ML20209A402

3 to Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant

ML20209A402
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Site:	Millstone
Issue date:	06/22/2020
From:	Dominion Energy Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
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CONTAINS SECURITY RELATED INFORMATION.

PLEASE HANDLE PROPERLY Millstone Power Station Unit 3 Safety Analysis Report Chapter 1: Introduction and General Description of Plant This Document contains Security-Related Information.

When this page is separated from the document, it is Decontrolled.

Table of Contents tion Title Page INTRODUCTION ...................................................................................... 1.1-1 GENERAL PLANT DESCRIPTION ......................................................... 1.2-1 1 General........................................................................................................ 1.2-1 2 Site .............................................................................................................. 1.2-1 3 Structures .................................................................................................... 1.2-1 4 Nuclear Steam Supply System.................................................................... 1.2-2 5 Instrumentation and Control Systems......................................................... 1.2-4 6 Radioactive Waste Systems ........................................................................ 1.2-4 7 Fuel Handling ............................................................................................. 1.2-5 8 Turbine Generator and Auxiliaries ............................................................. 1.2-5 9 Electrical Systems....................................................................................... 1.2-6 10 Engineered Safety Features ........................................................................ 1.2-6 11 Cooling Water and Other Auxiliary Systems ............................................. 1.2-9 12 Reference for Section 1.2.......................................................................... 1.2-10 COMPARISON TABLES .......................................................................... 1.3-1 1 Comparison with Similar Facility Designs ................................................. 1.3-1 1.1 Comparison of Nuclear Steam Supply System........................................... 1.3-1 1.2 Comparison of Engineered Safety Features................................................ 1.3-1 1.3 Comparison of Containment Concepts ....................................................... 1.3-1 1.4 Comparison of Instrumentation Systems .................................................... 1.3-2 1.5 Comparison of Electrical Systems .............................................................. 1.3-2 1.6 Comparison of Radioactive Waste Systems ............................................... 1.3-2 1.7 Comparison of Other Reactor Plant Systems ............................................. 1.3-2 2 Comparison of Final and Preliminary Designs........................................... 1.3-2 IDENTIFICATION OF AGENTS AND CONTRACTORS...................... 1.4-1 1 Licensee's Subsidiaries ............................................................................... 1.4-1 2 Architect-Engineer...................................................................................... 1.4-1

tion Title Page 3 Nuclear Steam Supply System Manufacturer ............................................. 1.4-1 4 Turbine Generator Manufacturer ................................................................ 1.4-1 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION ........................................................................................ 1.5-1 GENERAL REFERENCES (HISTORICAL) ............................................ 1.6-1 DRAWINGS AND OTHER DETAILED INFORMATION ..................... 1.7-1 1 Electrical, Instrumentation, and Control Drawings .................................... 1.7-1 2 Piping and Instrumentation Diagrams ........................................................ 1.7-1 3 Loop and Systems Diagrams ...................................................................... 1.7-1 4 Other Detailed Information (Special Reports and Programs)..................... 1.7-1 CONFORMANCE TO NRC REGULATORY GUIDES .......................... 1.8-1 N NSSS CONFORMANCE TO NRC REGULATORY GUIDES ............. 1.8N-1 STANDARD REVIEW PLAN DOCUMENTATION OF DIFFERENCES .......................................................................................... 1.9-1 TMI ACTION ITEMS .............................................................................. 1.10-1 1 MATERIAL INCORPORATED BY REFERENCE ............................... 1.11-1

List of Tables mber Title 1 Design Comparison 2 Comparison of Engineered Safety Features 3 Comparison of Containment Concepts 4 Comparison of Containment Atmosphere Pressure Sensor Parameters 5 Comparison of Reactor Coolant Pump Bus Protection 6 Comparison of Engineered Safety Feature Actuation Signals 7 Comparison of Emergency Generator And Steam Generator Auxiliary Feedwater Pump Start Signals 8 Comparison of Process And Effluent Radiation Monitoring Systems 9 Comparison of Area Radiation Monitoring Systems 10 Comparison of Airborne Radiation Monitoring Systems 11 Comparison of Electrical System Parameters 12 Comparison of Radioactive Liquid Waste Systems 13 Comparison of Radioactive Gaseous Waste Systems 14 Comparison of Radioactive Solid Waste Systems 15 Comparison of Other Reactor Plant Systems 16 Comparison of Final And Preliminary Information 1 Topical Reports as General References (Historical) 1 Electrical, Instrumentation, And Control Reference Documentation 2 Piping And Instrumentation Diagrams 3 Omitted 4 Special Reports And Programs 1 NRC Regulatory Guides N-1 NRC Regulatory Guides 1 Summary of Differences From SRP 2 SRP Differences And Justifications

-1 TMI Action Items

List of Figures mber Title

-1 Not Used

-2 Plot Plan 3 (Sheets 1-3) Piping & Instrumentation Diagram Legend

s Final Safety Analysis Report (FSAR) has been prepared with the guidance of Regulatory de 1.70, Revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear er Plants, LWR Edition, dated November 1978. The report is intended to be responsive to the de, to existing regulations, and to NUREG-75/087, Standard Review Plan for the Review of ety Analysis Reports for Nuclear Power Plants, LWR Edition.

INTRODUCTION s report was submitted in support of an application by the companies listed in the General rmation Section of the application (the Applicants) for a Class 103 permit for a facility rating license to operate a nuclear power plant, designated as Millstone Nuclear Power ion - Unit 3 (Millstone 3). This plant is located on a site in the town of Waterford, New don County, Connecticut, on the north shore of Long Island Sound.

lstone 3 uses a pressurized water type nuclear steam supply system (NSSS) furnished by stinghouse Electric Corporation (WNES) and a turbine-generator furnished by the General ctric Company (GE). The remainder of the unit, including a subatmospheric reactor tainment, was designed and constructed by the Applicants, with the assistance of their esentative, Northeast Utilities Service Company (NUSCo.), and their architect-engineer, ne & Webster Engineering Corporation (SWEC).

core was originally designed for a warranted power output of 3,411 MWt, which was the inal license application rating. This output, combined with the reactor coolant pump heat put of 14 MWt, gave an NSSS warranted output of 3,425 MWt.

core has been re-analyzed for a power output of 3,650 MWt which, when combined with the sed reactor coolant pump output of 16 MWt, gives a 100% NSSS power output of 3,666 MWt.

steam and power conversion equipment, including the turbine-generator, has the capability to erate a maximum calculated gross output of approximately 1,296 MWe. When the NSSS is rating at its warranted output of 3,666 MWt, the net electrical output is approximately 1,245 e.

project schedule was based on a fuel loading date of November 1, 1985, and an anticipated mercial operation date of May 1, 1986. The Low Power License (5 percent) was issued by the C November 25, 1985, the Full Power License was issued January 31, 1986, and the Unit ame commercially operational April 23, 1986.

001, Millstone Units 1, 2 and 3 operating licenses were transferred from Northeast Nuclear rgy Company to Dominion Nuclear Connecticut, Inc. (DNC).

C is an indirect wholly-owned subsidiary of Dominion Energy, which is in turn owned by minion Resources, Inc. (DRI). Virginia Power, which is the licensed owner and operator of the th Anna and Surry nuclear stations, is also a subsidiary of DRI.

&P and DNC.

FSAR will retain references to Northeast Utilities and Northeast Nuclear Energy Company uments/activities when they are used in a historic context and are required to support the plant nsing bases.

n license transfer, all records and design documents necessary for operation, maintenance, decommissioning were transferred to DNC. Some of these drawings are included (or renced) in this FSAR. These drawings often have title blocks (or drawing numbers) which list theast Nuclear Energy Company or Northeast Utilities Service Company (et. al). In general, changes to these title blocks will be made at this time. Based on this general note, these wings shall be read as if the title blocks list Dominion Nuclear Connecticut, Inc.

s section includes a summary description of the principal characteristics of the site and a cise description of Millstone 3.

1 GENERAL lstone 3 incorporates a four loop closed cycle pressurized water type nuclear steam supply em (NSSS); a turbine generator and electrical systems; engineered safety features; radioactive te systems; fuel handling systems; structures and other on site facilities; instrumentation and trol systems; and the necessary auxiliaries required for a complete and operable nuclear power ion. The site plan (Figure 2.1.4) and the plot plan (Figure 1.2-2) show the general ngement of the unit.

ng and instrumentation diagrams are included throughout this document with the appropriate em descriptions. Symbols and abbreviations used in the diagrams are illustrated on ure 1.2-3.

h respect to the numbers, graphs, and drawings included within this report, the normal rance permitted by good engineering practice is intended. Where operating parameters are sually important, such items have been included in the technical specifications.

2 SITE site, approximately 500 acres in area, is on the north shore of Long Island Sound and on the side of Niantic Bay. It is located in the Town of Waterford, Connecticut, about 3.2 miles west-thwest of New London and about 40 miles southeast of Hartford. The surrounding area is arily residential with some commercial and industrial uses.

lstone 1 and 2 are also located on the site. Millstone 1 is a permanently defueled boiling water tor. Millstone 2 uses a two-loop pressurized water reactor supplied by Combustion ineering, Inc., with a rated thermal power level of 2,700 MW; the architect-engineer was htel Corporation. Section 2.1 contains a more detailed description of the site and surrounding s.

3 STRUCTURES lstone 3 major structures are the containment structure, auxiliary building, fuel building luding decontamination facilities), waste disposal building, engineered safety features ding, main steam valve building, turbine building, service building, control building, technical port center, emergency generator enclosure, containment enclosure building, warehouse 5 luding the condensate polishing waste treatment facility), auxiliary boiler enclosure, and ulating and service water pumphouse. Section 3.8.4.1 describes the general arrangement of e structures.

urry Power Stations 1 and 2, North Anna Power Stations 1 and 2, and Beaver Valley Power ion 1. Following the loss-of-coolant accident (LOCA), described in Section 15.6.5, the tainment remains above atmospheric pressure.

containment structure is housed within the containment enclosure building, which along with ctures adjacent to the containment, forms the boundary of the supplementary leak collection release system (SLCRS). The SLCRS establishes a subatmospheric pressure in the tainment enclosure building and contiguous structures. See Section 6.2.3 for a further cription.

seismic criteria used in the design of the structures and equipment for Millstone 3 are cribed in Section 3.7.

4 NUCLEAR STEAM SUPPLY SYSTEM nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water-type tor and four closed reactor coolant loops connected in parallel to the reactor vessel. Each loop tains a reactor coolant pump and a steam generator, two loop isolation valves, an isolation ass valve, and a bypass line. The NSSS also contains an electrically heated pressurizer and iliary systems.

h pressure water circulates through the reactor core to remove heat generated by the nuclear in reaction. The heated water exits from the reactor vessel and passes via the coolant loop ng to the steam generators. Here, it releases heat to the feedwater to generate steam for the ine generator. The cycle is completed when the water is pumped back to the reactor vessel.

entire coolant system is composed of leaktight components to ensure that all fluids are fined to the system.

reactor core is of the multi-region type. All fuel reactor assemblies are mechanically tical, although the fuel enrichment is not the same in all assemblies. These assemblies rporate the rod cluster control concept in canless 17 x 17 fuel rod assemblies using a spring grid to provide support for the fuel rods. The reactor moderator has a negative temperature fficient of reactivity at full power at all times throughout core life.

he typical initial core loading, three fuel enrichments are used. Fuel assemblies with the hest enrichments are placed in the reactor core outer region, and the two groups of lower chment fuel assemblies are arranged in a selected pattern in the central region. In subsequent elings, one-third of the fuel is discharged from the central region and fresh fuel is loaded into outer region of the reactor core. The remaining fuel is arranged in the central two-thirds of the tor core in such a manner as to achieve optimum power distribution.

cluster control assemblies are used for reactor control and consist of clusters of cylindrical orber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the tor core, each cluster of absorber rods is attached to a spider connector and drive shaft, which

reactor coolant pumps are vertical, single-stage, centrifugal pumps of the shaft-seal type.

steam generators are Westinghouse Model F vertical U-tube units which contain Inconel es. The Model F steam generator includes features such as improved tube support plate design high circulation ratio which are designed to minimize most forms of corrosion, sludge dup, and chemical attack. Integral moisture separation equipment reduces the moisture of m to one-quarter percent or less.

of the pressure containing and heat transfer surfaces in contact with reactor water are stainless l clad or stainless steel except the steam generator tubes and fuel tubes, which are Inconel and aloy, respectively. Reactor core internals, including control rod drive shafts, are primarily nless steel.

re are two double-disc, motor-operated, loop isolation valves in each loop, one located ween the reactor vessel and the steam generator and the other between the reactor vessel and reactor coolant pump of each of the four loops. The isolation bypass valves, also double-disc, or-operated valves, are located in a bypass line connecting the two loop isolation valves in h loop.

electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant em pressure during normal operation, limits pressure variations during plant load transients, keeps system pressure within design limits during abnormal conditions. In addition, it vides indication of and maintains reactor coolant system water inventory.

auxiliary systems, provided as part of the NSSS, charge the reactor coolant system and add eup water, purify reactor coolant water, provide chemicals for corrosion inhibition and tivity control, remove decay heat when the reactor is shut down, and provide for emergency ty injection.

er auxiliary systems supporting the NSSS but not part of the NSSS provide the following:

1. System components cooling

2. Fuel pool cooling

3. Reactor coolant water and other auxiliary system fluid sampling

4. Venting and draining the reactor coolant system and other auxiliary systems

5. Emergency containment depressurization spray and combustible gas control

6. Maintaining the containment atmosphere pressure at sub-atmospheric levels

8. Preparation of solid wastes for disposal

9. Process, store, and supply reactor coolant system boric acid

10. Component ventilation

11. Instrument and valve operator air 5 INSTRUMENTATION AND CONTROL SYSTEMS instrumentation and control for the reactor protection system, engineered safety features ation system, and other safety related systems meet the requirements of IEEE 279-1971, iteria for Protection System for Nuclear Power Generating System. In addition, other licable criteria are met as described in Sections 3.1 and 7.1.2.

nonsafety related instrumentation and controls accomplish reliable control and allow nitoring of the plant status without degradation of safety related instrumentation. Section 7.7 cribes the design details.

reactor is controlled by a coordinated combination of chemical shim, mechanical control s, and temperature coefficients of reactivity. The control system allows the unit to accept step changes of 10 percent and ramp load changes of 5 percent per minute over the load range of ercent to 100 percent power under nominal operating conditions subject to xenon limitations.

trol of the reactor and the turbine generator is accomplished from the control room, which tains all instrumentation and control equipment required for startup, operation, and shutdown, uding normal and accident conditions. The turbine generator controls are designed for manual ration; the operator selects the load setpoint and loading rate. The NSSS automatically follows turbine generator, on decreasing power, from loads of 100 to 15 percent power. The operator s manual action to match the NSSS to the turbine generator load, for load increases between and 100% power. If, during rapid turbine generator loading (5 percent per minute), the onse of the control rods and chemical shim is not adequate to supply the needed reactivity, the tor coolant temperature automatically drops to supply more reactivity. If a reactor coolant low rating temperature is reached, turbine generator loading is stopped automatically.

6 RADIOACTIVE WASTE SYSTEMS ioactive wastes are collected, processed, and disposed of in a safe manner complying with ropriate regulations, in particular, NRC Regulations 10 CFR 20, 10 CFR 50 Appendix I, 10 R 61, 10 CFR 71, 49 CFR 171-178, 10 CFR 100, and General Design Criteria 60 and 64 ctions 3.1.2.60 and 3.1.2.64). There are three interrelated radioactive waste treatment systems:

oactive liquid waste, radioactive gaseous waste, and radioactive solid waste. Chapter 11 cribes these systems.

harge off site. The process operations available to treat the liquid wastes are filtration and ineralization. The process descriptions and flow diagrams illustrate the number and sequence rocessing steps to be applied to each type of liquid waste (Section 11.2).

eous wastes, consisting of hydrogen streams and air streams containing various levels of oactivity, are treated before release to the environment. Through the use of degasification and fication of reactor coolant letdown, the consequences of any reactor coolant leakage are imized. This degasification and purification process produces hydrogen waste gas streams, ch are passed through charcoal decay beds to provide adequate holdup time for the decay of le gases and the removal of iodines. The decay beds are followed by high efficiency iculate air filters to ensure particulate removal. Aerated waste gas streams, produced by other ses of unit operation, are released to the environment via the Millstone stack. A process flow ram, illustrating the processing steps for gaseous waste, appears in Section 11.3.

radioactive solid waste system provides holdup, packaging, and storage facilities for the ntual off site shipment and ultimate disposal of solid radioactive waste material. Available cess operations are: solidification of liquid wastes, holdup for decay, sluicing and dewatering resins, encapsulation of miscellaneous solid wastes, and compacting of compressibles.

visions for shielding during the processing and shipment of solid wastes are included in the gn of the radioactive solid waste system. A process flow diagram illustrates the processing handling sequences for solid wastes generated by the plant (Section 11.4).

7 FUEL HANDLING reactor is refueled by equipment designed to handle spent fuel under water from the time the nt fuel leaves the reactor vessel until the spent fuel is placed in a cask for shipment from the

. Underwater transfer of spent fuel provides an optically transparent radiation shield as well as liable source of coolant for the removal of decay heat produced by the spent fuel. New fuel is sferred to the fuel pool using the new fuel handling crane and is loaded into the reactor using same equipment that handles the spent fuel. (Section 9.1.4) 8 TURBINE GENERATOR AND AUXILIARIES turbine is a tandem-compound, six-flow, 1,800 rpm unit with 43 inch last stage blades. Two bination moisture separator-reheaters remove moisture and superheat the steam between the h and low pressure turbines. The turbine generator is discussed, in detail, in Section 10.2.

turbine generator plant and associated steam and power conversion systems are capable of a percent load rejection without producing a reactor trip. This is accomplished by dumping m into the condenser through the turbine bypass system which reduces the transient to within NSSS transient response capability.

turbine control system is an electrohydraulic control (EHC) system capable of remote manual utomatic control of acceleration and loading of the unit at preset rates, and holding speed and

ingle-pass deaerating surface condenser installed in three sections, two 100 percent design acity steam jet air ejector units, three 50 percent design capacity condensate pumps, three m generator feedwater pumps (two turbine-driven and one motor-driven), two 50 percent gn capacity motor-driven and one 100 percent design capacity turbine-driven steam generator iliary feedwater pumps, three trains of feedwater heaters, each having six stages, and a full condensate demineralizer polishing system are provided. Any combination of two steam erator feedwater pumps is adequate to support 100% power operation. Condenser circulating er is provided by six circulating water pumps. The turbine plant component cooling system vides cooling water for lubricating oil coolers, generator hydrogen coolers, and other turbine t auxiliary heat loads.

9 ELECTRICAL SYSTEMS electric power system includes the electrical equipment and connections required to generate deliver electric power to the 345 kV system. This system also includes station service trical equipment to provide electric power to support station auxiliaries during normal ration, startup, shutdown, and accident conditions.

major component in the system is the turbine-driven main generator. The electric power put from this generator is stepped up in a transformer bank and delivered to the 345 kV tchyard for distribution to the utility grid.

station service equipment consists of switchgear, load centers, motor control centers, ac vital nonvital buses, and battery systems. The normal source of station service power is provided he main generator through normal station service transformers. Startup and shutdown station ice power is provided by a preferred off site source from the 345 kV switchyard through the n and normal station service transformers with the generator breaker open, or by the rnative off site source from the 345 kV switchyard through the reserve station service sformers. In the event of an accident with loss of both normal and off site sources, an on site rgency power system, consisting of two redundant diesel engine-driven generators, provides er to the emergency 4,160V buses within 11 seconds after receiving a start signal. These el engine-driven generators are sized to provide required power to all safety related ipment.

10 ENGINEERED SAFETY FEATURES ineered safety features (ESF) are provided to mitigate the consequences of postulated dents including a loss-of-coolant accident (LOCA) resulting from large and small pipe breaks.

ESF systems provided for Millstone 3 have sufficient redundancy and independence of ponents and power sources so that, under the conditions of the postulated accident, the

1. Prevent radiation release to the outside environment from exceeding the limit specified in 10 CFR 50.67.

2. Provide core cooling to prevent excessive metal-water reaction, to limit the core thermal transient, and to maintain the core in a coolable geometry.

lstone 3 is independent of Millstone 1 and 2 with respect to ESF. The systems provided for lstone 3 are summarized below.

Containment Structure steel lined reinforced concrete containment structure provides a barrier against the escape of ion products. It is designed to operate at approximately atmospheric pressure and can hstand the pressures and temperatures resulting from a spectrum of LOCAs and secondary em breaks. The containments response following the accident is similar to other atmospheric Rs. The structure and all penetrations, including access openings, are of proven design ction 6.2.1).

Emergency Core Cooling System emergency core cooling system (ECCS) provides borated water to cool the reactor core owing a major LOCA. This is accomplished by the automatic injection of water from the ty injection accumulators into the reactor coolant loops and by the automatic pumping of a ion of the refueling water storage tank contents into the loops via the charging pumps, the ty injection pumps, and the residual heat removal pumps. After the injection mode of rgency core cooling, long term core cooling is maintained by recirculating the water from the tainment structure sump by the containment recirculation pumps, through the containment rculation coolers, and into the reactor coolant loops directly and via the charging and safety ction pumps (Section 6.3).

Containment Heat Removal System containment heat removal system consists of the quench spray and the containment rculation systems. Following the postulated DBA, the containment pressure is reduced by loying both systems.

quench spray system sprays borated water from the refueling water storage tank (RWST).

recirculation spray system draws suction from the containment sump, the content of which sists of the primary or secondary system effluent and the quench spray.

sump water pH is controlled to be above 7.0 to improve effectiveness of fission products oval (Section 6.5.2), and for materials compatibility (Section 6.1.1).

Supplementary Leak Collection and Release System owing a postulated accident, particulate and gaseous radioactive material is ducted from the tainment enclosure structure and the buildings contiguous to the containment structure to the plementary leak collection and release system (SLCRS), where it is filtered and discharged to atmosphere through an elevated stack rather than through a ground-level vent. SLCRS is not ited for a postulated fuel handling accident.

Containment Isolation System containment isolation system isolates piping lines which penetrate the containment boundary hat, in the event of a LOCA, radioactivity is not released to the environment through these s.

lines are either isolated passively (check valves or locked closed manual valves) or isolated matically by receipt of a safety injection signal (SIS), a containment isolation (phase A and ignal, or a steamline isolation (SLI) signal (Section 6.2.4).

Engineered Safety Features Actuation System engineered safety features actuation system (ESFAS) monitors selected parameters and rmines whether predetermined safety limits have been exceeded. If they have been exceeded, ESFAS initiates action to mitigate the abnormal occurrence (Chapter 7).

Habitability Systems Millstone 3 control room envelope is equipped with an intake isolation system designed to ect the plant operators from the presence of hazardous substances outside the control room elope.

trol room envelope makeup air is supplied via redundant air filtration trains designed to meet requirements of Regulatory Guide 1.52 (Section 6.4).

Inservice Inspection ASME Section III, Code Class 1, 2, and 3 systems and components which require inservice ection and testing are designed, fabricated, and erected to meet the inspection requirements of ME Section XI (except where specific written relief has been granted by the Commission suant to 10 CFR 50.55a). The inservice inspection program includes baseline preservice mination and periodic inservice inspection and testing to ensure the operability and integrity ll systems classified Class 1, 2, and 3 pursuant to 10 CFR 50.55a (Sections 5.2.4 and 6.6).

ling water and other auxiliary systems provided in Millstone 3 are outlined below. Their gn criteria and details are described in Chapters 9, 10, and 11.

1. The chemical and volume control system 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 corrosion 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 water 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 system is capable of producing primary-grade water and concentrated boric acid solution for station reuse or disposal.

3. Radioactive fluid degasification, liquid concentration, and waste solidification for disposal are provided by the radioactive gaseous waste, radioactive liquid waste, and radioactive solid waste systems.

4. The service water system provides 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 water system, an intermediate cooling system, transfers heat from systems containing reactor coolant or other radioactive or potentially radioactive liquids to the service water system, and provides a source of safety grade cooling water to systems 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.

spent fuel pool.

8. The reactor plant vent and drain systems collect potentially radioactive fluids and gases from various plant systems and transfer these fluids and gases to the boron recovery system or to the appropriate waste disposal system.

9. Individual ventilation systems are provided for the containment and other structures. The containment ventilation system recirculates and 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 heat resulting from the operation of the turbine generator and main condensers. The 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 throughout 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. Dryers are provided in the instrument air system and the containment instrument air system.

13. The fire protection system furnishes the capacity to extinguish any probable fires which might occur at Millstone 3. The system includes a water system, a CO2 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 principal 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 storage tanks and to carry condensate from various heating and processing equipment associated with both Unit 2 and Unit 3 during normal plant operations. The system is nonnuclear safety (NNS).

12 REFERENCE FOR SECTION 1.2

figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.

1 COMPARISON WITH SIMILAR FACILITY DESIGNS cipal features of the design of Millstone 3 at the time of application for an operating license e similar to those that were evaluated and approved by the NRC staff for other reactors under struction, operation, or review. Comparison of notable similarities and differences to North a 1 and 2, Surry 1 and 2, Comanche Peak 1 and 2, W. B. McGuire 1 and 2, Maine Yankee, and jan was provided in a series of comparison tables in this section. This section is retained for orical purposes.

design of this facility was based on proven technology attained during the development, gn, construction, and operation of pressurized water reactors of similar or identical types. The

, performance characteristics, and other information represented a standard design that was ineered for the particular requirements of the utility system and site characteristics.

1.1 Comparison of Nuclear Steam Supply System esign comparison of major parameters or features of Millstone 3 with similar plants was ented in Table 1.3-1. The following plants were used in the comparison:

1. Comanche Peak Units 1 and 2, Docket Number 55-445, -446

2. W.B. McGuire Units 1 and 2, Docket Number 50-369, -370

3. Trojan, Docket Number 50-344 1.2 Comparison of Engineered Safety Features le 1.3-2 compared the design data for the Millstone 3 engineered safety features (ESF) ems with similar systems in Trojan (for NSSS scope of supply - ECCS), and North Anna er Station Units 1 and 2 (for balance of plant). The ESF systems compared were the rgency core cooling, containment depressurization, hydrogen recombiner, and supplementary collection and release systems.

se units were chosen for comparison because Millstone 3 systems were similar in design.

jan and North Anna 1 and 2 obtained operating licenses and are currently in commercial ration.

1.3 Comparison of Containment Concepts le 1.3-3 summarized the design and operating parameters described in Section 6.2.1 for the lstone 3 containment concept. The table provided a comparison with similar data for atmospheric containments from the FSAR for North Anna 1 and 2, and the FSAR for Surry 1

2. These references were selected because the units used the Stone & Webster subatmospheric tainment design. Surry 1 and 2 and North Anna 1 and 2 are in commercial operation.

les 1.3-4 through 1.3-10 compared Millstone 3 instrumentation with similar instrumentation in th Anna 1 and 2. These units were chosen for the comparison because they have rumentation similar in design to that of Millstone 3, they have obtained operating licenses, and currently in commercial operation.

1.5 Comparison of Electrical Systems le 1.3-11 compared the Millstone 3 electrical system parameters with similar systems of lstone 2 and Maine Yankee Atomic Power Station. The electrical systems compared were the kV transmission, ac power, ac vital bus, 125 volt DC, and emergency power systems. These s were chosen for the comparison because they have electrical systems similar in design to lstone 3. They have obtained operating licenses, and they are currently in commercial ration.

1.6 Comparison of Radioactive Waste Systems les 1.3-12, 1.3-13, and 1.3-14 compared the radioactive waste systems features for Millstone 3 h similar systems in North Anna 1 and 2 and Surry 1 and 2. These units were chosen for the parison, because they had radioactive waste systems similar in design to Millstone 3, have ined operating licenses, and are currently in commercial operation.

1.7 Comparison of Other Reactor Plant Systems le 1.3-15 summarized the final design and operating parameters for the major reactor plant ems. This table compared the Millstone 3 data with systems data in similar nuclear power ts (North Anna 1 and 2).

2 COMPARISON OF FINAL AND PRELIMINARY DESIGNS le 1.3-16 detailed the significant design changes that were made since the submittal of the R, through the issuance of the operating license.

Millstone 3 FSAR Chapter/

Parameter or Feature Section Millstone 3 Comanche Peak W. B. McGuire Troja Reactor Core Heat Output (MWt) 4, 5, 15.0 3,411 3,411 3,411 3,41 Minimum DNBR for Design Transients 4.1, 4.4, 15.0 > 1.30 > 1.30 > 1.30 > 1.

Total Thermal Flow Rate (10 lb/hr) 4.1, 4.4, 5.1 140.8 140.3 140.3 132.

Reactor Coolant Temperatures (°F) 4.1, 4.4 Core Outlet 620.6 620.8 620.8 619.

Vessel Outlet 617.2 618.2 618.2 616.

Core Average 590.5 589.4 589.4 585.

Vessel Average 587.1 588.2 588.2 584.

Core Inlet 557.0 558.1 558.1 552.

Vessel Inlet 557.0 558.1 558.1 552.

Average Linear Power (kW/ft) 4.1, 4.4 5.44 5.44 5.44 5.44 Peak Linear Power for Normal Operation (kW/ft) 4.1, 4.4 12.6 12.6 12.6 13.6 Heat Flux Hot Channel Factor, FQ 4.1, 4.4, 15.0 2.32 2.32 2.32 2.50 Fuel Assembly Array 4.1, 4.3 17 x 17 17 x 17 17 x 17 17 x Number of Fuel Assemblies 4.1, 4.3 193 193 193 193 Uranium Dioxide Rods Per Assembly 4.1, 4.3 264 264 264 264 Fuel Weight as Uranium Dioxide (lb) 4.1, 4.3 222,739 222,739 222,739 222,739

Millstone 3 FSAR Chapter/

Parameter or Feature Section Millstone 3 Comanche Peak W. B. McGuire Troja Number of Grids Per Assembly 4.1, 4.3 8-Type R 8-Type R 8-Type R 8-Type Rod Cluster Control Assemblies 4.1, 4.3 Number of Full/Part Length 61/- 53/- 53/8 53/8 Absorber Material Hf Hf Ag-In-Cd/B4C

• Ag-In-C Clad Material SS ** SS SS SS Clad Thickness (inches) 0.0185 0.0185 0.0185 0.0185 Equivalent Core Diameter (inches 4.1, 4.3 132.7 132.7 132.7 132.

Active Fuel Length (inches) 4.1, 4.3 144 143.7 143.7 143.

Fuel Enrichment (Weight Percent) 4.1, 4.3 Unit 1 Unit 2 Region 1 2.40 1.60 1.40 2.10 2.10 Region 2 2.90 2.40 2.10 2.60 2.60 Region 3 3.40 3.10 2.90 3.10 3.10 Number of Coolant Loops 5 4 4 4 4 Total Steam Flow (10 lb/hr) 5.1 15.05 15.14 15.14 15.0 Reactor Vessel 5.3 Inside Diameter (inches) 173 173 173 173 Inlet Nozzle Inside Diameter (inches) 27.5 27.5 27-1/2 27.5 Outlet Nozzle Inside Diameter (inches) 29 29 29 29

Millstone 3 FSAR Chapter/

Parameter or Feature Section Millstone 3 Comanche Peak W. B. McGuire Troja Number of Reactor Closure Head Studs 54 54 54 54 Reactor Coolant Pumps 5.4.1 Horsepower 7,000 7,000 7,000 6,00 Capacity (gpm) 100,400 99,000 99,000 88,5 Head (feet) 289 288 288 277 Steam Generators 5.4.2 Model F D D 51 Heat Transfer Area (ft2) 55,000 48,300 48,000 51,5 Number of U-Tubes 5,626 4,578 4,674 3,38 Residual Heat Removal 5.4.7 Initiation Pressure (psig) 425 425 425 400 Initiation/Completion Temperature (°F) 350/120 350/140 350/140 350/

140 Component Cooling Water Design Temperature 95 105 95 95

(°F)

Cooldown Time After Initiation (hr) 20 16 16 16 Heat Exchanger Removal Capacity (10 Btu/hr) 35.27 39.1 34.15 34.2 Pressurizer 5.4.10

Millstone 3 FSAR Chapter/

Parameter or Feature Section Millstone 3 Comanche Peak W. B. McGuire Troja Heatup Rate Using Heaters (°F/hr) 55 55 55 55 Internal Volume (ft3) 1,800 1,800 1,800 1,80 Pressurizer Safety Valves 5.4.13 Number 3 3 3 3 Maximum Relieving Capacity (lb/hr) 420,000 420,000 420,000 420,000 Accumulators 6.3 Number 4 4 4 4 Operating Pressure, Minimum (psig) 600 600 600 600 Minimum Operating Water Volume Each (ft3) 950 950 950 870 Centrifugal Charging Pumps 6.3 Number 3 2 2 2 Design Flow (gpm) 150 150 150 150 Design Head (feet) 5,800 5,800 5,800 5,80 Safety Injection Pumps 6.3 Number 2 2 2 2 Design Flow (gpm) 425 425 425 425 Design Head (feet) 2,680 2,680 2,500 2,50

Millstone 3 FSAR Chapter/

Parameter or Feature Section Millstone 3 Comanche Peak W. B. McGuire Troja Residual Heat Removal Pump 5.4.7, 6.3 Number 2 2 2 2 Design Flow (gpm) 4,000 3,800 3,000 3,00 Design Head (feet) 350 350 375 375 Instrumentation and Controls 7 *** *** *** ***

Chemical and Volume Control 9.3.4 Total Seal Water Supply Flow Rate, Nominal 32 32 32 32 (gpm)

Total Seal Water Return Flow Rate, Nominal 12 12 12 12 (gpm)

Letdown Flow, Normal/Maximum (gpm) 75/120 75/120 75/120 75/1 Charging Flow, Normal/Maximum (gpm) 55/100 55/100 55/100 55/1 NOTE:

• The Ag-In-Cd on Unit 1, B4C on Unit 2

• • SS = Stainless steel

• • • The instrumentation and control systems discussed in Chapter 7 for Millstone 3 are functionally similar to those systems implemented in Comanche Peak, W. B. McGuire, and Trojan.

ergency Core Cooling System (Section 6.3) Millstone 3 Trojan arging Pump ed for high pressure safety injection)

Number 3 2 Design capacity each (gpm) 150 150 Design total developed head (feet) 5,800 5,800 ety Injection Pump (used for intermediate ssure safety injection)

Number 2 2 Design capacity each (gpm) 425 425 Design total developed head (feet) 2,500 2,500 sidual Heat Removal pump ed for low pressure safety injection)

Number 2 2 Design capacity each (gpm) 4,000 3,000 Design total developed head (feet) 350 375 fety Injection Accumulator Number 4 4 Total volume each (ft3) 1,350 1,350 Water volume (ft3, minimum) 950 870 Operating pressure (psig, minimum) 600 600 ntainment Depressurization Systems Millstone 3 North Anna 1 and 2 ction 6.2.2) ench Spray Pump Number 2 2 Design capacity each (gpm) 4,000 2,000 Design total developed head (feet) 291 265 Containment Recirculation Pump

Number 4 outside 2 out / 2 in a containment Design capacity each (gpm) 3,950 3,700/3,300 Design total developed head (feet) 342 287/269 Millstone 3 North Anna 1 and 2 ntainment Recirculation Cooler Number 4 4 UA per cooler (Btu/hr-°F) 3.865 x 106 3.79 x 106 Recirculation flow (gpm) 3,950 3,500 Service water flow (gpm) 6,500 4,500 fueling Water Storage Tank Volume (gal) 1,206,556 480,000 Temperature (°F, maximum) 75 50 drogen Recombiner System (Section 6.2.5)

Number 2 2 Flow rate (scfm, each) 50 50 pplementary Leak Collection and Release System ction 6.2.3)

Number of filter trains 2 None Flow rate (scfm, each) 9,700 None

a. Out - outside containment In- inside containment

(Section 6.2.1)

North Anna Surry Millstone 3 1 and 2 1 and 2 pe Subatmospheric Subatmospheric Subatmospheric (feet) 140 126 126 erall height (feet) 200 191 173 e volume (ft3) 2.3 x 106 1.825 x 106 1.73 x 106 ximum design pressure (psig) 45 45 45 sign temperature (°F) 280 280 280 lculated peak pressure psig) 38.49 44.1 a 44.98 b actor Coolant System:

uid volume 11,695 9,874 9,874 cluding pressurizer) (ft3) mperature (mass average) (°F) 583.5 586.8 574.5 ncrete Thickness:

ertical wall 4 feet 6 inches 4 feet 6 inches 4 feet 6 inches ome 2 feet 6 inches 2 feet 6 inches 2feet 6 inches ntainment structure leak rate 0.30 (0-24 hrs) 0.1 0.1

/day) 0.15 (24-720 hrs) 0.0 0.0

a. Based on Tagami condensing heat transfer coefficient

b. Based on Uchida condensing heat transfer coefficient

SENSOR PARAMETERSA North Anna Pressure Sensors Millstone 3 1 and 2 ntainment Atmosphere High Pressure Transmitter (Millstone 3, gh-1)

Number of channels 3 3 Logic matrix 2/3 2/3 Approximate setpoint (psia) 19.7 15.0 ntainment Atmosphere Intermediate High-High Pressure ansmitter (Millstone 3; High-2)

Number of channels 3 3 Logic matrixÅ 2/3 2/3 Approximate setpoint (psia) 19.7 20 ntainment Atmosphere High-High Pressure Transmitter illstone 3; High-3)

Number of channels 4 4 Logic matrix 2/4 2/4 Approximate setpoint (psia) 24.7 25.0

a. The numbers stated for components or system performance do not represent the maximum/minimum acceptable or required values to support system operation. Setpoints are stated in the Technical Specifications.

BLE 1.3-5 COMPARISON OF REACTOR COOLANT PUMP BUS PROTECTION A Millstone 3 North Anna 1 and 2 dervoltage Reactor Coolant Pumps Buses Number of channels N/A 3 Logic matrix N/A 2/3 Approximate Setting (V) N/A 70% of 4,160 (2,912) derfrequency Reactor Coolant Pumps Buses Number of channels N/A 3 Logic matrix N/A 2/3 Approximate Setting (Hz) N/A 54-59 actor Coolant Pump Shaft Low-Low Speed Number of channels 4 N/A Logic matrix 2/4 N/A Approximate Setting (rpm) (later) N/A actor Coolant Pump Shaft Low Speed Number of channels N/A N/A Logic matrix N/A N/A Approximate Setting (rpm) N/A N/A

a. The numbers stated for components or system performance do not represent the maximum/minimum acceptable or required valves to support system operation. Settings are stated in the Technical Specifications.

SIGNALS Signal Actuation Millstone 3 North Anna 1 and 2 ety Injection Signal (SIS)

Low pressurizer pressure coincident with No No low pressurizer level Low pressurizer pressure Yes Yes High main steam line differential pressure No Yes Low steam line pressure Yes No High main steam flow coincident with low No Yes main steam pressure or low temperature average High containment atmosphere pressure Yes Yes Manual initiation Yes Yes ntainment Isolation Phase A (CIA) Signal Safety injection signal (SIS) Yes Yes Manual initiation Yes Yes ntainment Isolation Phase B (CIB) Signal Containment Depressurization Actuation DA) Signal High-High containment atmosphere Yes Yes pressure (Millstone High-3)

Manual initiation Yes Yes am Line Isolation Signal High main steam flow coincident with low No Yes steam pressure or low main reactor coolant temperature average High-High containment pressure No No Intermediate High-High containment Yes Yes atmosphere pressure (Millstone; High-2)

High steam pressure rate Yes No Low steamline pressure Yes No Manual initiation Yes Yes

GENERATOR AUXILIARY FEEDWATER PUMP START SIGNALS Component Millstone 3 North Anna 1 and 2 ergency generator auto Emergency bus under-voltage Emergency bus under-voltage rt signals Safety injection signal (SIS) SIS am generator auxiliary All steam generator feedwater dwater pump pumps tripped otor-driven) 2/4 - lo-lo level trip any steam 2/3 - lo-lo level trip any steam generator (1/4 matrix) generator (2/3 matrix) coincident with reactor coolant coincident with reactor coolant loop cold leg stop valve open loop hot leg stop valve open or reactor coolant loop cold leg stop valve open Sequenced safeguards signal Undervoltage reserve station service power SIS SIS am generator auxiliary 2/3 - undervoltage on station dwater pump (turbine- service bus ven) auto start signals 2/4 - steam generators lo-lo 2/3 - steam generator lo-lo level (2/4 matrix) coincident level trip (2/3 matrix) with reactor coolant loop cold coincident with reactor coolant leg stop valve open loop hot leg stop valve open or reactor coolant loop cold leg stop valve open TE:

numbers stated for components or system performance do not represent the maximum/

imum acceptable or required values to support system operation.

