ML17272A375

Requests Addl Info to Complete Review of FSAR Sections on Instrumentation & Control Sys.Requests Revision of FSAR Sections & Adoption of Relevant Resolutions Developed in Similar Reviews

ML17272A375
Person / Time
Site:	Columbia
Issue date:	03/28/1979
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
NUDOCS 7904160321
Download: ML17272A375 (32)

Text

C March 28, 1979 Docket No.: 50-397 Nr. Neil 0. Strand Mashington Public Power Supply System 300 George Mashington May P.O.

Box 968 Richland, Mashington 99352

SUBJECT:

FIRST ROUND QUESTIONS ON THE MNP-2 OL APPLICATION - ICSB In our review of your application for an operating license for the MNP-2 facility, we have identified a need for additional information which we require to complete our review.

The specific requests are contained in Enclosure 1 to this letter and are the eighth set of our round-one questions; additional requests related to other portions of the MNP-2 facility will be sent during the next several weeks.

In order to maintain our present

schedule, we need a completely adequate response to all questions in Enclosure 1 by June 18, 1979.

r The attached set of round-one questions represents the review effort of the Instrumentation and Controls Systems Branch.

Mhere we have been able to formulate statements of staff positions (RSP),

we have included these in Enclosure l.

Mith respect to this set of round one questions, we are c'oncerned that the present status of those sections of your FSAR which relate to the design of the instrumentation and control systems of the MNP-2 facility, is inadequate to complete our review in support of the issuance of an operating license.

This is reflected in the total number of questions asked to date in this area of review (i.e.,

113 questions covering 37 pages)

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Of even greater concern to us is that many of

these questions have been previously asked on the dockets of several faci.l.ities which are very similar to the MNP-2 facility (e.g.,

Zimmer and LaSalle);

In particular, about three-quarters of our 113 questions to date on the MNP-2 docket leave been asked previously.

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YIr. Neil 0. Strand this repetitious transmittal of similar, if not identical, questions on each succeeding application for an operating license is both burden-some to the NRC staff and wasteful of our limited resources.

te believe that this large number of questions is equally, if not more, burden-some to you.

It is our belief that you have made a conscientious effort to upgrade the overall quality of your FSAR since your application was initially tendered in Yarch 1977.

However it is also our belief that you should have continued your efforts to update those sections of your FSAR related to the design of the MNP-2 instrumentation and controls systems to ensure consistency with the FSAR's for similar facilities for which we are nearing completion of our review.

Accordingly, to resolve this particular problem, we request you to a.

Revise Section 7 and those other portions of the FSAR which relate to the design of the instrumentation and control systems (e.g.,

Sections 3.10, 3.11, 15 and 16).

In particular, the content of Sections

7. 6 and ?. 7 of the FSAR does not provide the information we need to complete our review.

This revision should follow the guidance contained in Regulatory Guide 1.70 and in our Standard Review Plan, NUREG-75/087.

b.

Follow the guidance provided by the staff in its previous questions and statements of staff positions on similar facilities, including the staff' evaluations in its SER's.

c.

Adopt for the MNP-2 facility, where practicable, the resolutions which were developed by other applicants and the staff for problems common to a number of plants presently undergoing an OL review.

Accordingly, we believe that you should take the initiative to revise your application to remove the repetitious design deficiencies and the conflicting or vague statements in your FSAR.

The main thrust of your effort should be to provide sufficient information to minimize the staff questions and positions.

Such action on your part will permit us to concentrate out safety review effort on those features of your design which are truly unique to the MNP-2 facility and provide for better utilization of our mutual resources.

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Mr. Neil 0. Strand w 3 To assist you in updating your FSAR, we are enclosing a list which relates the questions on your docket to similar questions on the dockets of similar facilities (Enclosure

2) ~

We are also enclosing a tabulation

{Enclosure 3) which cross-references our concerns discussed above (e.g.,

conflicting or confusing statements in the WNP-2 FSAR, design problems similar to those of other facilities and failures to meet industry standards) wfth the individual questions in Enclosure l.

Since we recognize that you may wish to discuss in detail, the round-one questions and that you may wish to discuss at a management

level, our three specific requests outlined above, we request that your representatives meet with us in mid-April.

Please contact with Mr. Lynch at (301) 492-7831, to arrange this meeting. If you require any discussion or clarification of the enclosed requests or of the subject matter of this'etter prior to this meeting, please contact us.

Sincerely, Origina1 signed by:-

D. B. Vassallo, Assistant Director for Light Water Reactors Division of Pr'oject Management

Enclosures:

As stated cc:

See attached sheet Distribution Docket File LWR (f4 Reading D. Lynch NRC PDR

'Local PDR R.

