ML20099H973
| ML20099H973 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 11/30/1984 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19269B210 | List:
|
| References | |
| NUDOCS 8503190552 | |
| Download: ML20099H973 (10) | |
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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)
I as part of its continuing efforts toward design and licensing standardization of nuclear power plants.
f O RESAR-SP/90 is a standard safety analysis report submitted initially for Preliminary Design Approval (PDA) in accordance with Appendix 0, " Standardization of Design; Staf f Review of Standard Designs," to l
Part 50 of Title 10 of the Code of Federal Regulations (hereinaf ter 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.
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I WAPWR/I&C/EP I.l-1 NOVEMBER, 1984 2252e:ld 8503190552 850227 PDR ADOCK 05000601 K PDR
O 1.2 GENERAL PLANT DESCRIPTION 1.2.2 Principal Design Criteria RESAR-SP/90 is designed to comply with 10CFR Part 50, Appendix A. " General Design Criteria for Nuc' lear Power Plants." The specific applications of General Design Criteria to RESAR-SP/90 are discussed in Section 3.1 of RESAR-SP/90 PDA Modulo 7, " Structural / Equipment Design."
1.2.3 Plant Description 1.2.3.6 Plant Instrumentation and Control Systems The plant instrumentation and control systems are described in detail in Chapter 7. The following is a summary of these systems.
i 1.2.3.6.1 Intearated Control Svstem
.The purpose of the MAPWR integrated control system is to regulate and maintain the plant operating conditions within prescribed limits over the entire oper-ating range. Those parameters which are monitored and controlled include RCS temperature, neutron power distribution, RCS pressure, pressurizer water level, steam generator water level, and nuclear-thermal power mismatch.
The integrated control system is comprised of the following systems which per-form control functions in order to maintain safe conditions during startup, operation, and shutdown:
o Advanced Power Control System (APCS)
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The advanced power control system provides an integrated control of these systems such that the core axial power distribution and other parameters are maintained automatically. Rather than controlling just a single mechanism such as bcron concentration or control rod posi-tion, this control system will provide an integrated response to reactivity control using the following subsystems:
HAPWR/I&C/EP l.2-1 NOVEMBER, 1984 2252e:ld
O Rod Control System l
The rod control system is designed to maintain nuclear power and reat-tor coolant tempera'ture, without challenging the protection systems, during normal operating transients. To maintain temperature within a desired control band, neutron absorbing control rods are inserted or withdrawn f rom the core.
Boron Control System The - boron control system maintains the reactor coolant boron concen-tration either automatically as directed by the APCS or by the oper-ator in such a manner that the axial nuclear power distribution and other operating conditions are maintained.
Gray Rod Control System The gray rod assemblies are used in conjunction with control rods and other mechan' .or controlling reactivity. They are either fully inserted or withdrawn under automatic control.
o Pressurizer Pressure Control The pressurizer pressure control system acts to maintain or restore l
the pressurizer pressure to the nominal operating value during normal operation or following transients.
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O o Pressurizer Water level Control The pressurizer water level control system regulates and maintains or restores pressurizer water level to its required value.
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O o Steam Generator Water Level Control System The steam generator water level control system maintains the steam generator water level within operating limits during steady state operation, and during normal transients. The water level control system also restores normal water level following a plant trip.
o Steam Dumo Control The steam dump control system controls an intentional release of steam bypassing the turbine to prevent a reactor trip following a sudden i loss of electrical load. The system ensures that stored energy and residual heat are removed following a reactor trip so that the plant can be brought to equilibrium no-load conditions without actuation of The steam dump control system is the steam generator safety valves.
also used for maintaining the plant at no-load or low load conditions and to facilitate controlled cooldown of the plant.
