ML13017A431

Issuance of Amendment Revise Technical Specification 3.10.1, Inservice Leak, and Hydrostatic Testing Operation

ML13017A431
Person / Time
Site:	Perry
Issue date:	04/18/2013
From:	Michael Mahoney Plant Licensing Branch III
To:	Kaminskas V FirstEnergy Nuclear Operating Co
Michael Mahoney, NRR/DORL, 415-3867
References
TAC ME8048
Download: ML13017A431 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 18, 2013 Mr. Vito A. Kaminskas Site Vice President FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant Mail Stop A-PY-A290 P.O. Box 97,10 Center Road Perry, OH 44081-0097

SUBJECT:

PERRY NUCLEAR POWER PLANT, UNIT NO.1 - ISSUANCE OF AMENDMENT RE: REVISE TECHINCAL SPECIFICATION 3.10.1, "INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION" (TAC NO. ME8048)

Dear Mr. Kaminskas:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 163 to Facility Operating License No. NPF-58 for the Perry Nuclear Power Plant, Unit No, 1. This amendment revises the technical specifications (TS) in response to your application dated February 22,2012, and supplemented by letter dated March 8, 2013.

This amendment revised TS 3,10.1, "Inservice Leak and Hydrostatic Testing Operation," and the associated TS Bases to expand its scope to include provisions for temperature excursions greater than 200 degree Fahrenheit as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in MODE 4.

A copy of the Safety Evaluation is also enclosed, The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Michael Mahoney, R ject Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440

Enclosures:

1. Amendment No, 163 to NPF-58

2. Safety Evaluation cc w/encls: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION CORP.

OHIO EDISON COMPANY DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 163 License No. NPF-58

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for license filed by FirstEnergy Nuclear Operating Company, et ai., (the licensee, FENOC) dated February 22,2012, and supplemented by letter dated March 8, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and reg ulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby amended to read as follows:

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(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 163 are hereby incorporated into this license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jeremy S. Bowen, Chief Plant Licensing Branch 1I1~2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: April 18, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 163 FACILITY OPERATING LICENSE NO. NPF*58 DOCKET NO. 50*440 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License NPF-58 License NPF-58 Page 4 Page 4 TSs TSs 3.10.1 3.10.1

-4 renewal. Such sale and leaseback transactions are subject to the representations and conditions set forth in the above mentioned application of January 23, 1987, as supplemented on March 3, 1987, as well as the letter of the Director of the Office of Nuclear Reactor Regulation dated March 16, 1987, consenting to such transactions. Specifically, a lessor and anyone else who may acquire an interest under these transactions are prohibited f.rom exercising directly or indirectly any control over the licenses of PNPP Unit 1. For purposes of this condition the limitations of 10 CFR 50.81, as now in effect and as may be subsequently amended, are fully applicable to the lessor and any successor in interest to that lessor as long as the license for PNPP Unit 1 remains in effect; these financial transactions shall have no effect on the license for the Perry Nuclear facility throughout the term of the license.

(b) Further, the licensees are also required to notify the NRC in writing prior to any change in: (i) the terms or conditions of any lease agreements executed as part of these transactions; (ii) the PNPP Operating Agreement; (iii) the existing property insurance coverage for PNPP Unit 1; and (iv) any action by a lessor or others that may have an adverse effect on the safe operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now and 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 reactor core power levels not in excess of 3758 megawatts thermal (100% power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 163, are hereby incorporated into the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions

a. FirstEnergy Nuclear Generation, LLC and Ohio Edison Company Amendment No. 163

Inservice Leak and Hydrostatic Testing Operation 3.10.1 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation LCD 3.10.1 The average reactor coolant temperature sp,ecified in Table 1.1-1 for MODE 4 may be changed to 'NA." and operation considered not to be in MODE 3: and the requirements of LCD 3.4.10. "Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown." may be suspended to allow reactor coolant temperature> 20DOF:

• For performance of an inservice leak or hydrostatic test.

• As a conse9uence of maintaining adequate pressure for an inservice leak or hydrostatic test. or

• As a consequence of maintai*ning adequate pressure for control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test.

provided the following MODE 3 LCOs are met:

a. LCO 3.3.6.1. "Primary Containment and Orywell Isolation Instrumentation." Functions 2.a, 2.c. 2.e and 2.h of Table 3.3.6.1-1;

b. LCO 3.6.1.1. "Primary Containment-Operating";

c. LCO 3.6.1. 2. "Primary Containment Air Locks";

d. LCO 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)";

e. LCD 3.6.1.11. "Containment Vacuum Breakers"; and

f. LCO 3.6.1.12, "Containment Humidity Control".

