ML20128D387
| ML20128D387 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/31/1985 |
| From: | Dennis Bley, John Stetkar, Stillwell D PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.) |
| To: | |
| Shared Package | |
| ML20128D367 | List: |
| References | |
| PLG-0367, PLG-367, NUDOCS 8507050098 | |
| Download: ML20128D387 (269) | |
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{{#Wiki_filter:m PLG-0367 j Revised i SOUTH TEXAS PROJECT PRELIMINARY SCOPING STUDY RESULTS s Prepared for HOUSTON LIGHTING AND POWER COMPANY Houston, Texas May 1985 [BPRB82!8888lk Pickard,Lowe andGarrick,Inc. Engineers e Applied Scientists e Management Consultants Newport Beach, CA Washington, DC ___ ]
i . . . , PLG-0367 Revised SOUTH TEXAS PROJECT PRELIMINARY SCOPING STUDY RESULTS Principal Investigator Dennis C. Bley Task Leaders John W. Stetkar Daniel W. Stillwell Other Key Technical Contributors Rick Grantom (HL&P) Ali Mosleh Shawn S. Rogers (HL&P) Harold F. Perla Vicki M. Bier Kathleen C. Ramp Mardyros Kazarians Shobha B.Rao Thomas J. Miksch! Martin B. Sattison Alfred Torri Technical Review Stan Kaplan Donald J. Wakefield Douglas C. Iden Project Management Karl N. Fleming Harold F. Perla B. John Garrick Prepared for HOUSTON LIGHTING AND POWER COMPANY Houston, Texas May 1985 Pickard,Lowe andGarrick,Inc. Engineers e Applied Scientists e Management Consultants Newport Beach, CA Washington, DC
ACKNOWLEDGMENT The South Texas Project study team is indebted to those not identified on the title page who made invaluable contributions to the successful completion of this report. Mr. Richard P. Murphy, the project manager from Houston Lighting and Power Company, made an outstanding contribution through his project management support and coordination of the essential participation of several different HL&P organizations. These organizations included nuclear services, project engineering, nuclear licensing, reactor operations, technical support, operator training, and maintenance organizations that support the South Texas Project Electric Generating Station as well as HL&P power supply planning and engineering and system load control. Through their supply of information, interactions with the study team, and review of project deliverables, the objective of achieving substantial realism and accuracy in modeling the STPEGS plant and system was achieved. The authors extend a special appreciation to the noteworthy contributions of those responsible for publishing this report on a very tight schedule and without compromising report quality as well as those who provided valuable administrative support. iii 0097H052185
ABSTRACT Houston Lighting and Power Company, as project manager for the South Texas Project Electric Generating Station, embarked on this Preliminary Scoping Study early in 1984 as a preliminary analysis of I plant safety using probabilistic methods. The study was performed solely to satisfy internal HL&P needs for timely feedback of risk management insights into the process of completing construction on STPEGS and preparing the plant for operation. With the design nearing its final form and the operating staff beginning the process of training and procedure development, the time was ripe for a review of the integrated plant--the interconnected hardware systems and their operating environment. The Preliminary Scoping Study provides that analysis. The Preliminary Scoping Study employs an advanced probabilistic risk assessment methodology developed and applied by Pickard, Lowe and Garrick, Inc., as expert technical consultant to HL&P. The work has benefited from substantial participation and support from the HL&P engineering, operations, and training organizations. The objectives of the study are aimed toward the development of a plant risk model uniquely applicable to STPEGS, both to provide early insights into the risk sensitivities for design and procedural purposes and to be continually used throughout plant life as a living risk management tool. The completion of the Preliminary Scoping Study has resulted in the development of the nucleus of a plant model, a preliminary safety quantification, and the identification of early insights into risk-sensitive factors about STPEGS. iv 0097H052185
COIttPuy iie"s'<- 'ix 'i"x n &i er it o ii"x i7"" ii""s'"". Tex"' 77""> <7's> "2"2ii HL&P Perspective Houston Lighting & Power Company, as Project Manager for the South Texas Project Electric Generating Station, initiated this Preliminary Scoping Study early in 1984 as part of a phased program aimed at developing a risk model through programmatic activities consistent with the development of a Level 1 PRA. This Preliminary Scoping Study supports this objective through the development of a p rel imi n a ry model for analyzing plant sa fety . The primary objectives of the Preliminary Scoping Study were to provide early insights into the design and operational sensitivities related to plant safety and to provide a comprehensive training vehicle for HL&P pe rsonnel in the probabilistic safety technology. These goals have been accompitshed through the completion of this preliminary model, _ which this report documents. During the time this Study was being completed, calculations were performed to provide additional information related to temperature rise on loss of EAB HVAC and the decision was made to install fail-closed isolation valves in the Supplemental Containmont Purge Subsystem. No additional activities are indicated at this time other than further development of the plant model through the reduction of uncertainties. In planning and implementing the STP plant model development program, the technology transfor required for HL&P personnel to assume a larger responsibility in the expansion of the model was provided through the integration of HL&P personnel into the PL&G Study team. The next phase will consist of an in-house program, supplemented by outsido resources, key areas. Periodically the program will be reevaluated to to reduco uncertaintios in doftne the lovel of activity required to support STP noods. v
CONTENTS Section Page ACKNOWLEDGEMENT iii ABSTRACT iv HL&P PERSPECTIVE y LIST OF TABLES AND FIGURES ix LIST OF ACRONYMS xii 1 INTRODUCTION 1-1 1.1 The Preliminary Scoping Study Project 1-1 1.2 PRA Perspective: What is a Preliminary Scoping Study? 1-3 1.2.1 Risk Management Perspective: Examples from Other Studies 1-6 1.2.2 Perspective on the Preliminary Scoping Study 1-7 1.3 STP/ Preliminary Scoping Study Project Perspective 1-10 1.4 Report Guide 1-12 1.5 References 1-13 2 RESULTS 2-1 2.1 What is the Likelihood of Core Melt? 2-2 2.2 What is the Likelihood of Offsite Impact? 2-4 2.3 Identification and Quantification of Uncertainty 2-5 2.3.1 Data Uncertainties 2-6 2.3.2 Uncertainties in Assumptions 2-7 2.3.3 Completeness 2-12 2.4 Quantification for One Example Assumption Set 2-12 2.4.1 Assumptions 2-13 2.4.2 Recovery Analysis 2-14 2.4.3 Principal Contributors to the Results 2-15 2.4.4 Sensitivity to Supplementary Purge Options 2-18 2.5 Impact of Assumption Sets on Dominant Sequences 2-19 2.6 References 2-20 3 STPEGS RISK MODEL OVERVIEW 3-1 3.1 Basic Concepts and Definitions in PRA 3-1 3.1.1 Probability and Frequency 3-1 3.1.2 Hazard and Risk 3-4 3.1.3 Decision Analysis 3-6 3.2 Risk Model Structure 3-6 3.2.1 Qualitative Description of STPEGS Risk Model 3-6 3.2.2 Logical Structure of a Risk Model 3-8 3.3 Matrix Formulation of a Risk Model 3-9 3.4 Decomposition of Risk and Cause Tables 3-10 3.5 Overview of STPEGS Scoping Study Risk Model 3-11 3.5.1 Plant Model Limitations 3-11 3.5.2 Containment Model Limitations 3-12 3.5.3 Site Model Limitations 3-12 vi 0097H052185
CONTENTS (continued) Section Page 3.6 References 3-13 4 EVENT SEQUENCE MODEL 4-1 4.1 Overview - Initiating Events, Auxiliary and Frontline System Models 4-1 4.2 Initiating Events 4-2 4.3 Auxiliary Systems Model 4-3 - 4.3.1 General Infonnation 4-3 4.3.2 Event Tree Description 4-4 4.3.3 Dependencies Between Main Line Systems and Auxiliary Systems 4-12 4.3.4 Quantification of the Auxiliary Trees 4-12 4.3.5 Auxiliary Systems Event Tree Results 4-12 (7 4.4 Frontline Systems Model 4-13 % 4.4.1 The General Transient Event Sequence Model 4-13 4.4.2 The Small LOCA Event Sequence Model 4-17 4.4.3 The LT1 Long-Term Response Event Sequence Model 4-18 4.4.4 The LT2 Long-Term Core Melt Response Event Sequence Model 4-19 5 SYSTEMS ANALYSIS 5-1 5.1 Systems Analysis Approach 5-1 5.2 Summary of Systems Analyzed 5-2 5.3 Example Systems Sumary - Essential Cooling Water System 5-2 5.4 References 5-4 l 6 SURVEY OF EXTERNAL EVENTS AND INTERNAL PLANT HAZARDS 6-1 l 6.1 Overview 6-1 6.2 Seismic Events 6-1 6.3 In-Plant Hazards and Spatial Interactions 6-2 6.3.1 Electrical Auxiliary Building 6-3 6.3.2 Mechanical Auxiliary Building 6-4 6.3.3 Fuel Handling Building 6-4 6.3.4 Reactor Containment Building 6-5 6.3.5 Diesel Generator Building 6-5 6.3.6 Isolation Valve Cubicle 6-5 6.3.7 Essential Cooling Water Intake Structure 6-5 6.3.8 Other Buildings and the Yard 6-5 6.4 Other External Events 6-6 6.4.1 Aircraft Hazard Analysis 6-6 6.4.2 Turbine Missile Risk 6-10 6.4.3 Tornado Wind and Missile Risk 6-11 6.4.4 Hazardous Chemical Analysis 6-15 6.4.5 External Flooding 6-16 6.5 References 6-17 vii 0097H052085
CONTENTS (continued) Section Page 7 SCOPING CONTAINMENT RESPONSE ANALYSIS 7-1 7.1 Introduction 7-1 7.2 STPEGS Containment Design 7-1 7.3 Plant Damage States 7-2 l 7.4 Release Categories 7-3 ! 7.5 The C Matrix 7-3 7.6 Site Consequence Analysis 7-4 APPENDIX A: RESOLUTION OF HL&P COMMENTS ON STP PROBABILISTIC SAFETY ASSESSMENT BASELINE STUDY INTERIM REPORT DATED JULY 1984 A-1 APPENDIX B: RESOLUTION OF HL&P COMMENTS ON STP PROBABILISTIC SAFETY ASSESSMENT BASELINE STUDY READING DRAFT { DATED DECEMBER 1984 B-1 APPENDIX C: RESOLUTION OF HL&P COMMENTS ON STPEGS DRAFT REPORT DATED FEBRUARY 1985 C-1 viii 0097H052085
LIST OF TABLES AND FIGURES Table Page 1-1 Dominant Accident Sequences for Seabrook Phase I Results 1-16 1-2 Summary of Accident Sequences with Significant Risk and Core Melt Frequency Contributions from Phase II of the Seabrook PRA with Applicable Phase I Sequences 1-17 2-1 Initiating Event Frequencies for STPEGS Scoping Study Quantification 2-21 2-2 Relationships of Scenarios to Offsite Impact Categories 2-22 2-3 Accident Sequences With Significant Contributions to Core Melt Frequency Based On Point Estimate Example Case Assumptions and STPEGS Model 2-23 2-4 Definition of Supplemental Purge Isolation Valve Options Quantified Using STPEGS Model 2-25 E 2-5 Sensitivity of Results to Supplementary Purge Isolation Valve Design Options 2-26 4-1 Catalog of Potential Initiating Events (Taken from the Seabrook Station Probabilistic Safety Assessment) 4-20 4-2 Initiating Events Quantified in the STPEGS Study 4-22 j 4-3 Auxiliary Systems Included in Auxiliary Event Trees 4-23 4-4 Matrix of Auxiliary System to Auxiliary System Interdependencies 4-25 4-5 Auxiliary Systems Not Shown in the Plant Auxiliary System Model s 4-27 4-6 Electric Power Tree Unique End States 4-28 4-7a Effect of Electric Power End States on Other Auxiliary Systems (General Transient) 4-32 4-7b Effect of Electric Power End States on Other Auxiliary Systems (Small LOCA) 4-7c Effect of Electric Power End States on Other Auxiliary Systems (Loss of Offsite Power) 4-39 4-8 Auxiliary Tree End States 4-40 4-9 Auxiliary Tree End States 1 4-42 4-10 Matrix of Auxiliary System to Main Line System Interdependencies 4-56 4-11 Auxiliary Systems Impact Vectors 4-58 4-12 Impact Vector Definition 4-59 4-13 General Transient Event Tree Success Criteria 4-61 4-14 Coding for Transfer States Between Early and Long-Term (LT-1) Response Event Trees 4-65 4-15 Coding for Transfer States Between Early and Long-Term (LT-2) Core Melt Response Event Trees 4-66 4-16 LT1 Long-Term Event Tree Success Criteria 4-67 5-1 STPEGS Systems Partially or Fully Analyzed in the Preliminary Scoping Study 5-5 5-2 Other STPEGS Systems Under Consideration for a Complete Level 1 Analysis 5-7 5-3 Essential Cooling Water System Component Dependencies 5-8 5-4 Essential Cooling Water System Loads 5-9 ix 0097H052185
LIST OF TABLES AND FIGURES (continued) Table Page 5-5 Normal Alignment for Essential Cooling Water, Component Cooling Water, and Essential Chilled Water Used for the Preliminary Scoping Study 5-10 5-6 Components Included in Essential Cooling Water System Model Blocks 5-11 5-7 Essential Cooling Water System Unavailability Expressions 5-12 5-8 Example Calculation of Hardware Unavailability for Essential Cooling Water System Model Block PA 5-13 5-9 Example Calculation of Essential Cooling Water System Unavailability 5-14 6-1 Aircraft Accidents and Accident Rates: U.S. Air Carriers, 1970 through 1979 6-19 6-2 Fatal Accident Rates for U.S. General Aviation Aircraft 6-20 6-3 Probabilities f2 and f3 of Targets Due to Units 1 and 2 Turbine Missiles (Shear Failure) 6-21 6-4 Probabilities f2 and f3 of Targets Due to Units 1 and 2 Turbine Missiles (Shear and Rotation Failure) 6-22 6-5 Tornado Windspeed Fractions 6-23 7-1 Plant Damage States for the STPEGS Preliminary Scoping Study 7-5 7-2 Definition of Release Category Sets Based on Containment Failure Modes 7-6 7-3 Definition of Release Categories for STPEGS Preliminary Scoping Study 7-7 7-4 C Matrix for STPEGS Preliminary Scoping Study Case I: No Water Access to Reactor Cavity; Debris Not Coolable 7-8 7-5 C Matrix for STPEGS Preliminary Scoping Study Case II: Water Access to Reaction Cavity; Debris Coolable 7-9 7-6 S Matrix for Case I - The Conditional Probabilities of Impact Categories Given Release Categories (Corresponds to C Matrix for Case I) 7-10 7-7 S Matrix for Case II - For Conditional Probabilities of Impact Categories Given Release Categories (Corresponds to C Matrix for Case II) 7-11 7-8 The CS Product Matrix--Conditional Probabilities of Impact Categories Given Plant Damage Sites 7-12 Figure 1-1 Examples of PRA Applications 1-18 1-2 Comparison of Core Melt Frequency Results Obtained in the Preliminary and Full Assessment Phases of the SSPSA 1-19 1-3 Seven Options to Defining the Scope of a Risk Analysis 1-20 2-1 Probability Distribution for Core Melt Frequency 2-26 2-2 Probability Distributions for Scenario Group Frequencies 2-28 2-3 Probability Distributions for Impact Category Frequencies 2-29 x 0097H052085
LIST OF TABLES AND FIGURES (continued) Figure 3-1 Probability' Curve Against Failure Frequency Based on Salesman s Statement and Perceived Bias of Salesman 3-14 3-2 State of Knowledge Probability Curve After Learning Neighbor's Experience 3-14 3-3 Pcint Estimate Risk Curve with No Uncertainty Quantification 3-15 3-4 Family of Risk Curves with Uncertainty Quantified 3-15 3-5 Role of PRA in Decision Analysis Process 3-16 3-6 Block Diagram Structure of a Full Scope Level 3 Risk Model 3-17 3-7 Standard Form of Accident Sequences in Full Scope, Level 3 Risk Model 3-18 3-8 Matrix Formulation and Risk Assembly Process 3-19 3-9 Overall View of the PRA Assembly Process Showing Relationships of Pinch Points, Event Trees, Frequency Vectors, and Transition Matrices 3-20 3-10 Progressive Steps in Risk Decomposition 3-21 4-1 STPEGS Plant Event Sequence Model 4-70 4-2 Auxiliary System Event Tree Relationships 4-71 4-3 Electric Power Event Tree - No Loss of Offsite Power Condition 4-72 4-4 Simplified Electric Power Event Tree for Loss of Offsite Power 4-79 4-5 Auxiliary Systems Event Tree for Offsite Power Available 4-80 4-6 Auxiliary Systems Event Tree for Loss of Offsite Power Condition 4-86 4-7 Event Sequence Diagram Symbology 4-91 4-8 General Transient Event Sequence Diagram 4-92 4-9 Simplified General Transient ESD Using Only the Event Tree Top Events 4-95 4-10 General Transient Event Tree 4-96 4-11 Small Loca Event Tree 4-100 4-12 LT1 Long-Term Response ESD 4-101 4-13 Frontline Event Tree LT1 4-102 4-14 LT2 Long-Term Core Melt Response ESD 4-104 4-15 Frontline Event Tree LT2 4-105 5-1 STPEGS Systems Requiring Little or No Further Analysis 5-16 5-2 South Texas Project Systems Analysis Summary Outline 5-17 5-3 Essential Cooling Water System Piping and Instrumentation Diagram 5-18 5-4 Essential Cooling Water System Block Diagram Model 5-20 6-1 Seismic Analysis 6-24 6-2 Geometry for Impact Probability Model 6-27 xi 0097H052085
n i I LIST OF ACRONYMS 4 Abbreviation Definition AFW . auxiliary feedwater AFST ' auxiliary feedwater storage tank CCWS ccmponent cooling water system DGB diesel generator building EAB electrical auxiliary building ECCS emergency core cooling system ECH essential chilled water ECW essential cooling water EDG emergency diesel generator ESD
-ESF event sequence diagram engineered safety features j
ESFAS engineered safety features actuation system ET event tree FHB fuel handling building FMEA failure modes and effects ar.alysis FSAR final safety analysis report J HL&P Houston Lighting and Power Company HVAC heating, ventilating, and air conditioning system IVC isolation valve cubicle LOCA loss of coolant accident MAB mechanical auxiliary building M0V motor-operated valve NRC Nuclear Regulatory Commission NREP National Reliability Evaluation Report PLG Pickard, Lowe and Garrick, Inc. PORV power-operated relief valve PRA probabilistic risk assessment PWR pressurized water reactor QA quality assurance RCB reactor containment building RCP reactor coolant pump RCS reacter coolant system RiiR residual heat removal RSS Reactor Safety Study RWST refueling water storage tank xii 0097H052085
LIST OF ACRONYMS (continued) Abbreviation Definition SI safety injection SMA Structural Mechanics Associates, Inc. SSSP solid state protection system STPEGS South Texas Project Electric Generating System TMI Three Mile Island xiii 0097H052085 i.
- 1. INTRODUCTION This report documents the results of the South Texas Project Electric Generating Station Preliminary Scoping Study of plant safety--a limited application of probabilistic risk assessment technology. The study provides a preliminary risk model of the integrated STP plant, a first-cut quantification, and early insights into the sensitivities for design and procedural purposes. It can serve as the foundation for a practical risk management program for STPEGS.
1.1 THE PRELIMINARY SCOPING STUDY PROJECT Houston Lighting and Power Company, as project manager for STPEGS, conceived of the Scoping Study as a means to examine the integrated response of the plant to unplanned departures from steady-state operations. Such an integrated analysis was especially appropriate at this time. Individual system designs, all meeting their own independent design criteria, were nearing completion. They could be used to develop corresponding system models, which could be linked together to study interactions and whole-plant response. Moreover, significant construction remained and the operating staff was just beginning to train and write procedures on integrated plant operations. The Scoping Study described in this report began on March 28,1984 to satisfy internal HL&P needs for timely feedback of risk management insights into the process of completing construction on STFEGS and preparing the plant for operation. PLG assumed the lead responsibility as expert technical consultant to plan, direct, and conduct the study and to begin the process of providing HL&P with PRA technology. This technology transfer has enabled HL&P's direct participation in the study and can develop at HL&P the type of intimate knowledge of the STPEGS risk model that will be necessary for a variety of risk management applications. To help focus the technology transfer effort, two HL&P engineers were assigned full time to the project team for the duration of the Scoping Study. In addition, the engineering, operations, and training organizations at HL&P provided plant documentation, answered numerous technical inquiries, and reviewed project deliverables. HL&P's project manager coordinated this effort and represented the third full-time HL&P engineer participating on the project team. The results of this study, therefore, reflect the contributions of both HL&P and PLG. Nevertheless, PLG was given and fully assumes the responsibility for the results of this Scoping Study and is fully prepared to explain the results. STPEGS design information included in this report is derived from design data available during the period from April to July 1984. Changes that have been made in the normal course of plant design since that time generally have not been analyzed or incorporated into the study results. Specific cases where more recent information has been included are
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explicitly identified in the HL&P comments attached as Appendices A and B. 1-1 0097H052185
The ANS/IEEE PRA Procedures Guide (Reference 1-1) defines three levels of PRA coverage. A Level 1 PRA analyzes only the performance of the plant systems, whereas a Level 2 PRA also considers core and containment phenomenology. A Level 3 PRA includes a complete consequence analysis. The Procedures Guide describes the effort required to conduct and fully document a state-of-the-art analysis at each of the three levels. Such analyses strive for completeness in coverage of initiating events and the modeling of the ensuing scenarios as well as the development of thorough, stand-alone documentation. This Preliminary Scoping Study is much more limited in coverage than any of the PRAs discussed above. The core of work in this effort could be expanded into a complete Level 1 PRA at a later date. The analytical coverage has been limited to those initiating events and aspects of plant response judged most important by the study team. Such limitations lead to broad uncertainties in the study results in comparison to a full-scope assessment. Documentation in this report has been limited to that required to present the results and their bases. The validity of the current study in describing the risk comes from the experience of its analysts--their involvement in many full-scope PRAs, their familiarity with the plant, and their first-hand knowledge of plant operations. Its utility in risk management comes from its completeness as a model of the integrated plant and its flexibility: completeness in terms of how thoroughly it models the interrelationships among plant systems at an appropriate level of detail and flexibility in terms of ease of expansion to include analytical refinements; i.e., the ability to take important parts of the model to the full-scope level to answer particular risk-related questions requiring more detail. Although the Preliminary Scoping Study is complete, it is not as detailed as a full-scope assessment. The objectives of the Preliminary Scoping Study can be summarized as follows: o Provide a basis for timely feedback of risk management insights into the process of completing construction on STPEGS. e Develop a cadre of experienced experts in the STP plant model, systems, and operations at PLG and HL&P. e Develop the nucleus of an STPEGS plant risk model consisting of event sequence diagrams and event trees for a limited set of initiating events, abbreviated systems models, a generic data ~ base, a simplified containment model, and a qualitative analysis of the impact of event and accident sequences. e Provide a state-of-knowledge estimate of the frequency of accidents i involving severe core damage, plant damage states, and impact l categories; i.e., an estimate including uncertainty. ' e Develop a basis for optimizing the allocation of resources to complete the development of risk assessment and risk management capability at HL&P. 1-2 0097H052085
The purpose of this report is to document the successful completion of the Scoping Study objectives. The uncertainties in the numerical results presented in this report have been quantified and are large because the model is incomplete and both the state of knowledge of the study team and the extent of supporting documentation are not as strong as they would be following a full-scope study on a completed plant. Rather than draw undue emphasis in this report to the numerical results, the authors have attempted to use these results to lend credence to the qualitative engineering conclusions and recommendations that were developed. The point value numerical results would most assuredly change at the completion of a full-scope PRA. However, past PLG experience with similar studies provides the authors with confidence that a major portion of the principal risk contributors and risk-sensitive factors of STPEGS have been identified in this study. This report incorporates the authors' response to HL&P written comments on earlier review drafts. As required by HL&P Policies and Practices (Reference 1-2), the appendices contain those comments and our resolution of them. 1.2 PRA PERSPECTIVE: WHAT IS A PRELIMINARY SCOPING STUDY? With the growth of the nuclear industry and the accumulation of a substantial body of operating experience, a relatively new method of safety assessment has developed that differs from the safety analyses traditionally employed in the licensing process. This method is known as probabilistic risk assessment. Its development for nuclear power plant application follows its successful application in the aerospace industry. The introduction of PRA was inspired by the assessment that PRA can be a powerful decision-making and risk management tool (References 1-3 through 1-6). PRA looks quantitatively at both the consequences that could potentially result from nuclear reactor accidents and the likelihoods of such occurrences. Consequences of all scenarios are calculated, rather than simple bounding cases. Then, for each consequence level, PRA calculates the likelihood of occurrence rather than assuming that some events are " incredible" and all others have equal weight. Combining consequences, likelihood of consequences, and uncertainty enables the ranking of pricrities for accident scenarios with respect to risk. Also, PRA establishes a framework for putting the risk of nuclear plant operation into context with other public risks. An identifying characteristic of PRA is its probabilistic rather than deterministic viewpoint toward safety. The use of probabilities and frequencies provides a means for dealing with random elements such as variation in certain environmental conditions (e.g., occurrence of earthquakes and floods) and certain internal plant failures exhibiting random behavior. It also provides a framework for quantifying uncertainty in otherwise deterministic accident simulation and consequence models. PRA has also evolved a technology for handling rare events that expands the completeness of accidents considered. Consideration of man-machine interfaces in PRA is also given emphasis. 1-3 0097H052085
Final results of a PRA are expressed in the " probability of frequency" format; i.e., as a family of curves giving the frequency of exceeding damage levels and the uncertainty in these frequencies. This family of curves embodies the three-dimensional aspects of risk: uncertainty, likelihood, and consequences. A common misconception about PRA is that its validit', or usefulness strongly hinges on having large quantities of data available. In fact, modern PRA accounts for variability in data (or lack of data) by assigning probability distributions to model these uncertainties and examining the effects of these distributions on the results. The outcome is a probability distribution of the results. These distributions describe the results in terms of means, medians, modes, high and low percentile values, and other distribution parameters of interest. Oversimplified thinking by some people in the past has led to the identification of the central tendency prediction (i.e., mean, median, or mode) as the " bottom-line" result. This thinking ignores the distribution that embodies the statement of uncertainty. The uneasiness about the accuracy of the central estimate wrongly leads to a discounting of the value of PRA, ironically because of a concern about the uncertainties. These uncertainties are an important consideration because the results may be telling us to study the issue further before making decisions about plant modifications or other actions to reduce risk. PRA has been developed and applied in a number of activities. A partial list of examples is presented in Figure 1-1. Included are PRAs on chemical plants and commercial aircraft systems (References 1-7 through 1-11). However, the most common application of PRA has been to assess the safety of nuclear power plants (References 1-12 through 1-21, for example). The first of these was the Reactor Safety Study (Reference 1-12), which was completed in 1975. It calculated the risk to the public from the operation of 100 (then current design) light water reactors in the United States based on a plant specific analysis of two plants on a composite of many different existing sites. The finished document formed a basis for risk methodology discussion, criticism, review, and improvement. Its influence on PRA continues to be felt to this day. More recent PRAs--for example, those performed on Zion, Indian Point, Seabrook, and Midland (References 1-18 through 1-21)--have incorporated i methodology improvements, some of which emanate from RSS critiques such as that provided by the Lewis panel (Reference 1-22). These improvements i include more complete analysis of dependent failures and human I interactions, uncertainty quantification methods, methods for assembling and dissecting the results, containment and core response analysis, modeling of external events (earthquakes, fires, floods, etc.), and incorporation of the site-specific topography, emergency preparedness plans, and changing weather patterns in the site model. In this report, a PRA incorporating all these features is termed a " full-scope, Level 3 PRA," in accordance with the ANS/IEEE PRA Procedures Guide (Reference 1-1). 1-4 0097H052085
Two impacts of the above methodological advances are worth noting. One is a more accurate specification of the contributors to risk. The methodology allows us to identify the contributors to risk and observe in increasing detail what is driving the risk level. This is vital for decision making on design or procedural options and other risk management actions by the utility. Knowing what the risk is and the fine structure of that risk enables its control and effective management. A second impact of recent methodological advances is to enhance the usefulness of PRA in risk management and in the regulatory process. The latter includes conformance with regulatory safety goals (Reference 1-23), post-TMI accident licensing requirements (Reference 1-24), environmental impact reports, and emergency preparedness plans (Reference 1-25). Despite the limitations of this STP Preliminary Scoping Study, both impacts of methodological advances are clear in the results presented in Section 2. Note further that the Scoping Study is a much more useful description of the risk from STP than any existing full-scope PRA. The risk profiles from other PRAs cannot be used for STPEGS. Recent experience indicates that risk profiles are even more plant-specific than realized following the early PRAs. A striking example is the difference in risk levels and dominant contributors between Indian Point Units 2 and 3, which are similar units located on the same site (Reference 1-19). Indeed, the results presented in Section 2 for STPEGS indicate several plant-specific factors not seen in prior work. It now becomes clear why a plant-specific PRA is performed. In particular, the ultimate reason for doing a risk assessment is that there are underlying decisions to be made. The risk assessment provides vital input to the decision-making process. A complete decision analysis requires not only an assessment of risk but also an assessment of costs and benefits. Only the risk assessment input to the decision-making process is addressed in this report. Importantly, assessments of risk, cost, and benefit should be done for each available decision option. If the decision in question is whether to modify a plant or its procedures for operation and maintenance, PRA can be most helpful in the following way: After the final risk curves have been assembled, the methodology permits a clear examination of risk contributors from several different perspectives. The structure of the risk model allows us to determine risk contributors in successive levels of detail. With this detail, we are in a position to identify options ttat can reduce risk in a cost-effective manner and, conversely, recognize that some proposed changes can have no effect on risk. Thus, the quav itative presentation of risk, before and after any proposed change, a11w3 us to decide whether the change is effective or warranted. It also allows us to provide a perspective by comparison with other sources of risk and with various proposed " safety goals" or " acceptable risk criteria." The PRA procedures also allow us to evaluate plant changes that take place, for example, as the plant ages. The idea is to be aware of any new contributors that might be significant in the future. Risk reduction may result from changes in specific plant components, personnel training, procedures, safeguards, containment, or emergency 1-5 0097H052085
I plans. The plant-specific and site-specific risk model being developed in this project is designed to accommodate any and all such changes in the decision analysis. 1.2.1 RISK MANAGEMENT PERSPECTIVE: EXAMPLES FROM OTHER STUDIES Although the preoccupation of PRA thus far has been with risk quantification and, hence, full-scope PRA projects, there have been numerous applications of PRA to resolve particular issues and to implement risk management (Reference 1-26). These applications afford the opportunity to identify with concrete examples the important interface between PRA and the decision-making process. They also 1 demonstrate how the performance of a limited-scope or full-scope PRA enhances the understanding of the safety significance of design features and identifies design weaknesses. One important application was the investigation of several proposed design modifications in a PWR plant. These modifications included a refractory core ladle; a filtered, vented containment; and the addition of hydrogen recombiners, all of which had been selected for consideration prior to the performance of the PRA. In the course of the PRA, it was readily identified that a fourth option, a diesel-driven containment spray pump modified to be independent of AC power, would not only cost considerably less, but would effect a greater reduction in an already very low risk level than the three costly alternatives that had been proposed prior to the PRA. More importantly, the results supported the decision option to leave the plant the way it was. A second example of a risk management action enhanced by a PRA pertains to the issue of backfitting a third auxiliary feedwater system pump in a PWR plant to meet one of its post-TMI requirements. The detailed analysis of dependent failures involving support systems in this PRA determined that the number and type of pumps in the original (two-pump) design, which included one motor and one turbine-driven pump, was not the key to this system's contribution to risk. The key was that both pumps were dependent on an electrically powered chilled water system. The third pump was installed in the turbine building so that it would be independent of the chilled water system. The merit of this aspect of the design had not been appreciated prior to the performance of an integrated plant-level PRA. i There are several other examples of this type in which use of PRA models led to enhanced risk management. Many of these provided a basis for identifying a more cost-effective solution than otherwise would have been obtained. The most important lesson from these applications is that PRA not only provides a means of evaluating the risk significance of different decision alternatives, it also helps in defining what decision options should be considered. The above management and technical lessons from earlier PRAs have been taken into account in planning and conducting this scoping effort for STPEGS. Because the risk analysts have conducted or participated in most of the recent industry-sponsored PRA projects, the methodology used has i 1-6 0097H052085
benefited from the most recent advances to the state of the art. This methodology is fully described in Chapter 4 of the Seabrook Station Probabilistic Safety Assessment (Reference 1-20) and the mathematical bases are documented in Reference 1-27.
~
1.2.2 PERSPECTIVE ON THE PRELIMINARY SCOPING STUDY The basic idea of the Preliminary Scoping Study is to gain a substantial portion of the benefit of a full-scope PRA for a small fraction of the cost. One aspect of the full-scope PRA approach that has prevented a rapid movement toward application to all plants has been the considerable costs associated with their performance. A full-scope PRA such as that completed on Seabrook can require as much as 20 to 25 man-years of analysis, documentation, and review, including the necessary utility support. A second concern has been that when the PRA is conducted in one pass from beginning to end, there is little benefit to be derived for the plant owner until near the end of the project when the first results appear. In response to the above concerns, a phased approach to PRA was developed and successfully demonstrated on the Seabrook Station PRA. Since then, we have adopted and refined several variations on this basic idea on PRAs now in progress on Beznau (Swiss plant), Three Mile Island Unit 1, and Salem Unit 1. This Scoping Study is similar to the first-phase, limited-scope analysis of those phased projects. Having observed that the simple " delta on WASH-1400" approach was unsuccessful (the dominant contributors from this type of study never matched the dominant contributors from full-scope PRAs on the same plants), PLG developed a phased approach in which experienced analysts learn all they can about the plant during the scoping phase. Working at the client's plant and engineering offices, they study all available documentation; talk with engineers, operators, and maintenance personnel; and inspect the plant. l Within the first 2 months, they develop detailed intersystem dependency l models, a detailed support system model, and a detailed event tree for l general transient initiating events suitable for use in the final risk assessment. Only coarse system models are prepared, but based on our extensive exparience, they include all important components. Quantification relies on suitable data and system analyses from other studies and includes an appropriate allowance for common cause failures. Four of these preliminary scoping studies have been performed to date, and the results have been most encouraging. For the only one that has completed the full-scope effort, the dominant contributors identified in the scoping phase were very much in line with those eventually found from the full assessment, even though the uncertainty in the quantitative results is much reduced in the final assessment. A full-scope risk assessment is a large, complex project and the questions it seeks to answer are broad and open-ended. Futhermore, we cannot know a priori what will be important at any specific plant. Therefore, it is difficult to carefully control and focus project effort and costs. The preliminary study helps to control costs by focusing attention on the areas where detailed analysis will be most beneficial in 1-7 0097H052085
br' , h- l 1 realistically reflecting the plant risk. Systems or initiating events that will never be significant contributors to risk can be identified during the preliminary scoping assessment so that relatively little effort can be spent on them during the full assessment. Especially important, the preliminary assessment can reveal troublesome phenomena j that must be evaluated in more detail before the full assessment event 1 trees can be completed. For instance, in the Seabrook PRA, the preliminary scoping phase found that great uncertainty existed about the amount of reactor coolant pump seal leakage that could occur following a total loss of component cooling water. As a result, this event was treated conservatively and appeared as an important contributor to the frequency of severe core damage. After this was revealed, the Westinghouse Owners Group was able to initiate pump seal leakage tests. If this had not been found until later, when full quantification results became available, it would have been difficult to initiate the same tests without serious disruption of project budget and schedule. More importantly, an appropriate emphasis on scenarios involving pump seal LOCAs in the SSPSA risk model was made possible. In our past experience with the limited-scope analyses, two different approaches were used to develop uncertainty distributions in the scoping phase. Method 1 consisted of simply fitting a lognormal distribution to upper and lower bound estimates of core melt frequency. The upper bound was taken to be the point estimate from the preliminary risk model with no credit for operator recovery actions, such as offsite power recovery or manual scram following an ATWS. The lower bound was a subjective estimate by the study team of the lower bound core melt frequency attainable by a modern light water reactor. In Method 2, uncertainties were propagated through the preliminary risk model with an allowance for those risk contributors (such as external events) that were left out or treated in a cursory manner. As Figure 1-2 illustrates for the Seabrook PRA, the full assessment results were represented by a narrower distribution than the results of Method 1 and were situated near the center of the uncertainty distribution obtained by Method 1. This is logical, since these results represent an enhanced state of knowledge and therefore the distribution should be narrower. The fact that the distribution is near the center of the Method 1 result indicates that the degree of conservatism embodied in the estimate of the upper bound in Method 1 was balanced by the degree of optimism in the estimate of the lower bound. In contrast, Method 2 appears to have understated the effects of uncertainty on the low side and is high relative to the final results. This can be explained by conservatisms in the preliminary model that were not factored into the quantification, the chief such conservatism being the treatment of the pump seal LOCA. In the preliminary phase, an unmitigated pump seal LOCA was assumed to lead to core uncovery in 30 minutes compared with nearly 4 hours in the full assessment. This led to an underestimate of the effects of operator recovery in the preliminary study. Therefore, the Method 1 approach to quantifying uncertainty seemed to more reasonably represent the state of knowledge that existed at the completion of the scoping phase for Seabrook. 1-8 0097H052085
I i In the current STP Scoping Study, we have followed the more rigorous propagation of uncertainties used in Seabrook's Method 2, and have allowed for the possibility of excessively conservative success criteria in the best estimate case by assigning probabilities to 14 possible definitions of these criteria--some less stringent and some more. We believe this approach superior to those used earlier; i.e., it should more accurately represent our true state of knowledge or uncertainty. With regard to the qualitative insights and list of cominant risk contributors developed in the preliminary phase for Seabrook, a comparison with the final results reveals many similarities and a few differences. The similarities include the appearance of station blackout scenarios in both lists of dominant accident sequences, a high conditional frequency of delayed overpressurization failure of the containment given core melt in both cases, and the prominence of sequences involving failure of the primary component cooling water system. The preliminary results also provided an early indication of the importance of assumptions about the behavior of the reactor coolant pump seals after a loss of seal injection and cooling. The most significant difference between the two sets of results was the failure of the preliminary phase to identify the risk significance of the interfacing systems LOCA. The importance of the interfacing systems LOCA with respect to early fatality risk in the scoping phase of Seabrook was masked by a simplified site (consequence) model that overestimated the consequences of delayed overpressurization relative to those resulting from containment bypass. Since the frequency of delayed overpressurization of the containment was estimated to be much greater than the frequency of the interfacing systems LOCA, the risk of the latter scenario was masked by that of the fonner. In summary, the preliminary scoping analysis provided a good perspective on core melt frequency, a first cut at the dominant risk contributors, and an improved allecation of resources for the full assessment. On the other hand, a first-cut, preliminary risk assessment cannot be regarded as a substitute for a full-scope PRA, or even for a full-scale plant model. A better perspective on the relationship between the limited and full-coverage analyses can be seen by comparing the accident sequences id5ntified in the two phases of the Seabrook PRA. The top six accident sequences identified in the preliminary phase of the Seabrook PRA are listed in Table 1-1. These sequences resulted from a plant model of a comparable level of detail to that in the final phase, but covering only a small set of initiating events: general transients, loss of offsite power, loss of service water, loss of primary component cooling water, large LOCA, and interfacing systems LOCA. In contrast, 58 initiating events were analyzed in the final phase, including the subdivision of the general transient category into 14 separate transient events. The preliminary initiators covered both internal causes and allowances for some external events. The list of sequences from the preliminary phase indicated a high importance for station blackout, transients both with and without scram, and failures of the component cooling system. This stemmed from the treatment of pump seal LOCA as discussed above. 1-9 0097H052085
I The top 10 sequences with respect to core melt frequency in the final phase of the Seabrook PRA are listed in Table 1-2. As shown, the top l sequence from the preliminary phase was confirmed; the same station ; blackout scenario also ranked first in the final phase. Different ! variations of station blackout involving service water system failures as the cause of diesel generator failure also appeared high on the list in the final results. Similar success was achieved in forecasting the importance of the PCC system and ATWS events to the final risk results for Seabrook. In summary, we are very encouraged about the capability to anticipate the most important results of a full probabilistic risk assessment in a relatively short preliminary scoping analysis. Any following analysis can thus focus on the sequences that matter most. Figure 1-3 graphically depicts the relationship among options for limited-scope and full-scope studies. The current report deals with the Scoping Study requantification option. It is suitable for supporting risk management decisions on specific issues, but should not be confused with more complete and realistic full-scope studies. Readers are warned against using point estimate results of this study independent of their broad uncertainty bounds. Comparison of point estimate Scoping Study results with point estimates from completed full-scope PRAs is especially inappropriate. Such comparisons will yield unreliable, almost surely incorrect and misleading conclusions. The Preliminary Scoping Study is incomplete, uses conservative approximations when they do not numerically impact core melt frequency results, and makes significant assumptions about the future content of plant procedures. Its results have great uncertainty and must be used only with great care and judgment. Nonetheless, when the limitations are understood, these results should provide an important basis for risk management at STPEGS. 1.3 STP/ PRELIMINARY SCOPING STUDY PROJECT PERSPECTIVE To effect project control, a Preliminary Scoping Study Project Plan (Reference 1-28) was developed and published in the first month of the study. This plan defined the project tasks, schedule, flow of documents, and detailed allocation of manpower resources needed to complete the Scoping Study. The project plan provided the basis for measuring progress, which is documented in monthly progress reports. 1 The scope of work was organized into the following technical and administrative tasks to effect project management: Number Task 1 Plant Familiarization 2 Initiating Event Identification 3 Systems Analysis 4 Support System Model Development 5 Frontline System Model Development 6 Survey of External Events and Spatial Interactions i 1-10 0097H052085
.?; I
, . I
. Number Task (continued) 7 Data Analysis 8 Risk Quantification 9 Technology Transfer 10 Definition of Phase II Approach 11 Interim Report Preparation
'12 Project Management i 13 Technical Review 14 Quality Assurance 15~ Administration and Support 16 Authorized Meetings The above tasks are highly interrelated and most were performed in an integrated fashion by the project team. During the first 2 weeks, the plant familiarization task (Task 1) began at PLG offices in California with the transfer and review of a large volume of plant documentation.
Then, over a 3-week period, the nucleus of the team, under the direction of the principal investigator, was moved to the Houston area. During this period, the team members met with HL&P engineering, operations, and training personnel at the HL&P offices at 5400 Westheimer, Southpoint,
; and at the STP site while performing technical work in Tasks 1 through 5. This intensive effort accelerated plant familiarization and enabled access to information not readily attainable from the plant documentation. This visibility of the team in the Houston area helped stimulate thinking among the HL&P organizations about the mutual benefits to be realized from interactions with the effort. The remaining 11 weeks of the Scoping Study were conducted at PLG offices in California in completion of the tasks leading to the development of the nucleus of a plant risk model and preliminary quantitative results.
The technical quality of the Preliminary Scoping Study and subsequent-phases of this effort has been and will continue to be held to the highest standards. Six principal approaches were followed to ensure high , technical quality: e The assignment of highly competent, experienced personnel to the team. , e The use of state-of-the-art methods, subject to limitation in scope. e The documentation of models, input data, computer programs, and other facets of the analysis. e The involvement of owner / operator engineers and operators to ensure that the models accurately described the plant and its operating environment. , e The conduct of independent technical reviews. e The use of comprehensive QA procedures to document that technical standards have been achieved. ] o 1-11 0097H052085
A PLG technical review board chaired by Dr. Stan Kaplan performed independent technical reviews of all project deliverables and received presentations on the results from the project team. These reviews, plus those of the analysts, task leaders, project manager, and HL&P, were documented according to the QA procedures in Reference 1-29. The QA procedures, which include those elements of 10CFR50, Appendix B, judged by PLG to be relevant and useful for PRAs, include procedures for computer program verification and documentation, document control, procurement, QA audits, and technical reviews. The QA program was administered by the PLG QA manager. The QA procedures are believed to be generally more strict than those normally followed in PRAs. For example, a computer program (SETS) used extensively in Nuclear Regulatory Comission-sponsored PRAs, such as the Interim Reliability Evaluation Program had not been independently verified until PLG QA procedures were applied to it in PRA projects carried out by PLG. 1.4 REPORT GUIDE The results of the Preliminary Scoping Study are presented in Section 2. These results include numerical estimates of the frequencies of severe core damage accidents and radioactivity release states with different potential for offsite consequences. Also, the major contributors to risk and to core melt frequency are presented with a statement of the important issues for resolution and other qualitative conclusions. The matrix formalism used to assemble and disassemble the results and other key aspects of PRA methodology are briefly summarized in Section 3. This methodology presentation briefly reviews full-scope, Level 3 PRA metheds, and spells out in additio,1a1 detail the limitations of the Scoping Study risk model. The remaining sections in the report document the application of the Scoping Study methodology to STPEGS. The plant event sequence model, or plant model, is discussed in Section 4. This section includes the selection and quantification of the Scoping Study initiating events and a modularized event sequence model with separate modules to cover the responses of the auxiliary systems and the frontline systems. The systems analysis task is described in Section 5. It includes the l qualitative analysis of all STPEGS systems, categorization of systems for risk model inclusion and exclusion, the quantification of the l unavailabilities of subsystems (generally, the redundant trains of systems), and operator actions that correspond with the event tree top ! events. Section 5 includes one example of the 87 systems analysis , summaries that were prepared during the Scoping Study and provided to l HL&P as separate deliverables. Additionally, it tabulates significant qualitative results at the systems level. f The preliminary survey of external events that was conducted as part of the Scoping Study is documented in Section 6. This section is organized into separate subsections for seismic events, in-plant hazards and spatial interactions, and other external events. This survey includes qualitative analyses of events such as seismic events, internal fires and l j 1-12 0097H052085
I l floods, and external flooding due to hurricanes, which were assessed to have some chance of impacting the results. Section 6 also includes ) bounding analyses of events that were found to have negligible risk ' contributions, including events such as aircraft crash, turbine missile, tornado missile, and hazardous chemical releases. Allowances for all , external events and spatial interactions were made in the quantification I of uncertainty in the results presented in Section 2. The scoping analysis of the containment response is documented in Section 7. It describes the containment design and defines the plant damage states tracked in the plant model in Section 4. These states set the needed input parameters for the containment analysis. Two C matrices relating plant damage states to release categories are developed, one for each of two possible building configurations. Section 7 concludes with a qualitative analysis of offsite impact that was achieved simply by grouping the analyzed accident sequences according to the potential for affecting risk to the public.
1.5 REFERENCES
1-1. American Nuclear Society and Institute of Electrical and Electronic Engineers, "PRA Procedures Guide; A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants," sponsored by the U.S. Nuclear Regulatory Commission and the Electric Power Research Institute, NUREG/CR-2300, April 1983. 1-2. Houston Lighting and Power Com any, " Policies and Practicies: Content of Consultant Reports,p' Policy No. 502, Rev. O, July 1, 1983. 1-3. Farmer, F. R., "The Growth of Reactor Safety Criteria in the United Kingdom," Anglo-Spanish Nuclear Power Symposium, Madrid, Spain, November 1964. 1-4. Garrick, B. J., and W. C. Gekler, " Reliability Analysis of Engineered Safeguards," Nuclear Safety, Vol. 8, No. 5, September-October 1967. 1-5. Garrick, B. J., " Principles of Unified Systems Safety Analysis," l Nuclear Engineering and Design, Vol .13, No. 2, pp. 245-321,1970. l 1-6. Mulvihill, R. J., et al., "Probabilistic Methodology for the i Safety Analysis of Nuclear Power Plants," USAEC Report San-570-2, February 1966. l 1-7. Stewart, R. M., "The Application of Modern Safety and Reliability i Methods to the Design and Operation of Protective Systems for l Large Potentially Hazardous Chemical Plants," 1974 Engineering Foundation Conference on Process Design, Operation, and Control for Safety and Reliability, Henniker, New Hampshire, July 1974. l 1-8. Warren, S. V., " Safety Assessment of Systems for Landing Aeroplanes in Bad Visibility," United Kingdom Civil Aviation Authority. 1-13 0097H052085
1-9. Garrick, B. J., and S. Kaplan, " Cost-Benefit Estimate of Transporting Spent Nuclear Fuel by Special Trains," Transactions of American Nuclear Society, Probabilistic Analysis of Nuclear Reactor Safety, Newport Beach, California, May 8-10, 1978. 1-10. " Experience of Reliability Work Within Aviation Technology," CDL-Report, Draft, 1978. 1-11. "Canvey: An Investigation of Potential Hazards from Operations in the Canvey Island /Thurrock Area," U.K. Health and Safety Executive, May 1978. 1-12. U.S. Nuclear Regulatory Commission, " Reactor Safety Study: An Assessment of Accident Risks in U.S. Nuclear Power Plants," WASH-1400 (NUREG-75/014), October 1975. 1-13. The Federal Ministry of Research and Technology, " German Risk Study--Summary," Federal Republic of Germany, August 15, 1979, English translation in EPRI-NP-1804:SR,1980. 1-14. Pickard Lowe and Garrick, Inc., "0PSA, Oyster Creek Probabilistic Safety Analysis, Plant Analysis Update," prepared for GPU Nuclear Corporation, PLG-0253, December 1982. 1-15. Fleming, K. N., et al., "HTGR Accident Initiation and Progression Analysis Status Report, Phase II Risk Assessment," DOE Report GA-A15000, General Atomic Company, April 1978. 1-16. " Clinch River Breeder Reactor Plant Risk Assessment Report," Report No. CRDRP-1, March 1977. 1-17. SAI, " Crystal River-3 Safety Study," prepared for the U.S. Nuclear Regulatory Commission, December 1981. 1-18. Pickard, Lowe and Garrick, Inc., Westinghouse Electric Corporation, and Fauske & Associates, Inc. " Zion Probabilistic Safety Study," prepared for Commonwealth Edison Company, September 1981. 1-19. Pickard, Lowe and Garrick, Inc., Westinghouse Electric Corporation, and Fauske & Associates, Inc., " Indian Point Probabilistic Safety Study," prepared for the Power Authority of the State of New York and Consolidated Edison Company of New York, Inc., March 1982. 1-20. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probablistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0300, December 1983. 1-21. Pickard, Lowe and Garrick, Inc., " Midland Nuclear Plant Probabilistic Risk Assessment," prepared for Consumers Power Company, May 1984. 1-14 0097H052085
1-22. Lewis, H. W., et al., " Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Comission," NUREG/CR-0400, September 1978. 1-23. U.S. Nuclear Regulatory Comission, " Safety Goals for Nuclear Power Plants," NUREG-0880, May 1982. 1-24. U.S. Nuclear Regulatory Comission, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," NUREG-0718, March 1981. 1-25. U.S. Nuclear Regulatory Comission, " Criteria for Preparation and Evaluation and Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654, Revision 1, November 1980. 1-26. Garrick, B. J., " Lessons Learned from First Generation Nuclear Plant Probabilistic Risk Assessments," Workshcp on Low-Probability /High-Consequence Risk Analysis, Arlington, Virginia, June 15-17, 1982. 1-27. Kaplan, S., G. Apostolakis, B. J. Garrick, D. C. Bley, and K. Woodard, " Methodology for Probabilistic Risk Assessment of Nuclear Power Plants," PLG-0209, June 1981. 1-28. K. N. Fleming, " Baseline Study Project Plan," STPEGS Plant Risk Model Development Program for Houston Lighting and Power Company, PLG-0361, April 1984. 1-29. Pickard, Lowe and Garrick, Inc, " Quality Assurance Manual," PLG-0223, July 1933. 1-15 0097H052085
TABLE 1-1. DOMINANT' ACCIDENT SEQUENCES FOR SEABROOK PHASE I RESULTS Ranking Ranking by by Core Early S'9"'"C' Risk Contribution
#"* " '7 Independent Dependent Frequency Melt Initiating Response Plant Release
* "'" " e n e Categon * '"
h End S te Fa s F u s* ac Early Recovery year)** Fatalities FmlMs 1 Loss of A23 Fail 33 2RW Onsite Power SSPS, ESFAS, 3.0 x 10-3 1 1 Offsite Power PCC, EAH, CS, (3.0 x 10-4) SI, RH, CBS, RCP Seal LOCA 2 Loss of A4 M 4D 2RW None CS, SI, EAH, 9.9 x 10-5 2 2 Component RH, and CBS (9.9 x 10-5) Cooling (long-term) 3 General A0 M 4A CG Reactor Trip None 3.3 x 10-4 No Early Low Transient (3.3 x 10-5) Fatalities 7 4 General A23 Fail 3D 2RW Onsite Power, SSPS, ESFAS, 3.0 x 10-4 3 3 6-* Transient Offsite Power PCC, EAH, CS, (3.0 x 10-5) SI, RH, CBS, RCP Seal LOCA 5 General A12 W 4A 8B ESFAS Signal CS, SI, RH, 2.6 x 10-4 No Early Low Transient (both trains) EFW (2.6 x 10-5) Fatalities 6 General A4 M 4D 2RW Component CS, SI, EAH, 1.4 x 10-4 4 4 Transient Cooling Water RH, and CBS (1.4 x 10-5) (both trains) (long-term) TOTAL 4.1 x 10-3 (5.0 x 10-4) Total All 2.7 x 10-4 Remaining (1.2 x 10-4) Sequences
- Indicated are specific systems at Seabrook Station.
** Numbers shown without parentheses are frequency without recovery; numbers in parentheses are with recovery.
TABLE 1-2. 'SL' -
- OF ACCIDENT SEQUENCES WITH SIGNIFICANT RISK AND CORE MELT FREQUENCY CONTRIBUTIC. -
.0M PHASE II 0F THE SEABROOK PRA WITH APPLICABLE PHASE I SEQUENCES Sequence Ranking In ting Additional Sy Failures /
- Resulting Dependent Failures reque y Latent My s C ,re (per reactor year) g,yg Nam Nam W ace Risk Risk Loss of Offsite Onsite AC Power. No Recovery of AC Fower Component cooling, hfgh pressure makeup 3.3-5 1 1
- 1 Power Before Core Damage (ECCS). reactor coolant pop seal LOCA.
cont 2f rswent flitration and heat removal. Loss of Offstte Service Water, lh Recovery of Offsf*e Onstte EC pawer, component cooling, high 9.2-6 2 2 *. 1 Power Power and few pressure makeup (ECCS). reactor coolant pop seal LOCA. containment flitretion and heat removal. , Small LOCA Residual Heat Removal None. 8.9-6 3 * * ** Control Roon None Componant cooling, high and low pressure 8.7 6 4 3
- 2 Fire makeup (ECO5l. reactor coolant pop seal '
LOf_A. containment filtration and heat j removal. 3 j Loss of Mafn Solid State Protection System Reactor trip, emergency feedwater, high 8.3-6 5 4
- 3 i Feedwater and low pressure makeup (ECCS),
t w contairment flitration and heat removal. I N Steam Line N Operator Failure to Estab11sh Long-Tera 5.6-6 6 *
- t Brest inside Heat Removal Containment Heat Removal Reactor trip Component Cooling Nf gh and low pressure makeup (ECCS), 4.6-6 7 5
- 6 reactor coolant pop seal LOCA.
containment filtration and heat removal. Loss of Offsite Train A Onsite Power, Train B Service Train 8 onsite power, component cooling. 4.4-6 8 6
- 1 Power 1 dater, No Recovery of AC Power Before high and low pressure makeup (ECCS),
Core Damage reactor coolant pump seal LOCA. i containment filtration and heat removal. Loss of Offstte Train 8 Onsite Power Trafn A Se mice Train A onstte power, component coolf ng. 4.4-6 9 7
- 1 Power Water, 40 Recovery of AC Power f.efore high and low pressure makeup (ECCS),
Core Dan *ge reactor coolant pump seal LOCA, containment ifitration and heat removal. PCC Area Ff re Mone Companent coolic high and low pressure 4.1-6 10 8
- 2 maken (ECC'). reach: ; Vaa pA., sea.
LO3. containment flitration, and heat removal .
- Negligible contributton to risk.
**!dentiffed but ranked low in Phase I.
TNot included in Phase I initiators. NOTE: Esponentf al notation is indicated in abbrevlated form; f.e., 3.3 5
- 3.3 x 10-5, A
i 0078H122684 1 1
e Protective systems for chemical plants - Imperial Chemical Industries, Ltd. (Reference 1-6). e Automatic landing systems for aircraft - United Kingdom Civil Aviation Authority (Reference 1-7). e Transport of radioactive materials (Reference 1-8). e Commercial aircraft accident experience (Reference 1-9). e Operation and growth of the Canvey Island, United Kingdom Industrial Complex (a major nonnuclear facility involving the production and storage of chemical, flammable, toxic, and hazardous materials) - Canvey, United Kingdom Health and Safety Executive (Reference 1-10). e Operation of 100 nuclear reactor power plants in the United States, the most ambitious of these projects, examining two plants in great detail - Reactor Safety Study, WASH-1400 (Reference 1-11). e Gennan Risk Study - the collective risk from the operation of the German nuclear power plants is analyzed (Reference 1-12). e OPSA, Oyster Creek Probabilistic Safety Analysis - a comprehensive risk analysis of the Oyster Creek Nuclear Power Plant (Reference 1-13). e Accident initiation and progression analysis - a probabi:istic risk assessment of the high temperature gas-cooled reactor (Reference 1-14). e Clinch River Breeder Reactor Plant Risk Assessment Report - a risk assessment of the Clinch River Breeder Reactor (Reference 1-15). e Crystal River-3 Safety Study - a safety study to determine the expected frequency of selected accident sequences associated with the operation of the Crystal River 3 power plant (Reference 1-16). o Zion Probabilistic Safety Study - a comprehensive risk analysis of the Zion nuclear power plant (Reference 1-17). e Indian Point Probabilistic Safety Study - a comprehensive risk analysis of the Indian Point nuclear power plants, Units 2 and 3 (Reference 1-18). e Seabrook Station Probabilistic Safety Assessment - a comprehensive risk analysis of Seabrook Station Units 1 and 2 (Reference 1-19). e Midland Nuclear Plant PRA (Reference 1-20). FIGURE 1-1. EXAMPLES OF PRA APPLICATIONS 0077H122684
i LOWER
- BOUND UPPER
- BOUND MEDIAN MEAN
__ I e
- s l METHOD 1 1 1
, PRELIMINARY SCOPING STUDY l METHOD 7 % J i
FULL ASSESSMENT i I i I I i 10'8 10-5 10'4 10'3 10-2 CORE MELT FREQUENCY (EVENTS PER REACTOR YEAR) )(
- LOWER AND UPPER BOUNDS CORRESPOND, RESPECTIVELY, 1 TO STH AND 95TH PERCENTILES OF THE UNCERTAINTY DISTRIBUTIONS.
FIGURE I-2. COMPARISON OF CORE MELT FREQUENCY RESULTS UBTAINED IN THE PRELIMINARY AND FULL ASSESSMENT PHASES OF THE SSPSA
i LIMITED SCOPE OPTIONS SCOPING STUDY OUALITATIVE > QUALITATIVE RISK MANAGEMENT ANALYSIS PERSPECTIVES V-SCOPING STUDY OUALITATIVE > PRELIMINARY, QUANTITATIVE RISK ANALYSIS MANAGEMENT PERSPECTIVES V
^"A PLAN > DEFINITION OF RISK MANAGEMENT PROCESS, ALLOCATION OF PRA RESOURCES SCOPING STUDY m REQUANTIFICATION r RISK MANAGEMENT DECISIONS ON SPECIFIC ISSUES FULL SCOPE 3
OPTIONS V FULL-SCOPE PLANT ANALYSIS > RISK MANAGEMENT DECISIONS BASED (LEVEL 1 PRA) ON CORE DAMAGE FREQUENCY, OVALITATIVE RISK PERSPECTIVES V FULL SCOPE CONTAINMENT l, ANALYSIS > RISK MANAGEMENT DECISIONS BASED ON RELEASE STATE FREQUENCIES AND (LEVEL 2 PR A) CORE MELT FREQUENCY l V FULL-SCOPE SITE ANALYSIS > RISK MANAGEMENT DECISIONS BASED (LEVEL 3 PR A) ON FULLY OUANTITATIVE RISK INFORMATION, NUMERICAL COMPLIANCE WITH SAF ETY GOALS CONFlRMED FIGURE 1-3. SEVEN OPTIONS TO DEFINING THE SCOPE OF A RISK ANALYSIS 1-20
- 2. RESULTS The two reactor units at STPEGS are essentially identical and have certain shared structures. The degree of similarity is sufficient to permit the results to be applied to either reactor unit insofar as single-unit operations are concerned. Therefore, the results presented in this section apply to each unit viewed separately. The units for expressing accident frequency are " events per reactor year."
The results are presented in two parts to provide answers to the following two questions:
- 1. What is the likelihood of core melt?
- 2. What is the likelihood of offsite impact?
These questions provide a structure for the analytical work of this study and a framework for organizing the numerical results. Preliminary answers to these questions are provided by this Preliminary Scoping S tudy. The first answer is based on a detailed study of the plant and simplified plant models. The second answer is based on a survey of the STPEGS containment and extrapolations from existing containment and site analyses based on expert judgment. To answer the important questions asked by the Scoping Study, we begin by identifying those events disturbing steady-state operation of the plant. The selection process for these " initiating events" is described in Section 4.2. A list of the initiating events analyzed in this study and their mean frequencies is given in Table 2-1. In our computations, the information in Table 2-1 is organized into a vector called the 4 vector. The use of matrix algebra to assemble and analyze the risk contributors is explained in Section 3 but is not a prerequisite to understanding the results in this section. It is important to distinguish among several broad groups of initiating events that have significantly different impacts. First are plant transients and loss of coolant accident (events 1 and 3 in Table 2-1). Some of these occur frequently, but the standby safety systems have been cesigned for such events and have a good chance to work. The rest of the initiators are really " common cause" events, which is to say that in addition to initiating a sequence of events, they degrade the performance of systems useful in stabilizing the plant. The interfacing systems LOCA (event 2) is very unlikely to occur. Nevertheless, because it degrades safety injection and recirculation cooling, it leads directly to core melt. Furthermore, it bypasses containment, increasing the potential for offsite impacts. Thus, it is worth including and has been found to be significant for other plants. The support system faults (events 4 through 7) degrade and fail key safety systems. Although their frequencies are much lower than simple transients, their occurrence can have significant impact. Current design , practice that emphasizes train separation can exacerbate their effects. 2-1 0099H052085
I Finally, the external events and internal plant hazards (events 8 through 16) can have modest to very far reaching common cause effects. They are mostly very rare events. Current design practice is aimed at reducing the impacts of these events and PRA results have generally shown that this practice has been quite successful. The accident sequences following the initiating events, as well as the models tracking them to core melt, and plant damage states are described in Sections 4 and 7. Section 7 also discusses phenomenological paths through the containment model to release categories and their relationship to offsite impacts. These relationships are summarized in i Table 2-2, where we see that scenarios bypassing or failing the i containment and having no containment spray to scrub fission products have significant effects. Additionally, if a large breach of containment occurs with no containment spray before or just after core melt, significant early impact can result. 2.1 WHAT IS THE LIKELIHOOD 0F CORE MELT? a I Since the principal inventory of radioactive material at STPEGS is held within the zircaloy clad uranium dioxide fuel assemblies in the reactor l core, there can be no significant release of this material from the 3 reactor core unless there is core damage. A small proportion of this inventory is contained in the gaps between the fuel and cladding. The remainder, a much larger proportion, is trapped in the ceramic matrix of ' the fuel, which would have to be raised toward its melting temperature (about 5,000*F) before a major release could occur. In order for the integrity of the containment to be challenged as a consequence of a
- severe damage event, it is necessary for core damage to proceed to the point of full-scale melting of the fuel and penetration of the molten core debris through the bottom of the reactor vessel. Thus, our first concern is with the likelihood of severe core damage or damage of a significant number of these assemblies. A second concern is with the likelihood of full-scale core melt. !
3 The level of resolution of the plant damage states used to categorize the j accident sequences in the STP Scoping Study event trees does not permit a 1 distinction between severe core damage and full-scale core melting that j proceeds to penetration of the reactor vessel. In other words, any i accident sequence that involves severe core damage to the point of onset of significant fuel melting is assumed in this study to result in core melt and vessel penetration. In earlier studies, the possibility of specifying additional plant states to distinguish between core melting and core damage short of melting was considered. The idea was rejected, however , upon finding that the time interval between onset of core damage and full scale fuel melting is short compared with the time interval between the initiating event and the time of core damage for risk significant scenarios. Therefore, there was a physical basis for the a assumption that, given the onset of core damage, the conditional likelihood of core melt approaches unity. Thus, the terms " core melting" and " severe core damage" are used interchangeably in the oiscussion of the results presented below and in the remainder of the report. j Depending on the scenario analyzed, the molten core debris may be cooled 2-2 j UO99H052085
r within the reactor cavity and does not necessarily melt through the containment basemat as assumed in some prior PRA studies, such as the Reactor Safety Study. Because of uncertainties, it is not possible to estimate a meaningful point value of the frequency of accident scenarios. Hence, we express our results for core melt frequency in terms of probability distributions. These distributions specify a range of possible core melt frequency values and probability weights for each value within this range. Figure 2-1 presents the probability distribution for the frequency of core melt estimated the South Texas Project Preliminary Scoping Study. As can ge seen from the figure, the median core melt frequency is about 3 x 10- events per reactor year. In other words, there is roughly a 50% chance that further study would find the core mel South Texas plant as designed to be less than 3 x 10 g, and frequency a 50% for the chance that the actual frequency will be found to be grgater. The mean frequency of core melt is slightly greater than 1 x 10-4 per reactor
.vear, or about once in 700 reactor years of operation.
