Text

Amend 81 to License NPF-3,changing Surveillance Requirement 4.4.9.1.2 & Associated Table 4.4-5 to Reflect Changes in Cycle Length & Fuel Loading Scheme
	ML20101L277
Person / Time
Site:	Davis Besse
Issue date:	12/17/1984
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Toledo Edison Co, Cleveland Electric Illuminating Co
Shared Package
ML20101L281	List: ML20101L277; ML20101L292;
References
	NPF-03-A-081 NUDOCS 8501020182
	Download: ML20101L277 (9)
	v • d • e

[furoyM UNITED STATES

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TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 81 License No. NPF-3 1.

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

A.

The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated August 27, 1984, complies with the standards and requirements 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 Comission; C.

There is reasonable assurance (i) that the activities authorized by

'this amendment can be conducted without endancerina 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 common 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 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-3 is hereby amended to read as follows:

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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 81, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGU ATORY COMMISSION oh ' F. Stolz, Chief Op ating Reactors Branch #4

' ision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December l_7, 1984

ATTACHMENT TO LICENSE AMENDMENT NO.g1 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

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, i. i 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0gCi/ gram Dose Equivalent 1131 DAVIS-BESSE, UNIT 1 3/4 4-23 m

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE 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.

A maximum heatup of 100'F in any one hour period, and b.

A maximum cooldown of 100*F in any one hour period.

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 fracture toughness properties 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS T

and pressure to less than 200*F and 500 psig, respectively, within t8E9following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 surv_eillance specimens rep'resentative of the vessel materials shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5.

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. 81

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60 100 140 180 220 260 300 340 380 420 INDICATED REACTOR COOT. ANT SYSTEM TEMPERATURE. Tc, *F Figure 3.4-1 Reactor Coolant System Pressure-Temperature Heatup and Cooldown Limits for Inservice Leak and Hydrostatic Tests for the First 5 EfPY

ty TABLE 4.4-5 D

E REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCIIEDULE a,

E Sequence Time of Withdrawal E

Ci First Earliest of:

1.5 EFPY; capsule fluence > 5 x 1018,fe,2; r

highest RT f an encapsulated material equals SOF.

HDT Second Earliest of: 3 EFPY; capsule fluence midway between that of the first and third capsules.

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Third Earliest of: 6 EFPY; capsule fluence corresponds to that of the EOL fluence of the reactor vessel 1/4T location.

l Fourth Schedule to be submitted for NRC approval prior to Cycle 6 operation.

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Fifth Schedule to be submitted for NRC approval prior to C,ycle 6 operation.

Sixth Schedule to be submitted for NRC approval prior to Cycle 6 operation.

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REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were detennined for those materials for which sufficient amounts of material were available.

The unirradiated impact properties and residual elements of the beltline region materials are listed in Bases Table 4-1.

The adjusted reference teipperatures are calculated by adding the predicted radiation-induced ART [the respective neutron fYb. The predicted ARTce and copper and and the unirradiated RT usiN arp calculated Bases Figure 4-1 illustrates the calculated peak neutron fluence, at several locations through the reactor vessel beltline region wall and at j

the center of the surveillance capsules as a, function of exposure time.

Bases Figure 4-2 illustrates the design curves for predicting the radiation-induced 6RT as a function of the material's copper and phosphorus content an$Ddeutron fluence. The adjusted RT

's of the beltline region materials at the end of the fifth full phNr year are listed in Buses Table 4-1.

The adjusted RT

's are given for the 1/4T and 3/4T (T is wall thickness) vessel wall YNations. The assumed RT of the closure head region is 40*F and the outlet nozzle steel forginhT is 60 F.

During cooldown at the higher temperatures, the limits are imposed by thennal and loading cycles on the steam generator tubes. These limits are segments D-E and D-F of the limit lines on Figures 3.4-2 and 3.4-4, respectively. These limits will not require adjustments due to the neutron fluences.

Figure 3.4-2 presents the pressure-temperature limit curve br 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-temperature limit curve for nonnal cooldown.

Figure 3.4-4 presents the pressure-temperature limit curves for heatup and cooldown for inservice leak and hydrostatic testing.

All pressure-temperature limit curves are applicable up to the fifth 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.

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, REACTOR C0OLANT SYSTEM 8ASES The number of reactor vessel 4rradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5.t.-The withdrawal schedule is based on four considerations:

(a) uncover possible technical anomalies as early in life as they can be detected (and of first fuel cycle), (b) define the material properties needed to perfom 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 the inplace annealing becomes necessary (d) provide material property data corresponding to the reactor vessel

.line surface conditions at the end of service. The withdrawal schedule of Table 4.4-5 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. Table 4.4-5 is designed to meet the requirements of ASTM E185-82.

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Amendment flo. 81 i

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ML20101L277

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