Iowa State University Capstones, Theses and Retrospective Theses and Dissertations Dissertations 1983 Quantitative assessment of human contribution to risk in nuclear power plants Abdallah Ahmad Ezzedin Iowa State University Follow this and additional works at: https://lib.dr.iastate.edu/rtd Part of the Nuclear Engineering Commons Recommended Citation Ezzedin, Abdallah Ahmad, "Quantitative assessment of human contribution to risk in nuclear power plants " (1983). Retrospective Theses and Dissertations. 7669. https://lib.dr.iastate.edu/rtd/7669 This Dissertation is brought to you for free and open access by the Iowa State University Capstones, Theses and Dissertations at Iowa State University Digital Repository. It has been accepted for inclusion in Retrospective Theses and Dissertations by an authorized administrator of Iowa State University Digital Repository. For more information, please contact [email protected]. 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In all cases we have filmed the best available copy. Universi^ Micrcxilms International 300 N. 2EEB RD., ANN ARBOR, Ml 48106 8316315 Ezzedin, Abdallah Ahmad QUANTTTATIVE ASSESSMENT OF HUMAN CONTRIBUTION TO RISK IN NUCLEAR POWER PLANTS Iowa State University PH.D. 1983 University Microfilms Intsmâtioriâ! m N. Zeeb Road. Ann Arbor. MI 48106 PLEASE NOTE: In all cases this material has been filmed in the best possible way from the available copy. Problems encountered with this document have been identified here with a check mark V . 1. Glossy photographs or pages 2. Colored illustrations, paper or print 3. Photographs with dark background 4. Illustrations are poor copy 5. Pages with black marks, not original copy 6. Print shows through as there is text on both sides of page 7. Indistinct, broken or small print on several pages 8. Print exceeds margin requirements 9. Tightly bound copy with print lost in spine 10. Computer printout pages with indistinct print ^ 11. Page(s) lacking when material received, and not available from school or author. 12. Page(s) seem to be missing in numbering only as text follows. 13. Two pages numbered . Text follows. 14. Curling and wrinkled pages 15. Other University Microfilms international Quantitative assessment of human contribution to risk in nuclear power plants by Abdallah Ahmad Ezzedin A Dissertation Submitted to the Graduate Faculty in Partial Fulfillment of the Requirements for the Degree of DOCTOR OF PHILOSOPHY Major; Nuclear Engineering Approved; Signature was redacted for privacy. In Charge of Major Work Signature was redacted for privacy. For the Major Dapartm^t Signature was redacted for privacy. For the Graduate College Iowa State University Ames, Iowa 1983 ii TABLE OF CONTENTS Page LIST OF ABBREVIATIONS USED vi 1. INTRODUCTION 1 2. LITERATURE REVIEW 5 3. PERFORMANCE SHAPING FACTORS 15 3.1. External PSFs 15 3.1.1. Situational characteristics (SCs) 15 3.1.2. Task and equipment characteristics 22 3.1.3. Job and task instruction 29 3.2. Stressors 30 3.2.1. Psychological stress 31 3.2.2. Physiological stress 33 3.3. Internal PSFs 35 3.4. PSFs and its Effect on Human Reliability 38 4. SAFETY ASPECTS OF NUCLEAR POWER PLANTS (PWR) 41 4.1. Introduction 41 4.2. PWR System Description 42 4.3. The Nature of Nuclear Power Plant Accidents 46 4.3.1. Loss of coolant accidents 48 4.4. Reactor Transients 56 5. S2C ACCIDENT SEQUENCE AND ITS HUMAN CONTRIBUTION TO RISK IN NUCLEAR POWER PLANTS 58 5.1. Introduction 58 5.2. SgC Accident Sequence Description 58 5.3. SgC Accident Sequence Systems Description 60 5.3.1. CSIS description and function 60 5.3.2. CSRS description and function 63 5.3.3. CHRS description and function 65 5.3.4. SHAS description and function 67 iii Page 6. ESTIMATION OF UNAVAILABILITY OF THOSE SYSTEMS INVOLVED IN SgC ACCIDENT SEQUENCE 70 6.1. Introduction 70 6.2. Method of Calculation 70 6.2.1. Distribution of human performance 70 