ML20155K505
| ML20155K505 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1988 |
| From: | Giitter J, Mckenna T NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-1210, NUREG-1228, NUDOCS 8811010272 | |
| Download: ML20155K505 (174) | |
Text
{{#Wiki_filter:_ _ - . - - _. - - - - NUREG-1228 l Source Term Estimation During Incident Response to Severe 1 Nuclear Power Plant Accidents i l ! , .i -l [ l i U.S. Nuclear Regulatory i Commission ; ! Office for Analysis and Evaluation of Operational Data ! T. J. McKenna, J. G. Giitter !
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Of fice, Post Office Box 37082, Washington, DC 20013 7082
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Referenced documents avaslable for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, cicculars, information notices, inspection and investigation notices; Licensee Evert Reports; vendor reports and correspondance; Com.nission capers; and applicant and licensee documents and correspondence. The following documents in the NUREG series ate ava:lable for purchase from the GPO Sales Program: formal NRC stafI and contractor reports, NRC sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information E rvice include NUREG series reports and technical reports prepared by other federal agencies and repons prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and sper.ial technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited. Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Copies of industry codes and standards used in a substantive manner in the NRC regutatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Magland, and are available there for reference use by the public. Codes and standards are usually copyrighted and rnay be purchased from the originating orgonization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018. _ _ _ __ _ J
NUREG-1228 Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents M:nuscript Completed: Cetober 1988 D:ta Published: October 1988 T. J. McKenna, J. G. Glitter Division of Operationel Assessment Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Wcshington, DC 20666 l p* . s, i J i i ___,____y __ - -- ,-- - - -- - - -- '~ - - ' '
ABSTRACT This document presents a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an l event. The various methods of estimating radionuclide release to the environ-ment (source terms) as a result of an accident at a nuclear power reactor are , discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant acoidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estima-tion of accident source terms and their consequences off site is presented. l i l 4 l l i 1 1 t i t
- i l
a NUREG-1228 iii
l l l l l TABLE OF CONTENTS em ABSTRACT ............................................................... iii FOREWORD .......................................... xiii ACKN0WLEDGEMENTS........................................................ ............... ..... xv ACRONYMS AND INITIALISMS.......................................... ..... xvii 1 INTRODUCTION ...................................................... 1-1 1.1 Objective.................................................... 1-1 1.2 Prerequisites ................................. 1-1 1.3 Background ................................................... ............... 1-1
- 1. 4 Bases for Source Term Estimation ............................. 1-3 1.4.1 Source Term Estimates Based on Effluent Monitors ...... 1-4 1.4.2 Source Term Estimation Based on Accidents Analyzed as Part of Licensing ..................................... 1-6 ,
1.4.3 Source Term Estimation Cased on Severe Accident ' Consequence Studies ................................... 1-6 1.4.4 Source Term Estimates Based or. Detailed Analysis of Plant Conditions ...................................... 1-8 1.4.5 3ource Term Estimates Based on Precalculated Assump-tions of Dominant Accident Conditions ................. 1-10 2 RADIONUCLIOES IMPORTANT IN SOURCE TERM ASSESSMENT .................2-1 2.1 Objective.................................................... 2-1 2.2 Fission Products ............................................. 2-1 2.2.1 Source of Fission Products ............................ 2-1 2.2.2 Inventory of Fission Products in the Core . . . . . . . . . . . . . 2-1 2.2.3 Fission Products Important to Offsite Consequences .... 2-1
- 2. 3 Inventory Assumptions To Be Used for Source Term Estimation .. 2-5 3 RADIONUCLIDE TRANSPORT MECHANISMS UNDER NORMA CONDIT IONS . . . . . . . . . . . . . . . . . . ...................... . . . . . . . . . . . . . . . . L3-1 OPE RATING 3.1 Objective.................................................... 3-1 3.2 Barriers to Fiss'an Product Release ..........................
3.3 Radionuclide Trans 3-1 Cladding Gap ..... ............................................ port From the Fuel Matrix Into the Fuel-3-1 3.4 Radionuclide Transport Fr s the Fucl Into the Coola ........ 3-4 3.5 Routine Effluent Releasas .........................nt ........... 3-10 NUREG-1228 v
TABLE OF CONTENTS (Continued) P_ag i t 4 ESTIMATING RADIONUCLIDE TRANSPORT MECHANISMS UNDER SEVERE ACCIDENT CONDITIONS ............................................... 4-1 4.1 0blective..................................................... 4-1 4-1
! 4.2 Maor Considerations .........................................
