ML17212A064
| ML17212A064 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/2017 |
| From: | Dominion Nuclear Connecticut |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17212A038 | List:
|
| References | |
| 17-208 | |
| Download: ML17212A064 (322) | |
Text
MPS-3 FSARMillstone Power Station Unit 3 Safety Analysis Report Chapter 1 MPS-3 FSAR 1-i Rev. 30CHAPTER 1-INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Table of ContentsSection Title Page
1.1INTRODUCTION
......................................................................................1.1-11.2GENERAL PLANT DESCRIPTION.........................................................1.2-11.2.1General........................................................................................................1.2-11.2.2Site..............................................................................................................1.2-11.2.3Structures....................................................................................................1.2-11.2.4Nuclear Steam Supply System....................................................................1.2-21.2.5Instrumentation and Control Systems.........................................................1.2-41.2.6Radioactive Waste Systems........................................................................1.2-41.2.7Fuel Handling.............................................................................................1.2-51.2.8Turbine Generator and Auxiliaries.............................................................1.2-51.2.9Electrical Systems.......................................................................................1.2-61.2.10Engineered Safety Features........................................................................1.2-61.2.11Cooling Water and Other Auxiliary Systems.............................................1.2-81.2.12Reference for Section 1.2..........................................................................1.2-101.3COMPARISON TABLES..........................................................................1.3-11.3.1Comparison with Similar Facility Designs.................................................1.3-11.3.1.1Comparison of Nuclear Steam Supply System...........................................1.3-11.3.1.2Comparison of Engineered Safety Features................................................1.3-11.3.1.3Comparison of Containment Concepts.......................................................1.3-11.3.1.4Comparison of Instrumentation Systems....................................................1.3-21.3.1.5Comparison of Electrical Systems..............................................................1.3-21.3.1.6Comparison of Radioactive Waste Systems...............................................1.3-21.3.1.7Comparison of Other Reactor Plant Systems.............................................1.3-21.3.2Comparison of Final and Preliminary Designs...........................................1.3-21.4IDENTIFICATION OF AGENTS AND CONTRACTORS......................1.4-11.4.1Licensee's Subsidiaries...............................................................................1.4-11.4.2Architect-Engineer......................................................................................1.4-11.4.3Nuclear Steam Supply System Manufacturer.............................................1.4-11.4.4Turbine Generator Manufacturer................................................................
1.4-1 MPS-3 FSAR Table of Contents (Continued)
Section Title Page 1-ii Rev. 301.5REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION........................................................................................1.5-11.6GENERAL REFERENCES (HISTORICAL)............................................1.6-11.7DRAWINGS AND OTHER DETAILED INFORMATION.....................1.7-11.7.1Electrical, Instrumentation, and Control Drawings....................................1.7-11.7.2Piping and Instrumentation Diagrams........................................................1.7-11.7.3Loop and Systems Diagrams......................................................................1.7-11.7.4Other Detailed Information (Special Reports and Programs).....................1.7-11.8CONFORMANCE TO NRC REGULATORY GUIDES..........................1.8-11.8NNSSS CONFORMANCE TO NRC REGULATORY GUIDES.............1.8N-11.9STANDARD REVIEW PLAN DOCUMENTATION OF DIFFERENCES..........................................................................................1.9-11.10TMI ACTION ITEMS..............................................................................1.10-11.11MATERIAL INCORPORATED BY REFERENCE...............................
1.11-1 MPS-3 FSAR 1-iii Rev. 30CHAPTER 1-INTRODUCTION AND GENERAL DESCRIPTION OF PLANT List of Tables Number Title1.3-1Design Comparison1.3-2Comparison Of Engineered Safety Features1.3-3Comparison Of Containment Concepts1.3-4Comparison Of Containment Atmos phere Pressure Sensor Parameters1.3-5Comparison Of Reactor C oolant Pump Bus Protection 1.3-6Comparison Of Engineered Sa fety Feature Actuation Signals1.3-7Comparison Of Emergency Generator A nd Steam Generator A uxiliary Feedwater Pump Start Signals1.3-8Comparison Of Process And Effl uent Radiation Monitoring Systems1.3-9Comparison Of Area Radi ation Monitoring Systems1.3-10Comparison Of Airborne Radiation Monitoring Systems1.3-11Comparison Of Electr ical System Parameters1.3-12Comparison Of Radioactive Liquid Waste Systems1.3-13Comparison Of Radioactive Gaseous Waste Systems1.3-14Comparison Of Radioactive Solid Waste Systems1.3-15Comparison Of Other Reactor Plant Systems1.3-16Comparison Of Final A nd Preliminary Information1.6-1Topical Reports as Genera l References (Historical)1.7-1Electrical, Instrumentation, A nd Control Reference Documentation1.7-2Piping And Instrumentation Diagrams1.7-3Omitted1.7-4Special Reports And Programs1.8-1NRC Regulatory Guides1.8N-1NRC Regulatory Guides1.9-1Summary Of Diff erences From SRP1.9-2SRP Differences And Justifications1.10-1TMI Action Items MPS-3 FSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
1-iv Rev. 30 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT List of Figures Number Title1.2-1Not Used1.2-2Plot Plan 1.2-3(Sheets 1-3) P&ID Legend MPS3 UFSAR1.1-1Rev. 30CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANTThis Final Safety Analysis Report (FSAR) has been prepared with the guidance of Regulatory Guide 1.70, Revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, dated November 1978. The report is inte nded to be responsive to the guide, to existing regulations, and to NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition.
1.1 INTRODUCTION
This report was submitted in support of an application by the companies listed in the General Information Section of the application (the Applicants) for a Class 103 permit for a facility operating license to operate a nuclear power plant, designated as Millstone Nuclear Power Station - Unit 3 (Millstone 3). This plant is located on a site in the town of Waterford, New London County, Connecticut, on the nor th shore of Long Island Sound.Millstone 3 uses a pressurized water type nuclear steam supply system (NSSS) furnished by Westinghouse Electric Corporation (WNES) and a turbine-generator furnished by the General Electric Company (GE). The remainder of the unit, incl uding a subatmospheric reactor containment, was designed and constructed by the Applicants, with the assistance of their representative, Northeast Utilities Service Company (NUSCo.), and their architect-engineer, Stone & Webster Engineering Corporation (SWEC).The core was originally designed for a warranted power output of 3,411 MWt, which was the original license application rating. This output, combined with the reactor coolant pump heat output of 14 MWt, gave an NSSS warranted output of 3,425 MWt. The core has been re-analyzed for a power output of 3,650 MWt which, when combined with the revised reactor coolant pump output of 16 MWt, gives a 100% NSSS power output of 3,666 MWt.All steam and power conversion equipment, including the turbine-generator, has the capability to generate a maximum calculated gross output of approximately 1,296 MWe. When the NSSS is operating at its warranted output of 3,666 MWt, the net electrical output is approximately 1,245 MWe.The project schedule was based on a fuel loading date of November 1, 1985, and an anticipated commercial operation date of Ma y 1, 1986. The Low Power License (5 percent) was issued by the NRC November 25, 1985, the Full Power License was issued January 31, 1986, and the Unit became commercially operational April 23, 1986. In 2001, Millstone Units 1, 2 and 3 operating licenses were transferred from Northeast Nuclear Energy Company to Dominion Nucl ear Connecticut, Inc. (DNC).DNC is an indirect wholly-owned subsidiary of Dominion Energy, which is in turn owned by Dominion Resources, Inc. (DRI). Virginia Power, which is the licensed owner and operator of the North Anna and Surry nuclear stati ons, is also a subsidiary of DRI.
MPS3 UFSAR1.1-2Rev. 30The transmission and distribution assets on the site will continue to be owned by Connecticut Light and Power (CL&P) and will be operated under an Inte rconnection Agreement between CL&P and DNC.The FSAR will retain references to Northeast Utilities and Northeast Nuclear Energy Company documents/activities when they are used in a historic context and are required to support the plant licensing bases.Upon license transfer, all records and design documents necessary for operation, maintenance, and decommissioning were transferred to DNC. Some of these drawings are included (or referenced) in this FSAR. These drawings often have title blocks (or drawing numbers) which list Northeast Nuclear Energy Company or Northeast Utilities Service Company (et. al). In general, no changes to these title blocks will be made at this time. Based on this general note, these drawings shall be read as if the title bl ocks list Dominion Nuclear Connecticut, Inc.
MPS3 UFSAR1.2-1Rev. 30
1.2 GENERAL
PLANT DESCRIPTIONThis section includes a summary description of the principal characteristics of the site and a concise description of Millstone 3.
1.2.1 GENERALMillstone
3 incorporates a four loop closed cycle pressurized water type nuclear steam supply system (NSSS); a turbine generator and electrical systems; engineered safety features; radioactive waste systems; fuel handling systems; structures and other on site facilities; instrumentation and control systems; and the necessary auxiliaries required for a complete and operable nuclear power station. The site plan (Figure 2.1-4) and the plot plan (Figure 1.2-2) show the general arrangement of the unit.
Piping and instrumentation diagrams are included throughout this document with the appropriate system descriptions. Symbols and abbreviations used in the diagrams are illustrated on Figure 1.2-3.With respect to the numbers, graphs, and drawings included within this report, the normal tolerance permitted by good engineering practice is intended. Where operating parameters are unusually important, such items have been included in the technical specifications.
1.2.2 SITEThe
site, approximately 500 acres in area, is on the north shore of Long Island Sound and on the east side of Niantic Bay. It is located in the Town of Waterfor d, Connecticut, about 3.2 miles west-southwest of New London and about 40 miles southeast of Hartford. The surrounding area is primarily residential with some commercial and industrial uses.Millstone 1 and 2 are also located on the site. Millstone 1 is a permanently defueled boiling water reactor. Millstone 2 uses a two-loop pressurized water reactor supplied by Combustion Engineering, Inc., with a rated thermal power level of 2,700 MW
- the architect-engineer was Bechtel Corporation. Section 2.1 contains a more detailed description of the site and surrounding areas.1.2.3 STRUCTURESMillstone 3 major structures are the containment structure, auxiliary building, fuel building (including decontamination facilities), waste disposal building, engineered safety features building, main steam valve buildi ng, turbine building, service build ing, control building, technical support center, emergency generator enclosure, containment enclosure building, warehouse 5 (including the condensate polishing waste treatment f acility), auxiliary boiler enclosure, and circulating and service water pumphouse. Section 3.8.4.1 describes the general arrangement of these structures.
MPS3 UFSAR1.2-2Rev. 30The reactor is operated inside a reinforced concrete containment structure maintained at a subatmospheric pressure between 10.6 and 14.0 psia.
The containment concept is similar to those of Surry Power Stations 1 and 2, North Anna Power Stations 1 and 2, and Beaver Valley Power Station 1. Following the loss-of-coolant accident (LOCA), described in Section 15.6.5, the containment remains above atmospheric pressure.
The containment structure is housed within the containment enclosure building, which along with structures adjacent to the containment, forms the boundary of the supplementary leak collection and release system (SLCRS). The SLCRS establishes a subatmospheric pressure in the containment enclosure building and contiguous structures. See Section 6.2.3 for a further description.The seismic criteria used in the design of the structures an d equipment for Millstone 3 are described in Section 3.7.
1.2.4 NUCLEAR
STEAM SUPPLY SYSTEMThe nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water-type reactor and four closed reactor coolant loops connected in parallel to the reactor vessel. Each loop contains a reactor coolant pump and a steam generator, two loop isolation valves, an isolation bypass valve, and a bypass line. The NSSS also contains an electrically heated pressurizer and auxiliary systems.High pressure water circulates through the reactor core to remove heat generated by the nuclear chain reaction. The heated water exits from the reactor vessel and passes via the coolant loop piping to the steam generators. Here, it releases heat to the feedwater to generate steam for the turbine generator. The cycle is completed when the water is pumped back to the reactor vessel. The entire coolant system is composed of leaktight components to ensure that all fluids are confined to the system.The reactor core is of the multi-region type. All fuel reactor assemblies are mechanically identical, although the fuel enrichment is not the same in all assemb lies. These assemblies incorporate the rod cluster control concept in canless 17 x 17 fuel rod assemblies using a spring clip grid to provide support for the fuel rods. The reactor moderator has a negative temperature coefficient of reactivity at full pow er at all times throughout core life.In the typical initial core loading, three fuel enrichments are used. Fuel assemblies with the highest enrichments are placed in the reactor core outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, one-third of the fuel is discharged from the central region and fresh fuel is loaded into the outer region of the reactor core. The remaining fuel is arranged in the ce ntral two-thirds of the reactor core in such a manner as to achieve optimum power distribution.
Rod cluster control assemblies are used for reactor control and consist of clusters of cylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the reactor core, each cluster of absorber rods is attached to a spider connector and drive shaft, which MPS3 UFSAR1.2-3Rev. 30is raised and lowered by a drive mechanism mounted on the reactor vessel head. Downward movement of the rod cluster control after trip is by gravity.
The reactor coolant pumps are vertical, single-stage, centrifugal pumps of the sh aft-seal type.The steam generators are Westinghouse Model F vertical U-tube units which contain Inconel tubes. The Model F steam genera tor includes features such as improved tube support plate design and high circulation ratio which are designed to minimize most forms of corrosion, sludge buildup, and chemical attack. Integral moisture separation equipment reduces the moisture of steam to one-quarter percent or less.All of the pressure containing and heat transfer surfaces in contact with reactor water are stainless steel clad or stainless steel except the steam gene rator tubes and fuel tube s, which are Inconel and Zircaloy, respectively. Reactor core internals, including control rod drive shafts, are primarily stainless steel.There are two double-disc, motor-operated, loop isolation valves in each loop, one located between the reactor vessel and the steam generator and the other between the reactor vessel and the reactor coolant pump of each of the four loops. The isolation bypass valves, also double-disc, motor-operated valves, are located in a bypass line connecting the two loop isolation valves in each loop.
An electrically heated pressuri zer connected to one reactor cool ant loop maintains reactor coolant system pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions. In addition, it provides indication of and maintains reactor coolant system water inventory.The auxiliary systems, provided as part of the NSSS, charge the reactor coolant system and add makeup water, purify reactor coolant water, provide chemical s for corrosion inhibition and reactivity control, remove decay heat when the reactor is shut down, and provide for emergency safety injection.
Other auxiliary systems supporting the NSSS but not part of the NSSS provide the following:1.System components cooling2.Fuel pool cooling3.Reactor coolant water and other auxiliary system fluid sampling4.Venting and draining the reactor coolant system and other auxiliary systems5.Emergency containment depressurizati on spray and combustible gas control6.Maintaining the containment atmosphere pressure at sub-atmospheric levels MPS3 UFSAR1.2-4Rev. 307.Collecting, processing, and dispos ing of liquid and gaseous wastes8.Preparation of soli d wastes for disposal9.Process, store, and supply r eactor coolant system boric acid10.Component ventilation11.Instrument and valve operator air
1.2.5 INSTRUMENTATION
AND CONTROL SYSTEMSThe instrumentation and control for the reactor protection system, engineered safety features actuation system, and other safety related systems meet the requirements of IEEE 279-1971, "Criteria for Protection System for Nuclear Power Generating System." In addition, other applicable criteria are met as described in Sections 3.1 and 7.1.2.
The nonsafety related instrumentation and controls accomplish reliable control and allow monitoring of the plant status without degradation of safety re lated instrumentation. Section 7.7 describes the design details.The reactor is controlled by a coordinated combination of chemical shim, mechanical control rods, and temperature coefficients of reactivity. The control system allows the unit to accept step load changes of 10 percent and ramp load changes of 5 percent per minute over the load range of 15 percent to 100 percent power under nominal ope rating conditions subject to xenon limitations.Control of the reactor and the turbine generator is accomplished from the control room, which contains all instrumentation a nd control equipment required fo r startup, operation, and shutdown, including normal and accident conditions. The turbin e generator controls are designed for manual operation; the operator selects the load setpoint and loading rate. The NSSS automatically follows the turbine generator, on decreasing power, from loads of 100 to 15 percent power. The operator takes manual action to match the NSSS to the turbine generator load, for load increases between 15% and 100% power. If, during rapid turbine generator loading (5 percent per minute), the response of the control rods and chemical shim is not adequate to supply the needed reactivity, the reactor coolant temperature automatically drops to supply more reactivity. If a reactor coolant low operating temperature is reached, turbine generator loading is stopped automatically.
1.2.6 RADIOACTIVE
WASTE SYSTEMSRadioactive wastes are collected, processed, and disposed of in a safe manner complying with appropriate regulations, in particular, NRC Regulations 10 CFR 20, 10 CFR 50 Appendix I, 10 CFR 61, 10 CFR 71, 49 CFR 171-178, 10 CFR 100, and General Design Criteria 60 and 64 (Sections 3.1.2.60 and 3.1.2.64). Th ere are three interrelated radio active waste treat ment systems: radioactive liquid waste, radioactive gaseous waste, and radioactive solid waste. Chapter 11 describes these systems.
MPS3 UFSAR1.2-5Rev. 30The radioactive liquid waste system (LWS) collects, classifies, and processe s all radioactive waste liquids generated during plant operation and refueling, either for recycle within the plant or for discharge off site. The process operations available to treat the liquid wastes are filtration and demineralization. The process descriptions and flow di agrams illustrate the number and sequence of processing steps to be applied to each type of liquid waste (Section 11.2).Gaseous wastes, consisting of hydrogen streams an d air streams containing various levels of radioactivity, are treated before release to the environment. Through the use of degasification and purification of reactor coolant letdown, the consequences of any reactor coolant leakage are minimized. This degasification and purification process produces hydrogen waste gas streams, which are passed through charcoal decay beds to provide adequate holdup time for the decay of noble gases and the removal of iodines. The decay beds are followed by high efficiency particulate air filters to ensure particulate removal. Aerated waste gas streams, produced by other phases of unit operation, are released to the environment via the Millstone stack. A process flow diagram, illustrating the processing steps for gaseous waste, appears in Section 11.3.
The radioactive solid waste system provides holdup, packaging, a nd storage facilities for the eventual off site shipment and ultimate disposal of solid radioactive waste material. Available process operations are: solidification of liquid wastes, holdup for decay, sluicing and dewatering of resins, encapsulation of miscellaneous solid wastes, and compacting of compressibles. Provisions for shielding during the processing and shipment of solid wastes are included in the design of the radioactive solid waste system. A process flow diagram illust rates the processing and handling sequences for solid wastes generated by the plant (Section 11.4).
1.2.7 FUEL HANDLINGThe reactor is refueled by equipment designed to handle spent fuel under water from the time the spent fuel leaves the reactor vessel until the spent fuel is placed in a cask for shipment from the site. Underwater transfer of spent fuel provides an optically transparent radiation shield as well as a reliable source of coolant for the removal of decay heat produced by the spent fuel. New fuel is transferred to the fuel pool using the new fuel handling crane and is loaded into the reactor using the same equipment that handl es the spent fuel. (Section 9.1.4)
1.2.8 TURBINE
GENERATOR AND AUXILIARIESThe turbine is a tandem-compound, six-flow, 1,800 rpm unit with 43 inch last stage blades. Two combination moisture separator-reheaters remove moisture and superheat the steam between the high and low pressure turbines. The turbine gene rator is discussed, in detail, in Section 10.2.The turbine generator plant and associated steam and power conversion systems are capable of a 40 percent load rejection without producing a reactor trip. This is accomplished by dumping steam into the condenser through the turbine bypass system which reduces the transient to within the NSSS transient response capability.The turbine control system is an electrohydraulic control (EHC) system capable of remote manual or automatic control of acceleration and loading of the unit at preset rates, and holding speed and MPS3 UFSAR1.2-6Rev. 30load at a preset level. The EHC provides a normal overspeed protection system and an emergency overspeed protection system to limit turbine overspeed.A single-pass deaerating surface condenser installed in three sections , two 100 percent design capacity steam jet air ejector units, three 50 percent design capacity condensate pumps, three steam generator feedwater pumps (two turbine-driven and one motor-driven), two 50 percent design capacity motor-driven and one 100 percent design capacity turbine-dr iven steam generator auxiliary feedwater pumps, three trains of feedwater heaters, each having six stages, and a full flow condensate demineralizer polishing system are provided. Any combination of two steam generator feedwater pumps is adequate to support 100% power operation. Condenser circulating water is provided by six circulating water pumps. The turbine plant component cooling system provides cooling water for lubricating oil coolers, generator hydrogen coolers, and other turbine plant auxiliary heat loads.
1.2.9 ELECTRICAL
SYSTEMSThe electric power system includes the electrical equipment and connections required to generate and deliver electric power to the 345 kV system. This system also incl udes station service electrical equipment to provide electric power to support st ation auxiliaries during normal operation, startup, shutdown, and accident conditions.The major component in the system is the turbine-driven main generator. The electric power output from this generator is stepped up in a transformer bank and delivered to the 345 kV switchyard for distribution to the utility grid.
The station service equipment consists of switchgear, load centers, motor control centers, ac vital and nonvital buses, and battery systems. The normal source of station service power is provided by the main generator through nor mal station service transformers. Startup and shutdown station service power is provided by a preferred off site source from the 345 kV switchyard through the main and normal station service transformers with the generator breaker open, or by the alternative off site source from the 345 kV switchyard through the reserve station service transformers. In the event of an accident with loss of both normal and off site sources, an on site emergency power system, consisting of two redundant diesel engi ne-driven generators, provides power to the emergency 4,160V buses within 11 seconds after receiving a start signal. These diesel engine-driven generators are sized to provide required power to all safety related equipment.
1.2.10 ENGINEERED SAFETY FEATURESEngineered safety features (ESF) are provided to mitigate the consequences of postulated accidents including a loss-of-coolant accident (LOCA) resulting from large and small pipe breaks.
The ESF systems provided for Millstone 3 have sufficient redundancy and independence of components and power sources so that, under the conditions of the postulated accident, the MPS3 UFSAR1.2-7Rev. 30systems maintain the integrity of the containm ent structure and accomplish the following even when operating with a postulated single active failure:1.Prevent radiation release to the outside environment from exceeding the limit specified in 10 CFR 50.67.2.Provide core cooling to pr event excessive metal-water reaction, to limit the core thermal transient, and to maintain the core in a coolable geometry.Millstone 3 is independent of Millstone 1 and 2 with respect to ESF. The systems provided for Millstone 3 are summarized below.Containment Structure The steel lined reinforced concrete containment structure provides a barrier against the escape of fission products. It is designed to operate at approximately atmospheric pressure and can withstand the pressures and temperatures resulting from a spectrum of LOCAs and secondary system breaks. The containments response following the accident is similar to other atmospheric PWRs. The structure and all penetrations, in cluding access openings, are of proven design (Section 6.2.1).Emergency Core Cooling System The emergency core cooling system (ECCS) provides borated water to cool the reactor core following a major LOCA. This is accomplished by the automatic injection of water from the safety injection accumulators into the reactor coolant loops and by the automatic pumping of a portion of the refueling water storage tank contents into the loops via the charging pumps, the safety injection pumps, and the residual heat removal pumps. After the injection mode of emergency core cooling, long term core cooling is maintained by recirculating the water from the containment structure sump by the containmen t recirculation pumps, through the containment recirculation coolers, and into the reactor coolant loops directly and via the charging and safety injection pumps (Section 6.3).
Containment Heat Removal System The containment heat removal system consists of the quench spray and the containment recirculation systems. Following the postulated DBA, the containment pressure is reduced by employing both systems.
The quench spray system sprays borated water from the refueling water storage tank (RWST).The recirculation spray system draws suction from the containment sump, the content of which consists of the primary or secondary system effluent and the quench spray. The sump water pH is controlled to be above 7.0 to improve effectiveness of fission products removal (Section 6.5.2), and for mate rials compatibility (Section 6.1.1).
MPS3 UFSAR1.2-8Rev. 30The pH is controlled by the dissolution of trisodium phosphate crystals (stored in baskets) in the containment sump water.
Supplementary Leak Collection and Release System Following a postulated accident, particulate and gaseous radioactive material is ducted from the containment enclosure structure and the buildings contiguous to the containment structure to the supplementary leak collection and release system (SLCRS), where it is filtered and discharged to the atmosphere through an elevated stack rather than through a ground-level vent. SLCRS is not credited for a postulated fuel handling accident.
Containment Isolation System The containment isolation system isolates piping lines which penetrate the containment boundary so that, in the event of a LOCA, radioactivity is not released to the environment through these lines.The lines are either isolated passively (check valves or locked closed manual valves) or isolated automatically by receipt of a safety injection signal (SIS), a containment isolation (phase A and B) signal, or a steamline isol ation (SLI) signal (Section 6.2.4).Engineered Safety Features Actuation System The engineered safety features actuation system (ESFAS) monitors selected parameters and determines whether predetermined safety limits have been exceeded. If they have been exceeded, the ESFAS initiates action to mitigate the abnormal occurrence (Chapter 7).
Habitability Systems The Millstone 3 control room envelope is equipped with an intake isolation system designed to protect the plant operators from the presence of hazardous substances outside the control room envelope.Control room envelope makeup air is supplied via redundant air filtration trains designed to meet the requirements of Regulat ory Guide 1.52 (Section 6.4).Inservice Inspection All ASME Section III, Code Class 1, 2, and 3 systems and components which require inservice inspection and testing are designed, fabricated, and erected to meet the inspection requirements of ASME Section XI (except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a). The inservice inspection program includes baseline preservice examination and periodic inservice inspection and testing to ensure the operability and integrity of all systems classified Class 1, 2, and 3 pursuant to 10 CFR 50.55a (Section 5.2.4 and 6.6).
MPS3 UFSAR1.2-9Rev. 301.2.11 COOLING WATER AND OTHER AUXILIARY SYSTEMS Cooling water and other auxiliary systems provided in Millstone 3 are outlined below. Their design criteria and details are described in Chapters 9, 10, and 11.1.The chemical and volume control syst em performs the following functions:a.Fills the reactor coolant system.b.Provides a source of high pressure water for pressurizing the reactor coolant system when cold.c.Maintains the water level in the pressurizer.d.Reduces the concentration of corrosi on and fission products in the reactor coolant.e.Adjusts the boric acid concentration of the reactor coolant for chemical shim control.f.Provides high pressure seal wate r for the reactor coolant pump seals.2.The boron recovery system concentrates and stores borated radioactive water from reactor coolant letdown (chemical and volume control system) processed through the gaseous waste system. Processing by an evaporator, ion exchanger, filters, and demineralizers in the boron recovery sy stem is capable of producing primary-grade water and concentrated boric acid solution for st ation reuse or disposal.3.Radioactive fluid degasification, liquid c oncentration, and waste solidification for disposal are provided by the radioactive ga seous waste, radioactive liquid waste, and radioactive solid waste systems.4.The service water system pr ovides cooling water for heat removal from the reactor plant auxiliary systems during all modes of operation and from the turbine plant auxiliary systems during normal operation.5.The reactor plant component cooling wa ter system, an intermediate cooling system, transfers heat from systems containing reactor coolant or other radioactive or potentially radioactive liquids to the service water sy stem, and provi des a source of safety grade cooling water to sy stems which have this requirement.6.The turbine plant component cooling system, also an intermediate cooling system, transfers heat from various turbine plant equipment to the service water system.
MPS3 UFSAR1.2-10Rev. 307.The fuel pool cooling and purification sy stem removes residual heat from spent fuel stored in the spent fuel pool and purif ies the water in the refueling cavity and spent fuel pool.8.The reactor plant vent and drain systems collect potentially radioactive fluids and gases from various plant systems and tran sfer these fluids and gases to the boron recovery system or to the appr opriate waste disposal system.9.Individual ventilation sy stems are provided for th e containment and other structures. The containment ventilation system recirculates a nd cools containment air. The ventilation and air-conditioning system servicing the main control room provides uninterrupted service, even under accident conditions.10.The circulating water system removes h eat resulting from the operation of the turbine generator and main condensers. Th e service water system provides cooling water from the ultimate heat sink for systems and components which require an ensured supply of cooling water under all conditions.11.The domestic water system receives water from the Town of Waterford, Conn., public water system and distributes it th roughout the unit for power plant system makeup and potable water needs.12.The compressed air systems consist of the service air system, instrument air system, and containment air system. Drye rs are provided in the instrument air system and the containment instrument air system.13.The fire protection system furnishes th e capacity to extinguish any probable fires which might occur at Millstone 3. Th e system includes a water system, a CO 2 system, a halon system, and portable fire extinguishers.14.The reactor and turbine plant sampling system have the capability for sampling all normal process systems and prin cipal components listed in Tables 9.3-1 and 9.3-2 for laboratory analysis.15.The Auxiliary Steam System is designed to supply steam for building heating and freeze protection for outdoor water stor age tanks and to carry condensate from various heating and processing equipment a ssociated with both Unit 2 and Unit 3 during normal plant operations. The sy stem is nonnuclear safety (NNS).
1.2.12 REFERENCE FOR SECTION 1.2 MPS3 UFSAR1.2-11Rev. 30 FIGURE 1.2-1 NOT USED MPS3 UFSAR1.2-12Rev. 30Withheld under 10 CFR 2.390 (d)(1)
FIGURE 1.2-2 PLOT PLAN
MPS3 UFSAR1.2-13Rev. 30 FIGURE 1.2-3 (SHEETS 1-3) P&ID LEGEND The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.
MPS3 UFSAR1.3-1Rev. 30
1.3 COMPARISON
TABLES
1.3.1 COMPARISON
WITH SIMILAR FACILITY DESIGNSPrincipal features of the design of Millstone 3 at the time of application for an operating license were similar to those that were evaluated and approved by the NRC staff for other reactors under construction, operation, or review. Comparison of notable similarities and differences to North Anna 1 and 2, Surry 1 and 2, Comanche Peak 1 and 2, W. B. McGuire 1 and 2, Maine Yankee, and Trojan was provided in a series of comparison tables in this section. This section is retained for historical purposes.The design of this facility was based on proven technology attained during the development, design, construction, and opera tion of pressurized water reactors of similar or identical types. The data, performance characteristics, and other info rmation represented a standard design that was engineered for the particular requirements of the utility system and site characteristics.
1.3.1.1 Comparison of Nuclear Steam Supply SystemA design comparison of major parameters or features of Millstone 3 with similar plants was presented in Table 1.3-1. The following plants were used in the comparison:1.Comanche Peak Units 1 and 2, Docket Number 55-445, -4462.W.B. McGuire Units 1 and 2, Docket Number 50-369, -3703.Trojan, Docket Number 50-344 1.3.1.2 Comparison of Engineered Safety FeaturesTable 1.3-2 compared the design data for the Millstone 3 engineered safety features (ESF) systems with similar systems in Trojan (for NSSS scope of supply - ECCS), and North Anna Power Station Units 1 and 2 (for balance of plant). The ESF systems compared were the emergency core cooling, containment depressurization, hydrogen recombiner, and supplementary leak collection and release systems.These units were chosen for comparison because Millstone 3 systems were similar in design. Trojan and North Anna 1 and 2 obtained operating licenses and are currently in commercial operation.
1.3.1.3 Comparison of Containment ConceptsTable 1.3-3 summarized the design and operating parameters described in Section 6.2.1 for the Millstone 3 containment concept. The table provided a comparison with similar data for subatmospheric containments from the FSAR for North Anna 1 and 2, and the FSAR for Surry 1 and 2. These references were selected because the units used the Stone & Webster subatmospheric containment design. Surry 1 and 2 and North Anna 1 and 2 are in commercial operation.
MPS3 UFSAR1.3-2Rev. 30 1.3.1.4 Comparison of Instrumentation SystemsTables 1.3-4 through 1.3-10 compared Millstone 3 instrumentation with similar instrumentation in North Anna 1 and 2. These units were chosen for the comparison because they have instrumentation similar in design to that of Millstone 3, they have obtained operating licenses, and are currently in commercial operation.
1.3.1.5 Comparison of Electrical SystemsTable 1.3-11 compared the Millstone 3 electrical system parameters with similar systems of Millstone 2 and Maine Yankee Atomic Power Station. The electrical systems compared were the 345 kV transmission, ac power, ac vital bus, 125 volt DC, and emergency power systems. These units were chosen for the comparison because they have electrical systems similar in design to Millstone 3. They have obtained operating licenses, and they are currently in commercial operation.
1.3.1.6 Comparison of Radioactive Waste SystemsTables 1.3-12, 1.3-13, and 1.3-14 compared the radioactive waste systems features for Millstone 3 with similar systems in North Anna 1 and 2 and Surry 1 and 2. These units were chosen for the comparison, because they had radioactive waste systems similar in design to Millstone 3, have obtained operating licenses, and are currently in commercial operation.
1.3.1.7 Comparison of Other Reactor Plant SystemsTable 1.3-15 summarized the final design and operating parameters for the major reactor plant systems. This table compared the Millstone 3 data with systems data in similar nuclear power plants (North Anna 1 and 2).
1.3.2 COMPARISON
OF FINAL AND PRELIMINARY DESIGNSTable 1.3-16 detailed the significant design changes that were made since the submittal of the PSAR, through the issuance of the operating license.
MPS3 UFSARMPS3 UFSAR1.3-3Rev. 30TABLE 1.3-1 DESIGN COMPARISON Parameter or FeatureMillstone 3 FSAR Chapter/SectionMillstone 3Comanche PeakW. B. McGuireTrojanReactor Core Heat Output (MWt)4.0, 5.0, 15.03,4113,4113,4113,411 Minimum DNBR for Design Transients4.1, 4.4, 15.0
> 1.30> 1.30> 1.30> 1.30Total Thermal Flow Rate (10 lb/hr)4.1, 4.4, 5.1140.8140.3140.3132.7 Reactor Coolant Temperatures (°F)4.1, 4.4Core Outlet620.6620.8620.8619.5 Vessel Outlet617.2618.2618.2616.8 Core Average590.5589.4589.4585.9 Vessel Average587.1588.2588.2584.7 Core Inlet557.0558.1558.1552.5 Vessel Inlet557.0558.1558.1552.5 Average Linear Power (kW/ft)4.1, 4.45.445.445.445.44
Peak Linear Power for Normal Operation (kW/ft)4.1, 4.412.612.612.613.6Heat Flux Hot Channel Factor, F Q4.1, 4.4 15.02.322.322.322.50Fuel Assembly Array4.1, 4.317 x 1717 x 1717 x 1717 x 17Number of Fuel Assemblies4.1, 4.3193193193193Uranium Dioxide Rods Per Assembly4.1, 4.3264264264264 Fuel Weight as Uranium Dioxide (lb)4.1, 4.3222,739222,739222,739222,739 MPS3 UFSARMPS3 UFSAR1.3-4Rev. 30Number of Grids Per Assembly4.1, 4.38-Type R8-Type R8-Type R8-Type RRod Cluster Control Assemblies4.1, 4.3Number of Full/Part Length61/-53/-53/853/8 Absorber MaterialHfHfAg-In-Cd/B 4C *Ag-In-CdClad MaterialSS **SSSSSSClad Thickness (inches)0.01850.01850.01850.0185Equivalent Core Diameter (inches4.1, 4.3132.7132.7132.7132.7 Active Fuel Length (inches)4.1, 4.3144143.7143.7143.7 Fuel Enrichment (Weight Percent)4.1, 4.3Unit 1 Unit 2 Region 12.401.601.402.102.10 Region 22.902.402.102.602.60 Region 33.403.102.903.103.10Number of Coolant Loops5.04444 Total Steam Flow (10 lb/hr)5.115.0515.1415.1415.07 Reactor Vessel5.3Inside Diameter (inches)173173173173
Inlet Nozzle Inside Diameter (inches)27.527.527-1/227.5Outlet Nozzle Inside Diameter (inches)29292929TABLE 1.3-1 DESIGN COMPARISON (CONTINUED)Parameter or FeatureMillstone 3 FSAR Chapter/SectionMillstone 3Comanche PeakW. B. McGuireTrojan MPS3 UFSARMPS3 UFSAR1.3-5Rev. 30Number of Reactor Closure Head Studs54545454Reactor Coolant Pumps5.4.1Horsepower7,0007,0007,0006,000 Capacity (gpm)100,40099,00099,00088,500 Head (feet)289288288277Steam Generators5.4.2ModelFDD51Heat Transfer Area (ft 2)55,00048,30048,00051,500Number of U-Tubes5,6264,5784,6743,388Residual Heat Removal5.4.7Initiation Pressure (psig)425425425400 Initiation/Completion Temperature (°F)350/120350/140350/140350/140Component Cooling Water Design Temperature
(°F)951059595Cooldown Time After Initiation (hr)20161616Heat Exchanger Removal Capacity (10 Btu/hr)35.2739.134.1534.2Pressurizer5.4.10 Heatup Rate Using Heaters (°F/hr)55555555TABLE 1.3-1 DESIGN COMPARISON (CONTINUED)Parameter or FeatureMillstone 3 FSAR Chapter/SectionMillstone 3Comanche PeakW. B. McGuireTrojan MPS3 UFSARMPS3 UFSAR1.3-6Rev. 30Internal Volume (ft 3)1,8001,8001,8001,800Pressurizer Safety Valves5.4.13Number3333Maximum Relieving Capacity (lb/hr)420,000420,000420,000420,000 Accumulators6.3Number4444 Operating Pressure, Minimum (psig)600600600600Minimum Operating Water Volume Each (ft 3)950950950870Centrifugal Charging Pumps6.3Number3222 Design Flow (gpm)150150150150 Design Head (feet)5,8005,8005,8005,800Safety Injection Pumps6.3 Number2222 Design Flow (gpm)425425425425 Design Head (feet)2,6802,6802,5002,500 Residual Heat Removal Pump5.4.7, 6.3TABLE 1.3-1 DESIGN COMPARISON (CONTINUED)Parameter or FeatureMillstone 3 FSAR Chapter/SectionMillstone 3Comanche PeakW. B. McGuireTrojan MPS3 UFSARMPS3 UFSAR1.3-7Rev. 30NOTE: *The Ag-In-Cd on Unit 1, B 4 C on Unit 2**SS = Stainless steel***The instrumentation and control systems discussed in Chapter 7 for Millstone 3 are functiona lly similar to those systems implemented in Comanche Peak, W. B. McGuire, and Trojan.Number2222Design Flow (gpm)4,0003,8003,0003,000 Design Head (feet)350350375375Instrumentation and Controls7.0************
Chemical and Volume Control9.3.4Total Seal Water Supply Flow Rate, Nominal (gpm)32323232Total Seal Water Return Flow Rate, Nominal (gpm)12121212Letdown Flow, Normal/Maximum (gpm)75/12075/12075/12075/120Charging Flow, Normal/Maximum (gpm)55/10055/10055/10055/100TABLE 1.3-1 DESIGN COMPARISON (CONTINUED)Parameter or FeatureMillstone 3 FSAR Chapter/SectionMillstone 3Comanche PeakW. B. McGuireTrojan MPS3 UFSAR1.3-8Rev. 30TABLE 1.3-2 COMPARISON OF ENGINEERED SAFETY FEATURES Emergency Core Cooling System (Section 6.3)Millstone 3 Trojan Charging Pump (used for high pressure safety injection)Number32 Design capacity each (gpm)150150Design total developed head (feet)5,8005,800Safety Injection Pump (used for intermediate pressure safety injection)Number22Design capacity each (gpm)425425 Design total developed head (feet)2,5002,500 Residual Heat Removal pump (used for low pressure safety injection)Number22Design capacity each (gpm)4,0003,000Design total developed head (feet)350375 Safety Injection Accumulator Number44Total volume each (ft 3)1,3501,350Water volume (ft 3 , minimum)950870Operating pressure (psig, minimum)600600Containment Depressurization Systems (Section 6.2.2)
Millstone 3 North Anna 1 and 2 Quench Spray PumpNumber22 Design capacity each (gpm)4,0002,000Design total developed head (feet)291265 Containment Recirculation Pump MPS3 UFSAR1.3-9Rev. 30Number4 outside containment 2 out / 2 in aDesign capacity each (gpm)3,9503,700/3,300Design total developed head (feet)342287/269 Millstone 3 North Anna 1 and 2 Containment Recirculation Cooler Number44UA per cooler (Btu/hr-
°F)3.865 x 10 6 3.79 x 10 6Recirculation flow (gpm)3,9503,500Service water flow (gpm)6,5004,500Refueling Water Storage TankVolume (gal)1,206,556480,000Temperature (°F, maximum)7550Hydrogen Recombiner System (Section 6.2.5)Number22Flow rate (scfm, each)5050Supplementary Leak Collect ion and Release System (Section 6.2.3)Number of filter trains2None Flow rate (scfm, each)9,700None a.Out - outside containment In- inside containmentTABLE 1.3-2 COMPARISON OF ENGINEERED SAFETY FEATURES (CONTINUED)
MPS3 UFSAR1.3-10Rev. 30TABLE 1.3-3 COMPARISON OF CONT AINMENT CONCEPTS (Section 6.2.1)
Millstone 3North Anna 1 and 2 Surry 1 and 2TypeSubatmosphericSubatmosphericSubatmosphericID (feet)140126126Overall height (feet)200191173 Free volume (ft 3)2.3 x 10 6 1.825 x 10 6 1.73 x 10 6Maximum design pressure (psig) 454545 Design temperature (°F)280280280Calculated peak pressure psig)38.4944.1 aa.Based on Tagami condensing heat transfer coefficient 44.98 b b.Based on Uchida condensing heat transfer coefficientReactor Coolant System:
Liquid volume (including pressurizer) (ft 3)11,6959,8749,874Temperature (mass average) (°F)583.5586.8574.5 Concrete Thickness:Vertical wall4 feet 6 inches4 feet 6 inches4 feet 6 inchesDome2 feet 6 inches2 feet 6 inches2feet 6 inches Containment structure leak rate
(%/day)0.30 (0-24 hrs)0.10.10.15 (24-720 hrs)0.00.0 MPS3 UFSAR1.3-11Rev. 30TABLE 1.3-4 COMPARISON OF CONTAINMENT ATMOSPHERE PRESSURE SENSOR PARAMETERS A a.The numbers stated for components or syst em performance do not represent the maximum/minimum acceptable or required values to support system operation. Setpoints are stated in the Technical Specifications.Pressure SensorsMillstone 3North Anna 1 and 2 Containment Atmosphere High Pressure Transmitter (Millstone 3, High-1)Number of channels33Logic matrix2/32/3 Approximate setpoint (psia)19.715.0 Containment Atmosphere Interm ediate High-High Pressure Transmitter (Millstone 3; High-2)Number of channels33Logic matrixÅ2/32/3 Approximate setpoint (psia)19.720 Containment Atmosphere High-High Pressure Transmitter (Millstone 3; High-3)Number of channels44Logic matrix2/42/4 Approximate setpoint (psia)24.725.0 MPS3 UFSAR1.3-12Rev. 30TABLE 1.3-5 COMPARISON OF REACTOR COOLANT PUMP BUS PROTECTION A a.The numbers stated for components or syst em performance do not represent the maximum/minimum acceptable or required valves to support system operation. Settings are stated in the Technical Specifications.Millstone 3North Anna 1 and 2 Undervoltage Reactor Coolant Pumps BusesNumber of channelsN/A3 Logic matrixN/A2/3Approximate Setting (V)N/A70% of 4,160 (2,912)Underfrequency Reactor Coolant Pumps Buses Number of channelsN/A3 Logic matrixN/A2/3Approximate Setting (Hz)N/A54-59 Reactor Coolant Pump Shaft Low-Low SpeedNumber of channels4N/ALogic matrix2/4N/AApproximate Setting (rpm)(later)N/A Reactor Coolant Pump Shaft Low SpeedNumber of channelsN/AN/ALogic matrixN/AN/A Approximate Setting (rpm)N/AN/A MPS3 UFSAR1.3-13Rev. 30TABLE 1.3-6 COMPARISON OF ENGINEERED SAFETY FEATURE ACTUATION SIGNALS Signal ActuationMillstone 3North Anna 1 and 2 Safety Injection Signal (SIS)
Low pressurizer pressure coincident with low pressurizer levelNoNoLow pressurizer pressureYesYesHigh main steam line differential pressureNo Yes Low steam line pressureYesNo High main steam flow coincident with low main steam pressure or low temperature averageNoYesHigh containment atmosphere pressureYesYesManual initiationYesYes Containment Isolation Phase A (CIA) Signal Safety injection signal (SIS)YesYes Manual initiationYesYes Containment Isolation Phase B (CIB) Signal or Containment Depre ssurization Actuation (CDA) Signal High-High containment atmosphere pressure (Millstone High-3)YesYesManual initiationYesYesSteam Line Isolation Signal High main steam flow coincident with low steam pressure or low main reactor coolant temperature averageNoYesHigh-High containment pressureNoNo Intermediate High-High containment atmosphere pressure (Millstone; High-2)YesYesHigh steam pressure rateYesNoLow steamline pressureYesNo Manual initiationYesYes MPS3 UFSAR1.3-14Rev. 30NOTE: The numbers stated for components or system performance do not represent the maximum/minimum acceptable or required va lues to support system operation.TABLE 1.3-7 COMPARISON OF EMERGENCY GENERATOR AND STEAM GENERATOR AUXILIARY FEEDWATER PUMP START SIGNALS Component Millstone 3 North Anna 1 and 2 Emergency generator auto start signalsEmergency bus under-voltageEmergency bus under-voltageSafety injection signal (SIS)SISSteam generator auxiliary feedwater pump (motor-driven)All steam generator feedwater pumps tripped 2/4 - lo-lo level trip any steam generator (1/4 matrix) coincident with reactor coolant
loop cold leg stop valve open 2/3 - lo-lo level trip any steam generator (2/3 matrix) coincident with reactor coolant loop hot leg stop valve open or reactor coolant loop cold leg stop valve openSequenced safeguards signalU ndervoltage reserve station service powerSISSISSteam generator auxiliary feedwater pump (turbine-driven) auto start signals 2/3 - undervoltage on station service bus 2/4 - steam generators lo-lo level (2/4 matrix) coincident with reactor coolant loop cold
leg stop valve open 2/3 - steam generator lo-lo level trip (2/3 matrix) coincident with reactor coolant loop hot leg stop valve open or reactor coolant loop cold leg stop valve open MPS3 UFSAR1.3-15Rev. 30TABLE 1.3-8 COMPARISON OF PR OCESS AND EFFLUE NT RADIATION MONITORING SYSTEMSMonitor Number of LocationsMillstone 3North Anna 1 and 2Aerated vent particulate01Aerated vent gas01 Ventilation vent particulate01Ventilation vent gas11Ventilation vent high range (particulate and gas)10 Hydrogenated vent1N/ASupplementary leak collection10Condenser air ejector11
Containment recirculation cooler service water outlet24 Component cooling heat exchanger service water discharge 01Liquid waste11Steam generator blowdown sample13Auxiliary condensate10 Turbine building floor drains10Reactor plant component cooling water subsystem11Reactor Coolant Letdown: Gross activity02
Reactor Coolant Letdown:
Specific fission product activity 00Circulating water discharge01Liquid waste evaporator01Service water discharge01Service water reservoir01 Control building inlet20 Regenerant evaporator (removed from service)10Waste neutralizing sump10Steam line monitors50 MPS3 UFSAR1.3-16Rev. 30TABLE 1.3-9 COMPARISON OF AREA RADIATION MONI TORING SYSTEMS Monitor Number of LocationsMillstone 3North Anna 1 and 2Containment structure low range41Containment structure high range41Manipulator crane11 Incore instrumentation transfer area11Decontamination area11New fuel storage area11 Spent fuel pool pit bridge and hoist11Auxiliary building control area01Sample room11 Control room11Laboratory (Service building)11Auxiliary building general area80
Equipment decontamination area (Service Building)10 MPS3 UFSAR1.3-17Rev. 30TABLE 1.3-10 COMPARISON OF AIRB ORNE RADIATION MONITORING SYSTEMS Number of LocationsMonitorMillstone 3North Anna 1 and 2 Auxiliary building lower levels particulate20Auxiliary building lower levels gas20
Auxiliary building upper levels particulate30Auxiliary building upper levels gas30Charging pumps cubicles particulate10 Charging pumps cubicles gas10Fuel building particulate10Fuel building gas10 ESF building particulate10ESF building gas10Waste building particulate10 Waste building gas10Control room particulate10Control room gas10 Containment structure particulate11Containment structure gas11Ventilation vent sample particulate a a.The ventilation vent sample particulate and gas monitors have the capability to take a sample from any one of eight different areas , some of which are equivalent to areas being monitored by separate Millstone 3 monitors.
11Ventilation vent sample gas11Leak collection area gas10 Leak collection area particulate10 MPS3 UFSARMPS3 UFSAR1.3-18Rev. 30TABLE 1.3-11 COMPARISON OF ELECTRICAL SYSTEM PARAMETERS Systems and Components Millstone 3 Millstone 2 Maine YankeeTRANSMISSION SYS TEM (SECTION 8.2)Transmission Lines to Circuits4 at 345 kV3 at 345 kV2 at 345 kV (Total for 3 Units)(Total for 2 Units)2 at 115 kV AC POWER SYSTEMS (SECTION 8.3.1)
Main transformer2 at 630 MVA1 at 945 MVA2 at 430 MVA Normal station service transformer1 at 50 MVA (6.9 kV)1 at 45 MVA1 at 30 MVA1 at 40 MVA (4.16 kV)1 at 20 MVAReserve station service transformer1 at 50 MVA (6.9 kV)1 at 45 MVA1 at 30 MVA1 at 45 MVA (4.16 kV)1 at 20 MVAAC VITAL BUS SYSTEM (SECTION 8.3.1)
Distribution Cabinets644 Inverters4 at 25 kVA 4 at 15 kVA4 at 10 kVA1 at 30 kVA 1 at 60 kVA 125-V DC SYSTEM (SECTION 8.3.2)
Unit batteries (125V)2 at 1650 Ah2 at 750 Ah2 at 2300 Ah 2 at 2550 Ah1 at 1200 Ah2 at 1800 Ah 4 at 200 amp6 at 400 amp4 at 250 amp MPS3 UFSARMPS3 UFSAR1.3-19Rev. 30 2 at 50 amp 3 at 200 amp (installed spares)
EMERGENCY POWER SYSTEM (SECTION 8.3.1)
Emergency generator (continuous rating)2 at 4986 kW2 at 2750 kW2 at 2500 kW Emergency 4.16 kV Buses2 at 2000/3000 amp2 at 2000 amp2 at 1200 ampTABLE 1.3-11 COMPARISON OF ELECTRICAL SYSTEM PARAMETERS (CONTINUED)Systems and Components Millstone 3 Millstone 2 Maine Yankee MPS3 UFSARMPS3 UFSAR1.3-20Rev. 30TABLE 1.3-12 COMPARISON OF RADIOACTIVE LIQUID WASTE SYSTEMS Millstone 3North Anna 1 & 2Surry 1 and 2Type of Processing EvaporationYesYesYes DemineralizationYesYesYes FiltrationYesYesYes Treatment of Radioactive Wastes High Activity WastesEvaporat ion and/or demineralization and filtration, if requiredSameSameLow Activity Wastes Including Contaminated Shower Drains Filtration (evaporation and subsequent operations are optional)SameSameRegenerant Chemical WasteFiltration if requiredNoneNoneSteam Generator BlowdownBlowdown piped to flash tank where steam is drawn off to 4th point heater and liquid is drawn to main condenser for treatment by the
condensate demineralizersBlowdown is cooled and sent to clarifier for flocculation, sedimentation, and filtration.
The sediment is sent to the
solid waste disposal system and the liquid discharged through the liquid waste
monitor to the circulating discharge tunnel.
Blowdown is cooled and then either released or demineralized and filtered and recycled to the main
condenser Equipment High Level Waste Drain Tanks Number222 MPS3 UFSARMPS3 UFSAR1.3-21Rev. 30Capacity (gal/each)26,000 *5,0002,390Low Level Waste Drain Tanks 222Capacity (gal/ea)4,000 *5,0002,874Regenerant Evaporator Feed Tanks (removed from service)N/AN/ANumber2Capacity (gal/each)13,500Contaminated Shower Drain TankNumber022 Capacity (gal/each)1,400 (including Laundry)1,230 (including Laundry)Waste Test TankNumber222 Capacity (gal/each)24,0001,500548 Regenerant Chemical Evaporator (removed from service)N/AN/ANumber1Trays, number8 Capacity (gal/each)35Waste EvaporatorTABLE 1.3-12 COMPARISON OF RADIOACTIVE LIQUID WASTE SYSTEMS (CONTINUED)Millstone 3North Anna 1 & 2Surry 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-22Rev. 30NOTE: *Include condensate demineralizer, liquid wa ste system tank volumes. The numbers stat ed for components or systems performance do not represent the maximum/mini mum acceptable or required valu es to support system operation. Number111Trays, number800 Capacity (gal/each)3566Demineralizers NumberTwo waste evaporators, one regenerant chemical evaporator (removed from service)
One waste evaporator distillate polishing, two clarifier demineralizersOne waste evaporator distillate polishingTypeMixed bedMixed bedMixed bed Capacity (ft 3)3517/45 respectively17Effluent FiltersNumber221 Capacity (gpm/each)505075 Filter element typeWound fiberWound cotton fiberWound synthetic fiberTABLE 1.3-12 COMPARISON OF RADIOACTIVE LIQUID WASTE SYSTEMS (CONTINUED)Millstone 3North Anna 1 & 2Surry 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-23Rev. 30TABLE 1.3-13 COMPARISON OF RADIOACTIVE GASEOUS WASTE SYSTEMS Millstone 3North Anna 1 and 2Surry 1 and 2Type of Treatment DegasificationYesOccurs in boron recovery system Occurs in boron recovery systemDecay of noble gases in high activity gas streamYes YesYes Filtration of low activity gas streams recombinersGas streamsYesYes NoRecombinersStreamsYesYes NoTreatment of Streams Continuous degasification of reactor letdownYesYesYes Degasification of letdown to boron recovery systemYesYesYes Decay method for gases stripped in degasifier Adsorption on charcoal for minimum 60-day xenon
decay before recycle or release to the environmentRecombination of hydrogen stream for reduction storage
in gas decay tank, then charcoal filtration before release to atmosphereRecombination of hydrogen stream for reduction storage
in gas decay tanks before release to the environmentLow activity air streams (nonventilation streams)
Release through Millstone stack Filtration (charcoal and HEPA filters) and release thru vent on top of Unit 1
containment Filtration (charcoal and particulate filters) and release to environment MPS3 UFSARMPS3 UFSAR1.3-24Rev. 30 Equipment Degasifier Number121 (common to both units)
Capacity (gpm/ea)150240Process Gas CompressorNumber222 Capacity (scfm/ea)31.52.5 (Currently abandoned in place)Process Gas Charcoal AbsorbersNumber200Capacity - Charcoal (lb/ea)13,500--Gas Decay TanksNumber022 Capacity (ft 3/ea)-462434Pressure (psig)-115 - the inside tank115 4 - the outside tankGas Surge Tank (Process Gas Receiver)Number111 Capacity (ft 3)101515.7Waste Gas RecombinersTABLE 1.3-13 COMPARISON OF RADIOACTIVE GASEOUS WASTE SYSTEMS (CONTINUED)Millstone 3North Anna 1 and 2Surry 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-25Rev. 30 Number011Capacity (scfm)1.51.31Pressure (psia)1422Waste Gas CompressorsNumber022Capacity (scfm/ea) 01.51.5 Pressure (psig)150120Filter for low activity aerated activity aerated gaseous waste effluentsNoneActivated charcoal with HEPA after-filter Activated charcoal with HEPA after-filterTABLE 1.3-13 COMPARISON OF RADIOACTIVE GASEOUS WASTE SYSTEMS (CONTINUED)Millstone 3North Anna 1 and 2Surry 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-26Rev. 30TABLE 1.3-14 COMPARISON OF RADIOACTIVE SOLID WASTE SYSTEMS Millstone 3North Anna 1 and 2Surry 1 and 2Type of Treatment Solidification/DewateringYesYesYes DrummingYesYesYesSolidification agentCementUrea formaldehydeCementInputs (Type) Treated Boron evaporator bottomsYesYesYesWaste evaporator bottomsYesYesYes Regenerant evaporator bottomsEvaporator removed from serviceN/AN/AFiltersYesYesYesResinsYesYesYes
Miscellaneous wastes (contaminated clothing, tools, paper products, etc.)YesYesYes Equipment Spent resin hold tank: Number111 Capacity (gal)3,0001,8002,019 Spent resin dewatering tank: Number111 Capacity (gal)500500619 Spent resin transfer pump filter:
MPS3 UFSARMPS3 UFSAR1.3-27Rev. 30 Number111Capacity (gpm)150100150TypeWound fiberWound fiberWound fiber Evaporator bottoms tank:Number11N/A Capacity (gal)3,1501,100 Shipping container (with cask):Capacity 50 ft 3 50 ft 3 55 gallon drumType All solid waste is 50 ft 3 or greater shipping containers All solid waste is shipped in 50 ft 3 containers except compressible waste which is in 55 gallon drums All waste in 55 gallon drumsTABLE 1.3-14 COMPARISON OF RADIOACTIVE SOLID WASTE SYSTEMS (CONTINUED) Millstone 3North Anna 1 and 2Surry 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-28Rev. 30TABLE 1.3-15 COMPARISON OF OTHER REACTOR PLANT SYSTEMS Operating ParametersSystems with ComponentsMillstone 3North Anna 1 and 2 Fuel Pool Cooling and Puri fication System (Section 9.1.3)
Fuel Pool Cooling Pumps:Number22 Design capacity (gpm)3,5002,750 Design total head (ft)9280 Fuel Pool Coolers:Number22 Duty per heat exchanger (Btu/hr)27,700,00056,800,000 Fuel pool cooling flow (gpm)3,5002,750 Component cooling flow (gpm)1,8003,350 Number of cores cooled15-1/2 (3048 Fuel Assemblies)1-1/3 Fuel pool temperature, normal (°F)150140 Reactor Plant Component Cooling System (Section 9.2.2.1)
Reactor Plant Component Cooling Pumps:Number34 Design capacity (gpm)8,1008,000 Design total head (ft)284190 MPS3 UFSARMPS3 UFSAR1.3-29Rev. 30 Reactor Plant Component Cooling Heat Exchangers:Number32 Duty per unit exchanger (Btu/hr)76,000,00052,000,000
Reactor plant component cooling water flow (gpm)8,1009,000Service water flow (gpm)8,00010,500 System Design Basis (safety related)Reactor cooldown to 120
°F in 24 hrReactor cooldown to 140
°F in 16 hrDuty per unit exchanger (Btu/hr)76,000,00052,000,000
Reactor plant component cooling water flow (gpm)8,1009,000Service water flow (gpm)8,0000,500 System Design Basis (safety related)Reactor cooldown to 120
°F in 24 hrReactor cooldown to 140
°F in 16 hrService Water System (Section 9.2.1)
Service Water Pumps:Number46 Design capacity (gpm)15,00011,500 Design total head (ft)120127
Design temperature, maximum (°F)8095 Boron Recovery System (Section 9.3.2)
Type of Treatment:TABLE 1.3-15 COMPARISON OF OTHER REACTO R PLANT SYSTEMS (CONTINUED) Operating ParametersSystems with ComponentsMillstone 3North Anna 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-30Rev. 30 Degasification of liquid entering liquid entering system Occurs in radioactive gaseous waste systemYesStorage of liquid prior to processingYesYesEvaporationYes, in boron evaporatorYes, in boron evaporator
Demineralization of boron evaporator distillateYes, if requiredYes, if requiredSystem Effluents:
Boron evaporator distillatePrimarily recycled to primary grade water storage tank; remainder to radioactive liquid waste system for dischargePrimarily recycled to primary grade water storage tank; remainder to radioactive liquid waste system for dischargeBoron evaporator bottomsRecycl ed to boric acid tanks or drummed in radioactive solid waste system Recycled to boric acid tanks or drummed in radioactive solid waste system Cesium Removal Ion Exchanger:Number22 Resin Capacity (ft 3)3545 Boron recovery tanks:Number23 Capacity (gal)150,000120,000 Boron Evaporator Subsystem:TABLE 1.3-15 COMPARISON OF OTHER REACTO R PLANT SYSTEMS (CONTINUED) Operating ParametersSystems with ComponentsMillstone 3North Anna 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-31Rev. 30Evaporator typeForced circulationForced circulation Number12 (for 2 units)
ReboilerExternalExternal Trays85 Capacity (gpm)2520Boron Test Tanks:Number22 Capacity (gal) 12,00020,000 Boron Demineralizers:Number22Resin capacity (ft 3)3545TABLE 1.3-15 COMPARISON OF OTHER REACTO R PLANT SYSTEMS (CONTINUED) Operating ParametersSystems with ComponentsMillstone 3North Anna 1 and 2 MPS3 UFSARMPS3 UFSAR1.3-32Rev. 30TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION FSAR SectionSignificant Changes Since PSAR PSAR Reference2.5.3 Thirteen faults have been uncovered in the rock excavation mapping since November 1979.2.5.33.1.1.3 Application of the single fail ure criterion has been upgraded to be consistent with regulatory requirements and industry standards and practices.
