ML23129A269

1 to Updated Safety Analysis Report,Section I, Introduction and Summary

ML23129A269
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Site:	Cooper
Issue date:	04/20/2023
From:	Nebraska Public Power District (NPPD)
To:	Office of Nuclear Reactor Regulation
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USAR I - INTRODUCTION AND

SUMMARY

PAGE 1.0 PROJECT IDENTIFICATION I-1-1 1.1 Identification and Qualification of Contractors I-1-1

1. 1. 1 Applicant I-1-1 1.1.2 Engineer-Const ructor I-1-2 1.1.3 Nuclear Steam Supply System Supplier I-1-2 1.1.4 Turbine-Genera tor Supplier I-1-3 2.0 DEFINITIONS I-2-1 3.0 METHODS OF TECHNICAL PRESENTATION I-3-1 3.1 Purpose I-3-1 3.2 Radioactive Material Barrier Concept I-3-1 3.3 Organization of Contents I-3-1 3.4 Format Organization of Chapters I-3-1 4.0 PRINCIPAL DESIGN CRITERIA I-4-1 4.1 Principal Design Criteria - Functional Classification I-4-1 4.1.1 General Criteria I-4-3

4. 1. 2 Power Generation Design Criteria, ( Planned Operation) I-4-3 4 .1. 3 Power Generation Design Criteria, (Abnormal Operational Transients) I-4-4 4 .1. 4 Nuclear Safety Design Criteria, (Planned Operation) I-4-4

4. 1. 5 Nuclear Safety Design Criteria, (Abnormal Operational Transients) I-4-4

4. 1. 6 Nuclear Safety Design Criteria, (Accidents) I-4-5 4 .1. 7 Nuclear Safety Design Criteria, (Special Event) I-4-7 4.2 Principal Design Criteria, System-By-Syste m I-4-7 4.2.1 General Criteria I-4-7 4.2.2 Nuclear System Criteria I-4-7 4.2.3 Power Conversion Systems Criteria I-4-8

4. 2. 4 Electrical Power System Criteria I-4-8 4.2.5 Radioactive Waste Disposal Criteria I-4-9 4.2.6 Nuclear Safety Systems and Engineered Safety Features Criteria I-4-9 4.2.6.1 General I-4-9 4.2.6.2 Containment and Isolation Criteria I-4-9 4.2.6.3 Emergency Core Cooling Systems (ECCS)

Criteria I-4-10 4.2.6.4 Standby Power Criteria I-4-10 4.2.7 Reactivity Control Criteria I-4-10 4.2.8 Process Control Systems Criteria I-4-11 4.2.8.1 Nuclear Steam Supply System Process Control Criteria I-4-11

4. 2. 8. 2 Power Conversion Systems Process Control Criteria I-4-11 4.2.8.3 Electrical Power System Process Control Criteria I-4-11 4.2.9 Auxiliary System Criteria I-4-12 4.2.10 Shielding and Access Control Criteria I-4-12 4.2.11 Structural Loading Criteria I-4-12 i-1-1 02/06/01

USAR I - INTRODUCTION AND

SUMMARY

(Cont'd)

PAGE 5.0 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS I-5-1 5.1 Quality Group Classificati ons I-5-1 5.2 Loading Classificat ion I-5-1 6.0 COMPARISON OF PRINCIPAL DESIGN CHARACTERISTICS I-6-1 6.1 Nuclear System Design Characteris tics I-6-1 6.2 Power Conversion Systems Design Characteris tics I-6-1 6.3 Electrical Power Systems Design Characteris tics I-6-1 6.4 Containment Design Characteris tics I-6-1 6.5 Structural Design Characteris tics I-6-1 7.0 STATION RESEARCH DEVELOPMENT AND FURTHER INFORMATION; REQUIREMENTS AND RESOLUTIONS

SUMMARY

I-7-1

8.0 REFERENCES

FOR CHAPTER I I-8-1 i-1-2 02/06/01

USAR LIST OF FIGURES (At End of Section I)

Figure No. Title I-1-1 Reactor Heat Balance - Rated I-2-1 Relationship Between Safety Action and Protective Action I-2-2 Relationship Between Protective Functions and Protective Actions I-2-3 Relationship Between Different Types of Systems Actions and Objectives I-5-1 Relationship of Various Categories of Systems i-2-1 02/06/01

USAR LIST OF TABLES Table No. Title I-3-1 Documents Incorporate d Into the CNS USAR by I-3-4 Reference I-4-1 Unacceptable Safety Results for Station Event I-4-2 Categories I-6-1 Comparison of Nuclear System Design I-6-2 Characteris tics I-6-2 Comparison of Power Conversion System Design I-6-10 Characteris tics I-6-3 Comparison of Electrical Power Systems Design I-6-11 Characteris tics I-6-4 Comparison of Containment Design I-6-12 Charaterist ics I-6-5 Comparison of Structural Design I-6-14 Characteris tics I-7-1 Cooper Nuclear Station Topical Reports I-7-2 Submitted to the AEC in Support of Docket I-7-2 Cooper Nuclear Station ACRS Concerns - I-7-5 Resolution I-7-3 Cooper Nuclear Station AEC Staff Concerns - I-7-7 Resolutions I-7-4 AEC-ACRS Concerns on Other Related Dockets - I-7-9 Cooper Capability for Resolution i-3-1 02/06/01

USAR I - INTRODUCTION AND

SUMMARY

1.0 PROJECT IDENTIFICATION The original Final Safety Analysis Report (FSAR) was submitted in 1971 in support of the application of the Nebraska Public Power District (formerly known as Consumers Public Power District), for a facility operating license for the Cooper Nuclear Station (CNS), located approximately 21/2 miles south of the town of Brownville, Nemaha County, Nebraska, for power levels up to 2381 MWt under Section 104 (b) of the Atomic Energy Act of 1954, as amended, and the regulations of the Atomic Energy Commission (redesignated as the Nuclear Regulatory Commission after the Energy Reorganization Act of 1974) set forth in Part 50 of Title 10 of the Code of Federal Regulations (10CFR50).

Following Refueling Outage 24, a Measurement Uncertainty Recapture (MUR) power uprate license amendment was approved in accordance with 10CFR50, Appendix K. This allowed thermal power to be increased to 2419 MWt.l 2 l This Updated Safety Analysis Report (USAR) meets the content requirements of 10CFR50.71(e). The USAR is updated and reported to the NRC in accordance with the requirements of 10CFR50.71(e).

Based on the design capability of the upgraded HP Turbine, LP Turbines, MUR, and the Main Generator electrical distribution and support systems, the station at rated power is designed to provide a gross electrical output of 835. 5 MWe and a net electrical output of approximately 815 MWe.

Historical dates of interest relating to CNS are as follows:

Publicly Announced August, 1966 Contract Awarded (NSSS) April, 1967 Construction Permit Application July, 1967 Construction Permit Issued June, 1968 Operating License Application February, 1971 Operating License Issued January 18, 1974 Fuel Loading January 22, 1974 Initial Criticality February 21, 1974 Commercial Operation July 1, 1974 MUR Power Uprate Approved June 30, 2008 The Nebraska Public Power District (NPPD) owns and operates the station. The station was designed by Burns and Roe, Inc. General Electric Company (GE) furnished the Nuclear Steam Supply System (NSSS) and Westinghouse Electric Corporation furnished the Turbine Generator set.

CNS uses a single cycle, forced circulation, boiling water reactor (GE BWR-4) substantially similar at the time of the license application to the TVA Browns Ferry Nuclear Power Station (AEC Docket 50 259/260). A heat balance showing the major parameters of Nuclear Steam Supply System for the design power condition is shown in Figure I-1-1. A plot plan for the CNS is shown on Burns and Roe Drawing 4003.

1.1 Identification and Qualification of Contractors This section contains historical information as indicated by the italicized text. USAR Section I-3. 4 provides a more detailed discussion of historical information. The information being presented in this section as historical has been preserved as it was originally submitted to the Atomic Energy Commission in the CNS FSAR.

1.1.1 Applicant As Owner, the applicant, Nebraska Public Power District, has engaged the below noted contractors to engineer and construct the Cooper Nuclear Station. However, irrespective ofthe contractual responsibilities discussed below, Nebraska Public Power District is the sole applicant for the facility license, and as owner and applicant is responsible for the design, construction, and operation of the Cooper Nuclear Station.

I-1-1 11/29/16

USAR "Effective January 1, 1970, Consumers Public Power District changed its name to Nebraska Public Power District and properties of Platte Valley Public Power and Irrigation District and Nebraska Public Power System were merged into Nebraska Public Power District. This change of name was recognized by an Order Changing Name of Company in this Docket No. 50-298 dated January 14, 1970."

Nebraska Public Power District (NPPD) is a public corporation and political subdivision of the State ofNebraska engaging in generation, transmission, distribution, and sale of electric energy.

The newly consolidated utility, NPPD, serves retail and wholesale electric customers and irrigation customers in 85 ofNebraska's 93 counties.

The Platte Valley Public Power and Irrigation District and Nebraska Public Power System were previously engaged mainly in electric transmission and distribution and lightly in hydroelectric and fossil-fueled electric generation facilities. These former utilities had no previous background in nuclear-fueled electric generation.

Prior to change of its name, Consumers Public Power District (CPPD) was a public corporation and political subdivision of the State of Nebraska, organized in 1939 to engage in generation, transmission, distribution, and sale ofelectric energy. CPPD had been actively engaged in the application ofnuclear energy to electric generation since 1957 when CPPD agreed to operate the Hallam Nuclear Power Facility for the Atomic Energy Commission.

CPPD successfully operated that facility until the decision was made to dismantle the nuclear portion of the plant.

CPPD subsequently performed the work required to retire the facility for the AEC. It is principally these former CPPD personnel with nuclear experience that are involved with the construction and will operate the Cooper Nuclear Station.

With its training and experience in the electric industry in general and in nuclear electric generation in particular, Nebraska Public Power District is well qualified to design, construct and operate the Cooper Nuclear Station. In addition, NP PD retained General Electric Company to provide project management services.

1.1.2 Engineer-Constructor The Nebraska Public Power District has retained Burns and Roe, Inc., to provide engineering and construction management services for the design and construction of the station, integrating the items furnished by General Electric Company and Westinghouse Electric Corporation with complete balance ofplant items. Burns and Roe, Inc., is also responsible for all procurement specifications. Burns and Roe, Inc., has been continuously engaged in construction or engineering activities since 1935.

Burns and Roe was founded in 1932. It was incorporated in 1935 as Burns and Roe, Inc. Burns and Roe, Inc., has been active in the fields ofpower generation and distribution, sea water and brackish water desalination, waste water renovation, environmental quality control, chemical and industry processing, and laboratory and testing facilities. Between 1955 and 1971, when the original FSAR was filed, Burns and Roe, Inc., completed or is currently providing engineering, design and/or construction management services for over 50 thermal power generating units, representing more than 11,400,000 kilowatts of new generating capacity of which more than 4,800,000 kilowatts is nuclear.

1.1.3 Nuclear Steam Supply System Supplier General Electric Company has been awarded the contract to design, fabricate, and deliver the nuclear steam supply system and nuclear fuel for the Cooper Nuclear Station as well as to provide technical direction for installation and startup of this equipment. The General Electric Company has been engaged in the development, design, construction, and operation of boiling water reactors since 1955. Operating boiling water reactors designed and built by General Electric include Oyster Creek Unit I, Vallecitos Boiling Water Reactor, Nine Mile Point Unit 1, Dresden Units 1 and 2, Humboldt Bay, Big Rock Point, KRB (Germany), Tarapur (India), KAIIL (Germany),

JPDR (Japan), Tsuruga (Japan), Millstone Unit 1, Monticello Unit 1, and SENN (Italy). Among the domestic reactors of General Electric design now under construction are Brown's Ferry Units I, 2, and 3, Dresden Unit 3, Quad Cities I-1-2 02/06/01

USAR Units 1 and 2, Vermont Yankee Unit 1, and Peach Bottom Units 2 and 3. Thus, General Electric has substantial experience, knowledge, and capability to design, manufacture, andfurnish technical assistance for the installation and startup of the reactor.

1.1.4 Turbine-Generator Supplier The District has awarded a contract to Westinghouse Electric Corporation to design, fabricate, and deliver the turbine generator for Cooper Nuclear Station as well as to provide technical assistance for installation and startup of this equipment. More recently, Westinghouse Electric Corporation was awarded a contract by the District to erect this turbine generator unit. This now gives Westinghouse an overall responsibility for the unit. Westinghouse Electric Corporation has a long history in the application of turbine generators in nuclear power stations going back to the inception of commercial electrical power production utilizing nuclear facilities. Westinghouse furnished the turbine generator unit for Shippingport No. 1. This unit was shipped in 1956. Westinghouse also furnished the turbine generator unit for Yankee Atomic Power Company Rowe No. 1. This unit was shipped in 1959. Another unit currently in operation is at the Southern California Edison Company, San Onofre No. 1. The Connecticut Yankee, Haddam Neck No. 1 unit went critical July, 1967. Between 1967 and 1972 Westinghouse has firm orders to ship 28 turbine generator units in addition to the unit for Nebraska Public Power District for application to nuclear cycles. Inlet pressures ofthese units vary between 750 psig and 1000 psig and temperatures vary from saturation to approximately 40 ° superheat. The ratings ofthese units range from 500,000 kW to 1,090,000 kW Westinghouse is therefore competent to design,fabricate, deliver, and erect the turbine-generator set and to provide technical assistance for the startup of this equipment.

I-1-3 02/06/01

USAR 2.0 DEFINITIONS The following definitions apply to the terms used in the USAR. See CNS Technical Specifications Section 1.1 for additional definitions.

Abnormal Occurrence - Abnormal occurrence refers to the occurrence of any station condition that:

(1) Exceeds an Allowable Value as established in the Technical Specifications, or (2) Violates a Limiting Condition for Operation as established in the Technical Specifications, or (3) Causes any abnormal operational transient, or (4) Causes any uncontrolled or unplanned release of radioactive material from the site.

Abnormal Operational Transient - An abnormal operational transient includes the events following a single equipment malfunction or a single operator error that is reasonably expected during the course of planned operations and is one of the design basis events. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors initiating the events in this category.

Accident - An accident is a single event, not reasonably expected during the course of station operations, that has been hypothesized for analysis purposes or postulated from unlikely but possible situations, and is one of the design basis events, and that causes or threatens a rupture of a Radioactive Material Barrier. A pipe rupture qualifies as an accident; a fuel cladding defect does not.

Achieving Criticality - See Planned Operation.

Achieving Shutdown - See Planned Operation.

Activated Device - An activated device is a mechanical module in a system used to accomplish an action. An activated device is controlled by an Actuation Device. See Figure VII-1-1.

Active Component - An active component is a device characterized by an expected significant change of state or discernible mechanical motion in response to an imposed design basis load demand upon the system. Examples include switches, relays, valves, pressure switches, turbines, transistors, motors, dampers, pumps, analog meters, etc.

Active Fuel - The portion of a fuel rod that contains fuel pellets (either enriched or unenriched).

Actuation Device An actuation device is an electrical or electromechanical module in a control and instrumentation system controlled by an electrical decision output used to produce mechanical operation of one or more Actuated Devices to accomplish the necessary action. See Figure VII-1-1.

Alteration of the Reactor Core - See Core Alteration.

Alternate Shutdown Capability - Alternate shutdown capability is the control capability provided to safely shutdown the reactor if an event, such as a fire, disables the normal control circuits to the Control Room or causes evacuation of the Control Room.

I-2-1 09/26/05

USAR Anticip ated Transien t Without Scram (ATWS) - Anticipa ted Transien t Without Scram is an Abnormal Operatio nal Transien t followed by the failure of the reactor protecti on system. See 10CFR50 .62.

Availab ility - Availab ility is the probabi lity that a structur e, System, or compone nt (SSC) is capable of performi ng its specifie d function when called upon.

Cold Shutdown Conditio n - The reactor is in MODE 4 or MODE 5 as defined in the Technic al Specific ations.

Compone nt - Compone nts are items from which the system is assemble d (e.g., resistor s, capacito rs, wires, connecto rs, transist ors, switches ,

springs, pumps, valves, piping, heat exchang ers, vessels, etc.).

Cooldown - See Planned Operatio n.

Core Alterati on - See Technic al Specific ations.

Core Fuel to Water Total Power - The core fuel to water total power is the sum of:

(1) The instanta neous integral , over the entire fuel clad outer surface, of the product of heat transfer area incremen t and position dependen t heat flux, and (2) The instanta neous rate of energy depositi on by neutron and gamma reaction s in all the water and core compone nts except fuel rods in the cylindri cal volume defined by the active core height and the inner surface of the core shroud.

Core Operatin g Limits Report (COLR) - See Technica l Specific ations.

