ML17275B230

Forwards Responses to Containment Sys Branch 810316 Request for Addl Info.Responses Will Be Incorporated Into FSAR Amend within 4 Months.Remainder of Responses Re Hydrogen Recombiner Will Be Submitted to NRC by 810918

ML17275B230
Person / Time
Site:	Columbia
Issue date:	09/04/1981
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
References
G2-81-269, GO2-81-269, NS-L-02-81-CDT, NS-L-2-81-CDT, NUDOCS 8109110415
Download: ML17275B230 (72)

Text

REGULA'TO NFORMATION DISTRIBUTION TEil (RIDS)

ACCESSION'BR:8109110415 DOC ~ DATE': 81/09/04 NOTARIZED:'" NO DOCKKfl' FACIL~:50.-307 WPPSS Nuclear Pr BYNAME'UTHOR ojectr Unit'r. Wash'ington Public" Powe> 050.00397 AUTH. AFFILIATION BOUCHEY<G.D, Washington Public Power Supply System RECIP,NAME<<RECIPIENT AFFILtIATION SCHWKNCKRgA ~ LHcensing, Branch 2 SUBJECT!'orwards" responses to Containment Sys Branch 810316 request for addi info. Responses will be incorporated into FSAR~ amend within>> 4 months;Remainder of responses re hydrogen recombiner wil.l be submitted to NRC by 810918 CODEi: B001S COPIES RECEiI VED: LR'R ENCL>> 'ISTRIBUTION PSAR/FSAR AMDTS and Related Correspondence SIZE'0-'ITLEt:

NOTES:2 copies al,l matl:PM.1 cy:BWR I.RG PM(H Faulkner). 05000397 RECIPIENT COPIES RECIPIENT COPXEG I O'ODE/NAMEI LTTR ENCL~ ID LTTR>> ENCL>>

IG.T'Iai>>: A/D L'ICENSNG LIC BR 42 LA 1

1 0

0 CODE/NAMEIC BR A2 BC AULUCKgR ~ '4 1 1

0 1

INTERNAL>> ACC ID EVAI BR26'HEM 1 1 AUX SYS BR 27 1 1 E>>lG BR 11 1 1 CONT SYS BR 09 1 1 CORE PKRF, BR 10, 1 1 EFF TR SYS BR12>> 1 1 EQUIP QUAL BR43i 3 3 FEMA REP DIV 39 1 GEOSCIENCKS 28 2 2 HUM FACT ENG 40 1 1 HYD/GEO- BR 30. 2' ILC SYS'R 16 1 1<<

ILK< 06" 3 3 IE/KPDB 35 1 IE/EPLB'6" 3 LCC GUID BR 33 1- 1.

LIC- QUAI BR 32! 1 1 MATL>> ENG BR 17 1 MECH ENG BR 18, 1 1 MPA 1 0 OELD 1 0 OP LIC BR 34 1 PO'"ER SYS BR 19 1 1 PROC/TST'EV 20 1 1 QA BR 21>> 1 1 RA SS BR22'1 1 1 REAC SYS BR 23>> 1 1 1 1 SITti ANAL BR 24 1 1 T ENG BR25 1 1 EXTERNAL>>.'CRS 41>> 16 16 LPDR 03 1 1 NRC PDR>> 0 2"., 1 1 NSIC 05 1 1.

NTIS 1 1 gp 'i '( NS)

AL NURBER OF COPIES REQUIRED:: LiTTR ~ ENCL'

4 )

II I

~

Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 September 4, 1981 G02-81-269 Mr. A. Schwencer, Chief NS-L-02-81-CDT-053 Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission Washington D.C. 20555

Subject:

SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 RESPONSES TO CONTAINMENT SYSTEMS BRANCH QUESTIONS

Reference:

Letter, RL Tedesco to RL Ferguson, "Request for Addi-tional Information Regarding the WNP-2 Facility (CSB)"

dated March 16,'.1981.

Enclosed are sixty (60) copies of responses to the Containment Systems Branch questions transmitted to the Supply System by the referenced letter. These responses will be incorporated into the FSAR in an amendment within four months. The remaining four hydrogen recombiner responses will be transmitted to the Nuclear Regulatory Commission by September 18, 1981.

Very truly yours, G. D. Bouchey Director, Nuclear Safety GDB/CDT/ldm Enclosure cc: WS Chin BPA AD Toth - NRC RO NS Reynolds - Debevoise & Liberman J Plunkett - NUS Corporation R Auluck - NRC NY OK Earle B&R, RO EF Beckett - NPI WNP-2 Files 8109110415 810904 PDR ADDCK 05000397 A PDR

MNP-2 211.006 The NRC changed this question number to 010.049

'WNP-2 Q. 211.107 The NRC changed this question number to 010.037

WNP-2 Q. 211.108 The NRC changed this question number to 010.038.

C C WNP-2 Q- 211.109 The NRC changed this questi'on number to 010.039-

WNP-2 Q. 211.111

.(5.2.2)

Arti.cle NB-7200'verpressure Protections of the ASNE Boiler and Pressure VesseL CodesSection III'equires that an overpressure protection report'e provided. No overpressure report could be found in the FSAR. Pro-vide this report.

Response

Five copies of the Overpressure Protection Report are submitted via separate transmittaL. The report is'ir-tualLy reporduced verbatim in Section 5.2.2 of the FSAR.

The response to Question 211.049 addresses the commit.-.,

ment to update this report to conform to a more recent anaLyticaL model (ODYN code) and to account for recir-cuLation pump t rip. WPPSS committed to reperform. the applicable Limiting transients in the responser and to update the FSAR.

WNP-2 Q. 211 .114 (5.2)

Subsection'5 ~ 2.2.4.1 of the FSAR states that each Safety/

Relief Valve is provided with a device to counteract the ef fects of, backpressure which results, in the discharge Line when the valve is open and discharging steam. What type of device is provided? Describe the device and what effects would be anticipated if the device were to

'aiL.

Response

There is not a singular backpressure balancing devices but there is an integrated feature in each Safety/Relief Valve to counteract the effects of backpressure when the valve is open backand discharging steam. To prevent this backpressur'e from affecting the vaLve's spring Lift set pointi each valve has a bellows and a balancing piston.

The beLLows isolates steam in the valve discharge chamber from the valves's internalsi and prevents dis-charging steam from affecting the valve's set point.

If- the bellows failsi the balancing piston serves as a f'unctional "up by presenting an ef f ective piston area to the back pr essure equal to the valve seat area.

This reduces the acting spring Load on the disc insert by the amount of back pressure Load additive to the spring set pressure Load acting on the disc holders thus balancing (neutralizing) it so that there is no net back pressure effect on the set (popping) point.

FSAR Figure 5.2-10 has been revised to show the bellows and balancing pistons and the FSAR text in Subsection 5.2.2.4.1 has been revised to include a reference to the bellows and balancing piston arrangement.*

• Draft FSAR page changes attached.

g g

~ E

tiI P/pc< w'tel 88~ Ciyaaa AtihcHBD )

LEVER I

iiP Q AIR CYLINDER SET PRESSURE SPR ING(S]

I I I I I 0

BACK PRESSURE BALANCING DEVICE STEAM DISCHARGE DISC I III I

(/ STEAM INI.ET WASHINGTON PUBLIC POWER SUPPLY SYSTEM SCHENTIC OF SAFETY VALVE WITH AUXILIARY FIGURE NUCLEAR PROJECT NO. 2 ACTUATING DEVICE. 5.2-10

I IFTIHG HE C Hh DEISM LEvER SET PRESSURE AEIJVSTiHG eOLT THRVSTIICP RltIG

• IIOhO~tTF R, VIPER SPRltlG&PPNEQ BONNET VgNP f'ISTOw.TYPE f9E'UtIQTIC LOPE EP, SWING HhSHE R, ACTUhTOR

'EHSL'f CALAHCItIG PISfbH

~pl t 4EL~

~ci Q IILCAVOOWHPPJVST IIIG RltlG

$ 4)I.'EtIOIO,PIIO QIR cONTR01 OISG Vhi.VF, hSSEtI8L'f DISGHh+G E, IIOZZLf HOYLE 600'f RitI6 (I ~

~

I

~ ~ I~

4A WNP-2 u ('-

A schematic of the main safety/relief valve is shown in Figure 5.2-10.

operation:

It is opened by blither of two modes of a ~ The spring mode of operation which cons'ists of direct action of the steam pressure against a spring-loaded disk that will pop open when the "

valve inlet pressure force exceeds the spring force. Figure 5.2-9 diagrams the valve vs. time characteristic.

lift

b. The power actuated mode of operation which consists of using an auxiliary actuating de-vice consisting of a pneumatic piston/cylinder and mechanical linkage assembly which opens the valve by overcoming the spring force, even with valve inlet pressure equal to zero psig.

