ML20198G494

Forwards Rev 1 to Draft SER & Request for Addl Info Based on Info Received from Applicant Up to & Including Amend 15 to PSAR

ML20198G494
Person / Time
Site:	Washington Public Power Supply System
Issue date:	03/04/1975
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Moore V Office of Nuclear Reactor Regulation
References
CON-WNP-1049 NUDOCS 8605290623
Download: ML20198G494 (14)

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j _,. -. A. Mode, Assistant Director for Light Water Reactors, Croup 2, RL fpsg.[6 4 REVISION 1 TO DRAFI SAFETY EVALUATION REPORT A'JD REQUEST FOR fp/

ADDIffALINFORMATION 4

Plot Narse: WPPSS 1 & 4 I

Deckat Nos.:

50-460 and 50-513 I

Licensing Stage: CP r

NSSS Supplier Babcock & Wilcox Architeet Engineer: United Engineers and Constructors l

Containment Type: Dry l

Responsible Branch & Project Manager: LWR 2-3; T. Cox Requested Completion Date: February 28, 1975 l

Applicant's Response Dates April 4, 1975 l

Review Status:

Incomplate Enclosed is Revision.1 to the Draft Safety Evaluation Report and a Request I

for Additional Informatiori from the Containment Systems Branch for the Ueshington Public rower Supply System, Units 1 & 4.

This revision and requer.c are based on information received from the applicant up to and l

including Amendment 15, r,, s_

i Ve noted '15s our Draft safety. Evaluation Report, dated November 26, 1974, that there vote three major outstanding areas. Since that tima tho applicant p

has psovided additional information in subsequent amendments to the PSAR.

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l We hava reviewed this information and are providing you with the enclosed i

i revised Draft Safety Evaluation Report. The summary of the results of our l

evaluscion are as follows:

8605290623 750304 DR ADOCK 0500 o

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Containment Pressure Annivsis E

An indicated to you in our previous correspondence on thia plant, the applicant is not using methods to predict nasa and energy telease rates to the containment-which we have found nesentabla

,on.pthe: si:ailar plants. The latest hadr2ent (Anendment 15) vns l

y,provid43 us on February 21, 1975 and described thoso new methods; l

.,however, the information contained was not sufficient to allow us l

to conclude on its acceptability. Our p)ehlem is essnatially l

with the time period about 1,000 seconds after tho accident.

l The applicant is indicating significantly lower nass and energy release rates than we have reviewed in the past and has not p.om#

adequately justified the manner in which he has calculated these

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V. A. Moore MAR 4 N l

rates. We have so stated in the enclosure and have also included a request for additional information which is also attached. We i

may note that, different from other sindar plants, the mass and i

energy release rates are important at this time (i.e.,1000 seconds) l because the WPPSS containment spray capacity is significantly lower j

than for other plants. A possible resolution to this matter is that the applicant increase spray capacity.

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2.

Subcompartment Pressure Analysis The applicant provided us with the analysis of the subcompartment pressure response in Amendment 15 received by us on February 21. 1975.

i We have reviewed this information and have been able to justify the design pressure proposed for the pressurizer cavity. However, j

the applicant has not been able to justify the analysis of the j

steam generator compartment, pipe penetaation and reactor cavity.

j We have calculated the pressure responses for these ommpartinents and our results are higher than the applicant's. In our esicu-i lations we have used the loss coefficients provided by the applicant which we believe to be lower than acceptable. We have discussed this l

matter with the applicant and he intends to reevaluate his method l

of analysis. We therefore have concluded that this matter will be resolved in a supplement to the Safety Evaluation Report and have included additional requests for information which is enclosed.

3.

Combustible Gas Control System l

The applicant has provided additional information related to the j

Combustible Gas Control System which adequately resolves our j

concerns and we have so concluded in our revision to the Draft j

Safety Evaluation Report.

We note that a draft supplement to the SER has been scheduled by the Project Manager for April 25, 1975. Meeting this schedule will be dependent on the l

applicant's date for submitting the enclosed Request for Additional Information.

l As a guide we will need about a month to completo the review after receiving the applicant's response.

Oristaat afreed by:

Robert I. Tedesco w

Robert L. Tedesco, Assistant Director for Containment Safety Division' of Tecimical Review

Enclosures:

As stated r

cc:

See next page

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V. A. Moore -

MAR 4 Y i

l ces A. Giambusso W. McDonalet F. Schroeder S. Hanauer D. Eisenhut J. Glynn

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I 3/3/75 3/3/75 3/ 3 /75 Foran AEC.)la (Rev. 9 33) AECM 02+0 ano cas se e s.es.

