ML20070D320

Forwards Responses to 820820 Acceptance Review Questions Re OL Application,Scheduled for NRC Review in Nov.Marked-up FSAR Pages Included Where Changes Became Necessary Due to Responses

ML20070D320
Person / Time
Site:	Satsop
Issue date:	11/30/1982
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Knighton G Office of Nuclear Reactor Regulation
References
GO3-82-1228, NUDOCS 8212150029
Download: ML20070D320 (22)

Text

4 Washington Public Power Supply System P.O. Box 968 3000 GeorgeWashnigton Way Richtand, Washington 99352 (509)372-5000 Docket No. 50-508 G03-82-1228 i'ovember 30, 1982 Mr. G. W. Knighton, Chief Licensing Branch No. 3 US Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

NUCLEAR PROJECT 3 RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS

References:

a) Letter 0. G. Eisenhut to R. L. Ferguson, dated 8/20/82 b) Letter #G03-82-830 G. D. Bouchey to..

R. Denton, dated 8/20/82 c) Letter #G03-82-1085 G. D. Bouchey to J. D. Kerrigan, dated 11/22/82 Reference a) transmitted a set of questions generated during the NRC's acceptance review of the WNP-3 Operating License Application (reference b).

Reference c) represents the initial Supply System response to these questions and provided a schedule for those cases where our evaluations were not yet complete.

This letter transmits those responses scheduled to be provided for NRC review in Novc;.ber.

In those cases where it is considered necessary or desirable to amend the FSAR due to our responses, we have provided marked up FSAR pages which show the changes which will be ir? I

'M in a subsequent amendment.

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$DDI N666 csF2 8212150029 821130 PDR ADOCK 05000508 A

PDR

Mr. G. W. Knighton Page 2 November 30, 1982 RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS If you require additional information or clarification, the Supply System point of contact for this matter is Mr. K. W. Cook, Licensing Project Manager (206/482-4428 ext. 5436).

)3D3 eu '

G. D. Bouchey, anager Nuclear Safety and Regulatory Programs AJM/ss Attachnents:

1.

NRC Question No. 241 (2.4.11.1) 2.

NRC Question No. 480.5(6.2.4) 3.

Request For Additonal Information - (Item 10) cc:

D. J. Chin - Ebasco NY0 N. S. Reyonds - D&L E. F. Beckett - NPI J. A. Adans - NESCO D. Smithpeter - BPA Ebasco - Elma WNP-3 Files - Richland A. A. Tuzes - Cemb. Engr.

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NRC QUESTIONS NO. 240.1 (2.4.11.1)

For non-safety-related water supplies, demonstrate that the supply will be adequate for a 100-year drought, or show that it is sufficient to not cause unacceptably frequent use of the emergency systems.

Include a discussion of low flow in the Chehalis River.

RESPONSE

As shown in Figures 2.4-28 (Chehalis River (at site) - 7 Consecutive Days Low Flow Frequently Analysis), and 2.4-33 (Chehalis River (at site) Day Low Flow Frequency Analysis), the 7-consecutive days flow of 550 cfs has a return period of 6 years, while the corresponding 1-day flow has a return period of 4 years. Therefore, the Chehalis River water supply will not be adequate for the 100-year drought since the State of Washington imposes a withdrawal limi-tation if the daily average river flow falls below 550 cfs. However, the above return periods show that the frequency of low flow periods resulting in withdrawal limitations is not unacceptable. The discussion of low flow is shown in Sections 2.4.11.1, 2.4.11.3 and Appendix 2.4A.

The record of estimated daily flows for the period 1930-1981 contains 25 periods of stream flows less than 550 cfs, which occurred in 13 of the 52 years. The durations of the periods range from 1 to 21 consecutive days.

In the 13 years, six years contain more than one period, which typically occur so close together that they should be considered as one combined event with re-spect to plant operation.

In the 52-year period of record, there is a total of 212 days with average daily flow less than 550 cfs (an average of 4 days per year). Thirty-nine (39) years of this period have no occurrences of flow less than 550 cfs.

