ML20003H367

Safety Evaluation Re Bases for Safety Relief Valve Setpoints

ML20003H367
Person / Time
Site:	FitzPatrick
Issue date:	04/28/1981
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20003H362	List: ML20003H361 ML20003H363 ML20003H366 ML20003H367
References
NUDOCS 8105050667
Download: ML20003H367 (3)

Text

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ATTACHMEST II SAFETY EVAL"ATIO5 RELATED TO BASES FOR SArETY/ RELIEF VALVE SETPOINTS l

l POh'ER AUTHORITY OF THE STATE OF NEW YORK JAMES A.

FITZPATRICK NUCLEAR POWER PLANT f

DOCKET NO. 50-333 APRIL 28, 1981 l

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BTososobM

Section I - Description of Modification The proposed change to the Technical Specifications is shown in Attachment I.

The change is in the Bases for the numerical distribution of safety / relief valve setpoint values.

The numerical distribution of safety / relief valve setpoints shown in Section 2.2.1.B of the Technical Specifications, ( 2@ 1090, 2@ 1105, 7 @ 1140), is justified by General Electric analyses.

During the Fall of 1978, several analyses were performed to evaluate structural safety margin for several combinations of SRV actuation.

All of those analyses, including one that assumed all eleven valves simultaneously hot popped, showed acceptable stresses, except for the torus support column & the weld between the column & the torus.

Since then, saddles have been installed under the torus to reduce column loads.

These saddles reduce column & weld joint loads to about 25% of the values calculated in 1978.

The safety analyses presented in Supplement No. 1 to the General Electric report NEDo-24129-1, referred to in the NRC's Safety Evaluation for Amendment 43 to the JAF Technical Specifications, justifies the 2-2-7 setpoint configuration.

The NRC's discussion in the Safety Evaluation for Amendment 43 concludes that the structural acceptance criteria set forth in the Mark I Short Term Program are satisfied.

j Section II - Purpose of Modification The purpose of the modification is to expand the 1.2 and 2.2 Bases, justifying distribution of valves among the categories of setpoints.

Section III - Impact of the Change There is no impact in the change to the Technical Specifications, since the change will merely provide additional information to the Bases justifying the distribution of SRVs among the three categories of setpoints.

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, 4 Section IV - Impicmentation of the Modification i

The modification as proposed will not. impact the ALARA or Fire l

Protection Program at JAP.

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Section V - Conclusion _

l a) will not change The incorporation of these modifications:

the probability nor the consequences of an accident or mal-to safety as p function of equipment important will.not increase tha-b) ated in the Safety Analysis Report; possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report;w i

and c) does not constitute basis for any Technical Specification, and d) l an unreviewed safety question.

Section VI - References a

(a)

JAF FSAR f

(b)

JAF SER

" Raised Safety Relief Valve Setpoint Reanalysis for the James A. FitzPatrick Nuclear Power Plant for Reload

Zull, L.M.,

(c)

No.

2", NEDO-24129-1 Supplement 1, September 1978 i

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