L-22-193, Request for Exemption from Specific Provisions in 10 CFR 50 Appendix H

Request for Exemption from Specific Provisions in 10 CFR 50 Appendix H

ML23045A144
Person / Time
Site:	Beaver Valley
Issue date:	02/14/2023
From:	Blair B Energy Harbor Nuclear Corp
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-22-193
Download: ML23045A144 (1)

Text

energy Energy Harbor Nuclear Corp.

Beaver Valley Power Station harbor P.O. Box 4 Shippingport, PA 15077 Barry N. Blair 440-280-7300 Site Vice President, Beaver Valley Nuclear February 14, 2023 L-22-193 10 CFR 50, Appendix H 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 Request for Exemption from Specific Provisions in 10 CFR 50 Appendix H Pursuant to 10 CFR 50.12, "Specific exemptions," Energy Harbor Nuclear Corp. is requesting a permanent exemption from a requirement of Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements," for Beaver Valley Power Station, Unit No. 1 (BVPS-1). Appendix H,Section IV.A requires, in part, that each capsule withdrawal and associated test results must be the subject of a summary technical report that is to be submitted to the Nuclear Regulatory Commission (NRC) within eighteen months of the date of capsule withdrawal, unless an extension is granted by the Director, Office of Nuclear Reactor Regulation.

An exemption to this requirement is requested for capsule A, which is a supplemental capsule that was fabricated and installed in response to pressurized thermal shock (PTS) concerns related to the BVPS-1 reactor vessel materials. Since the PTS concern was identified in 1996, additional evaluation has identified that the limiting BVPS-1 reactor vessel beltline material is not as brittle as originally anticipated. Because capsule A was developed to alleviate a PTS concern that is no longer applicable to BVPS-1, Energy Harbor Nuclear Corp. is requesting NRC approval of an exemption from the requirement to submit a technical report documenting the capsule testing results within eighteen months of the capsule withdrawal that is required by 10 CFR 50, Appendix H, Section IV.A.

Energy Harbor Nuclear Corp. is requesting approval of the attached exemption by February 29, 2024, since surveillance capsule A is scheduled to be withdrawn from the reactor vessel during the fall refueling outage 2024.

Beaver Valley Power Station, Unit Nos. 1 and2 L-22-193 Page2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Manager - Fleet L icensing, at (330) 696-7208.

Barry N. Blair

Attachment:

Exemption Request cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment L-22-193 Exemption Request Page 1 of 13

Subject:

Request for Exemption from Specific Provisions in 10 CFR 50, Appendix H, at Beaver Valley Power Station, Unit Nos. 1 and 2 TABLE OF CONTENTS 1.0 PURPOSE

2.0 BACKGROUND

3.0 PROPOSED EXEMPTION 4.0 JUSTIFICATION OF EXEMPTION 5.0 ENVIRONMENTAL ASSESSMENT

7.0 REFERENCES

Attachment L-22-193 Page 2 of 13 1.0 PURPOSE Pursuant to 10 CFR 50.12, "Specific exemptions," Energy Harbor Nuclear Corp. is requesting a permanent exemption from a requirement of Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements," for Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2).

2.0 BACKGROUND

10 CFR 50, Appendix H, Section IV.A states:

Each capsule withdrawal and the test results must be the subject of a summary technical report to be submitted, as specified in § 50.4, within eighteen months of the date of capsule withdrawal, unless an extension is granted by the Director, Office of Nuclear Reactor Regulation.

In 1996, a pressurized thermal shock (PTS) evaluation of the BVPS-1 beltline plate material was performed that identified that the end of license BVPS-1 reference temperature for pressurized thermal shock (RTPTS) values were approaching the current PTS screening criteria identified in 10 CFR 50.61. As a result of this evaluation, the decision was made to broaden the fracture toughness data set to be used in future surveillance evaluations. This increase in the data set was accomplished by assembling surveillance capsule A, which is comprised of previously irradiated BVPS-1 surveillance materials, unirradiated BVPS-1 surveillance materials, and materials from the Fort Calhoun and St. Lucie surveillance programs.