MONITORING SYSTEMS Number of Locations Monitor Millstone 3 North Anna 1 and 2 rated vent particulate 0 1 rated vent gas 0 1 ntilation vent particulate 0 1 ntilation vent gas 1 1 ntilation vent high range (particulate and gas) 1 0 drogenated vent 1 N/A pplementary leak collection 1 0 ndenser air ejector 1 1 ntainment recirculation cooler service water outlet 2 4 mponent cooling heat exchanger service water 0 1 charge uid waste 1 1 am generator blowdown sample 1 3 xiliary condensate 1 0 rbine building floor drains 1 0 actor plant component cooling water subsystem 1 1 actor Coolant Letdown: Gross activity 0 2 actor Coolant Letdown: Specific fission product 0 0 ivity culating water discharge 0 1 uid waste evaporator 0 1 rvice water discharge 0 1 vice water reservoir 0 1 ntrol building inlet 2 0 generant evaporator (removed from service) 1 0 aste neutralizing sump 1 0 am line monitors 5 0

Number of Locations Monitor Millstone 3 North Anna 1 and 2 ntainment structure low range 4 1 ntainment structure high range 4 1 nipulator crane 1 1 ore instrumentation transfer area 1 1 contamination area 1 1 w fuel storage area 1 1 ent fuel pool pit bridge and hoist 1 1 xiliary building control area 0 1 mple room 1 1 ntrol room 1 1 boratory (Service building) 1 1 xiliary building general area 8 0 uipment decontamination area (Service Building) 1 0

SYSTEMS Number of Locations Monitor Millstone 3 North Anna 1 and 2 xiliary building lower levels particulate 2 0 xiliary building lower levels gas 2 0 xiliary building upper levels particulate 3 0 xiliary building upper levels gas 3 0 arging pumps cubicles particulate 1 0 arging pumps cubicles gas 1 0 el building particulate 1 0 el building gas 1 0 F building particulate 1 0 F building gas 1 0 aste building particulate 1 0 aste building gas 1 0 ntrol room particulate 1 0 ntrol room gas 1 0 ntainment structure particulate 1 1 ntainment structure gas 1 1 ntilation vent sample particulate a 1 1 ntilation vent sample gas 1 1 ak collection area gas 1 0 ak collection area particulate 1 0

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.

Systems and Components Millstone 3 Millstone 2 Maine Yan TRANSMISSION SYSTEM (Section 8.2)

Transmission Lines to Circuits 4 at 345 kV 3 at 345 kV 2 at 345 kV (Total for 3 Units) (Total for 2 Units) 2 at 115 kV AC POWER SYSTEMS (Section 8.3.1)

Main transformer 2 at 630 MVA 1 at 945 MVA 2 at 430 MVA Normal station service transformer 1 at 50 MVA (6.9 kV) 1 at 45 MVA 1 at 30 MVA 1 at 40 MVA (4.16 kV) 1 at 20 MVA Reserve station service transformer 1 at 50 MVA (6.9 kV) 1 at 45 MVA 1 at 30 MVA 1 at 45 MVA (4.16 kV) 1 at 20 MVA AC VITAL BUS SYSTEM (Section 8.3.1)

Distribution Cabinets 6 4 4 Inverters 4 at 25 kVA 4 at 15 kVA 4 at 10 kVA 1 at 30 kVA 1 at 60 kVA 125-V DC SYSTEM (Section 8.3.2)

Unit batteries (125V) 2 at 1650 Ah 2 at 750 Ah 2 at 2300 Ah 2 at 2550 Ah 1 at 1200 Ah 2 at 1800 Ah 4 at 200 amp 6 at 400 amp 4 at 250 amp

Systems and Components Millstone 3 Millstone 2 Maine Yan 2 at 50 amp 3 at 200 amp (installed spares)

EMERGENCY POWER SYSTEM (Section 8.3.1)

Emergency generator (continuous rating) 2 at 4986 kW 2 at 2750 kW 2 at 2500 kW Emergency 4.16 kV Buses 2 at 2000/3000 amp 2 at 2000 amp 2 at 1200 amp

Millstone 3 North Anna 1 & 2 Surry 1 and 2 Type of Processing Evaporation Yes Yes Yes Demineralization Yes Yes Yes Filtration Yes Yes Yes Treatment of Radioactive Wastes High Activity Wastes Evaporation and/or demineralization Same Same and filtration, if required Low Activity Wastes Including Filtration (evaporation and Same Same Contaminated Shower Drains subsequent operations are optional)

Regenerant Chemical Waste Filtration if required None None Steam Generator Blowdown Blowdown piped to flash tank where Blowdown is cooled and sent Blowdown is cooled an steam is drawn off to 4th point heater to clarifier for flocculation, then either released or and liquid is drawn to main condenser sedimentation, and filtration. demineralized and filte for treatment by the condensate The sediment is sent to the and recycled to the ma demineralizers solid waste disposal system condenser and the liquid discharged through the liquid waste monitor to the circulating discharge tunnel.

Equipment High Level Waste Drain Tanks Number 2 2 2

Millstone 3 North Anna 1 & 2 Surry 1 and 2 Capacity (gal/each) 26,000

• 5,000 2,390 Low Level Waste Drain Tanks 2 2 2 Capacity (gal/ea) 4,000

• 5,000 2,874 Regenerant Evaporator Feed Tanks N/A N/A (removed from service)

Number 2 Capacity (gal/each) 13,500 Contaminated Shower Drain Tank Number 0 2 2 Capacity (gal/each) 1,400 (including 1,230 (including Laundry) Laundry)

Waste Test Tank Number 2 2 2 Capacity (gal/each) 24,000 1,500 548 Regenerant Chemical Evaporator N/A N/A (removed from service)

Number 1 Trays, number 8 Capacity (gal/each) 35 Waste Evaporator

Millstone 3 North Anna 1 & 2 Surry 1 and 2 Number 1 1 1 Trays, number 8 0 0 Capacity (gal/each) 35 6 6 Demineralizers Number Two waste evaporators, one One waste evaporator One waste evaporator regenerant chemical evaporator distillate polishing, two distillate polishing (removed from service) clarifier demineralizers Type Mixed bed Mixed bed Mixed bed Capacity (ft3) 35 17/45 respectively 17 Effluent Filters Number 2 2 1 Capacity (gpm/each) 50 50 75 Filter element type Wound fiber Wound cotton fiber Wound synthetic fi NOTE:

• Include condensate demineralizer, liquid waste system tank volumes. The numbers stated for components or systems performan do not represent the maximum/minimum acceptable or required values to support system operation.

Millstone 3 North Anna 1 and 2 Surry 1 and 2 Type of Treatment Degasification Yes Occurs in boron recovery Occurs in boron recovery system system Decay of noble gases in high activity gas Yes Yes Yes stream Filtration of low activity gas streams Gas streams Yes Yes recombiners No Recombiners Streams Yes Yes No Treatment of Streams Continuous degasification of reactor Yes Yes Yes letdown Degasification of letdown to boron Yes Yes Yes recovery system Decay method for gases stripped in Adsorption on charcoal for Recombination of hydrogen Recombination of hydrogen degasifier minimum 60-day xenon stream for reduction storage stream for reduction storage decay before recycle or in gas decay tank, then in gas decay tanks before release to the environment charcoal filtration before release to the environment release to atmosphere Low activity air streams Release through Millstone Filtration (charcoal and Filtration (charcoal and (nonventilation streams) stack HEPA filters) and release particulate filters) and thru vent on top of Unit 1 release to environment containment

Millstone 3 North Anna 1 and 2 Surry 1 and 2 Equipment Degasifier Number 1 2 1 (common to both units)

Capacity (gpm/ea) 150 240 Process Gas Compressor Number 2 2 2 Capacity (scfm/ea) 3 1.5 2.5 (Currently abandoned in place)

Process Gas Charcoal Absorbers Number 2 0 0 Capacity - Charcoal (lb/ea) 13,500 - -

Gas Decay Tanks Number 0 2 2 Capacity (ft3/ea) - 462 434 Pressure (psig) - 115 - the inside tank 115 4 - the outside tank Gas Surge Tank (Process Gas Receiver)

Number 1 1 1 Capacity (ft3) 10 15 15.7 Waste Gas Recombiners

Millstone 3 North Anna 1 and 2 Surry 1 and 2 Number 0 1 1 Capacity (scfm) 1.5 1.31 Pressure (psia) 14 22 Waste Gas Compressors Number 0 2 2 Capacity (scfm/ea) 0 1.5 1.5 Pressure (psig) 150 120 Filter for low activity aerated activity None Activated charcoal with Activated charcoal with aerated gaseous waste effluents HEPA after-filter HEPA after-filter

Millstone 3 North Anna 1 and 2 Surry 1 and 2 Type of Treatment Solidification/Dewatering Yes Yes Yes Drumming Yes Yes Yes Solidification agent Cement Urea formaldehyde Cement Inputs (Type) Treated Boron evaporator bottoms Yes Yes Yes Waste evaporator bottoms Yes Yes Yes Regenerant evaporator bottoms Evaporator removed from N/A N/A service Filters Yes Yes Yes Resins Yes Yes Yes Miscellaneous wastes (contaminated Yes Yes Yes clothing, tools, paper products, etc.)

Equipment Spent resin hold tank:

Number 1 1 1 Capacity (gal) 3,000 1,800 2,019 Spent resin dewatering tank:

Number 1 1 1 Capacity (gal) 500 500 619 Spent resin transfer pump filter:

Millstone 3 North Anna 1 and 2 Surry 1 and 2 Number 1 1 1 Capacity (gpm) 150 100 150 Type Wound fiber Wound fiber Wound fiber Evaporator bottoms tank:

Number 1 1 N/A Capacity (gal) 3,150 1,100 Shipping container (with cask):

Capacity 50 ft3 50 ft3 55 gallon drum Type All solid waste is 50 ft3 or All solid waste is shipped All waste in 55 gallon greater shipping containers in 50 ft3 containers except drums compressible waste which is in 55 gallon drums

Operating Parameters Systems with Components Millstone 3 North Anna 1 and 2 Fuel Pool Cooling and Purification System (Section 9.1.3)

Fuel Pool Cooling Pumps:

Number 2 2 Design capacity (gpm) 3,500 2,750 Design total head (ft) 92 80 Fuel Pool Coolers:

Number 2 2 Duty per heat exchanger (Btu/hr) 27,700,000 56,800,000 Fuel pool cooling flow (gpm) 3,500 2,750 Component cooling flow (gpm) 1,800 3,350 Number of cores cooled 15-1/2 (3048 Fuel Assemblies) 1-1/3 Fuel pool temperature, normal (°F) 150 140 Reactor Plant Component Cooling System (Section 9.2.2.1)

Reactor Plant Component Cooling Pumps:

Number 3 4 Design capacity (gpm) 8,100 8,000 Design total head (ft) 284 190

Operating Parameters Systems with Components Millstone 3 North Anna 1 and 2 Reactor Plant Component Cooling Heat Exchangers:

Number 3 2 Duty per unit exchanger (Btu/hr) 76,000,000 52,000,000 Reactor plant component cooling water flow (gpm) 8,100 9,000 Service water flow (gpm) 8,000 10,500 System Design Basis (safety related) Reactor cooldown to 120°F in 24 hr Reactor cooldown to 140°F in 16 hr Duty per unit exchanger (Btu/hr) 76,000,000 52,000,000 Reactor plant component cooling water flow (gpm) 8,100 9,000 Service water flow (gpm) 8,000 0,500 System Design Basis (safety related) Reactor cooldown to 120°F in 24 hr Reactor cooldown to 140°F in 16 hr Service Water System (Section 9.2.1)

Service Water Pumps:

Number 4 6 Design capacity (gpm) 15,000 11,500 Design total head (ft) 120 127 Design temperature, maximum (°F) 80 95 Boron Recovery System (Section 9.3.2)

Type of Treatment:

Operating Parameters Systems with Components Millstone 3 North Anna 1 and 2 Degasification of liquid entering liquid entering system Occurs in radioactive gaseous Yes waste system Storage of liquid prior to processing Yes Yes Evaporation Yes, in boron evaporator Yes, in boron evaporator Demineralization of boron evaporator distillate Yes, if required Yes, if required System Effluents:

Boron evaporator distillate Primarily recycled to primary grade Primarily recycled to primary grade water storage tank; remainder to water storage tank; remainder to radioactive liquid waste system for radioactive liquid waste system for discharge discharge Boron evaporator bottoms Recycled to boric acid tanks or Recycled to boric acid tanks or drummed in radioactive solid waste drummed in radioactive solid waste system system Cesium Removal Ion Exchanger:

Number 2 2 Resin Capacity (ft3) 35 45 Boron recovery tanks:

Number 2 3 Capacity (gal) 150,000 120,000 Boron Evaporator Subsystem:

Operating Parameters Systems with Components Millstone 3 North Anna 1 and 2 Evaporator type Forced circulation Forced circulation Number 1 2 (for 2 units)

Reboiler External External Trays 8 5 Capacity (gpm) 25 20 Boron Test Tanks:

Number 2 2 Capacity (gal) 12,000 20,000 Boron Demineralizers:

Number 2 2 Resin capacity (ft3) 35 45

PSA FSAR Section Significant Changes Since PSAR Refere 2.5.3 Thirteen faults have been uncovered in the rock excavation mapping since November 1979. 2.5.3 3.1.1.3 Application of the single failure criterion has been upgraded to be consistent with regulatory 3.1.1.2 requirements and industry standards and practices.

3.3.2 Postulated tornado missiles and their characteristics have been revised to conform to 3.3.2 Regulatory Guide 1.76 dated April 1974.

3.5.1.4 Conformance to Regulatory Guide 1.117, Rev. 1, dated April 1978. Have revised the 3.5.1.4 spectrum of tornado generated missiles used for plant design.

3.7B.4.2 The number of seismic instrumentation packages has been increased and those locations 3.7.4 revised.

3.8.1 Structural ring added around the containment structure to maintain isolation from the 3.8.2.1 surrounding rock.

3.8.4 1. Revised plan and design of waste disposal building to conform to Regulatory Guide 1.143, 3.8.2 Rev. 1, dated October 1979.

2. Description of railroad canopy adjacent to fuel building added. 3.8.1 3.11 Class IE electrical equipment has been qualified to IEEE 323-1974 and IEEE 344-1975 3.11 requirements.

5.2.4 Inservice methods have been changed from closed circuit television monitors to ultrasonic 5.2.8 scanners in the reactor vessel.

5.2.5 Air cooler outlet temperature and reactor coolant system make up rates are no longer 5.2.7 monitored.

5.4.2 Model D steam generators have been replaced by Model F steam generators. 5.4.2

PSA FSAR Section Significant Changes Since PSAR Refere 5.4.7 Contents have been revised to include cold shutdown requirements of Regulatory Guide 5.5.7, 1.139, Rev. 0, date May 1978.

6.3 6.3 6.2.1.7 Temperature monitoring for containment sump is not Q.A. Category I as implied in PSAR 7.5.3.2 Section 7.5.3.2.

6.2.6.2 Added description of electrical penetrations leakage rate tests. 6.2.6.4 6.3 Boron injection tank (BIT) has been eliminated. 6.3 6.4 1. Pressurization system split into two banks of 100% capacity each. 6.4

2. Deletion of vinyl chloride detection.

3. Air-conditioning units which serve areas outside the control room pressure envelope have been relocated from this control room area to outside the control room pressure envelope.

6.4.6 Only one smoke detector is provided on the control building ventilation inlet. The response to Response PSAR Question 9.57 stated redundant detectors would be provided. PSAR Question 7.3.1, 1. Change in pressurizer pressure required for actuation of safety injection signal (SIS). 7.3.4.3, 1.3.1.4 1.3.1.4

2. Change in steamline and containment pressure required for actuation of steamline isolation signal.

7.4 Millstone 3 design now has the capability for a safety grade cold shutdown from the auxiliary 7.4, shutdown panel.

3.1.2.19 3.1.2.19 7.8 Addition of ATWS Mitigation System Actuation Circuitry, conformance to 10 CFR 50.62. None 8.3.1 1. Millstone 2 electrical power tie has been deleted. 8.3.1

PSA FSAR Section Significant Changes Since PSAR Refere

2. 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 been 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 circumstances, 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 motor control centers is no longer required.

Instead these nonsafety circuits are considered as safety circuits within this equipment and subsequently changed to nonsafety status after going through qualified isolation devices.

8.3.2 1. Addition of a second 125V battery and its associated equipment to the normal (nonsafety 8.3.2 related) dc power system.

2. 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. 9.1.2 9.2.1 Addition of motor control center (MCC) and rod drive area air-conditioning (a/c) units as 9.2.1 loads to the service water system (SWP). This addition requires two booster pumps to meet flow requirements to the a/c units.

9.2.2.1 1. The component cooling system no longer supplies cooling water to the reactor coolant pump 9.2.2.1 (RCP) motor air cooler and the control rod drive mechanism (CRDM) shroud coolers. Loads transferred to chilled water system.

PSA FSAR Section Significant Changes Since PSAR Refere

2. Containment air recirculation coolers are supplied with chilled water except during a CIA or LOP when component cooling water is automatically supplied. They were previously supplied with component cooling water during winter conditions when there were no thermal bypass lines around the reactor plant 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. 9.2.2.2

2. 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 system (CCP) - chilled water system (CDS) interaction.

9.2.3 Ultrafiltration system (UF modules, cleaning solution skid, pH solution skid, cartridge filter, 9.2.7 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 forwarded to the demineralizer system.

9.2.4 Hydropneumatic pumps deleted. Electric hot water in lieu of steam. Domestic water service 9.2.6 throughout plant.

9.2.8 Addition of 200 gpm deaerator system to the primary water system to lower oxygen level in 9.2.4 the primary water to comply with Westinghouse requirements.

9.3.1.1 Addition of two shutdown air compressors to improve plant operability following a loss of 9.3.3.1 power.

9.4.0 Air-conditioning units which serve areas outside the control room pressure envelope have 9.4.1 been relocated from this control room area to outside the control room pressure envelope.

PSA FSAR Section Significant Changes Since PSAR Refere 9.4.1.5 1. The control building isolation signal (CBI) has been changed; it is no longer initiated by all 9.4.1.5 safety injection signals.

2. Both trains of control building air-conditioning do not start automatically 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 block 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 open to the atmosphere. The expansion tank is manually filled and has an overflow connection. Therefore, a high level alarm is not required.

9.4.1 1. Manual diversion of exhaust air systems to filtration units occurs on receipt of high 9.4.5 radioactivity or CIA signal.

2. Addition of safety related dampers in supply air ductwork.

9.4.5 1. The emergency generator enclosure ventilation system uses supply fans, not exhaust fans. 9.4.10

2. Fans are single speed instead of two-speed.

3. Air in the emergency generator enclosure can now be recirculated.

9.4.8 1. The auxiliary building filter system will filter the waste disposal building exhaust air instead 9.4.3 of main filter bank of the SLCRS.

2. The air supply equipment for the waste disposal building has been relocated from the auxiliary building to the roof of the waste disposal building.

PSA FSAR Section Significant Changes Since PSAR Refere

3. The waste disposal building ventilation system has been redesigned to reflect the new building layout. The system now utilizes three 50-percent units rather than two 100-percent units.

9.4.10 Addition of hydrogen recombiner heating, ventilation and air-conditioning system. None 9.5.1 and Change in scope attributed to increased level of fire protection design and analysis as 9.5.1 documented in the Millstone 3 Fire Protection Evaluation, June 1977 as amended.

Fire Protection Evaluation Report (June 1977) 9.5.4.5 1. Fuel oil storage tank: local fuel oil level indication deleted. (Remote indication on emergency 9.5.4.5 generator panel).

2. Manual control changed from main control board to emergency generator panel.

3. Fuel oil transfer pump discharge: Local pressure 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 10.3 pump turbine.

2. Addition of bypass valves around the main steam pressure relieving valves.

10.4.3 The PSAR stated that the exhaust from the gland seal condenser exhaust would be passed 11.3, through a charcoal filter and released from the auxiliary boiler blowdown vent stack. Now Figure 11.

there is no charcoal filter and the release is from the turbine building roof.

PSA FSAR Section Significant Changes Since PSAR Refere 10.4.5 A circulating water pump is not automatically tripped in the event of low flow in the Response associated condenser discharge line coincident with high level in the condenser discharge PSAR strainer pit. Manual shutdown was selected as the alternative to automatic controls as a more Questions dependable method to prevent flooding in the turbine building in the event of a condenser 10.2, 10.8 expansion joint failure. Circulating water flows are 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.

10.4.8 1. System redesigned with a single blowdown tank and no longer contains a blowdown tank 10.4.6 condenser.

2. Blowdown liquid drains to the steam generator blowdown tank. The steam generator blowdown tank drains to the condenser (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 cavitating venturies to limit flow to faulted steam 10.4.3 generator.

2. Addition of service water system 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. 11.2.5 11.4 The radwaste solidification system has been changed from an urea formaldehyde system to a 11.5.6 Dow polymer solidification system.

11.5 Addition of radiation monitors for post accident monitoring. 11.4 14.2 Entirely rewritten to conform to Regulatory Guide 1.70, Rev. 3. 14.2

PSA FSAR Section Significant Changes Since PSAR Refere 17.1.2 The Stone & Webster QA Program was revised on March 31, 1975 to conform to NRC 17.1.2 approved Stone & Webster Topical Report SWSQAP 1-74A, Revision N/A, dated December 31, 1974.

17.1.3 The original QA Program implemented by WNES for Millstone 3 was described in Chapter 17.1.3 17 of the Millstone 3 PSAR. During the design and initial procurement activities for Millstone 3, the upgrading of the WNES QA Program reflected changes in regulatory requirements and industry standards. These changes have culminated in the present WNES QA Program as presented in WCAP-8370, Revision 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.

1 LICENSEE'S SUBSIDIARIES minion Nuclear Connecticut, Inc. (DNC) is responsible for the operation, maintenance, and ing of Millstone 3. DNC is an indirect wholly owned subsidiary of Dominion Energy, which is urn wholly owned by Dominion Resources, Inc.

2 ARCHITECT-ENGINEER ne & Webster Engineering Corporation (SWEC) in Boston, Massachusetts, provided ineering design and construction management services for Millstone 3. SWEC is an ineering and construction firm serving the electric utility industry in the design and struction of all types of power stations.

3 NUCLEAR STEAM SUPPLY SYSTEM MANUFACTURER stinghouse Electric Corporation (Westinghouse) was responsible for supplying the NSSS and fuel load for Millstone 3.

stinghouse has designed, developed, and manufactured nuclear facilities since the 1950s, inning with the world's first large central station nuclear power plant (Shippingport), which produced power since 1957. Completed or presently contracted commercial nuclear capacity ls in excess of 97,000 MW. Westinghouse pioneered new nuclear design concepts, such as mical shim control of reactivity and the rod cluster control concept, throughout the last two ades. Among the company's own related manufacturing facilities are the Columbia Plant, lear Fuel Division, the largest commercial nuclear fuel fabrication facility in the world, the sacola Plant which fabricates reactor internals and steam generators, and the Cheswick Plant ch produces control rod drive mechanisms and reactor coolant pumps.

4 TURBINE GENERATOR MANUFACTURER turbine generator was manufactured by General Electric Company (GE). Design of the ine generator was under the direction of the Steam Turbine-Generator Products Division ted in Schenectady, New York.

has extensive experience manufacturing turbine generators for nuclear and nonnuclear lications and has supplied them for 36 operating nuclear power plants. These include 11 surized water reactors (PWR), 24 boiling water reactors (BWR), and 1 high temperature gas tor (HTGR). GE is also providing turbine generators for 36 nuclear power plants in various es of construction. These include 19 PWRs and 17 BWRs.

s section has been deleted in its entirety.

le 1.6-1 lists topical reports which provided information that was filed separately with the lear Regulatory Commission (NRC) in support of this and similar applications.

information contained in Section 1.6 is retained for historical purposes. The review status es previously included in this section were presented for information only and were accurate he time of license application. Since the review status codes are not required by Regulatory de 1.70, are no longer applicable, and could lead to misinterpretation, they were removed in 7 as part of a general FSAR upgrade and clarification of information.

NRC Refere Report Number Title Submittal Sectio WCAP-2048 The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel July 1962 4.3 Elements WCAP-2850 Single Phase Local Boiling and Bulk Boiling Pressure Drop Correlations April 1966 4.4 (Proprietary)

WCAP-7916 June 1972 (Non-proprietary)

WCAP-2923 In-Pile Measurement of U02 Thermal Conductivity 1966 4.4 WCAP-3269-8 Hydraulic Tests of the San Onofre Reactor Model June 1964 4.4 WCAP-3269-26 LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM - Sept 1963 4.3, 15.0, 7094 WCAP-3385-56 Saxton Core II Fuel Performance Evaluation, of Mass Spectrometric and Radio- July 1970 4.3, 4.4 Chemical Analyses of Irradiated Saxton Plutonium Fuel WCAP-3680-20 Xenon-Induced Spatial Instabilities in Large PWRs (EURAEC-1974) March 1968 4.3 WCAP-3680-21 Control Procedures for Xenon-Induced X-Y Instabilities in Large PWRs Feb 1969 4.3 (EURAEC-2111)

WCAP-3680-22 Xenon-Induced Spatial Instabilities in Three-Dimensions (EURAEC-2116) Sept 1969 4.3 WCAP-3696-8 Pressurized Water Reactor pH - Reactivity Effect Final Report (EURAEC-2074) Oct 1968 4.3 WCAP-3726-1 PU02 - U02 Fueled Critical Experiments July 1967 4.3 WCAP-6065 Melting Point of Irradiated U02 Feb 1965 4.4

NRC Refere Report Number Title Submittal Sectio WCAP-6069 Burnup Physics of Heterogenous Reactor Lattices June 1965 4.4 WCAP-6073 LASER - A Depletion Program for Lattice Calculations Based on MUFT and April 1966 4.3 THERMOS WCAP-6086 Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Aug 1969 4.3 Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium WCAP-7015 Revision 1 Subchannel Thermal Analysis of Rod Bundle Cores Jan 1969 4.4 WCAP-7048-PA The PANDA Code Jan 1975 4.3 (Proprietary)

WCAP-7757-A (Non-proprietary)

WCAP-7198-L Evaluation of Protective Coatings for Use in Reactor Containment April 1969 6.1N (Proprietary)

WCAP-7825 Dec 1971 (Non-proprietary)

WCAP-7213-P-A The TURTLE 24.0 Diffusion Depletion Code Feb 1975 4.3 (Proprietary)

WCAP-7758-A 15.0, 15.

(Non-proprietary)

WCAP-7308-L Evaluation of Nuclear Hot Channel Factor Uncertainties Dec 1971 4.3 (Proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-7810 (Non-proprietary)

WCAP-7359-L Application of THINC Program to PWR Design Aug 1969 4.4 (Proprietary)

WCAP-7838 Jan 1972 (Non-proprietary)

WCAP-7477-L Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems March 1970 5.2 (Proprietary)

WCAP-7735 Aug 1971 (Non-proprietary)

WCAP-7488-L Solid State Logic Protection System Description March 1971 7.2 (Proprietary)

WCAP-7672 June 1971 7.3 (Non-proprietary)

WCAP-7518-L Radiological Consequences of a Fuel Handling Accident Dec 1971 15.7 (Proprietary)

WCAP-7828 (Non-proprietary)

WCAP-7536-L Seismic Testing of Electrical and Control Equipment (High Seismic Plants) Dec 1971 3.10 (Proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-7821 (Non-proprietary)

Supplements 1-6 WCAP-7558 Seismic Vibration Testing with Sine Beats Oct 1971 3.10 WCAP-7588 Revision An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Jan 1975 15.4 1-A Reactors Using Spatial Kinetics Methods WCAP-7623 Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Dec 1970 5.4 Steel Plate WCAP-7667-P-A Interchannel Thermal Mixing With Mixing Vane Grids Jan 1975 4.4 (Proprietary)

WCAP-7755-A (Non-proprietary)

WCAP-7695-P-A DNB Tests Results for New Mixing Vane Grids (R) Jan 1975 4.4 (Proprietary)

WCAP-7958-A (Non-proprietary)

WCAP-7695 DNB Test Results for R Grid Thimble Cold Wall Cells Jan 1975 4.4 Addendum 1-P-A (Proprietary)

WCAP-7958 Addendum 1-A (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-7706-L An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients Feb 1971 4.6 (Proprietary)

WCAP-7706 7.1 (Non-proprietary) 7.2 WCAP-7769 Overpressure Protection for Westinghouse Pressurized Water Reactors Oct 1971 15.2 WCAP-7798-L Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Jan 1972 6.1N (Proprietary) Accident Environment WCAP-7803 (Non-proprietary)

WCAP-7800 Revision 5 Nuclear Fuel Division Quality Assurance Program Plan Nov 1984 4.2, 17.1 WCAP-7806 Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Dec 1971 4.3 Poison Rods WCAP-7811 Power Distribution Control of Westinghouse Pressurized Water Reactors Dec 1971 4.3 WCAP-7832 Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Dec 1973 5.4 Combined LOCA Plus SSE Conditions WCAP-7836 Inlet Orificing of Open PWR Cores Jan 1972 4.4 WCAP-7870 Neutron Shielding Pads May 1972 3.9N

NRC Refere Report Number Title Submittal Sectio WCAP-7907 LOFTRAN Code Description June 1972 6.3, 15.0 15.1, 15.

15.3, 15.

15.5, 15.

WCAP-7908 FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod June 1972 15.0, 15.

15.3, 15.

WCAP-7912-P-A Power Peaking Factors Jan 1975 4.3 (Proprietary)

WCAP-7912-A 4.4 (Non-proprietary)

WCAP-7913 Process Instrumentation for Westinghouse Nuclear Steam Supply System (4-Loop March 1973 7.3, 7.2 Plant Using WCID 7300 Series Process Instrumentation)

WCAP-7921-AR Damping Values of Nuclear Power Plant Components May 1974 3.7N WCAP-7941-P-A Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Jan 1975 4.4 (Proprietary) Grid WCAP-7959-A (Non-proprietary)

WCAP-7950 Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant July 1972 3.7N Accident WCAP-7956 THINC-IV An Improved Program for Thermal-Hydraulic Analysis of Rod June 1973 4.4 Bundle Cores WCAP-7964 Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor June 1971 4.3

NRC Refere Report Number Title Submittal Sectio WCAP-7979-P-A TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code Jan 1975 15.0 (Proprietary)

WCAP-8028-A 15.4 (Non-proprietary)

WCAP-7988 Application of Modified Spacer Factor to L. Grid Typical and Cold Wall Cell Oct 1972 4.4 (Proprietary) DNB WCAP-8030-A (Non-proprietary)

WCAP-8054 Application of the THINC-IV Program to PWR Design Oct 1973 4.4 (Proprietary)

WCAP-8195 (Non-proprietary)

WCAP-8163 Reactor Coolant Pump Integrity in LOCA Sept 1973 5.4 WCAP-8170 Calculational Model for Core Reflooding After a Loss of Coolant Accident June 1974 6.2 (Proprietary) (WREFLOOD Code)

WCAP-8171 15.6 (Non-proprietary)

WCAP-8174 Effect of Local Heat WCAP-8202 Flux Spikes on DNB in Non-Uniform Heated Aug 1973 4.4 (Proprietary) Rod Bundles WCAP-8202 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8183 Revision 6 Operational Experience With Westinghouse Cores June 1977 4.2 WCAP-8200 Revision 2 WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a July 1974 15.6 (Proprietary) Multi-Loop PWR WCAP-8261 Revision 1 (Non-proprietary)

WCAP-8218-P-A Fuel Densification Experimental Results and Model for Reactor Application March 1975 4.1 (Proprietary)

WCAP-8219-A 4.2, 4.3, (Non-proprietary)

WCAP-8236 Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Jan 1974 3.7N (Proprietary)) Coolant Accident WCAP-8288 Dec 1973 4.2 (Non-proprietary WCAP-8236 Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and March 1974 3.7N Addendum 1 Loss of Coolant Accident (Proprietary)

WCAP-8268 April 1974 Addendum 1 (Non-proprietary)

WCAP-8252 Revision 1 Documentation of Selected Westinghouse Structural Analysis Computer Code July 1977 3.9N WCAP-8253 Source Term Data for Westinghouse Pressurized Water Reactors July 1975 11.1 Amendment 1

NRC Refere Report Number Title Submittal Sectio WCAP-8255 Nuclear Instrumentation System Jan 1974 7.2, 7 WCAP-8264-P-A Westinghouse Mass and Energy Release Data for Containment Design June 1975 6.2.1 (Proprietary)

WCAP-8312-A (Non-proprietary)

Revision 2 WCAP-8278 Hydraulic Flow Test of the 17x17 Fuel Assembly Feb 1974 4.2 (Proprietary)

WCAP-8279 4.4 (Non-proprietary)

WCAP-8296-P-A Effect of 17x17 Fuel Assembly Geometry on DNB Feb 1975 4.4 (Proprietary)

WCAP-8297 (Non-proprietary)

WCAP-8298-P-A The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing Jan 1975 4.4 (Proprietary)

WCAP-8299-A (Non-proprietary)

WCAP-8301 LOCTA-IV Program: Loss of Coolant Transient Analysis June 1974 15.0 (Proprietary)

WCAP-8305 15.6 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8302 SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of June 1974 6.2.1 (Proprietary) Coolant WCAP-8306 15.0, 15.

(Non-proprietary)

WCAP-8303-P-A Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model July 1975 3.9N (Proprietary) Tests WCAP-8317-A (Non-proprietary)

WCAP-8324-A Control of Delta Ferrite in Austenitic Stainless Steel Weldments June 1975 5.2 WCAP-8327 Containment Pressure Analysis Code (COCO) June 1974 15.6 (Proprietary)

WCAP-8326 (Non-proprietary)

WCAP-8330 Westinghouse Anticipated Transients Without Trip Analysis Aug 1974 4.3, 4.6, 15.2, 15.

15.8 WCAP-8339 Westinghouse ECCS Evaluation Model -Summary July 1974 6.2.1 15.6 WCAP-8340 Westinghouse ECCS - Plant Sensitivity Studies July 1974 15.6 (Proprietary)

WCAP-8356 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8341 Westinghouse ECCS Evaluation Model Sensitivity Studies July 1974 15.6 (Proprietary)

WCAP-8342 (Non-proprietary)

WCAP-8359 Effects of Fuel Densification Power Spikes on Clad Thermal Transients July 1974 4.3 WCAP-8370/7800 Westinghouse Water Reactor Divisions Quality Assurance Plan Nov 1984 17.1.

Revision 10A/6A WCAP-8377 Revised Clad Flattening Model July 1974 4.2 (Proprietary)

WCAP-8381 (Non-proprietary)

WCAP-8385 Power Distribution Control and Load Follow Procedures Sept 1974 4.3 (Proprietary)

WCAP-8403 4.4 (Non-proprietary)

WCAP-8424 Revision 1 An Evaluation of Loss of Flow Accidents Caused by Power System Frequency June 1975 15.3 Transients in Westinghouse PWRs WCAP-8446 17x17 Drive Line Components Tests -Phase IB, II, III, D-Loop Drop and Dec 1974 3.9N (Proprietary) Deflection WCAP-8449 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8453-A Analysis of Data from the Zion (Unit 1) THINC Verification Test May 1976 4.4 WCAP-8471 Westinghouse ECCS Evaluation Model -Supplement Information April 1975 15.6 (Proprietary)

WCAP-8472 (Non-proprietary)

WCAP-8472 (Non-proprietary)

WCAP-8498 Incore Power Distribution Determination in Westinghouse Pressurized Water July 1975 4.3 Reactors WCAP-8516-P UHI Plant Internals Vibration Measurement Program and Pre and Post Host April 1975 3.9N (Proprietary) Functional Examinations WCAP-8517 (Non-proprietary)

WCAP-8536 Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry With 22-Inch Grid May 1975 4.4 (Proprietary) Spacing WCAP-8537 (Non-proprietary)

WCAP-8565-P-A Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies July 1975 15.6 (Proprietary)

WCAP-8566-A (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8584 Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Feb 1976 4.6 (Proprietary) Actuation System WCAP-8760 (Non-proprietary)

WCAP-8760 (Non-proprietary)

WCAP-8587 Revision 2 Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Feb 1979 3.10N, 3 Electrical Equipment WCAP-8587 Equipment Qualification Data Packages Nov 1978 3.10N, 3 Supplement 1 WCAP-8622 Westinghouse ECCS Evaluation Model -October 1975 Version Nov 1975 15.6 (Proprietary)

WCAP-8623 (Non-proprietary)

WCAP-8624 General Method of Developing Multi-frequency Biaxial Test Inputs for Bistables Sept 1975 (Proprietary)

WCAP-8695 Aug 1975 (Non-proprietary)

WCAP-8682 Experimental Verification of Wet Fuel Storage Criticality Analyses Dec 1975 4.3 (Proprietary)

WCAP-8683 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8691 Fuel Rod Bowing Dec 1975 4.2 (Proprietary)

WCAP-8692 (Non-proprietary)

WCAP-8693 Delta Ferrite in Production Austenitic Stainless Steel Weldments Jan 1976 5.2 WCAP-8708 MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Feb 1976 3.9N (Proprietary) Thermal-Hydraulic-Structure System Dynamics WCAP-8709 (Non-proprietary)

WCAP-8720 Improved Analytical Models Used in Westinghouse Fuel Rod Design Oct 1976 4.2 (Proprietary) Computations WCAP-8785 (Non-proprietary)

WCAP-8768 Revision 2 Safety-related Research and Development for Westinghouse Pressurized Water Oct 1978 1.5, 4.2, Reactors, Program Summaries-Winter 1977-Summer 1978 WCAP-8780 Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational May 1976 3.9N Tests on Trojan 1 Power Plant WCAP-8846-A Hybrid B4C Absorber Control Rod Evaluation Report Sept 1976 15.3 WCAP-8892-A 7300 Series Process Control System Noise Tests June 1977 7.1 (Non-proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-8929 Benchwork Problem Solutions Employed for Verification WECAN Computer June 1977 3.9N Code WCAP-8963 Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis Nov 1976 4.2 (Proprietary)

WCAP-8964 Aug 1977 (Non-proprietary)

WCAP-8970 Westinghouse ECCS Small Break October 1975 Model April 1977 15.6 (Proprietary)

WCAP-8971 (Non-proprietary)

WCAP-8976 Failure Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Sept 1977 4.6 Control System WCAP-9168 Westinghouse Emergency Core Cooling System Evaluation Model - Modified Sept 1977 15.6 (Proprietary) October 1975 Version WCAP-9169 (Non-proprietary)

WCAP-9179 Revision 1 Properties of Fuel and Core Component Materials Sept 1977 4.2 (Proprietary)

WCAP-9224 July 1978 (Non-proprietary)

WCAP-9220-P-A Westinghouse ECCS Evaluation Model, 1981 Version Dec 1981 15.6 (Proprietary)

NRC Refere Report Number Title Submittal Sectio WCAP-9221-P-A (Non-proprietary)

Revision 1 WCAP-9227 Reactor Core Response to Excessive Secondary Steam Release Jan 1978 15.1 WCAP-9485 PALADON - Westinghouse Nodal Computer Code Dec 1978 4.3 (Proprietary)

WCAP-9486 (Non-proprietary)

WCAP-9600 Report on Small Break Accidents for Westinghouse NSSS System June 1979 5.4.1 WCAP-10858P-A AMSAC Generic Design Package July 1987 7.8 Rev. 1 RP-8A (SWEC) Radiation Shielding Design and Analysis Approach for Light Water Reactor May 1975 12.2, 12.

Power Plants SWSQAP-1-74A Standard Nuclear Quality Assurance Program Dec 1974 17.1.

(SWEC)

QA-1 Revision 3A Quality Assurance Program March 1979 17.1.

(NUSCO)

1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS le 1.7-1 identifies the safety related electrical, instrumentation, and control drawings used on lstone 3.

2 PIPING AND INSTRUMENTATION DIAGRAMS le 1.7-2 identifies the piping and instrumentation diagrams (P&ID) used on Millstone 3. These rams are included throughout the FSAR in conjunction with specific system descriptions.

bols and abbreviations used in the diagrams are illustrated on Figure 1.2-3. A more complete ng of Drawing & Figure Numbers for the P&IDs is contained in the Generation Records rmation System (GRITS), and for those drawings used as Figures in the FSAR, refer to the mary Table of Contents, and Effective Pages List.

3 LOOP AND SYSTEMS DIAGRAMS s table has been deleted.

4 OTHER DETAILED INFORMATION (SPECIAL REPORTS AND PROGRAMS) le 1.7-4 identifies special reports and programs referenced in the FSAR and submitted arately from the FSAR.