The pneumatic operator is so arranged that it will not prevent the valve disk from lifting steam if itifmalfunctions inlet pressure reaches the spring lift set pressure.

For overpressure safety/relief valve operation (self-actuated or spring lift mode), the spring load establishes the safety valve opening setpoint pressure and is set to open at set-points designated in Table 5.2-2. In accordance with the 5

ASME Code, full lift in this mode of operation is attained at a pressure no greater than 3% above the setpoint.

To prevent. backpressure from affecting S~a lift~ ~~u' p e spring a

set-td Vav~

point, each valve is provided with to counteract the effects of backpressure which results in the discharge line when the valve is open and discharging steam. g--

The safety function of the safety/relief valve is a backup to the relief function described below. The spring-loaded valves are designed and constructed in accordance with ASME III, NB 7640 as safety valves with auxiliary actuating devices.

For overpressure relief valve operation (power actuated mode),

each valve is provided with a pressure sensing device which operates at the setpoints designated in Table 5.2-2. When the set pressure is reached, it 'operates a solenoid air valve which in turn actuates the pneumatic piston/cylinder and linkage aaaembly .to open the valve.

( 4 44 g Ctk~

+)yggaIlj g I 4

g ~a 1C.

pg ~ ~~ .C.).M444 g+ Q ~~ Q ~CO h

WNP-2 Q. 211 .121 (5.2.5)

Subsection 5.2.5.5.5 of the FSAR states that the Leak detection system wilL satisfactori ly detect unidentified Leakage of 5 gpm. Subsection 7.6.2.4.2.1.2 states that the sensitivity and response time for each portion of the Leak detection system for detection of unidentified leakage is one gallon per minute in Less than one hour (excluding airborne systems) . Resolve this inconsistency.

Response

The infor'mat ion for sensitivity and response time of the Leak detection system is now contained in subsection 7.6.2.4.bi rather than subsection 7.6.2.4.2.1.2.

The 5 gpm discussed in. Section 5.2.5.5.5 refers to the maximum expected instantaneous Leakge rate. The second paragraph on page 5 '-46 of subsection 5.2.5.5.5 has been corrected to. read as follows-*

The Leak detection system sensitivity and response t.ime is such that an unidentified Leakage rate increase of one gpm in Less than one hour will be detected.

• Draft FSAR page change attached.

4 I 0

WNP-2 N 13 1

The unide'ntified leakage rate limit is based, with an adequate margin for contingencies, on t he crack size large enough to propagate rapidly. The establi,".hed limit is sufficiently low so that, even if the entire unidentified leakage rate were coming from a single crack in the nuclear system process barrier, corrective action could be taken before the integrity of the barrier would be threatened with significant compromise. +"~i a<

p;y,'$~ ~itd resp~ae 5'me iS Suc,h The leak detection systems uniden-tifiedAovleakage r'VI'/I rak4  !~crease a5 one gp~ y4 less %a<

One bc a fa,~Had >

Sensitivity, including sensitivity testing and response time of the leak detection system, and the criteria for shutdown if leakage limits are exceeded, are covered in Table 7.6-7.

5.2.5.6 Differentiation Between Identified and Unidentified Leaks: 'I Section 5.2.5.1 describes'-the systems that are monitored by the leak detection system; The ability of the leak detection system to differentiate between identified and unidentified leakage is discussed in 5.2.5.1 and 7.6.1.3.

5.2.5.7 Sensitivity and Operability Tests Testability of the leakage detection system is contained in 7.6.

5.2.5.8 Safety Interfaces The Balance of Plant-GE Nuclear Steam Supply System safety

'interfaces for the leak detection system are the signals from the monitored balance of plant equipment and systems which are part .of the nuclear system process barrier,, and associated wiring and cable lying outside the Nuclear Steam Supply Equipment. These balance-of-plant systems and equipment include the main steam line tunnel, the safety/relief valves, and the turbine building sumps.

5.2.5.9 Testing and Calibration Prov'isions for Testing and Calibration of the leak detection system are covered in Chapter 14, "Initial Tests and Operation".

5.2-46

WNP-2 Q. 211.122 The NRC changed this question number to 010.050.

WNP-2 Q. 211.123 The NRC changed this question number to 010.051.

WNP-2 Q. 211.125 The NRC changed this question number to 010.052.

WNP-2 Q. 211 .126 The NRC changed this question number to 010.053

WNP-2 Q. 211.130 The NRC changed this question number to 010.044.

WNP-2 Q. 211 .1 31 The NRC changed this question number to 010.045.

WNP-2 Q 211.133 The NRC changed thi s question number to 010.046.

MNP-2 R. 211 .134 The NRC changed this question number to 010.047.

WNP-2 Q 211.135 The NRC changed this question number to 010.048.

WNP-2 Q 211.138 (4 6.4.1)

Provide the common mode failure probability value for the control rod drive and the standby Liquid control systems.

Response.

'A Fault Tree Analysis was completed for both of these systems'nd the calculated unreliabiLtiy is Less than 10 "/reactor year. This unreliabiltiy is an estimate of 'the failure* to fully insert the control rods into the cores'ombined with a failure to inject boron into the .vessel by .the SLCS.

• Failure is defined to be'non"insertion of CRDs in the following manner. >50% in a "checkerboard pattern" r

)31% in a random patterns or )4% in a cluster.

WNP-2 Q. 211.150 (1 5. 0)

Provide a Listing of the transients and accidents in Chapter 15 for which operator action is required in order to mitigate the consequences. For corrective actions required prior to 20 minutesr provide just" i f i cat i on.

Response

For the design basis accident events (i.e.i cited in Chapter 15'he required operator action LOCA's) and its justification are detailed in the responses to

~

the staf f Questions 211.59 and 211.65.

For all anticipated transients cited in Chapter 15m no operator action is assumed in less than 10 minutes to mitigate the consequences of the event or to prevent the pLant from exceeding safety. design Limits. Oper-ator action is allowed and utilized after 10 minutes in order to maintain the plant:

a) In a steady state condition; b) Initiate safe and orderly shut-down; c) Naneuver plant from condition that would neces-sitate safety action; or d) Reduce the impact on pLant system operation due to a single operator error or a single equipment maL-function.

In no case would the operator's action or non"action resuLt in an unacceptabLe ef feet on the health and safety of the general public. This operator action for transients certainly is justifiabLe since it is his (or her) normaL operational assignment

WNP-2 Q. 211.152 (15.0)

In relation to Figure 15.0-2i confirm the following items for all transients in Chapter 15.0 which require control rod insertion to prevent or Lessen plant damage.

a) The scram curve used in Chapter 15.0 anaLyses (Figure 15.0-2) has a total reactivity worth of

$ 37-05 and is the nominal conservatism factor of 0.8.

b) The slowest aLLowable scram insertion speed 'was for the scram curve applied to Chapter 15.0 analyses.

c) The end of cycle 1 scram curve has a total reac-tivity worth of $ 40.21 and is identified incorrectly'n Figure 15.0-2.

