ML20097A373
| ML20097A373 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/01/1992 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20097A369 | List: |
| References | |
| NUDOCS 9206020303 | |
| Download: ML20097A373 (666) | |
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{{#Wiki_filter:1 GE Nuclear Energy l ww tu : cc w p ,,z ,u < _ .. , O May 30,1992 ) MFN No.12192 1 Docket No. STN 52-001 1 SLK-9274 I Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Robert C. Pierson, Director Standardization and Non-Power Reactor Project Directorate Subjec: Tier I Design Certification Material for the GE ABWR Design Stage 3 Submittal Enclosed are thirty .our (34) copies of the Stage 3 GE ABWR Tier I Design Certification material. This Stage 3 materialincludes the following: A - Design descriptions and proposed inspections, tests, analyses and acceptance () criteria (ITAAC) for all of the ABWR systems for which design certification is being sought
- Tier I entries for generic issues such as equipment qualification and radiation p otection. This generic material includes technical issues for which certification veill be based on approval of design acceptance criteria (DAC).
Interface requirements and associated ITAAC as called for by 10 CFR Part 52 for those portions of the plant for which design certification is not being sought.
- A detinition of the site-related parameters which have been used as input to the ABWR design process.
This submittal represents fulfillment of the GE commitment to submit Stage 3 Tier 1 material by the end of May 1992. l l The enclosed material does not include the following information: ! 1. Comprehensive responses to ITAAC-related NRC comnients received by GE in the last few weeks. I kJ 9206020303 920530 PDR ADOCK 05200001 A PDR 1
Document Control Desk Docket No. 52 001 ( U.S. Nuclear Regulatory Commission May 29,1992 - MFN No.12192 Page 2
- 2. Tier I material for the Human Factors Engineering (HFE)/ main control room design process. This issue is being actively reviewed by GE-NRC and by mutual agreement is not addressed in the enclosed Stage 3 submittal. The HFE material-will be added to the attached Tier I document when a mutually accepteble version is inplace.
- 3. Any proposed road map material beyond that presented in the Stage 2 submittal.
GE believes there is an uncerstanding that preparation of road maps is an after-the fact exercise which can best be undertaken following resolution of NRC cominents on the Tier I material. l GE views the enclosed document as draft in the sense that we fully anticipate technical interactions with the staff and modifications of the enclosed material similar to the process which has occurred on earlier Tier I submittals. OE believes mutually acceptable schedules should now be established for NRC comments /GE responses on the entire Tier 1 package. We believe this schedule discussion should be an agenda item for the upcoming June S GE/NRC management meeting in San Jose. Sincerely,-
.I t
- P.W. Marriott, Manager
-: Regulatory and Analysis Services M/C 444, (408) 925-6948 cc: F. A. Ross (DOE)
N. D. Fletcher (DOE) C. Posiusny, Jr. (NRC) R. C. Berglund (GE) J. F. Quirk (GE) ~ l I O a l l
ABWR oesign Documcnt O 7" b ' * ' "**"** 1.1 General Plant Description 2.1 Nuclear Steam Supply 2.1.1 Reactor Pressure Vessel System 2.1.2 Nuclear Boiler System 2.1.3 Reactor Recirculation System 2.2 Control and Instrument ! 2.2.1 Rod Control and Information System 2.2.2 Control Rod Drive System 2.2.3 Feedwater Control System 2.2.4 Standby Liquid Control System 2.2.5 Neutron hionitoring System 2.2.6 Remote Shutdown System 2.2.7 Reactor Protection System 2.2.8 Recirculation Flow Control System 2.2.9 Automatic Power Regulator System 2.2.10 Steam Bypass and Pressure Control System 2.2.11 Process Computer System 2.2.12 Refueling Platform Control Computer 2.2.13 CRD Removal hiachine Control Computer -p d 2.3 Radiation hionitoring 2.3.1 Process Radiation hionitoring (PRhi) System 2.3.2 Area Radiation hionitoring System 2.3.3 Dust Radiation hionitoring System 2.3.4 Containment Atmospheric hionitoring System 2.4 Core Cooling 2.4.1 Residual Heat Removal System 2.4.2 - High Pressure Core Flooder (HPCF) System 2.4.3 Leak Detection and Isolation System 2.4.4 Reactor Core Isolation Cooling System 2.5 Reactor Senicing Equipment 2.5.1 Fuel Senicing Equipment 2.5.2 Mixellaucuus Senicing Equipmcnt 2.5.3 Reactor Pressure Vessel Senicing Equipment 2.5.4 RPV Internal Senicing Equipment 2.5.5 Refueling Equipment 2.5.6 Fuel Storage Facility 2.5.7 Under-Vessel Senicing Equipment 2.5.8 CRD hiaintenance Facility 2.5.9 Internal Pump Maintenance Facility 2.5.10 Fuel Cask Cleaning Facility p() 2.5.11 Plant Start-up Test Equipment 2.5.12 Insenice Inspection Equipment
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ABWR oesign occument 2.6 Reactor Auxilian-qj 2.6.1 Reactor Water Cleanup System 2.6.2 Fuel Pool Cooling and Cleanup System 2.6.3 Suppression Pool Cleanup System 2.7 Control Panels 2.7.1 Main Control Room Panel 2.7.2 Radioactive Waste Control Panel 2.7.3 local Control Panels 2.7.4 Instrument Racks 2.7.5 Multiplexing System 2.7.6 Local Conu ol Box 2.8 Nuclear Fuel 2.8.1 Nuclear Fuel 2.8.2 Fuel Channel 2.8.3 Control Rod 2.9 Radioactive Waste 2.9.1 Radwaste System 2.10 Power Cycle 2.10.1 Turbine Main Steam System 2.10.2 Condensate Feedwater and Condensate Air Extraction System g 2.10.3 Heater Drain and Vent System i 2.10.4 Condensate Purification System x 2.10.5 Condensate Filter Facility 2.10.6 Condensate Demineralizer 2.10.7 Main Turbine 2.10.8- Turbine Control System 2.10.9 Turbine Gland Steam System 2.10.10 Turbine Lubricating Oil System 2.10.11 Moisture Separator Heater 2.10.12 Extraction System 2.10.13 Turbine Bypass System 2.10.14 Reactor Feedwater Pump Driver 2.10.15 Turbine Auxiliary Steam System ; 2.10.16 Generator ) 2.10.17- Hydrogen Cas Cooling System 1 2.10.18 Generator Cooling System 2.10.19 Generator Sealing Oil System 2.10.20 Exciter 2.10.21 Main Condenser 2.10.22 Off-Gas System 2.10.23 Circulating Water System 2.10.24 Condenser Cleanup Facility p 2.11 Station Auxiliary V 2.11.1 Makeup Water (Purified) Systern 2.11.2 Makeup Water (Condensate) System ii 6/1/92
ABWR Design Document (~'N 2.11.3 Reactor Building Cooling Water System V 2.11.4 Turbine Building Cooling Water System 2.11.5 HVAC Normal Cooling Water System 2.11.6 HVAC Emergency Cooling Water System l 2.11.7 Oxygen Injection System i 2.11.8 Ultimate Heat Sink l 2.11.9 Reactor Senice Water System 2.11.10 Turbine Service Water System 2.11.11 Station Senice Air System 2.11.12 Instrument Air System 2.11.13 High Pressure Nitrogen Gas Supply System 2.11.14 Heating Steam and Condensate Water Return System 2.11.15 House Boiler 2.11.16 Hot Water Heating System 2.11.17 Hydrogen Water Chemistry System j 2.11.18 Zinc injection System i l 2.11.19 Breathing Air System I l 2.11.20 Process Sampling System i 2.11.21 Freeze Protection System . 2.11.22 Iron Injection System l l 2.12 Station Electrical 2.12.1 Electrical Pov,er Distribution System 2.12.2 Unit Auxiliary Transfurmer 2.12.3 Isolated Phase Bus 2.12.4 Nonsegregated Phase Bus 2.12.5 Metal Clad Switchgear 2.12.6 Power Center 2.12.7 Motor Control Center ; 2.12.8 Raceway System 2.12.9 Grounding Wire-2.12.10 Electrical Wiring Penetrations 2.12.11 Combustion Turbine Generator 2.12.12 Direct Current Power Supply 2.12.13 Emergency Diesel Genemtor System (Standby AC Power Supply) 2.12.14 Reactor Protection System Alternate Current Power Supply 2.12.15 AC Power Supply And AC Instrument and Control Power Supph-Systems 2.12.16 Instrument and Control Power Supply 2.12.17 Communication System 2.12.18 Lighting and Senicing Power Supply Systems 2.13 Power Transmission l 2.13.1 Reserve Auxiliary Transformer 2.14 Containment and Environmental Control 2.14.1 Primary Containment System 2.14.2 Containment Internal Structures 2.14.3 Reactor Pressure Vessel Pedestal iii 6/1/92 i
ABWR oesign oocument n 2.14.4 Standby Gas Treatment System () 2.14.5 2.14.6 PCV Pressure and Leak Testing Facility Atmospheric Control System 2.14.7 Drywell Cooling System 2.14.8 Flammability Control System 2.14.9 Suppression Pool Temperature Monitoring System 2.15 Structures and Senicing 2.15.1 Foundation Work 2.15.2 Turbine Pedestal 2.15.3 Crancs and Hoists 2.15.4 Elevator 2.15.5 Heating, Ventilating and Air Conditioning 2.15.6 Fire Protection System 2.15.7 Floor Leakage Detection System 2.15.8 Vacuum Sweep System 2.15.9 Decontamination System 2.i5.10 Reactor Building 2.15.11 Turbine Building 2.15.12 Control Building 2 15.13 Radwaste Building 2.15.14 Senice Building 2.16 Yard Stn ctures and Equipment l (j, 2.16.1 Stack 2.16.2 Oil Storage and Tnmsfer System 2.16.3 Site Security l l 3.1 Equipment Qualification (EQ) 3.2 Instrument Setpoint Metlu>dology 33 Piping Design 3.4 Safety System Logic and Control l 3.5 Sofware Development 3.6 Human Factors Engineering 3.7 Radiation Protection 3.8 Reliability Assurance Program 3.9 Welding 4.1 Ultimate Heat Sink 4.2 Offsite Power System 4.3 Potable and Sanitary Water System 4.4 Turbine Service Water System 4.5 Reactar Service Water System 4.6 Makeup Water Preparation System
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ABWR Design Document 4.7 Communication System O 4,8 Airborne Particulate Radiation Monitoring 4.9 Site Security 5.0 Site Parameters APPENDIX A LEGEND FOR FIGURES APPENDIX B TIER 2 ITAAC CORREI.ATION MATRICES 1 0
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ABWR oesign oocument 1.0 Introduction GE has applied f oi design certification of the Advanced Boiling Water Reactoi (ABWR) design under the provisions of lOCFF Part 51 As endorsed by the NRC commissioners, the design certification process is proceeding on the basis of a tiered approach Tier I will he the certified Rule and will include a description of the principal design bases and principal features of the certified design. Moic specifically, Tier 1 will include: (1) A design description togethei with inspections. test.s, analyses and acceptance criteria (ITAAC) entries foi cach of the approximatelv 100 systems in the ABWR f acility for which design certification is being sough t. (2) Tier 1 enuies for selected generic issues such as equipment qualification (EQ) and radiation protection. This generic material includes technical issues for which certification will be based on approvid of design acceptance criteria (DAC). (3) Interface requirements and the associated ITAAC as called for by 10CFR Part 52 for those ponions of the plant for which design certification is not being sought. O (4) A definition of the site-related parameters which have been used as input to the ARWR design process. These site-related design parameters have been selected with the intent they envelope conditions at most potential sites in the United States. The purpose of this report is to present the Tier 1 material fm the ABWR. Tier 2 will encompass the 1:uger body of design material submitted as part of the certification application . documented in the plant Safety Analysis Report I (SAR). Tier 2 material is not addressed in this report. This document is structured as follows: Section 1,1-A top-level General Plant Description intended to be a Tier 1 design description entn; This material provides a broad overview of the plant r.nd addresses general features and characteristics not covered by the more detailed system material which forms the bulk of the Tier 1 design descriptions. Examples ofissues addressed in Section 1.1 are the site plot plan, facility thennal and electrical power output, and major plant thermal-h>draulic parameters. No ITAAC are proposed for the technical entries in Section 1.1. 1.0 6/1/92 1 1
ABWR oesign Docwnent Section 2-A design description together with ITAAC entries foi cach of the apo oximately 100 srstems within the scope of the AllWR design for which d c<.ign certification is being sought. Section 3-Tier lentries that f all into the categoiy of generic. This type of enuy addresses tethnicalissues that span muhiple ABWR systems and is most appropriately handled in a single Tier 1 location. Section 3 includes a matrix showing which generic entries apply to each of the ABWR vstems. In selcued areas of the plant for which design details are (for legitimate reasons) unantilable. Design Acceptance Crii i (DAC) are being prepared. The DAC approach involves certification of the design process and is being applied to areas of the plant design for which design details are not available at the time of design certification. Section 3 inc'udes Tier 1 material that is in this category. Section 4-10 CFR Part 52 requires that ihe ITAAC include methods f or verifying interface requir ements. The latter are defined in 10 CFR Part 52 as the technical requirements to be met by those portions of the plant for which design certification is not being sought. Section 4 contains ITAAC entries required for compliance with this provision. Section 5-Design of the ABWR requires quantified ndues of many site-related characteristics such as tornado strength, flood height, and earthquale accelerations. Since it is intended the certified ABWR design be referenceable g for a wide range of sites, it has been necessary to specify a set of site design parameters enveloping the conditions which will occur at most potential power plant sites in the United States. Section 5 defines this envelope of sit.: conditions. It is intended that any facility which references the certified design will utilize a site where the actual site-specific conditions are within the defined envelope. Appendix A-A legend which defines the svmbols used to prepare simplified figures of systems presented in Section 2. Appendix Ib-GE is preparing indices which will be used to identify the relationship between Tier 2 (the SAR) entries and the Tier 1 ITAAC material. The intent of this index material is to provide a " road map" which will indicate which ITAAC entries are being used to verify kcy parameters defined in the SAR. O l 10 6/1/92
ABWR Design Documen' 1.1 General Plant Descr:ption The following is a t umman of the Adumred Boiling Water Reactor (AllWR) Standard Plant prii.cipal design features and principal dc<ign criteria. ABWR Standard Plant Scope The ABWR Standard Plant includes all buildings which are dedicated exclusively or priruarily to housing systems and the equipment related to the nuclear system or controls access to this equipment and systems. There ,ne the such buildings nithin the scope of the ABWR Standard Plant: (1) Reactor Building (including contaimnent) (2) Senice Building i (3) Control Building (4) Turbine Building (5) Radwaste Building g In addition to the buildings and their contents, the ABWR Standard Plant Q provides the supporting facilities shown in Figure 1.1. The principal plant structures include the following: 1 (}) Reactor Building-includes the containment, drywell, and major portions of the Nuclear Steam Supply System (NSSS), steam tunnel, refueling area, diesel generators, essential power, non-essential power, c.mergency Core Cooling Systems (ECCS), Heating. Ventilation and Air Conditioning System (HVAC), and supporting systems. (2) Senice Building-personnel facilities, and portions of the non essential IWAC. (3) Control Building-includes the control room, the computer facility, the cable tunnels, some of the plant essential switchgear, some of the essential power, reactor building water system and the essential HVAC System. l (4) Turbine Building-houses all equipment associated with the main turbine generator. Other auxiliary equipment is also located in this building. 1.1 6/1/92
ABWR Design Document (5) Radwaste Building-houses all equipment associated with the g collection and processing of solid and liquid radioactive waste w generated by the plant. Number of Plant Units For the purpose of this design certification, a single standard plant is described. Type of Nuclear Steam Supply This plant will have a boiling water reactor (BWPJ nuclear steam supply system (NSSS) designed by GE and designated as the Advanced Boiling Water Reactor (ABWR). Type of Containment The ABWR will have a low-leakage containment vessel comprised of the dqwell and pressure suppression chamber. The containment vessel 4 a cylindrical steel-lined reinforced concrete stmcture integrated with the Reacu>. 3uilding. Core Thermal Power Levels The information presented in this design certification pertains to one reactor unit with a rated power level of 3926 MWt and a design power level of 4005 MWt. The station utilizes a single <ycle, forced <irculation BWR designed to operate at a gross electrical power output of approximately 1356 MWe and an electrical power output of appro).imately 1300 MWe. Principal Design Parsmeters L Rated power (MWt) 3,926 Design power (MWt) (ECCS design basis) 4,005 Rated steam flow rate, Kg/hr at 215.6 C (FW temp) 7.64x10 0' Rated core coolant flow rate (Kg/hr) 52.2x10 6 RCPB design pressure (Kg/cm2g) 87.9 ! RCPB design tempemture ( C) 302 Containment internal design pressure (Rg/cm 2g) 3.16 Number of fuel assemblies 872 l Number of control rods 205 l Number of internal pumps 10 i O 1.1 6/1/92
ABWR Design Document 2.0 Tier 1 Material for ABWR Systems N This section pimides Tier 1 material for each of the altWR systems within the scope of the certified design. The listing of svstems to be addiessed in Tici 1 is devired irom and con.patible with, the AltWR systems indentified in Table 3.21 of the plant Safety Analysis Report. In most cases. cach system has (as a minimum) a Tier 1 design description which is intended to be the technical description of the facility that will appear in Tici l of the Certification Rule. Most systems also base entries defining the inspections tests, analyses and acceptance criteria (ITAAC) called for by 10 CFR Part 52. 9 Notice For a number of AllWR systems addressed in this doctunent, the Tier 1 design description is accompanied by a schematic diagram of the system configtu ation. The diagrams include simplified system piping and instrume:.tation diagmms for hydraulic / pneumatic systems; simplified one-line diagnuns for electrical systems; and simplified outline drawings for selected equipment items. These diagrams are for the purpose ofillustrating the principal design features of the ABWR systems and their relationship to each other. The simplified figures are not to scale and are not intended to be exact representations of the detailed system configurations that will be utilized in any facility referencing the certified design. The proposed AlnVRTier 1 materialincludes numericalinfonnation for aspects of the design such as equipment perfonnance, material compositions, structantl uimensions and system configurations. Where appropriate, this numerical information includes the allowable range and/or tolenmces. In those cases-where allowable variations are not specifically quantified, the stated value should be considered nominal with tolerances based on accepted industry practices as they apply to the parameter being considered. C 2.0 6/1/92
ABWR oesign Document o 2.1 Nuclear Steam Supply U 2.1.1 Reactor Pressure Vessel System Design Description The Reactor Pressure Vessel (RPV) System consists of (1) the reactor pressure vessel and its appurtenances, supports and insulation, and (2) the reactor internals enclosed by the vessel, exchiding the core, in-core nuclear instrumentation, reactor internal pumps, and connot rod drives. The reactor coolant pressure boundary (RCPB) portion of the RPV System retains integrity as a tadioactive material banier during nonnal operation and following abnormal operational transients and design basis accidents (DBAs). ! Certain RPV internals support the core, Good the core during a DBA, and support instrumentation utilized during a DBA. Other RPV internals direct coolant Dow, separate steam, hold material surveillance specimens, and support instrumentation utilized for normal operation. The RPV System provides guidance and support for the control rod drives , (CRDs). It also admits and distributes the sodium pentaborate from the Standby l Liquid Control (SLC) System. The RPV System restrains the CRD in order to preunt the ejection of the control
- rod connected with the CRD in the event of a postulated failure of the RCPB l
associated with the CRD housing. A restraint is also provided for the reactor internal pump (RIP) in order to prevent it from becoming a missile in case of a postulated failure of the RCPB associated with the reactor internal pump. t l The major plant design parameters are listed in Section 1.1. The con 6guration l of the RPV System is shown on Figure 2.1.la, with key dimensions presented in Table 2.1.lb, and the acceptable variations in these dimensions in Table 2.1.lc. The RPVS parameters (postulated break areas) used in LOCA analyses are r identified in Table 2.1.ld. 1 , 1 l Reactor Pressure Vessel, Appurtenances, Supports and Insulation The reactor pressure vessel (RPV), as shown schematically in Figure 2.1.la, consists of a vertical, cylindrical pressure vessel of welded constructicn, removable top head and head closure bolting and seals. The vessel includes the l cylindrical shell, Dange, bottom head, reactor internal pump (RIP) casings, ' penetrations, brackets, nozzles, venturi shaped How restrictors in the steam p outlet nozzles, and the shroud support, which includes the pump deck forming i d the partition between the RIP suction and discharge. The shroud support is an j ascembly consisting of a short vertical cylindrical shell, a horizontal annular l 2.1 G/1/92 1
ABWR Design Docwnent pump deck plate and vertical stilt legs. This support canies the weight of peripheral fuel assemblies, neutron sources core pla , top guide, shroud, and shroud head with steam separators. It also laterally supports the fuel assemblies and the pump diffusers. The shroud support also sustains the differential , pressures. The control rod drives (CRDs) aie mounted into the CRD housings. Sodium
}xntaborate solution from the SLC System enters the vessel via one of the two i high pressure core flooding (HPCF) lines a d is distributed through the sparger connected to the line.
The CRD housings are inserted thr ough and connected to the CRD penetrations (stub tubes) in the reactor vessel bottom head. The in-core neutron Dux monitor j housings are inserted through and connected to the bottom head. A Danged nozzle is provided in the top head for bdting of the flange associated with the instntmentation for vibration test ofinterstis. l The integral reactor vessel skirt supports the vessel on the RPV pedestal Steel anchor bolts extend through the pedestal and secure the Gange of the skin to the pedestal. RPV stabilizers are provided in the upper portion of the RPV to resist horizontal loads. Lateral supports among the CRD housings and iu< ore g housings are provided by restrai-ts which, at the periphery, are supported off W the CRD housing restraint beams. A restraint consisting of a pair of energy absorbing rods is provided to prevent a RIP from being a missile in case of a failure in the casing weld with the bottom head penetration. The restraint is connected to lugs on the RPV bottom head and the RIP motor cover. The RPV insulation is supported from the biological shitId wnll sunounding the vessel. Insulation for the upper head and Dange is supported by a steel frame t independent of the vessel and piping. Insulation access panels and insulation around penetmtions are designed for case ofinstallation and removal for vessel insenice inspection and maintenance operation. The RCPB portion of the RPV and appurtenances and the suppans (RPV skirt, stabilizer and CRD housing /in core housing restraints and beams) are classified as Quality Group A, Seismic Category I. The design, materials, manufacturing, l fabrication, testing, examination, and inspection used in the construction of l these components meet the requirements of ash!E Code Class 1 vessel and supports respectively. The shroud support is classified as Quality Group C, Seismic Category I, and designed and fabricated to ASNIE Code Class CS (core g support structures). Hydrostatic test of the RPVis perf ormed in accordance with W the requirements for ASNIE Code Chtss 1 vessels. The cmnponents are code-stamped according to their code chtss. 2.1.1 6/1/92
ABWR ocsign occument r The materials used in the RCPit portion of the RPV and apponenances (or their ( equivalents) will be used: AShlE $4533. Type IL Class 1 (platch SA-50S. Class 3 (forging); SA 508, Class 1 (forgingh Sit-166. Type 600 t UNS 06600, f orgingh SA. 182, F316L (maximum carbon 0.0209 ) or F316 i maxinmm carbon 0 020% and niuogen from 0.060 to 0.120E forgingh and SA-540, Grade B23 or B24 (bolting). The materials of the low alloy plates and f orging used in consu uction of the RPV are mehed to fine grain practice and are supplied in quenched and tempered condition Vacuum degassing is performed to lowei the hydrogen level and improve the cleanliness of the low alloy steels Electroslag welding is not applied for structural welds. Picheat and interpass temperature 3 employed for welding of low alloy steet . ret oi exceed the values given in ash!E, Section 111, Appendix 1t Post-weld heat treatment ,t 593"C minimum is applied to all low-alloy steel welds. Pressure boundary welds are giwn an ultrasonic examination in addition to the radiographic examination performed during fabrication. The ultrasonic examination method, including calihmtion, insinnuentation, scanning sensitivity, and coverage, is based on the requirements imposed by ash 1E, p Section XI, Appendix 1. Acceptance standards are equivalent or more restrictive N than required by ash 1E, Section XI. A stainless steel weld overlay is applied to the interior of the cylindrical shell and the steam outlet noule. Other noules and the RIP motor casing do not have cladding. The bottom head is clad with Ni-Cr-Fe alloy. The RIP penetrations are clad with Ni-Cr-Fe alloy or, alternatively, stainless steel. , The fracture toughness tests of pressure boundary ferritic materials, weld metal and heat-affected zone (HAZ) are perfonned in accordance with the requirements for ASNIE Code Class 1 vessel. Both longitudinal and transverse specimens are used to detennine the minimum upper shelf energy (USE) level of the core beltline materials. Separate, unirradiated baseline specimens are used to detennine the transition temperature curve of the core beltline base materials, weld metal, and HAZ. For the vessel material surveillance program, specimens are manufactured frorn , the material actually used in the reactor beltline region and weld typical of those in the heldine region, thus representing base metal, weld material, and the weld HAZ material. The plate and weld specimens are heat treated in a manner which simulates the actual heat treatment peifonned on the core region shell plates of the completed vessel. Each in-reactor surveillance capsule contains Charpy V-s notch specimens of base metal, weld metal, and ilAZ material, and tensile specituens from base metal and weld inctal. Brackets are welded to the vessel 2.1.1 6/1/92
ABWR nesinn occurnent (ladding in the core belt region for retention of the detachable holders, each of which contains a number of the specimen capsules. Neutron dosimeters and tempcature monitors are located within the capsules. L Access for examinations of the installed RPV is incorporated imo the design of the vessel, biological shield wall, and vessel insulation. Reactor Pressure VesselInternals The major reactor internal components included in the RPV Systern are: (1) Core Support Stmetures Shroud, shroud support (integnd to the RPV and includinp the internal pump deck), core plate, top guide, fuel supports (orificed furi supports and peripherai fuel supports), and control rod guide tubes.. (2) Other Reactor Internals Control rods, ferdwater spargers, RHR/ECCS low pressure Gooding spargers, ECCS high pressure core Gooding spargers and coupling, in-core guide tubes and stabilizers, core plate differential pressure (DP) lines, suneillance specimen hc!ders, shroud head and steam separators assembly, and steam dryer assembly, A general assembly drawing of these reactor internal components is shown in Figure 2.1,la. The core support stmctures locate and support the fuel assemblics, form partitions within the reactor vessel to sustain pressure different.ials across the partitions, and direct the now of the coolant water. The shroud support, shroud, and top guide make up a stainless steel cylindrical assembly that provides a partition to separate the upward Gow of coolant through the core from the downward recirculation flow. This partition separates the core region from the downcomer annulus. The core plate consists of a circuhtr stainless steel plate with round openings and is stiffened with a rim and beam structure. The core plate provides lateral support and guidance for the control rod guide tubes, in< ore Oux monitor guide tubes, peripheral fuel supports and startup neutron sources. The last two itenis are also supported vertically by tne core plate.
- The top guide consists of a circular plate edth square openings for fuel and with a cyliridrical side forming an upper shroud extension. Each opening provides
. lateral support and guidance foi fourfue! assemblics or,in the case of per;pheral ,
fuel, less thar. four fuel assemblies. Holes are provided in the bottom, where the 2.1 1 4 6/1/92 !~
ABWR Design Document sides of the openinga intersect, to anc hor the in core instrumentation detectors O- and start-up neutron sources. The fuel suppom are of two types: 1) peripheral and 2) orificed.The peripheral ; fuel supports are located at the outer edge of the active core and are not adjacent to conuol rods. Each peripheral fuel support supports one peripheral fuel assembly and contains an orifice to proside coolant Gow to the fuel assembly. , I Each orificed fuel suppor t suppm ts fom fuel assemblics veitically upmud and horizontally and contains four orifices to provide < oolant flow distribution to each fuel assembly. The orificed fuel supports rest on the top of the control rod guide tubes (CRGTs), which are supported laterally by the core plate. The
- control rods pass through cruciform openings in the center of the orificed fuel i
support. l The CRGTs located inside the vessel extend from the top of the CRD housings up through holes in the core plate. Each guide tube is designed as the guide for the lower end of a control rod and as the support for an orificed fuel support. This locates the four fuel assemblies surrounding the control rwl. The lower end of the guide tube is supported by the CRD housing, which,in turn, transmits the weight of the guide tube, fuel supports, and fuel assemblies to the reactor vessel bottom head. The CRGTs also contain holes, ri -r the top of the CRGT and below the core plate, for coolant flow to the orificed fuel supports. The CRGT base is provided with a device for coupling the CRD with it. The CRD l is restrained from ejection, in the case of a stub tube weld Iailure, by the coupling of the CRD with the CRGT base; in this event, the Gange at the top of the guide tube will contact the core plate and restrain the ejection. The coupling will also j prevent ejection if the housing fails at the stub tube weld; in this event, the guide j tube remains supported on the intact upper housing. l The control rods are cruciform-shaped neutron absorbing members that can be inserted or withdrawn from the core by the CRD to control reactivity and reactor j power. Each of the two feedwater lines is connected to dirce spargers via three RPV j nozzles. The feedwater spargers, which also function as ECCS high or low j pressure Gooding spargers (depending upon their connection to the line l l designated to receive high pressure or low pressure coolant Gooding supply, l [ resputively), are stainless steel headers located in the mixing plenum above the l downcomer annuhis. Each sparger, in two halves, with a tee connected in the middle, is fitted to each feedwater nozzle with the tee. The sparger tee inlet is connected to the RPV nozzle safe end by a double thermal sleeve arrangement. Feedwater flow enters the center of the spargers and is discharged radially inward to mix the cooler feedwmer with the down omer flow from the steam separators and steam dryer. l 2.1.1 6/1/92
ABWR oesign Document The design feature of the two residual heat removal (RHlO shutdown cooling g system spargers, which also function as ECCS low pressuie flooding (1.PFlJ W spargers,is similar to that of the feedw.uer spargers. Two lines of RHR shutdown cooling system enter the reactor vessel through the two diagonally opposite nozzles and connect to the spargers. The sparger tee inlet is connected to the RPV nozzle safe end by a ther mal sleeve arrangement. The two ECCS high pressure core flooding (HPCF) spargets and couplings are the means for directing high pressure ECCS Dow to the upper end of the core. Each of the two HPCF lines enters the reactor vessel through a diagonally opposite nozzle with a thennal sleeve arrangement. The curved sparger, including the connecting tee, is located around the inside of an i is supported hv the cylindrical portion of the top guide. The sparger tee is connected to the thermal sleeve by the HPCF coupling. In-core guide tubes (ICGTs) protect the in-core flux monitoring instrumentation from flow of water in the bottom head plenum. The ICGTs extend from the top of the in core housing to the top of the core plate. The Irx al power mnge monitoring (1.PRht) detectors for the Power Range Neutron hionitoring (PRNht) System and the detectors for the Startup Range Neutron hionitoring (SRNht) System are inserted through the guide tubes. Two levels of stamless steel stabilizer latticewor k of clamps, tie bars, anJ spacers give lateral support and rigidity to the guide tubes. The stabilizers are connected to the shroud and shroud support. The core plate differential pressure (DP) lines enter the reactor vessel through reactor bottom head penetrations. Four pairs of the core plate DP lines enter the head in four quadrants through four penetrations and terminate immediately above and below the core plate to sense the pressure in the region outside the bottom of the fuel assemblies and below the cc,re plate during normal operation. Surveillance specimen capsules, which are held in capsule holders mentioned earlier, are located at three azimuths at a com5/26/92 mon elevation in the core beltline region. The capsule holders are non-safety-related internals. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding in order to allow their rernoval and reattachment. i The shroud head and steam separators assembly includes the connecung standpipes and forms the top of the core discharge mixture plenum. The steam ) dryer assembly removes moisture from the wet steam leaving the steam j separators. The extracted moisture flows down the dryer vanes to the collecting l troughs, then flows through tubes into the dc>wucomer annulus. The shroud ' head and steam separators assembly and the steam dryer assembly are non-saferv- , related internals. ! l 2.1.1 6/1/92 l
ABWR Design Dr.;wnent The core support structures are classified as Quality Group C, Seismic Categon-
- 1. The design, materials, manufacturing. fabrication, examination, and inspection used in the constniction of the core support structures incet the requircinents of ASME Code Claw CS structures. These structures are code stamped accordingly. Other reactor internals are designed per the guidelines of ASME Code NG-3000 and are constnicled so as not to adversely affect the integrity of the core support structures as icquired by NG-1122.
Special controls aie exercised when austenitic stainless steel is used f or construction of RPV internals in order to avoid cracking during servic e , Design and constniction of the RI'Y internals assure that the internals can withstand the effects or tlow. induced vibration (FIV). Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.la provides a definition of the inspection, tests, and/or anahses together with associated acceptance criteria which will be undertaken for the Reactor Pressure Vessel System.
~
l l l I i I l l f l 2.1.1 -7 f41/92 l l
Table 2.1.1a: Reactor Pressure VesscI System } Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. Svstem configuration of the Reactor 1. Vic.ual field inspactions wi!! be conducted 1. The installed configuration of the RPV Pressure Vessel (RPV) System is shown on of the installed RPV System key System will be considered acceptable if it Figure 2.1.1a. Key dimensions are components identified in Section 2.1.1 and complies with Figures 2.1.1a, b, and c, presented in Table 2.1.1b, with design Figure 2.1.1a. Tables 2.1.1b and c and Section 2.1.1 details of RPV lower plenum and core arrangement in Figures 2.1.1b and 2.1.1c, respectively.
- 2. The RPVS parameters used in LOCA 2. Visual Celd ir spections will be conducted 2. The installed configurations of the RPVS analyses are identified in Table 2.1.1d. of the inspectiori locations ider ~ed in featues identified in Table 2.1.1d are Table 2.1.1d acceptable if the associated areas are as noted in the tab!e.
- 3. The reactor coolant pressure boundary 3. Inspections wil! be conducted of ASME 3. Existence of necessary ASME Code p (RCP.8) portion of the RPV and Code required documents and the Code required documents and the Code stamps appurtenances and their supports are stamp on the components. on the components confirm that the classified as Quality Group A, Seismic components in the RCPB portion of the Category 1. These components are RPV and the supports, and the core designed, fabricated, examined, and support structures are designed, fabricated hydrotested in accordance with the rules of and examined as ASME Code Class 1 and ASME Code Class 1 vessel or component CS, respectively. This also~ confirms that support, and are code stamped the RPV is hydrotested per the ASME Code accordingly. The core support structures Class 1 requirements.
are Ouality Group C, Seismic Category 1 and are designed, fabricated, and examined in accordance with the rules of ASME Code Class CS structures, and are code-stamped accordingly.
- 4. The RCPB of the RPV System retains its 4. A hydrostatic test of the RCPB will be 4. The results of the hydrostatic test must integrity under internal pressure that will conducted in accordance with ASME Code conform with the regt.irements in the be experienced during the servic3. requirements. ASME Code.
E '1 9 O O
(. 1 t [ Table 2.1.1a: Reactor Pressure Vessel System (Continued) Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspectior.s, Tests, Analyses Acceptance Criteria
- 5. The materials used for the RCPB portion of 5. Inspection will be conducted of the records 5. Records of the materials and processes the RPV and appurtenances ere low and of materials, fabrication, and examination must confirm that the requirements s high alloy steels with certain additional used in construction of the RCPB and specified for the RCPB in Section 2.1.1 are requirements for construction (Section austenitic stainless steel reactor internals. satisfied and that the manufacture and !
2.1.1). Special controls are exercised when fabrication of the RPV internals made of austenitic stainless steel is used for austenitic stainless steel avoid potential for [ construction of RPV internals in order to cracking in service. j avo i d cracking during service. f i
- 6. The ferritic materials used in the RCPB 6. Fracture toughness tests of the ferritic 6. Records of the fracture toughness data of '
portion of the RPV and appurtenances are base, weld and heat-affected zone (HAZ) the RCPB ferritic materials must confirm not susceptible to brittle fracture under metal used in the RCPB will be conducted that 1) the requirements of the ASME Code pressure during the service, in accordance whh the requirements for are met, and 2) the reactor vessel beltline ! ASME Class 1 components. materials will not be susceptible to brittle E fracture during the service.
- 7. Specimens for the surveillance program 7. Inspection will be conducted of the records 7. The specimens, with respect to location are selected from the vessel base metal of the spe< mens selected from the reactor and orientation, types (tensile or Charpy V-and weld metal. beltline region. notch), and quantities, must meet the i requirements of ASTM E-185.
- 8. Analysis for vibration prediction is 8. A vibration test will be conducted of the 8. Reactor vessel internals vibration is performed to assure that design and reactor internals to verify the adequacy of considered acceptable when results of the construction of the RPV interna's can the internals design, manuf acture, and vibration measurement are compared with ,
withstand the effects of flow-induced assembly with respect to the potential results of the vibration prediction analysis vibration (FIV). The design analysis is effects of FlV. The first-of-a-kind prototype to verify compliance with design limits, basede predicted values of FIV loads.The internals will be flow tested by vibration and when inspection of the intemals ; vibration prediction analysis may be instrumentation followed by inspection for indicate no sign of damage, loose parts, or t upgraded by available test data. damage. The internals in subsequent excessive wear in the prototype test. The plants will be flow tested, but without vibration of reactor internals in subsequent i vibration instrumentation, followed by plants is considered acceptable when t inspection for damage. inspection of the internals indicate no sign
, . of damage, loose parts. or excessive wear.
i i f I
" Table 2.1.1a: Reactor Pressure Vessel System (Continued) a inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspos tions, Tests, Analyses Acceptance Criteria
- 9. Access fr". )xaminations of the RPV is 9. Visual inspection will be conducted of 9. Provisions for access in the design of the incorporzd into the design of the vessel, accessibility for examinations of the vessel vessel, biological shield wall, and vessel biological shield wall and vessel insulation. and welds. insulation shall be, in the minimum, as follows:
The shield wall and vesset insulation behind the shield wall must be spaced ! away from the RPV outside surface. Access for the insertion of automated devices must be provided through removable insulation panels at the top of the shield wall and at access ports at reactor vessel nozzles. Access to the RPV welds above the
; top of the biological shield wall must be .
o provided by removable insulation panels. , The closure head must have removable ' insulation to provide access for manual ultrasonic examinations of its welds. Access to the bottom head to shell weld must be provided through openings in the RPV support pedestal and removable insulation panels around the lower cylindrical portion of the vessel. Access must be provided to partial penetration nozzle welds (i.e., CRD penetrations, instrumentation norries and recirculation internal pump penetra n welds) for performance of visual examinations. Access must be provided for examination of the attachment weld between the support skirt knuckle (forged integrally on 3 the shell ring) and the RPV support skirt. Access must be provided to the balance of
]
w the support skirt for pei formance of visual examination. O O O
l ABWR Design Document Table 2.1.1b: Key Dimensions of RPVS Components O) G Elevation / Nominal Dimension Value Description (Figure 2.1.1a) (mm) RPV inside diameter (inside cladding) G 7112.0 RPV wall thickness in beltline (without cladding) H 174.0 RPV bottom head inside invert A 0.0 Top of RPV flange F 17703.0 RPV support skirt bottom B 3200 0 RPV stabilizer connection E '3766.0 Shroud outside diameter L 5550.0 Shroud wall thickness M 50.8 Steam nozzle ID at pipe connection K 642.0 Core plate support / Top of shroud middle flange C 4695.2 Top guide support / Top of shroud top flange D 9351.2 Shroud support legs (Fig. 2.1.1b) Nx0 153.0x662.0 Control rod guide tube OD P 273.05 l l l O 2.1.1 -1 1 - 6/1/92
ABWR Design Document Table 2.1.1c: Acceptable Variations of Dimensions and Elevations g Elevation /
' Dimension Variatic-)
Description (Figure 2,1.1a) (mm) RPV inside diameter (inside cladding) G , 250.0 RPV wall thickness in beltline (without cladding) H + 14.0/-3.0 RPV bottom head inside invert A Reference 0.0 Top of RPV flange F 125.0 RPV support skirt bottom B +50.0/-10.0 RPV stabilizer connection E 115.0 Shroud outside diameter L 20.0 Shroud wall thickness M 12.0 Steam nozzle ID at pipe connection K +8.0/-0.0 Core plate support / Top of shroud middle flange C 10.0 Top guide support / Top of shroud top flange D 113.1 Shroud support legs (Fig. 2.1.1b) Nx0 16.0 (for N and Q) Control rod guide tube OD P 12.5 O 2.1.1 6/1/92
i ABWR aesign Document Table 2.1.1d: RPVS Parameters ssedin LOCA Analyses l Postulated Break Area Line inspection Location (mm7) Steamlint, Flow element throat diameter in the steam outlet no::le. 98480l
- _ _ _ _ . _ _ _ _ _ _ _ _ _ __ _ _ _ ___.y Feedwater loside diameters of flow noriles on the three sparge rs of a 83890 line for the total flow area.
RHR Injection inside diameters of flow nottles on the three spargere of a 20530 line for the total flow area. High Pressure Core inside diameters of flow notries on the three spargers of a 9200 Flooder line for the total flow area. RHR Shutdown Insloe diameter of an RHR shutdown outlet norile. 79150 Cooling Drain inside diarneter ckthe bcttom head drain outlet nozzle at 2030 the inside surf ace of the head. Note: The areas calculated from the inspectiorra shall not exceed the postulated break areas by 5 percent. g a O I 2.1. Cv'1/92
ABWR 0: sign Docunwnt O L VIB R A T10N.___.. -- > 73, ,,- . , g _ INST RUMENTATION ly h [l} M.,*d CLOSUPE HEAD N' i k E//f k, / i STE AM FLOW HESTRICTOR STEAM ('*4 Y E R g , _3,
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RESTRAINT BEAL i l - l Figure 2.1.1a Reactor Pressure Vessel Systv. Key Features j 2.1.1 -14 6/1/92
ABWR 0: sign occument
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l ABWR oosign Document I l NOTE: The arrangement is shown for quarter core only. Rotational i symmetry apphes. l t 1 r, 3
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ABWR oesign oocument 2.1.2 Nuclear Boller System ! Design Description General System Description The primary functions of the Nuclear Boiler System (NBS) are: (1) to delive: steam from the Reactor Pressure Vessel (RPV) to the hiain Steain Systein (htSS). (2) deliver feccheater from the Condensate, Feedwater, and Air Extraction System to the RPV, (3) over}uessm e protection of the Reactor Coolant Picssme Boundary (RCPB), (4) automatic dcInessuritation of the RPV in the event of a Loss of Coolant Accident (IDCA) where the RPVdoes not de}nessurire rapidly and the high pressure makeup systems fail to adequately maintain the water level in the RPY, and (5) with the exception of snonitoring the neutron flux, providing the instnimentation necessary to monito the conditions in the RPV. This includes the RPV pressure, metal temperature, and water level instrumentation. Figures 2.1.2a and 2.1.2b show the general configurution of the Siain Steam Lines (hiSLs), the Safety / Relief Valves (SRW), arul the SRV discharge lines. The SRW perform the dual function of overpresr te protection and automatic deptessuriz.. tion of the RPV. Figure 2.1.2c show the genend configuration of the Fee <hvater (FW) lines. 1 The htSLs are designed to direct steam from the RPV to the hiss, the FW lines to direct feedwater from the FW System to the RPV, and the RPV instrtunentation to monitor the conditions within the RPV, over the full range of reactor power opemtion. The NBS contains the v;dves necessary for isolation of the 51SLs, fecchvater lines, and their drain lines at the primary containment boundary. The NBS also contains the RPV head vent line and non<ondensable gas removal line. Main Steam Llnes l The NBS does not contain all of the htSLs. The NBS contains only the portion of the h1SLs from their connection to the Reactor Pressure Vessel (RPV) to the boundary with the hiain Steam System (htSS), which occurs at the seismic l interface located downstrearn of the outboard Stain Sterun isolation Valves ! (htSIW). I The main steam lines are Quality Group A from the RPV out to and including pd the outboard hfSIVs, and Quality Group B from the outboard htSIW to the l 2.1.2 6/1/92
ABWR Design Document 1 tm hine stop valves. Thev are Seismic Category I hom the scactor pressure sessel out to the scismic interface. gl I To supjxui the stfety asialysis. the total steam voluine of the stcain lines, f roin the RPV to the main steam tuihine stop valves and tuihine hvpass vahes, shall be gicater than or equal to 113.2 in 3 1
)
MSL flow limiter : Each SISL has a flow limiter. The $1S1. Ilow limiter consists of a flow iestricting venturi which is located in each RPY htSL outlet nonic. The restric tor limits the coolant blowdown rate fiom the RPV in the event a SiSL bicak occurs outside the containment to a (choke) flow rate equal to or less than 200% of rated steam flow at 72.1 kg/cm 2g upstream pressure. The h1SL flow limiter also serves as a flow element to monitor the h!SL flow. Instnnuents lines are provided to monitor the pressure at the throat of the hlSi flow limiter. The RPV steam dome pressme instnnnent lines are used to pr ovide the pressure upstream of the h!SL llow limiter. The hiSL flow limiters are designed to limit the loss of coolant from the RPY following a SISL rupture outside the containment to the extent that the RPV water level remains high enough to provide cooling within the time required to close the hf SIVs. h The htSL flow limiter has no moving parts. Main Steam Isolation Valves ! Two isolation valves are welded in a horizontal run of each of the four main steam lines; one valve inside of the dowell, and the other is near the outside of' the primary containment pressure boundag. The htSIVs are Y-pattern globe valves. The main disc or poppet is attached to the lower end of the stem. Normal steam flow tends to close the valve, and higher inlet pressure tends to hold the valve closed. The Y-pattern permits the inlet and outlet pasuges to be streamlined; this minimites pressure drop during normal j steam flow. The primary actuation mechanism utilizes a pneumatic cylinder; the speed at which the valve opens and closes can be adjus.ed. Ile:ical springs around the spring guide shafts will close the udve if gas pressuie in the actuating cylinder is reduced. The 51SIV quick closing speed is 2 3 and s 4.5 seconds when N2 or air pressure is admitted to the upper piston compartment. The vahe can be test closed with 2.1.2 6/1/92
ABWR 0: sign occument a 4540 second slow closing speed by adinitting N.; oi air to both the upper .uul p\ low piston coinpaitments. Feedwater Lines The Nil 5 does not contain all of the ITV lines. The Nits contains only t he poition of the ITY lines froin the sciunic intrif ace locaird upsticain of the Stotos. Operated Yalves ($10Vs) to thch connections to the RPV. Figure 2.1.2c shows the portion of the ISV lines within the Nils. The ITV piping consists of two 550A (22-inch) diaincter lines fioin the FW supply heades. Isolation of each line is accomplished by two containinent isolation nihes consisting of one check valve inside the drywell and one positive closing check udve outside the containtnent. Also included in this portion of the line is a rnanual inaintenance udte located between the inboard isolation udte and the reactor noule. The feedwate line upstrearn of the outboard isolation udve contains a icinote, rnanual, hiotor-Operated (h10) gate ndve, and a scisinic interface restiaint. The outhoard isolation valve and the 510 gate valve piovide a quality group transitional point in the fecchnner lines. The fecchvater piping is Quality Group A froin the RPV out to and including the outboard isolation udve. Quality Group 11 from the outbo.ud isolation udve to
\ and including the AfD gate ndve, and Quality Group I) upstr carn of the $10 gate udve. The fecchvater piping and all connected piping of 65A (21/2 -inch) or larger no'ninal size is Seismic C.uegory I f rom the RPV to the seismic interface.
Safety / Relief Valves The nuclear pressure relief systein consists of SRVs located on the h1SI.s hetween the RPV and the first isolation valve, i.e. the inboard AISIV, within the drywell. These valves protect against overpressure of the nuclear system. The rated capacity of the pressure. relieving devices shall be sullicient to prevent a rise in pressure within the [notected vessel of more than 110'7e of the design pressure (1.10 x 87.9 kg/cm g c 96.7 kg/cm'g) for design basis events which cause the RPV pressure to rise. The SRV discharge line is designed to achieve critical Dow conditions through the udve, thus provi, ling How independence to discharge pipe losses. Each SRV has its own dischargt : line. The SRV discharge lines tenninate at the quenchers located below the surface of the suppression pool. 2.1.2 -3 C./1/92
ABWR oesign Docwnent The SRW provide three snain protection luiu tions: (1) Ovel])leMule salcly ()})ct~.ltion: The nilves Ittrictioni as saf ety nihes aint opell to [11 event uticlear systen) <1WIpleMuti7atio!) - they af(* sell-acttlatilig by inlet sica!n pleMule il llot allcady sIg!)dled ()pell foi rellel operatioll. The safety (steam ptessulc) sinxle <>l <>l>craticiti is iliitiated wheti dir c< t alid klicicasilig static killet sica!!! pleut!!c tivelcoines llte It'stlaillilly Spring atid frictional foices actillg .tgaillst the inlet steam plessin e at the main disc or pilot disc and the main diw moves in the opening dhection.The condition at which this action is initiated coriespomis to the set-pressure value stamped on the nameplate of the SRV. (2) Overinessure relief opemtion: The udves aic opened using a pneumatic actuator upon receipt of an automatic or manually initiated signal to reduce pressure or to limit preMure rise. The relief (power actuated) mode of operation is initiated when an electrical signal is received at any of the solenoid udves located on the pneutnatic actuator assembly. The solenoid valve (s) will open, allowing pressurized air to enter the lower side of the pneumatic cylinder wl'ich g pushes the piston and the rod upw;uds. This action pulls the lif ting W mechanism of the main or pilot disc thereby opening the valve to allow steam to discharge through the SRV until the inlet pressure is near or equal to zero. For overpressure relief utive operation (power-actuated nu>de), pressure sensors on the RPV genemte a RPV high pressure trip signal which is used to initiate opening the SRVs. When the set pressure is reached, the SRV power-actuated relief solenoid is energiicd, which admits pneumatic pressure to the SRV actuator, thereby opening the SRV. l The SRV pneumatic operator is so armnged that, if it malfunctions, it will not prevent the SRV from opening when steam inlet pressure reaches the spring lift setpoint. l (3) Depressurization opemtion: The Automatic Depressurization System l (ADS) utives open automatically as part of the Emer gency Core Cooling System (ECCS) for events involving small breaks in the nuclear system l process barrier. Eight of the eighteen SRVs are designated as ADS valves and are capable of opemting from either ADS logic or safety /relieflogic signals. Automatic depressurization by the ADS is provided to reduce the I l 2.1.2 4- c!1l92 l l l , ._.
1 1 ABWR oesign occunwnt ica< tor pressure during e l A WA in which the liigh Piewuie Coic lbodei (llPCF) Systein and/or the Reacto Con c Isolation (:oohng (RCIC) System ;u e unable to iestoic watei level. This allows inateup of (oic cooling water by the low picuuie inateup system (the Inu Picssure Flooder (1.PFl.) Mode of the Residual lleat Removal (RilR) System). The Al)S < onsists of icthuulant it ip chaimels arnmged in two sep.u ated logics that contiol two separat":.olenoid-operated gas pilots. Al)S I aint ADS 2, on ca( h ADS SRV. Either pilot can opeiate the ADS valve. 'I hese pilots contr ol the pneumatic piewuie applied by the e cumulator s .uul the High hessure Nitrogen Gas Supply (llPIN) System. The - instnnnentation and logic power is obtained liom the Safets System logic and Contiol (SSI C) Division I and 11 Sensors froin all four divisions anal Division I control logic 10: low reactor water level and high dipvell preuure initiate ADS 1 pilots. and sensors froin all four divisions and Division 11 mitiate ADS 2 pilots. cither of which will initiate the opening of the ADS SRVs. The reactor venel low water level initiation setting f or ADh is pie. selected ' pressurire the scactor vesselin time to allow adc<1uate cooling o. ne f uel by the networ k of ECCS following a LOCA. Timely depressurization of the reactor vesselis provided if the icactor water level diops below preset limits together with an indication that high drywell pressure has occuned, which signifies there is a loss of coolant into the containment with insufficient high pressure makeup to _ maintain reactor water level. For bicaks outside the contaimnent, timely depressurization of the scactor vesselis provided if the icactor water level drops below preset limits for a time period sullicient f on the ADS high drywell pressure bypav timer and the ADS timer to time out. All SRVs have individual non-safety related accumulators. In addition, those with ADS function have a separate safety-related larger capacay accumulators with separate redundant gas power actm The ADS accumulators are sized to operate the SRV two tiines with the drywell pressure at 70% of design gauge pressure following failure of the pnemnatic supply to the accuinulator. The SRVs can be operated individually in the power-actuated nu>de by icinote manual switches located in the main control rootu. 2.1.2 C/1/92
ABWR oesign Document NBS Instrumentation The purpose of the NBS RPVinstnanentation is to inonitor and provide c ontiol input f or operation vuiables during plant operation. The NBS contains the instnunentation for monitoring the tea ( tor pressure, metal tempenuure, and water level. The reactor pressure and water lesel inst 4uments are used by multiple ftoding Water Reactoi (BWR) systeins, both safety iclated and non-safety related. Piessure indicators and tnmsmitters detect scactor vesselinter nal pressme frorn the same instniinent lines used for riscasuring icactoi veswl water level. RPV coolant teinperatures are determined by incasuring saturation pressure (which gives the saturation temperature), outlet flow teinperatuie to the Reactor Water Cleanup (CUW), and RPV bottom head drain line temperature. Reactor vessel outside surface (metal) temperature are measured at the head flange and the bottom head locations. Temperatures needed for operation and for operating liinits are obtained from these measurements: During nonnal operation, either reactor stcain saturation temperature aini/or inlet temperatures of'he scactor coolant to the CUW System and the RPV bottom head drain can be used detecmine the RPV coolant temperuure. Figure 2.1.2e shows the water level and RPV penetrations for each water level range. The mstnaments that sense the water level are all differential pressure devices calibrated for a specific RPV presnue (and corresponding lk}uid temperatur e) conditions. The water level measurement design is the cosidensate reference chamber type. Instnunent zero for all the RPV water level ranges is the top of the active fuel. The following is a description of each water level range shown on Figure 2.1.2c. (1) Shutdown Range Water Level. This range is used to monitor the reactor water level during shutdown condition when the reactor system is flooded for maintenance and head removal. The two RPV instnunent penetrations elevations u. sed for this watet level measurement are located at the top of the RPV head and the instrument tapjust below the dryer skirt. (2) Narrow Range Water I.evel. This range is used to monitor reactor water level during normal power operation. This ranges uses the RPV taps at the elevations near the top of the steam outlet nozzles and the taps at the elevation near the bottom of the dryer skirt. The Feedwater Control (FDWC) System uses this h range for its water level control and indication inputs. 2.1.2 6- 6/1/92
ABWR oesign Document (3) Wide Range Water i esel.
,\
This range i, used to inonitoi se.ition wate lesel foi esents where the water level exreeds the sange of the murow range water level instnnnentation, and is used to generate the low icactor water lesel trip signals which indicate a potential !.( WA. This range uses the RPV taps at the elevations nea the top of the sicam outlet nonles ;uul the tap below the Top of the At tive Fuel (TAF). (4) Fuel Zone Range Wate: 1.crel. This range is provided f oi the post an ident monito ing, aint provides the capabihty to monitor the scartor water level below the wide range water levelinstrumentation. This range uses the RPV taps at tiu-elevations near the top of the stcain outlet noules and the taps below the TAF (above pump deck L The Nits contains the insinunent lines to monitor the differentialI ncmne across the RPV ptunp deck and core support plate.1 hc instnnnentation which actually perfonns these functions is located within the Reciirulation Flow Control (RFC) System. i The SRVs are provided with guition sensors which provide positive indication of SRV disk / stem position. Thernu> couples are located in the discharge exhaust pipe of the SRVs. The temperature signal goes to a multipoint iccorder with an alarm and will be activated by any temperature in excess of a set temperatuie signaling that one of the SRV seats has started to leak. The Nits also contains the dnwell pressure instnnnentation used to genemte the safety related high drywell pressure trip 1.OCA signal, which is used by many of the safety telated systems to initiate safety actions. The Reactor Piotection System (RPS) utilizes this signal as a scnun initiation signal. The Leak Detection L and Isolation System (LDS) utilires this signal to initiate containment isolation. The Emergency Core Cooling Systems (ECCSs) utilizes this signal as a system
- initiation signal.
l Control room indication and/or alarms are provided for the important plant parameters monitored by the Nits. l l \ 2.1.2 7- 6'132
. _ . - .. .. -__ _. . . . - - . - - . .- = ._ _ -. . -
ABWR oesign oocwnent ASME Code Requirements The major mechanical components are designed to ineet American Societs of hiechanical l{ngineens ( ASMD Code Requirement.s as shc wn helow: ASME Design Conditions Component Code Class Pressure Temperature FW lines from the MOVs to the 2 87.9 kg'cm?g 3n2'C outboard containment isolation (1250 psig) (5'5"F) check valves 2 FW lines from the outboard 1 87.9 ka'cm 9 302 C containment isolation check (1250 psig) (5750F) valve to the RPV Feedwater (FW) line outboard 1 87.9 kg'cm?g 302"C containment isolation check (1250 psig) (5750F) valve Main Steam Isolation Valves 1 96.7 kg/cm2g 308 C (MSIVs) (1375 psig) (586.*F) Safety / Relief Valves (SRVs) 1 96.7 kg'cm2g 308 C (1375 psig) (586. F) Main Steam Lines (MSLs), from 1 87.9 kg/cm20 302 C Reactor Pressure Vessel (RPV)to (1250 psig) (575 F) outboard MSIVs MSLs from the outboard MSIVs 2 87.9 kO /cm2g 302 C to the seismic interface restraint (1250 psig) (575*F) SRV discharge line piping, from 3 38.0 kg/cm2g 250"C the SRVs to the diaphragm floor (540 psig) (482 F) SRV discharge line piping, from 2 38.0 kg/cm2g 250 C the diaphragm floor to the (540 psig) (482"F) suppression pool surf ace Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.2 proddes a definition of the instructions, tests and/or analyses together with associated acceptance criteria rhich will be undertaken for the NILS. O 2.1.2 6/132 i
Table 2.1.2: Nuclear Boiler System l i inspections, Tests, Analyses and Acceptance Critena ' Certified Design Commitment Inspections, Test, Analysis Acceptance Criteria i
- 1. A simplified configuration of the Main 1. Visual field inspection will be conducted to 1. The system configuration is in accordance !
Steam Lines (MSts), and Feedwate; (FW) confirm that the installed equipment is in with Figures 2.1.2a,2.1.25, and 2.12c. lines within the Nuclear Bc, iter System compliance with the design configuration (NBS) scope, and the Safety / Relief Valve defined in Figures 2.1.2a,2.12b and 2.1.2c. (SRVs) and the Safety / Relief Valve (SRV) discharge lines, as described in Section 2.1.2 and shown in Figures 2.1.2a,2.1.2b, and 2.1.2c.
- 2. The Reactor Coolant Pressure Boundary 2. Inspections will be conducted of ASME 2. The components have appropriate ASME .
(RCPB) portions of the NBS are classified Code required documents and the Code - Code, Section ill, Class 1 certifications and as American Society of Mechanical stamp on the actual components to verify Code Stamps. l Engineers (ASME) Code Class 1. They are that they have been manufactured porthe ! designed, fabricated, examined and relevant ASME requirements. i
'? hydrotested per the rules of the ASME Code, Section liL r This includes the MSLs from the Reactor Pressure Vessel (RPV) to and including the !
i outboard Main Steam isolation Valves (MSIVs), the FW Iines from the outboard ' positive closing check valves to the RPV.
- 3. Each Main Steam Line (MSL) shalt have a 3. Using the as-bunt dimensions, perform an 3. Analysis confirms that the MSL flow l flow limiter located in the RPV MSL outlet analysis which shows that the MSL flow limiters perform their intended function. l nozzle. The MSL flow limiter shall limit the limiters satisfy the requirement. !
coolant blowdown rate from the RPV in the event of r MSL break to a (choke) flow rate I equal to, or less than 200% of rated steam flow at 72.1 kg/cm2g upstream pressure. 4 Each MSL flow limiter has taps for two 4. Visual inspection will be conducted to 4 Inspection confirm that the MSL flow l q instrument lines. These instrument fines confirm that the MSL instrument lines have instrument lines have been installed.
$ are used for monitoring the flow through been installed in compliance with design ,
each MSL. commitment. I
?
[ Table 2.1.2: Nuclear Boiler System (Continued) w inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Test, Analysis Acceptance Criteria
- 5. The total steam line volume from the RPV 5. Using the as designed configuration of the 5. Calculations confirms that the steam line to the main steam turbine stop valves and steam lines perform calculations to volume satisfies the design requirement.
steam bypass valves shall bs greater than determine the main steari. '.ine volume. or equal to 113.2 m 3
- 6. The MSIVs meet the requirements of 6. Inspections will be conducted of ASME 6. The MSIVs have appropriate ASME Code.
ASME Code, Section Ill. Code required documents and the Code Section !!!, Class 1 certifications and code Stamp on the actual components to verify stamps. that they have been manufactured per the relevant ASME requirements.
- 7. The Main Steam Isciation Valve (MSIV) 7. Pre-operational tests will be conducted to 7. Pre-operational tests confirms that the closing time shall be between 3 and 4.5 demonstrate proper operation of the MSIVs satisfy the closure time seconds when N2 or air is admitted into the MSIVs, INit, ding verification of the closure requirement.
L valve pneumatic actuator. time. 9
- 8. The SRVs meet the requirements of ASME 8. Inspections will be conducted of ASME 8. The SRV have appropriate ASME Code, Code Section 111. Code required documents and the Code Section !!!, Class 1 certif cations and code Stamp on the actual components to verify stamps.
that they have been manufactured per the relevant ASME requirements.
- 9. There shall be 18 SRVs mounted on the 9. Inspections will be conducted to confirm 9. Inspection confirms that the SRVs have the MSLs as shown in Figure 2.1.2a. The that the SRVs have the required (nominal) required capacities and set pressures required spring set pressure and capacities spring set pressure and (minimum) identified on their name plates.
are given in Table 2.1.2n. %e SRVs shall capacity on the SRV nameplate. meet the opening performance shown in inspections confirms that the proper Figure 2.1.2f. Visual inspections will be condtsted to capacity and set pressure SRV has been confirm that a!! 18 SRVs have been mounted in its correct location. installed in their proper locations. Confirm that the selected SRV model Review of the qualification test data for the satisfies the performance re uirements, particular SRV model selected to e nfirm $ that the opening performance complies 8 with the requirements. O O O
h-t [ Table 2.1.2: Nuclear Boiler System (Continued) b
- Inspections, Tests, Analyses and Acceptance Criteria i
Certified Design Commitment inspections, Test, Analysis Acceptance Criteria
- 10. The SRVs shall be provided with 10. Inspection will be performed that the SRVs 10. Inspection confirms that the SRVs have instrumentation which will provide positive have positive position indication positive position indication.
indication (i.e. by direct measurement) oi instrumentation, and that the SRV position. instrumentation has been properly ! connected.
- 11. A simplified configuration of the Automatic 11. Visual field insoection will be conducted to 11. The configuration is in accordance with r Depressurization System (ADS) SRVs and confirm that the instslied equipment is in Figure 2.1.2d. !
the non. ADS SRVs as described in Section compliance with Figure 2.1.2d. 2.1.2 and Figure 2.1.2.d. There are 8 ADS SRVs and 18 non-ADS SRVs. !
- 12. Upon receipt of a either a high drywell 12. Logic and instrument functional testing 12. The drywell pressure and RPV water level 5 pressure trip signal current with a RPV low shall be performed to demonstrate that the instrumentation, es weil as the ADS logic, f water level 1 trip signal of sufficient ADS logic performs as required. functions as required to generate the ADS :
duration for the ADS timer to time-out, or a '
- initiation signal.
RPV low water level 1 trip signal of ! sufficient duration for the ADS high drywell i pressure bypass timer and the ADS timer to time-out, the ADS logic generates a ADS initiation signal to the SRV ADS solenoids.
; 13. The SRV discharge lines shall terminate at 13. Visual inspections will confirm that the SRV 13. Inspection confirms that the SRV discharge l 2 the quenchers located below the surf ace of discharge line quenchers have been line quenchers have been installed.
the suppression pool. installed. j 14. The RPV shall be provided with instrument 14. Visual inspections wi!! be performed to 14. Inspection confirms that the ! lines and instrumentation necessary to confirm that the instrument lines and inswumentation has been property ! monitor the RPV steam dome pressure and instrumentation for the RPV steam dome installed. i
, the RPV water level from the Bottom of the pressure,the RPV shutdown range watar Active Fuel (BAF) to top of the steam dome. level, the RPV narrow range water level, the RPV wide rae water level, and the l RPV fuel zone range water level sensors 3 has been property installed.
1 B { i- 15. For the safety related NBS instrumentation, 15. Instrument functional testing shall be 15. The instrumentation functions as required. [ the instrumentation be capable of performed to demonstrate that the perforr ' Mg its necessary function. instrumentation performs as required. I ! t
#1 s .
ej Table 2.1.2: Nuclear Boiler System (Continued) is Inspections, Tests, Analyses and Acceptance Criteria inspections, Test, Analysis Acceptance Criteria Certified Design Commitment
- 16. Inspection shall be performed which 16. Inspection confirms that the in ' Mant
- 16. Control room indication / alarms are plant parameters have been indicated and/
provided for the important plant confirms that the important plant parameters monitored by the NBS are or alarmed in the main control room. parameters monitored by the NBS. indicated and/or alarmed in the main control room. e 6 ~ O O O
ABWR oesign Document 1 l MAIN STEAM LINES h h h k l l ASME l l OUTBOARD 1 i CODE l l MAIN STEAM ISOLATION l I CLASS 2I l l I l ; ; VALVE A Y CONTAINMENT i ASME WALL l l woeCLASS l 1l g
/
l l l l i i l l INBOARD DRYWELL MAIN STEAM i ISOLATION N VALVE
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SR SRV SRV ' 'h SRV
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SRV SRV SRV SRV FAV [,' _ MSL _ MSL _ ~ REACTOR VESSEL M S L ,_ __ _ - ~ ~ ,M S L SRV _- ~~__ o Figure 2.1.2a Safety / Relief Valves and Steamline 2.1.2 -13 6/1/92
~ N 1 3 SRV INBOARD MAIN , (TYPICAL) [ STEAM ISOLATION
- l VALVE
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SEISMIC ,!; NON-SEISMiv i CATEGORY l CATEGORY I f 1 2 i I -- F TO MAIN ' g __ ___ __ _ - p- -+ STEAM MAIN STEAM TURBtNE g REACTOR UNE A FROM VESSEL i FROM t STEAMLINES OUTBOARD MAIN STEAM ISOLATION
! STEAMUNES l B.C &D f ACE VALVE { gl l l l l l
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CODE =
= NONCODE l CLASS 1 4 1 :
i l WETWEL'.- i [ i i i lA ou m { SUPPRESSION POO' ' l 1 l l );f i _. i Figure 2.1.2b Stesmline O O O
ABWR vesign Document I FEEDWATER LINES V y
/ / NN j Q SEISMIC INTERFACE E S l l
_1 J T FROM RCIC, I i FROM RHR, C UW, AN D - - - - - M, ASME h- -- -- CUW, AND CRD CODE l CRD l CLASS 2 I I I A J - - - - J V CONTAINMENT WALL
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Figure 2.1.2c Feedwater Line 2.1.2 - 1 Er G/1/92
ABWR oesign occunwnt l
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l HPIN NBS NBSj HPIN
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l l l l l l l l l l 1 FROM FROM FROM POWER ADS ADS ACTUATED DIV. I DIV.11 REllEF LOGIC LOGIC LOGIC l JSRV ADS SRV O HPIN NBS g ,, i: { RELIEF T J RV s ACCUMULATOR f A I I l FROM - POWER
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i ACTUATED TSRV RELIEF 4 LOGIC NON ADS SRV O Figure 2.1.2d Safety / Relief Valve Pneumatic Lines 2.1.2 6/1/92
l ABWH Design Document O I i
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D $ DRYER SKIRTg _ g ,, ; N 6 8,'; N -
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# M'MVMiRR; !NSTRUMENT
! READOUT RANGE l / \
4 y INSTRUMENT TAP RANGE l fs J 1 i Figure 2.1.2e Water Level Range Definition 2.1.2 17 6/1/92
ABWR Design Document 9 100 -- - 1 I e i O g SAFETY VALVE
$ OPEfJlfJG
.J l CHARACTERISTICS 8 l 50 st l E I i
l 9
> l I
6 I VALVE !
" 4, STROKE k-! -
TIME l 0 4 0.3 b-1 TIME (sec) 11 -TIME AT WHICH PRESSURE EXCEEDS THE VALVE SET PRESSURE Figure 2.1.2f Safety action Valve Lift Characteristics 2.1.2 0/1/92
1 ABWR oesign Document 2.1.3 Reactor Recirculation System Design Description The Reacto Recirculation Spiem (RRS) includes an airangement of lo teac .r internal (scal-less) purups (RIP) with wet motors mounted in the bottoin of tlu-RIT as shown in Figure 2.1.3. The RlPs circulate coolant tinough the ie.u tor core at v.uiabic flow rates, which varv teactor power approximatelv 70-1000. One coolant flow inte is cosmolled hv the Rnisculation Flow Contiol (RFC) System. The RFC System includes the Adjustable Speed Drive (ASD) RIP trip (RPT) function and core flow measuternent. Tier 1 information for the RFCS is in Section 2.2.8. In addition to providing core coolant flow during normal reactor operation, the RIPS and associated equipment are designed to (1) have flow coastdown characteristics that provide an adequate fuel thermal margin during plant transients, and (2) maintain reactor coolant picssure boundag (RCPil) integrity during adverse combinations of loading during abnormal, accidem, and special event conditions. The only safetyqelated portion of the RRS is the bottom motor cover bolted to the RIP motor housing. The RIP motor housing is part of the RPV described in Section 2.1.1. The motor cover is part of the RCPit and is designed to Seismic Category I and Quality Group A (similar to the RPV). The design, materials, manufacturing. fabrication, testing, examination, and inspection used in the construction of the cover and cover bolts and nuts meet requirements of AS$1E Code Class 1 vessels. The motor cover and cover bolt materials are low or high _ alloy steels. Hydrostatic test of the covers and bolts after fabrication and in the plant is performed in accordance with the requirements for AS$1E Code Class 1 vessels. The RIP design panuncters are:
- RIP hiotor Cover Design Pressure (kg/cm2 g) 87.9
- RIP hiotor Cover Design Tem erature (*C) 302 8
- Individual RIP rated Flow (m /hr) 2 6912
- Rated Total Developed Head (TDH) for RIP 'm) 2 32.6 The RIP and core flows are measured in various plant operating nuxles with the RFCS (Section 2.2.8).
The Recirculation Slotor Cooling (Rh!C) Subsystem (Figure 2.1.3) provides forced circulation with an auxiliary cornbination thrust bearing-irnpeller mounted on the bottom of the motor rotor,inside the inotor housing. The 2.1.3 -1 6/1/92
.___ _-____ --. - = _- - - . _ __ - -. . -. . . - ._ ._ _. ..
1 ABWR oesign occument innpeller foires cooling wates tloongh the anotor i.ulial be.nings aint wiiulings
.uul to the inotoi ( oohng heat exchanger. The RSIC heat eu h.uigeis .u c lot ated /'
inulci the RP\' close to the RIP inotors. The RNIC is classified as Quality Gioup
- 11. For plant availability. the RNIC is designed to the s.une p.u.unctris as the RPV.
The RIPS icceive power tinough theit inulividual ASI)s horn the plant non-essential power wstein. As shown in Figure " 1.3. the RIPS aie inounted in the KPV bottoin head. The inotor cooling heat cu bangers are located inside the RPY pedestal .nljar ent to the RIP inotors. Each RIP instnnnentation includes speed, vibration, and RN1C temperature transinitters is iiulicated in the Stain Control Rooin. In-Scirice Inspection (ISI) of the inotor cover aiul holt.s can be perfonined during plant shutdown when the inotors are removed fo: t outine inaintenance. Inspections, Tests, Analyses and Acceptance Criteria Table '2.1.3 provides a definition of the instructions. tests, aiul/or analyses together with associated acceptance criteria which will be un<leitalen for the RRS. l l O l 2.1.3 2 6/1/92
o b i I l [ Table 2.1.3: Reactor Recirculation System w f I Inspections, Tests, Analyses and Acceptance Criteria [ t Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria [ i l 1. System configuration of the Reactor 1. Visual field inspections will be conducted 1. Tho installed configuration of the RRS will ; j Recirculation System (RRS) as described in of the installed RRS key components be considered acceptable if it complies [ ] Section 2.13 is shown on Figure 2.13. identified in Section 2.13 and Figure 2.13. with Figu<e 2.13 and Section 2.13. ! i
- 2. RIPS toertia provide adequate reactor fuel 2. Factory measurements will determine the 2. These mertta measurements confirms RIP
! thermal margin during plant transients. mechanicalinertia of the RIP / motor inertia in these factory conditions. rotating assembly. , ] 3. The reactor coolant pressure boundary 3. Inspections wi!! be conducted of ASME 3. Existence of necessary ASME Code , ) (RCPB) motor covers are classified as Code required documents and the Code required documents and the Code stamps ! ) Quality Group A, Seismic Category 1. The stamp on the cover ai J bolts. on the components confirm that the RCPB ! I covers and bolts are designed, fabricated, cover and bolts are designed, fabricated. i l examined, and hydrotested in accordance and examined as ASME Code Class 1. i 1 with the rules of ASME Cods Class 1
- @ vessels and are code stamped accordingly.