APPLICABILITY: MODE 4 with average reactor coolant temperature> 200°F.

PERRY UNIT 1 3.10-1 Amendment No. 163

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• • • • 1< SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 163 TO FACILITY OPERATING LICENSE NO. NPF-58 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION CORP.

OHIO EDISON COMPANY PERRY NUCLEAR POWER PLANT, UNIT NO.1 DOCKET NO. 50-440

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated February 22, 2012 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML12054A061), FirstEnergy Nuclear Operating Company (FENOC, the licensee), requested changes to the technical specifications (TS) for Perry Nuclear Power Plant (PNPP), Unit No.1.

The proposed amendment revises TS limiting conditions for operation (LCO) 3.10.1, "Inservice Leak and Hydrostatic Testing Operation," and its associated TS Bases, to expand its scope to include provisions for temperature excursions greater than 200 degrees Fahrenheit CF) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in MODE 4.

The licensee's application states, "this change is consistent with NRC approved Revision 0 to Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-484, Use of TS 3.10.1 for Scram Time Testing Activities." The availability of the TS 3.10.1 revision was announced in the Federal Register on October 27, 2006 (71 FR 63050) as part of the consolidated line item improvement process.

2.0 REGULATORY EVALUATION

2.1 Inservice Leak and Hydrostatic Testing The reactor coolant system (RCS) serves as a pressure boundary and serves to provide a flow path for the circulation of coolant past the fuel. In order to maintain RCS integrity,Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)

Enclosure

-2 requires periodic hydrostatic and leakage testing. Hydrostatic tests are required to be performed once every 10 years and leakage tests are required to be performed each refueling outage. Title 10 of Code of Federal Regulations (10 CFR) Part 50, Appendix G, states that pressure tests and leak tests of the reactor vessel that are required by Section XI of the ASME Code must be completed before the core is critical.

The PNPP is a General Electric boiling-water reactor (BWR/6) designed facility. Standard Technical Specifications (STS) NUREG-1434, "General Electric Plants, BWR/6, Revision 4,"

specifies requirements for LCO 3.10.1 which are similar to those stated in PNPP's TS LCO 3.10.1. The STS LCO was created to allow for hydrostatic and leakage testing to be conducted while in MODE 4 with average reactor coolant temperature greater than 200 degree Fahrenheit (OF) (same value in PNPP's TS), provided certain secondary containment LCOs are met.

The licensee's application states:

The PNPP (TS) are different from the Standard Technical Specifications (STS).

TS 3.10.1 in the STS contains requirements to satisfy the MODE 3 limiting conditions for operation associated with secondary containment. The PNPP TS 3.10.1 MODE 3 requirements are associated with primary containment control.

Due to the PNPP specific design, the more appropriate method for establishing a boundary for control of fission products is the primary containment. The primary containment requirements were included in the PNPP TS when PNPP converted to the Improved STS in 1995 (References 4, 5, and 6). There are no additional TS changes required since TSTF-484 does not impact the MODE 3 req uirements.

TSTF-484 allows a licensee to implement LCO 3.10.1 while hydrostatic and leakage testing is being conducted should the average RCS temperature exceed 200 OF during testing. This allowance does not alter current requirements for hydrostatic and leakage testing as required by Appendix G to 10 CFR Part 50.

2.2 Control Rod Scram Time Testing Control rod's function are to control reactor power level and provide adequate excess negative reactivity to shut down the reactor from any normal operatillg or accident condition at any time during core life. The control rods are scrammed by using hydraulic pressure exerted by the control rod drive (CRD) system. Criterion 10 of Appendix A to 10 CFR Part 50, states that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. The scram reactivity used in design basis accidents and transient analyses is based on an assumed control rod scram time.