Table 1 lists the dates of occurrences of daily flows less than 550 cfs. The two low flow events of the longest duration occurred in 1951 (33 days less than 550 cfs over a 48-day combined event) and 1967 (38 days less than 550 cfs over a 47-day combined event). Figure 1 shows the -esults in terms of percent of the years of record in which a given duration of an event or com-bined event is equalled or exceeded.

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TABLE 1 OCCURRENCES OF DAILY FLOWS LESS THAN 550 CFS, CHEHALIS RIVER AT SITE 1930 - 1981 Combined Event Dates Consecutive Days Duration, Days Aug. 28, 1935 1

-1 Sept. 12-30, 1938 19 19 Aug. 21-26,1944 6

Aug. 28-Sept. 1, 1944 5

Sept. 6-13, 1944 8

24 Sept. 9-13,1949 5

5 Aug. 8-10, 1951 3

Aug. 16-28, 1951 13 Sept. 3-7, 1951 5

Sept. 13-24, 1951 12 48 Oct. 3-20,1952 18 18 Sept. 16-23, 1953 8

Sept. 25-27, 1953 3

12 Sept. 6-8, 1956 3

3 July 30-31,1958 2

(not included)

Aug. 12-28, 1958 17 Sept. 6-14, 1958 9

33 Sept. 10-11, 1965 2

2 Aug. 19-26, 1%6 8

Sept. 4-10, 1966 7

23 Aug.13-Sept. 2,1%7 21 Sept. 4-9, 1967 6

Sept. 17-21, 1%7 5

Sept. 23-28, 1967 6

47 Aug. 14-Sept. 2, 1970 20 20

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FORMSetIDEW?79

NRC QUESTION NO. 480.5 (6.2.4 )

Provide an evaluation of your conformance to Branch Technical Position CSB 6-4.

Identify and justify any deviations.

RESPONSE

The design of the WNP-3 containment Vent System (CVS), conforms with the pro-visions of BTP CSB 6-4 (see FSAR Subsection 9.4.6.6).

The following is the status of the analy;es required by Branch Technical Position 5 of CSB 6-4 to justify the Containment Vent System design:

Sa. The analysis of the radiological consequences of a loss-of-coolant acci-dent is completed (see Attachment A). FSAR Subsections 6.2.4 and 9.4.6.6 will be amended to reflect Attachnent A.

Sb. An analysis which demonstrates the acceptability of the provisions made to protect structures and safety-related equipment; e.g., fans, filters and ductwork, located beyond the vent system isolation valves againct loss of function from the environment created by the escaping air and steam is not deemed necessary, because the isolation valve is designed to close quickly enough to preclude any significant release to the vent system.

3c. An analysis of the reduction' in the containment pressure resulting from a loss-of-coolant accident will be complete by June 1983. FSAR Subsections 6.2.1.5 and 6.3 will be amended as shown in Attachment B to reflect this.

5d. An analysis is being performed on the operability of the CVS isolation valves for the spectrum of design basis pressures and flows against which the valves must close and the associated allowable leak rates. The re-sults of the analysis will be submitted via a report by August 1983.

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ATTACHMENT A I

1288W-3 WNP-3 b*

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{y; Valve closure times (less than or equal to 10 seconds) assure isolation as f ast as possible, while at the same time, not imposing excessively severe

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requirements on the design of the valve and valve eperators. An analysis will be perf ormed to verif y that the five second closure time for the eight inch containment vent system isolation valves, is fast enough to satisf y radiological concerns and ECCS back pressure concerns when postulating a LOCA during containment venting.

Isolation valves inside containment are located between the secondary sh'ield9 wall and the containment vessel. The secondary shield wall serves as a biological shield and a missile barrier.

