ML19298B490

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Amendment 28 to Updated Final Safety Analysis Report, Chapter 1, Introduction and Summary
ML19298B490
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 10/04/2019
From:
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation
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ML19298B540 List:
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Download: ML19298B490 (171)
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Text

BFN-26

1.0 INTRODUCTION

AND

SUMMARY

TABLE OF CONTENTS 1.1 Project Identification .................................................................................................................. 1.1-1 1.1.1 Identification and Qualification of Contractors ............................................................ 1.1-1 1.1.2 Licensing Basis Documents ....................................................................................... 1.1-3 1.2 Definitions.................................................................................................................................. 1.2-1 1.3 Methods of Technical Presentation ........................................................................................... 1.3-1 1.3.1 Purpose ..................................................................................................................... 1.3-1 1.3.2 Radioactive Material Barrier Concept ........................................................................ 1.3-1 1.3.3 Organization of Contents ........................................................................................... 1.3-1 1.3.4 Format Organization of Sections ................................................................................ 1.3-2 1.3.5 Power Level Basis for Analysis of Abnormal Operational Transients and Accidents ............................................................................................................ 1.3-3 1.4 Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation ................... 1.4-1 1.4.1 Introduction ................................................................................................................ 1.4-1 1.4.2 Classification Basis .................................................................................................... 1.4-1 1.4.3 Use of the Classification Plan .................................................................................... 1.4-2 1.5 Principal Design Criteria ............................................................................................................ 1.5-1 1.5.1 Principal Design Criteria Classification-by-Classification ........................................... 1.5-1 1.5.2 Principal Design Criteria, System-By-System ............................................................ 1.5-7 1.6 Plant Description ....................................................................................................................... 1.6-1 1.6.1 General ...................................................................................................................... 1.6-1 1.6.2 Nuclear Safety Systems and Engineered Safeguards ............................................... 1.6-9 1.6.3 Special Safety Systems ............................................................................................. 1.6-15 1.6.4 Process Control and Instrumentation ......................................................................... 1.6-15 1.6.5 Auxiliary Systems....................................................................................................... 1.6-19 1.6.6 Shielding .................................................................................................................... 1.6-22 1.6.7 Implementation of Loading Criteria ............................................................................ 1.6-22 1.7 Comparison of Principal Design Characteristics ........................................................................ 1.7-1 1.7.1 Nuclear System Design Characteristics ..................................................................... 1.7-1 1.7.2 Power Conversion Systems Design Characteristics .................................................. 1.7-1 1.7.3 Electrical Power Systems Design Characteristics ...................................................... 1.7-1 1.7.4 Containment Design Characteristics .......................................................................... 1.7-1 1.7.5 Structural Design Characteristics ............................................................................... 1.7-2 1.7.6 Discussion of Core Design Improvement ................................................................... 1.7-2 1.0-i

BFN-26

1.0 INTRODUCTION

AND

SUMMARY

TABLE OF CONTENTS (Cont'd) 1.8 Summary of Radiation Effects ................................................................................................... 1.8-1 1.8.1 Normal Operation....................................................................................................... 1.8-1 1.8.2 Abnormal Operational Transients .............................................................................. 1.8-1 1.8.3 Accidents ................................................................................................................... 1.8-1 1.9 Plant Management .................................................................................................................... 1.9-1 1.10 Quality Assurance Program ....................................................................................................... 1.10-1 1.11 Identification-Resolution of Construction Permit Concern - Summary ....................................... 1.11-1 1.11.1 General ...................................................................................................................... 1.11-1 1.12 General Conclusions ................................................................................................................. 1.12-1 1.0-ii

BFN-26 INTRODUCTION AND

SUMMARY

LIST OF TABLES Table Title 1.3-1 List of FSAR Engineering Drawings 1.3-2 Engineering Drawings Cross-Reference List 1.4-1 BWR Safety Engineering Concept for Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.4-2A Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.4-2B Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.7-1 Comparison of Nuclear System Design Characteristics 1.7-2 Comparison of Power Conversion Systems Design Characteristics 1.7-3 Comparison of Electrical Power System Design Characteristics 1.7-4 Comparison of Containment Design Characteristics 1.7-5 Comparison of Containment Design Characteristics 1.11-1 Browns Ferry Nuclear Plant Topical Reports Submitted to the AEC in Support of Docket 1.11-2 Browns Ferry Nuclear Plant AEC-ACRS Concerns - Resolutions 1.11-3 Browns Ferry Nuclear Plant AEC-Staff Concerns - Resolutions Units 1 and 2 1.11-4 AEC-Staff Concerns - Resolutions Unit 3 1.11-5 AEC-ACRS Concerns On Other Dockets - Resolutions 1.11-6 AEC-ACRS Concerns On Other Dockets - Capability for Resolution 1.0-iii

BFN-26 INTRODUCTION AND

SUMMARY

LIST OF ILLUSTRATIONS Figure Title 1.2-1 Relationship Between Safety Action and Protective Action 1.2-2 Relationship Between Protective Functions and Protective Actions 1.2-3 Relationships Between Different Types of Systems, Actions, and Objectives 1.3-1 Piping and Instrument Symbols 1.3-2 General Symbols Flow Diagram 1.6-1 Equipment Plans - Roof 1.6-2 Equipment Plans - Elevations 664 and 639 1.6-3 sht 1 Equipment Plans - Elevations 621.25 and 617 (Unit 1) 1.6-3 sht 2 Equipment Plans - Elevations 621.25 and 617 (Unit 2) 1.6-4 Equipment Plans - Elevations 606 and 604 1.6-5 Equipment Plans - Elevations 593 and 586 1.6-6 Equipment Plans - Elevations 565 and 557 1.6-7 Equipment Plans - Elevations 541.5 and 519 1.6-8 sht 1 Equipment - Transverse Section (Unit 1) 1.6-8 sht 2 Equipment - Transverse Section (Unit 2) 1.6-8 sht 3 Equipment - Transverse Section (Unit 3) 1.6-9 Equipment - Longitudinal Section 1.6-10 Equipment - Longitudinal Sections 1.6-11 Equipment Plans - Roof and Elevations 664 and 639 1.6-12 Equipment Plans - Elevations 621.25, 617, 606 and 604 1.6-13 Equipment Plans - Elevations 593, 586, 565, and 557 1.6-14 Equipment Plans - Elevations 541.5 and 519 1.6-15 Equipment Plans - Roof Elevations 664 and 639 1.6-16 Equipment Plans - Roof Elevations 621.25 and 617 1.6-17 Equipment Plans - Elevations 557, 565, 541.5 and 519 1.6-18 Equipment Refueling Floor Laydown Space 1.6-19 Equipment Refueling Floor Laydown Space 1.6-20 Deleted 1.6-21 Deleted 1.6-22 Deleted 1.6-23 General Arrangement Plan - Elevations 595, 580, and 578 1.6-24 General Arrangement Plans - Elevations 565 and 546 1.0-iv

BFN-28 INTRODUCTION AND

SUMMARY

LIST OF ILLUSTRATIONS (Cont'd)

Figure Title 1.6-25 General Arrangement - Sections 1.6-26 Equipment - Plans and Sections 1.6-27 Equipment - Plan 1.6-28 Reactor Heat Balance - 3952 MWt 1.6-29 sht 1 Turbine-Generator Heat Balance - Rated Power (Unit 2) 1.6-29 sht 2 Turbine-Generator Heat Balance - Rated Power (Unit 3) 1.6-29 sht 3 Turbine-Generator Heat Balance - Rated Power (Unit 1) 1.6-30 General Plant Systems Flow Diagram 1.0-v

BFN-28 INTRODUCTION AND

SUMMARY

1.1 PROJECT IDENTIFICATION This Final Safety Analysis Report is in support of the application of the Tennessee Valley Authority (TVA), herein designated as the applicant, for facility operating licenses for a three-unit nuclear power plant located at the Browns Ferry site in Limestone County, Alabama, for initial power levels up to 3293 MWt each, under Section 104(b) of the Atomic Energy Act of 1954, as amended, and the regulations of the Atomic Energy Commission set forth in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50). The FSAR is now maintained up-to-date and used as a complete and accurate description of the Browns Ferry Nuclear Plant as constructed and as modified since.

The facility is designated as the Browns Ferry Nuclear Plant, hereinafter referred to as the plant.

Commercial operation of each unit began on the following dates: unit one on August 1, 1974, unit two on March 1, 1975, and unit three on March 1, 1977.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 were subsequently uprated by five percent from 3293 MWt to 3458 MWt. In August 2017, license amendments were issued for all three units for a core thermal power uprate from 3458 MWt to 3952 MWt.

As used throughout this document, Atomic Energy Commission (AEC) is equivalent to the Nuclear Regulatory Commission (NRC) formed under the Energy Reorganization Act of 1974.

1.1.1 Identification and Qualification of Contractors Irrespective of any contractual responsibilities with any suppliers, the Tennessee Valley Authority is the sole applicant for the facility licenses and as owner and applicant, is responsible for the design, construction, and operation of the plant.

1.1.1.1 Applicant The TVA power system is one of the largest in the United States. TVA is primarily a wholesaler of power, operating generating plants, and transmission facilities, but no retail distribution systems. The TVA transmission system contains over 17,000 miles of lines. TVA supplies power over an area of about 80,000 square miles in parts of seven southeastern states, containing more than 2.3 million residential, farm, commercial, and industrial customers.

1.1-1

BFN-28 TVA has pioneered in erecting large generating units. Examples are the 1,150-megawatt unit at the Paradise Steam Plant; the 1,300-megawatt units at the Cumberland Steam Plant; and the two 1,170-megawatt units at the Sequoyah Nuclear Plant; and two 1,170-megawatt units at the Watts Bar Nuclear Plant. A total of over 67 individual steam generating units have been designed, constructed, and placed in operation by TVA in the past 35 years.

Much of TVA's experience has been gained from early and continuing participation in nuclear power studies. In 1946, TVA participated in the Daniels power pile study at Oak Ridge and the work of the Parker Committee, which surveyed prospects of nuclear power application. In 1953, TVA started developing a nuclear power staff and began a more detailed study of possible uses of nuclear power on its own system. In 1960, TVA agreed to operate the Experimental Gas-Cooled Reactor for AEC at Oak Ridge, and developed a technical and operating staff. Many of these trained and experienced people were assigned to TVA engineering and operating organizations that have been or are directly involved in the planning, design, construction, and operation of the Browns Ferry Nuclear Plant.

1.1.1.2 Engineer-Constructor TVA acts as its own engineer-constructor.

Since 1949, TVA has designed and constructed a number of projects including twelve major coal-fired steam plants, consisting of 63 individual generating units.

TVA has an experienced, competent nuclear plant design organization, including a large number of engineers with many years of steam plant experience. TVA also has a similarly experienced construction organization which has had extensive experience in the construction of large steam plants. A comprehensive quality assurance program has been developed to assure that the plant has been designed and constructed and will be operated to adequate standards of quality.

1.1.1.3 Nuclear Steam Supply System Supplier General Electric Company was awarded a contract to design, fabricate, and deliver the nuclear steam supply system and nuclear fuel for the plant, as well as to provide technical direction for installation and startup of this equipment. General Electric (GE) 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 the Vallecitos Boiling Water Reactor, Dresden Unit 1, Humboldt Bay, Big Rock Point, KRB (Germany), KAHL (Germany), JPDR (Japan), SENN (Italy), Oyster Creek Unit 1, and Dresden Unit 2. Among the domestic reactors of General Electric design are Millstone Point Unit 1, Dresden Unit 3, Quad-Cities Units 1 and 2, Monticello Unit 1, Vermont Yankee Unit 1, Peach Bottom Units 2 and 3, Pilgrim, Hatch Units 1 and 2, Brunswick Units 1 and 2, 1.1-2

BFN-27 Cooper, Duane Arnold, and Fitzpatrick. Thus, General Electric has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation, startup, and support of the normal operation of the reactor.

1.1.1.4 Turbine-Generator Supplier The applicant awarded a contract to General Electric to design, fabricate, and deliver the turbine generators for the plant as well as to provide technical assistance for installation and startup of this equipment. General Electric has a long history in the application of turbine generators in nuclear power stations going back to the inception of nuclear facilities for the production of electrical power and has furnished the turbine-generator units for most of its BWR nuclear steam supply contracted stations.

General Electric was competent to design, fabricate and deliver the turbine-generator units and to provide technical assistance for the installation and startup of this equipment.

1.1.2 Licensing Basis Documents The following documents are typical documents submitted periodically to NRC.

Implementation of changes to these documents without NRC approval may be controlled by regulation or the plant license. The following list provides references for the review and approval requirements for the listed documents.

Document Regulation Or Requirement Updated Final Safety 10 CFR 50.71(e)

Analysis Report (UFSAR)

Technical Requirements 10 CFR 50.59 Manual and Tech. Spec. Bases Organizational Topical Report 10 CFR 50.54(a)(3)

Quality Assurance Plan 10 CFR 50.54(a)(3)

Offsite Dose Calculation Manual (ODCM) Tech. Spec., Section 5.5.1 Physical Security Plan 10 CFR 50.54(p)

Radiological Emergency Plan (REP) 10 CFR 50.54(q)

Core Operating Limits Report (COLR) Tech. Spec., Section 5.6.5 1.1-3

BFN-25 1.2 DEFINITIONS The following definitions apply to the terms used in the Safety Analysis Report.

1. Radioactive Material Barrier - A radioactive material barrier includes the systems, structures, or equipment that, together, physically prevent the uncontrolled release of radioactive materials. The four barriers are identified as follows:
a. Reactor Fuel Barrier - The uranium dioxide fuel is sealed in a zirconium cladding tube.
b. Nuclear System Process Barrier - The nuclear system process barrier includes the systems of vessels, pipes, pumps, tubes, and similar process equipment that contain the steam, water, gases, and radioactive materials coming from, going to, or in communication with the reactor core. The actual boundaries of the nuclear system process barrier depend upon the status of plant operation.

For example, process system isolation valves, when closed, form part of the barrier. The steam-jet ejector offgas path forms a planned process opening in the barrier during power operation.

Because the nuclear system process barrier is designed to be divided by isolation valve action into two major sections under certain conditions, this barrier is considered in two parts as follows:

(1) Nuclear system primary barrier - This barrier includes the reactor vessel and attached piping out to and including the second isolation valve in each attached pipe. In various codes and standards used in the industry, this barrier is sometimes referred to as the "primary system pressure boundary,"

(2) Nuclear system secondary barrier - This barrier is that portion of the nuclear system process barrier not included in the nuclear system primary barrier.

c. Primary Containment - The primary containment is defined as the drywell in which the reactor vessel is located, the pressure suppression chamber, and process line reinforcements out to the outermost containment isolation valve outside valve outside the containment wall.

Portions of the nuclear system process barrier may become part of the primary containment, depending upon the location of a postulated failure. For example, a closed main steam isolation valve is part of the 1.2-1

BFN-25 primary containment barrier when the postulated failure of the main steam line is inside the primary containment.

d. Secondary Containment - The secondary containment is the reactor building, which completely encloses the primary containment. The reactor building ventilation system and the standby gas treatment system constitute controlled process openings in this barrier.
2. Radioactive Material Barrier Damage - Radioactive material barrier damage is defined as an unplanned, undesirable breach in a barrier, except that the operation of a main steam relief valve does not constitute barrier damage.
3. Nuclear System - The nuclear system 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 coming from or going to the reactor core. The nuclear system includes the following:

Reactor vessel Reactor vessel internals Main steam 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 Core standby cooling systems Reactor water cleanup system Feedwater system piping between the reactor vessel and the first valve outside the primary containment.

4. 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 to generate electrical power.

Thus, the word "safety" is used to identify aspects of the plant which are considered to be of primary importance with respect to safety.

5. Power Generation - The phrase "power generation," 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 the mission of the plant - to generate electrical power - as opposed to concerns considered to be of primary safety importance. Thus, the phrase "power generation" is used to identify aspects of the plant which are not considered to be primary importance with respect to safety.

1.2-2

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6. Operational - The adjective "operational," along with its noun and verb forms, is used in reference to the working or functioning of the plant, in contrast to the design of the plant.
7. Scram - Scram refers to the rapid insertion of control rods. A scram is initiated either automatically in response to the detection of undesirable conditions or manually by the control room operator.
8. Limiting Safety System Setting (LSSS) - The limiting safety system setting is a setting on instrumentation which initiates the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.
9. Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.
10. Safety Limit - The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of the 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.
11. Normal Operation - Normal operation is normal plant 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 plant 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, and refueling outage.

The following planned operations are identified:

a. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage.

1.2-3

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b. Achieving Criticality - Achieving criticality includes all the plant actions which are normally accomplished in bringing the plant 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 begins where achieving criticality ends and includes all plant actions which are normally accomplished in approaching nuclear system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator.
d. Power Operation - Power operation begins where heatup ends and includes continued operation of the plant at power levels in excess of heatup power.
e. Achieving Shutdown - Achieving shutdown begins where power operation ends and includes all plant actions normally accomplished in achieving nuclear shutdown (more than one rod subcritical) following power operation.
f. Cooldown - Cooldown begins where achieving shutdown ends and includes all plant actions normally accomplished in the continued removal of decay heat and the reduction of nuclear system temperature and pressure.
12. Incident - An incident is any event--abnormal operational transient, accident, special event, or other event, not considered as part of planned operation.
13. 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 plant operations. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors initiating the events in this category.
14. Abnormal Occurrence - Abnormal occurrence refers to the occurrence of any plant condition that:
a. Causes any abnormal operational transient, or
b. Violates a limiting condition for operation as established in the technical specifications, or
c. Exceeds a limiting safety system setting as established in the technical specifications, or 1.2-4

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d. Causes any uncontrolled or unplanned release of radioactive material from the site.
15. Accident - An accident is a single event, not reasonably expected during the course of plant operations, that has been hypothesized for analysis purposes or postulated from unlikely but possible situations, 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.
16. Design Basis Accident - A design basis accident is a hypothesized accident the characteristics and consequences of which are utilized in the design of those systems and components pertinent to the preservation of radioactive material barriers and the restriction of radioactive material release from the barriers. The potential radiation exposures resulting from a design basis accident are greater than any similar accident postulated from the same general accident assumptions. For example, the consequences of a complete severance of a recirculation loop line are more severe than those resulting from any other single pipeline failure inside the primary containment.
17. Special Event - A special event that neither qualifies as an abnormal operational transient nor an accident but that is postulated to demonstrate some special capability of the plant or its systems.
18. Safety Action - A safety action is an ultimate action in the plant that is essential to 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 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, core standby cooling, and reactor shutdown from outside the control room. See Figures 1.2-1 and 1.2-3 and Tables 1.4-2A and 1.4-2B.
19. Power Generation Action - A power generation action is an action in the plant that is essential to the avoidance of specified conditions considered to be of primary significance to the plant mission--the generation of electrical power.

The specified conditions are those that are directly related to the following:

a. The ability to carry out the plant mission--the generation of electrical power--through planned operation,
b. The avoidance of conditions that would limit the ability of the plant to generate electrical power, and 1.2-5

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c. The avoidance of conditions that would prevent or hinder the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

There are power generation actions associated with planned operation, abnormal operational transients, accidents, and special events. See Figure 1.2-3.

20. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.
21. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.
22. Safety System - A safety system is any system, group of systems, component, or group of components the actions of which are essential to accomplishing a safety action. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
23. Process Safety System - A process safety system is a safety system the actions of which are essential to a safety action required during planned operation. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
24. Nuclear Safety System - A nuclear safety system is a safety system the actions of which are essential to a safety action required in response to an abnormal operational transient. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
25. Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
26. Protection System - Protection system is a generic term that may be applied to nuclear safety systems and engineered safeguards. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
27. Special Safety System - A special safety system is a safety system the actions of which are essential to a safety action required in response to a special event. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.
28. Power Generation System - A power generation system is any system the actions of which are not essential to a safety action, but which are essential to a power generation action. Power generation systems are provided for any of the following purposes:
a. To carry out the mission of the plant--generate electrical power--through planned operation, 1.2-6

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b. To avoid conditions which would limit the ability of the plant to generate electrical power, and
c. To facilitate and expedite the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

29. 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 that have objectives are safety systems. See Figure 1.2-3.

30. 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 plant. This includes objectives that are specifically established so the plant can fulfill the following purposes:
a. The generation of electrical power through planned operation,
b. The avoidance of conditions that would limit the ability of the plant to generate electrical power, and
c. The avoidance of conditions that would prevent or hinder the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

See Figure 1.2-3.

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.

31. Analytical Objective - An analytical objective describes the purpose or intent of a portion of the Safety Analysis Report presenting an analysis.
32. 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 by the system results in conditions for which plant safety actions would be inadequate.

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33. Power Generation Design Basis - The power generation design basis for a power generation system states in functional terms the unique design requirements that establish the limits within 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.
34. Safety Evaluation - A safety evaluation is an evaluation that shows how the system satisfies the safety design basis. A safety evaluation is performed for those systems having a safety design basis. Safety evaluations form the bases for the technical specifications and establish why specific safety limitations are imposed.
35. Power Generation Evaluation - A power generation evaluation is an evaluation that 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. 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.
36. Operational Nuclear Safety Requirements - An operational nuclear safety requirement is a limitation or restriction on either the value of a process variable or the operability of a plant system. Such operational nuclear safety requirements must be observed in the operation (not necessarily at power) of the plant to satisfy specified operational nuclear safety criteria. The aggregate of all operational nuclear safety requirements defines an operational framework within which actual plant operations must remain.
37. Rated Power - Rated power refers to operation at a reactor thermal power of 3952 MWt. Rated power is also termed 100 percent power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.
38. Design Power - Design power refers to the power level used in safety and licensing analyses which support operation at rated power. Power corresponds to 3952 MWt. For radiological dose analyses provided in Section 14.6, design power has been assumed to be 3952 MWt.

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39. Single Failure - A single failure is a failure that can be ascribed to a single causal event. Single failures are considered in the design of certain systems and are presumed in the evaluations of incidents to investigate the ability of the plant to respond in the required manner under degraded conditions. The nature of single causal event to be presumed depends on the risk of the event being evaluated. Reasonably expected single failures are presumed as the cause of abnormal operational transients. Single failures of passive equipment are assumed sometimes to be the causes of accidents. Safety actions essential in response to abnormal operational transients and accidents must be carried out in spite of single failures in active equipment.

In any case, a single failure includes the multiple effects resulting from the single causal event.

40. Operable - Operability - A system, subsystem, division, component, or device shall be Operable or have operability when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
41. Operating - A system or component is operating when it is performing its intended functions in its required manner.
42. Operating Cycle - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.
43. Deleted.
44. Mode - A Mode shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning, specified as follows, with fuel in the reactor vessel.

1.2-9

BFN-25 MODE TITLE REACTOR MODE AVERAGE REACTOR SWITCH POSITION COOLANT TEMPERATURE (F) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot NA Standby 3 Hot Shutdown(a) Shutdown 212 4 Cold Shutdown(a) Shutdown 212 5 Refueling(b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

45. Deleted.
46. Deleted.
47. Deleted.
48. Deleted.
49. Deleted.
50. Deleted.
51. Place in Isolated Condition - Place in isolated condition means conduct an uninterrupted normal isolation of the reactor from the main (turbine) condenser including the closure of the main steam isolation valves.
52. Deleted.
53. Deleted.
54. Deleted.
55. Refueling Outage - Refueling outage is a period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and 1.2-10

BFN-25 surveillance, a refueling outage shall mean a regular scheduled refueling outage.

56. Core Alteration - Core Alteration shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be Core Alterations:
a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and
b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of Core Alterations shall not preclude completion of movement of a component to a safe position.

57. Risk - Risk is the product of the probability of an event and the adverse consequences of the event.
58. Reliability - Reliability is the probability that an item will perform its specified function without failure for a specified time period in a specified environment.
59. Unreliability - Unreliability is the probability that a component or system will fail to perform its specified action for a specified time period in a specified environment. (The sum of reliability and unreliability equals unity.)
60. Availability - Availability is the probability that a system will be functional at any randomly selected instant.
61. Unavailability - Unavailability is the probability that component or system will not be functional at any randomly selected instant. (The sum of availability and unavailability equals unity.)
62. Repair Rate - The repair rate is the number of repairs completed per unit time.
63. Failure Rate - The failure rate is the number of failures per unit time.
64. Test Duration - The test duration is the elapsed time between test initiation and test termination.
65. Test Interval - The test interval is the elapsed time between the initiation of identical tests.

1.2-11

BFN-25

66. Active Component - 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 are: switch, relay, valve not remaining in a stationary position, pressure switch, turbine, transistor, motor, damper, pump, and analog meter.
67. 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 are: cable, piping, valve in stationary position, resistor, capacitor, fluid filter, indicator lamp, cabinet, and case.
68. Operating Basis Earthquake - That earthquake which produces the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional.
69. Design Basis Earthquake - That earthquake which produces the vibratory ground motion for which those features of the nuclear power plant necessary to shut down the reactor and maintain the plant in a safe condition without undue risk to the health and safety of the public are designed to remain functional.
70. Deleted.
71. Deleted.
72. Probable Maximum Flood - The Probable Maximum Flood (PMF) is the hypothetical flood (peak discharge, volume, and hydrograph shape) that is considered to be the most severe reasonable possible, based on comprehensive hydrometerological application of probable maximum precipitation, and other hydrologic factors favorable for maximum flood runoff, such as sequential storms and snowmelt. The PMF design level at the Browns Ferry site is 572.5 feet.

The term Maximum Possible Flood (MPF) has also been used in Browns Ferry design documents, however the preferred term for all Browns Ferry design is PMF. (See also Appendix 2.4.A, Probable Maximum Flood).

73. Emergency Core Cooling Systems (ECCS) are defined as:
a. High Pressure Coolant Injection System (HPCI),
b. Automatic Depressurization System,
c. Core Spray System, and 1.2-12

BFN-25

d. Low Pressure Coolant Injection System (LPCI) (an operating mode of the Residual Heat Removal System).

The term Core Standby Cooling Systems (CSCS) has also been used in the FSAR, design documents, and plant procedures to describe the same systems. The terms ECCS and CSCS may be used interchangeably.

1.2-13

r------,

I SAFETY I I SYSTEM B I I I L-- __ ...J r-----, ,------,

I SAFETY I PROTECTIVE I SAFETY I I SYSTEM A I ACTION B I SYSTEM C I I I IL _ _ _ _ ...JI L-- __ ...J 1

PROTECTIVE PROTECTIVE ACT ION A ACTION C SAFETY ACTION CONCEPT rSTANDBY 1 I I A-C I POWER I I SYSTEM L __ T__ ...JI PROVIDE r------, A-C r-LOW---,

I CORE I POWER 1 PRESSURE 1 I SPRAY I 1 COOLANT 1

I L __

SYSTEM T__ ...JI T__

ILINJECTION

__ _JI I

PUMP WATER PUMP WATER TO CORE TO CORE CORE STBY

' - - - - - - - COOL ING ( LOW ...-..,___ _ ___.

