L-20-059, License Amendment Request to Extend License Expiration Date by Recapturing Low-Power Testing Time
| ML20086K773 | |
| Person / Time | |
|---|---|
| Site: | Perry  (NPF-058) |
| Issue date: | 03/26/2020 |
| From: | Payne F Energy Harbor Nuclear Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-20-059 | |
| Download: ML20086K773 (17) | |
Text
energy inQrbr Energy Harbor Nuclear Corp.
Pe:Try Nuclear Power Plant 10 Center Road P.O. Boat 97 Perry, Ohio 44081 Frank R. Payne Site Vice President, Perry Nuclear March 26, 2020 L-20-059 440-280-5382 10 CFR 50.90 AITN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant Docket No. 50440, License No. NPF-58 License Amendment Request to Extend License EXDiration Date Bv RecaDturina Low-Power Testing Time Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Energy Harbor Nuclear Corp. requests an amendment to the Perry Nuclear Power Plant, Unit No. 1 (PNPP) facilfty (full power) operating license No. NPF-58 expiration date to recapture low-power testing time. Specifically, the expiration date of PNPP's fulLpower operating license (FPOL) would be revised such that it would expire 40 years from the date of issuance of the FPOL, as opposed to 40 years from the date of issuance of the fuel loading and low-power testing license No. NPF45, as permitted by 10 CFR 50.51, "Continuation of license." If this amendment is approved, the FPOL would be extended for approximately 7.7 months, and would expire on November 7, 2026 instead of March 18, 2026.
Energy Harbor Nuclear Corp,'s evaluation of the proposed amendment is enclosed.
Approval of the proposed amendment is requested by March 31, 2021. 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 -
Nuclear Licensing and Regulatory Affairs, at 330-315-6810.
I declare under penalty of perjury that the foregoing is true and correct. Executed on March A, 2020.
.-.,. `
Perry Nuclear Power Plant L-20-059 Page 2
Enclosure:
Evaluation of the Proposed Amendment cc: NRC Region III Administrator NRC Resident Inspector NRC Project Manager Branch Chief, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
Utility Radiological Safety Board
Enclosure L-20-059 Evaluation of the Proposed Amendment (14 pages follow)
Evaluation of the Proposed Amendment Page 1 of 12
Subject:
Proposed Amendment of Operating License Condition 2.H, License Expiration Date TABLE OF CONTENTS 1.0
SUMMARY
DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.1.1 Reactor Pressure Vessel 2.1.2 Structures 2.1.3 Mechanical Equipment 2.1.4 Electrical Equipment 2.2 Current License Requirement 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change
3.0 TECHNICAL EVALUATION
3.1 Evaluation 3.1.1 Reactor Pressure Vessel 3.1.2 Structures 3.1.3 Mechanical Equipment 3.1.4 Electrical Equipment 3.1.5 Quality Assurance and Maintenance Programs 3.1.6 Aging Effects 3.2 Conclusion
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements / Criteria 4.1.1 Regulations and Requirements 4.1.2 Relevant NRC Documents 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions
5.0 ENVIRONMENTAL CONSIDERATION
S ATTACHMENT License Page Markup
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 2 of 12 1.0
SUMMARY
DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, an amendment to the Perry Nuclear Power Plant, Unit No. 1 (PNPP) Facility Operating License, No. NPF-58, expiration date is requested to recapture low-power testing time. The proposed amendment would change the facility operating license expiration date shown in license condition 2.H from March 18, 2026 to November 7, 2026.
2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The following sections describe design features and programs relevant to the proposed amendment. More specifically, the sections describe the design and construction of the affected safety related reactor pressure vessel, structures, mechanical equipment, and electrical equipment.
2.1.1 Reactor Pressure Vessel The reactor pressure vessel was designed and fabricated in accordance with the requirements of Section III, Class 1, of the American Society of Mechanical Engineers (ASME) Code edition, and addenda applicable at the time of order placement, Winter 1972. Operating limitations of the ASME Code and of 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," also apply. The reactor pressure vessel and the reactor coolant system were designed to allow inspections in accordance with Section XI of the ASME Code. Industry experience with steel structures confirms a service life in excess of 40 years.