Boyd D. Vassallo F. Williams S.

Varga M. Service R. Mattson D.

Ross J. Knight R. Tedesco R.

DeYoung V. Moore W. Kreger M. Ernst R. Denise OELD IE (3) bcc:

J.

Buchanan, NSIC T. Abernathy, TIC ACRS (16)

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Washington Public Power Supply System ccs:

Joseph B. Knotts, Jr.,

Esq.

Debevoise 4 l.iberman 700 Shoreham Building 806 Fifteenth Street, N.

M.

Washington, 0.

C.

20005 Richard Q. Quigley, Esq.

Washington Public Power Supply System 3000 George Washington Way P. 0.

Box 968

Richland, Washington 99352 Nepom 8 Rose Suite 101 Kellogg Building 1935 S.

E. Washington Nilwaukie, Oregon 97222 Ns.

Susan N. Garrett 7325 S.

E. Steele Street

Portland, Oregon 94206 Nr. Greg Oarby 807 So. Fourth Avenue Pasco,.Washington 99301 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Nr. 0. K. Earle Licensing Engineer P. 0.

Box 968

Richland, Washington 99352

ENCLOSURE 1

030.0 INSTRUMENTATION AND CONTROL SYSTEMS BRANCH 031. 080 (1. 2. 1)

(3. 10)

'3.'ll)

(3. 1.1. 2)

(4.4.3)

(6.2.4)

(6.7.1)

(7. 1. 2)

(7.2. 1)

(7.'3'.1)

(7. 6. 1)

'a 0 b.

Clarify the discrepancy between the definition of passive failures in electrical, instrumentation, and control systems in Sections 1.2. l. 1'. 1. 1.2. L and 3. 11. 2.3, 6.2. 4. 1. 1, and 6. 7. l. 2; c of the FSAR Clarify the discrepancy between your response to Item 031.011(b) and Figures 7.2-lb and 7.2-lc of the FSAR.

Your FSAR still contains many conflicting or confusing statements which must be resolved so that our review may proceed.

Accordingly, provide the following information:

(7. 6. 2)

(8. 2. 1)

(8. 3. 1)

(031. 001)

C.,

d.

Revise Section 4.4.3.3.3 of the FSAR to provide the actual values of the measured parameters which are to be used in the WNP-2 facility.

This section indicates "typical" values; at the OL stage of review, we require actual values.

Clarify the discrepancy between the 25 percent pump speed interlock value described in Section 4.4.3.3.3.

1 and the 20 percent value which is given in Figure 7.7-7c of the FSAR.

e.

Describe the primary and secondary modes of operation of the isolation valves which are referenced in Section 6.2'.2 of the FSAR.

Clarify the discrepancy between the description of the solenoid valves in Section 6.2.4.2 of the FSAR and the design which is presented in your response to Item 031.001(h).

g.

Clarify the discrepancy between the isolation valve arrangement which is described in Section 6.2.4.3.2. l. 2. 1 and that which is shown in Figures 5.4-9a and 7.4-1a of the FSAR.

h.

Provide a cross-reference between GE diagram numbers (EXX-XXXX)which are used in the FSAR diagrams and are included in the list of references on these

diagrams, and the WNP-2 figures.

Clarify the reference to Figure 1.2-5 of the FSAR which is contained in Section 7.2. l. 1.4.2.e by specifying the exact physical location and arrangement 031-25

k.

of the turbine governor oil line pressure switches and their sensing lines.

Indicate in Section 7.2. l. 1.4.4.5 of the FSAR, the delay time before the" reactor mode switch scram is automatically bypassed.

Clarify the discrepancy in the instrumentation range between note 4 of Table 7.1-2 and line 3 of Table 7.2-1 of the FSAR (i.e., reactor vessel lower water level).

m.

n.

Clarify the discrepancy between the response to Item 031.001(s) and the content of drawing 807E180TC, Sheets 1 through 9.

Clarify the reference in Section 7.3. 1.2.7 of the FSAR to Sections

8. 2. 1 and 8. 3. 1.

This clarification should clearly state the range of voltage and frequency for which all Class lE instrumentation and control equipment is qualified and the range of voltage and frequency to which it will be exposed in the MNP-2 fac i 1 ity.

Clarify in Section 7.6.1.4.2 of the FSAR, the divisional assignments which are made for the motor-generator sets of the reactor protection system.

Specifically, justify the designation of these busses as "critical."

o.