3.1.1.23.3.2 Postulated tornado missiles and their characteristics have been revised to conform to Regulatory Guide 1.76 dated April 1974.
3.3.23.5.1.4 Conformance to Regulatory Guide 1.117, Rev. 1, dated April 1978. Have revised the spectrum of tornado generated missiles used for plant design.
3.5.1.43.7B.4.2 The number of seismic instru mentation packages has been increased and those locations revised.3.7.43.8.1 Structural ring added around the containment structure to maintain isolation from the surrounding rock.
3.8.2.13.8.4 1.Revised plan and design of waste disposal building to conform to Regulatory Guide 1.143, Rev. 1, dated October 1979.
3.8.22.Description of railroad canopy adjacent to fuel building added.3.8.13.11 Class IE electrical equipment has be en qualified to IEEE 323-1974 and IEEE 344-1975 requirements.3.115.2.4 Inservice methods have been changed from cl osed circuit television monitors to ultrasonic scanners in the reactor vessel.
5.2.85.2.5 Air cooler outlet temperature and reactor coolant system make up rates are no longer monitored.
5.2.75.4.2 Model D steam generators have been replaced by Model F steam generators.5.4.2 MPS3 UFSARMPS3 UFSAR1.3-33Rev. 305.4.7 Contents have been revised to include co ld shutdown requirements of Regulatory Guide 1.139, Rev. 0, date May 1978.
5.5.7, 6.3 6.36.2.1.7 Temperature monitoring for containment sump is not Q.A.
Category I as implied in PSAR Section 7.5.3.2.
7.5.3.26.2.6.2 Added description of electrical penetrations leakage rate tests.6.2.6.46.3Boron injection tank (BIT) has been eliminated.6.36.41.Pressurization system split into two banks of 100% capacity each.6.42.Deletion of vinyl chloride detection.3.Air-conditioning units which serv e areas outside the control r oom pressure envelope have been relocated from this control room area to outside the control ro om pressure envelope.6.4.6Only one smoke detector is provided on the control building ventilation inlet. The response to PSAR Question 9.57 stated redund ant detectors would be provided.
Response to PSAR Question 9.577.3.1,1.Change in pressurizer pressure required for actuation of safety injection signal (SIS).7.3.4.3, 1.3.1.4 1.3.1.42.Change in steamline and containment pressure required for actuation of steamline isolation signal.7.4 Millstone 3 design now has th e capability for a safety gr ade cold shutdown from the auxiliary shutdown panel.
7.4, 3.1.2.19 3.1.2.197.8 Addition of ATWS Mitigation System Actuation Circuitry, conformance to 10 CFR 50.62.NoneTABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-34Rev. 308.3.1 1.Millstone 2 electrical power tie has been deleted.8.3.12.Improvements in cable separation to comply with Regulatory Guide 1.75, Rev. 2, dated Sept. 1978.3.The capacity of the emergency generators has been increased from 4,300 kW to 4,986 kW.4.The emergency generator enclosure has been relocated closer to the control building.5.The second off site source of power has b een changed from having a capacity for one emergency system to having a capacity for both emergency systems and all nonsafety systems. The unit now has two independent off site sources each with the capacity for all emergency and normal loads during any circ umstances, and each source is now immediately (0.1 second or less) available. 6.Physical and electrical separation of nonsafety from safety circuits within safety related 4,160V switchgear, 480V load centers and mo tor control centers is no longer required.
Instead these nonsafety circuits are considered as safety circuits within this equipment and subsequently changed to nonsafety status af ter going through qualifi ed isolation devices.8.3.2 1.Addition of a second 125V battery and its as sociated equipment to the normal (nonsafety related) dc power system.
8.3.22.Compliance to Regulatory Guide 1.75, Rev. 2, dated Sept. 1978.9.1.2 Large increase in amount of spent fuel that is able to be stored.9.1.29.2.1 Addition of motor control center (MCC) and rod drive area air-conditioning (a/c) units as loads to the service water system (SWP). This addition requires two booster pumps to meet flow requirements to the a/c units.
9.2.19.2.2.1 1.The component cooling system no longer s upplies cooling water to the reactor coolant pump (RCP) motor air cooler and the control rod drive mechanism (CRDM) shroud coolers. Loads transferred to chilled water system.
9.2.2.1TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-35Rev. 302.Containment air recirculation coolers are supplied with chilled water except during a CIA or LOP when component cooling water is automa tically supplied. They were previously supplied with component cooling water duri ng winter conditions when there were no thermal bypass lines around the reactor plan t components cooling heat exchangers.3.One component cooling pump (CCP) can supply all four reactor coolant pump coolers, if only one CCP pump is available.9.2.2.2 1.Urea formaldehyde air-conditioning unit load deleted.9.2.2.22.RCS pump oil cooler, thermal barrier and bearing loads have been transferred to the reactor component cooling system.3.Added process vent cooler load.
4.Reactor plant component cooling water sy stem (CCP) - chilled water system (CDS) interaction.9.2.3 Ultrafiltration system (UF modules, cleaning solution skid, pH solution skid, cartridge filter, permeate tank, and pumps) have been added to the water treating system. UF membranes will remove all colloidal material and large organic molecules from the raw water. The ultrafiltered water is then forwar ded to the demineralizer system.
9.2.79.2.4 Hydropneumatic pumps deleted. Electric hot water in lieu of steam. Domestic water service throughout plant.
9.2.69.2.8 Addition of 200 gpm deaerator system to the primary water system to lower oxygen level in the primary water to comply with Westinghouse requirements.
9.2.49.3.1.1 Addition of two shutdown air compressors to improve plant operability following a loss of power.9.3.3.19.4.1 Air-conditioning units which serv e areas outside the control r oom pressure envelope have been relocated from this control room area to outside the control ro om pressure envelope.
9.4.1TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-36Rev. 309.4.1.5 1.The control building isolation signal (CBI) has been changed; it is no longer initiated by all safety injection signals.
9.4.1.52.Both trains of control building air-conditi oning do not start automati cally on receipt of a safety injection signal (SIS). Both trains receive a CBI signal; however, one train is normally running and the standby system will not start unless there is a fault in the running system. The CBI signal serves to bloc k a manual stop of the running system.Since Trains A and B air-conditioning units do not operate simultaneously, the units discharge into common ductwork designed to accommodate airflow from one train only. 3.The chilled water expansion tank level is alarmed on low level only, not high-low. The chilled water system was originally designed as a pressurized system and a high level alarm was necessary. However, the system is now ope n to the atmosphere. The expansion tank is manually filled and has an ove rflow connection. Therefore, a high level alarm is not required.9.4.2 1.Manual diversion of exhaust air systems to filtration units occu rs on receipt of high radioactivity or CIA signal.
9.4.52.Addition of safety related dampers in supply air ductwork.9.4.6 1.The emergency generator enclosure ventilation system uses supply fans, not exhaust fans.9.4.102.Fans are single speed instead of two-speed.3.Air in the emergency generator enclosure can now be recirculated.9.4.9 1.The auxiliary building filter system will filter the waste disposal buildi ng exhaust air instead of main filter bank of the SLCRS.
9.4.32.The air supply equipment for the waste dis posal building has been relocated from the auxiliary building to the roof of the waste disposal building.TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-37Rev. 303.The waste disposal building ventilation syst em has been redesigned to reflect the new building layout. The system now utilizes three 50-percent units rather than two 100-percent units.9.4.11 Addition of hydrogen recombiner heating, ventilation and air-conditioning system.None9.5.1 andChange in scope attributed to increased le vel of fire protection design and analysis as documented in the Millstone 3 Fire Pr otection Evaluation, June 1977 as amended.
9.5.1 Fire Protection
Evaluation (June 1977)9.5.4.5 1.Fuel oil storage tank: local fuel oil level indication delete
- d. (Remote indication on emergency generator panel).
9.5.4.52.Manual control changed from main control board to emergency generator panel.3.Fuel oil transfer pump discharge: Local pres sure indication deleted (local flow indication provided).10.3.2 1.MOVs have been replaced with AOVs in the steam supply system to the auxiliary feedwater pump turbine.
10.32.Addition of bypass valves around the ma in steam pressure relieving valves.10.4.3 The PSAR stated that the exhaust from the gl and seal condenser exhaust would be passed through a charcoal filter and released from the auxiliary boiler bl owdown vent stack. Now there is no charcoal filter and the release is from the turbine building roof.11.3, Figure 11.3.-1BTABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-38Rev. 3010.4.5 A circulating water pump is not automatically tripped in the event of low flow in the associated condenser discharge line coincident with high level in the condenser discharge strainer pit. Manual shutdown was selected as the alternative to automatic controls as a more dependable method to pr event flooding in the turbine building in the event of a condenser expansion joint failur
- e. Circulating water flows ar e not measured. High level in the condenser discharge pit will be alarmed in the control room to allow the operator to manually trip the circulating water pumps in the event of the postulated rupture.
Response to PSAR Questions 10.2, 10.810.4.8 1.System redesigned with a single blowdown tank and no longer contains a blowdown tank condenser.
10.4.62.Blowdown liquid drains to the steam generator blowdown tank. The steam generator blowdown tank drains to the c ondenser (closed cycle) or the circulating water discharge tunnel (open cycle).3.Blowdown lines are no longer isolated on a containment isolation Phase A (CIA) signal.4.Blowdown rate changed.10.4.9 1.Deletion of smart valves and addition of cavi tating venturies to limit flow to faulted steam generator.
10.4.32.Addition of service water sy stem supply to pump suctions.3.Auxiliary feedwater lines join main feedwater lines inside containment.11.2.3 Conformance to Regulatory Guide 1.112 dated April 19, 1976.11.2.511.4 The radwaste solidification system has been changed from an urea formaldehyde system to a Dow polymer solidification system.11.5.611.5 Addition of radiation monitors for post accident monitoring.11.414.2 Entirely rewritten to conform to Regulatory Guide 1.70, Rev. 3.14.2TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSARMPS3 UFSAR1.3-39Rev. 3017.1.2The Stone & Webster QA Program was re vised on March 31, 1975 to conform to NRC approved Stone & Webster Topical Repor t SWSQAP 1-74A, Revision N/A, dated December 31, 1974.
17.1.217.1.3The original QA Program implemented by WNES for Millstone 3 was described in Chapter 17 of the Millstone 3 PSAR. During the desi gn and initial procurement activities for Millstone 3, the upgrading of the WNES QA Program reflected changes in regulatory requirements and industry standa rds. These changes have culminated in the present WNES QA Program as presented in WCAP-8370, Revi sion 7A. This revision of the WNES QA Program applies to activities within the WNES scope performed for Millstone 3 which were initiated after January 1, 1975. Section 3.1.3 and WCAP 8370, Revision 7A, include WNES positions on regulatory guides for Millstone 3.
17.1.3TABLE 1.3-16 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (CONTINUED)FSAR SectionSignificant Changes Since PSAR PSAR Reference MPS3 UFSAR1.4-1Rev. 30
1.4 IDENTIFICATION
OF AGENTS AND CONTRACTORS
1.4.1 LICENSEE'S
SUBSIDIARIESDominion Nuclear Connecticut, Inc. (DNC) is responsible for the operation, maintenance, and testing of Millstone 3. DNC is an indirect wholly owned subsidiary of Dominion Energy, which is in turn wholly owned by Dominion Resources, Inc.
1.4.2 ARCHITECT-ENGINEERStone & Webster Engineering Corporation (SWEC) in Boston, Massachusetts, provided engineering design and construction management services for Millstone 3. SWEC is an engineering and construction firm serving the electric utility industry in the design and construction of all types of power stations.
1.4.3 NUCLEAR
STEAM SUPPLY SYSTEM MANUFACTURERWestinghouse Electric Corporation (Westinghouse) was responsible for supplying the NSSS and first fuel load for Millstone 3.Westinghouse has designed, developed, and manufactured nuclea r facilities since the 1950s, beginning with the world's first large central station nuclear power plan t (Shippingport), which has produced power since 1957. Completed or presently contracted commercial nuclear capacity totals in excess of 97,000 MW. Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Among the company's own related manufacturing facilities are the Columbia Plant, Nuclear Fuel Division, the largest commercial nuclear fuel fabrication facility in the world, the Pensacola Plant which fabricates reactor internals and steam generators, and the Cheswick Plant which produces control rod drive mech anisms and reactor coolant pumps.
1.4.4 TURBINE
GENERATOR MANUFACTURERThe turbine generator was manufactured by General Electric Company (GE). Design of the turbine generator was under the direction of the Steam Turbine-Generator Products Division located in Schenectady, New York.GE has extensive experience manufacturing turbine generators for nuclear and nonnuclear applications and has supplied them for 36 operating nuclear power plants. These include 11 pressurized water reactors (PWR), 24 boiling water reactors (BWR), and 1 high temperature gas reactor (HTGR). GE is also providing turbine generators for 36 nuclear power plants in various stages of construction. These include 19 PWRs and 17 BWRs.
MPS3 UFSAR1.5-1Rev. 30
1.5 REQUIREMENTS
FOR FURTHER TECHNICAL INFORMATIONThis section has been deleted in its entirety.
MPS3 UFSAR1.6-1Rev. 30
1.6 GENERAL
REFERENCES (HISTORICAL)Table 1.6-1 lists topical reports which provided information that was filed separately with the Nuclear Regulatory Commission (NRC) in s upport of this and similar applications. The information contained in Section 1.6 is retained for historical purposes. The review status codes previously included in this section were presented for information only and were accurate at the time of license application. Since the review status codes are not required by Regulatory Guide 1.70, are no longer applicable, and could lead to misinterpretation, they were removed in 1997 as part of a general FSAR upgrad e and clarification of information.
MPS3 UFSARMPS3 UFSAR1.6-2Rev. 30TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERENCES (HISTORICAL)
Report NumberTitle NRC SubmittalReference Section(s)WCAP-2048The Doppler Effect for a Non-Uniform Temperature Di stribution in Reactor Fuel ElementsJuly 19624.3 WCAP-2850 (Proprietary)
Single Phase Local Boiling and Bulk Boili ng Pressure Drop CorrelationsApril 19664.4 WCAP-7916 (Non-proprietary)
June 1972WCAP-2923In-Pile Measurement of U0 2 Thermal Conductivity19664.4 WCAP-3269-8Hydraulic Tests of the San Onofre Reactor ModelJune 19644.4 WCAP-3269-26LEOPARD - A Spectr um Dependent Non-Spatial De pletion Code for the IBM -
7094Sept 19634.3, 15.0, 15.4 WCAP-3385-56Saxton Core II Fuel Performance Evaluation, of Mass Spectrometric and Radio-Chemical Analyses of Irradiated Saxton Plutonium FuelJuly 19704.3, 4.4WCAP-3680-20Xenon-Induced Spatial Instabilities in Large PWRs (EURAEC-1974)March 19684.3 WCAP-3680-21Control Procedures for Xenon-Induced X-Y Instabilities in Large PWRs (EURAEC-2111)Feb 19694.3 WCAP-3680-22Xenon-Induced Spatial Instabilities in Three-Dimensions (EURAEC-2116)Sept 19694.3 WCAP-3696-8Pressurized Water Reactor pH - Reactivity Effect Final Report (EURAEC-2074)Oct 19684.3 WCAP-3726-1PU0 2 - U0 2 Fueled Critical ExperimentsJuly 19674.3 WCAP-6065Melting Point of Irradiated U0 2Feb 19654.4 MPS3 UFSARMPS3 UFSAR1.6-3Rev. 30WCAP-6069Burnup Physics of Heterogenous Reactor LatticesJune 19654.4 WCAP-6073LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOSApril 19664.3 WCAP-6086Supplementary Report on Evaluation of Mass Sp ectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through CuriumAug 19694.3 WCAP-7015 Revision 1Subchannel Thermal Analysis of Rod Bundle CoresJan 19694.4 WCAP-7048-PA (Proprietary)The PANDA CodeJan 19754.3 WCAP-7757-A (Non-proprietary)
WCAP-7198-L (Proprietary)
Evaluation of Protective Coatings for Use in Reactor ContainmentApril 19696.1N WCAP-7825 (Non-proprietary)
Dec 1971WCAP-7213-P-A (Proprietary) The TURTLE 24.0 Diffusion Depletion CodeFeb 19754.3 WCAP-7758-A (Non-proprietary) 15.0, 15.4 WCAP-7308-L (Proprietary)
Evaluation of Nuclear Hot Channel Factor UncertaintiesDec 19714.3 TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-4Rev. 30 WCAP-7810 (Non-proprietary)
WCAP-7359-L (Proprietary)
Application of THINC Program to PWR DesignAug 19694.4 WCAP-7838 (Non-proprietary)
Jan 1972 WCAP-7477-L (Proprietary) Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply SystemsMarch 19705.2 WCAP-7735 (Non-proprietary)
Aug 1971 WCAP-7488-L (Proprietary) Solid State Logic Protection System DescriptionMarch 19717.2 WCAP-7672 (Non-proprietary)June 19717.3 WCAP-7518-L (Proprietary)
Radiological Consequences of a Fuel Handling AccidentDec 197115.7 WCAP-7828 (Non-proprietary)
WCAP-7536-L (Proprietary) Seismic Testing of Electrical and Control Equipment (High Seismic Plants)Dec 19713.10N TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-5Rev. 30 WCAP-7821 (Non-proprietary)
Supplements 1-6WCAP-7558Seismic Vibration Testing with Sine BeatsOct 19713.10N WCAP-7588 Revision 1-A An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics MethodsJan 197515.4 WCAP-7623Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel PlateDec 19705.4 WCAP-7667-P-A (Proprietary) Interchannel Thermal Mixing With Mixing Vane GridsJan 19754.4 WCAP-7755-A (Non-proprietary)WCAP-7695-P-A (Proprietary) DNB Tests Results for New Mixing Vane Grids (R)Jan 19754.4 WCAP-7958-A (Non-proprietary)
WCAP-7695 Addendum 1-P-A (Proprietary) DNB Test Results for R Grid Thimble Cold Wall CellsJan 19754.4 WCAP-7958 Addendum 1-A (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-6Rev. 30 WCAP-7706-L (Proprietary) An Evaluation of Solid State Logic Reactor Protection in Anticipated TransientsFeb 19714.6 WCAP-7706 (Non-proprietary) 7.1 7.2 WCAP-7769Overpressure Protection for Westinghouse Pressurized Water ReactorsOct 197115.2 WCAP-7798-L (Proprietary) Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Accident EnvironmentJan 19726.1N WCAP-7803 (Non-proprietary)WCAP-7800 Revision 5Nuclear Fu el Division Quality Assurance Program PlanNov 19844.2, 17.1.3WCAP-7806Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison RodsDec 19714.3 WCAP-7811Power Distribution Control of Westinghouse Pressurized Water ReactorsDec 19714.3 WCAP-7832Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE ConditionsDec 19735.4 WCAP-7836Inlet Orificing of Open PWR CoresJan 19724.4 WCAP-7870Neutron Shielding PadsMay 19723.9N TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-7Rev. 30WCAP-7907LOFTRAN Code DescriptionJune 19726.3, 15.0, 15.1, 15.2, 15.3, 15.4, 15.5, 15.6WCAP-7908FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO 2 Fuel RodJune 197215.0, 15.2, 15.3, 15.4WCAP-7912-P-A (Proprietary) Power Peaking FactorsJan 19754.3 WCAP-7912-A (Non-proprietary) 4.4 WCAP-7913Process Instrumentation for Westinghouse Nuclear Steam Supply System (4-Loop Plant Using WCID 7300 Series Process Instrumentation)March 19737.3, 7.2WCAP-7921-ARDamping Values of Nuclear Power Plant ComponentsMay 19743.7N WCAP-7941-P-A (Proprietary) Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane GridJan 19754.4 WCAP-7959-A (Non-proprietary)WCAP-7950Fuel Assembly Safety Analysis fo r Combined Seismic and Loss of Coolant AccidentJuly 19723.7N WCAP-7956THINC-IV An Improved Program for Thermal-Hydraulic Anal ysis of Rod Bundle CoresJune 19734.4 WCAP-7964Axial Xenon Transient Tests at the Rochester Gas and Electric ReactorJune 19714.3 TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-8Rev. 30WCAP-7979-P-A (Proprietary)
TWINKLE - A Multi-Dimensional Neutron Kinetics Computer CodeJan 197515.0 WCAP-8028-A (Non-proprietary) 15.4 WCAP-7988 (Proprietary)
Application of Modified Spacer Factor to L. Grid Typical and Cold Wall Cell DNBOct 19724.4 WCAP-8030-A (Non-proprietary)
WCAP-8054 (Proprietary)
Application of the THINC-IV Program to PWR DesignOct 19734.4 WCAP-8195 (Non-proprietary)WCAP-8163Reactor Coolant Pump Integrity in LOCASept 19735.4 WCAP-8170 (Proprietary)
Calculational Model for Core Refloodi ng After a Loss of Coolant Accident (WREFLOOD Code)June 19746.2 WCAP-8171 (Non-proprietary) 15.6 WCAP-8174 (Proprietary) Effect of Local Heat WCAP-8202 Flux Sp ikes on DNB in Non-Uniform Heated Rod BundlesAug 19734.4 WCAP-8202 (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-9Rev. 30WCAP-8183 Revision 6Operational Experience With Westinghouse CoresJune 19774.2 WCAP-8200 Revision 2 (Proprietary) WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWRJuly 197415.6 WCAP-8261 Revision 1 (Non-proprietary)WCAP-8218-P-A (Proprietary)
Fuel Densification Experimental Results and Model for Reactor ApplicationMarch 19754.1 WCAP-8219-A (Non-proprietary) 4.2, 4.3, 4.4 WCAP-8236 (Proprietary))
Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant AccidentJan 19743.7N WCAP-8288 (Non-proprietaryDec 19734.2 WCAP-8236 Addendum 1 (Proprietary)
Safety Analysis of the 8-Grid 17x17 Fu el Assembly for Combined Seismic and Loss of Coolant AccidentMarch 19743.7N WCAP-8268 Addendum 1 (Non-proprietary)
April 1974WCAP-8252 Revision 1Documentation of Selected Westinghouse Structural Analysis Computer CodeJuly 19773.9N WCAP-8253 Amendment 1Source Term Data for Westinghouse Pressurized Water ReactorsJuly 197511.1 TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-10Rev. 30WCAP-8255Nuclear Instrumentation SystemJan 19747.2, 7.7 WCAP-8264-P-A (Proprietary) Westinghouse Mass and Energy Release Data for Containment DesignJune 19756.2.1 WCAP-8312-A (Non-proprietary)
Revision 2 WCAP-8278 (Proprietary) Hydraulic Flow Test of the 17x17 Fuel AssemblyFeb 19744.2 WCAP-8279 (Non-proprietary) 4.4 WCAP-8296-P-A (Proprietary) Effect of 17x17 Fuel Assembly Geometry on DNBFeb 19754.4 WCAP-8297 (Non-proprietary)WCAP-8298-P-A (Proprietary) The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal MixingJan 19754.4 WCAP-8299-A (Non-proprietary)
WCAP-8301 (Proprietary) LOCTA-IV Program: Loss of Coolant Transient AnalysisJune 197415.0 WCAP-8305 (Non-proprietary) 15.6 TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-11Rev. 30 WCAP-8302 (Proprietary) SATAN-VI Program: Comprehensive Space-Ti me Dependent Analysis of Loss of CoolantJune 19746.2.1 WCAP-8306 (Non-proprietary) 15.0, 15.6WCAP-8303-P-A (Proprietary) Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model TestsJuly 19753.9N WCAP-8317-A (Non-proprietary)WCAP-8324-AControl of Delta Ferrite in Austenitic Stainless Steel WeldmentsJune 19755.2 WCAP-8327 (Proprietary)
Containment Pressure Analysis Code (COCO)June 197415.6 WCAP-8326 (Non-proprietary)WCAP-8330Westinghouse Anticipated Transients Without Trip AnalysisAug 19744.3, 4.6, 15.1, 15.2, 15.4, 15.8 WCAP-8339Westinghouse ECCS Evaluation Model -SummaryJuly 19746.2.1, 15.6 WCAP-8340 (Proprietary) Westinghouse ECCS - Plant Sensitivity StudiesJuly 197415.6AE WCAP-8356 (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-12Rev. 30 WCAP-8341 (Proprietary) Westinghouse ECCS Evaluation Model Sensitivity StudiesJuly 197415.6 WCAP-8342 (Non-proprietary)WCAP-8359Effects of Fuel Densification Power Spikes on Clad Thermal TransientsJuly 19744.3 WCAP-8370/7800 Revision 10A/6AWestinghouse Water Reactor Divisions Quality Assurance PlanNov 198417.1.3 WCAP-8377 (Proprietary) Revised Clad Flattening ModelJuly 19744.2 WCAP-8381 (Non-proprietary)
WCAP-8385 (Proprietary)
Power Distribution Control and Load Follow ProceduresSept 19744.3 WCAP-8403 (Non-proprietary) 4.4 WCAP-8424 Revision 1An Evaluation of Loss of Flow Accidents Caus ed by Power System Frequency Transients in Westinghouse PWRsJune 197515.3 WCAP-8446 (Proprietary) 17x17 Drive Line Components Tests -Phase IB, II, III, D-Loop Drop and DeflectionDec 19743.9N WCAP-8449 (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-13Rev. 30WCAP-8453-AAnalysis of Data from the Zion (Unit 1) THINC Verification TestMay 19764.4 WCAP-8471 (Proprietary) Westinghouse ECCS Evaluation Model -Supplement InformationApril 197515.6 WCAP-8472 (Non-proprietary)
WCAP-8472 (Non-proprietary)WCAP-8498Incore Power Distribution Determination in Westinghouse Pressurized Water ReactorsJuly 19754.3 WCAP-8516-P (Proprietary) UHI Plant Internals Vibration Measurem ent Program and Pre and Post Host Functional ExaminationsApril 19753.9N WCAP-8517 (Non-proprietary)
WCAP-8536 (Proprietary) Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry With 22-Inch Grid SpacingMay 19754.4 WCAP-8537 (Non-proprietary)WCAP-8565-P-A (Proprietary) Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity StudiesJuly 197515.6 WCAP-8566-A (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-14Rev. 30 WCAP-8584 (Proprietary) Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Actuation SystemFeb 19764.6 WCAP-8760 (Non-proprietary)
WCAP-8760 (Non-proprietary)WCAP-8587 Revision 2Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical EquipmentFeb 19793.10N, 3.11N WCAP-8587 Supplement 1Equipment Qualification Data PackagesNov 19783.10N, 3.11N WCAP-8622 (Proprietary) Westinghouse ECCS Evaluation Model -October 1975 VersionNov 197515.6 WCAP-8623 (Non-proprietary)
WCAP-8624 (Proprietary)
General Method of Developing Multi-frequency Biaxial Test Inputs for BistablesSept 19753.10N WCAP-8695 (Non-proprietary)
Aug 1975 WCAP-8682 (Proprietary) Experimental Verification of Wet Fuel Storage Criticality AnalysesDec 19754.3 WCAP-8683 (Non-proprietary)TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-15Rev. 30 WCAP-8691 (Proprietary) Fuel Rod BowingDec 19754.2 WCAP-8692 (Non-proprietary)WCAP-8693Delta Ferrite in Production Austenitic Stainless Steel WeldmentsJan 19765.2 WCAP-8708 (Proprietary) MULTIFLEX - A FORTRAN-IV Com puter Program for Analyzing Thermal-Hydraulic-Structure System DynamicsFeb 19763.9N WCAP-8709 (Non-proprietary)
WCAP-8720 (Proprietary) Improved Analytical Models Used in Westinghouse Fuel Rod Design ComputationsOct 19764.2 WCAP-8785 (Non-proprietary)WCAP-8768 Revision 2Safety-related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries-Winter 1977-Summer 1978Oct 19781.5, 4.2, 4.3WCAP-8780Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on Trojan 1 Power PlantMay 19763.9N WCAP-8846-AHybrid B 4C Absorber Control Rod Evaluation ReportSept 197615.3 WCAP-8892-A (Non-proprietary) 7300 Series Process Control System Noise TestsJune 19777.1 WCAP-8929Benchwork Problem Solutions Employed for Verification WECAN Computer CodeJune 19773.9N TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-16Rev. 30 WCAP-8963 (Proprietary)
Safety Analysis for the Revised Fuel Rod Internal Pressure Design BasisNov 19764.2 WCAP-8964 (Non-proprietary)
Aug 1977 WCAP-8970 (Proprietary) Westinghouse ECCS Small Break October 1975 ModelApril 197715.6 WCAP-8971 (Non-proprietary)WCAP-8976Failure Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Control SystemSept 19774.6 WCAP-9168 (Proprietary) Westinghouse Emergency Core Cooling Sy stem Evaluation Model - Modified October 1975 VersionSept 197715.6 WCAP-9169 (Non-proprietary)
WCAP-9179 Revision 1 (Proprietary)
Properties of Fuel and Core Component MaterialsSept 19774.2 WCAP-9224 (Non-proprietary)
July 1978WCAP-9220-P-A (Proprietary) Westinghouse ECCS Evaluation Model, 1981 VersionDec 198115.6 TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSARMPS3 UFSAR1.6-17Rev. 30WCAP-9221-P-A (Non-proprietary)
Revision 1WCAP-9227Reactor Core Response to Excessive Secondary Steam ReleaseJan 197815.1 WCAP-9485 (Proprietary) PALADON - Westinghouse Nodal Computer CodeDec 19784.3 WCAP-9486 (Non-proprietary)WCAP-9600Report on Small Break Accidents for Westinghouse NSSS SystemJune 19795.4.15 WCAP-10858P-A Rev. 1AMSAC Generic Design PackageJuly 19877.8 RP-8A (SWEC)Radiation Shielding Design and Analysis Approach for Light Water Reactor Power PlantsMay 197512.2, 12.3 SWSQAP-1-74A (SWEC)Standard Nuclear Quality Assurance ProgramDec 197417.1.2 QA-1 Revision 3A (NUSCO)Quality Assurance ProgramMarch 197917.1.1TABLE 1.6-1 TOPICAL REPORTS AS GENERAL REFERE NCES (HISTORICAL) (CONTINUED)
Report NumberTitle NRC SubmittalReference Section(s)
MPS3 UFSAR1.7-1Rev. 30
1.7 DRAWINGS
AND OTHER DETAILED INFORMATION 1.7.1 ELECTRICAL, INSTRUMENTATI ON, AND CONTROL DRAWINGSTable 1.7-1 identifies the safety related electrical, in strumentation, and control drawings used on Millstone 3.
1.7.2 PIPING
AND INSTRUMENTATION DIAGRAMSTable 1.7-2 identifies the piping and instrumentation diagrams (P&ID) used on Millstone 3. These diagrams are included throughout the FSAR in conjunction with specific system descriptions. Symbols and abbreviations used in the diagrams are illustrated on Figure 1.2-3. A more complete listing of Drawing & Figure Numbers for the P&IDs is contained in the Generation Records Information System (GRITS), and for those drawings used as Figures in the FSAR, refer to the Summary Table of Contents, and Effective Pages List.
1.7.3 LOOP AND SYSTEMS DIAGRAMS This table has been deleted.
1.7.4 OTHER
DETAILED INFORMATION (SPECIAL REPORTS AND PROGRAMS)Table 1.7-4 identifies special reports and programs referenced in the FSAR and submitted separately from the FSAR.
MPS3 UFSAR1.7-2Rev. 30TABLE 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL REFERENCE DOCUMENTATION (Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-39001Westinghouse System or TitleWestinghouse Drawing Number15001Process Control Block Diagram7244D80 Sh. 115002Process Control Block Diagram7244D80 Sh. 215003Process Control Block Diagram7244D80 Sh. 3 15004Process Control Block Diagram7244D80 Sh. 415005Process Control Block Diagram7244D80 Sh. 515006Process Control Block Diagram7244D80 Sh. 6 15007Process Control Block Diagram7244D80 Sh. 715008Process Control Block Diagram7244D80 Sh. 815009Process Control Block Diagram7244D80 Sh. 9 15010Process Control Block Diagram7244D80 Sh. 1015011Process Control Block Diagram7244D80 Sh. 1115012Process Control Block Diagram7244D80 Sh. 12 15013Process Control Block Diagram7244D80 Sh. 1315014Process Control Block Diagram7244D80 Sh. 1415015Process Control Block Diagram7244D80 Sh. 15 15016Process Control Block Diagram7244D80 Sh. 1615017Process Control Block Diagram7244D80 Sh. 1715018Process Control Block Diagram7244D80 Sh. 18 15019Process Control Block Diagram7244D80 Sh. 1915020Process Control Block Diagram7244D80 Sh. 2015021Process Control Block Diagram7244D80 Sh. 21 15022Process Control Block Diagram7244D80 Sh. 2215023Process Control Block Diagram7244D80 Sh. 2315024Process Control Block Diagram7244D80 Sh. 2415025Process Control Block Diagram7244D80 Sh. 2515026Process Control Block Diagram7244D80 Sh. 26 MPS3 UFSAR1.7-3Rev. 3015027Process Control Block Diagram7244D80 Sh. 2715028Process Control Block Diagram7244D80 Sh. 2815029Process Control Block Diagram7244D80 Sh. 2915030Process Control Block Diagram7244D80 Sh. 30 15031Process Control Block Diagram7244D80 Sh. 3115032Process Control Block Diagram7244D80 Sh. 3215033Process Control Block Diagram7244D80 Sh. 33 15034Process Control Block Diagram7244D80 Sh. 3415035Process Control Block Diagram7244D80 Sh. 3515036Process Control Block Diagram7244D80 Sh. 36 15037Process Control Block Diagram7244D80 Sh. 3715038Process Control Block Diagram7244D80 Sh. 3815039Process Control Block Diagram7244D80 Sh. 40 15040Process Control Block Diagram7244D80 Sh. 3915041Process Control Block Diagram7244D80 Sh. 4115042Process Control Block Diagram7244D80 Sh. 42 15043Process Control Block Diagram7244D80 Sh. 4315044Process Control Block Diagram7244D80 Sh. 4415045Process Control Block Diagram7244D80 Sh. 45 15046Process Control Block Diagram7244D80 Sh. 4615047Process Control Block Diagram7244D80 Sh. 4715048Process Control Block Diagram7244D80 Sh. 48 04021NIS Source Range Functional Block Diagram5655D4904022NIS Source Range Functional Block Diagram5655D5004023NIS Source Range Functional Block Diagram5655D5104023NIS Source Range Functional Block Diagram5655D5207011Safeguard Test Cabinets8758D57 Sh. 1TABLE 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL REFERENCE DOCUMENTATION (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-39001Westinghouse System or TitleWestinghouse Drawing Number MPS3 UFSAR1.7-4Rev. 3007012Safeguard Test Cabinets8758D57 Sh. 207013Safeguard Test Cabinets8758D57 Sh. 3 07014Safeguard Test Cabinets8758D57 Sh. 407015Safeguard Test Cabinets8758D57 Sh. 5 07016Safeguard Test Cabinets8758D57 Sh. 607017Safeguard Test Cabinets8758D57 Sh. 707018Safeguard Test Cabinets8758D57 Sh. 8 07019Safeguard Test Cabinets8758D57 Sh. 907020Safeguard Test Cabinets8758D57 Sh. 1007021Safeguard Test Cabinets8758D57 Sh. 11 07022Safeguard Test Cabinets8758D57 Sh. 1207023Safeguard Test Cabinets8758D57 Sh. 1307025Safeguard Test Cabinets8758D57 Sh. 15 07026Safeguard Test Cabinets8758D57 Sh. 1607027Safeguard Test Cabinets8758D57 Sh. 1707028Safeguard Test Cabinets8758D57 Sh. 18 07029Safeguard Test Cabinets8758D57 Sh. 1907030Safeguard Test Cabinets8758D57 Sh. 2004002Logic Diagram108D684 Sh. 1 04003Logic Diagram108D684 Sh. 204004Logic Diagram108D684 Sh. 304005Logic Diagram108D684 Sh. 4 04006Logic Diagram108D684 Sh. 504007Logic Diagram108D684 Sh. 604008Logic Diagram108D684 Sh. 704009Logic Diagram108D684 Sh. 804010Logic Diagram108D684 Sh. 9TABLE 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL REFERENCE DOCUMENTATION (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-39001Westinghouse System or TitleWestinghouse Drawing Number MPS3 UFSAR1.7-5Rev. 3004011Logic Diagram108D684 Sh. 1004012Logic Diagram108D684 Sh. 1104013Logic Diagram108D684 Sh. 1204014Logic Diagram108D684 Sh. 13 04015Logic Diagram108D684 Sh. 1404016Logic Diagram108D684 Sh. 1504017Logic Diagram108D684 Sh. 16 04018Logic Diagram108D684 Sh. 1704019Logic Diagram108D684 Sh. 1904020Logic Diagram108D684 Sh. 18TABLE 1.7-1 LOGIC DIAGRAMS (Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title022-0171-6AReactor Trips Control L ogic Description Logic Diagram022-0181-6BReactor Trips Control L ogic Description Logic Diagram022-0191-6CReactor Trips Control L ogic Description Logic Diagram022-0201-6DReactor Trips Control L ogic Description Logic Diagram022-0211-6EReactor Trips Control L ogic Description Logic Diagram022-0221-6FReactor Trips Control L ogic Description Logic Diagram022-0231-6GReactor Trips Control L ogic Description Logic Diagram022-0241-6HReactor Trips Control L ogic Description Logic Diagram022-0251-6JReactor Trips Control L ogic Description Logic Diagram022-0261-6KReactor Trips Control L ogic Description Logic Diagram022-0271-6LReactor Trips Control L ogic Description Logic DiagramTABLE 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL REFERENCE DOCUMENTATION (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-39001Westinghouse System or TitleWestinghouse Drawing Number MPS3 UFSAR1.7-6Rev. 30022-0281-6MReactor Trips Control L ogic Description Logic Diagram022-0291-6NReactor Trips Control L ogic Description Logic Diagram 022-0301-6PAMSAC Logic Diagram022-0311-6QAMSAC Logic Diagram 025-0012-1.1ACirculating Water Pump Breaker025-0022.1-1BCirculating Water Discharge & Water Box Valves Control Logic Diagram025-0032.1-1CCirculating Water Discharge & Water Box Valves Control Logic Diagram025-0042.1-1DCirculating Water Discharge & Water Box Valves Control Logic Diagram025-0052.1-1ECirculating Water Discharge & Water Box Valves Control Logic Diagram025-0062.1-1FLogic Diagram Circulating Water Pump025-0072.1-1GLogic Diagram Circulating Water Pump025-0082.1-1HLogic Diagram Circulating Water Pump 025-0092.1-1JLogic Diagram Circulating Water Pump036-0013-1.1ATurbine Bypass Control Logic Diagram036-0023-1.1BTurbine Bypass Control Logic Diagram 036-0033-1.1CTurbine Bypass Control Logic Diagram036-0043-1.1DTurbine Bypass Control Logic Diagram036-0053-1.1ETurbine Bypass Control Logic Diagram 036-0063-1.1FTurbine Bypass Control Logic Diagram036-0073-1.1GTurbine Bypass Control Logic Diagram036-0083-1.1HTurbine Bypass Control Logic Diagram 036-0093-1.2AMain Steam Isolat ion Control Logic Diagram036-0103-1.2BMain Steam Isolat ion Control Logic Diagram036-0113-1.2CMain Steam Isolat ion Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-7Rev. 30036-0123-1.2DMain Steam Isolat ion Control Logic Diagram036-0133-1.2EMain Steam Isolat ion Control Logic Diagram061-0013-9AAux Stm Sys Control Logic Diagram061-0023-9BAux Stm Sys Control Logic Diagram 120-0016-1.1AMotor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram120-0026-1.1BMotor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram120-0036-1.1CMotor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram120-T0196-1.1DMotor Drive Stm Gen Pump & Discharge Valve Logic Diagram120-0046-1.2AFdwtr Sys Control Logic Diagram 120-0056-1.2BFdwtr Sys Control Logic Diagram120-0066-1.2CFdwtr Sys Control Logic Diagram120-0076-1.2DFdwtr Sys Control Logic Diagram 120-0086-1.2EFdwtr Sys Control Logic Diagram126-0016-2.1AMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0026-2.1BMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0036-2.1CMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0046-2.1DMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0056-2.1EMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0066-2.1FMotor Drive Aux Fdwtr Pump & Recirc Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-8Rev. 30126-0076-2.1GMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0086-2.1HMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0096-2.1JMotor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram126-0106-2.2ATurbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram126-0116-2.2BTurbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram126-0126-2.2CTurbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram5757-3.2Aux Fdwtr Pump & Driv e Lube Oil Control Logic Diagram143-0018-9AEmergency Diesel Genera tor Fuel Control Logic Diagram143-0028-9BEmergency Diesel Genera tor Fuel Control Logic Diagram148-0019-1AReactor Plant Component Cooling Water Control Logic Diagram148-0029-1BReactor Plant Component Cooling Water Control Logic Diagram148-0039-1CReactor Plant Component Cooling Water Control Logic Diagram148-0049-1DReactor Plant Component Cooling Water Control Logic Diagram148-0059-1EReactor Plant Component Cooling Water Control Logic Diagram148-0069-1FReactor Plant Component Cooling Water Control Logic Diagram148-0079-1GReactor Plant Component Cooling Water Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-9Rev. 30148-0089-1HReactor Plant Component Cooling Water Control Logic Diagram148-0099-1JReactor Plant Component Cooling Water Control Logic Diagram148-0109-1KReactor Plant Component Cooling Water Control Logic Diagram165-0019-2AChilled Water Syst em Control Logic Diagram165-0029-2BChilled Water Syst em Control Logic Diagram165-0039-2CChilled Water Syst em Control Logic Diagram165-0049-2DChilled Water Syst em Control Logic Diagram165-0059-2EChilled Water Syst em Control Logic Diagram165-0069-2FChilled Water Syst em Control Logic Diagram165-0079-2GChilled Water Syst em Control Logic Diagram165-0089-2HChilled Water Syst em Control Logic Diagram165-0099-2JChilled Water Syst em Control Logic Diagram165-0109-2KChilled Water Syst em Control Logic Diagram173-0019-4ACharging Pumps Cooling Control Logic Diagram173-0029-4BCharging Pumps Cooling Control Logic Diagram173-0039-4CCharging Pumps Cooling Control Logic Diagram 173-0049-4DCharging Pumps Cooling Control Logic Diagram179-0019-5ASafety Injection Pumps Cooling Control Logic Diagram179-0029-5BSafety Injection Pumps Cooling Control Logic Diagram184-0019-7ATurbine Plant Component Cooling Water Control Logic Diagram184-0029-7BTurbine Plant Component Cooling Water Control Logic Diagram184-0039-7CTurbine Plant Component Cooling Water Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-10Rev. 30184-0049-7DTurbine Plant Component Cooling Water Control Logic Diagram184-0059-7ETurbine Plant Component Cooling Water Control Logic Diagram193-0019-10AService Water System Control Logic Diagram193-0029-10BService Water System Control Logic Diagram193-0039-10CService Water System Control Logic Diagram 193-0049-10DService Water System Control Logic Diagram193-0059-10EService Water System Control Logic Diagram193-0069-10FService Water System Control Logic Diagram 193-0079-10GService Water System Control Logic Diagram193-0089-10HService Water System Control Logic Diagram193-0099-10JService Water System Control Logic Diagram 193-0109-10KService Water System Control Logic Diagram193-0119-10LService Water System Control Logic Diagram193-0129-10MService Water System Control Logic Diagram 224-00112-1AInstrument Air Control Logic Diagram224-00212-1BInstrument Air Control Logic Diagram224-00312-1CInstrument Air Control Logic Diagram 224-00412-1DInstrument Air Control Logic Diagram224-00112-1AInstrument Air Control Logic Diagram224-00212-1BInstrument Air Control Logic Diagram 224-00512-1EInstrument Air Control Logic Diagram224-00612-1FInstrument Air Control Logic Diagram224-00712-1GInstrument Air Control Logic Diagram224-00812-1HInstrument Air Control Logic Diagram224-00912-1JInstrument Air Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-11Rev. 30234-00312-3CContainment Instrument Air Control Logic Diagram267-00114-1ANitrogen System Control Logic Diagram267-00214-1BNitrogen System Control Logic Diagram267-00314-1CNitrogen System Control Logic Diagram 267-00414-1DNitrogen System Control Logic Diagram283-00115-03AFire Protection Low Press CO 2 Logic Diagram283-00215-03BFire Protection Low Press CO 2 Logic Diagram283-00315-03CFire Protection Low Press CO 2 Logic Diagram283-00415-03DFire Protection Low Press CO 2 Logic Diagram283-00515-03EFire Protection Low Press CO 2 Logic Diagram320-00122-1AReactor Plant Ventil ation Control Logic Diagram320-00222-1BReactor Plant Ventil ation Control Logic Diagram320-00322-1CReactor Plant Ventil ation Control Logic Diagram320-00422-1DReactor Plant Ventil ation Control Logic Diagram320-00522-1EReactor Plant Ventil ation Control Logic Diagram320-00622-1FReactor Plant Ventil ation Control Logic Diagram320-00722-1GReactor Plant Ventil ation Control Logic Diagram320-00822-1HReactor Plant Ventil ation Control Logic Diagram320-00922-1JReactor Plant Ventil ation Control Logic Diagram320-01022-1KReactor Plant Ventil ation Control Logic Diagram320-01122-1LReactor Plant Ventil ation Control Logic Diagram320-01222-1MReactor Plant Ventil ation Control Logic Diagram320-01322-1NReactor Plant Ventil ation Control Logic Diagram320-01422-1PReactor Plant Ventil ation Control Logic Diagram320-01522-1QReactor Plant Ventil ation Control Logic Diagram320-01622-1RReactor Plant Ventil ation Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-12Rev. 30320-01722-1SReactor Plant Ventil ation Control Logic Diagram320-01822-1TReactor Plant Ventil ation Control Logic Diagram320-01922-1UReactor Plant Ventil ation Control Logic Diagram320-02022-1VReactor Plant Ventil ation Control Logic Diagram320-02122-1WReactor Plant Ventil ation Control Logic Diagram320-02222-1XReactor Plant Ventil ation Control Logic Diagram320-02322-1YReactor Plant Ventil ation Control Logic Diagram320-02422-1ZReactor Plant Ventil ation Control Logic Diagram320-02522-1AAReactor Plant Ventil ation Control Logic Diagram320-02622-1ABReactor Plant Ventil ation Control Logic Diagram339-00122-7ADiesel Gen Bldg Ventilation Control Logic Diagram339-00222-7BDiesel Gen Bldg Ventilation Control Logic Diagram339-00322-7CDiesel Gen Bldg Ventilation Control Logic Diagram 342-00122-8AYard Structure Ventilation Control Logic Diagram342-00222-8BYard Structure Ventilation Control Logic Diagram342-00322-8CYard Structure Ventilation Control Logic Diagram 342-00422-8DYard Structure Ventilation Control Logic Diagram342-00522-8EYard Structure Ventilation Control Logic Diagram342-00622-8FYard Structure Ventilation Control Logic Diagram 342-00722-8GYard Structure Ventilation Control Logic Diagram342-00822-8HYard Structure Ventilation Control Logic Diagram342-00922-8JYard Structure Ventilation Control Logic Diagram 342-01022-8KYard Structure Ventilation Control Logic Diagram342-01122-8LYard Structure Ventilation Control Logic Diagram342-01222-8MYard Structure Ventilation Control Logic Diagram342-01322-8NYard Structure Ventilation Control Logic Diagram342-01422-8PYard Structure Ventilation Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-13Rev. 30342-01622-8SYard Structure Ventilation Control Logic Diagram345-00122-9AControl Bldg A/C Control Logic Diagram345-00222-9BControl Bldg A/C Control Logic Diagram345-00322-9CControl Bldg A/C Control Logic Diagram 345-00422-9DControl Bldg A/C Control Logic Diagram345-00522-9EControl Bldg A/C Control Logic Diagram345-00622-9FControl Bldg A/C Control Logic Diagram 345-00722-9GControl Bldg A/C Control Logic Diagram345-00822-9HControl Bldg A/C Control Logic Diagram345-00922-9JControl Bldg A/C Control Logic Diagram 345-01022-9KControl Bldg A/C Control Logic Diagram345-01122-9LControl Bldg A/C Control Logic Diagram345-01222-9MControl Bldg A/C Control Logic Diagram 345-01322-9NControl Bldg A/C Control Logic Diagram345-01422-9PControl Bldg A/C Control Logic Diagram 366-00122-12AChilled Water Syst em Control Logic Diagram366-00222-12BChilled Water Syst em Control Logic Diagram366-00322-12CChilled Water Syst em Control Logic Diagram366-00422-12DChilled Water Syst em Control Logic Diagram366-00522-12EChilled Water Syst em Control Logic Diagram366-00622-12FChilled Water Syst em Control Logic Diagram366-00722-12GChilled Water Syst em Control Logic Diagram366-00822-12HChilled Water Syst em Control Logic Diagram366-00922-12JChilled Water Syst em Control Logic Diagram36622-12KChilled Water Syst em Control Logic Diagram36622-12LChilled Water Syst em Control Logic Diagram36622-12MChilled Water Syst em Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-14Rev. 30373-00122-16AHot Water Heating Logic Diagram373-00222-16BHot Water Heating Logic Diagram373-00322-16CHot Water Heating Logic Diagram373-00422-16DHot Water Heating Logic Diagram 373-00522-16EHot Water Heating Logic Diagram587-00122-26AESF Bldg Ventilation Control Logic Diagram587-00222-26BESF Bldg Ventilation Control Logic Diagram 587-00322-26CESF Bldg Ventilation Control Logic Diagram587-00422-26DESF Bldg Ventilation Control Logic Diagram587-00522-26EESF Bldg Ventilation Control Logic Diagram 587-00622-26FESF Bldg Ventilation Control Logic Diagram587-00722-26GESF Bldg Ventilation Control Logic Diagram587-00822-26HESF Bldg Ventilation Control Logic Diagram 587-00922-26JESF Bldg Ventilation Control Logic Diagram 592-00122-27AContainment Structure Vent Control Logic Diagram592-00222-27BContainment Structure Vent Control Logic Diagram 592-00322-27CContainment Structure Vent Control Logic Diagram592-00422-27DContainment Structure Vent Control Logic Diagram592-00522-27EContainment Structure Vent Control Logic Diagram 592-00622-27FContainment Structure Vent Control Logic Diagram592-00722-27GContainment Structure Vent Control Logic Diagram 595-00122-28AMain Stm Valve Bldg Ventilation Control Logic Diagram 595-00222-28BMain Stm Valve Bldg Ventilation Control Logic Diagram 391-00122-33AHot Water Preheating Sys Control Logic Diagram391-00222-33BHot Water Preheating Sys Control Logic Diagram 708-00124-3AReserve Station Service Breaker Controls Control Logic DescriptionTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-15Rev. 30708-00224-3BReserve Station Service Breaker Controls Control Logic Diagram708-00324-3CReserve Station Service Breaker Controls Control Logic Diagram708-00424-3DReserve Station Service Breaker Controls Control Logic Diagram708-00524-3EReserve Station Service Breaker Controls Control Logic Diagram708-00624-3FReserve Station Service Breaker Controls Control Logic Diagram708-00724-3GReserve Station Service Breaker Controls Control Logic Diagram708-00824-3HReserve Station Service Breaker Controls Control Logic Diagram708-00924-3JReserve Station Service Breaker Controls Control Logic Diagram708-01024-3KReserve Station Service Breaker Controls Control Logic Diagram709-00124-4AMedium Voltage Bus Tie Breaker Controls Control Logic Diagram709-00224-4BMedium Voltage Bus Tie Breaker Controls Control Logic Diagram713-00124-8ALow Voltage Switchgear Supply Breaker Controls Control Logic Diagram713-00224-8BLow Voltage Switchgear Supply Breaker Controls Control Logic Diagram721-00124-9.2AEmergency Gen Breaker Controls Control Logic Diagram721-00224-9.2BEmergency Gen Breaker Controls Control Logic Diagram721-00324-9.2CEmergency Gen Breaker Controls Control Logic Diagram721-00424-9.2DEmergency Gen Breaker Controls Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-16Rev. 30722-00124-9.3AEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00224-9.3BEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00324-9.3CEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00424-9.3DEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00524-9.3EEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00624-9.3FEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00724-9.3GEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00824-9.3HEmergency Diesel Ge n Control & Protection Control Logic Diagram722-00924-9.3JEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01024-9.3KEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01124-9.3LEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01224-9.3MEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01324-9.3NEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01424-9.3OEmergency Diesel Ge n Control & Protection Control Logic Diagram722-01524-9.3PEmergency Diesel Ge n Control & Protection Control Logic Diagram 723-00124-9.4AEmergency Gen Load Sequence Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-17Rev. 30723-00224-9.4BEmergency Gen Load Sequence Control Logic Diagram723-00324-9.4CEmergency Gen Load Sequence Control Logic Diagram723-00424-9.4DEmergency Gen Load Sequence Control Logic Diagram723-00524-9.4EEmergency Gen Load Sequence Control Logic Diagram 723-00624-9.4FEmergency Gen Load Sequence Control Logic Diagram723-00724-9.4GEmergency Gen Load Sequence Control Logic Diagram723-00824-9.4HEmergency Gen Load Sequence Control Logic Diagram 723-00924-9.4JEmergency Gen Load Sequence Control Logic Diagram723-01024-9.4KEmergency Gen Load Sequence Control Logic Diagram723-01124-9.4LEmergency Gen Load Sequence Control Logic Diagram 723-01224-9.4MEmergency Gen Load Sequence Control Logic Diagram723-01324-9.4NEmergency Gen Load Sequence Control Logic Diagram723-01424-9.4PEmergency Gen Load Sequence Control Logic Diagram 723-01524-9.4QEmergency Gen Load Sequence Control Logic Diagram723-01624-9.4REmergency Gen Load Sequence Control Logic Diagram723-01724-9.4SEmergency Gen Load Sequence Control Logic Diagram 723-01824-9.4TEmergency Gen Load Sequence Control Logic Diagram723-01924-9.4UEmergency Gen Load Sequence Control Logic Diagram723-02024-9.4VEmergency Gen Load Sequence Control Logic Diagram 723-02124-9.4WEmergency Gen Load Sequence Control Logic Diagram723-02224-9.4XEmergency Gen Load Sequence Control Logic Diagram723-02324-9.4YEmergency Gen Load Sequence Control Logic Diagram 723-02424-9.4ZEmergency Gen Load Sequence Control Logic Diagram 714-00124-10ABattery Power Supply Co ntrol Logic Functional Diagram714-00224-10BBattery Power Supply Control Logic Diagram714-00324-10CBattery Power Supply Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-18Rev. 30714-00424-11AInstrument and Control AC Power Supply Control Logic Functional Diagram714-00524-11BInstrument and Control AC Power Supply Control Logic Functional Diagram714-00624-11CInstrument and Control AC Power Supply Control Logic Functional Diagram 724-00124-12.5ASynchronizing Chec k Control Logic Diagram724-00224-12.5BSynchronizing Chec k Control Logic Diagram724-00324-12.5CSynchronizing Chec k Control Logic Diagram724-00424-12.5DSynchronizing Chec k Control Logic Diagram 404-00125-1.1AReactor Coolant Pumps Control Logic Diagram404-00225-1.1BReactor Coolant Pumps Control Logic Diagram404-00325-1.1CReactor Coolant Pumps Control Logic Diagram 404-00425-1.1DReactor Coolant Pu mps Control Logic Diagram 404-00525-1.2APressurized Control Logic Diagram404-00625-1.2BPressurized Control Logic Diagram 404-00725-1.2CPressurized Control Logic Diagram404-00825-1.2DPressurized Control Logic Diagram404-00925-1.2EPressurized Control Logic Diagram 404-01025-1.2FPressurized Control Logic Diagram404-01125-1.2GPressurized Control Logic Diagram404-01225-1.2HPressurized Control Logic Diagram 404-01325-1.JPressurized Control Logic Diagram404-01425-1.KPressurized Control Logic Diagram404-01525-1.LPressurized Control Logic Diagram 404-01625-1.4APressurizer Relief Tank Control Logic Diagram404-01725-1.4BPressurizer Relief Tank Control Logic Diagram 504-01825-1.5AReactor Coolant Isolation Valves Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-19Rev. 30504-01925-1.5BReactor Coolant Isolation Valves Control Logic Diagram504-02025-1.5CReactor Coolant Isolation Valves Control Logic Diagram414-00126-2.1AReactor Coolant Le tdown Control Logic Diagram414-00226-2.1BReactor Coolant Le tdown Control Logic Diagram414-00326-2.1CReactor Coolant Le tdown Control Logic Diagram414-00426-2.1DReactor Coolant Le tdown Control Logic Diagram414-00526-2.2AVolume Control Tank Control Logic Diagram 414-00626-2.2BVolume Control Tank Control Logic Diagram414-00726-2.2CVolume Control Tank Control Logic Diagram414-00826-2.2DVolume Control Tank Control Logic Diagram 414-01026-2.3ACharging Pumps Control Logic Diagram414-01126-2.3BCharging Pumps Control Logic Diagram414-01226-2.3CCharging Pumps Control Logic Diagram 414-01326-2.3DCharging Pumps Control Logic Diagram414-01426-2.3ECharging Pumps Control Logic Diagram414-01526-2-3FCharging Pump s Control Logic Diagram414-01626-2.3GCharging Pumps Control Logic Diagram414-01726-2.3HCharging Pumps Control Logic Diagram414-01826-2.3JCharging Pumps Control Logic Diagram 414-02926-2.5AReactor Makeup & Boric Acid Blender Control Logic Diagram414-03026-2.5BReactor Makeup & Boric Acid Blender Control Logic Diagram414-03126-2.5CReactor Makeup & Boric Acid Blender Control Logic Diagram414-03226-2.5DReactor Makeup & Boric Acid Blender Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-20Rev. 30414-03326-2.5EReactor Makeup & Boric Acid Blender Control Logic Diagram414-03426-2.5FReactor Makeup & Boric Acid Blender Control Logic Diagram414-03526-2.5GReactor Makeup & Boric Acid Blender Control Logic Diagram414-03626-2.6AReactor Coolant Pumps Seal Water Control Logic Diagram414-03726-2.6BReactor Coolant Pumps Seal Water Control Logic Diagram434-00127-2AHigh Pressure Safety Injection Control Logic Diagram434-00227-2BHigh Pressure Safety Injection Control Logic Diagram434-00327-2CHigh Pressure Safety Injection Control Logic Diagram 434-00427-2DHigh Pressure Safety Injection Control Logic Diagram434-00527-2EHigh Pressure Safety Injection Control Logic Diagram434-00627-2FHigh Pressure Safety Injection Control Logic Diagram 434-00727-2GHigh Pressure Safety Injection Control Logic Diagram434-00827-2HHigh Pressure Safety Injection Control Logic Diagram434-00927-2JHigh Pressure Safety Injection Control Logic Diagram 434-01027-2KHigh Pressure Safety Injection Control Logic Diagram434-01127-2LHigh Pressure Safety Injection Control Logic Diagram441-00127-3ALow Pressure Safety Injection Control Logic Diagram441-00227-3BLow Pressure Safety Injection Control Logic Diagram441-00327-3CLow Pressure Safety Injection Control Logic Diagram441-00427-3DLow Pressure Safety Injection Control Logic Diagram 441-00527-3ELow Pressure Safety Injection Control Logic Diagram441-00627-3FLow Pressure Safety Injection Control Logic Diagram441-00727-3GLow Pressure Safety Injection Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-21Rev. 30441-00827-3HLow Pressure Safety Injection Control Logic Diagram 449-00127-7AResidual Heat Removal Control Logic Diagram449-00227-7BResidual Heat Removal Control Logic Diagram449-00327-7CResidual Heat Removal Control Logic Diagram 449-00427-7DResidual Heat Removal Control Logic Diagram449-00527-7EResidual Heat Removal Control Logic Diagram449-00627-7FResidual Heat Removal Control Logic Diagram 449-00727-7GResidual Heat Removal Control Logic Diagram449-00827-7HResidual Heat Removal Control Logic Diagram449-00927-7JResidual Heat Removal Control Logic Diagram 449-01027-7KResidual Heat Rem oval Control Logic Diagram 454-00127-10AContainment Vac uum Control Logic Diagram454-00227-10BContainment Vac uum Control Logic Diagram 457-00127-11AContainment Recirc Control Logic Diagram457-00227-11BContainment Recirc Control Logic Diagram457-00327-11CContainment Recirc Control Logic Diagram 457-00427-11DContainment Recirc Control Logic Diagram457-00527-11EContainment Recirc Control Logic Diagram457-00627-11FContainment Recirc Control Logic Diagram 457-00727-11GContainment Recirc Control Logic Diagram457-00827-11HContainment Recirc Control Logic Diagram457-01927-11JContainment Recirc Control Logic Diagram 457-01027-11KContainment Recirc Control Logic Diagram457-01127-11LContainment Recirc Control Logic Diagram 463-00127-12AQuench Spray Control Logic Diagram463-00227-12BQuench Spray Control Logic Diagram463-00327-12CQuench Spray Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-22Rev. 30463-00427-12DQuench Spray Control Logic Diagram463-00527-12EQuench Spray Control Logic Diagram463-00627-12FQuench Spray Control Logic Diagram 470-00127-13ADBA Hydrogen Recombiner Control Logic Diagram 470-00227-13BDBA Hydrogen Recombiner Control Logic Diagram470-00327-13CDBA Hydrogen Recombiner Control Logic Diagram470-00427-13DDBA Hydrogen Recombiner Control Logic Diagram 470-00527-13EDBA Hydrogen Recombiner Control Logic Diagram719-00127-17ASafety Injection Ac tuation Control Logic Diagram719-00227-17BSafety Injection Ac tuation Control Logic Diagram719-00327-17CSafety Injection Ac tuation Control Logic Diagram 716-00127-18AContainment Spray Ac tuation Control Logic Diagram716-00227-18BContainment Spray Ac tuation Control Logic Diagram716-00327-18CContainment Spray Ac tuation Control Logic Diagram720-00127-19AContainment Isolation Control Logic Diagram 720-00227-19BContainment Isolation Control Logic Diagram720-00327-19CContainment Isolation Control Logic Diagram720-00427-19DContainment Isolation Control Logic Diagram 720-00527-19EContainment Isolation Control Logic Diagram720-00627-19FContainment Isolation Control Logic Diagram720-00727-19GContainment Isolation Control Logic Diagram 482-01231-1.1High Level Waste Drn. TK/Pump 482-00231-1.2AWaste Eval Reblr. Pump/Waste Dist. Pump482-00331-1.2BWaste Eval Reblr. Pump/Waste Dist. Pump482-00431-1.2CWaste Eval Reblr. Pump/Waste Dist. PumpTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-23Rev. 30482-00531-1.2DWaste Eval Reblr. Pump/Waste Dist. Pump482-00631-1.2EWaste Eval Reblr. Pump/Waste Dist. Pump482-01831-1.3AWaste Test TK/Pump482-01931-1.3BWaste Test TK/Pump 482-02031-1.3CWaste Test TK/Pump 482-02131-1.4AWaste Evap Bot Pump/Waste Bot. Coolant Pump482-02231-1.4BWaste Evap Bot Pump/Waste Bot. Coolant Pump482-02331-1.4CWaste Evap Bot Pump/Waste Bot. Coolant Pump482-02431-1.4DWaste Evap Bot Pump/Waste Bot. Coolant Pump 482-02531-1.4EWaste Evap Bot Pump/Waste Bot. Coolant Pump 482-02631-1.5ALow Level Wast. Drn. TK/Pump482-02731-1.5BLow Level Wast. Drn. TK/Pump 492-00131-2.1ADegasifier & Recirc Pumps Control Logic Diagram492-00231-2.1BDegasifier & Recirc Pumps Control Logic Diagram492-00331-2.1CDegasifier & Recirc Pumps Control Logic Diagram 492-00431-2.1DDegasifier & Recirc Pumps Control Logic Diagram492-00531-2.1EDegasifier & Recirc Pumps Control Logic Diagram492-00631-2.1FDegasifier & Recirc Pumps Control Logic Diagram 492-00731-2.1GDegasifier & Recirc Pumps Control Logic Diagram492-00831-2.1HDegasifier & Recirc Pumps Control Logic Diagram 516-00132-3AReactor Plant Gaseous Drains Control Logic Diagram516-00232-3BReactor Plant Gaseous Drains Control Logic Diagram516-00332-3CReactor Plant Gaseous Drains Control Logic Diagram516-00432-3DReactor Plant Gaseous Drains Control Logic Diagram 522-00132-4AReactor Plant Aerated Drains Control Logic Diagram522-00232-4BReactor Plant Aerated Drains Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-24Rev. 30522-00332-4CReactor Plant Aerated Drains Control Logic Diagram522-00432-4DReactor Plant Aerated Drains Control Logic Diagram522-00532-4EReactor Plant Aerated Drains Control Logic Diagram640-00132-5.1ATurbine Plant Misc Drains Control Logic Diagram 640-00232-5.1BTurbine Plant Misc Drains Control Logic Diagram640-002A32-5.1B1Turbine Plant Misc Drains Control Logic Diagram640-00332-5.1CTurbine Plant Misc Drains Control Logic Diagram 640-00432-5.1DTurbine Plant Misc Drains Control Logic Diagram 546-00132-13ASteam Generator Blowdown System Control Logic Diagram546-00232-13BSteam Generator Blowdown System Control Logic Diagram546-00332-13CSteam Generator Blowdown System Control Logic Diagram55433-1Containment Leakage Moni toring Control Logic Diagram55733-2Containment Atmosphere Monitoring Control Logic Diagram 560-00134-1AFuel Pool Cooling & Purification System Control Logic Diagram560-00234-1BFuel Pool Cooling & Purification System Control Logic Diagram560-00334-1CFuel Pool Cooling & Purification System Control Logic Diagram560-00434-1DFuel Pool Cooling & Purification System Control Logic Diagram560-00634-1EFuel Pool Cooling & Purification System Control Logic Diagram567-00135-1APrimary Grade Water Control Logic Diagram567-00235-1BPrimary Grade Water Control Logic Diagram567-00335-1CPrimary Grade Water Control Logic DiagramTABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-25Rev. 30567-00435-1DPrimary Grade Water Control Logic DiagramTABLE 1.7-1 ONE-LINE DIAGRAMS - CLASS 1E (Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Dwg.
