ML20210S910

Forwards Containment Sys Branch Draft SER Input Re Containment Sys.Sys Acceptable

ML20210S910
Person / Time
Site:	Satsop
Issue date:	06/26/1975
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Deyoung R Office of Nuclear Reactor Regulation
References
CON-WNP-1700 NUDOCS 8605290504
Download: ML20210S910 (15)

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DRAFT SAFETY EVALUATION REP 0Kr (CONTAINMENT SYSTEMS _)

WPPSS NUCLEAR PROJECTS NOS. 345

, DOCKET NOS. 50-508/509 i

6.2.1 Containment Functional Desian The, containment (reactor building) is a steel-lined, prestrersed concrete structure with a net free volume of 3,400,000 cubic feet.

The containment structure houses the nuclear steam supply system, which includes the reactor, steam generators, reactor coolant pumps l

and pressuria.or, as well as certain components of the plant's engineered safety, feature systens. The containment is designed for an internal pressure of 44 psig and a temperature of 257'F.

The reactor containment is completely enclosed by a seismic Class I shield building. The shield building is a medium leakage reinforced concrete structure and is designed to provide biological shielding d.uring normal operations and LOCA conditions, protection for the containment from atmospheric conditions and external missiles, and a means for collection and filtration of fission product leakage from the containment vessel following a 14CA.

The applicant has analyzed the containment pressure response to postulated accidents in th,e following manner. Mass and energy release i

e rates to the containment for postulated reactor coolant system pipe breaks were based upon the data provided in CESSAR. These data were used as input to the CONTEMPT computer code (references 1 and 2), which perform transient thermodynamic calculations with appropriate con-sideration of containment heat removal systems and structural heat sinks to calculate the containment pressure response. The CONIEMPT

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• I computer code used by the applicant is basically the same code used by us to perform containment pressure analyses.

The applicant has described the methods used to determine the contain-ment design pressure in the Preliminary Safety Analysis Report (PSAR).

A spectrum of reactor coolant system break locations and sizes and steam line breaks were considered. A postulated double-ended pipe rupture at the pump suction of the reactor coolant system results in the high-1 est containment pressure and is the design basis loss-of-coolant accident for the containment design. The applicant assumed the operation of only one containment spray system train during the postulated design basis accident.

For the design basis LOCA the applicant calculated a peak containment l

pressure of about 37 psig. We have also analyzed the containment response to the DBA, and have calculated's peak containment pressure of 39 psig. The containment design pressure provides about a 12%

margin above our peak calculated pressures therefore, we conclude that the containment design pressure is adequate.

For the main steam line break analysis, the applicant used the SCN-III code to calculate the mass and energy release rates. The applicant based the analysis on conservative asstsoptions, such as assuming that the total inventory between the feedwater isolation valves and the steam generator is added to the ruptured steam generator and discharged to the containment as saturated steem. The calculated maximum con-tainment pressure for the main steam line break accident is 31.9 psig, l

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~3-which is considerably less than the DBA maximum pressure. We have also analyr.ed the containacnt pre.ssure respon,se due to a steam line f ailure and calculated a pressure of 32.3 pels. Based on our review, we conclude that the applicant's analysis was done in a conservative manner, and is acceptable.

The applicant has analysed the pressure response of tarious cr/ptain-ment interjor compartmeats to postulated high encrpy line breaks.

The compartments investigated include the reactor cavity, pressuriser compartment, and steam generator camps:'tpent. The CK-rLASil-4A code was used to calculate the mano and energy release rates for the worst credible breaks assignable to each compartment.

The RELAP-3 code was used by the applicant to calculate the subcompiirtnant pressure responses; results are suematised tu Table 1 belcw. The applicant conservatively used the maximum caleviated absolute pressure to determine the maximum differential pressures for the stbcompartments; I

i.e., the applicant noglected pressure increases on the outsido curf aces of structures. Design prrssures for the subcompartnerits will be 40%

greater than the peak calculated ditterential pressures.

TAFLE 1 RESULTS OF SUSCOMPARTM_E*lT ANALYM Postulated Peak Calculated Piping Bifferential Compartmen t Failutf Preneure_(inid)*_

Steam Generator Compartment Single-EnM 37 Hot Leg Split Pressuriser Compartment Double-inded 60 Gurge Lino fireak Reactor Cavity Singlo-Ended Hot 192 Leg Split

• Maximum pressure minus initial containment presouro (14.7 psia).

