Indian Point 3 Nuclear Power Plant Systems Interaction · 2012. 12. 4. · INDIAN POINT 3 NUCLEAR...

Indian Point 3 Nuclear Power Plant

Systems Interaction Study- Report Volume 1

COPY NUMBER _ __

8312060173 831130 PDR ADOCK 05000286 P PDR

NewYorkPower Authority

NEW YORK lOWER AUTHORITY

INDIAN POINT 3 NUCLEAR POWER PLANT

SYSTEMS INTERACTION STUDY

Prepared for the New York Power Authority

By

Thasco Services Incorporated

Project Identification No. PASN IP-3 SIS

INDIAN POINT 3 NUCLEAR POWER PLANT SYSTEMS INTERACTION STUDY

.PURCHASER:

OWNER:

OPERATING COMPANY:

PROJECT:

UNIT NO.

LOCATION

NEW YORK POWER AUTHORITY



INDIAN POINT NO. 3 NUCLEAR POWER PLANT

3 NORMAL MW: 965

r:. BUCHANAN, NEW YORK

Prepared under the supervision of- Project ManagerR Ciorgio

Report Status

Original RI

Date Verified By

June 30, 1983 W A Griswold November 10, 1983 T P Ruggier~p

Pages Affected

All

2-2, 3-1 thru -3 4-1, 5-1 thru -5, 5-8, 5-10, 5-12, 6-1, 6-3 thru -6, 6-9 thru -11, 7-2, 7-3, Table 6-1, 5-3, Sys Descp 18-1, 19-1 & Analysis Result 1-1

r

NEW YORK POWER AUTHORITY INDIAN POINT 3 NUCLEAR POWER PLANT


CONTENTS

VOLUME I

METHODOLOGY

TITLE

INTRODUCTION

General Content Work Responsibilities and Locations

BACKGROUND

OBJECTIVE AND SCOPE

Premise Objectives Scope

STUDY TEAM ORGANIZATION

NYPA Team Organization Ebasco Team Organization

METHODOLOGY

Purpose Initial Activities Identification of Primary Systems Identification of Secondary Systems

(Auxiliary Systems) Interconnected Systems Interactions Definition General Process Scope of System Failures Excluded Failures Included Failures

SECT ION

Chapter 1 1.0

PAGE

2-1Chapter 2 2.0

Chapter 3 3.0

3.1 3.2 3.3

Chapter 4 4.0

Chapter 5 5.0

5.1 5.2 5.2.1 5.2.2

5.3 5.3.1 5.3.2 5.3.3 5.3.4 5.3.4.1 5.3.4.2

3-1 3-2 3-2

4-1 4-2

5-1 5-1 5-2 5-2

5-3 5-3 5-3 5-4 5-4 5-4 5-5



CONTENTS (Cont'd)

VOLUME I

METHODOLOGY

SECTION TITLE PAGE

Chapter 5 5.0 METHODOLOGY (Cont'd)

5.4 Nonconnected Systems Interactions 5-5

5.4.1 Definitions 5-5 5.4.2 Failure Events 5-5 5.4.3 Interaction Identification 5-6 5.4.4 Discovery of Nonconnected Systems Interactions 5-8 5.4.4.1 Seismically Induced Spatially Coupled Interactions 5-8 5.4.4.2 System Walkdown-Search Phase 5-9 5.4.4.3 Evaluation Phase 5-10 5.4.4.4 Area Specific Interactions 5-11

5.5 Induced Operator Error 5-11

5.5.1 General 5-11 5.5.2 Identification 5-12 5.5.3 Evaluation 5-13

Chapter 6 6.0 EVALUATION CRITERIA 6-1

6.1 Interconnected Systems 6-1 6.2 Nonconnected Systems 6-2 6.2.1 Event/Source Credibility Evaluation 6-2 6.2.2 Source Evaluation Criteria 6-2 6.2.3 Comparison with NRC Guidelines 6-6 6.2.4 Interaction Effects Evaluation Criteria 6-7 6.2.4.1 Categories of Interaction Effects or 6-9 6.2.4.2 Methods of Evaluation 6-9 6.2.4.3 Evaluation of Direct Interaction Effects or 6-10 6.2.4.4 Evaluation of Secondary Effects

Cascading Influences 6-11 6.2.5 Modification Criteria 6-12



CONTENTS (Cont'd)

VOLUME I

METHODOLOGY

TITLE

QUALITY ASSURANCE

General Analysis Criteria and Basis Internal Analysis Reviews Analysis Verification Data Package Records Types of Documents Storage and Security of Records Audits Audit Results, Evaluation and Report Reaudi t Audit Program Review

REFERENCE DOCUMENTS

LOGIC DIAGRAMS

FUNCTIONAL TABLES

AUXILIARY DIAGRAMS

SYSTEM DESCRIPTIONS

INTERACTION SUMMARY

SHEETS 1 to 4

SHEETS 1 to 4

SHEETS 1 to 23

SYSTEMS 1 to 23

Systems Interactions Induced by Externally Generated Missiles.

2.0 Systems Interactions & Pipe Whip.

Induced by Internally Generated Missiles

Systems Interactions Induced by Severe Environment (other than

flooding) Resulting from Natural Phenomena.

Systems Interactions Induced by Severe Environment within Class I Structures.

PAGESECTION

Chapter 7 7.0

7.1 7.2 7.3 7.4 7.5 7.6 7.7 7.8 7.9 7.10 7.11 7.12

7-1

7-1 7-1 7-1 7-2 7-2 7-2 7-3 7-3 7-3 7-3 7-3 7-3

Chapter 8 8.0



CONTENTS (Cont'd)

VOLUME I

INTERACTION SUMMARY

SECTION TITLE PAGE

5.0 Systems Interactions Induced by the Effects of Flooding Due to

Natural Phenomena.

6.0 Systems Interactions Induced by the Effects of Internally

Generated Flooding.

7.0 Central Control Room Review (including HVAC/instrument air).

8.0 Induced Operator Error Interaction Summary.

9.0 Systems Interactions Commonality.

10.0 Auxiliary Feedwater System Interactions.

Appendix 1

Appendix 2

Volume 2

1-Rod Control System

2-Safety Injection System

Volume 3

2-Safety Injection System (Cont'd)

3-Chemical and Volume Control System

Volume 4

3-Chemical and Volume Control System (Cont'd)

Volume 5




CONTENTS (Cont'd)

VOLUME I

INTERACTION SUMMARY

Volume 6


Volume 7

3-Chemical and Volume Control System (Cont'd) 4-Residual Heat Removal System

Volume 8

4-Residual Heat Removal System (Cont'd)

5-Reactor Coolant System

Volume 9

5-Reactor Coolant System (Cont'd)

6-Containment Spray

Volume 10

8-Service Water System

Volume 11

8-Service Water System (Cont'd)

Volume 12

8-Service Water System (Cont'd)

9-Component Cooling Water System

Volume 13

9-Component Cooling Water System (Cont'd)

Volume 14

9-Component Cooling Water System (Cont'd)



CONTENTS (Cont'd)

VOLUME I

INTERACTION SUMMARY

Volume 15

9-Component Cooling Water System (Cont'd) 10-Nitrogen Backup System

Volume 16

11-Diesel Generators System

Volume 17

12-DC System 13-Containment Isolation System

Volume 18

14-Main Steam System 15-Main Steam & Feedwater Isolation System

Volume 19

16-Hydrogen Recombiners System 17-Reactor Protection System

Volume 20

17-Reactor Protection System (Cont'd) 18-Electrical Trays

Volume 21

18-Electrical Trays (Cont'd) 19-HVAC System 20-Nitrogen Actuation System 21-Containment Recirculation System

Volume 22

22-Electrical Distribution System 23-Auxiliary Feedwater System

Volume 23

23-Auxiliary Feedwater System (Cont'd)



CONTENTS (Cont'd)

VOLUME I

INTERACTION SUMMARY

Volume 24

Evaluations of Interaction Credibility (EIC)

Induced Operator Error Analyses (IOEA)

Span Evaluations

Volume 25

Photographs

vii


SYSTEM'INTERACTION STUDY

CHAPTER 1

1.0 INTRODUCTION

1.1 General Content

This report documents a study conducted for Indian Point 3 nuclear power plant in an effort to identify and evaluate systems interactions to enhance the level of safety associated with the continued operation of the plant. It was prepared for the New York Power Authority by Ebasco Services Incorporated and consists of 25 volumes. Described in this report are the methodology used to identify and evaluate systems interactions and the overall results of the study (Volume 1), and the application of these methods and criteria to the 23 systems determined to be important to plant safety (volumes 2-24). Volume 25 contains photographs taken in support of the study.

1.2 Work Responsibilities and Locations

The majority of the work was conducted at the Indian Point 3 nuclear power plant site and in Ebasco's Lyndhurst, NJ office with special support from the Applied Physics department in Ebasco's World Trade Center office. The work necessitated frequent trips between the Lyndhurst office and the site and required special training for Ebasco employees to allow them entry into the radiological areas of the plant. Communications between responsible authorities of the New York Power Authority and members of the Ebasco study team during the course of the study were limited to conceptual presentations, contract negotiations, quality assurance audits, progress reports, and some typical plant walk down inspections to ensure that the results were representative of an independent third party effort.


SYSTEM INTERACTION STUDY

CHAPT~ER 2

2.0 BACKGROUND

From a historical point of view it is noted that the Nuclear Regulatory Commission's (formerly AEC) General Design Criteria (GDC) for nuclear power plants and the Indian Point 3 Nuclear Power Plant (IP-3) design were developed concurrently during the late 1960's and early 1970's. The GDC are now incorporated in the NRC's regulations as Appendix A to 10CFR Part 50.

While Criteria 2, 3 and 4 of the GDC require that structures, systems and components important to safety be able to accommodate natural phenomena such as earthquakes, the effects of fires, and other environmental effects without loss of capability to perform their intended safety functions, the systems interaction issue was not specifically raised as a potential concern until the Advisory Committee on Reactor Safeguards (ACRS) formally raised the question in 1974.

In 1977 systems interaction formally appeared as NRC Generic Task Action Plan A-17. The first phase of this NRC plan has just recently been completed with the publication of the Sandia Report "Final Report PHASE I Systems Interaction Methodology Applications Program". TMI-2 events have to a large extent been factored into this systems interaction plan. Additional details on the regulatory developments on systems interaction are found in:

a) Generic Task Action Plan A-17 (NUREG 0606 Rev. 2) Systems Interaction In Nuclear Power Plants.

b) N UREG 0510

Identification of Unresolved Safety Issues Relating to Nuclear Power Plants.

c) NRC Information Notice 79-22

Potential Interactions Between Non-Safety Related Control Systems and Safety Systems.

d) NUREG 0585

TMI-2 Lessons Learned Task Force Final Report Recommendation 9 - Review of Safety Classifications and Qualifications.

2-1

2.0 BACKGROUND (Cont'd) R1

e) NUREG 0660

Action Plans for Implementing the Recommendations of the

President's Commission and Other Studies of TMI-2 Accident.

TASK II.C.I - Systems Engineering, Reliability Engineering and

Risk Assessment.

The NRC in the spring of 1981 distributed three reports prepared by

independent laboratories which address the different methodologies being utilized by various utility groups, consultants, etc. They are, NUREG/CR-1859, UCRL-53016, Systems Interaction: "State-of-the-Art Review and

Methods Evaluation", prepared by Lawrence Livermore Laboratory for NRC-ONRR, November, 1980; NUREG/CR-1901, BNL-NUREG-51333, "Review and Evaluation of System Interactions Methods", prepared by Brookhaven National Laboratory for

NRC-ONRR, January, 1981, NUREG/CR-BMI-2055, R-2 "Report on Review of Systems Interaction Methodologies", prepared by Battelle Columbus Laboratories for NRC-ONRR, January, 1981.

Discussions between the industry via the Atomic Industrial Forum (AIF) and NRC continued throughout 1981 without specific guidelines being generated for use in evaluating systems interactions. With the formation of the Systems

Interaction Section of the Reliability and Risk Assessment Branch (RRAB) the first evidence of specific guidance was formulated. Over the past and specifically during that period of time while the preliminary issue of this report was being reviewed by the SI Section of RRAB, there has been an effort to provide a reasonable boundary for the overall SI topic and thereby focus in on those key issues which would make an SI study both meaningful and effective.

It is important to recognize that the systems interaction study is an attempt to reevaluate in a systematic fashion those potential events whose direct effect could reduce the plant safety margin. The criteria employed are considered new only to the extent that effects of nonsafety systems on safety systems are considered in a more thorough fashion. Currently, neither the NRC

nor any industry body (such as AIF or ANS) have published any accepted methodology for performing systems interaction analysis.

During review of the preliminary issue of this report by the SI Section of RRAB, it was agreed explicitly that the threshold for identification of adverse SI's would be a nonsafety system or component failure that leads to the defeat of one train of a safety system or engineered safety feature, even if the remaining train(s) of the affected safety system or ESF could perform the intended safety function. This is the most stringent application of the

single failure criterion currently used in the licensing review process.

2-2



CHAPTER 3

3.0 OBJECTIVES AND SCOPE

3.1 Premise

In order to derive a working definition of systems interaction, it is necessary to consider a number of associated concepts. In the design of a nuclear power plant, provisions are made to make the release of radioactivity to the environment an extremely unlikely event by providing independent ways in which a safety function can be performed. These provisions are expressed in terms of redundancy and diversity so that multiple independent system failures would not necessarily result in a safety function failure. Systems which support safety functions may be designed to interact with each other. These interactions are intentional. An "interaction" of concern results when the conditions in one system unintentionally affect the ability of another system to perform its safety function. Therefore, system interactions are those events that affect the safety of the plant by one system acting upon one or more other systems in a manner not intended by design. For the purposes of this study the emphasis is on nonsafety to safety types of interactions. A systems interaction study involves (1) the systematic search for hidden or inadequately analyzed couplings that link safety and nonsafety systems in the reactor plant, and (2) the evaluation of the effects of nonsafety system failures propogated into safety systems by such couplings.

To accurately identify the SI relationship to plant safety, four safety functions are defined: (1) Achieve and maintain reactor subcriticality (2) Remove decay heat (3) Maintain containment integrity (4) Maintain reactor coolant pressure boundary.* These basic safety goals are assumed to be inviolable and solely necessary to the safe operation of the plant during normal and emergency conditions. SI's are identified only for those systems essential to the four basic safety goals and are evaluated with respect to them.

Systems interactions generally fall into 2 categories, process connected and nonconnected. Process connected (also called interconnected) interactions link nonsafety to safety systems by design, through piping, instrument tubing, cables and control wiring. Induced operator error interactions are also considered process coupled, where the operator serves as the link instead of a fluid or electrical current. Induced operator error is discussed in more detail in section 5. Nonconnected interactions depend on external effects (outside of the safety systems' physical boundaries) to establish the link, such as collision, flooding, missiles, tornado, or steam line rupture.

*Where used in this study the terms safety function and safety goal are synonymous.

3.0 OBJECTIVES AND SCOPE (Cont'd) Ri

3.2 Objectives

The objectives of this study were (1) to develop the methodology and evaluation criteria, based upon conservative engineering judgement, used to identify and evaluate systems interactions and (2) to apply these criteria to a systems interaction review of 23 systems that have been identified as essential to the four safety goals. The results of the study were to provide a list of systems interactions whose effects on the safety goals were potentially degradable and to identify the manner in which the SI's affected them.

3.3 Scope

For the "process" connected portion of this study, a "dependency-analysis" technique was used as the primary means of identifying systems interactions. Their evaluations were based on systems operation required to support the four safety functions.

Induced operator error interaction analyses were performed on a selected group of indications, which were considered most likely to cause induced operator error. Like fluid or electrically coupled process interactions, FMEA's were used for evaluations.

For the "nonconnected system"' portion of the study, Ebasco investigated the possibility of adverse interactions transported via spatial or physical proximity considerations during design basis events such as earthquake, tornado, fire, high energy pipe rupture, internal or external flooding and internally or externally generated missiles. These latter events were investigated for interactions via plant walkdowns and by a review of reports previously prepared on these subjects. Evaluations were prepared using FMEA techniques, again based on system design and operation with respect to satisfying the four basic safety goals. Flooding, tornado, missiles, and environmental design base events are presented separately from seismically induced collision interactions in this report.

The consequential repair of system interactions identified in the study and evaluated as unacceptable fell outside the scope of this study. However, some discussion is contained in the Interaction Summary on the available methods of resolving the outstanding interactions.

3.0 OBJECTIVES AND SCOPE (Cont'd)

3.3 Scope (Cont'd)

A suggestion was made by the NRC SI staff to investigate the use of the IP simulator to uncover and treat system interaction dependencies of the "first-order" type. It was believed that a training simulator could accurately model at least direct interconnections between safety and nonsafety front line systems and their support systems and that it may be possible to do a comprehensive and systematic analysis of their failure effects. As a preliminary step, the SI staff was invited to participate in initial trials on September 23-24, 1981 at the Indian Point simulator facility. Subsequent to those trials an arrangement was made to further investigate the use of the IP simulator for specific malfunctions modeled into the simulator. This activity was accomplished on October 29, 1981. In general the results obtained during the initial trials and the malfunction tests confirmed that the use of the simulator does not effectively uncover SI's between safety and nonsafety systems. A complete summary report of the activities of the initial trials and specific malfunction tests that followed are presented in Appendix 1.

In addition, the NRC staff emphasized that the consideration of operating experience was an important element in the systems interaction analysis and should be treated in the IP-3 SI study. To this end, the study was expanded to include a review of Significant Occurrence Reports. The results of the review are located in Appendix 2.

Not considered within the scope of this study were fuel handling and storage systems, refueling systems, fire detection and protection systems, security systems, operator training, plant administration, radiation detection and control, plant chemistry, flow induced vibration, lubrication, operator induced failure, sabotage and equipment unavailability due to maintenance.

3.0



CHAPTER 4

4.0 STUDY TEAM ORGANIZATION

4.1 NYPA Team Organization

The Power Authority retained Ebasco Services Incorporated for the performance

of a systems interactions study for the Indian Point 3 Nuclear Power Plant.

The Nuclear Support PWR (NS-PWR) Division of the Nuclear Generation (NG)

Department in the White Plains office (WPO) had the primary responsibility for

accomplishing the Systems Interaction Study.

The Power Authority's Director of Project Engineering appointed a Task Force

Leader from his staff with the concurrence of the Vice President-Nuclear, Support PWR, and the Executive Vice President-NG. For the Systems Interaction Study the Power Authority's IP-3 Supervisory Engineer was the Task Force Leader.

The Task Force Leader was responsible for monitoring and controlling day to

day activities and for ensuring a sound multi-disciplinary review of work done by Ebasco. This was accomplished by choosing the following Power Authority personnel to be part of the review team:

Senior Operations & Maintenance Engineer-Nuclear Operations &

Maintenance NS-PWR, NG, WPO.

Nuclear Licensing Engineer-Nuclear Licensing NS-PWR, NG, WPO.

Senior Nuclear Engineer-Design & Analysis Division, Engineering Dept. WPO.

Senior Structural Engineer-Design & Analysis Division, Engineering

Dept. WPO.

Site Engineer-Technical Services Department, IP-3 Site

QA Engineer-QA Dept, WPO.

The plant walk through team was comprised of the Task Force Leader or Nuclear

Operations Engineer, and the Ebasco Systems Interaction multi-disciplinary team personnel.

Figure 4-1 indicates the structure of the Power Authority's reporting

relationships among the Systems Interaction Task Force.

STUDY TEAM4 ORGANIZATION (Cont'd)

4.2 Ebasco Team Organization

Within the Ebasco Organization, the Systems Interaction Study for the Indian Point 3 Nuclear Power Plant was administered by the Mechanical Engineering Department under the direction of a Project Manager. Personnel from various Ebasco engineering and design disciplines were assigned to the project and took functional directions from the Project Engineer.

Figure 4-1 indicates the reporting relationships among Ebasco engineering and design personnel who fulfilled the key roles in the Systems Interaction Study.

The responsibilities within the Ebasco Study Team are outlined as follows:

Project Manager

The role of the Ebasco Project Manager was to provide central leadership, planning, scheduling, budgeting and coordination of all services supplied by Ebasco to the Power Authority in addition to developing and administering controls to achieve schedule and budget compliance.

Systems Interaction Project Engineer

The role of the Project Engineer who reported to the Project Manager, was to provide advice, guidance, and support to the Project Team in performance of their function, manage the overall engineering effort, and integrate the multiple engineering activities.

His responsibilities included the following:

a) Writing the System Interaction Study description.

b) Coordinating the efforts of other Ebasco engineering and design disciplines who prepared the study, preparing implementation procedures, determining study inspection and evaluation criteria, and reviewing evaluations proposed by the Interaction Team.

C) Providing functional and technical direction to the Interaction Team.

d) Verifying the evaluations proposed by the Interaction Team.

e) Preparing interim reports and the final program report.

f) Communicating the activities of the Interaction Team and the results of the program to the Project Manager.

STUDY TEAM ORGANIZATION (Cont'd)

4.2 Ebasco Team Organization (Cont'd)

Systems Interaction Project Engineer (Cont'd)

The Project Engineer used in-house engineering and design disciplines to recommend technical decisions, provide administrative assistance, recommend resolutions, and provide analysis as needed. All engineering and design disciplines reported to the Project Engineer.

Project Quality Assurance Engineer (PQAE)

The PQAE was responsible for the implementation of the Quality Assurance Program for the System Interaction Study. He reported directly to the Chief Quality Assurance Engineer and had the authority and responsibility to identify quality related problems, to initiate or recommend solutions to control conformances until properly dispositioned and to verify implementation of approved dispositions. For a description of the quality assurance program, see chapter 7.

Interaction Team

The interaction team members were required to have considerable experience in their area of assignment and to have been involved with various aspects of plant and system operations on other nuclear projects. In addition, members had experience with projects similar to this study.

a) The Interaction Team comprised the following discipline Lead Engineers and their staffs:

(1) Mechanical Engineering (2) Instrumentation and Control Engineering (3) Electrical Engineering (4) Quality Assurance

b) The discipline Lead Engineers were selected from the staff of the engineering and design departments and were under the technical direction of the Discipline Chief Engineer.

c) Special assitance was obtained as required from other Ebasco departments, including Applied Physics & Licensing.

4-3

4.0

Figure 4-1 NEW YORK POWER AUTHORITY


NUCLEAR GENERATION DEPARTMENT

QUALITY ASSURANCE & RELIABILITY

SYSTEMS INTERACTION TASK FORCE TEAM

ENGINEERING DEPARTMENT

WSMT~ av~u mc GUALITY

mum.. .o...ua.~ Ga.i~~ STRUC1IAAI. ELEcTRICAL AUURM~I

miamna mostdagE pgouu,.



CHAPTER 5

5.0 METHODOLOGY

5.1 Purpose

ThiB section describes the methodology and documentation used in performing

the systems interaction study for the Indian Point 3 Nuclear Power Plant.

Through an evaluation of state-of-the-art methodology techniques, it was concluded that no single method could perform an adequate review to determineadverse systems interactions. However, all of the methods evaluated included the common process of "sifting-out" adverse systems interactions by 1) selecting specific systems for detailed evaluation, 2) identifying dependencies, and 3) evaluating the systems interactions through the determination of their relative importance to safety. It is this process which provided the foundation for performing a systems interaction study for the Indian Point 3 Nuclear Power Plant.

In this study the application of single failure criterion was used very conservatively. The reason for this decision was that the design of most safety related systems included the concept of "single failure". If a potentially unacceptable interaction, for example an interaction caused by the seismic induced failure of a nonseismic component, existed within a given safety related system, the subsequent application of the single failure criterion may render that system inoperable. In this study no credit was taken for redundancy of components within any given system. Credit for redundancy was taken at the system level, providing that the redundant system had no interactions caused by the same event (i.e. seismically induced).

5.2 Initial Activities

The primary task of this study was to determine if adverse systems interactions could occur and if so, whether or not a degradation of the reactor core or the release of unacceptable levels of radioactivity to the site environs could result. Those .conditions considered to be adverse and to have a significant potential for~ leading to core damage or to the release of unacceptable levels of radiation to the environs are, 1) failure to achieve or maintain reactor subcriticality, 2) failure to remove decay heat, 3) failure of the reactor coolant system pressure boundary and 4) failure of containment integrity. For the purposes of this study the aforementioned conditions are defined as Basic Safety Goals or Safety Functions. To relate these safety goals to the plant systems necessary to support them, the functional tables (FT), logic diagrams (LD), and auxiliary diagrams (AD) were developed. They were the primary documents used to identify the primary and support systems (including nonsafety systems) that were analyzed for interactions. The following two paragraphs explain how the safety goals/plant systems' relationships were established.

5.0 METHODOLOGY (Cont'd) RI

5.2 Initial Activities (Cont'd)

5.2.1 Identification of Primary Systems

The first step in "sifting-out" or searching for adverse systems interactions was the selection of primary (first line) systems, which was accomplished by developing a functional table and logic diagram for each of the four safety goals. The logic diagrams and functional tables were based upon the FSAR, system descriptions, mechanical flow diagrams, and electrical schematic, block and wiring diagrams. In addition, important information about system interfaces was obtained at the site by inspecting physical facilities and by meeting with plant personnel familiar with the design, operation and maintenance of the systems.

On the functional tables were listed each of the basic safety goals and the corresponding primary systems required to perform the functions. These systems were classified as primary or first line because they directly performed a specific function required to accomplish one or more of the basic safety goals. As the safety systems and their functions were identified on the functional tables, they were arranged separately on the logic diagrams (LD) to indicate alternative success paths leading to the required safety goals.

The logic diagrams and functional tables were both developed at the system level to allow the simple definition of those systems requiring study. After system identification, the study was performed at the component level through the use of matrices.

5.2.2 Identification of Secondary Systems (Auxiliary Systems)

After completion of the FT & LD for each of the four goals, each identified primary system was analyzed to determine the specific supporting systems necessary for it to perform its safety goals. These supporting systems provide for such requirements as motive power, cooling, and instrument air supply. The system descriptions, flow diagrams, one line diagrams, FSAR, etc were used to determine every sequence in which a safety system was required, thereby ensuring that all safety systems, and hence all support requirements were identified. Note that support of the primary systems was regarded as the support required to allow the primary systems to perform only their safety functions. For example, some valves in a primary system required instrument air to operate, but since operation of the valves was not required for a safety goal, the instrument air system was not identified as a secondary system. The systems required to provide support of the primary systems are defined as secondary or auxiliary systems. The auxiliary systems are displayed on the auxiliary diagrams (AD).

5.0 METHODOLOGY (Cont'd). Ri

5.2 Initial Activities (Cont'd)

5.2.2 Identification of Secondary Systems (Auxiliary Systems) (Cont'd)

An auxiliary diagram was prepared for each primary system to indicate the supporting requirements for the system. Also, an auxiliary diagram was prepared for each auxiliary system to indicate systems which in turn support the functions of the auxiliary system. In developing the auxiliary diagrams determinations were made on whether the support system designs were functionally redundant to their associated primary systems by reviewing design information about the plant. After completing the auxiliary diagrams the logic diagrams were reviewed in conjunction with them for the four postulated safety goals to ensure that all safety sequences in which the subject auxiliary safety system appeared were identified.

5.3 Interconnected Systems Interactions

5.3.1 Definition

Interconnected systems were defined as those mechanical and electrical complexes which were process coupled to one another physically via piping, instrumentation tubing or electrical wiring. Included in this definition was HVAC equipment, which although not physically connected, might be necessary to support the continuous safe operation of interconnected systems, e.g., an air handling unit which had been specifically designed to cool an essential safety related pump/motor set.

5.3.2 General

For interconnected mechanical or electrical complexes (process coupled systems), dependency analyses formed the basis for the systems interaction evaluations. The FT's, LD's and AD's described the combinations of systems which were searched for unacceptable process couples that could result in degradation of any of the four basic goals, (ie, reactor-subcriticality, decay heat removal, reactor coolant pressure boundary, or containment integrity) and possibly lead to core damage or the release of unacceptable levels of radioactivity to the environs. These documents were the vehicles by which the identification and evaluation of interconnected systems interactions were made.

In addition to the above, as a supplemental device for the evaluation of identified SI's, consideration was given to the use of fault trees on individual systems already available from the Z/IP-3 PS5 analysis. The SI's, however, did not require the use of fault trees for evaluation.


5.3 Interconnected Systems Interactions (Cont'd)

5.3.3 Process

For each primary and secondary system in this study an initial screening

process was employed to determine the instances in which safety/seismic systems interfaced with nonsafety/nonseismic systems at the component level. The basic documents for this effort were the flow diagram, system description, one-line diagram and instrument list. Each safety/seismic component was examined for piping, tubing, and electrical connections to nonsafety/nonseismic components and the results were placed on a list which represented a tabular mapping of the 23 systems important to the 4 basic safety goals. This list was called the interconnected matrix (IM). For each connected component that was indicated as nonsafety/nonseismic an initial evaluation was made by postulating specific failures and determining if the failures could be tolerated by the safety component. If the initial evaluation yielded potentially unacceptable results then a Failure Modes and Effects Analysis (FMEA) was performed.

5.3.4 Scope of System Failures

5.3.4.1 Excluded Failures

- Operator Induced Failures - Equipment Unavailability Due to Testing or Maintenance - Sabotage

The subject-of operator induced failures In system interaction studies was

excluded from consideration. The subject of the operator's influence on plant

safety was not, however, neglected since much of the available industry resources has been directed at improving operator training, developing advanced simulators, improving the human-machine interface through additional instrumentation and control room re-evaluation, and the development of improved operational procedures.

Equipment unavailability due to testing or maintenance was also excluded from

the scope of this study since technical specifications limit the time that safety related equipment may be removed from service.

Sabotage is not capable of being evaluated in an SI analysis. Therefore it

was not included in this study.


5.3 Interconnected Systems Interactions (Cont'd)

5.3.4 Scope of System Failures (Cont'd)

5.3.4.2 Included Failures

The scope of failures included those caused by adverse interactions of

interconnected systems and components that resulted as a direct consequence of off-normal events or actions for which the effected equipment had been prescribed to operate. The off-normal events or actions considered in this study were:

- Loss of Power (both motive and control power of the

electrical, hydraulic and pneumatic type)

- Seismic Induced Failures

- Loss of Cooling (including HVAC equipment)

5.4 Nonconnected Systems Interactions

5.4.1 Definition

Nonconnected systems were defined as all seismic and nonseismic mechanical,

electrical, instrumentation, and civil systems which were associated to each other by physical arrangement (spatial coupling) regardless of system function and independent of interconnected interactions.

5.4.2 Failure Events

As a basis for considering systems interactions of nonconnected systems for the events described herein, it was assumed that the structures, systems and components which were required to satisfy any of the four safety goals would not be prevented from carrying out their required safety functions because of nonconnected interactions caused by the event induced failures. Furthermore, spatial couplings were assumed to occur because of the event induced failures of components which were not qualified to withstand the events.

In this report "events" capable of causing nonconnected interactions included

the following:

1) Earthquake: Up to and including the safe shutdown

earthquake.

5-5

5.0 METHODOLOGY (Cont'd)

5.4 Nonconnected Systems Interactions (Cont'd)

5.4.2 Failure Events (Cont'd)

2) Pipe Failure: Pipe whip, jet impingement, jet reaction, severe environment (temperature, pressure, humidity only)

3) Physical Impact: Missiles generated internally and externally.

4) Flooding: Internal failures (pipe and tank failure) or external effects due to rain, snow, etc.

5) Tornado Depressurization/Overpressurization

6) Fire

7) Loss of Offsite Power

8) LOCA or Main Steam Line Break

5.4.3 Interaction Identification

Unlike connected systems interactions, nonconnected systems interactions did not lend themselves to study by one single method. This was because some types of interactions, for example those caused by the failure of nonseismic components, were system specific, while other types like external missiles, were area specific. For both types of interactions it was important to classify nonconnected (spatially coupled) systems, components and structures as either "sources" or "targets". Equipment which required protection from potential event induced interactions were designated as targets. Structures, systems and components of primary and auxiliary systems defined in section 5.2 of the study were considered targets.

The sources of detrimental interactions were defined as nonseismically supported or nonqualified structures, systems and components which, by their proximity or connection to targets, could interact through physical, mechanical, electrical or environmental means to compromise the integrity or operability of the targets. For events such as flooding, tornado or other natural phenomena the sources were identified as those specific events.

5.0 METHODOLOGY (Cont'd) /


5.4.3 Interaction Identification (Cont'd)

By definition, a nonconnected interaction occurred whenever the event induced behavior of a nonqualified component (source) affected a nearby primary or secondary system component (target), making it necessary to identify first the targets with their associated sources in order to identify potentially adverse interactions. Pairings of targets and sources were based on physical proximity which was assessed by a field walkdown team. If it could be established by inspection that no credible failure mode could be induced in the sources by events of credible severity then no interaction existed. If there was a credible failure mode and a physical proximity, then a potentially adverse nonconnected interaction was identified.

Electrical cables within the safety related target cable trays were not individually identified as targets nor were they individually evaluated. This was done to combine the cables within the trays as common targets.

In general, event induced interactions were identified as being in one or more of the following categories:

a) Contact (collision) with an unsupported source that would

compromise operability of the target.

b) Fluid leakage from one or more sources that would degrade the

environment of the target component and thereby prevent it from functioning properly.

c) Contact between a missile generated by a nonsafety related source and an initial target that would compromise the pressure boundary of a second connected target component.

d) Contact between a missile generated by a nonsafety related source and an initial target that would compromise operability of a second connected target component.

e) Failure of nonsafety related electrical equipment that would compromise the operability or integrity of a target equipment.

f) Secondary effects or cascading influences (mechanical, electrical or fluid) caused by any of the above interactions were considered to the extent that they resulted in the failure or misoperation of a specific safety system target(s). Cascading influences did not include multiple train collisions or secondary missiles, (missiles caused by missiles).

5.0 METHODOLOGY (Cont'd) R1


5.4.4 Discovery of Nonconnected Systems Interactions

5.4.4.1 Seismically Induced Spatially Coupled Interactions

Each safety related system component identified in the interconnected matrix was listed on a corresponding nonconnected matrix (NCM) as a target component. This document was utilized to record the results of the plant walkdowns and later identify spatial interactions. After listing the potential targets a multistep program was used to study the seismically induced interactions. The steps were as follows:

- Search for interactions by means of an in-plant walkdown inspection. (Matrix).

- Preliminary evaluation of the interactions to determine first level credibility. (Evaluation Sheet).

- Failure Modes and Effects Analyses (FMEA) of the target components. (FMEA).

- Final evaluation of the source/target interactions using specific source failure modes and engineering judgement to determine credibility of the interactions. (EIC).

- Analyses of the sources to determine if they could withstand seismic loadings without impacting the targets. (Source Analysis).

5.4.4.2 System Walkdown-Search Phase

The first step in the discovery of seismically induced spatial interactions was a system walkdown inspection performed by an interdisciplinary team of experienced engineers using the previously prepared nonconnected matrix. During this inspection, a stringent criterion, described in chapter 6, was used to postulate source/target proximity and initial source failure credibility.

The target components listed on the nonconnected matrix were located by the interaction team and inspections were made of the areas around the targets for potential sources using the predetermined criterion. Each source component found in the vicinity of the item was examined for possible failure by any or all of the specific mechanisms listed in the criteria acting singly and in combination. If failure was postulated a determination was made during the inspection or afterwards by further analyses on the credibility of the interaction with the target equipment.

Once the field system evaluation had been completed the following information was documented for each interaction.



5.4.4.2 System Walkdown-Search Phase (Cont'd)

a) Location of each potential interaction, using background drawings and photograph numbers.

b) Targets and sources involved in the potential interaction identified on the nonconnected matrix form and documented on the nonconnected evaluation sheet. An interaction number which included the specific target and source under consideration, the target responsible discipline and the target system was included on the evaluation sheet.

c) The specific criteria used for the evaluation (which included

the type of interaction) documented on the evaluation sheet.

d) A photographic record of each identified interaction on the matrix. The photograph number was a unique number which identified the source in the photograph volume of this report. On the photograph itself, a small arrow indicated the general location of the sources.

e) Recommendation of the interaction team. This took the form of one of the following:

(1) Based on the first level criteria the interaction would

or could not occur.

(2) Recommend further evaluation. (Potentially Unacceptable)

5.4.4.3 Evaluation Phase

After the preliminary evaluation a number of interactions were classified as potentially unacceptable; that is to say that a more detailed evaluation was required. Potentially unacceptable interactions were further evaluated to determine if failure of a particular target component would violate one or more of the four basic safety goals and if, based on specific failure modes of the source, the interaction was credible.

The tools for performing these analyses were:

- Failure Modes and Effects Analysis (FMEA) of the target component.

- Interaction Credibility Evaluation (EIC) of the source/target

interaction.

- Span Evaluation of the source pipe.

5-9



5.4.4.3 Evaluation Phase (Cont'd)

For each potentially unacceptable interaction either an FMEA or EIC or both

was performed. The FMEA was performed by postulating and analyzing those target failures (with respect to the four safety goals) which could be caused by interaction with the source component. The EIC was performed by postulating specific individual failure modes of the source component and making a decision as to whether interaction with the target(s) was credible.

Interactions that were initially determined to be potentially unacceptable by FMEA were evaluated using the EIC and interactions that were initially determined to be potentially unacceptable by EIC were evaluated using the FMEA. An interaction was acceptable if it passed either FMEA or EIC evaluations. If an interaction was determined to be potentially unacceptable by both FMEA and EIC and if the source component was piping or conduit, then an evaluation of the support span was performed. Interactions still classified as potentially unacceptable by all three methods were marked as such on the system tabulation sheets for future disposition. Interactions which remained potentially unacceptable were also listed by common source (i.e. for each source, a list of unacceptably interacting targets).