To get some feeling,for the source of the core melt frequency curve, Figure 2-2 shows the contributions from internal events and the
. combination of external events and the interfacing systems LOCA (dominated by external event scenario frequencies). It is clear that the external event-generated sequences have only a small impact on the total core melt frequency. Section 2.3.2 explains the assumptions used in the various point value quantifications of internal events using the STP model and identifies uncertainties associated with those assumptions. It explains how the uncertainties in assumptions as well as other uncertainties were quantified and propagated through the results to develop the STP Scoping Study probability of frequency curves.
Section 2.4 provides a detailed list of the most significant scenarios contributing to one example assumption set result. Returning to the probability of core melt frequency, Figure 2-1, the highest values are a result of the small chance that the most pessimistic
- assu'nptions that were censidered in this study apply combined with the l possibility that our most pessimistic data for operator recovery and l hardware failure also apply. Conversely, the lowest values result from l - the small chance that our most optimistic assumptions and data apply.
l Each possible combination of assumptions and data is weighted by the i probability that it represents the "true state of the plant system," a l system including hardware, procedures, and trained personnel that have ! not yet been built, written, or trained. Substantial uncertainty is j inherent in the Scoping Study results. The current value of the mean frequency of core melt reflects the large degree of uncertainty associated with the Scoping Study rather than a strong degree of evidence that a core melt is actually so likely to occur. Note that the distribution is quite broad. According to the - results of our analysis, the core mel frequency has a 90% chance of l beingintheintervalbetween3x10gand5x10-3 4 2-3 i 0099H052085
?
a i 2.2 WHAT IS THE LIKELIHOOO OF 0FFSITE IMPACT? ^ A wide spectrum of different radioactive material releases can be ' postulated. Each such release could have a different magnitude and composition of radionuclides, time-dependent release rate, thermal energy of release, and other factors that influence the calculation of accident impact. Release categories have been defined that group scenarios by these parameters. For a given release, the impact to the public is dependent on many external factors such as the directions of the wind at the time of the hypothesized accident, the atmospheric stability, weather conditions, the time of day, day of the week, season of the year, speed and effectiveness of evacuation, and many other factors. As stated earlier, we have grouped all these effects into three impact categories as shown in Table 2-2. Figure 2-3 decomposes the probability distribution for the frequency of core melt into the contributions to each impact category. Category II, with potential for latent impacts, and Category III, with little public risk, provide similar contributions to core melt frequency. They arise from related but significantly different scenario groups. Both come primarily from RCP seal LOCA events. In the case of Category III, electric power remains available, making it possible to operate containment spray and scrub fission products from the containment a tmosphere. Category II scenarios lead to the failure of onsite AC power; thus, sprays are unavailable following melt and the frequency of these scenarios are somewhat lower. The Category I scenario group with potential for early and latent impacts have much lower frequency and occur primarily when the containment supplementary purge valves remain open with no containment spray. 1 Details of the scenarios are given in Section 2.4. Consider the following as a greatly simplified summary of the basic scenario groups: o Category III. A loss of cooling water (ECW or CCW) event leads to the loss of charging pumps and RCP thermal barrier cooling. An RCP seal LOCA results, followed by successful safety injection. When the shift to the recirculation mode occurs on low RWST level, no cooling is available. Core melt and eventual containment failure occur because decay heat is not removed. Containment sprays operate. Small containment bypass paths may exist. e Category II. The loss of all AC power leads to an RCP LOCA with no safety injection, no Containment spray, and no fan Coolers. The scenario group is initiated either by a long term failure of EAB HVAC (both normal and smoke purge modes) or an unrecovered loss of offsite power followed by failure to deliver onsite diesel generator power. Core melt and eventual containment overpressure are guaranteed. Small containment bypass paths may exist. Scenarios with large bypass paths through the supplemental purge valves with containnent sprays operable make a small contribution to this category. e Category I. Scenarios similar to Category II in which the supplementary purge valves fail to close either because of valve problems or a failure of ESF actuation signal combinea with the 2-4 0099H052085
._ . -_ - _ - - . . _ - - - ~
failure of the operators to generate a signal. Note that no power is s l available to close the MOVs inside containment. The interfacing l systems LOCA makes a small contribution to this category. 2.3 IDENTIFICATION AND QUANTIFICATION OF UNCERTAINTY Assessing the risk from extremely rare events such as the potential accident sequences considered in a PRA is subject to significant uncertainties. In view of this, the authors have adopted an approach in which the expression of uncertainty is a fundamental consideration in presenting the results. As explained more fully in Section 3, this consideration is embodied in the definition of risk itself. The risk associated with potential accidents is defined by a list of accident sequences, an assessment of the likelihood and impact of each sequence, and a statement of the uncertainty. What is particularly 'special aDout the approach adopted by the authors is that every attempt is made to express' this statement of uncertainty quantitatively. While the quantification of uncertainty is an important element in any PRA results, it is especially important in this limited-coverage Scoping Study. The Preliminary Scoping Study includes many sources of uncertainty. Some are identical to those found in complete Level 1 studies on operating plants such as uncertainty in basic human error rates. Many are identical to those found in complete Level 1 studies on other plants in the construction stage with no operating history, such as plant-to-plant variability in equipment failure rates, uncertainty in common cause failure data for two-train systems, and uncertainty in the exact plant layout needed for fire analysis. Others are common to limited-scope studies such as incompleteness in coverage of initiating events, incompleteness in supporting analyses (e.g., equipment fragilities to temperature and seismic excitation and room heatup analyses), and simplified systems analyses. Finally some are peculiar to the STPEGS design and site such as uncertainty in common cause data for three-train and four-train systems and the capability of the RHR pumps to operate following a LOCA, or a bleed and feed operation that creates a hot, moist environment inside containment. l The general approach to quantification of uncertainty in the Preliminary Scoping Study results consisted of the following steps: e Step 1. Point estimate quantification of STPEGS risk model using a
" reasonable" set of assumptions for internally initiated sequences only.
e Step 2. Determination of the principal contributors to the point estimate result.
- e Step 3. Identification of key sources uncertainty in the results of l 5teps 1 and 2 to include uncertaintics in data, models, success criteria, plant behavior under accident conditions and sequences and contributors missing from the model.
2-5 0099H052085
- - . --. . - .- - =_ ]
i e Step 4. Construction of a simplified risk model for propagation of sources of uncertainty. e Step 5. Quantification of the effects of uncertainty in terms of probability distributions of the frequencies of accident sequences. In applying Step 3 to the STPEGS risk model, the following major categories of uncertainties were identified. o Data Uncertainties. These include uncertainties in the data used to quantify the risk model, especially that associated with the common cause events and operator actions in the dominant accident sequences _ of Table 2-3.
. e Uncertainties in Assumptions. The most important uncertainties in e assumptions were found to be the success criteria for HVAC systems, the thermal transient response of rooms resulting from degraded HVAC
- equipment, the effects of elevated temperatures on critical components, and the performance of RCP seals under loss of CCW and AC electric power conditions.
e Completeness. The approach to quantify these uncertainties was to make an allowance for missing sequences based on the results of completed, full-scope PRAs on plants similar- to STPEGS. L The key uncertainties in the study and the methods used to account for i them in the.quantification process are described in the following sections. 2.3.1 UATA UNCERTAINTIES 4 The Scoping Study used the generic PLG data base for PWRs developed in . the PRAs on Seabrook (Reference 2-1), Midland (Reference 2-2), and other plants was used without detailed review for applicability to STPEGS. It is expected that, if a Level 1 PRA were completed for STPEGS, the data base for component failure rates, maintenance frequencies and durations, ! and human errors would not change very.much in relation to the data used in the Scoping Study results. The most important type of data that could change considerably is the common cause event data. A common cause event is a failure of two or more redundant components due i to a cause other than the failure of another component. Examples of such Cduses are design, manufacturing, and Construction errors, erroneous procedures, human errors in following procedures, and environmental stresses. Common cause events are modeled primarily through the use of common cause parameters that express proportions of the component failure rates attributable to varying degrees of common cause failure. In a three-train or four-train system, such as the AFW system, several common 2 cause parameters must be quantified for each redundant component and each failure mode applicable to the irodel. In a completed Level 1 PRA, these 3 parameters are quantified by screening each event in a dependent events data base for a given component for assessment of applicability and
- degree of impact for each system in the risk model. In this fashion, design and operation-specific common cause parameters are quantified.
4 2-6 0099H052085
l s In this Scoping Study, common cause parameters were assumed, based on prior studies. Unfortunately, these parameters are known to exhibit large variability between designs and the results of systems quantification are quite sensitive to variations in these parameters. Generally. they are more sensitive to changes in these parameters than to variations in failure rates. In reviewing the quantifications of the. dominant sequences, common cause parameters for the following components either dominated or made significant contributions to the accident sequence frequencies. s' e Essential Cooling Water Pumps and Motor-0perated Valves e Component Cooling Water Pumps and Motor-0perated Valves e Essential Chilled Water Pumps and Chillers e EAB HVAC Fans and Dampers e Class 1E Diesel Generators e AFWS Pumps and Motor-0perated Stop ': heck Valves In addition to the above, there are manycother components whose quantification was sensitive to assur.otions regarding common cause parameters. A large portion of theiuncertainty now appearing in the results in Figure 2-2 for internally initiated accident sequences is associated with the relatively large uncertainties in these common cause parameters. A second category of uncertainty particularly important in the Scoping Study was the quantification of human actions. This is reflected in the facts that no in-depth human actions analysis was performed to quantify the operator actions and that detailed procedures to carry out such actions are not yet available for STPEGS. While the Westinghouse Emergency Response Guidelines were useful to help characterize several operator actions, no procedures were available for effecting others such as " smoke purge" HVAC operation. On balance, human action made the ' second most important contribution to the quantification of internally initiated sequences, insofar as the data-related uncertainties are concerned. - 2.3.2 UNCERTAINTIES IN ASSUMPTIONS Uncertainties in the study's assumptions about the RCP seal LOCA and the impact of degraded HVAC capability have been quantified using the basic PRA approach set forth in Section 3. We simply list the possibilities for the true state of the world (the scenarios, si), the probability that each is the true state (pt), and the consequence associated with that state (xj). Then, the set of triplets i<sj, p1, xj>J is the complete representation of our state of knowledge and is the result of the PRA. Here, each scenario, sj, is a discrete set of . assumptions; the probability, pj, represents the collective judgment of tN stuoy team that the set of assumptions is the true set; and the consequence, xt, is the set of core melt sequence frequencies that would occur if the assumption set is the true set. 2-7 0099H052085
l The judgments expressed in the probabilities reflects discussions with HLdP engineers, discussions with Bechtel and Westinghouse in the presence of HL&P engineers, and a review of limited Bechtel calculations of EAB heatup as well as the previous analytical a*:d operational experience of the team. Note also that under each set of assumptions, a different list of dominant sequences is possible. For the RCP seal LOCA, two discrete cases were considered. Based on previous review of the Westinghouse Owners' Group seal LOCA studies, current conversations with Westinghouse, and a consideration of modeling capabilities in the Scoping Study, the following two assumptions were quantified: Assumption Probability S1 - RCP seal LOCA occurs in 2 hours. 0.22 S2 - RCP seal LOCA occurs in 16 hours. 0.78 For the case when 16 hours are available, additional recovery modes are possible: for example, operators could rig alternative cooling for the RCP thermal barriers (exact method would depend on the availability of power and the cause of the original loss of cooling) or they could adapt the existing Emergency Response Guideline for loss of all AC power to this situation and reduce RCS pressure and temperature to prevent the seal LOCA. Recovery analysis under assumption S2 evaluated such actions. For the HVAC success criteria, many possibilities exist. Fcr example, the EAB heatup calculations for the hottest day in the data base showed that with two trains of smoke purge, a peak temperature of 113*F coincided with the peak outside air temperature. At the end of 24 hours, EAB temperature fell to 109*F (16*F above initial value). No calculations were carried out for the next 24 hour temperature cycle. Calculations also showed that a single ECH train in normal operation would hold average temperature below 117*F for 24 hours. However, temperature was still rising. In both cases, equipment failure might still be possible if the operators are unable to restore normal HVAC, if the outside temperature were higher, if local hotspots exist in the EAB, or if some key equipment fails at lower temperature. Furthermore, outside air brought into the plant by smoke purge might be so humid that moisture-induced failures occur. On the other hand, electric power might be maintained and core melt averted even when the calculations show temperatures exceeding 120*F. For example, combinations of favorable weather conditions, exact HVAC status, and operator actions (with no procedures) such as stripping buses to reduce heat loads, rigging temporary blowers, and restoring or replacing failed equipment may work. Key equipment may not really fail until much hotter. n 2-8 0099H052085
Because of all these wide-ranging uncertainties, a two-step process was used to select and evaluate possible sets of assumptions. First, based on available design information and supplementary calculations, we selected three discrete HVAC success criteria as follows: EAB HEAT LOAD IS BASED ON THE NUMBER 0F AC BUSES ENERGIZED HVAC Success Criteria Assumptions 3 2 1 Hi - Smoke purge effective. Number of fan 2 1 1 trains required (well supported case). H2 - Smoke purge effective. Number of fan 1 1 1 trains required (optimistic case). H3 - Smoke purge ineffective. Normal HVAC required with the following equipment: Fan trains. 2 2 1 Tons of chiller capacity required. 750 600 450 If the normal HVAC criteria of assumption 3 are met (and they will be met if no system failures have occurred), no failures caused by HVAC can occur. The need for smoke purge arises only when the fan and chiller requirements of H3 are not satisfied. Second, even if the " correct" success criteria (to really avoid equipment failure due to ambient temperature rise) based on limiting environmental conditions are known precisely, AC power may not fail. Uncertainties in actual environmental conditions (weather, etc.), heat loads (initially running equipment decay as equipment fails sequentially, etc.), and operator response can affect EAB heatup. To address these possibilities, we considered three discrete assumptions, trying the fragility of AC power to the " correct" success criteria and assessed the likelihood of each: Fragility Assumption Probability F1 - AC power is guaranteed to fail if the HVAC success criteria are not met. (Prcbability of success is 0 if success criteria not met.) 0.170
Fragility Assumption P robability F2 - If the HVAC success criteria are not met, power may fail or succeed depending on the existing heat load, exact state of HVAC, outside air temperature, and effectiveness of ad hoc measures used by operators to control temperature (probability of success is 0.5). 0.710 F3 - AC power will not fail even if all HVAC (including smoke purge) fails. Minimal actions by the operators in response to rising teirperature will occur. (Probability of success is 1.0 if success criteria not met.) 0.120 For our uncertainty analysis, it is necessary to combine the effects of the HVAC success criteria assumptions with the fragility assumptions. To do so, we first assess the probability of H1, H2, or H3, being the
" correct" success criteria, conditional on (given that) F1 (then F2, then F3) is the fragility assumption:
This is the Conditional Probability of Each Under This Fragility HVAC Success Criteria Assumption H1 H2 H3 Smoke Smoke Normal HVAC Purge (2,1,1) Purge (1,1,1) Required F1: AC Fails if 0.64 0.20 0.16 Success Criteria Not Met F2: 50/50 Chance of 0.53 0.37 0.10 AC Failure if Success Criteria Not Met F3: AC Not Failed by 0 1 0 HVAC Failure 2-10 0099H052085
Combining the probability of each fragility assumption with the conditional probability of each success criteria, we obtain the joint probability of each combined HVAC assumption set: The Joint Probability of Each Combined Fragility and Success Criteria Assumption Set
- H1 H2 H3 Smoke Smoke Normal HVAC Purge (2,1,1) Purge (1,1,1) Required
^ F1: AC Fails if .11 .03 .03 Success Criteria Not Met F2: 50/50 Chance of .38 .26 .06 AC Failure if Success Criteria Not Met F3: AC Not Failed 0 .12 0 by HVAC Failure P Finally, the 7 possible HVAC assumption sets are combined with the 2 RCP seal LOCA assumptions to yield the 14 quantified assumption sets: Probability That Mean Frequency of Core This is the "Right" Melt Given These Assumption Set Assumption Set
- Assumptions S2 F3 H2 0.09 1.70 x 10-4 S2 F2 H2 0.20 2.87 x 10-4 S2 F1 H2 0.03 3.97 x 10-4 S2 F2 H1 0.30 5.17 x 10-4 S2 F1 H1 0.09 8.67 x 10-4 S1 F3 H2 0.03 9.73 x 10-4 S1 F2 H2 0.06 1.09 x 10-3 S1 F1 H2 0.01 1.20 x 10-3 4
S1 F2 H1 0.08 1.32 x 10-3 S1 F1 H1 0.02 1.67 x 10-3 S2 F2 H3 0.05 5.65 x 10-3 S1 F2 H3 0.02 6.45 x 10-3 S2 F1 H3 0.02 1.11 x 10-2 S1 F1 H3 0.01 1.19 x 10-2 The distribution of uncertainty in assumptions, which includes a distribution on frequency for each assumption set, has been propagated
*The joint probabilities do not sum to 1.00 because of round-off error.
2-11 0099H052085
i through the analytical results given in Sections 2.1 and 2.2. The different assumption sets generated distinct sets of dominant sequences. . Dominant sequences are discussed in Sections 2.4 and 2.5. ' 2.3.3 COMPLETENESS The final major source of uncertainty that was quantified in the Preliminary Scoping Study results was that associated with the completeness of the risk model. In addressing the issue of completeness, it is important to recognize that only selected internally initiated events were propagated through the detailed event sequence model. The most important initiating events not accounted for in the internal events analysis based on the results of published PRAs are external events and internal plant hazards. Tne approach taken in the Scoping Study to quantify the uncertainty results associated with these events was to estimate, based on a limited state of knowledge, their possible range of contribution. In effect, this required the construction of a simplified risk model for external events and in-plant hazards and a quantification based on the results of completed PRAs on similar plants such as Seabrook. The higher level of uncertainty associated with this abbreviated treatment is reflected in much broader distributions for these Scoping Study results in Figure 2-2. The simplified model for these events consists of two types of accident sequences: (1) events resulting in a nonrecoverable loss of offsite power combined with an independent loss of onsite AC power, and (2) events that lead directly to core melt because of the damage caused by the external event or in-plant hazard. Tne events considered in this analysis are listed in Table 2-1 as events 8 through 16, along with their mean occurrence frequencies. The nonrecoverable loss of offsite power events, Nos.13,15 and 16, were combined with events from the internal events analysis (such as loss of all three Class 1E diesel generators and loss of EAB HVAC equipment in violation of ventilation success criteria) to provide accident sequences which were assigned to impact Category II. Tne remaining external events were assigned directly~ to core melt and impact Category II. The quantification of these events was based on the results of completed PRAs with a subjective stretching of the distributions to account for the limited extent of the analysis. The results presented in Figure 2-2 reflect a large degree of uncertainty for this class of events and help to put the overall results into proper perspective. It is expected that a completed Level 1 PRA for STPEGS would exhibit vastly reduced ranges of uncertainty for this class of events. 2.4 QUANTIFICATION FOR ONE EXAMPLE ASSUMPTION SET The results for core melt frequency in Figure 2-1 were obtained using the matrix approach to risk assembly for each assumption set. The approach is described in Section 3. It first leads to the development of point estimates of the frequencies of a large number of accident sequences. These sequences are categorized with respect to whether core melt occurs, ! and for core melt sequences, they are further grouped on the basis of the l resulting plant damage states, release states, and impact categories. A ' 2-12 0099H052085
sequence of matrix operations is used to assemble and decompose the results to identify risk contributions. A result of this process is the listing of important accident scenarios that contribute to risk. A word of caution is appropriate before proceeding to the example set resul ts. Recall that the example set is only one of several possible cases considered in the full state-of-knowledge uncertainty treatment of Section 2.3. It is, in fact, assumption set S1, F2, H1 and is only assigned about an 8% chance of being the true state of the plant. Therefore, if an attempt is made to compare the example set with point estimates derived from the curves of this section, only partial agreement can be expected. For example, the relative magnitudes of Categories II and III change. (Category III is larger in the example set and smaller when uncertainty is quantified.) The primary reason is that a lower probability is assigned to the example set assumption that an RCP seal LOCA will develop in 2 hours than to an assumption that it takes much longer. With more time and with available equipment and power, the chance for the operators to successfully intercede increases and the probability of developing the LOCA decreases. Therefore, scenarios that proceed to melt via the seal LOCA alone (such as those in the scenario group described for Category III) become relatively less important. The reader must wonder why we chose one of tne less likely assumption sets as the example case. Presenting the most likely set would be more desirable. The reason is historical. One assumption set that appeared most likely early in the project was selected for detailed analysis. The results for the example case are more complete than for other assumption sets and remain our selection for detailed presentation. Unfortunately, toward the end of tne project, with more complete information in hand and the full group of assumption sets defined for final quantification, the picture has changed. Nevertheless, it is still reasonable to present the example case. It is most complete and scenario definitions from the example case served as the basis for calculating changes under changed assumptions. 2.4.1 ASSUMPTIONS The example set calculation is based on a single set of assumptions, l which had been adopted as a reasonable set for developing the basic point estimate results of the study. The key assumptions are the following: l l 1. EAB HVAC success criteria as a function of the Class 1E ESF buses energized (assumption H1): Normal Mode Smoke Purge Mode Buses l Energized Fan Trains Tons of Chiller Fan Trains j Capacity Required 3 2 750 2 I 2 2 600 1 1 1 450 1 l 2-13 0099H052085 i b.
- 2. Even if the HVAC success criteria are not met, there is assessed a 50% chance that no loss of AC power will occur (assumption F2).
- 3. Following loss of seal injection and thermal barrier cooling, the RCP I seals will degrade in about 2 hours, causing a LOCA of several '
hundred gpm per pump (assumption S1). 2.4.2 RECOVERY ANALYSIS In PRA, the usual approach is to include only those operator actions in the progression of scenarios that are required by procedure. In a second pass (" recovery analysis"), additional operator actions are considered on l a scenario-by-scenario basis. There are two important reasons for i adopting this approach. The first addresses the question of dependency. i operator response to an adverse condition is heavily scenario-dependent. l The number of operators available, the other demands competing for their i attention, the time window available for action, the specific cues i identifying the problem and the distractions that create the potential for misperceptions are just some of the unique aspects of each scenario. Failure to account for these dependencies can lead to overly optimistic evaluation of human response. The second reason for this approach addresses the practical realities of analysis. There are limitless , combinations of possible human actions during any scenario. If any ' analysis is to reach closure, the cases considered must be limited to the important classes of actions that can be frequent enough to impact study results. The Scoping Study has followed the usual practice and a recovery analysis has been performed to improve the level of realism of the results. Any j sequence contributing at least 1% of the example set core melt frequency i has been examined and, if reasonable, additional operator response (" recovery") has been incorporated. The kinds of recovery actions j considered include the following: e Starting redundant trains of support equipment such as ECW trains, CCW trains, etc. e Shifting HVAC to the smoke purge mode if sufficient chiller trains are unavailable. e Manually generating a safeguards signal if the automatic system fails. o Manually opening AFWS valves if the motor operators fail, e Restoring offsite power following a LOSP. e demote closing of the motor-operated supplementary purge valves, which are inside containment, if power is available. These recovery analyses have been applied to the dominant scenarios of ; Section 2.4.3 below. 1 2-14 0099H052085
2.4.3 PRINCIPAL CONTRIBUTORS TO THE RESULTS The ultimate benefits of a risk model such as the one developed for STPEGS in this Scoping Study are the quantitative and qualitative bases that it provides for effecting risk management. One such basis is an understanding of the relationship between the results and important characteristics of the plant design, its operational procedures, and the state of knowledge regarding how it behaves under a spectrum of abnormal and accident conditions. The essence of this relationship is conveyed in terms of the principal contributors to the results. The PRA methodology used to develop and quantify the STPEGS risk model has been carefully designed to facilitate tracing of the important contributors as well as to generate the numerical results themselves. A summary of this metnodology is provided in Section 3. The elemental form of the results of a risk assessment is a listing of event sequences or scenarios, an estimate of the frequency and impact of each sequence, and a quantification of the uncertainty associated with these estimates. There are several different ways to express the contributors to the results. One way is to put the accident sequences into groups based on sequence characteristics such as initiating event, plant damage state or impact. One such a breakdown of contributors was provided in Figure 2-2 which showed that internally initiated accident sequences were dominant with respect to core melt frequency. Another such breakdown was shown in Figure 2-3 with the message that, when grouped by impact potential, only a small contribution of the core melt frequency _comes from sequences assigned to the most severe impact category, Category I. Another perspective on contributors is provided by tracing specific accident sequences through the risk model from initiating event to termination in a plant state. By following specific paths through the event sequence model that are responsible for the major portion of the frequency of each group of sequences defined previously, important system failures, human actions, design features and moceling uncertainties can be identified. In Table 2-3, a listing of the most important accident sequences under the assumption set S1, F2, H1 is provided. These sequences are ranked in terms of their contribution to core melt frequency. This table describes individual paths through the STPEGS plant model in three parts: (1) the initiating event, (2) additional plant and operator actions contributing to sequence frequency and plant state, and (3) functionally dependent failures resulting from events in Parts 1 and 2. Because of the limitations in scope and objectives of this Scoping Study, detailed tracing of contributors was only possible for the internally initiated group of accident sequences. In terms of the example set point estimate results, this group comprised about 87% of the total core melt frequency. The 11 sequences in Table 2-3 cover more than 95% of the core melt frequency associated with the internally initiated group, or roughly 82% of the total point estimate core melt frequency. The remaining 18% of the core melt frequency not shown in this table is distributed over a large number of internally initiated sequences, each having a small frequency and externally initiated sequences that were not subdivided into individual sequences in this Scoping Study analysis. 2-15 0099H052085
As evident in Table 2-3, support system failures and operator actions dominate the results at the event sequence level of analysis. Take, for example, the first sequence, which is initiated by a general transient event such as a turbine trip from full power. In normal plant operation, there is one running train of essential cooling water, one train in standby and ready to start automatically if the normally running train fails, and one train in the "0FF" position, which can be brought online via remote manual action. (All three trains of ECW .;.re signalled to start automatically upon generation of a safety injWtion signal.) In this first sequence, the normally running train of EtW fails to continue running following the initiating event and the standby train fails to start on demand. The operator then atteopts to start the "0FF" train of ECW and, when it fails, puts the EAB HVAC system into the " smoke purge" or open loop mode of operation. The latter action successfully prevents damage to EAB and control room envelope equipment and, therefore, AC power availability is maintained. However, all three trains of ECW are lost. The integrity of the reactor coolant pump seals is dependent on the success of at least one train of ECW, because with no ECW, there can be no component cooling water and, therefore, the RCP thermal barrier coolers and the charging pumps that provide RCP seal injection cease functioning. Also, because the ECW through the CCW provides heat removal for the ECCS/RHR heat exchangers and the containment fan coolers, ccre recirculation cooling and containment heat removal functions are lost. The high head safety injection pumps provide ECCS injection, but when ECCS shif ts to recirculation, no cooling is available to the RHR heat exchanger and core melt follows. Note that the dependent failures in Part 3 of the sequence definition occur with a conditional frequency of 1, given the events in Parts 1 and 2, because of functional intersystem dependencies in the plant design. Hence, the frequency of each accident sequence is determined by the events that occur in Parts 1 and 2. The second ranking sequence in Table 2-3 begins with a general transient event such as turbine trip and includes support system failures and important operator actions as with sequence 1. In this case, the support system exhibiting failures is the EAB HVAC system. In normal plant operation before the initiating event, there are two normally running l trains of HVAC fans and the third train is in the "0FF" position so that only operator action or safety injection signal can initiate a start signal. In this second sequence, one of the two normally running HVAC fan trains fails to continue running following the initiating event. Because of the need to properly line up the ECW, essential chilled water and HVAC fan trains, the operator takes action to start the "0FF" trains of ECW, ECH, and EAB HVAC. Subsequently, however, a second train of EAB HVAC fails and that, according to the assumptions for this example set, violates the success criterion for EAB ventilation. As noted earlier for the example set of assumptions, successful EAB ventilation requires two trains of EAB HVAC fans and dampers when three buses are energized. Variations in the results due to uncertainties in these and other important success criteria assumptions are described and quantified in , Sections 2.3 and 2.5. Because of the violation of the EAB ventilation l criterion for sequence 2, AC power is assumed lost and, therefore, so is l all electrically driven equipment, including ECW and CCW. The dependent failures listed in part 3 for this sequence include those for sequence 1, j loss of ECCS injection capability and the inability to close any - initially open containment isolation valve with an AC motor operator. l l 2-16 0099HU52085
i The third sequence in Table 2-3 is similar to sequence 1 except that, instead of losing all three trains of ECW following a general transient initiator, all three trains of CCW are lost. Sequences 4 and 5 correspond with sequences 1 and 3 except that, instead of ECW or CCW loss subsequent to a general transient initiating event, the loss occurs as 1 the initiating event itself. Sequence 6 is a loss of offsite power initiating event followed by failure of all three Class 1E diesel generators and failure to recover offsite power within 2 hours. The example set point estimate assumptions reflected in the results of this table include the assumption that the RCP seal LOCA occurs quickly and is severe enough to result in core uncovery and damage within 2 hours. The possibility of a more slowly developing seal LOCA with core uncovery in 16 hours is considered in the quantification of uncertainty as described in Sections 2.3 and 2.5. The prolonged loss of AC power postulated in this sequence results in the same set of dependent failures as sequence 2. I Accident sequence 7 is the same as sequence 2 except that an additional train of' equipment is postulated to fail the "0FF" train of ECW. As with sequence 2, the EAB ventilation success criteria are violated and the dependent failures include loss of AC electrically driven equipment. Sequence 8 is the same as sequence 7 except that it is the "0FF" train of ECH that fails, rather than the "0FF" train of ECW; however, the resulting dependent failures are the same. Sequence 9 is the first sequence encountered in which key equipment 3. failures are located within frontline safety systems, rather than support
; systems. This is typical of PRA results for modern PWR plants like
! STPEGS. This ' sequence begins with a small LOCA initiating event, 2 successful reactor trip and ECCS injection, and operator action to isolate the RWST and to enable switchover to core recirculation cooling. As a result of valve failures of various types; e.g., a check valve failing to close in a non-isolated RWST injection path, core recirculation cooling fails and core melt occurs. However, all containment systems function properly for this sequence, largely due to the proper functioning of support systems. In sequence 10, as with several higher ranking sequences, equipment failures in HVAC systems are postulated following a general transient initiating event. Unlike the higher ranking sequences, in which favorable operator actions are postulated, in this sequence the operator fails to start the "0FF" trains of HVAC and support equipment and fails to initiate " smoke purge" HVAC operation. The dependent failures for this sequence are like sequence 2 because of the assumed loss of EAB ventilation.
-The final sequence presented in Table 2-3 is only the second sequence presented in which crucial equipment failures occur in the frontline systems and not in the support systems. It is initiated by a general transient initiating event and a postulated failure of all four trains of the auxiliary feedwater system. The operator attempts to manually start the AFW train with the steam turbine-driven pump, but nonrecover pump or 2-17 0099H052085
M MOV failures are encountered. The operator then fails to initiate bleed and feed cooling in time to be effective and core melt ensues. The availability of support systems along this sequence precludes additional dependent failures in part 3 of the sequence description. The impact category assignments for each of the above 11 sequences are listed in the last column of Table 2-3. None of these 11 sequences were assigned to impact Category I, which is associated with the most severe levels of impact, because of the lower frequency associated with these sequences. A Category I impact requires early, gross failure of the containment due to either an interfacing systems LOCA or an initially open and failed-open supplementary purge containment isolation valve and failure of containment spray. The total frequency of Category I sequences for the example set was only 6.6 x 10-b; therefore, none of these appear in Table 2-3. Of those sequences that do appear, the ones that have no AC electric power belong to Category II and the ones with AC electric power to Category III, the most benign impact category. Category II sequences are typified by a delayed containment overpressurization and loss of sprays, while Category III includes containment intact, basemat melt-through and overpressurization with spray cases. . 2.4.4 SENSITIVITY TO SUPPLEMENTARY PURGE OPTIONS An important objective of the Preliminary Scoping Study was to provide information about the sensitivity of the results to a number of options associated with the supplementary purge isolation valves. Six options were investigated as defined in Table 2-4. What was varied among these options were the number _ of fail-closed type valves in each of the two supplementary purge penetrations and the fraction of time all the valves
- in these penetrations are left initially open (i.e., are assumed to be open at the time of the initiating event). Fail closed-type valves have the characteristic of initiating closure upon loss of AC electric power,
; whereas both fail-closed and fail-as-is valves initiate closure upon receipt of a containment isolation signal. Among these options, the
, example set results described in the previous sections correspond with option 2. ' 1 The sensitivity of the results to the different supplementary purge containment isolation valve options required the STPEGS risk model to be 4 requantified separately for each option. Top events in the event trees associated with containment isolation had to be requantified for each case and the modeling of front line system to support system interdependencies had to be mocified as appropriate. The results in terms of point estimates for the frequencies of impact categories and core melt are shown in Table 2-5. As seen in this table, there is no variation in core melt frequency because the design options considered , are independent of core cooling. Also, the differences are very minor for impact Category III and only one significant departure from the example set result was noted in Category II; namely, option 1. As
; expected, the principal sensitivity in the results to these options was observed in impact Category I.
2-18 0099H052085 i
- , - . , . _ . . _ _ . _ . _ _ _ ~ _ - . . , ~ . _ - - _ _ _ . . ._._ _,.. - - _ _ .,.. .__._ _ . -,_-.__._-_,.-..mm
. As noted above, Category I sequences include the interfacing systems LOCA and sequences with failed-open supplementary purge isolation valves and failure of containment sprays. The departure in Category II observed in option 1 is due to the difference in frequency associated with sequences involving failed-open supplementary purge isolation valves with sprays working. Because the assessment of interfacing systems LOCA was not affected by these options, all the variations in impact Category I are due to variations in the frequency of sequences with failed-open supplementary purge isolation valves and failed containment sprays.
Starting from the example set of option 2, the frequency of impact Category I increases by almost 2 orders of magnitude for option 1, which mostly reflects the reliability assumed for one fail-closed type valve. Surprisingly, when moving to option 3, which has a considerably greater - cost, the reduction in Category I frequency is minimal. This is because of the relatively high contribution to both of these cases from sequences with electric power available, in which the failed-closed feature is unimportant. Options 4, 5 and 6 correspond with options 1, 2, and 3 in terms of contributors, but the frequencies are lower to correspond with the smaller fraction of time the valves are initially open. It is very important to note that the impact category assignments in Table 2-3 are applicable to the example set (option 2 only). For example, in the case of option 1, the sequences in Table 2-3 assigned to Category II would have to be reassigned to Category I, since the loss of all AC power would leave the initially opened supplementary purge isolation valves open and the containment sprays failed. 2.5 IMPACT OF ASSUMPTION SETS ON DOMINANT SEQUENCES As we move through the assumption sets defined in Section 2.3.2, it is obvious that the dominant sequences of Section 2.4.3 will change. A straight-forward summary of those changes is provided below. First, consider the RCP seal LOCA. All but 2 of the scenarios (numbers 8 and 11) progress to core damage by a seal LOCA combined with failure of safety injection or recirculation cooling. If the seal LOCA takes longer
- to occur (assumption S2), then in the example case (SI), the frequency of core damage due to these scenarios decreases because the operator has more time to take corrective action as discussed in Section 2.3.2. The i S1/S2 assumptions have no impact on consequence category.
! Second, consider the HVAC Success Criteria Assumptions H1, H2, and H3 [ (see Section 2.3.2). Assumption H1 is the example set. Under ! Assumption H2, some loss of ventilation scenarios leading to core melt in
- - the example set would become successes. Thus, the contributions from the i
loss of all AC power scenario groups (like sequence 2 in Table 2-3) would decrease and the frequencies of core melt and especially impact Category II would decrease. Under assumption H3, all the smoke purge recovery actions of the example set would be ineffective. Frequent
- sequences. now going to success would go to core melt and the frequencies of melt and Categories I and II would increase.
2-19 0099H052085
Finally, consider the fragility assumptions on electric power, F1, F2, amd F3. The example set uses the most likely assumption F2. Under F1 (i.e., AC power fails if the HVAC success criteria are not met), the chance of core melt via HVAC induced failure of AC power would increase. Scenarios similar to number 2 in Table 2-3 become more frequent and begin to dominate the loss of ECW scenarios (number 1 in Table 2-3). Under F3, AC power is essentially independent of HVAC. Scenario number 2 and those like it would disappear, and the frequency of core melt would be substantially reduced.
2.6 REFERENCES
2-1. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-300, December 1983. 2-2. Pickard, Lowe and Garrick, Inc., " Midland Probabilistic Risk Assessment," prepared for the Consumers Power Company, May 1984. 9 l l l l 2-20 0099H052085
_ - _ _ _ _ _ _ ~_ ._ _. .___ _ . -- _ ..._. - - _ -. - _ _ - TABLE 2-1. . INITIATING EVENT FREQUENCIES FOR STPEGS SCOPING ST0DY QUANTIFICATION Frequency Group Initiating Event. Category Code .(events per reactor year) Loss.of Coolant Inventory 1. Small LOCA SLOCA 1.7-2 '
.2. Interfacing Systems LOCA ' ISLOCA 1.6-7 Transients 3. General. Transient- GT 1.1+1 Common Cause Initiating Events Support System Faults 4. Loss of Offsite Power LOSP 9.5-2
- 5. Loss of Essential Chilled Water LECH 7.3-3
- 6. Loss of Component Cooling Water LCCW 3.9-5
'? 7. Loss of Essential Cooling Water LECW 1.3-4 .
.3 External Events 8. Aircraft Crash Leading to Core Damage AC 7.0-8 ;
- 9. Turbine Missile Leading to Core Damage TM 5.7-8
- 10. Tornado Excessive Wind Leading to Core Damage TRW 1.1-8
- 11. Tornado Missile Leading.to Core Damage .TRM 7.4-8
- 12. Hazardous Chemical Release Leading to HCR. 7.6-6 i Control Room Inhabitability
- 13. Seismic-Induced LOSP SLOSP 2.0-5
- 14, Other Events Leading to Core Damage- OCM 1.2 I Internal Plant Hazards 15. Fire-Induced LOSP FLOSP 5.2-4 l 16. Flood-Induced LOSP FLEP 2.6-4 i I
NOTE: Exponential notation is indicated in abbreviated form; i.e., 1.7-2 = 1.7 x 10-2, ) CAUTION: PRELIMINARY RESULTS- } IMPORTANT UNCERTAINTIES ! DESCRISED IN THis SECTION , i t , 0089H122884
TABLE 2-2. RELATIONSHIPS OF SCENARIOS TO 0FFSITE IMPACT CATEGORIES Impact Description Typical Category Scenarios I Significant potential for A large containment leak path early and latent health exists at the time of core melt effects. with containment sprays failed. II Significant potential for A small containment leak path latent health effects and, exists at the time of core melt at most, only small with containment sprays failed numbers of early health (late overpressure containment effects. failure is also possible if there is no containment heat removal); a large containment leak path with sprays operating; or containment intact at core melt but no containment sprays or heat removal--late overpressure containment failure or basemat melt-through with fission products " burped" to atmosphere between basemat and concrete mud mat. III Potential for at most Any other scenario--most likely only small numbers of both containment spray and heat latent health effects. removal are available. CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCE RTAINTIES DESCRl8EO IN THIS SECTION l i i 2-22 0089H122884
TABLE 2-3. ACCIDENT SEQUENCES WITH SIGNIFICANT CONTRIBUTIONS TO CORE MELT FREQUENCY BASED ON POINT ESTIMATE EXAMPLE CASE ASSUMPTIONS AND STPEGS MODEL Sheet 1 of 2 Event Sequence Definition
" #9"'"
Core Melt Part 3 - Functionally Impact Part 2 - Plant Failures and Operator equency Frequency Dependent Failures Category Part 1 - 9"*"#* 9"'"## * " ** ###
* # " " 9 Resulting from Events Initiating Event and Plant Damage State in Parts 1 and 2 1 General Transient Failure of running and standby trains of RCP seal LOCA, core 3.7 x 10-4 III ECW, operator starts "0FF" ECW train recirculation cooling and which fails, and starts "SHOKE PURGE," containment heat removal.
AC power and ECCS injection availability maintained. 2 General Transient Failure of one of two running trains of EAB ventilation, AC power, 2.2 x 10-4 II EAB HVAC fans, operator starts *0FF" RCP seal LOCA, ECCS trains of ECW, ECH, and EAB HVAC fans and, injection, containment heat subsequently, a second train of EAB HVAC removal, motor-operated fans fails, containment isolation valves. Failure of running and standby trains of RCP seal LOCA, core 1.6 x 10-4 III 7 3 Ceneral Transient CCW, operator starts "0FF" trains of recirculation cooling and y CCW and ECW; subsequently, the third containment heat removal. train of CCW fails AC power and ECCS l injection availability maintained. l 4 Loss of All Three Operator starts "SM0KE PURGE," AC power RCP seal LOCA, core 1.3 x 10-4 III ECW Trains and ECCS injection availability recirculation cooling and maintained. containment heat removal. RCP seal LOCA, core 3.9 x 10-5 ggg 5 Loss of All Three No additional failures, AC power and CCW Trains ECCS injection availability maintained. recirculation cooling and containment heat removal. 3.5 x 10-5 gg 6 Loss of Offsite Failure of all three standby Class IE AC power, RCP seal LOCA, ECCS Power diesel generators, operator fails to injection, containment heat recover offsite power (no credit for removal, motor-operated diesel generator recovery), containment isolation valves. 7 General Transient Failure of one of two running trains of EAB ventilation, AC power, 3.3 x 10-5 gg EAS HVAC fans, operator starts "0FF" RCP seal LOCA ECCS trains of ECW, ECH, and EAB HVAC fans. injection, containment heat Subsequently, the "0FF" train of ECW and removal, motor-operated a second train of EAB HVAC fans fail. CAUTION: PRELIMIN ARY RESULTs IMPORT ANT UNCERTAINTIES DESCRIBED IN THis sECTION 0089H022685
TABLE 2-3 (continued) Sheet 2 of 2 Event Sequence Definition
" 'S"*"#'
Part 3 - Functionally Impact Frequency p ., Part 2 - Plant Failures and Operator pg Frequency Initiating Event
"" " " 9 "* "# ' 9"'"## Resulting from Events and Plant Damage State MP 1W2 8 General Transient Failure of one of two running trains of EAB ventilation, AC power, 3.2 x 10-5 gg HVAC fans, operator starts "0FF" trains RCP seal LOCA, ECCS of ECW, ECH, and EAB HVAC fans. injection, containment heat Subsequently, the "0FF" train of ECH containment isolation valves.
and a second train of EAB HVAC fans fail. 9 Small LOCA ECCS switchover to core recirculation Core recirculation cooling. 1.6 x 10-5 ggg cooling attempted but sump recirculation valves fail; all other systems available except that noted in Part 3. 10 General Transient Fallure of one of two running trains of EAB ventilation AC power, 1.4 x 10-5 gg EAB HVAC fans or one running 300-ton RCP seal LOCA. ECCS ro chiller, operator fails to start "0FF* injection, containment heat r'o trains of ECW and ECH, failure to start. removal, motor-operated A "SM0KE PURGE." containment isolation valves, 11 General Transient Failure of all four trains of AFW, None 1.2 x 10-5 ggg operator resets AFW turbine and manually opens motor-operated stop/ check valve in turbine-driven AFW train, turbine-driven AFW pump train still fails to start, operator fails to initiate bleed and feed cooling, all other systems available. CAuff 0N: PRELIMINARY REsULTS-IMPORTANT UNCERTAINTIES DEsCRISED IN THIS sEcTION
l TABLE 2-4. DEFINITION OF SUPPLEMENTAL PURGE ISOLATION VALVE OPTIONS QUANTIFIED USING STPEGS MODEL i Number of Fail Closed Type Case (ho r p r year V lves Valves Per Penetration Initially Open 1 0 1.00 (8,766) 2* 1 1.00 (8,766) 3 2 1.00 (8,766) 4 0 .06 (500) 5 1 .06 (500) 6 2 .06 (500)
*Also corresponds with example. case results.
CAUTION: PRELIMINARY RESULTS-i, IMPORTANT UNCERTAINTIES i DESCRIBED IN THIS SECTION i l 0089H122884
TABLE 2-5. SENSITIVITY OF RESULTS TO SUPPLEMENTARY PURGE ISOLATION VALVE DESIGN OPTIONS Accident Frequency (events / reactor-year)** Design Option
- Impact Category Core I II III Mel t
- 1. Zero FC, Two M0V 4.0-4 2.0-4 8.3-4 1.4-3 100% Open
- 2. One FC, One M0V 6.6-6 5.9-4 8.3-4 1.4-3 100% Open (base case)
- 3. Two FC, Zero M0V 6.3-6 5.9-4 8.3-4 1.4-3 100% Open
- 4. Zero FC, Two MOV 3.3-5 5.7-4 8.4-4 1.4-3 6% Open
- 5. One FC, One MOV 7.0-7 6.0-4 8.4-4 1.4-3 6% Open
- 6. Two FC, Zero M0V 6.7-7 5.9-4 8.4-4 1.4-3 6% Open
*FC = fail-closed type valve l HOV = motor-operated valve (fail-as-is) '
- All results are point estimates. ,
1 CAUTION: PRELIWNARY RESULTS IWORTANT UNCERTAINTIES DESCR18ED IN THIS SECTION I l 2-26 0089H122884
CAUTION. PRELiesetARY CESULTS-IAAPORTANT UNCE%TEINTIES CESCRIBE3 IN THIS SECTION STH PERCENTILE = 3 x 10-5 MEDIAN = 3 a 104 95TH PERCENTILE = 5 m 10-3 g ] I I ~ t l l l A i l i I e l l l E
' l l l l l l 1 1 l
I 1 g ,i MEAN = 1 a 10-3 l I I l l i l I I i I i i l i i i i l I i i l i 1 I I I I I 10'O 10-5 10 4 10-3 10-2 10-I CORE MELT FREQUENCY (EVENTS PER REACTOR YEAR) FIGURE 2-1. PROBABILITY DISTRIBUTION FOR CORE MELT FREQUENCY
CAUTION: PRELIM & NARY RESULTS-IMPOR TANT UNCE RTAINTIES OESCRISED IN THIS SECTION CORE MELT
/ g/
\
s I
\
t g EXTERNALS +V \
, r p INTERNALS O
> f g E I i I t
\
e s l \ 1 I N I b i \
/
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/
l l 1 __ ' I I I ' 16 0 10'8 10-I 10-6 10-b 10-4 10-3 10-2 10'I SCENARIO GROUP FREQUENCIES (EVENTS PER REACTOR YEAR) FIGURE 2-2. PROBABILITY DISTRIBUTIONS FOR SCENARIO GROUP FREQUENCIES
CAUTION. PRELIMINARY RESULTS IMPORTANT UNCERTAINTIES DESCRISED IN THis SECTION LEGEND CATEGORY 4 POTENTaAL FOR EARLY AND LATENT HE ALTH EFFECTS CATE GORY II POTENTIAL FOR EARLY AhD. AT MOST.
$ MALL huMSE RS OF LATENT HEALTH EFFECTS CORE MELT CATECORY ess POTENTBAL FOR AT uoST,$ MALL NUMSERS OF LATENT HE ALTH EFFECTS
/ \
IMPACT / \ CATEGORY 48 / I IMPACT I '
/
CATEGORV at f I t IMPACT I \ 9 CATEGORY I I \ g \
> 8
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I \ 5 I \ t I
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* /
\
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I \ l m /
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s \ N / \
- / \
l \
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i 1YO 10-8 t __--# f f l %* 10'3 104 10-6 gg-4 10 4 10-2 10'I IMPACT CATEGORY FREQUENCIES (EVENTS PE R REACTOR YEAR) FIGURE 2-3. PROBABILITY DISTRIBUTIONS FOR IMPACT CATEGORY FREQUENCIES
- 3. STPEGS RISK MODEL OVERVIEW The purpose of this section is to provide a sufficient description of the STPEGS. risk model to enable an understanding of how the results of the previous section were developed. This presentation's flow is from the general to the specific. Thus, the basic concepts and definitions of PRA are presented first, followed by a description of the architecture of a general plant risk model.- To provide a proper perspective for viewing
-the limited-scope risk model of the Preliminary Scoping Study, the general plant risk model is described for a full-scope, Level 3 completed PRA. With this perspective, the particular capabilities and limitations of the Scoping Study risk model and a limited-scope, Level 1 PRA are described. A more in-depth' discussion of the major elements of the STPEGS risk model can be found in the remaining sections of this report.
3.1 BASIC CONCEPTS AND DEFINITIONS IN PRA s
-There are several concepts that need to be understood in order to successfully transform generic plant operating experience and STPEGS design-specific data into an overall model of plant risk. The following sections highlight the most important and basic of these concepts and serve as-an introduction to the STPEGS risk model. Additional concepts ,
are incorporated, as appropriate, by reference. , 3.1.1 PROBABILITY AND FREQUENCY To describe the reliability performance of equipment or components, a simple model of the failure process is used. This model is commonly referred to as a " random grocess" or " stochastic" model. In this model, an object fails " randomly over time (i.e., with no discernible pattern) and with a constant failure rate, A. The stochastic model can be grouped into two types of models, each of which gives rise to a specific definition of a type of failure rate. If the equipment is idle until called upon to perform, at which time it may or may not succeed to perform its function, a binomial model is assumed. In this model, the likelihood of observing x failures in n demands (or, alternatively, x failures out of n components, all of which are demanded) L(x), is L(x) = xg("n x)! A (1~A)" The failure rate, A, in this case is approximately the number of failures x occurring in n demands, or A*f 3-1 0101H052085
l l l l If the equipment is in constant use (that is, always on demand), the binomial model generalizes to the Poisson model. The likelihood of observing x failures in a time T is expressed by A x y L(x) = g e i Here, the failure rate A is characterized by the number of failures, x, within the given time period, T, or A=f The exponential distribution models the probability of survival over a period T for the stochastic process; i.e., R=e 4T where R is the survival probability. The failure rate A is a parameter of these models; that is, A is a constant characterizing the exponential distribution of failures. Hence, the Poisson model for n components corresponds with an exponential distribution for the time of failure of each component. The following are properties of the exponential distribution that afford a more formal definition of the parameter A and additional insight into this basic model of component reliability. The probability density function, f(t)dt, defines the probability of failure in the time interval [t,t + dt]. For the exponential distribution f(t)dt = A e-A tdt = R(t)Adt Note that the definition of the reliability function, R(t) = e-At, is , the probability of no failures in the time interval [0,t]. This helps explain the formal definition of A as Adt = the conditional probability of failure in the time interval [t, t + dt] given that it survives up to time t. What typifies the exponential distribution is that A is a constant. i The mean time to failure, t, of the exponential distribution is given by ' t= o tf(t)dt = f i 3-2 0082H122784
When it is known that failure has occurred sometime in the interval [0, T], the conditional mean time to failure, tT, is given by tf(t)dt tT=
= ; Tu f o
Another basic measure of equipment performance is its availability; i.e., the probability that the equipment is capable of performing its function at a point in time. This contrasts with reliability, which is the proba-bility that no failures occur over a time interval (referred to as the mission time in reliability engineering parlance). In risk assessment, the average availability, or its complement, unavailability, (averaged over time) is often used, which corresponds to the proportion of time the equipment is "up," as opposed to " failed" or being repaired. When the exponential failure model is used to model failure and the process of restoration is either exponential or of fixed duration, the average equipment availability A is given by A A= =1-t + (MTTR) A+p where MTTR=meantimetorepair=f. p = repair rate. For the Poisson and binomial models, the parameter A approximates the long-term frequency of failure of the equipment. The exact value of the failure frequency is uncertain, and this uncertainty must be quantified. Our approach to quantify the uncertainty is to use a " Bayesian" method to combir.e one's " prior" state of knowledge in the form of probability versus frequency curves with experimental evidence or observations. In the Bayesian methodology, a " prior" probability distribution characterizes the state of knowledge about the possible values of A prior to any actual measurements of the failure rate. In this context, the assertion that the probability is .9 that A is greater than 2 is taken to mean that the person expressing this probability is willing to accept nine-to-one odds that A is indeed greater than 2 [in informal betting terminology, nine-to-one " odds" corresponds to a probability of 1/(9 + 1)]. The probability distribution for the value of A is developed by essentially asking the odds a person is willing to accept at each value of the frequency A. The probability distribution can be updated explicitly using the appropriate model assumption as data about the value of A become available. Thus, data may change the odds that one is willing to accept 3-3 0082H122784
on the values of the failure frequency A. As an example, a person's prior knowledge of the failure rate of a pump, based on the pump salesman's assertion of a failure rate of 2, might look like Figure 3-1. (This figure includes the fact the person believes the failure rate is more likely to be 3 or 4 rather than 2 as the salesman asserted.) If it is found that the same sort of pump has been operating at a neighboring plant with an average failure rate of 2.5 per year over the last 10 years, the updated (or " posterior") probability distribution on the value of A would look like Figure 3-2. The uncertainty in the failure frequency can then be propagated through the underlying stochastic model to give ranges of uncertainty for times to first failures, etc. 3.1.2 HAZARD AND RISK Hazard is defined as a source of danger, loss, or injury. An automobile, for example, is a hazard. Driving an automobile poses a risk due to the chance of crashes, but the risk would be lessened by the use of seat belts, airbags, crash helmets, and other safeguards and the hazard remains unchanged. The risk would also be lessened by reduction or avoidance of car usage or by enhanced driver training. Thus, both preventive and mitigative factors influence the level of risk. Risk can thus be viewed as proportional to the hazard presented and inversely proportional to the effectiveness of the safeguards used and precautions taken. The concept of risk, therefore, involves the concepts of both danger and uncertainty; that is, uncertainty about whether or when any consequences will occur. Thus, to quantify the notion of risk associated with a particular hazard (or " accident sequence" or " scenario"), say the ith scenario si, both the level xq of damage or injury and the frequency 4j with which scenario i i s expected to occur must be determined. For complex systems such as nuclear power plants, the overall risk can be expressed as a set of triplets ranging over all possible scenarios, each triplet being expressed as l
<si, &i, xj>
where sj labels the ith scenario, 4j the frequency of its occur-rence, and xj the damage level associated with the scenario. One very basic definition of risk is simply the set of all such triplets, or symbolically i l risk E R = {<sj, 4j, xj>} all i The need for completeness that is implied in this definition cannot be fulfilled rigorously in any practical application of PRA. Fortunately, however, PRA methods have been used that provide a high degree of confidence that overlooked or missing scenarios do not make significant contributions to risk. 3-4 0082H022085
The development of a risk model that provides these triplets can be simply thought of as the process in which the answers to three fundamental safety questions are determined: e What can go wrong? e What is the likelihood of these different sequences? e What are the damage levels associated with these sequences? To answer these three questions for STPEGS, a structured thinking process has been employed. This process begins with a systematic identification and categorization of the scenarios that might lead to significant damage to the plant. Each scenario is then analyzed to determine its frequency of occurrence and its consequences. Each scenario consists of an initiator, which might be a system failing, a pipe breaking, or a human error, and several manually and automatically actuated actions and passive processes that determine the consequences of the scenario. In principle, each scenario can be divided into four parts:
- 1. The initiating event.
- 2. The subsequent performance of the plant systems.
- 3. The phenomenological events that occur in the core and containment after the initiation of core damage.
- 4. The weather-related and evacuation-related events.
A Level 1 PRA analyzes only the performance of the plar t systems, while a Level 2 PRA also considers core and containment phenonknology and a Level 3 PRA includes a complete consequence analysis. These three levels of scope are represented by the plant model, the containment model, and the " site" (really offsite) model, respectively. This means that each . Level 3 PRA scenario consists of three fragments, one from each of these three models. The events in each fragment are defined as being conditional on a certain set of events having previously occurred. Therefore, after a set of interfaces or " pinch points" between the three models has been agreed upon, each model can be developed separately. The form typically used in presenting Level 3 risk assessment results is the complementary cumulative or frequency of exceedance form. In this form, the frequencies of all scenarios exceeding a particular level of damage are summed. Curves are then plotted for each damage type to show the frequency of exceeding each level of Jamage. Figure 3-3 illustrates a curve of o frequency of exceeding damage level x, (x) versus x for a set of similar scenarios (or defined radioactivity release). The scenario risk is described by the entire curve, which includes all possible damage levels and the frequency of , exceeding each level. As will be explained more fully in Section 3.3, the assessment of damage levels in the Scoping Study is limited to the qualitative assessment of the potential of each analyzed release state to produce various impacts. 3-5 0101H052085
To evaluate the overall plant risk as would be performed in a full-scope, Level 3 PRA, risk curves for the set of accident scenarios are comDined into an overall risk curve, which has the same form as in Figure 3-3. This is done by summing for each damage level xi the corresponding 4j on each scenario curve. The risk curve illustrated in Figure 3-3 incorporates variations l'n the frequency of damage levels from scenarios through the plant, containment, and site models. Uncertainty in the parameters and phenomena incor-porated in the three models is set forth in a family of risk curves indexed by the probability P of a particular curve's occurrence as illustrated in Figure 3-4. This family of curves is analogous to the probability distribution of the failure frequency A, b'st it describes a surface rather than a single curve. The interpretation of the proba-bility associated with each curve represents the analyst's "confiderce" that each curve is the " correct" curve in .1.ight of the underlying uncertainties. Alternatively, the curves may be labe?ed by the cumula-tive probability as !n Figure 3-4. With this interpretation, the curve shown, for example, may be read as "with probability .9, the frequency of exceeding damage level x1 is no greater than 41 " 3.1.3 DECISION ANALYSIS The results of a probabilistic risk assessment are really only an input into a more general decision model, described by Figure 3-5. At each point of a decision, there are two or more options from which' to choose, including the options to not decide or to seek more information. Options i (including not deciding or seeking more information) have associated with them uncertain outcomes of costs, benefits, and risk. These uncertain-ties must be expressed by probability curves such as those shown in Figure 3-5. The outcomes for each option, then, may be regarded as a triplet
<C, B, R>, each element of which is characterized by a probability curve (or surface in the case of risk), and the objective of the decision maker is to choose the mort desirable triplet. The object of PRA is to provide the risk information to the decision analysis, not to actually
- make these decisons.
3.2 RISK MODEL STRUCTURE 3.2.1 QUALITATIVE DESCRIPTION OF STPEGS RISK MODEL A PRA is basically a listing and analysis of scenarios, and a full-scope. PRA can contain literally billions of scenarios depending on how finely the scenarios are described. Accounting for all event tree paths incorporated > lnto a plant risk model requires specific modeling and quantification of all these sceanrios. To provide a logic structure to the qualitative progression of an accident scenario, the overall risk model can be thought of as three linked models: the plant model, the containment model, and the site model, as shown in Figure 3-6. A single j accident scenario progressing to offsite consequences spans all three of . these models. For most accident scenarios, the input to the containment l l 3-6 i 0082H022085
and site models depends only on the state of the plant or containment and } not on the history of the arrival to that state. This fact enables much of the work on the three models to be done independently and in parallel by the various analysts, which is especially useful since different expertise is needed for each of the three models. 3.2.1.1 The Plant Model Billions of possible scenarios must be enumerated in the plant model. To do this requires detailed modeling of the plant, its systems, its components, and their interdependencies. Physical and human interactions with the plant that can affect the frequency of occurrence of an accident scenario must also be included. Event frequencies and their associated uncertainties are quantified using historical evidence in both nuclear and nonnuclear experience when applicable. The plant model contains all the systems reliability aspects, including the engineered safety features of the containment. The containment model (explained below) deals with only the issues of containment response once core damage or melt occurs. A more in-depth presentation of the STPEGS plant model can be found in Section 4. 3.2.1.2 The Core and Containment Model The core and containment model represents the subsequent progression of a scenario once core damage or melt is experienced. The outcome of the scenario is principally determined by the physical processes of the scenario (for example, the pressure and temperature response, the cooling of core debris, etc.) as well as the passive response of the containment structure itself. The containment event tree models the scenario in approximate chronological order and gives special consideration to effects that unique and specific plant features have on the accident simulation. The results of the model are a continuation of the scenario structure, expressed by release categories, quantification of their frequencies, and a source term for estimating accident impacts. Section 7 contains a discussion of the Scoping Study core and containment model. 3.2.1.3 The Site Model The site model represents the progression of scenarios from the release categories, output from the containment model, to the actual offsite impacts. The site model in Level 3 PRA uses Monte Carlo simulation rather than the event trees used in the piant and containment models to describe the extremely large number of meteorological states that could exist at the time of a release and the randomly changing nature of thesa states. The computer code CRACIT uses a plume dispersion model for the release, and site-specific demographic, geographic, and meteorological data as well as evacuation route information in simulating the possible effects. For each particular release, the model traces the movement of radioisotopes, their possible deposition on the gr ound, and their possible effect on the population. 3-7 0101H052085
The output of the site model estimates the damages that could occur as a result of the release. As noted earlier, only a qualitative site analysis was performed in the Scoping Study. The results are briefly summarized in Section 7. 3.2.2 LOGICAL STRUCTURE OF A RISK MODEL The first step in the development of a risk model is to identify
" initiating events" that may, depending on the response of the plant, lead to core damage. These are identified using several independent approaches including a fault tree analysis of the plant energy balance, a
" master logic diagram" (which is another form of a fault tree), failure modes and effects analysis of plant systems, and cross-checks against reactor operating experience, events identified in other PRAs, plant design documents, and the systems analyses.
Once the initiating events are identified, scenarios or accident sequences that could result are identified using a " plant event tree." As explained in Section 4, the plant event tree is actually a network of event tree modules. The top events of each event tree represent the responses of the various plant systems so that each path through the tree represents an event sequence. In this way, the event tree embodies a truth table of all significant success / failure comoinations of the plant systems. At the end of each sequence, the plant either is in a stable, recovered condition or has suffered some core damage. A set of plant states yj is defined, and each path through the tree is assigned to one of these states. This point in the analysis is called a " pinch point." Once a scenario has reached this point, its further development depends only on the plant state yj and not on how that state was reached. Each plant state is carefully defined so that further analysis is the same
' whether that state was reached because of a LOCA or loss of offsite power, etc.
the subsequent events in the Given scenario that arethe plant is in represented bystate yj,tainment event tree." the " con The methodology used is analogous to the plant model. There are, however, subtle differences in the way in which the event tree frequencies and probabilities are quantified, as explained more fully in Section 5. The entry states to the containment event tree are the plant damage states, and the top events of the tree correspond with various containment phenomena, such as hydrogen burning, debris bed cooling, etc. At the end of each sequence, core damage has resulted in a release of varying magnitude. A set of release categories pk, representing types, quantities, timings, and elevations of radioisotopes released, is defined f ano each path through the containment tree is assigned to one of these states. This is another pinch point in the sense that the impacts of release category pk are independent of how the release category was reached. In a full-scope, Level 3 PRA, the impacts of the various radioactive releases are analyzed using a site-specific atmospheric model and the code CRACIT. The results are " conditional frequencies" skg, which express the likelihood of a damage level xx or greater to the public, given that release category pk has occurred. 3-8 0101H052085
The four sets of pinch points defined above enable the convenient modularization of the event sequence model into three event sequence
" fragments." These fragments correspond to portions of the entire event sequence that are modeled in the plant model, core and containment response model, and site model. This provides the basis for the standard form of defining event sequences in the risk model, which is illustrated in Figure 3-7.
3.3 MATRIX FORMULATION OF A RISK MODEL The key idea in the matrix formulation of the risk (Reference 3-1), which is illustrated in Figure 3-8, is that the event trees may be considered equivalent to transition matrices: the trees define the likelihood of moving from various input states to various output states. In the plant event trees, the input states are the initiating events, i, and the output states are the plant states, yi. From the event tree, frequency of being in
- the number mij, representing
, given that initiating the conditionali event has occurred, can be plant state yi[he trees can then be represented by a matrix M composed of calculated.
these mij. Because the plant event tree is very large, it is actually a set of event trees that are connected by several sets of intermediate pinch points (see Section 4 for a discussion of this point). Similarly, the containment event tree can be represented by a C matrix whose entries Cjk represent the conditional frequency of having a release category pk given that the plant was in state yj. In a full-scope, Level 3 PRA, the site matrix S is not derived explicitly from an event tree. Its elements skg are derived from Monte Carlo simulations of the interaction between time dependent meteorological conditions and evacuation, but the interpretation of these matrix elements is again the conditional frequency of having damage level xg given that release category pk occurred. Products of these matrices retain the transition frequency interpretation: e The product matrix (MC) is the transition matrix from initiating events to release categories. The 1,kth element of this product matrix is the conditional frequency that a release of type k occurs given that an initiating event i has occurred. e The product matrix (CS) is the transition matrix from plant states to damage levels. The j,tth element of this product matrix is the conditional frequency that a damage level E occurs given that the jth plant state has occurred. e The product matrix (MCS) is the transition matrix from initiating events to damage levels. The 1,tth element of this product matrix is the conditional frequency that damage level E occurs given that initiating event i has occurred. To determine the unconditional frequengies associated with the various 1 s{ates, an " initiating event vector" 4 must be introduced.
$ is a row vector whose entries 4} denote the frequency of occurrence of each initiating event i.