Number 25212-30SWEC Drawing Number 12179-EE-Diagram Title0011AMain One-Line Diagram Power Distribution Composite0271AD480 V MCC One-Line Diag ram (Bus 32-1R, 1W) Aux Bldg (Sh. 3)0631AE480 V MCC One-Line Diagram (Bus 32-5G, 5H, 5T, 5U) Circulating Water Pump-house0401AH480 V MCC One-Line Diagram (Bus 32-2F, 4T) ESF Bldg (Sh.
1)0411AJ480 V MCC One-Line Diagra m (Bus 32-2J, 3U, 4U) ESF Bldg (Sh. 2)0591AK480 V MCC One-Line Diagram Bus 32-3D, 1L, 1T, 1U) Diesel Enclosure & Auxiliary Bldg0471AQ480 V MCC One-Line Diagra m (Bus 32-2R, 2W) Rod Control Area (Sh. 2)0611AS480 V MCC One-Line Diagram (Bus 32-2T, 2U) Control Bldg0651AT480 V MCC One-Line Diagram (Bus 32-3A, 3P, 3T) Turbine Building (Sheet 4)0761BA125 V dc and 120 V ac Distribution Sys Composite0771BB125 V dc One-Line Diagram Batteries 301A-1 and 301A-2 0781BC125 V dc One-Line Diagram Batteries 301B-1 and 301B-20791BD125 V dc One-Line Diagram Batteries 301C-10801BE125 V dc One-Line Diagram Battery 301D-A0811BF120 V ac One-Line Diagram Vital Bus I and IIITABLE 1.7-1 LOGIC DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-28SWEC Drawing Number 12179-LSK-Diagram Title MPS3 UFSAR1.7-26Rev. 300821BG120 V ac One-Line Diagram Vital Bus II and IV0831BH120 V ac One-Line Diagram Vital Bus Back-Up 1 and 20871BM125 V Misc dc One-Line Diagram Battery 10881BN125 V Misc dc One-Line Diagram Battery 2 (Sh. 1) 1061BR125 V Misc dc Line Diagram Batteries 1 & 21071BS125 V dc One-Line Diagram Batteries 1 & 2108 (Sh. 1)1BT125 V Misc dc One-Line Diagram 122 (Sh. 3)1CC120 V Misc ac One-Line Diagram126 (Sh. 7)1CG120 V Misc ac One-Line Diagram127 (Sh. 8)1CH120 V Misc ac One-Line Diagram 128 (Sh. 9)1CJ120 V Misc ac One-Line Diagram129 (Sh. 10)1CK120 V Misc ac One-Line Diagram125 (Sh. 12)1CM120 V Misc ac One-Line Diagram 0041DMain One-Line Diagram 4,160 & 480 Normal0281EF480 V One-Line Diagram (Bus 32Y, X) [3EJS*US-4A, 4B]30030IEH480 V One-Line Diagram Emergency Diesel Generator 480 V Distribution Panels0181K4.16 kV One-Line Diagram Bus 34C [3ENS*SWG-A(-0)]
(Sh. 1)0191L4.16 kV One-Line Diagra m Bus 34C [3ENS*SWG-A(0)]
(Sh. 2)0201M4.16 kV One-Line Diagra m Bus 34D [3ENS*SWG-B(P)]
(Sh. 1)0211N4.16 kV One-Line Diagra m Bus 34D [3ENS*SWG-B(P)]
(Sh. 2)0551U480 V One-Line Diagram (Bus 32T, U) [3EJS*US-1A, 1B]0321V480 V One-Line Diagram (B us 32S, V) [3EJS*US-2A, 2B]0331W480 V One-Line Diagram (Bus 32R, W) [3EJS*US-3A, 3B]TABLE 1.7-1 ONE-LINE DIAGRAMS - CLASS 1E (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Dwg.
Number 25212-30SWEC Drawing Number 12179-EE-Diagram Title MPS3 UFSAR1.7-27Rev. 30TABLE 1.7-1 ELEMENTARY DIAGRAMS (Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram TitleSh. 2A2ADevice Numbers and General NotesSh. 2B2BCode Information GuideSh. 2C2CElementary Diagram SymbolsSh. 2D 2DElementary Diagram Symbols Sh. 2E2EElementary Diagram SymbolsSh. 2F2FElementary Diagram SymbolsSh. 2G2GOne-Line Diagram Symbols Sh. 2H2HOne-Line Diagram SymbolsSh. 2J2JOne-Line Diagram SymbolsSh. 2K2KOne-Line Diagram Symbols Sh. 2L 2LLegend and Location SymbolsSh. 2M2MLegend and Location SymbolsSh. 2N2NLegend and Location Symbols Sh. 2P2PLegend and Location SymbolsSh. 2Q2QTyp isolation CKT & presentation on ESK320013AControl Switch Contact Diagram 320013BControl Switch Contact Diagram320013CControl Switch Contact Diagram320013DControl Switch Contact Diagram 320013EControl Switch Contact Diagram320013FControl Switch Contact Diagram320013GControl Switch Contact Diagram 320013HControl Switch Contact Diagram320013JControl Switch Contact Diagram320013KControl Switch Contact Diagram320013LControl Switch Contact Diagram MPS3 UFSAR1.7-28Rev. 30320013MControl Switch Contact Diagram320013NControl Switch Contact Diagram320013PControl Switch Contact Diagram320013QControl Switch Contact Diagram 320013RControl Switch Contact Diagram320013SControl Switch Contact Diagram320013TControl Switch Contact Diagram 320013UControl Switch Contact Diagram320013VControl Switch Contact Diagram320013WControl Switch Contact Diagram 320013XControl Switch (OIM) Contact Diagram320013YControl Switch Contact Diagram320013ZControl Switch Contact Diagram 320013AAControl Switch Contact Diagram320013ABControl Switch Contact Diagram320013AC1Control Switch Contact Diagram 320013AC2Control Switch Contact Diagram320013ADControl Switch Contact Diagram320013AEControl Switch Contact Diagram 320014AAAOutline Isolator Cabinets320014AAH01Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP320014AAH02Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP320014AAH03Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP320014A01Main Control Board Outline320014BA03Auxiliary Shutdown Panel 3RPS*PNLAS320014BA04Auxiliary Shutdown Panel 3RPS*PNLASTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-29Rev. 30320014BA05Auxiliary Shutdown Panel 3RPS*PNLASSh. 4BA64BA6Auxiliary Shutdow n Panel (ASP) 3RPS*PNLAS (Sh.6)Sh. 4BA74BA7Auxiliary Shutdow n Panel (ASP) 3RPS*PNLAS (Sh.7)Sh. 4BA84BA8Auxiliary Shutdow n Panel (ASP) 3RPS*PNLAS (Sh.8)Sh. 4BA94BA9Auxiliary Shutdow n Panel (ASP) 3RPS*PNLAS (Sh.9)320014B02MB1 Front Section320014B03MB1 Front Section 320014B04MB1 Front Section320014B05MB1 Front Section320014B06Main Control Board Front Section MB1 320014B07Main Control Board Front Section MB1320014B08Main Control Board Front Section MB1320014B09MB1 Front Section 320014B10Main Control Board Front Section MB1320014B11MB1 Front Section320014C02Main Control Board Front Section MB2 320014C03Main Control Board Front Section MB2320014C04Main Control Board Front Section MB2320014C05Main Control Board Front Section MB2 320014C06Main Control Board Front Section MB2320014C07Main Control Board Front Section MB2320014C08Main Control Board Front Section MB2 320014C09Main Control Board Front Section MB2320014C10Main Control Board Front Section MB2320014C11Main Control Board Front Section MB2320014D02Main Control Board Front Section MB3TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-30Rev. 30320014D03Main Control Board Front Section MB3320014D04Main Control Board Front Section MB3320014D05Main Control Board Front Section MB3320014D06Main Control Board Front Section MB3 320014D07Main Control Board Front Section MB3320014D08Main Control Board Front Section MB3320014D09Main Control Board Front Section MB3 320014D11Main Control Board Front Section MB3320014E02Main Control Board Front Section MB4320014E03Main Control Board Front Section MB4 320014E04Main Control Board Front Section MB4320014E05Main Control Board Front Section MB4320014E06Main Control Board Front Section MB4 320014E07Main Control Board Front Section MB4320014E08Main Control Board Front Section MB4320014F02Main Control Board Front Section MB5 320014F03Main Control Board Front Section MB5320014F04Main Control Board Front Section MB5320014F05Main Control Board Front Section MB5 320014F06Main Control Board Front Section MB5320014F07Main Control Board Front Section MB5320014F08Main Control Board Front Section MB5 320014F09Main Control Board Front Section MB5320014F10Main Control Board Front Section MB5320014F11Main Control Board Front Section MB5320014F13Main Control Board Front Section MB5TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-31Rev. 30320014G02Main Control Board Front Section MB6320014G03Main Control Board Front Section MB6320014G04Main Control Board Front Section MB6320014G05Main Control Board Front Section MB6 320014G06Main Control Board Front Section MB6320014G07Main Control Board Front Section MB6320014G07AMain Control Board Front Section MB6 320014H02Main Control Board Front Section MB7320014H03Main Control Board Front Section MB7320014H04Main Control Board Front Section MB7 320014H05Main Control Board Front Section MB7320014H06Main Control Board Front Section MB7320014H07Main Control Board Front Section MB7 320014H08Main Control Board Front Section MB7320014H09Main Control Board Front Section MB7320014H10Main Control Board Front Section MB7 320014H12Main Control Board Front Section MB7320014J02Main Control Board Front Section MB8320014J03Main Control Board Front Section MB8 320014J04Main Control Board Front Section MB8320014J05Main Control Board Front Section MB8320014J06Main Control Board Front Section MB8 320014J07Main Control Board Front Section MB8320014J08Main Control Board Front Section MB8320014J09Main Control Board Front Section MB8320014J13Main Control Board Front Section MB8TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-32Rev. 30320014K02Main Control Board Front Section MBIR320014K03Main Control Board Front Section MB1R320014K04Main Control Board Front Section MB1RSh. 4LA34LA3Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA44LA4Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA54LA5Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA64LA6Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA7 4LA7 Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA84LA8Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA94LA9Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA104LA10Main Vent and Air Cond Panel 3HVC*PNLVP1Sh. 4LA114LA11Main Vent and Air Cond Panel 3HVS*PNLVP1Sh. 4LA124LA12Main Vent and Air Cond Panel 3HVS*PNLVP1 Sh. 4LA134LA13Main Vent and Air Cond Panel 3HVS*PNLVP1Sh. 4LA144LA14Main Vent & Air Cond Panel 3HVS*PNLVP1Sh. 4LA154LA15Main Vent and Air Cond Panel Sh. 4L24L2Main Control Board Rear Section MB2RSh. 4L34L3Main Control Board Rear Section MB2RSh. 4L44L4Main Control Board Rear Section MB2RSh. 4M24M2Main Control Board Rear Section MB3RSh. 4M34M3Main Control Board Rear Section MB3RSh. 4N24N2Main Control Board Rear Section MB4R Sh. 4N34N3Main Control Board Rear Section MB4RSh. 4N44N4Main Control Board Rear Section MB4RSh. 4P24P2Main Control Board Rear Section MB5RSh. 4P34P3Main Control Board Rear Section MB5RTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-33Rev. 30Sh. 4P44P4Main Control Board Rear Section MB5RSh. 4P54P5Main Control Board Rear Section MB5RSh. 4P64P6Main Control Board Rear Section MB5RSh. 4P74P7Main Control Board Rear Section MB5RSh. 4P84P8Main Control Board Rear Section MB5RSh. 4Q24Q2Main Control Board Rear Section MB6RSh. 4Q34Q3Main Control Board Rear Section MB6R Sh. 4Q44Q4Main Control Board Rear Section MB6RSh. 4Q54Q5Main Control Board Rear Section MB6RSh. 4Q64Q6Main Control Board Rear Section MB6R Sh. 4R24R2Main Control Board Rear Section MB7RSh. 4R34R3Main Control Board Rear Section MB7RSh. 4R44R4Main Control Board Rear Section MB7R Sh. 4R54R5Main Control Board Rear Section MB7RSh. 4R64R6Main Control Board Rear Section MB7RSh. 4S24S2Main Control Board Rear Section MB8RSh. 4S34S3Main Control Board Rear Section MB8RSh. 4S44S4Main Control Board Rear Section MB8RSh. 4S54S5Main Control Board Rear Section MB8RSh. 4S64S6Main Control Board Rear Section MB8RSh. 4S74S7Main Control Board Rear Section MB8RSh. 4U14U1Post Accident Sampling Panel 3SSP*PNL3 Sh. 4U24U2Post Accident Sampling Panel 3SSP*PNL3Sh. 4U34U3Post Accident Sampling Panel 3SSP*PNL3Sh. 4U44U4Post Accident Sampling Panel 3SSP*PNL3Sh. 4U54U5Post Accident Sampling Panel 3SSP*PNL3TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-34Rev. 30Sh. 4YA34YA3Auxiliary Relay Rack (AR4) 3RPS*RakotxaSh. 4YA44YA4Auxiliary Relay Rack (AR4) 3RPS*RakotxaSh. 4YA54YA5Auxiliary Relay Rack (AR4) 3RPS*RakotxaSh. 4ZA34ZA3Auxiliary Rela y Rack (AR5)3RPS*RakotxbSh. 4ZA44ZA4Auxiliary Rela y Rack (AR5) 3RPS*RakotxbSh. 4ZA54ZA5Auxiliary Rela y Rack (AR5) 3RPS*RakotxbSh. 4ZB4ZBAux Relay & Cont Panel 3HVC*PNL CHLIA Cont Bldg Chld Wtr Sys-Train "A" AuxSh. 4ZC4ZCAux Relay & Cont Panel 3HVC*PNL CHIB Cont Bldg Chld Wtr Sys-Train "B"Sh. 4ZD24ZD2Transfer Switch Panel Train A (TSPA) 3 ES*PNL TSASh. 4ZD34ZD3Transfer Switch Panel Train A (TSPA) 3 ES*PNL TSASh. 4ZD44ZD4Transfer Switch Panel Train A (TSPA) 3CES*PNL TSASh. 4ZE24ZE2Transfer Switch Panel Train A (TSPA) 3CES*PNL TSASh. 4ZE34ZE3Transfer Switch Panel Train A (TSPA) 3CES*PNL TSASh. 5A5ATyp Med Voltage Swgr ACB Sh. 5BB5BB4.16 KV Norm Stat Service BKrSh. 5BC5BCNorm Sta Svce Bkr [3NNS-ACB-BN] 35A3-34B-2Sh. 5BD5BDRsv Sta Svce Bkr [3ENS*ACB-AR] 23SA3-34C-2 Sh. 5BE5BERsv Sta Svce Bk r [3ENS*ACB-BR] 23SA-3-34D-2Sh. 5BF5BFBus Tie Bkr [3ENS*ACB-TA] 34C-1T-2Sh. 5BG5BGBus Tie Bkr [3ENS*ACB-TB] 34D-1T-2 Sh. 5BT5BTUS Fdr Bkr [3ENS*ACB-AA], 34C3-2Sh. 5BU5BUUS Fdr Bkr [3ENS*ACB-AB], 34C4-2Sh. 5BV5BVUS Fdr Bkr [3ENS*ACB-BA], 34D2-2Sh. 5BW5BWUS Fdr Bkr [3ENS*ACB-BB], 34D3-2Sh. 5BX5BXUS Fdr Bkr [3ENS*ACB-AC], 34C5-2TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-35Rev. 30Sh. 5BY5BYUS Fdr Bkr [3ENS*ACB-BC], 34D4-2Sh. 5CJ5CJService Wtr Pump [3SWP*P1A]Sh. 5CK5CKService Wtr Pump [3SWP*P1B]Sh. 5CL5CLService Wtr Pump [3SWP*P1C]
Sh. 5CM5CMService Wtr Pump Sh. 5CN5CNCntmt Recirc Pump [3RSS*P1A]Sh. 5CP5CPCntmt Recirc Pump [3RSS*P1B]
Sh. 5CQ5CQCntmt Recirc Pump [3RSS*P1C]Sh. 5CR5CRCntmt Recirc Pump [3RSS*P1D]Sh. 5CS5CSCharging Pump P3A [3CHS*P3A]
Sh. 5CT5CTCharging Pump P3B [3CHS*P3B]Sh. 5CU5CUCharging Pump P3C (Swing) [3CHS*P3C]Sh. 5CV5CVCharging Pump P3C (Swing) [3CHS*P3C]
Sh. 5DA 5DAReactor Plant Comp Cooling Wtr Pp [3CCP*P1A]Sh. 5DB5DBReactor Plant Comp Cooling Wtr Pp [3CCP*P1B]Sh. 5DC5DCReactor Plant Comp C ooling Wtr Swing Pp [3CCP*P1C]Sh. 5DD5DDReactor Plant Comp C ooling Wtr Swing Pp [3CCP*P1C]Sh. 5DE5DEResidual Heat Removal Pump P1A [3RHS*P1A]Sh. 5DF5DFResidual Heat Removal Pump P1B [3RHS*P1B]
Sh. 5DG5DGQuench Spray Pump P3A (3QSS*P3A)Sh. 5DH5DHQuench Spray Pump P3B (3QSS*P3B)Sh. 5DJ5DJSafety Injection Pump P1A [3SIH*P1A]
Sh. 5DK5DKSafety Injection Pump P1B [3SIH*P1B]Sh. 5DR5DREmer Diesel Gen Bkr [3ENS*ACB-G-A] 15G-14U-2Sh. 5DS5DSEmer Diesel Gen Bkr [3ENS*ACB-G-B] 15G-15U-2Sh. 5DX5DXStm Gen Aux Fdwtr Pp Mot Driven P1A [3FWA*P1A]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-36Rev. 30Sh. 5DY5DYStm Gen Aux Fdwtr Pp Mot Driven P1B [3FWA*P1B]Sh. 5DZ5DZCntrl Bldg Chilled Wtr Chiller CHL1A [3HVK*CHL1A]Sh. 5EA5EACntrl Bldg Chilled Wtr Chiller CHL1B [3HVK*CHL1B]Sh. 5EB5EBEmer Gen Neutral Bk r [3ENS-ACB-GNA] 15G-14U-2NSh. 5EC5ECEmer Gen Neutral Bk r [3ENS-ACB-GNB] 15G-15U-2NSh. 5EX5EXElem Diag Misc Ckts Sta 3VCE Bkr ControlSh. 5EY5EY4.16 kV - US Fdr Bkr [3ENS*ACB-AD]
Sh. 5EZ5EZ4.16 kV - US Fdr Bkr [3ENS*ACB-BD]Sh. 5FD5FDAH D.G. Bkr Backup ProtectionSh. 6A6ATypical 480 V ACB Sh. 6M6MEmer Supply Bkr [3EJS*ACB-AA]Sh. 6N6NEmer Supply Bkr [3EJS*ACB-AB]Sh. 6P6PEmer Supply Bkr [3EJS*AC]
Sh. 6Q6QEmer Supply Bkr [3EJS*ACB-BA]Sh. 6R6REmer Supply Bkr [3EJS*ACB-BB]Sh. 6S6SEmer Supply Bkr [3EJS*ACB-BC]
Sh. 6AG6AGFuel Pool Cooling Pump [3SFC*P1A]Sh. 6AH6AHFuel Pool Cooling Pump [3SFC*P1B]Sh. 6AN6ANPressurizer Htrs BU GP A [3RCS*PH1A]Sh. 6AP6APPressurizer H Ht rs BU GP B [3RCS*PH1A]Sh. 6AV6AVCrdm Shroud [3HVU-FN2A]Sh. 6AW6AWCrdm Shroud Fan [3HVU-FN2B]
Sh. 6AY6AYInst Air Comp Fdr BkrSh. 6BA6BACntmt Air R ecirc Fan [3HVU-FN1A]Sh. 6BB6BBCntmt Air R ecirc Fan [3HVU-FN1B]Sh. 6DD6DDServ Wtr Pp Str and Backwash V [3SWP*STR1A], [3SWP*MOV24A]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-37Rev. 30Sh. 6DE6DEServ Wtr Pp Str and Backwash V [3SWP*STR1B], [3SWP*MOV24B]Sh. 6DF6DFServ Wtr Pp Str and Backwash V [3SWP*STR1C], [3SWP*MOV24C]Sh. 6DG6DGServ Wtr Pp Str and Backwash V [3SWP*STR1D], [3SWP*MOV24D]Sh. 6DX6DXEmer Gen Fuel Oil Xfer pps [3EGF*P1A,C]Sh. 6DY6DYEmer Gen Fuel Oil Xfer Pps [3EGF*P1B,D]Sh. 6GD6GDBoric Acid Transfer Pump [3CHS*P2A]Sh. 6GE6GEBoric Acid Transfer Pump [3CHS*P2B]
Sh. 6GH6GHCharging Pumps Cooling Pumps [3CCE*P1A,B]Sh. 6GT6GTCont Bldg Chilled Wtr Pump [3HVK*P1A]Sh. 6GU6GUCont Bldg Chilled Wtr Pump [3HVK*P1B]
Sh. 6GV6GVCont Bldg Chilled L ube Oil Pumps [3HVK*P3A] & [3HVK*P3B]Sh. 6LD6LDCntmt Recir Wt r Spray Hdr Isol VV20ASh. 6LE6LECntmt Recir Wt r Spray Hdr Isol VV20BSh. 6LF6LFCntmt Recir Wt r Spray Hdr Isol VV20CSh. 6LG6LGCntmt Recir Wt r Spray Hdr Isol VV20DSh. 6LH6LHCont Recir Pump Suct Isol VV23ASh. 6LJ6LJCont Recir Pump Suct Isol VV23BSh. 6LK6LKCont Recir Pmp Suct Isol VV23C Sh. 6LL6LLCont Recir Pump Suct Isol VV23DSh. 6LM6LMRss to Rhr Cross Connect MV8837ASh. 6LN6LNRss to Rhr Cross Connect MV8838ASh. 6LP6LPRss to Rhr Cross Connect MV8837BSh. 6LQ6LQRss to Rhr Cross Connect MV8838BTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-38Rev. 30Sh. 6LS6LSQuench Spray Header Isol Valve (3QSS*MOV34A)Sh. 6LT6LTQuench Spray Header Isol Valve (3QSS*MOV34B)Sh. 6MF6MFRefueling Wtr Stor Tk to SIP [3SIH*MV8806]Sh. 6MG6MGSIP Suction [3SIH*MV8923A]
Sh. 6MH6MHSIP Suction [3SIH*MV8923B]Sh. 6MJ6MJSIP Disch HDR Isol Valve [3SIH*MV8821A]Sh. 6MK6MKSIP Disch Hdr Isol Valve [3SIH*MV8821B]
Sh. 6ML6MLSIP to Cold Legs [3SIH*MV8835]Sh. 6MN6MNHi Press SIP Mini Flow Isol [3SIH*MV8813]Sh. 6MP6MPSuction Hdr Cross Connection [3SIH*MV8807A]
Sh. 6MQ6MQSuction Header Cr oss Connection [3SIH*MV8807B]Sh. 6MR6MRSIP to Hot Legs [3SIH*MV8802A]Sh. 6MS6MSSIP to Hot Legs [3SIH*MV8802B]
Sh. 6MV6MVBoron Inj Tk Disch Isol [3SIH*MV8801A]Sh. 6MW6MWBoron Inj Tk Disch Isol (3SIH*MV8801B)Sh. 6MX6MXRwst to Rhr P1A Isolation 3SIL*MV8812A Sh. 6MY6MYRwst to Rhr P1B Isolation [3SIL*MV8812B]Sh. 6MZ6MZRhr to Cold Leg Isol [3SIL*MV8809A]Sh. 6NA6NARhr To Cold Leg Isol [3SIL*MV8809B]
Sh. 6NB6NBAccumulator Isolation [3SIL* MV8808A]Sh. 6NC6NCAccumulator Isolation [3SIL*MV8808B]Sh. 6ND6NDAccumulator Isolation [3SIL*MV8808C]
Sh. 6NE6NEAccumulator Isolation [3SIL*MV8808D]Sh. 6NF6NFRhr P1A to Charging Pump [3SIL*MV8804A]Sh. 6NG6NGRhr P1B to Charging Pump [3SIL*MOV8804B]Sh. 6NH6NHResidual Ht Removal to Hot Legs [3SIL*MV8840]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-39Rev. 30Sh. 6NJ6NJCross Connection Is olation VV22A (3RHS*MV8716A)Sh. 6NK6NKCross Connection Isolation Valve [3RHS*MV8716B]Sh. 6NU6NUHI Press SIP Mini Flow Isol [3SIH*MV8814]Sh. 6NV6NVHi Press SIP Mini Flow Isol [3SIH*MV8920]Sh. 6NW6NWSI Charging Sucti on Cross Connection [3SIH*MV8924]Sh. 6PA6PARcp Seal Wtr Isol [3CHS*MV8109A]Sh. 6PB6PBRcp Seal Wtr Isol [3CHS*MV8109B]
Sh. 6PC6PCRcp Seal Wtr Isol [3CHS*MV8109C]Sh. 6PD6PDRcp Seal Wtr Isol [3CHS*MV8109D]Sh. 6PE6PECharging Pump Mini-flow Isol Valve [3CHS*V8110]Sh. 6PF6PFCharging Pump Mini-flow Isol Valve [3CHS*MV8111]Sh. 6PG6PGCharging Pump to Rx Clt Sys Isol Valve [3CHS*MV8105]Sh. 6PH6PHCharging Pump to Rx Clt Sys Isol Valve [3CHS*MV8106]
Sh. 6PJ6PJBoric Acid Fltr to Charging Pmp Valve [3CHS*MV8104]Sh. 6PK6PKVol Cont Tk Outlet Isol VV112BSh. 6PL6PLVol Cont Tk Outlet Isol VV112C Sh. 6PM6PMRefueling Wtr Stor Tk to Charging Pmp VV112D Sh. 6PN6PNRefueling Wtr Stor Tk to Charging Pmp V112ESh. 6PP6PPRx Clt Pmp Seal Wtr Isol [3CHS*MV8112]Sh. 6PQ6PQRx Clt Pmp S eal Wtr Isol [3CHS*MV8100]Sh. 6QR6QRRHS Inlet Isolation Valve [RHS*MV8701C]Sh. 6QS6QSRHS Inlet Isolation Valve [3RHSMV8702C]Sh. 6QT6QTRHS Inlet Isolation Valve 3RHSMV8701ASh. 6QU6QURHS Inlet Isolation Valve 3RHSMV8702BSh. 6QV6QVRHS Inlet Isolation Valve 3RHSMV8701B Sh. 6QW6QWRHR Inlet Isolation Valve 3RHS*MC8702ATABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-40Rev. 30Sh. 6QX6QXRHR Pmp P1A Mini-flow Recirc VV610Sh. 6QY6QYRHR Pmp P1B Mini-flow Recirc VV611Sh. 6RJ6RJCntrl Rm Area A/C Units ACU1A, 1BSh. 6RL6RLBattery Rms 1, 2&5 ACU3A, 3B Sh. 6RM6RMCable Spread & Swgr, Equip Rm ACU4A & 4BSh. 6RP6RPCntmt Bldg Chill Wtr Pp Aux Ckt ASh. 6RQ6RQCntmt Bldg Chill Wtr Pp Aux Ckt B Sh. 6RR6RRCntmt Bldg Air Cond Booster Pp (3SWP*P2A, B)Sh. 6RV6RVChiller Equip Space Supply Fan - FN2A, FN2BSh. 6RW6RWChiller Equip Space Exhaust Fan - FN7A, 7B Sh. 6RX6RXCntmt Bldg Emerg Vent Fa n - FN1A & Inlet Damper MOD33ASh. 6RY6RYCntmt Bldg Emerg Vent Fan - FN1B & Inlet Damper MOD33BSh. 6SB6SBAux Bldg Exh Fan & Assoc Dmprs
[3HVK*FNGA, AOD20A & 26A]Sh. 6SC6SCAux Bldg Fltr Exh Exh Fan & Assoc DmprsSh. 6SD6SDChg Pp Cub Sply Fan & Assoc Dmprs Sh. 6SE6SEChg Pp Cub Sply Fan & Assoc DmprsSh. 6SF6SFChg Pp Exh Fan & Assoc DmprsSh. 6SG6SGChg Pp Exh Fan & Assoc Dmprs Sh. 6SH6SHFuel Bldg Exh Fan & Assoc DmprsSh. 6SJ6SJFuel Bldg Exh Fan & Assoc DmprsSh. 6SK6SKSlcr Exh Fan & Assoc Dmprs Sh. 6SL6SLSlcr Exh Fan & Assoc DmprsSh. 6SM6SMMCC, Rod Cont & Cable Vault Area ACUSh. 6TD6TDPressurizer Relief Isol Valve [3RCS*MV8000A]Sh. 6TE6TEPressurizer Relief Isol Valve [3RCS*MV8000B]Sh. 6TF6TFReac Clnt LP1 Hot Leg Stop Valve (3RCS*MV8001A)TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-41Rev. 30Sh. 6TG6TGReac Clnt LP2 Hot Leg Stop Valve (3RCS*MV8001B)Sh. 6TH6THReac Clnt LP3 Hot Leg Stop Valve (3RCS*MV8001C)Sh. 6TJ6TJReac Clnt LP4 Hot Leg Stop Valve (3RCS*MV8001D)Sh. 6TK6TKReac Clnt LP1 Cold Leg Stop Valve (3RCS*MV8002A)
Sh. 6TL6TLReac Clnt LP2 Cold Leg Stop Valve (3RCS*MV8002B)Sh. 6TM6TMReac Clnt LP3 Cold Leg Stop Valve (3RCS*MV8002C)Sh. 6TN6TNReac Clnt LP4 Cold Leg Stop Valve (3RCS*MV8002D)
Sh. 6TP6TPReactor Coolant Loop 1 Bypass Leg Stop Valve (3RCS*MV8003A)Sh. 6TQ6TQReactor Coolant Loop 2 Bypass Leg Stop Valve (3RCS*MV8003B)Sh. 6TR6TRReactor Coolant Loop 3 Bypass Leg Stop Valve (3RCS*MV8003C)Sh. 6TS6TSReactor Coolant Loop 4 Bypass Leg Stop Valve (3RCS*MV8003D)Sh. 6TW6TWSwgr Area & Battery Rooms Supply Fans Sh. 6TX6TXBattery Rooms 1 & 2 Exhaust FansSh. 6TY6TYBattery Rooms 3 & 4 Exhaust Fans Sh. 6TZ6TZBattery Room 5 Exhaust Fan Sh. 6VK6VKInstrument RK & Com puter Rooms A/C Unit 3HVC*ACU2ASh. 6VL6VLInstrument RK & Com puter Rooms A/C Unit 3HVC*ACU2BSh. 6VM6VMStm Gen Aux Fdwtr Isol Valve [3FWA*MOV35A]
Sh. 6VN6VNStm Gen Aux Fdwtr Isol Valve [3FWA*MOV35B]Sh. 6VP6VPStm Gen Aux Fdwtr Isol Valve [3FWA*MOV35C]Sh. 6VQ6VQStm Gen Aux Fdwtr Isol Valve [3FWA*MOV35D]Sh. 6VV6VVAux Fdwtr Pump Turb Stm Supply Non-return Valve
[3MSS*MOV17A]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-42Rev. 30Sh. 6VW6VWAux Fdwtr Pump Turb Stm Supply Non-return Valve
[3MSS*MOV17B]Sh. 6VX6VXAux Fdwtr Pump Turb Stm Supply Non-return Valve
[3MSS*MOV17D]Sh. 6VY6VYMain Stm Press Rel Isol Valve [3MSS*MOV18A] (AO)Sh. 6VZ6VZMain Stm Press Rel Isol Valve [3MSS*MOV18B] (BP)Sh. 6WA6WAMain Stm Press Rel Isol Valve [3MSS*MOV18C] (CO)Sh. 6WB6WBMain Stm Press Rel Isol Valve [3MSS*MOV18D] (DP)Sh. 6ZH6ZHEGE MCC Fdr Bkr [3EHS*MCC1A1]Sh. 6ZJ6ZJCont Bldg MCC Fdr Bkr [3EHS*MCC1A2]
Sh. 6ZK6ZKAux Bldg MCC Fdr Bkr [3EHS*MCC3A1]Sh. 6ZL6ZLSfgrds Area MCC Fdr Bkr [3EHS*MCC1A4]Sh. 6ZM6ZMCirc & Serv Wtr Pp Hse MCC Fdr Bkr [3EHS*MCC1A5]
Sh. 6ZN6ZNEGE MCC Fdr Bkr [3EHS*MCC1B1]Sh. 6ZP6ZPCont Bldg MCC Fdr Bkr [3EHS*MCC1B2]Sh. 6ZQ6ZQAux Bldg MCC Fdr Bkr [3EHS*MCC3B1]
Sh. 6ZR6ZRSfgrds Area MCC Fdr Bkr [3EHS*MCC1B4]Sh. 6ZS6ZSCirc & Serv Pp Hse MCC Fdr Bkr [3EHS*MCC1B5]Sh. 6ZT6ZTRod Contmt Area MCC Fdr Bkr [3EHS*MCC2A1]
Sh. 6ZU6ZURod Contmt Area MCC Fdr Bkr [3EHS*MCC2B1]Sh. 6ZV6ZVTurb Bldg MCC Fdr Bkr [3EHS*MCC2A2]Sh. 6AAA6AAAContmt Recirc Cir Out VV57A Sh. 6AAB6AABContmt Recirc Cir Out VV57B Sh. 6AAC6AACContmt Recirc Cir Out VV57CSh. AADAADContmt Recirc Cir Out VV57DSh. 6AAF6AAFContmt Recirc Cir Supply VV54ASh. 6AAG6AAGContmt Recirc Cir Supply VV54BTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-43Rev. 30Sh. 6AAH6AAHContmt Recirc Cir Supply VV54CSh. 6AAJ6AAJContmt Recirc Cir Supply VV54DSh. 6AAK6AAKRx Plt Com pnt Clg Hx Supply VV50ASh. 6AAL6AALRx Plt Com pnt Clg Hx Supply VV50BSh. 6AAM6AAMTPCCW Hx Inlet Valve [3SWP*MOV71A]Sh. 6AAN6AANTPCCW Hx Inlet Valve [3SWP*MOV71B]Sh. 6AAU6AAUService Wtr Pmp Disch Valve [3SWP*MOV102A]Sh. 6AAV6AAVService Wtr Pmp Disch Valve [3SWP*MOV102B]Sh. 6AAW6AAWService Wtr Pmp Disch Valve [3SWP*MOV102C]Sh. 6AAX6AAXService Wtr Pmp Disch Valve [3SWP*MOV102D]Sh. 6AAY6AAYSI Pump CoolingSh. 6ABG6ABGDiesel Gen "A" Enc Ve ntilation Supply Fan [3HVP*FN1A,C]Sh. 6ABH6ABHDiesel Gen "B" Enc Ve ntilation Supply Fan [3HVP*FN1B,D]Sh. 6ABP6ABP480 V Motor Cont Emer Gen A&B Misc Equipment FdrSh. 6ABZ6ABZEmer Gen Crankcase Vac Pmp [3EGD*P1A, B]Sh. 6ACA6ACAEmerg Gen "A" Air Compr [3EGA-C1A, C2A]
Sh. 6ACB6ACBEmer Gen "B" Air Compr [3EGA-C1B, C2B]Sh. 6ACL6ACL480 V Mc Diesel Gen Enc+ "A" Ventilation Dampers
[3HVP*MOD20A, 23A, 26A, 20C]Sh. 6ACM6ACM480 V Mc Diesel Gen Enc+ "B" Ventilation Dampers
[3HVP*MOD20B, 23B, 26B, 20D]Sh. 6ACN6ACNDiesel Gen "A" Enc Ve ntilation Supply Fan [3HVP*FN1C]Sh. 6ACP6ACPDiesel Gen "B" Enc Ve ntilation Supply Fan [3HVP*FN1D]Sh. 6ACZ6ACZService Water Pum phouse Exh Fans & Inlet Dampers 3HVY*FN2A&B; 3HVY*AOD23A&BSh. 6ADH6ADHMain Steam Valve Bldg Exh Fan 1BSh. 6ADR6ADRSI, QS & RHR Pps Area A/C Units [3HVQ*ACUS1A,B]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-44Rev. 30Sh. 6ADS6ADSContmnt Recirc Pp & Clr Area A/C Units [3HVQ*ACUS2A,B]Sh. 6ADU6ADUAux Fdwtr Pp Area Emer Sply Fans [3HVQ*FN5A,B]Sh. 6ADV6ADVAux Fdwtr Pp Area Emer Vent Dmprs
[3HVQ*MOD26A1, B1,C1]Sh. 6ADW6ADWAux Fdwtr Pp Area Emer Exh Fans [3HVQ*FN6A,B]Sh. 6ADX6ADXCharging Pp Aux Lube Oil Pp (3CHS*P6C)Sh. 6AEA6AEACharging Pp Area Inlet & Outlet Air Dampers
[3HVR*MOD49A,50A, 49B*50B]Sh. 6AEB6AEBCharging Pp Area Inlet & Outlet Air Dampers
[3HVR*MOD49C1, 50C1,49C2,50C2]Sh. 6AEC6AECElem Diag 120V ac Aux & Fuel Bldg Vent Dampers
[3HVR*MOD28A,28B]Sh. 6AFA6AFABus 32T Tie Bk to Bus 32S 32T11-2 [3EJS*ACB-T11]
Sh. 6AFB6AFBBus 32S Tie Bkr to Bus 32R 32S11-2 [3EJS*ACB-S11]Sh. 6AFC6AFCBus 32U Tie Bkr to Bus 32V 32U11-2 [3EJS*ACB-U11]Sh. 6AFD6AFDBus 32V Tie Bkr to Bus 32W 32V11-2 [3EJS*ACB-V11]Sh. 6AFE6AFEAux Bldg Filter Unit [3HVR*FLT1A]Sh. 6AFF6AFFAux Bldg Filter Unit [3HVR*FLT1B]Sh. 6AFG6AFGFuel Bldg Filter Unit [3HVR*FLT2A]Sh. 6AFH6AFHFuel Bldg Filter Unit [HVR*FLT2]Sh. 6AFK6AFKCont Bldg Chiller Condenser Inlet Temp Cont Valves (3SWP*TV35A&B)Sh. 6AFL6AFLCntmt Recirc Pp Miniflow Valve [3RSS*MOV38A]Sh. 6AFM6AFMCntmt Recirc Pp Miniflow Valve [3RSS*MOV38B]Sh. 6AFZ6AFZContainment Open Pressure Tap Isolation Valve
[3LMS*MOV40A]Sh. 6AGA6AGAContainment Open Pressure tap Isolation Valve
[3LMS*MOV40B]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-45Rev. 30Sh. 6AGB6AGBContainment Open Pressure Tap Isolation Valve
[3LMS*MOV40C]Sh. 6AGC6AGCContainment Open Pressure Tap Isolation Valve
[3LMS*MOV40D]Sh. 6AGD6AGDAux Fd Pps Area Emer Vent Dmprs
[3HVQ*MOD26A2, B2, C2]Sh. 6AGF6AGFCirculating Water Valve 3SWP*MOV115ASh. 6AGG6AGGCirculating Water Pump Lube Water Valve 3SWP*MOV115BSh. 6AGH6AGHContainment Instr Air Isolation Valve [3IAS*MOV72]Sh. 6AGX6AGXFuel Bldg Vent Dampers [3HVR*MOD72A,B]Sh. 6AGY6AGYAux Fdwtr Pp Area Emer Sply FanSh. 6AHB6AHBSuppl Leak Coll Rel Sys Flt HtrsSh. 6AHL6AHLContmt ATM Monit Inside Contmt Isol Valve [3CMS*MOV24]Sh. 6AJH6AJHElem Diag 480 V MC Reactor Vessel to Excess Ltdn Valve (3RCS*V8098)Sh. 6AJJ6AJJElem Diag 480 V MC Boric Acid Gravity Feed Valve (3CHS*V8507A)Sh. 6AJK6AJKElem Diag 480 V MC Boric Acid Gravity Feed Valve (3CHS*MV8507B)Sh. 6AJL6AJLElem Diag 480 V MC LPSI Charging Pp Suct Valve (3CHS*V8468A)Sh. 6AJM6AJMElem Diag 480 V MC LPSI Charging Pipe Suct Valve (3CHS*MV8468B)Sh. 6AJN6AJNElem Diag 480 V MC Charging Pp Recirc Valve (3CHS*MV8111B)Sh. 6AJP6AJPElem Diag 480 V MC Charging Pp Recirc (3CHS*MV8111C)Sh. 6AJQ6AJQElem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV8438A)Sh. 6AJR6AJRElem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV8438B)TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-46Rev. 30Sh. 6AJS6AJSElem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV-8438C)Sh. 6AJU6AJUElem Diag 480 V MC Contmt Air Recirc Coil Sply Valve (CCP*MOV222)Sh. 6AJV6AJVElem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV223)Sh. 6AJW6AJWElem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV224)Sh. 6AJX6AJXElem Diag 480 V Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV225)Sh. 6AJY6AJYElem Diag 480 V Sply Valv e Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV226)Sh. 6AJZ6AJZElem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV227)Sh. 6AKA6AKAElem Diag 480 V MC Contmt Air Recirc Coil Sply Valve (3CCP*MOV228)Sh. 6AKB6AKBElem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV229)Sh. 6AKC6AKCElem Diag 480 V MC Charging Hdr Isol Valve (3CHSX-MV8116)Sh. 6AKE6AKEContmt Isol Valve (3CCP*MOV45A)Sh. 6AKF6AKFContmt Isol Valve (3CCP*MOV45B)Sh. 6AKG6AKGContmt Isol Valve (3CCP*MOV48A)Sh. 6AKH6AKHContmt Isol Valve (3CCP*MOV48B)Sh. 6AKJ6AKJContmt Isol Valve (3CCP*MOV49A)Sh. 6AKK6AKKContmt Isol Valve (3CCP*MOV49B)Sh. 6AKL6AKLElem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74ASh. 6AKM6AKMElem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74BTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-47Rev. 30Sh. 6AKN6AKNElem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74CSh. 6AKP6AKPElem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74DSh. 6AKX6AKXElem Diag 480 V ESF Bldg MCC Fdr Bkr (3EHS*MCC1B3)Sh. 6AKY6AKYElem Diag 480 V Em er Supply Bkr (3EJS*ACB-AD)Sh. 6AKZ6AKZElem Diag 480 V Em er Supply Bkr [3EJS*ACB-BD]Sh. 6ALA6ALAElem Diag 480 V Bus 32Y Tie to Bus 32R 32T 4A-2
[3EJS*ACB T4A]Sh. 6ALB6ALBElem Diag 480 V Bus 32X Tie Bkr to Bus 32W 32T4B-2
[3ETS*ACB-T4B]Sh. 6ALC6ALCElem Diag 480 V US Spare MCC Fdr Bkr [3EHS*MCC4A1]Sh. 6ALD6ALDElem Diag 480 V US Spare MCC Fdr Bkr
[3EHS*Bkr MCC4B1]Sh. 6ALG6ALGElem Diag 480 V MC (3SWP-3A & B)Sh. 6ALH6ALHElem Diag 480 V MC (3STR 2A & B)Sh. 6AMB6AMBElem Diag 480 V MC Cold Shutdown Air Compressor [3IAS-C2A]Sh. 6AMC6AMCElem Diag 480 V MC Cold Shutdown Air Compressor [3IAS-C2B]Sh. 6AMG6AMGMn Stm Valve Bldg Vent Exh Fan DampersSh. 6AMH6AMHMn Stm Valve Bldg Vent Exh Fan DampersSh. 6AMJ6AMJMn Stm Valve Bldg Vent Inlet DampersSh. 6AMK6AMKMn Stm Valve Bldg Vent Inlet DampersSh. 6BAA6BAAAir Sample PumpSh. 6BAB6BABElem Diag 480 V MC Charging Pp Miniflow Cntrl Valve
[3CHS*MV8512A]Sh. 6BAC6BACElem Diag 480 V MC Charging Pp Miniflow Cntrl Valve
[3CHS*MV8512B]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-48Rev. 30Sh. 6BAD6BADElem Diag 480 V MC Charging Pp Miniflow Cntrl Valve [3CHS*MV8511A]Sh. 6BAE6BAEElem Diag 480 V MC Charging Pp Miniflow Cntrl Valve [3CHS*MV8511B]Sh. 6BAF6BAFElem Diag 480 V Fuel Bldg Fltr Exh Fan [3HVR*FN10A2]Sh. 6BAG6BAGElem Diag 480 V Fuel Bldg Fltr Exh Fan [3HVR*FN10B2]Sh. 6BAN6BANElem Diag. Flow to Monitor for Pzr SVsSh. 6BAW6BAWCRDM Fan 3-HVU-FN2A PENSEC PROT BkrSh. 6BAX6BAXCRDM Fan 3-HVU-FN2B PENSEC PROT BkrSh. 6BAZ6BAZCont Struc Air Recirc FN 3-HVU-FN1ASh. 6BBA6BBACont Struc Air Recirc FN 3-HVU-FN 1B PENE Sec Prot BkrSh. 6BBE6BBE120 V DC Halon Disch TimerSh. 7J7J4.16 kV Bus 34C [3NNS-SWG-A] Aux Ckt Sh. 7L7L4.16 kV Bus 34D [3ENS*SWG-B] Aux CktSh. 7Q7QEmer Diesel Gen Bkr
[3ENS*ACB-G-A] Aux Circuit 15G-14U-2Sh. 7R7REmer Diesel Gen Bkr
[3ENS*ACB-G-B] Aux Circuit 15G-15U-2Sh. 7W7WEmerg Diesel Clr Outlet Valves [3SWP*AOV39A&B]Sh. 7AD7ADStm Jet Air Ejector Stm Valves [3ASS-AOV22A,B]Sh. 7AJ7AJAux Fdwtr Alt Suct Valves [3FWA*AOV23A,23B]Sh. 7AL7ALDwst Htr Circ Line Isol Valve [3FWA*AOV25, 26]
Sh. 7AM7AMRefuel Wtr Recirc Pp Suct Isol Valve (3QSS*AOV27, 28)Sh. 7AN7ANAccumulator Nit Test Line Isol SIP Hot Leg Test Line
[3SIL*CV8880] [SIL*CV8825]Sh. 7AP7APTest Line Header
[3SIH*CV8964] & [3SIH*8871]Sh. 7AQ7AQAccumulator Gas Sply & Vent [3SIL*AV8875A] Accumulator Fill Line Isol [3SIL*AV8878A]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-49Rev. 30Sh. 7AR7ARAccumulator Gas Sply & Vent [3SIL*AV8875B]Accumulator Fill Line Isol [3SIL*AV8878B]Sh. 7AS7ASAccumulator Gas Sply & Vent [3SIL*AV8875A]Accumulator Fill Line Isol [3SIL*AV8878C]Sh. 7AT7ATAccumulator Gas Sply & Vent [3SIL*AV8875D]Accumulator Fill Line Isol [3SIL*AV8878D]Sh. 7AU7AURHR Pmp Hot Leg Test Line [3SIL*CV8890A, CV8890B]
Sh. 7AY7AYMn Stm Isol Trip Valve [3MSS*CTV27A]Sh. 7AZ7AZMn Stm Isol Trip Valve [3MSS*CTV27B]Sh. 7BA7BAMn Stm Isol Trip Valve [3MSS*CTV27C]
Sh. 7BB7BBMn Stm Isol Trip Valve [3MSS*CTV27D]Sh. 7BC7BCMn Stm Line Isol Bypass Valve [3MSS*HV28A]Sh. 7BD7BDMn Stm Line Isol Bypass Valve [3MSS*HV28B]
Sh. 7BE7BEMn Stm Line Isol Bypass Valve [3MSS*HV28C]Sh. 7BF7BFMn Stm Line Isol Bypass Valve [3MSS*HV28D]Sh. 7BG7BGSIP Hot Leg Test Li ne [3SIH*CV8824] & [3SIH* CV8881]Sh. 7BH7BHStm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22A]Sh. 7BJ7BJStm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22B]Sh. 7BK7BKStm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22C]Sh. 7BL7BLStm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22D]Sh. 7BP7BPRx Pl Gaseous Drs Inside Isol Valve [3DGS*CTV24]Sh. 7BQ7BQRx Pl Gaseous Drs Outside Cntmt Isol Valve [3DGS*CTV25]Sh. 7BR7BRCntmt Drs Isol Valv es [3DAS**CTV24] & [3DAS*CTV25]Sh. 7BT7BTCntmt Drs Isol Valve [3DAS*CTV25]Sh. 7CM7CMChg to Rx Cool Sys Isol Valves [3CHS*AV8147, 8146]Sh. 7CS7CSRx Coolant Makeup C ont Aux, Circ, RPS-RAK Aux A,CTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-50Rev. 30Sh. 7CT7CTBA Inj Valve to BA Blender VV 110A BA Mkup Inj Valve to Charging Pp Hdr VV 110B Sh. 7CU7CUBa Mkup Wtr Inj Valve to BA Blender VV 111A BA Dil Inj Value to Vol Cntrl Tk VV 111BSh. 7CW7CWVol Cntrl Tk Isol & Inert Valves [3CHS*AV8101, LCV112A]
Elem Diag 480 VSh. 7CY7CYEmerg Bux 32R, S, T, Y Aux Ckt Elem Diag Sh. 7CZ7CZ480 V Emerg Bus Bus 32U, V, W, X Aux CktsSh. 7DF7DFCntmt Purge Norm & Fltr Exh DmprsSh. 7DG7DGChg Pp & CCW Norm & Fltr Exh Dmprs Sh. 7DH7DHWaste Dspl & Pipe Tunnel Norm & Fltr Exh DmprsSh. 7DJ7DJAux Bldg Norm & Fltr Exh DmprsSh. 7DK7DKAux Bldg Norm & Fltr Exh Dmprs Sh. 7DL7DLFuel Bldg Fltr & Bypass DmprsSh. 7DV7DVReac Coolant Letdn Drn Valves & Excess Letdn Isol Valves [3RCS*AV8032, 8153]Sh. 7DW7DWPressurizer Power Power Relief Valves [3RCS*PCV455A, 456]Sh. 7DW17DW1Pressurizer Pwr Relief U/V 3RCS*456Sh. 7DX7DXLetdown Line Isol Valves [3RCS*LCV459, 460]
Sh. 7DZ7DZReactor Coolant System Spray VV 8145 Pressurizer Relief Tank Press VV 469Sh. 7EK7EKChg Pump Test Line Isol [3SIH*CV8843] & [3SIH*AV8882]Sh. 7EM7EMAccumulator Fill Line 3SIH*CV8888 Skip Cold Leg Test Line 3SIH*CV8823Sh. 7EZ7EZPri Gr Wtr Cntmt Isol Valves [3PGS*CV8026, 8046]Sh. 7FG7FG4.16 kV Bus [3ENS*SWG-A] Aux CircuitSh. 7FH7FH4.16 kV Bus [3ENS*SWG-B] Aux Circuit Sh. 7FP7FPMain Steam Line Drains [3DTM*AOV63, 64]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-51Rev. 30Sh. 7FS7FSMain Steam Line Drains [3DTM*AOV29A,B]Sh. 7FT7FTMain Steam Line Drains [3DTM*AOV29C, D]Sh. 7FU7FUMain Steam Line Drains [3DTM*AOV61A, B]Sh. 7FV7FVMain Steam Line Drains [3DTM*AOV61C, D]
Sh. 7HK7HKSG Fdwtr Bypass Valve & Cntrl Valves Auxiliary Ckt [3FWS*LV550]Sh. 7HL7HLSG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV560]
Sh. 7HM7HMSG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV570]Sh. 7HN7HNSG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV580]Sh. 7HP7HPSG Mn Fdwtr Flow Cntrl Valve [3FWS-FCV510]
Sh. 7HQ 7HQSG Mn Fdwtr Flow Cntrl Valve [3FWS-FCV520]Sh. 7HR7HRSG Mn Fdwtr Flow Cntrl Valve (3FWS*FCV530)Sh. 7HS7HSSG Mn Fdwtr Flow Cntrl Valve 3FWS*FCV540Sh. 7HX7HXTurb Bypass Cntrl Sys Aux CktSh. 7JC7JCLetdown Orifice Isol Valve [3CHS*AV8149A]Sh. 7JD7JDLetdown Orifice Isol Valve [3CHS*AV8149B]
Sh. 7JE7JELetdown Orifice Isol Valve [3CHS*AV8149C]Sh. 7JK7JKChg Pps Clrs Out Crossover Valves [3CCE*AOV26A,B]Sh. 7JL7JLChg Pmps Disch Crossover Valves [3CCE*AOV30A,B]Sh. 7JN7JNSG Fdwtr Isol Valve [3FWS*CTV41A]Sh. 7JP7JPSG Fdwtr Isol Valve [FWS*CTV41B]Sh. 7JQ7JQSG Fdwtr Isol Valve [3FWS*CTV41C]
Sh. 7JR7JRSG Fdwtr Isol Valve [3FWS*CTV41D]Sh. 7JY7JYCntmt Inst Air Isol Valve [3IAS*CTV15]Sh. 7KJ7KJ120 VAC Cont Bldg Aux CktsSh. 7KK7KKCVS Isol Valves (3CVS*CTV20A,B)Sh. 7KL7KLCVS Isol Valves (3CVS*CTV21A,B)TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-52Rev. 30Sh. 7KQ7KQServ Wtr Pps Train A & B Aux Ckts Sh. 7KT7KTSG #1 Blwdwn Sample Isol VV19ASh. 7KU 7KU SG #2 Blwdwn Isol VV19BSh. 7KV7KVSG #3 Blwdwn Isol VV19C Sh. 7KW7KWSG #4 Blwdwn Sample Isol VV19DSh. 7KX7KXPzr Vapor Space Sample Isol VVs 20&21Sh. 7KZ7KZRx Clnt Hot Leg Sample Isol VVs 26&27 Sh. 7LA7LARx Clnt Cold Leg Sample Isol VVs 29&30Sh. 7LB7LBSI Inj Accumulator Sa mple Isol VVs 32&33 Aux CircuitSh. 7LC7LCPzr Rlf Tk Gas Space Sample Isol Valves (3SSR*CV8026, 8025)Sh. 7LX7LXNon-Sfty Hdr Sply Rtn Valves [3CCP*AOV10A, B,&19A,B]Sh. 7MA7MACntmt Isol Valves [3CCP*CTV49A,B]Sh. 7MB7MBResidual Heat Removal Outlet Valves [3CCP*AOV66A&B]Sh. 7MC7MCThermal Barrier Coolant Return Valves [3CCP*AOV178A&B]Sh. 7MD7MDThermal Barrier Coolant Return Valves [3CCP*AOV178C&D]Sh. 7ME7MEComp Cooling Cross Conn VVs 179A&B Sh. 7MF7MFComp Cooling Cross Conn VVs 180A&BSh. 7MG7MGNon-Sfty Hdr Sply & Rtn Values
[3CCP*AOV194A, B, & 197A,B]Sh. 7MW7MWCtmt Spray Test Inter CircuitarySh. 7MX7MXDiesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)Sh. 7MY7MYDiesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)Sh. 7MZ7MZDiesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)Sh. 7NA7NA125V Hx Temp Cont VIV 3CCP*SOU32A & 32C-1TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-53Rev. 30Sh. 7NB7NB125V Hx Temp Cont VIV 3CCP*SOU32B & 32C-2Sh. 7PB7PBControl Room Ventilation Outlet Air Isol VVs 20,21 [HVA*AOV20,21]Sh. 7PC7PCControl Room Ventilation Outlet Air Isol VVs 22,23
[HVC*AOV22,23]Sh. 7PD7PDControl Room Ventilation Outlet Air Isol VVs 25,26
[HVC*AOV25,26]Sh. 7PJ7PJPurge Supply Fan Inlet Damper AOD135 Emerg Vent Air Return Damper AOD119Sh. 7PK7PKCntrl Bldg Makeup Ai r Dmprs, (3HVC*AOD 27A,B)Sh. 7PL7PLAir Stor Tk 1A Outlet VVs SOV 74A & 74BSh. 7QD7QDTurbine-Driven Aux Fdwtr Pp Aux Oil Pp (3FWL*PS), Trip & Throttle Valve (3MSS*M5V5)Sh. 7QE7QEMn Stm Rel & Drain Valves [3MSS*PV20A, 3SVV-AOV20A]Sh. 7QF7QFMn Stm Rel & Drain Valves [3MSS*PV20B, 3SVV-AOV20B]Sh. 7QG7QGMn Stm Rel & Drain Valves [3MSS*PV20C, 3SVV-AOV20C]
Sh. 7QH7QHMn Stm Rel & Drain Valves [3MSS*PV20D, 3SVV-AOV20D]Sh. 7QM7QMChilled Wtr Cntmt Isol Valves [3CDS*CTV38A,B]Sh. 7QN7QNChilled Wtr Cntmt Isol Valves [3CDS*CTV39A,B]Sh. 7QP7QPChilled Wtr Cntmt Isol Valves [3CDS*CTV40A,B]Sh. 7QT7QTReactor Coolant Letdown Valves [3CHS*TCV12G, AV8143]Sh. 7QU7QULetdown Line Isol Valves [3CHS*CV81528160]
Sh. 7QV7QVAux Bldg Inlet Dampers [3HVR*AOD33A, 35A]Sh. 7QW7QWAux Bldg Inlet Da mpers [3HVR*AOD33B, 35B]Sh. 7QX7QXPipe Tunnel In let Dampers [3HVR*AOD85,86]Sh. 7QY7QYCntmt Purge Inlet Da mpers [3HVR*AOD55A/B & 174A/B]Sh. 7QZ7QZChg & Compnt Clg Pps Temp Cntrl Dmprs
[3HVR*AOD45B1, B2, C1, C2]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-54Rev. 30Sh. 7RB7RBReac Clnt Loop Stop Valve Aux Buffering CircuitSh. 7RC7RCReac Clnt Loop Stop Valve Aux Buffering CircuitSh. 7RF7RFTDAFW Pmp Mtr Speed changr 3FWA*M7Sh. 7RG7RGESF Bldg Vent Dampers Sh. 7RH7RHESF Bldg Vent DampersSh. 7RK7RKContainment Purge Inlet & Outlet Isol Valves [3HVU*CTV32A&B], [3HVU*CTV33A&B]Sh. 7RM7RMContainment Atmosphere MonitoringSh. 7RZ7RZSG Level Aux Ckts: Mn Turb TripSh. 7SC7SCSG Chemical Feed Pp Isolation Valves [3SGF*AOV24A&B]Sh. 7SD7SDSG Chemical Feed Pp Isol Valve [3SGF*AOV24C&D]Sh. 7SF7SFMisc Level Ind Lights DAS SystemSh. 7SH7SHRx Pl Gas Vents Hdr Cntrl Trip Valves (3VRS*CTV20,21)Sh. 7SM7SMGas Waste to Unit 1 St ack Isol Dmprs (3GWS*AOD78A,B)Sh. 7SP7SPPzr Relief Tk Nitrogen Sply Isol Valves (3GSN*CV8033, CTV105)Sh. 7SS7SSESF Manual Actuation CktsSh. 7ST7STESF Manual Reset CktsSh. 7SX7SX4.16 kV Bus 34C (3ENS*SWG-A) Undervoltage (Hi STPT) Trip CktSh. 7SY7SY4.16 kV Bus 34D (3ENS*SWG-B) Undervoltage (Hi STPT) Trip CktSh. 7SZ7SZ4.16 kV Bus 34C&D (3 ENS*SWG-A&B) Undervoltage (Hi STPT) RelaysSh. 7TA7TACntrl Bldg Isol (Train A) Sh. 1Sh. 7TB7TBCntrl Bldg Isol (Train A) Sh. 2Sh. 7TC7TCCntrl Bldg Isol (Train B) Sh. 1 Sh. 7TD7TDCntrl Bldg Isol (Train B) Sh. 2TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-55Rev. 30Sh. 7TN7TNContainment Fire Protection Water Isolation Valves 3FPW &
CTV 48,49Sh. 7TP7TPESF Aux Relays (Train A)Sh. 7TQ7TQESF Aux Relays (Train A)
Sh. 7TR7TRESF Aux Relays (Train B)Sh. 7TS7TSESF Aux Relays (Train B)Sh. 7TY7TYAccumulator N2 Line Isol Valve 3SIL*CVS968Sh. 7TZ7TZChilled Wtr Cntmt Isol Valves [3CDS*CTV91A,B]Sh. 7UF7UFAuto and Manual Rod Withdrawal BlockSh. 7UK7UKAccumulator Gas Sply & Vent Valves (3SIL*SV8875F&G)
Sh. 7UL7ULAccumulator Gas Sply & Vent Valves (3SIL*SV8875F&G)Sh. 7UM7UMRx Vessel Head Vent Isol Valve (3RCS*SV8095A1B)Sh. 7UN7UNRx Vessel Head Vent Isol Valve (3RCS*SV0896A1B)
Sh. 7UR7URElem Diag Water Feed to Chlorination System
[3WTC*AOV25A&B]Sh. 7UT7UTElem Diag SI Accumulator Tk Vent Valves Sh. 7UU7UUElem Diag ESF Manual Reset CasketsSh. 7UV7UVElem Diag Reactor Reactor Vessel Head Vent ValvesSh. 7UW7UWElem Diag Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27A)
Sh. 7UX7UXElem Diag Mn Stm Trip Valve Aux Ckt (3MSS*CTV27B)Sh. 7UY7UYElem Diag Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27C)Sh. 7UZ7UZElem Diag Mn Stm Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27D)Sh. 7VA7VAAnalog Isolation CktsSh. 7VB7VBAnalog Isolation CktsSh. 7VC7VCElem Diag Change Hdr Flow ValvesSh. 7VD7VDElem Diag (3DTM*AOV63B, 64B) Main Steam Line DrainsTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-56Rev. 30Sh. 7VE7VEElem Diag (3DTM*AOV64B, 64C) Main Steam Line Drains Sh. 7VG7VGElem Diag ASP Lamp Test Aux CktsSh. 7VH7VHElem Diag ASP Lamp Test Aux Ckts Sequencer Status LightsSh. 7VJ7VJElem Diag 125 V dc Cntmt Purge Inlet Dampers (3HVR*AOD55B & 174B)Sh. 7VK7VKElem Diag 125 V dc 125 V dc Digital Isolator CircuitsSh. 7VM7VMElem Diag 120 V ac Main Vent & Air Cond Panel Lamp Test Ckt, Train "A"Sh. 7VN7VNElem Diag 120 V ac Main Vent & Air Cond Panel Lamp Test Ckt, Train "B"Sh. 7VQ7VQElem Diag 120 V ac Main Vent & AirSh. 7VX7VXDC Digital Isol Circ 3BYS-PNL-5 & 6Sh. 7VZ7VZElem Diag Digital Isolator Circuits Sh. 7WB7WBElem Diag 120 V ac MCB Lamp Test CktSh. 7WC7WCElem Diag 120 V ac MCB Lamp Test CktSh. 7WD7WDElem Diag 120 V ac MCB Lamp Test Ckt Sh. 7WF7WFElem Diag 120 V ac MCB Lamp Test CktSh. 7WG7WGElem Diag 120 V ac MCB Lamp Test CktSh. 7WJ7WJElem Diag 120 V ac MCB Lamp Test Ckt Sh. 7WL7WLElem Diag 120 V ac MCB Lamp Test CktSh. 7WM7WMElem Diag 120 V ac MB4Sh. 7WP7WPElem Diag 120 V ac MB3 Sh. 7WR7WRElem Diag 120 V ac MB3Sh. 7WS7WSElem Diag 120 V ac MB2Sh. 7WU7WUElem Diag 120 V ac MB2Sh. 7WV7WVElem Diag 120 V ac MB1Sh. 7WX7WXElem Diag 120 V ac MB1TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-57Rev. 30Sh. 7XA7XAAux Fdwtr Control ValvesSh. 7XB7XBAux Fdwtr Control ValvesSh. 7XC7XCAux Fdwtr Control ValvesSh. 7XD7XDAux Fdwtr Control Valves Sh. 7XJ7XJElem Diag 125 V dc Chilled Water Diff Press Cntrl Valve
[3HVK*POV32A, 32B]Sh. 7XK7XKElem Diag 125 V dc Inst Rack & Comp Rms A/C Unit Temp Cntrl ValveSh. 7XL7XLElem Diag 125 V dc Cntrl Rm A/C Unit Temp Cntrl Valve
[3HVK*TV41A, 41B]Sh. 7XM7XMElem Diag 125 V dc E&W Swgr Rms A/C Units Temp Cntrl Valve [3HVK*TV76A, 76B]Sh. 7XN7XNElem Diag 125 V dc E&W Swgr Rms Backup A/C Units Temp Cntrl Valve [3HVK*TV77A, 77B]Sh. 7XP7XPChg Pps Clg Valves Sh. 7XQ7XQElem Diag (3HVZ*MOD20A, 21A)
Sh. 7XR7XRElem Diag (3HVZ*MOD20B, 21B)Sh. 7XV7XVElem Diag 120 V ac MB4Sh. 7XW7XWElem Diag 120 V ac MCB Lamp Test Ckt Sh. 7XZ7XZElem Diag C DA Signal Reset Aux CktSh. 7AAG7AAGElem Diag Control Bldg Chilled Wtr Isol ValvesSh. 7AAH7AAHElem Diag Control Bldg Chilled Wtr Isol ValvesSh. 7AAJ7AAJElem Diag Control Bldg Chilled Wtr Isol ValvesSh. 7AAK7AAKElem Diag Control Bldg Chilled Wtr Isol ValvesSh. 7AAP7AAPAux Fdwtr Cntrl Valves [3FWA*HV31A, 32A & 36B]Sh. 7AAQ7AAQAux Fdwtr Cntrl Valves [3FWA*HV31B, 32B & 36A]Sh. 7AAR7AARAux Fdwtr Cntrl Valves [3FWA*HV31C, 32C &36D]Sh. 7AAS7AASAux Fdwtr Cntrl Valves [3FWA*HV31D, 32D & 36C]TABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-58Rev. 30Sh. 7ABC7ABCElem Di ag Prst Accid Simpl VVS 3SSP*SOV1A, 1BSh. 7ABD7ABDElem Di ag Prst Accid Simpl VVS 3SSP*SOV1C, 1DSh. 7ABE7ABEElem Di ag Prst Accid Simpl VVS 3SSP*SOV2A, 2BSh. 7ABF7ABFElem Di ag Prst Accid Simpl VVS 3SSP*SOV3, 5Sh. 7ABG7ABGElem Diag Post-Accident Sample Valves 3SSP*CTV7,8Sh. 7ABP7ABPElem Diag Cntmt Recirc Isol Valves 3SSP*SOV25A,25BSh. 7ABU7ABUElem Diag Rx Trip on Turb Trip IsolSh. 7ABV7ABVElem Diag Rx Trip on Turb Trip InputsSh. 7ABZ7ABZAux Stm to Aux Bldg Isol Valves [3ASS-AOV102A & B]Sh. 7ABZ17ABZ1Aux Stm to Aux Bldg Temp Switches Sh. 7ACA7ACAAux Bldg Hot Wtr Htg Sys Inlet & Outlet ValvesSh. 7ACB7ACBAux Bldg Hot Wtr Htg. Sys Inlet & Outlet ValvesSh. 7ACD7ACDElem Diag 125 V dc Hot Wtr Preheating Sply & Return IsolSh. 7ACE7ACEElem Diag 125 V dc Hot Wtr Preheating Sply & Return IsolSh. 7ACF7ACFFuel Bldg Air Su ply Isol Damper 3HVR*184Sh. 7ACG7ACGDig Isol CK &
for Fuel Bldg A/S Isol DmprSh. 7ACL7ACLElem Diag 125 V DC Demin Wtr Stor TK to Aux Fd Wtr Pmp Suc 3FWA*AOV61A, 61BSh. 7ACM7ACMAux Fd Wtr Pmp Disch Crossover 3FWA*ALU62A, 62BSh. 7ACP7ACPChrg Pmp Comp Cooling Pmp Cubicle to Aux Bldg Flt Damper 3HVR*MOD 46A, 46BSh. 7ACQ7ACQAux Bldg Exh Fn Var Inlet Valve 3HVR*MOD 140A, 140B Sh. 7ACW7ACWMn. Stm Bldg Exh Fn Dampers 3HVV*AOD50A2, 50B2Sh. 7ACY7ACYTr A Reset for low-low Stm Gen Level on Aux Fd StartSh. 7ACZ7ACZTr B Reset for low-low Stm Gen Level on Aux Fd StartSh. 7ADA7ADARHR Hx Flow Contro l Safety Grade Cold ShutdownSh. 8BA8BARes Sta Svce X Fmr A Prot (3RTX-XSR-A) three lineTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-59Rev. 30Sh. 8DA8DASynchronizing Sh. 8DB8DBSynchronizingSh. 8DC8DCSynchronizingSh. 8HC8HCRSV Sta Svce Bckup Prot Transf Trip RCV CK 3SPRN03Sh. 8HG8HGRsv Sta Svce Pri Prot Xfmr Trip RcvrSh. 8JB8JBNorm Sta Svce Xfmr Bu ProtSh. 8JC8JCNorm Sta Svce Xfmr Bu Prot Sh. 8JD8JDRsv Sta Svce Xfmr Pri ProtSh. 8JF8JFRsv Sta Svce Xfmr Pri ProtSh. 8KA8KAEmer Diesel Gen Excita [3EGS*G-A]Sh. 8KB8KBEmer Diesel Gen Excita [3EGS*G-B]Sh. 8KC8KCEmer Diesel Gen Engine Control [3EGS*G-A]Sh. 8KD8KDEmer Diesel Gen Engine Control [3EGS*G-A]Sh. 8KE8KEEmer Gen "A" G overnor [3EGS*EG-A] 15G-14USh. 8KF8KFEmer Diesel Gen Engine Control [3EGS*G-B]Sh. 8KG8KGEmer Diesel Gen Engine Control [3EGS*G-B]Sh. 8KH8KHEmer Gen Governor [3EGS*EG-B] 15G-15USh. 8KJ 8KJElem Diag Emerg Dies el Gen 3EGS*EG-A, Shutdown CktSh. 8KK8KKElem Diag Emerg Dies el Gen 3EGS*EG-B, Shutdown CktSh. 8KL8KLAlt AC DG 125V DC Hyd Govnr Cntr 3BGS-BG-ASh. 11A11AReactor Trip Breaker [3RPS*ACB-RTA] Sh. 11B11BReactor Trip Breaker [3RPS*ACB-RTB]
Sh. 11C11CReactor Trip Bypass Bkr [3RPS*ACB-BYA]Sh. 11D11DReactor Trip Bypass Bkr [3RPS*ACB-BYB]Sh. 11H11HNuclear Inst SysTABLE 1.7-1 ELEMENTARY DIAGRAMS (CONTINUED)(Refer to Plant Document Control for latest Document Rev. and Date)
NUSCO Drawing Number 25212-32001 SWEC Drawing Number 12179-ESK-Diagram Title MPS3 UFSAR1.7-60Rev. 30TABLE 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS (See Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document C ontrol for Latest Document Rev and Date)
FSAR Figure NumberPiping and Instrumentation Diagram Title NUSCO Number1.2-3Legend: Piping and Instrumentation Diagram269005.1-1Reactor Coolant System26902 5.4-5Low Pressure Safety Injection System269126.2-36Quench Spray and Hydrogen Recombiner System269156.2-37Low Pressure Safety Injection System26912 6.2-53Containment Monitoring System269546.3-2High Pressure Safety Injection System269139.1-6Fuel Pool Cooling and Purification System269119.2-1Service Water System269339.2-2Reactor Plant Component Cooling System269219.2-3Reactor Plant Chilled Water System26922 9.2-4Safety Injection Pump and Neutron Shielding Tank Cooling System 269149.2-5Charging Pump Sealing and Lubrication System269059.2-6Condensate Demineralizer Liquid Waste System269299.2-7Water Treatment System269209.2-8Domestic Water and Sanitary System26947 9.2-9Condensate System269269.2-10Turbine Plant Component Cooling System269349.2-11Primary Grade Water System26919 9.3-1Compressed Air System269389.3-2Reactor Plant Sampling System269449.3-3Turbine Plant Sampling System269439.3-4Radioactive Gaseous Waste System269099.3-5Reactor Plant Gaseous Drains System26907 MPS3 UFSAR1.7-61Rev. 309.3-6Radioactive Liquid Waste and Aerated Drains System269069.3-7Reactor Coolant Pump Seals System269039.3-8Chemical and Volume Control System26904 9.3-9Boron Recovery System26908 26908 269089.3-10Post Accident Sample Sheet269559.4-1Control Building Heating, Ventilation, and Air-Conditioning System 269519.4-2Reactor Plant Ventilation System269489.4-3Turbine Plant Ventilation System269509.4-4Office, ESF, and MSV Building Heating, Ventilation, and Air-Conditioning System 269529.4-5Containment Structure Ventilation System269539.4-6Service Building Ventilation System26949 9.4-7Auxiliary Boiler and Ventilation System269369.4-8Hot Water Heating System269379.4-9Technical Support Center HVAC26956 9.5-1Fire Protection System269469.5-2Emergency Generator Fuel Oil System269179.5-3Emergency Generator Cooling, Starting Air, and Lube Oil System 269169.5-5Nitrogen and Hydrogen System2693910.2-1Electro-Hydraulic Control System2694010.2-2Turbine Generator and Feed Pump Oil Generator System 26941 10.2-3Turbine Generator Support System 2694210.3-1Main Steam and Reheat System26923TABLE 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS (CONTINUED)(See Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document C ontrol for Latest Document Rev and Date)
