View source for L-17-111, License Amendment Request to Modify Technical Specifications 4.2.1 and 5.6.3 and a 10 CFR 50.12 Exemption Request to Implement Optimized Zirlo Fuel Rod Cladding

{{Adams
| number = ML17100A269
| issue date = 04/09/2017
| title = License Amendment Request to Modify Technical Specifications 4.2.1 and 5.6.3 and a 10 CFR 50.12 Exemption Request to Implement Optimized Zirlo Fuel Rod Cladding
| author name = Richey M
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000334, 05000412
| license number = DPR-066, NPF-073
| contact person = 
| case reference number = L-17-111
| document type = Letter type:L, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 38
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:FENOC'             2' Beaver Valley Power Station P.O. Box 4 RrstEnergy Nuclear Operating Company                                                     Shippingport, PA 15077 Marty L. Richey                                                                                 724-682-5234 Site Vice President                                                                       Fax: 724-643-8069 April 9, 2017 L-17-111                                                       10 CFR 50.90 10 CFR 50.12 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==SUBJECT:==

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 License Amendment Request to Modify Technical Specifications 4.2.1 and 5.6.3 and a 10 CFR 50.12 Exemption Request to Implement Optimized ZIRLO' Fuel Rod Cladding Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) is requesting an amendment to both the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS) facility operating licenses. The proposed amendment would modify the BVPS technical specifications (TS) to allow the use of Optimized ZIRLO' as an approved fuel rod cladding material. This change is consistent with the Nuclear Regulatory Commission (NRC) approved use of Optimized ZIRLO' fuel rod cladding material as documented in the safety evaluation included in Addendum 1-A to Westinghouse Electric Company LLC topical report WCAP-12610-P-A & CENPD-404-P-A, "Optimized ZIRLO'."
To support this amendment, FENOC is requesting an exemption from certain requirements of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models," in accordance with 10 CFR 50.12, "Specific exemptions." This exemption request relates solely to the specific type of cladding material specified in these regulations for use in light water reactors. As written, the regulations presume use of either Zircaloy or ZIRLO fuel rod cladding. The exemption is required since Optimized ZIRLO' has a slightly different composition than Zircaloy or ZIRLO.
An evaluation of the request for licensing action is provided in Enclosure A, while the exemption request is provided in Enclosure B. FENOC is requesting NRC staff approval by April 11, 2018, which will support a spring outage at BVPS, Unit No. 1. Implementation of the amendment by FENOC is planned within 90 days of its approval.

Beaver Valley Power Station, Unit Nos. 1 and 2 L-17-111 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.
I declare under penalty of perjury that the foregoing is true and correct. Executed on April~, 2017.

==Enclosures:==

A. Evaluation of a Request for Licensing Action B. Exemption Request cc:  NRC Region I Administrator NRC BVPS Resident Inspector NRC BVPS Project Manager Director, BRP/DEP Site Representative, BRP/DEP

Enclosure A L-17-111 Evaluation of a Request for Licensing Action (Twenty-eight pages follow)

Evaluation of a Request for Licensing Action Page 1 of 13

==Subject:==
Request for licensing action to revise Beaver Valley Power Station, Unit Nos. 1 and 2, Technical Specifications 4.2.1, "Fuel Assemblies," and 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," to incorporate the use of Optimized ZIRLO '
fuel rod cladding material.
1.0    

==SUMMARY==
DESCRIPTION 2.0    DETAILED DESCRIPTION

==3.0    TECHNICAL EVALUATION==

==4.0    REGULATORY EVALUATION==

4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 4.4 Conclusions

==5.0     ENVIRONMENTAL CONSIDERATION==

==6.0    REFERENCES==

Attachments:
: 1. Beaver Valley Power Station, Unit Nos. 1 and 2, Proposed Technical Specification Changes (Mark-up)
: 2. Beaver Valley Power Station, Unit Nos. 1 and 2, Proposed Technical Specification Changes (Re-typed - For Information Only)

Page 2 of 13 1.0 

==SUMMARY==
DESCRIPTION This evaluation supports a request to amend Operating License Nos. DPR-66 and NPF-73 for Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS).
The proposed amendment would revise the BVPS Technical Specifications (TS) to allow the use of Optimized ZIRLO ' fuel rod cladding material. The current acceptable fuel rod cladding material is identified in TS 4.2.1, "Fuel Assemblies." The proposed amendment would revise TS 4.2.1 to add Optimized ZIRLO ' to the approved fuel rod cladding materials, to correct the spelling of the word Zircaloy, to add the word "clad" after the phrase "Optimized ZIRLO'," and to add a registered trademark designator to the word ZIRLO. TS 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," would be revised to add Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ," to the list of analytical methods used to determine the core operating limits previously reviewed and approved by the Nuclear Regulatory Commission (NRC).
Additionally, in order to use Optimized ZIRLO ' fuel rod cladding material in future core reload applications for BVPS, an exemption from the provisions of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models," is required. An exemption request pursuant to 10 CFR 50.12, "Specific exemptions," accompanies this submittal.
Optimized ZIRLO ' was developed to meet the needs of longer operating cycles with increased fuel discharge burnup and fuel duty. Fuel rod internal pressure (resulting from the increased fuel duty, use of integral fuel burnable absorbers, and corrosion and temperature feedback effects) has become more limiting with respect to fuel rod design criteria. Reducing the associated corrosion buildup and thus minimizing temperature feedback effects provides additional margin to the fuel rod internal pressure design criterion. Compared to ZIRLO, the lower tin content and microstructure difference of Optimized ZIRLO ' provides a reduced corrosion rate while maintaining the benefits of mechanical strength and resistance to accelerated corrosion from abnormal chemistry conditions.
2.0 DETAILED DESCRIPTION The following revisions are proposed for the BVPS TS. Revise TS 4.2.1, "Fuel Assemblies," to include Optimized ZIRLO ' , correct the spelling of the word Zircaloy, add the word "clad" after the phrase "Optimized ZIRLO ' ," and add a registered trademark designator to the word ZIRLO.