3-9 0082H022085 1 _ _ _ _ _ _ _ _ _ _ _-__-_-______-___________A
These frequencies can be quantified with the aid of a " thought experiment" in which identical plants are imagined to undergo the various initiating events and the number of occurrences of such events per plant year are counted. When multiplied by the plant matrix M, the result is the " plant state vector" $Y whose elements 4V denote the (unconditional) frequency of occurrence of plant state yj. This, in turn, multiplied by the containment matrix, C, results in $P, the " release vector." When
$P is multiplied by the site matrix, S, the final result is the damage vector 4X where 4{ is the frequency of exceeding damage state xg.
This " assembly process" can be sumarized as follows: 4Y = 4I M (frequencies of the plant damage-states)
$P = $3C = &l MC (frequencies of the release categories) 4X= 4pS = 4YCS = 4 IMCS (exceedance frequencies as a function of damage level, i.e., the points of the risk curve) l This equation is the " master assembly equation," or the "Kaplan equation," for processing the results from a full-scope risk assessment.
Figure 3-9 summarizes the relationship between the event tree model and the matrix formalism, forming the basis of the assembly equation. 3.4 DECOMPOSITION OF RISK AND CAUSE TABLES If the initiating event vector 4I is written as a diagonal matrix I 0 4 1; CD= 2., I 0 *N the 1.jth element of the product matrix 4fM is the frequency of occurrence of the jth plant state resulting fromI the ith initiating event. Comparison of this i,jth element of &nM with the.jth element of 4Y gives the fraction of the total frequ4ncy gf the jth plant state attributable to the ith initiating event. The vector 4' is actually the c91umn sum of t is product matrix. In the same way, the product matrices &fMC and MCS show the contributions that each initiating event makes to a r lease category frequency and damage level frequency, respectively. Contributions of the plant state to release categories and damage levels a d of the release states to damage levels can be determined by defining 7 and C,
&p and di CS, and 4 S.
gonal matrices0 $ and4y,andformingtheproducts The overall logic of the risk ecomposition is shown in Figure 3-10. The steps below matrix operation & M are performed with the aid of a computer program called MAXIMA which traces through the event sequence logic to identify paths between the initiating events and plant damage states that make major risk contributions. The results of an application of the risk decomposition process to the STPEGS risk model was presented in Section 2. 3-10 0082H022085
3.5 OVERVIEW OF STPEGS PRELIMINARY SCOPING STUDY RISK MODEL In the Preliminary Scoping Study, we have not achieved the kind of full-scope PRA conceptually defined above. To understand what the results of this preliminary study really mean, it is essential that the limitations in scope be fully understood. 3.5.1 PLANT MODEL LIMITATIONS The principal parameters of the plant model that are subject to possible limitation in the Scoping Study are (1) completeness in the choice of initiating events, (2) completeness in the definition of accident sequences, (3) completeness and degree of rigor in the quantification of accident frequency, (4) assumptions about the dependence of safety related equipment on HVAC systems and assumptions regarding the development and progression of the RCP seal LOCA. The quantification of initiating events in the Scoping Study was limited to a set of 16, as described in Section 4.2. This set of initiators, while typical for a Phase I-type preliminary risk quantification, is much less extensive than the corresponding list for a completed, full-scope PRA such as that recently completed for Seabrook Station (Reference 3-2). The initiating event set analyzed in the Seabrook Station PRA included 58 events that were subjected to plant model quantification. The 16 for STPEGS were selected by the Scoping Study team as those likely to be important, based on prior PRA experience and a careful examination of the plant design. The second principal parameter, completeness in the definition of accident sequences, appears to be limited in the Scoping Study only to the extent that the choice of initiating events is limited. For the initiating events in the Scoping Study, therefore, the definition of accident sequences is about as complete as would be expected for a full-scope study. One exception to this is that scenarios initiating transients without reactor trip were only partially integrated into the Scoping Study quantification, even though the event tree logic for these sequences was in fact completed. With respect to the degree of rigor in the quantification of accident frequency, the third parameter of the plant model, the Scoping Study was developed from a full-scope treatment of intersystem functional dependencies and simplified plant-specific system models. Departures from the full-scope quantification methodology could give rise to variations in the results. For example, neither a complete treatment of test and maintenance nor a complete propagation of uncertainty distributions for component unavailability parameters was performed in the Scoping Study. In addition, design specific failure rates and common cause failure parameters were based on generic data which was not reviewed nor screened for detailed applicability to STPEGS. Based on experience with the Seabrook PRA and the degree of familiarity with STPEGS achieved in this Scoping Study, these quantification variabilities are not expected to produce many surprises. However, a completed Level 1 PRA would enable the reduction in the range of uncertainty, especially with regard to common cause parameters. 3-11 0101H052085
Finally, the assumptions regarding vulnerability co degradation of HVAC and susceptibility RCP seal LOCA are crucial to the study results. In the Scoping Study, the study team used their juj.gment of the likelihood l of possible true states of those conditions (based on experience, partial I calculations, and discussions with vendors, architect / engineer and utility personnel, and testing organizations) to include the full range of possibilities in the uncertainty calculations. Additional analysis or testing of these phenomena would reduce the range of uncertainty in the results. A discussion of the impact of these model limitations and uncertainties on the results is presented in Section 2. 3.5.2 CONTAINMENT MODEL LIMITATIONS The scope of the containment analysis in the Scoping Study was very modest in relation to a completed, Level 2 or 3 PRA. Only a very small portion of the Scoping Study manpower resources--less than 2%--was devoted to the containment analysis. However, the relatively lower scope in this area was balanced by the fact that no rigorous quantification of accident consequences was performed. Only that limited analysis necessary to establish the relative impacts of accident sequences was performed in the Scoping Study containment analyses. The scope of work in a completed Level 2 containment analysis normally includes, among other tasks, the definition of plant damage states and release categories, the computer simulation of plant and containment transient response to a wide spectrum of accident scenarios, the analysis of the physical processes of core melt sequences, the analysis of containment failure modes, the calculation of the source terms for offsite release, and the construction and quantification of the containment event tree. 1 By contrast, the coverage of this Scoping Study was limited to the definition of plant damage states and release categories, qualitative survey of containment structaral and systems features, and a judgmental quantification of the containment, or C matrix, without explicit quantification of a containment event tree. These limitations, while not permitting a quantification of accident consequences per se, do permit the Scoping Study accident sequences to be put into perspective. A more in-depth discussion of plant damage states and release categories is presented in Section 7. 3.5.3 SITE MODEL LIMITATIONS Of the three major elements of the STPEGS risk model, the site model was the most limited in scope in relation to a completed, full-scope (Level 3) PRA. Rather than probabilistically calculating the l consequences of accident sequences in each release category through a Monte Carlo simulation of the interaction between releases, meteorological, transport, and evaculation scenarios, only a very limited qualitative analysis was performed. Of the release categories that were defined, only a gross grouping by relative impact was performed. Based on experience with previous PRAs, the accident sequences belonging to each release category were placed into one of three " impact categories:" J e Category I. Significant potential for early and latent health effects. 3-12 0101H052185
e Category II. Significant potential for latent health effects and possioly small numbers of early health effects. e Category III. Potential for, at most, only small numbers of latent health effects. This grouping enabled the addition of a qualitative public impact perspective to the results that otherwise would be limited to the quantification of core damage frequency. Section 7 includes the results of the assignment of accident sequences to impact cateogries. The above discussion highlights some of the more important limitations in the STPEGS risk model. Those limitations that are particularly risk sensitive have been defined in the presentation of the Scoping Study results in Section 2.
3.6 REFERENCES
3-1. Kaplan, S., "A Matrix Theory Formalism for Event Tree Analysis-- Application to Nuclear Risk Analysis," Risk Analysis, Vol. 2, No.1. March 1982. 3-2. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0300, December 1983. i i l 3-13 0101H052085 l-
P C a 5 O C s ai 3 O E px(A)
, i , e 0 2 x
1 3 4 5 FAILURE FREQUENCY FIGURE 3-1. PROBABILITY CURVE AGAINST FAILURE FREQUENCY BASED ON SALESMAN'S STATEMENT AND PERCEIVED BIAS OF SALESMAN P h 5 E h a 5 g px(A) E
, i i 0 4 x
1 2 3 5 ! FAILURE FREQUENCY FIGURE 3-2. STATE OF KNOWLEDGE PROBABILITY CURVE AFTER LEARNING NEIGHBOR'S EXPERIENCE 3-14
0 a g ei _________ o I h i E I l I I I I Xg X DAMAGE LEVEL FIGURE 3-3. P0 INT ESTIMATE RISK CURVE WITH NO UNCERTAINTY QUANTIFICATION P
#3 --___ - __ ___ 0 90 l
i I 0.50 1 I l 0.10 l t X X g DAMAGE LEVEL FIGURE 3-4. FAMILY OF RISK CURVES WITH UNCERTAINTY QUANTIFIED 3-15
IDENTIFY OUANTIFY DETERMINE MAKE DECISION -
~
COSTS, BENEFITS, 2 UTILITY OF - OPitMAL - OPTIONS AND RISKS EACH OPTION DECISION CA* P PA (*8 COST. c 84= P PA (b) > U4=UKC,84,RA>Im g BENEFIT, b [ PRA 4 P Al 'I POINT DAMAGE OF DECISION ,
\ OPTION 8
. , , OPTIMAL DECISION =
, MAX (U g,U*---UI 8 N e e e g
- I 04 0,
! 04 C" N P COST 8N. p , UN=Ut<Cy,8N,RN>I BENEFIT PRA v_ FIGURE 3-5. ROLE OF PRA IN DECISION ANALYSIS PROCESS
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E VENT 5E QUENCE 0 EVENTSEQUENCE : . 5siE CON 5f OUE NCE S MODE L MODE L MODEL
- m sysygug HOUAN ENTERNAL CON T AINME NT ACCtDENT $OURCE MODE LS 9- INTE R ACitON -> Events -> F Alt uRE TERM ECONOMIC TOPOCR APHY* ""V910G'C^l MODE L S 4 SIM UL AllON .e -> iMrACT e- DE por.n Arnv, DU50 MODELS MOUEL MODE L QUAN- ,
MODEL ME T E OHOLOG v Af 70P45E a TIFICATION l l uopggs a u ., DAIA PE ANT GJ 8ASE UNIOut SITE SPECIFIC I FEAIURES fEATunt5AND >-a E V ACUA140N PLANS N ( PL Ara T uGOE L j ( CONTAINMENT MOOEL j ( $1TE MOOEL j FIGURE 3-6. BLOCK DIAGRAM STRUCTURE OF A FULL-SCOPE LEVEL 3 RISK MODEL
- 1 SPECIFICATION OF THE INITIATING EVENT.
2 RESPONSE AND STATUS OF SUPPORT SYSTEMS (SUCCESS / FAILURE COMBINATIONS). PLANT - MODEL 3 EARLY RESPONSE AND STATUS OF FRONTLINE SYSTEMS (SUCCESS / FAILURE COMBINATIONS). 4 LONG TERM RESPONSE OF FRONTLINE SYSTEMS (SUCCESS / FAILURE COMBINATIONS). ( 5'(- SPECIFICATION OF PLANT DAMAGE STATE. PHENOMENOLOGICAL RESPONSE OF (DEGRADED) i CO A NMENT ^ l
RESPONSE
7 RESPONSE OF CONTAINMENT STRUCTURE. SPECIFICATION OF RELEASE CATEGORY; b8 i.e., SEVERITY OF RADIOACTIVE RELEASES. l 9 METEOROLOGICAL DISPERSION SEQUENCE. SITE MODEL 10 EVACUATION / EMERGENCY ACTION RESPONSE. U SPECIFICATION OF PUBLIC HEALTH AND PROPERTY DAMAGE LEVELS. EVENT SEQUENCE PINCH POINTS FIGURE 3-7. STANDARD FORM 0F ACCIDENT SEQUENCES IN A FULL-SCOPE, LEVEL 3 RISK MODEL 3-18
FREOUENCY vEcToas: <' Tr ,; . i . ._i ,v-r4.,;.._ #a - r ,;. ,;. i + = - r+;..;.._i (INlTIATING (PLANT STATE VECTOR) (RELEASE VECTOR) [ DAMAGE VECTOR) EVENT VECTOn) $ w / %, / 2:::,. l TR ANSITION " I l " 12 * * * 'll 812 * *
- s y g s 12 * *
- MATRICES: u. " 28 .
21 .
'28
. . C. . .
- g. .
(PLANT M ATRlxl (CONTAINMENT MATRlxl (SITE MATRIX) Y G; ASSEMDt.Y PROCESS: dY = ,IM 4# = 9 YC = p'MC
# Y
+ * ~ & S ~ 6 CS ~ 4'MCS FIGURE 3-8. MATRIX FORMULATION AND RISK ASSEMBLY PROCESS
1 P PIN C H IE; Yj Pk Xg POINTS:
- FINAL INITI ATING PLANT' RELEASE DAMAGE l , EVENT STATE CATEGORY STATE r , r , r ,
PLANT g ,_ EVENT CONTAINMENT TREE 9 EVENT ll SEGMENT TREES CRACIT MOD E LS: - CALCULATIONS PLANT MODEL CONTAINMENT MObEL SITE MODEL P 1P 1r 9r 3r W FREQUENCY VECTORS: ,I . [p , pI I
,,,j (V.lp Y,4Y ,,,,] 98 . [,8,9 ,,,j ,m . [, u, , ,,,g
* (INITIATING (PLANT STATE VECTOR) (RELEASE VECTOR) (DAMAGE VECTORI EVENT VECTORI q, q, ,
m,m12 g ***
*11 *12 ***
*11 s 12 ***
TR ANSITION <
- 21 * *21 e '21
- MATRICES: * + . . . .
- e . . .
(PLANT MATRIX) ICONTAINMENT MATRIX) (SITE MATRIX) i ASSEMBLY PROCESS: I pY = eM
=
,8 . (YC 0'MC 9" = S* =
(YCS = 'MCS FIGURE 3-9. OVERALL VIEW 0F THE PRA ASSEMBLY PROCESS SHOWING RELATIONSHIPS OF PINCH POINTS, EVENT TREES, FREQUENCY VECTORS, AND TRANSITION MATRICES
i l DECOMPOSITION STEP INFORMATION OBTAINED SELECT DAMAGE INDEX TO DECOMPOSE (e.g., EARLY FATALITIES) 1r r , PERFORM MATRIX OPERATION RELATIVE CONTRIBUTIONS OF
> INITIATING EVENTS TO DAMAGE
@D MCS LEVEL FREQUENCY t J ir r ,
PERFORM MATRIX OPERATION RELATIVE CONTRIBUTIONS OF 0 RELEASE CATEGORIES TO S DAMAGE LEVEL FREQUENCY - D t i 1r r , PERFORM MATRIX OPERATION RELATIVE CONTRIBUTIONS
? OF INITIATING EVENTS TO 4DMC RELEASE CATEGORIES c >
1r r m PERFORM MATRIX OPERATION R ELATIVE CONTRIBUTIONS
> OF INITIATING EVENTS TO M PLANT DAMAGE STATES D
L ; 1P EXAMINE PATHS THROUGH PLANT PARTICULAR SCENARIOS AND EVENT TREES BETWEEN PARTICULAR m SYSTEM FAILURES THAT INITIATING EVENTS AND PLANT CONTR:BUTE TO IMPORTANT STATES ELEMENTS OF M MATRIX w J 1r RELATIVE CONTRIBUTIONS OF EXAMINE CAUSE TABLES FOR HARDWAR E, COMMON CAUSE, EACH IMPORTANT SYSTEM ' MAINTENANCE, AND HUMAN ALLURE ERROR TO SYSTEM FAILURE , ir F 7 EXAMINE UNDERLYING DATA BASE, EVIDENCE, AND MODEL m INSIGHTS TO ENHANCE USED TO G ENERATE SYSTEM RISK MANAGEMENT FAILURE CONTRIBUTIONS FIGURE 3-10. PROGRESSIVE STEPS IN RISK DECOMPOSITION 3-21
l
- 4. EVENT SEQUENCE MODEL 4.1 OVERVIEW - INITIATING EVENTS, AUXILIARY AND FRONTLINE SYSTEM MODELS
'The plant-level event sequence model is divided into four separate parts, as illustrated in Figure 4-1. The first part of the plant event sequence analysis enumerates the initiating events; i.e., those events upsetting equilibrium conditions such that standby systems must operate to maintain core cooling. The selected initiating events are discussed in Section 4.2.
The second part of the plant event sequence analysis evaluates the response of the various auxiliary (or support) systems to the specific initiating event category in question. Auxiliary systems include electric power supplies, various cooling systems, control and protection systems, and the reactor trip function itself. They are modeled separately from the so-called "frontline" systems, since they are required to support several frontline systems and, therefore, represent a potential source of dependent failures. A single failure in an auxiliary system can give rise to multiple failures in frontline systems because of functional dependencies. The auxiliary system model is an event tree that is used to determine the conditional frequencies of various unique auxiliary system states, given the specific initiating event in question. Each state has an associated frequency of occurrence and an impact vector describing the downstream effects the auxiliary system state has on the frontline systems. The auxiliary system model is described in Section 4.3. The third part of the plant event sequence analysis evaluates the early
-response of the frontline equipment and the plant operators to the specific initiating event in question. "Early" refers to the timing relative to the initiating event. The sequences included in the early response models terminate in one of three possible conditions:
o Successful long-term decay heat removal by either the steam generators or a residual heat removal loop. O Successful core cooling and inventory control by injection of the contents of the refueling water storage tank into the reactor coolant system. Transfer is then made to the LTl long-term response model. LTl addresses long-term recirculation cooling and, if long-term cooling is not established, the operability of containment systems for accident mitigation. o Inadequate core cooling, which leads to a core melt condition, and which transfers to the LT2 long-term core melt response model that addresses the operability of containment systems for accident mitigation. The fourth part of the plant event sequence analysis evaluates the long-term response of the plant if recirculation cooling is called for or, if core melt occurs, the operability of containment systens. In some 4-1 0091H022285
cases, the containment systems are modeled prior to the occurrence of core melt in instances where core cooling is provided, partly by l operation of containment systems. The sequences in the long-term response models terminate in either successful core cooling or specific plant damage states that define conditions important to the core and containment response analysis activities (as noted in Figure 4-1). PDSs for STPEGS are discussed in Section 7.3 and the frontline systems models are developed in Section 4.4. A hot shutdown condition is included as a possible success state for those initiating events after which it can be expected that a return to power operation will occur. This avoids the need to artificially assume, as in some other PRA studies, that each successful sequence must result in a cold shutdown condition. For initiating events that are expected to require an extended shutdown for inspection and/or repair (such as LOCA events), the success states are assumed to be cold shutdown conditions. The mission time selected for system operation in the event sequence models is 24 hours, with the plant conditions stabilized and with every reason to expect continued long-term core heat removal. Should running failures occur after the 24-hour or mission time, the temperatures are quite low and the core decay heat rate is sufficiently small that considerable time is available to provide whatever corrective action may be required. 4.2 INITIATING EVENTS The purpose of this section is to describe the selection of initiating events for the Preliminary Scoping Study. As described more fully above, an initiating event is a disturbance to steady-state plant conditions. It can begin a sequence of events having the potential to lead to core damage. Because knowledge of this potential for damage implies an analysis of the ensuing sequences, possible initiating events are selected by searching for all events that disturb the steady state. Within the initiating events groups, specific events can degrade the performance of systems useful in restoring the plant to stable conditions. Such " common cause" initiating events can often be identified by hypothesizing " external" or environmentally induced disturbances and employing failure mode and effects analysis to find support system failure modes impacting plant operations. In selecting the set of initiating events for the Scoping Study, the goals are the following: e Identify the core melt frequency and our uncertainty about it. o Quantify preliminary contributors to core melt frequency. In previously completed, full-scope PRAs, logical inductive and deductive i procedures have evolved for the selection of initiating events. Two deductive procedures are the master logic diagram and heat balance fault l 4-2 0102H052085
l l i tree method. Both of these procedures can be described as plant-level, l. qualitative fault tree analyses. An inductive procedure for identifying ! plant specific common cause initiating events is failure modes and ' effects analyses. This method is normally applied to plant support systems and provides enhanced assurance of completeness, especially with respect to the common cause initiators. Experience using the above procedures on four-loop Westinghouse PWRs with large, dry containments similar to STPEGS and the collective experience < for other PRAs and related studies provided a long list of potential initiating events. For example, the 58 Seabrook initiators, listed in Table 4-1, served as an important starting point for the selection of initiators for the STPEGS Study. The next. step in the process of selecting initiators was to fully account for specific and unique features of the STPEGS plant and systems. The-focal point of this step in the Scoping Study was the qualitative ! analysis of initiating event potential in all STPEGS systems. This analysis was documented in the systems summaries that were completed as part of the plant familiarization task. i . On the basis of the plant-specific evidence and experience with previous PRAs, and in view of the limited scope and objectives of the STP study, a list of five initiating event groups was selected for quantification. These events are presented in Table 4-2 and include 2 loss of coolant inventory initiators,1 general transient category, and 13 common cause i initiating events. The last group includes four support system faults, seven external events, and two internal plant hazards. By comparing Table 4-2 with Table 4-1, the limits of the Scoping Study are evident. The degree of analysis performed for each of the initiating events in Table 4-2 varied substantially from case to case. Initiators 1 (Small l LOCA), 3 (General Transient), 4 (Loss of Offsite Power), 5 (Loss of Essential Chilled Water), 6 (Loss of Component Cooling Water), and 7 (Loss of Essential Cooling Water) were analyzed in conjunction with a complete set of plant event trees. Initiator 2 (Interfacing Systems LOCA) did not require propagation through the event tree logic because it could be determined that it would result in a particular plant damage , state irrespective of the response of plant systems. External events 8 through 12 were analyzed in the Scoping Study in the form of bounding analyses that are described in Section 6. The remaining initiating
- events were taken into account in the quantification of uncertainties about the Scoping Study results. These results were presented in Section 2.
4.3 AUXILIARY SYSTEMS MODEL 4.3.1 GENERAL INFORMATION The analysis of the detailed plant initiating event specific event trees described in Section 4.4 is greatly facilitated and simplified by excluding the various auxiliary systems that support the successful 4-3 0102H052085
i operation of the main line systems and functions included in the detailed plant event trees. Because these auxiliary systems are required to operate following any initiating event, they are analyzed and quantified in separate auxiliary system event trees for each general category of plant initiating event. The auxiliary systems included in the auxiliary systems event trees are identified in Table 4-3. i The auxiliary systems have been modeled in three detailed event trees
, that account for the intersystem dependencies among the various auxiliary i
systems. The event trees are the following: i e Electric Power System Event Tree i e Auxiliary Systems Event Tree for Electric Power States with Offsite ! , Power Available e Auxiliary Systems Event Tree for Loss of Offsite Power Figure 4-2 presents in block diagram form the relationships among the three auxiliary systems event trees. Three event trees are required to present and quantify the STPEGS auxiliary system and states because of the large number of combinations of auxiliary system individual train failures. Table 4-4 presents the relationships among the various auxiliary systems. This matrix is used to develop the detailed auxiliary systems event trees. Table 4-4 lists each train of a particular auxiliary system and the effect of its failure upon other auxiliary systems. An "X" indicates a guaranteed failure of the auxiliary systems train if the associated train fails. The plant auxiliary systems that are not listed in Table 4-4 (but are included in the Scoping Study) are presented in Table 4-5 along with their function and the section where they are analyzed. In general, the systems shown in Table 4-5 only affect a single main line system or function and are included in the analysis of that system or function. 4.3.2 LVENT TREE DESCRIPTION 4.3.2.1 Electric Power Systems Event Tree Figure 4-3 presents the electric power system event tree developed for the STPEGS risk analysis. This event tree questions-the status of the offsite grid; the generator output breaker; the standby transformers; 13.8 kV buses IF, 1G, 1H, and IL; Class 1E 4,160V buses E1A, ElB, and E1C; Class 1E DC distribution buses E1All, E1811, and E1C11; and the emergency diesel generators DG-11, DG-12, and DG-13. This model was constructed assuming a plant transient has occurred that requires a trip
- of the main generator from the grid. The various methods of assuring power to the Class 1E 4,160V AC distribution system are shown explicitly in this model, some of which require operator action. Each sequence through the event tree is assigned a unique end state that describes the status of the plant electric power system. These end states are defined in Table 4-6.
'T 4-4 i 0102H052085 l
4.3.2.1.1 Top Event Descriptions e Event [0G]. This top event questions the availability of the offsite power grid. This event may fail as a result of the grid disturbance associated with the initiating event and subsequent generator trip, or may be a guaranteed failure for the loss of offsite power initiating events. Success of this event assures a source of power - to the standby transformers for both units and, if the generator breaker functions, to unit auxiliary transformer for the affected unit. Failure of this event leaves the emergency diesel generators as the only source of power for the Class 1E 4,160V distribution buses E1A, ElB, and E1C. e Event [UA]. This top event models the availability of the unit auxiliary transfermer to supply power to the 13.8 kV buses IF,1G, 1H, and 1J and the operation of the generator output breaker. Success of this event assures a supply of power to these non-Class 1E main buses and Class 1E 4,160V bus E1A. Failure of this top event causes a loss of power to the main bus sections of buses IF,1G,1H, and 1J and the standby bus section of bus IF, which normally supplies power to 4,160V Class 1E bus E1A. Backup power is available to the 13.8 kV buses from either standby transformer. Operator action is necessary to line up these standby power sources. e Event [S1]. This top event models the availability of the standby transformer for Unit 1. This transformer normally supplies power to the standby bus sections of 13.8 kV buses 1G and 1H and, through the auxiliary engineered safety feature transformers, to 4,160V Class 1E buses ElB and E1C. Success of this event assures a supply of power to these bus sections. Failure of this top event causes a loss of power to these bus sections. Backup power to these bus sections can be supplied by the unit auxiliary transformer, if available, through the bus section crosstie breakers, or from the standby transformer for Unit 2. Operator action is necessary to line up these power sources. o Event [S2]. This top event models the availability of the standby transformer for Unit 2. This transformer can supply power to the standby bus sections of 13.8 kV buses IF,1G, and 1H. This top event is only questioned upon failure of Top Event S1. Operator action is necessary to line up this power source. Failure of this source in conjunction with failure of Event S1 causes a loss of power at the 13.8 kV buses supplied by these transformers. o Event [1F]. This top event models the continued availability of 13.8 kV bus IF. Success of this event assures a power supply through the auxiliary ESF transformer supplying 4,160V Class 1E bus E1A. Failure of this event causes a loss of this source of power to bus EIA. e Event [1G]. This top event models the continued availability of 13.8 kV bus 1G. This top event is similar to Top Event 1F. This bus normally supplies 4,160V Class 1E bus ElB. I 4-5 0091H022085
e Event [1H]. This top event models the continued availability of l 13.8 kV bus 1H. This top event is similar to Top Event 1F. This bus ' normally supplies 4,160V Class 1E bus E1C. e Event [1L]. This top event models the availability of the emergency transformer and associated bus. This bus supplies an alternative source of power to the auxiliary ESF transformers in the event of loss of the normal AC power feed. No credit is taken for this transformer in the Scoping Study. e Event [EA]. This top event models the availability of 4,160V Class 1E bus E1A, given a source of non-Class 1E power. Success of this event implies power is available to all equipment and buses supplied by this bus. Failure of this top event is assumed to result in a nonrecoverable loss of power to the equipment supplied by this bus due to the failure modes involved (primarily bus fault). This top event is not questioned if power from the non-Class 1E distribution system is not available. Under this condition, bus faults and other failures that result in sustained loss of this bus are included with failure of the emergency diesel generator, which supplies bus E1A. e Event [EB]. This top event models the availability of 4,160V Class 1E bus ElB and is similar to Top Event [EA]. e Event [EC]. This top event models the availability of 4,160V Glass 1E Dus E1C and is similar to Top Event [EA]. e Event [DA]. This top event models the availability of Class 1E DC bus E1A11 after an initial loss of the AC feed to the battery chargers. This top event is not questioned if AC power to the chargers is supplied by the offsite grid. If power is lost to 4,160V Class 1E bus E1A, DC bus E1A11 must supply control power for bus E1A and the 4,160V breakers that must function to ensure transfer of AC power from the offsite grid sources to this emergency diesel generator. Success of this event assures DC power for control of the emergency diesel generator and the breakers that must function. Failure of this event is assumed to result in failure of bus E1A and failure of all equipment supplied by this bus. With no AC power to the battery charger for longer than 2 hours, the associated battery is assumed to be failed due to the design capacity of the battery. Class 1E DC bus E1011 is also supplied power from Class 1E AC bus E1A. Failure of this battery after a loss of AC power at bus E1A is included in the failure frequency for the turbine-driven auxiliary feedwater pump. e Event [08]. This top event models the availability of Class 1E DC bus ElB11 and is similar to Top Event [DA]. 4-6 0102H052085
e Event [DC]. This top event models the availability of Class 1E DC bus ElG11 and is similar to Top Event [DA]. e Event [GA]. This top event models the availability of emergency diesel generator DG-11 af ter a loss of power to 4,160V Class 1E bus E1A. Success of this top event assures a source of AC power to , the equipment supplied by this bus. Failure of this emergency diesel generator fails all equipment that receives power from bus E1A. This top event includes E1A bus failures, load sequencer failure, and emergency diesel generator support system failures that lead to a failure of diesel generator 11. e Event [GB]. This top event models the availability of diesel generator 12 and is similar to Top Event [GA]. e Event [GC]. This top event models the availability of diesel generator 13 and is similar to Top Event [GA]. Figure 4-4 presents the electric power systems event tree that is used to determine the unavailability of the Class 1E electric power systen after a loss of offsite )ower. This event tree only includes Top Eventr [DA], [DB],[DC],[GA], :GB], and [GC]. 4.3.2.1.2 Transfer of Information from Electric Power Tree to Other Auxiliary System Event Trees The unique end states assigned to the various electric power tree sequence and contributions of bus failures have different effects on the systems and top events contained in the other auxiliary system event trees. Table 4-7 presents in matrix form an overview of these effects. An "X" in a column for a unique end state signifies guaranteed failure of the associated auxiliary system train. j 4.3.2.2 Other Auxiliary Systems Event Tree - Offsite Power Available l Figure 4-5 presents the other auxiliary systems event tree for the t condition offsite power available. This event tree questions the status of the SSPS, the three trains of the ESFAS, the three trains of the essential cooling water system, the three trains of the CCWS, and the three trains of the essential chilled water system. This model was , constructed under the following assumptions: o ECW train A is operating; train C is on standby; train B starts j automatically in response to a safety injection actuation signal or load sequences start signal only. e CCW train A is operating; train C is on standby; train B starts automatically in response to a safety injection actuation signal or load sequences start signal only. l 0 ECH trains A and B are operating with cooling water supplied from ECW train A; ECH train C starts automatically in response to a safety injection actuation signal or load sequencer start signal only. I t 4-7 l 0091H022085
1 e EAB emergency auxiliary building ventilation trains A and B are operating; train C starts automatically in response to a safety injection actuation signal or load sequences start signal only. The status of the third train of ECW, CCW, and ECH is always questioneif, although for some initiating events, these trains can only be started by operator action. Each sequence through.this event tree is assigned a unique end state that describes the, status of the auxiliary systems shown in this model. These end states are presented and defined in Tables 4-8 and 4-9. The top events of this model are defined below: e Event [SS]. This top event questions the status of the SSPS. Two trains of SSPS are provided, train R and train S. Operation of either train ensures the appropriate actuation signals are sent to all trains of ESFAS and at least one reactor trip breaker. Failure of this top event results in failure of all train of the ESFAS and failure of .the reactor trip function if offsite power is available. This systc# is affected by the status of the Class 1E DC system because the instrumentation that provides the input signals is powered from the Class 1E DC distribution system through the 120V vital AC distribution channels, and because the SSPS cabinets are powered from the 120V AC vital AC distribution system. e Event [EA]. This top event models the availability of ESFAS train A. The ESFAS receives actuation signals from the SSPS. The types of signals and the output required depend on the initiating event under consideration. Success of this top event implies automatic signals are present at the necessary equipment. Failure of this top event is defined as failure to provide signals for all necessary functions provided by the ESFAS. From this definition, failure to provide a single function (e.g., containment spray actuation) fails the associated ESFAS actuation train. This ESFAS train is affected by the status of Class 1E DC bus E1A11. Failure of this DC bus fails ESFAS train A. e Event [EB]. This top event models the availability of ESFAS train B and is sfiiiilar to Top Event [EA]. This ESFAS train fails if Class 1E DC bus ElB11 fails. e Event [EC]. This top event models the availability of ESFAS train C and is similar to Top Event [EA]. This ESFAS train fails if Class 1E DC bus E1C11 fails, o Event [WA]. This top event models the availability of ECW train A and its associated auxiliary systems. For purposes of analysis, this train is assumed to be operating prior to the initiating event, providing cooling water for train A CCW and condensing water for the ECH chillers in train A and train B. Should the operating ECW or CCW pump fail, the standby ECW and CCW pump trains start automatically. For purposes of analysis, train C ECW and CCW pumps are selected for i standby operation. Success of this top event implies a continued 4-8 0091H022085
i l l supply of cooling water for CCW train A and ECH trains A and B. . Failure of this top event fails CCW train A and results in an ! automatic trip of the ECH chillers in trains A and B if ECW train B is not operating. This train is affected by the status of power at Class 1E AC bus E1A and Class 1E DC bus E1A11. e Event [WB]. This top event models the availability of ECW train B and is similar to Top Event [WA]. This ECW system is assumed to start automatically in response to a safety injection actuation signal or load sequencer start signal only. Failure of this top event fails CCW train B and, with failure of Top Event [WA], ECH trains A and B. This train is affected by the status of power at Class 1E AC bus ElB and Class 1E DC bus E1811. e Event [WC]. This top event models the availability of ECW train C i and is similar to Top Event [WA]. This ECW train is assumed to be selected for standby operation and will start automatically on failure of ECW or CCW train A. In addition, this train starts automatically in response to a safety injection actuation signal or load sequencer start signal. Failure of this top event fails CCW train C and ECH train C. ECH train C fails due to the assumed lineup for the train C chiller condensing water supply. This top event is affected by the status of power at Class 1E AC bus E1C and Class IE DC bus E1C11. o Event [CA]. This top event models the availability of CCW train A and its associated auxiliary systems. For purposes of analysis, this train is assumed to be operating prior to the initiating event. Should the operating ECW or CCW pump fail, the standby ECW and CCW pumps start automatically. For parposes of analysis, train C ECW and CCW pumps are selected for standby operation. Success of this top event implies a continued supply of CCW to the train A CCW loads and, depending on the initiating event, to the CCW loads that are supplied by the common header. Failure of this top event fails the equipment cooled by CCW train A. This train is affected by the status of power at Class IE AC bus EIA and Class 1E DC bus E1A11, and by the status i of ECW train A. l o Event [CB]. This top event models the availability of CCW train B and is similar to Top Event [CA]. This CCW train is assumed to start automatically in response to a safety injection actuation signal or load sequencer start signal only. Failure of this top event fails the equipment cooled by CCW train B. This train is affected by the status of power at Class 1E AC bus ElB and Class 1E DC bus ElB11, and ! by the status of ECW train B. l l e Event [CC]. This top models the availability of CCW train C and is j similar to Top Event [CA). This CCW train is assumed to be selected for standby operation and will start automatically on failure of ECW or CCW train A. In addition, this train starts automatically in response to a safety injection actuation signal or load sequencer start signal. Failure of this top event fails the equipment cooled by CCW train C. This train is affected by the status of power at 4-9 0091H022085
Class 1E AC bus E1C and Class 1E DC bus E1C11, and by the status of , ECW train C. ' e Event [SA]. This top event models the availability of ECH train A. i Each ECH train contains two water chillers, one rated at 150 tons of 1 air conditioning, the other rated at 300 tons of air conditioning. l Success of this top event implies chilled water is at the design j temperature and flow is supplied to the systems serviced by this train. The chillers in the ECH train are automatically tripped if condensing water flow is lost for more than 5 seconds. If condenser water flow is restored, the chillers automatically restart. Condensing water for this ECH train is assumed to be normally supplied by ECW train A and cross-connected to ECW train B. Failure of both ECW trains A and B is assumed to result in failure of ECH train A. Failure of this ECH train is assumed to result in (1) long-term failure of the equipment supplied only by this chiller train (primarily pump and valve room coolers) and (2) loss of one-half of the heat removal capacity for those systems supplied by any ECH train (primarily the EAB area ventiliation systems). This train is affected by the status of power at Class 1E AC bus E1A and Class 1E DC bus E1A11, and by the status of ECW trains A and B. e Event [SB]. The top event models the availability of ECH train B and is similar to Top Event [SA]. Condensing water for this chiller is assumed to be supplied normally by ECW train A and cross-connected to ECW train B. This train is affected by the status of power at Class 1E AC bus ElB and Class 1E DC bus E1811, and by the status of ECW trains A and B. e Event [SC]. This top event models the availability of ECH train C and is similar to Top Event [SA]. This ECH train is assumed to be off. This train starts automatically in response to safety injection actuation or load sequencer start signals only. Condensing water for this chiller is assumed to be lined up to CH train C only. Operator action would be required to cross-connect the condensing water supply to another ECW train. Operator action is also necessary to start this chiller in response to the failure of chiller train A or 8 if safety injection actuation has not occurred. This ECH train is affected by the status of power at Class IE AC bus E1C and Class 1E DC bus E1C11, and by the status of ECW train C. e Event [E_V]. This top event models the availability of the EAB HVAC system. The system consists of three trains of supply and exhaust fans and associated cooling units. Each train is supplied chilled water from the associated ECH train. During normal plant operation, two fan trains are operating to satisfy tha cooling requirements of the EAB. The third EAB fan train starts automatically in response to safety injection actuation or load sequencer start signals only. The three EAB HVAC . rains are affected by the following auxiliary systems: EAB HVAC Train A. Class 1E AC power at E1A, Class 1E DC power at EIA11, ECW trains A and B, and ECH train A. 4-10 0091H022085
EAB HVAC Train B. Class 1E AC power at ElB, Class 1E DC power at E1811, ECW trains A and B, and ECH train B. EAB HVAC Train C. Class 1E AC power at E1C, Class 1E OC power at E1C11, ECW train C, and ECH train C. 4.3.2.3 Other Auxiliary Systems Event Tree - Loss of Offsite Power Figure 4-6 presents the other auxiliary systems event tree for the condition loss of offsite power. This event tree, including the systems modeled, is similar to the other auxiliary system event tree for the condition offsite power available. The differences in the top event descriptions and the resulting effect on the event tree structure are described below: e Event [WA]. Failure of Top Event [WA] results in loss of cooling to diesel generator 11 with a subsequent loss of power to Class 1E AC bus E1A. Equipment powered from E1A falls. ECW train A must restart after power is available. e Event [WB]. Failure of Top Event [WB] results in loss of cooling to diesel generator 12 with a subsequent loss of power to Class 1E AC bus E18. Equipment powered from ElB fails. ECW train B starts automatically when power is available. e Event [WC]. Failure of Top Event [WC] results in loss of cooling to diesel generator 13 with a subsequent loss of power to Class 1E AC bus E1C. Equipment powered from bus E1C fails. ECW train C starts automatically when power is available. e Event [CA]. CCW train A restarts when power is available to Class 1E AC bus ElA. e Event [CB]. CCW train B starts automatically when power is available to Class IE AC bus ElB. o Event [CC]. CCW train C starts automatically when power is available to Class 1E AC bus E1C. ! e Event [SA]. CH train A restarts automatically when power is available to Class 1E AC bus E1A. l e Event [SB]. CH train B restarts automatically when power is l available to Class 1E AC bus ElB. 1 i e Event [SC]. CH train C starts automatically when power is available to Class 1E AC bus E1C. e Event [EV]. All three EAB HVAC trains start automatically when power is availaole at their respective Class 1E AC buses. i l I 4-11 ! 0091H022285 L
4.3.3 UEPENDENCIES BETWEEN MAIN LINE SYSTEMS AND AUXILIARY SYSTEMS The dependencies identified for the various main line systems and functions on the plant auxiliary systems are presented in a matrix format in Table 4-10. An "X" in a column for a particular main line system or function indicates failure of that function, given failure of the associated auxiliary system train. For SSPS and ESFAS failures, an "X" implies failure of the automatic start of the associated main line
- function.
4.3.4 QUANTIFICATION OF THE AUXILIARY TREES i Three initiating events were chosen as representative of all initiating events for the Scoping Study quantification of the auxiliary trees. l These initiating events are a general plant transient requiring a reactor trip and auxiliary feedwater actuation from the ESFAS, a loss of offsite , power requiring diesel generator operations, and a small LOCA requiring actuation of the emergency core cooling systems. The auxiliary system failure frequencies presented in Section 5.4 were used to quantify the auxiliary systems event trees for the three initiating events used in the Scoping Study. The computer code MAXIMA was used to assemble the results of the auxiliary systems quantification by initiating event, electric power end state, and auxiliary systems end states. The assembled results were used to determine the auxiliary systems impact vectors described below. 4.3.5 AUXILIARY SYSTEMS EVENT TREE RESULTS One purpose of the auxiliary systems quantification in the Scoping Study was to determine a unique set of auxiliary system impact vectors for us in quantifying the main line event trees. An impact vector defines an auxiliary system end state with a certain effect on the main line systems. A unique impact vector is a combination of all auxiliary system impact vectors having the same effect on the main line systems. For the Scoping Study analysis,17 unique impact vectors are defined for the auxiliary systems. These impact vectors and their associated auxiliary systems failures are presented in Table 4-11. All auxiliary tree end states presented in Table 4-9 were mapped into one of these impact vectors. Table 4-12 presents a mathematical representation of these impact vectors and their effects on the various main line event tree top events. 3 The impact vector having the highest frequency is " AUX," which describes , the state in which all auxiliaries are available. Impact vector "9," which describes the loss of EAB HVAC, has the next highest frequency. For this point estimate analysis, impact vector "9" is assumed to result in the loss of plant-essential AC power. Impact vector "9" and impact vector P, which is associated with loss of essential cooling and loss of j component cooling with electric power available, were found to contain the most important contributors to core melt frequency as described more fully in Section 2. l I i i 4 4-12 0102H052085
i 4.4 'FRONTLINE SYSTEMS MODEL
~
Four event sequence models have been developed to analyze the plant response to three major initiating event categories that are to be considered in the Scoping study. These event sequence models are as follows: Model Events Analyzed General Transient ' General Transient Loss of Offsite Power Loss of Essential Chilled Water Loss of Component Cooling Water Loss of Essent,lal Cooling Water Small LOCA Small LOCA ' Long-Term Response (LT1) Sequences Successfully Exiting the SLOCA and General Transient Models Following Injection of the RWST Long-Term Core Melt Sequences Exiting the SLOCA and Response (LT2) General Transient Models in a Core Melt Condition The general transient and small LOCA models analyze the early response of the plant, whereas the LT1 and LT2 models analyze the long-term response. Each model is described below, first in the form of an event sequence diagram, then in the form of an event tree. The ESus are made up of several explanatory blocks with symbology as noted in Figure 4-7. ESus are useful in describing in a more general and easily understood manner the various sequence paths; ESDs are used for documentation purposes and developing the event trees but do not easily lend themselves to direct quantification. The event trees illustrate the same logic information portrayed in an ESO but in a different manner; event trees explicitly show each possible sequence and are used in the actual quantification process. The top events in the event trees correspond to groups of event blocks in the ESD. A node is shown in the event tree when a top event is questioned in a specific sequence. If the questioned top event is successful, the sequence continues to the right of the node; if the top event fails, the sequence continues downward from the node. When transfer is made from the early response model to either the LT1 or LT2 long-term response models, certain information that will affect the long-tenn response must be specified as a long-term model entry condition. 4.4.1 THE GENERAL TRANSIENT EVENT SEQUENCE MODEL The general transient event sequence model is used to evaluate the early response of a broad range of initiating events including reactor trip, 4-13 0102H052085
turbine trip, loss of main feedwater flow, and loss of load and offsite power. The impact of the specific initiating event category on both the auxiliary systems and the frontline systems is determined as part of the systems analysis activity, and the corresponding information is included as an integral part of the event tree quantification. The general l transient event sequence diagram is shown in Figure 4-8. The events shown in each block of the ESD are, for the most part, sel f-explanatory. The ESD events are numbered 1E through 32E. The ESD events are combined into functional groups (within dashed lines in Figure 4-8) and numbered 1 through 19 in the simplified ESD of Figure 4-9. That diagram provides an easy-to-follow logic of the top event represented in the general transient event tree (Figure 4-10). This event tree provides the logic in a readily quantifiable form. The following discussion axplains some of the events in the general transient ESD and its relationship to the simplified ESD and the event tree. Reasons for the branching logic of the event tree are also given. The general transient ESD begins with the transient itself, then in event IE asks whether the reactor has tripped. This event is actually handled in the auxiliary systems model described in Section 4.3. Following the event sequence diagram beyond the reactor trip event leads to the question of whether steam demand is secured. This is accomplished either by tripping the turbine or shutting the MSIVs. If successful, the decay heat must be removed by steaming the steam generators through either the turbine bypass or the atmospheric steam dumps, or via the steam generator safety valves. In cases where excessive steam demand continues, overcooling events occur and questions are raised concerning i pressurized thermal shock. In event 14E, the high head injection system responds to the overcooling depressurization of the reactor coolant system and the question asked is: "Does the operator control high head injection?" If not, pressurized thermal shock conditions exist and the integrity of the
- reactor vessel is questioned. If intact, the event sequence continues to event 6E with the constraint that steam-driven auxiliary feedwater and main feedwater pumps are not available. If vessel integrity is lost, {
i core melt is assumed and it is questioned whether containment spray I provides water to flood the reactor cavity region. l l For cases where no steam generator heat removal occurs, the sequence l branches to event 29E for bleed and feed cooling. In the cases with l successful steam removal, questions regarding the availability of l feedwater are then asked. These questions include events SE, 6E, 21E, 22E, 25E, and 26E. With no feedwater available, the sequence again branches to bleed and feed at event 29E. With no auxiliary feedwater, if the operator does not recognize this condition and respond properly, core melt ensues. Again, questions are asked about the containment spray. In scenarios with successful feedwater, it is possible that a pressurizer PORV opens and, if so, must reclose. Failure here is equivalent to a small LOCA. With the PORV reseated, other questions of potential loss of RCS inventory arise. Event 8E includes a variety of potential leak paths; e.g., reactor coolant pump seal leak LOCA via the letdown system 4-14 0091H022085
either to the VCT or to the containment via the 600 psia relief valve; and a low flow LOCA via the reactor coolant pump seal return lines. Under LOCA conditions, a branch is made to event 33E for RCS makeup. On successful inventory control, it is questioned whether the operator controls feed, what the effects of failure to control feed are, and finally, long-term questions of stabilization. Bleed and feed event 29E involves the operator opening two pressurizer PORVs and su3 plying high head injection. Success along this path can be either via tne close loop residual heat removal system or through a branch to long-term tree LT1 and open loop recirculation. In core melt cases, containment spray is questioned and we branch to LT2. Event 33E follows the LOCAs in event 8E and questions whether the operator controls the LOCA by decreasing RCS pressure. If so, continued cooldown is asked via event 9E. If the operator failed to depressurize, it questions whether response to the small LOCA condition is successful, i.e., did we have high head injection and either closed loop recirculation or open loop via LT1? Some of the events in the ESD need not be quantified for various reasons. The following discussion takes us from the ESD to the top events of the event tree as depicted in the simplified ESD. For steam relief, we have decided to consider only the atmospheric steam dumps, event 12E for several reasons. First, there is a variety of conditions under reduced auxiliary system states where turbine bypass will not work, of ten due to loss of condensate flow and instrument air arising from loss of nonvital AC power. Also, the atmospheric steam dumps are very reliable; thus, including the turbine bypass is not essential to obtaining accurate results. Again, because of atmospheric steam dump reliability, questions of the steam generator safety valve are not, in general, essential. Moreover, use of the safety valves is an unusual operating mode. They may stick open and lead to depressurization of a steam generator, raising questions in the operators' minds as to whether continued use of the affected steam generators is warranted. As a result of safety valves sticking open, the operator loses direct control of steam generator pressure and reactor coolant system temperature. Therefore, only steam relief via the ASDs is included in event tree tops 4 and 5, and turbine trip and ItSIV closure are grouped as a single event for the event tree top event 1. In overcooling scenarios, events 14E and 15E are combined into a single event 2, for the event tree. Here, high head injection is almost guaranteed to work. The only serious question is whether the operator can control high head injection. A rather large group of events are combined into event tree events 4, 5, and 6. First, as discussed earlier, all steam relief is modeled simply as the opening of the atmospheric steam dump on the steam generators. Next, a combination of ASDs, and situations of two or more auxiliary feedwater trains operating are combined into event 4 of the event tree. Given the failure of event 4, one train of auxiliary feedwater combined with atmospheric steam dump makes up event 5. The main feedwater system l 4-15 l 0091H022085
is guaranteed to trip under normal situations when the turbine bypass automtically brings the RCS average temperature to the no-load Tave setpoint. It is not very likely the operator will need to turn to the startup feed pump. Finally, it is assumed that if two or more trains of auxiliary feedwater work, the pressurizer PORVs will not be demanded. To allow the chance that PORVs are at times demanded, we ask for their opening and reclosure for every situation in which only one train of auxiliary feedwater is working. This is event 6 in the event tree. It is unlikely that the operators will overfill the steam generators (event 9E). Even if they do, it is extremely unlikely that the main steam lines fail; and, even if they fail, the only substantial effect contributing to core melt is the loss of turbine-driven feed pumps. For these reasons, we have decided not to model events 9E, 27E, and 28E. The remaining events provide much more obvious linkage between the ESD and event tree tops. For example, each train of ECCS can be affected by common failures, and some common failures can even couple trains. These effects are modeled in event tree tops 10,11, and 12 to account for trainwise common failures among two trains and common failures among three trains such as RWST failures. Each train is modeled separately because of trainwise effect between the ECCS common equipment, the high head injection equipment and the containment spray equipment. These interdependencies are also important in the long-term trees for further common recirculation equipment and low head injection equipment for each train. Now, examining the generalized transient event tree in Figure 4-10, we find the same top events just discussed. Table 4-13 gives the success criteria for each of the top events. These are modeled and quantified in the systems analysis of Section 5. For scenarios in the general transient tree that branch to either the LT1 tree or the LT2 tree, it is essential to keep track of the boundary conditions that are of importance in those later trees. Tables 4-14 and 4-15 give the coding used for transfer state between carly and long term response event trees. For example, LTIE transfer state branches from an early event tree with two high head safety injection pumps failed (one due to an ECCS common effect) and no loss of RCS inventory control (i.e., event 0I succeeds). Proceeding directly to the general transient event tree, the normal expected sequence of events is event sequence 1 in which all events succeed. If event ON fails, i.e., the operator fails to provide the long-term cooling, core melt surely occurs, and in sequences 2 through 28 , we branch to the LT2 tree. In defining each branch, we keep track of the I number of containment spray system failures and whether these failures L are due to common effect. This is important for defining whether water [ can be present in the reactor vessel cavity. The presence of water is I important in assessing the plant damage states as described in Section 7.2. If event 01 fails and we have a loss of reactor inventory control, i.e., either a reactor coolant pump seal LOCA, a LOCA via the letdown system, or a LOCA via the reactor coolant pump seal return line, safety injection l l i 4-16 0091H022085
is required. If both safety injection and closed loop residual heat removal cooling are successful, we are in a success state on event sequence 29. If closed loop residual heat removal is not successful, we branch via sequence 30 to LT1G to consider long-term open loop recircu-
.lation. Sequences 31 through 93 represent variations on the scheme with 0,1, 2, or 3 high head injection pumps failed either directly or through comon failures in EA, EB, and EC as well as all possible failure states for containment spray injection for the core melt scenarios.
Should AF fail (less than two auxiliary feedwater trains and steam relief are available) we can proceed to a success state in sequence 94 if F1 (one auxiliary feedwater train), PR (pressurizer power-operated relief valve opens and closes), 01 and 0A are successful. If 01 fails, sequences 122 through 186 duplicate sequences 29 through 93. If ON fails, sequences 95 through 121 duplicate sequences 1 through 28. If PR fails, a small LOCA exists because either the PORV stuck open or it did not open and rising pressure caused a leak elsewhere in the reactor coolant system. Sequences 187 through 251 represent the early tree response to the small LOCA condition and are identical with sequences 29 through 93 except that the LT1 and LT2 codes are given for no 0I failure. The OI branch is not asked in this case because, first, a LOCA already exists and, second, release paths outside containment are also tracked in event CI in the LT1 and LT2 trees. With failure at AF and F1, no auxiliary feedwater is available so heat removal via the steam generators fails. In this case, the only route to success is bleed and feed cooling via event 08. PR is not asked because, although the PORVs may indeed lift, bleed and feed requires they both be
-manually opened; otherwise, core melt will follow. Event 01 is not asked since, as above event OB induces a LOCA and if OB fails, core melt will ensue; also, release paths outside containment induced by the OI branch are also tracked.in the LT trees in event CI. When event OB is successful and the operator decides to initiate bleed and feed cooling and open the PORVs, sequences 252 through 316 are identical to sequences 187 through 251, which is the response to a small LOCA. If OB
. fails, core melt is guaranteed and sequences 316 through 343 are identical to sequences 2 through 28.
- When TT fails (i.e., steam demand is not' secured and a rapid cooldown l ensues) the operator should control high head injection event OH. If the operator is successful, scenarios 344 through 686 look exactly like the 343 scenario already described. If the operator fails to control high head injection but the reactor vessel remains intact, the same 343 sequences are replicated as sequences 687 through 1,029. If the reactor vessel fails through the pressurized thermal shock, core melt is
. guaranteed and scenarios 1,030 through 1,056 are replicates of L sequences 2 through 28.
l 4.4.2 THE SMALL LOCA EVENT SEQUENCE MODEL The small LOCA event sequence model is used to evaluate the early response to small LOCA events. Small LOCAs include unisolated breaks or
- openings in the RCS pressure boundary of a sufficient size so that the l
l 4-17 i 0091H022085
4 1 charging pump is unable to maintain adequate RCS inventory (implying leak 4 rates in excess of 120 gpm) but not large enough to be considered medium LOCAs. The effective diameter of small LOCA breaks range from 3/8 inch to 2 inches. Breaks smaller than 3/8-inch diameter are considered to be leaks that result in an orderly plant shutdown. Although the equipment required to prevent core melt during a small LOCA may vary somewhat depending on the size of the break within the range considered, conserva-tive success criteria will be used. Accordingly, the flow out the break will be considered too small to remove sufficient energy compared to the ! core decay heat, so it is assumed that at least one PORV will need to be opened if bleed and feed cooling is required. One of three high head injection pumps is required for adequate inventory makeup. The small LOCA model is a special case of the events included in the general transient model. Therefore, only a discussion of the restric-tions on the general transient model leading to the small LOCA model is given here. The success criteria for the top events in the small LOCA event tree (Figure 4-11) are identical for similar events in the general
- . transient tree except as discussed below.
The first difference between the general transient and small LOCA event trees is that events AF and F1 of the general transient are replaced by the single event F1 in the small LOCA tree. This is because a LOCA already exists so that examination of the PORV lifting and reclosing is not essential.- Thus, for the small LOCA event F1 represents the operation of one or more trains of auxiliary feedwater combined with steam relief from a steam generator receiving feedwater. Event PR is no longer required. Again, since a LOCA already exists event 01 is not required. Event ON is not required either, because success in the case of small LOCAs requires actions of the operator in the long-term response tree LT1. The only remaining difference is that event OR, closed loop residual heat removal cooling, has not been included in the small LOCA event tree because the residual heat removal pumps are located inside the containment and are not qualified for a hot, moist atmosphere that would exist following a LOCA. The logic for branching is as described under general transient. 4.4.3 THE LT1 LONG-TERM RESPONSE EVENT SEQUENCE MODEL The LT1 long-term response event sequence diagram, Figure 4-12, is 4 entered frem either the early response general transient or small LOCA models for sequences where a significant portion of the RWST has been injected into the vessel and addresses the transfer to either high or low , pressure recirculation cooling. In the ESD we first question whether fan ' coolers provide long-term containment, and hence recirculation, cooling. If the fan coolers work, all that it is required to reach a successful end state is that the common recirculation equipment and one out of three high head safety injection pumps operate. On failure of the common recirculation equipment no long-term containment spray is possible and only the containment isolation questions are asked. If the high head safety injection system fails, low head recirculation is possible; if that fails, both containment spray and containment isolation are important. l 4-18 0091H022085
If the fan coolers fail, successful core cooling requires that reactor pressure be low enough for low head recirculation through the residual heat removal heat exchanger to cool the core. If low head recirculation is not possible prior to core melt it could still provide long term containment sump cooling after reactor vessel melt-through permits low head flow. To prevent pressure buildup in the containment, fission product heat must be removed from the containment atmosphere, either via operating containment spray system or one operating fan cooler. In the baseline model we have not tracked the remote chance that only one containment fan cooler is operating with no containment spray pumps and one train of low head recirculation with a residual heat removal heat exchanger. With this discussion, the events in the LT1 ESD are self-explanatory and directly convert into the LT1 event tree (Figure 4-13). The only additional information in the event tree involves the tracking of the separate trains of common recirculation equipment and low head safety injection. Table 4-16 gives success criteria of the LT1 event tree sequences in accordance with the definitions of Section 7.2. 4.4.4 THE LT2 LONG-TERM CORE MELT RESPONSE EVENT SEQUENCE MODEL The LT2 long-term response core melt event sequence diagram, Figure 4-14, is entered from either the general transient or the small LOCA model for sequences that have resulted in or are leading to core melt. The model questions whether sufficient RWST water has been injected into the containment to provide debris bed cooling as well as whether containment functions that can mitigate the offsite consequences (namely, containment isolation, spray, and heat removal) are working. The ESD is a subset of the LT1 ESO and requires no further explanation. The corresponding event tree shown in Figure 4-15 and the top event success criteria are as given for LT1. I i l l l l I l 4-19 0091H022085 L _ _ _ _ _ _ - -
TABLE 4-1. CATALOG 0F P0TENTIAL INITIATING EVENTS (TAKEN FROM THE SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT) Sheet 1 of 2 Group Initiating Event Categories Selected for Separate Quantification Loss of Coolant 1. Excessive LOCA Inventory 2. Large LOCA
- 3. Medium LOCA
- 4. Small LOCA
- 5. Interfacing Systems LOCA
- 6. Steam Generator Tube Rupture General 7. Reactor TripTransients
- 8. Turbine Trip
- 9. Total Main Feedwater Loss
- 10. Partial Main Feedwater Loss
- 11. Excessive Feedwater Flow
- 12. Loss of Condenser Vacuum
- 13. Closure of One MSIV
- 14. Closure of All MSIVs
- 15. Core Power Excursion
- 16. Loss of Primary Flow
- 17. Steam Line Break Inside Containment
- 18. Steam Line Break Outside Containment
- 19. Main Steam Relief Valve Opening
- 20. Inadvertent Safety Injection Common Cause Initiating Events Support 21. Loss of Offsite Power System Faults 22. Loss of One DC Bus
- 23. Total Loss of Service Water
- 24. Total Loss of Component Cooling Water Seismic 25. 0.79 Seismic LOCA Events 26. 1.0g Seismic LOCA
- 27. 0.29 Seismic Loss of Offsite Power
- 28. 0.39 Seismic Loss of Offsite Power
- 29. 0.49 Seismic Loss of Offsite Power
- 30. 0.59 Seismic Loss of Offsite Power
- 31. 0.79 Seismic Loss of Offsite Power
- 32. 1.0g Seismic Loss of Offsite Power l
CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 l 0013H022085 4-20
TABLE 4-1 (continued) Sheet 2 of 2 Group Initiating Event Categories Selected for Separate Quantification Common Cause Initiating Events (continued) Fires 33. Cable Spreading Room - PCC Loss
- 34. Cable Spreading Room - AC Power Loss
- 35. Control Room - PCC Loss
- 36. Control Room - Service Water Loss
- 37. Control Room - AC Power Loss
- 38. Electrical Tunnel 1
- 39. Electrical Tunnel 3
- 40. PCC Area
- 41. Turbine Building - Loss of Offsite Power Turbine 42. Steam Line Break Missile 43. Large LOCA
- 44. Loss of Condenser Vacuum
- 45. Control Room Impact
- 46. Condensate Storage Tank Impact
- 47. Loss of PCC Tornado 48. Loss of Offsite Power and One Missile Diesel Generator
- 49. Loss of PCC
- 50. Control Room Impact Aircraft 51. Containment Impact Crash 52. Control Room Impact
- 53. Primary Auxiliary Building Impact i Flooding 54. Loss of Offsite Power l 55. Loss of Offsite Power and l One Switchgear Room l 56. Loss of Offsite Power and Two Switchgear Rooms
- 57. Loss of Offsite Power and Service Water Pumps l
Others 58. Truck Crash into Transmission Lines I l l CAUTION: PRELIMINARY RESULTS-IMPORTANT ONCER TAINTIES ( DESCRl8ED IN SECTION 2 001311022085 4-21
TABLE 4-2. INITIATING EVENTS QUANTIFIED IN THE STPEGS STUDY Initiating Event Categories Code Group Selected for Separate Designator Quantification Loss of Coolant 1. Small LOCA SLOCA Inventory 2. Interfacing Systems LOCA ISLOCA General Transients 3. General Transient GT Common Cause Initiating 2 vents Support System 4. Loss of Offsite Power LOSP Faul ts 5. Total Loss of Essential TLECH Chilled Water
- 6. Loss of Component Cooling Water
- 7. Loss of Essential Cooling Water External Events 8. Aircraft Crash AC
- 9. Turbine Missile TBM
- 10. Tornado Excessive Wind TW
- 11. Tornado Missile TM
- 12. Hazardous Chemical Release HCR
- 13. Seismic Events EQ
- 14. External Flooding EFL Internal Plant 15. Fire IF Hazards 16. Internal Flood IFL CAUTION: PRf LIMINARY RESULTS-IMPORTANT UNCERTAINTIES l DESCRISED IN SECTION 2 )
i I
~
0013H122884
1 TABLE 4-3. AUXILIARY SYSTEMS INCLUDED IN AUXILIARY EVENT TREES Sheet 1 of 2 System Function ELECTRIC POWER SYSTEM Non-Class IE Provides power to Class 1E AC distribution Distribution systems during normal and transient plant operations., Provides power for certain nonsafety systems analyzed in the Scoping Study. 1E - AC Distribution Provides power to other auxiliary systems and plant main line systems to allow mitigation of plant transient events. Includes emergency diesel generators. 1E - DC Distribution Provides control power for large AC loads and various solenoid-operated valves. Provides normal source of power for the 120V vital AC distribution systems. Provides power for the engineered safety features actuation system, emergency diesel generator operation, and other DC loads. SOLID STATE PROTECTION Receives input from various plant monitoring SYSTEM systems; using this input, sends signals to the engineered safety features actuation system, the reactor trip system, the Class 1E AC distribution system, and other functions. The signals developed depend on the input parameters and the plant initiating event. ENGINEERED SAFETY Receives actuation signals from the SSPS and, FEATURES ACTUATION using master relay / slave relay combinations, SYSTEM sends these signals to the equipment that must operate to mitigate the initiating event. ESSENTIAL COOLING WATER Provides cooling water from the ECW pond to SYSTEM the component cooling water heat exchangers, essential chilled water chiller condensers, I and emergency diesel generator coolers, and returns the water to the ECW pond. COMP 0NENT COOLING WATER Supplies cooling water to the reactor coolant SYSTEM pumps, centrifugal charging pumps, residual heat removal pumps and heat exchangers, the containment fan cooler units under accident
, conditions, and other process loads to remove l generated heat.
CAUTION: PRE LIMINARY RESULTS-lMPORTANT UNCERTAINTIES 0103H052085 4-23 DEsca'sto 'N stcTioN L .
TABLE 4-3 (continued) Sheet 2 of 2 System Function ESSENTIAL CHILLED WATER Provides chilled water to various safety SYSTEM related ventilation system cooling coils to provide suitable environmental conditions for continued equipment operation and for the plant operators. i ELECTRICAL AUXILIARY Supplies cooled air to vital equipment BUILDING MAIN AREA located in the main area of the EAB. HVAC SYSTEM REACTOR TRIP SYSTEM Provides negative reactivity to shut down the fission process. Although a main line function, reactor trip is quantified with the . plant auxiliary systems because of its direct relationship with the non-Class 1E distribution system and the SSPS. CAUTION: PRELIMINARY RESULTS-IhMORTANT ONCERTAINTIES DESCRIBED IN SECTION 2 0013H122884 4-24
3 I
.sw /
I i i TABLE 4-4. MATRIX 0F AUXILIARY SYSTEM 1 0 # Class It Class IE
- EDGs $$PS 4.160V AC Buses DC Buses 1 E18 E1C EIAH EIGH UCH E4H MH E2 KU Sys afn Cffsite Grfd (a) (a) (a) (b) (b) (b) (b) (c) (c) (c) 4.160Y Bus EIA - (f) (f) (g) 4.160V Bus E18 . (f) (g) 4.160V Bus E1C . (f) (g)
DC Sus EIA11 . I (f) f (f) DC Bus E1911 - K f DC Bus ETC11 . I i 1 DC Bus E1011 EDG DG11 (j) (f) (f) . Eoc DG12 (j) (f) . EDG DG13 (j) (f) .
