ML17272A379

Forwards First Round Request for Addl Info Re Facility OL Application from Matls Engineering,Reactor Sys & QA Branches

ML17272A379
Person / Time
Site:	Columbia
Issue date:	03/07/1979
From:	Varga S Office of Nuclear Reactor Regulation
To:	Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
NUDOCS 7904210047
Download: ML17272A379 (38)

Text

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Docket No:

50-397 Distribution E. Williams J.,Knight Local,PDR S.

Varga R. Tedesco Docket File D. Lynch'.

DeYoung Lt<R,II4 File M. Service V. Moore R.

Boyd R. Mattson R. Vollmer D. Vassallo D.

Ross M. Ernst March 7, 1979 R. Denise ELD IE (3)

Mr. Neil 0. Strand Washington Public Power Supply System 300 George Washington Nay P. 0. Box 968

Richland, Washington 99352

Dear her. Strand:

bcc:

J.

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

SUBJECT:

FIRST ROUND QUESTIONS ON THE NNP-2 OL APPLICATION hlTEB, RSB, AND QAB In our review of your application for an operating license for the NNP-2 facility, we have identified a needd for additiona1 information which ~ce require to complete our review.

The specific requmsts are contained in the enclosure to this letter and are the fifth set of our round one questions; additional requests related to other portions of %he NNP-2 facility will be sent during this month.

In order to maintain our present tentative

schedule, we need a completely adequate response to all questions in the enclosure by May 28, 1979.

The attached set of round one questions represent the review effort of the Materials Engineering Branch, the Reactor Systems Branch and the Quality Assurance Branch.

llthere we hive been able Ao formulate statements of staff positions (RSP),

we have included these in our questions.

Please contact us if you require any discussion or clarification of the enclosed requests.

Sincerely,

'7804&.cc (7

Enclosure:

Request for

~itional Inform S even A. Varga, Ch'ief ight Water Reactors Branch No.

4 Division of Project Management Ap ct;:

See next page OFF'~

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UNITEDSTATES NUCLEAR REGULATORY COMMISSiON WASHINGTON, D. C. 20555 Docket No:

50-397 March 7, 1979 Mr. Neil O. Strand washington Public Power Supply System 300 George Washington stay P.

O.

Box 968 Richland; Washington 993S2

SUBJECT:

FIRST ROUND QUESTIONS ON THE 9INP-2 OL APPLICATION MTEB, RSB, AND QAB In our review of your application for an operating license for the NNP-2 facility, we have identified a need for additional information which we require to complete our review.

The specific requests are contained in the enclosure to this letter and are the fifth set of our round one questions; additional requests related to other portions of the l(NP-2 facility will be sent during this month.

In order to maintain our present tentative

schedule, we need a completely adequate response to all questions in the enclosure by May 28, 1979; The attached set of round one questions represent the review effort of the Materials Engineering

Branch, the Reactor Systems

'ranch and the Quality Assurance Branch.

1<here we have been able to formulate statements of staff positions l'RSP),

we have included these in 'our questions.

Please contact us if you require any discussion or clarification of the enclosed requests.

S'erely

Enclosure:

Request for Additional Information Ci.

ven Va ga, kg Light Water React s Branch No.

4 Division of Project Management cc:

See next page

f S

Washington Public Power Supply System ccs:

Joseph B. Knotts, Jr.,

Esq.

Debevoise 8 Liberman 700 Shoreham Building 806 Fifteenth Street, N.

W.

Washington, D. C.

20005 Richard Q. Quigley, Esq.

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

Box 968

Richland, Washington 99352 Nepom 5 Rose Suite 101 Kellogg Building 1935 'S.

E. Washington Milwaukie, Oregon 97222 Ms. Susan M. Garrett 7325 S.

E. Steele Street

Portland, Oregon 94206 Mr. Greg Darby 807 So. Fourth Avenue

Pasco, Washington 99301 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Mr. 0. K. Earle Licensing Engineer P. 0.

Box 968

Richland, Washington 99352

EiNCLOSURE STATEMENT OF REGULATORY STAFF POSITIONS AilD RE UEST FOR ADDITIONAL INFORMATION NPPSS NUCLEAR PLANT NO.

2 DOCKET NO.

50-397

MATERIALS ENGINEERING BRANCH Materials Inte rit Section We require that your inspection program for Class 1,

2 and 3

components be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, paragraph (g) published in the February 12,

1976, issue of the FEDERAL REGISTER.

Accordingly, submit the following information:

a 0 A preservice inspection plan which is consistent with the required edition of the ASME Code.

This inspection plan should include any exceptions you propose to the Code requirements.

b.

An inservice inspection plan submitted within six months of the anticipated date for commercial operation.

This preservice inspection plan will be reviewed to support the safety evaluation report finding regarding your compliance with preservice and inservice inspection requirements.

Our determination of your compliance will be based on:

c.

That edition of Section XI of the ASME Code referenced in your PSAR or later editions of Section XI referenced in the FEDERAL REGISTER that you may elect to apply.

d.

All augmented examinations estabished by the Commission when added assurance of structural reliability was deemed necessary.

Examples of augmented examination requirements can be found in the NRC positions on:

(1) high energy fluid systems ln Section 3.2 of the Standard Review Plan (SRP),

NUREG-75/087; (2) turbine disk integrity in Section 10.2.3 of the SRP; and (3) the feedwater inlet nozzle inner radii.

Your response to this item should define the applicable edition(s) and subsections of Section XI of the ASME Code.

If any of the examination requirements of the particular edition of Section XI you referenced in the PSAR cannot be met, a request for relief must be submitted, including complete technical justification to support your request.

Detailed guidelines for the preparation and content of the inspection programs to be submitted for staff review and for relief requests are attached as Appendix B to Section 121.0 of our review questions.

121 "7

APPENDIX 8 TO SECTION 121.0 GUIDANCE FOR PREPARING PRESERVICE AND INSERVICE INSPECTION PROGRAMS AND RELIEF RE VEST PURSUANT TO 10 CFR 50.55a A.

Descri tion of the Preservice/Inserv'ice Ins ection Pro ram This program should cover the requirements set forth in Section 50.55a(g) of 10 CFR Part 50 and the ASME Boiler and Pressure Vessel

Code,Section XI, Subsections IWA, IWB, IWC and IWD.

The guidance provided in this enclosure is intended to illustrate the type and extent of information that should be provided for NRC review.

It also describes the information necessary for "request for relief" of items that cannot be fully inspected to the requirements of Section XI of the ASME Code.

By utilizing these guidelines, licensees can significantly reduce the need for requests for additional information from the NRC staff.

B.

Contents of the Submittal The information listed below should be included in the submittal:

1.

For each facility, include the applicable date of the ASME Code and the appropriate addendum date.

2.

The period and interval for which this program is applicable.

3.

Provide the proposed codes and addenda to be used for repairs, modifications, additions or alterations to the facility which might be implemented during this inspection period.

Indicate the examinations that you have exempted under the rules of Section XI of the ASME Code.

A reference to the applicable paragraph of the code that grants the exemption is satisfactory.

The inspection requirements for exempt components should be stated (e.g., visual inspection during a pressure test).