P ~

Clarify the discrepancy between Figure 5.2-6 and Tables 7.2-1, 7.3-2, 7.3-3, 7.3-4 and 7.3-5 of the FSAR with regard to the low level set point and range.

Clarify the discrepancies in the pressure trip setting between the Amendment 1 revision of Table 6.3-2 and other submittals of information in the FSAR such as Table 7.3-3 for the spray valve differential pressure.

Clarify the discrepancy between your statement in Section 7.6. 1.8. 1.2 of the FSAR regarding the uniqueness of the RPT and the statement in Item 37 of Table 7. 1-2 which claims your RPT is identical to that of Zimmer.

Clarify the reference to four RPS divisions in Section 7.6. 1.8.3.2 of the FSAR.

It is our understanding that there are only two RPS divisions.

031 "26

s.

Clarify the discrepancy between the content of Sections

3. 11 and 7. 6. 2.8. 2. l. 1.4 of the FSAR.

031. 081

'3.io)

(3. >>)

031. 082 (7.6.2) 031. 083 (3. 11. 3)

(3. 11A)

(031.006)

(031.056)

(031.059) t.

Clarify the discrepancy between Sections

3. 10 and

7. 6.

2'8. 2. l. 5 of the FSAR.

(Note that the RPT system is not listed in Table 3. 10-1.)

u.

Clarify the references in Section 7.6. 1.7.8 of the FSAR to Table 3. 11-4 for the reactor and control building environments.

Identify each type of relay in the WNP-2 facility which must be energized or which must remain energized, during a seismic event.

For each of these relay types, provide the following information:

(1) the minimum voltage at which it must operate; (2) the voltage at which it was seismically qualified; (3) the normal operating voltage; and (4) the locations and functions of this type of relay.

Where a particular relay was not qualified by test or was not tested in both the energized and de-energized state, justify the seismic qualifica-tion of the relay.

Demonstrate that the safety-related equipment discussed in Section 7.6.2.3.2.

1 of the FSAR satisfies the requirements of General Design Criteria 1, 2, 3, 4, 13, 14, 16, 19, 23, and 55.

Provide this demonstration of compliance with the requirements of Appendix A to 10 CFR Part 50 in other sections of the FSAR where it is missing.

Neither your response to Item 031.006 nor Appendix 3. 11A of the FSAR satisfy our need for additional information on equipment qualification.

In order to ensure that your environmental qualification programs conforms with General Design Criteria 1, 2, 4 and 23 of Appendix A and Section III and XI of Appendix 8 to 10 CFR Part 50, and to the national standards (e.g.,

IEEE Std 323-1971) mentioned in the Acceptance Criteria contained in Section

3. 11 of the Standard Review Plan, NUREG-75/087, provide an amended response to Item 031.006 for:

a.

The logic equipment for the standby gas treatment system (031.006, item (d) of the second paragraph).

b.

The following sensors:

(1) the rod block monitor

'low transmitters; (2) the main steam line tunnel temperature thermocouple; and (3) 822-N024A.

031-27

031.084 (3. 11A) 031. 085 (7. 6. 1)

(7. 7. 1) 031. 086 (7.4.1)

(7.4. 1)

(9.3.5) 031. 087 (3. i1.2) c.

All items listed in guestions 031.056 and 031.059.

The specification requirements of Table 3. 11A-1 of the FSAR are incomplete since they do not address the maximum and minimum values of all of the parameters which are cited in Section 3(7) of IEEE Std 279-1971.

Accordingly, provide the required data for all Class lE components.

Several systems (e.g.,

the safety/relief valve discharge line temperature monitoring system and the reactor vessel head leak detection system) are listed in both Sections 7.6 and 7.7 of the PSAR.

However, Section 7.6 should describe only those systems required for safety while Section 7.7 should describe only those systems not required for safety.

In conformance with the guidance contained in Section 7.7 of Regulatory Guide 1.70, Revision 2, safety-related systems should not be listed in Section 7.7 of the FSAR.

Accordingly, revise your FSAR to eliminate such ambiguous design descrip-tions of safety-related systems and nonsafety-related systems.

The standby liquid control system (SLCS) is designated in Section

7. 4. l. 2. 3. 1 of the FSAR as a special plant capa-bility event system in the WNP-2 facility.

To assure the availability of the SLCS, you have provided in parallel, two sets of those components required to actuate this system.

However, our review indicates that you have not provided redundant heating systems.

Additionally, the heating equipment supply emergency bus is neither identified nor is it redundant.

We have concluded, therefore, that your statement in Section 9.3.5.3 of the FSAR that "...a single failure will not prevent system operation..."

is not true.

