ML17272A196

Forwards Responses to Questions from Csb,Core Performance Branch,& Geosciences Branch,As Well as Responses or Updates to Previous Questions from Csb.(Only 1 Copy of Ref Repts Sent,Filed in Geosciences Branch,259pp.)

ML17272A196
Person / Time
Site:	Columbia
Issue date:	11/21/1978
From:	Renberger D WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 7811280265
Download: ML17272A196 (182)

Text

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'r'Rc,+ f REGULATORY INFORMATION DISTRIBUTION STEM DOCKET NBR: 5P-397 WPPSS RECXPIENT:

VARGA S.A.

ORIGINATOR-RENBERGER D. L.

C COMPANY NAME:

WA PUB PWR SUPPLY 'SYS

SUBJECT:

DOC DATE-ACCESSION NBR:

COPIES RECEIVED:

LTR ENCL SIZE-ll3 Forwards responses to questions from CSB Core Performan Geosciences

Branch, as well as res onses or u dates to from CSB.

(Only l copy of ref re ts sent Filed in Geoscience 259 pp.)

DISTPIpNTTO~ CODEX'OOt DISTRTRUTIO'4 TITLE)

P%4>/FSAR A~DT 8 4ND RELATED CORRESPONDENCE NOTARXZED g 4!4E ASST BR CHTFF PROJ

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OELD OPERAvpR I.IC WV E~ERQF!!CY PLA~>

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ENr, MECH rut'R STRUCTI!RAL EMG RR HATERTAL F<6 RR AD FOR REAC SFTY ENCL?

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ORIGINATOR:

REGULATORY INFORMATION DISTRIBUTION STEM DOC DATE:

ACCESSION NBR:

COPIES RECEIVED:

SUBJECT:

LTR SIZE:

ENCL FOR ACTION REACTnP SYSTEMS SR ANALYSTS RR CORE tFuFntii"ANCE ~rt AD FDO PLAt'T SYSTEMS AUXIL T A>Y SYS BR CONTAth.>EhT SYSTEMS I

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SYSTE"~S RR PO<<ER SYS Rtt AD Fnt~

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 OEV%)UM POR:

TERA Corp.

PROD:

SUBJECT:

US iiRC/TTDC/Distribution Services Branch Special Document Handling Requirements O 1.

Please use the following special distribution list for the attached document.

2.

The attached document requires the following special considerations:

Do uot send oversize enclosure to the HRC PDR.

Q Only one oversize enclosure uas received - please return for Regulatory Pile storage.

g Proprietary inzomation - sand arridavit only to the

.'IRC PDR father:

(speci=y)

( q Pep:r-pea)c<

RPWi cc:

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'Z'ZDC/DSB Authoriz 'ature

Washington Public Power Supply System A JOINT OPERATING AGENCY P. O. BOX 958 3000 GEO. WASHINOTON WAY RICHLAND, WASHINOTON 99353 PHONE (509) 375 5000 November 21, 1978 G02-78-259 Docket Ilumber 50-397 P Director, Office of Nuclear Reactor Regulation U. S. Regulatory Commission Washington, D.

C.

20555 Attention:

Mr. S.

A. Varga, Chief Branch No.

4 Division of Project Management

Subject:

WPPSS NUCLEAR PROJECT NO.

2 RESPONSES TO CONTAINMENT SYSTEMS

BRANCH, CORE PERFORMANCE

BRANCH, AND GEOSCIENCES BRANCH UESTIONS

Reference:

Letter, SA Varga, NRC, to WPPSS, "Additional Acceptance Review guestions for the WPPSS Nuclear Plant No. 2",

dated September 18, 1978.

Dear Mr. Varga:

Attached please find sixty (60) copies of the responses to the questions from Containment Systems Branch (CSB), Core Performance

Branch, and Geosciences Branch submitted to WPPSS in the referenced letter.

This submittal is consistant with the schedule provided you at the October 10th meeting (copy attached).

Also included are responses or updates to previous questions from CSB where they were required.

SA Varga Page 2

These responses will be formally inserted in the FSAR as an amendment after submittal of the responses to the Instrumentation and Control Branch questions (scheduled for January 1).

Very truly yours',

DLR:OKE:cph attachment D.

L.'RENBERGER

Assistant Director Technology CC:

JJ Verderber - BSR, w/o responses JJ Byrnes - B8R, w/o responses RC Root - BRR Site, w/o responses HR Canter - BSR, w/o responses D.

Roe - BPA. w/o responses FA MacLean - GE, San Jose, w/o responses I. Littman - WPPSS, NY, w/o responses E.

Chang - GE, San Jose, w/5 copies of responses J. Ellwanger - B&R, w/5 copies of responses NS Reynolds - Debevoise 8 Liberman, w/1 copy of responses

~

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IV.

SCHEDULE Containment Systems Branch Nov 15 (18 questions)

IGC Branch Jan 1

(25 questions)

Reactor Systems Branch Dec 15 (4 questions)

Core Performance Branch Nov 17 (3 questions)

Geosciences Branch (4 questions)

Nov 15 Hydrology/Meteorology Branch Dec 15 (1 question)

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Docket No. 50-397 Responses to Containment Systems

Branch, Core Performance

Branch, and Geosciences Branch guestions STATE OF WASHINGTON )

)

ss COUNTY OF BENTON

)

D. L.

RENBERGER, Being first duly sworn, deposes and says:

That he is the Assistant Director, Generation and Technology, for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that he is authorized to submit the foregoing on behalf of said applicant; that he has read the foregoing and knows the contents thereof; and believes the same to be true to the best of his knowledge.

Qc)

DATED O'O~lae~ J7 1978 D. L.

RENBERGER On this day personally appeared before me D. L.

RENBERGER to me known to be the individual who executed the foregoing instrument and acknowledged that he signed the same as his free act and deed for the uses and purposes therein mentioned.

GIYEN under my hand and seal this M~ day of

, 1978.

g

'I Notary Public in and f r the State of Washington Residing at

REVISIONS TO PREVIOUSLY SUBMITTED CSB QUESTIONS

@go-S~~

Oackat + gg g]gg Caustral 4'g6~lzr 8

of Bocumsnt:

BECULQGRY MC'AHFl1E

PlNP-2 Provide the secondary containment pressure time response for the design basis accident.

List and discuss all assumptions'ade in this analysis.

Res onse:

The analysks is continuing along the outline stated in Amendment No. 1.

However, the effort is more time consuming than originally contemplated.

It is now estimated a reponse will be forthcoming in February 1979.

'0

s WNP-2 0 22.10 (6.2-6)

Provide the following information, with regard to the leakage rate testing of the Type C containment isolation valve (as defined in Appendix J to LOCFR Part 50):

a e For each fluid line that penetrates the con-

tainment, show schematically the isolation valve arrangement and the design provisions that, will permit the isolation valves to be leak tested.

Indicate the direction in which the valves will.be leak tested.

Xdentify the containment isolation valves that will'not be subjected to Type C leak testing, and provide justification for not leak testing these valves.

Resnonse a ~

b.

All fluid lines which penetrate the primary containment and the associated containment isolation valves which wilL be subject to

,Type C testing are illustrated in Figure 6."2-3L(a-t).

These figures also indicate how the Type C tests will be conducted.

The only lines which have not been identified as subject to Type C testing are:

1..A11 instrument lines (vss<K (o.z Ie, IIorI 27-)

2.

cRD insert 'and withdrawal lines(.TAou.(,K-I poorE g) 3.

TIP lines CT~~

Nun. 29'j 4.

RRC hydraulic controL lines (7W~ (n.~-~~, +>< ~~3

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NOTES ON TYPE C TESTING (ISOLATION VALVE LEAKAGE TESTING)-

1.

TYPE C TESTING IS PERFORtiED BY APPLYING A DIFFERENTIAL PRESSURE IN THE SAME DIRECTION AS SEEN BY THE VALVES DURING CONTAINMENT ISOLATION.

2.

TYPE C TESTING IS PERFORtiED BY PRESSURIZING BETWEEN THE TWO-PIECE DISK GATE VALVE.

3.

4

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TYPE C TESTING IS PERFORMED BY PRESSURIZING BETWEEN THE ISOLATION VALVES.

THE TEST YIELDS CONSERVATIVE RESULTS SINCE THE INBOARD GLOBE VALVE IS PRESSURIZED UNDER THE SEAT DURING THE TEST;

WHEREAS, DURING CONTAINtKNT ISOLATION/

IT IS PRESSURIZED ABOVE THE SEAT.

1 TYPE C TESTING IS PERFORMED BY PRESSURIZING BETWEEN THE ISOLATION VALVES.

THE TEST YIELDS EQUIVALENT RESULTS FOR THE INBOARD GATE OR BUTTERFLY VALVE.*

5.

6.

TYPE C TESTING IS PERFORMED BY PRESSURIZING THE ISOLATION VALVE IN THE OPPOSITE DIPECTION AS WHEN THE VALVE PERFORMS CGNTAINt1ENT ISOLATION.

SI:FACE THE ISOLATION VALVE IS A GAT" VALVE, THE TEST YIELDS EQUIVALENT, RESULTS.*

TYPE C TESTING IS PERFORMED BY PRESSURIZING BETWEEN THE ISOLATION VALVES.

THE TEST YIELDS EQUIVALENT RESULTS. FOR THE INBOARD GATE VALVE.'*'HE ONE INCH GLOBE VALVE WILL

'AVE TEST'RESSURE APPLIED UNDER THE SEAT HOWEVER I THE DIFFERENCE BETWEEN TESTING A ONE INCH GLOBE VALVE OVER OR UNDER THE SEAT IS CONSIDERED NEGLIGIBLE' THE GATE Ai)D BUTTERFLY VALVES ARE EQUALLY LEAK TIGHT If< EITlfER OIRECTIOH BECAUSE OF SYNf'fETRY OF DESIGfl A!ID BECAUSE OF COl'1STRUCTIOif.

THIS FACT HAS BEEfl'OI'fFIRI1ED BY REVIEt'J OF LEAKAGE T~DATA AND OTHER Il(FORI'NTIOH SUPPLIED BY THE VALVE "'1Ai'lUFACTURERS o I,

/

WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLZAR PROJECT NO 2.

NOTES ON TYPE C TESTING FIGURE 6.2-3la

CSP-V-l C SP-V-3 AO CSP-V-Z, CSP-V-4 AO FOR X-4&ONLY 5E.E. F18.&.2-31'.

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REACTOR t=EEDWATER, LINES WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO 2.

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WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO 2

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0 22.12 For those isolation systems which satisfy the requirements of Criterion 57 of the Design Criteria, provide adequate information to demonstrate that these systems meet the requirements of a closed system as stated in Section 8.9 of BTP-CSB 6-3.

RESPONSE

The only lines presently listed as falling under GDC 57 (see 6.2.4.3.2.3) are the hydraulic control lines for the RRC flow control valve.

Note 28 of the revised Table 6.2-16 evaluates these lines against GDC-57.

See response to question 22.027 for revised Table 6.2-16.

RESPONSES TO CSB QUESTIONS*

• per letter, NRC to MPPSS, Sept.

18, 1978.

E t'ai 022. 013 6.2.1 Provide the following additional information with respect to the secondary containment:

a.

Appropriate plant elevation and section drawings for those structures and areas that will be maintained at negative pressure following a postulated loss-of=coolant accident and that were considered in the dose calculation model.

b.