5.4.4.4 Area Specific Interactions

Area specific interactions included:

- Interactions caused by tornado or external environments

- Interactions caused by flooding both internal and external.

- Interactions caused by severe internal environment.

- Interatctions caused by missles both internal and external.

The study of these items required a review of previous licensing information and studies. Note that the system interaction study did not repeat previous evaluations if the criteria were similar to or could be extended to the criteria of this study. If information was not available a detailed study was carried out. Such was the case with internal flooding. The study of area specific interactions also included consideration of intercompartmental interaction effects, such as flooding, steam, and cable and piping pass through. Information on these subjects gathered from plant walkdowns was utilized wherever possible.

5.5 Induced Operator Error

5.5.1 General

Although interactions usually connected nonsafety systems to safety systems by physical means (spacial-collision, process connections, or environmental) the

5-10


5.5 Induced Operator Error (Cont'd)

5.5.1 General (Cont'd)

operator could also serve as that connection. That type of SI was termed "induced operator error" (IOE) and involved a set of circumstances in which (1) a nonsafety system failure caused loss (particularly massive loss) of normal control instumentation display, and (2) the operator was assumed to act correctly (procedurally speaking) on the basis of incorrect readings produced by the initiating failure. Induced operator error interactions were evaluated as were process connected or spatially connected interactions. If the interactions affected any of the four basic safety goals in an adverse manner, they were considered potentially unacceptable. It should be noted that induced operator error was component-initiated and not operator-initiated, so it should not be confused with "operator induced error". Induced operator error interactions were identified and evaluated as described below, following the block diagram in Table 5.1.

5.5.2 Identification

To identify possible induced operator error interactions, it was necessary only to identify the indications which could induce the operator to err. Three basic assumptions were made for those indications:

(1) Failure of the indications must have been caused by failure of the nonsafety instruments or components which provided the operating signals. Nonsafety indications were distinguished from all other indications by defining them as those associated with components attached to 'piping outside of the seismic Category I boundaries, as shown on the flow diagrams. Exceptions included the reactor coolant pumps instrumentation which did not affect any of the four basic safety goals, and the Reactor Protection System instrumentation, which was considered to be entirely safety related.

(2) The failed indications must have induced the operator to affect the four basic safety goals through his actions and since the operator was assumed to act correctly by procedure, then the failed indications must have directed him to the operation of the 23 systems important to the safety goals. To do so, the indications were assumed to be part of those 23 systems.

5-11



5.5.2 Identification (Cont'd)

(3) The failed indications must have been those by which the operator normally controlled the plant in its operational and emergency modes. These were the operator's "first line" of indications located in the control room. Local indications which by design were used for testing, performance muonitoring, or local manual control were considered outside of the scope of induced operator error.

If all three assumptions were met, then there existed the potential for induced operator error interactions and consideration was given to the chain of events which could possibly lead to a degradation of the basic safety goals. This chain of events was evaluated with respect to operations of the 23 systems important to the four basic safety goals.

5.5.3. Evaluation

Once a nonsafety indication has failed in the control room and the operator has observed the indication and assumed it to be correct, the process which leads to an induced operator error has commenced. Since IOE was not concerned with the mechanical or electrical cause of the failure, the evaluations of the induced operator error process did not take the failure into consideration. Instead they started with the operator's action, or non-action, depending on the mode in which the indication failed (high, low, on, off) and the procedural guidance which the operator followed. Most often the worst case evaluations involved a failure which induced the operator to believe that the system under observation was operating normally, leading to operator non-action. This was usually the case with annunciators and component condition indicating lights which were the majority of indications confronting the control room operator.

When procedural guidance was not specific, the evaluations were conducted using "good operating judgement" which required an examination of the operating characteristics of each system and the evaluation of multiple operator action paths. To control the scope of the evaluations, speculations into improbable operator actions and "what-if" determinations were avoided. Further efforts were taken to avoid a design review of the control room panels, control wiring, supports, and ergonomic layout.

5-12



5.5.3. Evaluation (Cont'd)

Evaluations were written on Induced Operator Error Analysis sheets, using an

abbreviated FMEA format similar to the form used for the evaluation of nonconnected interactions. Results were labeled as either acceptable or as having the potential for causing an induced operator error interaction and were identified with the safety goal that could be affected. Evaluations were considered acceptable (no possible IOE interaction) if the interaction did not affect the four basic safety goals or if there was a safety indication on the central control room panel, functionally redundant to the nonsafety indication under evaluation. Safety indications were not considered to fail, and the operator was assumed to make use of them, when available. Using all available indications is representative of good operating judgement.

5-13

INDUCED OPERATOR ERROR (IOE) INTERACTIONS

PMEMISE:

IDENTIFY INDICATION:

NONSAFETY INSTRUMENT OR COMPONENT

CONTROL ROOM CONTROL BOARD LOCATION

SYSTEM IMPORTANT TO 4 BASIC SAFETY GOALS I

DETERMINE ACTION:

EVALUATE:

TABLE NO. 5.1

5-14



CHAPTER 6

6.0 EVALUATION CRITERIA

As stated in Section 3.0 a systems interaction analysis involves the systematic search *for hidden systems interactions (couplings) and the evaluation of the effects of nonsafety system failure propagated into the safety systems by such interactions. Section 5.0 provided the methodology for searching for and evaluating adverse systems interactions and was characterized as being more comprehensive than the original IP3 licensing process. Chapter 6 prescribes the evaluation criteria or acceptance criteria used in identifying and in analyzing identified potentially unacceptable systems interactions consistent with Section 5.0.

6.1 Interconnected Systems

The evaluation of interconnected systems interactions and their effects on plant safety were based upon satisfying the system operability requirements as reflected in the FT's, LD's and AD's. Effects on safety system operation induced by random failures of safety related components were not included in the scope of this study. Furthermore, single failure criterion, as applied for Failure Modes and Effects Analysis of safety targets was assumed valid only on the system level. (See 5.1 for further discussion on single failure.)

Induced operator error interactions were evaluated and presented separately from the other interconnected systems interactions due to different criteria, specifically redundancy on the component functional level and the selective identification of nonsafety components. If a nonsafety related component was found to affect the four safety goals through an induced operator action, a search was made to locate a safety related indication that was functionally redundant to the nonsafety indication. If-functional redundancy existed, and assuming the operator utilized all available indications, then further evaluation of the induced operator error interaction was unnecessary. Also, nonsafety indications found outside of the boundaries of the 23 systems were excluded from evaluation.

6-1

6.0 EVALUATION CRITERIA (Cont'd)

6.2 Nonconnected Systems

6.2.1 Event/Source Credibility Evaluation

Potential sources were evaluated to determine if the postulated events could credibly lead to detrimental interaction with targets. The situations which were studied were categorized below. These generic categories were expanded to a more detailed list from which event information was taken for the system nonconnected evaluation forms and the evaluation of interaction credibility forms.

a) Event will not lead to interaction because of defensible qualification of the sources by analysis, test, or experience with the same or similar items..

b) Events may lead to damage or failure of. the sources, but the credible failure modes are not a threat to the safety goals of the target.

c) Events may lead to a credible failure mode of the sources which have the potential to cause adverse interactions threatening to the safety goals.

6.2.2 Source Evaluation Criteria

The following criteria provided minimum guidance for evaluation of sources of seismically induced events:

a) Structural Source Evaluation

Any nonsafety related structural element determined to be a potential source was assumed to fail, unless seismic qualification by analysis, test or comparison to similar previously qualified elements had been performed to ensure integrity.

b) Mechanical Source Evaluation

The following is the set of failure modes for mechanical equipment which was considered when evaluating potential sources. In addition to the specific failures below, complete loss of motive power for all source equipment and loss of control power was postulated. When examining the possibility of collision due to target/source proximity during the walkdown inspections, the relative motion between the sources and targets was considered.

6.0

6.0 EVALUATION CRITERIA (Cont'd) R1

6.2 Nonconnected Systems (Cont'd)

6.2.2 Source Evaluation Criteria (Cont'd)

b) Mechanical Source Evaluation (Cont'd)

- Overturning of tanks, pumps, filters or other unsupported

equipment where the center of gravity location as measured

from the base was longer than one-half the base width in all directions. Each direction was evaluated independently. A horizontal acceleration equivalent to at least that value

associated with the plant SSE, would be required to overturn an unsupported component whose height was less than 1/2 base

width from the base. Overturning was not considered where the

distance from the base to the center of gravity was small.

- All non-seismically qualified valves, pumps, tanks and vessels were assumed to fail in the "worst credible mode" possible.

(i.e., partial failure of valves and operation of pumps below

design flow rate). The "worst credible mode" was based on conservative engineering judgement.

c) Electrical Source Evaluation

Several categories of failure type were considered with regard

to seismic effects on electrical sources (equipment and cabling). They were:

c)-l Electrical Equipment

- Overturning of unsupported equipment where the center of gravity location as measured from the base was longer than one-half the base width in all directions. Each direction was evaluated independently. The same considerations discussed in regard to overturning of mechanical equipment apply to electrical equipment, i.e., overturning was assumed only for cases where the distance to the center of gravity was significant compared to the base width.

- All nonseismically qualified electrical equipment was

assumed to fail in the worst credible mode possible. The "worst credible mode" was based on sound engineering judgement.

6.0 EVALUATION CRITERIA (Cont'd) R1


6.2.2 Source Evaluation Criteria (Cont'd)

c) Electrical Source Evaluation (Cont'd)

All nonseismically supported electrical equipment was assumed to be a source of the "worst possible" physical and electrical interaction.

c)-2 Cable Trays

Seismically Supported Cable Trays

Cable trays that were determined to be seismically supported/restrained were assumed to remain physically intact in the event of an SSE (i.e., they do not become a source) and also that they could develop no electrical faults as built.

- Non-Seismically Supported Cable Trays

A non-seismic cable tray in the vicinity of essential

safety related equipment was a potential source and assumed to collapse. Also cables contained within the tray were assumed to develop electrical faults. The "vicinity" was defined by the criteria assumed and

illustrated in Figures 6-1 & 6-2.

c)-3 Conduits

Non-seismically supported/restrained conduits were

assumed to be the source of mechanical and electrical interactions in an SSE.

d) HVAC Source Evaluation

Non-seismically supported ductwork that runs directly over essential safety related targets was considered a source of potential interaction. The interaction boundary envelope is illustrated in Figure 6-3.

While considering systems interaction of HVAC systems, the effects of ductwork crimping and adverse operation (or non-operation) of non-safety related fans that might spread combustible or toxic fumes through the ductwork was considered.

EVALUATION CRITERIA (Cont'd)

6.2.2

Physical Impact: All non-seismically designed/supported piping running in the vicinity of targets could fail and physically impact the target within the pipe's volume of influence. The volume of influence was defined as five (5) pipe diameters or five (5) feet whichever was greater, laterally from the pipe center line. The pipe was assumed to fail anywhere along the piping run, during or post SSE. This criterion is based upon conservative engineering judgement and is illustrated in Figure 6-4. Additionally, for the evaluation of interaction credibility (EIC), specific failure modes were postulated separately (center of span, change in direction, hanger tear out).

Nonconnected Systems (Cont'd)

Source Evaluation Criteria (Cont'd)

d) HVAC Source Evaluation (Cont'd)

- Failure of in-line HVAC equipment followed the source evaluation criteria for mechanical equipment. Support failure resulting in tipping, falling, sliding or overturning may occur. Overturning was assumed possible when the distance as measured from the base to the center of gravity was more than one-half the width of the base. Each direction was evaluated independently.

e) Piping System Source Evaluation

- Random failures of high energy pipes, jet impingement effects, flooding effects and internal missile analyses were not included except in the cases where these effects were seismically induced.

- All piping and associated components identified as essential safety related components fell under the category of targets. Also they were assumed to be seismically supported or restrained and hence could not become seismically induced sources.

- Non-seismically designed piping was considered as sources in the following context:

6.0

EVALUATION CRITERIA (Cont'd)

Nonconnected Systems (Cont'd)

Source Evaluation Criteria (Cont'd)

e) Piping System Source Evaluation (Cont'd)

Flooding:

Environmental:

A non-seismic piping run in the vicinity of target equipment was assumed to have a circumferential or longitudinal rupture during or post SSE that could flood the room (attention was paid to the instrumentation cabinets, motors, etc. in the room), or flood any cable tray runs immediately above or below the piping run.

Piping failures or a resulting chain interaction could cause unacceptable environmental conditions enveloping target equipment, (e.g., auxiliary steam line

failures could result in a steam environment with elevated temperatures and humdity). Specific targets could either cease functioning or malfunction in this environment.

6.2.3

6.2.3.1

f) Instrumentation and Control, Source tvaluation

All instrumentation was assumed to malfunction in the "worst credible mode". Instrumentation that was not seismically mounted was assumed to fail structurally. The "worst credible mode" was based on conservative engineering judgement.

Comparison With NRC Guidelines

The following criteria provided guidance for evaluation of sources for pipe failure induced events.

The criteria provided by the NRC Standard Review Plans 3.6.1 and 3.6.2 with companion Branch Technical Positions BTP APCSB 3-1, MEB 3-1 and RG 1.46 were used to evaluate systems interactions associated with pipe failure induced

events Table 6-1 summarizes the acceptance criteria for external and internal challenging events relative to the system, component or structure being evaluated.

6.2.3.2 The following criteria provided guidance for evaluation of sources

for missile (internally and externally) generated induced events.

The criteria provided by the NRC Standard Review Plans 3.5.1, 3.5.2 and 3.5.3 were used to evaluate systems interactions associated with the effects of

6.0

6.2

6.2.2



6.2.3 Comparison With NRC Guidelines (Cont'd)

internally and externally generated missile systems interactions. Table 6-1 summarizes the acceptance criteria for challenging events relative to the system component or structure being evaluated.

6.2.3.3 The following criteria provided guidance for evaluation of sources associated with flooding induced events.

The criteria provided by the NRC Standard Review Plans 3.4.1 and 3.4.2 were used to evaluate adverse systems interactions associated with the effects of flooding. Table 6-1 summarizes the acceptance criteria for challenging events relative to the system, component or structure being evaluated.

6.2.3.4 The following criteria provided guidance for evaluation of sources resulting from the effects of severe environment.

The criteria provided by the NRC Standard Review Plans 3.3, 3.4, 3.5, 3.6 and 3.11 were used to evaluate systems interactions resulting from severe environmental conditions. In addition the guidance provided by IE Bulletin 79-QIB was used to the degree practicable for this evaluation. Table 6-1 summarizes the acceptance criteria for challenging events relative to the system, components or structure being evaluated.

6.2.3.5 Specific criteria used for the evaluation of area specific interactions are discussed as applicable in Sections 1.0 through 6.0 of the Summary.

6.2.4 Interaction Effects Evaluation Criteria

6.2.4.1 Categories of Interaction Effects

Interactions that were considered were direct physical interactions caused by such events as target impact from a falling or moving source. Evaluations of these interactions had to take into account the typical target effects of these interactions, as listed below.



6.2.4.1 Categories of Interaction Effects (Cont'd)

Mechanical:

- valve failing to operate

- piping rupture

- pump failing to operate

Electrical:

- loss of control power

- loss of motive power

- unwanted energization

Pneumatic:

- loss of pressure (loss of control)

- unwanted pressurization

- jet impingement

- hostile gas

Hydraulic:

- loss of pressure

(a) loss of control

(b) loss of lubrication

- unwanted pressurization

- jet impingement (environmental effect)

- flooding

- hostile fluids

6.0 EVALUATION CRITERIA (Cont'd) Rl


6.2.4.1 Categories of Interaction Effects (Cont'd)

Environmental (area intensive)

- elevated temperatures

- humidity

- radiation

6.2.4.2 Methods of Evaluation

Interactions were evaluated for their impact on the required safety goals of identified targets. The following are acceptable methods of evaluation of identified interactions.

a) Target Operability Evaluation:

One approach to resolution of an interaction was to show that the target's safety goal was not impaired. This was accomplished by studying the means by which impairment occurred and the possible effects of the impairment. For example, a pneumatically operated valve which might have been required to close during shutdown failed due to falling equipment which severed the air line to the valve operator. If the valve was a "fail open" type, then shutdown capability was compromised, but if the valve was a "failed closed" type, then shutdown capability was not compromised even though the air supply was lost. In this example it would have also been necessary to consider the consequences of crimping the air line, which would lock in air pressure to the operator. This example is typical of the reasoning process that was necessary in the evaluation of interactions for which a Substantial degree of conservative engineering judgement was used. In other cases where the target failure modes were found to, be unimportant to the results of the evaluation, they were not examined in detail. Decisions based on judgement, along with the rationale, are documented on the FMEA evaluation sheets.

6-9

6.0 'EVALUATION CRITERIA (Cont'd) R1


6.2.4.2 Methods of Evaluation (Cont'd)

b) Source Behavior Evaluation:

For the initial discovery process, simple conservative failure criteria were applied. A source pipe was assumed to fail at any location in order to allow it to fall so that it impacted the target. No credit was taken for existing structures which might limit drop distance or redirect its path. In this manner a large number of interactions were screened very quickly. Interactions failing this initial evaluation were then re-evaluated by a closer study of the source.

The last step in the source evaluation was to, if possible, determine if the source was capable of withstanding the postulated event. This was done using criteria in 6.2.4.3 and 6.2.4.4 and conservative engineering judgement. Where there was the slightest possibility of disabling damage to the target, the interaction was considered adverse.

6.2.4.3 Evaluation of Direct Interaction Effects

For evaluations of the effects of spatially coupled interactions the following guidance was established. For evaluations not covered by the criteria below, then the pertinent criterion was developed on a case-by-case basis, using the same level conservatism-as used by the criteria listed here.

a) Dynamic effects of breaks in piping were evaluated using the criteria in Section 6.2.3.1. One criterion, for example, was that no damage would result if the target pipe size is at least equal to the size of the source pipe and the wall thickness of the target pipe was at least equal to that of the source pipe.

b) Direct impact of missiles or falling objects on structures and components were evaluated when necessary using the criteria of Section 6.2.3.2. Care was taken to consider such appurtenances as instruments, power connections, cooling and lubrication connections.

c) Direct impact of missiles or falling objects on HVAC ducts were evaluated on a case-by-case basis.

6-10

6.0 EVALUATION CRITERIA (Cont'd) RI


6.2.4.3 Evaluation of Direct Interaction Effects (Cont'd)

d) Flooding effects of broken or leaking pipes were evaluated using the criteria of Section 6.2.3.3.

e) The effects of fire were evaluated using the criteria of Section 6.2.3.4.

f) Environmental effects of broken or leaking piping, tanks, etc. were evaluated by comparison of the estimated environment with the target's qualification profile. Helpful criteria and data are contained in Section 6.2.3.4.

6.2.4.4 Evaluation of Secondary Effects or Cascading Influences

Two types of secondary effects or cascading influences were considered, chain-reaction failures and degraded operation.

For the chain-reaction events where the operation of one major component affected the operation of other major components in a designed sequence, the criteria for evaluation of events were all similar to the direct interactions and were successively applied to each member of the chain. Each step in the chain scenario had an associated probability of less than one and judgement w as applied to consider only credible scenarios.

Degraded operation concerned itself with controlling components and power supplies which may affect major components. For example, in order for the plant to safely shut down, it is necessary for the safe shutdown valves to operate in the required manner or to fail in the required position. For this to occur the control systems must remain intact after the interaction event, or else be damaged only in such a way as to allow valve failure in the. required position. If an air operated valve is required to fail in a certain mode, it should do so on loss of air. If, however, the air line between the control device and the valve were to be impacted during a seismic event, the line might be crimped closed. This could prevent the venting of air and thereby prevent the valve from failing in its proper mode. The degraded components in this case might be the pneumatic control tubing, whose loss of pressure affects a major component.

Similarly, in electrically operated devices, a non-qualified component could impact the signal cable and cause damage which would adversely affect proper operation.

6-11



6.2.4.4 Evaluation of Secondary Effects or Cascading Influences (Cont'd)

To consider secondary process effects, process tubing, instrumentation, electrical cables and cable trays requiring protection from potentially unacceptable interactions were identified, examined for interactions, and evaluated for adverse effects to primary and secondary system components.

6.2.5 Modification Criteria

Modifications may be required to resolve identified nonconnected systems interactions which may degrade the four safety goals. While not performed as part of this study, they are the logical next step in seeking a resolution of discrepancies found during the study. These modifications may be any of the following:

a) Modification of the source to eliminate the adverse behavior by bracing, supporting, or reinforcing the source component.

b) Shielding or relocation of the target to preclude the physical interaction.

c) Modification of the target to permit retention of the required safety goals in spite of the interaction.

d) Alteration of system design to provide alternate means of accomplishing the safety goals.

e) ,Modification of the operating procedure such that the interaction could not cause a degradation of the associate~d safety goals.

The criteria for structural or mechanical modifications to safety related components are the same as documented for safety related structures and equipment. For relocation or modification of non-safety related equipment, the criterion for acceptability is that the modified configuration, when re-evaluated for interactions using the evaluation criteria previously stated, is found to have resolved the original interaction and not created any new interactions.

6-12

. RlINDIAN POINT8CLAR POWER PLANT SYSTEMS INTERACTION STUDY

NON-CONNECTED SYSTEMS INTERACTION-EVALUATION CRITERIA TABLE 6-1

ACCEPTANCE CRITERIA RELATIVE LOCATION OF CHALLENGING EVENT RELATIVE

TO SYSTEM COMPONENT BEING EVALUATEDCOMMON CAUSE EVENT

A. Earthquake

B. Pipe Failure

C . Missile

D. Flooding

FOLLOW-ON EVENT

a. structural failure b. pipe failure C. flooding d. severe environment e. missile

a. missile b. flooding c. severe environment d. structural

a. pipe failure b. flooding C. fire d. severe environment e. structural failure

a. severe environment b. structural failure

flooding f Ire temperature humidity radiation wind missile depressurization

EXTERNAL

See Chapter 6.2.2

" structure, system, component capable of withstanding the resulting effects of pipe whip, jet impingement, flooding and severe environment

" structure, system, component capable of withstanding the resulting effects of pipe failure, flooding, fire, severe environment, impact

o structure/compartment designed to adequately prevent flooding entry

o structure/compartment capable of withstandIng the resulting environmental condition

o no communicatinR paths

INTERNAL

See Chapter 6.2.2

" Whip restraints o Barriers/Shields " Separation

o Barriers/Shields o Separation

o Drainage system capable of handling maximum expected flood rate

o Components capable of functioning in submerged (flooded) environment

o Equipment/component qualified to the resulting environmental condition

o Compartment environment controlled

E. Severe

CORRESPONDING STANDARD REVIEW PLAN/REGULATORY GUIDE CRITERIA

lOCFR part 50, Appendix A GDC 2 Regulatory Guide 1.29 "Seismic Design Classification" Standard Review Plan 3.2.1 Seismic Classification

Standard Review Plan 3.6.l/APCSB.3-1 Standard Review Plan 3.6.2/MEB 3-1 Regulatory Guide 1.46

Standard Review Plan 3.5.1 Standard Review Plan 3.5.2 Standard Review Plan 3.5.3

Standard Review Plan 3.4.1 Standard Review Plan 3.4.2

Standard Review Plan 3.3 Standard Review Plan 3.4 Standard Review Plan 3.5 Standard Review Plan 3.6 Standard Review Plan 3.11

INTERACTION-ENVELOPE SOUNDARY

u z

ThRGIET

EU

ka

-I

EBASCO SERVICES INCORPORATEDi INIA POINTURE DIV.N DR. 5ON5MS IWEKAXTION STUODATE? 2"D - I weAicriN criT~iA.

ISCALE WMI I ~EPcLi CA&L CABSLE -MW&S

INTE.AC11C ENVELOPE BOUNDARY

TARGCET

EBASCO SERVICES INCORPORATED I - N ________________ FIGURE

DIV. IW !- DR..,un;A DATE CSTEI,,NTZ CTON STUDY

DATE"TES IN.CCTION MA-.2. SCALE .--..Ou .'.. LecftJc &L .AeoL 'T

(

TEE

NON-SEISMI SUPPORTS

S5TEMSIA4TERA&CTION SThD*I( IWTERACTI ON MRTER1b4 HVA~r- WqCWO,%.

NON.5EFWI' PMPE OR

ELEL CONUrr

5 FEET CWrHnE.VER IS -. .REhTER)

OR ELEC. CONDUIT I NON-5E| lMIC, SUPPORTS

--t4ERCTION 2 ENVELOPE 'BOUN DARY

EBASCO SERVICES INCORPORATED i ' DIAN POINT No. 1 DIV. ME DR..- m NUCLEAR PWt PLANT ITE -. m s I4TERACTION STUDr &-4

DATE~~ Ca5~ I,4TERACflON4 CluTe.%IA SCAL f1M &.w Pipe oR slec.T coNuIs

'SOURC-E-i

E~d ~

liE Zll

%J



CHAPTER 7

7.0 QUALITY ASSURANCE

In order to assure that the Systems Interaction Study project meets the requirements of the Quality Assurance Program, the Project Quality Assurance Engineer assigned qualified internal auditors to review the inprocess activities of the SIS Project personnel. Results of these audits were distributed to the SIS Project Manager and the PQAE for information and corrective action if required.

7.1 General

The contractor had in operation, over the duration of this contract, a system or program with supporting procedures as necessary which addressed the Quality Assurance requirements imposed in this section, and elsewhere in the contract.

7.2 Analysis Criteria And Basis

The basic document which identifies the applicable regulatory requirements, Sdesign bases, codes and standards and other criteria is a requirement and is specified in the Contractor's Engineering Procedures. The basic document was prepared at the outset of any analysis effort by the responsible engineering organization, and revised and approved in accordance with the applicable engineering procedure. The initial issue of the basic document was not intended to provide all the detailed requirements to be incorporated into analysis documents, but to provide sufficient basic requirements to permit the analysis process to proceed. As additional criteria were developed, they were incorporated into the basic document, following the review and approval prescribed by the applicable engineering procedures.

7.3 Internal Analysis Reviews

Internal Analysis Reviews were used during the analysis process, as defined in applicable engineering procedures, to assure acceptability of the document prior to the analysis verification process. These reviews consisted of checking and approval of applicable calculations, system descriptions, design specifications, and other documents as required. Errors and deficiencies identified were documented and appropriate corrective action instituted to preclude repetition, as specified in engineering procedures.

7.4 Analysis Verification

Analysis verification was performed on all final analysis documents by individuals or groups other than those who performed the original analysis work, in accordance with applicable engineering procedures, to assure compliance with applicable analysis bases, regulatory requirements, codes and standards. The authority and responsibilities of personnel who performed these reviews were identified and controlled by engineering procedures.

The analysis verification process was performed in accordance with a plan which considered the importance of the engineering tasks, structure, system, or component, and was accomplished through one or more of the following actions:

- Independent review or special review of analysis documents

- Alternate calculations necessary to assure the analysis met the specified criteria.

The results of all analysis verification activities were documented, and reviewed by cognizant management personnel in accordance with applicable procedures.

7.5 Data Package

A formal engineering data package review was established and consisted of (as a minimum) the following:

A. Identify and list documentation on a Document List with revisions which were included in the data package.

B. The final review by the responsible engineers including sign-off on the Document List, which attested that the engineering scope, including the support engineering data, relative to the assigned tasks had been completed

C. A final supervisory review and approval sign-off of the engineering data package.

7.6 Records

Analysis documents and reviews, records and changes thereto were collected, stored, aind maintained as part of the official project records, in accordance with applicable record procedures.

7.7 Types Of Documents

For the purposes of control, three types of documents were considered as a minimum:

- Documents supplied or referenced by the Power Authority for use by the Contractor.

7-2

7.7 Types Of Documents (Cont'd)

- Technical documents prepared by the Contractor for delivery to the Power Authority, including design analysis report, drawings, and design documents.

- Documents for the control of contractor and sub-contractor activities, including the Engineering Manual, Quality Assurance Manual, and Procedures.

7.8 Storage And Security Of Records

Records were stored in facilities which minimized the potential for their destruction by fire, flooding, theft, and deterioration from environmental conditions.

7.9 Audits

Audits were performed by the Contractor on activities which related to the quality of the service performed, in accordance with a quality assurance procedure.

These audits included internal audits of the project organization and interfacing organizations. The audit procedure included provisions for planning, performance, evaluation, and reporting.

7.10 Audit Results, Evaluation and Report

Audit results were documented, and then reviewed with management having responsibility in the area audited. The audit report included the following:

- Description of audit scope

- Identification of auditors

- Persons contracted during the audit

- Summary of audit results, including an evaluation statement regarding the effectiveness of the quality assurance program, applicable to the areas audited.

- Description of identified non-conformances

- Recommendations for corrective action, when applicable

7.11 Reaudit

When warranted by-audit results, deficient areas were subject to reauditing on a timely basis to verify the effective implementation of corrective action.

7.12 Audit Program Review

Audit data was analyzed by the Chief Quality Engineer on a periodic basis and reported to responsible management personnel for reviews and assessment.



CHAPTER 8

8.0 REFERENCE DOCUMENTS

8.1 Reference Documents

1. FSAR - Consolidated Edison Company of New York Indian Point Nuclear Generating Unit No. 3. (Sup 32 Nov '75) (Initial evaluation)

2. Answers to AEC Questions. (Sup 31 Oct '75) (Initial evaluation)

3. Safety Evaluation Report by the Director of Licensing U S Atomic Energy Commission in the Matter of Con Ed Company of New York, Inc. 1973.

4. Final Safety Analysis Report, Indian Point 3, Nuclear Power Plant IP3 FSAR Update Rev 0. 7/82.

5. Fire Protection Safety Evaluation Report by NRC for IP3 (3-6-79)

6. Review of Indian Point Station Fire Protection Program Vol I. Dec 1976. (Rev I April '77)

7. Vol II Prepared in response to APCSB 9.5-1, Appen. (Rev 1 April '77)

8. Indian Point Station Unit No. 3 - System Descriptions. (Rev 1)

9. Probabilistic Risk Assessment: Pickard Lowe & Garrick Inc. (Oct 7, '80)

10. PASNY-Indian Point 3 Instrument Bus Failure Analysis (IE 79-27) EDS Job Number 0900-007-831, Report No. 02-0900-6 & 7. (Rev 0, 6-13-80)

11. Review of Evaluation of Systems Interactions Methods, NUREG/CR-1901, BNL-NUREG-51333. (Jan'81)

12. Systems Interaction: State of the Art, Review and Methods Evaluation, NUREG/CR-1859, UCRL-53016. (Jan '81)

13. Final Report Phase I, Systems Interaction Methodology Applications Program. NUREG/CR-1321, SAND 80-0384 AN. (April '80)

14. Description of the Systems Interaction Program for Seismically-Induced Events: Diablo Canyon Units 1 & 2. (R-4 Aug 29, '80)

Safety Evaluation Report (Sup 11 Oct '80)

15. TMI Items Approved for Implementation, NUREG 0737. (10-31-80)

8.1 Reference Documents (Cont'd)

16. Identification of Unresolved Safety Issues Relating to Nuclear Power Plants; Report to Congress, NUREG 0510. (2-12-79)

17. Unresolved Safety Issues Summary: AQUA Book NUREG-0606 Vol 4, No. 1. (2-19-82); Task A-17, Systems Interactions in Nuclear Power Plants.

18. TMI-2 Lessons Learned, Task Force Final Report, Recommendation 9-Review of Safety Classifications and Qualifications. NUREG 0585 (Oct '79)

19. Action Plans for Implementing the Recommendations of the Presidents Commission & Other Studies of TMI 2 Accident. Task II. C.1, Systems Engineering Reliability Engineering & Risk Assessment, NUREG 0660. (May '80)

20. BTP MEB-3-1: Postulated Break and Leakage Locations in Fluid Systems Piping Outside Containment. (5-76)

21. BTP APCSB 3-1: Protection Against Postulated Piping Failures in Fluid Systems Outside Containment. (5-76)

22. BTP APCSB 9.5-1 Guidelines for Fire Protection for Nuclear Power Plants. (5-76)

23. PRA Procedures Guide, A Guide to the Performance of Probabilistic Risk Assessment for Nuclear Power Plants, NUREG/CR-2300. (Rl 4.5.82)

24. NSSS Instrumentation and Control Equipment, SO Int 320, Run Date 4-14-76, (64 pages).

25. Instrument Schedule Secondary Plant (Oct '71), Con Ed Dwg. No. D-202540 (Sheets I to 95 & 142)

26. Westinghouse Electric Corp Conventional Line Schedule Class 1 Piping, 10 sheets UE&C Dwg. No. 9321-C-23003-1, Con Ed Dwg No. C-202787. (Rev 1)

27. Westinghouse Electric Corp Nuclear Line Schedule (Sheets 1 to 39), UE&C Dwg No. 9321-C-27413-4, Con Ed Dwg No. C-202251. (6-77)

28. Westinghouse Electric Corp Valve & Specialty List (Sheets 1 to 104), UE&C Dwg Nos. 9321-LL-20433-8; 9321-05-20433. (6-77)

29. WEDCO IPP3 Valve Report, Sheets 1 to 354. (6-20-74)

30. Indian Point 3 Alarm Response Procedures (latest rev. 11-21-80)

31. Indian Point 3 System Operating Procedures (12-2-82)

32. Indian Point 3 Off Normal Operating Procedures (4-26-82)

8-2


33. Indian Point 3 Plant Operating Procedures (12-2-82)

34. Indian Point 3 Alarm Response Procedures (6-9-82)

35. Indian Point 3 Plant Emergency Procedures (8-23-82)

36. UE&C, Indian Point Unit 3 Pipe Supports, Design Guidelines (7-24-79)

37. WAH-4, General Administrative; Seismically Supported Cable Trays; Tray Loadings Used in Original IP3 Design. (Telecon IUP 2224, M 1056) (11-14-79)

38. Indian Point Generating Station Unit No. 3, Cable Tray - Seismic Analysis (APD No. 17186) (3-29-73)

39. Indian Point 3 Nuclear Power Plant; PASNY Work Authorization WAH-68/71; IE Bulletins 79-02 and 79-07 (Nl-1051 UE&C Memo) (11-20-79)

40. Indian Point 3 Tech Specs for Protective Instrumentation CCR Ventilation Controls Sketch. (PASNY to Ebasco Letter IP-JMC-323 of 2-10-83)

41. Indian Point 3 Air Operated Containment Isolation Valve List. Air Operated Valves Effected by Plant Emergency and Emergency Safeguards Procedures. Off Normal Operating Procedures for Loss of Instrument Air. Shutdown Steps for Loss of Instrument Air. (PASNY to Ebasco Letter IPO-83-02, dated 2-8-83)

42. Potential Interactions Between Non-Safety Related Control Systems and Safety Systems; NRC Information Notice 79-22.

43. Report on Review of Systems Interaction Methodologies; NUREG/CR-BMI-2055, R-2.

44. Environmental Qualification of Class IE Equipment; IE Bulletin 79-01B.

45. Code of Federal Regulations, Title 10 - Energy.

46. Zion/Indian Point Generating Unit No. 3, Probabilistic Risk Assessment; Pickard, Lowe and Garrick, Inc. - DRAFT (Used in initial preparations).

47. American National Standard N18.2 (ANS 51.1) Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.

48. NRC Regulatory Guide; 1.46, Protection Against Pipe Whip Inside Containment (5.73).

8-3


49. NRC Standard Review Plan; 3.3.1, Wind Loadings (7.81).

50. NRC Standard Review Plan; 3.3.2, Tornado Loadings (7.81).

51. NRC Standard Review Plan; 3.4.1, Flood Protection (7.81).

52. NRC Standard Review Plan; 3.4.2, Analysis Procedures (7.81).

53. NRC Standard Review Plan; 3.5.1.1, Internally Generated Missiles (Outside Containment) (7.81).

54. NRC Standard Review Plan; 3.5.1.2, Internally Generated Missiles (Inside Containment) (7.81).

55. NRC Standard Review Plan; 3.5.1.3, Turbine Missiles (7.81).

56. NRC Standard Review Plan; 3.5.1.4, Missiles Generated by Natural Phonomena (7.81).

57. NRC Standard Review Plan; 3.5.1.5, Site Proximity Missiles (Except Aircraft) (7.81).

58. NRC Standard Review Plan; 3.5.1.6, Aircraft Hazards (7.81).

59. NRC Standard Review Plan; 3.5.2, Structures, Systems, and Components to be protected from Externally Generated Missiles (7.81).

60. NRC Standard Review Plan; 3.5.3, Barrier Design Procedures (7.81); and Appendix A (7.81).

61. NRC Standard Review Plan; 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment (7.81); and BTP ASB-3-1 (7.81).

62. NRC Standard Review Plan; 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping (7.81) and BTP MEB-3-1 (7.81).

63. NRC Standard Review Plan; 3.11, Environmental Design of Mechanical and Electrical Equipment (7.81).

64. NRC Standard Review Plan; 9.5.1, Fire Protection Program (7.81) and BTP CMEB 9.5.1 (7.81) and Appendix A (7.81).

8-4

8.2 Reference Drawings

TITLE

PLOT PLAN

CONTAINMENT ENLARGED CONCRETE PLAN @ EL 99'0

T/B MISC PIPE SUPPORTS

CONTAINMENT BLDG SLABS AT FAN FLOOR EL 68'-0 SH 1

CONTAINMENT BLDG SLABS AT FAN FLOOR EL 68'-0 SH 2

SHIELD WALL AREA

SHIELD WALL AREA

CONTAINMENT BUILDING REACTOR COOLANT PUMP SUPPORTS

CONTAINMENT BUILDING FAN FLOOR FRAMING AT EL 68'0

CONTAINMENT BUILDING ANNULUS FRAMING SH 1

T/B TURB CRANE RUNWAY & HTR BAY ROOF PLAN

T/B OPER FL EL 53'-0 LAYDOWN PLAN FOR DISMANTLED UNIT

T/B MAIN ROOF PLAN

DOOR SCHEDULE & MISC DETAIL

CONTAINMENT & DIESEL GENERATOR BLDG

CONTR & DIESEL GEN BLDG'S DOOR OPENING DET'S & SCH

ELEC PENETR TUNNEL & PERSONNEL ENTR SHIELDING BLDG

FAN2 HOUSE & PIPE PENETRATION AREA PLANS

FAN HOUSE & PIPE PENETRATION AREA PLAN AT 54'-0 & 67'-0

FAN HOUSE & PIPE PENETR AREA SECT'S & DET'S SHT No

FAN HOUSE & PIPE PENETR AREA SECT'S & DETAILS

SHIELD WALL AREA

SHIELD WALL AREA

SHIELD WALL AREA

SHIELD WALL AREA

DRAWING NO.