(Note that this matter has been resolved in similar facilities by providing redundant heating systems.)

Accordingly, provide a modified design of the SLCS which satisfies the single failure criterion.

Alternatively, justify your present design.

With regard to Section

3. 11.2.3 of the FSAR, provide the following additional information and clarifications:

a.

Provide a copy of the procedures for the following aging simulations:

(1) thermal; '(2) radiation; (3) operation; and (4) seismic.

b.

Provide justification for the aging temperature which was used with respect to the maximum normal environmental conditions which are listed in Table 3.11-1 of the

FSAR, 031-28

Indicate the thermal aging acceleration rate and provide the basis for this rate.

Indicate the thermal aging time used for each plant location listed in Table 3.11-1 of the FSAR which contains a valve that has been qualified in accordance with IEEE Std 382-1972.

Identify the valves which are so qualified.

Provide information similar to that requested in Items (b) through (d) above for radiation aging.

In addition, describe how the effect of the neutron fluences was considered.

Provide your criteria for determining the limits of an actuator family including:

(1) the definition of the limits of an actuator family; (2) the criteria which were used to a'ssure that the sample valve operator is a valid representative of the family; and (3) a demonstration of how the criteria were applied.

Provide a table of the following information for all Class 1E valve actuators in the WNP-2 facility:

(1) the equipment specifications in accordance with Section 3 of IEEE Std 382-1972; (2) an identification of the family membership; and (3) an identification of the samples.

Indicate the number of operating cycles to which each test specimen was subjected.

Indicate the frequency range which was used in the seismic qualification and aging of the samples.

(Note that the frequency range permitted by IEEE Std 382-1972 does not agree with our acceptance criteria contained in Paragraph II.l.a of Section

3. 10 of our Standard Review Plan, NUREG-75/087.

We will require conformance with our'positions in this latter document.)

Describe how you assure that equipment not qualified for all service conditions, will not spuriously operate during exposure to service conditions, including excessive exposure times during which this equipment is not required to function to mitigate the effects of accidents on other events.

031-29

031

~ 088 (6. 2. 4)

(7. 6.1)

(7. 7. 1)

(9.3.5) 031.089 (1.7)

(6.7.3) 031.090 (7.2.1)

(F7.2-la)

(031.032),

031. 091 (F7.2-9)

Clarify the discrepancies between the following sections and figures of the FSAR with regard to isolation of the reactor water cleanup system when the standby liquid control system is initiated:

(1) Sections 6.2.4.3.2. l. 1.7, 7'. 1.8,

7. 6. 1.4. 3. 6, and 9. 3. 5.2; (2) Figures

7. 3-lla, 7.4-3,.7.7-14; and (3) Table 7.3-13.

The single failure analysis presented in Section 6.7.3.1 of the FSAR is inadequate.

Accordingly, revise this section to include single failures of electrical components such as the spurious closing of relay contacts on K4.

(Refer to GE Dwg 851E708TD).

Provide the electrical schemati'c and one-line drawings of this system for our review.

In Section 7.2. l. 1.2 of the FSAR, you state that the reactor protection system (RPS) is Class lE.

However based on our review of similar facilities (e. g.,

Zimmer ),

we believe that this statement is incor rect since the WNP-2 motor-generator sets of the RPS are probably not Class 1E equipment.

Accordingly, correct the discrepancy between Sections

7. 2. l. 1. 1 and 7.2. l. l. 2 of the FSAR regarding the qualification of the RPS motor-generator sets.

Alternatively, demonstrate that all components of the RPS are Class lE equipment.

Additionally, describe how the design and implementation of your RPS satisfies the requirements of Section 6.6 of IEEE Std 379-1972, with special emphasis on the last paragraph of this section.

In facilities similar to WNP-2, the wiring from the RPS relay contacts 14A and 14C, via cabinet penetration Y, and 14E and 14G, via cabinet epentration Z, appears on terminal strip CC.

(Refer to the GE drawing 807E166TU.)

This wiring is powered from two separate Class lE d-c busses.

Insufficient physical separation was provided between these 'busses on terminal strip CC, the associated

cables, and in penetrations Y and Z, which also serve the plant process computer system.

Our concern is that there may be insufficient physical separation in the RPS cabinets of the WNP-2 facility since it is our understanding that they are being manufactured by the same vendor.

Accordingly, if this same problem exists in the RPS cabinets of the WNP-2 facility, we will require you to provide an acceptable design for the routing of Class IE circuits inside the RPS cabinets.