Your proposed technical specification limit for leakage which may bypass the filters of the standby gas treatment system (e.g., valve leakage and guard pipe leakage);

and c.

A discussion of the testing methods which will be used to verify that the systems provided are capable of reducing and maintaining the pressure within all secondary contain-ment volumes to a negative pressure of 0.25 inches (water guage).

RESPONSE

Information on the secondary containment is provided in 6.2.3; secondary containment functional design.

Specifically, a.

The reactor building is the secondary containment, refer to Figures 1.2-3 thru 1.2-7 and'3.8-1 and -Z.

b.

Refer to response to NRC question 022.6.

The isolation valves on the lines which have been identified as potential secondary containment bypass paths will be tested to ensure the total bypass leakage rate is equal to or less than

.4 scfh as outlined in 6.2.3.3.

c.

Refer to 6.2.3.4 and 7.3.1.1.7.

The reactor building contains sufficient openings to maintain uniform pressure throughout the building, refer to previously listed figures.

• 6.2.3 will be changed as per the attached draft.

.RIP-2 rate of one psi per second.

The effects of the design basis tornado pressures on the structure axe discussed in 3.3, and tornado gene ated.

missiles are discussed in 3.5..

g.

The reactor building 's designed for all probable combinations of the 'design basis wind and the design. baiis tornado velocities and associated difference's of pressure within the structure and atmospheriq pressure outside the structure.

For structural'design purposes the various pres-sures considered acting within the secondary con-tainmeat, structure include:.

C 1)

A negative'nternal pressure of 0.25 inches of water under which the secondary contain-

.ment normally operates.

2)

A negative internal pressure of 0.012 psig, relative to the outside atmosphere, which exceeds the 0.25 inches of water in g(I), to account for any uncertainty in pressure measurement and to account for any negative pressures actually developed by the standby gas treatment system or by unknown causes.

3)

A positive pressure of 0.25 psig relative to the outside atmosphere, to account for any positive pressure transient which the secondary containment may experience fol-lowing a postulated pipe break in the secondaxy containment, to account for any'utward positive differential pressures created by wind loads, and to account for any uncertainty in pressure measurements.

6.2.3.2

System Design

Refer to Figures for genera arrangement drawings of the reactor building showing plan "

and elevat'on views of:the boundary of the structure.

Also refer to Figures 3.8-1 and 3.8-2.

Refer to Table 6 2-12, Secondary Containment Design and Performance Data, for the design and performance data for the secondary containment structure.

6.2-45

22.014 Note 2 of Table 6.2-16 states that the suppression pool serves as an isolation barrier to the environment.

It is our position that this is unacceptable.

RESPONSE

The penetrations that reference note number 2 in Table 6.2-16 meet General Design Criteria 56 and include a single containment isolation valve and are closed systems outside containment.

This provides.double barrier containment isolation and the water in the suppression chamber is not considered as part of containment isolation.

The closed systems outside containment are protected from missiles, are Seismic Category 1, Safety Class 2,

and have a design temperature and pressure at least equal to that of the containment.

flote number 2 of Table 6.2-16 is being deleted.

Table 6.2-16 will be revised to include the General Design Criteria for all penetrations.

(See response to question 22.027).

g.

22. 15 Provide a detailed drawing showing the locations of the containment spray headers relative to the internal structures.

RESPONSE

The containment spray system is discussed in 6.5.2.+

The suppression chamber spray header is shown in Figure 3.5-3.

The lower drywell spray header is shown in Figures 3.5-4, 3.5-21, 3.5-22, 3.5-25, 3.5-.26, 3.5-27 and 3.5-28.

The upper drywell spray header is shown in Figures 3.5-16, 3.5-17, and 3.5-18.

• See draft on following page

riMP 2

6.5.2 CONTAXNMENT SPRAY SYS'ZEi4 6.5.2.1 Design'Bases The containment spray system is capable of quickly reducing containment pressure during the post-accident period of a LOCA through condensation of"@team in the drywell and through cooling of the non-conden'sab3;e gases in the free volume above the suppression pool.

Contai,nment spray is not required to prevent overpressurization of'he containment (see 6.2.1).

The containment spray system is not used for fission product removal from the containment atmosphere.

I

~

I 6.5.2.2 System Design The drywell spray consists of two independent 3.oops and spray headers.

The suppression chamber spray consists of one spray header supplied from two otherwise independent loops.

Since the water source for all containment spray is the suppression pool, the spray system is a closed loop cooled by the RHR heat exchangers.

The rated flows for drywell and suppression chamber sprays are 7450 gpm/loop and 450 gpm/loop, respec-tively.

Containment spray is a subsystem of the RHR System (5.4.7).

The drywell spray 'valves are electrically 'terlocked to a3.low actuation of the drywell spray only when there is a high dry-well pressure signal present.

After a high drywell pressure signal is present, a second electrical interlock prevents actuation of either the drywell or the suppression chamber spray lines until the corresponding LPCX injection valve is shut.

A procedural restriction proh'bits the operators during the irst ten minutes ollowing a LOCA from closing a

LPCX in-jection valve and interrupting core cooling (Refer to 6.2.2.2).

Containment spray must be initiated and secured by ooerator action.

Hydrogen mixing whicn resu3.ts from containment spray operation is discussed in 6.2.5.

SVpp<C55lOA C313pvlgeI 6.5-12

22. 016 Provide the value of the external design pressure for the containment structure.

RESPONSE

The primary containment is designed for a total external pressure of 4 psid;

however, since the compressed insulation between the concrete biological shield and the containment exerts a uniform 2 psid external pressure-half of the total external pressure differential allowed - the reactor building pressure may be no greater than 2 psi above the primary contain-ment pressure.

This value is given in Table 6.2-1, "Containment Oesign Parameters",

and Section

6. 2. 1. 1.2.

22. 017 Provide the assumptions and initial conditions for the activation of both drywell spray loops following a postulated loss of coolant accident (LOCA) that has purged all the drywell noncondensable gases into the suppression chamber; provide the same information for the inadvertent drywell spray activation at normal operation conditions.

This information is requested so that we may perform an independent evaluation of the reverse differential pressure across the drywell floor to establish the degree of conservatism in the small steam line break analysis.

RESPONSE

The assumptions used for activation of both drywell spray loops following a small steam line break are discussed in 6.2.1.1.4.

Initial conditions are given in Table 6.2-19*, Initial Conditions Employed in Negative Pressure Design Evaluation, and Table 6.2-1, Containment Design Para-meters.

The same information should be used for inadvertent activation of drywell spray during normal operating conditions except that assumption d.

and e. (in 6.2.1.1.4) are not applicable.

• See next page for draft

Table 6.2-19 Initial Conditions Employed in Negative Pressure Design Evaluation A.

Containment preincident conditions used for sizing internal vacuum breakers (wetwell to drywell)

1. Pressure, psig

2. Temperature, F

3. Relative Humidity,Ã Drywell (DW)

.75 135 20 Suppression Chamber (WW)

.75

'50 100 8.

Containment preincident conditions used for sizing external vacuum breakers (reactor building to wetwell)

1. Pressure, psig

2. Temperature, F

3. Relative Humidity,

'5 Drywell (DW) 0 150 30 Suppression Chamber (WW) 0 50 100

NNP-2 6.2.1.1.4 Negative Pressure Design Evaluation The limiting transient for the wetwell-to-drywell (WW-DW) and reactor building-to-wetwell (RB-WW) vacuum breaker systems is simultaneous operation of both drywell spray loops after a small LOCA.

This transient has been deter-mined by analysis to be more severe than any of the following:

a.

ECCS flow from a recirculation line break b.

One drywell and one wetwell spray line activation following a small LOCA c.

Xnadvertant operation of one drywell spray line during normal operating conditions The analysis performed for the case of simultaneous opera-tion of both drywell loops after a small LOCA made the following conservative assumptions:

a.

Drywell spray flow of 8400 gpm from each loop.

This flow corresponds to runout flow for the RHR pumps.

b.

100% spray efficiency c.

50 F spray temperature d.

All non-condensible gases are purged into the wetwell as a result of the LOCA.

e.

The drywell is full of steam at a pressure corresponding to wetwell pressure plus the hydrostatic head corresponding to the down-comer submergence.

Drywell spray is not reauired to maintain the primary con-tainment below design pressure nor is it required for con-tainment cooling.

Xf following a small LOCA all the non-condensible gases are purged into the wetwell, actuation of one of the two drywell sprays will rapidly depressurize the drywell.

Actuation of the second drywell spray with the -drywell full of steam is neither necessary nor desir-able and represents the limiting transient for the vacuum breaker systems.

lylALQ.I ~oaclI4tbfls U58J Lr1

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iaBIIe g.2.-/'l.

Q. 022.018 (6.2.1)

I He require your compliance with our proposed position on containment steam bypass for small breaks.

The details are contained in the Containment Systems Branch Technical Position, "Steam Bypass for Hark II Containments,"

a copy of which is enclosed.

Q. 031.070 (RSP)

(6.2.2.2)

(6.5.2.2)

(7.3.1.1)

(031.001)

(031.011)

It is the staff's position that insufficient time is available for the operator to reliably take. the manual actions which are necessary to initiate suppression pool spray during a small break.

The staff has established the requirement for automatic initiation of suppression pool spray for the Hark II containment.

Accordingly, we require you to provide a Class IE automatic control system for each suppression pool spray system.

Response

The WNP-2 design meets the intent'f the proposed CSB Branch Technical Position on "Steam Bypass for Nark II Containments".

The history of the question of steam bypass on HNP-2 is extensive, dating back to January 1972.

Questions 5.4, 5.22, and 5.24 to the PSAR all respond to the con-cern.

The SER (pp 63-65) summarized the NRC position on the issue at the CP stage and noted that HPPSS agreed to study additional means to mitigate the consequences or minimize the potential for bypass leakage.

This was formally documented as a Post CP item in the notes of a NRC-WPPSS meeting held on October 17-18, 1973 (Reference 1).

In the notes WPPSS committed to submitting a report on the matter.

In August 1974, Reference 2 trans-mitted the WPPSS report WPPSS-74-2-R5, "Drywell to Wetwell Leakage Study", satisfying the commitment.

The NRC requested additional information concerning the report in Reference (3).

References (4) and (5) provided HPPSS responses to the NRC questions.

Reference (6) indicated that Structural Engi-neering Branch found the applicable WPPSS responses accept-able.

WPPSS has no record of eedback from Containment

Systems Branch on the responses to its questions but assumed in Reference (5) that, in the absence of feedback, the post CP item was resolved.

Accordingly, IlPPSS has gone ahead with construction in these areas based on the above correspondence.

A point by point discussion summarizing the lrjPPSS design capabilities to mitigate Bypass Leakage problems based on the above correspondence and with respect to the oro-posed Branch Technical Position is given below:

1.

NRC Proposed Requirement:

Allowable bypass capability on the order of 0.05 ft (A/ ~

2

~Res ense:

As documented in reference 5 and the PSAR, the maximum allowable bypass leakage capacity is A/v K=.028 ft2 using conservative calculational techniques and assumptions.*

MPPSS, therefore, believes the existing calculations meet the intent of A/~K = 0.05 ft.

2.

t<RC Pro osed Re uirement:

An automatic system should be provided to initiate automatic wetwell sprays.