Response

a) The scram curve used in Chapter 15.0 analyses with a total reactivity worth of $ 37 .05 is quite conser-vative compared to the nominal scram curve. For any transients =since the neutron fLux would drop sig-nificantLy within 2 to 3 seconds after scram'he excess negative reactivity introduced after this short period of time has negligible effect on the peak values of the important parametersr e.g.i neutron f Luxr surface heat flux or vessel pressure.

The initial portion of the scram curve used for analyses bounds the nominaL scram curve multiplied by the conservatism factor of 0.8 in order to assure the coverage of the transient effect with the intended conservatism.

b) The scram time characteristic shown in Figure 15.0-2 is derived from the Technical Specification scram time. The slowest allowable scram insertion speed was used for, the scram curve applied to Chapter 15.0 analyses.

c) The $ 40.21 is the correct total reactivity worth for the end of cycle 1 scram curve and is incor-rectly Labeled in Figure 15.0"2. Figure 15.0-2 has been corrected'

• Draft FSAR page change attached.

$ 40.21

$ 37.05 z END OF CYCLE I S" RAM CURVE O

I 0$

67B SCRAM ROD DRIVE 30 zV I

0$ I l

'oI z C O

Ill o 0.4 ~ I C

l' o 40

~ ~

II I

SCRAM CURVE USED LN ANAI.YSIS 10 z

oO 0

0 TIME laccI

)5, G-22 RnSHIHGTOX'PUBLIC POt ER SUPPLY SYSTEH SCRN POSITION AkP Ri~CTIVITY FI GURr.

AR I-RDJECT hO. 2 C'AHA .7 i Pi ST1CS '5. G-"'

Q'. 211 .1 64 (15.1.1.2.2)

On page 15.1-2i it is stated that the thermal power monitor (TPN) is the primary protection system for mitigating the consequences of the transient resulting from Loss of feedwater heating ~ A description of this monitors which typically involves the flow-weighted APRN scram in conjunction with a 6-second time con-stant circuits was not.found in the WNP"2 FSAR. Pro-vide this description in sufficient detail to permit evaluation of the TPN for WNP-2.

If the time constants which affects scram initiation.

by the TPNr is Less than the effective time constant for the WNP 2 fueL for this type of transients the TPN should provide a conservative'easure of the time var-iation is surface heat flux; Howevers if the time constant is appreciably Larger than that'or the fuels the fixed APRN trip without a time constant would pro-vide the scram protection. The resulting NCPR woul'd then be Less than that predicted for the TPN scram which has a Lower setpoint.

There is no current provision .in the Technical Spec-i f i cat ions -for surveil lance of this t ime constant c'ir-cuit. It is the staff 's position that credit be taken only for,the fixed APRN scram in Chapter 15 unless the TPN is approved by the staff and appropriate Limiting conditions for operation and survei llanc.e requirements are incorporated in the TechnicaL Specifications for WNP-2.

a) Provide an analysis of the "Loss of feedwater assuming credit only for the fixed heat-'ng",transient APRN scram. This is a more conservative it will result in a more severe transient approach'ecause due to a higher fixed APRM scram setpoint.

b) Revise NSOA Figure 15A.6-21 to indicate the high fLux scram signal occurs from the fixed APRN scram instead of the TPN.

c) Re-evaluate single failure criteria in Sect ion 15.1.1.2.3 without taking credit for the TPN.

WNP-2

Response

Sections 7.6.1.4.3 and 7 2.",.1.b.1.b of the WNP-2 FSAR-describe the thermal power monitor function of the Neutron Nonitoring System (NNS). Table 7.6-3 APRM System Trips and Figure 7.6-10m APRN Circuit Arrange-ment for RPS Inputs provide additionaL information on the TPN setpoints and trip actions. These descriptions permit the evaluation of the TPN's application to WNP-2 In a response to an earlier questions (211.089) the Supply System outlined its intentions with respect to technical specifications surveiLlance requirements and Limiting conditions for operation for the TPN. The TPN components are safety grade qualified (SC"2i quality class Ii seismic categoryI) and the system is designed to be single failure proof. For these reasonsi it is appropriate to take credit for the TPN scram during the "Loss of feedwater heating" (LFWH) event. Re-analysis of the LRWH event and revision of the appropriate NSOA figure without TPN is unjustified. The evaluation in Section 15.1.1.2.3 remains accurate.*

• Draft FSAR page change attached.

'0 rqNP -2 15.1.1.2 Sequence of Events and Systems Operation l5.1.1.2.1 Sequence of Events Tables 15.1-1 and 15.1-2 lists the sequence bf events for this transient and its effect on various par'ameters is shown in Figure 15.1-1 and 15.1-2.

15.1.1.2.1.1 Identification of Operator Actions In the automatic flux/flow control mode, the reactor settles

,out at a lower recirculation flow wi,th no change in steam output. An average power range monitor (APRM) neutron flux or thermal power alarm 'will alert the operator that he must insert control rods to get back down to the rated flow control line, or that he must reduce flow if in the manual mode. The operator must determine from existing tables the maximum allowable T-G output with feedwater heaters out-of service. .If reactor scram occurs, as it does in manual flow control mode, the operator must monitor the reactor water level and pressure controls and the T-G auxiliaries during coastdown.

15.1.1'.2.2 Systems Operation In establishing the expected. sequence of events and simu-lating the plant performance, it was assumed that normal functioning occurred in the plant instrumentation and controls, plant 'protection and reactor protection systems.

The thermal power monitor (TPM) is the primary protection system grip in mitigating the consequen of this event;

~PCkscr~p8~ of ge 7s'm ia previcfM IA I. bb l.b 7 0 ~ ~.~-.

y,g, Required operation of engineered'safeguard features (ESF) is not expected for either'of these transients.

15.1.1.2.3 The Effect of Single Failures and Operator Errors These two events generally lead to an increase in reactor power level. The 'TPM mentioned in 15.1.1.2.2 is the mitigating system and is designed to be single failure proof.

15.1-2

WNP-2 Q. 211.168 (15.14.2.1 .1)

For the "inadvertent opening of a safety/reLief valve" transients include the time at which suppression pool temperature alarms and Technical Specification Limit are attained in event Table 15.1-5.

Response

WNP-2 is currently in the process of analyzing t,he Suppression Pool (SP) temperature response for various transientsr including a stuck open relief valve. The

'specific transients to be analyzed along with the

~

methodologyi assumptions and initial conditions. are outlined in NUREG-0783. Among the initial conditions assumed in these analyses is that the. suppression pool temperature is at the technicaL specification -Limit.

For WNP-2r this Limit- is 90o Fahrenheiti which is also the SP high temperature alarm point. Table 15.1 "5 has been revised to indicate that at time zero the plant is operating at the maximum technical specificationr suppression pooL temperat'ure and the high suppression pool temperature alarm is received in the control room.*

Upon completion of the SP temperature response analysis' the FSAR will be revi sed to incorporate the finaL results.

• Draft FSAR page change attached.

(ca emcee~>

%erg/ < ~

TABLE 15 . 1-5 SEQUENCE OF EVENTS FOR INADVERTANT SRV OPENING Time Event 0 Initiate opening of 1 safety relief valve 0 which remains open throughout the event.

~O10 Min Operator attempts to close valve fail.

Operator scrams the plant and MSIV closure occurs (worst case).