4.3 Estimating Core Fission Product Inventory .................... 4-2 4.4 Estimating Fission Product Inventory Released From the Core .. 4-3 4.4.1 Normal Fuel Pi n Leakage (600*F) . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4.2 Fuel Cladding Rupture / Gap Release (1300'-2100*F) ...... 4-5 4.4.3 Grain Boundary Release (3000*F),and Melt Release ! l (In-Vessel) (4500'F) .................................. 4-9 ; i 4.4.4 Reactor Vessel Melt-Through (Ex-Vessel Melting) ....... 4-11 l 4.4.5 Form of Release From the Fuel ......................... 4-11 - 4 4.5 Movement of Fission Products From the Core to Atmosphere (Re-duction Factors) ............................................. 4-11 4.5.1 Basic Fission Product Reduction Mecharisms ............ 4-13 i 1 4.5.1.1 Gaseous Release Reduction Mechanisms ......... 4-13 + 4.5.1.1.1 Plateout in the Reactor Coolant i System ............................ 4-13 4.5.1.1.2 Removal of Aerosols and Particu- , lates Suspended in large Volumes... 4-13 ; 4.5.1.1.2.1 Natural Removal i Reduction Factors .... 4-14
< 4.5.1.1.2.2 Spray Removal Reduc- [
tion Factors ......... 4-14 } 4.5.1.1.3 Filters ...................., ... 4-14
) 4.5.1.1.3.1 BWR Standby Gas :
- Treatment System ..... 4-16 l l 4.5.1.1.3.2 PWR Containment Re- i circulating Filter !
System ............... 4-16 !
. 4.5.1.1.3.3 PWR Auxiliary Build- l ing Filter Systems ... 4-17 1
i
- 4. 5.1.1. 4 BWR Suppression Pool .............. 4-17 1 4.5.1.1.5 FWR Ice Condenser ................. 4-18 i
4.5.1.2 Liquid Release Reduction Mechanisms .......... 4-18 i-
) l ; 4.5.2 Basic Release Pathways and Characteristics ............ 4-22
{ i i 3 NUREG-1228 vi i U
TABLE OF CONTENTS (Continued) P.gte i 4.5.2.1 Reactor Coolant S Barrier) ........ystem (Third Fission Product
............................. 4-22 4.5.2.1.1 PWR Reactor Coolant S Breaks / Leaks ........ystem .............. 4-22 4.5.2.1.2 PWR P ~er-Operated Relief Valves .. 4-35 4.5.2.1.3 PWR team Generator Tube Rupture... 4-35 4.5.2.1.4 PWR and BWR Bypass Accidents ...... 4-36 4.5.2.1.5 PWR and BWR Vessel Melt-Through ... 4-37 4.5.2.1.6 PWR and BWR Leakage From Process .
Lines ............................. 4-38 l 4.5.2.1.7 BWR Reactor Coolant System ' Breaks / Leaks ...................... 4-40 4.5.2.1.8 BWR Automatic Depressurization System and/or Safet Relief Valves ............y ................ 4-40 l
~
l 4.5.2.2 Primary Containment Fourth Fission Product , Barrier)...........(.......................... 4-40 ' 4.5.2.2.1 Design Leakage .................... 4-41 4.5.2.2.2 Isolation Failures ................ 4-42 4.5.2.2.3 Catastrophic Containment Failures .......................... 4-42 ' 4.5.2.2.3.1 Hydrogen Detonation / Burns ................ 4-43 4.5.2.2.3.2 Steam Explosions, Direct Heating, and Missile: ............. 4-46 l 4.5.2.2.3.3 Containment Over-l pressurization ....... 4-46 1 4.5.2.2.4 Containment Bypass ................ 4-47 4.5.2.2.5 Controlled Venting of Ccatainment ....................... 4-47 4.5.2.3 O the r Ba r r i e r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-48 4.6 Plant Instrumentation ........................................ 4-49 4.6.1 Radiation Monitors ........................... 4-49 4.6.2 Grab Samples .......................................... ......... 4-53 4.6.3 Key Plant Parameters .................................. 4-54 4.6.3.1 Core Temperature ............................. 4-54 4.6.3.1.1 Pressurized-Water Reactor ......... 4-54 4.6.3.1.2 Boiling-Water Reactor ............. 4-56 NUREG-1228 vii
TABLE OF CONTENTS (Continued) l 1 Paa! 4.6.3.2 Reactor Coolant Temperature and Pressure ..... 4-56 4.6.3.3 PWR Pressurizer Level ........................ 4-56 4.6.3.4 Reactor Vessel Water Level ................... 4-57 4.6.3.4.1 BWR Water Level Indicator ......... 4-57 4.6.3.4.2 PWR Vessel Coolant Level Detectors .,....................... 4-57 4.6.3.5 Engineered Safety Feature and Critical Safety Function Status Indication ................... 4-57 4.6.3.6 Containment Isolation Status ................. 4-58 ! 4.6.3.7 Containment Atmosphere Temperature and : Pressure ..................................... 4-58 5 SOURCE TERM DETERMINATION ......................................... 5-1 1 5.1 Objective.................................................... 5-1 5.2 Introduction ................................................. 5-1 5.3 Sample Source Term Calculation ............................... 5-2 5.3.1 Step 1: Gather / Assess Plant Information .............. 5-2 5.3.2 Step 2: Estimate the Fission Product Inventory Released From the Core ................................ 5-3 J 5.3.3 Step 3: Estimate the Fission Products Available for t Release ............................................... 5-3 ! 5.3.4 Step 4: Estimate 1-Hour Release ...................... 5-4 l 5.4 Use of Event Trees To Estimate Release ....................... 5-6 i 5.5 Comparison With Current Source Ters Results .................. 5-28 5.6 Estimate the Uncertainties ................................... 5-28 6 REFERENCES ........................................................ 6-1 , APPENDIX A PROCEDURE FOR ESTIMATING OFFSITE CONSEQUENCES BASED ON PLANT CONDITIONS r 1 1 List of Figures P,a21 i 1.1 Source ters estimation based on effluent monitors ............... 1-5 l' 1.2 Relationship of plant conditions to WASH-1400, siting source terms and consequences .......................................... 1-7 l 1.3 Sample MENU-TACT flov ........................................... 1-9 , 1.4 Examples of major aerosol / particulate reduction mechanisms ...... 1-11 i 2.1 Relative importance to bone marrow dose of the radioactive j elements found in the ccre of a reactor given a major release ... 2-4 l 2.2 Inventory of Kr-85 in core of a 1000 MWe PWR . . . . . . . . . . . . . . . . . . . . 2-5 NUREG-1228 viii
