ML18191A810

Wppss Nuclear Project No. 2 Post-Construction Permit Item Drywell/Wetwell Leakage Study

ML18191A810
Person / Time
Site:	Columbia
Issue date:	04/28/1976
From:	Renberger D Washington Public Power Supply System
To:	Butler W Office of Nuclear Reactor Regulation
References
Download: ML18191A810 (19)

Text

NRC FORM 195 U.S.t NUCLEAR REGULATORV COMMISSION DOCKFT NUMOER I2 76I 50-397 FILE NUMBER NRC DISTRIP'jTION FDR PART 50 DOCKET MATERIAL FROM: DATE OF DOCUMENT TO:

Richland, Hashington 4-23-76 D.L. Renberger RECEIVED W~Butl er

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5-6-76 OLETTER QNOTORIZED PROP INPUT FORM NUMOER OF COPIES RECEIVED E$ ORIGINAL 5UNC LASSIF I ED OCOPV 40 DESCRIPTION ENCLOSURE Ltr.. notarized 4-23-76..Ltr. re. our 1-14-76 Post construction permit,itepl,DryweI 1/wetwell ltr'their 2-25-75 ltr, and our 5-].5-.75 ltr... Leakage Study. ......

'trans the following....... (1 Sigaed 5 30 Carbon'ys. Received) q

.DO NOT REMOVE JI.GKNPVI7L DG@D j PLANT NASH; Hamford 8 2 SAFETY FOR ACTION/INFORMATION ENVIRO ASSIGNED AD: ASSXGNED AD .

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PROJECT MANAGER: PROJECT MANAGER :

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Washington Public Power Supply System A JOINT OPERATING AGENCY P. 0. BOX 968 3000 GSO. WASHINGTON WAT RICHI.ANO. WASHINOTON SSSS2 PHONC (509) 946 9681 April 23, 1976 G02-76-156 Docket 50-397 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Dr. Walter R. Butler, Chief Light Water Reactors Project Branch 1-2

Subject:

WPPSS NUCLEAR PROJECT NO. 2 POST-CONSTRUCTION PERMIT ITEM DRYWELL/WETWELL LEAKAGE STUDY

References:

1) Letter, W. R. Butler to J. J. Stein, Transmitting. Request for Additional Information, dated January 14, 1975.

(GI2-75-10)

2) Letter, J. J. Stein to A. Giambusso, entitled, "Response to Request for Additional Information," Drywell/Wetwell Leakage Study, dated February,25, 1975. (G02-75-52)

3) Letter, W. R. Butler to J. J. Stein, same subject, dated May 15, 1975. (GI2-75-75)

Dear Dr. Butler:

Attached are the-responses to the NRC Containment Safety questions of Reference l. Reference 2 provided responses to your Structural Branch questions. Your concurrence to the Reference 2 responses was received in Reference 3. The attachment should complete the resolution of open NRC concerns related to this Post-Construction Permit Item.

Forty (40) copies of the attachment are being submitted for your review.

Very truly yours, D. L. RENBERGER, Assistant Director Generation and Technology DLR:GLG:cab Attachments

WPPSS NUCLEAR PROJECT NO. 2 Post-Construction Permit Item Drywell/Wetwell Leakage Study STATE OF WASHINGTON ss COUNTY OF BENTON )

D. L. RENBERGER, Being first duly sworn, deposes and says: That he is the Assistant Dir ector, Generation and Technology, for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that he is authorized to submit the foregoing on behalf of said applicant; that he has read the foregoing and knows the contents thereof; and believes the same to be true to the best of his knowledge.

DATED + ><~ +~, 1976 D. L; RENBERGER On this day personally appeared before me D. L. RENBERGER to me known to be the individual who executed the foregoing instrument and acknowledged that he signed the same as his free act and deed for the uses and purposes therein mentioned.

under my hand and seal this 49~ day of t'IVEN

, 1976.

Notary Public in and for e State of Washington Residing at

WNP-2 QUESTION 1:

The analysis of allowable bypass leakage for small primary system breaks given in the Drywell to Wetwell Leakage Study (WPPSS-74-2-R5) is an endpoint type calculation which does not track the pressure-temperature transient in the drywell and wetwell. Current plants with pressure suppression contain-ment (i.e., Mark III) do such a transient calculation which is based on a physically reasonable, yet still conservative, model and provides a more accurate estimate of the containment bypass capability. Therefore, the results of such a similar analysis (wetwell pressure, wetwell temperature and suppression pool temperature as a function of time) should be provided.

Include the effect of heat sinks in the wetwell and indicate the allowable bypass capability of the con'tainment for small breaks on the basis of this type of analysis.

RESPONSE

A transient analysis using the CONTEMPT-LT(1) computer code has been performed. The code was'modified to include the mass and energy transfer to the suppression pool from relief valve dis-charge. The limiting case is a very small reactor system break which will not automatically result in reactor depressurization.