2.1.2 Structures As discussed in Updated Safety Analysis Report, Section 3.1, Conformance with General Design Criteria, PNPP is in compliance with the general design criteria of 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants." The NRC reviewed the design, fabrication, and construction of PNPP structures, systems, and components important to safety, and the quality assurance program for operation of the facility when the plant was licensed for low-power operation. The NRC's safety evaluations approving the programs and their implementation with respect to these structures are documented in the NRC's Safety Evaluation Report for PNPP (NUREG-0887) and its supplements. Industry experience with concrete and steel structures confirms a service life in excess of 40 years. The recapture period requested by the proposed amendment represents approximately 1.6 percent of the 40-year service life of the plant.
The major codes and specifications used in the design and construction of the Category 1 concrete and steel structures were ACI 318-71, "Building Code Requirements for Reinforced Concrete," and the American Institute of Steel Construction (AISC) 1969 specification, "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings." The foundations of the seismic Category 1 structures are reinforced concrete designed to meet ACI 318-71 requirements. The design and construction of Category 1 concrete and steel structures will continue to
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 3 of 12 comply with these codes and specifications (that are specified in the Updated Safety Analysis Report) during the requested 7.7 months of additional power operation.
Sections 3.8.2, 3.8.3, and 3.8.4 of NUREG-0887 state that the criteria that were used in the analysis, design, and construction of seismic Category 1 structures at PNPP account for anticipated loading and postulated conditions that may be imposed on the structures during their service lifetime.
2.1.3 Mechanical Equipment ASME Code Section XI Equipment that is safety-related is classified as ASME Code Class 1, 2, or 3 and is subject to the inservice inspection requirements of Section XI and 10 CFR 50.55a, except where the NRC has granted written relief from these requirements. These requirements apply throughout the service life of a plant.
Technical Specifications 10 CFR 50.36 requires the establishment of limiting conditions for operation for certain equipment. Limiting conditions for operation specified in plant technical specifications are the lowest functional capability or performance levels of equipment required for safe operation of the facility. This equipment is subject to surveillance requirements relating to test, calibration, or inspection that are also specified in plant technical specifications to assure the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and the limiting conditions for operation will be met.
10 CFR 50.65 During normal plant operations (including normal shutdown operations) 10 CFR 50.65 requires monitoring of the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components can fulfill their intended functions.
10 CFR Part 50, Appendix J Leakage test requirements are specified in 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Appendix J includes Type A testing requirements for the overall containment structure and Type B and C testing requirements for penetrations and valves, respectively. These tests ensure that leakage from the reactor containment does not exceed allowable leakage rates specified in the PNPP Technical Specifications and the integrity of the containment structure is maintained during its service life.
Mechanical Equipment Qualification Mechanical equipment that is required to perform a design safety function is designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents in accordance with GDC 4.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 4 of 12 2.1.4 Electrical Equipment Electrical equipment that is required to perform a design safety function is designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents in accordance with GDC 4.
Environmentally qualified electrical equipment requires scheduled maintenance to prevent the components from exceeding their qualified life, and periodic testing to locate components that may have failed. Nonscheduled maintenance is performed as a result of out-of-specification performance, and any malfunctioning part is repaired or replaced by a part that is equivalent to the original part.
2.2 Current License Requirement Facility Operating License No. NPF-58, Condition 2.H, for PNPP states that:
This license is effective as of the date of issuance and shall expire at midnight on March 18, 2026.
2.3 Reason for the Proposed Change The proposed amendment request to recapture low-power testing time and thereby extend the PNPP facility operating license expiration date by 7.7 months would permit additional time for PNPP to operate and additional time to prepare a license renewal application.
