-10231, and BAW-10084. All other limits, conditions, and requirements are verified herein and on a reload specific basis.

-Reference 10: NRC approval is requested to apply the existing spent fuel pool analysis of record to the storage of the AGORA-5A-I LTAs in the SPS spent fuel pool.

An exemption request from the requirement of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling systems for Light-Water Nuclear Power Reactors," and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models," for use of M5 cladding is required. The exemption request is provided in Attachment I of this letter. The proposed changes do not require relief from any other regulatory requirements.

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Serial Number 15-438 Docket Nos. 50-280/281 Attachment 2 4.2 Precedent While the use of AREVA M5TM cladding has numerous precedents as described in and Dominion reload methods were specifically approved for application to the AREVA Advanced Mark-BW fuel design, there is no precedent that covers all aspects of the AGORA-5A-I fuel design proposed in this request.

Also, FdH margin of approximately 5% was used to ensure LTAs remained non-limiting in the Dominion submittal for Optimized ZIRLO LTAs at Millstone Unit 3. Discussion of the approximately 5% FdH margin was provided to the NRC in response to Requests for Additional Information in correspondence dated November 10, 2003 (Reference 9).

5.0 Conclusions The AGORA-SA-I lead test assemblies are mechanically, thermal hydraulically, and neutronically very similar in design to the Westinghouse fuel design that comprises the remainder of the core. The reload core design for SPS cycles which incorporate the lead test assemblies will meet all applicable design criteria, and will not result in any changes to the SPS Units 1 and 2 operating and safety analysis limits. The existing safety analyses based on the resident Westinghouse fuel design will remain applicable for cores incorporating the AGORA-5A-I lead test assemblies.

Based on the considerations discussed above, there is reasonable assurance that (1) the health and safety of the public will not be endangered by the operation of SPS with the AREVA LTAs with M5 fuel rod cladding, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the requested changes will not be inimical to the common defense and security or to the health and safety of the public.

6.0 References

1. VEP-FRD-42(P), Revision 2.1-A, "Reload Nuclear Design Methodology,"

August 2003.

2. DOM-NAF-2(P)-A, Revision 0.3, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," September 2014.

3. EMF-92-1 16(P)-A, Revision 0, Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Designs," December 2011.

4. BAW-10240(P)-A, Revision 0, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," May 2004.

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Serial Number 15-438 Docket Nos. 50-280/28 1 Attachment 2

5. BAW-10084(P)-A, Revision 3, "Program to Determine In-Reactor Performance of BWFC Fuel Cladding Creep Collapse," July 1995.

6. BAW-10231(P)-A, Revision 1, "COPERNIC Fuel Rod Design Computer Code,"

January 2004.

7. WCAP-8762(P)-A, Revision 0, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," July 1984.

8. NRC Correspondence, "Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Request for Technical Specification Revisions Related to the Core Operating Limits Report (TAC Nos. ME2591 and ME2592)," dated October 19, 2010.

9. NRC Correspondence, "Millstone Power Station, Unit No. 3, Response to Request for Additional Information Regarding Exemption to Use a Low Tin Cladding (TAC No. MB9897)," dated November 10, 2003.

10.NRC Correspondence, "Surry Unit 1 and 2, Issuance of Amendments RE:

Increased Enrichment of Reload Fuel (TAC Nos. MA0122 and MA0123)," dated June 19, 1998.

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