Ioss of power to each motor-operated containment isolation val.e is detect,ed and annunciated in the Control Room. All other power-operated contai...nent isolation valves are designed to f ail in the position of greater saf ety.

m. 8 7 r..

wm-The isolation valves on both the on-line, eight inch' and the' ref'ullidg," f orty eight inch containment vent and purge system are designed to' f ail-closed. The on-line vent system is isolated by the containment high radiation signal. The Containment Purge System is isolated by the containment high radiation" signal' and the refueling pool high radiation signal... Au override capabliity exists for the purge function (see Table 6.2.4-1).

These isolation' signals ' aire in :,.-

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additi n to the CIAS.] eidli/),".h( cloyMGS/Mp}?g2/M{jQSdhi

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'bbsDh'i$hbetuate the isolation valves are discussed in,Section 7.3.

g,g Included is the diversity and setpoints of the parameters sensed. ~ The design

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maxJ of the Containment Isolation System is such that either a low pressurizer pressure or a high containment pressure initiates the CIAS, MSIS bi the SIAS.

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q. :; W;M' y,Q y as the outside containment isolation val've. [Except ll A check valve is not used ; } e as described in Subsection 6.2.4'.1.2 f or isolation of instfrumenti [

1 penetra tions. When used as the inside containment isolation valve, the check l

valve is designed and installed to have its disc seated when the differential pressure across the seat is zero.

Containment vent lines are provided with screens at the containment opening to prevent postulated debris f rom interf ering with the closure of the isolation val ve s. Similarly, the recirculation system sump is provided with screens o ensure debris does not interf ere with the operation of the Safety Injection System (SIS) or Containment Spray System (CSS).

When a component (e.g., inside containment isolation valve) is part of a contsinment isolation barrier and also part of the reactor coolant pressure boundary, it is built to Saf ety Class 1 criteria since the higher criteria go ve rns.

The set pressure of all relief valves which se. ve as isolation valves is much grea ter than 1.5 times containment design pressu e.

The set pressures for the relief valves on penetrations 1, 2, 3, 4 and 29 are all in excess of 1200 psig. The setpoint f or the relief valves on penetrations 27 and 28 is 435 psig.

"a y

6.2-129 Amendment No. 1, (10/32)

1247W-13 unP-3 a4go.<r FSAR.

...g The air exhausted by the Containment Vent and Purge Systems is monitored by j

t:*M)?

plant Vent Stack No. I radiation sonitors. Four ambient radiation monitors are located around the refueling pool in the containment. These monitors are

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seismic Category I and Class lE qualified. Upon detection of high radiation l1, level above the preset value of the monitors, the monitors generate a' signal.

I to close the Containment Vent and Purge Systems isolation wives and provide an alarm in the Control Room.

Section 11.5 and Subsection 12.3.4 provide further details of radiation monitoring for the containment system..

9.4.6.6.4 Tests and Inspections er The preoperational and periodic testing of all system components assure the reliability of the system. Verification of air flow and start-up of each f an are perf ormed during initial balancing and testing to ensure that.the svstem i design parameters are met.

Testing of air-cleaning units, which provide' the exhaust f unction f or the Containment Vent and Purge Systems is' performei..in..

4 accordance with the requirements of ANSI-N510. _ Containment Penetrat' ions 'and '

~ l1" Containment isolation Valves 2PV-B018SB, 2PV-B017SB, 2PV0B016SA,2PV-5112SA, 2PV-B111SB, 2PV-Bil3SB, 2PV-B064SA, 2PV-B164SB, 2PV-B019SA and 2PV-5021SA are tested periodically within the plant technical specification'(ised Chapter.16)'

7 to assure that the leakage of Containment Isolation Sy' stem is held within the isolation wives will be conducted on a. periodic basis,to assure tha,t. tIEih,j, [#@j Verification of. operation of the_c[on' tai,ninenty:b' 77 g allowable leakage limit.

'n valves will close within the acceptable closure time upon simula11on of a'L'.a'X,

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Containment Isolation or High Radiation Signal.