The system should meet the standards of an Engineering Safety Feature including redundancy and diversity and be actuated automatically ten minutes following a LOCA. If the RHR system is used for this purpose, it must be analyzed to assure no degradation of its ECCS function.

~Res onse:

liPPSS asserts that manual initiation is sufficient since the drywell floor will be routinely tested and evaluated against a Tech Spec limit of A/~K = 0.0045 ft,

a level at 2

which no operator action is required for the spectrum of small break sizes.

(Reference 5 - see

>43 below for testing details).

• The FSAR currently lists the capability as 0.026 ft This 2

is from a GE analysis and the FSAR is being amended to reflect the latest calculations (see attached draft change).

0

The construction, design, quality control, and surveillance requirements on the drywell floor give it the same level of safety as the containment itself.

Reference 4 and Part VI of reference 2 showed that through-wall cracks will not develop through the concrete slab under postulated desiqn conditions including the SSE and that leakage in excess of that accounted for due to permeability would not be possible.

Reference 6 indicated the NRC Structural Engineering 8ranch's acceptance of these responses.

Accordingly, l<PPSS sees no reason to assume that an A/JK of.0045 ft is exceeded any more than there would be 2

reason to assume the design containment leak rate of.5/ per day is exceeded.

Calculations documented in Reference 5

using the CONTEMPT - LT computer code were used in computing the maximum allowable leakage rate of AgK =.028 ft, six times the Tech Spec limit.

In the calculation over 167

.minutes was available for operator action before drywell design pressure was exceeded.

Accordingly, a requirement that an automatic system be provided is unnecessary.

3.

NRC Pro osed Re uirement:

A single preoperational high pressure leakage test should be performed and periodic low pressure tests at each refueling outage with an acceptance criterion of 10/ of the bypass capability at the test pressure.

~Res onse:

The intent of this proposed requirement has been committed to by l(PPSS.

A single preoperational leakage test will be conducted with the downcomers capped at 15 psid and 25 psid (the design drywell to wetwell differential pressure).

At each refueling outage a low pressure opera-tional test will be performed as a Tech Spec Surveillance Requirement to verify.0045 ft.

Details of the nature of 2

this test are discussed in question 5.22 to the PSAR but will be summarized here since the specific numbers have been since updated.

Routine Leak Testin and Ins ection: During entry to the drywell at each refueling outage, accessible drywell to wetwell barrier surfaces will be visually inspected to ascertain any possible leak paths.

Vacuum relief valves will be visually inspected to insure they are clear of foreign material.

At each refueling outage, before the primary system is pres-surized, after all these containment inspections are

complete, and after the vacuum breakers are exercised, the following test shall be carried out:

The drywell will be pressurized to at least 1.0 psi above the wetwell.

After an adequate stabiliza-tion period, the drywell to wetwell leakage rate will be measured.

The acceptance criterion will correspond to an equivalent leakage capacity (A/~K) of 0.0045 ft, which is 16% of the allow-able leakage.

If a greater leakage rate is found, the containment shall be entered and the cause deter'mined and corrected and the test repeated.

4.

NRC Pro osed Re uirement:

Vacuum relief valves should have redundant position indicators with indication and redundant alarms in the control room.

The vacuum breakers should be operability tested at monthly intervals to assure free movement.

~Res onse:

MPPSS meets this requirement with the current design.

Each vacuum breaker penetration consists of two discs in series, each disc with redundant position indication which display in the Control Room.

Each vacuum breaker disc wi 11 be equiped with an exercising mechanism and each disc will be exercised at a frequency equivalent to the testing of ECCS valves.

References 1.

Letter, MR Butler, NRC> to JJ Stein,

MPPSS, "Meeting Summary October 17-18,

1973, dated November 26, 1973.

2.

Letter, MPPSS to NRC, G02-74-17, dated August 9, 1974.

3.

Letter, NRC to MPPSS, dated January 14, 1975.

4.

Letter, MPPSS to NRC, G02-75-52, dated February 2S; 1975.

5.

Letter, MPPSS to NRC, G02-76-156, dated April 23, 1976.'.

Letter, NRC to MPPSS, dated tray 15, 1975.

FSAR Change to Section 6.2.1.1.5.4 connected with SCH 78-25 and NRC questions 22.018 and 31.070

a.

Flow through the postulated leakage path is pure steam.

For a given leakage path, if tne leakage flow consists of a mixture of liquid. and vapor, the total leakage mass flowrate is higher but the steam flowrate is less than for the case of pure steam leakage.

Since only.the steam enter-ing the suppression chamber free space results in the additional containment pressurization, this is a conservative assumption.

b.

There is no condensation of the leakage flow on either the suppression pool surface or the containment and vent system structures.

Since condensation acts to reduce 0he suppression chamber press'ure, this is a conservative assump-tion.

For an actual containment tnere will be condensation, especially for the larger primary system. break where vigorous agitation at the pool surface will occur during blowdown.

6.2.1.1.5.4 Analytical Results

\\

The containment has been analyzed-to determine the allowable leakage between the drywell and suppression chamber.

Figure 6.2-17a.shows the allowable leakage capacity (A// K} as a

function of primary system break area.

A is the area of the leakage flow path and K is the total geometric loss coefficient associated with the leakage flow path.

Figure 6.2-17~is a composite of two curves.

Xf the break area is greater than approximately 0.4 square feet, natural reactor depressurization will rapidly terminate the trans-ient.

For break areas less than 0.4 square feet,

however, continued reactor'lowdown limits the allowable leakage to small values.

'he, maximum allowable leakage'apacity is at A/V K =.026 square feet.

Since a typical geometric loss factor would be three or greater, the maximum allowable flow path would be about

.052 square feet. This corresponds to a 3 inch line size.

p z. ~ sc"a-

~ Q 6.2.1.1.6 Suppression Pool Dynamic Loads A generic discussion of the suppression pool dynamic loads and asymmetric loading conditions is given in Nark.

XE Dynamic Forcing Function Information Report, Reference 6.2-4.

A unique plant assessment of these dynamic loads is made in WNP-2 Design Assessment

Report, Reference 6.2-5.

6.2.1.1.7 Asymmetric Loading Conditions See comment in 6.2.1.1.6.

6.2-30

1)

(Insert for page 6.2-30)

Burns and Roe, Inc. confirmed the results of the above analysis py GE in reference 6.2-7.

Further investigation into the transient nature of the problem was then 'undertaken at the request of the NRC.

A transient analysis using the CONTEMPT-LT(Ref. 6.2-8) computer code was performed.

The code was modified to include the mass and energy transfer to the suppression pool from relief valve discharge.

The limiting case was a very small reactor system break which would not automatically result in reactor depressurization.

For this limiting case, it was assumed that the response of the plant operators was to shut the reactor down in an orderly manner at 100 F/hr cooldown rate.

No other operator action was accounted for.

Heat sinks considered were such items as major support steel inside containment, the reactor

pedestal, the diaphragm floor and support columns and the steel and concrete of the primary containment.

Based on this analys'is, the allowable bypass leakage (A/ J K) was 0.028 ft.

The drywell pressure 2

transient is shown in Figure 6.2-17b along with the corresponding curves of wetwell pressure, wetwell temperature and suppression pool temperature.

The allowable bypass leakage of 0.028 ft is well above the maximum 2

possible containment bypass leakage..

Periodic testing will be per-formed to confirm that the containment bypass leakage does not exceed A/~K = 0.0045 ft.

Figure 6.2-17c presents the resulting containment transient for A/ ~K = 0.0045 ft The peak containment pressure shown in Figure 6.2-17c is well below the containment design pressure.

1

bNP 2

6.2.7 REFERENCES 6.2-1

James, A. J.,

"The General Electric Pressure Suppression Containment Analytical klodel",

April 1971, (NEDO-10320).

6. 2-2 6~2 3

6. 2-4

6. 2-5 6.2-6

James, A. J.,

"The General Electric Pressure Suppression Containment Analytical i~lodel",

Supp3.ement 1,

May 1971 (NEDO-10320).

i~loody,'. J.,

"Maximum Two-Phase Vesse2.

Blowdown from Pipes", Topical Report APED-4824, General Electric Company, 1965.

"MK II Containment Dynamic Forcing Functions Information Report (Revision 2)",. General Electric and Sargeant and Lundy, NEDO-21061, September 1976 "Plant Design Assessment Report for SRV and LOCA Loads (Revision

1) ", Washington Public Power Supply

System, February 1978.

J.

D. Duncan and J. E. Leonard, "Emergency Cooling in BNR's Under Simulated Loss-of-Coolant (B!7R PLECMP) Final Report," FEAP-13197, Genera3. Electric, June 1971.

6.2-7 WPPSS REPORT, "Drywell to Wetwell Leakage Study", WPPSS-74-2-RS,

~

~

~

~

July, 1974.

(Submitted to NRC by WPPSS to HRC, Ltr. G02-74-17, dated Aug. 9, 1974).

f pigmeat, L. 'L.;

$Bgner, R. J.; Niederauer, G

F ; Obenchain, L. F.

CONTEilPT-LT--A COMPPZER PROGRA'I FOR PRZDZCTZNG CONTAI'iiii'=NT PRESSURE-TEM ERATUR" RESPONSE TO A. LOSS-OF-COOLANT ACCIDENT, ANCR-1219, Aeroject Nuclear Company, J "ne, 1975.

6. 2-88

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6.2.1 Provide the indormation requested in paragraphs B.l.a, B.l.d, B.l.g, 8.4 and 8.5 of Branch Technical Position CSB 6-4, "Containment Purging. During Normal Plant Operation".

RESPONSE

Refer to the response to previous question 022.1.

The valves in the containment purge and vent lines are containment isolation valves.

The design criteria for the isolation valves is discussed in 6.2.4, Contain-ment Isolation System, and 6.2.6, Containment Leakage Testing.

In addition paragraph 6.2. l. 1.8.2 has been revised to expand the definition of reactor operation and include information relative to purge isolation valve qualification/

WNP-.":

6.2.1.1.8.2 Primary Containment Purging The primary containment above the drywell floor is provided with a purge system to reduce residual contamination prior to personnel access.

This system is designed to produce a

purge rate equivalent to 3 air changes per hour of the net free volume.

The drywell is purged once a year during scheduled refueling shutdown period and as required for inspection.

The drywell purge rate is 10,500 cfm.

Provision is made to automatically route a reduced pu ge rate of 4400 cfm to the standby gas treatment system if residual airborne contamination is higher than allowable limits for direct release to the atmosphere.

Purge air is taken from the reactor building ventilation sup-ply duct through two 30" normally closed isolation valves into the primary containment.

Purge air is extracted from the drywell through two 30" normally closed isolation valves and is routed to one of two systems.

The discharge can be routed through a normally closed isolation valve to the reactor building exhaust air plenum or to the standby gas treatment system (Figures 3.2-15 and -18).

Xf a high air-borne activity occurs, the radiation monitors at the exhaust air plenum would cause the reactor buildings ventilation and primary containment purge systems to isolate.

Provision is also made to purge the suppression chamber sec-tion of the primary containment.

Purge air is taken from the reactor building supply duct through two 24" normally closed isolation valves into the suppression chamber.

Purge air is extracted from the suppression chamber through two 24" normally closed isolation valves and routed to the exhaust air plenum or standby gas treatment system in the same manner as the drywell purge exhaust.

