Text

Forwards Util Responses to Round One,Set Three Questions from Auxiliary Sys Branch.Responses Will Be Incorporated Into FSAR
	ML17272A453
Person / Time
Site:	Columbia
Issue date:	05/16/1979
From:	Renberger D; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Varga S; Office of Nuclear Reactor Regulation
References
	GO2-79-99, NUDOCS 7905250537
	Download: ML17272A453 (137)
	v • d • e

/(

t I

REGULATOR

%FORMATION DISTRIBUTION S~h.

EN (RIDS)

ACCESSION NBR: 7905250537 DOC. DATE: 79/05/ib NOTARIZED:

NQ DOCKET I FAC lL: 50-397 MPPSS NUCLEAR PROJECT, UNl'T 2, MASHINGTQN PUBLIC PQME 05000397 AUTH. MANE AUTHOR AFFILIATlQN RENBERGERi D. L.

MASHINGTQN PUBLIC PQMER SUPPLY SYSTEN REC IP. NANE RECIPIENT AFFILIATION VARGAiS. A.

LIGHT MATER REACTORS BRANCH 4

SUBJECT:

FGRMARDS RESPONSES TO ROUND QNEi SET THREE QUESTIONS FRQN AUXILIARYSYS BRANCH.

ESPONSES MELL BE INCORPORATED lN FSAR.

(~

DISTRIBUTION CODE:

B001 CQP IES RECElVED: LTR ~ ENCL ~~ SIZE:

8 TITLE: PSAR/FSAR ANDTS AND RELATED CORRESPONDENCE.

NOTES:

BILE PR iag4 b8L I ~

REC IP IENl CQP lES ID CODE/NANE LTTR ENCL ACTION:

05 PN H LP/4<~

1 1

BC ~~@

1 0

REC IP IENT ID CODE/MANE AD <<5&fmo LA ~~~ cf CQP IES LTTR ENCL 0

1 0

Ob I Z.E 09'EOSCIEN BR NECH ENG BR 13 NATL ENG BR ib ANALYSIS BR 18 AUX SYS BR 20 I 5 C SYS BR 22 AD SITE TECH 27 EFFL TRT SYS 2'P KIRKMGQD AD PLANT SYS AD SITE ANLYSIS NPA EXTERNAL: 03 LPDR 30 ACRS 1

1 2

2 1

2 2

1 1

4 4

1 1

1 0

0 0

02 NRC PDR 08 OPERA LIC BR 10 GAB 12 STRUC ENG BR 15 REAC SYS BR 17 CORE PERF BR 1'P CONTAIN SYS 21 PQMER SYS BR 2b ACCDNT ANLYS 28 RAD ASNT BR AD FQR ENG AD REAC SAFETY DIRECTOR NRR GELD 04 NSIC 1

1 1

0 1

0 Hbb ~~

Q-QC 5 r4OM 5

~IS MAY 80%7S TOTAL NUNBER QF COPIES REQUIRED:

LTTR 49 ENCL 39

.'i ~

P dl II ff I

e

- rg>>

1< q lt Tv y ll I

t k

I ~

I t~

e+>>

~

Washington Public Power Supply System JOINT OPERATING AGENCY P. O. 8OX 958 3000 GEO. WAIHINCTOH WAY RICHLAND. WASHIHOI'ON 99352 PHONC (509) 375 5000 Docket No. 50-397 May 16, 1979 G02-79-99 Director, Office of Nuclear Reactor Regulation U.

S.

Nuclear Regulatory Cormission Washington, D.

C.

20555 Attention:

Subject:

Reference:

Mr. S.

A. Varga, Chief

'Branch No.

4 Division of Project Management WPPSS NUCLEAR PROJECT NO.

2 RESPONSES TO ROUND ONE

UESTIONS, SET THREE - ASB

Letter, S. A. Varga (NRC) to N. Strand (WPPSS), "First Round questions on WNP-2 OL Application - ASB,H dated January 13, 1979.

Dear Mr. Varga:

Attached please find sixty (60) copies of'he responses to the rouhd one, set three questions representing the Auxiliary Systems Branch.

Also included are the responses to a few open items from a previous set.

The responses to these questions will be incorporated formally into the FSAR in an amendment within four months.

Very truly yours, D. L.

RENBERGER Assistant Director Technology DLR:SAG:sg

Attachment:

Responses to Round 1 guestions (60) cc:

I. Littman - WPPSS, NY - wo/att JJ Verderber - B&R, NY-JJ Byrnes - B&R, NY RC Root - B&R, Site HR Canter -'8R, NY C. Bryant -

BPA E.

Chang - GE, San Jose w/att (4)

FA MacLean - GE, San Jose (1)

J. Ellwanger - B&R, NY (5)

NS Reynolds - Debevoise

& Liberman w/att (1)

WNP-2 Files - w/att (1)

>Osegoqy t

STATE OF WASHINGTON)

)

ss COUNTY OF BENTON D. L.

RENBERGER, Being first duly sworn, deposes and says:

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

DATED

, 1979 D. L.

RENBERGER

, 1979.

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.

GIVEN under my hand and seal this /

ay of Notary Public in ~nd or the S

t of Washington ~

Residing at

a

.r v' I

WNP-2 Responses to:

Auxi1iary Systems Branch questions (10.10 - 10.34)

WNP-2 g.

010.10 3.4.1 Demonstrate that all piping and electrical penetrations in safety-related structures that are below the level of the Probable Maximum Flood, are water tight.

~Res onse:

As stated in 3.4.1.4.1 the plant site grade is higher than the design basis flood elevation resulting from the probable maximum precipitation (PMP) event.

Due to the short duration of the PMP flood, the ground water level at the plant site is not affected.

As stated in 3.4.1.4,2, piping and electrical penetrations are above the design basis ground-water level and are therefore not sealed against groundwater pressure.

L 0

WNP-2 g.

010.11 3.5 He require you to provide an evaluation of the environmental effects resulting from a postulated failure of the main steam lines and the main feedwater line.

Your evaluation should demonstrate conformance with our requirements that:

a.

Those compartments and tunnels which house the main steam lines, the feedwater lines, including the isolation valves for these lines, are designed to withstand the environmental effects (pressure, temperature and humidity) and the potential flooding resulting from a postulated crack equivalent to the flow area of a single-ended pipe rupture in these lines.

b.

The essential equipment located within these compartments, including the main steam line isolation valves and the feed-water valves and their associated valve operator s, are capable of operating in the environment resulting from the crack postulated in Item (a) above.

c.

If the forces resulting from this postulated crack could cause the structural failure of these compartments, the consequent failure of these compartments will not jeopardize the safe shutdown of the plant.

d.

The remaining portion of the pipe in the tunnel between the outboard safety valve and the Turbine Building meet the guidelines of Branch Technical Position ASB 3-7, "Protection Against Postulated. Piping Failures in Fluid Systems Outside Containment", with respect to the stress levels in this portion of the pipe and with respect to the location of the postulated break points.

We further require that you submit an analysis of the sub-compartment pressure buildup following a postulated pipe break, including the structural.-:evaluation of the affected sub-compartments, to demonstrate that the design of the pipe tunnel conforms with our positions as stated above.

If you cannot demonstrate conformance with our positions in this matter, indicate any design changes which may be required to comply with our positions.

This evaluation should demonstrate that the methods used to calculate the pressure transient in the sub-compartments outside of the primary containment are the same as those for sub-compartments inside the containment for postulated pipe break.

Demonstrate that the margin against a structural failure resulting from the pressure transient, are the same as those in sub-compartments inside the primary containment.

If you propose to use methods of analysis for sub-compartments outside of containment which are different from those used inside containment, demonstrate that the methods of analysis for sub-compartments outside containment

wNp-2 assure adequate design margins.

Identify the computer codes and the assumptions regarding the mass and energy release rates which you used in your analysis.

Provide sufficient, design data so that we may perform independent calculations.

~Res onse:

The complete response to this question will be supplied in July 1979.

The structural adequacy of the steam tunnel and the environmental conditions in the steam tunnel following a pipe break in a main steam line were previously evaluated based on a double-ended guillotine pipe break and instantaneous venting of the blowout panels.

This analysis is currently being reevaluated using using the RELAP 4 code in line with the conditions in this question including single-ended breaks for both feedwater and main steam.

It is expected that the results from the original analysis based on a double-ended break of the main steam line will be shown to be bounding.

WNP-2 01 0.12 Provide the results of your evaluation of the jet impingement forces and the environmental effects, including pressure, temperature,

humidity, and flooding, resulting from a postulated failure of the main steam and main feedwater systems in the turbine building.

This evaluation should address only those safety-related components, systems and struc-tures, if any, in (or immediately adjacent to) the turbine building (e.g., the walls of the auxiliary building).

~Res onse:

It has been determined that the only items with safety-related functions in the Tur'bine Building are some RPS sensor inputs from the Main Steam

System, NSIY isolation logic inputs from the Main Steam

System, and the Tower Nake-up Transformers located in the basement of the Turbine Building which are required to function only for the Design Basis Tornado event.

This last item is remote from the steam and feedwater lines (being located at the basement grade level of the building) and has been evaluated to have adequate protection from tornado missiles and internal flooding (see the responses to questions 10.25 and 10.34*).

In addition, there is cabling for the condensate storage tank level sensors whcih provide for auto-switching of HPCS from the storage tank to the suppression pool.

The routing of this cabling is currently through the turbine building, but is under design review to insure its adequate protection from accidents.

Appropriate design changes will be made as a consequence of this, evaluation.

Accordingly, the only items of concern are the RPS and MSIY isolation logic sensor inputs.

Due to their nature they cannot be made immune from pipe-break effects.

However, no analysis has been performed of the specific effects of a steam line or feedwater break in the Turbine Building on this equipment since it has been determined that the complete loss of all this equip-ment could occur for these events without the loss of capability to bring the plant to a cold shutdown or mitigate the radiological consequences of such an incident even assuming a single failure in the safety systems that remain unaffected.

The electrical cable connected with this safety related equipment in the corridors separating the Turbine Building, Reactor Building, and Radwaste Building would be exposed to temperatures and pressure effects of a postulated failure of the main steam or feedwater lines in the Turbine Building, but the exposure conditions would be for less than the design environmental requirements contained in the purchase specifications for the cable.

10.34 is a circulating water break which is conservative for a flooding event.

No other safety-related equipment is located in an area which would be vulnerable to the environmental effects of a pipe break in the Turbine Building.

The only safety related structures adjacent to the Turbine Building are the Reactor Building and Radwate-Control Building.

A pipe break in a main steam or feedwater line in the Turbine Building would result in transitory pressurization of the corridors between the Turbine Building, Reactor Building, Radwaste-Control

Building, and Diesel-Generator Building.

Air and steam would be forced into these corridor s through openings in the south wall of the Turbine-Generator Building, and through the seismic gap between the Turbine Building, Reactor Building, and Radwaste-Control Building.

No compartmental pressurization analysis is required to determine peak pressures and temperatures in the corridors due to the large volume of the Turbine Building, and the fact that the metal siding and exterior doors into the Turbine Building are not leak-tight and are not designed to with-stand more than a minimal pressure differential, the peak pressures seen by the reinforced concrete walls of the Reactor Building and Radwaste-Control Building would not exceed the structural capacity of the walls.

The doors to the control room are low-range blast

doors, designed to withstand a pressure differential of 3 pounds per square inch, which is considered adequate to maintain control room habitability as discussed in 3.6.1.12.

It should be noted that the response to this question is directed towards the Turbine Building as a whole and does not cover the steam tunnel.

The response to question 10.11 will address this area.

e

WNP-2

(}.

010,13 3.6 For postulated pipe breaks, you have not provided the information required to determine:

1)

The mechanism which terminates the resulting blowdown; or, 2)

The period of time over which blowdown occurs.

Accordingly, for each postulated pipe break or leakage crack indicate the time over which blowdown occurs and identify the mechanism which either terminates the blowdown or limits the amount of blowdown flow.

These mass and energy flow rates will be used to evaluate the peak pressures and temperatures in compartments and structures following a postulated break of the high energy pipes inside these structures.

~Res onse:

(

Except for the main steam isolation valves which terminate blowdown flow from the reactor building side of pipe breaks in the main steam line, and check valves in the reactor feedwater lines, which terminate blowdown flow from the reactor building side of pipe breaks in the reactor feedwater lines, no mechanism terminates flow except exhausting of the inventory of fluid in the line following the pipe break.

Where blowdown flow is not automatically terminated by isolation valves or check valves as described

above, the duration of the blowdown event as the inventory of fluid in a line is exhausted is not, considered in the analysis of peak compartmental pressure and temperature.

To evaluate the peak pressures and temperatures in compartments and structures following a postulated break of the high energy pioes inside these structures, the blowdown analysis is extended far beyond the initial transient until the blowdown flow becomes steady or decreases continuously.

The duratioo.. of'he analysis is therefore sufficient to.correctly predict the peak pressures and temperatures in these compartments and structures.

For a postulated pipe break or leakage crack in the main steam lines outside primary containment, the flow from the reactor side of the break is terminated by the closing of the main steam isolation valves in each of the four main steam lines.

The main steam isolation valves start to close at 0.5 seconds after the break and are fully closed at or prior to 5.5 seconds after the break, as given in Table 15.6-6.

I,

HNP-2 For a postulated break or leakage crack in the reactor feedwater lines outside primary containment, the flow from the reactor side of the break is terminated by the closing of'he check valves in each of the two reactor feedwater lines.

The check valves start to close when the direction of the flow reverses, and the flow from the reactor side of the break is therefore terminated within a fraction of a second.

HNP-2 You state in Section 3.6.1.1.1 of the FSAR, that fluid piping systems which the staff would classify as high-energy lines are considered by you to be moderate-energy systems if:

(1) their fluid temperatures are below 200oF and; (2) the fluid pressure is generated by centrifugal pump instead of a fluid reservoir.

(The staff classification system states that the fluid temperature must be less than 200oF and the fluid pressure must be less than 275 psi for a system to be designated as moderate-energy.)

Accordingly, demonstrate that these sytems do not contain enough energy to cause pipe whip.

