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NUREG/CR-4551, Vol. 2, Rev. 1, Part 1, "Evaluation of Severe Accident Risks: Quantification of ... PDF

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8524 J. A. Wackerly 1I11111 11111 11111 IIIII Ill IIIIII IOI II I IlllHillllllIlll IIIII loll 11111lI1l l111 NUREG/CR-455 1 8232-21 /071149 SAND86-1309 1lI11l11lllll1Il ll1Hlllll 1l 1l1111111111 1l1/111 1111 1111llllll1 IllI ll Vol. 2, Rev. 1, Part 1 - 00000001 Evaluation of Severe Accident Risks: Quantification of Major Input Parameters Expert Opinion Elicitation on In-Vessel Issues Prepared by I'. 1'. Harper, R. J. Breeding, T. D. Brown, J. J. Gregory, A. C. Payne, E. D. Gorham, C. N. Amos Sandia National Laboratories Operated by Sandia Corporation Prepared for U.S. Nuclear Regulatory Commission AVAIUBILITY NOTICE Availabilrty of Reference Materials Cited in NRC Publicabons Most documents clted In NRC publicatlons wlll be avallable from one of the followlng sources: 1. The NRC Public Document Room, 2120 L Street. NW. Lower Level, Washington, DC 20555 2. The Superlntendent of Documents, U.S. Government Prlntlng Office. P.O. Box 37082, Washington. DC 20013-7082 3. The Natlonal Technlcal Information Servlce, Sprlngfield, VA 22161 Although the listing that follows represents the majority of documents clted In NRC publications. It Is not Intended to be exhaustive. Referenced documents avallable for lnspectlon and copying for a fee from the NRC Publlc Document Room include NRC correspondence and Internal NRC memoranda; NRC Office of lnspectlon and Enforcement bulletins, clrculars. lnformatlon notices, lnspectlon and lnvestlgatlon notices: Llcensee Event Reports; ven- dor reports and correspondence: Commlsslon papers; and applicant and licensee documents and corre- spondence. The following documents In the NUREG series are avallable for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings. and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulatlons In the Code ol Federal Regulations. and Nuclear Regulatory Commlssion lssuances. Documents avallable from the National Technlcal Information Servlce Include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commls- slon. forerunner agency to the Nuclear Regulatory Commlsslon. Documents available from public and speclal technlcal llbrarles include all open literature Items, such as books, journal and perlodlcal articles, and transactlons. federal Register notlces. federal and state leglsla- tlon. and congresslonal reports can usually be obtalned from these ilbraries. Documents such as theses, dissertations, forelgn reports and translations, and non-NRC conference pro- ceedings are avallable for purchase from the organlzatlon sponsorlng the publicatlon clted. Slngle coples of NRC draft reports are avallable free, to the extent of supply, upon wrltten request to the Office of Information Resources Management. Dlstributlon Sectlon, U .S . Nuclear Regulatory Commlsslon. washington. DC 20555. Coples of Industry codes and standards used in a substantlve manner In the NRC regulatory process are malntalned at the NRC Llbrary. 7920 Norfolk Avenue, Bethesda. Maryland, and are avallable there for refer- ence use by the public. Codes and standards are usually copyrighted and may be purchased from the orlglnatlng organlzatlon or. If they are American Natlonal Standards, from the American National Standards instltute. 1430 Broadway. New York, NY 10018. I D ISCLAl M ER NOTICE This report was prepared as an amunt of work sponsored by an agency of the United Stales Government. Neitherthe United Slates Government nor any agency thereof, or any of their employees, makes any warranty, expresed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of such use,o f any information, apparatus, product OT process disclosed in this report, or represents that its us by such third party would not infringe privately owned rights. NUREGICR-4551 SAND86-1309 Vol. 2, Rev. 1, Part 1 Evaluation of Severe Accident Risks: Quantification of Maj or Input Parameters Expert Opinion Elicitation on In-Vessel Issues Manuscript Completed: November 1990 Date Published: December 1990 Prepared by F. T. Harper, R. J. Breeding, T. D. Brown, J. J. Gregory, A. C. Payne, E. D. Gorham, C. N. Amos' Sandia National Laboratories Albuquerque, NM 87185 Prepared for Division of Systems Research Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A1853 'Science Applications International Corporation, Albuquerque, NM ABSTRACT This report records