Page 3 of 13 The revised TS wording, with changes in bold, follows:
The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO ' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
Revise TS 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," to add Westinghouse topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ," to the list of analytical methods.
The revised TS wording, with changes in bold, follows:
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
WCAP-8745-P-A, "Design Bases for the Thermal Overtemperature t:.T and Thermal Overpower t:.T Trip Functions,"
WCAP-12945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis,"
(For Unit 1 only) WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification,"
WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"
WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

Page 4 of 13 WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,"
WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997 (Westinghouse Proprietary),
WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"
WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology,"
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ."

==3.0 TECHNICAL EVALUATION==

Westinghouse topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A "Optimized ZIRLO ' ," (Reference 1) provides the details and test results of Optimized ZIRLO ' compared to ZIRLO fuel rod cladding material. The topical report also contains the material properties to be used in various models and methodologies when analyzing Optimized ZIRLO ' fuel rod cladding. The NRC staff's safety evaluation (SE) for the Optimized ZIRLO TM topical report (Reference 2) contains ten conditions and limitations. FirstEnergy Nuclear Operating Company (FENOC) will comply with these conditions and limitations as follows:
: 1. Until rulemaking to 10 CFR Part 50 addressing Optimized ZIRLO ' has been completed, implementation of Optimized ZIRLO ' fuel clad requires an exemption from 10 CFR 50.46 and 10 CFR Part 50, Appendix K.
RESPONSE: A request for the required exemption from 10 CFR 50.46 and 10 CFR 50, Appendix K accompanies this submittal.
: 2. The fuel rod burnup limit for this approval remains at currently established limits:
62 [gigawatt days per metric tonne of uranium (GWd/MTU)] GWd/MTU for Westinghouse fuel designs and 60 GWd/MTU for [Combustion Engineering (CE)]
CE fuel designs.
RESPONSE: For any fuel using Optimized ZIRLO ' fuel rod cladding, the maximum fuel rod burnup limit for Westinghouse fuel designs will continue to be 62 GWd/MTU until such time that a new fuel rod burnup limit is approved for use.
BVPS only uses Westinghouse fuel designs. The fuel burnup limit will be confirmed as part of the normal reload design process.

Page 5 of 13
: 3. The maximum fuel rod waterside corrosion, as predicted by the best-estimate model, will [satisfy proprietary limits (included in topical report and proprietary version of the SE)] of hydrides for all locations of the fuel rod.
RESPONSE: The maximum fuel rod waterside corrosion for fuel using Optimized ZIRLO ' fuel rod cladding will be confirmed to be less than the specified proprietary limits for all locations of the fuel rod. Evaluations are performed to confirm that the appropriate corrosion limits are satisfied as part of the normal reload design process.
: 4. All the conditions listed in previous NRC SE approvals for methodologies used for standard ZIRLO and Zircaloy-4 fuel analysis will continue to be met, except that the use of Optimized ZIRLO ' cladding in addition to standard ZIRLO and Zircaloy-4 cladding is now approved.
RESPONSE: The Optimized ZIRLO ' fuel rod analysis will continue to meet all conditions associated with approved methods. Confirmation of these conditions is required as part of the normal reload design process.
: 5. All methodologies will be used only within the range for which ZIRLO and Optimized ZIRLO ' data were acceptable and for which the verifications discussed in Addendum 1 and responses to [requests for additional information (RAls)] RAls were performed.
RESPONSE: The application of ZIRLO and Optimized ZIRLO ' in approved methodologies will be made consistent with the approach accepted in WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ,"
July 2006. Confirmation of these conditions is required as part of the normal reload design process.
: 6. The licensee is required to ensure that Westinghouse has fulfilled the following commitment: Westinghouse shall provide the NRC staff with a letter(s) containing the following information (based on the schedule described in response to RAI #3, Reference 3):
: a. Optimized ZIRLO ' [lead test assembly (LTA)] LTA data from Byron, Calvert Cliffs, Catawba, and Millstone.
: i. Visual ii. Oxidation of fuel rods iii. Profilometry iv. Fuel rod length
: v. Fuel assembly length

Page 6 of 13
: b. Using the standard and Optimized ZIRLO ' database including the most recent LTA data, confirm applicability with currently approved fuel performance models (e.g., measured versus predicted).
Confirmation of the approved models' applicability up through the projected end of cycle burnup for the Optimized ZIRLO ' fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLO ' , sufficient LTA data may only be available to confirm the models' applicability up through 45 GWd/MTU. In this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLO ' fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations, since sufficient LTA data up through the burnup limit should be available within a few years.
RESPONSE: Westinghouse has provided the NRC with information related to test data and models. As a result, the NRC has confirmed that this condition has been satisfied as stated in letter dated August 3, 2016 (Reference 4). No further information is necessary in response to this condition.
: 7. The licensee is required to ensure that Westinghouse has fulfilled the following commitment: Westinghouse shall provide the NRC staff with a letter containing the following information (based on the schedule in response to RAI #11, Reference 3):
: a. Vogtle growth and creep data summary reports.
: b. Using the standard ZIRLO and Optimized ZIRLO ' database including the most recent Vogtle data, confirm applicability with currently approved fuel performance models (e.g., level of conservatism in Westinghouse rod pressure analysis, measured versus predicted, predicted minus measured versus tensile and compressive stress).
Confirmation of the approved models' applicability up through the projected end of cycle burnup for the Optimized ZIRLO ' fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLO TM, sufficient LTA data may only be available to confirm the models' applicability up through 45 GWd/MTU. In

