ML20062C166

Forwards Proposed Agenda & List of Questions from Instrumentation & Control Sys Branch for Upcoming Instrumentation & Control Meetings.Proposed Schedule for Conducting Meetings Requested within 2 Wks

ML20062C166
Person / Time
Site:	River Bend
Issue date:	07/28/1982
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	William Cahill GULF STATES UTILITIES CO.
References
NUDOCS 8208050180
Download: ML20062C166 (22)

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JUL 2 E tgy-Tu Docket Nos. 50-458/459 Mr. William J. Cahill, Jr.

Senior Vice President River Bend Nuclear Group Gulf States Utilities Company P. O. Box 2951 Beaumont Texas 77704

Subject:

Agenda Items for Instrumentation and Control Systems Meetings with Gulf States Utilities - River Bend, Units 1 and 2 The purpose of this letter is to provide a proposed agenda (Enclosure 1) for the first of several instrumentation and Control (I&C) meetings to be held with Gulf States Utilities on River Bend Station, Units 1 and 2.

In addition, a set of questions (Enclosure 2) is provided to obtain the information required to give the staff a better understanding of the design bases and design implementation of I&C systems at River Bend.

It is suggested that each meeting be held at the Architect Engineer's (Stone & Webster - S&W) offices, so that the applicable I&C drawings will be readily accessible.

It is also suggested that each meeting in-clude the minimum number of participants necessary to discuss the topics listed on the meeting agenda and to address the set of questions to be provided prior to each m2eting.

In addition, unresolved issues and unanswered questions from previous meetings would be discussed. NSSS vendor (General Electric - GE) representatives familiar with the instru-mentation and controls at River Bend within the GE scope of supply should be present at the first meeting as emphasis will be placed on drawing reviews for these systems and the similarity of these systems to pre-viously approved designs (e.g., Clinton, Perry, etc.).

It is antici-pated that each of the I&C meetings will last from two to three days.

The questions contained in Enclosure 2 were developed by the Instrumentation and Control Systems Branch (ICSB) in conjunction with the Argonne National Laboratory (ANL) techncial staff and resulted from evaluation of informa-tion presented in Chapter 7 of the River Bend FSAR. Additional I&C ques-tions which arise during the review process will be provided at later dates, and therefore, Enclosure 2 should not be considered to be a final list of questions prior to issuance of a Safety Evaluation Report. The applicant's responses to questions, where applicable, should be supplemented by appro-priate I&C drawings (including P& ids, electrical schematics, logic diagrams, and other detailed instrument drawings).

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l We request that you provide us within 2 weeks of receipt of this letter, your proposed schedule to conduct these meetings.

If you have any ques-tions concerning this matter, please call R. Perch, NRC Project Manager.

i Sincerely, A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing

Enclosures:

As stated cc: See next page i

DISTRIBUTION:

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River Bend Mr. William J. Cahill, Jr.

Senior Vice President River Bend Nuclear Group Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 ATTN: Mr. J.E. Booker cc:

Troy B. Conner, Jr., Esquire Conner and Wetterhahn 1747 Pennsylvania Avenue, N. W.

Washington, D. C.

20006 Mr. William J. Reed, Jr.

Director - Nuclear Licensing Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 Stanley Plettman, Esquire Orgain, Bell and Tucker Beaumont Savings Building Beaumont, Texas 77701 William J. Guste, Jr., Esquire Attorney General State of Louisiana Post Office Box 44005 State Capitol Baton Rouge, Louisiana 70804 Richard M. Troy, Jr., Esquire Assistant Attorney General in Charge 4

State of Louisiana Department of Justice 234 ~ Loyola Avenue i

New Orleans,' Louisiana 70112 1

A. Bill Beech Resident Inspector Post Of fice Box 1051 St. Francisv111e, Louisiana 70775 e

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-- a ENCLOSURE 1 RIVER BEND UNITS 1 & 2 I&C MEETING #1 AGENDA I. ' INTRODUCTIONS AND GENERAL DISCUSSION Scope of ICSB review and discussion of the review process.

Status of the River Bend Station and approximate schedules for completion of unfinished items.

Discussion of answers to questions on Enclosure 2.

II. SIMILARITY OF THE RIVER BEND DESIGN TO OTHER GE PLANTS (WHICH HAVE BEEN PREVIOUSLY REVIEWED)

Discussion of systems listed in FSAR' Table 7.1-2.