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4 We have performed confirmatory subcompartment pressure response i

calculations using the RELAP-3 computer code and have predicted pressures in good agreement with the applicant's results. We l

therefore conclude that the subcompartment design pressures are acceptable.

We have evaluated the containment system functional design in the light of the General Design Criteria stated in 10 CFR Part 50 of the

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Commission's Regulations, particularly. Criteria 16 and 50. We conclude that the containment and interior compartment designs are i

acceptable. However, we will require the applicant to demonstrate prior to plant operation the functional capability of the containment vacuum relief system.

6.2.2 Containment Heat Removal System The containment heat removal system will consist of a containeent spray system to reduce the containment pressure following postulated high energy line break accidents.

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The containment spray system will consist of two separate spray trains of equal capacity. All active components of the system will be located outside the containment vessel to facilitate maintenance operations.

j Missilo protection will be provided by physical separation of equipment.

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The system will be seismic Category I.

The containment spray purp recirculation intakes in the containment emergency sump will be en-closed by a screen assembly to prevent the entry of debris which could clog the spray nozzles. The protective screen assembly design in

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I consistent with the guidelines of Regulatory Guide 1.82.

l A high-high containment pressure signal in coincidence with a l

safety injection actuation signal will automatically actuate the con-tainment spray system. The system design will permit manual operation of pumps and valves from the control room. The spray pumps will initially take suction from the borated water storage tank. When the water in the tank reaches a low level, e switchover from injection to recirculation will be initiated automatically.

The applicant has provided an analysis which demonstrates that sufficient net positive suction head will b2 available to the spray pumps for both the injection and recirculation modes of operation.

The analysis performed is consistent with t!.e guidelines of Regulatory Guida 1.1.

Ba6ed on our review of the containment heat removat systems, we conclude that the system designs are consistent with the requirements of General Vesign Criteria 38, 39 and 40, and are therefore acceptable.

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The secondary containment is e low leakage concrete arructure which protects the containment vessel from external missiles, provides biological shielding, and provides a means of c9ntro111ng radioactive fission products in the event of leakage f rom the primary containment.

[14e annulus volume formed by the primary and secondary containment is approximately 765.000 ft. An annulus vacuum maintenance system

_will maintain the annulus at a negative press 9ra during normal plant

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operation. Upon receipt of a containment isolation signal the annulua vacuum maintenance systea will shut down. The shield building ventilation system will then start and continue to maintain the annulus at a negative pressure.

The shield building annulus will normally be maintained at a negative pressure of -10 in. w.g..

At the OL stage, the applicant will be required to demonstrate the functional capability of the shield building annulus to maintain the prescribed negative pressure.

The applicant has calculated the pressure response in the shield building annulus following the design basis LOCA. The applicant conservatively assumed the failure of the annulus vacuum maintenance system and operation of only one of two shield building ventilation system trains. The applicant used the ATEMPT code, a mcdified version of the CONTEMPT code, to perform the analysis. The applicant calcu-lared that the annulus pressure would rise to -3.8 in. w.g..

Based on our review, we conclude that the applicant's analysis is acceptable and that a negative pressure would be maintained in the shield building annulug following the containment design basis accident.

The ECCS area within the auxiliary building will also be maintained at a negative pressure folicwing a LOCA to prevent unfiltered leakage of airborne radioactive' materials to the outside atmosphere. The applicant will be required to demonstrate the functional capability of the exhaust system serving the ICCS erea such th.at a negative pressure of at least -1/4 in, v.g. will occur within 30 seconds after system af artup.

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The applicant has identified containment leak paths which could bypass the secondary containment and leak directly to the enviornment. The applicant has further committed to limit the total bypass leakage to 2.8% of the containment design leak rate, and perform local leak rate tests in accordance with the requirements of Appendix J to 10 CFR Part 50. We will conclude on the acceptability of the applicant's test program to measure the bypass leakage upon completion of the NRC Standard Technical Specifications for the plant.

6.2.4 Containment Isolation System The containment isolation system will be designed to automatically isolate piping systems that penetrate the containment to prevent outleakage of the containment atmosphere following postulated accidents. Double barrier protection, in the form of closed systems and isolation valves, will be provided to assure that no single active failure will result in the loss of containment integrity.

The containment isolation provisions, including the isolation valving and penetration piping, will be seismic Category I.

Containment isolation will be initiated by the containment isolation actuation signal on high containment pressure. Main steam and feedwater isolation is initiated by the main steam isolation signal.

High radiation signals are also used to isolate the containment vessel purge system lines.