TABLE OF CONTENTS (Continued) P_agg 3.1 Fission product barriers to release of radioactive material for a PWR ........................................................... 3-2 3.2 Cross section of fuel pellet illustrating cracking . . . . . .. ... . .. . 3-3 l 3.3 Typical PWR fuel assembly (A) and fuel pin (B) .................. 3-6 i 3.4 Typical iodine spike following shutdown ......................... 3-8 l 3.5 Typical cesium spike following shutdown.......................... 3-9 l 4.1 Fission product release rate constants from fuel - smoothed I curves .......................................................... 4-10 4.2 Comparison of removal of airborne aerosols and particulates by natural effects and sprays ...................................... 4-15 4.3 Typical pathways for release of radionuclides to the atmosphere from a steam generator tube rupture ............................. 4-20 4.4 Conditions in radionuclide release pathways following a "U" tube steam generator tube rupture to the environment ............ 4-21 4.5 PWR dry containment simplified release pathways ................. 4-23 4.6 PWR ice condenser containment simplified release pathways ....... 4-24 4.7 BWR Mark I simplified release pathways .......................... 4-25 4.8 BWR Mark II simpli fied rel ease pathways . . . . . . . . . . . . . . . . . . . . . . . . . 4-26 4.9 BWR Mark III simpli fied release pathways . . . . . . . . . . . . . . . . . . . . . . . 4-27 4.10 TMI accident radioisotope release pathways ...................... 4-39 4.11 Comparison of early containment failure probabilities ........... 4-44 4.12 Low-temperature and pressure flammability and detonation limits of hydrogen, air, and steam mixtures ..................... 4-45 4.13 Response (cps) of selected process monitors during the TMI accident ........................................................ 4-51 4.14 Response of selected area radiation monitors during the TMI accident ........................................................ 4-52 4.15 THI-2 accident: map of core exit temperature ('F), 8:00 a.m. - 9:00 a.m., March 28, 1979 ....................................... 4-55 4.16 Relationship among critical safety functions, maintaining fission product barriers, and preventing a release .............. 4-59 5.1 Release pathway for sample calculation case ..................... 5-5 5.2A PWR large, dry or subatmospheric containment release event tree for a gap release from the core ................................. 5-7 5.28 PWR large, dry or subatmospheric containment release event tree for a grain boundary release from the core ...................... 5-8 5.2C PWR large, dry or subatmospheric containment release event tree for a melt release from the core ................................ 5-9 5.3A PWR ice condenser containment release event tree for a gap release from the core ........................................... 5-10 5.3B PWR ice condenser containment release event tree for a grain bounda ry rel ease f rom the core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-11 5.3C PWR ice condenser containment release event tree for a melt release from the core ........................................... 5-12 5.4A A PWR steam generator tube rupture event tree for a release of normal and 100X spike coolant .................................... 5-13 NUREG-1228 ix
TABLE OF CONTENTS (Continued)
.P,ag 5.4B PWR steam generator tube rupture event tree for a release of coolant contaminated with a gap release from the core ........... 5-14 5.4C PWR steam gneerator tube rupture event tree for a release of coolant contaminated with a grain boundary release from the core ............................................................ 5-15 5.40 PWR steam generator tube rupture event tree for a release of coolant contaminated with a melt release from the core .......... 5-16 ,
5.5A BWR containment, dry-well leakage / failure release event tree for a gap release from the core ................................. 5-17 5.5B BWR containment, dry-well leakage / failure release event tree for a grain boundary release from the core ...................... 5-18 5.50 BWR containment, dry-well leakage / failure release event tree for a melt release from the core ................................ 5-19 5.6A BWR containment, wet-well leakage / failure release event tree for a gap release from the core ................................. 5-20 5.6B BWR containment, wet-well leaka0e/ failure release event tree for a grain boundary release f rom the core . . . . . . . . . . . . . . . . . . . . . . 5-21 5.6C BWR containment, wet-well leakage / failure release event tree for a melt release from the core ................................ 5-22 5.7A PWR and BWR containment bypass release event tree for a gap release from the core ........................................... 5-23 5.78 PWR and BWR containment bypass release event tree for a grain boundary release from the core .................................. 5-24 1 5.7C PWR and BWR containment bypass release event tree for a melt release from the core ........................................... 5-25 5.8 Comparison of source term projection results for station '. bl ac kout scenari os at Surry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-29 List of Tables 2.1 Factors that have the greatest effect on radionuclide inventory ....................................................... 2-2 2.2 Fission product inventories (Ci/HWe) ............................ 2-3 3.1 Helting point and boiling point (*F) of selected materials ...... 3-5 3.2 Typical coolant concentrations (pCi/g) .......................... 3-7 3.3 Range of total BWR/PWR airborne effluent release (curies) for 1980 ............................................................ 3-11 3.4 Release rate (Ci/sec) needed to meet benchmarks ................. 3-11 4.1 Core release fraction assumptions (1-hour release) .............. 4-4 4.2 Average LWR fuel, cladding, and coolant operational temperatures .................................................... 4-5 4.3 PWR baseline coolant concentrations for various levels of core damage .......................................................... 4-6 NUREG-1228 x