]
- 4. The RCPB cover and bolts retain their 4. A hydrostatic test of the RCPB, including 4. The hydrostatic test results must conform integrity under internal pressure that will the covers and their bo!:s, will be {
with the ASME requirements. j i be experienced during the service. conducted in accordance with the ASME [ i Code requirements. ' 3 5. The materials used for the RCPB motor 5. Inspection will be conducted of the RCPB 5. Records of the materials and processes covers and bolts are proven low and high motor covers and bolts records of must confirm that the requirements f i alloy steels with'certain additional materials, fabrication, and examination specified for the RCPB covers and bolts are f requirements for constr uction, as identified used in construction of the covers and satisfied. [ in Section 2.1.1. bolts. [ t
- 6. The RCPB covers and bolts ferritic 6. Fracture toughness tests of the ferritic 6. Records of the fracture toughness data of meterials are not susceptible to brittle materials will be conducted in accordance the RCPB ferritic materials must confirm ,
fracture under pressure during service. with the requirements for ASME Class 1 that the ASME Code requirements are met. : components l
- 7. The Recirculation Motor Cooling (RMC) 7. Factory or preoperational tests will be 7. Detectors in the RMC Subsystem confirm [
4 e forced circulation transfers the heat from performed to determine that the RMC will that the temperatures of the RMC water i each RIP motor to its heat exchanger. adequately remove the motor heat within and motor temperatures are acceptable. f' the RMC design limits. r i l I i I i !
; r l
ABWR 0: sign oocwnent i.2 Control and Instrument 2.2.1 Rod Control and Information System
- i. Sign Description The non-saf ety design bases of the axl wntr ol .uul infor mation system (RCIM is to reliably provide:
(1) The capability to control icactos powei level by controlling the movement of control rods in reactor cou in manual, semiautomated, and automated modes of plant operations. (2) Controls for sorne RCIS bypass and suncillance test functions, and summary information of control iods position aint status on the RCIS Dedicated Operator Interface (1x )l). (3) Transmission of Fine blotion Control Rod Drives (FSICRD) status and control rods position and statu% data to other plant systems (e.g., the plant process computer system). (4) Automatic control rod nm-in function of all operable control rods A following a scmm (scram follow function). b (5) Automatic enforcement of rod movement blocks to prevent potentially undesimble rod movements (these blocks do not have an eilect on scram insertion function). (6) Control capability for insertion of all control rods by an alteniate and diverse method [ Alternate Rod Insertion (ARI) function). (7) The capability to enforce a preestablished sequence for control nxi movement when reactor power is below the low power setpoint. (8) The capability oa enforce fuel opemting thennal limits when reactor power is above the low power setpoint. (9) The capability to send flow nmback signals to Recirculation Flow Control System on detecting an all-rods-in condition. (10) Tbc capability to provide for Selected Control Rod Run In (SCRRI) function for core thennal-hydmulic stability control. The RCIS is classified as a Non-Nuclear-Safety (NNS) system,it has a control design basis only, and is not required for the safe and orderly shutdown of the plant. A failure of the RCIS will not result in gross fuel damage, llowever, the rod block function of RCIS is important in limiting the consequences of a rod 2.2 6/1/92
ABWR Design occument l withdrawal crior, and pievention of local Inel operating thennal ihnits siolanons g I dtning nonnal plant operations. Thricinic RCIS is designed to ineet single- T f ailure ciiteria and to be highly icliaine. The RCIS consists of seteml difIcient types of cabinets (or panels), which contain special elecuonic/ electrical equipment modules, and a dedicated operator interface on the main conuol panel in the control suom. The RCIS block diagnun is shown in Figtne 2.2.1 which depicts the major iomponents of the RCIS, their interconnections, and inteilat es with other AllWR svstems. The RCIS is a dual redundant sptem that consists of two indepciulent channels foi noimal control rod position monitoring and connol iod movements. Tbc two channels receive the same but separate input signals and per form the s.une exact functions. For normal functions of RCIS, the two c hannels must always he in agreement and any disagreement between the two channels results in iod block. However, the protective function logic of RCIS (i.e., iod block) is designed such that the detection of a rod block condition in only one channel of RCS would result in a rod block. There are four types of electronic /clectrical cabinets that make up the RCIS. They are: (1) Rod action control cabinets (RACC) (2) Remote conununication cabinets (RCCs) (3) Fine motion driver cabinets (FAIDCs) (4) Rod brake controller cabinets (RitCCs) in addition, RCIS includes a fiber-optic dual channel hiultiplexing Networ k that is used for tmnsmission of rod position and status data from RCCs to the RACCs, and rod block / movement command from the RACCs to RCCs. A sununan-description of each of above is provided below. (1) Rod Action Control Cabinets (RACC) There are two RACCs in the control room RACC Channel A and RACC Channel 11 that provide for a dual redundant architecture. Each RACC consist.s of three main functional subsystems, as follows: (1) Automated Thermal 1.imit hionitor (ATI.51) (2) Rod Worth hiinimizer (RW.\1) (3) Rod Action and Position lufor mation (RAPI) 2.2.1 G/1/92 i.
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ABWR vesign Document ( '.' ) Reinote umanunication Cabinets (RCC) The icinote < ononunication cabinets (RCGs) contain a dual channel file contial inodule (I CM) and several dual channel nod sence inodules (RSMst The FChi interf aces wit.. the RSMs aiul RAPI. (3) hne Motion 1)iiver Cabinets (FMI)C) The fine anotion diives cabinets (FMI)Cs) consist of sevcial stepping inotor diiven inodules. Each steppine snotor driver inodule t ontaitn an electionic convester/ invert ( r that (or.verts the incorning 3-phase AC power into DC and then inverts the DC power to v.uiable vohage/ fic(juency AC powet that is supplied to FMCRD stepping rnotois. For each convester/intertt r, these exists an Invester Controlles (IC) that contiols the duration of power supplied to tlus stepping inotors uiuler the coininand of RSMs. (4) Rod israke Controller Cabinets (RiiCC) The rod brake controller cabinets (RitCCs) contain electiical power
.upplies, electronic (or relay) logic, and other associated electrical equipinent for the proper operation of the FMCRD bnites. Signals foi brake disengagement / engagement are received from the associated rod server nuxlules. The hnike controller logic provides two separate (channel A and channel 11) bnike status signals to the associated iod server inodule.
(5) RCIS Multiplexing Network The RCIS multiplexing network consists of two independent channels (channel A and channel it) of fiber-optic conununication links between the RACCs (channel A and channel 11), and the dua' channel file control nuxlules located in the remote communicatirn cabinets. l The plant essential multiplexing network interfisces with FMCRD dual l redundant separation switches (A/ll) and provides the appropriate status signals to the RACC that is used in the RCIS logic for initiating rod block signals if a sepamtion occurs. The essential multiplexing network is not part of the RCIS scope. (6) RC'S Power Sources RCIS equipment derive theii awer from two different sour ces. Fine Motion Driver Cabinets and Rod lirake Controller Cabinets det ive their power from the plant divisional Class lE power smuces that are backed l l 2.2.1 3- cv1/92
ASWR oesign Document up bv plant diesel generatois. All other l'.CIS c<[uijunent deriw their power from the plot Ln-Class lE uninter ruptible AC pimei systern Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.1 provides a defh. ;.on of the inspections, tests, and/(n anah ses; together with associated acceptance criteria, which will be used by RCIS. O O 2.2.1 6/1!92
m s G kpJ. { Table 2.2.1: Rod Control and Information System ~ Inspections, Tests, An# /ses and Acceptance Criteria ' l Certified Design Commitment inspection, Tests, Analyses Acceptance Criteria
- 1. Proper separation is maintained between 1. Visual field inspection of installed 1. Visual field inspections in conjunction with the non-safety RCIS and sa'ety systems equipment and review of as-built RCIS and review of drawings confirm that proper that interface with RCIS. interfacing systems drawings to ensure separation is n aintained.
that implemented isciation methods meet the design requirements.
- 2. The RCIS is designed to meet single failure 2. Preoperational tests will be conducted to 2. Observation of RCIS continued operation criteria. The two channels of RCIS are confirm channel redundancy, channel when one channel is disabled, that one independent of each other,in the cense protective function independence, and channel can cause a rod block, and that it that each channel can independently cause channel agreement for normal RCIS takes the agreement of the two channels to a rod block; and that for normal RCIS operations. cause normal movement of control rods.
functions of control rods movements and control rods position monitoring, the two channels must be in agreement.
- 3. RCIS design is capable of continued 3. Preoperational tests will be conducted to 3. Observation of RCIS_ continued operation operatien when different subsystems of confirm RCIS bypass capabilities and to when different subsystems are bypassed RCIS are bypassed. RCIS bypass interlock confi+m proper functioning of the bypass and RCIS hypass interlock logic pre.enting logic precludes a bypass state that would interlock logic. a bypass state that would violate bypass render RCIS inoperational. rules as specified in RCIS design documentation.
- 4. When reactor power level is below low 4. Preoperational tests of RCIS willinclude a 4. Observation of rod block signals by RWM, power setpoint, the Rod Worih Minimirer sufficient number of attempts to withdraw / when an out sequence rod withdraw / insert (RWM) of RCIS enforces control rod insert control rods that are both in- is attempted, given that reactor power is I withdrawal and insertion sequence to compliance and not-in-compliance with the below low power setpoint.
comply with a preestablished pa,(ern,in known preestablished pattern, at below order to minimize the consequences of a and above the low power setpoint. rod drop accident, by issuing a rod i sn movement block signal whenever an out of j sequence rod movement is attempted. c
i i ( [ Table 2.2.1: ReI- -----------
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Rod Server Module Channel A Logic l Channel B Logic INDIVIDUAL Jk J Ld Jk JL ROD LOGIC V T & CONTROL l Inverter Controller) (TYP. OF 205) i V V T Reed Switch RodR d e Stepping Motor Senser Module Controller Driver Module JL l jt5l' k. 5k. l/ .$. /$. ./.,?.//./. .
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? FIGURE 2.2.1: ABWR ROD CONTROL AND INFORMATION SYSTEM BLOCK DIAGRAM =
0 Figure 2.1.3 Control and Instrument 2.2.1 8- 6/1/92
ABWR Design Document 2.2.2 Control Rod Drive System
- Design Description The Control Rod Drive (CRD) System is composed of tluce major elements: (1) the elecuo-hydraulic fine motion control rod drive (DICRD) mechanisms, (2) the hydraulic conuol unit (HCU) assemblics, and (3) the Connol Rod Drive Hydraulic (CRDH) Subsystem. The FNICRDs proside elecu ic-motor <lriven positioning for normal insertion and withdrawal of the control rods and hydraulic-powered rapid control rod insertion (scram) for abnormal operating conditions. Simultaneous with scram, the FSICRDs also provide electric-motor driven run-in of all control rods as a path to rod insertion that is diverse from the hydraulic-powered scram. The hydraulic power required f or scram is provided by high pressure water stored in the individualIICUs Each HCU is designed to scrmn two FhtCRDs. The HCUs also proside the flow path for purge water to the associated diives during normal operation. The CRDH Subsystem supplies lugh pressure demineralized water which is regulated and distributed to proside charging of the HCU scram accumulators and purge water flow to the FA1CRDs.
During power operation, the CRD System controls changes in core reactivity by movement and positioning of the neutron absorbing control rods within the core in fine increments via the FMCRD electric motors, which are operated in response to control signals from the Rod Control and Information System (RCIS). The CRD System provides rapid control rod insertion (scram) in response to manual or automatic signals from thc Reactor Protection System (RPS), so that no fuel damage results from any plant mmsient. There are 205 FMCRDs mounted in housings welded into the reactor vessel bottom head. A schematic of the drive is shown in Figure 2.2.2.a. Each FMCRD has a movable hollow piston tube that is coupled at its upper end, inside the reactor vessel, to the bottom of a control rod. The piston is designed such that it can be moved up or down, both in fine increments and continuously over its entire range, by a ball nut and ball screw driven at a nominal speed of 30 mm/ sec by the electric stepper motor. In response to a r cram signal, the piston rapidly inserts the controi . rod into the core hydraulically using stored energy in the HCU scram accumulator. The scram water is introduced into the drive through a scram inlet connection on the FMCRD housing, and is then discharged directly into the reactor vessel via clearances between FMCRD parts. The FMCRD scram time requirements with the reactor pressure as measured at the vessel bottom below 1085 psig are: (D (j Percent Insertion Time (sec) 10 s 0.42 40 f 1.00 2.2.2 G/1/92
ABWR oesign Document Percent Insertion Time (see) 60 51 A4 100 $ 2.80 The BlCRD design includes an electro-mechanical bruke on the motor drive shaft and a ball check valve at the point of connection uith the scnun inlet line. These features prevent contial rod ejection in the event of a f ailure of the scram insert line. An internal housing support is prosided to prevent ejection of the BlCRD and ita attached control rod in the esent of a housing failure. It utilizes the outer tube of the drive to provide support. The outer tube, which is welded r to the drive middle Gange, attaches by a bayonet lock to the control rod guide tube (CRGT) base. The CGRT, being supported by the lower core plate, in turn. prevents any downward movement of the drive. The DiCRD is designed to detect sepamtion of the control rod from the drive mechanism. Two redundant and separate Class 1 E switches detect separation of either the control rod from the hollow piston or the hollow piston from the ball nut. Actuation of either switch will cause an inunediate rod block and initiate an alarm in the control room, thereby preventing the occunence of a rod drop accident. There are 103 HCUs, each of which provides sufncient volume of water stored at high pressure in a pre-charged accumulator to scram two BlCRDs at any reactor pressure. Figure 2.2.2.b shows the major HCU components. Each accumulator is connected to its associated DiCRDs by a hydraulic line that includes a normally-closed scram valve. The scram valve opens by spring action but is normally held closed by pressurized control air. To cause scram, the RPS provides a de-energizing reactor trip signal to the solenoid <yerated pilot valve that vents the control air from the scram valve. The system is ' fail safe' in that loss of either electrical power to the solenoid pilot valve or loss of control: pressure causes scram. The HCUs are housed in the secondary containment at the basemat elevation. This is a Seismic Category I stnicture, and the HCUs aie protected from external natural phenomena such as earthquakes, tornados, hurricanes and floods, as well as from internal postulated accident phenomena. In this area, the HCUs are not subject to conditions such as missiles, pipe whip, and discharging Guids. The CRDH Subsystem design provides the pumps, valves, filters instrumentation, and piping to supply the high pressure water for charging the HCUs and purging the DICRDs. Figure 2.2.2.b shows the major system equipment. Two 100% capacity pumps (one on standby) supply the HCUs with water from the condensate treatment system and/or condensate storage tank for charging the accumulators and for supplying BICRD purge water. The CRDH Subsystem equipment is housed in the Seismic Category I reactor building to protect the system from floods, tornadoes, and other natural phenomena. 2.2.2 -2 6/1/92
I ABWR Design Document The CRD System includes control room indication and alanns to allow for b,A monitoring and control during design basis operational conditions, including system flows, temperatures and pressures, as well as v,dve position indication r,nd pump on/off status for those instnunents and components shown in Figure 2.2.2.b, with the exception of simple check valves. Class 1E pressure mstnunentation is provided on the HCU charging water header to monitor header perfonnance. The pressure signals from this instrumentation are provided to the RPS, which willinitiate a scrmn if the header pressure degmdes to a predetermined low pressure setpoint. This feature assuies the capability to scram and safely shut down the reactor before HCU accumulator pressure can degrade to the level where scram performance is adversely alfected following the loss of charging header pressure. The BICRD electric motors are powered from a dedicated non-divisional 480 VAC power center fed by the Division 16.9 kV Class 1 E bus as the first source of standby power and by the nor -divisional combustion turbine generator as the second backup source. The CRD pumps, valves, and controls are powered from two separate trains of 6.9-kV offsite power with automatic transfer to the combustion turbine generator upon loss of preferred power. l ( Components of the system that are required for scram (DiCRDs, HCUs and ! scram piping), are classified Seismic Categog I. The balance of the system equipment (pumps, valves, filters, piping, etc.) is classified as non-Seismic Categog I, with the exception of the Class 1E charging water header pressure instrumentation, which is Seismic Categog I. The major mechanical components are designed to meet ASME Code requirements as shown below: ASME Design Conditions Component Code Class Pressure Temperature BICRD (RCPB parts) 1 87.9 kg/cm2g 302 C Scram Piping 2 190 kg/cm g 2 66 C HCU (scram related parts) 190 kg/cm g 2 66 C ( 2 CRD Pumps non-Code 190 kg/cm g 2 66 C CRDHS Piping, Valves non-Code 190 kg/cm g 2 66 C l l The CRD System is separated both physically and electrically from the Standby Liquid Control (SLC) System. Inspections, Tests, Analyses and Acceptance Criteria l p) C This section provides a definition of the inspections, tests, and/or analyses. together with associated acceptance criteria, which will be undenaken for the CRD System. 2.2.2 6/1/92
Table 2.2.2: Control Rod Drive System } inspections, Tests, Analyses and Acceptance Cri2eria Certified Design Commitment inspections, Tests, Analyses Acceptance C:iteria
- 1. A simplified configuration of the Control 1. Visual field inspections will be conducted 1. The system configuration is in accordance Rod Drive (CRD) System as described in to confirm that the installed CRD System with Figure 2.2.2.b.
Sec' ion 2.2.2 is shown in Figure 2.2.2.b. equipment is in compliance with the design configuration defined in Figure 2.2.2.b.
- 2. The RCPB portions of the FMCRD (midde 2. Inspections will be conducted of ASME 2. The components have appropriate ASME flange, spool piece, :nounting bolts, seal Code required documents and the Code Code, Section til, Class 1 certifications and housing) are classified as ASME Code stamp on the actual components to verify Code stamps.
Class 1. They are designed, fabricated, that they have been manufactured per the examined, and hydrotested per the rules of relevant ASME requirements. ASME Code, Section Ill.
- 3. The scram-related parts of the HCU and the 3. Inspections will be conducted of ASME 3. The components have appropri3te ASME p scram piping are classified as ASME Code Code required documents and the Code Code, Section lit, Class 2 cartifications and Class ,". They are designed. fabricated, stamp on the actual components to verify Code stamps.
examined, and hydrotestad per the rules of that they have been manufactured per the ASME Code, Section 111. relevant ASME requiremern;
- 4. The installed FMCRDs and HCUs shall be 4. Scram tests will be conducted during the 4. The observed / measured scram times are:
capable of providing control rod scram preoperational testing program to confirrn Percent Ir.sertion Time (sec) time performance within specified limits. proper operation of HCUs and assoda*ed 1010.42 vaivaa. includmg scram timing 40 s 1.00 demonstrations with the .eactor at 60$; 1.44 atmospheric pressure. 100 f 2.80
- 5. The nominal FMCRD motor-driven rod 5. Functional tests will be performed for each 5. The coserved/ measured motor-driven motion speed shall be 30 mm/sec. FMCRD during the preoperational testir.g FMCRD speed is 30 mm/see t10%
program to confirm that drive speed complies with the design commitment.
- 6. The FMCRD electromechanical brake and 6. Functional tests of the brake and check 6. The brake holdino torque is within ball check valve shall be capable of valve will be performed for each FMCRD specified limits. The ball check valve
$ performing their rod ejection prevention during the preoperational testing program. actuates to scal the scram inlet port under $ functiors as identified in Section 2.2 2. conditions of reverse flow. 9 O O
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g Table 2.2.2:' Control Rod Drive System (Continued)l 4-Inspections, Tests, Analyses and Acceptance Criteria ; Certified Design Commitment Inspections, Testsf Analyses . Acceptance Criteria .[ , 7. The FMCRD outer tube, which is welded to 7. Visual. inspection of the actualinstalled 7. Inspection confirms that a bayonet lock is j the drive middle flange, shall bayonet lock equipment shall confirm the FMCRD is in provided.
- to the CGRT base to form the intern:.1 compliance with the design commitment. i housing Support for prevention of rod
- ejection in the event of a CRD housing ' I failure.
i
- 8. The FMCRD separation switches shall 8. The separation switch operation shall be - 8. The switches actuate the control room t detect separation of the control rod from., ' tested as part of the drive functionaltesting alarm when exercised. .
[ the drive mechanism and initiate a control conducted during the preoperational [ room alarm.The separation switches are testing program. ' classified Class 1E. ,
- 9. The pressure instrumentation on the HCU 9. Logic snd instrument functional testing 9. The pressure instrumentation functions as charging water header for monitoring HCU shall be performed to demonstrate that low required to generate low pressure scram 'i accumulator charging pressure shall signal charging header pressure will generate a signals to the RPS.
the RPS to initiate a scram if charging scram by the RPS. 4 pressure is low. i i , 10. CRD System equipment can be powered 10. System tests will be conducted after 10. The installed equipment can be powered from the standby AC power supplies as ' installation to confirm that the electrical ' from standby AC power supplies. c described in Section 2.2.2. power supply configurations are in l conipliance with c'esign commitments. I m I b 1
._. ___ _______o
ABWR oesign Document O B AYONET COUPLING-
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- GENERATOR Figure 2.2.2a Fino Motion Control Rod Drive Schematic 2.2.2 6/1/92
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Figure 2.2.2b Control Hod Drive System L _ _ - .
ABWR Design Document n 2.2.3 Feedwater Control System U Design Description The Feedwater Control (FDWC) System controls the flow of teedwater into the reactor pressme vessel (RPV) to maintain the water levelin the vessel within predetennined limits during all plant openning modes. The FDWC System may operate in either single-or three-element control modes. At low reactor powers (when steam now is either negligible or else measutement is below scale), the FDWC System utilizes only water level measurement in single-element control mode. When steam flow is negligible, the Reactor Water Cleanup (CUW) System dump valve Dow can be controlled by the FDWC System in single-element nu>de in order to counter the efrects of density changes during heatup and purge flows into the reactor. At bigher powers, the FDWC System in three-element control mode uses water level, main steamline flow, main feedwater line flow, and feedpump suction How measurements for water level control. The FDWC System control structure is shown in Figure 2.2.3. The FDWC System is a power genention (control) system with operation range between high water level and low water level trip setpoints. It is classified as nonsafety-related. This system is not required for safety purposes, nor is it required to operate after the design basis accident. This system is also required O to operate in the normal plant environment for power generation purposes only. Reactor vessel narrow range water level is measured by three identical, independent sensing systems. For each level measurement channel, a differential pressure transmitter senses the difIerence between the pressure caused by a constant reference column of water and the pressure caused by the variable height of water in the reactor vessel. The FDWC fault tolerant digital controllers (FTDCs) will detennine one validated narrow range level signal using the three level measurements a inputs to a signal validation algorithm. The validated narrow range water level is indicated on the main control panel and is continuously recorded in the main control room. The steam flow in each of four main steamlines is sensed at the RPV nozzle venturi's. The Multiplexing (MUX) System signal conditioning algorithms process the venturi differential pressures and provide steam How rate signals to the TTDCs for validation. These naidated measurements are summed in the FTDCs to give the total steam How rate out of the vessel. The total steam Dow rate is Ladicated on the main control panel and recorded in the main control room. Feedwater flow is sensed at a single Dow element in each of the two feedwater g lines. The MUX System signal conditioning algorithms process the flow element s differential pressure and provide feedwater now rate signals to the FTDCs. These validated measurements are summed in the FTDCs to give the total feedwater 2.2.3 -1 6/1/92
l i ABWR Design Document licne rate into the vessel. The total feedwater flow rate is indicated on the main g ! control panel an; recorded in the main control room. T I i Feedpump suction flow is sensed at a single flow element upstream ot each ; feedpump. The MUX System signal conditioning algorithms process the flow element differential pressure and provide the suction flow rate meastu ements to the FTDCs. The fet dpump sut tion Gow rate is compared to the demand flow for that pump, and the resulting error is used to adjust the actualm in the disc < tion necessaiy to reduce that error. Feedpump speed change and low flow control valve position control are the flow adjustment techniques involved. Three modes of feechvater flow control (and, thus, level control) are provided: (1) single-element control; (2) three-element control; and (3) manual connol. Each FTDC will e.xecute the control . software for all three of the control modes. Actuator demands from the redundant FTDCs will be sent over the MUX System to field voters which will determine a single demand to be sent to each actuator. Each feedpump speed or control valve position demand may be controlled either automatically by the control algorithms in the FTDCs or manually from the main control panel through the FTDCs. Three-element automatic control is provided for normal operation. Three-element control utilizes wiuer level, feedwater flow, steam flow, and feedpump g flow signals to determine the feedpump demands. The total feedwater now is W subtracted from the total steam flow signal, yielding the vessel flow mismatch. The flow mismatch, sununed with the conditioned level error from the master level controller, provides the demand for the master flow controller. The master flow controller output provides the demand for the feedpump flow loops which send a pump speed demand signal to the adjustable speed drives (ASD) for the feedpump. In the single-element control mode, which is employed at lower fecchvater flow rates, only conditioned level error is used to determine the feedpump demand. The master level controller conditions the level error and sends it directiv to the , feedpump ASDs, and/or low flow control valve actuator, When the reactor water inventory must be decreased (e.g., during very low steam flow rate conditions), the CUW System dump valve is controlled by the FDWC system in single-element control. Reactor water is dumped through the CUW System to the condenser. Each feedpump flow control actuator can he controlled ' manually' f rom the main control panel by selecting the manual mode for that feedpump. In manual mode, the operator may increase or decrease the demand that is sent directiv to the ASD of the chosen feedpump. The FDWC System also provides interlocks and control functions to other systems. When the reactor water level reaches the high level trip setpoint, the 2.2.3 6/1/92
l l ABWR Design Document c FDWC system simultaneously annunciates a control room alarm, sends a trip signal to the turbine control system to trip the turbine generator, sends trip signals to all feedpumps, and closes the main feedwater discharge valves. This interlock is enacted to protect the turbine from damage from high moisture content in the steam caused by excessive can7 aver, while preventing wmter level from rising any higher. The FDWC System will send a signal to the main steamline condensate drain valves to open when steam flow rate is below a pre <letennined setpoint. This also protects the turbine from damage caused by excessive moisture in the steamline. The FDWC System will send a trip signal to the Recirculation Flow Control (RFC) System when reactor water level reaches this low level serpoint, The RFC System will runback the reactor internal pumps (RIPS) if this low level signal coincides with a feedpump trip signal provided to the RFC System by the Feechvater and Condensate System. The RIP runback will aid in avoiding a low water level scram by reducing the reactor steaming rate. Feedwater flow is delivered to the reactor vessel through a combination of three adjustable speed turbine-driven feedpumps and a low flow control valve. Each adjustable speed drive can also be controlled by its manual / automatic transfer p station, which is part of the Feedwater and Condensate System. A Low Flow d Control Valve (LFCV) is also provided in parallel to a common discharge line from the feedpumps. The LFCV can also be controlled by the manual / automatic transfer station which is part of the Feedwater and Condensate System. The FDWC System is not required for safety purposes, nor is it required to operate after the design basis accident. This system is required to operate in the normal plant emironment for power generation purposes only. The FDWC System is powered by redundant uninterruptable power supplies (UPS). No single power failure will result in the loss of any FDWC System functions. Controllers to be used for the FDWC System shall be triplicated, fault tolerant digital type with self-test and diagnostic capabilities. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.3 provides definition of the inspection, tests, and/or analysis, together with associated acceptance criteria which will be undertaken for the Feedwater Control System. O 2.2.3 6/1/92
Table 2.2.3: Feedwater Control System
}
inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Ana!yses Acceptance Criteria
- 1. This system automatically maintains 1. Perform tests to confirm that the system 1. The FDWC System must maintain water reactor water level within operational can maintain reactor water levelin all level between high and low trip setpoints limits, by regulating feedwater flow and control modes. (see Figure 2.2.3).
reactor water cleanup system dump flow. Operate system in each mode of controls: This system shall use single-element control (level only), or three-element - Single Element control (level, steam flow and feedwater - Three Element flow), or manual control. - Manual
- 2. This system must be powered by 2. I oss of power tests shall demonstr ate no 2. There is no loss of FDWC System function redundant uninterruptable power supplies. loss in R //C System function. by loss of any power supply.
- 3. The water level shall be measured by three 3. Inspection and testing will show the three 3. The FDWC System s.onforms to Figure L. identical, independent sensing systems. identical and independent sensing 2.2.3, and the input signal is indepertdent systems. of the output signal response.
- 4. Triplicated. fault tolerant digital controllers 4. Inspect FTDCs and perform validation 4. The FTDC, self-test and on-line diagnostics (FTDCs) with self test and diagnostic testing. test features are capable of identifying and capabilities shall be used. isolating failures of process sensors,I/O cards, buses, power supplies, pr ocessors and inter-processors communication paths down to the machine !evel.
- 5. The RP System shall monitor reactor weer 5. Perform test to confirm that the high water 5. High water level trip signal is issued to the level and in the event that high water level level trip signal is properly issued. Turbine Control System and Feedwater and is reached, shall issue trip signals to the Condensate System when reactor water Turbine Control System to trip turbine reaches high level.
generator, Feedwater and Condensate Systems to trip feedpumps and close discharge valves.
- 6. This system shall monitor reactor water 6. Perform test to confirm that the low water 6. Low level trip signal is issued to the
'S level and, in the event that low water level level trip signal is properly issued. Recirculation Flow Control System when M is reached, 'll issue trip signal to roactor water reaches low level. Recirculation Flow Control System (RFC System logic determines need for RIP runback). O O O
(m e~ , ,e
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u_ N u O HIGH l 1 LEVEL - O 40% REP A , FLOW SUCTION LOW TRIP FLOW C A TOTAL y STEAM TRIM 3E FLOW C n COf4 TROLLER P M'A 4 RFP A DEMAt4D WATER LEVEL C g 1E l REPO SUCTIOt1 f FLOW C *E V 71M T fNA 4 RFB B DEMAND MASTER CONTROLLER 4
'_ EV EL MASTER g y -.o
! SETPOINT C - LEVEL n m OW W
" v v CONTROLLER CONTROLLER 4k REPC u 3E
" TRIM P M/A l SUCTION O -O RFP C DEMAND CONTROLLER FLOW % 4 1E 10TAL m FEEDWAT ER FLOW C - m m LFCV c M/A --O LFCV DEMAND l
LFCV 4 I CONTROLLER dp C
- cuW CUW > DUMP FNA I CUW D'JMP qOW VALVE DE*AAND i
DUMP FLOW C COf0 ROL!.ER
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I Figure 2.2.3a FWC Control Alporithm
ABWR Design Document O M
>d FEEDPUMP FLOW TRANSMITTER v
m O
+
FEEDWATER PUMPA l ASD l DEMAND O- ---d FEEDPUMP FLOW TRAN0MITTER FROM LOW PRESSURE
] M STEAM FLOW TRANSMITTER FEEDWATER 4 HEATERS C O FEEDWATER k>
l ASD l FEEDWATER FLOW em PUM P B DEMAND O- --- d TRANSMITTER LEVEL TRANSMITTER M E] LOW FLOWO- ------ CONTROL b ~
]
VALVE l -{> < ( -- DEMAND l FEEDWATER FLOW
} TRANSMITTER LOW FLOW CONTROL M VALVE bs LOW FLOW CONTROL VALVE CUW DUMP dp XMTR FLOW VALVE CUW DEMAND O- -------- ----
1i FEEDPUMP FLOW TRANSMITTER L--- A 5 " CUW DUMP FEEDWATER PUMPC O ] )) FLOW TRANSMITTER DEMAND @ l l ASD l L _ _ _ _J Figure 2.2.3b FWC Piping and instrumentation 2.2.3 6/1/92
ABWR oesign Document _ f- 2.2.4 Standby Liquid Control System t The Statulbv 1.iquid Control (SLC) Systein is designed to inject neuuon absorbing poison using a boron solution into the icactor and thus pt ovide b.vt. up reaClot shtlidt)Wn capabihty illdependent of the !)olmal teJctivity ront!(d system based on inser tion of control iods mto the core. The St.C Sssicm is capable of operation over a wide range of icactor pressuie conditkus up to and including the clewed pressures associated with an anticipated plant transient coupled w',th a f aihire to se: ram (ATWS). The SLC System is designed to bring the t eactot, at any time in a cycle. and ai all conditions, hom full powes to n subcritical condition, with the scactor in the most reacthe xenonJree state, without control rod movement. The system will inject the minimum requhed ooron solution in 61 minutes. The SLC System (Figure " 1) consists of a boron solution stomge tank, two posithe displacement pumps, two motor-operated injection valves which are provided in pandlel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The horated colution is discharged through the 'If high pressure co e flooder (HPCF) subsystem sparger. Key equipment pt rformance acquirements are: (1) Pump flow (minunum) 100 gpm with lxnh pumps running l (2) hiaxirnum reactor pressure 1250 psig (for injection) (3) Pumpable volume in storage 6100 U.S. gal tank (minimum) l The required volume of solution contained i', the storage tank is dependent upon the solution concentmtion, and this concentration can vary during reactor l operations. A required baron solution volume /concentmtion relationship is used to define acceptable SLC System storage tank conditions during plant l operation. The SLC System is automatically initiated during an ATWS. An ATWS condition ! exists when either of the following occurs: l l (1) High RPV pressure (1125 psig) and Average Power Range Nlonitor ! (APRN1) not down scale for 3 minutes, or (2) Low RPV level (i.evel 2) and APR.\1 not down scale for 3 m:nutes. s 2.2.4 C/1/92 l
ABWR Design Document When the St.C System is automatically initiated to inject a liquid neutron g absorber into the reactor, the follmei.g devices aie actuated: W (1) The two injection valves are opened (2) The two storage tank discharge vahes aie opened. (3) The two injection pumps are start d. (4) The reactor water cleanup isolation valves are closed. The SLC System can also be manually initiated from the main conuol room. When it is manually initiated to inject a liquid neutton absorber into the reactor, _ the following devices are actuated by each switch-
- l) One of the two injection v;dves is opened.
(2) One of the two storage tank discharge valves is opened. (3) One of the two injection pumps is started. (4) One of the reactor water cleanup isolation valves is closed. The SLC System provides bomted water to the reactor core to compensate for the various reactivity effects during the required conditions. These effects h' include xenon decay, elimination of steam voids, changing water density due to the reduction in water temperature, Doppler effect in uranium, changes in neutron leakage, and changes in control rod worth as boron affects neutron migration length. To meet this objective, it is necessary to inject a quantity of _ boron which produces a minimum concentration of 850 ppm of natural boron in the reactor core at 70 F. To allow for potential leakage and imperfect mixing in the reactor system, an additional 25% (220) is added to the above requirement. The required concentration is thus achieve! = unur.6for dilution in the RPV with normal water leven and including the volume in the RHR shutdown cooling piping. This quantity of baron solution is the amount which is above the pump suction shutofflevel in the tank. thus allowing for the portion of the tank volume which cannot be injected. The pumps are capable of producing discharge pressure to inject the solution into the reactor when the reactor is at high pressure conditions corresponding to the system relief valve actuation (1560 psig), which is above peak ATWS pressure. The SLC System inclr. .es sufncient control room indication to allow for the g necessary monitoring and control during design basis operational conditions W This includes pump discharge pressure. storage tank liquid level and temperature, as well as valve open/close and pump on/off indication for those 2.2.4 6/1/92 l l l
ABWR Design Document p components shinen on Figure 224 (with the exception ot the simp:e check () valves). The SLC System usesa divolved solution of codium pentaborate as the neutron-absoibing poison. Tlus solution is held in a storage tank which has a heater to maintain solution temperature above the saturation temperatute. The heater is capable of automatic operation and automatic shutof f to inaintain an acceptable solution temperature. The SLC System solution tank, a test water tank, the two positive displacement pumps, and associated valving are all located in the secondag containment on the floor elevation below the operating floor. This is a Scisinic Category I structure, and the SLC System equipment is protected from phenomena such as earthquakes, tornados, hurricanes. and floods, as well as from internal postulated a cident phenomena. In this ;uca, the St C System is not subject to conditions such as missiles, pipe whip, and discharging fluids. The pumps, heater, valves, and controls are powered from the standby power supply or nonnal ofTsite power. The pumps and valves are powered and controlled from separate buses and circuits so that single active failure will not prevent system operation. The power supplied to one motor-operated injection valve, storage tank discharge valve, and injection pump is powered from Division I,48 VAC. The power supply to the other motormperated in,jection valve, storage tank outlet valve, and injection pump is poweied from Division II,
\
480 VAC. The power supply to the tank heaters and heater controls is connectable to a standby power source. The standby power source is Class 1E from an on-site source and is independent of the off-site power. Components of die S1 C System which are required for injection of the neutron absorber into the reactor are classified Seismic Catego y I. The major mechanical components are designed to meet ASME Code requirements as shown below: ASME . Design Conditions Cornponent Code Class Pressure Tetuperature Stomge Tank 2 Static Head 150 F Pump 2 1560 psig 150 F Injecdon Valves ~1 1560 psig 150 F Piping Inboard of 1 1250 psig 575'F Injection Valves 1 O , 2.2.4 6/1/92 i
ABWR oesign Document Piping and coinponents not acquired f or the injection of the neutron absorbei g (e.g.. test tank, sampling systern line. and storage tank vent) are classified Non- W Nuclear Safety (NNS). 1)csign provisions to pennit ssstem testing include a test tank aiul associated piping and ndves. The tank can he supplied with deminerali: d water which can be pumped in a closed loop through either pump or injected into the scactor, The SI.C System is separated both physically and electrically f rom the Control Rod Drive System. Inspections, t ests, Analyses and Acceptance Criteria 3 Table 2.2.4 provides a definition of the inspections, tests, and/'or analyses, togeth n with associated acceptance criteria, which will be undertaken for the SLC System. O M O 2.2.4 4- 6/1/92
{ Table 2.2.4: Standby Liquid Control System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The minimum average poison. 1. Construction records, revisions and plant 1. It must be shown the SLC System can concentration in the reactor after operation visual examinations will be undertaken to achieve a poison concentration of 850 ppm of the SLC System shall be equalto or assess as-built parameters listed below for or greater, assuming a 25% dilution due to greater than 850 ppm. compatibility with SLC System design non-uniform mixing in the reactor and calculations. If necessary, an as-built SLC accounting for dilution in the RHR System analysis will be conducted to shutdown cooling systems. This demonstrate that the acceptance criteria concentration must be achieved under are met. system design basis conditions.
Critical Parameters: This requires that the SLC System meet the following values:
- a. Storage tank pumpable volume
- a. Storage tank puropable volume range
- b. RPV water inventory at 7(TF 6100-6800 gal. ,
- c. RHR shutdown cooling system water b. RPV water inventory 5 1.00 x 106 tb inventory at 70*F
- c. RHR shutdown coolin system inventory s 0.287 x 10 lb
- 2. A simplified system configuration is shown 2. Inspections of installation records, together 2. The system configuration is in accordance in Figure 2.2.4. with plant walkdowns, will be conducted to with Figure 2.2.4.
confirm that the installed equipment is in compliance with the design configuration defined in Figure 2.2.4. R e F)
- g. Table 2.2.4: Standby Liquid Control System (Continued) a inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 3. The SLC System shall be capable of 3. System preoperation tests will be 3. It must be shown that the SLC System can delivering 100 gpm of solution with both conducted to demonstrate acceptable automatically inject 100 gpm (both pumps pumps opa 'ating against the elevated pump and system performance. These running) against a reactor pressure of 1250 pressure unditions which can exist in the tests will involve establishing test psig with simulated ATWS conditions. It reacter during events involving SLC conditions that simulate conditions which must also be shown that the SLC System System initiation. will exist during an SLC System design pumps can pump the entire storage tank
' basis event. To demonstrate adequate Net pumpable volume.
Positive Suction Head (NPSH), delivery of rated flow will be confirmed by tests conducted at conditions of low level and maximum temperature in the storage tank, and the water will be injected from the storage tank to the RPV. 6> 4 The system is designed to permit in-service 4 Field tests will be conducted after system 4. Using normally installed controls, power funct;onal testing of the SLC System. installation to confirm that in-service supplies and other auxiliaries, the system system testing can be performed. has the capability to perform:
- a. Pump tests in a closed loop on the test tank.
- b. RPV injection tests using demineralized water from the test tank.
- 5. The pump, heater, valves and controls can 5. System tests will be conducted after 5. The installed equipment can be powered be powered from the standby AC power installation to confirm that the electrical from the standby AC power supply.
supply as described in Section 2.2.4. power supply configurations are in compliance with design commitments.
- 6. SLC System components which are 6. See Generic Equipment Qualification 6. See Generic dquipment Qualification required for the injection of the neutron verification activities (ITA). Acceptance Criteria (AC).
absorber into the reactor are classified m Seismic Category I and qualified for 2 appropriate environment for locations 8 where installed. O O O
O O O u VENT PRIMART CONTAINMENT NNS STORAGE 2 TANK
% H TE A L
NNSl 2 H A SAMPLING e SYSTEM
._ g_g_ (L.
HEATER
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__3 1 1 '
- -N - - ----- -- H PC F 'B' i 1
----- h. SUCTION VALVES M M (WITH POSITION j INDICATION) 3 i_.___t,.t____ _ ____ M 7' sME =
l I L ____ CODE 1 CLASS I
--------------~~~--!_______!
h--Ik pf-- I I i INJECTION l l l
! ; VALVES I PUMPS I
' I I j (WITH POSITION .
l l I I I t INDICATION) J I
- 1-- L___I i l M M i i i L--- h - -L,f---!
l I I NNSI t---- i Y u_____ r---- ASME I
! l CODE CLASS 2 ! l 2 8 I i P
(_tif1S el _ TEST $i TANK l l l 1 Figure 2.2.4 Standby Liquid Control System (Standby Mode)
ABWR Design Document g 2.2.5 Neutron Monitoring System (J . Design Description The neutnm monitoring system (NNIS) for the Advanced lioiling Watei Re..,.or (AllWR) is a neutron monitoring and protection system. The piimary functions of the system are to: (1) monitor the thei mal neutron flux in the icactos coic as reactor power information, (2) provide trip signals to the scartor protection system to initiate scartor scnim under excessive neutron flux (and thermal power) increase condition or neutron flux fast rising condition, and (3) provide power information to the operator and other plant control or process systems. The NNIS is classified as a safety related system. The salety related subsystems of the NhtS consist of the startup range neutron monitor (SRNN1), the local power nmge monir r (1.PRht), and the average power range monitor (APRN1). The LPRh! and the APRh1 together are also called power range neutron monitor (PRNhi). The non-safety related subsystems consist of the automated traversing in< ore probe (ATIP) system and the multi <hannel rod block monitor (51RisNt) system. The NNIS detectors and the safety related electrical equipments of the system are classified as Safety Class 2 and 3, Seismic Category 1, an.1 as IEEE electrical categorv Class 1 E. m The SRNht monitors neutron flux from the source nmge (l x 103 nv) to (] approximately 15% of the rated power. The SRNNI subsystem has ten SRNN1 l l channels which are evenly distributed throughout the reactor core and assigned to four safety divisions. The SRNh1 detector is a fixed in< ore fission chamber sensor. Detector cables are separated according to different divisional assignment, connected to their designated pre-amplifiers located in the reactor building, and then transmitted to : ;nal processing electroni i nits. The SRNNI can generate a high neutron flux tiip or a short period trip s e al to initiate scram in time to prevent fuel damage resulting from anticipm d or abnormal operational transients. Trip signal outputs from the SRNht channels are divided and assigned to four safety divisions. Any single SRNhl channel trip will cause a trip in this safety division. The SRNhi channels are grouped into three bypass groups independent of their safety division assignment. Individual SRNN1 l channel can be bypassed to allow maintenance. llypassed SRNN1 will not cause the trip signal to be sent to the RPS. The LPRM monitors local neutron flux in the power nmge from 1% to 125% of the rated power, which overlaps with the SRNN1 monitoring range from 1% to at least 10% of the rated power. There are fifty two LPRN1 detector assemblies evenly distributed in the core, with four sensors per each LPRh1 assembiv. The p LPRht detector is a fixed in< ore fission chamber sensor. The LPRN1 ass ~mbly also contains a calibration tube for the ATIP detector to traverse. The 1.PRNt detector outputs are connected to the APRh1 signai conditioning units, where 2.2.5 rd1/92 l
ABWR Design Document the signals are processed and amplified. All 1.PD1 detector signals are divided and assigned to four APUI channels co responding to the four safety divisions. Signals in each channel are sununed and averaged to fonn an APRh1 signal. The APD1 is then calibrated to represent the core aserage power. The APRh1 can generate a high neutron flux trip, a simulated thennal power uip signal, or a mpid flow decicase trip signal, to initiate scnun in time to prevent fuel damage resulting from anticipated or abnonnal opennional conditions. Any two APRN1 trips out of the four APRN1 channels willinitiate a reactor scram trip in the RPS. A bypassed APRN1 channel will not cause a trip output sent to the RPS. One APRN1 channel can be bvpassed at any one time. Consequently, for both the SRNh1 and the APRhl, the redundancy criteria aic met such that in the event of a single failure under permissible SRNh1 or APRh! bypass conditions, safety protection function can still be performed. A typical NhtS division block diagram is shown in Figure 2.2.5. The ATIP is comprised of a set of three TIP machines. Other than the power probe detector itself, each machne has a drhe mechanism, a position indexing mechanism, and associated guide tubes fordetector traveling. Within each ATIP machine, the ATIP detector is traversed via guide tubes and tnrough desired index positions to the designated LPRN1 assembly calibration tubes. Flux readings along the axiallength of the core are obtained while the detector is traversed along the fuel region, with the signal data sent to an ATIP control unit for data processing and storage. The data are then sent to the process computer for calibration and performance calculations. The whole ATIP operation can be ftdly automated, with manual control capability. The h1RBhl utilizes a selected number of LPRh1 signals around each designated control rod to detect local power change during the rod withdnnval. If the averaged LPRN1 signal exceeds a preset rod block setpoint, a control rml block demand is issued. The setpoint is detennined based on analysis which assures that the fuel thennallimits do not violate the safety limits. Since it monitors more than one region, it is called the multi-channel rod block monitor. The $1RBh1 is a dual channel, highly reliable system. The NhtS provides trip signals to the RPS as part of the RPS safety protection ftmction inputs. All trip setpoints are adjustable. The SRNNI trip and the APRh1 u-ip are separate logics to the RPS, each interfacing with the RPS independently. Fail-safe logic is used for both subeystems. The NhtS bypass function is perfonned within the NhtS. The bypass functions of the SRNNI and the APRh1 me separate and independent from each other. Both the SRNh1 and the APRhl are designed to pennit functional testing during innal plant operation. Provisions exist to limit access to trip sctpoints ano calibnuion controls. All the SRNht, LPRh1 and APR$1 instruments are powered by four 120 VAC un-interruptible power supp' - (UPS) buses A. IL C. and 1) that correspond to the Ol 2.2.5 2- 6/1/92
ABWR oesign Document q four safety dhisions. Each bus supplies power to appr oximately one fourth of the V total number of detectors. Loss of a powei supply bus will cause the loss of the divisionalinstruments, including the SRNh1 and the Al'RAl. The power for the ATIP is supplied from the instnunent AC power source. The power supply for the AfRllhi is from the non<livisional 120 VAC UPS bus. The SRNat and LPRAI detectors and detector assemblics are designed to operate under normal and design basis abnormal wnditions. The SRNht pre-amplifiers which are located in the reactor building, and the NNIS instruments which are located in the control room, are designed to operate under all expected environmental conditions in those areas. The wiring, cables, and connectors in the drywell are designed for continuous duty under normal and design basis abnormal c onditions. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.5 provides a definition of the inspections, tests, and/or analyses with associated acceptance critena for the NhtS. I i t l 2.2.5 6/1/92
o,_.2 : s =- , . U Table 2.2.5: Neutron Monitoring System w Inspections, Tests, Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment
- 1. Inspection of certification documents from 1. The inspec* ion must confirm the follov.{ng
- 1. The power monitoring range for the SRNM detector performance range:
shall be from shutdown to at least 15% the detector manufacturers will be conducted to confirm the specifiad power a) SRNM: shutdown level to at least 15% of rated power. The power monitoring ranga rated power; for the LPRM shall cover an equivalent cor e monitoring range. b) LPRM: ir dividua: detactor overall range average power range frorr.1% to 125% of equivaief-t to : core a serage powe of 1% rated power. to 125% of rated power.
- 2. Inspcctions of installation records and 2 The system configuration is in accordance
- 2. A simplified system configuration is shown with Figure 2.2.5.
in Figure 2.2 5. plant walkdowns will be conducted to confirm that the installed equipment is in compliance with the design configuration , i defined in Figure 2.2.5. 1 The divisional assbment and separation 3. Inspections and plant wa* downs will be 3. The SRNM and the PRNM equipments )
- 1. 3. must be arranged so that the basic l of the four redundant SRNM andthe PRNM conducted to confirm the four division redundancy and the electrical and physical requirements of equipment physical and subsystems sha!! be properly electrical separation are met.
implemented. separation of the four division SRNM and j PRNM e ,v ments. . 4 SNM and W' I trip functions will be 4. On the installed equipment,the system l
- 4. The trip functions of the SRNM and APRM must be able to issue trip signals during are properly implemented as described in tested 9..%gh the testing of the eleurical somporwnt Other field tests wit! be also functional test for the foliowing trip Section 2.2.5. The system is designed to conducM uit r system installation to functions:
permit functional testing of the NMS al SRNM period trip; during normal plant operation. confirm system functional tering can be per!crmed. b) APRM upscale trip; c) AP9M thermal power upscale trip; d) APRM rapid flow decrease trip, e) SRNM and APRM inoperative trip
- 5. , tem tests will be conducted after 5. The instal!ed equipment is powered from
- 5. The Nsir ignments of the SRNM the four divisional Class TE UPS power anc. WIL i supply is provided by the .nstallation to confirm that the electrical power supply configuration is in sources.
four l' O VAC ' S buses. 3 complianc,. dth design commitments. IE o O O O d
f g Table 2.2.5: Neutron Monitoring System in Inspections, Tests, Analyses and Acceptance Criteria t Certified Design Commitment !nspections. Tests, Analyses Acceptance Caiteria
- 6. The SRNM and APRM trip setpoints of trip 6. Inspections and tests will be cor*.sucted to 6. The efectac equipment must be tested to signals to the RPS shall be adjustable and confirm that t!.e setpoints me adjos -ble show that th e setpoints are adjustable and .
properly implemented, using applicable and are properIV imisiemented. the setpoint values are preparty setpoint methodology. implemented according to plant technical specification. j
- 7. Provisions exist tu limit access to trip 7. Inspections and .asts will be conducted tu 7. Appropriate method of security controts setpoints and calibration controls. Confirm the existence of appropriats must exist to change trip setpoints or to security controls, calibrate the instruments.
h l e 8
ABWR ossign cocement DETLCTOR DETECTOR (2 OR 3I 1 20 27 52 T RANSMITTER eee . _..o e e l -- V y l l 1 1 I I -- i l A SRNM A REACTOR l I DUIL DING CLASS 1 E I l___ _____
~~ ~~~~~~
CLASS 1E ~t yy y y Y YY YY OtH RO g U V BUILDING I f lSnNul"~~~1, APRM I I N FROM INTERFACE I
-- UNIT
{ I I I i
! DEDICATED INTERFACE I l INT ERFACE UNIT (l'O) j l i L __ _..___ _ _ _ __ _ _ _ _ _ _ _ _ __ _ ._ __ _ _ _ ._ _ _ __ _ a O
V V V Y V V V V V V V RC&IS RPS BO RD MREM ANNUN o ATI C RL RECORD V JI RCatS ATLM DISPL CONSOL RC&lS: ROD CONTROL & INFORMATION SYSTEM ATLM: AUTOMATED THERMAL LIMIT MONITOR PART OF RC&ls RECIR CONTRL: RECIRCULATION FLOW & CONTROL SYSTEM CONTROL Figure . 2.5 Neutron Monitoring System O i t 2.2.5 -G- W1/92
AB~WR Design Document 2.2.6 Remote Shutdown System Design Description The Remote Shutdown System (RSS) for the Advanced lioiling Water Re.u tor (AllWR) provides iemote manual contiol of nonnal atul nuclear safety aclated systems necessaiv to ining the i.ractor to c old shutdown (orulitions in an orderly fashion hom outside the inain control room. No 1.oss of Coolant Accident (liCh scismic event, or otoer abnonnal plant condition, except loss of off site power. is assmned to occur coincident with the event acquiring the main contial room evacuation. The RSS has two divisional panels and associated controls and indicators for monitoring the following inted~ acing systems: (1) Residual Heat Removal System (RHR) (Pool cooling and shutdown cooling modes). (2) High Pressure Core Flooder Systen. (HPCF) (3) Nuclear 1. toiler System (Nits). Safety Relief Valves ( 1) Reactor Service Water System (RSW) (5) Reactor lluilding Cooling Water System (RCW) (6) Electrical Power Distribution System (EPDS) (7) Atmospheric Control Sys'em (AC) (8) Emergency Diesel Generator (D/G) (9) Stake-up Water Condensate Systern ( AtUWC) (10) Flammability Gas Control System (FCS) The RSS is classified as a safety-related system because it interfaces with nuclear safety-related equipment from other systems. The two remote shutdown control panels are Seismic Category I and are located in a single remote shutdown sta' ion in the Reactor lluilding. A physical barrier provides separation between the two panels The RSS provides remote control capability through control and transfer switches in the RSS panels which override the controls from main control room and tansfer rontrol to the RSS pam Is. Indication for plant parameters is also provided on the remote shutdown panels (\ to assure a safe and controlled shutdown of the plant. Figure 2.2.6 shows the RSS with the interfacing systems and control and indicatiori functions provided. 2.2.6 1 6/1/92
.. . . -- .. - = _ - - _ _ . _ . .. - -_
ABWR oesion Document Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.fiInosides a <!cfinition of the sisnal inspections, tests aiul/or analyses, together with associated acceptance criteria, which will be perfor med for the RSS. O 2.2.6 2 G/1/92
S
-{ Table 2.2.G: Remote Shutdown System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria
- 1. RSS provides a remote manual control of 1. Review o' as-built documentation and 1. RSS has the required plant system control the following interfacing systems to bring visualinspections of the RSS will be capablity.
the reactor to co!d shutdown conditions: performed. Testing of the RSS control functions will be performed.
- a. RHR (pool cooling and shutdown cooling modes)
- 6. HPCF
- c. NBS Safety Relief Valves
- d. RSW
- e. RCW
- f. EPDS
- g. AC
- h. D/G g i. MUWC
- j. FCS
- 2. The RSS has two divisional panels for 2. Visual inspections and documentation 2. The panels conform to their requirements monitoring and contro!!ing of the review to confirm the appropriate location, for c'ivisional separation and seismic i interfacing systems. The panels are isolation, and seismic capabilities of the criteria.They are located in a separate RSS
) physically separated and are located in a panels. statio.1. ! remote shutdown station.
- 3. RSS providesindication of plant 3. Visual inspections and review of as-built 3. The RSS has the required plant monitoring parameters in RSS panels to monitor a documentation relating tr> RSS monitoring capability.
- controlled shutdown of the plant. function.
l l . E
U - RllR A. 8 FLOW L> - HX. trJLET TEMP. A. B
- HX. If;LET VLV. POS. A. 8
- RX. LVL (WLR)
- HX. BYPASS VLV. - RX LVL (SHUT) 63 KV WC g. D
- CSP LEVEL VOLTAGt POSITk71 - RX PRESS.
i i i i
! HPCF ! . S.P TEMP ! - RCW A. B !
l FLOW l - SP LEVEL l FLOW l RUtrSTOP i I i I i I i i i I I i i I i i
- I gi al El 21 al eI el "I I "I "I 'l 'l "I dl
= 1 I I l l l l 1 8 Y Y Y Y Y Y Y Y O
5 RSS PLANT PARAMETERS z PANELS IrJDiCATORS A. 8 - CONTROL & TRANSFER SW',TCHES
- ~~~ ~
j, - VLV. POSITION INDICATION g - PUMP STOP rut 1
~
E 2 8 A A A A A A A I I I l l I i l i I i i I I I I i i I i i I I I Y Y Y Y Y Y Y E11 Rio B21 E22 P21 Pat FCS- A. B 1 RHR-A, B EPOS N.B HPCF-8 RCW- A. B RSW- A. B 49KV MC -SAFETY
- SUPP. POOL COOLING -480V P.C RELIEF MODE VALVES D'V 8 "
. sHurDOwn COounG MODE cy M
;S Figure 2.2.6 Remote Shutdown System O O O
ABWR oosign occument n 2.2.7 Reactor Protection System
! i V Design Description The Reactor Protection System (RPS) for the Advanced Boiling Water Reactor (ABWR) is a warning and trip systein wheie initial warning and tr ip decisions aie implemented uith sof tware logic instr.lled in microprocessors. The primary functions of this system is to proside prompt protection against the onset and consequences of events or conditions that threaten the integrity of the fuel barrier. To accomplish this, the system is designed to: (1) make the logic decisions related to warning and trip condinons of the individualinstnnnent channels, and (2) make the decision for system trip (cmengency icactor shutdown) based on coincidence of instrument channel trip conditions.
The RPS is classified as a safety protection system (i.e., as dif f ering from a icactor control system or a power generation system). The functions of the RPS and its components are safety-related. The RPS and the electrical equipment of the system are aise classified as Safety Class 3, Seismic Ca'egory I and as IEEE electrical category Chtss lE. liasic system parameters are: (1) Number ofindependent divisions of equipment 4 (2) Minimum number of sensors per trip variable 4 (at least one per division) (3) Number of automatic trip systems (one per division) 4 (4) Automatic trip logic used for plant sensor inputs 2mut-of4 (per division) (5) Separate automatic trip logic used for division 2mutof4 trip outputs l (6) Number of separate manual trip systems 2 l (7) Manual trip logic 2-ou tof-2 The RPS consists ofinstrument channels, trip logics, trip actuators, manual controls, and scnun logic circuitry that initiates rapid insertion of control rods (scram) to shut down the reactor for nituations that could result in unsafe reactor operating conditions. The RPS also establishes the required trip , conditions that are appropriate for the different reactor operating modes and l (% provides status and control signals to other systems and annunciators. The RPS l related equipment includes detectors, switches, microprocessors, solid-state logic circuits, relay type contactois, relays, solid-state load drivers, lamps, 2.2.7 6/1/92
ABWR Design Document \
\
l displap, signal transmission souteA circuits, and other equipment which aic required to execute the f unctions of the sptem. To ac complish its oserall , funct ion, the RPS utihees the functions of'he essential muhiplexing ssstem (Eh1S) and of portions of the Safety System 1.ogic and Contiol (SSI.C) System. As shown in Figure 2.2.7a. the RPS interf aces with the Neution N1onitoiing System (NNIS), the Process Radiation N1onitoring (PRRht) System, the Nuclear lloiler System (Nils), the Cono ol Rod th ive (CRD) System,ihe Rod Cono ol and Inf ormation System (RC&lSh the Rech culation Flow Control ( RFC) System. the Process Computer System, and with other plant systems and equipment. RPS components and equipment me separated or segiegated from process control system sensors. circuits and functions such as to minimize cont rol and protection system interactions. Any necessary imerlocks f rom the RPS to connol systems are through isolation desices. The RPS is a four-dhision system which is designed to proside reliable single-failure-proof capability to automatically or manually initiate a reactor sciam while maintaining protection against unnecess.uy scnuns resulting from single failures in the RPS. The RPS remaias single-failure proof even when one entire dhision of channel sensors is bypassed and/or when one of the four automatic RPS trip logic systems is outof-service. Equipment within the RPS is designed to fail into a trip initiating state or other safe state on loss of power or input signals or disconnection of portions of the system. The system also includes trip bypes g and isolated outputs for display, annunciation or perfonnance monitoring. RPS inputs to annunciators, recorders and the computer are electrically isolated so that no malfunction of the annunciating, recording, or computing eqmpment can functionally disable any portion of the RPS. The RPS related equipment is dhided into four redundant divisions of sensor (instnanent) channels, trip logics and trip actuators, and two divisions of manual scnun conuols and scram logic circuiuy. The automatic and manual scnun initiation logic systems are independent of each other and use diverse methods and equipment to initiate a reactor scram. The RPS design is such that, once a full reactor scnun has been initiated automatically or manually, this scram condition seals-in such that the intended fast insertion of control rods into the reactor core can continue to completion. After a time delay, the design allows operator action to return the RPS to normal. 1 Figure 2.2.7b shows the RPS dhisional separation aspects and the signal flow paths from sensors to scnun pilot valve solenoids. Equipment within a RPS related sensor channel consists of sensors (transducers or switches), I multiplexers, and digital trip modules (DThis). The sensors within each channel monitor for abnormal operating conditions and send either disciete histable (trip /no trip) or analog signals directly to the RPS related DTht, or else send analog output signals to the RPS related DThi by means of the iemote I multiplexer unit (Rh1U) within the associated division of essential multiplexing 2.2.7 61/92
ABWR oesign Document system (EhtS). The RPS telated histable switch type sensors or, in the ree of analog channeh, the RPS software logic wdl initiate reactor trip signah within the individual senso ch:umels, when any one or more of the comfitions listed below exist within the plant dming diffeient conditions of accctoi opeiation. and willinitiate reactor senun if coincidence logic is satisfied (the wst"m monitoring the pro (ess condition is indicated in brackets). (1) Tu:bine Stop Valves Closure (above 40% power levels) [ RPS) (2) Turbine Control Valves Fast Clostue (above 40% powei levelo [RPS] (3) Nh!S monitoicd SRNh1 and APRh1 conditions exceed acceptable limits [NSIS] (4) liigh hiain Ste:un Line Radiation [PRRh1 System] (5) liigh Reactor Pressure [ Nils] (6) low Reactor Water f.evel (I evel 3) [N11S] (7) High Dryweh Pressure [ NILS] (8) Stain Steam Line Isolation (h1SLI) (Run mode only) [NitSJ (9) 1.ow Control Rod Drive Accumulat r Charging Header Pressure [CRD] (10) Operator-initired hianual Scnun [RPS] The RPS outputs, NhtS outpua PRRhi System outputs, and the htSLI and nanual scram outputs are prc.ided direcdy to the RPS by hard-wired or fiber. optic signals. The Nils and the CRD System provide other sensor outputs dirough the EhtS. Analog-to<ligital conversion of these latter sensor output v; dues is done by EhtS equipment. The DThi in each division uses either the dtscrete bistable input signals, or compares the cun'ent values of the individual monitored analog variables with their trip setpoint values, and for each variable sentis a separate, discrete bistable (trip /no trip) output signal to the trip logic l units (TLU3) in all four dwisions of trip logics. The DThis and TLUs utilized by the RPS are microprocessor components within the SSLC System. RPS related equipment within a RPS division of trip logic consists of manual control switches, bypats units (ItPUs), trip logic units (TLUs), and output logic units (OLUs). The manual control switches and the BPUs, TLUs and OI.Us are component., of the RPS portions of the SSLC System. The v,uious manual 3 switches provide the operator a means to modify the RPS trip logic for special operation, maintenance, testing, and system reset. The bypass units perform bypass and interlock logic for the single division of channel sensors bypass function and for the single dhision TLU bypass function. The T1.Us perIor in the 2.2.7 3- G/1/92
ABWR oesign occument automatic scnun initiation logic. nornrdly checking foi twmint4,1-fom g coincidence of nip conditions in anv set of insu ument channel signals (oming W hom the f om4thision DTMs or hom isolated histable inputs hom all fom divisions of NMS equijuneat, and outputting a nip signalil any one of the two-out-of-f our coincidence checks is satisfied. T!.l' u ip decision logic in all Iour RPS T1.Us bc(omes a chet L for twocutof-ilure toim idence of uip conditions if any one division of c hannel sensors has been bypassed. The UI.Us perfor m the division nip. scal in, reset atul uip . cst f unctions. Tiip signals h om the OLUs within a single dhision ar e used to u ip the tiip actuators, which .n c last r esponse, histable, solid. state load driveis for automatic sc ram initiation, aini aic nip relays for air headei dump (backup scnun) inination. l oad driver outputs teggled by a division 01.l' interconnect with load driver outputs toggled by other division OLUs into two separate arrangements, which resuhs in twocut of-lour scram logic (i.e., s cactor scnun will occur il load driveis associated with any two or moic divisions receive trip signals). The isolated AC load drivers are fast response time, histable, sodd-state, high current interrupting devices. The operation of the load drivers is such that a trip signal on the input side will cicate a high impedance, cunent internipting condition on the output side. The output si.le of each load driver is electrically isolated from its input signal. The load driver outputs are arranged in the scnun logic circuitry, between the scram pilot valves' solenoids and the solenoids' AC power source, such that, when in a uipped state, the load drivers will cause deenergization of the scram pilot valve solenoids (scram initiation). Nonnally closed relay contacts are arranged in the two backup senim logic circuits, between the air header dump valve solenoid and air header dump valve DC solenoid power source, such that, when in a tripped state (coil deenergized), the relays will cause energization of the air header dump valve solenoids (air header dump initiation). Associated DC voltage relay logic is also utilized to ellect scnun reset pennissives and scnun-follow (control rod nm-in) initiation. The RPS design for the AIMR is testable f or conect iesponse and performance, in overlapping stages, either on line or off-line (to :niniinire potential of unwanted tiips). Access to bypass capabilitica of trip functions,inrtrument channels or a trip system and access to setpoints, calibration controls and test points are under adminio save control. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.7 provides a definition of the visual inspections, tests and/or analyses, together with as ociated acceptance criteria, which will be used by the RPS. O 2.2.7 -4 6/1/92
4 i ; i t { Table 2.2.7: REACTOR PROTECTION SYSTEM i Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria
- 1. RPS components and equipment are itept 1. Visual field inspections and analyses of 1. RPS equipment installation acceptable if l' separate from equipment associated with relationship of installed RPS equipment inspections, analyses and/or tests confirm process control systems, and of installed equipment of interfacing that any failure in process control systems process control systems (and/or tests of can not prevent RPS safety functions. I interfaces) to confirm that appropriate isolation methods have been used to satisfy separation and segregation
- requirements.
- 2. Fail-safe failure modes result upon loss of 2. Field tests to confirm that trip conditions 2. Acceptable if safe state conditions result j power or disconnection of compcnents. and/or t fars inhibits result upon loss of upon loss of power or disconnection of
) power on , ist cnnection of components, portions of the RPS. ; 3. Provisions exist to limit access to trip 3. Visual field inspections of the installed RPS 3. The RPS hardware /firmware will be f
, @ setpoints, calibration centrols and test equipment will be used to confirm the considered acceptable if appropriate points. existence of appropriate administratise methods exist to enforce administrative , I, controls. control for access to sensitive areas. I
- 4. The four redundant divisions of RPS 4. Inspections of fabncation and installation 4. Installed RPS equipment will be equipment and the four automatic trip
~
records and construction drawings or determined to conform to the documented { systems are independent from cach other visual field inspections of the installed RPG description of the design as depicte' in 3 except in the area of the required equipment will be used to confirm the Figure 2.2.7b. ! coincidence of trip logic decisions and are quadruple redundancy of the RPS and the j both electrically and physically separated electrical and physical separation aspects ; J from each other. Similarly, the two manual of the GPS instrument channels and the i 4 trip systems are separate and independent four automatic trip systems, as well as of each other and of the four automatic trip their diversity and independence from the j systems. two manual trip systems. ! I E N ' i
N Table 2.2.7: REACTOR PROTECTION SYSTEM (Continued) e u Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. It is possible to conduct verifications of 5. Preoperational tests will be conducted to 5. The installed RPS configuration, controls RPS operations, both on-line and off-line, confirm that system testing such as power sources and installations of by means of (a) individual instrument channel checks, channel functional tests, interfacing systems supports the RPS locic channel functional tests, (b) trip system channel calibrations, coincident logic tests system functionc! testing and the
, functional tests and (c) total system and paired control rods scram tests can be ooerability verification of design as functional tests. performed.These tests will involve follows:
simulation of RPS testing modes of operation. Interlocks associated with the a. Installed RPS he.-dware/firmware reactor mode switch positions, and wi*h initiates trip conditions in all four RPS other operational and maintenar.ce automctic trio systems upon bypasses or test switches, w9l be tested coincioence of trip conditions in two or and annunciation, display and logging more instrument channels associated functions vdli be cor. fir med. with the same trip variable (s). E
- b. nstalled system initiates full reactor trip and emergency shutdown (i.e.,
deenergization of both solenoids associated with all scram pilot vaives) upon coincidence of trip conditions in two or more of the four RPS automatic trip systems.
- c. Installed system initiates trip conditions in both RF* ianual trip systems if both manua. tr;p switches are operated or if the reactor mode switch is placed in the " shutdown
- position.
- d. Trip system (automatic and manual) trip conditions seal-in and protective
, actions go to completion. Trip reset 4 (after appropriate delay for trip
$ completion) requires deliberate operator action.
9 9 9
4 L
- e i
t ' \ N N -Table 2.2.7: REACTOR PROTECTION SYSTEM (Continued) 3 Inspections, Tests, Analyses and Acceptance Criteria l 4 Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria ;
- 5. (Continued)
- ' e. Installed system energizes both air
- header dump (backup scram) valves of I the CRD hydrtulic system, and initiates CRD motor run-in, concurrent only with a full scram condition.
- f. When not bypassed, trips result upon ,
1 loss or disconnection of portions of the j 4 system.When bypassed. ! inappropriate trips do not result. ; 7,
- g. Installed system provides isolated i l status and control signals to data ;
J logging, display and annunciator l systems. . 4 i i
- h. Installed system demonstrates j operational interi'ocks (i.e., trip inhibits or permissives) required for different
; conditions of reactor operation. I l i j 6. The RPS design provides prompt 6. Preoperational tests will be conducted to 6. The RPS hardware /firmware response to protection against the onset and [
} measure the RPS and supporting systems initiate reactor scram will be considered i i consequences of events or conditions that response times to: (a) monitor the variation acceptable if such response is threaten the integrity of the fuel barrier. j of the selected processes;(b) detect when demonstrated to be sufficient to assure
- trip setpoints have been exceeded; and, (c) that the specified acceptable fuel design i
{ } execute the subsequent protection actions limits are not exceeded. I { when coincidence of trip conditions exist. j [ l 5
~
I 4 i [ t j i
p [ LOCAL AREA T APRu m** Fu twa* MAIN CONTROL ROOM PLANT SENSORS M cs1 j]*j**. RPS LOGIC & CONTROL j NMS trop - 2 C71 C71 Svomssed (nrit W C71 C71 C71 C71 2 RPS RPS SR?aA W*Un Fluu Upsemie RPS RPS RPS RPS Re.w:*or tm Rem a S N t Paruj Manuai Onr T p MANUAL MAP &JAL Pa*ed 4kwfs
, Moo. B Y P8" ou+pu s stoP wt a SCRAM SCruu Sc-a= Tec 9 s.,tc., h5 . m Ht t Byoassef (not teW , , ear.,, e,.3 r r A e s.9ches T
i 3 Hf --- t , a h me --- t e D11 3 f pl PRRM 14 0 @ MSL Radaron - C71 s " H' ; 1r 1r RDS k 1 I SSLC RPS LOGtC PERFORMS:
+ AS n h r
- S~ C- Tm -
-System CoccMence Trp Dec.em S'M RE ! :
q-3S - C-** - s-** tv { i r-c2'3 L i CHD HCU Accumutaro,
- Mariusi Dvoort Trg CRO CT2
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E- m 2 I- .- Sen. n_- sys-a w_ og, gcg-- H23 M C C - Cae se.- - o an ' tv.n %we so.m riu v.sve ; ; [ e25R- e w w.v- toww --*- S - + % ws % NC3 ,R#f 1 Di*'_P_r;rf _______ C71 RPS LOGIC g>,v . n. macr ~n l Sr < MSN Pm on S ace.s , p REACTOR DTM $j . PROTECTION TLU
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- D"*P V*I'*S u v.Saw SYSTEM OLU C71 Tv.t no HTS 04 Presure Wmion r gg g dhv. 11, !!! IK' Pomen l Tu< bee 1st Stage Pressure y '
RPS w S.smic A nmy # i; W SSLC tog-c t d P"x *58*9 Pmcess.3 9 7 e@prW for O.N,, g e ,
$ ince*$ Se.ev Snw c11 m ar. Se ,ro w ng 1 __ - - n <,m, ; RcarS pc r* R* Ru+: # <
I f GLOSSARY w
- APRM - A.+ age Pome Range f.bmv l CRD - Cetd "ke Oma FOR 7cuW 4 CONCIDEPCE LOGfC 1 EOC - Enu 08 Cytie gI EuS -t*cw usp* m S e ; 1
, g .
{HCU - Hy+wc Cmts LW f f C11 RCIS Rod WW Weeni BWh When Bypass Se th Fw CRO iN E N*RN j HT3 - Hyt aR T*e Sys'e A g , % t ,p,% ,,'y,,,w g g p ,, , t'w *E ??o w AC W f **' s MCRP - Wn Cor*=$1 f% = Perss $ 7 , f MSfV - Wn Stear : inwr Van,= # ' k "" Y # "s ' E anSL - Wn S*ean . U= C91 P ocesa Cmtee - Oca axw 0*svavs Den torpng r g ,, i Nas - Wi a N#r sysn m 4 H11 MCRP-1 CW Co pes - Ope arew Ovavs. Awnc eioes %. v y;
!NMS - Nv *on Shoeng Syse '
f PRRM - P exess Rafdon L4r tormg Syee- 8 R42 aos a-w sr - ae - { L'CtrS - FW! Cor*M & Marmare Sce*~' W ""
- l e iRFC - Raetvide F h Ce*W Oyste w .
y l PS - ww.w rnem S,sv " p agra,1 h hpicd Fa rr One Of Four Divisions' l { IRPT - Rmreusse Pep Tv I f SANM - S*w'$ Rantje fMMi* o** SArCr [
} SSLC - S4*" Sys**m L(%< & Cu*c3
{ UPS Power Sepp' y i
-. _ _U W-arvuptN h
g Figure 2 2.7a Re r Protection System
ABWR oesign Document _ __- - . . . _ - - _ ~ .-. O Division 1 Division 2 Division 3 Division 4 Sensca sensor 3ensor aensor A B C D l I I i 1 l 1 1 Division 1 Raceway DiviWon 2 Raceway Dmsion 3 Raceway Division 4 Raceway DTM DTM DTM DTM A B C D ( L L.~.. , .,) .J ' / 's..- . -! .Ji ) (s~.'s.