Based on our review, we conclude that the containment isolation

' system design conforms to Cencral Design Criteria 54, 55, 56 and

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57, and the guidelines of Regulatory Guide 1.11, and is acceptable.

6.2.5 Combustible cas Control Systems _

Following a loss-of-coolant accident, hydrogen may accumulate inside the containment. The major sources of hydrogen generation include:

(1) a chemical reaction between the fuel ro'd cladding and the steam resulting from vaporization of emergency core cooling water, (2) corrosion of metals and paints by the spray solution, and (3) radiolytic decomposition of the cooling water in the reactor core and the containment sump.

The applicant's analysis of post-LOCA hydrogen generation following a loss-of-coolant accident is consistent with the guidelines of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations",

dated March 10, 1971 and indicates that the concentration in the containment would not reach 3 volume percent until approximately 14 days af ter a LOCA without recombiner operation. The recombiner will, however, be placed in operation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter a LOCA. We have, performed a similar analysis of hydrogen production and accumulation within the containment, and our results are in agreement with the applicant's.

In order to mitigate the potential for excessive hydrogen accumulation in the containment building, the applicant has proposed two redundant electric hydrogen recombiners located inside containment and a backup purge system. Operation of either recombiner will maintain the hydrogen concentratien in the containment below the limits specified

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in Regulatory Guide 1.7.

Each of the two 100% capacity electric recombiners is capable of processing 100 scfm of containment atmosphere.for post-accident hydrogen control. We have reviewed tests that have been conducted for a full-scale prototype and a production recombiner. The tests consisted of proof-of-principle tests, testing on a prototype re-combiner, environmental equalification testing and functional tests for a production recombiner.

(These tests are described in WCAP-7820 and its supplements 1-4.)

The results of these tests demonstrated that the recombiner should be capable of controlling the hydrogen in a post-LOCA containment environment. The ecombiner system will be designed to seismic Category I seismic criteria and to the IEEE requirements for an engineered safety feature.

The hydrogen purge system serves as a backup to the hydrogen recombiner system. The system will consist of two redundant subsystems and will be seismic Category I design. The hydrogen purge system will direct the containment vessel atmosphere into the shield building annulus where mixing and' holdup will occur prior to treatment by the shield building ventilation system. The hydrogen purge into the annulus will have negligible effect on the negative pressure maintained within the annulus. The containment will have redundant hydrogen sampling systems. The systems will have the capability of sampling and measuring the hydrogen concentration at the points where hydrogen may accumulate in the containment during all modes of operation.

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I-Based on our ieview of the systems provided for combustible gas control fo110 wing a postulated LOCA, we conclude that they conform to the guidelines of Regulatory Guide 1.7 and the requirements of General Design Criteria 41, 42 and 43, and are therefore acceptable.

6.2.6 - Containment Leakage Testing Program The containment design includes provisions and features to satisfy the testing requirements of Appendix J to 10 CFR Part 50.

The design of the containment penetrations and isolation valves will s

permit periodic leakage rate testing at the pressure specified in Appendix J.

Included will be those penetrations that have resilient seals and expansion bellows 3 such as persennel airlocks, equipment hatch, refueling tube blind flange, hot process line penetrations and electrical penetrations.

n The proposed reactor containment leakage testing program will comply with the requirements of Appendix J to 10 CFR Part 50.

Such compliance provides adequate a'ssurance that containment integrity can be verified throughout the service lifetime of the plant and that e

the leakage rates will be periodically checked on a timely basis to

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' assure that.they are within specified limits. Maintaining containment

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leakage ratia within such limits provides reasonable assurance that, in the event of any radioactivity releases within the containment vessel, the loss of containment atmosphere through potential leak

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paths will not be in excess of acceptable limits specified for the site.

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. t 6.2.7 ECCS Containment pressure Evaluation _

Appendix K to the 10 CFR Part 50 of the Commission's regulations requires that the effect of operation of all the containment installed pressure reducing systems and processes be included in ECCS evaluation.

For the purpose of ECCS evaluation it is conservative to minimize the containment pressure which affects the reflood rate in the core because of the resistance to steam flow in the reactor coolant loops; i.e.,

it will reduce the reflood rate.