5.

Identify the inspection and pressure testing requirements of the applicable portion of Section XI that are deemed impractical because of the limitation of design, geometry or materials of construction of the components.

Provide the information requested in the following section of this appendix for the inspections and pressure tests identified in Item 4 above.

121-8

C.

Re uest for Relief from Certain Ins ection and Testin Re uirements It has been the staff's experience that many requests for relief from testing requirements submitted by licensees have not been supported by adequate descriptive and detailed technical information.

This detailed information is necessary to:

(1) document the impracticality of the ASME Code requirements within the limitations of design, geometry and materials of construction of components; and (2) determine whether the use of alternatives will provide an acceptable level of quality and safety.

Relief requests submitted with a justification such as "impractical,"

"inaccessible,"

or any other categorical

basis, require additional information to permit the staff to make an evaluation of that relief request.

The objective of the guidance provided in this section is to illustrate the extent of the information that is required by the NRC staff to make a proper evaluation and to adequately document the basis for granting the relief in the staff's Safety Evaluation Report.

The NRC staff believes subsequent requests for additional information and delays in completing the review can be considerably reduced if this information is provided initially in the licensee's submittal.

For each relief request submitted, the following information should be included:

l.

An identification of the component(s) and/or the examination requirement for which relief is requested.

2.

The number of items associated with the requested relief.

3.

The ASME Code class.

4.

An identification of the specific ASME Code requirement that has been determined to be impractical.

5.

The information to support the determination that the require-ment is impractical; i.e., state and explain the basis for requesting relief.

6.

An identification of the alternative examinations that are proposed:

(a) in lieu of the requirements of Section NI; or (b) to supplement examinations performed partially in compli-ance with the requirements of Section NI.

7.

A description and justification of any changes expected in the overall level of plant safety by performing the proposed alternative examinations in lieu of the examination required by Section XI. If it is not possible to perform alternate 0

121-9

examinations, discuss the impact on the overall level of plant quality and safety.

For inservice inspection, provide the following additional information regarding the inspection frequency:

8.

State when the request for relief would apply during the inspection period or interval (i.e., whether the request is to defer an examination).

9.

State when the proposed alternative examinations will be implemented and performed.

10.

State the time period for which the requested relief is needed.

Technical justification or data must be submitted to support the relief request.

Opinions without substantiation that a change will not affect the quality level are unsatisfactory.

If the relief is requested for inaccessibility, a detailed description or drawing which depicts the inaccessibility must accompany the request.

A relief request is not required for tests prescribed in Section XI that do not apply to your facility.

A statement of "N/A" (not applicable) or "None" will suffice.

0.

Re uest for Relief for Radiation Considerations Exposures of test personnel to radiation to accomplish the examina-tions prescribed in Section XI of the ASt1E Code can be an important factor in determining whether, or under what conditions, an examina-tion must be performed.

A request for relief must be submitted by the licensee in the manner described above for inaccessibility and must be subsequently approved be the NRC staff.

We recognize that some of the radiation considerations wi 11 only be known at the time of the test.

However, the licensee generally is

aware, from experience at operating facilities, of those areas where relief will be necessary and should submit as a minimum, the following information with the request for relief:

1.

The total estimated man-rem exposure involved in the examination.

2.

The radiation levels at the test area.

3.

Flushing or shielding capabilities which might r educe radiation levels.

4.

A proposal for alternate inspection techniques.

5.

A discusson of the considerations involved in remote inspections.

0 121 "10

6.

Similiar welds in redundant systems or similar welds in the same system which can be inspected.

7.

The results of preservice inspection and any inservice results for the welds for which the relief is being requested.

'I 8.

A discussion of the consequences if the weld which was not

examined, did fail.

210. 0 REACTOR SYSTEMS BRANCH 211.02 (5. 2. 5)

(7. 6. 1)

(9. 3. 3) 211. 03 (5. 2. 5) 211.04 (5.2.5)

(7. 6. 2) 211. 05 (5.2.5)

Discuss the sump geometry, the accuracy of the leakage flow rate measurements, the monitoring interval and any other informa-tion required to demonstrate a sensitivity of one gallon per minute (gpm) per hour for the floor drain sump level monitoring systems.

The design of the WNP-2 facility routes drainage of both "hot" and "cold" reactor coolant leakage into the drywell equipment drain sump.

However, relatively hot sources of leakage water (e.g.,

the reactor vessel head flange, the vent drain and valve packings) may flash into steam which then must. be condensed before it can be drai.ned into the sump.

Accordingly, indicate what assurance there is that this steam from relatively hot sources will be condensed so that monitoring of this leakage may be performed in the drywell sump to detect these "hot" leaks.

Since leakage from those sources which are relatively cool is drained into the floor drain system, this system should be tested periodically for blocked lines.

Accordingly, discuss the surveillance program you propose for the WNP-2 facility to detect blockage of the floor drain system.

In conformance with the staff's positions in Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"

May 1973, you state in Section 7.6.2.4 of the FSAR that the radioactivity monitoring channels are qualified for operation following a Safe Shutdown Earthquake (SSE).

Confirm that all of the other leakage detection methods and/or systems will function properly following an Operating Basis Earthquake (OBE).

These other leakage detection systems include the drywell equipment sump and the floor drain sump, the sump

coolers, and the associated instrumentation and piping.

Provide a list of the normal and the maximum anticipated leakage rates through the reactor coolant pressure boundary (RCPB),

including the concentrations of radioactivity in this leakage flow, from both'identified and unidentified sources (e.g.,

the control rod drive flanges and the vent cooler drains) which are routed into the drain sumps.

211.06 (5.2.5)

(7.6.2)

(12.3.4)

Provide a detailed discussion of the sensitivity and response times, of the containment airborne radiation monitoring systems for a number of containment background activity levels.

The background activity levels which should be considered are those levels in the containment that would result from leakage through the RCPB assuming:

(1) relatively clean water in the 211-2

211.07 (7.6.2) reactor coolant system at the initial operation or the WNP-2 facility at power; and (2) the maximum level of activity in the reactor coolant permitted by the WNP-2 Technical Specifica-tions.

In responding to this item, assume both the normal and the maximum leakage rates identified in your response to Item 211.05 above.

Indicate your assumptions in estimating the response times of the containment airborne radiation monitoring systems (e.g.,

the preset alarm level for higher background leakage and the plateout factor).

Expand your discussion in Section 7.6.2.4 of the FSAR regarding the operability verification and the calibration of the leakage detection system which will be accomplished by comparing the results of diverse monitoring methods.

For example, if a radioactivity monitoring system is checked against the sump level and the flow monitoring system, indicate how the latter system is determined to have acceptable accuracy.

Confirm that calibration and operability tests will:

(1) be performed periodically during plant operation; and (2) be in compliance with the requirements of IEEE Standard 279-1971.

211. 08 (5. 2. 2)

(7.6.i) 211. 09 (5.2.5)

(7.6.2)

In Section

7. 6. l. 4 of the FSAR, you state that major components within the drywell which are sources of leakage by nature of their design (e.g.,

the sump seals, the valve stem packing, and the equipment warming drains),

are enclosed and the leakage is piped to an equipment drain sump and identified there.