PRESSURE)

EXAMPLE AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Relationsh i p between Safety Action and Protective Action FIGURE1.2 - 1

SAFETY SYSTEM PROTECTIVE PROTECTIVE FUNCTION A FUNCTION B INTRA-SYSTEM ACTIONS PROTECTIVE ACTION CONCEPT HIGH PRESSURE COOLANT INJECTION SYSTEM HIGH REACTOR DRYWELL VESSEL LOW PRESSURE WATER LEVEL START SYSTEM OPEN VALVES START TURBINE PUMP PUMP WATER TO CORE EXAMPLE AMENDMENT 16 BROWNS FERRY NUCL EAR PLANT FINAL SAFETY ANALYSIS REPORT Relationship between Protec ti ve Functions and Protective Act ions FIGURE 1 . 2- 2

DIR SAP'[TT BIR MESStON Pow:R CENERA TJON J TI S.VHY SYSTtNS PIIOlECTlVE ACTJ(ltS SAFETY M'.:rtONS SAFETY 08J[CTIV[S NOT[S : I. ONLY TIO SYSTEMS C-, [A01 TYl'E AM SHCIIN . Tlt£11[ WAY It WOil( THAN THIS NI.MIU Of SYSTEMS JN ANY CATEGOllr,

2. THERE MAY BE CASES IHERE THE SYSTEM LEVEL ACfJON IS IOENrtCAL TO Ill[ ULTIMATE ACTION [NTH£ PLANT IN SUCH A CA$[ THE JNTEIIIICOIATE SYSTEM U:vtl ACTtOtf N[EO HOT 1K IDENTJP'U:O .

AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT F !NAL SAFETY ANAL YS!S REPORT Relo t lon1hfp1 betwun 011hr1nt Types of Syatems, Act I ons, and Object i vea F !CURE 1 . 2-J

BFN-19 1.3 METHODS OF TECHNICAL PRESENTATION 1.3.1 Purpose The original purpose of this Final Safety Analysis Report (FSAR) was to provide the technical information required by Section 50.34 of 10 CFR 50 to establish a basis for evaluation of the plant with respect to the issuance of facility operating licenses.

The FSAR is to serve as the one document which will provide a complete and up-to-date description of the Browns Ferry Nuclear Plant as constructed and as modified since. In accordance with 10 CFR 50.71(e), as amended, the FSAR will be kept up-to-date through the issuance of amendments. An effective page listing will be placed just ahead of the FSAR Table of Contents. This listing will ensure that each copy in use within TVA is maintained in the most up-to-date condition possible.

Except where otherwise specified, the information in this report is applicable to all three units of the Browns Ferry Nuclear Plant.

1.3.2 Radioactive Material Barrier Concept Because the safety aspects of this report pertain to the relationship between plant 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 plant 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 plant.

1.3.3 Organization of Contents The Final Safety Analysis Report is organized into 14 major chapters, each of which consists of a number of sections. A system for classifying the various aspects of the BWR with respect to safety is given in section 1.4 (fourth section in Chapter 1). This classification system is fundamental to assessing the adequacy of the plant with respect to the relative importances of different safety concerns. The principal architectural and engineering criteria, which define the broad frame of reference within which the plant is designed, are set forth in section 1.5. Section 1.6 presents a brief description of the plant in which the nuclear safety systems and engineered 1.3-1

BFN-19 safeguards are separated from the other plant systems, so that those systems essential to safety are clearly identified.

Chapters 2 through 13 present detailed information about the design and operation of the plant. The nuclear safety systems and engineered safeguards are integrated into these sections according to system function (core standby cooling, control),

system type (electrical, mechanical), or according to their relationship to a particular radioactive material barrier. Chapter 3 (Reactor) describes plant components and presents design details that are most pertinent to the fuel barrier. Chapter 4 (Reactor Coolant System) describes plant components and systems that are most pertinent to the nuclear system process barrier. Chapter 5 describes the primary and secondary containments. Thus Chapters 3, 4, and 5 are arranged according to the four radioactive material barriers.

The remainder of the chapters group system information according to plant function (radioactive waste control, core standby cooling, power conversion, control) or system type (electrical, structures). Chapter 14 (Plant Safety Analysis) provides an overall safety evaluation of the plant which demonstrates both the adequacy of equipment designed to protect the radioactive material barriers and the ability of the safeguard features to mitigate the consequences of situations in which one or more radioactive material barriers are assumed damaged.

1.3.4 Format Organization of Sections Sections are numerically identified by representing their order of appearance in a chapter by two numbers separated by a decimal point; e.g., 3.4 is the fourth section in Chapter 3. Sections are further divided into subsections by numbers separated by decimal points (3.4.1, 3.4.1.1, etc.). Pages within each section are consecutively numbered (3.4-1, 3.4-2, etc.).

Tables are identified by the section number followed by a decimal point and the number of the table according to its order of mention in the text; e.g., Table 7.5-3 is the third table of section 7.5. Drawings, pictures, sketches, curves, graphs, and engineering diagrams are identified as Figures and are numbered in the same manner as tables. Figures 1.3-1 and 1.3-2 defines the meanings of piping and instrumentation symbols used in the figures of this report. Table 1.3-1 provides a list of all design and plant system figures appearing in the FSAR with the FSAR figure number cross-referenced to the engineering drawing number.

1.3-2

BFN-19 The general organization of a section describing a system or component is as follows:

Objective Design Basis Description Evaluation Inspection and Testing To clearly distinguish the safety versus operational 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 1.2. 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 plant was performed to systematically identify the operational limitations or restrictions which were to be observed with regard to certain process variables and certain plant systems to satisfy specified nuclear safety operational criteria during the initial operational fuel cycle. The method used for this analysis is described in Appendix G. Subsequent nuclear safety operational analyses have been and will continue to be generated for the plant as they are necessitated by plant modifications.

Sections presenting information on topics other than plant 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. References are cited either at the bottom of a page or at the end of a subsection. Most of the references are cited as a particular technical basis for BWR plant design and analysis, but some are specifically applicable to the Browns Ferry Nuclear Plant. The references in this category generally provide a full development and analysis of some aspect of GE BWR plant technology. These special references are incorporated by reference into the safety analysis report, thereby becoming part of the license application.

1.3.5 Power Level Basis for Analysis of Abnormal Operational Transients and Accidents For those abnormal operational transients and accidents for which high power operation increases the severity of the results, the analyses assume plant operation at design power as an initial condition. For those events for which an initial condition 1.3-3

BFN-19 of low or intermediate power level operation renders the most severe results, the analyses presented in this report represent the most severe case within the operating spectrum.

1.3-4

BFN-28 Table 1.3-1 Sheet 1 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

1.3-1 Piping and Instrument Symbols 104R900 1.3-2 General Symbols Flow Diagram 0-47E800-2 1.6-1 Equipment Plans - Roof 47E200-1 1.6-2 Equipment Plans - Elevations 664 and 639 0-47E200-2 1.6-3 Equipment Plans - Elevations 621.25 and 617 Sheet 1 1-47E200-3 Sheet 2 2-47E200-3 1.6-4 Equipment Plans - Elevations 606 and 604 0-47E200-4 1.6-5 Equipment Plans - Elevations 593 and 586 0-47E200-5 1.6-6 Equipment Plans - Elevations 565 and 557 0-47E200-6 1.6-7 Equipment Plans - Elevations 541.5 and 519 0-47E200-7 1.6-8 Equipment - Transverse Section Sheet 1 1-47E200-8 Sheet 2 2-47E200-8 Sheet 3 3-47E200-8 1.6-9 Equipment - Longitudinal Section 0-47E200-9 1.6-10 Equipment - Longitudinal Sections 0-47E200-10 1.6-11 Equipment Plans - Roof and Elevations 664 and 639 3-47E200-11 1.6-12 Equipment Plans - Elevations 621.25, 617, 606 and 604 0-47E200-12 1.6-13 Equipment Plans - Elevations 593, 586, 565, and 557 3-47E200-13 1.6-14 Equipment Plans - Elevations 541.5 and 519 3-47E200-14 1.6-15 Equipment Plans - Roof Elevations 664 and 639 0-47E200-15 1.6-16 Equipment Plans - Roof Elevations 621.25 and 617 0-47E200-16 1.6-17 Equipment Plans - Elevations 557, 565, 541.5 and 519 0-47E200-17 1.6-18 Equipment Refueling Floor Laydown Space 0-47E200-18 1.6-19 Equipment Refueling Floor Laydown Space 0-47E200-19 1.6-23 General Arrangement Plan - Elevations 595, 580, and 578 0-47W215-1 1.6-24 General Arrangement Plans - Elevations 565 and 546 0-47W215-2 1.6-25 General Arrangement - Sections 0-47E215-3 1.6-26 Equipment - Plans and Sections 0-47W220-1 1.6-27 Equipment - Plan 3-47E220-2 1.6-29 Turbine Generator Heat Balance Values Sheet 1 2-47K1110-13 Sheet 2 3-47K1110-13 Sheet 3 1-47K1110-32 1.6-30 General Plant Systems Flow Diagram 0-47E800-1 2.2-4 Location of Principal Plant Structures 0-10E201-7

BFN-25 Table 1.3-1 (Contd)

Sheet 2 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

2.4A Figure 25 Channel Relocation West of Cooling Towers, Typical Sections 10H237 2.4A Figure 26 Channel Relocation West of Cooling Towers Sheet 1 0-10H243 Sheet 2 10H242 2.5-S1 Foundation Investigation Grid Low Level Radwaste Storage 0-10E151-7 2.5-17 Mechanical Instruments and Controls 0-47E600-121 2.5-19 Rock Excavation 41N703 3.4-8a CRD Hydraulic System - Mechanical Control Diagram Sheet 1 2-47E610-85-1 Sheet 2 2-47E2610-85-2 Sheet 3 3-47E610-85-1 Sheet 4 3-47E3610-85-2 Sheet 5 1-47E610-85-1 3.4-8b Control Rod Hydraulic System - Flow Diagram 0-47E820-1 3.4-8c Control Rod Drive Hydraulic System - Flow Diagram 2-47E820-2 3.4-8d CRD Hydraulic System - Mechanical Control Diagram 2-47E2610-85-5 3.4-8e Control Rod Drive Hydraulic System - Flow Diagram 3-47E820-2 3.4-8f Control Rod Drive Hydraulic System - Mechanical Control Diagram 3-47E610-85-5 3.4-8g CRD Hydraulic System - Mechanical Control Diagram 1-47E1610-85-2 3.4-8h CRD Hydraulic System - Mechanical Control Diagram 1-47E1610-85-5 3.8-1 Standby Liquid Control System Flow Diagram 2-47E854-1 3.8-2 Standby Liquid Control System Mechanical Control Diagram 2-47E610-63-1 3.8-3 Standby Liquid Control System - Flow Diagram 1-47E854-1 3.8-5 Standby Liquid Control System - Flow Diagram 3-47E854-1 3.8-6 Standby Liquid Control System - Mechanical Control Diagram 3-47E610-63-1 3.8-8 Standby Liquid Control System - Mechanical Control Diagram 1-47E610-63-1 4.2-1 Reactor Vessel 2-104R935-1 4.2-3 Reactor Vessel 3-104R935-1 4.2-4 Reactor Vessel 1-104R935-1 4.3-2a Nuclear Boiler Flow Diagram Sheet 1 1-47E817-1 Sheet 2 2-47E817-1 Sheet 3 3-47E817-1 4.4-6 T-Quencher for Safety/Relief Discharge 2-47W401-5 4.4-7 Mechanical Main Steam Relief Valve Vent Piping 3-47E401-5 4.4-8 Mechanical Main Steam Relief Valve Vent Piping 1-47W401-5 4.5-1 Primary Steam Piping 1-729E229-1 4.5-2 Primary Steam Piping 2-729E229-4 4.5-3 Primary Steam Piping 3-729E229-4 4.7-1a Reactor Core Isolation Cooling System Flow Diagram 2-47E813-1 4.7-1b Reactor Core Isolation Cooling System, Mechanical Control Diagram 2-47E610-71-1

BFN-25 Table 1.3-1 (Contd)

Sheet 3 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

4.7-1c Reactor Core Isolation Cooling System - Flow Diagram 3-47E813-1 4.7-1d Reactor Core Isolation Cooling System - Mechanical Control Diagram 3-47E610-71-1 4.7-1e Reactor Core Isolation Cooling System - Mechanical Control Diagram 1-47E610-71-1 4.7-1f Reactor Core Isolation Cooling System - Flow Diagram 1-47E813-1 4.9-1 Reactor Water Cleanup System Flow Diagram 2-47E810-1 4.9-2 Reactor Water Cleanup Demineralizer Flow Diagram 2-47E837-1 4.9-3 Reactor Water Cleanup System, Mechanical Control Diagram 2-47E610-69-1 4.9-5 Reactor Water Cleanup System - Flow Diagram 3-47E810-1 4.9-6 Reactor Water Cleanup Demineralizer - Flow Diagram 3-47E837-1 4.9-7 Reactor Water Cleanup System - Mechanical Control Diagram 3-47E610-69-1 4.9-8 Reactor Water Cleanup System - Flow Diagram 1-47E810-1 4.9-9 Reactor Water Cleanup Demineralizer - Flow Diagram 1-47E837-1 4.9-10 Reactor Water Cleanup System - Mechanical Control Diagram 1-47E610-69-1 5.2-2a Primary Containment System, Mechanical Control Diagram Sheet 1 1-47E610-64-1 Sheet 2 2-47E610-64-1 Sheet 3 3-47E610-64-1 5.2-2b Primary Containment System, Mechanical Control Diagram 2-47E610-64-2 5.2-2c Primary Containment System, Mechanical Control Diagram 2-47E610-64-3 5.2-2d Primary Containment System - Mechanical Control Diagram 3-47E610-64-2 5.2-2e Primary Containment System - Mechanical Control Diagram 3-47E610-64-3 5.2-2f Primary Containment System - Mechanical Control Diagram 1-47E610-64-2 5.2-2g Primary Containment System - Mechanical Control Diagram 1-47E610-64-3 5.2-6a Containment Inerting System, Mechanical Control Diagram Sheet 1 1-47E610-76-1 Sheet 2 2-47E610-76-4 Sheet 3 2-47E610-76-1 Sheet 4 3-47E610-76-1 Sheet 5 3-47E610-76-4 Sheet 6 0-47E610-76-1 Sheet 7 1-47E1610-76-3 5.2-6b Primary Containment Cooling Temperature Monitoring System - Mechanical Control Diagram 1-47E610-80-1 5.2-6c Primary Containment Cooling Temperature Monitoring System, Mechanical Control Diagram 2-47E610-80-1 5.2-6d Primary Containment Cooling Temperature Monitoring System - Mechanical Control Diagram 3-47E610-80-1 5.2-7 Containment Atmosphere Dilution System, Flow Diagram Sheet 1 1-47E862-1 Sheet 2 2-47E862-1 Sheet 3 3-47E862-1 5.2-8 Containment Atmosphere Dilution System, Mechanical Control Diagram Sheet 1 1-47E610-84-1 Sheet 2 2-47E610-84-1 Sheet 3 3-47E610-84-1

BFN-25 Table 1.3-1 (Contd)

Sheet 4 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

5.3-3a Heating and Ventilation Air Flow Diagram 1-47E865-1 5.3-3b Heating and Ventilating - Standby Gas Treatment System, Flow Diagram 0-47E865-11 5.3-3c Heating and Ventilating Air Flow, Flow Diagram 2-47E2865-12 5.3-3d Heating and Ventilation - Flow Diagram 3-47E865-12 5.3-9 Standby Gas Treatment System, Mechanical Control Diagram 0-47E610-65-1 6.4-1 HPCI System - Mechanical Control Diagram 2-47E610-73-1 6.4-2 Core Spray System, Flow Diagram 2-47E814-1 6.4-3 HPCI System - Mechanical Control Diagram 3-47E610-73-1 6.4-4 Core Spray System - Flow Diagram 3-47E814-1 6.4-5 HPCI System - Mechanical Control Diagram 1-47E610-73-1 6.4-6 Core Spray System, Flow Diagram 1-47E814-1 7.2-1 Reactor Protection System, Single Line 2-45E702-4 7.2-2 Reactor Protection System, Auxiliary Instrument Room Panel 2-791E167RF 7.2-3 Reactor Protection System, Single Line 1-45E701-3 7.2-7a Reactor Protection System Instrument Engineering Diagram 2-729E136-3 7.2-7b Reactor Protection System Instrument Engineering Diagram (Unit 3) 3-729E136-3 7.2-7c Reactor Protection System - Single Line 3-45E703-3 7.2-7d Reactor Protection System Instrument Engineering Diagram (Unit 1) 1-729E136-3 7.2-8 Reactor Protection System, Auxiliary Instrument Room Panel 1-791E167 7.2-9 Reactor Protection System Auxiliary Instrument Room Panel 3-791E167 7.3-1 Nuclear Boiler Flow Diagram Sheet 1 1-47E817-1 Sheet 2 2-47E817-1 Sheet 3 3-47E817-1 7.4-1b High Pressure Coolant Injection System, Flow Diagram Sheet 1 2-47E812-1 Sheet 2 3-47E812-1 Sheet 3 1-47E812-1 7.4-5a ECCS Preferred Pump Logic - Mechanical Control Diagram 2-47E610-75-3 7.4-5d Pre-ACD and Common ACD Signal - Mechanical Control Diagram 2-47E610-75-2 7.4-5i ECCS Preferred Pump - Mechanical Control Diagram 1-47E610-75-3 7.4-5l Pre-ACD and Com ACD Signal - Mechanical Control Diagram 1-47E610-75-2 7.4-5m ECCS Preferred Pump Logic - Mechanical Control Diagram 3-47E610-75-3 7.4-6a Residual Heat Removal System, Flow Diagram Sheet 1 2-47E811-1 Sheet 2 1-47E811-1 Sheet 3 3-47E811-1 7.4-6b Residual Heat Removal System, Mechanical Control Diagram Sheet 1 2-47E2610-74-1 Sheet 2 2-47E2610-74-2 Sheet 3 1-47E610-74-1A

BFN-25 Table 1.3-1 (Contd)

Sheet 5 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

Sheet 4 3-47E610-74-1 Sheet 5 3-47E610-74-2 7.4-7b ECCS Preferred Pump Logic - Mechanical Control Diagram 2-47E610-74-3 7.4-7i Pre-ACD and Com ACD Signal - Mechanical Control Diagram 3-47E610-75-2 7.4-7p ECCS Preferred Pump - Residual Heat Removal System - Mechanical Control Diagram 1-47E610-74-3 7.5-1a Startup Range Neutron Monitoring System, Instrument Engineering Diagram 2-105E1512-1 7.5-1b Startup Range Neutron Monitoring System, Instrument Engineering Diagram 3-105E1512-1 7.5-1c Startup Range Neutron Monitoring System, Instrument Engineering Diagram 1-105E1512RE-1 7.5-2 SRM/IRM Neutron Monitoring Unit 729E946-1 7.5-11a Power Range Neutron Monitoring System, Instrument Engineering Diagram 2-105E1512-2 7.5-11b Power Range Neutron Monitoring System, Instrument Engineering Diagram 3-105E1512-2 7.5-11c Startup Range Neutron Monitoring System - Instrument Engineering Diagram 1-105E1512RE-2 7.5-13 Power Range Neutron Monitoring Unit 0-729E989-1 7.5-23a Neutron Monitoring System Physical Arrangement 0-729E761-1 7.5-23b Neutron Monitoring System Physical Arrangement 0-729E761-2 7.5-23c Neutron Monitoring System Physical Arrangement (Unit 1) 1-729E761-1 7.5-23d Neutron Monitoring System Physical Arrangement (Unit 1) 1-729E761-2 7.8-1 Sheet 1 - Reactor Feedwater System Mechanical Control Diagram 2-47E610-3-1 Sheet 2 - Reactor Water Recirculation System Mechanical Control Diagram 2-47E610-68-1 Sheet 3 - Reactor Feedwater System - Mechanical Control Diagram 3-47E610-3-1 Sheet 4 - Reactor Water Recirculation System - Mechanical Control Diagram 3-47E610-68-1 Sheet 5 - Reactor Feedwater System - Mechanical Control Diagram 1-47E610-3-1 Sheet 6 - Reactor Water Recirculation System - Mechanical Control Diagram 1-47E610-68-1 7.8-3 Reactor Vessel Temperature Monitoring System Physical Arrangement 0-47E600-91 7.10-2 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-1 7.10-3 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-2 7.10-4 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-3 7.10-5 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-1 7.10-6 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-2 7.10-7 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-3 7.10-8 Feedwater Control System - Mechanical Control Diagram 1-47E610-46-1 7.12-2a Radiation Monitoring System Mechanical Control Diagram Sheet 1 2-47E610-90-1 Sheet 2 2-47E610-90-2 Sheet 3 0-47E610-90-4 Sheet 4 0-47E610-90-20 Sheet 5 1-47E610-90-1 Sheet 6 3-47E610-90-1 Sheet 7 0-47E610-90-21 7.12-2b Radiation Monitoring System Mechanical Control Diagram Sheet 2 0-47E610-90-2

BFN-25 Table 1.3-1 (Contd)

Sheet 6 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

Sheet 4 2-47E610-90-3 Sheet 5 1-47E610-90-3 Sheet 6 3-47E610-90-3 8.3-2 Electrical Equipment General Arrangement Plan 0-75W200 8.3-2a Browns Ferry Nuclear Arrangement of Transmission Lines LC48417-1 8.3-4 Gen. 1 and 500-kV Switchyard - Main Single Line 0-45E1504 8.3-5 Generator 2 and 500-kv Switchyard - Main Single Line 0-45E1505 8.3-6 500-kV Switchyard - Main Single Line 0-45E1506 8.3-6a 161-kV Switchyard - Main Single Line 0-45E506 8.4-1a Normal Auxiliary Power System Key Diagram 0-15E500-2 8.4-1b Standby Auxiliary Power System Key Diagram 0-15E500-1 8.4-2 Normal and Standby Auxiliary Power System Key Diagram 3-15E500-3 8.4-3 4160-V Shutdown Auxiliary Power Schematic Diagram 0-45E765-1 8.5-1 Diesel Generator Panel - One-Line Diagram 0-731E747-1 8.5-2 Sheet 1 - Diesel Starting Air System Diesel Generator A Flow and Control Diagram 0-47E861-1 Sheet 2 - Diesel Starting Air System Diesel Generator A Flow Diagram 0-47E861-1A Sheet 3 - Cooling System and Lubricating Oil System Standby Diesel Generator A Flow Diagram 0-47E861-5 Sheet 4 - Diesel Starting Air System Diesel Generator 3A- Flow Diagram 3-47E861-1 Sheet 5 - Diesel Starting Air System Diesel Generator 3A- Flow Diagram 3-47E861-1A Sheet 6 - Cooling System and Lubricating Oil System Diesel Generator 3A - Flow Diagram 3-47E861-5 8.5-3a Fuel Oil System - Flow Diagram 0-47E840-1 8.5-3b Fuel Oil System - Flow Diagram 0-47E840-3 8.5-4a 4160-V Shutdown Board A - Single Line 0-45E724-1 8.5-4b 4160-V Shutdown Board 3EA - Single Line 3-45E724-6 8.5-4c 4160-V Shutdown Board B - Single Line 0-45E724-2 8.5-4d 4160-V Shutdown Board C - Single Line 0-45E724-3 8.5-4e 4160-V Shutdown Board D - Single Line 0-45E724-4 8.5-4f 4160-V Shutdown Board 3EB - Single Line 3-45E724-7 8.5-4g 4160-V Shutdown Board 3EC - Single Line 3-45E724-8 8.5-4h 4160-V Shutdown Board 3ED - Single Line 3-45E724-9 8.5-5 480-V Shutdown Board 2A - Single Line 2-45E749-3 8.5-6 480-V Shutdown Board 2B - Single Line 2-45E749-4 8.5-7a 480-V Reactor MOV Board 2A - Single Line 2-45E751-1 8.5-7b 480-V Reactor MOV Board 2A - Single Line 2-45E751-2 8.5-7c 480-V Reactor MOV Board 1A - Single Line 1-45E751-1 8.5-7d 480-V Reactor MOV Board 1A - Single Line 1-45E751-2 8.5-7e 480-V Reactor MOV Board 3A - Single Line 3-45E751-1 8.5-7f 480-V Reactor MOV Board 3A - Single Line 3-45E751-2 8.5-8a 480-V Reactor MOV Board 2B - Single Line 2-45E751-3 8.5-8b 480-V Reactor MOV Board 2B - Single Line 2-45E751-4

BFN-25 Table 1.3-1 (Contd)

Sheet 7 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

8.5-8c 480-V Reactor MOV Board 1B - Single Line 1-45E751-3 8.5-8d 480-V Reactor MOV Board 1B - Single Line 1-45E751-4 8.5-8e 480-V Reactor MOV Board 3B - Single Line 3-45E751-3 8.5-8f 480-V Reactor MOV Board 3B - Single Line 3-45E751-4 8.5-9a 480-V Reactor MOV Board 2C - Single Line 2-45E751-5 8.5-9b 480-V Reactor MOV Board 2C - Single Line 2-45E751-6 8.5-9c 480-V Reactor MOV Board 3C - Single Line 3-45E751-5 8.5-9d 480-V Reactor MOV Board 3C - Single Line 3-45E751-6 8.5-10 480-V Reactor MOV Board 2D - Single Line 2-45E751-8 8.5-11 480-V Reactor MOV Board 2E - Single Line 2-45E751-11 8.5-11a 480-V Reactor MOV Board 3D - Single Line 3-45E751-9 8.5-11c 480-V Control Bay Vent Board A - Single Line 0-45E736-1 8.5-11d 480-V Control Bay Vent Board B - Single Line 0-45E736-2 8.5-12a 480-V Diesel Auxiliary Board A - Single Line 0-45E732-1 8.5-12b 480-V Diesel Auxiliary Board A - Single Line 0-45E732-2 8.5-12c 480-V Diesel Auxiliary Board 3EA - Single Line 3-45E732-5 8.5-13a 480-V Diesel Auxiliary Board B - Single Line 0-45E732-3 8.5-13b 480-V Diesel Auxiliary Board B - Single Line 0-45E732-4 8.5-13c 480-V Diesel Auxiliary Board 3EB - Single Line 3-45E732-6 8.5-13d 480-V Standby Gas Treatment Bd - Single Line 0-45E733-1 8.5-13e 480-V Diesel Auxiliary Board B - Single Line 0-45E732-8 8.5-25 480-V Shutdown Board 1A - Single Line 1-45E749-1 8.5-26 480-V Shutdown Board 1B - Single Line 1-45E749-2 8.5-27 480-V Shutdown Board 3A - Single Line 3-45E749-5 8.5-28 480-V Shutdown Board 3B - Single Line 3-45E749-6 8.6-1a Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-1 8.6-1b Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-2 8.6-1c Instrument and Controls DC and AC Power Systems Key Diagram (Safeguards Information - Located in Drawing Control) 0-45W710-3 8.6-1d Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-4 8.6-1e Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-5 8.6-1f Instrument and Controls DC and AC Power System Key Diagram 0-45E710-7 8.6-2b Shutdown Boards 250-V Battery and Chargers - Single Line 0-45E709-1 8.6-2c Shutdown Boards 250-V Battery and Chargers - Single Line 3-45E709-2 8.6-3 Engineered Safeguards and RCIC 250-V DC System Separations Scheme Block Diagram 0-731E744-1 8.6-5 DC Board 9-9 One-Line Diagram 0-731E723-1 8.6-6 Control Room DC Board - Single Line 0-55E715-2 8.7-1 Instrumentation and Controls AC Power System, One-Line Diagram 0-731E752-1 8.7-3 Unit Preferred AC Power System, One-Line Diagram 0-731E751-1 8.7-4a AC Board 9-9 One-Line Diagram 0-731E753-1 8.7-4b AC Board 9-9 Preferred and Nonpreferred Loads One-Line Diagram Sheet 1 0-731E753-2