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9.4.6.6.5 Inst rumenta tion n= sal 1

The Containment Vent and Purge Systems are provided with monitoring and alarm fs.,.,

_ _'" provisions to enable the operatorito detect any "malfunc'tidn[i(MW " "DM.Tf:$F

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thEs'ystin The following variablis~ are monitored in the Control Room: M

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Air flow alarm to annunciate an alarm due to air flow failure of the

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Vent and Purge System f ans.

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Dif f erential pressure alarm to annun'ciate'when tihe pre'dsu're dif ferenes

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across the makeup air filter of the Purge System exceeds' a preset wlue.

c)

Low temperature alarm f or the makeup air of the Purge System.

d)

High radiatic:. alarm f or the containment atmosphere and f or the purge air through the plant Vent Stack No.1.

Sis [br~cl %N N b;b" U 0A S Eak:o!oy d b ase kna MTh&Sg 4-4,Po,; k sk Sec. A sch 's

,a Mk 9.4-90 Amendment No.

1, (10/82) 1._

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t Insert 1 9.4.6.6.6 RADIOLOGICAL DOSE ANAYSIS (Branch Technical Position (BTP) CSB 6-4, Position 5.a)

Introduction SRP 6.2.4 and BTP CSB-6-4 (Fosition 5.a), require an analysis of the radio-logical consequences of a loss-of-coolant accident.

The analysis is done for i

a spectrum of break sizes. The fission products in the containment atmosphere j

are assumed to be released through the.open Containment Vent System isolation valves during the maximum interval required for valve closure. The purpose of the subject analysis is to provide a detailed evaluation of offsite exposures resulting from such releases, i

Calculation of Releases and Offsite Doses The most serious design basis loss-of-coolant accident is a double ended guil-i i

lotine pipe break in the primary coolant system. Upon breakage, the rapid depressurization results in iodine spiking in the primary coolant and flashing i

of a mixture of steam and water into the containment atmosphere. Figure 6.2-1 presents the containment pressure transient analysis following LOCA. This figure presents an upper estimate of the containment pressure transient assum-i ing the Centainment Vent System isolation valves are closed at the time of the accident.

l If the eight inch CVS lines (one for exhaust and one for makeup) are open at the time of a LOCA, a mixture of air and flashed primary coolant would be vented directly to the environment until the lines are isolated.

In order to i

calculate the total amount of radionuclides released, an estimate of the total l

release of steam during this time period was made. The actuation of isolation valves begins at 19.7 psia, and using Figure 6.2-1, this occurs at approxi-mately 1.0 second post LOCA. The valve closure time is assumed to be 5 sec-onds. Since instrument delay is negligible, a total isolation time of 6 sec-l onds has been assumed in the analysis and used to determine the quantity of i

steam released.

l The concentration of radionuclides in the steam released prior to isolation

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has been assumed to be 60ICi/g dose equivalent curie of I-131.

Based on the above steam releases and radionuclide concentrations in the steam, the offsite doses were calculated. The atmospheric dispersion factor of 6.3 x 10-4 sec/m3 has been used at the Exclusion Area Boundary. A breathing rate of 3.47 x 10-4 m3/sec has been assumed. The assumptions and parameters used in the analysis have been summarized in Table 9.4.6-2.

The incremental Containment Vent System contribution as well as total (CVS plus LOCA) thyroid doses post LOCA have been presented in Table 9.4.6-3.

The exposures are within 10CFR100 guideline values.

The exposures presented in Table 9.4.6-3 are limiting for the cases where pipe l

break size would be smaller than that for DBA LOCA.

In such cases, the pres-I sure inside the containment would not build up as rapidly. The actuation of valve closure would occur based on a radiation monitor located at the exhaust, and the releases to the environment would be smaller than that fcr a DBA.

LOCA. The resulting offsite doses would be less than those presented in Table 9.4.6-3.

Therefore, a detailed analysis of such cases is not necessary.

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TABLE 9.4.6-2 ASSUMPTIONS AND PARAMETERS USED IN THE RADIOLGICAL DOSE ANALYSIS 1.