Page 7 of 13 this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLO ' fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations since sufficient LTA data up through the burnup limit should be available within a few years.
RESPONSE: Westinghouse has provided the NRC with information related to test data and models. As a result, the NRC has confirmed that this condition has been satisfied as stated in letter dated August 3, 2016 (Reference 4). No further information is necessary in response to this condition.
: 8. The licensee shall account for the relative differences in unirradiated strength (yield strength (YS) and ultimate tensile strength (UTS)] YS and UTS between Optimized ZIRLO ' and standard ZIRLO in cladding and structural analyses until irradiated data for Optimized ZIRLO ' has been collected and provided to the NRC staff.
: a. For the Westinghouse fuel design analyses:
: i. The measured, unirradiated Optimized ZIRLO ' strengths shall be used for
[beginning of life (BOL)] BOL analyses.
ii. Between BOL up to a radiation fluence of 3.0 X 1021 [neutrons/centimeter2 (n/cm2)] n/cm2 (E > 1 (Million electron volts (MeV)] MeV), pseudo-irradiated Optimized ZIRLO ' strength set equal to linear interpolation between the following two strength level points: at zero fluence, strength of Optimized ZIRLO ' equal to measured strength of Optimized ZIRLO ' and at a fluence of 3.0 X 1021 n/cm2 (E > 1 MeV), irradiated strength of standard ZIRLO at the fluence of 3.0 X 1021 n/cm2 (E > 1 MeV) minus 3 ksi.
iii. During subsequent irradiation from 3.0 X 1021 n/cm2 up to 12 X 1021 n/cm2, the differences in strength (the difference at a fluence of 3.0 X 1021 n/cm2 due to tin content) shall be decreased linearly such that the pseudo irradiated Optimized ZIRLO ' strengths will saturate at the same properties as standard ZIRLO at 12 X 1021 n/cm2.
: b. For the CE fuel design analyses, the measured, unirradiated Optimized ZIRLO ' strengths shall be used for all fluence levels (consistent with previously approved methods).
RESPONSE: The Optimized ZIRLO ' fuel rod analysis for BVPS will use the yield strength and ultimate tensile strength as modified per conditions and limitations 8.a.i., 8.a.ii., and a.a.iii. Confirmation of this condition is required as part of the BVPS reload design process. BVPS uses a Westinghouse fuel design and not a CE fuel design, therefore, Condition 8.b does not apply.

Page 8 of 13
: 9. As discussed in response to RAI #21 (Reference 3), for plants introducing Optimized ZIRLO ' that are licensed with LOCBART or STRIKIN-11 and have a limiting [peak cladding temperature (PCT)] PCT that occurs during blowdown or early reflood, the limiting LOCBART or STRIKIN-11 calculation will be rerun using the specified Optimized ZIRLO ' material properties. Although not a condition of approval, the NRC staff strongly recommends that, for future evaluations, Westinghouse update all computer models with Optimized ZIRLO ' specific material properties.
RESPONSE: BVPS is not licensed to use the LOCBART or STRIKIN-11 codes.
Therefore, this condition does not apply.
: 10. Due to the absence of high temperature oxidation data for Optimized ZIRLO ' , the Westinghouse coolability limit on PCT during the locked rotor event shall be
[proprietary limits included in topical report and proprietary version of safety evaluation].
RESPONSE: Confirmation of this condition is required as part of the normal reload design process.

==4.0 REGULATORY EVALUATION==

FirstEnergy Nuclear Operating Company (FENOC) proposes to revise the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS) Technical Specifications (TS) to allow the use of Optimized ZIRLO ' fuel rod cladding material. The current acceptable fuel rod cladding material is identified in TS 4.2.1, "Fuel Assemblies." The proposed amendment would revise TS 4.2.1 to add Optimized ZIRLO ' to the approved fuel rod cladding materials, correct the spelling of the word Zircaloy, add a clarification to TS 4.2.1 by adding the word "clad" after the phrase "Optimized ZIRLO'," and add a registered trademark designator to the word ZIRLO. TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," is revised to add Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ," to the analytical methods used to determine the core operating limits previously reviewed and approved by the Nuclear Regulatory Commission (NRC).
4.1 Significant Hazards Consideration FENOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below.

Page 9 of 13
: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would allow the use of Optimized ZIRLO ' clad nuclear fuel at BVPS. The NRC approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO TM ," prepared by Westinghouse Electric Company LLC (Westinghouse), which addresses Optimized ZIRLO ' fuel rod cladding and demonstrates that Optimized ZIRLO ' fuel rod cladding has essentially the same properties as currently licensed ZIRLO fuel rod cladding. The use of Optimized ZIRLO ' fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The fuel cladding itself is not an accident initiator and does not affect accident probability. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes and do not affect the fuel cladding design. Use of Optimized ZIRLO' meets the fuel design acceptance criteria and hence does not significantly affect the consequences of an accident.
Therefore, the proposed amendment 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 kind of accident from any accident previously evaluated?
Response: No.
The use of Optimized ZIRLO ' fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes and do not affect the fuel cladding design. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A demonstrated that the material properties of Optimized ZIRLO ' fuel rod cladding are similar to those of ZIRLO fuel rod cladding. Therefore, Optimized ZIRLO ' fuel rod cladding will perform similarly to ZIRLO fuel rod cladding, thus precluding the possibility of the fuel rod cladding becoming an accident initiator and causing a new or different kind of accident.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 10 of 13
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment will not involve a significant reduction in the margin of safety.
NRG-approved Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of the Optimized ZIRLO ' fuel rod cladding are similar to those of ZIRLO fuel rod cladding. Optimized ZIRLO ' fuel rod cladding is expected to perform similarly to ZIRLO fuel rod cladding for normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The use of Optimized ZIRLO ' fuel rod cladding will not result in adverse changes to the operation or configuration of the facility. The correction of a typographical error, the addition of a word for clarification of the TS, and the addition of a registered trademark designator are administration changes that do not affect the fuel cladding design.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Based on the above, FENOC concludes that the proposed amendment 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.
4.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c) requires that TS, in part, include design features; and administrative controls. The amendment proposes revisions to the TS design features and the administrative controls sections.
10 CFR Part 50, Appendix A, General Design Criterion (GDC) 10, "Reactor design,"
requires that the "reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences." Specified acceptable fuel design limits are established to ensure that the fuel is not damaged, that is, the fuel rods do not fail and the fuel system dimensions remain within operational tolerances.
GDC 35, "Emergency core cooling," requires, in part, that a "system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts."