BFN-25 Table 1.3-1 (Contd)

Sheet 8 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

Sheet 2 2-731E753-2 8.7-4c AC Board 9-9 Instrumentation and Control One-Line Diagram Sheet 1 1-731E753-3 Sheet 2 2-731E753-3 Sheet 3 3-731E753-3 8.7-4d Control Room DC Board - Single Line 0-55E715-1 9.2-3a Radwaste System - Flow Diagram 0-47E830-1 9.2-3b Radwaste System - Flow Diagram 0-47E830-2 9.2-3c Radwaste System - Flow Diagram 0-47E830-3 9.2-3d Radwaste System - Flow Diagram 0-47E830-4 9.2-3e Radwaste System - Flow Diagram 0-47E830-5 9.2-3f Radwaste System - Flow Diagram 0-47E830-6 9.2-3g Radwaste System - Flow Diagram 0-47E830-7 9.2-3h Radwaste System - Flow Diagram 0-47E830-8 9.2-3i Radwaste System - Flow Diagram 0-47E830-9 9.2-3j Radwaste System - Mechanical Control Diagram 0-47E610-77-1 9.2-3k Radwaste System - Mechanical Control Diagram 0-47E610-77-2 9.2-3l Radwaste System - Mechanical Control Diagram 0-47E610-77-3 9.2-3m Radwaste System - Mechanical Control Diagram 0-47E610-77-4 9.2-3n Radwaste System - Mechanical Control Diagram 0-47E610-77-5 9.2-3o Radwaste System - Mechanical Control Diagram 0-47E610-77-6 9.2-3p Radwaste System - Mechanical Control Diagram 0-47E610-77-7 9.2-3q Radwaste System - Mechanical Control Diagram 0-47E610-77-8 9.2-3r Radwaste System - Mechanical Control Diagram 0-47E610-77-9 9.2-3s Radwaste System - Mechanical Control Diagram 0-47E610-77-10 9.2-3t Radwaste System - Mechanical Control Diagram 0-47E610-77-11 9.5-1 Offgas System Flow Diagram Sheet 1 2-47E809-2 Sheet 2 2-47E809-3 Sheet 3 3-47E809-2 Sheet 4 3-47E809-3 Sheet 5 1-47E809-2 Sheet 6 1-47E809-3 9.5-2 Offgas System - Flow Diagram 3-47E809-4 9.5-3 Offgas System Flow Diagram 2-47E809-4 9.5-4 Offgas System Flow Diagram 1-47E809-4 10.3-1 High Density Spent Fuel Rack - Unit 1 1-C5445E-102 10.3-2 High Density Spent Fuel Rack - Unit 2 C5445E-103 10-3-3 High Density Spent Fuel Rack - Unit 3 3-C5445E-104 10.5-1a Fuel Pool Cooling System, Flow Diagram 2-47E855-1 10.5-1b Fuel Pool Cooling and Demineralizer System, Mechanical Control Diagram

BFN-25 Table 1.3-1 (Contd)

Sheet 9 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

Sheet 1 2-47E610-78-1 Sheet 2 0-47E610-78-1 Sheet 3 1-47E610-78-1 Sheet 4 3-47E610-78-1 10.5-1c Fuel Pool Cooling System - Flow Diagram 1-47E855-1 10.5-1d Fuel Pool Cooling System - Flow Diagram 3-47E855-1 10.5-2 Fuel Pool Filter Demineralizer, Flow Diagram Sheet 1 2-47E832-1 Sheet 2 0-47E832-1 Sheet 3 1-47E832-1 Sheet 4 3-47E832-1 10.6-1a Reactor Building Closed Cooling Water System, Flow Diagram 1-47E822-1 10.6-1b Reactor Building Closed Cooling Water System, Flow Diagram 2-47E822-1 10.6-1c Reactor Building Closed Cooling Water System - Flow Diagram 3-47E822-1 10.7-1a Raw Cooling Water, Flow Diagram Sheet 1 1-47E844-1 Sheet 2 2-47E844-1 Sheet 3 3-47E844-1 10.7-1b Raw Cooling Water, Flow Diagram Sheet 1 1-47E844-2 Sheet 2 2-47E844-2 Sheet 3 0-47E844-3 Sheet 4 3-47E844-2 Sheet 5 2-47E844-3 Sheet 6 3-47E844-3 Sheet 7 1-47E844-3 10.7-2 Raw Cooling Water System, Mechanical Control Diagram Sheet 1 2-47E610-24-1C Sheet 2 2-47E610-24-1D Sheet 3 0-47E610-24-3 Sheet 4 3-47E610-24-1 Sheet 5 3-47E610-24-2 Sheet 6 1-47E610-24-1 10.9-1a RHR Service Water System, Flow Diagram Sheet 1 1-47E858-1 Sheet 2 2-47E858-1 Sheet 3 3-47E858-1 10.9-1b Raw, Potable, Demineralized Residual Heat Removal, Emergency 0-17W300-5 Equipment Cooling Water and Compressed Air 10.9-2a RHR Service Water System, Mechanical Control Diagram 2-47E610-23-1 10.9-2b RHR Service Water System, Mechanical Control Diagram 0-47E610-23-2 10.9-2c RHR Service Water System, Mechanical Control Diagram 0-47E610-23-3

BFN-25 Table 1.3-1 (Contd)

Sheet 10 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

10.9-2d RHR Service Water System - Mechanical Control Diagram 3-47E610-23-1 10.9-2e RHR Service Water System - Mechanical Control Diagram 1-47E610-23-1 10.10-1a Emergency Equipment Cooling Water, Flow Diagram 1-47E859-1 10.10-1b Emergency Equipment Cooling Water, Flow Diagram 3-47E859-2 10.10-1c Emergency Equipment Cooling Water, Flow Diagram 2-47E859-1 10.10-1d Emergency Equipment Cooling Water - Flow Diagram 3-47E859-1 10.10-2 Emergency Equipment Cooling Water System, Mechanical Control Diagram 1-47E610-67-1 10.10-3 Emergency Equipment Cooling Water System, Mechanical Control Diagram Sheet 1 3-47E610-67-3 Sheet 2 2-47E610-67-2 Sheet 3 3-47E610-67-2 Sheet 4 0-47E610-67-2 10.12-1 Heating and Ventilating Air Flow, Flow Diagram 2-47E865-3 10.12-2a Ventilating and Air Conditioning Air Flow, Flow Diagram 0-47E865-4 10.12-2b Ventilation and Air Conditioning Air Flow, Flow Diagram 2-47E2865-4 10.12-2c Ventilating and Air Conditioning Air Flow, Flow Diagram 3-47E865-4 10.12-3 Heating and Air Conditioning Hot and Chilled Water, Flow Diagram 0-47E866-3 10.12-4 Heating and Ventilating Air Flow, Flow Diagram 0-47E865-6 10.12-5 Heating, Ventilating, and Air Conditioning Air Flow 0-47E865-8 10.12-6 Heating, Ventilating, and Air Conditioning Air Flow 3-47E865-8 10.12-7 Heating and Ventilating Air Flow System - Flow Diagram 1-47E865-3 10.12-8 Heating and Ventilating Air Flow - Flow Diagram 3-47E865-3 10.12-9 Flow Diagram Air Conditioning Chilled Water 0-47E866-9 10.14-1 Control Air System, Mechanical Control Diagram Sheet 1 1-47E610-32-1 Sheet 2 2-47E610-32-1 Sheet 3 3-47E610-32-1 10.14-2a Compressed Air, Station Service, Flow Diagram 0-47E845-1 10.14-2b Compressed Air, Station Service, Flow Diagram 0-47E845-2 10.14-4 Control Air System, Mechanical Control Diagram Sheet 1 - Control Air System - Mechanical Control Diagram 2-47E610-32-2 Sheet 2 - Control Air System - Flow Diagram 2-47E2847-5 Sheet 3 - Control Air System - Mechanical Control Diagram 3-47E610-32-2 Sheet 4 - Control Air System - Mechanical I & C - Flow Diagram 3-47E3847-5 Sheet 5 - Control Air System - Mechanical I & C - Flow Diagram 1-47E1847-6 Sheet 6 - Control Air System - Mechanical Control Diagram 1-47E610-32-2 10.17-1a Sampling and Water Quality Systems, Mechanical Control Diagram 2-47E610-43-1 10.17-1b Sampling and Water Quality Systems, Mechanical Control Diagram 2-47E610-43-2 10.17-1c Sampling and Water Quality Systems, Mechanical Control Diagram Sheet 1 0-47E610-43-3 Sheet 2 2-47E610-43-3

BFN-25 Table 1.3-1 (Contd)

Sheet 11 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

10.17-1d Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-1 10.17-1e Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-2 10.17-1f Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-3 10.21-1 Flow Diagram - PASS 2-47E867-3 10.21-2 Mechanical Control Diagram - PASS 2-47E610-43-5 10.21-3 Sampling and Water Quality System - Flow Diagram 3-47E867-3 10.21-4 Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-5 10.21-5 Sampling and Water Quality System - Flow Diagram 1-47E867-3 10.21-6 Sampling and Water Quality System - Control Diagram 1-47E610-43-5 11.1-1a Main Steam - Flow Diagram 2-47E801-1 11.1-1b Main Steam - Flow Diagram 2-47E801-2 11.1-1c Main Steam - Flow Diagram 3-47E801-1 11.1-1d Main Steam - Flow Diagram 3-47E801-2 11.1-1e Main Steam - Mechanical Flow Diagram 1-47E801-1 11-1-1f Main Steam - Flow Diagram 1-47E801-2 11.6-1 Condenser Circulating Water - Flow Diagram 2-47E831-1 11.6-2 Condenser Circulating Water - Flow Diagram 1-47E831-2 11.6-3 Condenser Circulating Water - Flow Diagram Sheet 1 1-47E831-3 Sheet 2 2-47E831-3 Sheet 3 3-47E831-3 Sheet 4 2-47E831-2 Sheet 5 3-47E831-2 11.6-4 Condenser Circulating Water - Flow Diagram 3-47E831-1 11.6-5 Condenser Circulating Water - Flow Diagram 1-47E831-1 11.6-6 Condenser Circulating Water - Flow Diagram 0-47E831-5 11.7-1 Condensate Demineralizers - Flow Diagram 2-47E833-1 11.7-2 Condensate Demineralizers - Flow Diagram 3-47E833-1 11.7-3 Condensate Demineralizers - Flow Diagram 1-47E833-1 11.8-1 Reactor Feedwater - Flow Diagram Sheet 1 2-47E803-1 Sheet 2 2-47E803-5 Sheet 3 3-47E803-1 Sheet 4 - Mechanical RPV Level Sensing Lines Instruments and Controls 3-47E803-5 Sheet 5 - Reactor Feedwater - Flow Diagram Mechanical RPV Level Sensing Lines Instruments 1-47E803-1 and Controls Sheet 6 - Reactor Feedwater - Flow Diagram Mechanical RPV Level Sensing Lines Instruments 1-47E803-5 and Controls 11.9-1a Condensate - Flow Diagram 2-47E804-1 11.9-1b Condensate Storage and Supply System - Flow Diagram Sheet 1 1-47E818-1 Sheet 2 2-47E818-1 Sheet 3 3-47E818-1

BFN-25 Table 1.3-1 (Contd)

Sheet 12 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

11.9-2 Condensate and Demineralized Water Storage Systems - Mechanical Control Diagram 0-47E610-2-2 11.9-4 Condensate - Flow Diagram 3-47E804-1 11.9-5 Condensate - Flow Diagram 1-47E804-1 12.2-2 Reactor Building General Plans and Sections - Units 1 and 2 41N700 12.2-3 Reactor Building General Plans and Sections - Units 1 and 2 41N701 12.2-4 Reactor Building General Plans and Sections - Units 1 and 2 0-41E702 12.2-5 Reactor Building General Plans and Sections - Unit 3 - Sheet 1 3-41E1000 12.2-6 Reactor Building General Plans and Sections - Unit 3 - Sheet 2 3-41N1001 12.2-21 Reactor Building Structural Steel - Typical Cross Section 0-48E408 12.2-22a Reactor Building Crane Runway - Plans and Sections 0-48E402 12.2-22b Reactor Building Crane General Arrangement 0-PA-2422 12.2-22c Reactor Building Crane - General Arrangement 0-44N220 12.2-23 Reactor Building Sacrificial Shield Wall - Plans and Elevation 0-48E445 12.2-24 Sheet 1 - Structural Steel Drywell Floor Framing - El. 563 ft 1/2 in. 1-48E442-1 Sheet 2 - Structural Steel Drywell Floor Framing - El. 563 ft 1/2 in. 2-48N442 Sheet 3 - Reactor Building Drywell Floor Framing - El. 563 ft 1/2 in. 3-48E442-1 12.2-25 Sheet 1 - Structural Steel Drywell Floor Framing - El. 584 ft 9-1/2 in. 1-48E443-1 Sheet 2 - Reactor Building Drywell Floor Framing - El. 584 ft 9-1/2 in. 2-48E443 Sheet 3 - Reactor Building Drywell Floor Framing - El. 584 ft 9-1/2 in. 3-48N443 12.2-42 Turbine Building, General Plans and Sections, Sheet 1 0-41E200 12.2-43 Turbine Building, General Plans and Sections, Sheet 2 0-41E201 12.2-44 Turbine Building, General Plans and Sections, Sheet 3 41N202 12.2-45 Turbine Building, General Plans and Sections, Sheet 4 0-41E203 12.2-46 Turbine Building, General Plans and Sections, Sheet 5 41N203-1 12.2-47 Turbine Building, Concrete Floor Live Loading, Sheet 1 41N600 12.2-48 Turbine Building, Concrete Floor Live Loading, Sheet 2 0-41E601 12.2-49 Turbine Building, Concrete Floor Live Loading, Sheet 3 41N602 12.2-50 Turbine Building, Structural Steel, Typical Cross Section, 0-48E320 Units 1 and 2 12.2-51 Turbine Building, Structural Steel, Typical Cross Section, Unit 3 48N321 12.2-52 Reinforced Concrete Chimney, Location Plan and Details 0-10E300 12.2-60 Reinforced Concrete Chimney, Foundation Outline 0-10N302 12.2-61 Reinforced Concrete Chimney Interior Above El. 568 0-10N312 12.2-62 Miscellaneous Steel Offgas Stack Exhaust Vent 0-18E211 12.2-64 Concrete, General Plans and Sections, Sheet 1 0-41E569 12.2-65 Concrete, General Plans and Sections, Sheet 2 41N570 12.2-66 Architectual Floor Plans 0-46W426 12.2-67 Architectual Elevations 0-46E425 12.2-69 Concrete - General Outline Features, Sheet 1 0-31E201 12.2-70 Concrete - General Outline Features, Sheet 2 0-31N202 12.2-71a Concrete General Arrangement 0-31E200

BFN-25 Table 1.3-1 (Contd)

Sheet 13 List of FSAR Engineering Drawings Engineering Figure Title Drawing No.

12.2-72a Cooling Tower System, Concrete General Arrangement 0-31E400-1 12.2-72b Cooling Tower System, Concrete General Arrangement 0-31N400-2 12.2-72c Water Supply Cooling Tower System, Concrete General Arrangement 0-31N400-3 12.2-73 Diffuser Pipes from Discharge Conduits 0-31N327 12.2-74 Intake Channel Gate Structure 0-31N410-2 12.2-75b Cooling Tower System - Concrete Arrangement Sheet 1 0-31E418-1 Sheet 2 0-31E418-2 Sheet 3 0-31E418-4 12.2-76 Diesel Generator Building, and Standby Gas Treatment Building, General Plans and Sections 0-41E572 12.2-80 Equipment Access Lock Doors 0-44E235-1 12.2-81 Diesel Generator Building, Concrete General Plans and Sections 3-41E595 12.2-82 Concrete - Offgas Treatment Building, General Plans and Sections 0-10E400 12.2-83 Concrete Pipe Plan, Slab and Walls Outline 0-41E598-1 12.2-84 Gate Structure No. 2, Sheet 1 31E420-1 12.2-85 Gate Structure No. 2, Sheet 2 31E420-2

BFN-25 Table 1.3-2 Sheet 1 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-10E151-7 2.5-S1 0-10E201-7 2.2-4 0-10E300 12.2-52 0-10E400 12.2-82 0-10H243 2.4a, Figure 26 Sheet 1 0-10N302 12.2-60 0-10N312 12.2-61 0-15E500-1 8.4-1b 0-15E500-2 8.4-1a 0-17W300-5 10.9-1b 0-18E211 12.2-62 0-31E200 12.2-71a 0-31E201 12.2-69 0-31E400-1 12.2-72a 0-31E418-1 12.2-75b Sheet 1 0-31E418-2 12.2-75b Sheet 2 0-31E418-4 12.2-75b Sheet 3 0-31N202 12.2-70 0-31N327 12.2-73 0-31N400-2 12.2-72b 0-31N400-3 12.2-72c 0-31N410-2 12.2-74 0-41E200 12.2-42 0-41E201 12.2-43 0-41E203 12.2-45 0-41E569 12.2-64 0-41E572 12.2-76 0-41E598-1 12.2-83 0-41E601 12.2-48 0-41E702 12.2-4 0-44N220 12.2-22c 0-44E235-1 12.2-80 0-45E506 8.3-6a 0-45E709-1 8.6-2b 0-45E710-1 8.6-1a 0-45E710-2 8.6-1b 0-45E710-4 8.6-1d 0-45E710-5 8.6-1e 0-45E710-7 8.6-1f 0-45E724-1 8.5-4a 0-45E724-2 8.5-4c 0-45E724-3 8.5-4d 0-45E724-4 8.5-4e

BFN-28 Table 1.3-2 (Contd)

Sheet 2 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-45E1504 8.3-4 0-45E1505 8.3-5 0-45E1506 8.3-6 0-45E732-1 8.5-12a 0-45E732-2 8.5-12b 0-45E732-3 8.5-13a 0-45E732-4 8.5-13b 0-45E732-8 8.5-13e 0-45E733-1 8.5-13d 0-45E736-1 8.5-11c 0-45E736-2 8.5-11d 0-45E765-1 8.4-3 0-45W710-3 8.6-1c 0-46E425 12.2-67 0-46W426 12.2-66 0-47E200-2 1.6-2 0-47E200-4 1.6-4 0-47E200-5 1.6-5 0-47E200-6 1.6-6 0-47E200-7 1.6-7 0-47E200-9 1.6-9 0-47E200-10 1.6-10 0-47E200-12 1.6-12 0-47E200-15 1.6-15 0-47E200-16 1.6-16 0-47E200-17 1.6-17 0-47E200-18 1.6-18 0-47E200-19 1.6-19 0-47E215-3 1.6-25 0-47E600-121 2.5-17 0-47E600-91 7.8-3 0-47E610-2-2 11.9-2 0-47E610-23-2 10.9-2b 0-47E610-23-3 10.9-2c 0-47E610-24-3 10.7-2 Sheet 3 0-47E610-43-3 10.17-1c Sheet 1 0-47E610-65-1 5.3-9 0-47E610-67-2 10.10-3 Sheet 4 0-47E610-76-1 5.2-6a Sheet 6 0-47E610-77-1 9.2-3j 0-47E610-77-10 9.2-3s 0-47E610-77-11 9.2-3t 0-47E610-77-2 9.2-3k 0-47E610-77-3 9.2-3l

BFN-28 Table 1.3-2 (Contd)

Sheet 3 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-47E610-77-4 9.2-3m 0-47E610-77-5 9.2-3n 0-47E610-77-6 9.2-3o 0-47E610-77-7 9.2-3p 0-47E610-77-8 9.2-3q 0-47E610-77-9 9.2-3r 0-47E610-78-1 10.5-1b Sheet 2 0-47E610-90-2 7.12-2b Sheet 2 0-47E610-90-4 7.12-2a Sheet 3 0-47E610-90-20 7.12-2a Sheet 4 0-47E610-90-21 7.12-2a, Sheet 7 0-47E800-1 1.6-30 0-47E800-2 1.3-2 0-47E820-1 3.4-8b 0-47E830-1 9.2-3a 0-47E830-2 9.2-3b 0-47E830-3 9.2-3c 0-47E830-4 9.2-3d 0-47E830-5 9.2-3e 0-47E830-6 9.2-3f 0-47E830-7 9.2-3g 0-47E830-8 9.2-3h 0-47E830-9 9.2-3i 1-47E831-1 11.6-5 0-47E831-5 11.6-6 0-47E832-1 10.5-2 Sheet 2 0-47E840-1 8.5-3a 0-47E840-3 8.5-3b 0-47E844-3 10.7-1b Sheet 3 0-47E845-1 10.14-2a 0-47E845-2 10.14-2b 0-47E861-1 8.5-2 Sheet 1 0-47E861-1A 8.5-2 Sheet 2 0-47E861-5 8.5-2 Sheet 3 0-47E865-11 5.3-3b 0-47E865-4 10.12-2a 0-47E865-6 10.12-4 0-47E865-8 10.12-5 0-47E866-3 10.12-3 0-47E866-9 10.12-9 0-47W215-1 1.6-23 0-47W215-2 1.6-24 0-47W220-1 1.6-26 0-48E320 12.2-50

BFN-28 Table 1.3-2 (Contd)

Sheet 4 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-48E402 12.2-22a 0-48E408 12.2-21 0-48E445 12.2-23 0-55E715-1 8.7-4d 0-55E715-2 8.6-6 0-729E286-1 7.9-23 0-729E761-1 7.5-23a 0-729E761-2 7.5-23b 0-729E989-1 7.5-13 0-731E723-1 8.6-5 0-731E744-1 8.6-3 0-731E747-1 8.5-1 0-731E751-1 8.7-3 0-731E752-1 8.7-1 0-731E753-1 8.7-4a 0-731E753-2 8.7-4b Sheet 1 0-75W200 8.3-2 0-PA-2422 12.2-22b 1-45E701-3 7.2-3 1-45E749-1 8.5-25 1-45E749-2 8.5-26 1-45E751-1 8.5-7c 1-45E751-2 8.5-7d 1-45E751-3 8.5-8c 1-45E751-4 8.5-8d 1-47E1847-6 10.14-4 Sheet 5 1-47E200-3 1.6-3 Sheet 1 1-47E200-8 1.6-8 Sheet 1 1-47W401-5 4.4-8 1-47E610-3-1 7.8-1, Sheet 5 1-47E610-23-1 10.9-2e 1-47E610-24-1 10.7-2 Sheet 6 1-47E610-32-1 10.14-1 Sheet 1 1-47E610-32-2 10.14-4 Sheet 6 1-47E610-43-5 10.21-6 1-47E610-46-1 7.10-8 1-47E610-63-1 3.8-8 1-47E610-64-1 5.2-2a Sheet 1 1-47E610-64-2 5.2-2f 1-47E610-64-3 5.2-2g 1-47E610-67-1 10.10-2 1-47E610-68-1 7.8-1, Sheet 6 1-47E610-69-1 4.9-10 1-47E610-71-1 4.7-1e

BFN-28 Table 1.3-2 (Contd)

Sheet 5 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 1-47E610-73-1 6.4-5 1-47E610-74-1A 7.4-6b Sheet 3 1-47E610-75-2 7.4-5l 1-47E610-75-3 7.4-5i 1-47E610-74-3 7.4-7p 1-47E610-76-1 5.2-6a Sheet 1 1-47E610-78-1 10.5-1b Sheet 3 1-47E610-80-1 5.2-6b 1-47E610-84-1 5.2-8 Sheet 1 1-47E610-85-1 3.4-8a, Sheet 5 1-47E610-90-1 7.12-2a Sheet 5 1-47E610-90-3 7.12-2b Sheet 5 1-47E801-1 11.1-1e 1-47E801-2 11-1-1f 1-47E803-1 11-8-1 Sheet 5 1-47E803-5 11.8-1 Sheet 6 1-47E804-1 11.9-5 1-47E809-2 9.5-1 Sheet 5 1-47E809-3 9.5-1 Sheet 6 1-47E809-4 9.5-4 1-47E810-1 4.9-8 1-47E811-1 7.4-6a Sheet 2 1-47E812-1 7.4-1b, Sheet 3 1-47E813-1 4.7-1f 1-47E814-1 6.4-6 1-47E817-1 4.3-2a Sheet 1 1-47E818-1 11.9-1b Sheet 1 1-47E822-1 10.6-1a 1-47E831-1 11.6-5 1-47E831-2 11.6-2 1-47E831-3 11.6-3 Sheet 1 1-47E832-1 10.5-2 Sheet 3 1-47E833-1 11.7-3 1-47E837-1 4.9-9 1-47E844-1 10.7-1a Sheet 1 1-47E844-2 10.7-1b Sheet 1 1-47E844-3 10.7-1b Sheet 7 1-47E854-1 3.8-3 1-47E855-1 10.5-1c 1-47E858-1 10.9-1a Sheet 1 1-47E859-1 10.10-1a 1-47E862-1 5.2-7 Sheet 1 1-47E865-1 5.3-3a 1-47E865-3 10.12-7

BFN-28 Table 1.3-2 (Contd)