The containment is instantaneously filled with steam following LOCA.

2.

The pressure in the containment as a function of time following LOCA is observed as given in Figure 6.2-1.

3.

The steam is released from both makeup and exhaust lines.

4.

The steam escaping from the containment expands adiabatically.

5.

Inlet to the 8" line is abrupt. The resulting velocity head loss is 0.5.

6.

The following mathematical model given in the Chemical Engineers' Hand-book by Perry and Chilton (Fif th Edition) is used to calculate the max-imum mass velocity hypothetically attainable on isothermal expansion.

an appropriate factor from Figure 5-29 (b) of the referenced handbook is applied to obtain the actual mass velocity.

Gci=Po9/

gc x M/2.71E x Rx To '

Where:

Po = Absolute pressure in the large chamber, lb force /ft2 gc = Dimensional constant, 32.17 (lb) (ft)/(1b force) (sec2)

M = Molecular weight, lb/lb-mole R = Gas constant, 13 % (ft) (lb. force)/(lb-mole) (OR)

To = Absolute temperature, OR = 460 + OF Gci = The maximum mass velocity hypothetically attainable on isothermal 2

expansion, lb/(sec) (ft )

7.

Concentration of I-131 in the 60 coolant ( C1/g) 8.

Partitien coefficient 1

9.

Mixing in the containment 0

10.

Plate out factor 0

11.

X/Q (sec/m3) EAB (0-2 hr) 6.3 x 10-4 i

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TABLE 9.4.6-3 RELEASES TO ENVIRONEMENT AND OFFSITE DOSES FOLLOWING DBA LOCA' 1.

Releases to Environment a)

Steam (Kg) 36 b)

Dose Equivalent Curies of I-131 2.2 l

2.

Offsite Thyroid Doses at EAB (Rem) l a)

Incremental Purge Contribution 0.7 b) 2 hr LOCA lose with isolation 250.0 TOTAL 250.7

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ATTACHMENT B l

& ngoS 12 0'M-7 WNP23 FSAR 6.2.1.3 Mass and Energy Release Analyses f or Postulated Loss-of-Coolant 8.'....

. Accident s

p;;.;;;;;;

See CESSAR-F Subsection 6.2.1.3.

6.2.1.4 Mass and Energy Release Analysis f or Postulated Secondary System Pipe Ruptures Inside Containment See CESSAR-F Subsection 6.2.1.4 6.2.1.5 Minimum Containment Pressure Analysis for Perf omance Capability Studies on Emergency Core Cooling System See CESSAR-F Subsection 6.2.1.5.

Applicability of CESSAR-F enveloping 5 % '" N p

enalysis is curreatly under confimatory review.

L 6.2.1.6 Te sting and Inspection Preoperational and periodic tests are conducted to insure the functional capability of the containment and associated structures, systems and components.

Included are the Integrated Iaak Rate Test (see Subsection 6.2.6), Shield Building Imak Rate Tbst and Operational Tests on mechanical equipment that are required to operate following a pipe break.

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A' preoperat[on'ak Nisual in'spe tibN'a'n'd test shall he ' conducted to verify the

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structural integrity of the Shield Building. The test is conducted by

' tent 11ating the Shield Building at a flow rate of 10,000 cfm or less and A4 insuring that the differential press' re between the Shield Building and the Q~;7 u

autside _ atmosphere is a negttive 1/4 inch vg. or greater. The results of the

• iisual inspection, any repairs made thereof, and of the test shall be recorded in accordance with the test procedure.

Visual inspeN. O j?$hH { ~ ions [p'rddpirational'and periodic tests are performed at t

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frequenc'y described in and to the requirements and acceptability of the Technical Specifications (Chapter 16).

6.2.1.7 Instrumentation Application Pressure ' sensing instruments monitor the containment atmosphere and initiate CIAS, SI AS and CSAS according to the logic discussed in Section 7.3.