Are the systems listed as being similar in design to Perry identical to the Perry design or are there specific differences in the designs (although similar) which should be pointed out?

III. SPECIFIC I&C ITEMS (REVIEW STATUS AND DISCUSSION OF DESIGN DETAILS AND STAFF POSITIONS)

Discussion and walk-through of the station electricaT distribution system with emphasis on emergency and vital buses as background for addressing various Chapter 7 concerns.

4' TMI Action Plan Items (See NUREG-0737):

1)

II.D.3 2)

II.F.1 (4,5,&,6) 3)

II.F.3

4) II. K. 3.18 5)

II K.3.21 Remote Shutdown System (Compliance with GDC 19 and SRP Section 7.4; ICSB Position)

Diversity of Interlocks (Compliance with ICSB Branch Technical Position #3).

Discussion of the four I&C concerns addressed by letter dated April 16, 1981 from R. Tedesco (NRC) to E. L. Draper (GSU). These concerns are:

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1) Concerns related to IE Bulletin 80-06

2) Concerns related to IE Bulletin 79-27

3) Control systems failures resulting from HELBs

4) Multiple control system failures due to common power source or sensor (including sensor impulse lines) malfunctions.

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t IV. DRAWING REVIEW (WALK-THROUGH) 0F G.E. SAFETY-RELATED SYSTEMS To answer specific questions concerning the RPS, RHR, HPCS, LPCS, ADS, and RCIC.

Standby Gas Treatment System t

Standby Service Water System 1,

4 Control Building Air Conditioning System i

Safety Relief Valves Other G.E. systems with a new design as listed in FSAR Table 7.1-2 I

(time permitting)

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1 ENCLOSURE 2 RIVER BEND STATION UNITS 1 & 2

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4 ICSB REVIEW QUESTIONS I

421.001 Describe how the remote shutdown station design at River Bend complies (7.4) with GDC 19 as interpreted in Section 7.4 of the SRP. Clarification of l

these requirements was provided in the "ICSB Position for Remote Shutdown Capability for River Bend" issued to W. Cahill, Jr.- (GSU) from A. Schwencer l

(NRC) by letter dated May 17, 1982.

421.002 Provide a detailed response to the concerns' addressed by IE Bulletin 80-06 (7.3)

(Engineered Safety Feature (ESF) Reset Controls) issued to operating reactors March 13, 1980. For all safety-related equipment which does not

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remain in its emergency mode following an ESF reset, provide adequate 1

justification for the change of state of each piece of equipment or proposed

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corrective actions to prevent such changes (e.g., equipment returning to l

its normal operational status).

If the LRG II position paper regarding IE l

Bulletin 80-06 is determined to be applicable to River Bend, this fact must be documented. A response to IE Bulletin 80-06 was requested by

. letter dated April 16, 1981 (Request for Additional Information from OL Applicants.Regarding Four Instrumentation and Control Systems Concerns)

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from R. Tedesco (NRC) to E. L. Draper (GSU).

j 421.003 If reactor controls and vital instruments derive power from common (7.5) electrical distribution systems, the failure of such electrical dis-tribution systems may result in an event requiring operator action concurrent with failure of important instrumentation upon which these operator actions should be based.

IE Bulletin 79-27 addresses several concerns related to the above subject. You are requested to t

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1 provide information and a discussion based on each IE Bulletin 79-27

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concern. 'Also, you are to: -

1) Confirm that all a.c. and d.c. instrument buses that could l

l affect the ability to achieve a cold shutdown condition were reviewed.

Identify these buses.

2) Confirm that all instrumentation and controls required by emergency shutdown procedures were considered in the review.

Identify these instruments and controls at-the system level of

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detail.

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3) Confirm that clear, simple, unambiguous annunciation of loss of power is provided in the control room for each bus addressed in item 1 above.

Edentify any exceptions.

4) Confirm.that the effect of loss of power to each load on each bus identified in item 1 above, including ability to reach cold shutdown, was considered in the review.

5) Confirm that the re-review of IE Circular No. 79-02 which is required by Action Item 3 of Bulletin 79-27 was extended i

to include both Class IE and non-class IE inverter supplied instrument or control buses.

Identify these buses or confirm that,they are included in the listing required by Item 1, above.

This item was also addressed in the April 16, 1981 letter from R. Tedesco i

to E. L. Draper (See question 421.002).