9321-F-i0023

9321-F-10903

93 21-F-11203

9321-F-i1943

93 21-F-11953

9321-F-12303

9321-F-12313

9321-F-12883

93 21-F-12933

9321-F-12 963

9321-F-13 713

9321-F-13773

93 21-F-13783

9321-F-13793

93 21-F-13813

9321-F-13823

93 21-F-13853

9321-F-13933

93 21-F-13943

9321-F-13963

93 21-F-13993

9321-F-14863

93 21-F-14873

9321-F-14883

93 21-F-14893

8-5

REV.

8.2 Reference Drawings (Cont'd)

TITLE

SHIELD WALL AREA

T/B & HTR BAY G.A. OPERATING FL.-EL 53'-0

T/B & HTR BAY G.A. MEZZ FL.-EL 36'-9

T/B & HTR BAY G.A. GROUND FL.-EL 15'-0

T/B & HTR BAY G.A. CROSS SECTION



T/B-TWO BAY EXTENSION GA PLANS & ELEV

INTAKE STRUCTURE G.A. PLAN

INTAKE STRUCTURE G.A. SECTION

AUX FEED PUMP BLDG GA PLANS-SH #1

AUX FEED PUMP BLDG GA SECTIONS-SH #2

FLOW DIAGRAM-SYMBOLS

FLOW DIAGRAM-MAIN STEAM

FLOW DIAG.-COND & BOILER FEED PUMP SUCTION

FLOW DIAG.-BOILER FEEDWATER

FLOW DIAG.-EXTRACTION STEAM

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

DIAG.-FLASH EVAPORATOR

DIAG.-HEATER DRAINS & VENTS

DIAG.-MOISTURE SEP. & REHTR. DRAINS & VENTS

DIAG.-BOILER FD PMP TURBINE STM LINES D&V

DIAG.-CONDENSER AIR REM. & WTR BOX PRIMING

DIAGRAM-CIRCULATING WATER

DIAG-JACKET WTR TO DIESEL GENERATORS

DIAG-STARTING AIR TO DIESEL GENERATORS

DIAG-FUEL OIL TO DIESEL GENERATORS

DIAG-EXTRACTION STM TRAP SYSTEM

DIAG-GLAND SEALING FOR VALVES & PUMPS

DIAG-SERV. & CLG WTR RIVER WTR & FRESH WTR

DRAWING NO.

9321-F-14903

9321-F-20043

9321-F-20053

9321-F-20063

9321-F-20073

9321-F-20083

9321-F-20093

9321-F-20103

9321-F-20113

9321-F-20123

9321-F-20143

9321-F-20153

9321-F-20163

9321-F-20173

9321-F-20183

9321-F-20193

9321-F-20203

9321-F-20213

9321-F-20223

9321-F-20233

9321-F-20243

9321-F-20253

9321-F-20263

9321-F-20283

93 21-F-20293

9321-F-20303

9321-F-20313

9321-F-20323

9321-F-20333

8-6

REV.


TITLE

DIAGRAM-CITY WATER

DIAGRAM-STATION AIR

DIAGRAM-INSTRUMENT AIR

DIAG-LUBE OIL

DIAG-CHEMICAL FEED

DIAGRAM-CHLORINATION

DIAGRAM-HYDROGEN & CO2

DIAGRAM-MAIN STM TRAPS-SH #1

DIAGRAM-MAIN STM TRAPS-SH #2

YARD AREA-H 2 & CO2 PIPING-SH #1

T/B & HTR BAY H2 & CO2 PIPING-SH #2

TURB HALL & SUPER HTR BLDG UNIT #1 PIPING TIE-INS

FOR UNIT #2 PIPING SYSTEMS

YARD AREA-WEST OF CONT BLDG MAIN STM PPG

YARD AREA-WEST OF CONT BLDG M.S. PPG SECT'S & ELEV

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

YARD AREA

T/B & HTR

T/B & HTR

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

MAIN

MAIN

MAIN

COND

COND

CON D

COND

COND

COND

MAIN

ST1- PPG

STM PPG

STM PPG

& BOILER F]

& BOILER F]

& BOILER F]

& BOILER F]

& BOILER F]

& BOILER F]

STM PIPING

PMP PP

PMP

PUP

PMP

PMP

SUCTION

SUCTION

SUCTION

SUCTION

SUCTION

SUCTION

BOILER FEED PIPING

BAY GLAND SEAL PPG

BAY EXTRACTION STM PPG

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

FLOW

DRAWING NO.

93 21-F-20343

9321-F-20353

9321-F-20363

9321-F-20373

9321-F-20383

9321-F-20393

9321-F-20403

9321-F-20413

9321-F-20423

9321-F-20443

9321-F-20453

9321-F-20483

9321-F-20493

9321-F-20503

9321-F-20513

9321-F-20523

9321-F-20533

9321-F-20543

9321-F-20553

9321-F-20563

9321-F-20573

9321-F-20583

9321-F-20593

9321-F-20603

9321-F-20613

9321-F-20673

9321-F-20693

REV.

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B


TITLE

YARD AREA-WEST OF CONT BLDG MAIN STM PIPING

T/B HTR BAY EXTRACTION

T/B HTR BAY EXTRACTION

STM PPG EL SHT NO. 4

STM PPG EL SHT NO. 5

MAIN STEAM PIPING

YARD AREA

T/B & HTR

T/B & HTR

T/B & HTR

T/B & HTR

T/B & HTR

H.P. TURB

HTR DRNS & VENTS

HTR DRNS & VENTS

HTR DRNS & VENTS

HTR DRNS & VENTS

HTR DRNS & VENTS

HTR DRNS & VENTS

HTR DRNS & VENTS

OIL DRAIN & VENT

COMPOSITE

HTR DRNS & VENTS

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG SHT NO. 8

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

T/B

NO. 1

NO. 2

SHT NO. 1

SHT NO. 2

S.J.A.E. BLOWER VENT PPG

BAY SER & CLG WTR PPG RIVER WTR SYSTEM




BAY SER & CLG WTR PPG CLOSED SYSTEM

DRAWING NO.

9321-F-20703

9321-F-20713

9321-F-20723

9321-F-20733

9321-F-20743

9321-F-20753

93 21-F-20763

9321-F-20773

93 21-F-20783

9321-F-20793

9321-F-20803

9321-F-20813

9321-F-20833

9321-F-20843

9321-F-20863

9321-F-20873

9321-F-20883

9321-F-20903

9321-F-20913

9321-F-20923

9321-F-20933

9321-F-20963

9321-F-20973

9321-F-20983

9321-F-20993

9321-F-21003

9321-F-21023

8-8

REV.

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

HTR

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

BAY

WATER BOX PRIMING PPG SHT

WATER BOX PRIMING PPG SHT

HTR DRN TK OUTLINE DWG

COND AIR REMOVAL PPG PLAN

COND AIR REMOVAL PPG PLAN

PRIMING SYSTEM PPG


TITLE

T/B & HTR BAY SER & CLG WTR PPG CLOSED SYSTEM

CONTROL BLDG SERV & CLG WTR PPG CLOSED SYSTEM

INTAKE STRUCTURE SERV WTR PPG RIVER WTR SYS

INTAKE STRUCTURE SERV WTR PPG RIVER WTR SYS

MAIN STEAM PIPE RESTRAINT DETAILS

MAIN STEAM PIPE RESTRAINT DETAILS

INTAKE STRUCTURE CHLORINATION PPG SHT NO. 1

INTAKE STRUCTURE CHLORINATION PPG SlT NO. 2

BOILER FEED PPG RESTRAINT DETAILS

YARD AREA STATION & INSTRUMENT AIR PPG

T/B HTR BAY STATION & INSTRUMENT AIR PPG

HTR BAY & CONTROL BLDG STA & INSTRU AIR

CONTROL BLDG INSTRUMENT AIR PPG

T/B HTR BAY ARRGT

T/B HTR BAY REHTR BALANCE LINE PPG (VOID)

AUX FEED PUMP BLDG TURB SUPPLY & EXHAUST PPG

AUX BLDG FEEDWATER PPG SHT NO. 2

T/B & AUX FEED PUMP BLDG

T/B YARD AREA MS SYS TRAP PPG SHT NO. 1




T/B HTR BAY TURB OIL CONDITIONING SHT NO. 1

T/B HTR BAY TURB OIL CONDITIONING SHT NO. 2

HTR BAY FLASH EVAP PPG SHT NO. 1

HTR BAY FLASH EVAP PPG SHT NO. 2

T/B HTR BAY & YARD AREA SAFETY VALVE DRN PPG

T/B HTR BAY MISCELLANEOUS PPG SHT NO. 1

DRAWING NO.

93 21-F-21033

9321-F-21053

93 21-F-21063

9321-F-21073

9321-F-21083

9 321-F-21093

93 21-F-21103

9321-F-21113

93 21-F-21123

9 321-F-21133

93 21-F-21143

9 321-F-21153

9321-F-21163

9321-F-21173

93 21-F-21193

9321-F-21253

93 21-F-21273

9321-F-21283

93 21-F-212 93

9 321-F-21303

9321-F-21313

9 321-F-21323

93 21-F-21333

9321-F-21343

93 21-F-213 53

9 321-F-21363

93 21-F-21373

9 321-F-21383

REV.


TITLE

T/B HTR BAY MISCELLANEOUS PPG SHT NO. 2

T/B HTR BAY MOISTURE SEPAR DRN TANKS

T/B HTR BAY PPG BELOW FL EL 15'-0

T/B & YARD AREA FLASH EVAP PPG

YARD AREA WEST OF CONT COMP PPS

T/B HTR BAY EXTRACTION STM SYS-TRAP PPG SHT NO. 1

T/B HTR BAY EXTRACTION STM SYS-TRAP PPG SHT NO. 2

HTR BAY HD TK CLOSE CLG SYS

AUX BLDG FL & WELL SLEEVES SHT NO. 1

AUX BLDG FL & WELL SLEEVES SHT NO. 2

SAFETY VALVE SETTINGS FOR STM GEN SECONDARY SIDE

YD AREA RESTRAINT & SUPPLY DES LINES 4 SlT NO. 2

YD AREA RESTRAINT & SUPPLY DES LINES 5,6,7&8

AUX BLDG RESTRAINT & SUPT DES LINES 1071

T/B HTR BAY-BFP TURB EXHAUST STM PPG

AUX BLDG RESTRAINT & SUPPT DES LINES 1074, 1075 & 1076

AUX BLDG RESTRAINT & SUPPT DES LINES 1072 & 1073

T/B HTR BAY-PPG THRU COND NECK

TRANSF COMP PPG UNDERGROUND

TRANSF YD CITY WATER FOR AUX FD PUMPS

CONTROL BLDG & PAB REST & SUPPT DESIGN LINE 11-SS

PAB RESTRAINT & SUPPT DES LINE 1155

YD AREA RESTRAINT & SUPPT DES LINES 1081 THRU 1085

YD AREA RESTRAINT & SUPPT DES LINES 1028,1029,1070,1080

DGB AREA RESTRAINT & SUPPT DES LINES 1093 THRU 1101,1223


DIESEL GEN BLDG GENERAL ARRG'T

DIESEL GEN BLDG GENERAL ARRG'T

CONDENSATE STORAGE TANK DETAILS

DRAWING NO.

93 21-F-21393

9 321-F-21413

93 21-F-214 53

9 321-F-21473

93 21-F-21493

9321-F-21503

9321-F-21513

9321-F-21723

93 21-F-22003

9 321-F-22013

93 21-F-22023

9321-F-22043

93 21-F-22063

9 321-F-22073

93 21-F-22253

9321-F-22283

93 21-F-22293

9 321-F-22333

93 21-F-223 63

9 321-F-22373

93 21-F-22393

9 321-F-22403

93 21-F-22423

9321-F-22433

93 21-F-22443

9321-F-22453

93 21-F-22503

9 321-F-22513

93 21-F-22523

8-10

REV.


TITLE

T/B HTR BAY-TUNNEL VENT PPG

AUX FEED PUMP BLDG ADD TO CHEM FEED

YD AREA & AUX B.F. PUMP BLDG

CONTR BLDG HD TK CLOSED CLG SYS

DIESEL GEN BLDG SERV WTR & STARTING AIR PPG

DIESEL GEN BLDG FUEL OIL & JACKET WTR PPG

DIESEL GEN BLDG FUEL OIL & JACKET WTR PPG

DIESEL GEN BLDG EXHAUST SYSTEM

INTAKE STRUCTURE CW PPG SHT NO. 1

COND & BLR FD PUMP SUCTION ELEC FREEZE PROTECT

BLR FW ELECTRICAL FREEZE PROTECTION

SERV & CLG WTR-RIV & FRESH WTR ELEC FREEZE PROT

CHLORINATION ELEC FREEZE PROTECTION

CIRC WTR ELEC FREEZE PROTECTION

FLASH EVAP ELEC FREEZE PROTECTION

MAIN STEAM ELEC FREEZE PROTECTION

FUEL OIL TO DIESEL GEN ELEC FREEZE PROT.

MAIN STEAM TRAPS ELEC FREEZE PROT.

JACK WTR TO DIESEL GEN ELEC FREEZE PROT.

F.D. STATION AIR PPG

YD AREA COND PPG SHT

YD AREA COND PPG SHT

AUX BLDG RESTRAINT &






ELEC FREEZE PROT.

NO. 1

NO. 2

SUPPT DES LINES 1001






& 1007

& 1008

& 1005

& 1006

DRAWING NO.

93 21-F-22533

9321-F-22543

9321-F-22553

9 321-F-22563

93 21-F-22573

9321-F-22583

93 21-F-22593

9 321-F-22603

93 21-F-22623

9321-F-22703

93 21-F-22 713

9321-F-22723

93 21-F-22733

9 321-F-22843

93 21-F-227 53

9 321-F-22773

9321-F-22783

9 321-F-22793

93 21-F-22803

9 321-F-22813

93 21-F-22873

9321-F-22838

93 21-F-22903

9321-F-22913

93 21-F-22923

9321-F-22933

93 21-F-22943

9 321-F-22953

8-11

REV.

F.D.

F.D.

F.D.

F. D.

F.D.

F.D.

F.D.

F.D.

F.D.

F.D.


TITLE

AUX BLDG RESTRAINT & SUPPT DES LINES 1014,1015,1030&1031

AUX BLDG RESTRAINT & SUPPT DES LINES 1011,1012,1013

CONVENTIONAL LINE SCH. CLASS 1 PPG SHTS 1-10

SHT 9-REV. 6, SHT 10-REV. 3

SECONDARY PLANT-PIPE WHIP RESTRAINT SCHEDULE

YD AREA RESTRAINT & SUPPT DES LINES 3 SHT 2, 1022&1023

AUX BLDG RESTRAINT & SUPPT DES LINES 1120,1121,1123,1124

1125 & 1126

PAB RESTRAINT & SUPPT DES LINES 1155 SHT NO. 4




SHIELD WALL RESTRAINT & SUPPT DES LINES 1026 & 1027

DGB RESTRAINT & SUPPT DES LINES 1045,1048,1049,1111,1114

DGB RESTRAINT & SUPPT DES LINES 1051 THRU 1055, 1057

THRU 1061,1063,1069


DGB RESTRAINT & SUPPT DES






YD AREA RESTRAINT & SUPPT

LINES 1046,1047,1056,1062

1064,1065, THRU 1068

LINES 1112,1113,1115 & 1116

LINES 1108,1109 & 1110

LINES 1105,1106 & 1107

LINES 1102,1103 & 1104

LINES 1077,1078 & 1079

DES LINES 1156

CONTROL BLDG RESTRAINT & SUPPT DES LINES 1181 & 1182

CONTROL BLDG RESTRAINT & SUPPT DES LINES 1183,1184,

1185,1204,1205

DRAWING NO.

9321-F-22963

9321-F-22973

9321-F-23003

9321-F-23013

9321-F-22023

9321-F-23033

9321-F-23043

9321-F-23053

9321-F-23063

93 21-F-23073

9321-F-23083

93 21-F-23093

9321-F-23103

9321-F-23113

93 21-F-23123

9321-F-23133

9321-F-23143

93 21-F-23153

.9321-F-23163

93 21-F-23173

9321-F-23183

93 21-F-23193

9321-F-23203

8-12

REV.


TITLE

CONTROL BLDG RESTRAINT &



SUPPT DES LINES

SUPPT DES LINES

SUPPT DES LINES

1186

1190

1196

THRU 1189

THRU 1195

THRU 1202


PAB RESTRAINT & SUPPT DES

PAB RESTRAINT & SUPPT DES


T/B RESTRAINT & SUPPT DES





T/B RESTRAINT & SUPPT DES





YD RESTRAINT

CONTROL BLDG

CONTROL BLDG

CONT BLDG GA

CONT BLDG GA

CONT BLDG GA

CONT BLDG GA

CONT BLDG GA

& SUPPT DES

RESTRAINT &

RESTRAINT &

EL 95'-0

EL 68'-0

EL 46'-0

SECTIONS

SECTIONS

LINES 1155 SHT NO. 8

LINES 1155 SHT NO. 9

DES LINES 1157 SHT NO. 1

LINES 1131,1132,1133,1134 &

1139

DES LINES 1135 THRU 1138

LINES 1214,1215 & 1216

DES LINES 1157 SHT NO. 2

LINES 1117,1118,1119,1217&1218

LINES 1219 THRU 1222

LINES 1033, 1034 & 1035

LINES 1036, 1039 & 1042

LINES 1037, 1040 & 1043

LINES 1038, 1041 & 1044

LINES 2 SHT NO. 2, 1020 & 1021

SUPPT DES LINES 1225,1227,1229

1230

SUPPT DES LINES 1223 SHT NO. 2

CONT BLDG GA SECTIONS

DRAWING NO.

9321-F-23213

9321-F-23223

9321-F-23233

9321-F-23243

9321-F-23253

9321-F-23263

9321-F-23273

9321-F-23283

9321-F-23293

9321-F-23303

9321-F-23313

9321-F-23323

9321-F-23333

9321-F-23343

9321-F-23353

9321-F-23363

9321-F-23373

9321-F-23383

9321-F-23413

9321-F-23423

9321-F-25013

9321-F-25023

9321-F-25033

9321-F-25063

9321-F-25073

9321-F-25083

8-13

REV.

3

3

3


TITLE

CONT BLDG REFUELING EQUIP LAYDOWN & STORAGE

PAB GA EL 15'-0, 32'-6, 34'-0 & 41'-0

PAB GA SECTIONS PIPE TRENCH GA

FAN ROOM & PIPE TRENCH GA

FUEL STORAGE BLDG GA *PRINT

PAB GA EL 55'-0 & 73'-0

PAB GA

WASTE HOLDUP TANK PIT

FAN ROOM GA

NUCLEAR TANK FARM GA

NUCLEAR TANK FARM GA

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

BLDG-COMP PPG AT REACTOR COOL PUMP #31




BLDG-COMP PPG OUTSIDE REFUELING CANAL WALLS SHT NPl

BLDG-COMP PPG OUTSIDE REFUELING CANAL WALLS SHT NP2

BLDG RTD PPG FOR PRIMARY COOLANT SYS

BLDG PRI COOL SYS LOOP TO STM GEN #31



CONT BLDG PRIM COOL PRESSURIZER PPG (NORTH 1/2)

CONT

CONT

CONT

BLDG

BLDG

BLDG

PRIM COOL PRESSURIZER PPG (SOUTH 1/2)

MAIN STM PPG-STM GEN #31 & 32


DRAWING NO.

93 21-F-25093

9 321-F-25103

93 21-F-25113

9 321-F-25123

93 21-F-25133

9 321-F-25143

93 21-F-25153

9321-F-25163

93 21-F-25173

9 321-F-25183

93 21-F-25223

9321-F-25233

93 21-F-25263

9321-F-25273

93 21-F-25283

9321-F-25293

93 21-F-25303

9321-F-25313

93 21-F-25323

9321-F-25333

93 21-F-25343

9321-F-25353

9321-F-25373

9321-F-25383

9321-F-253 93

9321-F-25403

8-14

REV.


TITLE

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG



RTD PPG FOR PRIM COOL SYS. SH NO. 2


PRIM COOL PRESS SAFETY RELIEF VALVE PPG

PRIM COOL PRESS SAFETY RELIEF VALVE PPG


PRESS SAFETY RELIEF VALVES RESTR. DETAILS

BOILER FD PPG-STM GEN #31 & 32


BOILER FD PPG-STM1 GEN #31 & 32


ST11 GEN BLOWDOWN TREATMENT SYSTEM

CONT BLDG STM GEN BLOWDOWN PPG

CONT BLDG STM GEN BLOWDOWN PPG

STM GEN NITROGEN CIRCULATION SYS.

BLG-COMP

BLG-COMP

BLDG AUX

BLDG AUX

BLDG AUX

BLDG AUX

CONT BLDG AUX

PPG IN RESIDUAL HT EX. ROOM

PPG IN RESIDUAL HT EX. ROOM

COOLANT SYS

COOLANT SYS

COOLANT SYS SECTIONS

COOLANT SYS SECTIONS

COOLANT SYS PPG AT REACTOR COOLING

SUPP BLOCK

PAB COMPOSIT PPG ARRG'T EL 55'-0 SHT #5



PAB COMPOSIT PPG ARRG'T EL 55'-O SHT #2

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

8-15

REV.DRAWING NO.

93 21-F-25413

9321-F-25423

93 21-F-25433

9321-F-25443

93 21-F-25453

9321-F-25463

93 21-F-25473

9321-F-25483

93 21-F-25493

9 321-F-25503

93 21-F-25513

9 321-F-25523

9321-F-25573

9 321-F-25583

93 21-F-25593

9 321-F-25603

93 21-F-25613

9321-F-25623

93 21-F-25633

9321-F-25643

93 21-F-25653

9321-F-25663

93 21-F-25673

9 321-F-25683

9321-F-25693

9 321-F-25703

93 21-F-25713

CONT

CONT

CONT

CONT

CONT

CONT


TITLE


PAB COMPOSIT PPG ARRT'T EL 55'-0 SlT #4

PAB DEMIN PPG

PAB DEMIN PPG

FUEL STORAGE BLDG AUX COOLANT SYS

FUEL STORAGE BLDG AUX COOLANT SYS

PAB DEMIN PPG

PAB DEMIN PPG

PAB DEMIN PPG

CONT BLDG CVCS

CONT BLDG CVCS

CONT BLDG CVCS

CONT BLDG CVCS

PAB COMP PPG ARRG'T IN RHR ROOM

PAB COMP PPG ARRG'T EL 73'-0




PAB COMP PPG ARRG'T AT RHR PUMPS

PAB COMP PPG ARRG'T IN PIPING BAY

PAB COMP PPG ARRG'T IN PIPING BAY

PAB DEMIN PPG

PAB GAS DECAY TANK PIPING

PAB GAS DECAY TANK PIPING

SAMPLING SYS PPG

VENT & DRAIN DETAILS FOR NUCLEAR PIPING

DRAWING NO.

9321-F-25723

9 321-F-25733

9321-F-25743

9321-F-25753

93 21-F-25763

9 321-F-25773

93 21-F-25783

9321-F-25793

9321-F-25803

9321-F-25813

9321-F-25823

9321-F-25833

93 21-F-25843

9 321-F-25853

9321-F-25863

9321-F-25873

93 21-F-25883

9321-F-25893

93 21-F-25903

9 321-F-25913

93 21-F-25923

9321-F-25933

93 21-F-25943

9321-F-25953

93 21-F-25973

9 321-F-25993

8-16

REV.


TITLE

BLDG LOCATION OF PIPE SLEEVES

BLDG PRESS. RELIEF VALVE DRAIN PPG

DIAG. TAGGED INSULATION FOR ISI

BUILDING SAFETY INJECTION SYS MODIFICATIONS

BLDG SI PPG FROM SUMP TO EX PUMP SUCTION

CONT

CONT

FLOW

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

ARRG'T IN

ARRG'T IN

FARM COMP

FARM COMP

FARM COMP

FARM COMP

FILTER ROOM

FILTER ROOM

PPG SHT NO.

PPG SHT NO.

PPG SHT NO.

PPG SHT NO.

PAB COMP PPG

PAB COMP PPG

NUCLEAR TANK

NUCLEAR TANK

NUCLEAR TANK

NUCLEAR TANK

PAB COMP PPG ARRG'T AT EL 73'-0

PAB COMP PPG ARRG'T AT EL 73'-0

ARRG'T OF ENCLOSURE TANK FOR SIS SUMP LINE VALVE

DRAWING NO.

93 21-F-26033

9 321-F-26043

9321-F-26053

9321-F-26063

9321-F-26133

9 321-F-26143

93 21-F-26153

9321-F-26163

9321-F-26173

9321-F-26223

9321-F-26233

9 321-F-26243

93 21-F-26253

9321-F-26263

93 21-F-26273

9 321-F-26293

93 21-F-26303

9 321-F-26313

9321-F-26323

9 321-F-26333

93 21-F-26343

9 321-F-26363

93 21-F-26373

9321-F-26423

8-17

REV.

SAFETY INJECTIONS SYS




CONT SPRAY SYS (SOUTH 1/2)

CONT SPRAY SYS (NORTH 1/2)

CONT SPRAY SYS SECTIONS

CONT SPRAY SYS SECTIONS

SAFETY INJECTION SYS SECTIONS

PPG COMP SECTIONS

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG


TITLE

CONT BLDG WASTE DISPOSAL SYS (NORTH 1/2)

CONT BLDG WASTE DISPOSAL SYS (SOUTH 1/2)

VALVE LEAKOFF PPG

VALVE LEAKOFF PPG

HOLDUP TANK PIT & PIPE TRENCH FLOOR & HUB DRAIN

CONT BLDG SERV WTR HEADER DRAINS

CONT BLDG SERV WTR HEADER DRAINS

GROSS FAILED FUEL DETECTOR PPG SYS

FLOW DIAG POST ACCIDENT CONT SAMPLING SYS

HOLDUP TANK PIT COMP PPG

HOLDUP TANK PIT COMP PPG

PIPE WHIP RESTRAINTS SHT








PAB COMP PPG ARRG'T AT SIS PUMPS

PAB COMP PPG ARRG'T AT SIS PUMPS

PAB COMP PPG ARRG'T AT EL 15'-0 FOR SPENT RESIN

CHEMICAL DRAIN & SUMP TANKS

PAB COMPOSITE PPG ARRG'T IN RAD. TUNNELS


DRAWING NO.

9321-F-26453

9321-F-26463

93 21-F-26473

9321-F-26483

93 21-F-26493

9 321-F-26503

93 21-F-26 513

9 321-F-26523

93 21-F-26533

9321-F-26553

9321-F-26563

9 321-F-26573

93 21-F-26583

9 321-F-26593

9321-F-26603

9 321-F-26613

93 21-F-26623

9 321-F-26633

93 21-F-26643

9321-F-26663

93 21-F-266 73

9321-F-26683

9321-F-26703

9321-F-26713

8-18

REV.


TITLE



TRENCH AREA PIPE WHIP RESTRAINTS

COMPOSITE PPG IN TRENCHES




PAB COMP PPG ARRG'T AT CONT SPRAY PUMPS & ADDITIVE

TANK AT EL 41'-0

PAB COMP PPG AT MONITOR TANK PUMPS & PPG WTR PUMPS

PAB SERV AIR & CITY WTR PPG

PAB SERV AIR & CITY WTR PPG

CONT BLDG POST ACCIDENT CONT AIR SAMPLING SYS

CONT BLDG SERVICE AIR, CITY WATER PIPING

CONT BLDG DEMISTER & COOLING COIL DRAIN PIPING

CONT BLDG DEMISTER & COOLING COIL DRAIN PIPING

CONT BLDG SERVICE WATER PIPING




FUEL STORAGE BLDG SERVICE AIR & CITY WTR PPG

PAB FLOOR & HUB DRAINS EL 15'-0, 34'-0, 41'-0

PAB FLOOR & HUB DRAINS EL 53'-0, 73'-0

CONT BLDG HYDROGEN RJECOMBINER PIPING

SERV WTR IN YARD AREA

SERV WTR PPG IN YARD AREA

SERV WTR PPG IN YARD AREA

SERVICE WATER PIPING BACKUP SUPPLY

DRAWING NO.

93 21-F-26723

9321-F-26733

93 21-F-26753

9321-F-26763

9321-F-26773

9321-F-26783

9321-F-26793

9321-F-26823

9321-F-26833

93 21-F-26843

9321-F-26853

93 21-F-26863

9321-F-26873

93 21-F-26883

9321-F-26893

93 21-F-26903

9 321-F-26913

93 21-F-26923

9 321-F-26933

93 21-F-26953

9321-F-26963

93 21-F-26973

9321-F-26983

93 21-F-27003

9321-F-27013

9321-F-27023

9 321-F-27033

8-19

REV.


TITLE

NUCLEAR EQUIPMENT HYDROGEN PIPING

CONT BLDG IN-CORE INSTR. PPG SUPPORTS G.A.

CONT BLDG IN-CORE INSTR. PPG SUPPORTS DETAILS

CONT BLDG CHARCOAL FILTER SPRAY PIPING

CONT BLDG CHARCOAL FILTER SPRAY PIPING

CONT BLDG PIPING PENETRATIONS DETAILS OF HOT LINES

CONT BLDG PIPING PENETRATIONS DETAILS OF COLD LINES

CONT BLDG PIPING PENETRATIONS DETIALS OF PURGE LINES

SPARE CONNECTIONS

CONT BLDG PPG PENET. DETAILS OF FUEL TRANSFER

CONT

CONT

CONT

FLOW

FLOW

FLOW

FLOW

FLOW

F.D.

FLOW

F.D.

F.D.

F.D.

F.D.

F.D.

FLOW

FLOW

BLDG ARRG'T OF PIPING PENETRATIONS

BLDG PIPING PENETRATIONS SCHEDULE

BLDG PIPING PENETRATIONS SCHEDULE

DIAG WASTE DISPOSAL SYS SHT NO. 1

DIAG AUX COOLANT SYS

DIAG AIR COOLING SYS FOR HOT PENETRATIONS

DIAG SERV WTR SYS NUCLEAR STM SUPPLY PLANT

DIAG NITROGEN TO NUCLEAR EQUIPMENT

PRIM M.V. WTR SYS NUC. STM SUPPLY PLANT

DIAG AUTO. GAS ANALYZER SYS.

PENET & LINER WELD JOINT CHANNEL PRESS. SYS

AUX STM & CONDENSATE FOR NUC. EQUIP.

NUCLEAR EQUIPMENT DRAINS

STM GENERATOR BLOWDOWN SYS.

WASTE DISPOSAL SYS SH #2

DIAG PPG AT REACTOR COOLANT PUMPS

DIAG SAFETY INJECTION SYS

DRAWING NO.

9321-F-27053

9321-F-27063

9321-F-27073

9321-F-27093

9321-F-27013

9321-F-27123

9321-F-27133

9321-F-27143

9321-F-27153

9321-F-27163

9321-F-27173

9321-F-27183

9321-F-27193

9321-F-27203

9321-F-27213

9321-F-27223

9321-F-27233

9321-F-27243

9321-F-27253

9321-F-27263

9321-F-27273

9321-F-27283

9321-F-27293

9321-F-27303

9321-F-27343

9321-F-27353

8-20

REV.


TITLE

F.D. CHEM & VOLUME CONTROL SYS.

F.D. CHEM & VOLUME CONTROL SYS.

F.D. REACTOR COOLANT SYSTEM

HOT PENETRATIONS COOLING SYS PIPING



FLOW DIAG. SAMPLING SYS

F.D. ISOLATION VALVE SEAL WATER SYS.

F.D. REACTOR COOLANT SYS SH #2

CONT BLDG-SERV WTR PPG FOR RECIRC FAN

CONT BLDG-SERV WTR PPG FOR RECIRC FAN

DIAG

DIAG

DIAG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

ISOLATION

ISOLATION

MOTOR COOLERS

MOTOR COOLERS

SAFETY INJECTION SYS SH #2

AUX COOLANT SYS

HYDROGEN RECOMBINER SYS

MN STM & BLR FD PIP RESTRS ARRG'T SH #6

ARRG'T PIP RESTRS MW STM & BLR FD SH #1


ARRG'T PIPRESTRS MW STM & BLR FD SH #1


ARRG'T PIP RESTRS PRESSURIZER SURGE LL

MAIN STM & BLR FO PIP RESTRS DET'S SH 1

BLR FD PIPE RESTRS DET'S SH 7

MAIN STM & BLR FD PIPE RESTR'S DET'S SH 2

VALVE SEAL WTR PIPING SH #1

VALVE SEAL WTR PIPING SH #2

DRAWING NO.

9321-F-27363

9321-F-27373

9321-F-27383

9321-F-27423

9321-F-27433

9321-F-27443

9321-F-27453

9321-F-27463

9321-F-27473

9321-F-27483

9321-F-27493

9 321-F-27503

93 21-F-27513

9321-F-27533

9321-F-27543

9321-F-27583

9321-F-27593

9321-F-27603

9321-F-27613

9321-F-27633

9321-F-27643

9321-F-27663

9321-F-27673

9321-F-27683

9321-F-27693

8-21

REV.

FLOW

FLOW

FLOW

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT


TITLE

MN STM & BLR FD PIPE RESTR'S DET'S SH 3



MN STM CABLE RESTR'S SH #2

BLDG LEAK RATE TEST & PRESS SYS.

LEAK RATE TEST ARRG'T OF EQUIP, & PIPING

CONT

CONT

CONT

CONT

F.D.

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

CONT

BLDG

BLDG

BLDG

BLDG

CONT

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

ANNULUS

ANNULUS

ANNULUS

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG

PPG

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

COMPOSITE

(SE

(SE

(SE

( SE

(NE

(NE

(NW

(NW

(SW

(SW

(NE

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

QUAD)

TURBING ARRG'T REACTOR COOLANT LOOPS l&2

BOUNDARY CHECK VALVE TESTING SH #1

TURBING ARRG'T REACTOR COOLANT LOOPS 3&4

BOUNDARY CHECK VALVE TESTING SH #2

REACTOR COOLANT SYS BOUNDARY CHECK VALVE TESTING

SH #1

REACTOR COOLANT SYS BOUND CHK VA TESTING SHT #2

REACTOR COOLANT SYS BOUND CHK VA TESTING SHT #3

CABLE SCHEMATIC MAIN POWER GENERATOR,

TRANSFORMERS AND TIE-LINES

DRAWING NO.

9321-F-27703

9321-F-27713

9321-F-27763

9321-F-27773

9321-F-27783

9321-F-27793

9321-F-27803

9321-F-27813

9321-F-27823

9321-F-27833

9321-F-27843

9321-F-27853

9321-F-27863

9321-F-27873

9321-F-27883

9321-F-27893

9321-F-27903

9321-F-28043

9321-F-28053

9321-F-28063

9321-F-28073

9321-F-28083

9321-F-30033

8-22

REV.

ANNULUS

ANNULUS

ANNULUS

ANNULUS

ANNULU S

ANNULUS

ANNULUS SM.

ANNULUS SM.

CONT BLDG


TITLE

SINGLE LINE DIAG 480V MCC SHT #1

SINGLE LINE DIAG 480V MOTOR CONTROL CENTERS

SINGLE LINE DIAG 480V MCC & INSTR BUSES

THREE LINE DIAGRAM LOW VOLTAGE

SINGLE LINE DIAG D-C SYSTEM

BLOCK DIAG. 3 LINE DIAG-LOW VOLTAGE

MAIN THREE LINE DIAGRAM

GROUNDING LAYOUT FUEL STORAGE BLDG & WASTE

HOLD-UP TANK PIT

LIGHTING & P.A. SYS T/B EL 36'-9

LIGHTING & P.A. SYS T/B EL 53'-0

AUXS BOIL FD PUMP AREA LIGHTING

LIGHTING & PA SYS CONTAIN BLDG EL 46'-0 & 68'-0

LIGHTING & PA SYS CONTAIN BLDG EL 95'-0

LIGHTING & PA SYS PAB EL 41'-0, 34'-0 & 15'-0

LIGHTING PANELS & CIRCUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

LAYOUT

LAYOUT

LAYOUT

LAYOUT

LAYOUT

LAYOUT

LAYOUT

LAYOUT

CONDUIT LAYOUT

TURBO-GEN

TURBO-GEN

TURBO-GE N

TURBO-GEN

TURBO-GE N

TURBO-GEN

TURBO-GEN

DIAG SHT #1

& HTR BAY EXT SHT #1



& HTR BAY EL 36'-9 SHT #1

& HTR BAY EL 36'-9 SlT #2

& HTR BAY EL 36'-9 SHT #3

BAY EXT EL 15'-0 & 36'-9

FUEL STORAGE BLDG

& DET'S CATHODIC PROT INTAKE STR

DRAWING NO.

9321-F-30043

9321-F-30053

9321-F-30063

9321-F-30073

9321-F-30083

9321-F-30093

9321-F-30113

9321-F-30163

9321-F-30233

9321-F-30243

9321-F-30283

9321-F-30293

9 321-F-30303

93 21-F-30353

9321-F-30413

9321-F-30683

9321-F-30693

9321-F-30703

9321-F-30713

9321-F-30723

9321-F-30733

9321-F-30743

9321-F-30803

9321-F-30873

8-23

REV.