TABLE OF CONTENTS (Continued) Paa' i 4.4 BWR baseline coolant concentrations for various levels of core i damage .......................................................... 4-7 l 4.5 Summary of particulate / aerosol gaseous reduction mechanisms ..... 4-12 1 4.6 Typical BWR standby gas treatment system suction locations ...... 4-16 , 4.7 Summary of ma;or PWR release pathway reduction mechanisms ....... 4-28 4.8 Summary of ma;or BWR release pathway reduction mechanisms ....... 4-31 4.9 Key to release pathway references on Figures 4.5 through 4.9 .... 4-34 4.10 Assumed containment and steam generator tube rupture escape fraction for 1 hour ............................................. 4-37 4.11 Sample containment design pressures and failure pressures ....... 4-47 5.1 1-mile doses for release of various core and coolant inventories ..................................................... 5-26 5.2 Comparison of event tree doses with CRAC whole-body dose projections for WASH-1400 PWR release categories at 1 mile ...... 5-30 i l NUREG-1228 xi
l FOREWORD l 1 In the event of a severe nuclear power accident, the initial public protective l measures or plant action to mitigate the accident may be revised based on an estimate of possible offsite consequences. This requires a method to relate key accident conditions to the distance to which the U.S. Environmental Protec-tion Agency's Protective Action Guides (PAGs) may be exceeded and the distance to which early health effect or early death dose levels are possible. The U. S. Nuclear Regulatory Commission (NRC) headquarters Protective Measuras Team recognized that the current methods used to estimate radionuclide releast (source term) consequences for severe accidents are inadequate. These methods either require sn accident te fit. a specific sequence already analyzed or to run a detailed computer code for wi.ich there are no commonly agreed-on set of assumptions. Both methods require a p rson with a detailed knowledge of all aspects of source term research, a detailed knowledge of plant conditions, and considerable time. Experience has shown in many cases that the events either do not fit a specific sequence already analyzed or that the analyst does not have sufficient time or information to conduct a detailed analysis. Further-more, there is limited experience with source term assessment in the NRC regions and in licensee or State / local response organizations. Therefore, it ic unlikely that early responders could benefit from these methods. Moreover, the results of a headquarters analysis often would be difficult to explain or reproduce since they are not based on commonly accepted and documented assumptions. The Incident Response Branch was requested to develop a simple method of relat-ing basic plant conditions to the distance to which PAGs may be exceeded and early health effect or early death dose levels may be possible. This document and procedure are the result of this effort. This document is written to be used by personnel without a detailed source term background and explains the basics of source term estimation. The resulting simple method and incorporated in a stand-alone procedure (Appendix A) presented uses a set ofincondition Section 5 trees that relate dominant reactor accident conditions to consequences. These trees present, in a form that can easily be related to reactor plant conditions, the results of calculations performed by the models used in the NRC Oper.stions Center. The goal is to have a common method to quickly bound the possible con-sequences of an accident available to all responders. The method and assumptions presented in this document will be reviewed on an ongoing basis and revised as appropriate to reflect advances in our understanding of severe accident source terms. Please provide your comments to Incident Response Branch. , f fMA R. Lee Spessard, Director Division of Operational Assessment Office for Analysis and Evaluation of Operational Data NUREG-1228 xiii
l l ACKNOWLEDGEMENTS Walter Pasedag and Patrick Easely (formerly of the U.S. Nuclear Regulatory Commission) were responsible for recognizing the need for this work and for providing the basic concepts on which it is based. 1 [
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i 1 r i NUREG-1228 xy
ACRONYMS AND INITIALISMS i ADS automatic depressurization system ANS American Nuclear Society ANSI American National Stendards Institute ARMS area radiation monitoring system ATWS anticipated transient without scram B&W Babcock & Wilcox BWR boiling-water reactor CPWG Containment Performance Working Group (NRC) CRF core release fraction j DCH direct containment heating ECCS emergency core cooling system EF escape fraction E0P emergency operating procedure EPA Environmental Protection Agency ESF engineered safety feature (s) FDA Food and Drug Administration FPI fission product inventory HEDL Hanford Engineering Development Laboratory HEPA high-efficiency particulate air i H00 Headquarters Operations Officer (NRC) HP health physics 1 LOCA loss-of-coolant accident LWR light-water reactor MSIV main steam isolation valve NRC U.S. Nuclear Regulatory Commission OTSG once-through steam generator l PAG Protective Action Guides (U.S. Environmental Protection Agency) PORV power-operated relief valve PWR pressurized-water reactor RCS reactor coolant system RDF reduction factor RG regulatory guide , RHRS residual heat removal system l l l NUREG-1228 xvii i
58GTS standby gas treatment system SG steam generator SGTR steam generator tube rupture SRV safety relief valve SST siting source term TMI Three Mile Island Nuclear Power Station l l
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1 NUREG-1228 xvitt
