ML20207S258

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Chapter 1, Introduction & General Description of Plant, to Nonproprietary RESAR-SP/90 Westinghouse Advanced PWR, Module 10, Containment Sys
ML20207S258
Person / Time
Site: 05000601
Issue date: 11/30/1986
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292G939 List:
References
NUDOCS 8703190197
Download: ML20207S258 (14)
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1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

The Westinghouse Electric Corporation (hereinafter referred to as Westinghouse) has developed this Reference Safety Analysis Report (RESAR-SP/90) for the Westinghouse Advanced Pressurized Water Reactor (WAPWR) as part of its continuing efforts toward design and licensing standardization

! of nuclear power plants. RESAR-SP/90 is a standard safety analysis report submitted initially for Preliminary Design Approval (PDA) in accordance with Appendix 0, " Standardization of Design; Staff Review of Standard Designs," to Part 50 of Title 10 of the Code of Federal Regulations (hereinafter referred to as 10CFR). The ultimate objective is to obtain a Final Design Approval (FDA) of RESAR-SP/90 followed by a rulemaking proceeding and design certification, i

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O G703190197 870309 DR ADOCK 0500 1 WAPWR-CS 1.1-1 NOVEMBER, 1986 5723e:1d
1.2 GENERAL PLANT DESCRIPTION i O 1.2.2 Principal Design Criteria RESAR-SP/90 is designed to comply with 10 CFR Part 50, Appendix A. " General Design Criteria for Nuclear Power Plants." The specific applications of General Design Criteria to RESAR-SP/90 are discussed in Section 3.1 of
RESAR-SP/90 PDA Module 7, " Structural / Equipment Design".

1.2.3 Plant Description j

The plot plan for the WAPWR nuclear power plant facility is shown on Figure 1.2-1 of RESAR-SP/90 PDA Module 3, " Introduction and Site". This plot plan shows the following major building and structures: 1) the containment building, 2) the reactor external building, 3) the turbine building, 4) the

waste disposal building, 5) the service building, and 6) the administration building. Not shown on this plot plan are the service water / cooling water building / structures and the ultimate heat sink. The containment building and the reactor external building essentially contain all the buildings, structures, systems and components that are included in the WAPWR NPB scope.

All the other major buildings listed above and shown on Figure 1.2-1 are excluded from the NPB scope. There are a few exceptions, however, where the l

NPB scope waste processing system would be located in the waste disposal l

building and the NPB scope Technical Support Center (TSC) would be located in the service building.

l The general arrangement drawings for the NPB scope containment building and reactor external building are provided in Figure 1,2-2, sheets 1 thrugh 9 of RESAR-SP/90 PDA Module 3. As shown on these drawings, the containment and the reactor external building are constructed on a ( ) meter integral +(a,e)

(common) basemat. The ( ) meter diameter spherical steel containment vessel +(a,c)

(SSCV),a ( ) meter diameter reinforced concrete shield building and a +(a,c) ,

wrap-around reactor external building (RE/B) represents the major structure located on this common basemat. The two key reference elevations for the O

WAPWR-CS 1.2-1 NOVEMBER, 1986 5723e:1d

nuclear island are the [ ] meter operating floor elevation and the'[ s -) +(**c) meter grade elevation. The [ ,) meter grado elevation corresponds to the "*'*)

containment floor elevation.

The [ ] meter diameter SSCV contains the WAPWR 3816 MWt four loop reactor +(a,c) coolant system (RCS). The major components of the RCS are the reactor vessel, four reactor coolant pumps, four steam generators, the pressurizer, and the pressurizer relief tank (PRT). Several major components of the engineered safety systems are also located in the SSCV, such as four containment O recirculation units, four accumulators, four core reflood tanks, four residual heat removal (RHR) heat exchangers, and a [ _ ] gallons emergency water +C*>c) storage tank (EWST).