Sheet 6 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 1-47E867-3 10.21-5 1-47E1610-85-2 3.4-8g 1-47E1610-85-5 3.4-8h 1-47E1610-76-3 5.2-6a, Sheet 7 1-47K1110-32 1.6-29 Sheet 3 1-48E442-1 12.2-24 Sheet 1 1-48E443-1 12.2-25 Sheet 1 1-104R935-1 4.2-4 1-105E1512RE-1 7.5-1c 1-105E1512RE-2 7.5-11c 1-729E136-3 7.2-7d 1-729E229-1 4.5-1 1-729E761-1 7.5-23c 1-729E761-2 7.5-23d 1-729-E761-1 7.5.23c 1-729E761-2 7.5.23d 1-731E753-3 8.7-4c Sheet 1 1-791E167 7.2-8 1-C5445E-102 10.3-1 2-104R935-1 4.2-1 2-105E1512-1 7.5-1a 2-105E1512-2 7.5-11a 2-45E702-4 7.2-1 2-45E749-3 8.5-5 2-45E749-4 8.5-6 2-45E751-1 8.5-7a 2-45E751-11 8.5-11 2-45E751-2 8.5-7b 2-45E751-3 8.5-8a 2-45E751-4 8.5-8b 2-45E751-5 8.5-9a 2-45E751-6 8.5-9b 2-45E751-8 8.5-10 2-47E200-3 1.6-3 Sheet 2 2-47E200-8 1.6-8 Sheet 2 2-47E610-23-1 10.9-2a 2-47E610-24-1C 10.7-2 Sheet 1 2-47E610-24-1D 10.7-2 Sheet 2 2-47E610-3-1 7.8-1 Sheet 1 2-47E610-32-1 10.14-1 Sheet 2 2-47E610-32-2 10.14-4 Sheet 1 2-47E610-43-1 10.17-1a 2-47E610-43-2 10.17-1b 2-47E610-43-3 10.17-1c Sheet 2

BFN-28 Table 1.3-2 (Contd)

Sheet 7 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 2-47E610-43-5 10.21-2 2-47E610-46-1 7.10-2 2-47E610-46-2 7.10-3 2-47E610-46-3 7.10-4 2-47E610-63-1 3.8-2 2-47E610-64-1 5.2-2a Sheet 2 2-47E610-64-2 5.2-2b 2-47E610-64-3 5.2-2c 2-47E610-67-2 10.10-3 Sheet 2 2-47E610-68-1 7.8-1 Sheet 2 2-47E610-69-1 4.9-3 2-47E610-71-1 4.7-1b 2-47E610-73-1 6.4-1 2-47E610-74-3 7.4-7b 2-47E610-75-2 7.4-5d 2-47E610-75-3 7.4-5a 2-47E610-76-1 5.2-6a Sheet 3 2-47E610-76-4 5.2-6a Sheet 2 2-47E610-78-1 10.5-1b Sheet 1 2-47E610-80-1 5.2-6c 2-47E610-84-1 5.2-8 Sheet 2 2-47E610-85-1 3.4-8a Sheet 1 2-47E610-90-1 7.12-2a Sheet 1 2-47E610-90-2 7.12-2a Sheet 2 2-47E610-90-3 7.12-2b Sheet 4 2-47E801-1 11.1-1a 2-47E801-2 11.1-1b 2-47E803-1 11.8-1 Sheet 1 2-47E803-5 11.8-1 Sheet 2 2-47E804-1 11.9-1a 2-47E809-2 9.5-1 Sheet 1 2-47E809-3 9.5-1 Sheet 2 2-47E809-4 9.5-3 2-47E810-1 4.9-1 2-47E811-1 7.4-6a Sheet 1 2-47E812-1 7.4-1b Sheet 1 2-47E813-1 4.7-1a 2-47E814-1 6.4-2 2-47E817-1 4.3-2a Sheet 2 2-47E818-1 11.9-1b Sheet 2 2-47E820-2 3.4-8c 2-47E822-1 10.6-1b 2-47E831-1 11.6-1 2-47E831-2 11.6-3 Sheet 4

BFN-28 Table 1.3-2 (Contd)

Sheet 8 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 2-47E831-3 11.6-3 Sheet 2 2-47E832-1 10.5-2 Sheet 1 2-47E833-1 11.7-1 2-47E837-1 4.9-2 2-47E844-1 10.7-1a Sheet 2 2-47E844-2 10.7-1b Sheet 2 2-47E844-3 10.7-1b Sheet 5 2-47E854-1 3.8-1 2-47E855-1 10.5-1a 2-47E858-1 10.9-1a Sheet 2 2-47E859-1 10.10-1C 2-47E862-1 5.2-7 Sheet 2 2-47E865-3 10.12-1 2-47E867-3 10.21-1 2-47E2610-74-1 7.4-6b Sheet 1 2-47E2610-74-2 7.4-6b Sheet 2 2-47E2610-85-2 3.4-8a Sheet 2 2-47E2610-85-5 3.4-8d 2-47E2847-5 10.14-4 Sheet 2 2-47E2865-12 5.3-3c 2-47E2865-4 10.12-2b 2-47K1110-13 1.6-29 Sheet 1 2-47W401-5 4.4-6 2-48E443 12.2-25 Sheet 2 2-48N442 12.2-24 Sheet 2 2-729E136-3 7.2-7a 2-729E229-4 4.5-2 2-731E753-2 8.7-4b Sheet 2 2-731E753-3 8.7-4c Sheet 2 2-791E167RF 7.2-2 3-C5445E-104 10.3-3 3-104R935-1 4.2-3 3-105E1512-1 7.5-1b 3-105E1512-2 7.5-11b 3-15E500-3 8.4-2 3-41E1000 12.2-5 3-41E595 12.2-81 3-41N1001 12.2-6 3-45E703-3 7.2-7c 3-45E709-2 8.6-2c 3-45E724-6 8.5-4b 3-45E724-7 8.5-4f 3-45E724-8 8.5-4g 3-45E724-9 8.5-4h

BFN-28 Table 1.3-2 (Contd)

Sheet 9 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-45E732-5 8.5-12c 3-45E732-6 8.5-13c 3-45E749-5 8.5-27 3-45E749-6 8.5-28 3-45E751-1 8.5-7e 3-45E751-2 8.5-7f 3-45E751-3 8.5-8e 3-45E751-4 8.5-8f 3-45E751-5 8.5-9c 3-45E751-6 8.5-9d 3-45E751-9 8.5-11a 3-47E200-11 1.6-11 3-47E200-13 1.6-13 3-47E200-14 1.6-14 3-47E200-8 1.6-8 Sheet 3 3-47E220-2 1.6-27 3-47E401-5 4.4-7 3-47E610-23-1 10.9-2d 3-47E610-24-1 10.7-2 Sheet 4 3-47E610-24-2 10.7-2 Sheet 5 3-47E610-3-1 7.8-1 Sheet 3 3-47E610-32-1 10.14-1 Sheet 3 3-47E610-32-2 10.14-4 Sheet 3 3-47E610-43-1 10.17-1d 3-47E610-43-2 10.17-1e 3-47E610-43-3 10.17-1f 3-47E610-43-5 10.21-4 3-47E610-46-1 7.10-5 3-47E610-46-2 7.10-6 3-47E610-46-3 7.10-7 3-47E610-63-1 3.8-6 3-47E610-64-1 5.2-2a Sheet 3 3-47E610-64-2 5.2-2d 3-47E610-64-3 5.2-2e 3-47E610-67-2 10.10-3 Sheet 3 3-47E610-67-3 10.10-3 Sheet 1 3-47E610-68-1 7.8-1 Sheet 4 3-47E610-69-1 4.9-7 3-47E610-71-1 4.7-1d 3-47E610-73-1 6.4-3 3-47E610-74-1 7.4-6b Sheet 4 3-47E610-74-2 7.4-6b Sheet 5 3-47E610-75-2 7.4-7i 3-47E610-75-3 7.4-5m

BFN-28 Table 1.3-2 (Contd)

Sheet 10 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-47E610-76-1 5.2-6a Sheet 4 3-47E610-76-4 5.2-6a Sheet 5 3-47E610-78-1 10.5-1b Sheet 4 3-47E610-80-1 5.2-6d 3-47E610-84-1 5.2-8 Sheet 3 3-47E610-85-1 3.4-8a Sheet 3 3-47E610-85-5 3.4-8f 3-47E610-90-1 7.12-2a Sheet 6 3-47E610-90-3 7.12-2b Sheet 6 3-47E801-1 11.1-1c 3-47E801-2 11.1-1d 3-47E803-1 11.8-1 Sheet 3 3-47E803-5 11.8-1 Sheet 4 3-47E804-1 11.9-4 3-47E809-2 9.5-1 Sheet 3 3-47E809-3 9.5-1 Sheet 4 3-47E809-4 9.5-2 3-47E810-1 4.9-5 3-47E811-1 7.4-6a Sheet 3 3-47E812-1 7.4-1b Sheet 2 3-47E813-1 4.7-1c 3-47E814-1 6.4-4 3-47E817-1 4.3-2a Sheet 3 3-47E818-1 11.9-1b Sheet 3 3-47E820-2 3.4-8e 3-47E822-1 10.6-1c 3-47E831-1 11.6-4 3-47E831-2 11.6-3 Sheet 5 3-47E831-3 11.6-3 Sheet 3 3-47E832-1 10.5-2 Sheet 4 3-47E833-1 11.7-2 3-47E837-1 4.9-6 3-47E844-1 10.7-1a Sheet 3 3-47E844-2 10.7-1b Sheet 4 3-47E844-3 10.7-1b Sheet 6 3-47E854-1 3.8-5 3-47E855-1 10.5-1d 3-47E858-1 10.9-1a Sheet 3 3-47E859-1 10.10-1d 3-47E859-2 10.10-1b 3-47E861-1 8.5-2 Sheet 4 3-47E861-1a 8.5-2 Sheet 5 3-47E861-5 8.5-2 Sheet 6 3-47E862-1 5.2-7 Sheet 3

BFN-28 Table 1.3-2 (Contd)

Sheet 11 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-47E865-12 5.3-3d 3-47E865-3 10.12-8 3-47E865-4 10.12-2c 3-47E865-8 10.12-6 3-47E867-3 10.21-3 3-47E3610-85-2 3.4-8a, Sheet 4 3-47E3847-5 10.14-4 Sheet 4 3-47K1110-13 1.6-29 Sheet 2 3-48E442-1 12.2-24 Sheet 3 3-48N443 12.2-25 Sheet 3 3-729E136-3 7.2-7b 3-729E229-4 4.5-3 3-731E753-3 8.7-4c Sheet 3 3-791E167 7.2-9 104R900 1.3-1 10H237 2.4A Fig 25 10H242 2.4A Fig 26, Sh 2 31E420-1 12.2-84 31E420-2 12.2-85 41N202 12.2-44 41N203-1 12.2-46 41N570 12.2-65 41N600 12.2-47 41N602 12.2-49 41N700 12.2-2 41N701 12.2-3 41N703 2.5-19 47E200-1 1.6-1 48N321 12.2-51 729E946-1 7.5-2 C5445E-103 10.3-2 LC48417-1 8.3-2a

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-t-AMENDMENT 16 BROWNS FERRY NVCLEAR PLANT

l NAL SAFETY 4NA.l vs IS REPORT FIGURE 1 .3-1

600~ Z-0083Lt-O V'l L9 SYMBOLS FLOW DIAGRAM DRAWING INDEX THROTTLING VALVE (GATE) -El- INSTRUMENT POT 47E800-SERIES ......... FLOW DIAGRAM GENERAL PLANT SYSTEMS & SYMBOLS 47E801-SERIES ......... FLOW DIAGRAM MAIN STEAM H 47E802-SERIES ......... FLOW DIAGRAM EXTRACTION STEAM THROTTLING VALVE (GLOBE) FILTER 47E803-SERIES ......... FLOW DIAGRAM REACTOR FEEDWATER 47E804-SERIES ......... FLOW DIAGRAM CONDENSATE 3-WAY BLOCK VALVE 47E8D5-SERIES ......... FLOW DIAGRAM HEATER DRAINS, VENTS & MISC PIPING ADJUSTABLE FIXED 47E807-SERIES ......... FLOW DIAGRAM TURBINE DRAINS & MISC PIPING LOUVERS LOUVERS 47E808-SERIES ......... FLOW DIAGRAM HYDROGEN WATER CHEMISTRY SYSTEM

---t><l- GATE VALVE (NORMALLY OPEN) 47E809-SERIES ......... FLOW DIAGRAM OFF-GAS SYSTEM COMBINATION AIR FILTER AND 47E810-SERIES ......... FLOW DIAGRAM REACTOR WATER CLEANUP SYSTEM PRESSURE REGULATOR 47E811-SERIES ......... FLOW DIAGRAM RESIDUAL HEAT REMOVAL SYSTEM GATE VALVE (NORMALLY CLOSED)

~

- BACKFLOW PREVENTER ASSEMBLY: COMMON BODY DOUBLE CHECK VALVE TYPE WITH SHUTOFF f-f-8)=)-)------jJ PRIMARY CONTAINMENT PENETRATION OR WALL AND FLOOR SLEEVES 47E812-SERIES ......... FLOW DIAGRAM HIGH PRESSURE C[X)LANT INJECTION SYSTEM 47E813-SERIES ......... FLOW DIAGRAM REACTOR CORE ISOLATION COOLING SYSTEM 47E814-SERIES ......... FLOW DIAGRAM CORE SPRAY SYSTEM GATE VALVE (NORMALLY CLOSED)

WITH ORIFICE THROUGH WEDGE TEST ANNUALLY VALVE ON EACH SIDE-REQUIRES 3 UNIDS

--mm- BREAKDOWN ORIFICE NOTE:

CONTAINMENT PENETRATIONS DENOTED BY X-NUMBERS IE. X-29 47E815-SERIES ......... FLOW DIAGRAM AUXILIARY BOILER SYSTEM 47E816-SERIES .... , .... FLOW DIAGRAM LUBRICATING OIL SYSTEM 47E817-SERIES ......... FLOW DIAGRAM REACTOR WATER RECIRCN, DRAINS, VENTS & BLOWDOWN SYSTEMS 47E818-SERIES ......... FLOW DIAGRAM CONDENSATE STORAGE AND SUPPLY SYSTEM GLOBE VALVE (NORMALLY OPEN) - BACKFLOW PREVENTER ASSEMBLY; DIFFERENTIAL PRESSURE (DP) TYPE, COMMON BODY DOUBLE --§-- STRAIGHTENING VANES TUBING TO PIPING TRANSITION 47E819-SERIES ......... FLOW DIAGRAM EMERGENCY HIGH PRESSURE MAKEUP PUMP SYSTEM 47E820-SERIES ......... FLOW DIAGRAM CONTROL ROD DRIVE HYDRAULIC SYSTEM

~

CHECK VALVE (WITH DP RELIEF VALVE) AND SHUTOFF VALVE ON EACH SIDE-REQUIRES 3 UNIDS

,,,, 1/2" Cl .SOD

,,,, (FRACTIONAL SIZE DENOTES PIPE AND 47E821-SERIES ......... FLOW DIAGRAM CHEMICAL CLEANING SYSTEM 47E822-SERIES ......... FLOW DIAGRAM REACTOR BUILDING CLOSED C[X)LING WATER SYSTEM DECIMAL SIZE DENOTES TUBING) 47E825-SERIES ......... FLOW DIAGRAM BREATHING AIR SYSTEM Ll GLOBE VALVE (NORMALLY CLOSED) 47E830-SERIES ......... FLOW DIAGRAM RADWASTE SYSTEM TEST ANNUALLY DESUPERHEATER 47E831-SERIES ......... FLOW DIAGRAM CONDENSATE CIRCULATING WATER SYSTEM 8

47E832-SERIES ......... FLOW DIAGRAM FUEL POOL FILTER/DEMINERALIZING SYSTEM BALANCING VAL VE

, BACKFLOW PREVENTER ASSEMBLY; SEPARATE BODY DEVICE MOUNTED ON MAIN CONTROL ROOM PANEL 47E833-SERIES ......... FLOW DIAGRAM CONDENSATE DEMIN~RALIZERS 47E834-SERIES ......... FLOW DIAGRAM MAKEUP WATER TREATMENT SYSTEM 47E835-SERIES .. ,, ..... FLOW DIAGRAM POTABLE WATER DISTRIBUTION SYSTEM G

TEST ANNUALLY DOUBLE CHECK VALVES WITH SHUTOFF VALVE ON EACH SIDE-REQUIRES 4 UNIDS ---l I BLIND FLANGE (SEE NOTE 3) 47E836-SERIES ......... FLOW DIAGRAM RAW SERVICE WATER & FIRE PROTECTION SYSTEM 47E837-SERIES ......... FLOW DIAGRAM REACTOR WATER CLEANUP DEMINERALIZER SLIDE OR BLAST GATE VALVE 47E838-SERIES ......... FLOW DIAGRAM RAW WATER CHLORINATION SYSTEM

-lNf- FLEXIBLE CONNECTION f-><,__--i><---f CAPILLARY TUBING 47E839-SERIES ......... FLOW DIAGRAM HYPOCHLORITE SYSTEM 47E84D-SERIES ......... FLOW DIAGRAM FUEL OIL SYSTEM (SEE NOTE 3) 47E841-SERIES ......... FLOW DIAGRAM GLAND SEAL WATER SYSTEM 47E842-SERIES ......... FLOW DIAGRAM INSULATING OIL SYSTEM THREE WAY VALVE MAIN PROCESS LINES 47E843-SERIES ......... FLOW DIAGRAM CO2 STORAGE, FIRE PROT & PURGING SYSTEM 47E844-SERIES ......... FLOW DIAGRAM RAW CCXJLING WATER SYSTEM (FLOW DIRECTION) 47E845-SERIES ......... FLOW DIAGRAM COMPRESSED AIR-STATION SERVICE SYSTEM PNEUMATIC TUBING 47E846-SERIES ......... FLOW DIAGRAM DEMINERALIZER BACKWASH AIR SYSTEM DIAPHRAGM OPERATOR 'J< ,('

(' N FOUR WAY VALVE .J " (SEE NOTE 3) 47E847-SERIES ......... FLOW DIAGRAM CONTROL AIR SYSTEM 47E848-SERIES ......... FLOW DIAGRAM VACUUM PRIMING SYSTEM AUXILIARY PROCESS LINES 47E849-SERIES ......... FLOW DIAGRAM HYDROGEN SYSTEM FOR GENERATOR COOLING 47E850-SERIES ......... FLOW DIAGRAM FIRE PROTECTION & RAW SERVICE WATER SYSTEM CHECK VALVE 47E851-SERIES ......... FLOW DIAGRAM DRAINAGE-TURBINE. DIESEL GENERATOR. OFFICE ELECTRIC MOTOR OPERATOR f' :n "'

,.:: .J(' NITROGEN PIPE AND SERVICE BUILDINGS AND STACK

.J '" OR TUBING 47E852-SERIES ......... FLOW DIAGRAM DRAINAGE-REACTOR BLDG (SEE NOTE 3) 47E853-SERIES ......... FLOW DIAGRAM AUXILIARY BOILER FEEDWATER SECONDARY TREATMENT STOP CHECK VALVE CLOSED DRAIN (CRW) 47E854-SERIES ......... FLOW DIAGRAM STANDBY LIQUID CONTROL SYTEM 47E855-SERIES ......... FLOW DIAGRAM FUEL POOL COOLING SYSTEM 47E856-SERIES ......... FLOW DIAGRAM DEMINERALIZED WATER SYSTEM CONSTANT HEAD POT 47E857-SERIES ......... FLOW DIAGRAM CONDENSER TUBE CLEANING SYSTEM ANGLE VALVE (NORMALLY OPEN) VALVE POSITIONER 47E858-SERIES ......... FLOW DIAGRAM RHR SERVICE WATER SYSTEM OPEN DRAIN (ORI) 47E859-SERIES .. ,, ..... FLOW DIAGRAM EMERGENCY EQPT COOLING WATER SYSTEM 47E860-SERIES ......... FLOW DIAGRAM CONTAINMENT INERTING SYSTEM 47E861-SERIES ......... FLOW DIAGRAM DIESEL STARTING AIR SYSTEM 47E862-SERIES ......... FLOW DIAGRAM CONTAINMENT ATt.OSPHERE DILUTION SYSTEM F

ANGLE VALVE {NORMALLY CLOSED) ( ~ ( 47E865-SERIES ......... FLOW DIAGRAM HVAC AIR FLOW SYSTEM EXPLOSIVE OPERATOR HEAT TRACING

} ~ } 47E866-SERIES ......... FLOW DIAGRAM BUILDING HEATING & AIR CONDITIONING HOODED SAMPLE STATION 47E870-SERIES ......... FLOW DIAGRAM WARM WATER WASTE HEAT DEMONSTRATION 47E873-SERIES ......... FLOW DIAGRAM AUXILIARY DECAY HEAT RE~AL SYSTEM ANGLE GLOBE VALVE 47E885-SERIES ......... FLOW DIAGRAM GENERATOR COOLING SYSTEMS CYLINDER OPERATOR FAN WITH t.OTOR DAMPER MANUAL WITH RELIEF VALVE f-/-1 LOCKING QUADRANT NOTES:

SOLENOID OPERATOR CONTROL AIR SUPPLY. PRESSURE 1. DELETED TO BE SET AT MANUFACTURER'S REQUIREMENT FOR THE DEVICE 2. DELETED ANGLE RELIEF VALVE HAND CONTROLLED VALVE BEING SUPPLIED 3. SEEDS E18.3.3 FOR ADDITIONAL INSTRUMENT NOMENCLATURE.

FIRE DAMPER

4. INFORMATION ON THIS DRAWING ORIGINALLY DEPICTED ON 47W801-1.

FLOAT OPERATED VALVE GRAB SAMPLE , ~~

~ ~

,f-~~~~~--1,

AIR FLOW IN DUCT, AIR VOLUME (CFM)

SHOULD BE USED TO INDICATE DIRECTIONAL y ATMOS VENT/EMERGENCY RELIEF

5. GRILLE WILL BE DEFINED AS HAVING FIXED DIFFUSERS (NOT ADJUSTABLE MANUALLY OR BY MOTOR).
6. REGISTER WILL BE DEFINED AS HAVING ADJUSTABLE DIFFUSERS (MANUALLY OPERATED).

FLOW WITH AND/OR WITHOUT CFM NOTED 7. THIS DRAWING ESTABLISHES SYMBOLS TO BE USED ON THE BROWNS FERRY CAD RESTORED DRAWINGS. SYMBOLIC REPRESENTATION DOES NOT DEPICT NEEDLE VALVE PINCH VALVE SIGHT FLOW INDICATOR (SEE NOTE 3)

BACK DRAFT DAMPER (IN DUCTWORK) I~ I I SINGLE PASS HEAT EXCHANGER EXACT SCALE.

8. USE OF SYMBOLS NOT DEPICTED ON THIS DRAWING SHALL FOLLOW THE GUIDELINES ADDRESSED IN DES 7.01-GENERAL DRAFTING PRACTICES.
9. DELETED E

BACK DRAFT DAMPER DOUBLE PASS (AT DOOR-GRILLE. WALL ALTITUDE VALVE / ' OPENING, NOT IN DUCT- HEAT EXCHANGER FT WORK, ETC.)

ORIFICE OR NOZZLE TYPE FLOW ELEMENT DIAPHRAGM VALVE WITH TRANSMITTER (SEE NOTE 3)

FE DCXJR/WALL TRANSFER GRILLE BUTTERFLY VALVE (NORMALLY OPEN) , FLAME ARRESTOR RETURN OR EXHAUST BUTTERFLY VALVE (NORMALLY CLOSED) GRILLE PRESSURE INDICATOR (SEE NOTE 3) CONDENSING POT MUD VALVE FIRE HOSE COMBINATION --v"-1~ SUPPLY GRILLE REVERSE CURRENT VALVE PLUG VALVE PRESS. RESTRICTING ANGLE VALVE SPRINKLER/NOZZLE/ORIFICE

---v'-~ RETURN OR EXHAUST DUCTWORK OPENING (NO GRILLE/REGISTER) 8 INSULATED FLANGE D

ODORIZER C? AIR RELEASE VALVE STOP COCK (CO2 SYSTEM)

---v'-~ SUPPLY DUCTWORK OPENING (NO REGISTER/DIFFUSER)

POSITIVE DISPLACEMENT EXCESS FLOW CHECK VALVE C::s PUMP FIRE HYDRANT

---v'-1~ SUPPLY REGISTER WRENCH OPERATED VALVE FIRE Ig__.....-------1 COOLING COIL SERVICE VALVE RETURN REGISTER AIR WRENCH OPERATED VALVE IN CABINET (OPTIONAL)

HEATING COIL AIR FLOW DIRECTION (DIFFUSER)

VACUUM RELIEF VALVE FIRE HOSE ANGLE VALVE W/ PRESSURE RESTRICTING C VALVE / ENDUCTOR/EJECTOR QUICK OPENING VALVE -

SPRING LOADED QUICK DISCONNECT WITH DOUBLE END SHUTOFF

-to-r BALL VAL VE QUICK DISCONNECT WITH SINGLE DIAPHRAGM OPERATED DAMPER FAIL AS IS END SHUTOFF r-1 VACUUM CLEANING INLET VALVE

/

VACUUM CLEANING INLET VALVE QUICK DISCONNECT (FLOOR) ' ~,,

TUBE UNION OR CONNECTOR ANGLE CHECK VALVE , V--f' 0 ,

DIAPHRAGM OPERATED DAMPER FAIL OPEN IN PROCESS LINE FLANGE CONNECTION REDUCER OR REDUCTION POWER CONTROL VALVE ORIFICE OR NOZZLE B

'"'c.-;::,."

RUPTURE DISC ,

V--f', DIAPHRAGM OPERATED DAMPER FAIL CHECK VALVE W/ ORIFICE

~<1111[ SERVICE CONNECTION X

CLOSED IN PROCESS LINE AMENDMENT 28 J CAP OR PLUG (SCREWED OR WELDED)

GENERAL PRESSURE REDUCING REGULATOR FAN (VENTILATION)

WITH EXTERNAL PRESSURE TAP -0]- IMPULSE TRAP WITH STRAINER BROWNS FERRY NUCLEAR PLANT SELF CONTAINED PRESSURE SINGLE STRAINER A/C UNIT FINAL SAFETY ANALYSIS REPORT REGULATING VALVE TWIN STRAINER GENERAL SYMBOLS A GATE VALVE WITH GLAND SEAL Hi() {} BUCKET TRAP PUMP FLOW DIAGRAM FIGURE 1 . 3-2 8 7 6 5

  • 4 3

BFN-16 1.4 CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION 1.4.1 Introduction To fully evaluate the many aspects of the design and operation of the boiling water reactor plant, it is necessary to classify the various systems, criteria, design bases, and operating requirements in light of specified personnel (including the public) hazard considerations. A system has been developed which allows classification of any BWR aspect-criterion, system, design basis, or operating requirement-relative to either personnel hazard or the plant mission (the generation of electrical power).

Table 1.4-1 illustrates the concept used in the classification process. The concept applies to the total plant: design and operation. A major distinction is made between those BWR aspects which are most pertinent to personnel hazard and those which are most pertinent to the plant mission-the generation of electrical power. Those aspects most pertinent to personnel hazard would appear under the "safety consideration" side (left side) of the table, and the aspects most pertinent to the plant mission would appear under the "power generation" side (right). All plant components contribute in some measure to safety, but those classified under "power generation" considerations are considerably less important to safety than those items classified under "safety" considerations. Therefore, the right and left sides of the table represent a major difference in importance to safety.