Following a loss-of-coolant accident, the pressure in the containment building will be increased by the addition of steam and water from the primary reactor system to the containment atmosphere. Subsequently, following the initial blowdown, heat transfer from the core, primary metal structures, and steam generators to the ECCS water, will produce additional steam. This steam together with any ECCS vater spilled from the primary system will flow through the postulated break into the containment. This energy will be released to the containment during both the blowdown and later ECCS operational phases; i.e., reflood and post-reflood.

l Energy removal occurs within the containment by several means. Steam I

condensation on the containment walls and internal structures serves as a passive energy heat sink that becomes effective early in the blowdown transient. Subsequently, the operation of the containment heat removal systems such as containment sprays and f an coolers will remove steam from the containment atmosphere. When the steam removal

' rate exceeds the rate of steam addition from the primary system, the

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contalment pressure will decrease from its maximum value.

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The applicant has referenced the ECCS containment pressure calculations in CESSAR and has compared. the WPS-3 and 5 containment to the parameters used in the CESSAR calculations. The values used in CESSAR for the containment net free volume, passive heat sinks and operation of the containment heat removal system were shown to be more conservative than the design values for the WPS-3 and 5 plants.

The ECCS evaluation for CESSAR including the containment pressure calculations is currently under review by the staff and we will report our conclusions in a supplement to our CESSAR Safety Evaluation.

Provided that the CESSAR containment pressure calculations for ECCS are found to be in accordance with Appendix K to 10 CFR 50 of the Conraission's regulations, we conclude that these calculations are acceptable for use by WPS-3 and 5.

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BIBLIOGRAPIIY OF REFERENCE MATERIAL l'

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R. J. Wagner and L. L. Wheat, CONTEMPT-LT Users M'a' ual, Interin Report n

I-124-74-12.1, Acrojet Nucicar, August 1973.

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2.

L. C. Richardson, L. J. Finnegan, R. J. Wagner, and J. M. Waage, CONTEMPT - A Computer Procram for Predicting, the containment Pressure-Temperature Response to a Loss-of-Coolant Accident, IDO-17220, Phillips Petroleum Company, June 1967.

3.

R. M. Shepard, H. W. !bssrc, R.11. Mark, and P. J. Docherty, " Resting-house Mass and Enercy Release Data for Containnent Design," Topical Report, WCAP 8312, Westinghouse Electric Corporation, March 1974.

4.

D. C. Slaughterbeck, Comparison of Analytical Techniques Used to Determine Distribution of Mass and Energy in the Liquid and Vapor Regions of 'a P'.!!! Contain ent Followine a Loss-of-Coolant Accident, Special Interin Report, Idaho Nuc1 car Corporation, January 1970.

5.. R. C. Schnitt, G. E. Bingham, and J. A. Norbert, Siculated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment--

.PFinal Report, IN-1403, lddito Nuclear Corporation, December 19 70.

6.

D. C. Slaughterbeck, Review of !! cat Transfer coefficients for Condensing Steam in a Containment Buildin

• Following a Loss-of-Coolant Accident, IN-13SS, Idaho Nuclear Corj. oration, September 1970, 7.

T. Tagami, Interis Report on Safety Assessnents and Facilitics Establishr.cnt Project in Japan for Period Endine June, 1965 (No. 1),

Prepared for _the National Reactor Testing Station, February 28, 1966, (unpublished work).

6.'

H. Uchida, A. Oyama, and Y. Toga, " Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors," in Prococdinns of the Third International Conference on the Peaceful Uses _ of Atonic Encrev lic1d in Cencva, August 31-September 9. 1964, Volume 13, Session 3.9, New York:

United Nations,1965 (A/ Conf. 28/P/436)(May 1964), pp.93-104.

9.

W. H. Rettig, G. A. Jaync, l'. V. Moore, C. E. Slater, and M. L. Uptmor, RELAP A Computer Program for Reactor Blowdown Analysis, IN-1321, Idaho Nuc1 car Corporation, June 1970.

10.

F. J. Moody, " Maximum Flow Rate of a Single Components Two-Phase Mix-turc," Vol. 87, pp.134, Journal of Ilcat Transfer, February 1965.

11.

L. F. Parsly, Desien Considerations of Reactor Containment Snrav Systems, Part VI, The lleatine of Spr.r/ Drops in Air-Stcan At=cspheres, USAEC Report ORNL-TM-2412, January 1970.

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12.

H. F. Coward.and C. W. Jones, Linits of Flars. ability of Cases and Vapors, Bureau of Mine Bulletin 503, 1952.

13.

A. O. Allen, The Radiation Chenistry of Water and Aqueous Solutions, Van Nostrand Co., 1961.

14. ANS Standard ANS-5.1, Decay Enerey Release Rates Following Shutdown of Uranium-Fuel Thernal Reactors (DRAFT),. Ancrican Nuclear Society,-

Hinsdale, Illinois, October 1971.

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