The capacity of the EWST is dictated by the water volume required to fill the relatively large refueling canal associated with the WAPWR RCS. Section views A-A and B-B on sheets 8 and 9 of Figure 1.2-2 of RESAR-SP/90 PDA Module 3, and the plan elevation [ ] meters on sheet 3 depict the conical shape of the +(*'C)

EWST. The EWST is a stainless steel lined tank that is located below the O nominal containment floor level of elevation [ ] meters. A 2500 ppe boron +(a,c) concentration is maintained in this tank during normal plant operation. In the event of an accident, the four integrated safeguards system (ISS) low head l and four ISS high head pumps would take direct suction from the EWST and l

provide the required flow to the RCS and the containment spray headers. Only after all the lower elevations within the containment are flooded, would water return to the EWST via the [ ] inch diameter spillways shown on plan +(**c) elevation [ ] meters (sheet 4 of Figure 1.2-2 of RESAR-SP/90 PDA Module 3). +(**c)

The reactor external building (RE/B) essentially contains all the NPB scope systems and components not located inside the SSCV. The RE/B is located on the [ ] meters common basemat and it extends 360' around the secondary +(a,e) containment (shield building). The equipment located in the RE/B has been arranged to: 1) separate the non-safety equipment from the safety related equipment; 2) separate the Train A components from the Train B components; and

3) separate the radioactive (clean) components from the non-radioactive (dirty) couponents.

O WAPWR-CS 1.2-2 NOVEMBER, 1986 5723e:1d

The RE/B general arrangement drawings (Figure 1.2-2, sheets 1 thru 9 of c d RESAR-SP/90 PDA Module 3), show the safety related equipment generally located between building column line (A) and (H). The non-safety related equipment is generally located from column line (H) to column line (Q).. For RE/B p electrical train separation, Train A equipment has generally been located to V the right of the RE/B centerline and Train B equipment is located to the left i of the RE/B centerline. The majority of non-safety related component areas are located in radioactive control areas and the majority of safety-related

' component areas are located in non-radioactive centrol areas. The only safety-related component areas that are classified as dirty areas are the four ISS safeguard component areas (SCA) located in the shadow area beneath the sphere, between elevation [ ] meters and elevation [ ] meters. +(a,c)

It should be noted that the RE/B boundary does include the building volume commonly referred to as the shadow area beneath the sphere. This building volume below elevation [ ] meters and between the primary containment +(a,c)

(SSCV) and the secondary containment (shield building) is subdivided into j sevan dedicated and totally separated zones. One of these seven zones is

, dedicated to the non-safety related chemical and volume control system (CVCS) pumps, valves, and piping. Two of the zones are dedicated to the two I emergency feedwater system (EFWS) subsystems, while the remaining four zones serve as the four ISS safeguard component areas (SCA). Sheets 1, 2 and 4 of Figure 1.2-2 of RESAR-SP/90 PDA Module 3 depict the complete separation of these seven zones at elevation [ ] meters. +(a,e)

Several key areas of the RE/B area: 1) the main control room (MCR) located at elevation [ ] meters in the southeast corner of the RE/B, 2) the Train A and +(a,c)

B diesel generator rooms located at elevation [ ] meters and in separate +(a,c) wings of the RE/B, 3) the Train A and B Class IE switchgear rooms located at elevation [ ] meters and in separate wings of the RE/B; 4) the fuel +(a,c) l handling area located in the north wing of the RE/B and extending from l elevation [ ] meters'(grade) to elevation [ ] meters; 5) the main steam +(a,c)

I tunnel located in the south wing of RE/B and extending from elevation [ ] +(a ,c)

I meters to elevation [ ] meters; 6) the electrical penetration areas +(a,c)

O WAPWR-CS 1.2-3 NOVEMBER, 1986 5723e:1d

located on elevation [ ] meters and in the southeast and southwest +(a,c) quadrants of the RE/B; 7) the emergency feedwater storage tanks located in the south wing of the RE/B and extending from elevation [ ] meters to elevation +(a,c)

( ) meters; 8) the CCW heat exchangers located in the south wing of the +(a,c)

RE/B at elevation [ ] meters; 9) the CCW pumps located directly below the +(a,c)

CCW heat exchangers at elevation [ ] meters, and 10) the majority of the +(a,e)

HVAC equipment located at elevation [ ] meters.

It should be noted that the space. between the primary containment building (SSCV) and the secondary containment building (shield building), above elevation [ ] meters is not considered part of the RE/B. This space is +(a,c) designated the annulus area. It should also be noted that the [ ] meter +(a,c) elevation coincides with the top of the concrete cradle which supports the spherical containment. Therefore, the building volume below the concrete cradle is considered part of the RE/B while the building volume above the concrete cradle is considered the annulus area.