Down the left side of Table 1.4-1 are listed the various types of plant operation, including events resulting in transients and accidents. An allowance is made for a special event in the left column to enable the classification of criteria, systems, and operational requirements not otherwise classifiable. The left-hand column is actually a gross probability scale. Planned operation is certain, abnormal operational transients are reasonably expected, and accidents are very improbable. Any special events would have to be fitted into the probability scale as appropriate. The left-hand column might ultimately develop into a quantified probability scale.

The rectangular spaces formed under the safety considerations heading and the power generation heading represent potential classification categories for BWR criteria, systems, and operational requirements. This classification concept, when applied, allows an accurate distinction between the importances of the various aspects of BWR design and operation.

1.4.2 Classification Basis Tables 1.4-2A and 1.4-2B present the basis for classifying various BWR items. The format of the tables is similar to that used in Table 1.4-1, which presented the classification concept. A list of unacceptable results is given within each 1.4-1

BFN-17 classification category. The unacceptable results represent a set of master criteria, from which the design and operation of the BWR can be consistently evaluated.

The only unacceptable results listed for the power generation consideration (Table 1.4-2B) are those that are more restrictive than those for the safety consideration (Table 1.4-2A).

In the various columns inside each classification category, generic labels are assigned to the specific elements which appear or would appear, if listed, in the column. A generic label is given only to facilitate discussion and identification of a group of elements united by their common classification. Beneath the generic names are listed some of the more illustrative BWR items which can be classified in the different columns. Some of the listed items are the limits and restrictions found in the technical specifications. Technical specifications are limited to those concerns that are only on Table 1.4-2A.

Classification analyses have been performed to establish the essentiality of the various BWR systems to the avoidance or prevention of the listed unacceptable results. Such analyses consider any applicable criteria requiring redundancy or specified levels of functional reliability in the avoidance of unacceptable results.

Once a system is classified, it is evaluated with reference to the criteria applicable to the group in which it performs an essential action. A classification analysis is not the same as a plant safety analysis. A classification analysis takes no credit whatever for the system under study; whereas, a plant safety analysis represents the true response of the whole plant to an event under specified analytical assumptions.

1.4.3 Use of the Classification Plan Because Tables 1.4-2A and B permits the classification of any BWR criterion, system, or operational requirement into one or more of the classification categories, the plan facilitates a plantwide safety overview. The plan explains the reasons for the differences in the designs of apparently similar systems by relating the actions of the systems to specified unacceptable results. With the design complete, the classification plan is used to establish operational requirements and procedures whose differences are consistent with the different importances of unacceptable results.

It should be noted that a system may be classified in several categories. This occurs because classification is the result of a functional analysis of the plant. When classified in more than one category, a system must satisfy all of the requirements for each category with regard to its contributions to the various safety actions within each of the categories.

1.4-2

BFN-21 TABLE 1.4-1 BWR SAFETY ENGINEERING CONCEPT FOR CLASSIFICATION OF BWR SYSTEMS, CRITERIA AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION Type of Operation Safety Considerations Power Generation Considerations or Event

1. Planned Operation In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems, and operational requirements power generation, criteria, plant actions, systems and pertinent to safety during planned operation. This space operational requirements pertinent to the production of represents the aspects of the BWR which must be considered to electrical power during planned operation. Process assure that the BWR operator can operate the plant within systems and normal operational procedures would be specified safety limitations. Certain process indicators, process classified here.

variable limits and limits on the release of radioactive material would be classified here.

2. Abnormal Operational In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for Transients criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to abnormal operational transients. operational requirements pertinent to the ability to produce Certain protection systems, safety limits, and limiting safety electrical power as that ability is affected by abnormal system settings would be classified here. operational transients. Certain systems not used for planned operation would be classified here.
3. Accidents In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to accidents. Engineered operational requirements pertinent to the ability to produce safeguards would be classified here. electrical power as that ability is affected by accidents.

Design considerations and post-accident procedures provided to enable the plant to be used for power generation after an accident would be classified here.

4. Special Event In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to the stated special event. Safety operational requirements pertinent to the ability to produce systems provided especially for the special event would be electrical power as that ability is affected by the stated classified here. special event. Systems and procedures provided to enable the plant to be returned to power operation following the special event would be classified here.

BFN-21 Table 1.4-2A (Sheet 1)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

1. Planned 1-1 The release of Nuclear Safety Design Safety Action- Safety Systems- Operational Nuclear Safety Operation radioactive material Criteria-Type S-1 Type S-1 Type S-1 Requirements-Type S-1 to the environs to such an extent that the Nuclear Safety Process Safety Action Process Safety Systems Operational Nuclear Safety limits of 10CFR20 Operational Criteria- (A Category of Safety (A Category of Safety Limits-Type S-1 are exceeded. Type S-1 Action) Systems)

Process Safety Indication of Process Indicators Technical Specifications-1-2 Fuel failure to such an Design Criteria Variables Type S-1 extent that were the freed fission products Process Safety Rod Worth Monitoring Rod Worth Minimizer Process Safety Limits to the environs via the Operational Criteria Program of Process Computer normal discharge paths Rod Pattern Control Limiting Conditions for for radioactive material, Various Industry Radwaste Systems for Operation for Indicators limits of 10CFR20 Codes Control of Process would be exceeded. Variables Process Radiation Monitors Radioactive Material Radwaste Criteria Release Limits Control Rod Control Control Rod Drive System 1-3 Nuclear System stress Loading Criteria Refueling Block in excess of that (Normal Conditions) Reactor Manual Rod Pattern Limits allowed for planned Control Rod Control Control System operation by applicable industry codes. Refueling Block Limiting Conditions for Refueling Interlocks Operation for Radwaste Systems Core Shutdown Control 1-4 The existence of a plant Reactor Protection System condition not considered Radwaste (Manual Scram) Nuclear System Leakage by plant safety analysis. Limits Isolation Condensate Storage System Neutron Monitoring System

BFN-21 Table 1.4-2A (Sheet 2)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

2. Abnormal 2.1 The release of radioactive Nuclear Safety Design Safety Action-Type S-2 Safety Systems-Type S-2 Operational Nuclear Operational material to the environs Criteria-Type S-2 Safety Requirements-Transients to such an extent that Scram Protection System (Generic Type S-2 the limits of 10CFR20 Nuclear Safety Term) are exceeded. Operational Criteria- Pressure Relief Type S-2 Nuclear Safety Systems Operational Nuclear Safety Core Cooling (A Category of Protection Limits-Type S-2 Systems) 2.2 Any fuel failure calculated as a result of the transient. Various Industry Codes Containment Reactor Protection System Technical Specifications-Cooling (RHRS) (Scram) Type S-2 IEEE-279 Primary Containment Control Rod Drive System (Scram) Safety Limits Secondary Availability Goals Containment Neutron Monitoring System (IRM, APRM) Limiting Safety System Loading Criteria Settings (Upset Conditions) Pressure Relief System 2.3 Nuclear system stress in Limiting Conditions for excess of that allowed for Single Failure Reactor Vessel Isolation Operation for Protection transients by applicable Criterion Control System Systems industry codes.

Testability Criteria High Pressure Coolant Surveillance Requirements Injection System for Protection Systems D-C Power System Standby A-C Power

BFN-21 Table 1.4-2A (Sheet 3)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

3. Accidents 3-1 Radioactive Nuclear Safety Design Safety Action-Type S-3 Safety Systems-Type S-3 Operational Nuclear Safety material release Criteria-Type S-3 Requirement-Type S-3 tosuch an extent that the guideline Scram Protection Systems (Generic values of 10CFR50.67 Nuclear Safety Term) Operational Nuclear Safety would be exceeded. Operational Criteria - Core Cooling Limits-Type S-3 Type S-3 Engineered Safeguards 3-2 Fuel cladding temperatures Containment Reactor Protection System Technical Specifications-in excess of 2200F for Control Rod Drive System Type S-3 Pipe Breaks Neutron Monitoring System 3-3 Nuclear system Various Industry Codes Pressure Relief System (Main pressure in excess Containment Cooling Steam Relief Valves) Limiting Safety System of that allowed for IEEE-279 Reactor Vessel Isolation Settings accidents by applicable Control System industry codes. Availability Goals Stop Control Rod Primary Containment Isolation Limiting Conditions for Ejection Control System Operation for Protection Primary Containment Systems Secondary Containment 3-4 Containment stresses Loading Criteria Limit Reactivity Main Steam Line Isolation sufficient to produce (Emergency and Insertion Rate Valves Surveillance Requirements containment failure Faulted Conditions) Main Steam Line Flow Restrictor for Nuclear System when containment is Pressure Relief High Pressure Coolant Injection required. System Automatic Depressurization 3-5 Overexposure to Single Failure Criteria Reactor Vessel System radiation of operating Isolation Low Pressure Coolant Injection personnel in the Testability Criteria Core Spray System control room. Primary Containment RHRS (Containment Cooling)

Isolation Control Rod Velocity Limiter

BFN-21 Table 1.4-2A (Sheet 4)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

3. Accidents 3-5 Overexposure Testability Criteria Secondary Containment Control Rod Drive Housing Surveillance requirements (Cont.) to radiation Isolation Supports for nuclear systems of operating Standby Gas Treatment System personnel in Standby A-C Power System the control Treatment of Fission D-C Power System room. Products Main Steam Line Radiation Monitoring System Restriction of Coolant Reactor Building Ventilator Loss Rate Radiation Monitoring System RHR Service Water System Control Room Isolation 3-6 Peak enthalpy of fuel in excess of 280 cal/gm for the control rod drop accident.

BFN-21 Table 1.4-2A (Sheet 5)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

4. Special Event 4-1 The inability to Nuclear Safety Design Safety Action - Type S-4 Safety Systems-Type S-4 Operational Nuclear Safety Loss of bring the reactor Criteria-Type S-4 Requirements-Type S-4 Habitability to the shutdown Special Safety Action Special Safety Systems of the Control condition by manip-Room ulation of the local controls and Nuclear Safety Operational Nuclear Safety equipment which Operational Criteria- Shutdown From Outside Local Controls Outside Limits-Type S-4 are available out- Type S-4 Control Room Control Room side the control room. Special Safety Design Technical Specifications-Criteria Cooldown from Outside Local Indicators Outside Type S-4 Control Room Control Room 4-2 The inability to bring the reactor Special Safety Limiting Conditions for to the cold shut- Operational Criteria Condensate Storage System Operation for Special Safety down condition from Systems outside the control room. Reactor Core Isolation Surveillance Requirements for Cooling System Special Safety Systems Pressure Relief System Reactor Protection System Control Rod Drive System RHR (containment Cooling)

BFN-21 Table 1.4-2A (Sheet 6)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

5. Special Event- 5-1 The inability to Nuclear Safety Design Safety Action-Type S-5 Safety Systems-Type S-5 Operational Nuclear Safety Inability to shut down the Criteria-Type S-5 Requirements-Type S-5 Shut Down reactor independent Special Safety Action Special Safety Systems Reactor With of control rods Operational Nuclear Safety Control Rods Limits-Type S-5 5-2 The inability to Nuclear Safety maintain the Operational Criteria Shutdown Without Control Standby Liquid Control reactor in the Type S-5 Rods System shutdown condition Technical Specifications-independent of Special Safety Design Maintain Shutdown During RWCU Isolation Type S-5 control rods Criteria Reactor Cooldown Limiting Conditions for Operation for Special Safety Systems Special Safety Operational Criteria Surveillance Requirements for Special Safety Systems

BFN-21 Table 1.4-2B (Sheet 1)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results

1. Planned Operation 1-1 Inability to generate Power Generator Design Power Generation Action- Power Generator Systems- Operational Power Generator electrical power Criteria-Type PG-1 Type PG-1 Type PG-1 Requirements-Type PG-1 1-2 Fuel Failure Power Generator Operational Power Generator Operational Criteria - Limits-Type PG-1 Type PG-1 1-3 Inability to Perform Process Action Process Systems Routine Maintenance (A Category of Power (A Category of Power with Plant at Power Generation Action) Generator Systems)

Process Design Criteria Normal Operating Procedures 1-4 Inability to Optimize Fuel Performance Process Operational Indications of Process Indicators Maintenance Procedures Criteria Variables 1-5 Inability to Respond Process Operations Process Computer System Calibration Procedures to Changes in Power Demand Fuel Performance Recirculation Flow Control Refueling Procedures Calculations System 1-6 Inability to Shut Down Power Level Control Reactor with Control Reactor Manual Control Rods in the Normal Consideration of Exhaust System Manner Steam Control Rod Drive System Feedwater System Turbine-Generator Main Condenser

BFN-21 Table 1.4-2B (Sheet 2)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results

2. Abnormal 2-1 Fuel Failure Power Generation Design Power Generation Action Power Generation Systems Operational Power Generation Operational Criteria-Type PG-2 Type PG-2 Type PG-2 Requirements-Type PG-2 Transients 2-2 The Lifting of Main Steam Power Generation Operational Power Generation Relief Valves Operational Criteria Limits-Type PG-2 Type PG-2 2-3 Conditions Requiring the Rod Block Reactor Manual Control Opening of the Reactor System (Rod Block)

Vessel for Inspection or Pressure Relief Repair Pressure Relief System Normal Operating Procedures Refueling Block 2-4 Inability to Return to Scram Refueling Interlocks Post Transient Recovery Power Operation Procedures 2-5 Inadvertent Criticality Core Cooling Reactor Protection Refueling Restrictions During Refueling System (RPS)

Electro Hydraulic Control (EHC) System Reactor Core Isolation Cooling (RCIC)

BFN-21 Table 1.4-2B (Sheet 3)

BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Power Generation Required to Avoid Required to Avoid Types of Requirements (Where More Restrictive Unacceptable Results Unacceptable Results to be Observed in Operation Type of Operation Than Unacceptable Types of Applicable (Where Not Required (Where Not Required of Plant to Avoid or Event Safety Results) Criteria as a Safety Action) as a Safety Action) Unacceptable Results

3. Accidents 3-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems - Operational Power Generation to Power Operation Criteria-Type PG-3 Type PG-3 Type PG-3 Requirements-Type PG-3 Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-3 Type PG-3 Post Accident Recovery Procedures
4. Special Event 4-1 Inability to Return Power Generation Design Power Generation Action- Power Generation Systems- Operational Power Generation Loss of to Power Operation Criteria-Type PG-4 Type PG-4 Type PG-4 Requirements-Type PG-4 Habitability of the Control Power Generation Operational Power Generation Room Operational Criteria - Limits-Type PG-4 Type PG-4 Post Event Recovery Procedures
5. Special Event 5-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems- Operational Power Generation Inability to to Power Operation Criteria-Type PG-5 Type PG-5 Type PG-5 Requirements-Type PG-5 Shut Down Reactor With Control Rods Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-5 Type PG-5 Post Event Recovery Procedures

BFN-19 1.5 PRINCIPAL DESIGN CRITERIA There are two ways of considering principal design criteria. One way is to consider the criteria on a system-by-system (or system group) basis. The second way is to consider criteria classification-by-classification as given in Tables 1.4-2 A and B.

In the classification-by-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 classification 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; thus, the less restrictive (but more important) criteria pertaining to the system in the classification approach will be masked by the more restrictive (and less important) criteria.

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

1.5.1 Principal Design Criteria Classification-By-Classification The principal architectural and engineering criteria for the design and construction of the plant are summarized below. The criteria are grouped according to the classification plan given in Tables 1.4-2 A and B. 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 this report.

Criteria pertaining to operation of the plant are given in Appendix G.

1.5.1.1 General Criteria Applicable Classifications Criteria PG-1,S-1,S-2,S-3 1. The plant shall be designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner. The plant design shall be in accordance with applicable codes and regulations.

S-1,S-2,S-3 2. The plant 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 1.5-1

BFN-19 applicable regulations pertaining to the release of radioactive materials are not exceeded.

S-1,S-2,S-3,S-4 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.

S-1,S-2,S-3 4. Adequate strength and stiffness with appropriate safety factors shall be provided so that a hazardous release of radioactive material shall not occur.

1.5.1.2 Power Generation Design Criteria, Type PG-1 (Planned Operation)

1. The nuclear system shall employ a General Electric 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. (Deleted).
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 shielding and cooling for spent fuel.

1.5-2

BFN-19

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 safeguards.

1.5.1.3 Power Generation Design Criteria, Type PG-2 (Abnormal Operational Transients)

1. The fuel cladding, in conjunction with other plant 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 in the reactor core for any abnormal operational transient. The capacity of such systems shall be adequate to prevent fuel clad damage.
3. Heat removal systems shall be provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems shall be adequate to prevent fuel clad damage.
4. Standby electrical power sources shall be provided to allow removal of decay heat under circumstances where normal auxiliary power is not available.
5. Fuel handling and storage facilities shall be designed to prevent inadvertent criticality.

1.5.1.4 Nuclear Safety Design Criteria, Type S-1 (Planned Operation)

1. The Plant shall be designed so that fuel failure during planned operation is limited to such an extent that, were the freed fission products released to the environs via the normal discharge paths for radioactive materials, the limits of 10 CFR 20 would not be exceeded.
2. The reactor core shall be designed so that its nuclear characteristics exhibit no tendency toward a divergent power transient.
3. The nuclear system shall be so designed that there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.
4. Gaseous, liquid, and solid waste disposal facilities shall be so designed that the discharge and offsite shipment of radioactive effluents can be made in accordance with applicable regulations.

1.5-3

BFN-19

5. The design shall provide means by which plant operations personnel can be informed whenever limits on the release of radioactive material are exceeded.
6. Sufficient indications shall be provided to allow determination that the reactor is operating within the envelope of conditions considered by plant 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 plant operation.

1.5.1.5 Nuclear Safety Design Criteria, Type S-2 (Abnormal Operational Transients)

1. The plant shall be so designed that fuel failure as a result of any abnormal operational transient is limited to such an extent that, were the freed fission products released to the environs via the normal discharge paths for radioactive materials, the limits of 10 CFR 20 would not be exceeded.
2. Those portions of the nuclear system which form part of the nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients.
3. Nuclear safety systems shall act to assure that no damage to the nuclear system process barrier 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 plant operations personnel.
5. Essential safety actions shall be carried out by equipment of sufficient redundance 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.
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 control room.

1.5-4

BFN-19

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.

1.5.1.6 Nuclear Safety Design Criteria, Type S-3 (Accidents)

1. Those portions of the nuclear system which form part of the nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following accidents. For accidents in which one breach in the nuclear system process barrier is postulated, such breach shall not cause additional breaches in the nuclear system process barrier.
2. Engineered safeguards shall act to assure that no damage to the nuclear system process barrier results from internal pressures caused by an 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 plant operations personnel.
4. Essential safety actions shall be carried out by equipment of sufficient redundance 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.
5. Features of the plant which are essential to 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.
6. The design of engineered safeguards shall include allowances for environmental phenomena at the site.
7. Provision shall be made for control of active components of engineered safeguards from the control room.

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8. Engineered safeguards shall be designed to permit demonstration of their functional performance requirements.
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 containments, in conjunction with other engineered safeguards, shall act to prevent the radiological 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 penetrates the primary containment structure, and which could serve as a path for the uncontrolled release of radioactive material to the environs, shall be automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.
18. Core Standby Cooling Systems shall be provided to prevent excessive fuel clad temperatures as a result of a loss-of-coolant accident.

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19. The Core Standby Cooling Systems shall provide for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.
20. The Core Standby Cooling Systems shall be diverse, reliable and redundant.
21. Operation of the Core Standby Cooling Systems shall be initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the plant.
22. Standby electrical power sources shall have sufficient capacity to power all engineered safeguards requiring electrical power.
23. The control room shall be shielded against radiation so that occupancy under accident conditions is possible.

1.5.1.7 Nuclear Safety Design Criteria, Type S-4 (Special Event In the event that the control room becomes inaccessible, it shall be possible to bring the reactor from power range operation to cold shutdown (Mode 4) by manipulation of the local controls and equipment which are available outside the control room.

1.5.1.8 Nuclear Safety Design Criteria, Type S-5 (Special Event)

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 normal operating condition, and subsequently to maintain the shutdown condition.

1.5.2 Principal Design Criteria, System-By-System The principal architectural and engineering criteria for design are summarized below on a system-by-system or system group basis. The system-by-system presentation facilitates the understanding of the actual design of any one system, but significant distinctions in the importance to safety of different criteria pertaining to a system cannot be made clear, as they are in the classification-by-classification presentation. To make consistent judgments regarding plant safety, the classification-by- classification approach to criteria must be used.

In the system-by-system presentation of criteria, only the most restrictive of any related criteria are stated for a system. Where the most restrictive criterion is one which is classified as a power generation consideration in Table 1.4-2B, less 1.5-7

BFN-19 restrictive, but more important, safety criteria may be hidden (not stated) in the system-by-system presentation.

1.5.2.1 General Criteria

1. The plant shall be designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner. The plant design shall be in accordance with applicable codes and regulations.
2. The plant 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.

1.5.2.2 Nuclear System Criteria

1. The nuclear system shall employ a General Electric boiling water reactor to produce steam for direct use in a turbine-generator.
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 nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients and accidents. For accidents in which one breach in the nuclear system process barrier is postulated, such breach shall not cause additional breaches in the nuclear system process barrier.
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 plant 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 circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems shall be adequate to prevent fuel clad damage.

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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 of highest reactivity worth fully withdrawn and unavailable for insertion.
8. The nuclear system shall be so designed that there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.
9. The reactor core shall be so designed that its nuclear characteristics exhibit no tendency toward a divergent power transient.

1.5.2.3 Power Conversion Systems Criteria

1. Appropriate power conversion systems shall be provided to efficiently convert the heat energy of the steam produced in the reactor vessel to mechanical energy for turning a generator to produce electrical power.
2. Means shall be provided for furnishing makeup (feedwater) to the reactor vessel to allow continued operation.

1.5.2.4 Electrical Power Systems Criteria

1. A generator capable of efficiently producing electric power shall be provided.
2. Electrical power for protection systems and engineered safeguards shall be available from two offsite sources so that no single failure in the facility can result in loss of offsite power.

1.5.2.5 Radioactive Waste Disposal Criteria

1. Gaseous, liquid, and solid waste disposal facilities shall be designed so that the discharge and offsite shipment of radioactive effluents can be made in accordance with applicable regulations.
2. The design shall provide means by which plant operations personnel can be informed whenever operational limits on the release of radioactive material are exceeded.

1.5.2.6 Nuclear Safety Systems and Engineered Safeguards Criteria 1.5.2.6.1 General

1. Nuclear safety systems shall act in response to abnormal operational transients to limit fuel damage such that, were the freed fission products 1.5-9

BFN-19 released to the environs via the normal discharge paths for radioactive material, the limits of 10 CFR 20 would not be exceeded.

2. Nuclear safety systems and engineered safeguards shall act to assure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients or accidents.
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 plant operations personnel.
4. Essential safety actions shall be carried out by equipment of sufficient redundance 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.
5. Features of the plant which are essential to 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 function to be performed.
6. The design of nuclear safety systems and engineered safeguards shall include allowances for environmental phenomena at the site (e.g., weather extremes and proximity to other high energy systems). Furthermore, electrical equipment in these systems shall be capable of performing their safety function as required under environmental conditions associated with all normal, abnormal, and plant accident operation.
7. Provision shall be made for control of active components of nuclear safety systems and engineered safeguards from the control room.
8. Nuclear safety systems and engineered safeguards shall be designed to permit demonstration of their functional performance requirements.

1.5.2.6.2 Containment and Isolation Criteria

1. A primary containment shall be provided that completely encloses the reactor vessel.
2. 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.

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3. It shall be possible to test primary containment integrity and leak tightness at periodic intervals.
4. A secondary containment shall be provided that completely encloses both the primary containment and fuel storage areas.
5. 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.
6. 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.
7. The primary and secondary containments, in conjunction with other engineered safeguards, shall act to prevent the radiological effects of accidents resulting in the release of radioactive material to the containment volumes from exceeding the guideline values of applicable regulations.
8. 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.
9. Piping that penetrates the primary containment structure, and could serve as a path for the uncontrolled release of radioactive material to the environs, shall be automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.

1.5.2.6.3 Core Standby Cooling Criteria

1. Core Standby Cooling Systems shall be provided to prevent excessive fuel clad temperatures as a result of a loss-of-coolant accident.
2. The Core Standby Cooling Systems shall provide for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.
3. The Core Standby Cooling Systems shall be diverse, reliable, and redundant.
4. Operation of the Core Standby Cooling systems shall be initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the plant.

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BFN-19 1.5.2.6.4 Standby Power Criteria

1. 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.
2. Standby electrical power sources shall have sufficient capacity to power all engineered safeguards requiring electrical power.

1.5.2.7 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.
2. In the event that the control room is inaccessible, it shall be possible to bring the reactor from power range operation to cold shutdown (Mode 4) by manipulation of the local controls and equipment which are available outside the control room.

1.5.2.8 Process Control Systems Criteria 1.5.2.8.1 Nuclear System Process Control Criteria

1. It shall be possible to manually control the reactor power level.
2. Control of the nuclear system shall be possible from a single location.
3. 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.
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 safeguards.

1.5.2.8.2 Deleted 1.5.2.8.3 Electrical Power Systems Process Control Criteria Controls shall be provided in the electrical power systems to protect against faults and to increase the reliability of incoming and outgoing power.

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BFN-19 1.5.2.9 Auxiliary Systems Criteria

1. Fuel handling and storage facilities shall be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel.
2. Means shall be provided to remove heat from process systems that is generated through operation of the plant.
3. Fire detection and protection systems capable of protecting the plant against all types of fires shall be provided.
4. Means shall be provided to adequately heat, ventilate, and air-condition plant buildings for personnel comfort and equipment protection.
5. Means shall be provided to furnish other auxiliary services as required for safe and efficient operation of the plant.

1.5.2.10 Shielding and Access Control Criteria

1. 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 plant operation.
2. The control room shall be shielded against radiation so that occupancy under accident conditions is possible.

1.5.2.11 Structural Loading Criteria Adequate strength and stiffness, with appropriate safety factors, shall be provided so that a hazardous release of radioactive material shall not occur. Details of implementation are given in Chapter 12 and Appendix C.

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BFN-27 1.6 PLANT DESCRIPTION 1.6.1 General 1.6.1.1 Site and Environs 1.6.1.1.1 Location and Size of Site The site contains approximately 840 acres and is located on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. It is approximately 30 miles west of Huntsville, Alabama.

1.6.1.1.2 Site Ownership The plant is located on property owned by the United States and in the custody of TVA.

1.6.1.1.3 Activities at the Site Activities at the site are those performed by TVA in operating the three-unit nuclear plant to produce electric power.