DOCUMENTATION (Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO Drawing Westinghouse mber 25212-39001 Westinghouse System or Title Drawing Number 15001 Process Control Block Diagram 7244D80 Sh. 1 15002 Process Control Block Diagram 7244D80 Sh. 2 15003 Process Control Block Diagram 7244D80 Sh. 3 15004 Process Control Block Diagram 7244D80 Sh. 4 15005 Process Control Block Diagram 7244D80 Sh. 5 15006 Process Control Block Diagram 7244D80 Sh. 6 15007 Process Control Block Diagram 7244D80 Sh. 7 15008 Process Control Block Diagram 7244D80 Sh. 8 15009 Process Control Block Diagram 7244D80 Sh. 9 15010 Process Control Block Diagram 7244D80 Sh. 10 15011 Process Control Block Diagram 7244D80 Sh. 11 15012 Process Control Block Diagram 7244D80 Sh. 12 15013 Process Control Block Diagram 7244D80 Sh. 13 15014 Process Control Block Diagram 7244D80 Sh. 14 15015 Process Control Block Diagram 7244D80 Sh. 15 15016 Process Control Block Diagram 7244D80 Sh. 16 15017 Process Control Block Diagram 7244D80 Sh. 17 15018 Process Control Block Diagram 7244D80 Sh. 18 15019 Process Control Block Diagram 7244D80 Sh. 19 15020 Process Control Block Diagram 7244D80 Sh. 20 15021 Process Control Block Diagram 7244D80 Sh. 21 15022 Process Control Block Diagram 7244D80 Sh. 22 15023 Process Control Block Diagram 7244D80 Sh. 23 15024 Process Control Block Diagram 7244D80 Sh. 24 15025 Process Control Block Diagram 7244D80 Sh. 25 15026 Process Control Block Diagram 7244D80 Sh. 26

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO Drawing Westinghouse mber 25212-39001 Westinghouse System or Title Drawing Number 15027 Process Control Block Diagram 7244D80 Sh. 27 15028 Process Control Block Diagram 7244D80 Sh. 28 15029 Process Control Block Diagram 7244D80 Sh. 29 15030 Process Control Block Diagram 7244D80 Sh. 30 15031 Process Control Block Diagram 7244D80 Sh. 31 15032 Process Control Block Diagram 7244D80 Sh. 32 15033 Process Control Block Diagram 7244D80 Sh. 33 15034 Process Control Block Diagram 7244D80 Sh. 34 15035 Process Control Block Diagram 7244D80 Sh. 35 15036 Process Control Block Diagram 7244D80 Sh. 36 15037 Process Control Block Diagram 7244D80 Sh. 37 15038 Process Control Block Diagram 7244D80 Sh. 38 15039 Process Control Block Diagram 7244D80 Sh. 40 15040 Process Control Block Diagram 7244D80 Sh. 39 15041 Process Control Block Diagram 7244D80 Sh. 41 15042 Process Control Block Diagram 7244D80 Sh. 42 15043 Process Control Block Diagram 7244D80 Sh. 43 15044 Process Control Block Diagram 7244D80 Sh. 44 15045 Process Control Block Diagram 7244D80 Sh. 45 15046 Process Control Block Diagram 7244D80 Sh. 46 15047 Process Control Block Diagram 7244D80 Sh. 47 15048 Process Control Block Diagram 7244D80 Sh. 48 04021 NIS Source Range Functional Block Diagram 5655D49 04022 NIS Source Range Functional Block Diagram 5655D50 04023 NIS Source Range Functional Block Diagram 5655D51 04023 NIS Source Range Functional Block Diagram 5655D52 07011 Safeguard Test Cabinets 8758D57 Sh. 1

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO Drawing Westinghouse mber 25212-39001 Westinghouse System or Title Drawing Number 07012 Safeguard Test Cabinets 8758D57 Sh. 2 07013 Safeguard Test Cabinets 8758D57 Sh. 3 07014 Safeguard Test Cabinets 8758D57 Sh. 4 07015 Safeguard Test Cabinets 8758D57 Sh. 5 07016 Safeguard Test Cabinets 8758D57 Sh. 6 07017 Safeguard Test Cabinets 8758D57 Sh. 7 07018 Safeguard Test Cabinets 8758D57 Sh. 8 07019 Safeguard Test Cabinets 8758D57 Sh. 9 07020 Safeguard Test Cabinets 8758D57 Sh. 10 07021 Safeguard Test Cabinets 8758D57 Sh. 11 07022 Safeguard Test Cabinets 8758D57 Sh. 12 07023 Safeguard Test Cabinets 8758D57 Sh. 13 07025 Safeguard Test Cabinets 8758D57 Sh. 15 07026 Safeguard Test Cabinets 8758D57 Sh. 16 07027 Safeguard Test Cabinets 8758D57 Sh. 17 07028 Safeguard Test Cabinets 8758D57 Sh. 18 07029 Safeguard Test Cabinets 8758D57 Sh. 19 07030 Safeguard Test Cabinets 8758D57 Sh. 20 04002 Logic Diagram 108D684 Sh. 1 04003 Logic Diagram 108D684 Sh. 2 04004 Logic Diagram 108D684 Sh. 3 04005 Logic Diagram 108D684 Sh. 4 04006 Logic Diagram 108D684 Sh. 5 04007 Logic Diagram 108D684 Sh. 6 04008 Logic Diagram 108D684 Sh. 7 04009 Logic Diagram 108D684 Sh. 8 04010 Logic Diagram 108D684 Sh. 9

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO Drawing Westinghouse mber 25212-39001 Westinghouse System or Title Drawing Number 04011 Logic Diagram 108D684 Sh. 10 04012 Logic Diagram 108D684 Sh. 11 04013 Logic Diagram 108D684 Sh. 12 04014 Logic Diagram 108D684 Sh. 13 04015 Logic Diagram 108D684 Sh. 14 04016 Logic Diagram 108D684 Sh. 15 04017 Logic Diagram 108D684 Sh. 16 04018 Logic Diagram 108D684 Sh. 17 04019 Logic Diagram 108D684 Sh. 19 04020 Logic Diagram 108D684 Sh. 18 TABLE 1.7-1 LOGIC DIAGRAMS (Refer to Plant Document Control for latest Document Rev. and Date)

SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-017 1-6A Reactor Trips Control Logic Description Logic Diagram 2-018 1-6B Reactor Trips Control Logic Description Logic Diagram 2-019 1-6C Reactor Trips Control Logic Description Logic Diagram 2-020 1-6D Reactor Trips Control Logic Description Logic Diagram 2-021 1-6E Reactor Trips Control Logic Description Logic Diagram 2-022 1-6F Reactor Trips Control Logic Description Logic Diagram 2-023 1-6G Reactor Trips Control Logic Description Logic Diagram 2-024 1-6H Reactor Trips Control Logic Description Logic Diagram 2-025 1-6J Reactor Trips Control Logic Description Logic Diagram 2-026 1-6K Reactor Trips Control Logic Description Logic Diagram 2-027 1-6L Reactor Trips Control Logic Description Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-028 1-6M Reactor Trips Control Logic Description Logic Diagram 2-029 1-6N Reactor Trips Control Logic Description Logic Diagram 2-030 1-6P AMSAC Logic Diagram 2-031 1-6Q AMSAC Logic Diagram 5-001 2-1.1A Circulating Water Pump Breaker 5-002 2.1-1B Circulating Water Discharge & Water Box Valves Control Logic Diagram 5-003 2.1-1C Circulating Water Discharge & Water Box Valves Control Logic Diagram 5-004 2.1-1D Circulating Water Discharge & Water Box Valves Control Logic Diagram 5-005 2.1-1E Circulating Water Discharge & Water Box Valves Control Logic Diagram 5-006 2.1-1F Logic Diagram Circulating Water Pump 5-007 2.1-1G Logic Diagram Circulating Water Pump 5-008 2.1-1H Logic Diagram Circulating Water Pump 5-009 2.1-1J Logic Diagram Circulating Water Pump 6-001 3-1.1A Turbine Bypass Control Logic Diagram 6-002 3-1.1B Turbine Bypass Control Logic Diagram 6-003 3-1.1C Turbine Bypass Control Logic Diagram 6-004 3-1.1D Turbine Bypass Control Logic Diagram 6-005 3-1.1E Turbine Bypass Control Logic Diagram 6-006 3-1.1F Turbine Bypass Control Logic Diagram 6-007 3-1.1G Turbine Bypass Control Logic Diagram 6-008 3-1.1H Turbine Bypass Control Logic Diagram 6-009 3-1.2A Main Steam Isolation Control Logic Diagram 6-010 3-1.2B Main Steam Isolation Control Logic Diagram 6-011 3-1.2C Main Steam Isolation Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 6-012 3-1.2D Main Steam Isolation Control Logic Diagram 6-013 3-1.2E Main Steam Isolation Control Logic Diagram 1-001 3-9A Aux Stm Sys Control Logic Diagram 1-002 3-9B Aux Stm Sys Control Logic Diagram 0-001 6-1.1A Motor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram 0-002 6-1.1B Motor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram 0-003 6-1.1C Motor Drive Stm Gen Pump & Discharge Valve Control Logic Diagram 0-T019 6-1.1D Motor Drive Stm Gen Pump & Discharge Valve Logic Diagram 0-004 6-1.2A Fdwtr Sys Control Logic Diagram 0-005 6-1.2B Fdwtr Sys Control Logic Diagram 0-006 6-1.2C Fdwtr Sys Control Logic Diagram 0-007 6-1.2D Fdwtr Sys Control Logic Diagram 0-008 6-1.2E Fdwtr Sys Control Logic Diagram 6-001 6-2.1A Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-002 6-2.1B Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-003 6-2.1C Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-004 6-2.1D Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-005 6-2.1E Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-006 6-2.1F Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 6-007 6-2.1G Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-008 6-2.1H Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-009 6-2.1J Motor Drive Aux Fdwtr Pump & Recirc Control Logic Diagram 6-010 6-2.2A Turbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram 6-011 6-2.2B Turbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram 6-012 6-2.2C Turbine-Driven Aux Fdwtr Pump & Recirc Index Control Logic Diagram 5 7-3.2 Aux Fdwtr Pump & Drive Lube Oil Control Logic Diagram 3-001 8-9A Emergency Diesel Generator Fuel Control Logic Diagram 3-002 8-9B Emergency Diesel Generator Fuel Control Logic Diagram 8-001 9-1A Reactor Plant Component Cooling Water Control Logic Diagram 8-002 9-1B Reactor Plant Component Cooling Water Control Logic Diagram 8-003 9-1C Reactor Plant Component Cooling Water Control Logic Diagram 8-004 9-1D Reactor Plant Component Cooling Water Control Logic Diagram 8-005 9-1E Reactor Plant Component Cooling Water Control Logic Diagram 8-006 9-1F Reactor Plant Component Cooling Water Control Logic Diagram 8-007 9-1G Reactor Plant Component Cooling Water Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 8-008 9-1H Reactor Plant Component Cooling Water Control Logic Diagram 8-009 9-1J Reactor Plant Component Cooling Water Control Logic Diagram 8-010 9-1K Reactor Plant Component Cooling Water Control Logic Diagram 5-001 9-2A Chilled Water System Control Logic Diagram 5-002 9-2B Chilled Water System Control Logic Diagram 5-003 9-2C Chilled Water System Control Logic Diagram 5-004 9-2D Chilled Water System Control Logic Diagram 5-005 9-2E Chilled Water System Control Logic Diagram 5-006 9-2F Chilled Water System Control Logic Diagram 5-007 9-2G Chilled Water System Control Logic Diagram 5-008 9-2H Chilled Water System Control Logic Diagram 5-009 9-2J Chilled Water System Control Logic Diagram 5-010 9-2K Chilled Water System Control Logic Diagram 3-001 9-4A Charging Pumps Cooling Control Logic Diagram 3-002 9-4B Charging Pumps Cooling Control Logic Diagram 3-003 9-4C Charging Pumps Cooling Control Logic Diagram 3-004 9-4D Charging Pumps Cooling Control Logic Diagram 9-001 9-5A Safety Injection Pumps Cooling Control Logic Diagram 9-002 9-5B Safety Injection Pumps Cooling Control Logic Diagram 4-001 9-7A Turbine Plant Component Cooling Water Control Logic Diagram 4-002 9-7B Turbine Plant Component Cooling Water Control Logic Diagram 4-003 9-7C Turbine Plant Component Cooling Water Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 4-004 9-7D Turbine Plant Component Cooling Water Control Logic Diagram 4-005 9-7E Turbine Plant Component Cooling Water Control Logic Diagram 3-001 9-10A Service Water System Control Logic Diagram 3-002 9-10B Service Water System Control Logic Diagram 3-003 9-10C Service Water System Control Logic Diagram 3-004 9-10D Service Water System Control Logic Diagram 3-005 9-10E Service Water System Control Logic Diagram 3-006 9-10F Service Water System Control Logic Diagram 3-007 9-10G Service Water System Control Logic Diagram 3-008 9-10H Service Water System Control Logic Diagram 3-009 9-10J Service Water System Control Logic Diagram 3-010 9-10K Service Water System Control Logic Diagram 3-011 9-10L Service Water System Control Logic Diagram 3-012 9-10M Service Water System Control Logic Diagram 4-001 12-1A Instrument Air Control Logic Diagram 4-002 12-1B Instrument Air Control Logic Diagram 4-003 12-1C Instrument Air Control Logic Diagram 4-004 12-1D Instrument Air Control Logic Diagram 4-001 12-1A Instrument Air Control Logic Diagram 4-002 12-1B Instrument Air Control Logic Diagram 4-005 12-1E Instrument Air Control Logic Diagram 4-006 12-1F Instrument Air Control Logic Diagram 4-007 12-1G Instrument Air Control Logic Diagram 4-008 12-1H Instrument Air Control Logic Diagram 4-009 12-1J Instrument Air Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 4-003 12-3C Containment Instrument Air Control Logic Diagram 7-001 14-1A Nitrogen System Control Logic Diagram 7-002 14-1B Nitrogen System Control Logic Diagram 7-003 14-1C Nitrogen System Control Logic Diagram 7-004 14-1D Nitrogen System Control Logic Diagram 3-001 15-03A Fire Protection Low Press CO2 Logic Diagram 3-002 15-03B Fire Protection Low Press CO2 Logic Diagram 3-003 15-03C Fire Protection Low Press CO2 Logic Diagram 3-004 15-03D Fire Protection Low Press CO2 Logic Diagram 3-005 15-03E Fire Protection Low Press CO2 Logic Diagram 0-001 22-1A Reactor Plant Ventilation Control Logic Diagram 0-002 22-1B Reactor Plant Ventilation Control Logic Diagram 0-003 22-1C Reactor Plant Ventilation Control Logic Diagram 0-004 22-1D Reactor Plant Ventilation Control Logic Diagram 0-005 22-1E Reactor Plant Ventilation Control Logic Diagram 0-006 22-1F Reactor Plant Ventilation Control Logic Diagram 0-007 22-1G Reactor Plant Ventilation Control Logic Diagram 0-008 22-1H Reactor Plant Ventilation Control Logic Diagram 0-009 22-1J Reactor Plant Ventilation Control Logic Diagram 0-010 22-1K Reactor Plant Ventilation Control Logic Diagram 0-011 22-1L Reactor Plant Ventilation Control Logic Diagram 0-012 22-1M Reactor Plant Ventilation Control Logic Diagram 0-013 22-1N Reactor Plant Ventilation Control Logic Diagram 0-014 22-1P Reactor Plant Ventilation Control Logic Diagram 0-015 22-1Q Reactor Plant Ventilation Control Logic Diagram 0-016 22-1R Reactor Plant Ventilation Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 0-017 22-1S Reactor Plant Ventilation Control Logic Diagram 0-018 22-1T Reactor Plant Ventilation Control Logic Diagram 0-019 22-1U Reactor Plant Ventilation Control Logic Diagram 0-020 22-1V Reactor Plant Ventilation Control Logic Diagram 0-021 22-1W Reactor Plant Ventilation Control Logic Diagram 0-022 22-1X Reactor Plant Ventilation Control Logic Diagram 0-023 22-1Y Reactor Plant Ventilation Control Logic Diagram 0-024 22-1Z Reactor Plant Ventilation Control Logic Diagram 0-025 22-1AA Reactor Plant Ventilation Control Logic Diagram 0-026 22-1AB Reactor Plant Ventilation Control Logic Diagram 9-001 22-7A Diesel Gen Bldg Ventilation Control Logic Diagram 9-002 22-7B Diesel Gen Bldg Ventilation Control Logic Diagram 9-003 22-7C Diesel Gen Bldg Ventilation Control Logic Diagram 2-001 22-8A Yard Structure Ventilation Control Logic Diagram 2-002 22-8B Yard Structure Ventilation Control Logic Diagram 2-003 22-8C Yard Structure Ventilation Control Logic Diagram 2-004 22-8D Yard Structure Ventilation Control Logic Diagram 2-005 22-8E Yard Structure Ventilation Control Logic Diagram 2-006 22-8F Yard Structure Ventilation Control Logic Diagram 2-007 22-8G Yard Structure Ventilation Control Logic Diagram 2-008 22-8H Yard Structure Ventilation Control Logic Diagram 2-009 22-8J Yard Structure Ventilation Control Logic Diagram 2-010 22-8K Yard Structure Ventilation Control Logic Diagram 2-011 22-8L Yard Structure Ventilation Control Logic Diagram 2-012 22-8M Yard Structure Ventilation Control Logic Diagram 2-013 22-8N Yard Structure Ventilation Control Logic Diagram 2-014 22-8P Yard Structure Ventilation Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-016 22-8S Yard Structure Ventilation Control Logic Diagram 5-001 22-9A Control Bldg A/C Control Logic Diagram 5-002 22-9B Control Bldg A/C Control Logic Diagram 5-003 22-9C Control Bldg A/C Control Logic Diagram 5-004 22-9D Control Bldg A/C Control Logic Diagram 5-005 22-9E Control Bldg A/C Control Logic Diagram 5-006 22-9F Control Bldg A/C Control Logic Diagram 5-007 22-9G Control Bldg A/C Control Logic Diagram 5-008 22-9H Control Bldg A/C Control Logic Diagram 5-009 22-9J Control Bldg A/C Control Logic Diagram 5-010 22-9K Control Bldg A/C Control Logic Diagram 5-011 22-9L Control Bldg A/C Control Logic Diagram 5-012 22-9M Control Bldg A/C Control Logic Diagram 5-013 22-9N Control Bldg A/C Control Logic Diagram 5-014 22-9P Control Bldg A/C Control Logic Diagram 6-001 22-12A Chilled Water System Control Logic Diagram 6-002 22-12B Chilled Water System Control Logic Diagram 6-003 22-12C Chilled Water System Control Logic Diagram 6-004 22-12D Chilled Water System Control Logic Diagram 6-005 22-12E Chilled Water System Control Logic Diagram 6-006 22-12F Chilled Water System Control Logic Diagram 6-007 22-12G Chilled Water System Control Logic Diagram 6-008 22-12H Chilled Water System Control Logic Diagram 6-009 22-12J Chilled Water System Control Logic Diagram 6 22-12K Chilled Water System Control Logic Diagram 6 22-12L Chilled Water System Control Logic Diagram 6 22-12M Chilled Water System Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 3-001 22-16A Hot Water Heating Logic Diagram 3-002 22-16B Hot Water Heating Logic Diagram 3-003 22-16C Hot Water Heating Logic Diagram 3-004 22-16D Hot Water Heating Logic Diagram 3-005 22-16E Hot Water Heating Logic Diagram 7-001 22-26A ESF Bldg Ventilation Control Logic Diagram 7-002 22-26B ESF Bldg Ventilation Control Logic Diagram 7-003 22-26C ESF Bldg Ventilation Control Logic Diagram 7-004 22-26D ESF Bldg Ventilation Control Logic Diagram 7-005 22-26E ESF Bldg Ventilation Control Logic Diagram 7-006 22-26F ESF Bldg Ventilation Control Logic Diagram 7-007 22-26G ESF Bldg Ventilation Control Logic Diagram 7-008 22-26H ESF Bldg Ventilation Control Logic Diagram 7-009 22-26J ESF Bldg Ventilation Control Logic Diagram 2-001 22-27A Containment Structure Vent Control Logic Diagram 2-002 22-27B Containment Structure Vent Control Logic Diagram 2-003 22-27C Containment Structure Vent Control Logic Diagram 2-004 22-27D Containment Structure Vent Control Logic Diagram 2-005 22-27E Containment Structure Vent Control Logic Diagram 2-006 22-27F Containment Structure Vent Control Logic Diagram 2-007 22-27G Containment Structure Vent Control Logic Diagram 5-001 22-28A Main Stm Valve Bldg Ventilation Control Logic Diagram 5-002 22-28B Main Stm Valve Bldg Ventilation Control Logic Diagram 1-001 22-33A Hot Water Preheating Sys Control Logic Diagram 1-002 22-33B Hot Water Preheating Sys Control Logic Diagram 8-001 24-3A Reserve Station Service Breaker Controls Control Logic Description

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 8-002 24-3B Reserve Station Service Breaker Controls Control Logic Diagram 8-003 24-3C Reserve Station Service Breaker Controls Control Logic Diagram 8-004 24-3D Reserve Station Service Breaker Controls Control Logic Diagram 8-005 24-3E Reserve Station Service Breaker Controls Control Logic Diagram 8-006 24-3F Reserve Station Service Breaker Controls Control Logic Diagram 8-007 24-3G Reserve Station Service Breaker Controls Control Logic Diagram 8-008 24-3H Reserve Station Service Breaker Controls Control Logic Diagram 8-009 24-3J Reserve Station Service Breaker Controls Control Logic Diagram 8-010 24-3K Reserve Station Service Breaker Controls Control Logic Diagram 9-001 24-4A Medium Voltage Bus Tie Breaker Controls Control Logic Diagram 9-002 24-4B Medium Voltage Bus Tie Breaker Controls Control Logic Diagram 3-001 24-8A Low Voltage Switchgear Supply Breaker Controls Control Logic Diagram 3-002 24-8B Low Voltage Switchgear Supply Breaker Controls Control Logic Diagram 1-001 24-9.2A Emergency Gen Breaker Controls Control Logic Diagram 1-002 24-9.2B Emergency Gen Breaker Controls Control Logic Diagram 1-003 24-9.2C Emergency Gen Breaker Controls Control Logic Diagram 1-004 24-9.2D Emergency Gen Breaker Controls Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-001 24-9.3A Emergency Diesel Gen Control & Protection Control Logic Diagram 2-002 24-9.3B Emergency Diesel Gen Control & Protection Control Logic Diagram 2-003 24-9.3C Emergency Diesel Gen Control & Protection Control Logic Diagram 2-004 24-9.3D Emergency Diesel Gen Control & Protection Control Logic Diagram 2-005 24-9.3E Emergency Diesel Gen Control & Protection Control Logic Diagram 2-006 24-9.3F Emergency Diesel Gen Control & Protection Control Logic Diagram 2-007 24-9.3G Emergency Diesel Gen Control & Protection Control Logic Diagram 2-008 24-9.3H Emergency Diesel Gen Control & Protection Control Logic Diagram 2-009 24-9.3J Emergency Diesel Gen Control & Protection Control Logic Diagram 2-010 24-9.3K Emergency Diesel Gen Control & Protection Control Logic Diagram 2-011 24-9.3L Emergency Diesel Gen Control & Protection Control Logic Diagram 2-012 24-9.3M Emergency Diesel Gen Control & Protection Control Logic Diagram 2-013 24-9.3N Emergency Diesel Gen Control & Protection Control Logic Diagram 2-014 24-9.3O Emergency Diesel Gen Control & Protection Control Logic Diagram 2-015 24-9.3P Emergency Diesel Gen Control & Protection Control Logic Diagram 3-001 24-9.4A Emergency Gen Load Sequence Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 3-002 24-9.4B Emergency Gen Load Sequence Control Logic Diagram 3-003 24-9.4C Emergency Gen Load Sequence Control Logic Diagram 3-004 24-9.4D Emergency Gen Load Sequence Control Logic Diagram 3-005 24-9.4E Emergency Gen Load Sequence Control Logic Diagram 3-006 24-9.4F Emergency Gen Load Sequence Control Logic Diagram 3-007 24-9.4G Emergency Gen Load Sequence Control Logic Diagram 3-008 24-9.4H Emergency Gen Load Sequence Control Logic Diagram 3-009 24-9.4J Emergency Gen Load Sequence Control Logic Diagram 3-010 24-9.4K Emergency Gen Load Sequence Control Logic Diagram 3-011 24-9.4L Emergency Gen Load Sequence Control Logic Diagram 3-012 24-9.4M Emergency Gen Load Sequence Control Logic Diagram 3-013 24-9.4N Emergency Gen Load Sequence Control Logic Diagram 3-014 24-9.4P Emergency Gen Load Sequence Control Logic Diagram 3-015 24-9.4Q Emergency Gen Load Sequence Control Logic Diagram 3-016 24-9.4R Emergency Gen Load Sequence Control Logic Diagram 3-017 24-9.4S Emergency Gen Load Sequence Control Logic Diagram 3-018 24-9.4T Emergency Gen Load Sequence Control Logic Diagram 3-019 24-9.4U Emergency Gen Load Sequence Control Logic Diagram 3-020 24-9.4V Emergency Gen Load Sequence Control Logic Diagram 3-021 24-9.4W Emergency Gen Load Sequence Control Logic Diagram 3-022 24-9.4X Emergency Gen Load Sequence Control Logic Diagram 3-023 24-9.4Y Emergency Gen Load Sequence Control Logic Diagram 3-024 24-9.4Z Emergency Gen Load Sequence Control Logic Diagram 4-001 24-10A Battery Power Supply Control Logic Functional Diagram 4-002 24-10B Battery Power Supply Control Logic Diagram 4-003 24-10C Battery Power Supply Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 4-004 24-11A Instrument and Control AC Power Supply Control Logic Functional Diagram 4-005 24-11B Instrument and Control AC Power Supply Control Logic Functional Diagram 4-006 24-11C Instrument and Control AC Power Supply Control Logic Functional Diagram 4-001 24-12.5A Synchronizing Check Control Logic Diagram 4-002 24-12.5B Synchronizing Check Control Logic Diagram 4-003 24-12.5C Synchronizing Check Control Logic Diagram 4-004 24-12.5D Synchronizing Check Control Logic Diagram 4-001 25-1.1A Reactor Coolant Pumps Control Logic Diagram 4-002 25-1.1B Reactor Coolant Pumps Control Logic Diagram 4-003 25-1.1C Reactor Coolant Pumps Control Logic Diagram 4-004 25-1.1D Reactor Coolant Pumps Control Logic Diagram 4-005 25-1.2A Pressurized Control Logic Diagram 4-006 25-1.2B Pressurized Control Logic Diagram 4-007 25-1.2C Pressurized Control Logic Diagram 4-008 25-1.2D Pressurized Control Logic Diagram 4-009 25-1.2E Pressurized Control Logic Diagram 4-010 25-1.2F Pressurized Control Logic Diagram 4-011 25-1.2G Pressurized Control Logic Diagram 4-012 25-1.2H Pressurized Control Logic Diagram 4-013 25-1.J Pressurized Control Logic Diagram 4-014 25-1.K Pressurized Control Logic Diagram 4-015 25-1.L Pressurized Control Logic Diagram 4-016 25-1.4A Pressurizer Relief Tank Control Logic Diagram 4-017 25-1.4B Pressurizer Relief Tank Control Logic Diagram 4-018 25-1.5A Reactor Coolant Isolation Valves Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 4-019 25-1.5B Reactor Coolant Isolation Valves Control Logic Diagram 4-020 25-1.5C Reactor Coolant Isolation Valves Control Logic Diagram 4-001 26-2.1A Reactor Coolant Letdown Control Logic Diagram 4-002 26-2.1B Reactor Coolant Letdown Control Logic Diagram 4-003 26-2.1C Reactor Coolant Letdown Control Logic Diagram 4-004 26-2.1D Reactor Coolant Letdown Control Logic Diagram 4-005 26-2.2A Volume Control Tank Control Logic Diagram 4-006 26-2.2B Volume Control Tank Control Logic Diagram 4-007 26-2.2C Volume Control Tank Control Logic Diagram 4-008 26-2.2D Volume Control Tank Control Logic Diagram 4-010 26-2.3A Charging Pumps Control Logic Diagram 4-011 26-2.3B Charging Pumps Control Logic Diagram 4-012 26-2.3C Charging Pumps Control Logic Diagram 4-013 26-2.3D Charging Pumps Control Logic Diagram 4-014 26-2.3E Charging Pumps Control Logic Diagram 4-015 26-2-3F Charging Pumps Control Logic Diagram 4-016 26-2.3G Charging Pumps Control Logic Diagram 4-017 26-2.3H Charging Pumps Control Logic Diagram 4-018 26-2.3J Charging Pumps Control Logic Diagram 4-029 26-2.5A Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-030 26-2.5B Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-031 26-2.5C Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-032 26-2.5D Reactor Makeup & Boric Acid Blender Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 4-033 26-2.5E Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-034 26-2.5F Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-035 26-2.5G Reactor Makeup & Boric Acid Blender Control Logic Diagram 4-036 26-2.6A Reactor Coolant Pumps Seal Water Control Logic Diagram 4-037 26-2.6B Reactor Coolant Pumps Seal Water Control Logic Diagram 4-001 27-2A High Pressure Safety Injection Control Logic Diagram 4-002 27-2B High Pressure Safety Injection Control Logic Diagram 4-003 27-2C High Pressure Safety Injection Control Logic Diagram 4-004 27-2D High Pressure Safety Injection Control Logic Diagram 4-005 27-2E High Pressure Safety Injection Control Logic Diagram 4-006 27-2F High Pressure Safety Injection Control Logic Diagram 4-007 27-2G High Pressure Safety Injection Control Logic Diagram 4-008 27-2H High Pressure Safety Injection Control Logic Diagram 4-009 27-2J High Pressure Safety Injection Control Logic Diagram 4-010 27-2K High Pressure Safety Injection Control Logic Diagram 4-011 27-2L High Pressure Safety Injection Control Logic Diagram 1-001 27-3A Low Pressure Safety Injection Control Logic Diagram 1-002 27-3B Low Pressure Safety Injection Control Logic Diagram 1-003 27-3C Low Pressure Safety Injection Control Logic Diagram 1-004 27-3D Low Pressure Safety Injection Control Logic Diagram 1-005 27-3E Low Pressure Safety Injection Control Logic Diagram 1-006 27-3F Low Pressure Safety Injection Control Logic Diagram 1-007 27-3G Low Pressure Safety Injection Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 1-008 27-3H Low Pressure Safety Injection Control Logic Diagram 9-001 27-7A Residual Heat Removal Control Logic Diagram 9-002 27-7B Residual Heat Removal Control Logic Diagram 9-003 27-7C Residual Heat Removal Control Logic Diagram 9-004 27-7D Residual Heat Removal Control Logic Diagram 9-005 27-7E Residual Heat Removal Control Logic Diagram 9-006 27-7F Residual Heat Removal Control Logic Diagram 9-007 27-7G Residual Heat Removal Control Logic Diagram 9-008 27-7H Residual Heat Removal Control Logic Diagram 9-009 27-7J Residual Heat Removal Control Logic Diagram 9-010 27-7K Residual Heat Removal Control Logic Diagram 4-001 27-10A Containment Vacuum Control Logic Diagram 4-002 27-10B Containment Vacuum Control Logic Diagram 7-001 27-11A Containment Recirc Control Logic Diagram 7-002 27-11B Containment Recirc Control Logic Diagram 7-003 27-11C Containment Recirc Control Logic Diagram 7-004 27-11D Containment Recirc Control Logic Diagram 7-005 27-11E Containment Recirc Control Logic Diagram 7-006 27-11F Containment Recirc Control Logic Diagram 7-007 27-11G Containment Recirc Control Logic Diagram 7-008 27-11H Containment Recirc Control Logic Diagram 7-019 27-11J Containment Recirc Control Logic Diagram 7-010 27-11K Containment Recirc Control Logic Diagram 7-011 27-11L Containment Recirc Control Logic Diagram 3-001 27-12A Quench Spray Control Logic Diagram 3-002 27-12B Quench Spray Control Logic Diagram 3-003 27-12C Quench Spray Control Logic Diagram

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SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 3-004 27-12D Quench Spray Control Logic Diagram 3-005 27-12E Quench Spray Control Logic Diagram 3-006 27-12F Quench Spray Control Logic Diagram 0-001 27-13A DBA Hydrogen Recombiner Control Logic Diagram 0-002 27-13B DBA Hydrogen Recombiner Control Logic Diagram 0-003 27-13C DBA Hydrogen Recombiner Control Logic Diagram 0-004 27-13D DBA Hydrogen Recombiner Control Logic Diagram 0-005 27-13E DBA Hydrogen Recombiner Control Logic Diagram 9-001 27-17A Safety Injection Actuation Control Logic Diagram 9-002 27-17B Safety Injection Actuation Control Logic Diagram 9-003 27-17C Safety Injection Actuation Control Logic Diagram 6-001 27-18A Containment Spray Actuation Control Logic Diagram 6-002 27-18B Containment Spray Actuation Control Logic Diagram 6-003 27-18C Containment Spray Actuation Control Logic Diagram 0-001 27-19A Containment Isolation Control Logic Diagram 0-002 27-19B Containment Isolation Control Logic Diagram 0-003 27-19C Containment Isolation Control Logic Diagram 0-004 27-19D Containment Isolation Control Logic Diagram 0-005 27-19E Containment Isolation Control Logic Diagram 0-006 27-19F Containment Isolation Control Logic Diagram 0-007 27-19G Containment Isolation Control Logic Diagram 2-012 31-1.1 High Level Waste Drn. TK/Pump 2-002 31-1.2A Waste Eval Reblr. Pump/Waste Dist. Pump 2-003 31-1.2B Waste Eval Reblr. Pump/Waste Dist. Pump 2-004 31-1.2C Waste Eval Reblr. Pump/Waste Dist. Pump

(Refer to Plant Document Control for latest Document Rev. and Date)

SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-005 31-1.2D Waste Eval Reblr. Pump/Waste Dist. Pump 2-006 31-1.2E Waste Eval Reblr. Pump/Waste Dist. Pump 2-018 31-1.3A Waste Test TK/Pump 2-019 31-1.3B Waste Test TK/Pump 2-020 31-1.3C Waste Test TK/Pump 2-021 31-1.4A Waste Evap Bot Pump/Waste Bot. Coolant Pump 2-022 31-1.4B Waste Evap Bot Pump/Waste Bot. Coolant Pump 2-023 31-1.4C Waste Evap Bot Pump/Waste Bot. Coolant Pump 2-024 31-1.4D Waste Evap Bot Pump/Waste Bot. Coolant Pump 2-025 31-1.4E Waste Evap Bot Pump/Waste Bot. Coolant Pump 2-026 31-1.5A Low Level Wast. Drn. TK/Pump 2-027 31-1.5B Low Level Wast. Drn. TK/Pump 2-001 31-2.1A Degasifier & Recirc Pumps Control Logic Diagram 2-002 31-2.1B Degasifier & Recirc Pumps Control Logic Diagram 2-003 31-2.1C Degasifier & Recirc Pumps Control Logic Diagram 2-004 31-2.1D Degasifier & Recirc Pumps Control Logic Diagram 2-005 31-2.1E Degasifier & Recirc Pumps Control Logic Diagram 2-006 31-2.1F Degasifier & Recirc Pumps Control Logic Diagram 2-007 31-2.1G Degasifier & Recirc Pumps Control Logic Diagram 2-008 31-2.1H Degasifier & Recirc Pumps Control Logic Diagram 6-001 32-3A Reactor Plant Gaseous Drains Control Logic Diagram 6-002 32-3B Reactor Plant Gaseous Drains Control Logic Diagram 6-003 32-3C Reactor Plant Gaseous Drains Control Logic Diagram 6-004 32-3D Reactor Plant Gaseous Drains Control Logic Diagram 2-001 32-4A Reactor Plant Aerated Drains Control Logic Diagram 2-002 32-4B Reactor Plant Aerated Drains Control Logic Diagram

(Refer to Plant Document Control for latest Document Rev. and Date)

SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 2-003 32-4C Reactor Plant Aerated Drains Control Logic Diagram 2-004 32-4D Reactor Plant Aerated Drains Control Logic Diagram 2-005 32-4E Reactor Plant Aerated Drains Control Logic Diagram 0-001 32-5.1A Turbine Plant Misc Drains Control Logic Diagram 0-002 32-5.1B Turbine Plant Misc Drains Control Logic Diagram 0-002A 32-5.1B1 Turbine Plant Misc Drains Control Logic Diagram 0-003 32-5.1C Turbine Plant Misc Drains Control Logic Diagram 0-004 32-5.1D Turbine Plant Misc Drains Control Logic Diagram 6-001 32-13A Steam Generator Blowdown System Control Logic Diagram 6-002 32-13B Steam Generator Blowdown System Control Logic Diagram 6-003 32-13C Steam Generator Blowdown System Control Logic Diagram 4 33-1 Containment Leakage Monitoring Control Logic Diagram 7 33-2 Containment Atmosphere Monitoring Control Logic Diagram 0-001 34-1A Fuel Pool Cooling & Purification System Control Logic Diagram 0-002 34-1B Fuel Pool Cooling & Purification System Control Logic Diagram 0-003 34-1C Fuel Pool Cooling & Purification System Control Logic Diagram 0-004 34-1D Fuel Pool Cooling & Purification System Control Logic Diagram 0-006 34-1E Fuel Pool Cooling & Purification System Control Logic Diagram 7-001 35-1A Primary Grade Water Control Logic Diagram 7-002 35-1B Primary Grade Water Control Logic Diagram 7-003 35-1C Primary Grade Water Control Logic Diagram

(Refer to Plant Document Control for latest Document Rev. and Date)

SWEC Drawing USCO Drawing Number 12179-mber 25212-28 LSK- Diagram Title 7-004 35-1D Primary Grade Water Control Logic Diagram TABLE 1.7-1 ONE-LINE DIAGRAMS - CLASS 1E (Refer to Plant Document Control for latest Document Rev. and Date)

SCO Dwg. SWEC Drawing Number Number 25212-30 12179-EE- Diagram Title 1 1A Main One-Line Diagram Power Distribution Composite 7 1AD 480 V MCC One-Line Diagram (Bus 32-1R, 1W) Aux Bldg (Sh. 3) 3 1AE 480 V MCC One-Line Diagram (Bus 32-5G, 5H, 5T, 5U)

Circulating Water Pump-house 0 1AH 480 V MCC One-Line Diagram (Bus 32-2F, 4T) ESF Bldg (Sh.