Page 11 of 13 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP), Section 4.2, "Fuel System Design," provides regulatory guidance, in part, to the NRC staff for the review of fuel rod cladding materials and fuel system. In addition, the SRP provides guidance for compliance with the applicable GDCs. According to SRP Section 4.2, the fuel system safety review provides assurance that:
* The fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs),
* Fuel system damage is never so severe as to prevent control rod insertion when it is required,
* The number of fuel rod failures is not underestimated for postulated accidents, and
* Coolability is always maintained.
By letter dated June 10, 2005 (Reference 2), the NRC staff issued a safety evaluation approving Addendum 1 to Westinghouse topical report WCAP-12610-P-A & CENPD-404-P-A, "Optimized ZIRLO', " wherein the NRC staff approved the use of Optimized ZIRLO' as an acceptable fuel rod cladding material for Westinghouse and Combustion Engineering fuel designs.
10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," provides the acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors. 10 CFR 50, Appendix K, "ECCS Evaluation Models," establishes the required and acceptable features of the ECCS evaluation models.
FENOC evaluated the proposed amendment with respect to the aforementioned regulations and determined that to use Optimized ZIRLO ' fuel rod cladding material requires an exemption from 10 CFR 50.46 and 10 CFR 50, Appendix K. An exemption request accompanies this submittal and provides the basis and justification for exemption from these regulations.
4.3 Precedent Several licensees; including Seabrook Station, Unit No. 1 (Reference 5); have requested license amendments to incorporate the use of Optimized ZIRLO' fuel rod cladding that have been approved by the NRC. The proposed amendment for BVPS is similar to the Seabrook Station, Unit No. 1 amendment in that it requests to incorporate the use of Optimized ZIRLO' fuel rod cladding. The differences between the two amendments appear to be administrative in nature, in that in the Seabrook Station, Unit No.1 COLR contains the approval dates of the listed methodologies, while the BVPS COLR does not.

Page 12 of 13 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVI RONMENTAL CONSIDERATION The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion 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 amendment.

==6.0 REFERENCES==
: 1. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ,"
July 2006 (Non-Proprietary Version stored as WCAP-14432-A & CENPD-404-NP-A, Addendum 1-A) (Accession No. ML062080569).
: 2. Letter from H. N. Berkow (USNRC) to J. A. Gresham (Westinghouse), "Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A & CENPD-404-P-A,
    'Optimized ZIRLO ', '" June 10, 2005 (Accession No. ML051670403).
: 3. Letter from J. A. Gresham (Westinghouse) to USNRC, "Westinghouse Responses to NRC Request for Additional Information (RAls) on Optimized ZIRLO ' Topical -
Addendum 1 to WCAP-12610-P-A and CENPD-404-P-A (Non-Proprietary), TAC No.
MB8041," LTR-NRC-05-26, May 18, 2005 (Accession No. ML051440548).
: 4. Letter from K. Hsueh (USNRC) to J. A. Gresham (Westinghouse), "Satisfaction of Conditions 6 and 7 of the U. S. Nuclear Regulatory Commission Safety Evaluation for Westinghouse Electric Company Addendum 1 to WCAP-12610-P-A & CENP-404-P-A, 'Optimized ZIRLO ' ,' Topical Report," August 3, 2016 (Accession No. ML16173A354).

Page 13 of 13
: 5. Letter from J. G. Lamb, USNRC, to K. Walsh, NextEra Energy, "Seabrook Station, Unit No. 1 - Issuance of Amendment Regarding the Use of Optimized ZIRLO ' Fuel Rod Cladding Material (TAC NO. MF2410)," March 5, 2014 (Accession No. ML13213A143).

Attachment 1 Beaver Valley Power Station, Unit Nos. 1 and 2, Proposed Technical Specification Changes (Mark-up)
(Four pages follow)

(3)      FENOC, pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)      FENOC, pursuant to the Act and 1 0 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)      FENOC, pursuant to the Act and 1 O CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 1 0 CFR Chapter 1 :
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)      Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)      Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. . are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)      Auxiliary River Water System (Deleted by Amendment No. 8)
Amendment No.
Beaver Valley Unit 1                                                Renewed Operating License DPR-66

(b)       Further, the licensees are also required to notify the N RC in writing prior to any change in: (i) the term or conditions of any lease agreements executed as part of these transactions; (ii) the BVPS Operating Agreement, (iii) the existing property insurance coverage for BVPS Unit 2, and (iv) any action by a lessor or others that may have adverse effect on the safe operation of the facility.
C.      This renewed operating license shall be deemed t o contain and i s subject to the conditions specified in the following Commission regulations set forth in 1 O CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)     Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)     Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 4--85, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. 4--85 Beaver Valley Unit 2                                             Renewed Operating License NPF-73