$ $$PS Trafn R ,
1 55P5 Train 5 ESFAS Train A ESFAS Trafn B 4 ESFAS Train C ECW Trafn A I ECW Train B I ECW Train C I
, CCW Train A i CCW Train B CCW Train C I ECH Trafn A (q (q
l lq) q) i 'q)
'q)
(q) (q) (q' (q i q) q) () () (r) (r)
i i
ECH Train 8 () ECH Train C (ql lq) iq) (q) (q i
'q) (r) 's
. EA8 HVAC (q) (q) (q) (q) (q) (q) a Reactor Trip
%etters in parentheses refer to notes on the following page.
j **An *x* denotes a guaranteed fallure of the systems train f f the associated trafn falls. d t 1 0013H122884 f 1 i W W. '_l i>_ _ - _ - - _ - - _ _ _ - - _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ . _ _ . - _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ _
e )AUXILIARYSYSTEMINTERDEPENDENCIES* Sheet 1 of 2 T5FAS Essential Cooling Water CCW Essential ChtIIed ifater EA8 Reactor HVAC Trip Trata Trata Train Trata Trata Train Train Train Train Trafn Trata Train A B C A B C A B C A 8 C (4) (d) (d) (d) (d) (d) (d) (d) (d) id) (e) X** X X (h) X X X X X X I (h) I X X X l$$ X X X X X 1 h l I 1:1 !!! !!! !!!
. (.) (.i (.)
. (m) (m) (m) (n)
- X toi
,,)
Aperture Card
- X X (p) l
~
T1
~
APERTURE
. CARD I
l l yogosooct-a CAUTION: PR8LIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 4-25
TABLE 4-4 (continued) Sheet 2 of 2
- a. For a loss of the offsite grid, bus failures represented by Top Events [EA], [EB],
and [EC] are included in the frequency of failure of the respective emergency diesel generator.
- b. DC buses are only questioned following a loss of AC power to the associated battery charger (s).
- c. .With a loss of the offsite grid, or loss of power to a Class lE 4,160V AC bus, the ,
EDGs receive automatic start signals. The load sequencers are included with the , associated EDGs. l .d. With a loss of offsite power, these auxiliary system trains receive automatic start j signals from the load sequencer when the EDG is supplying power to the associated i~ bus.
- e. With a loss of offsite power, the control rod drive motor generator sets lose power. Successful reactor trip for this case does not require reactor trip breaker operation,
- f. With no AC power to the batte chargers, the batteries are assumed to fail after 2 hours (rated batter ).
- g. Top Events [EA], [EB]y capaci represent bus failure or distribution failures that
,and[
cannot be recovered by EDG operation.
- h. Loss of one EA8 HVAC train.
- 1. Loss of power at a Class lE DC bus results in a loss of power to the associated 120V vital AC power distribution channel. The associated instrumentation is designed to
" fail safe."
I j. If the EDGs are required, the bus failures represented by Top Events (EA], [EB], I and [EC] are included with the frequency of failure of the associated EDG. j k. Loss of a single SSPS train results in a loss of one of two input, available to the ESFAS train. I 1. Loss of a single SSPS results in a loss of trip signal to the associated reactor trip breaker. t c. Loss of an ESFAS train results in no automatic start signals to the associated ECW l train, CCW train. ECH train, and ventilation train for those plant initiating events
- that require safety injection actuation. If the ECW and CCW trains selected for
' standby operation are associated with the failed ESFAS train, the failure of this ESFAS train has no effect on the operation of the ECW and CCW train. With concurrent loss of offsite power, failure of an ESFAS train does not affect the l operation of the associated ECW, CCW, and ECH trains because the load sequencer will
- provide start signals to these trains.
l n. Train C of EAB HVAC does not start automatically if safeguards actuation is required.
- 6. For purposes of quantification ECH trains A and B are assumed to be operating with
) chiller condenser water supplied by ECW train A. With loss of ECW train A, ECW j train B can supply chiller condenser water to ECH trains A and B without valve i lineup changes. ECH train C is assumed to be lined up to receive chiller condenser l water only from ECW train C.
- p. Loss of train C of EA8 HVAC.
- q. Failure of a single operating ECH train will result in increasing temperatures in i the areas supplied by the electrical auxiliary building main area ventilation i system. This may have an effect on the operation of the Class lE switchgear and I batteries supplied by this EA8 ventilation system. Operator action is necessary to
! start the third ECH train if offsite power is available and no safety injection
- actuation signal is expected.
( r. Failure of a single operating ECH train will result in increasing temperature in the ! area supplied by the EAB main control room ventilation system. This may have an effect on the operation of the SSPS and other equipment cooled by this ventilation system. Operator action is necessary to start the third ECH train if offsite power is available and no safety injection actuation signal is expected. [ f 1 CAU floN: PRELIMINARY RESULTS-IMPoRTANT UNCERTAINTIES DEsCRieEO IN sECTION 2 I i b 4-26 0013H122884
TABLE 4-5. AUX 1LIARY SYSTEMS NOT SHOWN IN THE PLANT AUXILIARY SYSTEM MODELS' System Function and Where Included EAB Main Control Room HVAC Supplies cooled air to equipment. located in the System main control room HVAC envelope. Not analyzed in the Scoping Study. MAB Supplementary Coolers Supplies cooled air to the charging pump cubicles
' System and to the component cooling pump cubicles.
Included with the associated equipment.
-MAB Supplementary Fan Coil Supplies cooled air to the essential chiller areas.
Cooling System Included with the associated cssential chilled water' train. FHB Supplementary Cooler Supplies cooled air to the safety injection system System pump and valve area cooler. Included with the associated safety injection system train. Penetration Space HVAC Supplies cooled air to the three elevations of the System Cooling Subsystem electrical penetration area. Not analyzed in the Scoping Study. Isolation Valve Cubicle and Supplies air to the auxiliary feedwater pump Pips Penetration Area cubicles and valve cubicle. Included with the
, Ventilation System auxiliary feedwater system analysis.
Diesel Generator Building Removes heat from the associated emergency diesel Emergency Ventilation generator building during operation of the System emergency diesel generator. Included with the analysis of the emergency diesel generators in the electric power analysis. Essential Cooling Water Provides a suitable environment for the essential Intake Structure Ventilating cooling water pumps and associated equipment. System Included with the analysis of the essential cooling water system. Plant Instrument Air Provides air for the operation of various process control valves and supports plant operation. Not analyzed in the Scoping Study. CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCE RTAINTIES DESCRISED IN SECTION 2 0103H052085 4-27
CAUTION: PRELIMINA%Y RESULTS-IMPORTANT UNCE7.TAINTIES DESCRISE313 SECTION 2 TABLE 4-6. ELECTRIC POWER TREE UNIQUE END STATES Sheet 1 of 4 l End State Definition Designation j AC AC power is available from the offsite grid. Operating equipment continues to run. ACS AC power is available to the Class 1E equipment from the emergency diesel generators. Previously operating equipment must restart. Bil Initial power loss to Class 1E AC bus E1A. Power is restored from the offsite grid or the emergency diesel generator. Equipment powered from bus E1A receives start signals from the bus load sequencer. Previously operating equipment powered from bus EIA must restart. B12 Initial power loss to Class 1E AC bus ElB. Similar to end state Bil. B13 Initial power loss to Class 1E AC bus E1C. Similar to end state Bil. B21 Initial power loss to Class 1E AC buses E1A and ElB. Power is restored from the offsite grid or the emergency diesel generators. Previously operating equipment must restart. B22 Initial power loss to Class 1E AC buses E1A and E1C. Similar to end state B21. B23 Initial power loss to Class 1E AC buses ElB and E1C. Similar to end state B21. All Loss of Class 1E AC bus E1A due to bus faults or distribution failures. Offsite power is available. A12 Loss of Class 1E AC bus E18. Similar to end state All. Offsite power is available. A13 Loss of Class 1E AC bus E1C. Similar to end state All. Offsite power is available. C11 Loss of Class 1E DC bus E1A11. Offsite power is available. Loss of Class IE AC bus EIA by definition. C12 Loss of Class 1E DC bus ElB11. Offsite power is available. Loss of Class 1E AC bus ElB by definiticn. C13 Loss of Class 1E DC bus E1C11. Offsite power is available. Loss of Class 1E AC bus E1C by definition. A181 Loss of Class 1E AC bus EIA, initial power loss with restoration for Class 1E AC bus ElB. Offsite power is available. , A182 Loss of Class 1E AC bus ElB, initial power loss with restoration for Class 1E AC bus E1A. Offsite power is available. A103 Loss of Class 1E AC bus E1A, initial power loss with restoration for Class 1E AC bus E1C. Offsite power is available. A104 Loss of Class 1E AC bus E1C, initial power loss with restoration ! for Class 1E AC bus E1A. Offsite power is available. A1BS Loss of Class IF AC bus ElB, initial power loss with restoration for Class 1E AC bus E1C. Offsite power is available. A186 Loss of Class 1E AC bus E1C, initial power loss with restoration i for Class 1E AC bus ElB. Offsite power is available. AB21 Loss of Class 1E AC bus EIA, initial power loss with restoration < for Class 1E AC buses ElB and E1C. Offsite power is available. 0013H122884 4-28
CAUTION: PRELIMINAR Y L ESULTS-IMPORTANT UNCELTAl% TIES DESCZl8E3 83 5ECTION 2 TABLE 4-6 (continued) Sheet 2 of 4 End State Definition Designation AB22 Loss of Class 1E AC bus ElB, initial power loss with restoration for Class 1E AC buses EIA and E1C. Offsite power is available. AB23 Loss of Class 1E AC bus E1C, initial power loss with restoration for Class 1E AC buses E1A and ElB. Offsite power is available. B1C1 Initial power loss with restoration for Class 1E AC bus E1A, loss of Class 1E DC bus ElB11. Offsite power is available. B1C2 Initial power loss with restoration for Class 1E AC bus ElB, loss of Class 1E DC bus E1A11. Offsite power is available. 81C3 Initial power loss with restoration for Class 1E AC bus EIA, loss of Class 1E DC bus E1C11. Offsite power is available. B1C4 Initial power loss with restoration for Class 1E AC bus E1C, loss of Class 1E DC bus E1A11. Offsite power is available. B1C5 Initial power loss with restoration for Class 1E AC bus ElB, loss of Class 1E DC bus E1C11. Offsite power is available. B1C6 Initial power loss with restoration for Class 1E AC bus E1C, loss of Class 1E DC bus E1811. Offsite power is available. B2C1 Initial power loss with restoration for Class 1E AC buses E1A and ElB, loss of Class 1E DC bus E1C. Offsite power is available. B2C2 Initial power loss with restoration for Class 1E AC buses E1A and E1C, loss of Class 1E DC bus ElB. Offsite power is available. 82C3 Initial power loss with restoration for Class 1E AC buses ElB and E10, loss of Class IE DC bus E1A. Offsite power is available. Eli Loss of offsite power and Class 1E AC bus EIA. E12 Loss of offsite power and Class 1E AC bus ElB. E13 Loss of offsite power and Class 1E AC bus E1C. Dil Loss of offsite power and Class 1E DC bus EIA11. D12 Loss of offsite power and Class 1E DC bus ElB11. D13 Loss of offsite power and Class 1E DC bus E1C11. A21 Loss of power at Class 1E AC buses EIA and ElB. Offsite power is available. A22 Loss of power at Class 1E AC buses E1A and E1C. Offsite power is available. A23 Loss of power at Class 1E AC buses ElB and E1C. Offsite power is available. A2B1 Loss of power at Class IE AC buses E1A and ElB, initial power loss with restoration for Class 1E AC bus E1C. Offsite power is available. A2B2 Loss of power at Class 1E AC buses E1A and E1C, initial power l loss with restoration for Class 1E AC bus ElB. Offsite power is i available. A283 Loss of power at Class 1E AC buses ElB and E1C, initial power loss with restoration for Class 1E AC bus EIA. Offsite power is available. A1C1 Loss of power at Class 1E AC bus E1A and Class 1E DC bus E1811. Offsite power is available. l t l 0013H122884 4-29
CAUTION: PRELIMINARY EESULTS-IMPORTANT UNCEITAINTIES DESCRISED IN SECTION 2 TABLE 4-6 (continued) Sheet 3 of 4 End State Definition Designation l A1C2 Loss of power at Class 1E AC bus ElB and Class 1E DC bus E1A11. Offsite power is available. A1C3 Loss of power at Class 1E AC bus E1A and C1 ass 1E DC bus E1C11. Offsite power is available. A1C4 Loss of power at Class 1E AC bus E1C and Class 1E DC bus E1A11. Offsite power is available. A1CS Loss of power at Class 1E AC bus ElB and Class 1E DC bus E1C11. Offsite power is available. A1C6 Loss of power at Class 1E AC bus E1C and Class 1E DC bus ElB11. Offsite power is available. ACB1 Loss of power at Class 1E AC bus E1A and Class 1E DC bus ElB11, initial power loss with restoration at Class 1E AC bus E1C. Offsite power is available. ACB2 Loss of power at Class 1E AC bus E1A and Class 1E DC bus E1C11, initial pcwer loss with restoration at Class 1E AC bus ElB. Offsite power is available. ACB3 Loss of power at Class 1E AC bus ElB and Class 1E DC bus E1A11, initial power loss with restoration at Class 1E AC bus E1C. Offsite power is available. ACB4 Loss of power at Class 1E AC bus ElB and Class 1E DC bus E1C11, initial power loss with restoration at Class 1E AC bus E1A. Offsite power is available. ACBS Loss of power at Class 1E AC bus E1C and Class 1E DC bus E1A11, initial power loss with restoration at Class 1E AC bus ElB. Offsite power is available. ACB6 Loss of power at Class 1E AC bus E1C and Class 1E DC bus ElB11, initial power loss with restoration at Class 1E AC bus EIA. Offsite power is available. C21 Loss of power at Class 1E DC buses E1A11 and ElB11. Offsite power is available. C22 Loss of power at Class 1E DC buses E1A11 and E1C11. Offsite power is available. C23 Loss of power at Class 1E DC buses E1811 and E1C11. Offsite power is available. BC21 Initial loss of power with restoration at Class 1E AC bus E1A, loss of power at Class 1E DC buses ElB11 and E1C11. Offsite power is available. BC22 Initial loss of power with restoration at Class 1E AC bus ElB, loss of power at Class 1E DC buses E1A11 and E1C11. Offsite power is available. BC23 Initial loss of power with restoration at Class 1E AC bus E1C, loss of power at Class 1E DC buses E1A11 and E1811. Offsite power is available. E21 Loss of offsite power and Class 1E AC buses E1A and ElB. E22 Loss of offsite power and Class 1E AC buses E1A and E1C. 0013H122884 4-30
CAUTIOle: PRELIM 6 MARY RESULTS-I gespORTAssT U8eCE ~,7A180 TIES DESCResED IN SECTIOff 2 TABLE 4-6 (continued) l Sheet 4 of 4 End State Definition Designation E23 Loss of offsite power and Class 1E AC buses ElB and E1C. DIE 1 Loss of offsite power, Class 1E DC bus E1A11, and Class 1E AC bus ElB. DIE 2 Loss of offsite power, Class 1E DC bus ElB11, and Class 1E AC bus E1A. DIE 3 Loss of offsite power, Class 1E DC bus E1A11, and Class 1E AC bus E1C. DIE 4 Loss of offsite power, Class 1E DC bus E1C11, and Class 1E AC bus E1A. DIES Loss of offsite power, Class 1E DC bus E1811, and Class 1E AC bus E1C. 01E6 Loss of offsite power, Class 1E DC bus E1C11, and Class 1E AC bus ElB. D21 Loss of offsite power and Class 1E DC buses E1A11 and E1811. D22 Loss of offsite power and Class 1E DC buses E1A11 and E1C11. D23 Loss of offsite power and Class 1E DC buses E1811 and E1C11. A2C1 Loss of Class 1E AC buses E1A and ElB, and Class 1E DC bus E1C11. Offsite power is available. A2C2 Loss of Class 1E AC buses E1A and E1C, and Class 1E DC bus E1811. Offsite power is available. A2C3 Loss of Class 1E AC buses ElB and E1C, and Class 1E DC bus E1A11. Offsite power is available. AC21 Loss of Class 1E AC bus E1A, and Class 1E DC buses E1811 and E1C11. Offsite power is available. AC22 Loss of Class 1E AC bus ElB, and Class 1E DC buses E1A11 and E1C11. Offsite power is available. AC23 Loss of Class 1E AC bus E1C, and Class 1E DC buses EIA11 and E1811. Offsite power is available. E3 Loss of all Class 1E AC power, with or without a loss of offsite power. DE21 Loss of offsite power, Class 1E DC power at bus E1A11, and Class 1E AC power at buses ElB and Elc. DE22 Loss of offsite power, Class 1E DC power at bus ElB11, and Class 1E AC power at buses E1A and E1C. DE23 Loss of offsite power, Class 1E DC power at bus E1C11, and Class 1E AC power at buses EIA and ElB. 02E1 Loss of offsite power, Class 1E DC buses E1A11 and ElB11, and Class 1E AC bus E1C. 02E2 Loss of offsite power, Class 1E DC buses E1A11 and E1C11, and Class 1E AC bus ElB. 02E3 Loss of offsite power, Class 1E DC buses E1811 and E1C11, and Class 1E AC bus E1A. D3 Loss of all Class 1E DC power, with or without a loss of offsite power. 0013H122884 4-31
TABLE 4-7a. EFFECT OF ELECTRIC POWER END STATES ON OTHER AUXILIARY SYSTEMS (GENERAL TRANSIENT) Sheet 1 of 4 Engineered E Safety Essential Component Electric SSPS Features Cooling Cooling EAB Power Water HVAC End Actuation Water Water System System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV AC ACS (a) (a) (a) (a) (a) (a) (a) (a) (a) (b) Bil (a) (a) (a) (a) (a) (a) (c) B12 (a) (a) (a) (c) B13 (a) (a) (a) (d) B21 (a) (a) (a) (a) (a) (a) (a) (a) (e) 822 (a) (a) (a) (a) (a) (a) (a) (f) 823 (a) (a) (a) (a) (a) (a) (f) All XXX (a) XXX (a) XXX (g) (g) (h) A12 XXX XXX XXX (g) (h) A13 XXX XXX XXX C11 XXX XXX (a) XXX (a) XXX (g) (g) (h) C12 XXX XXX XXX XXX (g) (h) C13 XXX XXX XXX XXX A1B1 XXX (a) (a) XXX (a) (a) XXX (a) (g) (i) A1B2 (a) XXX (a) (a) XXX (a) (a) XXX ( (i) A1B3 XXX (a) XXX (a) XXX (g) (g) a) A184 (a) XXX (a) XXX (a) (a) XXX (c) A185 XXX (a) XXX (a) XXX (a) A1B6 (a) XXX (a) XXX (a) XXX AB21 XXX (a) (a) XXX (a) (a) XXX (a) (a) AB22 (a) XXX (a) (a) XXX (a) (a) XXX (a) AB23 (a) (a) XXX (a) (a) XXX (a) (a) XXX B1C1 XXX (a) XXX (a) (a) XXX (a) (a) XXX (g) (1) B1C2 XXX XXX (a) (a) XXX (a) (a) XXX (a) (g) (i)
- a. Equipment indicated starts or restarts automatically.
- b. All EAB HVAC fans start.
- c. One previously operating EAB HVAC fan train restarts.
- d. Third EAB HVAC fan train starts. Three trains operating.
- e. Operating EAB HVAC fan trains restart.
- f. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- g. Operator action is necessary to start or restart this equipment.
- h. One EAB HVAC fan train operating.
- i. Restart of previously operating EAB HVAC fan train, operator action is necessary to start a second EAB HVAC fan train.
CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES , DESCRISED IN SECTION 2 0013H122884 4-32
TABLE 4-7a (continued) Sheet 2 of 4 Engineered Elsctric Safety Essential Component $j"1 EAB SSPS Features Cooling Cooling Water Power HVAC End Actuation Water Water System System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV B1C3 XXX (a) XXX (a) XXX (a) (a) XXX (c) 81C4 XXX XXX (a) XXX (a) XXX (g) (a) B1C5 XXX (a) XXX (a) XXX (a) XXX B1C6 XXX XXX (a) XXX (a) XXX (a) B2C1 XXX (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) 82C2 XXX (a) XXX (a) (a) XXX (a) (a) XXX (a) (a) B2C3 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (a) E11 XXX (a) (a) XXX (a) (a) XXX (a) (a) (a) E12 (a) XXX (a) (a) XXX (a) (a) XXX (a) (a) E13 (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) D11 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (a) 012 XXX (a) XXX (a) (a) XXX (a) (a) XXX (a) (a) D13 XXX (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) A21 XXX XXX (a) XXX XXX (a) XXX XXX (g) (g) A22 XXX (g) XXX XXX (g) XXX XXX (g) XXX (g) A23 XXX XXX XXX XXX XXX XX (h) A2B1 XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) A2B2 XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) A283 (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) A1C1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (g) (h) A1C2 XXX XXX XXX (a) XXX XXX (a) XXX XXX (g) (h) A1C3 XXX XXX (g) XXX XXX (g) XXX XXX (g) XXX (h) A1C4 XXX XXX (g) XXX XXX (g) XXX XXX (g) XXX (h) A1C5 XXX XXX XXX XXX XXX XXX XXX (h) A1C6 XXX XXX XXX XXX XXX XXX XXX (h) ACB1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h)
- a. Equipment indicated starts or restarts automatically.
- b. All EAB HVAC fans start.
- c. One previously operating EAB HVAC fan train restarts.
- d. Third EAB HVAC fan train starts. Three trains operating.
- c. Operating EAB HVAC fan trains restart.
- f. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- g. Operator action is necessary to start or restart this equipment.
- h. One EAB HVAC fan train operating.
- 1. Restart of previously operating EAB HVAC fan train, operator action is necessary to start a second EAB HVAC fan train.
CAUTION: PRELIMINARY RESULTS-IMPORT ANT UNCE RTAINTIES DESCRittD IN SECTION 2 0013H122884 4-33
TABLE 4-7a (continued) l 1 Sheet 3 of 4 Engineered Safety Essential Component Essential Electric SSPS Features Cooling Cooling Chilled EAB P r Water HVAC Actuation Water Water System System 3 SS EA EB EC WA WB WC CA CB CC SA SB SC EV ACB2 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) ACB3 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) ACB4 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) ACBS XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) ACB6 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) C21 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (g) (g) C22 XXX XXX XXX (g) XXX XXX (g) XXX XXX (g) XXX (g) C23 XXX XXX XXX XXX XXX XXX XXX XX (h) BC21 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) BC22 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) BC23 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) E21 XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) E22 XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) E23 (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) DIE 1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) DIE 2 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) DIE 3 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) DIE 4 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) D1ES XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) DIE 6 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) 021 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (h) D22 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (h) D23 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (h) A2C1 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX A2C2 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX A2C3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX AC21 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX AC22 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX
- a. Equipment indicated starts or restarts automatically.
- b. All EAB HVAC fans start.
- c. One previously operating EAB HVAC fan train restarts.
- d. Third EAB HVAC fan train starts. Three trains operating.
- e. Operating EAB HVAC fan trains restart.
- f. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- g. Operator action is necessary to start or restart this equipment.
- h. One EAB HVAC fan train operating.
- i. Restart of previously operating EAB HVAC fan train, operator !
action is necessary to start a second EAB HVAC fan train. i CAUTION: PRELIMINARY RESULTS-IMPORTANT ONCERTAINTIES DESCRieED IN SECTION 2 0013H122884 4-34
TABLE 4-7a (continued) Sheet 4 of 4 Engineered E n Safety Essential Component Electric SSPS Features Cooling Cooling 1l EAB Power Water HVAC Actuation Water Water System End System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV AC23 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX E3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XX DE21 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX DE22 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX DE23 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX D2E1 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX D2E2 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX D2E3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX D3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XX
- a. Equipment indicated starts or restarts automatically.
- b. All EAB HVAC fans start.
- c. One previously operating EAB HVAC fan train restarts.
- d. Third EAB HVAC fan train starts. Three trains operating.
- e. Operating EAB HVAC fan trains restart.
- f. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- g. Operator action is necessary to start or restart this equipment.
- h. One EAB HVAC fan train operating.
- i. Restart of previously operating EAB HVAC fan train, operator action is necessary to start a second EAB HVAC fan train.
CAUTION: PREL!MINARY RESULTS-IMPORTANT UNC tRTAINTitS DESCR15ED IN $tCTION 2 0013H122884 4-35
TABLE 4-7b. EFFECT OF ELECTRIC POWER END STATES ON OTHER AUXILIARY SYSTEMS (SMALL LOCA)(a) Sheet 1 of 3 Engineered Essential Safety Essential Component E ectric SSPS Features Cooling Cooling C ed p9 9 , End Actuation Water Water System System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV AC (a) (a) (a) (a) (a) (b) ACS (a) (a) (a) (a) (a) (a) (a) (a) (a) (c) Bil (a) (a) (a) (a) (a) (a) (a) (a) (a) (b) 012 (a) (a) (a) (a) (a) (a) (b) B13 (a) (a) (a) (a) (a) (a) (b) 821 (a) (a) (a) (a) (a) (a) (a) (a) (a) (d) B22 (a) (a) (a) (a) (a) (a) (a) (a) (a) (d) B23 (a) (a) (a) (a) (a) (a) (d) All XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) A12 XXX (a) XXX (a) XXX (a) (e) A13 (a) XXX (a) XXX (a) XXX (e) C11 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) C12 XXX XXX (a) XXX (a) XXX (a) (e) C13 XXX (a) XXX (a) XXX (a) XXX (e) A181 XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) A182 (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) A183 XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) A184 (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) A185 XXX (a) XXX (a) XXX (a) (e) A186 (a) XXX (a) XXX (a) XXX (e) AB21 XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) AB22 (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) AB23 (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) B1C1 XXX (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) BIC2 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) 01C3 XXX (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) B1C4 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) 01C5 XXX (a) XXX (a) XXX (a) XXX (e)
- a. All standby equipment starts or restarts automatically.
- b. All EAD HVAC fans start.
- c. Third EAB HVAC fan train starts. Three trains operating.
- d. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- e. Two EAB HVAC fan trains operating.
- f. One EAB HVAC fan train operating.
CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 0013H122884 4-36
TABLE 4-7b (continued)(a) Sheet 2 of 3 Engineered E Safety Essential Component Electric SSPS Features Cooling Cooling 1 EAB P r Actuation Water Water Sy t System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV B1C6 XXX XXX (a) XXX (a) XXX (a) (e) B2C1 XXX (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) B2C2 XXX (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) B2C3 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) E11 XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) E12 (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) E13 (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) D11 XXX XXX (a) (a) XXX (a) (a) XXX (a) (a) (e) D12 XXX (a) XXX (a) (a) XXX (a) (a) XXX (a) (e) D13 XXX (a) (a) XXX (a) (a) XXX (a) (a) XXX (e) A21 XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) A22 XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) A23 XXX XXX XXX XXX XXX XXX (f) A2B1 XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) A2B2 XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) A2B3 (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) A1C1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) A1C2 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) A1C3 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) A1C4 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) A1C5 XXX XXX XXX XXX XXX XXX XXX (f) A1C6 XXX XXX XXX XXX XXX XXX XXX (f) ACB1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) ACB2 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) ACB3 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) ( f) ACB4 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) ACBS XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) ACB6 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) C21 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) C22 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) C23 XXX XXX XXX XXX XXX XXX XXX XXX (f) BC21 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f)
- a. All standby equipment starts or restarts automatically,
- b. All EAB HVAC fans start.
- c. Third EAB HVAC fan train starts. Three trains operating.
- d. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- e. Two EAB HVAC fan trains operating.
- f. One EAB HVAC fan train operating. C4urioN: ensuwNAny nesutts.
IMPORTANT UNCERfAINTIES DESCRISED IN stCTION 2 0013H122884 4-37
TABLE 4-7b (continued)(a) 1 Sheet 3 of 3 Engineered Essential Safety Essential Component Chilled Electric SSPS Features Cooling Cooling EAB Pwer Actuation Water Water Water HVAC End System System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV BC22 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) BC23 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) E21 XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) E22 XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) E23 (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) DIE 1 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) DIE 2 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) DIE 3 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) DIE 4 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) DIES XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) 01E6 XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) D21 XXX XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) (f) D22 XXX XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX (f) 023 XXX XXX (a) XXX XXX (a) XXX XXX (a) XXX XXX (f) A2C1 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X A2C2 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X A2C3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X AC21 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X AC22 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X AC23 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X E3 XXX XXX XXX XXX XXX XXX XXX XXX XXX X DE21 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X DE22 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X DE23 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X D2E1 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X D2E2 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X : D2E3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X D3 XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX XXX X
- a. All standby equipment starts or restarts automatically,
- b. All EAB HVAC fans start.
- c. Third EAB HVAC fan train starts. Three trains operating.
- d. Previously operating EAB HVAC fan train restarts, third EAB HVAC fan train starts. Three trains operating.
- e. Two EAB HVAC fan trains operating.
- f. One EAB HVAC fan train operating.
CAUTION: PRELIMINARY RESULTS- ) IMPORTANT UNCERTAINTit$ ! DESCRISED IN StCTION 2 0013H122884 4-38
TABLE 4-7c. EFFECT OF ELECTRIC (LOSS OF 0FFSITEPOWER POWER) a) END STATES (ON OTHER AUX Engineered Essential Safety Essential Component Chilled Electric SSPS Features Cooling Cooling EAB l Power Water HVAC End Actuation Water Water System System State SS EA EB EC WA WB WC CA CB CC SA SB SC EV ACS (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) E11 XX (a) (a) XX (a) (a) XX (a) (a) (b) E12 (a) XX (a) (a) XX (a) (a) XX (a) (b) E13 (a)-(a) XX (a) (a) XX (a) (a) XX (b) 011 XX XX (a) (a) XX (a) (a) XX (a) (a) (b) D12 XX (a) XX (a) (a) XX (a) (a) XX (a) (b) D13 XX (a) (a) XX (a) (a) XX (a) (a) XX (b) E21 XX XX (a) XX XX (a) XX XX (a) (c) E22 XX (a) XX XX (a) XX XX (a) XX (c) E23 (a) XX XX (a) XX XX (a) XX XX (c) DIE 1 XX XX XX (a) XX XX (a) XX XX (a) (c) DIE 2 XX XX XX (a) XX XX (a) XX XX (a) (c) DIE 3 XX XX (a) XX XX (a) XX XX (a) XX (c) 01E4 XX XX (a) XX XX (a) XX XX (a) XX (c) DIES XX (a) XX XX (a) XX XX (a) XX XX (c) D1E6 XX (a) XX XX (a) XX XX (a) XX XX (c) 021 XX XX XX XX (a) XX XX (a) XX XX (a) (c) D22 XX XX XX (a) XX XX (a) XX XX (a) XX (c) D23 XX XX (a) XX XX (a) XX XX (a) XX XX (c) E3 XX XX XX XX XX XX XX XX XX XX DE21 XX XX XX XX XX XX XX XX XX XX XX DE22 XX XX XX XX XX XX XX XX XX XX XX DE23 XX XX XX XX XX XX XX XX XX XX XX D2E1 XX XX XX XX XX XX XX XX XX XX XX XX D2E2 XX XX XX XX XX XX XX XX XX XX XX XX D2E3 XX XX XX XX XX XX XX XX XX XX XX XX D3 XX XX XX XX XX XX XX XX XX XX XX XX XX XX
- a. All equipment starts or restarts automatically.
- b. Two EAB HVAC fan trains operating.
- c. One EAB HVAC fan train operating.
CAUTION: PRELIMINARY RESULTS-IMPORTANT ONCERTAINTIES DESCR' SED IN SECT 10N 2 0013H122834 4-39
TABLE 4-8. AUXILIARY TREE END STATES Sheet 1 of 2 Description St e A Loss of one train of the engineered safety features actuation system. B Loss of two ESFAS trains. C Loss of three ESFAS trains. D(a) Loss of one essential cooling water train. E Loss of one ECW train [not the same as ESFAS train (s)]. F(b) Loss of two ECW trains. G Loss of two ECW trains [not the same trains as ESFAS train (s)]. H(c) Loss of three ECW trains. I(d) Loss of one ECW train after loss of offsite power or loss of power at the associated bus. J(d) Loss of one ECW train after LOSP [not the same train as ESFAS train (s)]. K(d) Loss of two ECW trains after LOSP. L(d) Loss of two ECW trains after LOSP [not the same trains as ESFAS train (s)]. M Loss of three ECW trains after LOSP. N Loss of one component cooling train. O Loss of two CCW trains. P Loss of three CCW trains. Q Loss of one CCW train (not the same train as ESFAS train (s)]. R Loss of two CCW trains (not the same trains as previous systems train failures). S Loss of the solid state protection system output. T(e) Loss of one essential chilled water system train. U Loss of two ECH trains. V Loss of three ECH trains. W Loss of one ECH train (not the same train as previous systems train failures),
- a. Loss of one ECW train fails the associated CCW train.
- b. Loss of two ECW trains fails the associated CCW trains and one associated ECH train.
- c. Loss of three ECW trains fails all CCW and ECH trains.
- d. Loss of an ECW train with LOSP fails the associated emergency diesel generator resulting in no power to the associated 4160V AC bus.
- e. The end states for ECH are based upon the following initial conditions:
(1) ECW train A operating; ECW train C in standby (2) ECH trains A and B are operating (both trains can be supplied from either ECW train A or train B) (3) ECH train C can only be suppied from ECW train C. CAuflON: PRELIMINARY RESULTS-IMPORTANT UNCENTAINTIES DE$CRISED IN SECTION 2 0013H122884
TABLE 4-8 (continued) Sheet 2 of 2 Description St e X Loss of two ECH trains [not the same trains as previous systems train failure (s)]. Y Loss of one ECH train (different train from preceding systems failures). Z Loss of two ECH trains (different trains from preceding systems failures). 1 Loss of one ECW train after AC power is lost at the associated bus and loss of a second ECW train. Not for LOSP. 2 Loss of one ECW train after AC power is lost at the associated bus and loss of the remaining ECW trains. Not for LOSP. 3 Loss of one ECW train after AC power is lost at the associated bus (not the same train as previous systems train failures) and loss of a second ECW train. Not for LOSP. 4 Loss of one ECW train after AC power is lost at the associated bus and loss of a second ECW train [not the same trains as previous systems train failures]. Not for LOSP. 5 Loss of one ECW train after AC power is lost at the associated bus (not the same train as previous systems train failures) and loss of the remaining ECW trains. Not for LOSP. 6 Loss of two ECW trains after AC power is lost at the associated buses and loss of the third ECW train. Not for LOSP. 7 Loss of two ECW trains after AC power is lost at the associated buses [not the same trains as previous systems train failures] and loss of the third ECW train. Not for LOSP. 9 Loss of the electrical auxiliary building ventilation system (long-term failure of the electric power systems).
- a. Loss of one ECW train fails the associated CCW train.
- b. Loss of two ECW trains fails the associated CCW trains and one associated ECH train.
- c. Loss of three ECW trains fails all CCW and ECH trains.
- d. Loss of an ECW train with LOSP fails the associated emergency diesel generator resulting in no power to the associated 4160V AC bus.
- e. The end states for ECH are based upon the following initial conditions:
(1) ECW train A operating; ECW train C in standby (2) ECH trains A and B are operating (both trains can be supplied from either ECW train A or train B) (3) ECH train C can only be suppied from ECW train C. CAUTION,. PRELIMINARY RESULTS. IMPORT ANT UNCERTAINTIES DESCRIBED IN StCTION 2 0013H122884 1
cauttose: PattmousAmy ptsuLis-meronfAsef uesctOTAssists DESCRetEO ffe sECTIOes a TABLE 4-9. AUXILIARY TREE END STATES 1 Sheet 1 of 14 End Auxiliary Systems and Numbers I States of Trains Affected AUX All auxiliaries available. T 1-ECH U 2-ECH V 3-ECH N 1-CCW l NT 1-CCW 1-ECH CCW, ECH same train. NW 1-CCW 1-ECH CCW, ECH different trains. NU 1-CCW 2-ECH CCW, one train ECH same. NX 1-CCW 2-ECH CCW, ECH different trains. NY 1-CCW 3-ECH 0 2-CCW OT 2-CCW 1-ECH One train CCW, ECH same. OW 2-CCW 1-ECH CCW, ECH different trains. 00 2-CCW 2-ECH CCW, ECH same trains. OX 2-CCW 2-ECH One train CCW one train ECH different. OV 2-CCW 3-ECH P 3-CCW PT 3-CCW 1.ECH PU 3-CCW 2-ECH PV 3-CCW 3-ECH D 1-ECW 1-CCW DT 1-ECW 1 :CW 1-ECH ECW, ECH same train. DW 1-ECW 1-CCW 1-ECH ECW, ECH different trains. DU 1-ECW 1-CCW 2-ECH ECW, one train ECH same. DX 1-ECW 1-CCW 2-ECH ECW, ECH different trains. DV 1-ECW 1-CCW 3-ECH DO 1-ECW 2-CCW 00T 1-ECW 2-CCW 1-ECH ECW, ECH same train. DOW 1-ECW 2-CCW 1-ECH ECW, ECH different trains; one train CCW, ECH same. 00Y 1-ECW 2-CCW 1-ECH CCW, ECH different trains. 000 1-ECW Z-CCW 2-ECH ECW, ECH same trains. 00X 1-ECW 2-CCW 2-ECH One train CCW, one train ECH dif ferent. DOZ 1-ECW 2-CCW 2-ECH ECW, ECH different trains. DOV 1-ECW 2-CCW 3-ECH DP 1-ECW 3-CCW DPT 1-ECW 3-CCW 1-ECH ECW, ECH same train. DPW 1-ECW 3-CCW 1-ECH ECW, ECH different trains. DPU 1-ECW 3-CCW 2-ECH ECW, one train ECH same. DPX 1-ECW 3-CCW 2-ECH ECW, ECH different tra:ns. DPY 1-ECW 3-CCW 3-ECH I 1-ECW 1-CCW 1-ECH (after LOSP) IU 1-ECW 1-CCW 2-ECH (after LOSP) IV 1-ECW 1-CCW 3-ECH (after LOSP) 4-42 0013H120684
CAUT9000: PRtLlassetARY RESULT 3 stEPORYAasTuseCE TAsseTatS Ot9CRl980 las SECTIOct 2 TABLE 4-9 (continued) Sheet 2 of 14 End Auxiliary Systems and Numbers States of Trains Affected 10 1-ECW 2-CCW 1-ECH af ter LOSP) IOU 1-ECW 2-CCW 2-ECH after LOSP) CCW, ECH same trains. 10X 1-ECW 2-CCW 2-ECH af ter LOSP) One train CCW, one train ECH different. 10V l-ECW 2-CCW 3-ECH (af ter LOSP) IP l-ECW 3-CCW 1-ECH (after LOSP IPU 1-ECW 3-CCW 2-ECH (af ter LOSP IPV l-ECW 3-CCW 3-ECH (after LOSP F 2-ECW 2-CCW 1-ECH FU 2-ECW 2-CCW 2-ECH ECW, ECH same trains. FX 2-ECW 2-CCW 2-ECH One train ECW, one train ECH different. FY 2-ECW 2-CCW 3-ECH FP 2-ECW 3-CCW 1-ECH FFU 2-ECW 3-CCW 2-ECH ECW, ECH same trains. FPX 2-ECW 3-CCW 2-ECH One train ECW, one train ECH different. FPV 2-ECW 3-CCW 3-ECH K 2-ECW 2-CCW 2-ECH (after LOSP) KV 2-ECW 2-CCW 3-ECH (af ter LOSP) KP 2-ECW 2-CCW 2-ECH (after LOSP) KPV 2-ECW 2-CCW 3-ECH (af ter LOSP) 1 2-ECW 2-CCW 1-ECH (af ter LOP one bus) Loss of AC power at one bus. IU 2-ECW 2-CCW 2-ECH (af ter LOP one bus) ECW, ECH same trains. IX 2-ECW 2-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. IV 2-ECW 2-CCW 3-ECH (after LOP one bus) IP 2-ECW 3-CCW 1-ECH (after LOP one bus) I PU 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. IPX 2-ECW 3-CCW 2-ECH (after LDP one bus) One train ECW, one train ECH different. IPV 2-ECW 3-CCW 3-ECH (af ter LOP one bus) 1: 3-ECW 3-CCW 3-ECH M 3-ECW 3-CCW 3-ECH (af ter LOSP) 2 3-ECW 3-CCW 3-ECH (after LOP one bus) Loss of AC power at one bus only. 6 3-ECW 3-CCW 3-ECH (after LDP 2-buses) Loss of AC power at 1 two buses only. 9 Electrical Auxiliary Building Yentilation A 1-ESFAS AT l-ESFAS 1-ECH ESFAS, ECH same train. AW 1-ESFAS 1-ECH ESFAS ECH different trains. AU 1-ESFAS 2-ECH ESFAS, one train ECH same. AX l-ESFAS 2-ECH ESFAS, ECH different trains. AV l-ESFAS 3-ECH AN 1-ESFAS 1-CCW ESFAS, CCW same train. ANT 1-ESFAS 1-CCW 1-ECH ESFAS, ECH same train. ANW 1-ESFAS 1-CCW 1-ECH ESFAS, ECH different trains. ANU 1-ESFAS 1-CCW 2-ECH ESFAS, one train ECH same. ANX l-ESFAS 1-CCW 2-ECH ESFAS ECH different trains. ANY l-ESFAS 1-CCW 3-ECH AQ 1-ESFAS 1-CCW ESFAS, CCW different trains. AQT l-ESFAS 1-CCW 1-ECH ESFAS, ECH same train. 0013H120384 4-43
CAuflON: PRELitetNARY RESULTS 14APOR7 ANT Utsct1Tf.INTIES OSSCRISEO IN SECTIOft 2 TABLE 4-9 (continued) Sheet 3 of 14 End Auxiliary Systems and Numbers States of Trains Affected AQW 1-ESFAS 1-CCW 1-ECH CCW, ECH same train. AQY l-ESFAS 1-CCW 1-ECH ESFAS, CCW, ECH different trains. AQU 1-ESFAS 1-CCW 2-ECH ESFAS, one train ECH same; CCW, other train ECH same. AQX 1-ESFAS 1-CCW 2-ECH ESFAS, one train ECH same; CCW, other train ECH different. AQZ 1-ESFAS 1-CCW 2-ECH ESFAS, ECH different trains. AQV l-ESFAS l-CCW 3-ECH A0 1-ESFAS 2-CCW ESFAS, one train CCW same. A0T 1-ESFAS 2-CCW 1-ECH ESFAS, ECH same train. A0W 1-ESFAS 2-CCW 1-ECH ESFAS, ECH different trains; one train CCW, ECH same. A0Y l-ESFAS 2-CCW 1-ECH CCW EC trains.Hdifferent AOU 1-ESFAS 2-CCW 2-ECH CCW, ECH same trains. A0X l-ESFAS 2-CCW 2-ECH ESFAS, one train ECH same; CCW, ECH different trains. A0Z 1-ESFAS 2-CCW 2-ECH ESFAS, ECH different trains. A0V l-ESFAS 2-CCW 3-ECH AR l-ESFAS 2-CCW ESFAS, CCW different trains. ART 1-ESFAS 2-CCW 1-ECH ESFAS, ECH same train. ARW 1-ESFAS 2-CCW 1-ECH ESFAS, ECH different trains. ARU 1-ESFAS 2-CCW 2-ECH ESFAS, one train ECH same. ARX l-ESFAS 2-CCW 2-ECH ESFAS, ECH different trains. ARY l-ESFAS 2-CCW 3-ECH AP l-ESFAS 3-CCW APT l-ESFAS 3-CCW 1-ECH ESFAS, ECH same train. APW 1-ESFAS 3-CCW 1-ECH ESFAS, ECH different trains. APU 1-ESFAS 3-CCW 2-ECH ESFAS, one train ECH same. APX l-ESFAS 3-CCW 2-ECH ESFAS, ECH different trains. APY l-ESFAS 3-CCW 3-ECH AD l-ESFAS 1-ECW 1-CCW ESFAS, ECW same train. ADT 1-ESFAS 1-ECW 1-CCW 1-ECH ESFAS ECH same train. ADW 1-ESFAS 1-ECW 1-CCW 1-ECH ESFAS, ECH dif ferent trains. ADU 1-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, one train ECH same. ADX l-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, ECH different trains. ADV l-ESFAS 1-ECW 1-CCW 3-ECH ADO l-ESFAS 1-ECW 2-CCW ADOT l-ETFAS 1-ECW 2-CCW 1-ECH ESFAS, ECH same train. ADOW 1-ESEAS 1-ECW 2-CCW 1-ECH ESFAS, ECH different trains; one train CCW, ECH same. 0013H120384 4-44
CAUTIOft: PRELie008sARY RESULTS-leAPORTANT UOsCe27A8NTIE$ DESCReetO IN SECTIOff 2 TABLE 4-9(continued) Sheet 4 of 14 End Auxiliary Systems and NLambers N tes States of Trains Affected AD0Y l-ESFAS 1-ECW 2-CCW 1-ECH CCW EC trains.Hdifferent ADOU 1-ESFAS 1-ECW 2-CCW 2-ECH CCW ECH same trains. ADOX l-ESFAS 1-ECW 2-CCW 2-ECE ESFIS,onetrainECH same; one train CCW, one train ECH different. ADOZ 1-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH different trains. ADOV l-ESFAS 1-ECW 2-CCW 3-ECH ADP l-ESFAS 1-ECW 3-CCW ADPT 1-ESFAS 1-ECW 3-CCW 1-TCH ESFAS, ECH same train. ADPW 1-ESFAS 1-ECW 3-CCW 1-ECH ESFAS, ECH different trains. ADPU 1-E>FAS 1-ECW 3-CCW 2-ECH ESFAS, one train ECH same. ADPX l-ESFAS 1-ECW 3-CCW 2-ECH ESFAS, ECH different trains. ADPY 1-ESFAS 1-ECW 3-CCW 3-ECH AE 1-ESFAS 1-ECW 1-CCW ESFAS, ECW different trains. AET 1-ESFAS 1-ECW 1-CCW 1-ECH ESFAS, ECH same train. AEW 1-ESFAS 1-ECW 1-CCW 1-ECH ECW, ECH same train. AEY l-ESFAS 1-ECW 1-CCW 1-ECH ESFAS, ECW, ECH different trains. AEU 1-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, one train ECH same; ECW, other train
. ECH same.
AEX l-ESFAS 1-ECW 1-CCW 2-ECH ECW EC trains.Hdifferent AEZ 1-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, ECH different trains. AEV 1-ESFAS 1-ECW 1-CCW 3-ECH AE0 1-ESFAS 1-ECW 2-CCW ESFAS, one train CCW same. AE0T 1-ESFAS 1-ECW 2-CCW 1-ECH ESFAS ECH same train. AE0W 1-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH same train. AE0Y 1-ESFAS 1-ECW 2-CCW 1-ECH CCW EC trains.Hdifferent AEOU 1-ESFAS 1-ECW 2-CCW 2-ECH CCW, ECH same trains. AE0X l-ESFAS 1-ECW 2-CCW 2-ECH ECW EC trains.Hdifferent AE0Z 1-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH different trains. AE0V 1-ESFAS l-ECW 2-CCW 3-ECH AER l-ESFAS 1-ECW 2-CCW ESFAS, CCW different trains. AERT l-ESFAS 1-ECW 2-CCW 1-ECH ESFAS, ECH same train. AERW 1-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH same train. AERY l-ESFAS 1-ECW 2-CCW 1-ECH ESFAS, ECW, ECH
"., different trains.