Additionally, provide justification for your analysis of flooding based on the moderate-energy crack criteria rather than basing your analysis on the full break required by the high-energy break criteria.

~Res onse:

The energy of the blowdown fluid from a break in a pressurized fluid system is a function of the pressure at the exit plane, the mass flog rate and the area of the fluid jet.

The blowdown process of the 200 F

water from high pressure to atmospheric pressure can be considered as adiabatic.

Since the water is subcooled, it will not flash during the compression transient of the blowdown.

Therefore, the water jet remains in the liquid phase and behaves like an incompressible fluid.

At the beginning of the decompression transient, immediately following the break, a decompression wave is formed and travels through the fluid at sonic velocity (approximately 5,142 ft/sec) to the pressure source which, in this case, is the centrifugal pump.

Due to the reduction in required

head, the flow rate accelerates rapidly increasing accordingly to the characteristics of the piping system until a new equilibrium is established.

Since the system operating pressure is derived solely from the centrifugal pump, the complete system is depressurized after the break and the energy supplied to the pump is completely transmitted to the fluid in terms of velocity head.

For an incompressible fluid in an open

system, the energy of the water jet is proportional to the velocity head only.

Hence, the thrust of the water jet from a break in this class of piping systems may be calculated at the exit plane of the jet using the following formula:

F =

AV /g.

MNP-2

>ihere F

= jet force normal to target, Lbf p

= fluid density, Lb /ft3 A = flow path cross-sectional area, ft2 V

= velocity of jet, ft/sec g = 32.179 Lb

- ft/Lb

- Sec 2 c

f iNaximum jet thrust may be obtained at the exit plane of the pump.

Assuming a friction coefficient of 1.5 to include only contraction and expansion

losses, two representative examples representing the highest head and highest flow cases in the plant are presented below to demonstrate that this class of high pressure systems do not cause pipe whip events if the system pressure is. derived from a centrifugal pump.

Piping Designation Pipe Schedule Operating Pressure Operating Temperature Jet Flow Jet Force Jet Pressure at Break Plane CRD Pum Dischar e

2" CRD (2) - 4 160 1,439 psig 100 F

284 gpm 50 Lbf 22.3 psi Condensate Pump Discharoe*

20" Cond (2) -

1 40 142 psig 109.4 F

19,000 gpm 1,800 Lbf 6.5 psi

By pressure and temperature criteria, this is classified as a

moderate energy break.

However, in question 110.18 the NRC requested consideration of condensate piping as a high energy system.

For the maximum 10'pan which exists between the pipe supports the above pressures result in pipe stresses which are below the minimum for formation of a plastic hinge and thus pipe whip will not occur.

I V

Therefore,,for piping systems with system pressure derived from a centrifugal pump, the system is treated as a moderate energy system and flooding analysis is performed based on postulated flow from a controlled leakage crack.

h

bZfP-2 P'age 1 of 2 Q. 010. 15 (RSP) be require that you modify the main steam line isolation valve leakage control system (HSIV-LCS) to satisfy the staff pos-itions contained in Regulatory 'Guide 1.96, Rev.

1, "Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Nater Reactor Nuclear Power Plants",

June 1976.

Speci-fically, we require that:

a)

The design of the 'iSIV-LCS permits its actuation within 20 minutes after a postulated loss-of-coolant accident.

b)

The leakage control system for the. valve stems on the main steam line be designed to the same standard as the HSIV-LCS, and c)

Operation of the HSIV-LCS during normal plant operation be prevented by inter-locks capable of functioning after a postulated single failure.

in the inter-locking system.

'Response:

'a) b)

See revised page 6.7-2.

See revised page 6..7-11.

Also: 'direct response)

Stem packing leakage from the outboard main steam isolation valves is directed

~o equipment. drain funnels located in the steam tunnel.

The leak-off piping is classified Nuclear Class 2 up to and including the first manual block valve.

As stated in 6. 7. 3m, leakage from the packing seals large enough to pressurize the steam tunnel 'and blow out the water seal traps in the equipment drain system would vent. into areas of tne reactor building for subsequent processing by the standby gas treatment system.

Refer to Figure 3.2-2',

Zone G2, which depicts the stem packing leakage piping.

c)

Refer to qoestion 031.076 response.

draft page attached

PL%C A

b.

The HSXV-LCS and necessary subsystems a'e cap-able of per orming their safety function, when necessary, considering the design basis LOCA effects including:

(1) internally generated missiles,'2) the dynamic effects associated with oipe whip and jet forces from the event and (3) normal operating, and accident-caused local environmental conditions consistent with-the event.

Ce The MSXV-LCS is capable of performing its intended function

. fo13.owing any single active component failure (including failure of any one of the main steam line isolation valves to close).

The MSXV-LCS is capable of performing its in-tended function fol3.owing a loss of all off-site power coincident. wiA the pos"ulated design basis LOCA.

e. 'he MSXV-LCS 's designed w'th sufficient capa-city and capability to control the leakage from the main steam lines consistent with containment ntegrrt~derWhe conditions assoc~ed with e postulated design'>asis LOCA.Y' l

f.

The MSXV-XCS is manually initiated and is

'esigned to permit actuation '.

q+ <ng +i~8 following the postu-

~~

~e~~con~4ent~iW~-oa~~~>re-th ze~

C 4

Xnst~entata.on d controls necessary for the functioning of the HSXV-LCS are designed in ac-cordance with standards applicable to nuclear plant safety-re3.ated instrumentation and control systems.

The HSXV-LCS controls are provided with inter-locks actuated from appropriately designed safety systems or circuits to prevent inadver-tent MSXV-LCS operation.

6.7-2

Steam leaks into the steam tunnel escape the steam tunnel through the equipment drain system and are directed to the reactor building where the radioactive gases are subsequently processed by the standby gas treatment system.

The MSIV-LCS does not process MSIV stem packing leaka Stem packing leakage from the zqyin steam isolation valves is directed to equa.pment drain funnels located in the steam tunnel.

These equipment drains are routed to the reactor build-ing equipment drain sump.

Low leakage from the stem packing would condense in the piping to the equipment drain.

Leakage large enough to pres-surize the steam tunnel and blow out the water seal traps in the equipment drain system would vent into areas of the reactor building for sub-sequent processing by the standby gas treatment system.

All interconnections between MSIV-LCS and other plant systems do not affect the intended function of the MSIV-LCS.

These interconnections and their safety related actions are as follows:

(1) Inlet 14"MSLC(2) -4 lines for each inboard main steam isolation valve share common 14"MS(9)-4 drain lines.

Motor operated drain valves MS-V-67 A through D close automatic-ally by the containment isolation system on a scram signal.

Thus these lines would be isolated prior to placing the MSIV-LCS in operation after a LOCA.

(2) Inlet 14"MSLC(3) -4 line shares the out-board main steam line isolation valve drain header 3 "MS(20) -4.

Motor operated valve MS-V-20 isolates this header from 14"MSLC (3) -4.

This valve is only used during reactor startup to warm up the main steam lines t'o the turbine.

During normal plant operation it is closed.

Isolation of this valve is, therefore, ensured during a loss of coolant accident and subsequent utili-zation of the MSIV-LCS system.

6.7-11

WNP-2 010.16 9.0 Identify all safety-related equipment that could be exposed to, or affected by, dust storms.

Describe how you propose to assure the proper functioning of this equipment during dust storms.

Provide a description of the methods which will be used to prevent the blockage of vital air supplies to safety-related equipment {e.g.,

clogging of the air filter of the Diesel Generators).

In your response to this question, provide a cross-reference to your response to 372.8.

~Res onse:

l.

Essentially all safety related equipment that could be affected by severe dust storms are contained within plant areas served by the HVAC systems for the reactor building, control room/

cable spreading room/critical switchgear areas, standby service water pumphouses and diesel generator building.

The only safety related system exposed directly to severe dust conditions are the service water spray ponds.

a)

The normal air intakes for the reactor building and control room/cable spreading room/critical switchgear are located 130 feet and 85 feet above ground level, respectively.

At these intake locations the dust loadings will be 10 to 15 percent of ground level dust loads

{See Figure 2.3-5 and the response to guestion 372.8 for representative dust loads)

~

All intake air is processed through either automatic roll type filters or replaceable filter elements in the air handling units before entering the air distribution systems for the reactor building, control room, cable spreading

room, and critical switchgear areas.

An air washer is also included in the reactor building air handling unit.

Pressure differential across the filter units is annunciated when filter replacement is required.

With the intake locations and filtration, discussed

above, the amount of dust entering the reactor building, control room, cable spreading room and critical switchgear areas will not degrade the operating capability of safety related equipment in these areas.

b)

The standby service water pumphouses have unfiltered outside air intakes and some dust may be expected to enter the pumphouses during severe dust conditions.

The amount of

dust, however, should be limited since the pumphouse HYAC systems will be shut down during normal plant operations with the intakes and exhaust openings restricted by dampers.

1 of3

0

WNP-2 The dust loading of the air drawn into the pumphouses, when the HVAC systems are operating, should be less than maximum ground level dust loads since the air intakes are located above the service water spray ponds and feed into a plenum before entering the intake fan.

Any dust which enters will settle out within the pumphouses without blocking vital air passages.

c)

Entry of dust into the service water pumphouse will not affect the operation of safety related equiPment.

Any equipment that could be affected by dust is either provided as sealed units, located in dust proof cabinets or protected by dust proof coatings.

The diesel generator building outside air intake is located at grade level with air filters located within the building at 15 feet above grade.

The coranon air filter bank processes all ventilation air into the diesel generator building.

During a worst case dust storm, as defined in 2.3.1.2.1' 5.2, the maximum estimated dust load will be 8.9 mg/m3 for an 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months duration.

After particle impaction and re-entrainment (due to intake louvers) is accoun)ed for, the calculated dust load to the filters is 6.44 mg/m Without taking any credit for particle settling in reduced velocity area before filter bank the filter will be subjected to a maximum of 0.231 8/F2 of dust.

The filter bank consists of two (2) filters (prefilter and final filter) in series with a common pressure switch which will alarm when filters need changing.

The prefilte~ normal maximum resistance is 0.50" W.G.

(equal to 0.047 8/F ).

The final fil:ter normal maximum resistance is 1.00" W.G.

(equal to 0.142 ~/F2).

During severe dugt storm conditions the prefilter can be loaded to 0.129

></F~ or 1.00" W.G.

During the postulated dust storm an initial filter alarm would require a complete filter change followed by a maximum of two (2) prefilter changes as filter alarms are annunciated.

The 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months loading on the diesel air filters is calculated to be 2102.4 grams.

The capacity of the air filters is 5000 grams or 2 '78 times the severe dust storm loading.

This response is an elaboration of the response given in part (d) to guestion 40.26.

2of3

MHP-2 1

P d) The ultimate heat sink transient analysis was performed assuming 6" of sedimentation at the bottom of the spray ponds (see 9.2.5).

In addition, no credit is taken in the analysis for the volume of viater within the sand traps which prevent sedimentation from being swept into the pump pits.

3of3

0

MNP-2 Q.

010.17 9,1.2 The design of your spent fuel rack includes a neutron absorbing material encapsulated in stainless steel.

However, recent experience at some spent fuel pools has shown that the stainless steel cladding may bow out due to the internal pressure of gases generated by the irradiation of the neutron absorbing material in the spent fuel pool.

This bowing of the steel cladding has caused the spent fuel assemblies'o become lodged in the spent fuel racks.

Accordingly, describe the method (e.g

~, venting the stainless steel plates.to release any evolved gases) you propose to prevent this from occurring in the WNP-2 spent fuel pool.

~Res onse:

Bowing of the steel cladding is not expected'o occur since the neutron absorber plates utilized in the !(NP-2 racks have been shown through testing not to.offgas when irradiated by a gamma source.

These plates manufactured by Electroschmelzwerk

- Kempten (ESK) differ substantially in composition and manufacturing process from the type of plates which underwent decomposition at Connecticut Yankee.

Because of the nonoff-gassing characteristic of these plates, venting of the racks is not planned'or additional information on the offgassing tests, refer to the response to question 010. 18 (9. 1.2).

0

WNP-2 Q

010.1S

~9.1.2 In Section 9.1.2 of the FSAR, you list the test results involving radiation,

thermal, seismic and borated water testing of the boron carbide plates.

Oescribe the procedures used for these tests.

Alternatively, provide a cross-reference to any of these test proce-dures which have previously been accepted by the NRC staff on another application.

~Res onse:

When the FSAR was originally written, the manufacturer of the boron carbide plates had not been identified.

Accordingly, data from previously licensed plates was used based on a program description and results of the qualification tests conducted on boron carbide neutron absorber plates submitted to the NRC under the Connecticut Yankee docket 50-213, letter D.C. Switzer to R.A. Purple dated April 15, 1976.

Subsequently ESK was selected as the manufacturer of the plates.

With the exception of the fuil scale seismic test, essentially all described tests have been performed by ESK for the plates of their manufacture.

8ecause of the similarity in physical characteristics with the plates previously tested and because Nodules of Rupture tests show plates wi 11 withstand two times calculated seismic stresses, repetition of shaker table testing was not deemed necessary.

Test results for the ESK plates were submitted to the NRC under the Kewaunee docket 50-305 in a letter E.W.

James to V. Stello dated September 5, 1978.

As a result of our decision to use ESK plates, section 9.1.2 is being revised.*

See attached draft

bNP-2 d.

Shielding for the spent fuel storage arrangement is sufficient to protect plant personnel from exposure to radiation in excess of 10 CFR Part 20

. limits.

Since provisions for portable shielding are not provided in the drywell, administrative control is used during refueling operations to avoid overexposure of personnel as the result of a postulated fuel drop accident such as a

drop occurring on the reactor seal plate.

9.1.2.1.2 a 0 Power Generation Design Bases Spent fuel storage space in the fuel storage pool is for 2658 fuel assemblies b.

Spent fuel storage racks are designed and arranged so that uel assemblies can be handled efficiently during refueling operations.

9.1.2.2 Facilities Description 9.1.2.2.1 Spent =uel Storage Racks Spent fuel storage racks provide a place in the fuel pool for storing the spent fuel discharged from the reactor vessel.