i l 1 l l 1 INTRODUCTION i 1.1 Objective , t To enable the reader to I specify the role of radionuclide release (source term) assessment in ! l emergency response l illustrate the necessity of recognizing and identifying the great l i uncertainties associated with performing a source ters assessment l 1 ; 2 . understand the utility of the various source term estimation methods for ' l use during an accident 5 1.2 Prerequisites ! l For the reader to understand the concepts in this manual, a basic understanding ! of reactor systems is required along with an understanding of protective ac- I 1 tions related to severe reactor (core damage) accidents. It is highly recom- ; l mended that Volumes 2 and 4 of NUREG-1210. "Severe Reactor Accident Incident :
; Response Training Manual," be studied before reading this document, i i -
i 1.3 Backaround i I j ) A radioactive release (source term) for accidents involving major core damage i that could result in early injuries or deaths to the public near the plant can i always be projected. As discussed in NUREG-1210. Volumes 2 and 4, those pro-jections are based on an assumption of a major early release (containment fail- [ ure). A wide range of these types of accidents has been analyzed. It may be L impossible to produce an analysis during a severe accident that will provide L
- additional insight into possible consequences and appropriate offsite protective !
4 actions. The results of the vast amount of analysis of severe accidents have been incorporated into the current NRC guidance on protective actions for core
! damage accidents. Therefore, initial protective action decisions for core i
damage accidents (general emergency) should never be based on source term or
! dose calculations performed at the time of the accident. These decisions must (
j l be based on predetermined plans based on NRC guidance. In t'act, for core ; damage accidents, the source term calculations required for initial decisions : have already been done, forming the basis for the current guidance. I I i For a severe nuclear power plant accident (core damage), the immediate prede-j termined protective actions are taken based on in plant indicators of core and ( i containment conditions (e.g., core and containment temperatures and pressures). i If core damage is projected or exists, the population near the plant (2-5 i i miles) should be evacuated, and people within a 10-mile radius of the site l
! should be sheltered. This protective action strategy was determined based on ,
considerations of consequence analyses for a wide range of core damage acci-2 dents involving containment failure, bypass, or leakage. These actions would I ) NUREG-1228 1-1 ! 1 1 L-_ . - _ _- _ _ - , - - , . . - _ . . - - - . _ _ _ _ _ _ - . - - - . . - - - _---,____.__---.\
provide adequate immediate protective a:tions for most core damage accidents. However there may be core damage accidents that would warrant taking additional protectIveactionsorthatwouldrequirevariousplantresponseoptions(e.g., venting) to be assessed. These assessments may require promptly assessing acci-dent source terms during the accident. 4 The first step in a dose assessment is to determine the amount of various radio-nuclides that are postulated or estimated to be released to the environment. This characterization of radionuclides that may be released to the environment, in conjunction with release rate and height, is referred to as the "source term." In the past, these dose assessments have been based on estimates of releases provided by plant radiation monitors. However, this approach is inadequate under i severe accident conditions, for the reasons discussed in Volume 2 of NUREG-1210. i i In addition, in a severe accident situation, it would be undesirable to wait until the release occurs before taking the necessary additional protective action because effective protective action requires prompt implementation. Therefore, an attempt should be made to project the magnitude of a release before it occurs. To be useful, these projections should be based on a best-estimate assessment of the source term and not on artificial assumptions intended only for licensing purposes. There has been a tendency in the past to utilize the source term as-sumptions in Regulatory Guides (RGs) 1.3 and 1.4, which indicate that 100% of the noble gases, 25% of the halogens, and 1% of the other fission products are re-leased to the containment. These assumptions should not be used to characterize an actual accident. The regulatory guide assumptions also tend to reinforce the erruneous assumptions that only noble gases and iodine would be released during a severe accident. By their very nature, severe accidents involve conditions that make the predic-
, tion of the source term very difficult: the plant is beyond design conditions, i
instrument readings may be unreliable, accident progression is unpredictable,
; specific in plant conditions may not be known or can change quickly and unex-i pectedly. In addition, even if plant conditions and accident progression were 1 completely understood, the ability to project the source term would be very limited because of the limited understanding of source term physics and chemis-try. The result is that for a severe accident there is little hope of actually 4 predicting the source term; only approximations of source term with large un-
! certainties can be produced. Because of these great uncertainties, there is little purpose being served by performing complex detai.sd assessments that ] consider secondary or little understood effects. Recent studies of the uncer-
- tainties associated with source term estimation indicate that source term oro-
) jections based on accident conditions are only accurate within a factor of 100 or more, even if all of the accident conditions are known (NUREG-0956 and i j NUREG-1150). Furthermore, experience has indicated that the details of acci- ! dent conditions usually are not available until the event is terminated--thus