Design Features 4.0 4.0   DESIGN FEATURES
: 4. 1  Site Location The Beaver Valley Power Station is located in Shippingport Borough, Beaver County, Pennsylvania, on the south bank of the Ohio River. The site is approximately 1 mile southeast of Midland, Pennsylvania, 5 miles east of East Liverpool, Ohio, and approximately 25 miles northwest of Pittsburgh, Pennsylvania. The Unit 1 exclusion area boundary has a minimum radius of 2000 feet from the center of containment. The Unit 2 exclusion area boundary has a minimum radius of 2000 feet around the Unit No. 1 containment building.
4.2   Reactor Core Zircaloy, ZIRLo*, or Optimized ZIRLOTM clad 4.2 . 1 Fuel Assemblies 1 57 fuel assemblies. Each assembly shall consist of a matrix of
* fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable N RC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2   Control Rod Assemblies The reactor core shall contain 48 control rod assemblies. The control material shall be silver indium cadmium as approved by the N RC.
4.3   Fuel Storage 4.3. 1  Criticality 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 as specified in LCO 3.7. 1 4, "Spent Fuel Pool Storage,"
: b. Unit 1 Kett  0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9. 1 2 of the UFSAR, Beaver Valley Units 1 and 2                       4.0 - 1                      Amendments 278 I 161

Reporting Requirements WCAP-12610-P-A & CENPD-404-P-A, Addendum 1 -A, "Optimized ZIRLO'."                                     5.6 5.6    Reporting Requirements 5.6.3     CORE OPERATING LIMITS REPORT (COLR) (continued)
WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"
WCAP-1 6045-P-A, Addendum 1 -A, "Qualification of the N EXUS Nuclear Data Methodology.,.,"
As described in reference documents listed above, when an initial assumed power level of 1 02% of RATED THERMAL POWER is specified in a previously approved method, 1 00.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM) .
Caldon, Inc. Engineering Report-BOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM ..J TM System" Caldon, Inc. Engineering Report-1 60P, "Supplement to Topical Report ER-BOP: Basis for a Power Uprate with the LEFM ..J TM System"
: c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.4         Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and LCO 3.4. 1 2, "Overpressure Protection System (OPPS)"
: b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
NRC Letter, "Beaver Valley Power Station, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure and Temperature Limits Report (TAC Nos. MB33 1 9 and MB3320)," dated October 8, 2002.
Beaver Valley Units 1 and 2               5.6 - 3                       Amendments 291 ,' 178

Attachment 2 Beaver Valley Power Station, Unit Nos. 1 and 2, Proposed Technical Specification Changes (Re-typed - For Information Only)
(Nine pages follow)

jFor Information Only I (3)      FE NOC, pursuant to the Act and 1 0 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)      FENOC, pursuant to the Act and 1 0 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)      FENOC, pursuant to the Act and 1 0 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 1 0 CFR Chapter 1 :
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)      Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)      Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)      Auxiliary River Water System (Deleted by Amendment No. 8)
Amendment No.
Beaver Valley Unit 1                                                Renewed Operating License DPR-66

! For I nformation Only I (b)     Further, the licensees are also required to notify the NRC in writing prior to any change in: (i) the term or conditions of any lease agreements executed as part of these transactions; (ii) the BVPS Operating Agreement, (iii) the existing property insurance coverage for BVPS Unit 2, and (iv) any action by a lessor or others that may have adverse effect on the safe operation of the facility.
C.      This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations set forth in 1 O CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)     Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)     Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No.
Beaver Valley Unit 2                                           Renewed Operating License NPF-73

Design Features
! For Information Only I 4.0 4.0     DESIGN FEATURES 4.1     Site Location The Beaver Valley Power Station is located in Shippingport Borough, Beaver County, Pennsylvania, on the south bank of the Ohio River. The site is approximately 1 mile southeast of Midland, Pennsylvania, 5 miles east of East Liverpool, Ohio, and approximately 25 miles northwest of Pittsburgh, Pennsylvania. The Unit 1 exclusion area boundary has a minimum radius of 2000 feet from the center of containment. The Unit 2 exclusion area boundary has a minimum radius of 2000 feet around the Unit No. 1 containment building.
4.2     Reactor Core 4.2 . 1 Fuel Assemblies The reactor shall contain 1 57 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO ' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4 .2.2  Control Rod Assemblies The reactor core shall contain 48 control rod assemblies. The control material shall be silver indium cadmium as approved by the NRC.
4.3     Fuel Storage 4.3 . 1 Criticality 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 as specified in LCO 3.7. 1 4, "Spent Fuel Pool Storage,"
: b. Unit 1 l<eff :,;; 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9. 1 2 of the UFSAR, Beaver Valley Units 1 and 2                             4.0 - 1                          Amendments       I

Reporting Requirements
! For Information Only I                                                                         5.6 5.0     ADM INISTRATIVE CONTROLS 5.6      Reporting Requirements The following reports shall be submitted in accordance with 1 0 CFR 50.4.
5.6.1           Annual Radiological Environmental Operating Report
                                                          - NOTE -
A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.
The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 5 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I , Sections IV. B.2, IV. B.3, and IV.C.
5.6.2           Radioactive Effluent Release Report
                                                          - NOTE -
A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 1 0 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 1 0 CFR 50.36a and 1 0 CFR Part 50, Appendix I, Section IV. B. 1 .
5.6.3           CORE OPERATI NG LIM ITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
SL 2. 1 . 1 , "Reactor Core Safety Limits" LCO 3 . 1 . 1 , "SH UTDOWN MARG I N (SOM)"
LCO 3 . 1 .3, "Moderator Temperature Coefficient (MTC)"
Beaver Valley Units 1 and 2                      5.6 - 1                  Amendments 278 / 1 61