They are top entry racks, designed to maintain the spent fuel in a space geometry that precludes the possibility of criticality under both normal and abnormal conditions.

This is accomplished with the aid of neutron absorbing plates.

The location of the spent fuel pool within the plant is shown 'n Figur'e 1.2-6.

The spent fuel storage ack design, shown in Figure 9.1-2, consists of fuel storage cells which are sauare stainless steel tubes w'th neutron absorbing B4C plates between them.

A stainless steel'late grid at the top and the bottom of the tubes, to which the tubes are welded, form the tubes into racks and maintain center-to-center spacing between the tubes at 6.5 inches.

The racks are welded together into modules which are held firmly in place by seismic restraints attached between the rack modules and the pool wail.

The storage racks are made of sta'nless steel.

The sauare tube storage cells a

e 1/8 inch thick.

The neutro bs olates are 0.21 inches gbic~~~om-posed of B4C powder bonded g~he~4~rm a plate with un' 4

g

<~ 2 n sert 8 wi the remainder beir(

bander and voids.

The plate has been s

n by tests to be chemically inert in water and ther 'ally over the range of pool water temperat

~

that can occ r.

The plates a

seal welded in th vity betw en tug s

o p ent w

er 'rusion, wit a k and plate dim sionp'pecified to eclude the pla tripp ng past

/'ac otn ".

There are no ad bearing rec iremenks for the pla es.

Plate]integr'ty and meC areal p ope"tie] have Seen verifie by a compr ensive+est pr which i clud se'smjc test'n a

frequencies f om 7 to 33 Hz, er al-cycljng from r temperature th oug 350 F, scat 'or 16 ys in 200

="

lutions of bor' acid and d WW~ed water, and amma ir iation of a roximatel 2 x 1011 rads.

The pp tests sh ed no swelling or weight loss, no cracking or dime onal changes and verified the mechanical oroperties a

s d in the desi n.

Differen rack sizes are used (12 x 16 12 x 13 f 8 x 13, 7 x 18 and 11 x 16 ar ays) to take full advantage of the fuel storage space in tne pool (see F'gure 9.1-3).

The upper rack structures are welded to an elevated base plate which, in turn, is supported by a system of welded beams and stirfeners.

The base serves to support the weight of the fuel assemblies and to distribute the load on the pool floor.

The base plate contains an opening at each fuel assembly storage location which accommodates the fuel assembly lower nozzle.

Natural circulation of pool water flows upward through the lower nozzle and the fuel assembly to remove decay heat.

The storage cells are des'gned to prov'de lateral support for the stored assembl'es.

The seismic restraints are stainless steel turnbuckles lo-cated between the pool walls and the racks around the periph-ery of the pool (Figure 9.1-3)..

They are located at both the top and bottom of the rack and, once adjusted will trans-mit the seismic forces of the OBE and the SSE between the racks and the walls and remain functional.

The turnbuckles are connected at the wall to stainless steel bands which are embedded in the concrete wall and seal welded to the pool liner.

9.1.2.2.2 Spent Fuel Storage Pool The spent fuel storage pool is designed to withstand earth-quake loadings as a Seismic Category I structure.

Xt 's a

reinforced concrete structure completely lined w'th stain-less steel, which provides a leakproof membrane that is re-sistant to abrasion and damage dur"'ng normal and refueling operations.

The stainless steel liner plates are seamwelded 9.1-9

+sad

~

d d

The neutron absorber plates have nominal dimensions of 19 'inches long, 5.88 inches wide, and 0.2 inches thick.

They are composed of B<C 1 dd~

1"'d 11 1

form properties.

They have' nominal B

loading yf 0.0959 grams per 10 square centimeter of plate and a plate density go 0.05 1bs/in The 3

QL/'e n

/f

)e cert"o~ie~~

plate has been shown by tests t Qn wa~e and thermally stable over the range of pool water temperatures that can occur.

The plates are seal welded in a stainless steel cavity to prevent water intrusion.

There are n'o load bearing requirements for the plates.

Based on the results of the Modulus of Rupture tests, the plates will withstand approximately two times the calculated stresses caused by a postulated seismic event.

Plate integrity and mechanical properties have been verified by comprehensive tests.

These tests included Modulus of Rupture and Modulusd of Elasticity tests.

The Modulus of Rupture testing was

&re+

performed using a ~ point support method and was done on specimens at temperatures varying from ambient to 300 F, specimens soaked in water, 0

and irradiated specimens.

The Modulus of Elasticity was performed using a resonance procedure and was done at varying temperatures and after the plate had been irrmersed in water.

The tests showed no swelling, cracking or dimensional changes and provided verification of the plate mechanical properties required for the rack design.

In addition to the mechanical

tests, extensive irradiation induced offgassing tests have been performed using gamma sources.

These test results clearly indicate that the amount of offgassing is negligible and ut//

not cause

. rack distortion.

lNP-2 010. 19 RSP (9.1,.2)

In Section 9.1.2.3.3 of the FSAR, you state that the interlocks which prevent the 125 ton crane in the reactor building from traversing the spent fuel pool, are occasionally by-passed.

This by-passing is unacceptable.

Accordingly, we require you to modify your procedures so that the interlocks on the reactor building crane prevent the crane from traversing over the spent fuel pool whenever there is spent'uel in the pools

~Res onse:

Regulatory Guide 1.13, c.3 allows for movement of loads necessary for fuel handling over the spent duel.

Occasionally it is necessary to operate the reactor building crane over the spent fuel pool in conjunction with maintenance of fuel storage and fuel handling facilities, or other activities associated with fuel handling and storage.

Therefore it is necessary to retain the ability to bypass the interlocks and use admini-strative control procedures under those conditions.

Movement of objects in excess of the rack design drop load (one fuel assembly at four feet above the top of the fuel rack) will be 'prohibited.

The electrical interlocks are bypassed only by actuation of a cab-mounted key-lock switch.

See revised section 9.1.2.3.3 Appendix c.3 page 11, and revised FSAR Figure 9. 1-17 which shows the interlock-controlled restricted area for crane travel over the spent fuel pool.

.~9.1.2.3.3 Spent Fuel and Cask Handling m

~ 4 The 125 ton reactor buildin crane traverses the full length ox the refueling oor evel the reactor building.

The design of the e-fueling floor pxovides aisles on both-sides of the fuel pool for moving components past

{and not: over) the fuel storage pool.

~

4 I

9 ~far lock>

o~ the r cac74r g~;lgi~q c.r"organ pr even+

+reve~

dv'e~

7 ~< SP<< ~ ~~~

P e

'+pew(ucC -

C4 &0j/~+ j'g5~;c Wc'cj a reo For <ra nw

/rave,l i5 pic claw iw Fight w f ~ I l7

~

"c55dc( do (L modem >/- I 5 ut gcc'554~$ * @Pe<<d~

the C<<n~

FQe(

pOO(

6/'CP I ~ C~J~+<~~

~ I~ Pc~v ac,goCI'cPcl c i'f4 f'ar <

o Ccu 5'r.~S c hen P4c',-~ ~er. (dcS

<re kyg<55'<~<

C~~(S arC usual

prevo'n f-7'h e C'run w carry i~g loacls 7 4<8 a~e vro v n ececgary v6v-v uN 1 ha~J(i or 5+r gyp anA ~&I'eg

<re iW eycagg oF

~A, I ac/a c7gsy~,alrop loacl (~c Fvsl c55ee'r 'hip ah 7 ne 4p OF 7'he 7~6 I'CCCk).

5 e'~

~

f.I.Z.S 2. c

~

Transfer of fuel assemblies between the reactor well and the spent fuel pool is performed with the refueling platform (see 9.1.4.2.10.2)

The fuel grapple or the auxiliary fuel hoist may be used, depending on the transfer operation.

The grapple and hoist: are provided with load sensing and limiting devices designed to the following limits:

Fuel Grapple (lbs)

Auxiliary Fuel Hoist (lbs)

Load limiting switch 1200 1000 Load sensing switch 485 485 Stall torque or hoist syst: em 3000 3000 The load limiting features of the refueling platform grapple and auxiliary fuel. hoist will prevent damage to the fuel racks if a fuel assembly accidently engages a xack while being lift.ed.

These load limits provide a redundant safety feature since the fuel handling g apple is not lowered below the upper fuel rack and is designed to interface only with the fuel bail.

Thus, the possibility of inadvertent direct lifting of the racks with the grappLe is precluded.

Guard rails around the spent fuel pool prevent the falling of fuel handling area machinery into the pool.

Other objects that could conceivably fall into the pool will not transfer energy amounts.exceeding the specified limits of the fuel.

~

racks.

f~

r

Regulatory Guide 1'.13, Rev.

1, December 1975 Spent Fuel Storage Facility Design Basis Compliance or Alternate Approach Statement:

NNP-2 complies with the intent of the guidance set forth in this regulatory guide by an alte nate approach.

General Compliance or Alternate Approach Assessment:

A controlled leakage building is provided enclosing the fuel pool.

The building is not designed to with-stand extremely high winds, but leakage is suitably controlled during refueling operations.

The building is equipoed with a ventilation and filtration system which is designed to limit the potential consequences of the release of radioactivitv specified in Regulatory Guide 1.25 to those guidelines set forth in 10CFR100.

The movement paths of heavy objects such as the reactor pressure vessel

head, containment vessel head and the spent fuel cask are designed not to pass over the spent fuel pool.

Furthermore, the reactor building crane and its auxiliary hoist are prevented by means of interlocks from passing over any of the spent, fuel pool except the spent fuel cask area.

g~p~~g;~~

i&ifdrlockg I'5 pc'i ~i f1c(

out!~

c(u,ring Fide(

Ra~diIn~ anu'*rape,

~/)~re Ams

<rnu(

aHvn,-~,-s f-ra 4 u'c/'y c'~ ~~8'I <H.

Although all of the spent fuel cooling and cleanup system equipment is not Seismic Category I, the source of emergency makeup water is from the Seismic Category I standby service water system.

The fuel pool is designed so that no pipe break will drain water from the fuel pool.

Specific Evaluation

Reference:

Refer to 9.1.

C.3-11

IP 0

WNP-2 0

010. 20 RSP (9.1.2)

In Section 9.1.2 of the FSAR, you state that a portion of the fuel handling building above the refueling floor is constructed of sheet metal.

Accordingly, we require you to demonstrate that the spent fuel pool is housed in a Seismic Category I structure which can withstand the impact of tornado missiles.

~Res onse:

Table 3.2-1 states that the Reactor Building is designed to Seismic Category I requirements.

Section

3. 5. 1.4 states that Seismic Category I

structures are designed to include the effects of missiles generated by the design basis tornado.

Section 3,8.4 provides details of the design features of the Reactor Building and spent fuel storage pool.

Tables 3.8-15 and 3.8-16 provide the load combinations and load factors used in design of Seismic Category I structures.

Section 3,5.1.4 '.a discusses the tornado missile-resistant design features of the Reactor Building.

Section 3.3.2 discusses the design features of the Reactor Building for tornado wind loading.

As stated in 3.5.1.4.l.a and 9.1.2.3.5, which reference GE Topical Report APED-5696, the design basis tornado missile for the refueling floor has been evaluated and found to not have sufficient energy to damage the spent fuel or the equipment and structures in the pool.

HNP-2 010.21 9.1.3 Provide a cooling system and a source of makeup water for the spent fuel pool which are both designed to seismic Category I criteria in accordance with the staff positions contained in Regulatory Guide 1.13, Revision 1, "Spent Fuel Storage Facility Design Basis,"

December 1975.

~Res onse:

MNP-2 has a seismic Category I source of makeup water for the spent fuel pool from the seismic Category I standby service water system.

This is shown on Figure 9.1-4 and stated in section 9.1.3.3.

Cooling-under

-,emergency conditions for the fuel pool is supplied by evaporation of pool water.

Reg.

Guide 1.13, Rev.

1, makes no specific statements about requiring a seismic category I spent fuel pool cooling system, As a result, WNP-2 meets the applicable criteria of the Reg.

Guide and the intent of the question.

However, fur ther evaluation of the design in this area is ongoing due to the interaction of fuel pool cooling and post-LOCA secondary containment pressure-temperature response.

(See the response to question 312.18)

NNP-2 010. 22

~

9.2.1 Identify which valves are. used to isolate that portion of the plant service water system which is not designed to Seismic Category I criteria from that portion which is designed to these criteria.

Provide a failure modes and effects analysis for the plant service water system, assuming a seismic event has occurred.

~Res onse:

The plant service water system (TSW) is not required for safe shutdown and accordingly is not designed to Seismic Category I requirements.

A failure modes and effects analysis is not considered necessary.

The portions of the TSW system piping in the Reactor Building have been designed to Seismic Category I requirements so that they will not fall and damage safety related equipment.

The standby service water system (SW) is used for safe shutdown and is designed to Seismic Category I criteria.

The SW is discussed in 9.2.5.

The plant service water system (TSW) and the standby service water system (SW) are independent systems and are not connected, therefore there are no valves which are used to isolate these systems from each other.

10. 23 RSP (9.2.S)

Provide the results of your analysis of the capability of the ultimate heat sink to absorb heat over a thirty-day period following a

postulated design basis accident.

Indicate the total heat absorbed in the ultimate heat sink, including the sensible

heat, the station auxiliary system heat, and the decay heat released by the reactor core.

In particular, provide the following information in both tabular and graphical formats:

a.

The total integrated decay heat.

b.

The heat rejection rate and the integrated heat rejected by the station auxiliary systems, including all operating

pumps, ventilation equipment, diesels and other heat sources.

c.

The heat rejection rate and integrated heat rejected due to sensible heat removed from the containment and the primary system.

'd.

The total integrated heat rejected; i.e., the sum of the Items (a),

(b) and (c),

Additionally, provide the following information:

e.

The maximum allowable temperature of the inlet water taking into account the rate at which heat must be removed, the cooling water flow rate, and the capabilities of the respective heat exchangers.

f.

The required and available net positive suction head (HPSH) at the suction lines of the service water pumps at the minimum water level of the ultimate heat sink.