1) 1 1AJ 480 V MCC One-Line Diagram (Bus 32-2J, 3U, 4U) ESF Bldg (Sh. 2) 9 1AK 480 V MCC One-Line Diagram Bus 32-3D, 1L, 1T, 1U) Diesel Enclosure & Auxiliary Bldg 7 1AQ 480 V MCC One-Line Diagram (Bus 32-2R, 2W) Rod Control Area (Sh. 2) 1 1AS 480 V MCC One-Line Diagram (Bus 32-2T, 2U) Control Bldg 5 1AT 480 V MCC One-Line Diagram (Bus 32-3A, 3P, 3T) Turbine Building (Sheet 4) 6 1BA 125 V dc and 120 V ac Distribution Sys Composite 7 1BB 125 V dc One-Line Diagram Batteries 301A-1 and 301A-2 8 1BC 125 V dc One-Line Diagram Batteries 301B-1 and 301B-2 9 1BD 125 V dc One-Line Diagram Batteries 301C-1 0 1BE 125 V dc One-Line Diagram Battery 301D-A 1 1BF 120 V ac One-Line Diagram Vital Bus I and III

(Refer to Plant Document Control for latest Document Rev. and Date)

SCO Dwg. SWEC Drawing Number Number 25212-30 12179-EE- Diagram Title 2 1BG 120 V ac One-Line Diagram Vital Bus II and IV 3 1BH 120 V ac One-Line Diagram Vital Bus Back-Up 1 and 2 7 1BM 125 V Misc dc One-Line Diagram Battery 1 8 1BN 125 V Misc dc One-Line Diagram Battery 2 (Sh. 1) 6 1BR 125 V Misc dc Line Diagram Batteries 1 & 2 7 1BS 125 V dc One-Line Diagram Batteries 1 & 2 8 (Sh. 1) 1BT 125 V Misc dc One-Line Diagram 2 (Sh. 3) 1CC 120 V Misc ac One-Line Diagram 6 (Sh. 7) 1CG 120 V Misc ac One-Line Diagram 7 (Sh. 8) 1CH 120 V Misc ac One-Line Diagram 8 (Sh. 9) 1CJ 120 V Misc ac One-Line Diagram 9 (Sh. 10) 1CK 120 V Misc ac One-Line Diagram 5 (Sh. 12) 1CM 120 V Misc ac One-Line Diagram 4 1D Main One-Line Diagram 4,160 & 480 Normal 8 1EF 480 V One-Line Diagram (Bus 32Y, X) [3EJS*US-4A, 4B]

030 IEH 480 V One-Line Diagram Emergency Diesel Generator 480 V Distribution Panels 8 1K 4.16 kV One-Line Diagram Bus 34C [3ENS*SWG-A(-0)]

(Sh. 1) 9 1L 4.16 kV One-Line Diagram Bus 34C [3ENS*SWG-A(0)]

(Sh. 2) 0 1M 4.16 kV One-Line Diagram Bus 34D [3ENS*SWG-B(P)]

(Sh. 1) 1 1N 4.16 kV One-Line Diagram Bus 34D [3ENS*SWG-B(P)]

(Sh. 2) 5 1U 480 V One-Line Diagram (Bus 32T, U) [3EJS*US-1A, 1B]

2 1V 480 V One-Line Diagram (Bus 32S, V) [3EJS*US-2A, 2B]

3 1W 480 V One-Line Diagram (Bus 32R, W) [3EJS*US-3A, 3B]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 2A 2A Device Numbers and General Notes

. 2B 2B Code Information Guide

. 2C 2C Elementary Diagram Symbols

. 2D 2D Elementary Diagram Symbols

. 2E 2E Elementary Diagram Symbols

. 2F 2F Elementary Diagram Symbols

. 2G 2G One-Line Diagram Symbols

. 2H 2H One-Line Diagram Symbols

. 2J 2J One-Line Diagram Symbols

. 2K 2K One-Line Diagram Symbols

. 2L 2L Legend and Location Symbols

. 2M 2M Legend and Location Symbols

. 2N 2N Legend and Location Symbols

. 2P 2P Legend and Location Symbols

. 2Q 2Q Typ isolation CKT & presentation on ESK 001 3A Control Switch Contact Diagram 001 3B Control Switch Contact Diagram 001 3C Control Switch Contact Diagram 001 3D Control Switch Contact Diagram 001 3E Control Switch Contact Diagram 001 3F Control Switch Contact Diagram 001 3G Control Switch Contact Diagram 001 3H Control Switch Contact Diagram 001 3J Control Switch Contact Diagram 001 3K Control Switch Contact Diagram 001 3L Control Switch Contact Diagram

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title 001 3M Control Switch Contact Diagram 001 3N Control Switch Contact Diagram 001 3P Control Switch Contact Diagram 001 3Q Control Switch Contact Diagram 001 3R Control Switch Contact Diagram 001 3S Control Switch Contact Diagram 001 3T Control Switch Contact Diagram 001 3U Control Switch Contact Diagram 001 3V Control Switch Contact Diagram 001 3W Control Switch Contact Diagram 001 3X Control Switch (OIM) Contact Diagram 001 3Y Control Switch Contact Diagram 001 3Z Control Switch Contact Diagram 001 3AA Control Switch Contact Diagram 001 3AB Control Switch Contact Diagram 001 3AC1 Control Switch Contact Diagram 001 3AC2 Control Switch Contact Diagram 001 3AD Control Switch Contact Diagram 001 3AE Control Switch Contact Diagram 001 4AAA Outline Isolator Cabinets 001 4AAH01 Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP 001 4AAH02 Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP 001 4AAH03 Fire Transfer Switch Panel (FTSP) 3CES*PNLFTSP 001 4A01 Main Control Board Outline 001 4BA03 Auxiliary Shutdown Panel 3RPS*PNLAS 001 4BA04 Auxiliary Shutdown Panel 3RPS*PNLAS

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title 001 4BA05 Auxiliary Shutdown Panel 3RPS*PNLAS

. 4BA6 4BA6 Auxiliary Shutdown Panel (ASP) 3RPS*PNLAS (Sh.6)

. 4BA7 4BA7 Auxiliary Shutdown Panel (ASP) 3RPS*PNLAS (Sh.7)

. 4BA8 4BA8 Auxiliary Shutdown Panel (ASP) 3RPS*PNLAS (Sh.8)

. 4BA9 4BA9 Auxiliary Shutdown Panel (ASP) 3RPS*PNLAS (Sh.9) 001 4B02 MB1 Front Section 001 4B03 MB1 Front Section 001 4B04 MB1 Front Section 001 4B05 MB1 Front Section 001 4B06 Main Control Board Front Section MB1 001 4B07 Main Control Board Front Section MB1 001 4B08 Main Control Board Front Section MB1 001 4B09 MB1 Front Section 001 4B10 Main Control Board Front Section MB1 001 4B11 MB1 Front Section 001 4C02 Main Control Board Front Section MB2 001 4C03 Main Control Board Front Section MB2 001 4C04 Main Control Board Front Section MB2 001 4C05 Main Control Board Front Section MB2 001 4C06 Main Control Board Front Section MB2 001 4C07 Main Control Board Front Section MB2 001 4C08 Main Control Board Front Section MB2 001 4C09 Main Control Board Front Section MB2 001 4C10 Main Control Board Front Section MB2 001 4C11 Main Control Board Front Section MB2 001 4D02 Main Control Board Front Section MB3

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title 001 4D03 Main Control Board Front Section MB3 001 4D04 Main Control Board Front Section MB3 001 4D05 Main Control Board Front Section MB3 001 4D06 Main Control Board Front Section MB3 001 4D07 Main Control Board Front Section MB3 001 4D08 Main Control Board Front Section MB3 001 4D09 Main Control Board Front Section MB3 001 4D11 Main Control Board Front Section MB3 001 4E02 Main Control Board Front Section MB4 001 4E03 Main Control Board Front Section MB4 001 4E04 Main Control Board Front Section MB4 001 4E05 Main Control Board Front Section MB4 001 4E06 Main Control Board Front Section MB4 001 4E07 Main Control Board Front Section MB4 001 4E08 Main Control Board Front Section MB4 001 4F02 Main Control Board Front Section MB5 001 4F03 Main Control Board Front Section MB5 001 4F04 Main Control Board Front Section MB5 001 4F05 Main Control Board Front Section MB5 001 4F06 Main Control Board Front Section MB5 001 4F07 Main Control Board Front Section MB5 001 4F08 Main Control Board Front Section MB5 001 4F09 Main Control Board Front Section MB5 001 4F10 Main Control Board Front Section MB5 001 4F11 Main Control Board Front Section MB5 001 4F13 Main Control Board Front Section MB5

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title 001 4G02 Main Control Board Front Section MB6 001 4G03 Main Control Board Front Section MB6 001 4G04 Main Control Board Front Section MB6 001 4G05 Main Control Board Front Section MB6 001 4G06 Main Control Board Front Section MB6 001 4G07 Main Control Board Front Section MB6 001 4G07A Main Control Board Front Section MB6 001 4H02 Main Control Board Front Section MB7 001 4H03 Main Control Board Front Section MB7 001 4H04 Main Control Board Front Section MB7 001 4H05 Main Control Board Front Section MB7 001 4H06 Main Control Board Front Section MB7 001 4H07 Main Control Board Front Section MB7 001 4H08 Main Control Board Front Section MB7 001 4H09 Main Control Board Front Section MB7 001 4H10 Main Control Board Front Section MB7 001 4H12 Main Control Board Front Section MB7 001 4J02 Main Control Board Front Section MB8 001 4J03 Main Control Board Front Section MB8 001 4J04 Main Control Board Front Section MB8 001 4J05 Main Control Board Front Section MB8 001 4J06 Main Control Board Front Section MB8 001 4J07 Main Control Board Front Section MB8 001 4J08 Main Control Board Front Section MB8 001 4J09 Main Control Board Front Section MB8 001 4J13 Main Control Board Front Section MB8

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title 001 4K02 Main Control Board Front Section MBIR 001 4K03 Main Control Board Front Section MB1R 001 4K04 Main Control Board Front Section MB1R

. 4LA3 4LA3 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA4 4LA4 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA5 4LA5 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA6 4LA6 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA7 4LA7 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA8 4LA8 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA9 4LA9 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA10 4LA10 Main Vent and Air Cond Panel 3HVC*PNLVP1

. 4LA11 4LA11 Main Vent and Air Cond Panel 3HVS*PNLVP1

. 4LA12 4LA12 Main Vent and Air Cond Panel 3HVS*PNLVP1

. 4LA13 4LA13 Main Vent and Air Cond Panel 3HVS*PNLVP1

. 4LA14 4LA14 Main Vent & Air Cond Panel 3HVS*PNLVP1

. 4LA15 4LA15 Main Vent and Air Cond Panel

. 4L2 4L2 Main Control Board Rear Section MB2R

. 4L3 4L3 Main Control Board Rear Section MB2R

. 4L4 4L4 Main Control Board Rear Section MB2R

. 4M2 4M2 Main Control Board Rear Section MB3R

. 4M3 4M3 Main Control Board Rear Section MB3R

. 4N2 4N2 Main Control Board Rear Section MB4R

. 4N3 4N3 Main Control Board Rear Section MB4R

. 4N4 4N4 Main Control Board Rear Section MB4R

. 4P2 4P2 Main Control Board Rear Section MB5R

. 4P3 4P3 Main Control Board Rear Section MB5R

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 4P4 4P4 Main Control Board Rear Section MB5R

. 4P5 4P5 Main Control Board Rear Section MB5R

. 4P6 4P6 Main Control Board Rear Section MB5R

. 4P7 4P7 Main Control Board Rear Section MB5R

. 4P8 4P8 Main Control Board Rear Section MB5R

. 4Q2 4Q2 Main Control Board Rear Section MB6R

. 4Q3 4Q3 Main Control Board Rear Section MB6R

. 4Q4 4Q4 Main Control Board Rear Section MB6R

. 4Q5 4Q5 Main Control Board Rear Section MB6R

. 4Q6 4Q6 Main Control Board Rear Section MB6R

. 4R2 4R2 Main Control Board Rear Section MB7R

. 4R3 4R3 Main Control Board Rear Section MB7R

. 4R4 4R4 Main Control Board Rear Section MB7R

. 4R5 4R5 Main Control Board Rear Section MB7R

. 4R6 4R6 Main Control Board Rear Section MB7R

. 4S2 4S2 Main Control Board Rear Section MB8R

. 4S3 4S3 Main Control Board Rear Section MB8R

. 4S4 4S4 Main Control Board Rear Section MB8R

. 4S5 4S5 Main Control Board Rear Section MB8R

. 4S6 4S6 Main Control Board Rear Section MB8R

. 4S7 4S7 Main Control Board Rear Section MB8R

. 4U1 4U1 Post Accident Sampling Panel 3SSP*PNL3

. 4U2 4U2 Post Accident Sampling Panel 3SSP*PNL3

. 4U3 4U3 Post Accident Sampling Panel 3SSP*PNL3

. 4U4 4U4 Post Accident Sampling Panel 3SSP*PNL3

. 4U5 4U5 Post Accident Sampling Panel 3SSP*PNL3

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 4YA3 4YA3 Auxiliary Relay Rack (AR4) 3RPS*Rakotxa

. 4YA4 4YA4 Auxiliary Relay Rack (AR4) 3RPS*Rakotxa

. 4YA5 4YA5 Auxiliary Relay Rack (AR4) 3RPS*Rakotxa

. 4ZA3 4ZA3 Auxiliary Relay Rack (AR5)3RPS*Rakotxb

. 4ZA4 4ZA4 Auxiliary Relay Rack (AR5) 3RPS*Rakotxb

. 4ZA5 4ZA5 Auxiliary Relay Rack (AR5) 3RPS*Rakotxb

. 4ZB 4ZB Aux Relay & Cont Panel 3HVC*PNL CHLIA Cont Bldg Chld Wtr Sys-Train "A" Aux

. 4ZC 4ZC Aux Relay & Cont Panel 3HVC*PNL CHIB Cont Bldg Chld Wtr Sys-Train "B"

. 4ZD2 4ZD2 Transfer Switch Panel Train A (TSPA) 3 ES*PNL TSA

. 4ZD3 4ZD3 Transfer Switch Panel Train A (TSPA) 3 ES*PNL TSA

. 4ZD4 4ZD4 Transfer Switch Panel Train A (TSPA) 3CES*PNL TSA

. 4ZE2 4ZE2 Transfer Switch Panel Train A (TSPA) 3CES*PNL TSA

. 4ZE3 4ZE3 Transfer Switch Panel Train A (TSPA) 3CES*PNL TSA

. 5A 5A Typ Med Voltage Swgr ACB

. 5BB 5BB 4.16 KV Norm Stat Service BKr

. 5BC 5BC Norm Sta Svce Bkr [3NNS-ACB-BN] 35A3-34B-2

. 5BD 5BD Rsv Sta Svce Bkr [3ENS*ACB-AR] 23SA3-34C-2

. 5BE 5BE Rsv Sta Svce Bkr [3ENS*ACB-BR] 23SA-3-34D-2

. 5BF 5BF Bus Tie Bkr [3ENS*ACB-TA] 34C-1T-2

. 5BG 5BG Bus Tie Bkr [3ENS*ACB-TB] 34D-1T-2

. 5BT 5BT US Fdr Bkr [3ENS*ACB-AA], 34C3-2

. 5BU 5BU US Fdr Bkr [3ENS*ACB-AB], 34C4-2

. 5BV 5BV US Fdr Bkr [3ENS*ACB-BA], 34D2-2

. 5BW 5BW US Fdr Bkr [3ENS*ACB-BB], 34D3-2

. 5BX 5BX US Fdr Bkr [3ENS*ACB-AC], 34C5-2

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 5BY 5BY US Fdr Bkr [3ENS*ACB-BC], 34D4-2

. 5CJ 5CJ Service Wtr Pump [3SWP*P1A]

. 5CK 5CK Service Wtr Pump [3SWP*P1B]

. 5CL 5CL Service Wtr Pump [3SWP*P1C]

. 5CM 5CM Service Wtr Pump

. 5CN 5CN Cntmt Recirc Pump [3RSS*P1A]

. 5CP 5CP Cntmt Recirc Pump [3RSS*P1B]

. 5CQ 5CQ Cntmt Recirc Pump [3RSS*P1C]

. 5CR 5CR Cntmt Recirc Pump [3RSS*P1D]

. 5CS 5CS Charging Pump P3A [3CHS*P3A]

. 5CT 5CT Charging Pump P3B [3CHS*P3B]

. 5CU 5CU Charging Pump P3C (Spare) [3CHS*P3C]

. 5CV 5CV Charging Pump P3C (Spare) [3CHS*P3C]

. 5DA 5DA Reactor Plant Comp Cooling Wtr Pp [3CCP*P1A]

. 5DB 5DB Reactor Plant Comp Cooling Wtr Pp [3CCP*P1B]

. 5DC 5DC Reactor Plant Comp Cooling Wtr Swing Pp [3CCP*P1C]

. 5DD 5DD Reactor Plant Comp Cooling Wtr Swing Pp [3CCP*P1C]

. 5DE 5DE Residual Heat Removal Pump P1A [3RHS*P1A]

. 5DF 5DF Residual Heat Removal Pump P1B [3RHS*P1B]

. 5DG 5DG Quench Spray Pump P3A (3QSS*P3A)

. 5DH 5DH Quench Spray Pump P3B (3QSS*P3B)

. 5DJ 5DJ Safety Injection Pump P1A [3SIH*P1A]

. 5DK 5DK Safety Injection Pump P1B [3SIH*P1B]

. 5DR 5DR Emer Diesel Gen Bkr [3ENS*ACB-G-A] 15G-14U-2

. 5DS 5DS Emer Diesel Gen Bkr [3ENS*ACB-G-B] 15G-15U-2

. 5DX 5DX Stm Gen Aux Fdwtr Pp Mot Driven P1A [3FWA*P1A]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 5DY 5DY Stm Gen Aux Fdwtr Pp Mot Driven P1B [3FWA*P1B]

. 5DZ 5DZ Cntrl Bldg Chilled Wtr Chiller CHL1A [3HVK*CHL1A]

. 5EA 5EA Cntrl Bldg Chilled Wtr Chiller CHL1B [3HVK*CHL1B]

. 5EB 5EB Emer Gen Neutral Bkr [3ENS-ACB-GNA] 15G-14U-2N

. 5EC 5EC Emer Gen Neutral Bkr [3ENS-ACB-GNB] 15G-15U-2N

. 5EX 5EX Elem Diag Misc Ckts Sta 3VCE Bkr Control

. 5EY 5EY 4.16 kV - US Fdr Bkr [3ENS*ACB-AD]

. 5EZ 5EZ 4.16 kV - US Fdr Bkr [3ENS*ACB-BD]

. 5FD 5FD AH D.G. Bkr Backup Protection

. 6A 6A Typical 480 V ACB

. 6M 6M Emer Supply Bkr [3EJS*ACB-AA]

. 6N 6N Emer Supply Bkr [3EJS*ACB-AB]

. 6P 6P Emer Supply Bkr [3EJS*AC]

. 6Q 6Q Emer Supply Bkr [3EJS*ACB-BA]

. 6R 6R Emer Supply Bkr [3EJS*ACB-BB]

. 6S 6S Emer Supply Bkr [3EJS*ACB-BC]

. 6AG 6AG Fuel Pool Cooling Pump [3SFC*P1A]

. 6AH 6AH Fuel Pool Cooling Pump [3SFC*P1B]

. 6AN 6AN Pressurizer Htrs BU GP A [3RCS*PH1A]

. 6AP 6AP Pressurizer H Htrs BU GP B [3RCS*PH1A]

. 6AV 6AV Crdm Shroud [3HVU-FN2A]

. 6AW 6AW Crdm Shroud Fan [3HVU-FN2B]

. 6AY 6AY Inst Air Comp Fdr Bkr

. 6BA 6BA Cntmt Air Recirc Fan [3HVU-FN1A]

. 6BB 6BB Cntmt Air Recirc Fan [3HVU-FN1B]

. 6DD 6DD Serv Wtr Pp Str and Backwash V [3SWP*STR1A],

[3SWP*MOV24A]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6DE 6DE Serv Wtr Pp Str and Backwash V [3SWP*STR1B],

[3SWP*MOV24B]

. 6DF 6DF Serv Wtr Pp Str and Backwash V [3SWP*STR1C],

[3SWP*MOV24C]

. 6DG 6DG Serv Wtr Pp Str and Backwash V [3SWP*STR1D],

[3SWP*MOV24D]

. 6DX 6DX Emer Gen Fuel Oil Xfer pps [3EGF*P1A,C]

. 6DY 6DY Emer Gen Fuel Oil Xfer Pps [3EGF*P1B,D]

. 6GD 6GD Boric Acid Transfer Pump [3CHS*P2A]

. 6GE 6GE Boric Acid Transfer Pump [3CHS*P2B]

. 6GH 6GH Charging Pumps Cooling Pumps [3CCE*P1A,B]

. 6GT 6GT Cont Bldg Chilled Wtr Pump [3HVK*P1A]

. 6GU 6GU Cont Bldg Chilled Wtr Pump [3HVK*P1B]

. 6GV 6GV Cont Bldg Chilled Lube Oil Pumps [3HVK*P3A] &

[3HVK*P3B]

. 6LD 6LD Cntmt Recir Wtr Spray Hdr Isol VV20A

. 6LE 6LE Cntmt Recir Wtr Spray Hdr Isol VV20B

. 6LF 6LF Cntmt Recir Wtr Spray Hdr Isol VV20C

. 6LG 6LG Cntmt Recir Wtr Spray Hdr Isol VV20D

. 6LH 6LH Cont Recir Pump Suct Isol VV23A

. 6LJ 6LJ Cont Recir Pump Suct Isol VV23B

. 6LK 6LK Cont Recir Pmp Suct Isol VV23C

. 6LL 6LL Cont Recir Pump Suct Isol VV23D

. 6LM 6LM Rss to Rhr Cross Connect MV8837A

. 6LN 6LN Rss to Rhr Cross Connect MV8838A

. 6LP 6LP Rss to Rhr Cross Connect MV8837B

. 6LQ 6LQ Rss to Rhr Cross Connect MV8838B

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6LS 6LS Quench Spray Header Isol Valve (3QSS*MOV34A)

. 6LT 6LT Quench Spray Header Isol Valve (3QSS*MOV34B)

. 6MF 6MF Refueling Wtr Stor Tk to SIP [3SIH*MV8806]

. 6MG 6MG SIP Suction [3SIH*MV8923A]

. 6MH 6MH SIP Suction [3SIH*MV8923B]

. 6MJ 6MJ SIP Disch HDR Isol Valve [3SIH*MV8821A]

. 6MK 6MK SIP Disch Hdr Isol Valve [3SIH*MV8821B]

. 6ML 6ML SIP to Cold Legs [3SIH*MV8835]

. 6MN 6MN Hi Press SIP Mini Flow Isol [3SIH*MV8813]

. 6MP 6MP Suction Hdr Cross Connection [3SIH*MV8807A]

. 6MQ 6MQ Suction Header Cross Connection [3SIH*MV8807B]

. 6MR 6MR SIP to Hot Legs [3SIH*MV8802A]

. 6MS 6MS SIP to Hot Legs [3SIH*MV8802B]

. 6MV 6MV Boron Inj Tk Disch Isol [3SIH*MV8801A]

. 6MW 6MW Boron Inj Tk Disch Isol (3SIH*MV8801B)

. 6MX 6MX Rwst to Rhr P1A Isolation 3SIL*MV8812A

. 6MY 6MY Rwst to Rhr P1B Isolation [3SIL*MV8812B]

. 6MZ 6MZ Rhr to Cold Leg Isol [3SIL*MV8809A]

. 6NA 6NA Rhr To Cold Leg Isol [3SIL*MV8809B]

. 6NB 6NB Accumulator Isolation [3SIL* MV8808A]

. 6NC 6NC Accumulator Isolation [3SIL*MV8808B]

. 6ND 6ND Accumulator Isolation [3SIL*MV8808C]

. 6NE 6NE Accumulator Isolation [3SIL*MV8808D]

. 6NF 6NF Rhr P1A to Charging Pump [3SIL*MV8804A]

. 6NG 6NG Rhr P1B to Charging Pump [3SIL*MOV8804B]

. 6NH 6NH Residual Ht Removal to Hot Legs [3SIL*MV8840]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6NJ 6NJ Cross Connection Isolation VV22A (3RHS*MV8716A)

. 6NK 6NK Cross Connection Isolation Valve [3RHS*MV8716B]

. 6NU 6NU HI Press SIP Mini Flow Isol [3SIH*MV8814]

. 6NV 6NV Hi Press SIP Mini Flow Isol [3SIH*MV8920]

. 6NW 6NW SI Charging Suction Cross Connection [3SIH*MV8924]

. 6PA 6PA Rcp Seal Wtr Isol [3CHS*MV8109A]

. 6PB 6PB Rcp Seal Wtr Isol [3CHS*MV8109B]

. 6PC 6PC Rcp Seal Wtr Isol [3CHS*MV8109C]

. 6PD 6PD Rcp Seal Wtr Isol [3CHS*MV8109D]

. 6PE 6PE Charging Pump Mini-flow Isol Valve [3CHS*V8110]

. 6PF 6PF Charging Pump Mini-flow Isol Valve [3CHS*MV8111]

. 6PG 6PG Charging Pump to Rx Clt Sys Isol Valve [3CHS*MV8105]

. 6PH 6PH Charging Pump to Rx Clt Sys Isol Valve [3CHS*MV8106]

. 6PJ 6PJ Boric Acid Fltr to Charging Pmp Valve [3CHS*MV8104]

. 6PK 6PK Vol Cont Tk Outlet Isol VV112B

. 6PL 6PL Vol Cont Tk Outlet Isol VV112C

. 6PM 6PM Refueling Wtr Stor Tk to Charging Pmp VV112D

. 6PN 6PN Refueling Wtr Stor Tk to Charging Pmp V112E

. 6PP 6PP Rx Clt Pmp Seal Wtr Isol [3CHS*MV8112]

. 6PQ 6PQ Rx Clt Pmp Seal Wtr Isol [3CHS*MV8100]

. 6QR 6QR RHS Inlet Isolation Valve [RHS*MV8701C]

. 6QS 6QS RHS Inlet Isolation Valve [3RHSMV8702C]

. 6QT 6QT RHS Inlet Isolation Valve 3RHSMV8701A

. 6QU 6QU RHS Inlet Isolation Valve 3RHSMV8702B

. 6QV 6QV RHS Inlet Isolation Valve 3RHSMV8701B

. 6QW 6QW RHR Inlet Isolation Valve 3RHS*MC8702A

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6QX 6QX RHR Pmp P1A Mini-flow Recirc VV610

. 6QY 6QY RHR Pmp P1B Mini-flow Recirc VV611

. 6RJ 6RJ Cntrl Rm Area A/C Units ACU1A, 1B

. 6RL 6RL Battery Rms 1, 2&5 ACU3A, 3B

. 6RM 6RM Cable Spread & Swgr, Equip Rm ACU4A & 4B

. 6RP 6RP Cntmt Bldg Chill Wtr Pp Aux Ckt A

. 6RQ 6RQ Cntmt Bldg Chill Wtr Pp Aux Ckt B

. 6RR 6RR Cntmt Bldg Air Cond Booster Pp (3SWP*P2A, B)

. 6RV 6RV Chiller Equip Space Supply Fan - FN2A, FN2B

. 6RW 6RW Chiller Equip Space Exhaust Fan - FN7A, 7B

. 6RX 6RX Cntmt Bldg Emerg Vent Fan - FN1A & Inlet Damper MOD33A

. 6RY 6RY Cntmt Bldg Emerg Vent Fan - FN1B & Inlet Damper MOD33B

. 6SB 6SB Aux Bldg Exh Fan & Assoc Dmprs

[3HVK*FNGA, AOD20A & 26A]

. 6SC 6SC Aux Bldg Fltr Exh Exh Fan & Assoc Dmprs

. 6SD 6SD Chg Pp Cub Sply Fan & Assoc Dmprs

. 6SE 6SE Chg Pp Cub Sply Fan & Assoc Dmprs

. 6SF 6SF Chg Pp Exh Fan & Assoc Dmprs

. 6SG 6SG Chg Pp Exh Fan & Assoc Dmprs

. 6SH 6SH Fuel Bldg Exh Fan & Assoc Dmprs

. 6SJ 6SJ Fuel Bldg Exh Fan & Assoc Dmprs

. 6SK 6SK Slcr Exh Fan & Assoc Dmprs

. 6SL 6SL Slcr Exh Fan & Assoc Dmprs

. 6SM 6SM MCC, Rod Cont & Cable Vault Area ACU

. 6TD 6TD Pressurizer Relief Isol Valve [3RCS*MV8000A]

. 6TE 6TE Pressurizer Relief Isol Valve [3RCS*MV8000B]

. 6TF 6TF Reac Clnt LP1 Hot Leg Stop Valve (3RCS*MV8001A)

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6TG 6TG Reac Clnt LP2 Hot Leg Stop Valve (3RCS*MV8001B)

. 6TH 6TH Reac Clnt LP3 Hot Leg Stop Valve (3RCS*MV8001C)

. 6TJ 6TJ Reac Clnt LP4 Hot Leg Stop Valve (3RCS*MV8001D)

. 6TK 6TK Reac Clnt LP1 Cold Leg Stop Valve (3RCS*MV8002A)

. 6TL 6TL Reac Clnt LP2 Cold Leg Stop Valve (3RCS*MV8002B)

. 6TM 6TM Reac Clnt LP3 Cold Leg Stop Valve (3RCS*MV8002C)

. 6TN 6TN Reac Clnt LP4 Cold Leg Stop Valve (3RCS*MV8002D)

. 6TP 6TP Reactor Coolant Loop 1 Bypass Leg Stop Valve (3RCS*MV8003A)

. 6TQ 6TQ Reactor Coolant Loop 2 Bypass Leg Stop Valve (3RCS*MV8003B)

. 6TR 6TR Reactor Coolant Loop 3 Bypass Leg Stop Valve (3RCS*MV8003C)

. 6TS 6TS Reactor Coolant Loop 4 Bypass Leg Stop Valve (3RCS*MV8003D)

. 6TW 6TW Swgr Area & Battery Rooms Supply Fans

. 6TX 6TX Battery Rooms 1 & 2 Exhaust Fans

. 6TY 6TY Battery Rooms 3 & 4 Exhaust Fans

. 6TZ 6TZ Battery Room 5 Exhaust Fan

. 6VK 6VK Instrument RK & Computer Rooms A/C Unit 3HVC*ACU2A

. 6VL 6VL Instrument RK & Computer Rooms A/C Unit 3HVC*ACU2B

. 6VM 6VM Stm Gen Aux Fdwtr Isol Valve [3FWA*MOV35A]

. 6VN 6VN Stm Gen Aux Fdwtr Isol Valve [3FWA*MOV35B]

. 6VP 6VP Stm Gen Aux Fdwtr Isol Valve [3FWA*MOV35C]

. 6VQ 6VQ Stm Gen Aux Fdwtr Isol Valve [3FWA*MOV35D]

. 6VV 6VV Aux Fdwtr Pump Turb Stm Supply Non-return Valve

[3MSS*MOV17A]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6VW 6VW Aux Fdwtr Pump Turb Stm Supply Non-return Valve

[3MSS*MOV17B]

. 6VX 6VX Aux Fdwtr Pump Turb Stm Supply Non-return Valve

[3MSS*MOV17D]

. 6VY 6VY Main Stm Press Rel Isol Valve [3MSS*MOV18A] (AO)

. 6VZ 6VZ Main Stm Press Rel Isol Valve [3MSS*MOV18B] (BP)

. 6WA 6WA Main Stm Press Rel Isol Valve [3MSS*MOV18C] (CO)

. 6WB 6WB Main Stm Press Rel Isol Valve [3MSS*MOV18D] (DP)

. 6ZH 6ZH EGE MCC Fdr Bkr [3EHS*MCC1A1]

. 6ZJ 6ZJ Cont Bldg MCC Fdr Bkr [3EHS*MCC1A2]

. 6ZK 6ZK Aux Bldg MCC Fdr Bkr [3EHS*MCC3A1]

. 6ZL 6ZL Sfgrds Area MCC Fdr Bkr [3EHS*MCC1A4]

. 6ZM 6ZM Circ & Serv Wtr Pp Hse MCC Fdr Bkr [3EHS*MCC1A5]

. 6ZN 6ZN EGE MCC Fdr Bkr [3EHS*MCC1B1]

. 6ZP 6ZP Cont Bldg MCC Fdr Bkr [3EHS*MCC1B2]

. 6ZQ 6ZQ Aux Bldg MCC Fdr Bkr [3EHS*MCC3B1]

. 6ZR 6ZR Sfgrds Area MCC Fdr Bkr [3EHS*MCC1B4]

. 6ZS 6ZS Circ & Serv Pp Hse MCC Fdr Bkr [3EHS*MCC1B5]

. 6ZT 6ZT Rod Contmt Area MCC Fdr Bkr [3EHS*MCC2A1]

. 6ZU 6ZU Rod Contmt Area MCC Fdr Bkr [3EHS*MCC2B1]

. 6ZV 6ZV Turb Bldg MCC Fdr Bkr [3EHS*MCC2A2]

. 6AAA 6AAA Contmt Recirc Cir Out VV57A

. 6AAB 6AAB Contmt Recirc Cir Out VV57B

. 6AAC 6AAC Contmt Recirc Cir Out VV57C

. AAD AAD Contmt Recirc Cir Out VV57D

. 6AAF 6AAF Contmt Recirc Cir Supply VV54A

. 6AAG 6AAG Contmt Recirc Cir Supply VV54B

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6AAH 6AAH Contmt Recirc Cir Supply VV54C

. 6AAJ 6AAJ Contmt Recirc Cir Supply VV54D

. 6AAK 6AAK Rx Plt Compnt Clg Hx Supply VV50A

. 6AAL 6AAL Rx Plt Compnt Clg Hx Supply VV50B

. 6AAM 6AAM TPCCW Hx Inlet Valve [3SWP*MOV71A]

. 6AAN 6AAN TPCCW Hx Inlet Valve [3SWP*MOV71B]

. 6AAU 6AAU Service Wtr Pmp Disch Valve [3SWP*MOV102A]

. 6AAV 6AAV Service Wtr Pmp Disch Valve [3SWP*MOV102B]

. 6AAW 6AAW Service Wtr Pmp Disch Valve [3SWP*MOV102C]

. 6AAX 6AAX Service Wtr Pmp Disch Valve [3SWP*MOV102D]

. 6AAY 6AAY SI Pump Cooling

. 6ABG 6ABG Diesel Gen "A" Enc Ventilation Supply Fan [3HVP*FN1A,C]

. 6ABH 6ABH Diesel Gen "B" Enc Ventilation Supply Fan [3HVP*FN1B,D]

. 6ABP 6ABP 480 V Motor Cont Emer Gen A&B Misc Equipment Fdr

. 6ABZ 6ABZ Emer Gen Crankcase Vac Pmp [3EGD*P1A, B]

. 6ACA 6ACA Emerg Gen "A" Air Compr [3EGA-C1A, C2A]

. 6ACB 6ACB Emer Gen "B" Air Compr [3EGA-C1B, C2B]

. 6ACL 6ACL 480 V Mc Diesel Gen Enc+ "A" Ventilation Dampers

[3HVP*MOD20A, 23A, 26A, 20C]

. 6ACM 6ACM 480 V Mc Diesel Gen Enc+ "B" Ventilation Dampers

[3HVP*MOD20B, 23B, 26B, 20D]

. 6ACN 6ACN Diesel Gen "A" Enc Ventilation Supply Fan [3HVP*FN1C]

. 6ACP 6ACP Diesel Gen "B" Enc Ventilation Supply Fan [3HVP*FN1D]

. 6ACZ 6ACZ Service Water Pumphouse Exh Fans & Inlet Dampers 3HVY*FN2A&B; 3HVY*AOD23A&B

. 6ADH 6ADH Main Steam Valve Bldg Exh Fan 1B

. 6ADR 6ADR SI, QS & RHR Pps Area A/C Units [3HVQ*ACUS1A,B]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6ADS 6ADS Contmnt Recirc Pp & Clr Area A/C Units [3HVQ*ACUS2A,B]

. 6ADU 6ADU Aux Fdwtr Pp Area Emer Sply Fans [3HVQ*FN5A,B]

. 6ADV 6ADV Aux Fdwtr Pp Area Emer Vent Dmprs

[3HVQ*MOD26A1, B1,C1]

. 6ADW 6ADW Aux Fdwtr Pp Area Emer Exh Fans [3HVQ*FN6A,B]

. 6ADX 6ADX Charging Pp Aux Lube Oil Pp (3CHS*P6C)

. 6AEA 6AEA Charging Pp Area Inlet & Outlet Air Dampers

[3HVR*MOD49A,50A, 49B*50B]

. 6AEB 6AEB Charging Pp Area Inlet & Outlet Air Dampers

[3HVR*MOD49C1, 50C1,49C2,50C2]

. 6AEC 6AEC Elem Diag 120V ac Aux & Fuel Bldg Vent Dampers

[3HVR*MOD28A,28B]

. 6AFA 6AFA Bus 32T Tie Bk to Bus 32S 32T11-2 [3EJS*ACB-T11]

. 6AFB 6AFB Bus 32S Tie Bkr to Bus 32R 32S11-2 [3EJS*ACB-S11]

. 6AFC 6AFC Bus 32U Tie Bkr to Bus 32V 32U11-2 [3EJS*ACB-U11]

. 6AFD 6AFD Bus 32V Tie Bkr to Bus 32W 32V11-2 [3EJS*ACB-V11]

. 6AFE 6AFE Aux Bldg Filter Unit [3HVR*FLT1A]

. 6AFF 6AFF Aux Bldg Filter Unit [3HVR*FLT1B]

. 6AFG 6AFG Fuel Bldg Filter Unit [3HVR*FLT2A]

. 6AFH 6AFH Fuel Bldg Filter Unit [HVR*FLT2]

. 6AFK 6AFK Cont Bldg Chiller Condenser Inlet Temp Cont Valves (3SWP*TV35A&B)

. 6AFL 6AFL Cntmt Recirc Pp Miniflow Valve [3RSS*MOV38A]

. 6AFM 6AFM Cntmt Recirc Pp Miniflow Valve [3RSS*MOV38B]

. 6AFZ 6AFZ Containment Open Pressure Tap Isolation Valve

[3LMS*MOV40A]

. 6AGA 6AGA Containment Open Pressure tap Isolation Valve

[3LMS*MOV40B]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6AGB 6AGB Containment Open Pressure Tap Isolation Valve

[3LMS*MOV40C]

. 6AGC 6AGC Containment Open Pressure Tap Isolation Valve

[3LMS*MOV40D]

. 6AGD 6AGD Aux Fd Pps Area Emer Vent Dmprs

[3HVQ*MOD26A2, B2, C2]

. 6AGF 6AGF Circulating Water Valve 3SWP*MOV115A

. 6AGG 6AGG Circulating Water Pump Lube Water Valve 3SWP*MOV115B

. 6AGH 6AGH Containment Instr Air Isolation Valve [3IAS*MOV72]

. 6AGX 6AGX Fuel Bldg Vent Dampers [3HVR*MOD72A,B]

. 6AGY 6AGY Aux Fdwtr Pp Area Emer Sply Fan

. 6AHB 6AHB Suppl Leak Coll Rel Sys Flt Htrs

. 6AHL 6AHL Contmt ATM Monit Inside Contmt Isol Valve [3CMS*MOV24]

. 6AJH 6AJH Elem Diag 480 V MC Reactor Vessel to Excess Ltdn Valve (3RCS*V8098)

. 6AJJ 6AJJ Elem Diag 480 V MC Boric Acid Gravity Feed Valve (3CHS*V8507A)

. 6AJK 6AJK Elem Diag 480 V MC Boric Acid Gravity Feed Valve (3CHS*MV8507B)

. 6AJL 6AJL Elem Diag 480 V MC LPSI Charging Pp Suct Valve (3CHS*V8468A)

. 6AJM 6AJM Elem Diag 480 V MC LPSI Charging Pipe Suct Valve (3CHS*MV8468B)

. 6AJN 6AJN Elem Diag 480 V MC Charging Pp Recirc Valve (3CHS*MV8111B)

. 6AJP 6AJP Elem Diag 480 V MC Charging Pp Recirc (3CHS*MV8111C)

. 6AJQ 6AJQ Elem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV8438A)

. 6AJR 6AJR Elem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV8438B)

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6AJS 6AJS Elem Diag 480 V MC Charging Hdr Isol Valve (3CHS*MV-8438C)

. 6AJU 6AJU Elem Diag 480 V MC Contmt Air Recirc Coil Sply Valve (CCP*MOV222)

. 6AJV 6AJV Elem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV223)

. 6AJW 6AJW Elem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV224)

. 6AJX 6AJX Elem Diag 480 V Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV225)

. 6AJY 6AJY Elem Diag 480 V Sply Valve Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV226)

. 6AJZ 6AJZ Elem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV227)

. 6AKA 6AKA Elem Diag 480 V MC Contmt Air Recirc Coil Sply Valve (3CCP*MOV228)

. 6AKB 6AKB Elem Diag 480 V MC Contmt Air Recirc Clg Coil Sply Valve (3CCP*MOV229)

. 6AKC 6AKC Elem Diag 480 V MC Charging Hdr Isol Valve (3CHSX-MV8116)

. 6AKE 6AKE Contmt Isol Valve (3CCP*MOV45A)

. 6AKF 6AKF Contmt Isol Valve (3CCP*MOV45B)

. 6AKG 6AKG Contmt Isol Valve (3CCP*MOV48A)

. 6AKH 6AKH Contmt Isol Valve (3CCP*MOV48B)

. 6AKJ 6AKJ Contmt Isol Valve (3CCP*MOV49A)

. 6AKK 6AKK Contmt Isol Valve (3CCP*MOV49B)

. 6AKL 6AKL Elem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74A

. 6AKM 6AKM Elem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74B

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6AKN 6AKN Elem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74C

. 6AKP 6AKP Elem Diag 480 V MC Mn Stm Press Relief Cntrl Bypass Valve 3MSS*MOV74D

. 6AKX 6AKX Elem Diag 480 V ESF Bldg MCC Fdr Bkr (3EHS*MCC1B3)

. 6AKY 6AKY Elem Diag 480 V Emer Supply Bkr (3EJS*ACB-AD)

. 6AKZ 6AKZ Elem Diag 480 V Emer Supply Bkr [3EJS*ACB-BD]

. 6ALA 6ALA Elem Diag 480 V Bus 32Y Tie to Bus 32R 32T 4A-2

[3EJS*ACB T4A]

. 6ALB 6ALB Elem Diag 480 V Bus 32X Tie Bkr to Bus 32W 32T4B-2

[3ETS*ACB-T4B]

. 6ALC 6ALC Elem Diag 480 V US Spare MCC Fdr Bkr [3EHS*MCC4A1]

. 6ALD 6ALD Elem Diag 480 V US Spare MCC Fdr Bkr

[3EHS*Bkr MCC4B1]

. 6ALG 6ALG Elem Diag 480 V MC (3SWP-3A & B)

. 6ALH 6ALH Elem Diag 480 V MC (3STR 2A & B)

. 6AMB 6AMB Elem Diag 480 V MC Cold Shutdown Air Compressor

[3IAS-C2A]

. 6AMC 6AMC Elem Diag 480 V MC Cold Shutdown Air Compressor

[3IAS-C2B]

. 6AMG 6AMG Mn Stm Valve Bldg Vent Exh Fan Dampers

. 6AMH 6AMH Mn Stm Valve Bldg Vent Exh Fan Dampers

. 6AMJ 6AMJ Mn Stm Valve Bldg Vent Inlet Dampers

. 6AMK 6AMK Mn Stm Valve Bldg Vent Inlet Dampers

. 6BAA 6BAA Air Sample Pump

. 6BAB 6BAB Elem Diag 480 V MC Charging Pp Miniflow Cntrl Valve

[3CHS*MV8512A]

. 6BAC 6BAC Elem Diag 480 V MC Charging Pp Miniflow Cntrl Valve

[3CHS*MV8512B]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 6BAD 6BAD Elem Diag 480 V MC Charging Pp Miniflow Cntrl Valve

[3CHS*MV8511A]