NUREG-1434 currently contains surveillance requirements (SR) to conduct scram time testing when certain conditions are met in order to ensure that Criterion 10 of Appendix A to 10 CFR Part 50 is satisfied. STS and PNPP's SRs 3.1.4.1 and 3.1.4.4 require scram time testing to be conducted following a shutdown greater than 120 days, and scram time testing to be conducted following work on the CRD system or following fuel movement within the affected core cell,

-3 respectively. Both SRs must be performed at reactor steam dome pressure greater than or equal to 950 pounds per square inch gauge (psig) and prior to exceeding 40 percent rated thermal power (RTP).

TSTF-484 allows modification of LCO 3.10.1 to allow SRs 3.1.4.1 and 3.1.4.4 to be conducted in MODE 4 with average reactor coolant temperature greater than 200 of. This modification does not alter the means of compliance with Criterion 10 of Appendix A to 10 CFR Part 50. Scram time testing would still be performed in accordance with LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown." As stated by the licensee in its application, the proposed changes are in accordance with the TSTF-484, Revision O.

3.0 TECHNICAL EVALUATION

The existing provisions of LCO 3.10.1 allow for hydrostatic and leakage testing to be conducted in MODE 4 with average RCS temperature greater than 200 of, while imposing MODE 3 secondary containment requirements. Under the existing provision, LCO 3.10.1 would have to be implemented prior to hydrostatic and leakage testing. If LCO 3.10.1 was not implemented prior to the subject testing, as a result, the testing would have to be terminated if average RCS temperature exceeded 200 of during the conduct of the test. TSTF-484 modifies TSs to allow a licensee to implement LCO 3.10.1 while hydrostatic and leakage testing is being conducted, should the average RCS temperature exceeds 200 OF during testing. The modification wi" allow completion of testing without the potential for interrupting the test in order to (a) reduce reactor vessel pressure, (b) cool the RCS, and (c) restart the test below 200 of. Since the current LCO 3.10.1 allows testing to be conducted while in MODE 4, with average RCS coolant temperature greater than 200 of, the proposed change does not introduce any new operational conditions beyond those currently allowed.

SRs 3.1.4.1 and 3.1.4.4 require that control rod scram time be tested at reactor steam dome pressure greater than or equal to 950 psig and before exceeding 40 percent RTP. Performance of control rod scram time testing is typically scheduled concurrent with inservice leak or hydrostatic testing while the RCS is pressurized. Because of the number of control rods that must be tested, it is possible for the inservice leak or hydrostatic test to be completed prior to completing the scram time test. Under existing provisions, scram time testing must be suspended if the testing cannot be completed during the LCO 3.10.1 required inservice leak or hydrostatic test. Additionally, if LCO 3.10.1 is not implemented and average RCS temperature exceeds 200 of while performing the scram time test, scram time testing must also be suspended. In both situations, scram time testing is resumed during startup and is completed prior to exceeding 40 percent RTP.

The allowance provided by TSTF-484, enables the licensee to complete scram time testing initiated during inservice leak or hydrostatic testing. As stated earlier, since the current LCO 3.10.1 allows testing to be conducted while in MODE 4, with average RCS temperature greater than 200 of, the proposed change does not introduce any new operational conditions beyond those currently allowed. It should be noted that completion of scram time testing prior to reactor criticality and power operations results in a more conservative operating philosophy with attendant potential safety benefits.

- 4 It is acceptable to perform other testing concurrent with the inservice leak or hydrostatic test provided that this testing can be performed safely and does not interfere with the leak or hydrostatic test. However, it is not permissible to remain in LCO 3.10.1 solely to complete such testing following the completion of inservice leak or hydrostatic testing and scram time testing.

Since the tests are performed with the reactor pressure vessel (RPV) nearly water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low. Small leaks from the RCS would be detected by inspections before a significant loss of inventory occurred. In addition, two low-pressure emergency core cooling systems (ECCS) injection/spray subsystems are required to be operable in MODE 4 by TS 3.5.2, "ECCS Shutdown". In the event of a large RCS leak, the RPV would rapidly depressurize and allow operation of the low pressure ECCS. The capability of the low pressure ECCS would be adequate to maintain the fuel covered under the low decay heat conditions during these tests.

Also, LCO 3.10.1 requires that secondary containment and standby gas treatment system be operable and capable of handling any airborne radioactivity or steam leaks that may occur during performance of testing.

Variations or deviations from the TS changes described in TSTF-484. Revision 0:

1. The existing provisions of LCO 3.10.1 allow for hydrostatic and leakage testing to be conducted in MODE 4 with average RCS temperature greater than 200 OF, while imposing MODE 3 secondary containment requirements.