Indicate what you mean by the term "sump seals" (i.e., did you intend to state pump seals).

Discuss what monitors are available

.to the operator to permit.him to identify the source of leakage; i.e., whether the leakage is from the "sump" (or pump) seals or the equipment warming drains or from any other component leakage sources which drain into the drywell equipment drain tank.

Indicate whether there are any sumps within the drywe'll which must be filled before the sump drain flow is routed to the equipment drain tank.

In conformance with the staff's positions in Regulatory Guide 1.45, you state that provisions will be made to monitor systems connected to the reactor coolant pressure boundary for signs of intersystem leakage.

Provide a detailed discussion of these provisions, including an identification of all potential intersystem leakage paths; e.g.,

the leakage from the primary coolant system to the residual heat removal (RHR) system and the emergency core cooling system (ECCS) injection line.

Identify the instrumentation used in each path which will provide positive indication of intersystem leakage in the affected system.

211-3

211. 10 (5.2.5) 211. 11 (5. 5. 2)

(App. C. 2)

Operating experience at some boiling water reactors (BWRs) has shown that the high pressure coolant injection system (MPCI) and the reactor core isolation cooling system (RCIC) have been rendered inoperable due to inadvertent leak detection isolations which were caused by a high differential temperature signal from the equipment room area.

These isolations occurred when there was a relatively sharp drop in the outside temperature.

In Section 5.4.6. l.. 1.1 and Table 5.2-9 of the FSAR, you indicate that the WNP-2 facility also has this type of isolation for the RCIC system and the steam side of the RHR system.

Provide a discussion of the modifications that have been, or will be, made to prevent inadvertent isolations of this type which affect the availability and reliability of the RCIC and the RHR systems.

Additionally, indicate the trip settings you propose for isolation of the WNP-2 RHR and RCIC systems due to high area temperature in terms of degrees Fahrenheit above the ambient temperature.

Discuss the method you propose to avoid this problem.

Show that the differential temperature setting could not be set too low, thereby causing an inadvertent isolation when the RCIC and the RHR systems are needed.

You deviate from the staff's positions in Regulatory Guide 1.29, Revision 3, "Seismic Design Classification,"

September

1978, by not designing the component cooling water portions of the reactor recirculation pumps to seismic Category I criteria.

The basis you state in the FSAR for this deviation is that these pumps do not perform a safety function.

Provide additional justification for this position and show that a loss of component cooling water to the recirculation pumps would not lead to unacceptable consequences.

211. 12 (4.i)

(5.4.6)

(5.4.7)

Describe the provisions incorporated into the WNP-2 facility to protect the RCIC and the RHR systems from cold weather and from dust storms and to assure satisfactory operational performance under any adverse meteorological conditions.

In this discussion, include consideration of the standby liquid control system and the control rod drive (CRD) hydraulic system and any other sources of water for these systems (e.g.,

the condensate storage tank and the standby service water).

211. 13 (3. 5. 1)

In evaluating the potential for missiles due to failures of pressurized components, you state that thermowells and sample probes have been assessed against criteria discussed in Section 3.5. l. 1.2 of the FSAR.

Indicate which specific criterion and basis has been considered in determining that the thermowells and sample probes are not credible missiles.

Provide justifica-tion for omitting other pressurized components such as blank 211-4

flange assemblies and pressurized vessels or bottles (e.g.,

the safety/relief valve air accumulators and the nitrogen accumulator tanks) from your assessment of potential missiles.

211. 14 (3. 5. 1)

Discuss the potential for missiles inside the containment due to falling objects (e.g., electrical hoists or any unrestrained equipment) for the following events:

(1) routine maintenance; (2) reactor operation; and (3) a postulated loss-of-coolant accident (LOCA).

211. 15 (3.5.1)

With regard to missiles generated by a p5'stulated failure of a rotating component, show by analysis that the impeller fragments resulting from an overspeed condition in the recirculation pump during a postulated LOCA, will not penetrate the pump case.

Provide a study demonstrating that the probability for significant damage to. safety-related components or systems inside containment resulting from impeller missiles which might be ejected out the open end of the broken pipe, is acceptably low. If you reference a similar study on another

docket, demonstrate the appropriateness of referencing this study for the WNP-2 facility.

211. 16 (3.5.1)

Based on our review of the design integrity of nuclear power plant piping systems, we have noted several failures of safety valve headers which caused the valves to become missiles.

(Refer to NUREG-0307, "Review and Assessment of Research Relevant to Design Aspects of Nuclear Power Plant Piping Systems,"

July 1977).

Since you address only the credibility of valve bonnets and stems as potential missiles, provide justification in Section 3.5. 1.2 of the FSAR for concluding that the safety valve header and valve cannot be considered to be credible missiles.

Your statement in Section 3.5. 1. 1.3. 1 of the FSAR that bonnet ejection is highly improbable and that bonnets are not considered to be credible missiles, is not supported.

Show that should a large valve component become a

missile, containment penetration would not occur.

Discuss the provisions incorporated into the WNP-2 facility (e. g., equipment separation and redundancy) to preclude damage to the systems necessary to achieve and maintain a safe plant shutdown.

211. 17 (4.6.1)

Provide information demonstrating that the loss of the operating CRD pump at low reactor pressure (i.e., less than 500 psig) will not result in depressurization of the accumulator, thereby causing a loss of the reactor's scram capability.

If the accumulator check valves were to leak following loss of the operating CRD pump, provide an estimate of the time interval before the reactor scram capability becomes marginal.

In responding to this question, provide your basis for this 211-5

estimate.

Describe a test program or procedure which would provide assurance that operation of these check valves is acceptable over the lifetime of the WNP-2 facility.

211. 18 Experience at some operating BWR's indicates that failures can (4.6. 1)

, occur in the collet fingers of the CRD mechanism.

In order to resolve this problem, some BWR facilities under construction have installed a revised collet retainer design.

However, you do not address this particular problem in your FSAR nor do you discuss its resolution., Accordingly, confirm whether the revised collet retainer design will be incorporated into the CRD mechanisms of the WNP"2 facility.

Revise Table 1.3-8 of the FSAR as required.

211. 19 (4.6.l.)

We note in the third item of Table l. 3-8 of the FSAR that you intend to cut and cap the CRD return line as a resolution of the stress corrosion problem in this line.

Discuss the impact of this modification on the plant.

In particular, provide additional information addressing, but not limited to, the following items:

a ~

Compare the reactor vessel makeup capability when either one or two CRD pumps are operating, before and after the proposed modification.

Additionally, provide a commitment to perform preoperational testing to verify the modified flow capability.

b.

Provide a commitment to perform preoperational testing to verify the individual performance of the modified CRD

. components and those portions of the CRD system that are potentially affected by the cut and capped CRD return line (e.g.,

the equalizing valves, filters, scram. times and the settling function).

C.

If you choose to add new equalizing valves, discuss the potential effect on drive speeds throughout the lifetime of the WNP-2 facility.

In particular, evaluate whether the CRD system can be adversely affected by a voiding of the drive exhaust header after a postulated single failure.

d.