. 6BAE 6BAE Elem Diag 480 V MC Charging Pp Miniflow Cntrl Valve

[3CHS*MV8511B]

. 6BAF 6BAF Elem Diag 480 V Fuel Bldg Fltr Exh Fan [3HVR*FN10A2]

. 6BAG 6BAG Elem Diag 480 V Fuel Bldg Fltr Exh Fan [3HVR*FN10B2]

. 6BAN 6BAN Elem Diag. Flow to Monitor for Pzr SVs

. 6BAW 6BAW CRDM Fan 3-HVU-FN2A PENSEC PROT Bkr

. 6BAX 6BAX CRDM Fan 3-HVU-FN2B PENSEC PROT Bkr

. 6BAZ 6BAZ Cont Struc Air Recirc FN 3-HVU-FN1A

. 6BBA 6BBA Cont Struc Air Recirc FN 3-HVU-FN1B PENE Sec Prot Bkr

. 6BBE 6BBE 120 V DC Halon Disch Timer

. 7J 7J 4.16 kV Bus 34C [3NNS-SWG-A] Aux Ckt

. 7L 7L 4.16 kV Bus 34D [3ENS*SWG-B] Aux Ckt

. 7Q 7Q Emer Diesel Gen Bkr [3ENS*ACB-G-A] Aux Circuit 15G-14U-2

. 7R 7R Emer Diesel Gen Bkr [3ENS*ACB-G-B] Aux Circuit 15G-15U-2

. 7W 7W Emerg Diesel Clr Outlet Valves [3SWP*AOV39A&B]

. 7AD 7AD Stm Jet Air Ejector Stm Valves [3ASS-AOV22A,B]

. 7AJ 7AJ Aux Fdwtr Alt Suct Valves [3FWA*AOV23A,23B]

. 7AL 7AL Dwst Htr Circ Line Isol Valve [3FWA*AOV25, 26]

. 7AM 7AM Refuel Wtr Recirc Pp Suct Isol Valve (3QSS*AOV27, 28)

. 7AN 7AN Accumulator Nit Test Line Isol SIP Hot Leg Test Line

[3SIL*CV8880] [SIL*CV8825]

. 7AP 7AP Test Line Header [3SIH*CV8964] & [3SIH*8871]

. 7AQ 7AQ Accumulator Gas Sply & Vent [3SIL*AV8875A]

Accumulator Fill Line Isol [3SIL*AV8878A]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7AR 7AR Accumulator Gas Sply & Vent [3SIL*AV8875B]

Accumulator Fill Line Isol [3SIL*AV8878B]

. 7AS 7AS Accumulator Gas Sply & Vent [3SIL*AV8875A]

Accumulator Fill Line Isol [3SIL*AV8878C]

. 7AT 7AT Accumulator Gas Sply & Vent [3SIL*AV8875D]

Accumulator Fill Line Isol [3SIL*AV8878D]

. 7AU 7AU RHR Pmp Hot Leg Test Line [3SIL*CV8890A, CV8890B]

. 7AY 7AY Mn Stm Isol Trip Valve [3MSS*CTV27A]

. 7AZ 7AZ Mn Stm Isol Trip Valve [3MSS*CTV27B]

. 7BA 7BA Mn Stm Isol Trip Valve [3MSS*CTV27C]

. 7BB 7BB Mn Stm Isol Trip Valve [3MSS*CTV27D]

. 7BC 7BC Mn Stm Line Isol Bypass Valve [3MSS*HV28A]

. 7BD 7BD Mn Stm Line Isol Bypass Valve [3MSS*HV28B]

. 7BE 7BE Mn Stm Line Isol Bypass Valve [3MSS*HV28C]

. 7BF 7BF Mn Stm Line Isol Bypass Valve [3MSS*HV28D]

. 7BG 7BG SIP Hot Leg Test Line [3SIH*CV8824] & [3SIH* CV8881]

. 7BH 7BH Stm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22A]

. 7BJ 7BJ Stm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22B]

. 7BK 7BK Stm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22C]

. 7BL 7BL Stm Gen Blwdn Cntrl Isol Valve [3BDG*CTV22D]

. 7BP 7BP Rx Pl Gaseous Drs Inside Isol Valve [3DGS*CTV24]

. 7BQ 7BQ Rx Pl Gaseous Drs Outside Cntmt Isol Valve [3DGS*CTV25]

. 7BR 7BR Cntmt Drs Isol Valves [3DAS**CTV24] & [3DAS*CTV25]

. 7BT 7BT Cntmt Drs Isol Valve [3DAS*CTV25]

. 7CM 7CM Chg to Rx Cool Sys Isol Valves [3CHS*AV8147, 8146]

. 7CS 7CS Rx Coolant Makeup Cont Aux, Circ, RPS-RAK Aux A,C

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7CT 7CT BA Inj Valve to BA Blender VV 110A BA Mkup Inj Valve to Charging Pp Hdr VV 110B

. 7CU 7CU Ba Mkup Wtr Inj Valve to BA Blender VV 111A BA Dil Inj Value to Vol Cntrl Tk VV 111B

. 7CW 7CW Vol Cntrl Tk Isol & Inert Valves [3CHS*AV8101, LCV112A]

Elem Diag 480 V

. 7CY 7CY Emerg Bux 32R, S, T, Y Aux Ckt Elem Diag

. 7CZ 7CZ 480 V Emerg Bus Bus 32U, V, W, X Aux Ckts

. 7DF 7DF Cntmt Purge Norm & Fltr Exh Dmprs

. 7DG 7DG Chg Pp & CCW Norm & Fltr Exh Dmprs

. 7DH 7DH Waste Dspl & Pipe Tunnel Norm & Fltr Exh Dmprs

. 7DJ 7DJ Aux Bldg Norm & Fltr Exh Dmprs

. 7DK 7DK Aux Bldg Norm & Fltr Exh Dmprs

. 7DL 7DL Fuel Bldg Fltr & Bypass Dmprs

. 7DV 7DV Reac Coolant Letdn Drn Valves & Excess Letdn Isol Valves

[3RCS*AV8032, 8153]

. 7DW 7DW Pressurizer Power Power Relief Valves [3RCS*PCV455A, 456]

. 7DW1 7DW1 Pressurizer Pwr Relief U/V 3RCS*456

. 7DX 7DX Letdown Line Isol Valves [3RCS*LCV459, 460]

. 7DZ 7DZ Reactor Coolant System Spray VV 8145 Pressurizer Relief Tank Press VV 469

. 7EK 7EK Chg Pump Test Line Isol [3SIH*CV8843] & [3SIH*AV8882]

. 7EM 7EM Accumulator Fill Line 3SIH*CV8888 Skip Cold Leg Test Line 3SIH*CV8823

. 7EZ 7EZ Pri Gr Wtr Cntmt Isol Valves [3PGS*CV8026, 8046]

. 7FG 7FG 4.16 kV Bus [3ENS*SWG-A] Aux Circuit

. 7FH 7FH 4.16 kV Bus [3ENS*SWG-B] Aux Circuit

. 7FP 7FP Main Steam Line Drains [3DTM*AOV63, 64]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7FS 7FS Main Steam Line Drains [3DTM*AOV29A,B]

. 7FT 7FT Main Steam Line Drains [3DTM*AOV29C, D]

. 7FU 7FU Main Steam Line Drains [3DTM*AOV61A, B]

. 7FV 7FV Main Steam Line Drains [3DTM*AOV61C, D]

. 7HK 7HK SG Fdwtr Bypass Valve & Cntrl Valves Auxiliary Ckt

[3FWS*LV550]

. 7HL 7HL SG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV560]

. 7HM 7HM SG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV570]

. 7HN 7HN SG Fdwtr Cntrl Valve Bypass Valve [3FWS*LV580]

. 7HP 7HP SG Mn Fdwtr Flow Cntrl Valve [3FWS-FCV510]

. 7HQ 7HQ SG Mn Fdwtr Flow Cntrl Valve [3FWS-FCV520]

. 7HR 7HR SG Mn Fdwtr Flow Cntrl Valve (3FWS*FCV530)

. 7HS 7HS SG Mn Fdwtr Flow Cntrl Valve 3FWS*FCV540

. 7HX 7HX Turb Bypass Cntrl Sys Aux Ckt

. 7JC 7JC Letdown Orifice Isol Valve [3CHS*AV8149A]

. 7JD 7JD Letdown Orifice Isol Valve [3CHS*AV8149B]

. 7JE 7JE Letdown Orifice Isol Valve [3CHS*AV8149C]

. 7JK 7JK Chg Pps Clrs Out Crossover Valves [3CCE*AOV26A,B]

. 7JL 7JL Chg Pmps Disch Crossover Valves [3CCE*AOV30A,B]

. 7JN 7JN SG Fdwtr Isol Valve [3FWS*CTV41A]

. 7JP 7JP SG Fdwtr Isol Valve [FWS*CTV41B]

. 7JQ 7JQ SG Fdwtr Isol Valve [3FWS*CTV41C]

. 7JR 7JR SG Fdwtr Isol Valve [3FWS*CTV41D]

. 7JY 7JY Cntmt Inst Air Isol Valve [3IAS*CTV15]

. 7KJ 7KJ 120 VAC Cont Bldg Aux Ckts

. 7KK 7KK CVS Isol Valves (3CVS*CTV20A,B)

. 7KL 7KL CVS Isol Valves (3CVS*CTV21A,B)

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7KQ 7KQ Serv Wtr Pps Train A & B Aux Ckts

. 7KT 7KT SG #1 Blwdwn Sample Isol VV19A

. 7KU 7KU SG #2 Blwdwn Isol VV19B

. 7KV 7KV SG #3 Blwdwn Isol VV19C

. 7KW 7KW SG #4 Blwdwn Sample Isol VV19D

. 7KX 7KX Pzr Vapor Space Sample Isol VVs 20&21

. 7KZ 7KZ Rx Clnt Hot Leg Sample Isol VVs 26&27

. 7LA 7LA Rx Clnt Cold Leg Sample Isol VVs 29&30

. 7LB 7LB SI Inj Accumulator Sample Isol VVs 32&33 Aux Circuit

. 7LC 7LC Pzr Rlf Tk Gas Space Sample Isol Valves (3SSR*CV8026, 8025)

. 7LX 7LX Non-Sfty Hdr Sply Rtn Valves [3CCP*AOV10A, B,&19A,B]

. 7MA 7MA Cntmt Isol Valves [3CCP*CTV49A,B]

. 7MB 7MB Residual Heat Removal Outlet Valves [3CCP*AOV66A&B]

. 7MC 7MC Thermal Barrier Coolant Return Valves [3CCP*AOV178A&B]

. 7MD 7MD Thermal Barrier Coolant Return Valves [3CCP*AOV178C&D]

. 7ME 7ME Comp Cooling Cross Conn VVs 179A&B

. 7MF 7MF Comp Cooling Cross Conn VVs 180A&B

. 7MG 7MG Non-Sfty Hdr Sply & Rtn Values

[3CCP*AOV194A, B, & 197A,B]

. 7MW 7MW Ctmt Spray Test Inter Circuitary

. 7MX 7MX Diesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)

. 7MY 7MY Diesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)

. 7MZ 7MZ Diesel Gen "B" Enclosure Vent Outlet Damper (3HVP*MOD20F,H)

. 7NA 7NA 125V Hx Temp Cont VIV 3CCP*SOU32A & 32C-1

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7NB 7NB 125V Hx Temp Cont VIV 3CCP*SOU32B & 32C-2

. 7PB 7PB Control Room Ventilation Outlet Air Isol VVs 20,21

[HVA*AOV20,21]

. 7PC 7PC Control Room Ventilation Outlet Air Isol VVs 22,23

[HVC*AOV22,23]

. 7PD 7PD Control Room Ventilation Outlet Air Isol VVs 25,26

[HVC*AOV25,26]

. 7PJ 7PJ Purge Supply Fan Inlet Damper AOD135 Emerg Vent Air Return Damper AOD119

. 7PK 7PK Cntrl Bldg Makeup Air Dmprs, (3HVC*AOD 27A,B)

. 7PL 7PL Air Stor Tk 1A Outlet VVs SOV 74A & 74B

. 7QD 7QD Turbine-Driven Aux Fdwtr Pp Aux Oil Pp (3FWL*PS), Trip &

Throttle Valve (3MSS*M5V5)

. 7QE 7QE Mn Stm Rel & Drain Valves [3MSS*PV20A, 3SVV-AOV20A]

. 7QF 7QF Mn Stm Rel & Drain Valves [3MSS*PV20B, 3SVV-AOV20B]

. 7QG 7QG Mn Stm Rel & Drain Valves [3MSS*PV20C, 3SVV-AOV20C]

. 7QH 7QH Mn Stm Rel & Drain Valves [3MSS*PV20D, 3SVV-AOV20D]

. 7QM 7QM Chilled Wtr Cntmt Isol Valves [3CDS*CTV38A,B]

. 7QN 7QN Chilled Wtr Cntmt Isol Valves [3CDS*CTV39A,B]

. 7QP 7QP Chilled Wtr Cntmt Isol Valves [3CDS*CTV40A,B]

. 7QT 7QT Reactor Coolant Letdown Valves [3CHS*TCV12G, AV8143]

. 7QU 7QU Letdown Line Isol Valves [3CHS*CV81528160]

. 7QV 7QV Aux Bldg Inlet Dampers [3HVR*AOD33A, 35A]

. 7QW 7QW Aux Bldg Inlet Dampers [3HVR*AOD33B, 35B]

. 7QX 7QX Pipe Tunnel Inlet Dampers [3HVR*AOD85,86]

. 7QY 7QY Cntmt Purge Inlet Dampers [3HVR*AOD55A/B & 174A/B]

. 7QZ 7QZ Chg & Compnt Clg Pps Temp Cntrl Dmprs

[3HVR*AOD45B1, B2, C1, C2]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7RB 7RB Reac Clnt Loop Stop Valve Aux Buffering Circuit

. 7RC 7RC Reac Clnt Loop Stop Valve Aux Buffering Circuit

. 7RF 7RF TDAFW Pmp Mtr Speed changr 3FWA*M7

. 7RG 7RG ESF Bldg Vent Dampers

. 7RH 7RH ESF Bldg Vent Dampers

. 7RK 7RK Containment Purge Inlet & Outlet Isol Valves

[3HVU*CTV32A&B], [3HVU*CTV33A&B]

. 7RM 7RM Containment Atmosphere Monitoring

. 7RZ 7RZ SG Level Aux Ckts: Mn Turb Trip

. 7SC 7SC SG Chemical Feed Pp Isolation Valves [3SGF*AOV24A&B]

. 7SD 7SD SG Chemical Feed Pp Isol Valve [3SGF*AOV24C&D]

. 7SF 7SF Misc Level Ind Lights DAS System

. 7SH 7SH Rx Pl Gas Vents Hdr Cntrl Trip Valves (3VRS*CTV20,21)

. 7SM 7SM Gas Waste to Unit 1 Stack Isol Dmprs (3GWS*AOD78A,B)

. 7SP 7SP Pzr Relief Tk Nitrogen Sply Isol Valves (3GSN*CV8033, CTV105)

. 7SS 7SS ESF Manual Actuation Ckts

. 7ST 7ST ESF Manual Reset Ckts

. 7SX 7SX 4.16 kV Bus 34C (3ENS*SWG-A) Undervoltage (Hi STPT) Trip Ckt

. 7SY 7SY 4.16 kV Bus 34D (3ENS*SWG-B) Undervoltage (Hi STPT) Trip Ckt

. 7SZ 7SZ 4.16 kV Bus 34C&D (3ENS*SWG-A&B) Undervoltage (Hi STPT) Relays

. 7TA 7TA Cntrl Bldg Isol (Train A) Sh. 1

. 7TB 7TB Cntrl Bldg Isol (Train A) Sh. 2

. 7TC 7TC Cntrl Bldg Isol (Train B) Sh. 1

. 7TD 7TD Cntrl Bldg Isol (Train B) Sh. 2

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7TN 7TN Containment Fire Protection Water Isolation Valves 3FPW &

CTV 48,49

. 7TP 7TP ESF Aux Relays (Train A)

. 7TQ 7TQ ESF Aux Relays (Train A)

. 7TR 7TR ESF Aux Relays (Train B)

. 7TS 7TS ESF Aux Relays (Train B)

. 7TY 7TY Accumulator N2 Line Isol Valve 3SIL*CVS968

. 7TZ 7TZ Chilled Wtr Cntmt Isol Valves [3CDS*CTV91A,B]

. 7UF 7UF Auto and Manual Rod Withdrawal Block

. 7UK 7UK Accumulator Gas Sply & Vent Valves (3SIL*SV8875F&G)

. 7UL 7UL Accumulator Gas Sply & Vent Valves (3SIL*SV8875F&G)

. 7UM 7UM Rx Vessel Head Vent Isol Valve (3RCS*SV8095A1B)

. 7UN 7UN Rx Vessel Head Vent Isol Valve (3RCS*SV0896A1B)

. 7UR 7UR Elem Diag Water Feed to Chlorination System

[3WTC*AOV25A&B]

. 7UT 7UT Elem Diag SI Accumulator Tk Vent Valves

. 7UU 7UU Elem Diag ESF Manual Reset Caskets

. 7UV 7UV Elem Diag Reactor Reactor Vessel Head Vent Valves

. 7UW 7UW Elem Diag Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27A)

. 7UX 7UX Elem Diag Mn Stm Trip Valve Aux Ckt (3MSS*CTV27B)

. 7UY 7UY Elem Diag Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27C)

. 7UZ 7UZ Elem Diag Mn Stm Mn Stm Isol Trip Valve Aux Ckt (3MSS*CTV27D)

. 7VA 7VA Analog Isolation Ckts

. 7VB 7VB Analog Isolation Ckts

. 7VC 7VC Elem Diag Change Hdr Flow Valves

. 7VD 7VD Elem Diag (3DTM*AOV63B, 64B) Main Steam Line Drains

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7VE 7VE Elem Diag (3DTM*AOV64B, 64C) Main Steam Line Drains

. 7VG 7VG Elem Diag ASP Lamp Test Aux Ckts

. 7VH 7VH Elem Diag ASP Lamp Test Aux Ckts Sequencer Status Lights

. 7VJ 7VJ Elem Diag 125 V dc Cntmt Purge Inlet Dampers (3HVR*AOD55B & 174B)

. 7VK 7VK Elem Diag 125 V dc 125 V dc Digital Isolator Circuits

. 7VM 7VM Elem Diag 120 V ac Main Vent & Air Cond Panel Lamp Test Ckt, Train "A"

. 7VN 7VN Elem Diag 120 V ac Main Vent & Air Cond Panel Lamp Test Ckt, Train "B"

. 7VQ 7VQ Elem Diag 120 V ac Main Vent & Air

. 7VX 7VX DC Digital Isol Circ 3BYS-PNL-5 & 6

. 7VZ 7VZ Elem Diag Digital Isolator Circuits

. 7WB 7WB Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WC 7WC Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WD 7WD Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WF 7WF Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WG 7WG Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WJ 7WJ Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WL 7WL Elem Diag 120 V ac MCB Lamp Test Ckt

. 7WM 7WM Elem Diag 120 V ac MB4

. 7WP 7WP Elem Diag 120 V ac MB3

. 7WR 7WR Elem Diag 120 V ac MB3

. 7WS 7WS Elem Diag 120 V ac MB2

. 7WU 7WU Elem Diag 120 V ac MB2

. 7WV 7WV Elem Diag 120 V ac MB1

. 7WX 7WX Elem Diag 120 V ac MB1

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7XA 7XA Aux Fdwtr Control Valves

. 7XB 7XB Aux Fdwtr Control Valves

. 7XC 7XC Aux Fdwtr Control Valves

. 7XD 7XD Aux Fdwtr Control Valves

. 7XJ 7XJ Elem Diag 125 V dc Chilled Water Diff Press Cntrl Valve

[3HVK*POV32A, 32B]

. 7XK 7XK Elem Diag 125 V dc Inst Rack & Comp Rms A/C Unit Temp Cntrl Valve

. 7XL 7XL Elem Diag 125 V dc Cntrl Rm A/C Unit Temp Cntrl Valve

[3HVK*TV41A, 41B]

. 7XM 7XM Elem Diag 125 V dc E&W Swgr Rms A/C Units Temp Cntrl Valve [3HVK*TV76A, 76B]

. 7XN 7XN Elem Diag 125 V dc E&W Swgr Rms Backup A/C Units Temp Cntrl Valve [3HVK*TV77A, 77B]

. 7XP 7XP Chg Pps Clg Valves

. 7XQ 7XQ Elem Diag (3HVZ*MOD20A, 21A)

. 7XR 7XR Elem Diag (3HVZ*MOD20B, 21B)

. 7XV 7XV Elem Diag 120 V ac MB4

. 7XW 7XW Elem Diag 120 V ac MCB Lamp Test Ckt

. 7XZ 7XZ Elem Diag CDA Signal Reset Aux Ckt

. 7AAG 7AAG Elem Diag Control Bldg Chilled Wtr Isol Valves

. 7AAH 7AAH Elem Diag Control Bldg Chilled Wtr Isol Valves

. 7AAJ 7AAJ Elem Diag Control Bldg Chilled Wtr Isol Valves

. 7AAK 7AAK Elem Diag Control Bldg Chilled Wtr Isol Valves

. 7AAP 7AAP Aux Fdwtr Cntrl Valves [3FWA*HV31A, 32A & 36B]

. 7AAQ 7AAQ Aux Fdwtr Cntrl Valves [3FWA*HV31B, 32B & 36A]

. 7AAR 7AAR Aux Fdwtr Cntrl Valves [3FWA*HV31C, 32C &36D]

. 7AAS 7AAS Aux Fdwtr Cntrl Valves [3FWA*HV31D, 32D & 36C]

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 7ABC 7ABC Elem Diag Prst Accid Simpl VVS 3SSP*SOV1A, 1B

. 7ABD 7ABD Elem Diag Prst Accid Simpl VVS 3SSP*SOV1C, 1D

. 7ABE 7ABE Elem Diag Prst Accid Simpl VVS 3SSP*SOV2A, 2B

. 7ABF 7ABF Elem Diag Prst Accid Simpl VVS 3SSP*SOV3, 5

. 7ABG 7ABG Elem Diag Post-Accident Sample Valves 3SSP*CTV7,8

. 7ABP 7ABP Elem Diag Cntmt Recirc Isol Valves 3SSP*SOV25A,25B

. 7ABU 7ABU Elem Diag Rx Trip on Turb Trip Isol

. 7ABV 7ABV Elem Diag Rx Trip on Turb Trip Inputs

. 7ABZ 7ABZ Aux Stm to Aux Bldg Isol Valves [3ASS-AOV102A & B]

. 7ABZ1 7ABZ1 Aux Stm to Aux Bldg Temp Switches

. 7ACA 7ACA Aux Bldg Hot Wtr Htg Sys Inlet & Outlet Valves

. 7ACB 7ACB Aux Bldg Hot Wtr Htg. Sys Inlet & Outlet Valves

. 7ACD 7ACD Elem Diag 125 V dc Hot Wtr Preheating Sply & Return Isol

. 7ACE 7ACE Elem Diag 125 V dc Hot Wtr Preheating Sply & Return Isol

. 7ACF 7ACF Fuel Bldg Air Suply Isol Damper 3HVR*184

. 7ACG 7ACG Dig Isol CK & for Fuel Bldg A/S Isol Dmpr

. 7ACL 7ACL Elem Diag 125 V DC Demin Wtr Stor TK to Aux Fd Wtr Pmp Suc 3FWA*AOV61A, 61B

. 7ACM 7ACM Aux Fd Wtr Pmp Disch Crossover 3FWA*ALU62A, 62B

. 7ACP 7ACP Chrg Pmp Comp Cooling Pmp Cubicle to Aux Bldg Flt Damper 3HVR*MOD 46A, 46B

. 7ACQ 7ACQ Aux Bldg Exh Fn Var Inlet Valve 3HVR*MOD 140A, 140B

. 7ACW 7ACW Mn. Stm Bldg Exh Fn Dampers 3HVV*AOD50A2, 50B2

. 7ACY 7ACY Tr A Reset for low-low Stm Gen Level on Aux Fd Start

. 7ACZ 7ACZ Tr B Reset for low-low Stm Gen Level on Aux Fd Start

. 7ADA 7ADA RHR Hx Flow Control Safety Grade Cold Shutdown

. 8BA 8BA Res Sta Svce X Fmr A Prot (3RTX-XSR-A) three line

(Refer to Plant Document Control for latest Document Rev. and Date)

NUSCO SWEC Drawing Drawing Number Number 5212-32001 12179-ESK- Diagram Title

. 8DA 8DA Synchronizing

. 8DB 8DB Synchronizing

. 8DC 8DC Synchronizing

. 8HC 8HC RSV Sta Svce Bckup Prot Transf Trip RCV CK 3SPRN03

. 8HG 8HG Rsv Sta Svce Pri Prot Xfmr Trip Rcvr

. 8JB 8JB Norm Sta Svce Xfmr Bu Prot

. 8JC 8JC Norm Sta Svce Xfmr Bu Prot

. 8JD 8JD Rsv Sta Svce Xfmr Pri Prot

. 8JF 8JF Rsv Sta Svce Xfmr Pri Prot

. 8KA 8KA Emer Diesel Gen Excita [3EGS*G-A]

. 8KB 8KB Emer Diesel Gen Excita [3EGS*G-B]

. 8KC 8KC Emer Diesel Gen Engine Control [3EGS*G-A]

. 8KD 8KD Emer Diesel Gen Engine Control [3EGS*G-A]

. 8KE 8KE Emer Gen "A" Governor [3EGS*EG-A] 15G-14U

. 8KF 8KF Emer Diesel Gen Engine Control [3EGS*G-B]

. 8KG 8KG Emer Diesel Gen Engine Control [3EGS*G-B]

. 8KH 8KH Emer Gen Governor [3EGS*EG-B] 15G-15U

. 8KJ 8KJ Elem Diag Emerg Diesel Gen 3EGS*EG-A, Shutdown Ckt

. 8KK 8KK Elem Diag Emerg Diesel Gen 3EGS*EG-B, Shutdown Ckt

. 8KL 8KL Alt AC DG 125V DC Hyd Govnr Cntr 3BGS-BG-A

. 11A 11A Reactor Trip Breaker [3RPS*ACB-RTA]

. 11B 11B Reactor Trip Breaker [3RPS*ACB-RTB]

. 11C 11C Reactor Trip Bypass Bkr [3RPS*ACB-BYA]

. 11D 11D Reactor Trip Bypass Bkr [3RPS*ACB-BYB]

. 11H 11H Nuclear Inst Sys

ee Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document Control for Latest Document Rev and Date)

FSAR Figure NUSCO Number Piping and Instrumentation Diagram Title Number 1.2-3 Legend: Piping and Instrumentation Diagram 26900 5.1-1 Reactor Coolant System 26902 5.4.5 Low Pressure Safety Injection System 26912 6.2-36 Quench Spray and Hydrogen Recombiner System 26915 6.2-37 Low Pressure Safety Injection System 26912 6.2-53 Containment Monitoring System 26954 6.3-2 High Pressure Safety Injection System 26913 9.1-6 Fuel Pool Cooling and Purification System 26911 9.2-1 Service Water System 26933 9.2-2 Reactor Plant Component Cooling System 26921 9.2-3 Reactor Plant Chilled Water System 26922 9.2-4 Safety Injection Pump and Neutron Shielding Tank Cooling 26914 System 9.2-5 Charging Pump Sealing and Lubrication System 26905 9.2-6 Condensate Demineralizer Liquid Waste System 26929 9.2-7 Water Treatment System 26920 9.2-8 Domestic Water and Sanitary System 26947 9.2-9 Condensate System 26926 9.2-10 Turbine Plant Component Cooling System 26934 9.2-11 Primary Grade Water System 26919 9.3-1 Compressed Air System 26938 9.3-2 Reactor Plant Sampling System 26944 9.3-3 Turbine Plant Sampling System 26943 9.3-4 Radioactive Gaseous Waste System 26909 9.3-5 Reactor Plant Gaseous Drains System 26907

ee Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document Control for Latest Document Rev and Date)

FSAR Figure NUSCO Number Piping and Instrumentation Diagram Title Number 9.3-6 Radioactive Liquid Waste and Aerated Drains System 26906 9.3-7 Reactor Coolant Pump Seals System 26903 9.3-8 Chemical and Volume Control System 26904 9.3-9 Boron Recovery System 26908 26908 26908 9.3-10 Post Accident Sample Sheet 26955 9.4-1 Control Building Heating, Ventilation, and 26951 Air-Conditioning System 9.4-2 Reactor Plant Ventilation System 26948 9.4-3 Turbine Plant Ventilation System 26950 9.4-4 Office, ESF, and MSV Building Heating, Ventilation, and 26952 Air-Conditioning System 9.4-5 Containment Structure Ventilation System 26953 9.4-6 Service Building Ventilation System 26949 9.4-7 Auxiliary Boiler and Ventilation System 26936 9.4-8 Hot Water Heating System 26937 9.4-9 Technical Support Center HVAC 26956 9.5-1 Fire Protection System 26946 9.5-2 Emergency Generator Fuel Oil System 26917 9.5-3 Emergency Generator Cooling, Starting Air, and 26916 Lube Oil System 9.5-5 Nitrogen and Hydrogen System 26939 10.2-1 Electro-Hydraulic Control System 26940 10.2-2 Turbine Generator and Feed Pump Oil Generator System 26941 10.2-3 Turbine Generator Support System 26942 10.3-1 Main Steam and Reheat System 26923

ee Summary Table of Contents for a Complete List of NUSCO Numbers and Sheets for each FSAR Figure or refer to Plant Document Control for Latest Document Rev and Date)

FSAR Figure NUSCO Number Piping and Instrumentation Diagram Title Number 10.3-2 Turbine Plant Miscellaneous Drains System 26945 10.3-3 Chemical Feed System 26931 10.4-1 Condensate System 26926 10.4-2 Condenser Air Removal and Waterbox Priming System 26927 10.4-3 Extraction Steam and Turbine Gland Seal and 26924 Exhaust System 10.4-4 Circulating Water System 26932 10.4-5 Condensate Demineralizer Mixed Bed System 26928 10.4-6 Feedwater System 26930 10.4-7 Feedheater and MSR Vents and Drains System 26925 10.4-9 Auxiliary Steam, Feedwater, and Condensate System 26935 11.2-1 Radioactive Liquid Waste and Aerated Drains System 26906 11.2-1 Condensate Demineralizer Liquid Waste System 26929 11.3-1 Radioactive Gaseous Waste System 26909 11.4-1 Radioactive Solid Waste System 26910 12.3-5 Containment Monitoring System 26954

TABLE 1.7-3 OMITTED TABLE 1.7-4 SPECIAL REPORTS AND PROGRAMS information contained in Table 1.7-4 is retained for historical purposes. Information provided was relevant at the time of operating license application.

Transmitted at Time of FSAR Submittal A. Failure Modes and Effects Analysis (FMEA)

B. Fire Protection Evaluation Report (FPER)

C. Millstone Nuclear Power Station Emergency Plan D. Modified Amended Security Plan (MASP) (Under Separate Enclosure)

E. Millstone 3 Design Basis Response to Regulatory Guide 1.97, Revision 2 Previously Transmitted Report on Faults and Soil Features Mapped in the Discharge Tunnel Excavation Transmitted after Submittal A. Probabilistic Safety Study (PSS)

B. Control Room Design Review (Chapter 18)

C. Inservice Inspection Program D. Environmental Qualification of Electrical Equipment Report (EEQ)

E. Environmental Qualification of Mechanical Equipment Report (MEQ)

Program to be Reviewed at Millstone 3 Equipment Qualification Documentation (EQD)

le 1.8-1 lists NRC Division 1 Regulatory Guides. This table is annotated whenever a new or sed Regulatory Guide is invoked or adopted. It identifies applicable FSAR sections, and cates the degree of Millstone 3 compliance.

F R.G. S No. Title Degree of Compliance Ref 1.1 Net Positive Suction Head for Emergency Core (1) Comply with the following exception: 6.

Cooling and Containment Heat Removal System For the recirculation system phase, the vapor pressure of the water in the sump 6.

Pumps is assumed to be equal to the containment pressure.

(Rev. 0, November 2, 1970)

The 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 (1) See Section 1.8N. 5.

(Rev. 0, November 2, 1970) 1.3 Assumptions Used for Evaluating the Potential Not applicable.

Radiological Consequences of a Loss-of-Coolant Applicable only to BWRS.

Accident for Boiling Water Reactors (Rev. 2, June 1974) 1.4 Assumptions Used for Evaluating the Potential Comply 12 Radiological Consequences of a Loss-of-Coolant R.G. 1.4 is only used for evaluation of original plant shielding design.

Accident for Pressurized Water Reactors (Rev. 2, June 1974) 1.5 Assumptions Used for Evaluating the Potential Not applicable.

Radiological Consequences of a Steam Line Break Applicable only to BWRS.

Accident for Boiling Water Reactors (Rev. 0, March 10, 1971) 1.6 Independence between Redundant Standby (On site) Comply 8.

Power Sources and Their Distribution Systems (Rev. 8.

0, March 10, 1971) 1.7 Control of Combustible Gas Concentrations in (1) 6.

Comply Containment Following a Loss-of-Coolant Accident (Rev.3, May 2003) 1.8 Personnel Selection and Training (Rev. 1-R, Compliance is as described in the QAPD Topical Report 13 May 1977) 17

F R.G. S No. Title Degree of Compliance Ref 1.9 Selection of Diesel Generator Set Capacity for Comply, with the following exceptions: 8.

Standby Power Supplies (Rev. 2, December 1979) (2) 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).

1.10 Mechanical (Cadweld) Splices in Reinforcing Bars Withdrawn:

of Category I Concrete Structures (Rev. 1, January 2, 3.

1973)

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.

F R.G. S No. Title Degree of Compliance Ref The 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.

F R.G. S No. Title Degree of Compliance Ref The 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.

1.11 Instrument Lines Penetrating Primary Reactor Comply 6.

Containment (Rev. 0, March 10, 1971) 1.12 Instrumentation for Earthquakes (Rev. 1, April 1974) Comply, except as noted in FSAR section 3.7.4.1. 3.

1.13 Spent Fuel Storage Facility Design Basis (1) Comply, except for Fuel Building Ventilation and Filtration requirements. 9.

(Rev. 1, December 1975) Fuel building ventilation and filtration systems are not credited in the radiological accident analyses.

1.14 Reactor Coolant Pump Flywheel Integrity (Rev. 1, (1) 5.

See Section 1.8N.

August 1975) 1.15 Testing of Reinforcing Bars for Category I Concrete Withdrawn:

Structures 3.

(Rev. 1, December 28, 1972) Withdrawal of this Guide is not intended to alter any prior or existing licensing commitments based on its use.

1.16 Reporting of Operating Information - Appendix A Comply. Monthly Operating Reports superseded by Generic Letter 97-02 and 16 Technical Specifications (Rev. 4, August 1975) License Amendment No. 223.

1.17 Protection of Nuclear Power Plants Against Comply, with the following clarification: 13 Industrial Sabotage (Rev. 1, June 1973) The 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.

1.18 Structural Acceptance Test for Concrete Primary Withdrawn: 3.

Reactor Containments (Rev. 1, December 28, 1972) Withdrawal of this Guide is not intended to alter any prior or existing licensing commitments based on its use.

F R.G. S No. Title Degree of Compliance Ref 1.19 Nondestructive Examination of Primary Withdrawn:

Containment Liner Welds 3.

(Rev. 1, August 11, 1972) Withdrawal of this Guide is not intended to alter any prior or existing licensing 3.

commitments based on its use. A position statement follows.

1. Position C.1.c Leak 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.d Where 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.

1.20 Comprehensive Vibration Assessment Program for (1) 3.

See Section 1.8N.

Reactor Internals During Preoperational and Initial Startup Testing (Rev. 2, May 1976)

F R.G. S No. Title Degree of Compliance Ref 1.21 Measuring, Evaluating, and Reporting Radioactivity Degree of compliance to RG 1.21 is defined for the various elements contained 11 in Solid Wastes and Releases of Radioactive within the guidance in the Radiological Effluent Monitoring and Off site Dose 11 Materials in Liquid and Gaseous Effluents from Calculation Manual (REMODCM). Exception is being taken to RG 1.21 12 Light-Water-Cooled Nuclear Power Plants (Rev. 1, requirement to estimate population doses as well as individual doses. The June 1974) 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.

Online monitors for all potentially 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.

1.22 Periodic Testing of Protection System Actuation (1) Comply, with the following exception: 7.

Functions (Rev. 0, February 17, 1972) RG 1.22 requires that where the ability of a system to respond to an accident 7.

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.

1.23 On site Meteorological Programs (Rev. 0, February Comply 2.

17, 1972)

F R.G. S No. Title Degree of Compliance Ref 1.24 Assumptions Used for Evaluating the Potential Not applicable.

Radiological Consequences of a Pressurized Water No tanks are used for the storage of radioactive gases. However, an analysis of Reactor Radioactive Gas Storage Tank Failure (Rev. the most severe potential radioactive gaseous release is presented in FSAR 0, March 23, 1972) Chapter 15.

1.25 Assumptions Used for Evaluating the Potential Regulatory Guide 1.25 is not used for fuel handling accident analyses. Current 15 Radiological Consequences of a Fuel Handling analyses are based on Regulatory Guide 1.183.

Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Rev. 0, March 23, 1972) 1.26 Quality Group Classifications and Standards for (1) 3.

Comply, with the following exceptions:

Water-, Steam-, and Radioactive-Waste Containing Components of Nuclear Power Plants (Rev. 3, February 1976)

1. 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.

F R.G. S No. Title Degree of Compliance Ref

4. 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 seismically qualified.

1.27 Ultimate Heat Sink for Nuclear Power Plants Comply 2.

(Rev. 2, January 1976) 9.

1.28 Quality Assurance Program Requirements (Design 1. (1) Prior to 12/14/05, Millstone Unit 3 complied as follows:

and Construction) Construction - Millstone 3 complied with R.G. 1.28, Rev. 0.

(Rev. 2, February 1979) (3) Operation - Millstone 3 complies with R.G. 1.28, Rev. 2.

2. Current compliance is as described in the QAPD Topical Report. 17 1.29 Seismic Design Classification (Rev. 3, September (1) Comply 3.

1978) 1.30 Quality Assurance Requirements for the Installation, No longer comply - 12/14/05 (QA standards are described in QAPD Topical 17 Inspection, and Testing of Instrumentation and Report.)

Electric Equipment (Rev. 0, August 11, 1972)

F R.G. S No. Title Degree of Compliance Ref 1.31 Control of Ferrite Content in Stainless Steel Weld (1) 4.

Comply, with the following clarification:

Metal The control of ferrite content in stainless steel weld metal meets the 4.

(Rev. 2, May 1977) (2) requirements of R.G. 1.31, Rev. 1, dated June 1973, or Rev. 3, dated May 1977, 5.

on a case-by-case basis.

Rev. 1 requirements are met by our method of qualifying welding procedures 5.

and controlling heat input during welding. This is subsequently verified by 6.

magnetic measurements of production welds. 10 Rev. 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-deposited, 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.

1.32 Criteria for Safety-Related Electric Power Systems Comply, with the following exception: Battery Discharge Test must be 8.

for Nuclear Power Plants performed at the specified testing frequency as described in the Surveillance (Rev. 2, February 1977) (3) Frequency Control Program.

8.

8.

F R.G. S No. Title Degree of Compliance Ref 1.33 Quality Assurance Program Requirements Compliance is as described in the QAPD Topical Report. 13 (Operation) 13 (Rev. 2, February 1978) 17 1.34 Control of Electroslag Weld Properties Not applicable.

(Rev. 0, December 28, 1972) There electroslag welding process for fabricating components was not used.

1.35 Inservice Inspection of Ungrouted Tendons in Not applicable.

Prestressed Concrete Containment Structures There is no prestressed concrete containment structure.

(Rev. 2, January 1976) 1.36 Nonmetallic Thermal Insulation for Austenitic 1. (1) Prior to 12/14/05, Millstone 3 complied, with the following clarifications 5.

Stainless Steel and exceptions:

(Rev. 0, February 23, 1973) Position C.1 states the packaging and shipping requirements of this guide. 6.

In lieu of controlled packaging and shipping, receipt inspection and tests are required, by specification.

This 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 production test outlined in the Guide.

2. Current compliance is as described in the QAPD Topical Report. 17 1.37 Quality Assurance Requirements for Cleaning of (1) 4.

1. Prior to 12/14/05, Millstone 3 complied.

Fluid Systems and Associated Components of 2. Current compliance is as described in the QAPD Topical Report. 17 Water-Cooled Nuclear Power Plants (Rev. 0, March 16, 1973) 1.38 Quality Assurance Requirements for Packaging, 1. (1) Prior to 12/14/05, Millstone 3 complied. 17 Shipping, Receiving, Storage, and Handling of Items 2. Current compliance is as described in the QAPD Topical Report.

for Water-Cooled Nuclear Power Plants (Rev. 2, May 1977)

F R.G. S No. Title Degree of Compliance Ref 1.39 Housekeeping Requirements for Water-Cooled 1. Prior to 12/14/05, Millstone 3 complied as follows: 12 Nuclear Power Plants (Rev. 2, September 1977) 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 1.40 Qualification Tests of Continuous Duty Motors (1) 3.

