ML20116D293

Proposed Tech Specs Table 4.4-5 Re Reactor Vessel Matl Surveillance Program - Withdrawal Schedule,Reflecting Removal of Schedule for Withdrawal of Reactor Vessel Matl Specimens

ML20116D293
Person / Time
Site:	McGuire, Mcguire
Issue date:	10/28/1992
From:	DUKE POWER CO.
To:
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ML20116D284	List: ML20116D281 ML20116D293
References
NUDOCS 9211050292
Download: ML20116D293 (4)

Text

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4 ATTACHMENT 2 DUKE POWER COMPANY McGUIRE NUCLEAR STATION e

PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS l

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Qp c}menfb Py Io f 3 REACTOR COOLANT SYSTEM BASE 5-PRESSURE / TEMPERATURE LIMITS (Continued)

The fracture toughness properties of the ferritic materia.ls in the reactor vessel are determined in accordance with the NRC Standard Review Plan, Asis E185 73, and in accordance with adoitional reactor vessel requirements.

These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section !!! of the A$ME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924 A; " Basis for Heatup and Cocidown Limit Curves # April 1975."

value of the nil-ductility reference temperature, R'iHeatup and cooldown NOT' 't

'h' '"U Of th' ef fective full power years (EFPY) of service life identifiect on the applicable technical specification figure.

The 10 EFPY service iife period is chosen socn that the limiting RTNOT at the 1/4T location in the core region is greater tnan j

the RT f the limiting unirradiated material, The selection of such a NOT limiting RTNOT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

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

Reactor operation and resulta" fast neutron (E greater than 1 MeV) irradiation can cause an increase in we RT Therefore, an adjusted reference temperature, NOT.

based upon the-fluence, cooper content, and phosph_ ate content of the material in question, can be predicted using Figure B 3/4.4-1 cnd the largest value of ARTNOT.

For Unit 1, the adjusted reference termperature has been computed by Regulatory Guide 1.99, Revision 2.

For Unit 2, the adjusted reference temperature has ~een computed as discussed in WCAP-11029.

The heatup and u

cooldown limit curves of Figures 3.4-2, 3.4-3 3.4-4 and 3.4-5 include predictec adjustments for this shift in-R!NOT at the end of the identified' service life.

Adjustments for possible errors in the pressure and temperature sensing instruments are included when stated on the applicable figure.

Values of ART determined in this manner may be used until the results NOT from the material surveillance program, evaluated according to ASTM E185, are availabie.

Capsules will be removed in accordance with the requirements of A$TH E185-73 and 10 CFR 50, Appendix H.

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The lead factor represents the relationship between the fast neutron flux density at the location of,the capsule and the inner wall of the pressure vessel.

Therefore, the results obtained from the surveillance spt-imens can be used to predict the future radiation damage to

'the pressure vessel material by using the lead factor and the withdrawal time of the capsule.

The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ART NOT NOT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cool-down rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50,.and these metnoas are discussed in detail in WCAP-7924-A.

Amendment No. M0(Unit 1.

McGUIRE - UNITS 1 and 2 B 3/4 4-8 Amendme No.12(Uni t 2:

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Pay 2oC3 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITINC CONDITION FOR OPERATION

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3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accorcance with the limit lines shown on Figures 3.4-2, 3.4-3, 3. 4-4, and 3.4-5 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

l Maximum heatup rates as specified in Figures 3.4-2 and 3.4-3 l

-b.

Maximum cooldown rates as specified in Figures 3.4-4 and 3.4-5

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A maximum temperature chanc? of less than or equal to 10*F in acy c.

1-hour period during inservice hydrostatic and lean 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; perform 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,yg and pressure to less than 200*F_and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 TN Reactor Coolant System temperature and pressure shall be l-determined to te 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 determine changes in material properties, as required by 10 CFR 50, Appendix H.i+ eccerdance-ith tha rheduk i-T:bh 4r4-t The results of these examinations shall be.used to update Figures i

3.4-2, 3.4-3, 3.4-4, and 3.4-5.

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McGUIRE - UNITS _1 and 2' 3/4 4-30 Amendment No. (Unit 2)

Amendment No.100. (Unit 1)

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