TITLE

CHLORINATION-ARRGT & CONDUIT LAYOUT

CONDUIT LAYOUT INTAKE STRUCTURE SHT #3

CONDUIT DETAILS SHT #4

CONTROL ROOM OPERATING DEST

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

CONDUIT

DETAILS SHT #1

DETAILS MANHOLES #31 & 32

DETAILS SHT #2

DETAILS SHT #3

DETAILS LAYOUT INTAKE STRUCTURE SHT #2

LAYOUT AUX BOIL FEED PUMP AREA SHT #1

DETAILS WASTE HOLD-UP TANK PIT

DETAILS FUEL TRANS SYS FUEL STOR & V.C.

DETAILS TRANSFER YARD AREA

CONTROL BLDG LOC OF SLEEVES &

CONTROL BLDG LOC OF SLEEVES &

CONDUIT DET MANHOLES 35 & 36

SCHEM DIAG 69KV SWGR 31

SCHEM DIAG 69KV SWGR 32

CONDUIT DET MANAHOLE 34

SCHEM DIAG 480V SWGR 31


CONDUIT DET MANHOLES 31A,


SCHEM DIAG 480V MCC 32






OPNGS

OPNGS

SHT #1

SHT #2

31B, & 33

DRAWING NO.

9321-F-30883

9321-F-30903

9321-F-30913

9321-F-30923

93 21-F-30933

9 321-F-30943

93 21-F-30953

9321-F-30963

9321-F-30983

9321-F-31003

9321-F-31023

9321-F-31063

9321-F-31073

9321-F-31103

9321-F-31113

9321-F-31123

9321-F-31133

9321-F-31143

9321-F-31153

9321-F-31173

9321-F-31183

9321-F-31203

9321-F-31213

9321-F-31223

9321-F-31233

9321-F-31243

9321-F-31253

9321-F-31263

9321-F-31273

8-24

REV.


TITLE

SCHEM

SCHEM

SCHEM

SCHEM

SCHEM

CABLE

DIAG 480V MCC 38

DIAG 480V MCC 39

DIAG MISC SOL VALVES

DIAG DIESEL GEN AUX

DIAG MISC A-C CIRCUITS

SCHEM-SOL VALVES

CONDUIT LAYOUT LEAK RATE TEST CONT BLDG

SCHEM DIAG MISC D-D CIRCUITS

CABLE ARRGT REACTOR HEAD SHT #2

CONDUIT DETAILS MANHOLE 37

LHTG & TRAY PLANS & SECT BRIDGE EL 53'-0 CONN UNITS

CONDUIT DETAILS MANHOLE 38

WIRING DIAG-118V AC INSTR BUS PANEL 31 & 32

WIRING DIAG-118V AC INSTR BUS PANEL 33 & 34

WIRING DIAG-125V DC DISTR PANELS 31, 32, 33 & 34

WIRING DIAG-DIESEL GEN'S 31, 32, & 33

CABLE ARRGT REACTOR HD SHT #3

CONDUIT LAYOUT CONTR RM AIR CONDITIONING

CONDUIT LAYOUT DIESEL GENERATOR BLDG

BATTERY #33 ARRGT & DETS

CONDUIT CONN SCH CONTROL CONDUITS FOR 480V S.G.

CONDUIT LAYOUT FOR 480V SWGR CONTROL CABLES

CONDUIT LAYOUT RADIATION SHIELD DOORS PAB STOR AREA

TERMINAL BOX XE1 ROD CONTROL LEADS

TERMINAL BOX XD9 ROD CONTROL LEADS

CONDUIT & TRAY CONN SCH CNTRL BLDG SHT #1



CABLE TRAY SUPPORT DETAILS

DRAWING NO.

93 21-F-31283

9 321-F-31293

93 21-F-31313

9321-F-31333

93 21-F-31373

9321-F-31383

93 21-F-31393

9 321-F-31403

93 21-F-31433

9 321-F-31463

9321-F-31613

9 321-F-31623

93 21-F-31993

9321-F-32003

93 21-F-320 73

9 321-F-32203

93 21-F-32453

9321-F-32673

93 21-F-32703

9321-F-32733

93 21-F-32783

9321-F-32793

93 21-F-32803

9 321-F-32813

93 21-F-32823

9321-F-32883

93 21-F-32893

9321-F-32903

93 21-F-32953

8-25

REV.


TITLE

LAYOUT AUX BOILER FEED PUMP AREA

& TRAY CONN SCH TURBO GEN BLDG EL 15'-0 SHT #1



BAY & CONTR BLDG ROOF DRNS PLAN

BAY & CONTR BLDG ROOF DRNS SECT & DETAIL

BAY FLOOR & HUB DRNS PLAN EL 15'-0

BAY FLOOR & HUB DRNS PLAN EL 36'-0 SECT & DET

T/B FIRE PROTECTION-STANDPIPE PLANS

T/B FIRE PROTECTION-STANDPIPE SECT & DUCT

T/B FIRE PROTECTION-STANDPIPE SECTIONS

MISAC DRN PLT AREA PLANS, SECT & DET

MISC BLDG'S ROOF DRAINS

POTABLE CITY WATER

CNTL BLDG EL 15'-0 AIR CONDG EQUIP RM PLANS

T/B HTR BAY HEATING PLANS

T/B HTR BAY HEATING PLANS

T/B VENTILATION PLAN, SECTIONS, DETAILS

T/B HTR BAY HEATING PLANS & SECT

F.D/ VENT SYS-CONTAINA, PAB & FUEL STOR BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

BLDG

RECIRC

RECIRC

RECIRC

RECIRC

AIR RECIRC

AIR RECIRC

BLDG AIR

BLDG AIR

RECIRC SYS

ABOVE EL 95'-0

ABOVE EL 85'-0

ABOVE EL 68'-0

SECTIONS A-A

SECTIONS B-B

SECTIONS

HANGERS & SUPPORTS

RECIRC SYS HANGERS & SUPPORTS

CONT BLDG EMERGENCY SHOWER & EYE WASH PIPING

CONDUIT

CONDUIT

CONDUIT

CONDUIT

T/B HTR

T/B HTR

T/B HTR

T/B HTR

8-26

DRAWING NO.

9321-F-32963

9321-F-33273

9321-F-33283

9321-F-33293

9321-F-40013

9321-F-40023

93 21-F-40033

9321-F-40043

93 21-F-40083

9321-F-40093

9321-F-40103

9321-F-40113

9321-F-40123

9321-F-40153

93 21-F-40173

9 321-F-40183

93 21-F-40193

9321-F-40203

9321-F-40213

9321-F-40223

9321-F-40233

9321-F-40243

9321-F-40253

9321-F-40273

93 21-F-40283

9 321-F-40293

93 21-F-40303

9321-F-40313

9321-F-40323

REV.

CON T

CONT

CONT

CONT

CONT

CONT

CONT

CONT


TITLE

FAN HOUSE PAB, PSB & CB XH., PURGE & DILUTION

FANS SECT.

PAB H&V PLANS ELEV 15'-0, 34'-0 & 41'-0

PAB H&V PLANS ELEV 55'-0, 73'-0

PAB H&V SECTIONS

PAB H&V SECTIONS

PAB H&V SECTIONS

FUEL STORAGE BLDG HEATING & VENTILATION PLANS

FUEL STORAGE BLDG H&V SECTIONS & DETAILS

MISC PLANT AREAS & BLDG VENTILATION SYSTE21S

AUX FEED PMP BLDG ROOF, FL & HUB DRNS PLANS & SECT.

DIESEL GENERATOR BLDG HEATING & VENTILATION

DIESEL GEN. BLDG FIRE PROTECTION-SPRAY SYS

PLANT STM HEATING STM & COND. DISTRIBUTION HEADERS



ELEC. TUNNEL, MCB & D.G. BLDGS VENT.-SCHEMATIC

ELEC. TUNNEL VENTILATION EQUIP ROOM (EL 65'-o)

CONTROL BLDG VENTILATION FAN ROOM (El 27'-0)

FLOW DIAG AUX STM SUPPLY & COND RETURN SYS.

PAB HEATING & VENTILATION

CONTROL ROOM (EL 53'-0) AIR CONDITIONING

T/B ELEVATOR MACH. RM. & LUBE OIL STRG RM.

EMERG. SHOWER & EYE WASH PPG-T/B, PAB, & AFPB

DRAWING NO.

9321-F-40343

9321-F-40363

9321-F-40373

9321-F-40383

9321-F-40393

9321-F-40403

9321-F-40413

9321-F-40423

9321-F-40433

9321-F-40473

93 21-F-40483

9321-F-40493

93 21-F-40513

9321-F-40523

9321-F-40533

9321-F-40543

9321-F-40553

9321-F-40563

9321-F-40573

9 321-F-40583

93 21-F-40593

9321-F-40613

9321-F-40643

8-27

REV.


TITLE

CONTROL BLDG EL 15'-0 & EL 33'-0 HEATING & VENT

CONTROL BLDG AIR COND EQUIP RM., CLG WTR, STM & COND

ELEC TUNNEL FIRE PROTECTION SPRAY SYS SECT.

HEATED ENCLOSURE FOR CHEM. PMPS

SHIELD WALL AREA ENCLOSURE FOR M.S. & BF PIPES HTG

SHIELD WALL AREA ENCLOSURE FOR M.S. & BF PIPES HTG

ROOF DRAIN

SHIELD WALL AREA ENCLOSURE FOR M.S., & BF PIPES HTG

ROOF VENT

AUX STM & COND. LINES BETWEEN UNIT 1 & UNIT 3

POST ACCIDENT CONT. VENTING SYSTEM

FUEL STORAGE BLDG EMERGENCY EXHAUST SYS.

FAN HOUSE PAB EXH & CB PURGE FILTER SYS.

FAN HOUSE PAB EXH & CB PURGE FILTER SYS.

FAN HOUSE CB PRESSURE RELIEF FILTER SYS.

FAN HOUSE PAB EXH, CB PURGE & CB PRESS RELIEF

FILTER SYS.

FIRE PROT. SYS. PAB EXH, CB PURGE & CB PRESS

RELIEF FILTER SYS.

FIRE PROT. SYS. PAB EXH, CB PURGE & CB PRESS

RELIEF FILTER SYS.

DRAINS PAB EXH, CB PURGE & CB PRESS RELIEF

FILTER SYS

DRAINS PAB EXH, CB PURGE & CB PRESS RELIEF

FILTER SYS

RESTRAINT & SUPPORT DESIGN STD PROCEDURES

DRAWING NO.

9321-F-40653

9321-F-40683

9321-F-40703

9321-F-40733

9321-F-40753

9321-F-40763

9321-F-40773

9321-F-40783

9321-F-40793

9321-F-40803

93 21-F-40823

9321-F-40833

93 21-F-40843

9321-F-40853

9321-F-40863

9321-F-40873

9321-F-40883

9321-F-40893

9321-F-50003

8-28

REV.


TITLE

INSTR AIR SUPPLY RESTRAINT & SUPPT DES SHT #8

CONT BLDG RESTRAINT & SUPPT DESIGN LINE 71

PAB RESTRAINT & SUPPORT DESIGN

CONT BLDG RESTRAINT & SUPPORT DESIGN

CONT BLDG RESTRAINT & SUPPORT DESIGN LINE 70

CONT BLDG RESTRAINT &



SUPPORT DESIGN

SUPPORT DESIGN

SUPPORT DESIGN

TRANSMITTER RACKS PPG ARRG'T-SH #1 INSTRUMENTATION



CABLED TUBING ARRG'T SH #1 INSTRUMENTATION


CABLED TUBING JUNCTION BOX ARRG'T INSTRUMENTATION

INSTR

INSTR

INSTR

INSTR

INSTR

INSTR

AIR SUPPLY SH #1 INSTRUMENTATION

AIR SUPPLY SH #2 INSTRUMENTATION

PIPING SCHEMATICS SH #1 INSTRUMENTATION




INSTR PIPING SCHEMATICS SH #5 INSTRUMENTATION

INSTR PIPING SCHEMATICS SH #6 INSTRUMENTATION


DRAWING NO.

9321-F-50633

9321-F-51313

93 21-F-52593

9 321-F-53083

93 21-F-55173

9321-F-55783

93 21-F-55793

9 321-F-55803

93 21-F-70023

9321-F-70033

9321-F-70043

9321-F-70053

9321-F-70063

9321-F-70073

93 21-F-70083

9321-F-70093

9321-F-70103

9321-F-70113

9321-F-70123

9321-F-70133

9321-F-70143

9321-F-70153

9321-F-70163

8-29

REV.


TITLE

LEVEL CONTROL PPG SH #1 INSTRUMENTATION

LEVEL CONTROL PPG SH #2 INSTRUMENTATION

PENET & LINER WELD JOINT CHANNEL PRESS PPG-INSTR

STM & WTR NALYSIS SY. SAMPLING DIA INSTRUMENTATION

STM & WTR ANALYSIS SYS. SAMPLE PPG Sl #1

INSTRUMENTATION

EXTR. STM FREE FLOW REVERSE CURRENT VALVES CONTR. SYS

LOCAL MOUNTED INSTR. SUPPORT DETAILS INSTRUMENTATION

PRESSURE GAGE & SWITCH DETAILS INSTRUMENTATION

PRIM. PLANT INSTR. PPG & SUPPORTS SH #1

INSTRUMENTATION

PRIM. PLANT INSTR. PPG & SUPPORTS SH #2

INSTRUMENTATION

CONT BLDG INSTR. ARRG'T SH #1 INSTRUMENTATION

CONT BLDG INSTR. ARRG'T SH #2 INSTRUMENTATION

WASTE HOLD-UP PIT & NUC. TANK FARM ARRG'T

INSTRUMENTATION

CABLED TUBING SCHEMATIC INSTRUMENTATION

AUX BOILER FEED PUMP ROOM INST PPG SH #1

INSTRUMENTATION

TRAVELING SCREENS DIFFERENTIAL LEVEL CONTROL

SYS INSTRUMENTATION

PENET. & LINER WELD JOINT CHANNEL PRESS PPG SH #1

INSTRUMENTATION


INSTRUMENTATION


INSTRUMENTAT ION

DRAWING NO.

9321-F-70173

9321-F-70183

93 21-F-70193

9321-F-70203

9321-F-70213

9321-F-70223

93 21-F-70233

9321-F-70243

9321-F-70253

9321-F-70263

9321-F-70273

9321-F-70283

9321-F-70293

9321-F-70303

9321-F-70313

9321-F-70323

9321-F-70333

9321-F-70343

9321-F-70353

8-30

REV.


TITLE

LEVEL CONTROL PIPING SH #3 INSTRUMENTATION



INSTRUMENTATION

PAB INSTR ARRG'T SH #1 INSTRUMENTATION



CONT BLDG PPG PENET TRENCH INSTR ARRG'T

STM & WTR ANALYSIS SYS SAMPLE PPG SH #2

INS TRUMENTAT ION

RADIATION MONITORING INSTALLATION DETAILS

INSTRUMENTATION


VARIABLE WEIR CONTROL PPG ARRG'T INSTRUMENTATION

INSTR. PPG SCHEMATICS SH #7 INSTRUMENTATION



TRANSMITTER RACKS PPG ARRG'T SH #4 INSTRUMENTATION

PENET & LINER WELD JOINT CHANNEL PRESS PPG SH #5

INSTRUMENTAT ION

AUX BOILER FEED PUMP ROOM INSTR PPG SH #2


NUCLEAR PLANT CONTROL VALVE HOOK-UP DETAILS

INS TRUMENTAT ION



DRAWING NO.

9321-F-70373

9321-F-70383

93 21-F-70393

9321-F-70403

9321-F-70413

9 321-F-70423

93 21-F-70433

9 321-F-70443

9321-F-70453

9321-F-70463

93 21-F-70473

9321-F-70483

93 21-F-70493

9321-F-70503

9321-F-70513

9321-F-70523

9321-F-70533

93 21-F-70553

9321-F-70563

9321-F-70573

93 21-F-70583

8-31

REV.


TITLE

CONT BLDG TEST PRIM SENSOR LOCATION INSTRUMENTATION


STARTUP FEEDWATER BYPASS PIT RACK INSTRUMENTATION

CONT BLDG INSTR ARRG'T SH #i1 INSTRUMENTATION

CONT BLDG INSTR ARRG'T SH #2 INSTRUMENTATION

AUX BOILER FEED PUMP ROOM INSTR PPG-SH #2

INSTRUMENTATION

RAD MONITORING INSTALLATION DETAILS-IODINE 131

MONITOR INSTRUMENTATION

PAG G.A. ELEV 55'-0 INSTRUMENTATION

FUEL STORAGE BLDG G.A. INSTRUMENTATION

STM GENERATOR BLOWDOWN SAMPLE PANEL INSTRUMENTATION

EQUIPMENT ARRG'T CONTROL BLDG-SH #1 INSTRUMENTATION

DRAWING NO.

93 21-F-70603

9321-F-70613

9321-F-70623

9321-F-70633

93 21-F-70643

9321-F-70653

9321-F-70683

9321-F-70693

93 21-F-70703

9321-F-70713

93 21-F-70723

8-32

REV.

8.3 Specification List

Pipe Support Design Guides Cable and Tray Support Criteria Electrical Workmanship Spec

Steam Generator Reactor Coolant Pumps and Vessel Ring Support

Roughing, HEPA and Carbon Filters

Electrical Unit Heaters

Plant Heating, Ventilating and Air Conditioning Systems

Containment Building Equipment Hatch and Containment Liner, Personnel Locks

Auxiliary Feed Pump and Turbine Drive

Piping Specifications

Hanger, Anchors and Supports

Service Water Spec

Overhead Cranes and Polar Cranes

8.4 Equipment Specification

Containment Spray Pumps E Spec

Reactor Containment Fan Cooling Systems Containment Centrifugal Fans E Spec

#EI-6009

#93 21-05-12-10

#9321-05-45-12

#9321-05-45-20

#9321-05-45-24

#93 21-05-225-1

#9321-05-238-1

#9321-05-248-18

#9321-05-248-27

#9321-05-248-35

#9321-05-257-1

#676428

#677113

8-33

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- PIZESSURIZEe AUXILIARY SPEW V (WlO 2C5 !,2EAYS

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-I OP Z R-R PUMPS (LARRE LOCA) _FLOUJ PAW FOR. MAKEUP UMgAER. -NIURA.L CIRCULAIONa -IOF5 PUMPS 0oF4 S/'S(LOSS

OF NORMXl. FW) PEQUIRES PO ER OPERTED A,'TOSPHENPIC ZELIEF VALVES.

-POJER OPERATED ATMOSPHERIC RELIEF VALVES USED IN COWJUIJCTIOQ NH AUK U-.EEDA ER_ EMERGENCf BORATION (CrMIC L REACTIVITY CONTROL)

-INJECIOWl OF 50R.O(CHEMICAL REACTIVITY COnTrOL- ecs oz SECOWDAZY -YSIEM FEEAX.S )

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U) I5OLAIE AW LINE RUPITUE 1 ) SAFEIY tNJECIOKI SYSIE- -CHARGING LINE CHECK VALVES OZ LEA&AAC-E4 MAIKIT&dl -AUX SPRAY CHECK VALVE

R_ PRE5!URE -EXCESS LETDOWN ISOLATION .)CHEMICALVoLUmE 6,COUi..oL VALVES 213A ( 213Be

SYSTEU -LETDOWN ISOLATION VALVES LCV-4S9 ( LCV-4.O -CHECK VALVES ON SAFElY

5)IeACTOR. COOLAUT SYSTEM INJECTION LINES -POWER OPERATED RELIEF VALVES PCV-455C OR 456

-PRESSURIZER CODE SAFETY VALVES

IEDASCO SERVICIES INCORPORATED POWJF - AI, l N01ITY STATE Or lFW yol t vl ! Qm to KE.L W"A.10-9 1 DIAWI rOW1J IO.5 W, Pp

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CLOSEDSYSTEM IMSIDE COWAb.N

8) EDUCE CONJTAIINJEWT 0 COWNjTAIMEWT SpehvY SYSTEM 01?-1IOF z CS PUMPS a 1 OF 5 FAkw PRESSU2E t TEVPEIATURE COWTNIMEWI -ECIP-CULAIOm COOLE-.5 OQ0,

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2.) PP SYS'TEM AVJD/OR -I .ECIPCULATE- SPP Ay AUO EUPTUMP SAFE'[1 IJJECTIOW SYSIEM LIQUID AF'TEE EUPTYI IW THE RtWT

-IOFZ PECC PUMPS OZ Ior- z eH PUMVP

-IO- 5 H-IGH PPESSURE- PUUPS (LARGE LOCA, OW.LY)

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ELECTRICAL TRAY SYSTEM (SEE AD-IS)

-PROVIDES CABLE ROUTING & SUPPORT

ELECTRICAL DISTRIBUTION SYSTEM (SEE AD-22) I

INTERRUPTS POWER TO THE MOTORGENERATOR SETS

/

OPERATOR ACTION

-PROVIDES MANUAL REACTOR TRIP

:) : I

ROL

CONTROL ROOM VENTILATION (SEE AD-7)

- CONTROL ROOM COOLING

-SENDS SIGNAL TO TRIP REACTOR & DROP CONTROL RODS

EBASCO SERVICES INCORPORATED J AT..HOZITY STATE OF ,JE( Y012K DIV. .- DR..E APPROVED IkJD w P0UT 4O5 WP ADDATE] CH.A ,. ' I SCALE 14-' i7f . AU A. DI, W-9A.4 I

+

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REACTOR PROTECTION SYSTEM (SEE AD-17)

ELECTRICAL TRAY SYSTEM, CAD-I8)


-PROVIDES ELEC POWER

REACTOR PROTECTION SYSTEM

(SEE AD-I7) O PER

PROVIDES SAFETY /!ACTIC INJECTION SIGNAL MANU

OPERi

ATOR )N

ATION

COMPONENT COOLING SYSTEM (SEE AD- 9)

-PROVIDES COOLING FOR PUMPS & HEAT EXCHANGERS

RESIDUAL HEAT N _ REMOVAL

I SYSTEM(SEE AD-4)'

-HEAT EXCHANGER -,-LOW PRESSURE PUM

-CONTROL ROOM VENTILATION (SEE AD-7)

-CONTROL ROOM COOLING

PS

DC POWER SYSTEM (SEE AD-12)

-PROVIDE CONTROL POWER

EBASCO SERVICES INCORPORATED POWER AU"THORITY G7ATE OF NEW'YORK DIV.i DR.N APPROVED IN POINT NO.l!N..PP

DI. D. A Sk, rET'Y INJECTION 6YSTEM AD-2 DATE r- AUILIA' DI4,,GA SCA LE tJ o e - 0 119 '# A

+*

ELECTRICAL DISTRIBUTION SYSTEM (SEE AD-22)

SAFETY INJECTIO SYSTEM

g

CONTROL ROOM VENTILATION

(SEE AD-7)


COMPONENT COOLING SYSTEM

(SEE AD-9)

%-FLUID DRIVE COOLERS

CHEMICAL VOLUME I CCONTPflI

TEM

ELECTRICAL DISTRI BUTION SYSTEM (SEE AD 22)

PROVIDES ELECTRICAL POWER

ELECTRICAL TRAV7 SYSTEM (AD-I )

-PROVIDE ROUTING SUPPORT

CABLE &

DC POWER SYSTEM

(SEE AD-12)


OPERATOR ACTION

- MANUAL OPERATION

EBASCO SERVICES INCORPORATED PDWER AUT<ORITy &TATE OF NEW YORK

DIV.* E (: DR.V.'Z- APPROVED INDN NPO0NT NO."!N.RRA AHEMICkL VOLI)ME CON7ROL , A D-3 DATE 1-- CFAL-F A'',. -UXILIA ,, ./4,C ad.i

SCALE tJ " a .1 1%1 V-, I

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ELECTRICAL TRAY SYSTEM (SEE AD-18)

-PROVIDE CABLE ROUTING & SUPPORT

ELECTRIC AL DISTRIBUTION SYSTEM

(SEE AD-22)

-PROVIDES ELECTRICAL POWER


l(SEE AD-7)

CONTROL ROOM COOLING

COMPONENT COOLING SYSTEM (SEE AD-9)

-PROVIDES COOLING TO PUMPS & HEAT EXCHANGERS -

.RESIDUA REMOVAL

I' ]REACTOR L H EAT PROTECTION'

SYSTEM SYSTEM SYSTEM I (SEE AD-17)

' ' -STARTS PUMPS

OPE RATOR

ACTION

-MANUAL OPERATION



+*

I

EBASCO SERVICES INCORPORATED POWER JTNORITY 67,TE O NEW YO4

DI V M DR.V.LZ APPROVED INDIAN POINT NO.IN. P, REIDUAL HEAT REM0VAL $TUDY A D-L4 DATE C.J-- - . r,- /. AUILIARY DI&GRAM SCALE I.JOW. E o..,,w I vv Jz-fK


(SEE AD-7)


ELECTRICAL TRAY SYSTEM

(SEE AD-IS) -PROVIDE CABLE

ROUTING & SUPPOR

-REMOVE CORE HEAT AUXILIARY FEEDWATER SYSTEM .

(SEE AD-23)

T

SAFETY , INJECTIONI SYSTEM (SEE AD-2)

ELECTRICAL DISTRIBUTION SYSTEM (SEE AD- 22)

- PROVIDE S ELECTRICAL POWER

NITROGEN TO THE P.O.R.V (SEE AD-20)

-SAFETY RELATED MOTIVE POWER FOR P.O.R.V'S

/ /

/

RESIDUAL HEAT REMOVAL SYSTEM (SEE AD-4)

-REMOVE DECAY HEAT

COMPONENT COOLING

(SEE AD-9)

-RCS SEALS \OPERATOR ACTION

-MANUAL OPERATION

EBASCO SERVICES INCORPORATED POWER AU"THORITY &TTE OFNEW Y)DOK SAPPROVED INDIAN POINT NO.N.PIP.

REACTOR CODLPNT ;';r'Ti AD-5 SCATLE- CA 4.,/ ] lAU)ILIARY DIAGRAM SCALE4ci ____ ___ __I__w_ _]__/__

CHEMICAL VOLUME & CONTROL SYSTEM

(SEE AD-3 )

-SEAL RC PUMPS (-NORMAL SHUTDOWN

MAKE - U.P)

DC POWER SYSTEM

(SEE AD-12)


/


(SEE AD-7)

-CONTROL ROOM COOL!ING


CSEE AD-IS)


SAFETY INJECTION SYSTEM

(SEE. AD - 2)

-RWST SUCTION TO SPRAY PUMPS

- RECIRCULATION PUMPS FOR LONG TERM SPRAY

REACTOR PROTECTIO SYSTEM (SEE AD-17)

SPRAY AC TUAT ION SIGNAL (HI HI CONT PRESS.)

I I

OPERATOR -ACTION

MANUAL OPERATION

ELECTRICAL POWER

"" RHR SYSTEM (SEE AD-4)

-RHR PUMPS & HEAT EXCHANGERS

DC POWER SYSTEM

(SEE AD-I12)


EBASCO SERVICES INCORPORATED POWER A4UTH0DPITY STATE OF NEW YORK

DIV.AE-W DR.Y-l APPROVED INIIN POINT NO.$N..I DATEJL85CK.. A-F , CONT/kNMENT CPP\kY SYSTEM AD-6

AE___ __ HAJ, LI Ay DIA.A., SCALE 00?'JE )f/; I 901'v ]-'(


N

0 0 u~.. r

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4

-MANUAL ISOLATION

TRAYSYS ,,OPERATOR (SEE AD-I8) ATO ACTION

-CABLE ROUTING SUPPORT

SERVICE WATER SYSTEM

I(SEE AD-B)

-COOLING FOR AC UNITS

INSTRUMENT AIR SYSTEM

-CONTROL TEMPERATURE

-ELECTRICAL POWER

EBASCO SERVICES INCORPORATED POWER AUT'NoIITr5TA7E OF W&J-W YORK

DIv.MA (-H DR.LT 5 APPROVED IWJDIA. Po!l4T WJO $ WPP D AT E L&CHI AL. CO&~L ;0M 'VE-K~rILATivw %o e AD-7


SYSTEM

ELECTRICAL DISTRIBUTION SYSTEM

(SEE AD-22)

ii:

ELECTRICAL

DISTRIBUTION SYSTEM (SEE AD-22)

PROVIDE

ELECTRICAL

POWER

ELETRIALCEA

OPERATORISEV E I SACTION i WATER I

ESYSTEM 11l -MANUALL

OPERATION

DC -- ELECTRICAL TRAY POWER SYSTEM SYSTEM

(S EE AD - 12) (SEE AD-IB)

-PROVIDE CONTROL -PROVIDE CABLE POWER ROUTING & SUPPORT

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SERVICE WATER SYSTEM ( SEE AD-8)

OPERATOR ACTION

-MANUAL OPERATION

-COOLS CC HEAT EXCHANGERS ROVIDES

ELECTRICAL

POWER

[ I - -']TRAY

'COMPONENT I (SEE A I COOLING - -PROVIDE

SYSTEM ROUTINC L V3 SUPPOR

CONTROL ROOM VENTILATION (SEE AD-7)


'ICAL YSTEM D-18)

CABLE

T


-PROVIDES CONTROL POWER

EBASCO SERVICES INCORPORATED POWER AUTHORITY STATE OF NEWYORK.

DVJA D Z AINDIAN POINT NO.lN.;P. DIVA-E.I DR.Iz . APPROVED COMPONENT COOLING WATER SYSTEM AD-9 SCALE QI"JE ,, . A.-/4L.. -


-MANUAL OPERATOR

OPERATOR ACTION

NITROGEN I BACKUP

,SYSTEM

N9 NO SUPPORT SYSTEMS

EBASCO SERVICES INCORPORATED PDWE RUT"ORIT'Y CTP, E OF NEW YOR(.

DIV."'f=C DR.V._ APPROVED INDIAN POINT NO.IN?.P. DATEl/t.-LCK4' ~ NITROQSN BACKUP SYS"FM MAD-1O SCALEONE .40 (] fWF AUJXIL ' DIAGRAM

jI~ u~2:

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OPERATOR ACTION

-MANUAL OPERATION

I

DC POWER

SYSTEM (SEE AD-12)

-FIELD FLASHING DIESEL CONTROLS

L I

ELECTRICAL TRAY

SYSTEM (SEE AD-I B)

-PROVIDE CABLE ROOTING & SUPPORT

EBASCO SERVICES INCORPORATED

SERVICE WATER SYSTEM

(SEE AD-B)

-COOLS DIESEL JACKET & LUBE OIL

II U

FouiJiz AU'NHOITY -T ,E or- kJEb YODZV NJDIAw PONjT WO 5 WPP

OIEsEL GEEAQ TO 5YSTEkA AJXILIA.% DIAz.4p PAI

AD-Ila

DIESEL GENERATOR

SYSTEM

F_ - - -

ELECTRICAL DISTRIBUTION

SYSTEM (SEE AD-22)

-PROVIDES DC LOADS & ELECTRICAL

TRAY SYSTEM CHARGING (SEE AD-I8) POWER


S DC J IPOWER I

EBASCO SERVICES INCORPORATED POW, AUTIOZITY STATE OF WEW YOM.

DIV ME. DR-.. 3 APPROVED INDIA PoIkIT W A-D P2 DATEITSICK-A F c cr Plow AD- 12 CALECJLIA.Zy D4. M

SCALUEO"t OAK _____

-PROVIDE ROUTING

DC POWER SYSTEM

(SEE AD-12)


CABLE & SUPPORT

ELECTR DISTRIB

SYST (SEE A

-PROVIDE POWER

ICAL UTION EM D-22)

ELECTRICAL

EBASCO SERVICES INCORPORATED

DIV. Er- 11 DR. APPROVED DATE V CHKA, E / 8'

SCALE WOcJJE-.i 1-W-9l 1 ik

U U

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C0QTkIkEmrT ISOL4,TIOwJ S'rTEVA A'JYJIAZ DlA6R4AW

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(SEE AD-18)

CONTAINMENT ISOLATION

SYSTEM

AD-19

. M

NITROGEN BACKUP SYSTEM (SEE AD-IC

ELECTRICAL DISTRIBUTIC SYSTEM (SEE AD-22)

-ELECTRICAL POWER

DC SYSTEM (SEE AD-12)

-PROVIDE CONI POWER

-SAFETY MOTIVE POWER FOR POWER OPERATED ATMOSPHERIC RV'S

N MAIN MAIN STEAM STEAM SYSTEM 1

7) VALVES S YSTE IP 7Ai

N OPERATOF ACTION

I-MANUAL V,

CLOSURE -PROVIDE CABLE

SUPPORT & ROUTING

rROL

z ALVE

ELECTRICAL TRAY SYSTEM. (AD-I8)

EBASCO SERVICES INCORPORATED POWER AUTNORITY STATE O NEWYORK.

DIV.m.L:-'- DR.V-7-- APPROVED INDIAN POINT NO.'IPP. AD -14 DATE 1ibC .,& S MAIN STEAM 5YSTEMAD SCALE4OC I41* 1-/ wm p p- L PI IA lIG-,X

+*REACTOR PROTECTI( SYSTEM (SEE AD-I

______


I ELECTRICAL DISTRIBUTION SYSTEM (AD-22)

- PROVIDE INSTRUMENT POWER

REMOVES HEAT FROM S/G

- -- __________ I

EBASCO SERVICES INCORPORATED POW-.RAUTHOPUT'( STATE OF NEWYORK DIV.M':4-- DR.V.. APPROVED INDIAN POiNT NO.IN.P. DATEI LL EE,., -I , FEERWTER TEM AD-15 SCALE NOCJF .j ri,

E*

ELECTRICAL TRAY SYSTEM (SEE AD-I8)

-iI

K'FEEDWAT SYSTEM

AUXILIARY

FEEDWATER

I

CONTAINMENT RECIRCULATION FANS (SEE AD-21)

-RECOMBINER SUCTION

-PROVIDE ELECTRICAL POWER

TRUCK -TO MAINTAIN RECOMBINER FLAME

I I I OXYGEN ISUPPLY

I I TRUCK

IM -TO START & HYDROGEN lMAIN MAINTAIN RECOMBINER HYDROGEN I STABLE

SUPPLY I RECOMBINER FLAME

ELECTRICAL TRAY SYSTEM (SEE AD-18)


NITROGEN I SUPPLY I -SUPPLY PNEU

-- J VALVES & INSTRUMENTS ASSUMING NO

OPERATOR INSTRUMENT ACTION AIR

- MANUAL OPERATION

DC SYSTEM (SEE AD-12)


EBASCO SERVICES INCORPORATED POWER AUTNOR1TY CTATE DFP7NEW YOR.K

DIV.M L- DR.VZ- APPROVED INDIN POINT N).INPR DATE I7CHRAF I UYNDD EN P, ECOMBINER S TEMI AD-16

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ELECTRICAL DISTRIBUTION SYSTEM(SEE AD-22)

4

ELECTRI CAL DISTRIBUTION SYSTEM (SEE AD-22)

-PROVIDES AC POWER

DC POWER SYSTEM

(SEE AD-12)

-PROVIDES POWER

ri REACTOR

W-PRO TE CTI O NI ,SYSTEM I (

OPERATOR ACTION I

-PROVIDES

MANUAL OPERATION


(SEE AD-7)

:ONTROL ROOM :OOLING

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TRAY SYSTEM

L

HVAC (SEE AD-19)

-PROVIDES

CABLE COOLING

V II

EBASCO SERVICES INCORPORATED POWER AUTHOFITY STATE OF NEW YORK

DIV. MECH DR.PLC- APPROVED INDIAN POINT NO 5 DATEJ I/. AE k T , TRAY 5YSTEM AD-18 SCALE NONE _ AUXILIARY DIACARM

1*

TRAY SYSTEM

(SEE AD-18)

-CABLE ROUTING & SUPPORT

HVAC

CELEC TUNNEL&CONT BLDG)

ELECTRICAL DISTRIBUTION SYSTEM

(SEE AD-22)

-ELECTRICAL POWER

W U

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a

I

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SUPPORTSYSTEMI SYSTEMS

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NITROGEN ACTUATION

4,.

NO

SERVICE WATER SYSTEM (SEE AD-B)

ELECTRICALI TRAY SYSTEM

L(SEE AD-I8)I

-PROVIDE CABLE I ROUTING & SUPPORT

-PROVIDE COOLING

p- --- -i1 I

CONTAINMENT I RECIRCULATION I

FAN UNITS I

-PROVIDE ELECTRICAL POWER


EBASCO SERVICES INCORPORATED POLUEp A UTTHh ATE OF WEW yOR,

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DIESEL GENERATOR (SEE AD-II)

PROVIDES EMERGENCY ELECTRICAL POWER

OFFSITE I POWER SYSTEM

PROVIDES L PREFERRED SOURCE L OF EMERGENCY ELECTRICAL POWER ELECTRICAL

P:

fRAY SYSTEM (SEE AD- IS)

PROVIDES CABLE ROUTING & SUPPORT

REACTOR PROTECTION SYSTEM

(SEE AD-17)

OPERATOR ACTION

MANUAL

OPERATION

DC POWER -"I SYSTEM

(SEE AD- 12)

PROVIDES

CONTROL POWER & EMERGENCY POWER TO INVERTERS


(SEE AD 7)


I II

EBASCO SERVICES INCORPORATED P.JE2 A.HOP-kIT'Y -7A (OFIEW YOIr

DIV.tAF-"e DR.-- APPROVED I w DI POIKIT kWO J.IJPP DATEA .CKAF ETL2ECT2IC7L DITP-..UTIOJ lPME AD-22 DAE D.- A- AUY.IUNO.I AGRAM SCALE NotIJ-_ JQ i €.