Comply Installed Inside the Containment of Water-Cooled 8.

Nuclear Power Plants (Rev. 0, March 16, 1973) 1.41 Preoperational Testing of Redundant On site Electric Comply 8.

Power Systems to Verify Proper Load Group 8.

Assignments (Rev. 0, March 16, 1973) 1.42 Interim Licensing Policy As Low As Practicable for Withdrawn Gaseous Radio Iodine Releases from Light-Water-Cooled Nuclear Power Plants 1.43 Control of Stainless Steel Weld Cladding of Low- (1) Comply 5.

Alloy Steel Components (Rev. 0, May 1973) 1.44 Control of the Use of Sensitized Stainless Steel (1) 4.

Comply, with the following clarification:

(Rev. 0, May 1973) The intent of Paragraph C.6 is complied with in varying degrees as follows: 4.

5.

1. Field fabrication and erection of ASME III piping complies with Paragraph 5.

C.6, except that the ASTM A708-74 standard is used to perform the 6.

intergranular corrosion testing. The radius of the bend specimen is as 10 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.

F R.G. S No. Title Degree of Compliance Ref R.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.

Paragraph 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 required, the above controls assure that base material will not be severely sensitized during welding and meets the intent of Paragraph C.6.

F R.G. S No. Title Degree of Compliance Ref

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 specific control reduces the possibility of a severely sensitized heat-affected zone during 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 unsensitized 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 components of the RCPB piping system are castings and, because of their delta ferrite content, are highly resistant to stress corrosion cracking.

1.45 Reactor Coolant Pressure Boundary Leakage Comply with the following interpretation given to Regulatory Position C.5: 5.

Detection Systems (Rev. 0, May 1973)

The 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 1 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 /> Pumped Capacity Monitoring System Containment Atmosphere Humidity, temperature or pressure Humidity, Temperature and monitoring of the containment Pressure atmosphere are considered as alarms or indirect indication of leakage to the containment

F R.G. S No. Title Degree of Compliance Ref Containment Atmosphere Gaseous 1 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 and Particulate Radioactivity the equilibrium activity of the reactor Monitoring 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 rate Comply with the following interpretation given to Regulatory Position C.7 5.

Indicators 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.

1.46 Protection Against Pipe Whip Inside Containment (1) 3.

Comply with the following clarifications and exceptions:

(Rev. 0, May 1973) Paragraph C.1.b At intermediate locations between terminal ends selected by either of the following criteria:

1. At each pipe fitting, welded attachment, and valve; or

2. 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 Sm in lieu of

F R.G. S No. Title Degree of Compliance Ref Any intermediate locations...conditions exceed 2.0 Sm for ferritic steel and 2.4 Sm 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; or

2. 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 Sh

+ S a) in lieu of Any intermediate locations...conditions exceed 0.8 (Sh + Sa).

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).

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 coefficient as determined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account as applicable, in the reduction of jet discharge.

F R.G. S No. Title Degree of Compliance Ref 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.

Certain 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 modifications 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.

F R.G. S No. Title Degree of Compliance Ref 1.47 Bypassed and Inoperable Status Indication for Comply, with the following clarifications: 7.

Nuclear Power Plant Safety Systems (Rev. 0, May 1. An indicator of bypass/inoperability, located in the control room, will be 1973) 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.

1.48 Design Limits and Loading Combinations for (1) Comply, with the following exception: 3.

Seismic Category I Fluid System Components (Rev. Class 2 and 3 piping analysis to follow criteria specified in 1971 ASME III 0, May 1973) 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.

1.49 Power Levels of Nuclear Power Plants (Rev. 1, Comply, with the following exception: 1.

December 1973) The analyses and evaluations for those conditions described in Position C.2 15 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.

F R.G. S No. Title Degree of Compliance Ref The 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 power 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.

1.50 Control of Preheat Temperature for Welding of Low- (1) 5.

Comply, with the following exception:

Alloy Steel (Rev. 0, May 1973)

In 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.

1.51 Inservice Inspection of ASME Code Class 2 and 3 Withdrawn Nuclear Power Plant Components 1.52 Design, Testing, and Maintenance Criteria for Comply, with the following clarifications and exceptions: 6.

Engineered-Safety-Feature Atmosphere Cleanup Paragraph C.2.g System Air Filtration and Adsorption Units of Light- The filter trains are not instrumented to monitor, alarm and record flow rates in Water-Cooled Nuclear Power Plants the control room. Flow through the filters are verified on a monthly basis. The (Rev. 2, March 1978) 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:

F R.G. S No. Title Degree of Compliance Ref

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 conditioners, 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)

Fuel 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)

F R.G. S No. Title Degree of Compliance Ref 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 interconnecting 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.

F R.G. S No. Title Degree of Compliance Ref

2. 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 N509-1980 requirements stated in Paragraph 4.6.3, which calls for the use of design static pressure (working pressure) in testing for duct leakage.

F R.G. S No. Title Degree of Compliance Ref 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:

• 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

F R.G. S No. Title Degree of Compliance Ref 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 demonstrated 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

F R.G. S No. Title Degree of Compliance Ref The 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.

F R.G. S No. Title Degree of Compliance Ref Documentation 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.

F R.G. S No. Title Degree of Compliance Ref This value is in agreement with ANSI N509-1980.

Paragraph 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: NR All Class II dampers on ESF atmosphere cleanup system air filtration and Qu adsorption units have been tested for leakage rates except for two backdraft 46 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.

F R.G. S No. Title Degree of Compliance Ref 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:

System P Across HEPA & P Across Charcoal Housing (in. wg) (in. wg)

Supplementary Leak Collection & 6.25 7.75 Release Control Room Emergency Air 6.75 9.86 Filtration Aux. Bldg. Filter 6.80 9.80 Paragraph 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.

F R.G. S No. Title Degree of Compliance Ref Exception 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 across 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

F R.G. S No. Title Degree of Compliance Ref The Auxiliary Building Filtration System activated 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.

1.53 Application of the Single-Failure Criterion to (1) 3.

Comply, with the following clarifications:

Nuclear Power Plant Protection Systems (Rev. 0, 1. Regulatory Position C.1 June 1973)

Due 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.2

F R.G. S No. Title Degree of Compliance Ref The 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 necessary to achieve protective functions is demonstrated via the testing program described above.

3. Regulatory Position C.3 Single 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.

F R.G. S No. Title Degree of Compliance Ref 1.54 Quality Assurance Requirements for Protective 1. (1) Prior to 12/14/05, Millstone 3 complied, with the following clarification 6.

Coatings Applied to Water-Cooled Nuclear Power and exception:

Plants (Rev. 0, June 1973) 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 1.55 Concrete Placement in Category I Structures Withdrawn (Rev. 0, June 1973) 3.

Withdrawal 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.

F R.G. S No. Title Degree of Compliance Ref The 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 retain 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.

1.56 Maintenance of Water Purity in Boiling Water Not applicable. Only applicable to BWRs.

Reactors (Rev. 1, July 1978) 1.57 Design Limits and Loading Combinations for Metal Not applicable.

Primary Reactor Containment System Components Only applicable to those plants with a metal primary reactor containment.

(Rev. 0, June 1973)

F R.G. S No. Title Degree of Compliance Ref 1.58 Qualification of Nuclear Power Plant Inspection, 1. (1) Prior to 12/14/05, Millstone 3 complied. 17 Examination, and Testing Personnel (Rev. 1, 2. Current compliance is as described in the QAPD Topical Report.

September 1980) 1.59 Design Basis Floods for Nuclear Power Plants Comply with the following clarification: 2.

(Rev. 2, August 1977) Position C.1 specifies designing hardened protection for all safety related structures, systems, and components. Position C.1 requires hardened 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.

1.60 Design Response Spectra for Seismic Design of (1) Not applicable.

Nuclear Power Plants (Rev. 1, December 1973) 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 consider this Regulatory Guide. PSAR Section 2.6.2.7 specifies the response spectra used for the design of Millstone 3.

F R.G. S No. Title Degree of Compliance Ref 1.61 Damping Values for Seismic Design of Nuclear (1)

Not applicable.

Power The NRC granted that Millstone Nuclear Power Station Unit 3 will not be Plants (Rev. 0, October 1973) 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.

1.62 Manual Initiation of Protective Actions (Rev. 0, (1) Comply, with the following clarification and exceptions: 7.

October 1973)

1. Regulatory Position C.1

a. 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 support 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 two 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.2 All equipment that contributes to the protective action will be initiated at the system level.

F R.G. S No. Title Degree of Compliance Ref There 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 standby conditions, the turbine-driven auxiliary pump feedwater control valves are maintained normally fully open.

F R.G. S No. Title Degree of Compliance Ref During 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.3 Switches 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.4

a. 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.6 Manual initiation portions of the protection system are designed 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.

F R.G. S No. Title Degree of Compliance Ref 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.

This 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.

1.63 Electric Penetration Assemblies in Containment Comply, with the following clarifications: 8.

Structures for Light-Water-Cooled Nuclear Power 1. The single failure provision of R.G. 1.63 shall apply to both Class 1E and Plants (Rev. 2, July 1978) 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, which ensure a highly reliable design will be used for guidance in the protection system design.

F R.G. S No. Title Degree of Compliance Ref 1.64 Quality Assurance Requirements for the Design of (1)

1. Prior to 12/14/05, Millstone 3 complied as follows:

Nuclear Power Plants (Rev. 2, June 1976) (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 1.65 Materials and Inspections for Reactor Vessel Closure (1) 5.

See Section 1.8N.

Studs (Rev. 0, October 1973) 1.66 Nondestructive Examination of Tubular Products Withdrawn 1.67 Installation of Overpressure Protection Devices (1) Comply 5.

(Rev. 0, October 1973) 1.68 Initial Test Programs for Water-Cooled Nuclear Millstone 3 initial startup test program complies with R.G. 1.68 with 14 Power Plant exceptions as stated in FSAR Section 14.2.7.

(Rev. 2, August 1978) 1.68.1 Preoperational and Initial Startup Testing of Not applicable.

Feedwater and Condensate Systems for Boiling Applicable only to BWRs.

Water Reactor Power Plants (Rev. 1, January 1977) 1.68.2 Initial Startup Test Program to Demonstrate Remote Comply 14 Shutdown Capability for Water-Cooled Nuclear Power Plants (Rev. 1, July 1978) 1.68.3 Preoperational Testing of Instrument and Control Air Millstone 3 initial startup test program complies with R.G. 1.68.3 with 9.

Systems (Rev. 0, April 1982) exceptions and clarifications as stated in FSAR Section 14.2.7.

14 1.69 Concrete Radiation Shields for Nuclear Power Plants Comply 12 (Rev. 0, December 1973)

F R.G. S No. Title Degree of Compliance Ref 1.70 Standard Format and Content of Safety Analysis Comply 1.

Reports for Nuclear Power Plants (Rev. 3, November 1978) 1.71 Welder Qualification for Areas of Limited (1) Comply, with the following clarification: 5.

Accessibility An acceptable alternative to this position is contained in the following 5.

(Rev. 0, December 1973) exception: In lieu of Paragraphs C.1 and C.2a, all applicable welds of 10 limited accessibility are volumetrically inspected to the requirements and standards of ASME III, Class 1.

1.72 Spray Pond Piping Made from Fiberglass- Not applicable.

Reinforced Thermosetting Resin (Rev. 2, November Fiberglass pipe is not used for QA Category I applications on Millstone 3.

1978) 1.73 Qualification Tests of Electric Valve Operators (1) Comply 3.

Installed Inside the Containment of Nuclear Power 8.

Plants (Rev. 0, January 1974) 1.74 Quality Assurance Terms and Definitions (Rev. 0, (1) 17 Comply February 1974) 17 1.75 Physical Independence of Electric Systems (1) 7.

Comply, with the following exceptions and clarifications:

(Rev. 2, September 1978)

1. General (Clarification 8.

Ventilated tray covers are considered equivalent to solid tray covers.

F R.G. S No. Title Degree of Compliance Ref 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.

3. 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.

F R.G. S No. Title Degree of Compliance Ref 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 conduit 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.

Coordination 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)

F R.G. S No. Title Degree of Compliance Ref 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.

5. Position C.7 (Section 4.6 of IEEE 384)

F R.G. S No. Title Degree of Compliance Ref 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 instrument 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.

F R.G. S No. Title Degree of Compliance Ref

2. Power cables (from the above 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 instrument 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.

F R.G. S No. Title Degree of Compliance Ref A minimum 1 inch separation (or a barrier) is required between redundant wire bundles or Class 1E and non-Class 1E wire bundles.

Test 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.

1.76 Design Basis Tornado for Nuclear Power Plants Comply 3.

(Rev. 0, April 1974) 1.77 Assumptions Used for Evaluating a Control Rod (1) Comply with the following exception. Regulatory Position C.3 no longer 15 Ejection Accident for Pressurized Water Reactors applies and has been replaced by R.G. 1.183.

(Rev. 0, May 1974) 15 1.78 Assumptions for Evaluating the Habitability of a Comply, with the following clarifications: 6.

Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release (Rev. 0, June 1974)

The 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 exceeds 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:

F R.G. S No. Title Degree of Compliance Ref Manual control room ventilation isolation, coating of concrete 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.

Sealing of all pipes, ducts, and electrical penetrations into the control room envelope.

Compression seals for access doors and equipment removal hatches 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 Millstone 3 initial startup test program complies with R.G. 1.79 Rev. 1 with 14 Systems for Pressurized Water Reactors (Rev. 1, exceptions as stated in FSAR Section 14.2.7.

September 1975) 1.80 Preoperational Testing of Instrument Air Systems Withdrawn 1.81 Shared Emergency and Shutdown Electric Systems Not applicable. 8.

for Multi-Unit Nuclear Power Plants (Rev. 1, Applicable only to those plants which share emergency and shutdown January 1975) (3) electrical systems in a multi-unit plant, which Millstone 3 does not.

1.82 Sumps for Emergency Core Cooling and (1) 6.

Comply, with the following clarifications:

Containment Spray Systems (Rev. 0, June 1974)

F R.G. S No. Title Degree of Compliance Ref C.1 The recirculation spray pumps take suction from a single sump. The 6.

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 exempt from passive failure. (See Section 3.1.1.3)

C.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.

1.83 Inservice Inspection of Pressurized Water Reactor (1) 5.

Comply Steam Generator Tubes (Rev. 1, July 1975) 16 4.

1.84 Design and Fabrication Code Case Acceptability - (1) 5.

Comply ASME Section III, Division 1 1.85 Materials Code Case Acceptability - ASME (1) Comply 5.

Section III, Division 1 1.86 Termination of Operating Licenses for Nuclear Comply 5.

Reactors (E (Rev. 0, June 1974) 1.87 Guidance for Construction of Class 1 Components in Not applicable.

Elevated-Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1593, 1594, 1595, and 1596) (Rev. 1, June 1975)

Applicable only to Elevated Temperature Reactors

F R.G. S No. Title Degree of Compliance Ref 1.88 Collection, Storage, and Maintenance of Nuclear (1)

1. Prior to 12/14/05, Millstone 3 complied as follows:

Power Plant Quality Assurance Records (Rev. 2, Construction October 1976) (3) 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 1.89 Qualification of Class 1E Equipment for Nuclear (1) Class 1E equipment other than that within the NSSS Scope of Supply 3.

Power Plants (Rev. 0, November 1974) complies with R.G. 1.89, dated November 1974, with the following clarifications:

1. 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 equipment 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.

F R.G. S No. Title Degree of Compliance Ref The clarification as stated above is similar to the position taken by the IEEE.

1.90 Inservice Inspection of Prestressed Concrete Not applicable.

Containment Structures with Grouted Tendons (Rev. Millstone 3 does not have a prestressed concrete containment structure.

1, August 1977) 1.91 Evaluation of Explosions Postulated to Occur on Comply 2.

Transportation Routes Near Nuclear Power Plants (Rev. 1, February 1978) 1.92 Combination of Modes and Spatial Components in (1) Comply, with the following clarification: 3.

Seismic Response Analysis (Rev. 0, December 1974) (2)

The 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.

1.93 Availability of Electric Power Sources (Rev. 0, 8.

December 1974) 1. One emergency diesel generator may be inoperable for up to 14 days. 8.

2. Preventive, as well as corrective, maintenance 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).

F R.G. S No. Title Degree of Compliance Ref 1.94 Quality Assurance Requirements for Installation, 1. Prior to 12/14/05, Millstone 3 complied, with the following exceptions:

Inspection and Testing of Structural Concrete and a. Millstone 3 will comply with the requirements of ANSI N45.2.5-1974 3.

Structural Steel During the Construction Phase of except that correlation testing shall be performed in accordance with the Nuclear Power Plants applicable paragraphs of Section 6.11 of N45.2.5-1978.

(Rev. 1, April 1976) (3) b. 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 17 Report.)

1.95 Protection of Nuclear Power Plant Control Room Comply, with the following clarifications and exceptions: 6.

Operators Against an Accidental Chlorine Release In accordance with R.G. 1.78, R.G. 1.95, and Section 2.2, no off site chlorine (Rev. 1, January 1977) 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.

1.96 Design of Main Steam Isolation Valve Leakage Not applicable.

Control Systems for Boiling Water Reactor Nuclear Applicable only to BWRs.

Power Plants (Rev. 1, June 1976) 1.97 Instrumentation for Light-Water-Cooled Nuclear (1) Millstone 3 compliance with Regulatory Guide 1.97, Rev. 2, is found in 7.

Power Plants to Assess Plant Conditions During and specification SP-M3-IC-022 titled, Millstone 3 Design Basis to Respond to Following an Accident Regulatory Guide 1.97, Rev. 2.

(Rev. 2, December 1980) 7.

1.98 Assumptions Used for Evaluating the Potential Not applicable.

Radiological Consequences of a Radioactive Offgas Applicable only to BWRs.

System Failure in a Boiling Water Reactor (Rev. 0, March 1976)

F R.G. S No. Title Degree of Compliance Ref 1.99 Effects of Residual Elements on Predicted Radiation (1) 5.

See Section 1.8N.

Damage to Reactor Vessel Materials (Rev. 1, April 1977) 1.100 Seismic Qualification of Electric Equipment for (1) Replacement items meet the original criteria of IEEE-344-87 (Endorsed by Nuclear Power Plants (Rev. 1, August 1977) Regulatory Guide 1.100, Rev. 2.)

1.101 Emergency Planning for Nuclear Power Plants Comply 13 (Rev. 2, October 1981) 1.102 Flood Protection for Nuclear Power Plants Comply with the following clarification: 3.

(Rev. 1, September 1976)

Position 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.

1.103 Post-Tensioned Prestressing Systems for Concrete Withdrawn Reactor Vessels and Containments (Rev. 1, October 1976)

F R.G. S No. Title Degree of Compliance Ref 1.104 Overhead Crane Handling Systems for Nuclear Withdrawn Power Plants The 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.

1.105 Instrument Setpoints (Rev. 1, November 1976) (1) 7.

Comply 7.

7.

1.106 Thermal Overload Protection for Electric Motor on Comply 7.

Motor-Operated Valves (Rev. 1, March 1977) (3) 8.

1.107 Qualifications for Cement Grouting for Prestressing Not applicable.

Tendons in Containment Structures (Rev. 1, February 1977) 1.108 Periodic Testing of Diesel Generator Units Used as Comply, with the following clarifications and exceptions: 8.

On site Electric Power Plants (Rev. 1, August 1977)

Section 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.

If 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.

F R.G. S No. Title Degree of Compliance Ref 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.

1.109 Calculation of Annual Doses to Man from Routine Comply 13 Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I (Rev. 1, October 1977) 1.110 Cost-Benefit Analysis for Radwaste Systems for Comply Light-Water-Cooled Nuclear Power Reactors (Rev.

0, March 1976) 1.111 Methods for Estimating Atmospheric Transport and Comply 2.

Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors (Rev. 1, July 1977) 1.112 Calculation of Releases of Radioactive Materials in Comply 11 Gaseous and Liquid Effluents from Light-Water- 11 Cooled Power Reactors 15 (Rev. 0, April 1976) 1.113 Estimating Aquatic Dispersion of Effluents from Comply 2.

Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I (Rev. 1, April 1977) 1.114 Guidance on Being Operator at the Controls of a Comply 13 Nuclear Power Plant (Rev. 1, November 1976) 1.115 Protection Against Low-Trajectory Turbine Missiles Comply, with the following clarification: 3.

(Rev. 1, July 1977)

F R.G. S No. Title Degree of Compliance Ref 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.

1.116 Quality Assurance Requirements for Installation, 1. (1) Prior t 12/14/05, Millstone 3 complied. 17 Inspection, and Testing of Mechanical Equipment 2. Current compliance is as described in the QAPD Topical Report.

and Systems (Rev. 0-R, May 1977) 1.117 Tornado Design Classification (Rev. 1, April 1978) Comply, with the following clarification: 3.

1. 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:

Protection 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...

F R.G. S No. Title Degree of Compliance Ref 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.

1.118 Periodic Testing of Electric Power and Protection (1) 7.

Comply Systems 8.

(Rev. 2, June 1978) (3) 1.119 Surveillance Program for New Fuel Assembly Withdrawn Designs 1.120 Fire Protection Guidelines for Nuclear Power Plants Comply, with the following clarification:

(Rev. 1, November 1977) 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.

1.121 Bases for Plugging Degraded PWR Steam Generator (1) See Section 1.8N. 5.

Tubes (Rev. 0, August 1976) 1.122 Development of Floor Design Response Spectra for Not applicable. 3.

Seismic Design of Floor-Supported Equipment or The implementation section of R.G. 1.122 states that the guide will be used in Components the evaluation of construction permit applications. The mathematical methods (Rev. 1, February 1978) 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:

F R.G. S No. Title Degree of Compliance Ref

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.

1.123 Quality Assurance Requirements for Control of (1)

1. Prior to 12/14/05, Millstone 3 complied as follows:

Procurement of Items and Services for Nuclear Construction Power Plants (Rev. 1, July 1977)

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 1.124 Service Limits and Loading Combinations for Class Comply, with the following exceptions and additions: 3.

1 Linear-Type Component Supports (Rev. 1, January 1. The following paragraph should be added to Regulatory Position C.3: 5.

1978)

C.3.c. The bending stress limits Fb 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 Sy or 0.55 Su for compact sections, 0.75 Sy or 0.63 Su for doubly symmetrical members with bending about the minor axis, 0.6 Sy or 0.5 Su for box-type flexural members and miscellaneous members.

The paragraph added to Regulatory 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:

F R.G. S No. Title Degree of Compliance Ref 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 Sy or Su at temperature. The increased allowable permitted for shear stress in bolts shall not exceed 0.42 Su at temperature.

The 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 increased allowables are achieved by limiting bolting tensile stress to the less of 0.7 Sy or Su at temperature and bolting shear stress to 0.42 Su at temperature. The 0.7 Su limit is well recognized in Section III of the Code. The average shear strength of bolting material is about 0.62 Su 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:

F R.G. S No. Title Degree of Compliance Ref 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:

Supports 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.

F R.G. S No. Title Degree of Compliance Ref 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 severe operating conditions.

1.125 Physical Models for Design and Operation of Comply Hydraulic Structures and Systems for Nuclear Power Plants (Rev. 1, October 1978) 1.126 An Acceptable Model and Related Statistical (1) See Section 1.8N. 4.

Methods for the Analysis of Fuel Densification (Rev.

1, March 1978)

F R.G. S No. Title Degree of Compliance Ref 1.127 Inspection of Water-Control Structures Associated Not application. 9.

with Nuclear Power Plants (Rev. 1, March 1978) Applies to water control structures; i.e., dams, 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 any abnormal conditions in these structures.

1.128 Installation Design and Installation of Large Lead Comply 8.

Storage Batteries for Nuclear Power Plant (Rev. 1, 8.

October 1978) (3) 1.129 Maintenance, Testing, and Replacement of Large Comply, except the battery discharge test frequency, which is controlled by the 8.

Lead Storage Batteries for Nuclear Power Plants Surveillance Frequency Control Program - Note: This Reg. Guide endorses 8.

(Rev. 1, February 1978) (3) 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.

1.130 Design Limits and Loading Combinations for Class Comply, with the following clarifications and exceptions:

1 Plate-and-Shell-Type Component Supports (Rev. 1. Regulatory Position C.3 should be revised as follows: Service limits for 1, October 1978) 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.

F R.G. S No. Title Degree of Compliance Ref

2. 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 combinations described in Regulatory 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 main 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 severe operating conditions.

1.131 Qualification Tests of Electric Cables, Field Splices, Comply and Connections for Light-Water-Cooled Nuclear Power Plants (Rev. 0, August 1977) 1.132 Site Investigations for Foundations of Nuclear Comply with the following exceptions and clarifications: 2.

Power Plants (Rev. 1, March 1979)

F R.G. S No. Title Degree of Compliance Ref

1. Paragraph B.5: Surveys of horizontal deviation are made in all boreholes that are used for cross-hold seismic tests.

The 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 and 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 materials, 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.

F R.G. S No. Title Degree of Compliance Ref

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.

The 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 investigations 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.

1.133 Loose-Part Detection Program for the Primary Comply, with the following exceptions:

System of Light-Water-Cooled Reactors (Rev. 1, May 1981)

1. Position C.1.g The 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

F R.G. S No. Title Degree of Compliance Ref 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.c Calibration 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.

1.134 Medical Evaluation of Nuclear Power Plant Comply Personnel Requiring Operator Licenses (Rev. 1, March 1979) 1.135 Normal Water Level and Discharge at Nuclear Comply, with the following clarification:

Power Plants (Rev. 0, September 1977)

R.G. 1.59, Rev. 2, is used to determine initial water levels for design basis flood analysis.

1.136 Material for Concrete Containments Not applicable.

(Rev. 2, June 1981)

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.

F R.G. S No. Title Degree of Compliance Ref Withdrawal of the above Guides does not alter any prior existing licensing commitments based on their use.

1.137 Fuel-Oil Systems for Standby Diesel Generators Comply with the following clarifications and exceptions: 9.

(Rev. 1, October 1979)

The 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.

This 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 following 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.

1.138 Laboratory Investigations of Soils for Engineering Comply, with the following clarifications and exceptions: 2.

Analysis and Design of Nuclear Power Plants (Rev.

0, April 1978)

1. Paragraph C.1.c: Standards used to calibrate laboratory test equipment are of a known higher accuracy than the test equipment, rather than four times more accurate than the working instrument.

F R.G. S No. Title Degree of Compliance Ref Standards 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 replacement 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.

F R.G. S No. Title Degree of Compliance Ref The 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 for 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.

F R.G. S No. Title Degree of Compliance Ref

8. 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.

1.139 Guidance for Residual Heat Removal (Rev. 0, May (1) 3.

Comply 1978)

F R.G. S No. Title Degree of Compliance Ref 1.140 Design, Testing, and Maintenance Normal Comply, with the following clarifications and exceptions: 9.

Ventilation Criteria for Exhaust System Air Paragraph C.2.c Filtration and Adsorption Units of Light-Water- The Technical Support Center filter train is instrumented to indicate, but not to Cooled Nuclear Power Plants (Rev. 1, October 1979) 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.

F R.G. S No. Title Degree of Compliance Ref Paragraph 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 system 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 demonstrated 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.

F R.G. S No. Title Degree of Compliance Ref 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 The 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 inspection 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

F R.G. S No. Title Degree of Compliance Ref 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 & P Across Charcoal Housing (in. wg) (in. wg)

Containment Air Filters 5.94 7.54 Technical Support Center Air 5.25 6.45 Filters Paragraph 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.

F R.G. S No. Title Degree of Compliance Ref 1.141 Containment Isolation Provisions for Fluid Systems (1) 5.

Comply, with the following exception:

(Rev. 0, April 1978) On the containment recirculation system, the only closed system outside 5.

containment, vent/drain valves and branch connections are not lock closed.

6.

1.142 Safety Related Concrete Structures for Nuclear Not applicable. The following Regulatory Guides have been incorporated into Power Plants (Other than Reactor Vessels and R.G. 1.142, Rev. 0, and have been withdrawn:

Containments)

• 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of Category 1 (Rev. 0, April 1978) Concrete Structures.

• 1.15 Testing of Reinforcing Bars for Category 1 Concrete Structures.

• 1.55 Concrete Placement in Category 1 Structures.

• 1.103 Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments.

1.143 Design Guidance for Radioactive Waste 1. Prior to 12/14/05, Millstone 3 complied, with the following clarifications 3.

Management Systems, Structures, and Components and exceptions: 11 Installed in Light-Water-Cooled Nuclear Power Section C.1.1.1 - All demineralizers have been designed, procured, and 11 Plants (Rev. 1, October 1979) manufactured in accordance with the ASME Section VIII Boiler and 11 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.

Section 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.

F R.G. S No. Title Degree of Compliance Ref

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 required, 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.

Section 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 socket-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 Category 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).

F R.G. S No. Title Degree of Compliance Ref

2. Current compliance is as described in the QAPD Topical Report 17 1.144 Auditing of Quality Assurance Programs for Nuclear 1. (1) Prior to 12/14/05, Millstone 3 complied as follows:

Power Plants (Rev. 1, September 1980) Construction Millstone 3 complies with ANSI N45.2.12-1977, as implemented by Appendix VII of the Millstone 3 Quality Assurance Program Manual.

Operation - Comply 2.

2. Current compliance is as described in the QAPD Topical Report. 17 1.145 Atmospheric Dispersion Models for Potential Comply 13 Accident Consequence Assessments at Nuclear Power Plants (Rev. 0, August 1979) 1.146 Qualification of Quality Assurance Program Audit 1. (1) Prior to 12/14/05, Millstone 3 complied. 17 Personnel for Nuclear Power Plants (Rev. 0, August 2. Current compliance is as described in the QAPD Topical Report.

1980) 1.147 Inservice Inspection Code Case Acceptability Comply ASME Section XI Division 1 (Rev. 0, February 1981) 1.148 Functional Specification for Active Valve The majority of active valves used at Millstone 3 were purchased prior to the Assemblies in Systems Important to Safety in issuance of Regulatory Guide 1.148. However, Millstone 3 meets the intent of Nuclear Power Plants Regulatory Guide 1.148 in that all active valves relied upon to perform a safety (Rev. 0, March 1981) 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.

F R.G. S No. Title Degree of Compliance Ref 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.

1.149 Nuclear Power Plant Simulators for Use in Operator Millstone 3 complies with R.G. 1.149, Rev.3. 1 Training (Rev. 3, October 2001) 1.150 Ultrasonic Test of Reactor Vessel Welds During Comply, per response to NRC Question Q250.3. 5.

Preservice and Inservice Examinations (Rev. 1, February 1983) 1.152 Criteria for Digital Computers in Safety Systems of Compliance is as described in the QAPD Topical Report. 1.

Nuclear Power Plants (Rev. 1, January 1996) 1.155 Station Blackout Comply 8.

1.163 Performance-Based Containment Leak-Test Comply Program 1.183 Alternative Radiological Source Terms for Complies with the following exceptions: 6.5 Evaluating Design Basis Accidents at Nuclear Power Section 3.2, Table 3, footnote 11 - Non-LOCA Fraction of Fission Product Reactors (Rev. 0, July 2000) 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 11 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 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-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.

F R.G. S No. Title Degree of Compliance Ref 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.

1.194 Atmospheric Relative Concentrations for Control Comply 2.3 Room Radiological Habitability Assessments at Nuclear Power Plants (Rev. 0, June 2003) 1.196 Control Room Habitability at Light-Water Nuclear Comply, with the following exceptions and clarifications: 6.4 Power Reactors (Rev. 0, May 2003)

1. RG 1.196 calls for evaluating, testing and maintaining the Control Room Emergency Filtration System per RG 1.52 Rev. 3. Millstone 3 complies with RG 1.52 Rev. 2, with the clarifications and exceptions as noted in this table.

2. RG 1.196 calls for performing an evaluation, and also periodic assessments, of the impact of hazardous chemical release on control room habitability using the methodology of RG 1.78 Rev. 1 (which replaced the previous revisions of RG 1.78 for hazardous gas release and RG 1.95 for chlorine release). Millstone 3 complies with RG 1.78 Rev. 0 and RG 1.95 Rev. 1, with the clarifications and exceptions as noted in this table.

3. RG 1.196 calls for determining control room radiological dose per RG 6.4 1.183 Rev. 0 for plants employing alternative source term methodology.

Millstone 3 complies with RG1.183 Rev. 0, with the exceptions as noted in this table.

F R.G. S No. Title Degree of Compliance Ref 1.197 Demonstrating Control Room Envelope Integrity at Comply, with the following exceptions and clarifications:

Nuclear Power Reactors (Rev. 0, May 2003)

1. With respect to Position C.1 for performance of an integrated Control Room Envelope (CRE) inleakage test:

• Appropriate application of ASTM E741 shall include the ability to take minor 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 CRE and developing compensatory actions in the event of excessive CRE inleakage:

• Vulnerability assessments for radiological, hazardous chemical and smoke, and emergency ventilation system testing were completed as documented in the UFSAR and other licensing basis documents. The exceptions to the Regulatory Guides referenced in RG 1.196 (i.e., RG 1.52, RG 1.78 and RG 1.183), which were considered in completing the vulnerability assessments, are documented in the UFSAR current licensing basis.

NOTE:

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 sections of the later revisions of the Guide.

3. Millstone 3 addresses a later revision of the Guide than is required by the implementation section of the Guide.

le 1.8N-1 lists the NRC Division 1 Regulatory Guides that were in effect during the time of lication for an Operating License. It identifies applicable FSAR sections, and indicates the SS scope of compliance for Millstone 3.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.1 Net Positive Suction Head for In predicting the NPSH available, Westinghouse assumes that the vapor pressure of 6.2.2, Emergency Core Cooling and the liquid in the sump is equal to the containment pressure, i.e., that the liquid is at a Containment Heat Removal System saturated condition, in calculating the available NPSH for the recirculation mode.

Pumps This assumption meets the intent of Regulatory Guide 1.1, since no credit is taken for (Rev. 0, November 2, 1970 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.

1.2 Thermal Shock to Reactor Pressure ISSUE: 5.2.3.

Vessels General Design Criterion 35 specifies design and operating conditions necessary to (Rev. 0, November 1970) assure that the reactor coolant pressure boundary will 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.

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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 inherent in reactor pressure vessels. Additional testing will be done to demonstrate the behavior of vessels under combined pressure and thermal shock loadings.

Details of progress and results obtained in the HSST program 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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.

1.7 Control of Combustible Gas Westinghouse will conform to all assumptions established in Regulatory Guide 1.7 6.2.5 Concentrations in Containment and thereby comply with General Design Criterion 41.

Following a Loss-of-Coolant Accident (Rev. 2, November 1978) 1.13 Spent Fuel Storage Facility Design Comply. 9.1.2 Basis (Rev. 1, December 1975) 1.14 Reactor Coolant Pump Flywheel Comply, with the following exceptions: 5.4.1.

Integrity (Rev. 1, August 1975)

1. Post-Spin Inspection

FSA R.G. Sectio No. Title Degree of Compliance Refere 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

FSA R.G. Sectio No. Title Degree of Compliance Refere 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 combined 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 isotropy 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

FSA R.G. Sectio No. Title Degree of Compliance Refere Vacuum Treatment - The requirements for vacuum melting and degassing 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 turbine generator during anticipated 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.

1.20 Comprehensive Vibration Assessment For each prototype reactor internals design, a program of vibration analysis, 3.9.2 Program for Reactor Internals During measurement, and inspection has been developed and reviewed by the NRC. This is Preoperational and Initial Startup documented in WCAP-7879.

Testing The reactor internals similar to the prototype design will be subjected during hot (Rev. 2, May 1976) 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.22 Periodic Testing of Protection System Periodic testing of the actuation equipment and actuated equipment of the reactor trip 7.2 Actuation Functions system and the engineered safety features actuation system is in agreement with the 7.3.2.

(Rev. 0, February 17, 1972) provisions of Regulatory Guide 1.22.

Where the ability of a system to respond to a bona fide accident 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 plant

2. 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; and

3. 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 switches

2. Turbine

3. 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)

FSA R.G. Sectio No. Title Degree of Compliance Refere

9. Charging and letdown isolation valves (close) 10.Spray header isolation valves (open) 11.CVCS suction valves - Normal (close) 12.Instrument air to containment isolation valves (close) 13.Chillwater supply and return containment isolation valves (close) 1.26 Quality Group Classifications and The definitions of safety classes found in ANSI N18.2a are utilized by Westinghouse. 3.2.2 Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants (Rev. 3, February 1976) 1.28 Quality Assurance Program The Westinghouse Quality Assurance Plan is presented in WCAP-8370, Rev. 9A, Requirements (Design and Westinghouse Quality Assurance Program and WCAP-7800, Nuclear Fuel Construction) Division Quality Assurance Program Plan.

(Rev. 2, February 1979) 1.29 Seismic Design Classification Westinghouse classifies each component important to safety as Safety Class 1, 2, or 3 3.2.1 (Rev. 3, September 1978) 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:

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, and

4. Routine post seismic procedures would disclose loss of the safety function.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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.

1.31 Control of Ferrite Content in Stainless The welding of austenitic stainless steel is controlled to mitigate the occurrence of 4.5.1 Steel Welding microfissuring or hot cracking in the weld. Although published data and experience (Rev. 3, April 1978) 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 4.5.2.

documented in the technical literature that the presence of delta ferrite is one of the 5.2.3.

mechanisms for reducing the susceptibility of stainless steel welds to hot cracking.

5.3.1.

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 10.3.6 lies somewhere between 0 and 3 percent delta ferrite.

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 components.

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere The fabrication and installation specifications require welding procedure and welder qualification in accordance with Section III, and include the 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 information required by Section III.

The 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 analysis 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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, including 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 nondestructive examination methods according to Section III rules.

To assure the reliability of these controls, Westinghouse has performed 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 Regulatory 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.

Welds 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 nondestructive acceptance standards.

1.36 Nonmetallic Thermal Insulation for The Westinghouse practice follows the recommendations of Regulatory Guide 1.36 5.3.2 Austenitic Stainless Steel (Rev. 0, but is more stringent in several respects as discussed below.

February 1973)

FSA R.G. Sectio No. Title Degree of Compliance Refere The nonmetallic thermal insulation used on the reactor coolant pressure boundary is 6.1.1 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.

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.

1.37 Quality Assurance Requirements for The Westinghouse position for all Water Reactors Division (WRD) on this 4.5.1 Cleaning of Fluid Systems and Regulatory Guide is documented in WCAP-8370.

Associated Components of Water- 17.1.2 Cooled Nuclear Power Plants (Rev. 0, 17.1.3 March 1973) 17.2

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.38 Quality Assurance Requirements for The Westinghouse position for the WRD NSSS scope of supply on Regulatory Packaging, Shipping, Receiving, Guide 1.38 is presented in WCAP-8370, WRD Quality Assurance Plan. The Storage, and Handling of Items for Nuclear Fuel Division (NFD) position on this Regulatory Guide is presented in Water-Cooled Nuclear Power Plants WCAP-7800, NFD Quality Assurance Program Plan.

(Rev. 2, May 1977) 1.40 Qualification Tests of Continuous-Duty There are no Class 1 continuous duty motors installed inside containment in the Motors Installed Inside the Containment NSSS scope of supply.

of Water-Cooled Nuclear Power Plants (Rev. 0, March 1973) 1.43 Control of Stainless Steel Weld Westinghouse performs a qualification test on any high heat input welding process 5.3.1.

Cladding of Low-Alloy Steel (such as the submerged arc wide-strip welding process or the submerged arc 6-wire Components process) used to clad coarse or fine grained SA-508 Class 2 material. This test (Rev. 0, May 1973) 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.

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.

1.44 Control of the Use of Sensitized It has been and continues to be Westinghouse practice to use processing, 4.5.1 Stainless Steel preoperational cleaning, packaging, and shipping controls to preclude adverse effects 4.5.2.

(Rev. 0, May 1973) of exposure to contaminants on austenitic stainless steel materials. Furthermore, 5.2.3.

Westinghouse strongly 6.1.1.1 recommends rigorous control of reactor coolant 5.3.1.

system water chemistry to prevent the intrusion of aggressive species.

6.1.1.

10.3.6 Austenitic 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:

FSA R.G. Sectio No. Title Degree of Compliance Refere

1. Solution annealed or water quenched, or

2. 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, and

2. 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere The 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.

1.46 Protection Against Pipe Whip Inside The criteria implemented in the evaluation of the main reactor coolant loop is based 3.6 Containment on draft ANS Standard 20.2, Design Basis for Protection Against Pipe Whip, and (Rev. 0, May 1973) is documented in WCAP-8172-A, Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop.

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.48 Design Limits and Loading Westinghouse meets and will continue to meet the requirements of General Design 3.9.3 Combination for Seismic Category 1 Criterion 2 and will thereby meet the intent of Regulatory Guide 1.48. The analytical Fluid System Components (Rev. 0, May and experimental procedures which Westinghouse will use to demonstrate structural 1973) 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.

Structural Integrity To ensure the structural integrity of fluid 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.