Deep Dose Equivale nt (DDE) - Deep dose equivale nt applies to external whole-bo dy exposure . It is the dose equivale nt at the tissue depth of 1 cm (1000 mg/cm 2 ) . See 10CFR20 .

Design Basis Acciden t - A design basis accident is a hypothe sized accident the charact eristics and conseque nces of which are utilized in the design of those systems and compone nts pertinen t to the preserva tion of Radioac tive Materia l Barriers . The potentia l radiatio n exposure s resultin g from a design basis acciden t are greater than any similar accident postulat ed from the same general acciden t assumpt ions and could result in potentia l off site exposure compara ble to the guidelin e exposure of 10CFRl00 , or 10CFR50 .67 (Fuel Handling Acciden t or Loss of Coolant Acciden t).

Design Basis Events - Conditio ns of normal operatio n, includin g anticipa ted operatio nal occurren ces, design basis accident s, external events, natural phenomen a for which the plant must be designed to function to and ensure (1) the integrit y of the reactor coolant pressure boundary (2) the capabil ity to shut down the reactor and maintain it in a safe shutdown conditio n (3) the capabil ity to prevent or mitigate the conseque nces of acciden ts that could result in potentia l offsite exposure s compara ble to the guidelin e exposure of 10CFRl00 , or 10CFR50 .67 (Fuel Handling Acciden t or Loss of Coolant Acciden t).

Design Power - Design power means a steady- state power level of 2486 thermal megawat ts. This is 104.4% of Rated Power (105% of rated steam flow).

I-2-2 02/05/10

USAR Engineered Safety Feature (ESF) - An ESF is a safety-related SSC that performs a safety action necessary to maintain the consequences of postulated accidents within acceptable limits. Also known as Engineered Safeguard.

Environmental Qualification (EQ) - Environmental qualification is the generation and maintenance of evidence to assure that electrical equipment important to safety and within a harsh environment will operate upon demand to meet system performance requirements. See 10CFR50.49.

Essential - Essential functions, structures, systems and components are equivalent to safety-related functions, structures, systems and components.

Fire Safety Analysis - A Fire Safety Analysis is the document required by NFPA 805, Section 2.7.1.2, that includes fire hazards identification and nuclear safety capability assessment, on a fire area basis, for all fire areas that could affect the nuclear safety or radioactive release performance criteria defined in Chapter 1 of that Standard. The CNS Fire Safety Analysis comprises the following calculations:

NEDC 11-084 - Fire Safety Analysis for Fire Area CB-A NEDC 11-085 - Fire Safety Analysis for Fire Area CB-A-1 NEDC 11-086 - Fire Safety Analysis for Fire Area CB-B NEDC 11-087 - Fire Safety Analysis for Fire Area CB-C NEDC 11-088 - Fire Safety Analysis for Fire Area CB-D NEDC 11-089 - Fire Safety Analysis for Fire Area IS-A NEDC 11-090 - Fire Safety Analysis for Fire Area DG-A NEDC 11-091 - Fire Safety Analysis for Fire Area DG-B NEDC 11-092 - Fire Safety Analysis for Fire Area RB-A NEDC 11-093 - Fire Safety Analysis for Fire Area RB-B NEDC 11-094 - Fire Safety Analysis for Fire Area RB-CF NEDC 11-095 - Fire Safety Analysis for Fire Area RB-DI NEDC 11-096 - Fire Safety Analysis for Fire Area RB-E NEDC 11-097 - Fire Safety Analysis for Fire Area RB-FN NEDC 11-098 - Fire Safety Analysis for Fire Area RB-J NEDC 11-099 - Fire Safety Analysis for Fire Area RB-K NEDC 11-100 - Fire Safety Analysis for Fire Area RB-M NEDC 11-101 - Fire Safety Analysis for Fire Area RB-N NEDC 11-102 - Fire Safety Analysis for Fire Area RB-P NEDC 11-103 - Fire Safety Analysis for Fire Area RB-T NEDC 11-104 - Fire Safety Analysis for Fire Area RB-V NEDC 11-105 - Fire Safety Analysis for Fire Area TB-A NEDC 11-106 - Fire Safety Analysis for Fire Area TB-C NEDC 11-107 - Fire Safety Analysis for Fire Area YD NEDC 13-009 - Fire Safety Analysis for Fire Area DW NEDC 14-043 - Fire Safety Analysis for Entire Power Block NEDC 10-080 Fundamental Fire Protection Program and Design Elements (B-1 Table)

NEDC 11-019 - Nuclear Safety Capability Assessment (NSCA)

NEDC 10-062 - NFPA 805 Radioactive Release Review Fuel Damage - Fuel damage is perforation of the fuel cladding which would permit the release of fission products into the reactor coolant.

Fuel Zone Zero - Fuel Zone Zero corresponds to the highest Top of Active Fuel of all fuel types that may be used at CNS, up to a maximum active fuel length of 150". (See also Top of Active Fuel.)

Heatup - See Planned Operation.

Incident - An incident is any event, abnormal operational transient, or accident not considered as part of planned operation.

I-2-3 04/08/15

USAR Incident Detection Circuitry - Incident Detection Circuitry includes those trip systems which are used to sense the occurrence of an incident. Such circuitry is described and evaluated separately where the incident detection circuitry is common to several systems.

Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit a signal related to the plant parameter monitored by that instrument channel. A channel terminates and loses its identity where individual channel outputs are combined in logic.

See Figure VII-1-1.

Locked Open/Closed - Locked Open/Closed means to maintain a component in an established position by administrative controls using a mechanical lock to physically restrain the component to its proper position.

Logic Logic is that array of components in a control and instrumentation channel which combines individual bistable output signals to produce decision outputs. See Figure VII-1-1.

Maximum Anticipated Thermal Output - The maximum anticipated thermal output is the thermal energy output at Design Power.

Module - Any assembly of interconnected components in a control and instrumentation circuit which constitutes an identifiable device, instrument, or piece of equipment.

Notch - A notch is a control rod blade position corresponding to the grooves cut in the CRDM index tube. There are six inches between notches.

Nuclear Safety Operational Analysis - A nuclear safety operational analysis is a systematic identification of the requirements for and the limitations on station operation necessary to satisfy nuclear safety operational criteria.

Nuclear Safety Operational Criteria - A nuclear safety operational criteria is a set of standards which were used to provide input for the proposed CNS Technical Specifications contained in Appendix B of the FSAR.

Nuclear Safety System - A nuclear safety system is a safety system the action of which is required for a safety action that is necessary in response to an abnormal operational transient. See Figure I-2-3.

Nuclear Stearn Supply System (NSSS) - The NSSS generally includes those systems most closely associated with the reactor vessel which are designed to contain or be in communication with the water and steam corning from or going to the reactor core. The NSSS includes the following:

Reactor vessel Vessel internals Reactor core Main Stearn lines from reactor vessel to the isolation valves outside the Primary Containment Neutron Monitoring system Reactor Recirculation system Control Rod Drive system Residual Heat Removal system Reactor Core Isolation Cooling system Emergency Core Cooling systems Reactor Water Cleanup system Reactor Feedwater system piping between the reactor vessel and the first valve outside the Primary Containment Safety and Relief Valve system I-2-4 04/08/15

USAR Operating - Operating means a system, subsystem, train, component, or device is performing its intended function in its required manner.

Operational - The adjective operational, along with its noun and verb forms, is used in reference to the working or functioning of the station, in contrast to the design of the station.

Passive Component - A device characterized by an expected negligible change of state or negligible mechanical motion in response to an imposed design basis load demand upon the system. Examples include cable, piping, valves in stationary positions, resistors, capacitors, fluid filters, indicator lamps, cabinets, cases, etc.

Planned Operation - Planned operation is the normal station operation under planned conditions in absence of significant abnormalities. Operations subsequent to an incident (transient, accident, or special event) are not considered planned operations until the actions taken in the station are identical to those which would be used had the incident not occurred. The established planned operations can be considered as a chronological sequence:

refueling outage - achieving criticality - heatup - power operation

- achieving shutdown - cooldown - refueling outage.

The following planned operations are identified:

a. Refueling Outage - A planned operation, consisting of the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling, which includes all the actions associated with a normal refueling outage:

(1) Planned, physical movement of core components such as the fuel and control rods.

(2) Refueling surveillance and testing operations.

(3) Planned maintenance.

For the purpose of designating the frequency of testing and surveillances, a refueling outage shall mean a regularly scheduled refueling outage.

b. Achieving Criticality - Achieving criticality is a planned operation which includes all the actions which are normally accomplished in bringing the station from a condition in which all control rods are fully inserted to a condition in which nuclear criticality is achieved and maintained.

c. Heatup - Heatup is the planned operation which begins where achieving criticality ends and includes all actions which are normally accomplished in approaching nuclear steam supply system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator and bypass valve operation.

d. Power Operation - Power operation is the planned operation which begins where heatup ends and includes continued operation of the station at power levels in excess of heatup power.

e. Achieving Shutdown Achieving shutdown is the planned operation which begins where power operation ends and includes all actions normally accomplished in achieving nuclear shutdown (reactivity equivalent to more than one rod subcritical) following power operation.

I-2-5 04;os11s I

USAR

f. Cooldown - Cooldown is the planned operation which begins where achieving shutdown ends and includes all actions normally accomplished in the continued removal of decay heat and the reduction of nuclear steam supply system temperature and pressure.

Power Generation - The phrase power generation, when used to modify words such as objective, design basis, action, and system, indicates that the objective, design basis, action, or system is related to the mission of the station, which is to generate electric power, as opposed to concerns considered to be of primary safety importance. Thus, the phrase power generation is used to identify aspects of the station which are not considered to be of primary importance to safety.

Power Generation Action - A power generation action is an action in the station which is required for the avoidance of specified conditions considered to be of primary significance to the station mission, the generation of electrical power. The specified conditions are those that are most directly related to the following:

(1) The ability to carry out the station mission, the generation of electrical power, through planned operation.

(2) The avoidance of conditions which would limit the ability of the station to generate electrical power.

(3) The avoidance of conditions which would prevent or hinder the return to conditions permitting the use of the station to generate electrical power following an abnormal transient, accident, or special event.

There are power generation actions associated with planned operation, abnormal operational transients, accidents, and special events. See Figure I-2-3.

Power Generation Design Basis - The power generation design basis for a power generation system states in functional terms the unique design requirements which establish the limits within which the power generation objective shall be met. A safety system may have a power generation design basis which states in functional terms the unique design requirements which establish the limits within which the power generation objective for the system shall be met.

Power Generation Evaluation - A power generation evaluation is an evaluation which shows how the system satisfies some or all of the Power Generation Design Bases. Because power generation evaluations are not directly pertinent to public safety, they are generally not included in the USAR. However, where a system or component has both safety and power generation objectives, a power generation evaluation can be used to clarify the safety versus power generation capabilities.

Power Generation Objective - A power generation objective describes in functional terms the purpose of a system or component as it relates to the mission of the station. This includes objectives which are specifically established so the station can fulfill its Power Generation Actions. A system or piece of equipment has a power generation objective if it is a Power Generation System. A safety system can have a power generation objective, in addition to a safety objective, if parts of the system are intended to function for power generation purposes. See Figure I-2-3.

I-2-6 04;os11s I

USAR Power Generation System - A power generation system is any system the action of which is not required for a safety action, but which is required for a power generation action. See Figure I-2-3.

Power Operation - See Planned Operation.

Primary Containment - The primary containment is a Radioactive Material Barrier consisting of the drywell in which the reactor vessel is located, the pressure suppression chamber, and process lines out to the first isolation valve outside the containment wall. Portions of the reactor coolant pressure boundary may become part of the primary containment, depending upon the location of a postulated failure. For example, a closed main steam line isolation valve is part of the primary containment barrier when the postulated failure of the main steam line is inside the primary containment.

Primary Containment Integrity - Primary Containment Integrity means that the Technical Specification Primary Containment Limiting Conditions for Operation are satisfied.

Process Control Program - The CNS Process Control Program ( PCP) establishes the processing conditions for assuring the solidification, dewatering or stabilization of CNS radioactive waste streams produced from the CNS liquid radioactive waste treatment system and from activities producing radioactive waste requiring solidification, dewatering or stabilization such as decontamination system resins, irradiated components and highly contaminated equipment. The PCP ensures that processing of radioactive waste containing liquid, which is subject to the requirements of 10CFR61, is consistent with the requirements specified in the Cooper Nuclear Station Offsite Dose Assessment Manual (ODAM) .

Process Safety System - A process safety system is a safety system which performs a safety action that is necessary during planned operation. See Figure I-2-2.

Protection System - Protection system is a generic term which may be applied to nuclear safety systems and engineered safety features. See Figure I-2-3.

Protective Action - A protective action is an ultimate action at the system level which contributes to and is required for the accomplishment of a safety action. System level actions which are required to accomplish reactor scram, reactor vessel isolation, containment isolation, pressure relief, automatic depressurization, and emergency core cooling are some of the protective actions. See Figures I-2-1, I-2-2, and I-2-3.

Protective Function - A protective function is a function which encompasses the monitoring of one or more station variables or conditions and the associated initiation of intra-system actions which eventually result in protective action. See Figure I-2-2.

Radioactive Material Barrier - A radioactive material barrier includes the systems, structures, or equipment that together physically prevent the uncontrolled release of radioactive materials. These barriers include the Reactor Fuel Barrier, Reactor Coolant Pressure Boundary, Primary Containment, and Secondary Containment.

Radioactive Material Barrier Damage - Radioactive material barrier damage is defined as an unplanned, undesirable breach in a Radioactive Material Barrier. Operation of a relief or safety valve does not constitute barrier damage.

I-2-7 04;os11s I

USAR Rated Thermal Power (RTP) See Technical Specification 1.1 definition. This is also termed 100% power in the USAR and Maximum Power Level per Operating License Condition 2.C. (1). Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear steam supply system pressure refer to the values of these parameters when the reactor is at Rated Thermal Power.

Reactor Coolant Pressure Boundary - The portion of the Nuclear System consisting of the reactor vessel and attached piping out to and including the second isolation valve in each attached pipe. Also known as Nuclear System Process Barrier.

Reactor Fuel Barrier - The reactor fuel barrier is the Radioactive Material Barrier consisting of the uranium dioxide fuel sealed in metal cladding.

Refuel Mode The refueling mode is defined in Technical Specification Table 1.1-1 as MODE 5.

Refueling Outage - See Planned Operation.

Reliability - Reliability is the probability that an i tern will perform its specified function without failure for a specified time period in a specified environment.

Risk - Risk is the product of the probability of an event and the adverse consequences of the event.

Rod Density - Rod density is the fraction or percent of control rods fully inserted into the core as determined by the total number of notches inserted into the core divided by the total number of notches inserted when all control rods are inserted.

Run Mode - Power operation is defined in Technical Specification Table 1.1-1 as MODE 1.

Safe Shutdown Systems and Components - Those components or portions of systems used to achieve and maintain a cold shutdown condition.

Safety - The word safety, when used to modify such words as objective, design basis, action, and system, indicates that the objective, design basis, action, or system is related to concerns considered to be of primary safety significance, as opposed to the plant mission of generating electrical power. Thus, the word safety is used to identify aspects of the station which are considered to be of primary importance with respect to safety.

Safety Action - A safety action is an ultimate action in the station which is required for the avoidance of specified conditions considered to be of primary safety significance. The specified conditions are those that are most directly related to the ultimate limits on the integrity of the radioactive material barriers or the unplanned or uncontrolled release of radioactive material. There are safety actions associated with planned operation, abnormal operational transients, accidents, and special events. Safety actions include such actions as the indication to the operator of the values of certain process variables, reactor scram, emergency core cooling, and reactor shutdown from outside of the control room. See Figures I-2-1 and I-2-2.

Safety Design Basis - The safety design basis for a safety system states in functional terms the unique design requirements which establish the limits within which the safety objective shall be met. A Power Generation System may have a safety design basis which states in functional terms the unique design requirements that ensure that neither planned operation nor operational failure I-2-8 04;os;15 I

USAR by the system results in conditions for which station safety actions would be inadequate.

Safety Evaluation - A safety evaluation is an evaluation which shows how the system satisfies the safety design basis. In the USAR, safety evaluations are provided for those systems having a safety design basis. Safety evaluations may form the bases for the Technical Specifications and establish why specific limitations are imposed. For station modifications and design changes, safety evaluations are performed on safety and non-safety systems in accordance with CNS procedures.

Safety Limit - The safety limits are limits within which the reasonable maintenance of the fuel cladding integrity and the reactor coolant system integrity are assured. Violation of such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

Safety Objective - A safety objective describes in functional terms the purpose of a system or component as it relates to conditions considered to be of primary significance to the protection of the public. This relationship is stated in terms of radioactive material barriers or radioactive material release.

The only systems which have safety objectives are safety systems. See Figure I-2-3.

