ML20087M670

Application for Amend to License NPF-3,revising Tech Specs to Reflect Changes in post-accident Sampling Sys & post- Accident Instrumentation for in-core Thermocouple & Reactor Coolant Hot Leg Level

ML20087M670
Person / Time
Site:	Davis Besse
Issue date:	03/22/1984
From:	Crouse R TOLEDO EDISON CO.
To:
Shared Package
ML20087M663	List: ML20087M662 ML20087M670 ML20087M672
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM GL-83-37, TAC-54388, TAC-54528, NUDOCS 8403300156
Download: ML20087M670 (7)

Text

APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Facility Operating License No. NPF-3, together with the Safety Evaluation for the requested change.

The proposed changes include:

1. Section 6.8.4 and

2. Tables 3.3-10 and 4.3-10 1

By /s/ R. P. Cronno .;

Vice President, Nuclear ,

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Sworn and subscribed before me this 22nd day _of March, 1984.

/s/ Nora Lynn Flood Notary Public

) My Conunission Expires September 1,1987.

8403300156 840322 SEAL yDRADOCK 05000346 PDR

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I Docket No. 50-346 -

l License No. NPF-3 Serial No. 1032

' larch 22,1984 a

Attachment I I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Section 6.8.4 A. Time required to Implement . This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request in response to Mr.

D.G. Eisenhut's letter concerning NUREG 0737 Technical Specifications (Generic Letter No. 83-37). The attached anendment request is for administrative controls of the Post-Accident Sampling System.

C. Safety Evaluation (See Attached) 4 D. Significant Hazard Consideration (See Attached) 1 4

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. t SAFETY EVALUATION i

This amendment request is to add Technical Specification requirements to maintain administrative controls over post-accident sampling capabilities j as requested by the NRC letter dated November 1, 1983 Log No. 1402 (Ref.

I NUREG-0737 Item II.B.3).

1 The safety function of the post-accident sampling system is the capability to obtain samples of reactor coolant and containment atmosphere under accident conditions. Further the system allows personnel the capability to obtain samples promptly without incurring a radiation exposure in excess of the limits.

Proposed additions to the Technical Specifications will ensure that a program to administer post-accident sampling is maintained. This request will add to the safety function of the system by providing guidelines for:

1) Training of personnel 2) Procedures for sampling and analysis 2 3) Provisions for maintenance of sampling and analysis equipment.

Therefore, Based on the above this is not an unreviewed safety question.

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SIGNIFICANT HAZARD CONSIDERATI "

The amendment request for maintaining administrative control over post-accident sampling capabilities does not represent a Significant Hazard.

NUREG-0737 (Item II.B.3) requires administrative controls to ensure the post-accident sampling capability under accident conditions.

The Post-Accident Sampling System provides the capability to obtain reactor coolant and containment atmosphere samples under accident conditions. The system allows for samples to be taken in a timely manner, and ensures personnel radiation exposure remains within applicable limits.

The proposed Technical Specifications will ensure that a program to administer post-accident sampling is properly maintained. The overall safety function of the system will be enhanced due to guidelines providing for 1) training of personnel, 2) maintenance of sampling and analysis equipment, and 3) procedures for sampling and analysis (i.e. , system operation).

The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (48 FR 14870).

One of the examples of actions involving no significant hazards considerations related to a change that constitutes an additional limitation, restriction, or control not presently included in the i technical specifications: for example, a more stringent surveillance requirement. (Example ii)

Based on the above information, t,his amendment request would not

1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) involve a significant reduction in a margin of safety.

Therefore, based on the above, the requested license amendment does not '

present a Significant Hazard.

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Docket No. 50-346 License No. NPF-3 Serial No. 1032 March 22,1984 -

Attachment II I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Tables 3.3-10 and 4.3-10 A. Time required to Implement . This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request in response to Mr. D.

G. Eisenhut's letter dated November 1, 1984 (Log No. 1455) concerning NUREG 0737_ Technical Specification (Generic Letter 83-37). The attached amendment request is for Post-Accident Instrumentation for In-Core Thermocouples and Reactor Coolant Hot Leg Level (Wide Range).

C. Safety. Evaluation (See Attached) l D. Significant Hazard Consideration (See Attached) w I

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4 l SAFETY EVALUATION This amendment request to incorporate additions to the Technical Specification for Post-Accident Instrumentation concerning core-exit thermocouples and the Reactor Coolant Hot Leg Level System.

4 The safety function of the Reactor Coolant Hot Leg Level System is to meesure the hot leg water level during natural circulation or during

, periods when natural circulation is interrupted. The safety function of the In-Core Thermocouples is to provide an indication of core uncovery.

The attached Technical Specification which provides limiting conditions for operation will ensure the availability of these systems during post-

accident conditions. Also, the surveillance frequency is adequate to verify that this operability is maintained in the applicable modes.

. Therefore, this request is not an unreviewed safety question.

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l SIGNIFICANT HAZARD CONSIDERATION The amendment request for Post-Accident Instrumentation concerning core-exit thermocouples and the Reactor Coolant Hot Leg Level System does not represent a Significant Hazard. The NRC issued Generic Letter 83-37, dated 11/1/83, (Log No. 1402) requiring NUREG-0737 Technical Specifications be provided for the RCS Hot Leg Level System and In-Core Thermocouples.

During RCS natural circulation, or periods when natural circulation is interrupted, the Hot Leg Level System will provide an accurate indication of reactor coolant water level. The in-core thermocouples will provide temperature indication in the various quadrants of the core, and thus serve as indicators of reactor vessel water level in the event the core becomes uncovered.

The Technical Specification Limiting Conditions for Operation (LCO) will ensure the availability of both systems during a post accident condition.

In addition, the frequency for system surveillance is adequate such that system operability is maintained for all applicable modes.

The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (48 FR 14870).

One of the examples of actions involving no significant hazards considerations related to a change that constitutes an additional limitation, restriction, or control not presently include in the technical specifications: for example, a more stringent surveillance requirement.

(Example ii)

Based on the above information, this amendment request would not

1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; ot

3) involve a significant reduction in a margin of safety.

Therefore, based on the above, the requested license amendment does not present a Significant Hazard, ej a/12

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