1.50 Control of Preheat Temperature for Westinghouse considers that this Guide applies to ASME Section III, Class 1 5.2.3.

Welding of Low-Alloy Steel (Rev. 0, components.

May 1973)

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

FSA R.G. Sectio No. Title Degree of Compliance Refere The Westinghouse position is that this guide requirement 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.

1.53 Application of the Single Failure Performance of a Single Failure Analysis - The principles described in the IEEE 3.1.1 Criterion to Nuclear Power Plant Standard were used in the Power Plant Protection Systems design of the protection Protection Systems systems on Westinghouse plants. For documentation of applicable analysis, refer to (Rev. 0, June 1973) 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:

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, and

4. 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere Scope of Analysis - The regulatory guide requires that the single failure analysis extend beyond the scope of IEEE-Standard 279-1971 and include actuation 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.

1.54 Quality Assurance Requirements for The Westinghouse NSSS equipment located in the for containment building is 6.1.2 Protective Coatings Applied to Water separated into four categories to identify the applicability of this Regulatory Guide to Cooled Nuclear Power Plants (Rev. 0, various types of equipment. These categories of equipment are as follows:

June 1973) 17.1 Category 1 - Large equipment Category 2 - Intermediate equipment Category 3 - Small equipment Category 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 Tanks

4. Manipulator Crane The 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:

FSA R.G. Sectio No. Title Degree of Compliance Refere

1. Preparation of vendor procedures

2. Use of specific coatings systems which are qualified to ANSI N101.2

3. Surface preparation

4. Application of the coating systems in accordance with the paint manufacturer's instructions

5. Inspections and nondestructive examinations

6. Exclusion of certain materials

7. Identification of all nonconformances

8. Certifications of compliance The 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 rig

3. Head lifting rig

4. Electrical cabinets The 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:

FSA R.G. Sectio No. Title Degree of Compliance Refere

1. Use of specific coatings systems which are qualified to ANSI N101.2

2. Surface preparation

3. Application of the coating systems in accordance with the paint manufacturer's instructions The 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 protective 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. Transmitters

2. Alarm horns

3. Small instruments

4. Valves

5. Heat exchanger supports These 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere Category 4 - Insulated or Stainless Steel Equipment Category 4 equipment consists of the following:

1. Steam generators - covered with wrapped insulation

2. Pressurizer - covered with wrapped insulation

3. Reactor pressure vessel - covered with rigid reflective insulation

4. Reactor coolant piping - stainless steel

5. Reactor coolant pump casings - stainless steel The 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.

1.58 Qualification of Nuclear Power Plant The Westinghouse portion for the WRD NSSS scope of supply on Regulatory Guide 17.1.2 Inspection, Examination, and Testing 1.58 is presented in WCAP-8370, WRD Quality Assurance Plan. The Nuclear Fuel 17.1.3 Personnel Division position on this Regulatory Guide is presented in WCAP-7800, NFD 17.2 (Rev. 1, September 1980) Quality Assurance Program Plan.

1.60 Design Response Spectra for Seismic The design response spectra of Regulatory Guide 1.60, Revision 1, are acceptable to 3.7N Design of Nuclear Power Plants Westinghouse with the following exception:

(Rev. 1, December 1973)

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.

1.61 Damping Values for Seismic Design of The damping values listed in Regulatory Guide 1.61 Design of Nuclear Power are 3.7N Nuclear Power Plants (Rev. 0, October acceptable to Westinghouse for plants using 3D seismic analysis. However, one 1973) 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.62 Manual Initiation of Protective Actions There are four individual main stop valve momentary control switches (one per loop) 7.3.2.

(Rev. 0, October 1973) 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.

Manual 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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.

1.64 Quality Assurance Requirements for the The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 17.1.2 Design of Nuclear Power Plants 1.64 is presented in WCAP-17.1.3 8370 WRD Quality Assurance Plan. The 17.1.3 (Rev. 2, June 1976) Nuclear Fuel Division position on this Regulatory Guide is presented in 17.2 WCAP-7800, NFD Quality Assurance Program Plan.

1.65 Materials and Inspections for Reactor Westinghouse is in agreement with Regulatory Guide 1.65 with the following 5.3.1 Vessel Closure Studs (Rev. 0, October exception for the material and tensile strength guidelines:

1973)

FSA R.G. Sectio No. Title Degree of Compliance Refere

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere The 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).

1.67 Installation of Overpressure Protection The scope of Regulatory Guide 1.67 is limited to the design of open discharge 5.4.11 Devices systems. Installation of overpressure protection devices is in the balance of plant (Rev. 0, October 1973) (BOP) and therefore this guide is not in the Westinghouse NSSS scope of supply.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.71 Welder Qualification for Areas of Westinghouse practice does not require qualification or requalification of welders for 5.2.3 Limited Accessibility areas of limited accessibility as described by the Guide and has provided welds of 5.3.1.

(Rev. 0, December 1973) high quality. 10.3.6 Westinghouse believes that limited accessibility qualification 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 similar 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.

1.73 Qualification Tests of Electric Valve The qualification programs for Westinghouse WRD supplied Class IE electric motor 3.11N Operators Installed Inside the operators, solenoid valves, and limit switches described in WCAP-8587 and Containment of Nuclear Power Plants WCAP-9688 meet the requirements of Regulatory Guide 1.73.

(Rev. 0, January 1974) 8.3.1 1.74 Quality Assurance Terms and The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 17.1.2 Definitions 1.74 is presented in WCAP-8370, WRD Quality Assurance Plant. The Nuclear (Rev. 0, February 1974) Fuel Division position on this Regulatory Guide is presented in WCAP-7800, NFD Quality Assurance Program Plan. 17.2 1.75 Physical Independence of Electric Westinghouse takes exception to the Regulatory Guide 1.75 in several areas as 7.1 Systems discussed below. These issues have been presented to the Regulatory Staff and are (Rev. 2, September 1978) not resolved at this time.

8.3.1.

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere

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)

Regulatory 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.

1.77 Assumption Used for Evaluating a The result of the Westinghouse analysis shows compliance with the Regulatory 15.4.2 Control Rod Ejection Accident for Position given in Section C.1 of Regulatory Guide 1.77. In addition, Westinghouse 15.4.7 Pressurized Water Reactors (Rev. 0, complies with the intent of the assumptions given in Appendix A of the Regulatory May 1974) Guide.

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.

1.82 Sumps for Emergency Core Cooling and The Robust Fuel Assembly (RFA) implemented in Cycle 7 (Region 9) includes the 6.2.2.

Containment Spray Systems debris resistant bottom nozzle (DRBN) and the protective bottom grid (P-Grid) fuel (Rev. 0, June 1974) 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).

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.83 Inservice Inspection of Pressurized Access 5.4.2.

Water Reactor Steam Generator Tubes The Westinghouse steam generator design permits access to steam generator tubes 16.3 /

(Rev. 1, July 1975) for inspection, plugging, or other repair.

4.4.5 Baseline Inspection Westinghouse 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 Limits The detailed requirements for inservice inspections are delineated in the Applicant's Technical Specifications.

1.84 Design and Fabrication Code Case Westinghouse believes that code cases should be treated as any other part of the 5.2.1.

Acceptability - ASME Section III, code and therefore accepted as approved by the ASME council.

Division 1 (Rev. 15, May 1979)

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere For Class 2, 3, and CS components, only code cases approved by the ASME council have been used.

1.85 Materials Code Case Acceptability - See compliance for Regulatory Guide 1.84 above. 5.2.1.

ASME Section III, Division 1 (Rev. 15, May 1979) 1.88 Collection, Storage, and Maintenance of The Westinghouse position for the WRD NSSS scope of supply on Regulatory Guide 17.1.2 Nuclear Power Plant Quality Assurance 1.88 is presented in WCAP-8370, WRD Quality Assurance Plan. The Nuclear Fuel 17.1.3 Records (Rev. 2, October 1976) Division position on this Regulatory Guide is presented in WCAP-7800, NFD 17.2 Quality Assurance Program Plan.

1.89 Qualification of Class IE Equipment of For Westinghouse NSSS Class IE equipment, Westinghouse meets IEEE Standard 3.11N Nuclear Power Plants 323-1974 (including IEEE 323a-1975 position statement of July 24, 1975) and (Rev. 0, November 1974) Regulatory Guide 1.89 by an appropriate combination of any or all of the following:

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.

1.92 Combination of Modes and Spatial Westinghouse will use the alternate method described in Section 3.7 to combine 3.7.2 Components in Seismic Response modal responses in the evaluation of seismic responses for Millstone 3.

Analysis (Rev. 0, December 1974) 1.97 Instrumentation for Light-Water-Cooled Westinghouse does not believe that adoption of the full scope of instrumentation 7.4 Nuclear Power Plants to Assess Plant proposed by Regulatory Guide 1.97, Revision 2, is necessary for safe operation of a 7.5 Conditions During and Following an Westinghouse pressurized water reactor. Nevertheless, Westinghouse has established Accident a design basis with respect to Regulatory Guide 1.97, Revision 2, in order to support (Rev. 2, December 1980) 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:

FSA R.G. Sectio No. Title Degree of Compliance Refere

1. 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere

- 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 withstand 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

FSA R.G. Sectio No. Title Degree of Compliance Refere 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 for attaining safe shutdown.

3. Implementation - Section D The 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.

1.99 Effects of Residual Elements on Justification of the Westinghouse position on Revision 0 and Revision 1 of the Guide 5.3 Predicted Radiation Damage to Reactor is detailed in References 1 and 2, respectively.

Vessel Materials (Rev. 1, April 1977)

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:

FSA R.G. Sectio No. Title Degree of Compliance Refere

1. Regulatory Position C.1 The basis, as well as the scope, of the guide for predicting adjustment of reference temperature as given in Regulatory Position C.1 are inappropriate since the data base used was incomplete and included some 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).

3. 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.

1.100 Seismic Qualification of Electric Westinghouse qualifies equipment to the requirements of IEEE 344-1971.

Equipment for Nuclear Power Plants Westinghouse has performed extensive testing to demonstrate the adequacy of (Rev. 1, August 1977) 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).

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.105 Instrument Setpoints Technical Specifications provide the margin from the nominal setpoint to the 7.1.2.

(Rev. 1, November 1976) technical specification limit. The allowances between the technical specification limit and the safety limit include the following items:

a. The inaccuracy of the instrument 7.2.2.

b. Process measurement accuracy 16.3/

16.4.3

c. Uncertainties in the calibration

d. The potential transient overshoot determined in the accident analyses (this may include compensation for the dynamic effect), and

e. 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 accident conditions. The need for qualification testing will be evaluated 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 locking 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.

FSA R.G. Sectio No. Title Degree of Compliance Refere 1.116 Quality Assurance Requirements for The subject of the Regulatory Guide is not in the Westinghouse NSSS scope of 17.1.2 Installation, Inspection, and Testing of supply. The Westinghouse position for the WRD NSSS scope of supply on 17.2 Mechanical Equipment and Systems Regulatory Guide 1.116 is presented in WCAP-8370, WRD Quality Assurance (Rev. O-R, May 1977) Plan. This Regulatory Guide is not applicable within NFD.

1.118 Periodic Testing of Electric Power and Westinghouse will make clear the distinction between recommendations and 7.1.2.

Protection Systems (Rev. 2, June 1978) requirements when addressing criteria. The position is as follows:

Westinghouse defines Protective Action Systems to mean electric, instrumentation, 8.1.6 and controls portions of those protection systems and equipment actuated and controlled by the protection system.

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.

Status, 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.

1.121 Bases for Plugging Degraded PWR Millstone 3 complies with Regulatory Guide 1.21, Rev. 0, August 1976 with the 5.4.2 Steam Generator Tubes (Rev. 0, August following clarifications:

1976)

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.

FSA R.G. Sectio No. Title Degree of Compliance Refere

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.

b. 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 threshold 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 names and references rather than the actual codes.

FSA R.G. Sectio No. Title Degree of Compliance Refere

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 are normally done to entirely different bases and criteria are not within the scope of this guide.

1.123 Quality Assurance Requirements for The Westinghouse position for the WRD NSSS scope of supply in Regulatory Guide 17.1.3 Control of Procurement of Items and 1.123 is presented in WCAP-8370, WRD Quality Assurance Plan. The Nuclear 17.2 Services for Nuclear Power Plants (Rev. Fuel Division position on this Regulatory Guide is presented in WCAP-7800, NFD 1, July 1977) Quality Assurance Program Plan.

1.126 An Acceptable Model and Related Millstone 3 does not use the model and related statistical method for the analysis of 4.2 Statistical Methods for the Analysis of fuel densification that is presented in Regulatory Guide 1.126.

Fuel Densification (Rev. 1, March 1978)

The 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.

1.139 Guidance for Residual Heat Removal Millstone 3 meets the requirements of BTP RSB 5-1 and SRP 5.4.7 except as (Rev. 0, May 1978) indicated in FSAR Sections 5.4.7.1 and 5.4.7.2.4.

1.141 Containment Isolation Provisions for Westinghouse's containment isolation philosophy for fluid systems complies with the 5.4.5 Fluid Systems guidance provided by ANSI N271-1976 and/or Regulatory Guide 1.141 with the 5.4.6 (Rev. 0, April 1978) following exceptions and/or clarifications: 6.2.4

FSA R.G. Sectio No. Title Degree of Compliance Refere

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.

Sections 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 generated. Westinghouse design criteria for the initiation of containment isolation do not include a requirement for diversity in the primary system variables for any given design condition or event.

FSA R.G. Sectio No. Title Degree of Compliance Refere 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 sensed that the Regulatory Guide unjustifiably ignores the benefits of alternate methods of obtaining different means of actuation. Westinghouse believes the standard gives appropriate guidance 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 safety 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 the seal injection lines following an accident. Each line is, however, equipped with a remote manual containment isolation valve which the operator can close when the charging pumps have completed their safeguard function.

1.144 Auditing of Quality Assurance Program The Westinghouse position for the WRD NSSS scope of supply on Regulatory 17.1.2 for Nuclear Power Plants (Rev. 0, Guide 1.144 is presented in WCAP-8370, WRD Quality Assurance Plan. The January 1979) Nuclear Fuel Division position on this Regulatory Guide is presented in WCAP-7800, NFD Quality Assurance Program Plan. 17.1.3 17.2 1.146 Qualification of Quality Assurance The Westinghouse position for the WRD NSSS scope of supply on Regulatory Program Audit Personnel for Nuclear Guide 1.146 is presented in WCAP-8370, WRD Quality Assurance Plan. The Power (Rev. 0, August 1980) Regulatory Guide is not applicable within NFD.

Millstone 3 FSAR was reviewed against NUREG-0800 at the time of application for an rating license to satisfy the requirements of 10 CFR 50.34(g). The following two tables were eloped to identify deviations from SRP acceptance criteria and provided a justification for e deviations. The differences noted were not construed as variances from regulation, rather, documented the deviations from acceptance criteria as stated within the NRC's internal ew guide (SRP) for safety analysis reports.

le 1.9-1 summarized the differences between the Millstone 3 FSAR and NUREG-0800.

le 1.9-2 presented the FSAR differences from NUREG-0800 and their justifications.

rmation contained in this section (Tables 1.9-1 and 1.9-2) has been retained for historical poses.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 2.1.3 (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 - Zones of alterations, irregular These features are not discussed. 2.5.1.2 weathering, structural weakness; unrelieved residual stresses in rock; unstable material or areas considered unstable.

2.5.2 (Rev. 1) II (2.5.2.1) - Magnitude designations. Earthquake magnitudes not identified as Mb, ML, or Ms. Tables 2.5.2-and 2.5.2-4 II (2.5.2.4) - Earthquake return period. There is no probabilistic determination of earthquake return 2.5.2.4 period for the largest earthquakes in each province.

II (2.5.2.7) - Probability of exceeding acceleration of No estimate given for probability of exceeding acceleration 2.5.2.7 OBE. of OBE.

2.5.3 (Rev. 2) II (2.5.3.1) - Offshore geologic investigation. No detailed offshore geologic investigation of 5-mi radius 2.5.3.1 has been done.

II (2.5.3.4) - Fault location. Location and investigation of all faults within 5 mi has not 2.5.3.4 been done.

II (2.5.3.6) - Age of faults. Age documentation of all faults within 5 mi has not been 2.5.3.6 done.

2.5.4 (Rev. 2) II (2.5.4.2) - Analyses for saturated soils and clays No FSAR table lists values of parameters used in analyses 2.5.4.2 underlying the site. of liquefaction potential, behavior, or static and dynamic behavior.

3.2.2 (Rev. 1) II - Use of RG 1.26 as the acceptance criteria for Millstone 3 uses the classification system provided in ANS 3.2.2 defining Quality Groups for components important to 18.2.

safety.

3.4.1 (Rev. 2) III.3 - Postulated failures of nonseismic Category I FSAR does not directly address these postulated failures. 3.4.1, 3.6 and non-tornado protected tanks.

3.4.2 (Rev. 2) II. 1 - Use of DBF or highest groundwater level in DBF or normal groundwater level used. 3.4.2 design.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 3.5.1.3 II - Missile protection for cold shutdown maintenance Cold shutdown maintenance components not specified as 3.5.1.3 (Rev. 1) components. targets.

3.5.1.5 (Rev.1) III.3 - Definition of PP. Missiles that produce secondary missiles which could 3.5.1.5, damage vital equipment not considered. 2.2.3.2.1 3.5.1.6 (Rev. III.2 - Inflight crash rate of 4x10-10. NUREG-75/087 crash rate of 3x10-9 used. 3.5.1.6 2) 3.6.1 (Rev. 1) BTP ASB 3-1, B.1.a(1) -Requires an arbitrary split be FSAR does not commit to postulate this arbitrary split. 3.6.1.3.3 postulated on the main steam and the feedwater systems at a location proximate to essential systems.

BTP ASB 3-1, B.1.a(2) -Suggests main steam and Main steam and feedwater pipes are routed in the vicinity of 3.6.1.3.3 feedwater pipes not be routed in the vicinity of the the control room.

control room.

BTP ASB 3-1, B.2.a -States that essential systems and In the aux steam and hot water heating systems, only the 3.6.1.3.1 components should be designed to meet the seismic electrical detection and actuation devices for the isolation design criteria of R.G. 1.29. valves are qualified to Class 1E requirements and located in a seismic Cat. I Building.

3.6.2 (Rev. 1) III.2.a - Requires use pressure and temperature values FSAR uses internal pressure, and temperature conditions in 3.6.2.2.1 corresponding to the greater contained energy at hot the piping system during reactor operation at 100 percent standby or at 102 per cent power. power.

III.2.a - Requires that the allowable capacity for FSAR uses an allowable of 80 percent of energy absorbing 3.6.2.2.1 crushable material shall be limited to 80 percent of its capacity based on static testing.

rated energy absorbing capacity as determined by dynamic testing.

BTP MEB 3-1, B.1.e -States particular criteria for FSAR does not postulate cracks in high energy piping. 3.6.2.1.2 postulating through wall leak age cracks in high energy Class 1, 2, 3, and non-nuclear piping.

3.7.2 (Rev. 1) II.4 - Finite element and half space representation of FSAR does not address half space representation. 3.7.2 subgrade soil stiffness.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti II.11 - Consideration of accidental torsion by Additional seismicity not addressed. 3.7B.2.11 assuming additional seismicity of +/-5 per cent of maximum building dimension.

3.7.3 (Rev. 1) II.2.l (1) - Interaction of Category I piping extended to FSAR discussion is restricted to piping. 3.7B.3.13 (BOP Scope) address on a system basis.

3.7.3 (Rev. 1) II.2.g - Combination of closely spaced modes should Westinghouse combines closely spaced modes as described 3.7N.3.7 (NSSS Scope) be in accordance with Regulatory Guide 1.92. in FSAR Section 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 Article CC-3000 of ASME III, Division 2 was not not used. 3.8.1.4.1 for design of containment structure tangential shear.

II.4.j - Ultimate capacity of reactor containment. Ultimate capacity of the reactor containment is not 3.8.1 discussed.

II.5 - Article 3000 of ASME III, Division 2 for loads, Article 3000 of ASME III, Division 2 not used. 3.8.1.2 load combinations, and stress allowables.

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, 3.8.1.2 II.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 3.9.1 (Rev. 2) III.1 - Plant conditions identified as design Levels FSAR identifies plant conditions as normal, upset, 3.9B.1.1 (BOP Scope) A,B,C,D. emergency, and faulted.

III.4 - Methods used in stress analysis of components. FSAR contains no justification for methods used 3.9B.1.4 3.9.1 (Rev. 2) II.2 - Computer codes used in design and analysis of Only a brief description of computer codes used by 3.9N.1.2 (NSSS Scope) seismic Category I components. Westinghouse is given.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 3.9.2 (Rev. 2) II.1.d - List snubbers on systems which experience FSAR does not provide a list of snubbers. 3.9B.2 (BOP Scope) sufficient thermal expansion.

II.1.e and f - Tests to verify thermal expansion/ FSAR does not provide a description of tests. 3.9B.2 vibration measurements.

II.2 - Seismic sub- system analysis. Information is not contained in FSAR Section 3.9B.2. 3.9B.3 3.9.2 (Rev. 2) II.2.e - Criteria for combining closely spaced modes. Westinghouse method is provided in FSAR 3.7N.3.7 (NSSS Scope) Section 3.7N.3.7.

3.9.3 (Rev. 1) II.1 - Stress limit criteria. FSAR does not reflect the stress limit criteria. 3.9B.3.1 (BOP Scope)

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) II.1 - Design criteria for internal parts of components Westinghouse does not provide criteria for the nonpressure 3.9N.3 (NSSS Scope) such as valve discs and pump shafts. boundary portions of ASME Code Class 1, 2, and 3 components in the FSAR.

Appendix A, 1.3.3 -Design basis pipe break (DBPB) Westinghouse defines DBPB as a faulted, not emergency 3.9N condition.

Appendix A, 3.1 - Stress limits and loading FSAR does not provide a table defining stress limits and 3.9N combinations for core support structures. loading combinations for core support systems.

3.10 (Rev. 2) II.1.a (14) (b) iii and iv - LOCA induced hydraulic There is no analysis of LOCA induced hydraulic functions 3.10B.1 (BOP Scope) forcing functions or differential pressures upon valve or differential pressures upon valve discs or pump rotors.

discs or pump rotors.

II.5.c - Seismic Qualification Report format. Seismic Qualification Report format was not used to 3.10B.1 document seismic qualification.

II - Mechanical equipment seismic and operability FSAR does not address mechanical equipment seismic and 3.10B.1 qualification. operability qualification.

II.1 - Verify operability of pumps and valves during all Pump operability has only been performed by analysis. 3.10B.1 operational conditions by test and analysis.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti II.5.b - Listing of systems necessary to perform This list is not included in FSAR Section 3.10. 3.10B.1 functions out lined in SRP 3.10.

3.10 (Rev. 2) II - Mechanical equipment seismic and operability FSAR does not address mechanical equipment seismic and 3.10N (NSSS Scope) qualification. operability qualification.

II.1.a(2) - Testing of equipment in the operational Flow loads are not superimposed on seismic loads for valve 3.10N condition. operability tests.

II.1.a(8) - Fixture design for seismic tests. The seismic qualification testing configurations are 3.10N designed to represent the typical plant installation.

II.1.a(10) - Static testing of pump or valve assemblies. End loadings are not applied and all dynamic amplification 3.10N effects are not included in the static deflection test for active valves.

II.1.a(14)(a) - Operability of active pumps and valves. Operability of active pumps and valves are not covered in 3.7N, 3.9N FSAR Section 3.10N.

The Millstone program utilizes a combination of test and 3.9N analysis to demonstrate operability for active valves.

No specific tests are done on valve gate, disc assemblies, 3.9N motor, etc.

II.1.a(14)(b) viii - Use of R.G. 1.92 for combination of Westinghouse utilizes the methods defined in FSAR 3.7N multimodal and multidirectional responses. Section 3.7 for combining closely spaced modes.

II.1.b(3) - Testing of supports. Seismic testing of all supports is not conducted. 3.10N II.1.c - Seismic and dynamic testing according to For some mechanical components, aging and sequence 3.10N IEEE 323-1974. testing was not included.

II.3 - Requirements for central files. Requirements for central files are not addressed in FSAR 3.10N Section 3.10N.

II.5.b(1) - Requirement for a list of systems necessary This list is not included in FSAR Section 3.10N. 3.10N, Chapt to perform the functions outlined in SRP 3.10. 6 and 7

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti II.5.b(2) - Description of the results of any in-plant Actual test results are not included in the FSAR. 3.10N tests.

II.5.c - Contents of Seismic Qualification Report Westinghouse does not maintain such a report for Millstone 3.10N (SQR). 3.

3.11 (Rev. 2) II - Mechanical equipment qualification. FSAR does not address mechanical equipment qualification. 3.11B.1 (BOP Scope)

II - NUREG-0588 methodologies. NUREG-0588 methodologies are not strictly followed. 3.11B.1 3.11 (Rev. 2) I.1 - Requirement for a list of systems necessary to This list is not included in FSAR Section 3.11N. Chapters 6 (NSSS Scope) perform functions outlined in SRP 3.11. and 7, 3.7N, 3.9N, 3.11N II - NUREG-0588. No reference is made in FSAR Section 3.11N to the results 3.11N of NUREG-0588 study.

Scope of SRP 3.11 -Environmental qualification of Environmental qualification of mechanical electrical 3.9N mechanical and electrical equipment. equipment is not addressed in FSAR Section 3.11N.

4.5.1 (Rev. 2) II.4 - On site cleaning requirements. On site cleaning requirements are not addressed in FSAR 1.8, 17.1.2 Section 4.5.1.

5.2.1.1 II - Compliance with 10 CFR 50.55a. FSAR used ASME III 1971 Ed. through Summer 1972 5.2.2.1 (Rev. 2) addenda instead of Winter 1972 addenda for design and fabrication of loop bypass valves.

Westinghouse uses ANS standards rather than R.G. 1.26. 1.8N, 3.2.

5.2.5 (Rev. 1) III.7 - Leakage detection testing requirements. FSAR does not completely address testing of the 5.2.5.2.7, unidentified leakage sump system. 9.3.3 5.3.1 (Rev. 1) II.6.c(3) - Capsule removal schedule. The tentative capsule removal schedule is not identical to 5.3.1.6 (NSSS Scope) the removal schedule described in 10 CFR 50, Appendix H, II.C.3.b.

5.4.1.1 II.2.b - Normal operating temperature is at least 100°F The FSAR states that the RTNDT is no higher than 10°F. 5.4.1.1.3 (Rev. 1) above the RTNDT.

(NSSS Scope)

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 5.4.2.1 BTP MTEB 5-3 -Secondary side chemistry program. Free hydroxide concentration is not measured. Procedure 5.4.2.1 (Rev. 2) number and basis not supplied for chemical analysis.

(NSSS Scope)

BTP MTEB 5-3, II.2 -Discussion of clean metal FSAR does not address this concern. 5.4.2.1 condition prior to startup.

II.B.2 - Access to remove sludge by lancing from tube Tube lancing from tube support plates is not discussed in the 5.4.2.1 support plates. FSAR.

6.2.1 (Rev. 2) II.f (6.2.1.1A) - Adequate margin above the maximum Actual margin of external pressure analysis not specifically 6.2.1.1.3.5 (BOP Scope) expected external pressure. addressed.

II.B.2 (6.2.1.2) -NUREG-0609. NUREG-0609 was not addressed. 6.2.1.2 6.2.1 (Rev. 2) II.2 (6.2.1.5) - BTP CSB 6-1 Section B.3b Westinghouse values for the long-term post-blowdown 6.2.1.5 (NSSS Scope) recommends conservative condensing heat transfer condensing heat transfer coefficients are nonconservative.

coefficients.

II.B.3.e (6.2.1.3) - Models for calculating mass and Model described in FSAR differs from Westinghouse model 6.2.1.3.3, energy releases for containment design basis referenced in the SRP. 6.2.1.3.4, calculations. 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 Millstone 3 switchover is manual. 6.2.2 mode.

II.1.e (6.5.2) - pH between 8.5 and 10.5 for fission Millstone 3 is designed to a minimum pH of 7.0 6.2.2 product control.

6.2.3 (Rev. 2) II.D.1 and II.D.2 -Discussion heat transfer analysis FSAR does not provide a discussion of these subjects. 6.2.3 and high energy line considerations.

6.2.5 (Rev. 2) II.3 - Mixing characteristics of the containment (plant Millstone 3 references the analyses of plants with similar 6.2.5.3 specific analyses). designs.

II.11 - Containment hydrogen monitor requirements. Hydrogen monitors will comply with these requirements by 6.2.5 core load.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 6.3 (Rev. 1) II - Discussion of non-safety grade interactions with FSAR Section 6.3 does not discuss this subject. 3.6 ECCS.

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 Single failure (active) criteria. Valves isolating air inlet ducting located in series. 6.4.3 A (Rev. 2)

Item 6 - No manual action credit allowed for repairs Isolation valves can be manually opened within 10 minutes, 6.4.3 until two hours. thus, credit for manually opening the valves within one hour should be acceptable.

6.5.1 (Rev. 2) II - Compliance to Regulatory Guide 1.52. See exceptions listed in Section 1.8. 6.5.1 II - Continuous indication and recording of air flow This is not provided. 6.5.1 for ESF filtration units.

II - Flow sensors in ESF filtration units. These are not provided. 6.5.1 II - Compliance to Section 8.3.1.6 of ANSI Three ESF filter train systems do not comply. 6.5.1 N510-1980.

II - Compliance to Sections 8.3.1.5, 8.3.1.6, and Three ESF filter train systems do not comply. 6.5.1 8.3.1.7 of ANSI N510-1980.

6.5.2 (Rev. 1) Containment spray as a fission product cleanup Containment spray system requirements are not discussed in 6.5.2 system. Section 6.5.2.

7. 2 (Rev. 2) BTP ICSB 26 - Sensor qualification. Sensors for reactor trip on turbine trip when power level is 7.2.1.1.2 50% or more are not seismically qualified.

7.5 (Rev. 2) III.6 - NUREG-0696 compliance. The Safety Parameter Display System, and the Emergency 7.5.3 Response Facilities are not discussed.

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 8.3.1 details.

9.1.2 (Rev. 3) III.2.e - Evaluation of lighter load drops at maximum This evaluation has not been performed. 9.1.2.3 heights.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti 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 9.1.3.2 ORIGEN2) computer code and credit for evaporative cooling instead of BTP ASB 9-2.

9.1.3 (Rev. 1) III.1.d - For maximum normal heat load, the pool The maximum temperature for a normal heat load is 150°F. 9.1.3.3 temperature should be kept at or below 140°F 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 The decay time for the maximum heat load is based on the 9.1.3.3 decay 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 />.

9.1.4 (Rev. 2) III.6 - Evaluation of lighter load drops at maximum This evaluation has not been performed. 9.1.4.1 heights.

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-coolant test for reactor coolant-pumps. Reactor-coolant-pumps have not been tested for the 20 9.2.2 minute time requirement.

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 9.4.4 least 20 feet above grade elevation.

II.5 - Detection and Control of airborne contamination Only normal building ventilation is monitored. 9.4.4.2 leakage from the system.

III.3.b - Tornado protection. No protection of ductwork from negative pressure due to 9.4.4.4 tornado.

BTP CMEB Refer to the Fire Protection Evaluation Report, Fire Protectio 9.5-1 (Rev 2) Appendix B, for a comparison of Millstone 3 design to Evaluation BTP CMEB 9.5-1 guidelines. 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 9.5.4.2 turbulence of sediments.

Correspondi SRP Section Specific SRP Acceptance Criteria Summary Description of Difference FSAR Secti II.2 (III.6.a) - Missile protection. The fill lines for the diesel generator fuel oil vaults are 9.5.4.2 protected from missiles.

9.5.8 (Rev. 2) II.4.g (III.8) - Reducing airborne particulate material. The bottoms of the fresh air intakes are all located at least 20 9.5.8 feet above grade elevation.

10.2.3 (Rev. 1) II.1 - FATT and Charpy V-notch energies. GE does not provide data for FATT and Charpy V-notch 10.2.3 energies to compare with SRP.

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 2, Item 5 - Spent fuel pool treating system. No automatic termination of effluents. 9.3.2, 9.4.

Table 2, Items 16 and 17 -Steam generator blowdown No automatic termination of effluents. 11.5.2.3.3 system.

12.2 (Rev. 2) I.2 - Tabulation of concentrations of airborne Only normal operation and anticipated operational 12.2.2 radioactive materials. occurrences are addressed.

13.5.2 II.C.2 - ANSI/ANS 3.2-1981, Section 5.3 FSAR uses ANSI N18.7-1976/ANS 3.2, Section 5.3 13.5.2 14.2 (Rev. 2) II.4 - Categories of reportable occurrences that are FSAR does not provide categories of occurrences. 14.2 repeatedly being experienced at other facilities.

15.4.6 Entire SRP. FSAR does not address this accident scenario. 15.4.6 15.4.8 (Rev. 1) III - Stresses should be evaluated to emergency Westinghouse considers a faulted condition as stated in 15.4.8 conditions for these accidents. ANSI N18.2.

15.6.5 (Rev. 2) II.3 - TMI Action Plan, II.K.3.30 and II.K.3.31. No modifications have been made to the small break LOCA 15.6.5.3 model.

15.7.3 III.1.a - Radionuclide inventory in failed components FSAR analyzed postulated tank failure using 1% fuel 2.4.13.3, defects. 15.7.3.2

P 2.1.3 P TITLE: POPULATION DISTRIBUTION 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.

Justification for difference from SRP Distances 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 distances were used in both documents.

P 2.5.1 P TITLE: BASIC GEOLOGIC AND SEISMIC INFORMATION Actual differences between FSAR and SRP

1. SRP 2.5.1, Paragraph II.4.b, requires a discussion on zones of alterations, irregular weathering, and structural weakness. 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 discussion on unstable materials or areas considered unstable due to physical properties. These are not discussed in FSAR 2.5.1.

Justification for differences from SRP

1. Major alteration zones, irregular weathering, and structural weakness do not exist at the site.

2. There is no history of stress relief problems 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.

P 2.5.2 P TITLE: VIBRATORY GROUND MOTION Actual differences between FSAR and SRP

1. Magnitudes of earthquakes shown in FSAR Tables 2.5.2-3 and 2.5.2-4 are not identified as Mb, ML, or Ms as specified in SRP 2.5.2, Paragraph II(2.5.2.1).

2. FSAR Section 2.5.2.4 does not give a probabilistic 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 estimate 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).

Justification for differences from SRP

1. The magnitude of most earthquakes is based on the relationship between empirical intensity and magnitude. It is not possible to be precise in terms of Mb or ML for historical earthquakes. None of the magnitudes listed in the FSAR text are surface wave magnitudes (Ms).

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.

P 2.5.3 P TITLE: SURFACE FAULTING Actual differences between FSAR and SRP

1. FSAR Section 2.5.3.1: A detailed offshore 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 location and investigation of every fault within 5 miles of the site has not been performed 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 required in SRP 2.5.3, Paragraph II(2.5.3.6).

Justification for differences from SRP

1. 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 included 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 million 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 identifying every fault within a 5-mile radius of the site was not considered necessary.

3. Because of the above justifications, no detailed offshore geologic 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.

P 2.5.4 P TITLE: STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS Actual difference between FSAR and SRP SRP 2.5.4, Paragraph II(2.5.4.2), asks for presentation 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 table listing values of parameters used to perform the analyses mentioned above.

Justification for difference from SRP Although the FSAR does not contain a specific table listing values of parameters used to perform the required analyses, the data is available in applicable subsections of Section 2.5.4 as listed below:

• Liquification potential - FSAR Section 2.5.4.8 and its subsections contain values of parameters used in evaluating liquification potential of the underlying soils beneath Category I structures.

• Consolidation behavior - No clay soils are present at the site. The method of analysis to determine elastic settlement of Category I structures on bedrock, glacial till and structural backfill is contained in FSAR Section 2.5.4.10.2. Values used in the analysis to determine settlement of underlying soils are contained in FSAR Section 2.5.4.4.3.

• Static and dynamic behavior - Values of parameters used for evaluating static and dynamic behavior of underlying 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.

P 3.2.2 P TITLE: SYSTEM QUALITY GROUP CLASSIFICATION Actual difference between FSAR and SRP SRP 3.2.2, Subsection II, utilizes Regulatory Guide 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 provided in ANS 18.2. FSAR Section 3.2.2 provides a cross reference between the ANS safety classifications and the Quality Groups defined in Regulatory Guide 1.26.

Justification for difference from SRP The 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 Regulatory Guide 1.26 Quality Group classification system provided a correlation between Quality Groups and the classification 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.

P 3.4.1 P TITLE: FLOOD PROTECTION Actual difference between FSAR and SRP SRP 3.4.1, Paragraph III.3, discusses the review 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.

Justification for difference from SRP Millstone 3 does not have QA Category I tanks that are nonseismic. Postulated 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 located outside safety related structures in areas that would preclude flooding of safety related equipment.

Non-QA Category I tanks within safety related structures are not considered to contain sufficient inventory to cause flooding of safety 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 damage due to potential flooding that would result if the non-QA Category I tanks were postulated to fail in a seismic event.

P 3.4.2 P TITLE: ANALYSIS PROCEDURES 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 been historically reported for the site. FSAR 3.4.2 states that structures located above the DBF level are designed for the hydrostatic effects of uplift and water pressure resulting from the DBF or normal groundwater, whichever is most severe.

Justification for difference from SRP FSAR Section 2.5.4.6 describes the groundwater conditions for the Millstone 3 site and includes a description of the low permeability of the bedrock as well as the overlying glacial till. Because of the low permeability of these materials at the site the groundwater level would not significantly change.

Therefore, the normal groundwater level and its associated dynamic effects are sufficient for 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 hydrostatic loading on the structure foundations.

P 3.5.1.3 P TITLE: TURBINE MISSILES 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.

Justification for difference from SRP Evaluation of turbine missiles, as described in FSAR Section 3.5.1.3, has concluded that the probability of a turbine missile being generated and causing damage to a safety related system or component is lower than what is recommended in Regulatory Guide 1.115.

P 3.5.1.5 P TITLE: SITE PROXIMITY MISSILES (EXCEPT AIRCRAFT)

Actual difference between FSAR and SRP SRP 3.5.1.5, Paragraph III.3, states that the definition of Pp includes probability of missiles that produce secondary missiles which could damage vital equipment. This was not considered in the FSAR analysis.

Justification for difference from SRP FSAR 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 (Pp = 1) provides a total probability which is still within the acceptance criteria provided in the SRP.

P 3.5.1.6 P TITLE: AIRCRAFT HAZARDS 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.

Justification for differences from SRP A response to this difference has been provided in the Applicant's response to NRC Acceptance Review Request Number 311.4.

P 3.6.1 P TITLE:PLANT DESIGN FOR PROTECTION AGAINST POSTULATED PIPING FAILURES IN UID SYSTEMS OUTSIDE CONTAINMENT Actual differences between FSAR and SRP

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. The split must be postulated regardless 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 room. 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 criteria of Regulatory Guide 1.29. The BTB defines essential systems and components as those required to shut down the reactor and mitigate the consequences 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.

Justification for differences from SRP

1. Essential systems, components, or structures 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 discussion in FSAR Section 3.6.1.3.3.

3. Redundant isolation capability is provided for both systems with Category 1E qualified detection and actuation 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.

Given a pipe break of the auxiliary steam or hot water heating systems in a safety related building, isolation of the affected system is capable assuming the most limiting single failure. The isolation valves are normally open and fail in the closed position. To ensure continued isolation capability following the postulated pipe break, plant operating procedures require manual valves to be closed to isolate the affected piping.

P 3.6.2 P TITLE:DETERMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS SOCIATED WITH THE POSTULATED RUPTURE OF PIPING Actual differences between FSAR and SRP

1. SRP 3.6.2, Paragraph III.2.a, states that pressure and temperature values should correspond to the greater contained energy at hot standby or at 102 percent power. FSAR Section 3.6.2.2.1 states that pressure and temperature values associated with reactor 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 commit to dynamic testing to determine energy absorbing capacity.

3. BTP MEB 3-1, B.1.e, states particular criteria for postulating 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.

Justification for differences from SRP

1. 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 absorbing material are not significant. The application is absorbing kinetic energy from pipe whip through 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.

P 3.7.2 P TITLE: SEISMIC SYSTEM ANALYSIS Actual differences between FSAR and SRP

1. SRP 3.7.2, Paragraph II.4, requires consideration 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 consideration of accidental torsion by assuming an additional seismicity of 5 percent of the maximum building dimension at the level under consideration. This is not addressed in the FSAR.

Justification for differences from SRP

1. 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 finalized 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.

P 3.7.3 P TITLE: SEISMIC SUBSYSTEM ANALYSIS 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.

Justification for difference from SRP Additional information will be provided in an amendment to the FSAR.

P 3.7.3 P TITLE: SEISMIC SUBSYSTEM ANALYSIS 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 described in FSAR Section 3.7N.3.7.

Justification for difference from SRP The 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.