10.5 Min RCIC or HPCS'nitiate.

20 Min Operator activates RHR in suppression pool cooling mode.

5 Hours Shutdown and cooldown completed.

Pool ee) &CA N]tCC /filip c)pE ag Ior /cc pl yc g bcpl/ rcsslan pool hip h krnpc ro*rc a /or.w r

Ak oclgrcWen poo I Is considcrc'd 4 bc a f /Ae ~u gioTum fe> ~pc<. +c~~cr~krC kr con 6n uouq p~c~ ~cicero.&on ecpI 4 0</ ~ pool c~l)n g liy ac/'vie C.

WNP-2 Q. 211.180 Question deleted.

WNP-2

a. 211.189 (15.2.9.3)

For the "failure of BHR shutdown cooling" events spec-ific input parameters for the models used to evaluate blowdown rate and suppression pool temperature are shown in Table 15.2-13 along with the analytical results in Figures 15.2-16'17'18'nd -19. In connection with this'rofide the foLLowing information:

a) Identify the analytical models used to eva L uat e b lowdown rate and suppress i on pooL temperature.

b) Re'vi se Table 15.2-13 to inc lude al L the input parameters for the models to be identified in step a) and provide just-ification that the input parameters are conservative.

In additions it is indicated that only a qualitative evaluation of the "faiLure of RHR shutdown cooling" transient is provided because the core behavior has been analyzed in Section 15.2.6. Update the FSAR to indicate a quantitative analysis has been provided.

Response

a) The analytical computer codes used to evaluate bLowdown rate and suppression pool temperature response are descr ibed in NEDO-10320 and NEDO-10320 Supplement 1r "General ELectric Pressur e Suppression Cont'ainment Analytical Nodelr" and in NEDE-20877~ "Long Term Containment Response for BWR."

b) Table 15.2-13 has been proveded to show the key param,eters which ralate to the transient analysis.* Providing a complete List of inputs would be impractical. The short and long term responses were obtained from the models referenced in a) above. Parameters in which variations might have significant ef fe'ct upon the results were selected at the most conservative design basis values (e.g.i minimum suppression pooL mass) to maximize the containment pressure and temperature response. If some area of input is of speciaL

WNP-2 interests it can be provided upon specific request.

Section 15.2.9 of the ~>NP-2 FSAR has been revised to remove the reference to the "qualitative evaluation."*

f

• Draf t FSAR page changes yttached.

Ec8~ /o@

o At 100 psig RPV pressure, actuates ADS to complete blowdowng and the operator establishes e reactor-cooling paN ns described in the notes for Figure 15.2-11.

Time recpxired to initiate &e necessary steps to maintain reactor pressure and level co".trol is approximately 10 minutes.

/

'15.2.9.2.2 System Operation Plant instrumentation and control is assumed to be functioning normally except, as noted. Zn this evaluation credit is taken~

for the plant nd reactor protection systems and/or the ESP utilized.

- C 15.2.9'.2.3 The Effect of Single Failures and Operator Errors. "

The worst case single failure (Loss of Division Power) hai =:.-:~;";.l'- ':-"-"....

already been analyzed'in thi.s event. Therefore,'o =single ':-'.-..-."..=,":.'.:,i':

failure or operator error can make the 'consequences of this:-"!.-':-'";":"-

-event any worse. See 15A for 'discussion of this subject.'": '.-

~ .. '.'

'15.2.9'3

~ " .. ', ....,-

Core and System Performance "'.:;-.-

~ .'.."-.

An event, that can'irectly cause reactor vessel water tempera- "-;::;-.":.--.:.

ture increase io one in which the energy removal rate is less than the decay heat rate. The applicable event i.s loss of RHR.'."!-'.:.- "."."'.'

shutdown cooling. This event can occur only during the low -..:4 .,':-:

pressure portion of a normal reactor shutdown and cooldown",",=-'-'=-.-.~j'-'- "'-:..:;.

when the RHR system is operating in the shutdown cooling mode..:"-'"'-.'-.-.=-.-',.:..-"

The'.earliest, time the'hutdown system can."be actuated 'is 2-.3'~.."':.<=:.'~"'-".-.

.'.:.: hours. after..shutdown.isinitiated.';,.During .this.time. HCPR,re-;;.,:~,'.:4>S- '

~ ':.chins" high. md nucleate. boiling"heat. trusser:.-.ia".no'exceeded:-'~"=::y, at any time.'herefore the core therma1. safity.margin re-M:.= .":-;:~~.";.".,

mains essentially unchanged. The 10-minute time period approximated for operator action is an estimate of how long it i

would take the operator to initiate the necessar ime by which he must initiate iction. Oyf ag

'on a i t e va a o is ro d be in tea sip&t TRAh)SIC'V7 Q4ap c)~ A ~ + ' .i 0 +

seckio Is. Z. ~~

15 '-66

c. ~

I 4l L'~

15.2.9.4 Qu li at ve Results a

0 For most single failures that could result in loss of shutdown cooling, no unique safety actions are required. In these cases, shutdown cooling is simply re-establ ished using the redundant shutdown cooling loop . 'In cases where the RHR shut-down cooling suction line, valves cannot be opened, alternate paths are available to accompl sh the shutdown cooling func-tion (Figure 15.2-10). An evaluation has been performed.

assuming a failure that disables the RHR shutdown cooling suc-tion line valves.

This evaluation demonstrates the capability to safely transfer fission product decay heat and other residual'eat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant ....

pressure boundary are not exceeded.

The alternate cooldown path chosen to accomplish the shutdown cooling function utilizes the RHR and'ADS or normal.rel'ief"':"-'-.-';.-".':.,'".'alve systems (see Reference 15.2-3 and Figure-.15.2-11').'he:-'.- " -'..=.',"-:"::-: ~"-':

alternate shutdown systems're capable 'of performing the func- ".',.'.;".':.::;.:",:.'-.::,'.",,

tion of transferring heat from .the reactor to tabb'nvironment :-:="."'. ",':,.:-,;;,:.-.=':-"

using only safety systems. The systems are'capable of '.'

bringing the reactor to a cold shutdown in approximately 36 "':.'-.'-- .'.-.:.";-';=..(;-,".'

hours or less after the transient occurs.:=-';;;-".".'.=.'

'I

",'; '.".c.'.'- -", ..-.-'.."'"."..

I The systems have suitable redundancy in components such that.

even for onsite electrical power operation (offsite power is not available) the systems'afety function can ke -'.:"--.'..',

accomplished assuming an "additional single failure . The -'" .::,

'ystems c n be fully,operated from the main control'-room.:.';.:",';".-:'.-.-

The'.design. evaluation. is.divided: into. two phases:

" ....,:.power. operation to. approximately .100 psig vessel'ressure,~and-.~~-,."

'1) full.:--: '"~";~'="-.~.~" ~

" '.:. '2) approximately 106. psig vessel -pressure'"'to.cold. shutdown-~",'-'<".~';~~~ I (14.7 asia 200 Z) conditions.

15.2.9.4 1 Full Power to Approximately 100 psig Independent of the event that initiated plant shutdown (whether it be a normal plant shutdown or a forced plant shutdown), the reactor is normally brought to approximately 100 psig using either the main condenser or, in the case where the main condenser is unavailable, the RCXC/HPCS systems together with the nuclear boi:ler pressure relief system and.

the RHR heat exchanger in the. suppression pool cooling mode. ll 15 2-67

<J

. WNP-2 Q. 211.198 (6.3)

Expand the discussion in >ec(ion 6.3 to describe the design provisions that are incorporated to facilitate maintenance (including draining and fLushing) and con-tinuous operation of the gCCS .pumpsi seals'alvesi heat exchangers and piping rusn in the Long-term LOCA mode of operation considering that the water being recircu lated is potentiaLLy very radioactive.