This analysis should demonstrate the capability of the ultimate heat sink to provide:

(1) an adequate water inventory; and (2) sufficient heat dissipation which will limit the essential cooling water operating temperatures within the design ranges of system components.

In this regard, we require you to use the methods contained in Branch Technical Position ASB 9-2 "Residual Decay Energy for Light Water Reactors for Long Term Cooling," when evaluating the residual decay energy release rate from the reactor core due to fission product decay and heavy element decay.

Assume an initial cooling water temperature based on the most adverse conditions possible during normal operations.

The meteorological conditions shoul.d be established following the

guidance contained in Position C.l of Regulatory Guide 1.27, Revision 1, "Ultimate Heat Sink for Nuclear Power Plants,"

triarch 1974.

~Res ense:

See revised Section 9.2.5* which provides the revised results of the analysis including the information requested above.

This revision also addresses the concerns of guestion 371.6.

See attached draft pages

WNP-2 V

The worst storm of these was storm No. 3.

While it was also shown in this study that once a given dust storm terminated,

'there existed a

5% probability that another one would occur within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months and a

50% probability that another one would occur within 30 days, none of the above six worst case dust storms had occurred within 30 days of each other.

Most had occurred in different years during the 1953-1970 study period.

The dust loading for storm No.

3 is conservative in terms of its being considered as the worst case storm for use in plant design evaluations.

As a result of the shorter storm dura-tions of the measured August 11, 1955, January 11,

1972, and April 1972 dust storms, their time integrated dust loadings at 5-6 feet above the ground are not worse than that computed for storm No.

3 2.3.1.2.2 Design Snow Load The American National Standards Institute (ANSI) in "Building Code Requirements for Minimum Design Loads in Buildings and other Structures" (19) provides weights of 100-year return period ground level snow packs for the site region.

The ANSI value of 20 pounds per square foot was used as the design snow load for all WNP-2 structures.*

Assuming a snow density (specific gravity) of 0.1 or 6.24 lbs/ft3, this design value corresponds to a snow depth of 3.2 feet.

The above snow load.. is conservative for the site as snow depth seldom exceeds six inches, and the greatest depth of 21 inches was recorded in February 1916.(4)

The weight of the 48-hour probable maximum winter precipitation can be determined from

.the data presented in Table 2.3-3.

Since the greatest snow-fall in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was 7.1 inches (January 1954) and a record depth of approximately 12 inches lasted four days (December 1964) these depths would correspond to snow loads of 3.7 and 6.24 lbs/ft2 respectively.

2.3.1.2.3 Meteorological Data Used for Evaluation of Ultimate Heat Sink~~'he meteorological data presented in Figures 2.3-7 to 2.3-9a and Tables 2.3-1, 2.3-5, and 2.3-7a-7h was used to evaluate the performance of the WNP-2 spray ponds in 9.2.5 with

Ice loading is included in this WNP-2 estimate.

- The teorolog' data used fo evaluation of e

UHS present her is urrently u e

oing review or com-pliance wx

.G.

1.

, Rev.

2.3-19

charge header of the pumps stops the jockey pump and starts one of the two main pumps on an increase in demand of system flow.

Upon a further increase in flow demand, above 140

gpm, the flow meter automatically starts the second pump.

When the flow demand decreases below 140 gpm, the second main pump stops.

If flow demand continues to decrease to below 50 gpm, the jockey pump starts and the main pump stops.

During the starting sequence if one pump fails to start, the sequence automatically continues to the next pump and a

local alarm and light indicate pump failure.

Upon indication of low potable water storage tank level, all pumps stop.

The reactor building potable water booster pumps are auto-matically cycled on and off by a pressure switch in the pressurizing tanks on the pump discharge in order to maintain header pressure between 20 and 50 psig.

All electric water heaters are thermostatically controlled to maintain the tank at the desired setpoint.

The hot water circulating pumps in the service building and radwaste building are cycled by a thermostat with sensor in the hot water recirculation line set to maintain the loop at a

minimum setpoint.

9.2.5 ULTIMATE HEAT SINK 9.2.5.1 Design Bases a.

The ultimate heat sink, a spray pond system, supplies cooling water to remove heat from all nuclear plant equipment which is essential for a safe and orderly shutdown of the reactor and to maintain it in a safe condition.

b.

The ultimate heat sink is capable of accomplishing its safety function for a normal cooldown or an emergency cooldown following a loss of coolant accident without the availability of off-site power.

The sink provides this cooling capability for a period of 30 days without outside makeup.

Provisions are made for replenishment of the sink to allow continued cooling capability beyond the initial 30-day period.

The sink will, accomplish its safety function despite the occurrance of the most severe site related natural events including earthquake,

tornado, flood drought.

or

9. 2-l6

WNP-2 The following worst. month meteorological data were used in 9.2.5 to establish the second through thirtieth day worst pond thermal performance and worst 30 day drift loss and evaporation(2~):

1)

July 9

August 8, 1961 at HMS, presented in Table

2. 3-7g (minimum heat transfer) 2)

July 2 August 1, 1960 at HMS, presented in Table 2.3-7h (maximum evaporation and drift loss)

I Diurnal -variations in dry bulb and wet bulb temperatures for both 30 day periods assumed that the hourly temperature vari-ation approximated a sine wave of one cycle in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (21).

The average wind speeds during both 30-day periods was approximately 5.5 mph.

The highest daily average wind speed for the 30-day mass loss eriod is 10.3 mph.

2.3.2 LOCAL METEOROLOGY

2. 3. 2. 1 Data Comparisons The local meteorology at the WNP-2 site can be descried from FSAR meteorological data procured during the period April 1, 1974 to March 31, 1976 from the permanent onsite 7 foot and 245 foot meteorological towers.

Data collected from the 245 foot WNP-2 tower have'een used for the short term (accident) and long term (routine) diffusion estimates.

Onsite meteoro-logical data were also obtained from'a temporary 23 foot, tower which commenced operation in April 1972 for the purpose of determining optimum cooling tower geometric orientation for performance during high wet bulb periods.

The 23 foot meteor-

~ ological tower data were also used with other regional data to establish the potential impact of proposed mechanical draft cooling tower atmospheric releases in the vicinity of WNP-2(22).

The permanent tower data have been compared where appropriate and possible, with simultaneously recorded and historical

.data obtained from the Hanford Meteorological Station (HMS) for the purpose of documenting the representative'ness of the two years of onsite meteorological measurements.

For the months of April through August 1974, 'comparisons have also

biP-2 c.

The ultimate heat sink is designed to satisfy the regulatory requirementq of Regulatory Guide 1.27 (Rev.

1). Z~ >pp~W < c ~d.

Scuk<ow

<-3-l 4 ~ 5 o

System Description

9.2.5.2 I

During all normal operating c ndt', including startups and normal shutdown, waste heat fr the reactor auxiliaries is transferred to the circulating water system.

Heat from this system is in turn rejected to the atmosphere by the normal plant cooling tower system.

Following any event that would prevent the use of the plant cooling towers, the heat rejection duties are transferred to the spray ponds.

The ultimate heat sink consists of two concrete ponds with redundant, pumping and spray facilities.

The pond and pumphouse arrangements are shown on Figure 9.2-11.

The ponds and pumphouses are designed to Seismic Category I requirements.

Standby service water (pSW) loop A draws water from pond A, cools the Division I eaui ment required for safe

shutdown, and discharges ~ the spray ring in pond B for heat dissipation.

Similarly, gSW loop B draws water from pond B, cools Division II equipment, an dischar es 'he spray ring in pond A.

The HPCS gSW system draws water from pond A, cools division III and dis-charges without spray into pond A.

A syphon between the ponds allows for water flow from one pond to the other.

. The spray system illustrated in Figure 9.2-11 consists of two annuli of spray trees one for'ach of the concrete ponds.

. Each annulus is 140.0 feet in diameter and contains 32 spray trees equally spaced (13.75 feet, between vertical centerlines) on the circumference.

The 'vertical trees are serviced by the annulus water pipe, 20 inches in diameter, mounted above the water level.

The annulus pipe is fed by the main'header from each respective pumphouse.

Each spray tree consists of a vertical riser pipe or trunk 8 inches in diameter and 7 hori-zontal limbs of l-l/2 inch pipe.

The limbs are attached to the riser at 2'8" intervals of heights and are rotated at, 90o subsequent angles from each other so that the arms resemble a counter-clockwise helix with increasing height.

The arms radial to the annulus are 4'6-7/16" long.

The lowermost arm is a tangent arm.

The arms tangent to the annulus pipe are 3'6" long.

Spray nozzles are located at the end of each arm and are connected by fittings so that/he orientation of every nozzle is radially inward with an angle of 55 upward from horizontal.

The nozzles are 1-1/2-CX-27-55 Whirljet nozzles supplied by Spraying Systems Company.

Since each

9. 2-17

1 4

tree nozzle is located at a different elevation, each nozzle pressure is different.

The uppermost nozzle water pressure is 17.0 psig, and the total water flow from a tree is approximately 300 gpm.

The HpCS gSil flow, 1192 gpm, is treated as. a straight heat dump in the thermal analyses.

The combined water volume of the spray ponds is adequate to provide cooling ~ater for 30 days without makeup.

Although the pond is not used for cooling during'normal operation, some small losses are to be expected due to normal evaporation from the surface and occasional blowdown needed to maintain water chemistry.

A gravity makeup line is provided from the circulating wate pumphouse to the spray ponds to automatically maintain the pond water at the required level.

The ponds can also be supplied directly from the plant makeup water pumps (see 10.4. 5).

Design parameters for the spray pond are given in Tables 9.2-1 and 9.2-2.

A standby service water pump is located in each spray pond pumphouse along with its associated equipment so that an

.accident, such as a fire or pipe break associated with one pump would not affect the operation of the redundant pump.

~ )

~ 'i

~

s p~

~~ERR&~

g,

~ the pump sump to prevent heavy debris from entering the pump

~ sump area.

A skimmer wall and fixed screen prevent. floating debris from entering the, pumps.

I A spray ring bypass is provided so that the water temperature may.be 'controlled during cold weather operation.

When the pond temperature drops below approximately

60oP, the spray ring may be bypassed by opening the dump valve returning water directly to the pond.

Tp prevent. adverse operation during freezing weather, all gSW piping and components are either below the frost line, within the heated pump houses, heat traced, or, in the case of the spray rings,. kept drained by'he return 'header

~

dump when not in operation.

'lg

s. Aaw3 9, ~N QQC)'

Y4 pvpQ @44" voo (4 ~

~~

u<

O" ~~ m~<<a. 4a};M&o~uiD

>'P

~a

-.~ ~

<'~'4~)<<'~~<<~'leiV~5.

jVVt. P)P~~

S.L-'l2.

s R.t-i'S) JA~VX W'g Rg, All&A

~ ~'hiJV. ~'Rcrn 'Q+'u" ski)~ ass&;9ZO '. M4~a

,i,g<~~gQ tJf>k)p )~,LQ0pg A cvA(5i 5 AV.N)U> ~a3i').

id>)~,0Jnk

)9i Xle.

e-- lu-H-I d.~

'rQ

) (O~~g W98g M+a

~

I

ll ~

HNP-2 9.2-5'3 Safety Evaluation An oriented spray cooling system (OSCS) is utilized for cooling the water inventory of the ultimate heat sink.

OSCS has been developed as a result of intensive analytical studies and experimental verification over a period of more than six years.

Details of the OSCS experimental and analytical developmental efforts are described in Topical

Report, Oriented Spray Cooling System(OSCS) fo Ultimate Heat Sink Application (UHS), I-R 100 which has been submitted for Nuclear Regulatory Commission staff review.

The meteoro-logical data'for the UHS is discussed in 2.3.1.2.3.

The thermal performance model is based on the correlation of the Canadys test data descr'bed in Section 3.1 'of Topical

Report, I-R 100.

The resulting KAV/L for this application is 2.66.

This includes a

10% derate of the KAV/L to cover conservatively the data scatter experienced at Canadys.

Since the KAV/L represents the performance of the specified geometry and nozzle pressure, the KAV/L combined with the meteorological da ta are sufficient to de termine the sys tern

" cooling performance. ~f5: Wv a~ au4~cu 8:Va The system model forpthermal performance ~ mass loss analy-sis was based on the following assumptions:

4 a.

The pond contains total inventory upon onset of LOCA less 0.5 feet for sedimentation of the pond basin.

b.

Water losses result only from drift, evapora-tion of the sprayed droplets, and evaporation.

due to heat rejection on the pond surface.

c.

All th'e heat transfer is accomplished by evaporation, none of the heat transfer is accomplished by sensible heat transfer.

().GLvr~

The, fa.rstgdayyof the thermal performan na ysxs m the worst single record (Table 9.2-4 Page 1 of 3).

The ~~

rough thirtieth days are the average meteorological conditions of the worst 30 day period. of record

~

I (Table 9.2 4 Page 2 of 3). Fe'~ 0~~<.abS~>~

a~umC,:A~

'. ~o8 i4 ~E,q~+.&cw 9.4

~~

e.

The fix'st through thirtieth day of the mass loss analysis are the average meteorological condi-tions of the worst 30 day period of record (See 2.3.1.2.3 and Table 9.2-4 Page 3 of 3)>

Vh due to drift of of the spray flow.

The spray flow is ba ed. on continuous operation of one spray ring".

,SP /g

)

)~~~a 9.2-19

WNP-2 Off-site power is lost and Division 2

diesel. fails. to start, resulting in a loss of csee~ the ~ spray headers.

C

~'""'N()

C-)

'ink

'is comprised

products, decay he sensible heat from and heat, removed f in the two operati loads are tabulate decay heat from fission from heavy elements, he. reactor coolant syst m the emergency equip t

divisions.

These t

in Table 9.2-8.

e average ind speed during t s 5.5 mph.

e highest daily ind speeds re lt in drift los 23% of the spra

flow, respect'ss of.73% of th spray flow, emonstrates that t spray pon s

nventory to meet dri losses r

igher than expected.

iurnal psychrometric data ave a

1957-1970) for each month d

st Laboratories "Climatogra thirty day mass oss period erage is 10.3 m

These s of approxi ely

.11% and ely.

