ML20064A333

Proposed Tech Specs Change 90-17 Removing Table 4.4-5 from Tech Spec 3/4.4.9.1

ML20064A333
Person / Time
Site:	Sequoyah
Issue date:	08/27/1990
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20064A332	List: ML20064A331 ML20064A333
References
NUDOCS 9009040044
Download: ML20064A333 (6)

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~l REACTOR COOLANT SYSTEM. ,j 3/4.4.9 PRESSURE / TEMPERATURE-LIMITS q e REACTOR COOLANT SYSTEM-LIMITING CONDITION FOR OPERATION y

3 3.4.9.1 The Reactor Coolant' System (except the pressurizer): temperature and j pressure shall be limited in accordance,with the limit lines shown on -

Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and. inservice-leak and hydrostatic testing with:

a. A maximum heatup of 100'F in any one hour period. I 1

b. A maximum cooldown of 100'F in any one hour period. I

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c. -A~ maximum temperature change of less than or equal to 5*F in any one hour period during inservice hydrostatic and leak testing operations

=above the heatup and cooldown limit curves.

APPLICABILITY: At all times. '

.h ACTION: '

.) !j With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30. minutes; perform an engineering eval'uation to- {

determine the ef fects of the out-of-limit condition on the structural integrity 'j of the Reactor Coolant System; determine that the Reactor Coolant System remains j acceptable for continued operation or be in at least HOT. STANDBY within the.next '

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, 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 avg 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 RE0VIREMENTS '

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4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determ, ed to be within the limits at least once per 30 minutes during system ,

'i n ~ - Mc!down, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens  !

attordance Uh he t N;mtshall abe .Mremoved and

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CFR to determine 50, Appendix H.! caanges in material properties,.e4 in

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di i '2 h 4.0 S. The results of these examinations shall be used to update figures 3.4-2 and 3.4-3.

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SEQUOYAH - UNIT 2 3/4 4-28 9009040044 900827 PDR ADOCK 05000328 P PDC '

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g EACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WIT E RAWL SCHEDUCE

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ENCLOSURE'2 PROPOSED TECHNICAL ~ SPECIFICATION. CHANGE SEQUOYAH NUCLEAR PLANT UNIT 2

-DOCKET NO. 50-328' (TVA-SQN-TS-90-17)-

DESCRIPTION AND JUSTIFICATION ~FOR DELETION OF TABLE 4.4-5.

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ENCLOSURE 2 Description of_ Change TVA proposes.to modify the Sequoyah Nuclear Plaat (SQN) Unit *2 Technical Specifications (TSs) to. remove Table 4.4-5 from TS 3/4.4.9.1.

J Amendment 87, which was issued on October 14, 1988, similarly resised the ,  !

-Unit 1 TSs.

Reason and Justification for Change 1

- TVA is required to- comply with 10 CFR Part 50, Appendix H . " Reactor Vessel 4 Material Surveillance' Program Requirements." The requirements of- (

Appendix H provide for NRC approval of a proposed withdrawal schedule before implementation. ~The requirements also identify the applicable .)

-American Society for Testing Materials (ASTM) and American Society of, Mechanical Engineers (ASME) Codes and the applicable reporting  !

requirements. Therefore, inclusion of Table 4.4-5 in TS 3/4.4.9.1 is. l redundant to requirements of 10 CFR Part 50, Appendix H. The deletion of 1 Table 4.4-5 is: consistent,with improvement of TSs by removing specifications that are redundant to regulation, as recommended by both the NRC Technical Specification Improvement Project and the Atomic i Industrial Forum (AIF) Subcommittee on Technical Specification Improvements in October 1985.

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In addition, the' data of Table 4.4-5 is contained in Section 5.4.3.7,:

" Reactor Vessel Material Surveillance Program Requirements," of the SQN Final Safety Analysis Report. Thus, the surveillance schedule is i available to NRC in an administratively controlled document that is  !

reviewed and revised on a regular basis.

. Environmental Impact Evaluation The proposed change request does not involve an unreviewed environmental ,

. question because operation-of SQN Units 1 and 2.in accordance with this '

change would not:

1. Result in a significant increase in any adverse environmental-impact previously evaluated in the Final Environmental Statement (FES) as modified byithe Staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or '

decisions of the Atomic Safety and Licensing Board.

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2. Result in a significant change in effluents or power levels.

- 3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.

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' PROPOSED TECHNICAL SPECIFICATION. CHANGE  ;

SEQUOYAH NUCLEAR PLANT-UNITS ~1'AND 2=

DOCKET NO. 50-328-  ;,

(TVA-SQN-TS-90-17) l

- DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION k l!

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ENCLOSURE 3-Significant Hazards Evaluation t

TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration.

based on-criteria. established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance with the proposed amendment will nots -

(1) Involve a-significant increase in the probability or-consequences of-an accident previously evaluated.

The proposed change revises TS-3/4.4.9.1 of the SQN Unit 2 TSs by_

deleting Table 4.4-5, " Reactor Vessel Material Survei11snee Program -

Withdrawal Schedule." Deletion of the subject-table from the TSs will not affect the reactor vessel surveillance program requirements as specified in 10 CFR Part 50, Appendix H. " Reactor Vessel Material.

Surveillance Program Requirements"; rather, the deletion will eliminate a license requirement that is redundant to regulation-requirements. The proposed amendment is therefore administrative in, nature and does not change plant hardware, plant operating setpoints or limits, or plant operating procedures. The potential for reactor

. vessel embrittlement affecting a postulated transient or accident conditions that have been previously evaluated ~is not increased as TVA is required to comply with 10 CFR Part 50, Appendix H.

Therefore, the proposed change involves no increase in,the probability or consequences of an accident previously evaluated.

(2) Create the possibi1~ity of a new or different kind of accident from any previously analyzed.

The proposed amendment is administrative in nature and does not change plant hardware, plant operating setpoints or limits, or plant operating procedures. -Also, the evaluation of reactor vessel embrittlement is not affected as TVA is required to comply with 10 CFR Part 50, Appendix H. Therefore, the proposed amendment does not create ,the possibility of a new or different kind-of accident' from any accident previously evaluated.

(3) Involve a significant reduction in a margin of safety.

The proposed. amendment is administrative in nature and does not involve a change in plant hardware, plant operating setpoints or limits, or plant operating procedures. The evaluation of reactor vessel embrittlement is not affected as TVA is required to comply with 10 CFR Part 50, Appendix H. Therefore, the proposed amendment involves no reduction in the margin of safety of the plant.

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