ML20101D320

Forwards Results of Util Review of 831212 Proposed Tech Spec Changes,Per 841107 Request.Meaning of Proposed Changes Not Altered by Errors.Encl Corrected Pages Should Be Incorporated Into Any Subsequent Amends

ML20101D320
Person / Time
Site:	Farley
Issue date:	12/13/1984
From:	Mcdonald R ALABAMA POWER CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
TAC-54449, TAC-54894, NUDOCS 8412240020
Download: ML20101D320 (8)

Text

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Mamng Address -

., .- Alabima Power Company -

_ 600 North 18th Strest '

i Post Offics Box 2641

. ~. Birmingham, Atabama 35291 Telephone 205 783-6090 R. P. Mcdonald Senior Vice President l

Fiintridge Building Alabama Pow.er .;

t%souvgrneecmc sgsm December 13, 1984 Docket Nos. 50-348

. 50-364 Director, Nuclear. Reactor Regulation

- U. S. Nudear Regulatory Commission Washington, D.C. 20555 t 4

Attention: - Mr. S. A. Yarga l

Joseph M. Farley Nuclear Plant - Units 1 and 2 Results of Review of i Proposed Technical Specification Changes Gentlemen:

1~

By letter dated October 23, 1984, Alabama Power Company advised the NRC Staff of an administrative error in a previously submitted technical specification change request,-dated December 12, 1983. By letter dated i November 7,1984, the NRC Staff issued the corrected technical specification page and requested that Alabama Power Company verify that each proposed technical specification change, which had not been issued by the NRC as an Amendment, be " proofed" to preclude further typographical errors.

Alabama Power Company has recently completed a review of each of -

the proposed technical' specification changes currently being reviewed by

• the NRC Staff. . Of the approximately 200 proposed Technical Specification and Bases pages reviewed, 5 pages were .found to have a i-

- typographical error. None of these typographical errors altered the

. meaning of any of.the proposed changes or made the proposed technical specifications difficult to understand. The corrected pages are L. provided as an Attachment to this letter. It is requested that the p .pages' in the Attachment be incorporated into-any subsequent amendments. ,

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Mr. S. A. 'Varga . .

December 13, 1984 U. S. Nuclear _ Regulatory Commission Page 2 Alabama Power Company recognizes and appreciates the efforts of the NRC Staff and the professionalism that has been exhibited in the review and approval of 33 License Amendments for-the Farley Nuclear Plant thus far in 1984. These efforts demonstrate that technical specification changes, which enhance the safe and orderly operation of a nuclear

- plant, can be submitted, adequately reviewed and approved by the NRC Staff in a timely manner.

If there are any questions, please advise.

Yours very tr ,

? v / /

R. P. Mcdonald RPM /CJ S:bdh-D4 Attachment cc: Mr. L. B. Long Mr. J. P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford I

ATTACHMENT' Corrected Pages Resulting From Review of Proposed Technical Specification Changes 4

Unit No. Page Submittal Date "Un it 1~ 3/4 8-12 May 3, 1983 Unit 2 3/4 8-15 May 3, 1983 Unit 1 B3/4 4 April-20, 1984 Unit 2 83/4~4-7 February 10, 1984 Unit 2 -B3/4 4-8 February 10, 1984 d

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ELECTRICAL POWER SYSTEMS j -SERVICE WATER BUILDING D.C. DISTRIBUTION - OPERATING

='

v LIMITING CONDITION FOR OPERATION

.3.8.2.5 ~ The following D.C. distribution systems shall be OPERABLE and ' '

energized: - <

, ' Traini"A" consistin'g of 125-volt D.C. Distribution Cabinet IM, 125-volt battery bank No.L1 and a full capacity charger.

Train'"B" consisting of 125-volt D.C. Distribution Cabinet IN,

125-volt

battery bank jo. 2 and a full capacity charger.

APPLICABILITY: MODES 1, 2, 3 and 4.

~

ACTION: With.one of the '125-volt distribution trains inoperable *,

restore the inoperable distribution system to OPERABLE status 4

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 in COLD SHUTDOWN 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

.n.... . ... .

4.8.2.5.1- Each D.C. train shall be determined OPERABLE and energized at least once per 7 days by arifying correct breaker alignment and indicated

power availability.

1 4.8.2.5.2- - Each 125-volt D.C. battery bank and charger shall be demonstrated OPERABLE:

1

a. At least once per 7 days by verifying that

1. The parameters in Table 4.8-2 meet the Ca'tegory A limits, and-l 2. The total battery terminal voltage is greater than or equal

.to 121.2-volts on float charge.

.b. ._At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110-volts, or battery overcharge with battery terminal voltage above 150-volts, by verifying that:

t

1. The parameters in Table 4.8-2 meet the Category B limits, /

• Except during performance of Surveillance Requirements 4.8.2.5.2.d, 4.8.2.5.2.e, and 4.8.2.5.2.c.5. During this test, one train may be inoperable until the battery is re::harged following completion of the battery discharge -test. i l

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'FARLEY-UNIT 1 3/4 8-12 AMENDMENT NO. e l-A L,-, ,_ .,J.- .-m. . , - - , - - , -, , , , - , , -

7 ,- _ ,

r EELECTRICAL POWER SYSTEMS-.