Safety-Related - Safety-related functions, structures, systems and components are those that are necessary to ensure:

(1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition, or

( 3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposure of 10CFRl00, or 10CFR50.67 (Fuel Handling Accident or Loss of Coolant Accident).

Safety System - A safety system is any system, group of systems, component, or group of components that accomplish a safety action. See Figure I-2-3.

Scram - Scram refers to the automatic rapid insertion of control rods in response to the detection of undesirable conditions.

Sealed-Open/Closed - Sealed-open/ closed refers to maintaining a component in an established position by administrative controls using a wire seal such that deliberate action is necessary to defeat the mechanism.

Secondary Containment - The secondary containment is the Radioactive Material Barrier consisting of the reactor building, which completely encloses the primary containment, the standby gas treatment system, and the Elevated Release Point (ERP).

Secondary Containment Integrity - Secondary Containment Integrity means that the Technical Specification Limiting Condition for Operation for Secondary Containment is satisfied.

I-2-9 04;os11s I

USAR Sensor - A sensor is that part of an instrument channel used to detect variations in the measured station variable or parameter. See Figure VII-1-1.

Setpoint - A setpoint is that value of the monitored plant variable or parameter which causes an instrument channel trip or a relief or safety valve to relieve pressure.

Shutdown - The reactor is shutdown when the effective neutron multiplication factor (Kett) is sufficiently less than 1. 0 that the full withdrawal of any one control rod could not produce criticality.

Shutdown Mode - The hot and cold shutdown modes are defined in Technical Specifications Table 1.1-1 as MODES 3 and 4 respectively.

Single Failure - A single failure is a failure that can be ascribed to a single causal event. Single failures of active components are considered in the design of certain systems and are presumed in the evaluations of incidents to investigate the ability of the station to respond in the required manner under degraded conditions. See 10CFRS0, Appendix A.

Special Event - A special event is an event which neither qualifies as an abnormal operational transient nor an accident and is not a design basis event but which is postulated to demonstrate some special capability of the system or systems.

Special Safety System - A special safety system is a safety system that performs an action that is necessary in response to a special event. See Figure I-2-3.

Startup/Hot Standby Mode - The startup mode is defined in Technical Specification Table 1.1-1 as MODE 2.

Station Blackout (SBO) The complete loss of alternating current (ac) electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e., loss of offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency ac power system).

See 10CFRS0.63.

Test Duration - The test duration is the elapsed time between test initiation and test termination.

Top of Active Fuel (TAF) - The highest elevation to which fuel pellets extend in the core (either enriched or unenriched). See also Fuel Zone Zero. Note - TAF is equivalent to the term "Top of Active Irradiated Fuel" used in the Technical Specifications and BASES.

Total Effective Dose Equivalent (TEDE) - The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). See 10CFR20.

Trip - A trip is the change of state of a bistable device in a control and instrumentation circuit which represents the change from a normal condition. A trip signal, which results from a trip, is generated in the channels of a Trip System and produces subsequent trips and trip signals throughout the system as directed by the logic.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order I-2-10 04/08/15

USAR to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

A trip system terminates and loses its identity where outputs are combined in logic. See Figure VII-1-1.

Ultimate Heat Sink - The ultimate heat sink is the heat dissipation means for the Unit to the environment, including the necessary retaining structure, and any connecting canals or conduits.

Unavailability - Unavailability is the probability that an SSC is incapable of performing its specified function when called upon. (The sum of availability and unavailability equals unity.)

Unsafe Failure - An unsafe failure is a failure that negates system operability and which, due to its nature, is revealed only when the instrument or control channel is functionally tested or attempts to respond to a real signal.

I-2-11 04/08/15

USAR 3.0 METHODS OF TECHNICAL PRESENTATION 3.1 Purpose The purpose of the original Final Safety Analysis Report (FSAR) was to provide the technical information required by 10CFR50.34(b) to establish a basis for evaluation of the station with respect to the issuance of a facility operating license. The purpose of the USAR is to update descriptions and analyses to reflect current plant operations as required by 10CFR50.71(e).

3.2 Radioactive Material Barrier Concept Because the safety aspects of this report pertain to the relationship between station behavior under a variety of circumstances and the radiological effects on persons off-site, the report is oriented to the radioactive material barriers. This orientation facilitates evaluation of the radiological effects of the station on the environment. Thus, the presentation of technical information is considerably different from that which would be expected in an operational manual, maintenance manual, or nuclear engineer's handbook.

The overriding consideration that determines the depth of detailed technical information presented about a system or component is the relationship of the system or component to the radioactive material barriers. Systems that must operate to preserve or limit the damage to the radioactive material barriers are described in the greatest detail. Systems that have little relationship to the radioactive material barriers are described only with as much detail as necessary to establish their functional role in the station.

3.3 Organization of Contents The USAR is organized into 14 chapters each of which consists of a number of sections. The principal architectural and engineering criteria, which define the broad frame of reference within which the station is designed, are set forth in Section I-4.0. The categories used for classifying the CNS SSCs with respect to safety are given in Section I-5.0.

Chapters I I through XI I I present detailed information about the design and operation of the station. The nuclear safety systems and engineered safety features are integrated into these chapters according to system function

( emergency core cooling, control), system type (electrical, mechanical) , or according to their relationship to a particular radioactive material barrier.

Chapter III describes station components and presents design details that are most pertinent to the fuel barrier. Chapter IV describes station components and systems that are most pertinent to the nuclear system process barrier. Chapter V describes the primary and secondary containments. Thus, Chapters III, IV, and V are arranged according to the four radioactive material barriers.

The remainder of the chapters group system information according to station function (radioactive waste control, emergency core cooling, power conversion, control) or system type (electrical, structures). Chapter XIV provides an overall safety evaluation of the station which demonstrates both the adequacy of equipment to protect the radioactive material barriers and the ability of the engineered safety features to mitigate the consequences of situations in which one or more radioactive material barriers are assumed damaged.

3.4 Format Organization of Chapters The format and content provided is generally based on "Guide to the Organization and Contents of Safety Analysis Reports" issued by the Atomic Energy I-3-1 02/06/01

USAR Commission in June, 1966. Each chapter is designated by a Roman numeral, e.g., I, II, etc. Each chapter is subdivided further and given an Arabic numeral, e.g., 1.0, 2.0, etc., and is individually paginated, i.e., page number I-4-5 is the fifth page of Chapter I, section 4.0. Sections are further subdivided by numbers following decimal points.

Tabulations of data appearing throughout the text are designated as "Tables" and are identified by Roman numerals corresponding to the chapter in which it appears and an Arabic number indicating its section and its sequence of tables appearing in that section, e.g., Table I-4-2, is the second table appearing in section 4.0 of Chapter I.

Applicable sketches, pictures, and plots are placed at the end of a chapter, and are identified as "Figures" by the chapter and section numbers and the sequential order of the drawing or diagram, e.g., Figure V-2-1. Elsewhere, reference to appropriate NPPD controlled drawings is made within the text, in order to illustrate or clarify the information presented. An equipment symbol chart for the station process and instrumentation drawings is shown on General Electric Drawing 197R567 for General Electric equipment scope and Burns and Roe Drawing 2001, Sheets 1 and 2, for Burns and Roe equipment scope.

The general organization of a chapter describing a system or component is as follows:

Objective Design Basis Description Evaluation Inspection and Testing To clearly distinguish the safety versus power generation aspects of a system, the objective, design basis, and evaluation titles are modified by the word "safety" or "power generation," according to the definitions given in Section I-2.0. Systems that have safety objectives are safety systems. A safety evaluation is included only when the system has a safety design basis; the evaluation shows how the system satisfies the safety design basis. A power generation evaluation is included only when needed to clarify the safety versus power generation aspects of a system that has both safety and power generation functions.

A nuclear safety operational analysis of the station has been performed to systematically identify the operational limitations or restrictions which must be observed with regard to certain process variables and certain station systems to satisfy specified nuclear safety operational criteria. The method used for this analysis is described in Appendix G. The resulting operational limits or restrictions formed the bases for proposed technical specifications which were submitted as Appendix B to the original FSAR. The present Technical Specifications may have various bases.

Chapters presenting information on topics other than systems or components are arranged individually according to the subject matter so that the relationship between the subject and public safety is emphasized.

Within each section of the text, applicable supporting technical material is referenced. These general references are cited either in the text or in a list of references at the end of a chapter. Documents that have been incorporated into the USAR by reference are underscored in the main text, as applicable, and are listed in Table I-3-1.

Certain text has been designated as historical information by being italicized in a distinctive font. Historical information was accurate at the time I-3-2 02/06/01

USAR of its initial submittal to the AEC/NRC (or as subsequently revised) and may have been relied on for licensing decisions made at the time. While historical information remains an important reference when evaluating proposed changes, tests, or experiments pursuant to 10CFR50.59, updating of this information is not generally required for compliance with 10CFR50.71(e). An exception to this is when contemporary information affects previous safety analyses conclusions relative to public health, and new safety analyses have been prepared with the results placed on the CNS docket as a result of NRC requirements. Historical information relates to: a) text that describes a completed physical milestone that is inherently dated, b) topics that are outside the responsibility of NPPD to control or influence as 10CFR50. 59 changes, tests, or experiments, or c) information that has been retained to provide historical perspective, but that has been superseded with equivalent up-to-date information.

I-3-3 02/06/01

USAR TABLE I-3-1 DOCUMENTS INCORPORATED INTO THE CNS USAR BY REFERENCE The following documents have been submitted on the CNS docket and are considered to be incorporate d by reference into the CNS USAR in accordance with the requirement s of 10CFR50. 32. The revision numbers for these documents were effective a maximum of 6 months prior to the most recent USAR Update filing. For documents subject to a programmati c change process, revisions and submittal of changes to the incorporate d document are made in accordance with the respective regulatory requirement governing the document.

Documents Subject to Update

1. NPPD Cooper Nuclear Station Quality Assurance Program For Operation Policy Document, Revision 28. Applicable regulation - 10CFR50.54( a).

2. Cooper Nuclear Station Fifth Ten-Year Interval Inservice Inspection Program and Third Ten-Year Interval Containment Inservice Inspection Program, Revision 4. Applicable regulation - 10CFR50.55a (g).

3. NPPD Emergency Plan for Cooper Nuclear Station, Revision 81. Applicable -

10CFR50. 54 (q).

4. NPPD Cooper Nuclear Station controlled drawings listed in Table I-3-1.

Applicable regulation - 10CFR50.59 (as applied to the CNS Drawing Control Program) .

5. Cooper Nuclear Station Off site Dose Assessment Manual for Gaseous and Liquid Effluents (ODAM), April, 2020. Applicable regulatory requirement - Technical Specificatio ns Section 5. 5. 1. c. 10CFR50. 59 and 10CFR50. 71 (e) apply to major changes to radioactive waste treatment systems (liquid, gaseous, and solid) as specified in ODAM Section D 5.5.

6. Cooper Nuclear Station Technical Requirement s Manual Limiting Conditions for Operation (TLCOs). Applicable regulation 10CFR50.59. License Condition 2.C. (4) applies to TRM Section 3.11.

7. Cooper Nuclear Station Plant Unique Analysis Report, February, 2007.

Applicable regulation - 10CFR50.59.

I-3-4 02/23/23

USAR TABLE I-3-1 CONTROLLED DRAWINGS INCORPORATED BY REFERENCE DRAWING NUMBER SHEET VENDOR TITLE REVISION DPPPG600204 3800 COOPER #1, BB96FA- 2XBB81-13.9m 2, UPGRADED HP, HEAT 5 SIEMENS ENERGY B008247-980 500 FLOW DIAGRAM 01 117C3346 INSTALLATION ELECTRICAL PENETRATION SEAL, MEDIUM GENERAL ELECTRIC 01 VOLTAGE POWER 117C3349 GENERAL ELECTRIC INSTALLATION PENETRATION SEAL, INDICATION & CONTROL 00 161F282BC 1 GENERAL ELECTRIC PROCESS DIAGRAM, CORE SPRAY SYSTEM 06 197R567 1 GENERAL ELECTRIC PIPING & INSTRUMENT SYMBOLS 00 197R576 1 GENERAL ELECTRIC ASSEMBLY REACTOR 04 2001 1 BURNS & ROE FLOW DIAGRAM, SYMBOLS & ABBREVIATIONS 20 2001 2 BURNS & ROE FLOW DIAGRAM SYMBOLS & ABBREVIATIONS 08 2002 1 BURNS & ROE FLOW DIAGRAM, MAIN, EXHAUST & AUXILIARY STEAM SYSTEMS 48 2002 2 BURNS & ROE FLOW DIAGRAM, MAIN, EXHAUST & AUXILIARY STEAM SYSTEMS 42 2004 1 BURNS & ROE FLOW DIAGRAM CONDENSATE & FEEDWATER SYSTEMS 36 2004 2 BURNS & ROE FLOW DIAGRAM, CONDENSATE & FEEDWATER SYSTEMS 51 2004 3 BURNS & ROE FLOW DIAGRAM, CONDENSATE & FEEDWATER SYSTEMS 69 2006 FLOW DIAGRAM, CIRCULATING & SCREEN WASH & SERVICE 1 BURNS & ROE 91 WATER SYSTEMS 2006 FLOW DIAGRAM, CIRCULATING & SCREEN WASH & SERVICE 2 BURNS & ROE 50 WATER SYSTEMS 2006 FLOW DIAGRAM, CIRCULATING & SCREEN WASH & SERVICE 3 BURNS & ROE 63 WATER SYSTEMS 2006 4 BURNS & ROE, NPPD FLOW DIAGRAM, CONTROL BUILDING, SERVICE WATER SYSTEM 65 2007 FLOW DIAGRAM, TURBINE BUILDING CLOSED COOLING WATER BURNS & ROE 84 SYSTEM I-3-5 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 2009 BURNS & ROE FLOW DIAGRAM, AIR REMOVAL SYSTEM 36 FLOW DIAGRAM, INSTRUMENT AIR, CONTROL & TURBINE 2010 1 BURNS & ROE, NPPD B7 BUILDING 2010 FLOW DIAGRAM, INSTRUMENT AIR, CONTROL & TURBINE lA BURNS & ROE, NPPD 16 BUILDINGS 2010 2 BURNS & ROE, NPPD FLOW DIAGRAM, INSTRUMENT AIR, REACTOR BUILDING 99 2010 3 BURNS & ROE, NPPD FLOW DIAGRAM, SERVICE AIR 48 FLOW DIAGRAM, INSTRUMENT AIR, RADWASTE & AUGMENTED 2010 4 BURNS & ROE, NPPD 28 RADWASTE BUILDINGS 2010 5 BURNS & ROE, NPPD INSTRUMENT & SERVICE AIR, MISCELLANEOUS DETAILS 27 2012 3 BURNS & ROE FLOW DIAGRAM, ELECTRIC HEATING BOILER SYSTEM 23 FLOW DIAGRAM, FIRE PROTECTION TURBINE GENERATOR 2016 1 BURNS & ROE 74 BUILDING FLOW DIAGRAM, FIRE PROTECTION, SERVICE BUILDINGS &

2016 lA BURNS & ROE, NPPD 09 YARD FLOW DIAGRAM, FIRE PROTECTION, CONTROL, RADWASTE AND 2016 lB BURNS & ROE, NPPD 05 AUGMENTED RADWASTE BUILDINGS 2016 lC BURNS & ROE, NPPD FLOW DIAGRAM, FIRE PROTECTION, REACTOR BUILDING 04 FLOW DIAGRAM, TURBINE & RADWASTE BUILDING, 2017 BURNS & ROE 19 CONTAMINATED FLOOR DRAIN & RADWASTE ROOF DRAIN SYSTEM FLOW DIAGRAM, TURBINE GENERATOR BUILDING & CONTROL 2018 BURNS & ROE 44 BUILDING, HEATING & VENTILATING FLOW DIAGRAM, MAIN CONTROL ROOM & CABLE ROOM &

2019 1 BURNS & ROE COMPUTER ROOM, HEATING & VENTILATING & AIR 50 CONDITIONING FLOW DIAGRAM CHILLED WATER SYSTEM, ELECTRICAL SHOP, 2019 2 BURNS & ROE, NPPD NONCRITICAL SWITCHGEAR, CONTROL BUILDING & OFFICE 13 BUILDING 2020 BURNS & ROE, NPPD FLOW DIAGRAM, REACTOR BUILDING, HEATING & VENTILATING 63 I-3-6 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 2021 FLOW DIAGRAM, RADWASTE BUILDING, HEATING &

BURNS & ROE 19 VENTILATING 2026 1 BURNS & ROE, NPPD REACTOR VESSEL INSTRUMENTATION, FLOW DIAGRAM 69 2027 FLOW DIAGRAM, LOOP A, REACTOR RECIRCULATION &