1.6.1.1.4 Access to the Site (See Figure 2.2-4)

The three-unit plant, including the intake and discharge canals, is enclosed by a security fence. Primary access to the plant area is by way of an access road through a security gate.

1.6.1.1.5 Description of the Environs (See Table 2.2-6)

The Browns Ferry site is located in an area where the land is used primarily for agriculture. Population densities are low, with a projected population of 33,340 within ten miles for the year 2020. There are no population centers of significance within ten miles of the plant. The low population zone is determined to be seven miles.

1.6.1.1.6 Geology The site is underlain by massive formations of nearly horizontal bedrock.

Historically, this region has been one of little structural deformation, and major folds and faults are entirely absent.

1.6-1

BFN-27 1.6.1.1.7 Seismology There has been no known major seismic activity originating in or near the site area.

The major seismic activity experienced at the site has been caused by distant major earthquakes.

1.6.1.1.8 Hydrology Groundwater movement in the area is from the plant site to the Tennessee River. A thick mantle of residuum in the site area retards the movement of shallow groundwater.

1.6.1.1.9 Regional and Site Meteorology The meteorology of the Browns Ferry site provides generally favorable atmospheric conditions for dispersion of plant emissions. The immediate terrain is flat and slightly undulating, with scattered 400- to 600-foot foothills. Thus, local entrapment or accumulation of emissions should not occur.

1.6.1.1.10 Design Bases Dependent Upon the Site and Environs

a. Offgas Systems The plant offgas systems are designed to maintain gaseous waste releases to the environment, during normal operation, at levels which assure that concentrations at the site boundary will be within the limits of 10 CFR 20. The effects of releases at or beyond the site boundary resulting from the design basis accidents will be within the reference values of 10 CFR 50.67.
b. Liquid Waste Effluents The plant Liquid Radwaste System is designed to maintain liquid waste releases to the environment at levels which comply with the plant's National Pollutant Discharge Elimination System (NPDES) permit limitations and assure that concentrations at the site boundary will be within the limits of 10 CFR 20.
c. Wind Loading Design A structural design capable of withstanding loadings resulting from a 100-mph sustained wind, except for Low Level Radioactive Waste Storage Facility (LLRWSF) which uses a wind of 95 mph, is considered appropriate. All Class I structures and equipment that are required to support and maintain safe shut down of all the units as a result of a tornado design basis event are designed to maintain their integrity when subjected to loading resulting from a 300-mph 1.6-2

BFN-28 tornado. The LLRWSF is designed for a 290 mph rotational speed at 150 feet radius and 70 mph translational speed. For the 600 foot tall reinforced concrete chimney, only the bottom 280 feet of the chimney is designed for the 300-mph tornado. The top 320 feet is designed only for the 100-mph sustained wind. See Section 12 for greater detail of design wind and tornado loadings.

d. Seismic Design The design of all Class I structures is based on a ground motion due to an acceleration of 0.10g (Operating Basis Earthquake). In addition, the design is such that the plant can be safely shut down during a ground acceleration of 0.20g (Design Basis Earthquake).
e. Flooding Plant grade is established at 565 feet above mean sea level. The probable maximum flood at Browns Ferry would reach El. 572.5, plus wind wave runup produced by a coincidental 45 MPH sustained wind speed.
f. Loss of Normal Heat Sinks (Downstream Dam Failure)

Failure of the downstream Wheeler Dam was assumed as part of the basis for the UHS. The assumed failure results in a consequential failure of the Wilson Dam further downstream.

If t h e Wheeler Dam downstream from the plant site were to fail, Wheeler Reservoir would pot ent ial ly drain down to a minimum elevation of 529 feet at the plant intake due to a high point in the Tennessee River basin downstream of BFN at approximately river mile 291.8. A pool of water approximately 1,000 feet wide and 7 miles long, containing a volume of about 69.6X10 6 cubic feet of water, would be available at the plant intake.

The resultant pool elevation is more than sufficient to maintain adequate flow and NPSH to the RHRSW p umps, which supply shutdown cooling water to the three units.

No credit is taken for normal effluent flow rates from the upstream Guntersville Dam. However, an evaluation of historical data for the upstream Guntersville Dam found that the daily flow rate exceeded 1,319 cubic feet per second (cfs) for greater than 99 percent of the historical time period and that the average daily flow rate was 33,500 cfs. Within the historical data, the lowest recorded daily flow rate was found to be 100 cfs, which is above the minimum required 1.6-3

BFN-28 RHRSW flow of 80 cfs for one unit responding to an accident and two units in shutdown.

Therefore, inflow contribution to the UHS from the upstream Guntersville Dam at the lowest recorded daily average flow rate combined with the dependable watershed drainages upstream of BFN will provide at least 100 cfs inflow to the UHS. The inflow to the UHS from the upstream sources provides additional assurance that resultant pool elevation is maintained at least 529 feet with more than the required 80 cfs of cooling water available.

g. Environmental Radiation Monitoring System The availability of past wind direction and persistence data and river flow records, along with knowing the location of population centers, has aided in the selection of monitoring locations and frequency of sampling.

1.6.1.2 Facility Arrangement The facility arrangement is shown in Figure 2.2-4. Plan and elevation views of the major buildings are shown in Figures 1.6-1 through 1.6-27.

1.6.1.3 Nuclear System Each nuclear system includes a single-cycle, forced-circulation, General Electric boiling water reactor producing steam for direct use in a steam turbine. A typical heat balance showing the major parameters of the nuclear system for the rated power condition is shown in Figure 1.6-28 (3952 MWt).

1.6.1.3.1 Reactor Core and Control Rods The fuel for the reactor core consists of uranium dioxide pellets made from slightly enriched uranium. These pellets are contained in sealed Zircaloy-2 tubes. These fuel rods are assembled into individual fuel bundles. The detailed description of fuel in the reactor core is given in Section 3.2 of the FSAR.

The description of the core for each unit is given in the current reload licensing document for that unit as described in FSAR Appendix N.

1.6-4

BFN-27 1.6.1.3.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structure, the steam separators and dryers, the jet pumps, the control rod guide tubes, distribution lines for the feedwater, core spray, and standby liquid control, the incore instrumentation, and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, feedwater lines, control rod drive housings, and core standby cooling lines.

Each reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure is 1050 psia (uprated) in the steam space above the separators. The vessel is fabricated of carbon steel and is clad internally (except for the top head) with weld overlay.

The reactor core is cooled by demineralized water which enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers, located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steam lines. Each steam line is provided with two isolation valves in series--one on each side of the primary containment barrier.

1.6.1.3.3 Reactor Recirculation System The Reactor Recirculation System pumps reactor coolant through the core to remove the energy generated in the fuel. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. Each loop has one motor-driven recirculation pump. Recirculation pump speed can be varied to allow control of reactor power level through the effects of coolant flow rate on moderator void content. For Unit 2 only, the two recirculation loops have a cross-connect line with one normally closed valve and one normally open valve to prevent pressure buildup between the valves.

1.6.1.3.4 Residual Heat Removal System The Residual Heat Removal System (RHRS) is a system of pumps, heat exchangers, and piping that fulfills the following functions.

a. Removal of decay heat during and after plant shutdown.
b. Injection of water into the reactor vessel following a loss-of-coolant accident rapidly enough to reflood the core and prevent excessive fuel clad temperatures independent of other core cooling systems. This is discussed in paragraph 1.6.2 (Nuclear Safety Systems and Engineered Safeguards).

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BFN-28

c. Removal of heat from the primary containment following a loss-of-coolant accident to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the water inside the primary containment. The redundancy of the equipment provided for containment cooling is further extended by a separate part of the RHRS which sprays cooling water into the drywell and pressure suppression pool.
d. Provide standby cooling.
e. Provide assistance for fuel pool cooling when required.

1.6.1.3.5 Reactor Water Cleanup System A Reactor Water Cleanup System, which includes a demineralizer arrangement, is provided to clean up the reactor cooling water, to reduce the amounts of activated corrosion products in the water, and to remove reactor coolant from the nuclear system under controlled conditions.

1.6.1.3.6 Reactor Core Isolation Cooling System The Reactor Core Isolation Cooling System (RCICS) provides makeup water to the reactor vessel whenever the vessel is isolated. The RCICS uses a steam-driven, turbine-pump unit and operates automatically to maintain adequate reactor vessel water level.

1.6.1.4 Power Conversion Systems The Power Conversion Systems use the steam produced in the reactor vessel to produce electrical power. Figure 1.6-29, Sheets 1, 2, and 3, shows the turbine generator heat balance for rated power conditions. Figure 1.6-30 is a flow diagram for general plant systems.

1.6.1.4.1 Turbine Generator The turbine is an 1,800-rpm tandem-compound, six-flow, nonreheat unit. It has a double-flow, high-pressure cylinder and three double-flow low-pressure cylinders.

Steam from the high-pressure cylinder passes through moisture separators before entering the low-pressure units. The turbine has five extraction stages for reactor feedwater 1.6-6

BFN-27 system heating. Turbine controls include an electric-hydraulic speed governor, overspeed protection, steam admission control valves, emergency stop valves, combined intermediate stop-intercept valves, bypass valves, and pressure regulators. The electrical generator is direct-driven, hydrogen-cooled with liquid-cooled stator, and is equipped with an automatic voltage regulator bus-fed from auxiliary transformers.

1.6.1.4.2 Turbine Bypass System The Turbine Bypass System is provided to pass steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the load permitted to pass to the turbine generator (such as during generator synchronization or following sudden load changes).

1.6.1.4.3 Main Condenser Three deaerating, single-pass, single-pressure, radial-flow-type surface condensers provide the primary heat sinks for each turbine-generator. Each condenser is located beneath one of the low-pressure turbines with the tubes oriented transverse to the turbine-generator axis. Baffling in the hotwell is arranged to ensure a minimum of 1.5 minutes retention time for the condensate.

1.6.1.4.4 Main Condenser Gas Removal and Turbine Sealing Systems Two 100-percent capacity steam jet air ejectors are provided for each unit to remove air and noncondensables from the main condensers during normal operation. A mechanical vacuum pump is provided for startup operation.

The Turbine Sealing System is provided to prevent steam leakage and air inleakage at the turbine seals.

1.6.1.4.5 Condenser Circulating Water System Seven mechanical-draft cooling towers are provided to dissipate waste heat to the atmosphere. Water is pumped through the main condenser to an open channel going to the towers of the circulating water pumps for each unit. Water is pumped to each cooling tower by lift pumps. The system is designed for open and helper modes of operation.

In the open mode, water is drawn into the circulating water pumping station forebay from Wheeler reservoir, pumped through the main condenser, and discharged back into the reservoir through a diffuser discharge system consisting of perforated metal pipes which extend across the reservoir channel to diffuse the warmer water from 1.6-7

BFN-27 the plant. In the helper mode, the water is pumped from the reservoir, through the plant, and into an open channel going to the cooling towers where it is pumped through the towers and is returned to the reservoir through the diffusers.

1.6.1.4.6 Condensate Filter/Demineralizer System This full-flow system removes dissolved and suspended solids from the condensate, providing high-quality water for the nuclear system. It consists of filter/demineralizer vessels containing filter elements which are coated with a mixture of powdered cation and anion exchange resins. These resins perform both the filtration and deionization functions.

1.6.1.4.7 Condensate and Reactor Feedwater Systems The Condensate and Reactor Feedwater Systems take suction from the main condensers and deliver demineralized water to the reactor vessel at an elevated temperature and pressure. Three vertical, centrifugal, motor-driven condensate pumps; three horizontal, centrifugal, motor-driven condensate booster pumps; and three horizontal, centrifugal, single-stage reactor feedwater pumps with variable-speed steam turbines are provided for these systems. Feedwater is controlled by varying the speed of the reactor feedwater-pump turbine-drives.

Five stages of feedwater heating are provided for each of the three feedwater streams. All heaters are of the two-pass, U-tube type.

1.6.1.5 Electrical Power Systems Each generator produces electrical power at 22-kV. This 22-kV generator output is transmitted through isolated-phase buses to a bank of three single-phase main power transformers, where the voltage is stepped up to 500-kV and transmitted to the 500-kV switchyard. The 500-kV switchyard connects the plant to the TVA 500-kV system. The plant has generator breakers so that startup and shutdown are from the 500-kV system. The 161-kV system is also available to provide plant startup and shutdown power.

1.6.1.6 Radioactive Waste Systems The Radioactive Waste Systems are designed to control the release of plant-produced radioactive material to within the limits specified in the ODCM and NPDES permits. The methods employed for the controlled release of those contaminants are dependent primarily upon the state of the material: liquid, solid, or gaseous.

1.6-8

BFN-27 1.6.1.6.1 Liquid Radwaste System The Liquid Radioactive Waste Control System collects, treats, stores, and disposes of all radioactive liquid wastes. These wastes are collected in sumps and drain tanks at various locations throughout the plant and then transferred to the appropriate collection tanks in the Radwaste Building for treatment, storage, and disposal. Wastes to be discharged from the system are processed on a batch basis, with each batch being processed by such method or methods appropriate for the quality and quantity of materials determined to be present. Processed liquid wastes may be returned to the condensate system or discharged to the environs through the circulating water discharge canal. The liquid wastes in the discharge canal are diluted with condenser effluent circulating water to achieve a permissible concentration at the site boundary.

Batches of low-conductivity liquid waste are processed through a filter and a waste demineralizer. Demineralizer effluent is sent to a waste sample tank. Depending upon the conductivity and level of radioactivity, the liquid may then be discharged to the circulating-water discharge canal or the cooling tower blowdown line, transferred to condensate storage tanks, returned for further processing through the waste demineralizer.

High-conductivity liquids are processed through a filter and are collected in a floor drain sample tank. If the concentration after dilution is less than or equal to the applicable limits, the filtered liquid may be discharged.

An alternate method of processing low and high conductivity liquid is the use of vendor supplied skid mounted equipment, interconnected to the permanent Radwaste System. Depending on effluent quality and plant needs, the water can be sent to either the waste sample tank or floor drain sample tank. Processing from the waste sample tank or floor drain sample tank is identical as described above.

Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with minimum personnel exposure. For example, tanks and processing equipment which will contain significant radiation sources are located behind shielding; and sumps, pumps, instruments, and valves are located in controlled access rooms or spaces. Processing equipment is selected and designed to require a minimum of maintenance.

Protection against accidental discharge of liquid radioactive waste is provided by valving redundance, instrumentation for detection with alarms of abnormal conditions, procedural controls, interlocks, and radiation monitor controlled valves 1.6-9

BFN-27 1.6.1.6.2 Solid Radwaste System With the Solid Radwaste System, solid radioactive wastes are collected, processed, and packaged for storage. Generally, these wastes are stored onsite until the short half-lived activities are insignificant. Solid wastes from equipment originating in the nuclear system are stored for radioactive decay in the fuel storage pool and prepared for reprocessing or offsite storage. Examples of these wastes are spent fuel, spent control rods, incore ion chambers, etc. Process solid wastes are collected, dewatered, and loaded in shielded containers for storage and shipping.

Examples of these solid wastes are spent demineralizer resins and filter aid.

Wastes such as paper, rags, and used clothing are placed into containers for storage and shipment.

1.6.1.6.3 Gaseous Radwaste System The Gaseous Radwaste System collects, processes, and delivers to the plant stack, for elevated release to the atmosphere, gases from each main condenser air ejector, startup vacuum pump, condensate drain tank vent, and steam packing exhauster.

Gases from each main condenser air ejector are passed through a preheater, a catalytic recombiner, a condenser, a moisture separator, and a dehumidification coil.

The gases then enter a decay pipe which provides a retention time of approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, during which N-16 and 0-19 decay to negligible levels. The gases are then passed through a cooler-condenser, a moisture separator, a reheater, a prefilter, six charcoal beds, an afterfilter, and mixed with dilution air, after which they are exhausted to the stack. The charcoal beds provide about 9.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> retention for krypton isotopes and 7.3 days retention for xenon isotopes. Gland seal and startup vacuum-pump gases are held up for approximately 1 3/4 minutes, to allow sufficient decay of N-16 and 0-19, and then passed directly to the stack for release.

1.6.2 Nuclear Safety Systems and Engineered Safeguards 1.6.2.1 Reactor Protection System The Reactor Protection System initiates a rapid, automatic shutdown (scram) of the reactor. This action is taken in time to prevent excessive fuel cladding damage and any nuclear system process barrier damage following abnormal operational transients. The Reactor Protection System overrides all operator actions and process controls.

1.6-10

BFN-27 1.6.2.2 Neutron Monitoring System Although not all of the Neutron Monitoring System qualifies as a nuclear safety system, those portions that provide high neutron flux signals to the Reactor Protection System do. The intermediate range monitors (IRM) and average power range monitors (APRM), which monitor neutron flux via incore detectors, signal the Reactor Protection System to scram in time to prevent excessive fuel cladding damage as a result of overpower transients.

1.6.2.3 Control Rod Drive System When a scram is initiated by the Reactor Protection System, it is the Control Rod Drive System that inserts the negative reactivity necessary to shut down the reactor.

Each control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high pressure water from an accumulator for each rod forces each control rod rapidly into the core.

1.6.2.4 Nuclear System Pressure Relief System A pressure relief system consisting of relief valves mounted on the main steam lines is provided to prevent excessive pressure inside the nuclear system following either abnormal operational transients or accidents.

1.6.2.5 [Deleted]

1.6.2.6 Primary Containment The design employs a pressure suppression primary containment which houses the reactor vessel, the reactor coolant recirculating loops, and other branch connections of the Reactor Primary System. The pressure suppression system consists of a drywell, a pressure suppression chamber which stores a large volume of water, connecting vents between the drywell and the pressure suppression chamber, isolation valves, containment cooling systems, and other service equipment. In the event of a process system piping failure within the drywell, reactor water and steam would be released into the drywell air space. The resulting increased drywell pressure would then force a mixture of air, drywell atmosphere, steam, and water through the vents into the pool of water in the pressure suppression chamber. The steam would condense in the pressure suppression pool, resulting in a rapid pressure reduction in the drywell. Air that was transferred to the pressure suppression chamber pressurizes the pressure suppression chamber, and is subsequently vented back to the drywell to equalize the pressure between the two vessels. Cooling systems are provided to remove heat from the reactor core, the drywell, and from the water in the pressure suppression chamber, and thus provide continuous cooling of the primary containment under accident conditions.

1.6-11

BFN-27 Appropriate isolation valves are actuated during this period to ensure containment of radioactive material, which might otherwise be released from the reactor containment during the course of the accident.

1.6.2.7 Primary Containment and Reactor Vessel Isolation Control System The Primary Containment and Reactor Vessel Isolation Control System automatically initiates closure of isolation valves to close off all potential leakage paths for radioactive material to the environs. This action is taken upon indication of a potential breach in the nuclear system process barrier.

1.6.2.8 Secondary Containment The secondary containment substructure consists of poured-in-place, reinforced concrete exterior walls that extend up to the refueling floor. The refueling room floor is also constructed of reinforced, poured-in-place concrete. The superstructure of the secondary containment above the refueling floor is a structural steel frame which supports metal roof decking, foamwall-stepped fascia panels, and insulated metal siding panels. The secondary containment structure completely encloses the primary containment drywells, fuel storage and handling facilities, and essentially all of the Core Standby Cooling Systems for the three units.

During normal operation and when isolated, the secondary containment is maintained at a negative pressure relative to the building exterior. Excessive pressure differentials are relieved by blowout panels in the metal siding.

1.6.2.9 Main Steam Line Isolation Valves Although all pipelines that both penetrate the primary containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steam lines, because of their large size, are given special isolation consideration. Two automatic isolation valves, each powered by both air pressure and spring force, are provided in each main steam line. These valves fulfill the following objectives:

a. Prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting from either a major leak from the steam piping outside the primary containment or a malfunction of the pressure control system resulting in excessive steam flow from the reactor vessel, and
b. Limit the release of radioactive materials by closing the primary containment barrier in case of a major leak from the nuclear system inside the primary containment.

1.6-12

BFN-27 1.6.2.10 Main Steam Line Flow Restrictors A venturi-type flow restrictor is installed in each steam line close to the reactor vessel. These devices limit the loss of coolant from the reactor vessel before the main steam line isolation valves are closed in case of a main steam line break outside the primary containment.

1.6.2.11 Core Standby Cooling Systems A number of standby cooling systems are provided to prevent excessive fuel clad temperatures in the event of a breach in the nuclear system process barrier that results in a loss of reactor coolant. The four Core Standby Cooling Systems are:

1. High Pressure Coolant Injection System (HPCI),
2. Automatic Depressurization System,
3. Core Spray System, and
4. Low Pressure Coolant Injection System (an operating mode of the Residual Heat Removal System) (LPCI).

1.6.2.11.1 High Pressure Coolant Injection System The HPCI System provides and maintains an adequate coolant inventory inside the reactor vessel to prevent fuel clad melting as a result of postulated small breaks in the nuclear system process barrier. A high-pressure system is needed for such breaks because the reactor vessel depressurizes slowly, preventing low-pressure systems from injecting coolant. The HPCI includes a turbine-pump powered by reactor steam. The system is designed to accomplish its function on a short-term basis without reliance on plant auxiliary power supplies other than the DC power supply.

1.6.2.11.2 Automatic Depressurization System The Automatic Depressurization System acts to rapidly reduce reactor vessel pressure in a loss-of-coolant accident situation in which the HPCI fails to automatically maintain reactor vessel water level. The depressurization provided by the system enables the low pressure standby cooling systems to deliver cooling water to the reactor vessel. The Automatic Depressurization System uses some of the main steam relief valves which are part of the nuclear system pressure relief system. The automatic main steam relief valves are arranged to open upon conditions indicating both that a break in the nuclear system process barrier has occurred and that the HPCI System is not delivering sufficient cooling water to the 1.6-13

BFN-27 reactor vessel to maintain the water level above a preselected value. The Automatic Depressurization System will not be automatically activated unless either the core spray or LPCI system is operating.

1.6.2.11.3 Core Spray System The Core Spray System consists of two independent pump loops that deliver cooling water to spray spargers over the core. The system is actuated by conditions indicating that a breach exists in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water onto the core and preventing excessive fuel clad temperatures following a loss-of-coolant accident.

1.6.2.11.4 Low Pressure Coolant Injection Low Pressure Coolant Injection is an operating mode of the Residual Heat Removal System (RHR) but is discussed here because the LPCI mode acts as an engineered safeguard in conjunction with the other standby cooling systems. LPCI uses the pump loops of the RHR to inject cooling water at low pressure into the reactor recirculation loops. LPCI is actuated by conditions indicating a breach in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced. LPCI operation, together with the core shroud and jet pump arrangement, provides the capability of core reflooding, following a loss-of-coolant accident, in time to prevent excessive fuel clad temperatures.

1.6.2.12 Residual Heat Removal System (Containment Cooling)

The containment cooling subsystem is placed in operation to limit the temperature of the water in the pressure suppression pool following a design basis loss-of-coolant accident. In the containment cooling mode of operation, the RHR main system pumps take suction from the pressure suppression pool and pump the water through the RHR heat exchangers, where cooling takes place by transferring heat to the RHR service water system. The fluid is then discharged back to the pressure suppression pool.

Another portion of the RHR is provided to spray water into the primary containment as an augmented means of removing energy from the containment following a loss-of-coolant accident. This capability is placed into service as required by manual operator action.

1.6.2.13 Control Rod Velocity Limiter A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its control rod drive.

1.6-14

BFN-27 The rate of reactivity insertion resulting from a rod drop accident is limited by this action. The limiters are passive components.

1.6.2.14 Control Rod Drive Housing Supports Control rod drive housing supports are located underneath the reactor vessel near the control rod housings. The supports limit the travel of a control rod in the event that a control rod housing is ruptured. The supports prevent a nuclear excursion as a result of a housing failure, thus protecting the fuel barrier.

1.6.2.15 Standby Gas Treatment System (SGTS)

The system provides a means of removing radioactive material from the secondary containment by filtration and exhausting to the atmosphere through the plant stack in the event of accidental release. Three trains, any 2 of which can provide 100 percent design flow, consisting of a moisture separator, heater, particulate and charcoal filters, and blower, are provided. The results of laboratory carbon sample analysis shall show 90 percent radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989. The blowers are powered from independent, safety-related power supplies. The SGTS is a Class I system.

1.6.2.16 Standby AC Power Supply The standby AC power supply consists of eight diesel generator sets. The diesel generators are sized so that they can supply all necessary power requirements for one unit under design basis accident conditions, plus necessary loads for safe shutdown of the other two units. The diesel generators are specified to start up and reach rated speed within ten seconds. The diesel generator system is arranged with eight independent 4160-V load buses, each connected to one diesel generator.

1.6.2.17 DC Power Supply Eleven 250-V batteries, associated chargers, and distribution systems (3 unit batteries, 3 station batteries, and 5 batteries supplying control power for the 4160-V and 480-V shutdown boards) are provided for the plant. The various safety-related loads derive normal power from the batteries or their associated battery charger through distribution boards.

1.6.2.18 RHR Service Water System The RHR Service Water System is a Class I system that consists of four pairs of pumps located on the intake structure for pumping raw river water to the heat 1.6-15

BFN-27 exchangers in the RHR System and four additional pumps for supplying water to the Emergency Equipment Cooling Water System.

1.6.2.19 Emergency Equipment Cooling Water System This Class I system distributes cooling water supplied by the RHR Service Water System to essential equipment during normal and accident conditions.

1.6.2.20 Deleted 1.6.2.21 Reactor Building Ventilation Radiation Monitoring System The Reactor Building Ventilation Radiation Monitoring System consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the Reactor Building. Upon detection of high radiation, the Reactor Building is automatically isolated and the Standby Gas Treatment System is started.

1.6.3 Special Safety Systems 1.6.3.1 Standby Liquid Control System Although not intended to provide prompt reactor shutdown, the Standby Liquid Control System provides a redundant, independent, and different way from the control rods to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition (Mode 4).

The SLC system is also required to supply sodium pentaborate solution for post-LOCA events that involve fuel damage to maintain the suppression pool pH at or above 7.0 for 30 days. The sodium pentaborate solution is credited as a buffering agent to offset the post-LOCA production of acids.

1.6.3.2 Plant Equipment Outside the Control Room Sufficient local controls are provided to allow the plant to be shut down from outside the control room. The plant design does not preclude bringing the plant to the cold shutdown condition (Mode 4) from outside the control room.

1.6-16

BFN-27 1.6.4 Process Control and Instrumentation 1.6.4.1 Nuclear System Process Control and Instrumentation 1.6.4.1.1 Reactor Manual Control System The Reactor Manual Control System provides the means by which control rods are manipulated from the control room for gross power control. Only one control rod can be manipulated at a time. The Reactor Manual Control System includes the controls that restrict control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.6.4.1.2 Recirculation Flow Control System The Recirculation Flow Control System controls the speed of the reactor recirculation pumps. Adjusting the pump speed changes the coolant flow rate through the core. This effects changes in core power level.