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! l 421.003 If contro,1 systems are exposed to the environment resulting from the (7.7) rupture of reactor coolant lines, steam lines or feedwater lines, the control systems may malfunction in a manner which woUld causiconsequences to be more severe than assumed in safety analyses.

I&E Information Notice i

79-22 discusses certain non-safety grade or control equipment, which if subjected to the adverse environment of a high energy line break, could impact the safety analyses and the adequacy of the protection functions performed by the safety grade systems.

The staff is concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, requested to per-form a review per the I&E Information Notice 79-22 concern to determine what, if any, design changes or operator actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the event beyond the FSAR analysis. Provide the results of your review including all identified probleas and the manner in which you have resolved them.

1 The specific " scenarios" discussed in the above referenced Information Notice are to be considered as examples of the kinds of interactions which might oc' cur. Your review should consider analogous interactions as relevant to the BWR design, i

This item was addressed in the April 16, 1981 letter from R. Tedesco to l

E. L. Draper.

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'l 421.005 If two or more control systems receive power or sensor information from i

(7.7) comon power sources or comni6n sensors (including common headers or impulse lines), failures of these power sources or sensors or. rupture /

4 plugging of a comon header or impulse line could result in transients or accidents more severe than considered in plant safety analyses.

A number of concerns have been expressed regarding the adequacy of safety systems in mitigation of the. kinds of control system failures that could actually occur at nuc5ar p'lants, as opposed to those

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analyzed in FSAR Chapter 15 safety analyses. Although,the Chapter 15 analyses are based on conservative assumptions regarding failures of single control s'ystems, systematic reviews have not been reported to demonstrate that multiple control system failures beyond the Chapter 15 analyses could not occur because of single events. Among the types of I

events that could initiate such multiple failures, the most significant i

are in our judgment those resulting from failure or malfunction:of power f

supplies or sensors comon to two or more control systems.

Ta provide assurance that the design basis event analyses adequately bound multiple control system failures you are requested to provide the following information:

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Identify those control systems whose failure or malfunctions could seriously impact plant safety.

2)

Indicate which, if any, of the control systems identified in (1) receive power from comon power sources. The power sources considered should include all power sources whose failure or malfunction cou.id 1

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1ead to failure or malfunction or more than one control system and f

should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centhrs.

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Indicate which, if any, of the control systems identified in (1) l i

receive input signals from common sensors. The sensors considered I

should include, but should not necessarily be limited to, common hydraulic headers or impuls~e-lines fdeding pressure, temperature, level or other signals to two or more control systems.

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Provide justification that any simultaneous malfunctions of the control systems identified in (2) and (3) resulting from failures

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or malfunctions of the applicable common power source or sensor j

are bounded by the analyses in Chapter 15 and would not require i

action or response beyond the capability of operators or safety systems.

This item was also addressed in.the April 16, 1981 letter from R. Tedesto (.NRC) to E. L. Draper (.GSU).

421.006 Table 7.1-3 of' the FSAR contains no references to the Branch s

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Technical Positions listed in Table 7.1 of the Standard Review Plan.

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The FSAR should identify and justify any exceptions taken to these Branch Technical Positions.

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421.007 In the discussion on conformance to Regulatory Guide (RG) 1.47, i

(7.1.2.4)

FSAR, Section 7.1.2.4, the statement is made that system level annunciation (7.5) upon actuation of bypass or test switches is not fully implemented into l

l the design.

Identify all safety-related systems in which this feature f

is not implemented in the design and discuss plans to bring ' ~ase systems into conformance with RG 1.47.

In addition, determine whether the bypass i

and inoperable status indication system complies with position B2 of ICSB l

t Branch Technical Position 21, discuss the design philosophy used in the selection of equipment / systems to be monitored, and provide a complete list of system automatic and manual bypasses within the BOP and NSSS scope of supply as it pertains to the recommendations of R.G.1.47.

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The design philosophy should describe as a minimum the criteria to be employed in the display of inter-relationships and dependencies on equip-l ment / systems and should verify that the bypassing or deliberately induced inoperability of any auxiliary or support system will automatically indicate all safety systems affected.

421.008, A review of Table 7.1-3 indicates that several of the t

j (7.1) protection systems do not comply with General Design Criteria 20, 21, 22, 23, 24, and 29 as required by Table 7.1 in the Standard I

Review Plan (SRP). Provide justification for each system which does not comply with the specified criteria.