Alternatively, demonstrate that our concern on this matter is not applicable to the WNP-2 facility.

031-30

031. 092 (F7. 2-9)

In facilities similar to WNP-2, the cabinet lighting circuit which is not treated as an associated circuit, crosses cabinet penetration 187 in RPS cabinet A, and as a

result, becomes associated with the containment isolation system wiring going to penetration 187.

Our concern is that the physical separation provided in the RPS cabinets of the WNP-2 facility may not satisfy either the requirements of IEEE Std 279-1971 or the WNP-2 separation criteria.

Accordingly, if this problem of physical separation exists in the RPS cabinets of the WNP-2 facility, we will require you to take the following corrective actions:

a.

Provide a modified design for the routing of non-Class lE circuits in RPS cabinet A which satisfies the separation criteria; b.

Review the design of all other Class lE cabinets for similar defects and indicate the cabinets which you reviewed.

031. 093 (7. 2. 1)

(7 2 2) c.

Advise us of your findings and plans for the modifica-tions necessary to satisfy the separation criteria.

d.

Identify and justify all exceptions which you may take to items (a), (b) or (c) above.

e.

Provide panel layout drawings and one line diagrams which show the routing and physical separation between the reactor trip sensors and:

(1) the high level cut-offs for the HPCS and RCIC; and (2) the post-accident reactor vessel level indication system.

Provide in Sections 7.2. l. 1 and 7.2.2.

1 of the FSAR, the design criteria and a description of the scram discharge volume switches and their qualification testing, including the following information:

(1) the manufacturer; (2) the type of float (i.e., whether it is self-equalizing or sealed);

(3) the float material and the magnet material; and (4) the qualification test conditions including the water temperature; the pressure; the duration of the test conditions; the number of test cycles; the period between test cycles; the extremes of external temperature,

pressure, and humidity; and the radiation source,

strength, and dose.

031. 094 (7.2.2)

Your response to Item 031,033 is incomplete since it 'does not indicate that the individual system level indicators 031-31

(7. 3. 2)

(031.033) 031.095 (7.3.2) 031. 096 (7 ~ 3 2) 031

~ 097 RSP (031.026) 031. 098 RSP (7.6)

(15.4. 1) can be actuated from the'ontrol room by the operator.

Accordingly, revise Sections

7. 2. 2. l. 2. l. 5,

7. 3.2. 1.2. 1. 6. 2, and 7.3.2.2.2. 1.5. 1.2 and all other similar sections of the FSAR, to describe the provisions you have incorporated into the design of the WNP-2 facility to satisfy Position C.4 of Regulatory Guide 1.47, "Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems,"

May

. 1973.

(Note that this position is not intended to address the testing of annunciators, but is'intended to provide manual initiation of system level indication of inoperable and bypassed status.)

Your discussion of how the instrumentation and controls satisfy the requirements of Section 4. 1 of IEEE Std 279-1971, is inadequate.

Indicate the pick-up and drop-out voltage values of the bus voltage relay.

Provide justification in Section 7.3:2.1 of the FSAR for indicating a loss of power to the motor starters by de-energizing the indicating lamps.

In your response, discuss how the reactor operator can distinguish between a failed lamp and a system bypass.

Demonstrate how your design of these indicating lights satisfies the require-ments of IEEE Std 279-1971 with regard to providing the reactor operator with timely and unambiguous information.

Your response to Item 031.026 is unacceptable.

It is our position that isolation devices which are used to provide electrical independence between Class lE and non-Class lE equipment, must:

(1) be designed,

'ualified, and implemented in accordance with all of the requirements for Class lE equipment; and (2) be an integral part of the system which they are intended to protect.

Accordingly, we require you to revise your response to Item 031.26 and to provide all the information requested in this previous request.

(Note that this matter has been resolved in similar facilities by modifying the design.)

It is our position that the use of the rod worth minimizer (RWM) is unacceptable for safety-related functions since it does not satisfy the requirements of IEEE Std 279-1971.

Accordingly, we require you to delete this system from Section 7.6 of the FSAR.

(Note that you claim credit for the RWM in Section 15.4. 1.2.2.

1 of the FSAR which implies that the RWM is a safety-related system.

031-32

031.099 (3.4)

(7.3.1)

(031.030) 031. 100 (7. 3. 2)

The response to Item 031.030(c) is incomplete since you do not discuss the consequences to electrical equipment in the event of internal flooding.

Section 7.3. 1.2.8.

1 of the FSAR is similarly incomplete.