Reporting Requirements
! For I nformation Only l                                                                     5.6 5.6      Reporting Requirements 5.6. 3      CORE OPERATING LIM ITS REPORT (COLR) (continued)
LCO 3 . 1 .5, "Shutdown Bank Insertion Limits" LCO 3.1 .6, "Control Bank Insertion Limits" LCO 3.2. 1 , "Heat Flux Hot Channel Factor (Fa(Z))"
LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( FfH )"
LCO 3.2.3, "Axial Flux Difference (AFD)"
LCO 3.3. 1 , "Reactor Trip System (RTS) Instrumentation" - Overtemperature and Overpower !),,T Allowable Value parameter values LCO 3.4. 1 , "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" LCO 3.9. 1 , "Boron Concentration"
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
WCAP-8745-P-A, "Design Bases for the Thermal Overtemperature !),,T and Thermal Overpower !),,T Trip Functions,"
WCAP-1 2945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis,"
(For Unit 1 only) WCAP-1 6009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
WCAP-1 02 1 6-P-A, "Relaxation of Constant Axial Offset Control/
Fa Surveillance Technical Specification,"
WCAP-14565-P-A, "VI PRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"
WCAP-1 261 0-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"
WCAP-1 5025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 1 7x1 7 Rod Bundles with Modified LPD Mixing Vane Grids,"
WCAP-1 3749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1 997 (Westinghouse Proprietary),
Beaver Valley Units 1 and 2                     5.6 - 2                  Amendments 291 / 1 78

jFor Information Only I                                                          Reporting Requirements 5.6 5.6    Reporting Requirements 5.6.3     CORE OPERATING LIM ITS REPORT (COLR) (continued)
WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"
WCAP-1 6045-P-A, Addendum 1 -A, "Qualification of the NEXUS Nuclear Data Methodology,"
WCAP-1 261 0-P-A & CENPD-404-P-A, Addendum 1 -A, "Optimized ZI RLO ' . "
As described in reference documents listed above, when a n initial assumed power level of 1 02% of RATED THERMAL POWER is specified in a previously approved method, 1 00.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).
Caldon, Inc. Engineering Report-BOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM ...J TM System" Caldon, I nc. Engineering Report-1 60P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM ...J TM System"
: c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.4         Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and LCO 3.4. 1 2, "Overpressure Protection System (OPPS)"
: b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
Beaver Valley Units 1 and 2                   5.6 - 3                          Amendments        I

jFor Information Only I                                                            Reporting Requirements 5.6 5.6       Reporting Requirements 5.6.4        Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIM ITS REPORT (PTLR} (continued)
NRC Letter, "Beaver Valley Power Station, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure and Temperature Limits Report (TAC Nos. MB33 1 9 and M B3320)," dated October 8, 2002.
WCAP-1 4040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
The methodology listed in WCAP-1 4040-NP-A was used with two exceptions:
* ASM E Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1 ."
* ASME, Section XI , Appendix G , "Fracture Toughness Criteria for Protection Against Failure," 1 996 version.
: c. The PTLR shall be provided to the N RC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.5            Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 1 4 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.6            Steam Generator (SG) Tube Inspection Report 5.6.6 . 1        Unit 1 SG Tube I nspection Report A report shall be submitted within 1 80 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5. 1 , "Unit 1 SG Program." The report shall include:
: a. The scope of inspections performed on each SG,
: b. Degradation mechanisms fou nd,
: c. Nondestructive examination techniques utilized for each degradation mechanism,
: d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
: e. Number of tubes plugged during the inspection outage for each degradation mechanism ,
: f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and Beaver Valley Units 1 and 2                        5.6 - 4                           Amendments        /

! For Information Only I                                                          Reporting Requirements 5.6 5.6      Reporting Requirements 5.6.6       Steam Generator (SG) Tube Inspection Report (continued) 5.6.6. 1        Unit 1 SG Tube Inspection Report (continued)
: g. The results of condition monitoring, including the results of tube pulls and in-situ testing.
5.6.6.2         Unit 2 SG Tube Inspection Report
: 1. A report shall be submitted within 1 80 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, "Unit 2 SG Program." The report shall include:
: a. The scope of inspections performed on each SG,
: b. Degradation mechanisms found,
: c. Nondestructive examination techniques utilized for each degradation mechanism,
: d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications,
: e. Number of tubes plugged or repaired during the inspection outage for each degradation mechanism,
: f. The number and percentage of tubes plugged or repaired to date, and the effective plugging percentage in each steam generator,
: g. The results of condition monitoring , including the results of tube pulls and in-situ testing , and
: h. Repair method utilized and the number of tubes repaired by each repair method.
: 2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, "Unit 2 SG Program," when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."
: 3. For implementation of the voltage-based plugging or repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise:
Beaver Valley Units 1 and 2                      5.6 - 5                           Amendments       I

! For I nformation Only!                                                         Reporting Requirements 5.6 5.6     Reporting Requirements 5.6.6.2    Unit 2 SG Tube Inspection Report (continued)
: a. If circumferential crack-like indications are detected at the tube support plate intersections.
: b. If indications are identified that extend beyond the confines of the tube support plate.
: c. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
: 4. A report shall be submitted within 90 days after the initial entry into MODE 4 following an outage in which the F* methodology was applied.
As applicable, the report shall include the following hot-leg and cold-leg tubesheet region inspection results associated with the application of F*:
a . Total number of indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.
: b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.
: c. The projected end-of-cycle accident-induced leakage from tubesheet indications.
Beaver Valley Units 1 and 2                    5.6 - 6                            Amendments        /

Enclosure B L-17-111 Exemption Request (Six pages follow)