Evaluate the effect, throughout the lifetime of the WNP-2 facility, of the added flow through such components as the drive exhaust header and the stabilizing lines.

In particular, discuss the increased potential for corrosion products from the carbon steel piping to be deposited as additional foreign matter in the drives.

211-6

e.

Discuss the potential for, an'd effect on, flow reversal through the directional control solenoid valve over the plant lifetime.

f.

Discuss the anticipated effect of your modifications to the CRD system on the DP settling function across the control rod drives to ensure latching of the rods after withdrawal.

211.20 (5. 4. 1)

(15. 0)

Appendices G and H of the LaSalle and Zimmer FSAR's, respectively, provide information on the recirculation flow control system.

State whether this information is applicable to the WNP-2 facility. If applicable, it should be referenced.

Otherwise, provide comparable information regarding this system, in your application.

211. 21 (5. 4. 1)

With respect to the recirculation flow control system:

a.

Provide justication for the 8 degree Fahrenheit subcooling limitation in operating the recirculation pump.

b.

In Section 5.4. 1.3 of the FSAR, you state that if the subcooling falls below 8 degrees Fahrenheit, the 60 Hz power supply is tripped to the 15 Hz power source to prevent cavitation of the recirculation

pump, the jet

pumps, and/or the flow control valve.

This temperature limitation on subcooling appears to initiate a two-pump trip transient.

Indicate whether the pump coastdown rate resulting from the above condition is more severe than the one assumed in the transient analysis in Chapter 15 of the FSAR.

If so, reanalyze the pump trip transient using the more severe pump coastdown rate.

Describe the consequences of a sudden increase in the recirculation pump speed which might occur, for example, due to an increase in the frequency of the power supply.

211. 22 (5.4. 1) 211. 23 (3 5)

(4. 6. 1)

Indicate the units associated with the value of CV shown in Figures 5.4-4a and 5.4-4b of the FSAR.

Provide assurance that the essential portions of the CRD system (i.e., the 1-inch supply and return piping located inside the containment) shown in Figures 3.5-20, 3.5-22, and 3.5-27 of the FSAR, are protected from the effects of high or moderate energy line breaks.

In responding to this item, consider postulated breaks such as ones in the high pressure core spray (HPCS) system, the feedwater injection system and the reactor coolant pressure boundary.

Our concern is whether pipe whip and/or jet impingement forces resulting from these postulated 211-7

breaks can impair the capability'of the CRD system to scram.

Additionally, provide an assessment of the damage which could occur to the cluster of CRD return and supply lines, including the effect on the scram capability, due to a postulated rupture of a single CRD supply or return line.

211. 24 (5.4. 7)

It is our position for all light-water reactors that the'HR system shall be capable of bringing the reactor to a cold shutdown condition using only safety-grade systems.

Confirm that this requirement is satisfied for the WNP-2 facility.

In responding to this request, include a consideration of the capability of the air supply system which is used to operate the RCIC steam and condensate control valves located at the RHR heat exchanger, when the RHR system is in the steam condensing mode.

211. 25 (5.4.7)

It is also our position for all light-water reactors that the RHR system shall be capable of bringing the reactor to a cold shutdown condition with only onsite or offsite power available, assuming the most limiting single failure.

In this regard, while we note that Figure 15.2-10 of the FSAR shows a number of available success paths to achieve a cold shutdown condition, vessel depressurization using the RHR system in the steam condensing mode is not shown.

(This latter mode is one of the success paths when offsite power is not available.)

Either correct this figure or justify this omission.

If vessel depressurization were to be achieved by manual actuation of the relief valve, indicate how many valves would have to be actuated.

Describe your plans for testing the alternate modes to achieve shutdown cooling.

Demonstrate that adequate passage of water through the safety/relief valves can be achieved and maintained when the alternate method is in use.

Indicate the quantity of air supplied, its source, and the time interval before the air is exhausted.

211. 26 (5.4. 7) 211. 27 (5.4.7)

In the shutdown cooling mode, the flush water valves are opened'nd closed outside the control room.

Identify in Section 5.4.7.2 of the FSAR, the local flush water valves which are operated and the source of this flush water.

Discuss the consequences if the operator were to omit this procedure and/or forget to close a local flush water valve and continue shutdown operations.

Include a discussion of the available interlocks in your response.

In Section 5.4.7. 1.3 of the FSAR, you indicate the specific RHR relief valves and the RHR design pressures used as the basis for providing relief capacity.

Expand your discussion by indicating the relief valve capacity, the nominal set points, 211" 8

211.28 (5.4. 7) the set point tolerance, and the ASME class designation of these valves and lines.

In addition, discuss the vulnerability of the RHR system to malfunctions which could result in over-pressurization of low pressure piping.

Support your evaluation by providing an outline of all operating procedures required to bring the plant to a cold shutdown condition from hot standby and the procedures for plant startup from cold shutdown.

Discuss the need and provide the design basis for incorporating a pressure interlock to prevent the connection of the RHR discharge piping to the primary system whenever the actual pressure difference across the discharge valve is greater than the design value for this pressure differential.

Identify the affected valves.

211.29 (5.4. 7)

Provide more detailed information in Section 5.4.7.1 of the FSAR regarding the actuation of the automatic minimum flow valves which are used to protect the RHR pumps from damage if these pumps were to be operated when the discharge valve is closed.

For example, state the flow rates which would initiate a signal to open and close the minimum flow valves.

Indicate whether the control system satisfies the requirements of IEEE Standard 279-1971.

211. 30 (5.4. 7)

In Figure 5.4-15 of the FSAR, you present the RHR pump charac-teristic curves.

However, two sets of curves are

shown, one for the maximum diameter piping and the other for the minimum diameter piping.

Indicate which of the head versus flow rate characteristics was used in the performance evaluation of the ECCS and the RHR system.

211. 31 (5.4.7)

In Table 5.4-3 of the FSAR, you indicate that the RHR isola-tion valves HOF008 and MOF009 are closed upon generation of a signal indicating reactor low water level.

It appears that you have mislabelled these valves in this table as "recircula-tion line suction" rather than as "RHR isolation."

Indicate whether this valve isolation signal is based on the same signal as the RHR pump actuation in the low pressure coolant injection system (LPCI) mode (i.e.,

a water level which is 1.0 foot above the active core).

If not, indicate the water level in the reactor pressure vessel at which the isolation signal is generated, thereby isolating the RHR suction valves.

Show that cooling of the reactor core can be maintained assuming a

pipe break outside the containment.

Assuming a pipe break outside containment in the RHR system when the plant is in a shutdown cooling mode, provide the following additional informa-tion:

211-9

a.

Identify the systems available for maintaining core cooling.

b.

Indicate the maximum discharge rate resulting from the postulated break and the time interval available for recovery based on the discharge rate and its effect on core cooling.

c.

~

Identify the alarms available to alert the operator in the event of such a break and show that sufficient time is available for operator action to prevent damage to safety-related systems.

d.

Indicate what recovery procedures are available.

e.

Following a postulated break in an moderate energy line, the single failure criterion should be applied in the manner discussed in Section 3.6.