1.6.4.1.3 Neutron Monitoring System The Neutron Monitoring System is a system of incore neutron detectors and out of core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level, for the entire range of flux conditions that may exist in the core. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux level indications during reactor startup and low power operation. The local power range monitors (LPRM) and average power range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. Rod block monitors (RBM) are provided to prevent rod withdrawal when reactor power should not be increased at the existing reactor conditions. The Traversing Incore Probe System (TIPS) provides a means for calibrating the LPRM portion of the neutron monitoring sensors.

1.6.4.1.4 Refueling Interlocks A system of interlocks that restricts the movements of refueling equipment and control rods when the reactor is in the refuel mode (Mode 5) is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling bridge, the refueling bridge hoists, the fuel grapple, control rods, and the service platform hoist.

1.6-17

BFN-27 1.6.4.1.5 Reactor Vessel Instrumentation In addition to instrumentation provided for the Nuclear Safety Systems and engineered safeguards, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. The instrumentation provided monitors reactor vessel pressure, water level, surface temperature, internal differential pressures and coolant flow rates, and top head flange leakage.

1.6.4.1.6 Process Computer System An online process computer is provided to monitor and log process variables, and to make certain analytical computations. The rodworth minimizer function of the computer prevents rod withdrawal/insertion under low power conditions, if the rod to be withdrawn/inserted is not in accordance with a preplanned pattern. The effect of the rod block is to limit the reactivity worth of the control rods by enforcing adherence to the preplanned rod pattern during startup or shutdown.

1.6.4.2 Power Conversion Systems Process Control and Instrumentation 1.6.4.2.1 Pressure Regulator and Turbine Generator Control The pressure regulation function of the turbine control system maintains control of turbine control valves to regulate pressure at the turbine inlet and therefore the pressure of the entire nuclear system. The turbine control system is an electrohydraulic control (EHC) system with an integral pressure regulation function.

When not in pressure control mode, the EHC system maintains a fixed load or speed of the turbine. In addition, the EHC system provides overspeed protection for large load rejections.

1.6.4.2.2 Feedwater Control System The three element controller is used to regulate the feedwater system so that proper water level is maintained in the reactor vessel. The controller uses main steam flow rate, reactor vessel water level, and feedwater flow rate signals. The feedwater control signal is used to control the speed of the steam turbine driven feedwater pumps.

1.6.4.3 Electrical Power Systems Process Control and Instrumentation Each generator neutral is grounded through a distribution transformer and a secondary loading resistor. Each generator is equipped with a shaftdriven alternator exciter, an exciter field circuit breaker, rectifiers, and voltage regulating equipment.

Current transformers are provided on the generator main and neutral terminals for relaying and metering.

1.6-18

BFN-27 Highspeed relays provide protection for the generator stator windings against faults.

Incoming power is received from the 500-kV and 161-kV systems. The TVA 161-kV network receives power via the 161-kV switchyard. Two 161-kV lines terminate at separate buses which are connected by a circuit breaker. Two common station service transformers are energized from these buses. Normally, the switchyard will be operated with the breaker closed and both transformers energized. Disconnect switches are provided to permit either incoming line to be isolated from the switchyard and both transformers supplied from the remaining line.

Output from the generators is fed into the TVA system by seven 500-kV lines via the 500-kV switchyard. The switchyard has a main and transfer zigzag bus arrangement. The two main bus sections are physically separated, and the transfer bus sections are separated from the main bus section by sectionalizing disconnect switches. Normally, the main and transfer bus sections are tied together through their respective disconnect switches.

1.6.4.4 Radiation Monitoring and Control 1.6.4.4.1 Process Radiation Monitoring Radiation monitors are provided on various lines to monitor either for radioactive materials released to the environs via process liquids and gases or for process system malfunctions. The following monitors are provided:

Main Stack Radiation Monitors, Air Ejector Offgas Radiation Monitor, Raw Cooling Water System Discharge Radiation Monitor, Reactor Building Closed Cooling Water System Radiation Monitor, Liquid Radwaste System Radiation Monitor, RHR Service Water System Radiation Monitors, and Plant Ventilation Exhaust Radiation Monitors.

1.6.4.4.2 Area Radiation Monitors A number of radiation monitors are provided to monitor for abnormal radiation at various locations in the Reactor Building, Turbine Building, and Radwaste Building.

These monitors annunciate alarms when abnormal radiation levels are detected.

1.6-19

BFN-27 1.6.4.4.3 Site Environs Radiation Monitors Radiation monitoring stations are provided to monitor the effects from natural and plant radiation sources. The stations employ appropriate devices to collect samples as well as measure direct radiation effects which can be used to determine changes in environmental radioactivity levels.

1.6.4.4.4 Liquid Radwaste System Control Liquid wastes to be discharged are handled on a batch basis, with protection against accidental discharge provided by procedural controls. Instrumentation with alarms to detect abnormal concentration and terminate release of liquid waste is provided.

1.6.4.4.5 Solid Radwaste Control The Solid Radwaste System collects, processes, stores, and prepares solid radioactive waste materials for offsite shipment. Wastes are handled on a batch basis, and radiation levels of the various batches are determined by the operating personnel.

1.6.4.4.6 Gaseous Radwaste System Control The Gaseous Radwaste System is continuously monitored by a radiation monitor located downstream of the recombiner system water separator, a monitor located downstream of the charcoal/absorbers but upstream of the afterfilters, and the main stack radiation monitor. Each of these monitors alarms on high radiation level.

In addition, a high level signal from the monitor downstream of the air ejectors automatically isolates the Gaseous Radwaste System by closing a valve in the line between the after-filters and the stack. This action causes an increase in condenser back pressure.

Hydrogen concentration in the gas downstream of the recombiners is continuously monitored. Although an explosion is not likely, temperature and pressure instrumentation in the line upstream of the decay pipe, in response to an explosion, causes valves downstream of the air ejectors to automatically isolate. These actions stop the supply of hydrogencontaining gas, and minimize release of radioactivity from a damaged filter. A main steamline high radiation condition will automatically close a valve between the main condensers and the mechanical vacuum pump. In addition, the mechanical vacuum pump is stopped.

1.6-20

BFN-27 1.6.5 Auxiliary Systems 1.6.5.1 Normal Auxiliary AC Power System The normal power source for unit auxiliaries is the 20.7- to 4.16-kV unit station service transformers. This source is connected to each unit generator's output leads. The startup power source for unit auxiliaries is the 500 kV system, with backup from the 161 kV switchyard through the common station service transformers.

1.6.5.2 Reactor Building Closed Cooling Water System The Reactor Building Closed Cooling Water System (RBCCWS) provides cooling water to designated auxiliary plant equipment located in the primary and secondary containments. The cooling water is available to the nuclear system auxiliaries under normal and accident conditions.

1.6.5.3 Raw Water Systems The Raw Cooling Water System is provided to remove heat from turbine associated equipment and accessories located in and adjacent to the Turbine Building, from the Reactor Building Closed Cooling Water System heat exchangers, and from other reactor associated equipment. The Raw Cooling Water System pumps are located in the Turbine Building and are supplied with river water from the condenser circulating water conduits. Three pumps are provided for each unit, with one spare provided for Units 1 and 2 and two spares for Unit 3.

A Raw Service Water System, consisting of four pumps, supplies river water from the condenser circulating water conduits for yard watering, cooling for miscellaneous plant equipment requiring small quantities of cooling water, washdown services in unlimited access areas, and provides a means of pressurizing the raw water Fire Protection System. The Raw Service Water System also serves as a charging source for the RHR Service Water and Emergency Equipment Cooling Water Systems.

1.6.5.4 Fire Protection Systems A high pressure, raw water Fire Protection System provides water for fixed water spray, water sprinkler, aqueous film forming foam, and water fog systems, and to fire hoses and hydrants located throughout plant buildings and the surrounding yard.

Fixed CO2, halon, and portable fire extinguishers furnish protection for hazards where use of water is not desirable. Fire detection, annunciation, and initiation systems are installed in selected areas of the Reactor Building, Control Building, 1.6-21

BFN-27 intake pumping station, cable tunnel to intake pumping station, Diesel Generator Buildings, and Turbine Building.

1.6.5.5 Heating, Ventilating, and Air Conditioning Systems Heating, Ventilating, and Air Conditioning Systems are provided for the Reactor Building, Turbine Building, Radwaste Building, and Control Building. The design of these systems varies; but in all cases, they maintain the indoor environment necessary for equipment protection and personnel comfort. In areas where significant airborne activity is expected, these systems limit the spread of contamination and filter the exhaust air before discharge.

1.6.5.6 New and Spent Fuel Storage A dry vault in the Reactor Building is provided for storage of new fuel. The new fuel is normally transferred directly to the spent fuel storage pool upon receipt. Fuel transfer during refueling is conducted underwater. Irradiated (spent) fuel is stored underwater in the Reactor Building until prepared for shipment from the site.

1.6.5.7 Fuel Pool Cooling and Cleanup System A Fuel Pool Cooling and Cleanup System is provided to remove decay heat from spent fuel stored in the fuel pool and to maintain a specified water temperature, purity, clarity, and level.

1.6.5.8 Control and Service Air Systems Clean, dry, control air is provided to pneumatically operated instruments and controls throughout the plant and yard. Each reactor unit has a drywell control air system that provides control air for the equipment inside its drywell. Service air outlets are provided throughout the plant.

1.6.5.9 Demineralized Water System A makeup demineralized water unit is used to furnish a supply of high purity water for makeup of the primary coolant systems, the Reactor Building Closed Cooling Water Systems, the pressure suppression chambers, and the Standby Liquid Control Systems. The water is also used for radioactive decontamination work and preoperational cleaning of reactor and piping systems.

1.6.5.10 Potable Water and Sanitary Systems These systems provide potable water from a nearby municipal water system for use in the plant plumbing systems and sewage treatment in a 65,000 gallon per day biological treatment system.

1.6-22

BFN-27 1.6.5.11 Equipment and Floor Drainage System Radioactive drainage from equipment leaks and from areas which may contain radioactive materials is collected and routed to shielded sumps. This waste is then pumped to drain collection tanks in the Radwaste Building, where it is treated and returned for reuse in the plant or discharged to the river.

Nonradioactive drainage is collected in drain sumps and discharged to the condenser circulating water discharge tunnels.

1.6.5.12 Process Sampling Systems These systems provide samples of process liquids and gases to obtain data from which the performance of the plant, items of equipment, and systems may be determined. Sampling is continuous or periodic as appropriate. These systems will function at all times and under all operating conditions.

1.6.5.13 Communications Systems An extensive, private telephone system, along with a paging system, sound powered telephone systems, and closed circuit television systems, provides complete communications throughout the plant.

1.6.6 Shielding Plant shielding allows personnel access to the plant to perform maintenance and carry out operational duties, with personnel exposures limited to the values given in Table 12.3-1.

1.6.7 Implementation of Loading Criteria When correctly installed in a suitable facility, structures, and equipment are designed to substantially resist mechanical damage due to loads produced by mechanical and thermal forces. For the purpose of categorizing mechanical strength designs for these loads, the following definitions are established.

a. Class I This class includes those structures, equipment, and components whose failure or malfunction might cause, or increase the severity of, an accident 1.6-23

BFN-27 which would endanger the public health and safety. This category includes those structures, equipment, and components required for safe shutdown and isolation of the reactor.

b. Class II This class includes those structures, equipment, and components which are important to reactor operation, but are not essential for preventing an accident which would endanger the public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade the integrity of any item designated Class I.

The loading conditions may be divided into four categories: (1) normal, (2) upset, (3) emergency, and (4) faulted conditions. These categories are generically described, and their meaning is expanded in quantitative, probabilistic language in Appendix C. The purpose of this expansion is to clarify the classification of any hypothesized accident or sequence of loading events. Event probability is used to establish meaningful and adequate safety factors for structural design so that the appropriate structural safety margins are applied.

1.6-24

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BFN-21 Figure 1.6-20 (Deleted by Amendment 21)

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BFN-21 Figure 1.6-22 (Deleted by Amendment 21)

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Figure 1.6-27

BFN-28 Figure 1.6-28 3952 MWt

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BFN-28 Figure 1.6-29 Sheet 2 Heat Balance Values (100%) Power Turbine Generator

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BFN-17 1.7 COMPARISON OF PRINCIPAL DESIGN CHARACTERISTICS This section highlights the principal design features of the plant 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 are historical and represent the plant as originally designed.

The parameter values presented in Tables 1.7-1 to 1.7-5 for the various nuclear plants are the values used in the design of these plants. Since the various owner-utilities were not contacted, no guarantee is given that these parameter values are current. The information contained in this section is, therefore, maintained for historical purpose only. More updated information can be found in the specific chapters dealing with specific topics.

1.7.1 Nuclear System Design Characteristics Table 1.7-1 summarizes the design and operating characteristics for Browns Ferry Nuclear Plant. The same characteristics are presented for the nuclear system of Duane Arnold Energy Center, Cooper Nuclear Station, Vermont Yankee Nuclear Power Station, and Hatch Nuclear Plant Unit 1.

1.7.2 Power Conversion Systems Design Characteristics Table 1.7-2 presents a summary of the power conversion systems design characteristics for the plant and compares these with Duane Arnold Energy Center, Cooper Nuclear Station, Vermont Yankee Nuclear Power Station, and Hatch Nuclear Plant Unit 1.

1.7.3 Electrical Power Systems Design Characteristics Table 1.7-3 is a summary and comparison of the electrical power systems design characteristics of the plant and the same four similar facilities.

1.7.4 Containment Design Characteristics Table 1.7-4 summarizes the design characteristics for the primary and secondary containments of the Browns Ferry Nuclear Plant. Design characteristics are also presented for the primary and secondary containment systems employed for Hatch Unit 1, Vermont Yankee Nuclear Power Station, Cooper Station, and Duane Arnold 1.7-1

BFN-17 Energy Center. In addition, data is given for the type, construction, and height of the elevated release point for the above plants.

1.7.5 Structural Design Characteristics Table 1.7-5 is a summary and comparison of the seismic and wind design factors considered in the structural design of Browns Ferry Nuclear Plant and the above similar plants.

1.7.6 Discussion of Core Design Improvement Numerous improvements have been made to the core design of Browns Ferry subsequent to receipt of the operating license for each of the three units. A general description of reload fuel designs presently used in Browns Ferry is given in Chapter 3. The specific fuel types loaded in each unit along with analytical results of the cycle-specific reload core design and licensing analyses are given in the applicable Supplemental Reload Licensing Report (SRLR). The current SRLR for each BFN unit is included in Appendix N of the FSAR.

1.7-2

BFN-17 TABLE 1.7-1 (Sheet 1)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD THERMAL AND HYDRAULIC DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Rated Power, MWt 3293 2436 1593 2381 1593 Design Power, MWt 3440 2537 1665 2500 1670 Steam Flow Rate, lb/hr 13.37 x 106 10.03 x 106 6.43 x 106 9.81 x 106 6.847 x 106 Core Coolant Flow Rate, lb/hr 102.5 x 106 75.5 x 106 48.5 x 106 74.5 x 106 48.5 x 106 Feedwater Flow Rate, lb/hr 13.315 x 106 10.445 x 106 6.43 x 106 9.81 x 106 6.77 x 106 Feedwater Temperature, F 378.4 387.4 372 367 420 System Pressure, Nominal in Steam Dome, psia 1020 1020 1020 1020 1020 Average Power Density, kW/liter 49.69/49.46/ 51.2 50.8 51.2 50.9 49.2 Maximum Thermal Output, kW/ft 18.5 (7x7)/13.4 18.3 18.37 18.5 18.5 (8x8)

Average Thermal Output, kW/ft 7.050 (7x7)/ 7.114 7.1 7.079 7.079 5.59 (8x8)

Average Heat Flux, Btu/hr-ft2 148937/142007/ 164,734 163,900 164,500 163,933 143635 Maximum UO2 Temperature, F 4430 4430 4430 4430 4430 Average Volumetric Fuel Temperature, F 1210 1210 1210 1210 1210 Average Fuel Rod Surface Temperature, F 560 560 560 560 560 Minimum Critical Power Ratio (MCPR)(1) >1.07 >1.9 >1.9 >1.9 >1.9 Coolant Enthalpy at Core Inlet, Btu/lb 521.3 526.2 522.9 520.1 525.6 Core Maximum Exit Voids Within Assemblies 79 79 79 79 79 Core Average Exit Quality, % Steam 13.2 13.9 13.6 13.2 14.3

BFN-17 TABLE 1.7-1 (Sheet 2)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD THERMAL AND HYDRAULIC DESIGN (Cont-d) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Design Power Peaking Factors Transverse Peaking Factor 1.4 1.4 1.4 1.4 1.405 Local Peaking Factor 1.24 1.24 1.24 1.24 1.24 Axial Peaking Factor 1.5 1.5 1.5 1.5 1.5 Total Peaking Factor 2.63 2.6 2.6 2.6 2.6 NUCLEAR DESIGN (First Core)

Water/UO2 Volume Ratio (Cold) 2.43 Type I 2.41 2.41 2.41 2.41 2.53 Type II & III Reactivity with Strongest Control Rod <0.99 <0.99 <0.99 <0.99 <0.99 Out, keff Moderator Temperature Coefficient At 68F, k/k - F Water -3.5 x 10-5 -3.5 x 10-5 -5.0 x 10-5 -3.5 x 10-5 -3.5 x 10-5 Hot, no voids, k/k - F Water -11.6 x 10-5 -11.6 x 10-5 -17.0 x 10-5 -11.6 x 10-5 -11.6 x 10-5 Moderator Void Coefficient Hot, no voids, k/k - % Void -8.7 x 10-4 -8.7 x 10-4 -1.0 x 10-3 -8.7 x 10-4 -8.7 x 10-4 At Rated Output, k/k - % Void -1.05 x 10-3 -1.05 x 10-3 -1.5 x 10-3 -1.05 x 10-3 -1.05 x 10-3 Fuel Temperature Doppler Coefficient At 68F, k/k - F Fuel -0.9 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 Hot, No Void, k/k - F Fuel -1.0 x 10-5 -1.2 x 10-5 -1.2 x 10-5 -1.2 x 10-5 -1.2 x 10-5 At Rated Output, k/k - F Fuel -0.9 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 Initial Average U-235 Enrichment, W/O 2.19% 2.30% 2.50% 2.15% 2.25%

Fuel Average Discharge Exposure, MWD/Ton 19,000 19,000 19,000 19,000 18,350 (6)

Nuclear Design (Reload Core) See applicable Nuclear Design Reports.

BFN-17 TABLE 1.7-1 (Sheet 3)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Fuel Assembly Number of Fuel Assemblies 764 560 368 548 368 Fuel Rod Array 7 x 7 or 8 x 8 7x7 7x7 7x7 7x7 Overall Dimensions, inches 175.98 175.98 175.98 175.98 175.98 Weight of UO2 per Assembly, pounds See applicable Undished - Undished - 487.4 Undished -

Nuclear Design 490.35 487.4 490.35 (6) Reports Dished - Dished -

483.42 483.42 Weight of Fuel Assembly, pounds 681 Undished - Undished - 682 Undished -

681.48 682 681.48 Dished - Dished -

674.55 674.55 Fuel Rods Number per Fuel Assembly 49 or 64* 49 49 49 49 (mixed cores) 1.483 Outside Diameter, inch 0.563 0.563 0.563 0.563 0.563 Clad Thickness, inch 0.032 0.032 0.032 0.032 0.032 Gap - Pellet to Clad, inch 0.006/0.009 0.006 0.006 0.006 0.006 Length of Gas Plenum, inches 16/9.48 16 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

  • Two different 8 x 8 fuel bundle arrangements are used. One uses 63 fuel rods and 1 water rod; the other uses 62 fuel rods and 2 water rods.

BFN-17 TABLE 1.7-1 (Sheet 4)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Fuel Pellets Material Uranium Dioxide Uranium Dioxide Uranium Dioxide Uranium Dioxide Uranium Dioxide Density, % of theoretical 94% 93% 93% 93% 93%

Diameter, inch 0.410 0.487 0.487 0.487 0.487 Length, inch 0.410 0.75 0.75 0.75 0.75 Fuel Channel Overall Dimension, inches (length) 166.906 166.906 166.906 166.096 166.906 Thickness, inch 0.080 0.080 0.080 0.080 0.080 Cross-Section Dimensions, inches 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core Assembly Fuel Weight as UO2, pounds 361,837 272,849 179,370 267,095 179,298 Zirconium Weight, pounds 140,397 96,370 63,300 94,305 63,300 (Zr.2 + Zr.4 Spacers)

Core Diameter (equivalent), inches 187.1 160.2 129.9 158.5 129.9 Core Height (Active Fuel), inches 144 - 150 144 144 144 144 Reactor Control System Method of Variation of Reactor Power Movable Control Movable Control Movable Control Movable Control Movable Control Rods and Variable Rods and Variable Rods and Variable Rods and Variable Rods and Variable Coolant Pumping Coolant Pumping Coolant Pumping Coolant Pumping Coolant Pumping Number of Movable Control Rods 185 137 89 137 89 Shape of Movable Control Rods Cruciform Cruciform Cruciform Cruciform Cruciform Pitch of Movable Control Rods 12.0 12.0 12.0 12.0 12.0

BFN-17 TABLE 1.7-1 (Sheet 5)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Reactor Control System (Cont'd)

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 Piston Locking Piston Locking Piston Locking Piston Locking Piston Supplementary Reactivity Control Grandolinia 156 Burnable Poison Flat, boron-stainless steel control curtains In-Core Neutron Instrumentation Number of In-Core Neutron Detectors (Fixed) 172 124 80 124 80 Number of In-Core Detector Assemblies 43 31 20 31 20 Number of Detectors Per Assembly 4 4 4 4 4 Number of Flux Mapping Neutron Detectors 5 4 3 4 3 Range (and Number) of Detectors Source Range Monitor Source to Source to Source to Source to Source to 0.001% power 0.001% power 0.001% power 0.001% power 0.001% power (4) (4) (4) (4) (4)

Intermediate Range Monitor 0.0001% to 10% 0.0001% to 10% 0.0001% to 10% 0.0001% to 10% 0.0001% to 10%

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

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

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

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

power (U1-6; U2-4; power (6) power (6) power (6) power (6)

U3-6)

BFN-17 TABLE 1.7-1 (Sheet 6)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD REACTOR VESSEL DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER 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. 20 - 11 18 - 2 17 - 2 18 - 2 15 - 3 Inside Height, ft-in. 73 1/2 69 - 4 63 - 1.5 69 - 4 66 - 4 Side Thickness (including clad) 6.313 5.531 5.187 5.531 5.625 Minimum Clad Thickness, inches 1/8 1/8 1/8 1/8 1/8 REACTOR COOLANT RECIRCULATION DESIGN Number of Recirculation Loops 2 2 2 2 2 Design Pressure Inlet Leg. psig 1148 1148 1175 1148 1148 Outlet Leg. psig 1326 1274 1274 1274 1268 CORE MECHANICAL DESIGN Design Temperature, F 562 562 562 562 562 Pipe Diameter Max. inches 28 28 28 28 22 Pipe Material 304/316 304/316 304/316 304/316 304/316 Recirculation Pump flow Rate, GPM 45,200 45,200 32,500 45,200 27,100 Number of Jet Pumps in Reactor 20 20 20 20 16 MAIN STEAM LINES Number of Steam 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 26 24 20 24 20 Pipe Material Carbon Steel (ASTM A155 KC70 or ASTM A106 Grade B)

BFN-17 TABLE 1.7-1 (Sheet 7)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE STANDBY COOLING SYSTEMS UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER (These systems are sized on design power)

Core Spray System Number of Loops 2 2 2 2 2 Flow Rate (gpm) 6250 at 4625 at 3000 at 4500 at 3020 at 105 psid 120 psid 136 psid 115 psid 127 psid High Pressure Coolant Injection system (No.) 1 1 1 1 1 Number of Loops 1 1 1 1 1 Flow Rate (gpm) 5000 4250 4250 4220 2980 Automatic Depressurization system (No.) 1 1 1 1 1 Low Pressure Coolant Injection (No.) 1 1 1 1 1 Number of Pumps 4 4 4 4 4 Flow Rate (gpm/pump) 10,800 gpm 7700 at 4800 at 7000 at 4800 at (1 pump per loop) 20 psid 20 psid 20 psid 20 psid 20,000 gpm (2 pumps per loop)

AUXILIARY SYSTEMS Residual Heat Removal System Reactor Shutdown Cooling (number of pumps) 4 4 4 4 4 Flow Rate (gpm/pump)(2) 10,000 7,700 7,000 7,700 4,800 Capacity (Btu/hr/heat exchanger)(3) 70 x 106 32 x 106 57.5 x 106 70 x 106 35 x 106 Number of heat exchangers 4 2 2 2 2 Primary Containment Cooling Flow rate (gpm)(4) 32,000 30,800 28,000 30,800 19,200

BFN-17 TABLE 1.7-1 (Sheet 8)

COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

(Parameters are related to Rated Power Output for a single plant unless otherwise noted)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD AUXILIARY SYSTEMS (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER RHR Service Water System Flow Rate (gpm/pump) 4,500 8,000 2,700 8,000 2,500 Number of pumps 12(5) 4 4 4 4 Reactor Core Isolation Cooling System Flow Rate (gpm) 616 at 400 at 400 416 at 416 1120 psid 1120 psid 1120 psid Fuel Pool Cooling and Cleanup system 6 6 6 6 6 Capacity (BTU/hr) 8.8 x 10 3.3 x 10 2.37 x 10 3.4 x 10 2.37 x 10 (1) The operating MCPR limits are subject to change from one cycle to the next and also from one part of the current cycle to the next. The appropriate value for MCPR may be obtained by consulting the applicable current Reload Licensing Amendment.

(2) Capacity during reactor flooding mode with three of four pumps running.

(3) Capacity during post-accident cooling mode with 165F shell side inlet temperature, maximum service water temperature, and 1 RHR pump and 1 RHR service water pump in operation.

(4) The existing design requires 16,000 gpm (2 pumps, 1 loop) to ensure torus water temperature is maintained within acceptable limits for following all postulated events.

(5) For all three units.

