ML17321A704

Amend 69 to License DPR-74,revising Tech Specs by Changing Heatup & Cooldown Curves to Reflect Most Current Reactor Vessel Matl Surveillance Capsule Analysis

ML17321A704
Person / Time
Site:	Cook
Issue date:	06/27/1985
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML17321A705	List: ML17321A704 ML17321A706
References
NUDOCS 8507120176
Download: ML17321A704 (6)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

69 License No.

DPR-74 1.

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

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated February 14, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 (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 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.

DPR-74 is hereby amended to read as follows:

850Ei27 050003ia PDR 8507i20i76 PDR ADOCK P

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 69, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLE R

REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 27, 1985 even g g, 6>f

'perating Reacto s Branch ¹I Division of Licen ng

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

69 FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NO. 50-316 Revise Appendix A as follows:

Remove Pa es 3/4 4-25 3/4 4-26 B3/4 4-6 Insert Pa es 3/4 4-25 3/4 4-26 83/4 4-6

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REACTOR COOLANT SYSTEM HEATUP LIMITATIONS 1 APPLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS. (MARGINS OF 60 PSIG AND 10'F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.)

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FIGURE 3.4.2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE UMITS VERSUS 100'F/HR RATE, CRITICALITYLIMITAND HYDROSTATIC TEST LIMIT jji!

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FIGURE 3 4-3 REACTOR COOLANT SYSTEM PRESSURE"TEMPERATURE LLIITS VERSUS COOLOOWN RATES

REACTOR COOLANT SYSTEM BASES 3/4. 4. 9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case.

There are several factors which influence the postulated location.

The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall.

During cooldown the bending stress profile is reversed.

In addition, the material tough-ness is dependent upon irradiation and temperature and.therefore the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.

The heatup limit curve, Figure 3.4>>2, is a composite curve which was prepared by determining the most conservative case, with either the inside 0

or outside wall controlling, for any heatup rate up to 100 F per hour.

The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall.

The heatup and cooldown curves were pre-pared based upon the most limiting value of the predicted adjusted reference temperature at the end of 12 EFPY.

The reactor vessel mategials have been tested to determine their initial RT g the results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference tem-

perature, based upon the fluence and copper content of the material in

question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include pre-dicted adjustment" for this shift in RTN at the cnd of 12 EFpY.

Qs well as adjustments for possible errors in the pressur>>

and temperature sensing NDT instruments-The heatup and cooldown curves are applicable to low leakage cores.

Do C.

COOK - UNIT 2 B 3/4 4-6 Amendment No. 69