Accordingly, provide a revised response to Item 031.030(c) which discusses the protection of Class lE equipment from internal flooding (e.g.,

a failure of either the main condenser cooling line or of the fire protection system).

In Table 7. 1-2 of the FSAR, you indicate that many of your instrumentation and control systems are identical to those of LaSalle and Zimmer.

Ouring the course of our review of these facilities which are similar to the WNP-2 facility, we encountered a number of errors in the smplementatson of the bas>c GE design.

Our concern ss that these same errors, or similar errors, could occur in implementing the electrical design of the WNP-2 facility.

In particular, we find that your analyses in Section 7.3.2. l. 2.3.

1 and 7'.2. 2. 2.3. 1. 1 of the FSAR, to determine compliance with the requirements of IEEE Std 279-1971 are too general in content.

We provide guidance for the information we need in Section 7.2 of the Standard Review Plan, especially in Appendix 7.2.A.

Specific examples of areas where we require additional information are presented in Items 031.081, 031.084, 031.091, and 031.092 of this enclosure.

Accordingly, provide more specific analyses of how you have implemented, in detail, the basic GE electrical design in the WNP-2 facility.

References to other sections of the FSAR are acceptable in lieu of repeating this information in Section 7.3.2.1.2.3.1.

031. 101 (7.3.2) 031. 102 (7. 3. 2)

In Section

7. 3. 2. 2. 2. 3. l. 4 of the FSAR, you state that:

"All components used in the isolation system have demon-strated reliable operation in... industrial applications."

Vague or general statements like this are unacceptable without the supporting basis'ccordingly, identify in Section 7.3.2 of the FSAR, the equipment which has been environmentally qualified by previous operating experience and, for each item, provide the basis for the extrapolation in accordance with the requirements of IEEE Std 323-1971.

Oescribe in Section 7.3.2.

1 of the FSAR, your proposed methods to provide for emergency operation of emergency switches and valves which are locked.

(Refer to your discussion on locked safety equipment in Section 7.3. 2. 1. 2. 3. 1. 14(d) of the FSAR. )

031"33

0

031. 103 (7 3)

(7.6)

(F7.6-5) 031. 104 (7.5)

(031.034)

L Your description of the area temperature monitoring system in Sections 7.3 and 7.6 of the FSAR, is insufficient.

Accordingly, provide the following additional information:

a.

Identify the interfaces between the Class lE and non-Class lE parts of this system.

b.

Describe how redundant components are electrically isolated and physically separated.

c.

Describe how the electrical isolation devices were qualified and indicate the range of this qualifi-cation in terms of voltages and currents.

d.

Provide the schematics for the Class lE portions of this system, including the isolation devices.

e.

Provide the bases and methods which were used to select the samples whjch were tested in accordance with the criteria identified in Item (c) above.

The description of the control room in Section 7.5 of the FSAR is incomplete as are the figures in this section.

(Refer to Item 031.034.)

Accordingly, provide a layout drawing of the control room showing in sufficient detail, the following information:

a.

The location and identification of each cabinet and panel.

b.

The location and routing of each conduit and cable tray and pan.

c.

The location and field of each emergency light.

d.

The location and identification of each indicator and relay contact which satisfies the requirements of Sections

4. 19 and 4.20 of IEEE Std. 279-1971.

031. 105 (7.4.1.4)

Your description of the procedure for reactor shutdown from outside the control room is inadequate.

Accordingly, provide the following additional information:

a.

Provide plant layout sketches which show where the switches are located.

031-34

0

b.

Describe the method which will be used to seal the transfer switches.

c.

Describe the'onsequences of an inadvertent actuation of one or more of the switches.

031. 106 (7.6.1) 031. 107 (7.6.2) 031. 108 (3 ll)

(4.4.3)

(5.'4.')

(7.6.1)

(7'.6.2) d; Identify and justify each transfer switch which is not wired to the bypass and inoperable status indication system.

e.

Describe the methods and indications available outside of the control room by which the operator can:

(1) verify relief valve operation; (2) determine the reactor pressure, the coolant level and the coolant temperature; (3) determine the suppression pool level and the temperature of the water in the pool (4) determi ne the containment pressure; and (5) determine the service water flow rate and the change in the coolant temperature through the RHR heat exchangers.

Confirm in Section 7.6.1.13 of the FSAR that the primary containment atmosphere monitoring system i

1 d'nsors, wall be seismically and environmentally qualified.

Identify and justify all exceptions.

Describe your methods to seismically qualify the d yw ll h d and ox e

ygen monstorsng system.

Indicate the required response spectra for which the system is qualified and identify the limiting component.