For this limiting case, it. is assumed that the response of the plant operators is to shut the reactor down in an orderly man-ner at 100 F/hr cooldown rate. Heat sinks considered were

'such items as major support steel inside containment, the re-

.actor pedestal, the diaphragm floor and support columns and the steel and concrete of the primary containment. Based on this analysis, the allowable bypass leakage (A/ JK) is 0. 028 ft . The drywell pressure transient is shown in Figure Ql-1 along with the corresponding curves of wetwell pressure, wet-well temperature and suppression pool temperature.

The allowable bypass leakage of 0.028 ft is well above the maximum possible containment bypass leakage. Periodic testing, as discussed in Question 5, will be performed to confirm that thy containment bypass leakage does not exceed A/JX = 0.0045 Figure Ql-2 pre~ents the resulting containment tra'nsient for A/JZ = 0.0045 ft . The peak containment pressure shown in Figure Ql-2 is well below the containment design pressure.

(1) Wheat, L. L.; Wagner, R. J.; Niederauer, G. F.; Obenchainf C. F., CONTEMPT-LT--A COMPUTER PROGRAM FOR PREDICTING CONTAINMENT PRESSURE-.TEMPERATURE RESPONSE TO A LOSS-OF-COOLANT ACCIDENT, ANCR-1219, Aeroject Nuclear Company, June, 1975.

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WNP-2 QUESTION 2:

On pages III-26 and 27 of the above Referenced Report it stated that the operators could actuate the drywell and wet-is

.well sprays or depressurize the reactor vessel via the pres-sure relief valves to terminate the bypass leakage transient.

Provide clarification of your position as follows:

(a) Justify that actuation of the spray system will terminate bypass leakage or mitigate its effects for the range of small primary system break sizes.

(b) Specify the suppression pool temperature at. the time that it would be anticipated to actuate the primary system pressure relief valves and the incremental temperature rise of the pool due to relief valve operation. Discuss the accepta-bility of potential pool dynamic loads due to relief valve blowdown over the given range of pool temperature and relief valve discharge, line mass flux.

RESPONSE

(a) The results presented in Figures Ql-1 and Ql-2 demonstrate the containment transient response without actuation of containment sprays. The transient results demonstrate that containment sprays are not required to terminate bypass leakage or mitigate its effects.

. (b) The results presented in Figures Ql-1 and Ql-2 demonstrate the containment transient response without operation of the ADS system to rapidly depressurize the reactor (greater than 100oF/hr).

The transient results demonstrate that actuation of ADS to rapidly depressurize the reactor is not required to terminate bypass leakage or mitigate its effect.

The main steam relief valves may be used by the operator to provide an orderly 100oF/hr shutdown.

In the event suppression pool temperature limits are approached during the transient, the operator may actuate either of two redundant modes of RHR suppression pool cooling 'to maintain acceptable pool temperature.

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WNP-2 QUESTION 3:

Page 1-4 of the above referenced report states that only 5%

of the RHR system flow can be diverted to the wetwell spray header, whereas current Mark III plants provide full RHR flow to the containment (wetwell) spray headers after a 10-minute delay to allow the RHR system to fulfillits ECCS function.

Discuss the feasibility of providing a similar design for your plant and indicate the increase in bypass capability that such a modification would yield.

RESPONSE

As noted in the response to question 2, containment sprays are not required to terminate bypass 'leakage or mitigate its effect. Consequently, no changes in the containment spray system are contemplated.

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WNP-2 QUESTION 4:

Page IV-2 of the referenced report indicates that redundant position switches will be provided on each vacuum breaker.

Specify the sensitivity of these switches and the total equivalent bypass area by, this amount.

if each vacuum breaker was unseated RESPONSE-The design and specification of the wetwell to drywell vacuum breaker valves incorporate the following devices to ensure that the valves are completely closed.

a. A positive closure mechanism to ensure closing of each valve after exercising.

b. Magnetic capture latches to ensure that the valve is completely closed.

c. Expandable seals around the seating periphery of the valve disc.

In the event that the above mentioned devices do not function adequately and the valve disc remains unseated by an amount equivalent to the limit switch tolerance, a flow bypass leak-age area will exist. Design specifications impose a maximum limit switch tolerance of 0.01 inches and vacuum breaker valve manufacturers offer valves having limit switch toler-ances of 0.006 inches. Using the design specification limit switch tolerance of 0.01 inches, the calculated total equiva-lent bypass leakage area (A/ JK) is 0.30 square inches for eight vacuum breaker valve sets and 0.35 square inches for nine vacuum breaker valve sets. The 0.30 square inches of bypass leakage area represents approximately 7Q% of the maxi-mum allowable leakage capacity (.028 square feet); and the 0.35 square inches, represents approximately 9% of the maxi-mum allowable leakage capacity.

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QUESTION 5:

Discuss any proposed periodic drywell-wetwell leak testing for your plant,.

RESPONSE

Periodic drywell-wetwell leakage tests will be performed to verify that not exceeded.

an equivalent leakage (A/ /Z) of 0.0045 ft is This equivalent leakage (A/ M) corresponds to 16% of the allowable leakage discussed in Question l.

Details of the periodic testing methods will be included in operating license Technical Specifications.

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