ML20126M492

Proposed Tech Spec Changes Providing New Reactor Vessel Pressure/Temp Limits for Heatup,Cooldown & Hydrostatic Test Due to Expiration of Current Pressure/Temp Limits

ML20126M492
Person / Time
Site:	Palisades
Issue date:	06/14/1985
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18051B427	List: ML20126M492
References
NUDOCS 8506200355
Download: ML20126M492 (8)

Text

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3.1.2 Heatup and Cooldown Rates (Contd)

(2)

(Contd) surveillance program capsule which was removed at t.he beginning of the Cycle 3.

For purposes of determining fluence at the reactor vennel beltline until a fluence of 1.3 x 1019nyt in realized at the inner vessel wall at the beltline region, the l

following basis is established:

5.9 x 10 nyt calculated at 19 the reactor vessel beltline for 2540 W r 40 yearn at an 80%

e load factor. This conversion han resulted in a correlation of 1.989 x 1012nyt per 1 Wd '

e (3)

The limit lines in Figuren 3-1, 3-2 and 3-3 are based on the requirements of Reference 9. Paragraphs IV.A.2 and IV.A.3.

l These lines reflect a preservice hydrostatic tent prennure of i

2400 psig and a vennel flange material reference temperature l

of 60*F.

Basis

(

All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and pren-nura changes. U} These cyclic loads are introduced by normal unit load transients, reactor trips and start-up and shutdown operation. During unit start-up and shutdown, the raten of temperature and pressure changen l

are limited.

A maximum plant heatup and cooldown rate of 100'F per hour is consintent with the design number of cycles and natisfien strean ilmita I

for cyclic operation.(

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The reactor vennel plate and material opposite the core has been purchased to a specified Charpy V-Notch tant renutt of 30 ft-lb or greater at an NDTT of + 10'F or 1enn.

The vennel weld han the highent l

l RT f plate, weld and HAZ materials at the fluence to which the NDT Figuren 3-1, 3-2 and 3-3 apply.00) The unirradiated RT han been NDT j

determined to be -56*F.U I}

An RT of -56*F in uned an an unirrad-l NDT isted value to which irradiation effects are added.

In addition, the plate han been 100% volumatrically inspected by ultranonic tent uning i

Proposed 3-5 t

P l

t 3.1.2 Heatup and Cooldown Raten (Contd)

Basis (Contd) both longitudinal and shear wave methods. The remaining material in

(

the reactor vennel, and other primary coolant system components, meets the appropriate design code requirements and specific component func-tion and has a maximus NDTT of +40'F.(5)

As a result of fant neutron irradiation in the region of the core, there will be an increase in the RT with operation. The techniques used to predict the integrated fast neutron (E > 1 MeV) fluxen of the reactor vennel are described in Section 3.3.2.6 of the FSAR and also in Amendment 13,Section II, to the FSAR.

Since the neutron spectra and the flux measured at the samples and reactor vennel inside radiun should be nearly identical, the mensured transition shift from a nample can be applied to the adjacent section of the reactor vennel for later stagen in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vennel will be obtained from the measured sample exposure by application of the calculated azimuchal neutron flux variation. The maximum integrated fant neutron (E > I MeV) exposure of the reactor ven-I9 sel in computed to be 5.9 x 10 nyt for 40 yearn' operation at 2540 MWg and 80% load factor. The predicted RT shift f r the base metal han NDT been predicted baned upon surveillance data and the appropriate US NRC l

Regulatory Guide.(0) The actual shift in RT will be established NDT periodically during plant operation by tenting of reactor vennel material samples which are irradiated cumulatively be necuring them near the innide l

wall of the reactor vennel an dancribed in Section 4.5.3 and Figure 4-11 of the FSAR. To compennate for any increase in the RT caused by irradia-tion, limita on the prennure-temperature relationship are periodically changed to stay within the strenn Ifmits during heatup and cooldown.

Reference 7 providen a procedure for obtaining the allowable loadinan for ferritic preneure-retaining materials in Clann I componenta. Thin l

procedure in based on the principles of linear elantic fracture mechan-l ica and involven a attenn intennity factor prediction which in a lower bound of neatic, dynamic and crack arrent critical valven.

The nerenn 3-6 Proposed L

3.1.2 Heatup and Cooldown Rates (Contd)

Basis (Contd) intensity factor computed is a function of RTg. operating tempera-ture, and vessel wall temperature gradients.

Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Refer-ence 7.

The limit lines of Figures 3-1 through 3-3 consider a 54 psi pressure allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline.