1 BURNS & ROE, NPPD 75 SUPPRESSION CHAMBER VENT SYSTEMS & CONNECTIONS 2027 FLOW DIAGRAM, LOOP B, REACTOR RECIRCULATION &

2 BURNS & ROE, NPPD 15 SUPPRESSION CHAMBER VENT SYSTEMS & CONNECTIONS 2028 FLOW DIAGRAM, REACTOR BUILDING & DRYWELL EQUIPMENT BURNS & ROE 56 DRAIN SYSTEM 2030 1 BURNS & ROE FLOW DIAGRAM, FUEL POOL COOLING & CLEAN UP SYSTEM 37 2030 2 BURNS & ROE FLOW DIAGRAM, FUEL POOL COOLING & CLEAN UP SYSTEM 17 2031 FLOW DIAGRAM, REACTOR BUILDING, CLOSED COOLING WATER 1 BURNS & ROE 24 SYSTEM 2031 FLOW DIAGRAM, REACTOR BUILDING, CLOSED COOLING WATER 2 BURNS & ROE, NPPD 66 SYSTEM FLOW DIAGRAM, REACTOR BUILDING, CLOSED COOLING WATER 2031 3 BURNS & ROE 35 SYSTEM 2032 FLOW DIAGRAM, HIGH CONDUCTIVITY PROCESS, FLOOR DRAINS 1 BURNS & ROE 32

& CHEMICAL & LAUNDRY WASTE 2032 FLOW DIAGRAM, HIGH CONDUCTIVITY PROCESS, FLOOR DRAINS 2 BURNS & ROE 22

& CHEMICAL & LAUNDRY WASTE 2032 FLOW DIAGRAM, HIGH CONDUCTIVITY PROCESS, FLOOR DRAINS 3 BURNS & ROE 16

& CHEMICAL & LAUNDRY WASTE 2032 FLOW DIAGRAM, HIGH CONDUCTIVITY PROCESS, FLOOR DRAINS 4 BURNS & ROE 26

& CHEMICAL & LAUNDRY WASTE FLOW DIAGRAM, HIGH CONDUCTIVITY PROCESS, FLOOR DRAINS 2032 5 BURNS & ROE 11

& CHEMICAL & LAUNDRY WASTE 2033 FLOW DIAGRAM, LOW CONDUCTIVITY PROCESSING, EQUIPMENT 1 BURNS & ROE 21 DRAINS 2033 BURNS & ROE FLOW DIAGRAM, LOW CONDUCTIVITY PROCESSING, EQUIPMENT 2 25 DRAINS I-3-7 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 2033 FLOW DIAGRAM, LOW CONDUCTIVITY PROCESSING, EQUIPMENT 3 BURNS & ROE 16 DRAINS 2033 FLOW DIAGRAM, LOW CONDUCTIVITY PROCESSING, EQUIPMENT 4 BURNS & ROE 17 DRAINS 2034 BURNS & ROE FLOW DIAGRAM, PLANT MAKEUP WATER TREATMENT SYSTEM 72 2035 1 BURNS & ROE FLOW DIAGRAM, CONDENSATE FILTER DEMINERALIZER SYSTEM 17 2035 2 BURNS & ROE FLOW DIAGRAM, CONDENSATE FILTER DEMINERALIZER SYSTEM 13 2035 3 BURNS & ROE FLOW DIAGRAM, CONDENSATE FILTER DEMINERALIZER SYSTEM 13 2035 4 BURNS & ROE FLOW DIAGRAM, CONDENSATE FILTER DEMINERALIZER SYSTEM 16 2036 1 BURNS & ROE, NPPD FLOW DIAGRAM, REACTOR BUILDING, SERVICE WATER SYSTEM A6 2037 FLOW DIAGRAM, HEATING & VENTILATION STANDBY GAS BURNS & ROE 74 TEATMENT & OFF GAS FILTERS FLOW DIAGRAM, REACTOR BUILDING, FLOOR & ROOF DRAIN 2038 1 BURNS & ROE 56 SYSTEMS 2039 BURNS & ROE, NPPD FLOW DIAGRAM, CONTROL ROD DRIVE, HYDRAULIC SYSTEM 65 2040 1 BURNS & ROE, NPPD FLOW DIAGRAM, RESIDUAL HEAT REMOVAL SYSTEM 83 2040 2 BURNS & ROE, NPPD FLOW DIAGRAM RESIDUAL HEAT REMOVAL SYSTEM, LOOP B 20 2042 1 BURNS & ROE FLOW DIAGRAM, REACTOR WATER CLEANUP SYSTEM 37 2042 2 BURNS & ROE FLOW DIAGRAM, REACTOR WATER CLEANUP SYSTEM 16 2042 3 BURNS & ROE FLOW DIAGRAM, REACTOR WATER CLEANUP SYSTEM 24 FLOW DIAGRAM, REACTOR CORE ISOLATION COOLANT &

2043 BURNS & ROE, NPPD 58 REACTOR FEED SYSTEMS FLOW DIAGRAM, HIGH PRESSURE COOLANT INJECTION &

2044 BURNS & ROE, NPPD 77 REACTOR FEED SYSTEMS 2045 1 BURNS & ROE, NPPD FLOW DIAGRAM, CORE SPRAY SYSTEM 58 2045 2 BURNS & ROE, NPPD FLOW DIAGRAM, STANDBY LIQUID CONTROL SYSTEM 21 2049 3 BURNS & ROE FLOW DIAGRAM, CONDENSATE SUPPLY SYSTEM 20 I-3-8 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 2050 BURNS & ROE GENERAL ARRANGEMENT, TURBINE BUILDING, BASEMENT FLOOR PLAN 17 2051 GENERAL ARRANGEMENT, TURBINE BUILDING, MEZZANINE BURNS & ROE 29 FLOOR PLAN 2052 GENERAL ARRANGEMENT, TURBINE BUILDING, OPERATING BURNS & ROE 42 FLOOR PLAN 2053 BURNS & ROE GENERAL ARRANGEMENT, TURBINE BUILDING SECTION AA 02 2054 BURNS & ROE GENERAL ARRANGEMENT, TURBINE BUILDING SECTION BB 03 2056 GENERAL ARRANGEMENT, INTAKE STRUCTURE PLANS &

BURNS & ROE 16 SECTIONS 2059 GENERAL ARRANGEMENT, REACTOR BUILDING, PLAN BELOW BURNS & ROE 05 GRADE 2060 GENERAL ARRANGEMENT, REACTOR BUILDING PLAN AT BURNS & ROE 15 ELEVATION 903-6 2061 GENERAL ARRANGEMENT, REACTOR BUILDING PLAN AT BURNS & ROE 12 ELEVATION 931-6 2062 GENERAL ARRANGEMENT, REACTOR BUILDING PLAN AT BURNS & ROE 07 ELEVATION 958-3 2063 GENERAL ARRANGEMENT, REACTOR BUILDING PLAN AT BURNS & ROE 07 ELEVATION 976-0 2064 GENERAL ARRANGEMENT, REACTOR BUILDING PLAN AT BURNS & ROE 10 ELEVATION 1001-0 2065 BURNS & ROE GENERAL ARRANGEMENT, REACTOR BUILDING SECTION AA 05 2066 BURNS & ROE GENERAL ARRANGEMENT, REACTOR BUILDING SECTION BB 05 2067 GENERAL ARRANGEMENT, RADWASTE BUILDING PLANS AT BURNS & ROE 13 ELEVATION 877-6 & 903-6 2068 GENERAL ARRANGEMENT, RADWASTE BUILDING, PLANS AT BURNS & ROE 13 ELEVATION 918-0 & 934-0 2069 BURNS & ROE GENERAL ARRANGEMENT, RADWASTE BUILDING SECTIONS 08 2072 BURNS & ROE GENERAL ARRANGEMENT, AUGMENTED RADWASTE BUILDING PLAN 05 I-3-9 02/22/21

USAR 2073 GENERAL ARRANGEMENT, AUGMENTED RADWASTE BUILDING BURNS & ROE 00 SECTIONS 2079 1 BURNS & ROE FLOW DIAGRAM, AUGMENTED LIQUID RADWASTE SYSTEM 14 2079 2 BURNS & ROE FLOW DIAGRAM, AUGMENTED LIQUID RADWASTE SYSTEM 31 2080 FLOW DIAGRAM, AUGMENTED LIQUID RADWASTE BUILDING BURNS & ROE 16 EQUIPMENT, CHEMICAL AND FLOOR DRAINS 2084 BURNS & ROE, NPPD FLOW DIAGRAM, STANDBY NITROGEN INJECTION SYSTEM 30 2298 BURNS & ROE RADWASTE BUILDING, CONVEYOR OPERATION AREAS PLANS 00 3001 BURNS & ROE MAIN ONE LINE DIAGRAM 37 3002 AUXILIARY ONE LINE DIAGRAM MCC Z, SWGR BUS lA & lB &

1 BURNS & ROE, NPPD 61 lE & CRITICAL SWGR BUS lF & lG 3003 2 BURNS & ROE, NPPD AUXILIARY ONE LINE DIAGRAM, MCC A & B & F & G 58 3004 AUXILIARY ONE LINE DIAGRAM, MOTOR CONTROL CENTERS C&

3 BURNS & ROE, NPPD 25 D & H & J & DGl & DG2 3005 AUXILIARY ONE LINE DIAGRAM, MOTOR CONTROL CENTERS M &

4 BURNS & ROE, NPPD 81 N & P & U & V &W AUXILIARY ONE LINE DIAGRAM, STARTER RACKS LZ & TZ, 3006 5 BURNS & ROE, NPPD MOTOR CONTROL CENTERS K & L & LX & RA & RX & S & T & 94 TX & X AUXILIARY ONE LINE DIAGRAM, MOTOR CONTROL CENTER E &

3007 6 BURNS & ROE, NPPD 85 Q & R & RB & Y 3009 1 BURNS & ROE, NPPD ONE LINE SWITCHING DIAGRAM, 12.5 KV RING BUS SYSTEM 64 3010 1 BURNS & ROE, NPPD VITAL ONE LINE DIAGRAM 91 3058 BURNS & ROE DC ONE LINE DIAGRAM 70 3070 BURNS & ROE ELECTRICAL SYMBOL LIST 13 3401 AUXILIARY ONE LINE DIAGRAM, MOTOR CONTROL CENTER CA &

BURNS & ROE, NPPD 33 CB & MR & OGl & OG2 4003, NF 13293 BURNS & ROE CIVIL, OVERALL SITE & VICINITY PLAN 46 I-3-10 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 4215 STRUCTURAL REACTOR BUILDING, SHIELDING BLOCKS, HPCI BURNS & ROE 03 HATCH 4259 1 BURNS & ROE, NPPD CONTAINMENT VESSEL, PENETRATION LOCATION 10 4259 lA BURNS & ROE, NPPD CONTAINMENT VESSEL PENETRATION SCHEDULE 03 4260 2A BURNS & ROE, NPPD CONTAINMENT SUPPRESSION CHAMBER PENETRATION LOCATION 00 4260 2B BURNS & ROE, NPPD CONTAINMENT SUPPRESSION CHAMBER PENETRATION SCHEDULE 05 4286 BURNS & ROE ENERGY ABSORBING PANEL LINER LOCATIONS 04 6000302,72-17 PIPING & INSTRUMENTATION DIAGRAM, AUGMENTED OFF GAS 1 COSMO DYNE 54 SYSTEM 6000302,72-17 2 COSMO DYNE P & I DIAGRAM, AUGMENTED OFFGAS SYSTEM 25 719E415BB 1 GENERAL ELECTRIC PIPING & INSTRUMENTATION DIAGRAM, NUCLEAR BOILER 15 GENERAL ELECTRIC, 719E4 79BB 1 PROCESS RADIATION MONITORING SYSTEM 13 NPPD 719E580BB 1 GENERAL ELECTRIC CONTROL ROD DRIVE HYDRAULIC SYSTEM 03 729El 74BB 1 GENERAL ELECTRIC RECIRCULATION FLOW CONTROL SYSTEM 01 729E211BB GENERAL ELECTRIC RESIDUAL HEAT REMOVAL SYSTEM, PROCESS DIAGRAM 12 729E222BB 1 GENERAL ELECTRIC REACTOR PROECTION SYSTEM MPL 05 02 GENERAL ELECTRIC, 729E222BB 2 REACTOR PROTECTION SYSTEM, FUNCTIONAL CONTROL DIAGRAM 04 NPPD GENERAL ELECTRIC, 729E222BB 3 REACTOR PROTECTION SYSTEM 04 NPPD 729E223BB 1 GENERAL ELECTRIC NEUTRON MONITORING SYSTEM 01 729E223BB 2 GENERAL ELECTRIC NEUTRON MONITORING SYSTEM 01 GENERAL ELECTRIC, 72 9E4 02BB 1 FUNCTIONAL CONTROL DIAGRAM, CORE SPRAY SYSTEM 06 NPPD GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, CONTROL ROD DRIVE 729E4 71BB 1 02 NPPD HYDRAULIC SYSTEM I-3-11 02/23/23

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 729E4 71BB 2 03 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 729E4 71BB 3 02 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 729E4 71BB 4 01 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 729E4 71BB 5 02 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 7 2 9E4 71BB 6 01 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, CONTROL ROD DRIVE HYDRAULIC SYSTEM, FUNCTIONAL 729E4 71BB 7 03 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, REACTOR CORE ISOLATION 729E517BC 1 04 NPPD COOLING SYSTEM GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, REACTOR CORE ISOLATION 729E517BC 2 06 NPPD COOLING SYSTEM GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, REACTOR CORE ISOLATION 729E517BC 3 06 NPPD COOLING SYSTEM, RCIC SYSTEM GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, HIGH PRESSURE COOLANT 729E589BB 1 06 NPPD INJECTION SYSTEM GENERAL ELECTRIC, HIGH PRESSURE COOLANT INJECTION SYSTEM, FUNCTIONAL 729E589BB 2 06 NPPD CONTROL DIAGRAM GENERAL ELECTRIC, HIGH PRESSURE COOLANT INJECTION SYSTEM, FUNCTIONAL 729E589BB 3 06 NPPD CONTROL DIAGRAM PROCESS DIAGRAM, REACTOR CORE ISOLATION COOLANT 729E719BC 1 GENERAL ELECTRIC 03 SYSTEM (RCIC) SYSTEM PROCESS DIAGRAM, HIGH PRESSURE COOLANT INJECTION 729E720BB GENERAL ELECTRIC 04 SYSTEM, HPCI 7 2 9E7 27BB 1 RECIRCULATION FLOW CONTROL SYSTEM, FUNCTIONAL CONTROL GENERAL ELECTRIC 01 DIAGRAM 7 2 9E7 27BB 2 GENERAL ELECTRIC RECIRCULATION FLOW CONTROL SYSTEM 00 I-3-12 02/20/19

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 729E727BB 3 GENERAL ELECTRIC RECIRCULATION FLOW CONTROL SYSTEM, FUNCTIONAL CONTROL DIAGRAM 02 729E989 1 GENERAL ELECTRIC POWER RANGE MONITORING UNIT MPL 07 01 730E100 1 GENERAL ELECTRIC NEUTRON MONITORING SYSTEM ARRANGEMENT 02 730E100 2 GENERAL ELECTRIC NEUTRON MONITORING SYSTEM ARRANGEMENT 01 730E140BB 1 GENERAL ELECTRIC RESIDUAL HEAT REMOVAL SYSTEM, FUNCTIONAL CONTROL DIAGRAM 10 730E140BB 2 GENERAL ELECTRIC RESIDUAL HEAT REMOVAL SYSTEM 05 730E140BB 3 GENERAL ELECTRIC RESIDUAL HEAT REMOVAL SYSTEM, FUNCTIONAL CONTROL DIAGRAM 10 730E148BB 1 GENERAL ELECTRIC PROCESS DIAGRAM REACTOR WATER CLEANUP SYSTEM MPL 12 00 730E149BB 1 GENERAL ELECTRIC NUCLEAR BOILER, FUNCTIONAL CONTROL DIAGRAM 05 730E149BB 2 GENERAL ELECTRIC FUNCTIONAL CONTROL DIAGRAM, NUCLEAR BOILER, MISCELLANEOUS SYSTEM 02 730El97BB 6B GENERAL ELECTRIC REACTOR RECIRCULATION CONTROLLER CONFIGURATION 07 730E805BA GENERAL ELECTRIC, 1 NEUTRON MONITORING SYSTEM, FUNCTIONAL CONTROL DIAGRAM 01 NPPD 730E805BA GENERAL ELECTRIC, 6 NEUTRON MONITORING SYSTEM, FUNCTIONAL CONTROL DIAGRAM 01 NPPD 730E805BA GENERAL ELECTRIC, 7 NEUTRON MONITORING SYSTEM, FUNCTIONAL CONTROL DIAGRAM 01 NPPD 730E923 GENERAL ELECTRIC INCORE STARTUP CHAMBER RETRACT MECHANISM, FIELD INSTALLATION 03 7 31E611 1 GENERAL ELECTRIC PRIMARY STEAM PIPING 04 791E257 3 GENERAL ELECTRIC ELEMENTARY DIAGRAM, FEEDWATER CONTROL SYSTEM 16 791E257 4 GENERAL ELECTRIC ELEMENTARY DIAGRAM, FEEDWATER CONTROL SYSTEM 31 80E1143 AUTOMATION INDUSTRIES FUEL STORAGE RACK 01 80E1148 AUTOMATION FUEL STORAGE RACK 13 & SEISMIC BRACING, INSTALLATION INDUSTRIES IN POOL 03 919D690BC 3 GENERAL ELECTRIC REACTOR VESSEL 01 I-3-13 03/10/15