Because the capsule lead factors are significantly higher in the Beaver Valley Power Station, Unit No. 2 (BVPS-2) reactor vessel than they are in the BVPS-1 reactor vessel, and since the inlet temperature is approximately the same at BVPS-2 as it is at BVPS-1, supplemental capsule A was installed in the BVPS-2 reactor vessel so that end of license extension fluence values would be accrued more efficiently. As a result, the withdrawal schedule for BVPS-1 surveillance capsule A is tracked on the BVPS-2 surveillance capsule withdrawal schedule.

The current BVPS-2 surveillance capsule withdrawal schedule was approved by the NRC in a safety evaluation dated July 17, 2014 (ML13242A266).

Since the BVPS-1 PTS concern was identified in 1996, additional evaluation identified that the limiting BVPS-1 reactor vessel beltline material is not as brittle as originally anticipated. In a safety evaluation dated July 2, 2018 (ML18164A082), the NRC approved revised RTPTS values for BVPS-1. The revised RTPTS values were calculated using revised unirradiated nil-ductility reference temperature (RTNDT) values that were calculated in accordance with American Society of Mechanical Engineers (ASME) code, Subarticle NB-2331, paragraph (a)(4). The revised BVPS-1 RTPTS

Attachment L-22-193 Page 3 of 13 values meet the PTS screening criteria identified in 10 CFR 50.61 through the end of the renewed operating license.

3.0 PROPOSED EXEMPTION Energy Harbor Nuclear Corp. proposes a permanent exemption to the 10 CFR 50, Appendix H, Section IV.A, requirement to submit a summary technical report, regarding capsule withdrawal and capsule test results, to the NRC within eighteen months of withdrawal for capsule A.

Capsule A will be disassembled, and the neutron dosimeters will be tested within one year after the capsule withdrawal to ensure that valid dosimetry measurements can be obtained prior to excessive radioactive decay of the dosimeters.

The capsule contents will be inventoried and placed in storage so that they are retrievable for future testing if the industry determines that it is necessary to support the coordinated pressurized water reactor (PWR) reactor vessel surveillance program guidelines identified in MRP-326, or if it is necessary to justify supplemental license renewals for BVPS-1.

Mechanical testing of capsule A will not be performed.

4.0 JUSTIFICATION OF EXEMPTION Technical discussion:

Compliance with the requirements of 10 CFR 50, Appendix H; 10 CFR 50, Appendix G; and 10 CFR 50.61 for the BVPS-1 60-year life were evaluated as described below.

The underlying purpose of 10 CFR 50 Appendix H is to monitor changes in the fracture toughness properties of ferritic materials from exposure to neutron irradiation and the thermal environment in the reactor vessel beltline region of light-water nuclear power reactors. The fracture toughness data obtained from the material surveillance program is subsequently used to assess the integrity of the reactor vessel.

With respect to the reactor vessel surveillance program, 10 CFR 50 Appendix H.III.B requires in part that the design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of the American Society for Testing and Materials (ASTM) E185 that is current on the issue date of the ASME code to which the reactor vessel was purchased. For reactor vessels purchased in or before 1982, later editions of ASTM E185 may be used, but including only those editions through 1982. This requirement is applicable to the original 40-year license period for BVPS-1.

For the period of extended operation,Section XI.M31 of NUREG-1801, Revision 2,

Attachment L-22-193 Page 4 of 13 provides guidance regarding reactor vessel surveillance programs, including recommended capsule withdrawal schedules.

Supplemental capsule A is not required to meet the requirements of ASTM E185-82 for the original 40-year license period nor NUREG-1801 Section XI.M31 for the period of extended operation.

For BVPS-1, Westinghouse Report WCAP-18102-NP, Revision 2, identifies that the change in the reference temperature of the limiting material at 50 effective full power years (EFPY) (60-year end-of-life) is greater than 200 degrees Fahrenheit. For reactor vessels that are in this category, ASTM E185-82 recommends that for the original 40-year plant license term, a minimum of five capsules are to be withdrawn, with the final capsule having a fluence of between one and two times the projected end-of-life fluence of the reactor vessel.