ELECTRICAL DISTRIBUTION

EGUIPMENT

m m

SELECTRICAL TRAY' SYSTEM (SE E AD -16)


MAIN STEAM SYSTEM (SEE AD-14)

STEAM SUPPLY TO AUX FEED pUMP TURBINE



CITYWATER

/-ALTERNATE WATER SOURCE

ELECTRICAL DISTRIBUTION SYSTEM (SEE AD-2

PROVIDES ELECTRICAL POWER

I-OPERATOR ACTION

- MANUAL OPERATION

NITROGEN BACKUP (SEE AD-ID)

AUX FEED -' WATER j F ACTUATION v SIGNAL.

-INITIATES AUX FEED WATER

IOTIVE POWER OR PNEUMATIC

'ALVES

L AUXILIARY I ,,PFEEDWATER S SYSTEM L

EBASCO SERVICES INCORPORATED -P.W kUl'NOP-l"IN '1A.TE OF WE:WYOer

DIV'MEC, DR.w._ APPROVED INDIAI O '['M joPP AD, -2 AU14 Ll 4,.2,." F .EEWx E:, YS.' DATE_1i CRAL. /4XILIA2_ 01.iZ

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1.0 Rod Control

1.1 Function & Applicability

The rod control system is required to provide the basic safety function of Achieving & Maintaining Reactor Subcriticality (FT-l)*. This is accomplished by the insertion of the Shutdown and Control Rod Cluster Control assemblies upon commands from the operator or from an automatic signal from Reactor Protection System.

1.2 Scope

1.2.1 The rod control system consists of components which can be subdivided into two categories

A) Internal to or a part of the reactor pressure vessel & head

assembly.

B) Components located remote from the RPV.

The study addresses itself to the B category of components, dealing specifically with the cable raceway system from the RPV to the Control Room and the rod control power supply system. The Rod Control System is described in System Description No. 16.0. . 1.2.2 All devices located in the Control Room are evaluated generical ly as part of control room review.

1.3 Description of Rod Control Power Supplies

Power to rod drive mechanisms is supplied by two motor generator sets operating from two separate 480 Volt, three-phase buses. The generators, driven by 150 hp induction motors, are paralleled through circuit breakers. Each generator is the synchronous type, rated at 438 Kva, 260 Volts line-to-line.

The a-c power is distributed to the power cabinets through the two series connected reactor trip breakers. Bypass breakers can be connected in parallel with the reactor trip breakers to facilitate on-line testing of the protection system. Their use is under administrative control and is not included in the scope of the study. The a-c power distribution lines downstream of the reactor trip breakers are run across the top of the power cabinets through a fully enclosed, three-phase, four-wire, plug-in bus duct. Power to each power cabinet is fed from the bus duct through three plug-in, fused disconnect switches serving the stationary, movable, and lift coil circuits of the mechanisms associated with the power cabinet.

The power cabinets contain equipment which converts t -he a-c supply to pulsed dc required by the mechanism coils. Each power cabinet can accomodate three

*Functional Table 1. See Volume I Tab entitled Functional Tables

groups, with a maximum of four mechanisms per group. (One group is designed to handle five mechanisms.) Design of the power cabinet permits motion of only one of the three-groups, with the other--two groups held in a stationary position. A dc hold panel is provided to'supply holding power to the stationary coils of a mechanism during maintenance operations.

A logic cabinet is provided to inform the power cabinets which group of rods is to be moved and translates speed and direction input signals into a form usable by the power cabinets.

2.0 Safety Injection System


The Safety Injection System (SIS) is required to accomplish the four basic functions as follows:

i) Achieve and Maintain Reactor Subcriticality (FT-I).*

The SIS provides makeup liquid from the Refueling Water Storage Tank to the Reactor Coolant System (RCS). Additionally SIS provides chemical reactivity control to provide sufficient reactivity shutdown margin and prevent an uncontrolled return to criticality. These functions are accomplished by the Accumulators in the event of a large LOCA or steam line break or; the Safety Injection Pumps in the event of a small or medium LOCA or a break in a secondary system (ie. Feedwater, Tube Rupture etc.) or unavilability of CVCS for normal shutdown.

ii) Maintain Containment Integrity (FT-2)*

In the event of a LOCA and subsequent to the emptying of Refueling Water Storage Tank (RWST) the Recirculation Pumps are required to recirculate the contents of the recirculation containment sump through the containment spray nozzles to reduce containment pressure and temperature. If the LOCA was large, the Safety Injection Pumps will also be required.

iii) Remove Decay Heat (FT-3)*

The Recirculation Pumps can be used in lieu of the RHR pumps to maintain circulation through the Reactor and the RHR heat exchangers.

iv) Maintain Reactor Coolant Pressure Boundary (FT-4)*

In the event of a break in one of the injection lines, the injection valves are required to isolate the break from the RCS.

The Safety Injection System is described in System Description No. 10.1.

2.2 Scope

2.2.1 For the purpose of this study the SIS includes:

- Accumulators

- Boron Injection Tank (BIT)

* Functional Tables 1 thru 4. See Volume 1 tab entitled Functional Tables

2-1

- Refueling Water Storage Tank (RWST)

Containment Recirculation Sump

Safety Injection Pumps

- Recirculation Pumps

Piping, Valves, Instrumentation and Electrical components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for SIS are shown on flow diagram FD 5209.02, Sheets 1 & 2.

2.2.2 -Support systems relied upon for SIS safety related functions are indicated on Auxiliary Diagram AD-2.

2.2.3 Systems which interface with the SIS but are not deemed necessary for accomplishing the safety related functions are:

i) Chemical Volume and Control System (CVCS)

CVCS serves to adjust boron concentration in the RWST and BIT. This is not a safety related function since boron concentration is assumed to be maintained within limits for plant operation.

ii) Primary Water System

Primary water is used to initially fill the RWST. The RWST must have sufficient volume for SIS function for plant operation.

iii) Nitrogen Supply Package

The system provides nitrogen to satisfy the pressure

requirements of the Accumulators. The Accumulators are isolated from the Nitrogen Supply by normally closed valves 891A through D. The Accumulators are assumed pressurized for plant operation.

iv) Auxiliary Steam & Electrical Heat Tracing

Auxiliary Steam is used to maintain minimum water temperature in the RWST and the outdoor pump suction lines are heat traced. It is assumed that any failure of the systems would be repaired long before minimum temperature limits are reached. (See answer to FSAR question 9.10).

If failure of Auxiliary Steam was due to an initiating event the S15 would empty the RWST long before minimum temperature

-was reached.

v) Sampling System (SS)

SS does not contribute to any of the four safety functions. The boundary of the study is the first normally closed sampling valve.

vi) Waste Disposal System (WDS)

The Waste Disposal System provides a method for draining the Accumulators. This is not a safety related function.

vii) Instrument Air System

All valves which are powered by instrument air will, upon loss

of air, fail in their safety related position (Response to FSAR question 9.19.3). Instrument air is therefore not relied upon for safety related function.

2.2.4 All devices located in the Control Room are evaluated generically as part of the Control Room review.

2-3

3.0 Chemical Volume and Control System


The Chemical and Volume Control System (CVCS) is required to provide two of the four basic safety functions as follows:

i) Achieve and Maintain Reactor Subcriticality (FT-l)*

Emergency boration by the CVCS provides a backup reactivity shutdown capability to shutdown the reactor independent of the control rod clusters. Also, the CVCS provides pressure control of the Reactor Coolant System through use of the Pressurizer Auxiliary Spray. The CVCS also provides makeup to the RCS via the Charging Lines (No RCS or Secondary Line Breaks).

ii) Maintain Reactor Coolant Pressure Boundary (FT-4)*

CVCS valves in the Letdown are required to isolate a Letdown rupture from the RCS.

CVCS is described in System Description No. 3.

3.2 Scope

3.2.1 For the purpose of this study CVCS includes:

- Charging Pumps

- Boric Acid Tanks & Transfer Pumps

- Auxiliary Pressurizer Spray

- Piping, Valves, Instrumentation and Electrical Components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

3.2.2 Included on the Interconnected and Nonconnected Matrices along with its respective piping and valves are:

- Regenerative Heat Exchanger

- Non-Regenerative Heat Exchanger

- Volume Control. Tank

- Boric Acid Blender

- Excess Letdown Heat Exchanger

- Seal Water Heat Exchanger

* Function Tables 1 & 4. See Volume 1 Tab entitled Functional Tables.

These items do not contribute to any of the four basic safety functions and, for the purpose of this study, are considered as non-safety related. However, their inclusion in the report was deemed necessary since the integrity of the piping is required to avoid degrading effects on the safety related components of the system.

3.2.3 The recycle system is not included within the scope of this study since it does not contribute to any of the basic safety related functions and is isolated from the rest of the CVCS.

The boundaries of this study for CVCS are shown on flow diagram FD 5209.03, Sheet 1.

3.2.4 Support systems which are relied upon for CVCS safety related functions are included on Auxiliary Diagram AD-3.

3.2.5 Systems which interface with CVCS but are not deemed necessary for accomplishing the safety related functions are:

i) Instrument Air System


of air, fail in their safety related position (Response to FSAR Question 9.19.3). Instrument Air is therefore not relied upon for safety related function.

The following will occur on loss of instrument air:

- HCV-133 - RHR to CVCS letdown fails closed.

- HCV-142 - CVCS charging fails closed.

- 201,202 Letdown containment isolation fail closed.

- Charging pump speed goes to maximum.

These occurances will effect the transition to RHR.

Emergency boration could be accomplished by seal injection or by the manual bypass around HCV-142. Since the letdown line becomes unavailable no flow path will exist between RHR and CVCS. However, RHR flow from the RCS, through the RHR heat exchangers and back to the RCS can be accomplished without instrument air.

The charging pump speed going to maximum would cause the pressurizer to fill. This is not serious due to the shrinking effect of the cooldown, and the requirement for pressurizer level to be at 85%. Manual control would prevent the pressurizer from completely filling.

3-2

ii) Waste Disposal System

This system provides for drainage of CVCS components for maintenance. It does not contribute to any safety related function.

iii) Nitrogen & Hydrogen

The volume control tank is used to strip hydrogen and reactor

fission gases from reactor coolant prior to cold shutdown for refueling. This is not a safety related function.

3.2.6 All devices located in the Control Room are evaluated generically as part of the Control Room Review.

4.0 Residual Heat Removal System


The Residual Heat Removal system (RHR) is required to accomplish three of the four basic safety functions as follows:

i) Achieve and Maintain Reactor Subcriticality (FT-I)*

The RHR provides makeup liquid, from the Refueling Water Storage Tank (RWST), to the Reactor Coolant System (RCS). This is accomplished by the RHR pumps after the injection by the SIS Accumulators in the event of a large LOCA.

ii) Maintain Containment Integrity (FT-2)*

In the event of a LOCA and after the RWST is empty, the RHR pumps can be used in lieu of the SIS recirculation pumps to recirculate the contents of the containment sumps through the Containment Spray Nozzles to reduce containment pressure and temperature.

iii) Remove Decay Heat (FT-3)*

The RHR pumps can be used in lieu of the SIS Recirculation Pumps to maintain circulation through the Reactor and the RHR Heat Exchangers.

RHR is described in System Description No. 4.2.

4.2 Scope

4.2.1 For the purposes of this study, RHR includes:

- RHR Pumps

- RHR Heat Exchangers

- Containment Sump

- Piping, Valves, Instrumentation and Electrical Components as

indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The study boundaries for RHR are shown, along with CCW, on flow diagram FD 5209.04, Sheets 1 & 2. Note that component cooling (system 9) is included on this flow diagram.

4.2.2 Support systems relied upon for RHR safety related functions are indicated on the Auxiliary Diagram AD-4.

4.2.3 All devices located in the Control Room are evaluated generically

as part of the Control Room review.

* Function Tables 1 thru 3. See Volume I Tab entitled Functional Tables

5.0 Reactor Coolant System


The Reactor Coolant System (RCS) is required to accomplish two of

the four basic safety functions as follows:

i) Achieve and Maintain Reactor Subcriticality (FT-l)*

The RCS removed decay heat by means of natural circulation through the Reactor and Steam Generators. In addition the RCS forms a flow path for makeup fluid. The pressurizer code safety valves or the power operated relief valves limit any pressure excursion in the RCS.

ii) Maintain Reactor Coolant System Pressure Boundary (FT-4)*

The pressurizer code safety valves or thepower operated

relief valves prevent overpressurization of the RCS.

The RCS is described in System Description No. 1.

5.2 Scope

5.2.1 For the purpose of this study RCS includes:

- Steam Generators

- Pressurizer

- Reactor Coolant Pumps

- Reactor Vessel

- Overpressurization Protection System

- Piping, Valves, Instrumentation and Electrical equipment as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The pressurizer relief tank and valve discharge piping is designated as Seismic Category II. It is not required for the safety related function of the RCS and is therefore not included within the scope of this study. The study boundary for RCS is shown on flow diagram FD 5209.05, Sheets 1 and 2.

5.2.2 Support systems relied upon for RCS safety related functions are

indicated on the Auxiliary Diagram AD-5.

*Function Tables 1 and 4. See Volume 1 Tab entitled Functional Tables

5.2.3 Systems which interface with the RCS but are not deemed necesssary for accomplishing the safety related functions are:

i) Sampling System

SS does not contribute to any of the safety functions. The boundary of the study is the first normally closed sampling valve.

ii) Waste Disposal System

The WDS provides for draining and venting of the RCS. WDS

does not contribute to the safety functions and hence is not included within the study boundary.


6.0 Containment Spray System

Function & Applicability

The Containment Spray System (CSS) is required to provide the basic safety function of Maintaining Containment Integrity (FT-2)*. This is accomplished by reducing containment pressure and temperature with the containment spray pumps discharging through the containment spray headers.

The Containment Spray System is described in System Description No. 10.2.

6.2 Scope

6.2.1 For the purposes of this study CSS includes:

- Containment Spray Pumps

- Spray Nozzles

- Liquid Jet Eductors

Piping, Valves, Instrumentation and Electrical components as

indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The study 1 and 2.

boundary of CSS is shown with SIS on flow diagram FD 5209.02, Sheets

6.2.2 Support Systems which are relied upon for CSS safety related functions are indicated on the Auxiliary Diagram AD-6.

6.2.3 The CSS removes iodine from the containment atmosphere by the addition of sodium hydroxide (NaOH) to the sprayed liquid. The addition of NaOH is not needed to accomplish a basic safety function as defined for this study. The Spray Addition Tank is therefore included in this study only to the extent that fluid may not be lost from the CSS.

but is not deemed All valves which in their safety

6.2.5. All devices located in the Control Room are evaluated generically as part of the Control Room review.

* Function Table 2. See Volume I Tab entitled Functional Tables

6.2.4 The Instrument Air System interfaces with CSS necessary for accomplishing the safety related functions. are powered by instrument air will upon loss of air, fail related position (Response to FSAR question 9.19.3).

7.0 Control Room Ventilation System


The Control Room Ventilation System (CRVS) is required to retain habitability of the Control Room during all modes of plant operation. The CRVS is described in System Description No. 11.

7.2 Scope

7.2.1 For the purpose of this study CRVS includes the Control Room Air Conditioning Units and Ductwork, Instrumentation and Electrical components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

7.2.2 Systems which are relied upon for CRVS safety related function are indicated on Auxiliary Diagram AD-7.

Note that instrument air is required for operation of the Control Room Ventilation System and is required to accomplish a safety function only for this system. Therefore, the instrument air piping from the supply piping to the Control Room Ventilation System as well as the compressors were studied to assure their availability.


8.0 Service Water System

8.1 Function & Applicability.

The Service Water System (SWS) provides cooling water to the

following safety related systems:

- Containment Air Recirculation System

- Component Cooling System

- Diesel Generator System

- Control Room Ventilation System

SWS is described in System Description No. 24.0.

8.2 Scope

8.2.1 For the purpose of this study SWS includes:

- Service Water Pumps

- Service Water Strainers

- Piping, Valves, Instrumentation and Electrical equipment as

indicated in the corresponding disciplines Nonconnected and

Interconnected Matrices.

8.2.2 Portions of SWS which service nonsafety related components, the

Conventional Plant Closed Cooling System and the Instrument Air Closed Cooling

System are not included within the scope of this study. Also, the Backup

Service Water Pumps, which are designated Seismic Category III and are not

missile protected, are not included within the scope of this study.

The boundaries of this study for SWS are shown on flow diagram FD 5209.08,

Sheet 1.

8.2.3 Support systems relied upon for SWS safety related function are

indicated on the Auxiliary Diagram AD-8.

8.2.4 Systems which interface with SWS but are not deemed necessary for

accomplishing its safety related function are:



of instrument air, fail in their safety related position

(Response to FSAR question 9.19.3). Instrument air is

therefore not relied upon to accomplish the SWS safety related

function.



9.0 Component Cooling Water System


The Component Cooling Water System (CCW) provides cooling for the following safety related services:

- RHR Heat Exchangers

- RHR Pumps

- Safety Injection Pumps

- Recirculation Pumps

- Charging pumps, fluid drive coolers and crankcase cooler.

A detailed description of the CCW is contained in System Description No. 4.1.

9.2 Scope

9.2.1 For the purpose of this study CCW includes:

- Component Cooling Surge Tanks

- Component Cooling Pumps

- Component Cooling Heat Exchangers

- Piping, Valves, Instrumentation, and Electrical components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for CCW are shown, along with RHR, on flow diagram FD 5209.04, Sheets 1 and 2.

9.2.2 Included on the Interfacing Component list is CCW piping and valving for cooling of:

- Reactor Coolant Pumps

- Excess Letdown Heat Exchangers

- CVCS Non-Regenerative Heat Exchanger

- Seal Water Heat Exchanger

- Boric Acid Recycle Evaporators and Condensate Coolers

- Sample Heat Exchangers

- Waste Evaporator Condenser'

Waste Gas Compressors

- Reactor Vessel Support Pads

- Spent Fuel Pit Heat Exchanger

- Primary Make-Up Heat Exchanger

Waste Disposal System - Liquid, Polishing-Demineralizer Heat Exchanger

- Gross Failed Fuel Detector

These items do not contribute to any of the four basic safety functions and,

for the purpose of this study are considered nonsafety. However, since these nonsafety items do not have any automatic means of isolation from the CCW headers, an interaction causing a rupture of a line servicing them may compromise safety related components served by CCW. Therefore, piping to and from the nonsafety components is considered to the extent that the lines may not rupture.

9.2.3 Support systems relied upon for CCW safety related functions are indicated on Auxiliary Diagram AD-9.

9.2.4 The Instrument Air System which interfaces with the CCW is not deemed necessary for accomplishing the safety related function. All valves which are powered by instrument air will, upon loss of air, fail in their safety related position (Response to FSAR question 9.19.3).



9-2

10.0 Nitrogen Backup System


The function of the Nitrogen Backup System (NBS) is to supply backup gas to the Pneumatic Valve Actuators that require air pressure to go to their proper safeguards position during accident conditions.

The valves in this category are; the auxiliary feedwater control valves and city water supply valves to the suction of the auxiliary feedwater pumps.

The Nitrogen Backup supply is also used for manual operation of the power operated atmospheric steam relief valves.

The NBS is described in System Description No. 29.7.

10.2 Scope

10.2.1 For the purpose of this study the NBS includes:

- One bank of three Nitrogen Cylinders.

Piping, Valves, Instrumentation and Electrical components on the Interconnected and Nonconnected Matrices.

10-1

11.0 Diesel Generator System


The Diesel Generator System (DGS) is required to provide a source

of backup power to safety related equipment following a loss of normal power supply. DGS is described in System Description 27.3.

11.2 Scope

11.2.1 For the purpose of this study DGS includes:

- Diesel Generator

- Jacket Water System

- Fuel Oil System

- Starting Air System

- Lube Oil System

- D G Building Ventilation

Piping, Valves, Instrumentation and Electrical components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for DGS are shown on flow diagram FD 5209.11,

Sheets 1 and 2.

11.2.2 Support systems relied upon for DGS safety related functions are indicated on Auxiliary Diagray AD-Il.

11.2.3 The city water system interfaces with the jacket water system of the Diesel Generator by providing makeup. The makeup is for fluid loss through primarily evaporation and leakage. Makeup is not considered a safety function since it is assumed that the Diesel Generator could operate for an extended period of time with the available water inventory of the jacket and expansion tank.


11-1

12.0 DC System


The dc system is required to accomplish each of the four basic safety functions in that as a minimum it provides:

1) Control power to all 480V Switchgear breakers.

2) Diesel Generator field flashing and control circuitry.

3) Power to solenoid operated valves and other miscellaneous control circuits.

4) Input to instrument power supplies.

5) Provides temporary emergency lighting.

Each of these functions are considered essential since they form an important part of the operation of the systems being reviewed in this study. This fact is evidenced by the number of auxiliary diagrams which reference the dc system.

12.2 Scope

12.2.1 The dc system is comprised of the following:

1) Four (4) 125 volt lead acid batteries Nos 31-34.

2) Four (4) Power Panels Nos 31-34.

3) Four (4) Distribution Panels Nos 31-34.

4) Battery Room Exhaust Fans.

5) Interconnecting cables and raceway.

A complete description of the dc system can be found in System Description No. 27.1.

12.2.2 All devices located in the control room are evaluated generically as part of the control room review. The battery chargers are not included within the scope of this study since they are not operable during a loss of offsite power and are considered to be nonsafety related components.

12.2.3 The support systems relied upon for successful operation of the dc system are indicated in Auxiliary Diagram AD-12.

12-1

13.0 Containment Isolation System


The Containment Isolation System (CIS) is required to Maintain .Containment Integrity (FT-2).* A detailed description of the CIS is provided in System Description No. 10.7.

13.2 Scope

13.2.1 For the purpose of this study CIS includes Piping, Valves, Instrumentation and Electrical components as indicated in the corresponding disciplines Nonconnected and Connected Matrices. It should be noted that many v .alves which are part of CIS are also part of other systems included in this study. Therefore, valves or piping which were studied as part of another system are not repeated in the study of CIS.

13.2.2 All safety related lines, valves and equipment of the Air Cooling System for Hot Penetrations are included in this study since this system is deemed necessary for the safe function of the Containment Isolation System.

13.2.3 Support systems relied upon for CIS safety related functions are indicated on Auxiliary Diagram AD-13.

13.2.4 Systems which interface with CIS but are not deemed necessary for

the safety related functions are:


All valves which are powered by instrument air will, upon loss of air, fail in their safety related position (response to FSAR question 9.19.3). Instrument air is therefore not relied upon to accomplish the CIS safety related function.

ii) Isolation Valve Seal Water System and Containment Penetration and Weld Channel Pressurization System

No credit has been taken for the operation of either system in the calculation of off-site accident doses (FSAR Section 6.5.1 and 6.6.1). These two systems were therefore not included within the scope of this study.

13.2.5 All devices in the Control Room are evaluated generically as part of the Control Room review.

*Functional Table 2. See Volume I Tab entitled Functional Tables

13-1

R1

14.0 Main Steam System


The Main Steam System (MSS) is required to Achieve and Maintain Reactor Subcriticality (FT-l)* in that it removes heat by means of the Power Operated Atmospheric Dump Valves or the Main Steam Safety Valves. MSS also provides motive steam for Auxiliary Feedwater Pump #32. The Main Steam System is described in System Description No. 18.0.

14.2 Scope

14.2.1 For the purpose of this study the MSS includes:

- Steam Generators

- Atmospheric Dump Valves

- Main Steam Safety Valves

- Main Steam Isolation Valves

- Piping, Valves, Instrumentation and Electrical Components indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The study boundaries for MSS are shown on flow diagram FD 5209.14, Sheet 1.

14.2.2 Support systems relied upon for MSS safety related functions are indicated on Auxiliary Diagram AD 14.

14.2.3 The Instrument Air System interfaces with the MSS but is not deemed necessary for accomplishing the safety related function. The atmospheric dump valves are supplied with motive power by the Nitrogen Backup System. The Main Steam Isolation Valves fail safe (fail closed) on loss of air. Instrument air is therefore not relied upon to accomplish the MSS sa fety related function.


*Function Table 1. See Volume I Tab entitled Functional Tables

14-1

15.0 Feedwater System


Auxiliary Feedwater

15.2

The Feedwater System (FWS) is required to provide a flow path for Feedwater and hence supports the safety function of the Auxiliary System. The FWS is described in System Description No. 21.

Scope

15.2.1 For the purpose of this study FWS includes Piping, Valves, Instrumentation and Electrical Components as indicated on the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for the FWS are shown on flow diagram FD 5209.15, Sheet 1 along with the feedwater portion of the Auxiliary Feedwater System.

15.2.2 Support systems relied upon for FWS safety related functions are indicated on Auxiliary Diagram AD-15.

15-1

15.1

R1

16.0 Hydrogen Recombiner System


The Hydrogen Recombiner is required to Maintain Containment Integrity (FT-4)*. It accomplishes this by reducing the concentration of hydrogen in the containment atmosphere. The Hydrogen Recombiner is described in System Description No. 10.9.

16.2 Scope

16.2.1 For the purpose of this study the Hydrogen Recombiner includes:

- Recombiners

- Hydrogen Stands

- Oxygen Stands

- Control Panel

- Piping, Valves, Instrumentation and Electrical Components as indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for the Hydrogen Recombiner are shown on flow diagram, FD 5209-16, Sheets 1 and 2.

16.2.2 Support systems relied upon by the Hydrogen Recombiners for accomplishing the safety related function are indicated on Auxiliary Diagram AD-16.

16.2.3 The Instrument Air System is relied upon for valve modulation, instrument intelligence transmission and control, and conversion of electrical signals to pneumatic signals. Nitrogen, supplied via a truck connection, can also provide the same functions. The Instrument Air System is not included on the Auxiliary Diagram since the Hydrogen Recombiner is required by the thirteenth day post accident, and it is assumed that this is adequate time for truck delivery of nitrogen., This assumption is justifiable since truck delivery is also relied upon for the supply of hydrogen and oxygen which is also required from system operation.


* Functional Table 4. See Volume I Tab entitled Functional Tables

16-1

17.0 Reactor Protection System


The functions of the Reactor Protection System (RPS) are as follows:

i) To protect the reactor core against fuel rod cladding damage caused by departure from nuclear boiling or high power density.

ii) To detect a failure of the reactor coolant system and initiate actions to contain any radioactive fission products.

iii) To generate alarms and initiate process signals for the control systems when trip conditions are approached.

The RPS is described in System Description No. 28

17.2 Scope

17.2.1 The RPS automatically trips the plant whenever the plant conditions monitored by nuclear and/or process instrumentation reach specified limits. The system also provides alarms that alert the plant operator when manual action is required to prevent a plant trip.

The RPS instrumentation initiate the following trip functions:

a) Overtemperature DT b) Overpower, c) High and Low Pressurizer Pressure d) High Pressurizer Level e) Reactor Coolant System Low Flow f) Nuclear Instrumentation High Flux g) Steam Flow/Feed Flow Mismatch in coincident with Low Steam

Generator Level h) High and High-High Containment Pressure

For the purpose of this study, some of the RPS instrumentation which generate the above trip functions were included as part of other systems such as the Reactor Coolant System.

17.2.2 Support systems relied upon for RPS safety related functions are indicated on Auxiliary Diagram AD-17.

17-1

18.0 Electrical Trays


The electrical trays included in the Systems Interaction study are those carrying power, control and instrumentation cables in the Reactor Containment Building, the PrimaryAuxiliary Building, the Control Building, the Fan House and the Cable Tunnels. The cable trays in the Turbine Building are not included in-this study.

The electrical trays function as raceways for routing power, control and instrumentation cables from components to components in the power plant. The trays are segregated by voltage class. Also, trays carrying cables of redundant safety systems are separated from each other as defined in the FSAR.

18.2 Scope

The trays system included in this study can be broadly divided into those in the Reactor Containment Building, the Primary Auxiliary Building, the Control Building, the Fan House and the Cable Tunnels. On each floor of these structures trays have been provided in a manner such that their damage from internal or external missiles is minimized. Also they are placed either near the ceiling or the walls to minimize damage-from equipment movement inside the buildings.

The cable trays are considered to be supported seismically if they meet the specific details provided on the electrical tray and conduit drawings. In discussions with the Power Authority regarding this topic, it was established that the use of "Unistrut" tray and conduit supports had previously been reviewed and found to satisfy the plant seismic criteria. For the purpose of this study all raceway components (regardless of the safety significance of the installed cables) utilizing these supports can not be classified as sources.

18-1

-19.0 HVAC Systems


The primary safety related function of the HVAC systems is to remove heat from operating equipment by maintaining safe ambient operating temperatures. The HVAC systems are described in System Description No. 11.

19.2 Scope

19.2.1 For the purpose of this study, the HVAC systems include ventilation systems for the Electrical Tunnels and the Control Building. Equipment, Ductwork, Louvers, Instrumentation and Electrical Components are indicated in the corresponding disciplines Nonconnected and Interconnected Matrices.

19.2.2 Ventilation for other safety related areas of the plant are evaluated in the following sections of this report:

Section 7.0 - Control Room Ventilation

Section 11.0 - Diesel Generators (Diesel Generator Room Ventilation)

Section 12.0 - D C Power (Battery Room Ventilation)

Section 21.0 - Containment Air Recirculation System

19.2.3 Support systems which are relied upon for HVAC systems safety related functions are included in Auxiliary Diagram AD-19.

19.2.4 HVAC for the Auxiliary Feed Pump Building was not included in this study. The Primary Auxiliary Building HVAC system has been determined not to be required for at least 24 hours based on letter JR Slotterback, UEC to C Pratt, PASNY & B Brandenburg, Con. Ed. dated 10/6/82.

19-1

20.0 Nitrogen Actuation System


The Nitrogen Actuation System (NAS) is required to provide safety related motive power to the Presurizer Power Operated Relief Valves (PORV) and therefore supports the safety functions of the Reactor Coolant System. The study boundary for NAS is shown on flow diagram FD 5209.20, Sheet 1.

20.2 Scope

20.2.1 For the purpose of this study NAS includes the Nitrogen Accumulators, Piping, Valves, and Instrumentation up to the PORV's and Electrical Components as indicated in the corresponding disciplines Interconnected and Nonconnected Matrices.

20.2.2 As indicated on Auxiliary Diagram AD-20, the NAS does not require any support systems for its safety related function.

20.2.3 functions.

The Nitrogen Supply Package is not deemed necessary for NAS safety

20.2.4 All devices located in the Control Room are reviewed generically as part of the Control Room review.

20-1

21.0 Containment Air Recirculation System


The Containment Air Recirculation System (CARS) is required to provide the basic safety function of Maintaining Containment Integrity (FT-2)*. It accomplishes this by reducing containment pressure and temperature and removes fision products from the containment atmosphere. CARS is described in System Description No. 10.3.

21.2 Scope

For the purpose of this study CARS includes:

- Fan Cooler Units

- Distribution Ducts

Fan Cooler Drain Collection and Measurement Piping and Valves as indicated on the Mechanical Interconnected and Nonconnected Matrices.

21.2.2 Support systems relied upon for CARS safety indicated on Auxiliary Diagram AD-21.

21.2.3 All devices located in the Control Room are as part of the Control Room review.

* Function Table 2.

related functions are

evaluated generically

See Volume I Tab entitled Functional Tables

21-1

21.1

21.2.1

22.0 Electrical Distribution Equipment


The primary function of the electrical distribution equipment is to provide an independent and redundant source of emergency power for the auxiliaries required for the four basic-safety functions. This function includes providing a reliable source of power for instrumentation and controls of the auxiliaries.

22.2 Scope

The electrical distribution equipment considered for the purpose of this study is that portion of the electrical AC system designated as Class I.

The electrical distribution equipment is comprised of three emergency generators and associated bus ducts, the 480 Volt switchgears busses 2A, 3A, 5A and 6A, the 480 Volt Motor Control Centers 35A, 36B and 36C, the diesel generator building 480 Volt bus racks, and the four static inverters and associated 118 Volt instrument busses 31, 32, 33 and 34. The 480 Volt switchgear consists of four bus sections designated 2A, 3A, 5A and 6A and is used to supply power to motors 400 horsepower and below, and to the Motor Control Centers.

Emergency power to the 480 Volt buses is provided by three diesels each of which is sized so that any two of the three units will have sufficient capacity to supply the engineered safeguards load.

The 480 Volt bus sections can be configured in the following manner to provide assurance of operation under all design conditions:

a) All four busses can be supplied from the 13.8 kV Buchanan substation.

b) All busses can be manually interconnected by tie breakers.

c) All four busses can be supplied by the gas turbine generator at Unit 1 or by the turbine generators at the Buchanan substation via the 13.8 kV gas turbine substation between Units 2 and 3.

The four 480 Volt busses are located on the 15' elevation in the Control Building. Busses 5A and 2A are housed in switchgear No. 31; busses 6A and 3A are in switchgear No. 32. Control Power is from 125 V DC Distribution Panel 31 (Bus 2A, 5A), and 125 V DC Distribution Panel 33 (Bus 3A, 6A).

MCC-36A, and 36B are located in the Primary Auxiliary Building, on elevation 55', and are Class I Components.

For a more detailed description of the Electrical System see System Description No. 27.

22-1

- 23.0 Auxiliary Feedwater System


The Auxiliary.Feedwater System (AFS) is required to Achieve and Maintain Reactor Subcriticality (FT-1)*. The AFS accomplishes this by removing heat from the Steam Generators. A description of the AFS is provided in System Description No. 21.

Scope

For the purposes of this study, AFS includes:

- Auxiliary Feedwater Pumps

- Condensate Storage Tank

- Piping, Valves and Instrumentation and Electrical Components as indicated on the corresponding disciplines Nonconnected and Interconnected Matrices.

The boundaries of this study for AFS are shown on Sheet 1; FD 5209.14, Sheet 1, along with MSS, and with FWS.

* Functional Table 1. See Volume I Tab entitled

flow diagrams FD 5209.23, FD 5209.15, Sheet 1, along

Functional Tables

23-1

23.1

23.2


INTERACTION SUMMARY

1.0 CONTENTS

This section contains a summary of the results of the systems interaction

study, followed by 10 separate sections for area specific interactions, induced operator error interactions, systems interactions, commonality and auxiliary feedwater system interactions as listed in Table S-1. Explanations are included with tables and figures where necessary. The summary itself includes a brief discussion of the study results for nonconnected and interconnected interactions with summary tables and recommended methods for resolution of the discrepancies found during the study.

The detailed results of the study (i.e., matrices, data sheets, evaluations) for nonconnected and interconnected interactions are located by system and responsible discipline in Volumes 2 through 23. The information for each system in those volumes is arranged as follows:

a) System description (preceding the mechanical discipline only)

b) Tabulation of results (b through k by discipline)

c) Interconnected interaction matrix

d) Interconnected interaction evaluation notes

e) Interconnected interaction evaluation sheets

f) Nonconnected interaction matrix

g) Nonconnected interaction evaluation notes

h) Nonconnected interaction evaluation sheets

i) FMEA evaluation notes

J) Interconnected interaction FMEA's

k) Nonconnected interaction FMEA's

1) System background drawings (following data sheets from all 3 disciplines)

m) System boundary diagrams

Volme 24 contains evaluation sheets for Interaction Credibility Evaluations CEIC), Span Evaluations, and Induced operator Error Analysis CIOEA) and Volume 25 contains the study photographs.

2.0 INTERCONNECTED INTERACTIONS

There were a total of 152 interconnected interactions identified during the initial review of system nonsafety-to-safety process couplings, all from the mechanical discipline (piping, valves, tanks, etc.) Although component couplings under the responsibility of both the electrical and the instrumentation & controls disciplines were reviewed in the same detail as the mechanical component couplings, no potentially unacceptable interconnected interactions were found by those disciplines.

Of the 152 interactions, 130 were resolved by initial evaluation and 12 were resolved by failure modes and effects analysis. A breakdown of the interconnected interactions is contained in Table S-2, and as can be seen from it, only 3 systems have interconnected interactions yet to be resolved. To further review the potentially unacceptable interactions, refer to the individual evaluation sheets which are listed by system in Volumes 2 through 23.

Note: On the initial evaluation sheets for interconnected interactions, a space was made to record the Method of Detection, which was found later, to have no value with regard to the identification or evaluation of the Interaction. Disregard all entries in that space.

3.0 NONCONNECTED INTERACTIONS

There were a total of 6221 nonconnected interactions identified during the plant walkdown inspections of the 23 primary and secondary systems. Of these, 2296 were initially evaluated as potentially unacceptable, to be further evaluated by FMEA, span evaluation, or EIC. Table S-3 is a summary listing of these interactions by system and responsible discipline with totals in both categories. The acceptance criteria utilized during the initial search was by design very conservative, so that little time would be spent in detailed evaluations which could later be performed on a selective basis. For this reason, it was expected that further analyses using more extensive criteria and sophisticated techniques would reduce the 2296 potentially unacceptables significantly. Since there were fewer sources than potentially unacceptable interactions, it was additionally expected that in-depth analyses of the supports of a few selected sources (i.e., span evaluations) would reduce the number of remaining interactions even further.

The 2296 potentially unacceptable interactions were analyzed using the 3 more detailed techniques, FMEA, span evaluation, and EIC. AB expected, the number was reduced by 73Z to 620 for which there are approximately 200 sources. Summaries of interactions by analysis technique, system, and discipline are contained in Tables S-4, S-5, and S-6 with totals in S-7. As can be seen in the tables, most of the resolutions resulted from analyzing target worth with FMEA's, (41% of the total). Following these, EIC's and span evaluations were equally effective in resolving the interactions, disposing of 12% and 13%', respectively of the total. Since the source support span evaluation was the most complex of the analysis techniques, it was performed last to reduce the number of times that it was applied. The low number of resolutions by span evaluation, 13%, reflects this. In addition, repairs accomplished as part of the auxiliary feedwater system corrections, discussed in summary Section 10.0, accounted for resolutions of 152 Interactions, or 7% of the total number of potentially unacceptable interactions.