USAR DRAWING NUMBER SHEET VENDOR TITLE REVISION 919D690BC 5 GENERAL ELECTRIC REACTOR VESSEL 01 GENERAL ELECTRIC, STANDBY LIQUID CONTROL SYSTEM, FUNCTIONAL CONTROL 920D225BB 1 03 NPPD DIAGRAM GENERAL ELECTRIC, FUNCTIONAL CONTROL DIAGRAM, AREA RADIATION MONITORING 921D796 1 02 NPPD SYSTEM CNS-MS-43 NPPD LEAKAGE PATHS FROM OUTBOARD MSIV'S 05 CNS-NBI-10 NPPD REACTOR WATER LEVEL INDICATION CORRELATION 06 CP00l 1 REACTOR CONTROLS CONTROL ROD DRIVE HYDRAULIC SYSTEM DIAGRAM 15 107B, AREA TEMPERATURE MONITOR SYSTEM, FOR NUCLEAR BOILER ILE70-3, 10105 BURNS & ROE 00 3A SYSTEM, LEAK DETECTION 107C, INSTALLATION DETAILS, TEMPERATURE DETECTORS FOR ILE70-3, ID105 BURNS & ROE 04 4 NUCLEAR BOILER SYSTEMS LEAK DETECTION 107E, INSTALLATION DETAILS, TEMPERATURE DETECTORS FOR ILE70-3, 10105 BURNS & ROE 00 6 NUCLEAR BOILER SYSTEMS LEAK DETECTION 107F, INSTALLATION DETAILS, TEMPERATURE DETECTORS FOR ILE70-3, 10105 BURNS & ROE 00 7 NUCLEAR BOILER SYSTEMS LEAK DETECTION NC29546 NPPD TRANSMISSION LINE ROUTES 02 NC44587 NPPD ROUTING OF 12.5 KV UNDERGROUND SYSTEM 18 RALPH A. HILLER SAA085 4 20 INCH BY 5 INCH MSIV ACTUATOR 01 COMPANY SK101670R CONTAINMENT PENETRATION ASSEMBLY, EXTERNAL 1 BURNS & ROE 02 PENETRATION CONTROLS CONTAINMENT PENETRATION X7AD, X9A, X9B, ASSEMBLY &

SKM200, M200 1 BURNS & ROE 00 EXTERIOR PENETRATION CONTROLS I-3-14 02/23/23

USAR 4.0 PRINCIPAL DESIGN CRITERIA In meeting the FSAR content requirement of 10CFR50.34(b) for describing the station design bases it is necessary to distinguish those station structures, systems, and components (SSCs) which are required to meet specified measures of safety from those that are not. To be valid, such a determination must be performed in a consistent, systematic manner. The principal design criteria, for general and nuclear safety, has been determined by the necessary functions of a system to respond to planned operation, abnormal operational transients, accidents, and special events. The actual design of SSCs reflects the criteria that pertain to it.

The design of Cooper Nuclear Station preceded issuance of 10CFR50 Appendix A "General Design Criteria for Nuclear Power Plants" and the established regulatory concepts of safety-related and non-safety-rela ted. However, CNS conformance to the proposed 10CFR50 Appendix A General Design Criteria is provided in Appendix F.

A systematic method has been developed to evaluate functionally the nuclear safety aspects of the BWR. The first step in this method is to specify in measurable terms the major judgements to be considered as the primary safety requirements. These top level judgements are offered as unacceptable safety results. Table I-4-1 describes the set of unacceptable safety results associated with each major category of station events. These unacceptable safety results are those used in evaluating the Cooper Nuclear Station.

Using the unacceptable safety results, the criteria for selecting the events of each category, and a consistent set of ground rules for evaluating the station events, it is possible to identify all the station actions required to avoid the unacceptable safety results. Such actions are called safety actions. Similarly, it is possible to identify the SSCs and functions required to avoid the unacceptable safety results. The station Nuclear Safety Operational Analysis (Appendix G) details this analysis method and presents the results. A major distinction is made between those BWR aspects which are required for the avoidance of the unacceptable safety results and those which are most pertinent to the station mission - the generation of electrical power.

Two methods are provided for describing the principal design criteria. The first method is by functional classification (either safety or power generation). The second is on a system-by-system (or system group) basis.

Safety analysis requires the information gained in the functional classification approach to criteria, but system description is more easily understood through the system-by-system method. In this section both approaches to criteria are given; both are useful.

4.1 Principal Design Criteria - Functional Classification In the functional classification approach the criteria must be stated in sufficient detail to allow placement of each criterion into one classification category. Thus, there may be closely related criteria pertaining to any given system in each category. This is a natural outgrowth of the functional (unacceptable result) approach to classification. The actual design of a system must reflect all of the criteria that pertain to it.

The principal architectural and engineering criteria for the design and construction of the station are summarized below. Some of the more general criteria are so broad that they are applicable, at least in part, to more than one classification. In these very general cases all of the affected classifications are indicated. Specific design bases and design features are detailed in other sections of the USAR.

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USAR TABLE I-4-1 UNACCEPTABLE SAFETY RESULTS FOR STATION EVENT CATEGORIES Event Category Unacceptable Safety Results

1. Planned Operation 1-1 The release of radioactive material to the environs to such an extent that the limits of 10CFR20 are exceeded.

1-2 Fuel failure to such an extent that were the freed fission products released to the environs via the normal discharge paths for radioactive material, the limits of 10CFR20 would be exceeded.

1-3 Nuclear system stress in excess of that allowed for planned operation by applicable industry code.

1-4 The existence of a station condition not considered by station safety analyses.

2. Abnormal Operational 2-1 The release of radioactive material to the environs Transients to such an extent that the limits of 10CFR20 are exceeded.

2-2 Any fuel failure calculated as a direct result of the transient analyses.

2-3 Nuclear system stress in excess of that allowed for transients by applicable industry codes.

3. Accidents 3-1 Radioactive material release to such an extent that the guideline values of 10CFRl00 (or 10CFR50.67 for Fuel Handling or Loss of Coolant Accident) would be exceeded.

3-2 Catastrophic failure of the fuel barrier as a result of exceeding mechanical or thermal limits.

3-3 Nuclear system stresses in excess of that allowed for accidents by applicable industry codes.

3-4 Containment stresses in excess of that allowed by industry standard or code when containment is required.

3-5 Overexposure to radiation of station operation personnel in the control room.

4 .A Special Event - 4-1 The inability to bring the reactor to the shutdown Ability to Shutdown condition by manipulation of the local controls and Reactor from Outside equipment which are available outside the control Control Room room.

4-2 The inability to bring the reactor to the cold shutdown condition from outside the control room.

B Special Event - 4-3 The inability to shutdown the reactor independent Ability to Shutdown of control rods.

Reactor Without Control Rods C Special Event - 4-4 Exceeding limits based on 10CFR50.62.

Ability to Mitigate the Consequences of an ATWS D Special Event - 4-5 Inability to cope with a station blackout for a Station Blackout specified duration in accordance with 10CFR50.63.

I-4-2 02/05/10

USAR 4 .1.1 General Criteria

1. The station shall be designed so that it can produce electric power in a safe and reliable manner. The station design shall be in accordance with applicable codes and regulations.

2. The station shall be designed in such a way that the release of radioactive materials to the environment is limited, so that the limits and guideline values of applicable regulations pertaining to the release of radioactive materials are not exceeded.

3. The reactor core and reactivity control system shall be designed so that control rod action shall be capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.

4. Adequate strength and stiffness with appropriate safety factors shall be provided so that a hazardous release of radioactive material shall not occur.

4 .1. 2 Power Generation Design Criteria, (Planned Operation)

1. The nuclear system shall employ a boiling water reactor to produce steam for direct use in a turbine-generator.

2. The fuel cladding shall be designed to retain integrity as a radioactive material barrier for the design power range.

3. The fuel cladding shall be designed to accommodate without loss of integrity the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

4. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions from plant shutdown to design power.

The capacity of such systems shall be adequate to prevent fuel clad damage.

5. Control equipment shall be provided for recirculation flow control to allow the reactor recirculation flow to be manually adjusted.

6. It shall be possible to manually control the reactor power level.

7. Control of the nuclear system shall be possible from a single location.

8. Nuclear system process controls shall be arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

9. Fuel handling and storage facilities shall be designed to maintain adequate cooling for spent fuel.

10. Interlocks or other automatic equipment shall be provided as a backup to procedural controls to avoid conditions requiring the functioning of nuclear safety systems or engineered safety features.

I-4-3 04/22/02

USAR 4 .1. 3 Power Generation Design Criteria, (Abnormal Operational Transients)

1. The fuel cladding, in conjunction with other station systems, shall be designed to retain integrity throughout any abnormal operational transient.

2. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated for any abnormal operational transient.

3. Standby electrical power sources shall be provided to allow removal of decay heat under circumstanc es where normal auxiliary power is not available.

4 .1. 4 Nuclear Safety Design Criteria, (Planned Operation)

1. The station shall be designed so that fuel failure during planned operation is limited to such an extent that if the freed fission products were released to the environs, via the normal discharge paths, the activity levels would be lower than permitted under 10CFR20.

2. The reactor core shall be designed so that its nuclear characteris tics do not contribute to a divergent power transient.

3. The nuclear system shall be designed so there is no tendency for divergent oscillation of any operating characteris tic, considering the interaction of the nuclear system with other appropriate station systems.

4. Gaseous, liquid, and solid waste disposal facilities shall be designed so that the discharge and off-site shipment of radioactive effluents can be made in accordance with applicable regulations .

5. The design shall provide means by which station operations personnel can be informed whenever limits on the release of radioactive material are exceeded.

6. Sufficient indications shall be provided to allow determinati on that the reactor is operating within the envelope of condition considered by station safety analysis.

7. Radiation shielding shall be provided and access control patterns shall be established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal station operation (see Section XII-3.1.1).

8. Fuel handling and storage facilities shall be designed to prevent inadvertent criticality .

9. Fuel handling and storage facilities shall be designed to maintain adequate shielding for spent fuel.

4 .1. 5 Nuclear Safety Design Criteria, (Abnormal Operational Transients)

1. The station shall be designed so that no calculated fuel failure occurs as a result of any abnormal operational transient.

2. Those portions of the nuclear system which form part of the reactor coolant pressure boundary shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients.

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USAR

3. Nuclear safety systems shall act to assure that no damage to the reactor coolant pressure boundary results from internal pressures caused by abnormal operational transients.

4. Where positive, precise action is immediately required in response to abnormal operational transients, such action shall be automatic and shall require no decision or manipulation of controls by operations personnel.

5. Required safety actions shall be carried out by equipment of sufficient redundancy and independence that no single failure of active components can prevent the required actions.

6. The design of nuclear safety systems shall include allowances for environmental phenomena at the site.

7. Provision shall be made for control of active components of nuclear safety systems from the Main Control Room.

8. Nuclear safety systems shall be designed to permit demonstration of their functional performance requirements.

9. Standby electrical power sources shall be provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

10. Standby electrical power sources shall have sufficient capacity to power all nuclear safety systems requiring electrical power.

4 .1. 6 Nuclear Safety Design Criteria, (Accidents)

1. Those portions of the nuclear system which form part of the reactor coolant pressure boundary shall be designed to retain integrity as a radioactive material barrier following accidents.

2. Engineered safety features shall act to assure that no damage to the reactor coolant pressure boundary results from internal pressures caused by accident.

3. Where positive, precise action is immediately required in response to accidents, such action shall be automatic and shall require no decision or manipulation of controls by operations personnel.

4. Required safety actions shall be carried out by equipment of sufficient redundancy and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE 279 is applicable, single failures of passive electrical components will be considered as well as single failure of active components in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components. Note: The one exception to this criterion at CNS is the Control Room Emergency Filter System (CREFS). The CREFS is designed to provide the safety function of maintaining Main Control Room habitability in the event of a design basis accident (see Item 23 below). However, the CREFS at CNS as designed is a single train system which does not meet the redundancy requirements described above, and therefore, is not completely single failure proof. For a more detailed description of the safety function of the CREFS, refer to Section X-10.4.

5. Features of the station which are required for the mitigation of accident consequences shall be designed so that they can be fabricated and erected to quality standards which reflect the importance of the safety action to be performed. Note: An exception to this criterion at CNS is the MSIV Leakage Pathway, and the Main Turbine Condenser complex. These Class IIS SSCs are credited for mitigation of DBA LOCA and CRDA dose consequences, but have been analyzed as capable of withstanding the seismic loadings of a Safe Shutdown Earthquake. SLC is also credited for the mitigation of a DBA LOCA even though it is not a safety-related system. For a more detailed discussion of the safety function of these SSCs, refer to Sections III-9, IV-11 and XI-3.

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USAR

6. The design of engineered safety features shall include allowances for environment al phenomena at the site.

7. Provision shall be made for control of active components of engineered safety features from the Main Control Room.

8. Engineered safety features shall be designed to permit demonstrati on of their functional performance requirement s.

9. A primary containment shall be provided that completely encloses the reactor vessel.

10. The primary containment shall be designed to retain integrity as a radioactive material barrier during and following accidents that release radioactive material into the primary containment volume.

11. It shall be possible to test primary containment integrity and leak tightness at periodic intervals.

12. A secondary containment shall be provided that completely encloses both the primary containment and fuel storage areas.

13. The secondary containment shall be designed to act as a radioactive material barrier under the same conditions that require the primary containment to act as a radioactive material barrier.

14. The secondary containment shall be designed to act as a radioactive material barrier, if required, whenever the primary containment is open for expected operational purposes.

15. The primary and secondary containment s, in conjunction with other engineered safety features, shall act to prevent the radiologica l effects of accidents resulting in the release of radioactive material to the containment volumes from exceeding the guideline values of applicable regulations .

16. Provisions shall be made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment .

17. Piping that both penetrates the primary containment structure and could serve as a path for the uncontrolle d release of radioactive material to the environs shall be automatical ly isolated whenever such uncontrolle d radioactive material release is threatened. Such isolation shall be effected in time to prevent radiologica l effects from exceeding the guideline values of applicable regulations .

18. ECCS shall be provided to prevent the fuel clad temperature from exceeding 2200°F as a result of a loss-of-coo lant accident and shall meet all acceptable criteria of 10CFR50.46.

19. The ECCS shall provide for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary.

20. The ECCS shall be diverse, reliable, and redundant.

21. Operation of the ECCS shall be initiated automatical ly when required.

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USAR

22. Standby electrical power sources shall have sufficient capacity to power all engineered safety features requiring electrical power.

23. The Main Control Room shall be shielded against radiation so that continued occupancy under accident conditions is possible.

4 .1. 7 Nuclear Safety Design Criteria, ( Special Event)

1. Backup reactor shutdown capability shall be provided independent of normal reactivity control provisions (control rods). This backup system shall have the capability to shut down the reactor from any operating condition, and subsequentl y to maintain the shutdown condition.

2. In the event that the Main Control Room becomes inaccessible ,

it shall be possible to bring the reactor from power range operation to a hot shutdown condition by manipulation of the local controls and equipment which are available outside of the Main Control Room. Furthermore , station design shall not preclude the ability, in this event, to bring the reactor to a cold shutdown condition from the hot shutdown condition.

3. Under conditions indicative of an anticipated transient without scram (ATWS), the reactor coolant recirculatin g pumps will be automatical ly tripped.

4. In the event the reactor protection system (RPS) fails to insert the control rods, an alternate rod insertion system that is independent from the RPS from sensor output to the final actuation device will initiate control rod insertion.

5. During an ATWS, the backup reactor shutdown system will produce reactor shutdown (independent of control rods) quickly enough to maintain acceptable core and primary containment conditions.

6. The station will be able to cope with a loss of all AC power (station blackout) for the credited duration in accordance with 10CFR50.63.

4.2 Principal Design Criteria, System-By-S ystem The principal architectur al and engineering criteria for design are summarized below on a system-by-sy stem or system group basis. The system-by-sy stem presentation facilitates the understandin g of the actual design of any one system.

In the system-by-s ystem presentatio n of criteria, only the most restrictive of any related criteria are stated for a system.

4.2.1 General Criteria

1. The station shall be designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner.

The design shall be in accordance with applicable codes and regulations .