1.2.3.6.2 Inteorated Protection System (IPS)
During normal operation, administrative procedures and the plant control sys-tems serve to maintain the reactor in a safe state, and in the case of a fault serve to prevent damage to the three barriers (fuel clad, reactor coolant sys-tem and reactor containment building) to avoid a release of radioactive mater-ial. Certain accident conditions may occur which can cause one or more of the three barriers to be threatened. The integrated protection system (IPS) mon-itors plant parameters and automatically initiates various protective func-tions to maintain the integrity of any of the three barriers. The IPS per-forms its functions by monitoring the plant parameters using a variety of sensors, performing calculations, comparisons and logic based on those sensor inputs and actuating a variety of equipment if parameter setpoints are exceeded.
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O 1.6 MATERIAL INCORPORATED BY REFERENCE The WAPWR Instrumentation and Control /F.lectric Power module incorporates, by O 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:
A -
U.S. Nuclear Regulatory Commission review complete; USNRC acceptance letter issued.
AE -
U.S. Nuclear Regulatory Comission accepted as part of the Westinqhouse emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.
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Submitted to USNRC as background information; not undergoing formal USNRC review.
0 -
On file with USNRC; older generation report with current validity; not actively under formal USNRC review.
U -
Actively under formal USNRC review.
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TABLE 1.6-1 O
MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review Report No. Title Number Reference to the NRC Status WCAP 7907-P-A LOFTRAN Code Description Rev 0 15.0 10/11/72 A WCAP 7908 FACTRAN - A FORTRAN-IV Rev 0 15.0 9/20/72 U Code for Thermal Tran-sients in a U0 2 Fuel Rod WCAP 7979- TWINKLE - A Multidimen- Rev 0 15.0 1/7/75 A P-A(P) sional Neutron Kinetics WCAP 8028-A Computer C$de WCAP 8301(P) LOCA-IV Program: Loss- Rev 0 15.0 7/12/74 AE WCAP 8305 of-Coolant Transient Analysis WCAP 8302(P) SATAN-IV Program: Compre- Rev 0 15.0 7/12/74 AE WCAP 8306 hensive Space-Time Depen-dont Analysis of Loss-of-Coolant WCAP 8370 Westinghouse Water Reactor Rev 9A 17.1 11/14/77 A Divisions Quality Assurance Plan WCAP 8567-P(P) Improved Thermal Rev 0 15.0 7/75 A WCAP 8568 Design Procedure WCAP 8846-A Hybrid B 4C Absorber Rev 0 15.0 10/77 A Control Rod Evtluation Report WCAP 8897 Bypass Logic for the Rev 1 7.1 10/77 0 '
WCAP 8898 Westinghouse Integrated ,
Protection System WCAP-8899 Model 414 Control System Rev 0 7.1 5/77 0 WCAP 8900 Signal Selection Device WCAP 9153 414 Integrated Protection Rev 0 7.1 8/77 0 WCAP 9154 Prototype Verification Program O
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O 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7.1 Electrical, Instrumentation, and Control Drawings Table 1.7-1 lists electrical, instrumentation, and control drawings that are considered to be necessary te evaluate the safety-related features pertaining to the MAPWR.
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l.7-1 NOVEMBER, 1984 l HAPWR/I&C/EP
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TABLE 1.7-1 O
ELECTRICAL. INSTRUMENTATION AND CONTROL DRAWINGS Fiaure No. Title 7.2-1 (14 sheets) WAPWR Functional Diagrams B.3-1 (2 sheets) WAPWR AC Main Single Line Diagram 8.3-2 WAPWR DC and 120 VAC Oneline Diagram l
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- O 1.8 CONFORMANCE WITH STANDARD REVIEW PLAN t 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 NRC Standard Review Plan (NUREG-0800) pertinent to the subject module. Table 1.8-1 provides this list for the " Instrumentation and Control / Electric Power" module.
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TABLE 1.8-1 O
STANDARD REVIEW PLAN DEVIATIONS SRP AcceDtance Criteria Deviation Section (During the licensing process, certain deviations with respect to the SRP acceptance criteria applicable to the Instrumentation and Control System and the Electric Power System will be listed here as appropriate).
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