2.4 Description of the Proposed Change The proposed amendment would revise the expiration date specified in PNPP facility operating license No. NPF-58, Condition 2.H, to recapture low-power testing time. The expiration date of PNPP's facility operating license would be revised so that it would expire 40 years from the date of issuance of the facility operating license, as opposed to 40 years from the date of issuance of the fuel loading and low-power testing license, as permitted by 10 CFR 50.51, Continuation of license. As shown on the attached license page markup, this proposed amendment would extend the facility operating license for approximately 7.7 months, so that the facility operating license would expire on November 7, 2026 instead of March 18, 2026.
License condition 2.H would be revised to read as follows:
This license is effective as of the date of issuance and shall expire at midnight on November 7, 2026.
3.0 TECHNICAL EVALUATION
3.1 Evaluation Current regulations require development and implementation of programs that ensure conditions adverse to quality, including degraded system, structure, or component function, are promptly identified and corrected. These programs include self-inspection, maintenance, and technical specification surveillance programs that monitor and test the physical condition of plant systems, structures, and components.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 5 of 12 For example, technical specifications include limiting conditions for operation, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Technical specifications also require surveillance requirements relating to test, calibration, or inspection to verify that the necessary quality of systems, structures, and components is maintained, that facility operation is within safety limits, and that limiting conditions for operation continue to be met. Furthermore, 10 CFR 50.55a requires, in part, that systems, structures, and components be tested and inspected against quality standards commensurate with the importance of the safety function to be performed, such as inservice testing and inservice inspections of pumps and valves.
Through these programs, degradation of components resulting from a number of different environmental stressors is identified, as well as degradation from inadequate maintenance or errors caused by personnel. Once a detrimental performance or condition caused by aging or other factors is revealed, mitigating actions are taken to fully restore the condition to its original design basis.
Consequently, there is considerable logic in ensuring that the design basis of systems, structures, and components is maintained through activities that ensure continued functionality. This process, including surveillance, is relied on in the current term to ensure continued operability (that is, to the greatest extent practicable, the intended design functions will be properly performed). The focus on maintaining functionality results in the continuing capability of systems, structures, and components, including supporting systems, structures, and components, to perform their intended functions as designed.
PNPP was designed and constructed to be operated over a 40-year service life. Design features allow for routine inspection of structures, systems, and components during this service life in accordance with NRC requirements and PNPP procedures.
Surveillance and inspection practices, which have been implemented in accordance with the PNPP Technical Specifications and ASME Code, respectively, and plant maintenance practices provide assurance that degradation in plant safety-related equipment will be identified and corrected to provide continued safe operation of the unit throughout the duration of the facility operating license, including the proposed license extension (that is, low-power testing recapture) period.
3.1.1 Reactor Pressure Vessel Over the operating life of a reactor pressure vessel, ferritic materials exposed to neutron irradiation will undergo changes in material properties and a decrease in fracture toughness. The decrease in fracture toughness is important because the ability of ferritic materials to resist failure caused by the propagation of a crack decreases with increasing irradiation. A surveillance program in accordance with 10 CFR Part 50, Appendix H, "Reactor Vessel Materials Surveillance Program Requirements," was developed to monitor changes in the fracture toughness properties of reactor pressure vessel materials. The purpose of the program is to help ensure reactor pressure vessel integrity by monitoring changes in the fracture toughness properties of the reactor
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 6 of 12 pressure vessel beltline materials. Input from this program is used to develop operating limits, in the form of pressure and temperature limit curves, which ensure adequate margin regarding brittle failure of the reactor pressure vessel and piping of the reactor coolant pressure boundary.
In 1998, pressure and temperature limit curves in Technical Specification 3.4.11, "Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits," were revised to incorporate specific pressure and temperature limits for the bottom head region of the reactor pressure vessel, separate from the curves for the core beltline region of the reactor pressure vessel (Accession No. ML021840331). The pressure and temperature curve limits are established in accordance with 10 CFR Part 50, Appendix G, and the values of adjusted reference temperatures and upper shelf energy are expected to remain within the limits of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," for at least 32 effective full power years of operation.