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REVISION 1 TO THI: DRAFT SAFETY EVALUATION (CONTAINIIENT SYSTDIS)

WASHINGTON PUBLIC POUP.R SUPPLY SYSTEMS, UNITS 1 & 4

'OCKET NOS. 50-460 & 50-513 D

6.2 containment Systems I

(no change) i 6.2.1 Containment Functional Design

- p The containment will consist of a steel-lined, reinforced concrete structure with a net free volume of 3,090,000 cubic feet. The con-tainment structure will house the nucicar steam supply system, in-cluding the reactor, steam generators, reactor coolant pumps, and pressurizer, as well as certain components of the plant engineered safety features systems. The containment is designed for an in-I ternal pressure of 52.0 psig and a temperature of 283.5'F.

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The applicant has analyzed the containment pressure response for i

the postulated loss-of-coolant accident in the following manner.

j Calculated mass and energy release rates from postulated pipe breaks i

to the containment were used as input data for the CONTRAST-S computer program, which perfonas transient thermodynamic calcula-t tions including the effects of containment heat removal systems and structural heat sinks to calculate the containment pressure.

The mass and energy released to the containment is considered in terms of the blowdown and post-blowdown phases of a loss-of-coolant accid ent. The b' lowdown phase (about 24 see) of the accident is that time interval inmediately following the occurrence of the postulated accident during which most of the energy contained i

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. in the reactor coolant system, including the primary coolant, metal

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and core stored energy is released to the containment. The post-blow-down phase (after about 24 sec) consists of the refill, reflood and i

post-reflood periods. The refill period (no refill time assumed) is

. that time during which the lower reactor vessel plenum is refilled to the bottom of the core by the ECCS. The reflood period (till about 140 sec) is that time during which the core is reflooded to its 10-foot elevation and when most of the stored energy is removed from 1

the steam generator located in the broken reactor pipe loop. The I

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stored energy in the steam generator in the intact reactor coolant loop is primarily removed during the post-reflood period (after about 140 see) of the accident.

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The CRAFI computer code was used by the applicant to determine the mass and energy addition rates to the containment during the blow-down phase and for part of the post-blowdown phase of the accident.

I However, the applicant has not adequately described the methods and assumptions used to determine the mass and energy release rates to l

the containment for the entire post-blowdown phase. We have requested 1

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additional information from the applicant on this matter and will i

conclude in a supplement to the Safety Evaluation Report.

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The applicant has also analyzed the containment pressure response I

resulting from a postulated f ailure of a main steam line within L

the containment, including consideration of possible single active i

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, f ailures in the feedwater isolation system. The applicant calcu-laced a peak containment pressure of 23.2 psig, which is well below the design pressure of the containment. We have reviewed the applicant's analysis and conclude that it is acceptable.

The applicant has analyzed the pressure response resulting from postulated pipe breaks within the containment interior compartments, such as the reactor vessel cavity, the primary shield pipe penetration, the steam generator compartments, and the pressurizer compartment.

f The applicant used the COMPRESS computer program to calculate the i

peak compartment pressure differentials. The applicant has set j

compartment design pressure differentials using at least a 40% margin between the maximum differential pressure calculated and design values i

used in the structural loading equations. We agree with this margin.

I The applicant has calculated pressure differentials of 5.3 psi for a i

l double-ended surge line rupture in the pressurizer compartment, I

20.2 psi for a double-ended hot leg rupture in a steam generator l

compartment, 180 psi for a single-ended hot leg rupture in the 1

i reactor cavity, and 371 psi in the primary shield pipe penetration.

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For the pressurizer compartment the applicant calculated a peak I

differential pressure of 5.3 psi. We have performed confirmatory i'

analyses and predicted pressure of the order of 4 psi which agrees l

with the applicant's results.

I For the steam generator compartment the applicant has not provided

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. sufficient information for use to conclude on the acceptability of the applicant's design differential pressure. We have requested additional information from the applicant on this matter and will report the results of our evaluation in a supplerent to the Safety Evaluation Report.

For the reactor cavity compartment, we have calculated a peak differential pressure significantly higher than the applicant's results and therefore cannot conclude that the design differential pressure is acceptable.

For the primary shield pipe penetration, the applicant has conserva-tively assumed a double-ended rupture of the cold leg. Although

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ve believe that the assumed break size may be overly conservative, we cannot evaluate that amount of conservatism associated with this assumption.

In performing confirmatory analysis, based on the applicant's assumed break size, uc calculate pressure higher than the applicant.

Hence, we cannot conclude that the design pressure for the pipe penetration is acceptable. We have discussed t!iis matter I

with the applicant and he intends to reevaluate his method of analysis to determine the reasons for the difference in calculated results.

In addition for the steam generator compartments, reactor cavity compartment, and primary shield pipe penetration analyses, the applicant has not adequately justified the loss coefficients and friction factors used. We have calculated peak differential pressures for the reactor i

~5-cavity compartment and primary shield pipe penetration that are higher than the applicant's results. Therefore, we are unable to conclude that the design differential pressures for these compart-I ments and pipe penetration are acceptable. We will require the applicant to justify the assumptions used in his analyses and will report on our il i

findings in a supplement to the Safety Evaluation Report.