part of the vast amount of information received during the expert judgment elicitation process that took place in support of the NUREG-1150 effort sponsored by the U.S. Nuclear Regulatory Commission. The results of the In-Vessel Expert Panel are presented in this part of Volume 2 of NUREG/CR-4551. The In-Vessel Panel considered six issues: 1. Temperature-Induced pressurized water reactor (PWR) Hot Leg or Surge Line Failure Before Vessel Breach; 2. Temperature-Induced Steam Generator Tube Rupture (SGTR) Before Vessel Breach ; 3. Boiling water reactor (BWR) In-Vessel Hydrogen Production; 4. BWR Bottom Head Failure; 5. PWR In-Vessel Hydrogen Generation; 6. PWR Bottom Head Failure. The report begins with a brief discussion of the methods used to elicit the information from the experts. The information for each issue is then presented in five sections: (1) a brief definition of the issue, (2) a brief summary of the technical rationale supporting the distributions developed by each of the experts, (3) a brief description of the operations that the project staff performed on the raw elicitation results in order to aggregate the distributions, (4) the aggregated distributions, and (5) the individual expert elicitation summaries. The individual expert elicitation summaries were written soon after the elicitation and were sent to the experts for review. They represent the raw results as received directly from the experts. iii/iv CONTENTS . .................................................. 1 INTRODUCTION 1.1 . ............................................ 2 EXPERT CREDENTIALS 2.1 . ................................................... 3 METHODOLOGY 3.1 ............................................ 3.1 Introduction 3.1 ......................... 3.2 Steps to Elicit Expert Judgment 3.2 ..................................... 3.3 Selection of Issues 3.2 .................................... 3.4 Selection of Experts 3.7 .................................... 3.5 Elicitation Training 3.7 .................................. 3.6 Presentation of Issues 3.10 .................. 3.7 Preparation and Discussion of Analyses 3.10 ............................................. 3.8 Elicitation 3.11 ................ 3.9 Recomposition and Aggregation of Results 3.12 .................................................. 3.10 Review 3.13 ........................................... 3.11 Documentation 3.13 . .......................................... 4 ELICITATION MEETINGS 4.1 . 5 RESULTS OF THE ELICITATION ON EACH IN-VESSEL ISSUE .......................................................5. 1.1 . 5.1 Issue 1 Temperature-Induced PWR Hot Leg or Surge ............. Line Failure Before Vessel Breach 5.1-1 . 5.2 Issue 2 Temperature-Induced Steam Generator Tube ............ Rupture (SGTR) Before Vessel Breach 5.2-1 . ............... 5.3 Issue 3 BWR In-Vessel Hydrogen Production 5.3-1 . ......................... 5.4 Issue 4 BWR Bottom Head Failure 5.4-1 . ............... 5.5 Issue 5 PWR In-Vessel Hydrogen Generation 5.5-1 . ......................... 5.6 Issue 6 PWR Bottom Head Failure 5.6-1 ........................................................ APPENDIX A A-1 ........................................................ APPENDIX B B-1 ........................................................ APPENDIX C C-1 V ............. ................ -.__....-. ......... ............ .__1.--... 1. FIGURES . ....................... 1 Back-End Documentation for NUREG-1150 xiv Issue 1 .......... 1-1 Expert A: Case 1: Induced Hot Leg Failure in PWRs 5.1-6 .......... 1-2 Expert C: Case 2: Induced Hot Leg Failure in PWRs 5.1-6 ................................ A-1 Induced Hot Leg LOCA. Case 1 5.1 .14 ................................ A-2 Induced Hot Leg LOCA, Case 2 5.1 .15 ............................... C-1 Expert C's Decomposition Tree 5.1-26 Issue 2 ................................ 2-1 PWR Temperature-Induced SGTR 5.2-3 A-1 Average Wall Temperature Versus Rupture Time ................................... for Steam Generator Tube 5.2-8 A-2 Average Wall Temperature Versus Rupture Time ........................................... for Hot Leg Pipe 5.2-8 A-3 Temperatures Through Hot Leg Wall at Nozzle.. ......................................... Hot Leg Connection 5.2-9 B-1 Average Wall Temperature Versus Rupture ................ Time for Steam Generator Tube (Inconel 600) 5.2-15 Issue 3 ............................... 3-1 Case la: Before Vessel Breach 5.3-7 ............................... 