- s. . ~.,s. s .-
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- s. s ~.F s, c .s , s s i . __ J l I i __ J l J Diviuon 1Itact any Divkion 2f(aceway Divioon 3Fiaceway Divis ion 4f .ac eway TLU A TLUB TLU C TLU D I I .._
l i I OLU A l l OLUB l l OLUC l l OLU D l O - i
~ -
I
~
i i l _~_.. - - - - _ _ _ me opt,e Cable - * - - i _ ._ _ .- a______ l 'I I ___I I I RPS-G3 RPS-G4 I l RPS-G1 I i I l RPS-G2 l l l l Raceways, I l l l l l l,i , , I I I I I Output I l Output l Same l Same l Sane l l Circuit l Circuit I as G1 l as G1 I as Gi I Group 1(A) Group 1 (D) g j l l l l Div. 2 Div. 3 I I I I ! I I I I I I Solonoid A I I I I sotonoid I
-y a we-- l l
l Vent l (Manual trip or test logic interfaces not shown) Inct. AirMb . .
- t '
.- o Exhaust **' Y ' - -
- To scram valve
__) Figure 2.2.7b Reactor Protection System 2.2.7 C/1/92
._, .-._...-., - - - , , , - . . , . - , - . . , . . . _ - . . . _ _ ~ _ . - . . _ , . . ._ - . .-
ABWR 0: sign Document 2.2.8 Flerirculation Flow Control System i
\ Design Description The Recin ulation llow Control (RFC) System conuoh ica( tor poner by controlling the recirculation flow sate of the reactor core wates. Reactor recirculation now and core flow is varied by nuxlulating the Recirculation Internal Pump (RIP) speeds / flows thtough the voltage and frequency modulation of adjustable speed drive outputs. Refer to Figure 2.2.8.
The RFC System consists of the triplicated process controller, solid state Adjustable Speed Drives (ASDs), switches, sensors, and alar.n devices provided foi operational manipulation of the ten RIPS and the surveillance of anociated equiprnent. Recirculation flow control is achieved either by manual operation, or by automatic operation if the power level is above approximately 70% of rated. The reactor internal pumps can be driven to operate anywheie between minimum speed and 100% of rated speed with the variable voltage, nuiable frequency power souice supplied by the ASDs. This system is a power generation system and is clauified as non-safety-related. The RFC System is designed to allow both automatic and manual operation. In the automatic mode called " Master Auto" mode
- Automatic load Fo!!owing q
, V (ALF) operation) the master contioller generates a demand signal for balancing out the load dernand enor to zero. This demand signal is forwarded to the 110w conuoller which generates a flow demand signal. The flow demand signal b adjusted by a flow demand set down function to lower the recirculation now when the sensed reactor flux is above 105E The speed controllers in the ASin generate speed demand based on the now demand from the flow contr oller. The speed demand causes adjustment of RIP motoi power input which changes the operating speed of the RIP and hence core flow and core power. This process continues until both the errors existing at the input of the Dow controller and
- master conuoller are driven to zero. The flow controller can remain in automatic even though the master controller is in manual.
The ieactor [x;we. change resulting from the change in iecirculation now caums the pressure regulator to reposition the turbine contial valves. If the cris;inal demand signal was a load / speed error signal, the turbine responds to the i change in reactor power level by adjusting the control ndves, and hence its power output, until the load / speed enor signal is reduced to zero, in the semi automatic mode, the operator sets the total core flow demand and ! the RFC System responds to maintain constant core flow. Core flow control is p achieved by comparing the core flow feedback, which is calculated from the core d plate differential pressure signals, with the operator supplied coie flow set point. J 2.2.8 1 Gl1/92
1 ABWR oesign Document i in total manual control, the operator can direttiv manipulate the RIP speeds. l'tunp speeds can he c antrolled indisidually or c ollectively. When individually g contio: led, putop speed demand is obtated through the openitor console and transmitted directly to the indisidual ASD f or pump ficquency control. In collective manual operation, a common speed set point is used for controlling cath RIP which has been placed in the GANG speed control mode.
)
The recirculation flow contiol system is aho used to (ontrol the start up of the f e.u toi kuternal puinps. To ininiinire thennal shock to the icactor vessel, the RFC System willi nevent uait up of an idle RIP if the temperature difference of the vessel bottom coolant to the saturated water teinperature corresponding to the steam dome pressure is above a predetermined value, in the even of either (a) turbine uip or generator load rejec tion above a piedetermined icactoi power level. (b) reactor pressme exceeds the high dome pressure trip set point,
- o. (c) reactor w;uer level drops below the Level 3 set point, logic will automatically be initiated to trip off a group of loui RIPS.
ASDs me used to provide electrical power and speed control to the pump motor s in the RIPS. The ASD ieceives electrical power froin a power plant hus at a constant AC voltage and frequency. The ASD < onverts this to a ariable frequency and voltage in accordance with the speed demand requested by the RFC Systc n controller. g The ASD is capable of supporting thice naxies of operation: start up, normal, and shutdown. When the start up mode is selected, the inverter output quickly steps up f rom iero to the required motor power conesponding to the minimum pump speed and holds at that output frequency. When the nonnal operation mode is selected, continuous output power frequency between minimum speed and 100% is allowed. The operation of the shutdown mode is exactly reverse that of the nonnal and start up mode: ASD output is automatically nunped to minimum speed fiequcncy, then stepped down to iero. The RFC System control functional logic is perfonned by a triply redundant, microprocessor ba. sed fault tolerant digital controller (FFDC). The FTDC consists of three identical processing channch working i,. pandlel to provide fault tolerant operation. The RFC System design consists of two main control k> ops, (1) the core flow loop, which modulates pump speed demand to provide the desired core flow rate, and (2) the automatic load following (Alf) which modulates the core flow demand in response to the load demand error In addition, pump speed in each RIP can be manually controlled individually or collectively. In the core flow control mode, sensed core flow calculated by the core plate differential pressure method is compared with the core flow demand supplied 2.2.8 C/1/92
ABWR oesign occument p by the operator or obtailled fioin the tuaster t onttoller, dependilig on the RFC systein operating inode. This flow crtor is input to the core flow controlles to drive putup speed dernand. In ALD inode, the inaster contioller neceives a load dennand signalIroin the stcain hypaAs and pressure control (SP&PC) systern in response to any cornbination of local operator load set point inputs, autotnatic generation control inputs, or grid load changes indicated by grid frequency v.uiation. When in local control, the operator's control panel ptosides the operator the capability to select the operating rnode of the systein and to initiate certain inanual actions. indicatiotis and alarms ate pro \ided to keep the operator infonned of the systein opennional inodes and equijnnent status, therchy allowing hini to quickly detennine the origin of any abnonnal conditions. Inspections Tests, Analysn and Acceptance Criteria Table 2.2.8 provides a defhstion of the inspec tions, tests, and/or analyses, together with the associated acceptance criteria, whic h will be undentalen for the RFC. O V P) c 2.2.8 3 6/1/92
y ~ m Table 2.2.8a: Recirculation Flow Control System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Test. Analyses Acceptance Criteria
- 1. The configuration of the RFC system is 1. Inspections of the as-built RFC system shall 1. Actual RFC system configuration, for those shown in Figure 2.2.8-1. be performed components shown. conforms with Figure 2.2.8-1.
- 2. Reactor intemal pumps (RIPS) will operate 2. Operation of the pumps at any speed 2. Pumps shall operator within design at any speed between minimum speed and between minimum speed and 100% of specificatien limits at any speed between 100% of rated speed. rated shall be performed. 10% and 100% of rated.
- 3. The RFC System may be operated in 3. The RFC System shall be operated at 3. The RFC System shall operate in automatic automatic mode above approximately power levels greater than approximately mode within design specification limits at 70% of rated power. 70% rated. any power level above approximately 70%
rated. i,
- 4. The RFC System rhall be used to control 4. The RFC System shall be operated in the 4. The RFC System shall operate within the the start up or shut down of the R!Ps. start up and shutdown modes. design specification limits in the start up and shutdcm modes. The pump shall be ramped from 0% to 30% and held and then shall be stepped down to 0%
- 5. The RFC System shall be interlocked so as 5. The RFC System shall be operated so as to 5. The RFC System shall prevent start up of to prevent start up of a idle RIP if the vessel start up an idle RIP when the vessel bottom an idle RIP under conditions speciied in bottom temperature is not within 144'F of temperature is not within 144'F of the the design specification.
the saturated dome pressure equivalent saturated dome pressure temperature temperature. equivalent.
- 6. A select group of Rips shall trip off in the 6. The RFC System shall be operated and the 6. The RFC System shalf oparate within event of either (a) turbine trip or generator following events shat! be simulated: design specification limits under all load rejection. (b reactor pressure e<ceeds ;
simulated fau!t conditions. high dome pressure inp set point, or generat rI d rejecdon (c) reactor level drops below Level 3. (c) !ow reactor dome pressure p (d) low vessel level O O O
ABWR oosign occwnent 2.2.9 Automatic Power Regulator System Design Description The Automatic Power llegulatc r ( APit) system is dawilied as a power generation system and is not icquired for safet). Safety events requiring control iod scram me sensed and conuolled by the salety-related scactor protection systein (ItPS), which is completely independent of the APit. The APR system contiols reactor ponci dring scarto stanup. power generation, and reactor shutdown by appu priate commands to change iod positions, or to change scarto iccirculation flow. The APR system also contiols the pr essure setpoint or tur bine bypass valve position during s cactor heatup and de-piessuriation (e.g. to control the scactor cool down rate). The automatic lower segulator system consists of redundant piocess controllers. Automatic power regulation is achieved by appr opriate contrei algorithms for dif feient phases of the scactor operation which include approach to criticality, heatup, reactor power increase, automatic load following, reactor power decrease, and icactor de-pressurization and cool down. The automatic power segulator system receives input f rom the plant process computer, power generation control system, the steam bypass and piessuie control system, and the opemtor's control console. The output demand signals f rom automatic power regulator system m e i sent to rod control and infonnation system to position the control rods, to the recirculation flow control system to change reactor coolant recirculation flow, and to the steam bypass and pressur e control system for automatic load following ope ations. The power genemtion system performs the ovendl plant startup, power operation, and shutdown functions. The automat.c poner regulation system performs only those functions associated with reactor power changes and with reactor pressure controller setpoint (or turbint hvpass valve position) changes during reactor heatup or de-pressurization. A simplified fenctional block diagnun of the automatic power regulation system is provided in Figure 2.2.9. The automatic power regulation system control functional logic is petionned by redundant, microprocessor-hased fault-tolemnt digital controllers (FrDC). The FFDC performs many functions, it reads a .d validates inputs from the non-essential multiplexing system (NEMS). h performs the specifir power control calculations and processes the pertinent alann and interlock functions, then updates all system outputs to the NEMS. To prevent computational divergence among the redundant processing channels, each channel performs a comparison check ofits calculated results with the other iedundant channels. The internal FTDC a chitecture features redundant multiplexing interfacing units for conununications between the NEMS and the FrDC processing
' chan nels.
2.2.9 cJ1/92
ABWR Design Document 1)aring norinal operation, the autoinatic power s egulation system inter f.u es with the oper ator's ( onsole Iti per f o: In its desired f unctions. The operator 's contnil panel for autoinatic plant startup, power operation, aiul slmidown f um is ns is p.u t of the power generation control system. The power generation conu M systern initiates dernand signals to nuious controllers to cant out the pic-defined contr ol functions. The f unctions awociated with icac toi power (onn ol are per fonned by the automatic powei regulation systein. For scactoi power contiol, the automatic power segulation system contains algoritluns that < an change reactor power hv control rt>d inutions. or hy tractor coolant recirculation flow ( hanges, but not both at the sune time.1)uring automath load following operation, the automatic power regulation system interiac es with the steam bypus azul piessure control system to coordinate main turbine and reactor power changes to accornplish load following. The normal mode of operation for the automatic power regulator system is automatic. If any system or component conditions are abnonnal during execution of the prescribed sequences of operation, the power generation control system will be automatically switched into the manual mo<le and the operator can manipulate conttol rods and recirculation now tinough the normal controls. A f ailure of the autornatic power regulation system will not prevent manual contiols of the reactor, nor willit prevent safe shutdown of the reactor. g The automatic power regulation system digital controllers are powered by redundant uninterruptible non class IE power supplies and sources. No single power failure williesult in the loss of any automatic power regulation systei.- function. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.9 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be used by APR. l l 9 l 2.2.9 2- 6/1/97
)'
{ Table 2.2.9: Automatic Power Regulator System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria ! 1. This system is powered by redundant 1. Test of loss of power due to single channel 1. There is no loss of APR function by loss of uninterruptible power supplies. power supply failure shall demonstrate no or e channel power supply. loss of APR function. 2 Triplicated, fault tolerant digital contro!!ers 2. Inspect FTDCs and perform validation 2. The fault tolerant digital controllers
- self I with self test and diagnostic capabilities testing. test and on-line diagnostic test features are shall be used. capable of icientifying and isolating failures t of input signah,1/O cards. buses, power supplies, processors and inter-processors ,
communicath a. -
- 3. The APR design provides automatic power 3. Preoperational tests wili be conducted to 3. The APR controls power automaticat;y ,
a control in different modes of operation confirm the automatic power regulation during various modes of operation. l capability in different inodes of operation. ' O f i f t I s r R 6 m
_ . _ _ - _ - _ - _ _ - - _ . _ . - _ . . = _ _ _ . _ _ . - - . _ . .. - _ _. - ABWR oesign occument l>uring normaloperation, the automatic power regulation w t the ulwrator's (onv >le to perh n m its desired f uncti s em inte fa< es with panel [of autolnatic plant AlaMup, power opef ation, onsis atu system initiates demarul signals to various y out the pte-o denned control functions. The functions associate ale pellollned hv the autolnatic powet legulationoisystem power connol fot I cactof pohet control, the automatic powei segulation system contains al gorithms that < an change tractos power hv conttol 1od luotions, or by teact or coolattt following opemtion, the automatic . autorna tic powei load 5 team bypass and pressure conti01 systein to c n eda < es with the reactor power changes to act omplish load followingouldinate m . The norntal automatic. If any system mode or c of operation for the automatic or sysicm is power reg execution of the prescribed sequences of operation, generation control .ystern will he automatically switched into the manual m o e and the operator can ruanipulate contial rods and recirculation rough the flow th nonnal controls. A f ailure of the automatic power regulation pievent manual contiols of the reactor, nor willit prevent reactor. will not te gsaf i The automatic power tegulation system digital controllers are powered by redundant uninterniptible non< lass 1E power supplies and sou ces. No single power function. failure will iesult in the loss of any automatic powe Inspections, Tests, Analyses and Acceptance Criteria Tat te 2.2.9 provides a definition of the inspections , or anah ses, tests and/ together with associated acceptance criteria whh h , will be used by APR. O' 2.2.9 2 1 r,'i m
ABWR oesign oocument q 2.2.10 Steam Bypass and Pressure Control System V . . Design Description The Steam Itypass and Pressure Control (Sit &PC) System is a non-safety selated system. It is a control system only, and c onsists of tluce redutulant f ault tolerant digital controllers (FI DCs) foi conuol algotilluns arul logic along with italitators and alanns for operator infot mation aint the non-safety-iciated powc supplies to power each FTDC. Itecause of the ssstein's triple redutulancy, it is possible to lose one logic channel without itupa< ting the system f unctions In addition, each FTDCis ec}uipped with self test aiul on-line diagnostic capabilities for identifying and isolating, falhu e of input / output signals, buses, power supplies, processors and interprocessor conununications. These on-line tests and diagnostics can be perfonned without inten ulning the nonnal control operation of the Sl\&PC Systern. The Sil&lT.Svstem occives input signah froin other systems and sensors as shown in Figure 2.2.10 and as follows: (1) Steam b>1m valve pcnition switches (2) Steem bypass valve servo conent senson s (3) TCS turbine trip sensors (4) TCS power / load unbalance relay operation (5) Turbine flypass System (Tits) hydnmlic power supply nouble sensors (6) Nuclear lloller System (Nils) Main Steam Isolation Valve (MSIV) position switches (7) Nits narrow and wide range dorne pressure transmitters (8) Steam Extraction System main condenser low vacuum sensors (9) Operator manual commands and manual swit( h positions The Sit &PC system prcnides output signals to: (1) Turbine Control System (TCS) i (2) Automatic Power Regulation (APR) System (3) Recirculation Flow Control System ' (4) Various related control room indicators and alarms , I (5) Pr ocess computer l l 2.2.1 C -1 G/1/92
ABWR oesign oocument The piiinaiy function of the Inenuie (onnol ponion of'the Sit &PC Systein is to efficiently control the scactoi systein picuure during plant staitupi 'midown, lxmen generation, and load following modes of plant operation. tinough control of tut hine control and/or ste.uu hypaw vahes. The system maintains plant stability duiing Inentue setpoint changet The Sil&PC System also has sescral sei ondai3 f unctions used during non-einengency situations and plant uansients, none of which are safety iciated. Additional scactor systern piewuie (ontiolluiu tions aic prosided by othei systerns when the MSIW are closed. The function of the steani bypass poition of the Sit &PC System is to control steam pressure by sending sicain dh et tly to the inain rotulensei whenever icactor steam production exceeds inabi to bine strain flow deinatul. The systein provides tiunsfer capalaility hetweed steann I)ypass valves and tuihiite control udves, and can acconunodate load ujection. Inspections, l'ests, Analyses and Acceptance Criteria ' Table 2.2.10 provides a definition of the inspections, tests, and/or analyses, together with associated acceptanc e criteria 3 hic h will be under taken f or the Sil&PC System. O 2.2.10 2 6/1/92
o O s Table 2.2.10: Steam Sypass and Pressure Control System i Es Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Acceptance Cdteria i l' T. Each FTDC is equipped with self-test and 1. Perform a on-line self-test with complete 1. The results of the self-test confirms system on-line diagnostic capabilities for diagnostics based on the parameters operation. identifying and isolating failure of input / shown in the design description I output signals, buses, power supplies, (Section 2.2.10). processors and interprncessor communications. These on-line tests and diagnostics can be performed without interrupting the normal control operation of the SB&PC System.
- 2. The system incorporates redundant control 2. The system shall be tested by simulating 2. The system continues to function during channels. .ai!ure of one operating controller. loss of one operating contro!!er.
- 3. The system is powered by redundant 3. Loss of one power supply shall 3. There is no loss of SB&PC functions by loss 4 uninterruptable power supplies. demonstrate no foss of functions of SB&PC of any one power supply.
system. i i E a
g N i 5 I I POWER SUPPLY p _ _ _ _ _ _ _. _, y f40 CLEAR BO!LER l 1f p _AUTOMATIC g _ _ _ _POWEP _ _ _,l g l R l 1_ _ _ SYSTEM _____; V +1_ _EGULATIOrJ______; SYSTEM r r r p _ _ EXTRACTIOfJ
. STEAM _ _ _ _ _ ,! 'I p RECiRCUI
--___ NJ _ _ ,!
FLOW SYSTEM l
> > > q j CC 4TR A SYSTEM l
_ _ _ _ _ _ _ _ _ . _-______a I _ _ _ _ _ _ _ ,i m m i m F r------ PROCESS I I l OPERATOR INTERFACE L_ COMPUTER l l l ---_____J l_ _ _COMMATJDS _____3 A A Y Y i I- - - - - - - - ,I g POWER GETJERATIOt1 r-----Ir----I 'I F-------'I COr: TROL ROOu l TURB:NE l! TURB;NE l l -~~----
# 3!
J l_ _COf4 TROL SYSTEMS ______; 1 COnTROt BYPASS l
; l SYSTEM lll SYSTEM g j t_____s t_____m f
- m )
Figure 2.2.10 Steam Bypgnd Pressure Control System g
ABWR Design Document p) 2.2.11 Process Computer System s U Design Description The Process Compt.ter Spiem (PCS) is a non-safety-related system. Its purpose is to promote efficient plant operation by: (1) performing the functbs and calculations neces.ary for the evaluation of plant operation; (2) providing a p rmanent historical record for plant operating activities and abnormal events; (3) prosiding analysis, evaluation and recommendation capabilities for start-up, normal operation, safe plant shutdown and abnormal operating and emergency conditions; (4) providing the ability to directly control certain non-safety-related plant equipment through on-screen technology. All division to division and safety to non-safety interfacing circuits are made up of fiber optic cables, which act as optical isolators for electrical separation. All power to the PCS is supplied by a non-safety related redundant, uninterruptible power supply. No single power falhue will cause the loss of any PCC function. The PCS has self-checking provisions. It performs diagnostic checks to
*tennine the operability of certain portions of the system hardware and performs internal programming checks to verify that input signals and selected pror mputations are either within specific limits or within reasonable boue14 The PCS is composed of two subsystems; the Performance Monitoring anci Control System (PMCS) and the Power Generation Control System (PGCS).
Neither of which serve a safety function. The PMCS znd the PGCS are functions of the PCS, implemented by various programmed routines. l Performance Monitoring and Control System The PMCS is a set of software routines for the PCS Input / Output Modules and j various CPUs to supply various functions and calculations. The basic input types include but are not limited to the following: (1) Various analog pressure signals from sensors on or in the Vessel, the l drywell,ind'idual equipment and the various plant buildings.
/m\
l U (2) Various analog tempermure signals from sensors on or in the Vessel, the l l drywell, individual eqmpment and the various plant buildings. 2.2.11 1- 6/1/92 l l
l ABWR Design Document (3) Various analog coolant and neam flow signals f rom sensors on or in the variou ptunps and pipes tinoughout the plant. (4) Wrious digital "On/Off' and "Open/ closed" signals from various switches and valve controllers throug' out the plant. (5) Various operator requests input through the nuious consoles. The basic output types include but are not limited to the following: (1) Plant Operating Conditions (2) Pacess Trends (3) Alarms (4) Results of Performance Calculations (5) Operator Requests (6) Switchyard Operating Conditions The types of calculations performed include but are not limited to the following: (1) Reactor core performance calculation O (2) Plant performance calculation The function types performed in addition to the calculations include but are not limited to the following: (1) Data Accumulation (2) Indication of Conuol Rod Position (3) Survei"ance test guide Power Gsmration Control System The PGCS is a function of the PCS. It is a software routine in which one or more CPUs act as a top level controller. It contains the algorithms for the automated control sequences associated with plant start-up, shutdown, and normal power generation. It receives the same type inputs as described in 2.2.11.1 and issues control commands and adjusts set-points of subloop controllers to support that automa: ion. The automation process is divided into phases corresponding to g plant start-ap, shutdown, and normal power generation. Each phase is then W divided into several break-points, or logical steps in plant operation. Automation proceeds under PCS control unni the end of a break-point division is reached, 2.2.11 6/1/92
ABWR Des gn Document p at which time, the operator must actuate a break-point switch before automation , can continue. Inspections, Tests Analyses and Acceptance Criteria Table 2.2.11 provides a definition of the inspections, tests, and/or analvses together with associated acceptance criteria which will be undertaken foi the l'CS. O t 1
%)
l l l l 1 l l I g )
%./
l 2.2.11 -3 G,'1/92
. ~ _ _
-l 1
y . Table 2.2.11: Process Computer System - Inspections, Tests,' Analyses and Acceptance Criteria'- Certified Design Commitment int.pections, Test, Analysis Acceptance Criteria
- 1. The PCS has self - checking provisions. It 1. Perform the self-checking test with ' 1. The results of the self-checking test'
. performs diagnostic checks to determine' i complete diagnostics, of the functions confirms satisfactory system operation.
the operability of certain portions of the listed in Subsection 2.2.11 using the
. system hardware and performs internal' manufacturer's operations and programming checks to verify that input maintenance manual. '
signals and selected program ' i computations are either within' specific ! limits or within reasonaSte bounds.
- 2. No single power failure will cause the loss 2. The system shall be tested by rimulating 2. The system continues to functinn during ; :
of any PCS function. failure of one of the system power . loss of one power supply. supplies.
- 3. In the event of the PCS going off-line or 3. The system shall be tested by simulating 3. Testing results conform to plant response L l 6 other troublet <cith the sy:, tem, PGCS is loss of the PCS. and stability requirements when the easily separaad from the control circuits systems are manually contro!Ied with the ,
and the plant will be safely controlled by system subloop controller. , i the subloop controllers. D r E j e e e 'i
l l ABWR Design cocwnent i eq 2.2.12 Refueling Platform Control Computer The Ref ueling I'latf orm will be computer controlled f r om the operator station on the relueling floor. Design Description-Control Computer The computer will provide X,Y and Z location of the ref ucling mast. .\tast and platform will be controlled by various limits on theb f unctions atul inovenients. Inspections, Tests, Analyses and Acceptance Criteria No entries for this system. O O 2.2.12 6/1/92
ABWR oesign Document f-- 2.2.13 CRD Removal Machine Control Computer NJ No Tier 1 entry for this system. l (3.J l 1 l 1~ 2.2.13 1- 6/1/92
ABWR Design Document 2.3 Radiation Monitoring w 2.3.1 P'rocess Radiation Monitoring (PRM) System Design Description The primary function of this system is to (a) monitor and record the various gaseous and liquid process streams and ellloent icleases (b) initiate alarms in the MCR to warn operating personnel to the high radiation activity, and (c) initiate the appropriate safety actions an:1 controls to prevent further radioactivity releases to the ensironment. This system provides both safety and non-safety instrumentation fcr radiological monitoring, sampling and analysis of identified piocess and effluents streams throughout the plant. The system monitors the radiation levels during normal, abnormal and accident plant conditions. The stack vent discharge and the standby gas treatment system (SGTS) are both equipped uith high range detectors for post accident monitoring oflevels up to 10 uc/cc. The process and effluent paths and/or areas as described herein are monitored for potential high radioacthity releases. The monitoring channch ofitems 1 through 4 below are provided for safety as Class IE instrumentation, while the O V rest of the process radiation instrumentation is considered non-essential which is provided to monitor plant operations. (1) Main steam line (MSL) tunnel area - 4 divisional channels The MSL tunnel area is continuously monitored for high gross ganuna radioactivity in the steam llow to the turbine. Reactor scram, MSIV closure, and main condeaser vacuum pump shutdown are automatically initiated on any two out of four channel trip. (2) Reactor Building ventilation exhaust - 4 divisional channels The air vent exhaust from the secondary containment is continuously monitored for gross gamma radioactivity. On high level, the standby gas treatment system is activated and the containment ventilation ducts are isolated on any two out of four channel trip. (3) Fuel handling area ventilation exhaust - 4 divisional channels The air vent exhaust from the fuel handling area is continuously monitored for gross gamma radioactisity. On high level, the standby gas p treatment system is activated and the fuel handling area ventilation ( ducts are isolated on any two out of four channel trip. 2.3 6/1S2
ABWR oesign Document (4) Control Building air intake supply - 4 dhisional ch.umels The air intake to the Connol Building is continuousiv monitored for gross gamma radioactivity. On high level, the ventilation ducts are isolated and the emergency air circulation system is activated on any two out of four channel trip. (5) Turbine Building ventilation exhaust - 4 channel The vent exhaust from the Turbine building is continuoudy monitored for gross ganuna radioactivity. The air exhaust from the ecluipment comparunent area and from the clean areas in the Tm hine Building aie each monitored by two redundant channels. Alarms are initiated on high radiation levels. (6) Charcoal vault ventilation exhaust - I channel The vent exhaust from the charcoal vault is continuously monitored for gross gamma radioactivity that may result from cracks in the activated charcoal beds. An alarm is initiated on high radiation. (7) Pre-treated main condenser off-gases - I channel The pre-treated main condenser oft-gases are continuously sampled O and monitored for gross gamma radioactivity. Alarms are initiated on , high mdiation and on abnormal sampling flow. Vial sampling is l provided for periodic isotopic analysis. (8) Post treated main condenser off-gases - 2 channels t l The treated oft-gases are continuously sampled and monitored for airborne mdioactivity by two gas 3:unplers and filters for collecting air particulates and halogens. Each gas sampler consists of a beta / gamma sensitive detector and a source check for periodic testing. On high radiation, the off-gases are routed through the entire charcoal bed for hold-up. On extremely high radiation, the off-gas discharge to the stack is isolated. Alarms are initiated on high mdiation levels and on l abnormal sampling flow. Vial sampling is provided for periodic isotopic l analysis. (9) Plant vent discharge - 2 channels The discharge through the stack is continuously sampled through an isokinetic probe and monitored for airborne radioactivity by two j redundant channels, each consists of a beta / gamma sensitive detector with a source check, a high-range ion chamber, and filters for collecting 2.3.1 2- 6/1/92
ABWR Design Document (p,/ air particulates and halogens. Sampling and collecting of nititun is also prosided. Alarms;u e initiated on high radiation lesch and on abnonaal sampling How. (10) Radwaste linilding ventilation exhaust - I channel The air vent exhaust f rom the Rachvaste lluilding is continuousl> sampled through an isokinetic probe and monito cd im airborne mdioactisity by a beta /ganuna sensitive detector with a source check and filters for collecting air particulates and iodine. A nitium monitor is also prosided for sample collection. Alarnu are initiated on high mdiation and on abnormal sampling How. (11) Radwaste liquid discharge - I channel The liquid waste discharge from the plant is continuousiv sampled and monitored by a liquid sampler,which consists of a scintillation detector, a source check and an ultra sonic cleaner. Alarms ine initiated on high mdiation levels and on abnormal sampling flow. On high radiation in the discharged waste, the flow to the environment is automatically terminated and alarmed. i (12) Drywell sump liquid discharge - 2 channels, one per sump The liquid discharge from each of the two drywell sumps is monitored i by an in-line ion chamber. On high mdiation, the discharge to the Radwaste Building is terminated and alarmed. (13) Standby gas treatment system (SGTS) discharge - 4 channels l The discharge from the SGTS to the stack is continuously sampled and monitored for airborne radioactivity by two gas chambers that are in series whh the tiow and by sampling filters for collecting air particulates i at.d halogens. Each gas sampler consists of a scintillation detector and a source check. A!so, radioactivity in the discharged gases are j continuously monitored for gamma radiation by two in-line high-mnge ion chambers. Alarms are initiated on high radiation levels.
- (14) Turbine gland steam condenser discharge - I channel The discharge from the main turbine gland steam condenser is continuously sampled and monitored for airborne r adioactivity by a gas chamber and by sampling filters for collecting air particulates and
( halogens. The gas sampler consists of a scintillation detector and a l source check for periodic testing of the detector. Vial sampling is 2.3.1 6/1/92
ABWR Design Document inovided for laboratory analvsis. Alarms are initiated on high radiation 1,.vels. (15) Intersystem radiation leakage - 3 channels, one per RCW system loop in'ersystem leakage into each loop of the reactor building closed cooling water sy> em is monitor ed bv an in-line sv ', nhtion detector for gioss ganuna s ar'.toactisity. An alann iv initiated cn. high radiation. Incation of the process radiation monitors is shown in the plant layout drawing of Figme 2.3,1 The radiation detectors are numbered according to the listing piovided above, inspections, Tests, Analyses and Acceptance Criteria Table 2.3.1 provides a definition of the inspections, tests and/or analyses together with the associated acceptance criteria which will be undertaken for the Process Radiation Monitoring System. O O 2.3.1 4- 6/1/92
O O O { Table 2.3.1: Process Radiation Monitoring (PRM) System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment . Inspections, Tests, Analyses Acceptance Criteria
- 1. The PRM is designed to continuously 1. Each detector shall be checked for 1. Proper detector calibration and sensitivity monitor the radiation levels in process and sensitivity and calibration based on are verified based on acceptable records effluent liquid and gaseous streams certified records and/or response tests and/or test results. Operational readiness throughout the plant. using either a portable gamma source or of each radiation monitor is verified by the the detector check source as required. Each monitor self test circuitry.
radiation monitor shall be visually checked for operational readiness.
- 2. The PRM is designed to initiate 2. The range of each radiation channel shall 2. Verification that the correct . onitor automatically the controls and safety be checked for the correct response using response is indicated at the acified actions as required to isolate and prevent sufficient simulated inputs. Also, the trip inputs for each channel. Also, confirmation ,
further releases of radioactivity. levels that initiate the safety actions and that the trips are initiated at the proper plant controls shall be verified. setpoints.
- 3. Each process radiation monitors is 3. The alarm setpoints of each radiation 3. Confirmation that alarm initiation occurs at designed to initiate alarms on high and low channel shall be verified using the the proper setpoint and when the monitor >
radiation levels and when the monitor adjustable trip output circuits and the INOP indicates gross failure. indicates gross failure (INOP trip). trip feature of the monitor.
- 4. The PRM is used to monitor radiation 4. Verify that the high range monitors of the 4. Verificction that each high range levels during normal, abnormal and plant vent discharge and the SGTS can monitoring channel including the accident plant conditions. detect gaseous effluents of levels up to 105 associated radiation monitor is capable of uc/cc. satisfying this requirement.
- 5. The PRM samples and monitor effluents 5. Verify that the sampling racks and 5. Operation of the sample racks is verified for noble gases and for collecting air associated equipment are operating within when the extracted air flow is normal and particulates, halogens and vial samples. specified limits to assure the extrac; ion of is within acceptable limits, valid and representative samples.
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- s 6. For the safety related functions, the PRM 6. Each required safety function shall be 6. Acceptance is based on satisfying the two provides 4 redundant divisional channels tested using various simulated signal out of four criteria for initiating the to initiate the required protective action on inputs to verify that the initiation of the required functions.
two out of four channel trip. protective action occurs only when any two out of four channels indicate trip.
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ABWR Design Document o 2.3.2 Area Radiation Monitoring System
! 1 Design Description The primary function of the Aca Radiation hionitoring (ARht) System is to monitor continuously gamma radiation leveh at various locations within the plant buddings, md to prWde em1:, warning to phnt [>crsannel when high radiation levels .ue detected so that appropriate actions can be taken to reduce ftuther exposure to radiation.
The AIG1 System comists ofloi al area indiation detectors, digita; radiation
.aonitors, and locai auxiliary units with audible alanus installed in selected key areas. Each inctntm, nted channel provide < trips on high radiation level, lack of detector response, and on gross failaic of the monitor. Ala ..s are actiwted on abnormalindicatiore:in the main control room (h1CR)as well as in the local areas where the >onic alarms are provided.
The gamma radiation levelis continuously monitored and recoided in each at ea where installed. An increase :n background radiation level is normally attribut.tble to either operational transients, maintenance activities, or to madvemat release of radioactivity. OV The instrumentation channels of the area radiation monitoring system are designed to detect exposure rates from 10# mR/hr to 10 R/hr during reactor l operation and during abnormal and accident conditions. The measuring range i and sensitivity of each channel are based on the expected background radiation level at the location where the detector is installed. The alarm trip setpoint for each channel is adjustable and will be based on the actual background radiation level measured at the detector location. i The system is classified as non-safety related. Power to the radiation monitors is l provided from the 125VAC non-essential vital source, which is available during loss of oft-site power. The system design is configured as shown in Figure 2.3.2. Inspections, Tests, Analyses and Acceptance Criteria l l Table 2.3.2 provides definition of the inspections, tests, and/or analysis together with associated acceptance criteria which will be undertaken for the ARN1 System.
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v l 2.3.2 6/1/92
{ Table 2.3.2a: D21 Area Radiation Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The instrumented area radiation 1. Visually inspect and verify that the 1. Verification that eat.h channel monitoring system channels are designed equipment f;r each instrumented channel configuration is in conformance with the to measure and record the gamma is properly configured, installed and required design. Also, that operational radiation levels at various locations in each functional. readiness of each channel is verified when of the plant buildings. indicated on the monitor.
- 2. Each channel measures and displays the 2. Each channel shall be tested across its 2. Successful channel operation wi!i be gamma dose rate across its monitoring monitoring range to verify response, demonstrated when the radiation monitor range and provides dosage level sensitivity, and calibration by using a indicates proper response and displays the indications in the main control room. portable gamma source traceable to the measurement within the required accuracy.
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- 3. Each channel activates alarms in the MCR 3. The alarm trip setpoints of each channel 3. Initiation of the appropriate alarms is and at local areas on indication of high shall be tested and validated. Initiation of confirmed at the required trip set;;oints.
@ radiation, lack of detector response, and the appropriate alarms in the MCR and in inoperative radiation monitor. local areas shall be verified. R = b O O O
ABWR oesign occument
'v' TYPICAL CH ANNEL TYPICAL CHANNEL
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U Figure 2.3.2 Area Radiation Monitoring System 2.3.2 -3 6/1/92
l l ABWR Design occument p 2.3.3 Dust Radiation Monitoring System k Not an AllWR system No entry. l O l 1 l l l l O 2.3.3 1 C/1/92
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_ABWR Design Document q 2.3.4 Containment Atmospheric Monitoring System b Design Description The primarv function of the Containment Annospheric Monito ing (G.ul) Sysicm is to monitor the atmosphere in the prim;uy containment for exc essive gamma radiation levels and foi high com entration of oxygen and hydrogen h sels during normal reactor operations and under post-accident conditions. The C.ul System is classified as a safety system. seismic Category 1, and piovides no control f unction. The safety function of the CAM System is to identify if a potentially explosive mixture of hydrogen and oxygen is building up in the primary containment during post-accident monitoring. and piovide concentration measmements to the operator for use in flanunability conuol. Also, the use of gamma monitor s nith high-range are piodded for post-accident monitoring. The CAM System consists of two independent but iedundant divisional subsystems (i and 11), which ate electrically and physically separated (Figure 2.3A). Each C.01 division provides measurement of the total gamma-ray dose rate and of the concentration of hydrogen and oxygen it vels in the drywell and/ or the suppression chamber during normal plant operation and following a LOCA event. The operation of each CAM Subsystem can be activated manually by the l operator during reactor operations, or it will be automatically activated by the LOCA signal, either on high dgwell pressure or on low teactor water level. In either mode, sampling is selected for the designated area. Two high-range radiation monitoring channels are provided per division, one for monitoring the radiation levelin the drywell and the other for monitoring the radiation level in the suppression chamber. Each channel provides continuous dosage rate measurements for display and recording in the control room. Alarms are activated on high radiation levels and when the snonitors fail and become inoperative. Each monitor has a measurement and display range of 1 to 10'R/hr. Each divisional hydrogen / oxygen monitoring channel conststs of a gas sa m pling rack used to extract samples of the atmosphere in the drywell (DW) or the suppression chamber (SC) and feeds the sample to a local gas analyzer for measurement and display in the control room. Alanns are activated on high gas content levels and for abnormal flow sampling. Each gas sampling rack is povided with gas calibration sources to verify operability of the individual gas monitors and for periodic calibration. Each hydiogen and oxygen monitor is capable of measuring gas contents up to 30% c,f solume and displays digitally the readout. 2.3.4 6/1/92
ABWR oesign oocument Power to each CAM subsystem is provided f rom the unintenuptable Class 1 E 120 VAC vital divisional sotuce. Inspections, Tests, Analyses and Acceptance Criteria Table 2.3.4 providt s a definition of the inspections, tests. and/or analysis, together with associated acceptance criteria which will be undertalen f or the Containment Atinospheric Monitoring Systern. O i r 9 2.3 1 6/1/92 l l
O O O
} Table 2.3.4: Containment Atmospheric Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria -
- 1. Each CAM subsystem is designed to be 1. Manually activate each subsystem and 1. Equipment readiness will be verified when operated manually. Air sampling and verify operational readiness of the each radiation morsitor and each sampling ra6ation monitoring are performed to radiation monitors and the sampling rack correctly indicates no failure and are check and analyze the air in the primary equipment. Select an air sample from the ready for operation. Correct valve containment for high levels of gas DW and SC, verify alignment of the sample alignment and normal air sampling will be concentrations and radiation. lines control valves, and eneck for normal indicated by the instrumentation.
air flow.
- 2. Each CAM subsystem is designed to be 2. In the auto mode, use simulated LOCA 2. Equipment readiness will be verified when activated automatically by a LOCA signal signals to 'aitiate operation of each CAM the same conditions stipulated under item for post-accident rnonitoring of the same subsystem and verify the sampling and #1 above are satisfied.
parameters as identified under item #1 monitoring operations for the conditions above. stipulated in item #1 above, a 3. Each radiation channel monitors and 3. Each channel shall be tested to verify 3. Successful channel operation will be displays the gamma dosage rate in the channel response at:d measurement by verified when each monitor provides the MIR in R/hr, and activates alarms on hinh using a portable gamma radiation source. required response and displays the sensed radiation levels or when the monitor fails. Tests shall be performed at least one point radiation level and initiates the appropriate at low end of the monitor range to verify alarms. , channel response and sensitivity. Perform l trip tests to validate the setpoints. ' Each CAM System gas sampler extracts an 4. Each hydrogen monitor shall be tested at 4 Monitor operability will be verified when ! air sample from the DW or the SC, analyzes least two known H2 concentration levels the response and display are compatible the hydrogen contents and displays the from 1 to 5 percent content using a with the tested gas levels. Confirmation measurement in the MCR in percent hydrogen gas calibrated source. The that the MCR alarms are initiated. , volume. High gas levels and abnormal channel response and readout shall be sampling w s" be alarmed in the MCR. verified. Perform trip tests for setpoint
- verification.
i
- 5. Each CAM System gas sampler extracts an 5. Each oxygen monitor shall be tested at 5. Monitor operability will be verified when air sample from the DW or the SC, analyzes least one known 02concentratios. level the response and display are compatible
, the contents for oxygen and displays the from 1 to 5 percent content using an with the tested gas levels. Confirmation j Q results in the 7 'CR in percent vo!ume. High oxygen gas calibration source.The channel that the MCR alarms are initiated.
O gas levels and abnormal sampling will be response and readout shall be verified. alarmed. Perform trip tests to verify the setpoints.
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NOTE: 3! THE SYSTEM IS AUTOVATICALLY 16 i INITIATED UPON RECEIPT OF A N I LOCA SIGNAL I Figure 2.3.4 Containment ospheric Monitoring System g .
ABWR Design Document n 2.4 CORE COOLING V 2.4.1 Residual Heat Removal System Design Description The Residual Hea; Removal (RilR) System is comprised of three divisionally separate subsystems that perf orm a variety of functions utilizing the following six basic modes of operation: (1) shutdown cooling, (2) suppression pool cooling, (3) wetwell and dowell spray cooling. (4) low pr essur e core flooder (LPFL), (5) fuel pool cooling, and (G) AC independent water addition.The configuration of each loop is shown on its P&lD in Figure 2.4.1 (aligned in the standby mode). The major functions of the various modes ofoperation include: (1) containment heat removal, (2) reactor decay h-at removal, (3) emergency reactor vessel level makeup and (4) augmented fuel, al coniing. In line with its given functions, portions of the system are a part of the ECCS r.etwork ami the containment cooling system. Additionally, portions of the RUR System are considered a part of the Reactor Coolant Pressure lloundag- (RCPB). The entire RHR System is designed to safety-related standards, although it perfonns some non-safety functions (i.e., those that are not taken credit for when evaluating design basis accidents). The safety-related nuxles of operation inchide: (1) low pressure flooding, (2) suppression pool cooling, (3) wetwell l spray cooling and (4) shutdown cooling. Non-safety-related modes of operation l include: (1)drywell spray ccoling, (2) AC independent water addition and (3) augmented fuel pool cooling. The RHR System also provides a backup, safety-related fuel poo! makeup capability. Ancillary modes of operation include minimum flow bypaso and full flow testing. The ECCS function of the RHR System is performed by the LPFL mode. Following receipt of a LOCA signal ( low reactor water level or high dowell pressure ), the RHR Synem automatically initiates and operates in the LPFL mode (in conjunction with the remainder of the ECCS network) to provide emergency makeup to the reactor vessel in order to keep the reactor core cooled l such that the criteria of 10 CFR 50.46 are met. The LPFL mode is accomplished by all three loops of the RHR System by transferring water from the suppression
- pool to the RPV, via the RHR heat exchangers. Although the LPFL mode is automatically initiated , it may also be initiated manually. The system will also l automatically revert to the LPFL mode of operation from any other test or operating mode upon receipt of a LOCA signal. Ench RHR loop's RPV injection mlve requires a low reactor pressure permissive signal whether being opened manually or automatically in response to a LOCA signal.
A Q The containment heat removal function in the AlnVR is performed by the Containment Cooling System, which is comprised of the low pressure core flooder (LPFL), suppression poc>l cooling, and wetwell and drywell spray cooling 2.4 6/1/92
ABWR Design Document modes of the RHR System. Following a LOCA, the energy present within the reactor primary sptem is dumped either directly to the suppression pool sia the SRVs, or indirectiv via the dmvell and connecting vents. Subsequently. fission product decay heat continues to add energy to the pool. The Contaimnent Coohng System is designed to limit the long-term bulk tempemture of the suppression pool, and thus limit the long-term peak temperatures and pressures within the wetwell and divwell regions of the containment to within their analyzed design limits with only two of the three loops in operation (i.e., worse case single failure). The cooling icquirements of the containment cooling function establish the necessan RHR heat exchanger heat removal capacity. The LPFL mode,in addition to its primary function of cooling the core, senes to cool the containment, as the heat exchanger is designed to ahays be in the loop. The dedicated suppression pool cooling mode is made availab.e in each of the three loops of the RHR System by circulating suppression pool water through the respective RHR heat exchanger and then directly back to the suppression pool. This mode of RHR is usually initiated manually but will also initiate automatically in response to high suppression pool temperature. The wetwell and dgwell spray modes of RHR are each available in only two of the three subsystems (loops B and C), Theser unctions are performed by drawing water from the suppression pool and delivering it to a common wetwell spray header and/or a common dnwell spray header both via the associated RHR heat exchanger (s) These containment spray modes of the RHR System are typically initiated manually, with the exception of automatic initiation of wetwell spray coincident with automatic suppression pool cooling. However, the dnwell spray inlet valves can only be opened if there exists high dowell pressure and if the RPV injection valves are fully closed. Wetwell and dowell sprays serve as an augmented method of containment cooling. Wetwell spray also senes to mitigate the consequences of steam bypassing the suppression pool. The normal operational mode of the RHR System is in the shutdown cooling mode of operation, which is used to remove decay heat from the reactor core. This mode provides the required safety related capability needed to achieve and maintain a cold shutdown condition, including consideration of the worst case system single failure. The RHR heat exchanger heat removal capacity requirements in this mode are bounded by containment cooling requirements. Shutdown cooling is initiated manually once the RPV has been depressurized below the system low pressure permissive. In this mode each loop takes suction from the RPV via its dedicated suction line, pumps the water through its respective heat exchanger, and returns the cooled water to the 'RPV. Two loops (B and C) discharge water back to the RPV via dedicated spargers, while the third loop (A) utilizes the vessel spargers of one of the two feedw;uer lines (FW-A).The heat renmved in the RHR heat exchangers is transported to the ultimate heat sink via the respective division of reactor cooling water and service water. Each shutdown cooling suction valve is interlocked with that loop's suppression 2.4.1 2- 6/1/92
ABWR oesign Document pool suction and discharge valves and wetwell sprav valve to prevent draining of ( the s cactor vessel to the suppression pool. Also, each slundown cooling suction salve is interlocked with. and automatically closes on. Iow reacion watei level, The augmented fuel pool cooling mode of the RHR System supplements / icplaces the normal fuel pool cooling system during inhequent conditions of high heat load. This mode is accomplished manuallv in one of two wap. When the scactor vessel head is temoved, the cavity 0ooded and the f oc! pool gates are removed, the RHR Spiem cook the fuel pool in the nonnal shutdown cooling mode. When the fuel pool is othensisc isolated Irom the reactor c nin, two loops (11 and C) of the RHR System can directly cool the pool hv taking suction hom and discharging back to the normal fuel pool cooling svstem. This cc.nnection also provides for emergency fuel pool makeup capability by supphing a safety-related makeup path to the fuel pool from a safety-iclated somcc (i.e., the suppression pool). One loop (C) of the RHR System also functions in an AC independem water addition mode. This mode provides a means of cross connecting the reactor building Dre protection system header to the RHR Systemjust outside the containment in the absence of the nonnal ECCS netwoiL and independent of the normal essential AC power distribution network. The connection is C accomplished by manually opening two in-series valves on the cross <onnection ( pipingjust upstream ofits tie-in to the normal RHR piping. Fire protection system water can be directed to either the RPV or the drywell spray sparger by manual opening of the loop C RHR injection valve or the two loop C dowell spray valves These three valves also nave manual hand whccis. The Grc water is supplied via the system's reactor building distribution header by either the direct diesel 4hiven Ore pump or 6 om an external source utilizing a dedicated connectionjust outside the reactor building. Each loop of the RHR System also has both a minimum flow mode and a full How test mode. The minimum now mode assures that there is pump Dow sufficient to keep the pump cool by opening a minimum flow valve that directs now back to the suppression pool anytime the pump is running and the main discharge valve is closed. Upon sensing that there is adequate flow in the pump main discharge line, the minimum Gow valve is automatically closed. In the full flow test mode, the system is essentially operated in the suppression pool cooling mode, drawing suction from and discharging back to the suppression pool. The RHR System is comprised of three separate loops or subsystems, each of which includes a pump and a heat exchanger, takes suction from either the RPV or the suppression pool, and directs water back to either the RPV or the suppression pool. Two of the three loops can divert a portion of the suppression pool return flow to a common wensell sprav sparger or direct the entire flow to a common drywell spray sparger. The divisional subsystems of the RHR System 2,4.1 6/U92
ABWR oesign Document are separated both mechanically and electrically, as well as being pin sicalh located in diffetent aicas of the plant to address icquiicments penaining to fisc protection and othei separation criteria. Each of the thice subsvstems is powered fiom a separate dhisional powei distribution bus that can be supplied from either an on site or oJ-site source. Cooling water to each dhision of RHR equipment (heat exchanger as well as pump and motor cooler s) is supplied hv the respective dhision of the tractor cooling wates (RCW) System. The RHR System also includes prosisions for containment isolation and R(.PB pressme isolation. The RHR System will maintain the capability to perfonn its intended safety-related nmctions either following a Saf e Snutdown Earthqueke (SSE) or during the environmental conditions imposed by a I.OCA, and in each cas assuming the worst case single faihne. The system will also acconunodate calculated movement and thermal stresses. The system is designed so that the pumps will have necessary head / flow characteristics and available NPSH greater than required NPSH for operating modes. The system can be powered from either normal off-site sources or by the emergency diesel genenuors. The RH R System is Seismic Categoiy I and is housed in the Seismic Categcny I reactor building to provide protection against tornadoes. floods, and other natural phenomena. The RHR pumps are motor-driven centrifugal pumps each capable of supplying at least 4200 gpm at 40 psid (dnwell to RPV). The pumps are ASME Code Class g 2 components with a design pressure of 500 psig and a design tempenuure of l 360 F. The pumps are interlocked from starting without an open suction path. The RHR pumps are protected from possible pump runcut conditions during operation. The RHR heat exchangers are horizontal U-tube /shell type each sized to provide a minimum effective heat removal capacity % coefficient) of l l5 Btu /sec F. The primary and secondan sides of the heat exchangers are ASME Code Class 2 and 3, respectively. The primarv side design temperature and pressure are 500 psig and 360"F, respectively. The seconda1T side design temperature and pressure are consistent with that of the RCW System. Each loop ( of the RHR System has its ownjockey pump to act as a keep-fill system for that loop's pump discharge piping. Thejockey pumps are ASME Code Class 2. l The RHR System piping and valves are ASME Code Class 1 or 2 as shown on the P&lD (Figures 2.4.la, b, c). The design pressure and temperature of piping and valves varies across the system. For that piping attached to the RPV, from the RPV out to and including the outboard containment isohtion valves, the design pressure and temperature are 1250 psig and 575 F, respectivelv. For other piping open to the containment atmosphere, out to and including the outboat d containment isolation valves, the design pressure and temperature are 45 psig and 219 F respectively. For piping and valves outside the containment isolation v;dves, the design pressure and temperature depends on whether it is located on the suction or dischargt side of the main pump Those portions on tbc suction 2.4.1 G/1/92
ABWR oesign Document side air r.ned at 300 psig and 360*F, while those portion < on the discharge side ( ate rated at 500 psig and 3f,0'E respecthcly. The low pressure portions of the shutdown cooling piping are protected f roin tull reactor pressuic by autornati< pressuie isolation nthrs that aie interlocked with icactor pressure. liigh reliabilits of this interlock is assuird by utituing foui separate and divisionalh independent pressure sensors in a Suit-of-4 logic. Additionally,in-series inhoard and outboard containtnent/piessure isolation nihrs in each loop aic powered f roin separate electrical divisions. Relief nihes are also provided foi piotection froin overpressinc. The RilR System includes conuol room indication to allow fe monitor ing .u ni control during design basis operational conditions, i.e., systt m flows, temperatur es and pressures, as well as valve open/close and purnp on/o!T indica ion for those instruinents and components shown on Figures ".4.la, h and c, with the exception of simple check udves and overprescure iclief vahes (of the check valves shown only the testable check udves downstream of each loop's RPV injectioe ndve has control room status indication). Inspections, Tests, Analyses and Acceptance Criteria This section provides a definition of the inspections, tests and/or analyses q together with associated acceptance criteria which will be undertaken for the
'V RH R System.
l i u 1 2A1 6/1/92
{ Table 2.4.1: Residual Heat Removal System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The configuration of the RHR System is 1. Inspections of the as-built RHR 1. Actual RHR System configuration, for shown in Figures 2.4.1a. b and c, which are configuration shall be performed. those components shown, conforms with each mechanically and electrically Figures 2.4.la, b and c and separation separated from each other. requirements.
- 2. The RHR System operates in the LPFL 2. The ECCS LOCA performance analysis for 2. RHR System actuation and operation is mode as part of the overal! ECCS network. assuring core cooling shall be validated by consistent with the ECCS performance RHR System functional testing, including analysis as follows:
demonstration that the LPFL moce (of each RHR loop)is capable of automatically a. RHR Flow (each loop) initiating and operating in response to a . 2 4700 gpm (at 40 psid) LOCA signal. b. Time to Rated Flow (each loop)
..s 3fi sec i 3. The RHR System operates in the 3. The primary containment performance 3. RHR System automatically actuates in the suppress;on pool cooling mode to limit the analysis for long-term peak pressure and suppression pool cooling mode as long-term temperature and pressure of the temperature shall be validated by RHR designed and RHR heat exchanger containment under post-LOCA conditions. System functional testing demonstrating performance is consistent with the the required flowrate through the heat contair. ment cooling system analysis as exchanger and by inspection of vendor test follows:
data demonstrating the heat exchanger's effective heat removal capability. a. Effective heat removal capability of each Automatic initiation in the suppression RHR Heat Exchanger (K coefficient) pool cooling mode will also be includes effects of RCW, RSW and UHS: demonstrated. ..? 195 Btu /sec*F.
- b. Tube side flow of each RHR Heat Exchanger
. 2 4200 gpm
- 4. A portion of the RHR System returr. flow (in 4. RHR System functional tests shall ba 4. RHR loops B and C each separately are loops B & C) can be diverted to the wetwell performed to demonstrate wetwell spray capable of providing wetwe!! spray flow spray header. flow capability. consistent with the suppression pool en bypass analysis as follows:
= w
- a. Wetwell spray flow.(each loop individucily)
. 2 500 gpm.
G 8 9
(vD Q V-g Table 2.4.1: Residual Heat Removal System (Continued) . a inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. The RHR System operate in the shutdown 5. RHR System functional tests shall be 5. RHR System (each loop) is capable of cooling mode to remove reactor core performed to demonstrate operation in the taking suction from and discharging back ,
decay heat and bring the reactor to cold shutdown cooling mode of operation. to the reactor pressure vessel [ Heat , shutdown conditions exchanger heat removal capability in thb mode is bounded by containment cooling requirements - ITAAC # 31
- 6. The RHR System (loops B and C) operates 6. RHR System functional tests shall be 6. RHR System (loops B & C) is capable of in the augmented fuel pool cooling mode performed to demonstrate operation in the taking suction from and discharging back to supply supplemental or replacement augmented fuel pool cooling mode of to the normal fuel pool cooling system. l cooling to the spent fuel storage pool operation. [ Required cooling capability in this mode under abnormal conditions. bounded by containment cooling requirements - ITAAC #3]
a 7. The RHR System (loop C) provides an AC 7. RHR System functional testing shall be 7. Flow capability exists for directing water independent water addition function. performed to demonstrate operation in the from the fire protection system to the RPV AC independent water addition mode of and drywell spra/ sparger, via the RHR operation. System (loop C), without power being available from the essential AC distribution systen The valves are capable of being opened by manual hand wheels.
- 8. The RHR System operates when powa ed 8. RHR System functional tests shall be 8. RHR System is capable of operating when from both normal off-site and emercy >cy performed to demonstrate operation when supplied by either power >urce.
on-site sources. supplied by either normal off-site power or the emergency diesel generator (s).
- 9. If already operating in any other mode, the 9. Using simulated inputs, log'ci and 9. RHR logic functions to automatically RHR System automatically reverts to the functional testing shall be performed to reconfigure the system to the LPFL mode LPFL mode in response to a LOCA signal demonstrate the RHR System's ability to of operation in resconse to a LOCA .,ignal automatically revert to the LPFL mode fr(m any other mode.
- 10. Using simulated inputs, logic and 10. Automatic isolation and interlock features e 10. Pressure isolation valves are provided to
')' protect low pressure RHrl piping from functional testing shall be performed to function upon receipt of input cignals. ;
being subjected to excessively high reactor demonstrate operation of automatic pressure. isolation and interlock functions of pressure isolation valves.
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. sr pa mlo s
T l;p a t la os le t noo f o e mh t mhc t, ce ot s e v ef rp l lv o a r p uis e s it h nlea pd cwv f at is s v e eh no a n mtc r t dei o yp m lyvr paips s yn e oe laat o r o l k t it m u or t t cvc s kc ee s edi v t c oc u k cu m ap it nea gr lagec h s op m s o a or loap j l r en j se C o e t mi t e e inp r u tvin hi t e eh t ae och t a s le sV inp zt n r t ut idn re e o iip l g e do . usf intsP t e i p a go ea h yxR i r n aa ue s e omgn ng vf t a ee D pwi t min el oi la vt oo t r peh r t st pu mko et o ol oe mo l d a t ne s ed t e soin yca mess . h sy )p l h nd ut fi if h R me r Snd r eme let a e wie f pi w Soo d . r so t it sh v r H uv R wt R Rl e R H on ysi r yd H g Hhlle e ulc Ri tn r i C mo Rd e Sl bo t R as cf d n s hinm O t v R a c ee s elly eee eu e a er ca i o E mf r hh T sp r Hts R er e h pr T opf u h a T s t h Tilni n
- 1. 2 3 4. 5 6 1 1 1 1 1 1 g ._ 3' p $~
.l llll' l lllfl! l flI
g Table 2.4.1: Residual Heat Removal Syste '(Continued) a inspections, Tests, Analyses and Acceptance Criteria
- Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 17. The RHR System full flow test mode allows 17. Functional tests will be performed to 17. Each RHR subsystem demonstrates futt periedic demonsteation of RHR capability demonstrate operation in the full flow test flo' ' .snctional capability while duri ig normal power operation. mode. appvoximating actual vessel injectiors conditions during operation in the full flow
, test mode. i
- 18. The RHR pumps have sufficient NPSH 18. Pump vendor records will be irspected and 18. Minimum pump NPSH available, as during postulated operating conditions. as-procured pump NPSH compared with determined based on as-built conditions design basis analysis assu.nptions. Actual and the results of vendor tests and/or system installation will be inspected, and analyses, exceeds as-pror ured pump appropriate measurements taken, to requirements and is consistent with design determine available pump NPSH. basis analyses requirements.
l
- 19. The RHR pumps have adequate head l flow 19. Pump vendor test records and calculations 19. RHR pumps,in as-installed system g characteristics. '
will be inspected, and as-insta"ed system configuration, demonstrate headiffow
~
flow testing conducted, to establish pump characteristics consistent with design f; asis head / flow characteristics. analyses assumptions. l 20. Control room indications are provided for 20 Inspections will be performed to verify 20. The instrurnentation is present in the RHR System parameters defined in Section presence of control room indicatioq for the control room as defined in Gection 2.4.1. 2.4.1. RHR System (Section 2.4.1). l I i e t l 8 i i
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l ASME 5. ' ASME I CODE CLASS 1 CODE CLASS 2 FEEDWATER *A* f
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,; I JOCKEY PUMP &
7 i I i I i I 2r r i' I __ L _ - L I--- _ 1 _ T _ _1 _ N* _E HK p } - 4 I p " -J A G TO FROM l g MAIN PUMP RCW *A* I RCW *A* I i Figure 2.4.1a Re-idual Removel (RHR-A) System g
O O O N-i. i I ,
- t TO FROM ,
F3C FPU 3 A l 3 R 7 I c - -fj f- - _ -M - _ > < R @ 12 I l g_--______________________g___ p . 4 ) 1 I I ' I I 2 1 _________________________ ______ .. q , _ _4 l 1 y 211 [ i _ _ _ _ _ ,_ _ y ; = __ y _ _ ____ l I h h ' _____________H__q h l k FM; g_______ __________ __t4______"_ l 1______ p ________,____j_________ ____ 4 t I h y[] t________ i 3'
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a FROM y TO { RCW *B* RCW *B* d i i I Figure 2.4.1b Residual Heat Removal (RHR-B) System f 1
N REACTOR BUILDING EXTERNAL CONNECTION {_ y _ _.. _ _ _ _ _ _
] '
FROM FIRE FROM TO PROTECTION FPC F)C A 3l 2 SYSTEM - - - - - ~ N -*l 3 [g [g 2 Y m ) s-- - s--- l m m _ .. _ _ } _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , 1 n I
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- g. _
g Hx I P T T P I A Y cn MAIN PUMP TO FROM RCW *C* RCW 'C' ! 8 Figure 2.4.1c Residual gt Removal (RHR-C) System
ABWR 0: sign oocunwnt 2.4.2 High Pressure Core Flooder (HPCF) System Design Description The liigh Picssuie Coic Flooder tilPCF) Svstem is t ompiised of two divisionath separate subsptems that provide cincigency makeup water to the scartor foi transient or !J M1A conditions. The configuiation of cat h loop is hown in Figure ".4.2 (aligned in the standhv mode). The llP( T Sptcin is a p.u t of the ECCS network. and portions of ine ustem are considered a pait of the scactoi coolant pressure bound.ur (RCPIR The entire llPC' Spi h h Wmed to s.dcty-related standaids. The I.CCS function of the i M +, . on nned by the Ifigh Prewure Floodei Mode, which floods the 1 e a..v t h r-actar f oi any sca< 101 pressure < ondition when an initiation smo.;. .aciwd. Ancillary modes of operation include litillirntiin flow hypass and !:dl flow testing. Following icceipt of an inhiation signal (low reactor wate level or high daywell pressure), the llPCF Systein automatically initiates aiul operates in the floodes inode in conjunction with the remainder of the ECCS netwoik. This emergency makeup to the reactor vessel contributes to keep the reactor core cooled so that the regulatory reqtiirements governing f uel perforinatice during a I A X1A are h)v met by the ECCS network The flooder rnode is accomplished by both loops of the IIPCF System by uansferring water from the Condensate Storage Tank (CST) or the Suppiession Pool (S/P) to the RPV. The flooder mode is the only automatically initiated mode of the llPCF System, but it may also he initiated manually. The system will automatically severt to the flooder mode of operation from the test mode upon receipt of an initiation signal. Each loop of the llPCF System also has both a miniinuin flow inode and a f ull flow test mode. The minimum flow mode assures that there is pump flow suflicient to keep the pump cool by opening a minimnin flow valve that directs flow back. to the S/P anytime the ptonp is iunning and the snain discharge vahe is closed. Upon sensing that there is adequate flow in the pump main discharge line, the minimum flow valve is automatically closed, in the full flow test mode, the system draws suction from the S/P and discharges back to the S/P. The HPCF System is comprised of two separate loops or subsystems, each of which includes a pump and takes suction from either the CST or the S/P, and directs water back to either the RPV or the S/P. The preferred suction source is the CST. Automatic suction transfer from the CST to the S/ P occurs with a CST low water level signal or with a S/P high watei level signal. The divisional subsystems of the llPCF System are separated both mechanically and electrically, O as well as being physically located in different areas of the plant to address requirements pertaining to lite protection and othei sep.u ation criteria. The HPCF System is separated both physically and electrically h om the RCIC System. 2A2 6/1/92
ABWR oesign occument Each of the ino subsystems is power ed f r om a sepanne Class 1 F.dhisional power g disuihution bus that can be supplied from eithen an omsite or of f4ite soun c. W Cooling water to each division of the HPCF putop ami motoi < oolet s is supplied by the respective dhision of the seat tor cooling watei (RCW) System. The 11PCF System aho includes provisions for containment isolat:on and RCPit pressure i isolation. The if PCF Systesa will maintain the capability to perfoun its intended safety-iciated furu tions either following a Safe Shutdown F.u tlujuake or during the emitonmental conditions imposed by a IMCA, and in each case assuming the worst case single failure. The system will also acconunodate calculated movement and thermal stresses. The system is designed so that available NPSil exceeds required NPSU for the pornps in all operating modes. The system can be powered from either nonnal off-site sources oi by the emesgency diesel generators. The H PCF System is Seismic Category I and is housed in the Seismic Category I icactor building to provide protection against tornadoes, floods, and other natural phenomena. The HPCF pumps are motor <hiven centrifugal pumps capable of supplying pressure at flow conditions at least equal to or greater than the udue conesponding to a stmight line between a scactor piessure of 1177 psid at 800 gpm and at a reactor pressure of 100 psid at 3200 gpm. The 1177 and 100 psid pressures are taken between the vessel and the air space of the compartment containing the source water for the pump. The puir.ps are ASSIE Code Class 2 l components with a design pressure of 1565 psig and a design temperature of 212 F. The pumps are interlocked from starting without an open suction path. The HPCF pumps are protected from possible pump nimout conditions in all operating modes. Each loop of HPCF utiliics a connection from the 51akeup Water System (Condensate) (StUWC), which remains open throughout plant operation to sene as a keep-fill system for that k>op's pump discharge piping. The HPCF System piping and valves are AShiE Code Class 1 or 2 as shown on Figure 2.4.2. The design pressure and temperature of piping and udves varies ( across the system. For that piping attached to the RPV, from the RPV out to the ( containment side (downstream side) of the outboard containment isolation l udves, the design pressure and temperature are 1250 psig and 576 F, respectively. The design pressure and temperature for the outboard l containment isolation valves are 1565 psig and 576 F, respectively. For other l piping open to the containment atmosphere, out to and including the outboard I (ontainment isolation valves, the design pressure and temperature are 45 psig and 219 F, respectively. For piping and udves outside the contaimnent isolation valves, the design piessme and temperature depends on whether it is located on I the suction or discharge side of the main pump. Those portions on the suction side are rated at 200 psig and 212 F, while those portions on the discharge side are rated at 1565 psig and 212 F, respectively. The low pressur" portions of the 2A2 2- C/1/92
ABWR oesign occunant shutdown cooling piping ;uc inotc< ted fiora l'ull icactor jnessuie in two ilu rk d udves in series or combinations of nor mally closed valves. Relici vahe . ;u e also juovided for protection noin overinessuie sesulting fioin high piessme valve leakage or water thennal expansion. The llPCF System includes Contiol Room indication to allow Ior the inonit o ing atul control during design basis opeiational wnditions, i.e., system flows uul jnessm es as well as udve open/ dose aiul puinp on/of f indication 10: t hose instnanents and components shown on Figtue 2.4.", with the esreption of siinple check udves anal overpresstue iclief valves (of the check vahes shown only the testable check udves downstrearn of each loop's RI'\' injection vahe has control room status indication), inspections, Tests, Analyses and Acceptance Criteria Table 2.4." provides a definition of the inspections, tests anul/or analyses together with associated acceptance criteria which will be undcitaken lot the liPCF System. l { i b 2.4.2 3- G/1/92
{ Table 2.4.2: High Pressure Core Flooder System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria [ 1 The two loop divisional configuration of 1. Inspections of the as-built HPCF 1. The actual two loop HPCF System the HPCF System is shown in Figure 2.42, configuration shall be performed. configuration, for those components which are each mechanically and shown, ccoforms with Figure 2.4.2 and electrically separated from each other. separation requirements.
- 2. The HPCF System operates in the flooder 2. The ECCS t.OCA performance analysis for 2. HPCF System actuation and operation is i mode as part of the overat! ECCS network. assuring core cooling shall be validated by consistent with the ECCS performance the HPCF System: analysis as follows: ;
- a. Demonstration that the flooder mode a. HPCF pump developed pressures of at (of each HPCF toop)is capable of least 1177 psid and 100 psid for flow automatically initiating and operating rates no less than 800 gpm and 3200 in response to an initiation signal. gpm, respectively, where the pressure b difference is between the RPV and the
- b. Analyses to demonstrate compliance air space of the compartment with acceptance criteria using as-built containing the source water for the functional performance test data and pump, and where the water construction dimensions. temperature is vaiued at 50*F.
- b. 36 seconds maximum allowed detay time from the initiating signal to rated flow available and the injection s alve e fully open.
- 3. The HPCF System operates when powered 3. HPCF System functional tests shall be 3. HPCF System is capable of operating when from both normal off-site and emergency performed to demonstrate operation when supplied by either power source.
on-site sources. supplied by either normal off-site power or the emergency diesel generator (s). 4 If already operating in any other mode, the 4 Using simulated inputs, logic and 4 HPCF logic functions to automatically HPCF System automatically reverts to the functional testing shall be performed to reconfigure the system to the flooder mode flooder mode in response to an initiation demonstrate the HPCF Systems ability to of operation upon receipt of an initiation e signal. automatically revert to the flooder mode signal. . } from any other mode. t l 9 9 9
} ( I ;l! i r > 1f' t jil[l fl i[l I - I t l ,
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)
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o a p e n s 2 s o t n yet rh o n c e n Hi to i l yum t Fit l I deor mae mh st ap u e eCa r owe b l p Si a d rf h la ot r a u doHpP e P dn ce t No T n e i vgg it h f p npk uo mfo or o ri nh i at dn it n ket) t c ng _ m ppy nt e c kieo t sne t epe i l oe ct c o c aC eow in t i ici m r aFi v t t uor lo u r aMW. d t wap io t f ar t e s f a p seUe r u m o sCs eP s s t n ehMi l lot sal s e p eo nep l vHc e zt C t t i ( f f nm o r eo v a iimu or ms l l _ n g l ae vr x e r e ed . eo r t en e lumo ad po an f en h et n uo i omg t i s e st sa mf e nye sl mdg sa D i osd e p n pt oSg e cn pl u d t arpt e owi oo a t muo t si t y c )e a r t sii r ydo u mts uo e lo c e l l e uh pt Set h Sidr pp i f swj . F f hr F na c C iw i i t ob Cnr e F n ni ss F ey Fl l Ca C r e el r uc t s u er u P uv Hm o P g Hi tn Co P c ded Cpit P si l P g segs st ns hi cnm er H ee n nm p Hwa ecp b Hn er i eoie o r r er Eaiomfr h a h pC u O t h! a ! h u P pb p Ts To( p Tac Td
. . . . . 0 5 6 7 8 9 1 4,
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g Table 2.4.2: High Pressure Core Flooder System (Continued) 4 Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Cnteria Certified Design Commitment
- 11. Procurement records and actual equipment 11. HPCF equipment has appropriate ASME,
- 11. HPCF mechanical equipment is built in accordance with ASME Code, Section lit shall be inspected to verify that applicable Section lit. Class 1 or 2 certifications in HPCF System components have been accordance with its proper class?fication requirements. f manufactured per the relevant ASME (Section 2.42).
requirements.
- 12. Inspections will be performed to verify 12. The instrumentation is present in the
- 12. Centrol room indications are provided for presence of control room indication for the contro! room (Section 2.41).
MPCF System parameters as specified in Section 2.4.2. HPCF System (Section 2.4.2).
- 13. Pump vendor test records and calculations 13. HPCF pumps,in as-installed system
- 13. The HPCF pumps have adequate head / flow will be inspected, and as-installed system configuration, demonstrate head / flow characteristics. characteristics consistent with design basis flow testing conducted, to establish pump head / flow characteristics. analyses assumptions.
- 14. System logic testing using simulated hput 14. Suction auto transfer occurs on Icw CST or b 14. HPCF pump suction automatically switches over from CST to the suppression pool on signals shall be performed to demonstrate high soppression pool water level.
Iow CST or high suppression pool water auto switch-over of suction source and level with override protection. override.