The suppression chamber purge rate is 7500 cfm.

The above systems are designed to purge either the drywell

'or the suppression chamber.

Provision is not made to purge both areas at rated flow simultaneously.

Only one vent line and one purge lin'e will be open at any one time during reactor operation.

The purge system may be used during reactor operation only for purging the primary containment prior to personnel entry.

Purge system operation during reactor operation<will be limited to 'ess than 1% of reactor operating time.

i"<~U~>"g +<+<p, hot standby,and ho% shukdou)n 6.2-32

<sobk~on All containment purge<valves, including the 2" bypass

valves, are designed to shut within four seconds of receipt of a containment isolation s igna1. ~~w-ahab-+gains-~Md~."on-merH~~ig~reesere-eZ 46-ps~

The containment isolation signals and the purge valves are part of the containment isolation system which is an ESF system.

Each purge line has two isolation valves.

These valves are opened by allow-ing compressed air to oppose a spring in the valve actuator.

On a loss of compressed air, loss of electrical signal, or on a containment isolation signal the valve is shut.

Zf the purge system were operating at the time of a LOCA, the system will automatically be secured.

The level of the activity released through the purge system before isolation would be limited to the activity present in the coolant prior to the accident since the purge system will be isolated be-fore any postulated fuel failure could occur.

6.2.1.1.8.3 Post LOCA The unit coolers are not required after a

LOCA since heat removal is then accomplished by the containment cooling

system, a subsystem of the RHR system, as described in 6.2.2.

Similarly, containment purge is not required following a LOCA.

Two 100$ redundant hydrogen recombiners are then placed in operation to ensure that the hydrogen buildup does not reach a dangerous level.

Any equipment located inside the primary containment which is required to operate subsequent to a LOCA has been designed to operate in the worst anticipated accident environment for the required period of time

{See 3.11).

6.2.3..1.9 Post Accident Monitoring A description of the post accident monitoring systems is pro-vided in 7.5.

6.2. 1.2 Containment Subcompartments The two areas within the primary containment considered sub-compartments are the area within the sacrificial shield wall and the area above the refueling bulkhead plate at elevation 583 All potential pipe breaks within the sacrificial shield wall have been evaluated.

The information is contained in Refer-ences 3.8-5, 3.8-6 and 3.8-7.

These references have been previously submitted to the NRC.

Two analyses are being performed to ensure the adequacy of the refueling bulkhead and inner refueling bellows at elevation 583'.

The first analysis, a break of the RClC head spray line, will determine the maximum downward loading due to pipe breaks, and the second analysis, a break of the RRC suction line, will determine the maximum upward loading.

These analyses will be incorporated into the FSAR by means of an amendment.

6.2-33 All p<<qe, isoIaho~ valves h"va bean anaIq+eai(>

~oalifiad 4 eIbse.

~Q~~~H a maxlmuvrl dewan diRere.+gal pre.ssor~

g4 fgQ psi in

NNP-2 Q 022.020 (6.2. 1)

Provide the test results and method of analysis utilized to determine the seismic sloshing loads discussed in 6.2.1.1.3.1 of the FSAR.

Provide the basis for the acceptance of these loads.

Res onse Please refer to 6.2.1.1.3.1, the information requested.~

3.8.2.4.3 and 3.8.2.4.3.2 for

~ i 2

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31

~.

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~

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I

I All of the analyses assume that. the primary system and con-tainment are initially at the maximum normal operating con-ditions.

References are provided that describe relevant experimental verification of the analytical models used to evaluate tho containment system response.

Seismic sloshing effects in the wetwell were evaluated for a

SSE and were found to be minimal.

The resulting vertical water displacement at the containment wall is. less than one foot.

The analytical results and method of analysis utilized to determine the seismic sloshing effects in the wetwell are discussed in 3.8.2.4.3.

6.2.1.1.3.2 Containment Design Parameters Table 6.2-1 provides a listing of the key design parameters of the primary containment system including the design

.. characteristics of the drywell, suppression pool and the

. pressure suppression vent system.

I Table 6.2-2 provides the performance parameters of the re-lated engineered safety feature systems which supplement the design conditions of Table 6.2-1 for containment cooling purposes during post blowdown long term accident operation.

~ Performance parameters given include those applicable to

~. full capacity operation and to those conservatively reduced capacities assumed for containment. analyses.

~

~

6.2.1.1.3.3 Accident Response Analysis 4 ~ The containment functional evaluation is based upon the con-sideration of several postulated accident conditions re-sulting in release of reactor coolant to the containment.

These accidents include:

I a ~

b.

An instantaneous guillotine rupture of a recirculation line, An instantaneous guillotine rupture of a main steam lineI c.

An intermediate size liquid line rupture, and d.

A small size steam line rupture.

Energy release from these accidents is reported in 6.2.1.3.

6.2-6

4

HNP-2 Nith the. formulation of an overall mathematical model which.

provides for the realistic response of the containment sys-

tem, response spectra and/or time histories are generated at the component interfaces, and at other desired points.

These component response spectra and/or time histories are used to perform detailed dynamic analyses of the individual com-ponents as previously mentioned.

Effects due to the presence of water in the suppression pool under earthquake excitations are established. following the procedures described in reference.3.8-2.

The additional loads due to sloshing effect are included as part of the design loads of the primary containment vessel.

The analytical results and methods of analysis utilized to determine the

,'eismic sloshing effects in the suppression chamber are dis-cussed in 3.8.2.4.3.2.

The model used for the seismic analysis and the method of seismic analysis to obtain the seismic

moments, shears, dis-

"- placements and floor response spectra are discussed in 3.7.

To obtain a detailed stress/strain analysis in local areas, the following additional methods are used.

~'n the dynamic analysis of the steel primary containment'vessel component, a dynamic mathematical model is formulated which

-. incorporates the general structural geometry and all signi-ficant boundary conditions present.

The design of the numer-ous. penetrations is such that any restraining forces on the steel primary containment vessel which could be developed

"'an be considered as negligible.

The effects of rotational inertia and shear deformations are also considered as neg-

"- ligible in the response of the steel containment vessel.

In the determination of the seismic response of the steel containment, vessel, damping effects are considered.

The

-" incorporation of damping into the dynamic analysis is facili-tated by the use of viscous (velocity proportional) damping.

The various damping values for both the operating basis earthquake-and SSE excitations for the steel containment vessel are discussed in 3.7.

The resulting equations of motion for the steel containment vessel are solved by the use of DACSR, a large capcacity com-

'uter program discussed in 3.12.

The solution algorithm used depends on the analytical method incorporated to evaluate the equations of motion for the system.

A complete dis-cussion of the solution technique is provided in 3.7.

I

3. 8-43

hNP-2 The results of the dynamic seismic analysis contain values for maximum translation and rotational displacements and accelerations, moments and shears, as well as response spectra and/or time histories at desired points throughout the steel containment vessel.

These resultant forces are Mew combined with the various loading conditions as described in 3.8.2.3.12 and in accord-ance with Sub-Article NE-3131 of Section XXI of the ASME Code.

These combined forces are used in the structural analysis of the various critical areas present within the steel primary containment vessel.

By using a response spectra and/or time history the cantilevered personnel

locks, as well as any

., other appurtenance, are dynamically analyzed as previously discussed.

I The resulting stress intensities due to the addition of seis-

..mic loads to the various loading conditions for the steel pri-

,. mary containment vessel and its appurtenances will be

, in accordance with the stress intensity limits as specified

'n Sub-Article NE-3131 of the ASME Code,Section III.

-..; 3-8.2

~ 4.3.1 Computer Program Utilized in the Seismic Dynamic Analysis The seismic dynamic analysis utilizes

DACSR, a large capacity

". computer program discussed in 3.12.

The program is capable of generating the reauired mass and stiffness matrices which

.. are recuired to represent the mass and stiffness of the actual

- structure.

I

.)

I 4

~

I

~.- The model of the structure and program solutions and output

~;:are discussed in 3.7 and in 3.12.

I

.3.8.2.4.3.2 Seismic Dynamic Analysis of Hater in Suppression Chamber (Sloshing Effects)

Tests were not performed to arrive at the seismically induced sloshing loads in the suppression chamber.

All of the loads

.: are arrived at by calculations.

The calculations provide the basis for the acceptance of these loads.

I The.method of analysis utilized to determine the seismic sloshing loads in the suppression chamber is taken from Chapter 6,

Dynamic Pressure on Fluid Containers, and Appendix F,

Dynamic Analysis of Fluids in Containers Subjected to Acceleration, both contained in reference 3.8-2.

I 3.8-44

)'RIP-2 Il Two separate analyses were performed, using the formulations given in the above referenced

document, as follows:

a.

Xn the first analyses, the entire suppression chamber is taken as a cylindrical rigid tank in plan having, a flat bottom as modeled in the above referenced

document, in lieu of the actual 2:1 bottom ellipsoidal head, and sup-ported on the foundation mat.

ln this analysis

~

the 'reactor pressure vessel (RPV) pedestal is excluded from the model, and the tank is con-

..idered as containing only the water to the full depth shown in Figure 3.8-1.

~

~

g ~

g

~

b.

In the second analysis, the RPV pedestal is included in the model.

To include the pedestal, the suppression chamber is modeled to consist of theoretical rectangular tanks in plan, of the minimum quantity and the maximum size that can be fitted or inscribed adjacent to each other within the annulus formed by the cylindrical wall of the suppression chamber and the concentric cylindrical RPV pedestal.

The tanks are each assumed as independent rigid bodies supported on the foundation mat, flat-bottomed and con-taining water to the full depth shown in Figure.3.8-1.

I

. ln both analyses, the structures are subjected to the maxi-

mum floor accelerations due to the Safe Shutdown Earthquake (SSE).

The acceleration values are obtained from the time history analysis performed for the reactor building given'n 3 ' '

I Both analyses yield water displacements, velocities, and impulsive and convective water pressures on the walls of the

-:. suppression chamber and the reactor building foundation mat.

'he first analysis, which considers the RPV pedestal excluded

~:.'rom the suppression

chamber, yields the maximum impulsive

~:. pressures.

The second analysis, which considers the RPV

~:'ncluded in the suppression

chamber, yields the maximum con-

vective pressures.

To obtain conservative values for the forces,'anding moments and overturning moments on the sup-pression chamber and foundation mat, the maximum impulsive forces from the first analysis and the maximum convective forces from the second analysis, are assumed to occur together.

3. 8-44a

'e

NNP-2 The following tabulation gives the analytical results obtained for the additional horizontal wall pressures due to SSE.

The-additional wall pressures are found to be negligible.

Distance below water Horizontal Wall Pressure due surface El. 466'-4 3/4" to SSE induced water sloshing (feat)

( si) 0 5

10 15 20 Below 20 0.30 1.56 2.57 3.30 3.74 5.84 l~;

The maximum vertical displacement (slosh height) and velocity

" of the oscillating water surface above the undisturbed equilibrium water surface elevation 466'-4 3/4" ka 9.5 inches

.Q at the, suppression chamber face.

This occurs in the second

" analysis which considers rectangular tanks with the RPV pedestal included.

s The period of water oscillation (time required for the water

. to oscillate one complete'ycle) is 6 seconds in the first analysis (circular tank) and 3.5 seconds in the second analysi (rectangular tanks).

