ML17321A450

Proposed Tech Specs,Changing Plant Heatup & Cooldown Curves to Reflect Recent Reactor Vessel Matl Surveillance Capsule Analyses Performed by Swri

ML17321A450
Person / Time
Site:	Cook
Issue date:	02/14/1985
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17321A449	List: ML17321A448 ML17321A450
References
NUDOCS 8502210036
Download: ML17321A450 (12)

Text

TT C

0 TOAE'R T

N N E DESC PT NS Ch e

Page 3/4 4-27; Figure 3.4-2 Page 3/4 4-28; Figure 3.4-3 Page B 3/4 4-6; Seotion 3/4 4.9 Page B 3/4 4-11; Section 3/4 4.9,

4. 10 Page 3/4 4-25; Figure 3.4-2 Page 3/4 4-26; Figure 3.4-3 Page B 3/4 4-6; Section 3/4 4.9 Figures 3.4-2 and Figure 3.4-3 are the heatup and cooldown curves for Unit Nos.

1 and 2.

These curves are composite curves prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 12 EFPY.

The revisions to the curves reflect the analyses done as a result of the capsule (Y) testing by Southwest Research Institute.

The assumptions and limitations used in the development of the curves are noted in the revised bases for both Units.

The bases for Pressure/Temperature limits were re>>written to refleot the current analyses.

Also, the Unit No.

1 bases was revised to read similarly to the Unit No.

2 writeup for consistency.

The proposed change may result in an additional limitation, restriction or control not presently included in the Technical Specifications and constitutes a change to conform to changes in the'federal regulations.

Me believe that the results of the change are clearly within all acceptable criteria since they are the direct result of an evaluation using methods acceptable to the NRC.

As such, the proposed changes are found not to involve a significant hazards consideration.

8502210036 8502i4 PDR ADOCK 050003i5 P

PDR

ATTACHMENT NO.

2 TO AEP:NRC:0894

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.or 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 s'urface 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 o'r outside wall controlling, for any heatup rate up to 60 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 oredicted adjusted reference temperature at the end of 12 EFPY.

The reactor vessel materials have been tested to determine their initial RT

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 NDT cause an increase in the RT Therefore, an adjusted reference tem-..

NDT

perature, based upon the fluence and cooper 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 cu res of Ficures 3.4-2 and 3.4-3 include ore-dicted adjustments for this shift in RT at the end of 12 EFPY, as well MDT as adjustments or possible errors in the pressure and temperature sensing instruments.

D. C.

COOK - UNIT 1 B 3/4 4-6 Amendment 'o.

REACTOR COOLANT SYSTEM BASES The actual shift in RT of the vessel matezial will be established

.NDT periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the zeactor vessel.

The heatup and cooldown curves must be recalculated when the 6 RT determined from the surveillance capsule is different from the calcuIated 6 RT for the equivalent capsule radiation exposure.

NDT NDT The pressure-temperature limit, lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been pro-vided to assure compliance with the minimum temperature requizements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERASELLTY of two PORYs, one PORV and the RHR safety valve, ar an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 188~F.

Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1)'he start of an idle RCP with the secondary water tempera-ture of the steam generator less than or equal to 504F above the RCS cold leg temperatures or (2) the start of a chargirig pump and its injection into a water solid RCS.

D. C.

COOK-UNIT 1 B 3/4 4-11 Amendment:

No.

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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 mateqials have been tested to determine their initial RT

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 fluPnce 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 adjustments for this shift in RT at the end of 12 EFpY, as well NDT as adjustments for possible errors in the pressur>>

and temperature sensing instruments.

D ~ C.

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

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ATTACHMENT NO.

3 TO AEP:NRC:0894

2600 2400 2200 2000 1800 16OO 1400 1200 1OOO 5

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'EACTOR COOLANT SYSTEM HEATUP LIMITATIONSAP PLICABLE FOR FIRST 12 EFFECTIVE FULL POWER

YEARS, (MARGINS Of 60 PSIG AND 10'F NE IN CLUDED FOR POSSIBLE INSTR)WENT ERRORs)

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FIGURE l.

PROPOSED HODIFICATION TO DONALD C.

COOK UNIT NO. l REACTOR COOLANT SYSTEH PRESSURE-TEHPERATUAE LIHITS VERSUS 60'F/IIOUR RATE CRITICALITY LIHIT AND HYDROSTATIC TEST LIHIT, l2 EFPY, FOR FRACTURE CONTAOt AT STRUCTURAL DISCONTINUITIES

SOUTH WEST RESEARCH INSTITUTE 408 T OffICE OIIAWKII28610

~ 8220 CULfBRA IIOAO ~ SAN AHTOIIIO, T8 XAS. VSA 74284

~ 16121 hhe 6111 ~ TtLE'X 78 ~ f267

~ilJ Pi IJT

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M~AP>'.et Pf Materials Sciences June 12, 1984 Hr. John R.

Je Mechanical Eng ering Department American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215

Subject:

Heatup P-T Limits for up to 12 EFPY Operation, D.

C.

Cook Unit No.

1

Reference:

Letter of May 30,

1984, E.

B. Norris to J.

R.

Jensen

Dear John:

Enclosed is a revised Figure 1 for placement in the referenced letter.

Although the 10'F allowance for instrument error had been in-cluded in the analysis described in the referenced letter, the -60 psi instrument error had not.

Therefore, I lowered the horizontal line from 625 psi to 565 psi.

I decided not to change Figure 2 of the referenced letter because the Figure 1 approach is acceptable and gives the plant operations more leeway.

In addition, any lower bound curve which I might draw, such as the dotted line in Figure 2, is strictly arbitrary.

Also enclosed are copies of the computer printouts which we used in the construction of the 12 EFPY P-T limit curves.

I have noted where the closure flange region limits impact on the normal values.

Should you have any questions, please do not hesitate to contact EBN:klc Enclosures Sincerely,

)+1 E.

B.

No is J.

Staff Engineer cc:

C.

E. Lautzenheiser, SwRI S,

)

SAN ANTONIO, TEXAS wITH officI$

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TexAs, AIIO wA5HIaaofoaa, 0 c,