Response

In response to items II.B.2 and II.D.1 of is currently evaluating the ECCSr as weLL as NUREG-0737'NP-2 RCIC and other systems required for Long-term coolingr considering that the water being recirculated may be potentiaLly very radioactive. One. of the objectives of this evaluation of to determine that the release of Large amounts of radioactive material wiLL not Limit personnel occupancy or degrade safety equipment by the radiation fields that may exist during and following the accident to the extent that'equired safety func-tions cannot be accomplished.

The evaLuation assumes the source terms recommended in ReguLatory Guide 1.3 and 1.7 and Standard Review Plan 15.6.5. It further assumes that these source terms are released instantaneously at the start of the accident. No one particu lar accident scenario is used; howevers all systems which receive automatic initiating signals are assumed to be running. If these systems take suction from the containment atmosphere or the suppression pools they are assumed to be contaminated.

The systems assumed to be contaminated are as follows:

Reactor, Core Isolation Cooling (RCIC)

Residual Heat Removal (RHR)

Low Pressure Core Spray (LPCS)

~

High Pressure Core Spray (HPCS)

Containment Atmospheric Control (CAC)

Nain Steam (up to the second isolation valve) (NS)

Nain Steam Isolation Valve Leakage Control System (Pj S I VL C)

WNP-2 Primary Containment Secondary Containment (due to Leakage from pri-mary containment and systems in secondary containment)

Standby Gas Treatment (SQT) irradiation The reactor buiLding is separated into radiation. zones.

Within each zoner aLL the significant con-taminated fLuid piping is Located. In additions aLL safety related equipment is identified and Located in each radiation zone. A "worst case" target is chosen and Located by inspection or by order-of-magnitude calculations for each potential "worst case" target.

Nexti the total integrated accident dose's into account direct shine from the contaminated calculated'aking system piping and from the primary containment building.

The accident dose is calculated using a time period of six months because the integrated dose from the con-taminated fluid systems does not increase significantly after six months. The accident dose is acced to the 40 year integrated. dose from normal plant operation.

It is this total integrated dose for the "worst case" total integrated doser then specif ic calculations are performed. If necessaryi the equipment is relocatedr replacedr or shielded.

In addition to determining that the equipment needed to mitigate an accident is not unduly degraded by the resulting radiation fields'he evaluation also identi-fies vital areas needed for post-accident operations (e.g.r Control Roomi Sampling Stations Technical Sup-port Centers etc.) and provides assurance that there is access to these vitaL areas.

An interim report wiLL be submitted to the NRC by the end of 1981 identifying aLL the assumptions and meth-odologies used to do the evaluation. Results of the evaluation in terms of accessibility and equipment reLiabi lity wiLL also be presented. As required by NUREG-0737'tem II.B.2r the final reports including an evaluation of'll safety-related equipment applicablei wiLL be completed and submitted at Least four months prior to issuance of'n operating License.

(

WNP-2 Q. 211 .202 A t ime'r i s used in each ADS Logic. The t ime delay set-ting before actuation of the ADS is Long enough that the HPCS system has time to operatei yet not so Long that the LPCI and core spray systems are unable to adequately cool the fuel if the HPCS system fails to start. Nanual reset circuits are provided for the ADS initiation signal and primary containment high pressure signals. By resetting these signals manuallyi the delay timers are recycled. The operator can use the reset pushbuttons to delay or prevent automatic opening of the "

relief valves if such delay or prevention is necessary.

The operator may aLso interrupt the depressurization at any time by 'the same action. The operator would make this decision based on an assessment of other plant conditions.

Discuss in detail any criteria to be given to the oper-ator (e.g.i in emergency proceduresi or operator trainin'g) that. wouLd form the bases for the operator's decision.

Discus's the consequences of interrupting ADS depressuri-zation prior to reaching the injection pressure for Low pressure systems. I

Response

At the present timer WNP-2 is still formulating, the emergency procedures. This effor wiLL involve incor-porating BWR Emergency Procedure Guidelines (EPG) wh ichi in parti provide the following criteria:

Nonitor reactor pressure vessel water level and pressure and primary containment temperatures and pressures from multiple indications.

2) If a safety function initiates automaticallyi assume a true initiating event has occurred unless other-wise confirmed by at Least two 2) independent'ndi-cations.

3) Do not secure or place an emergency core cooling system in manual mode unless by at Least two (2) independent indicationsi a) missoperation in auto-matic mode is confirmedi or b) adaquate core cooL-ing is assured. If an emergency core cooling sys" .

tern is placed in the manuaL modei it wiLL not initiate automatically. Nake frequent checks of th' initiating or controlling parameters. When manuaL operation is no Longer requiredi restore the system to automatic/standby mode if possible.

a 11 WNP-2 The Emergency Procedures as they now exist specifically mention resetting the ADS timers in two precedures:

4.8.1 HPCS Failure Step 4.8.1.4 Step A.9 and 4.8.0.2.4 Total Loss'f,AL L Feedwa'.er FLow Step 3.C.

Step 4.8.1.4r A-9 HPCS Fai Lure-

"If reactor water is being,res'tored and ADS timers have initiatedi reset the timers at paneL P.601.

"Note: ADS timers are not to be reset without permis-sion of the Shift Supervisor."

Step 4.8.0.2.4i 3 Total Loss of ALL feedwater FLow-

"If ADS timers

~ C have been initiatedr monitor reactor water Level.

If Level can be restored and maintained with-'HPCS and RCICi reset the ADS timers on Panel P.601."

"Note'. ADS =timers are not to be reset without permis" sion of the Shift Supervisor."

Resett ing of the ADS timer as weLL as the interruption o f AD'S depressurization are both covered under the EPG as des cribed above. The operator would not reset ADS unL-ess he would confirm either adequate core cooLing or,ADS system messoperation by at Least two independent indica t ions.

In additions the ADS function would onLy be required the H'igh Pressure (HP) makeup i

during the course of a plant trans en't in the event that systems fai Led.

WNP-2 Q. 211.203 (6.3)

Restricting orifices are commonly installed downstream of a pump to limit the maximum f Low rate that could occur and prevent pump damage if the pump discharge runout protection). It line wer e to fail (i.e.r pump is not clear whether or not restricting orif ice plates will be used for the LPCI system at WNP 2. Figures

'5.4 "13a and 5.4-13b show a rest ri ct ing or i f i ce in the injection piping of each LPCI Loop. Howevers note 9 on Figure 5.4-13a states that these orifices are recom-mended but not required.

Oescribe precautionary measures taken to reduce the potentiaL for.LPCI pump damage due to runout conditions.

Response

The metering orifice in the discharge Line does not serve as a restricting orifice.

The piping for each mode of RHR operation has been investigated to ensure that the resistance is Low enough to allow the rated f Lows given in Figure 5.4-14b yet high enough to prevent pump runout. Restricting orifices are necessary in the system test Lines to pre-vent excessive runout during suppression pool cooling and test modes and in the main discharge Line to pre" vent excessive runout for LPCI. and aLL other RHR modes.

'Engineering changes are currently being processed which will add these rest ricting orifices. Figure 3.2-6 wiLL be revised to indicate the Location of restricting orifices in the main discharge Line.

WNP-2 Q. 211 .204 (6.3)

Figures 6.3-53ai -53bi -54ai and -54b show the results of a break in a core spray line from the "Lead plant" analyses. The assumed single failure shown on the figures does not appear to be the most Limiting. It would appear that the LPCI dieseL-generator failure (division 2) would be more restrictive than, the LPCS diesel-generator failure (division 1) i i.e.r only,LPCI Loop A wuld be available to ref Lood the core. Explain why failure of the LPCI diesel-generator (division 2) does not result in a higher peak cladding temperature than that shown on Figure 6.3-54b.

