B16979, Forwards Responses to Questions Re Proposed Rev to TS Instrumentation Surveillances

Forwards Responses to Questions Re Proposed Rev to TS Instrumentation Surveillances

ML20199E865
Person / Time
Site:	Millstone
Issue date:	01/23/1998
From:	Bowling M NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20199E870	List: B16979, Forwards Responses to Questions Re Proposed Rev to TS Instrumentation Surveillances ML20199E874
References
B16979, NUDOCS 9802020264
Download: ML20199E865 (8)

Text

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Nht ""P' *"7 "d- (" "'e is6), weiertora, er 0638s Nuclear Energy Sistooe N cica, Powe, c:atio. ~

Northeast Nuclea, Energy Company P.O. Box 128 Waterford, CT 06385 0128 (860) 447 1791 Faz (860) 444 4271

'the Northewt Utilitice System JAN 2 31998 Docket No. 50-423 B16979 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Instrumentation Surveillances (PTSCR 3-30-97) l Reauest for Additional Information Northeast Nuclear Energy Company in a .790|letter dated October 15,1997]], requested a revision to Millstone Unit No. 3 Technical Specifications. The NRC Staff, in a lettet dated January 9,1998, requested additional information to assist their review of this submittal. Attached are our responses to the questions, if you have another questions, please contact Mr. D. A. Smith at (860) 437-5840.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Nh M. L. Bowling, .It.

A d

Millstone Unit No. 2 - Recovery Officer Sworn ta and subscribed before me  !

this c23 day of sceneea '

1998 QJ l

%%e ed t iLw.b i Nohiry Pubic My Commission expires b300. @ gen i Attachments cc: See Page 2 j; - v U 2 hh,k 9802020264 980123 PDR ADOCK 05000423 P PDR ,

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U.S. Nucl:ar Regul: tory Commission B16979\Page 2 cc: H. J. Miller, Region i Administrator J. W. Andersen, NRC Project Manager, Millstone Unit No. 3 A. C. Corne, Senior Resident inspector, Millstone Unit No. 3 W. D. Travers, Ph.D, Director, Special Projects Office 1

t

. . Docket No. 50-423 g3 Attachment 1 Millstone Nuclear Power Stat;on, Unit No. 3 Commitments i

January 1998 t , . . .

U.S. Nucbir Regul; tory Commission B16979%ttachment 1\Page 1 Enclosure List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document. Please notify the Manager - Regulatory Compliance at the Millstone iJuclear Power Station Unit No. 3 of any questions regarding this document or any associated regulatory commitments.

I Commitment Committed Date or Outage l

l NONE N/A s

% e a

i .

.. l Docket No. 50-434 B16979-Attachment 2 .

Millstone Nuclear Power Station, Unit No. 3 I

Request for Additional Information i

January 1998 p , . - -

I U.S. Nuclear Regul: tory Commission B16979%ttachment 2\Page 1 i- 0  :

! Millstone Nuclear Power Station, Unit No. 3

Proposed Revisicn to Technical Specification

! Instrumentation Surveillances (PTSCR 3-30-97)

I Reauest for AdditionalInformation i

a NRC Question 1

[ Proposed LCO statement (2.2.1) does not reflect its heading requirement and

[ does not refer to the table that lists the setpoints. Please justify.

NNECO's Response 1 Our intent in modifying the wording associated with LCO 2.2.1 was to make it more consistent with the wording provided in NUREG 1431, Revision 1.

However, in making this - change, we created an ' inconsistency within the Millstone Unit No. 3 Technical Specifications by removing the reference 'to Table 2.2-1. Therefore, we are requesting that our submittal of October 15, 1997, be modified by replacing page 2-4 with the page contained within Attachment 4 to this letter which corrects the above described inconsistency.

We have reviewed the above listed change to our submittal of October 15,1997 and have concluded that the change does not effect the Safety Assessment or Significant Hazards Consideration contained within the submittal.

NRC Question 2 Instrument operellity requirements are stated in LCO 3.3.1, why is it needed twice (proposed 2.2.1 and existing 3.3.1)? Please justify.

NNECO's Response 2

-The response and actions associated with Ouestion 1 resolves this question.

NRC Question 3 In the current TS, if the reactor trip system instrumentation or interlock setpoint is less conservative than the allowable value, the TS requires, adjusting the setpoint consistent with the trip setpoint value of Table 2.2-1 and determining within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that equation 2.2-1 was satisfied for the affected channel, or declaring the channel inoperable and following the action in Table 3.3.1. For the proposed changes, the bases states that if the value is found to be in excess of a'lowable value, but within the administratively controlled limit, the channel is i - . . . - - . _ , - . - _ - .

• et U.S. Nucirr Regul: tory Commission B16979\ Attachment 2\Page 2 operable and must be calibrated per the plant procedure. As such, the channel operability is controlled by an administrative procedure and not by TS. Please justify.

NNECO's Response 3 In the current Millstone Unit No. 3 Technical Specifications if the reactor trip system instrumentation or interlock setpoint is less conservative than the alloweble value, the Technical Specifications require:

adjusting the setpoint consistent with the trip setpoint value of Table 2.2-1 and determining witHn 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that equation 2.2-1 was satisfied for the affected channel, of.

declaring the channel inoperable and applying the applicable action a statement requirement of Specification 3.3.1 until the channel is restored to operable status with its setpoint adjusted consistent with the trip

! setpoint.

In actual practice, during the performance of reactor trip system instrumentation calibration suiveillances, the applicable channel is declared inoperable and the applicable action statement requirement applied. The channel is not declared operable until its setpoint is adjusted consistent with the Trip Setpoint value (Technical Specification 2.2.1 b.2).

Use of equation 2.2-1 and the Z, R, S and TA values is described in the Technical Specification Basis as an optional provicion for determining operability of a channel when its Trip Setpoint is found to exceed the Allowable Value.

While data on the as found condition of the components is taken during refuel channel calibrations and the equation calculated, this method has been found to be less reliable in identifying excessive component drift. Pest calibrations hnve shown that the Technical Specification allowance for rack drift (the difference between Nominal Trip Set Point and Allowable Value) is much larger than that >

observed during calibration / surveillance activities. Westinghouse has issued a Technical Bulletin stating that the past methods utilized to calculate the operability criteria noted in the Technical Specifications Allowable Velue result in criteria that are significantly greater than drift data statistical evaluation results.

This results in the pnnsibility of rack component failure prior to meeting or exceeding the currently identified Technical Specifications Allowable Values.

As-found calibration data is tyNcally not used to determine historical operaoility.

In the case of instrument drift, i; is not usually possible to determine when the instrument exceeded its allowed drift. The purpose of routine calibration has

,m

U.S. Nucbtr R;gul: tory Commission B16979%ttachment 2\Page 3

,. always been to restore an instrument's response to within a nominal band at a predetermined interval. That nominal band is determined by M&TE accuracy, equipment performance specifications and often actual operating experience.

It is more effective to control and monitor as-found calibration operability criteria Wa administrative procedures because we will be able to more accurately accomrnodate the effects of changes in M&TE type, equipment replacement, and operating experience. Furthermore, it allows us to establish more stringent acceptance criteria that will trigger a review and resolution of component performance problems before reaching any reportability threshold, in this process, when a loop and/or component is found outside its expected value (listed in plant surveillance procedures) it is identified and resolved via our corrective action program. This program includes an operability / reportability determination and the identification and implementation of appropriate corrective actions.

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