- 15. Functional testing using simulated input 15. HPCF auto shutdown on high reactor water
- 15. HPCF System auto shutdown on high level, and auto re-stsrt on low reactor reactor water level and auto re-start signals shall be performed on the system logic to demonstrate HPCF systems water leve!.
capability. I capability to automatically rhutdown on high reactor water level, and automatically re-start when low water level re-occurs. B O O O L ____
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l _ _ _ _ _ _ _ W _ _ _ _ _ _b 4 _ _ _ _.] I ! i Figure 2.4.2 High Pressure Core Flooder System
ABWR Design occument 2.4,3 Leak Detection and Isolation System O Design Description The priimn3 hmetion of this system is to detect and monitor leakage f rom the 2cactor coolant pressure houndan (RCI'li) and to initiate the appr opriate saf ety action to isolate the leakage soun e and pievent f urther radiological icleases Iloin tbc NCIit. The syste!!1 is desigiled h3 aun Ollalically initiate isolation (d the main steamlines and other process lines that connect to the containment. Isolation tesohs in closme of the apinojniate c ontaimnent inhoard and outboard isolation vahes The 1.DS functions include c ontainment isolation following a 13 CA event, monitoring of leakages inside and outside the primaiy containment, monitoring of identified and un identified leakages in the di3well, and annunciating excessive leakages in AlCil 1.DS is a four divisional safety system. The instrumentation that initiates contairunent isolation consist of four redundant divhional channels for each monitored plant variable. The logic design is such that any nemi-of-four channel trip will resuh in initiation of the appropriate isolation function. Various plant parameters are constantly monitored for indication of reactor coolant leakage such as flow, pressure, water level, temperm ore, radiation . . etc. All LDS safety related measurements are uansmined to the ndriopn>cessor based Safety System & Logic Control (SSI.C) System foi processing, setpoint i comparisons, and generation of the required trip signals that initiate the isolation functions. The LDS control and isolation signals are scaled-in and will require Inanual logic reset to return the logic to its norinal status. The following automatic control functions are provided by LDS: (1) Isolation of H21/htSIVs and $1SL drain vidves on low reactor water level (LI.5), high .\fSL flow in any steamline, high htSL tunnel area radiation, high ambient ternperature in 51SL tunnel aica or in turbine
- building, low main condenser vacuum, or low inlet turbine pressure as shown in Figure 2. l.3a, t
(2) Isolation of G31/CUW system process lines on low reactor watei level (L2), high ambient 51SL tunnel area temperature, high nutss differential flow, high ambient temperature in equipment areas, or when C41/SLCS is activated. (3) Initiation of T22/SGTS operation on high dnwell }nessure, low reactor water level (L3), or high radiation in the secondary containment. 2.4.3 1- 6/1/92
ABWR oesign Docuinent (4) Isolation of scactoi huihling l'41 z llVAC ventilation ssstein on lugh drywell pressure, low trat tot watei level (1.3), or high r.uli.uion in the secondaiv contaiinnent. (5) holation of contaisunent purge and vent lines of T31 < AC systein on high dnwell pressure, low icactor water lesel (1.3). oi high r.nhotion in the sect uularT contalinnent. (6) Isolation on high dnhll pr essuir or low leartoi water iesel (1.1) of the P21/RCW cooling water lines to di3well t ooleis and Ril' heat exchanger. . atul of the P"4 /ilNCW cooling watei lines to dn weil cooling systt in. (7) Isolation of El 1/ RilR shutdown cooling loops on high scactor picutne or low scactor water level (L3), aiul iaolation of cat h RilR shutdown cooling loop on high ann >ient teinperatuie in its equipinent area. (8) Isolation of E51/RCIC stcainline to the tuihine on high stcainline flow, low steainline pressure, high tuihine exhaust pressine, or high equipinent aica ternperatuie. (9) Isolation of G51/SPCU suppression pool clean-up systein on high dr)well pressure or low reactor wter level (1,3). (10) Isolation of T49/FCS flanunability contiol system lines on high dipsell pressure or low reactor water level (L3) (11) Isolation of the dnwell sumps drain lines an high dnwell pressure, low _ reactor w.uer level (L3), or high radioactivity in the drained liquid. (12) Isolation of fission products inonitor drywell sampling lines on high drywell pressure or low reactor water level (L"). (13) Initiation of C51/NMS ATIP withdrawal on high dipsell piessurc o: low reactor water lesel (L3). In addition to those functions specified above, the following parameters are continuously monitored by LDS for indication of leakages: (1) Condensate flow from the di)well air coolers - one flow channel (2) Ds)well surnp levels changes one level sensor per sump (3) Dnwell air temperature - four thermocouple channels (4) Valve stein leakages inside dr}well- one letuperatute sensor per valve 2.4.3 2- 6/1/92
ABWR oosign occument f~ (M Difleiential ainbient teinperatuie in equipinent .n e.n - one set of theiinocouples pei equi l onent aica ($1SI. tunnel. RilR. RCl(:;uul CUW areas) As shown in Figur es 2.4.3b aint d.1.DS consists of instruineni ( hannets and logic d units that initiates the isolation f unctions. Also, inannal( onn ois .u c piovided as k desciihed beolw for isolation and logic r eset,51SIV inode uinn ol aiul test. .uul f or channel bypasi 1.DS utilizes the essentialand non+su niialinuhiplexing svstenn 031S & N1'h1S) as appropiiate foi data conversion and transinission except foi the signals that control the $1SIV pilot solenoids which ;ue hant wited. _ LDS is a four division systein designed to piovide icliable single-taihuc pioof capability to autoinatically initiate the isolation functions. A single channel , f ailure or a loss of one divisional power to a single channel will not cause inadven tent isolation. Also, LDS incorporates logic provisions to pennit bypass of single division of channel at a tirne to facilitate inaintenanc e and icpaii. The inain condenser vacuum channels provided for 51SIV isolation can be bypassed inanually or autoinatically during allinodes of scactoi operations except when in the run inode to guaid against htSIV spuiions isolation. 1.DS provides the following control signals to each htSIV which contains thice pilot solenoids, #2 and #3 for inode control and # 1 ior test only as shown in Figure 2.4,3c: (1) Two<>ut-of-f our control signals to solenoids #2 and #3 t:_ open the valve. htSIV closuie is automatir on loss of signals to both solenoids - (2) Two divisional control signals to test solenoid #1 to exercise v;dee closure during norinal reactor operation. Division 1 or 3 is used to test the outboards 51SlVs and division 2 or 4 will test the inboard 51SIVs. , Also, l.DS provides three divisional trip signals (Div 1,1,', and 3) as shown in Figure 2.4.3b for isolation of the appn>priate containment isolation valves. The LDS design incorporates the folknving inanual control switches for isolation of the htSIVs, the containinent ' alation valves, at d the RCIC systenu (1) Four divisional hiSIV isolation switches - one per division Simultaneous closure of all the $1SIVs requires the use of two divisional switches, either divisions 1 and 4 or 2 and 3. v (2) Three l'W containtnent isolation switches - one pei division 1. 2 and 3 2A3 3- 'J ll9 J
ABWR oesign Document E.u h disisional switch will isolate all its respective divisiorud g containtnent isolation valves cucpt for the MSIVs and itCIC W (3) Two itCIC isolation switches - one per division I and 2 Either divisional switch will isolaic the stcain line to the f(CIC tuihine and cause turbine shutdown. Division I will isolate the inboard and disision 2 will isolate the ontbo.ud isolation sahes. l.ogic reset switches (9 total) are prosided on a disisional bases for inanual reset of the logic, coinplementing the noinber of isolation switches that are provided as described above, in addition to the isolation and logic reset switches, each .\lSIV is provided with a inode contiol switch (dual bank) which supplies division 1 and 2 signals for energizing its pilot solenoids #3 and #2. tespectively. Also, each MSIV is provided l with a test switch (dual bank) which supplies two divisional signals to its solenoid
#1 f or exercising valve closure to its 90% open position during reactor operation. l l
Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.3 provides a definition of the inspections, tests, and/or analyses i together with associated acceptance criteria for the 1.cak Detection and Isolation Systein. g! O 2 4.3 -4 6/1/92
4 Table 2.4.3: Leak Detection and Isolation System { j inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. LDS is designed as a safety system to 1. Each LDS channel sha!! be checked for 1. Proper channel calibration and response is I
detect leakage from the RCPB by proper calibration either by reviewing serified when the records and/or the test monitoring changes in plant parameters, certified records by performing channel results are considered acceptable. Also,
- alarm the high leakage levels in MCR, and response tests and/or by cross channel channel integrity and operability is verified initiate closure of the appropriate comparison.To check channel integrity and when the trip signal that initiates an alarm containment isolation valves. operabitity, a simulated signalinput shall and/or isolation occurred at the setpoint.
, be used to verify initiation of the 4 appropriate trip signal at the specified setpoints for alarming and/or isolation.
- 2. Four redundant safety divisional channels 2. Each LDS logic isolation function shall be 2. Acceptance is based on satisfying the l are previded to monitor each plant tested using various simulated signal required two out of four criteria for
! variable. The logic design is such that any inputs to verify that isolation occurs only initiating an isolation function. two oct of four channel trip will initiate an when any two or more out of the four i ? isolation, channels indicate trip. i 3. The LDS logic design permits bypass of a 3. While in channel bypass, each LDS logic 3. Acceptance is based on satisfying the single division of sensors at any one time isolation function shall be tested using required two out of three criteria for to permit test and maintenance during various simulated inputs to verify that initiating an isolation function while in the normal reactor operation without causing isolation occurs only when any two or bypass mode. outage. more out of three channels indicate trip. . 1. The LDS design provides each MS!V with a 4. Actuation of the MSIV mode switch shall 4 Acceptance is based on verifying valve I dual-bank mode control switch to open and cause the valve to open.The mode switch operation ut der the specified conditions.
- close the valve, and with a dual-bank test shall provide 2 control signal, Div 1 to pilot Div 1 and 2 mode switch control signals switch to exercise valve closure during solenoid #3 and Div 2 to pilot solenoid #2. when applied to MSIV pilot solenoids #3
! reac+or operation. Actuation of the test switch sha!! cause the and #2, respectively, cause the MSIV to valve to partially close to its 90% open fu!!y open. Either divisional control signals l position and
- hen return to normal. The from the test switch when applied to MSr/
test switch shall provide Div 1 and 3 for the pilot solenoid #1 cau:;e partial closure. , outboards and Div 2 and 4 for the inboard MSIVs. R c 8 4 J I
*g Table 2.4.3: Leak Detection and Isolation System (Continued) w Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. The LDS design provides control switches 5. 5. Confirmation that each specified manual to isoi=te separately the MS!Vs, RCIC and a. Simu'taneous closure of the MSIVs shall isolation function was properly the PCV as follows: occur when two MS!V switches are implemented. Also, closure of each valve is
- a. Four I,11SV switches, one/ division 1 to 4 actuated, Div 1 and 4 or Div 2 and 3. confirmed by its indicating position status
- b. Two RCIC switches, one/ division 1 and 2 b. RCIC system shall be isolated by either Div light , RED for close position and GREEN
- c. Three PCV switches, one/ division 1 to 3 1 or 2 switch. (Div 1 for inboard and Div 2 for open position.
for outboard valves.)
- c. Each PCV divisional switch shall isolate its respective divisional containment isolation valves.
- 6. The main condenser vacuum logic 6. Verify that each main condenser vacuum 6. Verification that each main condenser channels are bypassed during startup and channel can be manua!!y or automatically vacuum logic channel can be bypassad as shutdown to guard against spurious bypassed by the logic without causing indicated by the logic.
6 isolation. MStV trip under simulated conditions.
- 7. The LDS design provides divisional logic 7. The divisionallogic isolation char.nel of the 7 Confirmatim that the divisional logic for reset switches that are used for init:al set of MSIVs, RCIC, and the FCV sha!! be initially the specified functions are normally sat.
the logic to de-energize to trip and for logic . set. Four switches (1/Div) are provided for reset after the trip conditions have cleared. MSIV logic reset, two switches (1/Div) for RCIC logic reset, and three switches (1 for Div 1,2 and 3) for PCV logic reset.
- 8. LDS monitors identified and un-identified 8. Verify that the instrumented channels that 8. Confirmation that each instrumented leakages in the drywell and af arms in MCR monitor sump level changes, drywell channel is operable and that alarm ex cessive leakages. coolers condensate flow, and leakages setpoints are verified from valve stems are operable and the alarm setpoints are correctly set. l i
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ABWR 0: sign Document 2.4.4 Reactor Coro isolation Cooling System Design Description The Reactor Core Isolation Cooling (RCIC) biem. in conjunction with other systems, supplies inakeup water to the teactor piewmc vessel to assure that sullicient water inventory is maintained to per mit adequate core moling to take place during the following events: (1) Inssef<oolant archlent (I A JCA). (2) Vesselisolated and maintained at hot standby. (3) Vessel isolated and accompanied by loss of feedwater flow. (4) Complete plant shutdown with loss of normal feedwater system belone l the reactor is depressuriicd to a level where the shutdown cooling l mode of the RI1R System can be placed in service. (5) loss of all AC powc : l The RCIC System consists of a 100% capacity steam-driven tmbine which drives a 100% capacity pump assembly and the pump accessories. The sptem also V includes piping, ndves, and instnunentation necessary to provide several flow paths for system opemtion. The RCIC steam supply branches off from main steamline "B" (leaving the RPV) and goes to the RCIC tuihine with drainage provision to the main condenser. The turbine exhausts to the suppression pool with nicuum breaking protection. The primary source of RCIC suction supply is from the Condensate Storage Tank (CST). The suppression pool water is the secondary source of RCIC supply. Autoir tic switchover of makeup water source from the CST to the suppression pool (with override provision) is integrated in the system logic. CST and suppression pool suction valves are interlocked, and check udves are provided to safeguard accidental drainage of CST water to the suppression pool . RCIC pump discharge lines include the main discharge to the feedwater line, a test return line to the suppression pool, a pump minimum flow bypass line to the suppression pool, and a cooling water supply line to auxiliary equipment. The piping configuration and insuumentation are shown in Figure 2A.4. The RCIC System is a part of the ECCS network and is designed to safety-related standards, it is powered from Class lE DC sources (except the inboard steam supply isolation udve, whh h has Class 1E AC), and is designed to perform its function deprived of all Ju . sources. Although RCIC System design is safety related, it also performs some non-safety-related functions. The safety-related functions include emergency core cooling, in conjunction with the liigh Pressure Core Flooder (IIPCF) System. Automatic Depressurization System 2.4.4 -1 G/1/92
ABWR Design Document (ADS) and the Residual Heat Removal (RHR) System. As p.u t of this network, the RCIC Ststein can provide icactor makeup in the period while the scactor is still at high pressure af ter a sinall bicak has occuried. The non safety.iclated functions inchide providing makeup water to the teactor piessuic sessel (1) during transient events accompanied by loss of fcedwates, and W) dming a complete loss of all AC power (Station lilackout). l lhning norinal ope ation, the RCIC System is in its standby condition with the j motor-operated valves in their normally open or norinally t losed position (Figure 2AA). In this mode, the pump discharge line is kept lilled with water l supplied by the system head of the Condensate Makeup System to pievent waterhanuner in the discharge t ping i system when the RCIC System is initiated. Full flow f unctional testing may be performed with the RCIC pump taking suction froin and returning flow to the suppression pool. Should an initiation signal occur during test mode, the system configuration would automatically scalign to the vesselinjection mode. During tmnsient and LOCA events, RCIC System is autoinatically initiated upon receipt oflow reactor water level or high diywell pressure signal. The steam turbine <h-iven pump delivers water f rom the CST or hom the suppression pool to the reactor vessel via the feedwater line "B" and distributes it tiuough the feedwater sparger to promote mixing with hot water or steam within the reactor vessel. The RCIC turbine is driven by the portion of the decay heat steam from g the reactor vessel, and exhausts through a discharge sparger below the suppression pool water level. The imhine exhaust line penetmtes the containment > location about 1 meter above the suppiession pool maximum water level. Two vacumn breakers in series aie connected to the exhaust line (abme the suppression pool water level) in the wetwell air space, A check valve and a remute manually operated motorized valve installed in series outside the containment provides containment isolation function for the turbine exhaust line. When high reactor water level in the reactor vessel has been established, the vesselinjection valve and the steam supply admission valve to the turbine will close, causing the turbine to shut down. When the low reactor water level initiation signal re occurs, the RCIC System will automatically restart to provide the core cooling function. The RCIC turbine is automatically tripped (turbine trip and throttle valve isolated) upon receipt of anysignalindicating turbine overspeed, low pump suction pressure, high turbine exhaust pressure, or an autcrisolation signal from the Ixak Detection System (LDS). Once tripped, the spring closing mechanism latches and must be manually reset if the turbine needs to be re-started. This very same isolation signal (LDS) also isolates the RCIC steam supply isolation valves h 2.4.4 6/1/92
ABWR oesign occument to piovide primary contaisunent isolation. The 1.DS isolation signah ,uc as d follow: (1) A high piessuie diop ac ross a 110w device in the sic.nn supph line equindent to 300% of the stcach-state stcain flow. (2) A high RCIC area temperature. (3) A low scactor pressure (low steamline piesstu el (4) A high piessure between the RCIC tuihine exhaust iuptuie diaphiains. The RCIC System can also be manually initiated and shut down h un the main - control room as long as permioive inteilocks are present. In the event that all AC power sonices iue not available (St uion Blackouth tlu-RCIC System is designed to perfoon its cos e cooling hun tion foi at least 8 houis. Station batteries and CST water inventory are sited to support the S-bom RCIC operation. The RCIC room is designed such that room tempemture ch>cs not , reach the equipment maximum enviroinnentallimit for at least 8 hours without ioom cooling. The RCIC steam supply isolation valves aie nonnally open motoi-operated valves. These valves fail as-is (open) on loss of AC power, ther chy providing steam supply flowpath to the tuihine. During this event, the icactor pressure is controlled and maintained at the main steam safety / relief valve (SRV) set pressuie to assure an 8-hour steam supply to the RCIC tuibine. The RCIC System is designed to Seismic Category 1 icquiiements azul is housed in a Seismic Category I reactor building structure to provide protection fiom _ tornadoes, floods, and other natural phenomena. The RCIC System also includes provision for primaiy containment and RCPit pressure isolation. The RCIC piping system and udves aie Seisn'ic Categoiy 1 Quality Group B except for the steam supply piping, which is Seisinic Categoir 1, Quality Group A up to and including the outermost primary contaiinnent isolation udve. The inboard and outboard isolation nuves are powered from independent Class lE sources. The steam supply piping up to and including tlu-tuihine has a design pressure of 87.9 kg/cm2 g antio design temperature of 302 C, while the turbine exhaust piping is designea to 10 kg/cm 2g and 184*C The RC;1C pump discharge pipig up to the ir,j.ection udve is designed to 120 kg/cm g and 77'C. The injection se itselfis rated at 120 kg/cm g and 302*C Beyond the injection .alve, the di3charge piping portion that connects to the feedwater line is rated at 87.9 kg/cm;g and 302 C. The pump suction p?hg i is mted at 21 kg/cm2 g and 77'C ProteuLa of the low pressure suction piping (c from potential high reactor presore is accomplished by thice udves in series (testable check udve, injection udve and ptunp dischaige check udve) at the pump discharge line. 2.4.4 3- 6/1/92
ABWR oesign occwnent The RCIC tur bine which drives the inunp is a safety ', ,a pone n t, ahhough not ( ovcied by the ASM E Code. The glatul w al .s t.ot safety-iciated, but ' it is not essential for RCIC operation. The tuihine and its accessories are i seismit ally designed and analy/cd to withstand a design basis carthtjuake (1)lW L The turbine is designed to opemte at both high and low pressu.c conditions. The ininimmn ucarn inlet picsunes at high picuore condition is 82.8 kg> cm abs. and 10.5 kg/cin' abs for the low inesstue condition. The RCIC pump is designed to Seismic Category 1 Qiality Group IL The poinp is a constant flow centrif ugal type capable of providing an injection flow into the reactor vessel of at least IM' m3 'lu against a diff esential pressure of 82.8 kg/ctn 2 (drywell to RPV) within 30 seconds following receipt ofinitiation ;ignals. The suction piping configuration is designed such that adectuate NPSil is always available on all RCIC operating modes. Pump developed head is about 900 2 meters at 83.8 kg/cm abs and 186 metens at !1.6 kg/cm 2abs reactor pre.uute. The RCIC System includes control room indications and alanns to allow for the monitoring and control during the design basis operational conditions, i.e., system flows, temperatures, pressures, valve open/close and pump on/off conditions, and bypassed, override or inoperative status conditions. Inspections, Tests, Analyses and At ceptance Criteria Table 2 A A provides a definition of the inspections, tests and/or analyses together with associated acceptance criteria which will be undertaken for the RCIC System. i I O l 2.4.4 L C'1/92 I l l
~ ---
O O O Table 2.4.4: Reactar Core isolation Cooling System Inspections, Tests, Analyses rtnd Acceptance Criteria inspections, Tests, Analyses Acc6ptance Criteria Certified Design Commitment
- 1. Inspection of the as-built RCIC 1. Verification of the as-built system ,t in
- 1. The configuration of the RCIC System is configuration shall be performed. conformance with the as-designed shown in Figure 2. 4 4.
configuration (Figure 2.4.4).
- 2. Using simulated LOCA signal, functional 2. RC1C System automatically re-aligns to
- 2. The RCIC System automatica!!y re-aligns to vesselinjection mode upon receipt of vessel injection mode if a LOCA signal testing of the system logic shall be occurs whi!e the system is in test mode. performed to demonstrate the system's LOCA signal.
capability to revert to the vessel injection mode while in test mode. l RCIC pump is capable of delivering a flow 3. Vendor to conduct shop tests relating to 3. Verification of certified documentation
- 3. l rate of 2182 m3lhr against 82.8 kg/cm2d. pump performance. demonstrating that the pump will meet 2182 m3/hr against 82.8 kg/cm2d.
4 Vendor to conduct shop tests relating to 4. Valve closure occurs against design basis g 4. Steam supply isolation valves are capable valve operation during dc sign basis differential pressu re. of closure against the maximum design basis differential pressure. eversts. RCIC pump suction automatically switches 5. System logic testing shall be performed to 5. Suction auto transfer occurs on low CST or 5. over from CST to the suppression pool on demonstrate auto switchover of suction high suppression poof water level. low CST or high suppression pool water source and override. level with override provision. RCIC steam supply isolation valves fail as- 6. Field testing shall be performed to 6. Valves remain open upon removal of AC 6. demonstrate that the steam supply power. is (open) on loss of AC power. isolation valves (normally open motorized valves) will stay in the open position when AC power is lost. RCIC steam supply isolation valves isolate 7. Functional testing shall be pedormed en 7. Valves isolate within 530 seconds from 7. the system logic by simulating the auto- receipt of auto isolation signafs. upon receipt of auto 4 solation eignals from Leak Detection System in s 30 seconds. isolation signal from LOS.
- 8. Functional testing shall be performed on 8. RCIC auto shutdown on high reactor water
- 8. RCIC System auto shutdown on high p
the system logic to demonstrate RCIC levet, and auto re start on few reactor d reactor water level and auto re-start System's capability to automatn.affy water fevel. 3 capability. shutdown on high reactor water level, and automatically restart when low water level re-occurs m , . , ,
g Table 2.4.4: Reactor Core Isolation Cooling System (Continued) h Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria
- 9. RCIC mechanical equipment (except the 9. Procurement records and actual equipment 9. Certified documentation demonstrates turbine) is built in accordance with ASME shall be inspected to verify that applicable compliance with the appropriate ASME Code Section lit requirerr.. RCIC System components have been Code.
designed, manufactured and installed per relevant ASME Code.
- 10. Provision for control room alarnes, and 10. Inspection will be performed to verify 10. The control room alarms and indications indications vital for RCIC System presence of control room alarms and specified in Section 2.4.4.
operation. indications. 6 I 0
.^
t O O O
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^
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l O RPV ASME CODE ASME CODE . j I r , 2 2 i M M CLASS 1 CLASS 2 f - - jISOLATION_ISOLATION M_'_ , 4 __ j S!GNAL SIGNAL M FE [ ~ ~ ~ ~ ~ ~ ~ '~l ' V d . 9 2# j I--- i I NC h
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Figure 2.4.4 Reactor Core isolation Cooling (RCIC) System P&lD
ABWR oesign Document g 2.5 Reactor Servicing Equipment G 2.5.1 Fuel Survicing Equipment Design Description The fuel senicing equipment is that equipment requiied for nonnal planned reactor refueling outage. This equipment will be used with other gencial pl.mt equipment that is not covered here. Also listed with this equipment is the Refueling Platfonn and this is covered in section 2.5.5. Fuel senicing equipment has an "Non-Essential ClassiUcation", " Safety Class" of other," Quality Group" of Electrical Codes and a ' Seismic Category' of none. The only exception to this is the New Fuel Inspection Stand which is " Passive Essential" and " Safety" ci..ss 2. Fuel Prep Machine The spent fuel storage pool has two fuel prep machines mounted to the pool walls. These machines are used for the stripping reusable channels from spent fuel a ' for channeling of the new fuel. These machines also provide an under water inspection capability of the fuel. The fuelis mounted to the carriage which q has an upper travel stop. O New FuelInspection Stand This fixture is a staad that holds two fuel assemblies for receiving inspection. There is a movable work platform surrounding the stand which allows the technicians to perform the inspection. The stand is firmly attached the wall on the refueling Door. Channel Bolt Wrsnch Th 4 is a long socke; wrench that fastens and unfastens the cap screw holding the che anel to the fuel assemble. The wrench also captures the screw. Channel Handling Tool This is a manually operated channel grapple that uses the area boom on thejib crane to support the weight. Vacuum Sipper This is a fuel isolation container for the monitoring of suspected claddmg failures. Fission product gas leakage is sensed by the Iseta detector and monitoring console. 2.5 6/1/92
ABWR Design Document General Purpose Grapple Generally used for fuel handling and support by the areajib crane. Used primarily in conjunction with the fuel prep machine. Channel Handling Boom This is ajib crane located in the area of the fuel prep machines. It is used to convenientl) s - items between the fuel prep inachine and the storage racks. Im sections, Tests, Analyses and Acceptance Criteria No entries for this system. O i i l l l l O 1 l 2.5 1 6/1/92 1 l
ABWR oesign oocument 2.5.2 Miscellaneous Servicing Equipment Y Design Description This equipment is generally used independently of other senicing equipment. Equipment re(piirements are that they operate in a emitonment up to a depth of 33 meters. Other requiiements are that the equipment can be quickly decontaminated and can be stored with a minimum of manpower. Under Water Lights Three types oflights are used; A general aica, a local aica and a drop type light. Viewing Aids Three types of siewing aids are used The floating type is the simplest, the under water siewing tube is a 1540 power telescope and the last is a under water remotely contralled telesision camera with an internal light source. Under Water Vacuum Cleaner The vacuum cleaner rests on the pool floor and is completely seniceable there. _es The power and control comes from the refueling floor by cables and the cleaner
) has its own accessories, inspections, Tests, Analyses and Acceptance Criteria Table 2.5.2 prosides definition of the inspection, test and/or analyses together with associated acceptance criteria which will be undertaken for the Servicing Equipment.
O 2.5.2 -1 6/1/92
S Table 2.5.2: Miscellaneous Servicing Equipment inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspection, Test, Analysis Acceptance Criteria
- 1. Environment 33 meters of water. 2. Install vacuum cleaner in 33 meters of 1 Visually evaluate the cleaner operations.
water and give operation test. 2 Envi.onment 33 meters of water. 2. Install television camera in 33 meters of 1 Visually evaluate the camera operations. water and give operation test. 1 P 5 O O O
O ' O~ 1
- g _
t u . i 1 _0 q' / PRIMARY CONTAINMENT
- j i- FEEDWATER LINE ~B-MAIN STEAM 4 - _ .- -_
ph; - -[><}- - + LINE *B- .-- 1 RPV I - m ; M A
! M rt ASME CODE
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, < :-w l ASME CODE t
i CLASS 1 CLASS 2 J *2 t+2 g4 _ r ,
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i i t Figure 2.4 a no- a i.
1 l ABWR Design Document Dryer and Separator Strongback The strongback is used to move the steam dryer or shroud head with steam separator between the vessel and storage. The strongback has pneumatically operated pins to engage the steam dryer or shroud head. The suongback is riesigned in accordance with the AISC and has a Safety Factor of 10 or better with respect to the ultimate strength of the material. The strongback is pioof-tested at 125% of rated load and all welds are magnetic particle inspected attei load test. Head Strongback/ Carousel The suongback is a combination of strongback, circular monorail and circular storage tray. The strongback senices many functions with respect to the vessel head. The constniction of the strongback is in accordance with the applicable AISC requirements. The suongback is designed to provide a 15% impact allow;mce and a Safety Factor of 10 or better to the ultimate strength of the material. The strongback is also designed to meet the applicable Crane Manufacturers Association of America, Specification and be tested in accordance with applicable ANSI requirements. All welding will be in accordance with The ASME Iloiler and Pressure Vessel Code, Section IX, Weldm r Qualification and proofload testing will be preformed with magnetic-particle inspection before coating. Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.3 provides definition of the inspection, test, and/or analysis together with associated acceptance criteria which will be undertaken for the RPV Senicing Equipment. l l 2.5.3 6/1/92
m. 1 i 5 Table 2.5.3: RPV Servicing Equipment Inspections, Tests, Analyses and Acceptance Criteria ! Certified Design Commitment inspection, Test, Analysis Acceptance Criteria
- 1. All tools have 60. year life. 1. Examine the design data. 1. Design data shows compliance.
- 2. Steam line plug safety factor of 5. 2. Examine analysis. 2. Analysis shows safety factor of 5. '
- 3. Steam iine plug is constructed to ACM 3. Examine certification data report.I 3. Certification data report shows code. compliance with ACM code
- 4. Meet or exceed the requirements of tlw 4. Es amine certification analysis report. 4. Certification data shows compliance with 1
.AISC using the floor respnnse spectrum AISC.
method of seismic analysis for pedestal.
- 5. Pedestal approved coating. 5. Examine certification data report. 5. Data report shows compliance.
- 6. Rack to store 8 studs. 6. Visualinspection of rack capacity. 6. Visual inspection provides verification. [
t g, 7. Rack safety factor of S. 7. Examine analysis. 7. Analysis data provides verification. !
- 8. Rack is constructed to ACM code. 8. Examine certification data report. 8. Certification data report shows compliance with ACM code.
t
- 9. D/S strongback constructed to .ASIC code. 9. Examine certification data report. 9. Certification data report shows <
compliance with A;SC code.
- 10. D/S strongback safety factor of 10. 10 Examine analysis. 10. Analysis data shows safety factor of 10.
- 11. D/S strongback proof-tested at 125% of 11. Examine certification data report. 11. Certification data shows 125% of rated load rated foad. compliance. ;
- 12. D/S strongback welds are magnetic particle 12. Examine certification test results. 12. Certification data shows results of tests. ,
inspected. i
- 13. Head strongback construction in 13. Examine certification data report. 13. Certification data reports show compliance with AISC and ACM. compliance with AISC and ACM code.
i
- 14. Head strongback designed for 15% impact 14. Examine analysis. 14. Analysis data shows safety factor of 10 and '
g and safety factor of 10 or better. 15% impact allowc ice.
- h 15. Head strongback designed to meet 15. Examine Spec 70 and construction data 15. Data report shows compliance witn applicable crane manufacturers' report. aonlicable requirements. ,
association requirements. !
- Table 2.5.3
- RPV Servicing Equipment (Continued) u inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspection, Test, Analysis Acceptance Criteria
- 16. Head strongback testing in accordance 16. Examine test results and compare with 16. Test results verify compliance with with applicable ANSI requirements. ANSI B30.16. applicable ANSI requirements.
- 17. All welding in accordance with ASME 17. Examine certification data report. 17. Certification data report shows Boiler and Pressure Vessel code, Section compliance and verification of ASME BPV IX, welder qualification and magnetic code Section IX, welder qualification and particle inspection. magnetic particle inspection.
b c 8 O O O
ABWR Design Document 2.5.4 RPV Internal Servicing Equipment Design Description The instrutnent strongback is used to handle I.I'RNI, and SRNNI Dry Tube from the floor to the icactor well. The auxiliarT crane of the building crane shall be used to lift and rotate the stiongback from horizontal position to vertical position. Then this strongback is moved tumu ds the icactor well, lowered into the RI'V.11pper portion of the strongback remains above water level so as the workers on the refueling platform can perform the operation. The Instnunent Handling Tool is connected to the wire terminal of the auxiliary hoist of the tefueling platfor m receives LPRN1 or Dry Tube b om the strongback. Instrument Strongback The instrument strongback is used to support in-core dry tubes while being assembled into the open reactor vessel. Instrument Handling Tool The instnunent handling toolis used to grip the dry tube assembly for remoud from the reactor vessel. Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.4 provides definition of the inspections, tests, and/or analyses together with associated criteria which will be undertaken for the Internal Senicing Equipment. 2.5.4 -1 6/1/92
Table 2.5.4: RPV Internal Servicing Equipment u Inspections, Tests, Analyses and Acceptance Criteria inspection. Test, Analysis Acceptance Criteria Certified Design Commitment
- 1. Examine analysis data. 1. Analysis data provides verification of
- 1. Strongback to be used over reactor well.
safety factor of at least 10.
- 2. Examine analysis data. 2. Analysis data provides verification of
- 2. Handling tool used over reactor well.
safety factor of at least 10. 9 l
?>
1 O O O
ABWR Design Document O 2.5.5 Refueling Equipment i V The Reactor Ituilding is supplied with a refue..ag platibun (or fuel movement and senicing plus an auxiliary platfonn foi senicing operations fiom the vessel flange level. Design DescriptiorwRefueling Platform The refueling platfonn is a gantry crane, which spans the reactor vessel and the storuge pools on bedded tracts in the refueling floor. A telescoping mast and grapple suspended from a trolley system is used to lif t and orient fuel bundles for placement in the coie and/or storage racks. Control of the platforn is from an operator station on the refueling floor. A position indicating system and travel limit computer is prosided to ocate the grapple over the vessel core and prevent collision with pool obstacles. ~."wo auxiliary hoists, one main and one auxiliary monorail trolley-mounted, are provided for in-core senicing.The grapple position provides sufficient wuter shielding over the active fuel during transit. The mast grapple has a redundant load path so that no single component failure will result in a fuel bundle drop. Interlocks on the platform: (1) prevent hoisting a fuel bundle over the vessel with a control rod removed; (2) prevent collision with fuel pool w; dis or other structures; (3) limit travel of the fuel grapple: (4) interlock grapple hook engagement with hoist load and hoist up power; and (5) ensure correct sequencing of the transfer operation in the automatic or manual mode. Design Description-.-Auxiliary Platform The auxiliary platform provides a reactor flange level working surface for in-vesselinspection and reactor internals senicing, and pennits senicing access for the full vessel diameter. No hoisting equipment is provided with this platform, as this function can be performed from the refueling platfonn. The platform operates on tracks at the reactor vessel flange level and is lowered into position by the Reactor Building crane using the dryer / separator strongback. The platform power is supplied by a cable from the refueling floor elevation. Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.5 provides definition of the inspection, test, and/or analyses together with associated acceptance criteria which will be undertaken for the refueling platfonn. No entries are proposed for the auxiliary platform. ( \ 2.5.5 6/1/92
{ Table 2.5.5: Refueling Platform Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The refueling platform has two auxiliary 1. Perform load tests on both auxiliary hoists. 1. Both auxiliary hoists shall be load tested hoists having the capacity of 500 kg each. and hold 125% of rated load.
- 2. The platform is provided with controls and 2. Review of as-installed equipment and field 2. Using normal installed controls and power, interlocks which: tests will be conducted after the platform the platform meets required operating has been installed. characteristics.
- a. Maintain water shielding over fuel when grappled on mast.
- b. Allow no fuel movesnent over vessel when control rod is removed.
- c. Provide fuel grapple travel limit.
N d. Prevent collision with fuel pool walls and other structures,
- e. Interlock grapple hook engagement with hoist load and hoist up power.
- f. Insure automatic sequencing control for transfer operation.
a FJ e G 9
ABWR Design Document n 2.5.6 Fuel Storage Facility r i
\j Storage racks are required for the temporary and long term stoiage of fuel and associated equipment. Storage may be eithei wet or d:T, depending upon the item being stored.
Design DescriptiorwFuelStorage Racl<s Itacks provide storce for spent fuelin the Spent Fuel Storage l'ool in the licactor liuilding. The racks are top loading, with fuel bail exicoded above the rack, and shall have a minimum capacity of 27096 of the reactor core. The rack design piceludes the possibility of criticality under normal or abnor mal
^
conditions and maintains a subcriticality of at least 5?L AL. The rack arrangement and design prevents accklentalinsertion of fuel between adjacent racks and provides adequate water flow to prevent the water f rom exceeding < 212 F. The racks are structurally able to maintain a Safety Class 2 and Seismic Category 1. The racks are an Essential component performing a passive safety function. Design Description-New Fuel Storage Rack The new fuel an:1 spent fuel storage racks are the same type rack in design, O construction and height. The new fuel storage racks are located in a vault. The vault is a pit in the refueling floor that is fitted with a special cover which is in place when ever fuel is not being processed. The depth of the pit is such that, when fuel is racked, the bail is below the cover's plane, The pit is constnicted the same as the spent fuel poc' except that it contains a drain and is maintained dry. The new fuel storage racks s' ore approximately 1096 of one full core fuel load. - Inspections, Tests, Analys.n and Acceptance Criteria Table 2.5.6 provides a definition of he t inspection, tests, and/or analyses and associated acceptance criteria which will be undertaken for the f uel storage racks. 2.5.6 6/1/92
i I Table 2.5.S: Fuel Storage Racks [ Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. A full rack is subcritical by at least 5% ak, 1. ' Design documentation and records will be 1. The calculated k.S including biases and [
which includes uncertainty value and reviewed to confirm that required criticality ' uncertainties, will not exceed 0.95 under
- associated probability and confidence ' margin has been provided. As-installed normal and abnormal conditions.
level. equipment will be compared to design l documentation; reconciliation analyses will i be performed if necessary.
- 2. The cooling water in the spent fuel storage 2. . Documentation for the as-installed ra" ks 2. The combination of storage racks and pool shall be under 212*F when all storage will be reviewed to confirm that adequate support structure provides adequate flow :
positions are full. cooling will occur. to prevent water from exceeding 212 F. , 3. The structure. Its appurtenances and its 3. . inspections will be conducted of ASME 3. Existence of ASME Code required
- supports shall satisfy the ASME Class, . Code required documents and the code documents and the Code stamps on the
. Seismic Category and Quality Group stamp on the components.: components confirms that the structure i 1 k requirements commensurate with its and components have been designed. classification. analyzed, fabricated and examined in ' i accordance with the applicable ; requirements. i t t f 9 9 9 l __ m.
ABWR Design Document 2.5,7 Under Vessel Servicing Equipmei.' Design Description The functions of the unde -icactor vessel senicing equipment is to: (1) remove aiulinstall control rod drives; (2) install and remove the neution detectors; .uul (3) remove and install RIP .Nfotois. The equi [unent handling platfor m and the CRD handling equipment are poweied pneumaticahv. This equijnnent is chtssified as Non< ssential with a safety class of"Other", has no Seismic requirements and a general Industrial code quality grouping. These characteristics are valid except where noted othenvise. Under-Vessel Platform This is the working surface for equipment and personnel. The plattorm is polar and capable of rotating 360 degrees and is designed in accoidance with the applicable icquirements of OSI1A (Vol 37, No.202, Part 1910N), A!SC, ANSI-C-1, National Electric Code. Spring Reel The Spring Reelis used to pull guide tube seals and detectors during incore i O senicing. ! d [ Water Seal Cap i The Water Seal Cap is used to prevent leakage of the primary coolant during detector replacement. Incore Flange Seal Test Plug The Incore Flange Seal Test Plug is used to determine the pressure integrity of the incore flange O-ring seal. Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.7 provides definition of the inspections, tests and/or analyses together
- with associated acceptance criteria which will be undertaken for the Under-l vessel Senicing Equipment.
/
( 2.5.7 l
-1 6/1/92 1
l
{ Table 2.5.7: Under Vessel Servicing Equipment inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspection, Test, Analysis Acceptance Criteria
- 1. Platform designed in accordance with 1. Examine design data, OSHA regulation, 1. Visual inspection and design data show OSHA (Vol. 37, No. 202, Part 1410N). and visual inspect platform. compliance with OSHA.
- 2. Platform construction in accordance with 2. Examine design data, AISC specifications, 2. Visual inspection and design data show AISC. and visual inspect platform. compliance with AISC.
- 3. Same for ANSI-C-l.
- 4. Same for National Electric Code.
9 9 c 8 O O O
ABWR oesign Document p 2.5.8 CRD Maintenance Facility G Design Description The CRD inaintenance facility is designed and equipped to accolunuxlate the performance of fine inotion connol rod drive (F51CRD) maintenance related activities, including decontamination of the FhtCRD components. perfonnance of acceptance tests, and drive storage. The facility uses manual and/or reinote operation to reduce radiation exposure to plant personnel and to reduce contamination of surrounding equipment during operation. The CRD maintenance f acility is housed in secondary containment ne;u the lower drywell equipment the layout of the facility is designed to incicase the cf ficiency of the personnel, thereby reducing the number of wor kers required. inspections, Tests, Analyses and Acceptance Criteria No entries for this system. O i t i 1 \ l l
\.
1 2.5.8 6/1/92
ABWR Design Document n 2.5.9 Internal Pump Maintenance Facility 8 lV . . Design Description The Reactor Internal Pump (RIP) maintenance facility is located in the reactor building and is designed for peifonning maintenance work on the RIP :notor, including decontamination, in the assembled and disassembled states. The facility is equipped with all tools, crancs and fixtmes needed for inspection of inotor parts and motor heat exchanger tube bundles, Undervessel RIP handling tools are stored outside this area. Inspections, Tests, Analyses and Acceptance Criteria No entries for this systern. l LJ 2.5.9 1- 6/1/92
ABWR Design occument rm 2.5.10 Fuel Cask Cleaning Facility b Design Description This facility is located in two different areas of the plant. The receiving area of the plant will have facilities for: (1) Checking the cask for contamination. (2) Cleaning the cask of road dirt. (3) Inspection of the cask for darnage. ( 1) Attachment of the cask lifting yoke. (5) Removal of head bolts and attachment of head lifting cables. (6) Raising the cask to the refueling floor using the main building crane. Upper Cask Cleaning Facility The refueling floor will have facilities for: O V (1) A deep drainable pit with gate acess to the storage pool, t (2) A under water area for the storage of the cask head and lifting yoke. (3) A area for high pressure cleaning and decontamination. This area must be accessible for chemical and hand scrubbing, refastening the head andsmear tests. l l [ Inspections, Tests, Analyses and Acceptance Criteria l l No entries for this system i I I O 2.5.10 -1 6/1/92 l l - - . -
ABWR oesign Document 2.5.11 Plant Start up Test Equipment Design Dascription No Tier 1 entry for this system. O O 2.5.11 1 6/1/92
ABWR oesign Document (
\
2.5.12 inservice inspection Equipment Design Description A tvpical nuclear power plant facility utiliics a wide range ofinsenice inspection equipment much of which is equipment and matetials used in performance of sisual, surface and volutnetric examinations required by the AShlE Code, Section XI. Automated ultrasonic scanning equipment using multiple angle beam and straight beam transducers may be employed for volumetric examination of areas such as reactor pressure vessel welds and noule inner radii. The data from the automated examination is typically stored on optical disk or othei appropriate recording media foi subscquent computenassisted data analysis. hianual uitmsonic examination equipment may be employed to supplement the automated examination if necessan or to perform the volumenic examination of areas such as Ash 1E Class 2 vessel welds and nonic inner radii. Stanual ultrasonic ex mdriation equipment consists of an ultrasonic instnunent containing analog e dipul oscilloscope-style display and hand-held transduce s. Where more tLn on" angle bma examination is required due to the Class 2 vessel wall thickness, additional manual scans may be performed using ultrasonic transducers adjusted for the required angles of examination. Class 1 and 2 piping welds mac he euunined volumetrically using either A computerized, automated uhrasonic scanning equipment oc using manual O ukrasonic examination e 1tupment. Surflice examinations of ferritic vessels and piping may be perfore3ed using the magnetic particle examination method with either prod or yc ic type equipment. The magnetic particles may be either dry or may be in a wet suspensi,n and may be either Duorescent or colored for viewing in sisible light. Sunace examinations of non-magnetic vessel and piping welds mar be performed using either fluorescent or visible dye liquid penetrant materials. When fluorescent magnetic particles or liquid penetrant materials are used, portable ultraviolet lights are used for viewing. Eddy <urrent probe coils
- hi,en by automated scanning devices with computerized data acquisition l systems may be substituted for surface examinations where the component configuration or radiation condit' 3ns render other surface examinatian techniques impractical or undesirable. Visual examinations of Class 1 and 2 bolting and component supports and attachments on Class 1,2 and 3 piping and components may be conducted directly using simple aids such as mirrors and magnifying glasses. Remote visual examination equipment may be used for l
examination ofinterior surfaces of the reactor vessel and other components. l Rigid fixtures are sometimes used as an aid in performance of the remote reactor visual examinations. It is anticipated that these will be continuing beneficial advances in the technology ofinsenice inspection. As these enhanced technologies become available and proven, they will be applied (as appropriate) to inspection of the certified design. 2.5.12 6/1/92
1 ABWR oesign Document _ l Inspections, Tests, Analyses and Acceptance Criteria No entries for this system. O O 2.5.12 6/1/92
l l ABWR Design Document l O 2.6 REACTOR AUXILIARY IO 2.6.1 Reactor Water Cleanup System I Design Description ! The CUW System removes particulate and dissohed impurities from the reactm coolant by recirculating a ponion of the reaaor coolant through a tilter-demineralizer. The CUW System is designed to process a nominal flow of 2% of l rated feedwater flow, and is designed for 87.9 kg/cm 2g and 302*C. i The CUW System removes excess coolant from the reactor system dudng startup, shutdown and hot standby. The excess water is directed to the mdwaste ! or suppression pool. The CUW System also provides processed water to the l l reactor head spray noule for RPV cooldown. ) l l The CUW System reduces RPV temperature gradients by maintaining l circulation in the bottom head of the RPV during periods when the reactor internal pumps are unavailable. l The suction line through the PCPB contains two motor operated isolation valves. l which automatically close upon receipt of auto. isolation signal from the Leak Detection System and upon actuation of the SLC System. The auto. isolation
]v signal from the LDS consists of the following signals:
(1) lx x reactor water level. l (2) High ambient temperature in CUW equipment room. (3) High temperature differential between the air conditioning duct and in the CUW equipment room. (4) High flow differential between CUW System suction and discharge flows. i The suction valves (containment isolation valves) are designed to isolate against l a maximum differential pressure of 87.9 kg/cm2d within 30 seconds The l inboard valve is powered from Class IE Division 1 AC, while the outboard is fed
- from Class IE Division 2 AC bus.
l The CUW System is classified as a nonsafety system with a major portion of the I system located outside of the primary containment pressure boundary (PCPB) and automatically isolatable. System piping and components within the PCPB, including the suction piping up to and including the outboard suction isolation valve, and containment isolation valves, including interconnecting piping, are l ASME Section III, Seismic Category I, Quality Group A. All onsafety equipment ! is designed as Nonseismic, Quality Group C. Low pressure piping in the filter-i ! 2.6 1- 6/1/92
ABWR oesign Document I deminerali/cr aica, downsticam of the high picom e block vahes,is designed to Quality Group D. The Cl'W System is a single closed loop system (Figure 2.6.1) that takes suction from the icactor sessel bottom head drain line or the shutdown cooling suction line connection to RHR loop "B" CUW Gow passes through a regenerative heat exchangei (RHX) and two parallel nonregenerative heat exchangers (NRHX) to two pumps in parallel . The flow is discharged to two fiber <lemineralizers and returned through the ieg nerative heat exchanger to f eedwater lines "A" and "B". Each puinp, NRHX s of filter 43eminendizer is capable of 509 system capacity operation. Each filter-demineralizer vessel is installed in an indiddual shielded compartment with provisions for handling filter material. Inlet, outlet, vent, drain and other process valves are located outside the tiltei-demineralizer compartment in a separate shielded area together with the necessary piping and associated equipment. Process equipment and cunnols are arranged so that nonnal operations are conducted at a panel from outside the vessel or valve and pump compartment shielding walls. Penetrations through compartment walls are designed so that they preclude direct radiation shine. A remote, manually operated valve on the return line to the feedwater lines in the steam tunnel provides long-term leakage conual and reverse flow isolation is provided by a check uthe in the CUW piping. _ Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.1 provides a definition of the instructions, tests, and,'or analyses together with associated acceptance criteria which will be undertaken .'or the CUW System. O 2.6 1 6/1/92
O O O Table 2.6.1: REACTOR WATER CLEANUP SYSTEM { Inspections, Tests, Analyses and Acceptance Criteria l Certified Design Commitments Inspections, Test, Analysis Acceptance Criteria
- 1. The configuration of the CUW System is 1. Inspection of the as-built CUW 1. As-built CUW System configuration shown in Figure 2.4.1. configuration shall be performed. conforms with Figure 2.6.1.
- 2. Suction line isolation valves automatically 2. Field test will be conducted to confirm that 2. CUW isolates within 30 seconds when the isolate the CUW System upon SLCS the CUW System willisolate upon SLCS SLC System is actuated or when leak actuation, and receipt of auto-isolation actuation and receipt of leak detection detection limit is sensed by closing the signal from the Leak Detection System signal by applying a simulated signal to the primary containment pressure boundary within 30 seconds. isolation logic circuit. isolation valves.
- 3. CUW suction valves are designed to close 3. Procurement records shall be reviewed 3. Certified documentation demonstrates that against the maximum design basis and vendor to conduct shop test relating to the valves can close against a maximur-differential pressure. valve operibility during design basis differential pressure of 87.9 kg/cm2 d within condition. 30 seconds. ,
9 R e O
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*i -' ASME ASME i
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- ISOL. SGNAL
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' - + RCW F M M 3 l P %---
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NRHX
> 4 A 3o A 3 nco
~ W -
---> 1 99+ - + RCW
- 34 gg CONTAINMENT (PCPB) NRHX l
_3+ A - RCW NCt FILTER C TO RADWASTE g DEMINERALIZER g NCl3 ki 31NC - 4 g FILTER
' r TO SUPPRESSION POOL DEMINERALIZER 7
i
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if . - em d v , h ; i I ' e Figure 2.6.1 Reactor Wa4 Cleanup (CUW) System P&lD e ..
i ABWR oesign Document 2.6.2 Fuel Pool Cooline 2nd Cleanup System Design Description The Fuel Pool Cooling and Cleanup (FPC) System (Figure 2.6.2) removes decay heat genciated by the spent fuel assemblies in the spent fuel storage pool. It also maintains ihe watei quality and clarity by r emoving corrosion prmlucts, fission products. and otner iinpurities from the pool. The system also monitors fuel pool water level and n.aintains a water h vel above the fuel sufficient to provide shielding for norraal building occupaary. The FPC System process water flows from the spent fuel storag pool through skinuner weits into two surgt unks. It is drawn from the surge tanks by two circolating pumps arnmged in parallel, and is subsequently discharged thrragh a common header to two filter /deminerali/cr units arranged in pamlici. The dischange water than flows through a common header to two heat exchangers airanged in pandlel and cooled by s cactor building cooling water system, and then returns to the spent fuel storage pool. A bypass line is provided around the fiber / demineralizer portion of the system. Check valves are provide '. in the pool return lines to prevent the pools from siphoning in the event of pipe rupture. The primary operational mode of the FPC System is cooling of the spent fuel pool under normal heat load conditions after a normal refuchng operation. hi this mode, initially both pumps, both heat exchangers, and both filter / demineralizer units are used, llowever, as fuel decay heat decreases, only one p unp and one fiher/demineralizer is used. The fiher/demineralizer units may be bypassed in this mode. The pool temperature is kept at or below 52 C during this operating mode. When the fuel pool is loaded with nmre than the nonnal fuel batch, the system operates in the maximum heat load operating mode. Since the decay heat in this mode exceeds the exchanged heat capacity of the FPC System heat exchangers, RHR System heat exchangers are used to supplement the FPC System heat exchangers. The FPC System operates with both pumps, bcth heat exchangers and both filter /demineralizer units along with two RilR heat exchangers. The pool tempenuure is kept at or below 60 C during this operating mode. After an earthquake, the FPC System is operated with the filter /demineralizer units bypassed. Normal makeup water to the spent fuel storage pool ~ , provided by the non-safety-related Condensate (Mt'WC) Makeup System. A backup to the nonnat makeup system is aFo available from the nonsafety-related Suppression Pool Cleanup (SPCt') Svstem. Additionally, an (mergency safety-related, seismic \ category I makeup water to the spent fuel pool is provided via the FPC System connections to the Residuai Heat Removal (RHR) System, which draws water 2 0.2 1 6/1/92
ABWR nosion ouv.,mem fiorn the suppression pool. a safety-telated wate souice The segment of the i PC , Smem piping h om the RiiR Ststem interhce to the discharge of the fuel pool l is salen n 1.ned. The entine FPC System, with the exception of the filter /demincialiiers. is desi,:m d to Seismic Categort I and Quality Group C standards. The sy. stein can be poweied fiom eithei norinal oll site souices or by the on4hc power source. Tbe FPC System is h>cated in the s cactor buihling, a Seismic Category i, ik ad and tornadoe missile protected structtne. The FPC System pumps :ue motostriven centrifugal pmnps supplying at least T'O m3 /hr at a head of 80m. A low sm tion picuuie at the pump inlet will automatically stop that jnunp. The punip is also protected by an interlock l n a low pump discharge flow. The FPC System heat exchangers are horizontal U- l tube /shell type, each siicd to provide a minimum heat translei rate of 1.65x10 kcal/hr with a coohng water inlet temperature (shell side) of 35'C maximum, and the piocess water inlet teinperature (tube side) of at least STC The filtcr/demineralizer subsystem consists of filter and demineraliter unit 3 and supporting facilities for precoating of sesin, backwashing, and waste remond. 1 The FPC System hieludes conuol room indication to allow for the monitoring and control during design basis operational conditions,i.e. system flows temperatures, pressures, and pool tater level, as well as ndve open/close and pump on/off indication for those instruments and components shown on Figure 2.6.2. with the exception of check udves and manual valves. Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.2 provides a definition of the inspections, tests and/or analyses together with associated acceptance criteria which will be undertaken for the FPC System. O
?C2 2- CJ1/92
j -
}
4 i
% Table 2.6.2: Fuel Pool Cooling and Cleanup System ! ~ ;
i [
- Inspections, Tests Analyses and Acceptance Criteria [
i Certified Design C6mmitment inspections, Tests. Analyses Acceptance Criteria i
- 1. The confitparation of the FPC System is 1. Inspection of the es.bt.L SPC System 1. As-built FPC System conf %uration for !
shown in Figure 2.S_2. con'iguration shall be periormed. those components *hown conforms with Figure 2.62. i i t j 2. FPC pump is capable of delivering flow rate 2. Review of vendor design documents and 2. 8nstalled pump meets design flow l l of > 250 m3 thr against mn differential test results relating to pump performance. requirements. head. L , t l 3. The FPC System operates when powered 3. FPC System functional tes". mil be 3. FPC System is capable of operating when from both normal off-site and on-site performed to demonstrate operation when supplied by either power source. !
- sources. supplied by either normal off-site power or [
l from the on-site power source. i I l 4. The FPC System mechanical equipment, 4. Procurement records and actual equipment 4. Instaffed equipment meets the Seismic : l' ;, excepting filter /demineralizer, is built to shall be inspectml to verify applicable FPC Category I requ:rements and Quality Group Seismic Category I and Out"ty Group C System ccmponents have been designed. C standards. t standards. manufactured and installed per the i relevant stendards. , I ! 5. Control room indications are provided for 5. Inspections shall b4. performed to verify 5. The instruments are present in the control FPC System parameters. presence of control room indication for the room as specified in Section 2.G.2.
- FPC System (Section 2.6.2). ,
s { 6. The RHR System provides a safety-related 6. The FPC and RHR Systems combined 6. The combined system operation transfers f I makeup water source to the fuel pool. fonctional test shall be performed by makeup water from suppression pool to ; aligning the system such that RHR draws the fuel pool. l water from the suppression pool and
~
[ discharges into the fuel pool. L f i P f i
n
}
~ 4 RO i
, ! )SPCU) N l F X
) nHs o)-N i
'A x A
1 A l )RHR C)- (RHR l
-9
)SPCU) yg N (hHR C(
t I, ; N s' SKIMMER SKIMMER i
- i. SPENT FUEL SURGE CURGE / ;7; (puwc(
DRYER- ,7 - - TANK STORAGE TANK SEPARATOR pg T POOL AND
" W FUEL CASK _
i P -*- +j \ y h I c 5 i H v : xi I
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g,'.T A F FPC PUMPS A em FPC FILTER DEM;NERALIZERS
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1 Figure 2.6.2 Fuel Pool Co g and Cleanup (FPC) System ,
ABWR oestun occument . - 2.6.3 Suppression Pool Cleanup System lia # hscription The suppiewinn Pool Cleanup (SPCU) System piovides a i onlinuous punhing i Wate licalille it (sl the suppleMIoin jM M >l. Tlle M stelli l eilH a\rs Valitius kin}}ullikes by fihration, adurption and ion exchange pio( esws. Watei quabtv is inaintained at a quality equal to that of the f uel and equipment pook ( nhei f unctium of the SI CU system aic: (1) Piomics inakeup water to the spent f uel pool and to thr itcactor Building Cooling Wates (RCW) System suige tanks if the normal _ makeup (htt'\ ') is not available. Pa Piovides water ielill to Gic upper pook ]n n>r to o f ochng outage. The SPCU System consisu of a circulation piping, poinp, vahes, wntiols and instnunentation as shown in Figur e 2.6.3.The SPCU pump diaws appr oxiimcely 250 m3/hr (1100 gpm) of suppression pool w.ncr and directs it to the filter demineraliier (shared with the Fuel Pool Cooling and Cleanup system).Ticated water from the filter demineralizer is then delivered eithen hat k to the suppic:.sion pool or to the upper pools via dryer /sepaiato pit. The SPCU System is a Seismic Category I system designed to provide makeup water to the spent fuel pool and the RCW surge tanks Iollowing a seismic oi low of of fsite power event,if the nonnal makeup (511'W) is not avaih ble. During this mode, the filter demineraliter is bypassed and isolated. The SPCU pump takes makeup water from the Condensate Storage Tank (CST) and diverts it to the spent iurl pool and/or to the RCW sorge tanks. Suppievion pool water may also be used for makeup if a loss of coolant accident (IDCA) has not occuned. Power to the pump and associated valves for makeup after a seismic event will be connected to the emergency AC source. The SPCU System has no safety related function except the primary containment isolation function. Following receipt of PCV isolation signals (low water level or high drywell pressure), the SPCU suction valves and discharge valve to the suppression pool are automatically closed to accomplish containment isolation function. The SPCU system suction valves will also close on low suppiession pool water level signal. Containment isolation valves are safety related and are powered from redundant Class IE power sources. The SCPU suction and return lines, fiom the containment up to and including the outboard isolation udte are classified as Seismic Category I and Quality Group IL The remainder of the piping systern is (lassified as Seismic Categon 1 263 cd1/92
-. - - _ _ . .- . .. - - . _ . . _ - - - ~ . - . . - _ _ ._- _ _
ABWR Design Document ( uality Gniup ('. The filter deininerali/c por tion is non seisinic Quality Group The 51'('.l' Sutem design pieuure and teinperattue are as follow: Design Conditions Component Pressure Temperature Peping penetrating PCV up to the 3.16 kg!cm2g 104"C outboard isolation valve (45 psig) (219 F) Outboard isolation valves 16LO'cm 2g 66 C (230 psig) (150"F) SPCU pump, valves and the 19 kg/cm2g 66 C remainder of the piping system (230 psig) (150"F) The SPCU System is provided with instnunentation and contiols to allow SPCU operation over the full range of normal pl mt operation. Inspections, Tests, Analyses and Acceptance Criteria Table '2.6.3 provides definition of inspections, tests, andi or analyses together with associated acceptance criteria which will be undertaken for the SPCU syste rn. O 2 6.3 6/1/92
i [ i N O Table 2.6.3: Suppression Pool Cleanup System , Inspections, Tests, Analyses and Acceptance Criteria < i Certified Design Commitment inspectis.:s, Tests, Analyses Acceptance Criteria [ 1
- 1. The configuration of the SPCU System is 1. Inspection of the as-built SPCU 1. Verification of the as built system is in shown in Figure 2.6.3. configuration shall be performed. conformance with the as-designed ,
configuration (Figure 2.6.3). !
- 2. The SPCU PCV isolation valves isolate 2. Function al testing shall be performed on 2. Valves isolate upon receipt of auto upon receipt of auto isolation signals from the system logic by simulating the auto isolation signal.
the Leak Detection System. isolation signal from the Leak Detection ; i System. t , 3. The SPCU pump capable of delivering 3. The SPCU System functional tests shall be 3. The SPCU System delivering fio u to spent ! maket.p water to the spent fuel pool and to performed to demonstratf. spent fuel pool fuel pool or RCW surge tanks with suction < the RCVJ ::~a tanks. and RCW surge tanks makeup. from the suppression pool and/or Condensate Storage Tank. ! d t Y 4. The SPCU System capability to operate on 4. SPCU functional testing shall be perfo<med 4. Satisfactory SPCU operation with power j on-site emergency .AC power source to demonstrate operation when supplied supplied from on-site emergency AC l from on-site emergency AC power. power, i i
- i i !
i i i !
- e !
1 5
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i ' t I l
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m w TO SPENT m FUEL POOL ' - ' SPCU RCW 3 3 TO RCW 2 --- SURGE TAfJK ' ISOLATIOfJ TO D/S SIGt4AL PIT 1 > 4 RCV.' SPCU 3 3 FPC SPCU M 3 3 A FROM F'D UNIT *A* r1 l- >g< - - -l M :g: < -- i i ! ASME ASME SPCU FPC l CLASS 2 CLASS 3 3 ?JC b ! l l u A FROM FO
; utur B-PRIMARY CONTAINMEffT l __
ISOLATION ISOLATION SUPPRESSION SIGtJAL StGPJAL 3 SPCU FPC POOL FE 4 M % M
/ i 6; ,
e l F :: - > N O A T O F/D M LM JIT *A* l ~ 1_-m._-. FROM CST. - > >l4 M T2 HPCF SPCU FPC HPCF SPCU 2 3 Figure 2.6.2 Suppre n Pool Cleanup System
ABWR oesign Document
- 2.7 Conttol Panels
(' 2.7.1 Main Control Room Panel Design Description The Slain Connol floom Panel a comprised of separate stand.alone modules (e.g., Stain Control Panel, Large Olsplay Panel). Each panel module is seismitally qualified and provides grounding, and electrical independence and phpiral sep.uation between safety divisions and between safety divisions and non-essential components and wiring. Electric al powei to divisional" Vital" components is from the Vital AC Connol Power or hattery of the same electrical division. Power to the non-essential
" Vital" components is fiom the non-essential Vital AC Control Power or non-essential battery. Divisional, non4 ital coinponents are powered from the respective divisional AC Instnnnent Power and non<livisional, non-vital component.s are powered from non+ssential AC Instrument Power.
The Stain Control Roorn Panel and other main control room operator interfaces are designed to provide the operator with infort tion and controls needed to safely operate the plant in all operating modes, inclnding str rtup, refueling, safe shutdown, and maintaining the plant in a safe shutdown condition. The piocess O to be used during the implementation stage willincorporate accepted lluman Factor Engineering (life) principles in implemer. ting the hiain Control Room lluman-System luterface (llSI). Inspections, Tests, Analyses and A:ceptance Criteria Table 2.7.1, together with the Design Acceptance Criteria (DAC) in Table 3. l. defines the design process to be used for the 51ain Control Room Panel and other main control room operator interfaces. 2.7 1- 6/1/92
Z Table 2.7.1: Main Control Room Panels Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria
- 1. The Main Control Room Panels are 1. Inspections cf the as-built design 1. Panels are seismically qualified and seismically qualified. documentation and installed equipment installed.
will be performed. l l l
- 2. The Main Control Room Panels design 2. Inspections o' the as-built design 2. ElectricalIndependence and physical l l provides grounding and electrical documentation and installed equipment separation. and grounding of components independence and physical separation wi!I be performed. and wiring is provided.
between divisions and between divisions and non-divisional components and wiring.
- 3. The Main Control Room Panel components 3. Inspections of installed equipment will be 3. Panel components are powered from identified as " Vital" are powered from their performed. power supplies consistent with component I respective division or non-essential Vital classification and divisional assignment.
9 AC Control Power or battery. Non-vital components are powered from their respective divisional or non-essential AC Instrument Power Supplies.
- 4. A Design and Implementation Process, 4. See Table 3.6. 1. Design and implementation of the Main directed by a dedicated Man-Machine Control Room Panels and other main Interf ace System (MMIS) Design Team, will control room operator interfaces comply govern the implementation of the Main with the cri+2ria defined in Section 3.6.
Control Room Panei and other main control room operator interfaces. Human Factors Engineering principles wirl be employed to provide a Human-System Interf ace (HSI) for the Main Control Room Panels. 9 8 9 O O
ABWR oesign occument 2.7.2 Radioactive Waste Control Panel No entn. Covered by item 2.9.1. O } i lO 2.7.2 6/1/92 1 i_..___.. _ . . . . _ _ _ , . - _ _ _ _ . . . . . . _ _,. - . . _ , ___ - - . ~ . _
. . = _ - - - - . - -_- - . _ _ _ - _ -. .- - - _ _ . _ _ . - _ - _ . . _ _ - - - --
ABWR oesign oocument 2.7.3 Local Control Panels Design Description 1.oc al panels. ( ontrol boxes. aial insinonent racks are provided as protective housings and/or support muctuies for elecuical and elecuonic c(luipinent to f acilitate system oper.uions at the locallesel. They are designed f or unifonnin using sigid steel structures capable of maintaining structuralintegrity as required unde: seismic and p' ant dyn.unic conditions. The term " local panels" is assumed to include local control boxes in this document. l.ot al panels and racks used for plant protection systems are classified as Safety Related. They are located in a safety class structure in which there are no potential sources of missiles or pipe breaks that couldjeopardize redundant modules. Each salcty related panel / iack is seismic (Category 1) qualified and piovides grounding. and electrical independence and physical separation between safety divisions and nonessential components and wiring. Electrical power to divisional panels / racks is from AC or DC power sources of the same division as that of each panel / rack itself. Power to the nonessential panels / racks is from the non essential AC and/or DC sources. inspections, Tests, Analyses and Acceptance Critoria s Table '17.3 piovides a definition of the inspections. tests, anci or analyses l together with associated acceptance criteria which will be ur. artaken for the local control panels and racks. l i l l l l 2.7.3 1 6/1/92
' ;iI ; ir ic s t
t n m n e . is e ntn e n oe s . o p mpm om ed ogn a n k r ale sl a cs an t la m ic c sf yo o f r d ci s eht s r ia ewl 4 r, h w t a ri p g i t r dd nn ,i n on peion C aa ed . t e c nde n esis asi v ri c sd e u d n :ee d oi nd ts o i a nf ngo r v t al i e ncd p pa e u pdr en p es na ne c c de q das olin A t 1 n ,i ng ppo la1y i m ua pit e r r o laionr o sc yg ct i i ai c rf i r r w l ei tete tc a e s f a epd en nws aoa i a S(aC l E sa Ppc l r e . . . t 1 2 3 s i r k C c a e e R c t t b n n n h d a e e n t m m iw a p ip ip t s e s n l e c c i s y nqu nqu e ge ge m n A la i sd isd ip a n ee ee u P dn A dl d! q l a ts lt la lt !a e o r iu sn t iu s t d t s se b i bn le n e - - i la o s T, sd . sd a nd. C l a l y s a n ito n and ea e h nmhf nm t f ior t e a ior e t in f s o O c o A c e otafo otafo s . p nt r nt r nd L t s s n o nepe o nepe oe s itcrm 3 e i it cme e ub e ub it cme eo 7 T p cl s oi pcl pf s r s od e l 2 s In d w In dw Inp e n l o . . . b i t 1 2 3 a c T e p d d s s s n e l! e n k ed a r c id n , e n i a v a,na sl w n- a r o eon oop t dn . r pne n pcio s i h e ade sevi nis er r ot it m lsi f i kc d idiv a sh sni t m el n a a e
/r pen n nd -
t ni ow m a pq u ls e eo est o ) ed w n g. nin ve nnt oi C laI cy ai ed n pd is t ig n loro l pl b ni a cn w ai r mes vn s d g ee aios cr t n3 oioct. ct e k ec D t t lotc iao.r cpsnee d lae(C a f lerai isa a s oepv s / er sm r u _ ie f
- r. c yi n g,eis d tr; te rb ig n g n e a e las ni t
en on _ it r e s m C e f ai s es isi ed lac phi s t t na e de uoyt niep sw m lamek c h re T a 1 h T gpb c 2. r h eo oosa Lf er 3 r sc 9
} 9 r
. . _ _ .- . _._ _ . - - . - . _ _ - .. . _ _ _ - = - - - - - _ _ _ -
ABWR 0: sign Document 2.7,4 Instrument Racks No entry. Covered under Itern '2.7.3. O O 2.7.4 1 f/1/92
ABWR oesion Document 2.7.5 Multiplexing Design Description Essential Muldplexing System The Essential Multiplexing System (EMS) prosides distributed data acquisition and (ontrol networks to support the monitoring and control of the plant standby safety systems. EMS comprises electrical devices and circuitry, such as Remote Multiplexing Units (RMUs), transmission lines, and Contr ol Room Multiplexing Units (CMUs), that acquire data from remotc process sensors and discrete monitors located within the plant and multiplex the signals to Saft system logic and Control (SSI.C) equipment in the main control room an ., SSI.C processes the input signals and multiplexes output control sign:ds to the final actuators of driven equipment associated with safety systems. EMS is divided into four disisions of equipment, each with independent control of data acquisition, multiplexing, and control output functions. System timing is asynchronous among the four divisions. No common clock signalis transmitted among the disisions of multiplexing and no timing signals are exchanged. i lloth analog and discrete sensors are connected to RMUs in local areas, which p perform signal conditioning, analog-to-digital conversion for c ontinuous ( process inputs, change-of-state detection for discrete inputs, and message formatting prior to signal transmission. The RMUs are limited to acquisition of sensor data and the output of tontrol signals. Trip decisions and other control logic functions are performed in SSI.C processors in the main control room area. The RMUs transmit serial, time-mult i plexed data streams representing the status of the plant variables via liber optic cables to the control room CM Us. Ihta transmission is made over dual redundant channels. EMS design features automatic self-test and automatic reconfiguration after failure of one ch:mnel (either a cable break or device faihire). The system returns to normal operation after reconfiguration within one full scan period. If an RMU or CMU has failed, that unit will be removed from senice. Faults and their location are annunciated to the operator in the main control room. The CMUs demultiplex the data and prepare the signals for use in interfacing controllers of SSI.C or monitoring systems such as the process computer or display controllers. After the input data is processed in SSI.C, the resulting trip logic decisions are tra 'tted (for Engineered Safety Features functions only) as a serial, time-multiped data stream via EMS to RMUs in the local areas, where the digital data is converted to contact closures or other signals for actuation of motor control centers or other device controllers.The data streams aw dual redundant to prevent inadvertent ECCS equipment actuation after a hardware or software fault in one channel. The data reaching the RMUs is compared in 2-out-of-2 voting logic to confirm final output to the actuators. 2.7.5 1- 6/1'92
ABWR oesion Document . 4 Data can be transferred to non-safety systems for control or display through g l isolating fiber-optic data links and buffering desices (gateways or bridges, if W i required). Data transfer is made such that failures on the non safety side cannot inhibit operation of safety-related logic f unctions. Data cannot be transmitted from the non safety side to ENIS. ENIS is capable of data transfer at rates sufficient to satisiy the system time response requirements of safety system functions. Data throughput capability shall be at least 10 megabits per second. ; l ENIS starts and runs automatically upon application of system power, regardless l of the sequence in which power is applied to indisidual controllers. EN1S and SSI.C automatically establish conununications by detection of correct ruessage passing. Logic is prosided to prevent equipment activation outputs from occ urring until stable plant sensor data and inteilock permissive data are being received. loss of power causes a controlled transition to a safe state without trans:ents occurring that could cause inadvertent initiation or shutdown of driven equipmer, EN15 equipment is classified as s,re ty relateJ, Llas; ' E, and is scismically q ui.lified. h Testability ENIS includes test facilities in the control room that will monitor data transmission to ensure that data transport, routing, and timing specifications are accurate. Ilit error rate of each EMS network shall be better than I crror in 10* Out-of-tolerance parameters detected on-line for a particular input signal will result in an inoperative condition for that input into the trip logic processors of SSLC. Inspections, Tests, Analyses and Acceptance Criteria Table 2.7.5 prosides a definition of the sisual inspections, tests and analyses, together with associated acceptance criteria, which will be used by SSI.C. O 2.1.5 W1!92 d
I l 0 o o O Table 2.7.5: Multiplexing inspections. Tests, Analyses and Acceptance Criteria inspections. Tests. Analyses Acceptance Criteria Cer*ified Design Commitment Visualinspection of the installed 1. EMS contiguration is in accordance with
- 1. Four divisions of independent and 1.
equipment will confirm the identity and equipment arrangement shown in section redundant EMS instrumentation acquire 3.4 SSI.C ITAAC. The figures indecate the and transmit the safety-related sensor location of EMS instrumentation, equipment panels, and their required relationship of EMS to other inputs and control functions of the plant safety system processing equipment. standby safety systems and auxiliary interconnections: supporting systems. Visual inspection of installed equipment. 2. Installed configuration of EMS conforms to
- 2. EMS panels and processing equipment are 2.
