ML20101S800

Application for Amends to Licenses NPF-4 & NPF-7,proposing to Revise Wording of Reactor Coolant Sys Volume in Design Features Section & Eliminate Reference to Exact Volume in Bases Section

ML20101S800
Person / Time
Site:	North Anna
Issue date:	07/16/1992
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20101S802	List: ML20101S800 ML20101S804
References
92-467, NUDOCS 9207200066
Download: ML20101S800 (6)

Text

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VIHOINIA Ih.t:01 HIC AND POWl:H COMPANY Hirnnown,Vinuam unuu July 16, 1992 i

U.S. Nuclear Regulatory Commission Serial No. 92 467 Attention: Document Control Desk NAPS /RMN Washington, D.C. 20555 Docket Nos. 50 338 50 339 Licerise Nos. NPF 4 NPF 7 Gentistnen:

i y(RGINIA ELECTBlp AND POWER COMPANY-NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGES Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests amendments, in the form of changes to the Technical Specifications, to Operating License Numbers NPF 4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed chatiges will revise the wording of the reactor coolant system volume in the design features section and o!!mir-40 reference to an exact volume in the bases section.

A discussion of the proposed changes is provided in Attachment 1.

The proposed changes are presented in Attachment 2.

This request has asen revirned by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee.

It has been determined that this request does not involve an unreviewed safety question as

. defined in 10 CFR 50.59 or a significant hazards considoration as defined in 10

- CFF 70.92. The; basis'ior our determination that no signification hazards consideration is involved is presented in Attachmert 3.

Should you have any questions or require additional information, please cor. tact us at your earliest convenience.

Very truly yours, l

u J4r W. L. Stewart Senior Vice Presidsnt - Nuclear i.00UO3 Ob 9207200066 920716-i i

_hDR ADOCK 05000338 PDR

4 Attachments 1.

Discussion of Proposed Changes 2.

Proposed Technical Specification Chan0es for North Anna Units 1 and 2 3.

10 CFR 50.92 Significant Hazards Consideration ec:

U.S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N.W.

Suite 2900 Atlanta, GA 30323 Mr. M. S. Lesser NRC Senior Resident inspector North Anna Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219

COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. P. O'Hanlon, who is Vice President Nuclear Operations, for W. L. Stewart who is Senior Vice President - Nuclear, of Virginia

' Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statomonts in the document are true to the best of his knowledge and belief.

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x Discussion of Proposed Changes North Anna Units 1 and 2 -

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4 INTRODUCTION The proposed change to Technical Specification 5.4.2 revises the wording for the Reactor Coolant System (RCS) volume in the design features.

The proposed change would revise the description of the RCS volume from ".. 9957 110 cubic feet at a nominal Tava "525 F* to "...approximately 10,000 cubic feet of at nominal operating conditions.

The proposed change to the bases for i

Technical Specification 3/4.1.1.3 would change the phrase "will circulate an equivalent Reactor Coolant System volume of 9957 cubic feet" to "will circulate the Reactor Coolant System volume."

RACKGROUND 10 CFR 50.36 describes the purpose of the design features in the Technical Specifications.

Design features in the Technical Specifications are those features of the facility which, if altered or modified, would have a significant offect on safety. Technical Specification 5.4.2 does not meet this criterion as it is currently written.

NUREG 0452, " Standard Technical Specifications for Westinghouse PWRs,"

Revision 4, (STS) provides guidance to licensees when preparing Technical Specifications. One dnsign feature specified in the STS is the RCS volume. A calculated RCS voMme was obtained from the reactor vendor and included in the North Anna Technical Specificatinns consistent with STS guidance.

However, that calculated volume is not used in accident analyses. Rather, the transient system models used tu perform accident analyses divide the RCS into component mass energy cells for which volumes are specified. The definition and volume of these mass energy cells vary depending on the model used and the accident being analyzed. Thus, the methodology does not require that a single total RCS volume be specified as an accident analysis basis and revising the current description would have no significant offect on snfety.

Technical' Specification 5.4.2, as currently worded, also may be somewhat confusing. It refers to a Tava that is significantly lower than normal operating conditions. Also, citing an RCS volume at a Tavo of 525*F provides insufficient information for meeting a tolerance of plus or minus ten cubic feet. Other parameters, such as system pressure and pressurizer temperature, are also required to calculate the volume to within ten cubic feet. Those parameters are r

not clearly defined since the cited Tavg is not based on normal operating conditions.

DfSCRIPTION OF CHANGES The proposed changes apply to both North Anr.a Units 1 and 2.

The description of the RCS volume in Techaical Specification 5.4.2 is changed from ".. 9957 i10 cubic feet at a nomina' Tavg of 525?F" to "...approximately 10,000 cubic feet at nominal operatirig conditions." This is the RCS volume at nominal operating conditions rounded off to the nearest 1000 cubic feet. The proposed revision states a norninal volume (i.e., to within 1.5%) and does not misrepresent the accuracy nor the operating temperature. The change is consistent with the definition of design features in 10 CFR 50.36.

Technical Specification 3/4.1.1.3 requires the flow rate of reactor coolant through the RCS to be a 3000 gpm when a reduction in RCS boron concentration is being made. The bases for Technical Specifica"on 5/4.1.1.3 state that the minimum flow rate of at least 3000 gpm provides adequate mixing, prevents st atification and ensures that reactivity changes will be gradual during boron concentration reductions in the RCS. It then states that "a flow rate of at least 3000 gpm will circulate an equivalent RCS volume of 9957 cuble feet in approxir;.utely 30 minutes." This statement is changed to "a flow rate of at least 3000 gpm will circulate the Reactor Coolant System volume in approximately 30 minutes."

SAFETY SIGNIFICANCE The proposed change to Technical Specification 5.4.2 does not have any adverse effect on safety. It specifies the RCS volume commensurate with the accuracy required for describing a design feature. However, the proposed change preserves the intent of the design feature by assuring that a significant RCS volume change would be identified and evaluated for potential safety significance.

The proposed change to the bases of Technical Specification 3/4.1.1.3 does not have any adverse effect on safety. Deleting the phrase does not alter the meaning of the bases and climinates a potential conflict with the revised design features.

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