I 1.2.3.3 Containment Safeguards System 1.2.3.3.1 Safety Features i The safety features limit the potential radiation exposure to the public and to plant personnel following an accidental release of radioactive fission l products from the reactor system, particularly as the result of a loss-of-coolant accident (LOCA). These safety features fur.ction to localize,

! control, mitigate, and terminate such accidents, ensuring that 10 CFR 100 guidelines are not exceeded. The safety features include the following systems:

o Emergency core cooling system (ECCS) o Containment spray system o Containment fan cooler system I

O WAPWR-CS 1.2-4 NOVEMBER, 1986 5723e:1d

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o Annulus air cleanup system o Hydrogen recombiners I

1.2.3.3.1.1 Emergency Core Cooling System ,

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i The ECCS function is provided by the Integrated Safeguards System (ISS), which

! injects borated water into the reactor coolant system following a LOCA. This provides cooling to limit core damage, metal-water reactions, and fission O product release and ensures adequate shutdown margin regardless of The ISS also provides continuous long term, post-accident temperature.

cooling of thi core by recirculating borated water between the in-containment Emergency Water StorageTank(EWST)andthereactorcore. See Section 1.2.3.4 of RESAR-SP/90 PDA Module 3, " Introduction' and Site," for a more detailed discussion of the ISS. '

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I 1.2.3.3.1.2 Containment Heat Removal System l The functional performance objective of the containment heat removal system, as an engineered safety features system, is to reduce the containment '

temperature and pressure following a LOCA or main steam line break (MSLB) inside containment accident, by removing thermal energy from the containment  !

! atmosphere. These cocling systems also serve to limit offsite radiation  !

levels by reducing the pressure differential betwaen the containment f atmosphere and the external environment, thereby diminishing the driving force

! for the leakage of fission products from the containment to the environment. <

i Two separate systems are utilized to perform the containment heat removal function: the containment spray portion of the ISS and the containment fan cooler system. Those components within the ISS that perform a containment spray function are the four low head pumps, the Emergency Water Storage Tank (EWST) and the associated valves, piping, and instrumentation. Within the i containment spray ring headers are used to provide containment atmosphere coverage. The containment fan cooler system consists of four fan coolers, which are cooled by component cooling water. 1 WAPWR-CS 1.2-5 NOVEMBER, 1986 I 5723e:1d m.w-., -,e,_..-y-vwy,,_,.-, ,- _mw-- . _ - . . . . . , , --.m,..m__-.-----,,_me. - - - _ . - ,

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These two' systems combined with the containment passive he'at sinks are capable of removing sufficient sensible heat and subsequent decay heat from the containment following the hypothesized LOCA or main steam line break accident to maintain the containment design pressure, in accordance with 10CFR50, Appendix A, General Design Criteria 38, " Containment Heat Removal." ,

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i i 1.2.3.3.1.3 Annulus Air Cleanup System

The Annulus Air Cleanup System described in Subsection 9.4.5 of RESAR-SP/90 l PDA Module 13, " Auxiliary Systems," collects and processes potential airborne l

I contamination due to leakage from the steel containment system. This filtration limits the environmental activity levels following an accident.

This system also collects and processes potential airborne contamination i resulting from leakage in the ISS recirculation paths outside containment.

a i j 1.2.3.3.1.4 Hydrogen Recombiners I .

Fully redundant electrical hydrogen recombiners inside the containment reduce O the percentage of hydrogen in the post-accident containment atmosphere to below combustible levels.

t 1.2.3.8 Auxiliary Systems i

1 i 1.2.3.8.7 Plant Ventilation Systems i

Within the nuclear power block, separate ventilation systems are provided for the containment and reactor external buildings.

The containment is normally cooled by the Containment Fan Cooler System; this system is safety grade and is also used for post-accident containment

cooling. The containment further includes a system to cool the drive mechanisms located on the reactor vessel head, a reactor cavity cooling system, a preaccess filtration system, and a purge system. .

O WAPWR-CS 1,2-6 NOVEMBER, 1986 l 5723e:1d i

The HVAC system for the reactor external building is subdivided into a number of smaller subsystems, each tailored to a specific area. The Reactor External
Building Ventilation system basically serves the radiation controlled zone, l including the fuel handling and annulus areas. However, these areas are switched to the Annulus Air Cleanup System during accident conditions; this l system includes filters for control of radioactive leakage. The clean area of j the reactor external building includes safety class, redundant ventilation systems for the diesel generator rooms, the safety related equipment rooms and the main control room.