Radiation monitors which monitor containment atmosphere and isolate select i

containment penetrations are discussed in Subsection 7.3.1.

Instrumentation applications for the various engineered safety features associated with the containment, such as the Containment Ibat Removal System and the Combustible Gas Control System, are discussed in Subsection 7.3.1.

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6.2-20 Amendment No. 1, (10/82) l

INSERT 1 (480.5)

A minimum containment pressure analysis has been performed for WNP-3 and the resultant pressure response does not envelope the CESSAR-F pressure response. A WNP-3 specific pressure response, including a discussion of methodology and input parameters, is presented as discussed in Section 6.3.

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0;; C"e m -F S:: tic: 5.3[The Emergency Core Cooling System (ECCS) or Saf ety

($5uf. A. Injection System conforms to the CE interf ace requirements as described in CESSAR-F Subsections 6.3.1.3 ites A through 6.3.1.3 item Q.1, and Section 1.9 1

of this FSAR.

Figures 6.3-la through 6.3-lh show the Characteristic Curves for.the High Pressure and the Low Pressure Saf ety Injection Pumps (HPSI and LPSI).

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INSERT 2 (480.5)

CESSAR-F Section 6.3.3.2.4 allows reference to the results of the ECCS analysis performed therein if a higher minimum containment pres-sure than CESSAR-F Figure 6.2.1.5-3 can be demonstrated. As discussed in Section 6.2.1.5, the WNP-3 minimum pressure calculation does not meet this requirement. A WNP-3 specific ECCS analysis will be prepared to demonstrate com-pliance with 10CFR50.46.

Except for the results of the ECCS analysis the existing CESSAR-F descriptions of plant systems, ECCS calculations, methodologies, and analytical assumptions are applicable to WNP-3 and are referenced herein. The WNP-3 analysis will be complete by June 1983, the applicability of CESSAR-F will then be reevaluated and any dif-ferences will be identified on the WNP-3 docket. Past experience has shown that plant specific demonstrations of compliance with i

10CFR50.46 acceptance criteria do not result in changes to ECCS system designs.

It is there-fore expected that the WNP-3 reanalysis will confirm the adequacy of the existing design.

I

REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4

10) EFFECTS OF CONTAINMENT C0ATINGS AND SUMP DEBRIS ON ECCS AND CONTAINMENT SPRAY OPERATION A copy of the staff concerns on this issue, including a request for additional information which has been sent to a number of OL applicants, is provided as 0 (Attachment I).

RESPONSE

Items 1 through 3 will be discussed in the technical specification portion of the FSAR.

1.

A procedure will be written for, or will incorporate, an inspection of containment and the containment sumps for any materials which have the potential for becoming debris capable of blocking the containment sump screens. The procedure will be implemented just prior to each contain-ment isolation and startup.

2.

An inspection program will be implemented through the Corporate ISI Program Plan covering containment sump components, including screens, and intake structures per Regulatory Guide 1.82, Item 14.

3.

An Emergency Operating Procedure will be written to cover possible sump vortexing and blockage due to debris.

4.

In recirculation mode following a LOCA there are no operating fluid systems in the contaiment to rupture except the safety injection and containment spray systems which themselves are en essential part of the safe shutdown procedure and recirculation systems.

In no case would any high or moderate energy piping failure compromise the functional capability of the SIS sumps when they are required.

The SIS sump design, described in detail in FSAR subsection 6.2.2.2, l

meets all the requirements of Regulatory Guide 1.82.