FSAR Figure NumberPiping and Instrumentation Diagram Title NUSCO Number MPS3 UFSAR1.7-62Rev. 3010.3-2Turbine Plant Miscellaneous Drains System2694510.3-3Chemical Feed System2693110.4-1Condensate System26926 10.4-2Condenser Air Removal and Waterbox Priming System2692710.4-3Extraction Steam and Turbine Gland Seal and Exhaust System 2692410.4-4Circulating Water System2693210.4-5Condensate Demineralizer Mixed Bed System2692810.4-6Feedwater System26930 10.4-7Feedheater and MSR Vents and Drains System2692510.4-9Auxiliary Steam, Feedwater, and Condensate System2693511.2-1Radioactive Liquid Waste and Aerated Drains System2690611.2-2Condensate Demineralizer Liquid Waste System2692911.3-1Radioactive Gaseous Waste System2690911.4-1Radioactive Solid Waste System26910 12.3-5Containment Monitoring System26954TABLE 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS (CONTINUED)(See Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document C ontrol for Latest Document Rev and Date)
FSAR Figure NumberPiping and Instrumentation Diagram Title NUSCO Number MPS3 UFSAR1.7-63Rev. 30TABLE 1.7-3 OMITTED MPS3 UFSAR1.7-64Rev. 30TABLE 1.7-4 SPECIAL REPORTS AND PROGRAMSThe information contained in Table 1.7-4 is retained for historical purposes. Information provided here was relevant at the time of operating license application.1.Transmitted at Time of FSAR SubmittalA.Failure Modes and Effects Analysis (FMEA)B.Fire Protection Ev aluation Report (FPER)C.Millstone Nuclear Power Station Emergency PlanD.Modified Amended Security Plan (MASP) (Under Separate Enclosure)E.Millstone 3 Design Basis Respons e to Regulatory Guide 1.97, Revision 22.Previously Transmitted Report on Faults and Soil Features Mapped in the Discharge Tunnel Excavation3.Transmitted after SubmittalA.Probabilistic Safety Study (PSS)B.Control Room Design Review (Chapter 18)C.Inservice Inspection ProgramD.Environmental Qualification of Electrical Equipment Report (EEQ)E.Environmental Qualification of Mechanical Equipment Report (MEQ)4.Program to be Reviewed at Millstone 3 Equipment Qual ification Documentation (EQD)
MPS3 UFSAR1.8-1Rev. 30
1.8 CONFORMANCE
TO NRC REGULATORY GUIDESTable 1.8-1 lists NRC Division 1 Regulatory Guides. This table is annotated whenever a new or revised Regulatory Guide is invoked or adopted. It identifies applicable FSAR sections, and indicates the degree of Millstone 3 compliance.
MPS3 UFSARMPS3 UFSAR1.8-2Rev. 30TABLE 1.8-1 NRC REGULATORY GUIDES R.G. No.TitleDegree of Compliance FSAR Section Reference1.1 Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps (Rev. 0, November 2, 1970)(1) Comply with the following exception:
6.2.2For the recirculation system phase, the vapor pressure of the water in the sump is assumed to be equal to the containment pressure.
6.3The vapor pressure of the sump water cannot exceed the containment total pressure; therefore, assuming they are equal gives the limiting low value of available NPSH.1.2 Thermal Shock to Reactor Pressure Vessels (Rev. 0, November 2, 1970)(1) See Section 1.8N.5.2.3.31.3Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors (Rev. 2, June 1974)Not applicable.
Applicable only to BWRS.1.4Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors (Rev. 2, June 1974)Comply12.2.1.1R.G. 1.4 is only used for evaluation of original plant shielding design.1.5Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors (Rev. 0, March 10, 1971)Not applicable.
Applicable only to BWRS.1.6Independence between Redundant Standby (On site) Power Sources and Their Distribution Systems (Rev. 0, March 10, 1971)
Comply8.3.1.4 8.3.2 1.7Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident (Rev.3, May 2003)(1) Comply6.2.5 1.8Personnel Selection and Training (Rev. 1-R, May 1977)Compliance is as described in the QAPD Topical Report13.2 17.1 MPS3 UFSARMPS3 UFSAR1.8-3Rev. 301.9Selection of Diesel Generator Set Capacity for Standby Power Supplies (Rev. 2, December 1979)
(2)Comply, with the following exceptions:8.3.1 The magnetizing inrush current due to the four 4,160-480 V load center transformers may cause a momentary (3 to 5 cycles) dip in voltage prior to the first load block. This momentary voltage dip to levels outside that allowed by the Regulatory Guide for load sequencing is considered inconsequential to the successful loading of the standby generator unit.C.11 Section 6.5, Site Acceptance Testing, and Section 6.6, Periodic Testing, of IEEE Std. 387-1977 should be supplemented by R.G. 1.108. The Millstone 3 position on R.G. 1.108 has several clarifications. (See R.G. 1.108).1.10Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures (Rev. 1, January 2, 1973)Withdrawn:3.8.1.6.2 Withdrawal of this guide does not alter any prior or existing licensing commitments based on its use. A position statement follows.1.Reinforce bars with a radius curve of 60 ft.-0 in. or greater are tested at the sampling frequency specified in paragraph C4a.Reinforcing bars with a radius of curvature of less than 60 ft.-0 in. are tested using only sister splices with the following frequency for each splicing crew:One sister splice for the first 10 production splices.
Four sister splices for the next 90 production splices.Three sister splices for the next and subsequent units of 100 production splices.If any sister splice used for tensile testing fails to equal or exceed 125% of the minimum yield strength specified for the reinforcing bar or the average tensile strength of each group of 15 consecutive samples fails to equal or exceed the guaranteed minimum tensile strength of the reinforcing bar, the individual Cadwelder shall be stopped and the procedure in Section C.5 of the Regulatory Guide will be followed.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-4Rev. 30The second paragraph of Position C.3 of R.G. 1.10 states that production mechanical splice samples for tensile testing should not be used from curved reinforcing sections, and then refers to Paragraph 4b for sampling frequency. Paragraph 4b provides for a combination of production and sister splices which appears to be inconsistent with the requirement to use straight sister splices.In preparing the Regulatory Guide, the NRC (formerly USAEC) assumed that the Cadwelders performed splices in the horizontal, vertical, and diagonal directions on the same day. Thus, there would be occasional splices on straight vertical bars which could be alternated with the curve bar splicing to permit the frequency of testing in paragraph 4b of R.G. 1.10, which requires both production and sister splices. However, construction is very apt to perform splices in one position only for more splices than those requiring another set of tests.The NRC agreed that it would accept a testing program using only sister splices whenever curved bars with a radius of curvature less than 60 ft. are mechanically spliced. The 60 ft. was established at SWEC's suggestion as we have obtained satisfactory test results for splices on bars of this or greater radii. Both the NRC and SWEC concede that the value obtained by test at this curvature of rebar will be slightly less than if a straight pull could be made. However, if otherwise practical, the NRC does not like a testing program using only sister splices.2.If any completed mechanical splice fails to pass the visual inspection specified in Paragraph C.2 and the rate of splices that fail the visual inspection does not exceed 1 for each 15 consecutive observed splices, the sampling program will be started anew without requalifying the crew. If the failure rate exceeds 1 in 15, the crew will be re-qualified.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-5Rev. 30The NRC issued a memo on a meeting held in Bethesda, Maryland on May 8, 1973 between the Commission and Erico Products, Inc. on Cadwelds, stating that requalification of a splicer should not necessarily be based on a single visual inspection. The memo stated that statistical sampling procedures permit a discard sample. Since the tensile test sampling program is based on 1 in 15 consecutive samples, the same statistical sampling procedure is used for the visual inspection sampling program.1.11Instrument Lines Penetrating Primary Reactor Containment (Rev. 0, March 10, 1971)
Comply 6.2.41.12Instrumentation for Earthquakes (Rev. 1, April 1974)Comply, except as noted in FSAR section 3.7.4.1.3.7.4.11.13Spent Fuel Storage Facility Design Basis (Rev. 1, December 1975)(1) Comply, except for Fuel Building Ventilation and Filtration requirements. Fuel building ventilation and filtration systems are not credited in the radiological accident analyses.
9.1.21.14Reactor Coolant Pump Flywheel Integrity (Rev. 1, August 1975)(1) See Section 1.8N.5.4.1.11.15Testing of Reinforcing Bars for Category I Concrete Structures (Rev. 1, December 28, 1972)Withdrawn:3.8.1.6.2Withdrawal of this Guide is not intended to alter any prior or existing licensing commitments based on its use.1.16Reporting of Operating Information - Appendix A Technical Specifications (Rev. 4, August 1975)Comply. Monthly Operating Reports superseded by Generic Letter 97-02 and License Amendment No. 223.
161.17Protection of Nuclear Power Plants Against Industrial Sabotage (Rev. 1, June 1973)Comply, with the following clarification:13.6The plant security system is in compliance with 10 CFR 73.55, since it is a federal regulation and, therefore, supersedes the outdated R.G. 1.17.1.18Structural Acceptance Test for Concrete Primary Reactor Containments (Rev. 1, December 28, 1972)Withdrawn:3.8.1.7Withdrawal of this Guide is not intended to alter any prior or existing licensing commitments based on its use.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-6Rev. 301.19Nondestructive Examination of Primary Containment Liner Welds (Rev. 1, August 11, 1972)Withdrawn:3.8.1.2.3Withdrawal of this Guide is not intended to alter any prior or existing licensing commitments based on its use. A position statement follows.3.8.1.7.21.Position C.1.cLeak chase channels were used in areas of inaccessibility, such as the containment liner bottom or weld configurations in which shape discontinuities were evident, such as the personnel and equipment hatches and penetrations. Leak chase channels were pressurized to containment design pressure, and soapsuds were applied to the opposite side for a more sensitive test, since the channels were tested at a greater pressure difference than would have been possible with a vacuum box. In the case of the liner bottom weld seam, which was accessible from one side only, a vacuum box test was performed before the channels were "installed," where possible. However, to verify leak tightness, all the liner bottom welds were leak chase channel pressure tested as stated above.In either case, defects were repaired and the test repeated using the same techniques to ensure compliance with leak-tightness requirements.2.Position C.1.dWhere leak chase channels were installed, the leak-tightness of the liner/penetration to channel weld was verified by pressurizing the channel with Freon 22 (after evacuating the air from the channels) to the containment structure design pressure and examining these welds with a halogen detector. This method provides a more sensitive test and greater assurance that the completed structure will satisfy the leak rate test requirements.1.20Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing (Rev. 2, May 1976)(1) See Section 1.8N.
3.9.2TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-7Rev. 301.21Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants (Rev. 1, June 1974)Degree of compliance to RG 1.21 is defined for the various elements contained within the guidance in the Radiological Effluent Monitoring and Off site Dose Calculation Manual (REMODCM). Exception is being taken to RG 1.21 requirement to estimate population doses as well as individual doses. The calculation of maximum individual doses in combination with a land use census of the area around the plant is sufficient to assess the radiological impact of the plant on man and to implement the requirements of the REMODCM. Population doses will not be routinely calculated and will not be included in the annual reports to the NRC.11.5.1 11.5.2 12.3.4 Online monitors for all pote ntially significant paths for release of radioactive material are provided. For those effluent paths having two or more significant contributing sources, online monitoring of the contributing sources is provided as recommended in the Guide.1.22Periodic Testing of Protection System Actuation Functions (Rev. 0, February 17, 1972)(1) Comply, with the following exception:
7.2RG 1.22 requires that where the ability of a system to respond to an accident signal is intentionally bypassed for the purpose of testing, positive means should be provided to prevent expansion of the bypass condition to redundant or diverse systems, and each bypass condition should be individually and automatically indicated to the reactor operator in the main control room. Permanently wired interlocks and annunciator circuitry are required to fully comply with the preceding requirement. Test circuitry is provided for items 1 through 8, shown in Table 1.8N-1. Items 9 through 13 rely on administrative controls to provide indication and prevent
expansion of the bypass condition.7.3.2.21.23On site Meteorological Programs (Rev. 0, February 17, 1972)Comply 2.3.31.24Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure (Rev.
0, March 23, 1972)Not applicable.15.7.1.3No tanks are used for the storage of radioactive gases. However, an analysis of the most severe potential radioactive gaseous release is presented in FSAR Chapter 15.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-8Rev. 301.25Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Rev. 0, March 23, 1972)Regulatory Guide 1.25 is not used for fuel handling accident analyses. Current analyses are based on Regulatory Guide 1.183.15.7.4 1.26Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste Containing Components of Nuclear Power Plants (Rev. 3, February 1976)(1) Comply, with the following exceptions:
3.2.21.The safety class terminology of ANSI N18.2 and ANSI N18.2-a,1975 is used instead of the quality group terminology.
Thus, the terms Safety Class 1, Safety Class 2, Safety Class 3, and non nuclear safety (NNS) are used instead of Quality Groups A, B, C, and D, respectively.2.Regarding Regulatory Positions C.1.e and C.2.c, one safety valve designed, manufactured, and tested in accordance with ASME III Division 1 (i.e., a code safety valve) is considered acceptable as the boundary between the reactor coolant pressure boundary and a lower safety class or NNS line.3.Regarding Regulatory Positions C.1 and C.2, all instrument tubing, classified as Safety Class 2 or 3, is designed to ASME Section III rules, with Seismic Category I supports installed with a 10 CFR 50, Appendix B program as described in Section 3.2.3.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-9Rev. 304.Groundwater removal post accident via the porous concrete groundwater removal system is a safety-related function. The components within the Engineered Safety Features Building (ESFB) that are accessible/retrievable from the roof are classified as non-nuclear safety instead of Safety Class 3 commensurate with their importance to safety. The sump and piping in the ESFB are Safety Class 3, however, the sump is designed to the guidance of the AISC Code and the piping is designed to the ANSI B31.1 Code. The electrical power for the sump pump is supplied from a single safety related source and is routed via non-safety related power cables.5.The service water system supply and return piping for the post accident sample cooler is designed to ANSI B31.1 requirements and is seis mically qualified.1.27Ultimate Heat Sink for Nuclear Power Plants (Rev. 2, January 1976)
Comply2.4.11.6 9.2.51.28Quality Assurance Program Requirements (Design and Construction)
(Rev. 2, February 1979)
(3)1.(1) Prior to 12/14/05, Millstone Unit 3 complied as follows:
Construction - Millstone 3 complied with R.G. 1.28, Rev. 0.
Operation - Millstone 3 complies with R.G. 1.28, Rev. 2.2.Current compliance is as described in the QAPD Topical Report.17.11.29Seismic Design Classification (Rev. 3, September 1978)(1) Comply 3.2.11.30Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (Rev. 0, August 11, 1972)
No longer comply - 12/14/05 (QA standards are described in QAPD Topical Report.)17.1TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-10Rev. 301.31Control of Ferrite Content in Stainless Steel Weld Metal (Rev. 2, May 1977)
(2) (1) Comply, with the following clarification:
4.5.1The control of ferrite content in stainless steel weld metal meets the requirements of R.G. 1.31, Rev. 1, dated June 1973, or Rev. 3, dated May 1977, on a case-by-case basis.4.5.2.45.2.3.4Rev. 1 requirements are met by our method of qualifying welding procedures and controlling heat input during welding. This is subsequently verified by magnetic measurements of production welds.5.3.1.46.1.1.1 10.3.6.2Rev. 3 requirements are met by purchasing to specifications which require the manufacturer to test each heat and lot for delta ferrite content. The weld pad used is described in SFA 5.4 (1971 version). This weld pad differs in dimensions and in the chill bars (steel versus copper) from the weld pad described in Rev. 3 of R.G. 1.31. However, it should be noted that both weld pads are designed to produce undiluted weld metal for test purposes. Measurements are required of the "Delta ferrite content at 5 places on the as-deposited, undiluted weld metal using a magnetic device calibrated to a single set of traceable standards." Thus, the intent of the Regulatory Guide (Rev. 3) is met; i.e., to test as-depo sited, undiluted weld filler metal for each heat, lot, and process to be used in production.Since the program meets the intent of either version of R.G. 1.31, field welding of ASME III austenitic stainless steel has been performed with weld filler metal purchased as indicated above and no subsequent testing or production welds is performed.1.32Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants (Rev. 2, February 1977)
(3) Comply 8.1.5 8.3.1 8.3.21.33Quality Assurance Program Requirements (Operation)
(Rev. 2, February 1978)Compliance is as described in the QAPD Topical Report.13.413.5.1.117.1TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-11Rev. 301.34Control of Electroslag Weld Properties (Rev. 0, December 28, 1972)
Not applicable.There electroslag welding process for fabricating components was not used.1.35Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures (Rev. 2, January 1976)
Not applicable.There is no prestressed concrete containment structure.1.36Nonmetallic Thermal Insulation for Austenitic Stainless Steel (Rev. 0, February 23, 1973) 1.(1) Prior to 12/14/05, Millstone 3 complied, with the following clarifications and exceptions:
5.2.3Position C.1 states the packaging and shipping requirements of this guide. In lieu of controlled packaging and shipping, receipt inspection and tests are required, by specification.
6.1.1This consists of visual inspection for physical or water damage to all cartons. Damaged cartons are segregated. The potentially contaminated insulation is not accepted unless randomly selected samples from each carton are shown to be acceptable after being re-subjected to the pro duction test outlined in the Guide.2.Current compliance is as described in the QAPD Topical Report.17.11.37Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (Rev. 0, March 16, 1973)1. (1) Prior to 12/14/05, Millstone 3 complied.
4.5.12. Current compliance is as described in the QAPD Topical Report.17.11.38Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Rev. 2, May 1977)1. (1) Prior to 12/14/05, Millstone 3 complied.17.12. Current compliance is as described in the QAPD Topical Report.1.39Housekeeping Requirements for Water-Cooled Nuclear Power Plants (Rev. 2, September 1977)1.Prior to 12/14/05, Millstone 3 complied as follows:12.5.3 Construction - Millstone 3 complied with Rev. 0 of the Guide.
Operation - Millstone 3 complied with Rev. 2 of the Guide.2.Current compliance is as described in the QAPD Topical Report.17.1TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-12Rev. 301.40Qualification Tests of Continuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (Rev. 0, March 16, 1973)(1) Comply3.11 8.31.41Preoperational Testing of Redundant On site Electric Power Systems to Verify Proper Load Group Assignments (Rev. 0, March 16, 1973)
Comply 8.1 8.31.42Interim Licensing Policy As Low As Practicable for Gaseous Radio Iodine Releases from Light-Water-Cooled Nuclear Power PlantsWithdrawn1.43Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components (Rev. 0, May 1973)(1) Comply5.3.1.41.44Control of the Use of Sensitized Stainless Steel (Rev. 0, May 1973)(1) Comply, with the following clarification:
4.5.1The intent of Paragraph C.6 is complied with in varying degrees as follows:4.5.2.4 5.2.3.41.Field fabrication and erection of ASME III piping complies with Paragraph C.6, except that the ASTM A708-74 standard is used to perform the intergranular corrosion testing. The radius of the bend specimen is as specified in ASME IX with the weld metal-base metal interface located at the centerline of the bend. This meets the intent of Paragraph C.6.5.3.1.46.1.1.1 10.3.6.2R.G. 1.44 requires that an intergranular corrosion test, such as ASTM A262 Practice A or E, should be performed to evaluate sensitization of the heat affected zone in stainless steel weldments having a carbon content greater than 0.03%. The ASTM A262 is a very severe test and most of unstabilized stainless steel weldments with a carbon content above 0.03% would not pass it. According to the NRC Regulatory Standard Review Plan (Section 4.5.1), ASTM A708-74 test (previously A393) is acceptable for testing the qualification welds for degree of sensitization.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-13Rev. 30Paragraph 7.1.1 of ASTM A708-74 requires that the test specimen is bent 180
° over a diameter equal to the thickness. This requirement is applicable to the base metal, but ductility of the weld metal and the heat affected zone may not be that good even without any corrosion attack. his is recognized in ASME IX, which required testing of weldments over a diameter equal to four times the thickness.2.Shop fabrication of ASME III piping, tanks, and valves and field fabrication of ASME III tanks require control of heat input during welding so as to avoid severe sensitization of the weld zone. In addition, the maximum interpass temperature is limited to 350
°F. While no testing for intergranular corrosion during weld procedure qualification is re quired, the above controls assure that base material will not be severely sensitized during welding and meets the intent of Paragraph C.6.3.Fabrication of ASME III forged stainless steel instrumentation valves and ASME III components other than those identified in Items 1 and 2 above is performed in fabrication shops and requires a maximum interpass temperature of 350°F. While no testing for intergranular corrosion during weld procedure qualification is required, this specifi c control reduces the possibility of a severely sensitized heat-affected zone duri ng welding. In addition, the need for the shop fabricator to provide unsensitized heat-affected zone during welding. In addition, the need for the shop fabricator to provide unsensitiz ed components is specifically identified in all procurement specifications by requiring supplied material to be capable of meeting ASTM A262, Practice A or E. Shop practice generally recognizes the need to limit heat input during the welding through good fit-up, adequate welder accessibility, proper positioning, and close supervision. Finally, most of the pressure retaining comp onents of the RCPB piping system are castings and, because of their delta ferrite content, are highly resistant to stress corrosion cracking.1.45Reactor Coolant Pressure Boundary Leakage Detection Systems (Rev. 0, May 1973)Comply with the following interpretation given to Regulatory Position C.5:5.2.5TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-14Rev. 30The sensitivity and response time of each leakage detection system employed to collect unidentified leakage are as shown in the following table:
System Sensitivity and Response Time Containment Drain Sump Level or Pumped Capacity Monitoring System1 gpm in less than 1 hourContainment Atmosphere Humidity, Temperature and PressureHumidity, temperature or pressure monitoring of the containment atmosphere are considered as alarms or indirect indication of leakage to the containmentContainment Atmosphere Gaseous and Particulate Radioactivity Monitoring1 gpm in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, provided that the equilibrium activity of the reactor coolant is sufficiently high and the equilibrium activity of the containment atmosphere is below a level that would mask the change in activity corresponding to this leak rateComply with the following interpretation given to Regulatory Position C.75.2.5Indicators and alarms for each leakage detection system should be provided in the main control room. Procedures for converting sump level and pumped capacity indications to a common leakage equivalent should be available to the operators.
The calibration of the indicators should account for needed independent variables. Due to the numerous factors that can affect the readings from the Containment Atmosphere Gaseous and Particulate Radioactivity Monitors, they can not be used to reliably quantify a leak rate, although they are a very sensitive indicator of a leak. Factors affecting the monitor readings include RCS radioactivity levels, containment air radioactivity levels, radionuclide mix, location of leak and removal mechanisms.1.46Protection Against Pipe Whip Inside Containment (Rev. 0, May 1973)(1) Comply with the following clarifications and exceptions:
3.6Paragraph C.1.b TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-15Rev. 30At intermediate locations between terminal ends selected by either of the following criteria:1.At each pipe fitting, welded attachment, and valve; or2.At locations where the primary plus secondary stress intensities (circumferential or longitudinal) derived on an elastically calculated basis under loadings associated with specified seismic events and operational plant conditions exceed 2.4 S min lieu of "Any intermediate locations...conditions exceed 2.0 S m for ferritic steel and 2.4 S m for austenitic steel."Paragraph C.2.b At intermediate locations between terminal ends selected by either of the following criteria:At intermediate locations between terminal ends selected by either of the following criteria:1.At each pipe fitting, weld attachment, and valve; or2.At locations where either the circumferential or longitudinal stresses derived on an elastically calculated basis under the loadings associated with specified seismic events and operational plant conditions exceed 0.8 (1.2 S h + S a)in lieu of "Any intermediate locations...conditions exceed 0.8 (S h + S a)."Paragraph C.3.a, Footnote 10 Longitudinal breaks are assumed to result in axial split without pipe severance. Splits oriented (but not concurrently) at two diametrically opposed points on the piping circumference such that jet reactions cause out-of-plane bending of the piping configuration. Alternatively, a single split may be assumed at the section of highest stress as determined by a detailed stress analysis (e.g., finite element analysis).TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-16Rev. 30 The dynamic force of the fluid jet discharge should be based on circular or elliptical (2D x 1/2D) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coeffi cient as determined for a circumferential break at the same location. Line restrictions, flow limiters, po sitive pump-controlled flow, and the absence of energy reservoirs may be taken into account as applicable, in the reduction of jet discharge.in lieu of Footnote 10 of R.G. 1.46.Paragraph C.3.b, Footnote 11 Pipe whipping is assumed to occur in the plane defined by the piping configuration and geometry and to cause pipe movement in the direction of the jet reaction.in lieu of Dynamic forces resulting...cause whipping in any direction normal to the pipe axis (Footnote 11 of R.G. 1.46).Implementation of additional criteria is documented and approved by NRC as providing an acceptable level of plant safety. These criteria potentially include those bases provided by NRC Standard Review Plans 3.6.1 and 3.6.2, the approved topical reports and SARs submitted by NSSS, and the draft of ANSI-N176, Standard Design Basis for Protection of Nuclear Power Plants Against Effects of Postulated Pipe Rupture.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-17Rev. 30Certain provisions of R.G. 1.46 are impractical. While NRC is in the process of revising R.G. 1.46, the NRC Standard Review Plant 3.6.2, Paragraph 11.1 stated "if the criteria specified in R.G. 1.46 are impractical for a specific application, the criteria of Branch Technical Position BTP-MEB-3-1 will be considered," and "BTP-MEB-3-1 may be used for all applications, in lieu of References 3 and 4, at the option of the Applicants." The modifica tions or exceptions taken by Millstone 3 on position of R.G. 1.46, are based upon BTP-MEB-3-1. These positions were incorporated in the PSAR of WEPCO and SWESSAR, and the Safety Evaluation Reports (SER) issued by the NRC for these two projects found it to be acceptable.1.47Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems (Rev. 0, May 1973)Comply, with the following clarifications:7.5.31.An indicator of bypass/inoperability, located in the control room, will be provided for redundant or diverse portions of each safety system. Bypass indication will be provided for any deliberate action that renders a safety system inoperable.2. Bypass of redundant portions of engineered safety feature support systems warrants indicators that must be differentiated from safety system bypass indicators.1.48Design Limits and Loading Combinations for Seismic Category I Fluid System Components (Rev.
0, May 1973)(1) Comply, with the following exception:
3.9.3Class 2 and 3 piping analysis to follow criteria specified in 1971 ASME III Code, and all Addenda up to and including Summer 1973, except that piping systems other than safety injection and quench/recirculation spray use an increased allowable (design limit) of 2.4 S for plant faulted loading conditions.Components, except piping, use stress criteria given in Tables 3.9B-5 and 3.9B-7. A comparison of these valves and R.G. 1.48 are given in Tables 3.9B-6 and 3.9B-8.R.G. 1.48 has been superseded by NUREG 0800, SRP 3.9.3. Refer to FSAR Section 1.9, SRP 3.9.3 for further discussion.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-18Rev. 301.49Power Levels of Nuclear Power Plants (Rev. 1, December 1973)Comply, with the following exception:1.1The analyses and evaluations for those conditions described in Position C.2 and the analyses of the possible off site radiological consequences described in Position C.3 are made at an assumed core power level using the guidelines in Position C.3.15.0The margin specified in Position C.2 is equal to 1.02 times the licensed power level to allow for possible instrument errors in determining the powe r level. The words "...equal to 1.02..." are interpreted to mean "...equal to at least 1.02...". This interpretation allows the analyses and evaluations described in Position C.2 to be made, at an Applicant's discretion, at a somewhat higher power level to account for the margin which may be provided in turbine generator designs above rated capacity. This interpretation can be found in Position C.3, and is, therefore, also considered applicable to Position C.2.As described in FSAR Chapter 15, most safety analyses demonstrating that the DNB design criterion is met do not explicitly use 1.02 times the licensed thermal power for full power initial condition assumptions. 1.50Control of Preheat Temperature for Welding of Low-Alloy Steel (Rev. 0, May 1973)(1) Comply, with the following exception:5.2.3.3In cases where it is impractical to maintain preheat until a postweld heat treatment has been performed, the preheat temperature is maintained for sufficient time to assure that the residual hydrogen has effused from the weld zone This reduces the tendency to form cracks in the weldment and, therefore, complies with the requirements of this Guide.1.51Inservice Inspection of ASME Code Class 2 and 3 Nuclear Power Plant ComponentsWithdrawnTABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-19Rev. 301.52Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants (Rev. 2, March 1978)Comply, with the following clarifications and exceptions:6.5.1.2 Paragraph C.2.g The filter trains are not instrumented to monitor, alarm and record flow rates in the control room. Flow through the filters are verified on a monthly basis. The flow through the fuel building filters are not required to be verified.Paragraph C.2.h The following exceptions are taken to the requirement that "all instrumentation and equipment controls should be designed to IEEE 279":1.All instruments and equipment controls that sense or process one or more variables and that act to accomplish the protective function are designed in accordance with IEEE 279. These include sensors, signal co nditioners, logic, and actuation device control circuitry. (The protective function with which the subject guide is concerned is atmospheric cleanup to mitigate accident doses.)2.In addition, a very limited class of analog indicators may be designed in accordance with selected applicable paragraphs of IEEE 279.The basis for selecting specific indicators to be so designed, is their significance to safety.All paragraphs of IEEE 279 are applicable, except 4.12, 4.13, 4.15, 4.16, and 4.17.For this limited class of indicators, redundant analog channels are provided. One channel is recorded. The systems are designed to operate before and after, but not necessarily during, a safe shutdown earthquake.3.Annunciator functions are incorporated in overall system design. Annunciators are not safety-related; therefore, they are not designed in accordance with IEEE 279.Paragraph C.2.i (Clarification)
TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-20Rev. 30Fuel building exhaust system actuation is manual from the control room. Fuel building high radiation will annunciate in the control room if the radiation level reaches the predetermined setpoint during normal plant operation, as well as DBA.
When the radiation alarm from the particulate and gas monitor for the exhaust duct work has been verified, the filtration system may then be actuated manually.Paragraph C.2.j (Clarification)
Adsorber design provides for the replacement of charcoal by an external vacuum system. The structural design of the filter train will provide for the removal of the filter components on a cell by cell basis. Demisters, heaters, fans and casings will be decontaminated by wash down process; wash down liquid will drain to an aerated drain system.Paragraph C.2.l Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510-1980. Leak rate of 0.1 percent in accordance with Table 4-3 of ANSI N509-1980 is acceptable. However, ductwork leak tests shall be performed in accordance with the procedures delineated in Chapter 8, Leak Testing, of the Manual for the Balancing and Adjustment of Air Distribution Systems, published by SMACNA, 1967.All ESF and non-ESF ventilation systems inte rconnecting with safety-related filter trains comply with ANSI N509-1980 leakage requirements, with the following clarifications:1.The following fan systems will exceed ANSI N509-1980 leakage requirements on the suction side:*Containment purge exhaust fan.*Suction side on fuel building filters and fuel building normal exhaust fan.The leakage flow into the ductwork will have no significant effect on air cleaning effectiveness since the fans and filters have enough capacity to handle the additional flow.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-21Rev. 302.The fuel building ductwork on the discharge side of the filter exhaust fans will exceed ANSI N509-1980 leakage requirements. This is a radiation protection consideration because the system will exhaust air from contaminated areas then leak into relatively cleaner areas. The consequences of leakage from this system during normal operation is inconsequential since the activity in the duct will be less than 1 MPC. However, personnel access may have to be limited to the top floor of the auxiliary building, where the ductwork is located, following a fuel handling accident.3.The leakage on the suction side of the SLCRS exceeds ANSI requirements, but will have no adverse effect on maintaining negative pressure within the cubicles since the fan can accommodate the added flow.The leakage on the discharge side of the fan also exceeds ANSI requirements, but will be into the same areas being exhausted. This type of leakage affects only effectiveness of the SLCRS to draw down the pressure. The fans are sized to compensate for this leakage.Also, the leakage between the filter outlet and the fan suction which is located in the auxiliary building is in excess of ANSI requirements. The effect of this leakage on system performance has been evaluated and was found to be acceptable.4.Leak test boundaries for the control building ESF filters are defined in accordance with ANSI N509-1980, Appendix B, Figures B-3 and B-4, schemes 13 and 22, to be from the filter inlet to filter outlet.5.Millstone 3 complies with ANSI N509-1 980 requirements stated in Paragraph 4.6.3, which calls for the use of design static pressure (working pressure) in testing for duct leakage.However, fan peak pressure is not used during leakage testing of segments of ductwork which could be isolated by a closed damper or a clogged filter as implied by Paragraph 4.6.2 of ANSI N509-1980. The bases for the clarification are as follows:TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-22Rev. 30*Total clogging of filter or closure of a damper on fan suction will have greater impact on air cleaning effectiveness than the increased leakage resulting from fan shut-off pressure. Therefore, duct leakage is determined at maximum working pressure.*Closure of dampers located on the discharge side of the fan will result in low or no flow in the system, thus causing that particular fan to stop and the redundant one to start. Such action will preclude the possibility of any significant duct outleakage and eliminate this health/physics concern.Paragraph C.3.d (Clarification)
Filters have been purchased to ANSI N509-1976. Filter media will be subjected to velocities recommended by the HEPA filter manufacturer which exceeds ANSI N509-1976 requirements given in Section 4.3.1. The HEPA filter cell testing is conducted initially at the manufacturer's facilities and again after installation at the plant site. All HEPA filters furnished are equipped with face guards in accordance with MIL-F-51068. When installed in the filter housing, the HEPA filters and housing are inspected for defects and tested for leak-tightness in accordance with ANSI N510-1980.Paragraph C.3.e (Clarification)
Filter mounting frame is constructed and designed in accordance with the recommendations of Section 4.3 of ERDA 76-21, except for the frame tolerance guidelines in Table 4.2. The tolerances selected for HEPA mountings are sufficient to satisfy the bank leak test criteria of Paragraphs C.5.c and C.5.d of RG. 1.52, Rev.
2.Paragraph C.3.g TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-23Rev. 30Millstone 3 complies with ANSI N509-1980 paragraph 4.6.2.2 with respect to designing inlet units and components which can be isolated from the fan to withstand a peak negative pressure by ensuring that such isolation is precluded via the design control logic between the fans and the inlet dampers. Compliance with designing inlet units and components, as noted in the same paragraph with respect to the plugging of such components, is de monstrated via routine surveillance and subsequent filter replacement as necessary.Millstone 3 is in accordance with ANSI N509, except access to the control building filter units is not provided with hinged doors or inspection windows. Access is via 20-inch by 40-inch bolted panels. Other units are provided with hinged doors or bolted panels with inspection windows. There is no internal lighting.
Even though doors are not available to access both sides of each bank of components, Millstone 3 complies with the intent of the requirements of Paragraph 5.6 of ANSI N509-1976 to provide access to each side of each component of the ESF ventilation filtration systems filter housings for maintenance and testing.Paragraph C.3.h Exception is taken to the recommendations of Section 4.5.8 of ERDA 76-21 relative to drain sizes and arrangement. Normally closed manual valves, instead of water seals and traps, will be provided to control the discharge of the fire sprinkler flow. Sprinkler flow will be a timed discharge, and the water will be contained within the housing until it is removed to the liquid radwaste system at a controlled rate. Condensate from the moisture separator chamber will continually drain via normally open drain valves.Paragraph C.3.i TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-24Rev. 30The dwell time for the minimum 2 inches of the carbon adsorber unit of the Auxiliary Building Filtration System is 0.22 sec. All filters use a 4-inch thick charcoal bed which exceeds the minimum 2 inches recommended by R.G. 1.52, Rev. 2. The additional 2 inches will result in a dwell time of 0.43 sec. for the Auxiliary Building Filtration System. Carbon is purchased to Table 5-1 of ANSI N509-1980. Pre Generic Letter 99-02 testing of the charcoal was based on a maximum face velocity of 46 fpm and a 2-inch thick bed. Testing of new and used ESF filter system charcoal, post Generic Letter 99-02 (Ref. Amendment 184) uses ASTM D3803-89 testing standards with face velocities based on the greater of the individual system upper Technical Specification air flow limit (cfm) or 12.2 meters/min. (40 fpm)Paragraph C.3.k When conservative calculations show that the maximum decay heat generation from collected radioiodines is insufficient to raise the carbon bed temperature above 250°F with no system air flow, ESF atmosphere cleanup systems may be designed without a decay heat removal mechanism. (See FSAR Table 6.5-1 for applicability.)In addition, exception is taken to provide humidity control for the decay heat removal system cooling air flow which uses room air of less than 70
% relative humidity.Paragraph C.3.l System resistances will be determined in accordance with Section 5.7.1 of ANSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan blast area data necessary to calculate inlet and outlet losses, per AMCA 201, are the responsibility of fan manufacturers, and are not available from them.The following vibration performance criteria will be used:
Acceptable Alert Action 0.325 in/sec
> 0.325 in/sec
> 0.700 in/sec instead of ANSI N509-1976 Section 5.7.3.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-25Rev. 30Documentation will not be furnished in accordance with Section 5.7.5 where AMCA certification ratings are submitted. An engineering evaluation and a test determined that the SLCRS exhaust fans were correctly selected to operate, and do operate, on the stable portion of the fan curve during single or parallel operation.Paragraph C.3.n Exception is taken to Section 5.10.3.5 of ANSI N509-1976; ductwork, as a structure, will have a resonant frequency above 25 Hz, but this may not be true for the unsupported plate or sheet sections. The design provides for specification of the resonant frequency range of the support hangers. Specifying the resonant frequency of the unsupported plate or sheet has no meaning in the design.Exception is taken to Section 5.10.5 of ANSI N509-1976 on welding in accordance with AWS D1.1 or ASME Section IX. AWS D1.1 General Provisions Section stipulates that the code is not intended to apply to welding base less than 1/8 inch thick. Since ductwork thicknesses are below 1/8 inch, the AWS D9.1 code, Specification for Welding of Sheet Metal, is used.In addition, exception is taken to the following:Workmanship samples shall be inspected with liquid penetrant or magnetic particle on both root and face surfaces.Workmanship samples shall be inspected by macro-sectioning butt welds. Macro-sectioning is a satisfactory method to determine butt weld quality, including both root and face surfaces.Exception is taken to Section 5.10 of ANSI N509-1976 regarding allowable stresses for the structural analysis of ductwork. Requirements shall be as described in Section 5.10.3.3 of ANSI N509-1980.Exception is taken to Section 5.10.9 of ANSI N509-1976 regarding balancing of all duct systems. All duct systems shall be balanced to within
+/-10 percent of the specified design air flow.This value is in agreement with ANSI N509-1980.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-26Rev. 30Paragraph C.3.p Exception is taken to the provisions in Section 5.9 of ANSI N509-1976 of designing dampers to ANSI B31.1 and to using butterfly valves. Class B dampers may be designed and tested to meet the verification of strength and leak-tightness necessary for use in a contaminated air stream. (Note: This exception does not pertain to containment penetrations.)In addition, exception is taken to the following:NRC Question 460.12All Class II dampers on ESF atmosphere cleanup system air filtration and adsorption units have been tested for leakage rates except for two backdraft dampers on the SLCRS fan discharge. The size of the two untested backdraft dampers is bounded by both larger and smaller size dampers which have been satisfactorily tested to 50% of allowable leakage rates.