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421.009 Several previously reviewed BWR' installations, c.5., Grand Gulf,

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(7.1) include a start-up transient monitoring system to provide recordings of selected parameters during the start-up and warranty testing. There l

's no information in the FSAR which describes this system.

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system, or any similar system, is intended for use in the River Bend i

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j (a)

Identify all safety-related parameters which will be monitored i

j with the transient monitoring system during start-up.

j (b)

For each safety parameter identified above, provide a corcise description of how the associated circuitry merges or connects (either directly, or indirectly by means of 1-solationdevices) with the circuitry associated with the transient monitoring system.

Where appropriate, supplement this description with detailed I

electrical schematics.

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(c) Describe provisions of the design to prevent failures of this system from degrading safety-related systems.

t 421.010 Various instrumentation and control system circuits in the plant l

(7.1) (including the reactor protection system, engineered safety features

, actuation system, instrument power supply distribution system) rely on certain f

devices to provide electrical isolation capability in order to maintain the independence between redundant safety circuits and between safety l

circuits and non-safety circuits.

Therefore, provide the following information:

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(a)

Identify the types of isolation devices which are used j

as boundaries to isolate non-safety-grade circujts from..

f the safety-grade circuits or to isolate redundant safety-grade circuits.

(b)

Provide the acceptance criteria for each isolation device identified I

in response to part (a) above.

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(c)

Describe the type of testiri'"that was coriducted on the isolation g

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devices to ensure adequate protection against EMI,(i.e., noise),

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short-circuit failures, voltage faults, and/or surges.

421.011 The information provided in FSAR Appendix 1A (River Bend Station Positions l

( 7.1 ) on Post-TMI Requirements, NUREG-0737) is incomplete. Therefore, pro-(7.2) i 7.3 vide information as to how your design conforms with the following TMI t

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(7.5) Action Plan Items as described in NUREG-0737:

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(a)

II.D.3 Direct Indication of Relief-and Safety-Valve Position (b)

II.F.1 Accident Monitoring Instrumentation Positions (4), (5), and (6) l (c)

II.F.3 Instrumentation for monitoring accident conditions (R.G.1.97, Rev. 2)

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(d)

I I.K. 3.18 Modification of Automatic Depressurization System Logic In addition, provide a more explicit response for item II.K.3.21 (Restart of Core Spray and Low Pressure Coolant Injection Systems) including any 1

l proposed modification to the HPCS system (e.g., automatic restart capa-bility following termination by the operator).

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421.012 Describe in detail how physical separation is maintained between protection (7.1) channel circuits, protection logic circuits, and non-safety-related circuits.

(7.2)

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(7.3) The channel independence described in FSAR Section 8.3.1 (Referenced by both FSAR Sections 7.2 and 7.3 regarding independence of redundant protection i

j system channels) pertains to power cabling; not instrument channel and j

logic circuitry, i

I 421.013 Identify where instrument sensors..o[ transmitters supplying information to more than one protection channel or to both a protection channel and a

control channel are located in a common instrument line or connected to i

a common instrument tap. The intent of this item is to determine whether I

the failure of a instrument line or tap (such as break or blockage) in l

I conjunction with a single failure of a protective channel not associated I.

with the failed sensing line could prevent or delay protective action 1cng f

enough to result in unacceptable consequences. A 1-out-of-2 taken twice I

logic for combining channel trips to achieve protective actions appears i

to be particularly susceptible to this scenario.

If the LRG II position i

paper addressing this item is determined to be applicable to River Bend, this fact must be documented.

421.014 Provide an evaluation of the effects of high temperatures on reference (7.2) legs of water level measuring instruments subsequent to high-energy line

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(7.4) breaks, including the potential for reference leg flashing / boil off, the (7.5) j indication / annunciation available to alert the control room operator of erroneously high vessel level indications resulting from high temperatures, and the effects on safety systems actuation (e.g., delays).

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421.015 In the discussion in SectiorI7.2.2.2 concerning conformance to l

(7.2) Paragraph 4.15 of IEEE Standard 279, the statement is made that I

there are multiple set points within the RPS.