The refueling l

pool cavity drain (El 384.00 ft.) is located in the opposite side of l

the building from the SIS sumps, and the floors above the containment water surface (El 370.00 ft.) in the area of the SIS sump are mostly steel structures and gratings, No drain flow and channel of spray flow l

is expected to be released below or impinging on the water surface (El 370.00 ft) in the area of the SIS sump which can cause significant i

l vortical flow in the neighborhood of the sump. The relative locations of the refueling cavity drain and the SIS sumps are shown in FSAR l

Figure 1.2-3.

l l

Also, as stated in FSAR Subsection 6.2.2.2, the minimum water level in the containment following a Design Basis Accident will be approximately I

at El 370 feet. This corresponds to the total design capacity (654,000 gallons) of the RWSTs and safety injection tanks capacity (57,600 gal-lons). The flood level can go as high as El 372 feet if the maximum

RWST usable capacity (826,000 gallons) and Safety Injection tank cap-acity (57,600 gallons) is available and is injected into the contain-ment. The centerline of the recirculation suction piping is at El 353 feet - six inches. As a result, a minimum submergence of 16 feet - six inches is available to prevent any degrading effect such as vortexing.

5.1.

The SIS recirculation sumps are collecting reservoirs designed to pro-vide an adequate supply of water to the CSS and SIS during the recirc-ulation mode. The SIS recirculation sumps are seismic Category I structures and are designed in accordance with Regulatory Guide 1.82 Rev. O requirements. Two SIS Recirculation Sumps, each with sufficient capacity to serve one of the redundant trains for the CSS and SIS are provided.

5.2 FSAR Figures 1.2-3 and 1.2-8 provide plant and cross-sectional views of the arrangement of the containment recirculation sump system.

5.3.a The bottom of the sump is at El 351 feet which is below the lowest floor elevation inside the containment, exluding the reactor vessel cavity. The sumps are protected by two screen barriers. FSAR Figures 6.2-28, 6.2-29 and 6.2-30 show plant and elevation details of the screen arrangements. The outer screen barrier or trash rack, is stain-less steel grating with a 1 x 4 inch clear opening. This trash rack is provided to prevent large debris from reaching the inner screen. The inner screen is a finer meshed stainless steel screen 8 x 8 mesh with 0.90 inch opening capable of filtering out 1/11 inch or larger partic-ulates.

The limitation of the maximum particle size of debris that is allowed to go through the inner screen is placed by the ECCS system which re-circulates sump water through the reactor vessel. The largest size of particles that is allowed to go through the system without impairing system performance is 0.09 inch. This limitation is substantially smaller than the CSS limit nion placed by the orifice size of the spray nozzles. The CS nozzles are of the open throat type with a 3/8 inch orifice diameter. Therefore, clogging of the CS nozzles is unlikely since the size of particles allowed through the system is substantially smaller.

A description of the trash rack and the inner screens is provided in the response to 5.3a and FSAR Subsection 6.2.2.2.

The trash rack is provided to assure that NPSH is not affected by blockage of the SIS sump by preventing large debris from reaching the inngr screen. The free area of the inner screen is appror,imately 285 ftc (not taking any credit for the top section). Assuming that ong half of the screen area is blocked, the remaining free area (142.5 ftc) will be approx-imately 30 times larger than the area of the 30 inch suction piping.

Thus, debris collected on the screens would create a neglible pressure drop (FSAR Subsection 6.2.2.3).

5.3b) Further, the primary source of post accident debris that could be gen-erated inside the containment and which could possibly clog the SIS sump screens, would be the disintegration of thermal insulation that is provided for equipment and piping. The insulation will be designed to be non-reactive under the LOCA conditions. Therefore, except for post-ulated localized damage, the insulation will withstand the environ-mental conditions with no loss in structural integrity. As a result, loose insulation would not be a f actor in clogging the sump screens.

No other type of debris is expected to clog the screens or prevent in any way the flow of water.

Also, permanent insulation assemblies are attached by stainless steel straps and fasteners of the expansion type which prevent overstressing of the bands or damage to the insulation coverings due to thermal expansion of the equipment surface. Removable assemblies are attached by means of stainless steel buckles or other fasteners of the quick release type which vary depending upon installation requirements.

In addition, a procedure will be written for, or will incorporate, an inspection of the containment and the containment sumps for any mate-rials which have the potential for becoming debris capable of blocking the containment sump screens as stated in response 1.