Class 1E. safety-related, and seismica!!y test records, and analyses based on certified commitment. qualified. equipment location will confirm the quaiification state's of EMS. ! Inspections of fabrication and installation 3. The instaffed EMS equipment conforms to I
- 3. The four divisions of redundant 3.
instrumentation are physically and records and construction drawings or certified commitment. electrically separated from each other. visual field inspections of the installed EMS There are no interconnections among equipment will be use * .o confirm electrical and physical separation.
- c. divisions of EMS. Data communications to the process computer or display contro!!ers shall use an iso!ating transmission medium such as fiber optic cables.
- 4. System tests will be conducted after 4. The installed instrument channels are
- 4. The RMUs and CMUs in each operational with the power sources instrumentation division are powered installation to confirm that the electrical power supply configurations are in specified in the certified commitment.
indepandently from the divisional plant DC sources (Class 1E 125 VDC) compliance with design commitments. Factory tests for EMC will be conducted in 5. EMC performance of EMS is considered
- 5. EMS meets Electromagnetic Compatibility 5.
a controlled environment on individual m eptable if tests confirm that (EMC) requirements. Protection is provided electrormgnetic fields. static discharges. against the effects of: EMS equipment and on the integrated system configuration. and electrical surges do not atiect system
- a. Electromagnetic in' 'rference (EMI) capabilly to acquire and condition data,
- b. Radio Frequency Inten'erence (RFI) transmit formatted data, receive control
- c. Electrostatic Discharge IESD) EMC tests will also be conducted on the installed EMS configuration in the normal signals and send control outputs to final
- d. Electrical surge (Surge Withstand actuators.
Capability (SWC)1 plant operating environment. a w
t Table 2.7.5: Multiplexing (Continued) 6 Inspections, Tests, Analyses and Acceptance Criteria l Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria
- 6. EMS includes test facilities that will 6. Preoperational tests witi be conducted on 6. Operability of the instal:ed EMS equipment monitor data transmission to ensure that the installed EMS equipment. These tests is considered acceptable under the data transport, routing. and timing are will confirm the basic functionality of each following conditions (for each divison):
accurate. multiplexing component. The tests will a. Monitued output signals match include simulation of typical input simulated input signals for accuracy of parameters and monitoring of received signal conversion and transmission these transmitted parameters. Random time. i simulations will be used to test bit error b. Bet error rate is <10'9 rate, which shall be determined to be <10- c. Simulated data errors are detected and 9- annunciated to operator.
- 7. Full system test of EMS with S ,LC and 7. Preoperational tests will be conducted to 7. EMS support of the interfacing safety !
other interfacing systems connected verify safety system logic functions of each systems is considered acceptable if reactor ! confirms EMS response to safety system interfacing safety system. These tests will trip. containment isolation, and ECCS tests specified in each interfacing system verify support of SSLC and the saiety response of the installed equipment meet ITAAC. Testing is conducted on the four systems for scram capability, containment the acceptance criteria stated in each
- . divisions of EMS /SSLC simultaneously to isolation capabihty, and ECCS initiation interfacing system ITAAC. The response ;
verify 2-out-of-4 system operation. capability. The tests will include time of each control action and trip output , demonstration of ability to meet stated is within performance limits of each delay times and maximum response times. interfacing system. Tests will be conducted such that each display, alarm, annunciator, or other status Performance of SSLC for these same tests indicator for each system is shown to be also confirms EMS performance. , functional. See section 3.4 SSLC ITAAC. item 7 for the scope and method of testing. j i e , 6 ~ 6 9 9
Table 2.7.5: Multiplexing (Continued) [ in Inspections, Tests, Analyses and Acceptance Criteria inspections Tests. Analyses Acceptance Criteria Certified Design Commitment Tests will be conducted to verify that 8. EMS response to foss of power is
- 8. EMS provides safe-state response to loss 8.
graceful degradation of EMS system acceptable for the following conditions: of power source. a. Loss of one division of power does not outputs occurs upon momentary or long-term loss of one division of the DC power cause false output trip or inadvertent source or power to individual EMS initiation of final system actuators. compwnts. Tests will also confirm that Loss of power and foss of divisional reinitia zation of system or component mp signals are annunciated. after power is restored does not impair b. Loss of power to individual component normal system function. produces a safe-state output condition without extraneous false outputs (normally-energized outputs de-energize, normally-de-energized outputs remain deenergized).
- c. Restart (initialization) of component or system upon recovery of power does i not cause inadvertent output actior l (outputs remain in safe-state conddion T'
- 9. Preonerational tests will be conducted to 9. EMS response to instrument or cabie
- 9. EMS is fault-tolerant in each division and failure is acceptable for the followmg l provides capability for automatically verify that a single failure of a multiplexing component does not ims, air total system conditions- l reconfiguring after failure of an RMU, i function. Faults will be simulated and the a. A single cable break does not affect CMU, or interconnecting cable. r'etwork operation.
response raonitored Tests specified in item 6 will be repeated to confirm b. Loss of one RMU or CMU removes that operability of network. unit from service; network continues normal operation.
- c. Fault occurrence and notice of reconfiguration is displayed to operator.
- s
ABWR oasign oocument 2.7.6 Local Controf Box No entiy. Covered under liein 2.7.3. l O ; I O 2.7.6 -1 6/1/92 r
ABWR onion Document 2.8 Nuclear Fuel 2.8.1 Nuclear Fuel Design Description (Including Loose Parts Monitoring) Fuel design for the ABWR is not within the scope of the certified design. It is intended that the specinc fuel to be utilized in any facility which has adopted the cenified design be in compliance with U.S. NRC approved fuel design criteria. This strategy is intended to permit future use of enhanced / improved fuel designs as they become available. However, this approach is predicated on the as,sumption that future fuel designs will be extensions of the basic fuel technology that has been developed for boiling light water reactors. Key characteristics of this established LWR fuel technology are: (1) Uranium oxide based fuel pellea. , (2) Zirconium-based (or equhalent) fuel cladding. (3) All rnaterial selected on the basis of BWR operating conditions. (4) Multi-rod fuel bundles in an N lattice. (5) Fuel bundle inlet orificing to control bundle flow rates, core flow I- O distribution, and reactor coolant hydraulic characteristics. The following is a summary of the principal requirements which must be met by the fuel supplied to any facility utilizing the certified design. The ABWR design provides a loose Parts Monitoring System (LPMS) aimed at protecting the fuel against the potential efTects ofloose parts entrained in the reactor coolant flow. A discussion of the LPMS is included in this section.
- General Criteels (1) NRC approved analytical models and analysis procedures are applied, t
(2) New design features are included in lead test assemblies. (3) The generic post-irradiation fuel examination program approved by NRC is maintained. 2.8 1- W1/92
ABWR Design Documont Thermal Mechanical The fuel design thermal. mechanical analyses are performed for the following conditions: (1) Either worst tolerance assumptions are applied or probabilistic analyses are performed to determine statistically bounding results (i.e., upper 957c confidence). (2) Operating conditions are taken to bound the conditions anticipated during normal steady state operation and anticipated operational occurrences. The fuel design evaluations are preformed against the following criteria: (1) The fuel rod and fuel assembly component stresses, strains, and fatigue life usage are evaluated to not exceed the material ultimate stress or strain and the thennal fatigue capability. (2) Mechanical testing is performed to ensure that loss of fuel rod and assembly component mechanical integrity will not occur due to fretting wear. (3) The fuel rod and assembly component evaluations include consideration of metal thinning and any associated tempemture increase due to oxidation and the buildup of corrosion products to the extent that these influence the material properties and structural strength of the components. (4) The fuel rod internal hydrogen content is controlled dtuing manufacture of the fue) rod consistent with ASTM standards. (5) The fuel rod is evaluated to ensure that fuel rod bowing does not result in loss of fuel rod mechanical integrity due to boiling transition. (6) Loss of fuel rod mechanical integrity will not occur due to excessive cladding pressure loading. (7) The fuel assembly (including channel box), control rtxi and control rod drive are evaluated to assure control rods can be inserted when required. These evaluations consider the effect of combined safe shutdown earthquake (SSE) and loss-of<oolant accident (LOCA) loads. (8) 1.oss of fuel rod mechanical integrity will not occur due to cladding collapse into a fuel column axial gap. 2.8.1 2 6A!92
ABWR oesign Document (9) la u of f uel iod mechanical integnin will not occui due to pellet-s < l.ohhnu inet h. uncal intenation. Nuclear (1) A negatise 1) oppler scacthity coellicient is inaintained f or any opei ating cond. tion. ($ $ llegative cole IHoderatoT void learticity f oclfiCient f esulting IloW boiling in the active flow thannels is maintained for any operating conditions. (3) A negative nM/deratol telnpenitute coelliciellt is Illailltailled above 1101 st a n dhv. (4) For a super ptompt critical reactivity insertion accident originating fiom any operating condition, the net prompt reactivity feedback due to prompt heating of the moderator and fuel is negative. (5) A negative pow er coef ficient, as deterinined by calculating the reactielty change. due to an incremental power change from a steady-state base [mwer level, is maintained for all operating power levels above hot i f. standhv. . l (6) The plant mects the cold shutdown margin requirement. l (7) The effective multiplication factor Ur fuel designs stored under normal and abnonnal conditions is shown to meet fuel storage limits by dennonstrating that the peak uncontrolled lattice k-infinity calculated in t a normal reactor core configurations meets the limits for the storage l rac ks. l l Hydraulic Flow pressure drop characteristics are included in the calculation of the Operating Limit MCPit Because of the channeled configuration of BWR fuel assemblics, there is no bundle-to-bundle cross-Dow inside the core, and the only issue of hydraulic compatibility of various bundle types in a core is the bundle inlet Oow rate variation and its impact on margin-to-thennal limits. The coupled thennal. hydraulic-nuclear analyses performed to determine fuel bundle flow and power distribution uses the various bundle pressure loss coefficients to detennine the Dow distribution required to maintain a total core pressure drop boundary u condition to be applied to al' fuel bundle. The margin to the thermal limits of 2.8.1 d/92
ABWR Design oocwnent each fuel hundle is determined using this mnsistein set of cah ulated bundle flow and power. Loose Parts Monitoring System (Design Description) The 1 oose Parts hlonitoring System (1.F.\15) is dedgned to proside detection of loose metallic parts within the scactor piessuic vessel Detection I loose par ts tan piovide the time icquired to avoid oi nitigate saf ety-related oarnage to on ill tif tlllctio!M of pritualy sysicin (olilponents. The LP.\lS detects strm ture borne l sound that can indicate the presence of loose pas ts impacting against the t eactor pressure vesselinternals. The system alarms when the signal amplitude exceeds preset limits. The LPhtS detection system can ev;duate some aspects of selected signals, llowever, the system by itself will not diagnose the presence and location of a loose part. Review of LPh!S data hv an experienced LPS1 engineer is required to confirm the presence of a loose part. The LPhtS continuoudy monitors the reactor pressta e vessel and appurtenantes for indications ofloose pans. The LPats consists of sensois, cables, signal cotiditioning equiprnellt, alarming monitor, signal analysis and data acquisition equip!nellt, a!)d Calibratioll equipillent. The alarm setting is set knv enough to meet the sensitivity requirements, yet is designed to discriininate between Ilorillal haCkgrollnd noises alid the loose f.att ilnp:tct sigital to minimi/c spulious alar 1Hs. The array of LPhtS sensors consist of a set of sensor channels that are strategically mounted on the external surfare of the primary presmre boundary at various elevations and azimuths at natural collection regions for potential loose parts. General mounting locations are at the a) main steam outlet nonle, b) feedwater inlet nonle, c) core spray noules, and d) control rod drive housings. The online system sensitivity is such that the system can detect a metallic loose part that weighs between 0.25 lb to 30 lbs and impacts with a kinetic energy of 0.51t-lh on the inside surface of the reactor pressure vessel within 3 feet of a sensor. The LPhtS frequency range ofinterest is typically from 1 to 10 kHz. Frequencies lower than 1 kHz are generally associated with flow induced vibration signals or flow noise. The LPh!S includes provisions for both automatic and manual start-up of data acquistion equipment with automatic activation in the event the preset alert level is reached or exceeded. The system also initiates an alarm to the control room personnel when an alert condition is reached. O 2.8.1 6/1/92
ABWR Design Document Inspections, Tests, Analyses and Acceptance Criteria Table 2.8.la provides a definition of the inspec tions, tests, ain:for anal >ses together with associated acceptance aiiriia which will be undertaken for the fuel that will be pio;x> sed for the Iaciht). Tables 2.8.lb provides a definition of the inspections, tests and/or analyses together with associated acceptance criteria which will be utulertaken for 1.oose l' arts .\f onito ing Systern. i I I d [ l f 1 ( l I l [ l m l l 2.8.1 5- 6/1/92 l l
} Table 2.8.1a: Nuclear Fuel Inspections, Tests, Analyses and Acceptance Criteria Design Criteria inspections. Tests Analyses Acceptance Criteria
- t. Fuel design thermal-mechanical analyses 1. Manufacturing specifications and design 1 ihe analyses are determined to be are performed using either worst tolerance drawings will be reviewed to ensure proper applicable to the fue: being used in the assumptions or probabilistic analyses to input parameters are used in the analyses. core. ;
determine statistically bounding results (i.e., upper 95% confidence).
- 2. Fuel design thermal-mechanical analyses 2. Operating limits defining the maximum 2. Eva'uations demonstrate that the fuel are performad for operating conditions allowable fuel peilet operating power level satisfies specified acceptable fuel design anticip ted during normal steady-state as a function of fuel pellet exposure, will be u enits for the applicable thermal limits (e.g.,
operation and anticipated operational established to ensure actual fuel operation MAPLHGR). occurrences (AOOs). is maintained within the analysis bases.
- 3. The fuel rod and fuel assembly component 3 Exposure-dependent, th ermal-mechanical 3 a. For stress or strain, the Design Ratio is stresses, strains, and fatigue life usage are analyses wi!! be performed. s 1.0, p evaluated to not exceed the material v.here ultimate stress or strain and the thermal Duign Ratio -E'fectly1Str.ess f a*'
- capability. Stress Limit orEffective_ Strain Strain Limit
- b. For fatigue, the calculated fatigue duty is less than the material fatigue capriility.
J. Mechanical testing is performed to ensure 4 Testi ng or operating experience wii? de 4. The testing or experience demonstrates that loss of fuel rod and assembly used to determine whether the fuel that the fuel will not fail due to fretting. j component mechanical integrity will not assembly is susceptible to significant occur due to fretting wear in an fretting wear. environment free of foreign material.
- 5. The fuel rod and assembly component 5. The effects of cladding oxidation and 5. The evaluations demonstrate the fuel evaluations include consideration of metal corrosion product buildup on the fuel rod design is adequate for resisting the effects thinning and any associated temperature surface will be included in the evaluations of metal thinning and any associated
, increase due to oxidation and the buildup as appropriate. temperature increases due to oxidation d of c >rrosion products to the extent that and corrosion product buildup.
O these 5 fluence the material properties and structural strength of the components. O 9 9
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Table ~1.la: Nuclear Fuel (Continued) t inspections, Tests. Analyses and Acceptance Criteria . Design Criteria Inspections. Tests. Analyses Acceptance Criteria r
- 13. A negative core moderator void reactivity 13. The core moderator void reactivity 13. The calculated core moderator void coefficient resulting from boiling in the coefficient wi!I be calculated for an reactivity coefficient is negative. t active f!ow channels as maintained for any equilibrium core of the fuel design, at the operating condition. limiting point in the cycle, and covering a!I expected modes of operaticn.
- 14. A negative moderator temperature 14. T he moderator temperature coefficient 14. The calculated moderator temperature coefficiejnt is maintained above hot will be calculated for an equilibrium core of coefficient is n:egative above hot standby.
standby. the fuel design, at the limiting peint in the i cycle, and covering all expected modes of operation. 4
- 15. For a supor prompt critical reactivity 15. The prompt reactivity feedback will be 15. The calculated prompt reactivity feedback insertion accident originating from any calculated for an equilibrium core of the is negative.
3 operating condition, the net prompt fuel design, at the limiting point in the react;vity feedback due to prompt heating cycle, and covering all expected modes of , of the moderator and fuel is negative, operation. ; i
- 16. A negative power coefficient, as 16. The sign of the power coefficient will be 16. The power coefficient is negative.
determined by calculating the reactivity determined for an equilibrium core of the i change, due to an incremental power fuel design, at the limiting Joint in the , change from a steady-state base power cycle, and covering all expected modes of ! level, is maintained for all operating power operation. ; levels above hot standby..
- 17. The plant meets the cold shatdown margin 17. The cold shutdown margir wm t,e 17. The calculated cold shutdown margin is i requirement. calculated each cycle for the most reactiv6 greater than the value given in the condition with the most reactive control Technical Specifications.
rod in the full-out position. r j 13. The effective multiplication factor for fuel 18. The peak uncontrolled lattice multiplication 18. The effactive multiplication factor under dusigns stored under normal and f actor will be calculated in the normal normal conditions is less than 0.90 for abnormat conditions is shown to meet fuel reactor core configuration or the effective regular density racks and less than 0.95 for t e storage limits. multiplication factor of fuel stored under high density racks, and less than 0.95 for !
.3 normal and abnormal conditions will be abnormal conditions for regular and high calculated. density racks.
i 9 9 9
i , ii , . l l l I j Table 2.8.1a: Nuclear Fuel (Continued) t Inspections. Tests, Analyses and Acceptance Criteria l l inspections, Tests, Analyses Acceptance Criteria Design Criteria
- 19. Catculations of the OLMCFR willinclude 19. Flow pressure drop characteristics are
- 19. Flow pressure drop characteristics are verified as being included in the OLMCPR included in the calculation of the Operating flow pressure drop characteristics.
calculation. Limit Minimum Critical Power Ratio (OLMCPR). o a
)
___ i
5 Table 2.8.1b: Loose Parts Monitoring System inspections, Tests, Analyses and Acceptance Criteria CertifWd Design Commitment inspections, Tests, Analyses Acceptance. Criteria
- 1. Ar. LPMS is provided with detectors 1. Visual inspections will be conducted of the 1. An LPMS has ben provided.
located at natural collection regions for as-built facility to confirm that the LPMS is loose parts.The system includes the in place and operational. necessary signal processing sad related equipment.
- 2. 1 he LPMS shat! be capable of detecting a 2. System calibration tests will be performed 2. It must be shown that the LPMS can detect metallic loose part that weinhs from 0.25 lb to demonstrate system sensitivity. a metallic loose part that weighs from 0.25 to 30 lbs and impacts with a kinetic energy Ib to 30 lbs and impacts with a kinetic of 0.5 ft-Ib within 3 feet of each sensor. energy of 0 4 ft-Ib within 3 feet of each sensor.
l t A
=
0 9 O O
ABWR Desinn Document g- 2.8.2 Fuel Channel
\'
Design Description F hanimi design foi the AliWR is not within scope of the certified design it
< < ended that the specific fuel channel to be utili/cd in any f adlity which has ac-pied the certified design be in compliance with U.S. NRC approved fuel channel <lesign caitei;a. This strategy is intended to pern.it iuture c ; of enhanc edfimproved con'rol rod designs as they become available. However.
this appio,ch is predicated on the assumption that f utuie f uel channel designs will be miensions of the basic technology that has been developed for light water reactors. The ke) characteristic of this established ilWit fuel cl annel technologs is the use of ziironium based (or equivalent) fuel channels which pr eclude ( ross-flow in the core iegion. The following is a stunmany of the principal requirements which must be uwt by the f uel channel supplied to any facility using the critified design. G~ner al Criteria (1) The material of the fuel channel shall be shown to be compatible with the scactor environenent. O v (2) The channel will be evaluated to ensure that channel deflection does not preclude control rod drive operation. 1 ! (3) The efIccts of channel bow will be iccluded in the fuel rod critical power evaluations. Inspections, Tests, Analyses and Acceptance Criteria Table 2.8.2 provides a definition of the inspection, tesis and /or analyses together with associated acceptance crittria which will be undertaken for the l fuel channel that will be proposed for the facility. 1 i n%s 2.8.2 6/1/92
% Table 2.8.2: Fuel Channel i ~
Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The material of the fuel channel shall be 1. Testing or operating experience wi!! be 1. The testing or experience demonstrates shown to be compatible with the reactor used to determine whether the fuel the traterials are adequate.
environment. channel materials are compatible with the reactor environment.
- 2. The channel will be evaluated to ensure 2. The channel will be evaluated to detemine 2. Calculated channe; deflections a e not of that chcanel deflection does not preclude the expected amount of channel deflection. an amount great enough to preclude control rod drive oporation. control rod insertion.
- 3. The effects of channel bow will be included 3. An allowance will be included in the critical 3. Verification demonstrates that an in the fuel rod critical power evaluations. power calculation to account for the effects allowance for channel bow is included in of channel bow. the fuel rod critical power evaluations.
U m J O O O
ABWR oesign Document 2,8.3 Control Rod Design Description Cnntrot rod design for the AltWR is not within the scope of the ceitilied design. It is intended that the specific conu oliod to be utilized in any facuity whic h has adopted the certified design he in compliance w th U.S. NRC approved conuol ind design criteria. This strategs is intended to permit future use of enhanced - improved control rod oesigns as thev he( ome availahic. However, this approach is predicated on the assumption that f uture i onts iod designs will he extensions of the basic technology that has been oeveloped for light water icactors. Key characteristics of this established ilWR control rod technology are: (1) Control rods perform dual functions of powei distribution shaping and reactisity control. (2) The control rod has a crucif orm cross-sectional envelope shape. (3) The control rod has a coupling at the bottom for attachment to the } control rod drive. (4) The control rod has an upper bail handle for transporting. (5) The cr aciform cross section contains neutron poison materials which are either contained within or as part of the control rod structure. The following is a summary of the principal requirements which must be r.ut by the control rod supplied to any facility utilizing the certified design.
~
General Criteria (1) The control rod stresses, strains, and cumulative fatigue shall be evaluated to not exceed the ultimate stress or strain of the material. (2) The control rod shall be evaluated to be capable of insertion into the core during all modes of plant operation within the limits assumed in the plant analyses. (3) The material of the control rod shall be shown to be compatible with the reactor environment. (4) The reactivity worth of the control rod shall be included in the plant core analyses. 2.8.3 -1 S/1l92
ABWR Design Document (5) 1. cad Stureillance progr.uu shall be implemented if a change in design features such as new absorber material or structural material no" previously used in icactor cores could impact the f unction of the control rod. Inspections, Tests, Analyses and Acceptance Criteria Table 2.8.3 provides a definition of the inspection, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the control rod that will be proposed for the facility. O O 2.8.3 6/1/92
rm % U i Table 2.8.3: Control Rod { Inspections, Tests, Analyses and Acceptance Criteria Design Criteria inspections, Tests, Analyses Acceptance Criteria
- 1. The control rod stresst v strains, and - 1. Evaluations of loads due to shipping, 1. The ultimate stress and strain limits are not cumulative fatigue are evaluated to not handling, and expected operating modes exceeded, and cumulative fatigue does not exceed the ultimate stress or strain of the wil be performed. exceed a fatigue usage factor of 1.0.
material.
- 2. The control rod is evaluated to be capable 2. Evaluations will be performed of the effects 2. Calculated control rod clearances are of insertion into the core during all modes on control rod clearance of manufacturing sufficient to permit insertion.
of plant operation within the limits tolerances and swelling and irradiation assumed in the plant analyses. growth under expected operating modes. i 3. The material of the control rod shall be 3. Testing or operating experience will be 3. The testing or experience demonstrates shown to be compatible with the reactor used to determine whether the control rod the materials are adeqimte. environment. materials are compatible with the reactor environment. 4 The reactivity worth of the control rod shall Evaluations will be performed to determine
- 4. 4. 4. The calculated cold shutdown margin is be iccluded in the plant core analyses. the reactivity worth of the control rod. greater than the value given in the technical specifications.
O
I ABWR Design Document 1 2.9 Radioactive Waste ( 2.9.1 Radwaste System Design Description The liquid waste system collec ts. treats, monitors. and either recycles or discherges all radioactive liquid wastes within the plant.1 he solid waste tystem collects, sorts, monitors and either recycles or packages all radioactive solid w,stes within the plant. The radwaste system does not serve or support any safety function and has no safety design basis. Liquid Waste System The liquid waste system consists of three subsystems: the low conductivity waste system (l CW), the high conductivity waste system (HCW) and the detergent waste system (DW). The 1.CW system collects and processes clean radwaste, i. e., water of relatively low conductivity. Equipment drains and backwush transfer water are typical of wastes found in this subsystem. These wastes are collected, treated and monitored. If quality is adequate, the water is sent to the condensate storage tank. If not, it is reprocessed. The HCW system collects and processes dirty radwaste, i. e., water of relativery high conductivity and solids content. Floor drains are typical of wastes found in this subsystem. These wustes are collected, treated and monitored. If quality is adequate, the water is sent to the condensate storage tank. If not, it is reprocessed. Sometimes, the water is discharged following established procedures to maintain proper plant water balance. The DW system collects and processes detergent waste from personnel showers and laundry operations. These wastes are collected, filtered and monitored. If quality is adequate, the water is discharged. The liquid waste system provides one discharge line to the canal for the release of processed liquid waste. This line is provided with flow instrumentation, means of flow control and a radiation monitor. A high radiation signal from this monitor will close the discharge valve. The liquid waste system is provided with sample tanks to collect processed water with provisions to mix the contents and obtain samples for radiochemical analyses prior to discharge. Discharge can be made from only one sample tank at a time through a locked closed valve that is O under administrative control. 2.9 1- 6/1/92
ABWR oesign occument Solid Waste System The solid waste system consists of two subsystems: The dry active waste system (DAW) and the wet acthe waste system (WAW). The DAW system has an ar ea which is devoted to collecting and storing DAW and soiting it into reusable and nomeusable items. Reusable items are de< ontaminated as necessarv and scused. Nonreusable items are separated for further treatment. Combustible DAW is burned in an incinerator. Incombustible and compressible DAW are ieduced in volume using a compactoi. The processed DAW is packaged for shipment. The WAW system has tanks for collecting concentrated liquids, sludges or slunies and spent resins. The concentrated liquids are dried and solidified. The slurries and spent resins are either dried or dewatered. Packaging and transporting of the packaged wastes are in conformance with 10CFR61 and 49CFR 173, Subpart 1. Radiation monitors are provided to survey all waste packages. Individual components are provided with vents to assure that dust or contaminated air are not released to work spaces. A Piocess Control Program shall be prepared and approved by the NRC demonstrating that the cement-glass process complies with 10CFR61, Section 61.56. Table 2.9.1 prosides a definition of the inspections, test and/or analyses together with associated accs ptance criteria which will be used for the radwaste syste m. Inspections, Tests, Analyses and Acceptance Criteria l Table 2.9.1 provides a definition of the inspections tests, and/or analyses together with associated acceptance criteria which will be used for the radwaste syste m. t 2.9.1 2 6/1/92
- s. J m 5 Table 2.9.1: Radwaste System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The liquid waste system has only one 1. The as-built discharge line and controls 1. The as-built discharge line and controls discharge line provided with flow shall be inspected and tested. shall operate as designed.
instrumentation, means of flow control and a radiation monitor. A high radiation signal from this monitor will stop the discharge. Discharge can be made from only one sample tank at a time.
- 2. The components of the liquid waste 2. Inspections, tests and analyses shall be 2. All components shall meet the required system shall be provided which meet the performed as required on all as-built codes and standards.
codes and standards in Table 1, Regulatory system components. Guide 1.143,
- 3. Means shall bu provided to package and 3. As-built equipment for packaging and 3. The analysis shall show that the as-built
@ transport the solid wastes in conformance transporting solid wastes shall be means of packaging and transporting solid with 10CFR61 and 49CFR173, Subpart 1. inspected, tested, and analyzed. wastes can meet the requirements of the regulatory guides.
- 4. Individual comnonents shall be properly 4. The as-built componcats shall be inspected 4. All as-built components have been vented to nrevent the release of dust or to show that the release of ...r-borne
. properly vented.
contaminated air to work spaces. radioactivity has been prevented by venting.
- 5. A Process Control Program has been 5. An inspection shall show that an approved 5. A Process Control Program has been developed and approved for the cement- PCP is available and the as-built equipment approved by the NRC and the as-built glass solidification system. is capable of being operated in equipment is suitable for operation conformance with the PCP. following the PCP.
5" 5ra
ABWR Design oocument 2.10 Power Cycle q v 2.10.1 Turbine Main Steam System Design Description The Alain Steam (hfS) System (Figure 2.10.1) supplies steam generated in the reactor to the turbine. This Tier 1 entry addresses that portion of the SIS System that ranges between but does not include, the outermost containment isolation valves and the turbine stop valves. The SIS System is not required to effect or support rafe shutdown of the reactor or to perform in the operation of rector safety features; however, the S1S System is designed: (1) To comply with applicable codes and standards in order to accommodate operational stresses such as internal pressure and dynamic loads without rist of failures and consequential releases of mdioactivity in excess of the established regulatory limits. (2) To accommodate normal and abnormal emironmental limits. (3) To assure that failures of non-Seismic Category I equipment or structures, or pipe cracks or breaks in high or moderate piping in the ! his will not preclude functioning of safety-related equipment or structures in the plant. (4) With suitable access to pennit in service testing and inspections. The '.iS System main steam piping consists of four lines from the outboard main steamline isolation valves to the main turbine stop valves. The header arrangement upstream of the turbine stop valves allows them to be tested on-line with minimum load reduction and also supplies steam to the power cycle auxiliaries, as required. inspections, Tests, Analyser :-'d Acceptance Criteria Table 2.10.1 provides a definition of the insI>ections, tests, and/or analyses, together with associated acceptance criteria which sill be undertaken for the hts System. O 2.10 6/1/92
5 Table 2.10.1: P. Inspections, Tests, Analyses and Acceptance Criteria Certified Design Co nmitment inspections, Tests, Analyses Acceptance Criteria
- 1. Failures of non-Seismic (:ategory 1 1. Visual inspection of the MS System will be 1. No safety-related systems or structures are equipment or structures, or pipe cracks or performed. in the vicinity or are protected from failures breaks in high or moderete piping in the in the nonseismic portions of the MS MS System will not preclude functioning of System.
safety-related equipment or structures in the plant.
- 2. Access is provided for in-service testing 2. Visualin ,)ection of the MS System will be 2 Confirmation that required in-service and inspections. performed. ir.spections can be accomplished.
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ABWR Design Document 2,10.2 Condensate Feedwater and Condensate Air Extraction System b V The Condensate Feedwater and Condensate Ah Fatraction System (CFDWA) consists of two subsystems, the Condensate and Feedwater System and the $1ain Condenser Evacuation System ($1CES). Condensate an2 Feedwater System Design Description The function of the Condensate and Feedwater (CF) System is to teceive condensate from the condenser hotwells, supply condensate to the cleanup system, and deliver high purity feechvater to the reactor, at the required flow rate, pressme and temperature. Condensate is pumped from the main l condenser hotwell by the condensate pumps, passes through the feedwater l' heaters to the feedwater pmups, and then is pumped through the high pressure heaters to the nuclear Steam Supply System. The CF System boundaries considered here extend from the main condenser outlet to (but not including) the second isolation valve outside the containment. The CF System consists of the piping, valves, heat exchangers, controls and instnunentation, and the associated equipment and subsystems which supply the p reactor with heated feedwater in a closed steam cycle utilizing regenerative V feecheater heating. The CF System does not sene or support any safety function and has no safety design basis. System analyses show that failure of this system cannot compromise any safety-related systems or prevent safe sh .tdown. Portions of the system that are radioactive during operation are shielded with access control for inspections. Leakage is minimized with welded construction used wherever practicable. Relief discharges and operating vents are channeled through closed systems. Operational system redundancy is provided with respect to feedwater heaters, pumps, or control valves by using multi-string arrangements and provisions for isolating and bypassiag equipment and sections of the system. The majority of the condensate and feedwater piping considered in this section is located within the turbine building which contains no safety-related equipment or systems. The portion which connects to the second isolation valve outside the containment is located in the steam 'unnel between the turbine and (n; reactor buildings. This portion of the piping is a .alyzed for dynamic effects from postulated events and safety / relief valve discharges. 2.10.2 6/1/92
ABWR Design Document The entire system piping is analued for waterhammer loads that could intentially result from anticipated How transients. h Inspections, Tests, Analyses and Acceptance Criteria Table 2.lR2a provides a definition of the inspections, tese and/or analvses. together with associated acceptatu e criteria which will be undertaken for the CF Systein. Main Condenser Evacuation System Design Description Noncondensable gases are removed from the power cycle by the Alain Condenser Evacuation (51CE) System (Figure 2.10.2). The NICE System removes the hydrogen and oxygen produced by the radiolysis of water in the reactor, and other power cy cle noncondensable gases. and exhausts them to the offgas sr m during plant power operation. and to the turbine building comparunent exhaust system at the beginning of each startup. The StCE System does not sene or support any safety function and has no safety design basis. The AICE System is designed to Quality Group D. O The SICE System consists of two 100% capacity, double stage, steamjet air ejectors (SJAE) units (complete with intercondenser) for power plant operation, and a mechanical vacuum pump for use during startup. The last stage of the SJAE unit is nonnally in operation and the other is on standby. Steam supply to the second stage ejector is maintained at a minimum specified flow rate to ensure adequate dihition of the hydrogen and prevent the offgas from reaching the flammable limit of hydrogen. Steam pressure and flow is continuously monitored and controlled in the ejector steam supply lines. Redundant pressure controllets sense steam pressure at the second stage inlet and modulate the steam supply control valves upstream of the air ejectors. The steam flow transmitters provide inputs to logic devices. These logic devices provide for isolating the offgas Oow from the air ejector unit on a two-out-of-three logic, should the steam flow drop below acceptable limits for offgas stream dilution. The vacuum pump exhaust stream is discharged to the turbine building compartment exhaust system which provides for radiation monitoring of the system efuuents prior to their release to the nionitored vent stack and the atmosphere. 2.10.2 6/1/92
ABWR oesign occument fq The vacuum pump is tripped and its discharge valve is closed upon icceiving a V main steam high-high radiation signal. Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.2b provides a definition of the inspections, tests and/or analyses. together with associated acceptance criteria which will be undenaken foi t' e
.\1CE System l
l l I l l I ( 2.10.2 3- 6/1/92
I y Table 2.10.2a: Condensate and Feedwater System Inspections, Tests, Analyses and Acceptance Criteria Inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment
- 1. Review failure analysis design ' As-built conditions are same as the design
- 1. The CF System will be analyzed to show Nure will not compromise assumptions with respect to as-built assumptions used in the analysis.
that systen plant safety. condition.
- 2. Visual inspection of the CF System will be 2. The as-built CF System provides shielding
- 2. The CF System will be provided with performed. and access control.
shielding and access control.
- 3. Visual inspection of the CF System wit! be 3. Welded construction utilized as designed.
- 3. CF Syst3m leakage will be minimized by use of welded construction wherever performed.
practicable.
- 4. Visualinspection of the CF System will be 4. Relief valve discharges and operating vent
- 4. CF System relief valve discharges and lines are routed as required by certified operating vents will be channeled through performed.
design. closed systems. Simulated signals to verify operational S. The CF System remains operational. b 5. The CF System will operate with a 5. feedwater heater, pump or control valve status maintained. out-of-service. Failures of nonseismic Category I 6. Visual inspection of the CF System will be 6. No safety-related systems or structures are
- 6. in the vicinity or are protected from failure equipment or structures, or pipe cracks and performed.
in the nonseismic portions of the CF breaks in high- or moderate piping in the System. CF System will not preclude functioning of safety 7related equipment or structures in the plant.
'7. Review waterhammer analysis design 7. As-built conditions are same as the design
- 7. The CF System will be analyzed for assumptions with respect to as built assumptions used in the analysis.
potential waterhammer loads. condition. J O O O
Table 2.10.2b: Main Condenser Evacuation System - P n inspections, Tests, Analyses and Acceptance Criteria , Certified Design Commitment inspections, Tests, Analyses - Acceptance Criteria
- 1. , The offgas will be prevented from reaching ' '1. Tests will be conducted using simulated 1. Confirmation that the system isolates -
a flammable limit of hydrogen. signals to the SJAE flow control system. before flammability limits are reached. ,
- 2. Radioactive releases will be maintained 2. ' Tests will be conducted using simulated 2. Confirmation that the system isolates as within established limits. signals to the vacuum pump isolation required to limit releases.
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ABWR Design Document 2.10.3 Heater Drain and Vent System No Tier 1 entry for this system. O l l l l 2.10.3 -1 6/1/92
ABWR Design Document q 2.10.4 Condensate Purification System V Design Description The Condensate Purification (CP) System purities and treats the mndensate as required to maint.un reactor feedwater purity, using filtration to remc.e corrosion pioduct.s, inn exchange to remove condenser leakage and other impurities and water treatment additions to minimize co. ; ion / erosion reieases in the powei cycle. The CP Ssstem does not . serve or support any safety function and has no safety design basit The CP System is designed to Quality Group D standards. The CP Svstem consists of full flow high cf ficiency particulate filters followed by full flow deep hed demineralizers. Shielding is provided for the CP System. Vent gases and other wastes from the CP System are collected in controlled areas and sent to the radwuste system for treatment and/or disposal. ! The CP System is located in the turbine building, and piping or equipment failures will not affect plant safety. Inspections, Tests, Analyses and Acceptance Criteria Table 2.10 4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria which will be undertaken for the CP System. O O 2.10.4 6/1/92
j E O Table 2.10.4: Condensate Purification System 4 Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. Shielding will be provided for the CP 1. Visualinspection of the as-built CP System 1. Installed equipment is shielded in System. will be performed. accordance with certified design.
- 2. NO safety-related equipment will be in the 2. Visualinspection of the as-built CP System 2. Equipment is tocated as specified by vicinity of the CP System will be performed. certified design.
- 3. CP System wastes will be collected in 3. Visualinspection of the as-built CP System 3. Compliance with certified design controlled areas. will be performed. commitment.
9 O 5 m O O O
ABWR oesign Docuinezat 2.10.5 Condensate Filter Facility I No entiy. Covered by item 2.10A. O mes O 2.10.5 1 6/1/92
ABWR Design Document 2.10.6 Condensate Demineralizer No crit!T. Cover ed lyy ] tern '), j f).4, i l l t O l l l O 2.10.6 'I' 6/1/92
ABWR oesign oocument . 2.10.7 Main Turbine Design Description The main Tuibine Generator (TG) System converts the energy in steam f rom the nudeat ste.un supply system into electrical energy. The TG System does not sene nor support any safety function and has no safety design nasis. However, the TG System is a potential source of high encigv missiles that could damage safety related equipment or structures. The TG System is designed to prevent overspeed and thus minimize the [x>ssibility of high energy missile generation from TG System moving parts. The following component iedundancies are employed to guard against over speed: (1) Main stop v;dves/ Control valves. (2) Intermediate stop valves / Intercept valves (CIVs). (3) Priman speed control / Backup speed control. (4) Fast acting solenoid valves / Emergency trip fluid system (ETS). (5) Speed control /Overspeed trip / Backup overspeed trip. The TG System is enclosed within the turbine building, which contains no safety-related equipment or structures. The turbine generator is orientated within the turbine building to be inline with the reactor and control buildings to minimize the potential for any high energy TG System generated missiles from damaging any safety-related equipment or structures. Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.7 provides a definition of the inspections, tests, and/er analyses, , together with associated acceptance criteria which will be undertaken for the TG ( System. 2.10.7 6/1/92
E O Table 2.10.7: Main Turbine Generator System ~ ~2 Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The TG System will be designed to prevent 1. Visualinspection of the installed 1. Design provisions to prevent overspeed the forbine generator otor frorn exceeding equipment together with simulated testing are in place.
the design overspeed with redundant of the as-built overspeed protection instrumentation, controls and valving such system. that a single failure of any component will not cause the rotor speed to exceed its design value.
- 2. The turbine building will contain no safety- 2. Visua! inspection of the as built turbine 2. Turbine generator arrangements per related equipment or structures. The building and plant arrangements. approved plant design.
turbine generator will be orientated to minimize the potential for low trajectory high energy TG System missiles from damaging safety-related equipment or structures. m J 9 O O
ABWR Design Document i 2.10.8 Turbine Control System J L entrv. Covered .mder item 2.10.7. 1 I i i l l 4 i l 1 l 1 1 4 i l 1 i l i i i l 4 1 i l l 1 V 2.10.f3 6/1/92 l 4 I 1 i
ASWR oesign occument (' 2.10,9 Turbine Gland Steam System b] Design Description The Tuihine Gland Scaling (TGS) System prevents the escape of radioactive steam frorn the turhive shaft / casing penetrations and valve stems and prevents air inleakage through subatmospheric turhine glands. The TGS System consists of a sealing steam pressure regulator, sealing steam header, a gland steam condenser, with two full capacity exliaust blowers, and the associated piping valves and instmmentation. The TGS System does not sene or support any safety function and has no safety daign basis. The TGS System is designed to Quality Group D standards. The outer portion of all glands of the turbine and main steam valves is connected to the gland steam condenser, which is maintained at a slight vacuum by the exhauster blower. During plant operation, the gland steam condenser and one of the two installed 100% capacity motor <lriven blowers are in operation. The exhauster blower to the turbine huilding compartment exhaust system effluent stream is continuously monitored prior to being discharged. (O3 j During normal operation, the steam seal header is supplied from the main steam patb. The auxiliaq steam system provides a :100% steam supply backup when high radiation levels are detected in the blower exhaust or the main steam path source (s) are unavailable. A site specific radiological analysis will be required to determine what actions and at what level the TGSS steam supply shou!d be switched to the auxiliary source. Relief mlves on the seal steam neader prevent excessive seal steam pressure. Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.9 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria which will be undertaken for the TGS System. l 1 O l-l l 2.10.9 1- 6/1/92 l
If Table 2.10.9: Turbine Gland Steam System O
!O Inspections. Tests, Analyses and Acceptance Criteria Certifited Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. Radiological releases wi!! be maintained 1. Visualinspection of tise installed 1. System switches to auxiliary steam as within established limits. equipment coupled with a site-specific required to limit radiological releases.
radiological analysis and simulated signals to verify that the TC 3 System switches to auxiliary steam on high adiation levels. i l I 1 l 1 l t 8 l e o e ' ! -- i i ,
ABWR oesign occument - I I g 2.10.10 Turbine LubricatinD Oil System No Tici 1 entn im this system. Y O I O ' 2.10.10 1- 6/1/92
ABWR oesign on.;ument 2.10.11 M.)lsture Separator Heater m . m , i . ,,,, m , ,i,m ,r ,..,,,. l ) O '
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O 2.10,11 0/1/92
ABWR oesign occument 2.10.12 Extraction System N IICI I entt y for this systeta i
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l 1 l i O O 2.10.12 1 6/1/92
ABWR oesign oocument 2.10.13 Turbine Gypass System Design Description The Tin hine ltypau. fllu %teni pn widen capability to disch.uge inain sicain lonn the scactoi diserth to the (ondenses to ininiini/c step load iedot tion u.nisients ellet ts on the ie.n to ( oolant systein. The systein is also uwd to discilalge niaI:n stcain dtiling learlot ht tt stail(lhy drid t'uoldoWn opes atitins. The 'l14 Ststein does not sene or suppori any safety l'ont tion and has no sale:3 design basis. Theie is no safety-related equipinent in the vicinity of the Tit Sys'ern. All high eneigt lines of the Tit Suteni are located in the turbme buihling and no f ailure ol' high energy lines in the Tit Systein will aff ect safety :clated equijnnent. The Tit Systein consists of (l) a tluce-valve (hest that is connected to the snain sicandines upsticain of the tuihine stop valves, and (2) thice dunip lines that connect separately each regulating valve outlet to one condenser shell. The Tlt Systein is designed to bypass norninally 33% of the rated inain stcain flow dir ectly to the condenser. The Tit Systein, in coinbination with the s cactor systerns, provides the capability to shed 40% of the tin hine-generator rated load without reactor trip. Inspections, Tests, Analyses and Acceptance Criteria l Table 2.10.13 provides a definition of' the inspections, tests, and/or analyses. t together with associated au eptance criteria which will be undertaken far the tit Svstt in. l l l
\
l l l 2 10.13 fe1/92
"g Table 2.10.13: Turbine Bypass System Inspections, Tests, Analysss and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria
- t. Failure of high energy lines in the TB 1. Visual inspection of the instaIIed 16 1. Confirmation that high energy line breaks System will net affect safety-related System will be conducted. Will not jeopardize any safety-related equipment. equipment.
5 L O J O O O
ABWR oesign oocument 2.10.14 Reactor Feedwater Pump Driver No entn Cos cied undei Item 2.10.2. O . O 2,10.14 6'1/92
ABWR oesign Document 2.10.15 Turbine Auxillary Steam System
.%i Tics i c itn ior this .sysiciii.
O O 2.10,15 1 6/1/92
ABWR vesign Document l 2.10.16 Generator No critry. Coverect unicler Iterii '.!.10.7. i 1 l O t l O ' 2,10,16 1 6/1/92 l i 1
ABWR oesign Document 2.10.17 Hydrogen Gas Cooling System s , u . , <.,,,, , ,<,, ,i m _ <.,,.. O l O l 2,10.17 1- 6/1/92 l
ABWR oesign cocwnent 2.10.18 Generator Cooling System m . u .., i .,,,, , ,... .,,, _ ,,,. O O 2.10.18 1 6/1/92
ABWR oesis., oocument 2.10.19 Generator Sealing Oi! System No Tier 1 entn for this system. l O l l l 1 l O 2,10.19 1 6/1/92 l l l
ABWR Design oocument 2.10.20 Exciter No Tici l ciiiry 101 this systen). e 9 O O 2.10.20 -1 6/1/92
ABWR oesign Docwnent 2.10.21 Main Condenser ( Design Description The main condenser is designed to condense :uul deaerate the exhaust ste.un hom the snain ttabine and piuside a heat sinL foi the Tuihine liypass (Tit) Systent The main condenser does not scive or support any safety function and has no safety design basis. It is. howesci, designed with necess.uy shielding and controlled access to protect plant personnel from mdiation. The main condenser is a multi-shell type deaemting unit with a shell located directly beneath each of the low pressure tmbines. Each shell has tube bundles through which circulating water flows. The condensing steam is collected in the condenser hotwells (the lower shell portion) which proside suction to the condensate pumps. Since the main condenser opentes at a vacuum, any leakage is into the shell side of the main condenser. Tubeside or cir culating water inleakage is detected by measuring the conductivity of sample water extracted beneath the tube bundles, in addi tion, conductivity is continuonsly monitored at the discharge of the condensate pmnps and alarms provided in the main control room. , i In al} operational modes, the Condenser is at YaCuum and Consequently no radioactive releases can occur. Loss of vacuum sequentiai ly leads to control r oom ! alann turbine trip and eventually bypass and main steam isolation valve closure ! to prevent condenser overpressurization. Additionally, to avoid a turbine uip on l high condenser backpressure reactor recirculation runbar'- utomatically i initiated and, on a site specific basis setting, on a combinau - high condenser backpressure and loss of a circulating water pump. Ultimate overprotection is provided by rupture diaphragms on the turbine exhaust hoods. The instrumentation and control features that moni'or the perfonnance to ensure that the condenser is in the correct ope ating mode include:
-(1) Hotwell Water Level-Automatically controlled within preset limits.
During normal full load operation with nominal hotwell levels, the main condenser provides a four-minute active condensate .;torage volume and has a two-minute surge capacity. At minimum nonnal t l 2.10.21 1- 6/1/92 i
l l ABWR oesign Document operating hotwell watei leu l. and not mal f ull load condensate ih m cate. the corulenser prosides a two ininute minimtun holdup tiine f oi N 1ti decat. (2) Lindenser Picssure-Kc3 iveiall peiformance indicator that mitiates ala ms and trips at preset inels. (3) Inw Piessuic Tuibine Exhaust flood Temperature-Automata allt initiates tuil'me exhaust water sprays to piotect the turbine. (4) Inlet and Ollllet Citt ulating Water Tetnpelature-Monilot s perforinance onl) l (5) Condurtivity within the condensel and at the discharge of the condensate pumps--Initiates alarms at pieset levels. The main condenser potential for flooding is less than the Ciiculating Water (CW) Systein aint, consequently flooding protection is the same as the CW System (2.10.23). Condenser pressure indicator s are located above any potential flood level. Spray pipes and baffles aic designed to protect the main condenser internals fiom high eneigy flow input.s. h Hydrogen buildup during operation is provided by continunus evacuation of the main condenser. Hydrogen sources are excluded during shutdown. Inspections, Tests, Analyses and Acceptance Criteria _ Table 2.10.21 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria which will be undertaken for the main condenser. O 2.10.21 2- 6/1/92
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2 O Table 2.10.21: Main Condenser Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria I
- 1. Overpressurization of the condenser will 1. Tests will be performed using simulated 1. System isolation occurs. i be prevented by condenser isolation from signals to verify that the system isolates.
high energy sources. ;
- 2. Condenser pressure indicators and 2. Visualinspections of the as-built system 2. Installed equipment is in compliance with transmitters will be located above any will be conducted. the design commitment. .
potential flood fevels. t
- 3. Shielding and controlled access shall be 3. Visual inspections of the as-built system 3. Instclied equipment meets the shielding provided for the main condenser. will be conducted, and access control provisions of the certified design.
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1 ABWR oesign occument l i 2.10.22 Off Gas System b G Design Description The func tion of the Of fiGas Sutein (OGS) is to incat the gc. exhausted hom the Ina n tulbine ColldellSet s. via the SteamJet Air Ejectors (SJAE), to contial and minimize the iclease of gaseous : idioactivit) disch.u ged to the plant c!!vih tilln e n t. The OGS includes iedundant hydrogen / oxygen ratalytic recombiness and ambient temperature charcoal beds for process gas voleme reduction and radionuclide retention / decay. All of the OGS equipment, shown in Fig. 2.10.3, is located in the tm bine building. Ahhough the OGS is a non. safety-related system, the OGS is capab!c of withstanding an inte:nal hydrogen exploCun and is designed to ASME lloller and l'ressure Vessel Code Section VIII Division 1 and ANSI 1131.1 Piping Code. The OGS design includes leakage limits, internally through valve seats and exte nally into the plant, to seduce radioactive ieleases tiuough or out of the syst e m. p The OGS prousses the SJAE discharge during plant startup and normal plard operation before discharging the air flow to the plant vent. The OGS charcoal beds can operate in one of thtee modes: (1) llypass - All OGS flow bypasses the charcoal beds (used during startup). (2) Guard lled - All OGS flow passes through the Guard lled only. (3) Adsorber lieds - All OGS flow passes through the Guard lled and then through 4 parallel pairs of adsorber beds- each pair consisting of two beds in series. Hydrostatic tests of the OGS components r.nd entire OGS is performed at the factory and i, the plant in accordance with the applicable requirements for ASME Vill and ANSI I131.1. The OGS desigr. parameters aic: Design Pressure 350 psig Recombiner Shell Design Temperature 450*F Normal Flow Rate (after iecombiner) 15 scfm ?N Startup Flow Rate (after tecombiner) 250 scfm input Gas Activity 100,000 uCi/sec 2.10.22 6/1/92
ABWR Design Document Antoinatic operation of the UUS (aused by high radiation lesels downsticain of the chaicoal bed diwh.u ge is as follows. g (1) liigh iadiation will piovide an alann. (2) High high radiation will change the process t '.r flow path f roin hypao to flow tin ough the ch.nc oal beds. (3) liigh high high iadiation will shut the of fgas discharge valve. Inspections, Tests, Analyses and Acceptance Critoria Table 2.10 22 provides a delinition of the inspections, tests and/or analyses togethei with associated (iiteria which will be undertaken foi the OGS. O l 2.10.22 6/1/92
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Table 2.10.22: Off-Gas System j Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections Tests, Analyses Acceptance Criteria ! i
, 1. System configuration of the OGS as 1. Visual field inspections will be conducted 1. The installed configuration of the OGS will
! described in Section 2.10.22 is shown on of the installed OGS key componen's be considered acceptable if it complies l Figure 2.10.22. identified in Section 2.1%).22 and Figure with Figure 2.10.22 and Section 2.10.22. j _ .10.3. t 1
- 2. The OGS is designed to withstand internal 2. A hydrostatic test of the OGS will be 2. The hydrostath test results n.ust conform j hydrogen explosions. conducted in the plant in accordance with with the ASME and ANSI requirements.
j the ASME Vill-1 and ANSI B31.1 requirements.
- 3. The OGS is designed to minimize 3. Leak tests wi!! be performed according to 3. The leak test results must conform with the
! radioactive leakage through the OGS valve ANSI NOE Testing Standards. ANSI requirements, j seats and extemally into the plant. ,
;, 4 The OGS automatically controls the OGS 4 Preop tetis will be performed as follows: 4.
flow bypassing or through the charcoal j
- a. A simulated high charcoal gas discharge a. A Main Control Room altan will be adsorber beds depending on the radioactivity signal will give a Main sounded on an OGS discharge line i radioactivity levels m the OGS process gas Control Room (MCR) Marm. high radiation signal.
downstr eam of the charcoal beds. ;
- b. If the OGS process gas flow .is bypassing b. The OGS charcoal bed bypass valve the charcoal beds, a simulated high. operates correctly on a high-high OGS high charcoal gas discharge discharge radioactivity signal.
; radioactivity signal will close the c. The OGS discharge vrIve closes on a bypass valve and direct the gas flow high-high.high OGS discharge I
through the charcoal beds. radioactivity signal. j c. lf a simulated OGS gas discharge : I radioactivity signal reaches a high- [ j high-high level, the charcoal bed discharge valve will close. p 8 j !* i i i
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I V A I E6 i e E 2 @ E e E e DE 3 3 0 5 5 5 5 0 I NOTE: I I I SS S S S S S S S S I vAi.vE POSrToNS 43 4 7 4 4 7 7 7 7 [ SHOWN ARE FOR FULL I I I CHARCOAL FETRATG g $2 1 3 3 3 $ $ $ 5 3 g I V W W .W W W W W i i . W) I I sn 5 lM I I l - I I g _ _ _ _ _ _ _ _ _ _ _ _ _ CHARCOAL ADSORBER VAULT ___________--_______g Figure 2.10. -Gas System 9 1:e
l ABWR Design Document 1 2.10.23 Circulatinr; Water System Design Description The (:in ulating Water (CW) Systeni juotides a continuous supph of cooling water to the main condenser to iemose the heat icjected by the tmhine cycle and ausiliary systems. The CW System does not seire or suppr.rt any safety Iunction and has no safety design basis. To pievent flooding of the turbine building, the CW Systein is designed to autoinatically isolate in the event of gross system leakage. The circulating water pumps tue tripped and the pump and condenser v.dves are closed in the event of a system isolation signal from the condenses area high-high level switches. A c ondenser area high level alann is provided in the control room. A icliable logic scheme will be adopted to minimize potential for spurious isolation trips (e.g.,2-outef-3 logic). The CW System is designed and constructed in accordance with Quality Group
!) specifications.
The CW System consists of the following components (Figure 2.10.23): (1) Intake scieens located in a screen house (2) Pumps (3) Condenser water boxes (-1) Piping and valves (5) Tube-side of the main condenser (6) Water box fill and drain subsystem Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.23 provides a definition of the inspections. tests, and/or analyses,
- together with associated acceptance criteria which will be undertaken for the l CW System.
2.10 23 1 6/1/92
y Tcble 2.10.23: Circul: ting W:.t r System a
" Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria Floodinty of the turbine building will be 1. Visual inspe:: tion of the installed 1. CW System isolates upon receipt of an 1.
prevented by CW System isolation in toe equipment coupled with the analyses of isolation signal. event of gross system leakage. the leakage /floodir:g characteristics of the as-built CW System will be performed using simulated signats to verify system isolates on high fevel. 9 l l l l O I
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ABWR oesion occument 2.10.24 Condenser Cleanup Facility No Tici 1 entiy for titis svstein. l l O O 2.10.24 1- ft'1/92
ABWR nesian occument 2.11 Station Auxiliary 2.11.1 Mai "later (Purified) System Desigt Jescription i he Statrup Watei ( Purified) System (hll'Wl') ]n ovides ]nnilied inateup watei t<> the rondernate storage tatik, plant auxiliatY systeins atul the stilge tanks sh.o ed by the s eat tor cooling water aiul IWAG cmcr gency cooling water systems. The $1CWP system consists of distribution piping atul valves to systein useis thloughout the plant, Platit structutes that have hlCWP piping are shown in Figure ".11.1. Stakeup water is supplied to the system by the .N !akeup Water (Picp.u;uion) system located outside the ttu bine building. (The pieparation systein is tiot within the scope of the celtified design; see Section 4.6 f or interf are t equit eme nts.) The interfaces between the 51UWP systein and all safety-telated systems aic k>cated in the conttol building or s cactor building which aic Seismic Categor) I structutes. The pottions of the blUWP systein that couhl adversly impact stnictutes, systeins.Or components iluportant to safeiy during a scistnic event at e designed to assure their integrity under seismic loading resulting from a safe shutdown earthquake. In the event of a IDCA. the safety-related systems are O isolated hom the 51UWP system by automatic valves in the : fety-related system. Inspections, Tests, Analyses and Acceptance Criteria Tahic 2.11.1 piovides a definition of the inspections, tests. and/or analyses togethel vith associated acceptance criteiia which will be undertaken loc the " SIUWP systein. O 2.11 1- 6/1/92
i y Table 2.11.1: Makeup Water (Purified) System (MUWP) u inspections, Tests, Analyses and Acceptance Criteria inspections. Tests. Analyses Acceptance Criteria Certified Design Commitment The MUWP distribution piping is provided 1. Visualinspections of the MUWP 1. MUWP pping is provided to each of the 1. for all principal building structures as penetrations into and out of a!! principal principal building structures as shown in shown in Figure 2.11.1. building structures as shown in Figure Figure 2.11.1. 2.11.1 shall be performed.
- 2. Visualinspection of the MUWP system and 2. The MUWP and makeup water
- 2. The MUWP system purified water is provided by the makeup water makeup water (preparation) system (preparation) system connections are Sterface connections shaII be performed. provided.
(preparation) system. 1
- 3. Visualinspections of the surge tank and 3. The MUWP and surge tank connections are
- 3. The MUWP system provides makeup water to the surge tanks shared by the reactor MUWP connections shall be performed. provided.
l cooling water and HVAC emergency cooling water systems. h m a O O O
ABWR oesign Document _ O l MAKEUP l WAT ER j l (PREPARATIOtJ)
! SYSTEM l t_________J COtJDEtJSATE ..
SIORAGE I TANK
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RADWASTE TURBINE BUILDING BUILDING CONTROL SERVICE BUILDING BUILDING O REACTOR BUILDING 5 Figure 2.11.1 MUWP System O 2.11.1 3- 6/1'92
ABWR oesign occument 2.11.2 Makeup Water (Condensato) Systern Design Description The Makeup Watei (Omdensate) Systein (Ml'W( a piosidc3 ( omlensatc <[u. din w.uce to v;uious plant systems for both nonnal and eincigency operations. lhe MUWC spiem consists of a condensate stoiage tank. tluce paralle' pinnp units. ;uul disuibution piping and vahes to sptein user s tinoughout the plant. In addition to the MUWC pumps, thc teactor i nie isolation cooling (RCIC), contiol iod dihe (CRD), high pressure coic floodei (llPCF) and suppression pool cleanup (SI CU) pumps take suction f unn the conelensate storage tank. Makeup watei is supplied to the tank fiom the inakeup water (purified) system.
< onnol iod drive system, nulvaste systern and the condensate icturn line.
I (:ondensate stoiage tank water levelis indicated both locally and in the main conu ol room. Alarms are also provided in the main control room for high water level. Any tank oscrilow or drainage is sem to the iadwaste system for ocatment. i The MUWC dhtilhotion piping co nection to the condensate storage tank is located at a tank elevation that ensures adequate water sepply to the llPC1', RCIC and SPCU which are connecter 8 at a lower elevation, in the event of a hicak in the MUWC piping, the water volume hele,w the tank connection is sollicient to piovide makeup water to these essential systems. The prirnary MUWC systein reclulicinents are: Condensate Storage Tank: Total Volume (gal.) 5tio,000 (appr ox.) Essential Volume (gal.) :> 220.000 (appr ox ) (llelow MUWC pump lines) MUWC Pumps: Quantity 3 Capacity Each (gpm) 550 The MUWC system is not safety-related, llowever, the system incorporates features that assure reliable opetution over the full range of normal plant conditions. Inspections, Tests, Analyses and Ac:eptance Criteria Table 2.11.2 pmvides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the MUWC system O 2.112 6/1/92
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O. Table 2.11.2: Makeup Water (Condensate) System (MUWC) Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analysis Acceptance Criteria
- 1. The condensate storage tank has adequate !. Visual inspections of the condensate 1. The condensate storage tank has a tota.
capacity for plant requirements. storage tank volume will be performed. capacity of approx. 560.000 gallons. The essential vo'ume below the MtJWC puinp line connection is > 220,000 gations (approx.).
- 2. The .MUWC oumps provide adequate flow 2. inspections of vendor documentation will 2. Each of the three MtJWC pumps is capable t > meet system and plant requirements. include pump capacities. Flow tests will of delivering 550 gum to the system.
confirm that adequate flow is available to the system.
- 3. Condensate storage tank water level 3. Visualinspection of the control room water 3. Condensate storage tank water leve; i indication end alarms are provided in the level indication and alarm equipment will ..idication and alarms are provided in the main control room. be performed. main control room.
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ABWR oesion occument 2.11.3 Reactor Building Cooling Water System ( \ Design Description The Reactor Building Cooling Water (RCW) System distributes cooling ,er during various plant operating : nodes, r,s well as during shutdown, and during post-LOCA operation of the various safety systems. The system ren .es heat from plant auxiliaries and transfers it to the Ultimate Heat CJan (UHS) via the Reactor Service Water (RSW) System, The RCW System removes heat from the ECCS equipment inclrding the emergency diesel generators during a safe reactor shutdown cooling functiori. The RCW system is designed to perform its required safe reactor shutd .n cooling function following a postulated loss-of-coolant accident / loss-of-offsite power (LOCA/ LOOP), assuming a sing!e active failure in any mechanical or elecuical RCW subsystem or RCW support system. In case of a failute which disables any one of the three RCW divisions, the other twc, divisions meet plant safe shutdown requirements, including a LOCA or a LOOP, or both. Redundant isolation valves are aale to separate the essential portions of the RCW cooled components from the nonsafety-related RCW cooled componer,ts during a LOCA, to assure the integrity and safety functions of the safety-related parts of n the system. The isolation valves to the non-essential RCW System are 'd automatically or remote-manually operated, and their positions are indicated in : l the main control room. Each RCW dision includes two pumps which circulate RCW through the ; mrious equipment cooled by the RCW System and through three heat exr. hangers which transfer the RCW heat to the UHS via e RSW System. Each RCW division hiain Lontrol Room ($1CR) instrument indication includes main loop surge tank levei, main loop radiation and RHR HX flow and temperature. hICR controlincludes all hiOVs and AOVs shown on Figure 2.11.3. Normal surge tank hiUWP makeup is automatic or h1CR controlled. The three RCW tsain conFgurations are shown on Figure 2.11.3. The RCW System provides three similar complete trains ( A, B and C) which are mechanically and electrically separated. The RCW pumps and m1ves mr each RCW d vision are supplied electrical power from a different divuion of the ESF power syste m. The RCW AShfE Code classifications for different portions of the system are indicated on Figures 2.11.3a.c. The safety-related pordons of the RCW divisions are designed to Seismic Categog I and Qaality Group C, and are located Seismic Category I structures. 2.11.3 -1 6D/92
ABWR oesign Document During various plant operating modes, one RCW water pump and two heat exchangers aic nm mally operating in each dhision. Flow halancing piovisions aie included within each RCW division. Ptunp design panuneters aie: RCW A/B RCW C Design pressme (psig) 200 200 Design temperature (SF) 158 158 Discharge now rate (gpm/] ni.o 2 5.700 2 4.800 Pump total head trsig) 2 80 2 75 Heat exchan 3 . .acities ;n e cach: 6 2 45E litu/h 2 42E6 l\tu/h _ Connecticus to n radiation monitor are provided in each division to detect radioactive contamination resulting from : abe leak in one of the RHR exchangers, fut i pool exchangers, or other exchangers. The RCW pumps and heat exchangers are located in the tower Doors of the control building. The equipment cooled by the RCW divisions are located in the reactor building. turbine building, anct radwaste building, (Figures 2.11.3a<>. Tables P.11.3h, c, d show which equipment receives RCW Row during various plant operating and emergency modes. The tables also indicate how many heat exchangers are in se:vice in each mode. h L)uring normal plant operation, RCW Dows through equipment which is normally operating and requires cooling and all ECCS equipment, except RHR heat exchangers and ESF diesel generators, as shown by open or closed valves in Figure 2.11.3. _ F If a LOCA occurs. a second RCW ptunp and third he exchanger in each loop are pla ed in senice. Automatic or remote operated ismation valves will separate the RCW for the I.OCA required safety equipment f om the nonsafety-related equipment, if a RCW surge tank low water level signal occurs. The primaty containment RCW isolation valves automatically close if a LOCA occurs. 7.i9 3 6/1/92
ABWR Design Document l Alici a 1.U( A, the following sequence will be followed: l (1) If the nonsafety portion of the RCW System is available to the instrument air /senice air (lA/SA) compressors, the CRD pmnps and CUW pumps RCW flow to these nonsafety components is maintained (Figure 2.11.3). Flow is automatically shutoff to other non-essential equipment after the LOCA. (2) If the operator determines after the LOCA, from essential RCW inst rument; a, that the integrity of the non-safety RCW System to the above-mentioned compressors and pumps has been lost, he can shut the remote operated non-essential isolation valves shown in Figure 2.I1.3. If the surge tant water level reaches a low level, with or without LOCA, indicating loss of water out of the RCW System, isolation valves in the supply and return piping to the nonessential equipment will automatically close, including the compressors and pumps mentioned above. Without a I.OCA and with low surge tank standpipe water level, all running RCW pumps trip. For post-LOCA both RCW pumps continue running with low surge tank standpipe water level. The RCW/RSW heat exchanger design basis condition occurs duiing post-LOCA cooling of the containment via the RHR heat exchangers. The RCW pumps have the flow capacity to deliver required flow to the ECCS equipment in each division and the above-mentioned compre3 sors and pumps if the isolation valves cannot be closed. After a LOOP, the RCW pumps isolation valves and their control logic are automatically poweied by the emergency diesel generators. A separate surge tank is provided for each RCW division. Normal makeup water source to the surge tank is the Makeup Demineralized Water (MUWP) System. For LOCA conditions, the Suppression Pool Cleanup (SPCU) System provides a backup surge tank water supply. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.3a provides a deGnition of the inspections, tests, and/or analyses together wah associated acceptance criteria which will be and undertaken for the RCW System. /^\ 2.11.3 3 6/1/92
Table 2.11.3a: Reactor BuiMng Cooling Water (RCW) System u inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. System configuration, including key 1. Inspection of construction records will be 1. The system configuration conforms with components and flow paths, is shown in performed. Visual inspection (VI) will be Figure 2.11.3.
Figure 2.11.3. performed based on Figure 2.11.3.
- 2. Three RCW trains are mechanically and 2 Tests and VI of the three independm.t trains 2. Plant tests and VI confirm proper electrically independent. will be conducted which wiH include independence of three RCW divisions.
independent and coincident operation of the three trains to demonr" ate complete divisional separation.
- 3. During various modes of operation, the 3. Limited system hydraulic tests will be 3. The results confirm that the RCW has the RCW System has adequate hydraulic conducted according to available water f!ow capability specified by the capability for plant auxiilaries and the nonnuclear heat plant conditions.The tests certified design commitment, including primary containment required fer safe will demonstrate a safe plant shutdown safe shutdown operation with 1 RCW 6 shutdown following a design accident or with one RCW division out of service. division out of service.
transient. These safe shutdown requireme;nts are satisfied with only an y 2 of 't RCW divisions operating.
- 4. Isolation valves as shown in Figure 2.11.3 4. VI of the installed RCW System and RCW 4. Isolation valves are properly located as can automatically or remote manually preoperational tests as follows will be shot- n in Figure 2.113 and ' re separate the RCW for the essential completed: demonstrated to operate autornatically or equipment from the RCW for the non- remote manually to isolate RCW for non-essential equipment. a. Remote-manual operation of the essential from RCW for essential isolation valves from the main control equipment ccoled by the RCW System.
room.
- b. During simulated LOCA conditions, a simulated LOCA condition wid be combined with a simulated RCW surge tank water level signal to automatically close the isolation valves.
e
- c. A LOCA signal will shut RCW isolation valves which will shut off HCW flow to all non-essential equipment except the IA/SA compressors, CRD umps and CUW pumps.
O O O
O - 3 Table 2.11.3a: Reactor Building Cooling Water (RCW) System (Continued) Inspections, Tests, Analyses ant' Acceptance Criteria Certified Design Commitment !nspections, Tests, Analyses Acceptance Criteria
- 5. Without LOCA and with low surge tank 5. RCW System preoperational tests will be 5. The RCW pumps will trip or operate as standpipe water I.+ vel, both RCW pumps in performed as follows: follows:
that division trip. For nost LOCA, both RCW
- pumps will operate with low surge tank a. Simulate a surge tank standpipe low a. The running pump (s) will trip on surge standpipe water tevel, water level in the standpipe and tank standpipe low water level.
confirm the running pump (s) trip.
- b. With a LCCA condition signal, both
- b. During a simulated LOCA condition RCW pumps will continue to operate and a simulated surge tank standpipe with a simulated surge tank standpipe low water level signal, confirm that low water level signal. '
both RCW pumps will operate.
- c. Both RCW pumps start on simulated
- c. During low surge tank standpipe water LOCA signal.
level condition, a simulated LOCA ! 9 signal starts both divisional RCW , pumps.
- 6. A LOCA will result in the automatic start of 6. Tests simulating LOCA/ LOOP conditions 6. LOCA/ LOOP signal successfully starts j the second RCW pump in each division and will be conducted for the RCW System second RCW pump and initiates RCW/RSW i start flow through the third RCW/RSW Hx which confirm the RCW and its support Hx floa in each division including the !
in each division. systems will perform its function under following confirmations-those conditions. Tests will be conducted During LOCA/ LOOP (loss-of-coolant for the RCW, which confirm that after the a. Regardless of which RCW pun.p was t accident / loss of off-site power) conditions, LOOP, each division of RCW pumps and operating during normal operation
- RCW pumps and valves are powered by valves operates with the same division of before the LOCA, after the LOAC/ LOOP the emergency diesel generators (D/G). emergency D/G power and associated DC simulation occurs, the first and second control power sources. RCW pump will start automatically, powered by the emergency diesel generator.