Five of the eight BVPS-1 reactor vessel surveillance capsules (V, U, W, Y, and X) have been withdrawn and tested. Capsule X was withdrawn at the end of cycle 22, meeting the surveillance capsule withdrawal requirements for the original 40-year end of life that are defined in the 1982 edition of ASTM E185. The most recently withdrawn surveillance capsule (capsule X) was irradiated to a fluence level approximately equal to 42.24 EFPY of plant operation (linearly interpolated from WCAP-18102 Table 2-5).

A sixth surveillance capsule (capsule Z) is scheduled to be withdrawn after 39 EFPY, which corresponds to the peak vessel fluence at the end of the 60-year license renewal period, 5.89 x 1019 n/cm2 (E > 1.0 MeV). This meets the end-of-life capsule withdrawal requirements contained in NUREG-1801,Section XI.M31.

Even without the mechanical testing of surveillance capsule A, the neutron fluence exposure of the BVPS-1 reactor vessel will be bounded by the fluence level of the most recently tested surveillance capsule. The current analysis of record for BVPS-1 reactor vessel integrity, WCAP-18102-NP, Revision 2, utilizes the surveillance capsule X data that bounds the projected peak fluence value for the BVPS-1 reactor vessel through a projected 42.24 EFPY of plant operation. In accordance with the NRC-approved surveillance capsule withdrawal schedule, capsule Z will be withdrawn and tested at 39 EFPY of plant operation, and the BVPS-1 analysis of record for reactor vessel integrity will be updated to incorporate the results of surveillance capsule Z. At the withdrawal date, the capsule Z fluence will bound the projected peak reactor vessel fluence at the end of the 60-year license renewal period for BVPS-1.

In addition to the in-vessel capsules, neutron fluence at BVPS-1 is also monitored through the use of an ex-vessel neutron dosimetry (EVND) system. In accordance with ASTM E2956-14, "Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels," the BVPS-1 ex-vessel neutron dosimetry is replaced and analyzed every five refueling cycles. The results of the EVND analyses are evaluated

Attachment L-22-193 Page 5 of 13 against the current fluence projections to identify significant differences between the measured and projected values. The BVPS-1 EVND was last removed and replaced during the spring 2021 refueling outage.

Therefore, it is concluded that the capsule withdrawal requirements of ASTM E185-82 have been met for BVPS-1 through the end of its 40-year life, and that the capsule withdrawal requirements of NUREG-1801 Section XI.M31 will be met for BVPS-1 for the 60-year life, without the testing of surveillance capsule A.

The fracture toughness data obtained from the material surveillance program that is required by 10 CFR 50 Appendix H is subsequently used to assess the integrity of the reactor vessel as required by 10 CFR 50 Appendix G and 10 CFR 50.61. Since the withdrawal and testing of surveillance capsule A is not needed to meet the requirements of 10 CFR 50 Appendix H or NUREG-1801, it is concluded that the testing of surveillance capsule A is not needed to meet the requirements of 10 CFR 50 Appendix G or 10 CFR 50.61. The following discussion identifies that the requirements of 10 CFR 50 Appendix G and 10 CFR 50.61 will continue to be met even without the mechanical testing of surveillance capsule A.

Appendix G to 10 CFR 50 The underlying purpose of Appendix G to 10 CFR 50 is to provide an acceptable margin of safety against brittle failure of the reactor coolant system during any condition of normal operation to which the pressure boundary may be subjected over its service lifetime.

Appendix G to 10 CFR 50 - Upper-Shelf Energy (USE) 10 CFR 50 Appendix G requires that Charpy upper shelf energy must account for the effects of neutron radiation. Specifically, 10 CFR 50 Appendix G.IV.1.a requires that reactor vessel beltline materials "must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68 J)."