Damper leakage will not impact on the air cleaning effectiveness of ESF systems.Paragraph C.4.a Exception is taken to full compliance with Section 2.3.8 of ERDA 76-21; i.e., the plant does not use any communications system, floor drains are as noted in Paragraph C.3.h above, decontamination areas and showers are not "nearby," filters are not used at duct inlets, and duct inspection hatches are not provided.Paragraph C.4.b Partial compliance, with a minimum spacing between filter frame of 2 ft.-6 in. instead of a minimum of 3 ft. This is deemed adequate since replacement of filter elements would be minimal due to system function, use, and location.Paragraph C.4.d (Clarification)
ESF atmosphere cleanup systems are run a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month. However, if the field data confirms that it is unnecessary to run the trains 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month to reduce the amount of moisture present on the filters, this decisions will be reconsidered.Paragraph C.5.a TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-27Rev. 30Visual inspection is performed in accordance with ANSI N510-1980.Paragraph C.5.b Test for air flow distribution to the HEPA filters and the adsorbers shall be conducted in accordance with Section 8 of ANSI N510-1980 with the following exceptions and clarifications:Paragraph 8.3.1.6 - Test shall be conducted to verify the following pressure drop values across the combined HEPA filter and charcoal adsorber banks and across the entire filter housing:
System P Across HEPA & Charcoal (in. wg)P Across Housing (in. wg) Supplementary Leak Collection &
Release6.257.75Control Room Emergency Air Filtration6.759.86Aux. Bldg. Filter6.809.80Paragraph 8.3.1.7 - Exception is taken to the requirement regarding testing at 50 percent of the pressure drop value. Testing the filter at this condition will require the removal of certain filter components with potential damage due to mishandling. The above condition is not considered a design parameter of the system.Paragraph 8.3.1.8 - Acceptance criteria for the control building shall be 1,225 cfm
+/-10 percent at clean filter condition, and 1,000 cfm
+/-10 percent at dirty filter condition. Calculations have demonstrated that the control room pressurization and dose limits are not affected by the above range of air flow.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-28Rev. 30Exception is taken to the SLCRS fan flowrate acceptance criteria of
+ or 10% requirement of ANSI N510-1980. Instead a range of flows is specified; the higher end of which is established by the capacity of the filter train and the lower end by an Inservice Test.Paragraph 8.3.2.3 - Air distribution tests ac ross the prefilter and moisture separator banks are not required by project specifications.Paragraph C.5.c HEPA Filter DOP testing is conducted in accordance with ANSI N510-1980.Exception taken to the requirement that test should be performed at least once per 18 months. Based on acceptance of 24 month fuel cycle; this test can be extended to at least once per refueling interval.Paragraph C.5.d Charcoal adsorber leak testing with refrigerant is conducted in accordance with ANSI N510-1980.Exception taken to the requirement that test should be performed at least once per 18 months. Based on acceptance of 24 month fuel cycle; this test can be extended to at least once per refueling interval.Paragraph C.6.a/b TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-29Rev. 30The Auxiliary Building Filtration System ac tivated carbon adsorber section has a 4 inch bed and operating face velocity of 46 fpm (0.43 sec residence time) based on nominal design air flows. The accident dose analysis in Chapter 15 of the FSAR is based on a 95 percent decontamination efficiency. Table 2 of R.G. 1.52, Rev. 2 assigns a 95 percent decontamination efficiency for an activated carbon sample having a methyl iodide penetration of less than 1 percent. Therefore, within 31 days after removal, a 4 inch laboratory sample from the installed sample canisters of ESF filtration systems activated carbon absorber, need only demonstrate a removal efficiency of 99 percent for methyl iodide when tested in accordance with ANSI N510-1980 at 80
°C and 70 percent relative humidity. Pre Generic Letter 99-02 Technical Specifications demonstrate removal efficiency of 99.825%. Technical Specification testing of used ESF filter system charcoal, post Generic Letter 99-02 (Ref. Amendment 184), uses ASTM D3803-89 testing standards assuring charcoal efficiency of 97.5% or greater.Table 2, Note C Table 2, Note C of R.G. 1.52, Rev. 2 states that testing should be performed ". . . (2) at least once per 18 months thereafter. . ."Exception taken to the requirement that test should be performed at least once per 18 months. This test is to be performed at least once per refueling interval.1.53Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (Rev. 0, June 1973)(1) Comply, with the following clarifications:
3.1.11.Regulatory Position C.1Due to the trial-use status of the source document, IEEE 379 1972, departure from certain provisions may occur. The phrase "any and all combinations of nondetectable failures" in Paragraph 3(3) of IEEE 379 is interpreted to mean that an accumulation of single nondetectable failures taken collectively,culminate in a nondetectable combination.2.Regulatory Position C.2TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-30Rev. 30The protection system, as defined by IEEE 279-1971, incorporates the capabilities for test and calibration as set forth in Paragraphs 4.9 and 4.10 of IEEE 279-1971.Final actuation devices (as defined by IEEE 379-1972) are capable of periodic testing in accordance with R.G. 1.22. Those final actuation devices which cannot be fully tested during reactor operation (for reasons as stated in regulatory positions 4a through 4c of R.G. 1.22) are subjected to a partial test with the unit online and full operational testing during reactor shutdown.Taken as a whole, the operability of all active components ne cessary to achieve protective functions is demonstrated via the testing program described above.3.Regulatory Position C.3Single switches supplying signals to redundant channels are designed with at least 6 inch separation or with suitable barriers between redundant circuits.4.Compliance with single-failure criteria is verified based on a collective analysis of both the protective system, as defined in IEEE 279-1971, and the final actuation devices or actuators, as defined in IEEE 379-1972.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-31Rev. 301.54Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants (Rev. 0, June 1973) 1.(1) Prior to 12/14/05, Millstone 3 complied, with the following clarification and exception:
6.1.2 Clarification
Compliance will not be invoked for equipment of a miscellaneous nature and all insulated surfaces. Due to the impracticability of imposing Regulatory Guide requirements to the standard shop process used in painting valve bodies, handwheels, electrical cabinetry and control panels, loudspeakers, emergency light cases and other miscellaneous equipment, the Regulatory Guide will not be invoked for these items since the total surface area for such items is relatively small when compared to the total surface area for which the requirements are imposed.Exception Quality Assurance Program recommendations stated in R.G. 1.54 are followed except for the inspection defined in Section 6.2.4 of ANSI N101.4-1972. Inspection is in accordance with ANSI N5.12-1974, Section 10. Testing of coating materials is performed in accordance with ANSI N101.2, or ASTM D3911 as noted in Section 6.1.2.1.2.Current compliance is as described in the QAPD Topical Report.17.11.55Concrete Placement in Category I Structures (Rev. 0, June 1973)Withdrawn3.8.1.6.1Withdrawal of this Guide is not intended to alter any prior licensing commitments based on its use. A position statement follows:1. Shop detail drawings for the reactor containment mat, shell, containment intervals, spent fuel pool, and dome reinforcement are checked by the designer. All other reinforcing shop details are checked by engineers at the job site.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-32Rev. 30The Regulatory Guide calls for all shop drawings to be checked by the designer. The mat, shell, and dome of the reactor containment structure are checked by the designer, as these are complicated arrangements. Further, the large size bars normally used require special detailing practices to permit bending and to satisfy development of strength requirements. Therefore, these detail drawings are checked by designers who have had previous experience with the large bars. Details for conventionally reinforced structures are normally checked by engineers in the field. This allows the field to review the proposed locations of construction joints.2.A slump of 4 inches is used for mass concrete in areas where the density of reinforcing steel requires a more plastic mix for placement.3.Curing of the concrete for the reactor containment shell and dome conformed to Chapter 12 of ACI 301 instead of Subsection CC-4240 of ACI 359. Both have the same requirements for temperature and duration, but Chapter 12 allows curing compounds to reta in the moist environment. Curing compounds are not used on the internal structures of the reactor containment.4.The ACI and ASTM specifications are supplemented as necessary with mandatory requirements relating to types and strengths of concrete, minimum concrete densities, proportioning of ingredients reinforcing steel requirements, joint treatments, testing requirements, and quality control.1.56Maintenance of Water Purity in Boiling Water Reactors (Rev. 1, July 1978)Not applicable. Only applicable to BWRs.1.57Design Limits and Loading Combinations for MetalNot applicable.Primary Reactor Containment System Components (Rev. 0, June 1973)Only applicable to those plants with a metal primary reactor containment.1.58Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel (Rev. 1, September 1980)
- 1. (1) Prior to 12/14/05, Millstone 3 complied.17.12. Current compliance is as described in the QAPD Topical Report.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-33Rev. 301.59Design Basis Floods for Nuclear Power Plants (Rev. 2, August 1977)Comply with the following clarification:2.4.2.2Position C.1 specifies designing hardened protection for all safety related structures, systems, and components. Position C.1 requires harden ed protection to be passive and in place, as it is to be used for flood protection, during normal plant operation.Flood protection for each service water pump cubicle is provided in part by a watertight door and a cubicle sump drain line. During normal operations, the drain line of each cubicle is open and the door of one cubicle only is open. Plant procedures require isolation of the sump drain line and watertight door of each cubicle on approaching severe weather. These actions are performed prior to sea level rising to the level which requires entry into the LCO for Technical Specification 3.7.6.A Technical Specification to ensure closing of the cubicle watertight doors in advance of a potential flooding event was required by Section 2.4 of the SER for MP3. Section 2.4 of the SER concluded that the guidelines of RG 1.59 had been met. The NRC therefore recognized that the cubicle watertight doors may be open during normal operations and found this to be acceptable.1.60Design Response Spectra for Seismic Design of Nuclear Power Plants (Rev. 1, December 1973)(1) Not applicable.The NRC granted that Millstone Nuclear Power Station Unit 3 will not be required to comply with Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, dated October 1973.Plants docketed prior to April 1, 1973 are not required to consid er this Regulatory Guide. PSAR Section 2.6.2.7 specifies the response spectra used for the design of Millstone 3.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-34Rev. 301.61Damping Values for Seismic Design of Nuclear Power Plants (Rev. 0, October 1973)(1) Not applicable.The NRC granted that Millstone Nuclear Power Station Unit 3 will not be required to comply with R.G. 1.61, Damping Values for Seismic Design of Nuclear Power Plants, dated October 1973.Plants docketed prior to April 1, 1973 are not required to address this Regulatory Guide. FSAR Tables 3.7B-1 and 3.7N-1 list the damping values used with Seismic Design of Millstone 3.1.62Manual Initiation of Protective Actions (Rev. 0, October 1973)(1) Comply, with the following clarification and exceptions:7.3.2.2.71.Regulatory Position C.1a.Manual initiation at the system level is interpreted to mean no more than three operator actions will be required to initiate at least one train, division, or channel of final actuation devices, including su pport systems.b.Engineering judgment will be exercised to assure that a minimum of operator actions are required to achieve system level manual initiation without unnecessarily jeopardizing the return to operation of the power plant. For protective actions that significantly affect return to operation, or for those protective actions that may, if inadvertently initiated, result in a less safe plant condition, operator actions on tw o control devices will be required.c.Designs requiring more than two operator actions per train, division, or channel to achieve protective action are to be limited to those actions required only in the long term and will be evaluated on a case-by-case basis.2.Regulatory Position C.2All equipment that contributes to the protective action will be initiated at the system level.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-35Rev. 30There is no auxiliary feedwater system low-low SG water level manual initiation design feature which replicates the automatic protective action design feature. Except for the safety injection and containment depressurization manual initiation design features which will initiate motor-driven auxiliary feedwater pumps, auxiliary feedwater manual initiation requires that the operator start the pumps and isolate steam generator blowdown lines. During normal power operation, plant operating and surveillance procedures require that all auxiliary feedwater valves be aligned in the positions which provides a path to the steam generators. No operator action to perform auxiliary feedwater valve position changes is required to establish a flow path to the SG. The motor-driven AFW pump automatic initiation system closes the steam generator blowdown sample valves (3SSR*CTV19A-D). Engineering judgment is exercised to conclude that steam generator blowdown sample line isolation is inconsequential to the accomplishment of the AFW safety function because these lines are 3/8 inch lines and have insignificant flow capacity. Therefore, the sample valves are not counted as a required operator action to manually initiate the auxiliary feedwater system protective action.At below 10% rated thermal power levels, plant operators may throttle or close flow control valves (3FWA*HV31A/B/C/D) for steam generator water inventory control (or to support system alignment or restoration from normal use during startup, shutdown, and hot standby). These control valves do not receive an open signal from safety injection nor containment depressurization system manual initiation design feature. In this case, manual operator action is credited to open these control valves, if required, and to support auxiliary feedwater system operability. During auxiliary feedwater system normal use during startup, normal shutdown, and hot st andby conditions, th e turbine-driven auxiliary pump feedwater control valves are maintained normally fully open.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-36Rev. 30During startup, normal shutdown, and hot standby conditions, the motor-driven auxiliary feedwater pumps may be aligned to take suction from the non-safety grade condensate storage tank (CST). Motor-driven auxiliary feedwater pump suction automatically switches to the demineralized water storage tank (DWST), including isolation from the CST, in the event of an SIS, LOP, CDA, two of four low-low water level condition in any one steam generator, or AMSAC signal. Relative to a low-low SG level condition operator action to realign the motor-driven auxiliary feedwater pump suction source may be required in this mode of operation.3.Regulatory Position C.3Switches for manual initiation are located in the control room in such a manner as to permit deliberate expeditious action by the operator.4.Regulatory Position C.4a.Equipment common to both manual and automatic initiation will be minimized. Where manual and automatic action sequencing functions and interlocks that contribute to the protective action are common, component or channel level initiation will also be provided in the control room.b.Manual initiation portions of the protection system meet the single failure criterion.c.Manual initiation portions of the protection system will not impair the ability of the automatic system to meet the single failure criterion.5.Regulatory Position C.6Manual initiation portions of the protectio n system are design ed such that once initiated, a protective action at the system level (initiation of the final action device associated with a given protective function) goes to completion.Having gone to completion (i.e., when sufficient breakers are closed or sufficient MOVs or other actuators are operated), a device shall be returned to its pre-initiation status only by deliberate operator action. This action shall be similar in nature for all protection systems.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-37Rev. 30This design is in compliance with the applicable section of IEEE 279 (Paragraph 4.16).In addition, manual initiation is provided to allow the operator to take early action based on observation of plant parameters. It is not to be treated as a "backup" to automatic features. Operator actions will not be required to compensate for single failures.1.63Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants (Rev. 2, July 1978)Comply, with the following clarifications:8.31.The single failure provision of R.G. 1.63 shall apply to both Class 1E and non-1E overcurrent protection devices.2.An acceptable method of compliance with the "Single Failure Criterion" of R.G. 1.63 may be the use of redundant or backup interrupting devices. Tripping coordination between primary and backup interrupting devices is not required.3.While satisfying the "Single Failure Criterion" in IEEE 279-1971, Section 4.2, the overcurrent protective devices are not required to comply with other criteria listed in IEEE-279-1971.4.Unless required for other considerations, the protection schemes and fault isolating devices need not be Class 1E or seismically qualified for protection of the penetrations.Overcurrent protection devices are not within the scope of IEEE 279-1971 as written. However, those principles developed in IEEE 279, whic h ensure a highly reliable design will be used for guidance in the protection system design.1.64Quality Assurance Requirements for the Design of Nuclear Power Plants (Rev. 2, June 1976)
(3) 1.(1) Prior to 12/14/05, Millsto ne 3 complied as follows:
Construction - Millstone 3 complied with Rev. 0 of the Guide Operation - Millstone 3 complied with Rev. 2 of the Guide.2.Current compliance is as described in the QAPD Topical Report.17.11.65Materials and Inspections for Reactor Vessel Closure Studs (Rev. 0, October 1973)(1) See Section 1.8N 5.3.1TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-38Rev. 301.66Nondestructive Examination of Tubular ProductsWithdrawn1.67Installation of Overpressure Protection Devices (Rev. 0, October 1973)(1) Comply5.4.11.31.68Initial Test Programs for Water-Cooled Nuclear Power Plant (Rev. 2, August 1978)Millstone 3 initial startup test program complies with R.G. 1.68 with exceptions as stated in FSAR Section 14.2.7.14.2.71.68.1Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants (Rev. 1, January 1977)
Not applicable.
Applicable only to BWRs.1.68.2Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants (Rev. 1, July 1978)
Comply14.2.71.68.3Preoperational Testing of Instrument and Control Air Systems (Rev. 0, April 1982)Millstone 3 initial startup test program complies with R.G. 1.68.3 with exceptions and clarifications as stated in FSAR Section 14.2.7.
9.3.114.2.71.69Concrete Radiation Shields for Nuclear Power Plants (Rev. 0, December 1973)
Comply12.3.21.70Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Rev. 3, November 1978)Comply 1.11.71Welder Qualification for Areas of Limited Accessibility (Rev. 0, December 1973)(1) Comply, with the following clarification:
5.2.3An acceptable alternative to this position is contained in the following exception: In lieu of Paragraphs C.1 and C.2a, all applicable welds of limited accessibility are volumetrically inspected to the requirements and standards of ASME III, Class 1.5.3.1.410.3.6.2TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-39Rev. 301.72Spray Pond Piping Made from Fiberglass-Reinforced Thermosetting Resin (Rev. 2, November 1978)Not applicable.Fiberglass pipe is not used for QA Category I applications on Millstone 3.1.73Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants (Rev. 0, January 1974)(1) Comply3.11.2 8.3.11.74Quality Assurance Terms and Definitions (Rev. 0, February 1974)(1) Comply17.1.217.21.75Physical Independence of Electric Systems (Rev. 2, September 1978)(1) Comply, with the following exceptions and clarifications:
7.11.General (Clarification 8.3.1.4Ventilated tray covers are considered equivalent to solid tray covers.Short lengths of cable (generally less than 10 feet) enclosed in a protective wrap of woven silicon dioxide are considered to be protected from electrically induced problems in adjacent cables to the same degree as the same cable in an enclosed raceway. The protective wrap of woven silicon dioxide (trade name -SIL-TEMP) is 54 mils thick and is wrapped longitudinally around cable(s) with a 50% overlap to ensure that cable(s) is enclosed by one thickness of the protective wrap. Metal clad cable, type MC, utilized in low energy, 120 V AC and 125 V DC nominal, circuits and in low density applications is considered adequate protection. As such, the minimum separation between these cables and other cables, or raceway (where required) is 1 inch. These cables are further described as follows:1.Type MC cable is a factory assembly of conductors, each individually insulated, enclosed in a metallic sheath of interlocking tape or a smooth or corrugated tube.2.Largest conductor size number 10 AWG.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-40Rev. 303.No more than three number 10 AWG conductors with remaining conductors of smaller size.4.Aluminum sheath cable (a Type MC cable in which the aluminum is continuously welded) may have an overall jacket of neoprene or hypalon.2.Position C.1 The power circuits for the non-Class 1E pressurizer heaters, control rod drive mechanism cooling fans, and containment air recirculation fans connected to Class 1E power sources are provided with two separate Class 1E breakers connected in series. In addition, the interconnecting cables (i.e., from power source to load) are identified by the same color code as the Class 1E power source to which they are connected.Power circuits for other non-Class 1E equipment connected to Class 1E power sources are provided with two separate Class 1E breakers or fuses connected in series. In addition, the interconnecting cables are identified by the same color code as the Class 1E power source to which they are connected (i.e., from power source up to and including the second breaker). In general, cable from the second breaker to the load are routed in rigid conduit, or routed in raceway of the same color as the power source. Refer to Table 8.3-3 for description of the routing of the interconnecting cable.The controlled routing (i.e., continuation of the circuit with the same color code or continuation of the circuit in rigid conduit) ensures the physical and electrical independence of the power circuit beyond the Class 1E isolation device (i.e.,
batter charger, isolation transformer, two series connected interrupting devices circuit breakers, fuses) or circuit breakers that trip on accident or loss-of-power signals.For those cables not routed in rigid co nduit or identified with the same color beyond the isolation device, the use of a Class 1E isolation device ensures the electrical independence of the Class 1E power source to the non-Class 1E equipment.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-41Rev. 30Coordination between the two series connected Class 1E breakers or fuses is not required. Coordination between the two series connected Class 1E breakers or fuses and the Class 1E main supply breaker is provided.3.Position C.4 (Clarification)
Associated circuits are identified by the same color code as the Class 1E circuit with which they are associated. This color code exists up to and including an isolation device, except as discussed under Position C.1.Associated circuits meet all other requirements of Class 1E circuits up to and including the isolation device.4.Position C.6 (Clarification)
Analyses of potential hazards in Section 5.1.1.1 of IEEE 384 are accomplished as follows:1.The high pressure piping and missile analyses are described in FSAR Sections 3.6 and 3.5, respectively.2.The fire protection analyses are outlined in FSAR Section 9.5.1 and the Fire Protection Evaluation Report.3.Cable that is not flame retardant is enclosed in a dedicated raceway for the entire length of the run.4.The building design for flooding is described in FSAR Section 3.4.
Analysis for establishing minimum separation distances as allowed for Section 5.1.1.2 of IEEE 384 for use in areas defined by Section 5.1.3 and 5.1.4 of IEEE 384 is accomplished by the following:Test and analysis performed to determine the separation requirements between Class 1E and non-Class 1E is presented in Wyle Test Report No. 47506-02.On a case-by-case basis, Wyle Test Report No. 47506-02 will be utilized to justify acceptable deviations to the electrical separation criteria between Class IE circuits as well as Class IE and non-Class IE circuits. These deviations located in areas outside of panels, will be listed in Specification SP-EE-076.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-42Rev. 305.Position C.7 (Section 4.6 of IEEE 384)
Where plant arrangements preclude maintaining the minimum separation distance between Class 1E and non-Class 1E circuits, either the Class 1E or the non-Class 1E circuit shall be run in an enclosed raceway or a barrier shall be provided between the circuits. Other barriers may be installed as illustrated in Figures 2 through 5 of IEEE 384-1974. The minimum distance from barriers or between enclosed raceway and exposed circuits is 1 inch. (Exception: The copper feeders to the reactor coolant pumps and circulating water pumps will meet the separation requirements of R.G. 1.75 and IEEE 384-1974.)The minimum separation between Class 1E and non-Class 1E enclosed raceways of "X," "C," and "K" service is 1/8 inch.Position C.9 Proposed splices in raceways will be evaluated on a case-by case basis and documented in Specification SP-EE-076.6.Position C.10 Class 1E cable and raceways shall be marked at intervals not exceeding 15 feet. The 5-ft. requirement is a typographical error which has been confirmed by the NRC.7.Position C.12 1.Power cables that supply power to in strument rack room and control room distribution panels, limited to 120 V AC and/or 125 V DC, are:*Enclosed in rigid conduit in the cable spreading room. The rigid conduit is either aluminum or steel.*Enclosed in rigid conduit with flexible conduit at entrance to the panels in the instrument rack room and control room.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-43Rev. 302.Power cables (from the ab ove distribution panels) to facilities serving the control room and instrument rack room, limited to 120 V AC and/or 125 V DC, are enclosed in rigid conduit except at entrance/exit to floor sleeves in the cable spreading room, instrument rack room and control room, and at entrance to equipment in the instru ment rack room and control room.3.Other power cable (4,160 V, 480 V, and 120 V AC service) that traverses the cable spreading room are enclosed in rigid steel conduit.4.The loss of the above cables or the control room, instrument rack room, or the cable spreading room due to the design basis event fire will not compromise the capability to achieve cold shutdown as outlined in the Fire Protection Evaluation Report.5.The Millstone 3 design utilizes a single cable spreading room.8.Position C.16 (Section 5.6.2 of IEEE 384) 1.The minimum 6 inch separation (or a barrier) applies to spacing between exposed terminals, contacts, and equipment of redundant Class 1E circuits for testing and maintenance purposes. A minimum of 1 inch separation (or a barrier) is required between redundant wire bundles or Class 1E and non-Class 1E wire bundles. The minimum of 1 inch separation is sufficient since the control boards are protected from and/or are not subject to hazards such as external fire, flooding, high energy piping, and missiles. Internal electrical fires are not considered a hazard due to fire retardant materials and low energy application.2.For internal to control room panels and cabinets (specifically 3CES*MCB-MB1 through MB8 and 3HVS*PNLVP1), the minimum separation distance between redundant Class 1E and non-Class 1E circuits is as follows:A minimum 1 inch separation (or a barrier) between exposed contacts or terminals.A minimum 1 inch separation (or a barrier) is required between redundant wire bundles or Class 1E and non-Class 1E wire bundles.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-44Rev. 30Test and Analysis performed to determine the reduced separation requirements within control room panels (specifically 3CES*MCB-MB1 through MB8 and 3HVS*PNLVP1) for redundant Class 1E or Class 1E and non-Class 1E is presented in Wyle Test Report 46317-1.1.76Design Basis Tornado for Nuclear Power Plants (Rev. 0, April 1974)
Comply3.3.2.11.77Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors (Rev. 0, May 1974)(1) Comply with the following exception. Regulatory Position C.3 no longer applies and has been replaced by R.G. 1.183.15.4.215.4.71.78Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release (Rev. 0, June 1974)Comply, with the following clarifications:6.4.4.2The assumptions used for identifying chemicals potentially hazardous to the control room are in accordance with R.G. 1.78, dated June 1974. Chemicals not known or projected to be present within a 5 mile radius of the reactor facility are not considered in the evaluation, and no specific design features are provided for the chemicals listed in Table C-1 of the Regulatory Guide. Hazardous chemicals that are known or projected to be used, transported, or stored within 5 miles of the reactor facility are considered in the evaluation of control room habitability. If the potential buildup of a specific hazardous chemical is slow, so that the time from detection to incapacitation is greater than 2 minutes, human detecting is used as appropriate. If the potential buildup of a specific hazardous chemical exceed s the toxic limit, automatic detection and isolation, low leakage design features, and pressurization, if necessary, are provided to ensure that the control room remains habitable. In this case, specific design features are included:Manual control room ventilation isolation, coating of conc rete and concrete block surfaces of control room with a suitable surface treatment to reduce leakage due to porosity, cracks, and construction joints in the control room.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-45Rev. 30Sealing of all pipes, ducts, and electri cal penetrations into the control room envelope.Compression seals for access doors and equipment removal hatc hes in the control room.In order to ensure control room habitability for design basis accidents, the following are provided:Maintenance of 0.125 inch wg positive pressure.Two tight butterfly dampers in series in each intake.These features provide the control room operator with the ability to isolate the control room. Hazards which could threaten control room habitability have been evaluated. The evaluation is consistent with the guide, and all hazards within 5 miles of the site are included. As a result of this evaluation, no commitment to provide hazardous chemical detectors has been made.Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors (Rev. 1, September 1975)Millstone 3 initial startup test program complies with R.G. 1.79 Rev. 1 with exceptions as stated in FSAR Section 14.2.7.14.21.80Preoperational Testing of Instrument Air SystemsWithdrawn1.81Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants (Rev. 1, January 1975)
(3) Not applicable.8.1Applicable only to those plants which share emergency and shutdown electrical systems in a multi-unit plant, which Millstone 3 does not.1.82Sumps for Emergency Core Cooling and Containment Spray Systems (Rev. 0, June 1974)(1) Comply, with the following clarifications:6.2.2.2C.1 The recirculation spray pumps take suction from a single sump. The sump and strainer were designed to eliminate any credible failure mechanisms which would require installation of a redundant sump or strainer and is considered especially qualified for service and exem pt from passive failure. (See section 3.1.1.3)6.2.2.4.2 TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-46Rev. 30C.3 The strainer is made of hollow fins constructed of perforated plate which is structurally robust. The strainer fins are located above the floor, which protects them from large rolling debris. Therefore trash racks are not required.C.8 The strainer is fully submerged at the start of the spray pumps. There is no top deck, however, there is non QA cover plate over the strainer fins to protect the strainer from damage during outages.1.83Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes (Rev. 1, July 1975)(1) Comply5.4.2.216.3 /4.4.51.84Design and Fabrication Code Case Acceptability - ASME Section III, Division 1(1) Comply5.2.1.21.85Materials Code Case Acceptability - ASME Section III, Division 1(1) Comply5.2.1.21.86Termination of Operating Licenses for Nuclear Reactors (Rev. 0, June 1974)
Comply 5.8 (EROLS)1.87Guidance for Construction of Class 1 Components in Elevated-Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1593, 1594, 1595, and 1596) (Rev. 1, June 1975)
Not applicable.Applicable only to Elevated Temperature ReactorsTABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-47Rev. 301.88Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records (Rev. 2, October 1976)
(3) 1.(1) Prior to 12/14/05, Millsto ne 3 complied as follows:
Construction Millstone 3 complies with R.G. 1.88, Rev. 0, August 1974 as noted in Appendix VII of the Millstone 3 Quality Assurance Program Manual.
Operation Millstone 3 complies with R.G. 1.88, Rev. 2, October 1976, with the exception that the records storage vault door and hardware have a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rating instead of the recommended 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> minimum stated in ANSI N45.2.9. This position is noted in the Quality Assurance Topical Report.2.Current compliance is as described in the QAPD Topical Report.17.11.89Qualification of Class 1E Equipment for Nuclear Power Plants (Rev. 0, November 1974)(1) Class 1E equipment other than that within the NSSS Scope of Supply complies with R.G. 1.89, dated November 1974, with the following clarifications:3.11.21.Determination of the radiation dose used for qualification of Class 1E plant equipment will take into account design features, such as the location of equipment within or outside the containment, fission product cleanup of the containment atmosphere by the containment spray system, local shielding, the time period required for equipm ent operation and spatial location.These design features will be applied in a conservative manner to realistically determine the radiation doses to which the devices must be qualified in addition to the other environmental factors.The radiological source term as defined in R.G. 1.7 includes a conservative margin well in excess of that required to qualify Class 1E components.2.Qualification testing of organic materials in beta radiation environments will not be required. The effect of beta radiation will be accounted for by a weighted addition of calculated gamma and beta doses and specifying a given does in rads with qualification testing in a gamma environment treated as sufficient qualification.The clarification as stated above is similar to the position taken by the IEEE.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-48Rev. 301.90Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons (Rev. 1, August 1977)
Not applicable.Millstone 3 does not have a prestre ssed concrete containment structure.1.91Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants (Rev. 1, February 1978)
Comply2.2.3.1.11.92Combination of Modes and Spatial Components in Seismic Response Analysis (Rev. 0, December 1974) (2) (1) Comply, with the following clarification:
3.7.2The combination of modes and spatial components in seismic response analysis satisfies the requirements of R.G. 1.92, Rev. 1, dated February 1976. The time history dynamic analysis uses three statistically independent (maximum correlation factor of 0.2) orthogonal ground accelerations (two horizontal and one vertical) of the prescribed earthquake input simultaneously as prescribed in R.G. 1.92, Rev. 1.Computation of structural responses due to input of three simultaneous earthquakes is more accurate and simpler than computing them for three separate earthquakes.1.93Availability of Electric Power Sources (Rev. 0, December 1974) 8.11. One emergency diesel generator may be inoperable for up to 14 days.8.32.Preventive, as well as corrective, main tenance will be performed during plant operation within the constraints of the appropriate Technical Specification allowed outage time.3. Millstone Unit 3 station batteries No. 3 & 4 (301A-2 & 301B-2 respectively) have a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOT (Allowable Out of Service Time).TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-49Rev. 301.94Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants (Rev. 1, April 1976)
(3) 1.Prior to 12/14/05, Millstone 3 complied, with the following exceptions:a.Millstone 3 will comply with the requirements of ANSI N45.2.5-1974 except that correlation testing shall be performed in accordance with the applicable paragraphs of Section 6.11 of N45.2.5-1978.3.8.1.6b.Admixture manufacturer shall submit certified test data confirming admixture complies with ASTM C260 when tested in accordance with ASTM C233. For each production lot shipped, the manufacturer shall certify that the admixture is similar to the material represented by the test data.2.No longer comply - 12/14/05 (QA standards are described in QAPD Topical Report.)17.11.95Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release (Rev. 1, January 1977)Comply, with the following clarifications and exceptions:6.4.4.2In accordance with R.G. 1.78, R.G. 1.95, and Section 2.2, no off site chlorine storage or transportation is close enough to the plant nor frequent enough to be considered a hazard. There is no on site chlorine that is considered a hazard under R.G. 1.95. A sodium hypochlorite biocide system is used and no specific control room design features are provided for chlorine.1.96Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants (Rev. 1, June 1976)
Not applicable.
Applicable only to BWRs.1.97Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Rev. 2, December 1980)(1) Millstone 3 compliance with Regulatory Guide 1.97, Rev. 2, is found in specification SP-M3-IC-022 titled, Millstone 3 Design Basis to Respond to Regulatory Guide 1.97, Rev. 2.
7.4 7.5TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-50Rev. 301.98Assumptions Used for Evaluating the Potential Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor (Rev. 0, March 1976)
Not applicable.
Applicable only to BWRs.1.99Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials (Rev. 1, April 1977)(1) See Section 1.8N.
5.31.100Seismic Qualification of Electric Equipment for Nuclear Power Plants (Rev. 1, August 1977)(1) Replacement items meet the original criteria of IEEE-344-87 (Endorsed by Regulatory Guide 1.100, Rev. 2.)1.101Emergency Planning for Nuclear Power Plants (Rev. 2, October 1981)
Comply13.31.102Flood Protection for Nuclear Power Plants (Rev. 1, September 1976)Comply with the following clarification:3.4.1Position C.1.c for Incorporated Barriers states that the plant should be designed and operated to keep doors necessary for flood protection closed during normal operation.Incorporated Barriers for each service water pump cubicle includes a watertight door and an isolated sump drain line. During normal operations, the drain line of each cubicle is open and the door of one cubicle only is open. Plant procedures require isolating the sump drain line and watertight door of each cubicle on approaching severe weather. These actions are performed prior to sea level rising to the level which requires entry into the LCO for Technical Specification 3.7.6.A Technical Specification to ensure closing of the cubicle watertight doors in advance of a potential flooding event was required by Section 2.4 of the SER for MP3. Section 2.4 of the SER concluded that the guidelines of Regulatory Guide 1.102 had been met. The NRC therefore recognized that the cubicle watertight doors may be open during normal operations and found this to be acceptable.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-51Rev. 301.103Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments (Rev. 1, October 1976)Withdrawn1.104Overhead Crane Handling Systems for Nuclear Power PlantsWithdrawnThe withdrawal of Regulatory Guide 1.104 does not affect licensing commitments for the design of single-failure-proof cranes made on the basis of the Guide.1.105Instrument Setpoints (Rev. 1, November 1976)(1) Comply7.1.2.1.97.2.2.2.17.3.1.2.61.106Thermal Overload Protection for Electric Motor on Motor-Operated Valves (Rev. 1, March 1977)
(3) Comply 7.18.3.1.1.51.107Qualifications for Ceme nt Grouting for Prestressing Tendons in Containment Structures (Rev. 1, February 1977)
Not applicable.1.108Periodic Testing of Diesel Generator Units Used as On site Electric Power Plants (Rev. 1, August 1977)Comply, with the following clarifications and exceptions:8.3.1Section C.2(a)2: Proper operation for design-accident-loading-sequence will be demonstrated under conditions as close to design as possible.Section C.2(a)9: Comply as stated in the ERRATA dated September 1977.
Section C.2(d): If the number of failures in the last 20 valid tests is less than or equal to one, the test frequency is once per 31 days.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-52Rev. 30If the number of failures in the last 20 valid tests is greater than or equal to two, the test frequency is once per seven days.If the number of failures in the last 100 valid tests is less than or equal to four, the test frequency is once per 31 days.If the number of failures in the last 100 valid tests is greater than or equal to five, the test frequency is once per 7 days.The above testing clarification was approved by the NRC in License Amendment 64, issued March 9, 1992.Section C2(a): These tests will be conducted at the frequency specified in the Surveillance Frequency Control Program.
1.109Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I (Rev. 1, October 1977)
Comply13.3.11.110Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (Rev. 0, March 1976)
Comply1.111Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors (Rev. 1, July 1977)Comply2.3.5.2.31.112Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors (Rev. 0, April 1976)
Comply11.2.1.311.3.315.7TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-53Rev. 301.113Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I (Rev. 1, April 1977)Comply2.4.121.114Guidance on Being Operator at the Controls of a Nuclear Power Plant (Rev. 1, November 1976)
Comply13.1.31.115Protection Against Low-Trajectory Turbine Missiles (Rev. 1, July 1977)Comply, with the following clarification:3.5.1.3.3.The turbine missile analysis complies with the Guide to the extent that the calculated total probability hazard rate is less than 1.0E-7 per annum. This is achieved by (1) assigning a 1.0E-2 value to strike and damage probability product and (2) implement maintenances and testing program to maintain the probability of turbine failure less than 1.0E-5 per annum.1.116Quality Assurance Requirements for Installation, Inspection, and T esting of Mechanical Equipment and Systems (Rev. 0-R, May 1977)
- 1. (1) Prior t 12/14/05, Millstone 3 complied.17.12. Current compliance is as described in the QAPD Topical Report.1.117Tornado Design Classification (Rev. 1, April 1978)Comply, with the following clarification:3.2.51.Paragraph 3 Appendix, "Structure, Systems, and Components of Light-Water-Cooled Reactors to be Protected Against Tornados."The statement:"3. The reactor core and individual fuel assemblies, at all times, including during refueling"Is clarified as:TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-54Rev. 30Protection during refueling is provided by the containment equipment hatch while the concrete missile shield blocks are removed. A probabilistic analysis demonstrated that the mean value of a tornado missile impacting the equipment hatch during refueling meets the NRC's acceptance criterion of less than or equal to 10
-6 per year.2.Paragraph 4.(4) Appendix, "Structure, Systems, and Components of Light-Water-Cooled Reactors to be Protected Against Tornados."The statement:"4. Systems or portions of systems that are required for...(4) mitigating the consequences of a tornado-caused PWR streamline break..."
Is interpreted as:Protection of systems and components for which credit is taken in the analysis of PWR steamline break outside containment.Clarification of Paragraph 4.(4) results from telephone communication between L. P. Walker, SWEC, and G. Chipman, USNRC (AAB) June 19, 1978. Mr.
Chipman, who is the author of Paragraph 4, stated that the intent of this requirement is to ensure tornado missile protection for valves, lines, water storage tanks (e.g., demineralized water storage, etc.) that are required to mitigate a steamline break outside of containment.1.118Periodic Testing of Electric Power and Protection Systems (Rev. 2, June 1978)
(3) (1) Comply7.1.2.4 8.1.61.119Surveillance Program for New Fuel Assembly DesignsWithdrawn1.120Fire Protection Guidelines for Nuclear Power Plants (Rev. 1, November 1977)Comply, with the following clarification:Since R.G. 1.120 has been deleted from SRP 9.5.1, the provisions of BTP CMEB 9.5-1, Rev. 2, July 1981 are followed.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-55Rev. 301.121Bases for Plugging Degraded PWR Steam Generator Tubes (Rev. 0, August 1976)(1) See Section 1.8N.
5.4.21.122Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components (Rev. 1, February 1978)Not applicable.3.7.2.5The implementation section of R.G. 1.122 states that the guide will be used in the evaluation of construction permit applications. The mathematical methods suggested in the Guide were not applied to the original Millstone 3 design since the development of the floor design response spectra had been completed prior to the issuance of the guide in February 1978.A description of the Millstone 3 methodology, using techniques which were state-of-the-art at the time of the ARS Development follows:1.Millstone 3 seismic design is based on three independent orthogonal components of earthquake motion as described in Section 3.7.1.1.2.Millstone 3 ARS peaks are broadened plus and minus 15% in all cases, and the broadened peaks are bounded by vertical lines.3.For both symmetrical and non symmetrical structures, the floor response spectra corresponding to the direction of input is used.1.123Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Rev. 1, July 1977) 1.(1) Prior to 12/14/05, Millsto ne 3 complied as follows:
Construction Comply with the alternative that certain standard or non engineered items may be procured without seller qualification as described in Section 7 of the Millstone 3 Quality Assurance Program Manual.
Operation Comply as described in the Quality Assurance Topical Report.2.Current compliance is as described in the QAPD Topical Report.17.11.124Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports (Rev. 1, January 1978)Comply, with the following exceptions and additions:3.61.The following paragraph should be added to Regulatory Position C.3:5.4.14TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-56Rev. 30C.3.c. The bending stress limits F b resulting tension and bending in structural members as specified in Appendix XVII 2214 of Section III, Div. 1, should be the smaller value of 0.66 S y or 0.55 S u for compact sections, 0.75 S y or 0.63 S u for doubly symmetrical members with bending about the minor axis, 0.6 S y or 0.5 S u for box-type flexural members and miscellaneous members.
The paragraph added to Re gulatory Position C.3 is necessary because of an apparent oversight in applying the 5/6 factor to bending stress allowables.2.The second paragraph under Regulatory Position C.4 should be replaced with the following:However, all increases (i.e., those allowed by NF-3231.1 (a), XVII-2110 (a), and F-1370 (a) shall always be limited by XVII-2110 (b) of Section III. The critical buckling strengths defined by XVII-2110 (b) of Section III should be calculated using material properties at temperature.The increased allowable permitted for tensile stress in bolts shall not exceed the lesser of 0.70 S y or S u at temperature. The increased allowable permitted for shear stress in bolts shall not exceed 0.42 S u at temperature.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-57Rev. 30The third sentence in the second paragraph of Regulatory Position C.4 prohibits the application of the increased allowables presently permitted by NF-3231.1 (a) and F-1370 (a) to Service Limits A or B for bolted connections. The danger of applying the increases presently allowed by Subsection NF has been pointed out at Subsection NF has been pointed out at Subsection NF Committee meetings.
The Millstone 3 position asserts that maximum safe incre ased allowables are achieved by limiting bolting tensile stress to the less of 0.7 S y or S u at temperature and bolting shear stress to 0.42 S u at temperature. The 0.7 S u limit is well recognized in Section III of the Code. The average shear strength of bolting material is about 0.62 S u according to test data, with a standard deviation of 0.033. Results indicate that the ratio of shear strength to tensile strength is independent of the bolt grade. Curves showing this appear on page 50 of Guide to Design for Bolted and Riveted Joints by J. W. Fisher. Test data are given in a paper by J. J. Wallaert and J. W. Fisher, Shear Strength of High-Strength Bolts, Journal of the Structural Division, ASCE, Volume 91, ST3, June 1965. 3.Paragraph C.5.a should be revised as follows:"The stress limits of XVII-2000 of Section III, and Regulatory Position 3 of this Guide, should not be exceeded for component supports designed by the linear elastic analysis method. These stress limits may be increased according to the provisions of NF-3231.1 (a) of Section III and Regulatory Position 4 of this Guide when effects resulting from constraint of free-end displacement and anchor motion are added to the loading combination."Loads developed by anchor motions are also deformation limited and, as such, are considered to be grouped in the same category as loads from restraint of free-end displacement. The resulting stresses are essentially of the secondary type.4.Regulatory Position C.8 should read as follows:TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-58Rev. 30Supports for the "active" components that are required only during an emergency or faulted plant condition and that are subjected to loading combinations described in Regulatory Positions C.6 and C.7 should be designed within the design limits described in Regulatory Position C.5 or other justifiable design limits. These limits should be defined by the design specification and stated in the SAR, such that the function of the supported system will be maintained when they are subjected to the loading combinations described in Regulatory Positions 6 and 7.Regulatory Position C.8 is revised as shown because this section implies that the lower stress limits associated with the Design Levels A and B Service Limits must be used for any component support that serves a safety-related function during an Emergency or Faulted Plant condition. This would seem to imply that a main coolant pump support, which constitutes a passive element in the main coolant loop, would have to be designed to meet the Design, Level A and B Service Limits during an Emergency or Faulted (LOCA) plant condition. This would require that a snubber providing restraint on an RHR line would have to be designed to the Design, Levels A and B Service Limits during an Emergency or Faulted plant condition. If this is the intent, it is a severe departure from current practice. Only active components, such as valves, whose operation is required for safe shutdown during an Emergency or Faulted condition, have been required to meet design stress limits for these plant conditions. Levels C and D service limits have been considered adequate to assure pressure boundary integrity under the more se vere operating conditions.1.125Physical Models for Design and Operation of Hydraulic Structures and Systems for Nuclear Power Plants (Rev. 1, October 1978)
Comply1.126An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification (Rev. 1, March 1978)(1) See Section 1.8N.
4.2TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-59Rev. 301.127Inspection of Water-Control Structures Associated with Nuclear Power Plants (Rev. 1, March 1978)Not application.9.2.1Applies to water control structures; i.e., da ms, reservoirs, conveyance facilities, etc., specifically for use in conjunction with a nuclear power plant. Millstone 3 is located on Long Island Sound and, as such, does not require dams or reservoirs for water impoundment.The intake and discharge structures are Category 1 and, as such, are built to stringent design and construction requirements. Normal maintenance during the course of operation of the plant would detect an y abnormal conditions in these structures.1.128Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plant (Rev. 1, October 1978)
(3) Comply 8.1.6 8.3.21.129Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants (Rev. 1, February 1978)
(3) Comply - Note: This Reg. Guide endorses IEEE 450-1975 with an additional requirement that the "Service Test" be performed in addition to the "Performance Discharge Test" during refueling outages. The Technical Specification basis references IEEE Std. 450-1980 for battery capacity test procedures and schedule. Sections 5 and 6 of 450-1980 replace 450-1975. Guidance on bypassing weak cells, if required, is in accordance with section 7.4 of IEEE 450-2002. The balance of 450-1975 applies to MP3.8.1.6 8.3.21.130Design Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports (Rev. 1, October 1978)Comply, with the following clarifications and exceptions:1.Regulatory Position C.3 should be revised as follows: "Service limits for component supports designed by linear elastic analysis should always be limited by the critical buckling strength. The critical buckling strength should be calculated using material properties at temperature. Conservative factors of safety for flat plates and for shells should be maintained for each design and service limit. The allowable stress for Service Limit D should not exceed two-thirds of the critical buckling stress."TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-60Rev. 302.Regulatory Position C.7 should read as follows: "Support for 'active' components that are required only during an emergency or faulted plant condition and that are subjected to loading combinations described in Regulatory Position C.5 and C.6 should be designed within the design limits described in Regulatory Position C.4 or other justifiable design limits. These limits should be defined by the design specification and stated in SAR, such that the function of the supported system will be maintained when they are subjected to the loading comb inations described in Regula tory Positions C.5 and C.6."Regulatory Position C.7 implies that the lower stress limits associated with the Levels A and B Service Limits must be used for any component support that serves a safety related function during an Emergency or Faulted (LOCA) plant condition. This would seem to imply that a main coolant pump support, which constitutes a passive element in the ma in coolant loop, would have to be designed to meet the Levels A and B Service Limits during an Emergency or Faulted plant condition. If this is the intent, it is a severe departure from current practice. Only active components, such as valves, whose operation is required for safe shutdown during an Emergency or Faulted condition, have been required to meet design stress limits for these plant conditions. Levels C and D Service Limits have been considered adequate to assure pressure boundary integrity under the more se vere operating conditions.1.131Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear
Power Plants (Rev. 0, August 1977)
Comply1.132Site Investigations for Foundations of Nuclear Power Plants (Rev. 1, March 1979)Comply with the following exceptions and clarifications:2.51.Paragraph B.5: Surveys of horizontal deviation are made in all boreholes that are used for cross-hold seismic tests.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-61Rev. 30The requirements for surveys of vertical deviation appear to be in error. Horizontal deviations must be measured in borings used for cross-hole seismic surveys in order to determine the true distance between energy source and receiver.2.Paragraph C.1, Item (5): Only typical time-distance plots should be included.Time-distance plots are not usually included on geologic profiles. A typical time-distance plot is more appropriate.3.Paragraph C.2: Test results of field permeability tests an d borehole logging should be tabulated and/or graphed.Field and laboratory test results are presented in tables and figures especially designed for these tests. Boring logs do not typically include field and laboratory test results.4.Paragraph C.3: Measurement of water or drilling mud levels in borings not required in all cases.Water or drilling mud levels in some mate rials, such as clays, may give false information about groundwater levels. A sufficient number of observations wells monitor groundwater levels.5.Paragraph C.6: Continuous undisturbed samples will be taken in compressible or normally consolidated clays only if required for geotechnical analysis.The need for continuous undisturbed samples is a matter of engineering judgment and is evaluated for each case.6.Appendix C: Spacing and depth requirements for borings under structures and pipelines are not in conformance for all cases.Borings for structures and pipelines were performed prior to issuance of the regulatory guide. It is considered that sufficient data are available concerning foundation conditions, bedrock quality, and bedrock contours from existing borings and the extensive geologic bedrock mapping program.7.Paragraph B.2s: No explorations have been conducted off site.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-62Rev. 30The area surrounding the Millstone site has been mapped by the U.S. Geological Survey and the Connecticut State Geological and Natural History Survey.The closest fault shown on any of these maps is approximately 10 miles northeast of the site in the Uncasville quadrangle. Detailed mapping of the 5 mile radius could uncover faults similar to those uncovered during the excavations of Millstone 3. However, detailed investig ations at the Millstone site have demonstrated the incapability of these faults. Studies performed at other sites have verified that no capable faults are known to exist in New England.1.133Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors (Rev. 1, May 1981)Comply, with the fo llowing exceptions:1.Position C.1.gThe loose-part detection system need not be qualified to an OBE, however, the equipment inside containment has been demonstrated to be functional following an OBE.OBE qualification is in excess of the requirements placed on other alarms of equal importance in disclosing failures during plant operation.2.Position C.5.a The location of the required sensors is not needed in the Technical Specifications since their locations are delineated in Section 4.4.6.4 of MNPS-3 FSAR.3.Position C.5.cCalibration will be verified at least once per fuel cycle or every 18 months, whichever is greater.This calibration period is adequate because the loose-part detection system is not safety-related.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-63Rev. 301.134Medical Evaluation of Nuclear Power Plant Personnel Requiring Operator Licenses (Rev. 1, March 1979)
Comply1.135Normal Water Level and Discharge at Nuclear Power Plants (Rev. 0, September 1977)Comply, with the following clarification:R.G. 1.59, Rev. 2, is used to determine initial water levels for design basis flood analysis.1.136Material for Concrete Containments (Rev. 2, June 1981)
Not applicable.The following Regulatory Guides have been incorporated into R.G. 1.136, Rev. 2, and have been withdrawn:*1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures.*1.15 Testing of Reinforcing Bars for Category I Concrete Structures.*1.18 Structural Acceptance Test for Concrete Primary Reactor Containments.*1.19 Nondestructive Examination of Primary Containment Liner Welds.*1.55 Concrete Placement in Category I Structures.*1.103 Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments.Withdrawal of the above Guides does not alter any prior existing licensing commitments based on their use.1.137Fuel-Oil Systems for Standby Diesel Generators (Rev. 1, October 1979)Comply with the following clarifications and exceptions:9.5.4The Millstone 3 design provides approximately 3-day diesel fuel storage tanks which are interconnected with normally closed valves. The interconnected tanks provide approximately a 6-day supply of fuel oil for either of the diesel generators.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-64Rev. 30This was approved by the NRC in License Amendment No. 97 issued October 17, 1994.The EDG Fuel Oil System does not meet the recommended configuration in Section 6.3 of ANSI N159 concerning strainers. A single type strainer is in the discharge of the A and B pumps and no strainer is in the discharge of the C and D pumps. Procedural controls and sampling from the TK1A and TK1B tanks assure that fuel oil quality meets or exceeds the EDG manufacturer's acceptance criteria.
In addition, the followi ng exception is taken:Analysis of the fuel properties listed in the applicable specifications are completed within 30 days after fuel addition rather than 2 weeks.Section C.1.e(2): The requirement of "pressure testing of the fuel oil system to a pressure 1.10 times the system design pressure at 10 year intervals..." has been deleted. This was approved by the NRC in License Amendment 110 issued May 1, 1995.1.138Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants (Rev.
0, April 1978)Comply, with the following clarifications and exceptions:2.51.Paragraph C.1.c: Standards used to calibrate laboratory test equipment are of a known higher accuracy than the test eq uipment, rather than four times more accurate than the working instrument.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-65Rev. 30Standards used to calibrate geotechnical laboratory equipment need only be more accurate, rather than at least four times more accurate, than the working instrument. Soils and rocks are materials whose properties vary widely within the same deposit or formation. In addition, certain physical properties of soils and rocks are greatly affected by sampling and by preparation for testing in the laboratory. The geotechnical engineer takes these natural variations and sampling/preparation effects into account, and exercises considerable judgment in assigning the material properties to be used in an analysis. (This differs from manufactured materials, where dimensions and physical properties are maintained within a narrow range by the manufacturing process.) Calibration of geotechnical laboratory equipment to higher standards than currently in use is not justified.2.Paragraph C.1.d: Index and classification tests are not performed on all soil and rock samples.Classification of soil and rock samples is performed by visual manual techniques. Index and classification tests are performed on representative samples to confirm the visual-manual classifications.3.Paragraph C.2: Moisture seals are not periodically checked, but are renewed as needed.Tube samples are inspected for obvious leakage when they are received. The moisture seals are inspected in detail when a tube has been selected for testing. Each sample is examined when it is extruded, and any evidence of drying in the tube is noted on the sample description log. This procedure is sufficient to evaluate whether drying has occurred. Samples that appear to have dried are not tested. Periodic inspection and replacem ent of moisture seals would be time consuming and would not provide any better protection against testing samples whose water content has changed.4.Paragraph C.2: Duration of storage is not specifically reported for each test.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-66Rev. 30The duration of storage can be calculated from the boring logs, where the sampling date is given, and the laboratory test data sheets, where the date of testing is reported. Therefore, it is unnecessary to make a separate listing of storage time.5.Paragraph C.3.a: Classification tests are not performed on every undisturbed test specimen of soil or rock.Visual-manual techniques are the primary means of classifying soil and rock samples. Classification tests are performed on representative samples as necessary to confirm the results of the visual-manual classifications. Therefore, it is not necessary to perform classification tests on all undisturbed samples have changed during shipment, storage, and handling.6.Paragraph C.3.2: Measurements and control tests are not performed to determine whether undisturbed samples have changed during shipment, storage, and handling.Undisturbed samples are visually inspected as they are opened and extruded, to determine whether there has been any change in sample length within the tube or if there are any signs of sample disturbance. Results of these inspections are reported on the sample description log. This procedure and examination of the laboratory test results fo r possible indications of sample disturbance, is sufficient to determine whether sample disturbance has occurred. Therefore, the sample does not have to be measured, weighted, or otherwise tested during shipment, storage, and handling.7.Paragraph C.3.b: A discussion of the validity of results on scalped materials is not presented as part of the laboratory test data.The laboratory test results indicate which portion of the sample has been scalped. Reasons for expecting test results to be valid are not given as part of the laboratory test data. A competent reviewer of the test results would understand the effect of scalping on the results of specific tests.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-67Rev. 308.Paragraph C.4.A (2): Results of tests with B-values less than 0.95 may be used for analysis in certain cases.For very stiff or hard clays, it may not be possible to achieve a B-value of 0.95. If results of such a test are used, the B-value is reported and the probable effect on results of the test could be evaluated by a competent reviewer.9.Paragraph C.5.a: Soil and rock identifications and descriptions are not documented so that an independent review can be performed.The only way to provide for independent review of all soil and rock descriptions is to provide the samples to the reviewer, without testing them. This is obviously impossible. The current practice of recording sample descriptions and determining index properties of representative samples is consistent with good engineering practice.10.Paragraph C.5.a: Anomalous test data are not reported if they are caused by sample disturbance or equipment malfunction.Anomalous test data caused by sample disturbance or equipment malfunction are not reported because the data do not reflect the true properties of the material in the field. However, records of such tests are maintained as part of the laboratory records.11.Appendix B, Relative Density: The frequency of the vibratory table is not adjusted.The frequency of the vibratory table cannot be adjusted, because it depends upon the fixed frequency of the input current.1.139Guidance for Residual Heat Removal (Rev. 0, May 1978)(1) Comply3.1.2.34TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-68Rev. 301.140Design, Testing, and Maintenance Normal Ventilation Criteria for Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants (Rev. 1, October 1979)Comply, with the following clarifications and exceptions:9.4Paragraph C.2.c The Technical Support Center filter train is instrumented to indicate, but not to alarm, filter flow within the habitability zone. The containment air filtration system is not instrumented to monitor or alarm filter flow. Surveillance procedures verify the flow rates of these two infrequently operated filter trains once per refueling interval.Paragraph C.2.f: Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510-1980. A leak rate of 1.0% in accordance with Table 4-3 of ANSI N509-1980 is acceptable. However, ductwork leak tests shall be performed in accordance with the procedures delineated in Chapter 8, Leak Testing, of the Manual for the Balancing and Adjustment of Air Distribution System published by SMACNA (First Ed., dated 1967).Paragraph C.3.b (Clarification): The Technical Support Center filter has been purchased to ANSI N509-1980, while the containment filters have been purchased to ANSI N509-1976. Filter media will be subjected to velocities recommended by the HEPA filter manufacturer which exceeds ANSI N509-1976 or 1980 requirements, as applicable, given in Section 4.3.1.The HEPA filter cell testing is conducted initially at the manufacturer's facilities and again after installation at the plant site. All HEPA filters furnished are equipped with the face guards in accordance with MIL-F-51068. When installed in the filter housing, the HEPA filters and housing are inspected for defects and tested for leak-tightness in accordance with ANSI N510-1980.Paragraph C.3.c: For HEPA filters and adsorber mountings, the requirements of ANSI N509-1976 or 1980, as applicable, Section 5.6.3, will be complied with except for the tolerance requirements. The tolerances for HEPA filters and adsorber mounting frames is sufficient to pass the bank leak tests of Paragraphs 5.c and 5.d of the Guide.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-69Rev. 30Paragraph C.3.e: Exception is taken to the recommendations of Section 4.5.8 of ERDA 76-21 relative to drain sizes and arrangement. All drains are capped, and no permanent deluge sy stem is provided.Millstone 3 complies with ANSI N509-1980 paragraph 4.6.2.2 with respect to designing inlet units and components which can be isolated from the fan to withstand a peak negative pressure by ensuring that such isolation is precluded via the design control logic between the fans and the inlet dampers. Compliance with designing inlet units and components, as noted in the same paragraph with respect to the plugging of such components, is de monstrated via routine surveillance and subsequent filter replacement as necessary.Paragraph C.3.f: Exception is taken to Section 5.10.3.5 of ANSI N509-1976 or 1980, as applicable; ductwork, as a structure, will have a resonant frequency above 25 Hz, but this may not be true for the unsupported plate or sheet sections. Ductwork testing shall be subject to the limitations of Paragraph C.2.f above.Paragraph C.3.g: The charcoal for the Technical Support Center filters is purchased in accordance with the requirements of Regulatory Guide 1.52, Paragraph C.3.i.Paragraph C.3.1 Exception is taken to the provisions in Section 5.9 of ANSI N509-1976 of designing dampers to ANSI B31.1 and to using butterfly valves.The non-ESF filtration systems are the containment air filtration system and the Technical Support Center filtration system. The first system is an internal recirculation system consisting of two 50% capacity air cleaning trains. Damper leakage will not impact on the air cleaning effectiveness of this system.The second system filters outside supply air before introducing it into the Technical Support Center. The intake damper will be open during system operation, and the discharge damper will leak filtered air into the room, which will not impact on air cleaning effectiveness of the system.Paragraph C.3.m TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-70Rev. 30The Technical Support Center filter is designed, constructed, and tested to Section 5.3 of ANSI N509-1980.Paragraph C.4.a Exception is taken to full compliance with Section 2.3.8 of ERDA 76-21; i.e., no communication system is used, decontamination areas and showers are not "nearby," filters are not used at duct inlets, and duct inspect ion hatches are not provided. Additionally, the Technical Support Center filter complies with Section 4.7 of ANSI N509-1980 rather than 1976.Paragraph C.4.b Partial compliance, with a minimum spacing between filter frame of 2 ft.-6 in. instead of a minimum of 3 feet. This is deemed adequate since replacement of containment filter elements would be minimal due to system function, use, and location.The Technical Support Center filter does not comply with this requirement since all components of the filter are easily accessible from outside the filter.Paragraph C.4.c The Technical Support Center complies with Section 4.11 of ANSI N509-1980.Paragraph C.5.a Visual inspection is performed in accordance with ANSI N510-1980.
Paragraph C.5.b Test for air flow distribution to the HEPA filters and the adsorbers shall be conducted in accordance with Section 8 of ANSI N510-1980 with the following exceptions and clarifications.Paragraph 8.3.1.6 - Test shall be conducted to verify the following pressure drop values across the combined HEPA filter and charcoal adsorber banks and across the entire filter housing.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-71Rev. 30 System P Across HEPA & Charcoal (in. wg)P Across Housing (in. wg)Containment Air Filters5.947.54Technical Support Center Air Filters5.256.45Paragraph 8.3.1.7 - Exception is taken to the requirement for testing at 50% of the pressure drop valve. Testing the filter at this condition will require the removal of certain filter components with potential damage due to mishandling. Above condition is not considered a design parameter of the system.Paragraph 8.3.2.3 - Air distribution test across the prefilter and moisture separator is not required by the project specification.Paragraph C.5.c HEPA filter DOP testing is conducted in accordance with ANSI N510-1980.Paragraph C.5.d Charcoal adsorber leak testing with refrigerant is conducted in accordance with ANSI N510-1980.1.141Containment Isolation Provisions for Fluid Systems (Rev. 0, April 1978)(1) Comply, with the following exception:
5.4.5 On the containment recirculation syst em, the only closed system outside containment, vent/drain valves and branch connections are not lock closed.