- b. Regardless of which two RCW/RSW Hx's were operating before the LOCA, ,
C after the LOCA/l.OOP occurs, the RCW L
;$ motor-operated valve on the third Hx discharge will open automatically. -
ABWR ocsign Document Table 2.11.3b: Reactor 9eilding Cooling Water Consumers Division A Hot Emergency Operating Normal Shutdown Shutdown Standby (LOCA) Mode / Operstmg , " "'g Y at 4 hours at 20 hors ino loss of (Suppression Components Conditions AC) U ** I A Pool at 97'C) RCW/RSW Heat 2 3 3 2 3 3 Exchangers in Service EGSENTIAL
- E mergency Die- -- - -- --
X X sel Generator A RHR Heat -- X X - X X Exchanger A FPC Heat X X X X X X Exchanger A Others (essen- X X X X X X tial)* NON ESSENTIAL RWCU Heat X X X X X - Exchanger inside Drywell* X X X X X -- Others (non. X X X X X X essential)* (1) (X)- Equipment receives RCWin this mode. (-) Equipment does not receive RCW in this mode. (2) HECW refrigerator, room coofers (FPC pump, RHR, RCIC, SGTS, FCS, CAMS), RHR motor and seal coolers. (3) Drywell(A & C) and RIP coolers. (4) instruments and service air coolers: RWCU pump cooler, CRD pump oil, and RIP Mg sets. O 2.11.3 6/1/92
AEWR oesign Document Table 2.11.3c: Reactor Buildin0 Cooling Water Consumers Division B Oe t g Star Shutdown Shutdown y g LOCA)
, O rat g Condrtions (loss of AC (S"PP^i "
Component s Pool at 9TC) AC) RCW/RSW 2 3 3 2 3 3 HoatExchangers in Service ESSENTIAL m Emergency Die- - - X X _ _ _ sol Generator B RHR Hont X X X X Excher.gerB f PC HeatEx- X X X X X X changorB Othois (osson- X X X X X X tial)* NON-ES SENTI AL , RWCU Heat X X X X X - Exchanger inside Drywell* X X X X X - Others (non- X X X X X X essentiall* (1) (X) - Equiprnant receives RCW in this taode. (-)- Equipment does not roceivo RCW in this mode. (2) HECW t efrigerator, room coolors, s cPC r. ump, RHR, RCIC, SGTS, FCS, CAMS), RHR motor and seal coolers. (3) Drywell (B) and RlP coolors. (4) Reactor Building sampling coolers; LCW sump coolers (in drywell and reactor building), RIP MG sets and RWCU pump coolers. O 2.11.3 7 6/1/92
ABWR Design Document _ Table 2.11.3d: Reactor Building Cooling Water Consumers Division C g Hot Emergen cy Operating Normal N Shutdown Shutdown Standby (LOCA) Mode / Operating 3, , (Suppression at 4 hours at 20 hours (no loss of Component. Cond'tiens r II fAC AC) Pool at 97'C) RCW/RSW 2 3 3 2 3 3 HoatF xc hangers in Service ESSENDAL Of Emergency Die- X X sol Generator 8 RHR Heat X X -- X X ExchangerB Others (essen- X X X X X X tialF NON ESSENTIAL Others (non- X X X X X X essential)0' (1) (X) Equipment receives RCW in this mode. (-) - Equipment does not receive RCW in this mode. (2) HECW refrigerator, room coolers, motor coolers, and mechanical seal coolers for RHR and HPCF. (3) Instrument and service air coolers, CRD pump oil c ooler, radwaste components, HSCR condenser, and turbine building sampling coolers. O 2.11.3 6/1'92
ABWR oesign Document s (_/ MUWP 3 (A Fd TE SURGE T ANK ss RHRhr (Reacto. Building) U , (Rextor Building) - em H l L - l EMERGENCY D'G ) (Reactor Bu!! ding) FPC HX U - - >- i (Reactor Building) SPCU MUWP _, OTHERS (ESSENTI AL) (Roactor and Control Building) FF y w %s CRD & CUW PUUPS * , o.. ~ c- - (Reactor Building)
~ .
qNC l --(NCl3 i__ ____ l ___ ___ OTHE RS (NON-ESSENTIAL) ,___,____ i (Reactor, Radwaste and Turbine Building) l l b IA!SA COMPRESSORS _ _ _ _ _ _ _ _ _ _ (Turbine Building) I - t
- ,M - DRYWELL EQUIPMENT
~ ~
NOj 2 qNC NCl2 (NC h CONTAINMENT CONTAINMENT RCW Hx ! i (centre emnoi .J C RS RSW RCW PUMP (Control Buikiing) e scentroI Pmi no RSr RSW
,, RCW Hx C (c-e,,a an n.,s RSW RCW PUMP RS (Control BuiWing) ~J Figure 2.11.3a RBCW Division - A 2.11.3 9 6/1/92 l
1
ABWR oesign Document O MUWP A i 3 TE M] SURGE TANK ss RHR Hx iJ ,- (Reactor Building) U (Reactor Building) - C L i EMERGENCYD'G (Reactor Building) M FPC HX - (Reactor Build:ng) SPCU i M'JWP OTHERS (ESSENTI AL) (Reactor and Control Building) A M M g _,___ CUW PUMP __ q NC i (Reactor Building) , NCj 3 ___ ___ OTHERS (NON-ESSENTI AL) u_______ g , i (Rea. tor, Radwaste and Turbine Building) , ! l ! i l l M Y UM ' e {12) iL- O A-N1- DRYWELL EQUIPMENT ~ NCl2 f qNC NCl2 4NC CONTAINMENT CONTAINMENT RCW Hx i f Cor' rol Bui4nal C RSV RSW RCW PUMP (Control Buildng) l (Control Bush na) RS RSW f l G RCW Hx (Control BuildiW
- C RSW RSW RCW PUMP (Control Buid ng)
O Figure 2.11.3b RBCW Division - B 2.11.3 6/1/92
ABWR nesign Ducoment /^\ \ MUWP NO F
,T M SURGETANK j s RHR Hx mI ,' (Roactor Building) U
^ (Reactor Building)
-e L -
EMERGENCY D G ' (Reactor Building)
~
w SPCU h MUWP _._ OTHERS (ESSENTIAL) . . (Reactor and Control Li1 ding) I CRD PUMP
~ _ _p .__ __
(9st - Mr Building) NCl3 , _ __ _ __. _%_NC
- I M ) ,
OTHEFS (NON-ESSENTI AL) ,_________ i (Reactor, Radwaste and Turbine Buildings y l 8 \ L__.______.___ IA/SA COMPRESSORS _ _ _ _ _ _ _ _ . _ (Tubinc Building)
- V 4D s
RCW Hx __ (Control Buikfinoi C RS RSW RCW PUMP (Control Building) PCW Hx _ JC ontrol Bueno) RSW~& RSW u RCW Hx C - (Control Buildnoi RSW - h RSW RCW PUMP (Control Beding)
)
v ' Figura 2.11.3c RBCW Division - C 2.11.3 6/1/92
ABWR Design Document 2.11.4 Turbine Building Cooling Water System Design Description The Tmbine Ituilding Cooling Water (TCW) Systern provides cc ciosion- i inhibited demineralized cooling water to the turbine island auxiliaiy equipment. The TCW System consists of one surge tan: one chemical addition tank,iluce parallel pumps each of 50 percent capacity, three parallel heat exchangers each of 50 percent capacity, associated coolers, piping, valves, controls, and instrumentation. The TCW System rejects heat into the Turbine Senice Water (TSW) System. The surge tank piovides makeup water for continuous system operation. The makeup water flow to the surge rank is initiated automatically bv low sur ge tank water level and continues until normal water level is established. During normal plant operation, two of the thice 50 percent capacity putnps circulate water tinough the shell side of two of the three 50 percent heat exchangers in senice. The standby TCW pump is automatically started on detection oflow TCS System pump discharge pressure. The standi>y TCS System heat exchanger when requiied, is placed in senice manually. The system configuration is shown in Figure 2.ll A. Surge tank high and low levels, and TCW pump discharge pressure are alarmed i O in the TCW local panel as well as in the mam control ~-n. l O l The TCW system is classified as a non-safety-related, non-seismic system. Inspections, Tests, Analyses and Acceptance Criteria Tab!c 9.11 A p.ovides a definition of the inspections, tests, and/or analyses together with ase.ociated acceptance criteria which will be undertaken for the TCW system i O 2.11/. 1 6/1/92 i
$ Tcbb 2.11.4: Turbine Building C:oling W:t:r Syst:m Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Test, Analysis Acceptance Criteria
- 1. The configuration of the TC.S System is 1. Inspection of the as-built TCW System 1. As-built TCW System configuration for shown in Figure 2.11.4. configuration shall be performed. those components shown, conforms with Figure 2.11.4.
6 cn 5 r$ O O O
O O O ta C O SURGE TA?JK CHEMICAL ADDITION (N LC 1ANK _ 2_ MAKE-UP oMg ' WATER V W m . FROM TURB.lNE ISLAND AUXILIARY EQUIPMENT TCW PUMPS 4H g l 'M (jO l h ,i m TO TURBINE ISLAND AUXILIARY EOUIPMENT CKI 'd .. igi i i 93 Mhl C4 iKi i i
- LEGEND p NC - NORMALLY CLOSED l LC - LEVEL COfJTROLLER Hw: w!
Wlh l 1 I- !\! TCW HEAT EXCHANGERS NXl - 1 I !\l 5 lkl NC Figure 2.11.4 Turbine Building Cooling Water (TCW) System i _ _ _ _ _
. . . . , , ) ,g-
ABWR oesign Document g 2.11.5 HVAC Normal Cooling Water System G . . Design Description The ll\'AC normal cooling water sptem (HNCW) delivers chilled water to the moling coils of the di)well coolers. each building supply unit and the local ali conditionen s to maintain design thermal emironments during normal and upset plant conditions. Tht HNCW system consists of five 25G chillers, each with pumps, sening a conunon chilled water distribution system connected to the chilled wuter cooling coils in the dnwell coolers, each building supply unit and the local air conditioners. Condenser cooling is provided by the turbine building cooling wuter system. The chiller and pump sets have a thrze ,vay mixing valve or a lion control valve to maintain the desired temperatme. The system uses nukeup water f rom a surge tank which is shared between the HNCW, turbine cooling wuter and hot water heating systems. , 1 Tne HNCW system is not safety-related. However, the portions of the HNCW system which penetrate the primary containment are provided with isolation valves and penetrations which are designed to Sei:mic Category I, ASME code, Section III, Safety Class 2, Quality Group B requirements. The isolatica . valves may be manually operated from the control room except when a LOCA signal is j present. Each chiller and pump unit is designed to meet the following requirements: Chiller Capacity Each (BTU /h) 8.9S .6 l The HNCW system is capable of removing the maximum heat loading during annual shutdown with one of the five pump and chiller units in standby. In case of a chiller or pump trip, the standby unit is automatically started. Flow switches also prohibit any chiller from operating unless there is water flow through the evaporator and condenser. Inspections, Tests, Ana'yses and Acceptance Criteria Table 2.11.5 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the HNCW system. a 2.11.5 6/1/92
N Table 2.11.5: HVAC Normal Coc!ing Water (HNCW) System inspections, Tests, Analyse
- and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The standby chiller and pump unit 1 Tests simulating chiller and pump failures 1. Each chiller and pump unit acting s a automatically starts when an operating will b , conducted for each chiller and standby unit successfully starts upon a trip chiller or pump trips. pump unit. signal from any one of the operating units.
- 2. The HNCW cooling capacity is capable of 2. Inspections of verJor documentation will 2. Each chiller unit shall have an effective removmg the design heat loads. include chiller ; .d pump capacities. Flow heat removal capacity of 8.93E6 BTU /h.
tests will confirm that adequate flow is available to the system
- 3. Isolation valves on the supply and return 3. Tests simulating a LOCA signal shall be 3. Visual inspections shall confirm that tne lines theough the primary containment conducted on the HNCW system logic, isolation valves close upon a LOCA signal.
shall close on a LOCA signal. g, All isolation valves in the HNCW system Visual .nspections shall confirm that the isolation valves may be manually operated shall be opened and closed from the main isolation valves open and close from the from the control room. cc.itrol room switches. main contrc. .oom switches. m Y O O O
ABWR Design Document 2,11,6 HVAC Emergency Cooling Water System Design Description The HVAC Emergency Cooling Water (HECW) System delivers chilled water to the control building essential electrical equipment room coolers, the diesel f generator zone coolers, and the main control room coolers during shutdown of the icactor, normal operating modes. and abnormal s cactor conditions including 1.OCA. The HECW System consists of three mechanically separated divis i ons (Figure 211.6). Each division provides cooling to one control building er% electrical equipment room and one diesel generator zone. Ei+
- s. T m "C' also provides coolmg to the main con. col room. Pow . ~ r + "d ; e" division from independant Class 1E sources.
HECW division "A" consists of one pump, one refrigerath.n 1,4 instrumentation, and distribution piping and valves to the cooling coils. , Dhisions "B" and "C" are similar except that two parallel pumps and refrigeration units are used. Surge tanks and condenser coolant Gow are provided by the corresponding dhision of the RCW System. A chemical addition tank is shared by all HECW dhisions. Makeup water is supplied from the makeup water (Purified) system at the surge tanks. The surge tanks are capable of replacing system water losses for more than 100 days during an emergency. The refrigeration and pump units are designed to meet the following requirements: (1) Refrigerator Capacity (BTU /ht) 2.3x10 6 (2) Pump Capacity (gpm) 256 All major system components are loc:.ted in the control building except for the diesel generator zone cooling coils, which are in the reactor building. There are no primary or secondary containment penetrations within the system. In addition, the system layout is designed to permit periodic in-service inspection of all system components to assure the integrity and capability of the system. Piping and valves for the .HECW System, as well as the cooling water lines from the RCW System, are designed to Seismic Category I and ASME Code, Section Ill. Class 3 and Quality Group C requirements. The classification extends up to
/ and i..cluding the block valves for the chemical addition tank. The only non- \ safety-related portion of the system is the chemical addition tank and the piping from the tank to the block valves.
2.11.6 6'1/92
ABWR Design Document The HECW System is capable of reinosing all heat loads with one if the f our pionp and relrigerator tmits from dhision "it" and "C' in standbv. I he st.uulin ictrigerator is equipped with an inteilock which automaticallt st.u ts the unit h upon failure of the operating ref riger.noi. Flow switches prohibit the icf rigerators from operating unless there is wates flow through the evaporator and condenser. The refrigerator units can be controlled indhidualh fiom the main control room by a remote manual switch. i The 'iECW System is designed to perform its required safe scartor shutdown rooling function following a postulated losmf-coolant accident / loss of of fsite , power (l.OCA/1.OOP), assuming a single active failure in any mecl anical or electrical division. In case of a failure which disables any one of the three HE' divisions, the other two divisions meet phnu safe shutdown icquirements. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.6 provides a definition of the inspections, test.s. and/or analyses together with associated acceptance criteria which will be undertaken for the 11ECW System. O l l l 9 l 2.11.6 6/1!92 l l
F C
]
e, Table 2.11.6: HVAC Emergency Cooling Water (HECW) System inspections, Tests, Analyses and Acceptance Criteria
- Certified Design Commitment . Inspections, Tests, Analyses Acceptance Criteria
- 1. The system configuration include 3 key 1 Inspection of construction records will be 1. The system cor9iguration conforms with components and flow paths as shown in performed. Visual inspection (VI) will be Figure 2.11.6.
. Figure 2.11.6. performed based on Figure 2.11.6.
- 2. The HECW divisions are mechanically and 2. Tests and VI of the divisions will include 2. Plant tests and VI confirm proper electrict.Ily independent. independent and coinddent operation of independence of each HECW division.VI ,
the three divisions to demonstrate confirm Class 1E power sources for each The HECW divisions are powered by complete divisional separation. VI will HECW division. mdependent Class 1E sources. check for independent Class 1E power ! sources. !
- 3. he standby refrigerator and pump units 3. Tests simulating high temperature cooling 3. Refrigerator cod pump units acting as automatically start upon high temperature . water and operating pump failure will be standby units start upon a high cooling water or failure of the operating conducted for each refrigerator and pump temperature cooling water or operating
@ units. unit in divisions "B* and "C*. Tests pump failure signal. Refrigerator and pump simulating main control room switch units are operable frr m main control room The refrigerator units can be controlled signals will be conducted for the sign als.
individually from the main control room. refrigerator units.
- 4. The HECW cooling capacity is capable of 4. Inspections of vendor documentation will 4. Each refrigeration unit shall have an .
removing the heat loads on the system. include refrigeration and pump capacities. effective heat removal capacity of 2.3 x 108 Flow tests will confirm that adequate flow Btu /hr at 256gpm. Each pump is capable of is available to the system. delivering 256 gpm to the system. i 9 e
ABWR oesign oocument RCW!HECW - A SURGETANK
$ D!G ZONE (A)
COOLING COILS (Reactor Building) e ESSENTIAL ELECTRICAL EQUIPMENT ROOM (A) COOLING ColLS (Control Building) O i HECW HECW HECW (B) (C) CHEMICAL g ADDITION h TANK V HECW N REFRIGERATOR l l (Control Building) HECW h PUMP RCW 4-- RCW (Control Building) O Figure 2.11.6a HECW Division - A 2.11.6 4- 6/1/92
ABWR vesign Document V RCWiHECW B SURGE TANK J'. D'G ZONE (B) COOUNG COILS (Reactor Building) i MAIN CONTROL A ROOM COOLING COILS (Control Building) ESSENTIAL ELECTRICAL EQUIPMENT ROOM (B) COOLING ColLS (Control Building) HECW
- - - i=^N TANK V
HECW N REFRIGERATOR -1 l C (Control Building) HECW JL PUMP R CW 4-- RCW (Control Building) HECW ' N- REFRIGERATOR l l (O (Controi Building) HECW JL PUMP RCW 4-- RCW (Control Building) O Figure 2.11.6b HECW Division B 2.11.c .s. s/1/92
ABWR oesign Document RCWiHECW C SURGE TANK JL D/G ZONE (C) COOLING COILS (Reactor Building) MAIN CONTROL ROOM , . COOLING COILS (Control Building) ESSENTIAL A j
~
ELECTRICAL EQUIPMENT , ROOM (C) COOLING COILS (Control Building) HECW
~
ADD ON TANK V HECW O N REFRIGERATOR ' l (Control Building) HECW JL PUMP RCW-4-- RCW (Control Building) HECW O N REFRIGERATOR l l (Control Building) HECW Jk PUMP RCW 4-- RCW (Control Building) Figu e 2.11.6c HECW Division - C 2.11.6 -G- G/1/92
ABWR oesign oocument 2.11.7 Oxygen injection System m , u <.,- > <.,,,, , ,<>,. ,, ,,<.,,,. i
)
( 2,01.7 , 6/1/92
ABWR oesign Document 2.11,8 Ultimate Heat Sink O ie'c<<><< i'e>". c"v<><a 8 1 >><>e 4 i. A O O 2.11.8 1 6/1/92
ABWR oesign Document 2.11.9 Reactor Service Water System Design Description The function of the Reactor Senice Wates (RSW) System is to remove heat hom the Reactor lluilding Cooling Water (RCW) Systern and reject this heat to the Ultiinate Heat Sink (UHS). The portions of the RSW System that air in the control building are within the AltWR scope and those portions of the RSW System that are outside of the c ontrol building are not in the AllWR scope and are described as interface requirements. The system is classified as safety related and has three separate divisions. __ Design Description Inside the control building, the senice water piping enters divisionally separted aicas and rooms and is sent to and from the RCW/RSW heat exchangers which are part of the RCW system. The requirements in the last two paragraphs of this section also apply to the portion of the RSW system in the control building. interface Roquirements Outside the control building. the pumps, strainers, valves, instmments, and controls are located in the US pump house. Piping connects those portions of the RSW system in the UHS pump house and the control building.
~
The total heat removal capacity of the RCW, RSW, and UHS is sufficient to temove heat loads associated with emergency shutdown and post-LOCA core and containment cooling. The system also removes heat during normal plant , operation and shutdcei. The RSW system is designed to perform its functions taking into consideratio site specific factors. These factors include adequate NPSH for the RSW pumps under all water level fluctuations in the UHS, tendency for organic or macrobial fouling and means for their control and component and piping materials compatible with the UHS water. Measures to prevent flooding in the control building af tei a pipline break will be included. O 2.11.9 cV1/92
ABWR Design Docwnent so ainers, valves, their per supplies and controls in the UHS ptunp house which g is a Seismic Categorv 1 structure and away from high energy piping systems. W lt a loss of piefericd poiver occus s, all RSW system pumps and he.u exchangers will automatically be laced in seivice using diesel generator power. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.9 provides a definition of the inspections, tests, and/oi analvses together with associated acceptance criteria for the RSW system. O O 2.11.9 6'1/92 l l l
0 O N Table 2.11.9: Reactor Service Water System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. Adequate pump NPSH is available for the 1. Available NPSH and requirements wi" be 1. Adequate NPSH is available under all RSW pumps under all anticipated water determined by analysis and compared to anticipated conditions.
level conditions in the UHS. the NPSH of the as-built pumps.
- 2. Provisions will be made to prevent organic 2. Analyses of potential fouling problems in . 2. Design provisions are in place to . - <lude and macrobial fouling. the UHS water source shall be performed unacceptable fouling or degradation of the and compared to the as-built provisions to RSW System performance.
prevent fouling.
- 3. Proper materials for RSW components and 3. Analyses of potential corrosion problems 3. The design has appropriate anti-corrosion piping will be selected. in the UHS water source shall be features.
performed and compared to the g, capabilities of the as-built equipment.
- 4. Provisions will be made to prevent contro! 4. An analysis will be performed of a pipe 4. The control building flooding shall not building flooding if a pipeline break occurs break in the control building using affec+ any other RCW division or any other in or near the control building. conservative assumptions The extent of safety related equipment or areas.
flooding will be estimated based on as-built component characteristics and site-specific UHS.
- 5. RSW System can remove the heat from 5. The heat removal capacity of the RSW S. The heat removal capacity of the RSW RCW System following a LOCA. divisions will be compared with the heat divisions is adequate to remove the heat removal requirements of the RCW System from the divisions of the RCW System.
divisions by evaluation of the as-built components.
- 6. The RSW divisions are separated 6. Inspections and analyses will be performed 6. The RSW divisions are separated mechanically and electrically and are of the mechanical and electrical separation mechanically and electrically and are protected from the events listed in the and the measures to protect the RSW protected against events listed in the De6]n Description. components and piping. Design Description.
M
ABWR oesign Document 2.11,10 Turbine Service Water System Design De'cription The Turbine Senice Water (TSW) System pr ovides cooling water to the Tm hine lluilding Cooling Water (TCW) System heat exchange s to transfer heat from the TCW Systera to the power cycle heat sink during normal and shutdown conditions. During normal operation the TSW System supplies cooling water to the TCW System heat exchangers at a temperature not exceeding 38 degrees C 1 The portion of the TSW System located inside the turbine building is within the scope of certified design, and the portion of the TSW System located outside the turbine building is not within the scope of certified design. , 1 The portion of the TSW System that is within the scope of certified design consists of TSW water supply and discharge lines and manual isolation valves supplying cooling water to the TCW system heat exchangers.The portion of the TSW System not within the scope of certified design has redundant pumping capacity and supplies sufficient water flow to the portion within the scope of certified design. The system configuration of the portion within tbc scope of certified design is t shown in Figure 2.11,10. ! l ) The TSW System is classified as a non-safety-related, non-seismic system. I Inspectiorm Tests, Analyses and Acceptance Criteria Table 2.11,10 prosides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the TSW System. ba 2.11.10 -1 6/1/92
(' Table 2.11.h Turbine Service Water System Inspections, Tests Analyses and Acces:tance Criieria Certified Design Commitment 'nspections. Testc. Analyses Acceptance Criteria
- 1. The configuration of the TSV System is 1. Inspection of the as-built TSW System t. As-built TSW System configuration for shown in Figure 2.11.10. configuration tvithin the scope of certified those componeats within the scope of design shall be performed. certified dasign, conforms with Figure 2.11.10.
- 2. The TSW System certified design portion 2. Inspection of the TSW System portion not 2. The TSW System portion not witSin the recebes sufficient flow with redundant within the scope of certified design. scope of certified <fesign supplies suffreient pump copecity. water flow and has redundant pumping capacity.
is 9 O
@ O 9
)
r. l i . n l \ l $ l l ,, : i ( , i ! l i t t ! P~_____-I r l w2 I e w= a r, . w, j l_ 3 i- _1 i l Y , }l - FROM TSW SYSTEM PORTION r- - l NOT WITHIN 4 l 4 TO POWER i SCOPE OF X l ! X CYCLE HEAT [ CERTIFIED 4 I_,- - p -l_ ] + SINK ! DJSIGN ! r 6 l V l __- _ _ _ _ -r- [ ' ' X X l l i l_ p -l_1 j jl ' V f l l TCW SYSTEU
+ HEAT EXCHANGERS J
t 1 I r t i i t m !
@ Figure 2.11.10 Turbine Service Water System (Portion Within Scope of Certified Design) o t i !
I~ i i s
ABWR oesign oocument 2.11.11 Station Service Alr System i Design Description
'I he Station Service Air (SA n Sutem is designed to puwide a continuous supply of (ointaessed a.* of suitable quality for genend plant use, S; nice air is priinarily used toi t.u.k spaiging, liltes /deinincralifen backwashing, ait openited tools, and ,ther senites icq siring air quality lower than that of the instrument Air Ststem. Another SA System iunt tion is to provine pneumatir backup to the lA System in the event IA system pressme is lost.
The SA System consists ni two air compicuing trains, an ali receiver tank, two tiains of dryer / filter s, distribution p; ping, valves, control and instnunentation. F.ach of the two air compressois is sired to provide 50% of the peak air consumption. One of the two (ompiessors is normally operating while the other i< on standby. The star.Jby compressor automatically starts when the pressure in the neceiver tank drops below the low pressure setting and automatically stops when the normal operating pressure has recovered. The SA System has no safety related iunction except the containment penetration which is icquired to maintin contaimnent integrity. The containment penetration portion is designed to Seismic Category I, Quality l Group It it consists of a check valve inside containment and a manually operated valve on the outside. This manual vahe is h>cked closed during nonnal plant operation and is opened only during refueling to admit service air inside the containment. The SA distribution piping system is non-seismic, and is designed to Quality Group D. Inspections, Tests, Analyses and Acceptance Criteria l l Table 2.11.11 provides definition of inspections, tes's, and/or analyses together with associated acceptance ciiteria which will be undertaken ior SA system. l l l i I l l 2 11.11 .1- 6/1/92 l y--' " 1r-T yyy *---'y-
. _ T 1mv u--r--y-tm- - =- -. -- w- -- - - - - -
y Table 2.11.11: Station Service Air System '
. , 3ctions, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteris
- 1. The ccafiguration of the SA system is as 1. Inspection cf the as-built SA system 1. Verification of the as bt,4Mt system is its shown in Figure 2.11.11. config..ation shall be performed. conformance with the as-desiyced configuration (Figure 2.11.1'l ,
- 2. The SA system outboard isolation valve 2. Actual testing shall be performed to 2. Manual closure of the outboard isolation can be manually closed. demon .' rate that the outboard isolation valve. !
valve can be manually closed. , t i i R i
.O O O
O O O t M > 5 i 5 ! ASME ASt.1E TJC CLASS 2 CLASS 2 tJC l u __m _ __ __ _ _ _ _ _ _ _
' I l l Y Y V Y Y- y y V y v v y i l TO AIR HOSE REACTOR BLDG EOU1P cot 4 TROL BLDG EQUIP RADWASTE BLDG EOUTP CONtJECTIONS i i {
PRIMARY l CONTA!NMENT REACTO'1 BLDG. CONTROL BLDG. RADWASTE BLDG. q A I 43 l l [- ! Y Y Y P4STRUMENT 4 -- N - - 3 i'
. TURBINE BUILDING EQtJiP AIR SYSTEM IA SA i
A (' AFTER ' COOLER O ] A --d A}- A - Ml - ! v i COMPRESSOR MOISTURE I AIR GRYER/ ] SEPARATOR ----> - FILTER 1 AFTER I Al l COOLER O I , B --d B H B N - [hk}., ag V AIR RECEIVER j COMPRESSOR MOISTURE AIR DRYER / 1 SEPARATOR FILTER TURBINE BLDG. e i 9
; ;S i Figure 2.11.11 Station Service Air System ;
I i I s
ABWR oesign Document 2.11.12 Instrument Air System Design Description The Instnnnent Air (IAi spicm is designed to proside a contimnius supply of (lean, dn and oil free compr essed ali f or pneumatic equipment, vah es, controls i
. uni instnnuentation outside the priman containtnent.
l l Poition of the lA System distribution piping penetrates the containtuent but is nonnally isolated f rom the 1A suppl > by a normally closed air operated valce. Pnemnatic supply to this line is normally supplied by the liigh Pressui e Niuogen Gas Supph- (HPIN) System which supplies the nitrogen gas f or the containment annosphere so the containment is normally inert. In the event HPIN pressure is lost. lA System provides air backup to the equipment that requires nitrogen innle wntainment by remote manual alignment of the associated vahes to the lA Sptein. During refueling, IA aho piovides air supply to equipment that icquires inside containment through this line. The 1A System consists of two air compressing trains, an air receiver tank, two dning trains in parallel, distribution piping, valves, control and instnnnentation. Each compresting train consists of suction filter,100% oil free , air compressor, after cooler and moisture separator. One of the two cornpressors ! is normally operating while the other is on standtg. The standby compressor l I
- automatically starts when the pressure in the receiver tank drops below the low piessuie setting and automatically stops when the nonnal operating pressure has iecoveicd. In the event of an unusual drop of air receiver pressure, the SA Systern provides ali supply backup to the instnnnent air users.
1 The lA System has no safety related f unction except the containment l penetration which is required to maintain containment integria The containment pencuation portion is designed to Seismic Category 1. Quality Group It it consists of a check vahe inside contamment and a motor operated udte on the outside. The IA distiibution piping system is non-sciemic, and is designed to Quality Group D. The IA System is connected to emer gency power for continued operation during a loss of off-site power event. Inspections, Tests, Analyses anct Acceptance Criteria Table 2.11.12 provides definition of inspections, tests, and/or analyses together witn associated acceptance criteria which will be undertaken for IA system. i i 2.11.12 6/1/92 l I. _ . _ , . . , _ _ . _ _ _ _ , . . _ , , , _ . _ .
h' Table 2.11.12: Instrument Air lystem a Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analpes Acceptanta Criteria
- 1. The configuration of the IA System is 1. Inspection of the as-built IA System 1. Verification of the a} built conformance i nwn in Figure 2.11.12. configuratnn shall be performed. with the as-designed configuration (Figure 2.11.12).
- 2. The IA System outboard isolation valve 2. Functior al testing shall be performed on 2. Valve isolates upon receipt of auto closes epon receipt of auto isolation signa? the system logic by simulating the auto isolation signal. >
from the leak Detection System. isolation signal from the Leak Detection , and Isolation System. l
- 3. The lA Systes capability to operate en on-
. 3. lA System functional testing shall be 4. Satisfactory IA System operation wnh site emergency AC power source. performed to demonstrate operation when power supplied from on-site emerg sney [
i supplied from on-site emergency AC AC power sources f power sources. [ 9 i i i CD i O O O
I i O- O O n PRfMARY COtJT. EQUIP 1 I 4 i ~ I
}~ 3 I
+ Y Y T --
d ( REACTOR BLDG EQU:P - - - NC c p u Y 7 Y I Y 7 Y } 4- I E!SOLATED Y l i SIGNAL CONTOLBLDG SERVICE BL'JG RADWASTE BLDG I i ig_ g _, 1 EQUtp i EQUIP EQUIP l 3 ASME NC i CLASS 2 ! PRIMARY HPIN !A 1 l CONTAINMENT REACTJR BLDG. CONTROL BLDG. y SERVK:E BLDG.
' RADWASTE BLDG l f I
@ FROM _ AW _
j' i SERVICE AIR Y Y Y TURBINE BLDG EQUIP SA
- A 3 '
JL A AFTER ~ ~ y _ COOLER O ' _ .N (- - A --- Q A A -d A F- A N l v i AIR DRYER k CCMPRESSOR MOISTURE l SEPARATOR -- hSUCTION FILTER
/ AFTER I A I I COOLER O I
~
8 --! B l-- B N I _y__ B . - _ V AIR RECEIVER COMPRESSOR MOISTURE AIR DRYER SEPARATOR TURBINE BLDG. m . G? l
~ i Figure 2.11.12 Instrument Air System i i
f
. _ 1
_ _ _ _ _ _ . . _ _ __ __ m __ A8WR oesign Document 2.11.13 High Pressu e Nitrogen Gas Supply System Design Description The High Picsstue Nitrogen Gas Supph t HPIN) System is designed to pioride nin ogen gas to pneumatic equipment inside primary containment. The llPIN Sutem consnts of two independent subsystems one being safety-related and the other non4afen-iclated. The nonaalen-iclated por tion icceives its nitrogen gas source hom the Atruospheric Connol (AC) Svstem and distributes it inside containment for the following couipment: (1) irlief f unction accumulators of main steam safety / relief valves. (Y) nin ogen operated udves and instnunents inside c ontainment. (3) leak detection system radiation monitor calibration. N) Automatic Depressuri/ation System ( ADS) funcuan accumulatms of the main steam safety / selief udves to compensate leakage durin nonnal I operation. Following a LOCA, nitrogen supply to the ADS f unction accumulators are supplied by the safety-related liPIN subsystem. The safety elated subsystem consists < t two redundant divisions supplied f rom high pressure nitrogen gas storage bottles. Each division is mechanically and electrically separated from th" I other. One division supplies nitrogen to half of the ADS designated safety / relief vahes and the other division for the remaining half.The nitrogen storage bottles suppiv valve is normally closed with key lock control switch nonnally in " auto" mode. Remote manual closure and opening can only be accompl ished with the key. The supply valve automatically opens in response to low pressure condition i in the ADS accumulator supply line. During this emergency mode of operation, I power to the safety-related HPIN subsystem is automatically switched to dwisional emergency AC power sources. Separations between the safety-related and the non-safety related portions of the HPIN System are provided by motor operated shutoff valves that automatically close on low pressure condition in the ADS and non-ADS SRV accumulator supply lines. The non-safety-related portion is d signed to non-seismic class, Quality Group D, while the safety-related portion is Safety Class 3, Seismic Category 1, Quality Group C, Electrical ( lass 1E. The shutoff valves separating safety-related from the non-safety-related portions are Seismic Category 1. Quality Group C design All primary containment penetrations meet Seismic Categmy 1. Quality Group C design 1equirementS, 2.11.13 1 6/1/92
ABWR oesign Document . l' oin the ninogen gas bottles up to the pressure ieducing valve is orsigned to 2no Lg o in g 12N45 psig >. This is tr ue foi both divisions of the !IPIN safety-g iclated subssstein. The reinaindet of the llPIN $ntenu including s.tfetprelated 2
.unt non-safety-s elated portions is rated at IS kg ~cm g (50 psig).
The llPIN Svstern is prosided with instnnnentation and < ontiols to monitor the
.ssstern design basis operation. These include high and low presstn e alanns, indications, vahe I:osition status lights, and othen s.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.13 i n ovides definition of inspections. tests, and/or analyses together with asswiated acceptance criteria which will be undertaken for the llPIN Systein. O l O z,,.,, . ,,, 1 i i
y Table 2.11.13: Remote Shutdown System ,
'-- t Inspections, Tests, Analyses and Acceptance Criteria !
Certified Design Commitment inspections, Tests Analyses Acceptance Criteria !
- 1. The configurata,o of the HPIN Syutem is 1. Inspection of the as. built HPIN System 1. Verification of the as-built conformance ,
shown in Figure 2.11.13. :nfiguration sha!l be performed. with the as-designed configuration (Figere ! 2.11.13).
- 2. The nitrogen gas bottles supply valve 2. Using simulated high and low pressure 2. Automatic opening and c!csing of the ;
automatica"y apens on low pressure and signals, functional testing of the system nitrogen gas bottles supply valve. l automatically closes on high pressure logic shall be performed to demonstrate conditions at the ADS accumulator supply automatic opening and closing capability ' i line. of the nitrogen gas bottles supp!y valve , with the control switch in " auto
- r.iode.
- 3. The nitrogen gas bottles supply valve 3. Demonstrate remote manual actuation of 3. Remote manual open/close actuation !
remote manual operability. the nitrogen gas bottles supply valve from from the main control room with key. No the m An control room wit 5 key. valve actuation when key is not used. ' O i
- 4. The safety-to-non-safety related interface 4 Functional testing utilizing simulated 4. Auto closure of the safety-to-non-safety l
shutoff valves automatically c!cse on low signals shall be performed to demonstrate interface shutoff valves. pressure condition on the ADS and non- auto closure of the safety-to-non-safety v ADS accumulator supply lines. interface shutoff valves on low pressure condition at the ADS and non-ADS l accumulator supply lines.
- 5. The safety-related portion of HPIN System 5. Demonstrate automatic power switching 5. HPIN System power switching and HPIN automatically switches power to and HPIN System operability when Systems operability on emergency AC emergency AC oa loss of normal power supplied from the emergency AC scurces. sources. ,
supply. I
- 6. HPIN outboard containment isolation 6. Demonstrate remote manuel closure 6. Valves remote manual closure from the l valves remote manual closure capability. capability of tha HPIN cutboard main control room.
containment isolation valves from the main contro! room.
- 7. I'rovision for control room alarms, and 7. Inspection shall be performed to verify 7. The control room alarms and indications
<n indicatiens vital for HP!N operation. presence of control room alarms and specified in Section 2.11.13. , g mdications. !
o , 1 i
4 pJ INSIDE PCV f r f r o A p g - vvvvv g3 m on T [DC
@ -M > ADS SRVs l
ASME ASME ,
' 3 + D'i 2 CLASS 3 CLASS 2 3 M I NITROGEN GAS BOTTLES DIVISION 2 1 I
=
M [l __ M y ["t,- ^ M - h r in3-- - ELG - -
-k i Dt2 -W ,
> "s 3
S I C3 TO INSTRU-i NC 2 2 31 AC HPIN MENT AIR NC
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3 h " i AAAAA U M [ -[ i H 4 ADS u v v v.V SRVs
.s g J EDC CXX}- -- N > Div.1 8
a 3 2 2 3 l YYYY _ NITRO 6EN GAS BOTTLES DIVISIO'l 1 c> N, Figure 2.11.12 High Pressure Nitrogen Gas Supply System ' O O O
1 ABWR oesign Document 2.11.14 Heating Steam and Condensate Water Return System Design Description The lie.uing ste.un aiul Coiulensate Wate Return (liSCR) Systein supplies heating ste.un fioin the house boiler for general pLnt use and ietoveis wndens.ue to the boiles feedwater tanks. The systein consists of piping, valves and aunt iaird (ontiols uul instnunentation. Tlic ilS(X sssterii is classified as a tior.-salet3-rehited. iiori-seisitiic systeiii. Inspections, Tests, Analyses and Accepsance Criteria No Tier 1 ITAAC ior this syste:n. O T O 2.11.14 6/1/92
ABWR oesign Document -. 2.1'i.15 House Boiler Design Description The Ib use ik>iler (1115) Systern consist 3 of the in >use lioileis. ietxnter s, fecclwatei c oinponents, b< ih . riter treatinent and u>ntn>l devices. The 1111 cystein supplies turbine gland steatn and heating ste.un The lilt .spiern is classified as a non-safetv iciated. non seistnic systeni. Inspections, Tests, Analyses and Acceptance Criteria No Tier 1 ITAAC for this syste: I 1 i 1 I l 2.11.15 6'1/92
ABWR Design Document 2.11.16 Hot Water Heating System Design Description The llot Water lleating (11Wiii Systein is a closed loop hot water supply to the vuions heating coils of the li\ AC Syste ms. The liWil Systern inc ludes two heat exchangers. a bat Lup heat eu hanger, sing an ; cheinical addition tanks, and esoriated equipn> cut, control and instnunentation. Tlic 11Wii Svstern is classified as a non-salety-renated, non-seismic system. Inspections, Tests, Analyses and Acceptance Criteria No Tier 1 ITAAC for this systein. p O 2 11.16 -1 6/1/92
ABWR Design oocwnent 2.11.17 Hydrogen Water Chemistry System Design Description The imhogen watei (immistn (HWC) system is used, along with othe - measin es, to icduce the hLelihood of conosion faihu es whit h would adsen ly afic( t plant availahihiv. BWR icartor coolant is elcmineralized water, typically containing 100 to "00 parts pei billion (ppb) dissolved oxygen hom the radiolytic decomposition of water. The function of the HWC system is to reduce the dissohed oxygen in t'ie seat tu water to less than "O pph by the addiuon of hydrogen to the feedwater. 'l bl.s reduction has been demonstrated to be highly elfcctive in the mitigation of the potential for intergranula suess conosion cracking (IGSCC) of sensitized aintenitic stainless steels. The concentration of hydrogen and oxygen in the main steam line and eventuallt in the main condenser is altered in this process. This leaves an excess of hydrogen in the main condenser that would not base equivalent oxygen to combine with in the of fgas system. To maintain the of fgas system near its normal operating characteristics, the HWC provides a flow rate of oxygen equal to approximately one-half the hydrogen flow rate, injected into the offgas system upsn. am of the secombiner. The HWC cystem is composed of hydrogen and oxygen supply systems, systems to inject hydiogen in the fecchvater and oxygen in the oilgas and subsystems to monitor the effecti,eness of the HWC system. A number of automatic control features are provided in the system to minimize the need for operator attention and to improve performance. Such controls include automatic variation of injection flow rates with reactoi power and automatic shutdown for several alann l conditions. The hydiogen water c hemistry system is non-nuclear and non-safety-related. It is iequired to be safe and reliable, consistent with the requirement of using hydrogen gas. The hydrogen piping in the turbine building is designeo to Seismic Category I requirements. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11,17 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the iIWC system. 2.11.17 1 6/1/92
y Table 2.11.17: Hydrogen tvate-Chemistry System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The simphfied system conaguration of the 1. Inspection of installation rere . ;s together 1. The as-built HWC system configuration is HWC system is shown in Figure 2.11.17. with plant walkdowns will be cunducted to in accordance with Figure 2.11.17.
confirm that the installed equipment is in compliance with the design configuration defined in Figure 2.11.17.
- 2. The nieans of storing antf handling 2. Perform a c>afety review of system 2. Reviews and inspections verify that the !
hydrogen shall t>e safe and reliable, operating procedures relating to the equipment and procedures for the storage ' cusistent with normal industry practices storage and handling of hydrogen and a and handling of hydrogen 3re safe and l for prevention of hydrogen fires and safety inspection of all hydrogen reliable. explosions. processing equipment. 9 3. The hydrogen piping in the turbine 3. Procurerr.ent records, design documents 3. Records, documents and inspections verify building shall be designed to Seismic and actual equipment shall be inspected to that the hydrogen pig.ing in the turbine Cateoory I requirements. verify that the requirer %ents are met. building is designed to Seismk Category I requirements. I k l i O O O
4 i o i n 1
? ,
U i t i l H ISOLATIOf4 7 FEE'- VATER VALVE jk l HYOROGEN N' l Supply ' INJECTION ; f.*ODULE j l ! T V i CoNrROL 4--------- invc i ROOM EFFECTIVENESS
- imc 4- - - - - - SUBSYSTEMS CONTROL l PANEL i H2AREA
! uONiroRs --------* +---------, ' e k A i I i i I I 1 i i I l 1 l I 4---- h l OXYGEN O* m OFF-GAS SUPPLY - > INJECT M -
+ 0 3ANALYZER l MODULE O, ISOLATION VALVE OFF-GAS jf > RECOMBINER 1f OFF-GAS PROCESS LINES e - - - - - -
SIGNAL LINES Figure 2.11.17 Hydrogen Water Chemistry System
ABWR Ocsign Document ' 2.11.18 Zinc injection System No Tier 1 entiy for this wstern. O 1 I \ l I ' l l l l I
\
l O !, I 2.11.18 6/1/92 I
ABWR oesign occument 2.11.19 Breathing Air System Design Description The bre.uhing ab system provides a continuous supply of ali to wo:Lers within c ontainment. The bre.ithing air system does not serve or support any safety function and has no safety design basis. The bicathing air systern takes air from the senice air sp<cin. purifies it and makes it available to workers. Inspections, Tests, Analyses and Acceptance Criteria Table '.'.11.19 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria which will be nsed foi the bicathing air systein. i l l \
? 11.19 Cv'1/92 1
_ , , , , 9,-
, - - - , 9
i y Table 2.11.19: Breathing Air System Inspections, Tests, Analyser and Acceptance Criteria , Certified Desigt. Requirement inspection Test, Analyses Acceptance Criteria ,
- 1. Quality of air produced by breathing air 1. Air quality shall be tested and compared 1. Air quality standards met.
system meets applicable OSHA standards. with OSHA standards. , 4 t h' l i L m
)
O O O ,
ABWR oosign occument 2.11.20 Process Sampling System Design Description The Process Sampling ( Pm Swein is designed to provide s.unpling of all principal fluid process stic.uns associated with plant operation. Representative s.unples are taken foi analnis . uni pr ovide the analytical inioi mation equir ed to monitor plant and equipment perfonnance. The PS System consists of: (1) Pennanently installed sampling noules and sample lines. (2) Sampling panels with analyzers and asw,ciated sampling equipinent. (3) Provisions for local grah sampling. (4) Permanent shielding. (5) Casks for storing and transporting samples. The seismic design and quality group classifications of sample lines and their components shall confonn to the classification of the system into which they aic connected, up to and including the block valve (or ndves), or, in the case of the ( reactor water sampling lines, the second isolation valve. The downstream s;unpling lines are Quality Group D. . Sampling is available from the post accident sampling station (PASS) following a LOCA or ATWS event. All PASS sampling valves are operated remotely. The PASS isolation vidves are operated from the main control room using Class 1E power sources. All other valves are operated from the local control panel with two offsite power supplies. Inspections, Tests, Analyses and Acceptance Cr'teria Table 2.11.20 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be used by the PS System, p 2.11.20 1 0d1/92
x / w - --
" Table 2.11.20:
Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests. Analyses Acceptance Criteria Certified Design Commitment
- 1. Visual inspection (VI) will confirm that a 1. The post-accident sampling station is
- 1. The system has the capability to perform provided.
post-accident sampling. post-accident sampfing station (PASS) is provided.
- 2. VI will include *.he isolation valve electrical 2. Plant tests and VI confirm Class 1E power
- 2. The PASS isolation valves are connected to sources and proper isolation valve Class 1E sc.,urces. connections.
operation under LOCA signals. The PASS isolation valves may be opened Tests simulating a LOCA signal will be for sampling during an accident without performed while the isolation valves are removing the LOCA signal. operated.
- 3. VI of the PASS will review the presence of 3. Shielding and transporting casks are
- 3. The PASS provides shielding and sample sample shielding. provided at the PASS.
transporting casks. Y l e T
'E r
O O O
ABWR oesign Document 2.11.21 Freeze Protection System Design Description The h eere l'iotection Mstein pnisides insulation, sic.un. .uul electrical heating for all extcinal tanks and piping that nias ficcie < luring wintei weathen. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.21 piosides definition of the inspection, tests, ainl/or analysis. togethei with associated criteria which will be undeitaken for the Freeze i l'it eter tion Mstent. 1 l l l l l , l \ l 1 l l l O 2,11.2 ', 1 6/1/92
] Table 2.11.21: Freeze Protection System inspections. Tests, Analyses, and Acceptance Criteria Certified Design Commitment Inspections, Tests. Analyses Acceptance Criteria
- 1. Provide insulation, steam, and electrical 1. Visual inspection will be conduded to 1. Confirmation that the required freeze heating for all external tanks and piping confirm that the insulation, steam, and protection has been installed.
that may freeze during winter weather. electrical heat provisions are insta!!ed as required. i i i I f 9 2 t O O O .
ABWR Design Document 2.11.22 fron injection System No Tier i entry hir this ssqcin. l l l \ l l l IO l l I l O 2.11.22 ,3, 6l1/92
ABWR Design occwnent 2.12 Station Electrical 2.12.1 Electrical Power Distribution Systern Design Description The plant Elec uical Power Distribution (1 PD) System is a complete three-load group distribution system with two independent off-site power sources (Normal Preferred and Alternate Preferred), the Alain Turbine Generator, three on-site Standbv Power Somtes (Emergency Diesel Generators), and a Combustion Turbine Generator (CTG) located on-site. This three-load group configuration, with multiple power sources, reduce, the challenge to plant salety systems by incicasing plant icliabinty. Any one of the three-load groups can safely shut down the plant and maintain safe shut down. The CTG provides an additional diver se power supply to back up safety system power supplies,if needed (Figure 2.12.1 a ) . During normal plant opemtion, the main generator supplies power to the blain Power Tnunformer (h1PT) and the three Unit Auxiliary Transformers (UATs) through a main generator output breaker and an isolated Phase Ilus. When the main generator is off-line, power is supplied to the UATs from the SIPT(Normal Preferred Power). Each of the three UATs supplies power to a separate load group. One winding l' of each tnmsformer supplies power to one nonessential medium voltage (6.9 kV) Power Generation (PG) switchgear and through a bus tie breaker to a Plant , investment Protection (PIP) switch gear.The second winding supplies power to ! a second non+ssential Power Generation medium voltage switchgear and to an l Essential Safety System (ESS) medium vohage switchgear. Power from the UATs to the medium voltage switchgear of the three non-essentialload groups and to the first set of medium voltage circuit breakers feeding the three essential medium voltage switchgear is supplied through Non-Segregated Phase !!uses. The Reserve Auxiliary Transfonner (RAT) is the Alternate Preferred Power source and is preferably lined up to supply power to one of the three ESS switchgear. One winding of the transformer can supply power directly to three non essential Power Genemtion (PG) and three non-essential Plant invettment Protection (PIP) medium voltage switchgear and through bus tie circuit breakers to the other three non-essential Power Generation medium voltage switchgear. The second winding can supply power to all three Essential Safety System medium voltage switchgear (Divisions I,II,111). Power from the RAT to all of the medium voltage switchgear of the three essential and non<ssential load groups is provided through Non-Segregated Phase Iluses. t Each ESS (Divisions 1, II, Ill) medium voltage switchgear is normally supplied power fr om its associated UAT or from the RAT. In addition to these power 2.12 6/1/92
ABWR Design Document soon es each E5S medium voltage switchgeat is prosided with its own dedicaird Standin Power supph. In the esent of low vohage on the switchgear bus (e g.. loss of oil-site power i, the awociated Emer geng l>icsel Generator auton.atically stas .s alHl. allel aulHinQ that all other inplit lcedet hicakers are open, automaticalh < onnects to the bus to supply emer gency pow"r. Each emergency bus can ab he supplied powei hom the CTG. All hu. transfer operations to the Class 1E buses except for the automatic connection of each dedicated emelgencs diesel generator, are manual only. Each had group of nonessential medium voltage switchgear ir, supplied power from its associaird UAT with an ahernate supply ham the RAT . In addition to these powes sources, the tlu ce non essential PIP :nedium voltage switchgear can he c onnected directly to the CTG. On loss of vohage to a pre-selected PIP bus, the CTG will automatically start and, aften asstuing that all other input feeder and bus tie becakers are open, automatically conon ts to the affected bus. Ilowevu, oniv the two pieselected buses of the three huses will connect automatically to tb" CTG. All other non-essential bus transfers are manual only. i Sfedium voltage .\tetal Clad Switchgear (h!/C) supply power to large loads (typically larger than 300 kW) and one or more medium voltage (6.9 kV) to low voltage (450V) Power Center Switchgear (P/C) transformers in the same non-essential load gioup or salety division. Power Center Switchgear supply power to medium size loads (typically between 100 to 300 kW) and multiple low voltage g (480V) hiotor Control Centers (hfCC) in the same non-essentialload group or salcty division. Stotor Control Centers supply power to smaller loads (typically less than 100 kW), including lighting,120 VAC instnnnents, power, and control equipment. With one exception, ESS switchgear and non-essential system switchgear are not inteironnected enept by common non-essential medium voltage power supplies. The one exception is the Fine Station Control Rod Drive Power Center Switchgear. One of the essential medium voltage switchgear supplies the preferred power to the non-essential Power Center through a series of one feeder circuit breaker and one tmnsfer switch. The feeder breaker is Chus lE. De feeder cable, transfer switch, Power Center transformer, and interconnecting cable to the Power Center bus input breaker are classified as Associated 1E. One of the non-esse - al medium voltage PIP switchgear supplies the ahernate power to the non-essential Power Center through a series of one feeder circuit breaker and the transfer switch. The feeder breaker, feeder cable (to the contacts of the transfer switch), tmnsformer output breaker and Power Center are non-essential. Automatic transfer of the transfer switch is only from l the essential to the non-essential power source on loss of essential bus voltage. Transfer back is manual only ( Figure 2.12.lb). 2.12.1 6/1/92 l l l
ABWR oesign Document Unit Auxiliary Transformers : The size of each Unit Auxiliarv Transformei (UAT, is selected such that it will proside the iull power requiieinents of its associated load gioup without exceeding it.s air / oil rating during normal 100'1 plant operaan (e.g., all du ce
} load grot.ps available) and will not exceed is t- , ed aii ' oil rating with one load % group out of senve during 100% plant operat.on. Each tiansformer has two i secondan windings as de.sciihed above and will proside power at 6.9 kV with a nominal mput voltage of 27 kV. Transfonner impedance is selected to limit the '
output voltage decicase to a nutximum of 209 during the starting e motors and to limit the f ault current to less than the maximum inte. .ing capacity of the circuit breakers while maintaining the required bus voltage regulation. The tluce UATs are separated irom each other and from the Mam Power Transformer by aadow fire walls. The UATs are also teparated from the w L , rve Auxiliary Transionner by a minimum o! 50 reco Each UAT is piovided wnn its own oil pit and drain. Giounding and ;ightning protection is provided. 5 Reserve Auxiliary Transformer: The size of the Resenr Auxiliary Transfonner (RAT) is selected such that one of the a condary windings will provide the power requirements of the loads on one ,, tw aon essentiai load group at 100% plant power operation and the second h winding will provide t e power requirements of all three divisions of Essential Safety Syste.u (ESS) loads without exceeding its air / oil rating. The transformer mtio and impedance is selected to provide 6.9 kV (+/-10%) with a maxinuun frequency variation of o -2% at 0.9 powei factor loaa and a maimum voltage decrease of 20% during the narting oflarge motors, assuming nominalinput vohage ar.d frequency. /s frequency variation of two cycles is acceptable dtuing [ periods ofinstability of the input. Ingedance is also selected to limit the fault current to less than the maximum intermpting capacity of the circuit breakers while maintaining the required bus regulation 1ne RAT and its input fecaer., are separated from the Main Pewer Transfo i er and its input feeders and from the UATs by a minimum of 50 feet. The RAT is provided with its own oil pit and drain Grounding and lightning protection is ptovided. Switchgear and Breakers: The Main Genuator Circuit Breaker is sized to handle the main generator full load output at a nominal voltage of 27 kV and to interrupt the maximum ' v.lculated fault current occurring at the breaker. It is equipped with redundant trip coils supplied from separete non-ssential on-site 195 VfsC batteries and is h>cated approximately michray between the Main Generator and the Main Power Transfonner. 2 12.1 G/1/92
l ABWR oesign Document Each f eeder hom a tJAT to its respective ESS swirrhgear is prosided with a stub g bus and rin uit breaker to facilitate the transition hom the Non-Segregated W , Phee lius to cable. All .\letal Clad and Power Centers switchgear, and hiotor Connol Centers are identified according to their Essentiality, Load Group or Disision and their voltage lesel (ti.9 kV,480V) and are physically separated accordingly. Divisional switchgear ar e qualified Essential Class 1E and are lot aied in Seismic Category I structures and in their respective divisional electrical equipment rooms or fire areas. Essential equipment rooms and fire areas ar e sepanned by turce-hou fn c baniers. Non-essotial switchgear are separated by appropriate distances between nonessential load groups. Switchgear and associated transformers (e.g., Power Centers) are selected for their intended senice and loac, iequhements and are rated to sustain the maximum calculated fault current under all modes of operation until the fault is cleared. Feroer and load cirruit breakers are sized and rated to provide the load requirements under all expected operating modes and are capable of interrupting their maximum calculaterl fault currents. Both switchgear and associated transformers are grounded. In addition, each r. .;dium voltage Sletal C!ad switchgear is pr osided with a Safety Ground Circuit Breaker which is racked-out during normal operation and is interlocked with bus voltage and its related bus feeder breakers to prevent inadvertem *sure. The breaker is annunciated in the main control room when it is in the racked-in position. Switchgear and motor control centers are provided with the manufacturer's recommended fault current and protective devices as required by the fault cunent and breaker coordination analysis performed during the implementation stage of the design. Fault current ar.d breaker coordination analysis for Class 1E equipment is coordinated with the nonessential equipment load groups. Analyses consider the impedante ofinterconnecting cables and buses, and load cables. Control and instrumentation power for each switchgear is provided from the associated divisional or n,n-essential power train 125 VDC battery. For power circuits prmding power th.ough primary containment ' penetrations, a redundant overcurrent protective device is prosided in series witF the circuit breaker if the calculated fault current could exceed the maximum continuous current rating of the penetration. In addition to the normal protective features, zone-select interlocks are provided on the input feeder breakers to the essential switchgear supplying power to the Fine 5fotion Control Rod Drive (FMCRD) Power Center. The interlocks are provided to delay tripping the essential switchgear input feeders until the normal overcurrent device on the feeder to the non<ssential FhtCRD Power Center has had time to trip and clear any fault. Electrical power generation and distribution parameters needed to assurc plant O reliability and safe shutdown am provided in the hiain Control Room and to the 2.12.1 6/1/92
ABWR Design Document _ Remote Shin wn Scstem. These paranieters inchide pouer distribution system bicaker positions. voltages, amperes, kVA. LW. and power facto- ~a addition, remote conn ol of wlected power generation circuit breakers, including synchronizing capability, is provided in the control room. Phase Buses and Cables: The Isolated Phase Bus is seh eted to cany th lain Generator full load output at a nominal'!7 LV and rated to sustain the maximmn calculated fault current until lhe fauh is cleaird. Disconnect links are piosided in the feeds to each of 5 the UATs to f acilitate maintenance and isolate a f aulty transformer. A main generator bicaker is also prosided as described above. The Isolated Phase Bus housing is grounded at both the hiaia Generator and the Main Power Transformer ends of the bus. The Non-Segregated Phase Buses are selected to cany the fullload at 6.9 kV to which they will be subjected under all medes at operation and are rated to sustain the maximum calculated fault current until the fauit is cleared. Buses are idendfied ac< cding to voltage level and load group and are grounded at the same point as the suitchgear to which they connect. Power Distribution System cables are selected for site and insulation based on their voltage, senice load routing, and environmental conditions, including tempemture, humidity, and radiation, to which they may be exposed. Ratings i an 1 loading of the .3 elected cables assures that they can sustain the maximum calculac d fault currents to which they may be subjected until the fault is cleared. Cable impedance is corJdered in the overall distribution system protection analysis which will be performed during the implementation stage of the design. Selection and application of cables is intended to assure a life expectancy of 60 years. Cables are identified according to voltage levels, non<ssential load group, essential division, and function. Independence and Separation: Electrical independence of equipment is provided by three separate load groups which are functionally redundant and capable of supporting plant operation at 50% ofits rated output. There are no automatic connections between the load . groups. Each load group is supplied by a separate power source unless connected to the Combustion Turbine Generator. Electricalindependence of Euential Safety Systems is provided by three separate safety divisions with their own dedicated emergency diesel generator. There are no automatic bus ties or power supply transf ers bet 4 ccen diviJons. Normal Preferred Power to each division is from a separate non-essent al power transformer. T he only on line connection between a afety division and a non-essential load is the divisional power feed (as described above) r. the FMCRD Power Center. 2.12.1 5- 6/1/92
ABWR oesign Document Transformer and switchgear separation is described above. Essential Saf ety g d' vision cables are routed in Seismic Categon I structures and dedicated
. W disisional raceways which are separated from each other such that tolerance is prosided for a complete burnout of a single fhe area. Non esseritial cables, if routed with divisional cables. are treated as Class 1E Associated. Cables of different divisions are not touted through a coremon hostile aiea except where justified by analysis (e.g., priman containment).
Separation of non-essential Normal Preferred and Alternate Preferred Power feeders is maintained by iouting through difIerent areas of the turbine and reactor buildings and by distance when routing across the control building. Non-essentialload group separation between feeders from the UATs to the divisio rd switchgear is provided by routing cables in separate raceways. A separate CTG feeder cable is provided to each of the three divisional and three non-essential switchgear to f acilitate maintenance and fault clearance. CTG feeders to the reactor building follow a similar routing scheme to that used for the Alternate Preferred Power feeders. In addition to the above separation, raceways are separated according to voltage levels and functions within divisions and load groups (e.g., low voltage control cables are routed separate from medium voltage power cables). Raceway are g provided with grounding connections. W Grounding: TI electrical grounding system is comprised of: (1) an instrument grounding networt Mr grounding ofinstrumentation and computer systems; (2) an equipr em grounding network for grounding electrical equipment (e.g., switchgear, motors, distribution panels, cables, etc.) and selected mechanical components (e.g., fuel tanks, chemical tanks, etc.); (3) a lightning protection network for protection of structures, transformers and other equipment k>cated outside buildings; and (4) a plant grounding grid. All grounding networks are insulated from each other and separately grounded to the plant grounding grid outside the structures. All grounding networks and equipment are low resistance grounded except the main generator, the cmergency diesel generators, and the FTG, which are high resistance grounded to maximize availability. All components requiring grounding are identified and provided with grounding connections. Inspections, Tests,,,nalyses and Acceptance Criteria Table 2.12.1 provides a definition of the aspection, test, and/or analysis together with associated acceptance criteria which will be undertaken for the Electrical Power Distribution Systmn. 2.12.1 6/1/92
U Table 2.12.1: Electrical Power Distribution System M. Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The elects. cal power distribution system is 1.a inspections of the distribution system 1.a Each of the three-load groups is supplied a three-load group distribution system with configuration will be performed to confirm power from a separate UAT.
two off site power supplies, three on site each load group is supplied by a separate emergency generators, a Combustion Unit Auxiliary Transformer. Turbine Generator located on site, and the main generator "sith its output circuit breaker. An isolated Phase Bus connects the Main 1.b Inspections of the isolated Phase Bus and 1.b isolated and Non-segregated Phase Buses, Generator to the Main Power Transformer Non-segregated Phase Bus Installations, with associated main generator breaker and Uait Auxiliary Transformers (UATs) including the main generator breaker and and disconnect links, are provided. through the Main Generator Breaker and disconnect links to the UAT, will be through disconnect links to the UATs. performed. U Non-segregated Phase Buses connect the UATs and the RATS to their associated sw;tchgear breakers and the first in-line breakers providing power to the Essential Safety System (ESS$ switchgear. Each UAT connects to two non-essential 1.c *nspections of the transformer and ot!.er 1.c The transformers, emergency diesel i Power Generation switchgear and one ESS power sources and their power feeders will generators, and Combustion Turbine [ switchgear it' its c en load group. be performed to confirm their location and Generator are located in accordance wh ! connections to the t.pecified switchgear. the certified design and connect to the :
- The RAT connects co tbree Power specified switchgear.
Generation, three P P, and three FSS l switchgea r. i l The CTG connects to the three non-essential PIP and the three ESS switch,sar. Each EDG only connects to its own ESS 3 switchgear. (See Figures 2.12.1a and b) 8
3 Table 2.~r 1: Electrical Power Distribution System (Continued) N inspections, Tests, Analyses and Acceptance Criteria Certified Design Coinmitmant inspections. Tests. Analyses Acceptance Criteria
- 2. Unit Auxiliary Transformers are sized to 2.a inspection of load assignments v.ill be 2.a Transformer namepf ate ratings will not be provide full load requirements within their performed to assure transformer exceeded during two and three-load group air / oil rating for 100% plant operation (with nameplate ratings will not be exceeded operating modes.
all three-load groups available) and will not with all expected loads operating during exceed their forced air / oil rating with one either the two or three-load group load group out of service. cperating mode. U ATs have two secondary windings and 2.b inspections and tests will be conducted to 2.b Transformer ratios provide output voltages will provide a nominal voltage of 6.9 kV confirm that transformer ratios provide that are consistent svith input voltage 3 and with a nominalinput voltage of 27 kV. cutput voltages on both windings that are output voltages do not decrease below Output voltage will not exceed a 20% consistent with the input voltage. 20% of nominal voltage when motors with decrease from nominal during motor the largest starting currents are started starting to assure at least the required during expected load condiiions. minimum voltage at connected motor de terminals.
- 3. The Reserve Auxiliary Transformer is sized 3.a inspection of load assignments will be 3.a Transformer nameplate ratings will not be to provide the full load requirements of performed to assure that transformer exceeded when supplying power to one one complete non-essentialload group aameplate ratings will rat be exceeded non-essential load group and three and a!! three Essential divisions without with all expected loads operating. Essential divisions.
exceeding its sir / oil rating. The RAf has two secondary windings and 3.b inspections end terts will i,a conducted to 3.b Transformer rauos provide'octput voltages will provide a nominal output voltage of confirm that transformer ratios provide that are consistent with int.ut voltages and G.9 kV */-10% with the nominalinput output voltages on both win < lings that era output voltages do not decrease below voltage provided. Output voltags will not consistent with the input voltage. 20% of nominal voltage when motors with exceed 5 20% decrease from nominal the largest starting currents are started during motor starting to assure at least the during expected load conditions. required minimum voltage at the connected motor terminals. O O O
O O ' g' Table 2.12.1: Electrical Power Distribution System IContinued)
~'
Inspections, Tests, Analyses and Acceptance Criteria s Certifid Design Commitment inspections, Tests, Analyses Acceptance Criteria *
- 4. Breakers are capable of Mterrupting the 4.a inspection of the connecte:! toad 4.a Transformers, switchgear, motor control maximum fault co rent to which they may requirements and breal.er wrdination centers, phase buses, and cables are be subjected. Switchgear, Motor Control schem3 will be performed m confirm the capable of sustaining the maximo.'n fault Centers, Isolated and Non-segregated selection of the electrical power currents to which they may be subjected [
Phase Buses, and cablet, are selected to distribution system components and until the fault is cleared. Circuit breakers sustain the maximum fault currents to cables and their capability to limit and clear are rated to interrupt the maximum fault which they may be subjected until the fault fauits. currents to which they may be subjected. k is cleared. l Transformers are sized to limit maximum f ault currents while maintaining required voltage regulation. Ccbles are sized and insulation selected to 4.b Inspection of the distribution system cable 4.b Cable selection is consistent with the cable 6 accommodate the load requirements, type selection criteria will be performed to selection i.riteria and will perform their of service, and env9onmental conditions to assure that sizing and insulation selection intended service. which they may be subjected. of catJes is consistent with the load and environment to which they may be subjected. Switchgear and motor control center 4.c inspection of power distributior: system 4.c Power distribution system protective protection devices and breaker control protective devices and control power devices and control power sources are power is provided from the 125VDC battery supplies will be performed. either internal to the switchgear or from of the same division or load group or the the 12S VDC battery of the same division or power is internal to t!'e switchgear. The non essentialload group.The main main generator breaker control power is generator breaker control power is provided from two separate, on-site, non- supplied from two separate or.-sit *, ,- l essential 125VDC batteries. essential 125 VDC batteries. Redundant overcurrent devices are 4.d inspection of the redundant overcurrent 4.d Redundant overcurrent devices are provided for cables entering primary devices on cables penetrating the primary provided, when required, on all electrical containment through penetrations, when containment will be performed. cables penetrating the primary , g re. quired containment.
~'
8
[ Table 2.12.1: Electrical Power Distribution System (Continued) P ~ Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitmerit inspections, Tests, Analyses Acceptance Criteria
- 5. Electrical independence is maintained 5.a inspections of the Essential Safety Systems 5.a There are no bus ties betweer Essential between Essential Safety Systems. will be performed to confirm their Safety Systems.
independence Electrical independence is maintained 5.t> Inspections of the non-essential load 5.b There are no automatic bus 'ies between betvecen non-essential load groups. groups will be performed to confirm their non-essential load groups. independence. Electrical indepe ,dence is maintained 5.c Inspection of the configuration and 5.c The configuration and protection between Essential Safety Systems .ind orotection scheme employed for the two employed on the essential and nca-non-essentialload groups. The one pcwer sources providing power to the essential feeders to the FMCRD Power exception is the two power supplies to the FMCRD Power Center will be performed. Center provide the required electrical Fine Motion Control Rod Drive (FMCRD) inc' pendence and separation. Powe. Centers h 6. All switchgear, phase buses, and cables are 6.a inspections of switchgear, phase buses, 6.a Po.ier distribution system components identified according to Essential Safety power distribution and control cables will and cables are identified according tn Division, Divisional Association, Non- be performed to onfirm that they are division, association, load group, voltage essential load group, voltage level and, identified according to their Essential level, and function, when required, function; and are separated division, divisional association, non-accordingly, essentiat ioad group, voltage levels, and functions. 9 n 8 O O O
(. .( g Table 2.12.1: Electrical Power Distribution System (Continued)
~
nnspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria
- 6. (Continued) 6.b inspection of the locations, separation, and 6.b Essential power distribution system ,
ESS divisional equipment is Class 1E and is identification of Essential power components are tocated in Seismic located in Seismic Categcry 1 structures distribution system componer's and Category 1 structures and separated and divisional equipment areas which are raceways will be performed. divisional raceways and fire araas. physically separated by three hour fire Separation is provided between divisions, barriers. Divisional cables are Class 1E and voltage levels, and functions. are routed in Seismic Category 1 structures and dedicated raceways which are 6.c inspections will be performed to identify all 6.c Class 1E Associated circuits are identified ' separated s:Jch that tolerance is provided associated circuits and comply with Class 1E requirements 1 , for complete burnout of a single' fire area. Cables of different divisions are not routed through a common hostile area except where justified. Non-essential cables,if , routad with Essential cables are Class 1F 3 c Associated. Non-essentialload group equipment is i non-esser.tial and separated by distance. The Normal Preferred Power and Alternate P.eferred Power feeders are routed through different areas of the Turbine and Reactor Buildings and by distance when crossing the Control Building. The three Normal Preferred Power feeders 6.d Inspections of the separation provided for 6.d Separation is provided between the are separated by routing in separate the Normal Preferred Power, Alternate Normal Preferred Power feeders and the raceways. Preferred Power, and Combustion Turbine Alternate Preferred Power and CTG feeders Generator feeders will be performed. and between the Normal Preferred Power CTG feeders follow a similar routing feeders of the three load groups. scheme as that used for the Alternate - Preferred Power feeders to separate them from the Normal i' referred Power. m c o r t i
g Tabla 2.12.1: Electrical Power Distribution Syst:m (Continued) N Inse sctions, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspectinns. Tests, Analyses Acceptance Critaria
- 6. (Continued) 6.e Inspection of the separat;on between the 6.e A minimum 50-ft separation is provided UATs are separatW Nm each other and Main Power Transformer, the Unit Auxiliary between the Main Power and Reserve from the Main Povvy - ansformer by Transformers, and the Reserve Auxiliary Auxiliary Transformers, and between the shadow fire walls and irom the RAT by a Transformer will be performed. Unit Auxiliary Transformers and the RATS.
minimum of 50 feet. The Main Power The Main Power and RAT transmission ' Transform RAT and its feeder by a lines are separated by a minimum of 50 minimum of 50 feet, feet. Shadow fire walls separate the UATs from each other and from the Main Power Essential division and non essentialload Transformer. group raceways are separated according to voltage levels and functions, when required. Medium voltage power cables are not routed in the same raceway ac
. control cat.les.
- 7. The electrical groundino system is 7.a inspection and tests will be performed on 7.a Grounding networks and lightning comprised of separate grounWng and the grounding networks and lightning protection systems are insulated from each lightning protection networks.These protection system to confirm that they are other and connected to the plant networks are lnstrument, equipment, insulated from each other and low grounding grid out4de struc*ures.
lightning protection, and a plant grounding resistance grour"Jed and that all Equipment requiring grounding is grid. The instrument equipment, and equipment requiring grounding are identified and low resistance grounded lightning protection networks are insulated identified. except for the Main Generator, the from each other and separately connected Emergency Diese! Gene rato.s, and the to the plant grounding grid outside the Combustion Turbine Generator, which are structures. high resistance grounded. All electrical and mechanical components requiring grounding are identified and low resistance grounded to the appropriate ground;ng network. The Main Gent.rator, Emergency Diesel Generators, and Combustion Turbine Generator are high
$ resistance grc inded 'o maximize :
@ availability. 1 Equipment located outside structures are grounded locally and pr<wided with lightning protection, when required.
O O O
O J f g Tat 'e 2.12.1: Electrical Power Distribution System (Continued)
~'
Inspections, Tests, Anclyses and Acceptance Criteria Cs rtified Cesign Comrnitment Inspections, Tests, Analyses Acceptance Criteria
- 7. (Continued) 7.b inspection and test of the Safety Grouna 7.b Safety Ground Circuit Breakers are Medium voltage switchgear are provided Circuit Breakers protection scheme will be interlocked with the bus voltage and bus with a Safety Ground Circuit Breaker which performed. input feeder breaker positions to prevent -
is interlocked with the bus voltage and the inadvertent clost re. Annuncietion is bus input feeder breakers and is provided ia the main control room when a annunciated in the coqtrol room when in breaker is in the racked-in posilier tSe racked in position
- 8. Power distribution syr. tem remote contros 8. Inspections of the controls and information 8. Necessary controls and information are parameter informati;n, and annunciators provided to the Main Control Room and provided in the Main Control Roora for safe are provided :n the Main Control Room and Remote Shutdown System will be operation and Safety Shutdown of the to the Remote Shutdown System for performed to assure plant control and plant, required plant operation and safety information needs are provided for plant shutdown of the plant. operation and safe shutdown.
L Y 9. All Bus transfer operations are manual 9. Testing will be performed to confirm that 9. Bus transfers are automatic for Safety only, except for automatic bus transfer on all bus transfers are manual only, except System transfers to the Emergency Diesel the div:sional buses form their normal for the specified automatic bus trarsfers on Generators, PIP bus transfers to the ' power supplies to their respective the Emergency, Plant investment Combustion Turbine Generator, FMCRD i Emergency Diesel Generators, automatic Protection, and FMCRD Power Center Pnwer Center transfer to the non-essential bus transfer on tne Plant Investment switchgear when bus low voltage occurs. power source. Bus transfers occur on bus Prntection buses from their normal power low v.,ltage. All other bus transfers are by supplies to the Combustion Turbine rnanual operation only, j Generator, and the automatic bus transfer from tha Essential divisional bus to the Plant Investment Protection bus for the Fine Motion Control Rod Drive Power Center. All aut(,matic bus transfers are dead bus transfers and are initiated on bus low voltage.
- 10. Essential Class 1E valve motors fed from 10. Testing will be performed to assure Class 10. Class 1E valve motor thermal overloads are i
, the Motor Control Centers are provided IE vcive motor thermal overloads will be bypassed on receiving a LOCA signal and a with thermal overload protection which is bypassed when a LOCA signal is received are operable under all other conditions 3 bypassed during a Loss of Coolant and are operable under all other Accident (LOCA) on'y. Ti.e thermal conditions.
overload bypass is separa$ely testable.
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u DIV I DIV 11 OlV lit MCC LOADS ARE TYPICAL OF CNE OR MOHE Figure 2.12.1b Electrical Power Distribution System
ABWR oesign Document 2.12.2 Unit Auxiliary Transformer
,n No entry. C(,vered by item 2.12.1.