In section 7.6.2.13.3.5.19 of the FSAR, you indicate that the post-LOCA containment monitors are in continuous operation.

This appears to be a change from previous and ox boiling water reactor (BWR) designs in which th h d ygen subsystems were activated by an accident e

y rogen signal.

Discuss this new feature of your design.

Figures 5.5-2, 7.7-7 and 7.7"8 and Sections

3. 11, 4.4.3.3,

5. 4. 1, 7. 6. 1. 8 and 7. 6. 2. 8 of the FSAR contain many discrepancies and are, therefore, unacceptable.

Provide a consistent set of drawings and other information which represents the design of the reactor recirculation system for the WNP-2 facility.

In your response to this item:

a.

Provide setpoint information (i.e., the range of the instrument, its accuracy and its setpoint) for the following items:

(1) the low total feedwater permis-sive (H13"P634); (2) the steam line recirculation 031-35

pump differential temperature (K634); (3) C001A rated speed permissive for CB3A Trip 2; (4) a pump speed greater than 15 percent but less than 40 percent; (5) C001A less than rated voltage permissive for closing CB2A; (6) the generator protective trip voltage; and (7) the reactor power permissive for low speed start.

b.

Identify which control option of note 7 in Figure 7.7-8 of the FSAR is applicable to the MNP-2 facility.

C.

Identify the inputs which are received from reference document No.

4 of Figure 7.7-7a of the FSAR (i.e.,

C12-1050).

Provide this reference and clarify the function of these trips and the ATMS trips shown on Sheet f of Figure 7.7-7.

d.

Clarify the discrepancy between the ATMS trips shown in Figure 7.7-7 of the FSAR and the logic description given in Section 7.6.1.8.1.

e.

Indicate the signal source for the "permissive when low speed auto start sequence is not activated."

(Refer to Figure 7.7-7 of the FSAR.)

f.

Indicate the signal source for the "transfer to high speed initiated" auxiliary device.

(Refer to Figure 7.7-7 of the FSAR.)

g.

Describe the initating circuitry for, and the location of, the hydraulic line containment isolation valves.

h.

Clarify the discrepancy between the setpoint stated in Sections 4.4.3.3.3.a and 5.4.1.3 of the FSAR.

i.

Provide justification for not environmentally qualify-ing the 6.9 Kv switchgear.

031. 109 (7. 7. 1)

Provide in Section 7.7. 1.2 of the FSAR, the results of a failure mode and effects analysis for the reactor manual control system analyzer.

Identify the design features which are provided to detect these failures.

Describe the test procedures, including the test frequency, which will be used to detect these failures.

031. 110 (7. 7. 1)

(15)

The information presented in Sections 7.7.1.5 and 15 of the FSAR is incomplete with regard to load following operations.

Accordingly, describe the interfaces between 031-36

031. 111 (9. 5. 2) the system dispatcher and the WNP-2 control systems (e.g.,

the turbine-generator and the recirculation flop control systems.)

The description of the intra-plant radio system inspection and testing in Section 9.5.2.2 of the FSAR is inadequate.

Oescribe the preoperational and periodic testing which assures that radio transmissions will not cause spurious operation of relays

and, as a result, negate the protective function of Class lE equipment.

(This question is similar to Item 031. 123 on the LaSalle docket.)

031. 112 (7. 6. 1)

(7.6.2) 031. 113 RSP (15.4. 1. 2)

Oescribe in Sections

7. 6. l. 1 and 7. 6. 2.

1 of the

FSAR, how the power cables and the refueling interlock circuits are separated on the refueling crane.

It is our position that the rod sequence control system does not satisfy the requirements of IEEE Std 279-1971 and, therefore, is unacceptable for the prevention of a control rod withdrawal accident.

Accordingly, we require you to provide a modified design for the WNP-2 facility.

031-37

ENCLOSURE 2 WNP-2 ICSB UESTIONS PREVIOUSLY ASKED ON OTHER DOCKETS MNP-2 031. 001 031. 004 031. 005 031. 006 031. 008 031. 009 031. 012 031. 013 031. 014 031. 015 031. 016 LaSalle 031. 003 031.007 031. 013 031

~ 017 031. 027 031. 028 031. 083 031. 142 031. 149 031. 084 031.004 031. 005 031. 142 031. 010 031. 087 031. 009 031. 020 Zimmer 221.004 221. 005 221. 015 221. 035 221. 040 221.220 221. 222 221. 266 221.267 221. 222 221. 210 221. 185 221. 028 221. 254 221. 029 221. 036 Sus uehanna 032.016 032.029 032. 019 032. 004 032.023 Grand Gulf 031. 044 031. 014 031. 021