6.2.2. Availability theory 72 6.2.3. Unavailability estimations 77 6.2.4. Discussion of the results 84 7. OPERATOR ERROR RATES ESTIMATION 87 7.1. Introduction 87 7.2. Operator Error Rate Estimates 87 7.3. Valve Mispositioning of Those Systems and Subsystems Involved in S2C Accident Sequence 92 7.3.1. Human error rates estimated for CSIS 92 7.3.2. Human error rates estimated for CSRS 99 7.3.3. Human error rates estimated for SHAS 102 7.3.4. Human error rates estimated for CHRS 107 7.4. Comparison with Other Estimates of Human Error Probabilities 111 7.5. Discussion of the Results 118 8. QUANTITATIVE EVALUATION OF SYSTEMS INVOLVED IN S2C ACCIDENT SEQUENCE 122 8.1. Introduction 122 8.2. Fault Tree Analysis and its Applications 123 8.2.1. Fault Tree Analysis (FTA) 123 8.2.2. Event symbols 124 8.2.3. Application of fault trees 128 iv Page 8.3. CSIS Fault Tree "Insufficient Fluid Flow" 133 8.3.1. CSIS fault tree description 133 8.3.2. PREP and KITT codes description 148 8.3.3. The PREP run for the CSIS fault tree 149 8.4. KlTT-1 Results for the CSIS Fault Tree 172 8.4.1. Case 1: WASH-1400 172 8.4.2. Case 2: LERs 178 8.4.3. Case 3: NUREG/CR-1278 182 8.4.4. Case 4: Sensitivity of CSIS relia­ bility to human errors 186 8.4.5. Discussion of the results 192 8.5. CHRS Unavailability Fault Tree 193b 8.5.1. CHRS fault tree description 193b 8.5.2. The PREP run for the CHRS un­ availability fault tree 202 8.5.3. KITT-1 results for CHRS un­ availability fault tree 224 9. SUMMARY AND CONCLUSIONS 236 10. RECOMMENDATION FOR FURTHER WORK 239 11. REFERENCES 241 12: ACKNOWLEDGMENTS 247 13. APPENDIX A: CHARACTERIZATION OF OPERATOR "ERRORS" FROM ANALYZED LICENSEE EVENT REPORTS (LERs) 248 14. APPENDIX B: DERIVED HUMAN ERROR PROBABILITIES AND RELATED PERFORMANCE SHAPING FACTORS 277 15. APPENDIX C: DESCRIPTION OF THE PREP CODES AND ITS INPUT DATA 284 15.1. Input Data Description 287 15.1.1. Input group 1 287 15.1.2. Input group 2 288 15.1.3. Input group 3 288 V Page 16. APPENDIX D; DESCRIPTION OF THE KITT CODES AND ITS INPUT DATA 291 17. APPENDIX E; KITT-1 OUTPUT RESULTS FOR THE CSIS FAULT TREE 300 18. APPENDIX F; KITT-1 OUTPUT RESULTS FOR CHRS FAULT TREE 317 vi LIST OF ABBREVIATIONS USED BWRs Boiling water reactors CHRS Containment heat removal system CI Containment integrity CLCS Containment limiting control system CSIS Containment spray injection system CSRS Containment spray recirculation system ECCS Emergency core coolant system EP Experienced personnel ESFs Engineered safety features ESSs Engineering safeguard systems FTA Fault tree analysis GPM Gallon per minute HE Heat exchanger HEPS Human error probabilities HPIS High pressure injection system HPRS High pressure recirculation system LERs Licensee event reports LOCAs Loss of coolant accidents LPIS Low pressure injection system LPRS Low pressure recirculation system LWRS Light water reactors NPPs Nuclear power plants NSRG Nuclear safety research group PARR Post accident radioactivity removal vii PAHR Post accident heat removal PRAs Probabilistic risk assessments PSFs Performance shaping factors PWRs Pressurized water reactors RCS Reactor coolant system RT Reactor trip RWST Refueling water storage tank SHAS Sodium hydroxide addition system THERP Technique for human error rate prediction TMI Three Mile Island 1 1- INTRODUCTION A high-technology society places a premium on its de­ mand for energy sources. Of the alternative available energy sources at the present time, nuclear power is one of the most efficient and economical sources. However, there is widespread concern (founded or unfounded) over the safety of nuclear power plants. Much of this concern surfaced as a result of the accident in March of 1979 at the Three Mile Island (TMI) generating station near Harrisburg, Pennsylvania. This accident was the result of a complex set of interactions involving design deficiencies, equipment failure and human errors.