O 1 4 l-l 2.12.2 ,3 gm2 l
ABWR oesign Document 2.12.3 isolated Phase Bus k No entry. Covered by item 2.12.1. O l l l l l l l 1 0 2.12.3 -1 6/1/92 i
' no h n tiesign <> cument ABWR resign oorement __
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- ABWR Design Document ABWR oesiga cocament 2.12.8 Raceway System 4
) 2.12.9 Grounding Me
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ABWR Design Document 2.12.10 Electrical Wiring Penetrations Design Description Electrical wiring penetmtions are provided for Primaty Containment, anci Secondary Containment, and fire barriers. Primary Containment penetrations are leak tested f or mechanical integrity in accoidance with the leak test requirements of the primaty containment. Primary Containment Penetrations All electrical cables penetrating primary containment are provided with redundant overcurre nt devices (e.g. fuses) in series with the circuit breakers when the maximum fault current can exceed the continuous cmrent rating of the penetration. The redundant overcunent devices are provided as backup protection for fault currents in the penetration in the event of circuit breaker overcurrent or fault protection failure. Redur. dant overcurrent protection devices are located such that a failure of one device will not disable the other. When a both redundant overcurrent devices are active devices (e.g. circuit breakers), separate trip coil power supplies are provided. Primary containment penetrations are sepamted between divisions by 3 hour fire barciers (e.g. separate rooms and floors) outside the containment and by distance inside the incrted containment. Divisional and non-divisional penetration sepamtion is N maintained in the same manner as raceway sepamtion. Voltage groupings in penetrations is the same as that employed in raceways. Secondary Containment and Fire Barrier Penetrations Secondary containment electrical penetrations are provided for conduit and other raceways through secondary containment walls, floors, between fire areas, and for bottom entry through ure barriers into panels and switchgear. Integrity is maintained between fire areas by filling the penetmtion area around cables and around the raceway with a fire retardant foam. Electrical penetrations are curbed when penetrating floors and cable tray ri::ars are self draining to prevent water column bu;1 dup in the riser. Penetrations in radiation areas are otTset on each side of the barrier to prevent radiation streaming through the penetration. inspections, Tests, Analyses and Acceptance Criteria Table 2.12.10 provides a definition of the Inspections, Tests, and/or Analysis, together with the associated Acceptance Criteria which will be undertaken for the Electrical Wiring Penetrations. Table 2.14.1, Primary Containment System, provides a definition of the n) Inspections, Tesu, and/or Analysis, together with the associated Acceptance 2.12.10 6/1/92
AB' N R ocsign Document Criteria which will be undertaken for the Electrical Wiring Penetration. tak g testing. W O O 2,12.10 -2 6/1/92
1 i y Table 2.12.10: Electrical Wiring Penetrations u Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. Electrical cabler oenetrating primary 1.a For each primary containment penetratico 1.a Analyses show that redundant fault current -
containment are provided with redundant requiring redundant fault current , protection devices will prevent fault . overcurrent prctection devices when the " protection, analyses of the fault clearing - currents from exceeding the continuous fault current car exceed the maximum time curves for the primary and seconda y current rating ei primary containment j continucus current rating of the . overcurrent interrupting devices plotted electrical oenetations. ; penetration.The overcurrent devices are against the thermal capability come of the > located such that a failuce of one of the penetration will be performed to assure ! fault current devices cannot prevent the that the coordination of the devices will ; r functioning of the redundant device.lf both provide the necessary penetration fault -l redundant devices require control power current protection. ! . for tripping, they will be provided from I separate contrcl power sources. In b inspection of the electrical documentation 1.b Redundant electrical penstratiun fault
.]
addition, primar y containment will be performed to assure that the failure current protection devices are i7stailed ;; b - penetrations are separated between of one redundant overcurrent devi::es will such that the failum of one device will not j divisions by 3 hour fire barriers outside not disable the function of the other and disable the .edundent device. Redundant }
- containment and by distance inside that the redundent overcurrent devices devices are provided control power for !
containment. Divisional and non-divisional (e.g. circuit breakers) are provided control tripping from separate sources when both .; penetrations are separated in the same power for tripping from separate sources. devices require tripping power. j manner as raceway separation. Voltage 1.c Inspection of the primary containment 1.c Primary containment penetretions are separation is n antamed consistent with penetration locations will be performed to separated between divisions outsida the I the vol* age levds identified for raceways. assure that electrical penetrations outside containment by 3 hour tire barriers and by . the containment are separated between distance inside containment. Divisional , divisions by 3 hour fire barriers and by and non-divisional penetrations are - ; distance inside C containment and that separated in the same mar ner a raceways. t divisionai and vion-divisional penetration separation is the same as that used for , raceways. 3 m m a mm. m. m - a - + __ _
" Table 2.12.10: Electrical Wiring Penetrations (Continued) e 5 Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 2. Electrical cable raceways penetrating 2. InspecNon of secondary containment and 2. Secondary containment and fire Larrier secondary containment or fire barrier walls fire barrier electrical penetrations will be electrical penetrations seats are provided and floors are provided with penetraticn performed to assure they are installed in and radiation streaming between areas is seals. Penetration integrity is maintained accordance with design installati3a prevented.
by filling the area around the cables and specifications and prevent radiation around the raceway with a fire retardant streaming through the penetration in foam. Raceway floor risers are curbed and radiation areas. self draining to prevent a water column buildup in the raceway. Electrical penetrations in radiation aress are offset on each side of the barrier to prevent radiation streaming. b 5" c 8 O O O
ABWR L.;gn occument g 2.12.11 Combustion Turbine Generator V Design Descn.p tion The Combustion Turbine Genentor (CTG) is a non-essential standby power source located on-site within the turbine island. The turbine generator unit is sized to proside standby electrical power to any two of the non-essential plant investment protection (FIP) buses or one PIP bus and one Essential Safety System (Division) bus and their associated loads at a nominal voltage of 6.9 kV and 60 cycles during loss of off-site power to the bus. The CTG is not required for safe shutdown or maintenance of safe shutdma of the plant under any condition. Transfer to the CTG power supply is automatic for either or both of a preselected pair of PIP buses on loss of bus voltage. Transfer of the CTG power supply to any one of the divisional safety buses is manual and only perfonned after assuring that the safety-relt ed power source has failed and no more than one PIP bus is being powered by the CTG. The CTG is provided with an output disconnert switch for maintenance and feeds a stub bus where indhidual cables are connected to provide power to any of the three non-essential PIP buses or three essential divisional buses. In the unlikely event of multiple ,,ower source failures. this configuration aho provides, with appropriate controls, the capability of using the CTG feeder cables as a dg vehicle for connecting any powei . aurce to any load (Figure 2.12.11). The CTG unit is a skid mounted unit. It is equipped with its own auxiliary control and support systems (e.g., hydraulic start, excitation, lubrication, cooling, intakt and exhaust, control and protective systems, control panel, etc.). Fuel is provided from an external fuel storage tank similar to that provided for an emergency diesel generator. Fuel is the same type and quality as that used by the diesel generators. The CTG is designed to automatically start on a decrease of bus voltage to 70% of nominal, on either of the two preselected PIP buses, and be up to rated conditions and ready to load within a specific start time after recching a start signal. The CTG will automatically provide power to the preselected PIP buses I only. l CTG voltage and frequency regulation is the same as that provided by the non-essential 6.9 kV power disuii>ution system. Sudden applications of large loads l will not result in a voltage decrease from nomimd voltage greater than 25E l Analysis to detennination the need, if any, for load sequmcing will be l performed during the implementation stage of the design. Controls, instrumentation, and alarms are provided in the control room to manually control and monitor the performance of the CTG. 2.12.11 6/1/92 l
ABWR Design Document The GTG is high resistance grounded to maximize availability. Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.11 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria that will be undertaken for the CTG. O 2,12.11 6/1/92
O O 3 Table 2.12.11: Combustion Turbine Generator l Inspections, Tests, Analyses and Acceptance Criteria l Certified Design Commitment inspections, Tests, Analyses Acceptar.ce Criteria f
- 1. The Combustien Turbine Generator is 1.a Inspection will be performed to confirm 1.a The combined maximum operating load of capable of supplying fullload power to any that the maximum expected combined the two heaviest loaded buses do not
( two Plant investment P.otection (PIP) loads on the two heaviest loaded buses are exceed the rated power output (according buses or any one plant investment within theioar' rating of the combustion to the r.ameplate rating) of the combustion protection bus aM any one essential safety turbine generator. generator. I but (Figure 2.12.11). 1.b Testing will be conducted by sychronizing 1.b The unit produces rated output at rated j the combustion generator to the off-site voltage and frequency for a minimum of 24 system and increasing its output to its full hours. smomentary transients excepted). I load condition. l i 2. Sudden a9plications of large loads will not 2. Testing will be conducted by suddan 2. The sudden application of the largest load i result in more than a 25% voltage decrease application of the fargest load block. block to the unit does not cause a voltage l
@ from nominal voitage. decrease in excess of 25% from nominal
( voltage. t , 3. Controls, instrumentation, and alarms are 3. inspection of instrumentation and testing 3. The unit can be controlled, loaded 7d provided in the control room to operate will be conducted by operation of the monitored from the main control room. ' and monitor performance of the Combustion Turbine Generator from the combustion generator. main control room. i i is ; i i
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r_ _ ABWR Design occument ._ 2.12,12 Direct Current Power Supply Design Description The plant Direct CmTent (DC) Power Supply System consists of safety-related 125 VDC and non-safety-related 125 VDC and 250 VDC powcr supply systems. 3 The system batteries are comprised ofindustrial type stomge cells. (See Figures 2.12.12a, b, and c.) Safety-Related DC Power The safety-related DC power distribution system consists of four Class 1E, ( ' ~ separated and electricallyindependent, dhisional distribution systems located in Seismic Category I structures. Each division contains its own 125 VDC batterj, associated central diatribution panel, motor contol center (if needed for larger loads), and distribution panek local to the supplied loads. Each batteg is separately housed in its dhisional area and in a ventilated room apart from its chargers and distribution equipment. Each battery is selected such that its $ warranted capacity will provide 100 percent ofits design loads at the end-of-installed-life with a minimum allowable voltage of 105 VDC. The batteries in safety divisions 2,3, and 4 are sized to supply all required loads for a minimum of 2 hours without recharging and the battery in dhision 1 is sized to supply required loads (including RCIC loads) for 8 hours of coping dunng station blackout. The dhision 1,2, and 3 batteries are each provided with a normal batten charger supplied from a motor control center (MCC) in the same dhision. The dhision 4 batten charger is supplird from the dhision 1 MCC. In addition, a standby battery charger is shared betricen dhisions 1 and 2 and a second standby charger is shared between dhisions 3 and 4. The battery charger circuit breakers are interlocked such that paralleling between dhisions, either at ; the AC supply inputs or DC outputs, is prevented, Each batteg charger is a self ; load limiting batten replacement type and is sized to supply tlw normal steady - state loads while restoring the batten to a full cnarged state at a maximum j charging voltage of 140 VDC. J Non Safety blated DC Power The non-sw, related DC power distribution system consists of three non-divisional 125 '/DC distribution systems (one per load group) and one 250 VDC distribution system. Each 125 VDC system contains it., wa bauery, associated central distribution panel, and distribution panels local to the supplied loads. Each battery is separately housed in a ventilated room apart from its chargers and distribution equipment. Each batten is selected such that its warranted capacity will provide 100 percent of its design loads at the end-of-installed-life with a minimum allowable voltage of 105 VDC. The batteries in each load group are sized to supply all required loads for a minimum of 2 hours without recharging. Each battery is prmided with a normal battery charger supplied 2.12.12 1- 6/1/92
ABWR Design Document from a motor control center (MCC) in the same load group. In addition, a standby battery charger is shared between all three load groups such that it can be powered from an MCC in any load group and feed any of the three 125 VDC h non-essential central distribution panels for load supply or batteiy charging. Each batteg charger is a selfload limiting battery replacement type and sired to supply the normal steady state loads while restoring the batteg to a full charged state at a maximum charging voltage of 140 VDC. The batteg charger circuit breakers are interlocked such that paralleling between any load group at the AC supply inputs or paralleling batteries is prevented. This batteg charger interlock mnfiguration, with additional interlocks on the central disuibution panel bus tie circuit breakers, prosides the ability for any batteg or batte:y charger to supply any central distribution panel while preventing the batteries from being paralleled. In addition to the 125 VDC non-essential power distribution systems, a single 250 VDC non essential power distribution system is provided to supply the plant computer systems and other non<ssential DC loads (e.g. turbine turning gear, lube oil pumps). The batteg is housed in a ventilated rc,om separate from its batten chargers and distribution panels. The battery is selected such that its warranted capacity will provide 100 percent of its design loads at the end-of-installed-life with a minimun allowable voltage of 210 VDC and sized to supply all required loads for a minimum of 2 hours without recharging. Two batteg g chargers are provided. The norrnal charger is powered from either of two W different non<ssentialload group Power Centers (P/C) through an interlocked, manual bus transfer device to prevent paralleling of the load group P/Cs. A smaller standby battery charger is powered from a control budding motor control center (MCC). The battery charger outputs are interlocked to preven: paralleling the chargers. Each battery charger is n self load limiting bat - ; replacement type. The normal battery charger i; sized to supply the normal steady state loads while restoring the battery to a full charged state at a maximum charging voltage of 280 VDC. The standby battery charger is sized to provide the normal lod, during battery and normal charger maintenance. The DC motor control centers, central distribution panels, and local distribution panels are identified according to their essentiality (e.g. essential division 1,2,3,4 or non-essentialload group A,B,C) and are located in their respective electrical equipment rooms or fire areas. Essential equipment rc oms and fire areas are separated by three hour fire barriers. MCCs and panels .Tre selected for their intended service and load requirements and are rated to sustain the maximum calculated fatut current under a'l modes of operation until the fault is cleared. Feeder and load circuit breakers are sized and rated to provide the load requirements under all expected operating modes and are capable of interrupting their maximum calculated fault currents. Switchgear and panels are grounded in accordance with the plant grounding specification. MCCs are provided with the manufactures recommended fault current and protective 2.12.12 6/1/92
ABWR oesign Document desices ar required by the fault current and breaker coordination analysis performed during thc implementation stage of the design. Control and instrumentation power for each switchgear is provided from the associated divisional or non-divisional battery. For circuits providing power through primary containment penetrations, a redundant overcurrent protecthe desice (generally a fuse) is provided in series with the circuit breaker if the calculated fault current could exceed the maximum continuous current rating of the penetration. Electrical power distribution parameters needed to assure plant reliability and safe shutdown (as detennined during the implementation stage of the design) are provided in the Main Control Room. Power disuibution ! system cables are selected for size and insulation based on their voltage, sersice load, routing, and environmental conditions (e.g. temperature, humidity, radiation) to which they may be exposed. Ratings and loading of the selected cables assures that they can sustain the maximum calculated fault currents to which they may be subjected until the fault is cleared. Cable impedance is considered in the overall distribution system protection analyses which will be performed during the implementation stage of the design. Selection and application of cables is intended to assure a life expectancy of 60 years. Cables are identified according to function (e.g. poveer, control), and essentiality (e.g. color coded) and are routed in the appropriate divisional or non.cssentialload group raceways. Inspecticans, Tests, A6.alyses and Acceptance Criteria l Table 2.12.12 provides a definition of the inspection, test, and/or analysis together with the associated acceptance criteria which will be undertaken for the direct current (DC) power supply batteries and battery charging systems. Table 2.12.1, Electrical Power Distribution System, provides a definition of the inspection, test, and/or analysis together with the associated acceptance criteria which will be undertaken for the direct current (DC) power supply distribution systems (e.g. raceways, cable and other equipment identification, grounding). l l i (") 2.12.12 3- 6/1/92
Table 2.12.12: Direct Current (DC) Power Supply a laspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The safety.related DC Power Distribution 1.a inspections of the safety-related 1.a Each divisional battery is separately System consists of four Class 1E, separated distribution system arrangement will be housed in a ventilated area apart from its and electrically independent divisions performed to confirm each divisional distribution equipment and all equipment located in Seismic Category I structures. battery is separately housed in a ventilated is located in their respective divisional Each division contains its own 125 VDC area apart from its associated distribution areas in Seismic Category I structures.
industrial type storage battery, battery equipment and all equipment is located in charger, central distribution panel, motor its associated divisional areas in Seismic co.itrol center (when needed for large Category i structures. loads), and distribution panels local to the supplied loads. Each battery is separately 1.b Testing will be performed to verify that 1.b Testing confirms that each battery capacity housed in its respective divisional area and each battery capacity is sufficient to satisfy is sufficient to supply its safety load in a ventilated room apart from its battery the safety load demand profile under demand. chargers and distribution equ + ment. conditions of LOCA and loss of peferred power. Each battery is selected such that its warranted capacity wili provide 100 percent of its design loads at the end-of-installed-life with a minimum allowable voltage of 105 VDC. The Division 2,3, and 4 batteries are sized to supply all required loads for a minimum of 2 hours without recharging.The Division 1 battery Is sized to supply required loads for 8 hours of coping daring station blackout. a u O O O
im n /m (V ( ) i - U g Table 2.12.12: Direct Current (DC) Power Supply (Continued) e G Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria Each battery is provided with a divisional 1.c A load capacity analyses will be performed 1.c The load capacity analyses confirms that normal battery charger and a shared showing each battery terminal voltage and each battery supplies the design loads at or .
- standby battery charger. The chargers are ' worst case DC load terminal voltage at above the required minimum voltage and suppiied from a motor control center each step of the battery loading profile to is consistent with the manufacturer's (MCC) in i.hc same division as the battery, assure that the battery will provide a ampere-hour ratings for the battery at a 2 except for the division 4 battery chargars minimum 105 VDC for the duratic,n of the hour and 8 hour rate..
which are supplied from a division 1 MCC, profile. The standby battery chargers are shared. Divisions 1 and 2 share one standby 1.d Inspections of the normal and standby 1.d Battery charger nameplate ratings confirm , charger and divisions 3 and 4 share a battery charger ratings (as identified by their capacity to supply normal steady second standby charger. Battery charger their nameplates) will be performed to state loads and recharge the connected circuit breakers are interlocked such that confirm their capacity to supply normal battery at a maximum voltage of 140 VDC. paralleling between divisions, either at the steady loads and recharge their respective AC input or DC output, is prevented. Each battery at a maximum voltage of 140 VDC. battery charger is a self load limiting 1.e Tests will be performed to confirm that the 1.e Tests confirm thet divisions cannot be battery replacement type and is sized to battery charger interlocks will prevent paralleled, either AC and DC, through the supply normal steady state loads while paralleling AC or DC divisions. battery chargers. restoring the battery to a full charge state at a maximum charging voltage of 140 VDC. (See Figure 2.12.12a.) ,
- 2. The non-safety-related DC Power 2.a inspections of the non-safety-related 2.a Each of the batteries is separately housed Distribution System. consists of three non- distribution system arrangement will be in a ventilated area apart from its divisional 125 VDC (one per load group) performed to confirm each battery is distribution equipment.
systems and one 250 VDC distribution separately housed in a ventilated area systems. Each 125 VDC system contains its apart from its associated distribution own industrial type storage battery, central equipment. distribution panel, motor control center (when needed for large loads), and 3 distribution panels local to the supplied h loads. Each battery is separately housed in its respective ventilated room apart from its battery chargers and distribution equipment. ,
Table 2.12.12: Direct Current (DC) Power Supply (Continued) [ M U Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criterie Certified Design Commitment 2.b A load capacity analyses will be performed 2.b The load capacity analyses confirms that Each battery is selected such that its each battery supplies the design loads at or warranted capacity will provide 100 showing each battery terminal voltage and worst case DC load terminal voltage at above the required minimum voltage and percent of its design loads at the end-of- is consistent with the manufacturer's nach step of the battery loading profile to installed-life with a minimum allowable assure that the battery will provide a ampere-hour ratings for the battery at a 2 voltage of 105 VDC.The batteries are sized hour rate. to supply all required loads for a minimum minimum 105 VDC for the duration of the of 2 hours without recharging. profile. 2.c inspections of the normal and standby 2.c Battery charger nameplate ratings confirm Each battery is provided with a normal their capacity to supply normal steady battery charger supplied from a motor battery charger ratings (as idcntified by state loads and recharge the connected their nameplates) will be performed to control center (MCC) in the same load confirm their capacity to supply normal battery at a maximum voltage of 140 VDC. group. A standby battery charger is shared v:ith all three load groups.The standby steady loads and recharge their respective battery charger can be powered from an battery at a maximum voltage of 140 VDC. p MCC in any one of the three load groups and supply any of the three non-essential 2.d Tests will be performed to confirm that the 2.d Tests confirm that AC load groups or DC central distribution panels for oad supply T batteries cannot be paralleled. battery charger interlocks will prevent or battery charging. Battery charger and paralleling AC load groups ct DC batteries. central distribution panel feeder and bus tie circuit breakers are interlocked such that paralleling load groups at the AC supply inputs or paralleling batteries is prevented. Each battery charger is a self load limiting battery replacement type and is sized to supply normal steady state foads while restoring the battery to a full charge state at a maximum charging voltage of $40 VDC. (See Figure 2.12.12b.) o n 8 O O O
+ .
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" ' Table 2.12.12: Direct Current (DC) Power Supply (Continued)
. .~ .
U : Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 3. The non-safety.related 250 VDC Power - 3.a inspections of the non-sa'ety-related : 3.s The battery is separately housed in a
- Distribution System consists of one 250 distribution system arrangement will be ventilated area apart from its distribution VDC industr!al type storaca batterv. central performed to confirm the battery is equipment.
distribution panel, motor control center, separately housed in a ventLted area and distribution panels local to the apart from its associated distribution supplied loads. The battery is separately ' equipment. housed in a ventilated room apart from its battery chargers and distribution . ! equipment. The battery is selected such that its 3.b A load capacity analyses will be performed 3.b The load capac*ty analyses confirms that warranted capacity will provide 100 showing banery terminal voltage and the battery supplies the design loads at or. percent of its design loads at the end.of. worst case DC load terminal voltage at above the required minimum voltage and installed-life with a minimum allowable each step of the battery loading profile to is consistent with the manufacturer's 9 voltage of 210 VDC.The battery is sized to assure that the battery will provide a ampere-hour ratings for the battery at a 2 , supply ali required loads for a minimum of minimum 210 VDC for the duration of the hour rate. j 2 hours without recharging. profile. The battery is provided with a normal . 3.c Inspections of the normal and standby 3.c Battery charger nameplate ratings confirm l battery charger ratings (as identified by their capacity to supply normal steady - battery charger supplied l rom two different non-essential load group Power Centers (P/ their nameplates) will be performed to state loads and the normal charger's C) through an interlocked bus transfer confirm their capacity to supply normal capacity to recharge the battery at a device to prevent paralleling AC load steady loads and the normal charger's maximum voltage of 280 VDC while groups.The battery charger is a self load capacity to recharge the battery at a supplying loads. limiting battery replacement type and is maximum voltage of 280 VDC while i sized to supply normcl steady state loads supplying loads. while restoring the battery to a full charge 3.d Tests will be performed to confirrr Gthe 3.d Tests confirm that AC load groups or state at a maximum charging voltage of battery charger interfocks will pre wr " battery chargers cannot be paralleled. 280 VDC. A smaller standby battery paralleling AC load groups or battery charger, powered from a control building chargers. MCC, is also provided and sized to supply normal steady state loads during battery -
$ maintenance. The two battery charger
@ . outputs are interlocked to prevent paralfeling chargers.
2 (See Figure 2.12.12c.) t
ABWR 0: sign Document g 2.12.13 Emergency Diesel Generator System (Standby AC Power Supply) Y Design Description The Class 1E diesel generators comprising the Dhision 1,11, and 111 standby AC power supplies are designed to restore power to their respecthe Class 1E distribution system dhisions as required to achieve safe shutdown of the plant and/or to mitigate the consequences of a lossef-coolant accident (LOCA) in the event of a coincident loss of normal electrical power, Fach of the three divisions of the AC power system has its own diesel generator. The major loads consist of the follmeing systems for all thiee dhisions: Residual Heat Removal (RHR) System, Reactor Building Cooling Water (RCW) System, HVAC Emergency Cooling Water (HECW) System, and Reactor Senice Water (RSW) System. In addition, Dhisions 11 and 111 include the Higu Pressure Core Flooder (HPCF) System loads. (The Dhision 1 RCIC System is also part of the ECCS network, but is steam-driven and therefore does not present s Jnificant load to the diesel generator.) Each Class IE diesel generator, with its auxiliary systems (i.e, Fuel Oil Storage and Transfer System, Jacket Cooling Water System, Starting Air System, Lubrication System, and Combustion Air Intake and Exhaust System), supplies
/7 standby AC power to various Class 1E loads through the 6.9 kV and 480V systems, b The 480V system,in turn, supplies power to the UPS and battery charger for the dhision's 120 VAC and 125 VDC safety loads. (The low voltage portion does not significantly contribute to diesel generator loading, but is included with "other 480V loads" per Figure 2.12.13.) Each is physically and electrically isolated from the other dhisions. No automatic interconnection is provided between the Class ~~
1E dhisions. Each diesel-generator set is operated independently of the other sets, and is connected to the utility power system by manual control only during testing or for bus transfer. A failure of any component of one diesel generator set will notjeopardize the capability of either of the two remaining diesel generator sets to perform their functions. The diesel generators and their essential support equipment are classified Seismic Category 1, and are qualified for the environments where located. All components except for the fuel storage tanks and fuel transfer equipment are located within the Reactor Building. Each diesel generator unit is rated at 6.9 kV,60 Hz, and is capable of automatically starting, aculcrating, attaining rated frequency and voltage within 20 seconds, and supplying its loads in the sequence and timing specified in the plant design documents. In addition, each diesel generator is capable of starting, accelerating and running its largest motor at any time after the automatic
- loading sequence is completed, assuming that the motor had failed to start
() initially. Each diesel generator unit is also reliability tested by the manufacturer. 2.12.13 1 6/1/92
ABWR Design Document The diesel generators start automatically on loss of hus voltage. Under-voltage g sensors are used to start each diesel engine in the event of a sustained drop in W bus voltage be!ow 70% of the nominal 6.9 kV mting of the bus. Low-water level sensors and dipvell high-pressure sensors in each division are also used to initiate the respective diesel start under accident conditions. However, the diesels uill remain on standby (i.e., running at rated voltage and frequency, but unloaded) unless the bus under-voltage sensors trigger the need for bus transfer to the diesel supply. Manual start capability (without need of DC power) is also provided. Each dieselis supplied by its own independent fuel storage tank, which is located in an area protected from natural phenomena. This tank has a fuel capacity sufIicient to operate its diesel for a ;wriod of seven days while the diesel generator is supplying maximum post-1.OCA load demand. A day tank is also provided for each diesel, and is located in the Reactor Building. The day i nk has a fuel capacity sufficient for approximately 8 hours of full-load opemtions. Low-level sensors on the day tank actuate dual motor <lriven transfer pumps to replenish the day tank supply from the storage tank The standby AC power supplies are designed such that testing and inspection of equipment is possible during both normal and shutdewn plant conditions. Each standby AC power supply is composed of a three-phase synchronous generator and exciter, the diesel engine, the engine auxiliaries (inchiding the fuel tanks), and the control panels. Figure 2.12.13 shows the emergency diesel genemtor system interconnections between the offsite power supplies and the diesel generator standby AC power supplies for Divisions I, II, and III. l The transfer of each Class 1 E bus to its standby power supply is automatic, should i this become necessary, on loss ofits offsite power. After the circuit breaker ! connecting the bus to the preferred power supply is open, large motors are kept on the bus for parallel coastdown and optimal residual voltage decay, When the mitage decays to an acceptable level, major loads are tripped from the Chtss 1E bus, except for the Class IE 480V unit substation feeders. Then the diesel-l generator breaker is closed when the required genemtor voltage and frequency j are established. The large motor loads are later re-applied sequentially and automatically to the bus after closing of the diesel-generator breaker. Each diesel generator is capable of being started or stopped manually from the main control room. Start /stop control and bus transfer control may be transferred to a local control station in the diesel generator room. Control room indications are provided for system parameters. Each diesel generator, when operating other than in test mode, is independent of the preferred power supply. Additional interlocks to the LOCA and loss <>f-2.12.13 6/1/92
ABWR oesign occument power sensing circuits terminate parallel operation tests and cause the diesel O generator to ievert and reset to its automatic control system if either signal appears during a test. A lockout er maintenance nu>de removes the diesel generator from senice. The inopemble status is indicated in the control oom. Devices monitor the conditions of the diesel generators and effect action in accordance with one of the following categories: s1) conditions to trip the diesel engine even under 1.OCA; (2) conditions to trip the diesel engine except unde LOCA: (3) conditions to trip the generator breaker but not the diesel, and (4) ccnditions which are only annunciated. Inspections, Tests, Analyses and Acceptance Criteria _ Table 2.12.13 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the emergency diesel generators and their auxiliary systems. , o O O 2.12.13 6/1/92
Table 2.12.13: Emergency Diesel Generator System u Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The three diesel ganerator trains are 1. Tests and verification inspection will be 1. Plant tests and verification inspection for mechanically and electrically independent. conducted which will include independent physicallocation confirm proper and coincident operation of the three trains independence of three diesel generator -
to demonstrate complete divisional divisions. separation. i 2. All components essential to the operation 2. See Ganeric Equipment Qualification 2. See Generic Equipment Qualification of the diesel generators are Seismic verification activities (ITA). Act.eptance Criteria (AC). Category I and qualified for the appropriate environment for locations where installed.
- 3. The three diesel generators are capable of 3a. Confirmatory inspection will be oerformad 3a. The maximum loads expected to occur for i supplying sufficient AC pow'er to achieve to assure the maximum design loads each division (according to nameplate safe shutdown of the plant and/or to expected to occur for each division are ratings) sha!I not exceed 90% of the rated p mitigate the consequences of a LOCA in within the ratings of the corresponding power output of the diesel generator.
the event of a coincident loss of normal diesel generator. power (Figure 2.12.13.). 3b. Testing will be conducted by synchronizing 3b. Each of the three units shall produce rated ' each diesel generator to the plant offsite power output at 20.8 PF for a pedod of 224 power system and increasing its output hours (momentary transients excepted). power level to its fully rated load condition. Each unit will then experience full load rejection by tripping the toad and verifying the unit does not trip.
- 4. Each diesel generator is rated at 6.9 kV, 4. Perform a test of each diesel generator to 4. Each diesel penerator attains a voltage of three phase,60 Hz; and is capable of confirm its ability to attain rated frequency 6.9 kVi10%, and a frequency of 60 Hr12%
attaining rated frequency and voltage and voltage, within 2'J seconds after application of a within 20 seconds after receipt of a start stare signal. signal. E w O O O
p (. m, " Table 2.12.13: Emergency Diesel Generator System (Continued) ~ d Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. In the event of a loss of normal power, each 5. The automatic and manual start sequences 5. Each of the three units starts from each diesel generator unit is capable of starting will be tested for each diesel generator automatic and remote manual signal,then (both manually and automatically), unit. accelerates and properly sequences its accelerating, and supplying its loads in the loads. Each local manual signal also starts proper sequence and timing specified in the corresponding unit, but does not the plant design documents. it is also initiate load sequencing.The automatic capable of recovery following trip and load sequence begins at s20 seconds and restart of its largest load. ends $65 seconds. Following application of each load, the bus voltage will not drcp more than 25% measured at the bus.
Frequency shsil be restored to within 2% of - nominal, and voltage shall be restored to within 10% of nominal within 60% of each load-sequence time interval. In addition, P the unit's largest motor load shall be tripped and restarted after the unit has completed its sequence, and the bus voltage shall recover to 6.9 kVi10% at 6012% Hz within 10 seconds.
- 6. Each diesel generator unit is capable of 6. Each unit will be tested and the air receiver 6. Black-start capability is demonstrated manually starting without the need for - tank capacities shat! be analyzed to assure following one successful manual start, external electrical power. The air receiver its black-start capability is functional. acceleration, and bus energization for each tanks have sufficient capacity for five starts of the three units without assist from any without recharging. extemal electric power. Following black start, each unit's receiver tanks shall have sufficient air remaining for four more starts.
- 7. Interlocks to the LOCA and loss-of-power 7. Interlocks for the standby AC power 7. While in a parallel test mode, each unit will sensing circuits terminate parallel system will be tested. revert and reset to its automatic control operation tests and ccuse the diesel system following individual application of e generator to revert and reset to its a simulated LOCA signal and a simulated 5
automatic control system if either signal loss-of-power signal. appears during a test.
" Table 2.12.13: Emergency Diesel Generator System (Continued) d Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria
- 8. Devices monitor the conditions of the 8. Using simulated signals, protective 8. Successful circuit testing will be confirmed diesel generators, and effect action in interlocks and annunciations will be tested for the individual diesel generator accordance with one of the following to assure they perforra their functions, in protective sensors according to the categories: (1) conditions to trip the diesel accordance with the four categorical following:
engine even under LOCA,(2) conditions to conditions described. trip the diesel engine except under LOCA, Cpteaory 1 Ser$qrs; Annunciations and (3) conditions to trip the generator breaker diesel engin trip signalswill be confirmed but not the diesel, and (4) conditions which in combinre <ith a simulated LOCA are only annunciated. signal. Cateaory 2 Sensors: Annunciations and diesel engine trip signals will be confirmed without a LOCA, but trips will be bypassed when a simulated LOCA signalis present. 9 CDtegorv 3Jensors; Annunciations and generator circuit breaker trip signals will be confirmed. Categerv 4 San _.sprs; Annunciation signats will be confirmed.
- 9. Each diesel has its own 7-day fuel storage 9a. Visual inspection and calculation of 9a. Tank inspections and calculations confirm tank, and its own 8-hour capacity day tank capacities for each tank shall be perfort. -
. proper capacities of the storage and day ;
which is replenished by the storage tank. tanks.These shall be sufficient for full-load operation of each respective diesel generator for 7 days, and 8 hours, respectively. 9b. The fuel transfer system shall be tested. 9b. Transfer system operation for each division will be confirmed by actuating both pumps from the day tank level sensors and e observing proper flow into the day tanks. c: 8 9 O O
O
" Table 2.12.13: Emergency Diesel Generator System (Continued)
~.
4 O Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Accepts:nce Criteria certified Design Commitment l
- 10. The manufacturer's test documents shall 10. Visualinspection of manufacturer's test
- 10. The manufacturer has conducted l be, visually inspected. documents confirms the required reliability i reliability testing on the units. testing has been performed, and that the diesel generator has passed the test requirements.
- 11. Control indications are provided for D/G 11. Inspections wi:t be performed to verify 11. The designated instrumentition is present presence of control room indication for the in the control room.
system parameters. D/G systr,m. i l l I' i l I l l m i u
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Figure 2.12.13 Emergency Diesel Generator System interconnections 2.12.13 8- 6/1/C2
ABWR oesign Document 2.12.14 Reactor Protection System Alternate Current Power Supply ( $ A._,/ Not an ABWR system. No entry, t' O o/ y 2.12.14 6/1/92
ABWR D: sign Document
. - . - 2.12.15 AC Power Supply And AC Instrument and Control Power Supply Systems
\
. - Design' Description Vital AC Power Supply System The Vital AC Power Supply System as shown in Figure 2.12.15a is comprised of a Class 1E safety-related system, a non-Class lE, non-safety-related system, and a non-Class IE, non-safety-related computer system. Each system provides power to those " vital" instrument and control circuits for which continuity of power is desimble.
The safety-related Vital AC Power Supply System provides uninterruptable, regulated 120VAC power to the four dhisions of the Class 1E Safety System Logic and Control (SSLC) System. Each of the four dhisions contains its own constant voltage constant frequency (CVCF) static inverter power supply. Normal Power to each CVCF is supplied from a 480VAC Motor-Control Center (MCC) in the same dhision, except for the Dhinon IV CVCF, which is supplied power from the Dhision I MCC. Backup Power for each CVCF is supplied from the 125VDC battery of the same dhision. Each CVCF output is provided to distribution panels
- local to the circuits powered. Dhisional CVCFs and their respective distribution panels are electrically independent and physically separated between dhisions l
l= DU and are appropriately identified. The Class 1E Vital AC Power Supplies and their distribution panels are located in Seismic Category I structures. Dhisional CVCF power distribution is arranged such that the loss of a single CVCF power supply will not result in an inadvertent reactor shutdown. The non-safety-related Vital AC Power Supply system provides uninterruptable, regulated 120VAC power to the non-safety-related logic and control circuits
' important to the continuity of power plant operation. There is a CVCF static inverter power supply in each of the three non essential load groups. Normal Power to each CVCF is supplied from a 480VAC MCC in its associated load group. Each MCC recches power from the Plant hwestment Protection (PIP) bus in thb associated load group. Backup power to each CVCF is supplied from the non<ssential 125VDC battery of the same load group. CVCF output is provided to distribution panels local to the circuits powered. Each load group .
CVCF and its respective distribudon panels are electrically independent from the other load groups and are appropriately identified.
~ The non-safety-related Vital AC Computer Power Supply system provides-uninterruptable, regulated 120VAC power to the non-safety-related plant =
computers. This system contains two non-essential CVCF static inverter power supplies. Normal Power to each CVCF is supplied from a different load group
. :480VAC Power center (P/C). Each P/C receives power from the PIP bus in its associated load group. Backup power to both CVCF power supplies is from the non-essential 250VDC battery CVCF power output is provided to distributictn 2.12.15 6/1/92 '4 i ,.m., -- . v r r . , + - , , y
ABWR oesign Document panels local to the circuits powered. Each CVCF load group and its respective g distribution panels are electrically independent from the other load groups and W are appropriately identified. Each CVCF contains an alternate power supply for maintenance of the invener or to supply power in the event of inverter falhu e. The alternate pow supply is a voltage-regulating stepdown transf ormer, which receives power from the same 480VAC power source as the normal power supply. Each invener is synchroni..ed in both frequency and phase with its alternate power supply to avoid unacceptable voltage spikes during transfer from the inverter to the alternate supply. Automatic tmnsfer between the three CVCF power sources within a load group occurs as necessary to maintain a regulated output. Manual transfer between each CVCF power source is also provided. AC Instrument and Control Power Supply System The AC Insuument and Control Power System is shown in Figure 2.12.15b and is comprised of both a Class 1E safety-related system and a non-Class lE, non-safety-related system. Both systems provide 120VAC power to "non-vital" instrument and control power loads which can sustain a power in -:rruption during a loss of offsite power (I.OOP) event. The Class 1E safety-related AC Instrument and Control Power Supply system is comprised of a transformer and distribution panels in each of the three safety-related divisions. Each transformer is supplied power from a 480VAC MCC within its dhision and provides power to distribution panels local to the circuits powered. The transformers and distribution pa els within each division are electrically independent and physically separated from each other and are appropriately identified. The Class 1E power supply system components are located in Seismic Category I stmctures. The non-Class IE, non-safety-related AC Instrument and Control Power Supply system is comprised of a transformer and distribution panels local to the circuits powered. The transformer is supplied power from either of two 480VAC MCCs through a manual transfer switch. Each MCC is powered from a different non-essential load group PIP bus. Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.15 provides a dermition of the bspections, tests and/or analyses, ! together with associated acceptance criteria which will be undertaken for the Vital AC Power Supply System and Instrument and Control Power Supply System. O 2.12.15 6/1/92
7., 0,.m N.O 0 3 Table 2.12.15:- Vital AC Power Supply and AC instrument and Control Power Supply Systems . Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 1.- A Class 1E Vital AC Constant Volteye; 1. Inspections will be performed to confirm 1. Each of the four divisional Class 1E CVCFs Constant Frequency (CVCF) Powerjupply that the four Class 1E CVCFs and their and associated distribution panels are and associated distribution panels are s associated distribution pane!s are located located in Seismic Category I structures, provided in each of the four Instrument in Seismic Category I structures, identified, identified, and electrically independent and and Control Safety Divisions.The CVCFs and that each division is electrically physically separated. and associated distribution panels are independent and physically separated from located in Seismic Category I structures, the other divisions. identified, and electrically independent and physically separated from each other.
- 2. Each Class 1E CVCF receives power from 2. Inspections will be performed to confirm 2. Each Class 1E CVCF receives power from the MCC and 125VDC battery in the same that the AC and DC power sources for each the MCC and 125VDC battery in the same division, except Division IV, which is Class 1E CVCF is from its associated division, except the CVCF in Division IV supplied AC power from the same division division, except the CVCF in Division IV which is supplied AC power from the same Y that provides the battery charger for the which is supplied AC power from the same division that provides the battery charger Division IV battery. division that provides the battery charger for the Division IV battery.
for the Division IV battery.
- 3. Each Class 1E CVCF inverter provides a 3. Inspections and tests will be conducted to 3. Each Class 1E CVCF provides the required 120VAC regulated voltage and frequency confirm the automatic transfer within the output regulation duririg norma! operation, output and its alternate power supply same division and output regulation of the automatic and manual transfer operations.
within the same division provides a Class 1E CVCFs. Manual transfer will be regulated voltage output.The CVCF tested. automatically transfers between power sources within the same division to maintain the required output. Manual transfer is also provided
- 4. A non-Class 1E Vitai AC Constant Voltage 4. Inspections will be performed to confirm 4. Eac! of the three non-Class 1E CVCFs and Constant Frequency (CVCF) Power Supply that the three non-Class 1E CVCFs and their associated distribution panels are and associated distribution panels are associated distribution panels e e identified and efectrically independent provided in each of the three non-essentist identified and are ele *trically independent from each other.
e load groups for instruments and Controls from each other.
$ important to the cGntinuity of power plant operation. The CVCFs and associated distribution panels are identified and electrically independent from each other.
~
4 2 c '3 4 Table 2.12.15: Vital AC Power Supply and AC Instrument and Control Power Supply Systems (Continued) P G Inspections, Tests, Analyses and Acceptance Criteria , Sertifisd Design Commitment . inspections, Tests, Analyses Accept &nce Criteria 9 3, 5.ach rion-Cass 1E CVCF receives power 5. Inspections will be performed to confirr.2 5. Each non-Class 1E CVCF receives power from the MCC and 125VDC battery in the that the AC and DC power sources for each from the MCC arsd 125VDC battery in the same fuad group. non-Ciass 1E CVCF is from its associated same non-essential loed group.. j
^'
non-essential load gcoup.
- 6. Each non Cass 1E CVCF inverter provides - 6. Inspections and tasts will be conducted to 6. Each non-Class 1E CVCF provides the ..
cenfirm the (tutomatic transfer and output . a 120VAC regulatad volicge and frequency required output regulatic n during norma! output and its alternate power supply regulation of the non-Cass 1E CVCFs. operation, automatic and manual transfer provides e regulated voltage oQtput. The Manual transfer will be tested. operations. CVCF automatically transfers between power sources to maintain the required ; e outpot. Manual transfer is also provided. 7 Two non Class 1E Vitaf AC Constant . 7. . Inspections wil! be performed to confirm 7. Each of the two non-Cass 1E computer l g Voltage Constant Frequency (CVCF) Power that the two non-Cass 1E computer CVCFs CVCFs eM Gssociated distribution panels Supplies and associated distribution and their associated clistribution panels are are identified and .3fectrically independent ; par **>ls are provided for the non-essential identified and 6re electricolly independent from each other. ; plant computers. The CVCFs and from eact; other. associated distribution panels are identif;ect and electrically independent from each other. ! S. Each non-Qar s 1E computer CVCF receives 8. Inspections will be performed to confirm 8. Each non.Oess 1E cc.mputer CVCF receives power from the P/C in t.Se same Ic3d group that the AC power sources for each non- power frem the PlCin The same nouf !
, and from the non-essential 250VDC Class 1E computer CVCF is from its essential load group and from the non- !
batterv. essociatnd non-essentist load proup and essential 250VDO battery. ; from the non-essential 250VDC b3ttery. }
- 9. Each nonA: lass 1E computer CVCF inverter .
- 9. Inspections and tests will be anducted to 9. Each non-Class 1E computer CVCF prUvides a 12DVAC regulated voltage and confirm the automatic transfer and output provides the regt. ired output fcguladon
, ' frequencf Outcut and its attemate pnwar regulation of the non-Cass 1E computer durir>g normal operation, autornatic and suppiy provides a regulated Voltage CVCFs. Manuct transfer will be tested. menud transfer operationss ! output. The CVCF automatically transfers
$ between ;xiwer sources to maintair: the ,
p 3 ' required output. Manual transfer is a:so ,
] provided. (See Figure 2.12.15a.) i i
9 9 9 3 a, . . . . ..; _. , , . --. .L..
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%f q) t Tabla 2.12.15: Vital AC Power Supply and AC Instrument and C mtrol Pover Supply Systems (Continued).
v E Inspections,Tes'ts, Analyses and Acceptance Critoria Certified Decign Commitment Inspections' Tests, Anaines Acceptance Criteria
- 10. Three Class 15120VAC Instrument and 10. Inspections will be performed to confiren 10. Each cf the three C: ass 1E instrument and Contro8 Power Se. Mies and associated that th3 three Class 1E Instroment and Cor. trol Power Suppliss and associated distribution paneis 3re provided for the ' Contro! Power supolics and their distribution panels are focated in Scismic "non-dtal" esser ist safety-r6:ated associated distribution panels are located Category 1 structures, ide ntified, and instrument and control circuits whhh can in Seismic Catagery I structures; identified, electrically independent and physically .
sustain a power interruption on loss on and are electricalyindepandent and scparated from each other. offsite power (LOOP). The instnament and physical?y Oparated from each other. Contro: Power Supplies and associated distribution paaels ars located in Seismic Category t structures, identified, and Sectrically independent and physically separ9tsd from each other.
- 11. Each Class 1E In strument and Control 11. Inspections w:ll be perforrned to confirm 11. Each Class 1E Anstrement ared Control p Power Supply receives power from the that the power sources for each Class 1E Power Supply receives power only from MCC ir* the same division. Instrument and Contiof Power Supply is the MCC 6 the same safety division and from the MCC of the same safety division the transformer ratio provides a Morr.irtal and that ths #ransfornMr ratio provides a 12CVAC output, nominal 120VAC output.
- 12. The non-Class 1E :20VAC instrument and 12. Inspectior*.s and test will be condur,tod to 12. The non-Class 1E 120VAC Instrument and Control Power Supply and associated confirm that the two powe* sources for the Cor trol Power Suppiy is powered from two distribution panels is provided for the non-C4s31E 12DVAC Instrumern und MCCs in different non-essential load
'non-vital", nonessentiglinstrument and Control Power Supply are frorn seca.ste groups and the manual transfers power control circuits which can sustain a powcr load groups and that the manual transfer between the two power sources.
interruption a LOOP. The Power Supply switch will transfer power between receives input power from either of 'wo s.oGrces. 480VAC non-essentLI MCCs through a manual transfer swirth. The MCCs are powered from Plant investment Protection (PIP) buses in sepa. ate icad groups. (See Figure 2.12.15b.) e
P'
" Table 2.12.15: Vital AC Power Supply and AC Instrument and Control Power Supply Systems (Continued) -
G Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses , Acceptance Criteria
- 13. Each Vital AC Power Supply and AC 13. Inspections wi:1 be performed to confirm 13. Each Vital AC Power Supply and AC instrument and Control Power Supply is that each Vital AC Power Supp!y and AC Instrument and Control Power Supply is '
sized to supply the full load requirements Instrument and Control Power Supply is si':ed (as determined by the nameplate of its connected loads. sized (as determined by the nameplate rating) to supply the full load requirements rating) to supply the full load requirements of its connected loads. of its connected !oads. i' a N O O O
m
'h CLASS 1E VITAL NON CLASS 1E VITAL' COMPUTER VITAL AC PCWER SUPPLY ' AC POWER SUPPLY AC POWER SUPPLY DIVISIONAL NON DIVISIONAL NON DIVISIONAL 125VDC 480VAC. 12SVDC 480VAC 250VDC - 480VAC DIST PANEL ,
MCC , DIST PANEL , MCC DIST PANEL P/C b b) b)- b) b) b) b). b) b)
? ? ? ? ? ? ? ? ?)
VOLTAGE VOLTAGE VOLTAGE CVCF REGULATING CW REGULATING CVCF REGULATING 9 RER ERER INVERTER TRANSFORMER TRANSFORMER TRANSFORMER r 6
- 6) 6) 6) e i
- 6) 6) 6) 6)
, f REACTOR CONTROL 7 f REACTOR CONTROL f f) fff COMPUTERS < BUILDING . BUILDING BUILDING BUILDL.G ; s __/ \ / \ / : 1YPICAL OF 4 ' TYPICAL OF 3 TYPICAL OF 2
- 1 PER DIVISION 1 PER NONESSENTIAL LOAD GROUP FROM DIFFERENT NONESSENTIAL g (DIV I, !!, al, IV) (LOAD GROUP A. B. C) LOAD GROUPS P/CS i D i
~
i i Figure 2.12.15a Vital AC Power Supply i
ABWR oesign occum:nt O SAFETY RELATED NON-SAFETY RELATED INSTRUMENT POWER INSTRUMENT POWER CLASS 1E NON CLASS 1E 2 30VAC MCC .BOVAC MCC 480VAC MCC 5 A) o 5) g)'NTentOCxe() _ _ _ _ . _ .) g m i ? 6 6) o o)- 6) o 6) 6) o o yed SB j " " " j TYPICAL OF 3 TYPICAL OF 1 1 PER DIVISION SUPPLY FROM DIFFERENT (DIV 1, li, Ill) LOAD GROUPS 9 Figure 2.12,15b AC Instrument and Control Power 2.12.15 8 6/1/92
ABWR D: sign Docum:nt (N 2.12.16 Instrumont and Control Power Supply
'sm ) No entg. Covered under Item 2.12.15.
l t i v i
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2.12.16 6/1/92 a
i ABWR oesign Document 2.12.18 1.lghting and Servicing Power Supply Systems
%)-
Design Description The plant lighting system is comprised of four independent lighting systems. They are the' Normal Lighting System, the Standby Lighting System, the Emergency Lighting System, and the Guide Lamp Lighting System. The Normal Lighting system is non-Class 1r The other three lighting systems are comprised , of both safety-related and non-safety-related subsystems. The Normal Lighting System is AC and non essential and prosides up to 50% of_ the lighting needed for operation, inspection, and repairs during normal plant
, operation and is installed throughout the plant in non-essential equipment areas, except for the passageways and stainvelh. Normal Lighting is generally supplied from the non-essential Power Generation (PG) buses. In the non-essential equipment areas, the Normal Lighting is supplemented (a minimum of -
50%) by the non-safety-related Standby Lighting System. Lighting from a single load group is acceptable for localized high intensity lighting and lighting in
. small rooms where only a limited number of fixtures are needed. Non -ssential senice outlets and internal lighting for non-essential panels is provided by the Normal Lighting system. In passageways and stairwells leading to non-essential y equipment areas, the lighting is supplied from two difTerent load groups of the
, non-safety-related standby Ligliting System. With this configuration, non-
^ '
- essential equipment areas recche 100% of their lighting from two different power sourcesc The non-safety-related AC Standby Lighting System is comprised of lighting from three non essential load groups. Each load group is supplied from a
- different Plant Investment Protection (PIP) bus which is connectable to the non-essential Standby Power Supply (Combustion Turbine Generator (CTG)). The
- non-safety-related Standby Lighting System supplies a minimum of 50% of the lighting needs of the non essential equipment areas and 100% of the iighting in passageways and stairwells leading to non-essential equipment areas (as -
described above);in addition, the non-safety-related Standby Lighting System n supplies up to 50% of the lighting needs m -essential equipment areas and in g passageways and stairwells leading to essential equipment areas. The remainder-of the lighting (a minimum of 50%) in the essential equipment areas and in passageways and stairwells leading to them is supplied from the safety-related Standby Lighting System. The non-safety-related Lighting in the essential equipment axeas and the passageways and stairwells leading to them is supplied from the same non<ssential load group as the essential load group (Safety Dhision) in the same area. , The safety-related Adandby Lighting System is comprised oflighting from three essential Safety fehisions. Each c t he three ssential dhisions is supplied l. I 2.12.18- -1 6/1 % !f 3,. -
ABWR Design Docwnent power from the Chss 1E dhisional bus, which is connectable to the Space Emergency Diesel benemiot (DQ of the essential Standhs Powei Supph m its g tespective dhision. Exh safett-related Standhv Lighting Sptem supplies a minimum of 50% of the lighting needs of the essential equipment atcas in its respective dhision and of the passageways and stainvells leading to its iespective equipment areas. The essentiallighting in the batten toom and other Instnnnent and Contiol meas of Dhision IV is supplied hom .be safety-ielated Standby Lighting System of the same division as other divisional equipment supplying the areas (e.g., hauen chargers). The Slain Control Room lighting is supplied from the same two dhisions of the safety-telated Standby Lighting System as the dhisions supphing the Main Contial Room Heating. Ventilation, and Air Conditioning (HVAQ. The remainder of the lighting (up to 500 in - the essential equipment aicas and the passagewavs and stainvells leading to them is supplied from the non-safety-iclated Standby Lighting System in the same load group as the safety-related Lighting System. With this configuration, essential equipment areas receive 100cl of their lighting needs from two different Standby Lighting power supplies. The above described AC lighting configuration permits retaining appioximatelv 50% of the lighting illumination in all passageways, stainvells and essential equipment areas dining lighting maintenanc- or loss of a load group. Illumination from 50% of the lighting is adequate to observe equipment and support personnel movement. (See Figure 2.12.18a) h The Emergency Lighting Systems provide DC powerc< ip lighting to prevent total blackout in areas which are occupied or mu accupied during periods when AC lighting is lost until the Normal or Standy, t.ighting Systems - are energized. The Emergency Lighting Systems, therebore. are not required to provide the same levels ofillumination as the normal standby systems. The non-safety-related Emergency Lighting System provides emergency lighting needs to the Radwaste lluilding control room (RWil), the Combustion Turbine Generator (CTG) area and control room, and the non-essential electrical equipment areas (both AC and DC) Lighting power for the RWil control room is supplied from the non-essential 250VDC battery. Lighting power for the non-essential electrical equipment rooms is supplied from the 125VDC battery in the same non-essentialload group as the equipment in the room. Lighting power for - the non-essential CTG is supplied from one of the non-essential ~125VDC batteries. The safety-related Emergency Lighting System provides the emergency lighting needs to the Main Control Room, the Remote Shutdown Panel room, the Emergency Diesel Generator areas and control rooms, and the essential electrical equipment rooms (both AC and DC). Lighting power for the identiDed essential areas is supplied from the 125VDC batten in the same 2.12 18 W1/92
ABWR oesign Document divisions as the area. The lighting power to the Main Contiol Room is supplied fiom tuo 12WDC batteries in the same division as the safetu elated Standbv Lighting souices for the control room. (See Figure 21214) Guide Lamps are piovided for stairways, exit routes, and major control areas such as the main control room, mdwaste control room and remote shutdown panel areas. The guide lamps are self contained, batten pack units, suitable for operation in the emironment of the areas in which they are locard. The units contain a rechargeable battery with a minimum 8-hour capacity and a battery charger supplied from the Standby Lighting System of the area in which they are located. Guide Lamps are Seismic Categc y 1 and are Class 1E when located in safety-related areas. All lighting systems are designed to prmide lighting intensities consistent with the lighting needs of the areas in which they are hcted. The lighting design considers the effects of glare and shadows on control panels, sideo display devices, and other equipment, and the mirror effects on glass and pools. I.ighting and other equipnient maintenance,in addition to the safety of personnel, plant equipment, and operation is considered in the design. Areas containing flammable materials M.g., battery rooms, fuel tanks, etc.) have j explosion-prooflighting systems. Areas subject to high moisture have water-q proof installations (e.g. dr>well, wash-down areas). Plant AC lighting systems are l V generally of the fluorescent type with mercury lamps provided for high ceiling l and yard lighting, except where breakage could introduce mercury into the l reactor coolant system. Incandescent lamps are used for DC lighting systems and above the reactor, fuel pools, and other areas where lamp hreakage could introduce mercury into the reactor coolant. Lighting systems and their distribution panels and cabies are identified
- according to their essentiality and type. Safety-related Llighting systems which are Class lE, are located in Seismic Category I structures, and are electrically independent and physically separated. Cables are routed in their respective i dhisional raceways. Normal Lighting is separated from Standby Lighting. DC l lighting cables are not routed with any other cables.
Plant Senice buses supply power and heavy duty senice outlets to equipment not generally used during normal plant power operation (e.g., turbine building and refueling floor cranes, welding eq'ignt). Senice outlets have grounded ! connections and the outlets in wet . oist areas are supplied from breakers ! with ground current detection. l l l l l 2.12.18 3 6/1/92
ABWR oesign occument inspections, Tests, Analyses and Acceptance Criteria Table 2.12.18 piovides ; definition of the inspc< tions, tests, and /oi analyses, togethei with associated acceptance ( riteiia which will be undertalen f oi the lighting and service power systenis.
~i
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]
9 2.12.18 4- G/1/92
(sv) y Table 2.12.18: Lighting and Service Power System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Accepsnce Criteria
- 1. The AClightingin non-essentialequipment 1. Inspections and tests will be conducted to i Two different AC lighting systems supply areas is supplied from two different confirm that two different AC lighting 100% of the lighting needs in the non-
, lighting power sources. AC Normal systems supply 100% of the lighting needs essential equipment areas and in the Lighting supplies up to 50% of the lighting in non-essential equipment areas and in passageways and stairwells leading to and nc i-safety-related AC Standby the passageways and stairwells leading to them At least 50% of the lighting is Lighting supplies the remainder of the them, and at least 50% of the lighting is supplied by a non-safety-related AC lighting needs (a minimum of 50%). The supplied from a Non-Safety- Related AC Standby Lighting System. Localized high lighting in passageways and saairwells to Standby Lighting System. intensity lighting and hghting in small non-essential equipment areas is supplied ' rooms is from a single source.
- from two non-safety-related Standby Lighting Systems from different non-ewential load groups. High intensity lighting and lighting in small rooms may be irom a single lighting system.
- 2. The AC lighting in essential equipment 2. Inspections and tests will be conducted to 2. Two different AC Standby Lighting j areas and the lighting in passageways and confirm that two different AC Standby Systems in the same load groep supply
] stairwells to essential equipment areas is Lighting Sy" ems in the same load group 100% of the lighting needs in the essential j supplied from two AC Stzndby Lighting supply 100% of the ligh+ing needs in equipment areas and in the passageways essential equipment areas and in the and stairwells lead:ng to them. At least Systems. AC safety-related Standby
- Lighting supplies a minimum of 50% of the passageways and stairwells leading to 50% of the lighting is suppliad by the l lighting and non-safety-related AC Standby them, and at least 50% of the lighting is safety-related AC Standby Lighting System Lighting supplies the remainder of the supplied from the safety-related AC in the same division as the essential lighting needs (up to 50%). Both the safety- Standby Lighting System in the same equipment area.
related and the non-safety-related Standby division as the essential equipment area. Lighting Systems are in the same , divisional or non-essential load group as the essential divisional area being supplied i lighting. p a 3
g Table 2.12.18: Lighting and Service Power System (Continued) y 4 Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 3. The three non-safety-related AC Standby 3. Insper+ ions will be performed to confirm 3. The three Non-Safety-Related AC Standby Lighting Systems are connectable to the that th three non-safety-related AC Lighting Systems can be supplied by the Combustion Turbine Generator (CTG) and Star.oby Lighting Systems are connectable Combustion Turbine Generator (CTG) and the three safety-related AC Standby Q the Combustion Turbine Generator that the three Safety-Related AC Standby Lighting Systems are connectable to their (CT G) ar:d that the three safety-related AC Lighting Systems can be supplied by their respective Emergency Diesel Generators Standby Lighting Systems are connectable respective DG.
(DG). Generally, the Normal Lighting to their respective DG. system is supplied from the non-essential Power Generation (PG) buses (see Figure 2.12.18al.
- 4. The non-safety-related DC Emergency 4. Inspections and tests will be conducted to 4 The non-essential 2SOVDC battery supplies Lighting system supplies lighting, at confirm that the non-essential 250VDC DC Emergency lighting to the Radwaste reduced illumination levels, to non- battery supplies DC Emergency Lighting to Building control room. The non-es cntial p '
essential areas which are occupied duri,c the Radwaste Building control room and 125VDC batteries st pply DC Emergency periods when AC lighting is lost. These Combustion Turbine Generator area and lighting to the non-essential AC and DC areas include the Radwaste Building (RWD; control room, and that the non-essential eM:trical equipment areas in their control room, the Combustion Turbine 125VDC batteries supply DC Emergency respective load groups. the Combustion Generator (CTG) area and control room, Lighting to the AC and DC non-essential Turbine Generator area and control room. and the non-essential AC and DC electrical electrical equipment areas in their Lighting is supplied from a non-essential equipment areas. The non-essential respective load groups. 125VDC battery. 250VDC battery supplies the DC lighting for the Radwaste Building and Combustion Turbine Generator. The lighting for the non-essential AC and DC electrical equipment areas is supplied from the non-essential 125VDC of the same load group as the equipment in the room. m a o O O O
p (m; '
'V (v} V " Table 2.12.18: Lighting and Service Power System (Conmnued) ~" Inspections, Tests, Analyses and Acceptance Criteria l
inspections, Tests, Analyses Acceptance Criteria l Certified Design Commitment l
- 5. Inspections and tests wilt be conducted to 5. An essential 125VDC battery supplies DC S. The safety-related DC Emergency Lighting Emergency Lighting to the Emergency confirm that the essential 125VDC battery system supplies lighting at reduced Diese! Generator area and controi room, illumination levels, to essential areas tupplies DC Emergency Lighting to the l Emergency Diesel Generator area and and the essential AC and DC electrical which are occupied during periods when equipment areas in the same safety AC lighting is lost. These areas include the control room, and the essentia! AC and DC electrical equipment areas in the same division.Two essential 125VDC batteries, in Main Control room. the Emergency Diesel the same divisions as the AC Standby Generator areas and control rooms, and safety division. Two essential 125VDC batteries,in the same division as the AC Lighting Systems, supply DC Emergency the essential AC and DC electrical Lighting to the Main Control Room.
equipment areas. Each essential 125VDC Stan-!by Lighting Systems, supply DC battery supplies the DC Emergency Emergency Lighting to the Main Control I Lighting for the Emergency Diesel Room. Generator area and control room, and the essential AC and DC electrical equipment area within its safety divi =;on. The Main
? Control room is supplieo DC Emergency Lighting from the two essential 125VDC batteries in the same division as the Safety-Related Standby Lighting source for the control room. (See Figure 2.12.18b.)
Inspections and tests will be conducted to 6. Guide Lamps are located in stairways. exit
- 6. Guide tsmps era provided for stairways. 6.
confirm that Guide Lamps are located in routes, and major control areas and exit routes, and major control areas. such contain 8-hour batteries, rechargeable as the Main Control rc,om and the stairways, exit routes, and major control areas and that they contain 8-hour from the AC Standby Lighting System in Radwaste Building Coatrol room.They are the same area. They are qualified Seismic batteries, rechargeable from the AC self contained onits with a minimum 8- Category I and are Class 1E in safety-hour battery pack and a battery charger Standby Lighting System in the same area. Seismic Category I and, when in safety- related areas. supplied from the AC Sindby Lighting System in the same area in which they are related areas. Class 1E status will als-> be located Guide Lamps are qualified Seismic confirmed. Cat igory I and are Class TE when located in a safety-related area. R
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8
~ Table 2.12.18: Lighting and Service Power System (Continued) 4 Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment
- 7. Inspection and tests will be conducted to 7. The lighting intensities are consistent with
- 7. A!! fighting systams are designed to the lighting needs of the area and intended provide the lighting intensities consistent confirm that lighting intensities are consistent with the lighting needs of the purpose of the lighting system. Seiacted with the lighting ne-da of the area and the area and intended purpose of the lighting lighting equipment as installed satisfiesthe intended purpose of the lightirig system. requirements of its intended application.
The effects of the fighting, such as glare system. Inspection of the selected lighting and shadows on equipment, and the mirror equipment and its instaIIation will be effacts on glass and pools, are considered performed to confirm that it satisfies the in the design. Lighting and other requ'rements of its intended application. equipment maintenance, in addition to environmental conditions (e.g., areas containing flammable materials. wet or moist areas, areas above the reactor and fuel pools) ar* considered in the selection and instsilation of lighting equipment. m
- 8. Inspections will be performed to confirm 8. Lighting equipment and cables are l
- 8. Lighting equipment, including distribution identified, electrically independent, and panels and cables, are identified according that lighting equipment and cables are identified, electrica!!y independent, and physically separated between safety to essentiality and type. Safety-related physically separated between safety divisions and between the Normal and l
lighting systems are C1 ass 1E, electrica;ly Standby Lighting Systems. Class TE i independent and physically separated, divisions and between the Normal and Standby Lighting Systems. The location of equipment and cables are located in and are located in Seismic Category i Class 1E equipment and cables in Seismic Seismic Category I structures and DC structures. Cables are routad in the cables are routed separate from ac cables. respective divisional raceways. Normal Category I structures cnd the separation Lighting is separated from Standby between AC and DC cables will afso be Lighting. DC lighting cables are not routed confirmed. with any other cables.
- 9. Inspections will be performed to confirm 9. Heavy duty service outlets are supplied
- 9. Heavy duty service outlets (e.g., welding from plant service buses and have outlets) are supplied from plant services that heavy duty service outlets are supplied from plant service buses and have grounded connection Outlets in wet or f ases and have grounded connections. moist ireas are supplied from breders Service outlets in wet or moist areas are grounded connections, and that outiets in wet or moist areas are supplied from with ground fault detection.
supplied from breakers with ground fault g> detection. breakers with ground fault protection. 8 O O O
!:0WN ii;WI Docuinent f.M D.u/c, baatmission l' ,1, e 'ieserse Auxiliary Transformer No c1111T. Co\ cled by itelin 2.l'.'.l .
O O 2.13 1 6/1/92
l ABWR onion 0:cumnt l l 2.14.1 Primary Containment System Design Description The primary containment system incorporates the dr>well and pressure suppression features of operating BWR plants and consists of a steel lined reinforced concrete containment structure fulfilling its design basis as a fission product barrier even at the increased pressur e associated with a main steam pipe rupture or a break in the largest fluid pipe. Main features include the upper and lower dnwell containment surrounding the reactor piessure vessel, a wetwell with a water filled supptession pool sening as a heat sink during nonnal, and abnormal operations and accidents. Refer to j Figure 2.14.1 Primary Containment System. l The primary containment volume is 259,563 cubic feet and is constructed as a right cylinder with an inside radius of 47' 7" and a reinforced concrete wall thickr. css of 6' 7" set on a reinforced concrete base mat 18' 0" thick and a wall height of 96 9" measured between the top of the base mat and the underside of
'the containment top slab. The primary containment top slab is 7' 2" thick except beneath the fuel pool, steam dryer / separator pool, and around the dqwell head opening where the slab is 7' 11" thick. The dnwell head opening is enclosed with a steel head removable for refueling operations. The dr>well design conditions O are 4!) psig and 340 degrees F. The wetwell design conditions are 45 psig and 219 degrees F.
The do,ull is comprised of two volumes: an upper dowell volume of 193,878 cubic feet, surrounding the reactor presstire vessel and housing the steam and feedwater piping, the safety /rclief valves, main steam drain piping and upper drywell coolers; and a lower dowell volume of 65,685 cubic feet, below the reactor pressure vessel housing the control rod drives, fine motion control rod drives, recirculation internal pumps, reactor cooling water heat exchangers, - neutron monitoring system, equipment platform, lower dowell coolers and two dryvell sumps.
'Ihe wetwell volume is 338,315 cubic feet, comprised of the suppression pool volume of 127,840 cubic feet at high water level or 126,427 cubic feet at low water level. The suppression chamber air space volume is 210,475 cubic feet at high 3
- water level and senes as the LOCA blowdown reservoir for the upper and lower
- drywell nitrogen and nonc@densables which pass through the ten 48 inch diameter drywell to wetweltvertical vents connecting with thirty 2.3 feet diameter,60 inch long horizontal vents with a total flow area of 125 square feet located at three levels below the suppression pool surface. The vent centerline submergence is 11.48 feet for the top row of horizontal vents,15.98 feet for the '
center row of horizontal vents and 20.48 feet for the bottom row of horizontal vents. The suppression pool water serves as the heat sink to condense steam _ 2.14.1 - 1- 6/1/92
ABWR oesion Document released into the dnwell during a LOCA, or steam from safetv ielief valve acthity or exhaust steam from reactor core injection coolant steam turhine operation. The 3' 11" thick dnwell diaphragm floor has steelliners on both top and bottom sides to minimize bypass leakage. Access to the upper dnwc!! is prosided through a double scaled personnellock and an equipment batch. The lower drwell access is prosided through a personnel access tunnel with a double scaled p-t'onnellock at the reactor building end. An equipment transfer tunnelis scaled by an equipment batch removable only during refueling or maintenance outages. These access tunnels span the suppression chamber. Access to the suppression chamber is prosided by a hatch located in the rcactor building. Prior to reactor operation the atmospheric control system in conjunction with the HVAC priman containment purge system and the dnwell cooling fans establish an inert gas ensironment in the primary (ontainment with nitrogen to limit the oxygen concentration to preclude combustion of hydrogen released during a LOCA. The primary containment leak rate mun he less than 0.5% per day based on a 39 psig post accident pressure. The primary containment design pressure is 45 psig. Additional nitrogen gas is added to the dnwell and the wetwell to maintain a positive prcssure and prevent air inleakage. High pressure nitrogen is also used for pneumatic controls inside the primary containment to avoid adding air to the inert atmosphere. g Refer to Section 2.14.2 for data on the Containment Internal Structures and Section 2.14.3 for data on the Reactor Pressure Vessel Pedestal. Design Bases The pressure appression containment structure has the capability to maintam its functionalintegrity during and following the peak transient pressures and temperatures caused by the worst 1 OCA pipe break postulated to occur simultaneously with loss of offsite power and a safe shutdown carthquake. The containment structure is designed to accommodate the full range ofloading conditions associated with normal and abnormal operations including LOCA related design loads in and above the suppression pool including negative pressure differences between the drywell, wetwell and reactor building. The containment structure has design features to withstand coincident fluidjet t forces associated with outfiow from the postulated rupture of any pipe within the l primary containment. l The containment structure has design features to accommodate flooding to sufficient depth above active fuel to permit safe removal of fuel assemblics from the reactor core after a postulated design basis accident. 2.14.1 2- 6/1/92
ABWR oesign 0:cument The containment structure is protected from and designed to withstand f~') hypothetical missiles from sources within the primary containment and pipe L' whip due to the uncontrolled motion of broken pipes. The centainment structure is configured to channel flow from oostulatt d pipe ruptures in the dnwell to the pressure supprcssion pool designed with the required vent submergence and water vohune to acconunodated the energy of the fluid released. The containment str ucture and penetration isolation system with concurrent operation of other accident mitigation systems, are designed to limit fission product Icakage during and following a postulated desiga basis accident to values well below leakage calculated for allowable ofTsite doses. I cakage tests are described below. The containment system has features for performing periodic leak rate teste at a reduced test pressure based on a 39 psig peak LOCA pressure initial leak test u establish primary containment leakage limit of 0.5% by weight per day of the primary containment free air volume. Type H tests measure loca' leakage, such as, individual air locks, hatches, dowell head, piping, electrical and instrument penetrations. Type C tests measure isolation valve leakage, and the Type A test measures the integrated containment leak rate. The indisidual and integrated preoperationalleak rates are recorded in the plant technical specnications for (]
'v comparison with the periodic leak rate test results. Periodic Type A integrated leak rate tests ar e conducted (three in a ten year period in nearly equalintervals with the third test at the ten yerr plant in-senice inspectiont Ily-pass leakage between the n.,well and the wetwell through the drywell diaphragm floor and the wetwell to dowell vacuum breakers is desig icd 0.05 square feet of area based on A over the square root of K, established by the preoperational test. The recorded value in the technical specifications is 0.005 square foot and is periodically tested and verified to be less than this rate and is conducted at a wetwell air chamber pressure that does not clear the dowell to wetwell vents.
A dowc" :o reactor vessel refueling bellows and reactor well platform are provided to seal off the dowell during refueling to enable the icactor well to be flooded prior to removal of the reactor steam separator and fuel bundle manipulations. Piping, cooling air ducts and return air vent openings in the reactor well platform must be removed, vents c '. and scaled watertight before filling the reactor well with water. The reiuding bellows also absorbs the movement of the vessel caused by operating temperature variations and seismic , actisity. U 2.14.1 -3 6/1/92
ABWR oesign occument The primary containnu nt isolation is accomplished with inhoard and outboard g isolation vakes on each piping lanetration which are signaled to close on W detection of high drrwell pressure or reactor low water lew1. Safety systems performing a post 1.OCA function are capable of opening their isolation vahes as needed. The dr>well hiced system prosides the means to reduce containment prt ssure following heat up of the dr>well during reactor startup. A containment vent system consisting of dual ruptur e disks in series are prosided l to relief containment overpressure and isolation valves are prosided for ! reclosure of the containment. l l The standby gas treatment system is conr.ectable to the containment purge ! exhaust sys tein in the event the containment contains airborne radioactisity requiring removal with the nitrogen iner t gas atmosphere prior to personnel entry of the dnwell and wetwell. Drywell coolers are prosided to remove heat released into the dr3well atmosphere during normal reactor operations. Dry sell and wetwell sprays are prosided to liinit temperature and pressure following an accident within the primary containment. g The Flammability Control System prmides redundant h>diogen recombiners :o remove any hydrogen releas d into the primary containment during a 1.OCA. Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.1 provides a definition of the inspection, test, and/or analyses togethen with the associated acceptance criteria which will he under taken for the Primary Containment System. ! O l 2.14.1 -4 fiU92
,m U,,
p V U Table 2.14.h Primary Containment System [ Inspections, Tests, Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment
- 1. Review the as-built Primary Containment 1 As-built Primary Containment System
- 1. The configuration of the Priraary instattations conform :o the configuration Containment System is shown in Figure System constn:ction records and conduct onsite inspections to confirm the for all components shown in Figure 2.14.la. 2.14.1a.
configuratsn is as shown in the design docu.nents.