4.0 DISCUSSION OF FINDINGS

In order to develop an understanding of the overall value of the results of this study and of the individual importance of each remaining potentially unacceptable interaction with respect to the safe operation of the plant, two characteristics of the study should be taken into consideration; they are the uniqueness of the systems interaction process and the conservatism built into the study.

First, it should be noted that the application of a systems interaction analysis on such a scale and under the criteria developed for this study and approved by the NRC and the ACRS, has never been performed. There are no previously written detailed guidelines on how the study should be conducted or how the results should be handled.

Second, the conservatism built into the study results in a more rigorous review of interactions than that performed during the original licensing basis. The scenarios postulated for the majority of the interactions are based on a chain of events whose chance of occurrence is remote. Not only must the nonsafetyto-safety interactions take place, but they must occur within the same time frame as an accident condition which requires the operation of selected safety systems and the interactions must have a detrimental effect on the operation of those systems.

5.0 INTERACTION DISPOSITION

Due to the uniqueness of this study, there is no prescribed approach for dispositioning the potentially unacceptable interactions found by the study effort (including interconnected, nonconnected, induced operator error, and area specific interactions). There are, however, a number of techniques, which if applied under a predetermined schedule, could resolve the remaining interactions in an efficient manner. Based on the variety of the results, it appears that the optimum application of these techniques will occur after an overall assessment of the results has been conducted which accounts for such parameters as relative importance to plant safety, operating schedule, cost of repair versus cost of evaluation, and commonality. Following the overall assessment, any combination of the techniques listed below may be utilized:

- Repair

a) Relocation of sources

b) Relocation of targets

c) Shielding of sources

d) Replacement or addition of qualified components

e) Resupport

- Dispositioning

a) Extensive Licensing evaluations with respect to individual system and component operation

b) Review of 1OCFR50 Appendix R modifications made or being planned

c) Explicit identification of cables in safety related trays for further detailed evaluation

d) PRA

e) Type testing or qualification

- Administrative

a) Procedural changes

b) Monitoring programs


SUMMARY SECT IONS

SECT ION

1.0

2.0

3.0

4.0

TABLE S-1

TITLE

SYSTEMS INTERACTIONS INDUCED BY EXTERNALLY GENERATED MISSILES

SYSTEMS INTERACTIONS INDUCED BY INTERNALLY GENERATED MISSILES & PIPE WHIP

SYSTEMS INTERACTIONS INDUCED BY SEVERE ENVIRONMENT (OTHER THAN FLOODING) RESULTING FROM NATURAL PHENOMENA

SYSTEMS INTERACTIONS INDUCED BY SEVERE ENVIRONMENT WITHIN CLASS I STRUCTURES

SYSTEMS INTERACTIONS INDUCED BY THE EFFECTS OF FLOODING DUE TO NATURAL PHENOMENA

SYSTEMS INTERACTIONS INDUCED BY THE EFFECTS OF INTERNALLY GENERATED FLOODING

CENTRAL CONTROL ROOM REVIEW (INCLUDING HVAC/INSTRUMENT AIR)

INDUCED OPERATOR ERROR INTERACTION SUMMARY

SYSTEMS INTERACTIONS COMMONALITY

AUXILIARY FEEDWAT ER SYSTEM INTERACTIONS

6.0

7.0

8.0

9.0

10. 0

-INDIAN POINT 3 NUCLEAR POWER PLANT SYSTEMS INTERACTION STUDY

SUMMARY OF INTERCONNECTED INTERACTIONS FINAL EVALUATIONS

INTERACTIONS TOTAL ACCEPTABLE BY INTERACTIO

NUMBER OF INITIAL ACCEPTABL INTERACTIONS EVALUATION BY FMEA

NS E INTERACTIONS

REMAINING

-TABLE S-2

SYSTEM NUMBER

TOTAL 152 130 12 10

(86%) (8%) (6%)


SUMMARY OF NONOONNECTED INTERACTIONS INITIAL EVALUATIONS

MECHANICAL INSTRUMENTATION & CNTL ELECTRICAL 70TAIS SYSTEM POTENTIALLY TOTAL NO. OF POTENTIALLY TOTAL NO. OF POTENTIALLY TOTAL NO. OF POTENTIALLY TOTAL NO. OF r POTENTIAL

UNACCEPTABLE INTERACTIONS UNACCEPTABLE INTERACTIONS UNACCEPTABLE INTERACTIONS UNACCEPTABLE INTERACTIONS UNACCEPTABLE

1 ROD ONTROL 0 0 0 0 2 6 2 6 33

2 SAFETY INJECTION 17 133 37 56 76 227 130 416 31

3 CHEMICAL VOL CNTL 232 665 125 152 79 201 436 1018 43

4 RESIDUAL HT REHL 13 99 22 30 34 98 69 227 30

5 REACTOR OOLANT 1 181 32 50 14 94 47 325 14

6 CONTAINMENT SPRAY 6 85 11 18 6 54 23 157 15

7 CONTROL RM VENT 39 42 1 1 26 67 66 110 60

8 SERVICE WATER 131 520 137 165 116 150 384 835 46

9 CDMPONENT CLG WTR 195 800 53 69 41 89 289 958 30

1O N 2 BACKUP 0 0 64 96 0 0 64 96 67

11 DIESEL GENERATOR 76 80 25 25 23 80 124 185 67

12 DC POWER 0 0 0 0 17 53 17 .53 32

13 CONT ISOLATION 51 84 0 0 33 44 84 128 66

14 MAIN STEAM 4 48 37 99 16 33 57 180 32

15 NS 6 FW ISOLATION 0 6 40 40 9 18 49 64 77

16 H 2 RECOMBINER 51 85 41 70 12 16 104 171 61

17 REACTOR PROTECTION 0 0 42 50 23 58 65 108 60

18 ELECTRICAL TRAYS 0 0 0 0 80 325 80 325 25

19 HVAC 0 4 0 1 5 17 5 22 23

20N 2 ACTUATION 0 26 0 14 0 6 0 46 0

21 CONTAINMENT RECIRC 8 12 0 1 0 25 8 38 21

22 ELECTRICAL DIST 0 0 0 0 0 131 0 131 0

23 AUXILIARY FEEDWATER 52 413 91 107 50 102 193 622 31

TOTALS 876 3283 758 1044 662 1894 2296 6221 37

TABLE S-3

0

INDIAN POINT 3 S EAR POWER PLANT SYSTEMS INT CTION STUDY

SUMMARY OF NONCONNECTED INTERACTIONS FINAL EVALUATIONS

MECHANICAL DISCIPLINE

SYSTEtl NO. OF ACCEPT BY ACCEPT BY ACCEPT BY ACCEPT BY TOTAL % TOTAL No. INTERACT FMEA EIC SPAN EVAL REPAIR ACCEPT ACCEPT REMAINING

1 0 0 0 0 0 0 - 0 2 17 6 2 7 0 15 88% 2 3 232 160 14 34 0 208 80% 24

4 13 12 0 0 0 12 92% 1 5 1 0 0 1 0 .1 100% 0 6 6 2 4 0 0 6 100% 0

7 39 6 8 0 0 14 36% 25 8 131 59 32 2 0 93 71% 38 9 195 63 18 49 0 130 67% 65

10 0 0 0 0 0 0 - 0 11 76 31 6 13 0 50 66% 26 12 0 0 0 0 0 0 - 0

13 51 28 13 0 0 41 80% 10 14 4 0 2 0 0 2 50% 2 15 0 0 0 0 0 0 - 0

16 51 22 0 18 0 40 78% 11 17 0 0 0 0 0 0 - 0 18 0 0 0 0 0 0 - 0

19 0 0 0 0 0 0 - 0 20 0 0 0 0 0 0 - 0 21 8 0 6 0 0 6 75% 2

22 0 0 0 0 0 0 - 0 23 52 1 7 24 19 51 98% 1

TOTALS 876 390 112 148 19 669 76% 207

TABLE S-4

0INDIAN POINT 3 NUC POWER PLANT SYSTEMS INTER ON STUDY


I & C DISCIPLINE

SYSTEM NO. OF ACCEPT BY ACCEPT BY ACCEPT BY ACCEPT BY TOTAL % TOTAL No. INTERACT FHEA EIC SPAN EVAL REPAIR ACCEPT ACCEPT REMAINING

1 0 0 0 0 0 0 r 0 2 37 8 6 4 0 18 49% 19 3 125 118 1 0 0 119 95% 6

4 22 9 3 3 0 15 60% 7 5 32 1 4 10 0 15 47% 17 6 11 11 0 0 0 11 100% 0

7 1 0 0 0 0 0 0% 1 8 137 58 22 0 .0 80 58% 57 9 53 33 2 3 0 38 72% 15

10 64 4 4 1 23 32 50% 32 11 25 19 0 3 0 22 88% 3 12 0 0 0 0 0 0 - 0

13 0 0 0 0 0 0 - 0 14 37 4 2 0 1 7 19% 30 15 40 40 0 0 0 40 100% 0

16 41 14 8 1 0 23 56% 18 17 42 4 0 12 0 16 38% 26 18 0 0 0 0 0 0 - 0

19 0 0 0 0 0 0 - 0 20 0 0 0 0 0 0 - 0 21 0 0 0 0 0 0 - 0

22 0 0 0 0 0 0 - 0 23 91 0 0 8 81 89 98% 2

TOTALS 758 323 52 45 105 525 69% 233

TABLE S-5

INDIAN POINT 3 * EAR POWER PLANT SYSTEMS INTERACTION STUDY


ELECTRICAL DISCIPLINE

SYSTEM NO. OF ACCEPT BY ACCEPT BY ACCEPT BY ACCEPT BY TOTAL 2 TOTAL No. INTERACT FMEA EIC SPAN EVAL REPAIR ACCEPT ACCEPT REMAINING

1 2 0 1 0 0 1 50% 1 2 76 44 6 18 0 68 89% 8 3 79 68 1 3 0 72 91% 7

4 34 .3 9 9 0 21 62% 13 5 14 0 13 0 0 13 93% 1 6 6 6 0 0 0 6 100% 0

7 26 5 5 0 0 10 38% 16 8 116 48 8 0 0 56 48% 60 9 41 7 5 23 0 35 85% 6

10 0 0 0 0 0 0 - 0 11 23 4 3 8 0 15 65% 8 12 17 0 6 0 0 6 35% 11

13 33 33 0 0 0 33 100% 0 14 16 0 8 0 0 8 50% 8 15 9 9 0 0 0 9 100% 0

16 12 0 0 12 0 12 100% .0 17 23 0 3 10 0 13 57% 10 18 80 6 29 14 0 49 61% 31

.19 5 5 0 0 0 5 100% 0 20 0 0 0 0 0 0 - 0 21 0 0 0 0 0 0 0

22 0 0 0 0 0 0 - 0

23 50 0 12 10 28 50 100% 0

TOTALS 662 238 109 107 28 482 73% 180

TABLE S-6

INDIAN POINT 3 NUCLj POWER PLANT SYSTEMS INTERAWN STUDY

SU MARY OF NONCONNECTFD INTERACTIONS FINAL EVALUATIONS

ALL DISCIPLINES

SYSTEM NO. OF ACCEPT BY ACCEPT BY ACCEPT BY ACCEPT BY TOTAL % TOTAL No. INTERACT FMEA EIC SPAN EVAL REPAIR ACCEPT ACCEPT REMAINING

1 2 0 1 0 0 1 50% 1 2' 130 58 14 29 0 101 78% 29 3 436 346 16 37 0 399 92% 37

4 69 24 12 12 0 48 70% 21 5 47 1 17 11 0 29 62% 18 6 23 19 4 0 0 23 100% 0

7 66 11 13 0 0 24 36% 42 8 384 165 62 2 0 229 60% 155 9 289 103 25 75 0 203 70% 86

10 64 4 4 1 23 32 50% 32 11 124 54 9 24 0 87 70% 37 12 i7 0 6 0 0 6 35% 11

13 84 61 13 0 0 74 88% 10 14 57 4 12 0 1 17 30% 40 15 49 49 0 0 0 49 100% 0

16 104 36 8 31 0 75 72% 29 17 65 4 3 22 0 29 45% 36 18 80 6 29 14 0 49 61% 31

19 5 5 0 0 0 5 100% 0 20 0 0 0 0 0 0 - 0 21 8 0 6 0 0 6 75% 2

22 0 0 0 0 0 0 - 0 23 193 1 19 42 128 190 98% 3

TOTALS 2296 951 273 300 152 1676 72.9% 620

TABLE S-7



1.0 SYSTEMS INTERACTIONS INDUCED BY EXTERNALLY GENERATED MISSILES

A review of the Indian Point 3 Final Safety Analysis Report (FSAR) Update, Rev. 0 and Safety Evaluation Report (SER), dated September 21, 1973, was conducted to determine if equipment required to accomplish one or more of the basic safety functions is protected from the effects of externally generated missiles.

As stated in the FSAR, equipment is protected from tornado missiles by either enclosure in a missile proof structure or by redundancy of equipment. Systems and equipment protected by enclosure or by redundancy are listed in FSAR Section 16.2.2.

One criterion used in this study postulates a LOCA concurrent with a single natural event, in this case a tornado. Using this criterion, safety injection would not be available since the refueling water storage tank is not in a tornado missile proof structure. Another study criterion postulates simultaneous multiple missiles which would render the service water system unavailable since the pump motors are unprotected.

Although unacceptable per the search criteria of this study, both the aforementioned interactions are considered acceptable based on the plant licensing criteria, as restated below:

I A tornado will not cause a Loss of Coolant Accident.

II A tornado will not impair the ability to safely shut the plant down.

III A tornado, following a Loss of Coolant Accident, will not impair the long term safety of the plant.

IP3 licensing basis postulates missiles occurring individually (i.e., multiple missiles are not considered).

A potentially unacceptable interaction existed with the PAD HVAC System. A .portion of the exhaust duct exists onto the PAB roof before entering the Fan House. This portion of duct is not missile protected nor does it have a redundant backup. However, it was subsequently found to be acceptable since the PAB HVAC System has been determined not to be required for at least 24 hours based on a letter from J R Slotterbach, UEC, to C Pratt, PASNY, andI B Brandenburg, Con Edison, dated 10/6/82.

Missiles produced by the turbine generator have not been evaluated in this study.



2.0 SYSTEMS INTERACTIONS INDUCED BY INTERNALLY GENERATED MISSILES & PIPE WHIP

During the walkdown segment of this study, observations were made relative to the potential for the generation of missiles from nonsafety related pressurized systems and nonsafety related rotating machinery and for pipe whip from seismic Category III systems, all of which could inflict damage to nearby safety related equipment. Also, the Indian Point 3 FSAR was reviewed to ascertain if any measures had been taken to prevent or mitigate the effects of missiles generated from nonsafety related sources. It was found in the review that all pressurized tanks (whether safety or nonsafety related) which were part of the original plant design, were evaluated in the response to FSAR Question 9.32 for their effect on seismic Category I equipment and structures. Although specifically included in the response, nitrogen cover gas bottles and replacement bottles not yet attached to the supply manifolds were evaluated during the walkdown inspections. The supports appeared to be insufficient to prevent overturning and consequent damage to the bottle valves which could result in unacceptable missiles.

In addition, an unacceptable interaction was discovered in the Central Control Room. An oxygen bottle used for emergency breathing is unrestrained allowing movement and tipover. If the valve/regulator on the tank were to be broken it could become a missile (see Photograph 1-322).

Pipe whip and jet impingement were addressed in the Analysis of High Energy Lines dated May 9, 1973. The Analysis included an evaluation of the effects of rupture in high energy pipe lines. However, the auxiliary steam line in the control building was not considered. Note that although the auxiliary steam system is by definition high energy with an operating temperature of 300 0F, the pressure is very low (50 psig). Therefore, spontaneous rupture was not postulated. Thru wall cracks were postulated and taken into account in the study of internal environmental effects and flooding.

2-1



3.0 SYSTEMS INTERACTIONS INDUCED BY SEVERE ENVIRONMENT (OTHER THAN FLOODING) RESULTING FROM NATURAL PHENOMIENA

A review of the Indian Point 3 Final Safety Analysis Report (FSAR) Update, Rev. 0, and the NRC (then AEC) Safety Evaluation Report (SER) was conducted to determine to what extent the structures, systems and components associated with safety related systems were protected against the effects of severe environment due to natural phenomena. As stated in Section 16.2 of the FSAR, the effects of tornado wind loadings and depressurization were considered in the design of all Category I structures.

Most safety related components are protected by enclosure within tornado proof structures. Those safety related components not located in tornado proof structures are protected by redundancy. It is not clear from the ESAR if outdoor safety related components themselves are designed to withstand tornado loads. However, this is acceptable for the reasons listed in summary Section 1.0.



4.0 SYSTEMS INTERACTIONS INDUCED BY SEVERE ENVIRONMENT WITHIN CLASS I STRUCTURES

Internally produced environmental effects were studied from the standpoint of: 1) pipe rupture of a nonsafety related line resulting in elevated temperature and humidity conditions and, 2) a fire effecting the environment of adjacent compartments. Data collected for the flooding analysis (summary Section 6.0) was used to ascertain if the environment could be adversely affected by rupture of a nonsafety related/nonseismic line or component.

Systems within the Reactor Containment were considered acceptable since any environmental effect generated by the rupture of a non-seismic line or component would be much less severe than that associated with a LOCA, main steam line break or feedwater line break. In the Primary Auxiliary Building, Diesel Generator Building, and Control Building all seismic Category III lines with operating temperatures higher than 120*F were considered potentially unacceptable since they could raise the temperature of the areas to an unacceptable level. Of particular note is the auxiliary steam piping, which runs through most compartments, including the Central Control Room. There were no unacceptable environmental induced interactions in the Auxiliary Feed Pump Building.

The Indian Point 3 Fire Protection Report, Revision 1, dated April 1977, was reviewed for information regarding fire effects on adjacent fire zones. Based on this review and on the plant walkdown inspections, there were no unacceptable interactions.



5.0 SYSTEMS INTERACTIONS INDUCED BY THE EFFECTS OF FLOODING DUE To NATURAL PHENOMENA

A review of the Indian Point 3 Final Safety Analysis Report (FSAR) Update, Rev. 0 and the NRC (then AEC) Safety Evaluation Report (SER), was conducted to determine to what extent the structures, systems and components associated with safety related systems were protected against the effects of flooding due to natural phenomena.

It was established that the most severe flooding condition at the site corresponds to a water elevation of 15 ft above mean sea level (MSL). As stated in the SER, this elevation is lower by three inches than the critical elevation at which water could start seeping into the lowest plant buildings. The staff concluded in their report that under the most extreme conditions the flood level could reach a level of 15.0 ft MSL, exclusive of wind-generated wave action.

In the event of wind-generated wave action in conjunction with extreme flooding, conditions, the plant will still be protected. Wind-generated action could raise the flooding level above plant grade in the vicinity of the service water pumps. In this unlikely event, the plant will be shut down in accordance with the Technical Specifications, and the service water pump areas will be protected. Other areas, such as the diesel generator area, will not require additional protection from the wind-generated waves due to wave dissipation on land.

Consequently, the combination of the elevation of the plant structures, and the Technical Specification requirements on plant operation and service water pump protection, result in acceptable conditions to protect the plant against f looding .



6.0 SYSTEMS INTERACTIONS INDUCED BY THE EFFECTS OF INTERNALLY GENERATED FLOODING

During the systems interaction study, the effects of internally generated floods were evaluated. Several types of flooding were considered, including the individual random failure of liquid containing components, the seismically induced failure of liquid containing components and the inadvertent operation of the fire protection systems.

Individual failure of tanks and piping as well as inadvertent initiation of fire protection systems were addressed in the Indian Point 3 Final Safety Analysis Report (FSAR) Update, Rev. 0, Chapters 9.6.2.2, 9.6.2.3 and 16.1. Seismically induced flooding was evaluated as part of the system interaction study.

Internal flooding, induced by a seismic event, was postulated to originate from seismic Category II and III piping systems. Although Category II piping can withstand the loadings of the design basis earthquake (see response to FSAR Question 5.24), Category II piping systems were not part of the Systems Interactions Study since they did not support the four basic safety goals. Consequently, it was assumed that one system interaction per flood area could occur for a Category II piping system that would result in a breach of the piping integrity. All Category III piping was postulated to break.

The Reactor Containment was not considered to have unacceptable interactions caused by a seismically induced flood since the flooding rate was found to be negligible compared to the design basis of the containment sumps, that basis being the largest primary or secondary system line break. Descriptions of the methods used for evaluating postulated flooding in the Primary Auxiliary Building, Diesel Generator Building, Control Building and Auxiliary Feedwater Pump Room are described in the following paragraphs.

6.1 Primary Auxiliary Building

Due to the complicated nature of the Primary Auxiliary Building's arrangement a flow chart of the postulated flooding is provided as Attachment 6-1.

Since the Indian Point 3 - Fire Protection Safety Report identified zones containing safety related components, it was used to develop the Flooding Analysis Diagram (FAD), showing the flow pattern of liquid through the flood areas. Each flood area was determined by its fluid level and contained within it one or more fire zones. Liquid from the non-Category I lines that discharged into the flood areas were assigned specific patterns of flow. By tracking the cascading flow through flood areas, the boundaries were defined and the final fluid levels were calculated.

To develop the Flooding Analysis Diagram and the final fluid levels, the following criteria were used:

1) 'Category I piping integrity does not fail.

2) Category II & III piping fails by guillotine break.

3) Category II &III piping fails in only one pl ace per flood area.

4) Fluid removal by the drainage system is considered negligible since adequate capacity of the system at the time of the event can not be assured.

5) Flow to separate areas. is divided equally.

6) Connected fire zones flood at an equal rate.

7) All steam discharge is assumed to immediately condense to water for the purpose of determining fluid level.

8) All gas and normally dry lines do not add to the flood.

9) Capacity of auxiliary steam and city water lines is assumed to discharge for ten minutes within the building before the inventory within the system is exhausted. This assumption was made since the external source of these systems is non-seismic and a break outside of the PAB is be more likely than a break within the PAR.

10) All piping from pumps discharges at the runout flow or, if not available, twenty percent above design flow.

11) All lines discharge at a constant rate until their source is

depleted.

12) Outside doors are not watertight.

13) Duplicate pipe breaks are eliminated when combining flood areas.

14) All curbs are four inches high.

15) The calculated floor area represents the surface area of the flooding fluid.

16)* The calculations are based on all grating and check plate being in

their design positions.

Using these criteria it was ascertained that fluid can accumulate as high as the top of the curb in those rooms where there is Category III piping or equipment. All other leakage, i.e., leakage in rooms that are not curbed or overflow from rooms with curbs, will cascade to the lower elevations of the PAB.

There are two areas of accumulation. One area is the pipe trench below the penetration area. In the pipe trench area, only the service water pipe trench within fire zone 61A will have appreciable flooding. Since only piping is contained within the area, it is unaffected by the flooding. A ventilation duct penetration and pipe openings in fire zone 5A preclude water accumulation in fire zones 5A and 62A as well as areas of fire zone 61A other than the service water pipe trench.

The second area is the 15 foot elevation which includes the RHR pump rooms. Analysis has shown that an accumulation of less than 3'-5" will occur after almost 10 hours of discharge from failed Category II and III pipes. Performance of the RHR pumps is not affected by flooding if the water level is less than 4 feet; therefore, this flood can be tolerated.

6.2 Diesel Generator & Control Buildings

The analysis and assumptions for the Diesel Generator and Control Buildings were similar those for the PAB except that a flow chart was not required since the flood areas did not combine. Credit was taken for the drainage system.

An analysis of flooding caused by failure of Seismic Category III piping in the turbine hall was made previously to determine if such flooding could affect equipment in the Control Building. This scenario is discussed in the FSAR Update Chapter 16.1.

In the SIS evaluation of flooding in the Control Building, it was determined that for the two pipes which may cause flooding, the auxiliary steam and the city water lines in the Central Control Room (Fire Zone 5), the postulated flooding will not produce enough water accumulation to affect any safety related component. For the CCR ventilation room (fire zone 35A), the flood level will reach approximtely 3-1/2 feet before the drainage system begins to reverse the rise in level. This can interact with the ventilation system of the Central Control Room since the air handling units are mounted on the floor.

The Diesel Generator Building has no interactions caused by flooding.

6.3 Auxiliary Feed Pump Building

Only one non-siesmic line will produce flooding in the pump room the hotwell makeup line. Flood waters originate from the condensate storage tank since drainage from the hotwell into the Auxiliary Feed Pump Building is not possible. The seismic Category III portion of the line can be isolated from the Category I portion of the line by a butterfly valve that is designed to close when a specified level is' reached in the condensate storage tank. This is to guarantee sufficient water inventory to support auxiliary feedwater system functions. A flood emanating from this line can, however, affect the operation of the auxiliary feedwater system since detection of a line rupture may not be apparent in sufficient time to isolate such a rupture. On-going studies are being performed to upgrade the design of this line such that flooding will be prevented or mitigated.



7.0 CENTRAL CONTROL ROOM REVIEW

A walkdown inspection of the Central Control Room (CCR) was conducted to identify nonconnected interactions, including those caused by pipe whip, internally generated missiles, environmental conditions, and seismic events. Interactions from the first three causes are evaluated in other sections of the summary and seismically induced interactions are evaluated below.

The CCR ceiling and lighting were inspected and judged by the walkdown team to

be adequately supported (Photographs 1-320 and 1-321). Three storage cabinets and 2 file cabinets also identified during the walkdown inspection were evaluated as potentially unacceptable, in that they are capable of toppling

and possibly causing damage to safety related equipment. The location of those cabinets is shown on CCR background drawings, sheets 70 and 71. Further

identification can be made on Photographs 1-323, 1-324, and 1-325.

The CCR HVAC system was examined in detail with respect to interconnected and

nonconnected interactions. Four interconnected and fourty-two nonconnected

interactions were identified and evaluated under the study criteria as potentially unacceptable and are individually listed with all system 7 data

sheets in Volume 9. Data sheets on the interactions of the supporting systems

which may affect the CCR HVAC system (service water, electrical trays, and electrical distribution) are contained with their respective systems in

Volumes 2 through 23. Most nonconnected interactions affected ducts,

electrical power supplies, and dampers, which if induced together could

jeopardize the operation of the system in normal and incident modes.

Instrument air tubing, except where connected to dampers and other targeted

components, was either shielded or routed above potential sources. These

shielding and routing conditions extended outside of the HVAC room to the

piping connection at the 3 inch instrument air header. The remainder of the instrument air system was observed as not being shielded or routed in the same

manner. A significant break in an unprotected part of the system could cause enough loss of pressure to partially reposition dampers and control valves and

alter temperature control. Complete failure of instrument air pressure would cause dampers and cooling water control valves to return to their "failed

positions" due to air pistons or diaphragms being overcome by spring pressure. Specifically, dampers Dl/D2 and Fl/F2 would fail closed shutting

down both A/C units and both filter units fans. Dampers A and B would fail

closed, isolating the outside air inlet. Damper C would fail open as if in

the incident mode. The cooling water control valves to the A/C units would fail closed. Since the instrument air controls of the CCR HVAC system

normally bleed small amounts of air, check valves and air relays within the

system would not maintain dampers in position on loss of air. Furthermore, the failure positions of the dampers is such that neither normal nor incident

modes would be possible if fans were to be operated through control circuit overrides. Instrument air control is therefore considered as a potentially

unacceptable interconnected interaction with regard to CCR HVAC.



SUMMARY OF INDUCED OPERATOR ERROR ANALYSES

TOTAL NUM1BER OF ANALYSES

ACCEPTABLE ANALYSES

NUMBER OF POTENTIAL IOE INTERACTIONS

TOTAL

112

SYSTEM NUMBER

TOTAL 112



9.0 SYSTEMS INTERACTIONS COMIONALITY

The 620 nonconnected interactions which were evaluated by all 3 methods (FMEA, EIC, and span evaluation) as being unacceptable based on the criteria of this study, are listed following this page by fire zone and source.

Each interaction is identified by its unique number, source name, target name, and target functional description. Also included is the identification of the safety goal(s) affected by each interaction. The commonality summary may be used to assist in determining the methods of interaction resolution, their estimated costs, and their priority. Interconnected and area specific interactions are listed in other sections.

Fire Zone YARDShe1

NEW YORK POWER AUTHORITY INDIAN POINT 3 SYSTEMS INTERACTION STUDY

SUMMARY OF UNACCEPTABLE INTERACTIONS

Source NameInteraction Number Target Name

Safety Description Goal

Turbine Building L-23-63-1 6" - 1080 Return Line to Condensate Storage Tank

SAFETY GOALS

1. Achieve & Maintain Reactor Subcriticality. 2. Remove Decay Heat. 3. Maintain Reactor Coolant Pressure Boundary. 4. Maintain Containment Integrity-

Sheet 1

Fire Zone BIT





Light

Roof Drain

1-2-75-2

1-2-75-1

1-2-78-1

E-2-398-1 399-1 400-1 401-1 402-1

TIC 918

TIC 918

TW 917

Flex CNDS Box TD8 3/4" CND 7GQ Box ZQ9 3" CND 7BW

Boron Injection Tank Heaters Temperature Indication

Boron Injection Tank Heaters Temperature Indication

Boron Injection Tank Heaters, backup temperature control, to indication and alarm on safeguard panel in CCR

Box TD8, 3/4" conduit 7GQ. These route cables from TC-918.

1

-1

Sheet 2

Fire Zone VALVE PIT


INDIAN POINT 3 SYSTEMS INTERACTION STUDY SUMMARY OF UNACCEPTABLE INTERACTIONS



Pr 1191 Sensing Line

PT 1190

Sensing Line

PT 1190

PTC 1191 Sensing Line

Service water discharge header pressure (nuclear) to PI 1191R in CCR

Service water discharge header

pressure (conventional) to PI 1190R

Service water discharge header pressure (conventional) to PI 1190R

Service water discharge header pressure (nuclear) to PI 1191R in in CCR

Conduit

Light

1-8-119-4

1-8-120-1

Light 1-8-120-2

Light 1-8-119-3

Tareet Name

Sheet 3

Fire Zone INTAKE STRUCTURE





8" Screen Wash Hdr.

L-8-229-1

L-8-227-1

L-8-228-1

L-8-230-1

L-8-231-1

L-8-232-1

V-8-156-1

V-8-158-1

V-8-160-1

3 "-SWN-1083

3" SWN-1081

3" SWN-1082

3" SWN-1084

3" SWN-1085

3" SWN-1086

1/2" SWN-58

1/2" SWN-58

1/2" SWN-58

3" Automatic Vent Line off S W 1-4 Pump #33 discharge header 14" #1083.

3" Automatic Vent Line off S W Pump #31 discharge line 14" #1081.





Normally Open Valve, on Instr Conn Line off 3" Automatic Vent Valve Line (off 14" S W Header). Rupture of 1/2" line also considered.



1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

Sheet 4

Fire Zone INTAKE STRUCTURE



Source NameInteraction Number Tareet Name

Safety

Descrintio GoaDescrintion

1/2" SWN-58

1/2" SWN-58


Normally Open Valve, on Instr Conn Line off 3" AutomaticVent Valve Line (off 14" S W Header). Rupture of 1/2" line also considered.

V-8-164-1

V-8-166-1

Sheet 5

Fire Zone 1





Capillary tube for TIC-627A, which controls annunciation for component cooling water pump high inlet temperature.

Sheet 6

Light 1-9-38-1 TW-627A

Fire Zone 1A





6" Steam Line

4"-#675

L-9-116-4

V-9-155-4

V-9-156-4

4-AC-320

AC-1853A 3/4"

AC-1853B 3/4"

Line from Flash Evaporator Product Cooler to Suction Line for Component Cooling Pump #31. The flow and temperature indicators on Line #320 are local only.

Normally open manually operated valve on sensing line for FI 647.

Normally open manually operated valve on sensing line for FI 647.

Tareet Name

Sheet 7

Fire ZoneShe8


SUMMARY-OF UNACCEPTABLE INTERACTIONS



THIS PAGE INTENTIONALLY BLANK

Sheet 8

Fire Zone 3



Source NameInteraction Number TarRet Name


Component cooling local flow indicator and low-flow alarm, for RHR pump #32.

Sheet 9

Light 1-9-45-1 FIC-645

Fire Zone 3A



Source NameInteraction Number Ta rget Name


2" - Line 311 E-18-125-5 125-6 125-8

Tray FC Routes Electrical Tables

Sheet 10

Tarizet Name

Fire Zone 4





Mono Rail

L-9-130-2

1-9-46-1

L-4-47-1

L-9-125-2

L-9-126-2

L-9-127-2

L-9-128-2

L-9-129-2 1-AC-658

3/4-AC-658

FIC-646

8"-AC-654

I"-AC-335

I"-AC-336

I"-AC-657

3/4"-AC-657

RHR Pump 32 discharge to RHR HXRs

Provides component cooling to RHR Pump #31 seal cooling. Flow Indicator Controller FIC 646 alarms on the ACS panel in the CCR on low flow of 12 gpm. This alarm is common to the alarm for low flow from Pump #32 cooling. In event of such an alarm, a check will have to be made at the local flow indicators todetermine which pump loop is at fault.

Return line from RHR Pump #31 cooling to Line #52.

Provides component cooling to RHR Pump #32 seal cooling. Flow Indicator Controller FIC 645 alarms on the ACS panel in the CCR on low flow of 12 gpm. This alarm is common with FIC 646 for low flow from Pump #31 cooling block. In event of such an alarm, a check will have to be made at the local flow indicators to determine which pump loop is at fault.

Seal cooling for RHR Pump #32 Cooling Block.

Return from seal cooling of RHR Pump #32 cooling block to Line #52.

RHR Pump #32 seal cooling loop.

Component cooling local flow indicator and low-flow alarm, for RHR pump #31.

1,2,4

2

1

2

2

Light

Sheet 11

Fire Zone 5A





18" - #389 1-16-51-2 PT 1065 Sensing Line

Pressure transmitter for low pressure alarm for the hydrogen supply to the hydrogen recombiner system and the chemical and volume control system.

Sheet 12

Sheet 13




Safety

Source Name Number GoalImcrintf nn

THIS PAGE INTENTIONALLY LEFT BLANK

Fire Zone

Fire Zone 9



Source Name

Heat Tracing Box

Electric Heat Tracing Conduits

2" Conduit (2)

Interaction Number

E-2-310-1 316-1

E-2-319-2

L-2-7-7

L-2-19-6

Tareet Name

Flex Cnd

Flex Cnd

3/4 SI-161

3/4 SI-270


Flexible conduits housing power 1 and control cables for MOV-1852A. 1 On inlet side of Boron Injection Tank

Routes power and control cables 1 connecting MOV-1852B on inlet side of Boron Injection Tank

Line from Safety Injection Pump (#31) Discharge Line 4" - #56 to Refueling Water Storage Tank.

Recirculation Line

3/4" line from Pump (#33) dis

charge line 4" - #550 to line 4" #550 to line 4" #16, bypassing boron injection tank line 6" #550

Sheet 14

Fire Zone





Electric Heater

Light Fixture

Light Fixture

Light Fixture

*ight Fixture

E-11-33-2

E-11-71-1 E-II-73-1

E-11-78-1 E-11-80-1

E-1 1-5 9-1

E-11-148-1

E-1I-253A-I

L-11-73-2

Panels and Conduits

Cnt 1 Pnl

Cntl Pnl

Cntl Pnl

1 1/2" DF 1050

Starter (KF3) for Fuel Oil Pump #31 for day tank Diesel Gen #31, Day Tank

Starter for Air Compr DG 31 and PB W87

Panel RG 8 and PB W23 for

LCV-1207

Control panel PP 9 for Diesel

Generator #31

Control panel PQ 1 for Diesel Generator #32

Control panel PQ 2 for Diesel

Generator #33

Normal fill line for fuel oil

Day Tanks. The line accepts

discharge from each of the F 0 Transfer Pumps #31, 32, & 33

Loud Speaker

L-1 1-7 8-1

L-I1-79-1

L-1 1-80-1

V-11-84-2

V-I 1-84-1

2" DF-1077

3/4" DF-1102

3/4" DF-1104

Diesel Engine & Exhaust

Diesel Engine & Exhaust

Vent to atmosphere for Fuel Oil Day Tank #31

Fuel return line from D G #31 to Fuel Oil Day Tank #31. For venting D G system during starting, for system priming

Fuel oil supply line from Day Tank #31 to D G #31. To supply fuel oil to D G

31 Diesel Engine #31 & Exhaust System

31 Diesel Engine #31 & Exhaust System

1-4

1-4

1-4

1-4

10 Sheet 15

Fire Zone10Set 6





Emergency Light

Space Heater

Light s

Light Fixture

V-11-84-3

V-11-84-4

1-11-12-1

V-8-216-1

V- 8-218-1

V-8-2 21-1

Diesel Engine 31 Diesel Engine #31 & Exhaust & Exhaust System

Diesel Engine 31 Diesel*Engine #31 & Exhaust & Exhaust System

Generator Panel

SWN-77 1/2" Vent

SWV- 63 3/4** sRN

SWN-77 1/2" Vent

Starting and controlling of Diesel Generator #31

1/2" valve on vent connection to 6" inlet cooling water line to diesel generator. Venting valve. Rupture of vent line also considered.

3/4" Relief valve for 6" inlet cooling water line to diesel generator. Rupture of its connection to 6" line is also considered.

1/2" valve on vent connection to 67 inlet cooling water line to diesel generator. Venting valve. Rupture of vent line also considered.

1-4

1-4

10 Sheet 16

Fire Zone 11





Rod Control Devices

Control devices on Rod Control Cabinet front and rear are

required for manipulation of control rods and their automatic control.