The BVPS-1 USE values are documented in Westinghouse letter LTR-SDA-17-017, Revision 0, which utilizes material properties taken from WCAP-18102-NP. The BVPS-1 USE values are determined for both the beltline and extended beltline materials at 50 EFPY (which bounds the expected fluence levels at the end of the period of extended operation). LTR-SDA-17-017 identifies that the USE values for all the extended beltline materials will be maintained at no less than 50 ft-lb through the remainder of the 60-year life at BVPS-1. With respect to USE, the limiting BVPS-1 plate material (for plate B6903-1) uses surveillance data (Position 2.2 in Regulatory Guide (RG) 1.99, Revision 2). Section 2.2 of RG 1.99 identifies that USE determined from surveillance data should be used in preference to USE determined using Position 1.2 of RG 1.99, Revision 2.

Attachment L-22-193 Page 6 of 13 Additionally, in accordance with the NRC-approved surveillance capsule withdrawal schedule, BVPS-1 surveillance capsule Z will be withdrawn after 39 EFPY, and the USE evaluation will be updated in accordance with BVPS-1 Technical Specification 5.6.4. The fluence exposure of capsule Z will bound the fluence received by the BVPS-1 reactor vessel at the end of the 60-year license renewal period (5.89 x 1019 n/cm2).

Since all the BVPS-1 reactor vessel materials are projected to meet the 10 CFR 50 Appendix G USE criteria through the end of the 50-year life, and since surveillance capsule Z will provide surveillance data that bounds the reactor vessel fluence at the end of the 60-year life, it is concluded that the Charpy USE requirements of 10 CFR 50 Appendix G are met for BVPS-1 without the testing and analysis of supplemental capsule A.

Appendix G to 10 CFR 50 - Pressure - Temperature Limits 10 CFR 50 Appendix G requires that for reactor vessel beltline materials, the values of RTNDT must account for the effects of neutron radiation, including the results of the surveillance program of Appendix H. Specifically, Appendix G requires that pressure and temperature limits must be established for the reactor coolant pressure boundary during normal and hydrostatic or leak rate testing conditions.

The current BVPS-1 pressure and temperature (P-T) limits are applicable through 50 EFPY (end of license extension) and are documented in the BVPS-1 pressure and temperature limits report (PTLR). The latest revision to the BVPS-1 PLTR incorporated the results of the surveillance capsule X analysis, which are trending conservative. The P-T limit curves generated are based on the limiting cylindrical beltline material (lower shell plate B6903-1). WCAP-18102-NP, Revision 2, provides the technical basis for the current BVPS-1 P-T limits. Table 7-2 and Table 7-3 of WCAP-18102-NP, Revision 2, documents the adjusted RTNDT values for the BVPS-1 P-T limits. Note (d) of Table 7-2 and Note (d) of Table 7-3 identify that the adjusted reference temperature (ART) values were calculated in accordance with the methodology contained in RG, 1.99 Revision 2, and therefore account for the effects of neutron irradiation as required by 10 CFR 50 Appendix G.

Future updates to the BVPS-1 PTLR must be performed in accordance with Technical Specification 5.6.4. Technical Specification 5.6.4 requires that all PTLR updates must be performed in accordance with the methodology identified in WCAP-14040-A, Revision 4. Section 2.4 of WCAP-14040-A, Revision 4, identifies that the adjusted reference temperature for each material in the beltline region is calculated in accordance with RG 1.99, Revision 2. It is therefore concluded that all future PTLR updates for BVPS-1 (including surveillance capsule Z at 39 EFPY) will account for the effects of neutron irradiation as required by 10 CFR 50 Appendix G without the testing of supplemental surveillance capsule A. BVPS Technical Specification 5.6.4 provides the necessary administrative controls to ensure that any required pressure and

Attachment L-22-193 Page 7 of 13 temperature limit changes will be implemented in accordance with approved methodology, such that the requirements for P-T limits in Appendix G are satisfied.

10 CFR 50.61 The underlying purpose of 10 CFR 50.61 is to prevent potential failure of the reactor vessel as a result of PTS.