In as ing a drift he mass 1

s analysis contai ufficient water om wi speeds significantly g

over a 14 year period sented in Battelle North-of the Hanford Area",

anting e

1 for the month of July te

erature, 77.4oF.

ier, o analyses were run.

t excha er and the second The pond mperature is shown in 're 9.2-7.

OoF on the thi

'eth hour.

gure 9.2-8, and t mass

-3.

The diurnal os 'a-.

to the changes in so lded a peak pond tempera e

ur after the accident.

etween the two analyses is sion 1 or Division 2 power nitial pond temperature.

e a

ielded the highest initi pon sing the assumptions ated ea he first assumed a

ean RHR h ssumed a fouled t exchanger ransient for t first analysi he pond temp ture peaks at 8

he mass in tory is shown. in osses ar abulated in Table 9.

ions '- pond temperature are d

oad' The second analysis y'oF lower and at the same only significant difference a

Aw

~~5 analysis was conducted e

If the failure was postulated in Division 3 (HPCS) instead of Division 1 or 2, the peak pond temperature is ~%r lo~~Ii eCKSSw The Hpcs)sw flow is a straight heat dump; therefore, inasmuch as the spray pond is concerned, it raises rather than lowers the temperature transient.

9.2-20 I-psu, at>

~~I

~t0yt5 )C87~

0

!lNP-2 g.

The major heat loads considered are reactor core decay

heat, sensible heat from both the coolant and the reactor, fuel pool decay heat, pump work, and the heat removed from the station auxiliaries.

These heat loads are detailed in Table 9.2-8 and Figures 9.2-7b, -7c and -7d.

No credit was taken for heat sinks in the primary containment other than the suppression pool volume.

The actual average wind speed during the selected thirty day period for the mass loss analysis was 5.5 mph.

However, for conservatism, the -drift loss assumed in the analysis was based on five times the calculated drift value at the highest daily average wind speed of 10.2 mph.

The mass loss analysis thus demonstrates that the spray ponds contain sufficient water inventory to meet drift losses significantly higher than expected.

The analyses assume an initial temperature of 77 F.

This is approximately the highest monthly average temperature expected if the sprays are not operated.

To maintain the pond temperature below this limit, the spray headers will be operated and/or..r,iver.water make-up to the cooling towers will be diverted through the spray ponds.

Analyses have been performed which demonstrate that the above operations can maintain the spray pond below 77oF.

0

iI p~I~~ ~-,~4

'Sin~~~

T esulting peak SSW temperature, 87oF, predicted by the "worst ase" analysis is considerably below the 95oF S

temperatur~eassumed in. the analyses

'performed in 6

for containment heat removal.

The peak suppressi pool temp-eratures stated in 6.

and 6.2.2 are t efore conservative.

The SSW peak temperature,

ever, ex eeds the design bases SSW temperature used for HVAC pment by 2oF for a period of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

This increas as bee valuated.

Xt would result in a peak temper-ure for'hose s served by emergency HVAC e

~ment of, at most, 2oF hx, than was originally ulated.

However, since the temper re rise is sma nd exists for only a short period of time, x

as as sed as not being deleterious to equipment operation.

'Drift losses following loss of makeup to the ponds are con-trolled during two spray ring operation by bypassing the spray header on one pond whenever spray pond temperatures drop below approximately 80oF.

Continuous, simultaneous operation of both spray rings is not required after a LOCA.

Since the two gSW loops are redundant to each other, shutdown equipment when they determine that the peak temp-eratures have been past.

Zn addition, the difference between assumed and calculated drift losses for continuous operation of one spray ring, is more than 'adequate to account for drift losses from the operation of the second spray ring for several days after the accident.

Table 9.2-7 lists the available sources of makeup water to provide continued cooling beyond the initial 30-day period.

This table assumes that off-site power is restored within the 30 days.

No credit is taken for the water stored in the cooling tower basins.

However, it is expected that this water will not be instantaneously lost and will flow to the pond for the same period of time.

Table.9.2-7 also summarizes the effects of natural phenomena and'of a LOCA on the water supplies to the spray pond.

The possibility of a tornado pass'ng over the spray pond and removing a significant amount of water is considered a

credible event.

For this reason, the makeup water pump-house is designed to be tornado proof, with all piping and electrical power supply between the plant and the pumphouse 9.2-21

MNP-2 The resulting peak spray'ond:temperature, 88.6 F, predicted by the "worst case" analysis is considerably helot( the 95oF service water temperature assumed in the analysis performed in 6.2.1 for containment heat removal, adding further conservatism to the containment temperature and pressure transients therein presented.

The s~rvice water temperature,

however, exceeds the design basis temperature, 85 F, at the emergency reactor building and control room HVAC equipment for a short period of time as shown in Figure 9.2-7a.

This results in a peak temperature for some of the electrical equipment rooms served by emergency HVAC equip-ment of, at most, 3oF higher than the nominal lifetime rating for the equipment.

This has been assessed as not being deleterious to the equipment operation.

A sensitivity study was performed to determine the effect of the RHR heat exchanger effectiveness on the suppression pool and spray pond temperature transients.

The RHR heat exchanger effectiveness varies with the amount of fouling and with the flow rates.

RHR heat exchanger flows different from the rated values in Table 6.2-2 are anticipated only if the operator delays or fails to close the RHR heat exchanger shellside bypass valve as discussed in 6.2.2.3.

Anticipated variations in flow and fouling were determined to have essentially no affect on the spray pond temperature transient following a design basis LOCA, but were determined to have an impact on the suppression pool temperature transient.

The most severe postulated suppression pool temperature transient results from assuming a fully fouled RHR heat exchanger and no operator action to close the shellside bypass valve.

This suppression pool transient presented in Figure 9.2-7a, is slightly less severe than the suppression pool transient presented in 6.2.1 which assumed a steady 95oF service water temoerature and that the operator closed the RHR heat exchanger bypass valves.'he results of the mass loss analysis assuming an unfouled heat exchanger is shown in Figure 9.2-8 and is tabulated in Table 9.2-3.

The mass loss assuming a fouled heat exchanger is less

severe, but only by approximately 2,000 gallons.

0

NNP-2 underground.

Since it is not credible to assume an earth-quake coincident with a tornado, this system need not be Seismic Category X.

Two 12,500 gpm plan't makeup water pumps are provided, one powered from each emergency diesel gene a-tor.

Should pond water be lost duo to a tornado, one of these pumps will be started to provide makeup.

Valves are provided in the makeup water line to isolate the flow ~k the cooling tower and to ensure that it goes to the spray pond.

9.2.5.4 Testing and Inspection Requirements After completion of the spray pond, an inspection and test program has been established to ensure that the spray system will accomplish its safety function as discussed in 14.2.

All valves and piping in the system have been hydrostatically tested in the shop per ASME Section III, Class 3.

After in-

'tallation the system is hydrostatically tested and visually

~ inspected.

During plant operation the system is periodically tested.

Preservice and inservice inspections for the spray system will be in accordance with 6.6.

9.2.5.5 Instrumentation Requirements The spray pond is equipped with redundant level and tempera-ture sensors which are alarmed and indicated in the main control room as well as locally.

In the event that the spray pond level falls below the minimum level required for 30 days of cooling',

an alarm is sounded and makeup automatically is provided directly from the plant makeup water line to the spray pond.

High and low temperature alarms are provided.

In the event that the pond water temperature approaches the design limit, the spray system is initiated to lower the temperature.

Upon low water temperature signal, return water is dumped directly into the ponds to prevent spray trees and spray headers from icing o 9 '-22

NHP-2 TABLE 9.2-3'OTAL SPRAY POND MATER LOSSES AND CONTENT 30 DAYS AFTER LOCA EVENT Drift losses Spray evaporation Surface evaporation Total Remaining inventory 2,RvR) >bj I