Response

Assuming a high pressure core spray (HPCS) Line breaks the worst single failure is the LPCS diesel-generator (DG failure (division 1). With this failure only 2 LPCI Loops are avai Lable for ref Looding the core. 'The LPCI .is injected into the core bypass region and drains into the Lower plenum through specified Leakage paths (refer to NEDE 20566'ection= 3.3 p. II-14 for further details) . The flow allowed through these Leakage paths is insufficient to completely drain the flow from 2 LPCI Loops. Therefore for the LPCS DG failure case there is a buildup of LPCI f Low in the core bypass region. This water that builds up in the bypass region does not directly contribute to ref looding the core. These factors combine to produce correspondingly Longer ref looding times (and hence higher peak cladding temper-atures) for the LPCS DG failure case when compared to the LPCI DG failure case.

For the LPCI DG (division 2) failure case which Leaves 1 LPCI and 1 LPCS avai Lable for ref Looding the core from division 1i the f Low of the 1 LPCI is allowed to drain through the Leakage paths. Also'lthough Limited by .counter current flow Limiting (CCFL) con-siderationsr STET the LPCS flow passes through the fueL bundles and into the Lower plenum thereby providing core spray heat transfer and directly contributing to the refLooding of the core.

WNP-2 Q 211.211 (15.3.3)

The response to question 211.092 is unacceptable.

Explain why the DBA-LOCA event is indicated as conser-vatively bounding the pump seizure event when different acceptance cr iteria are used for each. The pump seizure event is evaluated based on exceeding 10CFR100 guide-lines whereas the main criterion for evaluating the DBA-LOCA event is a peak cladding t'emperature of 2200oF.

Coordinate this request with question 211.185.

Response

See response to question 211.185.

WNP-2 Q. 211.212 Our position on the emergency core cooling systems (ECCS) is that these systems should be disigned to withstand the failure of any single active or passive component without adversely affecting their long-term cooling capabilities. In this regard~ we are concerned that the suppression pooL in boiling water reactors (BWR's) may be drained by Leakage from isolation valves which may be rendered inaccessible by Localized radioactive contamination foLLowing a postulated Loss-of-coolant accident (LOCA) . According Lyr. indicate the disegn features in the WNP2 facility which wiLL con-tain Leakage from the first isolation valve in the ECCS Lines taking water (suction Lines) from the sup-pression pooL during the long-term cooling phase fol-Lowing a postulated LOCA.

Response

During normal operations Leakage is coLLected in sumps located in the ECCS and RCIC pump compartments in the reactor building and pumped to the radwaste building for processing. The ECCS and RCIC pump com-partments are watertight structuresi the waLLs of which rise to a Level above the suppression pool water Level.

For the worst conceivable Leak f rom the suppression pooL in which the water Level in the Largest of these pump compartments equalizes with the suppression pooL water Leveli at Least 12 feet of NPSH over NPSH requirements on the ECCS pump performance reports is maintained for each pump. The suppression pool water temperature was taken at 212oF.

Additionallyr the concerns in this question have been previously responded -to in NRC Question 212.003r in which WNP-2 committed to installing Class IE LeveL instru-ments in each ECCS pump room. These instruments wiLL be mounted just above floor Level to allow sufficient time for the operator to identify and isolate the faulted ECCS Line. Any ECCS Leak can be isolatedi including any packing failure on any ECCS pump suction valve.

WNP-2 Q. 022.069 (RSP)

Based on our review of the information presented in Section 6.2.1.1.5 of the FSAR .and your responset to item 022.018 which references your response to Item 031.070m we find that your discussion of steam bypass from the drywell to the wetwell for postulated smaLL steam Line breaksi is unaccptable. Specificallyi the maximum aLLowable bypass Leakage which you calculate (i.e.rA4K = 0-028 sq ~ ft.) i is not acceptable. Accord-inglyr we require that you design the WNP-2 contain-ment to have a bypass Leakage capability which sat-isf ies the provisions of Appendix I to Section 6.2.1.1.'C of the Standard Review Plan (SRP) (i.e.r Agg = 0.05 sq. ft.). Provide the appropriate discussionsr just-ifications and analyses to demonstr ate how you comply with the provisions of Appendix I cited above.

Response

Please ref er to revised 6.2.1.1.5.4 and Figure 6.2-17b.

Also refer to the revised response to Question 031.070 *

• Draft FSAR page changes attached.

6.2.3..-3..5.2 Reactor Blowdown Conditions and Operator

Response

In the highly unlikely event of a primary system leak in the drywell accompanied by a simultaneous open bypass path be-tween the drywell and suppression chamber, several postu-lated conditions may occur. Fox a given primary system bxcak area> the maximum allowable leakage capacity can be determined when the containment pressure reaches the design pressure at the end of reactor blowdown. The most limiting-conditions would occur for those primary system break sizes which do not cause rapid reactor depressuriz rather have long leakage duration. %~ reak sizsf - e' uhlan to terminate the reactor blowdown

< 'w 4.'* )

pr'tem break, therethe

~ntainmenc . sure as 4 W lee would be a fairly rapid rise in noncondensable gases in the dry-1 well are carried o o the suppression chamber. During this portion of the tran -'it is assumed that the plant, are unaware that a le path exists. Under i~4 A opexators these circumstances, the maximum pres the suppression chamber is appxoximately 27 at CGIl occur is 5+chd the pressure that would result, if all of the nonco This able

~ ~ ~

4 e4 Fox the maximum allowable leakage calculations, tors realize a leaka e it .

~~hssumed that the plant opera-oath exists I% e J re@ 4 4 ke 4 4

~ ae %4 Ca

~ was rr 4 6.2.1.1.5.3 Analytical Assumptions Shen calculating the allowable leakage capacities for a spectrum of break sizes, the following assumptions are made:

6.2-29

Inser&to pa e 6.2-29:

Following 'the postulated condition given above, there is an increase in drywell pressure which leads to drywell venting to the wetwell via the downcomers. Both noncondensibles and vapor are vented. Xf no bypass leakage exists, the maximum, suppression chamber pressure would be 28 psig, the pressure resulting from displacing all containment noncondensibles into the suppression chamber.

Insert to pa e 6.2-29:

after an alarm announces that the suppression chanher pressure has reached 30 'psig and prepare to take action. Containment sprays must be initiated with a delay no longer than 41 minutes or drywell pressure will exceed 45 psig. Containment pressure decreases immediately upon the starting of drywell spraying.

Once .the containment pressure is reduced by means of drywell spraying, the operators would proceed to depressurize the reac-tor'essel by means of either the main condenser or the relief valves.

WNP 2 APXNDMENT NO ~

MARCH 19 79

a. Plow through the. postulated leakage path is pure steam. Por a given leakage path, if the 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 b

in the additional containment pressurization, u s w~

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 the suppression chamber pressure, this is a conservative assump'

• . . tion. Por an actual containment there will be condensation, especially for the larger primary system break where vigorous agitation at the pool surface vill occur during blowdown.

6.2.1.1.5.4 Analytical Results 1

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 wi& the leakage flow path.

Figure 6.2-17a is a composite of 'two curves. If 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 b wdown its the allowable leakacre to s a> values.

K = .

maximum allowable leakage capacity square feet. Since a typical geometric loss

~ at factor wouTd be three or greater, the maximum allowable flow path would be about quare feet. his corresponds to a~ inch line size.