ENCLOSURE 2 Continued)

MNP-2 031. 017 031. 018 031. 021 031. 023 031. 024 031. 025 031. 026 031. 028 031. 030 031. 031 031. 032 031. 033 031. 034 031. 035 031. 036 LaSal 1 e 031. 011 031. 097 031. 018 031. 023 031. 021 031. 049 031. 069 031. 034 031. 035 031. 102 031

~ 031 031. 094 031. 098 031. 033 Zimmer 221. 021 221. 042 221. 038 221. 037 221. 061 221. 261 221. 044 221. 045 221. 050 221. 052 221. 257 221. 049 221. 058 Sus uehanna 032. 026 032.009 032.027 032. 037 032. 035 032. 016 Grand Gulf 031. 029 031. 013 031. 035 031. 036 031. 037

ENCLOSURE 2 (Continued)

MNP-2 031. 037 031. 038 031. 039 031. 040 031.041 031. 044 031. 045 031. 047 031. 049 031. 050 031. 052 031. 053 031. 055 031. 056 031. 075 LaSalle 031. 106 031. 108 031. 107 031. 044 031. 043 031. 073 031. 052 031. 112 031. 053 031. 114 031. 014 031. 054 031. 119

= 031. 177 031. 206 Zimmer 221. 010 221. 130 221. 053 221. 260 221. 060 221. 065 221. 069 221. 032 221. 088 221. 071 221. 373 032.038 032. 036 032. 040 032.041 032.044 032. 024 032. 046 032. 048 Grand Gulf 031. 032 031. 030 031.028

ENCLOSURE 2

Continued MNP-2 031. 079 031. 080 031. 081 031. 086 031.087 031. 088 031 F 089 031. 090 031. 091 031. 092 031

~ 095 031. 096 031. 098 031. 100 031. 101 LaSalle 031. 140 031. 001 031. 083 031. 213 031. 156 031. 012 031. 152 031. 045 Zimmer 221. 174 221.384 221. 225 221. 003 221. 362 221. 195 221. 258 221. 259 221. 057 221. 026 Sus uehanna 032. 016 032. 017 032.022 032. 018 032. 045 031. 028 Grand Gulf 031. 020 031. 043 031. 023 031. 038 031. 018

ENCLOSURE 2 Continued WNP-2 LaSalle Zimmer Sus uehanna Grand Gulf 031. 102 031. 103 031. 105 031. 106 031. 107 031. 108 031. 109 031. 110 031. 111 031. 112 031. 113 031. 104 031. 158 031. 050 031. 115 031. 213 031. 157 031. 120 031. 123 031. 243 031. 162 221. 005 221. 354 221. 062 221. 222 221. 366 221. 088 031. 041 NOTES:

/1-t1ultiple part questions on the MNP-2 docket overlap a number of questions on other dockets'

ENCLOSURE 3 DEFICIENCIES IN THE INSTRUMENTATION AND CONTROL SYSTEMS PORTION OF THE WNP-2 OL APPLICATION Type of

~0eficienc 1

~

Repetitive design deficiencies Round 1

uestion and Comment 031.083 Environmental qualification of Class 1E.equipment 031.086 - Redundancy for standby liquid control system 031.090 - qualification of the RPS motor-generator sets 031.097 - Electrical independence between Class lE and nonsafety-related equipment 2.

Failure to meet industry 031.084 - Compliance with IEEE Std. 279-1971 standards or follow staff 031.094 - Conformance with Regulatory Guide 1.47 guidance.

(The alternative 031.095 - Compliance with IEEE Std. 279-1971 is to provide justification.) 031.098 Compliance with IEEE Std. 279-1971 031. 100 Compliance with IEEE Std. 279-1971 3.

Failure to indicate compliance with the General Design Criteria 031.082 - Compliance with GDC 1, 2, 3, 4, 13, 14, 16, 19, 23 and 55 4.

Failure to respond to earlier questions 031.083(a)

- Equipment qualification per IEEE Std. 323-1971 031.099 - Flooding of Class lE equipment due to internal floods

ENCLOSURE 3 (Continued)

Type of

~Deficienc 5.

Conflicting or confusing statements Round 1

uestion and Comment 031.080 - For example, items (k), (p) and (9) 031.085 Categorization of Class lE equipment as both required and not required for safety 031.088 - Isolation of the reactor water cleanup system 031. 108 - Reactor recirculation system 6.

Vague statements 031. 101 - Reference to industrial applications without supporting basis.

Pl

'P