1 of the Standard Review Plan (SRP)

NUREG-75/087, and in Branch Technical Position APCSB 3-1, "Protection Against Postulated Piping Failures In Fluid Systems Outside Containment",

November 24, 1975.

211. 32 (5.4.7)

Discuss the system design provisions incorporated into the facility to prevent damage to the RHR pumps in the LPCI mode during pump runout conditions in the ECCS operating mode and in the test mode.

We note that Figures 5.4-13a, 5.4-13b and 5.4-14a of the FSAR indicate that a metering orifice is installed in the discharge lines.

Indicate whether this metering orifice can perform the same function as a restricting orifice. If not, it is our position that the discharge lines of the RHR pumps should incorporate a restricting orifice.

211. 33 (5.4. 7)

Provide a more detailed description, including the location, of the RHR pump suction strainer which is inside the suppression pool.

Indicate the pipe bends and the minimum height of the suppression pool water level above this strainer.

Show that the required net positive suction head (NPSH) at the center line of the RKR pump will be available when the pump is operating at its design conditions and at the most limiting operating conditions.

Discuss the size of particles that could pass through the strainer into the RHR pump passages.

Indicate the amount of material blockage it would take to significantly reduce the RHR pump suction flow from the suppression pool following a postulated LOCA.

211.34 Provide the process data (i.e., flow, temperature, and pressure) 211-10

(5.4. 7) 211. 35 (5.4. 7) 211. 36 (5. 4. 7) 211. 37 (5. 4. 7) for all modes of operation of the RHR system which you reference in Figure 5.4-14a of the FSAR; i.e.,

MPL Item No.

E12-1020.

Identify the pressure interlock set points of the RHR isolation valves F008 and F009 which are set to:

(l) prevent inadvertent opening to the low pressure suction piping; and (2) initiate valve closure when the reactor pressure is increasing.

Confirm that all valves performing an isolation function between the high pressure and low pressure boundary in the RHR system (e.g.,

check valves and motor-operated valves) meet the leak testing and inspection requirements of Section XI of the ASME code for Category A valves.

In this regard, it is our position that a combination of two or more check or motor-operated valves in series should have design provisions which permit individual leak testing of any two valves.

In Note 12 of Figure 5.14-13a of the FSAR, you state that:

"Between valves MOF008 and MOF009 consideration should be given to thermal expansion of the contained water."

Provide a commitment to incorporate a method for pressure relief between these two RHR isolation valves.

Alternatively, show by analysis that piping integrity would be maintained in the event that a

LOCA or steam line break occurred and the water trapped between these two valves, thermally expanded.

211.38 (5. 4. 7)

(14. 2. 12) 211.39 (5. 4. 7) 211. 40 (5.4. 7)

Provide the test acceptance criteria discussed in Section 14.2. 12. 1.7 of the FSAR regarding preoperational testing of the RHR system.

Operation of the RHR system in the steam condensing mode involves partial draining of one or both RHR heat exchangers and introduction of reactor steam into lines and heat exchangers which are initially cold.

Oescribe the methods (e.g.,

valve operation or air introduction) and the provisions you propose to prevent the occurrence of water hammer during initiation of operation in this mode and in the change to the pool cooling mode.

Indicate whether the jockey pump system shown in Figure 5.4-13a of the FSAR, can fill the lines to the injection valve in the core spray lines and the RHR lines (i.e., valves F016 and F042, respectively) when the RHR is in the steam condensing mode using one or both heat exchangers.

If not, indicate what procedures you propose to prevent water hammer following star tup of the core spray or RHR pumps.

Those pressure relief valves and lines which are designed to prevent overpressurization of the RHR system, are routed outside the containment before being returned to th'e suppres-

sion pool.

Discuss the design provisions incorporated into the WNP-2 facility to minimize the potential for water hammer in these lines.

State whether these relief lines are capable of withstanding both seismic and dynamic blowdown loads without suffering a loss of structural integrity.

211. 41 (5. 4. 7)

Discuss the procedures to be used in the WNP-2 facility which will minimize the potential for exceeding the allowable cooldown rate (i.e.,

a cooldown rate greater than 100 degrees Fahrenheit/hour) of the RHR system and the reactor coolant system when placing the WNP-2 facility in a shutdown cooling mode following normal shutdown or following an emergency shutdown.

211. 42 (5 ~ 4 ~ 7)

Discuss the reliability of the RHR pumps for long-term operation.

It is our position that long-term term reliability of these pumps should be demonstrated either by operational experience or by testing.

If you cite previous operational experience as the basis for qualifying these pumps in your response to this

question, identify any design differences in the pumps and indicate the operating conditions of the pump service life which is cited.

211. 43 (5. 4. 6) 211.44 (5. 4. 6)

Provide an RCIC pump performance curve that shows flow rate versus reactor vessel pressure.

Identify the most limiting operating condition for the RCIC pump and identify the NPSH margin under this condition.

When the steam isolation valves are temporarily closed for maintenance, it appears that it is possible for some steam condensate to remain in the lines leading to the RCIC steam turbine.

Discuss whether the amount of water condensed from steam can cause sufficient damage to the RCIC turbine to render the RCIC system incapable of delivering water to the reactor vessel as required.

Describe the design modifica-tions, if any, you propose to prevent water hammer occurring at the RCIC turbine exhaust.

211'5 (5.4.6)

An isolation signal will close a number of valves (i.e.,

F063 and F008) in the RCIC system, located inside and outside containment, branched off the main steam line.

However, the process and instrumentation drawing (P8 ID) for the RCIC (i.e.,

Figure 5.4-9a),

shows that these valves are keylocked open.

Explain this apparent discrepancy.

Additionally, evaluate the consequences of a postulated pipe, break downstream of the first or second isolation valve for steam flow rates which are greater than 300 percent of the steady-state steam flow.

Provide justification for the selection of this 300 percent 1 imit.

211. 46 (5.4. 6) 211.47 (15. 2. 9)

The acceptance criteria contained in Section 5.4.6 of the SRP state that "As a system which must respond to certain abnormal

events, the RCIC system must be designed to seismic Category I

standards, as defined in Regulatory Guide 1.29."

However, the condensate storage tank which is the normal supply of water for the RCIC, is not designed to seismic Category I criteria.

While the suppression pool provides an alternate source of water from a seismic Category I structure, the switchover to this alternate source in the WNP-2 facility requires operator action.

Any one of the following alternatives would be an acceptable approach for meeting the acceptance criterion cited above:

(1) a seismic Category I supply; (2) an automatic safety-grade switchover to a seismic Category I supply; or (3) a manual safety-grade switchover to a seismic Category I supply, if appropriately justified. It appears that you are proposi ng to use the third option.

If so, provide justification for the time required for the operator to perform a manual switchover.

If the third alternative is not proposed, identify and discuss which approach will be used in the WNP-2 facility.

Provide the suppression pool temperature and the reactor vessel temperature and pressure as a function of time for the alternate shutdown cooling modes (i.e., activity Cl and C2) described in Figure 15.2-11 of the

FSAR, assuming a failure of the normal RHR shutdown cooling mode.

Provide an estimate of the time required to achieve a cold shutdown condition for these alter-nate cooling paths.