10 CFR 50.61 requires in part that the RTPTS values for the reactor vessel beltline materials using the end of life fluence values must be less than 270 degrees Fahrenheit for plates, forgings, and axial weld materials, and less than 300 degrees Fahrenheit for circumferential weld materials.

The current BVPS-1 RTPTS values are documented in Appendix E of WCAP-18102-NP, Revision 2. All the BVPS reactor vessel beltline materials remain below the 10 CFR 50.61 screening criteria through the end of the 60-year license renewal period. The NRC reviewed and approved the RTPTS values contained in Revision 1 of WCAP-18102-NP in a safety evaluation dated July 2, 2018 (ML18164A082). There were no changes made to Appendix E between Revision 1 and Revision 2 of WCAP-18102-NP. It is therefore concluded that the RTPTS values for the BVPS-1 beltline materials are compliant with the requirements of 10 CFR 50.61 through the end of the license renewal period. Not performing the mechanical testing of capsule A will not affect the validity of the BVPS-1 RTPTS values.

Additionally, the limiting BVPS-1 beltline material (lower shell plate B6903-1) has 11.9 degrees of margin to the 10 CFR 50.61 screening criterion for RTPTS. RTPTS values are updated when surveillance capsules are withdrawn and analyzed because the projected fluence values used to calculate RTPTS are updated to account for actual fluence exposure from historical plant operation. When BVPS-1 surveillance capsule Z is withdrawn and tested in approximately 2027, the RTPTS values will also be updated from historical plant operation. It is anticipated that these values will not significantly differ from the projected BVPS-1 RTPTS values. Therefore, it is expected that the BVPS-1 RTPTS values will remain below the 10 CFR 50.61 screening criteria even after the surveillance capsule Z testing and analysis.

==

Conclusion:==

The proposed exemption request is acceptable because BVPS-1 will remain compliant with 10 CFR 50 Appendix H, ASTM E185-82, and NUREG-1801 Section XI.M31. The surveillance capsule testing and withdrawal requirements of 10 CFR 50 Appendix H are met without the mechanical testing of surveillance capsule A, and therefore the actions after capsule withdrawal and testing that are defined in 10 CFR 50 Appendix G and 10 CFR 50.61 will also be met without the mechanical testing of surveillance capsule A.

Attachment L-22-193 Page 8 of 13 Current USE and P-T limits remain applicable and appropriate for use at BVPS-1 through the end of the 60 year license renewal period.

Due to the additional evaluation that was documented in ML18164A082, the PTS concern that initiated the development of supplemental capsule A is no longer applicable for the BVPS-1 60-year license renewal period.

Regulatory discussion:

10 CFR 50.12, "Specific exemptions," states that the Commission may grant exemptions from the requirements of the regulations of this part provided that the exemption is authorized by law, will not present undue risk to the public health and safety, and is consistent with the common defense and security. The regulation also states that the Commission will not consider granting an exemption unless special circumstances are present.

Required criteria:

The requested exemption satisfies the required criteria as described below.

1. This exemption is authorized by law.

The analyses report required by 10 CFR Part 50; Appendix H is not required by statute.

The requested exemption is authorized by law in that no law precludes the activity covered by this request. The Commission has the authority under 10 CFR 50.12 to grant an exemption from the requirements of 10 CFR 50 upon proper justification. The request does not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commissions regulations. Therefore, granting an exemption is explicitly authorized by law.

2. This exemption will not present undue risk to the public health and safety.

Section IV.A of 10 CFR 50 Appendix H indicates that capsule withdrawal and associated test results need to be the subject of a summary technical report that is to be submitted to the NRC within 18 months of capsule withdrawal. Although capsule A will be removed from the BVPS-2 reactor vessel in accordance with the NRC approved withdrawal schedule, exemption is requested from the associated testing and report submittal activities. The exemption is requested based on a previous evaluation that determined that the original PTS concern is no longer applicable, and that the other BVPS-1 surveillance capsules provide adequate monitoring of vessel embrittlement throughout the period of extended operation. The regular BVPS-1 surveillance capsules meet the requirements ASTM E185-82 and NUREG-1801, Revision 2.