, The analytical results used for horizontal pressures in the

~.: suppression chamber due to the Operating Basis Earthquake

.: are one-half of the values obtained for SSE.

j 3.8.2.4.4 Protective Coatings Protective coatings are applied to all exposed steel surfaces of the primary containment vessel.

Surfaces embedded in con-crete are not coated.

Coating systems used on the inside of the primary containment vessel are selected on the basis of

... their ability to withstand not only normal operating con-ditions but design basis accident conditions as well.

The

~

coating is able to withstand a

DBA without being removed from the surface, so 'that it will not interfere with emergency pumping and spraying systems.

The coating systems are sub-jected to tests designed to determine their radiation resist-s

ance, decontaminability, resistance to decontamination chemicals, and resistance to accident conditions.

The g pic g. i~em per eeez~g resPec<Vi~t).

3.8-44b

0

)'s'NP-2 TABLE OF CONTENTS (Continued) 3.8.2.4.1.4 Personnel Access Lock, and the (Combined)

Equipment Hatch and CRD Removal Hatch

3. 8. 2. 4. 2 Computer Programs Utilized in Design and Analysis

3. 8. 2. 4. 3 Seismic Analysis Pacae

3. 8-40 3.8-41 3.8-42 3.8.2.5.3 3.8.2 '.4 3.8.2. 6 3.8.2. 6. 1 3.8.2.6.2 3.8.2.6.3 3.8.2.7 3.8.2.7.1 Peak Stresses Buckling Criteria for the Primary Containment Vessel Materials, Quality Control and Special Construction Techniques Materials I

I Quality Control:

1 Special Construction Techniques Testing and Inservice Inspection Requirements Inspection of Material and Parts for Fabrication 3.8.2.4.3.1 Computer Program Utilized in the Seismic Dynamic Analysis 3.8.2.4.3.2 Seismic Dynamic Analysis of Nater in Suppression Chamber (Sloshing Effects)

I I 3.8.2.4.4 Protective Coatings 3.8.2.5 Structural Acceptance Criteria I

r,.

3.8.2.5.1, Stress Limits for Design Loading Conditions 3.8.2.5.2 Primary and Secondary Stresses 3.8-44 3.8-44 3.8-44b 3.8-45 3.8-45 3.8-46 3.8-46

3. 8-46 I

3. 8-46 3.8-46 3.8-52 3.8-55 3.8-55 3.8-55 3.8.2.7.2 3.8.2.7.3 Testing of Primary Containment Vessel During Field Erection Testing of the Erected Primary Containment Vessel 3.8>>55 3.8-55

,'3-xxvi

g.

22.021

6. 2.1 With regard to all safety-related equipment located inside the containment building such as the control rod drive hydraulic system, the reactor vessel supports and all incore instrumentation

leads, we require that the environment be maintained within the maximum temperatures and rate of temperature changes for. which the equipment is qualified to operate.

Indicate whether the reactor building ventilation system (RBVS) is required to assist in maintaining an acceptable temperature range.

If it is, provide the following information regarding the RBYS:

a.

0ustification for not treating this system as an engineered safety feature.

b.

The results of an analysis that the RBVS will not be a potential source of missiles; demonstrate that the RBVS meets our pipe whip criteria.

c.

A discussion of the operating procedures to be initiated in the event that the RBVS is unavailable.

d.

The location of all temperature sensors associated with the operation of the RBVS.

RESPONSE

As discussed in 6.2.3 and 9.4.2 the RBVS is the ventilation system for the secondary containment and is designed to be automatically shutdown and isolated on receipt of an accident signal.

Except for the isolation portion of the RBVS, the system is not an engineered safety feature. All equipment in the reactor building required to bring the reactor to a

safe shutdown is either qualified to the accident environment of the secondary containment or enclosed in compartments which are cooled by the reactor building emergency cooling systems (see 3.11.1.1.2, 3.11.4.2, and 9.4.9).

Seismic Class I supports are provided for all RBVS and emergency cooling system ducting to preclude damage to any safety related components in the secondary containment during a

seismic event.

The primary containment is cooled by the primary containment cooling system as discussed in section 9.4.11.

I) should be noted that this system is not required for safe shutdown',1s not assumed to be operable for accident analysis or assessment of the primary containment accident environment.

22.022 Provide a detailed justification for the use of guality Group C valves for containment isolation.

Our position on this matter is that guality Group B valves should be used.

RESPONSE

All containment isolation valves are either guality Group A or B.

Table 7.3.13, is confusing in that there is a "class" column which reflected outdated WNP-2 project related'terminology and does not pertain to code classification.

This table will be deleted from the FSAR and reference will be made to Table 6.2-16.

(Reference response to question 22.027)..

022.023 6.2.5 Identify the seismic category and the quality group of the hydrogen monitoring system.

RESPONSE

The hydrogen monitoring system is seismic category I.

The system is guality Group B for the suction lines to the downstream side of the excess flow check valves and for the discharge line from the upstream side of the check valve.

Between these code breaks, including the monitor, the guality Group is D.

Refer to'igure 3.2-15, paragraphs 6.2.5.1 and 6.2.5.2.3, and the response to question 022.8 in Appendix D.

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MASHINGTON PUBLIC PO'!ER SUPPLY SYSTEti NUCLEAR PROJECT ic!O.

2 FLO/! DIAGRA 1 - REACTOR BUILDtf<G PRIHARY CO)STAIN!lENT COOLING 8( PURGING SYSTEi~l FICURE l 3.2-15

22.024 6.2.5 Section

6. 2. 5.2.

1 of the FSAR states that all hydrogen mixing will be accomplished by a natural convection process.

Provide a detailed analysis that will support your non-mechanical mixing assumption.

RESPONSE

An assessment of the natural convection process as well as mechanical means of forced convection is in progress.

A complete response to this question will be filed prior to April, 1979.

22. 025 Identify the location of the hydrogen sampling points in the drywell and the suppression chamber.

Identify the location of the suction and discharge points of the combustion gas. control system with-respect to local structures and equipment.

RESPONSE

Please refer to FSAR Table 7.6-12 and Figures 6.2-32 through 6.2-35 for the requested information.+

0

The system processes the primary containment atmosphere using a blower.

The con"tant speed blower draws a minimum of 155 scfm from the containment.

The aas first enters the water scrubber, where particulate matter, droplets and soluble trace impurities are removed from the gas by direct.

continuous contact with water in a packed bed column.

The gas passes upward through the column and loaves the scrubber at: t;he column top through a demister

pad, which prevents entrained water from leaving with the gas.

The water, with particulates and dissolved solids, leaves the bottom of the scrubber and is directed to the suppression pool.

The gas then enters the blower and is compressed 10 psi to provide flow through the system and connecting piping."

The gas then enters the preheater, where it is heated to main-tain a the mostagically controlled recombiner inlet tempera-ture between 500 F and 550 F.

The heated and diluted gas enters the catalytic recombiner where the hydrogen and a stoichiometric amount of oxygen react on the catalyst bed to form water vapor.

The catalyst bed operates between 550 F and 1130 F and provides essen-O~

0 tially 100$ conversion efficiency.

inlet temperature greater than 500 F prevents degradation of the catalyst bed from halogens that are present in the feed gas.

The hot recombiner effluent gas is then cooled below 150 F in the aftercooler.

The condensate is separated in the moisture separator and is routed to the suppression pool.

A portion of the recombiner discharge is recycled to t:he blower suction to minimize blower horsego::e when contain-ment pressure is equal to 3 psig or greater.

During system operation, the containment at:mosphere is drawn from the drywell and the recombiner effluent gas is

~discharged to the suppression charker.

Each 'hydrogen recombiner is skid mounted into an integral package having maximum dimensions of 11 feet long by 9 feet wide and 9 feet high.

All pressure containing equipment including piping between components is considered an exten-sion of the containment and is classified Quality Group B

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6.2-74

PklP-2 7.6-1.13.8 Containment Hydrogen and Oxygen Atmosphere samples from three locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two completely redundant and separate analyzers.

For precise locations of each sample point see Table 7.6-12.

Each gas analyzer cabinet contains a hydrogen and an oxygen analyzer with sample conditioning and sample programming means.

The programmer also admits standardizing gases peri-odically to calibrate the analyzers.

Vent gases are pumped back to the primary containment at all times.

The analyzers are single range, i.e.,

0-10% hydrogen and 0-25% oxygen.

System accuracy is +

2% of full scale.

The output signal from each analyzer is sent to individual recorders in the main control room.

Each analyzer has two adjustable alarm contacts which annunciate abnormal condi-tions in the main control room.

The analyzers are quality group classification D.

7. 6. 1. 13. 9 Suppression Chamber Pressure Suppression chamber pressure is recorded in the main control room from two separate pressure transmitter systems.

Range of recording is from 0-100 psig with an accuracy of + 2.0%

of span.

7.6.1.14 'uppression Pool Temperature Monitoring System instrumentation and Controls 7.6.1.14.1 System identification The suppression pool temperature monitoring (SPTM) system is designed to monitor suppression pool water temperature and alert the plant operator to the potentially hazardous condi-tion of elevated pool water temperature with subsequent severe structure vibration.

The instrumentation for the SPTM system is shown in Figure 3.2-8.

The following is a description of the suppression pool tem-perature monitoring system.

The power supply for the sup-pression pool temperature monitoring system is from the 120 VAC instrument buses.

7..6-65

SP

.TABLE 7.6-12 CONTAINMENT HYDROGEVL AND OXYGEN MONITORING SYSTEM SAMPLE POINT LOCATIONS qgyn.P gc POi~~

SRmPLK POiUT PENETRATION 5

AZIMUTH ELEVATION 74 75 76 78 80 sa 72G 72d 72e 82c 85d 85e 73d 84b 188

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22.026 In accordance with Appendix J of 10CFRSO we require that containment isolation valves for those systems not vented and drained during Type A tests, are to be tested in accordance with Section III.C of Appendix J and those results are to be reported to the commission.

RESPONSE

In general the containment isolation valves for those systems not vented and drained during Type A tests are being Type C tested.

The exceptions to this are listed in the response to question 22.10 along with the justification.

The results of these tests will be reported to the Commission as stated in FSAR Section 6.2.6.4.

W.7 ~

0

22.027 6.2.6 Augment table 6.2-16 to provide the information requested in Section 6.2.4.2, "Systems Design," of Regulatory Guide 1.70.

RESPONSE

Tables 6.2-13, 6.2-16 and 7.3-13 have been combined into one table, number 6.2-16*.

This table has been further expanded to include all the information required in Section 6.2.4.2 of Regulatory Guide 1.70.

Ch. 6 will be appropriately revised to reflect the revised table 6.2.16.

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WNP 2 Page 13 of 23 TABLE bi2" /(o ISOLATION SIGNAL CODES FOR TABLE &.2- /6

~si nal B*

(~r:>c ~)

Reactor vessel low water level~-

(A scram,toccurs at this level also.

This is the higher of the Qf'ac.Pow water level signals.)

Reactor vessel low water level

(%Pic 27.

1 ~

Q*

K*

High radiation Main steam Line break - Main steamline (steamline high space temperature or high steam flow).-

High drywell pressure (core standby cooling systems are started)

Line break in cleanup system high space tempera-ture.

Line break in RCIC system line to turbine (high

RCXC, p>pe.)'space temperature, high steam flow,'r low steam line pressure).