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O WAPWR-CS 1.2-7 NOVEMBER, 1986 5723e:1d

1.6 MATERIAL INCORPORATED BY REFERENCE The WAPWR Radiation Protection Module incorporates, by reference, certain topical reports. The topical reports, listed in Table 1.6-1, have been filed previously in support of other Westinghouse applications.

The legend for the review status code letter follows:

f A -

U.S. Nuclear Regulatory Commission review complete; USNRC

! acceptance letter issued.

AE -

U.S. Nuclear Regulatory Commission accepted as part of the Westinghouse emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.

B -

Submitted to USNRC as background information; not undergoing l formal USNRC review.

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On file with USNRC: older generation report with current validity; not actively urder formal USNRC review.

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Actively under formal USNRC review.

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'q TABLE 1.6-1 g MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review' Report No. Title Number Reference to the NRC Status WCAP-7709-L(P) Electric Hydrogen Recombiner Rev. 0 6.2.5.4 7/71 -

WCAP-7820(NP) for PWR Containments -

Equipment Qualification Report q (Through Supplement 7) -

6.2.5.4 8/77 -

WCAP-7907-P-A LOFTRAN Code Description Rev. 0 15.0 10/11/72 A WCAP-7908 FACTRAN-A FORTRAN-IV Code for Rev. 0 15.0 9/20/72 0 Thermal Transients in a U02 Fuel Rod -

WCAP-7979-P-A TWINKLE - A Multi-dimensional Rev. 0 15.0 1/7/75 A WCAP-8028-A Neutron Kinetics Computer Code WCAP-8170-P Calculational Model for Core Rev. 0 6.2.1 6/74 AE Reflooding After a loss-of-Coolant Accident (WREFLOOD Code)

, CAP-8305 W LOCTA-IV Program: Loss-of- Rev. 0 15.0 7/12/74 AE Coolant Transient Analysis WCAP-8306 SATAN-IV Program: Comprehensive Rev. 0 15.0 7/12/74 AE Space-Time Dependent Analysis of Loss of-Coolant WCAP-8312-A Westinghouse Mass and Energy Rev. 1 6.2.1 4/17/74 A Release Data for Containment Design WCAP-8326 Containment Pressure Analysis Rev. 0 6.2.1 12/74 AE Code - COC0 l WCAP-8370 Westinghouse Water Reactor Rev. 9A 17.1 11/14/77 A Divisions Quality Assurance Plan WCAP-8567-P Improved Thermal Design Procedure Rev. 0 15.0 6/76 A 4

WCAP-8568 WCAP-8776(NP) Corrosion Study for Determining -

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Hydrogen Generation From Aluminum and Zine During Post Accident Conditions WAPWR-CS 1.6-2 NOVEMBER, 1986 5723e:1d I- . _ - .-_ . . - - -. - - __ .

C TABLE 1.6-1 (Continued)

MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review Report No. Title Number Reference to the NRC Status WCAP-8822-P Mass and Energy Releases Follow- Rev. 0 6.2.1 9/76 U WCAP-8860 ing a Steamline Rupture WCAP-8846-A Hybrid B C Absorber Control Rev. 0 15.0 10/77 A O RodEvaldationReport WCAP-8936 Environental Qualification: Rev. 0 6.2.1 3/14/77 U Instrument Transmitter Temperature Transient Analysis WCAP-9220-P-A Westinghouse ECCS Evaluation - Rev. 1 6.2.1 12/81 A Model 1981 Version WCAP-10079 NOTRUMP: A Nodal Transient Rev. 0 6.2.1 10/10/83 U Small Greak and General Network Code O

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WAPWR-CS 1.6-3 NOVEMBER, 1986 5723e:1d

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1.8 CONFORMANCE WITH THE STANDARD REVIEW PLAN In accordance with 10CFR50.34(g), Table 1.8-1 of each PDA module identifies and evaluates deviations,from the acceptance criteria of those sections of the l

NRC Standard Review Plan (NUREG-0800) pertinent to the subject module. Table

! 1.8-1 provides this list for the " Containment Systems".

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TABLE 1.8-1

,O STANDARD REVIEW PLAN DEVIATIONS  ;

l SRP Acceptance Criteria Deviation Section i (During the licensing process, certain deviations with respect to the SRP acceptance criteria applicable to the Containment Systems module will be l

O listedhereasappropriate.)

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WAPWR-CS 1.8-2 NOVEMBER, 1936 1 5723e:1d

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