Sheet 17

Light E-l- 33-1

Fire Zone 12, 13

6,Sheet 18


SUMMARY.OF UNACCEPTABLE INTERACTIONS

Source Name

Electric Heating Unit

1 1/2" Cnd to Heater

Lights

Elect Heating Unit

Lights

1/2" Lighting Cnd

.1/4" Cnd to Heating Unit

Interaction Number

E-12-1-1 6-1

E-12-1-2 6-2

E-12-1-4

E-12-7-1 -1I

E-12-7-2

E-12-7-4

E-12-7-5 11-5

Target Name

Battery

Battery

Battery

Battery

Battery

Battery

Battery

Des er nt lon

Cable

Cable

Cable

Cable

Cable

Cable

Cable

Battery 31 DC power

Battery 31 DC power

Battery 31 DC power

Battery 32 DC power

Battery 32 DC power

Battery 32 DC power

Battery 32 DC power

Safety rcr m

which

which

which

which

which

which

which

supplies

supplies

supplies

supplies

supplies

supplies

supplies

1-4 1-4

1-4 1-4

1-4

1-4 1-4

1-4 1-4

1-4

1-4 1-4

Descrintion

Fire Zone 14



Source Name

Lights

Lights

Interaction Number

E-18-102-3

E-18-103-3

E-18-108-1

Target Name

Tray

Tray

Tray


Electric Tray CD

Electric Tray JD, DD, FD

Electric Tray CD

Sheet 19

Fire Zone

NEW YORK POWER AUTHORITY INDIAN POINT 3 SYSTEMS INTERACTION STUDY SUMMARY OF UNACCEPTABLE INTERACTIONS



3" Conduit 1-16-37-3

1-16-39-3

Communication Box

Space Heater, Steam & Condensate Piping

2" - 560

1-16-37-8

1-16-39-8

L-16-1-2

L-9-170-3

Pr-3

PT- 3 Signal Line

PT-3

PT- 3 Signal Line

1"-WD-572

1"-AC-248

Pressure transmitter for remote pressure indication and low pressure alarms for the oxygen supply to the hydrogen recombiner system.

Pressure transmitter for remote pressure indication and low pressure alarms for the oxygen supply to the hydrogen recombinersystem.



Supply from oxygen fill truck to 02 Stand #31.

From outlet- line 16"-53A of Com

ponent Cooling Heat Exchanger #32 to Component Cooling Surge Tank #32. Valve 764B at surge tank is normally closed. FT6OIB on Line #53A indicates in the CCR a combined flow with FT601A on outline line 16"-53 of Component Cooling Heat Exchanger #31.

17A Sheet 20

Fire Zone 17A





1-AC-249

I"-WD-572

Local Cntl Pnl Emerg Boration

From outlet line 16"-53 of Component Cooling Heat Exchanger #31 to Component Cooling Surge Tank #31. Valve 764A at surge tank is normally closed.

Supply from oxygen fill truck to 02 Stand #31.

Charging pump local control panel for controlling CVCS charging pumps

1 1/" City Water Line

3/4" Station Air

01" Aux Steam Inlet & 1" Condensate Ret

1/2" Ltg Conduits

E-18-144-1 145-1

E-18-144-2 145-2

E-18-149-2

thru 154-2

E-18-149-3

Tray

Tray

Tray

Electric Tray DB

Electric Tray DB

Electric Tray FC

Tray Electric Tray FC

Charging Pumps Speed Control Panel (Emerg Boration)

Charging Pumps Local Control Panel (Emerg Boration)

Charging Pumps Speed Control Panel

Charging Pumps Local Control Panel

L-9-171-3

Conduit

Light

L-1 6-1-1

E-3-153-2

Light

1-4 1-4

1-4 1-4

1-3-8-2

Light 1-3-10-3

Tareet Name

Sheet 21

Fire Zone 17A




Safety

Description Goal

Light 1-16-37-1

1-1 6-3 9-1

Nitrogen Bottle 1-16-41-11

1-16-44-11

1-1 6-45-11

PT3

PT 3 Signal Line

PT 1065 Signal Line

PT 5

PT 5 Signal Line


Pressure transmitter for remote pressure indication and low pressure alarms for the oxygen

supply to the hydrogen recombiner system.

Pressure transmitter for the low

pressure alarm for the hydrogen supply to the hydrogen recombiner system and the chemical and volume control system.

Pressure transmitter for remote pressure indication and low pressure alarms for the hydrogen supply to the hydrogen supply manifold.

Pressure transmitter for remote pressure indication and low pressure alarms for the hydrogen supply to the hydrogen supply manifold.

Sheet 22





1/2" Ltg Conduits E-18-160-1 Tray Electric Tray FC

Fire Zone 19A Sheet 23

Fire Zone Set2



Source NameInteraction Number Targzet Name

Safety De scription Goal

8" Screen Wash E- 8-53-2

54-2

55-2 56-4 57-4

Control Panel

Control Panel

Control Panel

Control panel for Service Water Pumps #31 and #32.

Control panel for Service Water Pumps #33 and #34.

Control panel Pumps #35 and and Box ZM4.

for Service Water #36, 1*' Conduit

Drain Header #1

Drain Header #2

Drain Header #3

V-8-167-1 168-1 169-1 170-1 171-1 172-1

V-8-167-2 168-2 169-2 170-2 171-2 172-2

V-8-167-6 168-6 169-6 170-6 171-6 172-6

Service Water Pumps

Service Water Pumps

Service Water Pump s

Lighting Circuit E-8-1-1

E-8-2-1

E-8-6-1

SW Pump Motor Terminal Box



Motor SW Pump box.

Motor SW Pump box

No. 31 and terminal

No. 31 and terminal

Motor SW Pump No. 32

Service Service Service Service Service Service



Water Water Water Water Water Water



Pump Pump Pump Pump Pump Pump



#31 #32 #33 #34 #35 #36

#31 #32 #33 #34 #35 #36

1-4

Sheet 24

Fire Zone 22 Set2



Source NameInteraction

Tar~t NRmPSafety

Dp ~rrf nt 1 r~n

Motor SW Pump No. 32E-8-7-1

E-8-1 1-1

E-8-12-1

E-8-1 6-1

E-8-1 7-1

E-8-21-1

E-8-22-1

E-8-2.6-1.

E-8-27-1

E-8-31- 2

E-8-53-3

E-8-54-3

E-8-55-3

Speaker











Control Panel

Control Panel

Control Panel

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

Source Name Number rTarget Name G-1

Motor SW Puimp No. 33 and terminal box.

Motor SW Pump No. 33 and terminal box.







Relay Panel for Service Water System.

Control panel for Service Water Pumps #31,and #32.

Control panel for Service Water Pumps #33 and 34.

Control panel for Service Water Pumps #35 and #36, 1" Conduit and Box ZM4.

Sheet 25

Fire Zone 22A





3/4-AC-732

AC-824G

AC-824H

'Portion of Line 732 which bypasses Boric Acid Evaporator Package.

Normally open manually operated globe valve on sensing line for FI-631B

Normally open manually actpated globe valve on sensing line for FI-631B.

L-9-217-3

V-9-268-3

V-9-269-3

Tareet Name

Sheet 26

12" - 728

Fire ZoneSheet 27


Source NameInteraction Number Tar .t N%.mP


Conduit 1-1/2" Phone Box

Conduit 1-1/2"

Conduit 1-1/2" I/P

Conduit 1"

0 Conduit 1"

Speaker

1-10-15-6

1-10-15-7

1-10-15-8

1-10-19-4

1-10-19-5

1-23-31-11

1-23-32-11

Piping to FCV405A,-B-C,-D

Piping to FCV405A,-B,-C,-D

Piping to FCV

405A,-B,-C,-D

Piping before

check valve

Piping beforecheck valve

PM 405A

PM 405B

Flow control valves for the auxiliary feedwater system from the steam turbine aux feedwater pump to the steam generators.



Piping downstream of the instrument air supply line check valve to the auxiliary feedwater system control valves.

Piping downstream of the instrument air supply line check valve to the auxiliary feedwater system control valves.

Turbine Driven Auxiliary Feedwater Pump No. 32 discharge flow control to Steam Generator No. 31

Turbine Driven Auxiliary Feedwater Pump No. 32 discharge flow control to Steam Generator No. 32

Number 12 Tar et Name

23




Wall Lite

1" Conduit

1-10-15-9

1-10-2-1

1-10-3-1

1-10-4-1

1-10-5-1

1-10-6-1

1-10-7-1

1-10-8-1

Piping to FCV 405A, B, C, D

PCV-1275

PCV-1274

PCV-1273

PC-1355

Root Valve for PC-1355

PCV-1284

PCV-1276

Flow Control Valves for the 2 Auxiliary Feedwater system from steam turbine and feedwater pump to the steam generators.

Pressure reducing valve (manual) 2 off the nitrogen cylinder before it enters the back-up system.



Low pressure alarm controller 2 for the nitrogen back-up system for the atmospheric steam dump valves and the auxiliary feedwater control valves.

Low pressure alarm controller 2 for the nitrogen back-up system for the atmospheric steam dump valves and the auxiliary feedwater control valves.

Releases high pressure in the 2 nitrogen back-up piping.

Regulates the pressure of the 2 nitrogen back-up system to a set pressure.

Fire Zone Sheet 28

Fire Zone 23



Source NameInteraction Number Taraet .Name

Safety ( ni~lDe ~!rint1on

1-10-9-1

1-10-10-1

1-10-11-1

3/4" Conduit 1-10-2-2

1-10-3-2

1-10-4-2

1-10-5-2

1-10-6-2

Valve Upstream of PCV-1276

Permits flow of nitrogen coming from PCV-1276.

Valve Downstream Permits flow of nitrogen coming of PCV-1276 from PCV-1276.

Bypass Valve PCV-1276

PCV-1275

PCV-1274

PCV-1273

PC-1355

Root Valve for PC-1355

Permits nitrogen flow to the auxiliary feedwater valveswhen PCV-1276 is inoperable.

Pressure reducing valve (manual) off the nitrogen cylinder before it enters the back-up system.

Pressure reducing valve (manual) off the nitrogen cylinder before it enters the back-up system.

Pressure reducing valve (manual). off the nitrogen cylinder before it enters the back-up system.

Low pressure alarm controller for the nitrogen back-up system for the atmospheric steam dump valves and the auxiliary feedwater control valves.

Low pressure alarm controller for the nitrogen back-up system for the atmospheric steam dump valves and the auxiliary feedwater control valves.

rTar2etName Goal

Sheet 29

Fire Zone 23




Safety

Description Goal

PCV-1284

PCV-1276

Valve Upstream of PCV-1276

Releases high pressure in the nitrogen back-up piping.

Regulates the pressure of the nitrogen back-up system to a set pressure.

Permits flow of nitrogen coming from PCV-1276.

Valve Downstream Permits flow of nitrogen coming of PCV-1276 from PCV-1276.

Bypass Valve PCV-1276

Permits nitrogen flow to the auxiliary feedwater valves when PCV-1276 is inoperable.

1-10-7-2

1-10-8-2

1-10-9-2

1-10-10-2

1-10-11-2

Tareet Name

Sheet 30

Fire Zone 24A




Safety

S...GoalDe scription

1 1/2" - #196 L-9-229-2

V-9-285-1

V-9-286-1

3/4-AC-472

AC-824A Signal Line

AC-824B

From inlet of Boric Acid Evaporator Package #31 to Relief Valve 821A. This bypass protects the Boric Acid Evaporator Package from overpressurization.

Normally open manually operated globe valve on sensing line for FI-631A.

Normally open manually actuated globe valve on sensing line for FI-631A.

Sheet 31

FireZone 27A

Set3NEW YORK POWER AUTHORITY

INDIAN POINT 3 SYSTEMS INTERACT ION STUDY SUMMARY OF UNACCEPTABLE INTERACTIONS



2" - 560

3/4" Station Air Header

Shower

* team Htr Piping

Space Heater

L-9-2 52-5

L-9-253-5

1-9-76-2

1-9-78-4

E-3-19 3-1 194-1 195-1

L-3-17 7-6

L-3-17 9-6

L- 3-181-6

1-AC--248

I-AC-24 9

PC-600QA

LT-628

J Box, Conduits

2" CH-202

3/4' CH-871

1" CH-202

From outlet of Component Cooling Heat Exchanger #32 via line 53A to Surge Tank #32.

From outlet of Component Cooling Heat Exchanger #31 via line 53 to Surge Tank #31.

Pressure controller annunciates and starts the stand-by pump on low pressure of the component cooling pump discharge into CCW loop 1.

Level transmitter for local and remote level indication, and high and low alarm for CCW surge tank No. 31.

Box V16, I" conduit 4VY, 1 1/4" conduit 4VZ route cables for MOV-333 on the emergency boration flow path

Boric acid solution supply line to charging pump suction header (Line #200)

Local sample and flushing line for an inlet line to B A blender, a non-safety function CVCS component

Boric acid solution supply line to charging pump suction header (Line #200)

2

1,3

1, 3

1, 3

Fire Zone Sheet 32

Fire Zone 27A




Safety (::n.21IDe .rri nt( n

L-3-182-6

L-3-183-6

L-3-18 6-6

L-3-187-6

L-3-188-6

L-3-189-6

L-3-190-5

V-3-287-5

V-3-288-5

V-3-321-6

1" CH-204

3/4" CH-873

2" CH-204

2" CH-284

2" CH-228

2" CH-227

2" CH-208

3/4" CH-202

2" CH-333

3/4" CH-399B

Normal RCS boration line

Local sample and flushing line for normal boration path (Line #204)

Normal supply of boric acid solution to reactor coolant. This line connects to an emergency boration supply line and also the charging pumps suction header; both a safety function portion of CVCS

Boric acid solution supply to BIT, a non-safety function. This line connects to Line #204, an emergency boration supply line, which is a safety function

Boric Acid Filter #31 bypass line

Downstream of Boric Acid Transfer Pumps

Emergency boration line connected to the charging pumps suction header (Line #200)

A drain valve for line 2" CH-208, an emergency boration supply line

Normally closed, motor operated valve located on an emergency boration line (2" CH-208)

Instrument root valve for PI-109, located downstream of B A Filter #31

1, 3

1, 3

1, 3

1

1

1, 3

Number Target Name Descri tion

Sheet 33

27A



Source Name

2" Space Heater Steam Lines (Amt-2)

Interaction Numbr

V-3-323-6

V-3-326-6

L-3-179-4

3/4" CH-399A

3/4" CH-399C

3/4" CH-871


A normally closed vent valve on 1 Boric Acid Filter #31

Instrument root valve for PI-108, 1 located upstream of B A Filter #31

Local sample and flushing line for an inlet line to B A blender, a non-safety functioning CVCC component.

L-3-181-4

L-3-182-4

L-3-183-4

V-3-287-3

V-3-321-4

V-3-323-4

V-3-326-4

1" Station Air Hdr L-3-200-15

Light 1-3-18-7

I" CH-202 Boric acid solution supply line to charging pump suction header (Line #200)

1" CH-204 Normal RCS boration line

3/4" CH-873 Local sample and flushing line for normal boration pat (Line #204)

3/4" CH-202 A drain valve for line 2" CH-208, an emergency boration supply line

3/4" CH-399B Instrument root valve for PI-109, located downsteam of B A Filter #31

3/4" CH-399A A normally closed vent valve on Boric Acid Filter #31

3/4" CH-399C Instrument root valve for P1-108, located upstream of B A Filter #31

3/4" CH-226 Boric Acid Tank #31 recirculation (Emerg Boration) line

LT 106 (Emerg Boration)

1, 3

1, 3

1, 3

1

Level Transmitter for Boric Acid Storage Tank #31. Level is indicated in-CCR and transmits indication to LC 106

Number Tar et Name

Fire Zone Sheet 34

T r pt .N m

Fire Zone 27A





.City Water Line 1-3-18-4

1-9-76-1

E-9-81-2 82-2 83-2 84-2

E-9-110-2

Light 1-9-76-3

Light 1-9-77-3

Light 1-9-79-3

LT 106 (Emerg Boration)

PC-600A

Flex Cond

Panel PP4

PC-600A

PC-600B

LT-629

Level Transmitter for Boric Acid Storage Tank #31. Level is indicated in CCR and transmits indication to LC 106.

Pressure controller annunciates and starts the standby pump on low pressure of the component cooling pump discharge into CCW loop 1.

Box AAED, 3/4" Conduit 4NH, 1" Conduit 4NHI routing cables connecting LT-628 which annunciates high and low levels of Component Cooling Surge Tank #31 in the CCR.

For Component Cooling Water System connected to TIC-627A which annunciates in theCCR or high temperature of Component Cooling Pump Loop 1 at the pump inlet.



Level transmitter for local and remote level indication, and high and low alarm for CCW surge tank No. 32.

Sheet 35 •

Fire Zone 35A





2" Steam Line

4" Sanitary Stack

E-7-2-5 3-5 6-5

E-7-7-3 7-5

E-7-13-2 14-2 14-4 16-2 17-2

E-7-50-2

51-2 52-2

E-7-53-2 54-2

L-7-4-7

L-7-5-7

L-7-6-7

Conduit

Conduit

Conduit

Conduit

Conduit

12" x 24" Carbon Filter Outlet Duct

12" x 22" Carbon Filter Outlet Duct

11" x 14" AC Unit Inlet Duct

1 1/4" Conduit 5SQ, 1" Conduit 5SL, 1 1/2" Conduit 5SC2 routing cables to Air Conditioning Unit ACU #31.

Control Panel and 1 1/2" Conduit A routing cables to the operating mechanism for Damper D1 on the inlet duct for Air Conditioning Unit ACU #31.

Pull Box, 1 1/4" Conduit 5SR, 1" Conduit 5SM and 1 1/4" Conduit 5SJ2 routing cables to Air Conditioning Unit ACU #32.

1" Conduit 5SN, Pull Box W75, and 1 1/2" Conduit 5ST routing cables to Filter Fan #31 which provides filtered air to the control room during incident conditions.

1" Conduit 5SP and 1 1/2" Conduit 5SS routing cables to Filter Fan #32 which provides filtered air to the control room during incident conditions.

Directs recirculated filtered air to the booster fans during incident conditions.

Directs recirculated filtered air to the booster fans during incident conditions.

During incident conditions, it directs recirculated filtered air to the air condition unit.

1

1-4

1-4

1-4

1-4

1-4

1-4

Sheet 36

FireZone3 5ASheet 37


SUMMARY OF UNACCE PTABLE I NTERA CT IONS

Source NameInteract ion

Target NameSafety

Number~N. ~-r~ nt-I nfl

L-7-7-7

V- 7-8-5

V- 7-9-5

V- 7-10-5

V- 7-11-5

1-7-3-1

E- 7-3 6-1

16" x 16" Inlet Duct

Damper F-i

Damper F-2

Carbon Filter Casing

Filter Fan 31

Flow Switch

1 1/4" CNDC

2" Auxiliary Steam

L-7-7-1

L-7-8-1

16 X 16 AC Inlet Duct

20 X 86 AC Inlet Duct

Air Condition Unit Inlet Duct 16" x 16".


Air condition unit inlet duct 16" x 16".

Damper F-i in 12" x 22" Duct. Damper in inlet duct for booster fan #31. Opened during incident conditions to provide filtered air to control room.

Damper F-2 in 12" x 22" Duct. Damper in inlet duct for booster fan #32 opened during incident conditions to provide filtered air to the control room if booster fan #31 fails to start.

Contains filters which filter outside air and recirculated air during incident condition.

Booster fan downstream of filter casing which provides filtered air to control room during incident conditions.

Air flow switch downstream of fan #31. In the event that booster fan #31 fails to start during the incident mode of operation, the air flow switch will start booster'fan #32 after a predetermined time delay.

1 1/4" Conduit housing cables connected to Damper Fl, in the inlet duct for Booster Fan #31. This is opened during incident conditions to provide filtered air to the control room.

1-4

1-4

1-4

1-4

1-4

2

3 5AFire Zone

Fire Zone 35ASet 8

NEW YORK POWER AUTHORITY INDIAN POINT 3 SYSTEM4S INTERACTION STUDY


Source NameInteract ion Number Tar2et Name


I"Condensate

L-7-1 3-1

L- 7-1 4-1

L- 7-1 5-1

V- 7-4-1

V-7-13-1

L-7-7-3

L-7-8-3

L-7-14-2

L-7-15-2

V-7-4-3

V- 7-5-3

14 x 46 AC Outlet Duct



Damper D-1

ACU 31

16x16 AC Inlet Duct

20x86 AC Inlet Duct

19x24 AC Outlet Duct

24x38 AC Outlet Duct

Damper D-1 in 20" x 86" Duct

Damper D-1 in 20" x 86 Duct

Directs flow from AC Unit (on elevation 15'-O') to the control room.

Air Condition Unit Outlet Duct 19" x 24".


-Damper D-1 in 20" x 86" Duct Damper in inlet duct for Air Condition Unit #31. Instrument air used to open.

Air Condition Unit #31.

Air Condition Unit Inlet Duct 16" x 16"




Damper in inlet duct for Air Condition Unit #31. Instrument air used to open.

Damper in inlet duct for Air Condition Unit #31. Instrument air used to open.

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

1-4

Tareet Name

35A Sheet 38

Fire Zone 35A Set3


SUJMMARY OF UNACCEPTABLE INTERACTIONS



1 1/2" Auxiliary L-7-8-2

V-7-5-2

20x86 AC

Inlet Duct

Damper D-2


Damper D-2 in 20" x 86" Duct. Damper in inlet duct for Air Condition Unit 1/32. Instrument air used to open.

Air Condition Unit #31.

ACU 32 Air Condition Unit #32.

V- 7-1 3-2

V-7-14-2

ACU 31

Sheet 39

Fire Zone 36A





FIC 1176 LP Sensing Line

DPC 1134C HP Sensing Line

FIC 1176 HP Sensing Line

DPC 1134C LP Sensing Line

SOV 1276A

Airlines from SOV 1276A to FCV 1176A

SOV 1276

Airlines from SOV 1276 to FCV 1176

SOV 1275

Diesel generators jacket water and 1 lube oil coolers service water flow controller.

Diesel generators jacket water and 1 lube oil coolers service water low flow switch, to CCR annunciator.

Diesel generators jacket water and 1 lube oil coolers service water flow controller.

Diesel generators jacket water and 1

lube oil coolers service water low flow switch.

Override solenoid for the pneumatically operated diesel generators coolers service water flow control valve.




Diesel generators coolers service

water flow control valve, pneumatic input shutoff solenoid.

Conduit 1-8-90-4

1-8-92-4

1-8-93-4

1-8-94-4

Conduit 0. 1-8-95-1

1-8-96-1

1-8-97-1

1-8-98-1

1-8-99-1

Sheet 40

Fire Zone 36A



Source NameInteraction Number Tarset Name


1-8-100-1

1-8-101-1

1-8-102-1

1-8-103-1

1-8-104-1

Wall Heater 1-8-95-3

1-8-96-3

1-8-97-3

Airline from FIC 1176 to SOC 1275

Airline from SOV 1275 to Positioners

SOV 1274

Airline from FIC 1176 to SOV 1274

Airline from SOV 1274 to Pbsitioners

SOV 1276A

Airlines from SOV 1276A to FCV 1176A

SOV 1276

Diesel generators jacket water and lube oil coolers service water flow controller pneumatic signal line.


Diesel generators coolers service water flow control valve, pneumatic input shutoff solenoid.





Override solenoid for the

pneumatically operated diesel generators coolers service water flow control valve.

Tar2et Name

Sheet 41

Sheet 42




Safety

Description Goal

Airlines from SOV 1276 to FCV 1176

SOV 1275

Airline from FIC 1176 to SOC 1275

Airline from SOV 1275 to Positioners

SOV 1274

Airline from FIC 1176 to SOV 1274

Airline from SOV 1274 to Pbsitioners

1-8"98-3

1-8-99-3

1-8-100-3








1-8-101-3

1-8-102-3

1-8-103-3

1-8-104-3

Tareet Name

36AFire Zone

Fire Zone 41A,43A,44A,45A,46A,48A



Source NameInteraction Number Target Nnm

Safetyflerr4rinn

Turbine Generator E-8-76-1 Building

E-8-77-1

E-8-78-1

E-8-79-1

E-8-80-1

E-8-81-1

E-8-82-1

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Flexible conduit, Box KS5, 3/4" conduit 3LS routing cable for PT-1190, which monitors service water pump discharge header pressure.



Flexible conduit, Box KS6, 3/4" conduit 3PC routing cables connected to PT-1191 for monitoring service water discharge header pressure.



Switch and I" flexible conduit routing cables connected to PC-1111A for service water discharge header pressure. Lowannunciator in CCR.

r Target Name Descri tion

Sheet 43




Source NameInteraction Number Taraet Name


E-8-83-1

E- 8-84-1

E-8-85-1

E-8-86-1

E-8-87-1

E-8-88-1

E-8-89-1

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Switch and 1" flexible conduit routing cables connected to PC-I11A for service water discharge header pressure. Lowannunciator in CCR.

Switch and I" Flexible conduit for PC-llllB for service water discharge header pressure. Lowannunciator in CCR.

Switch and 1" Flexible conduit for PC-1111B for service water discharge header pressure. Lowannunciator in CCR.

Switch and I" Flexible conduit connected to PC-1112A for service water discharge header pressure. Low-annunciator in CCR.

Switch and 1" Flexible conduit connected to PC-1112A for service water discharge header pressure. Low-annunciator in CCR.

Switch and 1" Flex conduit, Box Z36, 1" CND 2LD routing cable connected to PC-1112B for service water pump discharge water pressure. Hi-Annunciator in CCR.


Tareet Name

Sheet 44





Safety

Description Goal

E-8-90-1

E-8-91-1

E-8-92-i

E-8-93-1

E-8-94-1

E-8-95-1

E-8-96-1

E-8-97-1

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box



3/4" conduits 3LU, 3LUl, 3LU, 3LVI, 3LY, 3LW Boxes WH3 and XB1 routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LUl, 3LU, 3LVI, 3LY, 3LW Boxes WH3 and XBl routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LUI, 3LU, 3LVI, 3LY, 3LW Boxes WH3 and XBl routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LUl, 3LU, 3LV1, 3LY, 3LW Boxes WH3 and XB1 routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LUl, 3LU, 3LVI, 3LY, 3LW Boxes WH3 and XBI routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LU1, 3LU, 3LVI, 3LY, 3LW Boxes WH3 and XBl routing cables to PT-1190 and PT-1191.

Sheet 45




Source NameInteraction Number Tariet Name


E-8-98

E-8-99-1

E-8-100-1

E-8-101-1

E-8-102-1

E-8-l03-1

E-8-104-1

E-8-105-1

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

3/4" conduits 3LU, 3LU1, 3LU, 3LV1, 3LY, 3LW Boxes WH3 and XB1 routing cables to PT-1190 and PT-1191.

3/4" conduits 3LU, 3LUl, 3LU, 3LVl, 3LY, 3LW Boxes WH3 and XBl routing cables to PT-1190 and PT-1191.

3/4" conduits 3LT, 3LTl, 3LU, 3LUl and Box WH2 routing cables for PT-1190 and PT-1191.

3/4" conduits 3LT, 3LTI, 3LU, 3LU1 and Box WH2 routing cables for PT-1190 and PT-1191.




3/4" conduits 3LR, 3LRl, 3LT, 3LTl and Box W73 routing cables for PT-1190- and PT-1191.



E-8-106-1

E-8-107-1

Sheet 46

Fire Zone 41A.43A,44A,45AP46AP48A



Interactionfl 1 1VPD T rc~P N~m


E-8-108-1

E-8-109-1

E-8-110-1

E-8-111-1

E-8-112-1

E-8-113-1

E-8-114-1

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

Cond. J. Box

3/4" conduits 3LR, 3LRl, 3LT, 3LTl and Box W73 routing cables

for PT-1190- and PT-1191.

3/4" conduits 3LR, 3LRl, 3LT, 3LT1 and Box W73 routing cables for PT-1190- and PT-1191.

3/4" conduits 3LS, 3PC, 3LR, 3LRl

and Box WX3 routing cables for PT-1190 and PT-1190.

3/4" conduits 3LS, 3PC, 3LR, 3LR1


3/4" conduits 3LS, 3PC, 3LR, 3LRI


3/4" conduits 3LS, 3PC, 3LR, 3LRl


3/4" conduits 3LS, 3PC, 3LR, 3LR1 and Box WX3 routing cables for PT-1190 and PT-1190.

Snwrf-0 woma Number Tar et Name

Sheet 47

.Fire Zone 41A




SafetyDP-rrnt4Al,

Turbine Bldg NonSeismic Area

1-8-107-1

1-8-108-1

1-8-109-1

1-8-110-1

1-8-111-1

PT-II91

PT-Il91 Sensing Line

PC-IlllAS

PC-illlAS Sensing Line

PC-1111BS

Service water pumps discharge header pressure (nuclear) to PI-1191R in CCR.

Service water pumps discharge header pressure (nuclear) to PI-1191R in CCR.

Service water pumps discharge header pressure (nuclear) to CCR annunciator (Lo)

Service water pumps discharge header pressure (nuclear) to CCR annunciator (Lo)

Service water pumps discharge header pressure (nuclear) to CCR annunciator (Hi)

PC-IIIBS Sensing Line

PT-I190

PT-I190 Sensing Line

PC-1112AS

PC-1112AS Sensing Line

Service water pumps discharge 2

header pressure (nuclear) to CCR annunciator (Hi)

Service water pumps discharge header pressure (conventional) to PI-ll9OR in CCR.

Service water pumps discharge header pressure (conventional) to PI-I19OR in CCR.

Service water pumps discharge header pressure (conventional) to CCR annunciator.

Service water pumps discharge header pressure (conventional) to CCR annunciator.

1-8-112-1

1-8-113-1

1-8-114-1

1-8-115-1

1-8-116-i

L Source.Name Number G-1

Sheet 48

Fire Zone 41A




Safety

TreNae Descr nti Ao n

1-8-117-1

1-8-1 18-1

E-17-100-1

E-17-10 1-1

E-17-102-1

E-17-103-1

E-17-104-I

E-17-105-1

E-17-9 6-1

E-17-9 7-1

E -17-9 8-1

E-17-9 9-1

PC-1112BS

PC-1112BS Sensing Line

CND 30T

Pull Box M-28

CND 30S

Tray B8

CND 3DF

Tray 76B

CND-3DG

Pull Box M72

CND-3 QU

Pull Box V64

Service water pumps discharge header pressure (conventional) to CCR annunciator (Hi).

Service water pumps discharge header pressure (conventional) to CCR annunciator (Hi).

Routes cables for Reactor Protection System.










Sheet 49

De rr t n

Fire Zone 5ASet5



Source Name

18" - #5

Interaction Number

1-14-61-1

1-14-6 2-1

1-14-63-1

1-14-64-1

1-14-65-1

1-14-66-1

1-14-67-1

Target Name

PT 419B Sensing Line




PT 419C Sensing Line



Tareet Name


Steam Generator 31 main steam1 transmitter, Channel 2 input to the reactor protection system and for steam flow pressure compensation

Steam Generator 32 main steam1 pressure transmitter, Channel 2 input to reactor protection system

Steam Generator 33 main steam1 pressure transmitter, Channel 2 input to the reactor protection system

Steam Generator 34 main steam1 transmitter, Channel 2 Input to the reactor protection system

Steam Generator 31 main steam1 pressure transmitter, interlock input to the high steam flow protection logic and steam line differential-pressure protection logic

Steam Generator 32 main steam1 pressure transmitter, interlock input to the high steam flow protection logic and steam line differential pressure protection logic

Steam Generator 33 main steam1 pressure transmitter, interlock input to the high steam flow protection logic and steam line differential protection logic

52A Sheet 50

Fire Zone 52A Set5



Source NameInteraction

Tar2et NameSafety

Description Goal

1-14-68-1

1-14-65-2

1-14-66-2

1-14-67-2

1-14-68-2

1-14-57-3






PT 419A Sensing Line

Steam Generator 34 main steam pressure transmitter, interlock input to the high steam flow protection logic and steam line differential pressure protection logic


Steam Generator 32 main steam pressure transmitter, intelock input to the high steam flow protection logic and steam line differential pressure protection logic



Steam Generator 31 main steam pressure transmitter, Channel 1 input to the reactor protection system for steam flow pressure compensation

18" - #6

18" - #7

Number Tareet Name

Sheet 51

Fire Zone 52A Set5



Source NameInteraction Number

1-14-58-3

1-14-59-3

1-14-60-3

@s8" - #8 1-14-5 7-4

Target Name




Fr 419A Sensing Line


Steam Generator 32 main steamI pressure transmitter, Channel 1 input to the reactor protection system and for steam flow pressure compensation

Steam Generator 33 main steam1 pressure transmitter, Channel 1 input to the reactor protection system and for steam flow pressure compensation

Steam Generator 34 main steamI pressure transmitter, Channel 1 input to the reactor protection system and for steam flow pressure compensation

Steam Generator 31 main steam1 pressure transmitter, Channel 1 input to the reactor protection system and for steam flow pressure compensation




Steam Generator 32 main steam pressure transmitter, Channel 1 input to the reactor protection system and for steam flow pressure compensation



1-14-58-4

1-14-59-4

1-14-60-4

Tareet Name

Sheet 52

Fire Zone





1-14-61-4

1-14-62-4

1-14-63-4

1-14-64-4

Light 1-10-20-2

Storage Cabinet 1-1 0-20-3





Piping

Piping


Steam Generator 32 main steam pressure transmitter, Charwnel 2 input to the reactor protection system

Steam Generator 33 main steam pressure transmitter, Channel 2 input to the reactor protection system

Steam Generator 34 main steam pressure transmitter, Channel 2 input to the reactor protection system

Nitrogen back-up supply piping to the auxiliary feedwater control valves.

Nitrogen back-up supply piping to the auxiliary feedwater control valves.

Tar2et Name

52A Sheet 53

Fire Zone 57A





Light 1-10-21-1

1-10-24-6

Manometer

Roof Drain

6" Roof Drain

1-10-24-4

E-14-14-1 15-1 16-1 16-2 17-1 17-2 18-1 19-1

1-14-81-1

1-14-82-1

Piping to Steam Dump Panel (PCV-1134, PCV-1135)

Steam Dump Panel PCV-1136 and PCV-1137

Steam Dump Panel PCV-1136 and PCV-1137

P Box, Cond

MSIV 33 Control Station

MSIV 31 Control Station

Provides nitrogen supply for manual operation of the atmospheric relief valves (PCV-1134, PCV-1135)

Used for manual operation of atmospheric relief valves PCV-1136, PCV-1137.

Used for manual operation of atmospheric relief valves PCV-1136, PCV-1137.

PB X82, ZT6, 1" Conduits 7TD & 7TF housing cables and controls for Main Steam Stop Valve #33

Steam Generator isolation valve station

Steam Generator isolation valve station

33 main steam local control

31 main steam local control

3/4" normally closed venting valve for 28" MS Line #2. Vent line is on MSIV cap. Rupture of the 3/4" vent line is also considered.

V-14-4-1 5-1

MS-56-31 MS-57-31

Sheet 54

Fire Zone 57A




Safety GoalDescription

1-1/2" Drain Line 1-10-24-3 Steam Dump Panel PCV-1136 and PCV-1137

Used for manual operation of atmospheric relief valves PCV1136, PCV-1137.

Sheet 55

Fire Zone 58A


SUMMARY OF UN~ACCEPTABLE INTERACTIONS

Source NameInteraction Number Tar2et Name


10". Steam Line L-9-262-1

L-9-263-1

L-9-277-1

I-AC-657

I-AC-6 58

8-AC-168

Seal Cooling for RR Pump #32 cooling block.

Re turn from seal cooling of RH-R Pump #32 cooling block to line #52.

Boric Acid Evaporator Package #31 to line 16" - #52.

Sheet 56

Tareet Name

Fire Zone 59A



Source Name

Interaction Number

Safety Target Name Description Goal

Y - #45 L-9-297-1

L-9-298-1

L-9-304-1

L-9-305-1

L-9-306-1

L-9-297-2

L-9-298-2

L-9-304-2

L-9-305-2

2-AC-55

2-AC-55A

2-AC-650

2-AC-840

2-AC-841

2-AC-55

2-AC-55A

2-AC-650

2-AC-840

Cooling Return line from Recirc Pump #31 (SIS)


Supply line from #22 to suction of Auxiliary Component Cooling Pumps #33 & #34. Pumps discharge to line 54A supplies cooling water to Recirc Pump #32 (SIS).

Provides cooling water to Gross Failed Fuel Detector sample cooler from line #53A. F1657, downstream of sample cooler on line #841, indicates locally.

From Gross Failed Fuel Detector sample cooler to line #52A. F1657 provides local indication.



Supply line from #22 to suction of Auxiliary Component Cooling Pumps #33 & #34. Pumps discharge to line 54A supplies cooling water to Recirc Pump #32 (SIS).


3" - #46

Source Name

Sheet 57

Fire Zone 59A Set5




Safety

GoalDescription

2-AC-841

2-AC-129

2-AC-133

2-AC-840

2-AC-841

PT 3 Signal Line

2-AC -129

2-AC-133

2-AC-840

L-9-306-2

L-9-2 99-3

L-9-300-3

L-9-305-3

L-9-306-3

3" - #47


Cooling Line from line #22 to Steam Generator Sample Heat Exchanger Loop 34.

Cooling water Return line from Steam Generator Sample Heat Exchanger loops to line #18.




Cooling Line from line #22 to Steam Generator Sample Heat Exchanger Loop 34.

Cooling water Return line from Steam Generator Sample Heat Exchanger loops to line #18.

Provides cooling water to Gross Failed Fuel Detector sample cooler from line #53A. F1657, downs tream of sample cooler on line #841, indicates locally.

1-16-52-4

3" - #48 L-9-299-4

L-9-300-4

L-9-30 5-4

Sheet 58

Fire Zone 59A





2-AC-841

PT 3 Signal Line

1"-SI-68



Safety injection system line through containment penetration RR. Contains pressurizing N2 gas for accumulators.