Line break in RHR shutdown PIP<<+ (than

>><7to~ F~<<~)

Low main steamline pressure at inlet turbine (RUN mode only).

• These are the isolation functions of the primary contain-ment, and reactor vessel isolation system; other functions are given for information only.

'I jl l

N g

~

%P-2 Page 14 of 33 TABLE g, Z.- IG (Continued)

Siqnal S

U D~t Low drywell pressure High reactor vessel pressure f[

High temperature at outlet of cleanup system non-regenerative heat. exchanger Standby liquid control system actuated Reactor building ventilation exhaust plenum high radiation.

Remote manual switch located in main control room.

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T ese are t e isolatio'n functions of the primary contain-ment and reactor vessel isolation system; other functions are given for information only.

VlNP-2 Page 15 of 23 TABLE Q,Z - 8:(Continued)

ABBREVIATIONS/LEGEND Valve AO MO PC EHO SO Type air operated motor operated'ositive closing electro-hydraulic operated solenoid operated 2.

Location I

inside containment 0

outside containment Power to Open/Close AC AC electrical power DC DC electrical Power process flow

Process, pro PP process fluid overpressurization spr spring 4.

Normal Position 0

open C

close 5.

Process fluid N

water A

air S

steam plpMot-(c-pc.u> o 6.

Termination Zone TB RB Rad N

s8 turbine building reactor building radwaste building S fLViCC I50iCO)~<

I ~

j C

~

~ ~ p ifNP-2 Page 16 of 23 I

TABLE &7= C (Continued)

NOTES FOR TABLE These notes are keyed by number to correspond to number in parenthesis, in Table 7.3-13.

Type C testing is discussed in,+

Fig re.6.2-.31 which.shows t e i olation valve ar ngeme t. ~<'.

ygp ~6,m '7AM~ 2-I4 ~

~ C-1.

Main steam isolation valves require that bo h solenoid Bi&) 't pilots be de-energized to close valves.

Accumulator air pressure plus spring set act together to close valves i

when both pilots are de-energized.

Voltage failure at only one pilot does not cause. valve closure.

The valves are designed to fully close in less than 10 seconds.

2.

5iippression cooling valves have interlocks that allow them to be manually reopened after automatic closure.

This setup permits sApg~ss o~

Poc c.,

spray, for high drywell pressure conditions, and/or suppression water cooling.

When automatic signals are not present, these valves may be opened for test or operating convenience.

3.

Testable check valves are desiqned for remote opening with zero differential pressure across the valve seat.

The valves will close on reverse flow even though the test switches may be positioned for open.

The valves open when pump pressure exceeds reactor pressure even though the test switch may be positioned for close.

4 ~

ol rod hydraulic lines can be isolate solenoi tside the pr n ainment.

Lines that ezter ggP~D g~ ~~~-, Pz~.,~g~ ~f and terminate system tha

'esigned" to prevent out-lea Solenoid valves normal closed, but open on rod movement and during'reactor

.5.

Alternating current motor-operated valves required'for isolation functions are powered from the AC standby power buses.

Direct current operated isolation valves are powered from station batteries.

6.

All motor-operated isolation valves remain in the last position upon failure of valve power.

All air-operated valves close on motive air failure or in the safest position.

NNP-2 Page 17 of 23 TABLE 4. Z, -lC (Continued) 7.

The standard minimum closing rate is 12 inches per minute for gate valves and 4 inches'per minute for globe valves.

For example, a 12 inch gate valve will close in one minute.

8.

Reactor building ventilation exhaust plenum high radia-tion signal (Z) is generated by two trip units; this requires one unit at high trip or both units at down

-scale (instrument failure) trip, in order to initiate isolation.

9.

These are isolation signals of the PCRVIS.

Process con-trol signals'ontrolling containment isolation valves are not indicated here.

For process control signals, see the appropriate section pertaining to the subject system.

An example of this note is the RCIC pump mini-mum flow bypass line to the suppression pool (Flow Diag. N519, Pene.

No. 65).

The table indicates RM for isolation signal.

This valve (RCIC-V-19) receives a

process control signal as indicated in 7.4.1.1.3.3.

10.

Normal status position of valve (open or closed) is the position during normal power operation of the reactor (see Normal Position column).

The specified closure rates are as required for con-tainment isolation only.

12.

All isolation valves are Seismic I.

13.

Used to evaluate primary containment leakage which may bypass the secondary containment emergency filtration system.

See 6.2.3.2.

14.

Size indicated is containment side of relief valve when relief valve size is not equal on both sides.

15.

Leakage control system provided, see 6.7.

Qg)AN LC4%44<

OF ~~+~~+ ~~~i>~&i'5 Il/07 COAlsr04CCP gd+Qg.

pt,si~~

Basis L.6<A, ~Le 0, Z.S,Z..

I I<

Page 18 of 23 Table 6.2-16 (Continued) 17.

18.

19.

20.

Valve operability will be improved because the environmental conditions are better outside the primary containment from the standpoint of humidity, radiation, pressure and temperature transients, and post-LOCA pipe whip and jet impingement.

These lines connect to systems outside of the containment which meet the requirements for a closed system as set by NRC SRP 6.2.4,Section II, paragraph 3e.

These systems are considered an exten-sion of the primary containment.

Any leakage out of these systems will be processed by the standby pa~: treatment system.

Relief valve setpoint greater than 77.5 psig (1.5 times contain-ment design pressure).

Re 1 ief va1 ve setpo int is 75 ps ig.

21.

22.

Cannot be reshut after opening without disassembly.

See 6.2.4.3.2.2. T,.2 23.

See 6.2.4. 3. 2.2.2.

24.

25.

Due to redundancy within the emergency core cooling systems, some subsystems may be secured during the long term cooling period.

In addition RHR loops A and B have several discharge paths (LPCI, Drywell Spray, Suppression Chamber Spray, Suppression Pool Cooling) which the operator may select during the 30 day post-LOCA period.

Applicable portion of the flow diagrams 3.2-6, -7, -8, and -15 to be updated to reflect the configurations shown on Figures 6.2-31r and -3ls.

26.

27.

An air operator is provided on the check valve to enable the operator to positively close the check valve.

The check valve position indication will be utilized to determine need for operator action.

The air operator switch has an automatic return to neutral so the vacuum breaker function will not be impaired.

Instrument lines that penetrate primary containment conform to Regulatory Guide 1. 11.

The lines that connect to the reactor pressure boundary include a restricting orifice inside containment, are Seismic Category I and terminate in instruments that are Seismic Category I.

The instrument lines also include manual isolation valves and excess flow check valves or equivalent (see hydrogen monitor return lines).

These penetrations will not be type C tested since the integrity of the lines are continuously demonstrated during plant operations where subject to reactor operating pressure.

In addition all lines are subject to the type A test pressure on a regular interval.

Leaktight integrity is also verified with completion of functional and

I 0

27.

28.

Page 19 of 23 (continued) calibration surveillance activities as well as by visual inspection during daily operator patrols as applicable.

Penetrations X-76 and X-77 contain lines for the hydraulic control of the reactor recirculation flow control valve.

These lines contain corrosive hydraulic fluid used to position the reactor recirculation flow control valve.

These lines inside of the containment are Seismic Categ'ory 1

and guality Group B.

They are provided with failed closed~automatia isolation valves outside the containment which receive an auto-matic isolation signal on high drywell pressure.

These lines meet the requirement of General Design Criterion 57 and therefore require only single automatic isolation valves outside of the containment.

These lines also meet the require-ment of Standard Review Plan 6.2.4.

They are designed to Seismic Category 1,

Code Group B and the following criteria:

a.

do not communicate with either the reactor coolant system or the containment atmosphere, b.

are protected against missiles and pipe whip, c.

are designed to withstand temperatures at least equal to the containment design temperature, d.

are designed to withstand the external pressure from the containment structural acceptance

test, and e.

are designed to withstand the loss-of-coolant accident transient and environment.

Even if the failed closed valve were to not shut there will be no leakage of containment atmosphere through the hydraulic control lines since the piping inside the primary containment remains intact.

There are no active component failures which would compromise the integrity of the closed system ins!ide the primary containment.

Integrity of the closed system inside the primary containment is, essentially, constantly monitored since the system is under a

constant operating pressure of 1800 psig.

Any leakage through this system would be noticed because operation would be erratic and because of indications provided on the hydraulic control unit.

In additon, in order to perform Type C tests on these lines, the system would have to be disabled and drained of the corrosive hydraulic fluid.

This is considered to be detrimental to the proper operation fo the system in that possible damage could occur in establishing the test condition or restoring the system to normal.

These lines and associated isolation valves should therefore be considered to be exempt from Type C testing.

~

~

Paqe 20 of 23 Since the traversing incore probe (TIP) system lines do not com-municate freely with the containment atmosphere or the reactor coo ant, General Design Criteria 55 and 56 are not directly applicable to this specific class of lines.

The basis to which these lines are d

d s more closely described by General Design Criterion 54, esigne is mo ld which states in effect that isolation capability of a system shou be commensurate with the safety importance of that isolation.

Further-~

more, even though the failure of the TIP system lines presents no safety consideration, the TIP system has redundant isolation capa-bilities.

The safety features have been reviewed by the NRC for t

BWR/4 (Duane Arnold), BWR/5 (Nine Mile Point) and BWR/6 (GESSAR),

and it was concluded that the design of the containment isolation system meets the objectives and intent of the General Design Criteria.

Isolation is accomplished'by a seismically qualified solenoi.d-operated ball valve, which is normally closed.

To ensure isolation capabi ity, an explosive shear valve is installed in each line.

Upon receipt of a signal (manually initiated by the operator), this explosive valve will shear the -TIP cable and seal the guide tube.

When the TIP system cable is inserted, the ball valve of the selected tube opens automatically so that the probe and cable may advance.

A maximum of f'ice.%valves may be opened at any one time to conduct calibration, and any one guide tube is used, at most, a few hours

~

per year.

If closure of the line is required during calibration, a signal causes t'e cable to be retracted and the ball valve to close auto-matically after completion of cable withdrawal. If a TIP cable f ils to withdraw or a ball valve fails to close, the explosive shear valve is actuated.

The ball valve position is indicated in the control room.

The tu<P - g.

TIP system design specifications require that the maximum leakage rate of the ball and shear valves shall be in accordance with the Manufactureres Standardization Society (Hydrostatic Testing of Valves) ~

The ball valves are lOOX leak tested to the following criteria by the manufacturer:

Pressure Temperature Leak Rate A statistically chosen sample manufacturer to the following 0

62 psig 340 F

10 cm /s of'he shear valves is tested by the criteria:

Pressure Temperature Leak Rate 0 - 125 psig 3403 F 3 10 cm /sec STP The shear valves have explosive squibs and require testing to destruc-tion.

They cannot therefore be lOOX tested.

Page 21 of 23 29.

(continued)

As stated

above, the penetration is automatically closed following use.

During normal operation the penetration will be open approxi-mately eight hours per month to obtain TIP information.

If a failure occurred such as not being able to withdraw the TIP cable, the shear valve could be closed to isolate the penetrations.

Installation requirements are that the guide tube/penetration flang/ball and shear

~alve composite assemble not leak at a

rate greater than 10 std cc/sec at 80 psig.

Further leak testing of the shear valves is not recommended since destructive testing would be required.