AERU 1-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, one train ECH same; ECW, other train l ECH same. AERX 1-ESFAS 1-ECW 2-CCW 2-ECH ECW EC trains.Hdifferent AERZ 1-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH different l trains. AERV 1-ESFAS 1-ECW 2-CCW 3-ECH AEP l-ESFAS 1-ECW 3-CCW AEPT l-ESFAS l-ECW 3-CCW 1-ECH ESFAS, ECH same train. l 0013H120384 4-45 b
CAUTION: PRE LIMINA *.y CEgULT3 6MPORTANT UNCE";TAINTitt DESCRISED IN SECTION 2 TABLE 4-9 (continued) Sheet 5 of 14 End Auxiliary Systems and Numbers States of Trains Affected l AEPW 1-ESFAS 1-ECW 3-CCW 1-ECH ECW, ECH same train. l AEPY l-ESFAS 1-ECW 3-CCW 1-ECH ESFAS, ECW, ECH different trains. AEPU 1-ESFAS 1-ECW 3-CCW 2-ECH ESFAS, one train ECH same; ECW, other train ECH same. AEPX l-ESFAS 1-ECW 3-CCW 2-ECH ECW, ECH different trains. AEPZ 1-ESFAS 1-ECW 3-CCW 2-ECH ESFAS, ECH different trains. AEPV l-ESFAS 1-ECW 3-CCW 3-ECH AI l-ESFAS 1-ECW 1-CCW 1-ECH (after LOSP) ESFAS, ECW same train. AIU 1-ESFAS 1-ECW 1-CCW 2-ECH (after LOSP) AIV l-ESFAS 1-ECW 1-CCW 3-ECH (after LOSP) AIO l-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) AIOU 1-ESFAS 1-ECW 2-CCW 2-ECH (af ter LOSP) CCW, ECH same trains. AIOX l-ESFAS 1-ECW 2-CCW 2-ECH (af ter LOSP) One train CCW, one train ECH different. AIOV l-ESFAS 1-ECW 2-CCW 3-ECH (after LOSP) AIP l-ESFAS 1-ECW 3-CCW 1-ECH (after LOSP) AIPU 1-ESFAS 1-ECW 3-CCW 2-ECH (after LOSP) AIPV l-ESFAS 1-ECW 3-CCW 3-ECH (after LOSP) AJ 1-ESFAS 1-ECW 1-CCW 1-ECH (after LOSP) ESFAS, ECW different trains. AJU 1-ESFAS 1-ECW 1-CCW 2-ECH (af ter LOSP) ESFAS, one train ECH same. AJX l-ESFAS 1-ECW 1-CCW 2-ECH (after LOSP) ESFAS, ECH different trains. AJV l-ESFAS 1-ECW 1-CCW 3-ECH (after LOSP) AJO l-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) ESFAS, one train CCW same. AJOU 1-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) CCW, ECH same trains AJ0X l-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) ESFAS, ECH different trains. AJ0Y l-ESFAS 1-ECW 2-CCW 3-ECH (af ter LOSP) AJR l-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) ESFAS, CCW different trains. AJRU 1-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) ESFAS, one train ECH same. AJRX l-ESFAS 1-ECW ?-CCW 2-ECH (after LOSP) ESFAS, ECH different trains. AJRV l-ESFAS 1-ECW 2-CCW 3-ECH (after LOSP) AJP l-ESFAS 1-ECW 3-CCW 1-ECH (after LOSP) AJPU 1-ESFAS 1-ECW 3-CCW 2-ECH (af ter LOSP) ESFAS, one train ECH same. AJPX l-ESFAS 1-ECW 3-CCW 2-ECH (after LOSP) ESFAS, ECH different trains. AJPV l-ESFAS 1-ECW 3-CCW 3-ECH (after LOSP) AFT l-ESFAS 2-ECW 2-CCW 1-ECH ESFAS, one train ECW same; ESFAS, ECH same train. AFW 1-ESFAS 2-ECW 2-CCW 1-ECH ESFAS, ECH different trains. AFU 1-ESFAS 2-ECW 2-CC'i 2-ECH ECW, ECH same trains. AFX l-ESFAS 2-ECW 2-CCW 2-ECH ESFAS, one train ECH same. l AFV 1-ESFAS 2-ECW 2-CCW 3-ECH l AFPT l-ESFAS 2-ECW 3-CCW 1-ECH l AFPW 1-ESFAS 2-ECW 3-CCW 1-ECH ESFAS, ECH different I trains. AFPU 1-ESFAS 2-ECW 3-CCW 2-ECH l i 0013H120384 4-46 t
CAUTION: PRE LIMINAAY C ESULTS. IMPORTANT UNCEZTAINTIES oEscaisEo IN sEctioN l TABLE 4-9(continued) Sheet 6 of 14 End Auxiliary Systems and Numbers , States of Trains Affected AFPX l-ESFAS 2-ECW 3-CCW 2-ECH ESFAS, one train ECH same. AFPV l-ESFAS 2-ECW 3-CCW 3-ECH AG l-ESFAS 2-ECW 2-CCW 1-ECH ESFAS, ECW different trains. AGU 1-ESFAS 2-ECW 2-CCW 2-ECH ESFAS, one train ECH same. AGX l-ESFAS 2-ECW 2-CCW 2-ECH ESFAS, ECH different AGV 1-ESFAS 2-ECW 2-CCW 3-ECH AGP l-ESFAS 2-ECW 3-CCW 1-ECH ESFAS, ECH different trains. AGPU 1-ESFAS 2-ECW 3-CCW 2-ECH ESFAS, one train ECH same. AGPX l-ESFAS 2-ECW 3-CCW 2-ECH ESFAS, ECH different trains. AGPV l-ESFAS 2-ECW 3-CCW 3-ECH AK 1-ESFAS 2-ECW 2-CCW 2-ECH (after LOSP) ESFAS, one train ECW same. AKV l-ESFAS 2-ECW 2-CCW 3-ECH (af ter LOSP) AKP l-ESFAS 2-ECW 3-CCW 2-ECH (after LOSP) AKPY l-ESFAS 2-ECW 3-CCW 3-FCH (after LOSP) AL l-ESFAS 2-ECW 2-CCW 2-ECH (after LOSP) ESFAS, ECW different trains, ALV l-ESFAS 2-ECW 2-CCW 3-ECH (after LOSP) ALP l-ESFAS 2-ECW 3-CCW 2-ECH (after LOSP) ALPV 1-ESFAS 2-ECW 3-CCW 3-ECH (after LOSP) Al 1-ESFAS 2-ECW 2-CCW 1-ECH (after LOP one bus) ESFAS, ECW(LOP) same train. AlU 1-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. AIX l-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. AIV l-ESFAS 2-ECW 2-CCW 3-ECH (after LOP one bus) AIP l-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) AlPU 1-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. AlPX l-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. AIPY l-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) A3 1-ESFAS 2-ECW 2-CCW 1-ECH (efter LOP one bus) ESFAS,ECW(LOP) different trains. A3U 1-ESFAS 2-ECW 2-CCW 2-ECH (atter LOP one bus) ECW, ECH same trains. A3X l-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ESFAS, ECH different trains. A3Y l-ESFAS 2-ECW 2-CCW 3-ECW (after LOP one bus) A3P l-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) A3PU 1-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. A3PX l-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ESFAS, ECH different trains. A3PV l-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) A4 1-ESFAS 2-ECW 2-CCW 1-ECH (after LOP one bus) ESFAS, ECW different trains. A4U 1-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ESFAS, one train ECH same. A4X l-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ESFAS, ECH different trains. A4V l-ESFAS 2-ECW 2-CCW 3-ECH (after LOP one bus) A4P l-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) A4PU 1-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ESFAS, one train ECH same. A4PX l-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ESFAS, ECH different trains. 0013H120384 4-47
CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIE5 oEScRIsto :N SECTION 2 TABLE 4-9 (continued) Sheet 7 of 14 End Auxiliary Systems and Numbers States of Trains Affected A4PV l-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) AH 1-ESFAS 3-ECW 3-CCW 3-ECH l AM l-ESFAS 3-ECW 3-CCW 3-ECH (after LOSP) A2 1-ESFAS 3-ECW 3-CCW 3-ECH (after LOP one bus) ESFAS, ECW(LOP) same train. A5 1-ESFAS 3-ECW 3-CCW 3-ECH (after LOP one bus) ESFAS, ECW(LOP) different trains. A6 1-ESFAS 3-ECW 3-CCW 3-ECH (after LOP 2-buses) ESFAS, ECW(LOP) same train. A7 1-ESFAS 3-ECW 3-CCW 3-ECH (after LOP 2-buses) ESFAS both trains ECW(LOP) different. B 2-ESFAS BT 2-ESFAS 1-ECH One train ESFAS, ECH same. BW 2-ESFAS 1-ECH ESFAS, ECH different trains. BU 2-ESFAS 2-ECH ESFAS, ECH same trains. BX 2-ESFAS 2-ECH One train ESFAS, one train ECH different. BV 2-ESFAS 3-ECH BN 2-ESFAS 1-CCW One train ESFAS, CCW same. BNT 2-ESFAS 1-CCW 1-ECH CCW, ECH same trcin. BNW 2-ESFAS 1-CCW 1-ECH One train ESFAS, ECH same; CCW, ECH different trains. BNY 2-ESFAS 1-CCW 1-ECH ESFAS, ECH different trains. BNU 2-ESFAS 1-CCW 2-ECH ESFAS, ECH same trains. BNX 2-ESFAS 1-CCW 2-ECH One train ESFAS, one train ECH different. BNZ 2-ESFAS 1-CCW 2-ECH CCW, ECH different trains. BNY 2-ESFAS 1-CCW 3-ECH BQ 2-ESFAS 1-CCW ESFAS, CCW different trains. BQT 2-ESFAS 1-CCW 1-ECH One train ESFAS, ECH same. BQW 2-ESFAS 1-CCW 1-ECH ESFAS, ECH different trains. BQU 2-ESFAS 1-CCW 2-ECH ESFAS, ECH same trains. BQX 2-ESFAS 1-CCW 2-ECH ESFAS, one train ECH same; CCW, other train ECH same. BQV 2-ESFAS 1-CCW 3-ECH B0 2-ESFAS 2-CCW ESFAS, CCW same trains. 80T 2-ESFAS 2-CCW 1-ECH One train ESFAS, ECH same. BOW 2-ESFAS 2-CCW 1-ECH ESFAS, ECH different trains. BOU 2-ESFAS 2-CCW 2-ECH ESFAS, ECH same trains. BOX 2-ESFAS 2-CCW 2-ECH One train ESFAS, one train ECH different. B0V 2-ESFAS 2-CCW 3-ECH BR 2-ESFAS 2-CCW One train ESFAS, one train CCW different. BRT 2-ESFAS 2-CCW 1-ECH One train ESFAS, one train CCW, ECH same. BRW 2-ESFAS 2-CCW 1-ECH CCW, ECH different trains. BRY 2-ESFAS 2-CCW 1-ECH ESFAS, ECH different trains. 0013H120384 4-48
CAUTION: PRELitteNARY f ESULTS. IMPORTA817 UNCE27AINTIES DESCRISE3 IN SECTION 2 TABLE 4-9 (continued) Sheet 8 of 14 End Auxiliary Systems and Numbers States of Trains Affected BRU 2-ESFAS 2-CCW 2-ECH ESFAS, ECH same trains. BRX 2-ESFAS 2-CCW 2-ECH CCW, ECH same trains. BRZ 2-ESFAS 2-CCW 2-ECH One train ESFAS, one train ECH different. BRV 2-ESFAS 2-CCW 3-ECH BP 2-ESFAS 3-CCW BPT 2-ESFAS 3-CCW 1-ECH One train ESFAS, ECH same. BPW 2-ESFAS 3-CCW 1-ECH ESFAS, ECH different trains. BPU 2-ESFAS 3-CCW 2-ECH ESFAS, ECH same trains. BPX 2-ESFAS 3-CCW 2-ECH One train ESFAS, one train ECH different. BPV 2-ESFAS 3-CCW 3-ECH BD 2-ESFAS 1-ECW 1-CCW One train ESFAS, ECW same. BDT 2-ESFAS 1-ECW 1-CCW 1-ECH ECW, ECH same train. BDW 2-ESFAS 1-ECW 1-CCW 1-ECH One train ESFAS, ECH same; ECW, ECH different trains. BDY 2-ESFAS 1-ECW 1-CCW 1-ECH ESFAS, ECH different trains. BDU 2-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, ECH same trains. BDX 2-ESFAS 1-ECW 1-CCW 2-ECH One train ESFAS, one train ECH different. BDZ 2-ESFAS 1-ECW 1-CCW 2-ECH ECW EC trains.Hdifferent BDY 2-ESFAS 1-ECW 1-CCW 3-ECH BDO 2-ESFAS 1-ECW 2-CCW ESFAS, CCW same trains. BDDT 2-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH same train. BDOW 2-ESFAS 1-ECW 2-CCW 1-ECH Dne train ESFAS, ECH same; ECW, ECH different trains. 800Y 2-ESFAS 1-ECW 2-CCW 1-ECH ESFAS, ECH different trains. BDOU 2-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH same trains. BD0X 2-ESFAS 1-ECW 2-CCW 2-ECH Dne train ESFAS, one train ECH different. B00Z 2-ESFAS 1-ECW 2-CCW 2-ECH ECW, ECH different trains. BDDV 2-ESFAS l-ECW 2-CCW 3-ECH BDR 2-ESFAS 1-ECW 2-CCW One train ESFAS, one train CCW different. BDRT 2-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH same train. BDRW 2-ESFAS 1-ECW 2-CCW 1-ECH CCW, ECH different trains. BDRY 2-ESFAS 1-ECW 2-CCW 1-ECH ESFAS, ECH different trains. BDRU 2-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH same trains. BDRX 2-ESFAS 1-ECW 2-CCW 2-ECH CCW, ECH same trains. - BDRZ 2-ESFAS 1-ECW 2-CCW 2-ECH ECW, ECH different trains. BDRV 2-ESFAS 1-ECW 2-CCW 3-ECH BDP 2-ESFAS 1-ECW 3-CCW BDPT 2-ESFAS 1-ECW 3-CCW 1-ECH ECW, ECH same train. BDPW 2-ESFAS 1-ECW 3-CCW 1-ECH Cne train ESFAS, ECH same; ECW, ECH different trains. BDPY 2-ESFAS 1-ECW 3-CCW 1-ECH ESFAS, ECH different trains. BDPU 2-ESFAS 1-ECW 3-CCW 2-ECH ESFAS, ECH same trains. 0013H122884 4-49
I CAUTKMf: PRELinNNARY EESULTF lAAPORTANT UNCEITAINTIES DESCRISED IN SECTION 2 TABLE 4-9 (continued) Sheet 9 of 14 End Auxiliary Systems and Numbers N es States of Trains Affected BDPX 2-ESFAS 1-ECW 3-CCW 2-ECH One train ESFAS, one train ECH different. BDPZ 2-ESFAS 1-ECW 3-CCW 2-ECH ECW EC trains.Hdifferent BDPV 2-ESFAS 1-ECW 3-CCW 3-ECH BE 2-ESFAS 1-ECW 1-CCW ESFAS, ECW different trains. BET 2-ESFAS 1-ECW 1-CCW 1-ECH One train ESFAS, ECH same. BEW 2-ESFAS 1-ECW 1-CCW 1-ECH ESFAS, ECH different trains. BEU 2-ESFAS 1-ECW 1-CCW 2-ECH ESFAS, ECH same trains. BEX 2-ESFAS 1-ECW 1-CCW 2-ECH One train ESFAS, one train ECH different. BEV 2-ESFAS 1-ECW 1-CCW 3-ECH BER 2-ESFAS 1-ECW 2-CCW One train ESFAS, one train CCW different. BERT 2-ESFAS 1-ECW 2-CCW 1-ECH One train ESFAS, ECH same. BERW 2-ESFAS 1-ECW 2-CCW 1-ECH CCW, ECH different trains. BERY 2-ESFAS 1-ECW 2-CCW 1-ECH ESFAS, ECH different trains. BERU 2-ESFAS 1-ECW 2-CCW 2-ECH ESFAS, ECH same trains. BERX 2-ESFAS 1-ECW 2-CCW 2-ECH CCW, ECH same trains. BERZ 2-ESFAS 1-ECW 2-CCW 2-ECH Dne train ESFAS, one train ECH different. BERY 2-ESFAS 1-ECW 2-CCW 3-ECH BEP 2-ESFAS 1-ECW 3-CCW BEPT 2-ESFAS 1-ECW 3-CCW 1-ECH One train ESFAS, ECH same. BEPW 2-ESFAS 1-ECW 3-CCW 1-ECH ESFAS, ECH different trains. BEPU 2-ESFAS 1-ECW 3-CCW 2-ECH ESFAS, ECH same trains. BEPX 2-ESFAS 1-ECW 3-CCW 2-ECH One train ESFAS, one train ECH different. BEPY 2-ESFAS 1-ECW 3-CCW 3-ECH BI 2-ESFAS 1-ECW 1-CCW 1-ECH (after LOSP) One train ESFAS, ECH same. BIU 2-ESFAS 1-ECW 1-CCW 2-ECH (after LOSP) ESFAS, ECH same trains. BIX 2-ESFAS 1-ECW 1-CCW 2-ECH (af ter LOSP) One train ESFAS, one train ECH different. BIV 2-ESFAS 1-ECW 1-CCW 3-ECH (after LOSP) BIO 2-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) ESFAS, CCW same trains. BIOU 2-ESFAS 1-ECW 2-CCW 2-ECH (af ter LOSP) ESFAS, ECH same trains. BIOX 2-ESFAS l-ECW 2-CCW 2-ECH (after LOSP) One train ESFAS, one train ECH different. BIOV 2-ESFAS l-ECW 2-CCW 3-ECH (after LOSP) BIR 2-ESFAS l-ECW 2-CCW 1-ECH (af ter LOSP) One train ESFAS, one train CCW different. BIRU 2-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) ESFAS, ECH same trains. BIRX 2-ESFAS l-ECW 2-CCW 2-ECH (after LOSP) CCW, ECH same trains. BIRV 2-ESFAS 1-ECW 2-CCW 3-ECH (after LOSP) BIP 2-ESFAS 1-ECW 3-CCW 1-ECH (afterLOSP) BIPU 2-ESFAS 1-ECW 3-CCW 2-ECH (after LOSP) ESFAS, ECH same trains. BIPX 2-ESFAS 1-ECW 3-CCW 2-ECH (after LOSP) One train ESFAS, one train ECH different. BIPY 2-ESFAS 1-ECW 3-CCW 3-ECH (after LOSP) BJ 2-ESFAS 1-ECW 1-CCW 1-ECH (after LOSP) ESFAS, ECW different trains. BJX 2-ESFAS 1-ECW 1-CCW 2-ECH (af ter LOSP) One train ESFAS, one l train ECH different. 0013H120384 4-50
r i l CAUTKW: PRELinHNA;Y rESULTS-IMPORTANT UNCEITAINTIE$ DESCRISED IN SECTION 2 TABLE 4-9 (continued) i Sheet 10 of 14 j End Auxiliary Systems and Numbers Notes States of Trains Affected BJV 2-ESFAS 1-ECW 1-CCW 3-ECH (after LOSP) EJR 2-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) One train ESFAS, one train CCW different. BJ RX 2-ESFAS 1-ECW 2-CCW 2-ECH (af ter LOSP) CCW, ECH same trains. BJRZ 2-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) One train ESFAS, one train ECH different. BJRV 2-ESFAS 1-ECW 2-CCW 3-ECH (after LOSP) BJP 2-ESFAS 1-ECW 3-CCW 1-ECH (after LOSP) BJPX 2-ESFAS 1-ECW 3-CCW 2-ECH (after LOSP) One train ESFAS, one train ECH different. BJPV 2-ESFAS 1-ECW 3-CCW 3-ECH (af ter LOSP) BF 2-ESFAS 2-ECW 2-CCW 1-ECH ESFAS, ECW same trains. BFU 2-ESFAS 2-ECW 2-CCW 2-ECH ECW, ECH same trains. BFX 2-ESFAS 2-ECW 2-CCW 2-ECH One train ESFAS, one train ECH different. BFV 2-ESFAS 2-ECW 2-CCW 3-ECH BFP 2-ESFAS 2-ECW 3-CCW 1-ECH BFPU 2-ESFAS 2-ECW 3-CCW 2-ECH ECW, ECH same trains. BFPX 2-ESFAS 2-ECW 3-CCW 2-ECH One train ESFAS, one train ECH different. BFPV 2-ESFAS 2-ECW 3-CCW 3-ECH BGT 2-ESFAS 2-ECW 2-CCW 1-ECH One train ESFAS, one train ECW different; one train ESFAS, ECH same. BGW 2-ESFAS 2-ECW 2-CCW 1-ECH ESFAS ECH different trains. BGU 2-ESFAS 2-ECW 2-CCW 2-ECH ESFAS, ECH same trains. BGX 2-ESFAS 2-ECW 2-CCW 2-ECH ECW, ECH same trains. BGZ 2-ESFAS 2-ECW 2-CCW 2-ECH One train ESFAS, one train ECH different. BGV 2-ESFAS 2-ECW 2-CCW 3-ECH BGPT 2-ESFAS 2-ECW 3-CCW 1-ECH One train ESFAS, one l train ECW different; one train ESFAS, ECH same. BGPW 2-ESFAS 2-ECW 3-CCW 1-ECH ESFAS, ECH different trains. BGPU 2-ESFAS 2-ECW 3-CCW 2-ECH ESFAS, ECH same trains. BGPX 2-ESFAS 2-ECW 3-CCW 2-ECH ECW, ECH same trains. , BGPZ 2-ESFAS 2-ECW 3-CCW 2-ECH One train ESFAS, one I train ECH different. I BGPV 2-ESFAS 2-ECW 3-CCW 3-ECH BK 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOSP) ESFAS, ECW same trains. BKV 2-ESFAS 2-ECW 2-CCW 3-ECH (af ter LOSP) BKP 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOSP) BKPY 2-ESFAS 2-ECW 3-CCW 3-ECH (af ter LOSP) BL 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOSP) One train ESFAS, one train ECW different. BLY 2-ESFAS 2-ECW 2-CCW 3-ECH (after LOSP) BLP 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOSP) BLPV 2-ESFAS 2-ECW 3-CCW 3-ECH (after LOSP) 81 2-ESFAS 2-ECW 2-CCW 1-ECH (after LDP one bus) ESFAS ECW same trains. Blu 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ESFAS, ECH same trains. BlX 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) One train ESFAS, one train ECH different. BlV 2-ESFAS 2-ECW 2-CCW 3-ECH (after LOP one bus) B1P 2-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) BlPU 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ESFAS.ECH same trains. BlPX 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) One train ESFAS, one train ECH different. 0013H120384 4-51
CAUTION: PRELIMINA%Y CESULTS-IMPORTANT UNCERTAt% TIES DESCRISED IN SECTION 2 TABLE 4-9 (continued) Sheet 11 of 14 End Auxiliary Systems and Numbers es l States of Trains Affected BlPV 2-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) B3 2-ESFAS 2-ECW 2-CCW 1-ECH (after LOP one bus) One train ESFAS, one train ECW different; ; one train ESFAS, ECW(LOP) same. B3U 2-ESFAS 2-ECW 2-CCW 2-ECH (after LDP one bus) ESFAS, ECH same trains. B3X 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. B3V 2-ESFAS 2-ECW 2-CCW 3-ECW (after LOP one bus) B3P 2-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) One train ESFAS, one train ECW different; one train ESFAS, ECW(LDP) same. B3PU 2-ESFAS 2-ECW 3-CCW 2-ECH (af ter LDP one bus) ESFAS, ECH same trains. B3PX 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. B3PV 2-ESFAS 2-ECW 3-CCW 3-ECW (after LOP one bus) B4 2-E SFAS 2-ECW 2-CCW 1-ECH (after LOP one bus) ESFAS,ECW(LOP) different trains. B4X 2-ESFAS 2-ECW 2-CCW 2-ECH (after LDP one bus) ECW, ECH same trains. B4Z 2-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. B4V 2-ESFAS 2-ECW 2-CCW 3-ECH (after LOP one bus) B4P 2-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) ESFAS, ECW(LOP) different trains. B4PX 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. B4PZ 2-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. B4PY 2-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) BH 2-ESFAS 3-ECW 3-CCW 3-ECH BM 2-ESFAS 3-ECW 3-CCW 3-ECH (after LOSP) B2 2-ESFAS 3-ECW 3-CCW 3-ECH (after LOP one bus) One train ESFAS, ECW(LOP) same. 85 2-ESFAS 3-ECW 3-CCW 3-ECH (after LOP one bus) ESFAS,ECW(LOP) different trains. B6 2-ESFAS 3-ECW 3-CCW 3-ECH (after LOP 2-buses) ESFAS, ECW(LOP) same trains. B7 2-ESFAS 3-ECW 3-CCW 3-ECH (after LOP 2-buses) One train ESFAS, one train ECW(LOP) different. C 3-ESFAS CT 3-ESFAS 1-ECH CU 3-ESFAS 2-ECH CV 3-ESFAS 3-ECH CN 3-ESFAS 1-CCW CNT 3-ESFAS 1-CCW 1-ECH CCW, ECH same train. CNW 3-ESFAS 1-CCW 1-ECH CCW EC trains.Hdifferent i CNU 3-ESFAS 1-CCW 2-ECH CCW, one train ECH l same. CNX 3-ESFAS 1-CCW 2-ECH CCW EC trains.Hdifferent CNV 3-ESFAS 1-CCW 3-ECH CO 3-ESFAS 2-CCW COT 3-ESFAS 2-CCW 1-ECH One train CCW, ECH same. COW 3-ESFAS 2-CCW 1-ECH CCW, ECH different trains. CDU 3-ESFAS 2-CCW 2-ECH CCW, ECH same trains. COX 3-ESFAS 2-CCW 2-ECH One train CCW, one train ECH different. COV 3-ESFAS 2-CCW 3-ECH CP 3-ESFAS 3-CCW 0013H120384 4-52
CAUTION: PRE LIMINARY f ESULTS-IMPORTANT UNCE*.TAINTIES DESCRISED IN SECTION 2 TABLE 4-9 (continued) Sheet 12 of 1'4 End Auxiliary Systems and Numbers es States of Trains Affected CPT 3-ESFAS 3-CCW 1-ECH CPU 3-ESFAS 3-CCW 2-ECH CPV 3-ESFAS 3-CCW 3-ECH CD 3-ESFAS 1-ECW 1-CCW CDT 3-ESFAS 1-ECW 1-CCW 1-ECH ECW, ECH same train. CDW 3-ESFAS 1-ECW 1-CCW 1-ECH ECW EC trains.Hdifferent CDU 3-ESFAS 1-ECW 1-CCW 2-ECH ECW, one train ECH same. CDX 3-ESFAS 1-ECW 1-CCW 2-ECH ECW EC trains.H different CDY 3-ESFAS 1-ECW 1-CCW 3-ECH CD0 3-ESFAS 1-ECW 2-CCW CDDT 3-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH same train. CDOW 3-ESFAS 1-ECW 2-CCW 1-ECH ECW, ECH different trains; one CCW train, ECH same. CD0Y 3-ESFAS 1-ECW 2-CCW 1-ECH CCW, ECH different trains. CDOU 3-ESFAS 1-ECW 2-CCW 2-ECH CCW, ECH same trains. CDOX 3-ESFAS 1-ECW 2-CCW 2-ECH One train CCW, one train ECH different. CD0Z 3-ESFAS 1-ECW 2-CCW 2-ECH ECW, ECH different trains. CDOY 3-ESFAS 1-ECW 2-CCW 3-ECH CDP 3-ESFAS 1-ECW 3-CCW CDPT 3-ESFAS 1-ECW 3-CCW 1-ECH ECW, ECH same train. CDPW 3-ESFAS 1-ECW 3-CCW 1-ECH ECW, ECH different trains. CDPU 3-ESFAS 1-ECW 3-CCW 2-ECH ECW, one train ECH same. CDPX 3-ESFAS 1-ECW 3-CCW 2-ECH ECW EC trains.Hdifferent CDPY 3-ESFAS 1-ECW 3-CCW 3-ECH CI 3-ESFAS 1-ECW 1-CCW 1-ECH (after LOSP) CIU 3-ESFAS 1-ECW 1-CCW 2-ECH (after LOSP) CIV 3-ESFAS 1-ECW 1-CCW 3-ECH (after LOSP) CIO 3-ESFAS 1-ECW 2-CCW 1-ECH (after LOSP) CIOU 3-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) CCW, ECH same trains. CIOX 3-ESFAS 1-ECW 2-CCW 2-ECH (after LOSP) One train CCW, one train ECH different. I CIOV 3-ESFAS 1-ECW 2-CCW 3-ECH (af ter LOSP)
- CIP 3-ESFAS 1-ECW 3-CCW 1-ECH (after LOSP) l CIPU 3-ESFAS 1-ECW 3-CCW 2-ECH (af ter LOSP)
CIPV 3-ESFAS 1-ECW 3-CCW 3-ECH (after LOSP) CF 3-ESFAS 2-ECW 2-CCW 1-ECH CFU 3-ESFAS 2-ECW 2-CCW 2-ECH ECW, ECH same trains. CFX 3-ESFAS 2-ECW 2-CCW 2-ECH One train ECW, one train ECH different. CFV 3-ESFAS 2-ECW 2-CCW 3-ECH CFP 3-ESFAS 2-ECW 3-CCW 1-ECH CFPU 3-ESFAS 2-ECW 3-CCW 2-ECH ECW, ECH same trains. CFPX 3-ESFAS 2-ECW 3-CCW 2-ECH One train ECW, one train ECH different. CFPV 3-ESFAS 2-ECW 3-CCW 3-ECH CK 3-ESFAS 2-ECW 2-CCW 2-ECH (after LOSP) CKV 3-ESFAS 2-ECW 2-CCW 3-ECH (after LOSP) CKP 3-ESFAS 2-ECW 3-CCW 2-ECH (after LOSP) CKPV '3-ESFAS 2-ECW 3-CCW 3-ECH (after LOSP) C1 3-ESFAS 2-ECW 2-CCW 1-ECH (after LOP one bus) ClU 3-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. 0013H120384 4-53
CAuflow: PRE uassNACY CESULTS-IhrORTANT UNCECTAE3rtES DESCRl8ED IN SECT 10N 2 TABLE 4-9 (continued) l Sheet 13 of 14 End Auxiliary Systems and Numbers Notes States of Trains Affected l CIX 3-ESFAS 2-ECW 2-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. Cly 3-ESFAS 2-ECW 2-CCW 3-ECH (af ter LOP one bus) CIP 3-ESFAS 2-ECW 3-CCW 1-ECH (after LOP one bus) ClPU 3-ESFAS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. CIPX 3-ESFAS 2-ECW 3-CCW 2-ECH (af ter LOP one bus) One train ECW, one train ECH different. CIPV 3-ESFAS 2-ECW 3-CCW 3-ECH (after LOP one bus) CH 3-ESFAS 3-ECW 3-CCW 3-ECH CH 3-ESFAS 3-ECW 3-CCW 3-ECH (after LOSP) C2 3-ESFAS 3-ECW 3-CCW 3-ECH (after LOP one bus) C6 3-ESFAS 3-ECW 3-CCW 3-ECH (after LOP 2-buses) S SSPS ST SSPS 1-ECH SU SSPS 2-ECH SV SSPS 3-ECH SN SSPS 1-CCW SNT SSPS 1-CCW 1-ECH CCW, ECH same train. SNW SSPS 1-CCW 1-ECH CCW, ECH different trains. SNU SSPS 1-CCW 2-ECH CCW, one train ECH same. SNX SSPS 1-CCW 2-ECH CCW, ECH different trains. SNV SSPS 1-CCW 3-ECH SO SSPS 2-CCW SOT SSPS 2-CCW 1-ECH One train CCW, ECH same. S0W SSPS 2-CCW 1-ECH CCW EC trains.H different SOU SSPS 2-CCW 2-ECH CCW, ECH same trains. S0X SSPS 2-CCW 2-ECH One train CCW, one train ECH different. SDV SSPS 2-CCW 3-ECH SP SSPS 3-CCW SPT SSPS 3-CCW 1-ECH SPU SSPS 3-CCW 2-ECH SPV SSPS 3-CCW 3-ECH SD SSPS 1-ECW 1-CCW SDT SSPS 1-ECW 1-CCW 1-ECH ECW, ECH same train. SDW SSPS 1-ECW 1-CCW 1-ECH ECW EC trains.H different SDU SSPS 1-ECW 1-CCW 2-ECH ECW, one train ECH same. SDX SSPS 1-ECW 1-CCW 2-ECH ECW EC I trains.Hdifferent SDV SSPS 1-ECW 1-CCW 3-ECH i SD0 SSPS l-ECW 2-CCW SDOT .SSPS l-ECW 2-CCW 1-ECH ECW, ECH same train. l I SDOW SSPS 1-ECW 2-CCW 1-ECH ECW EC trains;Hdifferentone train CCW, ECH same. SD0Y SSPS 1-ECW 2-CCW 1-ECH CCW EC trains.Hdifferent S000 SSPS 1-ECW 2-CCW 2-ECH CCW, ECH same trains. SD0X SSPS 1-ECW 2-CCW 2-ECH One train CCW, one train ECH different. SD0Z SSPS 1-ECW 2-CCW 2-ECH ECW EC trains.Hdifferent SDOV SSPS 1-ECW 2-CCW 3-ECH SDP SSPS 1-ECW 3-CCW 0013H120384 4-54
CAUTION: PRELIMINA*,Y CESULTS. IMPORTANT UNCERTAl~1 TIES DESCRISED IN SECTION 2 TABLE 4-9 (continued) Sheet 14 of 14 End Auxiliary Systems and Numbers States of Trains Affected SDPT SSPS 1-ECW 3-CCW 1-ECH ECW, ECH same train. SDPW SSPS 1-ECW 3-CCW 1-ECH ECW EC trains.Hdifferent SDPU SSPS 1-ECW 3-CCW 2-ECH ECW, one train ECH same. SDPX SSPS 1-ECW 3-CCW 2-ECH ECW EC trains.Hdifferent SDPV SSPS 1-ECW 3-CCW 3-ECH SI SSPS 1-ECW 1-CCW 1-ECH (after LOSP) SIU SSPS 1-ECW 1-CCW 2-ECH (after LOSP) SIV SSPS 1-ECW 1-CCW 3-ECH (after LOSP) SIO SSPS 1-ECW 2-CCW 1-ECH (after LOSP) SIOU SSPS 1-ECW 2-CCW 2-ECH (after LOSP) CCW, ECH same trains. SIOX SSPS 1-ECW 2-CCW 2-ECH (after LOSP) One train CCW, one train ECH different. SIOV SSPS 1-ECW 2-CCW 3-ECH (after LOSP) SIP SSPS l-ECW 3-CCW 1-ECH (after LOSP) SIPU SSPS 1-ECW 3-CCW 2-ECH (after LOSP) SIPV SSPS 1-ECW 3-CCW 3-ECH (after LOSP) SF SSPS 2-ECW 2-CCW 1-ECH SFU SSPS 2-ECW 2-CCW 2-ECH ECW, ECH same trains. SFX SSPS 2-ECW 2-CCW 2-ECH One train ECW, one train ECH different. SFY SSPS 2-ECW 2-CCW 3-ECH SFP SSPS 2-ECW 3-CCW 1-ECH SFPU SSPS 2-ECW 3-CCW 2-ECH ECW, ECH same trains. SFPX SSPS 2-ECW 3-CCW 2-ECH One train ECW, one train ECH different. SFPV SSPS 2-ECW 3-CCW 3-ECH SK SSPS 2-ECW 2-CCW 2-ECH (after LOSP) SKY SSPS 2-ECW 2-CCW 3-ECH (after LOSP) SKP SSPS 2-ECW 3-CCW 2-ECH (after LOSP) SKPV SSPS 2-ECW 3-CCW 3-ECH (after LOSP) S1 SSPS 2-ECW 2-CCW 1-ECH (after LOP one bus) SIU SSPS 2-ECW 2-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. SIX SSPS 2-ECW 2-CL,W 2-ECH (after LOP one bus) One train ECW, one train ECH different. Sly SSPS 2-ECW 2-CCW 3-ECH (after LOP one bus) SIP SSPS 2-ECW 3-CCW 1-ECH (after LOP one bus) SlPU SSPS 2-ECW 3-CCW 2-ECH (after LOP one bus) ECW, ECH same trains. SlPX SSPS 2.rCW 3-CCW 2-ECH (after LOP one bus) One train ECW, one train ECH different. SlPY SSPS 2-ECW 3-CCW 3-ECH (after LOP one bus) SH SSPS 3-ECW 3-CCW 3-ECH SM SSPS 3-ECW 3-CCW 3-ECH (after LOSP) S2 S~as 3-ECW 3-CCW 3-ECH (after LOP one bus) ( 56 F.'S 3-ECW 3-CCW 3-ECH (after LOP 2-buses) I r l t 0013H120384 4-55
N e% s TABLE 4-10. MATRIX 0F AUXILIARY SYSTEM TO MAH - Maf n Lf ne Avallfary Feedwater Pops Steam Generator Relfef Valves Charging Pumps I ' inW $ p , Trfp Avaf1tary Systems 11 12 13 14 PV-7411 PV-7421 PV-7431 PV-7441 1A 18 1A 18 1C Offstte Crfd (a) (b) (b) l 4,160V But EI A X X X X X 4.160V Bus EIS X X X 4,160V Bus EIC X X X X DC Bus EIAll (e) X X 1 X DC Bus E1811 (e) X X X DC Bus EIC11 X X X X DC Bus E1011 X X EDG DG-11 X X X X X EDG DG-12 X X X EDG DG-13 X X X X [5FAS Trafn At (k) X (1) (f ) X ESFAS Trafn 8 (k) I X [5FAS Train C X (f) X [CW Train A (m) (m) (n) (n) ECW Traf n 8 (a) (a) (n) (n) ECW Trafn C (m) (m) (q) CCW Train A (a) (m) CCW Train B fn) (n) CCW Traf n C (n) (n) ECH Traf n A (q) ECM Train 8 (4) ECH Train C (4) EA8 HvAC (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) Reactor Trfp fr) a o n or prtcu$ tFor[f sEe
- an ne
$$PS and ESFAS failures, an *X* fnplies falfure f the automatic starto!rYnction of the accocf indfcates failure ated main Ifne of that function given failure of function.
w w
Also Available On-Aperture Card TI LINE SYSTEM INTERDEPENDENCIES* APERTURE CARD i f I
\
w pngs,, Sheet 1 of 2 P ress:rf rer Safety Contatanent Sa fety Reactor Containment Reactor Residual diftf Yalves injection $ Pray Injection Fan Cooling Units Coolant Heat Removal pgg Pops Recirculation Containment Pops Paps Isolation 3t55A PCV656A 1A 18 1C A. 8 C A B C 11A 12A 118 128 11C 12C A B C D A 8 C (b) (b) (b) (b) (b) (b) x** x x x (c) x x x x x x x x kN$ 1 x x (d) x x x x x X x x x x fd X l x x (d x (c) X X X x x x X X
!$$ x x x (d) x I
fh h h) ) h h hl h h) ( ) ) () ) d x x x x x I (j ) (j) fdl (3) (j) (d) (n) (n) (n) (n) (o) x x (p) (p) (p) (p) (n) (n) (n) (n) to) x x x gggg x (o) x x x (q) (g) (q) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) (s) CAUTION: PRELIMINARY RESULTS-t IMPORTANT UNCERTAINTIES DESCRISED IN SECTICN 2 O 4-56
- - .-- - . _ _ . - - - ~ +
TABLE 4-10 (continued) NOTES: Sheet 2 of 2
- a. Offsite grid failure as a result of an initiating event other than i loss of offsite power is assumed to guarantee successful turbine trip.
- b. The previously operating charging pump will restart when power is available to its Class 1E AC bus.
i c. Power is lost to the associated block valve. t
- d. Failure of the indicated auxiliary system train fails certain containment isolation valves.
- e. Loss of the indicated DC bus deenergizes one MSIV solenoid valve resulting in MSIV closure.
. f. Failure of both SSPS trains results in loss of turbine trip signals from the SSPS.
- g. Failure of both SSPS trains results in failure of the main steam isolation function.
- h. Failure of both SSPS trains results in failure of the automatic start function for the indicated equipment.
- 1. Failure of SSPS or the indicated ESFAS trains has an effect on the centrifugal charging pumps.
- j. Failure of both SSPS trains or the indicated ESFAS train results in no automatic start signal and no automatic open signal to the CCW 2 isolation valves for the associated reactor containment fan cooling units.
- k. Failure of ESFAS trains A and B results in failure of the main steam line isolation function.
- 1. Failure of ESFAS train A results in failure of the automatic start function for AFW pump 14 (turbine-driven pump).
- m. Loss of all ECW or CCW pumps fails the centrifugal charging pumps due i to loss of lubrication and cooling from the CCW systems. Loss of ECW or CCW train C and plant instrument air fails charging pump 1A. Loss of ECW or CCW trains A and B and plant instrumentation air fails charging pump 18.
- n. Loss of ECW trains A and B fails ECH trains A and B. Loss of an ECH train fails the associated safety injection and containment spray
, pumps in the long term (after the start of recirculation) due to the loss of room cooling.
- o. Failure of the indicated ECW or CCW train fails cooling to the associated residual heat removal heat exchanger.
- p. Failure of all ECW or CCW trains results in loss of reactor coolant pump motor and thermal barrier cooling. With charging pump failure, a reactor coolant pump seal loss of coolant accident may result.
- q. Loss of the indicated ECW or ECH train fails the associated safety injection and containment spray pumps in the long term (after the start of recirculation) due to the loss of room cooling.
- r. If reactor trip breakers fail to open, no automatic turbine trip signal from reactor trip will occur.
- s. Loss of EAB HVAC is assumed to result in loss of all AC power for the Scoping Study analysis with resulting loss of power to the systems indicated.
CAUTION: PRELIMINARY RESULTS. IMPORTANT UNCERTAINTIES DESCRIGED IN SECTION 2 0103H052085 4-57
TABLE 4-11. AUXILIARY. SYSTEMS IMPACT VECTORS Impact Auxiliary Event Tree
, Vector Definition Designation End States Included AUX All required auxiliary systems are available. AUX,D(GT),NT(GT).
NW(GT).DW(GT) 4 T Loss of one ECH train. T. DU(GT), DX(GT) N Loss of one CCW train. N, D, 00(GT) DOT (GT) DOW(GT) NT Less of one CCW train and the same ECH train. NT,DT,DOU(GT) . NW Loss of one CCW train and an opposite ECH NW, DW, D0Y(GT) train. NU Loss of one CCW train and two ECH trains. NU, DU, FU, FX, FV, NX, DX 0- Loss of two CCW trains. O D0, DPW(GT) H(GT) F Loss of two CCW trains and one ECH train, F, OT, DOT, OW, DOW
, only two traint affected.
P Loss of three CCW trains, RCP seal LOCA. P PT, FP, FPV, DPU, DPX I Loss of one AC power bus after a loss of I BI, CI, CIOU, Cl, , offsite power. CIV Loss of one AC power bus, offsite power is 1, 10 1
- available, and loss of a second ECW train.
1 IU Loss of one AC power bus and a second ECH IU, IV train after a loss of offsite power. IOU Loss of one AC power bus , a second CCW 100, lu, IOX, IV, train, and a second ECH train after a loss of IOV, IX offsite power. IP Loss of one AC power bus and all CCW trains IP,1P, IPU, IPU after a loss of offsite power, RCP seal LOCA. ! K Loss of two AC power buses. K,~AK, AL, BK, CK KY Loss of two AC power buses and the third ECH KV, AKV train. KP Loss of two AC power buses and the third CCW KP, 2. AKP , train. M Loss of all AC power M AM, BM, CM 9 Loss of the EAB ventilation system. 9 A Loss of one actuation signal. A. AT, AN ANT, AQW, AD, ADT, ADOW, AD0Y AU Loss of one actuation sfgnal and two ECH AU, AW, AX, AD0Y, AEY, trains. ADOU, ADOZ, AEOU, AEX, 1 AE0X, AE0V, AERX, AFU,
- AERZ, AERY, AEPY, AFV, AFPY, AGY AO Loss of one actuation signal and two CCW AO, AQ, ADO, ADPW, I
trains. AEPT, AH, AE, AET, I AE0T, AE0Y, AERT, ! AERY, AG ADW Loss of one actuation signal, one ECW train ADW, AEW, AFW and an opposite ECH train, two trains affected. AI Loss of one actuation signal and one AC power AI, AIO, AIOU, Al, l bus, sue train affected. AlV, AIU i M Loss of one actuation signal and one AC power M, MU, MX, MR, A4, bus, two trains affected. MO, MOU, A3, A3V, MRX, A4V B Loss of two actuation signals. B, BD, BDOW, BDR, BDRY, BERT, BDP C Loss of three actuation signals, or failure C, CN, CDW, CDO, CDOW,
- of SSPS. C00Y, CDOU, CDOZ, CFY, CDP All auxiliary event tree end states not shown above are included with auxiliary systems impact vector "M" because of the low frequency of occurrence of these end states.
CAUTION: PRELHIAINARY RESULT 5r inePORTANT UNCERTAINTIES DESCRISED IN SECTloN 2 0013H122884 4-58
1 CAUTION: PREUMINARY C ESULTS. IMPORTANT UNCEZTAINTIE S DESCRl8ED IN SECTION 2 TABLE 4-12. IMPACT VECTOR DEFINITION l l l l Sheet 1 of 2 j General l Transi nt IMPACT YECTORS Events Events AUX T U N NT NW NU 0 F P ! 1 IU IOU IP K KV KP M 9 A AU A0 ADW AI AJ B C TT 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 2 TT OH 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 OH VI O O O O O O O O O O 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 VI AF 0 0 0 0 C 0 0 0 0 0 2 2 2 2 2 1 1 1 1 1 2 2 2 2 2 1 1 1 AF F1 0 6 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 2 1 1 2 2 2 2 2 2 2 1 F1 PR 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 PR 01 0 0 0 2 2 2 2 2 2 1 2 2 2 2 1 2 2 1 1 1 2 2 2 2 2 2 2 1 OI ON 0 0 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 2 1 1 0 0 0 0 2 2 2 1 ON OB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 OB EA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 EA EB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 1 1 1 E8 EC 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 EC HA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 HA HB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 1 1 1 HB HC 0 0 0 0 0 0 0 0 0 0 0 0 0 0.0 0 0 0 1 1 0 0 0 0 0 0 0 1 HC OR 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 OR SA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 SA S3 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 l 1 1 SB SC 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 SC Small LOCA IMPACT VECTORS Tg Events p Ev;nts AUX T U N NT NW NU 0 F P I 1 IU IOU IP K KY KP M 9 A AU A0 ADW AI AJ B C TT 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 2 TT OH 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 OH VI O 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 VI F1 0 0 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 2 1 1 2 2 2 2 2 2 2 1 F1 08 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 OB EA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 EA EB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 1 1 1 EB EC 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 EC HA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 HA HB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 1 1 1 HB HC 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 HC SA 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 SA SB 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 1 1 1 0 0 0 0 0 1 1 1 SB SC 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 1 SC egend: 0 = auxiliary systems impact vector has no effect on the top event. 1 e auxiliary systems impact vector fails the top event. 2 a auxiliary systems impact vector affects the top event failure split fraction. 3 = auxiliary systems impact vector do not map to these top events because of previous failures. 4-59 380H1200784
1 I TABLE 4-12 (continued) Sheet 2 of &
~
IMPACT VECTORS Tree Top Top Events Events AUX T U N NT NW NU 0 F P I 1 IU IOU IP K KV KP M 9 A AU A0 ADW Al M B C FC 0 0 0 2 2 2 2 2 2 3 2 2 .2 2 3 2 2 3 3 3 2 2 2 2 2 2 2 3 FC RA 0 1 1 0 1 1 1 0 1 3 1 1 1 1 3 1 1 3 3 3 1 1 1 1 1 1 1 3 m RB 0 0 1 0 0 0 1 0 0 3 0 1 1 1 3 1 1 3 3 3 0 1 0 1 0 1 1 3 RB RC 0 0 1 0 0 0 1 0 0 3 0 0 1 1 3 0 1 3 3 3 0 1 0 0 0 0 0 3 RC SI O O O O O O O O O 3 0 0 0 0 3 0 0 3 3 3 0 0 0 0 0 0 0 3 SI OL 0 0 0 0 0 0 0 0 0 3 0 0 0 0 3 0 0 3 3 3 0 0 0 0 0 0 0 3 OL LA 0 1 1 1 1 1 1 1 1 3 1 1 1 1 3 1 1 3 3 3 1 1 1 1 1 1 1 3 LA LB 0 0 1 0 0 1 1 1 1 3 0 1 1 1 3 1 1 3 3 3 0 1 1 1 0 1 1 3 LB LC 0 0 1 0 0 0 1 0 0 3 0 0 1 1 3 0 1 3 3 3 0 1 0 0 0 0 0 3 LC RX 0 0 1 0 0 0 1 0 0 3 0 0 1 1 3 0 1 3 3 3 0 1 0 0 0 0 0 3 RX CS 0 2 1 0 2 2 1 0 2 3 2 2 1 1 3 1 1 3 3 3 2 1 2 1 2 1 1 3 CS CP 0 0 0 0 0 0 0 0 0 3 0 0 0 0 3 0 0 3 3 3 2 2 2 2 2 2 2 3 CP CI O O O O O O O 0 0 3 2 2 2 2 3 2 2 3 3 3 2 2 2 2 2 2 2 3 CI IMPACT VECTORS T@ Top Events Events AUX T U N NT NW NU 0 F P I 1 IU IOU IP K KV KP M 9 A AU A0 ADW AI M B C FC 0 0 0 2 2 2 2 2 2 1 2 2 2 2 1 2 2 1 1 1 2 2 2 2 2 2 2 1 FC RA 0 1 1 0 1 1 1 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 M RB 0 0 1 0 0 0 1 0 0 0 0 1 1 1 0 1 1 1 1 1 0 1 0 1 0 1 1 1 RB RC 0 0 1 0 0 0 1 0 0 0 0 0 1 1 0 0 1 0 1 1 0 1 0 0 0 0 0 1 RC CS 0 2 1 0 2 2 1 0 2 2 2 2 1 1 2 1 1 1 1 1 2 1 2 1 2 1 1 1 CS LA 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 LA LB 0 0 1 0 0 1 1 1 1 1 0 1 1 1 1 1 1 1 1 1 0 1 1 1 0 1 1 1 LB LC 0 0 1 0 0 0 1 0 0 1 0 0 1 1 1 0 1 1 1 1 0 1 0 0 0 0 0 1 LC RX 0 0 1 0 0 0 1 0 0 1 0 0 1 1 1 0 1 1 1 1 0 1 0 0 0 0 0 1 RX CP 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 2 2 2 2 2 2 1 CP CI O O O O O O O O O O 2 2 2 2 2 2 2 2 1 1 2 2 2 2 2 2 2 1 CI Lagend: 0 = auxiliary systems impact vector has no effect on the top event. 1 = auxiliary systems impact vector fails the top event. 2 = auxiliary systems impact vector affects the top event failure split fraction. 3 = auxiliary systems impact vector do not mar to these top events because of previous failures. CAUTION: PRELIMINARY RESULTS-IMPORTANT JNcERTAINTIES DESCRIBED IN SECTION 2 0103H052085 4-60
TABLE 4-13. GENERAL TRANSIENT EVENT TREE SUCCESS CRITERIA Sheet 1 of 4 E t Description Success Criteria TT Turbine Trip or Automatic turbine trip (four valves shut to Main Steam Isola- block steam flow) or automatic trip of all tion Trip four MSIVs on SI due to rapid cooldown. OH Operator controls Because the reactor coolant system High Head Injection refills and repressurizes quickly one high head injection starts and OH is assumed failed for the Scoping Study. VI Vessel Integrity Should PTS conditions occur (rapid cooldown and repressurization) there is a small chance of crack initiation and propagation. In such a case, a loss of coolant accident beyond the capability of the ECCS would lead to core melt. NOTE: One condition under which PTS occurs has not been included in the Scoping Study model--bleed and feed with the reactor coolant pumps secured; our judgment is that the chance of vessel failure will be negligibla compared to other contributors to all plant damage states. AF Two or More Trains Automatic operation of two, three, of Auxiliary Feed- or four AFW pumptrains and their water and associated ARVs. Steam Relief Relief F1 One Train of AFW Automatic operation of one AFW pump train and Steam Relief and its associated ARV given failure of auxiliary feedwater. PR Pressurizer Power- Automatic opening of either PORY and Operated Relief automatic closure of both PORVs. Failure Valve Opens to open a relief path on demand is assumed and Closes to lead to a small LOCA, and failure to reclose such a relief path is a small LOCA. OI RCS Inventory This is essentially a LOCA path. The LOCAs i Control considered occur primarily as a result of I _upport system and charging system failures. Success requi t that none of CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRIBED IN SECTION 2 l l 0103H052085 4-61 i l
TABLE 4-13 (continued) Sheet 2 of 4 E Descripdon Success CMeria t the following LOCA paths are established.
- 1. The reactor coolant pump seal LOCA, which occurs some time following loss of seal injection and loss of component cooling water to the reactor coolant pump thermal barrier.
- 2. Two distinct LOCA paths can occur via letdown system. If charging flow is lost and the letdown system does not isolate automatically, a flow path to either the containment or outside containment to the VCT will exist.
Motor-operated valves in the letdown system receive isolation signals on low pressurizer level. If none of the valves shut, a flow path exists outside containment to the VCT. If either letdown system isolation valve closes, there will be no LOCA. If the two isolation valves remain open and either one of the containment isolation valves shut, a 600 psi relief volve will open causing a LOCA inside containment.
- 3. The last of these LOCA pt ths is the seal return line that goes outside containment. This line should isolate on an SI signal.
If any of these LOCA paths is established, event OI fails. The two paths outside containment, the VCT or the seal return lines, also provide small paths for containment bypass in case of a core melt. LOCAs via these paths spill water outside containment so there is no possibility of containment sump recirculation cooling in the long-term. For the Scoping Study, we conservatively assume that if event OI fails a containment bypass path exists. If OI fails and high head injection succeeds, the scenario branches to the LT1 tree and requires CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 0103H052085 4-62
- . . .- -. .- ~. ._ . . _ . - . . . _ . . . . - .
d CAUTICII: PIIEL8tIIIsAftY RESULTS-IAAPOglTAfdT Uf0CE2TAlstTIES ossCRenam secTos TABLE 4-13 (continued) Sheet 3 of 4 E t Description Success Criteria i successful recirculation cooling. While this appears nonconservative for the two LOCAs outside containment, the maximum i combined flow rate via those LOCAs is only 100 gpm. At this flow rate, with high pressure injection working, several days are required before the RWST inventory is expended. In that time, many actions outside the scope of the present analysis could be carried out to stabilize the plant. Therefore, if anything, branching to the LT1 tree is slightly conservative. ON Operator Provides This event allows for the very slight Long Term Cooling chance that in a completely stabilized condition the operators may intercede improperly or neglect simple long-term actions for which a great many redundant i cues are available and induce failure of
- cooling in the long-term.
'B Operator Initiates The operator must recognize the loss of Bleed and Feed secondary cooling conditions, decide to carry out bleed and feed cooling, and actually follow the procedures for bleed
- and feed cooling. The procedures described in the ERGS have been used in evaluating
- operator response. Basically, the operator l must decide to carry out bleed and feed, &
l push the SI button to start high head injection, and turn the switch to open each of two pressurizer PORVs all within approximately 20 minutes following the loss of secondary cooling. This allows for the time required to boil dry the steam generators before RCS pressure would increase too high for successful high head injection. Success of this event also - requires that the two PORVs physically open on demand. EA ECCS Common A Train A of ECCS includes high head injection pump A, low head injection pump A, and containment spray pump A. All three pumps share a common header such that common ( valves in that header, maintenance on 0013H122884
CAUTION: PREUANNA2Y RESULTS-7,c"[o$s'Ec l" TABLE 4-13 (continued) Sheet 4 of 4 E t Description Success Criteria EA any of the three pumps, or ventilation to ' l the comon room can fail all three pumps. Failure of any of that comon equipment leads to failure of Top Event EA. EB ECCS Comon B All comon ECCS train B equipment must operate as above. Additionally, any common events that can fail two trains of ECCS (e.g., common cause failure of one supply valve in each train) are included in the
, quantification of this top event.
EC ECCS Comon C All common ECCS train C equipment must operate as above. Additionally, any common events that can fail all three trains of ECCS (e.g., loss of the RWST) are included in the quantification of this top event. HA High Head Safety Must start and successfully operate for Injection Pump A 24 hours. HB High Head Safety Same as Top Event HA; however, for train B, Injection Pump B any comon causes not included in EB that can fail, two trains of high head injection are quantified in this top event. i HC High Head Safety Same as Top Event HA; however, for train C, Injection Pump C any comon cause events not included in Top Event EC that can fail three trains of high head injection are quantified in this top event. i OR Closed Loop RHR For the Scoping Study, OR is assumed failed , Cooling since the RHR pumps are not qualified for ! high temperature or high humidity environment. SA Containment Containment spray pump train A must start and t Spray A run for 24 hours. SB Containment Same as Top Event SA; however, for train B, 2 Spray B any common cause events not included in HB that can fail two trains 07 containment spray are quantified. SC Containment Same as Top Event SA; however, for train C, Spray C any common cause events not modeled in EC that can fail three trains of containment spray are quantified. 0103H052085 4-64 l~
TABLE 4-14. CODING F6R TRANSFER STATES BETWEEN EARLY AND LONG-TERf1 (LT-1) RESPONSE EVENT TREES LT-1 Entry State Equipment Status 01 Succeeds 01 Fails A B C D E F G H I J K L Number of High Head 0 1 1 2 2 2 0 1 1 2 2 2 Safety Injection Failures (Top Events HA, HB, HC) Due to ECCS Common - - Yes - 1 2 - - Yes - 1 2 Failures (Top Events EA, EB, EC) Loss of RCS Inventory No No No No No Ne Yes Yes Yes Yes Yes Yes Control (Top Event 01) l NOTE: On branch LT-1, at least one high head safety injection pump works; water is in sump. l l i CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 l
~ @013H122884
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TABLE 4-15. CODING FOR TRANSFER STATES BETWEEN EARLY AND LONG TERM (LT2) CORE MELT RESPONSE EVENT TREES LT-2 Entry State Equipment Status 01 Succeeds OI Fails A B C D E F G H I J K L M N O P Q R S T Number of Containment 0 1 1 2 2 2 3 3 3 3 0 1 1 2 2 2 3 3 3 3 Spray Failures (Top Events SA, SB, SC) i Due to ECCS Conunon - - Yes - 1 2 - 1 2 3 - - 1 - 1 2 2 3 $ 1 Failures (Top Events EA ,EB, EC) Loss of RCS Inventory No No No No No No No No No. No Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Control (Top Event OI) CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2
l l TABLE 4-16. LT1 LONG-TERM EVENT TREE SUCCESS CRITERIA Sheet 1 of 3 Description Success Criteria E nt FC Fan Coolers Fan coolers a.e the primary source of containment cooling. Success requires that two out of six fan coolers operate and be supplied with cooling water. RA Recirculation Each ECCS train can operate in a Common A recirculation mode moving water from the containment sump to the reactor coolant system via the high head safety injection pump or the low head safety injection pump and through the containment spray system via the containment spray pump. Common sumps, suction piping, and valves supply each of the three ECCS pumps in train A. Any failures in this common equipment lead to failure of Top Event RA. RB Recirculation Same as Top Event RA for train B; also, Common B any common causes that can fail two trains of ECCS in the recirculation phase (e.g., common cause valve failures) are included in this top event. RC Recirculation Same as Top Event RA for train C; also, Commmon C any common cause events that can fail all three trains of ECCS in the recirculation model (such as failure of the RWST to be isolated) are included in this top event. SI High Head Safety If one out of three trains of high head Injection safety injection is operating, water will be supfi;ed to the reactor coolant system at high pressure. This keeps the core covered and prevents injection via the low head safety injection pump and this precludes cooling by the RHR heat exchanger. This event is really just a switch based on the entry state to the LT1 event tree which depends on high head safety injection pumps performance in Top Events HA, HB, and HC. CAUTION: PRE LIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 4-67
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TABLE 4-16 (continued) O. . W" ,* FU. ' e 4 -y. . .
. 4 ~ .; . .
. g ;o ..;-
i Sheet 2 of 3 t.t. 5 . ;. e .,
- pr 1 Top Event Description Success Criteria h;-; .#3 h . . .
g . .. s . .
;c . G.; . OL RCS Pressure Below When Top Event C has failed, RCS . .i: . C :~
t Low Head Safety pressure should below low head safety Injection Design injection design before the RWST runs g( 6 :." . y;S; empty unless the reactor coolant pumps W: < t-
~. are not running. In that case, high d. $ tN '.
temperatures in the reactor vessel head . g.l:~. '6. .., ' .. J region may not be adequately cooled and p . ' .5 , .. a steam bubble may form in that region, b.4.'. , A , , maintaining RCS pressure high. 74 .pj
".* Therefore, Top Event OL succeeds when '/ : rf %M y nonvital power is available to run N . ." -
.-- reactor coolant pumps, and Top Event OL f.M. .;.; . fails in auxiliary states when nonvital J. 1
?;.4.f.
- power is failed.
i
- g .p...m;. .
- LA Low Head Safety Low head safety injection pump train A i " k.h e
@ Injection A must start and operate for 24 hours. G.~ t[( ,'" .
- . G .. .
1.' LB Low Head Safety Same as Top Event LA for train B; also
'(y.4.}. AJ ~. ~
Injection B includes common cause events not
. modeled in Top Event RB that fail two 49$[.] "
i trains of low head safety injection. .
- s. 'y; LC Low Head Safety Same as Top Event LA for train C; also [g... 9. ~r; ~
3 Injection C includes common cause events not '('- modeled in event RC that fail all three [. .)
,t trains of low head safety injection. M]76
. u 1.: ! - f"i t
,, RX RHR Heat Exchanger Success requires component cooling ?
; water to be supplied to one RHR heat .-f.[ 2 Vj. "1 7. ..
exchanger that has low head X Wi . : :. recirculation flow; i.e., any one of -A.'. Top Events LA, LB, or LC has succeeded. 7, ; "7 's; 7; 8 .. (.'. - 91 i p. t.
. CS Containment Spray Two containment spray pumps are -
4 required to generate adequate spray 1.) , 7. Sp dispersal in the containment spray 'J %.'...; Q 8, system. Operability of the containment .. - :( W. spray pumps has been established in the . ;,. 5 i. early tree. Therefore, Top Event CS is y{.6.5. ' "'
- a switch conditional on entry state and VN .4.^
1 Top Events RA, RB, and RC. j,;;gM l' p g.; 5r d..
.--; p.; ,9 4 CAUTION: PRELIMINARY RESULTS- fh.NM-/ * ' , ".
IMPORTANT UNCERTAINTIES e
,, DESCR18ED IN SECTION 2 ~.ng A(.;..
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0013H122884 4-68 b;)[U., wa.r ..
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TABLE 4-16 (continued) Sheet 3 of 3 Description Success Criteria Eet CP Gross Containment The large containment purge lines must Isolation (greater be isolated or gross failure of than 3 inches) containment isolation occurs. Failure of this event r. quires that the purge line valves be open at the time of the initiating event and fail to close following that event. Closure is automatic on SI signal, but since these valves are i+0Vs, success depends on the state of electric power. CI Containment Isolation If any small line that connects the (less than 3 inches) containment to the outside atmosphere fails to isolate following safety injection signal, Top Event CI fails. Among the most likely failure paths are the letdown lines to the VCT and the reactor coolant pump seal return lines. CAUTION: PRELIMINARY RESULTS-IANORTANT UMCERTAINTIES DESCRISED IN SECTION 2 4-69