5.4.6 6.2.41.142Safety Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and
Containments)
(Rev. 0, April 1978)
Not applicable. The fo llowing Regulatory Guides have been incorporated into R.G. 1.142, Rev. 0, and have been withdrawn:*1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category 1 Concrete Structures.*1.15 Testing of Reinforcing Bars for Category 1 Concrete Structures.*1.55 Concrete Placement in Category 1 Structures.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-72Rev. 30*1.103 Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments.1.143Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants (Rev. 1, October 1979)1.Prior to 12/14/05, Millstone 3 complied, with the following clarifications and exceptions:3.1.2.63 11.2 Section C.1.1.1 - All demineralizers have been designed, procured, and manufactured in accordance with the ASME Sectio n VIII Boiler and Pressure Vessel Code However, overpressurization protection was not provided in
accordance with ASME VIII for 3LWS-DEMN1 and 3LWS-DEMN2. These demineralizers have been evaluated and determined that over pressurization is not considered a credible failure.11.3 11.4Section C.1.1.2 - Pipe and pipe fittings (valves are excluded) can be procured to an ASTM specification with the basis of acceptance of each item as indicated below. The item shall be manufactured without welding.1.Each item is marked with the applicable ASTM material specification by the original manufacturer. No material manufacturer's Certificate of Conformance (C of C), as defined in ANSI N45.2.10-1973, is required. Only markings from the original manufacturers are acceptable.2.Unmarked items with a Certificate of Conformance to the ASTM material specification are acceptable if shipped directly from the original manufacturer.3.Unmarked items received with no certification are non-destructively tested to ascertain the material of the item. Additional testing, if re quired, is specified by the project engineer.Section C.1.2.1 - Tank high level alarms are provided either locally or in the control room. Tank overflow is precluded by administrative controls where
alarms are not provided in the control room, filling is monitored at local control which has high level alarms.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-73Rev. 30Section C.1.2.5 - Rather than dikes or retention ponds to retain overflows, certain outside tanks are designed to have overflows piped directly to the waste disposal building sump and ultimately to the liquid radioactive waste system.Sections C.2.1.3 and C.5.1 - Portions of the gaseous waste system which store gaseous radioactive wastes have not been designed to seismic criteria. Rather, the building housing in the gaseous waste system is seismically designed in accordance with procedure set forth in FSAR Section 3.7B and releases from the buildings can be filtered by charcoal filters on a high radiation signal.Section C.4.3 - instead of butt-welded joints, belled end so cket-welded fittings are used for some of the radioactive waste management system piping of nominal size 2.5 to 4 inches. Installation of this piping was limited to liquid waste management systems (LWS, LWC, BRS) piping located outside of the Waste Disposal Building.Section C.5.2 - The seismic analyses of the buildings housing the radwaste systems use the Millstone 3 method of seismic analysis of Ca tegory 1 buildings (FSAR Section 3.7B) and do not use the data in R.G. 1.60 and R.G. 1.61 (see compliance for these Guides in this section).2.Current compliance is as described in the QAPD Topical Report17.11.144Auditing of Quality Assurance Programs for Nuclear Power Plants (Rev. 1, September 1980) 1.(1) Prior to 12/14/05, Millsto ne 3 complied as follows:
Construction Millstone 3 complies with ANSI N45.2.12-1977, as implemen ted by Appendix VII of the Millstone 3 Quality Assurance Program Manual.
Operation - Comply2.3 2.Current compliance is as described in the QAPD Topical Report.17.11.145Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants (Rev. 0, August 1979)
Comply13.3TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8-74Rev. 301.146Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Rev. 0, August 1980)1. (1) Prior to 12/14/05, Millstone 3 complied.17.12. Current compliance is as described in the QAPD Topical Report.1.147Inservice Inspection Code Case Acceptability ASME Section XI Division 1 (Rev. 0, February 1981)Comply1.148Functional Specification for Active Valve Assemblies in Systems Important to Safety in Nuclear Power Plants (Rev. 0, March 1981)The majority of active valves used at Millstone 3 were purchased prior to the issuance of Regulatory Guide 1.148. However, Millstone 3 meets the intent of Regulatory Guide 1.148 in that all active valves relied upon to perform a safety function are designed and analyzed to ensure their structural integrity and operability during the transients or events considered in the respective operating condition categories.The overall design process includes systems design, valve specifications, and quality assurance procedures. Many of the functional requirements are included in the systems design which dictates the type of valve required for the system application.Additionally, the NRC conducted a Pump and Valve Operability Review Team Audit in March 1985 following which the NRC concluded that the Millstone 3 valve operability program met the intent of Regulatory Guide 1.148.1.149Nuclear Power Plant Simu lators for Use in Operator Training (Rev. 3, October 2001)Millstone 3 complies with R.G. 1.149, Rev.3.13.21.150Ultrasonic Test of Reactor Vessel Welds During Preservice and Inservice Examinations (Rev. 1, February 1983)Comply, per response to NRC Question Q250.3.5.2.41.152Criteria for Digital Computers in Safety Systems of Nuclear Power Plants (Rev. 1, January 1996)Compliance is as described in the QAPD Topical Report.1.71.155Station Blackout Comply8.3.1.1.51.163Performance-Based Containment Leak-Test Program ComplyTABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance FSAR Section Reference MPS3 UFSAR1.8-75Rev. 301.183Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Rev. 0, July 2000)Complies with the fo llowing exceptions:Section 3.2, Table 3, footnote 11 - Non-LOCA Fraction of Fission Product Inventory in Gap - for the fuel handling accident, the fraction of the fuel rods exceeding the criteria of footnote 11 that would bound projected loading plans and operating strategies were modeled with the gap fractions listed in Regulatory Guide 1.25 (as modified by the direction of NUREG/CR-5009) instead of values from Table 3.App. A, Section 3.7- for control room dose analysis, the Technical Specification Containment leak rate is reduced by 50% after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> versus 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> listed in the Reg. Guide. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is used for off site dose analysis.Appendix E and Appendix F, Sections 5.1 - Transport - for the main steam line break, steam generator tube rupture, locked rotor accident, and rod ejection accident Regulatory Guide 1.183 requires primary-to-secondary leakage to be modeled as the leak rate limiting condition for operation specified in the technical specifications. The Regulatory Guide 1.183 requirements predates the Steam Generator Program, which currently limits primary-to-secondary operational leakage to RCS LCO 3.4.6.2 and requires that accidents be modeled with that accident induced leakage (Technical Specification 6.8.g.b.2). The accident induced leakage is consistent with the leak rate limiting condition for operation specified in the technical specifications prior the advent of the Steam Generator Program.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance MPS3 UFSAR1.8-76Rev. 30NOTE: 1.See FSAR Section 1.8N For Nuclear Steam Supply System (NSSS) scope of compliance for Millstone 3.2.Later revisions NA - Millstone 3 is exempted by the implementation se ctions of the later revisions of the Guide.1.194Atmospheric Relative Co ncentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants (Rev. 0, June 2003)
Comply1.196Control Room Habitability at Light-Water Nuclear Power Reactors (Rev. 0, May 2003)Comply, with the following exceptions and clarifications:1.RG 1.196 calls for evaluating, tes ting and maintaining the Control Room Emergency Filtration System per RG 1.52 Rev. 3. Millstone 3 complies wit h1.52 Rev. 2, with the clarifications and exceptions as noted in this table.2.RG 1.196 calls for performing an evaluation, and also periodic assessment sthe impact of hazardous chemical release on control room habitability usin gmethodology of RG 1.78 Rev. 1 (which replaced the previous revisions of R1.78 for hazardous gas release and RG 1.95 for chlorine release). Millston ecomplies with RG 1.78 Rev. 0 and RG 1.95 Rev. 1, with the clarifications aexceptions as noted in this table.3.RG 1.196 calls for determining control room radiological dose per RG 1.1 8Rev. 0 for plants employing alternative source term methodology. Millston ecomplies with RG1.183 Rev. 0, with the exceptions as noted in this table.1.197Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors (Rev. 0, May 2003)Comply, with the following exceptions and clarifications:1.With respect to Position C.1 for performance of an integrated Control Roo mEnvelope (CRE) inleakage test:*Appropriate application of ASTM E741 shall include the ability to tak eminor exceptions to the test methodology. These exceptions shall be documented in the test report.2.With respect to Position C.2 for establishing the licensing basis of the CR Edeveloping compensatory actions in the event of excessive CRE inleakage:*Vulnerability assessments for radiol ogical, hazardous chemical and smand emergency ventilation system testing were completed as document ethe UFSAR and other licensing basis documents. The exceptions to th e Regulatory Guides referenced in RG 1.196 (i.e., RG 1.52, RG 1.78 an d1.183), which were considered in completing the vulnerability assess mare documented in the UFSAR current licensing basis.TABLE 1.8-1 NRC REGULATORY GU IDES (CONTINUED)R.G. No.TitleDegree of Compliance MPS3 UFSAR1.8-77Rev. 303.Millstone 3 addres ses a later revision of the Guide th an is required by the implementation section of the Guide.
MPS3 UFSAR1.8N-1Rev. 30 1.8N NSSS CONFORMANCE TO NRC REGULATORY GUIDESTable 1.8N-1 lists the NRC Division 1 Regulatory Guides that were in effect during the time of application for an Operating License. It identifies applicable FSAR sections, and indicates the NSSS scope of compliance for Millstone 3.
MPS3 UFSARMPS3 UFSAR1.8N-2Rev. 30TABLE 1.8N-1 NRC REGULATORY GUIDES R.G. No.TitleDegree of ComplianceFSAR Section Reference 1.1Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps (Rev. 0, November 2, 1970In predicting the NPSH available, Westinghouse assumes that the vapor pressure of the liquid in the sump is equal to the containment pressure, i.e., that the liquid is at a saturated condition, in calculating the available NPSH for the recirculation mode. This assumption meets the intent of Regulatory Guide 1.1, since no credit is taken for any increase in containment pressure. It is also a reasonable assumption since the containment is a closed thermodynamic system and will remain at an equilibrium condition.6.2.2, 6.31.2Thermal Shock to Reactor Pressure Vessels (Rev. 0, November 1970)
ISSUE: General Design Criterion 35 specifies design and operating conditions necessary to assure that the reactor coolant pressure boundary w ill behave in a nonbrittle manner. To provide protection against loss-of-coolant accidents, present designs provide for the injection of large quantities of cold emergency coolant into the reactor coolant system. The effect on the reactor pressure vessel of this cold water injection is of concern because the reactor vessel is subjected to greater irradiation than other components of the reactor coolant pressure boundary and, thus, has a greater potential for becoming brittle. Regulatory Guide 1.2 describes a suitable program which may be used to implement General Design Criterion 35 to assure that the reactor pressure vessel will behave in a nonbrittle manner under loss-of-coolant accident conditions.5.2.3.3 NUCLEAR SAFETY POSITION
- Westinghouse follows all recommendations of the guide. The guide Position C.1 is followed by Westinghouse's own analytical and experimental programs, as well as by participation in the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory.Under the Heavy Section Steel Technology Program, a number of steel pressure vessels containing carefully prepared and sharpened surface cracks have been tested. Test specimens have been subjected to either hydraulic internal pressure loadings or thermal shock loadings. This program has validated analytical fracture mechanics techniques and has demonstrated quantitatively the margin of safety inhere nt in reactor pressure vessels.
Additional testing will be done to demonstrate the behavior of vessels under combined pressure and thermal shock loadings.
MPS3 UFSARMPS3 UFSAR1.8N-3Rev. 30Details of progress and results obtained in the HSST progra m are available in the Heavy Section Steel Technology Program Semiannual (quarterly, beginning in 1974) Progress Reports issued by Oak Ridge National Laboratory.Westinghouse is continuing to obtain fracture toughness data for reactor pressure vessel steels through internally funded programs, as well as industry sponsored work.Regulatory Position C.2 is followed inasmuch as no adverse changes have been made in approved core or reactor designs. A change to low leakage loading pattern core designs has been made on some plants. This change reduces the rate irradiation of the pressure vessel and is, therefore, beneficial.Regulatory Position C.3 is followed since the vessel design does not preclude the use of an engineering solution to assure adequate recovery of the fracture toughness properties of the vessel material. If additional margin is needed, the reactor vessel can be annealed. This solution was shown to be feasible by EPRI program RP1021-1, "Feasibility and Methodology for Thermal Annealing an Embrittled Reactor Vessel."Westinghouse is continuing to develop analytical capability for reactor vessel integrity evaluations. A conservative generic evaluation was performed in 1981 under the Westinghouse Owners Group for all domestic operating plants. This evaluation, summarized in WCAP-10019, "Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants," confirms that no near-term safety concern exists. Acceptability through end-of-life was not demonstrated by this report for all plants, and additional plant specific evaluations are being performed to remove some of the excessive conservatisms inherent in the generic report. Additional development work is also being performed in response to NRC questions arising from their review of this issue under Task Action Plan A-49.1.7Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident (Rev. 2, November 1978)Westinghouse will conform to all assumptions established in Regulatory Guide 1.7 and thereby comply with General Design Criterion 41.6.2.5 1.13Spent Fuel Storage Facility Design Basis (Rev. 1, December 1975)Comply.9.1.2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-4Rev. 301.14Reactor Coolant Pump Flywheel Integrity (Rev. 1, August 1975)Comply, with the following exceptions: 5.4.1.1 1.Post-Spin Inspection Westinghouse has shown in WCAP-8163, "Topical Report Reactor Coolant Pump Integrity in LOCA," that the flywheel would not fail at 290 percent of normal speed for a flywheel flaw of 1.15 inches or less in length. Results for a double ended guillotine break at the pump discharge with full separation of pipe ends assumed, show the maximum overspeed to be less than 110 percent of normal speed. Even with an assumed instantaneous loss of power to the reactor coolant pump, the maximum overspeed was calculated in WCAP-8163 to be about 280 percent of normal speed for the same postulated break. In comparison with the overspeed presented above, the flywheel is tested at 125 percent of normal speed. Thus, the flywheel could withstand a speed up to 2.3 times greater than the flywheel spin test speed of 125 percent provided that no flaws greater than 1.15 inches are present. If the maximum speed were 125 percent of normal speed or less, the critical flaw size for failure would exceed 6 inches in length. Nondestructive tests and critical dimension examinations are all performed before the spin tests. The inspection methods employed (described in WCAP-8163) provide assurance that flaws significantly smaller than the critical flaw size of 1.15 inches for 290 percent of normal speed would be detected. Flaws in the flywheel will be recorded in the prespin inspection program (see WCAP- 8163). Flaw growth attributable to the SPIN test (i.e., from a single reversal of stress, up to speed and back), under the most adverse conditions, is about three orders of magnitude smaller than what nondestructive inspection techniques are capable of detecting. For these reasons, Westinghouse performs no post-spin inspection and believes that prespin test inspections are adequate.2.Interference Fit Stresses and Excessive Deformation TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-5Rev. 30 Much of Revision 1 deals with stresses in the flywheel resulting from the interference fit between the flywheel and the shaft. Because Westinghouse's design specifies a light interference fit between the flywheel and the shaft, at zero speed, the hoop stresses and radial stresses at the flywheel bore are negligible. Centering of the flywheel relative to the shaft is accomplished by means of keys and/or centering devices attached to the shaft, and at normal speed, the flywheel is not in contact with the shaft in the sense intended by Revision 1. Hence, the definition of "Excessive Deformation," as defined in Revision 1 of Regulatory Guide 1.14, is not applicable to the Westinghouse design since the enlargement of the bore and subsequent partial separation of the flywheel from the shaft does not cause unbalance of the flywheel. Extensive Westinghouse experience with reactor coolant pump flywheels installed in this fashion has verified the adequacy of the design.Westinghouse's position is that co mbined primary stress levels, as defined in Revision 0 of Safety Guide 14 (C.2 (a) and (c)) are both conservative and proven and that no changes to these stress limits are necessary. Westinghouse designs to these stress limits and thus, does not have permanent distortion of the flywheel bore at normal or spin test conditions.3.Section B, Discussion of Cross Rolling Ratio of 1 to 3 Cross Rolling Ratio - Westinghouse's position is that specification of a cross rolling ratio is unnecessary since past evaluations have shown that ASME SA-533-B Class 1 materials produced without this requirement have suitable toughness for typical flywheel applications. Proper material selection and specification of minimum material properties in the transverse direction adequately ensure flywheel integrity. An attempt to gain isot ropy in the flywheel material by means of cross rolling is unnecessary since adequate margins of safety are provided by both flywheel material selection (ASME SA- 533-B Class 1) and by specifying minimum yield and tensile levels and toughness test values taken in the direction perpendicular to the maximum working direction of the material.4.Section C, Item 1a Relative to Vacuum-Melting and Degassing Process or the Electroslag Process TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-6Rev. 30Vacuum Treatment - The requirements for vacuum melting and dega ssing process or the electroslag process are not essential in meeting the balance of the Regulatory Position nor do they, in themselves, ensure compliance with the overall Regulatory Position. The initial Safety Guide 14 stated that the "flywheel material should be produced by a process that minimized flaws in the material and improves its fracture toughness properties." This is accomplished by using SA-533 material including vacuum treatment.5.Section C, Item 2b Westinghouse interprets this paragraph as follows:
Design Speed Definition Design speed should be 125 percent of normal speed or the speed to which the pump motor might be electrically driven by station turbin e generator during an ticipated transients, whichever is greater. Normal speed is defined as the synchronous speed of the AC drive motor at 60 Hz.6.Section C, Item 4b, Inservice Inspections (1) and (2)
Instead of the flywheel inspections required at approximately 3-year and 10-year intervals, an inspection will be performed by either a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (magnetic particle testing and/or penetrant testing) of exposed surfaces defined by the volume of the disassembled flywheels at least once every 10 years.1.20Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing (Rev. 2, May 1976)For each prototype reactor internals design, a program of vibration analysis, measurement, and inspection has been developed and reviewed by the NRC. This is documented in WCAP-7879.3.9.2 The reactor internals similar to the prototype design will be subjected during hot functional testing to the same system flow conditions imposed on the prototype design applicable, and for the same duration. Pre- and post test inspections will be conducted to assure that the internals are well behaved and that no excessive motion or wear are experienced.1.22Periodic Testing of Protection System Actuation Functions (Rev. 0, February 17, 1972)Periodic testing of the actuation equipment and actuated equipment of the reactor trip system and the engineered safety features actuation system is in agreement with the provisions of Regulatory Guide 1.22.7.2 7.3.2.2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-7Rev. 30Where the ability of a system to respond to a bona fide acci dent signal is intentionally bypassed for the purpose of performing a test during reactor operation, there are provisions so that the bypass condition may be automatically indicated to the reactor operator in the main control room by a separate indication for the train in test. Test circuitry does not allow two trains to be tested at the same time so that extension of the bypass condition to the redundant system is prevented.Actuation logic for the reactor trip system and for the engineered safety features actuation system is tested at power. Where actuated equipment is not tested during reactor operation, it has been determined that:1.There is no practical system design that would permit operation of the equipment without adversely affecting the safety or operability of the plant2.The probability that the protection system will fail to initiate the operation of the equipment is, and can be maintained, acceptably low without testing the equipment during reactor operation; and3.The equipment can routinely be tested when the reactor is shut down.The list of equipment that cannot be tested at full power so as not to damage equipment or upset plant operation is:1.Manual actuation switches2.Turbine3.Main steam line isolation valves (close)4.Main feedwater isolation valves (close)5.Feedwater control valves (close)6.Main feedwater pump trip solenoids 7.Reactor coolant pump seal water return valves (close)8.Charging header to cold leg isolation valves (close)9.Charging and letdown is olation valves (close)10.Spray header isolation valves (open)11.CVCS suction valves - Normal (close)TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-8Rev. 3012.Instrument air to containment isolation valves (close)13.Chillwater supply and return containment isolation valves (close)1.26Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants (Rev. 3, February 1976)The definitions of safety classes found in ANSI N18.2a are utilized by Westinghouse.3.2.2 1.28Quality Assurance Program Requirements (Design and Construction)
(Rev. 2, February 1979)The Westinghouse Quality Assurance Plan is presented in WCAP-8370, Rev. 9A, "Westinghouse Quality Assurance Program" and WCAP-7800, "Nuclear Fuel Division Quality Assurance Program Plan."1.29Seismic Design Classification (Rev. 3, September 1978)Westinghouse classifies each component important to safety as Safety Class 1, 2, or 3 and these classes are qualified to remain functional in the event of the Safe Shutdown Earthquake, except where exempted by meeting all of the below requirements. Portions of systems required to perform the same safety function as required of a safety class component which is part of that system shall be likewise qualified or granted exemption. Conditions to be met for exemption are:3.2.1 1.Failure would not directly cause a Condition III or IV event (as defined in ANSI N18.2-1973),2.There is no safety function to mitigate, nor could failure prevent mitigation of, the consequence of a Condition III or IV event,3.Failure during or following any Condition II event would result in consequences no more severe than allowed for a Condition III event, and4.Routine post seismic procedures would disclose loss of the safety function.Westinghouse agrees with Position C2 that establishes a second category of earthquake-resistant equipment but, primarily this affects proper methods of installation and anchoring of certain equipment such that non-Seismic Category I components do not cause loss of function of Seismic Category I components in an earthquake.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-9Rev. 301.31Control of Ferrite Content in Stainless Steel Welding (Rev. 3, April 1978)
The welding of austenitic stainless steel is controlled to mitigate the occurrence of microfissuring or hot cracking in the weld. Although published data and experience have not confirmed that fissuring is detrimental to the quality of the weld, it is recognized that such fissuring is undesirable in a general sense. Also, it has been well documented in the technical literature that the prese nce of delta ferrite is one of the mechanisms for reducing the susceptibility of stainless steel welds to hot cracking. However, there are insufficient data to specify a minimum delta ferrite level below which the material will be prone to hot cracking. It is assumed that such a minimum lies somewhere between 0 and 3 percent delta ferrite.4.5.1 4.5.2.4 5.2.3.4 5.3.1.4 10.3.6.2 The scope of these controls discussed herein encompasses welding processes used to join stainless steel parts in components designed, fabricated, or stamped in accordance with ASME B&PV Code,Section III, Class 1, 2, 3, and CS componen ts. Delta ferrite control is appropriate for the above welding requirements, except in the following cases: where no filler metal is used (for example, in electron beam welding and in autogenous gas shielded tungsten arc welding), where stainless steel filler metal is used for weld metal cladding, explosive welding, and welding using fully austenitic welding materials.The fabrication and installation specifications require welding procedure and welder qualification in accordance with Section III, and include th e delta ferrite determinations for the austenitic stainless steel welding materials that are used for welding qualification testing and for production processing. Specifically, the undiluted weld deposits of the "starting" welding materials are required to contain a minimum of 5 percent delta ferrite (or the equivalent Ferrite Number) as determined by chemical analysis and calculation using the appropriate weld metal constitution diagrams in Section III. When new welding procedure qualification tests are evaluated for these applications, including repair welding or raw materials, they are performed in accordance with Section III and Section IX. The results of all the destructive and nondestructive tests are reported in the procedure qualification record in addition to the informa tion required by Section III.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-10Rev. 30The "starting" welding materials used for fabrication and installation welds of austenitic stainless steel materials and components meet the requirements of Section III. The austenitic stainless steel welding material conforms to ASME weld metal anal ysis A-7 (designated A-8 in the 1974 Edition of the ASME code), type 308 or 308L for all applications. Bare weld filler metal, including consumable inserts, used in inert gas welding processes conform to ASME SFA-5.9, and are procured to contain not less than 5 percent delta ferrite in the deposit according to Section III. Weld filler metal materials used in flux shielded welding processes conform to ASME SFA-5.4 or SFA-5.9 and are procured in a wire-flux combination to be capable of providing not less than 5 percent delta ferrite in the deposit according to Section III. Welding materials are tested using the welding energy inputs to be employed in production welding.Combinations of approved heats and lots of "starting" welding materials are used for all welding processes. The welding quality assurance program includes identification and control of welding material by lots and heats, as appropriate. All of the weld processing is monitored according to approved inspection programs which include review of "starting" materials, qualification records, and welding parameters. Welding systems are also subject to quality assurance audits, in cluding calibration of gages and instruments, identification of "starting" and completed materials, welder and procedure qualifications, availability and use of approved welding and heat treating procedures, and documentary evidence of compliance with materials, welding parameters, and inspection requirements. Fabrication and installation welds are inspected using nondest ructive examination methods according to Section III rules.To assure the reliability of these controls, Westinghouse has perfo rmed a delta ferrite verification program, described in WCAP-8324, "Control of Delta Ferrite in Austenitic Stainless Steel Weldments," June 1974. The verification program has been approved as a valid approach to verify the Westinghouse hypothesis and is considered an acceptable alternative for conformance with the Interim Position on Regulato ry Guide 1.31. The Regulatory Staff's acceptance letter and topical report evaluation were received on December 30, 1974. The program results, which support the hypothesis presented in WCAP-8324, are summarized in WCAP-8693, "Delta Ferrite in Production Austenitic Stainless Steel Weldments," January 1976.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-11Rev. 30Welds made in accordance with the criteria discussed herein have continually resulted in sound production welds, which are free from detrimental fissuring and consistently conform to Section III nondestruc tive acceptance standards.1.36Nonmetallic Thermal Insulation for Austenitic Stainless Steel (Rev. 0, February 1973)The Westinghouse practice follows the recommendations of Regulatory Guide 1.36 but is more stringent in several respects as discussed below.5.3.2 The nonmetallic thermal insulation used on the reactor coolant pressure boundary is specified to be made of compounded materials which yield low leachable chloride and/or fluoride concentrations. The compounded materials, in the form of blocks, boards, cloths, tapes, adhesives, cements, etc, are silicated to provide protection of austenitic stainless steels against stress corrosion which may result from accidental wetting of the insulation by spillage, minor leakage, or other contamination from the environmental atmosphere. Each lot of insulation materials is qualified and analyzed to assure that all of the materials provide a compatible combination for the reactor coolant pressure boundary.6.1.1 The tests for qualification specified by the guide (ASTM C692-71 or RDT M12-1T) allow use of the tested insulation material if no more than one of the metallic test samples crack. Westinghouse rejects the tested insulation material if any of the test samples crack.The Westinghouse procedure is more specific than the procedures suggested by the guide, in that the Westinghouse specification requires determination of leachable chloride and fluoride ions from a sample of the insulating material. The procedures in the guide (ASTM D512 and ASTM D1179) do not differentiate between leachable and unleachable halogen ions.In addition, Westinghouse experience indicates that only one of the three methods allowed under ASTM D512 and ASTM D1179 for chloride and fluoride analysis is sufficiently accurate for reactor applications. This is the "referee" method, which is used by Westinghouse.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-12Rev. 301.37Quality Assuranc e Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (Rev. 0, March 1973)The Westinghouse position for all Water Reactors Division (WRD) on this Regulatory Guide is documented in WCAP-8370.4.5.1 17.1.2 17.1.3 17.21.38Quality Assuranc e Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Rev. 2, May 1977) The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.38 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Nuclear Fuel Division (NFD) position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."1.40Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (Rev. 0, March 1973)There are no Class 1 continuous duty motors installed inside containment in the NSSS scope of supply.1.43Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components (Rev. 0, May 1973)Westinghouse performs a qualification test on any high heat input welding process (such as the submerged arc wide-strip welding process or the submerged arc 6-wire process) used to clad coarse or fine grained SA-508 Class 2 material. This test follows the recommendations of Regulatory Position C.2 of Regulatory Guide 1.43. Production welding is monitored by the fabricator to ensure that essential variables remain within the limits established by the qualification. If the essential variables exceed the qualification limits, an evaluation will be performed to determined if the cladding is acceptable for use.5.3.1.4 Where Westinghouse permits the use of the submerged arc strip process on SA-508 Class 2 material, a two-layer technique is used to minimize intergranular cracking.1.44Control of the Use of Sensitized Stainless Steel (Rev. 0, May 1973)It has been and continues to be Westinghouse practice to use processing, preoperational cleaning, packaging, and shipping controls to preclude adverse effects of exposure to contaminants on austenitic stainless steel materials. Furthermore, Westinghouse strongly 6.1.1.1 recommends rigorous control of reactor coolant system water chemistry to prevent the intrusion of aggressive species.4.5.1 4.5.2.4 5.2.3.4 5.3.1.4 6.1.1.1 10.3.6.2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-13Rev. 30Austenitic stainless steel materials are utilized in the final heat treated condition required by the respective ASME Section II material specification for the particular type of grade or alloy. More specifically, the austenitic stainless steel materials are utilized in one of the following conditions:1.Solution annealed or water quenched, or2.Solution annealed and cooled by other means through the sensitization temperature range within approximately five minutes.It is generally accepted that these practices will prevent sensitization; Westinghouse has verified this by performing corrosion tests (ASTM 393) on as-received wrought material.The Westinghouse practice is that austenitic stainless steel materials of product forms with simple shapes need not be corrosion tested provided that the solution heat treatment is followed by water quenching. Simple shapes are defined as all plates, sheets, bars, pipe, and tubes, as well as forgings, fittings, and other shaped products which do not have inaccessible cavities or chambers that would preclude rapid cooling when water quenched. Stainless steel cast metal and weld deposits (including weld deposited safe ends), which contain a minimum of 5 percent ferrite, are not considered to be susceptible to sensitization and, therefore, are not corrosion tested. When testing is required, the tests are performed in accordance with ASTM A 262-70, Practice A or E, as amended by Westinghouse Process Specification 84201 MW. This process specification supplements the A262 specification since the latter does not define specimen removal location and does not adequately define bend testing criteria for thick and complex stainless steel raw material.The Westinghouse specification requires that:1.Specimens be removed from the same location from which mechanical test specimens are removed, and2.The bend test diameter must be 4X material thickness instead of 1X (Paragraph 36.1, ASTM A 262-70).This second modification is based on the fact that almost all stainless steel materials procured by Westinghouse are eventually welded, and the 4X thickness bend test diameter is required for weldments.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-14Rev. 30The heat affected zones of welded components must, of necessity, be heated into the sensitization temperature. However, severe sensitization can be avoided by control of welding parameters and welding processes. Westinghouse controls the heat input in all austenitic pressure boundary weldments by:1.Prohibiting the use of block welding.2. Limiting the maximum temperature to 350
°F.3.Exercising approval rights on all welding procedures.Westinghouse has demonstrated the importance of heat input and cooling rate by corrosion testing a number of production and qualification weldments. The tested weldments represented all major welding processes and included a variety of components and base metal thicknesses from 0.10 to 4.0 inches. Portions of only 2 out of 25 weldments exhibited sensitization; in both cases, sensitization was caused primarily by high heat inputs relative to the section thickness. If it becomes necessary to further assure that the present controls are effective in preventing sensitization, Westinghouse will conduct additional corrosion tests on qualification weldments.It is not normal Westinghouse practice to expose wrought unstabilized austenitic stainless steel materials to the sensitization range of 800 to 1,500
°F during fabrication other than welding. If, during the course of fabrication, the steel is inadvertently exposed to the sensitization temperature range, the material may be tested (as described in Regulatory Position 5 of the Guide) in accordance with ASTM A 262-70, as amended by Westinghouse Process Specification 84201 MW, or the material will be resolution annealed and water quenched or rejected.1.46Protection Against Pipe Whip Inside Containment (Rev. 0, May 1973)The criteria implemented in the evaluation of the main reactor coolant loop is based on draft ANS Standard 20.2, "Design Basis for Protection Against Pipe Whip," and is documented in WCAP-8172-A, "Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop."3.6 WCAP-8172-A has received NRC approval as providing an equivalent degree of protection as would be obtained by applying the criteria of Regulatory Guide 1.46.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-15Rev. 301.48Design Limits and Loading Combination for Seismic Category 1 Fluid System Components (Rev. 0, May 1973)Westinghouse meets and will continue to meet the requirements of General Design Criterion 2 and will thereby meet the intent of Regulatory Guide 1.48. The analytical and experimental procedures which Westinghouse will use to demonstrate structural integrity of fluid system components and operability of active components are discussed below. These procedures are intended to provide an alternate acceptable basis for the demonstration of compliance with General Design Criterion 2.3.9.3 Structural Integrity To ensure the structural integrity of fl uid systems components, the limits given in RESAR-3, Amendment 5, Section 5.2.1 will be used in the design and analysis of ASME Code Class 1 components.For ASME Code Class 2 and 3 components, the limits given in RESAR-3, Amendment 6, Section 3.9.2 will be used.The conservatism in the above limits and the associated ASME design requirements precludes any component structural failure.For a discussion on operability of active pumps and valves, see Section 3.9.1.50Control of Preheat Temperature for Welding of Low-Alloy Steel (Rev. 0, May 1973)Westinghouse considers that this Guide applies to ASME Section III, Class 1 components.5.2.3.3 The Westinghouse practice for Class 1 components is in agreement with the requirements of Regulatory Guide 1.50, except for Regulatory Positions 1(b) and 2. For Class 2 and 3 components, Westinghouse does not apply Regulatory Guide 1.50 recommendations.1.Regulatory Position 1(b)
The welding procedures are qualified within the preheat temperature ranges required by Section IX of the ASME Code. Westinghouse experience has shown excellent quality of welds using the ASME qualification procedures.2.Regulatory Position 2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-16Rev. 30The Westinghouse position is that this guide requireme nt is both unnecessary and impractical. Code acceptance low-alloy steel welds have been and are being made under present Westinghouse specified procedures. It is not necessary to maintain the preheat temperature until a post-weld heat treatment has been performed by the guide, in the case of large components. In the case of reactor vessel main structural welds, the practice of maintaining preheat until the intermediate or final post-weld heat treatment has been followed by Westinghouse. In either case, the welds have shown high integrity. Westinghouse practices are documented in WCAP-8577, "The Application of Preheat Temperature After Welding of Pressure Vessel Steel," which has been accepted by the NRC. 1.53Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems (Rev. 0, June 1973)Performance of a Single Failure Analysis - The principles described in the IEEE Standard were used in the Power Plant Protection Systems design of the protection systems on Westinghouse plants. For docum entation of applicable analysis, refer to WCAP-8584, Revision 1, which is a FMEA for the ESFAS and to WCAP-7706 for Reactor Trip System. The latter topical is in the format of a fault tree analysis rather than a failure mode and effects analysis (FMEA). Even though Regulatory Guide 1.53 proposes a FMEA as an acceptable format, the fault tree analysis format is considered equally acceptable and more useful for arriving at quantitative results. Subjects which are covered in the standard include:3.1.1 1.Identification of undetectable failures,2.Analyses of channel interconnections for failures which could compromise independence,3.Testing to determine independence between redundant parts of the protection system, and4.Analysis to show that no single failure can cause loss of function due to improper connection of actuators to a power source.The intent of the guide is met in these areas through existing design requirements.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-17Rev. 30Scope of Analysis - The regulatory guide requires that the single failure analysis extend beyond the scope of IEEE-Standard 279-19 71 and include actua tion and actuated equipment. WCAP-8584 does not extend to actuation and actuated equipment, but contains interface criteria in Appendix B that, when it is incorporated in the BOP design, assures that the FMEA results are equally applicable to actuation and actuated equipment.1.54Quality Assuranc e Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants (Rev. 0, June 1973)The Westinghouse NSSS equipment located in the for containment building is separated into four categories to identify the applicability of this Regulatory Guide to various types of equipment. These categories of equipment are as follows:6.1.2 17.1 Category 1 - Large equipment Category 2 - Intermediate equipment Category 3 - Small equipmentCategory 4 - Insulated/stainless steel equipment Category 1 - Large Equipment The Category 1 equipment consists of the following:1.Reactor Coolant System Supports 2.Reactor Coolant Pumps (motor and motor stand)3.Accumulator Tanks4.Manipulator CraneThe total exposed surface area for these items is approximately 20,830 (sq ft) for a four loop plant.Since this equipment occupies a large surface area and is procured from only a few vendors, it is possible to implement tight controls over these items.Westinghouse specifies stringent requirements for protective coatings on this equipment through the use of a painting specification in its procurement documents. This specification defines requirements for:1.Preparation of vendor proceduresTABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-18Rev. 302.Use of specific coatings systems which are qualified to ANSI N101.23.Surface preparation4.Application of the coating systems in accordance with the paint manufacturer's instructions5.Inspections and nondestructive examinations6.Exclusion of certain materials7.Identification of all nonconformances 8.Certifications of complianceThe vendor's procedures are subject to review by WRD Engineering personnel, and the vendor's implementation of the specification requirements is monitored during the Westinghouse QA Surveillance activities.This system of controls provides assurance that the protective coatings will properly adhere to the base metal during prolonged exposure to a post-accident environment present within the containment building. No loss of paint is anticipated.Category 2 - Intermediate Equipment The Category 2 equipment consists of the following:1.Seismic platform and tie rods 2.Reactor internals lifting rig3.Head lifting rig4.Electrical cabinetsThe total exposed surface area of these items is approximately 3,450 sq ft. Since these items are procured from a large number of vendors, and individually occupy very small surface areas, it is not practical to enforce the complete set of stringent requirements which are applied to Category I items. However, Westinghouse does implement another specification in its procurement documents. This specification defines to the vendors the requirements for:1.Use of specific coatings systems which are qualified to ANSI N101.22.Surface preparationTABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-19Rev. 303.Application of the coating systems in accordance with the paint manufacturer's instructionsThe vendor's compliance with the requirements is also checked during the Westinghouse QA Surveillance activities in the vendor's plant. Westinghouse believes that these measures of control provide a high degree of assurance that the pr otective coatings will adhere properly to the base metal and withstand the postulated accident environment within the containment building. However, to be conservative, Westinghouse has not taken credit for this in calculating the amount of paint which might peel or flake off in the post-accident environment.
Category 3 - Small Equipment Category 3 equipment consists of the following:1.Transmitters2.Alarm horns3.Small instruments 4.Valves5.Heat exchanger supportsThese items are procured from several different vendors and are painted by the vendor in accordance with conventional industry practices. Because the total exposed surface area is only 900 sq ft, Westinghouse does not believe it is necessary to specify further requirements. For purposes of estimating the amount of paint that might peel or flake off, Westinghouse has assumed that all of this material might come off.Category 4 - Insulated or Stainless Steel Equipment Category 4 equipment consists of the following:1.Steam generators - covered with wrapped insulation2.Pressurizer - covered with wrapped insulation3.Reactor pressure vessel - covered with rigid reflective insulation4.Reactor coolant piping - stainless steel5.Reactor coolant pump casings - stainless steelTABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-20Rev. 30The wrapped or rigid insulation captures and retains any paint which might come off the equipment surfaces, thereby preventing the paint from blocking the sump drains or interrupting the water flow in the containment spray system.1.58Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel (Rev. 1, September 1980)The Westinghouse portion for the WRD NSSS scope of supply on Regulatory Guide 1.58 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."
17.1.2 17.1.317.21.60Design Response Spectra for Seismic Design of Nuclear Power Plants (Rev. 1, December 1973)The design response spectra of Regulatory Guide 1.60, Revision 1, are acceptable to Westinghouse with the following exception:3.7N The damping values recommended and approved by the Staff in WCAP-7921-AR, "Damping Values of Nuclear Power Plant Components," are used in dynamic analysis of Westinghouse supplied equipment.1.61Damping Values for Seismic Design of Nuclear Power Plants (Rev. 0, October 1973)The damping values listed in Regulatory Guide 1.61 Design of Nuclear Power are acceptable to Westinghouse for plants using 3D seismic analysis. However, one exception is that of the large piping systems faulted conditions value of 3 percent critical. Higher damping values, when justified by documented test data, have been provided for in Regulatory Position C.2. A conservative value of 4 percent critical has therefore been justified by testing for the Westinghouse reactor coolant loop configuration in WCAP-7921, "Damping Values of Nuclear Power Plant Components," and has been approved by the Staff.3.7N 1.62Manual Initiation of Protective Actions (Rev. 0, October 1973)There are four individual main stop valve momentary control switches (one per loop) mounted on the control board. Each switch, when actuated, will isolate one of the main steam lines. In addition, there will be two system level switches. Each switch will actuate all four main steam line isolation and bypass valves of the system level. Manual initiation of switchover to recirculation is in compliance with Section 4.17 of IEEE Standard 279-1971 with the following comment.7.3.2.2.7 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-21Rev. 30Manual initiation of either one of two redundant safety injection actuation main control board mounted switches provides for actuation of the components required for reactor protection and mitigation of adverse consequences of the postulated accident. Manual safety injection actuation will initiate delayed actuation of sequenced started emergency electrical loads if a LOP signal is also present. The safety injection mode is completed when the residual heat removal (RHR) pumps automatically stop on receipt of a low-low RWST level signal. Refer to Section 6.3 for a discussion of the manual switchover from injection mode to cold leg recirculation mode. Manual operation of other components or manual verification of proper position as part of emergency procedures is not precluded nor otherwise in conflict with the above described compliance to paragraph 4.17 of IEEE Standard 279-1971 of the semiautomatic switchover circuits.No exception to the requirements of IEEE Standard 279-1971 has been taken in the manual initiation circuit of safety injection. Although paragraph 4.17 of IEEE Standard 279-1971 requires that a single failure within common portions of the protective system shall not defeat the protective action by manual or automatic means, the standard does not specifically preclude the sharing of initiated circuitry logic between automatic and manual functions. It is true that the manual safety injection initiation functions associated with one actuation train (e.g., train A) shares portions of the automatic initiation circuitry logic of the same logic train; however, a single failure in shared functions does not defeat the protective action of the redundant actuation train (e.g., train B). A single failure in shared functions does not defeat the protective action of the safety function. It is further noted that the sharing of the logic by manual and automatic initiation is consistent with the system level action requirements of the IEEE Standard 279- 1971, paragraph 4.17, and consistent with the minimization of complexity.
Although manual actuation of main steamline isolation (all valves), containment isolation (Phase A), and containment spray actuation is not within the NSSS scope, the same criteria herein described for the manual safety injection also applies to these aforementioned manual actuation functions in the balance of plant scope.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-22Rev. 301.64Quality Assurance Requirements for the Design of Nuclear Power Plants (Rev. 2, June 1976)The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.64 is presented in WCAP-17.1.3 8370 "WRD Quality Assurance Plan." The Nuclear Fuel Division position on this Regulatory Guid e is presented in WCAP-7800, "NFD Quality Assurance Program Plan."
17.1.2 17.1.317.21.65Materials and Insp ections for Reactor Vessel Closure Studs (Rev. 0, October 1973)Westinghouse is in agreement with Regulatory Guide 1.65 with the following exception for the material and tensile strength guidelines:5.3.1 1.Westinghouse has specified both 45 ft lb and 25 mils lateral expansion for control of fracture toughness determined by Charpy-V testing, required by the ASME Boiler and Pressure Vessel Code,Section III, Summer 1973 Addenda and 10 CFR Part 50, Appendix G (July 17, 1973, Paragraph IV.A.4). These toughness requirements assure optimization of the stud bolt material tempering operation with the accompanying reduction of the tensile strength level when compared with previous ASME Boiler and Pressure Vessel Code requirements.The specification of both impact and maximum tensile strength as stated in the guide results in unnecessary hardship in procurement of material without any additional improvement in quality.The closure stud bolting material is procured to a minimum yield strength of 130,000 psi and a minimum tensile strength of 145,000 psi. This strength level is compatible with the fracture toughness requirements of 10 CFR 50, Appendix G (July 1973, Paragraph 1.C),
although higher strength level bolting materials are permitted by the code. Stress corrosion has not been observed in reactor vessel closure stud bolting manufactured from material of this strength level. Accelerated stress corrosion test data do exist for materials of 170,000 psi minimum yield strength exposed to marine water environments stressed to 75 percent of the yield strength (given in Reference 2 of the Guide). These data are not considered applicable to Westinghouse reactor vessel closure stud bolting because of the specified yield strength differences and a less severe environment; this has been demonstrated by years of satisfactory service experience.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-23Rev. 30The ASME Boiler and Pressure Vessel Code requirement for toughness for reactor vessel bolting has precluded the guide's additional recommendation for tensile strength limitation, since to obtain the required toughness levels, the tensile strength levels are reduced. Prior to 1972, the Code required a 35 ft lb toughness level which provided maximum tensile strength levels ranging from approximately 155 to 178 kpsi (Westinghouse review of limited data - 25 heats). After publication of the Summer 1973 Addenda to the Code and 10 CFR Part 50, Appendix G, wherein the toughness requirements were modified to 45 ft lb with 25 mils lateral expansion, all bolt material data reviewed on Westinghouse plants showed tensile strengths of less than 170 kpsi.Additional protection against the possibility of incurring corrosion effects is assured by:1. Decrease in level of tensile strength comparable with the requirements of fracture toughness as described above.2.Design of the reactor vessel studs, nuts, and washers allowing them to be completely removed during each refueling permitting visual and/or nondestructive inspection in parallel with refueling operations to assess protection against corrosion, as part of the inservice inspection program described in Chapter 5.3. Design of the reactor vessel studs, nuts, and washers, providing protection against corrosion by allowing them to be completely removed during each refueling. The bolting materials are discussed in Chapter 5.4.Use of manganese phosphate or a vapor phase plating process.a.Use of Code Case 1605 does not constitute an issue between the NRC and Westinghouse inasmuch as use of this code case has been approved by the NRC via the guideline of Regulatory Guide 1.85 (see Revision 6, May 1976).1.67Installation of Overpressure Protection Devices (Rev. 0, October 1973)The scope of Regulatory Guide 1.67 is limited to the design of open discharge systems. Installation of overpressure protection devices is in the balance of plant (BOP) and therefore this guide is not in the Westinghouse NSSS scope of supply.5.4.11.3 1.71Welder Qualification for Areas of Limited Accessibility (Rev. 0, December 1973)Westinghouse practice does not require qualification or requalification of welders for areas of limited accessibility as described by the Guide and has provided welds of high quality.5.2.3 5.3.1.4 10.3.6.2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-24Rev. 30Westinghouse believes that limited accessibility qualif ication or requalification, which are additional to ASME Section III and IX requirements, is an unduly restrictive requirement for shop fabrication, where the welders' physical position relative to the welds is controlled and does not present any significant problems. In addition, shop welds of limited accessibility are repetitive due to multiple production of si milar components, and such welding closely supervised.For field application, the type of qualification should be considered on a case-by-case basis due to the great variety of circumstances encountered.1.73Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants (Rev. 0, January 1974)The qualification programs for Westinghouse WRD supplied Class IE electric motor operators, solenoid valves, and limit switches described in WCAP-8587 and WCAP-9688 meet the requirements of Regulatory Guide 1.73.3.11N 8.3.1 1.74Quality Assurance Terms and Definitions (Rev. 0, February 1974)The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.74 is presented in WCAP-8370, "WRD Quality Assurance Plant." The Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."
17.1.217.21.75Physical Independence of Electric Systems (Rev. 2, September 1978)Westinghouse takes exception to the Regulatory Guide 1.75 in several areas as discussed below. These issues have been presented to the Regulatory Staff and are not resolved at this time.7.1 8.3.1.4 1.Isolation Devices (Paragraph 3.8)
Regulatory Position: Interrupting devices actuated by fault current are not isolation devices.Westinghouse Position: Interrupting devices actuated by fault current are isolation devices when justified by test or analysis.2.Cable Spreading Area and Main Control Room (Paragraph 5.1.3)
Regulatory Position: Places additional severe restrictions on equipment in area.Westinghouse Position: The IEEE draft criteria are adequate.3.Instrument Cabinets (Paragraph 5.7)
TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-25Rev. 30Regulatory Position: Separation requirements for instrument cabinets are the same as those for control boards.Westinghouse Position: Separation requirements should not be the same for instrumentation racks and control boards because functional requirements are different. The IEEE draft criteria are adequate.Refer to WCAP-8892-A and FSAR Section 7.1.2.2.1 for further information.1.77Assumption Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors (Rev. 0, May 1974)The result of the Westinghouse analysis shows compliance with the Regulatory Position given in Section C.1 of Regulatory Guide 1.77. In addition, Westinghouse complies with the intent of the assumptions given in Appendix A of the Regulatory Guide.
15.4.2 15.4.7 However, Westinghouse takes exception to Position C.2, which implies that the Rod Ejection Accident should be considered as an emergency condition. Westinghouse considers this a faulted condition as stated in ANSI N18.2. Faulted condition stress limits will be applied for this accident.Westinghouse also complies with Position C.3 for dose calculations and uses the assumptions in Appendix B.1.82Sumps for Emergency Core Cooling and Containment Spray Systems (Rev. 0, June 1974)The Robust Fuel Assembly (RFA) implemented in Cycle 7 (Region 9) includes the debris resistant bottom nozzle (DRBN) and the protective bottom grid (P-Grid) fuel features (see Section 4.2). Due to these Region 9 fuel features, the minimum restriction at the fuel assembly inlet of approximately 0.075 in. is larger than the fine mesh screening for the sump (1/16 in. = 0.0625 in).6.2.2.2 1.83Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes (Rev. 1, July 1975)Access 5.4.2.2 The Westinghouse steam generator design permits access to steam generator tubes for inspection, plugging, or other repair.
16.3 /4.4.5 Baseline Inspection TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-26Rev. 30Westinghouse concurs with the option of the last paragraph of Section B, which permits the shop examination of tubing to serve as an adequate baseline inspection, provided that the examination is done in accordance with the requirements of the ASME Code,Section III, Subsection NB, Article 2550. The owner may, at his option, perform the inspection prior to operation of the plant in accordance with paragraph C.3.a. If the shop examination is chosen to serve as a baseline inspection, the technical details of the procedure should be presented to assure that the shop examination is no less sensitive than the succeeding inservice inspection technique. Sample Selection, Supplementary Sampling, Testing, and Acceptance LimitsThe detailed requirements for inservice inspections are delineated in the Applicant's Technical Specifications.1.84Design and Fabrication Code Case Acceptability - ASME Section III, Division 1 (Rev. 15, May 1979)Westinghouse believes that code cases should be treated as any other part of the code and therefore accepted as approved by the ASME council.5.2.1.2 For Class 1 components, Footnote 6 to 10 CFR 50.55a applies to Millstone 3. Code Case 1528 was used in the procurement of components. Westinghouse has applied for generic approval for the use of this code case via letter NS-CE-1228 of October 4, 1976. Uses of any other code case applicable to Code Class 1 components were identified in applicable safety reports for U.S.A. projects and concurrence of the Commission was obtained by its approval of Applicant's documents and issuance of Construction Permits.For Class 2, 3, and CS components, only code cases approved by the ASME council have been used.1.85Materials Code Case Acceptability -ASME Section III, Division 1 (Rev. 15, May 1979)See compliance for Regulatory Guide 1.84 above.5.2.1.2 1.88Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records (Rev. 2, October 1976)The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.88 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."
17.1.2 17.1.317.2TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-27Rev. 301.89Qualification of Class IE Equipment of Nuclear Power Plants (Rev. 0, November 1974)For Westinghouse NSSS Class IE equipment, Westinghouse meets IEEE Standard 323-1974 (including IEEE 323a-1975 position statement of July 24, 1975) and Regulatory Guide 1.89 by an appropriate combination of any or all of the following:3.11N Type testing, operating experience, qualification, by analysis, and ongoing qualification.This commitment was satisfied by implementation of WCAP-8587, "Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," as discussed in Section 3.11N.1.92Combination of Modes and Spatial Components in Seismic Response Analysis (Rev. 0, December 1974)Westinghouse will use the alternate method described in Section 3.7 to combine modal responses in the evaluation of seismic responses for Millstone 3.3.7.2 1.97Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Rev. 2, December 1980)Westinghouse does not believe that adoption of the full scope of instrumentation proposed by Regulatory Guide 1.97, Revision 2, is necessary for safe operation of a Westinghouse pressurized water reactor. Nevertheless, Westinghouse has established a design basis with respect to Regulatory Guide 1.97, Revision 2, in order to support its customers in responding to this licensing recommendation. The position stated below indicates where the Westinghouse design basis takes exception to the regulatory positions contained in Section C and the implementation criteria contained in Section D of the guide:7.4 7.5 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-28Rev. 301.Regulatory Guide Position C.1 a.The selection criteria specified by the NRC in Table 2 imply that all Type A variables should be Category I. Westinghouse permits Category 2 and 3 under Type A where a variable is employed in the Westinghouse Reference Emergency Operating Instructions for the sole purpose of providing preferred backup information.b.The guide contains (Sections 1.3.1b, 1.3.2.e, f, etc) recommendations concerning the methods of displaying information. Westinghouse has retained the intent of the Regulatory Guide display requirements (i.e., information should be immediately available, versus continuously displayed, etc) in the design basis, but has permitted the method of display to be optimally selected elsewhere in conjunction with the implementation of requirements contained in NUREG-0700, "Human Engineering Design Guidelines."c.The guide designates Type D and E key variables as Category 2. In addition, Westinghouse has designated as Category 2 those Type A, B, and C variables which provide preferred backup information and whose instrumentation will be subject to a high energy line break environment, when required, to provide the backup information.d.The qualification conditions specified for Category 2 contain the definitions for two subcategories by indicating that seismic qualification may be required for only those instruments associated with non safety systems need not be seismically qualified. In both cases, however, environmental qualification to Regulatory Guide 1.89 is specified. The logic for requiring environmental qualification, but not seismic, in certain cases is not explained and furthermore:-Regulatory Guide 1.89 itself requires seismic qualification as an essential part of the qualification sequence.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-29Rev. 30-The basis for identifying what instrumentation is part of a safety-related system is open to interpretation. For example, this could mean that only instrumentation that is an essential part of the safety system performance is included (i.e., actuation of safeguards). Alternatively, this definition could be interpreted to include instrumentation that merely monitors the performance of the safety-related system (i.e., status lights).As a consequence, Westinghouse does not support this subdivision of Category 2 and recommends that Category 2 instrumentation be qualified for at least the environment (seismic and/or environmental) in which it must operate to service its intended post-accident function.e.The NRC has indicated (Section 1.3.3.a) that Category 3 instrumentation should be of high quality commercial grade and should be selected to with stand the specified service environment. In addition, it is Westinghouse's position that Category 3 instrumentation should not be required to provide information to the operator when exposed to a hostile environment resulting from a high energy line break.f.Westinghouse does not believe it appropriate to specify periodic checking, testing, calibration, and calibration verification in accordance with Regulatory Guide 1.118 for commercial grade instrumentation and has deleted this requirement. The scope of Regulatory Guide 1.118 is restricted to periodic testing of the protection system and electric power systems for systems important to safety. As such, it should not be applied in relation to non safety equipment.2.Regulatory Position C.2 The NRC definition of Type D employs the terminology; plant safety systems and other systems important to safety. Westinghouse does not employ this terminology since it has no strict definition. Westinghouse has defined the scope of Type D to include safety systems employed for mitigating the consequences of an accident and for achieving subsequent plant recovery to a safe shutdown condition and other systems normally employed fo r attaining safe shutdown.3.Implementation - Section D TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-30Rev. 30The implementation section of the guide recognizes, in some part, the need to modify some of the requirements when considering backfit to operating plants or plants under construction. However, many of the regulatory guides referenced in the Regulatory Position (Section C) are not superseded by any statement under Section D. Westinghouse believes Regulatory Guide 1.97, Revision 2, should be revised to rectify this omission. It is the Westinghouse position that the criteria to be specified for Category I post-accident monitoring instrumentation for operating or licensed plants should be at least equivalent to those criteria originally specified for the plant's Class IE systems.1.99Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials (Rev. 1, April 1977)Justification of the Westinghouse position on Revision 0 and Revision 1 of the Guide is detailed in References 1 and 2, respectively.5.3 In summary, Revision 1 of the guide is substantially identical with Revision 0, with minor clarifications and inclusion of a new position C.2, which had previously been included in the Discussion section of Revision 0.The Westinghouse letter of comment on Revision 1 reiterates the comments of Revision 0 and includes further clarification of vessel material hardship imposed by the guide.The Westinghouse position, with respect to each of the guide positions, is as follows:1. Regulatory Position C.1 The basis, as well as the scope, of the guide for predicting adju stment of reference temperature as given in Regulatory Position C.1 are inappr opriate since the data base used was incomplete and included so me data which were not applicable.2.Regulatory Position C.2 Westinghouse is in agreement with the Guide Position C.2a. However, with respect to Guide Position C.2b, Westinghouse believes that Figure 2 of the Guide is incorrect since the upper shelf energy for the 6-inch thick ASTM A302B reference correlation monitor material reported by Hawthorne indicates essentially a constant upper shelf at fluences above 1 x 10 n/cm (Hawthorne).TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-31Rev. 303.Regulatory Position C.3 The Westinghouse position, with reference to the Guide Position C.3 controlling residual elements to levels that result in a predicted adjusted reference temperature of less than 200°F at end-of-life, is that the stresses in the vessel can be limited during operation in order to comply with the requirements of Appendix G to 10 CFR Part 50, even though the end-of-life adjusted reference temperature may exceed 200
°F. By applying the procedures of Appendix G to ASME Section III, the stress limits, including appropriate Code safety margin, can be met.
References Letter of Comment on Revision 0 of the Guide to the Secretary of the Commission by C.E. Eicheldinger, NS-CE-784, September 22, 1975.Hawthorne, J.R., "Radiation Effects Information Generated on the ASTM Reference Correlation - Monitor Steels," to be published.1.100Seismic Qualification of Electric Equipment for Nuclear Power Plants (Rev. 1, August 1977)Westinghouse qualifies equipment to the requirements of IEEE 344-1971. Westinghouse has performed extensive testing to demonstrate the adequacy of seismic qualification to the requirements of IEEE 344-1971. See Section 3.10 for a further discussion of the seismic qualification of electrical equipment.Replacement items meet the original criteria or either IEEE 344-75 or IEEE-344-87 (endorsed by Reg. Guide 1.100, Rev. 2).1.105Instrument Setpoints (Rev. 1, November 1976)Technical Specifications provide the margin from the nominal setpoint to the technical specification limit. The allowances between the technical specification limit and the safety limit include the following items:7.1.2.1.9 a.The inaccuracy of the instrument7.2.2.2.1 b.Process measurement accuracy16.3/16.4.3c.Uncertainties in the calibrationd.The potential transient overshoot determined in the accident analyses (this may include compensation for the dynamic effect), andTABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-32Rev. 30e.Environmental effects on equipment accuracy caused by postulated or limiting postulated events (only those systems required to mitigate consequences of an accident).Westinghouse designers choose setpoints such that the accuracy of the instrument is adequate to meet the assumptions of the safety analysis.The range of instruments is chosen based on the span necessary for the instrument's function. Narrow range instruments will be used where necessary. Instruments will be selected based on expected environmental and acci dent conditions. The need for qualification testing will be evalua ted and justified on a case basis.Administrative procedures coupled with the present cabinet alarms and/or locks provide sufficient control over the setpoint adjustment mechanism such that no integral setpoint securing device is required. Integral setpoint lock ing devices will not be supplied.The assumptions used in selecting the setpoint values in Regulatory Position C.1 and the minimum margin, with respect to the technical specification limit and calibration uncertainty, will be documented by Westinghouse. Drift rates and their relationship to testing intervals will not be documented by Westinghouse.1.116Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems (Rev. O-R, May 1977)The subject of the Regulatory Guide is not in the Westinghouse NSSS scope of supply. The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.116 is presented in WCAP-8370, "WRD Quality Assurance Plan." This Regulatory Guide is not applicable within NFD.
17.1.217.21.118Periodic Testing of Electric Power and Protection Systems (Rev. 2, June 1978)Westinghouse will make clear the distinction between "recommendations" and "requirements" when addressing criteria. The position is as follows:7.1.2.4 Westinghouse defines "Protective Action Systems" to mean electric, instrumentation, and controls portions of those protection systems and equipment actuated and controlled by the protection system.8.1.6 Equipment performing control functions, but actuated from protection system sensors is not part of the safety system and will not be tested for time response.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-33Rev. 30Status, annunciating, display, and monitoring functions, except those related to the Post Accident Monitoring Systems (PAMS) are considered by Westinghouse to be control functions. Reasonability checks, i.e., comparison between or among similar such display functions, will be made.Response time testing for control functions operated from protection system sensors will not be performed. Moreover, Nuclear Instrumentation sensors are exempt from testing since "their worst case response time is not a significant fraction of the total overall system response (i.e., less than 5 percent)." This exemption is permitted by IEEE-338.The standard Westinghouse protection system design does not include provisions which permit in-situ testing of process sensors.1.121Bases for Plugging Degraded PWR Steam Generator Tubes (Rev. 0, August 1976)Millstone 3 complies with Regulatory Guide 1.21, Rev. 0, August 1976 with the following clarifications:5.4.2 1.Regulatory Position C.1 Westinghouse interprets the term "Unacceptable defects" to apply to those imperfections resulting from service induced mechanical or chemical degradation of the tube walls which have penetrated to a depth in excess of the Plugging Limit.2.Regulatory Position C.2a(2) and C.2a(4)
Westinghouse will use a 200-percent margin of safety based on the following definition of tube failure. Westinghouse defines tube failure as plastic deformation of a crack to the extent that the sides of the crack open to a nonparallel, elliptical configuration. This 200-percent margin of safety compares favorable with the 300-percent margin requested by the NRC against gross failure.3.Regulatory Position C.2.b a.In cases where sufficient data exist to establish degradation allowance, the rate used will be an average time-rate determined from the mean of the test data.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-34Rev. 30b.Where requirements for minimum wall are markedly different for different areas of the tube bundle, e.g., U-bend area versus straight length in Westinghouse designs, two plugging limits may be established to address the varying requirements in a manner which will not require unnecessary plugging of tubes.4.Regulatory Position C.3.d(1) and C.3.d(3)
The combined effect of these requirements would be to establish a maximum permissible primary-to-secondary leak rate which may be below the threshol d of detection with current methods of measurement.Westinghouse has determined the maximum acceptable length of a through-wall-crack based on secondary pipe break accident loadings which are typically twice the magnitude of normal operating pressure loads. Westinghouse will use a leak rate associated with the crack and size determined on the basis of accident loadings.5.Regulatory Position C.3.e(6)
Westinghouse will supply computer code name s and references rather than the actual codes.6.Regulatory Position C.3.f(1)
Westinghouse will establish a minimum acceptable tube wall thickness (Plugging Limit) based on structural requirements and consideration of loadings, measurement accuracy, and, where applicable, a degradation allowance as discussed in this position and in accordance with the general intent of this guide. Analyses to determine the maximum acceptable number of tube failures during a postulated condition ar e normally done to entirely different bases and criteria are not within the scope of this guide.1.123Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Rev. 1, July 1977)The Westinghouse position for the WRD NSSS scope of supply in Regulatory Guide 1.123 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."