Conduits

3/4" Conduit

1" Conduit

6" - #864

1-8-30-2

1-8-31-2

1-16-52-1

1-16-52-2

L-13-53-4

V-13-83-4

FT-1125 LP Sensing Line

FT-1125 HP Sensing Line

PT 3 Signal Line

PT 3 Signal Line

3/4" - 866

Cap on Line 3/4" - 866

Containment recirc fan #35 service water outlet flow.




Line through containment penetration RR. Used for containment leak rate test, then capped or flanged closed.

Part of containment isolation boundary.

L-9-306-4

1-16-52-5

2" - #34 L-13-29-1

Sheet 59

Fire Zone 59A





Steam Generator Blowdown Lines

V-13-84-4

E-3-232-3 233-3 234-3

Cap on Line 3/4" - 867

P Box, Conduit

Part of containment isolation 4 boundary.

3/4" flexible conduit, Box FT-128, I 1" conduit 7DD/JB housing cables 1 connected to FT-128 which monitors 1 flow in the charging line for emergency boration.

Speaker

Emergency Light

Steam Generator Blowdown Xmtrs

* drogen Tanks

3" #830

E-18-178-1 180-1

E-18-184-2

1-3-62-3

1-3-63-3

L-16-21-2

L-13-13-2

L-13-17-2

V-13-24-1

1-8-36-2 36-3

1-8-37-2 37-3

Tray

Tray

Electric Tray JA

Electric Tray DD

FE 128 HP Sensing Line

FE 128 LP

1" WD-758

2" SA-34

2" IA-39

PCV-1228

FT 1124

FT 1123

Sensing line for FT 128

FE 128 LP sensing line

Flange connection for hydrogen fill truck to supply header.

Service air line through containment penetration Y.

Instrument air line through containment penetration Y.

A normally open, fail closed, air/hand operated gate containment isolation valve located on line 2"-IA-39 outside.



1-4 1-4

1-4

Sheet 60

Fire Zone 59A





1-8-38-2 38-3

1-8-39-2 39-3

1-8-40-2 40-3

1-17-47-3

1-17-48-3

1-17-49-3

1-17-50-3

1-17-51-3

1-17-52-3

L-13-13-3

FT 1121

FT 1122

FT 1125

PT 948A

PT 948B

PT 948C

PT 949A

PT 949B

PT 949C

2" SA-34




Containment pressure transmitter, Hi and Hi-Hi input to SI logic, and containment isolation and spray actuation logics.






Service air line through containment penetration Y.

3" #830A

Sheet 61

Fire Zone 59A





L-13-17-3

V-13-24-2

1-8-36-4

1-8-37-4

1-8-38-4

1-8-39-4

1-8-40-4

1-17-47-4

2" IA-39

PCV-1228

FT 1124

FT 1123

FT 1121

FT 1122

FT 1125

PT 948A

1-17-48-4

1-17-49-4

Instrument air line through containment penetration Y.

A normally open, fail closed, air/hand operated gate containment isolation valve located on line 2"-IA-39 outside.









PT 948B

PT 948C

Sheet 62

Fire Zone 59A





1-17-50-4

1-17-51-4

1-17-52-4

Light

01/2" - #95

3" - #830

1-9-90-1

1-9-95-1

1-17-53-3

1-17-54-3

PT 949A

PT 949B

PT 949C

FIC 633A

FIC 625






Local flow indication and remote low-flow alarm for cooling water to the SIS recirc pump No. 31 in CCW loop 1.

Flow indicator controller for the RCP thermal barrier cooling water return. a) Local flow indication b) Generation of signal to close

containment isolation valves (FCV-789 & -625) on hi-flow.

c) Activation of remote low-flow alarm.



Sheet 63

Fire Zone 59A





. - #83oA

1-17-55-3

1-17-56-3

1-17-57-3

1-17-58-3

1-17-53-4

1-17-54-4

1-17-55-4

1-17-56-4

















Sheet 64

Fire Zone 59A



Source NameInteraction Number TarRet Name






Sheet 65

1-17-57-4

1-17-58-4

Target Name

Sheet 66Fire Zone 70A

NEW.YORK POWER AUTHORITY INDIAN POINT 3 SYSTEMS INTERACTION STUDY


Interaction Number Tar2et Name

Safety

Description Goal

Lube Oil Piping

Lights (4)

L-9-5-3

L-9-8-3

1-5-41-5

1-5-49-1

1-5-50-1

1-5-51-1

1-9-11-1

1-9-12-1

1-5-27-1

1-5-29-1

1-5-33-1

1-5-34-1

1-5-35-1

L-9-13-5

1-AC-13C

1-AC-14

FIC 431 Sensing Line

FT 444 Sensing Line

FT 445 Sensing Line

FT 446 Sensing Line


FIC 619 LP Sensing Line

TE 430 A

TE 431 A

FT 434 Sensing Line

FT 435 Sensing Line

FT 436

Sensing Line

3/4"-AC-379

I-AC-13C to R C Pump #34 motor 2 cooler.

Return line from RCP #33 motor 2 cooler.

Reactor Coolant Loop #3 RTD mani- 1

fold, common return line flow.

Transmitter alarms in CCR at less than 90% normal flow.

Reactor Coolant System Loop #4 3

Flow Transmitter


Flow Transmitter


Flow Transmitter

Flow Indicator on return line 2

for bearing cooler coding water

from RCP #33.

Flow Indicator on return line 2

for bearing cooler cooling water

from RCP #33.

Reactor Coolant System Narrow 1

Range RTD (Hot Leg Loop #3)

Reactor Coolant System Narrow Range RTD (Hot Leg Loop #3)

Reactor Coolant Loop #3 Flow

Transmitter

Reactor Coolant Loop #3 Flow Transmitter

Reactor Coolant Loop #3 Flow Transmitter

Line from #21A thru Relief Valve #783C to floor trench

'"-379 0

Q wrr- WAMP Number Tar2et Name

Fire Zone 71A





Lights (4) 1-5-20-2

1-5-21-2

1-5-22-2

1-17-31-1

1-17-34-1

Speaker

2"L - WD338-01

2"L - WD338-02

1" Drain Line

E-4-124-3 125-3

E-4-124-4 125-4

E-4-125-6

FT 424 Sensing Line

FT 425 Sensing Line

FT 426 Sensing Line

LT 417C LT 417D Line SG

LT 427C LT 427D Line SG

and Sensing 1

and Sensing 2

Cond P Box

Cond P Box

Cond P Box

Reactor Coolant Loop #2 Flow Transmitter. Flow is indicated and alarmed on the SA panel and is used for low flow reactor trip signals.

Reactor Coolant Loop #2 Flow Transmitter.

Reactor Coolant Loop #2 Flow Transmitter.

Steam generator 31 level transmitter to reactor trip logic, steam generator level logic, and level recorder.

Steam generator 32 level transmitter to reactor trip logic, steam generator level logic, and level recorder.

1" Conduit 9CN, Pull Box XQ7. These route cable for MOV-731. For RHR pump normal loop suction.



I-AC-13A to RCP #31 motor cooler 2

1,2,4 1,2,4

1,2,4

1,2,4

1,2,4

I-AC-13A

Tarizet Name

Sheet 67

Lube Oil Piping L-9-16-6

Fire Zone 71A





Lube Oil Drain

L-9-24-6

L-9-27-6

1-5-25-1

L-9-22-6

L-9-31-1

L-9-32-1

1-AC-14B

1 1/2-AC-21B


1-AC-14A

2"AC-54A

2"AC-55A

I-AC-14B Return from RCP #32 motor cooler.

1 1/2-AC-21B Return from RCP #32 lower bearings.

Loop #2 RTD manifold, common return line flow. Transmitter alarms in CCR at less than 90% normal flow.

Return line from RCP #31 motor cooler

2-AC-54A to Recirc Pump #32 cooler (SIS)

2-AC-54A Cooling return line from Recirc Pump #32 (SIS)

Sheet 68

Fire Zone 72A



Snurce NameInteraction Number Target Name


3/8"L - #24 1" L - #30

2" Stn Air Header

Line #338 - 2"

1" Conduit

g Light

E-4-43-2 E-4-44-2

E-4-114-1 E-4-115-1

1-5-83-1

1-9-26-2

1-9-27-2

1-4-14-2

1-4-15-2

1-4-16-2

Light

1 1/2" Line 338

1-5-83-2

E-5-96-3

Box Conduit

Cond

PT 402, PT 413 Sensing Line

FIC-613

FIC-616

FE 946C HP Sensing Line

FE 946C LP Sensing Line

FT 946C Sensing Line

PT 402, PT 413 Sensing Lines

3/4" Conduit 8LK1

Box LZ8, 1" Conduit 8SK houses cables for FT-946C, which measures the flow to RCS Loop 2.

1" Conduit 8RKI and 1-1/4" Conduit 8RK2. These route power and control cables for MOV-730 for RHR normal loop suction.

Reactor Coolant System Loop #1 Hot leg Pressure Transmitters

Flow Indicator on return line for bearing cooler cooling water from RCP #31.

Flow Indicator on return line for bearing cooler cooling water from RCP #31.

Safety Injection Low Head Subsystem Flow to RCS Cold Leg Loop 2.



Reactor Coolant System Loop #1 Hot Leg Pressure Transmitters

Housing cables for FIC-421, flow transmitter, measures the flow in the common return line from hot and cold leg RTD manifolds.

1,2,4 1,2,4

Source Name Number Tareet Name

Sheet 69

Fire Zone 73A Set7

NEW YORK POWER AUTHORITY INDIAN POINT.3 SYSTEMS INTERACTION STUDY




Lights (3) E-18-36-1 37-1 39-1 40-1 43-1 53-1 54-1

Tray Cable Trays at Penetration1-

1-4

1-4 1-4 1-4 1-4 1-4 1-4

Sheet 70

Fire Zone 75A


SUMMARY OF UNACCEPTABLE INTERACTI ONS



E-18-341-2

E-18-353-3

E-18-355-1

Tray JA

Cross Tray JA/CA

Tray JD

Routes Electric Cables



Sheet 71

Light

Light

Fire Zone 76A Set7





Overhead Light 1-4-22-2

1-4-23-2

1-4-24-2

Light 1- 5-8 9-1

1-5-90-1

* tation Air Header 1-2-33-2

1-2-34-2

Relief Valve 318 1-2-33-3

1-2-34-3

FE 946B HP Sensing Line

FE 946B LP Sensing Line

FT 946B Sensing Line

PT 403) PT 443 Sensing Lines

PT 433 Sensing Line

FT 980 HP Sensing Line

FT 980 LP Sensing Line

FT 980 HP Sensing Line

FT 980 LP Sensing Line




Reactor Coolant System Loop #3 Pressure Transmitters.

Reactor Coolant System Loop #3 Pressure Transmitter.

Hi Press Sensing Line - Safety injection flow to Loop 3 cold leg

Lo, Press Sensing Line - Safety injection flow to Loop 3 cold leg

Hi Press Sensing Line - Safety injection flow to Loop 3 cold leg

Lo Press Sensing Line -.Safety injection flow to Loop 3 cold leg

Sheet 72

Fire Zone



Source Name

Light

Light

Light

RT Instrument a

ion Piping

Interaction Number

1-2-45-2

1-2-48-2

- 1-2-47-1

1-2-53-3

1-2-54-3

1-2-56-4

Target Name

LT 934A Sensing Line

LT 935A

Sensing Line

LT 935A

LT 935B

LT 935B Sensing Line

LT 934C


Safety Injection Accumulator Tank 1 31, Level, to LI 934A on Safeguards Panel SBF-1 in CCR

Safety Injection Tank 31 Level 1

Safety Injection Accumulator Tank 1 31 Level, to LI 935A on SI Supervisory Panel SMF

Safety Injection Accumulator Tank 1 32 Level, to LI 935B on SI Supervisory Panel SMF in CCR

Safety Injection Accumulator Tank 1 32 Level

Safety Injection Accumulator Tank 1 33 Level, to LI 934C on Safeguards Panel SBF-l in CCR

LT 934C Sensing Line

LT 935C



LT 935D

Safety Injection Accumulator Tank 33 Level

Safety Injection Accumulator Tank 33 Level, to LI 935C on SI Supervisory Panel SMF in CCR



Safety Injection Accumulator Tank 34 Level, to LI 935D on SI Supervisory Panel SMF in CCR

1-2-57-4

1-2-59-4

1-2-60-4

1-2-57-2

1-2-65-6

Light

Line #85

7 7A Sheet 73

Fire Zone 78A





Line 379 - I" 1-4-8-1

E-4-13-2

FT-640 LP Sensing Line

T. Box P37

Safety Injection Recirculation Sub- 2 System Header Flow to Loops 3 and 4 Cold Legs, for indication on the Auxiliary Cooling System Panel.

T. Box P37 for Recirculation Pump 1,4 Motor 321, which assures lo(ng term core cooling of reactor.

I" Line 507

. re Zone 79A

SG Blowdown Tank Room

E-4-31-1 -32-1

E-4-16-3

L-9-309-1

L-9-310-1

L-9-31 1-1

L-9-312-1

L-9-313-1

Box LHI

MOV 1802B Flex Cond

2-AC-129

1-AC-129

2-AC-133

1-AC-133

1-AC-6 31

3/4" conduit 8DV houses cable for FT-640, monitors flow from RHR Heat Exchanger #32.

Flexible conduit housing for MOV 1802B which is one of the valves on inlet side of Residual Heat Exchangers #31 and #32.

Steam Generator Sample Heat Exchangers Cooling Line.

Steam Generator Sample Heat Exchangers Cooling Line.

Cooling Water Return Line from Steam Generator Sample Heat Exchanger loops.

Cooling Water Return line from Steam Generator Sample Heat Exchanger loops.

Cooling water from line #129 to Steam Generator Sample Heat Exchanger loop #31.

Sheet 74

2

1,2,4

Tar2et Name

Fire Zone 79A


Source NameInteraction Number


Cooling water from line #129 to 2 Steam Generator Sample Heat Exchanger loop #32.

Cooling water from line #129 to 2 Steam Generator Sample Heat Exchanger loop #33.

Cooling Water Return from Steam 2 Generator Sample Heat Exchanger loop #33 to line #133.



From line #129 thru relief valve 2 785A to line #133. Bypasses Steam Generator Sample Heat Exchanger Loops for overpressurization protection.

0~cP q~g

3/4"-AC-637

S.G. Sample HX 31

S.G. Sample HX 32

S.G. Sample HX 33

S.G. Sample HX 34

Overpressurization bypass for Steam Generator Sample Heat Exchanger Loops.

S.G. Sample H-X #31

S.G. Sample H-X #32

S.G. Sample H-X #33

S.G. Sample H-X #34

Sheet 75

L-9-314-1

L-9-315-1

L-9-316-1

L-9-317-1

L-9-318-1

L-9-319-1

1-AC-632

I-AC-633

I-AC-634

1-AC-635

1-AC-636

1-AC-637

L-9-320-1

V-9-391-1

V-9-392-1

V-9-393-1

V-9-394-1

T rpp N m

Fire Zone 82A





1 1/2" Condensate Return

2" Steam Line

2" Breathing:Air

V-21-3-2

V-21-3-3

1-14-40-1

1-14-41-1

Fan Cooler #34 For Containment Cooling and Filteration.

Fan Cooler #34 For Containment Cooling and Filteration.

FT 439B HP Line

FT 439B LP Line

Main steam flow from Steam Generator 33


Sheet 76


INDIAN POINT 3 SYSTEMS INTERACTION STUDY


InteractionT= ,-oc~t T%1mD


1 2" Breathing Air

Space Heater, Steam & Condensate Piping

1-14-14-1

1-14-15-1

L-16-32-2

L-16-33-2

L-1 6-34-1

L-16-35-1

V-16-107-1

V-16-108-2

FT 439B HP Line

FT 439B LP Line

2" WD-573

3/4" WD-574

2" WD-575

3/4" WD-576

Recombiner #31

Recombiner #32



From H2 Stand #32, outside containment, to Recombiner #32

inside containment. Main flame

H2 supply line.

From H2 Stand #32, outside

containment, to Recombiner #32 inside containment. Pilot flame H2 supply line.


containment, to Recombiner #31

inside containment. Main flame supply line.


containment, to Recombiner #31

inside containment. Pilot flame H2 supply line.

Skid containing combustion chamber,

diluent manifold for reducing exhaust temperature, combustion air blower, and associated ductwork.

Skid containing combustion chamber,

diluent manifold for reducing exhaust temperature, combustion

air blower, and associated ductwork.


T- - N-no X7_



Source NameInteraction Number Tareet Narne

SafetyD rr t I L

Safety Injection Accumulator Tank 34 pressure, to PI 936D on Safeguards Panel SBF-l in CCR


Light 1-2-73-1 PT 936D

Fire Zone 88A





Emergency Light E-9-12-1

L-16-39-4

L-16-41-4

Emergency Light 1-16-35-2

Motor

3/4"-WD-574

3/4" WD-576

Recombiner No. 31 CP

Motor for Aux Coolant Pump #34, Box MX2 and 1" Conduit 6CU.

From H2 stand #32 outside containment, to Recombiner #32 inside pilot flame H2 supply.

From H2 stand #31 outside containment, to Recombiner #31 inside pilot flame H2 supply.

Combustor 31, which burns the hydrogen from the containment atmosphere following a LOCA.

FT-2B Flow transmitter for hydrogen

flow to combustor 32.



a) Activates annunciator window for hydrogen flow with comrbustor off.

b) Signals FRC-lB to maintain proper oxygen flow in relation to hydrogen flow.



Light 1-16-17-2

1-16-35-4

1-16-36-3

Sheet 79

Fire Zone




Safety GoalDescription

Light Fixture L-I1-32-1

L-I1-33-1

L-1 1-48-1

L-1 1-5 0-1

3/4-JW 1040

1-JW 1040

2-DF-1078

3/4 DF-1105

Jacket water discharge conn from Diesel Engine #32 to J W Expansion Tank #32. Returns J W to the tank (closed loop)

Jacket water inlet line from J W Expansion Tank to Diesel Generator #32. For jacket water inlit to D G

Vent to atmosphere for Fuel Oil Day Tank #32

Fuel return line from D G to Fuel Oil Day Tank #32.

venting D G system during for system priming

#32 For

starting,

3/4 DF-1107

Diesel Engine 32 &.Exhaust

SWN-77 (1/2" vent)

SWN-77 (1/2" vent)

SWN-63 (3/4" SRV)

Fuel oil supply line from Day Tank 1-4 #32 to D G #32. To supply fuel oil to D G

Diesel Engine #32 and Exhaust 1-4 System

1/2" valve on vent connection to 1-4 6" inlet cooling water line to diesel generator venting valve. Rupture of vent line also considered.

1/2" valve on vent connection to 1-4 6" inlet cooling water line to diesel generator venting valve. Rupture of vent line also considered.


1-4

L-II1-51-1

V-i1-45-3

V-8-207-1

V-8-209-1

V-8-211-1

1 01A Sheet 80

Fire one lASheet 81



Source NameInteract ion Number Target Name


Space Heater

Emnergency Light

Loud Speaker

Lights

V-11-45-1

V-11-45-2

V-11-45-4

1-11-47-1

Diesel Engine 32 & Exhaust



Generator Panel #32

Diesel Engine #32 and Exhaust System

Diesel Engine #32 and Exhaust

Diesel Engine #32 and Exhaust

Starting and controlling of the Diesel Generator #32

101AFire Zone

Fire Zone 102A Set8




Safety Descrivtion Goal

Light Fixture L-11-5-5

L-11-6-5

L-1 1-17-1

L-1 1-19-1

L-1 1-20-1

V-11-5-2

V-8-198-1

V-8-200-1

V- 8-202-1

3/4*' JW 1039

lI" 3W 1036

2" D F-1079

3/4"* D F 1108

3/4* D F 1110


SWN-77 (1/2" vent)

SWN-7 7 (1/2" vent)

SWN-6 3 (3/4* vent)

Jacket water discharge conn from 1-4 Diesel Engine #33 to J W Expansion Tank #32. Returns J W to the tank (closed loop)

Jacket water inlet line from J W 1-4 expansion tank to Diesel Generator #33. For jacket water inlet to D G

Vent to atmosphere for F 0 Day 1-4 Tank #33

Fuel return line from D G #33 to 1-4 Fuel Oil Day Tank #33. For venting D G system during starting. For system priming

Fuel oil supply line from flay 1-4 Tank #33 to D G #33. To' supply oil to D G

Diesel Engine #33 & Exhaust System 1-4

1/2" vent valve on connection to 1-4 6" inlet cooling water line to diesel generator venting valve. Rupture of vent line also considered.

1/2" vent valve on connection to 1-4 6" inlet cooling water line to diesel generator venting valve. Rupture of vent line also considered.


Sheet 82

Sheet 83Fire Zone 102A


INDIAN POINT 3 SYSTEMS INTERACTION STUDY


Interaction

Source name I MEMI. LfVL r

l's rfrn* 5~0

Loud Speaker

Space Heater

Emergency Light

Lights

V-1 1- 5-1

V-11-5-3

V-11-5-4

1-11-82-1




Generator Pane l 33

Diesel Engine #33 & Exhaust System



Starting and controlling of Diese-Generator #33

a' r'rf; ntI nfl

Safety Goal

1-4

1-4

1-4

2



10.0 AUXILIARY FEEDWATER SYSTEM INTERACTIONS

The auxiliary feedwater system was originally evaluated as part of a pilot

systems interaction study performed by Ebasco Services Incorporated and dated December, 1981. Many of the interactions found as a result of that study were

interactions also identified during the walkdown inspections of this study.

Since then, those interactions have been disposed of by the following methods:

a) Modification of the source component.

b) Analysis to determine that the source component will not fail

during a design basis earthquake.

c) Commitment by the Power Authority to modify the source component.

Where corrections have been completed, they are so indicated in the tabulations. Of the interactions identified during this study and not the

previous study, only 3 remain potentially unacceptable. They have been presented in Volume 23 as part of the results for future resolution.

10-1

APPENDIX 1

* NEW YORK POWER AUTHORITY INDIAN POINT NUCLEAR POWER PLANT


DEMONSTRATION OF IP-2/IP-3 SIMULATOR FOR POSSIBLE USE IN IP-3 SIS

I NTRODUTION:

This report describes a series of malfunction tests conducted by the Power Authority and demonstrated on the IP-2 simulator on the evening of October 29, 1981 to examine the capabilities of the simulator for use in the IP-3 Systems Interaction Study. The demonstration was conducted at Consolidated Edison's IP-2 Visitor Information Center /Simulator Training facility in Buchanan, N.Y.

Based on the results of these tests, this report provides conclusions relative to the capabilities of the simulator for use in systems interaction studies.

BACKGROUND:

During the NRC review of the methodology and criteria proposed in the IP-3 Systems Interaction Study, a concern was identified regarding diagnosing and/or predicting SI events for interconnected systems. The NRC SI staff requested that the Power Authority consider possible application of the Indian Point simulator in the treatment of "first-order" types of SI involving nonsafety control and power systems. The SI staff believed that because such a training simulator accurately models at least direct interconnections between safety and nonsafety front line systems and their support systems, it might be possible to do a more comprehensive and systematic analysis of their failure effects. In addition, the staff also noted that a training simulator would appear to be an almost ideal tool to be applied in treating in a systematic and comprehensive manner, nonsafety instrumentation display failure effects (i.e., induced operator error SI). The Power Authority agreed to investigate these possibilities and as a result of their investigation arranged for simulator time to examine some specific scenarios and failure combinations of particular interest.

On September 23 and 24, 1981, during two, one-hour sessions with members of the NRC SI staff, the IP-2 simulator facility was utilized to determine its capability to simulate first order, immediate reactions from predetermined system malfunctions.

MALFUNCTION TESTS:

All of the tests described herein are standard malfunctions programmed into the simulator computer and used in the normal operator training scenarios for the IP-2/IP-3 personnel. Each test was initiated from a 100% power level simulation. Immediately following the major upset initiated by the test malfunction, time on the simulator was "frozen" and the control board scrutinized for any apparent safety system interactions. When malfunctions resulted in normal designed safety system actions, they were not considered to. be systems interactions, e.g., the loss of 6.9 kV Bus 2 resulted in the

loss of one RCP which causes a reactor trip, all of which was designed and e Ixpected to occur. Prior to performing the malfunction tests, agreement was reached between the Power Authority and the NRC SI staff members present, that only first order reactions identified within a lapse time of 8 seconds or less would be completed. Long-term effects were discussed, but not necessarily allowed to run to completion because of the time lapse required.

CONCLUSIONS:

It can readily be seen from the attached Malfunction Test Tables that none of the tested malfunction scenarios indicated any systems interaction impact on safety related functions. In all cases, the simulator performed according to currently accepted design criteria. Such criteria however, does not demand the level of detail required to predict or diagnose systems interactions. The simulator is designed and programmed as a training tool, to be used to simulate accidents for operator instruction, based on current licensing acceptance criteria which does not include the scope of SI. To develop software programs of the level of complexity to adequately simulate systems interactions would require a comprehensive SI study utilizing the methodology currently proposed by the Power Authority. Therefore, for the simulator to display and realistically reflect systems interactions, all of the detailed SI studies would have to be performed to provide input to the computer program.

Thus, the current simulator and software programming is not capable of predicting or diagnosing SI any more accurately or comprehensively than the methodology currently proposed by the Power Authority.

0

SHEET I of 6




Event Half. Malfunction MalfunctionSafety-Related

Impact UC5LL*JJLIU~~~~t ItU fu- O~lL~Itf~fC*f5UUfL

I Steam Supply 22 Failure of Steam Dump

Control (52)

2 Steam Supply 23 Failure of Steam Dump

Control (30%)

3 Steam Supply 24 Failure of Automatic

Steam Dump Control (By-pass System)

4 Steam Supply 25 Gland Seal Regulator

Malfunction

5 Electrical

6 Turbine

26 Loss of External Electrical Load

27 Excessive Load Incident

7 Electrical 28 Station Blackout

B Electrical 29 Loss of 6.9 kV Bus 2

9 Electrical 30 Loss of 6.9 kV Bus 5

Dump valves fail and allow steam dumps to pass 52 of steam.

Dump valves fail and allow steam dumps to pass 302 of steam

False error signal cause slow opening of by-pass valves during temperature mode operation only.

Both Regulator Valves fail closed causing loss of sealing steam to LP Turbines and Main Feed Water Turbines.

External Load drops to zero instantaneously (Remote Breaker Buchanan Trips).

High Pressure oil drain line on Turbine Control Valve No. 3 servo positioner clogs causing valve to open completely.

Failure of Station Auxiliary Transformer.

Bus 2 is lost due to an electrical fault.

Bus 5 is lost due to an electrical fault.

X High steam 7, 82.

flow. MW decrease,

X High steam flow. MW decrease 18%. Tavg-Tref. deviation.

X Tavg-Tr'ef. deviation. Rod drive in.

X Steam seal press. to zero. Gradual loss of condenser vacuum.

X Instantaneous reactor trip.

X Load increase.

X 6900V power supply failure. Low vacuum trip. Diesels start. Loss of two vital busses, but no SR functions on them.

X Reactor trip due to loss of flow in one loop. Does not teed 480V Bus.

X Lose all loads. Simulation is such that diesels do not start immediately, therefore No. 21 & 22 containment recirc. fans and No. 71 SI pump are lost.

0.u System "o. lilie Description les .0 RLBLcLant Reactionsia-el~ nts

i. * . . a ,

I

SHEET 2 of 6


SYSTEMS INTERACIION STUDY

DEMONSTRATION OF IP-2/IP-3 SIMULATOR FOR

POSSIBLE USE IN IP-3 SIS

Half. Malfunction Malfunction

Safety-Related Impact

V.. MN R4 .ni fleant ReactionsiCommenta

31 Auto Voltage Regulator Failure, Excessive Lag VARS

32 Auto Voltage Regulator Failure, Excessive Lead VARS

12 Electrical 33 Automatic Voltage Regulator Failure on Main Generator

13 Electrical 37 Unit Auxiliary Transformer Heating Up

14 Component Cooling

15 Condensate and Feedwater

58 Loss of Auxiliary Coolant Water to Reactor Coolant Pump

65 Loss of Normal Feedwater

Voltage Regulator fails. Generator p'icka up lag VARS in ramp function of

150 KVARS/sec until field breaker trips.

Voltage Regulator fails. Generator picks up lead VARS in ramp function of 150 KVARS/sec until field breaker overloads.

Loss of Regulator (Buck-Boost) Control Action.

Unit Auxiliary Transformer, Primary

Cooling System, fails; Automatic Switchover fails also. Temperature increases until limit of 121lC is reached.

Valve 771A is closed due to remote

manual mia-position. Coolant water flow to RCP No. 21 thermal barrier and bearing oil coolers goes to zero, heat transfer rates at the heat exchanger and oil coolers will decrease to zero. RCP No. 21 temperatures will rise. "Bearing Coolers and RCP No. 21 LO Flow" alarm on CCR ACS panel will be annunciated.

Normal Feedwater Supply to the steam generators is lost. Resulting from both main boiler feed pumps tripping due to loss of turbine lube oil.

X Equipment heats up. trip.

Reactor

X With operator action, no SR impacts.

X Unit trip. reactor trip. AFW system auto-start.

Event

10 Electrical

11 Electrical

No. System No. Title scr pt on a Yes No Si nifleAnt RoArtinnalCommants

S

SHEET 3 of 6


SYSTEMS INTERACTION STUDY DEMONSTRATION OF IP-2/IP-3 SIMULAOR FOR


Half. Malfunction tdt~ n. Tirll

Malfunction Descrintion


Yes No Significant Reactions/Comments

16 Condensate and 67 Feedwater






Reduction in Feedwater Enthalpy Incident (sudden decrease in FW, Temperature)

Loss of one Feedwater Pump

Loss of Feedwater to No. 22 S/G

Loss of Feedwater to No. 23 S/C

Condenser Tube Leak (Variable)

Loss of Condensate Pump

Low Pressure Feedwater Heater By-Pass Valve FCV 1150 suddenly fails opened. This results in excess heat removal by the secondary system.

Feedwater Pump No. 21 fails due to thrust bearing wear. Results in a reduction of teedwater flow to the steam generators.

FCV 427 fails closed due to loss of signal from the Feedwater Control System.

FCV 437 fails closed due to loss of signal from the Feedwater Control System.

Circulating Water Tube in Condenser 21 springs a teak and the condensate will become contaminated with river water. The severity of this leak will increase with time.

Condensate Pump No. 22 fails due to a thermal overload. Severity of this malfunction is related to the operating power level

100% requires 3 pumps 852 requires 2 pumps 43% requires I pump

X Sudden cooldown.

X Reactor trip.

X Reactor trip.

X Reactor trip.

No reactor trip.

Event Nn -

No S stem No Title

0




Half. Malfunction System No. Titl.

MalfunctionSafety-Related

Impact

-..... -..kL~ .-- -- OLKInLL*U~flI. IW*a LLon|[tP..Otle







28 Chemical Volume 91 Control System



31 Chemical Volume 97 Cuntrol System

Hotwell Level Control Failure

Lose of vacuum

Gland Steam Condenser Tube Rupture

Loss of Heater Drain Pump

High Pressure Feedwater Heater Tube Leak

Loss of Circulating Water Pump

Dilution

Lose of Auto Makeup to Volume Control Tank

Non-Regenerative Heat Exchanger Coolant Loss

Loss of One Charging Pump

The Hotwell Level Control (LC 1128) located on Hotwell No. 22 will fail in the low direction. The hotwell level is normal but the level controller responds as if the hotwell level were low.

There is a leak in the wall of Condenser No. 23 resulting in a sudden and rapid

loss of condenser vacuum.

The Condensate Tube in the gland steam condenser ruptures.

Heater Drain Pump No. 21 fails due to a thermal overload.

There is a sudden large rupture of one of the feedwater tubes in Heater 26A.

Circulating Water Pump No. 25 fails due

to a thermal overload.

Primary Water Counter (YIC-lll) continues to count beyond set value and results in FCV-IIA remaining open.

Start switch makes faulty contact. Makeup does not start when demanded.

Failure of Auto Manual Controller HTC-130 causes Butterfly Valve TCV-130

in Auxiliary Coolant System to close.

One of the operating Charging Pumps

trips due to low oil pressure.

Reactor trip.

X No immediate reaction.

X Lose of Pb.

X No reactor trip. See Bus 5A failure.

X Cannot be modeled an simulator.

Automatic transfer to the gWST.

X Lost of CCW, VCT tamp increase, diversion required to by-pass ion exchanger.

X Low P on thermal barrier.

S

SHEET 4 st 6

Event No.


SYSTEMS INTERACTION STUDY DEMONSTRATION OF IP-2/IP-3 SIMULA7OR FOR


Half. Malfunction System No. Title


v. * u..

32 Chemical Volume Control System

33 Chemical Volume Control System

34 Turbine

35 Turbine

36 Turbine

37 Turbine

38 Turbine

102 Volume Control Tank Level Control Failure

103 Boric Acid Storage Tank Heater Failure

132 Turbine Control Valve Malfunction

133 Turbine Generator Malfunction (Vibration, Eccentricity).

134 Lube Oil Temperature High

135 Lube Oil Pressure Turbine

137 High Pressure Oil Malfunction

39 Instrument 142 Rapid Loss of Instrument Air

Level Transmitter LT-112 fails - no signal.

Electrical short circuit fails heater in No. 22 Boric Acid Tank

Control Valve No. 2 fails closed due to mechanical failure in servomotor.

Rotor shaft distortion causes abnormal readings on vibration/eccentricity recorder. Ramp function.

Temperature Control Valve (TCV 1102) fails closed causes reduced service water flow through oil cooler.

Break in LP oil line causes moderate but continuous decay in oil pressure from its steady state pressure to 372 of that value in 15 minutes.

High Pressure Oil regulating device opens causing immediate pressure drop which reduces control valve openings.

Large rupture of 3" Instrument Air for Nuclear Service line immediately downstream of Valve 1A-52. All pneumatic actuated instrumentation and valves respond to loss of instrument air. Equipment having own accumulators continue to function.

X VCT outlet valve closes. RWST makeup goes open.

X No immediate reaction, loss of temperature over time.

X Control rods drive in, no trip, unit stabilizes at lower power level.

x No trip.

X No trip. Manual action to restore.

Loss of load.

Containment isolation. All valves go to safe position. Letdown isolated.

Event No.

S

SHEET S of 6

Malfunction Dleserintinn

&= a as, ons omme Le Descri tion Yes No Si ;fic n- R -4 Ir

* I i*. I

SHEET 6 of 6


SYSTEMS INTERACTION STUDY DEMONSTRATION OF IP-2/IP-3 SIMULATOR FOR


Half. Malfunction System No. • Title

Malfunction Deneri ntnn


V.. Mn

40 Fire and Smoke 151 Detection






Fire in Control Building Elevation 33 ft.

Fire in Control Building Elevation 15 ft.

Fire in Electrical Tunnel Elevation 33 ft. to 68 ft.

Fire in Electrical Tunnel Area Elevation 58 ft.

Fire in Piping Penetration Area Elevation 51 ft.

Fire in Electrical Penetration Area Elevation 46 ft.

Zone 1 Indicator Lamp on PYR-A-LARH

panel illuminates.

Zone 2 Indicator Lamp on PYR-A-LARH panel illuminates.

Zone 3 Indicator Lamp on PYR-A-LAIR panel illuminates.

Zone 4 Indicator Lamp on PYR-A-LARM panel Illuminates.

Zone 5 Indicator Lamp on PYR-A-LARM panel Illuminates.

Zone 6 Indicator Lamp on PYR-A-LARM panel illuminates.

X PYR-A-LAIM alarm Illuminates, Fire Cable Tray Annunciator trips.

X PYR-A-LARE alarm illuminates, Fire Cable Tray Annunciator trips.

X PYR-A-LARM alarm illuminates, Fire Cable Tray Annunciator trips.

x PYR-A-LARM alarm illuminates, Fire Cable Tray Annunciator trips.

x PYR-A-LAtH alarm illuminates, Fire Cable Tray Annunciator trips.

x PYR-A-LARN alarm illuminates, Fire Cable Tray Annunciator trips.

Event No.

Yes No-8, on@ men Descri tion St nifica t R -4 Ir- .;.R4om4F4rmm Rmm o4 mmm ir^ o

is

APPENDIX 2


INDIAN POINT NUCLEAR POWER PLANT...


SIGNIFICANT OCCURRENCE REPORT REVIEW

BACKGROUND:

The NRC SI staff emphasized that the consideration of operating experience was

an important element in a systems interaction analysis and could be utilized

for two purposes. First, documented operating experience could be

extrapolated to predict systems interactions. Second, documented operating

experience could be used to demonstrate the feasibility of any proposed

methodology by comparing the results of an SI study with the documented

operating experience to show that SI's identified by the study were similar to

those that occurred in the past.

EFFORT:

Recognizing that the examination of historical data could assist in the

identification of future systems interactions, the SI study team conducted a

review of the Significant Occurrence Reports (SOR's) that had been reported

during 1980 and the first half of 1981 for IP3. The documents that were

examined, however, contained material which was outside the scope of the SI

study (i.e., operation errors on systems not included in the study), making

the results of the review inconclusive. The logic utilized to search for

potential SI's is shown on the attached diagram.

CONCLUSIONS:

While the examination of the historical data proved to be without merit for

this particular study, it may still be useful as an evaluation tool to judge

the relative effectiveness of other proposed SI studies. It should be noted

that only some SOR's are the results of systems' interactions and that their

utility will be limited.

-EBASCO SERVICES IN PRATED... W. .YORK

Client: New York Power Authority

Project: Indian Point Nuclear Power Plant. Systems Interaction Study

Subject: Significant Occurrence Report Review Logic Diagram

Does SOR Involve at Least One Safet'i Related System, Structure, or Component and One Nonsafety System, Component o--Structure?

Indian Point 3 Nuclear Power Plant Systems Interaction · 2012. 12. 4. · INDIAN POINT 3 NUCLEAR...

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