2. The station shall be designed in such a way that the release of radioactive materials to the environment will be as low as practicable , which will be lower than permitted by the applicable regulations pertaining to the release of radioactive materials.

4.2.2 Nuclear System Criteria

1. The nuclear system shall employ a General Electric boiling water reactor to produce steam for direct use in a turbine-gen erator.

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USAR

2. The fuel cladding shall be designed to retain integrity as a radioactive material barrier for the design power range and for any abnormal operational transient.

3. Those portions of the nuclear system which form part of the reactor coolant pressure boundary shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients and accidents.

4. The fuel cladding shall be designed to accommodate without loss of integrity the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

5. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions from shutdown to design power, and for any abnormal operational transient. The capacity of such systems shall be adequate to prevent fuel clad damage.

6. Heat removal systems shall be provided to remove decay heat generated in the core under circumstanc es wherein the normal operational heat removal systems become inoperative . The capacity of such systems shall be adequate to prevent fuel clad damage.

7. The reactor core and reactivity control system shall be designed so that control rod action shall be capable of bringing the core subcritical and maintaining it so, even with the rod at highest reactivity worth fully withdrawn and unavailable for insertion.

8. The nuclear steam supply system shall be designed so there is no tendency for divergent oscillation of any operating characteris tic, considering the interaction of the nuclear system with other appropriate station systems.

9. The reactor core shall be designed so that its nuclear characteris tics do not contribute to a divergent power transient.

4.2.3 Power Conversion Systems Criteria Components of the power conversion systems shall be designed to perform two basic objectives:

1. Produce electrical power from the steam coming from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater, with a major portion of its gases and particulate impurities removed.

2. Assure that any fission products or radioactivi ty associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.

4.2.4 Electrical Power System Criteria The station electrical power systems shall be designed to efficiently deliver the electrical power generated to the 345 kV transmissio n system.

Sufficient normal and standby auxiliary sources of electrical power shall be provided to attain prompt shutdown and continued maintenance of the I-4-8 02/06/01

USAR station in a safe condition . The capacity of the power sources shall be adequate to accomplis h all required engineere d safety feature functions under postulate d design basis accident condition s.

4.2.5 Radioacti ve Waste Disposal Criteria Gaseous, liquid, and solid waste disposal facilitie s shall be designed so that the discharge and off-site shipment of radioacti ve effluents will be in accordanc e with 10CFR20.

Process and discharge streams shall be appropria tely monitored and such features incorpora ted as may be necessary to maintain releases below the permissib le limits of 10CFR20.

4.2.6 Nuclear Safety Systems and Engineere d Safety Features Criteria

4. 2. 6 .1 General

1. Nuclear safety systems shall act in response to abnormal operation al transient s so that there will be no calculate d fuel failure.

2. Nuclear safety systems and engineere d safety features shall act to assure that no damage to the reactor coolant pressure boundary results from internal pressures caused by abnormal operation al transient s or accidents .

3. Where positive, precise action is immediate ly required in response to accidents , such action shall be automatic and shall require no decision or manipulat ion of controls by operation s personnel .

4. Required safety actions shall be carried out by equipment of sufficien t redundanc y and independe nce so that no single failure of active component s can prevent the required actions. For systems or component s to which IEEE 279 is applicabl e, single failures of passive electrica l component s will be considere d as well as single failure of active component s in recognitio n of the higher anticipat ed failure rates of passive electrica l component s relative to passive mechanica l componen ts.

5. Features of the station which are required for the mitigatio n of accident consequen ces shall be designed so that they can be fabricate d and erected to quality standards which reflect the importanc e of the safety function to be performed .

6. The design of nuclear safety systems and engineere d safety features shall include allowance s for environme ntal phenomena at the site.

7. Provision shall be made for control of active component s of nuclear safety systems and engineered safety features from the Main Control Room.

8. Nuclear safety systems and engineere d safety features shall be designed to permit demonstra tion of their functiona l performan ce requireme nts.

4.2.6.2 Containme nt and Isolation Criteria

1. A primary containme nt shall be provided to completel y enclose the reactor vessel. It shall be designed to act as a radioacti ve material barrier during and following accidents that release radioactiv e material into the primary containme nt. It shall be possible to test the primary containme nt integrity and leak tightness at periodic intervals .

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USAR

2. A secondary containment that completely encloses both primary containment and fuel storage areas shall be provided and shall be designed to act as a radioactive material barrier.

3. The primary and secondary containment s, in conjunction with other engineered safety features, shall act to prevent the release of radioactive material from within the containment volumes from exceeding the limiting values of applicable regulations .

4. Provisions shall be made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment .

5. Piping that both penetrates the primary containment structure and could serve as a path for the uncontrolled release of radioactive material to the environs shall be automatical ly isolated whenever such uncontrolle d radioactive material release is threatened. Such isolation shall be effected in time to prevent radiologica l effects from exceeding the limiting values of applicable regulations .

4.2.6.3 Emergency Core Cooling Systems (ECCS) Criteria

1. ECCS shall be provided to prevent the fuel clad temperature from exceeding 2200°F as a result of a loss-of-coo lant accident.

2. The ECCS shall provide for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary.

3. The ECCS shall be diverse, reliable, and redundant.

4. Operation of the ECCS shall be initiated automatical ly when required regardless of the availability of offsite power supplies and the normal generating system of the plant.

4. 2. 6. 4 Standby Power Criteria Standby electrical power sources shall be provided to allow prompt reactor shutdown and removal of decay heat under circumstanc es where normal auxiliary power is not available. They shall also provide sufficient power to all engineered safety features requiring electrical power.

4.2.7 Reactivity Control Criteria

1. Backup reactor shutdown capability shall be provided independent of normal reactivity control provisions. This backup system shall have the capability to shut down the reactor from any operating condition, and subsequently to maintain the shutdown condition. It shall also be able to produce a reactor shutdown (independent of control rods) quickly enough during postulated ATWS conditions to maintain acceptable core and primary containment conditions.

2. In the event that the Main Control Room is inaccessibl e, it shall be possible to bring the reactor from power range operation to a cold shutdown condition by manipulation of the local controls and equipment which are available outside of the Main Control Room.

I-4-10 02/06/01

USAR 4.2.8 Process Control Systems Criteria

4. 2. 8 .1 Nuclear Steam Supply System Process Control Criteria level.

1. It shall be possible to manually control the reactor power I

2. Control of the NSSS shall be possible from a single location.

3. NSSS process controls shall be arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

4. Interlocks or other automatic equipment shall be provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear safety systems or engineered safety features.

4. 2. 8. 2 Power Conversion Systems Process Control Criteria

1. Controls shall be provided to maintain temperature and pressure to below design limitations. This system will result in a stable operation and response to all allowable variations.

2. Controls shall be designed to provide indication of system trouble.

3. Control of the power conversion system shall be possible from a single location.

4. Sensors shall be provided to detect a loss of the main condenser.

5. Controls shall be provided to ensure adequate cooling of power conversion system equipment.

6. Controls shall be provided to ensure adequate condensate purity.

7. Controls shall be provided to regulate the supply of water so that adequate reactor vessel water level is maintained.

4. 2. 8. 3 Electrical Power System Process Control Criteria

1. Controls shall be provided to ensure that sufficient electrical power is provided for startup and normal operation and to attain prompt shutdown and continued maintenance of the station in a safe condition.

2. Control of the electrical power system shall be possible from a single location.

I-4-11 04/22/02

USAR 4.2.9 Auxilia ry Systems Criteria

1. Multipl e independ ent station auxiliar y systems shall be provided for the purpose of cooling and servicin g the station, the reactor and the station containm ent systems under various normal and abnorma l conditio ns.

2. Fuel handling and storage faciliti es shall be designed to prevent critica lity and to maintain adequate shieldin g and cooling for spent fuel.

4. 2 .10 Shieldin g and Access Control Criteria

1. Radiatio n shieldin g shall be provided and access control patterns shall be establis hed to allow the operatin g staff to control radiatio n doses within the limits of applicab le regulati ons in any mode of normal station operatio n ( see Section XI I-3 .1. 1) . The design and establish ment of the above shall include conditio ns which deal with fission product release from failed fuel elements and contami nation of station areas from system leakage.

2. The Main Control Room shall be shielded against radiatio n and have suitable environm ental control so that occupanc y under design basis accident conditio ns is possible .

4. 2 .11 Structu ral Loading Criteria The station structur es shall be designed to withstan d all applicab le loading conditio ns, includin g environm ental loads, so that a hazardou s release of radioac tive materia l shall not occur.

I-4-12 02/06/01

USAR 5.0 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS Certain structures, systems, and components (SSCs) perform active or passive functions that support the Safety Design Criteria described in Section I-4. The ways in which SSCs work together to avoid the Unacceptable Safety Results associated with Normal Plant Operations, Anticipated Operational Transients, Accidents, and Special Events is explained in the Nuclear Safety Operational Analysis (NSOA) of Appendix G. As explained in Appendix G, the NSOA provides a high level basis for classifying SSCs as being of primary importance to safety or power generation. Based on this and other analyses, classification groups have been established.

5.1 SSC Quality Group Classifications Appendix A describes the classification methodologies used by the NSSS supplier and the architect-engineer for fluid systems under their contract scope. The current methodology for classifying piping equipment pressure parts, including ASME classification, is also described in Appendix A. The NPPD Quality Assurance Program for Operation Policy Document specifies the criteria for application of the CNS Quality Assurance program to SSCs.

In response to Generic Letter 83-28, component level classifications have been established which use the terms "Essential" and "Non-Essential." These classifications are similar to the more common industry terms "safety-related" and "non-safety-related." An additional classification category is used for electrical equipment important to safety (both safety-related and non-safety-related) that is required to be environmentally qualified per 10CFR50.49. Figure I-5-1 shows the relationship of "safety-related" to "safety" and other terminologies used in the USAR.

5.2 Loading Classification Structures and equipment are designed to resist structural and mechanical damage due to loads produced by environmental and thermal forces. For the purpose of categorizing mechanical strength designs for these loads, the following definitions are established:

a. Class I - Structures and equipment whose failure could cause significant release of radioactivity or which are vital to a safe shutdown of the plant and removal of decay and sensible heat.

b. Class II - Structures and equipment which may be essential to the operation of the station, but which are not essential to a safe shutdown.

Section XII-2 describes these categories in detail and Appendix C discusses the structural loading criteria of Class I SSCs.

I-5-1 10/15/04

USAR 6.0 COMPARISON OF PRINCIPAL DESIGN CHARACTERISTICS This USAR section contains historical information as indicated by the italicized text. USAR Section I-3.4 provides a more detailed discussion of historical information. The factual information being presented in this section on the comparable plant data between CNS and other boiling water reactors has been preserved as it was originally submitted to the Atomic Energy Commission in the CNS FSAR. While no attempt has been made to update the information on the other BWRs, salient contemporary CNS information may be found in USAR Sections I-2, III-2, III-3, III-4, III-5, III-6, III-7, IV-3, IV-7, IV-8, V-2, V-3, VI-3, VI-4, VII-5, VII-9, VIII-2, VIII-3, VIII-5, VIII-6, X-5, X-8, XI-2, XI-3, XI-5, XI-6, XI-8, and Appendix C.

This section highlights the principal design features of the station and provides a comparison of the major features with other boiling water reactor facilities.

The design of this facility is based upon proven technology attained during the development, design, construction, and operation of boiling water reactors of similar or identical types. The data, performance characteristics, and other information presented herein represent a current firm design.

6.1 Nuclear System Design Characteristics Table 1-6-1 summarizes the design and operating characteristics for the Cooper Nuclear Station. The same characteristics are presented for the nuclear systems of the Vermont Yankee Nuclear Power Station, Monticello Nuclear Generating Plant, Duane Arnold Energy Center, and the Browns Feny Nuclear Plant, all of which have been reviewed by the Atomic Energy Commission and for which construction permits have been issued.

6.2 Power Conversion Systems Design Characteristics Table 1-6-2 summarizes the design and operating characteristics for the Cooper Nuclear Station.

These same characteristics are, again,presentedfor the Vermont Yankee Nuclear Power Station, Monticello Nuclear Generating Plant, Duane Arnold Energy Center, and the Browns Ferry Nuclear Plant.

6.3 Electrical Power Systems Design Characteristics The electrical power system design characteristics are presented in Table 1-6-3 together with those of the Vermont Yankee Nuclear Power Station, Monticello Nuclear Generating Plant, Duane Arnold Energy Center, and the Browns Ferry Nuclear Plant.

6.4 Containment Design Characteristics Table 1-6-4 summarizes the design characteristics for the primary and secondary containments ofthe Cooper Nuclear Station. Design characteristics are also presented for the primary and secondary containment systems employed for Vermont Yankee Nuclear Power Station, Monticello Nuclear Generating Plant, Duane Arnold Energy Center, and the Browns Ferry Nuclear Plant. In addition, data is given for the type, construction, and height of elevated release point for the above plants.

6.5 Structural Design Characteristics The seismic and wind loading design of the Cooper Nuclear Station are given in Table 1-6-5. The seismic and wind loading design of the Vermont Yankee Nuclear Power Station, Monticello Nuclear Generating Plant, Duane Arnold Energy Center, and the Browns Ferry Nuclear Plant are presented for comparison.

I-6-1 02/06/01

USAR TABLE 1-6-1 COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS*

(PARAMETERS ARE RELATED TO REFERENCE DESIGN THERMAL OUTPUT FORA SINGLE UNIT UNLESS OTHERWISE NOTED)

DUANE ARNOLD BROWNS FERRY THERMAL AND HYDRAULIC DESIGN VERMONT YANKEE MONTICELLO** ENERGY CENTER COOPER STATION UNITS 1 & 2 Reference Design Thermal Output, Mwt 1593 1670 1593 2381 3293 Maximum Anticipated Thermal Output, Mwt 1665 1670 1670 2500 3440 Steam Flow Rate, lb/hr 6.43 X 106 6.77x 106 6.843 X 106 9.81 X 106 13.38x106 Core Coolant Flow Rate, lb/hr 48.0 X 106 57.6 X 106 48.5 X 106 74.5x 106 102.5 X 106 Feedwater Flow Rate, lb/hr 6.43 X 106 6.75 X 106 6.822x 106 9.81 X 106 13.38x 106 Feedwater Temperature, °F 372 376.3 420 367.1 376.1 System Pressure, Nominal in Steam Dome, psia 1020 1020 1020 1020 1015 Average Power Density, kw/liter 50.8 40.6 50.9 51.2 50.18 Maximum Design Limit Output, kw/ft 18.37 17.5 18.5 18.5 18.4 Average Thermal Output, kw/ft 7.1 5.7 7.079 7.079 7.049 Maximum Heat Flux, Btu/hr-ft2 426,210 405,000 428,400 427,820 425,000 Average Heat Flux, Btu/hr-ft2 163,900 131,346 163,933 164,500 163,200 Maximum UO2 Temperature, °F 4380 4450 4380 4380 4380 Average Volumetric Fuel Temperature, °F 1100 900 1100 1100 1100 Average Fuel Rod Surface Temperature, °F 558 558 558 558 558 Minimum Critical Heat Flux Ratio (MCHFR) 21.9 21.9 21.9 21.9 21.9 Coolant Subcooling at Core Inlet, Btu/lb 27.1 27.0 24.4 29.9 28.7 Core Maximum Exit Voids Within Assemblies, % 79 76 79 79 79 Core Average Exit Quality, % Steam 13.6 12.0 14.3 13.2 13.2

• Items noted NA not available at the time ofprinting this table.

• • Data for this plant is for the "as-built" conditions reported in the final safety analysis.