17.1.317.21.126An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification (Rev. 1, March 1978)Millstone 3 does not use the model and related statistical method for the analysis of fuel densification that is presented in Regulatory Guide 1.126.4.2 TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSARMPS3 UFSAR1.8N-35Rev. 30The Regulatory Guide clearly states that, "The model presented in...this Guide is not intended to supersede NRC approved vendor models." Millstone 3 uses the Westinghouse fuel densification model presented in WCAP-8218 (Proprietary) which has been approved by the NRC. WCAP-8219 (Nonproprietary) and WCAP-8264 (Customer Version) are companions to the approved versions.1.139Guidance for Residual Heat Removal (Rev. 0, May 1978)Millstone 3 meets the requirements of BTP RSB 5-1 and SRP 5.4.7 except as indicated in FSAR Sections 5.4.7.1 and 5.4.7.2.4.1.141Containment Isol ation Provisions for Fluid Systems (Rev. 0, April 1978) Westinghouse's containment isolation philosophy for fluid systems complies with the guidance provided by ANSI N271-1976 and/or Regulatory Guide 1.141 with the following exceptions and/or clarifications: 5.4.5 5.4.6 6.2.4 1.The standard in Section 3.6.3 states that remote manual closure of isolation valves on ESF or ESF related systems is acceptable when provisions are made to detect possible failure of the fluid lines inside and outside containment. Although such provisions are outside Westinghouse scope of supply, Westinghouse is of the opinion that provisions to detect failure of fluid lines inside containment are unnecessary. Since redundant ESF capacity is provided and off site doses due to leakage inside containment are not a concern, Westinghouse does not require or provide for detection of failures in fluid lines inside containment.2.Section 3.6.4 states that a single valve and closed system outside containment is acceptable if the closed system is treated as an extension of the containment. Further, the standard requires that the valve and the piping between the valve and the containment be enclosed in a protective leak tight or controlled leakage compartment. The closed system is also required to be leak tested in accordance with 10 CFR 50 Appendix J unless it can be shown by inspection that system integrity is being maintained for those systems operating during normal plant operation at a pressure equal to or above the containment design pressure.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of ComplianceFSAR Section Reference MPS3 UFSAR1.8N-36Rev. 30Sections 4.2.5 and 4.4.6 of the standard as implemented by item C.3 of the Regulatory Guide are interpreted by Westinghouse to state that to preclude common mode failures, diversity is required in the parameters sensed from which isolation signals are generate dWestinghouse design criteria for the initiation of containment isolation do not include a requirement for diversity in the primary system variables for any given design conditio n or event.Westinghouse, however, utilizes different primary system variables to derive and generate the protection function for the first phase (A) of containment isolation. Some diversity is therefore available for this phase for a given event. The second phase of containment isolation (B), which isolates only component cooling water to the reactor coolant pump, is initiated only by a high containment pressure signal. Diversity is therefore not available for this containment isolation function. Westinghouse believes that restricting the standards recommendation (i.e., for diverse means of actuation) to diversity in parameters sen sed that the Regulatory Guide unjustifiably ignores the benefits of alternate methods of obtaining different means of actuation. Westinghouse believes the standard gives appropriate guid ance and the Regulatory Guide restriction should be deleted.The standard states in Section 1 that "If an accident occurred, fluid systems penetrating the containment would be isolated except those which are engineered safety related." With respect to this recommendation, the reactor coolant pump seal injection line is either isolated or not isolated, depending on whether the charging pump is used for safet y injection and the value of the pump header discharge pressure. On plants which utilize the charging pumps for safety injection, flow will be provided by the pumps through th e seal injection lines following an accident. Each line is, however, equipped with a remot emanual containment isolation valve which the operator can close when the charging pumps have completed their safeguard function.1.144Auditing of Quality Assurance Program for Nuclear Power Plants (Rev. 0, January 1979)The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.14 4 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, "NFD Quality Assurance Program Plan."1.146Qualification of Quality Assurance Program Audit Personnel for Nuclear Power (Rev. 0, August 1980)The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 1.14 6 is presented in WCAP-8370, "WRD Quality Assurance Plan." The Regulatory Guide is no tapplicable within NFD.TABLE 1.8N-1 NRC REGULATORY GUIDES (CONTINUED)R.G. No.TitleDegree of Compliance MPS3 UFSAR1.9-1Rev. 30
1.9 STANDARD
REVIEW PLAN DOCUMENTATION OF DIFFERENCESThe Millstone 3 FSAR was reviewed against NUREG-0800 at the time of application for an operating license to satisfy the requirements of 10 CFR 50.34(g). The following two tables were developed to identify deviations from SRP acceptance criteria and provided a justification for those deviations. The differences noted were not construed as variances from regulation, rather, they documented the deviations from acceptance criteria as stated within the NRC's internal review guide (SRP) for sa fety analysis reports.Table 1.9-1 summarized the differences be tween the Millstone 3 FSAR and NUREG-0800.Table 1.9-2 presented the FSAR differences from NUREG-0800 and th eir justifications.Information contained in this section (Tables 1.9-1 and 1.9-2) has been retained for historical purposes.
MPS3 UFSARMPS3 UFSAR1.9-2Rev. 30TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section2.1.3 (Rev. 2)II.6 - Population density within 30-mi radius.A distance of 60 km (37 mi) used for checking density.2.1.3.6 2.5.1 (Rev. 2)II. 4.b,c, and d - Zone s of alterations, irregular weathering, structural weakness; unrelieved residual stresses in rock
- unstable material or areas considered unstable.These features are not discussed.2.5.1.2 2.5.2 (Rev. 1)II (2.5.2.1) - Magnitude designations. Earthquake magni tudes not identified as M b , M L , or M s.Tables 2.5.2-3 and 2.5.2-4 II (2.5.2.4) - Earthquake return period.There is no proba bilistic determination of earthquake return period for the largest earthquakes in each province.
2.5.2.4 II (2.5.2.7) - Probability of exceeding acceleration of OBE.No estimate given for probability of exceeding acceleration of OBE.
2.5.2.7 2.5.3 (Rev. 2)II (2.5.3.1) - Offshore geologic investigation.No detailed offshore geologic investig ation of 5-mi radius has been done.
2.5.3.1 II (2.5.3.4) - Fault location.Location and investigation of all faults within 5 mi has not been done.
2.5.3.4 II (2.5.3.6) - Age of faults.
Age documentation of all faults within 5 mi has not been done.2.5.3.6 2.5.4 (Rev. 2)II (2.5.4.2) - Analys es for saturated soils and clays underlying the site.
No FSAR table lists values of parameters used in analyses of liquefaction potential, behavior, or static and dynamic behavior.2.5.4.2 3.2.2 (Rev. 1)II - Use of RG 1.26 as the acceptance criteria for defining Quality Groups for components important to safety.Millstone 3 uses the classification system provided in ANS 18.2.3.2.2 3.4.1 (Rev. 2)III.3 - Postulated failures of nonseismic Category I and non-tornado protected tanks.
FSAR does not directly address these postulated failures.3.4.1, 3.6 MPS3 UFSARMPS3 UFSAR1.9-3Rev. 303.4.2 (Rev. 2)II. 1 - Use of DBF or highest groundwater level in design.DBF or normal groundwater level used.3.4.2 3.5.1.3 (Rev.
1)II - Missile protection for cold shutdown maintenance components.
Cold shutdown maintenance co mponents not specified as targets.3.5.1.3 3.5.1.5 (Rev.1)III.3 - Definition of P P.Missiles that produce seconda ry missiles which could damage vital equipment not considered.
3.5.1.5, 2.2.3.2.1 3.5.1.6 (Rev.
2)III.2 - Inflight crash rate of 4x10
-10.NUREG-75/087 crash rate of 3x10
-9 used.3.5.1.6 3.6.1 (Rev. 1)BTP ASB 3-1, B.1.a(1) -Requires an arbitrary split be postulated on the main steam and the feedwater systems at a location proximate to essential systems.FSAR does not commit to postulate this arbitrary split.3.6.1.3.3 BTP ASB 3-1, B.1.a(2) -Suggests main steam and feedwater pipes not be routed in the vicinity of the control room.
Main steam and feedwater pipes are routed in the vicinity of the control room.
3.6.1.3.3 BTP ASB 3-1, B.2.a -States that essential systems and components should be designed to meet the seismic design criteria of R.G. 1.29.
In the aux steam and hot water heating systems, only the electrical detecti on and actuation devices for the isolation valves are qualified to Class 1E requirements and located in a seismic Cat. I Building.
3.6.1.3.1 3.6.2 (Rev. 1)III.2.a - Requires use pressure and temperature values corresponding to the greater contained energy at hot standby or at 102 per cent power.
FSAR uses internal pressure, and temperature conditions in the piping system duri ng reactor operation at 100 percent power.
3.6.2.2.1 III.2.a - Requires that the allowable capacity for crushable material shall be limited to 80 percent of its rated energy absorbing capacity as
determined by dynamic testing.FSAR uses an allowable of 80 percent of energy absorbing capacity based on static testing.
3.6.2.2.1 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-4Rev. 30BTP MEB 3-1, B.1.e -States particular criteria for postulating through wall leak age cracks in high energy Class 1, 2, 3, and non-nuclear piping.FSAR does not postulate cracks in high energy piping.3.6.2.1.2 3.7.2 (Rev. 1)II.4 - Finite element and half space representation of subgrade soil stiffness.
FSAR does not address half space representation.3.7.2 II.11 - Consideration of accidental torsion by assuming additional seismicity of
+/-5 per cent of maximum building dimension.Additional seismicity not addressed.3.7B.2.11 3.7.3 (Rev. 1) (BOP Scope)
II.2.l (1) - Interaction of Category I piping extended to address on a "system basis."
FSAR discussion is restricted to piping.3.7B.3.13 3.7.3 (Rev. 1) (NSSS Scope)II.2.g - Combination of closely spaced modes should be in accordance with Regulatory Guide 1.92.Westinghouse combines closely spaced modes as described in FSAR Section 3.7N.3.7.
3.7N.3.7 3.8.1 (Rev. 1)II.2 - Use of Regulatory Guide 1.136.FSAR does not reference Regulatory Guide 1.136.3.8.1 11.4.f - Use of ASME III, Division 2, Article CC-3000 for design of c ontainment structure tangential shear.
Article CC-3000 of ASME II I, Division 2 was not not used.3.8.1.4.1 II.4.j - Ultimate capacity of reactor containment.Ultimate capacity of the reactor containment is not discussed.
3.8.1 II.5 - Article 3000 of AS ME III, Division 2 for loads, load combinations, and stress allowables.Article 3000 of ASME III, Division 2 not used.3.8.1.2 3.8.3 (Rev. 1)II.2 - ACI 349-76.ACI 349-76 was not used.3.8.3.2, 3.8.1.2 3.8.4 (Rev. 1)II.2 - ACI 349-76.ACI 349-76 was not used.3.8.4.2, 3.8.1.2II.4.d - Design report format.FSAR does not use this format.3.8.4 3.8.5 (Rev. 1)II.4.b - ACI 349-76. rather than ACI349-76.ACI 318-71 was used 3.8.5.2, 3.8.1.2 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-5Rev. 303.9.1 (Rev. 2) (BOP Scope)III.1 - Plant conditions identified as design Levels A,B,C,D.FSAR identifies plant co nditions as normal, upset, emergency, and faulted.
3.9B.1.1 III.4 - Methods used in stress analysis of components.
FSAR contains no justification for methods used3.9B.1.4 3.9.1 (Rev. 2) (NSSS Scope)II.2 - Computer codes used in design and analysis of seismic Category I components.
Only a brief description of computer codes used by Westinghouse is given.
3.9N.1.2 3.9.2 (Rev. 2) (BOP Scope)
II.1.d - List snubbers on systems which experience sufficient thermal expansion.FSAR does not provide a list of snubbers.3.9B.2 II.1.e and f - Tests to ve rify thermal expansion/
vibration measurements.
FSAR does not provide a description of tests.3.9B.2 II.2 - Seismic sub- system analysis.Informati on is not contained in FSAR Section 3.9B.2.3.9B.3 3.9.2 (Rev. 2) (NSSS Scope)II.2.e - Criteria for combining closely spaced modes.Westinghouse method is provided in FSAR Section 3.7N.3.7.3.7N.3.7 3.9.3 (Rev. 1) (BOP Scope)II.1 - Stress limit criteria.FSAR does not reflect the stress limit criteria.3.9B.3.1 II.2 - Information on Class 3 safety/relief devices.FSAR does not address Class 3 safety/relief devices.3.9B.3 II.3 - Information on snubbers.Requirements not addressed in FSAR.3.9B.3 3.9.3 (Rev. 1) (NSSS Scope)II.1 - Design criteria for internal parts of components such as valve discs and pump shafts.Westinghouse does not provide criteria for the nonpressure boundary portions of ASME Code Class 1, 2, and 3 components in the FSAR.
3.9N.3 Appendix A, 1.3.3 -Desi gn basis pipe break (DBPB)Westinghouse defines DBPB as a faulted, not emergency condition.
3.9N Appendix A, 3.1 - Stress limits and loading combinations for core support structures.
FSAR does not provide a table defining stress limits and loading combinations for core support systems.
3.9N TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-6Rev. 303.10 (Rev. 2) (BOP Scope)
II.1.a (14) (b) iii and iv - LOCA induced hydraulic forcing functions or differential pressures upon valve discs or pump rotors. There is no analysis of LOCA induced hydraulic functions or differential pressures upon valve discs or pump rotors.
3.10B II.5.c - Seismic Qualification Report format.Seismic Qualification Report format was not used to document seismic qualification.
3.10B II - Mechanical equipment seismic and operability qualification.FSAR does not address mechanical equipment seismic and operability qualification.
3.10B II.1 - Verify operability of pumps and valves during all operational c onditions by test and analysis.Pump operability has only been performed by analysis.3.10B II.5.b - Listing of systems necessary to perform functions out lined in SRP 3.10.This list is not included in FSAR Section 3.10.3.10B 3.10 (Rev. 2) (NSSS Scope)II - Mechanical equipment seismic and operability qualification.FSAR does not address mechanical equipment seismic and operability qualification.
3.10N II.1.a(2) - Testing of equipment in the operational condition.
Flow loads are not superimposed on seismic loads for valve operability tests.
3.10N II.1.a(8) - Fixture design for seismic tests. The se ismic qualification test ing configurations are designed to represent the typical plant installation.
3.10N II.1.a(10) - Static testing of pump or valve assemblies.End loadings are not applied and all dynamic amplification effects are not included in the static deflection test for active valves.
3.10N II.1.a(14)(a) - Operability of active pumps and valves. Operability of active pumps and valves are not covered in FSAR Section 3.10N.
3.7N, 3.9N The Millstone program utilizes a combination of test and analysis to demonstrate operability for active valves.
3.9N TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-7Rev. 30 No specific tests are done on valve gate, disc assemblies, motor, etc.
3.9N II.1.a(14)(b) viii - Use of R.G. 1.92 for combination of multimodal and multidirectional
responses.Westinghouse utilizes the methods defined in FSAR Section 3.7 for combining closely spaced modes.
3.7N II.1.b(3) - Testing of supports.Seismic test ing of all supports is not conducted.3.10N II.1.c - Seismic and dynami c testing according to IEEE 323-1974.
For some mechanical components, aging and sequence testing was not included.
3.10N II.3 - Requirements for central files.Requirements for central files are not addressed in FSAR Section 3.10N.
3.10N II.5.b(1) - Requirement fo r a list of systems necessary to perform the functions outlined in SRP 3.10.This list is not included in FSAR Section 3.10N.3.10N, Chapters 6 and 7 II.5.b(2) - Description of the results of any in-plant tests.
Actual test results are not included in the FSAR.3.10N II.5.c - Contents of Seismic Qualification Report (SQR).Westinghouse does not maintain such a report for Millstone 3.
3.10N 3.11 (Rev. 2) (BOP Scope)II - Mechanical equipment qualification.FS AR does not address mechanical equipment qualification.3.11B II - NUREG-0588 methodologies. NUREG-0588 methodologies are not strictly followed.3.11B 3.11 (Rev. 2) (NSSS Scope)I.1 - Requirement for a list of systems necessary to perform functions outlined in SRP 3.11.
This list is not included in FSAR Section 3.11N.Chapters 6 and 7, 3.7N, 3.9N, 3.11N II - NUREG-0588.No reference is made in FSAR Section 3.11N to the results of NUREG-0588 study. 3.11N TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-8Rev. 30Scope of SRP 3.11 -Environmental qualification of mechanical and electrical equipment.
Environmental qualification of mechanical electrical equipment is not addressed in FSAR Section 3.11N.
3.9N 4.5.1 (Rev. 2)II.4 - On site cleaning requirements.On site cleaning requirem ents are not addressed in FSAR Section 4.5.1.
1.8, 17.1.25.2.1.1 (Rev.
2)II - Compliance with 10 CFR 50.55a.FSAR us ed ASME III 1971 Ed. through Summer 1972 addenda instead of Winter 1972 addenda for design and fabrication of loop bypass valves.
5.2.1.1 Westinghouse uses ANS standards rather than R.G. 1.26.1.8N, 3.2.2 5.2.5 (Rev. 1)III.7 - Leakage detection testing requirements.FSAR does not co mpletely address testing of the unidentified leakage sump system.
5.2.5.2.7, 9.3.3 5.3.1 (Rev. 1) (NSSS Scope)II.6.c(3) - Capsule removal schedule.The tentative capsule removal schedule is not identical to the removal schedule described in 10 CFR 50, Appendix H, II.C.3.b.
5.3.1.6 5.4.1.1 (Rev.
- 1) (NSSS Scope)II.2.b - Normal operating te mperature is at least 100°F above the RTNDT.The FSAR states that the RTNDT is no higher than 10
°F.5.4.1.1.3 5.4.2.1 (Rev.
- 2) (NSSS Scope)BTP MTEB 5-3 -Secondary side chemistry program. Free hydroxide concentration is not measured. Procedure number and basis not supplie d for chemical analysis.
5.4.2.1 BTP MTEB 5-3, II.2 -Discussion of "clean metal" condition prior to startup.FSAR does not address this concern.5.4.2.1 II.B.2 - Access to remove sludge by lancing from tube support plates.Tube lancing from tube support plates is not discussed in the FSAR.5.4.2.1 6.2.1 (Rev. 2) (BOP Scope)II.f (6.2.1.1A) - Adequate margin above the maximum expected external pressure.Actual margin of external pressure analysis not specifically addressed.
6.2.1.1.3.5 II.B.2 (6.2.1.2) -NUREG-0609.NUREG-0609 was not addressed.6.2.1.2 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-9Rev. 306.2.1 (Rev. 2) (NSSS Scope)II.2 (6.2.1.5) - BTP CSB 6-1 Section B.3b recommends conservati ve condensing heat transfer coefficients.Westinghouse values for th e long-term post-blowdown condensing heat transfer coef ficients are nonconservative.
6.2.1.5 II.B.3.e (6.2.1.3) - Models for calculating mass and energy releases for containment design basis calculations.Model described in FSAR differs from Westinghouse model referenced in the SRP.
6.2.1.3.3, 6.2.1.3.4, 6.2.1.3.5, 6.2.1.3.6 6.2.2 (Rev. 3)II.1.a (6.5.2) - Automatic switchover to recirculation mode.Millstone 3 switchover is manual. 6.2.2 II.1.e (6.5.2) - pH between 8.5 and 10.5 for fission product control.
Millstone 3 is designed to a minimum pH of 7.06.2.2 6.2.3 (Rev. 2)II.D.1 and II.D.2 -Discussion heat transfer analysis and high energy line considerations.
FSAR does not provide a discussion of these subjects.6.2.3 6.2.5 (Rev. 2)II.3 - Mixing character istics of the containment (plant specific analyses).
Millstone 3 references the anal yses of plants with similar designs.6.2.5.3 II.11 - Containmen t hydrogen monitor requirements.
Hydrogen monitors will comply with these requirements by core load.
6.2.5 6.3 (Rev. 1)II - Discussion of non-safety grade interactions with ECCS.FSAR Section 6.3 does not discuss this subject.3.6 6.4 (Rev. 2)II.5.b - Compliance to Regulatory Guide 1.95.The chlorine detectors are not Seismic Category I.6.4.3
6.4 Appendix
A (Rev. 2)
Single failure (active) criteria.Valves isolating air inlet ducting located in series.6.4.3 Item 6 - No manual action credit allowed for repairs until two hours.
Isolation valves can be manually opened within 10 minutes, thus, credit for manually opening the valves within one hour should be acceptable.
6.4.3 6.5.1 (Rev. 2)II - Compliance to Regulatory Guide 1.52.See exceptions listed in Section 1.8.6.5.1 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-10Rev. 30 II - Continuous indication and recording of air flow for ESF filtration units.This is not provided.6.5.1 II - Flow sensors in ESF filtration units.These are not provided.6.5.1 II - Compliance to Section 8.3.1.6 of ANSI N510-1980.Three ESF filter train systems do not comply. 6.5.1 II - Compliance to Sections 8.3.1.5, 8.3.1.6, and 8.3.1.7 of ANSI N510-1980. Three ESF filter train systems do not comply.6.5.1 6.5.2 (Rev. 1)Containment spray as a fission pr oduct cleanup system. Containment spray system requirements are not discussed in Section 6.5.2.
6.5.2 7. 2 (Rev. 2)BTP ICSB 26 - Sensor qualification.Sensors for reactor trip on turbine trip when power level is 50% or more are not seismically qualified.
7.2.1.1.2 7.5 (Rev. 2)III.6 - NUREG-0696 compliance.The Safety Parameter Display System, and the Emergency Response Facilities are not discussed.
7.5.3 8.3.1 (Rev. 2)II.4.f - Compliance to NUREG/CR-0660.NUREG/CR-0660 is not addressed. See Table 1.9-2 for details.8.3.1 9.1.2 (Rev. 3)III.2.e - Evaluation of lighter load drops at maximum heights.This evaluation has not been performed.9.1.2.3 9.1.3 (Rev. 1)II.1.d (4) - BTP ASB 9-2 decay heat removal.Decay heat removal is based on DECOR (based on ORIGEN2) computer code a nd credit for evaporative cooling instead of BTP ASB 9-2.
9.1.3.2 9.1.3 (Rev. 1)III.1.d - For maximu m normal heat load, the pool temperature should be kept at or below 140
°F The maximum temperature for a normal heat load is 150°F.9.1.3.3 9.1.3 (Rev. 1)III.1.h (ii) - Maximum heat load is after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decayThe decay time for the maximum heat load is based on the heat removal capacity of the spent fuel pool heat exchangers and varies from 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> to 349 hours0.00404 days <br />0.0969 hours <br />5.770503e-4 weeks <br />1.327945e-4 months <br />.
9.1.3.3 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-11Rev. 309.1.4 (Rev. 2)III.6 - Evaluation of lighter load drops at maximum heights.This evaluation has not been performed.9.1.4.1 9.2.1 (Rev. 2)III.3.d - Location of radiation monitors.No manual valve in series with motor operated valve.9.2.1 9.2.2 (Rev. 1)II.3.e - Loss-of-c oolant test for reactor coolant-pumps.Reactor-coolant-pumps have not been tested for the 20 minute time requirement.
9.2.2 9.4.1 (Rev. 2)II.4 - Compliance to Regulatory Guide 1.95.The chlorine detectors are not Seismic Category I.9.4.1.3 9.4.5 (Rev. 2)II.4 - Protection from The bottoms of the fresh air intakes are not all located at least 20 feet above grade elevation.
9.4.5 II.5 - Detection and C ontrol of airborne contamination leakage from the system.Only normal building ventilation is monitored.9.4.5.2 III.3.b - Tornado protection.No protection of ductwork from negativ e pressure due to tornado.9.4.5.4 BTP CMEB 9.5-1 (Rev 2)Refer to the Fire Prot ection Evaluation Report, Appendix B, for a comparison of Millstone 3 design to BTP CMEB 9.5-1 guidelines.
Fire Protection Evaluation Report 9.5.4 (Rev. 2)II.4.b - Regulatory Guide 1.137. Millstone 3 has two 3.5-day capacity fuel oil tanks.9.5.4.2 III.5 - Turbulence of sediments.There are no tank design features which minimize turbulence of sediments.
9.5.4.2 II.2 (III.6.a) - Missile protection.The fill lines for the dies el generator fuel oil vaults are protected from missiles.
9.5.4.2 9.5.8 (Rev. 2)II.4.g (III.8) - Reduc ing airborne particulate material.
The bottoms of the fresh air in takes are all located at least 20 feet above grade elevation.
9.5.8 10.2.3 (Rev.
1)II.1 - FATT and Charpy V-notch energies.GE does not provide data for FATT and Charpy V-notch energies to compare with SRP.
10.2.3 10.3 (Rev. 2)III.5.d - Tabulation of all flow paths.FSAR does not tabulate this information. 10.3 11.5 (Rev. 3)Table 1, Item 6 - Fuel storage area ventilation. No automatic termination of effluents. 11.5.2.2.9 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSARMPS3 UFSAR1.9-12Rev. 30Table 2, Item 5 - Spent fuel pool treating system.No automatic termination of effluents.9.3.2, 9.4.2 Table 2, Items 16 and 17 -Steam generator blowdown system.No automatic termination of effluents.11.5.2.3.3 12.2 (Rev. 2)I.2 - Tabulation of concentrations of airborne radioactive materials.
Only normal operation and anticipated operational occurrences are addressed.
12.2.2 13.5.2II.C.2 - ANSI/ANS 3.2-1981, Section 5.3FSAR uses ANSI N18.7-1976/ANS 3.2, Section 5.313.5.2 14.2 (Rev. 2)II.4 - Categories of reportable occurrences that are repeatedly being experienced at other facilities.FSAR does not provide categories of occurrences.14.2 15.4.6Entire SRP.FSAR does not address this accident scenario.15.4.6 15.4.8 (Rev.
1)III - Stresses should be evaluated to emergency conditions for these accidents.Westinghouse considers a faulte d condition as stated in ANSI N18.2.
15.4.8 15.6.5 (Rev.
2)II.3 - TMI Action Plan, II.K.3.30 and II.K.3.31.No modifications have b een made to the small break LOCA model.
15.6.5.3 15.7.3 III.1.a - Radionuclide inventory in failed components FSAR analyzed postulated tank failure using 1
% fuel defects.2.4.13.3 15.7.3.2 TABLE 1.9-1
SUMMARY
OF DIFFERENCES FROM SRP (CONTINUED)SRP SectionSpecific SRP Acceptance CriteriaSummary Description of DifferenceCorresponding FSAR Section MPS3 UFSAR1.9-13Rev. 30TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS SRP 2.1.3SRP TITLE: POPULATION DISTRIBUTIONA.Actual difference between FSAR and SRP Population distribution wheels conform to the metric radius distances defined in the Environmental Standard Review Plans (ESRPs) and were used to assure conformance with the EROLS. Consequently, a distance of 60 km (37 miles) is used for checking density in FSAR Section 2.1.3.6 instead of a distance of 30 miles specified in SRP 2.1.3, Paragraph II.6.B.Justification for difference from SRPDistances were defined to correspond to those being used in the EROLS per the instructions of the ESRP (NUREG-0555) which, at the time of the EROLS preparation, was the most current document offering guidance for EROLS preparation. To assure consistency between the EROLS and FSAR, metric distan ces were used in both documents.
SRP 2.5.1SRP TITLE: BASIC GEOLOGIC AND SEISMIC INFORMATION A.Actual differences between FSAR and SRP1.SRP 2.5.1, Paragraph II.4.b, requires a di scussion on zones of alterations, irregular weathering, and structural wea kness. These are not discussed in FSAR 2.5.1.2.SRP 2.5.1, Paragraph II.4.c, requires unrelieved residual stresses in the rock to be addressed. This is not addressed in FSAR 2.5.1.3.SRP 2.5.1, Paragraph II.4.d, requires a discus sion on unstable materials or areas considered unstable due to physical properties. These are not discussed in FSAR 2.5.1.B.Justification for differences from SRP1.Major alteration zones, irregular weat hering, and structur al weakness do not exist at the site.2.There is no history of stress relief pr oblems in the area and there were none evident during excavations.3.Unstable materials and areas considered unstable due to physical properties are discussed in FSAR Sections 2.5.4 and 2.5.5.
MPS3 UFSAR1.9-14Rev. 30 SRP 2.5.2SRP TITLE: VIBRATORY GROUND MOTIONA.Actual differences between FSAR and SRP1.Magnitudes of earthquakes shown in FSAR Tables 2.5.2-3 and 2.5.2-4 are not identified as M b , M L , or M s as specified in SRP 2.5.2, Paragraph II(2.5.2.1).2.FSAR Section 2.5.2.4 does not give a probabi listic determination of earthquake return period for the largest earthquakes in each province as specified in SRP 2.5.2, Paragraph II(2.5.2.4).3.FSAR Section 2.5.2.7 does not give an estima te for the probability of exceeding the acceleration level of the 0BE during the 40-year operating life of the plant as specified in SRP 2.5.2, Paragraph II(2.5.2.7).B.Justification for differences from SRP1.The magnitude of most earthquakes is based on the relationship between empirical intensity and magnit ude. It is not possible to be precise in terms of M b or M L for historical earthquakes. None of the magnitudes listed in the FSAR text are surface wave magnitudes (M s).2.Refer to the Applicant's response to the NRC Acceptance Review Request Number 230.1.3.Refer to the Applicant's response to the NRC Acceptance Review Request Number 230.2.
SRP 2.5.3SRP TITLE: SURFACE FAULTINGA.Actual differences between FSAR and SRP1.FSAR Section 2.5.3.1: A detailed offshor e geologic investigation of 5-mile radius has not been attempted as required in SRP 2.5.3, Paragraph II(2.5.3.1).2.FSAR Section 2.5.3.4: The loca tion and investigation of every fault within 5 miles of the site has not been perf ormed as required in SRP 2.5.3, Paragraph II(2.5.3.4).3.FSAR Section 2.5.3.6: Age documentation of every fault within the 5-mile radius has not been performed as requi red in SRP 2.5.3, Paragraph II(2.5.3.6).TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-15Rev. 30B.Justification for differences from SRP1.A study of the regional and site geology (on land) was performed, as discussed in FSAR Sections 2.5.1 and 2.5.3. This incl uded extensive on site geologic mapping and age determinations of faults on site. Because the on site geologic investigation did not reveal any recent faulting or unusual features, an offshore geologic investigation was not considered necessary.2.FSAR Section 2.5.3.4: The Millstone geologic study shows that the last period of faulting at the site occurred approximately 142 milli on years ago and was related to Triassic-Jurassic rifting or older events. According to the United States Geologic Survey (USGS) geological maps of the area, faults outside the site but within the 5 mile radius, would also be associated with these periods of tectonism. Therefore, locating and identify ing every fault within a 5-mile radius of the site was not considered necessary.3.Because of the above justifications, no detailed offshore ge ologic investigation within a 5 mile radius was considered necessary, nor was it considered necessary to investigate and determine the age of every fault within 5 miles of the site.
SRP 2.5.4SRP TITLE: STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONSA.Actual difference between FSAR and SRP SRP 2.5.4, Paragraph II(2.5.4.2), asks for presenta tion of a table listing the values of parameters used in the analyses of the following properties for saturated soils and clays that underlie the site:
- Liquification potential
- Consolidation behavior
- Static and dynamic behavior FSAR Section 2.5.4.2 does not contain a specific tabl e listing values of parameters used to perform the analyses mentioned above.B.Justification for difference from SRP Although the FSAR does not contain a specific tabl e listing values of parameters used to perform the required analyses, the data is available in applicable s ubsections of Section 2.5.4 as listed below:TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-16Rev. 30
- Liquification potential - FSAR Section 2.5.4.8 and its subsections contain values of parameters used in evaluating liqui fication potential of the underlying soils beneath Category I structures.
- Consolidation behavior - No clay soil s are present at the site. The method of analysis to determine elastic settleme nt of Category I structures on bedrock, glacial till and structural backfill is contained in FSAR Subsection 2.5.4.10.2. Values used in the analysis to dete rmine settlement of underlying soils are contained in FSAR Subsection 2.5.4.4.3.
- Static and dynamic behavior - Values of parameters used for evaluating static and dynamic behavior of underlyi ng soils beneath Category I structures are contained in FSAR Sections 2.5.4.2.5, 2.5.4.2.6, and 2.5.4.5.2.
SRP 3.2.2SRP TITLE: SYSTEM QUALITY GROUP CLASSIFICATION A.Actual difference between FSAR and SRP SRP 3.2.2, Subsection II, utilizes Regulatory Gu ide 1.26 as the acceptance criteria for defining the Quality Groups for components important to safety. In lieu of this Regulatory Guide, Millstone 3 utilizes the classification system provi ded in ANS 18.2. FSAR Section
3.2.2 provides
a cross reference between the ANS safety cla ssifications and the Quality Groups defined in Regulatory Guide 1.26.B.Justification for difference from SRPThe ANS classification system implemented for Millstone 3 has been endorsed by industry as an acceptable alternative to Regulatory Guide 1.26. This classification system has been used on many other plants and has been accepted by the NRC's Mechanical Engineering Branch. Additionally, SRP 3.2.2, Subsection III, indicates that the NRC will accept alternatives to the Regulator y Guide 1.26 Quality Group clas sification system provided a correlation between Qua lity Groups and the clas sification system used by the applicant is provided in the FSAR. As noted above, such a correlation has been provided in Section 3.2.2 of the FSAR.
SRP 3.4.1 SRP TITLE: FLOOD PROTECTIONA.Actual difference between FSAR and SRP SRP 3.4.1, Paragraph III.3, discusses the revi ew of postulated failure of nonseismic Category I and nontornado protected tanks. FSAR Section 3.4.1 does not address the postulated rupture effects of these tanks.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-17Rev. 30B.Justification for difference from SRP Millstone 3 does not have QA Category I tanks that are nonseismic. Po stulated failure of nontornado protected tanks has been considered during the review of moderate energy lines as described in FSAR Section
3.6. Items
in this category are lo cated outside safety related structures in areas that would preclude flooding of safety related equipment. Non-QA Category I tanks within safety related structures ar e not considered to contain sufficient inventory to cause flooding of sa fety related equipment. In addition, safety related equipment required for safe shutdown of the plant is located in cubicles, or on elevated platforms which would preclude da mage due to potential flooding that would result if the non-QA Category I tanks were postulated to fail in a seismic event.
SRP 3.4.2SRP TITLE: ANALYSIS PROCEDURESA.Actual difference between FSAR and SRP SRP 3.4.2, Paragraph II.1, requires that the design basis flood (DBF) or the highest groundwater level and the associated dynamic effects, if any, used in the design shall be the most severe ones that have b een historically reported for the site. FSAR 3.4.2 states that structures located above the DB F level are designed for the hydrostatic effects of uplift and water pressure resulting from the DBF or normal groundwater , whichever is most severe.B.Justification for difference from SRP FSAR Section 2.5.4.6 describes th e groundwater conditions for the Millstone 3 site and includes a description of the low permeabilit y of the bedrock as well as the overlying glacial till. Because of the low permeability of these materials at th e site the groundwater level would not significantly change. Ther efore, the normal groundwater level and its associated dynamic effects are sufficient fo r the design of the foundations of the site structures.
Figure 2.5.4-37 shows the map of the stabilized groundwater level contours that were used as the basis for determining the hydrosta tic loading on the structure foundations.
SRP 3.5.1.3
SRP TITLE: TURBINE MISSILESA.Actual difference between FSAR and SRP SRP 3.5.1.3, Subsection II, requires missile protection for components needed to maintain the reactor in cold shutdown. Cold shutdown maintenance components are not specified as targets in the FSAR.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-18Rev. 30B.Justification for difference from SRPEvaluation of turbine missiles, as described in FSAR Section 3.5.1.3, has concluded that the probability of a turbine missile being genera ted and causing damage to a safety related system or component is lower than what is recommended in Regulatory Guide 1.115.
SRP 3.5.1.5 SRP TITLE: SITE PROXIMITY MISSILES (EXCEPT AIRCRAFT)A.Actual difference between FSAR and SRP SRP 3.5.1.5, Paragraph III.3, states that the definition of P p includes probability of missiles that produce secondary missiles which coul d damage vital equipment. This was not considered in the FSAR analysis.B.Justification for difference from SRPFSAR Section 2.2.3 provides the analysis of site proximity missiles (except aircraft). Since the probability stated in this section is sufficiently below the acceptance criteria, consideration of secondary missiles was not deemed necessary. Inclusion of secondary missiles (P p = 1) provides a total probability which is still within the acceptance criteria provided in the SRP.
SRP 3.5.1.6 SRP TITLE: AIRCRAFT HAZARDS A.Actual difference between FSAR and SRP SRP 3.5.1.6, Paragraph III.2, indicates an inflight crash rate of 4 x 10
-10 per year. FSAR analysis uses the old SRP crash rate of 3 x 10
-9.B.Justification for differences from SRPA response to this difference has been pr ovided in the Applicant's response to NRC Acceptance Review Request Number 311.4.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-19Rev. 30 SRP 3.6.1SRP TITLE:PLANT DESIGN FOR PROTECTION AGAINST POSTULATED PIPING FAILURES IN FLUID SYSTEMS OUTSIDE CONTAINMENTA.Actual differences between FSAR and SRP1.BTP ASB 3-1, B.1.a(1), requires an arbi trary split be postulated on the main steam and the feedwater systems at a lo cation proximate to essential systems.
The split must be postulated regardle ss of whether the break exclusion requirements of BTP MEB 3-1, Item B.
1.6 are met. The FSAR does not commit to postulate this split.2.BTP ASB 3-1, B.1.a(2), states that main steam or feedwater piping should not be routed in the vicinity of the control r oom. The FSAR states that the main steam and feedwater pipes are routed in the vicinity of the control room.3.BTP ASB 3.1, B.2.a states that essential systems and components should be designed to meet the seismic design cr iteria of Regulatory Guide 1.29. The BTB defines essential systems and components as those required to shut down the reactor and mitigate the c onsequences of a postulated piping failure without off site power. FSAR Section 3.6.1.3.1 identifies two high-energy line break isolation systems for the auxiliary steam and hot water heating systems where the isolation valves are in a nonseismic piping system in a nonseismic area.B.Justification for differences from SRP1.Essential systems, components, or stru ctures are not located within the main steam or feedwater containment penetration area. Environmental effects are of no consequence. The design basis for environmental effects in these areas is given in FSAR Section 3.11, Appendix B.2.Pipe rupture restraints are provided to prevent main steam pipe whip into the control building wall. The feedwater pipe does not impact the control building in accordance with the disc ussion in FSAR Section 3.6.1.3.3.3.Redundant isolation capability is provi ded for both systems with Category 1E qualified detection and actu ation devices located in the Auxiliary Building. In order to provide optimal isolation capability, the isolation valves are located in nonseismic buildings. These valves are not fully seismically qualified. For a nonseismic system in a nonseismic building, seismic qualification of a component in that system is not feasible. However, location of the isolation valves in the nonseismic building is the only practical manner in which to provide the isolation function.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-20Rev. 30 Given a pipe break of the auxi liary steam or hot water heating systems in a safety related building, isolation of the affected system is capable assuming th e most limiting single failure. The isolation valves are normally open and fail in the closed position. To ensure continued isolation capability following th e postulated pipe break, plant operating procedures require manual valves to be closed to isolate the affected piping.
SRP 3.6.2SRP TITLE:DETERMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPINGA. Actual differences between FSAR and SRP1.SRP 3.6.2, Paragraph III.2.a, states that pressure and temperature values should correspond to the greater contained ener gy at hot standby or at 102 percent power. FSAR Section 3.6.2.2.1 states that pressure and temperature values associated with reac tor operation at 100 percent power are used.2.SRP 3.6.2, Paragraph III.2.a, states that the allowable capacity of crushable material shall be limited to 80 percent of its rated energy dissipating capacity as determined by dynamic testing at loading rates within
+/-50 percent of the specified design loading rate. FSAR Section 3.6.2.2.1 commits to 80 percent of energy absorbing capacity but does not co mmit to dynamic tes ting to determine energy absorbing capacity.3.BTP MEB 3-1, B.1.e, states particular criteria for postu lating through wall leakage cracks in high energy piping. FSAR Section 3.6.2.1.2 does not commit to postulate through wall leakage cracks in high energy piping.B.Justification for differences from SRP1.The FSAR commitment includes the hot standby mode of normal operation. The differences in internal pressure and temperature between 102 percent power and 100 percent power are not significant. 2.Dynamic effects on crushable energy abso rbing material are not significant. The application is absorbing kinetic energy fr om pipe whip throu gh relatively small distances. The impact velocities are small so that energy absorbing capacity based on static test data is acceptable.3.High energy line pipe breaks are more limiting environmentally than high energy through wall leakage cracks in any area where essential systems are located.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-21Rev. 30 SRP 3.7.2SRP TITLE: SEISMIC SYSTEM ANALYSISA.Actual differences between FSAR and SRP1.SRP 3.7.2, Paragraph II.4, requires cons ideration of and an envelope of responses from finite element and half space representations of subgrade soil stiffness. FSAR Section 3.7.2 does not address half space representation.2.SRP 3.7.2, Paragraph II.11, requires cons ideration of accidental torsion by assuming an additional seismicity of 5 percent of the maximum building dimension at the level unde r consideration. This is not addressed in the FSAR.B.Justification for differences from SRP1.Millstone 3 has committed to use finite element representation of soil stiffness at the CP Stage as described in PSAR Section 3.7.1. Studies that have been conducted on the only structure which is completely soil-founded, the emergency generator enclosure, indicate that the finite element results provided more severe results than the half space representation.2.Millstone 3 designs were fi nalized prior to the development of this SRP criteria.
Guidance found in 10 CFR 50 regulation does not require continuous upgrading of the codes and standards used in the design.
SRP 3.7.3 SRP TITLE: SEISMIC SUBSYSTEM ANALYSISA.Actual difference between FSAR and SRP (BOP Scope)FSAR Section 3.7B.3 does not describe the seismic analysis procedures used to account for the seismic motion of non-Category I systems in the seismic design of Category I systems as specified in SRP 3.7.3, Paragraph II.2.l(1).
The FSAR currently describes only the seismic analysis procedures used to account for the seismic motion of non-Category I piping in the seismic design of Category I piping.B.Justification for difference from SRP Additional information will be provide d in an amendment to the FSAR.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-22Rev. 30 SRP 3.7.3SRP TITLE: SEISMIC SUBSYSTEM ANALYSISA.Actual difference between FSAR and SRP (NSSS Scope)
SRP 3.7.3, Paragraph II.2.g, requires the closely spaced modes be combined in accordance with Regulatory Guide 1.92. Westinghouse combines closely spaced modes in accordance with the methods describe d in FSAR Section 3.7N.3.7.B.Justification for difference from SRPThe Westinghouse methods for combining closely spaced modes represent an alternative to Regulatory Guide 1.92 which has been accepted by the NRC's Structural Engineering Branch and Mechanical Engineering Branch on specific plant dockets. The Westinghouse position on combining closely spaced modes has been accepted on the Seabrook, Catawaba, SNUPPS, Byron, and Comanche Peak dockets.
SRP 3.8.1 SRP TITLE: CONCRETE CONTAINMENTA.Actual differences between FSAR and SRP1.FSAR Section 3.8.1 does not reference Re gulatory Guide 1.136 as specified in SRP 3.8.1, Paragraph II.2. 2.FSAR Section 3.8.1 does not discuss the ultimate capacity of the reactor containment with respect to failure m odes as described in SRP 3.8.1, Paragraph II.4.j.3.Millstone 3 did not use Article 3000 of ASME III, Division 2, for loads, load combinations, and stress allowables as described in SRP 3.8.1, Paragraph II.5.4.Millstone 3 did not use ASME III, Di vision 2, Article CC-3000 for the analysis and design of the containment structure tangential shear as described in SRP 3.8.1, Paragraph II.4.f.B.Justification for differences from SRP1.Regulatory Guide 1.136 does not apply to Millstone 3. See FSAR Section 1.8 for position on Regulatory Guide 1.136.2.The ultimate capacity of the reactor c ontainment with respect to failure modes has been considered in the PRA study, wh ich will be submitted as a separate report.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-23Rev. 303.ASME III, Division 2, wa s not available at the time of the Millstone 3 Construction Permit. ACI 318 and AISC-1969 Ed. were the codes used. ASME III, 1971 Ed., with Addenda through Su mmer 1973, Subsections NC and NE were used as a guide. Guidance found in 10 CFR 50 regulation does not require continuous upgrading of the codes and standards used in the design.4. ASME III, Division 2, wa s not available at the time of the Millstone 3 Construction Permit. The procedure used for analysis and design of the containment structure tangential shear , as described in FSAR Section 3.8.1.4.1, meets the intent of SRP Section 3.8.1, Paragraph II.4.f.
SRP 3.8.3 SRP TITLE:CONCRETE AND STEEL INTER NAL STRUCTURES OF STEEL OR CONCRETE CONTAINMENTSA.Actual difference between FSAR and SRP ACI 349-76 was not used as desc ribed in SRP 3.8.3, Paragraph II.2.B.Justification for difference from SRPThis code was not in effect at the time of the Construc tion Permit. ACI 318, AISC-1969 Ed.
and ASME III 1971 Ed. through Summer 1973 a ddenda were the codes used. Guidance found in 10 CFR 50 regulation does not requir e continuous upgrading of the codes and standards used in the design.
SRP 3.8.4SRP TITLE: OTHER SEISMIC CATEGORY I STRUCTURESA.Actual differences between FSAR and SRP1.ACI 349-76 was not used during the design stage of Millstone 3 as described in SRP 3.8.4, Paragraph II.2.2.SRP 3.8.4, Paragraph II.4.d, addresses the use of the design report format presented in Appendix C to this SRP. Th e Applicant's design information is not in this format.B.Justification for differences from SRP1.The ASME III 1971 Ed. through Summer 1973 addenda and AISC-1969 Ed. codes were in effect during the design stage of Millstone 3. Guidance found in 10 CFR 50 regulation does not require c ontinuous upgrading of the codes and standards used in the design.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-24Rev. 302.The material described in Appendix C of this SRP can be found in the design criteria and design calculations which are contained in an auditable file located at the Millstone 3 site.
SRP 3.8.5SRP TITLE: FOUNDATIONSA.Actual difference between FSAR and SRP ACI 318-71 was used rather than ACI 349-76 as specified in SRP 3.8.5, Paragraph II.4.b.B.Justification for difference from SRPACI 349-76 was not in effect at the time th e construction permit was issued. Guidance found in 10 CFR 50 regulation does not requir e continuous upgrading of the codes and standards used in the design.
SRP 3.9.1SRP TITLE: SPECIAL TOPICS FOR MECHANICAL COMPONENTSA.Actual differences between FSAR and SRP (BOP Scope)1.FSAR Section 3.9B.1.1 identifies plant conditions as normal, upset, emergency, and faulted, whereas SRP 3.9.1, Pa ragraph III.1, requires them to be identified as Design Level A, B, C, and D.
Also, allowables used in stress anal ysis are not based on service limits.2.SRP 3.9.1, Paragraph III.4, requires the FSAR to include justifications as well as the demonstration of acceptability of stress strain curves employed.However, FSAR Section 3.9B.1.4 only describes the methods and the extent to which these methods are employed in the stress analysis of components, and references ASME Section III provisions.B.Justification for differences from SRP1.Millstone 3 design is based on ASME III 1971 Ed., which defined plant conditions as normal, upset, emergency and faulted as opposed to Level A, B, C, and D.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-25Rev. 302.The piping and components associated with reactor coolant pressure boundary are designed, analyzed, and built to the requirements of ASME Section III. The rigorous requirements of ASME Section I II to be satisfied for Code Class 1 systems and components ensure that the requirements of SRP 3.9.1, Paragraph III.4, are met. For noncode components, FSAR Section 3.9B.1.4.5 refers to the analytical procedures used for ASME code components noted in FSAR Section 3.7B.3.1.1.
SRP 3.9.1SRP TITLE: SPECIAL TOPICS FOR MECHANICAL COMPONENTSA.Actual difference between FSAR and SRP (NSSS Scope)
SRP 3.9.1, Paragraph II.2, requires a consider able amount of information for all the computer codes used in th e design and analysis of Se ismic Category I components. Westinghouse only provides a br ief description of the computer codes used by Westinghouse for component design and anal ysis in FSAR Section 3.9N.1.2. Additional information required by the SRP for the com puter codes referred to in FSAR Section 3.9N.1.2 is provided by reference to WC APs-8252 and 8929. Computer codes used by Westinghouse vendors are not included in the FSAR.B.Justification for difference from SRP The information requested by the SRP for referenced Westinghouse computer codes is provided in WCAPs-8252 and 8929. Both of these documents have been submitted to the NRC for review. WCAP-8252 has been a pproved by the NRC, and WCAP-8929 is currently under review by Oak Ridge National Laboratory.Vendor computer codes are not included in the FSAR because of the large number of codes used and the proprietary nature of this vendor information. Westinghouse assures the acceptability of vendor computer codes through quality assurance audits at vendor facilities (as described in Chapter 17 of the FSAR) a nd the technical review of various design documents submitted by vendors to Westinghouse.
The NRC's Mechanical Engineering Branch ha s interpreted this NRC guideline to be applicable only to computer codes used by th e major contractors (i.e., NSSS supplier/AE).
SRP 3.9.2SRP TITLE:DYNAMIC TESTING AND ANALYSIS OF SYSTEMS, COMPONENTS, AND EQUIPMENTA.Actual differences between FSAR and SRP (BOP Scope)TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-26Rev. 301.FSAR Section 3.9B.2 does not provide a list of snubbers on systems which experience significant thermal expa nsion, as required by SRP 3.9.2, Paragraph II.1.d.FSAR does not provide a description of the tests to be conducted to verify thermal expansion/vibration measurements as desc ribed in SRP 3.9.2, Paragraph II.1.e and f.2.Information required by SRP 3.9.2, Paragr aph II.2, is not co ntained in FSAR Section 3.9B.2B.Justification for differences from SRP1.The Technical Specifications will provide details on snubber testing and a list of safety related snubbers.2.Information requested is provided in FSAR Section 3.7B.3.
SRP 3.9.2SRP TITLE:DYNAMIC ANALYSIS AND TESTI NG OF STRUCTURES, SYSTEMS, AND EQUIPMENTA.Actual difference between FSAR and SRP (NSSS Scope)
SRP 3.9.2, Paragraph II.2.e, defines criteria for combining closely spaced modes. The Westinghouse method for combining closely spac ed modes is provide d in FSAR Section 3.7N.3.7.B.Justification for difference from SRPThe Westinghouse methods for combining closely spaced modes represent an alternative to Regulatory Guide 1.92 which has been accepted by the NRC's Structural Engineering Branch and Mechanical Engineering Branch on specific plant dockets. The Westinghouse position on combining closely spaced modes has been accepted on the Seabrook, Catawaba, SNUPPS, Byron, and Comanche Peak dockets.
SRP 3.9.3 SRP TITLE:ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURESA.Actual differences between FSAR and SRP (BOP Scope)1.FSAR Section 3.9B.3.1 does no t reflect the stress limi t criteria of SRP 3.9.3, Paragraph II.1, and Appendix A.2.SRP 3.9.3, Paragraph II.2, requires inform ation on Class 3 safe ty/relief devices along with Classes 1 and 2. As written, the FSAR does not specifically address Class 3 safety and relief devices.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-27Rev. 303.The requirements of SRP 3.9.3, Paragraph II.3, for snubbers have been fulfilled but have not been included here.B.Justification for differences from SRP1.The allowable values utilized by M illstone 3 for piping systems meet the requirements stipulated in ASME S ection III, 1971 Ed. through Summer 1972 addenda.Components, except piping, use stress cr iteria of ASME Section III, 1974 Ed., and Code Cases 1606, 1607, 1635, and 1636. These code cases were approved for use in Regulatory Guide 1.84.2.The design and analysis requirements of Class 3 safety/pre ssure relief devices are the same as those of Class 2 as described in FSAR Section 3.9B.3.3. 3.Information on snubbers will be discussed in FSAR Chapter 16, Technical Specifications.
SRP 3.9.3SRP TITLE:ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTUREA.Actual differences between FSAR and SRP (NSSS Scope)1.SRP 3.9.3, Paragraph II.1, requires that de sign criteria for internal parts of components, such as valve discs and pump shafts, be provided. Westinghouse does not provide criteria for the nonpr essure boundary portions of ASME Code Class 1, 2, and 3 components in the FSAR.2.SRP 3.9.3, Appendix A, Paragraph 1.3.3, de fines the design basis pipe break (DBPB) as an emergency condition.
For ASME Code Class 1, 2, and 3 components and component supports, We stinghouse defines the DBPB as a faulted condition (see loading comb ination tables in Section 3.9N).3.SRP 3.9.3, Appendix A, Paragraph 3.1, requi res that stress limits and loading combinations be provided for core support structures. The FSAR does not currently provide tables defining load co mbinations and stress limits for core support structures.B.Justification of differences from SRP1.Westinghouse does not consider it appropriate to provide this type of detail in the FSAR. Westinghouse employs good engin eering practice in defining design criteria for critical internal parts. Generally, for critical internal parts of components, equipment specifi cations limit stresses to th e criteria defined in the ASME Code or to the yield strength of the material.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-28Rev. 302.Westinghouse defines the DBPB as a faul ted condition event consistent with the criteria defined in ANS 18.2. Additionally, Westinghouse considers the stress limits and analysis methods for faulte d conditions defined in the ASME Code and FSAR Section 3.9N to be sufficiently conservative to assure the structural integrity and operability of components when subjected to faulted condition loads including the DBPB.3.A response to this difference will be provided in an amendment to the FSAR.
SRP 3.10SRP TITLE:SEISMIC AND DYNAMIC QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENTA.Actual differences between FSAR and SRP (BOP Scope)1.SRP 3.10, Paragraph II.1.a(14)(b)iii and iv, requires an analysis of LOCA-induced hydraulic forcing functions or differential pressures upon valve discs or pump rotors. FSAR Section 3.10 does not contain this analysis.2.Seismic Qualifica tion Report format, as specifie d in SRP 3.10, Paragraph II.5.c, was not used to document seismic qualification.3.Mechanical equipment seismic and operabi lity qualification is not addressed in FSAR Section 3.10B, as specifi ed in SRP 3.10, Subsection II.4.SRP 3.10, Paragraph II.1, requires a combinat ion of test and analysis to verify the operability of pumps and valves during all plant ope rational conditions.
Pump operability has only be en performed by analysis.5.SRP 3.10, Paragraph II.5.b, requires a list of systems necessary to perform the functions outlined in SRP 3.10. This list is not included in FSAR Section 3.10B.B.Justification of differences from SRP1.Components and equipment within the system where a LOCA occurs are considered to be rendered inoperable after this event. Valves adjacent to break exclusion areas are supported and/or rest rained to maintain stresses within allowable limits to assure operability.2.Controlled seismic qualification files containing equivalent information are maintained.3.Seismic qualification of mechanical eq uipment is described in FSAR Sections 3.7B and 3.9B. Pump and va lve operability qualificati on is discussed in FSAR Section 3.9B.3.2.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-29Rev. 304.The seismic and operabili ty qualification programs de scribed in FSAR Section 3.9B.3.2 provide adequate assurance of proper equi pment performance under all required conditions.5. Safety related mechanical and electric systems are listed in FSAR Table 3.2-1 and described in FSAR Chapters 6 and 7.
SRP 3.10SRP TITLE:SEISMIC AND DYNAMIC QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENTA.Actual differences between FSAR and SRP (NSSS Scope)1.FSAR Section 3.10N currently addresse s only Category I in strumentation and electrical equipment. Mechanical equipment seismic and operability qualification is not discussed in Section 3.10N as specified in SRP 3.10, Subsection II.2.SRP 3.10, Paragraph II.1.a(2), requires that equipment should be tested in the operational condition and that loadi ngs simulating normal plant conditions should be superimposed on seismic a nd dynamic loads. This includes flow induced loads and degraded flow conditions. For the tests performed by Westinghouse, operational conditions are in cluded where practical, simulated in some manner, or addresse d by analysis. Flow load s are not superimposed on seismic loads for valve operability tests.3.SRP 3.10, Paragraph II.1.a(8), requires th at fixture design for seismic tests should simulate actual service mounting and should not cause any extraneous dynamic coupling to the test item. West inghouse seismic qual ification testing configurations are designed to represent the typical plant installation for the tested component.4.If the dynamic testing of a pump or valve is impractic al, static testing of the assembly is acceptable if conducted in accordance with SRP 3.10, Paragraph II.1.a(10). However, end loadings are not applied and all dynamic amplification effects are not included in the static deflection tests for active valves.5.FSAR Section 3.10N does not cover operabili ty of active pumps and valves as specified in SRP 3.10, Paragraph II.1.a(14)(a).TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-30Rev. 306.FSAR Section 3.9N discusses the appr oach for seismic and operability qualification of safety related mech anical equipment. SRP 3.10, Paragraph II.1.a(14)(a), only allows analysis to be used for demonstrating structural integrity. It further states that operabi lity of active pumps and valves must be demonstrated by test or a combinati on of test and analysis. The Millstone program utilizes a combination of test and analysis to demonstrate operability for active valves.7.SRP 3.10, Paragraph II.1.a(14)(a), require s that all complex active components other than pump and valve bodies (simple and passive elements) should be tested for operability. The Millstone 3 valve ope rability program utilizes tests or a combination of test and analysis to demonstrate operability, but no specific tests are done on valve gate, disc assemblies, motors, etc.8.SRP 3.10, Paragraph II.1.a(14)(b) viii, requires the use of Regulatory Guide 1.92 for combination of multimodal and mult idirectional responses in analyses. Westinghouse utilizes the methods define d in FSAR Section 3.7 for combining closely spaced modes.9.SRP 3.10, Paragraph II.1.b(3), states th at supports should be tested with equipment installed or with dummy weights installed. Where seismic testing is conducted, the equipment is mounted as it is installed in the plant (this includes supports). However, seismic testi ng of all supports is not conducted.10.SRP 3.10, Paragraph II.1.c, requires th at seismic and dynamic testing be performed in sequence in accordance with IEEE 323-1974. For some mechanical components, aging and sequence testing wa s not included as part of the seismic and operability testing. 11.SRP 3.10, Paragraph II.3, spells out requi rements for central files that are not addressed in FSAR Section 3.10.12.SRP 3.10, Paragraph II.5.b(1), requires a list of systems necessary to perform the functions outlined in SRP 3.10, Subsection I, be included in FSAR Section 3.10.
This list is not included in FSAR Section 3.10N.13.SRP 3.10, Paragraph II.5.b(2), requires that a description of the results of any in-plant tests used to confirm qualification of equipment be included in the FSAR.
Actual test results are not included in the FSAR.14.SRP 3.10, Paragraph II.5.c, requires a se ismic qualification report. Westinghouse does not maintain such a report for Millstone 3.B.Justification for differences from SRPTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-31Rev. 301.The guidance provided in NUREG-0800 doe s not conform to the guidelines of Regulatory Guide 1.70 (Standard Format and Content Guide). The Millstone 3 FSAR covers seismic qualification for safety related mechanical equipment in Sections 3.9N and 3.7N. Valve operabili ty is addressed in FSAR Section 3.9N.3.2.2.Full operational testing conditions are not included in testing because performing such a test is impractical. For example, when static deflection tests on valves are performed, the P across the valve disc is simulated. However, the test is not performed with the valve in a flow loop. As stated above, Westinghouse addresses operational cond itions other than by test.The active valve operability program defined in FSAR Se ction 3.9N.3.2 outlines the program for demonstrating operabili ty under all required plant conditions.