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= 158 DE23 158 w 159 D12 159 160 DIES 160 w 161 D1E2 161 l
162 DE22 162 w-r- 163 D23 163 w-r- 164 D2E3 164 w 165 D11 165 166 DIE 3 166 167 DIE 1 167 I 168 DE21 168 w w- 169 D22 169 l
= 170 D2E2 170 w-w 171 D21 171 g
172 D2E1 172 w-r-r- 173 D3 173
------------------ 174 XFR6 174
[ I
------------------ 175 XFR6 Nfe== N7 -
176 823 182 19Q [ = - F-177 A186 178 B1C5 191 192
= =-aw--= = = 179 A185 193 180 A23 194
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182 AS21 196 183 A282 197
- = w- 184 AC82 198 MWNft- N : -
185 A291 199 186 E3 200
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------------------ 188 XFR6 202 I a fr-
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, 199 E3 220 w-==- 200 A2C2 221 l . i ------------------ 201 XFR6 222 I
i l j l ------------------ 202 XFR7 230 I FIGURE 4-3 (Sheet 4 of 7) 4-75
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-------------------------- 262 XFR9 347
= ------------------ 263 XFR3 520 g
------------------ 264 XFM4 528 g
=m- - .- 265 B22 536 266 A184 537 i
. - . r- 267 B1C3 538
- == = =g 268 A922 539 269 A293 540
= t- 270 ACB4 5 41 M*--mm-we - M 271 A193 542 272 A22 543
= r r- 273 A1C3 544
- me- aa- w* - = - 274 A291 545 275 E3 546
= = t- 276 A2C1 547
------------------ 277 XFRS S48 I
m: m: - r 278 8 21 556
= 279 A1ST. 557 r = w-me-- 280 B1Cl 558
== = = g
= 281 AB23 559
= 282 ATJ3 560
= = w-sor- 283 MB6 561
= n = = g = 284 A181 562
= 285 A21 563
= = w- se*- 286 A1C1 564
= = = g
= 287 A232 565
= 288 E3 566
= = w -ses-- 289 A2C2 567
------------------ 290 XFR7 568 l = M*-a: r 291 ACS 576
[ g 292 AB23 293 AB22 577 578 294 A293 579
- F- 295 B2C1 580 g
= 296 ACB4 581
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CAUTION: PRELIMINARY RESULTS-FIGURE 4-3 (Sheet 6 of 7) DESCRigED IN SECTION 2 4-77
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------------------------ 55 XFR4 661 1 ------------------------ 56 XFw4 9 91
------------------------ 57 XFT54 1321 1 ------------------------ 58 XFR4 1651
------------------------ 59 XFR4 1981 1 ------------------------ 60 XFR4 2311 w-a-w------------------------- 61 XFR4 2641 CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 FIGURE 4-5 (Sheet 2 of 6) 4-81
430 , 3NO 530 3NG 930 3NG S30 3NO S30" 3NO 930' S30' 3NO 3NO 930- 3NO S30' 3NO ND 81v13 40 Sgv13 NO' 91v13 ND 8av13 NO' 91v13 NO' 8iv13 S1vlE NO' NO' S1v13 NO' 91v13 I yna CCI Y 998 V MI 8 lCEI v 19CI 8 4681 E ECBI 3 5991 3 2 6 CCE 6- 99E 6 66E 6 1CZE 6 19CE 6 16BE 6 ECXE 6 C 1 CCC va 99C yM 59tE 6 MC 81 1 CEC VM 89CC 81 16BC 8M ECIC 31 29tC 31 9 6 CCt 6 996 6 669 6 1C29 6 89Ct 6 5689 6 ECID & C 1 ' CCC VM - 99C va 2999 6 46C 81 ICCC wM 89CC SM 56BC 81 ECIC 31 59tC 31 9 6 CC9 6 999 6 669 6 1C29 6 39C9 6 8689 6 5C39 6 i n CCI vn 994 yO 5999 6 ML 86 1CE4 VX 89CL SM 1684 8X 5C14 3n 5964 3n 8 6 CCB 6 998 6 668 6 IC28 6 I9C8 6 6 1 1688 6 ECIB 6 5968 6 CC6 WN 996 VM 666 SM ICE 6 va 89C6 81 1686 81 fO 6 CtO 6 940 6 5Cle 31 59t6 31 3000 6 ICC0 6 8990 6 8660 6 ECEO 6 39CO 6 11 n CtI vO 94I VX 100t sI ICCI YO
-IZ 6 CtE 6 94E 6 1002 6 5991 8n 5661 8X ECEI 3A 39CI 3n ICCE 6 899E 6 166C 6 EEEE 6 59CE 6 lC n teC va 94C V0 100C 8X ICCC WO 899E 8X ID 6 Ctt 6 949 6 166C SA ECEC 3A 29CC 3A 5009 6 ICCt 6 3999 6 8669 6 5CEW 6 59C9 6 1C A ctC vA 94C vA 300C SA ICCC VA 19 6 Ct9 6 899C SA 166 C 8A ECEC 3A 59CG 3A 949 6 1009 & ICC9 6 8999 6 1669 6 ECC9 6 29C9 6 84 N Ct4 YN 944 VD 5004 8N ICC4 vO 5994 8N 88 6 5664 80 5CEE 3N 59C4 3N cte . 6 948 6 1008 6 ICC8 6 8998 6 166B 6 ECCB 6 E9CS 6 86 NA Ct6 WN1 946 VOM 5006 SNA ICC6 VOM 5996 SNA 8 6M SOM 30 6 CCO ~ 6 900 6 EC26 3N1 29C6 3NA IOIO 6 ICtO 6 3940 6 2000 6 ECCO 6 5990 6 E1 NM CCI WNM 981 VDA f0II 8NM ICDI v0A 894I GNA EE 6 CC5 6 5008 801 ECC1 3NM 2991 3NM 9BE e TOIE 6 ICtE 6+ 8945 6 500E 6 ECCE 6 3995 6 3C NO CCC YNO 9BC WOD ICIC SNO ICDC v02 894C GNX EOOC SOX 29 6 CCt 6 989 6 ECCC 3NO 399C 3NA 1089 6 1Ctt 6 3949 6 3009 6 ECCt 6 2999 6 3C NM CCC NfM 98G VDA ICIC SNA ICtC VDA 894C I 29 6 200C 801 ECCC 3NN 599C 3NM CC9 6 989 6 8089 6 1Ct9 6 I949 6'NM 3009 6 ECC9 6 2999 6 34 NO CC4 wNO 984 WOZ 1084 8NX lCt4 v00 1944 GNA 3004 803 28 6 CCB 6 988 6 CCC4 3NO 5994 3NA SOIB 6 1CtB 6 8948 6 500B 6 ECCB 6 5998 6 36 Nk CC6 WNI 986 VCX IOt6. SNZ lCt6 WOX 8946 SN2 3006 600 CO 6- C90 6 960 6 5CC6 3NK 8996 3NX 80:0 6 ICCO 6 3980 6 3080 6 ECTO 6 5940 6 C NA C91 YNA 961 WOA !028 ENA ICC1 WOA 5991 SNA CE' 6- 5083 SCA ECtI 3NA 5945 3NA C9E 6 96E 6 fO E 6 ICCE 6 19BE 6 208E 6 ECtE 6 394E 6 CC N C9C vO 96C WN 102C 8N ICCC vO 199C 80 501C 8N Ct 6 C99 6 ECtC 3N 59lC 3N 969 6 8029 6 ICCt 6 I9Bt 6 2019 6 ECtt 6 5949 6
- CC NM C9C v01 96C wNM 302C SNM ICCC VDA 898C 801 EDIC UNA C9 6 C99 6 ECDC 3NM 394C 3NM 969 6 1029 6 ICC9 6 8999 6 3019 6 ECt9 6 2949 6 C( NA C91 v0N - 962 vNI 1024 GNI ICCL VOM 8984 5CM 3014 8NA CB 6 C9E 6 ECt4 3NA 3944 3NI 968 & IO2B 6 ICCB 6 8988 6 5088 6 ECtS 6 5948 6 C6 NO C96 v00 966 YNA I026 SNO ICC6 W02 1986 801 3086 SNk 90 6 C40 6 5Ct6 3ND 5946 3NA 400 & f OCO - 6 IC90 6 1960 6 5020 6 ECCO 6 3980 6 91 NM CLI VDA 401 wNM 10CI ENA IC98 VDA 8963 801 2021 ONM tE 6 5CCB 3NM 5991 3NM CLE 6 40E 6 10CC 6 IC9E 6 596E 6 EOCE 6 ECGE 6 898E 6 tC NM CLC WON LOC VNX loCC GNI IC9E WCX' 99 6 896C 000 2OCC ENI ECCC 3NR 398C 3NN C49 6 406 6 50C6 6 lC96 6 8969 6 5029 6 5CCt 6 399t 6 DC Nn CLC v02 LOC vNA IOCC GNX 1C9C v00 896C 803 50EC SNA 19 6 C49 6 ECCC 3NO 298G 3NO 409 6 IOC9 6 IC99 6 1969 6 2029 6 ECC9 6 5989 6 91 NA CLL WOA 404 YNA 1002 SNA 8094 VDA 8964 80A 2022 SNA 98 6 CCC4 3NA 2984 3NA CIB 6 408 6 loCB 6 1C98 6 8968 6 20:8 6 ECCB 6 8988 6 ta 0 C46 VD 406 vO 10C6 80 IC96 WW 3966 8W CO 6 50:6 8k ECC6 30- 5996 30 C80 6 410 6 3060 6 fC40 6 1100 6 50CO 6 EC90 6 CB 01 CBI v01 2960 6 llI VOM 3098 801 1CLI vuM 5401 BWA 30C1 SMA CC98 301 5961 301 CC 6 CBE 6 ttE 6 tote 6 ittE 6 340E 6 CC 01 3OCE 6 EC9E 6 E963 6 CBC voM liC v01 totC 801 IC4C vuM ILOC SUA EOCC SUA EC9C 301 296C 301 tt 6 CBt 6 lit 6 tot 9 6 fC49 6 8409 6 2009 6 CC on EC9# 6 3969 6 CBC von 41C v00 809C 800 fCLC VWX ILOC SuX EOCS SuX EC9C 300 596C 30A C9 6 C89 6 4I9 6- !099 6 fC49 6 3409 6 CI OM 50C9 6 EC99 6 5969 6 CB4 VDA lid WDA 8094 SOM tct4 WWk 8404 8uM EOCL sum EC94 30M 5964 30M CB 6 CBe 6 418 6 8098 6 lCLB 6 C6 on ILOB 6 EDCB 6 EC90 6 5968 6 C86 von 456 V02 1096 SOX itt6 WWO 8406 sun 50C6 guZ 5096 30R 9L 6 060 6 4W 6 2966 30E 1060 6 ECBO 6 8410 6 dutu 6 4Ltu a cav a 8 l nk MI vol iM WDI toc 1 SON ECBI WWn illI euf oel Sgn ECLI 3OX 5401 30E
' 9'2 6 C65 6 43E 6 locE 6 1CBE 6 illE 6 E0tE 6 5Ct5 6 EIDE 6 i 9C DA C6C v0A 4ZC VDA TOCC EDA 1CBC vuA ttiL SWA E0tC SMA ECtC 30A 540C - 30A 6 99, 6 C69 6 429 6 10CD 6 1C89 6 8489 6 5099 6 ECtt 6 5409 6 9C N C6C vO 4EC VC IOCC 80 1CBC WN ILEC ON E0tC SN ECIC 3N EIOC 3N . 99 6 C69 6 429 e IOC9 6 IC09 6 3419 f 2099 6 5C49 6 5409 6 9# NN C64 WDA 424 VDA loCE 801 fCBL WNM 1434 ONM 3094 SNA CCt4 3NM 5404 3NM 98 6 C60 6 428 & IOCG 6 ICBB 6 ttiB 6 20t8 6 EELB 6 5408 6 96 NM C46 v0A 436 v01 IOC6 801 ICB6 WNM 8486 SNA 50t6 SNM EC46 3Nf 5406 3NM
- 40 6 900 6 4cO 6 8090 6 IC60 6 3420 6 2OCO 6 5C80 6 5410 6 48 NR tol von 4CI WOX 8091 00A IC63 YNX 1423 SN2 50CI 8NI ECBI 3NX lE 6 WCE 6 EllI 3NX 4CE 6 809E 6 IC6E 6 14ZE 6 EOCE 6 5C0E 6 54!E 6 4C NA DOC VOM 4CC VOM 809C SOM IC6C YNA flEC GNA EOCC SNA ECBC 3NA EliC 3N1 49 6 tot 6 4C* 6 1096 6 IC69 6 842D 6 50Ct 6 ECGD 6 E41# 6 4C NO DOC v00 4CC v02 I09C 803 IC6C YNO llEC SNA 2OCC SNI ECBC 3NA 49 6 909 6 EllC 3ND LC9 e 8099 6 lC69 6 3429 6 50C9 6 ECB9 6 5489 6 44 Nn 904 V02 4CL von 1091 SON IC64 WND 1454 GNR 3OCL SNO EC84 3NA 54I4 3NA 4B 6 tcO 6 4CB 6 8098 6 lC68 6 1438 6 EOC8 6 ECBB 6 ELIB 6 46 NA 906 WOA 4C6 VDA 5096 SOA IC66 YNA 14E6 SNA 50C6 GNA CCe6 3NA 3486 3NA 80 6 tfO 6 493 6 8040 6 8900 6 liCO 6 2090 6 EC60 6 ELEC 6 81 0 911 vO 461 VW 8011 8m 8901 vO itcI 80 5091 8W EC68 30 54:1 30
! 82 6 ttE 6 ttE 6 1043 6 ltOE 6 giCC 6 5092 6 EC6E & ELEC 6 8C 03 ' tlc v01 teC WWM SOLC SWA ttoC vDN 84CC 801 309C 0sA EC6C 301 542C 301 89 6 tit 6 499 6 8049 6 8909 6 14Ct 6 3099 6 EC69 6 ELED 6 BC OM tlc YDA ttC WWA tolC SUM ttoC W0A 14CC SOM 509C 8WM EC6C 30M circ 30M avnivom: avinmuvu A uiSnns-3I9083 #-S )syaaq E OJ 9( nnovivninwo m vimrias GiS3W80 01N ts3ilO= E
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SEG. END SEQ END SEG. END SEQ END SEO. END SEO. END SEO END SEQ END SEG END NO STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE 86 9 416 9 746 9 1076 9 1406 9 1736 9 2066 9 2396 9 2726 9 87 OX 417 A0X 747 ARU 1077 DRU 1407 A0Z 1737 004 2067 DRZ 2397 COX 2727 COX 88 9 418 9 748 9 1078 9 1408 9 1738 9 2068 9 2398 9 2728 9 89 OT 419 AOW 749 ARW 1079 DRY 1409 AOT 1739 00T 2069 DRT 2399 COT 2729 COT 90 9 420 9 750 9 1080 9 1410 9 1740 9 2070 9 2400 9 2730 9 91 OU 421 AOU 751 ARX 1081 DRX 1411 AOU 1741 DOU 2071 DRX 2401 CDU 2731 COU 92 9 422 9 752 9 1082 9 1412 9 1742 9 2072 9 2402 9 2732 9 93 Ox 423 AOZ 753 ARU 1083 BRZ 1413 AOX 1743 DOX 2073 DRV 2403 COX 2733 COX 94 9 424 9 754 9 1084 9 1414 9 1744 9 2074 9 2404 9 2734 9 95 OV 425 A0V 755 ARV 1085 DRV 1415 ADV 1745 00V 2075 DRV 2405 COV 2735 COV 96 9 426 9 756 9 1086 9 1416 9 1746 9 2076 9 2406 9 2736 9 97 0 427 AR 757 AO 1087 DR 1417 AO 1747 DR 2077 DO 2407 CO 2737 CO 98 9 428 9 758 9 1088 9 1418 9 1748 9 2078 9 2408 9 2738 9 99 OW 429 ART 759 AOY 1099 DRW 1419 A0Y 1749 DRW 2079 DOW 2409 COW 2739 COW 100 9 430 9 760 9 1090 9 1420 9 1750 9 2000 9 2410 9 2740 9 108 OT 431 ARW 761 ACT 1091 DRT 1421 ADW 1751 DRY 2001 DDT 2411 COT 2741 COT 102 9 432 9 762 9 1092 9 1422 9 1752 9 2082 9 2412 9 2742 9 103 CX 433 ARU 763 AOX 1093 DRU 1423 A0Z 1753 DRZ 2003 DOX 2413 COX 2743 COX 104 9 434 9 764 9 1094 9 1424 9 1754 9 2004 9 2414 9 2744 9 105 OT 435 ARW 765 AOW 1095 DRY 1425 ACT 1755 DRT 2085 DOT 2415 COT 2745 COT 106 9 436 9 766 9 1096 9 1426 9 1756 9 2086 9 2416 9 2746 9 107 OX 437 ARU 767 A0Z 1097 DRZ 1427 A0X 1757 DRU 2007 DOX 2417 COX 2747 COX 108 9 438 9 768 9 1098 9 1428 9 1758 9 2000 9 2418 9 2748 9 109 OU 439 ARX 769 AOU 1099 DRX 1429 AOU 1759 DRX 2089 DOU 2419 COU 2749 CDU 110 9 440 9 770 9 1100 9 1430 9 1760 9 2090 9 2420 9 2750 9
!!! OV 441 ARV 771 AOV 1101 DRV 1431 A0V 1761 DRV 2091 DOV 2421 COV 2751 COV 112 9 442 9 772 9 1102 9 1432 9 1762 9 2092 9 2422 9 2752 9 113 P 443 AP 773 AP 1103 DP 1433 AP 1763 DP 2093 DP 2423 CP 2753 CP 114 9 444 9 774 9 1104 9 1434 9 1764 9 2094 9 2424 9 2754 9 115 PT 445 APT 775 APW 1105 DPT 1435 APW 1765 DPT 2095 DPW 2425 CPT 2755 CPT 116 9 446 9 776 9 1106 9 1436 9 1766 9 2096 9 2426 9 2756 9 117 PT 447 APW 777 APT 1107 DPT 1437 APW 1767 DPW 2097 BPT 2427 CPT 2757 CPT 118 9 448 9 778 9 1108 9 1438 9 1768 9 2098 9 2428 9 2758 9 119 PU 449 APU 779 APU 1109 DPU 1439 APX 1769 DPX 2099 DPX 2429 CPU 2759 CPU 120 9 450 9 780 9 1110 9 1440 9 1770 9 2100 9 2430 9 2760 9 121 PT 451 APW 781 APW 1111 DPW 1441 APT 1771 DPT 2101 DPT 2431 CPT 2761 CPT 122 9 452 9 782 9 1112 9 1442 9 1772 9 2102 9 2432 9 2762 9 123 PU 453 APU 783 APX 1113 DPX 1443 APU 1773 DPU 2103 DPX 2433 CPU 2763 CPU 124 9 454 9 784 9 1114 9 1444 9 1774 9 2104 9 2434 9 2764 9 125 PU 455 APX 785 APU 1115 DPX 1445 APU 1775 DPX 2105 DPU 2435 CPU 2765 CPU 126 9 456 9 786 9 1116 9 1446 9 1776 9 2106 9 2436 9 2766 9 127 PV 457 APV 787 APV 1117 DPV 1447 APV 1777 DPV 2107 DPV 2437 CPV 2767 CPV 128 9 458 9 788 9 1118 9 1448 9 1778 9 2108 9 2438 9 2768 9 129 DT 459 ADT 789 AEW 1119 DDT 1449 AEW 1779 DDT 2109 DEW 2439 CDT 2769 CDT 130 9 460 9 790 9 1120 9 1450 9 1780 9 2110 9 2440 9 2770 9 131 DU 461 ADV 791 AEU 1121 DDU 1451 AEZ 1781 DDX 2111 DEX 2441 CDU 2771 CDU 132 9 462 9 792 9 1122 9 1452 9 1782 9 2112 9 2442 9 2772 9 133 DU 463 ADU 793 AEZ 1123 DDX 1453 AEU 1783 DDU 2113 DEX 2443 CDU 2773 CDU 134 9 464 9 794 9 1124 9 1454 9 1784 9 2114 9 2444 9 2774 9 135 DV 465 ADV 795 AEV 1125 DDV 1455 AEV 1785 DDV 2115 DEV 2445 CDV 2775 CDV 136 9 466 9 796 9 1126 9 1456 9 1786 9 2116 9 2446 9 2776 9 137 DDT 467 ADOT 797 AEDW 1127 DDOT 1457 AERW 1787 DDRT 2117 DERW 2447 CDOT 2777 CDOT 138 9 468 9 799 9 1128 9 1458 9 1788 9 211e 9 2448 9 2778 9 139 DOU 469 ADOU 799 AEDU 1129 DDOU 1459 AERZ 1789 DDOX 2119 DERX 2449 CDOU 2779 CDOU 140 9 470 9 000 9 1130 9 1460 9 1790 9 2120 9 2450 9 2780 9 141 DOX 471 ADOX 801 AE0Z 1131 DDOX 1461 AERU 1791 DDRU 2121 DERZ 2451 CDOX 2781 CDOX 142 9 472 9 802 9 1132 9 1462 9 1792 9 2122 9 2452 9 2782 9 143 DOV 473 ADOV 803 AEOV 1133 DDOV 1463 AERV 1793 DDRV 2123 DERV 2453 CDOV 2783 CDOV 144 9 474 9 004 9 1134 9 1464 9 1794 9 2124 9 2454 9 2784 9 145 DOT 475 ADOT BOS AERW 1835 DDRT 1465 AECW 1795 DDOT 2125 DERW 2455 CDOT 2785 CDOT 146 9 476 9 806 9 !! 36 9 1466 9 1796 9 2126 9 2456 9 2786 9 147 00X 477 ADOX 807 AERU 1137 DDRU 1467 AE0Z 1797 DDOX 2127 DERZ 2457 CDOX 2787 CDOX 148 9 478 9 808 9 1138 9 14a8 9 1798 9 2128 9 2438 9 2788 9 149 DOU 479 ADOU 809 AERZ 1139 DDRX 1469 AEOU 1799 DDOU 2129 DERX 2459 CDOU 2789 CDOU 150 9 480 9 810 9 1140 9 1470 9 1800 9 2130 9 2460 9 2790 9 151 DOV 481 ADOV 811 AERV 1141 DDRV 1471 AEDV 1801 DDOV 2121 DERV 2461 CDOV 2791 CDOV 152 9 482 9 812 9 142 9 1472 9 1802 9 2132 9 2462 9 2792 9 153 DPT 483 ADPT 813 AEPW 1143 DDPT 1473 AEPW IBv3 DDPT 2133 DEPW 2463 CDPT 2793 CDPT 154 9 484 9 814 9 1144 9 1474 9 1804 9 2134 9 2464 9 2794 9 155 DPU 485 ADPU B15 AEPU 1145 DDPU 1475 AEPZ 1805 DDPX 2135 DEPX 2465 CDPU 2795 CDPU 156 9 486 9 016 9 1146 9 1476 9 1806 9 2136 9 2466 9 2796 9 157 DPU 487 ADPU 817 AEPZ 1147 DDPX 1477 AEPU 1807 DDPU 2137 CCPX 2467 CDPU 2797 CDPU 150 9 488 9 818 9 1148 9 1478 9 1800 9 2138 9 2468 9 2798 9 159 DPV 489 ADPV 819 AEPV 1149 DDPV 1479 AEPV 1809 DDPV 2139 DEPV 2469 CDPV 2799 CDPV
, 160 9 490 9 820 9 1150 9 1480 9 1810 9 2140 9 2470 9 2000 9 161 D 491 AE 821 AD 1151 DD 1481 AE 1911 DE 2141 DD 2471 CD 2001 CD 162 9 492 9 822 9 1852 9 1482 9 1812 9 2142 9 2472 9 2002 9 163 EW 493 AET 823 ADW 1153 DDW 1483 AEY 1813 DET 2143 DDY 2473 CDW 2003 CDW 164 9 494 9 824 9 1854 9 1484 9 1814 9 2144 9 2474 9 2004 9 165 DT 495 AEW 825 ADT 1155 DDT 1485 AEW 1015 DEW 2145 DDT 2475 CDT 2005 CDT 166 9 496 9 826 9 1156 9 1486 9 1816 9 2146 9 2476 9 2006 9 167 DU 497 AEU B27 ADU 1157 DDU 1487 AEZ 1817 DEX 2147 DDX 2477 CDU 2007 CDU 168 9 498 9 028 9 1150 9 1488 9 1810 9 2148 9 2478 9 2808 9 169 DW 499 AEY e29 ADW 1159 DDY 1489 AET 1019 DET 2149 DDW 2479 CDW 2009 CDW 170 9 500 9 830 9 1160 9 1490 9 1020 9 2150 9 2480 9 2810 9 CAUTION: PRELIMINARY RESULTS-FIGURE 4-5 (Sheet 4 of 6) mPORTANT u=CERTAmfiEs DESCRIBED IN SCCTION 2
SEO END SEQ END SEQ END SEO. END SEQ. END SEQ. END SEO. END SEO. END SEO. END NO. STATE NO. STATE NO. STATE NO. STATE NO STATE NO. STATE NO STATE NO. STATE NO. STATE 171 DX 501 AEX 831 ADX !!61 DDZ 1491 AEX 1821 DEU 2151 DDZ 2481 CDX 2811 CDX 172 9 502 9 832 9 1862 9 1492 9 1822 9 2152 9 2482 9 2812 9 173 DU 503 AEZ 833 ADU 1863 DDX 1493 AEU 1823 DEX 2153 DDU 2483 CDU 2813 CDU 174 9 504 9 834 9 1864 9 1494 9 1824 9 2154 9 2484 9 2014 9 175 DV 505 AEV 835 ADV 1165 DDV 1495 AEV 1825 DEV 2155 DDV 2485 CDV 2815 CDV 176 9 506 9 836 9 1166 9 1496 9 1826 9 2156 9 2486 9 2816 9 177 DO 507 AE0 837 ADO 1167 0D0 1497 AER 1027 DER 2157 DDR 2487 CD0 2817 CDO 178 9 508 9 838 9 1168 9 1498 9 1828 9 2158 9 2488 9 2818 9 179 DOW 509 AEOT 839 ADOW 1169 DDOW 1499 AERY 1829 [ERT 2159 DDRY 2489 CDOW 2819 CDOW 100 9 510 9 840 9 1870 9 1500 9 1830 9 2160 9 2490 9 2820 9 181 DOT 511 AE0W 841 ADOT 1171 DDOT 1501 AERW 1831 DERY 2161 DDRT 2491 CDOT 2821 CDOT 182 9 512 9 842 9 1172 9 1502 9 1832 9 2162 9 2492 9 2822 9 183 000 513 AEOU 843 ADOU 1173 DDOU 1503 AERZ 1833 DERX 2163 DDRX 2493 CDOU 2823 CDOU 184 9 514 9 844 9 1174 9 1504 9 1834 9 2164 9 2494 9 2824 9 185 DOY 515 AEOY 845 AD0Y 1175 DDOY 1505 AERT 1835 DERW 2165 DDRW 2495 CD0Y 2825 CDOY 186 9 516 9 846 9 1176 9 1506 9 1836 9 2166 9 2496 9 2826 9 187 DOZ 517 AEDX 847 ADOZ 1177 DCO2 1507 AERX 1837 DERU 2167 DDRZ 2497 CDOZ 2827 CDOZ 188 9 518 9 848 9 1178 9 1508 9 1838 9 2168 9 2498 9 2828 9 139 DOX 519 AE02 849 ADOX 1179 DDOX 1509 AERU 1839 DERZ 2169 DDRU 2499 CDOX 2829 CDOX
- 19) 9 520 9 850 9 1100 9 1510 9 1840 9 2170 9 2500 9 2830 9 191 DOV 521 AE0V 851 ADOV 1181 DDOV 1511 AERV 1841 DERV 2171 DDRV 2501 CDOV 2831 CDOV 192 9 522 9 852 9 1182 9 1512 9 1842 9 2172 9 2502 9 2832 9 193 DO 523 AER 853 ADO 1183 DDR 1513 AE0 1843 DER 2173 DD0 2503 CD0 2833 CDO 194 9 524 9 854 9 1184 9 1514 9 1844 9 2174 9 2504 9 2834 9 195 00Y 525 AERT 855 AD0Y 1185 DDRW 1515 AE0Y 1D45 DERW 2175 BD0Y 2505 CD0Y 2035 CDOY 196 9 526 9 856 9 !!B6 9 1516 9 1846 9 2176 9 2506 9 2836 9 197 DOT 327 AERW B57 ADOT 1187 DDRT 1517 AEDW 1847 DERY 2177 DDOT 2507 CDOT 2837 CDOT 198 9 528 9 858 9 1888 9 1518 9 1848 9 2178 9 2508 9 2838 9 199 DOX 529 AERU 859 ADOX 1189 DDRU 1519 AE02 1849 DERZ 2179 DDOX 2509 CDOX 2839 CDOX 200 9 530 9 860 9 1190 9 1520 9 1850 9 2180 9 2510 9 2840 9 201 DOW 531 AERY 861 ADCW 1191 DDRY 1521 AEDT 1851 DERT 2181 DDOW 2511 CDOW 2841 CDOW 202 9 532 9 862 9 1192 9 1522 9 1852 9 2182 9 2512 9 2842 9 203 DOZ 533 AERX 863 ADOZ 1193 DDRZ 1523 AE0X 1853 DERU 2183 DDOZ 2513 CDOZ 2843 CDOZ 204 9 534 9 864 9 1194 9 1524 9 1854 9 2184 9 2514 9 2844 9 205 DOU 535 AERZ 865 ADOU 1195 DDRX 1525 AEOU 1855 DERI 2185 DDOU 2515 CDOU 2845 CDOU 206 9 536 9 866 9 1196 9 1526 9 1856 9 2186 9 2516 9 2046 9 207 DOV 537 AERV 867 ADOV 1197 DDRV 1527 AE0V 1857 DERV 2187 DDOV 2517 CDOV 2847 CDOV 208 9 538 9 868 9 1198 9 1528 9 1858 9 2188 9 2518 9 2848 9 209 DP 539 AEP 869 ADP 1199 DDP 1529 AEP 1859 DEP 2189 DDP 2519 CDP 2049 CDP 210 9 540 9 870 9 1200 9 1530 9 1860 9 2190 9 2520 9 2850 9 211 DPW 541 AEPT 871 ADPW 1201 DDPW 1531 AEPY 1861 DEPT 2191 DDPY 2521 CDPW 2851 CDPW 212 9 542 9 872 9 1202 9 1532 9 1862 9 2192 9 2522 9 2852 9 213 CPT 543 AEPW 873 ADPT 1203 DDPT 1533 AEPW 1863 DEPW 2193 DDPT 2523 CDPT 2853 CDPT 214 9 544 9 874 9 1204 9 1534 9 1864 9 2194 9 2024 9 2854 9 215 DPU 545 AEPU 875 ADPU 1205 DDPU 1535 AEPZ 1865 DEPX 2195 DDPX 2525 CDPU 2855 CDPU 216 9 546 9 876 9 1206 9 1536 9 1866 9 2196 9 2526 9 2856 9 217 CPW 547 AEPY 877 ADPW 1207 DDPY 1537 AEPT 1867 DEPT 2197 DDPW 2527 CDPW 2857 CDPW 218 9 548 9 878 9 1208 9 1538 9 1868 9 2198 9 2528 9 2858 9 219 DPX 549 AEPX 879 ADPX 1209 DDPZ 1539 AEPX 1869 DEPU 2199 DDPZ 2529 CDPX 2859 CDPX 220 9 550 9 880 9 1210 9 1540 9 1870 9 2200 9 2530 9 2860 9 221 DPU 551 AEPZ 881 ADPU 1211 DDPX 1541 AEPO 1871 DEPX 2201 DDPU 2531 CDPU 2861 CDPU 222 9 552 9 882 9 1212 9 1542 9 1872 9 2202 9 2532 9 2062 9 223 DPV 553 AEPV 883 ADPV 1213 DDPV 1543 AEPV 1873 DEPV 2203 DDPV 2533 CDPV 2003 CDPV 224 9 554 9 884 9 1214 9 1544 9 1874 9 2204 9 2534 9 2864 9 225 F 575 AFT 885 AFW 1215 DF 1545 AG 1875 DGT 2205 DGW 2535 CF 2865 CF 226 9 556 9 886 9 1216 9 1546 9 1976 9 2206 9 2536 9 2866 9 227 FU 557 AFU 887 AFU 1217 DFU 1547 AGX 1877 DGX 2207 DGX 2537 CFU 2867 CFU 228 9 558 9 888 9 1218 9 1548 9 1878 9 2208 9 2538 9 2868 9 229 FX 559 AFX 889 AFX 1219 DFX 1549 AGU 1879 DGU 2209 DGZ 2539 CFX 2869 CFX 230 9 560 9 890 9 1220 9 1550 9 4880 9 2210 9 2540 9 2870 9 231 FV 561 AFV 891 AFV 1221 DFV 1551 AGV 1881 DGV 2211 DGV 2541 CFV 2871 CFV 232 9 562 9 892 9 1222 9 1552 9 1882 9 2212 9 2542 9 2872 9 233 FP 563 AFPT 893 AFPW 1223 DFP 1553 AGP 1883 DGPT 2213 DGPW 2543 CFP 2873 CFP 234 9 564 9 894 9 1224 9 1554 9 1884 9 2214 9 2544 9 2874 9 235 FPU 565 AFPU 895 AFPU 1225 DFPU 1555 AGPX 1885 DGPX 2215 DGPX 2545 CFPU 2875 CFPU 236 9 566 9 896 9 1226 9 1556 9 1886 9 2216 9 2546 9 2876 9 C37 FPX 567 AFPX 897 AFPX 1227 DFPX 1557 AGPU 1887 DGPU 2217 DCPZ 2547 CFPX 2877 CFPX 238 9 568 9 898 9 1228 9 1558 9 1888 9 2218 9 2548 9 2878 9 239 FPV 569 AFPV 899 AFPV 1229 DFPV 1559 AGPV 1889 DGPV 2219 DCPV 2549 CFPV 2079 CFPV 240 9 570 9 900 9 1230 9 1560 9 1890 9 2220 9 2550 9 2880 9 241 D 571 AE 901 AE 1231 DE 1561 AD 1891 DD 2221 DD 2551 CD 2001 CD 242 9 572 9 902 9 1232 9 1562 9 1892 9 2222 9 2552 9 2882 9 243 DW 573 AET 903 AEY 1233 DET 1563 ADW 1893 DDW 2223 DDY 2553 CDW 2883 CDW 244 9 574 9 904 9 1234 9 1564 9 1894 9 2224 9 2554 9 2884 9 245 DW 575 AEY 935 AET 1235 DET 1565 ADW 1895 DDY 222S DDW 2555 CDW 2885 CDW 246 9 576 9 906 9 1236 9 1566 9 1896 9 2226 9 2556 9 2986 9 247 DX 577 AEX 907 AEX 1237 DEU 1567 ADX 1897 DDZ 2227 DDZ 2557 CDX 2887 CDX 248 9 578 9 408 9 1238 9 1568 9 1898 9 2228 9 2558 9 2888 9 249 DT 579 AFW 909 AFW 1239 DEW 1569 ADT 1899 DDT 2229 DDT 2559 CDT 2099 CDT 250 9 500 9 910 9 1240 9 1570 9 1900 9 2230 9 2560 9 2890 9 FIGURE 4-5 (Sheet 5 of 6) CAUTION: PREUMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRIBED IN ;ECTION 2 4-84
SEQ. END SEQ. END SEO. END GEO. END SEO. END SEO END SEG. END SEG. END SEO. END NO. STATE NO STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE 231 DU 581 AEU 911 AEZ 1241 DEX 1571 ADU 1901 DDU 2231 DDX 2561 CDU 2891 CDU 252 9 582 9 912 9 1242 9 1572 9 1902 9 2232 9 2562 9 2892 9 253 DU 583 AEZ 913 AEU 1243 DEX 1573 ADU 1903 DDX 2233 DDU 2563 CDU 2893 CDU 254 9 584 9 914 9 1244 9 1574 9 1904 9 2234 9 2564 9 2094 9 4 255 DV 585 AEV 915 AEV 1245 DEV 1575 ADV 1905 DDV 2235 DDV 2565 CDV 2895 LDV 256 9 586 9 916 9 1246 9 1576 9 1906 9 2236 9 2566 9 2896 9 257 DO 587 AE0 917 AER 1247 DER 1577 ADO 1907 000 2237 DDR 2567 CD0 2897 CD0 258 9 588 9 910 9 1248 9 1578 9 1900 9 2238 9 2568 9 2898 9 259 DOW 589 AEDT 919 AERY 1249 DERT 1579 ADOW 1909 DDOW 2239 DDRY 2569 CDOW 2099 CDOW 260 9 590 9 920 9 1250 9 1580 9 1910 9 2240 9 2570 9 2900 9 261 D0Y 591 AEOY 921 AERT 1251 DERW 1581 AD0Y 1911 DD0Y 2241 DDRW 2571 CD0Y 2901 CD0Y 262 9 592 9 922 9 1252 9 1582 9 1912 9 2242 9 2572 9 2902 9 263 DOZ 593 AE0X 923 AERX 1253 DERU 1583 ADOZ 1913 DDOZ 2243 DDRZ 2573 CDOZ 2903 CDOZ 244 9 594 9 924 9 1254 9 1584 9 1914 9 2244 9 2574 9 2904 9 265 DOT 595 AEDW 925 AERW 1255 DERY 1505 ADOT 1915 DDOT 2245 DDRT 2575 CDOT 2905 CDOT 1 2&& 9 596 9 926 9 1256 9 1586 9 1916 9 2246 9 2576 9 2906 9 267 DOU 597 AEOU 927 AERZ 1257 DERX 1587 ADOU 1917 DDOU 2247 DDRX 2577 CDOU 2907 CDOU 268
- 598 9 928 9 1258 9 1588 9 1918 9 2248 9 2578 9 2900 9 i 229 DOX 599 AE0Z 929 AERU 1259 DERZ 1589 ADOX 1919 DDOX 2249 DDRU 2579 CDOX 2909 CDOX 270 9 600 9 930 9 1260 9 1590 9 1920 9 2250 9 2580 9 2910 9 271 DOV 601 AE0V 931 AERV 1261 DERV 1591 ADOV 1921 DDOV 2251 DDRV 2581 CDOV 2911 CDOV 272 9 602 9 932 9 1262 9 1592 9 1922 9 2252 9 2582 9 2912 9 273 DO 603 AER 933 AE0 1263 DER 1593 ADO 1923 DDR 2253 DD0 2583 CD0 2913 CD0 274 9 604 9 934 9 1264 9 1594 9 1924 9 2254 9 2584 9 2914 9 275 DOY 605 AERT 935 AE0Y 1265 DERW 1595 AD0Y 1925 DDRW 2255 DD0Y 2585 CD0Y 2915 CD0Y 276 9 606 9 936 9 1266 9 1596 9 1926 9 2256 9 2586 9 2916 9 j 277 DOW 607 AERY 937 AEDT 1267 DERT 1597 ADOW 1927 DDRY 2257 DDOW 2587 CDOW 2917 CDOW 4
278 9 608 9 938 9 1268 9 1598 9 1928 9 2258 9 2588 9 2918 9 279 DOZ 609 AERX 939 AE0X 1269 DERU 1599 ADOZ 1929 Dkt 2259 DDOZ 2589 CDOZ 2919 CDOZ 280 9 610 9 940 9 1270 9 1600 9 1930 9 2260 9 2590 9 2920 9 281 DOT 611 AERW 941 AE0W 1271 DERY 1601 ADOT 1931 DDRT 2261 DDOT 2591 CDOT 2921 CDOT 283 9 612 9 942 9 1272 9 1602 9 1932 9 2262 9 2592 9 2922 9 283 DOX 613 AERU 943 AE02 1273 DERZ 1603 ADOX 1933 DDRU 2263 DDOX 2593 CDOX 2923 CDOX 284 9 614 9 944 9 1274 9 1604 9 1934 9 2264 9 2594 9 2924 9 235 000 615 AERZ 945 AEOU 1275 DEPX 1605 AD00 1935 DDRX 2265 DDOU 2595 CDOU 2925 CDOU 284 9 616 9 946 9 1276 9 1606 9 1936 9 2266 9 2596 9 2926 9 287 DOV 617 AERV 947 AEDV 1277 DERV 1607 ADOV 1937 DDRV 2267 DDOV 2597 CDCV 2927 CDOV 283 9 618 9 948 9 1278 9 1608 9 1938 9 2268 9 2598 9 2928 9 289 DP 619 AEP 949 AEP 1279 DEP 1609 ADP 1939 DDP 2269 DDP 2599 CDP 2929 CDP t 290 9 620 9 950 9 1280 9 1610 9 1940 9 2270 9 2600 9 2930 9 291 DPW 621 AEPT 951 AEPY 1281 DEPT 1611 ADPW 1941 DDPW 2271 DDPY 2601 CDPW 2931 CDPW 292 9 622 9 952 9 1282 9 1612 9 1942 9 2272 9 2602 9 2932 9 293 DPW 623 AEPY 953 AEPT 1283 DEPT 1613 ADPW 1943 DDPY 2273 DDPW 2603 CDPW 2933 CDPW 294 9 624 9 954 9 1284 9 1614 9 1944 9 2274 9 2604 9 2934 9 295 DPX 625 AEPX 955 AEPX 1285 DEPU 1615 ADPX 1945 DDPZ 2275 DDPZ 2605 CDPX 2935 CDPX 276 9 626 9 956 9 1286 9 1616 9 1946 9 2276 9 2606 9 2936 9 297 DPT 627 AEPW 957 AEPW 1287 DEPW 1617 ADPT 1947 DDPT 2277 DDPT 2607 CDPT 2937 CDPT 298 9 628 9 958 9 1288 9 1618 9 1948 9 2278 9 2608 9 2938 9 299 CPU 629 AEPU 959 AEPZ 1289 DEPX 1619 ADPU 1949 DDPU 2279 DDPX 2609 CDPU 2939 CDPU 300 9 630 9 960 9 1290 9 1620 9 1950 9 2200 9 2610 9 2940 9 301 DPU 631 AEPZ 961 AEPU 1291 DEPX 1621 ADPU 1951 DDPX 2281 DDPU 2611 CDPU 2941 CDPU 302 9 632 9 962 9 1292 9 1622 9 1952 9 2282 9 2612 9 2942 9 303 DPV 633 AEPV 963 AEPV 1293 DEPV 1623 ADPV 1953 DDPV 2233 DDPV 2613 CDPV 2943 CDPV , 304 9 634 9 964 9 1294 9 1624 9 1954 9 2284 9 2614 9 2944 9 305 F 635 AFT 965 AG 1295 BGT 1625 AFW 1955 BF 2285 DCW 2615 CF 2945 CF 306 9 636 9 966 9 1296 9 1626 9 1956 9 2286 9 2616 9 2946- 9 4 307 FX 637 AFX 967 AGU 1297 DGU 1627 AFX 1957 DFX 2287 DGZ 2617 CFX 2947 CFX 308 9 638 9 968 9 1298 9 1628 9 1958 9 2288 9 2618 9 2948 9 309 FU 639 AFU 969 AGX 1299 DGX 1629 AFU 1959 DFU 2289 DGX 2619 CFU 2949 CFU 310 9 640 9 970 9 1300 9 1630 9 1960 9 2290 9 2620 9 2950 9 311 FV 641 AFV 971 ACV 1301 DGV 1631 AFV 1961 DFV 2291 DGV 2621 CFV 2951 CFV 312 9 642 9 972 9 1302 9 1632 9 1962 9 2292 9 2622 9 2952 9 313 FP 643 AFPT 973 AGP 1303 DGPT 1633 AFPW 1963 DFP 2293 DGPW 2623 CFP 2953 CFP 314 9 644 9 974 9 1304 9 1634 9 1964 9 2294 9 2624 9 2954 9 315 FPX 645 AFPX 975 AGPU 1305 DGPU 1635 AFPX 1965 DFPX 2295 DGPZ 2625 CFPX 2955 CFPX 316 9 646 9 976 9 1306 9 1636 9 1966 9 2296 9 2626 9 2956 9 3i7 FPU 647 AFPU 977 AGPX 1307 DGPX 1637 AFPU 1967 DFPU 2297 DGPX 2627 CFPU 2957 CFPU 318 9 648 9 978 9 1308 9 1638 9 1968 9 2298 9 2628 9 2950 9 319 FPV 649 AFPV 979 AGPV 1309 DGPV 1639 AFPV 1969 DFPV 2299 DGPV 2629 CFPV 2959 CFPV 320 9 650 9 980 9 1310 9 1640 9 1970 9 2300 9 2630 9 2960 9 331 FU 651 AGX 981 AFU 1311 DGX 1641 AFU 1971 DGX 2301 DFU 2631 CFU 2961 CFU 322 9 652 9 982 9 1312 9 1642 9 1972 9 2302 9 2632 9 323 FV 2962 9 653 AGV 983 AFV 1313 DGV 1643 AFV 1973 DGV 2303 DFV 2633 CFV 2963 CFV 324 9 654 9 984 9 1314 9 1644 9 1974 9 2304 9 2634 9 325 FPU 2964 9 655 AGPX 985 AFPU 1315 DGPX 1645 AFPU 1975 DGPX 2305 DFPU 2635 CFPU 2965 CFPU 326 9 656 9 986 9 1316 9 1646 9 1976 9 2306 9 2636 9 2966 9 327 FPV 657 AGPV 987 AFPV 1317 DGPV 1647 AFPV 1977 DGPV 2307 DFPV 2637 CFPV 2967 CFPV 328 9 658 9 988 9 1318 9 1648 9 1978 9 2308 9 2630 9 2968 9 329 H 659 AH 989 AH 1319 DH 1649 AH 1979 DH 2309 DH 2639 CH 2969 CH 330 9 660 9 990 9 1320 9 1650 9 1980 9 2310 9 2640 9 2970 9 l i l CAUTION: PRELIMINARY RESULTS-t FIGURE 4-5 (Sheet 6 cf 6) ,uPORTA=T u=CERTAi= TIES DESCRISED IN SECTION 2 I t 4-85 t
(fCD4D: GF GUARANTEED FAILURE
< m o ,.
o o o g g g 3 CAUTION: PRELIMINARY RESULTS-
, Sh Dh Oh 4MPORTANT UNCE RTAINTIES K < m o < m o < m u d h h DESCR18ED IN SECTION 2 8z 8z z a 5E BE BE ~o E z z z 5 g wk G 5 5 5 $5 $$ $5 lgE E $ (N $6 $5 dh a gs ! T s eE "E 99 U@m E lE ME s! s!
E$ 8 5 g S8 E5 E uf d b b Um U$ h5 5 b5 UI Um Om de ss wA e w: Cs sA ss EV 8 y e d8 gy it EA Es EC CA CC SC % 1 1 2 1 3 3 I 4 4 5 5 I 6 6 g 7 7 8 8 9 9 g 10 10 11 11 g 12 12 13 13 l 14 14 g 15 15 16 16 17 XFR1 17 18 XFR1 33 l _ _ _ _ _ _ _ _ _ . 19 XFR1 49 20 XFR1 65 1 ---------* 21 XFR1 81 22 XFR1 97 i _________. 23 XFR1 113 F 24 129 w[- w 25 130 g 26 131 27 132 w 28 133 g 29 134 e 30 135 31 136 w----------- 32 XFR2 137 y_.______. 33 XFR2 145 w----------- 34 XFR2 153 w e 35 161 [ ' g 36 37 162 163 38 164 r g 39 165 40 166 41 167 g 42 168 43 XFR3 169 y _ _ _ _ . 44 XFR3 177 l - - - - - - - - - - 45 XFR3 185 w-w e-r 46 193 47 194 F-r 48 195 g 49 196 y_y- _-. -_ _ . 50 XFR5 197 FIGURE 4-6. AUXILIARY SYSTEMS EVENT TREE FOR LOSS OF 0FFSITE POWER CONDITION (Sheet 1 of 5) 4-86
LEGEND: GF GUARANTEE 3 FA: LURE
< m o o o o E E E
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w a E E g 83 3 { Z Z Ob Ob Ob g g ,5, i i i ge gs gi ge ge g5 gE g5 g5 $! 8 E $8 E
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s il NUW "l w 58 l it ss EA ES EC WA WB WC CA CS CC SA $8 SC EV I F 51 2 01 v[ ' g 52 53 202 203 54 204 55 205
' 206
!!6 57 207 58 108 w-------- 59 XFR4 209 F-------- 60 X5R4 217 g
w-------- 61 XFR4 225 y w-w -w ; 62 233 63 234 w y 64 235 65 236 F-r- - - - - - - - 66 XFR6 237 y-w 67 2 41 w-w[ 68 69 242 243 l 70 244 w- r- - - - - - 71 XFR7 245 y-w-r-w-r-e-w- 72 240 l
- ---------------------- 73 XFR8 250
' ------------------------ 74 Xm8 4H I ------------------------ 75 Xm8 74
------------------------ 76 XFR8 997 I ------------------------ 77 XFR8 1246 78 XFR8 1405 I ---_- _--- ------ _----_--
--- -- ---------- --- -- 79 xma 1744 y-y-y------------------------- 80 XFR8 1993 CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCR18ED IN SECTION 2 FIGURE 4-6 (Sheet 2 of 5) 4-87
r SEQ END SEQ END SEO END SEG. END SEQ. END BEO. END SE Q. END SE Q. END SEO END NO. STATE NO. STATE NO. STATE NO. STATE NO- STATE NO. STATE NO. STATE NO. STATE No. STATE I AUX 250 A 499 A 748 B 997 A 1246 B 1493 3 2 9 251 9 SOC 9 749 9 1744 C 1993 C 998 9 3247 9 8496 9 3745 9 1994 9 3 T 232 AT Sol AW 750 BT 999 AW 1248 BT 4 9 I497 BW 1746 CT 1995 CT 253 9 502 9 758 9 2000 9 1249 9 1498 9 1747 9 5 7 254 AW 1996 9 S03 AT 752 BT 1001 AW 1250 BW 1499 BT 1748 CT 1997 CT
& 9 2SS 9 504 9 753 9 8002 9 1238 9 1500 9 1749 9 1998 9 7 U 256 AU SOS AU 754 BU 1003 AX 1252 BK 1501 BX 1750 CU 1999 CU 8 9 237 9 506 9 753 9 2004 9 3253 9 9 7 1502 9 1751 9 2000 9 258 AW S07 AW 756 BW 1003 AT 1254 ST 1503 BT 1752 CT 2001 CT 10 9 259 9 508 9 757 9 1006 9 1255 9 1504 9 11 U 1753 9 2002 9 260 AU SO9 AX 758 BX 1007 AU 1256 BU 1505 BX 1754 CU P903 CU
$2 9 268 9 510 9 739 9 1008 9 1257 9 1506 9 13 U 1755 9 2004 9 262 AX Sit AU 760 BX 1009 AU 1258 BX 1507 BU 1756 CU 14 9 263 9 Sl2 9 2005 CU 761 9 1010 9 1259 9 1508 9 1757 9 2006 9 13 V 264 AV S13 AV 762 BV 1011 AV 1260 BV 1509 BV 16 9 26S 9 1758 CV 2007 CV Sie 9 763 9 1012 9 1261 9 1510 9 1759 9 2008 9 17 N 266 AN 313 AG 764 BN 8013 AG 1262 BN 8511 BQ 18 9 267 9 1760 CN 2009 CN 316 9 765 9 1014 9 1263 7 1512 9 8768 9 2010 9 19 NT 268 ANT S17 AQW 766 BNT 101S AQW 1264 BNT 1513 BOW 8762 CNT 2011 CNT 20 9 269 9 318 9 767 9 1016 9 1265 9 1584 9 23 NW 1763 9 2012 9 270 ANW S19 AQT 768 BNW lot? AQY 8266 BNY 158S BQT 1764 CNW 2013 CNW 22 9 271 9 520 9 769 9 1018 9 1267 9 1516 9 1765 9 2014 9 23 NU 272 ANU S21 AQU 770 BNU 1019 AG2 1268 BNX IS$7 BQI 1766 CNU 2015 CNU 24 9 273 9 S22 9 771 9 1020 9 1269 9 1518 9 1767 9 23 NW 274 ANW
. 2016 9 S23 AQY 772 BNY 1028 AQT 1270 BNW 1919 BQT 1768 CNW 2017 CNW 26 9 275 9 324 9 773 9 1022 9 1271 9 1522 9 1769 9 2018 9 27 NU 276 ANU S23 AQ2 774 BNX 1023 AQU 1272 BNU 1521 BQX 1770 CNU 2019 CNU 28 9 277 9 526 9 775 9 8024 9 1273 9 8522 9 29 NX 1778 9 2020 9 278 ANX $27 AQX 776 DNZ 102S AQX 1274 BNZ 1523 BQU 1772 CNX 2021 CNX 30 9 279 9 528 9 777 9 1026 9 1273 9 1524 9 1773 9 38 NV 280 ANV 2022 9 S29 AQV 778 BNV 1027 AQV 1276 BNV $$23 80y 1774 CNV 2023 CNV 32 9 281 9 S30 9 779 9 1028 9 1277 9 1526 9 33 N 1775 9 2024 9 282 AG S31 AN 7BO BN 1029 AQ 1278 BQ $327 DN 8776 CN 2025 CN 34 9 283 9 532 9 781 9 1030 9 1279 9 1528 9 33 NW 1777 9 2026 9 284 AQT S33 ANW 782 BNW $031 AQY 1280 BQT IS29 BNY 1778 CNW 2027 CNW 36 9 2GS 9 S34 9 793 9 1032 9 1281 9 1530 9 ;779 9 37 NT 286 AQW 2028 9
$35 ANT 764 BNT 1033 AQW 1282 BQW 1531 BNT 1780 CNT 2029 CNT 30 9 287 9 S36 9 785 9 1034 9 1283 9 1532 9 39 NU 1781 9 2030 9 288 AQU S37 ANU 786 BNU 1035 AQZ 1284 BQX $333 BNI 1782 CNU 2033 CNU 40 9 289 9 538 9 787 9 1036 9 128S 9 8534 9 1783 9 el NW 2032 9 290 AQY 539 ANW 7e8 BNY 1037 AGT 1286 BQT 8535 BNW 1784 CNW 2033 CNW 42 9 291 9 S40
- 789 9 1038 9 1287 9 1536 9 43 NX 1785 9 2034 9 292 AQX 541 ANX 790 BN1 1039 AQX 1283 BQU 1537 BNZ 1786 CNX 2035 CNX 44 9 293 9 S42 9 791 9 1040 9 1289 9 1538 9 1787 9 2036 9 45 NU 294 AQZ S43 ANU 792 BNI 3041 AQU $290 BQX 1939 BNU 1788 CNU 2037 CNU 46 9 295 9 S44 9 793 9 1042 9 1291 9 1540 9 47 NY 8789 9 2038 9 296 AQV S45 ANV 794 BNV 1043 AQV 1292 BQV 1548 BNV 1790 CNV 2039 CNV 48 9 297 9 S46 9 795 9 8044 9 1293 9 1542 9 49 0 1791 9 2040 9 298 A0 547 A0 796 B0 104S AR 1294 BR 1543 BR 3792 CD 2048 CD SO 9 299 9 548 9 797 9 5046 9 1295 9 3544 9 3793 9 58 07 2042 9 300 ACT 549 A0W 798 BOT 1047 ARW 3296 BRT 1545 BRY 1794 COT 2043 COT
$2 9 308 9 550 9 799 9 1048 9 1297 9 1546 9 1795 9 2044 9
$3 07 302 A0W SSI ACT 800 907 1049 ARW 8298 BRY 1547 BRT 1796 COT 204S COT 54 9 303 9 S$2 9 B;l 9 10$O 9 1299 9 IS48 9 SS OU 1797 9 2046 9 304 AOU $53 AOU B02 BOU 1051 ARI 1300 BRM 1549 BRI 3798 CDU 2047 CDU 56 9 305 9 SS4 9 803 9 1052 9 1308 9 ISSO 9 57 OW 1799 9 2048 9 306 A0Y SSS AOY 804 BOW 1053 ART 8302 BRW IS$1 PRW 1800 COW 2049 COW
$8 9 307 9 S$6 9 805 9 1054 9 1303 9 1552 9 1801 9 S9 OX 2000 9 308 A0X SS7 A02 806 BOX 1055 ARU 1304 BRU ISS3 BRI 1802 COI 205l COI 60 9 309 9 SSB 9 807 9 3056 9 8303 9 ISS4 9 1803 9 61 CX 20S2 9 310 A02 SS9 AOX B08 BOX 1037 ARU 8306 BRZ 1535 BRU 1804 COX 20S3 COI 62 9 318 9 360 9 809 7 8038 9 8307 9 RSS6 9 ISOS 9 2054 9 63 OV 312 ADV 361 ADV 810 BOV 1059 ARV 1300 BRV ISS7 BRV 1806 COV 20$$ COV 64 9 313 9 S62 9 818 9 8060 9 1309 9 1558 9 1807 9 20$6 9 63 N 314 AQ 363 AG B12 BQ 1061 AN 3310 DN IS59 BN 8808 CN 66 9 20S7 CN 313 9 364 9 813 9 8062 9 1313 9 1560 9 1809 9 20S8 9 67 NW 316 AQT 565 AQY B14 B07 1063 ANW 1312 BNW $568 DNY 1810 CNW 68 9 317 9 20$9 CNW 566 9 BBS 9 1064 9 1313 9 1562 9 1811 9
, 69 NW 318 AQY 567 AQT B16 807 106$ ANW 2060 9 1314 BNY $$63 BNW 1812 CNW 2068 CNW 70 9 319 9 $68 9 817 9 1066 9 1315 9 1564 9 1813 9 2062 9 78 NX 320 AQX S69 AQX 818 BQU 8067 ANX 1316 BNI 1565 BN2 1814 CNN 2063 CNN 7P 9 328 9 570 9 819 9 1068 9 1317 9 1566 9 191S 9 2064 9 73 NT 322 AQW 571 AQW B20 SQW 1069 ANT 8318 BNT IS67 BNT 1816 CNT 2065 CNT 74 9 323 9 572 9 828 9 1070 9 8319 9 1568 9 1817 9 2066 9 75 NU 324 AQU S73 AGI B22 BQX 1071 ANU 1320 BNU 1569 DNX 1818 CNU 2067 CNU 76 9 323 9 574 9 823' 9 1072 9 1328 9 1570 9 1819 9 2068 9 77 NU 326 AQ2 575 AQU 824 BQX 1073 ANU $322 BNX IS71 BNU 1820 CNU 2C69 CNU 78 9 327 9 576 9 825 9 1074 9 1323 9 1572 9 1828 9 2070 9 79 NV 328 AQV S77 AQV B26 BQV 107$ ANV 8324 BNV 1573 BNV 1822 CNV B0 9 329 9 2071 CNV 578 9 827 9 1076 9 8325 9 1574 9 1823 9 2072 9 Bt 0 330 A0 S79 AR 828 BR 8077 A0 1326 BO $$75 BR 8824 CD 2073 CD B2 9 331 9 58u 9 829 9 1078 9 1327 9 1576 9 1825 9 83 01 2074 9 332 AOT 581 ARW 830 BRT 1079 A0W 1328 BOT 1577 BRY 1826 COT 207S COT 84 9 333 9 $82 9 831 9 8080 9 1329 9 1578 9 1927 9 2076 9 BS OW 334 A0Y $83 ART 832 BRW 1088 AOY 1330 BCW 1579 BRW $828 COW 2077 COW CAUTION: PRE LIMINARY RESULTS-FIGURE 4-6 (Sheet 3 of 5) IWORTANT UNCERTAINTIES CESCRISED IN SECTION 2 4-88 l
SEG. END SEO END SES END SEG. END SEO. END SEG. ENO SEG. END SEG. END SEG. END No. STATE NO. STATE NO. STATE NO- STATE NO. STATE NO. STATE NO. STATE NO. STATE NO. STATE 86 9 335 9 SO4 9 833 9 1002 9 1331 9 1580 9 1829 9 2079 9 87 0E 336 ADI SOS ARU B34 BRU 1003 A02 1332 BOX 1581 BR2 1830 COI 2079 COI es 9 337 9 586 9 B33 9 1084 9 8333 9 1582 9 1831 9 2000 9 99 Of 338 AOW $87 ARW B36 BRY 108S ACT 8334 BOT 1583 BRT 1832 COT 2001 COT 90 9 339 9 588 9 G37 9 1086 9 1333 9 1584 9 1833 9 2082 9 93 OU 340 AOU S89 ARI S39 BRI 1087 AOU 8336 BOU ISBS BRI 1834 CDU 2083 CDU 92 9 341 9 590 9 B39 9 1089 9 3337 9 ISB6 9 1833 9 2004 9 93 OI 342 A0Z S98 ARU G40 BRZ 1089 AOI 1339 BOX 1587 BRU 1836 COI 2005 COI 94 9 343 9 392 9 841 9 5090 9 1339 9 1588 9 1937 9 2006 9 95 OV 344 ADV 593 ARV G42 BRV 1091 ADV 1340 BOV 8589 BRV 1038 COV 2007 COV 96 9 34S 9 594 9 843 9 1092 9 1348 9 1590 9 1839 9 2000 9 97 0 346 AR 593 A0 844 BR 3093 A0 1342 BR 1591 B0 1940 CD 2009 CD 98 9 347 9 596 9 845 9 1094 9 1343 9 1592 9 1841 9 2090 9 99 OW 348 ART $97 AQY G46 BRW 109S A0Y 1344 BRW 8593 i OW 1942 COW 2091 COW 100 9 349 9 590 9 847 9 1096 9 1345 9 1594 9 1843 9 2092 9 808 GT 3S0 ARW 599 ACT 849 BRT. 1097 A0W 1346 BRV B593 BOT 1844 COT 2093 COT 102 9 351 9 600 9 B49 9 3099 9 1347 9 IS96 9 184S 9 2094 9 103 OI 352 ARU 608 AOI 850 BRU 8099 A02 1348 BRZ 1597 BOI 1846 COI 2093 COI 104 9 353 9 602 9 858 9 8100 9 8349 9 8598 9 1847 9 2096 9 105 07 354 ARW 603 A0W 852 BRY 1801 ACT 8350 BRT IS99 007 IB40 COT 2097 COT 306 9 3SS 9 604 9 BS3 9 5102 9 1351 9 1600 9 1849 9 2099 9 107 OR 356 ARU 60S A02 854 BRZ B103 ACK 83S2 BRU 3601 BOX 1850 COI 2099 COI 108 9 3S7 9 606 9 BSS 9 1804 9 1333 9 1602 9 1851 9 2100 9 109 DU 3SS ARI 607 Ar"J 856 BRI 1803 ACU , 13S4 BRI 1603 BOU 1852 CDU 2101 CDU llo 9 359 9 608 9 BS7 9 1106 9 1353 9 1604 9 ISS3 9 2102 9 til OV 360 ARV 609 ADV 859 BRV 8807 ADV 8356 BRV 1605 BOV 1854 COV 2103 COV 112 9 361 9 630 9 839 9 8108 9 1357 9 1606 9 BBSS 9 2104 9 113 P 362 AP 688 AP B60 BP 1809 AP 1358 BP 1607 BP IBS6 CP 2103 CP lie 9 363 9 682 9 Bel 9 1810 9 1339 9 1609 9 1857 9 2106 9 383'PT' 364 APT 613 APW 862 BPT lill APW 1360 BPT 8609 BPW 1858 CPT 2107 CPT lie 9 363 9 614 9 863 9 til2 9 1361 9 1680 9 1859 9 2108 9 317 PT 366 APW 61S APT 964 BPT 8113 APW 1362 BPW 1611 BPT 1860 CPT 2109 CPT liG 9 367 9 616 9 863 9 til4 9 1363 9 1612 9 leal 9 Ello 9 189 PU 36G APU 617 APU 866 BPU lil5 API 1364 BPI 1613 BPI 1862 CPU 2181 CPU 320 9 369 9 618 9 867 9 tila 9 136S 9 late 9 1863 9 2112 9 128 PT 370 APW 689 APW B68 BPW 1887 APT 1366 BPT 1615 BPT 1864 CPT 2813 CPT 822 9 371 9 620 9 869 9 til9 9 1367 9 1616 9 1863 9 2154 9 123 PU 372 APU 628 API 970 BPI 1819 APU 1368 BPU 1617 BPI 1866 CPU 211$ CPU 824 9 373 9 622 9 871 9 1120 9 1369 9 tels 9 1867 9 Elle 9 325 PU 374 A8I 623 APU 872 BPI 1821 APU 1370 BPI 1619 BPU 1868 CPU 2187 CPU
$26 9 375 9 624 9 B73 9 1822 9 1371 9 1620 9 1869 9 2119 9 127 PV 376 APV 62S APV 874 BPV 8123 APV 1372 BPV B628 BPV 1970 CPV 2819 CPV 128 9 377 9 626 9 975 9 8824 9 1373 9 1622 9 1871 9 2120 9 129 3 378 At 627 AJ 876 BI il2S AJ 1374 BI 8623 BJ IB72 Cl 2121 CI 130 9 379 9 629 9 G77 9 1126 9 1373 9 1624 9 1873 9 2122 9 331 10 380 AIU 629 AJU 879 Blu 1827 AJE 3376 BII 1623 BJI 1874 CIU 2123 CIU
$32 9 381 9 630 9 879 9 1828 9 1377 9 1626 9 1973 9 2124 9 133 IU 382 AIU 631 AJI 800 BII 1829 AJU 1379 Blu 1627 BJE 1976 CIU 2125 CIU 134 9 383 9 632 9 881 9 1830 9 3379 9 3628 9 1877 9 2126 9 83S BV 384 AIV 633 AJV 882 Blv 1831 AJV 1380 BIV 1629 BJV 1978 CIV 2127 CIV 136 9 385 9 634 9 Bel 9 1832 9 1388 9 1630 9 1879 9 2128 9 337 10 386 AIO 63S AJO 894 BIO 1133 AJR 1382 BIR 8631 BJR 1800 CIO 2129 CIO 139 9 387 9 636 9 B8S 9 1134 9 1383 9 16 2 9 3888 9 2130 9 139 IOU 380 AIOU 637 AJ00 886 BIOU 113S AAI 1384 BIRI 1633 BJRI 1882 CIDU 2138 CIOU 440 9 309 9 639 9 887 9 1836 9 1385 9 1634 9 1883 9 2132 9 848 10E 390 AION 639 AJOI 888 BIDI 3937 AJRU 1386 BIRU 1635 BA2 5884 CIOI 2133 CICI 142 9 398 9 640 9 809 9 1838 9 1387 9 1636 9 1803 9 2134 9 143 10V 392 AIOV 641 AJOV 990 SIC 8 1839 AJRV 1388 BIRV 1637 BJRV 1086 C10V 2133 CIOV 144 9 393 9 642 9 898 9 8140 9 8389 9 1639 9 1087 9 2136 9 145 10 394 AIO 643 AJR 992 BIR ll48 AJO 1390 BIO 1639 BJR 8989 CIO 2137 CIO
- 846 9 395 9 644 9 893 9 1842 9 8391 9 1640 9 1889 9 2138 9 147 IDI 396 A!ON 64S AJRU 894 BIRU 1843 AJOI 1392 BIDI 1648 BJR2 1990 CION 2139 CION 848 9' 397 9 646 9 895 9 1844 9 1393 9 1642 9 1891 9 240 9 349 IOU 399 AIOU 647 AJRI 896 BIRI 1843 AJOU 1394 BIOU 1643 BAE 1992 CIOU 2148 CIOU ISO 9 399 9 649 9 897 9 1146 9 8395 9 1644 9 1893 9 2142 9 138 IOV 400 AIOV 649 AJRV 898 BL% 1847 AJOV 1396 BIOV 164S BJRV 1894 CIOV 2143 CIOV ftS2 9 401 9 650 9 899 9 1848 9 1397 9 1646 9 1895 9 2144 9 IS3 IP 402 AIP 631 AJP 900 BIP 8149 AJP 1398 BIP 8647 BJP $896 CIP 2443 CIP 154 9 403 9 652 9 901 9 1830 9 1399 9 164B 9 1897 9 2146 9 RSS IPU 404 A!PU 6S3 AJPU 902 BIPU 1831 AJPI 1400 BIPI 1649 BJPI 1998 CIPU 2147 CIPU IS6 9 40S 9 654 9 903 9 8192 9 1408 9 1650 9 1899 9 2149 9 157 IPU 406 AIPU 6SS AJPI 904 BIPI 1853 AJPU 1402 BIPU 1651 B JP N 1900 CIPU 2149 CIPU ISO 9 407 9 656 9 903 9 1854 9 8403 9 16S2 9 1901 9 2150 9 IS9 IPV 400 AIPV 6S7 AJPV 906 BIPV llSS AJPV 8404 BIPV $653 BJPV 1902 CIPV 2158 CIPV 160 9 409 9 659 9 907 9 8356 9 1403 9 3654 9 1903 9 2152 9 let I 410 AJ 659 At 909 BI 1857 AJ $406 BJ 165S BI 1904 CI 2153 Cl 162 9 481 9 660 9 909 9 1859 9 3407 9 teSe 9 190S 9 2154 9 863 IU 412 AJJ 661 AIU 980 Bly 1839 AJE 8409 BJI 1657 BII 1906 CIU 215S CIU 164 9 413 9 662 9 918 9 1860 9 1409 9 8659 9 1907 9 2156 9 865 10 484 AJI 663 A!U 912 BII Stel AJU 1410 BJI 1659 Stu 8909 CIO 2157 CIO '
l&& 9 415 9 664 9 953 9 8862 9 1488 9 1660 9 1909 9 2159 9 *
'867 IV 416 AJV 665 AIV 914 BIV 1163 AJV $412 BJV 1661 BIV 8910 Ctv 2139 Civ B68 9 487 9 664 9 919 9 1864 9 8413 9 1662 9 1988 9 2160 9 169 to 419 AJO 667 AID 936 BIO 316S AJR 1484 BJR 3663 BIR 1982 CIO 2168 C10 170 9 489 9 668 9 917 9 4166 9 4483 9 1664 9 1983 9 2167 9 c^ufioN: enEuulNAny nESutTs.
FIGURE 4-6 (Sheet 4 of 5) IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2 4-89
SEQ END SEG END SEG. END SEG. END SEG. END SEG. END SEO. ENO SEG END SEG END NO. STATE NO STATE NO. STATE NO. STATE NJ. STATE NO. STATE NO. STATE NO. STATE NO. STATE 171 ICU 420 AJOU 669 AIOU 988 BIOU 1867 AAN 3416 BAE 1665 BIRI 1914 CIOU 2163 CIOU I 172 9 I 421 9 670 9 939 9 1868 9 1417 9 1666 9 1985 9 2164 9 173 IOI 422 AJ0X 671 AIDI 920 010r !!69 AAU 1418 BA2 1667 BIRU 1986 CION 216S CION 174 9 423 9 672 9 928 9 1870 9 1489 9 1668 9 1987 9 2166 9 173 10v 424 AJOV 673 AIOV 922 BIOV !!71 AJRV 1420 BAV 1669 BIRV 1918 CIOV 2167 CIOV 176 9 423 9 674 9 923 9 1872 9 1428 9 1670 9 1919 9 2168 9
, 177 to 426 AJR 673 AIO 924 BIR 8173 AJO 1422 BJR 3671 BIO $920 CIO 2169 CIO 178 9 427 9 676 9 92S 9 8174 9 1423 9 1672 9 1921 9 2370 9 ! 179 ION 428 AJRU 677 AIOX 926 BIRU 1873 AJ0X 8424 BJRZ 1673 8801 1922 C10I 2171 CION r ISO 9 429 9 678 9 927 9 3176 9 1425 9 3674 9 1923 9 2172 9 181 IOU 430 AJRI 679 AIOU 928 BIRI 1877 AJOU 1426 4JRI 1675 BIQU 1924 C10U 2173 CIOU 182 9 431 9 680 9 929 9 1878 9 1427 9 1676 9 192S 9 2374 9 883 IOV 432 AJRV 681 AIOV 930 BIRV 1879 AJOV 1428 BJRV 1677 BIOV 1926 CIOV 2173 CIOV j 884 9 433 9 682 9 931 9 1880 9 1429 9 1678 9 1927 9 2876 9 185 IP 434 AJP 683 AIP 932 BIP 1881 AJP 8430 BJP 1679 BIP 1928 CIP 2377 CIP I 306 9 433 9 684 9 933 9 !!B2 9 1433 9 1680 9 1929 9 2178 9 187 IPU 436 AJPU 683 AIPU 934 BIPU 1883 AJPI 1432 BJPN 3688 BIPN 3930 CIPU 2179 CIPU 188 9 437 9 686 9 933 9 1884 9 8433 9 1682 9 1931 9 2180 9 189 IPU 438 AJPI 687 AIPU 936 BIPI 3183 AJPU 1434 BJPN $683 BIPU 1932 CIPU 2381 CIPU 190 9 439 9 688 9 937 9 1986 9 1433 9 1684 9 1933 9 2182 9 191 IPV 440 AJPV 689 AIPV 938 BIPV $187 AJPV 1436 BJPV 168S BIPV 1934 CIPV 2183 f!PV 192 9 448 9 690 9 939 9 1888 9 1437 9 1686 9 8935 9 2184
- 193 K 442 AK 691 AK 940 DK 1889 AL 1438 BL 1687 BL 1936 CR 2183 CK 194 9 443 9 692 9 941 9 1890 9
- 1439 9 1688 9 1937 9 2186 9 19$ KV 444 AKV 693 AKV 942 BKV 8893 ALV 1440 SLV 1689 BLV 1938 CAV 2187 CRV 196 9 445 9 694 9 943 9 1892 9 5441 9 1690 9 1939 9 2888 9 197 kP 446 #KP 695 AKP 944 BKP 8893 ALP 1442 BLP 1698 BLP 1940 CKP 2189 CKP 198 9 447 9 696 9 945 9 1894 9 1443 9 1692 9 1948 9 2190 9 199 KPV 448 AKPV 697 AKPV 946 BKPV 1893 ALPV 1444 BLPV 1693 BLPV 1942 CKPV 2191 CKPV
, 200 9 449 9 698 9 947 9 1896 9 144S 9 1694 9 1943 9 2192 9 208 3 450 AJ 699 AJ 948 BJ 1897 At 1446 Bt 1695 BI 1944 CI 2193 CI
' 202 9 458 9 700 9 949 9 1898 9 1847 9 1696 9 8945 9 2194 9
.203 IU 4S2 AJU 708 AJE 950 BJX 1199 A!U 1448 BIU 1697 B!I 1946 CIU 2193 CIU 1204 9 433 9 702 9 951 9 1200 9 1449 9 3698 9 1947 9 2196 9 20S IU 454 AJI 703 AJU 932 BJE 3201 AIU 1450 BII 1699 Blu 1948 CIU 2197 CIU 206 9 4SS 9 704 9 953 9 8202 9 1438 9 1700 9 1949 9 2198 9 207 IV 456 AJV 70S AJV 954 DJV 8203 AIV 1452 Blv 1701 BIV 1950 CIV 2199 CIV 208 9 457 9 706 9 95S 9 1204 9 1433 9 1702 9 1938 9 2200 9 209 10 458 AJO 707 AJR 956 DA 120S AIO 1454 BIO 1703 BIR 1952 C10 2201 CIO 280 9 459 9 708 9 957 9 1206 9 1433 9 8704 9 8953 9 2202 9 288 10U 460 AJOU 709 AJRR 958 BJRI 1207 AIOU 14Se BIOU 170S BIRI 1954 C100 2203 CIOU 282 9 461 9 780 9 959 9 1208 9 1457 9 1706 9 1955 9 2204 9 283 IDI 462 AJOI 711 AJRU 960 BJRZ 1209 AIDI 1438 BIDI 1707 BIRU 1956 CICI 2205 CICI 214 9 463 9 782 9 961 9 1210 9 1439 9 1708 9 1957 9 2206 9 21S 10V 464 AJOV 713 AJRV 962 BJRV 1258 AIOV 1460 BIOV 1709 BIRV 1958 CIOV 2207 CIOV 216 9 46S 9 714 9 963 9 1212 9 1461 9 1780 9 1959 9 2208 9 237 IO 466 AA 78S AJO 964 BJR $283 A!O 1462 BIR $711 BIO 1960 CIO 2209 CIO 2ie 9 467 9 786 9 96S 9 1214 9 1463 9 1712 9 1961 9 2250 9 289 10E 468 AJRU 787 AJOI 966 BA! 1213 At01 1464 BIRU 1713 BIDI 1962 CICI 2281 CION 220 9 469 9 788 9 967 9 1216 9 8465 9 1714 9 1963 9 2282 9 221 100 470 AJRX 719 AJOU 968 BJRX $287 AIOU 1466 BIRI 1715 BIOU 8964 CIOU 2283 CIOU 222 9 471 9 720 9 969 9 1288 9 1467 9 1786 9 1965 9 2254 9 223 IOV 472 AJRV 728 AJOV 970 BJRV 1219 A100 $468 BIRV 3717 BIOV 1966 CIOV 2213 CIOV 224 9 473 9 722 9 971 9 1220 9 1469 9 87 9 1967 9 2216 9 22S IP 474 AJP 723 AJP 972 BJP B221 AIP 1470 BIP 17. t BIP 1968 CIP 2217 CIP 226 9 47S 9 724 9 973 9 1222 9 1478 9 8720 9 1969 9 2288 9 227 IPU 476 AJPU 725 AJPX 974 BJPI $223 A!PU 1472 BtPU 1721 BIPI 1970 CIPU 2289 CIPU 228 9 477 9 726 9 97S 9 1224 9 $473 9 1722 9 1978 9 2220 9 229 IPU 478 AJPN 727 AJPU 976 BJPI 122S AIPO 1474 BlPI $723 BIPU 1972 CIPU 2221 CIPU 230 9 479 9 728 9 977 9 1226 9 1473 9 3724 9 1973 9 2222 9 238 IPV 480 AJPV 729 AJPV 978 BJPV 1227 AIPV 1476 BIPV 8725 BIPV 1974 CIPV 2223 CIPV 232 9 481 9 730 9 979 9 1228 9 1477 9 1726 9 197S 9 2224 9 233 K* 482 AK 731 AL 980 BL 1229 AK 1478 BK 1727 BL 1976 CK 2223 CK 234 9 483 9 732 9 988 9 8230 9 3479 9 1728 9 1977 9 2226 9 235 KV 484 AKV 733 ALV 982 BLV 1231 Akv 1480 BKV $729 BLV 1978 CRV 2227 CKV 236 9 485 9 734 9 903 9 1232 9 8488 9 1730 9 1979 9 2228 9 237 KP 486 AKP 73S ALP 984 BLP 8233 AKP 1482 BeP 1731 BLP 1980 CKP 2229 CkP 238 9 487 9 736 9 985 9 1234 9 1483 9 1732 9 1981 9 2230 9 239 KPV 488 AKPV 737 ALPV 986 BLPV $235 AKPV 1484 BKPV 1733 BLPV 1982 CMPV 2238 CKPV 240 9 489 9 738 9 987 9 1236 9 1485 9 1734 9 1983 9 2232 9 248 K 490 AL 739 AK 988 BL 1237 AK 1486 BL 1735 BK 1984 CR 2233 CK 242 9 491 9 740 9 989 9 1238 9 1487 9 1736 9 198S 9 2234 9 243 KV 492 ALV 748 AKV 990 BLV 1239 ARV 8488 BLV 1737 BKV 1986 CKV 2233 CKY 244 9 493 9 742 9 998 9 1240 9 1489 9 1738 9 1987 9 2236 9 245 KP 494 ALP 743 AKP 992 BLP 1248 AKP 8490 BLP 8739 BKP 1989 CKP 2237 CKP 246 9 495 9 744 9 993 9 1242 9 8498 9 1740 9 1989 9 2238 9 247 kPV 496 ALPV 745 AKPV 994 BLPV 1243 AKPV 1492 BLPV 1748 BKPV 1990 CKPV 2239 CKPV 248 9 497 9 746 9 995 9 1244 9 1493 9 8742 9 8998 9 2240 9 249 M 498 AM 747 AM 996 BM 1245 AM 8494 BM 1743 SM 1992 CM 2241 CM CAUTION: PRELIMINARY MESULTS- ,
IMPORTANT UNCERTAINTIES FIGURE 4-6 Sheet 5 of 5) DESCRISED IN SECTION 2 1 2 4-90 1
l l INITIATING EVENT BLOCK PHENOMENA BLOCK SYSTEM FUNCTION OR EVENT BLOCK A = AUTOMATIC INITIATION M = MANUAL INITIATION
--> N = EVENT NUMBER SUCCESS P = PHENOMENA OR FAILURE OF A NORMALLY OPERATING
/ ', A 4 EOUiPMENT N # \ - SHOWN IF TRANSFERS TO
- FAILURE /= J THIS BL OCK OCCUR V
I SUCCESS OR STABLE STATE BLOCK TRANSFER TO ANOTHER ESD N TRANSFER TO EVENT BLOCK N PLANT DAMAGE END STATE FIGURE 4-7. EVENT SEQUENCE DIAGRAM SYMBOLOGY 4-91 f
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GENERAL TRANSIENT Y
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S1 TURBINE TRIP ~) STE AM RELIEF RCS OPERATOR PORV HOTSTAND9Y R MSIV
+ > 2 AFWPs >
INVENTORY CONTROL
-> LONG TERM ON STE AM R COOLING GENElsATORS 1 4 7 8 NO OPE RATORS AVAILABLE. ONLY AUTOMATIC ACTIONS V U 'I I ' l SMALL LOCA NTROLS > *IAF OP N - SEALS AND/ R ECCS c Q CONT. -'CWET, gpg CLOSES LETDOWN LINE + COMMON g -
SPRAY B / RE LIEF VALVE A - A - 2 5 6 10- 12a 17 19a DRY l SMALL LOCA THIS IS NOT NE'CESSARILY DRY, HH1 CLOSED ONTO START AT VMT U V U A OPE R ATOR I I OPE R ATOR b U1 SURVI ES PTS - CO ECCS C @ HIGH HEAD C @ ESTAB OSED COLD ST ANDBY CHALLENGE PORVs OPEN > COMMON R INJECTION B RHR 3
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. DENOTES SEQUENCES LEADING TO CORE MELT DRY RAY DRY B
; INDICATES TH AT DETAILED MODEL MUST A -
' KE EP TRACK OF TR AINWISE ECCS/SPR AY INFORMATION (IS MAINF!OW INPORTANTM 17-19b l CAUTION: PRELIMINARY REsULTS-IMPORTANT UNCE RTAINTIES FIGURE 4-9. SIMPLIFIED GENERAL TRANSIENT ESD USING ONLY DEsCRisto iN sECTION :
THE EVENT TREE TOP EVENTS m_
LEGEND: GF GUARANTEED FAILURE E GS GUARANTEED SUCCESS h" I l < a o < a u NN NOT NECESSARY C ! D o < m o E I n l3 e$! !" i d E5 ! ! ! : i g g g B zu ! g '$
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= < :it < < 2 m z g" $' $' <
s E i G 8tlst ii H ii igs ! ! ! ! ! ! e 8 8 i IE TT CH vt AF f A PR 01 ON 08 EA EB EC HA HB HC OR SA $8 SC r 1 S 1 [ I
= :- 2 LT2A 2
=[ 3 LT28 3 4 LT2a 4 5 LT23 5 6 LT2S 6 7 LT2D 7 8 LT2D 8 9 LT2G 9
-=-m-w = w- 10 LT2C 10 w- 11 LT2E 11 w- 12 LT2E 12 w- 13 LT2H 13
= - r- : r 14 LT2C 14 15 LT2E 15
-w 16 LT2E 16 17 LT2H 17
= 7-w = w-w- 18 LT2F 18
, w-r- 19 LT2l 19 i e-a-=-m-w 20 LT2C 20 21 LT2E 21 22 LT2E 22 23 LT2H 23
-F-m-w-=-w w- 24 LT2F 24 w- 25 LT21 25 w-w = : w-w 26 LT21 26 27 LT21 27
-w-w-w-w-r-w-w- 28 LT2J 28
- - :- :- 29 S 29
=:[- I
= = =
===
30 LT1C 31 S 30 g 31
=== 32 LT1H 32 r = = 33 5 33 r = ^- 34 LT1H 34
- = = 35 5 35 g
= = = 36 LT1J 36
= ; = 37 5 37
=== 38 LT1H 38 39 5 39 b == == =:- 40 L11J 40
= : ; 41 S 41
- : : 42 LTIJ 42
-w 43 LT2K 43 44 LT2L 44 45 LT2L 45 48 LT2N 46 47 LT2L 47 48 LT2N 48 49 LT2N 49 50 LT20 50 1
FIGURE 4-10. GENERAL TRANSIENT EVENT TREE C^uTioN: entumiNany nasvus. l (Sheet 1 of 4) '"""*C'"*"'8 l DESCal8ED IN $ECTION 2 4-96
LEEND: GF GUARANTEED FAILURE GS GUARANTEED SUCCESS E h f 4 = u < m u NN NOT NECESSARY
$p E C %d 5
b
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;. sg sg
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g = w Q ifs i ! ais3 Ei E ii i0 ! ! ! ! ! ! ! 5 B $ it l 77 CH vi ArlF1 PR Of CN 09 EA ES EC HA HB HC OR SA SS SC w c' w- 51 S 51 m-***-w- 52 LT11 52
-w := w- 53 S 53 I
- = w- 54 LT1K 54 w = := w- 55 S 55 m-m--w- 56 LT1K 56
-w-w w- 57 LT2M 57 w- 58 LT20 58 w- 59 LT20 59 w- 60 LT2R 60
= w-a* 61 S 61 m-w-au- 62 LT11 62 m-w-am- 63 S 63 N*-w-e 64 LT1K 64
--w m-w -am- 65 S 65 m-w-m-- 66 LT1K 66
-w- W 67 LT2M 67 68 LT20 68
-w 69 LT20 69 70 LT2R 70
- -w m-w-a- 71 S 71 m*-w-w- 72 LT1L 72
--w-w-w -w- 73 LT2P 73 w-w- 74 LT2S 74 w w -- -
75 S 75 w-m-n*- 76 LTII 76 w - 77 5 77 Lw - 78 LT1K 78 w -- 79 S 79 w-e-a- 80 LT1K 80
-w-w 81 LT2M 81 82 LT20 82 83 LT20 53 84 LT2R 84
-v w w- m-w- 85 S 85 Lw =-w- 86 LT1L 86
-w-wqw w-w-
87 LT2P 88 LT25 87 88 w-w w-w-e- 89 S 89 w-w-*+- 90 LTIL 90 w-w 91 LT2P 91 92 LT2S 92
-w-w-w-w-w-w-r- 93 LT2T 93 CAUTION: PRELIMINARY RESULTS.