Leak testing of the ball valves also is not recoranended since the guide tube terminates in a sealed indexer housing which is kept under-a positive pressure by a nitrogen or air purge.

The purge make up will be indicative of the system leakage.,

Note that the TIP ball valve is normally closed and thus is a part of the leakage barrier being monitored.

Consequently the personnel exposure required to conduct Type C tests from inside the con-tainment is not warranted.

Page 22 of 23 NOTE 4 The isolation provisions for the CRD lines are commensurate with the essential requirement that the control rods are inserted on a scram.

Isolation of the hydraulic lines is provided by check valves 115 and 138 and solenoid valves 120, 122, and 123 on the hydraulic control units (HCU) and by air operated valves F010, F011 on the scram discharge header (see Figures 4.6-5a and b).

The HCU's and scram discharge headers as well as the hydraulic lines themselves are Seismic I, and are qualified to

'he appropriate accident environment.

The failure and scram position of all power operated valves are compatible with system isolation and, at the same time, rod insertion on a scram.

Addition of power operated isolation valves on the hydraulic lines themselves could prevent control rod insertion.

fianual isolation valves 101 and 102 allow for further isolation if it becomes necessary.

The hydraulic lines are small and terminate in the reactor building which is served by the standby gas treatment system.

The hydraulic lines and their manual isolation valves in the scram discharge header and its air operated valves are code group B.

The hydraulic control'nit (HCU) is a General Electric factory-assembled engineered module of valves, tubingf piping, and stored water which controls a single control rod drive by the application of precisely timed sequences of pressures and flows to accomplish slow insertion. or withdrawal of the control rods for power control, and rapid insertion for reactor scram..

Although the hydraulic control unit, as a unit, is field installed and connected to process piping, many of its internal parts differ markedly from process piping com-ponents because of the more complex functions they must provide.

Thus, although the. codes and standards invoked by Groups A, B, C, and D pressure integrity quality levels clearly apply at all levels to the interfaces between the HCU and the connecting conventional piping components (e.g., pipe nipples, fittings, simple hand valves, etc.), it is consid-ered that they do not apply to the specialty parts (e.g solenoid valves, pneumatic components, and instruments).

The design and construction specifications for th'e HCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but these codes and standards are supplemented with additional requirements for these parts and for the remaining parts and details.

For example,

'1} all welds are penetrant tested (pT),

r,

Page 23 of 23

2) all socket welds are inspected for gaos between pipe and socket bottom, 3) all. welding. is.performed by qualified

welders, and

4) all work is done per written procedures Quality Group D is generally applicable because the codes and standards invoked by that group contain clauses which permit the use of manufacturer's standards and proven design techniques which are not explicitly defi~ed within the codes of Quali'y Group A, B, or C.

This is supplemented by the QC techniques.

The CRD lines will be included in the type A test leakage since the reactor pressure vessel and the nonseismic portions of the CRD system are vented during the performance of the type A test.

The CRD insert and withdraw lines are compatible with the criteria intended by 10CFR50, Appendix J, for Type C testing, since the acceptance criterion for type C testing allows demonstration of fluid leakage rates by associated'ases.

c~

'F

022.028 Several of the loads presented in Table 3.4-1 of the Plant-Design Assessment Report (DAR) have been generated using computer codes which have not been reviewed by the HRC staff.

Provide a comp'lete description of your method of analysis for all codes pr esented in Appendix D of the OAR.

RESPONSE

Three computer codes have been used to develop short term LOCA hydrodynamic loads for NNP-2 plant assessment.

The three codes are the downcomer vent clearing analytical model computer code VENT, the pool swell analytical model computer code SWELL, and the LOCA bubble charging analytical model computer code BUBBLE.

Complete documentation on each of the above three codes is provided or refer-enced. in Appendix D of the DAR (see below for specific references).

Documentation on the WNP-2 load calculation procedure is provided in 3.2.1 of the DAR.

l.

,Vent Code a) Assumptions See Section D; 2 of the DAR.

b) Equations See Fig. D-l of the DAR.

c) Methodology See Fig., 0-1 of the DAR.

2.

Swell Code The'ssumptions, equations, and methodology for the swell code are identical to that described in reference D-l.

3.

Bubble Code The assumptions, equations, and methodology for the Bubble Code are identical to that described in reference D-T.

II

~k Jf V

WNP-2 rr/~

Q 022.029 Safety issues such as the proposed reductions in the pool boundary chugging loads and the safety/relief valve quenchers loads are being resolved generically and are not scheduled for resolution until about 1980.

Discuss your short term solutions for these types of problems which require a long period of time to be resolved.

~Res onse Two important suppression pool hydrodynamic loads that were described in DAR revision 1 as being under study to improve the load definitions are chugging loads on the pool boundary and quencher sazety relief valve air clearing loads on the pool boundary.

The definition of the chugging pool boundary load is based on data obtained by General Electric in the 4T test facility, which is representative of a full scale single cell in a Hark ZZ geometry.

Znfluence of certain test facility para-meters has been evident in the pressure measurements made during the test.

Zdentification of these parameters has been the subject of an analysis effort by Burns and Roe with the goal being to define a chugging forcing zunction, independent of the test facility, i.e., that could be applied in a Hark ZZ suppression pool at vent exits.

A computat'onal methodology and the associated computer code, has been developed by Burns and Roe; it al'ows application oz this-forcing function at.

downcomer vent exits in a Hark ZZ suppression pool and cal-culation of pool boundary loads and bui3.ding responses properly accounting for fluid/structure interaction effects.

The BGR computational methodology wi3.3. be discussed with NRC in early November.

The documentation required in support of the chugging forcing function definition an'd the computational methodology is scheduled for submittal for NRC review in the WNP-2 Design Assessment Report, revision 2.

The current safety relief valve quencher load definition in the DFPR is very conservative and results in high building response loads for WNP-2.

WPPSS is currently involved in a program that will utilize existing test data and the data from the ongoing Caorso safety relief val've tests to develop a

more realistic SRV quencher load definition.

As discussed in the NRC-WPPSS meeting of October 10, 1978, a report covering SRV load definition for WNP-2 is scheduled to be ava'lable for NRC review in Hay, 1979.

0

3'KP-2 Q 022.030 r

Provide a more detailed description of the significant modi-fications to the WNP-2 facility which are being made or scheduled to be made based upon the results of your, ongoing experimental and analytical efforts.

~Res onse Modifications made to the NNP-2 plant as a result of investi-gations of the SRV and LOCA phenomenon are listed below:

1)

Seven horizontal ring tee stiffeners added to the submerged circumference of the steel containment vessel.

2)

Redesign of the downcomer bracing system from a system of radial beams to a pipe truss system which includes braces for the suppression pool columns and lateral restraints for the SRV piping.

3)

Revised location of plat orms in the suppression pool and revision of their connection to the containment vessel.

4)

Revised location of wetwell to drywell vacuum breakers.

5)

Provided a auencher discharge device and support tower for each SRV line.

Added a redundant vacuum breaker on each SRV discharge piping line.

SRV piping was re outed to optimize the line air volumes.

6)

Provided additional stiffening in the area of many containment vessel piping penetrations.

7)

Redesign of piping systems resulting in additional pipe supports and snubbers.

8) 'owncomer pipes were locally reinforced where the SRV pipe penetrates the downcomer wall and the downcomer flanges were removed.

9)

Installation of a suppression oool temperature monitoring system.

10)

Revised the design of piping suction st ainers.

~

)'

RESPONSES TO CORE PERFORMANCE BRANCH QUESTIONS*

• per letter, NRC to HPPSS, Sept.

18, 1978.

r II

~

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,c MNP-2 5( i 7r 10/13/78 Revision 0

QUESTION 231.001 (4.2)

Section 4.2 of the MNP-2 FSAR references the General Electric Topical Report, NED0-20944, as the sole input for fuel design.

During our review of this GE topical, another topical report "BMR/4 and BMR/5 Fuel Design, Amendment 1,"

NEDO-20944-1P dated January

1977, was submitted.

It is our position that this report is applicable to the MNP-2 facility.

Accordingly, revise your FSAR to include both G.E. topical reports.

RESPONSE

NEDE-20944-1P (Amendment One to NEDE-20944-P) is applicable to MNP-2 and ah~d c~///

be referenced.

Q 231.001

~

H

WNP-2 4.2 FUEL SYSTEM DESIGN Information covering the following subjects in 4.2 are provided in topical report NEDO-20944*.

Proprietary infor-mation is contained in NEDE-20944-P*<

WNP-2 is a BWR 5 with 251 inch vessel, 764 fuel assemblies, loaded on C lattice.

Topical report paragraph, table and figure numbers are con-sistent with FSAR numbers, except that the initial digit (4) has been suppressed in the topical numbering.

4.2 FUEL SYSTEM DESIGN 4.2.1 General and Detailed Design Bases 4.2.1.1 General Design Bases 4.2.1.2 Detailed Design Bases

, 4.2.2 General Design Description 4.2.2.1 Core Cell

4. 2. 2. 2 Fuel Assembly 4.2.2.3 Fuel Bundle

4. 2. 2. 4 Reactivity Control Assembly 4.2.3 Design Evaluations 4.2.3.3.

Results of Fuel Rod Thermal-Mechanical Evaluations 4.2.3.2 Results from Fuel Design Evaluations 4.2.3.3 Reactivity Control Assembly Evaluation (Control Rods) 4.2.4 Testing and Inspection 4.2.4.1

Fuel, Hardware and P <<sembly 4.2.4.2 Testing and Inspection (Enrichment and Burnable Poison Concentrations) 4.2.4.3 Surveillance Inspection and Testing of Irradiated Fuel Rods 4.2.5 Operating and Developmental Experience 4.2.6 References NED0-20944, "BWR 4 and BWR 5 Fuel Design," October

1976, NEDE-20944P Proprietary Version

(~g(~ gfjy g fbi' ZO)+Q

) P (Q~AdA)clif 1 'fo JETE -'20 j4lf p)-

J 4 2-1

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QUESTION 231.002 (4.2)

WNP-2

('c ~/

-/,y- '(/

10/13/78 Revision 0

The NRC staff is concerned with the validity of fission product gas release calculations in most fluel performance

codes, including GEGAP-III, for a fuel burnup greater than 20,000 MWd/tU.

General Electric was informed of this con-cern on November 23,

1976, and was provided with a method of correcting fission product gas release calculati.ons for fuel burnups greater than 20,000 MWd/tU.

Since there was no question of the adequacy of GEGAP-III for fuel burnups below 20,000 MWd/tU, your calculations are acceptable only for that time in reactor core li,fe when the peak local burnup is less than 20,000 MWd/tU.

For fuel

. burnaps in excess of this specific value, GEGAP-III calculations and all other affected analyses, must be redone using the correction cited above.

Alternatively, you may submit a modified method which addresses the staff's concerns.

RESPONSE

)

NRC concern regarding the validity of fission gas release calculations for burnup greater than 20,000 MWd/tU was transmitted to General Electric in Ref-erence 1.

Reference 1.requested an analysis to describe the impact of higher fission gas release for G.E. operating power reactors (BWR 2-4 product line).

Reference 1 did not indicate that such analyses would be necessary for licensing support for operating plants or that the NRC would require the application of the fission gas release correction factor in future analyses.

Hence, the use of the correction factor is not part of the design-basis analysis.