Burns and Roe, Inc. confirmed the results of the above

. analysis by GE in Reference 6.2-7. Further investigation into the transient nature of the problem was then unde'Waken ss the z'squest sf the tTRC~~gg 6~ 2-30

~ ~ ~

Ag P

' ~

~ I ~

~~ ~~kg ~ ~

V WNP-2 AMENDMENT NQ. 3 March 1978 A transient analysis using the CONTEMPT-LT (Ref. 6.2-8) computer code was performed. The code was modified to in-clude 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 mange~ at 100oF/hr cooldown rate. No other operator actionsaccounted 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. Bpqed on this analysis, the allowable ft

~

bypass leaka e (A V K)~% s . The drywell pressure ansxent is shown in Figure 6.2-17b along with the cones-O.0~0 ponding curves of wetwell pressure, wetwell temperature and suppression pool temperature.

e~ &pe

~need ek Theaallowable bypass leakage of ~~

aa80 ft2 is well above the containment bypass leakage. Periodic testing will be performed to confirm that the containment bypass leakage does not exceed A/.VK = 0.0045 ft2. Figure 6.2-17c presents'the resulting containment transient for A/ V K = 0.0045 ft2. The peak containment pressure shown in Figure 6.2-17c is well below the containment design pressure.

6.2.1.1.6 Suppression Pool Dynamic Loads A'generic discussion of the suppression pool dynamic loads and asymmetric loading conditions - s given in Mark II Dynamic Forcing Function Information Report, Reference 6.2-4.

A unique plant assessment of these dynamic loads is made in HNP-2 Design Assessment Report, Reference 6.2-5.

6.2.1.1.7 Asymmetric Loading Conditions See comment in 6.2.1.3..6..

l gylfkiQ&8. QE dt glo+lI Sfl(cg+ +Q~ ~K Q~~

~~

pcs~( 5 Qccc<ds QUf)P~l~ f) ig> RhO %CA TO (RC83 cpm@~(

6.2-30a

g ~~J I

WNP-2 2LMENDJMNT NO ~ 3 maCS 1979 Page 2 of 5 A point by point discussion summarizing the NPPSS design capabilit:ies to mitigate Bypass Leakage problems based on the above correspondence and with respect to the proposed Branch Technical Position is given below:

NRC Proposed Requirement: Allowable bypass capability on the order of 0.05 ft (A/ ~K

>red Response: As Reference nd the FSAR, the maxim allowable bypass leakage v is A'VK se o450 I a4 2~ NRC Pro osed Recruirement: An automatic system should be provided to initiate automatic wet-well 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.

purpose, If it the RHR syst: em is used for this must be analyzed to assux'e no degradation of its ECCS function.

Response: HPPSS asserts that manual initiation xs sufficient since the drywell floor will be routinely tested and evaluated against a Tech Spec limit of A/VK = 0.0045 which no operator'ction is ft , a level at recpxired for the spectx~ of small break sizes. (Reference 5-see t3 below for testing details) .

The construction, design, cpzality contxol, and surveillance recpxirements on the drywell floor give it the same level of safety as the contain-ment itself. Reference 4 and Part VI of Reference -2 showed that through-wall cracks will rot develop through conditions the concrete slab including under postulated design 4a

'tKH c

e e NNP-2 DYLAN T NO 3 MARCH 1979 Page 3 of 5 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 Branch's acceptance of these responses. Accordingly, HPPSS sees no reason to assume that an A/M of 0.0045 ft is exceeded. any more than there would be reason

~

to assume the design containment leak rate of

.5% per day is exceeded. Calculations aocumented in Reference 5 ~~ the CONTEMPT - LT computer code were used. in computing the maxim leakage rate of A V K = ~M& ft2,

...,,p..

](@<minutes was available or operator

~

action allowab3.e times the .

e al ue q.[

drywell design press4e ingly, a was exceeded.

