GO2-81-524, Forwards Results of Suppression Pool Temp Transient analysis.In-plant Test Will Be Conducted to Measure Difference Between Local & Bulk Pool Temps During Main Steam Valve Discharge

Forwards Results of Suppression Pool Temp Transient analysis.In-plant Test Will Be Conducted to Measure Difference Between Local & Bulk Pool Temps During Main Steam Valve Discharge

ML17276A438
Person / Time
Site:	Columbia
Issue date:	12/15/1981
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML17276A445	List: GO2-81-524, Forwards Results of Suppression Pool Temp Transient analysis.In-plant Test Will Be Conducted to Measure Difference Between Local & Bulk Pool Temps During Main Steam Valve Discharge ML17276A449
References
RTR-NUREG-0763, RTR-NUREG-0783, RTR-NUREG-763, RTR-NUREG-783 GO2-81-524, NUDOCS 8112180281
Download: ML17276A438 (8)

Text

REGULATOR Y INFORMATION DISTRIBUTION -o YSTEM

(~R IDS)

'AOCE'SSION NBR:8112180281 DOC ~ DATEP 81/12/15, NOTARIZED; NO DOCKET FACIL,'50-397 KPPSS Nuclear 'Pr o J eats Unit '2~ Washington IPubl ic Powe 05000397

'AUTH INANE AUTHOR -AF F ILI A TI ON BOUCHEY>G.D, Kashington-IPublic 'Power Supply 'System

>REC IP ~ NAME RECIPIENT AFFIL'IATION SCHNENCERF A ~

Licensing Branch '2

SUBJECT:

Forwards results of suppression pool;temp.transient analysisB In~plant ~test will be conducted [to measure difference between local 8 bulk >pool temps during main steam valve "discharge.

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Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 December 15, 1981 G02-81-524 Mr. A. Schwencer, Chief Licensing Branch No.

2 Division of Licensing Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Schwencer:

BECBVEO 8

DEC >V~9St~,

1S @atm gripy'ty gylrg~q

'>BK K8Ãggyyg yicc

Subject:

NUCLEAR PROJECT NO.

2 SUPPRESSION POOL TEMPERATURE TRANSIENT ANALYSIS AND IN-PLANT SRV TEST The purpose of this letter is to transmit the results of the suppression pool temperature transient analysis for WNP-2, and to advise you that we will perform an in-plant test to measure the difference between local and bulk pool temperatures duving main steam relief valve discharge.

Both of these were identified as open items during the Containment Systems Branch review meeting in Richland in September

1981, and in the draft SER for WNP-2.

Results of the pool temperature transient analysis are contained in the attached Report No. 14057-U(D)-1, "Suppression Pool Temperature Analysis",

prepared for Washington Public Power Supply System Nuclear Project No.

2 by Stone and Webster Engineering Corporation.

This analysis indicates that for ths cases evaluated, the maximum suppression pool peak bulk tesperatuve is 198 F.

Fov wetwell airspace pressure of 0 psi gage, this is 35 F below saturation temperature at the quencher centerline ele~ation, and thus al-lows for a local-to-bulk temperature difference of 15 F, in accordance with the acceptance crjteria of draft NUREG 0783 for steam mass flow vates of less than 42 lbs/ft /sec.

NUREG 0763 provides guidelines fov determining whether plant-specific tests may be required to measure SRV discharge loads and pool tempevature grad-ients.

The stated policy is that new plant-specific tests will be required if plant parameters affecting loads and temperature distribution are sub-stantially different from those previously tested.

According to NUREG 0763, "applicants may be able to demonstrate that discharge conditions in their plants ave sufficiently similar to conditions previously tested,to obviate the need for any new tests or to curtail the scope of tests."

8112180281 8112i5 PDR ADQCK 05000397 A

PDR

Mr. A. Schwencer Page 2

December 15, 1981 G02-81-524 1.

discharge device geometry 2.

discharge line parameters:

line length, area, volume, quencher submergence, vacuum breaker size, available pool area per quencher.

steam flow rate quencher location and orientation, and pool geometry structural characteristics of containment 3.

4, 5.

Plant parameters defined in NUREG 0763 which must be evaluated for similarity to parameters previously tested are as follows:

In-plant tests performed in the Caorso plant in Italy and in the Tokai plant in Japan are directly applicable to WNP-2.

Both Caorso and Toka'i have Mark II containments geometrically similar to WNP-2, with SRV discharge line para-meters and quencher geometry essentially identical to WNP-2.

The bottom of both the Caorso and Tokai containments are flat, while the WNP-2 containment has a trapezoidal-shaped bottom.

The Caorso containment is reinforced con-crete, while Tokai utilizes a steel containment similar to WNP-2.

Data from both the Caorso and Tokai tests have been extensively evaluated on our behalf by Burns and Roe, Inc.

Of the two in-plant tests, Caorso has been the most thoroughly investigated due to the availability of data through the Mark II Owners group.

The quencher air clearing load definition for WNP-2 was based on detailed analysis of data from the in-plant SRV tests in Caorso, and confirmed by evaluation of data from the Tokai tests.

Differences between the Caorso

plant, conditions, and WNP-2, were accounted for in the load definition report sub-mitted to the NRC in August 1980

("SRV Loads-Improved Definition and Applica-tion Methodology for Mark II Containments").

NRC review of the SRV. load defi-nition for WNP-2 considered all of the plant parameters identified above which could affect quencher air clearing loads.