I e~

gal k, lb') >7.l 5-y-R~).3" gal

. n~v,e~~

0~110 ~72,h

~~~ gal

~ r~~

9.2-36

TABLE 9'-4 DIURNAL VARIATION IN METEOROLOGICAL DATA (FOR WORST SlNCLE DAY OF RECORD USED TO ANAfYZE THE POND TllERMAL RESPONSE ef~~

BAY FOLLOWING LOCA)

Hour Dry Bulb

( F)

De+

Point

( F)

Wet Bulb

( F)

dauf5, V-Wind Speed (ss9ph)

Solar Radiation (~)

Noon ls00 p.sts.

2s00 3s00 4 00 Ss00 6s00 7s00 8's 00 9s00 10s00 lls00 Midnight ls00 a.rs.

2s00 3s00 4 00 5500 6s00 7 00 8s00 9s00 los00 lls00 100'1 103.09 105.20 105.71 104.93 102.48 101915 98.27 96.21 90'2 91.33 91 49

".90.91 85.92 84.24 80.61 80.24 78 27 83.25 86.77 90.64 92'4

'95. 23 98 ~ 32

59. 4).

59.69 58.91 56.00 54.,11 55.88 56.05 56'3 56.59 60.53 57.68 60.48 58.03 59.17 57.28 56.21 58;48 59.55 62.99 62.91 61.09 62.00 63.36 62.40 72.98 73.58 73.96 72.80 71.7&

71.81 71.50 70.68 70.27 70.57 69.31 70.77 69.35 68.39 66.88 65.14 66.21 66.15 69.65 70 '7 70.83

~

71.90 73.38 73.73

-~0 9~'7 1+?l

~Yc Mq

$. 26 gl5 11&.5.

8&f9 12 J.)9 9%+5 4'glk 5

10 3C77 5 +9 0 bb t',so 290.81 282,71 261,30 226.27 180.98 127.56 70.89 16.86 0.00 0.00 '

~ 00 0.00 0.00 0.00 0 F 00 0.00" 0;DO 16.86 70.89 127.56 180.98 226.27 261. 30 282.71 Data based upon 10 July 1975.

0

~ tfl

TABLE 9.2-4 (Continued)

DIURNAL VARIATIOV Ill NETEOROI.OGICAL DATA (FOR DAY X TllRU 30 USED TO ANALYZ E POOD THEit.LV RESPONSE fOLLOWIDG LOCA)

)tour Noon 1:00 p.m.

2 F 00 3 F00 4:00 sz00 6:00 7:00 8:00 9s00 10:00 11:00 Hidnight 1:00 a.m.

2 00 3:00 4:00 5 00 6:00 7c00 8$ 00 9.00 10:00 11:00 Dry Bulb

( F) 95.40 96.80 97.30 96.80 95.40 93.10 90.10 86.60 82.80 79.00 75.60 72.50 70.20 68.80 68'0 68.80 70.20(

72.50 75.60 79.00 82.80 86.60 90.10 93.10 DcM Point

( F) 45.9 46 ~ 1 46.1 46.2 46.2 46.0 45.6 45.6 45.6 45 ~ 2" 45.6 46.0 46.2 46.0 46.3 46.1 46.2.

45.8 4 6.. 0

'46.6 45.8 45.6 45 ~ 8 45 '

wet Dulb

(

E')

65.5 66.0 66.2 66.0 65.5 64.7 63.6 62.3 61.0 59.6 58.4 57.3 56.5 56.0 55.8 56.0 56.5 57.3 sa.4 59.6 61.0 62.3

~ 63.6 64.7 Wind +

Speed fnph) 5.50 5.50 5.50 5.50 F 50 5.50 5.50 5.50 5.50 5.50 5.50 5.50 50 s.so 5.50 5.50 5.50 5.50 5.50 5.50 5.50 5.50 5.'0 5.50 Solar Radiation f

)

hr 290.81 282.71 261.30

~ 226,27 180.98 127.56 70.89 16.86 0.00 0.00 0.90 0 F 00 0.00 0.00 0.00

.0 ~ 00 C.OO 16.86 70.89 127; 56 180.98 226.27 261.30 282.71.

Data based upon average values for the period 9 July -

8 August. 1961.

Q\\

TABLE 9.2-4 (Continued)

I DIURHAL VARIATIOH IH HETEOROLOCICAL DATA (FOR DAY 1 1RU SEO T AHALYZE MA S LOSS FOLLOWING LOCA Hour Dry Bulb

( F),

Dess Point

( F)

Wet Bulb

( F)

Wind Speed (mph)

Solar Radiation (~)

Hoon ls00 p.as.

2s00

.3s00 4s00 5s00 6s00

.7s 00 8s00 9s00 10s00 lls00 Hidnight ls00 p.sa.

2s00 3s00 4s00 Ss00 6s00 7 00 8:00 9'00 10s00 11100

96. 40 98.00 98.50 98.00 96'0 93o90 90.70

. 86.90 82.90 78.90 75.10 71.90 69.40 67.80 67.30 67.80 69'0 71.90 75.10 78.90 82.90

. 86.90 90.70 93.90'2.50 43.50 43.50 43.50 42.50 42.00 42.00 40.50 40.00 40.00 39.00 39.00 39.00 39.00 39.00 39.00 39 F 00 39.50 39.00 40.00 40.00 40.70 42.00 42.20 64.70 65.40 65.60 65.40 64.70 63.70 62.30 60.70 59.00 57 30 55.70 54.30 53.30 52.60 52.40 52.60 53.30 54.30 55.70 57.30 59.00 60.70 62.30 63.70 lbs>

LOgg'I

>b.S()

0

(."l

~50 5

SO S~r50 5%0, 5.50 5.(50 S. 507 5~50 5.50 S.s)q 5.50 50 290e81 282.71 261.30 226.27 180.98 127.56 70.89 16.86 0.00 0 00 0.00 0 ~ 00 0.00 0 F 00 0.00 0.00 0 F 00 16.86 70.89 127.56 180.98 226.27 261. 30 282.71 Data based upon average values for the period 2 July - 1 August 1960

'~M Sr

I

Table 9'-8 Heat Loads Rates Used in UHS Anal sis I.

Core Decay Heat Load See Table 6.2-11 Reactor Coolant Sensible Heat Load The energy (414 x 10 BTU referenced to 32 F) of the reactor coolant is accounted for by starting the suppression pool at 150oF.

III.

Reactor Vessel,

Piping, and Core Sensible Heat Load Time (hours)

Rate(10 BTU/hr) 24 t>

24 8.14 negligible IV.

Metal-Mater Reaction Time (hours) 1 t

1 Heat Load Rate (10 BTU/hr)

.47 negligible V.

ECCS Pump Mork Load Time (hours) 8 8

Rate (10 BTU/hr) 12.35 5.49 VI, HPCS (Div. 3) Service Mater System Heat Load Time (hours)

Rate (10 BTU/hr) 8 t >

8 8.73

0

YII.

Constant Div.

1 Service Mater System Heat Load Time (hours)

Rate (10 BTU/hr) 6 t>0 VIII, Fuel Pool Heat Load Time (hours) 0<t<10 10

< t < 20 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 t

> 52 18.18 Rate (xl0 BTU/hr) 0

.5

,54

.76 1.09 1.41 1.74 1.96 2.39 2.61 2.82 3.04 3.26 3.48 3.69 3.86 4.02 4.07 4.13

I

Table 9.2-8 (continued)

Notes:

(1)

Rejected initially to the suppression pool and subsequently transferred by the RHR heat exchangers to the UHS.

(2)

RHR pump 1.93 x 10 BTU/hr 6

LPCS pump 3.56 x 106 BTU/hr HPCS pump 6.86 x 10 BTU/hr (3)

HPCS system and HPCS SW system shut down after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

LPCS system and RHR loop A maintain long-term cooling.

(4)

HPCS service water pump work

.13 x 10 BTU/hr HPCS diesel coolers 7.40 x 106 BTU/hr HPCS coolers (Table 9.2-5) 1.20 x 10 BTU/hr (5)

Div. I Service water pump work 3.82 x 10 BTU/hr 6

Div. I Diesel Generator 11.69 x 10" BTU/hr Coolers and misc. equip.

(Table 9.2-5) 2.67 x 106 BTU/hr (6)

Excludes fuel pool and RHR heat exchanger heat loads (7)

Added to the RHR service water system

'able 9.2-9 Inte rated Heat Data - Mt<P-2 UHS Re-anal sis T1Ne After LOCA Min.

Q Deca+')

Q Sen Q Aux Q Aux 2 Q Aux 3 Q Total q

SH 10 BTU 0

1 2

4

'10 20 40 90 120(2H) 240(4H) 360(6H) 480(8H) 720(12H) 960(16H) 1200(20H) 1440(10) 2160(139) 2880(20) 4320(3D) 5760 4D 11520(80) 14400(100) 17280(120) 23040(16D) 28800(20D) 34560(240 43200(30D) 0 3.51 4.28 5.57 8.72 13.02 20.26 35.16 43.03 70.65 94.84 117.0 157.6 194.9 229.9 263.1 354.5 435.3 577.2 702.3 816.2 922.0 1116 1292 1456 1756 2029 2282 2635 0

. 014

. 027

.054

.136

.271

.543 1.22

1. 63 3.26 4.88 6.51 9.77 13.02 16.28 19.54 19.54 19.54 19.54 19.54 19.54 19,54 19.54

19. 54

19. 54 19.54 19.54 19.54 19.54

. 020

.041

.083

.205

.413

.823 1.85 2.48 4.94 7.41 9.88 12.08 14.27 16.47 18.66 25.25 31.84 45.02 58.19 71.37 84.54 110.9 137.2 163.6 216.3 269.0 321.7 400.8 0

. 030

~ 061

.121

.303

.606 1.21 2.73 3.64 7.27 10.91 14.54 21.92 29.39 36.86 44 '5 68.67 94.64 148.2 201.7

'255. 3 308.8 415.9 523.0 630.1 844.2 1058 1273 1594 0

.015

.029

.058

.146

.291

.582 1.31 1.75 3.49

' '4 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6.98 6'8 6.98 6.98 0

3.59 4,44 5.89 9.51 14.62 23.45 42.32 52.57 89.66 123.3 155.0

?08.4 258.6 306.5 352.8 475.0 588.3 796.9 988.8 1169 1342 1669 1979 2276 2843 3383 3903 4656 0

.174

.355

.719 1.83 3.75 7.80 18.69 25.57 54.51 84.37 114.3 172.4 227,3 279.3 328.6 461.1 581.1 796.6 995.4 1182 1358 1689 2001 2300 2870 3412 3935 4689

I 0

Table 9.2-9 (continued)

(1)

Q Decay (2)

Q Sensible (3)

Q Auxiliary 1 (4)

Q Auxiliary 2 (5)

Q Auxiliary 3 (6)

Q Total Integrated core decay heat rejected to suppression pool.

Integrated sensible heat rejected by the reactor

vessel, piping, and core to the suppression pool.

Integrated heat from ECCS pump work rejected to the suppression

pool, Integrated heat from auxiliary systems rejected to division 1 service water system.

This heat includes all sources of heat into division 1

SW system except for the RHR heat exchanger.

The RHR heat exchanger transfers heat from the suppression pool to division 1

SW system.

Integrated heat from HPCS service water system.

This heat is a straight heat dump into spray pond A.

Sum of Q Decay, Q Sensible, Q Auxiliary 1, Auxiliary 2, and Q Auxiliary 3.

(7)

Q Service Mater Sum of Q Auxiliary 2 and the heat rejected by the RHR heat exchanger into Division 1 service water system, i.e., the sum of the heat rejected through the spray nozzles.

}'$}tp)!:.'}i}rgbI A,. ~ t k

y i

~

w

~

}

p I

I

'I Jfpg

}0 r

- r

~ 1

~%~

C

~tl 4C rA 0 g}

k~>> >n 4... W~t r.

~

~.

~

C

A i

V c

PH0 I

~

J

~ ~

~

1

~

4 4

S.OISAAITIIII!C 4!E 7520 H~Z t 4 ~ CICLCS

~44II41 I 4 AcvrrcL ~ circa co 4

4 1

'5

~

5 St ~ t!

4

~

5 4ttt ~

I ~

4

~

4 I'

5 41

~ t!,.

I t

4t 4

IJ4 U

iat -I!!

~. r

/~+~ ~

I

~. I 4

Ai I

r ~ ~'\\

~

4 5 thtl

~ 4 ~ I LILL la h

Iac'I."!

I L '! Ct 4

/0 4

5 L!.I; L~'C~~-

~ A 4 ~4

'll

~ A:=I

~II II~>.<< -5:I'Ui-IA:.'::.I

-"c=i:,'!ii t.=-'ILI Zd P.Ptc,tJ&ir

='4=I=-.::::'-'tr d

';"g J:J I.

. R..!$~0d i,:.

Q~

ASS j;t7 vw I

~

4 4

I+Ii 'I 1 Mt I '

..5~art Ci~&P!.

~ 'CIU I it 5~-.-=!:X.=:I-;-'.-I'I..cl'-:I:-i:5

~ 'I

-"4 IQ

~ 4 ~

cr, gc U h INJ \\

4

..l.tc

\\ ~

~ Irc h

~

~ ll ~

I lh I'l1 I

~

4 44IIN La

!II" i-I nC rl

!I:=. -Xtrl J 4-I ISS.)

441'al IJA i

. ii'

/0 t

44I 4 'I al f1 i.gi I

++

MP

'X"'

CMM" P.

~ l.

R~'- =

~ I I e e I-'I-LJ I I

' 4

~ I ~ ~

4 Cl i ~crl! U.",I !ac. Irt.:

41 ":.I 7.rl Jai p I.tw,

~

4 rh 4

~ I 4 ~

lel O'I I

LI(

Watt I

~ I'I:.

4 ar 4

I~

4 '

~

~rr I r 1

.I

.aa 4

artier

~ 4 Ic.i I',

I Iaa 4 I

!5 act'.141J ~

I!4"'LIca I ~

I'I 5

.IJi.'-. E:. '

4

'll I 'I L~

II Il I.4 ha

~

Li

~'I I I t,t', h!

~r.l ha

~

Ik I

! ~

al ~

la Ja

~I'4'I'I

~ 4 r 4

I

'I

/V55s.'W LI=

tWZ q+I;

~,.

4 4ar ia ta

'K: cc!

I g'44 4 ~

~ ) Itla 41 i 5'ii It cli

. E:-

4 ~

I4!

"-4'41 5 4

~ I

~ '

Ic

~

I 4'rr i I I

~

I cili I ",!I.crl -!. I1

~p I '

I

~

I 4 ~ ~

~

~ ~ I

~a'

~- ~ '

El 4

Cl~I I 4'I 41L

'Ia

'I ~

il=-I4'I

~

~ IIJ JJ=Ait.-..t

't

. '4 rt aai EIICrr 14 Pl -J.lg: J::

IL'

~

~' I

~

~ 'I W-I r ie I

~

lal ~I 4'I

".:1c IV' h ~ ~

at aaa' I

14 ata 4" 4~...

~

Li.c:=* =11

~aa I

(5 I

aa ~

al

'L L

11

>:14 F~'F.

I I

~

I~

~ I 4

LIJ

~

P!SI c

I ~

4

~..J Il 5

I'

~

I

~

4 I

~

I ~

I

~

I ~ I I

.! I tice I

I I

~ I ~ i!.- ~ I., I..I I

I I I"'r!t I

~

I ~

'IJ I ~ 'IC 'III "I

~

I

~

~ I hLI.. I till

~ 1.

I,l sl-I'"

I ar

~ I

~

I!.l It

!i '. I!;c I'l'~'

', IC,!4

~ I

':I, p

4'c

~-lt r

' !I;I

~ ~

I I I::.:-:I I

~

~ ~

I

~I

/0

/0 Time RF78R Z.ocH-reer~

I0 o/z7/7 q

///Pi.

4 c 4 h

~ I LI CU LOCArrtTHLIIC

~ 4d j520

,VLF 5 5 s crcLcs

~ttt tt LLL

~

acvrccL ~ cssca co.

~

5

~

5 SOI C5CS

~ 51 5

5 c

5 55551 C

5 5 T.SSI

~

S 5 '5SSI I

~

t L

. I A N~WO I

~ '

I'ICtttt I ~ I 7I.IJ. I 7 Ja.4 l'c

ll:I t

LC) t ~ ~ I I 'll' ';I, L+15 7

U I

~

~ t t

\\t IJ ~lt CL.rI<<

f~f I ++if Ll

~ ht 5

5 5

~ ~

~

t ~ I.-

~ It

~ i.

1' I't U I tt t V

'IILC I

crt 71st At 7

t 171%i ht Itt t 4 It I 15'

~ '

H

~ ~re 'I ll oJt I;-.sr"

~

I t

~ It tt sist 'riffI'tt*it!if

~lo Nstltt 4/1, IIL I ~ 1

~ lr~

Litt tl I L Ijg 11LCS

~

~ I~

ffttttttt'c I

'I 1 t tll

~I

.ii crt 45 ILII tttt lot tt N V tot Ii, Itl t I~

I

~

ilttISI. IrI: ~.

Iieet

,=-=furr=

~

I I 'I cr "licgg

~ U

~I

~

I It ~ I rl, I trt ~ ~

I

~

7

'-". icS.Uf:itf

g~H j=}='t1U

~

t I J

~

LUt't I

I t

~

I Ult tiUrtl t

~ I

~ ~ I I

I t'I ~ I I't vrrc

~ tt, Ut

~ r. Itt a.

i.fr L,

'5]L QQ s

~tt

~

~ ~

~ \\

~ t i

~

I t

~ ~

pc

~

It I tl ~ 't

~

t t ~ Il'

't."I

~

5 U ~

tll~

Jllctl I 'i' Slt'. r,r!

I ~

~

Lf to VJ

~,

~

L 11

~'I

~

V I tf5 is,

~

t

~'II I~

icl

.. ~

, ~I ~

~

7

~ I

~ t CW

~ It/

-:is I if,.iol'..

'51

~ it I I ill

~

~~t

~

I l Ilt

}

liiil I'c I

~ ~

7-lnZ PF 7E'R LdCrV - M/P rcr g )g 37/4 7!7f CCf /P C

~ ~

r.

~g

~

tl

WNP-2 Regulatory Guide 1.27; Rev. 2, January 1976 Ultimate Heat Sink for Nuclear Power Plants Compliance or Alternate Approach Statement:

WNP-2. does not comply with the guidance set forth in Revision 2 of this regulatory guide.

WNP-2 complies with the intent of the guidance set forth in Revision 1 of this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment:

The basic design and much of the construction of tne spray ponds was completed prior to the issuance of Revision 2 of this re ulatorv on a

30 day period with the worst ew point deoression and average w'nds during that period.

C>lv>~~

Two Seismic Category I spray ponds are used, each with a capacity of 6.5 million gallons each.

The makeup for these ponds is supplied from the pumphouse at the Columbia River.

The makeup water piping is buried under a minimum of 5 feet of Quality Class I fill.

The makeup water supply system is utilized only in the event of a design basis

tornado, and therefore, it is not designed and constructed to withstand the effects of the OBE and water. flow ba'sed on severe historical events in the region.

v~~

I~gc~i $

~j Specific Assessment

Reference:

l<S~~T Refer to 9.2.5.

3o~

p

,~~)

w ~

s C.3-22

MNP '

010.24 RSP (g.2.5)

Me require that you protect the sprays in the ultimate heat sink from the effects of tornados and tornado missiles.

~Res onse:

As discussed in Section 3.3.2.3, the NNP-2 UHS design provides for continuous water make-up to the spray ponds in the event that both the spray systems are rendered inoperable due"to tornado missiles.

Therefore the sprays are not required to be protected from the effects of. tornado missile since an alternate UHS operating mode (continuous Make-up) is available which is protected from the effects of tornadoes and tornado missiles

k 0

0.

010. 25 9.2.5 In the event that a tornado siphons water from the ultimate heat sink (UHS),

the make-up water pumps will replenish the UHS.

Demonstrate that the transformers located in the turbine building and the electric cabling which are both required to operate the; make-up

pumps, are protected from tornados and tornado missiles.

~Res ense:

As described in section 3.3.2.3, the THU transformers (TR-75-72 and TR-85-82) are located at gound level in the southeast corner of the turbine building where they are protected by the exterior walls of the turbine building, the reactor building to the south, the service building to the

east, and other reinforced concrete interior walls to the north and to the west, and are therefore not considered vulnerable to tornado missile impact.

As described in 3.5.2, electrical cabling to the TMU pumphouse is buried at sufficient depth in compacted backfill to provide protection against tornado missiles.

Electrical cabling

,rom each transformer is routed separately to two switchgear units at ground level in the southwest corner of the turbine building.

For missile trajectories which would jeopardize the Tl)U transformers, associated

cabling, and switchgear, the exterior walls of the turbine building provide adequate orotection against design basis missile penetration and spalling.

"Appropriate draft FSAR changes are attached.

I

The availability of essential electric power to the makeup water pumphouse systems is assured.

The electrical lines are underground with sufficient earth cover to resist tornado-generated missiles.

The electrical lines are installed in such manner as to provide two redundant electrical systems from the power source to the makeup water pumphouse.

The two electrical systems are physically separated to provide adequate missile protection of one system from the other.

At the one end of each system,'edundant power source transformers ~~p=evMed The terminal ends and trans-formers at the makeup water pumphouse are enclosed within the tornado-resistant pumphouse.

Manholes within each sys-tem are also designed to withstand tornado generated missiles.

The spray pond piping and supports are designed to withstand the effects of the design basis tornado.