The pressure and temperature limit curves in Technical Specification 3.4.11 added by Amendment 127 (Accession No. ML030700189) were based, in part, on an alternative methodology. The alternative methodology used ASME Code Cases N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division I." In developing the pressure and temperature limit curves, the PNPP neutron fluence calculations were also updated. These calculation updates were performed using the NRC-approved "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation" (General Electric Nuclear Energy Topical Report, NEDC-32983P-A, Revision 1). The pressure and temperature limit curves also incorporated a revised adjusted reference temperature. The pressure and temperature limit curves approved with Amendment 127 were calculated for 22 and 32 effective full power years of operation.
The most recent amendment (Amendment 168) of the pressure and temperature limit curves in Technical Specification 3.4.11 updated pressure and temperature limits due to issues involving a water level instrument nozzle and vacuum conditions in the reactor coolant system. For the amendment, compliance with the NRC approved methodology for developing P/T limits, including Appendix G, were maintained.
The reactor pressure vessel materials are discussed in Section 5.3.1 of the PNPP Updated Safety Analysis Report. Per PNPP Updated Safety Analysis Report, Section 5.3.1.6.1, Compliance with Reactor Vessel Material Surveillance Program Requirements, the program for implementation of the scheduling, withdrawal, and testing of the reactor pressure vessel material surveillance specimens is governed and controlled by the Boiling Water Reactor Vessel and Internals Project (BWRVIP)
BWRVIP-86-A, BWR Vessel Internals Project, BWR Integrated Surveillance Program (ISP) Implementation Plan, dated October 2002. This integrated surveillance program complies with the requirements of 10 CFR Part 50, Appendix H.
Reassessment of the validity of the pressure and temperature limit curves occurs as new data become available from the integrated surveillance program. Based on the current withdrawal schedule in the latest NRC-approved revision of BWRVIP-86-A, the
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 7 of 12 next scheduled integrated surveillance program surveillance capsule withdrawal applicable to the PNPP reactor pressure vessel material is not scheduled until after 2026.
As of March 2013, the PNPP reactor pressure vessel accumulated a total of 20 effective full power years of exposure. Looking at the expected future operating history, it is concluded that the proposed 7.7-month extension of the facility operating license expiration date to November 7, 2026 will remain within the current 32 effective full power year period of validity without reassessment for the integrated surveillance program surveillance capsule scheduled to be withdrawn after 2026.
Based on the above discussion, there is reasonable assurance that the PNPP reactor pressure vessel will continue to meet applicable Part 50 and PNPP license requirements during the additional 7.7 months of plant operation sought by this low-power testing recapture request.
3.1.2 Structures The NRC has approved use of the codes, standards, and specifications referenced in Section 2.1.2 for the PNPP design, analyses, and construction, as well as the PNPP Quality Assurance (QA) Program required by 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." Thus, there is reasonable assurance that the concrete and steel structures will continue to comply with applicable NRC regulations and PNPP facility operating license requirements during the proposed 7.7-month license term extension.
3.1.3 Mechanical Equipment Surveillance and maintenance requirements for mechanical equipment are in place to verify operability or to detect degradation and ensure that the equipment that does degrade is replaced or other corrective actions are taken. Surveillance, inspection, testing and monitoring requirements at PNPP, which apply during the plant's operating life, include the items described in Section 2.1.3.
Compliance with the codes, standards, and regulatory requirements to which the mechanical equipment were analyzed, constructed, tested, and inspected provides reasonable assurance that the structural integrity of equipment important to safety will be maintained during the operating life of the plant, including the additional 7.7 months of operating life requested in this amendment. Such compliance further ensures that any significant degradation of mechanical equipment is discovered, and the equipment is restored to an acceptable and operable condition.
3.1.4 Electrical Equipment Scheduled maintenance and periodic testing of safety related electrical equipment as described in Section 2.1.4 provides reasonable assurance that these components will perform their safety function if called upon.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 8 of 12 3.1.5 Quality Assurance and Maintenance Programs In licensing PNPP, the NRC reviewed the QA program and the conduct of operations.
The QA program for plant operations assesses how the plant organization is following procedures and meeting requirements for plant operation. This includes the plant maintenance program that assures the equipment is operable. In NUREG-0887, the NRC concluded that the QA program and program for use of operating and maintenance procedures were acceptable.