! a full understanding of the accident conditions will probably not be immediately available to the analyst performing the dose assessment. Thus, any result ob-J tained during an accident will have a possible error of 1 to 3 orders of magni- , tude or more. However, source teru studies do provide important insights into ' what accident conditions dominate the characteristics of a release. This infor-mation allows relative consequence comparisons to be made between different release pathways or conditions. The possible consequences of various accident ;
, sequences can be reked based on consideration of these dominant accident condi-i tions. For example, it is clear that there is considerably less risk from a l
l l NUREG-1228 1-2
I l l l release that has passed through the suppression pool in a boiling-water reactor i l (BWR) as compared with one that has not been filtered by the suppression pool, i I There also should be sufficient information to rank specific accidents relative i to their consequence potential. For example, it is clear that it requires major : damage to the core of a reactor or possibly the spent fuel pool and a fast and , substantially unfiltered pathway to the environment to result in early deaths ! l or injuries off site. Source term assessments conducted during an accident must r
- be based on fast, best-estimate calculations that account for the dominant '
i effects. If a change in assumpt'ons does not result in a change to the source term by at least 1 order of magnitude, it is not worth considering because it will provide no useful information. These assessments are best directed at
- comparisons of the potential offsite consequences from various possible accident i sequences. Because of the great uncertainties associated with specific source I i term estimates, these rankings do not require detailed consideration of offsite i i transport and dose. In most cases, the estimate of offsite consequences should j when discussing the results. Additionally, in !
l beused,andnotdoseestimates}bletoprojectasourcetermwithsufficient most cases, it will not be poss accuracy to estimate an offsite dose that can be reasonably expected to match 2 that resulting from the release. However, the relative risk of the accident t
] anditsprojectedsequencescanbejudged.
l I For the decision-makers to be able to use the source term estimate in their
! decision process, these uncertainties must be understood and their bases must
{ be clear. To support any decision by the decision-makers, the following tasks i j must be performed: i { i 1 (1) Only the dominant plant conditions that influence the potential risk to l
) the offsite population, resulting from the release, must be identified and -
l considered. ) (2) Basedoncurrentandprojectedplantconditions,abestestimateofthe !
' magnitude of any offsite releases and the assumptions on which these !
estimates are based must be provided for the sequences considered. ! (3) The confidence the analyst has in the estimate must be specified. A dis-cussion of the great uncertainties always associated with source term must be included, l (4) The reasonable bounds of the analyst's estimate must be specified. ( 1 (5) The upper and lower bounds of the release must be stated based on possible j i courses of the accident or changes in key assumptions as related to off-site consequences such as doses large enough to result in early health effects (50-100 rem) or to exceed the Environmental Protective Agency's 1 Protective Action Guides (1-5 rem whole body) (EPA-520/1-75-001). 1.4 Bases for Source Term Estimation J 1 There are five bases that may be used to estimate source term (radionuclide { release) from a severe reactor accident. These are (1) effluent monitor readings i NUREG-1228 1-3 i
(2) accident analysis contained in the safety evaluation report (3) various severe accident consequence studies such as the Reactor Safety Study (WASH-1400) l (4) detailed analysis of plant conditions conducted during an accident (5) precalculated estimates that relate dominant accident conditions to potential radionuclide releases (source terms) Each of these bases will be discussed briefly although this document concen-trates on providing a methodology for basis 5 above.
! 1.4.1 Source Term Estimates Based on Effluent Monitors Obviously if a release is out of a monitored pathway, the monitor could provide useful information on the size of the release. However, a monitor does not provide a direct measure of release rate (Ci/sec) for the various isotopes being released. As shown in Figure 1.1 an isotopic release rate estimate is ,
based on measured flow and activity (counts /sec) and assumptions about monitor ! efficiency and isotopic mixture. The nuclear plant operators have developed calibration values Lased on assumed release mixture that relate monitor mea-surements (counts /sec) to a specific isotopic release rate (e.g., 100 Ci/sec Kr-85). , On-line radiation monitors capable of measuring the noble gases released through plant vents were installed at nuclear facilities following the Three Mile Island, Unit 2, accident. However, on-line monitors for iodine and other r
; particulates were not considered practical. The amounts of iodine and particu- ,
lates in a release are normally determined through analysis of samples taken ' during the release. This could require several hours, t Use of effluent monitors as the sole basis of a source term estimate / projection has four disadvantages: , j (1) Major releases may bypass the monitors (e.g., major containment failure) l
- and therefore will not be characterized.
l (2) The mixture being released may not be that assumed in the calibration for ' the monitor. The actual composition of the release may not be known for l several hours until samples have been analyzed. ! (3) Effluent monitors provide source term estimates at the time of the release, j As discussed in Volumes 2 and 4 of NUREG-1210, this may be too late for : implementation of the most effective protective actiors. (4) Accident conditions may influence the monitor; for example, because of contamination of the monitors, they may indicate releases long after the <
- releases have stopped.
