ML19309H447

Forwards Responses to Anl 791220 Questions Re FSAR Chapter 7,concerning Instrumentation & Controls

ML19309H447
Person / Time
Site:	Byron, Braidwood
Issue date:	05/06/1980
From:	Peoples D COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8005130290
Download: ML19309H447 (7)

Text

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Commonwealth Edison one First National Plan Cheraqo mawns Address Reply to Post Office Hou 7r,7 Chica00, Illinois 60690 May 6, 1980 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Hegulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Responses to Questioris by Argonne National Laboratory NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457

Dear Mr. Denton:

On December 20, 1979 a meeting among representatives of Argonne National Laboratory, NRC Staff, Commonwealth Edison Company, Sargent & Lundy Engineers and Westinghouse Electric Corporation was held to discuss the content of Chapter 7 of the Byron /Braidwood FSAR which concerns instrumentation and controls. In the course of this

' meeting, a number of questions involving items within Westinghouse's scope were left unresolved. At! ached are Commonwealth Edison's

responses to those questions.

1 Please address any questions that you might have concerning this matter to this office.

One (1) signed orignal and fif ty-nine (59) copies of this transmittal are provided for your use.

Very truly yours, f $

U. L. Ped les Director of Nuclear Licensing DLP:WFN: rap attachment ec: Ira Charak, ANL w/att.

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I Questions by Argonne National Laboratory on Byron /Braidwood FSAR Chapter 7 At a meeting held at Argonne National Laboratory on December 20, 1979

, numerous questions were asked on Chapter 7 of the Byron /Braidwood FSAR concerning instrumentation and control . The majority of these ques-

!' tions were answered at the meeting by Commonwealth Edison Company, I

Sargent and Lundy, and llestinghouse. The following questions involve l items in Westinghouse scope that could not be readily addressed at the meeting.

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1. Should SNUPPS be referenced as a plant with similar instrumentation and control, since Byron /Braidwood is now ahead of SNUPPS in the NRC review process?

Response

The instrumentation and control systems discussed in Chapter 7 for Byron /

Braidwood are functionally similar to those systems implemented by Comanche Peak, W. B. McGuire, and Trojan. The instrumentation and control systems for all of these plants have been reviewed by the NRC and have been accepted in an NRC Safety Evaluation Report (SER). Trojan is currently in commercial operation. An operating license is anticipated in mid-1980 for McGuire. An SER is anticipated by the end of 1980 for Comanche Peak.

2. Definitions and their usage. Calibration and setpoint errors.

Response

A 11 umber of interrelated questions were asked concerning the definitions of errors and their usage, and the means employed to measure plant thermal I output and calibrate the neutron level detectors. These matters are con-9

sidered in a Westinghouse Topical Report in much greater detail than would be appropriate for inclusion in an FSAR. The subject report is:

WCAP-8567-P, " Improved Thermal Design Procedure," July 1975 by H. Chelemer, L. H. Boman and D. R. Sharp (Proprietary), and WCAP-8568, July 1975 (Non-Proprietary). Although WCAP-8567-P has been reviewed and accepted by the NRC, it is not as yet appropriate for inclusion as a reference in the Byron /Braidwood FSAR, but the information it contains does address the Argonne questions concerning the statistical treatment of errors and calibrations. Since the improved thermal design procedures are employed by Westinghouse for the analysis of plants using Optimized Fuel, WCAP-8567-P will be referenced when Byron /Braidwood files an Optimized Fuel amendment.

In a new Westinghouse topical report being finalized, the following definitions are being used to insure clear understanding of tolerances related to I&C Protection system setpoints. Although this report will not be referenced in Chapter 7 of the FSAR for Byron /Braidwood, these definitions should aid in understanding those already provided.

They are also believed to be consistent with the present definitions

' provided as well as the manner in which they are used in the discussions of Chapter 7 of the FSAR. The following definitQas are for information:

a. Trip Accuracy The tolerance band containing the highest expected value of the difference between (a) the desired trip point value of a process

' variable and (b) the actual value at which a comcarator trips (and thus actuates some desired result). This is the tolerance band, in

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percent of span, within which the complete channel must perform its , intended trip function. It includes comparator setting

e. Channel Accuracy The accuracy of an analog channel whic.h includes the accuracy of the primary element and/or transmitter and modules in the chain where calibration of modules intermediate in a chain is allowed to compensate for errors in other modules of the chain. Rack environmental effects are not included here to avoid duplication due to dual inputs, however, normal environment effects on field mounted hardware is included.

3. Numbering of Interlocks and Permissives Response '

Sequential numbering of interlocks and penmissives is not maintained because design control does not permit safety system interlock designa-tion to be duplicated unless the design functions are identical from plant to plant. Consequently, missing numbers indicate an interlock used in a nuclear plant with a dis-similar system function than Byron /

Braidwood.

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4. Overcurrent devices, type

Response

The overcurrent devices mentioned on page 7.2-12 are fuses and circuit breakers.

No reliance is placed on any overcurrent device for ensuring automatic initiation of the system nor would failure of these devices prevent the required safety action.

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5. Manual Reactor Trip and Safety Injection

Response

Manual initiation of Reactor Trip or Safety Injection energizes the shunt trip coil of the reactor trip breakers directly and in addition de-energizes the undervoltage coil of the reactor trip breakers thru the Solid State Protection System. This arrangement maets the requirements of Regulatory Guide 1.62 - Manual Initiation of Protective Actions.

6. Size of lines going to pressure sensors Response

• The lines to pressure sensors are typically stainless steel, 3/8 inch 00 lines with 0.065 inch walls.

7. What does the reactor coolant bypass line manifold for the temperature sensors look like?

Response

See drawing 583F499, Sub 2, attached.

8. How are display temperatures and protective temperatures derived from the sensors for Th ot, Tcold, aT, Tavg, etc. '

Response l 4

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WCAP-7913, referenced in Section 7.2 for information only, describes how the necessary control and protective temperature information is derived from the sensors in a manner that assures non-interaction between control and protection functions.

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• t 9.

Status of NRC Review of WCAP-7488

Response

WFAP-FJ ,

" Solid State Protection System," has been reviewed and approved cy

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1. or all applications having designs with simila r functional requirements for reactor protection.

10. Figure 7.2-1, sheet 4.

What do notations X>7%, etc. signify?

Response

All notes such es that shown on the subject sheet, X>7%, X>10%, etc.,

'are internal notations which have no functional relationship to the logic diagrams.

These notations will be deleted the next diagram submittal for the Byron /Braidwood FSAR - Chapter 7.

11. Revisions of Logic Diagrams, Fig. 7.2-1

_ Response

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The logic diagrams of Figure 7.2-1 in the FSAR have not yet been updated to agree with the corresponding logic diagrams currently ~ in effect. The FSAR will be amended at the earliest opportunity to delete reactor trip on loss of feedwater flow, for example, and incorporate other revisions.

12.

Does

,as the areactor?

trip on undervoltage trip the reactor coolant pump as well

. Response No.

Only the reactor is tripped on undervoltage.

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