Differences between Coarso and WNP-2 plant parameters were addressed by the Supply System in response to NRC questions during the licensing process.

Following discussions with the NRC during the Containment Systems Branch review meetings in Richland in September

1981, some modifications to the SRV load definition were made to account for minor differences in these parameters.

A summary of comparisons of the Caorso and WNP-2 significant plant parameters potentially affecting the quencher air-clearing load is provided below:

uencher Geometr The Caorso and WNP-2 quenchers are essentially identical.

(See FSAR question 22.053).

SRV Dischar e Line SRV discharge line lengths,

areas, volumes, and quencher submergence for Caorso and WNP-2 are similar.

Minor differences are accounted for in the response to CSB issue f47 from the September 1981 review of WNP-2.)

C y

~

Mr. A. Schwencer Page 3

December 15, 1981 G02-81-524 Vacuum Breaker Size The SRV discharge line vacuum breakers affect the reflood transient within the SRV line, and therefore the internal water level within the line for a subsequent SRV actuation.

The Caorso test"conditions included a diversity of SRV line initial water levels'ince the WNP-2 SRV load definition en-velopes the loads obse'rved at Caorso, differences in vacuum breakers are accounted for.

(See FSAR questions 22.054, 22.057, and the response to CSB issue 0'47 from the September 1981 review of WNP-2.)

Pool Area er uencher The pool surface area per quencher for WNP-2 is slightly larger in WNP-2 than in Caorso.

This difference would have no significant effect on the quencher air clearing load.

(See FSAR question 022.107, and the response to CSB issue 847 from the September 1981 review of WNP-2.)

Steam Flow Rate The steam flow rates in WNP-2 range from 236 to 252 ibm/sec.

Steam flow rates in the Caorso tests ranged from 238 to 244 ibm/sec.

The Caorso steam flow rates were within 1/ of the flow rates for the six lowest set-point SRV's at WNP-2, and within 2.5'A of the flow rates for the highest setpoint SRV at WNP-2.

These differences are not significant.

uencher Location and Orientation guenchers at both Caorso and WNP-2 are arranged around the suppression pool in an inner circle and an outer circle.

The outer quenchers in WNP-2 are farther away from the containment wall (9.95 feet) than the outer dis-charging quencher in the Caorso tests (7.5 feet).

Since bubble pressure attenuates with distance, using Caorso test pressures applied directly to the WNP-2 containment is conservative.

Pool Geometr Except for the trapezoidal-shaped bottom on WNP-2, Caorso and Tokai have essentially identical geometries to WNP-2.

Since the magnitude of the SRV quencher air clearing loads acting on the containment wall have been found to be primarily a function of proximity of containment to the quencher, the effect of the shape of the pool bottom is not significant.

Structural Characteristics of Containment Fluid/structure interaction (FSI) effects during the Caorso tests and the analytical methods used to extract rigid wall pressures from the test mea-surements are discussed in detail in the SRV load definition report sub-mitted to NRC.

As shown therein, the analytical model used to predict boun-dary pressures in a Mark II containment is in good agreement with Caorso test measurements.

Also discussed is the application of FSI effects to the

~

~

Mr. A. Schwencer Page44 December 15, 1981 G02-81-524 steel containment structure of WNP-2.

Differences between Caorso structural characteristics and WNP.-2 are thus accounted for in the SRV load definition.

(Also, see FSAR question 22.063.)

kt Based on this comparison of plant parameters, only minor differences between WNP-2 and Caorso are found to exist, and these differences are conservatively accounted f'r in the SRV load definition for WNP-2.

We have concluded that an in-plant test to confirm the adequacy of the quencher.

air-clearing load would not substantially add to the body of knowledge already obtained from other in-plant tests, and is therefore not required for WNP-2, per the guide-lines of NUREG 0763.

Suppression pool temperature response was also measured in the Caorso tests.

As previously mentioned, the only significant difference between WNP-2 and Caorso which could conceivably affect pool temperature gradients is the shape of the pool bottom.

Since the local-to-bulk pool temperature difference mea-sured in the Caorso tests, as reported in NED0-24798, was only 5oF, it does not appear likely that th~ temperature difference for WNP-2 would approach the allowable value of 15 F determined from the attached report.

However, because the NRC has questioned the influence of the trapezoidal-shaped pool bottom on flow characteristics and temperature distribution in the suppression pool during SRV discharge, and because of uncertainties which would be asso-ciated with a purely analytical approach to this problem, the Supply System commits to conducting an in-plant test to measure the local-to-bulk temperature difference.

Local temperature will be measured by temperature sensors mounted on the containment wall opposite the discharging

quencher, in accordance with the guidelines of draft NUREG 0783.

The existing suppression pool temperature monitoring system will be utilized in these tests, for measurement of both lo-cal and bulk pool temperatures.

G.

D. Bouchey Deputy Director Safety and Security EAF:kjf

Enclosure:

Report,No.

14057-U(D)-1 "Suppression Pool Temperature Analysis" cc:

R. Auluck - NRC EF Beckett - Nuclear Projects Inc.

WS Chin - BPA AI Cygelman - BSR Site (954W)

F. Eltawila NRC R. Feil - Resident Inspector JA Forrest - B8R RO SA Giusti - BPC (904)

ND Lewis - EFSEC, Olympia FA MacLean - GE, San Jose TA Mangelsdorf - BPC (982)

S. Smith - GE, San Jose RE Snaith - BSR NY JJ Verderber - BSR NY (w/I attachment)

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