The piping system 'cannot be protected from the impact, of tornado generated missiles.

In the event of missile damage to one of the pond spray headers, the alternate spray system which is 100% redundant is placed in operation.

In the event that both spray systems are rendered inoperative, the cooling tower makeup water system is placed into operation to provide continuous makeup to the spray ponds with Columbia River water, the temperature of which never exceeds 70 F.

The cooling tower makeup water system is provided with sufficient protection to prevent its loss of function in the event of a design basis tornado passing over the project site.

Since the makeup water flow rate exceeds that of the standby ser-vice water systems, and since'he makeup water temperature is substantially lower than the standby service water system design temperature of SS F, the continuous availability of cooling water at a maximum temperature of 70oF is assured.

The method of detection of spray pond header failure and procedures for alternate spray pond usage is described in 9.0.

'I Failure of non-tornado resistant cooling towers due to tornado loads does not endanger Seismic Category I structures since the plant arrangement provides sufficient distance between the cooling towers and Seismic Category I structures.

pgSoCra~

ger'fchp'~~ ~g Ca,blithe iS Pr>~<~~

(

Q4'o u r o A P~ +~br~

kcci'kCia~

erC-

~y'+~~i w5 p8g 44/~

fm c cp Q~

whig yuipmevu+, f+ ~ harv~

gpss~ ilimp b<)

7~

hui i'<~

3 ~ 3 7

I HNP-2 32IENDMENT NO.

1 July 1978 c.

Radwaste and Control Building The exposed exterior concrete walls and roofs.

housing safety related

systems, equipment and components, are designed to withstand the effects of the design basis tornado generated missiles.

Figures 1.2-3, 1.2-4, 1.2-5, 1.2-9 and 1.2-11 illustrate the radwaste and control building and their relative location in the plant complex.

d.

Standby Service Water Pumphouses and Spray Ponds The exterior walls of both 'pumphouses are con-structed 'of reinforced concrete and are 2'-4'hick, minimum.

This thickness is adequate

-..c withstand design basis tornado generated missiles.

In addition, the two pumphouses are redundant to each other.

In the event that one pumphouse is inoperable, the other.is capable of providing sufficient service water for safe shutdown.

The ability of the spray, ponds to tolerate t:.e'esign basis tornado generated missiles is dis-cussed in 3.3.2.3.

Figure 1.2-14 illustrates the pumphouses anc spray ponds.

e.

Makeup Hater Pumphouse The exterior walls and roof of the makeup water pumphouse are of reinforced concrete and are sufficiently thick to withstand the effects

.~f the design basis tornado generated missiles as discussed in 3.3.2.3.

The exterior walis are 2'-4"'thick and the roof slab is 1'-4" thick.

Figures 1.2-1 and 1.2-13 furnish its location and arrangement.

g~; Id,~)

C~~p~ ~,

is ~

7 ~~" 'm

'7

/oc~wM i~

Lci ~~ <

~

t

~ Q ~.pi g 5 Pf'a.Vr <<~/ 6y P~ I+ +<<C.i./

l8 ir cM5 iw 7<v cfC~5~ /

s/~6s uv ~r W~u g-,

e/~~=;>-=~

P C 5 19

~

] I~S/~<< ~'~

C f.rh;~

b k~-'.).-

I 4

f. All openings for heating, ventilation and air conditioning system fresh air intakes (FAI)"

and exhausts (EXH), in buildings housing safety-related equipment, are protected against ex-ternally generated missiles by means of shield walls as indicated in Table 3.5-6.

Examples are the louvred openings above the floor elevation 572'-0" in the north and south walls of the reactor building.

These open-ings are protected by a labyrinth of missile shield walls immediately inside the opening.

3.5.3 BARR1ER DESIGN PROCEDURES The design objectives emphasize missile containment and structural integrity without secondary missile generation.

Concrete missile barriers are designed in accordance with the modified Petry equation. (Reference 3.5-2).

In all cases, ex-ce t for barriers exposed to turbine missiles, a concrete thickness of twice the penetration thickness determined for

'n infinitely thick slab is provided to prevent perforation, spalling scabbing.

For discussion of turbine generated missiles se 3.5.1.3.

5 7~ 3.$ -5

~

Cf

.The formulae u'sed to determine penetration depths into steel barriers are given in 3.5.1.1.2.

The overall response of barriers subject, to impact are inves-tigated by the use of general energy equations given in "Introduction to Structural Dynamics", J.

M. Biggs (Reference 3.5-9).

Upon determination of penet ation depth and duration of impact, an effective dynamic force is computed.

The addi-tional calculation of the natural period of the target struc-ture and the selection of a ductility ratio facilitates the determination of the required structural resistance.

In this manner, missile impact is,translated to an equivalent static load in an effort to quantify bending moments and shear.

The detailed method used for'redicting the overall response of missile barriers, including the forcing function method of determining ductility in structural elements and the. basis for the ductility ratios used in the calculations, is provided in Appendix C of the report "Protection Against Pipe Breaks Out-side Containment" (Reference 3.5-13) that was presented to and approved by the NRC.

3.5-24

I I

TABLE 3'-5 DEPTHS OF MISSILE 'PENETR'ATI'ONSXNTO'ONCRETE

'ISSXLE 35'TILITY POLE TARGET Quality Class I structures up to 30'bove grade

'ENETRATXON DEPTH tI)

(in.)

STEEL ROD Quality Class I 3" diameter x

3 ft.

structures at long.

any elevation 3.5-32

l I l 0,

HNP-2 9.2.5 In Section 9.2.5 of the FSAR, you state that the two ponds which comprise the ultimate heat sink are connected by a siphon that allows water to flow from one pond to the other.

Demonstrate that a failure in this siphon line, or in one of the ponds, will not result in draininq of both ponds.

~Res onse:

The siphon between the two ponds is a Seismic Category I, guality Group C,30 inch pipe, whose centerline is 4 ft.

6 in. below the normal water level of the spray ponds.

Therefore a siphon line failure would be considered a passive failure.

Applying single failure criteria indicates that if the siphon f'ailed then both SM loops would be operating, thus keeping them at the same level.

If one of the SM loops fails, then an additional failure of the passive siphon is not considered credible.

The spray ponds are Seismic Category I structures located below grade with continuous waterstops in all joints and bounded with guality'lass I

high density backfill.

Both ponds together form the Ultimate Heat Sink, a concept which has been accepted on other plants that only have a

single pond which contains the redundant spray networks.

Failure of either Pond A or Pond B will result in drainage of the other pond, which results in the same consequence if the MNP-2 UHS were a single pond design.

However, as described above and in section 3.8,4.1.5 the spray ponds have been conservatively designed to preclude pond failure.

I

g 010.27 RSP (g.2.7)

Me require that you protect the standby service water system from tornado missiles.

~Res onse:

The standby service water system (except for the spray pond spray piping) is protected from tornado missiles.

The structures which house the standby service water systems (Reactor Building, DG Building, Control Building, and SW Pumphouse) have been designed to withstand design basis tornado generated missiles as described in section 3.5.1.4.1.

Buried portions of the standby service water system are protected from tornado missiles as described in Section 3.5.2, See the response to;question 10.24 as to why it is not necessary to protect the spray pond spray headers from tornado missiles.

l

WNP-2 0

010 28

~9. 3.4 Describe how flooding of safety-related equipment due to backflooding through the equipment and floor drainage

system, is prevented.

Demon-strate that those portions of the drainage system necessary to prevent backflooding (e.g.,

check valves) are designed to Seismic Category I

criteria and that their system function will be maintained, assuming a single active failure.

~Res onse:

It is assumed that the question is directed to FSAR section 9.3.3.2.2,1, Reactor Building Floor Drains, and not 9.3.4, Chemical and Volume Control System.

As shown on Figure 9.3-8, the floor drain piping in the reactor building drains to one of four sumps listed below.

F~10 i

S FDR-R-1 Room Location RHR A Pump Room Rooms Served RCIC RHR A FOR-R-2 FOR-R-3 RHR B Pump Room HPCS Pump Room RHR B

HPCS CRD FDR-R-4 RHR C Pump Room LPCS RHR C

Each of the four downcomers is equipped with instrumentation which alarms in the control room to tell the operator at which elevation an excess of water is collecting in the downcomer.

Each sump is equipped with level intrumentation which:

1) controls the sump

pumps,

2) alarms in the control room (on high sump level),

and

3) initiates closure of the isolation valves in the downcomers and in the piping between interconnected rooms.

Not currently shown on Figure 9.3-8 are Class lE level instrumentation to be installed just above floor level in each ECCS pump room.

This instrumenta-tion will alarm in the control room.

The floor drain system is analyzed against the potential sources of flooding within the reactor building, i.e. pipe break outside containment and passive failures in the ECCS during post-LOCA long term cooling.

Using the acceptance criteria for either event (Standard Review Plan 3.6.1 and Reactor Systems Branch Technical Position, Leak Detection Requirements for ECCS Passive Failures),

the floor drain system design is acceptable in mitigating the consequences of flooding ECCS pump rooms.

e

The effects of pipe breaks outside containment are addressed in Section

3. 6. 1. 11.4, i. e. ruptures in fluid systems have no effect on the ability to bring the reactor to a cold shutdown condition.

Single random active failures are assumed in the analysis and credit is taken for systems not affected by the flooding.

As stated in Section 3.6.1, these assumptions and the approach taken are consistent with the guidance of Branch Technical Position APCSB 3-1.

This is in conformance with the criteria of Standard Review Plan 3.6.1, March 1975.

The effects of passive failures in the ECCS during post-LOCA long term cooling is addressed in the response to question 212.003.

The largest passive failure has been identified as the total failure of an RHR pump seal and it is equivalent to a 23 gpm leak.

Class lE instru-mentation in each ECCS pump room will detect'the leak and give the operator at least 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months to identify and isolate the passive failure before it has any additional adverse effects on ECCS operation.

s J

WNP-2 FSAR UESTION 010.29 Demonstrate that the heating, ventilating and air conditioning systems (HVAC) for the engineered safety features are protected from tornado missiles,

RESPONSE

Except for standby service water piping, control room remote air intake piping and control room remote air intakes, the HVAC systems serving engineered safety features are all located within reinforced concrete structures designed to withstand the effects of design basis tornado missiles.

These structures include the reactor building, radwaste and control building, diesel generator building and standby service water pumphouses.

Design of the buildings, including protection for HVAC system air intakes and exhausts, is discussed in 3,5.

The standby service water piping runs between the standby service water pumphouses and the reactor building and supplies water to the cooling coils of critical HVAC equipment during a design basis accident.

The control room 'remote air intake lines run between the remote air intakes and the radwaste and reactor buildings, respec-tively, to supply control room pressurization and makeup air during accidents involving radioactive releases.

Since these piping runs are all covered to a depth of over five feet wi th Class I compacted earth fill, they are adequately protected from tornado missiles as discussed in 3.5.2.

The two control room remote air intakes are over 200 feet from any major plant structure and are located in a northwest and southeast direction, respectively, from the turbine-generator, radwaste and reactor bui lidng complex.

The intakes are of reinforced concrete construction designed to withstand design basis tornado missiles, The roof slab of each intake is 12 feet square and 2 feet thick with a grated 3 foot square opening for the intake air.

Eight 4" pipes surround the intake opening and serve both as barriers and as alternate air intakes in the event the grated opening is blocked or damaged.

The top of the roof slab is 15 inches above the surrounding grade level.

The walls and floor of the intake structure are 18 inches thick and are buried to a depth of approximatey 9 feet in Class I compacted earth fill.

An internal barrier is provided as additional protection for the intake piping and to assure an unobstructed path for the flow of control room air.

This barrier, which is 4 feei wide and 18 inches thick, is supported from two sidewalls and two one by two foot columns.

Details of the intake structure are shown on Figure 3,5-52,

To be supplied in FSAR Amendment No.

4

WNP-2 010.30 9.4.0 You state in )he FSAR )hat the outdoor design temperature range for the HVAC is 0

F to 105 F.

However, you also indicate on Page 9.<-2 of the FSAR, that the extreme outside temperature range is minus 27 F to

'15 F.

Provide the results of your analysis which demonstrate that the functional capability of safety-related equipment will not be impaired by the outdoor temperatures which would occur during these extreme meteorological conditions.

The effect of extreme low temperatures on safety-related equipment located outdoors should also be discussed.

~Res onse:

This question has been addressed in 9.4 (Page 9.4-2).

Even though the normal ou)side temaerature range for the'design of the H!ItAC systems was 0

F to 105 F, as stated in the FSAR, operation at the extreme conditions of -27 F and 115 F were also evaluated.

Where necessary, equipment was selected to assure operation of safety related systems during the extreme conditions.

As discussed in 9.4.7, 9.4.8 and 9.4.10, the heating equipment for the diesel generator building, diesel generator cable corridor and standby se~vice water pumphouses are capable of maintaining temperatures at 35 F or above during the extreme

-27 F conditions.

In sizing heating equipment for primary operating areas of the plant, such as, the control room, reactor building or turbine-generator

building, no credit was taken for heating available from plant lighting or operating equipment.

Including these additional heating sources, even in a shut-down mode, the existing heating equipment is adequate to maintain the areas served above minimum set temperatures during the extreme cold condition.

The extent and duration of any room temperature increases which !!I!ay result during operation at the extreme summer temperature of 115 F

with the existing cooling systems, will not be sufficient to degrade the operating capability of any safety related equipment.