Discuss how mode switch operation affects RPS set points.

l 421.016 During an earlier review (Hatch Unit 2), the staff questioned i

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(7.2) the adequacy of protection affor.ded.the Class lE ftPS against j

possible sustained overvoltage d'r~undervoltage conditions from I

the non-Class lE RPS power supply. Several similar plants (i.e.,

Perry and Grand Gulf) have provided an electrical protection assembly j

(EPA) between the RPS and its power sources. No reference to the EPA could be found in Chapter 7 or 8 of the River Bend FSAR. State whether the River Bend design will incorporate an EPA between the RPS and its l

power sources.

If so, the FSAR should be modified accordingly.

If not, l

describe how the RPS is protected from degraded voltage and frequency I

condi tions.

l 421.017 Provide and describe the following information for NSSS and B0P safety i

(7.2) related setpoints:

j (7.3)

(a)

Provide a detailed discussion and/or reference to the methodology

! i used in determining safety system set points.

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(b)

Discuss any differences between the referenced methodology and i

i the methodology used for River Bend.

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(c) Verify that environmental error allowances are based on the highest value determined in qualification testing or identify i

and provide adequate justification for eacF exception.

(d)

Identify any time limits on environmental qualification of instruments used for trip, post-accident monitoring or engineered l

safety features actuation. Where instruments are qualified for t

only a limited time, specify the time and basis for the limited time.

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421.018 Identify any sensors or circuits used to provide input signals (7.2) to the protection systems which are located or routed through (7.3)

(7.4) non-seismically qualified structures. This should include sen-sors or circuits providing input for reactor trip, emergency safeguards equipment such as the Emergency Core Cooling System, i

and safety-grade interlocks.

Verification should be provided that i

l the sensors and circuits meet IEEE 279 and are seismically and environmentally qualified.

Of particular concern is the instru-i j

mentation relied upon to transfer RCIC and HPCS pump suction from the CST to the suppression pool.

i 421.019 Provide a description of the recirculation pump trip circuitry (7.2) used to mitigate the effects of Anticipated Transients Without Scram (ATWS). This description should include the criteria to i

which this circuitry is designed and implemented.

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421.020 Discuss the design provisions provided for conducting response time (7.2) tests as described in IEEE 338-1977, "IEEE Standard Criteria for the 4'

(7.3)

(7.4)

Periodic Testing of Nuclear Power Generating Station Safety Systems,"

l as supplemented by Regulatory Guide 1.118, " Periodic Testing of Electric i

Power and Protection Systems." As a minimum, provide the following j

information:

(a)

Identify any safety-related systems that do not have provisions for response time testing.

(b)

Confirm that the technical specifications will provide detailed l

j requirements for the operator which insure that blocking of a selected protection function actuator circuit is returned to normal operation after testing.

f (c)

Discuss response time testing of B0P and NSSS protection systems i

i using the design criteria described in position C.12 of R.G.1.118 and Section 6.3.4 of IEEE 338. Confirm that the response time 3

l testing will be provided in the technical specifications.

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421.021 The discussion on Scram Discharge Instrument Volume (SDIV) Water Level, (7.2)

FSAR Section 7.2.1.1, states that four pressure (level) transmitters sense SDIV level and provide inputs to the RPS. However, Section l

4.6.1.1.2.4.2.5 states that both float sensing and pressure (level)

I sensing devices are used for the automatic scram function.

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this apparent discrepancy.

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421.022 It has been noted during past reviews that pressure switches or (7.3) other devices were incorporated into the final actuation control l

(7.4) circuitry for large horsepower safety-related motors which are used s

i to drive pumps. These switches or devices preclude automatic (safety I

signal) and manual operation of the motor / pump combination unless per-r missive conditions, such as lube oil pressure, are satisfied. According-1 I

ly, identify any safety-related motor / pump combinations which are used i

in the River Bend design that operate as noted above.

Describe the j

pressure switches or other permissive devices used, the pctential for failure (including common mode failure) of these devices to preclude safety functions, and the capability provided for testing these devices.

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421.023 During oQr review, it has become apparent that the logic for manual (7.3) initiation for several Engineered Safety Feature (ESF) systems is interlocked with permissive logic from various sensors.

In some cases it appears that the permissive logic is dependent upon the same sensors as those used for automatic initiation of the system. The staff's i

position is that the capability to manually initiate each safety system should be independent of permissive logic, sensors, and circuitry used for automatic initiation of that system such that-a single failure will not prevent the initiation of a protective function by both automatic and manual means (see Section 4.17 of IEEE-279).

