ML20100C487

Application for Amend to Licenses DPR-29 & DPR-30,revising Tech Specs Re Containment Pressure Setpoint & Deleting MSIV Surveillance.Fee Paid

ML20100C487
Person / Time
Site:	Quad Cities
Issue date:	11/27/1984
From:	Rybak B COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20100C490	List: ML20100C487 ML20100C492
References
9451N, NUDOCS 8412050483
Download: ML20100C487 (4)

Text

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[ ^ N Commonwealth Edison

~^ ) One First Nitionil Plaza Chicago, Illinois

'k 7 Address R: ply to: Post Offics Box 767

(_j Chicago, Illinois 60690 s

November 27, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

- Quad Cities Station Units 1 & 2 Proposed Technical Specification Amendment Revised Containment Pressure Setpoint and Deletion of Main Steam Line Isolation Valve Surveillance NRC Docket Nos.- 50-254 & 50-265

Dear Mr. Denton:

Pursuant to 10CFR 50.59, Commonwealth Edison proposes amendments to Appendix A, the Technical Specifications, for Operating Licenses DPR 29 & 30. These proposed changes raises the drywell high pressure trip point from 2.0 psig to 2.5 psig and deletes the existing bi-weekly-main' steam isolation valve surveillance.

A summary of the proposed changes are enclosed in Attachment 1. The proposed amendments can be found in Attachments 2 & 3 and were both On-site and Off-site reviewed. We have reviewed this amendment request and find that no significant hazards consideration exists. Our review is documented in Attachment 4.

Commonwealth Edison is notifying the State of Illinois of our request for this amendment by transmittal of a copy of this letter and,its attachments to the designated State Official.

In accordance with 10CFR170, a fee remittance of $150.00 is enclosed.

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'H. R. Denton November 27, 1984 Please direct any questions you may have concerning this matter to this office.

Three (3) signed originals and thirty-seven (37) copies of-

.this transmittal and its attachments are provided for your use.

Very truly yours, a

B Ryba Nuclear Licensing Administrator Attachments (1): Summary of Proposed Changes (2): Technical Specification Change to DPR-29 (3): Technical Specification Change to DPR-30 (4): Evaluation of Significant Hazards Consideration cc: Region III Inspector - Quad Cities R. Bevan - NRR M. C. Parker - Ill.

SUBSCRIBED and SWORN to befpre me.this,# W . day o f V)l t1u-rwiu s > , -1984

l.  ;!/ortn/ir A / la

! Notary Public l

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. ATTACHMENT 1 Justification for' Proposed-Changes High Drywell Pressure Setpoint The. proposed' change _is~tc raise the high drywell. pressure trip setpointzfrom-2.0.psig to 2.5 psig. Current Technical Specifications require a 1.2 psig drywell to suppression _ chamber differentialipressure. In practice the drywell is maintained at

! about:1.3.psig with the suppression chamber kept at atmospheric t

pressure._With the actual high drywell pressure trip setpoint maintained'at less than 2.0Lpsig to allow for instrument drift, actual margin.between the. trip point and normal drywell pressure is approximately 0.6 psig. This narrow margin has in the past led to spurious actuation.

The containment isolation pressure setpoint should be as low as possible without increasing the probability of inadvertent actuation of the isolation signal. Recent NRC staff guidance (see T.A. Ippolito_ letter to L. O. DelGeorge dated January 7, 1980) on the-resolution of'TMI Item II.E.4.2.5 " Containment Pressure l Setpoint" established a limit-of 3.0 psig as an acceptable isolation setpoint margin over normal containment pressure. Clearly raising F the high drywell pressure setpoint to 2.5 psig would fall within that acceptable margin.

The original accident analysis assumed a high drywell

. . pressure-setpoint of.2.0 psig with an unpressurized drywell. The i

initiation time of the ECCS system. depends primarily on the-rate of pressure rise in the drywell and the pressure difference between the setpoint and normal operation.. Since the first parameter is relatively unaffected by_this initial drywell pressure, and since the proposed setpoint change will result in a pressure difference i-between normal operation-and the setpoint less than that used in n~

previous LOCA analyses, the_ change will result-in a more: responsive

{ performance of ECCS following drywell pressurization than that in the original analyses.

The high drywell pressure trip signal is used to' initiate primary-containment isolation and serves as a backup or conjunctive

signal to initiate ECCS. This proposed change has been reviewed-

-with respect to the time to achieve containment isolation, the

_ performance of ECCS, and the containment response to a postulated

-LOCA. . The higher initial containment pressure will slightly improve In _ ECCSfpump performance due to the small increase in the NPSH F accompanied by a lesser increase in pun p discharge pressure. In

, addition, the. change in the containment isolation time and the i, '

containment pressure response will be small since they are primarily a? function'of the. differential pressure from drywell ambient and the.

itrip. setting.; The margins between the containment design pressure

- and temperature, and the calculated results for a spectrum of breaks

~ .is sufficiently large-to accommodate-the small changes associated 4

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with the higher setpoint. Fuel peak clad temperatures should be unaffected in the event of a DBA-LOCA by the 0.5 psig increase in containment pressure, as-the rate of discharge from a postulated double-ended pipe rupture would oe at choked flow conditions, and independent of discharge pressure.

.Therefore it is our conclusion that this change also falls within the bounds of the original accident analyses.

Bi-Weekly MSIV Surveillance The proposed change is to delete the bi-weekly main steam line valve partial closure test. The MSIV's have already shown a high degree of reliability as far as the exercise aspects of the bi-weekly test. As the monthly scram also adequately demonstrates proper valve movement, there is no need for the bi-weekly test.

Support of this position can be found in the Standardized BWR Technical Specifications - the STS does not impose any greater frequency of testing over that of other primary containment valves.

The only other-change is the note on page 3.7/4.7-10 in DPR-29 was removed as MO-220-2 was restored to operability.

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