3-2 Case lb: Before Vessel Breach 5.3-7 ............................... 3-3 Case 2a: Before Vessel Breach 5.3-8 ............................... 3-4 Case 2b: Before Vessel Breach 5.3-8 ............................... 3-5 Case 3a: Before Vessel Breach 5.3-9 ............................... 3-6 Case 3b: Before Vessel Breach 5.3-9 vi FIGURES (Continued) ................ A-1 Expert A's Decomposition for Cases la and lb 5.3-14 .................. A-2 Expert A's Assessments for Cases la and lb 5.3-16 A-3 Expert A's Assessments for Cases 2a and 2b .................... 3-17 A-4 Expert A's Assessments for Cases 3a and 3b .................... 3.18 . A-5 Calculated Cumulative Probability Distribution Functions For Total Hydrogen Production Through All Four States ....................................5. 3.19 B-1 Summary Results of Expert B .................................5. 3.26 ................... B-2 Timing of Hydrogen Production for Case la 5.3-26 ................... B-3 Timing of Hydrogen Production for Case lb 5.3-27 ................... B-4 Timing of Hydrogen Production for Case 2a 5.3-27 ................... B-5 Timing of Hydrogen Production for Case 2b 5.3-28 ................... B-6 Timing of Hydrogen Production for Case 3a 5.3-28 ..... C-1 Case 2a Assessments of Hydrogen Production for Expert C 5.3 -33 ..... C-2 Case la Assessments of Hydrogen Production for Expert C 5. 3.35 ...... C-3 Case 3 Assessments of Hydrogen Production for Expert C 5. 3.35 C-4 Timing of Hydrogen Production Using Judgments of Expert C for Case la ....................................5. 3.36 C-5 Timing of Hydrogen Production Using Judgments of Expert C ................................................5. 3.36 D-1 Expert D's Hydrogen Production Estimate .................................. in Two Stages for Case la 5.3-40 D-2 Timing of Hydrogen Production Using Judgments of Expert D ................................................5. 3-42 ............. D-3 Assessment for Hydrogen Production for Expert D 5.3 -44 Issue 4 A-1 Expert A's Decomposition Tree for BWR ........................................ Bottom Head Failure 5.4-7 vii ~-~~~ ... II_..__- ......... ............ ... -. . 1-. .. -.. ..,. ... ---. ........ FIGURES (Continued) Issue 5 .................. 5-1 Expert A: Percentage oE Oxidized Zirconium 5.5-14 .................. 5-2 Expert B: Percentage of Oxidized Zirconium 5.5-14 .................. 5-3 Expert C: Percentage of Oxidized Zirconium 5.5-15 .................. 5-4 Expert D: Percentage of Oxidized Zirconium 5.5-15 ...................... 5-5 Expert E: Amount of Hydrogen Generated 5.5-16 ............................. 5-6 Aggregate of Oxidized Zirconium 5.5-16 5-7 Case la: RCS Percentage of Zirconium Oxidized .......................................... when at 2500 psia 5.5-17 5-8 Case lb: RCS Percentage of Zirconium Oxidized .......................................... when at 2500 psia 5.5-17 5-9 Case IC: RCS Percentage of Zirconium Oxidized .......................................... when at 2500 psia 5.5-18 5-10 Case 2a: RCS Percentage of Zirconium Oxidized .................................. when at 1000 to 1500 psia 5.5-18 5-11 Case 2b: RCS Percentage of Zirconium Oxidized .................................. when at 1000 to 1500 psia 5.5-19 5-12 Case 2c: RCS Percentage of Zirconium Oxidized .................................. when at 1000 to 1500 psia 5.5-19 5-13 Case 3a: RCS Percentage of Zirconium Oxidized .................................... when at 150 to 500 psia 5.5-20 5-14 Case 3b: RCS Percentage of Zirconium Oxidized .................................... when at 150 to 500 psia 5.5-20 5-15 Case 4: RCS Percentage of Zirconium Oxidized ..................................... when at 40 to 200 psia 5.5-21 5-16 Case 5: RCS Percentage of Zirconium Oxidized .................................. when at 1000 to 1500 psia 5.5-21 ................................................... A-1 Case 1 5.5-26 ................................................... A-2 Case 2 5.5-26 ................................................... A-3 Case 3 5.5-27 .................................... D-1 Expert D's Decomposition 5.5-46 viii

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8524 J. A. Wackerly. I111111 11111 11111 . The report begins with a brief discussion of the methods used to elicit the information Assessment for Five U. S. Nuclear Power Plants," NUREG-1150,Vol.l,. Office of Systems Analysis Division at SNL in Livermore, CA, and the supervisor of the Reactor
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