Identify the initial pool and service water temperatures assumed in this analysis.

211.48 (5.2.2)

(5.4.7)

(6.3)

In the FSAR, you state that the volume of air stored in the pneumatic accumulator for each safety/relief valve is sufficient for one actuation of the power-operated relief valves.

You also state that the accumulator volume is sufficient for two actuations of the automatic depressurization system (ADS) valves.

However, a "noninterruptible" safety-grade source of air to actuate the ADS valves is required to terminate certain postulated transients and accidents without loss of the ADS function.

Demonstrate that an adequate supply of air will exist to operate the ADS valves for the following postulated accident conditions:

a.

The alternate method of achieving and maintaining a cold shutdown following a loss of offsite'power, concurrent with the worst single failure in the RHR system.

b.

A small break LOCA concurrent with the failure of the high pressure core spray which would then require the ADS valves to be actuated to:

(1) depressurize the reactor

vessel; and (2) maintain long-term cooling.

In your

response, also discuss your proposed procedures to replenish the coolant inventory for this particular postulated accident.

A small steam line break disabling the RCIC concurrent with a single failure of the HPCS that would require actuation of the ADS to depressurize the reactor vessel.

In your response, discuss the supply of air required to actuate the ADS valves to provide long-term cooling of the reactor core.

Additionally, for this specific postu-lated accident condition, indicate whether the reactor vessel inventory would be maintained above the shutoff head of the low pressure cooling system when the decay heat of the reactor core repressurizes the vessel.

420. 0 UALITY ASSURANCE BRANCH 422.0 Conduct of 0 erations 422.

1 422. 2 422. 3 (13. 1. 1) 422. 4 (13. 1. 1) 422. 5 (13. 1. 1)

You state that the Plant Superintendent is responsible for fire protection activities at the plant level.

Describe any further delegation of these responsibilities for the fire protection.

program such as maintenance of fire protection

systems, testing of fire protection equipment, fire safety inspections, fire fighting procedures, and fire drills:

Describe the proposed composition of your plant fire brigade.

Indicate the number of professional persons reporting to each of the following:

(1) Manager, Test and Startup Programs; (2)

Chief, Design Engineering; (3) Project Engineer, WNP-2; and (4)

Manager, guality Assurance.

Provide the resume of the persons filling the positions of Plant Operations Division Manager and Project Engineer - WNP-2.

Provide the requirements for the specific level of experience for your Mechanics, Electricians, and Technicians shown in Figure

13. 1-5 of the FSAR.

(Refer to subsections 4.5.2 and 4.5.3 of ANSI N18.1-1971.)

422. 6 (13. 1. 3)

Provide the resumes of the persons filling the positions of Nuclear Engineer and-I8C Engineer shown in Figure 13.1-5 of the FSAR.

422. 7 (13.4. 1)

Describe the functions and responsibilities, the quorum require-

ments, the meeting frequency, the author ity, and the r ecor dkeeping provisions of your Plant Operations Committee.

422. 8 (13.4.2) 422. 9 (13.43)

Describe the membership, responsibilities, authority, and method of operation of your Safety Review Board.

Describe your audit program for operational phase activities which will satisfy the requirements contained in Section 4.5 of ANSI N18.7-1976.

(Refer to subsection 6.5.2 of the NRC Standard Technical Specifications.)

422-1

420. 0 423. 0 423. 11 RSP UALITY ASSURANCE BRANCH Initial Test Pro ram Your replies to Items 423.2, 423.6, and 423.7 do not clearly identify the amount of participation by General Electric, Burns 8

Roe, and WPPSS personnel, other than identifying them as members of the Test Working Group and the Plant Operating Committee who prepare,

conduct, and review preoperational and startup tests.

Additionally, your responses do not indicate that all personnel involved in the preparation, conduct or review ot tests will be qualified.

In your responses, you should have clearly established minimum requirements for the qualification of supervisory and review positions.

Our position in this matter is that, in general, the minimum qualification requirements for individuals who direct preoperational tests or startup tests or who review test procedures are:

The minimum qualifications of individuals, at the time the individuals are assigned, who will direct or supervise the conduct of individual preoperational tests are:

l.

A bachelor's degree in engineering or the physical sciences (or the equivalent) and one year of applicable power plant experience.

At least three months of indoctrination/training in nuclear power plant systems and component operation of a nuclear power plant should be included in the applicable experience requirement.

The experience at a nuclear power plant should be at one which is substantially similar in design to the WNP-2 facility.

2.

Alternatively, a high school

diploma, or the equivalent, and four years of power plant experience.

Credit for up to two years of this four year experience may be given for related technical training on a one-for-one time basis.

At least three months of indoctrination/training in nuclear power plant systems and component operation of a nuclear power plant should be included in the applicable experience requirement.

The experience at a nuclear power plant should be at one which is substantially similar in design to the WNP-2 facility.

Minimum qualifications of individuals, at the time the individuals are assigned, who direct or supervise the conduct of individual startup tests are:

l.

A bachelor's degree in engineering or the physical sciences (or the equivalent) and two years of appli-cable power plant experience of which at least one year shall be applicable nuclear power plant experience.

423-3

2.

Alternatively, a high school diploma or the 'equivalent and five years of applicable power plant experience of which at least two years shall be applicable nuclear power plant experience.

Credit for up to two years of non-nuclear experience may be given for related technical training on a one-for-one time basis.

C.

Minimum qualifications of individuals, at the time the specific activity is to be performed, who are assigned to groups responsible for review and approval of preoperational and startup test procedures and/or review and approval of test results are:

l.

Eight years of applicable power plant experience with a minimum of two years of applicable nuclear power plant experience.

A maximum of four years of non-nuclear experience may be fulfilled by satisfactory completion of academic training at the college level.

Several sections of the FSAR, including Sections 14.2.4.4, 14.2.5.2, and 14.2.6. 1, reference the MPPSS Test and Startup Program Manual.

Incorporate the applicable portions of this manual into the FSAR.

Indicate the approximate numbers (by job position) and the approximate schedule relative to fuel loading, for providing test personnel.

In response to Item 423.4, you modified Section 14.2.4. 1.5 of the FSAR to address the matter of significant modifications or repairs to safety-related systems.

Define the term "significant modifications and repairs" and designate the group or individuals authorized to determine the significance of a modification or repair and to determine the requirements for retesting the affected system.

Indicate how modifications and repairs which are not considered significant, are to be controlled.

Revise Section 14.2. 11 of the FSAR to ensure that test procedures will be available not less than 60 days prior to fuel loading.

In Section 14.2.7 of the FSAR, you discuss conformance of test programs with applicable Regulatory Guides.

Expand this section to discuss conformance with Regulatory Guides 1.52, 1.56, 1.68.1, 1.68.2 and 1.108.

Modify the appropriate test descriptions to reflect the staff's positions in these Regulatory Guides.

423-4

423.17 In Section 14.2.10.1.4 of the FSAR, you refer to the preopera-tibnal testing listed in Table 14.2-4.

However, Table 14.2-4 is a list of startup tests.

Correct this discrepancy.

423. 18 423. 19 Several of your prerequisites for preoperational tests include the requirement that support systems must have the capability to verify their readiness to function.