Capsule A was a supplemental capsule and is not required to meet the intent of 10 CFR 50 Appendix H for the license renewal period, and therefore, this exemption will not present undue risk to the public health and safety.

Attachment L-22-193 Page 9 of 13

3. This exemption is consistent with the common defense and security.

Complying with the surveillance capsule reporting requirements of 10 CFR 50 Appendix H does not involve an activity that could potentially impact the common defense and security of the United States.Section IV.A of 10 CFR 50 Appendix H, requires the submittal of a technical report that provides withdrawn capsule test results within 18 months of the date of the capsule withdrawal. Surveillance capsule A is a supplemental capsule and is not required to meet the surveillance requirements identified in ASTM E185-82, NUREG-1801,Section XI.M31, or the regulation described above. Therefore, the common defense and security are not impacted by this exemption request.

Special circumstances:

Appendix H of 10 CFR 50, Section I, identifies that the data to be obtained through the Appendix H surveillance program "will be used as described in Section IV of Appendix G to part 50." The introduction to Appendix G states that "This appendix specifies fracture toughness requirements for ferritic materials ... to provide adequate margins of safety during any condition of normal operation ... "

The underlying purpose of 10 CFR 50, Appendix H (and Appendix G), is to provide adequate margins of safety to protect the reactor vessel from brittle fracture. Appendix H assists in protecting the reactor vessel from brittle fracture by requiring a surveillance program that monitors changes in the fracture toughness properties of the ferritic materials that make up the reactor vessel. The results of the surveillance program required by Appendix H are evaluated against the fracture toughness limits established in accordance with 10 CFR 50, Appendix G, Section IV.

In accordance with 10 CFR 50.12(a)(2)(ii), a special circumstance would involve the application of a regulation when it is not necessary to achieve the underlying purpose of the rule. This would apply to section IV.A of 10 CFR 50, Appendix H, that requires the submittal of a technical report that provides withdrawn capsule test results within 18 months of the date of the capsule withdrawal. The capsule testing requirements necessary to provide adequate margins of safety and protect the BVPS-1 reactor vessel from brittle fracture have been met for its 40-year life. The capsule testing requirements necessary to provide adequate margins of safety and protect the BVPS-1 reactor vessel from brittle fracture will be met for its 60-year life with the withdrawal and testing of surveillance capsule Z in accordance with the NRC approved surveillance capsule withdrawal schedule. Surveillance capsule A is a supplemental capsule and is not required to meet the surveillance requirements identified in ASTM E185-82 or NUREG-1801,Section XI.M31.

Attachment L-22-193 Page 10 of 13 5.0 ENVIRONMENTAL ASSESSMENT Energy Harbor Nuclear Corp. has determined the proposed exemption does not require an environmental review since it meets the eligibility criteria for categorical exclusion in 10 CFR 51.22(c)(25), as: (i) there is no significant hazards consideration; (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (iii) there is no significant increase in individual or cumulative public or occupational radiation exposure; (iv) there is no significant construction impact; (v) there is no significant increase in the potential for or consequences from radiological accidents; and (vi) the requirements from which the exemption is sought involves inspection or surveillance requirements. The information provided below supports the basis for this determination.

(i) No significant hazards consideration.

Section IV.A of 10 CFR 50, Appendix H, indicates that capsule withdrawal and associated test results need to be the subject of a summary technical report that is to be submitted to the NRC within eighteen months of capsule withdrawal. Although capsule A will be removed from the BVPS-2 reactor vessel in accordance with the NRC approved withdrawal schedule, exemption is requested from the associated testing and report submittal activities. The exemption is requested because the other reactor vessel surveillance capsules already meet the intent of 10 CFR 50 Appendix H by complying with the requirements of ASTM E185-82 and NUREG-1801 for the 60 year license renewal period without the testing of supplemental capsule A.