(6) See Appendix N

BFN-17 TABLE 1.7-2 COMPARISON OF POWER CONVERSION SYSTEMS DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

Browns Ferry Duane Arnold TURBINE GENERATOR Each Unit Hatch Unit 1 Vermont Yankee Cooper Station Energy Center Design Power, MWt 3440 2537 1665 2487 1670 Design Power, MWe 1152 849 564 836 597 Generator Speed, rpm 1800 1800 1800 1800 1800 6 6 6 6 6 Design Steam Flow, lb/hr 14.035 x 10 10.48 x 10 6.423 x 10 10.049 x 10 6.696 x 10 Turbine Inlet Pressure, psia 965 970 950 970 950 TURBINE BYPASS SYSTEM Capacity, percent of turbine design steam flow 25 25 100 25 25 MAIN CONDENSER 6 6 6 6 6 Heat removal capacity, Btu/hr 7,770 x 10 5,800 x 10 3,500 x 10 5,367 x 10 3,681 x 10 CIRCULATING WATER SYSTEM Number of Pumps 3 3 3 4 2 or more Flow Rate gpm/pump 220,000 185,000 117,000 162,500 130,000 or less CONDENSATE AND FEEDWATER SYSTEMS 6 6 6 6 6 Design Flow Rate, lb/hr 13.845 x 10 10.096 x 10 6.4 x 10 9.773 x 10 7.146 x 10 Number Condensate Pumps 3 3 2 3 2 Number Condensate Booster Pumps 3 - ---

Number Feedwater Pumps 3 2 2 2 2 Condensate Pump Drive AC power AC power AC power AC power AC power Condensate Booster Pump Drive AC power - - - -

Feedwater Pump Drive Turbine Turbine AC power Turbine AC power

BFN-17 TABLE 1.7-3 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD TRANSMISSION SYSTEM NUCLEAR PLANT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Outgoing lines (number-rating) 7-500kV 2-230kV 2-345kV 4-345kV 2-345kV NORMAL AUXILIARY AC POWER Incoming lines (number-rating) 2-161kV 2-230kV 2-345kV 1-115kV 2-345kV 1-230kV 1-69kV 3-161kV 1-115kV 1-4160kV Auxiliary transformers 3 1 1 1 2 Startup transformers 2 2 1 2 1 STANDBY AC POWER SUPPLY Number diesel generators 8 3 2 4 2 Number of 4160V Shutdown buses 8 3 2 2 2 Number of 480V Shutdown buses 6 4-660V 3 3 3 DC POWER SUPPLY Number of 125V or 250V batteries* 6 2 2 2 2 Number of 125V or 250V buses* 6 4 4 4 2

  • 3 of the 6 250V systems are qualified

BFN-17 TABLE 1.7-4 Sheet 1 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

BROWNS FERRY VERMONT COOPER Duane Arnold PRIMARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION Energy Center 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 56 56 56 56 56 Internal Design Pressure (psig)

Pressure Suppression chamber - 2 2 2 2 2 External Design Pressure (psig)

Drywell-Internal Design Pressure (psig) 56 56 56 56 56 Drywell-External Design Pressure (psig) 2 2 2 2 2 Drywell Free Volume (ft3) 159,000 146,400 134,000 145,430 130,930 Pressure Suppression chamber 119,000 101,410 99,000 109,810 94,630 Free Volume (ft3), minimum Pressure Suppression Pool Water 128,700 86,660 78,000 87,660 61,500 Volume (ft3), maximum Submergence of Vent Pipe Below 4 4 4 4 4 Pressure Pool Surface (ft), nominal Design Temperature of Drywell (F) 281 281 281 281 281 Design Temperature of Pressure 281 281 281 281 281 Suppression Chamber (F)

  • Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-4 (Cont'd)

Sheet 2 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD PRIMARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Downcomer Vent Pressure Loss Factor 4.1 6.21 6.21 6.21 6.21 Break Area/total Vent Area 0.017 0.019 0.019 0.019 0.019 Calculated Maximum Pressure After Blow- 49.6 45 35 46 45 down Drywell (psig)

Pressure Suppression chamber (psig) 27 28 22 28 29 Initial Pressure Suppression Pool 40 50 35 50 50 Temperature Rise (F)

Leakage Rate (% Free Volume/Day 0.5 0.5 0.5 0.5 0.5 at 56 psig and 281F SECONDARY CONTAINMENT Type Controlled Leakage, Controlled Leakage, Controlled Leakage, Controlled Leakage, Controlled Leakage, Elevated Release Elevated Release Elevated Release Elevated Release Elevated Release Construction Reinforced Reinforced Reinforced Reinforced Reinforced Concrete Concrete Concrete Concrete Concrete Upper Levels Steel Super- Steel Super- Steel Super- Steel Super- Steel Super-structure and structure and structure and structure and structure and Siding Siding Siding Siding Siding Roof Steel Decking Steel Steel Steel Steel with Builtup Sheeting Sheeting Sheeting Sheeting Composition Roof Internal Design Pressure (psig) +7 to -5 in. H2O 0.25 0.25 0.25 0.25 Design Inleakage Rate (% Free 100 100 100 100 100 Volume/Day at 0.25 inches H2O)

  • Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-4 (Cont'd)

Sheet 3 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

BROWNS FERRY VERMONT COOPER Duane Arnold SECONDARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION Energy Center ELEVATED RELEASE POINT Type Stack Stack Stack Stack Stack Construction Reinforced Steel Steel Steel Steel Concrete Height (above ground) 600 feet 150 meters 318 feet 100 meters 100 meters

  • Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-5 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry)

BROWNS FERRY VERMONT COOPER DUANE ARNOLD SEISMIC DESIGN NUCLEAR PLANT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Operating Basis Earthquake (horizontal g) 0.10 0.08 0.07 0.10 0.06 Design Basis Earthquake (horizontal g) 0.20 0.15 0.14 0.20 0.12 WIND DESIGN Maximum sustained (mph) 100 105 80 100 105 Tornadoes (mph) 300 300 300 300 300

BFN-21 1.8

SUMMARY

OF RADIATION EFFECTS 1.8.1 Normal Operation The gaseous and liquid radioactive waste systems are designed so that dose to any offsite person will not exceed that permitted within the limits specified in the Offsite Dose Calculation Manual (ODCM), applicable limits in the plant technical specifications, and technical requirements manual. The expectancy, based on operating experience, is that dose to any off-site person from gaseous waste discharge will not average more than a small fraction of the permissible dose, and that concentrations of liquid waste at the point of discharge will average less than 1% of the concentrations permitted by 10 CFR 20. Both effects are only a small fraction of the effect of natural background radiation.

1.8.2 Abnormal Operational Transients A design objective is to avoid fuel damage as a result of abnormal operational transients. Analyses of these events, which are described in the "Plant Safety Analysis", show that abnormal operational transients do not result in any significant increase of radioactive material release to the environs over that experienced during normal operation.

1.8.3 Accidents The ability of the plant to withstand the consequences of accidents without posing a hazard to the health and safety of the public is evaluated by analyzing a variety of postulated accidents. The calculated consequences of the design basis accidents, which result in the greatest potential off-site radiation exposures, are presented in Chapter 14. These doses are substantially below the guideline doses given in 10 CFR 50.67.

1.8-1

BFN-17 1.9 PLANT MANAGEMENT The summary of information originally presented in this section was superfluous and has been deleted. The Browns Ferry Nuclear Plant Organization is presented in Tennessee Valley Authority Topical Report, TVA-NPOD89-A Nuclear Power Organization Description.

1.9-1

BFN-16 1.10 QUALITY ASSURANCE PROGRAM Appendix D describes the comprehensive quality assurance plan that meets the requirements of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."

1.10-1

BFN-16 1.11 IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERN -

SUMMARY

1.11.1 General The information presented in this section and FSAR Appendix I had direct applicability to the licensing efforts expended by TVA at the time of the preparation and submittal of the FSAR. Section 1.11 and Appendix I are retained in the FSAR for historical and traceability reasons only.

The design of the General Electric boiling water reactors for this station is based upon proven technological concepts developed during the development, design, and operation of numerous similar reactors. The AEC in reviewing the Browns Ferry docket at the Construction Permit stage identified several areas where further R&D efforts were required to more definitely assure safe operation of this station. Also, both the AEC Staff and the Advisory Committee for Reactor Safeguards had, in their review of this and more recent reactor projects, identified several additional technical areas for which further detailed support information had to be obtained. All of these development efforts thus were of three general types: (1) those which pertain to the broad category of water-cooled reactors; (2) those which pertain specifically to boiling water reactors; and (3) those which have been noted particularly for a facility during the construction or operating permit licensing activities by the AEC Staff and ACRS reviews.

Appendix I of this FSAR provides a complete, comprehensive examination and discussion of each of these concern areas, indicating the resolution accomplished or planned at the time of FSAR preparation. A summary conclusion of this analysis is provided in this subsection by Tables 1.11-1 through 1.11-6. The concerns have been subdivided as follows:

a. Areas specified in the Browns Ferry AEC-ACRS Construction Permit Reports.

(See Table 1.11-2.)

b. Areas specified in the Browns Ferry AEC Staff Construction Permit Safety Evaluation Reports. (See Tables 1.11-3 and 1.11-4.)
c. Areas specified in other related AEC-ACRS construction permit and operating license reports. (See Table 1.11-5.)
d. Areas specified in other somewhat related AEC Staff construction permit and operating license safety evaluation reports. (See Table 1.11-6.)
e. Additional information available and supplied by General Electric on BWR.

(See Table 1.11-1.)

1.11-1

BFN-16 The scope of many of the areas of technology for items in a, b, and c above is discussed in Appendix I in detail and as part of an official response1 by General Electric to the various ACRS concern subjects.

General Electric has submitted many topical reports to the AEC in support of this application and those of other facilities. Table 1.11-1 provides a list of all topical reports submitted to the AEC on behalf of TVA while trying to obtain an operating license for Browns Ferry. Topical reports submitted to the NRC (and the former AEC) after receipt of the operating license are not included in this list but can be found from a search of TVA-NRC correspondence.

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

1.11-2

BFN-16 TABLE 1.11-1 (Sheet 1)

BROWNS FERRY NUCLEAR PLANT TOPICAL REPORTS SUBMITTED TO THE AEC IN SUPPORT OF DOCKET GE Report No. Title

1. APED 5286 Design Basis for Critical Heat Flux in Boiling Water Reactors (September 1966)
2. APED 5446 Control Rod Velocity Limiter (March 1967)
3. APED 5449 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 of ACRS Safety Concerns (April 1968) (Not Class I)
9. APED 5455 The Mechanical Effects of Reactivity Transients (January 1968)
10. 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 of Core 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)
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. SPED 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 of General Electric Boiling Water Reactor Main Steam Line Isolation Valves (March 1969)
22. APED 5756 Analytical Methods for Evaluating the Radiological Aspects of the General Electric Boiling Water Reactor (March 1969)
23. APED 5652 Stability and Dynamic Performance of the General Electric Boiling Water Reactor (April 1969)

BFN-16 TABLE 1.11-1 (Cont'd)

(Sheet 2)

BROWNS FERRY NUCLEAR PLANT TOPICAL REPORTS SUBMITTED TO THE AEC IN SUPPORT OF DOCKET GE Report No. Title

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. NEDO 10017 Field Testing Requirements for Fuel, Curtains and Control Rods (June 1969)
27. NEDO 10029 An Analytical Study on Battle Fracture of GE-BWR Vessel Subject to the Design Basis Accident (July 1969)
28. NEDO 10045 Consequences of a Steam Line Break for a General Electric Boiling Water Reactor (October 1969)
29. NEDO 10173 Current State of Knowledge High Performance BWR Zircaloy-Clad UO Fuel (May 2

1970)

30. NEDO 10139 Compliance of Protection Systems to Industry Criteria: General Electric BWR Nuclear Steam Supply System (June 1970)
31. NEDO 10179 Effects of Cladding Temperature and Material on ECCS Performance (June 1970)
32. NEDO 10174 Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor (May 1970)
33. NEDO 10189 An Analysis of Functional Common-Mode Failures in GE BWR Protection and Control Instrumentation (July 1970)
34. NEDO 10208 Effects of Fuel Rod Failure on ECCS Performance (August 1970)
35. NEDO 10320 The General Electric Pressure Suppression Containment Analytical Model (April 1971)
36. NEDO 10320 The General Electric Pressure Suppression Containment Analytical Supplement 1 Model(May 1971)
37. NEDO 10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors (April 1971)
38. NEDO 10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for Supplement 1 General Electric Boiling Water Reactors (April 1971)
39. NEDO 10349 Analysis of Anticipated Transients Without Scram (March 1971)

BFN-16 TABLE 1.11-2 (Sheet 1)

BROWNS FERRY NUCLEAR PLANT AEC-ACRS CONCERNS - RESOLUTIONS Identification Section No. AEC-ACRS Concern Browns Ferry Resolutions 1.2.2 Effects of Fuel Failure on CSCS Topical Report (GE-APED-5608)

Performance Topical Report (NEDO-10208 August 1970) 1.2.3 Effects of Fuel Bundle Flow Blockage Topical Report (GE-APED-5608)

Topical Report (NEDO-10174 July 1970) 1.2.4 Verification of Fuel Damage Limit Topical Report (GE-APED-5608)

Dresden 2/3 - Amendment 14/15 Topical Report (NEDO-10173 May 1970) 1.2.6 Effects of Cladding Temperature and Topical Report (GE-APED-5608)

Materials on CSCS Performance Topical Report (GE-APED-5458)

Topical Report (NEDO-10179 June 1970) 1.2.5 Quality Assurance and Inspection of FSAR (Incorporated in Design -

the Reactor Primary System Section 4 and Appendix D) 1.2.7 Control Rod Block Monitor Design FSAR (Incorporated in Design -

Sections 1, 7 and Appendix G)

Dresden 2/3 - Amendments 17/18 and 19/20 Brunswick 1/2 - Supplement 5 1.2.8 Station Startup Program Topical Report (GE-APED-5698)

FSAR (Incorporated in Design -

Section 13) 1.2.9 Main Steam Line Isolation Valve FSAR (Incorporated in Design -

Testing Under Simulated Accident Section 4)

Conditions Topical Report (GE-APED-5750)

Topical Report (GE-NEDO-10045)

Topical Report (GE-APED-5608) 1.2.10 Performance Testing of the Plant FSAR (Incorporated in Design -

Standby Diesel Generator System Section 8)

General Motors Report 1.2.11 Formulation of an In-Service FSAR (Incorporated in Design -

Inspection Program Section 4)

Technical Specifications -

Sections 3/4) 1.2.12 Diversification of CSCS Initiation FSAR (Incorporated in Design -

Signals Sections 6 and 7) 1.2.13 Control Systems for Emergency Power FSAR (Incorporated in Design -

Section 8) 1.2.14 Misorientation of Fuel Assemblies FSAR (Incorporated in Design -

Section 3) 1.2.15 Concern of Dr. Stephen H. Hanauer - FSAR (Incorporated in Design -

emergency power and Core Standby Sections 6, 8 and 14, Appendix I)

Cooling Systems

BFN-16 TABLE 1.11-2 (Cont'd)

(Sheet 2)

BROWNS FERRY NUCLEAR PLANT AEC-ACRS CONCERNS - RESOLUTIONS Identification Section No. AEC-ACRS Concern Browns Ferry Resolutions 1.2.16 Fuel Clad Disintegration Limitations FSAR (Incorporated in Design -

Section 6)

Topical Report (GE-APED-5608)

Dresden 2/3 - Amendment 7/8 1.2.17 General concerns with regard to reactors FSAR (Incorporated in Design -

of high power density and all large Appendix I) water-cooled power reactors

BFN-16 TABLE 1.11-3 (Sheet 1)

BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 2 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.2.1 Units 1 and 2 ACRS concerns FSAR Appendix I, subsection 1.2 1.3.2.2 Core Spray Cooling Effectiveness FSAR (Incorporated in Design Section 6)

Topical Report (GE-APED-5458) 1.3.2.3a Reliability of CSCS Injection Valves FSAR (Incorporated in Design -

Section 4, 6 and 7) 1.3.2.3b Diversification of the CSCS Initiation FSAR (Incorporated in Design -

Signals Sections 6 and 7) 1.3.2.3c Sequencing of CSCS FSAR (Incorporated in Design -

Sections 6 and 8) 1.3.2.3d Core Spray Cooling Effectiveness (See 1.3.2.2 above) 1.3.2.3e Performance Testing of the Standby FSAR (Incorporated in Design -

Diesel Generator System Section 8)

General Motors Report 1.3.2.3f Fuel Failure Modes FSAR (Incorporated in Design -

Sections 6, 7, 14 and Appendix A)

Topical Report (GE-APED-5652)

Topical Report (GE-APED-5756)

Topical Report (GE-APED-5448)

Topical Report (GE-APED-5528)

Topical Report (GE-APED-5455)

Topical Report (GE-APED-5608)

Topical Report (GE-APED-5458)

Topical Report (NEDO 10208 August 1970)

Topical Report (NEDO 10174 July 1970)

Topical Report (NEDO 10173 May 1970)

Topical Report (NEDO 10179 June 1970)

Dresden 2/3 Amendment 7/8 1.3.2.4 Control Rod Block Monitor Design FSAR (Incorporated in Design -

Sections 1, 7 and Appendix G)

Dresden 2/3 Amendments 17/18 and 19/20 Brunswick 1/2 Supplement 5 1.3.2.5 Core Cooling FSAR (Incorporated in Design -

Section 6)

Topical Report (GE-APED-5458) 1.3.2.6 Control Rod Worth Minimizer Topical Report (GE-APED-5449)

FSAR (Incorporated in Design -

Section 7) 1.3.2.7 Control Rod Velocity Limiter Topical Report (GE-APED-5446)

FSAR (Incorporated in Design -

Section 3)

BFN-16 TABLE 1.11-3 (Cont'd)

(Sheet 2)

BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 2 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.2.8 In-Core Nuclear Instrumentation Topical Report (GE-APED-5456)

Topical Report (GE-APED-5706)

FSAR (Incorporated in Design -

Section 7) 1.3.2.9 Jet Pump Development Topical Report (GE-APED-5460) 1.3.2.10.1 Core Analytical Models FSAR (Incorporated in Design -

Sections 6, 7 and 14)

Topical Report (GE-APED-5652)

Topical Report (GE-APED-5756)

Topical Report (GE-APED-5448)

Topical Report (GE-APED-5528)

Topical Report (GE-APED-5455)

Dresden 2/3 Amendment 10/11 1.3.2.10.2 Fuel Failure Modes (See 1.3.2.3f above) 1.3.2.10.3 Electrical Load Control Using FSAR (Incorporated in Design -

Variable Speed Reactor Coolant Section 7)

Recirculation System Pumps Startup Test Results Oyster Creek No. 1 Nine Mile Point No. 1 Dresden No. 2 Millstone No. 1 1.3.2.10.4 Diversification of the CSCS Initiation (See 1.3.2.3b above)

Signals 1.3.2.10.5 Main Steam Line Isolation Valve FSAR (Incorporated in Design -

Testing Under Simulated Accident Section 4)

Conditions Topical Report (GE-APED-5750)

Topical Report (GE-NEDO-10045)

Topical Report (GE-APED-5608) 1.3.2.10.6 Performance Testing of the Station (See 1.3.2.3e above)

Standby Diesel Generator System

BFN-16 TABLE 1.11-4 BROWNS FERRY NUCLEAR PLANT UNIT 3 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.3.1 Performance Testing of the Standby (See 1.3.2.3e above, Table 1.11-3)

Diesel Generator System 1.3.3.2 Reactor Building Basement Corner FSAR (Incorporated in Design -

Room Flooding Section 4) 1.3.3.3 Automatic Pressure Relief System FSAR (Incorporated in Design -

Initiation Interlock Sections 6 and 7) 1.3.3.4 Criterion 35 Intent FSAR (Incorporated in Design -

Section 4 Appendix A) 1.3.3.5 RPV-Stub Tube Design FSAR (Incorporated in Design -

Section 4)

Topical Report (GE-APED-5703) 1.3.3.6 Requirements for Further Technical (See Table 1.11-3)

Information from Unit 1 and 2 C.P.

1.3.3.7 CSCS Thermal Effects on The Reactor Topical Report (GE-NEDO-10029)

Vessel and Internals FSAR (Incorporated in Design -

Sections 3 and 4) 1.3.3.8 Depressurization Performance of HPCIS FSAR (Incorporated in Design -

Section 6)

Topical Report (GE-APED-5608)

Topical Report (GE-APED-5447) 1.3.3.9 Electrical Equipment Inside FSAR (Incorporated in Design -

Containment Section 7) 1.3.3.10 Primary System Leakage Detection FSAR (Incorporated in Design -

Sections 4 and 10)

BFN-16 TABLE 1.11-5 (Sheet 1)

AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions_

1.4.2 Ring Header Leakage Design FSAR (Incorporated in Design -

Sections 4, 5 and 6) 1.4.3 CSCS Thermal Effects on The Reactor Topical Report (GE-NEDO-10029)

Vessel and Internals FSAR (Incorporated in Design -

Sections 3 and 4) 1.4.4 Effects of Blowdown Forces on Reactor FSAR (Incorporated in Design -

Primary System Components Sections 3, 4 and Appendix C) 1.4.5 Separation of Control and Protection FSAR (Incorporated in Design -

System Functions Sections 6, 7 and Appendix A) 1.4.6 Instrumentation For Prompt Detection FSAR (Incorporated in Design -

of Gross Fuel Failures Section 7)

Brunswick 1/2 - Supplements 3 and 4 1.4.7 Design of Piping Systems to Withstand FSAR (Incorporated in Design -

Earthquake Forces Section 12 and Appendix C)

Dresden 2/3 - Amendment 13/14 1.4.8 LPCIS - Logic Control System Design FSAR (Incorporated in Design -

Section 6) 1.4.9 Reevaluation of Main Steam Line Break Topical Report (GE-APED-5608)

Accident Topical Report (NEDO-10045)

FSAR (Incorporated in Design -

Section 14) 1.4.10 Depressurization Performance of HPCIS FSAR (Incorporated in Design -

Section 6)

Topical Report (GE-APED-5608)

Topical Report (GE-APED-5447) 1.4.11 AEC General Design Criteria No. 35 FSAR (Incorporated in Design -

Intent Design Conformance Section 4) 1.4.12 Automatic Pressure Relief System FSAR (Incorporated in Design -

Initiation Interlock Sections 6 and 7) 1.4.13 Scram Reliability Study FSAR (Incorporated in Design -

Sections 3 and 7)

Study Results (To be Available Early 1970)

Brunswick 1/2, Supplement 6 1.4.14 Design Basis of Engineered Safety Topical Report (GE-APED-5756)

Features FSAR (Examined Capability of Design -

Section 14)

BFN-16 TABLE 1.11-5 (Cont')

(Sheet 2)

AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions 1.4.15 Hydrogen Generation Study Topical Report (GE-APED-5454)

Topical Report (GE-APED-5654)

Brunswick 1/2, Supplement 4 Study Results (To be Available Middle 1970) 1.4.16 Primary Containment Inerting Topical Report (GE-APED-5454)

Topical Report (GE-APED-5654)

FSAR (Incorporate in Design -

Sections 5 and 6)

Dresden 2 - ACRS Letter, 9/10/69) 1.4.17 Seismic Design and Analysis Models FSAR (Re-Confirmation of Design -

Section 12 and Appendix C)

Dresden 2 - Re-Confirmation Information (submitted October 1969) 1.4.18 Automatic Pressure Relief System FSAR (Incorporated in Design -

Single Component Failure Capability Sections 6 and 8)

Manual Operation 1.4.19 Matters of Current Regulatory Staff Applicant Discussion (a) Standby Gas Treatment System FSAR (Incorporated in Design -

Electrical and Physical Separation Sections 5, 7, and 8)

(b) Official, Issued Technical Specifi- Proposed Technical Specifications cations - License Appendix A Appendix B 1.4.20 Flow Reference Scram FSAR (Incorporated in Design -

Section 7) 1.4.21 Future Items of Consideration for Incorporation ....

(a) Radiolytic Decomposition of Topical Report (GE-APED-5454)

Cooling Water Topical Report (GE-APED-5654)

Brunswick 1/2, Supplement 4 Study Results (To be Available Middle 1970)

(b) Development of Instrumentation FSAR (Justified Design - Sections Vibration and Loose Parts 3, 4 and Appendix C)

Detection (c) Consequences of Water Contamination FSAR (Incorporated in Design -

Structural Material - LOCA Section 14) 1.4.22 Diesel Generator Synchronization FSAR (Incorporated in Design -

Considerations Sections 6, 7, and 8)

BFN-16 TABLE 1.11-5 (Cont'd)

(Sheet 3)

AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions 1.4.23 Development of Instrumentation FSAR (Justified Design - Section 4)

Primary Containment Leakage Technication Specification -

Detection System Increased Appendix B Sections 3 and 4)

Sensitivity Studies 1.4.24 Development of Instrumentation - FSAR (Justified Design - Sections 3, Vibration and Loose Parts 4, and Appendix C)

Detection Studies 1.4.25 CSCS - Leakage Detection, Protection, FSAR (Justified in Design - Sections and Isolation Capability 4, 10 and Appendix A)

Brunswick 1/2 - Supplement 4, C/R 6.4 1.4.26 Main Steam Lines - Standards For FSAR paragraph 1.4.2.6 Fabrication, Q/C and Inspection

BFN-16 TABLE 1.11-6 AEC ACRS CONCERNS ON OTHER DOCKETS - CAPABILITY FOR RESOLUTION Identification Section No. AEC-Staff Concern Capability for Resolution 1.5.2 Tornado and Missile Protection FSAR (Incorporated in Design -

GE BWR-Spent Fuel Storage Pool Sections 2, 10, and 12)

Topical Report (GE-APED-5696) 1.5.3 BWR System Stability Analysis FSAR (Incorporated in Design -

Section 7)

Topical Report (GE-APED-5652)

Topical Report (GE-APED-5640)

Peach Bottom 2/3 - Amendment 2

BFN-28 1.12 GENERAL CONCLUSIONS Based on the favorable plant site characteristics, on the design of the plant herein analyzed, on the criteria, principles, and design requirements pertinent to safety, on the calculated potential consequences of routine and accidental release of radioactive material to the environs, on the results of research and development programs, and on the technical competence of the applicant and his contractors, there is reasonable assurance that the Browns Ferry Nuclear Plant can be operated without endangering the health and safety of the public.

NRC has accepted these conclusions and has issued operating licenses for all three units of the Browns Ferry Nuclear Plant. The dates of issuance for each unit's operating license are as follows:

UNIT ONE - June 26, 1973 UNIT TWO - June 28, 1974 UNIT THREE - July 2, 1976 A subsequent license amendment for a five percent core thermal power uprate, from 3293 MWt to 3458 MWt, was issued for Units 2 and 3 on September 8, 1998, and for Unit 1 on March 6, 2007. On August 14, 2017, license amendments were issued for all three units for a core thermal power uprate from 3458 to 3952 MWt.

Operating data for each unit will be provided to the NRC. This operating data shall comply with the operating data (for each calendar month) as described in Generic Letter 97-02 Revised Contents of the Monthly Operating Report. This data will be provided by the last day of the month following the end of each calendar quarter.

This operating data may be provided by the use of an Industry Database (e.g., the Industrys Consolidated Data Entry Program (CDE) or other reports to prevent any gaps in the monthly operating statistics and shutdown experience provided to the NRC.

1.12-1

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