ML20247R470

Proposed Tech Specs Revising 18-month Surveillance Interval for Insp & Operability Testing of Reactor Vessel Internal Vent Valves

ML20247R470
Person / Time
Site:	Davis Besse
Issue date:	08/01/1989
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20247R463	List: ML20247R460 ML20247R467 ML20247R470
References
1674, NUDOCS 8908080014
Download: ML20247R470 (6)

Text

Docket tiumber 50-346

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THIS PAGE PROVIDED FORINFORMATION DE

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APPLICABILITY SURVEILLANCE REQUIREMENTS' A

4.0.1 ' Surveillance Requirements shall be applicable during the OPER-ATIONAL MODES or other conditions specified for individual Limiting E "dft'io5s' for Operation unless otherwise stated in an individual Sur-

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1 veillTncY Yiq~uirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified' time interval with:

a.

A maximum allowable extension not to exceed 25% of the sur-I veillance interval, and b.

.A total maximum combined interval time for any 3 consecutive

.g tests not to exceed 3.25 times the specified surveillance.

l interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation.

Exception to these requirements are stated in the individual Specifications.

Surveillance Requirements'do not have to be performed on inoperable equipment.

l 4.0.4 Entry into an OPERATIONAL MODE or other specified applicability I

condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a.

During the time period:

1.

From issuance of the Facility Operating License to the start of facility commercial op? ration, inservice testing i

of ASME Code Class 1, 2 and 3 pumps and valves rhall be l

performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition, and Addenda through Summer 1975, except where specific written relief has been granted by the Commission.

2.

Following start of facility commercial operation', inservice I

inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI.of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 55.55a(g)(6)(i).

b.

Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda'for the inservice inspection and testing activities required by the ASME l

Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

8908083014 890801 FDR ADOCK 05000346 p

PDC DAVIS-BESSE, UNIT 1 3/4 0-2 Amendment No. 71 i

Docket Number 50-346 License Number NPF-3

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IlilS PAGE PROVIDED Page'8 FORINFORMATIONON REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION l'

3.4.10.1-The structural integrity of 'ASME Code Class 1, 2 a6d 3 components shall be maintaineo in accordance with Specification 4.4.10.1.

l APPLICABILITY: All MODES.

ACTION:

With the structural integrity _of any ASME Code Class 1 component (s) a.

not confnrming to the above requirements, restore the structural integrity of the affected component (s) to within'its limit'or

, isolate.the affected component (s) prior ~ to increasing the Reactor Coolant System temperature more than.50'F above the minimum temperature required by NDT considerations..

b.

-With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit.or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.

c.

With the structural integrity of any ASME Code Class 3 component (s) l I

not conforming to the above' requirements, restore the structural integrity of the component (s) to within its limit or isolate the affected component (s) from service.

d..

The provisions of Specification 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS j

!'I 4.4.10.1 In addition to the requirements of Specification 4.0.5:

a.

The reactor coolant pump flywheels shall be inspected per the

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recommendations of Regulatory Position C.4.b. of Regulatory j

Guide 1.14, Revision 1, August 1975.

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DAVIS-BESSE, UNIT 1 3/4 4-30

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Docket Number 50-346 License Number NPF-3

'Scrial Number 1674 Attachment'1 Page 9 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.

Each internals vent valve shall be demonstrated OPERABLE

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1.

Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation, 2.

Verifying the valve is not stuck in an'open position, and 3.

Verifying through manual actuation that the valve is fully open when a force of 5,400 lbs..is applied l

vertically upward.

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tut se lete th_; the Cycle 5 ::f;:lis; ::::g:.

YThe read.sions of SpecMiention 'iA A nee not appl l cable.

DAVIS-BESSE, UNIT 1 3/4 4-31 Amendment No. 23, 95 J

Docket Number 50-346~

Lic,ense Number NPF-3

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' Serial' Number 1674 g

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. Attachment 1 FORRFORE110N ONLY

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REACTOR COOLANT SYS REM BASES t

3/4.4.10 STRUCTURAL INTEGRITY

'The inspection programs for ASME Code Class 1, 2 and 3 components, except uteam generator tubes, ensure that the structural integrity of these components will be maintained at an-acceptable level throughout the life of the plant.

To the extent applicable, the inspection prcgram for these l

components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

'l The internals. vent valves are provided to relieve the pressure generated i

by steaming in the core following a LOCA so that the core remains suffi-ciently covered.

Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that.the valves are not stuck open during normal operation, and 3) demonstrates that the valves are fully.

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open at the forces equivalent to the differential pressures assumed in the safety analysis, i

3.4.4.11 HIGH POINT VENTS

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l The Reactor Coolant System high point vents are installed per.NUREG-0737 item II.B.1 requirements..The operability of the system ensures capability of venting steam or noncondensable gas bubbles in the reactor cooling

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system to restore natural circulation following a small break loss of l

coolant accident.

DAVIS-BESSE, UNIT 1 B 3/4 4-13 Amendment No. 85

. Docket Number 50-346

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License Number NPF-3 Serial Number 1674 Page 1 SIGNIFICANT HAZARDS CONSIDERATION DESCRIPTION The purpose of this Significant Hazards Consideration is to review proposed changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating

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l License, Appendix A, Technical Specifications. The proposed changes involve revising the 18-month surveillance interval for inspection and operability testing of Reactor Vessel Internals Vent Valves.

The discussion below provides the Significant Hazards Consideration for the changes as proposed in the Technical Description (Attachment 1).

SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazard exists. A proposed amendment to an Operating License for a facility involves no significant hazards if operation of the facility in accordance with the proposed changes vould:

1) Not involve a significant increase in the probability or consequences of an accident previously evaluated; 2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) Not involve a significant reduction in a margin of safety.

Toledo Edison has reviewed the proposed changes as discussed in the Technical Description and determined that a significant hazard does not exist because operation of the Davis-Besse Nuclear Power Station, Unit Number 1 in accordance with these changes would:

Not involve a significant increase in the probability or consequences of an accident previously evaluated because the accident conditions and assumptions are not affected by the proposed Technical Specification changes. As discussed in the Technical Description, the proposed surveillance interval for inspection and operability testing of the RVVVs would still be frequent enough to ensure that the high reliability for proper operation of the system is unaffected. Furthermore, the proposed changes vill allow the required surveillance of RVVVs to be performed during sebcduled refueling outages, as was intended, and eliminate the need for mid-cycle shutdowns solely for the purpose of performing the surveillance (10CFR50.92(c)(1)).

Not create the possibility of a new or different kind of accident.from any accident previously evaluated because, as discussed in the Technical Description, the accident conditions and assumptions are not affected by the proposed Technical Specification changes.

On matters related to nuclear safety, all accidents are bounded by previous analysis. The proposed changes do not add to or modify any equipment or system design nor do they involve any changes in the operation of any plant system (10CFR50.92(c)(2)).

Not involve a significant reduction in a margin of safety because the proposed surveillance interval vould still be frequent enough to ensure that proper operation of the valves is unaffected. As discussed in the Technical l

l Docket Number 50-346 Li. cense Number NPF-3 Serial Number 1674 Page 2 Description, operating experience with these valves in the industry and at Davis-Besse has shown that they will remain operable for periods far greater than 18 months.

Equipment reliability and margin of safety vill be maintained. The proposed changes will allow the required surveillance of RVVVs to be performed during scheduled refueling outages, as was intended, and eliminate the need for mid-cycle shutdowns solely for the purpose of performing the surveillance (10CFR50.92(c)(3)).

On the basis of the above, Toledo Edison has determined that the proposed Technical Specification changes do not involve a significant hazard.

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