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i We have evaluated the containment system functional design in

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accordance with the General Design Criteria stated in 10 CFR Part 50 of the Commission's Regulations and, in particular, Criteria 16 I

i and 50. However, before we can conclude that the containment and I

interior compartment design pressures are adequate, we will need additional haformation regarding the applicant's method of calculating mass and energy release rates to the containment in the long term;

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i.e, beyond the time the CRAFT code is used, the method of calculating i

loss coefficients and friction factors for use in the subcompartnent analysis, and revised subcompcrement analyses. Uc vill report our conclusions on the acceptability of analyses and adequacy of design pressures in a supplement to the Safety Evaluation Report.

i 6.2.2 Containment Heat Removal System (no change) 1 6.2.3 Containment Air Purification and Cleanup System f

(not Contain^ ent Systems Branch responsibility) m i

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6.2.4 contairnent Isolation Svnten (no change) 6.2.5 Combustible cas control Systen Replace the last two paragraphs of Section 6.2.5 with the following:

" Based on our review of the systems to be provided for combustibic gas control following a postulated loss-of-coolant accident, we conclude that the syste=s will conform to the guidelines of Reguintory Guide 1.7, meet the intent of General Design Criteria 41, 42, cnd 43, and are therefore acceptabic."

6.2.6 Containnent Lenkage Testing (no change) 9 m

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REQUEST FOR ADDITIONAL INFOPPATION UPPSS 1 & 4 DOCKET NOS. 50-460 & 50-513 6.1 The following requests are related to the containment pressure analysis:

(1) For the containment pressure analysis the CRAFT conputer code was used to calculate the mass and energy release rate to the containment from the time of rupture to about 1000 seconds.

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At this time, another conputer program was used to predict the mass and energy rates to the containment.

Provide a detailed description of this computer code and the assumptions regarding decay energy and steam quenching. Discuss the applicability of your method to both hot and cold leg breaks, and for both partial and full ECCS operations.

(2) At approximately 3500 seconds, you have modified the mass and i

energy release rates to the containment that were predicted by Babcock and Wilcox.

Provide a detailed description of the method and assumptionc used to detcenine the mass and energy release ratos to the containment after 3500 seconds.

Include in your l

discussion the assumptions made regarding the temperature of l

fluid entering the core and temperature of spray water. Discuss j

the applicability of your method to both hot and cold leg breaks, I

and both partial and full ECCS operations.

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(3) We have noted that for the containment pressure response that the pressure is still relatively high at the time of recirculation.

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At this time th'c containment spray system will become icss effective in reducing the containment pressure than it was before because of the high temperature water that is being drawn from the sump. There could be a significant precsure incraase in the containment at this time. Therefore, provide a discussion of how the recirculation time was determined assuming various singic failure modes in the ECCS and containment spray system.

In your discussion, specify the arount of watwr remaining in the borated water storage tank when you suitch to the recirculation mode.

6.2 Subcompartment Analysis We calculate significantly higher pressures for the secan generator compartment, pipe penetration, and reactor cavity. As we have discussed with you, the pressures you calculate for the pipe breaks assumed are significantly lower than what oc calculate; however, we have not been abic to determine the reasons for these differences. Therefore, ne vill require the folleuing:

(1) Justification of the ecthods and assumptions used in the analynia;

and, i

(2) Justification of the vent loss coefficients and friction factors provided in Tables 6.2-45 through 6.2-47 including identification i

of each component used in determining the overall loas coefficient and the flow area to uhich this loss coefficient applies. Also i

provide a sample calculation of a loss coefficient and friction factor that typifies your method.

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r BIBLIOG;'.P:IY nF LP:Mi':':C7 Plu 1:7.:1 1.

R. J. 1 Tanner and L. L. Uheat, CO?:TN::nT-LT U::e.rs ihnual, Interim Report I-214-74-12.1, Aerojet Nuclear, Augunt, 1973.

~2.

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3.

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6.

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7.

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8.

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10.

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t 12.

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Report OR:'L-11:-2412 Janen::y 1970.

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J. J. Di !!uano, T. D. Andercon, R. E. Enhor, and R. L. Wr.terfield, Calcu19 tion of Distr.ne.e Pretcrn for Pu.< r and Tect Panctor Siten, TID-14844, USAEC March 23, 1962.

15.

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16.

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17.

A!!S St.andard A'i3-5.1, pac;1rnerm R 1< mic not<.n ro11ou turaggt Jn;.3 of Urnniten-Fuel Thor n1 Rocct ornJ!) fat".'1, h::crican !!uclec.; Society llinsdale, Illinola, Oct.ober,1971.

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