NRC inspections of the quality assurance and maintenance programs at PNPP show that these programs remain acceptable. The QA program implements the requirements of 10 CFR Part 50, Appendix B and applies to activities associated with structures, systems, and components which are safety related.
Therefore, implementation and use of the quality assurance and maintenance programs at PNPP provide reasonable assurance that, for the proposed license term extension, equipment important to safety will satisfy applicable NRC requirements and the PNPP facility operating license.
3.1.6 Aging Effects Aging effects are not safety-significant because the license recapture period, about 7.7 months for PNPP, is a small percentage (approximately 1.6 percent) of the overall license period.
3.2 Conclusion The requested amendment involves no physical changes to the design features or operation of the facility. The proposed amendment will not impact the design functions, or methods of performing or controlling design functions of structures, systems, and components. Nor will it affect the conduct of PNPP programs. As a result, the proposed amendment will not change accident analysis assumptions, or change, degrade, or prevent actions described or assumed in accidents evaluated and described in the PNPP Updated Safety Analysis Report. Therefore, the proposed amendment does not adversely affect public health and safety or result in an increase in the radiological consequences of any accident described in the PNPP Updated Safety Analysis Report.
Based on the above discussion, it is concluded that there are no safety issues that would preclude an additional 7.7 months of operation beyond the current facility operating license expiration date. This time period is insignificant from an aging effects perspective, particularly when considered in conjunction with the surveillance, inspection, and maintenance programs implemented to provide early indication of degradation in plant safety-related equipment. Ongoing maintenance and testing provide for continued safe operation of the unit throughout the duration of the facility operating license and would continue to do so during the additional 7.7 months of operation requested by this proposed amendment.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 9 of 12
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements / Criteria 4.1.1 Regulations and Requirements The proposed amendment has been evaluated to determine whether applicable regulations and requirements continue to be met as described below.
Section 103.c of the Atomic Energy Act of 1954, as amended, provides that a license is to be issued for a specific period not to exceed 40 years. 10 CFR 50.51 also specifies that each license will be issued for a fixed period of time not to exceed 40 years from the date of issuance. Also, 10 CFR 50.56 and 50.57 allow the issuance of an operating license pursuant to 10 CFR 50.51 after the construction of the facility has been substantially completed, in conformity with the construction permit and when other provisions specified in 10 CFR 50.57 are met.
Regulation 10 CFR, Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," provides the requirements for a quality assurance program.
Regulation 10 CFR, Part 50, Appendix G, "Fracture Toughness Requirements,"
specifies the fracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary, including reactor pressure vessels.
Regulation 10 CFR, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," specifies the requirements to monitor changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel.
Regulation 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," specifies the requirements for establishing a program for qualifying electrical equipment.
No changes to plant design, programs, or procedures that ensure compliance with the regulations and requirements cited above are proposed as part of the proposed license extension. Compliance with the codes, standards, and regulatory requirements to which mechanical and electrical equipment are analyzed, constructed, tested, and inspected provides adequate assurance that the structural integrity of equipment important to safety will be maintained during the operating lifetime of the plant and during the proposed license term extension. Any significant degradation of such equipment would be discovered, and the equipment restored to an acceptable, and operable, condition.
4.1.2 Relevant NRC Documents The following NRC documents are relevant to the proposed amendment:
- Commission paper SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated December 21, 1998.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 10 of 12
- Staff requirements memorandum (SRM) regarding SECY-98-296, "Staff Requirements - SECY-98-296 - Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated March 30, 1999.
In the SRM for commission paper SECY-98-296, "Staff Requirements - SECY-98-296 -
Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated March 30, 1999, the Commission established NRC policy regarding recapture of low-power testing time for nuclear power plants.
The Commission approved granting future license amendment requests from licensees provided the 40-year license term began with issuance of a low power operating license and a separate full power facility operating license was issued. The request to recapture time spent conducting low-power testing at PNPP falls within the scope of this Commission policy.