4

MNP-2 Q

010. 31 TRR44.. 44 The Radwaste Building chilled water system which is not designed to Seismic Category I criteria, is connected to the HVAC system at the control

room, and to the standby service water system.

Provide your analysis which demonstrates that the potential failure of the Radwaste Building chilled water system during an earthquake will not cause an unacceptable degradation of the control room HVAC system and the standby service water system.

~Res onse:

The Radwaste Building chilled water system is completely isolated from the standby service water system.

Both control room air handling units are Seismic Category I and are provided with two "N" stamped cooling coils.

One coil is connected to the Radwaste building chilled water system and the other coil is connected to the standby service water system.

Failure of the chilled water system will not adversely effect control room or the standby service water system.

Please see 9.4.1.2.1 for additional information.

I 1

HNP-2 010. 32 9.4. 7 Demonstrate that the ventilation system of the Diesel Generator fuel oil pump room is designed to Seismic Category I criteria, and receives power from the Class 1E buses.

~Res onse:

Please see 9.4.7.3, Figure 9.4-7 and Figure 8.3-1d for requested information.

WNP-2 010. 33 9.4. 7 Provide your analysis which demonstrates that the potential failure of the heaters in the Diesel Generator HYAC System which are not designed to Seismic Category I criteria, will not have an adverse affect on the functional capability of either the Diesel Generator or the Diesel Generator HYAC System.

~Res onse:

There are two types of heaters in the diesel generator

spaces, electric unit heaters in the diesel oil pump rooms and electric heating coils in the duct systems in the diesel engine rooms themselves, The electric unit heaters in the pump rooms are Seismic Category II.

These heaters are supported as Seismic Category I, however, and can fail in place without affecting any safety related equipment.

Oil pump room unit heaters are used only for maintenance during cold weather for personnel comfort.

The heating coils in the generator room themselves are Seismic Category I and are designated Class lE.

WNP-2 10.4.5 Your response to Item 010.09 is unacceptable.

Specifically, your analysis of flooding due to failure of the circulating water system is based on a crack whose area is equal to one-quarter of the pipe diameter times the pipe thickness (.5t X.5d).

Provide an analysis of flooding due to a postulated failure of the expansion joint in the circulating water system assuming a double-ended guillotine break at this location,

~Res onse:

The original response to Item 010.09 has been rewritten for clarity (see 10.4.5).*

The double-ended guillotine break referred to above was not considered.

The circulating water system is a moderate energy system by definition.

Therefore, in accordance with HRC Standard Review Plan Section 3.6.1, 3.6.2, and 10.4.5, and the associated Branch Technical Position HEB 3-1, the criteria for a postulated failure shall be a through - wall leakage crack of the type addressed in the written response (10',5).

In any

case, as stated at the end of 10.4.5, circulating water piping is located remote from any safety-related equipment.

The piping is located in a large room containing little other equipment and no safety-related equipment.

Accordingly, safety-related equipment is not vulnerable to environmental effects of a circulating water pipe rupture.

The pipe exits the room below grade in its routing to and from the cooling towers.

It should also be noted that the condenser is located on grade level.

Therefore, water above the floor elevation will drain outside and not collect other than in collection basins.

l0.4.5.3 Safety Evaluation The circuiat'ng water system is a non-safety related system.

Consecuently, the c'rculating water system is not designed to Seismic Category I requirements.'efer to 9.2.5 for a descript'on of the ultimate hea" sink which is designed to pe form safety-related func ions.

The condenser des'gn assures that.the pxessure on the tube side is always mainta'ned h'gher

.than the pressure

'on the shell side, thus eliminating leakage into the circulating water sys em should tube failure occur.

Consequentlv, the design of the circulating water system precludes radioactive

,leakage into the system.

Periodic injection of chlorine is performed for biocide treat-

ment, and sulfuric acid is added for scale-cox'rosion control within the circulating water svstem.

An analvsis of the trans-portation,

handling, storage, and utilization of chlorine 's presented in 6.4.

A detailed evaluation was performed to determine the effects of a postulated failure in the circulatincr water system in-side the turbine building.

ror this analysis a moce ate energy crack was postulated to occur in the circulating water system barrier, (e.g.,

the rubber evpansion joints) at the inlet to the main condensex.

The inlet side was selected because it yields the severest. results.

The entire condenser area is drained bv means of sumps (see Figure 9.3-9),

each equipped with duplex pumps.

Sumps T-2 and T-3, servicing the inlet and outlet of the condenser, each have 50 gpm pumps.

Hach of these sumps is ecuipped with a level alarm and 's therefore capab'e of detect'ng a circulating water system barrier failure.

The level alarm will annunciate in the main control room upon reaching

~ high level, providing a means of detecting the postulated ailure within 5 minutes.

The crack area for this postulated failure was assumed to be equal to 1/2 the pipe diameter t'mes 1/2 the pipe wall thickness.

(see

3. 6. 2:i. 4. 2. b) 10 '

17

g4e~ e The flow exit'ng from such a crack would b an orif'ce flow.

The head at expansion joint or normal " "

pump opera-tion at 186,000 gpm e ch was determined

( rom system energv gradients) to be 90 feet.

The flow for these condit'ons was calcula ed to be:

Q = 1,737 gpm The svs em has diffe ent operating pressures or the various modes of pump operation.

The piping was designed fo" an internal pressure of 60 ps'g, which is wel3. above the design energy gradient.

The motor operated 'nlet and outlet valves at the condenser are designed and, manufac ured "o close in 60 seconds to avoid excessive pressures caused bv fast valve c3.osure.

Therefore, rapid valve closure is not a considerat'on.

A e

closure of the inlet and outlet valves,

however, the system will be operatinc with 2/3 of the condenser capacity.

With 3 cir-culating water pump opezation and 2 sections o

tne con-denser 'n operat'on, the system flow as determined from the pump operating point diagram will be approximately 450,000 gpm.

Comparing the system energy gradients for this mode of operation to that when all three condenser un's are in operation, the resultant diz erence 'n pressures will be:

r At the inlet side, an inc ease oz approximately 4.3 ft. of head (2 psi) occurs At the outlet s'de, a decrease of approximately 5.2 ft of head (2 psi) occurs Detection of the postulated failure rrill occur within 5

minutes, as desc ibed above, by the annunciat on in the con rol room'f the sump high level alarm.

3:t is assumed that there will be a 15 minute time allowance for an opera-tor in the control room to check the circulating wate system barr'rs and close both the inlet and outlet valves of one unit of the condenser as may be recuired.

This closure is accomplished by the activation of a remote manual switch in the control room, and therezore no cont ol circuit'y time delays nor coastdown times are invo3.ved.

r"low wi3.1 continue, however, after valve closure for about 3.06 minutes at a decreasing rate, until the remaining water from tne condense is completely discharged.

10.4-17a

'I 1

L"Ilp-2 In the f'rst 5 minutes after a crack, 8,435 gallons of water w'll spill into the inlet basin.

The capacity of each basin and its capability to store excess flow we e calculated to be as follows:

a.

Inlet bas'n:

22,500 gallons from E' 436 to El.

441 b.

Outlet basin: 27,500'allons rom El.

436, to El. 441 c.

Net volume under condenser:

180,500 gallons from El.

433 to El. 441.

The time recuired to fillthe inlet basin, after a postu-lated crack occurs, is computed to be 13.3 minutes.

This includes the 50 gpm outflow from the sump pump.

The circu-lating water leakage flow w'll continue for 6.7 minutes after filling the inlet basin, until reach'g the total estimated shutoff time of 20 minutes.

It can be assumed that 10% of this water will flow out over the floor at El. 44',

and the remainder, about 10, 170 gallons, will flow into the condenser basin area.

Dur'ng this same time period, 4

sump pumps in the condense basin area will have alternately pumped out 670 gallons, leaving 9500 gallons or 0.42 feet of water in the condenser basin.

The rate of rise of water, therefore, is '0.021 ft/min during the firs 20 minutes after the postulated crack occurs.

Note that on high sump

'level, both pumps run simultaneously rather tnan alternately, thus doubl'ng the calculated outflow capacity.

After the valves are closed, the water contained in the condenser unit water box will continue to discnarge

"'o the area.

The ouantity of water remaining is e

ated to be 87,000 gallons.

The flow will vary w'th a de'nishing

head, the head going from about 25 feet to zero

~ee~.

Using a

20 ft head and the same orif'ce flow criteria, the rate of flow wil1 be approximately 8 19 gpm, discharging the re-maining water in about 106 minutes.

There will be an out-flow from all the sump pumps of 150 gpm, with 10% of the flow f om the crack again assumed to flow out over the floor.

The water will accumulate in the condenser bas'n at. about 590 gpm.

After 106 minutes, the water level in this basin w'll rise an additional 2.77 feet, or 0.0261 ft/m'n.

The total heigh" of water when the discharge has stopped is therefore 3.19 feet to El. 436.19.

10.4-17b

HHP-2 There are no safety-ralated system components that could be a fected by the flood elevati'on established above.

Addi"ionally, here are no safety-related electr'cal systems or svstem components that could be potent'ally submeraed.

1 s

etv-re ed elec

'cal sv gems route

-nrougi'he tu~ ine gen tor bu'ng en" r the bui'ng at aled 11 pene ations 'ide c.duits at 490 an termin ie in inst-went ra

.cs on >>~.

501 as w

1 as te

" nal bo es mount above

. 501.

he flao vaters ca ot rea th e

elev tions i the eve of a fa're as p -'late Discharge operat'on of water accumulated uncer the condenser shall be performed in accordance w'th radioactivitv checking recruirements for sump discharges.

10.4.5.4 Tests and Inspections'll system components, except the condenser, are accessible during operation and may be inspected visually.

The circu-lating water pumps are tested in accordance with the i-:ydraulic Institute Standards.

10.4-17c

WNP-2 Responses to previous questions:

Hydrology-Meteorology (371.6)

Geosciences Branch (360.4, 360.5)

c'

>NP-2

.. 371.6 Provide the results of a transient analysis to determine the adequacy of the ultimate heat sink spray ponds under emergency conditions, including consideration of the requirements for both the temperature and volume of the water.

(Refer to Regulatory Guide 1.27, Rev.

2, for guidance on this matter.)

Provide the basis for any assumptions used in your analysis and a discussion of your analytical techniques.

~Res onse:

The mass loss and thermal transient analyses for the UHS following a design basis LOCA are presented in revised section 9.2.5.

See the response to g 010.23 for additional information.*

draft changes to 9.2.5 are included with g 10.23.

0 0

MNP-2 g..

360.4 In the Meston Geophysical

Research, Inc. report,",gualitative Aeromagnetic Evaluation of Structures in the Columbia Plateau and adjacent Cascade mountain Area," March 28, 1978, Figure 13 shows several north to northwest trending aeromagnetic linears in the vicinity of Badger Mountain and Jump Off Joe Anticline.

However, the Meston report does not discuss the origin or interpretations of these particular linears.

The north trending linear crossing the Columbia River at the junction with the Snake River has an apparent offset of the magnetic low defining the Rattlesnake Hills anomaly.

Since these aeromagnetic linears trend toward the MNP-2 site, provide:

(1) an interpretation of these features, including but not limited to the potential'for their continuation to the north to near site area:

and (2) a discussion of the fault parameters, if such an interpretation is proposed.

~Res onse:

The concerns raised in this question relate. to recent information which post-dates the information now before the staff.

The reference letter proposes a meeting to update the staff with respect to this information.

As stated in the letter, a generic report is scheduled for early fall 1979,which will place this information in perspective and respond to the concerns -of this question.

Reference:

Letter, D.L. Renberger (MPPSS) to O.D. Parr (NRC), "Update of Geological Studies",

dated April 27, 1979.

i~

4

WNP-2 g.

360.5 Some of the data and discussions in the FSAR of those Columbia Plateau structures relevant to the WNP-2 site're slightly different from the information provided in Amendment 23 to the WNP 1

5 4 PSAR (Docket Nos.

50-460 and 50-513).

For example, with regard to the Wallula Gap Fault, your FSAR states that the "...probable fault movement occurred after the deposition of the Touchet beds, and thus less than 12,000 years ago."

However, in Appehdix 2RH.4 of the WNP 1

8 4 PSAR (Amendment 23), you indicate that the fault is older than the guaterna'ry Kennewick fanglomerate based on trenching.

Additionally, in this same amendment to the WNP 1

8 4 PSAR, you indicate that the faulting along the Horse Heaven Hill Anticline occurred about 3.5 million years before the present (mybp).

The WNP-2 FSAR does not discuss this particular point

'ut, rather, questions the existences of faulting along the Horse Heaven Hill Anticline and indicates that it could be the sole result of folding.

Clarify these apparent discrepancies and provide cross-references in the WNP-2 FSAR to the appropriate sections of the WNP 1

8 4 PSAR.

~Res onse:

The concerns raised in this question relate. ta. recent information

=- which post-dates the information now before the staff.

The reference letter proposes a meeting to update the staff with respect to this

. information..'s stated in the letter, a generic report is scheduled for early fall 1979 which will place this information in'erspective and respond to the concerns of this question.

Reference:

Letter, D.L. Renberger (WPPSS) to O.D. Parr (NRC), "Update of Geological Studies",

dated April 27, 1979.

4' E, C

ML17272A453

Contents

Text

SUBJECT:

Subject:

Reference:

Dear Mr. Varga:

Attachment:

Reference:

System Description

Reference:

RESPONSE

Reference:

Reference:

Navigation menu

ML17272A453

Text

SUBJECT:

Subject:

Reference:

Dear Mr. Varga:

Attachment:

Reference:

System Description

Reference:

RESPONSE

Reference:

Reference:

Navigation menu

Search