Energy Harbor Nuclear Corp. has evaluated the proposed exemption to determine whether a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed exemption involve a significant increase in the probability or consequence of an accident previously evaluated?

Response: No The proposed exemption has no effect on facility structures, systems, and components (SSCs), the capability of any facility SSC to perform its design function, or plant operations, and, therefore, would not increase the likelihood of a malfunction of any facility SSC or increase the consequences of previously evaluated accidents. The proposed exemption does not alter any assumptions or methodology associated with the previously evaluated accidents in the BVPS-1 Updated Final Safety Analysis Report. The proposed exemption will not affect the probability of occurrence of any previously analyzed accident.

Therefore, there is no increase in the probability or consequence of any previously evaluated accident.

Attachment L-22-193 Page 11 of 13

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

Response: No The proposed exemption does not involve a physical alteration of the facility. No new or different type of equipment will be installed, and there are no physical modifications to existing equipment associated with the proposed exemption.

Similarly, the proposed exemption would not physically alter any SSCs involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed exemption does not create the possibility of a new accident because of new failure modes associated with any equipment or personnel failures. No changes are being made to the facilities' normal parameters or in protective or mitigative action setpoints, and no new failure modes are being introduced.

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

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

Response: No The proposed exemption does not alter the design basis or any safety limits, nor does it impact station operation or any facility SSC that is relied upon for accident mitigation.

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

Based on the above, Energy Harbor Nuclear Corp. concludes that the proposed exemption 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.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed exemption from the testing of reactor vessel capsule A and associated report submittal activities will not result in any changes to systems, structures, or components that function to limit or monitor the release of plant effluents. The proposed exemption does not involve a physical alteration of the facility, and it will not impact operation of any plant system. Therefore, there is no significant change in the type or significant increase in the number of effluents that may be released offsite with the proposed exemption.

Attachment L-22-193 Page 12 of 13 (iii) There is no significant increase in individual or cumulative public or occupational radiation exposure.

The proposed exemption does not involve a physical alteration of the facility, and it will not impact operation of any plant system. There will be no significant increase in individual or cumulative public or occupational radiation exposure associated with this exemption.

(iv) There is no significant construction impact.

No construction activities are associated with the proposed exemption.

(v) There is no significant increase in the potential for or consequences from radiological accidents.

The proposed exemption does not involve a physical alteration of the facility, and it will not impact operation of any plant system.

The proposed exemption will not affect the ability to respond to or mitigate any previously evaluated accidents, or affect the radiological assumptions used in the evaluations. The exemption will not affect the consequences of any accidents previously evaluated. Therefore, the proposed exemption does not result in a significant increase in the potential for, or consequences of, a radiological accident.

(vi) The requirements from which an exemption is sought involve inspection or surveillance requirements.

The purpose of the exemption is to allow alternative action to the stated requirement in Section IV, Item A, in Appendix H of 10 CFR 50. The proposed exemption involves reactor vessel capsule testing and associated report submittal requirements.

Performance of the scheduled capsule testing is a surveillance requirement.

6.0 CONCLUSION

Pursuant to 10 CFR 50.12, the requested exemption is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security. Approval of this exemption request maintains alignment with the purpose of 10 CFR 50, Appendix H.

7.0 REFERENCES

1. Westinghouse Report WCAP-18102-NP, Revision 2, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," March 2021.

[ML21113A044].

Attachment L-22-193 Page 13 of 13

2. Electric Power Research Institute, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CVRSP) Guidelines (MRP-326, Revision 1 )," June 2021.

3. Letter from Robert G. Schaaf (NRC) to Eric A. Larson (FirstEnergy Nuclear Operating Company), "Beaver Valley Power Station, Units 1 and 2 - Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule," [ML13242A266],

July 17, 2014.

4. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706."

5. NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report,"

December 2010, [ML103490041].

6. ASTM E2956-14, "Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels.

7. Westinghouse Letter LTR-SDA-17-017 Revision 0, Beaver Valley Unit 1 Upper-Shelf Energy Values, November 16, 2017.

8. Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004, [ML050120209].