I-6-2 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY DESIGN POWER PEAKING FACTOR VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 &2 Transverse Peaking Factor 1.4 1.58 1.405 1.4 1.4 Local Peaking Factor 1.24 1.24 1.24 1.24 1.24 Axial Peaking Factor 1.5 1.57 1.5 1.5 1.5 Total Peaking Factor 2.6 3.08 2.6 2.6 2.6 Nuclear Design (First Core)

Water/UO2 Volume Ratio (Cold) 2.41 2.42 (Undished) 2.41 2.41 2.41 Reactivity with Strongest Control Rod Out, < 0.99 < 0.99 < 0.99 < 0.99 < 0.99 keff Moderator Temperature Coefficient At 68 °F, L1 KJK - °F Water -5.0x 10-5 -8.0 X 10-5 -3.5 X J0-5 -3.5 X J0-5 -3.5 X J0-5 Hot, No Voids, L1 KIK - °F Water -17.0x Jo-5 -17.0x 10-5 -Jl.6 X 10-5 -J J.6 X 10-5 -J J.6 X J0-5 Moderator Void Coefficient Hot, No Voids, L1 KJK - % Void -I.Ox 10-3 -I.Ox 10-3 -8.7 X 10-4 -8. 7 X 10-4 -8. 7 X 10-4 At Rated Output, L1 KJK - % Void -l.6x 10-3 -J.4 X 10-3 -J.05 X J0-3 -J.05 X 10-3 -J.05 X J0-3 Fuel Temperature Doppler Coefficient At 68 °F, L1 KJK - °F Fuel -J.3 X 10-5 -J.2 X 10-5 -J.3 X 10-5 -J.3 X 10-5 -J.3 X 10-5 Hot, No Voids, L1 KJK - °F Fuel -J.2 X 10-5 -J.2 X 10-5 -J.2 X J0-5 -J.2 X] 10-5 -J.2 X 10-5 At Rated Output, L1 KJK - °F Fuel sl.3 x 10-5 sl.2 x 10-5 sl.3 x 10-5 sl.3x 10-5 sl.3 x Jo-5 Initial Average U-235 Enrichment, W/O 2.29% 2.25% 2.25% 2.17% 2.19%

Fuel Average Discharge Exposure, Mwd/Ton 19,000 19,000 18,350 19,000 19,000 Fuel Assembly__

Number of Fuel Assemblies 368 484 368 548 764 Fuel Rod Array 7x 7 7x 7 7X 7 7x 7 7x 7 I-6-3 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY CORE MECHANICAL DESIGN (CONTINUED) VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Fuel Assemblr_ (Continued)

Overall Dimensions, Inches 175.88 174.88 175.88 175.88 175.88 Weight of UO2 per Assembly, Pounds Undished- Undished- Undished- 487.4 487.4 487.4 492.5 490.35 Dished- Dished-481.7 483.42 Weight of Fuel Assembly, Pounds Undished- Undished- Undished- 682 682 682 678.9 681.48 Dished- Dished-668 674.55 Fuel Rods Number per Fuel Assembly 49 49 49 49 49 Outside Diameter, Inch 0.562 0.563 0.562 0.562 0.562 Clad Thickness, Inch 0.032 0.032 0.032 0.032 0.032 Gap - Pellet to Clad, Inch 0.005 0.005 0.005 0.005 0.005 Length of Gas Plenum, Inches 16 11.24 16 16 16 Clad Material Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Cladding Process Free standing Free standing Free standing Free standing Free standing Loaded tubes Loaded tubes Loaded tubes Loaded tubes Loaded tubes Fuel Pellets Material Uranium Uranium Uranium Uranium Uranium Dioxide Dioxide Dioxide Dioxide Dioxide Density, % of Theoretical 93% 93% 93% 93% 93%

I-6-4 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY CORE MECHANICAL DESIGN (CONTINUED) VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Fuel Pellets (Continued)

Diameter, Inch 0.488 0.488 0.488 0.488 0.488 Length, Inch 0.5 0.5 0.5 0.5 0.5 Fuel Channel Overall Dimension, Inches (Length) 166.875 166.875 166.875 166.875 166.875 Thickness, Inch 0.085 0.080 0.085 0.085 0.085 Cross Section Dimensions, Inches 5.448 X 5.438 X 5.438 X 5.448 X 5.448 X 5.448 5.438 5.438 5.448 5.448 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core Assembly Fuel Weight as UO2, Pounds 179,370 238,370 179,298 267,095 372,373 Zirconium Weight, Pounds (Z-2 + Z-4 + Spacers) -63,300 -80,993 -63,300 ~94,305 ~131,476 Core Diameter (Equivalent), Inches 129.9 149 129.9 158.5 178.1 Core Height (Active Fuel), Inches 144 144 144 144 144 Reactor Control System Method of Variation of Reactor Power Movable Movable Movable Movable Movable Control Rods Control Rods Control Rods Control Rods Control Rods

& Variable & Variable & Variable & Variable & Variable Coolant Coolant Coolant Coolant Coolant Pumping Pumping Pumping Pumping Pumping I-6-5 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY CORE MECHANICAL DESIGN (CONTINUED) VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Reactor Control System (Continued)

Number of Movable Control Rods 89 121 89 137 185 Shape of Movable Control Rods Cruciform Cruciform Cruciform Cruciform Cruciform Pitch of Movable Control Rods, Inches 12.0 12.0 12.0 12.0 12.0 Control Material in Movable Rods B4C Granules B4C Granules B4C Granules B4C Granules B4C Granules Compacted Compacted Compacted Compacted Compacted in SS Tubes in SS Tubes in SS Tubes in SS Tubes in SS Tubes Type of Control Rod Drives Bottom entry, Bottom entry, Bottom entry, Bottom entry, Bottom entry, Locking Locking Locking Locking Locking Piston Piston Piston Piston Piston Number of Temporary Control Curtains 156 216 None None None Curtain Material Flat, boron -- Flat, boron --

stainless stainless steel steel In-Core Neutron Instrumentation Number of In-Core Neutron Detectors 80 96 80 136 172 Number of In-Core Detector Strings 20 24 20 31 43 Number of Detectors per String 4 4 4 4 4 Number of Flux Mapping Neutron Detectors 3 3 3 4 5 Range (and Number) of Detectors Source Range Monitor Source to Source to Source to Source to Source to

.01%power (4) 10% power (4) .001% power (4) 10% power (4) 10% power (4)

Intermediate Range Monitor .0001% to 10% 1% to 10% .0001% to 10% 1% to 10% 1% to 10%

power (6) power (8) power (6) power (8) power (8)

Local Power Range Monitor 5% to 125% 5% to 125% 5% to 125% 5% to 125% 5% to 125%

power (80) power (96) power (80) power (124) power (172)

Average Power Range Monitor 5% to 125% 5% to 125% 2.5% to 125% 5% to 125% 5% to 125%

power (6) power (6) power (6) power (6) power (6)

Number and Type of In-Core Neutron Sources 8 Sb-Be 5 Sb-Be 4 Sb-Be 5 Sb-Be 7 Sb-Be I-6-6 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY REACTOR VESSEL DESIGN VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS I & 2 Material Carbon Steel/Clad Stainless Steel (ASME SA-336 & SA-302B)

Design Pressure, Psia 1265 1265 1265 1265 1265 Design Temperature, °F 575 575 575 575 575 Inside Diameter, Ft-In. 17-1 17-2 15 - 3 18-2 20-11 Inside Height, Ft-In. 63 - 1.5 63-2 66-4 69- 4 72- 0 Side Thickness (Including Clad) 5.187 5.187 4.593 *3 5.53] 6.313 Minimum Clad Thickness, Inches 1/8 1/8 1/8 1/8 1/8 Reactor Coolant Recirculation Desig_n Number ofRecirculation Loops 2 2 2 2 2 Design Pressure Inlet Leg, Psig 1175 1148 1148 1148 1148 Outlet Leg, Psig 1274 1248 1268 1274 1326 Design Temperature, °F 562 562 562 562 562 Pipe Diameter, Inches 28 28 22 28 28 Pipe Material 304/316 304 304/316 304/316 304/316 Recirculation Pump Flow Rate 32,500 gpm 32,500 gpm 27,100 gpm 45,200 gpm 45,000 gpm Number ofJet Pumps in Reactor 20 20 16 20 20 Main Steam Lines Number ofSteam Lines 4 4 4 4 4 Design Pressure, Psig 1146 1146 1146 1146 1146 Design Temperature, °F 563 563 563 563 563 Pipe Diameter, Inches 20 18 20 24 26 Pipe Material Carbon Steel (ASTM Al 55 KC70 or ASTM Al 06 Grade BJ I-6-7 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY CORE STANDBY COOLING SYSTEMS VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 &2 (These systems are sized on maximum anticipated thermal output)

Core Spray System Number ofLoops 2 2 2 2 2 Flow Rate (Gpm} 3000 at 3020 at 3020 at 4500 at 6250 at 136 psid 307 psid 127 psid 115 psid 122 psid High Pressure Coolant In[ection System (No.) 1 1 1 1 1 Number ofLoops 1 1 1 1 1 Flow Rate (Gpm} 4250 3000 2980 4220 5000 Automatic Depressurization System (No.) 1 1 1 1 1 Low Pressure Coolant In[ection (No.) 1 1 1 1 1 Number of Pumps 4 4 4 4 4 Flow Rate (Gpm/Pump) 7000 at 4000 at 4800 at 7700 at 10,000 at 20 psid 20 psid 20 psid 20 psid 20 psid Residual Heat Removal Sy_stem Reactor Shutdown Cooling (Number of Pumps} 4 4 4 4 4 Flow Rate (Gpm/Pump)O) 7000 3600 4800 7700 10,000 Capacity (Btu/Hr/Heat Exchanger)(2) 57.5 X 106 57.5x 106 35 X 106 70 X 106 70 X 106 Number of Heat Exchangers 2 2 2 2 4 Primary Containment Cooling Flow Rate (Gpm) 28,000 16,000 19,200 30,800 40,000 OJ Capacity during reactor flooding mode with 3 of 4 pumps operating.

(21:apacity during post-accident cooiing mode with 165 °F shell side inlet temperature, maximum service water temperature, and one RHR pump and two service water pumps in operation.

I-6-8 02/06/01

USAR TABLE 1-6-1 (CONTINUED)

DUANE ARNOLD BROWNS FERRY AUXILIARY SYSTEMS VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Service Water Sy_stem Flow Rate (Gpm/Pump) 2700 3500 2500 4000 4500 No. of Pumps 4 4 4 4 4 Reactor Core Isolation Cooling_§J!._stem Flow Rate (Gpm) 400 400 416 416 616 Fuel Pool Cooling and Cleanup System Capacity (Btu/Hr.) NA 2.87x 106 2.37 X 106 3.4 X 106 8.8x 106 I-6-9 02/06/01

USAR TABLE 1-6-2 COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS DUANE ARNOLD BROWNS FERRY TURBINE-GENERATOR VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 &2 Design Power, Mwt 1665 1670 1670 2487 3440 Design Power, Mwe 564 545 597 836 1152 Generator Speed, RPM 1800 1800 1800 1800 1800 Design Steam Flow, lb/hr 6.423 X 106 6.77x 106 6.894 X 106 10.049x 106 14.049x 106 Turbine Inlet Pressure, Psig 950 950 950 970 965 TURBINE BYPASS SYSTEM Capacity, percent of turbine design steam flow 100 15 25 25 25 MAIN CONDENSER Heat removal capacity, Btu/hr 3500x 106 3760x 106 3681 X 106 5367.6x 106 7770 X 106 CIRCULATING WATER SYSTEM Number of Pumps 3 2 2 or more 4 3 Flow Rate gpm/pump 117,000 140,000 130,000 162,500 200,000 or less CONDENSATE AND FEED WATER SYSTEMS Design Flow Rate, lb/hr 6.4x 106 6.77x 106 7.143 X 106 9,773,000 13,999,000 Number Condensate Pumps 2 2 2 3 3 Number Feedwater Pumps 2 2 2 2 3 Condensate Pump Drive a-c power a-c power a-cpower a-c power a-cpower Feedwater Pump Drive a-c power a-c power a-cpower turbine turbine I-6-10 02/06/01

USAR TABLE 1-6-3 COMPARISON OF ELECTRICAL POWER SYSTEMS DESIGN CHARACTERISTICS DUANE ARNOLD BROWNS FERRY TRANSMISSION SYSTEM VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Outgoing lines (number-rating) 2-345 kV 2-345 kV 2-345 kV 4-345 kV 4-500 kV NORMAL AUXILIARY A-C POWER Incoming lines (number-rating) 2-345 kV 2-345 kV 2-345 kV 1-161 kV 2-161 kV 1-230 kV 2-230 kV 3-161 kV 1-69 kV 1-115kV 1-115kV 1-4160 kV Auxiliary transformers 1 1 2 1 2 Start-up transformers 1 1 1 1 2 Shutdown transformers 1 STANDBY A-C POWER SUPPLY Number diesel generators 2 2 2 2 3of4 Number of 4160 V standby buses 2 2 2 2 4 Number of 480 V standby buses 3 3 3 3 5 D-C POWER SUPPLY Number of 125 V or 250 V batteries 2 3 2 2 2 Number of 125 V or 250 V buses 4 3 2 4 4 I-6-11 02/06/01

USAR TABLE 1-6-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS DUANE ARNOLD BROWNS FERRY PRIMARY CONTAINMENT* VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS I &2 Type Pressure Pressure Pressure Pressure Pressure Suppression Suppression Suppression Suppression Suppression Construction Drywell Light bulb Light bulb Light bulb Light bulb Light bulb shape; steel shape; steel shape; steel shape; steel shape; steel vessel vessel vessel vessel vessel Pressure Suppression Chamber Torus; steel Torus; steel Torus; steel Torus; steel Torus; steel vessel vessel vessel vessel vessel Pressure Suppression Chamber - Internal Design Pressure (psig) +56 +56 +56 +56 +56 Pressure Suppression Chamber - External Design Pressure (psi) +2 +2 +2 +2 +l Drywell - Internal Design Pressure (psig) +56 +56 +56 +56 +56 Drywell Free Volume (ft3J 134,00 134,200 130,930 145,430 159,000 Pressure Suppression Chamber Free Volume (ft3) 99,000 98,280 94,630 109,810 119,000 Pressure Suppression Pool Water Volume (ft3) 78,000 68,000 61,500 87,660 85,000 Submergence of Vent Pipe Below Pressure Pool Surface (ft) +4 +4 +4 +4 +4 Design Temperature of Drywell ( °F) 281 281 281 281 281 Design Temperature of Pressure Suppression Chamber ( °F) 281 281 281 281 281 Downcomer Vent Pressure Loss Factor 6.21 6.21 6.21 6.21 6.21

• Where applicable, containment parameters are based on maximum anticipated power output.

I-6-12 02/06/01

USAR TABLE 1-6-4 (CONTINUED)

DUANE ARNOLD BROWNS FERRY PRIMARY CONTAINMENT (CONTINUED) VERMONT YANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS I &2 Break Area/Total Vent Area 0.019 0.019 0.019 0.019 0.019 Calculated Maximum Pressure After Blowdown Drywell (psig) 35 41 45 46 40 Pressure Suppression Chamber (psig) 22 26 29 28 25 Initial Pressure Suppression Chamber Temperature Rise ( °F) 35 50 50 50 32 Leakage Rate (% Free Volume/Day at 56 psig and 281 °F) 0.5 0.5 0.5 0.5 0.5 SECONDARY CONTAINMENT Type Controlled Controlled Controlled Controlled Controlled Leakage Leakage Leakage Leakage Leakage Elevated Elevated Elevated Elevated Elevated Release Release Release Release Release Construction Lower Levels Reinforced Reinforced Reinforced Reinforced Reiriforced Concrete Concrete Concrete Concrete Concrete Upper Levels Steel Super- Steel Super- Steel Super- Steel Super- Steel Super-structure & structure & structure & structure & structure &

siding siding siding siding siding Roof Steel Steel Steel Steel Steel Sheeting Sheeting Sheeting Sheeting Sheeting Internal Design Pressure (psig) 0.25 0.25 0.25 0.25 0.25 Design Inleakage Rate (%free Volume/Day at 0.25 inches H2O) JOO JOO JOO JOO JOO ELEVATED RELEASE POINT Type Stack Stack Stack Stack Stack Construction Reinforced Reinforced To be Steel Steel concrete concrete determined Height (above ground) 318feet 89 meters JOO meters JOO meters 200 meters I-6-13 02/06/01

USAR TABLEI-6-5 COMPARJSON OF STRUCTURAL DESIGN CHARACTERISTICS DUANE ARNOLD BROWNS FERRY SEISMIC DESIGN VERMONTYANKEE MONTICELLO ENERGY CENTER COOPER STATION UNITS 1 & 2 Maximum Design (horizontal g) Q07 Q06 0.060) QJ0 QJ0 Maximum Hypothetical Earthquake QJ4 QJ2 0.120) Q20 Q20 WIND DESIGN Maximum Sustained (mph) 80 JOO 105 JOO 100 Tornados (mph) 300 300 300 300 300 (IJon rock.

I-6-14 02/06/01

USAR 7.0 STATION RESEARCH DEVELOPMENT AND FURTHER INFORMATION; REQUIREMENTS AND RESOLUTIONS

SUMMARY

The design of the boiling water reactor for this station was based upon proven technologic al concepts developed during the development, design, and operation of numerous similar reactors. The AEC, in reviewing the Browns Ferry Nuclear Plant and Cooper Nuclear Station dockets at the constructio n permit stage, identified several areas where further research and development efforts were required to more definitely assure safe operation of this station. Also, both the AEC Staff and the Advisory Committee on Reactor Safeguards (ACRS) had in their reviews of this and other recent reactor projects, identified several additional technical areas for which further detailed support information should be obtained. All of these development efforts were of three general types: (a) Those which pertained to the broad category of water-cooled reactors, (b) those which pertained specifically to boiling water reactors, and (c) those which have been noted particularly for a facility during the construction permit licensing activities by the AEC Staff and ACRS reviews.