IMPORTANT ONCERTAINTIES DESCRIBED IN SECTION 2 FIGURE 4-10 (Sheet 2 of 4) 4-97
A GENERAL TRANSrENT EVENT TUR3:NE TR'P OR MSiv TR'P
@ OPERATOR CONTROLS HHf 5 VESSEL INTEGR'TY
~
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+
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C 5335* $7 o D (M T P O I 1 1 1 t 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 66666655555555554444 444444333333333322 yog8 NGGt . E S O N 54321 0987654321 0987654321 0987654321 098 NSFE G C R : E R TA P NGG N I R B E N E XXXxXLLLLSLLLSLLLLLSLSLSLLL5LLLLLSLSLS F F F FFTTTT TTT TTTTT T T TTT TTTTT IT T 3o v % i OUU:3 TAA D TU LI RRRRR2221 221 22Z2T T I 221 22221 1 RR 1 331 4 4 r FF 1 F HEE 0 E E C 1 FF HEEC E E C I N NC IM N N AA E NN S E E A CTT C R R EEE T T Y 1 2 d' Ed SEE I A 06332222222222222222222222222222222222 SJ3 O I R 3841 5554 4 444 44444333333333322222222221 yoyd A N NT E S 074721 0987654321 0987654321 0987654321 09 RSF 2 I E U YUA S LT C L' CU S ER S S E
LEGEND: GF GUARANTED FAILURE GS GUARANTED SUCCESS E g < a u < a o NN NOT NECESSARY g 15 C m 4 @g
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^:- ^:-
2 LT1B 3 LTIS 2 3
- 4 LDD 4 5 LTIS 5 y -
6 LUD 6 7 LT1D 7 8 LT2A 8 [ I 9 10 LT2a LT28 9 10 11 LT2D 11 12 LT2B 12 13 LT2D 13 14 LT2D 14 I 15 LT2G 15 7 :' w- 16 LTIC 16 w w- 17 LTIE 17 g W - w- 18 LT1E 18 w- 19 LT2C 19 w[ I F-w- 20 LT2E 21 LT2E 20 21 w- 22 LT2H 22 g 23 LTIC 23 w== 24 LTIE 24 w ; MS- w - N*-- 25 LT1E 25 w 26 LT2C 26 [ w I 27 LT2E 23 LT2E 27 28 29 LT2H 22 g w-t-N*--w- r- 30 LT1F 30 w-r- 31 LT2F 31 w w-w[w w-r- 32 LT21 33 LTIC 32 33 g w - - - 34 LTIE 34 w - 35 LT1E 35 36 LT2C 36 w[ I 37 LT2E 38 LT2E 37 38 39 LT2H 30 w E r : M LDF M g w-- 41 LT2F 41 w-w
-w-w[ w :- -
42 LT2! 43 LTIF 42 43 44 LT2F 44 w-r-e-e-e-r-e-w[ 45 LT2! 46 LT2J 45 46
---------------------- 47 XFR1 47 40 XFR2 93
[ I m*--==--w--------
**--w - N: -------
49 XFR3 50 XFR4 1 01 105
=*-w--w-------- 51 XFR5 100 I
w - = -- n -- w- - - - - . 52 XFR6 111 w -w-- w- y- - - - - . 53 XFR7 115 w-w-aww-r- - - 54 XFR8 117 w-w-w-w-r-r- SS LT2J 119 _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . es xm9 120 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . 57 XFR9 239 w-w----------------------- 58 XFR10 35a FIGURE 4-11. SMALL LOCA EVENT TREE cAurioN: rae uMiNAny assutts. IMPOR f ANT UNCE RTAINTIEs "C"'"'"**'"* 4-100
LTl
+
Su n.C ENT 5' Ix'5LERS ";cg";^T E5'M$'^'*" + "H5' PUMP 3: / yd,y,",g,; F AN COOLE RS T A.? 2 A.w 3 P 1P 1P CONTAI MENT EO a - M PUMP - PUMP DESIGN NO LE AK 4 A 6 8 l l 3, SCRUB LIAK MELT - MELT CONTAINME NT CONTAINMENT NO SCRUS SPRAY PUM* ?S^tATION COOL M LI^K m 7 NO ^ * ^ SCRUS
^ ?
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EIs"T TU HMS4TO O SUCCEED
>-* qp 3 P LATON IF -p ' T "M O BE LHS3 LHSI PUMP -) AT DECAY EAT PUMP DE SIGN EXCHANGER REMOVAL V EC RC D p.iR MX 2 A/w 2 A/M 4 y A 8 y A NO 1F NO COOL SCRUS LEAK R O AL M H P ST MELT N CTS FTE R + CONTAINMENT CONTAINMENT MELT "li'MOv"I^' ggLMELT E% ANGER lE E "I^' ,
5" ^" UMP SC3_Ug iSOtAToN ofAx 4 ^ 8 ^ ^
- ME LT h h NO COOL I
l ! CONTAINME NT NO SCRUS ISOLATION LEAK NO COO 8 ( ^ CAUTION: PRELIANNARY RESULTS-IMPORTANT UNCERTAINTIES FIGURE 4-12. LT1 LONG-TERM RESPONSE ESD oESCRieEo iN PferioN r I l l l
LEGEND: GF GUARANTEED FAILURE g GS GUARANTEED SUCCESS a NN NOT NECESSARY 8 *- g < a u , g $ w E 5 5 5 g c c E g v 4 e u 3 g3 l a o o u o ,, 2 l
=
UEg {=a n 5 5 5 a a m
$ ! W W W 2 5 5 5 5 6 5 8_
IE FC RA RB RC St CL LA LB LC RX CS CP Cl g 1 S 1 r r :- :- ax = 2 S 2 g
= = = m: = 3 S 3
= = = 4 S 4 5 2A 5 w[ 6 7
2G 2E 6 7 g 8 28 8 9 2H 9 10 2F 10 w-w-w-w- - - - - - - - - 11 XFR1 11 12 S 17 I
- 7 = = a: = 13 S 18 7 = N: = 14 S 19 g
w- r- - - - - - - - - 15 XFR1 20 w-w-w-w- - - - - - - - . 16 XFR1 26 g 17 S 32 w = - = == = 18 S 33 w g= = m--a: = 19 S 34 w--------- 20 XFR1 35 w-w-w- r- - - - - - - - - 21 XFR1 41 g
= w-w-= = - = 22 S 47 g
w--w = = :: = 23 S 48 w-a- w- - - - - - - - - 24 XFR1 49 w- w-w- r- - - - - - - - - 25 XFR1 55 m: = = = = == = 26 S 61 g wq= = =
=
=
=
- =
=
27 S 28 S 62 63 g w - -- ---. 29 XFR1 64 w_w-w-w- _ - _ _ _ _ . 30 XFR1 70 w --- = w = = :: . w--. 31 S 76 w w-= = me-*- 32 5 77 I y- w_ _ _ . _ - . 33 XFR1 78 y-w-y-y- _ .- . 34 XFR1 84 m - = :- = ax = 35 S 90 w-w - 36 S 91 w--------. 37 XFR1 92 w- w-e- w- - - - - - - - - 38 XFR1 98 w-e-w-w-w-= 39 28 104 40 2H 105 41 2F 106
= an 42 S 107
[ [ [ 43 2C 44 21 108 109 45 2E 110 46 2D 111 47 2J 112 48 2F 113 49 XFR2 114 50 XFR2 121 I - - _- - - _ - - _- Si xFR2 iza 52 XFR2 135 I _ _ _ - _ . 53 XFR2 142 _ _ _ _ _ _ _ _ _. 54 xFR2 149 I w______._. 55 xrR3 156 FIGURE 4-13. FRONTLINE EVENT TREE LT1 c,urios; eneuwmany nesutts. (Sheet 1 of 2) w eonT48t uscenraiNries orscnieso is section :
LEGEND: GF GUARANTEED FAILURE 5 GS GUARANTEED SUCCESS c5 NN NOT NECESSARY 4 e U 8 > w " 3 3 5 Y d 8 0 l r b "2 r CAUTION: PRELIMINARY RESULTS-0 0
, ~; $ $ IMPORTANT UNCERTAINTIES
- Z 2
9 3 8 "8 W
"8 m a a m a v E" =2 DESCRISED IN SECTION 2 W W i E 5 5 5 & 9 5_ 80 e re R4 ms Re s ca. LA ts Le Rx cs cp a 1 - - - - - - - -
56 XFR1 162 [ _ 57 XFR3 58 XFR4 168 174 I 59 XFR4 186
---------- 60 XFR4 198 I
- - - - - - - - - -- 61 XFR4 210 62 XFR4 222 I
63 XFR4 234 w - -- -. 64 XFR3 246 65 XFR8 252 r - - - - - - - - - - -- 66 XFR2 397 [ I y - _ _ _. _ _ _ w------------ 67 XFR2 68 XFR2 404 411 I w-w- - - - - - - - . 69 XFR3 418
- w- ----~~- 70 XFR1 424
_ _ . 71 xyg3 430
----w------------ 72 XFR5 436 w----------- 73 XFR5 448 I w-y - .
74 XFR3 460 75 XFR9 466 76 XFR2 535 [ w I 77 XFR2 78 XFR2 542 549 I w--------. 79 XFR3 556 w ---------- 80 XFR5 562 l ----------- 81 XFR5 574 w - - - - - - - - - -- 82 XFR5 586 I w- ------- 83 XFR3 598 84 XFR10 604 w-w------------ 85 XFR2 673 [ I w-w- r- - - - - - - - . w-w------------ 86 XFR3 87 XFR5 680 686 I w-a- w- - - - - - - - . 88 XFR3 698 y 89 XFR11 90 XFR2 704 735 [ I 91 XFR2 92 XFR2 742 749 I w- - - - . 93 XFR3 756 w ---------- 94 XFR5 762 I .. _ - - - - - - - -. 95 XFR5 774 i 96 Xm5 7M w--------. 97 XFR3 798
--------------------- 98 XFR12 804 y w- -- - .
99 XFR2 873 [ w w- r- - - - - - - - . 100 XFR3 W----------- 101 XFR5 880 886 I e- a- - - - - - - - . 102 XFR3 898
--------------------- 103 XFR13 904 w-w ---------- 104 XFR2 935
[ w-w I w--------. 105 XFR3 106 XFR5 942 948 w------- -. 107 XFR3 960
. 108 XFR14 966 w-w-w-w-r 109 20 997 I
110 2J 998 111 2F 999 FIGURE 4-13 (Sheet 2 Of 2) 4-103
LEGEND: GF GUARANTEED FAILURE a CS GUARANTEED SUCCESS 8 , , , NN NOT NECESSARY l I I I
' l lc ! c 5
5 l l l l$l 3 W l = M ilg5 il 35 E ! ! w h 8 $ $ $ I sh 8h IE FC RA R9 RC CS LA LB LC RX CP Q N r - 1 2A 1 [ I
=
2 2G 3 2E 2 3
=
=[ INN ---
4 28 5 2H 6 2F 4 5 6
== ------
7 XFR2 7
= ------
8 XFR2 10 g
=[ I
=
9 28 10 2H 11 2F 13 14 15 m r w = ------ 12 XFR1 16 g 2 w = ------ 13 XFR2 19 w : ------ 14 XFR2 22 g w r ------ 15 XFR3 25 y nn---w ------ 16 XFR1 28 V = ------ 17 XFR2 31 g w ,
= ------
18 XFR2 34 w ------ 19 XFR3 37 g N: w-w = ------ 20 XFR1 40 g w-w = ------ 21 XFR2 43 w-w w ------ 22 XFR3 46 w-ha ------ 23 XFR1 49 g W N: = ------ 24 XFR2 52 g
= ------
25 XFR2 55 g w ------ 26 XFR3 58 w- ==---w - ------ 27 XFR1 61 w w -
------ 28 XFR2 64 g
w w ------ 29 Xm3 67 w-r-a ------ 30 XFR1 70 w-w = ------ 31 XFR2 73 I w ------ 32 Xm3 76
= w-r-w w ------
33 XFR3 79
------ 34 XFR1 82
[I 35 2C 85 [ g 36 21 37 2E 86 87 38 XFR4 88 g
--------- 39 Xm4 94 40 XFR4 100 g
41 XFR4 106 42 XFR4 112 g 43 XFR4 118 w ------ 44 XFR5 124 g 45 XFR2 127 46 20 130 [ I 47 2J 48 2F 131 132 49 XFR6 133
= ------
50 XFR6 136 w+ ------ 51 XFR6 139 I 52 XFR6 142 53 XFR6 145 I 54 XFR6 148 55 2D w[ 151 I 56 2J 152 57 2F 153 FIGURE 4-15. FRONTLINE EVENT TREE LT2 CAurioN: ensu w NAny n Sutts. (Sheet 1 of 2) (WORTANT UNCERTAINTitS Of SCRittD IN SECTION 2 4-105
LEGEND: GF GUARANTEED FAILURE
< m u GS GUARANTEED SUCCESS 8 8 8 , , , a NN NO7 NECESSARY
- g -
W % % % g 5
$ 2 2 2 s = ! ! ! g a = = ! IE -E i i i ig i g M ! I i
3 1 a 8! 5 l!5 9 ! w w 9 5 s s 5 5 sa 53 IE FC RA R9 RC CS LA LS LC Rt CP O I w --------- 58 XFR4 154 59 XFR4 160 g 7 --------- 60 XFR4 166 w w ------ 61 XFR5 172 m ------ 62 XFR6 175 w h ------ 63 XFR6 178 w - ------ 64 XFR6 181 g w w ------ 65 XFR7 184 w --------- 66 XFR4 187
--------- 67 XFR4 193 w g 68 XFR4 199 w ------
69 XFR5 205 j w 7 ------ 70 XFR6 208
= ------
71 XFR6 211 w ------ 72 XFR6 214 l w ------ 73 XFR7 217 w-w --------- 74 XFR4 220 v w w --~~~~ 75 Xm5 D6 w w ------ 76 XFR6 229 l w w w ------ 77 XFR7 232 w --------- 78 XFR4 235 l --------- 79 XFR4 2 41 g
--------- 80 XFR4 247 w ------
81 Xm5 2M w ------ 82 Xm6 2M g v: ------ 83 XFR6 259 g 84 XFR6 262 85 XFR7 265 w g w --------- 86 XFR4 268 w w ------ 87 XFRS 274 w w - ~ ------ 88 Xm6 277 l w w ------ 89 XFR7 280 w-w g 90 XFR4 283 w ------ 91 XFRS 289 w-w g 92 XFR6 292 w ------ 93 XFR7 295 W w-w w w ------ 94 Xm7 298 CAUflON: PRELIMINARY RESULTS. IMPORfANT UNCERfAINilES DESCRieED IN SECTION 2 FIGURE 4-15(Sheet 2of2) 4-106
- 5. SYSTEMS ANALYSIS 5.1 SYSTEMS ANALYSIS APPROACH The purposes of systems analyses are to provide the qualitative knowledge about the plant systems needed to construct a plant risk model and to provide the input quantification of the event sequence model (i.e., to quantify top event split fractions in the plant event trees).
The systems analysis was conducted in two parts. The first part was an initial screening that encompassed all STP systems. The screening determined each system's normal and transient alleviating actions. These actions characterize the response of the plant to initiating events. Success requirements were established for all alleviating actions as discussed in Section 4.4. The second part of the systems analysis was a detailed analysis of those systems that passed the initial screening. One result of the initial screening was a classification of each system
- as either a "frontline" system or a " support" system. Each system was also classified as to whether further analysis was required. Those requiring no further analysis were so noted and their system summaries were filed.
The second portion of the baseline systems analysis determined the intersystem dependencies for the support system model and recorded the details required to properly model the system. Items covered were as follows: e Support Systems Needed e Systems Supported e Equipment Shared with Other Systems e Normal Automatic Actions e Normal Manual Actions e Operator Emergency / Recovery Actions e Controlling Station Locations, Indications, and Alarms e Testing and Maintenance Requirements e Technical Specifications, Limiting conditions for Operation, and Surveillance Also noted in a general comment section were any other system characteristics important to the system model. Once the systems were studied and documented and the event sequence diagrams developed, the systems were assigned to event tree top events and actions in a way that preserved dependencies between tops. Initiating events that would disable each system were also identified. 5-1 0104H052085
System logic models were considered next. When necessary, system boundaries were defined and block diagrams were drawn. These block diagrams were modified for each success criterion and for each support system state for a given top event. The final step in systems analysis was the estimation of conditional split fractions. This estimation was performed in one of two ways, depending on the systems involved:
- 1. For those systems that are very similar to their Midland or Seabrook counterparts and whose models were not significantly different (i.e.,
similar logic, success criteria, and support system states), the conditional split fractions from the Midland or Seabrook PRA (References 5-1 and 5-2) were used. / 2. Where significant differences in system logic were noted, block diagram logic models were used that included independent hardware failure, maintenance, testing, human error, and common cause failure terms. The unavailability expressions were then quantified using data from the Seabrook PRA data bases. 5.2
SUMMARY
OF SYSTEMS ANALYZED Tables 5-1 and 5-2 and Figure 5-1 document all plant systems considered in the Scoping Study and the disposition of the initial screening process described above. Those systems classified as important to risk were studied in detail and modeled to the level required for the Scoping S tudy. Table 5-1 lists these systems. In some cases, only portions of the system were incluced in the risk model. Table 5-2 lists those systems that were classified as not needing further analysis in the Scoping Study but whose further analysis might prove useful during risk management efforts. Systems that are not required for safety and offer little or no potential for affecting accident scenarios appear in Figure 5-1. . 5.3 EXAMPLE SYSTEMS
SUMMARY
- ESSENTIAL COOLING WATER SYSTEM A brief summary of the ECW's systems analysis is presented in this section of the report to familiarize the reader with the system analysis methodology used in the Scoping Study. Although the ECW's analysis provides a relatively simple example of the evaluation and quantification process, the same general methods were applied for each of the systems modeled in the event trees. To obtain a complete picture of the moceling of systems in the Scoping Study, it is helpful to also review the event sequence model in Section 4. The event sequence model represents an important interface between the risk assessment and the modeling of systems.
Information about each plant system was collected and documented in a system analysis summary. An example summary outline is shown in Figure 5-2. Important information about the ECW system is summarized in Tables 5-3, 5-4, and 5-5. 5-2 0104H052085
Figure 5-3 is the ECW's piping and instrumentation diagram. Figure 5-4 is a block program model of the ECWs used for system unavailability quantification. Table 5-6 lists the components included in each model block. The block diagram model forms the basis for the Boolean logic expressions used to quantify system unavailability. The block diagram portrays the
" success paths" of the system. These paths are combinations of component success states that enable successful functioning of the system. The success paths, which have the same logical information contained in a listing of the minimal cutsets, provide the basis for calculating system unavailability. Table 5-7 presents these expressions for the ECW system under each of the eight general boundary conditions of support system availability derived from the auxiliary systems' dependency information.
Quantification of ECW's unavailability is performed in two steps. The first step develops unavailability information for each system model block by evaluating the effects on each component in the block from hardware failures, testing, maintenance, human errors, environmental effects, and other causes. Table 5-8 illustrates an example calculation of the hardware failure contribution to the unavailability of one ECW system model block for a specific system operating condition. The second step of the quantification process develops system level unavailability results by combining the model blocks through the Boolean logic expression for each support system boundary condition. The system-level calculations account for all the causes for each block's unavailability, and they include combinations of causes affecting components in redundent system blocks. For example, in the ECW system model from Figure 5-4, the joint unavailability of model blocks PA and PB (pump trains 1A and 18) has contributions from each of the following causes: o Coincident hardware failures of pumps 1A and IB. e Maintenance of pump 1B and hardware failures of pump 1A. e Dependent failures of pumps 1A and 18 during operation. There are no contributions from maintenance of pump 1A, because it is the normally running pump and cannot be out of service for maintenance when a plant initiating event occurs. Table 5-9 illustrates an example ECW's
- unavailability calculation for a specific system operating condition i under one support system boundary condition.
l l The top events of the event trees were quantified using generic component l failure rate data, maintenance frequency and duration data, human error rates, and common cause failure rate data from References 5-1, 5-2, and other sources. Where specific component dependent failure rate data were not available from these studies, point estimate values of b = 0.05 and y = 0.50 were used. p is the fraction of component failures in which cne or more similar components are failed due to a shared, common cause. y is the fraction of the common cause failures that are shared by three or more similar components. These parameters are estimated from component failure event data at U.S. nuclear plants. This summary level quantification is consistent with a preliminary, scoping level analysis. Additional documentation would be needed for a comprehensive, stand-alone analysis. 5-3 0104H052085
5.4 REFERENCES
5-1. Pickard, Lowe and Garrick Inc., " Midland Probabilistic Risk Assessment," prepared for the Consumers Power Company, May 1984. 5-2. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0300, December 1983. i l i 0104H052085
iwcerar u= cram =rits
**'# TABLE 5-1. STPEGS SYSTEMS PARTIALLY OR FULLY ANALYZED IN THE PRELIMINARY SCOPING STUDY Sheet 1 Ff 2 Frontline (F) Included in System the Freliminary E "E or
- Disposition Support (S) * **
Scoping Study Reactor Contafrnent F Yes FC EP,ESFAS CCW Provides heat removal and limits Fan Coolers pressure of containment atmosphere after a large energy release inside containment. Full analysis required. Contain=ent Isolation F Yes CI.CP EP,ESFAS Isolates lines penetrating containment to prevent radioactive releases. Full analysis required. Emergency Core Cooling F Yes EP ESFAS,ECH,CCW The primary method of keeping the core e Reactor Water Storage Tank EA,EB EC covered in many scenarios. Provides Suction Lines alternate core cooling method e Contairment Sump Suction RA,RS,RC in bleed and feed. Provides long-Lines term core cooling in recirculation e High Head Safety Injection HA,HB,HC,SI scenarios. e Low Head Safety Injection LA.LB,LC,RX Auxiliary Feedwater F Yes AF,F1 EP,ESFAS.MS Backup to main feedwater for reactor m coolant system heat removal. Full m analysis required. Contairment Spray F Yes SA,SB,5C,CS EP,ESFAS ECH Removes fission products and limits pressure of containment atmosphere after a large energy release inside containment. Full analysis required. Component Cooling Water S Yes CA,CB,CC EP,ESFAS,ECW Provides cooling to vital components during both normal and accident conditions. Full analysis required. Essential Cooling Water S Yes WA.WB,WC EP,ESFAS Removes heat loads from safety related equipment to the ultimate heat sink. Full analysis required. Electric Pcwer S Yes This system supports all vital systems, o Offsite Grid OG The analysis includes all Class IE power e Generator Breaker, Main, UA supplies, associated switchgear and and Unit Auxiliary support systems, and DC Power. Transformer e Standby Transformer 51,52 e 13.8 kV Buses IF,1G,lH,ll e Vital 4.16 kV Buses EA,EB.EC EV e Vital DC Bases DA,DB,0C EV e Standby Diesel Generators GA,GB.CC ECW 0105H052085
CAUTION: PREUMikARY REsuLTs-twPORTANT UNC13AINTits
****""'"**C""*
TABLE 5-1 (continued) Sheet 2 of 2 Frontline (F) Included in EE System or the Preliminary Disposition Support (S) Scoping Study
'" * #8
- Reactor Trip System S Yes SS - Anticipated transients without scram (solid state protection not quantified in the Preliminary systea) Scoping Study. Full analysis required in a complete Level 1 PRA.
Engineered Safety 5 Yes EA,EB.EC EP Safety systems receive actuation signals Features Actuation from ESFAS. Full analysis required. Steam Duro Control F Yes AF,F1 IA,SDC.ESFAS, Provides reactor coolant system cooling (steam relief) EP.IA via secondary side. Steam relief includes steam generator PORVs steam generator safeties, and condenser vacuum. Ultimate Heat Sink - - - - Included in essential cooling water analysis. Primary Relief F Yes PR,08 EP Provides overpressure protection and is ut used in feed and bleed. Analysis & includes the pressurizer PORVs, block valves, and safeties. A complete Level 1 PRA may include the vessel head vent system. Essential Chilled Water S Yes SA,SB,5C EP.ESFAS, Removes heat from essential equipment ECW area HVAC unit coolers and rejects heat to the ECWS. Full analysis required. Electrical Auxiliary S Yes EV EP,ECH Supplies cooling to switchgear for Building HVAC safety related equipment. Room heatup calculations and additional equipment qualification analyse.; should be performed. Chemical and Volume F Yes OI.CI EP,ESFAS,CCW Provides reactor coolant system inventory Control control. The Preliminary Scoping Study includes charging pumps, normal charging line, letdown line to containment isolation valves, RCP seal injection and return lines, and RCP cooling. Main Steam F Yes TT ESFAS,EP The main steam isolation valves are analyzed in the Preliminary Scoping Study. Electrohydraulic Control F Yes TT EP This system is analyzed for turbine trip logic and turbine stop and governor valvre
"5'"C'*5'*" "2 FOR PHASE II ANALYSIS Frontline (F) Included in "EE System or the Preliminary Disposition
'" * #5
- Support (5) Scoping Study Main Feedwater F No None EP.MS.ESFAS This system is not analyzed in the Preliminary Scoping Study because it is automatically isolated after a reactor trip with low reactor coolant system T,,g.
Reacter Coolant F No None - This system is not explicitly analyzed. Loss of coolant accident analyses will account for the loss of this system. Residual Heat Removal F No OR EP,CCWS Not evaluated in the Preliminary Scoping Study due to lack of equipment qualification analyses supporting pump operation in high temperature and high humidity environments. Further analyses may be required in a complete Level 1 PRA to Wtify residual heat removal operation in uegraded conditions. Fuel Handling Building HVAC S No - - ECCS pumps cubicle coolers are included in ECCS model. m Centrol Room Envelope HVAC S No - EP ESFAS,ECH Effects of failure are bounded by EAB HVAC a nalysi s. Auxiliary Cooling Water S No - - Not analyzed for the Preliminary Scoping Study. System does not serve any equipment analyzed in the Preliminary Scoping Study. Instrtenent Air S No - EP,ACS Not analyzed for the Preliminary Scoping Study. Instrument air failure does not I affect any systems modeled in the Preliminary Scoping Study. Mechanical Auxiliary 5 Yes - - HVAC explicitly modeled with individual Building HVAC systems as appropriate. Diesel Generator S Yes - - Included in diesel generator analysis. Building HVAC Diesel Generator Fuel Oil 5 Yes - - Included in diesel generator analysis. Storage and Transfer Diesel Generator Closed S Yes - - Included in diesel generator analysis. Cooling Water Diesel Generator Starting - - - - Included in diesel generator analysis.
- Included in diesel generator analysis. l Diesel Generator Lubrication - - -
0105H0520S5
r o TABLE 5-3. ESSENTIAL COOLING WATER SYSTEM C011PONENT DEPENDENCIES AC DC ESFAS ESFAS ! Component Power Power Train Relay Pump 1A Bus E1A Bus E1A11 A Sequencer M0V0121 (discharge MCC E1A3 valve) Pump 1A Strainer MCC E1A3 FV6935 (blowdown Bus EIA11 A K817 valve) Traveling Screen 1A MCC E1A3 A K817 Screen Wash Booster MCC EIA3 A K817 Pump 1A FV6914 (screen wash Bus EIA11 valve) Vent Fan FN001 MCC E1A3 Vent Fan Fl1002 MCC EIA3 Pump 1B Bus ElB Bus E1811 B Sequencer MOV0137 (discharge MCC E183 valve) Pump 18 Strainer MCC E183 FV6936 (blowdown Bus ElB11 B K817 valve) Traveling Screen IB MCC E183 B K817 Screen Wash Booster MCC E183 8 K817 Pump 18 FV6924 (screen wash Bus ElB11 valve) Vent Fan FN003 MCC E183 Vent Fan FN004 MCC E183 Pump IC Bus E1C Bus E1C11 C Sequencer MOV0151 (discharge MCC E1C3 valve) Pump 1C Strainer MCC E1C3 FV6937 (blowdown Bus E1C11 C K817 ! valve) l Traveling Screen 1C MCC E1C3 C K817 Screen Wash Booster MCC E1C3 C K817 Pump 1C i FV6934 (screen wash Bus E1C11 valve) Vent Fan FN005 MCC E1C3 j Vent Fan FN006 MCC E1C3 l l CAufl0N: PRGLIMINARY Rt4ULTS-IMPORf ANT UNCERfA6NfitS DESCRie80 IN 88Cfl0N 2 i i 5-8 002811071684
TABLE 5-4. ESSENTIAL COOLING WATER SYSTEM LOADS ECWS Train Components Supplied Cooling Water 1A Standby Diesel Generator IA Standby Diesel Generator 1A Auxiliaries Skid Component Cooling Water Heat Exchanger IA Component Cooling Water Pump 1A Supplementary Cooler Train 1A Essential Chiller CH001* Train 1A Essential Chiller CH004* IB Standby Diesel Generator 1B Standby Diesel Generator 18 Auxiliaries Skid Component Cooling Water Heat Exchanger 18 Component Cooling Water Pump 18 Supplementary Cooler Train IB Essential Chiller CH002* Train IB Essential Chiller CH005* 1C Standby Diesel Generator 1C Standby Diesel Generator 1C Auxiliaries Skid Component Cooling Water Heat Exchanger 1C Component Cooling Water Pump 1C Supplementary Cooler Train IC Essential Chiller CH003* Train 1C Essential Chiller CH006*
- Essential chillers can be supplied from any ECWS train through crossties.
CAUTION: PRELIMINARY RESULTS-IhWORTANf uhCERTAINfitS Of SCRISED IN 88CTION 2 0084H122684 5-9
m TABLE 5-5. NORMAL ALIGNMENT FOR ESSENTIAL COOLING WATER, COMP 0NENT COOLING WATER, AND ESSENTIAL CHILLED WATER USED FOR THE PRELIMINARY SCOPING STUDY System Running Standby Off Train Train Train Essential Cooling Water A C B Component Cooling Water A C B Essential Chilled Water A+B C NOTE: Crosstie valves EWO265 and EWO274 are open to supply the chillers for essential chilled water trains A and B from ECWS train A. All other ECWS'crosstic valves are closed. CAufl0N: PRE Linh80AR Y RESULTS-14APORfANT UNCERf A4Nfitt DESCRestD IN SECit04 2 5-10 0028H071684
TABLE S-6. C0llPONENTS INCLUDED IN ESSENTIAL COOLING WATER SYSTEM MODEL BLOCKS llodel Block Components PA Pump 1A Check Valve EW0006 Discharge Valve M0V0121 Strainer 1A Traveling Screen 1A Yent Fan FN001 Vent Fan FN002 PB Pump 1B Check Valve EWOO42 Discharge Valve H0V0137 Strainer 1B Traveling Screen IB Vent Fan FN003 Vent Fan FN004 PC Pump 1C Check Valve EWOO79 Discharge Valve MOV0151 Strainer 1C Traveling Screen 1C Vent Fan FN005 Vent Fan FN006 CP Essential Cooling Pond (system common) CAufl0N: PRtLIMINARY RESULTS. 141POATANT UNCERfAINfitS 068CMINO IN StCTION 2 5-11 0028H071684
l I j TABLE 5-7. ESSENTIAL COOLING WATER SYSTEM UNAVAILABILITY EXPRESSIONS i ab e Unavailability Expression Sy[P[fgfva n None Qecw = 1.0 A Qecw = PA + CP B Qecw = PB + CP C Qecw = PC + CP l A+B Qecw = (PA)(PB) + CP A+C Qecw = (PA)(PC) + CP B+C Qecw = (PB)(PC) + CP A+B+C Qecw = (PA)(PB)(PC) + CP i F CAUTION: PRELleuleNARY RESULTS. l IMPORfANf UNCERfAINfItS DESCRief 0 lN SECTION 2 i i $ l l l 4 ! 5-12 0028H071684
1 i TABLE 5-8. EXAMPLE CALCULATION OF HARDWARE UNAVAILABILITY FOR ESSENTIAL COOLING WATER SYSTEM MODEL BLOCK PA (System Operating Condition: Pump 1A Restart Required) 4 Component Failure Mode Variable Failure Mission Pump 1A Fails to Restart PS 2.4-3/d Pump 1A Fails During Operation PR 3.4-5/h 24h Check Valve EW0006 Fails to Reopen CVO 2.7-4/d Check Valve EW0006 Transfers Closed CVC 1.0-8/h 24h MOV0121 Transfers Closed MVC 9.3-8/h 24h Strainer 1A Plugs SP 6.2-6/h 24h
. Traveling Screen 1A Plugs TP 6.2-6/h 24h Fan FN001 Fails to Restart FS 4.8-4/d Fan FN001 Fails During Operation FR 7.9-6/h 24h ^l Fan FN002 Fails to Restart FS 4.8-4/d Fan FN002 Fails During Operation FR 7.9-6/h 24h
- Dependent Fan Failure to Restart SFS .05 Dependent Fan Failure During Operation SFR .05 NOTES:
- 1. Exponential notation is i i.e., 2.4-3 = 2.4 x 10-3,ndicated in abbreviated form;
- 2. d = demand; h = hour.
QPA = PS+PR+CV0+CVC+MVC+SP+TP+(FS + FR)2,gp (p3)*0FR(FR) l l
= (2.4-3) + (3.4-5)(24) + (2.7-4) + (1.0-8)(24) + (9.3-8)(24) 2
+ (6.2-6)(24) + (6.2-6)(24) + {[(4.8-4) + (7.9-6)(24)3
+ (3.8-7)) + (.05)(4.8-4) + ( 05)(7.9-6)(24)
QPA = 3.8-3 CAufl0N: PRELIWNARY RESULTS. mWORTANT ONCERTAINfitt DESCR60tD IN SECflON J 5-13 0084H122884
? TABLE 5-9. EXAMPLE CALCULATION OF ESSENTIAL COOLING WATER SYSTEM UNAVAILABILITY [ Support System Boundary Condition: Support Available for ECWS Trains A and B (e.g., Power Available at Buses EIA and ElB)] (System Operating Condition: Pump 1A Restart Required) Model Hardware Block Unavailability PA 3.8-3 PB 8.1-3 CP 2.7-6 Component Variable U a$NS$kty Punp 1B PM 2.6-3
$$$ue Variable Failure Mission Rate Time Mode Pump Fails to Start PS 2.4-3/d Pump Fails During Operation PR 3.5-5/H 24h Check Valve Fails to Open CVO 2.7-4/d Fan Faf1s to Start- FS 4.8-4/d Fan Fails During Operation FR 7.9-6/h 24h NOTES:
- 1. Exponential notation is indicated in abbreviated form; i.e., 2.4-3 = 2.4 x 10-3,
- 2. d = demand, h = hour.
CAUflON; PRELIMINARY RESULTS-IMPORTANT UNCERf AINiltS DESCRISED IN SECTION 2 5-14 002811071684
1 TABLE 5-9 (continued) Dependent Failure Fraction Variable Point Estimate Pump Fails to Start BPS 1.1-1 Pump Fails During Operation BPR 7.6-2 Check Valve Fails to Open SCVO .05 Fan Fails to Start (two fans) 8FS .05 Fan Fails During Operation (two fans) 8FR .05 Intertrain Dependencies ST .05 Unavailability Expression: QECW = (PA)(PB) + CP QECW = (PA)(PB + PM) + CP + Sp3(PS) + BPR(PR) + BCV0(CV0)
+ BOT FS(FS) + 8 83 pg(FR)
= { (3.8-3)[(8.1-3) + (2.6-3)] + (9.3-6) + (1.2-6) + (5.7-8)
+ (3.5-14) + (4.8-12) + (1.4-8) + (1.8-11) + 5.6-15))
+ (2.7-6) + (1.1-1)(2.4-3) + (7.6-2)(3.4-5)(24) + ( .05)(2.7-4)
+ (.05)(.05)(4.8-4) + (.05)(.05)(7.9-6)(24)
QECW = 4.0-4 NOTES:
- 1. Exponential notation is indicated in abbreviated form; i.e., 1.1-1 = 1.1 x 10-1
- 2. Additional terms in braces { } account for the effects of correlated uncertainties between the failure rates for similar components in blocks PA and PB.
CAufl0N: Pattl4804ARY ASSULTS-148P047 ANT UNCERfAINfit8 000CR6000 IN 90CflON 2 5-15 0084H122884
Control Room Habitability Condensate Polishing Digital Rod Position Indication Demineralizers Reactor Coolant Leak Detection Potable and Sanitary Water Combustible Gas Control Station Sewage Treatment Containment Air Purification Sodium Hypochlorite Feed , and Cleanup Reactor Control Reactor Makeup Water Pressurizer Pressure Control Pressurizer Level Control Fuel Handling Spent Fuel Cooling and Cleanup Steam Generator Level Control Process Sampling Equipment and Floor Drain Service Air In-Core Instrumentation Plant Lighting Rod Control Turbine-Generator Reactor Makeup Water Main Condensate Communications Condenser Evacuation Fire Protection Turbine Gland Sealing Reactor Containment Building Turbine Bypass HVAC Turbine-Generator Building Steam Generator Blowdown HVAC Liquid Waste Processing Gaseous Waste Processing Boron Concentration Measurement Solid Waste Processing Boron Thermal Regeneration Radiation Monitoring P2500 Plant Computer Circulating Water Main Reservoir Auxiliary Steam Extraction Steam Secondary System Chemical Lube Oil Purification and Addition Transfer Makeup Demineralized Water Generator Stator and Cooling Generator Stator Water Cooling Vibration and Loose Isolated Bus Duct Cooling Generator Hydrogen Gas Vibration Monitoring Parts Monitoring i FIGURE 5-1. STPEGS SYSTEMS REQUIRING LITTLE OR N0 FURTHER ANALYSIS l
"AuflON: PRELIMINARY RESULTS-14MORTANf uMCERTAINTIES OtsCRISED IN $tCTION 2 i
0084H122684 5-16
._ __ __ _ _ _ _ _ - . _ _ . _ _ ~ - _ . . _ _ _ . _ _ _ _ _ _ . _ . . _ _ . - _ _ - . _ _ - . . .___
A. Evaluation for System Screening
- 1. References
- 2. System Function During All Normal Modes of Operation
- 3. Impact on Plant if Normal Functions Are Lost
- 4. Safety Functions
- 5. Events That This System Could Initiate
- 6. FSAR Success Requirements for Normal and Safety Functions (number of pumps or trains and timing) and for What Conditions They Are Needed
- 7. System Classification
- 8. Disposition B. Development of Intersystem Dependencies Matrices
- 9. Support Systems Needed
- 10. Systems Supported
- 11. Shared Equipment with Other Systems C. Detailed Description for Systems Analysis
- 12. Normal Automatic Actions (initiation logic, setpoints)
- 13. Normal Plant Manual Actions
- 14. Operator Emergency / Recovery Actions
- 15. Controlling Station Locations, Indications, and Alarms
- 16. Testing and Maintenance Requirements
- 17. Technical Specifications /LCOs/ Surveillance D. Interface with Event Sequence Task
- 18. Initiating Events Disabling This System
- 19. Event Tree Top Events Involving This System
- 20. Support System Boundary Condition States to be Modeled E. Comments / Questions and Answers F. Logic Models
- 21. Define Boundaries of Logic Model l 22. PRA Study Failure Conditions for Each Function Analyzed l 23. System Block Diagram
- 24. System Block Descriptions G. _ System Quantification
! 25. System Equations
- 26. Conditional Split Fraction FIGURE S-2. SOUTH TEXAS PROJECT SYSTEMS ANALYSIS SUMi%RY OUTLINE S-17 0028H071684
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- 2. INSTRUMENT TAKECFF PIPE AND Fill!NCS TO AND INCLUDING 0135. 03 36. 013$. SH!
THE ROOF VALVE SHALL COhf0RN 10 fHE MAIN LINE SPEC. 0140. 0141. 0142. 0145. Osts. 0149. Di
- 3. LOCATE FLANCE OutSIDE INTAKE STRUCTURE. 0150. 0353. 0154 0155. 015s. 0161
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DESIGNATOR En. 0196. 0131. 0207. 020s. 020s. 02:0. 0236 0237. 0238. Ficure 9.2.1-S. ALL INSTRUMENT $ ON THIS Ps!D ARE SEFARATION CROUP N UN E5s OfMERalSE DES!CNAtEn. 0233. 0012. 0028. 0032 003s. 004a. G. t M ESS OTHERdlSE NOTED. UNIT I IS SHOMN ON THl$ PAIO. 00ss. 00ss. 0073. FOR APPL ICA110N OF THIS Pal 010 IJN!! 2. USE UNIT 2 0065. 0102. 0 06. IDENTIFIC ATION NUMBERS FOR P!PINC. veLVES. 0110. 0015. 0031. INSTRUMENTS. ECUlPMENT. E TC. REFER TO ORAMlNCS 0035. 0039. 0058. SA310F00008 AND 9A310F00002 FOR IOE N T IF ICAi1.pt 0008. 0072. 0U7s. NUM8ER OETAILS. 0088. DIOS. CICS. 0113. 0211. 0212. 0213. 0214
- 7. hMEN APPtVING THl$ P41010 UNII 2. CHANGE ALL UNII 0215. 02ts. 0217. 0?!8 0219. NOTE:
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- 6. SURVEY OF EXTERNAL EVENTS AND INTERNAL PLANT HAZARDS
- 6.1 OVERVIEW
, In performing a full-scope plant risk model for a Level 1, 2, or 3 PRA, the inclusion of external events and internal plant hazards is logically phased after much of the work to develop the basic plant event sequence and systems models has been completed. The objectives of this task in the Preliminary Scoping Study are the following: e Screen a large number of potential external events and internal plant hazards using bounding risk estimates and knowledge of site-specific and plant-specific characteristics. e Provide a basis for quantifying the approximate range of possible accident frequencies that could result from these events (performed in Section 2). With respect to the above objectives, the results have been organized into three sections. In Section 6.2, seismic events are discussed. Although this discussion is purely qualitative, it provided a basis for quantifying uncertainties in the Scoping Study results in Section 2. A key finding is that the scope of the seismic analysis to fully document seismic risk could probably be reduced below the normal level in view of the low seismicity of the STPEGS and a relatively high safe shutdown earthquake level in relation to the seismicity. In Section 6.3, internal plant hazards are discussed. Again, this section is purely qualitative but provides part of the basis for the probability distributions in Section 2 that characterize the state of knowledge at the end of the Scoping Study about core melt and release frequencies. The most critical plant locations, especially for the i analysis of internal fires and floods were identified in this task. The remaining external events are discussed in Section 6.4. In this i section, bounding estimates are made for external events discussed in the
- FSAR. Based on these bounding estimates, it is unlikely that any J additional work would be needed to document that these events are l insignificant contributors to risk for STPEGS.
( 6.2 SEISMIC EVENTS A preliminary review of the FSAR and discussions with the systems analysis team were accomplished for the purpose of scoping the tasks to complete a fully documented PRA seismic analysis. Adequate site specific !- geologic and seismologic information exists in the final safety analysis t report to be used in the seismicity analysis. The FSAR indicates that l there are no tectonic faults in the vicinity of the site, and that there is an absence of historic seismic activity there. Postulated earthquake magnitudes are low and are not expected to present any significant frequuicy of acceleratieq 1evels that could damage plant components. t However, in order for PRA documentation to be complete, it would be 6-1 0106H052185 t
i necessary to develop probability of frequency exceedance curves for peak i ground acceleration at the STP. l The FSAR was also reviewed for general design information relative to fragilityofstructuresandequipment. Nothing was found that suggests i STP s facilities would have any particular weaknesses from predicted seismic events, but it must be remembered that physical inspection was not possible because of the early stage of plant construction. Although some summary design stress and capacity information is contained in the FSAR, additional information would be required in order to perform a detailed fragility analysis. This information--strong evidence of low seismicity and no indication of weak structures or equipment--was combined with our experience in seismic risk assessment and our new knowledge of the STP plant to estimate the possible range of frequency of seismically induced core damage events. The most useful experience was in PRAs of PWRs with similar capabilities and higher seismicity. 6.3 IN-PLANT HAZARDS AND SPATIAL INTERACTIONS f i This section describes a qualitative evaluation of the vulnerability of the STPEGS to internal plant environmental hazards, such as fires, l internal floods, smoke, steam environments, etc., that are triggered from within the plant. To provide a basis for this investigation, the STPEGS i plant model at the HL&P offices on Westheimer Avenue, Houston, Texas, was inspected and the following documents were reviewed: plant layout i drawings, an STPEGS fire hazards analysis report (Reference 6-1), and STPEGS FSAR (Reference 6-2). Because of the nature of this category of external events, the investigation in the Scoping Study was carried out on an
, location-by-location basis as opposed to an event-by-event basis.
! Several areas corresponding to fire zones within buildings that could be important to plant risk are identified below. These are areas where a l physical interaction associated with a particular hazard could disable i several vital pieces of equipment. Areas of particular interest are ! those capable of involving scenarios impacting equipment in multiple trains of a support or frontline system as well as those scenarios involving equipment in more than one support or frontline system. The Scoping Study identifies only the obvious potential paths of hazard i progagation among the locations. 1 Reference 6-1 describes the separation among the three safety trains for safe shutdown equipment. This description includes all the equipmekt ! items of interest to this study except the HHSI pumps, containment spray i pumps and their associated valves, piping, and electrical circuits. i- Also, the exact routing of cables is not yet determined for STPEGS; however, the fire hazard analysis report clearly indicates the location of cable trays and conduits by their train designations. 4 F I 6-2 0106H052185
Using the above referenced documents, a preliminary qualitative ! evaluation of the potential for risk significant spatial interactions , within the STPEGS buildings has been performed based on knowledge gained on the STPEGS plant during the Scoping Study. Each building is discussed separately below. 6.3.1 ELECTRICAL A,UXILIARY BUILDING The electrical auxiliary building houses most of the electrical circuits for plant control, safety-related equipment, and accident mitigation. The control room, the relay room, the safety-related switchgear rooms, the cable penetrations into the containment, the auxiliary shutdown panel, and other related areas such as cable spreading rooms and cable vaults are located in this building. Separation in this building is achieved primarily by dedicating each compartment to only one train. In a few compartments, some cable trays or conduits from a redundant train are present. In these cases, trays and conduits will be wrapped in 1-hour fire rated insulation and the area will be provided with an automatic fire suppression system. The exceptions to this rule are the control room, the relay room, the auxiliary shutdown area (designated as FA017.01), and cable spreading room B. The auxiliary shutdown panel is a standby system and is electrically isolated from the main electrical circuits. Therefore, all environmental hazards affecting this panel have only a small likelihood of adversely affecting safe plant shutdown. The control room and the relay room are not separated, and hazardous environments can severely impact the plant. The operators can use the auxiliary shutdown panel as a redundant point for plant control. Scenarios for the relay room and control room hazardous environments would be analyzed in detail to evaluate operator response (possible errors or significant delay in taking action) in a fully documented PRA. Experience in completed PRAs combined with knowledge of the STP design permits us to estimate the possible impacts of such scenarios. An example of the scenarios considered is the case of loss of all component cooling water pumps and operator delay (because of conflicting information on the control board) in tripping the reactor coolant pumps before seal damage occurs. Cable spreading room B contains a few control and instrumentation cable trays from train C. It is not clear which circuits are in these trays. Thus, there may be potential for disabling both trains B and C of some equipment. The redundant trains of electrical circuits in the EAB are generally well separated in terms of fire impact with the use of barriers. By contrast, high energy missiles and flooding hazards can penetrate these barriers and affect the redundant trains. Such hazards were not found in this building in this preliminary analysis. Thus, the spatial interaction analysis should primarily concentrate on the occurrence of fires and 7 floods in this building. In this regard, spatial interactions caused by 6-3 0106H052085
l 1 failures or unavailability of component cooling, HVAC, and chilled water systems have been explicitly integrated into the plant and systems analysis models. 6.3.2 MECHANICAL AUXILIARY BUILDING The mechanical auxiliary building houses the CCW system, the chemical and volume control system, and the waste processing systems as well as other mechanical systems for plant operation, monitoring, and shutdown. This building is similar to the EAB in the sense that in all fire areas and fire zones where redundant sets of cable trays appear, one set of trays will be wrapped in 1-hour rated fire barriers and the area will be provided with an automatic fire suppression system. An exception to this rule is fire area FA102.07, which contains all three CCW heat exchangers. For the MAB, only high energy hazards (such as water jet and pipe whip) and internal floods are deemed important. A rigorous investigation could be performed for the ways that the high energy hazard sources present in the MAB may disable redundant cable trays and conduits. The high energy hazards are deemed to have significant likelihood for penetrating the 1-hour rated fire barriers, ground cable trays, and conduits. An important source of flooding in the MAB is the essential water cooling system that provides cooling to the component cooling water systems heat exchangers. A break in the essential water cooling system can spill a very large volume of water into the MAB. The water can accumulate at Elevation 10'-0" of the MAB and jeopardize safety class components; e.g., CCW and charging pumps, etc. Such large breaks are unlikely to occur. For the MAB, the spatial interaction analysis should concentrate on scenarios involving high energy hazards, water sprays, and rising pools of water. 6.3.3 FUEL HANDLING BUILDIING The fuel handling building contains the high head safety injection system, the low head safety injection system, the containment spray system, and spent and new fuel handling equipment. The three redundant trains of the above-mentioned safety-related systems are separated by 3-hour rated fire barriers. The only exception is fire area FA303 at Elevation 4'-0", which contains all three trains of FHB exhaust and booster fans. The fire hazard analysis report does not indicate the location of cable trays and conduits for this building. Therefore, the routing locations of certain cables inside the FHB could have some impact on risk. If the redundant cables are also well separated, the spatial interaction analysis should concentrate on environmental hazard scenarios originating at the higher elevations of the building and impacting the three safety-related systems at Elevation 29'-0". 6-4 0085H022185
6.3.4 REACTOR CONTAINMENT BUILDING The reactor containment building houses the nuclear steam supply system, the residual heat removal pumps and heat exchangers, and che:nical volume l and control system related equipment. The RCB does not have any completely enclosed compartments. There are communication paths among all fire areas within this building. Separation among the redundant j -trains is achieved by distance (either vertical or horizontal). The building contents are well designed for most of the potential i environmental hazards, such as high energy line break, steam, and hydrogen combustion. Only very large fires can impact redundant cable
, trays and conduits.
6.3.5 DIESEL GENERATOR BUILDING The diesel generator building houses all three emergency diesel generators. The building has three separate sections. Each section houses one diesel generator and its support equipment. All significant environmental hazards generated from within the building are deemed to be contained within the section of origin and to impact only one diesel
- generator. Only extremely high energy sources, such as fire in a fuel storage tank, could conceivably impact two or more locations.
! 6.3.6 ISOLATION VALVE CUBICLE The isolation valve cubicle houses the main steam line isolation valves, the atmospheric steam relief valves, main feedwater piping, and the auxiliary feedwater system. Four auxiliary feedwater pumps are located
- at Elevation 10'-0" of the building. Except for a few cables, these pumps and their associated cabling are well separated from one anccher.
Fires affecting redundant cables in some of the fire areas within the building could affect risk results. Thre is some potential for steam line and feedwater breaks with propagation of steam and water to the separated sections of the building.
- 6.3.7 ESSENTIAL C0OLING WATER INTAKE STRUCTURE The essential cooling water pumps are located in the essential cooling
! water intake structure. Each pump train is in a completely separate section of the intake structure. There is no interaction among these sections. The spatial interaction analysis should include scenarios involving multiple events; e.g., separation is breached via a door left open and an environmental hazard occurs independently. 6.3.8 OTHER BUILDINGS AND THE YARD All other buildings of the plant not mentioned explicitly here do not contain safety-related equipment and do not pose any hazards to the buildings mentioned above. Therefore, they are given lower priority in the spatial interaction analysis. t 6-5 0085H022085
The yard area contains, among many nonsafety items, the auxiliary feedwater storage tank and the auxiliary transformers. No internally generated sources of hazard that could disable the auxiliary feedwater pump were identified. The auxiliary ESF transformer will be analyzed in detail in the spatial interaction analysis. The offsite power and diesel generator power leading to the engineered safety feature switchgear in the EAB are routed via an underground separated bus duct system. Therefore, no internally generated environmental hazards can disable them. 6.4 OTHER EXTERNAL EVENTS Several other external events are analyzed in this section. They consist of events that do not usually contribute significantly to the core melt frequency and often can be dismissed in a screening analysis. The following sections present the results of such screening analyses of risk associated with aircraft crash, turbine missile, tornado wind and missile, hazardous chemicals, and external flooding. 6.4.1 AIRCRAFT HAZARD ANALYSIS There are two airports within 10 miles of STPEGS. These are C-Level Farm, 9.5 miles west-northwest of the site, and Collegeport Airfield, located approximately 8.5 miles southwest. The latter is no longer in active use. However, an airfield has been constructed about 1/4 mile east of the old runway and will be used primarily for agricultural aviation. It is estimated that during the peak growing season, there will be approximately 100 takeoffs and landings per day. The C-Level Farm facility is also used for crop-dusting operations. During working seasons, there are 25 to 35 landings or takeoffs per day. There are also several small grass strips within 10 miles of the site used during aerial crop-dusting operations. Since the aircraft used for agricultural purposes are typically very light and the annual number of flights is not large, the risk associated with such aircraft activities is not considered significant. There are two low level federal airways within 10 miles of the site. These are airways V70 and V20 with centerlines 5 and 9 miles from the site, respectively. Of these two airways, only V70 has a significant contribution to the risk of aircraft because of its relative proximity to the site. 6.4.1.1 Analytical Model The frequency of aircraft crashes into different structures of the plant is estimated using the following model (Reference 6-3) L f = M [N Ad 4 I6*1) k i=1 j=1 93jjA pj 6-6 0085H022085
where fk = annual frequency of impact on the kth structure (events per year). M = number of different flight paths that take aircraft past the site. L = number of different type of aircraft that pass the site. Njj = annual number of operations of aircraft of type j to or from airport i or along airway 1. Aj = crash rate of aircraft of type j (accident / mile flown). dj = distance traveled by the aircraft while the plant site is within its potential impact area (miles). Ak j = effective impact area of the kth structure of the plant for aircraft of type j (square miles). Ap j = potential impact area for aircraft type j (square miles). The product Njjljdj is the number of aircraft a:cidents of type j within the defined distance segment dj per year that could potentially affect the plant from the ith airway or airport. The ratio Ak / oj is the probability of hitting a particular structure given tha he aircraft accident is in the vicinity of the site. The quantities d and An (the aircraft type index j is dropped for convenience) can be calculated by assuming that any given time a crash initiating malfunction occurs, there is an equal probability of crash termination anywhere in a sector of radial length, gh, and angular width, $, located directly in front of the aircraf t, where g is the 91ide distance per unit of altitude 1ost and h is the altitude. The situation is shown in Figure 6-2, where b is defined as the distance of closest lateral approach between the normal flight path of the aircraft and the site. Figure 6-2 shows that the distance d is given by 1 d= (gh)2 -b + b/ tan S (6.2) This quantity can be averaged over all allowable values of b. The result is d=fgb(f)/ Sin (f) (6.3) Ap is the area of the sector defined by the angle $ and radius gh A =(gh)2(f) (6.4) p 6-7 008511022085
1 1 6.4.1.2 Crash Rates i Crash rate statistics are provided either in the form of the number of crashes in the total number of miles or the number of hours flown by a particular type of aircraft. The latter can be converted to the former by multiplying the number of hours by an average speed for the type of aircraft under consideration. Table 6-1 shows 10 years' statistics for fatal accidents involving air carriers (Reference 6-4). The mean and the variance of the annual inflight crash rates are 1.51 x 10-9 and 4.39 x 10-19, respectively. These values are used to fit a lognormal distribution by matching moments method. Other characteristics of the distribution are as follows: 5th Percentile: 6.95 x 10-10 50th Percentile: 1.39 x 10-9 95th Percentile: 2.76 x 10-9 By using the data from 1970 to 1979 to construct our state of knowledge distribution for the crash ratio during the period of operation of STP, we are reflecting the possiblity of changes in the crash rates that might occur due to introduction of new technology. The accident rates for general aviation aircraft are given in Table 6-2 (Reference 6-5). The classification into single and multiple engine aircraft is due to the difference in their impact effects on the plant structures. The following values characterize the lognormal distributions chosen to represent our uncertainty concerning these rates: CRASHES PER MILE FLOWN Cha e 8 9 r u Single Engine Multiple Engine 5th Percentile 1.91 x 10-7 5.54 x 10-8 50th Percentile 2.27 x 10-7 7.14 x 10-8 95th Percentile 2.70 x 10-7 9.20 x 10-8 Mean 2.28 x 10-7 7.23 x 10-8 The distribution in each case is derived by using the mean and variance of the crash rates based on Table 6-2 as the mean and variance of a lognormal distribution. l 6-8 0085H022085
6.4.1.3 Number of Operations (N) The majority of aircraft activities near the site take place along the low level federal airway V70. The centerline of V70 has a closest approach of approximately 5 miles. According to Reference 6-2, a 1983 survey of flights in the vicinity of the site indicates that there are about 25 flights per day. Approximately half of these flights have altitudes less than 17,000 feet and are categorized as general aviation aircraft. The other half, with an altitude greater than 17,000 feet, are mainly air carriers. Therefore, the following values are estimated for the number of flights for these two categories of aircraft: General Aviation: N1 = 4,563 Air Carriers: N2 = 4,563 6.4.1.4 Exposure Parameters d and Ap To obtain the exposure parameters d and Ap as defined by Equations (6.3) and (6.4), several other parameters (namely, g, the glide rat.io; h, the altitude; and 4, the exposure angle) are needed for each category of aircraft. In-this analysis, a glide ratio of 17 was used for both categories of aircraft. It was also assumed that 4/2 > 90* for the two categories. The altitude of general aviation flights was taken to be about half of the 17,000 feet altitude, which was used to distinguish between two categories of aircraft. For the large aircraft flying above 17,000 feet, an average altitude of 23,500 feet was used. 6.4.1.5 Impact Area and Fragility of Different Structures The total exposed area of the plant is about 0.034 square miles, which consists of the plant area (approximately 0.0095 square miles), shadow area (about 0.013 square miles), and slide area (approximatelv 0.011). It is assumed that concrete structures can withstand impact of general aviation aircraft but would collapse upon impact by air carriers. 6.4.1.6 Impact Frequency The annual frequency of aircraft crashes into any structure at the plant was calculated in Equation (6.1) using values of tge various parameters obtained above. The total frequency is 6.95 x 10- per yegr, which is the sum of the general aviation contribution of 6.94 x 10-' and the air carrier contribution of 1.66 x 10-9 per year. Given that the impact area of the general aviation type of aircraft is relatively small, and that most critical structures of the plant can withstand the impact load of this type of aircraft that usually weigh less than 12,500 lbs., the frequency of substantial damage to any given safety-related structure of the plant and the impact of such damage woul magnitude smaller than approximately 7 x 10With. p besuch at least a low an order of initiating event frequency, aircraft crash-initiated scenarios are not anticipated to rank high among the contributors to core melt frequency. 6-9 0085H022005
_ . . ._ _ _ __ __ _.. . _ _ m _ __ _ _ _ _ _ _ . _ _ l 6.4.2 TURBINE MISSILE RISK l This section presents a screening analysis of the risk to the STP plant from missiles that can potentially be generated in the event of a steam turbine failure. 6.4.2.1 Turbine Missile Impact and Damage Frequency The frequency, f, _of serious damage to a specific system due to a- turbine missile is calculated from f=f1*f2*f3 (6.5) where fl = annual frequency of missile generation. i f2 = conditional probability'of a missile striking an essential i system given that a turbine missile has been generated. 1 f3 = conditional probability of unacceptable damage to tha system given that a missile strikes the system.
- 6.4.2.2 Frequency of Turbine Missile Generation, f1 The STP plant utilizes Westinghouse turbine generators. Each turbine is a four-casing, tandem compound six-flow reheat,1,800 rpm unit with 40-inch last stage blades (Reference 6-2). Failures of turbine generator rotating elements are generally categorized into (1) failure at or near operating speed, and (2) overspeed failure that results from failure of
. the steam admission control components. i An analysis of the likelihood of each of the above failure modes based on Westinghouse experience (Reference 6-6) provides the following estimates for missile generation frequency: Operating Speed: fl = 1.6 x 10-6 per unit / year i Overspeed: f{=1.7x10-10 per unit / year l More recent analyses based on generic data indicate that missiles can be generated at much higher frequencies. In this analysis, we use the generic estimates obtained by Bush and Heasier (Reference 6-7) fl = 1.1 x 10-4 per unit / year i f{=4.3x10-5 per unit / year It is noted thatused the values assumedanalysis above forinf1 the are greater than and fFSAR. (ReF(erence 6 the values in a similar The ! authors had already reviewed the estimates of Reference 6-8 as part of ' the SSPSA. Rather than use additional resources to review the FSAR
- estimates, the estimatcs of Reference 6-8 were used for this bounding analysis.
, 6-10 { 0085H022085 i l
6.4.2.3 Conditional Probability of Missile Impact, f2
-To obtain f2 , the conditional probability of a missile striking an essential system given turbine failure, the behavior of potential missiles ejected from the turbine must be analyzed, taking into account the kinetic energy and possible trajectories of the missiles as well as the location of potential targets. Such an analysis has been performed for STP and the results are reported in Reference 6-2 and shown in Tables 6-3 and 6-4 for two failure mechanisms. These tables also provide the estimated total frequencies of damage to each safety-related structure, which are obtained by using the appropriate values of fi ,
f 2, and f3 in Equation (6.5). Tables 6-1 and 6-2 show that the annual frequency of damage to any given target is not greater than 5.65 x 10-8, which is the estimated damage frequency for the Unit 2 isolation valve cubicle. However, damage to the structure does not necessarily mean damage to the equipment inside. Furthermore, even complete destruction of a single system is not likely to lead to core melt directly. Based on these considerations, we conclude that turbine missile initiated scenarios at STPEGS are insignificant contributors to the total core melt frequency. 6.4.3 TORNADO WIND AND MISSILE RISK This section presents the results of analysis of the risk associated with tornado wind and missiles. In general, winds can affect critical structures of the plant in two ways: o If wind forces exceed the load capacity of a building or other external structures, the incident walls or framing might collapse or the structure overturn from excessive loading. o In case of strong winds, such as in tornadoes, objects might be lifted and thrust as missiles against a critical facility that, if not designed to resist missile penetration, might be damaged and lose its function. The following section discusses the risk due to tornado wind load, followed by a tornado missile risk analysis. 6.4.3.1 Tornado Wind Risk The design basis tornado windspeed for critical structures of STPEGS is 360 mph. This windspeed is composed of a translational component of 70 mph and a rotational component of 290 mph (Reference 6-2). The annual i frequency, &, of excessive tornado wind load on structures can be found i using j & " $t * &vlt (6.6) ! where 4t is the annual frequency of a tornado striking the plant and i t is the fraction of tornadoes with peak windspeed greater than 1 6-11 0085H022085
The algorithm used to estimate 4t is (Reference 6-8)
&t=n* (6.7) where W = mean destructive path area of a tornado in square miles.