Exemption Request Page 1 of 6

==Subject:==
Request for exemption from certain requirements of 1 0 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 1 0 CFR 50, Appendix K, "ECCS Evaluation Models" to allow the use of Optimized ZI RLO ' fuel rod cladding material at Beaver Valley Power Station, Unit Nos. 1 and 2.
1 .0 PURPOSE

==2.0 BACKGROUND==

3.0 TECHNICAL JUSTIFICATION OF ACCEPTABILITY 4.0 JUSTI FICATION OF EXEMPTION

==5.0 CONCLUSION==

==6.0 REFERENCES==

Page 2 of 6 1.0 PURPOSE In accordance with 10 CFR 50.12, "Specific exemptions," FirstEnergy Nuclear Operating Corporation (FENOC) requests an exemption from the provisions of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models." The requested exemption would permit the use of Optimized ZIRLO ' fuel rod cladding material in future core reload applications for Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS). The regulations in 10 CFR 50.46 contain acceptance criteria for the emergency core cooling system (ECCS) for reactors that have fuel rods fabricated either with zircaloy or ZIRLO fuel rod cladding material. Concurrently, 10 CFR 50, Appendix K, paragraph I.A.5, requires the Baker-Just equation be used to calculate the rate of energy release, hydrogen generation, and cladding oxidation from the metal-water reaction in the core. The Baker-Just equation assumes the use of a zirconium alloy other than Optimized ZIRLO ' material. Therefore, an exemption is required from both 10 CFR 50.46 and 10 CFR 50, Appendix K to support the use of Optimized ZIRLO '
fuel rod cladding at BVPS. This exemption request relates solely to the specific cladding material identified in these regulations (fuel rods with zircaloy or ZIRLO cladding) and will provide for the application of 10 CFR 50.46 and 10 CFR 50, Appendix K acceptance criteria to fuel assembly designs utilizing Optimized ZIRLO ' fuel rod cladding at BVPS.

==2.0 BACKGROUND==

As the nuclear industry pursues longer operating cycles with increased fuel discharge burnup and fuel duty, the corrosion performance requirements for nuclear fuel cladding become more demanding. Optimized ZIRLO ' was developed to be more resistant to accelerated corrosion, due to abnormal chemistry conditions, than ZIRLO, while retaining its mechanical strength benefits. In addition, fuel rod internal pressure (resulting from the increased fuel duty, use of integral fuel burnable absorbers, and corrosion and temperature feedback effects) have become more limiting with respect to fuel rod design criteria. Reducing the associated corrosion buildup, and thus, minimizing the temperature feedback effects, provides additional margin to the fuel rod internal pressure design limit.
A Technical Specification (TS) amendment for BVPS is required to allow the use of Optimized ZIRLO ' fuel rod cladding. The amendment request accompanies this submittal.
3.0 TECHNICAL JUSTIFICATION OF ACCEPTABILITY Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-12610-P-A
& CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ," (Reference 1), provides the details and results of tests comparing Optimized ZIRLO ' and ZIRLO (as well as the material properties to be used in various models and methodologies when analyzing Optimized ZIRLO ' fuel rod cladding). The Nuclear Regulatory Commission (NRC)

Page 3 of 6 Safety Evaluation (SE) (Reference 2) for the topical report contains ten conditions and limitations. The first condition requires FENOC request an exemption from 10 CFR 50.46 and 10 CFR 50, Appendix K (which is being requested in this submittal) before implementing Optimized ZIRLO ' at BVPS. Westinghouse has provided the NRC with information related to test data and models (References 3 through 10) to address conditions and limitations 6, 7, and Ba. The NRC has confirmed by letter (Reference 11) that the data has satisfied conditions 6 and 7 and that no further information needs to be provided specific to conditions 6 and 7 when referencing WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A. Condition Ba must continue to be addressed as written in the NRC SE for WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A. Condition and limitation 9 is not applicable since BVPS is not licensed to use LOCBART or STRIKIN-11 codes. The remaining conditions and limitations are addressed with changes to the BVPS TS and the evaluations that are part of the BVPS reload design process. There are no additional commitments necessary to support NRC approval of this exemption request.
Future reload evaluations will ensure that acceptance criteria are met for the insertion of fuel assemblies composed of fuel rods clad with Optimized ZIRLO TM _ These fuel assemblies will be evaluated using NRG-approved methods and models.
4.0 JUSTIFICATION OF EXEMPTION 10 CFR 50.12, "Specific exemptions," states that the Commission may grant exemptions from the requirements of the regulations of this part provided two criteria are met.
These criteria are: (1) the exemption authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security; and (2) the Commission will not consider granting an exemption unless special circumstances are present.
The requested exemption to allow the use of Optimized ZIRLO ' fuel rod cladding material in addition to zircaloy or ZIRLO at BVPS satisfies these criteria as described below.
Criterion 1
: a. This exemption is authorized by law.
As required by 10 CFR 50.12 (a)(1), this requested exemption is "authorized by law." The selection of a specified cladding material in 1 O CFR 50.46 and implied in 10 CFR 50, Appendix K, was adopted at the discretion of the Commission consistent with its statutory authority. No statute required the NRC to adopt this specification. Additionally, the NRC has the authority under 10 CFR 50.12 to grant exemptions from the requirements of Part 50 upon showing proper justification. FENOC is not seeking an exemption from the acceptance and analytical criteria of 10 CFR 50.46 and 10 CFR 50, Appendix K. The intent of this request is solely to allow the use of criteria set forth in these regulations for application to the Optimized ZIRLO ' fuel rod cladding material.