Consequence / dose estimates based on effluent monitors are important, even con-sidering their shortcomings. Such estimates are of a known release while all other bases are for projected pathways. Estimates based on effluent monitor . 6 i i NUREG-1228 1-4 I l
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readings, when compared to actual field monitor'.ng data, can confirm or dis-count the existence of other release pathways. During events not involving core damage, effluent monitor readings may be the primary basis of source term estimation. 1.4.2 Source Term Estimation Based on Accidents Analyzed as Part of Licensing As part of the licensing process, analyses are conducted of various postulated accidents, some of which have the potential for offsite releases. These postu-lated accidents are listed in Section 15 of the Standard Review Plan (NUREG-0800). They include control rod drop accidents steam generator tube failures loss-of-coolant accidents (LOCAs) waste tank leak / failure fuel handing accidents spent fuel drop accidents anticipated transients without scram These analy:es are based on very conservative assumptions and, in many cases, unrealistic accident scenarios. The result is that for the accident conditions analyzed, the projected offsite doses are much grsater than realistic projections. In some cases, the differences between realistic and assumed calculations found ) in the Standard Review Plan could be several decades. Because of the unrealistic nature of these evaluations, they normally should not be used to estimate actual accident source terms or offsite doses. These analyses cannot even form an upper bound of offsite consequences since the containment is typically assumed to remain intact and only single failures are allowed. 1.4.3 Source Tern Estimation Based on Severe Accident Conwequence Studies Starting with the Reactor Safety Study (WASH-1400, 1975), studies have been ennducted to estimate the source terms that could result from various accidents invo ht.; core damage. The Reactor Safety Study analyzed many specific acci-dents and tne results were grouped into "release categories." Tables 1.a and 1.b of NUREG-1062 show the fission products released for each pressurized-water reactor (PWR) and BWR release ':ategory. These release categories are sometimes used to characterize the possible accident source terms by persons knowledgeable of probabilistic risk assessment (PRA) research. The specific accidents that are associated with a release category are typically described by symbols (e.g., AE-a). Tables 4 through 7 of NUREG-1062 provide a key to the set of letters used in WASH-1400 and other studies to identify spe-cific accidents. However, these designations normally are not understood by people not knowledgeable of PRA. In addition, a specifically studied, postulated accident sequence may not fit the actual accident conditions or the actua*i accident conditions may not be known sufficiently to be categorized. As an aid Figure 1.2 shows the relationship of basic accident conditions to WASH-1400 and other release categories and possible offsite consequences. This figure is in the form of an event tree. Moving from left to right, "yes/no" NUREG-1228 1-6
1 1 l l l t 1 Mm44T %4 Att t &H 0996ftf Cont STATUS C04?A14Wetut STATUS Svartte afAtys CAtt004'40 C04000Vt4C48 l i i t i(ATftash,,I T T I l Coat C0#4 WOLT aegy 8Amtv g g i papCOV4 AV Ot&L esAJoe g speaT omSUP9A46GK)4l 4 Sef wasw use g ' I 44 8 ' C04T&lg. s C04 Task, l #4414en-e;da s t l POOL l l Atts &S4 888 88847 l lpe Pvtk , l We, arf mage te a'4 T l Twe0V4M {8A4Wa0 l l pggyng g g , i i 1 I i i i
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PWG4 , $ 1P 89 0 l o a, , 1....f,... k0 S&BLV 64T 4 ' g j mt&4TM RPflCTS , w I s t I i rigure 1.2 Relationship of plant conditions to WASH-1400, siting source terms and consequences { i NUREG 't28 1-7 i
of answers to the accident conditions at the top result in a series of branche4 possible source term categories and offsite consequences. Since 1975, many additional studies have been conducted to correct shortcomings in the WASH-1400 analyses. In 1982, a set of source term categories (NUREG/ CR-2239) were constructed that also spanned the range of possible source terms associatedwithmajorreactoraccidents. These source terms are denoted as siting source terms (SSTs). Figure 1.2 also shows the relationship of plant conditions to these source terms. Figure 1.2 shows that knowledge of the status of major plant systems during an accident only allows a crude estimate of the source term. In the case of the most serious accident shown in Figure 1.2, several release categories (BWR/ This PWR 1 2-3) all are associated with the most severe accident conditions. may appear to be an unacceptable range, but this may be an accurate as possible during an actual accident. The NRC assessment of the Chernobyl source term However, the magnitude of the offsite con-wa.s that it was a PWR 1, 2, or 3. sequences for Chernobyl are not comparable to those projected for these release categories. This was the result of the release being in the form of a very high plume. This type of release is not thought to be possible at a U.S. reactor site. The WASH 1400 and SST release categories, if they can be related to the condi-tions during an accident by a knowledgeable analyst, or by figure 1.2, could provide a valuable tool in predicting the possible consequences of core damage accidents. NUREG-1062 provides a simple tool for relating WacB-1400 release meteorological categories (source terms) to offsite doses for a wide range o'. conditions. The advantages of this basis are that core damage accidents and unmonitored pathways are considered and various dominant accident scenarios can be compaired. However, in using this and all other source term estimation methods, the analyst must keep in mind the limitations and uncertainties involved. 