A = area of interest within which it is assumed the tornado could strike the site. ; n = mean number of tornado occurrences per year in this area, A. Tornado statistics for the site region indicate that from January 1951 through June 1978, 62 tornadoes have occurred within a 50-mile radius of the STP site (Reference 6-2). This is an average of 2.25 tornadoes per year. The same statistics also provide an estimate of 0.05 square miles for the tornado mean path area. Approximately 1,100 square miles of the area within 50 miles of the site is the Gulf of Mexico. Since the tornado data are normally provided for the areas over the land, the area of water was excluded in calculating the mean annual frequency of a tornado hitting the site. The result is 0.05 t = (2.25) 6,750 = 1.67 x 10-5 strike per year Tornado wind exceedance probability, &vlt, is more difficult to estimate because of the inaccuracy of indirect measuring techniques and the lack of a good analytical model for tornado behavior. An analysis of 4,582 tornadoes whose intensities were classified according to the Fujita F-scale is presented in Reference 6-9. Table 6-5 shows the histogram of frequencies of tornado windspeeds based on a Johnson SB distribution fit to the data for NRC tornado Region 1, which is applicable to the STP site. According to this distribution, the frequency of windspeed exceedance in Region 1 for tornado intensity F > F6 is obviously an upper bound for the frequency of windspeeds exceeding 360 mph. Therefore, 0.0005 was conservatively chosen as the value of 4vlt-Although no upperbound for the windspeed is indicated in this histogram, Reference 6-9 proposes a value of 300 mph as the maximum windspeed in Region 1. Other experts indicate that a tornado windspeed higher than 400 mph is not possible due to atmospheric friction. In this analysis, it is assumed that 400 mph is the maximum windspeed for Region 1 tornadoes. Finally, the annual frequency, 4, of excessive tornado windspeeds in excess of 360 mph is found by multiplying the values of &t and
&vl t-4 = (1.67 x 10-5)(5 x 10-4) = 8.33 x 10-9 per year 6.4.3.2 Tornado Wind Fragility of Structures As mentioned earlier, the design tornado windspeed for seismic Category I structures is 360 mph. According to Reference 6-2, to calculate tornado 6-12 0085H022035
wind load on such structures, maximum windspeed pressure, qmax, was obtained from the following formula 2 qmu(V) = 0.00256V where V is the total tornado windspeed. Therefore, for V = 360 mph, qmax(360) = 332 psf is obtained. For V = 400 mph, which was used as the maximum possible tornado windspeed, qmax(400) = 410 psf, which is higher than the design pressure calculated for a 360 mph windspeed by a factor of 1.23. The conservative factor of safety applied to material yield stress in order to obtain design allowable stresses was judged to be well within the margin of safety for Category I structures. Therefore, the lower end of tornado wind fragility curve for such structures is assumed to be in the vicinity of 400 mph. We conservatively assumed a step function fragility curve for wind load on the safety-related concrete structures at 400 mph. In other words, it is assumed that these structures do not fail under 400 mph wind load and that failure is certain above that value. This means that wind load is not likely to damage the safety-related concrete structures because, as discussed earlier, 400 mph is the maximum possible windspeed at the site. Therefore, the issue does not merit further investigation. There are some critical pieces of equipment outdoors that can be damaged at windspeeds far below 360 mph. For instance, power lines, transformers, and related equipment would be lost in weaker but more frequent tornadoes. This loss would result in a transient initiating event. The critical exterior metal vessels such as the refueling water storage tank and the auxiliary feedwater storage tank may also be subject to failure from negative or positive pressures generated by winds at tornado levels. However, these tanks are normally about 2/3 to 3/4 full when in service with resultant uniform internal pressures ranging to over 2,000 psf at the bottom walls. As long as they carry such a capacity,
! large external wind pressures cannot develop sufficiently to cause t
asymmetrical loads that would threaten buckling of the tanks, although the tanks top might be blown out from negative pressures. This, however, would not create buckling effects on the tank walls. Therefore, loss of contents from these metal vessels due to tornado wind load is highly l unlikely. 6.4.3.3 Tornado Wind-Initiated Scenarios l l Most tornadoes are capable of causing power lines to fail. Therefore, it can be reasonably assumed that offsite power is lost in a tornado event. However, additional failures caused by the tornado or other causes are required in order to have a core melt. As discussed, critical exterior l l 6-13 0085H022085 t
metal vessels are extremely unlikely to fail due to wind load. Moreover, even with the conservative assumption of step function fragility for safety-related concrete structures at a windspeed of 400 mph, the possible core melt scenarios will have frequencies less than 1.12 x 10-8 per year, which is the frequency of tornado winds exceeding 360 mph. The above analysis provides a good indication that the risk contribution from tornado wind scenarios is small. Based on the Preliminary Scoping Study Technical Review Board's recommendation, additional analyses considering scenarios involving nonrecoverable offsite power loss would be required to fully document the low contribution of tornado wind scenarios. 6.4.3.4 Tornado Missile Risk Tornado missile analysis involves information about the likelihood of a spectrum of available missiles in the plant vicinity, representation of the wind field in the tornado, and aerodynamic behavior relative to
" liftoff," and flight of the potential missile. The analysis leads to a spectrum of missiles and missile impact velocities with their respective probabilities. A detailed analysis that integrated all these effects for typical plant layouts has previously been performed (Reference 6-10).
The results of that work are considered to be reasonable gross estimates for the hazard of tornado missiles at STP. In Reference 6-10, calculations were made using tornado histories of each tornado region defined by the NRC. It used a typical two-unit plant layout to establish the target envelope and a 26-missile spectrum which includes the six missiles defined in the NRC Standard Review Plan, Section 3.5.1.4 (wood plank, steel pipe, steel rod, utility pole, automobile). In general, the 26-missile spectrum of Reference 1-21 is more conservative than the SRP spectrum with respect to damage potential. Calculations were made for the combined design life of a two-unit plant in which initially one unit is operational for 3 years while the other unit is being completed, followed by both units operating. Assuming 5,000 available missiles during the construction phase and 1,000 missiles during the operating phase, the study estimates the following mean values for the annual impact and damage frequency for all structures of a plant in NRC Region 1: o Case 1: One Unit Operating, Other Unit Being Built Mean: 7.51 x 10-6 e Case 2: Two Units Operating Mean: 3.33 x 10-6 The surface area of the plant studied in Reference 6-10 is about 500,000 square feet for each unit, which is nearly the same as the total exposed surface area of each STP unit. ! 6-14 0106H052085
However, the impact frequencies of Referenge 6-10 were calculated based on.a tornado strike frequency of 2.3 x 10-3 per year, tornado strike frequency at the STP site is 2.25 x 10 jhereas per year.the Adjusting the impact frequencies for the site-specific strike frequency, 7.36 x 10-8 and 3.24 x 10-8 are obtained for the annual impact frequency for the first and the second case, respectively. The above values are the annual frequency of inside wall scabbing for the safety-related structures (except the turbine building). All damages are believed to be localized and, therefore, it is very conservative to assume scabbing causes damage to all the contents. Even if it is assumed that such scabbing for Class 1 structures causes damage to enough vital components located near exterior walls and leads to core melt, the frequency of such an event is a negligible contribution to the total core melt frequency. As with the wind component of the tornado-initiated scenarios, the missile component appears to be a low contribution to ~r isk. Also, as with tornado wind, more work would be needed to fully document scenario involving a nonrecoverable loss of offsite power. 6.4.4 HAZARDOUS CHEMICAL ANAYLSIS This section presents a bounding analysis of the contribution to the risk from hazardous chemicals in the area surrounding the plant. According to Reference 6-2, there are four industrial facilities within 5 miles of the plant. Only two of these facilities store chemicals: the Crysen Terminal facility, located 4.8 miles from the plant, which has a storage capacity for 120,000 barrels of gasoline, and the Celanese Chemical Company facility, located 5 miles from the site. Because of its distance from the site, Crysen Terminal facility does not pose any hazard to the plant in the event of gasoline-air explosion. The chemicals stored at and shipped to and from the Celanese Chemical Company include five chemicals that could be considered potential hazards to the plant. These are anhydrous ammonia, hydrochloric acid, naptha, acetic acid and vinyl acetate. The chemicals are shipped to or from the plant via road FM 521 (nearest distance to the plant 0.89 miles) or the Colorado River (nearest distance to the plant 2.75 miles). To perform a bounding analysis, we use the results of a detailed analysis of the risk associated with hazardous chemicals for the Midland Nuclear Plant (Reference 1-21). The hazard to the Midland plant has been shown to be dominated by the Dow Chemical plant to the north and by the Dow Corning plant to the east. Both sites have storage tanks, rail lines, and pipelines where large inventories of a variety of hazardous chemicals are stored. The distance of most of these storage facilities to the Midland site is approximately 1 mile. Reference 1-21 calculated a total frequency of 7.55 x 10-6 per year for
.all relevant plant damage states resulting from relases of toxic chemicals from the Dow l The analysis was based on an annual
' frequency of 1.74 x 10 k ants.for toxic gas release due to dirgct storage tank failure and an annual frequency of between 1.8 x 10-'+ and 6-15 0085H022085
t 5.4 x 10-4 for indirect release depending on the chemical substance stored. The indirect release is the result of tank failure due to shock wave from the explosion caused by another tank failure or some other mechanism. Another factor considered in the analysis was the availability of the hazardous gas monitoring system for the control room air intake system. The results in general indicated that all of the frequencies of the resulting plant damage states were at least more than one order of magnitude smaller than the frequency of same plant damage states due to other initiators. There are several factors that would lead us to believe that the core melt frequency due to hazardous chemicals at STPEGS is bounded by the frequency calculated for the Midland plant. One factor is the quantity of the chemicals stored or shipped to or from at the Celanese Chemical Company, which is substantially smaller than that of Dow plants near Midland. Second, the distance of the Celanese plant to STPEGS is almost five times more than the distance between Midland and Dow plants, which i would make any vapor clouds with high concentration of hazardous material less likely to reach the STPEGS site. Also, similar to the Midland plant, STPEGS is equipped with detection, alarm, and automatic control room isolation for the hazardous material. It is therefore concluded that the frequency of core melt at STPEGS due to release of hazardous chemicals is much smaller than 7.55 x 10-6 per year and that no further analysis of the subject is needed. There are several potentially hazardous chemicals stored at the site. These are relatively small quantities of Anhydrous Ammonia, Ammonium Hydroxide and Hydrazine. Scoping analysis of the accident scenarios initiated by large releases of these subs melt frequency is no greater than 3 x 10 gances indicate per year. that the core This frequency is calculated based on jarge leaks of ammonia from its storage tank with a frequency of 2 x 10- per year; a probability of 0.16 that the wind direction would be N, NNW, or NW at the time of release, resulting in a high concentration of toxic gases in the control room; and 0.1 as the conditional probability of core melt given such an accident scenario. The conclusion is that the frequency of core melt scenarios initiated by release of onsite hazardous chemicals is bounded by other core melt scenarios. 6.4.5 EXTERNAL FLOODING The STPEGS site is located about 10 miles north of the Gulf of Mexico. The Colorado River to the east of the site passes about 2.75 miles from the plant at its nearest distance from the site. . The only major dam upstream from the site is Mansfield Dam. There is also a reservoir south of the site with surface area in excess of 14 square miles, wnich is used as a cooling pond. According to meterological data (Reference 6-2), tropical storms and hurricanes are frequent in the general area surrounding the site. The average frequency of such storms for the entire Texas cost is approximately one per year. In 1961, Hurricane Carla passed near the 6-16 0085H022185
site and dropped 17.10 inches of rain in Bay City,12 miles north-northeast of the site. Other hurricanes that passed near the site in recent years were Celia in 1970, Allen in 1980, and Alicia in 1983. This gives a frequency of 0.2 per year for occurrence of major hurricanes. Reference 6-2 states that a breech of the cooling reservoir embankment is capable of producing the maximum possible flood level at the site. The critical structures of the plant are designed to withstand the effects of such flooding. The possible combination of a major storm and failure of the cooling reservoir embankment may have potential for plant damage. In l fact, this combination represents the only identified external flood event with possible implications on risk. If we knew nothing about the l structure and fragility of the reservoir embankment, use of historical i generic dam fai' ure data would yield an annual frequency of major site flooding of 10 6 and a much lower frequency of core damage from such events would be expected. However, we do know considerably more about this embankment than some unspecified dam--information that makes us bel { eve that the frequency of site flooding would be much less than 10- per year. Amendment 43 to Reference 6-2 explains that significant measures have been taken to ensure positive protection of the embankment against failure. The cooling reservoir water level is controlled; i.e., it is not subject to uncontrolled increases in water level such as a dam collecting water from a water shed. The embankment conditions and hydrostatic pressure are monitored, yielding information that provides a basis to decrease embankment water level prior to danger of failure. There is a program for the reduction of hydrostatic pressure beneath the embankment through the use of relief wells at locations along the perimeter of the reservoir. The embankment is protected against errosion. Conservative calculations combining the effects of rain, wind setup, and wave runup show that overtopping will not occur. Additionally, the STP impoundment does not have a connection to dissimilar materials (e.g., the side of a hill) so that historical data that resulted in failure associated with this interface is not applicable.
6.5 REFERENCES
6-1. Houston Lighting & Power Company, " South Texas Project, Fire Hazards Analysis Report," April 1984. l 6-2. Houston Lighting & Power Company, " Final Safety Analysis Report, ! South Texas Project, Units 1 and 2, April 1984. , 6-3. U.S. Atomic Energy Commission, " Aircraft Consideration ! Preapplications Site Review by the Directorate of Licensing," in the matter of Portland General Electric Company, Boardman Nuclear Plant, Boardman, Oregon, Project Number 485. I
- 6-4. U.S. Department of Transportation, Federal Aviation l Administration, "FAA Statistical Handbook of Aviation, Calendar l Year 1979," December 1979.
l 6-17 0085H022185 1
6-5. National Transportation Safety Board, " Annual Review of Airport Accident Rates, Calendar Year 1980," NTSB-ARG-80-1, May 1980. 6-6. Shaffer, D. H. , S. C. Chay, D. K. McLain, and B. A. Powell,
" Analysis of the Probability of the Generation and Strike of Missiles from a Ni: clear Turbine," Mathematics Department, Westinghouse Research Laboratories, Westinghouse Electric Corporation.
6-7. Buch, S., and P. Heasier, " Probability of Turbine Missile C.. -ation," paper. presented at EPRI Steam Turbine Missile Disc Integrity Seminar, New Orleans, Louisiana, April 6-8, 1981. 6-8. Thom, H. C. S., " Tornado Probability," Monthly Weather Review, No. 91, pp. 730-736, 1963. Twisdale, L. 6-9. A., " Tornado Data Characterization and Windspeed Risk," Journal of Structural Division, Proceedings of ASCE, Vol .104, No. ST10, October 1978. 6-10. Twisdale, L. A., W. L. Dunn, and J. Cho, " Tornado Missile Risk Analysis," Electric Power Research Institute Inc., EPRI NP-768, May 1978. 6-18 0085H022185
TABLE 6-1. AIRCRAFT ACCIDENTS AND ACCIDENT RATES: U.S. AIR CARRIERS, 1970 THROUGH 1979 Number of Fatal Accidents Aircraft Year Landing Miles R own Cr R tes* Inflight and (103 ) Takeoff 1970 4 4 2,684,552 1.49-9 1971 6 2 2,660,731 2.25-9 1972 3 5 2,619,043 1.14-9 1973 5 4 2,646,669 1.89-9 1974 7 2 2,464,295 2.84-9 1975 1 2 2.477,764 0.40-9 1976 2 2 2,568,113 0.78-9 1977 4 1 2,684,072 1.49-9 1978 4 2 2,742,860 1.46-9 1979 4 2 2,889,131 1.36-9
- Accidents per aircraft mile flown.
NOTE: Exponential notation is indicated in abbreviated form; i.e., 1.49-9 = 1.49 x 10-9 0088H122784 6-19
TABLE 6-2. FATAL ACCIDENT RATES FOR U.S. GENERAL AVIATION AIRCRAFT Fatal Accident Rates Per Miles Flown Year Single Multiple All Types Engine Engine 1972 2.63-7 8.7-8 2.11-7 1973 2.52-7 8.2-8 2.09-7 1974 2.45-7 7.6-8 1.88-7 1975 2.30-7 6.9-8 1.71-7 1976 2.02-7 6.4-8 1.66-7 1977 2.03-7 5.1-8 1.59-7 1978 2.02-7 7.7-8 1.59-7 NOTE: Exponential notation is indicated in abbreviated form; i.e., 2.63-7 = 2.63 x 10-7, i l 0088H122784 6-20
TABLE 6-3. PROBABILITIES f2 AND f3 0F TARGETS DUE TO UNITS 1 AND 2 TURBINE MISSILES (SHEAR FAILURE) Turbine Generator Unit 1 Unit 2 Target
- f f f2 (10-3) f 3 2 3 RC8 1 1.4303 .0118 2.59-9 RC8 2 .3450 0.0 0.0 DGB 1 .3176 .1583 7.70-9 .1615 .7850 1.94-8 DGB 2 .5206 .5122 4.08-8 .3176 .1583 7.70-9 FHB 1 FHB 2 EAB 1 .2155 .1206 3.98-9 MEAB 2 .0190 0.0 0.0 AFW .2018 .8332 2.57-8 Tank 1 AFW Tank 2 IVC 1 .4072 .8178 5.09-8 IVC 2 .0377 .1709 9.79-10 ECW Intake .1350 .8089 1.67-8 Structure Total 1.2399 4.85-8 2.8689 1.27-7 i *RCB 1 designates reactor containment building Unit 1; DGB 1 designates diesel generator building Unit 1. etc. EAB designates mechanical / electrical auxiliary building, comprising the MAB and the EA8.
- NOTES
i
- 1. Blanks are for those targets either located outside the low trajectory missile
- strike zone or shaded by other targets.
, 2. Exponential notation is indicated in abbreviated fonn; f.e., 2.59-9 = 2.59 x 10-9 caufroN: PREumeNARY RESULTS-l IMPORTANT UNCERTAINTIES i DESCRIS40 IN SECTION 2
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6088H122784
l TABLE 6-4. PROBABILITIES f2 AND f3 0F TARGETS DUE TO UNITS 1 AND 2 TURBINE MISSILES (SHEAR AND ROTATION FAILURE) Turbine Generator Unit 1 Unit 2 Target
- f f f f2 (10~) 3 2 3 RC8 1 1.5660 0.0 0.0 RCB 2 .3779 0.0 0.0 DGB 1 .5288 0.0 0.0 0.0 0.0 0.0 DGB 2 .7882 .3782 4.56-8 .5288 0.0 0.0 FHB 1 FHB 2 MEA 8 1 .2534 .0073 2.91-10 MEA 8 2 - .0234 0.0 0.0 AFW .2066 .7849 2.48-8 Tank 1 AFW Tank 2 IVC 1 .4502 .8207 5.65-8 IVC 2 .0457 .1438 1.01-9 ECW Intake .1422 .7197 1.57-8 Structure Total 1.7640 4.66-8 3.1472 9.73-8
*RC81 designates reactor containment butiding Unit 1; DGB 1 designates diesel generator building Unit 1. etc. EAB designates mechanical / electrical auxiliary building, comprising the MA8 and the EAB.
- NOTES:
! 1. Blanks are for those targets either located outside the low trajectory missile strike , . zone or shaded by other targets.
- 2. Exponential notation is indicated in abbreviated form; i.e., 4.56-8 = 4.56 x 10-8, f
- cAurioso
- resussueAny nasutts-l twontaut uncanvasaties OEsCateEO IN sECTeoN 2 6-22 0088H122784
N TABLE 6-5. TORNADO WINDSPEED FRACTIONS I indspeed F-Scale Range Fe enc (Np g ) (mph) 0 40 to 72 0.2440 1 72 to 112 0.4241 2 112 to 157 0.2375 I 3 157 to 206 0.0735 4 206 to 260 0.0172 5 260 to 318 0.0032 6 318 to 380 0.0005
>6 > 380 f
0088H122784 6-23
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4 j 7. SCOPING CONTAINMENT RESPONSE ANALYSIS
7.1 INTRODUCTION
This section presents a scoping overview of the containment response and source term characteristics of the STPEGS design. The assessment is l based on an inspection of the containment design and an evaluation of its characteristics, relying to a large extent on PLG's experience from
~
detailed containment response analyses of Westinghouse PWRs in large, dry containments. This section defines the plant damage states used in the Preliminary Scoping Study, develops an experience-based C matrix,* and develops release category definitions that can be used to qualitatively assess the health impact potential of the various release states.- 7.2 STPEGS CONTAINMENT DESIGN i The STPEGS containment is a very large, dry PWR containment with a design pressure of 56.5 psig and a net free volume of 3.4 x 10 6 ft 3. The free volume is approximately 25% larger than other large, dry PWR containments with a comparable design pressure. This design feature will favorably extend the time when containment failure would occur in the absence of l heat removal by active systems during postulated core melt accidents. i The containment is steel-lined and post-tensioned. Our experience with this type of containment indicates that a realistic failure pressure can 4 be expected to be between 150 and 200 psia. The positive impact of high containment failure pressure on the containment response will be j discussed in Section 7.4. The reactor cavity in the STPEGS design has a total floor area of t 650 square feet, or about 30% larger than the Zion, Indian Point, or Seabrook reactor cavity floor area. The reactor cavity is level with the containment floor, which should facilitate the dispersal of debris out of i the reactor cavity, if a dispersal path exists. However, from the available information, it was not possible to determine the structural strength and watertightness of the doors that are to be installed at the , current access door to the reactor cavity, which is located at Elevation -11'. If this boundary around the reactor cavity is very strong and watertight, most of the core debris emanating from the reactor vessel may be trapped in the reactor cavity. In this case, the question of water accessibility to the cavity becomes important. Again, based on the information available, it was not possible to identify clear and unambiguous paths by which water (several hundred gpm) could access the reactor cavity to cool core debris on the floor. The access paths examined included: e Manway at Elevation -11' ' e Manway at Elevation -2' OThe C matrix, as described in Section 3, contains the results of the containment event tree quantification and is used to map plant damage states into release categories. 7-1 010eH052085
e Instrument Tube Penetrations e Reactor Cavity Floor Drains to Containment Sump e Orainage Around the Reactor Vessel from the Refueling Pit Floor Water accessibility to the reactor cavity is important because with the large cavity size, the debris bed is sufficiently shallow to be coolable even if all the core debris is trapped in the cavity. Debris coolability requires, however, that the debris is flooded with water, which requires that the water level equalizes between the reactor cavity and the outside area. Since it was not possible to establish a clear access path for water, it was decided to examine two configurations: o Case I. Debris is trapped in the reactor cavity, water has no access to the reactor cavity. e Case II. Debris is dispersed from the reactor cavity and water has access to flood the debris in the reactor cavity. A separate analysis including the C matrix and separate sets of release categories was performed for each case. 7.3 PLANT DAMAGE STATES The plant damage states define a discrete set of end states for accident , sequences that lead to core damage. As the pinch point between the plant model and the containment model, they define the end states for accident sequences through the plant event trees. They also define the initial conditions for the containment event tree. Therefore, two observations 1 are important: e The plant damage states must be defined so that all the plant event sequences collected within a given PDS have similar characteristics in the containment response model and can be represented by one representative initiator in the containment response model. e Even if a containment and source term analysis is planned for a future time, it is necessary to define the plant damage states as the plant model end states so that it will not be necessary to revise and requantify the plant model later when the containment analysis is performed. The plant damage states for the STPEGS Scoping Study are shown in Table 7-1. A PDS is identified by a number / letter designator ranging from 1A to 2J. States 1 and 2 distinguish sequences with RWST injection from those where only the RCS inventory is released to the containment. For "1" states, there will only be a few inches of water on the containment floor, whereas for the "2" states, the depth of water will be several feet. States A to D characterize intact containment configurations differentiated according to the availability of contair. ment heat removal and/or fission product removal. The success criteria are indicated in the bottom row. One train of sump recirculation or two fan cooling units are adequate for containment heat removal. Two spray recirculation trains are required for fission product scrubbing. 7-2 0108H052085
Containment states E and F represent large containment bypass configurations differentiated according to fission product scrubbing
.along the release path. Containment states G through I represent small containment bypass leak configurations differentiated according to the same criteria used for the intact containment states A through D.
7.4 RELEASE CATEGORIES The purpose of this section is to define a set of release categories appropriate to the STPEGS design, based on potential containment failure modes and radionuclide release characteristics. The intent is not to quantify these-release categories; rather, it is to associate them with health impact potential based on our experience with similar release categories from completed PRAs. Table 7-2 lists the six basic release category distinctions according to the containment failure modes, numbered 1 through 6. The "T" designator is used to indicate that the release categories are applicable specifically to STPEGS. For each of the six basic containment failure modes T1 to T6, release categories are defined according to the following three criteria: e A "C" designator means that containment heat removal is available to reduce the containment pressure and thus the driving force for leakage. O A "B" designator means that the containment spray system is not available to scrub fission products from the containment atmosphere. O A "V" designator means that a vaporization release has occurred as a result of debris penetration into the concrete basemat. Furthermore, release category set T1 is always assumed to include an oxidation release of fission products as a result of fine particulate fragmentation of the debris in the containment atmosphere. The full set of release categories used for Cases I and II is tabulated in Table 7-3 according to these abbreviations. 7.5 THE C MATRIX Table 7-4 shows the C matrix estimated for the Scoping Study for Case I, no water access to the reactor cavity. Table 7-5 shows the C matrix for Case II. The plant damage states are listed on the lef t-hand side of the l table, whereas the appropriate release categories are listed across the
- top. The' numerical values indicate the conditional probability that a given plant damage state will lead to a given release category. These numerical values were estimated on the basis of PLG's prior experience with detailed containment response analyses for large, dry PWR 3
containments. An c entry designates a very low value (less
- than 0.01). In past analyses, the c values usually were found to be l between 10-3 and 10-6 It is noted that for PDSs with intact
( containments (A through D), the early containment failure release l category T1 is very unlikely because of the very high anticipated I containment failure pressure that is well above the range of peak l 7-3 0108H052085
pressures expected from vessel failure blowdown pressure spikes or from hydrogen burns. Therefore, early containment failure can be expected to be extremely unlikely. The remaining entries basically reflect the competition between release categories for late overpressure failure (T3), basemat melt-through (T4), and intact containment (TS) as influenced by the availability of containment systems. Plant damage states that represent bypassed containments are assigned to the appropriate containment bypass release category (T2, T6) with a probability of 1.0. 7.6 SITE LONSEQUENCE ANALYSIS All release categories of the previous section have been related to those defined in the SSPSA. Based on a careful review of the consequence curves for each release category and some knowledge of the STPEGS site characteristics, we have assigned each release category to one of the following four coarse " impact categories": o Category I. Significant potential for early and latent health effects. e Category II. Significant potential for latent health effects and possibly small numbers of early health effects. e Category Ill. Potential for, at most, only small numbers of latent health effects. Two S matrices based on judgment are given in Tables 7-6 and 7-7, one for each C matrix defined earlier. These matrices simply show how release categories are mapped into the impact categories. Interestingly, the CS product matrix in Table 7-8 is the same for either pair, C ISg or CgIS gg. Therefore, on the basis of the analysis performed in the Scoping Study, it does not appear that the uncertainty about water flows into the reactor cavity that gave rise to the definition of Cases I and 11 is a particularly risk-sensitive issue. l 4 7-4 0108H052085
TABLE 7-1. PLANT DAMAGE STATES FOR THE STPEGS PRELIMINARY SCOPING STUDY Containment Intact at Core Melt Initiation Yes No RiiST Containment Leak Size Containment Functions Available Initiated >3 Inch Diameter At Vessel Equivalent
<3 Inch Diameter Equivalent Melt-Through Release Containment Functions CHR + FPR CHR Dnly FPR Only None Filtered Not Filtered CHR + FPR CHR FPR None State A B C D E F G H I J No 1 IB ' ID 1F lH IJ y Yes 2 2A 28 2C 2D 2E 2F 2G 2H 21 2J u,
~
"2/6 1/31 2/6 2/3 None 2/3 None
[2/6 1/3' 2/6 2/3 None 3 , \CFC SRC[ CFC CSR CSR (CFC SRC[ CFC CSR 2/3 [1/6 1/31 2/3 'l/6 1/3' CSR SRC[ CSR LCFC \CFC*SRC[ Legend: Plant Damage State Not Possible CHR = Containment Heat Removal FPR = Fission Product Removal SRC = Sump Recirculation Cooling CSR = Containment Spray Recirculation CAUTION: PRELIMINARY RESULTS. CFC = Containment Fan Coolers IIdORTANT UNCE RTAINTIES DESCRISED IN sECTION 2 0087H122884
r TABLE 7-2. DEFINITION OF RELEASE CATEGORY SETS BASED ON CONTAINMENT FAILURE MODES Release Category Description Set T1 Airborne release due to early containment failure. Source term includes oxidation release from fine l particulates involving 50% of core inventory as a result of debris fragmentation and dispersal into the containment atmosphere. T2 Early increase-in the containment design basis leak rate due to early pressure spike or small penetration , isolation failure. T3 Airborne release due to late overpressure failure of containment, which is due to either lack of containment heat removal or gas generation during concrete penetration. ' T4 Ground release due to concrete basemat melt-through of , the debris prior to aboveground containment shell i failure. T5 Containment integrity is maintained. Represents slow release to atmosphere at design leakage rate. T6 Containment is not isolated (large penetration) or is bypassed due to initiating event failures. Containment remains unisolated, resulting in continuous release. CAUTIOps: PRELitetNARY RESULTS-140pORfANT UNCERTA8NTIES 00$CRettD IN SECTIOff 2 7-6 0087H122884-
TABLE 7-3. DEFINITION OF RELEASE CATEGORIES FOR STPEGS PRELIMINARY SCOPING STUDY Release Containment Containment 0xidation Vaporization Category Case Containment Failure Mode Spray Operating Heat Removal Release Release TI II Early Containment Failure Yes - Yes No TIB II Early Containment Failure No -- Yes No T1V I Early Containment Failure Yes - Yes Yes TIBV I Early Containment Failure No - Yes Yes T2 II Early Increased Leak Rate Yes No No No T2C II Early Increased Leak Rate Yes Yes No No T2B II Early Increased Leak Rate No No No No T2CB II Early Increased Leak Rate No Yes No No T2BV I, II Early Increased Leak Rate No No No Yes T2CBV I, II Early Increased Leak Rate No Yes No Yes T2V I Early Increased Leak Rate Yes No No Yes T2CV I Early Increased Leak Rate Yes Yes No Yes T3 II Late Overpressure Yes No No No T3B II Late Overpressure No No No No y T3BV I, II Late Overpressure No No No Yes u T3V I Late Overpressure Yes No No Ycs ' T4 II Basemat Melt-Through Yes No No No T4BV I, II Basemat Melt-Through No No No Yes T4CBV I, II Basemat Melt-Through No Yes No Yes T4V I Basemat Melt-Through Yes No No Yes T4CV I Basemat Melt-Through Yes Yes No Yes T5 II Containment Intact Yes Yes No No TSB II Containment Intact No Yes No No TSBV I, II Containment Intact No Yes No Yes TSV I Containment Intact Yes Yes No Yes T6 II Large Containment Bypass Yes - No No T6B II Large Containment Bypass No - No No T6BV I, II Large Containment Bypass No No Yes T6V I Large Containment Bypass Yes - No Yes CAUTIOes: PRELIGAINARY RESULTS-IAAPORTANT U8sCERTAINTIES DESCRISED IN SECTION 2 0087H122884
TABLE 7-4. C MATRIX FOR STPEGS PRELIMINARY SCOPING STUDY CASE I: NO WATER ACCESS TO REACTOR CAVITY; DEBRIS NOT C00LABLE Plant STPEGS Release Category Damage State T1V T2V T2CV T3Y T4V T4CV TSV T6V T1BV T2BV T2CBV T3BV T4BV T4CBV TSBV T6BY 1B c* .05 c .7 .25 1D c c .9 .1 IF 1.0 1H 1.0 IJ 1.0 2A c .05 c .7 .25 y 2B c .05 c .7 .25 m 2C c .05 .85 .1 20 c c .9 .1 2E 1.0 2F 1.0 2G 1.0 2H 1.0 2I 1.0 2J 1.0
*c = negligible probability CAUTION: PRELIRAINARY RESULTS-IAAPORTANT UNCERTAINTIES DESCRISED los SECTION 2
TABLE 7-5. C MATRIX FOR STPEGS PRELIMINARY SCOPING STUDY CASE II: WATER ACCESS TO REACTION CAVITY; DEBRIS C00LABLE Plant STPEGS Release Category Damage State T1 T2 T2C T3 T4 T5 T6 T1B T2B T2CB T2BV T2CBV T3B T38V T4BV T4CBV TSB T5BV T6B T6BV IB c* c .05 c .7 .25 1D e c .9 .1 1F 1.0 1H 1.0 IJ 1.0 2A c .05 e c .95 2B c .05 e c .95 u 2C c .05 .95 c E 20 e c 1.0 c 2E 1.0 2F 1.0 2G 1.0 2H 1.0 21 1.0 2J 1.0
*c = negligible probability CAUTION: PRELieseNARY RESULTS-ItePORTANT useCERTAINTIES DESCR10ED IN SECTIOes 2 0087H122884
TABLE 7-6. S MATRIX FOR CASE I - THE CONDITIONAL PROBABILITIES OF IMPACT CATEGORIES GIVEN RELEASE CATEGORIES (Corresponds to C Matrix for Case I) Impact Category Release Category g gg ggy T1V 1 T2V 1* T2CV 1* T3V 1 T4V 1 T4CV 1 TSV 1 T6V 1 T1BV 1 T2BV 1 T2CBV 1 T3BV 1 T4BV 1** T4CBV 1 T5BV 1 T68V 1
- Inconsistent with Seabrook PRA, which conservatively equated these low frequency categories to the more severe and frequent T2BY. Although no quantitative site model calculations have been performed, operation of containment spray should eliminate the possibility of early effects and minimize latent effects.
**In this basemat melt-through category, no water is available to trap fission products and high containment pressures can cause fission products to " burp" to atmosphere between the failed basemat and the underlying concrete.
CAuflON: PMGLIMINARY RtSULTS-IMPOMf ANf uMCERTAINfits DESCRISED IN $$CilON 2 7-10 0087H122884
TABLE 7-7. S MATRIX FOR CASE II - FOR CONDITIONAL PROBABILITIES OF IMPACT CATEGORIES GIVEN RELEASE CATEGORIES (Corresponds to C Matrix for Case II) Impact Category Release Category g gg ggy T1 1 T2 1* T2C 1* T3 1 T4 1 TS 1 T6 1 T1B 1 T2B 1 T2CB 1 f T2BV 1 T2CBV 1 T3B 1 T3BV 1 T48V 1** T4CBV 1 TSB 1 TSBV 1 T6B 1 T60V 1
- Inconsistent with Seabrook, which conservatively equated these low frequency categaries to the more severe and frequent T2BV. Although no quantitative site model calculations have been performed, operation of containment spray should eliminate the possibility of early effects and minimize latent effects.
**In this basemat melt-through caution: entuminany nesutts.
category, no water is available to iveontaur uscantaimries trap fission products and high o,,cn,,,o , . erios , containment pressures can cause fission products to " burp" to atmosphere between the failed basemat and the underlying concrete mud mat. 0087H122784 7
l TABLE 7-8. THE CS PRODUCT MATRIX-CONDITIONAL PROBABILITIES 0F IMPACT CATEGORIES GIVEN PLANT DAMAGE STATES Plant Consequence Category Damage State g gg ggt i 1B 1
- 10 1 1F 1
- 1H 1 t
IJ 1 j 2A 1 i 2B 1 2C 1 , 2D 1
- 2E 1 j 2F 1 I
2G 1 2H 1 21 1 2J 1 i CAUTION: PRELIMINARY RESULTS-IMPORTANT UNCERTAINTIES DESCRISED IN SECTION 2
-2 0087H122884
APPENDIX A RESOLUTION OF HL&P COMMENTS ON STP PROBABILISTIC SAFETY ASSESSMENT BASELINE STUDY INTERIM REPORT DATED JULY 1984 Comment Number: 1 Report
Reference:
General Comment: Note caution related to " realism" contained in letter of transmittal and revieu the vording pf this Report accordingly.
Response
Appropriate revisions were made to the report, particularly to Section 1 (Introduction) and Section 2 (Results). These revisions emphasize the limitations of the baseline risk model, the substantial uncertainties, especially for a Baseline Study, and how conservative assumptions were taken into account in the quantification of uncertainty to provide a balanced perspective of the results. Comment Number: 2 Report
Reference:
General Comment: The Interim Report ehould summarine the resulte pf the Baseline Study as auch uithout progrannatic references to any future work, including the Nanagement Plan.
Response
Programmatic references to future work have been deleted from the report. Care has been taken to explain the limitations of the study. Areas where additional work would be needed to remove those limitations are identified. Comment Number: 3 Report
Reference:
General Comment: Figures contained in the Interim Report chould prominantly display a caution in their une and chould reference a acetion of the report that providee an appropriate discuccion as to the meaning to be ascribed to the reaulte at this etage of analyele. A-1 0081H021985
r j l Response: , A warning or caution label, " Preliminary Results for Project Scoping i Purposes Only" was applied to all results, figures and tables in the draft report. A new caution label; " Caution: Preliminary Results - ! Important Uncertainties Described in Section 2" has been applied to > these pages in the final draft. ! Comment Number: 4 Report
Reference:
General ! Comment: - An assessment of the consequences of radioactive releasee was not in the ocope of the Study and vac not performed per ca. Therefore att referenece in the Study to " consequences" or " health effecte" should be deleted and references should be made only to " release categorica. " The only exception should be in the definition of the '
" Release Categories" currently contained in Tabic 2-2. l t
Response: l L Assessment of consequences in the Baseline Study was limited to a ; qualitative grouping of accident sequences according to their approximate p"otential for offsite health effects. References to !
" consequences and " health effects" have been removed in accordance i with comment.
For this revised Baseline Study, the four consequence categories have been condensed into three impact categories as follows: Old Consequence New Impact Category Category I I II II III II IV III Old categories II and III have been combined because they were quite similar. Both provided substantial potential for latent effects; OLD Category II only gave a small potential for a very few early effects. Keeping them separate provided very little additional information and led to some confusion among reviewers. Comment Number _: 5 Report Reference _: General A-2 0081H122884
Comment: In scenarios reculting in station blackout, the availability of alternato courece of AC pouer chould be noted in mitigating the consequencco of the blackout. The Technical Support Center or Balance of Plant diecele are available ac auto-otart (on undervoltage) AC pouer sources. Additionally, the Emergency Transformer can be energined via operator action and through celective cuitching, placed on the ESF buses to regain AC pouer to vital loads (although 3 buses can be energined, only 2 full trains of icada can be handled by the transformer).
Response
The comment is noted. The B0P diesel supplies power to support the instrument air compressor which is required to operate the EAB dampers for the smoke purge mode of HVAC. It was modeled for loss of offsite power scenarios that could benefit from smoke purge cooling. In a completed, full-scope Level 1, 2, or 3 PRA, a detailed recovery analysis of station blackout is typically performed. An example of such a detailed blackout recovery analysis is found in the SSPSA (Reference A-1). In such an analysis for STPEGS, the 80P diesel generator would be fully taken into account. In the more limited-scope analysis of this Baseline Study, accident sequences involving postulated failure of all three Class 1E diesel generators were incorporated into the model and recovery of offsite power was included in the quantification of this sequence. As analyzed, this sequence made a small contribution to risk and to core melt frequency as indicated in Section 2. In a completed Level 1 PRA, there are several key uncertainties whose removal would be sought regarding the analysis of station blackout. The uncertainties include the possible variation between generic j diesel generator failure rates and common cause parameters and those uniquely appropriate for STPEGS, the performance of RCP seals with loss of CCW and the attendant loss of seal injection, and the effect of the B0P diesel. Resolution of these uncertainties is outside the scope of the Baseline Study and apparently not as important as some of the other sources of uncertainty mentioned in Section 2. Comment Number: 6 Report
Reference:
Abstract Comment: The purpoco of an abstract in to convoy the cubject of the report indicating ito intent, methodology, and generut statement of recutto and limitatione. The third and fourth centencco of the accond paragraph are inconcietent uith the intent of the abotract, and the cubject matter chould be racerved for tha report. The abotract should recognino llL&P ao the Prodcat flanager for the South Texac A-3 0081H122884
I Project and chould indicate that the Study vaa undertaken on HL&P's ovn initiative to provide carty insights as to the rick [censitivities] for design and procedural purpoces. The Study was not performed due to regulatory requiremente.
Response
The comment is noted and the suggested text modifications have been made. Comment Number: 7 Report
Reference:
Section 1 Comment: Before the Objectivea of the Baccline Study are stated, the introduction should describe the program undertaken by HL&P leading to the Interim Report and the preliminary nature of the rceutte contained in the Report. In thic connection, Probabilictic Rick Accesoment should be defined, the scope of the variot.o levela of PRA chould be diccuaced, the relationahip of the Baseline Study to a fuit coope PBA chould be clearly stated, and what is meant by a risk management progmm chould be indicated. The program undertaken by HLSP up to thic point has been compoced of the Basetino Study and technology trancfor, including training. Provide the Objectives of the Basetina Study only. The first and most important Objective of the Dacetine Study la to " Provide a banic for timely feedback of risk management insighta into the proccoo of completing construction on STPEGS," and should precede att other objectivas tieted.
Response
The comment is noted and the suggested text modifications have been made. Comment Number: 8 Report Reference _: Section 1, page 1-3 l Comment: The pamgraph eucceeding the tact of the objectivea chould be placed in the " Introduction" with a caution and a perepcotivo ao to the purpose and n; caning of numerical reculta at this stage of the rick modet development. It is alco appropriate that the introduction etcarty explain the meaning of the term " Point Ectimate", ito relationchip to the reculto, and a caution in making car.panicono at thic etage in the PRM devotopn.ent.
Response
The comment is noted and appropriate text modifications have been made. A-4 0081H122884
Comment Number: 9 Report
Reference:
Section 1 Comment: Scotione 1.2, 1.2.1 & 1.2.2 should be deleted in their entirety cince they provide general discuacione not related directly to the STPEGS etudy resulto. The Baccline Study doce not in any way comprice an analyaic of consequences and should not be related or compared to any previous full scope studica and the icvel of detail, documentation, degree of realism, methodology, or magnitude of reautta provided by those otudico. Judgmente and comparieone in either methodology or resulto are conaidered inappropriate due to ti e preliminary nature of and approach to the analycio of the Baseline Study as compared to other otudica.
Response
The comment is noted. These report sections have been rewritten to make use of a general PRA discussion that focuses on the specific objectives and limitations of the Baseline Study performed for STPEGS and on the use of PRA as a risk management tool. Comment Number: 10 Report
Reference:
Section 1.4 - Report Guide Comment: The timitatione of the Bacclinc Study rick model ahould be sumarized more prominantly in the Introduction.
Response
An increased visibility of the limitations of the Baseline Study now appears in the Introduction (Section 1) and in the Summary of Results (Section 2). Comment Number: 11 Report
Reference:
Section 4 - Event Sequence Model Comment: An overcooling trancient that recutta in reactor veccol rupture in included in the cuent acquence analycle. The Weatinghouac Ouner'o Group haa performed calculationa that demonctrato that total RTndT for STP vencela la very 200 (i.e., 88 degrcce F for Unit 1 and 68 degreco F for Unit 2 for 32 effective full pocer yearo). Therefore, it io not a concern for STP and chould be co indicatcd in the Report: (Ecference ST-QG-HL-90169). A-5 0081H122834
Response
The comment is noted. The inclusion of a question of reactor vessel integrity following certain overcooling transients has become standard practice in the construction of event sequence models at PLG. We recognize the cited reference as strong evidence that the potential for reactor vessel failure resulting from pressurized thermal shock to be low relative to other plants. However, the conservative bounding treatment of this phenomena in the Baseline Study has shown that such scenarios are negligible risk contributors at STPEGS. Similar conclusions were reached for Seabrook without the need for performing an accurate estimate of reactor vessel failure probability. Hence, we concur that PTS is not a concern for STPEGS from a PRA perspective. The only identified scenario in which cooling below the RTndT is possible is the one involving injection of cold RWST water following an RCP trip. With no loop flow, it is possible to subject the vessel wall to RWST temperature water. This scenario was not modeled in the Baseline Study, but from other PRAs, is not expected to make a significant contribution. Comment Number: 12 Report
Reference:
Section 5 - Systems Analysis - CVCS Letdown Isolation Valves Comment: The design of thic cyctem hac been changed--valuce LCV466 and LCV468 uitt both have fait-cloced pneumatic actuatore. [Ac a revicion to XL&P'a commento, instructione vore given not to change accumptione regarding the type of actuatore in the CVCS totdcun linee.]
Response
HL&P has subsequently notified PLG that these valves will be motor-operated. They have no significant impact on the frequencies of Categories I, II, or III. Comment Number: 13 Report
Reference:
Section 5 - Systems Analysis - SG PORVS Comment: The PORV'c have enough energy otored in their accumulatore to operate one full open and clocod cycle. Thio to cufficient to ensura the FORY vill fait in the clocod pocition on loca of pouer. In order to provida cooling of the atoam generator, tuo methode are availabic. A chort term method to to increaco the cecondary cidc precoura high enough to actuate a apring operated SRV. The cocond method le manual actuation of the FORV'o. A hand pump van procured uith the FORV'o chich can be connected to the PORY to open and clocc A-6 , 0081H122834
the valvo. The manner in which this hand pump'c use vill bc implemented hao not been determined in time for incorpontion into the Baceline Study.
Response
The steam generator PORVs are modeled as failing to the closed position on loss of AC power for their hydraulic pumps. Manual operation of the valves with the hand pump is not included in the Baseline Study. Local recovery actions will be evaluated, if necessary, in subsequent analyses if PORV failures contribute significantly to the Baseline Study results. However, the current results in Section 2 do not seem to be sensitive to this conservative treatment of PORV recovery. The current model includes operation of the steam generator safety valves for intermediate-term heat removal if the PORVs are not available. Comment Number: 14 Report
Reference:
Section 5 - Systems Analysis - Supplemental Purge Isolation Valves Comment: The dcoign of thic cyatem has been changed--tha actuatora on both contaiment intet valveo and both contairrent outlet valvce have been changed to failed cloccd actuators. The type of actuator, pneumatic or electro-hydmutic, haa not been detamined. (In a revielon to HMPa comenta, inctructione ucro provided to quantify the acnaitivity in the reautta to varicua accur.ptione regarding the dcoign and opemtion of the eupplemental purge icolation vaivaa].
Response
Subsequent to receipt of these comments, HL&P requested that the results in Section 2.4 include sensitivities to six different sets of assumptions regarding the design of the contai . ment isolation valves and the fraction of time they are left in the open initial position. Those sensitivities are provided. Comment Number: 15 Report
Reference:
Section 5 - Systems Analysis - AFWS Stop Check Isolation Valves Comment: Tha turbine-driven AFW pump atop-check vatoc io not identical to the valuca provided in tha thrce motor-driven AFW tmino, Uith rcepcot to precouro claac and actuator typc. Theoc valuca are normally clocad to reduca the poccibility of etcan binding in the AFW purma The potential for comon noda failure of att four vatvaa may la.affected in light of theco facto and accordingly may changa the analyaia in A-7 0081H122884
r-the Baceline Study ao it aceumed the came design in all four traine. The valvoo, motore and operatore vill be cpecified to take into consideration changes in procesc conditione due to leakage of upctream check valves. This is conoictent with the findings of AEOD report C-203, " Survey of Valve Operator-related Evente Occurring during 1978,1979 and 1980. "
Response
Consistent with the limited scope and objectives of the Baseline Study generic values of common cause parameters ( 6 y and a ) and values taken frcm other studies on similar plants were used throughout the systems analyses. The following values were used throughout for all components with a few exceptions ( 8 = .05, Y = .5, o = 1). These values are supported by generic data. In a completed Level 1 PRA, we normally use system and plant specific common cause parameters obtained by event-by-event screening of generic data. Our data b'ase for common cause failures of motor-operated valves includes some 42 events out of 400 reactor years of data in which two or more MOVs experienced failure or some degradation in performance due to a common, shared cause. This data has been documented in an EPRI report which will be completed in early 1985 (Reference A-2). A detailed screening of these data for applicability to both types of stop check isolation valves in the AFWS at STPEGS would be performed in a Level 1 PRA. A cursory review of this data, however, indicates that a large fraction of the applicable common cause events for MOVs would not have been prevented by the differences noted in the comment. A large number of the experienced common cause events were associated with defective torque and limit switch components, environmental stresses, and various human errors in test and maintenance. A recovery model was included in the baseline requantification to account for local manual operation of these valves. A residual contribution to failure of all four valves from mechanical binding was retained. A more quantitative resolution of ) this comment would be accomplished in a complete Level 1 PRA. ; Comment Number: 16 Report
Reference:
Section 5 - Systems Analysis - Essential Chilled Water System Comment: The ECIl cyatem providea chitted vator to ecicated room cootern in the NAB and FilB, an uctt ao the CAB flVAC and Control licom air handling unito. For room coolcra ecruiced by the ECil c. ' tem in the MAB and I'llB, one ECll train aupplica 100% heat removal capability, flouaver, for the EAR llVAC and Control licom Envelop flVAC, tuo traina of ECll are needed to remova 100% of the dccign bacia heat load. An analycia paa perfomed to calculate the rate of temperature inercace upon loco of tuo and three ECll t raina. The recutta arc documented in an attachment to the ccmenta trancmittat letter. A-8 0001H122884
l i l
Response
The information provided in Attachment A regarding thermal transient i response of EAB and control room envelope HVAC service areas, thermal capacity of equipment and the beneficial effects of smoke purge operation of the HVAC provides a significantly enhanced basis for assuming the impact of HVAC loss than was possible using information that was made available during the original Baseline Study. Significant revisions to the Baseline Analysis of HVAC were made to account for this enhanced state of knowledge and the effects on the results are evident in Section 2. Despite this enhanced perspective, significant uncertainties remain regarding the variation between local and average room temperature, the effects of humidity condensation in smoke purge, and the quantitative values of equipment fragilities as a function of temperature. These uncertainties should be emphasized in any continuation of the Baseline Study toward the completion of a Level 1 PRA. Comment Number: 17 Report
Reference:
Section 5 - System 5 - Systems Analysis - Containment! Purge Comment: The "targe" containment purge vaivaa vitt always be shut during plant operation at power by administrative procedure. The Supplemental Purgo Valvoo vitt be opened 100% of the time uhite at power to attou operator entry to the RCB on a daity, routino bacia. Ao indicated etacuhere, theco cupplemental purge valvco vitt fait clocad upon loca of AC. [Sco alco comment 14.]
Response
Comment noted and the distinction between these different sets of purge valves and assumptions regarding initial position conform to the comment. [See also response to comment 14.] Comment Number: 18 Report Reference _: Section 5 - System Analysis - Auxiliary Foodwater Storage Tank (AFST) Comment: The AFST, formerty termed the CST, ic currounded by a reinforced conarcto cylindrical atructura, covered by a reinforced conanota roof. .in addition, AFST in derigned for att credible loading combinatione, including dead loado, tive loado, earthquaha loado, normt vind loado and tornado loada (reference Bechtel Dacign Critoria 4S100!!Q1010), and in thereforo not concidered to ha "cubject i to failure". l A-9 008111122834
4 i l
Response
Comment noted. Our interpretation of the comment is that the design 1 criteria for the AFST are very stringent and that the likelihood of failure within these design bases is very low. Within the PRA 4 framework, the potential for failure modes such as clogged vents, debris clogging suction, and structural failures from beyond design basis conditions is normally considered for safety grade tanks such j as the AFST. The Baseline Study results support the view that
- failures of this tank are not a significant risk contributor.
Comment Number: 19 Report
Reference:
Systems Analysis - RHR i , Coment: The Jockey Pumpo in the RHR Syctem are being deleted. ECPH3 ceat otandpipcc vitt be uccd ao head-tanka to keep the EBR Rx'c futt of oater.
Response
) This design change is noted and the Baseline Study Risk model conforms to the comment.
Comment Number: 20 Report
Reference:
Section 6 - Survey of External Events Comment: The data uced in the external econta analyclo chould be conciatent }, Uith the STPEGS ESAR, e.g., crach ratcc, number of flighto, ceicmic acceteration, turbine micottee, valuco for dcatructivo overopeed and decign overopced, atrikea per year, etc. Independent development and ductification of parameters in to be avoided unteco PL&G hae cufficient data a information to juotify otherolce. R_esponse: i
- A very limited scope survey of external events was included in the i Baseline Study. To conserve resources, maximum possibio use was mado of the work done for the S$PSA, especially in regard to the bounding analysis of turbine missiles, aircraft crash and wind-driven missiles. The parameter valves taken from the SSPSA were used in i lieu of spending additional HL&P resourcos to review and verify the references cited in the FSAR. Since in all cases the parameter i' values used were conservative in relation to the FSAR values, this approach was deemed suitable for bounding purposos. In no cases were
- independent valvos generated without such a justification. All the
- affected events were found to be negligiblo risk contributors.
2 A-10 0081H122884
A.1 REFERENCES A-1. Pickard, Lowe and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hupshire and Yankee Atomic Electric Company, PLG-0300, December 1983. - A-2. Fleming, K. N., et al., "Clar'sification and Anal sis of Reactor Operating Experience Involving Dep'endent Events,y' prepared for the Electric Power Research Institute, Research Project 2169-4, to be published in 1955. - i
/
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e o 6 ( i e i A-11 0081H122984
i APPENDIX B RESOLUTION OF HL&P COMMENTS ON STP PROBABILISTIC SAFETY ASSESSMENT BASELINE STUDY READING DRAFT DATED DECEMBER 1984 l Comment Number: I hL&P suggeste that the value to be derived by undertaking more detailed evaluatione utitiaing PRA should be discussed in the Management Plan mther than in the "resulte" section. j Response: 1
. Except as required to clarify the limitations of the Baseline Study risk model, discussion of the value to be derived by undertaking more detailed evaluations utilizing PRA have been removed from the l Baseline Study report. See Appendix A, coment number 10. ! Comment Number 2:
1 hLaP suggeste a clear definition be included in Section 1.1 of the basic differences between the various levele of PBA analysee. 4
Response
Section 1.1 has been revised to clearly define the basic differences i between tL 3 various levels of PRA analysis. Comment Number 3: In Section 6.1 ond in Section 6.4.5, the effects of externat i flocaing are described focusing on the potential failures of the cooling reservoir embankment. A careful consideration of the STP embankment and ability to control cater level in the impoundment has been undertaken in substantial detait such as to make the use of generic statistical dam failure data inappropriate relative to STP. , The cooling reservoir cater levet is positively controtted and is not aubject to uncontrotted increases in cater levet auch as a dam collecting water from a cater ehed. Zhe monitoring of embankment conditione and hydrostatic pressure ateo yleide information that provides a basis to decrease embankment vater tevet prior to danger of failure. Given the ability to positively control the impoundment vater level, the abcence of an uncontratted fitt mechanism for the impounament, and the current program for the reduction of hydrostatic praecure through the use of relief velic at actected locatione along the embankment, it seemo inappropriate to apply hictoric dam failure rates to STP. Aaditionally, the STP impounament does not have a connection to diceimilar materlate; e.g., the side of a hitt so that historical aata that recuttea in failura associated with this interface le similarly inappropriate. Thue, to use historicat dam i 1: B-1 0093H052085
l failure data at the S2P site vould require that those failurce that i resultea from conditione chich cannot exist at STP be removed from the data base. We believe that information that recognizes the differences between STP and generic statistical dam failure probability should be recognized in the Baseline report. Attached is information recently published in the FSAR [ Amendment 43] that bears on embankment failure conditions that ue believe you should revieu in connection uith our comment. If you have any questione concerning this issue, ve vould be pleased to have our technical people discues this matter with your etaff.
Response
Section 6.4.5 has been revised to include the new information supplied by HL&P. We concur that the frequency of failure of the STP embankment should be lower than the generic dam failure frequency. Comment Number 4: The report should identify that the STP design information included in the report is derived from design data available during the period ' April to July 1984. Changee chich have been made in the normal cource of plant design since that time have not been explicitly analyzea or incorporatea in the Baseline resulte uith specific exceptione, such as heatup calculations and supplemental containment ' purge valves, that are explicitly noted in the appendixes to the ' report.
Response
Section 1.1 has been revised to identify sources of design data. _ Comment Number Sa: The discussion of HVAC cuccess criteria on Page 2-16 should be clarifica. The reference to "emoke purge" and its relationship to the conditione "belov 1170 at 24 hours" appear to be inconsistent vith the heatup calculations which indicate that 1 ECH train vould similarly comply, thue apparently conflicting with the success criteria enumerated on Page 2-6. In order to avoid this apparent contradiction with the example case HVAC success criteria, the distinction neeas to be made that in the emake purge moae, the peak temperature cae 1130 at 9 hours, chile the 1 ECH moae reeutted in 1170 at 24 hours ana increasing. Understanding of this section vould be improved by a discussion of the scenarios in which electric power vould be maintained utilizing 1 ECH train. An indication of some of these are providea in item 2 on Page 2-6. Response: i The error in referencing the heatup calculations has been corrected. The section describing the alternative success criterion (assumption l sets) has been rewritten to aid the readers' understanding. Each possible alternative has been assigned a probability of being the true case. B-2 , 0093H052085
Comment Number Sb: Genarat Section 2 - Please clarify if the reference to "EAB hVAC System" refers to both the "EAB Main Area HVAC System" and the
" Control Room hVAC System. "
Response
In Section 2 "EAB HVAC System" refers to the "EAB Main Area HVAC System." Our primary concern has been with loss of AC power to needed equipment due to overheating of circuit breakers (especially solid state protection devices) in the EAB. While loss of control room HVAC could cause problems, the operators would be directly aware of increasing temperatures and could shif t plant control to the auxiliary shutdown panels. Comment Number Sc: The reference to " Ton qi Chillers" on pages 2-6 and 2-17 should be reuritten as " Tone of Chiller Capacity Required."
Response
The two references have been rewritten. Comment Number 5d: Page 2-7, Section 2.3 2 - in the area marked with bullete, the final bullet refers to "notor-operated supplementary purge valves." Since the supplementary purge steam valve aesign has been changed, and the outboard valve is now an air-operated fail-closed valve, the statement should be clarified to refer to the " inboard" NOV.
Response
The section has been revised to indicate that the MOVs are inside containment. Comment Number Se: Page 2-9, Paragraph 2 - Given the failures indicated, there is no need to put the HVAC into the "emoke purge" mode. Please clarify. i Response: We concur. The analysis was correct and did not require placing HVAC into the smoke purge mode. The error in the sequence description has been corrected to agree with the existing analysis. Comment Number 5f: Page 2-10, Paragraph 3 - The discussion included of accident Sequences 7 and 8 in relation to Sequence 1 does not appear to agree uith the scenarios describea in Table 2-3. It appears that accident B-3 0093H052085
ecquencee 7 and 8 may be similar to accident sequence 1 in effdct, but not in sequence of evente. All the accident sequence descriptions should be reviewed for clarity.
Response
Accident sequences 7 and 8 are similar to sequence 2, not sequence 1. The text has been revised. Comment Number Sg: Page 2-10, Accident Sequence 7 describes scenarios where failure pf EAB HVAC fans may result in subsequent AC power losses. The Baseline study should state that the criteria used in EAB hVAC analysis is conservative with respect to the number of fan failures required to disable an EAB HVAC train and to delineate the assumptions and design considerations used to quantify EAB HVAC fan failure.
Response
We do not concur with this comment. First, no single criterion is used in EA3 HVAC analysis. Section 2.3.2 describes a group of 14 discrete assumption sets used in the Baseline Study. Some are conservative; some are very optimistic. All are weighted by the probability that they represent the true success criteria for the modeled scenarios. Note that we even allow some chance that the plant can survive with no HVAC operable. It is sometimes suggested that average success criteria should be used for probabilistic studies. We do not agree. Consider a case in which some unlikely, but possible, condition leads to adverse effects while average conditions do not. The truth would be that the adverse condition is unlikely, but possible. Probabilistic studies should determine how likely such conditions are, rather than mask them via
" average" or "best estimate" analysis. Conservatism is not desired, but a true picture of consequences and their probabilities.
One specific example of relevance for the STP HVAC system would be an assumption that a single supply fan would be sufficient if only one bus is energized. While it may very well succeed, it also might not. With no exhaust fan running, the supply fan could possibly overheat and trip. Such behavior has been observed elsewhere. Also, dampers may trip under this arrangement. While we have not attempted to define all combinations of specific HVAC operating modes, heat loads, equipment fragilities, and external conditions, we feel the 14 discrete assumption sets reasonably span the possibilities. The assigned probabilities (weights) represent a realistic, rather thaa a conservative, assessiaent of criteria for HVAC. B-4 0093H052085
Comment Number Sh: Page 2-15, Section 2.5.2, let Paragraph, last sentence - Zhe text of thic sentence is not clear.
Response
The text has been clarified. Comment Number 51: Page 2-17, the E2 rou does not add to 1 0.
Response
Due to a change in display format, the round-off error is no longer visible in this table, but it is in two others that have been added. Comment Number Sj: Page 2-23, Table 2-3, Erequency Rank 9 - The Sump Recirculation valves can be manually repositioned. Clarify uhether operator action vae included in the recovery model and its effect, if any, on current results.
Response
The scenario in question involves failure of the " sump recirculation path." In the model, that path fails if either all three sump recirculation valves fail to open or any one of the three RWST supply line check valves fails to close. We agree that the recirculation valves can be repositioned and that, given the remaining RWST inventory at switchover, sufficient time exists for such manual recovery. However, failure of the recirculation path is dominated by the second failure mode (failure of one-of-three check valves to close), so the indicated operator action would give no noticible improvement for this scenario. Failure of a check valve to close provides a direct path from containment to the RWST to atmosphere. Furthermore, the containment may be pressurized and the path to atmosphere involves the recirculation (safety injection) pump suction line. Flashing in that line and the RWST could cause failure of the recirculation pumps. In fact, in the Midland PRA, failure under this condition was assumed, based on information supplied by the architect / engineer. No alternative pump failure assumption or recovery by operator action (closing the associated MOV) was modeled for the following reasons: o If failure of the check valve to close will cause pump failure, little indication of the problem exists, and the time for operator response is short. The emergency procedures will probably call for manual closing of the three RWST isolation MOVs, but the time available to protect the pumps is not clear. B-5 0093H052085
o The number of additional assumption sets could be quite large. ! Damage could occur immediately, within several minutes, or over longer periods. Containment pressure can vary depending on break size and operability of containment safeguards. These possibilities multiply the existing number of assumption sets. o The scenario gives less than 1% contribution to core melt frequency. Thus, even if its frequency were reduced, it would have little effect on current results. Comment Number Sk: Page 2-26, 2-27, and 2-28, Figures 2-1, 2-2, and 2 The axie for frequencice presented in these figures should be changed to reflect powere of 10, not 1.0.
Response
The figures have been corrected. Comment Number 51: 1he notes regarding RHR operation on pages 4-64 and 5-7 should be reuritten to indicate that the environment in question is "high temperature and high humidity."
Response
The notes have been rewritten. Comment Number Sm: Pages 5-18 and 5-19, Figure 5 This Figure is not legible and should be replaced. PLG should take necessary steps to assure the legibility of all tables and figures. Foldoute should be used if necessary.
Response
The new figure supplied by HL&P has been used as a foldout. It is more legible than the earlier one. Comment Number Sn: Page 6-7, beation 6.3.2 - The third sentence pf the second paragraph shoula be revised to indicate " safety class componente (e.g., CCW and charging pumps, etc.) . "
Response
The sentence has been revised. B-6 0093H052085
Comment Number 50: Page 6-7, Section 6.3.3 - In the last sentence of the first paragraph, change " main exhaust fans" to "EHB exhaust and booster fane."
Response
The sentence has been revised. Comment Number Sp: Page 6-8, Section 6.3.5, last sentence - It is HL&P's position that fire in the Diesel Euel Storage Tank is not a ma,jor risk contributor and further analysis is not necessary.
Response
We agree that this fire is not a major risk contributor and have not called for further analysis. Comment Number Sq: The hazardous chemical analysis described in Section 6.4.4 does not mention any chemicale stored on site (Reference Table 2.2-5 in the ESAR). Please clarify whether consideration has been given in the report to the storage of hazardous chemicate on site.
Response
The section has been revised to include a discussion of chemicals stored on site. ) B-7 0093H052085
APPENDIX C RESOLUTION OF HL&P COMMENTS ON STP DRAFT REPORT DATED FEBRUARY 1985 LETTER COMMENTS Comment:
...in keeping with your terminology utilized on page 7 pf this draft, and to more accurately describe the state of completion of the work which this report documents, it is suggested that this report be referred to as a Preliminary Scoping Study, a preliminary analysis of plant safety using probabilistic methoas. This terminology should be employed throughout the report. Thus terminology which refore to the STP Preliminary Scoping Study as a probabilistic risk assessment, in the sense pf the levels defined in the PRA Procedures Guide (huhEG/CR-2300), or a baseline study, should be changed or otherwise omittea...
Response
We agree and the report has been so revised. Although the suggested title is not as informative as we might like, it is important to ensure that the current, limited-scope study is not confused with the more familiar PRA studies. Comment: To place this atuay in a perspective to HL&P's overall activities related to the STP Rick Noael Development Program, it is requested that the attached "hL&P Perspective" be inserted near the front pf the report and included in the Table pf Contente, clearly identified as an HL&P document. With this perspective this report's role as a Preliminary Scoping Study could be clarified.
Response
The "HL&P Perspective" is now included in the report's front matter. NUMBERED COMMENTS Comment Number 1: heferences that coula be interpreted to imply that the HL&P program is not going to be continuea, such as that in the first full paragraph on page 1-3, should be changed throughout the report.
Response
All references we could identify that were subject to such an interpretation have been changed. C-1 0110H052185
l Comment Number 2: In the sixth line of the last paragraph on page 2-2, the word " full" should be corrected to " fuel".
Response
The correction has been made. Comment Number 3: On page 2-10 in the continuation of the table at the top of the page, under F3, it appears that the last expression in parentheses should be corrected to read " Probability of success is 1.0 if success criteria are not met."
Response
The error has been corrected as suggested. Comment Number 4: In the tables on page 2-11, since the probabilities listed do not add to 1.0 due to round-off error, these tables should be footnoted to that effect.
Response
An appropriate footnote has been added. Comment Number 5: The first column of the bottom table on page 4-60 should be correctea to be entitled "LT II Tree Top Events."
Response
Both tables on page 4-60 were mislabeled. They have been corrected to read "LT-1" and "LT-2" as appropriate. C-2 0110H052185
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