Reference 2 provided G.E.!s response to the NRC request.

The NRC fission gas release correction was employed to modify the GEGAP(3) thermal performance code.

The modified GEGAP code was then employed to calculate the following parameters as a function of exposure for 7x7 and 8x8 fuel:

(1)

(2)

(3)

(4)

(5)

Percent of fission gas released Fuel rod internal pressure Pellet-to-cladding gap conductance at the peak power axial position Fuel centerline temperature at the peak power axial position Fuel volume average temperature at the peak power axial position Q 231.002

ii

These parameters have been compared with results of the standard (unmodified)

GEGAP code in Reference 2.

The only affected safety analyses, as indicated in Reference 2, were the loss-of-coolant analyses.

Although the calculations were not specifically performed

~j for the WNP-2 fuel, th>8x8 analysis performed for early reflooding plants will bound the MNP-2 case.

Consequently, based on the results indicated in Reference 2, the NRC fission gas release correction results in less than an 85 F increase in calculated peak cladding temperature at a target planar average exposure of 30,000 MWdlt (The WP-2 initial core is not expected to exceed 20,000 MWdlt).

References:

1.

Ross, Denwood F., letter to Dr. Glen Sherwood, November 23, 1976 2.

Sherwood, G. G., letter to Denwood F.

Ross, December 22, 1976 3.

"GEGAP III:

A Model for the Prediction of Pellet-Cladding Thermal Conductance in BMR Fuel Rods," NED0-20181, November 1973 Q 231.002

4~

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7 T ~

QUESTION 232. 001

( 15. 4. 1 )

13 10/13/78 Revision 0

General Electric has performed a generic analysis of the consequences of the continuous withdrawal of an out-of-sequence control rod during reactor startup.-

This analysis has been documented on the Hatch-2 docket (Docket No. 50-366).

Adopt this analysis either by reference or submit it in its entirety on your docket.

RESPONSE

The detailed analysis of the consequences of a RME in the startup range is provided in NEDM-23842, "Continuous Control Rod Withdrawal Transient in the Startup Range," April 18, 1978 by R. C. Stirn and J.

F. Klapproth.

Q 232.001

RESPONSES TO GEOSCIENCES BRANCH QUESTIONS*

• per letter, NRC to WPPSS, Sept.

18, 1978.

(1) one copy of the references is attached to the original of question 362.3

t I

(

i

360.0 362.0 Q. 362.1 GEOSCIENCES BRANCH Geotechnical En ineerin Section (2.5.4.5.3) and (2.5.H2.3)

Provide summaries of field test results which support the statement on page 2.5-137 of the FSAR that "... the relative density values were within the specified limits..."

RESPONSE

Summaries and an evaluation of the compacted fillplaced for support of all Seismic Category I Structures, except the Condensate Storage

Tanks, are provided in Reference 2.5-127

("Soil Compaction Evaluation of. Quality Class I Backfill, Washington Public Power Supply System Nuclear Project No.

2, Benton County, Washington" by Shannon 6 Wilson, Inc.,

May ll, 1976).

Relative density test summaries are presented on Figures B-2, C-3, C-4, C-5, and C-6 of Reference 2.5-127.

An evaluation and summary of relative density tests for the fillplaced to support the Condensate Storage Tanks is provided by Tables A, B, C, and D in Reference 2.5-127A

("Soil Compaction Evaluation Quality Class I Backfill Con-densate Storage Tanks Area, WPPSS Nuclear Project No.

2, Benton County, Washington",

by Shannon

& Wilson, Inc.,

February 15, 1977).

In both references, the evaluations and summaries demonstrate that the degree of frill compaction was within the specification requirements for relative density and is adequate for support of all Seismic Category I structures.

References 2.5-126,

127, and 127A are attached in response to Question 362.3.

I

g.

362.2 (2.5.4.5.2) and (2.5H.2.1)

On page 2.5H.2, third paragraph, line 10, revise your requirement for void ratio to read "A void ratio of less than 0.25 was specified"~.

Provide justification for this upper limit, including its relationship to the potential for liquefaction and/or settlement~.

In particular, demonstrate that the top of the Ringold gravel was acceptably dense prior to placement of construction fill.

Describe the excavation testing procedures recommended by the Foundation Engineer in Section 4.1la of Appendix 2.5F of the FSAR".

RESPONSE

On page 2.5H.2, third paragraph, line 10 should read:

"A void ratio equal to or less than 0.25 was specified." +

The upper void ratio limit of 0.25 was selected because it represents a very dense soil (Glacial till, very mixed-grained),

as listed on "Soil Mechanics in Engineering Practice" by Terzaghi and Peck, July 1955 (reproduced in Table 3, Appendix A, reference 2.5-127).

The data presented on Figure 2.5H-2 of the NNP-2 FSAR demonstrates that Ringold gravels do not fall within the gradation range of soils susceptible to liquefaction.

Also, very dense granular soils are not subject to liquefaction and/or excessive settlement.

All density tests to determine void ratio'were taken near the surface of the proof-rolled Ringold gravel after com-pleting six coverages with the specified roller.

All void ratio determinations met the specification requirements, thereby demonstrating that the surface of the Ringold gravel was acceptably dense prior to placement of construction fill.

"As the general excavation in the central plant area approached final grade, close observation by the engineer was required to verify that the very dense Ringold gravel had been reached.

This was accomplished as follows:

a)

The excavation was extended to an elevation just lower than those at which the very dense Ringold gravel had been encountered in the previous test borings.

Between borings, similar gravel was ex-posed.

b)

The very dense Ringold gravel exhibits a distinct tan or light gray-brown color.

In contrast, the

0 0

,H

overlying sands and gravels are gray or black.

Xn all

cases, excavation was continued until the distinct tan or light gray-brown soil was reached, which further dem'-

onstrated that the very dense Ringold gravel had been encountered.

c)

Due to the very dense nature of the Ringold gravel compared to the overlying sand and gravel, there was a noticeable and.distinct increase in the degree of difficulty required for excavation.

Also the move-ment of the construction equipment over the surface of the exposed Ringold gravel would produce a very dense stable surface compared to the overlying soils which would exhibit a loose and unstable condition.

Probing into the very dense Ringold gravel with a pointed steel rod reached refusal within a few inches.

further demonstrating that the very dense gravel zone had been reached.

d)

Testing to determine the in situ density and corres-ponding void ratio was performed in accordance with specification requirements to further demonstrate that the very dense Ringold gravel had been encountered.

Dewatering during excavation in this area was accomplished with a perforated 55-gallon oil drum installed in the deepest portion of the excavation for a pump sump.

Pumping continued during excavation in order to ob crve and determine that the excavation extended to the Hingold gravel and permit the removal of all loose and medium dense sandy oils and/or disturbed Bingold gravel.

Pumping continued after the excavation was complete and the area was backfilled except for the small area in the vicinity of the sump.

The pump and perforated oil drum were subse-quently removed and the small area backfilled quickly in order to keep placement ahead of the rising groundwater.

Lift thickness and compaction were performed in accordance with specification requirements, but coarser than average sandy material was used to provide maximum stability.

Back-f'lling in this area was carried at least two to three feet above the static groundwater level before this phase of back-filling was stopped.

The very dense Ringold gravel encountered in the bottom of the excavation was identified by its gradation and color.

tVhcreas the upper loose to medium dense sand is dark gray to black and composed mainly of basalt grains, the Ringold gravel is tan to brown, grades to 3-inch maximum size and contains quartzitic sand.

Typical gradations of 1?ingold

gravel, based on samples retrieved from the base of the cen-tral plant excavation, are shown on 1'igur e 2. SH-2.

In

addition, the specifications required identification of the Piqgqld grpvel also bc based on void ratio.

A void ratio

~-"~

presents results of the 12 void ratio measuremen"s taken in the base of the 7/NP-2 central plant excavation.

As shown, the maximum vo'd ratio, e, measured was 0.20, and the aver-age was 0.17.

2.5H.2.1.1 Geologic Napping of Excavation Slopes Excavation in the central plant area was accomplished in two

phases, as shown on Pigures 2.511-3 and 2.5H-4.

Detailed geologic profiles of the slopes developed from the mapping are presented on I'igures 2.SH-5 and 2.5H-6.

Geologic mapping of the excavation slopes was performed by geologists from Shannon and Wilson, Inc.

The slopes of the initial excavation wore mapped interm'ttently during the period of November 27, 1972 through January, 1973.

The slopes of the fi>>al excavation were mapped during the period from May 7, 1973 through May ll, 1973.

2.5H-2

l

Q.

362. 3 (2.5. 8)

Provide references 2.5-126 and 2.5-127.

RESPONSE

Reference 2.5-126 "Letter Report, Soil-Structure Inter-

action, Washington Public Power Supply System Hanford No.

2 Nuclear Station Richland, Washington" by E. D'Appolonia Consulting Engineers, Inc., August, 1972, is enclosed.

Reference 2.5-127 "Soil Compaction Evaluation of Quality Class I Backfill, Washington Public Power Supply System Nuclear Project No.

2 Benton County, Washington" by Shannon and Wilson, Inc.,

May 11, 1976 is also enclosed.

Reference 2.5-127A "Soil Compaction Evaluation Quality Class I Back-fillCondensate Storage Tanks Area WPPSS Nuclear Project No.

2 Benton County, Washington" by Shannon and Wilson, Inc., February 15, 1977, is also included and is hereby incorporated as a separate reference.

Q.

362.4 (2.5.4.10)

Describe the methods used to calculate the dynamic coefficient (KD) of 0.3 shown in Figure 2.5-69 of the FSAR.

Xndicate how the effects of compaction are included in the calcu-lations of lateral earth pressures.

RESPONSE

The dynamic coefficient (KD) was calculated from the Mon-onabe-Okabe analysis (Seed and Whitman, 1970).

Additional items considered in our calculation included:

1)

The assumption that the walls of the Seismic Category I structures would be rigid with resulting at-rest pressures.

2)

An assessment of model test results performed by Japanese investigators and summarized in Seed and

Whitman, 1970.

3)

A review of coefficients required by building codes around the world as summarized in Seed and Whitman, 1970.

The distribution of lateral earth pressure included on Figure 2.5-69, and discussed in Appendix 2.5F was based on experimental data presented in Seed and Whitman,

1970, and

Aggour, 1972.

The effects of compaction on the static and dynamic earth pressures are somewhat offsetting.

Heavy compactive effort is expected to increase the static pressure but would result in a decreased dynamic pressure (see Figure 13 and 19 in Seed and Whitman, 1970).

On the other hand, lesser com-pactive effort would result in lower static pressure, and potentially a higher pressure under dynamic loading.

There-fore, the static and dynamic coefficients presented in Figure 2.5-69 allow for any effects of compaction since they would take into account both the case of a less compacted material with a potentially higher pressure under dynamic loading or a denser compacted material with a corresponding lower pressure under dynamic loading.

References

Seed, H. B., and Whitman, R. V., 1970, "Design of Earth Retaining Structures for Dynamic Loads", Specialty Con-

ference, Lateral Stress in the Ground and Design of Earth-Retaining Structures, Soil Mechanics and Foundations Division, ASCE.

Aggour, M. S. l972, "Retaining Walls in Seismic Areas",

Dissertation submitted in partial fulfillment of the requirements for the degree of Doctor of Philosophy, University of Washington,

Seattle, Washington.