4.2 Precedent The proposed PNPP license period extension to recapture time spent conducting low-power testing is similar to a request submitted for the Clinton Power Station. Consistent with the policy established by the Commission in the March 30, 1999 SRM for SECY-98-296, the NRC staff approved a request (Accession No. ML19109A001) to amend the expiration date of the Clinton Power Station operating license issued under Atomic Energy Act Section 103.c to allow recapture of the time spent by the plant in low-power testing before issuance of their facility operating license.
4.3 No Significant Hazards Consideration The proposed amendment would revise the expiration date of the Perry Nuclear Power Plant (PNPP) facility operating license to base it upon the issuance date of the facility operating license instead of the issuance date of the low-power testing license.
Whether a significant hazards consideration is involved with the proposed amendment has been evaluated by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated because it does not involve a change to the design configuration or operation of the facility. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously analyzed in the PNPP Updated Safety Analysis Report.
PNPP was designed and constructed to ensure at least a 40-year service life. Design features provide for inspection of structures, systems, and components during this
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 11 of 12 service life. Surveillance, inspection, and maintenance practices, which have been implemented in accordance with the ASME Code and the PNPP Technical Specifications, provide assurance that any degradation in plant safety-related equipment will be identified and corrected to ensure continued safe operation of the unit throughout the duration of the facility operating license.
The low-power testing recapture period requested by this amendment is for 7.7 months.
This time period is insignificant from an aging effects perspective, particularly when considered in conjunction with the surveillance, inspection, and maintenance programs described above.
Therefore, the proposed amendment does not significantly increase 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 proposed amendment does not involve a physical alteration of plant systems, structures or components (no new or different type of equipment will be installed), or change parameters governing the manner in which the plant is operated and maintained.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment does not involve a physical alteration of plant systems, structures or components (no new or different type of equipment will be installed), or change parameters governing the way the plant is operated and maintained.
Margin of safety is associated with confidence in the ability of the fission product barriers (that is, fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the radiological dose to the public and control room operators in the event of an accident. The proposed amendment to the facility operating license has no impact on the margin of safety provided in the design and construction of the facility. In addition, the proposed amendment will not relax any of the criteria used to establish safety limits, nor will the proposed amendment relax safety system settings or limiting conditions for operation as defined in the PNPP Technical Specifications.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Based on the above, it is concluded 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.
Evaluation of the Proposed Amendment Perry Nuclear Power Plant, Unit No. 1 Page 12 of 12 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 Commissions 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 ENVIRONMENTAL CONSIDERATION
S A review has determined that 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 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.
Attachment License Page Markup (1 page follows)
Amendment No. 187 H. This license is effective as of the date of issuance and shall expire at midnight on March 18, 2026.
- 1. Based on the Commissions Order dated April 15, 2016 regarding the direct transfer from Ohio Edison Company (OE) to FirstEnergy Nuclear Generation, LLC (FENGen)*, FENOC** and FENGen* shall comply with the following conditions noted below.
A. On June 1, 2016, OE shall transfer to FENGen* the accumulated decommissioning funds held by OE for Perry. All such funds shall be deposited and held in FENGen* nuclear decommissioning trust for Perry along with the other funds maintained by FENGen* for Perry and shall be maintained in such trust outside of FENGens* administrative control and in accordance with the requirements of 10 CFR 50.75(h)(1).
B. The Support Agreement in the amount of $400 million from FirstEnergy Solutions Corp. (FE Solutions)*** described in the application dated June 30, 2015 shall be effective and consistent with the representations in the application. FENGen* shall take no action to void, cancel or modify the Support Agreement without the prior written consent of the NRC staff. FENGen* shall inform the Director of the Office of Nuclear Reactor Regulation, in writing, no later than 10 working days after any funds are provided to FENGen* by FirstEnergy Solutions*** under the terms of the Support Agreement.
FirstEnergy Nuclear Generation, LLC (FENGen) has been renamed Energy Harbor Nuclear Generation LLC.
- FirstEnergy Nuclear Operating Company (FENOC) has been renamed Energy Harbor Nuclear Corp.
- FirstEnergy Solutions Corp. (FE Solutions) has been renamed Energy Harbor Corp.
November 7