Provide a description of this readiness verification and indicate which individuals or groups are authorized to make this determination.

In our review of your preoperational test phase, we found that several systems and design features may not be scheduled for preoperational testing.

Our evaluation of your preoperational test program was based on a comparison of your proposed test program with the structures, systems and components of the WNP-2 facility that:

a.

will be relied upon for a safe shutdown and cooldown of the reactor under normal plant conditions; b.

will be relied upon for safe shutdown and cooldown of the reactor under faulted,

upset, or emergency conditions; c.

will be relied upon for establishing conformance with safety limits or limiting conditions for operation that will be included in the WNP-2 Technical Specifications; d.

are classified as engineered safety features or will be relied upon to support or assure the operation of engineered safety features within design limits; e.

are assumed to function, or for which credit is taken, in the accident analysis of the WNP-2 facility; and f.

wi 11 be utilized to process, store, control, or limit the release of radioactivity.

Accordingly, expand or modify the description of your preopera-tional test phase to address your plans relative to preoperational testing of the following:

1.

The logic, controls, valves, and components used in the condensate and feedwater heating systems.

2.

The valves in the automatic depressurization system (ADS), including demonstrations of operability using all alternate pneumatic supplies and a demonstration of the operability of the pneumatic air supply systems.

(Refer to Regulatory Guide 1.80.)

423"5

The logic of the'WNP-2 system which is used to mitigate the consequences of anticipated transients without scram (ATWS) events, including the instrumentation and controls and the hardware to mitigate the ATWS events.

The leak-tightness of the control room.

The diesel-generator air starting system.

The "keep-full" systems for the high pressure core spray (HPCS) system, the low pressure core spray (LPCS) system and the residual heat removal (RHR) system pumps.

The automatic transfer of suction from the condensate storage tank (CST) to the suppression pool for the HPCS system.

The temperature monitoring instrumentation and the heat tracing associated with the CST tank.

The manual isolation capability between the main condenser and the offgas system.

The manual operations (local-manual) of all valves or dampers that are provided with manual operators for those systems classified as engineered safety features.

Your response should indicate whether this wi 11 be done as a

part of each individual preoperational

test, as a test prerequisite, or as a construction acceptance test.

The timing test for the flow control valves of the recircu-lation system.

The leak-tightness tests for the emergency core cooling systems (ECCS).

The test firing of squib explosive devices in the traversing in-core probe (TIP) system and the standby liquid control (SLC) system.

The response time testing of engineered safety features including the initiating logic.

The heating, air conditioning and ventilating systems in the following areas:

the main control room cable spreading room/critical switchgear

area, the emergency diesel-generator building, the diesel-generator cable area corridor, the radwaste building, the reactor building emergency cooling system and critical electrical equipment area cooling system.

423-6

16.

The standby gas treatment system.

17.

The main steam line isolation valve leakage control system.

18.

The oriented spray cooling system of the ultimate heat sink.

We could not conclude from our review of the preoperational test phase and the test abstracts provided in Table 14.2 of the FSAR whether comprehensive testing is scheduled for several of the tests described in this table.

Accordingly, clarify or expand the description of the preoperational test phase to address the following:

a.

b.

Modify the individual test descriptions of the a-c and the d-c distribution systems or provide an integrated test description to verify proper load group assignments.

(Refer to Regulatory Guide 1.41.)

Provide your plans to verify that:

(1) the d-loads are consistent with the assumptions regarding the sizing of the batteries; and (2) the supplied loads remain operable at the minimum terminal voltage of the batteries which is equivalent to that measured in the initial and periodic load discharge tests.

Modify the test descriptions of the 250 Volt d-c, 125 Volt d-c, and 24 Volt d-c systems to include these testing requirements.

Provide acceptance criteria for these tests.

C.

State how operability of emergency loads using offsite power, will be demonstrated during the tests of the a-c and d-c systems.

d.

Modify the description of the primary containment leak rate test to address the progression of test pressures and the method of closure of the containment isolation valves.

Clarify whether the type B and C local leak tests will be conducted as a part of the construction testing or of the preoperational testing.

(Refer to Appendix J of 10 CFR Part 50.)

e.

Identify the testing you will perform to verify the amount of drywell floor bypass leakage.

Provide quanti-tative acceptance criteria for this test.

Provide your plans for testing the primary containment isolation system, including the response times of the containment isolation valves.

423-7

g.

Revise the test description of'he reactor protection system to provide your plans for assuring that the effects of interfacing hardware (e.g.,

snubbers and pulse dampers) located between measured variables and the input to the sensors for the reactor protection

system, do not compromise the requirements for the channel response time.

Provide acceptance criteria that reflect the effect of these interactions.

h.

Modify the description of the reactor recirculation system and control test to demonstrate the proper operation of the rate limiters of the recirculation flow control valve on the flow controllers.

Demonstrate that individual control valve stroke rates do not exceed the assumptions in your analysis of accidents in Section 15 of the FSAR.

Provide quantitative acceptance criteria for these tests.,

In our review of your proposed startup testing phase, we found that some tests may not fully conform with the staff positions described in Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Mater-Cooled Power Reactors,"

November 1973.

Describe how you will conform with the following staff positions in Appendix A to Regulatory Guide 1.68:

(1) C.2.f and C.2.i; (2) 0.2.a for the high pressure core spray system; (3) D.2 '

with respect to the operation of a bypass valve; (4) 0.2.o; (5) D.2. r with respect to the trip of two recircula-tion pumps when the plant is operating at 100 percent of rated power; and (6) D.2.v.

Modify Section 14.2 of the FSAR to specifically identify each startup test listed in Table 14.2 that you do not consider "essential" to demonstrate the operability. of structures,

systems, and components which meet any of the criteria listed bel ow.

a 0 Those that will be used for safe shutdown and cooldown of the reactor under normal plant conditions and for main-taining the reactor in a safe condition for an extended shutdown period.

b.

Those that will be used for safe shutdown and cooldown of the reactor:

(1) under transient conditions which are infrequent or moderately frequent events; (2) under postulated accident conditions; and (3) for maintaining the reactor in a safe condition for an extended shutdown period following such conditions.

C.

Those that will be used for establishing conformance with safety limits or limiting conditions for operation that will be included in the WNP-2 Technical Specifications.

423-8

d.

Those that are classified as engineered safety features or will be used to support operations, or ensure that operations of engineered safety features are within design limits.

e.

Those that are assumed to function, or for which credit is taken, in the WNP-2 accident analysis in Section 15 of the FSAR.

f.

Those that will be used to process, store, control, or limit the release of radioactive materials.

423.23 In our review of the test abstracts provided in your FSAR, we found they are not sufficiently descriptive to permit us to conclude that a comprehensive testing program is planned or that satisfactory test acceptance criteria have been established.

The individual tests abstracts should be modified as indicated below:

a ~

b.

Provide technical justification for the Level 2 acceptance criterion of + 7 percent of rated power for the calibration test of the average power range monitor (APRM).

Modify the test abstract for the reactor core isolation cooling (RCIC) system to provide for five cold, quick starts of the system.

Indicate the system conditions for a cold, quick start.