Appendix H of this SAR provides a comprehensi ve examination and discussion of each of these concern areas, indicating resolution. A summary conclusion of the analysis is provided in this section in Tables I-7-1 through I-7-4. The tables cover the following areas of concern:

a. Areas specified in the Cooper Nuclear Station ACRS constructio n permit letter. (Refer to Table I-7-2).

b. Areas specified in the Cooper Nuclear Station AEC Staff constructio n permit safety evaluation. (Refer to Table I-7-3).

c. Areas specified in other related AEC-ACRS constructio n and operating permit letters. (Refer to Table I-7-4).

The scope of many of the areas of technology for items in a, b, and c above is discussed in Appendix Hin detail as part of an official response[ll by the General Electric Company to the various concerns expressed by the ACRS.

The General Electric Company had submitted many topical reports to the AEC in support of this application and those of other GE-BWR facilities.

(Refer to Table I-7-1).

I-7-1 02/06/01

USAR TABLE 1-7-1 COOPER NUCLEAR STATION TOPICAL REPORTS SUBMITTED TO THE AEC IN SUPPORT OF DOCKET The information contained in this table is designated as historical as indicated by the italicized text. It represents a listing of Topical Reports which were submitted in support of the original Licensing process. No attempts are made to update this table to reflect Topical Reports submitted since receipt of the CNS Operating License.

USAR Section 1-3. 4 provides a more detailed discussion of the purpose of highlighting certain text with Italics.

GE Report No.

1. APED 5286 Design Basis for Critical Heat Flux Condition in Boiling Water Reactors (September, 1966)

2. APED 5446 Control Rod Velocity Limiter (March, 1967)

3. APED5449 Control Rod Worth Minimizer (March, 1967)

4. APED 5450 Design Provisions for In-Service Inspection (April, 1967)

5. APED 5453 Vibration Analysis and Testing of Reactor Internals (April, 1967)

6. APED 5555 Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RDB144A (November, 1967)

7. TR67SL211 An Analysis of Turbine Missiles Resulting from Last Stage Wheel Failure (October, 1967)

8. APED 5608 General Electric Company Analytical and Experimental Program for Resolution ofACRS Safety Concerns (April, 1968)

9. APED 5455 The Mechanical Effects ofReactivity Transients (January, 1968)

JO. APED 5528 Nuclear Excursion Technology (August, 1967)

11. APED 5448 Analysis Methods of Hypothetical Super-Prompt Critical Reactivity Transients in Large Power Reactors (April, 1968)

12. APED 5458 Effectiveness ofCore Standby Cooling Systems for General Electric Boiling Water Reactors (March, 1968)

13. APED 5640 Xenon Considerations in Design of Large Boiling Water Reactors (June, 1968)

14. APED 5454 Metal Water Reactions - Effects on Core Cooling and Containment (March, 1968)

15. APED 5460 Design and Performance of General Electric Boiling Water Reactor Jet Pumps (September, 1968)

16. APED 5654 Considerations Pertaining to Containment Inerting (August, 1968)

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USAR TABLE 1-7-1 (CONTINUED)

GE Report No.

17. APED 5696 Tornado Protection for the Spent Fuel Storage Pool (November, 1968)

18. APED 5706 In-Core Neutron Monitoring System for General Electric Boiling Water Reactors, Rev. 1 (April, 1969)

19. APED 5703 Design and Analysis of Control Rod Drive Reactor Vessel Penetrations (November, 1968)

20. APED 5698 Summary of Results Obtained From a Typical Startup and Power Test Program for a General Electric Boiling Water Reactor (February, 1969)

21. APED 5750 Design and Performance ofGeneral Electric Boiling Water Reactor Main Steam Line Isolation Valves (March, 1969)

22. APED 5756 Analytical Methods for Evaluating the Radiological Aspects ofthe General Electric Boiling Water Reactor (March, 1969)

23. APED 5652 Stability and Dynamic Performance ofthe General Electric Boiling Water Reactor (April, 1969)

24. APED 5736 Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards (April, 1969)

25. APED 5447 Depressurization Performance of the General Electric Boiling Water Reactor High Pressure Coolant Injection System (June, 1969)

26. NEDD 10017 Field Testing Requirements for Fuel, Curtains, and Control Rods (June, 1969)

27. NEDD 10029 An Analytical Study on Brittle Fracture of GE-BWR Vessel Subject to the Design Basis Accident (July, 1969)

28. NEDD 10045 Consequences of a Steam Line Break in a General Electric Boiling Water Reactor (July, 1969)

29. NEDD 10139 Compliance of Protection Systems to Industry Criteria; GE BWR NSSS (June, 1970)

30. NEDD 10173 Current State of Knowledge High Performance BWR Zircaloy-clad UO2 Fuel (May, 1970)

31. NEDD 10179 Effects of Cladding Temperature and Material on ECCS Performance (June, 1970)

32. NEDD 10174 Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor (May, 1970)

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USAR TABLE 1-7-1 (CONTINUED)

GE Report No.

33. NEDO 10189 An Analysis ofFunctional Common-Mode Failures in GE BWR Protection and Control Instrumentation (July, 1970)

34. NEDD 10208 Effects of Fuel Rod Failure on ECCS Performance (August, 1970)

I-7-4 02/06/01

USAR TABLE I-7-2 COOPER NUCLEAR STATION ACRS CONCERNS - RESOLUTION Appendix H Section No. ACRS Concern Cooper Resolution H-2.2 Station Foundation USAR (Incorporate d in Design -

Support Chapter II)

H-2.3 Emergency On-Site USAR (Incorporate d in Design -

Power System Chapter VIII)

H-2.4 AEC General Design USAR (Incorporate d in Design -

Criteria No. 35 Chapter IV, Appendix A and Intent Design Appendix F)

H-2.5.1 Effects of Fuel Topical Report (GE-APED-56 08)

Failure on Topical Report (GE-NEDO-10 208)

ECCS Performance H-2.5.2 Effects of Cladding Topical Report (GE-APED-56 08)

Temperature and Topical Report (GE-APED-54 58)

Materials on Topical Report (GE-NEDO-10 179)

ECCS Performance H-2.5.3 Control Systems for USAR (Incorporate d in Design -

Emergency Power Chapters VI, VII, and VIII)

H-2.5.4 Diversifica tion of USAR (Incorporate d in Design -

ECCS Initiation Chapters VI and VII)

Signals H-2.5.5 Main Steam Line USAR (Incorporate d in Design -

Isolation Valve Chapters IV)

Testing Under Topical Report (GE-APED-57 50)

Simulated Accident Topical Report (GE-NEDO-10 045)

Conditions Topical Report (GE-APED-56 08)

H-2.5.6 Misorientat ion of USAR (Incorporate d in Design -

Fuel Assemblies Chapter I II)

H-2.5.7 Effects of Fuel Topical Report (GE-APED-56 08)

Bundle Flow Blockage Topical Report (GE-NEDO-10 174)

H-2.5.8 Verificatio n of Fuel Topical Report (GE-APED-56 08)

Damage Limit Criteria Dresden 2/3 - Amendment 14/15 Topical Report (GE-NEDO-10 173)

H-2.5.9 Control Rod Block USAR (Incorporate d in Design -

Monitor Design Chapter VII) Dresden 2/3 -

Amendments 17/18 and 19/20, Brunswick 1/2 - Supplement 5 I-7-5 02/06/01

USAR TABLE I-7-2 (CONTINUED)

Appendix H Section No. ACRS Concern Cooper Resolutio n H-2.5.10 Quality Assurance USAR (Incorpor ated in Design -

and Inspectio n of Chapter IV, Appendix D and the Reactor Primary Appendix J)

System H-2.5.11 Formulati on of an USAR (Incorpor ated in Design -

In-Servic e Inspectio n Appendix J)

Program H-2.5.12 Station Start Up Topical Report (GE-APED- 5698)

Program USAR (Incorpor ated in Design -

Chapter XI I I)

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USAR TABLE I-7-3 COOPER NUCLEAR STATION AEC STAFF CONCERNS - RESOLUTIONS Appendix H Section No. AEC Concern Cooper Resolutio n H-3.3.1 Linear Heat Topical Report (GE-APED- 5608)

Generatio n Rate Fuel Dresden 2/3 - Amendment 14/15 Damage Limit Topical Report - (GE-NEDO- 10173)

H-3.3.2 Local Fuel Melting Topical Report (GE-APED- 5608)

Resulting from Inlet Topical Report (GE-NEDO- 10174)

Coolant Orifice Blockage H-3.3.3 Effect of Fuel Clad Topical Report (GE-APED- 5608)

Failure on Emergency Topical Report (GE-NEDO- 10208)

Core Cooling H-3.4.1 Core Spray Topical Report (GE-APED- 5608)

Effective ness Topical Report (GE-APED- 5458)

Topical Report (GE-NEDO- 10179)

H-3.4.2 Stearn Line Isolation Topical Report (GE-APED- 5608)

Valve Testing Topical Report (GE-APED- 5750)

Topical Report (GE-NEDO- 10045)

H-3.4.3 Adequacy of HPCI USAR (Incorpor ated in Design -

System as a Chapter VI)

Depressu rizer Topical Report (GE-APED- 5608)

Topical Report (GE-APED- 5447)

H-3.4.4 Electrica l Equipment USAR (Incorpor ated in Design -

Inside Containme nt Chapters V and VII)

H-3.4.5 Control Rod Worth Topical Report (GE-APED- 5449)

Minimizer USAR (Incorpor ated in Design -

Chapter VII)

H-3.4.6 Jet Pump Developme nt Topical Report (GE-APED- 5460)

H-3.4.7 , Rod Velocity Limiter Topical Report (GE-APED- 5446)

USAR (Incorpor ated in Design -

Chapter I II)

H-3.4.8 In-Core Neutron Topical Report (GE-APED- 5456)

Monitor System Topical Report (GE-APED- 5706)

USAR (Incorpor ated in Design -

Chapter VI I)

H-3.5.1 Failure of Passive USAR (Incorpor ated in Design -

Component s of ECCS Chapters IV and XII)

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USAR TABLE I-7-3 (CONTINUED)

Appendix H Section No. AEC Concern Cooper Resolution H-3.5.2 Thermal Shock Topical Report (GE-NEDO-10 029)

USAR (Incorporate d in Design -

Chapters III and IV)

H-3.5.3 Interchanne l Flow USAR (Incorporate d in Design -

Stability Chapters III and VII)

Topical Report (GE-APED-56 52)

Topical Report (GE-APED-56 40)

Peach Bottom 2/3 - Amendment 2 H-3.5.4 In-Service Inspection USAR (Incorporate d in Design -

Appendix J)

H-3.5.5 Primary System Leak USAR (Incorporate d in Design -

Detection Chapter IV)

I-7-8 02/06/01

USAR TABLE I-7-4 AEC-ACRS CONCERNS ON OTHER RELATED DOCKETS COOPER CAPABILITY FOR RESOLUTION Appendix H Section No. AEC-ACRS Concerns Cooper Resolution H-4.2 LPCIS-Logic Control The LPCI Loop Select Logic was System Design removed by DC 76-2. (See USAR Chapter VII)

H-4.3 Re-Evaluati on of Main Topical Report (GE-APED-56 08)

Steam Line Break Topical Report (NEDO-10045)

Accident USAR (Incorporate d in Design -

Chapter XIV)

H-4.4 Design of Piping USAR (Incorporate d in Design -

Systems to withstand Chapter XII and Earthquake Forces Appendices A and C)

Dresden 2/3 - Amendment 13/14 H-4.5 Fuel Clad USAR (Incorporate d in Design -

Disintegrat ion Chapter VI)

Limitations Topical Report (GE-APED-56 08)

Topical Report (GE-NEDO-10 179)

Pilgrim Amendment 14 H-4.6 Automatic Pressure USAR (Incorporate d in Design -

Relief System - Chapters VI and VII)

Initiation Interlock H-4.7 Effects of Blowdown USAR (Incorporate d in Design -

Forces on Reactor Chapters III and IV, and Primary System Appendix C)

Components H-4.8 Separation of Control USAR (Incorporate d in Design -

and Protection System Chapters VI and VII and Functions Appendix F)

H-4.9 Instrumenta tion for USAR (Incorporate d in Design -

Prompt Detection Chapters VII and XIV) of Gross Fuel Brunswick 1/2 - Supplements 3 Failures and 4 (The requirement for scram and containment isolation functions of the MSIV Radiation Monitors was removed by Amendment 158)

H-4.10 Scram Reliability USAR (Incorporate d in Design -

Study Chapters III and VII)

Topical Report (GE-NEDO-10 189)

Brunswick 1/2 - Supplement 6 H-4. 11 Design Basis of Topical Report (GE-APED-57 56)

Engineered Safety USAR (Examined Capability of Features Design - Chapter XIV)

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USAR TABLE I-7-4 (CONTINUED)

Appendix H Section No. AEC-ACRS Concerns Cooper Resolutio n H-4.12 Hydrogen Generatio n USAR (Incorpor ated in Design -

Study Chapter V)

H-4.13 Seismic Design and USAR (Re-Confi rmation of Design -

Analysis Models Chapter XII and Appendix C)

Dresden 2 - Re-Confir mation Informati on (Submitte d October, 1969)

H-4.14 Automatic Pressure USAR (Incorpor ated in Design -

Relief System - Chapters VI, VII and VIII)

Single Component Topical Report (GE-NEDO- 10139)

Failure Capabilit y -

Manual Operation H-4.15 Flow Reference Scram USAR (Incorpor ated in Design -

Chapter VI I)

H-4.16 Matters of Current Regulator y Staff -

Applicant Discussio n a) Standby Gas USAR (Incorpor ated in Design -

Treatment System Chapters IV, VII, and VIII)

Electrica l and Physical Separatio n b) Official, Issued Technical Specifica tions Technical Specifica tions -

License Appendix A H-4.17 Future Items of Considera tion for Incorpora tion ....

a) Radiolyti c Topical Report (GE-APED- 5454)

Decompos ition of Topical Report (GE-APED- 5654)

Cooling Water Brunswick 1/2, Supplemen t 4 Dresden 3, Amendment 23 b) Developme nt of USAR (Justified Design -

Instrumen tation - Chapters III, IV, and Vibration and Loose Appendix C)

Parts Detection c) Consequen ces of USAR (Incorpor ated in Design -

Water Contamina tion Chapter XIV)

- Structura l Material - LOCA I-7-10 02/06/01

USAR TABLE I-7-4 (CONTINUED)

Appendix H Section No. AEC-ACRS Concerns Cooper Resolution H-4.18 Development of USAR (Justified Design -

Instrumenta tion - Chapter IV)

Primary Containment Technical Specificatio ns -

Leakage Detection (Chapters III and IV)

Increased Sensitivity Studies H-4.19 Development of USAR (Justified Design Instrumenta tion - Chapter III, IV, and Appendix C)

Vibration and Loose Parts Detection Studies H-4.20 ECCS - Leakage USAR (Justified in Design -

Detection, Protection Chapters IV, X, and Appendix F) and Isolation Brunswick 1/2 - Supplement 4, Capability C/R 6.4 H-4.21 Main Steam Lines - USAR (Chapter IV, Appendices D Standards for and H)

Fabrication , Q/C, and Inspection H-4.22 Primary Containment USAR (Incorporate d in Design -

Inerting Chapters V and VI)

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USAR

8.0 REFERENCES

FOR CHAPTER I

1. Bray, A. P., et al., APED-5608, "The General Electric Company Analytical and Experimental Program for Resolution of ACRS Safety Concerns", April, 1968.

2. Cooper Nuclear Station License Amendment 231, dated June 30, 2008, Issuance of Amendment Re: Measurement Uncertainty Recapture Power Uprate.

I-8-1 08/07/08

USAR Figure I-5-1 Safety-Related Engineered Safety Design Basis Functions, Features Accidents (LOCA, Structures, RDA, MSLB, FHA) Systems and Components Protection Systems Safety-Related and Non-Safety-Safety Functions Abnormal Related Nuclear Safety Systems and Systems Operational Functions, Transients Structures, Systems and Components*

Special Safety Systems Special Events Non-Safety Related Process Safety Systems Non Protection Functions, Systems Structures, Planned Systems and Power Generation Operations Power Generation Components Systems Functions and Systems RELATIONSHIP OF VARIOUS CATEGORIES OF SYSTEMS (The same horizontal level of one column relative to another column denotes equivalency)

• Non-safety-related functions and SSCs may be credited in mitigating Abnormal Operating Transients provided the event is enveloped by safety-related functions and SSCs. Assuming the non-safety-related equipment fails, safety-related equipment must ultimately assure: a) the integrity of the Reactor Coolant Pressure Boundary, b) the capability of shutting down the reactor and maintaining it in a safe shutdown condition, and c) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10CFRl00 (or 10CFR50.67 for Fuel Handling Accident).

03/09/07