The licensee's application states:

The PNPP (TS) are different from the Standard Technical Specifications (STS).

TS 3.10.1 in the STS contains requirements to satisfy the MODE 3 limiting conditions for operation associated with secondary containment. The PNPP TS 3.10.1 MODE 3 requirements are associated with primary containment control.

Due to the PNPP specific design, the more appropriate method for establishing a boundary for control of fission products is the primary containment. The primary containment requirements were included in the PNPP TS when PNPP converted to the Improved STS in 1995 (References 4, 5, and 6). There are no additional TS changes required since TSTF-484 does not impact the MODE 3 requirements.

The licensee further states:

FENOC has reviewed the safety evaluation (SE) published on October 27,2006 (71 FR 63050) as part of the [consolidated line item improvement process] CLlIP Notice of Availability. FENOC has concluded that the technical justifications presented in the SE prepared by the NRC staff are applicable to PNPP using primary containment requirements and therefore justify this amendment for the incorporation of the proposed changes to the PNPP TS.

An NRC SE for license amendment number 69 dated June 23, 1995 (ADAMS Accession No. ML021830716) approved the change described above.

- 5 In response to a NRC staffs request for additional information dated February 22,2012 (ADAMS Accession No. ML12054A061), the licensee provided the technical and licensing bases for the exclusion of operability requirements in TS 3.10.1 for secondary containment and the annulus exhaust gas treatment (AEGT) system, in a letter dated March 8,2013 (ADAMS Accession No. ML13070A202). During the PNPP TS conversion, TS 3.10.1 was revised by replacing the secondary containment and standby gas treatment requirements with primary containment requirements, which were approved by the NRC by license amendment No. 69, dated June 23, 1995. In summary, the licensee's basis for the TS change is due to PNPP's specific design features, such that the use of the primary containment is the more appropriate boundary for control of fission products. Below is a brief description of PNPP's containment design and loss of-coolant accident (LOCA) analysis, and the NRC staffs assessment of the exclusion of operability requirements in TS 3.10.1 for secondary containment and the AEGT system.

PNPP's primary containment is a pressure retaining structure composed of a free-standing steel cylinder with an ellipsoidal dome, secured to a steel lined reinforced concrete foundation mat.

The primary containment is surrounded by a reinforced concrete structure (the shield building).

The primary containment is designed to contain radioactive material that might be released from the RCS following a LOCA. The primary containment ensures a high degree of leak tightness during normal operating and accident conditions.

The secondary containment at PNPP consists of the volume (or annulus) between the shield building and the primary containment. The secondary containment is designed to collect fission product leakage from the primary containment during and following a postulated design basis accident and delays the fission product leakage until processing it through the AEGT system to minimize the escape of radioactive particles to the environment, by maintaining the annulus air space at a slight negative pressure. The AEGT system has two subsystems, both consisting of a high efficiency particulate (HEPA) pre-filter, a charcoal adsorber, and a HEPA post-filter.

Mechanical and electrical lines that enter or exit the primary containment pass through penetrations located in both the primary containment and shield building walls. The majority of the penetrations bypass the secondary containment, and any leakage would not be processed by the AEGT system.

The design basis of the primary containment after a LOCA event is to maintain its integrity such that the design limits are not exceeded. The primary containment design limit at PNPP is 15 pounds per square inch and the calculated peak primary containment pressure during a design basis LOCA event, does not reach this pressure, thus the design limit is not exceeded. The radiological aspect of the LOCA analysis was revised as a result of applications of the revised accident source term at PNPP (license amendment No. 103, dated March 26, 1999 (ADAMS Accession No. ML021840462)), which includes postulating a fission product release to the containment. The licensee's dose calculations made a number of assumptions including: no credit for iodine removal by the AEGT system charcoal absorbers, increased engineered safety feature (ESF) system leakage outside of containment, and increased secondary containment bypass leakage. The analysis considered four potential fission product release pathways:

1) Main steam isolation valve leakage;

2) Secondary containment bypass leakages;

3) Post-accident water leakage from ESF system leakage outside of containment, and; and

- 6

4) Containment leakage.