requirement that an automatic system Accord- pi~ 786P~~

~~~",~K~

be provided is unnecessary., ~csRc&Q NRC Proposed Requirement: A single per-atxona1 high pressure 3.eakage test should be performed and periodic low pressure tests at each refueling outage with an acceptance cri-terion of 10% of the bypass capability at the test pressure.

Response: The intent of this proposed require-ment has been committed to by NPPSS. A single preoperational leakage test will be conducted with the downcomer's capped at 15 psid and 25 psid (the esign drywe3.1 to wetwel'if erential pressure). At each refuels g outage a low pressure operational test. will be performed as a Tech Spec Surveillance. Requirement to verify 0.0045 ft2. Details of the nature of 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 Inspection: During entry to the drywe11 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 031. 070-3

WNP.-2 AMENDMENT NO. 3

~J MARCH 1979 Page 5 of 5 References Letter, NR Butler, NRC, to JJ Stein, GPSS, "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 WPPSS, dated January 14, 1975.

4 ~ Letter GPSS to NRC g GO2 75 52 dated February 25 g 1975

5. Letter, NPPSS to NRC, GO2>>76-156, dated Apri3. 23, 1976.

I

6. Letter, NRC to GPSS, dated May 15, 1975.

gg.a. C?~~ Itn6.5.

Q p~ C~~'.z 031.070-5

~~

~

A, CRYW'ELL PR6&SuRF fb. wFTU&.l-i- PREssUH,6 53 c. Og WQSL4, TEvLPER,ATURK 0 ',4~FTN&U .'7FtfpgÃAVVQG:.::'::- ...

'd nH O

0 40NTAIltiME.N~ V~S+BL DF Sl64'PRR~SVP a) COps

~ C l/1

'If

~ g I ~

-goo N '0 V ps)

Ul IL t

I

,;I'j i goo

,'I I

I PRK%5uRB I

'i I, I TBMPF RA.TUR5.

"P5E A. j I II I I

~

('P) ..

/VS) 6)

I I

I I

}

I It ltt I:t

~

'j ,f!

I Q.oo

't

..:: ~

(LS)@II I Itl I i

j I

'I '

~

ll IL

.','I I'i 0r p)'rS fui1')PC>>cy I: lr~f I

P AgrgrPp'g pp'fip>> J

~

I'! fli I j

jI ~ I I. I Ii I

}

i Ij 'I lorna I 'Ijl illl LI I II

~

lit

~ t I

~

}'I I'I

'I

';jl

~ ~

<<j. I' t I ~I

~

~ I .I. :II tTI

~ ~~ j/f '.I-

~ It ~ ~ I ~ I t ~ ~~ ~

.I opggATUg- SIGNIFICANT POCL7IQhl ..

l HtlT Tl&B OP TAAHSIEQT HA%

<~~gWeg Nht)

~>~C 441J11$ 4ILS~ff PAL ~N ERG%0'EACt'aA

~t5YS tl lese.) A'] / l

- 4'/

< < A IJ J PII u 0 f

~

SP C

- .-" .". = 1c f'(Qf:III',g . >RlNCq R~buc.i")> 1'o -'.:.

CagTCuH~7 $ 'R6'-SU

WNP-2 Q. 022. 074 In note 31 of Table 6.2-16'ou indicate that primary t i c o n a nm e n t a n d r e'a c t o r v e s s e L i s o l a t i o n s i g n a L s a r e not desirable signals for initiating closure of the feedwater block vaLve. We find this approach accetable provided that the valve can be. manually c'Losed from the control room (i.e.,r remotely) if the control room oper-ator determines that continued makeup from the feedwater system is either unavailable or unnecessary. Discuss the information which will be available to the control room operator to alert him of the need to isolate the

'eewater system. Indicate the time interval which would then take the operator to complete this action.

Response

-The response to this question is. addressed under revised Section 6.2.4.3.2.1.1.1.*

• Draft revised FSAR page changes attached.

., ~

~ hZ E

.AMENDMENT HO..4

~

June 1979 o22 07'able 6.2-16 contains those influent pipes that comprise the reactor coolant pressure boundary and penetrate the contain-ment.

6.2.4.3.2.'1.1.1 Feedwatex Lines, The feedwater lines axe part of. the reactor coolant pressure boundaxy as they penetrate the dzywell to connect with the reactor pressuxe vessel. The"isolation valve inside the dxy-well is a y-pattexn check valve;.located as close as practic-able to the containment wall .Outside the containment is another y-pattern check valve located as close, as practicable to the containment wall and farther away from the containment is a motor operated gate valve. '-Should a break occur in the feedwatex'ine, the check valves -prevent sigriificant loss of reactor coolant inventory and offer immediate isolation. How-ever, .in case a loss-of-coolant accident occurs without a seismic event, the design allows the condensate and condensate booster pumps to supply feedwatex to the vessel thxough a by-pass line azound the reactor feed pumps - which a=e txipped on a loss of steam supply - as saon as the vessel is partially depressurized. For this reason, the outermost, gate valve does not autamatically isolate upon signal from the protection system. The gate valve meets the same environmental and seismic qualifications as the outboard isolation valve. The valve is capable of being remotely closed from the cantxol zoom to provide long-tenn leakage protection upon opexator WSE gT judgement that feedwater makeup is unavailable or unnecessary.

No credit is taken for feedwater flow in accessing core and containment response ta a loss-of-coolant accident.

6.2.4.3.2.1.1.2 HPCS Line The HPCS line penetrates the chill to inject directly into the reactor pzessuxe vessel. Isolation is provided by an air testable check valve, located inside the drywell with position incicated in the main contxol room, and xemote-manually actu-

~

ated pa"e valve located as close as practicable to the ex-terior walk of the containment. Long-term leakage control is maintamed by this gate valve. If a losswf-coolant accident occurred, this gate valve would receive an automatic signal to open.

6.2.4.3.2.1.1.3'PCX and LPCS Lines Satisfaction af isolatian criteria for the LPCI and the LPCS system is accomplished by use af remote-manually operated gate valves and check valves. Both types of valves axe nor-mally closed. with the gate valves receiving an automatic

Question 022.074 to 6.2-58 'nsert The opex'ator can determine if make-up from the feedwater system is unavailable by use of the feedwater flow indicator in the contxol room, which will show high flow.'.for,a feedwater pipe break or no flow for feedwatex pump trip.

The operator can also determin'e if'.make-up from the feedwater system is. unnecessary by verifying that, the ECCS is functioning. properly and the reactor water level is being adequately maintained. ECCS operation signals and reactor vessel water level indication are provided in the control room for operator information.

Since dur,to the check valves it is not necessary to immediately isolate the feedwater system for leakage mitigation, there is no need to alert the operator to initiate the feedwater isolation signal other than as described above. However, for long term isolation purposes, the opera-tor may close the motor>>operated gate valve at: any convenient time.

WNP-2 Q. 022.076 With regard to Note'-45 of Table 6.2-16 of the FSARi you should note that our acceptance criteria for containment isolation signals is set forth in Section 6.2.4r "Contain-ment Isolation Systemsr" of the SRP. Indicate whether your design for the WNP-2 .facility conforms to our acceptance criteria on this matter.

Response

The Recirculation pump seal, water supply Line does not close automatically because of the desirabi lity of main-taining the pump seaL water as Long as the'uppLy is available or the RRC pump is operating.

The Line 'wiLL be isolated manually from the controL room on indication of Loss of either the CRD or RRC pumps or a Loss of flow as indicated by the seaL water inLet and outlet pressure inst rumentat ion. AL inst rumentation indi cat ion L

is available to the operator in the controL room.

This design conforms to the requirements of the SRP 6.2.4 acceptance criteria in paragraph II.11.

p ( ~

0

> 0 I

~ 1

~ > ~

WNP-2

,Q. 022.080 (RSP)

Your response to question 022.050 in which you state that the hydrostatic or pneumatic test wiLL be repeated every ten years's not acceptable. It is our position that those systems identified as closed systems and which become an extension of the primary containmenti should be Leak tested during each reactor. shutdown for refuelingr or other convenient intervalsr but in no case at intervals greater than two years. Accordinglyr we require you to provide a commitment consistent with our position on this matter.

Response

Note 18 of Table 6.2-16 commits the Supply System to meeting the requirements for a closed sy'tem as set by SRP-6.2.4r Section IIr paragraph 3e which states in parts that "The closed system outside containment should be Leak tested unless it can be shown that the system integrity is being maintained during normal pLant operation." Of the five systems Listed as closed systems outside containment in question 022.050'our (LPCSr HPCSi RHRr RCIC) are equipped with water Leg pumps which maintain the system fuLL of water and pres-surized to about 85 psig during normaL operation. The sections of these pressurized systems outside containm ments thereforer are under a continuous Leak check at a pressure over twice that required by 10CFR50~

Appendix J. The pressurized system combined with periodic visual inspections are considered more than adequate to insure system integr ity during normal operation.

The Containment Atmosphere Control System (.CAC) is not pressurized during normal operation and w'ill be pneu-maticaLLy Leak testedi as requiredi during each reactor shutdown for refuelingi or other convenient intervalsr but in no case at intervals greater than two years.

Paragraph 6.2.6.5 of the WNP-2 FSAR wiLL be revised to show the additional test r equirements.

WNP-2

6. 2. 6.5 Special Testing Requirements The secondary containment sha11 be subjected to'ests prior initial fuel loading and at each refueling outage to as- 'o sure the maximum allowable leakage rate of 100% of. secondary containment free volume per day at -0..25 inches wa'ter gauge pressure with respect to outside atmospheric pressure. The test procedure for determining that the leakage rate does not exceed the maximum allowable. is summarized in 6.2.3.4.

See Chapter 16, Technical Specifications..'est .procedures for the MS?V-LCS are in 6.7.

ca l Hc H /DcoQG& K6<oQ j5'L Qp Q S Y s T5'~ $ w 4 C Lostq Q 0

CoMY'AlA.tee~> 5Hqv. pc' wA~c Lg~TQ>

6++4 76+ v5 L [40'QQQ gQ > ~r He% C,anvil.td.<~i l k) 7EPMOc.g p~+ o ~As' l nvCR.'obis G g.~g~e~ Qg ><

~/a e s ~ l << I~A lC. 7awzi~G %HE i< Q~ l~ Q~<<wotr ce P~q ~ mu (t>8 I ~ IQ [ P~~ (~

,3s 6.2-87 I

WNP-2 Q. 022.090 In the event of a smaLL steam Line breaks the exists for poten-.'.'ial steam to bypass the suppression pooL via the hydrogen controL system. Discuss the cap~

abil ity of the hydrogen'ecombiner system to condense this superheated steam. Discuss whether the after-cooler could be initiated'ndependently from the-recombiner system to condense Leaking steam. Discuss the design provisions in the WNP-2 facility to eliminate the potential for steam bypass.

Response

The hydrogen recombiner system containment isolation valves are normally closed and are opened'nly if the hydrogen concentration inside primary containment requires operation of the recombiner. Thereforer in the event of a smaLL steam Line break without hydrogen generationr the isolation valves remain closed and there is no potentiaL for steam bypass'hrough the hydrogen recombiner system.

The hydrogen recombiner system is designed to take suction from the dryweLL and discharge the processed stream into the suppression chamber. Existing dis-charge Line valves to the dryweLL and suction Line valves from the suppression chamber are key-Locked closed and their electricaL interlocks with the recom-biner are disconnected. The key Locks are Located on a control room panel for remote manual operationr when and if another mode of operation (based on hydro-gen concentration) is required. Refer to FSAR Section 6.2.5.2.3 and Figure 3.2-17. As di cussed in the response to NRC question 022.089'he hydrogen recom".

biner system is capable of handling a feed gas contain-ing steam. Thereforei in the event of a smaLL steam L ine brea'k with hydrogen generation and the recombiner system 'in operationi there, is no potential for steam bypass through the hydrogen recombiner system because any steam in the feed gas is condensed in the recom-biner's scrubber.

Based on the above discussion and the response to NRC quest ion 022.089'ndependent operation of the after-cooler to condense Leaking steam is not required.

~ a.

4 I