In addition, for calculational purposes, 5'F and 30 psi were taken as measurement error allowances for temperature and pressure, respectively.

By Reference 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting.

It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested.

At these locations, fluence attenu-ation and thermal gradients have been evaluated. During cooldown, the 1/4 thickness location is always more limiting in that the RT i'

NDT higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate.

Figures 3-1 through 3-3 define stress limitations only from a fracture mechanic's point of view.

Other considerations may be more restrictive with respect to pressure-temperature limits.

For normal operation, other inherent plant charac-teristica may limit the heatup and cooldown rates which can be achieved.

Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown raten to less than 60'F per hour.

The revised pressure-temperature limita are applicable to reactor vennel inner wall fluences of up to 1.3 x 1019nyt. The application of appropriate fluence attenuation factorn (Reference 10) at the 1/4 and 3/4 thicknean locationa results in RT a

en of 223 F and 170'F, NDT respectively, for the limiting weld material.

The criticality condition Propo m!

3-7

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3.1.2 Heatup and Cooldown Rates (Contd) g Bania (Contd) which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 352'F.

The most limiting wall location is at 1/4 thickness. The minimum criticality temperature, 352*F is the minimum permissible temperature for the inner-vice system hydrostatic pressure test. That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.

The restriction of heatup and cooldown rates to 100'F/h and the maint-enance of a pressure-temperature relationship to the right of the heatup, cooldown and innervice tent curves of Figuren 3-1, 3-2 and 3-3, respec-tively, ensuren that the requirements of References 6, 7, 8 and 9 are met.

The core operational limit applies only when the reactor is critical.

The criticality temperature is determined per Reference 8 and the core operational curves adhere to the requirements of Reference 9.

The innervice tent curven incorporate allowances for the thermal gradients associated with the heatup curve used to attain inservice test pressure.

Theea curves differ from heatup curves only with respect to margin for primary membrane stress.(I}

For heatup rates lenn than 60'F/h the hypothetical O'F/h (inothermal heatup) at the 1/4 T location in con-trolling and heatup curves converge. Cooldown curven crosa for various cooldown raten, thus a composite curve is drawn. Due to the shifts in i

RTNDT, NDTT requirements associated with nonreactor vessel materials are, for all practical purposes, no longer limiting.

Referencen (1) FSAR, Section 4.2.2.

(2) ASME Doller and Prennure Vennel Code,Section III A-200>.

(3)

Battelle Columbus Laboratories Report, "Palisaden Frenmure Vensel Irradiation Capsule Programt Unitradiated Hochanical Propertion "

August 25, 1977.

(4)

Battelle Columbus Laboratorien Report, "Palisaden Nuclear Plant Reactor Vessel Surveillance Program Capsula A-240," Harch 13, 1979 Aubmitted to the NRC by Consumern Power Company letter dated July 2, 1979.

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3.1.2 Heatup and Cooldown Raten (Contd)

Re.forences (Contd)

(5)

FSAR, Section 4.2.4.

(6)

US Nuclear Regulatory Connaission, Regulator Guide 1.99 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," July, 1975.

(7)

ASME Boiler and Pressure Vessel Code,Section III, Appendix G,

" Protection Against Non-Ductile Failure," 1974 Edition.

(8)

US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, " Pressure-Temperature Limits."

(9) 10 CFR Part 50, Appendix C, " Fracture Toughness Requirements "

May 31, 1983.

(10) US Nuclear Regulatory Commission Regulatory Guide 1.99 Draft Revision 2. April, 1984.

(11) Combustion Engineering Report CEN-189, December, 1981.

3.1.3 Minimum Conditions for Criticality a)

Except during low-power physica test, the reactor shall not be made critical if the primary coolant temperature in below 525'F.

b)

In no cane shall the reactor be made critical if the primary coolant temperature in below 352*F.

c)

When the primary coolant temperature is below the minimum tempera-ture specified in "a" above, the reactor shall be suberitical by an amount equal to or greater than the potential reactivity inner-tion due to depreneurization.

d)

No more than one control rod at a time shall be exercised or with-drawn until after a steam bubble and normat water level are entab-linhed in the prennuriser.

e)

Primary coolant boron concentration shall not be reduced until aftet a ateam bubble and normal water level are establinbed in the prennutiner.

Masta At the beginning of life of the initial fuel cycle, the moderator temperature coefficient la expected to be slightly negative at operating temperaturen with all control roda withdrawn,UI Ilowever, the uncer-tainty of the eniculation in such that it in possible that a slightly positive coefficient could extat.

3 12 Proposmi