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SERVICE WATER' BUILDING D.'C. DISTRIBUTION'- OPERATING ,

-LIMITING CONDITION FOR OPERATION

. ~ .

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.3.8.2.5- -The following D.C. distribution systems shall be OPERABLE and energized:

Train "A" consisting of 125-volt D.C. Distribution Cabinet 2M, 125-volt battery bank No.1 and a full capacity charger.

Train."B" consisting of- 125-volt D.C. Distribution Cabinet 2N, 125-volt battery bank No. 2 and a full capacity charger.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: With one of the 125-volt distribution. trains inoperable *,

restore the inoperable distribution system to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. i

(. '

SURVEILLANCE REQUIREMENTS l 4.8.2.5.1 Each D.C. train shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated l- power availability.

4.8.2.5.2 Each 125-volt D.C. battery bank and charger shall be demonstrated

. OPERABLE:

i

! a. At least once per 7 days by verifying that: i

[ 1. The parameters in Table 4.8-2 meet the Category A limits, and 1 2. The total battery terminal voltage is greater than or equal j 1 to 121.2-volts on float charge.

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110-volts, or battery overcharge with battery ter.ninal voltage above 150-volts, 3 by verifying that:

1

1. The parameters in Table 4.8-2 meet the Category B limits, 4

• Except during performance of Surveillance Requirements 4.8.2.5.2.d, l

' 4.8.2.5.2.e, and 4.8.2.5.2.c.5. During this test, one train may be inoperable until the battery is recharged following completion of the i- ~ battery discharge test. . ,

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FARLEY-UNIT 2! 3/4 8-15 AMENDMENT NO.

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REACTOR'COOLANTSYShEM BASES

4) The pressurizer heatup and cooldown' rates shall;not exceed 100*F/hr and

~

200*F/hr respectively. The spray. shall not be used.if the temperature L difference between the pressurizer and the spray fluid is greater than 320*F.

5).Systen pre' service hydrotestis an'd in-service l'eak and_ hydrotests shall be

. performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of theferritic materials in the reactor.

~~

- vessel are determined in accordance with ASTM E185-82.and in accordance with:

additional reactor' vessel: requirements. These properties are then evaluated in accordance with Appendix.G of the 1976 Summer Addenda to Section III of the--ASME Boiler and Pressure Yessel Code and the calculation methods j described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

i. Heatup and cooldown limit curves are calculated using the most limiting value:of the nil-ductility reference temperature,- RTndt, at the end of 7' 1

effective full power years of service life. The 7 EFPY service life period

!- is chosen such that the limiting RTndt at the 1/4T location in the core

, region is greater than the RTndt.of the limiting unirradiated material. The selection of such a limiting RTndt assures that all components in the l' Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTndt; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase-in the RTndt. Therefore, an adjusted reference temperature, based upon the fluence and copper centeht of the material in question, can be predicted using Figure B 3/4.4-1 and the reconsnendations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Raolation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTndt at the end of 7 EFPY (as well as

- 2 adjustments for possible errors in the pressure and temperature sensing instruments).

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. FARLEY1- UNIT 1 B 3/4 4-7 AMENDMENT NO.

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L q+p go % mry r8

REACTOR COOLANT SYSTEM

' BASES

4) The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and l 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.

5) System presersice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-82 and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

Heatup and cocidown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 5 effective full power years of service life. The 5 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt 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 RTndt; the results of these tests are shown in Table 8 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTndt. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Fi Regulatory Guide 1.99, Revision 1, gure B 3/4.4-1 Effects of Residual and the Elements recomendations on of Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTndt at the end of 4.3 EFPY (as well as l adjustments for possible errors in the pressure and temperature sensing instruments).

FARLEY - UNIT 2 B 3/4 4-7 AMENDMENT NO.

\

w REACTOR'C00LANT SYSTEM.

BASES Yalues of ARTndt determined in this manner may..be used until the results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordan&

with the requirements of ASTM E185-82 and 10CFR50, Appendix H. ~The surveillance specimen withdrawal schedule is shown in Table 4.4-5. The heatup and cooldown curves must be recalculated when the ARTndt

-determined from the surveillance capsule exceeds the calculated ARTndt 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 Vessel Code as required by Appendix G to 10CFR Part 50 and these methods are discussed in detail in WCAP-7924-A. -

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fractwe mechanics ~

(LEFM) technology.' In the calculation procedures 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,3 referred.to in Appendix G of ASME -Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure.

To assure that the radiation enbrittlement effects are accounted for in the calculation of the limit curves, the nost limiting value of the nil ductility reference temperature, RTndt, is used and this includes the. ,

radiation induced shift, ARTndt, corresponding to the end of the period for which heatup and cooldown curves are generated.

FARLEY - UNIT 2 B 3/4 4-8 AMENDMENT N0.

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