- 2. Verify from as-built documents the two 2. As-built documentation and calculations
- 2. Drywell free air volumes are: shall confirm the free 7ir volumes of Upper Upper Drywell: 5490 cubic meters. independent dryweil volumes less internal structures, piping. RPV and equiptrent and Lower D7weil are not more than the i Lower Drywell: 1860 cubic meters. design values. l (1220 cutic meters in upper drywell and 30 Drywell: 7350 cubic meters. l cubic meters in the lower drywell) are equal or less than the design free air l volumes.
The Upper Drywellis 29 M diameter,9 M high from diaphragro floor to ceiling. Lower Drywell is 10.6 M diameter.11.55 M from invert of RPV to tcp of basemat.
- 3. Verify from the as-built documents and 3. As-built documentation and calculations
- 3. Wetwell volumes are: shall confirm the water volumes of the Suppression Pool: low water level 3580 calculations the Wetwell less internal structures, piping and equipment (50 cubic Suppression Pool and the free air volume cubic meters; high water level: 3620 cubic of the Suppression Chamber are not less meters) is equal or more than the design meters. than the design values.
Suppression Chamber: high water level free air volume. The minor diameter is 14 M.the major diamet~ !s 29 M and the 5960 cubic meters. height from the basemat to the ceiling Total Wetwetl: 9580 cubic meters. (underside of the diaphragm floor)is 7 M. 4 Measure and by visual inspection verify the 4. Confirm the wetwell drawdown volume 4 Suppression pool water drawdown volume that can be contained in the lower dryweit due to holdup in the Lower Drywell is five 0.3 m diameter return paths from the lower drywell are not insta!!ed higher than below the five return paths in the drywell based on the Lower Drywell volume below to wetwell vertical vents is less than 815 the five return openings in the lower elevation (-}4550 mm through the wall into the vertical drywell to wetwell vents. cubic meters. Confirm the five vents are a drvwell wall connect into the drywe!i to minimum of 0.3 m diameter and conne.;t to wetwell vertical vents. Calculate the volume of water that could be contained below the five return paths. the drywell to wetwell vertical vents. R
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[ Table 2.14.1: Primary Containment System (Continued) .A ~ Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. The configuration of the Drywell Head is 5. Review documentation of the insta!!ed 5. Confirm the as-built configuration of the shown in Figure 2,14.1b drywell head and associated equipment for Drywell Head and associated equipment is compliance and (if applicable) the code designed, fabricated, installed and tested stamp on the hardware. in compliance with applicable codes and regulatory requirements.
- 6. Two air lock type personnel access hatches 6. Review as-built documentation, 6. Confirm the as-built personnel locks and and three equipment hatches are provided operational test reports and by visual equipment hatches are located as shown in through the primary containment wall are inspection of the installation and operation Figure 2.14.1b and are designed, shown in Figure 2.14.1b. of two air lock type personnel access units fabricated, installed and tested in and three equipment hatches determine compliance with applicable codes and compliance and the code stamp on the regulatory requirements.
hardware.
- 7. Primary containment leakage is minimized 7. Review as-built documentation, test 7. Confirm that liners havu been designed, with drywell and wetwell liners anchored reports and conduct visualinspection of all fabricated, instatted and leak tested.
to allinterior sides of primary containment primary containment liner welds at joints. ? perimeter walls, ceilings and floors. Wetted penetration sleeves and structural portions of suppression pool walls and interfaces. Verify tests of seals at the floors are steel lined with stainless steel drywell head, personnel locks and hatches. cladding. Both surfaces of the upper drywell diaphragm floor are lined, and the lower drywell floor is lined. The pedestal and reactor shield wall are constructed of steel with concrete fill. The drywell head and per::;onnel locks and hatches are steel with dcuble type testable seals.
- 8. Primary containment is designed as a 8. Review as-built documentation to verify 8. Confirm that primary containment Seismic Category I reinforced concrete construction moterials were tested to reinforced concrete structure, materials, structure. required stardards, placed and installed as and appurtenances have been designed, configured for the Seismic Category I fabricated, installed and tested in requirements. compliance wwellis provided through a personnel tunnel with a double scaled personnellock at the reactor building end. An equipment access tuunct is scaled by a component hatch removable only during refueling or maintenance outages. These access tunnels span the suppression chamber. Access to the suppression chamber is provided by a hatch located in the reactor building.
The primary containment internal structures are designed to withstand coincident fluidjet forces associated with outflow from the postulated rupture of any pipe within the primary containment. The primary containment internal structures are protected from and designed to withstand hypothetical missiles from sources within the primary containment and pipe whip due to the uncontrolled motion of broken pipes. The primary containment dr)well to wetwell vents are conngured to channel g flow from postulated pipe ruptures in the drywell to the pressure suppression W pool. These internal structures are designed with the required vent 2.14.2 2- 6/1/92
ABWR oesign occument p V submergence and water volume to accommodate the energy of the fluir' . leased into the suppression pool. Irakage between the wetwell and the drywell through the drywell diaphragm floor and the wetwell to drywell vacuum breakers is periodically tested and verified against the allowable values of the design, established by the preoperational test and recorded in the technical specifications. Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.2 provides the definition of the inspection, tests and/or analyses together with the associated acceptance criteria which will be undertaken for the Containment Internal Structures. I l (D V i l ID %)
.2.14.2 3 6/1/92
y Table 2.14.2: Containment Internal Structures e inspections, Tests, Analyses and Acceptance Criteria Inspections. Tests Analyses Acceptance Criteria Certified Design Commitment
- 1. F;eview the as-built Containment intemal 1 As-bui;t Containment Intemal Structures
- 1. The configuratton of the Cor'tainment installations conform to the configuration intemal Structures are shown in Figure Structures construction records and conduct onsite inspections to confirm the for all components shown in Figure 2.14.2a. 2.14.2a.
configuration is as shown on the design documents.
- 2. V,sually inspect the p!atform and verify 2. As-built functional tests and visual
- 2. The ;c ver dryweII equipment platform is a inspection shall confirm the equipmert circular structural steel assembly which that the CRD,incore monitoring and RIP equipment handling features are installed. platform is able to accomplish both the can be rotated on a continuous support rail CRD and incore monitor removal and mounted on independent columns and installation tasks and both the RIP removal beams. Special CRD and incore monitor and installation tasks.
handling equipment and RIP handling equipment are mounted on the platform.
- 3. Confirm the Lower Drywell RIP hoist f The RIP hoist and support beam is 3. Demonstrate and visually inspect the
- 3. satisfies its design requirements.
designed to tilt the RIP unit both from a capability of the hoist to tilt the RIP unit horizonta'i position to a vertical position from the horizontal to the vertical position A
~ and remove the RIP protective Cover. Also and vice versa. The hoist shall both install and remove the RIP protective cover. verify the capability to instai! the RIP cover and tilt the RIP unit from the vertical to the horizantal position and plact the RIP unit on the dolly.
Review documentation and visually inspect 4. Confirm the Lower Drywell restraint beam
- 4. The Lower Dryweil restraint beam is 4.
and the associated structures satisfy the designed to support the CRD housings, the the installed restraint beam and associated structures to veafy the CRD housings, design requirements for support of the FMCD units, the incore monitors, and CRD housings FMCD units,and incore stabilize the assembly during a seismic FMCD units and the incore monitors are prov;ded support. monitors. event.
- 5. Confirm the CRO hydraulic piping supports 5 The CRD h, 'raulic piping supports 5. Review the as-built documentation for the CRD hydraulic piping supports and visually are in their design locations and are extending from the control rod drive designed, fabricated and instalfed in housing
- to the primary containment inspect the supports for seismic anchors compliance with apphcable codes and penetrations are Safety Class 2. Quahty and braces.
regulatory requirements. Group B and Seismic Catagory L E
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~ . Tath. 2.14.2: Containment Internal Structures (Continued) e " Inspections Tests, Analyses and Acceptance Cnteria inspecteons, Tests Analyses Acceptance Criteria Certified Desinyt Commitment
- 6. . Ten RIP structural supports are designed 6. Review the as-built documentation of the - Si Confirm the as-built configuration of the R.lP supports and visually inspect the RIP structural supports and verify they are Safety Class 2 Quality Group B and installation for compliance and (if . designed, fabricated, installed and tested Seismic CategoryIto restrain the suspended motor end of the RIP units and applicable) the code stamp on the - in compliance with applicable codes and provide an anchor and guide for the hardware. regulatory requirements.
jacking screws used to lower and raiseihe
.. unit from and to the RPV mounting flanges.
- 7. teview the as-buiit documentation of the 7. Confirm the as-built configuration of the
- 7. Ten RIP heat exchanger supports and and veri *y they are designed, fabricated, associated cooling water piping supports RIP heat exchanger sepports and the cooling water piping and visually inspect instal!ad and tested in compliance with are designed to Safety Class 3. Quality the instattation for compliance and (if applicabre codes and regulatory Group C, and Seismic Category I.
applicable) the code stamp on the requirements, hardware. l The reactor pressure vessel skirt ring girder 8. Review the reactor pressure vessel skirt 8. Confirm the as-built configuration of the 8. ring girder support docurrwntation and reactor pressure vessel skirt ring girder support is designed to anchor the reactor ? inspect the installation for compliance and support and verify it is designed, J pressure vessel to the pedestal a+ the RPV (if appI4able) the code stamp on the fabricated, installed and tested in skirt flange. hardware. compliance with applicable codes and regulatory requirements. The under vessel head insulation support 9. Review the as-built records and inspect the 9. Confirm the RPV bottom head insulaticn
- 9. support is designed, fabricated and extends from the inside surface of the reactor pressure vessel bottom head insulation suppovt to verify their are no installed to hold the insulation in place vessel skirt down between the RIP motors with no opening of joints.
and the RIP heat exchangers and across open joints in the insulation and the the bottom of the reactor vessel bottom insulation is sealed at the CRD housings, nead. incore monitors, and RWCtj reactor drain pepmg.
- 10. Review the as-built records and inspect the 10.. Confirm the reactor shield wall is designed,
- 10. The reactor shield wall is a composite reactor shield wall placement and fabricated, and instaffed to meet the Safety double walled steel cylinder with concrete uniformity of the annutus space for RPV Class 2. Quality Gmuo B and Seismic fill supported on top of the drywell inspections. Category I requirements.
pedestal. The reactor shield wall surrounds the reactor and has an internal diameter of 9440 mm, an outside diameter of 10,600 mm and a thickness of 580 mm. The shield a wa21 is designed Safety Class 2. Quality ( Group B and Seismic Category I. m ,
e Table 2.14.2: Containmect Internal Structutes (Continued) F Inspections, Tests, A:ialyses and Acceptance Criteria Certified Design Commitment inspectians, Tests, Anatyset Acceptan<.w CrFeria
- 11. The diaphragm floor is a 1200 mm thick 11. Review the as-built constrxtion and test 11. Confirm the diaphragm Ucor is desigf rsd reinforced concrete structure with steel records to verify the design requirements and ConstructrxDo meet Ifn differente,1 liners on both top and bottom surfaces. have been met. Visually hapec0iht floor pvessore and ieakago hmQs arm comply The diaphragm floor is designed to Safety and ceiling liners and psetrations. with the code end regulato y requirements Class 2, Quality Group B and Seismic Category I. This flo3r separates the wetwell from the drywed and is designed to accommodate a differential pressure of 25 psid with a leak ate limited to 10% of the drywell free air volume over a 24-hour period. ,
- 12. Upper Drywell piping support structure is 12 Review as-built documentation o! the 12. Confirm the piping support strncture is designed Safety Class 3, Quality Grcup C ir: stalled piping support structure against cesigned, fabricated, and insta red in and Seismic Category I to support, anchor the design documentation and inspect me compliance with ateplicable codes and and guide safety related piping, and piping support structure and its ties to tlw regulatory requirements.
^,,
provide pipe whip restraints to protect this reactor shield wall and the upper drywell piping. wall. Verify the pipe restraints are in place. 13, A system of monorails is provided from the 13. Review the astuilt documentation for the 13. ConCrm the monorail system is designed. ceiling of the upper drywell to allow monorails and inspect the upper drywell fabncated, insta!!ed and tested to verify the removal and reinstallation ef the inboard for placement to verify a means is ava:lable required load capscity has been provided. main steam isolation valves the safety for equipment removal and replacement. relief valves and miscellaneous valves, and equipment located within the upper drywell.
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2 ABWR onion occument
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- 1. ORywEtt HE AD 1 M M to 2 2 RCCV W 6LL 3 DRYWELL NE AD FLANGl
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Figure 2.14.2b Primary Containment Configuration 2.14.2 8 S'1/92
ABWR oesign Document l l l i 2.14.3 Reactor Pressure Vessel Pedestal e
\ Design Description The reactor pressure vessel pedestallocated in the lower dnwell consists of a composite steel and (oncrete filled su ucture rising from the base mat to the reactor vessel skir t support bracket and reactor shield wall base support element. !
Cast into the pedestal base mat are the dr>well HCW (identified leakage) sump and dnwell 1.CW (unidentified leakage) sump. Sections of the unfilled steel 1 pedestal contain the upper and lower dnwell to suppression pool submerged vents and the upper to lower dnwell pipe and conduit spac es in the ret tangular portions of the dnwell to suppression pool vent openings. Welded to the pedestal wall are the personnel access and equipment transf er tunnels spanning the suppression chamber above the suppression pool. The pedestal carries the equipment platform perimeter support rails, undertessel control rod drive and fine motion control rod drive restraint beams. lower vessel and under vessel head insulation supports, reactor internal recirculation pump supports, and reac tor internal recirculation pump cover lifting hoist stub beam. Refer to Figure 2.14.3 Reactor Pressure Vessel Pedestal. Main features include: The lower dr)well pedestal structure has the capability to maintain its functional integrity during and following the peak transient pressures and temperatures caused by the worst i OCA pipe break postulated to occur simultaneously with loss of offsite power and a safe shutdown carthquake. The pedestal structure is designed to accommodate the full range ofloading conditions a.ssociated with normal and abnormal operations including the 1.OCA related design loads. Negative pressure difTerences between the dr>well and wetwell are accommodated by eight vacuum breaker valves, piping and sleevet, penetrating the pedestal wall. These 20-inch diameter vacuum breaker valves are designed to begin opening at a wetwell to dr)well pressure difTerential of 0.1 psig and are fully open at 0.5 psig. They return the non-condensable gases to the dowell. These gases passed with the stram released from the pipe break in the dr>well through the dr>well to wetwell vents into the suppression pool where the gases were released into the wetwell air space and increased the wetwell airspace pressure. The pedestal structure is designed to withstand coincident fluidjet forces associated with outflow from the postulated rupture of any pipe within the primary containment. g The pedestal structure is designed to accommodate flooding ' a dept . below Q the five return openings provided in the dnwell wetwell vertical vents. t l 2.14.3 1- W1/97 l
ABWR Design Document l l The pedestal structure is protected from and designed to withstand hypothetical g , missiles from sounes within the primary containment and pipe whip due to the W uncontrolled motion of broken pipes. . I The pedestal structure is configured to channel flow from postulated pipe ruptures in the lower dnwell to the pressure suppression pool designed with the required vent submergence and water volume to acconunodated the energy of the fluid released. The lower drTwell floor shall have a low cathon dioxide foiming type concrete to resist formation and release of carbon dioxide when in potential contact with the reactor corium during a severe accident. The pedccal structure and eight vacuum breaker penetrations with concurrent operation of other accident mitigation systems, are designed to limit suppression pool bypass leakage during and following a postulated design basis accident to values well below leakage calculated for allowable wetwell pressures and temperatures. Leakage tests are described below. Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.3 provides the definition of the inspcctions, tests and/or analyses together with the associated acceptance criteria which will be undertaken for the Reactor Pressure Yessel Pedestal. O 2.14.3 2- fv192
O O O Table 2.14.3: Reactor Pressure Vessel Pedestal
.~
w inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria
- 1. The Reactor Pressure Vessel (RPV) Pedestal 1. Review the as-built RPV Pedestal 1. As-built RPV Pedestal construction is shown in Figure 2.14.3a. construction records to verify it matches conforms to the design configuration the design. Inspection of the configuration shown in Figure 2.14.3a.
will verify the installation matches the design.
- 2. The RPV Pedestal is designed to Safety 2. Review the as built construction records to 2. Confirm by visualinspection of as-1:uilt Class 3. Quality Group C and Seismic verify the RPV Pedestal meets the design construction records the RPV Pedestal has Category I requirements. requirements been designed and built to comply wrth applicable codes and regulatory requirements.
- 3. The RPV Pedestal Drywell to Wetwell Vent 3. Review tl e as-built construction records to 3. Confirm the RPV Pedestal Drys - a to System consists of ten vents sized 1 M by 2 vMry the Drywell to Wetwell Vent System Wetwell vent System has been designed.
M between the Drywell Door at elevation is configu.ed as show.m on Figure 2.14.3a. fabricated and installed in accordance with 7350 mm and the Lower Drywell at inspect the constructioriinstallation of the the design documentation. c, elevation (-)1450 mm. Below this elevation Drywell to Wetwell Vent System to verify the ten vents are 1.2 M diameter down tr, the inst;'lation matches the design elevation El 13150 mm. Three rows of ten documo.itation. 0.7 M diameter horizontal vents 1500 mm in length are placed at centerline elevations of (-)12,390 mm, (-)11,020 mm and (-) 9.650 mm as shown in Figure 2.14.3a.
- 4. The RPV Pedestal Drywell to Wetwell Vent 4. Cor. firm the RPV Pedestal Drywell to 4 Confirm the RPV Pedestas Dryweil to System has been designed Safety Class 2. We well Vent System has been designed. Wetwell Vent System has been dessned.
Gu9hty Group B and Seismic Category I fabricat-d and installed in accordance with fabricated and installed to Safety Class 2. ' zerpsirements. the design documentation. Quality Group B and Seismic Category I requirements.
- 5. Five of the Drywell to Wetwell Verticai 5. Feview the as built construction 5. Confirm the five 0.3 M diameter return path Vents are provided with a 0.3 M diameter documentation of the installed return path vents have been installed at elevation b) retum paths ,3t elevation F) 1450 to insure vents and verify by inspection that these 1450 mm and connect to the 1.2 M the flood level in the lower drywell is vents have been insta!!ed in conformance diameter vertical vents. Also confirm these controlled.These retum path vents are with the design documentation. return path vents have been desigred.
designed to Safety Class 2. Quality Group fabricated, and insta!Ied to Safety Class 2 B. Seismic Category I requirements. Quality Group B and Seismic Category i SP requirements.
~
[ Table 2.14.3: Reactor Pressure Vessel Pedestal (Continued) r
" Inspections, Tests, Analyses and Acceptance Criteria inspections. Tests. Analyses Acceptance Criteria Certified Design Commitment
- 6. Review the as-built construction 6- Confirm the as-built con'iguration of the ti The Lower Drywell floor is scaled with a documentation of the Lower Drywell floor Lower Drywell floor contains the Low Level steel liner which surrounds the outer leak Waste Sump, the fligh Levei Waste sump detection tanks of both the Low Level liner and the Low Level Waste Sump and Waste Sump and the High Level Waste the High Level Waste Sump. and the floor has a steel 8 ner.
l Sump.
- 7. Visually inspection the ten lower drywell 7. Confirm by inspection of the as-built
- 7. The RPV Pedestal wall contains ten equally documentation and by visual observatien, spaced 100 mm Lower Drywell f!ooders flooders to verify they are 100 mm or larger equally spaced around the Lower Drywell there are ten
- O mm drywell fiooders with mounted in ten vertical vents 1 M above 500 DEG. F. mr .ed thermal plug valves the Lower Drywell floor. Each flooder is and are 1 M above the lower Drywell floor.
Verify each floooer is equipped with a mounted 1 M above the Lower Dryweil designed to release suppression pool water into the Lower Drywell when the thermal plug valve rated to open at a f?oor and equally spaced around the temperature of 500 DEG F. Also verify the pedestal wall at ten of the vertical drywell Drywell temperature reaches 500 DEG. F.(260 DEG.C.L The Lower Drywell flooders flooders are designed, fabricated, tested to wetwell vs.nts. Also conform the thermal and insta!!ed to Safety Class 2. Quality plug valves were designed. fabricated, are designed to Safety Class 2. Quality Group 8 aad Seismic Category i Group B and Seismic Category I installed and tested in accordance with the requirements and meet the severe accident requirements. design documentation. potential conditions.
~
- 8. Tests and visual inspection of each S. As-built operationa! tests and visual
- 8. During abnormal and accident conditions inspectons sha'l confirm that the Isolation dampers are designed to isolate 9. Demonstrate with a simufated. 9. Confirm the isolation 9.
52
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ABWR oesion 0:cument 2.14.4 Standby Gas Treatment System i 1 V The Standby Gas Treatment System (Sr~S) has the capability to filter the gaseous efiluent from the primary conta....nent or from secondary containment when required to limit the discharge of radioactisit/ to the ensironment. The SGTS is designed to accomplish the fo. owing: (1) hiaintain a negative pressure in the secondary containment, relative to the outdoor atmosphere, to control the release of fission products to the environment. (2) Filter airborne radioactisity (halogen and particulates) in the process stream to reduce off-site doses. (3) Ensure t",t failure of any active component, assuming loss of off-site power, cannot impair the ability of the system to perform its safety funcuon. The SGTS co",ists of two parallel and redundant trains of active equipment which share a sing J (ilter train. Suction is taken from above the refueling flor or from the primary containment sia the Atmospheric Control System. The
, discharge g.,es to the main plant stack.
l ,\ The SGTS consists of the following principle components: (1) Two independent dryer trains consisting of a moisture separator andan electric process heater. (2) Two independent process fans located upstream of the filter train. 1 (3) A filter train consisting of a prefilter, a high efliciency particulate air (HEPA) filter, a charcoal adsorber, a second HEPA filter, and space l heaters. l Instrumentation strictly required for monitoring the operation of the SGTS to mitigate off-site releases is provided in the main control room (htCR) on panel displays designed for that purpose. Instrumentation used for testing or maintenance is located at the local instrument rack. There are two basic parameters that are important to assure SGTS function, secondary containment pressure and charcoal adsorber inlet relative humidity. If the secondary containment pressure is less than the ambient pressure, any release from the plant passes through and is treated and monitored by the SGTS. C3 If the inlet relative humidity to the charcoal adsorber is less than or equal to V 70% then credit for a 99% efficiency may be taken. If the operator confirms the secondary containment pressure is negative with respect to ambient on all faces 2.14.4 -1 6/1/92
ABWR oesign occument of the building and the celative humidityisless than 70% entering the adsorber, then the system is functioning as intended to mitigate calculated off-site doses, The ABWR SGTS design provides four divisional differential pressure transmitters with high and low alarms monitoring secondary containment pressure with respect to ambient pressure outside cach of the four walls of the Reactor Building. In addition, four divisions of moisture measurement with high alarms are prosided in tbc filter housing upstrea,n of the charcoal adsorher, prosiding a direct measurement of relative humidity. These basic parameters each have main con rol room indicatic and alarm. Figure 2.14 A shows the major system components. Key equipment performance requirements are: (1) Fan capacity (minimum) 4000 scfm ( (2) Ihyer train outlet relative humiday <0% (3) Filter train charcoal weight (nominal) 1750 lb A slight negative pressure is normally maintained in the secondary containment by the Reactor Building HVAC ;ystem. Upon the receipt of a high primary containment pressure signal or a low reactor water level signal, or when high radioactivity is detected in the secondary containment or refueling floor g ventilation exhaust, the ';GTS is automatically actuated. Upon SGTS initiation, the secondarT containment is automatically isolated from the HVAC system. If SGTS operation is not confirmed, the redundant process fan and dryer train are automatically placed into sersice. In the event a malfunction disables an operating process fan or dryer train, the standby process fan and dryer train are manually initiated. Tne SGTS has independent, redundant active components. Should any active component fail, SGTS functions can be performed by t% redundant component. The SGTS is on standby during normal plant operation and may be manually initiated before or during primary containment purging (i.e., de-inerting) when required to limit the discharge of contaminants to the emironment. If purging through the HVAC could or does result in a trip from the ventilation exhaust radiation monitors, then de-inerting can be [re-) initiated at a reduced rate through the SGTS. Use of SGTS during de-inerting is very infrequent. The SGTS may be manuaily initiated whenever its use may be needed to avoid exceeding radiation monitor set points. Cooling cf the SGTS filters may be required to prevent the gradual accumulation of decay heat in the charcoal. This heat is generated by the decay of radioactive iodine adsorbed on the SGTS charcoal. The charcoal is typically h 2.14.4 6/1/92
ABWR oesign occument gw cooled by the mir from the process fan. A water deluge capability is prosided for (j fire protection. Water is supplied from the fire protection system and is connected to the filter train via a removable spool piece. The SGTS, except for the deluge,is designed and built to meet the requirements for Safety Class 3 equipment. The electrical desices ofindependent components l are powered from separate Class 1E electrical buses. The SGTS is designed to i Seismic Category I requirements and is housed in a Category I structure. The construction materials used for the SGTS are compatible with normal and accident emironments pastulated for the area in which the equipment is located. l Inspections, Tests, Analyses and Acccptance Criteria Table 2.14.4 prosides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the SGTS. r3 V a't 2.14.4 4- 6/1/92
y Table 2.14.4: Standby Gas Treatment System a Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tesis, Analyses Acceptance Criteria
- 1. The SGTS shall be capable of rnaintaining 1. System preoperation tests will be 1. It must be shown either SGTS fan can a negative pressure of at least 0.25 ;."hes conducted to demonstrate acceptable fan maintain the secondary containment at a w g. and the re!ative humidity of the and filter performance. These tests will be negative pressure of at least 0.25 inches process stream entering the filter train conducted at steady state conditions. w.g. with the associated dryer maintaining below 70%. Each SGTS fan capacity shall the outlet relative humidity below 70%
be at least 4000 scfm measured with With secondary containment not isolated, secondary containment not isolated, fan capacity shall be at least 4000 scfm.
- 2. A simplified system configuration as 2. Inspections of installation records tcgether 2. The system configuration is in accordance shown in rigure 2.14.4. with plant walk-downs will be conducted to with Figure 2.14.4.
confirm that the installed equipment is in compliance with the design configuraiion defined in Figure 2.14.4.
- 3. The dryer, fan and associated valves can be 3. System tests will be conducted after 3. The installed equipment can be powered powered from the standby AC power installation to confirm that the electrical from the standby AC power supply.
A supply as described in Section 2.2.4. power supply configurations are in compliance with design commitments. 4 SGTS components which are required to 4. Review associated documentation. 4. Components meet Seismic Category I maintain negative pressure in secondary requirements. containment are classified Seismic Category 1. R S ~ G O
i i to a
- Stack A
west from north wall ,
* " ' " " ' " " " " " " " ' Reactw _,,-
wall Atmospheric Building Gntrol dP i g System
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dP - l dP l L mmmmn east i south wall wall 2 3 Figure 2.14.4 Standby Gas Treatment System
1 l ABWR Design Document l f 2.14.5 PCV Pressure and Leak Testing Facility No Tier 1 ITAAC f or this system. O 2.14.5 1 6/1/92
+
1 l ABWR oesign Document l O 2.14,6 Atmospheric Control System \
~) Design Description The Atmospheric Control (AC) System is designed to establish and maintain an inert atmosphere within the primary contaimnent during all plant operating modes except during shutdown for refueling ur equipment maintenance and during limited periods of time to permit access for inspection at low reactor power.
The AC System is not safety-related with the exception of the primary containment isolation portion which is required to maintain cc>ntainment integrity. The AC System consists of a pressurized liquid nitrogen storage tank, a steam-heated nitrogen vaporizer, injection lines, exhaust lines, bleed line, overprcssure protection line, associated valves, controls and instnunentation. All AC System components are located inside the reactor building except the liquid nitrogen storage tank and the steam-heated nitrogen vaporizer which are located in the yard. The AC System is designed to non-seismic class, Quality Group D requiremem3. ( ) The priman containment penetrations up to and including the outermost
'"' isolation valves are designed to Seismic Categog 1, Quality Group B.
The AC System has several modes of opemtion, namely: (a) Inerting, (b) Makeup, (c) De-inerting, and (d) Overpressure Protection. The inerting process is perfe-nned during plant startup. This is accomplished by allowing large vo?ume ofliquid nitrogenio flow from the nitrogen stomge tank, vaporized and heated up to appropriate temperature by the steam-heated vaporizer. The vaporized and heated nitrogen gas : then injected into the drywell and into the wetwell air space through penetration nozzles. The containment atmosphere should be inert to about 3.5% Oxygen by volume within 24 hours. Following the inerting process, the makeup mode takes over to maintain the containmert in inert state after manual realignment of system valves. Small volume ofliquid nitrogen from the storage tank is heated up and gasified by an electric heater. The containment pressure is kept constant at slightly higher than the secondag containment. In response to changes in the containment pressure, the pressure control valve modulates (open/close) to provide nitrogen makeup thereby maintaining containment pressure. An increase in containment f) pressure over the normal operating range is controlled by venting through the U bleed line. During this mode, nitrogen makeup to the HPIN System is also 2.14.6 6/1/92
ABWR 09 sign Document provided. Isolation signal override capability aic prosided to the makeup vahes g such that conimued nitrogen makeup can be auomplished as required. W During plant shutdowu. the containment is de-inerted with breatable air to allow personnel accessinside the containment. Airis provided by the RBHVAC Splem utilizing fans to displace nitrogen. The oxvgen volumetric concentration in the containment should be at least 18 % within 24 hours. The AC System exhaust is directed to the RBHVAC exhaust line which undergo futration and rad monitoring before being discharged to the plant stack. In the event high radioactivity is detected during venting or purging, the AC System exhaust to the RBHVAC exhaust is isolated and the flow is diverted to the Standby Gas Treatment System (SGTS) for treatment before discharging to the plant stack. The AC System exhaust flow from the containment is through the bleed valve and the SGTS vent valve which aie both operable from the main control room with isolation signal override capability. The AC System is designed to relieve the wetwell air space sbould an overpressure condition develop. A piping reliefline with two normally open valves and two nipture disks is connectad to the AC System wetwell exhaust line. The reliefline is designed to passively relieve the wetwell vapor space pressure of about 5.6 kg/cm 2g (80 psig). The normally open valves can be remote manually g closed from the main control room to re-establish containment isolation W following the opening of the mpture disks as required. The AC System also includes instrumentation required for the operation of other safety-related systems. These instrumentation are as follows: (1) Suppression pool water level instnnnentation. (2) Differential pressure instnunent between drywell and wetwell. (3) Containment water level instrument, inspections, Tests, Analyses and Acceptance Criteria Table 2.14.6 provides a definition of the inspections, tests and/or analyses together with associated criteria which wil: be undertaken for the AC System. O 2.14.6 2- 6/1/92
2. l L y Table 2.14.6: Atmospheric Control System m
. inspections, Tests, Analyses ami Acceptance Criteria i
- Certified Design Commitment . inspections, Tests, Analyses Acceptance Criteria
- 1. The configuration of the AC System is . i 1. Inspection of tMas-built AC System 1. Verification of the as-built conformance shown in Figure 2.14.6. configuration shall be performedc ' with the as-designed configurotion (Figure l
, 2.14.6).1 -
' 2. The AC System PCVisolation valvesisolate - 2. Functional testing shall be performed on - 2. Valves isolate upon receipt of auto
, upon receipt of auto isolation signals from . the system logic by simulating the auto isolation signal within 30 seconds. the Leak Detection System within 30 ' isolation signal from the Leak Detection seconds. . System within 30 seconds. . 3.' ' The AC System PCV isolation valves fail 3. Field testing shall te performed to - 3. Valves c'osure upon reinoval of power an'd/ - close on loss of power and/or air supply to demonstrate that the AC System PCV or air supply, the valve operators. isolation valve will fail in the safe direction , 4 (close) when power and/or air supply are t removed.
.6
- 4. Override capability of the AC System . 4. Functional testing shall be perforrr:ed on 4. Opening of the makeup valves, bleed -
makeup valves, bleed valve, and the SGTS the system logic to demonstrate the valve, and the SGTS vent valve in the vent valve following an isolation event. override capability of the makeup valves, presence of an isolation signal. bleed valve and the SGTS vent valve following an isolation event utilizing a ! keylock switch from the main control room.
- 5. The containment overpressure protection 5. Vendor testing shall be performed to verify 5. Verification of vendor documents certifying I rupture disk will open when containment actual disk rupture pressure against the actual disk rupture pressure and that the pressure reaches to about 5.6 kg/cm2g (80 required pressure setting of 6.6 kg/cm2 g supplied rupture disks are all identical (part - !
psig). (80 psig).- of the same batch and are made from the l same metallic sheet). I
- 6. The containment overpressure protection 6. Field testing shall be performed to 6. Valves remote manual closure from the
- line isolation valves can be remote demonstrate remote manualclosure of the main centrol room. i manually closed to re-establish overpressure protection line isolation containment isolation. valves from the main control room. i s
sn 7. Provision for instrumentation specified in 7. Inspection shall be performed to verify 7. Presence of instrumentation specified in i 5 Section 2.14.8. presence of instrumentation specified in Section 2.14.6. ; Section 2.14.6 j
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ABWR 0: sign occument _ . _
,Q 2.14.7 Drywell Cooling System C/
Design Description The Dnwell Cooling (DWC) System is designed to maintain the average dr)well temperature at or below 57'C, and maximtun k> cal temperattu e at or below 65"C during normal plant operation. The system al .o maintains the average dr)well tempenume at or below 25"C during plant test oi maintenance period. The DWC System is an air / nitrogen recirculating system consisting of three fan coil units and two HVAC normal cooling water (HNCW) cooling units. A fan coil unit consists of a reactor building cooling water (RCW) cooling coil and a fan, and the HNCW cooling unit consists of a cooling coil only. Nonnally two of the three fan coil units, and both HNCW units are in operation. The third fan coil unit serves as a standby unit. The conditioned air / nitrogen is distributed to the surious zones in the dnwell. Each cooling unit is provided with a drain pan that collects water vapor condensed m er the cooling coil. The condenstate fr om each drain pan in collected in a common header and is piped to a leak detection system (LDS) Howmeter. The system configuration is shown on in Figure 2.14.7 During loss of off-site power (in the absence of loss of coolant accident (LOCA) n signal), the fans are automatically powered from the on-site source, and only Q RCW system water is available for cooling. The system is not operated in the presence of a 1.OCA signal. The entire DWC system is classified as a nonsafety-related, non-seismic system. Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.7 provides a definition of the inspection, tests and/or analyses together with assaciated acceptance criteria which will be undertaken for the DWC system. l'3 l G 2.14.7 6/1/92
t-Table 2.14.7: Drywell Cooling System { Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections,Tesk, Analyses Acceptance Criteria
- 1. The configuration of the DWC System is 1, inspection of the as-built DWC System 1. As-bui't DWC System configuration for shown in Figure 2.14.7. configuration shall be p.srformed- those components shown, conforms with Figure 2.14.7
- 2. The DWC Systern fans operate when 2. DWC System functionni test shall be 2. DWC System fans operate when supplied powered from both normal off-sk.e and on- performed to demonstrate fan operation by either power source.
site sources. , when supplied by either normal off-site power or from the on-site power source. 0 a n O 9 9
- k n ,, ,,,--c
(.) U C/ m _ m- - _ TO HNCW UNIT A N I TO HNCW UNIT B A A A RCW 4- - RCW + - RCW-RCW + -] RCW-l l RCW-l 11 11 _L1 TT ++ GD TT +4 GD TT +4 GD C C c r - C t C . C DRAIN DRAIN \ DRAIN PAN PAN PAN
- FCU-B. ::
FCU-A FCU-C :: c, DRYWELL COOLER qy DRAIN HEADER 3y yy
! LDS LDS LDS HNCW & HNCW 4 -
HNCW->- HNCW+ 11 1 I L L
---- '~ LEGEND:
g ' ' +VD
, VD C/C COOLING COtt C C GD GRAVITY DAMPER "M + > VD VOLUME DAMPER DWC FA WC FA FCU FAN Coil UNIT C C I N DRAIN PAN
,I DRAIN
! PAN
! HNCW UNIT-A HNCW UNIT-B e! l-k k i LDS LDS i
Figure 2.14.7 Drywell Cooling (DWC) System
FABWR ossign oocument
'1-2.14.8 : Flammability Control System.
\ -- Design Description The Flammability Control System (FCS) is prmided to contial the potential buildup of oxygen in the containment from design basis radiolysis of water, The primary containment during nonnal operation is purged with nitrogen and maintained in an oxygen deficient condition (53.5 volume percent) by the Aimospheric Control System ( ACS). The objective of these two systems together
. (ACS and FCS) is to preclude combustion of hydrogen and damage to essential equipment and structures.
The FCS consists of two identical thermal hydrogen recombiners, with associated piping, valves, controls and instrumentation. The recombiner units are located in the secondarycontainment and controlled from the main control room. Each recombiner removes gas from the drywell, recombines the oxygen with hydrogen, and returns the gas mixture, along with the condensate to the suppression chamber. After a LOCA, the system is manually actuated from the control room when high oxygen levels are indicated by the Containment 1 Atmospheric Monitoring System (CAMS). Once placed in operation, the system continues to operate until it is manually shut down when an adequate margin
- below the oxygen concentration design limit is reached.
L:
- 4_O Operation of either recombiner will provide effec ive control over the buildup -
of oxygen generated by radiolysis after a design basis LOCA. Independent dr)well and suppression chamber penetrations are provided for the two , recombiners. Each penetration has two normally closed isolatio'n valves; one air
'or nitrogen operated and one motor operated.
Each recombiner unit is an integral package. All pressure-containing r equipment; including piping between components, is considered an extension of the containmen't and therefore is designed to ASME Section III, Safety Chtss
' 2 requirements; The endre package is designed to meet Seismic Category I requirements. The recombiners are in separate rooms in the secondmy
< containment and at: protected from damage by flW, dre, tornadoes and pipe whip.
The recombiner unit consists of a blower, electric heater, reacdon chamber,
; water spray cooler, a water separator, piping, valves, controls and instrumentation. During operation of the system, gas is drawn from the drywell
- by the blower, and heated. Hydrogen and oxygen in the gas will be recombined
'into steam in' the reaction chamber and condem-d in the spray cooler. The -
. condensate and spray water, along with sorne of the gas, are returned to the wetwell. The rest of the gas is recycled through the blower.
1 2.14.8 - 1- 6/1/92
ABWR oesign Document The operation of the system can be tested from the control room. The test consists of energizing the blower and heaters and obsening system operation to - see if components are pedorming propeily. Flow and pressure meastuement desices are periodically calibrated. Cooling water required for oper: tion of the system after a 1.OCA is taken from the RHR System. Demineralized water is used for functional testing of the recombiner units. The cooling water is used to cool the water vapor and the residual gases leaving the recombiner prior to returning them to the Containment. Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.8 provides a defmition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be under1aken ft t the FCS. O l 2.14.8 2- 6/1/92
ABWR nesign Document 2,14.9 Suppression Pool Temperature Monitoring System
\ Design Description The Suppression Pool Temperature Stonitoring System (SPTN1) i? a Safety-Related system designed to proside the operator with suppression pool l temperature information, The SPT51 system is a r edundant system which is powered from two separate safety disisions. The SPThi system prosides indisidual temperature indication from both disisions in the main control room and at the remote shutdown station and bulk average temperatures from both disisions for indication, treno ng, recording and alarm in the main control room. The temperature sensors are arranged in six circumferential locations around the pool such that they are out of the direct path ofjet impmgement from the horizontal vents or SRV quenchers and still be in direct sight of a SRV discharge. Each sensor location contains four vertical sensors from each division so as to reliably measure the bulk average pool temperature. All temperature sensors are located below normal pool water level and at a sufficient distance from pool walls to proside accurate local temperature measurement. Sensors are physically separated between redundant divisions and terminated in moisture protectedjunction boxes in the wetwell for sensor replacement. Temperature signal processing is designed such that a sensor division can be bypassed for maintenance or calibration and any failed or uncovered sensor (e.g. pool water h -_ level below sensor) will be excluded from the bulk averaging process, identified and anntinciated. Divisional SPTS1 system outputs are electrically isolated when provided for use other than within its respective division (e.g, process computer).
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.9 provides a definition of the Inspections, Tests, arid /or Analysis, together with the associated Acceptance Criteria which will be undertaken for L the Suppression Pool Temperature Stonitoring System. u
- O 2.14.9'- 6/1/92
y Table 2.14.9: Suppression Poul Temperature Monitoring System to inspectios, Tests, Analyses and Acceptance Criteria ' Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Suppression Pool Temperature 1. - Inspection will be performed to confirm 1. Inspection confirms that two divisions of Monitoring System (SPTM) is a safety- that two divisions of individual individual temperature indication is related system designed to provide the temperature sensor indications are provided to the main control room and operator with two divisions of suppression provided in the main control room and at remote shutdown system and that bulk pool temperature indication. Individual the remote shutdown system and that bulk average temperatures are indicated, temperature indication from both average temperatures are indicated, trended, and recorded and that system divisions is provided in the main control trended, and recorded, and system failures failures are annunciated in the main room and at the remote shutdown system. are annunciated in the main control room. control roora.
Bulk average temperature indication, trending, and sensor failure or sensor uncovered annunciation is provided in the main control room.
- 2. The SPTM system temperature sensors are 2. Inspection will be perforrried to confirm 2. Inspection confirms that SPTM arranged in circumferential locations that SPTM temperature sensors are temperature sensors are provided in six around the pool in locations such that they provided in six locations around the pool locations around the pool out of the direct 9 are out of the direct path of jet out of the direct path of jet impingement path of jet impingement from horizontal impingement from horizontal vents and from horizontal vents and SRV quenchers vents and SRV quenchers and are in direct SRV quenchers and are still in direct sight and are in direct sight of an SRV discharge. sight of an SRV discharge.
of a SRV discharge.
- 3. Each location of SPTM temperature 3. Inspection wi!! be performed to confirm 3. Inspection confirms that each SPTM sensors contains four vertical sensors from that each SPTM temperature sensor temperature sensor location contains four each division, installed below normal pool location contains four vertical sensors from vertical sensors from each division, low water level. Sensors are separated each division, installed below normal pool installed below normal pool low water between divisions and terminated it' low water level, and are separated between level, and are separated between divisions moisture protected junction boxes its the divisions and terminated in moisture and terminated in moisture protected wetwell for sensor replacement. protected junction boxes in the wetwell. junction boxes in ths Wetwell.
- 4. Electricalindependence and physical 4. Inspection will be performed to confirm 4. Inspection confirms that electrical separation is maintained between divisions that electrical independence and physical independence and physical separation is of SPTM system componend and wiring separation is maintained between divisions maintained between divisions of SPTM and any output provided for use other than of SPTM system components and wiring system components and wiring and any within its respective division is electrically and any output provided for use other than output provided for use other than within isolated. within its respective division is electrically its respective division is electrically isolated. isolated.
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js gs - G V J .. ", - Table 2.14.9:-' Suppression Pool Temperature Monitoring System (Continued) > Inspections, Tests, Analyses and Acceptance Criteria ' Certified Design Commitment inspections Tests Analyses
, , , Acceptance Criteria E.' Each division of the SPTM system can be : 5.' Testing will be performed to confirm that 5. . Testing confirms that each division of the bypassed for maintenance, calibration, and each division of the SPTM system can be . . SPTM system can be bypassed for -
testing bypassed for maintenance, calibration, and maintenance, calibration, and testing - testing h ( w m_ . _ . _ _ _ _ . _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . . _ . . . _ _ _ _ .
- g. - -. - - _
i , ABWR Design Document __
?.15 Structures and Servicing (ms
'J 2.15.1 * *stion Work No enuy Covcied by item 215.10.
4 l 4 i ') v d 7,, .,. Sim
ABWR oesign Document gg 2.15.2 Turbine Pedestal l. (') No Tier 1 entry for this system. l l l l l l l 1 i O I O J 2.15.2
~1- si;jng
. ~ , _ . _ _ . , . _ . . . - , _ _ _ . _ . . . _ _ . - , . . _ _ , . . _ , , . . . , . . _ . . . . . _ . . . . . . . . . - . . ~ _ - , . .
l F ABWR onion 0:cumat 2.15.3 Crancs and floists Design Description The ABWR Certified Design makes use of many Grancs and Hoists for maintenance and refueling tasks;These cranes and hoists
- ave load ratings greater than the heaviest expected loads to be applied as identified by site specific equipment lists for each item to be serviced.
The RB Crane used during refueling /senicing as well as when the plant is online. Durin;; refueling /senicing, the crane handles the shield plugs, drywell
. and reactor vessel heads, steam dryer / separators, etc. Minimum crane coverage includes RB refueling floor laydown areas, and RB equipment storage pit.
During normal plant operation the crane used to handle new fuel shipping containers and the spent fuel shipping casks. Coverage include the new fuel
- vault, the RB equipment hatches, and the spent fuel cask loading and washdown pits.
ne RB crane interlocked to prevent mosement of heavy loads over the spent fuel storage portion of the sper.t fuel storage pool. These cranes are single - failure proof, and can hold their load in an SSE. The Upper Drywell Hoists are used during outages to senice valves and
' +p ^
equipment inside the Upper Drywell. These hoists are Seismic Category 1. ne Lower Drywell Hoists are used during outages to senice valves and $ equipment inside the Upper Drywell. These hoists are Seismic Category I. n ib ^ The Mainsteam Tunnel Hoists are used during outages to senice MSlYs and FWR"s inside the mainsteam tunnel, m . . Other Hoists and Cranes in the ABWR certified design are used to senice and g remove plant equipment. R
. Inspections, Tests, Analyses and Acceptance Criteria
,[ This section provides a definition of the inspections, tests, and/oi analyses f together with associated acceptance criteria which will be undta ten for the B Cranes and Hoists. Table 2.15.3 Cranes and Hoists inspection, Tests, Analyses and Acceptance Criteria. LO u l 2.15.3 6/1/92
y Table 2.15.3: Cranes and Hoists b inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. A!! cranes and hoists have a load rating in 1. Compare Crane asid Hoist load ratings in 1. Load Rating greater than the Heaviest excess of the heaviest expected load per purchase documents to verify that load Equipment To be lifted the site specific equipment lists for each ratings are in excess of expected lift.
item to be serviced.
- 2. The RB Crane interlocked to prevent 2. Review Purchase Documents for 2. The RB Crane is interlocked to prevent the carrying nieavy loads over the spent fuel compliance with interlock. carrying of heavy loads over the spent fuel portion of the spent fuel pool. portion of the spent fuel pool.
- 3. The RB Crane is single failure proof and 3. Review Purchase Documents for 3. RB Crane meets single failure proof criteria can hold its load under an SSE. comphance with redundancy and seismic and will not drop its load under an SSE.
capability requirements
- 4. Upper and Lower Drywe!! Hoists are 4 Review design / purchase documents. 4 Upper and Lower Drywell Hoists are Seismic Category 1. Seismic Category 1.
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[ }8WRoesignDocument fN - 2.15.4 Elevators
&J '
Design Description The AIMR Certified design makes use of elevators to move operational personnel and light loads vertically within the plant. Four elevators are located in the Reactor Building. Two in the corners of the
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sccondary containment at 45 degrees and 225 degrees from plant north. Another two are located in the clean zone outside secondary containment at - 3 degrees and 270 degrees.
- One elevator is located inside the control building of the AIMR Standard Plant.
It is located at 90 degrees near the entrance to control building from the senice building. One elevator is located inside the senice building of the ABWR Standard Plant. It is located outside the door of the control building. , An elevator is located inside each of the Turbine Building and the Radwaste Building.
. Inspections, Tests, Analyses and Acceptance Criteria y
y No entries for this' system. l L i ( f- .. O L 2.15.4 1 6/1/92
..m. --_ _.- ._ _ - - ._.- .- _
ABWR 0: sign Document p 2.15.5 Heating, Ventilating and Air Conditioning d Design Description Design Descriptions are provided for each of the following HVAC Systems: Connol Building, Control Room Habitability Area, Reactor Building, Turbine Building, Elecuical Building, Senice Building and Radwaste Building, Tables for the Inspections. Tests, Analyset and Acceptance Criteria are included with ten IWAC System Figures. Control Building HVAC Systems Control Building safety-related air conditioning systems other than the Control Room Habitabilhy Area, are designed to maintain 85'F,50% RH at a slight positive pressure to prmide efUcient work emironments for the operators and proper emironments for stmetures and equipment to insure it has the capability to perform every safety function considering the worst case single failure for all normal and abnormal reactor operating condidons and accident conditions. Air conditioning equipment to accomplish the above is designed to maintain efficient work emironments for the operators and proper emironments for equipment and stmetures. A U Major equipment consists of redundant supply fans, prefilters,80% efliciency filters, hot water heating coils, chilled water cooling coils, and recirculation / l exhaust fans, backdraft dampers, fire dampers, and air distribution ducts and accessories. Bird screens, dust and in3ect filters are provided to protect heating and cooling coil efficiency. Corrosion resistant materials are used in the fabrication of fans, coils, cabinets, plenums, air ducts and accessories (see Figure 2.15.f.a for a simplified design configuration). All safety-related HVAC systems are served from Class 1E power from either nonnal off-site sources or on-site emergency diesel generators. Electrical equipment rooms are maintained at a positive pressure, and air movement is designed to flow to the battery rooms maintained at a negative pressure by the exhaust fans. Rooms housing the motor-generator (MG) sets, which provide power to the reactor internal pumps, are cooled by individual fan coil cooling units. These non-safety related cooling units are powered from the same electrical source as the MG set served. The HVAC Normal Cooling Water System connects to each r fan coil unit cooling coil. l O I U Smoke detectors are provided to initiate an alarm to close the return air l dampers, open the fire zone damper bypassing th- exhaust fans and start the l l l 2.15.5 6/1/92 r
ABWR Design Document supply f ans to piessutize the Control Building compartments and discharge smoke through the exhaust louvers. The supply fans are located in mechanical rooms separate from the remainder of the Control Building compartments. The supplv and exhaust fans can be started from the Control Room or the hand-of f-automatic switches on the motor control center. These fans are powered irom Class IE Elecuical Divisions 1,2 or 3. Control Room Habitability Area HVAC System The Control Room is maintained at a positive pressure for most events,7f7F, 429 relative humidity (RH) and is continuously habitable during 1.OCA, chemical release, fire, safe shutdown earthquake. tornado, fiood, and other natural phenomena to insure that the operators can safely shut down the reactor and keep it in a safe shutdmyn condition.
.\lajor equipment conrists of redundant supply fans, prefilters,80% efficiency filters, hot water heating coils, chilled water cooling coils, and recirculation /
exhaust fans, backdraft dampers, fire dampers, and air distribution ducts and accessories. Bird screens, dust and insect filters are provided to protect heating and cooling coil efficiency. Corrosion resistant materials are used in the fabrication of fans, coils, cabinets, plenums, air ducts and accessories. The Control Room habitability equipment consists of redundant HEPA and charcoal filtration units designed to meet regulations addressing Control Room habitability during LOCA and other abnormal events. These units treat air from one of two widely separated air intakes with radiation monitors controlled to select the uir intake with the non<ontaminated air or isolate both in the event contaminants are present at both locations. Provisions are included for the future installation of site dependent toxic chemical monitors with controls capable of actuating the Control Room isolation dampers. The Control Room Habitability HVAC System is Seismic Category I, located in a Seismic Category I structure with air intakes and exhausts designed for protection from the effects of wind, rain, snow, tornados and tomado missiles (see Figure 2.15.5b for a simplified design configuration. All safety- related HVAC components are served from Chtss 1 E power from either normal off-site sources or on-site emergency diesel generators Smoke detectors are provided to initiate an alarm to close the return air dampers, open the fire zone damper bypassing the exhaust fans and start the supply Ians to pressurize the Control Room Habitability areas and discharge smoke through the exhaust louvers. The supply and exhaust fans are located in mechanical rooms separate from the remainder of the Control Room Habitability Area and can be started from the Control Room or the handoff-automatic switches on the motor control anter. These fans are powered from g Class 1 E Electrical Divisions 11 or III. 2.1E5 -2 6/1/92
ABWR oesign oocwnent
,q Reactor Building HVAC Systems O Reactw l\uihling Scrondan Containment is sened in*m non4afety related HVAC equipment located in th- Turbine Buihling and is designed to maintain temperatwes between 65 to 10PF,50% RH and hold a negative 0.25-inch water gauge pressure. Air suppbc and exhaust duct systems are balanced to cause air movement f rom clean areas to areas with potential airborne radioactive contamination. Reduadant Seconday Containment isolation dampers in sciies are prosided in the main air supply and exhaust ducts where they enter the Reactor Building. These isolation dampers close whenever high airborne radiation is detected in the exhaust duct or in the Refuciing Floor exhaust air intale, or when the fans fail or are not operating, These isolation dampers are safety related. Seismic Categog I with Seismic Categog i supports and have normally open, fail closed air operators powered from Class 1 E Electrical Divisions 1 or 11.
Seconday Containment air conditioning and heating equipment consists of three 50?e air sunply fans moving 100% outdoor air which is filtered with bag-type filters, heated with hot water coils or cooled with chilled water coils before the air is distributed through air ducts to and within the Secondary Containment. Exhaust air from the Reactor Building Secondag Containment fa compartments is collected in ducts, monitored for radiation and drawn to three V 507c exhaust fans discharging into the plant stack. Seismic Category I duct supports are provided where air ducts could fall on safety-related equipment. l The Primary Containment supply fan, filter and purge exhaust fan are not safety-l related and serve the Primag C(. ainment Atmo pheric Control Syswm (see Figure 2.15.5c for a simplified design configuration). Essential Equipment HVAC System is safety related and consists of cabinet cooling (HVH), units containing fans and coohng coils connected to the Reactor Cooling Water System. Individual HVH coolers are provided for each compartment housing the following safety-related equipment: (1) Emergency Core Cooling System (ECCS) consisting of three Residual Heat Removal (RHR) pumps and heat exchangers, (2)two High Pressure Core Flooding (HPCF) pumps: (3)one Reactor Core Injection Cooling (RCIC) steam turbine pump; (4)two Flammability Control System (FCS) recombiners; (5)two Standby Gas Treatment System (SGTS) filter /dger units and the two Containment Atmospheric Monitoring System (CAMS) equipment rooms. Each room cooler is controlled to start when the equipment sened starts or when the respective space thermostat calls for cooling. The main steam tunnel has a non-safety-related cabinet cooler (HVH) (Q containing cooling coils sened from the HVAC Normal Cooling Water System. U Two fans distribute air to the main steam (MS) and feedwater (FW) isolation valve areas. These units are inanually started from the main Control Room and 2.15.5 6/1/92
ABWR oesign Document are designed to keep the temperature below 140 F. Other non-safety-related cabinet coolers (HVH) wntaining f ans and cooling wits connected to the HVAC Normal Cooling Wate Svstem are piorided for the Refueling Machine Control Room, the Insenice inspection (ISI) Rooms and the Suppression Pool Cleanup System (SPT) Equipment Room. These cabinet cooling units are controlled to start when the space thermostat calls for cooling. Radiation monitors air provided in the air environment of the iefueling door and in the main air exhaust duct in the Reactor Building to cause closure of the main air supply and exhaust duct automatic isolation dampers whenever high airborne radiation occurs. This high radiation signal will also activate the Standby Gas Treatment System to maintain the negative 0.25-inch water gauge pressure within the Secondary Containment. Smoke removid from any compartment of the Secondary Containnu s.t is accomplished by operating all thiee air supply fans and all three air exhaust fans with their tilter bypau diunpers opened. Air exhaust now limiting dampers are actuated within the fire moras not experiencing the fire to pressurire these fire zones to limit smoke intrusion. 7 The remaining areas orthe Reactor Building outside of Secondan Containment are sened by individual HVAC supply and exhaust systems designed to keep the temperatures below 104 F. Electrical Equipment HVAC consists of three safety-related systems, Seismic Category I, Safety Class 3, Quality Group C and are powered from their respective Class iE Electrical Dhisions 1,2 or 3. Outdoor air and return air is - mixed. filtered, cooled, and distributed to maintain a slightly positive pressure in the electrical equipment rooms and a slightly negasive pressure in the Diesel Generator and Day Tank Rooms except when the diesel genemtors are nmning and their two emergency cooling fans operate to keep the temperature below 110*F. Smoke removal is accomplished by stoppir the exhaust fans, closing he return air damper and opening the exhaust fan by pass damper. Continuing t - operate the supply f ans pressurizes the areas served ar <' releases the smoke through the exhaust bypass duct to the outdoors (see Figure 2.15.5d for a simplified design configumtion), Reactor Internal Pump (RIP) Rooms are supplied recirculated air cooled by HVAC normal cooling water coils and distributed by fans and air ducts. The return air is dmwn into the RIP power supplies and control panels before being recooled. This RIP HVAC System is non-safety related and non-seismic except the air duct supports where safety related equipment is located (see Figure 2.15.5e for a simplified design configuration). 2.15.5 6/1/92
ABWR oesign oocument Fine Motion Control Rod Drive (FMCRD' .u toExchanger Control Panel Rooms are sen'ed by three fan coil units (FCU) with cooling water sup}ilied by
' the IWAC Noimal Cooling Water System. These FCUs are not safety related.
Turbine Building HVAC Systems Turbine Buikling is served from non-safety-related HVAC equipment located within the building to maintain less than 104 F,50% RH and a slightly negative pressure except in electric switchgear rooms. Air supply and exhaust duct systems shall be balanced to cause air movement from clean areas to areas with - potential airborne radioactive contamination. Turbine Building air conditioning and heating equipment consists of three 50% ventilation system air supply fans moving 100% outdoor air which is filtered with bag type filters, cooled with chilled water coils or heated with hot water coils . before the air is c'Stributed through air ducts to and within the Turbine Building. General exhaust air from the Turbine Building is collected in ducts connected to three 50% ventilation system exhaust fans with bag filters , discharging into the plant vent stack. Heat from the Turuine Operating Floor is removed by roof exhaust ventilating fans (see Figures 2.15.5f and 2.15.5g for the
?1 simplified design configurations).
Sepamte Lube 0:1 Area exhaust fans and ducts are provided to serve the LO l storage and pump rooms to remove lubricadng oil (LO) fumes and discharge them from the plant vent stack.
- Compartments with potential radioacthe contamination are collected in sepamte exhaust' ducts and moved by the compartment exhaust fans with bag filters and radiation monitors to the plant vent stack.
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Compartments housing heat releasing equipment are provided with multiple fan recirculation fan coil unit coolers'with cooling coils and filters to keep l temperatures below 104'F.
; Smoke removalis accomplished with operation _ of the Turbine Building roof
=
- power exhaust ventilators, supply fans with the return air damper closed, exhaust fans with their exhaust filter bypass dampers opened and fire zone smoke '
dampers positioned to create a positive pressure in the areas adjacent to the zone experiencing the fireTThe Turbine Building supply and exhaust fans can be' started from the Control Room or the on-off automatic switches on the motor : control center in the Electrical Building. Electrical Building HVAC Systems.-
.p 5b Redundant air supply units with filters, cooling coils and fans are provided to j maintain a positive pressure in the non safety related Electrical Switchgear 2.15.5 6/1/92 gr- .,. ., -r cy .a 4 r-gwr '-- egme.- 4i-m e e w-y-w+- + m 7y-- fr
ABWR oesign Document Rooms. Return, exhaust fans and e4 du,'< t>roside the ventilation. Recirculating g fan coil unit cooleis help maintain the tem,.- atm e below 104'F.in tl2 Electrical W Switchgear Rooms and the Air Comp:essor Room. A negative pressure in the At.xilian- Boiler Rooms and Combustion Gas Tm hine Generator Room is accomplished with roof exhausers (see Figure 2.15.5h for a simplified design configuration L Smoke removal is accomplished by closing the return air dampers and circulating all outdoor air within the Electrical Building. The Heating Boiler l Room and Combustion Turbine Generator Room are maintained at a negative pressure relative to the Electrical Switchgear Rooms, Chiller Room, Air Compressor Room and the stair towers which are maintained at a positive pressure. Fquipment rooms position their fire zone smoke damj>ers to increase pressurization when the fire is in an adjacent area. Supply and exheust fans can be started and dampers aligned from the Control Room or the hand-off-automatic switches on the motor control center. l l Service Building HVAC Systems j l The Senice Building is sened f rom non-safety-related HVAC equipment located within the building to maintain 72 F,50% R4 and a slightly negative pressure except in corridors and electrical equipment rooms (see Figure 2.15.5i Senic. Building HVAC Systems for a simplified design configumtion). Senice Building air supply to the nonradioactive area is provided with a mixture of outdoor and return air which is filtered, cooled, dehumidified or humidified and distributed by redundant fans through air ducts and diffusers to three reheat zones controlled by zone thermostats. Cooling is provided by the HVAC Normal Cooling ' Water System and reheat by the Hot Water Her.ing System. Air supply and exhaust duct system are balanced to cause air movenient from clean areas to areas with potential airborne radioactive contamination. Senice Building air supply to the potentially radioactive area is provided with 100% outdoor air which is filtered, cooled and distributed by redundant fans and air ducts to a single reheat zone controlled by a thermostat. The potentially radioactive area is maintained at a negative pressure by redundant exhaust fans which draw the exhaust air through filters before discharge to the vent stack. The exhaust air flow is controlled by a variable air operated damper with signals from a fiow meter and radiation monitor. Room cooling is supplemented by fan coil units with filters and cooling coils provided with HVAC normal cooling water. The Chemical Counting Room, Computer Room and Technical Support Center are provided with cooling units having redundant fans. The space temperature is controlled by thermostats modulating the HVAC normal cooling water valves. 2 15.5 6- 6/1/92
l ABWR oesign oocument p Smoke iemoval can be accomplished by clasing the non radioactive contiolled ( .uea ieno n aii d.unper to pienutire this anca and positioning the fine ione sinole dampei in the exhaust duct to hv pass the exhaust fans and semove the smoke through the exhaust lomers. T he nadioac tive (onuolled ;aca supplv and exhaust f ans (iiculate all outdoor air and normally maintain this area at a negative prewme compaird to the non-radioac tive contiolled area. The nulinactive controlled anca exhaust f ans can remove smoke fiom both the non-radioactise controlled area and the radioac tive contiolled area. Supoly and exhaust fam and return .dr and fisc zone dampers can be contiolleo fiom the Comiol Foom or f om the hand 4;ff automatic switches on the motoi contaal cen'er. Radwaste Building HVAC Systems The Radwaste Ituilding is se:Ted from non-safety-related ilVAC equipment located within the building to maintain 65 to 104'F. 50"i RH and a slightly negative pienuie except in the Radwaste Control Room. An supply and exhamt duct systeins are balanced to cause air movement fiom clean areas to areas with potential als borne radioactive contamination (see Figur e 2.15.5j for a shnplified design configuration). Radwaste Building air supply to potentially radioactive areas is provided with (v 100% outdoor air which is filtered, cooled, and distributed by redundant fans and air ducts to several reheat zones each controlled by a thermostat. The potentially radioactive area is maintained at a negative pressure by sedundant exhaust fans which draw the exhaust air through filters befoic discharge to the vent stack. The exhau t air flow is controlled by a variable ir operated damper with signals from a flow meter and radiation monitor. Radwaste Building process tanks are connected to a ta..' veru incufer sygem that equalizes air outflow from tanks being filled with air inFow needed for tanks being emptied. An) excess air is exhausted through a filter, radiation monitor and redundant exhaust fans to the plant vent stack. The Radwaste Control Room is maintained at a positive pies 3ure by varying the air flow to the redundant exhaust fans by a variable position damper. Smoke removal is accomplished by opening the exhaust fan by-pass damper to enable tl e dual Radwaste Building air supply fans to be started to pressurire all areas. Smoke is discharged to the stack. The supply and exhaust fans can be controlled from the Radwaste Building Control Room or the hand off autom;"ic switches on the motor control ce,ter. p) 1 l
?.% 5 7- Od1/92 l
1
. _ . _ _ .. I
l ABWR oesign occument l Inspections, Tests, Analyses and Acceptance Criteria Tlw folh ming taliles jn oside the insjwctions Tests. Analvseutui amici.oed l Aueptanc e Ciitriia which are to be accoinplished tot the plant il\'AC sptems. 1 Table System 2.15.Sa Control Building HVAC Systems i 2.15.5b Control Room Habitability Area HVAC System l 2.15.Sc Reactor Building HVAC Systems 2.15.5d Turbine Building HVAC Systems 2.15.5e E!ectrical Building HVAC Systems 1 2,15.5f Service Building HVAC Systems 2.15.59 Radwaste Building HVAC Systems O l t l l l O 2.15 5 8 6/1/92
a . n . x w m Table 2.15.5a: Contml Building HVAC System in Inspections, Tests, Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment
- 1. Inspections of the as-buitt HVAC System 1. As-built Control Building HVAC System
- 1. The configuratbn of'he Control Building installations conform to thi configuration HVAC Systems are shown in Figure construction records shall be performed.
Visualinspection of the configuration shall for all components shown in Figure 2.15.5a. be accomplished. 2.15.5a.
- 2. Tests and visualinspection of the three 2. As-built operational tests and visual
- 2. Three Contral Building HVAC trains are inspection shall confirm independance of mechanically and electrically independent. ind6 pendent trains will be conducted which will include independent and the three electrical devisions.
coincident operation of the three trains to demonstrate complete divisional separation.
- 3. Demonstrate and visually inspect the 3. Confirm that the Control Building exhaust
- 3. Exhaust fan bypass dampers are designed capability of each exhaust fan bypass fan bypass dampers are capable of being to enhance smoke removal from the aligned and cperated from inside or
@, Control Building in the event of a fire inside damper to open, retum air damper to close or outside the Control Building. Refer to and the exhaust fans to be stopped from outside the Control Room and able to the Control Room or aligned and remove smoke from the iontrol Building l Table 3.2b, Ventilation and Airborne '
Monitoring. positioned from outside the Control Room with their hand-off-automatic (H-O.A) switches in the motor control center (MCC) to remove smoke from the Control Building. Cont-ol Building HVAC equipment is 4. Review documentation of the installed 4. Confirm the system equipment is 4. equipment, instruments, ducts, piping and designed, fabricated, installed and tested desigr'ed to Safety Class 3, Quality Group in compli:nce with appficabie codes and C Seismic Category I requirements and is supports for compliance, and (if applicable) the Code Stamp on the hardware. regulatory requirements. Visuatly inspect powered from Class 1E Electrical Divisions 1, 2 or 3. the electrical installation to confirm C1.ss 15 Electrical Divisions 1. 2 and 3. R i
o d s f d r n n o e ol ss e o d mre t r u e la oAe ir e nog m uwt od e .ka h m o it la af l r l t ir t f n oRdoeo i ae amt ConF uie uk oainnea a Ho Ac n Vn n i w i ud sn vep d e i ssh vat dt aof t nat enio hi a r t r r ih ir d bo o r ie r b rotisn el aat spn of l ua is O i r et i o nd . nh ee t o t e hla h a n s at npt bh e oau o e aoe v ecCh t pt i ft la s sin smooat gh r s t smi o t la r e ie C otst er s sr t af ail emai t v ah mc sh b o l nnn i e t if i v t nl wt rd ndgr oer t c oie la ni l a oa t e et a emno n a it n od n c nsapt aa o ais it n nf n t p us r t p o l cl c a ol ige l it asah a ut a k eml s a sa lonl ep t a o e igym c ita lair ahs nnt t t dmn s ioisAt or u c f So rh t r oiat ig ny c esc e a si s et ay y i n A p nl e p sirt r cf it e ot e cl s i o oeo o na oi iaehh t it n ht oelim i o . t id r t t lt it it aommwio s. ml nleb n uttahb r raol. il c c m t 5 iu ew su e b phddf r o si r aieak t e i gpt f i a na odt e bib- h5 bpt nimbin s v t sai 1t s- nh se s sgt r oeo a nig s y i a r AHw2 Ait Aih niuiol oaccri an CdcHes S t e 1 2 3 4 C i r A C e n r V l e h o iat m o r H c df st o n it .e a o t n a n r t iat iast ear s r o o o f a od s e dcnd n r nd k mpe laet yC r t p neoe wedi t hd ea at nhli e A e s a ri t h t t cna e uat r a cih a c r nt t eme gt ih b am l y c e s Cncs o ei h dt on c gi a it s a a t c nn e air hi e do dhi t et t l t y Aipp i l i A la Vt c nms f oewo o cd t s i f orodr tahhi ebr et ba t r al o b a d n Huio r n o e ehi v n ei nl eo n t a laoHs s t n A lt s a c t l c it b pt f d oat a o e ehg u nd i i b a s t
- s. iunus a boie cl l e oe it ml d ps c r ut uOl ch oi men isa o t a e s
e s c V. b pwd ei t p om e oll s eCs e H s T amdl s sn inne olp s bni s n aah tewn i dg n asaAo y eeea n ei O l ml s i d amt i egf icatc i h eVc t o a n h t t smh s l a ia r ur o ut lcpe ce l a h ut inogs r e il wre Hs o mve n yr o A i t f oSon st is nune pd sel oee t el R c syr i f o v ni n ot . t aol a i
, e p nipa t et el t an vi gd o ht me .
aco ri v l o s s oi l ddli nr to d ak n t smR n r t s n ibeu r nnwesi t n n, i a amda nt nel o t ae dn a t io t n T e i cab et lig pblan i f t sphic einmpc 1 t or a s ae laik t rdog r ent nfn a yid ia oh mctnl oim. r t a o o e he o o s c s sdhi s e n C s InHsc ah o en Tiwcds oe e T epomiot sciar n a Dt CiR n b o . 5 i t 1 2 1 4 c 5 e e 1 p n y s h 2 s i t nyh o t n n i oigwi t t e e I mw l i b il t i t e l oo a ibh e a it b los a t n Rs oh it b dnt ita t oh T ob isfo f e l oi s Hn ad c a gf n o o t r m n Hih r t d o t t m d a dt i m n ote nly ed no eet al id n ob ni mCs o eS y Cc i . a c apr cCso gs n C ht y Ann t a A e r e ise o n f it Vae i dV f a dg r hd g oil cn nHoc a ei n is e n ba mee omp la mlet a r r s a ue D iiot o e ob s sds t a b b R hdn moal e a d uHa5 r l t oi r o R pe av lvme a ole ie 5 ob y nl oce . s v or f igd 1 r b r e l t ell nr ne n Rl a oac a it n n f r at r e a 2 gcn st aok ol oi c o i a C cCr e Cmt i eau oic r i nCni a r t n it l atrn m e wa e u eid a ai i O hV ig r oo5 THF Tt r le Dth r t r a IsCc
- 1. 2 3 4 yi n h PtN
t
,im V V V Table 2.15.5b: Control Room Habitability Area HVAC System (Continued) 2
?
" Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment Exhaust fan bypass dampers are designed 5. Demonstrate and visut lly inspect the 5. Confirm the Control Room smoke remova!
5. capability of each exhaust fan bypass equipment is capable of being afigned and to enhance smoke removal from the operated outside the Control Room and l Control Room in the event of a fire inside d?mper to be opened, each return air damper to be closed end the exhaust fans able to ret.,ove sc*oke from the Control or outside the Control Building. to be stopped by their remote manual Room. i switches (RMS) in the Control Room or the l hand- off-automatic switches in the motor control center (MCC) outside the Cont:vi Room. All outdoor air pressurization of the Control Room removes the smoke though l the exhaust louvers.
- 6. Test and visually inspect the air treatment 6. Confirm treatm.mt equipment is in G. Habitability air treatmerit equipment is equipment to demonstrate that all of the compliance with acceptance criteria of desigr*ed to meet the requirements of components are ready to perform their applicable standards relating its functiona!
applicable regulations and standards. Refer function in accordance with applicable performance. to Tab:e 3.2b Ventilation and Airborne Monitoring. standards.
- 7. Review documentation of the instaIIed 7. Confirm the system equipment is
- 7. Control Room Habitability Area HVAC equipment, instruments, ducts, piping and designed, fabricated, installed e,d tested equipment is designed to Safety Class 3,
( supports for compliance, and (if applicable) in compliance with applicab% codes and Ouality Group C, Seismic Category I the Code Stamp on the hardware. regulatory requirements. Visually inspect requirements and are powered from Class the electrical insta!!ation to confirm the 1E Electricat Divisions 2 or 3. Class 1E Electrical dvisions 2 and 3. E.
- l
Table 2.15.5c: Reactor Building Heating, Ventilating And Air Conditioning (HVAC) System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitrrent inspections, Tests, Analyses Acceptance Criteria
- 1. The configuretion of the Reactor Building 1. inspections of the as-built HVAC System 1. As-built Reactor Building Secondary Secondary Containment HVAC System is construction records shall be performed. Containment HVAC installation conforms showr* in Figure 2.15.5c. Visualinspection of the configuration to the configuration shown in Figure components shall be accomplished. 2.15.5c.
- 2. Secondary Containment dual i:,olation 2. Review the documentation of the as- 2. Confirm by visual inspection the isolation dampers of the main air supply and insta!!ed isolation dampers to verify dampers are designed, f abricated. installed exhaust ducts are designed to Safety Class compliance with the required standards and tested in complianc}}