Identify each safety system i

at River Bend which is interlocked as described above and provide proposed modifications or justification for the existing-design.

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421.024 It appears that the ADS solenoid valves and associated circuitry cannot be (7.3) tested with the plant at power.

Provide plans for the testing of these j

valves and circuits or provide justification for the existing design.

Identify all other ESF systems where either a portion of the actuation

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circuitry or an actuation device cannot be tested during reactor power

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421.025 In FSAR Table 7.1-2, (Similarity of Safety-Related Systems to Licensed I

(7.3) Reactors) the design of the Main Steam-Positive Leakage Control System I

j (MS-PLCS) is identified as being similar to the Hartsville MS-PLCS de-sign.

Construction of the Hartsville plants has been delayed. Table 7.1-2 should therefore be modified to either reference another plant having the same MS-PLCS design as River Bend or list this system as a new design.

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j 421.026 Inthe discussion of the High Pressure Core Spray (HPCS) System, I

(7.3) Section 7. 3.1.1.1.1, the statement is made that the HPCS provides makeup water to the reactor until the vessel water ievel reaches

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the high level trip (trip level 8) and there is no high drywell pressure j

signal present. The high drywell pressure interlock has been removed l

in other BWR HPCS designs since the potential exists for flooding of the j

l steam lines and subsequent damage to safety-related valves and primary l

system piping.

State whether t'e. Ger Bend _ design will be modified h

l to eliminate this interlock.

If so, revise the FSAR and associated I

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drawings accordingly.

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421.027 Identify any safety systems that are shar_ed by both units. Discuss l

design criteria for instrumentation and controls shared between units.

421.028 According to Table 3.10A-1 several seismically-qualified instruments l

l (7.3) and controls (i.e., hydrogen analyzers, HVAC flow switches, and i

radiation monitors) have not yet been selected.

Discuss the current status of these items.

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l 421.029 In the discussions concerning the design of the Standby' Liquid Control (7.4) System in FSAR Sections 9.3.5 and 7.4.1.3,. very little information is i

provided concerning the design of the heating system required to prevent l

precipitation of the sodium pentaborate from the solution during storage.

i Provide a more detailed discussion on the design of the heating system, including associated instrumentation and controls.

Include information concerning the power sources used for the instrumentation and heaters and

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any alarms used to indicate failure of the heating system.

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421.030 Demonstrate that the Safety / Relief Valve (SRV) low-low set point I

i (7.6) function is adequate given a single failure which could cause an additional SRV to open during the time for which only one valve is permitted to be open (i.e., on second and subsequent valve pops).

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421.031 Amend Section 7.6 of the FSAR to include a discussion on high f

(7.6) pressure / low pressure interfaces and associated interiocks. Discuss-g

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i how each of the high pressure / low pressure interfaces in your design i

conforms to the requirements of Branch Technical Position ICSB 3, I

" Isolation of Low Pressure Systems from the High Pressure Reactor Coolant System." Also, discuss how the associated interlock circuitry conforms to the requirements of IEEE 279.- The-discussion should include illustrations from applicable drawings.

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421.032 Table 7.1-2 indicates that the design cf the relief function of the (7.7)

Safety / Relief Valves is a new design. Describe how the River Bend f

design differs from that of previous designs, i.e., Perry and Grand Gul f.

421.033 Amend Section 7.7 of the FSAR to include a discussion of the process 1

(7.7) computer system.

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421.034 Identify all plant safety-related systems, or,srtions thereof, for (7.1) which the design is incomplete at this time and estimate when the design details, including electrical schematic drawings, will be avail-able for staff review of these systems.

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421.035 Identify. where microprocessors, multiplexers, or computer systems l

(7.1) are used in or interface with safety-related systems.

421.036 The staff has recently issued Revision 2 to Regulatory Guide 1.97, (7.5) " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Following an Accident."

l This revision reflects a number of major changes in post-accident l

instrumentation, and includes specific implementation requirements

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for plants in the operating license review stage.

Discuss the schedule

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for complying with this Regulatory Guide.

l 421.037 The FSAR information provided describing the separation criteria for (7.1) instrument cabinets and the main control board is insufficient.

Please r

discuss the separation criteria as it pertains to the design criteria of IEEE Std. 384-1977, Sections 5.6 and 5.7.

Detailed drawings should I

i be used to aid in verifying compliance with the separation criteria.

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