5.3c. Thermal pipe insulation for use inside containment has been specified, however the specific type and manufacturer has not been decided upon at the present time. This information will be provided by June, 1983.

5.3d. As stated previously, particles small enough to pass through the fine screens can pass through the system without deliterious effects. Pump operability is not expected to be impaired.

The FSAR will be revised to reflect this response,

l 1219'n'-10 Mc.f.M ;gljo

[

WNP-3 FSAR q

0 2)

Suction pipe lengths based on actual piping layouts.

[Ell,,,

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w.m..

d)

To assure that NPSH is not af f ected by blockage of the SIS sump, a

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t ra sh ra ck i s p ro vide d.

The trash rack encloses the SIS sump to i

prevent large debris f rom accumulating in the sump intake structure.

A fine inner screen with sufficient arca is provided inside the trash

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rack, to limit the maximum size particles that would be allowed to enter the system. The f ree area of the inner screen is approximately

[

285 f t2 (not taking any credit f or the top section). Assuming that P

one half gf the screen area is blocked the remaining f ree area I

I,sedd (142.5 f t ) will be approximately 30 times larger than the area of i

the 30 inch suction piping. Thus, debris collected on the screens

[

would create a neg11rible pressure drop.

Based on the foregoing the minimum available NPSH at the suction of the CS L

pumps during recirculation mode is as follass:

g f

Flow

• NPSH (awnil)

NPSH (Req)

Pe Pi Margin

.1 10,665 gpm 33.5 fe 12.5 fe 37.5 f e

'4 f t 268%

(:unout) l "he integrated energy content of the containment atmosphere and recirculation water as function of time following.the. postulated design basis less-of-coolant accident is shown on Figure 6.2-11.

The integrated energy obsorbed by the structural heat sinks and removed by the shutdown heat g

sxchanger is shown on Figure 6.2-11.

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Feilure mode and ef f ects analysis of the CSS is presented in Table 6.2.2-1.

c w-j The design of WNP-3 and WNP-5 does not employ a containment fan system as part cf the containmet heat renowl system during post-accident conditions.

6.2.2.4 Tests and inspections Preoperational testing of the CSS in conjunction with the chemical additive cubsystem is presented in Sabsection 6.5.2.4 In summary, these tests will l

dtmonstrate that, the systems are capable of fulfilling their design functions.

l The in-service inspection of the spray nozzles shall be limited to visual exami ution only. Since the size of the nozzles (3/8 it.ch crifice diameter) is substantially larger than the largest particle in the system, ( 09 inch) 3 clogging of the nozzles is unlikely, d

Pariodic testing and inspection of the system active components, i.e., pumps, j

cives, etc. will be performed in accordance with the. in-service inspection requirements of the ASME Code Section XI.

The in-service inspection and testing of the CS pumps will be performed by using the minimum recirculation line. A flow meter in the line is provided to check pump perf ormance.

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• Flow includes all ECCS and CS pumps.

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6.2-92 Amendment No. 1, (10/82)

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Ens'losure 4 Item 10 Insert 1 In recirculation mode following a LOCA there are no operating fluid systems in the contain-ment to rupture except the safety injection and containment spray systems which themselves are an essential part of the safe shutdown proce-dure and recirculation systems.

In no case would any high or moderate energy piping fail-ure compromise the functional capability of the SIS sumps when they are required.

The SIS sump design, described in detail in subsection 6.2.2.2, meets all the requirements of Regulatory Guide 1.82.

The refueling pool cavity drain (el. 384.00 ft.) is located in the opposite side of the building from the SIS sumps, and the floors above the containment water surface (el. 370.000 ft.) in the area of the SIS sump are mostly steel structures and gratings. No drain flow and channel of spray flow is expected to be released below or im-pinging on the water surface (el. 370.000 ft.)

in the area of the SIS sump which can cause significant vertical flow in the neighborhood of the sump. The relative locations of the refueling cavity drain and the SIS sumps are shown in FSAR Figure 1.2-3.

1 s