P 3.8.1 P TITLE: CONCRETE CONTAINMENT Actual differences between FSAR and SRP

1. FSAR Section 3.8.1 does not reference Regulatory 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 modes 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, Division 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.

Justification for differences from SRP

1. 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 containment with respect to failure modes has been considered in the PRA study, which will be submitted as a separate report.

3. ASME III, Division 2, was 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 Summer 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, was 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.

P 3.8.3 P TITLE:CONCRETE AND STEEL INTERNAL STRUCTURES OF STEEL OR CONCRETE NTAINMENTS

Actual difference between FSAR and SRP ACI 349-76 was not used as described in SRP 3.8.3, Paragraph II.2.

Justification for difference from SRP This code was not in effect at the time of the Construction Permit. ACI 318, AISC-1969 Ed. and ASME III 1971 Ed. through Summer 1973 addenda were the codes used. Guidance found in 10 CFR 50 regulation does not require continuous upgrading of the codes and standards used in the design.

P 3.8.4 P TITLE: OTHER SEISMIC CATEGORY I STRUCTURES Actual differences between FSAR and SRP

1. 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. The Applicant's design information is not in this format.

Justification for differences from SRP

1. 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 continuous upgrading of the codes and standards used in the design.

2. 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.

P 3.8.5 P TITLE: FOUNDATIONS 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.

Justification for difference from SRP ACI 349-76 was not in effect at the time the construction permit was issued. Guidance found in 10 CFR 50 regulation does not require continuous upgrading of the codes and standards used in the design.

P 3.9.1 P TITLE: SPECIAL TOPICS FOR MECHANICAL COMPONENTS 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, Paragraph III.1, requires them to be identified as Design Level A, B, C, and D.

Also, allowables used in stress analysis 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.

Justification for differences from SRP

1. 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.

2. 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 III 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.

P 3.9.1 P TITLE: SPECIAL TOPICS FOR MECHANICAL COMPONENTS Actual difference between FSAR and SRP (NSSS Scope)

SRP 3.9.1, Paragraph II.2, requires a considerable amount of information for all the computer codes used in the design and analysis of Seismic Category I components. Westinghouse only provides a brief description of the computer codes used by Westinghouse for component design and analysis in FSAR Section 3.9N.1.2. Additional information required by the SRP for the computer codes referred to in FSAR Section 3.9N.1.2 is provided by reference to WCAPs-8252 and 8929. Computer codes used by Westinghouse vendors are not included in the FSAR.

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 approved 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) and the technical review of various design documents submitted by vendors to Westinghouse.

The NRC's Mechanical Engineering Branch has interpreted this NRC guideline to be applicable only to computer codes used by the major contractors (i.e., NSSS supplier/AE).

P 3.9.2 P TITLE:DYNAMIC TESTING AND ANALYSIS OF SYSTEMS, COMPONENTS, AND UIPMENT Actual differences between FSAR and SRP (BOP Scope)

1. FSAR Section 3.9B.2 does not provide a list of snubbers on systems which experience significant thermal expansion, 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 described in SRP 3.9.2, Paragraph II.1.e and f.

2. Information required by SRP 3.9.2, Paragraph II.2, is not contained in FSAR Section 3.9B.2

Justification for differences from SRP

1. The Technical Specifications will provide details on snubber testing and a list of safety related snubbers.

2. Information requested is provided in FSAR Section 3.9B.3.

P 3.9.2 P TITLE:DYNAMIC ANALYSIS AND TESTING OF STRUCTURES, SYSTEMS, AND UIPMENT 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 spaced modes is provided in FSAR Section 3.7N.3.7.

Justification for difference from SRP The 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.

P 3.9.3 P TITLE:ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND RE SUPPORT STRUCTURES Actual differences between FSAR and SRP (BOP Scope)

1. FSAR Section 3.9B.3.1 does not reflect the stress limit criteria of SRP 3.9.3, Paragraph II.1, and Appendix A.

2. SRP 3.9.3, Paragraph II.2, requires information on Class 3 safety/relief devices along with Classes 1 and 2. As written, the FSAR does not specifically address Class 3 safety and relief devices.

3. The requirements of SRP 3.9.3, Paragraph II.3, for snubbers have been fulfilled but have not been included here.

Justification for differences from SRP

1. The allowable values utilized by Millstone 3 for piping systems meet the requirements stipulated in ASME Section III, 1971 Ed. through Summer 1972 addenda.

Components, except piping, use stress criteria 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/pressure 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.

P 3.9.3 P TITLE:ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND RE SUPPORT STRUCTURE Actual differences between FSAR and SRP (NSSS Scope)

1. SRP 3.9.3, Paragraph II.1, requires that design criteria for internal parts of components, such as valve discs and pump shafts, be provided. Westinghouse does not provide criteria for the nonpressure 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, defines the design basis pipe break (DBPB) as an emergency condition. For ASME Code Class 1, 2, and 3 components and component supports, Westinghouse defines the DBPB as a faulted condition (see loading combination tables in Section 3.9N).

3. SRP 3.9.3, Appendix A, Paragraph 3.1, requires that stress limits and loading combinations be provided for core support structures. The FSAR does not currently provide tables defining load combinations and stress limits for core support structures.

Justification of differences from SRP

1. Westinghouse does not consider it appropriate to provide this type of detail in the FSAR.

Westinghouse employs good engineering practice in defining design criteria for critical internal parts. Generally, for critical internal parts of components, equipment specifications limit stresses to the criteria defined in the ASME Code or to the yield strength of the material.

2. Westinghouse defines the DBPB as a faulted condition event consistent with the criteria defined in ANS 18.2. Additionally, Westinghouse considers the stress limits and analysis methods for faulted 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.

P 3.10 P TITLE:SEISMIC AND DYNAMIC QUALIFICATION OF MECHANICAL AND ELECTRICAL UIPMENT 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 Qualification Report format, as specified in SRP 3.10, Paragraph II.5.c, was not used to document seismic qualification.

3. Mechanical equipment seismic and operability qualification is not addressed in FSAR Section 3.10B.1, as specified in SRP 3.10, Subsection II.

4. SRP 3.10, Paragraph II.1, requires a combination of test and analysis to verify the operability of pumps and valves during all plant operational conditions. Pump operability has only been 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.1.

Justification of differences from SRP

1. 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 restrained to maintain stresses within allowable limits to assure operability.

2. Controlled seismic qualification files containing equivalent information are maintained.

3. Seismic qualification of mechanical equipment is described in FSAR Sections 3.7B.1 and 3.9B.1. Pump and valve operability qualification is discussed in FSAR Section 3.9B.3.2.

4. The seismic and operability qualification programs described in FSAR Section 3.9B.3.2 provide adequate assurance of proper equipment 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 and7.

P 3.10 P TITLE:SEISMIC AND DYNAMIC QUALIFICATION OF MECHANICAL AND ELECTRICAL UIPMENT Actual differences between FSAR and SRP (NSSS Scope)

1. FSAR Section 3.10N currently addresses only Category I instrumentation 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 loadings simulating normal plant conditions should be superimposed on seismic and dynamic loads. This includes flow induced loads and degraded flow conditions. For the tests performed by Westinghouse, operational conditions are included where practical, simulated in some manner, or addressed by analysis. Flow loads are not superimposed on seismic loads for valve operability tests.

3. SRP 3.10, Paragraph II.1.a(8), requires that fixture design for seismic tests should simulate actual service mounting and should not cause any extraneous dynamic coupling to the test item. Westinghouse seismic qualification 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 impractical, 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 operability of active pumps and valves as specified in SRP 3.10, Paragraph II.1.a(14)(a).

6. FSAR Section 3.9N discusses the approach for seismic and operability qualification of safety related mechanical equipment. SRP 3.10, Paragraph II.1.a(14)(a), only allows analysis to be used for demonstrating structural integrity. It further states that operability of active pumps and valves must be demonstrated by test or a combination 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), requires that all complex active components other than pump and valve bodies (simple and passive elements) should be tested for operability. The Millstone 3 valve operability 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 multidirectional responses in analyses. Westinghouse utilizes the methods defined in FSAR Section 3.7 for combining closely spaced modes.

9. SRP 3.10, Paragraph II.1.b(3), states that 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 testing of all supports is not conducted.

10. SRP 3.10, Paragraph II.1.c, requires that seismic and dynamic testing be performed in sequence in accordance with IEEE 323-1974. For some mechanical components, aging and sequence testing was not included as part of the seismic and operability testing.

11. SRP 3.10, Paragraph II.3, spells out requirements 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 seismic qualification report. Westinghouse does not maintain such a report for Millstone 3.

Justification for differences from SRP

1. The guidance provided in NUREG-0800 does 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 operability 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 conditions other than by test.

The active valve operability program defined in FSAR Section 3.9N.3.2 outlines the program for demonstrating operability under all required plant conditions. This program of conservative design, analysis, 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 accordance with these interface criteria would have been adequately considered during the test program.

4. Westinghouse places conservative restrictions 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 response is a concern. In most cases, the equipment is rigid and does not display dynamic amplification characteristics.

5. The Millstone 3 operability program is covered in FSAR Section 3.9N.3.2. The latest version of the SRP has included operability under Section 3.10 and has deleted it from Section 3.9.

6. Programs currently in effect for Millstone 3 utilize analysis for demonstrating operability of active mechanical equipment such as check valves. For some components (valves with extended structures, etc.), static deflection test programs are utilized in combination with analysis to demonstrate operability.

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 required conditions. For active valves, the present operability programs confirm 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 Section 3.9N.3.2 provides adequate assurance through the procedures employed that active valves will perform their required safety function under all required conditions.

8. The Westinghouse methods for combining 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 testing. The seismic and operability qualification programs implemented for Millstone 3 provide adequate assurance of proper equipment performance under all required conditions.

11. Westinghouse qualification documentation is maintained for the 40-year design life in engineering files at Westinghouse. These records are filed and stored in accordance with 10 CFR 50 Appendix B and Regulatory Guide 1.88 as defined in FSAR Chapter 17.

Seismic information 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 and7.

13. The test results for Westinghouse supplied 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 necessary to prepare an additional Seismic Qualification Report.

P 3.11 P TITLE:ENVIRONMENTAL QUALIFICATION OF MECHANICAL AND ELECTRICAL UIPMENT Actual differences between FSAR and SRP (BOP Scope)

1. Millstone 3 FSAR does not address mechanical 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.

Justification for differences from SRP

1. Preparation and submittal of information pertaining to environmental qualification of mechanical equipment is pending NRC rulemaking.

2. A summary comparison of NUREG-0588 will be provided with the Electrical Equipment Qualification Data Packages as a separate report.

P 3.11 P TITLE:ENVIRONMENTAL QUALIFICATION OF MECHANICAL AND ELECTRICAL UIPMENT Actual differences between FSAR and SRP (NSSS Scope)

1. SRP 3.11, Paragraph I.1, states that all mechanical and 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.

Justification for differences from SRP

1. The information requested in the SRP is located in different parts of the FSAR as listed below:

a. Safety Related Mechanical and Electrical Systems - FSAR Chapters 6 and7

b. Active Pumps and Valves - FSAR Tables 3.9N-11 and 3.9N-12

c. Class 1E Generic Components - FSAR Table 3.11N-1

d. Plant-Specific Class 1E Components - NUREG-0588 study for Millstone 3

e. Other safety related mechanical components - FSAR Sections 3.7N and 3.9N

2. 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 components for use under adverse environments.

This program for mechanical equipment 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.

P 4.5.1 P TITLE: CONTROL ROD DRIVE STRUCTURAL MATERIALS Actual difference between FSAR and SRP SRP 4.5.1, Paragraph II.4, addresses on site cleaning requirements which are not directly referenced in FSAR Section 4.5.1.

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.

P 5.2.1.1 P TITLE: COMPLIANCE WITH THE CODES AND STANDARDS RULE, 10 CFR 50.55a Actual differences between FSAR and SRP

1. The loop bypass valves are designed and 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.

Justification for differences from SRP

1. Updating the loop bypass valves to a later ASME code addendum would require additional cost and administrative cost burden without a compensating increase in the level of quality or safety. In addition, the actual hardware configuration would not be changed by upgrading to a later code addendum.

2. Components are classified commensurate 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.

P 5.2.5 P TITLE: REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION Actual difference between FSAR and SRP

1. FSAR Section 5.2.5 does not address the frequency of testing of the unidentified leakage sump system as required in SRP 5.2.5, Paragraph III.7.

Justification for difference from SRP

1. Testing requirements for the unidentified leakage sump system, will be addressed in the Technical Specifications.

P 5.3.1 P TITLE: REACTOR VESSEL MATERIALS 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 required in SRP 5.3.1, Paragraph II.6.c(3).

Justification for difference from SRP 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.

P 5.4.1.1 P TITLE: PUMP FLYWHEEL INTEGRITY (PWR)

Actual difference between FSAR and SRP (NSSS Scope)

SRP 5.4.1.1, Paragraph II.2.b, states that pump flywheel fracture toughness 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.

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 the 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.

P 5.4.2.1 P TITLE: STEAM GENERATOR MATERIALS 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, Paragraph II.2) discusses a clean metal conditions prior to startup. FSAR Section 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 support plates should be provided. This is not discussed in FSAR Section 5.4.2.1.

Justification for differences from SRP

1. 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 analysis are contained in the Millstone 3 Chemistry Manual.

2. A small amount of corrosion product (oxide) retards further run-away corrosion by acting as a barrier to continued corrosion attack. A metal clean condition is therefore unduly restrictive and no real benefit is derived 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 top 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 great degree on the support plates because of support plate flow slots and, in the case of Model F steam generators, quatrefoil tube support plate holes. These 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 selectively 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 the blowdown system. Any additional sludge which is not removed from the tubesheet by blowdown 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.

P 6.2.1 P TITLE: CONTAINMENT FUNCTIONAL DESIGN Actual differences between FSAR and SRP (BOP Scope)

1. Actual margin of external pressure analysis as required in SRP 6.2.1.1A, Paragraph II.f, is not specifically addressed in FSAR Section 6.2.1.1.

2. NUREG-0609 was not addressed in FSAR Section 6.2.1.2 as required in SRP 6.2.1.2, Paragraph II.B.2.

Justification for differences from SRP

1. Conservatism of the analysis provides for margin.

2. NUREG-0609 was issued subsequent to the analysis performed for Millstone 3, which used the NUREG-75/087 SRP criteria. FSAR Section 6.2.1.2 describes the methods used to perform the analysis.

P 6.2.1 P TITLE: CONTAINMENT FUNCTIONAL DESIGN Actual differences between FSAR and SRP (NSSS Scope)

1. BTP 6-1, Section B.3b, recommends conservative 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 Millstone 3 containment design.

Justification for differences from SRP

1. It has been determined that Westinghouse 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 release 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 Section 6.2.7. The current FSAR model is under review by the NRC staff.

P 6.2.2 P TITLE: CONTAINMENT HEAT REMOVAL SYSTEMS Actual differences between FSAR and SRP

1. SRP 6.5.2, Paragraph II.1.a, requires automatic switchover to recirculation 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 requirements for Heat removal only spray steam.

Otherwise, SRP 6.5.2 (Fission Product Removal) would not apply to Millstone 3.

Justification for differences from SRP

1. 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 are available to the operator so he can make the correct decision at the correct time. Description of operator action considerations during the switchover from the injection phase to recirculation phase is discussed in FSAR Section 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.

P 6.2.3 P TITLE: SECONDARY CONTAINMENT FUNCTIONAL DESIGN Actual difference between FSAR and SRP FSAR 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.

Justification for difference from SRP Refer to the Applicant's response to the NRC Acceptance Review Request Number 480.4.

P 6.2.5 P TITLE: COMBUSTIBLE GAS CONTROL IN CONTAINMENT Actual differences between FSAR and SRP

1. SRP 6.2.5, Paragraph II.3, requires a plant 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.

Justification for differences from SRP

1. FSAR Section 6.2.5.3 references the analyses of Surry Power Station, Units 1 and 2 which have a similar containment 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.

P 6.3 P TITLE: EMERGENCY CORE COOLING SYSTEM Actual difference between FSAR and SRP No reference is made in FSAR Section 6.3 to studies which demonstrate that nonsafety grade interactions cannot exist which could degrade the performance of the ECCS or its supporting systems as specified in SRP 6.3, Subsection II.

Justification for difference from SRP The 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.

P 6.4 P TITLE: CONTROL ROOM HABITABILITY SYSTEMS Actual difference between FSAR and SRP The chlorine detectors are not Seismic Category I nor Electrical Class 1E (IEEE 323-1974 qualified) as required by SRP 6.4, Paragraph II.5.b. They are redundant and classified as non-seismic.

Justification for difference from SRP The redundant chlorine detectors are located in a non-harsh environment (i.e., control equipment room, 50-104°F, 10-60 percent RH, 1.1 x 102 Rads). In the event of detector failure, the control room envelope is automatically isolated.

P 6.4 APPENDIX A P TITLE:ACCEPTANCE CRITERIA FOR VALVE OR DAMPER REPAIR ALTERNATIVE Actual differences between FSAR and SRP

1. The air inlet ducting is isolated by two low leakage air operated butterfly valves positioned in series. Since the valves are 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, indicates no manual action credit 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 />.

Justification for differences from SRP As 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. Following 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of air bottle pressurization, 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 the control room habitability 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.

P 6.5.1 P TITLE: ESF ATMOSPHERE CLEANUP SYSTEMS Actual differences between FSAR and SRP

1. SRP 6.5.1, Subsection II, lists 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 compliance to Regulatory Guide 1.52.

2. Continuous indication and recording of air flow for individual ESF filtration units is not provided as described in SRP 6.5.1, Subsection II.

3. Flow sensors are not provided for annunciating high air flows through ESF filtration units as described 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 three 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 resistance corresponding to 1.25 times design dirty filter condition. The three noncomplying filter 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 three 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 flow rates to remain constant while the system resistance varies from clean to 1.25 times dirty filter condition. The three noncomplying filter train systems are the SCLRS filter banks, the fuel building filter banks, and the control room emergency ventilation filter trains.

NOTE: The fuel building filter banks are no longer credited as an ESF System in the radiological accident analyses.

Justification for differences from SRP

1. Justifications for the exceptions taken 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 reliable system operation. Periodic surveillance tests ensure that system balancing is adequate 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 automatically start standby units. Sensors to detect high air flow conditions are not necessary since the system is balanced such that the flow is limited to ensure proper performance of the filtration units. Periodic surveillance tests ensure that system flow rates will not exceed unit design parameters.

4. During tests, fans cannot develop a performance greater than for what they were designed. Fan performance requirements were based on Regulatory Guide 1.52 as a design document and the referenced 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 conditions, 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 ANSI N509 do not, and need not, require such controls. The identification of system 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 described below for the three ESF systems.

SLCRS Filter Banks

The SLCRS is a standby ESF system designed to operate following a LOCA. Its function is to drawdown enclosures contiguous to the containment to a minimum negative pressure of 1/4 inch water gauge in 60 seconds after SIS, and maintain negative pressure conditions for a minimum of 30 days following 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 described 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 calculations, the station operating procedures, and the technical specifications. Also, periodic surveillance testing 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 system 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 exhaust 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 ventilation 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 room 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 condition 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.

The control room radiological habitability 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 condition. 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.

P 6.5.2 P TITLE: CONTAINMENT SPRAY AS A FISSION PRODUCT CLEANUP SYSTEM

Actual difference between FSAR and SRP The containment spray system requirements of SRP 6.5.2 are not discussed in FSAR Section 6.5.2.

Justification for difference from SRP No credit for containment spray fission product removal is taken in the Millstone 3 design.

P 7.2 P TITLE: REACTOR TRIP SYSTEM (RTS)

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.

Justification for difference from SRP The sensors are isolated by digital isolators to prevent degrading the reactor trip system.

P 7.5 P TITLE: INFORMATION SYSTEMS IMPORTANT TO SAFETY 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.

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.

P 8.3.1 P TITLE: AC POWER SYSTEMS (ON SITE)

Actual difference between FSAR and SRP NUREG/CR-0660 is not addressed in the FSAR as required by SRP 8.3.1, Paragraph II.4.f.

Justification for difference from SRP NUREG/CR-0660 considerations have been addressed 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.

P 9.1.2 P TITLE: SPENT FUEL STORAGE Actual difference between FSAR and SRP SRP 9.1.2, Paragraph III.2.e, requires an evaluation of lighter load drops at maximum heights. This evaluation has not been performed.

Justification for difference from SRP Electrical interlocks and load paths prevent any load from being carried over the spent fuel pool with the new fuel handling crane. Spent fuel bridge and hoist only carries fuel assemblies at their normal lifting height.

P 9.1.3 P TITLE: SPENT FUEL POOL COOLING AND CLEANUP SYSTEM Actual difference between FSAR and SRP

1. Decay heat removal is based on the DECOR 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 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 />, 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).

Justification for difference from SRP

1. Decay heat removal analysis is based on the DECOR computer code (based on ORIGEN2) and credit for evaporative cooling 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 scenarios

3. 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 result in greater heat removal capacities which permit larger heat loads to be placed in the pool and shorter decay times.

P 9.1.4 P TITLE: LIGHT LOAD HANDLING SYSTEM (RELATED TO REFUELING)

Actual difference between FSAR and SRP SRP 9.1.4, Paragraph III.6, requires an evaluation of lighter load drops at maximum heights. This evaluation has not been performed.

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.

P 9.2.1 P TITLE:STATION SERVICE WATER SYSTEM (NUCLEAR SERVICE COOLING WATER STEM)

Actual difference between FSAR and SRP SRP 9.2.1, Paragraph III.3.d, requires that radiation monitors be located on system discharge, and at components susceptible to the 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 recirculation coolers; however, there is no manual valve in series with the motor operated valve.

Justification for difference from SRP There is a radiation monitor located on the discharge side of each containment recirculation cooler which alarms in the control room on a high 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 operator malfunctions, the containment recirculation coolers may be isolated by closing the isolation valves on the containment recirculation system side of the coolers.

P 9.2.2 P TITLE:REACTOR AUXILIARY COOLING WATER SYSTEMS (COMPONENT COOLING TER AND AUXILIARY COMPONENT COOLING WATER SYSTEM)

Actual difference between FSAR and SRP

1. The Millstone 3 reactor coolant pump has not, at this time, been tested to the 20-minute time requirements as specified in SRP 9.2.2, Paragraph II.3.e.

Justification for difference from SRP

1. A program is underway by Westinghouse to comply with the testing requirements.

P 9.4.1 P TITLE: CONTROL ROOM AREA VENTILATION SYSTEM Actual difference between FSAR and SRP The chlorine detectors are not Seismic Category I nor Electrical Class 1E (IEEE 323-1974 qualified) as required by SRP 9.4.1, Paragraph II.4. They are redundant and classified as non-seismic.

Justification for difference from SRP The redundant chlorine detectors are used and located in a non-harsh environment (i.e., control equipment room, 50-104°F, 10-60 percent RH, 1.1x102 Rads). In the event of detector failure the control room envelope is automatically isolated.

P 9.4.5 P TITLE: ENGINEERED SAFETY FEATURES VENTILATION SYSTEM Actual differences between FSAR and SRP

1. 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 total ventilation system shall have the capacity to detect and control leakage 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 essential portions of the engineered safety features ventilation systems be protected from the effects of tornados. This SRP requirement also applies to the circulating and service water pumphouse and other yard structures ventilation system (FSAR Section 9.4.7) and the hydrogen recombiner building heating, ventilation, and air conditioning system (FSAR Section 9.4.10). Additionally, Regulatory Guide 1.76 requires 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 emergency generator enclosure are protected by tornado dampers.

Justification for differences from SRP

1. 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 equipment 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 changing the filter upon setpoint alarm annunciation will preclude excessive dust accumulation.

Each emergency diesel generator building ventilation intake has 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 recommended 20 feet above grade 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 allowing a direct path to the mechanical rooms emergency ventilation which utilizes outside air for cooling.

3. Only those ventilation systems that are required for habitability or heat removal requirements after a DBA or LOP were considered for installation of tornado dampers.

Safety related equipment whose ventilation systems do not have tornado dampers have been qualified for adverse conditions during short-term operation, as described in Section 3.11B.1. Plant operational procedures will address long-term ventilation outages.

P BTP CMEB 9.5-1 (Section 9.5.1)

P TITLE: GUIDELINES FOR FIRE PROTECTION FOR NUCLEAR POWER PLANTS er to the Fire Protection Evaluation Report, Appendix B, for a comparison of Millstone 3 design to P CMEB 9.5-1 guidelines.

P 9.5.4 P TITLE:EMERGENCY DIESEL ENGINE FUEL OIL STORAGE AND TRANSFER SYSTEM Actual differences between FSAR and SRP

1. SRP 9.5.4, Paragraph II.4.b, requires that each diesel generator be capable of operating continuously for 7 days. Each diesel 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.

Justification for differences from SRP

1. 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 off 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 (it was caused by salt contamination of insulators) in the future. The Applicant also notes that by running two diesel generators at part load (starting 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after a postulated accident occurred) and judiciously realigning one train of ESF loads between two operating diesel generators, individual engine operating time per each fuel oil storage tank can be extended to 5-6 days.

2. 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 sediment 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.

P 9.5.8 P TITLE:EMERGENCY DIESEL ENGINE COMBUSTION AIR INTAKE AND EXHAUST Actual difference between FSAR and SRP SRP 9.5.8, Paragraph III.8, states that a minimum of 20 feet should exist between the bottom of all fresh air intakes and the grade elevation. The actual design is not consistent with this requirement.

The bottom of the intake hoods is at elevation 40 feet-9 inches, which is 16 feet-3 inches above grade.

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.

P 10.2.3 P TITLE: TURBINE DISK INTEGRITY Actual difference between FSAR and SRP The 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.

Justification for difference from SRP GE considers data and calculations requested by the SRP to be proprietary information which, if requested, can be made available to the NRC under the provisions of 1 0CFR 2.790.

P 10.3 P TITLE: MAIN STEAM SUPPLY SYSTEM (MSSS)

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.

Justification for difference from SRP All flow paths are shown on the appropriate P&IDs.

P 11.5 P TITLE:PROCESS AND EFFLUENT RADIOLOGICAL MONITORING INSTRUMENTATION D SAMPLING SYSTEMS Actual differences between FSAR and SRP

1. SRP 11.5, Table 1, Item 6, requires an automatic control feature which automatically terminates effluents of the 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.

Justification for differences from SRP

1. During fuel handling activities, the fuel building ventilation is processed by the fuel building filtration units. Accident analysis indicates that the filters prevent the release of excessive amounts of radioactive effluent.

2. The spent fuel pool cooling and purification 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. Safety evaluations described in FSAR Section 9.1.3 show this to be adequate.

3. Monitoring of the steam generator blowdown system is provided by the reactor plant sampling system radiation monitor, 3SSR-RE08. An evaluation of the accident scenario for a steam 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.

The main steamline monitors would identify which steam generator is affected and operator action would close valves to prevent release of steam generator blowdown effluents.

P 12.2 P TITLE: RADIATION SOURCES Actual difference between FSAR and SRP

SRP 12.2, Paragraph I.2, requires tabulation of the calculated concentrations of radioactive material, by nuclide, expected during normal operation, anticipated operational occurrences, and accident conditions for equipment cubicles, corridors, and operating areas normally occupied by operating personnel. FSAR Section 12.2 does not tabulate the calculated concentrations of radioactive material expected during accident conditions.

Justification for difference from SRP During accident conditions, local surveys and measurements will be performed as required and exposures will be limited to the requirements of NUREG-0737.

P 13.5.2 P TITLE: OPERATING AND MAINTENANCE PROCEDURES Actual difference between FSAR and SRP SRP 13.5.2, Paragraph II.C.2, references Section 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.

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 described in the FSAR. Since the requirement for functional-based Emergency Operating Procedures is already explicitly addressed, and the approach is to comply with NRC approved procedure guidelines, little will be gained by committing to a partial standard which is not addressed within existing regulatory guides.

P 14.2 P TITLE: INITIAL PLANT TEST PROGRAMS - FSAR Actual difference between FSAR and SRP SRP 14.2, Paragraph II.4, requires the Applicant 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.

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 design is the application of the experience to the safety, reliability, and cost of Unit 3.

P 15.4.6 P TITLE:CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT RESULTS A DECREASE IN THE REACTOR COOLANT Actual difference between FSAR and SRP FSAR Section 15.4.6 does not address this accident scenario.

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 boron 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 Acceptance Review Request Number 440.8.

P 15.4.8 P TITLE: SPECTRUM OF ROD EJECTION ACCIDENTS (PWR)

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 faulted condition as stated in ANSI N18.2. Faulted condition stress limits are applied for this accident.

Justification for difference from SRP System overpressurization due to a rod ejection transient was evaluated in WCAP-7588, Revision 1-A, and received NRC acceptance in the Topical Report Evaluation.

P 15.6.5 P TITLE: LOCA INSIDE CONTAINMENT 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 and II.K.3.31, as required by SRP 15.6.5, Paragraph II.3.

Justification for difference from SRP The 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 analysis model to address the NRC concerns mentioned in II.K.3.30 and II.K.3.31 (see FSAR Section 1.10).

P 15.7.3 P TITLE:POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING TANK ILURES Actual difference between FSAR and SRP The FSAR has analyzed postulated tank failure using 1 percent fuel defects, while SRP 15.7.3, Paragraph III.1.a, uses 0.12 percent fuel defects.

Justification for difference from SRP FSAR analysis is more conservative.

le 1.10-1 presents the MNPS-3 positions on the PWR applicable items from the US Nuclear ulatory Commission's post-TMI action plan requirements for applicants for an operating nse, NUREG-0737, Enclosure 2, dated November 1980.

FSAR/TS Item and Title Position Reference I.A.1.1 Shift Technical Advisor MPS-3 meets the provisions of the Commission's Policy Statement TS 6.2.2 & 6.

for providing engineering expertise on shift (STA).

I.A.1.2 Shift Supervisor Administrative Duties MPS-3 meets the requirements of this item. 13.1.2.2 I.A.1.3 Shift Manning MPS-3 meets the requirements of this item. TS 6.2.2 I.A.2.1 Immediate Upgrade of RO and SRO MPS-3 meets the requirements of this item. 13.2 Training and Qualifications I.A.2.3Administration of Training Programs MPS-3 meets the requirements of this item. 13.2 I.A.3.1 Revise Scope and Criteria for This item is not applicable to MPS-3.

• Licensing Examinations I.B.1.2 Evaluation of Organization and MPS-3 meets the requirements of this item. 13.4.4 & TS Management 6.2.3 I.C.1 Short Term Accident Analysis and MPS-3 Emergency Operating Procedures are based on 13.5.2.1 Procedures Revision Westinghouse Owners Group emergency procedure guidelines which are approved by the NRC.

I.C.2 Shift and Relief Turnover Procedures MPS-3 meets the requirements of this item. 13.5.1.3 I.C.3 Shift Supervisor Responsibility MPS-3 meets the requirements of this item. 13.1.2 I.C.4 Control Room Access MPS-3 meets the requirements of this item. 13.5.1 I.C.5 Procedures for Feedback of Operating MPS-3 meets the requirements of this item. 13.3.5 Experience I.C.6 Procedures for Verification of Correct MPS-3 meets the requirements of this item. 13.5 Performance of Operating Activities

FSAR/TS Item and Title Position Reference I.C.7 NSSS Vendor Review of Procedures MPS-3 emergency operating procedures are based on NRC 13.5.2 approved Westinghouse Emergency Response Guidelines and therefore eliminates the requirements for additional NSSS vendor review of emergency operating procedures.

I.C.8 Pilot Monitoring of Selected Emergency MPS-3 emergency operating procedures are based on NRC 13.5.2 Procedures for NTOLS approved Westinghouse Emergency Response Guidelines and therefore eliminates the requirement for pilot monitoring of selected emergency procedures for near term operating license applicants.

I.D.1 Control Room Design Review A control room design review was performed for MPS-3 to meet the 18 requirements of this item. Modifications are reviewed in accordance with the Design Control Manual.

I.D.2 Plant Safety Parameter Display Console MPS-3 meets the requirements of this item. 7.5.1 I.G.1 Training during Low-Power Testing This item was completed prior to low power testing.

• II.B.1Reactor Coolant System Vents Safety grade reactor vessel and pressurizer venting capability is 5.4.15, 7.5 provided in the MPS-3 design.

II.B.2 Plant Shielding The MPS-3 plant shielding design is outlined in Chapter 12. 12.3 II.B.3 Post-Accident Sampling Not applicable to MPS-3 (Reference Technical Specifications

• Amendment number 201)

II.B.4 Training for Mitigating Core Damage MPS-3 has implemented a training program utilizing the INPO 13.2 guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage as the basis for the program.

FSAR/TS Item and Title Position Reference II.D.1 Testing Requirements for Reactor MPS-3 meets the requirements of this item as noted below with the 5.4.13 Coolant System Relief and Safety Valves exception with the exception of ATWS testing. MPS-3 meets the requirement of this item for the safety 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 valve functionability is demonstrated by application of test data and analyses.

II.D.3 Relief and Safety Valve Position The pressurizer PORVs have a reliable direct position indication in 5.4.13, 7.5 Indication the control room.

II.E.1.1 Auxiliary Feedwater System An auxiliary feedwater system reliability analysis has been 10.4.9 Evaluation performed as specified. Incorporation of cavitating ventures has solved any possibility of excessive flowrate; therefore, MPS-3 meets the requirements of this item.

FSAR/TS Item and Title Position Reference II.E.1.2 Auxiliary Feedwater System The present design of the MPS-3 auxiliary feedwater system 10.4.9, 7.3 Automatic. Initiation and Flow Indication incorporates the requirements to have reliable automatic initiation and flow indications in accordance with IEEE 279-1971. 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 may be in normal operation, feeding the steam generators to maintain Steam Generator inventory level. In this mode, there is no automatic function associated with the MDAFW pump flow control valves and operator action is required to open the MDAFW system control valves. 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. Manual initiation may require operator action to realign the motor-driven auxiliary feedwater pump suction source in this mode of operation.

II.E.3.1 Emergency Power Supply for The emergency power supply for pressurizer heaters meets the 8.3.1, 5.4.10.

Pressurizer Heaters requirements of this item.

II.E.4.1 Dedicated Hydrogen Penetrations The MPS-3 design includes redundant hydrogen recombiners and 6.2.4, 6.2.5 (Containment Design) analyzers outside the containment. Penetrations 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.

II.E.4.2 Containment Isolation Dependability MPS-3 meets the requirements of this item. 6.2.4, 7.3.1

FSAR/TS Item and Title Position Reference II.F.1 Additional Accident Monitoring MPS-3 meets the requirements of this item. Hydrogen monitors 11.5, 7.5, 6.2.1, Instrumentation meet the requirements of 10 CFR 50.44 using guidance of 6.2.5, TS 3.3.3.

Regulatory Guide 1.7. TS 3.3.3.6, TS3.6.4.1 II.F.2 Identification of and Recovery From MPS-3 utilizes a subcooled margin monitor system, and a core exit 4.4.6.5 Conditions Leading to Inadequate Core thermocouple system and Combustion Engineering heated junction Cooling thermocouple system to meet the requirements of this item.

II.G.1 Power Supplies for Pressurizer Relief Emergency power for pressurizer equipment meets the requirements 5.4.13 Valves, Block Valves, and Level Indicators for this item.

II.K.1.5 Review ESF Valves MPS-3 meets the requirements of this item. 13.5.2 II.K.1.10 Operability Status MPS-3 meets the requirements of this item 13.5.1 II.K.1.17 Trip per Low-Level B/S An MPS-3 pressurizer low pressure signal initiates both a reactor 7.2, 7.3 trip and the start of safety injection. Pressurizer low-level trips are not utilized on MPS-3.

II.K.2.13 Thermal Mechanical Report MPS-3 meets the requirements of this item. 5.2.3

FSAR/TS Item and Title Position Reference II.K.2.17 Potential for Voiding in the Reactor MPS-3 meets the requirements of this item

• Coolant System During Transients Westinghouse has performed a study which addresses the potential for void formation in Westinghouse designed nuclear steam supply systems during natural circulation 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 Evaluation 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 circulation cooldown/depressurization transients, and specifies those conditions under which upper head voiding may occur. These Westinghouse Owners Group generic guidelines have been submitted to the NRC (Jurgensen 1981e). The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant specific consideration) has been utilized in the implementation of MPS-3 plant specific operating procedures.

II.K.2.19 Sequential Auxiliary Feedwater Not applicable to MPS-3. The NRC has completed a generic review

• Flow Analysis 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).

FSAR/TS Item and Title Position Reference II.K.3.1 Installation and Testing of Automatic The addition of an automatic isolation system for the PORVs will

• Power-Operated Relief Valve Isolation System 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 automatic closure of the block valve may inhibit operator detection of a stuck open PORV.

II.K.3.2 Report on Overall Safety Effects of MPS-3 meets the requirements of this item.

• Power-Operated Relief Valve Isolation System 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 The licensee is responsible for ensuring that any failure of a PORV TS 6.9.1 Challenges 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.

FSAR/TS Item and Title Position Reference II.K.3.5 Automatic Trip of Reactor Coolant Automatic trip of reactor coolant pumps during LOCA is not

• Pumps During Loss-of Coolant Accident provided. Westinghouse has performed an analysis of delayed reactor coolant pump trip during small break LOCAs. This 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 information related to Generic Letter 85-12 concerning implementation of TMI Action 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 This item is applicable to B&W plants only and therefore does not

• Probability apply to MPS-3.

II.K.3.9 Proportional Integral Derivative The MPS-3 design does not include a proportional integral 7.7.1.5 Controller Modification derivative (PID) controller in the power-operated relief valve control circuit (see Figures 7.7-4 and 7.2-1, Sheet 11).

FSAR/TS Item and Title Position Reference II.K.3.10 Proposed Anticipatory Trip The MPS-3 design incorporates this trip modification. 10.4.4.1, 7.2.

Modification The NRC has raised the question 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 initial power level of 53 percent. All operating parameters (RCS and secondary temperature, pressure and flow) are at initial values without uncertainties, 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 operational 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 would be 1.124 psia which remains below the nominal secondary relief valve actuation setpoint of 1.140 psia.

II.K.3.11 Justification Use of Certain PORVs The PORVs used in the MPS-3 design are pilot-operated relief

• valves supplied by Garrett.

FSAR/TS Item and Title Position Reference II.K.3.12 Confirm Anticipatory Trip The MPS-3 design includes an anticipatory reactor trip on turbine 7.2.1.1.2 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).

II.K.3.17 Report on Outages of Emergency The ECCS Component Reliability and Tracking Program

• Core Cooling Systems Licensee Report and commitment has been satisfied. The required reliability information Proposed Technical Specification Changes required by NUREG 0737 on ECCS components involved in outages during the first five years of commercial operation has been compiled, 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 currently 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 During normal operation, seal injection flow from the chemical and

• Power on Pump Seals 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 automatically started and either seal injection flow or component cooling water to the 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 off 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.

FSAR/TS Item and Title Position Reference II.K.3.30 Revised Small Break MPS-3 meets the requirements of this item. 15.6.5 Loss-of-Coolant Accident Methods to Show Compliance With 10 CFR Part 50, App. K II.K.3.31 Plant Specific Calculations to Show MPS-3 meets the requirements of this item. 15.6.5 & TS Compliance With 10 CFR Part 50.46 6.9.1.6 III.A.1.1 Emergency Preparedness, Short Term Existing plans for Millstone site apply. 13.3 III.A.1.2 Upgrade Emergency Support The emergency support facilities identified in the Emergency Plan 13.3 Facilities are functional.

II.A.2 Emergency Preparedness Existing plans for the Millstone site apply. 13.3 III.D.1.1 Primary Coolant Sources Outside MPS-3 has implemented a program to reduce leakage from systems 5.2.5 & TS Containment outside containment that would or could contain highly radioactive 6.8.4.a fluids during a serious transient or accident to as-low-as practical levels.

III.D.3.3 Inplant I Radiation Monitoring Continuous air monitors with direct readout and alarm capabilities 11.5, 12.3, T are located in the MPS-3 control room and in the Technical Support 6.8.4.b and Emergency Operations Centers.

III.D.3.4 Control Room Habitability The requirements of this item have been addressed. 6.4, 2.2.3, 9.

FSAR/TS Item and Title Position Reference NOTES:

• Statement stands alone, no FSAR Section reference.

References for Table 1.10-1:

Anderson, T. M. (Westinghouse) September 26, 1980. Letter (NS-TMA-2318) to Eisenhut, D. G. (NRC).

Jurgensen, R. W. (Chairman, Westinghouse Owners Group) March 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 Trip During Small Loss-of-Coolant Accidents for Westinghouse Nuclear Steam Supply System.

WCAP-9804, February 1981. Probabilistic Analysis and Operational Data in Response to NUREG-0737 Item II.K.3.2 for Westinghouse NSSS Plants.

Jaffe, D. H. (NRC), March 29, 1989 Letter to Mroczka, E. J. (NNECO).

following is a list of material incorporated by reference in the Final Safety Analysis ort (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.

nformation incorporated by reference into the Final Safety Analysis Report is subject to the ate and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59 ss separate NRC change control requirements apply (e.g., 10 CFR 50.54(a)).