This program of conservative design, an alysis, and test provides adequate assurance that safety related equipment will perform the required safety functions under the appropriate plant conditions.3.Interface requirements are defined based on the test configuration and other design requirements. Installation is then completed in accordance with the component interface requirements. Any dynamic coupling effects that result from mounting the component in accordan ce with these inte rface criteria would have been adequately consid ered during the test program.4.Westinghouse places conservative restri ctions on the allowable piping loads transmitted to the valve or pump body such that these loads cannot cause
detrimental deflections of the active components. This restriction of allowable piping loads combined with the static deflection testing performed on active valves provides adequate assurance of valve performance and obviates the need to apply end loadings during the static deflection tests for active valves.The Westinghouse operability program for active valves addresses dynamic amplification effects by increasing the g loadings utilized in static deflection tests and analyses when dynamic equipment res ponse is a concern. In most cases, the equipment is rigid and does not displa y dynamic amplificati on characteristics.5.The Millstone 3 operability program is covered in FSAR Section 3.9N.3.2. The latest version of the SRP has includ ed operability under Section 3.10 and has deleted it from Section 3.9.6.Programs currently in effect for Millst one 3 utilize analysis for demonstrating operability of active mechanical equipment such as check valves. For some components (valves with extended struct ures, etc.), static deflection test programs are utilized in combination with analysis to demonstrate operability.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-32Rev. 30 The programs defined for active electrical equipment comply with the guidelines outlined in IEEE 344-197 Regulatory Guide 1.100 and provide adequate
assurance of operability under all require d conditions. For active valves, the present operability programs confir m the conservative design of these components and provide adequate assurance that these devices can perform their safety function under all required conditions.7.The program defined in FSAR Secti on 3.9N.3.2 provides adequate assurance through the procedures employed that acti ve valves will perform their required safety function under all required conditions.8.The Westinghouse methods for combin ing closely spaced modes has been previously justified and accepted by the NRC. 9.The supports of safety related equipment are adequately qualified utilizing test or analysis procedures.10.For electrical equipment discussed in WCAP-8587, the guidelines provided in IEEE 344-1975 and 323-1974 were followed during the qualification program.
For mechanical equipment, there are currently no official guidelines that dictate requirements for aging or sequence te sting. The seismic and operability qualification programs implemented for Mi llstone 3 provide ad equate assurance of proper equipment performa nce under all required conditions.11.Westinghouse qualification documentation is maintained for the 40-year design life in engineering files at Westinghouse.
These records are f iled and stored in accordance with 10 CFR 50 Appendix B and Regulatory Guide 1.88 as defined in FSAR Chapter 17. Seismic informati on for the NSSS supplied Class 1E equipment qualified under the WCAP-8587 program is contained the Equipment Qualification Data Packages and Test Reports which Millstone 3 maintains in their central file.12.The information requested on safety related mechanical and electrical systems is included in FSAR Chapters 6 and 7.13.The test results for Westinghouse suppl ied equipment are referenced in the FSAR.14.Seismic qualification of equipment is documented in test reports, analysis reports, calculations, etc., contained in Westinghouse files. The documentation maintained by Westinghouse satisfies existing regulatory requirements and, therefore, it is not considered neces sary to prepare an additional Seismic Qualification Report.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-33Rev. 30SRP 3.11SRP TITLE:ENVIRONMENTAL QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENTA.Actual differences between FSAR and SRP (BOP Scope)1.Millstone 3 FSAR does not address mechan ical qualification as required in SRP 3.11, Subsection II.2.NUREG-0588 methodologies are not strictly followed as required in SRP 3.11, Subsection II.B.Justification for differences from SRP1.Preparation and submi ttal of information pert aining to environmental qualification of mechanical eq uipment is pending NRC rulemaking.2. A summary comparison of NUREG-0588 wi ll be provided with the Electrical Equipment Qualification Data Packages as a separate report.SRP 3.11SRP TITLE:ENVIRONMENTAL QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENTA.Actual differences between FSAR and SRP (NSSS Scope)1.SRP 3.11, Paragraph I.1, stat es that all mechanical a nd electrical systems and equipment necessary to perform the functions listed in SRP 3.11, Subsection I, should be listed in FSAR Section 3.11. FSAR Section 3.11N does not include this list.2.No reference is made in FSAR Section 3.11N to the results of the NUREG-0588 study as specified in SRP 3.11, Subsection II.3.FSAR Section 3.11N is restricted to electrical equipment only. The environmental qualification of mechanical equipment is not addressed in this section as specified in SRP 3.11.B.Justification for differences from SRP1.The information requested in the SRP is located in different parts of the FSAR as listed below:a. Safety Related Mechanical and Elect rical Systems - FSAR Chapters 6 and 7 b. Active Pumps and Valves - FSAR Tables 3.9N-11 and 3.9N-12TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-34Rev. 30c. Class 1E Generic Components - FSAR Table 3.11N-1
- d. Plant-Specific Class 1E Component s - NUREG-0588 study for Millstone 3
- e. Other safety related mechanical components - FSAR Sections 3.7N and 3.9N2.NUREG-0588 study is not complete.
3.Mechanical equipment qualification for seismic and operability requirements are discussed in FSAR Section 3.9N.
In addition to the tests and analyses discussed in Section 3.9N, Westinghouse designs safety grade mechanical components to accommodate environmental effects through the stringent selection of materials utilized in safety grade mechanical components (e.g., stainless steel, etc.).Soft parts or consumables such as gaskets, seals, and O-rings are selected for use based on their capability to perform in a nuclear application and are maintained through inservice inspection and maintenance programs. In some cases, partial type tests or separate effects tests have been performed to demonstrate adequacy of selected materials or component s for use under adverse environments.
This program for mechanical equipmen t is based on a combination of design qualification tests and analyses and periodic in plant test and maintenance/
surveillance procedures. This program for qualification of safety related mechanical equipment provides adequate assurance that safety grade mechanical components will perform their required functions under all normal, abnormal, accident, and post accident conditions.
SRP 4.5.1 SRP TITLE: CONTROL ROD DRIVE STRUCTURAL MATERIALSA.Actual difference between FSAR and SRP SRP 4.5.1, Paragraph II.4, addresses on site clean ing requirements which are not directly referenced in FSAR Section 4.5.1.B.Justification for difference from SRP Compliance with the cleanliness requirements of Regulatory Guide 1.37 is described in FSAR Section 1.8 and Appendix VII of the SWEC Topical Report for the construction phase referenced in FSAR Section 17.1.2.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-35Rev. 30 SRP 5.2.1.1SRP TITLE: COMPLIANCE WITH THE CODES AND STANDARDS RULE, 10 CFR 50.55aA.Actual differences between FSAR and SRP1.The loop bypass valves are designed a nd fabricated to ASME III1 971 Ed.
through Summer 1972 addenda, whereas 10 CFR 50.55a requires the Winter 1972 addenda.2.SRP 5.2.1.1, Subsection II, indicates the use of Regulatory Guide 1.26 to meet GDC 1 and 10 CFR 50.55a. However, the NSSS (Westinghouse) uses ANS standards rather than Regulatory Guide 1.26.B.Justification for differences from SRP1.Updating the loop bypass valves to a la ter ASME code adde ndum would require additional cost and administrative cost burden without a compensating increase in the level of quality or safety. In addition, the actual hard ware configuration would not be changed by upgradi ng to a later code addendum.2.Components are classified commensura te with the safety function to be performed. FSAR Sections 1.8N and 3.2.2 discuss the Millstone 3 position on Regulatory Guide 1.26.
SRP 5.2.5SRP TITLE: REACTOR COOLANT PRESSURE BOUNDARY LEA KAGE DETECTIONA.Actual difference between FSAR and SRP1.FSAR Section 5.2.5 does not address the frequency of testing of the unidentified leakage sump system as requi red in SRP 5.2.5, Paragraph III.7.B.Justification for difference from SRP1.Testing requirements for the unidentified leakage sump system, will be addressed in the Technical Specifications.
SRP 5.3.1SRP TITLE: REACTOR VESSEL MATERIALSA.Actual difference between FSAR and SRP (NSSS Scope)The tentative capsule removal schedule is not identical to the removal schedule described in 10 CFR 50, Appendix H, II.C.3.b, as requi red in SRP 5.3.1, Paragraph II.6.c(3).B.Justification for difference from SRPTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-36Rev. 30 The schedule described in the FSAR defines more frequent capsule removal in the early years than required by Appendix H and provides adequate standby capsules to meet the Appendix H requirements in later years. This schedule is more conservative than that which is required.
SRP 5.4.1.1SRP TITLE: PUMP FLYWHEEL INTEGRITY (PWR)A.Actual difference between FSAR and SRP (NSSS Scope)
SRP 5.4.1.1, Paragraph II.2.b, states that pump fl ywheel fracture t oughness properties are acceptable if the normal operating temperature is at least 100
°F above the RTNDT. FSAR Section 5.4.1.1.3 states that the RTNDT is no higher than 10
°F.B.Justification for difference from SRP The pump flywheel will see an operating temperature of at least 110
°F once steady state operating conditions have been achieved. In th e actual plant environment, the temperature would likely be higher because of the proximity of heat sources such as the reactor coolant circulated through the pump and attached piping.
SRP 5.4.2.1SRP TITLE: STEAM GENERATOR MATERIALSA.Actual differences between FSAR and SRP (NSSS Scope)1.SRP 5.4.2.1 (BTP MTEB 5-3) requires analysis of free hydroxide. Also required are the reference procedures for chemical analysis. 2.SRP 5.4.2.1 (BTP MTEB 5-3, Paragra ph II.2) discusses a "clean metal" conditions prior to startup. FSAR Sect ion 5.4.2.1 does not discuss this concern.3.SRP 5.4.2.1, Paragraph II.B.2, states that access for tooling to remove sludge by lancing from the tube suppor t plates should be provided.
This is not discussed in FSAR Section 5.4.2.1.B.Justification for differences from SRP1.Free hydroxide will not be analyzed as no additional information on the condition of secondary water chemistry is gained by performing this analysis.
Reference procedures for chemical anal ysis are contained in the Millstone 3 Chemistry Manual.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-37Rev. 302.A small amount of corrosi on product (oxide) retards fu rther run-away corrosion by acting as a barrier to continued corr osion attack. A "metal clean" condition is therefore unduly restrictive a nd no real benefit is derive d from this requirement. Additionally, the interpretation and quality of visual techniques that may be used to identify a "metal clean" condition can be unreliable and, therefore, conflicting and/or inaccurate conclusions may be drawn about the surface condition of steam generator components.3.Sludge lancing is performed on the t op of the tubesheet to remove the accumulation of corrosion products from the bottom of the steam generator. The sludge accumulates in the low flow area on the top of the tubesheet. Sludge particles do not accumulate to such a gr eat degree on the support plates because of support plate flow slots and, in th e case of Model F steam generators, quatrefoil tube support plate holes. Th ese openings allow sludge to filter down (by gravity) to the bottom of the steam generator and accumulate in the low flow area on top of the tubesheet. By select ively placing the blowdown lines in this low flow area of the tubesheet, a large amount of the accumulated sludge can be removed during normal operation of th e blowdown system. Any additional sludge which is not removed from the tubesheet by blowdow n can be removed during an outage by sludge lancing. Access for sludge lancing of the tubesheet is made possible by access ports in the steam generator shell. Sludge lancing from the steam generator shell access ports is considered more effective than sludge removal from the tube support plates would be.
SRP 6.2.1SRP TITLE: CONTAINMENT FUNCTIONAL DESIGNA.Actual differences between FSAR and SRP (BOP Scope)1.Actual margin of external pressu re analysis as required in SRP 6.2.1.1A, Paragraph II.f, is not specifical ly addressed in FSAR Section 6.2.1.1.2.NUREG-0609 was not addressed in FS AR Section 6.2.1.2 as required in SRP 6.2.1.2, Paragraph II.B.2.B.Justification for differences from SRP1.Conservatism of the analysis provides for margin.
2.NUREG-0609 was issued subsequent to th e analysis performed for Millstone 3, which used the NUREG-75/087 SRP cr iteria. FSAR Section 6.2.1.2 describes the methods used to perform the analysis.
SRP 6.2.1SRP TITLE: CONTAINMENT FUNCTIONAL DESIGNTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-38Rev. 30A.Actual differences between FSAR and SRP (NSSS Scope)1.BTP 6-1, Section B.3b, recommends cons ervative condensing heat transfer coefficients which differ from the Westinghouse model.2.SRP 6.2.1.3, Paragraph II.B.3.e, identifies the Westinghouse model cited in Reference 18 of SRP 6.2.1 as acceptable. This model differs from the Westinghouse model actually used in the Millston e 3 containment design.B.Justification for differences from SRP1.It has been determined that West inghouse values for the long-term post-blowdown condensing heat transfer coefficients are nonconservative. However, the SRP guideline (BTP 6-1) for blowdown heat transfer is four times Tagami values. During blowdown, Westinghouse conservatively uses five times Tagami values. Consequently, the Westinghouse evaluation model for ECCS minimum containment pressure, as presented in Appendix A of WCAP-8339 (1974), has been approved by the NRC staff.2.The Westinghouse mass and energy rele ase model for containment design is described in FSAR Sections 6.2.1.3.3, 6.2.1.3.4, 6.2.1.3.5, and 6.2.1.3.6. The
references are listed in FSAR Sect ion 6.2.7. The current FSAR model is under review by the NRC staff.
SRP 6.2.2 SRP TITLE: CONTAINMENT HEAT REMOVAL SYSTEMSA.Actual differences between FSAR and SRP1.SRP 6.5.2, Paragraph II.1.a, requires automatic switchover to re circulation mode.
Millstone 3 switchover is manual.2.SRP 6.5.2, Paragraph II.1.e, requires a pH between 8.5 and 10.5 for fission product control. Millstone 3 is designed to a minimum pH of 7.0.
NOTE: SRP 6.2.2 refers to SRP 6.5.2 for requireme nts for "Heat removal only" spray steam. Otherwise, SRP 6.5.2 (Fission Pr oduct Removal) would not apply to Millstone 3.B.Justification for differences from SRPTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-39Rev. 301.Millstone 3 complies with BTP ICSB 20 which states that manual switchover from injection mode to recirculation mode is sufficient if adequate instrumentation and information display ar e available to the operator so he can make the correct decision at the correct time. Description of operator action considerations during the sw itchover from the injecti on phase to recirculation phase is discussed in FSAR Secti on 6.3.2.8. Instrumentation and controls available to the operator are provided in FSAR Section 7.5.2.The pH value of 8.5 to 10.5 is related to fission product control. Since no credit is taken for the Millstone 3 spray system for this purpose, a design basis minimum pH of 7.0 was chosen based on material considerations.
SRP 6.2.3SRP TITLE: SECONDARY CONT AINMENT FUNCTIONAL DESIGNA.Actual difference between FSAR and SRPFSAR Section 6.2.3 does not provide a discussion of heat transfer analysis and high energy line considerations as specified in SRP 6.2.3, Paragraphs II.D.1 and II.D.2, respectively.B.Justification for difference from SRP Refer to the Applicant's response to the NRC Acceptance Review Request Number 480.4.
SRP 6.2.5SRP TITLE: COMBUSTIBLE GAS CONTROL IN CONTAINMENT A.Actual differences between FSAR and SRP1.SRP 6.2.5, Paragraph II.3, requires a pl ant specific analysis of the mixing characteristics of the containment. FSAR Section 6.2.5 references the analyses of plants with a similar containment design.2.SRP 6.2.5, Paragraph II.11, states that the containment hydrogen monitor shall meet the requirements of NUREG-0737, Item II.F
.1; NUREG-0718; and the Appendix of Regulatory Guide 1.97.B.Justification for differences from SRP1.FSAR Section 6.2.5.3 references the analyses of Surry Power Station, Units 1 and 2 which have a similar containm ent design (FSAR Section 1.3) and for which the USAEC concluded in the Surry 1 and 2 Safety Evaluation Report that there is adequate mixing of hydrogen in the post-LOCA environment.2.By fuel load, Millstone 3 will have implemented hydrogen monitors which will comply with these requirements.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-40Rev. 30 SRP 6.3 SRP TITLE: EMERGENCY CORE COOLING SYSTEMA.Actual difference between FSAR and SRP No reference is made in FSA R Section 6.3 to studies which demonstrate that nonsafety grade interactions cannot exist which could de grade the performance of the ECCS or its supporting systems as specifie d in SRP 6.3, Subsection II.B.Justification for difference from SRPThe effects of failures in non safety related systems due to pipe whip, jet impingement, and/
or adverse environment are provided in FSAR Section 3.6.
SRP 6.4 SRP TITLE: CONTROL ROOM HABITABILITY SYSTEMSA.Actual difference between FSAR and SRP The chlorine detectors are not Seismic Category I nor El ectrical Class 1E (IEEE 323-1974 qualified) as required by SRP 6.4, Paragraph II.5.b. They are re dundant and classified as non-seismic.B.Justification for difference from SRP The redundant chlorine detectors are locate d in a non-harsh environment (i.e., control equipment room, 50-104
°F, 10-60 percent RH, 1.1 x 10 2 Rads). In the event of detector failure, the control room envel ope is automatically isolated.
SRP 6.4 APPENDIX ASRP TITLE:ACCEPTANCE CRITERIA FOR VALVE OR DAMPER REPAIR ALTERNATIVEA.Actual differences between FSAR and SRP1.The air inlet ducting is isolated by two low leakage air operated butterfly valves positioned in series. Since the valves ar e located in series, the arrangement does not meet single failure (active) criteria as described in SRP 6.4, Appendix A, first paragraph.2.SRP 6.4, Appendix A, Item 6, indicate s no manual action cr edit allowed for repairs until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.B.Justification for differences from SRPTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-41Rev. 30As described in FSAR Section 6.4.3, the primary function of the air inlet isolation valves is to isolate the control room, enabling the air bottle pressurization system to pressurize the control room envelope. Followi ng 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of air bottle pressu rization, one air bottle train will be exhausted. At this point , either the standby air bottle system can be used or the outside air pressurization system. Should either air inlet isolation valve fail to open automatically at this time, they are within th e control room habitab ility zone and can be manually opened within 10 minutes. Thus, credit for manually opening the isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> should be acceptable.NOTE:The air bottle pressurization system is no longer credited in radiological accident analyses.SRP 6.5.1SRP TITLE: ESF ATMOSPHERE CLEANUP SYSTEMSA.Actual differences between FSAR and SRP1.SRP 6.5.1, Subsection II, list s Regulatory Guide 1.52 as part of its acceptance criteria. Exceptions have been taken to this Regulatory Guide. See FSAR Section 1.8 for the Millstone 3 comp liance to Regulatory Guide 1.52.2.Continuous indication and recording of air flow for individual ESF filtration units is not provided as desc ribed in SRP 6.5.1, Subsection II.3.Flow sensors are not provided for annunciating high air flows through ESF filtration units as describe d in SRP 6.5.1, Subsection II.4.SRP 6.5.1, Subsection II, lists ANSI N510-1980 as part of its acceptance criteria with respect to in-place testing. The th ree ESF filter train systems listed below will not comply during in-place testing with Section 8.3.1.6 of ANSI N510-1980 which requires the establishment of design flow within
+/-10 percent of system design flow with a system resistan ce corresponding to 1.25 times design dirty filter condition. The three noncomplying filte r train systems are the SLCRS filter banks, the auxiliary building filter banks, and the control room emergency ventilation filter trains.5.SRP 6.5.1, Subsection II, lists ANSI N510-1980 as part of its acceptance criteria with respect to in-place testing. The th ree ESF filter train systems listed below will not comply during in-place testing with Sections 8.3.1.5, 8.3.1.6, and 8.3.1.7 of ANSI N510-1980 which require system fl ow rates to remain constant while the system resistance vari es from clean to 1.25 times dirty filter condition. The three noncomplying filter tr ain systems are the SCLRS filter banks, the fuel building filter banks, and the control room emergency ventilation filter trains.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-42Rev. 30NOTE:The fuel building filter banks are no longer credited as an ESF System in the radiological accident analyses.B.Justification for differences from SRP1.Justifications for the exceptions ta ken to Regulatory Guide 1.52 are found in FSAR Section 1.8.2.Continuous indication or recording of air flows through individual ESF filtration units is not necessary to ensure reliab le system operation. Periodic surveillance tests ensure that system balancing is ad equate to maintain operating flow rates through filtration units within design limitations. Additionally, dp alarm setpoints for each ESF filter train can be set to ensure that during all operating conditions flow is maintained within
+/-10 percent of design flow.3.Failure of system fans to function is sensed by flow sensors which annunciate low air flow conditions in the control room and automa tically start standby units. Sensors to detect high air flow conditions are not necessary since the system is balanced such that the flow is limite d to ensure proper performance of the filtration units. Periodic survei llance tests ensure that sy stem flow rates will not exceed unit design parameters.4.During tests, fans cannot de velop a performance greater than for what they were designed. Fan performance requirement s were based on Regulatory Guide 1.52 as a design document and the referenc ed ANSI N509, which do not, and need not, require the development of design flow rate at 1.25 times dirty filter pressure drop conditions. Dirty filter conditions recommended by the manufacturer are factored into the setpoint calculations of the differential pressure switches for annunciating filter changeout requirements. Thus, fan performance within the 10 percent tolerance beyond the setpoint pressure drop is unnecessary.5.In accordance with fan laws and the laws of fluid flow, as system resistance increases from clean to dirty filter c onditions, the system flow rate decreases unless volumetric capacity controls are incorporated into the design of the system to provide for constant system flow. Regulatory Guide 1.52 as a basic design document and the referenced AN SI N509 do not, and need not, require such controls. The identification of syst em design conditions and the definition of acceptable tolerances of system variables need to be determined by system function on a case-by-case basis as descri bed below for the three ESF systems.SLCRS Filter Banks TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-43Rev. 30 The SLCRS is a standby ESF system design ed to operate foll owing a LOCA. Its function is to drawdown enclosures contiguous to the containment to a minimum negative pressure of 1/4 inch water gaug e in 60 seconds after SIS, and maintain negative pressure conditions for a minimum of 30 days foll owing a LOCA. Flow rates shown on P&IDs are allowances used in design for purposes of equipment and duct sizing; actual flow rates will be determined by test to demonstrate the pressure drawdown characteristics de scribed above. Therefore, the design conditions of the SLCRS are a flow rate to be determined by test, and a system resistance based on dirty filter condition recommended by the filter manufacturer and incorporated into the setpoint calc ulations, the station operating procedures, and the technical specifications. Also, periodic surveillance te sting will ensure that the system operating variables are verified, and if necessary, manually adjusted, to be within the specified limits of the technical specifications.
Fuel Building Filter Banks (This design feature is no longer credited in radiological accident analysis and not an ESF atmospheric cleanup system, the following information is for historical documentation only.)
The fuel building exhaust and filtration sy stem is designed to draw and filter exhaust air during refueling. The function of the filter banks exhaust system is to mitigate the consequences of a fuel handling accident by filtering exhaust air and by preventing uncontrolled outleakage from the fuel building. The design condition of the system is an exhaus t flow rate of 41,360 cfm at a system resistance based on dirty filter condition of 10 inches water gauge. With clean filters the variable inlet vanes on the fans suction will be manually adjusted to provide the design flow rate of 41,360 cfm. This provides an exhaust flow rate of 2,360 cfm in excess of the supply air flow rate.
The fuel building exhaust and filtration system is not credited in the radiological analysis for fuel handling accidents. Control Room Emergency Ventilation Filter Trains The control room emergency ventilat ion system is a standby ESF system designed to be manually started one hour after a LOCA. Its function is to continue to maintain the pressurization of the control r oom habitability zone at a positive pressure of.125 inch water gauge with filtered air after the compressed air bottles have been depleted.
The design condition of the system is 1,000 cfm filtered flow at the system resistance corresponding to dirty filter c ondition of approximately 10 inch water gauge with an outdoor air makeup to recirculation air ratio of 3:1. With clean filter conditions, the system flow rate increases to 1,225 cfm with the same 3:1 ratio of makeup air to recirculation air.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-44Rev. 30 The control room radiological habita bility and the.125 inch water gauge pressurization requirements are both satisfied throughout the entire range of filter train performance from clean to dirty filter condition. Therefore, no volumetric flow control of the pressurization system is required. The system can be balanced at 1,000 cfm
+/-10 percent for dirty filter condition and allowed to increase beyond the +/-10 percent for clean filter conditi on. The dose analysis in Section 15.6.5.4 and Table 15.6-12 is based on 100 percent outside air makeup of 1,000 cfm as a worst case assumption.
SRP 6.5.2SRP TITLE: CONTAINMENT SPRAY AS A FISSION PRODUCT CLEANUP SYSTEMA.Actual difference between FSAR and SRP The containment spray system requirements of SRP 6.5.2 ar e not discussed in FSAR Section 6.5.2.B.Justification for difference from SRP No credit for containment spray fission product removal is taken in the Millstone 3 design.
SRP 7.2 SRP TITLE: REACTOR TRIP SYSTEM (RTS)A.Actual difference between FSAR and SRP The sensors (turbine low trip fluid pressure or all stop valves closed) for reactor trip on turbine trip when power level is 50 percent or more are not seismically qualified as specified in BTP ICSB 26.B.Justification for difference from SRP The sensors are isolated by digital isolators to prevent degrading the reactor trip system.
SRP 7.5SRP TITLE: INFORMATION SYSTEMS IMPORTANT TO SAFETY A.Actual difference between FSAR and SRP The Safety Parameter Display System and the Emergency Response Facilities are not discussed in the FSAR as required by SRP 7.5, Paragraph III.6.B.Justification for difference from SRP As mentioned in FSAR Section 7.5.3, these items are currently being finalized and will be provided in a future amendment.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-45Rev. 30 SRP 8.3.1 SRP TITLE: AC POWER SYSTEMS (ON SITE)A.Actual difference between FSAR and SRP NUREG/CR-0660 is not addressed in the FSA R as required by SRP 8.3.1, Paragraph II.4.f.B.Justification for difference from SRP NUREG/CR-0660 considerations have been a ddressed in responses provided to NRC questions. Refer to the 430 series of questions - Question 430.58 through Question 430.134 for details. This NUREG is only applicable to the emergency diesel engine and its support systems as described in Section 9.5.
SRP 9.1.2 SRP TITLE: SPENT FUEL STORAGEA.Actual difference between FSAR and SRP SRP 9.1.2, Paragraph III.2.e, requires an evalua tion of lighter load drops at maximum heights. This evaluation has not been performed.B.Justification for difference from SRPElectrical interlocks and load paths prevent any load from be ing carried over the spent fuel pool with the new fuel handling crane. Spen t fuel bridge and hoist only carries fuel assemblies at their normal lifting height.
SRP 9.1.3 SRP TITLE: SPENT FUEL POOL COOLING AND CLEANUP SYSTEMA.Actual difference between FSAR and SRP1.Decay heat removal is based on the DECO R computer code (based on ORIGEN2) and credit for evaporative cooling, not BTP ASB 9-2, as required by SRP 9.1.3, Paragraph II.1.d(4).2.The maximum temperature for a normal heat load is 150
°F, a single active failure at 150°F will cause an increase in temperature to approximately 155
°F before cooling is restored, not 140
°F as required by SRP
9.1.3 Paragraph
III.1.d.3.The decay time for the maximum heat load in the spent fuel pool is based on the heat removal capacity of the spent fuel pool h eat exchangers and varies from 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> to 349 hours0.00404 days <br />0.0969 hours <br />5.770503e-4 weeks <br />1.327945e-4 months <br />, not 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> as required by SRP 9.1.3, Paragraph III.1.h(ii).B.Justification for difference from SRPTABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-46Rev. 301. Decay heat removal analysis is based on the DECOR computer code (based on ORIGEN2) and credit for evaporative cool ing in order to get a more accurate value of decay heat loads.2.All SSCs associated with the Spent Fuel Pool have been evaluated and have been found to be acceptable for an increase over the SRP limit of 140°F. The decay heat of the fuel is removed and the water coverage of the fuel is maintained for all anticipated scenarios3.The decay time for the maximum heat load in the spent fuel pool is based on heat removal capacities that are dependent on the actual cooling water temperatures.
Colder cooling water temperatures resu lt in greater heat removal capacities which permit larger heat loads to be placed in the pool and shorter decay times.
SRP 9.1.4 SRP TITLE: LIGHT LOAD HANDLING SYSTEM (RELATED TO REFUELING)A.Actual difference between FSAR and SRP SRP 9.1.4, Paragraph III.6, requires an evalua tion of lighter load drops at maximum heights. This evaluation has not been performed.B.Justification for difference from SRP Electrical interlocks and load paths prevent carrying any load over the spent fuel pool with the new fuel handling crane. Spent fuel bridge and hoist only carry fuel assemblies at their normal lifting height.
SRP 9.2.1SRP TITLE:STATION SERVICE WATER SYSTEM (NUCLEAR SERVICE COOLING WATER SYSTEM)A.Actual difference between FSAR and SRP SRP 9.2.1, Paragraph III.3.d, requires that radi ation monitors be located on system discharge, and at components susceptible to th e leakage, and that these components can be isolated by one automatic and one manual valve in series. There are motor-operated valves at the inlet and discharge of the service water side of the containment re circulation coolers; however, there is no manual valve in series with the motor operated valve.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-47Rev. 30B.Justification for difference from SRPThere is a radiation monitor located on the discharge side of each co ntainment recirculation cooler which alarms in the control room on a hi gh radiation signal. The isolation valves at the inlet and discharge of the coolers are remote-manually operated. When the radiation monitor alarms in the control room, the operator can remotely close the isolation valves. In the event of loss of power to the valve motor operator, a hand wheel may be engaged to locally close the valve. If the valve operato r malfunctions, the containment recirculation coolers may be isolated by cl osing the isolation valves on the containment recirculation system side of the coolers.
SRP 9.2.2 SRP TITLE:REACTOR AUXILIARY COOLING WATER SYSTEMS (COMPONENT COOLING WATER AND AUXILIARY COMPONENT COOLING WATER SYSTEM)A.Actual difference between FSAR and SRP1.The Millstone 3 reactor c oolant pump has not, at this time, been tested to the 20-minute time requirements as sp ecified in SRP 9.2.2, Paragraph II.3.e.B.Justification for difference from SRP1.A program is underway by Westi nghouse to comply with the testing requirements.
SRP 9.4.1 SRP TITLE: CONTROL ROOM AREA VENTILATION SYSTEMA.Actual difference between FSAR and SRP The chlorine detectors are not Seismic Category I nor El ectrical Class 1E (IEEE 323-1974 qualified) as required by SRP 9.4.1, Paragraph II.4. They are redundant and classified as non-seismic.B.Justification for difference from SRP The redundant chlorine detect ors are used and located in a non-harsh environment (i.e., control equipment room, 50-104
°F, 10-60 percent RH, 1.1x10 2 Rads). In the event of detector failure the control room e nvelope is automatically isolated.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-48Rev. 30 SRP 9.4.5 SRP TITLE: ENGINEERED SAFETY FEATURES VENTILATION SYSTEMA.Actual differences between FSAR and SRP1.There are four fresh air intakes for engineered safety features ventilation systems that are not located 20 feet above grade elevation to prevent dust infiltration as required in SRP 9.4.5, Paragraph II.4.2.SRP 9.4.5, Paragraph II.5, states that the to tal ventilation syst em shall have the capacity to detect and control leakag e of airborne contamination from the system. The system presently provides for monitoring of the normal building ventilation and not the emergency ventilation.3.SRP 9.4.5, Paragraph III.3.b, requires essentia l portions of the engineered safety features ventilation systems be protected from the effects of tornados. This SRP requirement also applies to the circ ulating and service water pumphouse and other yard structures ve ntilation system (FSAR Se ction 9.4.8) and the hydrogen recombiner building heating, ventilat ion, and air conditioning system (FSAR Section 9.4.11). Additionally, Regulat ory Guide 1.76 requi res structures, systems, and components important to safety be protected against tornado pressure drop and tornado-generated missiles. To meet these requirements, tornado dampers should be provided to prevent the ductwork from collapsing. Only the control building and emergenc y generator enclosure are protected by tornado dampers.B.Justification for differences from SRP1.The four fresh air intakes are located in the ESF, the emergency diesel generator, and the hydrogen recombiner buildings. The intake in the ESF building, for ventilation of the mechanical equipmen t rooms and auxiliary feedwater pump rooms, has a centerline 15 feet-10 inches above grade. This intake is equipped with a filter rated at 55-60 percent NBS efficiency. Periodic surveillance of the pressure drop across the filter and ch anging the filter upon setpoint alarm annunciation will preclude excessive dust accumulation. Each emergency diesel generator building ventilation intake ha s a centerline 19 feet-6 inches above grade. The intake for hydrogen recombiner cubicle 1B in the hydrogen recombiner building has a centerline 18 feet-8 inches above grade. These intakes are sufficiently close to the recommende d 20 feet above grad e so as to minimize entrainment of dirt or dust.2.The emergency ventilation a/c system ductwork is designed to be of low leakage construction and only recirculates air, thereby, not allo wing a direct path to the mechanical rooms emergency ventilation which utilizes outside air for cooling.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-49Rev. 303.Only those ventilation systems that are required for habitability or heat removal requirements after a DBA or LOP were c onsidered for installation of tornado dampers. Safety related equipment whose ven tilation systems do not have tornado dampers have been qualified for adverse conditions during short-term operation, as described in Section 3.11B. Plant ope rational procedures will address long-term ventilation outages.
SRP BTP CMEB 9.5-1 (SECTION 9.5.1)
SRP TITLE: GUIDELINES FOR FIRE PR OTECTION FOR NUCLEAR POWER PLANTS Refer to the Fire Protection Evaluation Report, Appendix B, for a comparison of Millstone 3 design to BTP CMEB 9.5-1 guidelines.
SRP 9.5.4SRP TITLE:EMERGENCY DIESEL ENGINE FUEL OIL STORAGE AND TRANSFER SYSTEMA.Actual differences between FSAR and SRP1.SRP 9.5.4, Paragraph II.4.b, requires that ea ch diesel generator be capable of operating continuously for 7 days. Each di esel fuel oil tank at Millstone 3 has a 3.5 day capacity of fuel oil.2.There are no tank design features which minimize turbulence of sediments as specified in SRP 9.5.4, Paragraph III.5.3.The fill lines for the diesel generator fuel oil vaults are not missile protected as required by SRP 9.5.4, Paragraph III.6.a.B.Justification for differences from SRP1.The Applicant can obtain fuel oil from sources nearby within 24-hours after a need for such oil is identified. The Applicant also has determined that, for the
power plants within the Northeast Power Coordinating Council area, off site power can be restored to the site 95 percent of the time within a 24-hour period after it is lost. In addition, a loss of of f site power has occurred only once in the 14 years the Millstone switchyard has been in operation. Steps have been taken to preclude occurrence of such a LOP (i t was caused by salt contamination of insulators) in the future. The Applican t also notes that by running two diesel generators at part load (starting 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> s after a postulated accident occurred) and judiciously realigning one train of ES F loads between two operating diesel generators, individual engine operating time per each fuel oil storage tank can be extended to 5-6 days.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-50Rev. 302.Formulation of corrosive product sediment is minimized by means of a sump and a sump pump with suitable controls for removal of condensation. Additionally, the tank interiors are coated with epoxy resin to preclude corrosion. Filters of progressively smaller mesh size, some of which are alarmed for pressure drop, also assure fuel oil sedime nt does not become a problem. 3.Alternate ways to fill the tank are provided through the tank manhole or through the flame arrestor/vent line.
SRP 9.5.8SRP TITLE:EMERGENCY DIESEL ENGINE COMBUSTION AIR INTAKE AND EXHAUSTA.Actual difference between FSAR and SRP SRP 9.5.8, Paragraph III.8, states that a minimu m of 20 feet should exist between the bottom of all fresh air intakes and the grade elevation. The act ual design is not consistent with this requirement. The bottom of the inta ke hoods is at elevation 40 feet-9 inches, which is 16 feet-3 inches above grade.B.Justification for difference from SRP The actual design has the center line of the intake at 19 feet-6 inches above grade, and also employs air filter silencers to control dust. Periodic surveillance of the pressure drop across the air filter and changing the filter when necessary will preclude excessive pressure drop from dust.
SRP 10.2.3 SRP TITLE: TURBINE DISK INTEGRITY A.Actual difference between FSAR and SRPThe turbine manufacturer (GE) states that the turbine materials have the lowest FATT and highest Charpy V-notch energies available, but provides no data for comparison with SRP 10.2.3, Paragraph II.1.B.Justification for difference from SRP GE considers data and calcul ations requested by the SRP to be proprietary information which, if requested, can be ma de available to the NRC under the provisions of 1 0CFR 2.790.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-51Rev. 30 SRP 10.3 SRP TITLE: MAIN STEAM SUPPLY SYSTEM (MSSS)A.Actual difference between FSAR and SRP FSAR Section 10.3 does not tabulated and describe all flow paths that branch off the main steam lines between the main steam isolation valves and the turbine stop valves as required by SRP 10.3, Paragraph III.5.d.B.Justification for difference from SRP All flow paths are shown on the appropriate P&IDs.SRP 11.5SRP TITLE:PROCESS AND EFFLUENT RADIOLOGICAL MONITORING INSTRUMENTATION AND SAMPLING SYSTEMSA.Actual differences between FSAR and SRP1.SRP 11.5, Table 1, Item 6, requires an automatic control feature which automatically terminates effluents of th e fuel storage area ventilation system.
Monitor 3HVR-RE17 extracts a sample from this system but provides no automatic termination.2.SRP 11.5, Table 2, Item 5, also requires an automatic control feature, which automatically terminates effluents of the spent fuel pool treating system. No such provision is provided on Millstone 3.3. SRP 11.5, Table 2, Items 16 and 17, require an automatic control feature, which automatically terminates effluents of the steam generator blowdown system.B.Justification for differences from SRP1.During fuel handling activities, the fuel building ventilation is processed by the fuel building filtration units. Accident anal ysis indicates that the filters prevent the release of excessive amounts of radioactive effluent.2.The spent fuel pool cooling and purific ation is a closed system; therefore, termination of effluents is unnecessary.
Monitoring is accomplished using the reactor plant sampling system radiation monitor, 3SSR-RE08, and area radiation monitors surveying the fuel pool. Sa fety evaluations described in FSAR Section 9.1.3 show this to be adequate.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-52Rev. 303.Monitoring of the steam generator blow down system is provided by the reactor plant sampling system radiation monitor, 3SSR-RE08. An evaluation of the accident scenario for a st eam generator tube rupture shows that such an event would be identified by the air ejector system monitor, 3ARC-RE21 or the main steam line monitors, 3MSS*RE75-78. Th e main steamline monitors would identify which steam generator is affected and operator action would close valves to prevent release of steam generator blowdown effluents.
SRP 12.2SRP TITLE: RADIATION SOURCESA.Actual difference between FSAR and SRP SRP 12.2, Paragraph I.2, requires tabulation of th e calculated concentrat ions of radioactive material, by nuclide, expected during nor mal operation, antici pated operational occurrences, and accident conditions for equipment cubicles, corridors, and operating areas normally occupied by operating personnel.
FSAR Section 12.2 doe s not tabulate the calculated concentrations of radioactive material expected during accident conditions.B.Justification for difference from SRP During accident conditions, local surveys and measurements will be pe rformed as required and exposures will be limited to the requirements of NUREG-0737.
SRP 13.5.2 SRP TITLE: OPERATING AN D MAINTENANCE PROCEDURESA.Actual difference between FSAR and SRP SRP 13.5.2, Paragraph II.C.2, references Sect ion 5.3 of ANSI/ANS 3.2 - 1981 (Draft 7).
The FSAR is based on Section 5.3 of ANSI N18.7 - 1976/ANS 3.2 which is endorsed in Regulatory Guide 1.33.B.Justification for difference from SRP The NRC endorses the 1976 version of ANSI/
ANS 3.2, Section 5.3 in Regulatory Guide 1.33, Revision 2, which is in accordance with Regulatory Guide 1.70, Revision 3.Emergency Operating Procedures are developed based on the Westinghouse Owner's Group Emergency Response Guidelines as approved by the NRC. The Emergency Operating Procedures are functional-based as describe d in the FSAR. Since the requirement for functional-based Emergency Oper ating Procedures is already explicitly addressed, and the approach is to comply with NRC approved pro cedure guidelines, little will be gained by committing to a partial standard which is not addressed within existing regulatory guides.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-53Rev. 30 SRP 14.2 SRP TITLE: INITIAL PLANT TEST PROGRAMS - FSARA.Actual difference between FSAR and SRP SRP 14.2, Paragraph II.4, requires the Applic ant to have recognized categories of reportable occurrences that are repeatedly being experienced at other facilities. FSAR Section 14.2 does not provide categories of occurrences.B.Justification for difference from SRP The review of operating and testing experience in development of the test program is independent of categories and frequency of occurrence. The criterion for incorporation into Millstone 3 procedures or de sign is the applicat ion of the experience to the safety, reliability, and cost of Unit 3.
SRP 15.4.6SRP TITLE:CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT RESULTS IN A DECREASE IN THE REACTOR COOLANTA.Actual difference between FSAR and SRP FSAR Section 15.4.6 does not address this ac cident scenario.B.Justification for difference from SRP Based upon the evaluation of current industry reviews and an NRC internal review, the consequences and risk of an inadvertent bor on dilution is minimal. Based on cost benefit analysis, modifications also were not shown to be justifiable. Refer to the Applicant's response to the NRC Acceptan ce Review Request Number 440.8.
SRP 15.4.8
SRP TITLE: SPECTRUM OF ROD EJECTION ACCIDENTS (PWR)A.Actual difference between FSAR and SRP SRP 15.4.8, Subsection III, implies that the stresses should be evaluated to emergency conditions. Westinghouse considers this a fa ulted condition as stated in ANSI N18.2.
Faulted condition stress limits ar e applied for this accident.B.Justification for difference from SRP System overpressurization due to a rod eject ion transient was evaluated in WCAP-7588, Revision 1-A, and received NRC acceptance in the Topical Report Evaluation.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.9-54Rev. 30 SRP 15.6.5SRP TITLE: LOCA INSIDE CONTAINMENTA.Actual difference between FSAR and SRP No modifications have been made to the small break LOCA model in accordance with NUREG-0737, Items II.K.3.30 a nd II.K.3.31, as required by SRP 15.6.5, Paragraph II.3.B.Justification for difference from SRPThe small break LOCA analysis model, currently approved by the NRC, is conservative and in compliance with Appendix K of 10 CFR 50. However, Westinghouse believes that improvement in the realism of small break calculations is a worthwhile effort and has committed to revise its small break LOCA an alysis model to address the NRC concerns mentioned in II.K.3.30 and II.K.3.31 (see FSAR Section 1.10).
SRP 15.7.3SRP TITLE:POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING TANK FAILURESA.Actual difference between FSAR and SRP The FSAR has analyzed postulated tank failur e using 1 percent fuel defects, while SRP 15.7.3, Paragraph III.1.a, uses 0.12 percent fuel defects.B.Justification for difference from SRP FSAR analysis is more conservative.TABLE 1.9-2 SRP DIFFERENCES AND JUSTIFICATIONS (CONTINUED)
MPS3 UFSAR1.10-1Rev. 30 1.10 TMI ACTION ITEMSTable 1.10-1 presents the MNPS-3 positions on the PWR applicable items from the US Nuclear Regulatory Commission's post-TMI action plan requirements for applicants for an operating license, NUREG-0737, Enclosure 2, dated November 1980.
MPS3 UFSARMPS3 UFSAR1.10-2Rev. 30TABLE 1.10-1 TMI ACTION ITEMS Item and TitlePosition FSAR/TS ReferenceI.A.1.1 Shift Technical AdvisorMPS-3 meets the provisions of the Commission's Policy Statement for providing engineering expertise on shift (STA). TS 6.2.2 & 6.2.4 I.A.1.2 Shift Supervisor Administrative DutiesMPS-3 meets the requirements of this item.13.1.2.2 I.A.1.3 Shift ManningMPS-3 meets the requirements of this item.TS 6.2.2
I.A.2.1 Immediate Upgrade of RO and SRO Training and QualificationsMPS-3 meets the requirements of this item.13.2 I.A.2.3Administration of Training ProgramsMPS-3 meets the requirements of this item.13.2 I.A.3.1 Revise Scope and Criteria for Licensing ExaminationsThis item is not applicable to MPS-3.*I.B.1.2 Evaluation of Organization and ManagementMPS-3 meets the requirements of this item.13.4.4 & TS 6.2.3 I.C.1 Short Term Acci dent Analysis and Procedures RevisionMPS-3 Emergency Operating Procedures are based on Westinghouse Owners Group emergency procedure guidelines which are approved by the NRC.13.5.2.1 I.C.2 Shift and Relief Turnover ProceduresMPS-3 meets the requirements of this item.13.5.1.3 I.C.3 Shift Supervisor ResponsibilityMPS-3 meets the requirements of this item.13.1.2 I.C.4 Control Room AccessMPS-3 meets the requirements of this item.13.5.1 I.C.5 Procedures for Feedback of Operating ExperienceMPS-3 meets the requirements of this item.13.3.5I.C.6 Procedures for Verification of Correct Performance of Op erating ActivitiesMPS-3 meets the requirements of this item.13.5 MPS3 UFSARMPS3 UFSAR1.10-3Rev. 30I.C.7 NSSS Vendor Review of ProceduresMPS-3 emerge ncy operating procedures are based on NRC approved Westinghouse Emergency Response Guidel ines and therefore eliminates the requirements for additional NSSS vendor review of emergency operating procedures.
13.5.2 I.C.8 Pilot Monitoring of Selected Emergency Procedures for NTOLSMPS-3 emergency operating proced ures are based on NRC approved Westinghouse Emergency Response Guidel ines and therefore eliminates the requirement for pilot monitoring of selected emergency procedures for near term operating license applicants.
13.5.2 I.D.1 Control Room Design ReviewA control room de sign review was performed for MPS-3 to meet the requirements of this item. Modifications are reviewed in accordance with the Design Control Manual.
18 I.D.2 Plant Safety Parameter Display ConsoleMPS-3 meets the requirements of this item.7.5.1 I.G.1 Training during Low-Power TestingThis item was completed prior to low power testing.*II.B.1Reactor Coolant System VentsSafety grade reactor vessel and pressuri zer venting capability is provided in the MPS-3 design.
5.4.15, 7.5 II.B.2 Plant ShieldingThe MPS-3 plant shielding design is outlined in Chapter 12.12.3 II.B.3 Post-Accident SamplingNot applicable to MPS-3 (Reference Technical Specifications Amendment number 201)
- II.B.4 Training for Mitigating Core DamageMPS-3 has implemented a training program utilizing the INPO guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage as the basis for the program.
13.2 TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-4Rev. 30II.D.1 Testing Requirements for Reactor Coolant System Relief and Safety Valves MPS-3 meets the requirements of this item as noted below with the exception with the exception of ATWS testing. MPS-3 meets the requirement of this item for the safe ty valves via the EPRI test program and for the PORVs through extrapolation of EPRI test data to the MPS-3 specific PORV design. Verification of block va lve functionability is demonstrated by application of test data and analyses.
5.4.13 II.D.3 Relief and Safety Valve Position IndicationThe pressurizer PORVs have a reliable direct position indication in the control room.
5.4.13, 7.5 II.E.1.1 Auxiliary Feedwater System Evaluation An auxiliary feedwater system reliabil ity analysis has been performed as specified. Incorporation of cavitating ventures has solved any possibility of excessive flowrate; therefore, M PS-3 meets the requirements of this item. 10.4.9TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-5Rev. 30 II.E.1.2 Auxiliary Feedwater System Automatic. Initiation and Flow Indication The present design of the MPS-3 auxiliary feedwater system incorporates the requirements to have reliable automatic initiation and flow indications in accordance with IEEE 27 9-1971. During normal power operation, manual AFW initiation is accomplished by the operator starting the pumps and isolating Steam Generator blowdown lines. At below 10
% rated thermal power, the MDAFW pumps ma y be in normal operation, feeding the steam generators to maintain Stea m Generator inventory level. In this mode, there is no automatic functi on associated with the MDAFW pump flow control valves and operator ac tion is required to open the MDAFW system control valves. The motor dr iven auxiliary feedwater pumps may be aligned to take suct ion from the non-safety grade condensate storage tank (CST). Motor-driven auxiliary feedwater pump suction automatically switches to the demineralized wa ter storage tank (DWST), including isolation from the CST, in the event of an SIS, LOP, CDA, two of four low-low water level condition in any one steam generator, or AMSAC signal. Manual initiation may require operator action to realign the motor-driven auxiliary feedwater pump sucti on source in this mode of operation.
10.4.9, 7.3 II.E.3.1 Emergency Power Supply for Pressurizer HeatersThe emergency power supply for pressurizer heaters meets the requirements of this item.
8.3.1, 5.4.10.3 II.E.4.1 Dedicated Hydrogen Penetrations (Containment Design)
The MPS-3 design includes redundant hydrogen recombiners and analyzers outside the containment. Pe netrations for this equipment are dedicated to that service as described in Section 6.2.4 and 6.2.5. Use of
recombiners and analyzers is no longer a requirement for DBAs.
6.2.4, 6.2.5 II.E.4.2 Containment Isolation DependabilityMPS-3 meets the requirements of this item. 6.2.4, 7.3.1 TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-6Rev. 30II.F.1 Additional Accident Monitoring Instrumentation MPS-3 meets the requirements of this item. Hydrogen monitors meet the requirements of 10 CFR 50.44 using guidance of Regulatory Guide 1.7.11.5, 7.5, 6.2.1, 6.2.5, TS 3.3.3.10, TS 3.3.3.6, TS3.6.4.1II.F.2 Identification of and Recovery From Conditions Leading to Inadequate Core
CoolingMPS-3 utilizes a subcooled margin monitor system, and a core exit thermocouple system and Combustio n Engineering heated junction thermocouple system to meet th e requirements of this item.
4.4.6.5 II.G.1 Power Supplies fo r Pressurizer Relief Valves, Block Valves, and Level IndicatorsEmergency power for pressurizer e quipment meets the requirements for this item.
5.4.13 II.K.1.5 Review ESF ValvesMPS-3 meets the requirements of this item.13.5.2 II.K.1.10 Operability StatusMPS-3 meets the requirements of this item13.5.1 II.K.1.17 Trip per Low-Level B/SAn MPS-3 pressurizer lo w pressure signal initiates both a reactor trip and the start of safety injection. Pressuri zer low-level trips are not utilized on MPS-3.7.2, 7.3 II.K.2.13 Thermal Mechanical ReportMPS-3 meets the requirements of this item.5.2.3 TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-7Rev. 30II.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients MPS-3 meets the requirements of this item Westinghouse has performed a study which addresses the potential for void formation in Westinghouse desi gned nuclear steam supply systems during natural circ ulation cooldown/depressurization transients. This study has been submitted to the NRC by the Westinghouse Owners Group (Jurgensen 1981c) and is applicable to MPS-3. The Staff has accepted the study Reference - Safety Eval uation Report for Millstone 3, NUREG-1031. In addition, the Westinghouse Owners Group has
developed a natural circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head region during natural circul ation cooldown/depressurization transients, and specifies those c onditions under which upper head voiding may occur. These Westinghouse Owne rs Group generic guidelines have been submitted to the NRC (Jurge nsen 1981e). The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant specific consideration) ha s been utilized in the implementation of MPS-3 plant specific operating procedures.
- II.K.2.19 Sequential A uxiliary Feedwater Flow Analysis Not applicable to MPS-3. The NRC has completed a generic review on this subject and concluded that the concerns expressed in this item are not applicable to NSSS with inverted U-tubes such as the one utilized in MPS-3 (Varga 1981).
- II.K.3.1 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System The addition of an automatic isolation system for the PORVs will not be utilized for MPS-3. Modifications implemented under Item II.K.3.2 will reduce the probability of a LOCA caused by a stuck open PORV to an acceptably low level. Also automati c closure of the block valve may inhibit operator detection of a stuck open PORV.
- TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-8Rev. 30II.K.3.2 Report on Overall Safety Effects of Power-Operated Relief Valve Isolation System MPS-3 meets the requirements of this item.
A generic report responsive to this item was produced by Westinghouse (WCAP-9804, 1981). The post-TMI modifications discussed in the above report have been implemented. These modifications will reduce the probability of a small break LOCA caused by a stuck open PORV to an insignificant level relative to all other small break LOCA events. This report determined the frequency of a small-break LOCA caused by a stuck open PORV is reduced to about 2.1 x 10
-6 per reactor year for MPS-3 design. This is well below the WASH-1400 medium frequency of 10
-3 for a small break LOCA.
- II.K.3.3 Reporting SV and RV Failures and Challenges The licensee is responsible for ensuring that any failure of a PORV or safety valve to close will be reported promptly to the NRC. All challenges to the PORVs or safety valves will be documented in the annual report.
TS 6.9.1II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss-of Coolant Accident Automatic trip of reactor coolant pumps during LOCA is not provided. Westinghouse has performed an analysis of delayed reactor coolant pump trip during small break LOCAs. Th is analysis is documented in WCAP-9584 and WCAP-9585 (1979). In addition, Westinghouse has performed test predictions of LOFT Experiments L3-1 and L3-6. The results of these predictions are documented in Jurgensen (1981a,b,d).
NNECO provided additional informati on related to Generic Letter 85-12 concerning implementation of TMI Ac tion Item II-K.3.5 in letters dated September 16, 1985, November 19, 1985, and June 30, 1987. NRC Letter dated March 29, 1989 closes out this issue by finding the plant-specific RCP trip setpoint development acceptable. The letter also states there are no longer safety significant concerns for the plant specific information.
- II.K.3.7 Evaluation of PORV Opening Probability This item is applicable to B&W pl ants only and therefore does not apply to MPS-3.* TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-9Rev. 30 II.K.3.9 Proportional Integral Derivative Controller Modification The MPS-3 design does not include a proportional integral derivative (PID) controller in the power-operate d relief valve control circuit (see Figures 7.7-4 and 7.2-1, Sheet 11).
7.7.1.5 II.K.3.10 Proposed Anticipatory Trip Modification The MPS-3 design incorporates this trip modification.
The NRC has raised the questi on of whether the pressurizer power-operated relief valves would be actuated for a turbine trip without reactor trip below a power level of 51 percent which is the highest P-9 trip setpoint allowed per Technical Specifications Table 2.2-1. An analysis has been performed using realistic yet conservative values for the core physics parameters (primarily reactivity feedback coefficients and control rod worths), and a conservatively high in itial power level of 53 percent. All operating parameters (RCS and secondary temperature, pressure and flow) are at initial values without uncer tainties, corresponding to 53 percent power. The transient was initiated from a power level of 53 percent, which is the highest P-9 trip setpoint allowed per Technical Specifications plus 2 percent for power measurement uncertainty. This is a conservative starting point, and would bracket all transients initiated from a lower power level.
The core physics parameters used were the ones that would result in the most positive reactivity feedbacks (i.e., highest power levels). NSSS Control Systems (i.e., steam dump, rod control, pressurizer spray) are assumed to be operationa l in the automatic mode of control. Based upon the results from the analysis, the peak pressure reached in the pressurizer would be 2,328 psia. Which remains below the nominal PORV actuation setpoint of 2.350 psia. The result also indicate that the peak pressure reached in the steam generator woul d be 1.124 psia which remains below the nominal secondary relief valv e actuation setpoint of 1.140 psia.
10.4.4.1, 7.2.1 II.K.3.11 Justification Use of Certain PORVsThe PORVs used in the MPS-3 design are pi lot-operated relief valves supplied by Garrett.
- TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-10Rev. 30II.K.3.12 Confirm Anticipatory TripThe MPS-3 design includes an anticipatory reactor trip on turbine trip as discussed in FSAR Section 7.2.1.1.2 (Item 6). The logic for this trip is shown on Figure 7.2-1 (Sheet 16).
7.2.1.1.2 II.K.3.17 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification ChangesThe ECCS Component Reliability and Tracking Program commitment has been satisfied. The required reliab ility information required by NUREG 0737 on ECCS components involved in outages during the first five years of commercial operation has been co mpiled, formally documented, and is retrievable on site from Nuclear Document Services for review. The reliability trending and monitoring program committed to within the SER has been superseded by and is cu rrently performed under the existing Maintenance Rule program consistent with the requirements contained with 10 CFR 50.64.
- II.K.3.25 Effect of Loss of Alternating Current Power on Pump Seals During normal operation, seal injecti on flow from the chemical and volume control system is provided to cool the RCP seals, and the component cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals. In the event of loss of off site power, the RCP motor is de-energized and both of these cooling supplies are terminated; however, the diesel generators are automaticall y started and either seal injection flow or component cooling water to th e thermal barrier heat exchanger is automatically restored within seconds.
Either of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during loss of of f site power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. See Section 8.3 for diesel loading sequence.
- II.K.3.30 Revised Small Break Loss-of-Coolant Accident Methods to Show Compliance With 10 CFR Part 50, App. KMPS-3 meets the requirements of this item.15.6.5 TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSARMPS3 UFSAR1.10-11Rev. 30 II.K.3.31 Plant Specific Calculations to Show Compliance With 10 CFR Part 50.46MPS-3 meets the requirements of this item.15.6.5 & TS 6.9.1.6 III.A.1.1 Emergency Preparedness, Short TermExisting plans for Millstone site apply.13.3 III.A.1.2 Upgrade Emergency Support FacilitiesThe emergency support facilities identified in the Emergency Plan are functional.
13.3 II.A.2 Emergency PreparednessExisting plans for the Millstone site apply.13.3 III.D.1.1 Primary Coolan t Sources Outside Containment MPS-3 has implemented a program to reduce leakage from systems outside containment that would or c ould contain highly ra dioactive fluids during a serious transient or accide nt to as-low-as practical levels.
5.2.5 & TS 6.8.4.a III.D.3.3 Inplant I Radiation Monitoring Continuous air monitors with direct readout and alarm capabilities are located in the MPS-3 control room and in the Technical Support and Emergency Operations Centers.11.5, 12.3, TS 6.8.4.b III.D.3.4 Control Room HabitabilityThe requirements of this item have been addressed.6.4, 2.2.3, 9.4.1 TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition FSAR/TS Reference MPS3 UFSAR1.10-12Rev. 30 NOTES: *Statement stands alone, no FSAR Section reference.References for Table 1.10-1:Anderson, T. M. (Westinghouse) September 26, 1980. Lett er (NS-TMA-2318) to Eisenhut, D. G. (NRC).Jurgensen, R. W. (Chairman, Westinghouse Owners Group) Ma rch 3, 1981a. Letter (OG-49) to Ross, D. F., Jr. (NRC).Jurgensen, R. W. March 23, 1981b. Letter (OG-50) to Ross, D. F., Jr. (NRC).Jurgensen, R. W. April 20, 1981c. Letter (OG-57) to Check, P. S. (NRC).Jurgensen, R. W. June 15, 1981d. Letter (OG-60) to Check, P. S. (NRC).Jurgensen, R. W. November 30, 1981e. Letter (OG-64) to Eisenhut, D. G. (NRC).Rahe, E. P. (Westinghouse) November 25, 1981. Letter (NS-EPR-2524) to Eisenhut, D. G. (NRC).Varga, S. A. (NRC) June 29, 1981. Letter to Carey, J. J. (Duquesne Light Company).
WCAP-9584 (Proprietary) and WCAP-9585 (Non-proprietary), August 1979 "Analysis of Delayed Reactor Coolant Pump Tri pSmall Loss-of-Coolant Accidents for Westinghouse Nuclear Steam Supply System."
WCAP-9804, February 1981. "Probabilistic Analysis and Operational Data in Re sponse to NUREG-0737 Item II.K.3.2 for W e NSSS Plants."Jaffe, D. H. (NRC), March 29, 1989 Le tter to Mroczka, E. J. (NNECO).TABLE 1.10-1 TMI ACTION ITEMS (CONTINUED)Item and TitlePosition F R MPS3 UFSAR1.11-1Rev. 301.11 MATERIAL INCORPORATED BY REFERENCE The following is a list of mate rial incorporated by reference in the Final Safety Analysis Report (1): 1.Millstone Unit 3 Technical Requirements Manual (TRM). 2.As identified in the List of Figures, the engineering controlled plant drawings that are, coincidentally, MPS-3 FSAR Figures.
(1)Information incorporated by reference into the Final Safety Analysis Report is subject to the update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59 unless separate NRC change control requirements apply (e.g., 10 CFR 50.54(a)).