Page 4 of 6
: b. This exemption will not present an undue risk to public health and safety.
The reload design process ensures that the acceptance criteria are met for the insertion of fuel assemblies with fuel rods clad with Optimized ZIRLO ' . Fuel assemblies using Optimized ZIRLO ' fuel rod cladding will be evaluated using NRG-approved analytical methods and plant-specific models to address the differences in the cladding material properties. Thus, the granting of this exemption request will not pose an undue risk to public health and safety.
: c. This exemption is consistent with the common defense and security.
As noted above, this exemption request is only to allow the application of the aforementioned regulations to an improved fuel rod cladding material. The requirements and acceptance criteria will be maintained. The special nuclear material in these assemblies is required to be handled and controlled in accordance with approved procedures. Use of Optimized ZIRLO ' fuel rod cladding at BVPS will not affect plant operations and is consistent with common defense and security.
Criterion 2 Special circumstances support the issuance of an exemption.
10 CFR 50.12(a)(2) states that the Commission will not consider granting an exemption to the regulations unless special circumstances are present. This exemption request meets the special circumstances criteria of 10 CFR 50.12(a)(2)(ii), "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule." For BVPS, application of the subject regulations is not necessary to achieve the underlying purpose of the rule.
10 CFR 50.46 identifies acceptance criteria for ECCS performance at nuclear power plants. As part of the BVPS reload design process, Westinghouse performs an evaluation of the core using loss-of-coolant accident (LOCA) methods approved for the site to ensure that fuel assemblies with Optimized ZIRLO ' fuel rod cladding meet the LOCA safety criteria.
The intent of 10 CFR 50, Appendix K, paragraph I.A.5 is to apply an equation for rate of energy release, hydrogen generation, and cladding oxidation from a metal-water reaction that conservatively bounds all post-LOCA scenarios (the Baker-Just equation).

Page 5 of 6 Application of the Baker-Just equation has been demonstrated to be appropriate for Optimized ZIRLO ' . Due to the similarities in material composition of the Optimized ZIRLO ' and ZIRLO fuel rod cladding, the application of the Baker-Just equation will continue to conservatively bound all post-LOCA scenarios.

==5.0 CONCLUSION==

The 10 CFR 50.46 and 10 CFR 50, Appendix K regulations are currently limited in applicability to the use of fuel rods with zircaloy or ZIRLO cladding. 10 CFR 50.46 and 10 CFR 50, Appendix K do not apply to the proposed use of Optimized ZIRLO ' fuel rod cladding material since Optimized ZIRLO ' has a slightly different composition than zircaloy or ZIRLO. With the approval of this exemption request, these regulations will be applied to Optimized ZIRLO ' fuel rod cladding at BVPS.
Pursuant to 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 the common defense and security. Approval of this exemption request does not violate the underlying purpose of the rule. In addition, special circumstances exist to justify the approval of an exemption from the subject requirements.
6.0. REFERENCES
: 1. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO ' ,"
July 2006 (Non-Proprietary Version stored as WCAP-14432-A & CENPD-404-NP-A, Addendum 1-A) (Accession No. ML062080569).
: 2. Letter from H. N. Berkow (USNRC) to J. A. Gresham (Westinghouse), "Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A & CENPD-404-P-A,
    'Optimized ZIRLO ' ,"' June 10, 2005 (Accession No. ML051670403).
: 3. Letter from J. A. Gresham (Westinghouse) to USNRC, "SER Compliance with WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," LTR-NRC-07-1, January 4, 2007 (Accession No. ML070100388).
: 4. Letter from J. A. Gresham (Westinghouse) to USNRC, "SER Compliance with WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," LTR-NRC-07-58, November 6, 2007 (Accession No. ML073130560).

Page 6 of 6
: 5. Letter from J. A Gresham (Westinghouse) to USNRC, "SER Compliance with WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," LTR-NRC-07-58, Revision 1, February 5, 2008 (Accession No. ML080390452).
: 6. Letter from J. A Gresham (Westinghouse) to USNRC, "SER Compliance of WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," L TR-NRC-08-60, December 30, 2008 (Accession No. ML090080381).
: 7. Letter from J. A Gresham (Westinghouse) to USNRC, "SER Compliance of WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," LTR-NRC-10-43, July 26, 2010 (Accession No. ML102140214).
: 8. Letter from J. A Gresham (Westinghouse) to USNRC, "SER Compliance of WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A 'Optimized ZIRLO ' '
(Proprietary/Non-proprietary)," LTR-NRC-13-6, February 25, 2013 (Accession No. ML13070A189).
: 9. Letter from J. A Gresham (Westinghouse) to USNRC (Document Control Desk),
   "Submittal of Responses to Draft RAls and Revisions to Select Figures in LTR-NRC-13-6 to Fulfill Conditions 6 and 7 of the Safety Evaluation for WCAP-12610-P-A &
CENPD-404-P-A Addendum 1-A (Proprietary/Non-Proprietary)," LTR-NRC-15-7, February 9, 2015 (Accession No. ML15051A427).
: 10. Letter from J. A Gresham (Westinghouse) to USNRC (Document Control Desk),
    "Submittal of Response to Condition 8.a of the Safety Evaluation Report (SER) on WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, 'Optimized ZIRLO ' '
(Proprietary/Non-Proprietary)," LTR-NRC-15-84, September 29, 2015 (Accession No. ML15279A114).
: 11. Letter from K. Hsueh (USN RC) to J. A Gresham (Westinghouse), "Satisfaction of Conditions 6 and 7 of the U. S. Nuclear Regulatory Commission Safety Evaluation for Westinghouse Electric Company Addendum 1 to WCAP-12610-P-A & CENP-404-P-A, 'Optimized ZIRLO ' ,' Topical Report," August 3, 2016 (Accession No. ML16173A354).}}