ML20097F574

Proposed TS Section 4.4 Re Reactor Vessel Matl Surveillance Specimen Withdrawal

ML20097F574
Person / Time
Site:	Wolf Creek
Issue date:	06/11/1992
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20097F572	List: ML20097F569 ML20097F574
References
NUDOCS 9206150376
Download: ML20097F574 (6)

Text

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' Attachment IV to ET 92 0122 Page 1 of 6 ATTACIMENT IV PROPOSED TECIINICAL SPECIFICATION CilANGES i

9206150376 920611 PDR ADOCK 05000482 P PDR

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Attechment IV to ET 92-0122 Page 2 of 6

, LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE BIOUIREMENTS SECTIOn EfflE TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES........................................... 3/4 4-21 3/4.4.7 CHEMISTRY........................................... 3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............. 3/4 4-23 TABLE 4 4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS.................................... 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY................................... 3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >

1 pCi/ GRAM DOSE EQUIVALENT I-131............... 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................................ 3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.......................... 3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 7 EFPY......................... 3/4 4-30 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 7 ETPY......................... 3/4 4-31

-TABLE 4.4-5 hisAePE+n VC000fr MA9sMAir-suftveEhANC" DELE ED PnOO n1J' '-lMu&RAWMr-seHBDULC . . . . . . . . . . . . . . . . . . . . p/4 4 32 Pressurizer...................................... 3/4 4-33

. Overpressure Protection Systems................. 3/4 4-34 FIGURE 3,4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM.................. 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY................................ 3/4 4-37

.3/4.4.11 REACTOR COOLANk' SYSTEM VENTS........................ 3/4 4-38 3/4.5 EMERGENCY COLE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS........................................ 3/4 5-1 WOLF CREEK - UNIT 1 VIII Amendment No. 40

Attachm2nt IV to ET 92-0122 Page 3 of 6

, 3/4.4.9 PRESSURE / TEMPERATURE LIMITS PEACTOR 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 rigures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 60'T in any 1-hour period for indicated T avg less than or equal to 200'r,

b. A maximum heatup of 100'r in any 1-hour period for indicated Ta vg greater to 200'T,

c. A maximum cooldown of 100'T in any 1-hour period, and

c. A maximum temperature change of less than or equal to lo'r in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY - At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfore an engineering evaluation to determine the effects of the out-of-limit condition on the structural 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 reduce the RCS T av and pressure to less than 200'T and 500peig,respectively,withinthefollowing30 hours.

SURVEILLANCE REOUIREMENTS 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 shall be removed and examined, to deterinine _c_hAnge rial pronerttes,

_as required by 10 Crk Part 50, Appendix H@Hr-% aarr_ a o fit _ d:3 mfd QgdQ( J_.L5') - The results of these examinations shall be used to update - -

Figures 3.4-2, 3.4-3, and 3.4-4.

WOLF CREEK - UNIT 1 3/4 4-29 Amendment No. 40 is ,_4.4-5

Attachinent IV to ET 92-0122 Page 4.of 6 i l

.l

.- REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE  !

CAPSULE VESSEL LEAD  !

NUMBER LOCATIOl{ FACTOR WITHDRAkAL TIME fEFPY)  !

-U 58.5' 4.00 1st Refueling l 241' 3.69 5  ;

V 61' 3.69 9  !

i X 238.5' 4.00 15 i W 121.5' 4.00 Standby -

!: 2 301.5' 4.00 standby i

i e

a h

6 L

F L

T I

WOLF CREEK - UNIT 1 3/4.4-32

Attachmsnt IV to ET 92-0122 Page 5 of 6 R'EACTOR COOLANT SYSTEM PASES

[EESSURE/ TEMPERATURE LIMITS (Continued) l 1

values of ART HDT determined in this manner may be used until the results of the next scheduled capsule f rom the material surveillance program, evaluated

.. cording to ASTM E185, are available. Capsules will be removed in a ce vi to of ASg_L anQ0_ H.

~__r w..m. y 1,- An a s u m uw. , ~ u The ead ctor represents the Ielationship between the fast neutron ilux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained f rem the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead f actor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Veseel Code as required by Appendix G to 10 CFR Part 50.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedureo a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME section III as the reference flav, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this ref erence crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature,-RTNDT, is used and this includes the radiation-induced shift, ART NDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference strece intensity factor, KIRe for the metal temperature at that time. K is obtained from the reference f racture toughness curve, definedinAppenbfxGtotheASMECode. The Kgg curve is given by the equations KIR = 26.78 + 1.223 exp [0.0145(T-RTHDT

• I"U)} (1)

WOLF CREEK - UNIT I B 3/4 4-8 Amendment No. 40 REACTOR COOLANT SYSTEM

i Attachm2nt IV to ET 92-0122  !

Page 6 of 6 I BASES i

i COLD OVERPRESSURE (Continued)

RCP eliminates the possibility of a 50*r difference existing between indicated and actual RCS temperature as a result of heat transport effects. Considering instrument uncertainties only, an indicated RCS temperature of 350*F is suffi-ciently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pres-surizer safety valves provide acceptable and redundant overpressure protection.

The Maximum Allowed PORV Setpoint for the Cold overpressure Hitigation System will be updated based on the results of examinations of reactor vessel rt xH 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable Jevel throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975.

3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function.

The valve redundancy of the Reactor Coolant System vent paths serves to minindre the probability of inadvertent or irreversible actuation while ensuring that a single failure vent valve power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System _ vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of THI Action Plan Requirements," November 1980.

WOFL CREEK - UNIT 1 B 3/4 4-15 Amendment No. 40 1

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