The Level 1 acceptance criteria in your FSAR refer to operating restrictions presented in Figure 14.2-3 of the FSAR if these acceptance criteria are not met.

However, this figure does not contain these restrictions.

Provide the appropriate operating restric-tions if Level 1 criteria are not met.

C.

Expand the description of the provisions which ensure reproducibility of the core power distribution measured in the test of the TIP system, considering both random noise and geometric components.

Provide assurance that the process computer properly calculates the power distri-bution in the reactor core for both symmetrical and non-symmetrical rod patterns.

Provide technical justifica-tion for the Level 2 acceptance criteria.

d.

Expand the description in Section 14.2.12.3.16 of the FSAR, of the test methods in the selected process tempera-tures test.

Revise the acceptance criteria to be consistent with the stated purposes of the tests.

423-9

Modify or clarify the acceptance criteria for the system thermal expansion test during heatup of the WNP-2 facility to provide assurance that the design stress levels or fatigue limits will not be exceeded.

In the description of the core power-void mode test, indicate the mode of control (i.e., either auto or manual) of each of the principal control systems for each test condition.

Provide technical justifi.cation, or the

bases, for the Level 2 acceptance criterion to assure, if the acceptance criterion is just satisfied, that stable performance can be expected throughout core life.

In the description of the pressure regulator startup

test, indicate the mode of control (i.e., either auto or manual) of each of the other principal control systems for each test condition.

Provide technical justification, or the bases, for the Level 2 acceptance criterion (paragraph

1) to assure, if the acceptance criterion is just satisfied, that stable performance can be expected throughout core life.

Modify the description of the feedwater system startup test to indicate the mode of control (i.e., either auto or manual) of each of the other principal control systems for each test condition for the feedwater control setpoint changes.

Provide technical justification, or your bases, for the Level 2 acceptance criterion (paragraph

1) to assure, if the acceptance criterion is just satisfied, that stable performance can be expected throughout core life.

Additionally, modify the test description to include a feedwater heater trip and to specifically identify:

(1) the type of trip to be initiated; (2) the feedwater heater(s) involved; and (3) a discussion of how the planned trip compares with the worst case limiting event that could be caused by either a single equipment failure or by an operator error.

Modify the acceptance criteria for this latter transient to:

(1) identify the parameters or variables to be monitored; (2) provide assurance that the test results of this transient wi 11 be compared with the predicted response for the actual test case; and (3) provide quantitative acceptance

criteria, and their bases, for the comparison of the actual test results with the predicted response of those variables and parameters which will be monitored.

Modify the description of the turbine valve surveillance test to ensure that the rate of valve stroking and timing of the close-open sequence is consistent with the conditions which will be experience during surveillance tests.

423-10

Modify the acceptance criteria for the tests of the main steam isolation valves (MSIV's) which demonstrate that full closure of the MSIV's at 100 percent power is obtained, so as to:

(1) identify the parameters and variables that wi 11 be monitored; (2) provide assurance that the results of the transients tests will be compared with predicted response for the actual test case; and (3) provide quanti-tative acceptance

criteria, and their bases, for the comparison of the actual test results with the predicted response of those variables and parameters which will be monitored.

Additionally, provide acceptance criteria for the performance of the safety/relief valves and the RCIC during this transient.

Provide acceptance criteria for minimum values of individual valve closure times.

Modify the description of the turbine trip and generator load rejection tests to:

(1) indicate that both a turbine trip and a generator load rejection test will be conducted from approximately full power; (2) correct the Level 1

acceptance criteria to be consistent with your design; (3) identify the variables or parameters to be monitored for each trip; (4) provide assurance that the test results will be compared with the predicted response for the actual tests for each type'f trip; (5) provide quantita-tive acceptance

criteria, and their bases, for the comparison of the test results with the predicted response for the variables and parameters which wi 11 be monitor ed for each type of trip; and (6) provide acceptance criteria for grid stability with respect to both voltage and frequency, following generator load rejection trips.

Modify the abstract of the recirculation flow control startup test to specify the mode of control (i.e., either auto or manual) for each of the other principal control systems for each test condition where system stability checks will be conducted.

Provide technical justification, or your bases, to assure, if Level 2 acceptance criteria are just satisfied, that stable performance can be expected throughout core life.

Modify the description of the recirculation system startup test to define the various types of trips, including the single and double pump trips, to be conducted for each test condition and the manner in which the pumps will be tripped.

Additionally, modify the test description for the pump trips of the recirculation

pump, and provide the appropriate acceptance criteria for:

(1) flow coastdown; and (2) the transient setpoints for the APRM flow biased rod block and scram.

Provide stabi'lity criteria for the performance of the WNP-2 facility following these trips.

423-11

n.

Modify the abstracts of the loss of turbine-generator test and the, loss of offsite power test to:

(1) describe the initial plant conditions for each test, including the lineup of the plant's electrical system; (2) describe the type of trip to be conducted; (3) identify the variables, parameters and plant equipment to be monitored; (4) provide assurance that the test results will be compared with the predicted response for the actual test case; (5) provide quantitative acceptance

criteria, and their

bases, for the comparison of the test results with the predicted response for the variables and parameters which will be monitored; and (6) provide acceptance criteria for plant equipment required to function during or following these tests.

0.

In your de seri pti on of the tes t of the reactor water cleanup

system, you state that this test will be run in three modes as described in the system process diagram.

However, the system process diagrams (i. e.,

Figures 5.4-17a, 5.4-17b, 5.4-17c and 5.4-18) do not define the three modes.

Modify the Level 2 acceptance criteria in this test description to correspond with the information presented in Section 5 of the FSAR.

Additionally, expand this test description to discuss automatic isolation on the initiation of the standby liquid control system.

p.

In your description of the acceptance criteria for the residual heat removal (RHR) system test, you state that these acceptance criteria are based on flow rates and temperatures in the process diagrams.

However, the RHR system process diagrams do not contain this information.

Modify the acceptance criteria to include the necessary information.

Indicate the status or mode of operations of the plant control systems (i.e., either automatic or manual) for all transient tests.

Provide acceptance criteria for the performance testing of the plant control systems.

For example, modify the test descriptions of Items (f), (g), (1) and (m) of the preceding question.

In Section

5. 2. 2. 4.

1 of the FSAR, you state that the preopera-tional and startup testing of the safety/relief valves will include monitoring of the discharge line movement.

Modify the startup test description to reflect this commitment.

In Section 5.2.5.5.5 of the FSAR, you state that alarm points for the leak detection system will be determined analytically or will be based on measurements of appropriate parameters which will be monitored during the startup or preoperational 423-12

tests.

Modify the test description to identify these parameters and to indicate how you will establish the alarm points.

ln Section 6.3.2.2.3 of the FSAR, you state that during preopera-tional testing of the low pressure core spray (LPCS) system, the size of the discharge flow orifice will be established to limit system flow to acceptable values as described in the LPCS process diagram.

Modify the test description to reflect this commitment.

Provide a preoperational test description for the various modes and systems of the fire protection system.

Provide a schedule relative to the fuel loading date, which identifies the scheduled time for performing the low power testing and the power ascension testing.

423-13

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