ML20207H381

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Amend 116 to License NPF-3,revising & Deleting Listed Tech Spec Sections,License Conditions,Figures & Table
ML20207H381
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 08/19/1988
From: Perkins K
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207H386 List:
References
TAC-66699, NUDOCS 8808260358
Download: ML20207H381 (19)


Text

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4 UNITED STATES

'7'n NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

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TOLED0 EDIS0N COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET N0. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.116 License No. NPF-3 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amandment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated March 30, 1988, supplemented May 4, 1988 complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations 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 (1) 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 Comission's regulations;

)

D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and by amending paragraphs 2.C.(2) and 2.C.(3)(d) of Facility Operating License No. NPF-3 to read as follows:

8808260350 800819 DR ADOC;'. 0500 6

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, 2.C.(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.116, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

2.C.(3)(d) Prior to operation beyond 10 Effective Full Power Years, the Toledo Edison Company shall provide to the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection again.ct law temperature reactor coolant system overpressure avercs.

3.

This license amendment is effectivt as of its date of issuance and shall be implemented prior to entering M(de 2 following the fifth refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION g

Kenneth E. Perkins, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V, & Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: August 19, 1988

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ATTACHMENT TO LICENSE AMENDMENT NO.116 FACILITY OPERATING LICENSE N0. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Remove Insert 3/4 4-3 3/4 4-3 3/4 4-4a 3/4 4-4a 3/4 4-4b 3/4 4-4b 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 8 3/4 4-7 B 3/4 4-7 B 3/4 4-8 B 3/4 4-8 B 3/4 4-9 8 3/4 4-9 B 3/4 4-10 8 3/4 4-10 B 3/4 4-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-12

REACTOR COOLANT SYSTEM SAFETY VALVES AND ELECTROMATIC RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLf. with a lift setting of < 2525 PSIG.* When not isolated, the pressurizer electromatic relief valve shall have a trip setpoint of > 2390 PSIG and an allowable value of > 2390 PSIG.**

l APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5.

For the pressurizer electromatic relief valve a channel cali-bration check shall be performed every 18 months.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

Allowable value for channel calibration check.-

DAVIS-BESSE, UNIT 1 3/4 4-4 Amendment No. 33,60

2 REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 Decay Heat Removal System relief valve DH-4849 shall be OPERABLE with a lift setting of 5 330 PSIG* and isolation valves DH-11 and DH-12 open and control power to their valve operators removed.

APPLICABILITY: MODES 4 and 5.

ACTION:

A.

With DH-4849 not OPERABLE:

1.

Make the valve OPERABLE within eight hours; or 2.

a.

Within next one hour, disable the capability of both high pressure injection (HPI) pumps to inject water into the reactor coolant system; and b.

Within next eight hours:

1.

Disable the automatic transfer of makeup pump suction to the borated water storage tank on low makeup tank level; and 2.

Reduce makeup tank level to 5 73 inches and reduce reactor coolant system pressure and pressurizer level within the acceptable region on Figures 3.4-2a (in MODE 4) and 3.4-2b (in MODE 5).

B.

With DH-11 or DH-12 closed, open DH-21 and DH-23 within one hour.

C.

With the control power not removed from DH-11 and DH-12, remove the power to the valve operators at the Motor Control Centers within one hour.

SURVEILLANCE REQUIREMENTS 4.4.2 Decay Heat Removal System relief valve DH-4849 shall be determined OPERABLE:

per the surveillance requirements of Specification 4.0.5.

a.

b.

at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying either:

1.

isolation valves DH-11 and DH-12 open with control power removed from their valve operators; or 2.

valves DH-21 and DH-23 open.

The lift seiting pressure shall correspond to ambient conditions of the valve at nominal o;srating temperature and pressure.

DAVIS-BESSE, UNIT 1 3/4 4-3 Amendment No. 57,116

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DAVIS-BESSE, UNIT 1 3/4 4-4b Amendment Fe. $1,116

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l DAVIS-BESSE, UNIT 1 3/4 4-4a Amendment No. 88 116

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE /TEMPERATLRF LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant system (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

A maximum heatup of 50 F in any one hour period, and l

a.

b.

A maximum cooldown of 100*F in any one hour period with cold leg temperature > 270*F and a maximum cooldown of 50*F in any one hour period with cold leg temperature <270*F.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens representative of the vessel materials shall be removed and examined, to determine changes in material properties, at the intervals defined in BAW 1543A.

The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.

DAVIS-BESSE, UNIT 1 3/4 4-24 Amendment No. 8J, 116

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l DAVIS-BESSE, UNIT 1 B 3/4 4-7 Amendment No.116 i

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I REACTOR COOLANT SYSTEM BASES The closure head region is significantly stressed at 7.elatively low temperatures (due to mechanical loads resulting from bolt pre-load). This region largely controls the pressure-temperature listitations of the first several service periods. The outlet nozzles of the reactor vessel also affect the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle which can be two to three times the membrane stresses of the shell. Af ter the first several years of neutron radiation exposure, the RT temperature of the beltline region materials will be high enough so th!Ethebeltlineregionofthereactorvesselwillstarttocontrolthe pressure-temperature limitations of the reactor coolant pressure boundary.

For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained

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through a point-by point comparison of the limits imposed by the closure j

head region, outlet nozzles, and beltline region. The maximum allowable pressure is taken to be the lower pressure of the three calculated pressures.

The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The limit curves were prepared based upon the most limiting adjusted reference temperature of all the beltline region materials at the end of the tenth effective full power year.

The actual shift in RT f the beltline region material will be estab-NDT lished periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irrriiation samples and vessel inside the radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The limit curres must be recalculated when the ART differen: fromthecalculkk}dARTdetermined from the surveillance capsule is NDT r the equivalent capsule radiation 1

exposure.

DAVIS-BESSE, UNIT 1 B 3/4 4-13 Amendment No.116

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I DAVIS-BESSE, UNIT 1 B 3/4 4-9 Amendment No, 116

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REACTOR COOI. ANT SYSTEM BASES The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in BAW 1543A. The withdrawal schedule is based on four considerations:

l (a) uncover possible technical anomalies as early in life as they can be detected (end of first fuel cycle), (b) define the material properties needed to perform the analysis required by Appendix G to 10 CFR 50, (c) reserve two capsules for evaluation of the effectiveness of thermal annealing in the event inplace annealing becomes necessary, (d) provide material property data corresponding to the reactor vessel beltline conditions at the end of service.

This withdrawal schedule is specified to assure compliance with the requirements of Appendix H to 10 CFR 50.

Appendix H references the requirements of ASTM E185 for surveillance program criteria.

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l DAVIS-BESSE, UNIT 1 B 3/4 4-12 Amendment No. 82 116

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REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline regiou materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available. The adjusted reference temperatures are calculated by adding i

the piedicted radiation-induced ART and the unirradiated RT The proceduresdescribedinRegulatory60Ide1.99,Rev.2,wereusggT.g,,

predicting the radiation induced ART as a function of the material's copper and nickel content and neutrogD}1uence.

Figure 3.4-2 presents the pressure-temperature limit curve for normal heatup. This figure also presents the core criticality limits as required by Appendix G to 10 CFR 50. Figure 3.4-3 presents the pressure-temper-ature limit curve for normal cooldown. Figure 3.4-4 presents the pressure-temperature limit curves for beatup and cooldown for inservice leak and hydrostatic testing.

All pressure-temperature limit curves are applicable up to the tenth

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effective full power year. The protection against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.4-2, 3.4-3 and 3.4-4, 1

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DAVIS-BESSE, UNIT 1 B 3/4 4-11 Amendment No. 116