The licensee's radiological analysis, assumes the fission product leakage from the main steam lines to be released directly to the environment. The analysis also assumes that a percentage of primary containment leakage bypasses the secondary containment, in which a limit is provided on this value in the TS. In addition, leakage of water from ESF components located outside of the primary containment would release fission products outside of the secondary containment during the recirculation phase of long-term core cooling following the LOCA The only fission product release pathway considered in the radiological analysis, in which the leakage is collected in the secondary containment for processing, is the containment leakage. The other three release pathways are assumed to bypass the secondary containment and the AEGT system. However, even though the containment leakage release pathway is considered to enter the secondary containment, no credit is taken for iodine removal by the AEGT system charcoal absorbers.

The NRC staff reviewed the conditions used in the licensee's LOCA analysis and compared them to the plant conditions when TS 3.10.1 would be entered. For the LOCA analysis, the plant is assumed to be at high power, high reactor pressure, and high steam flow. The conditions for TS 3.10.1 is at the end of a refueling outage when the reactor vessel is nearly water solid, at relatively low decay heat values, in MODE 4 (Cold Shutdown), and the reactor core is not critical.

The relative stored energy in the reactor core during this period will be very low and if a leak would occur during this testing, there should be insufficient energy to pressurize the containment and provide the driving force for leakage approaching the primary containment design value.

The NRC staff agrees with the licensee's assessment that implementing primary containment requirements in lieu of secondary containment and AEGT system requirements establishes more robust containment conditions. Leakage into the secondary containment through the containment shell and other paths is expected to be minimal due to the limited energy conditions postulated during the period that TS 3.10.1 would be used. In addition, the licensee stated that this limited leakage could be processed by the fuel handling area ventilation system, which serves the building that is connected to the annulus (secondary containment) when the shield beams are not installed. Based on the evaluation above, the NRC staff finds the exclusion of operability requirements in TS 3.10.1 for secondary containment and the AEGT system acceptable.

In summary, the protection provided by the normally required MODE 4 applicable LeOs, in addition to the primary containment requirements required to be met by LCO 3.10.1) minimizes potential consequences in the event of any postulated abnormal event during testing.

In addition, the requested modification to LCO 3.10.1 does not create any new modes of operation or operating conditions that are not currently allowed. Therefore, the NRC staff finds the proposed change acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 [or changes a

-7 surveillance requirement]. The NRC staff has determined that the amendment involves no significant increase in the amounts, no significant change in the type of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occl,Jpational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (77 FR 43377; July 24,2012). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: R. Grover, NRR B. Lee, NRR Date of issuance: April 18, 2013

Mr. Vito A. Kaminskas Site Vice President7 FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant Mail Stop A-PY-A290 P.O. Box 97,10 Center Road Perry,OH 44081-0097 SUB...IECT: PERRY NUCLEAR POWER PLANT, UNIT NO.1 - ISSUANCE OF AMENDMENT RE: REVISE TECHINCAL SPECIFICATION 3.10.1, "INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION" (TAC NO. ME8048)

Dear Mr. Kaminskas:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 163 to Facility Operating License No. NPF-58 for the Perry Nuclear Power Plant, Unit NO.1. This amendment revises the technical specifications (TS) in response to your application dated February 22, 2012, and supplemented by letter dated March 8, 2013.

This amendment revised TS 3.10.1, "Inservice Leak and Hydrostatic Testing Operation," and the associated TS Bases to expand its scope to include provisions for temperature excursions greater than 200 degree Fahrenheit as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in MODE 4.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440

Enclosures:

1. Amendment No. 163 to N PF-58

2. Safety Evaluation cc w/encls: Listserv DISTRIBUTION:

PUBLIC LPL3-2 R/F RidsNrrPMPerry Resource RidsNrrDssSlsb Resource RidsOgcRp Resource RidsNrrLASRohrer Resource RidsNrrDorlLpl3-2 Resource RidsRgn3MailCenler Resource RidsNrrDorlDpr Resource RidsNrrDssScvb Resource RidsAcrsAcnw_MailCTR Resource Accession No'.. M113017A431 OFFICE LPL3-2/PM LPL3-2/LA DSS/STSB/BC DSS/SCVB/BC NAME MMahoney SRohrer REIliol RDennig DATE 4/11/13 4/11/13 4/16/13 4/17113 OFFICE LPL3-2/BC LPL3-2/PM NAME JBowen MMahoney DATE 4/17113 4/18/13 OFFICIAL RECORD COPY