1.4.4 Source Term Estimates Based on Detailed Analysis of Plant Conditions Computer codes (e.g., TACT, NUREG/CR 3287, NUREG/CR-4722) have been One developed by the NRC to predict releates (source terms) resulting from accidents. such code, TACT, allows the user to specify releases from the core, the patNay to the envirunnent (nodes), and conditions that may af fect the release (e.g. , These codes sprays). Figure 1.3 provides an example of the TACT code flow. require considerable detailed information about the plant (e.g., containaant volume), accident conditions (e.g., leak rates), and cases, In most effectiveness of source this detailed infer-term reduction mechanisms (e.g., spra',5). nation is not known during an acciden t. Moreover, these models, which are designed for use during severe accidents, are very crude compared to the sttte-of the-art models that have been developed to wstimate source terms for research purposes. As discussed in NUREG 0956, the state-of-the-art models are accurate only vithin a factor of 100 even if all Another the accident conditions are know disadvantage of these have greater than a factor of 100 uncertainties. codes is that they require considerable time and information, neither of which seems to be available during an accident. Finally, these codes are very flext-ble, and this also can be a problem if standard documented assumptions have not NUREG-1228 18
h r P FRACTION OF CORE INVENTORY ! OR f .i VOLUME OF COOLANT ' ( i lI l i NODE 1 FORCED FLOWS NODE 2 f m CONTAINMENT AUXILIARY
- BUILDING RP. CIRCULATION FILTERS REMOVAL AND i j
DECAY - FILTERS ! .- 2 SPRAY REMOVAL AND C DECAY LEAKAGE TO THE _= ENVIRONMENT 1 lf i I Figure 1.3 Sample MENU-TACT flow NUREG-1228 1-9 _ _ _ _ _ _ _ __. - - - - . ~, ___ ____. _ _. -
been established for the various model inputs. An additional problem with fle).ibility is that the results of these codes can be difficult to explain to others, to reproduce or compare with other analyses. This makes it very diffi-cult to use these codes to rank consequences of various accident sequences. These codes may be useful for analyzing lesser accidents for which conditions are well understood and when time allows this type of analysis. However, it is uncertain if the results of this approach would be any more accurate than using precalculated source terms. Considering the limited possible calculational improvement compared with the limitations, these codes do not currently appear to provide a useful response tool. 1.4.5 Source Term Estimates Based on Precalculated Assumptions of Dominant I Accident Conditions The remainder of this document describes, in general terms, the severe accident conditions that should dominate possible accident releases as a result of damage to the reactor core or primary coolant system. A method has been developed for estimating source term based on precalculated assumptions of dominant accident conditions. The basic assumptions of this method are that (1) there is a small set of accident conditions that dominate any severe accident release, (2) there are values that can characterize these dominant conditions, and (3) these con-ditions can be recognized / characterized during an actual accident. The following are the basic steps for source term estimation: (1) Estimate the inventory of fission products in the core. (2) Estimate the amount of fission products released from the core. (3) Identify the dominant release pathway. (4) Characterize the dominant mechanisms that will act to reduce the release. These would include filters, pools of water, sprays, or natural processes as illustrated in Figure 1.4. (5) Estimate the release rate. This method attempts to bridge the gap between using precalculated severe acci-dent source terms (WASH-1400) and conducting detailed calculations at the time of the event. In fact, this method arose from the observation that the same source conditions / accident assumptions were being analyzed over and over during events, drills, and exer.ises. The results will be a set of precalculated doses , I that can be used to compare possible consequences of various accident sequences. This method allows a large range of accident conditions and core damage states t to be analyzed based on a small set of predetermined assumptions that are well documented. The major disadvantage of this method, as with other methods, is the large uncertainties. However, a comparison (ranking) of the potential con-sequences associated with various accident scenarios will be possible. i NUREG-1228 1-10
l l WATER + SLOW FLOW FILTER + SLOW ~
) & DRY FLOW 1M 5% 1%
]__
, Q f,,
100 % l v NATURE + TIME (hrs) ' f h. 100 % q 90%' OHR 1 HR SPRAY + TIME (MINS) l [//I\D
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, (61 1 )
100 % 97 % 0 MIN 10 MINS Figure 1,.4 Examples of major aerosol /particulatr; reduction mechanisms
- NUREG-1228 1-11
2 RADIONUCLIDES IMFORTANT IN SOURCE TERM ASSESSMENT l 2.1 Objective To enable the reader to
+
identify the factors that have the greatest effect on radionuclide inven-tory in the reactor core identify the fission products that are important to offsite consequences 2.2 Fission Products To understand the significance of the radioactive materials in reactor accidents, it is necessary to describe how they originate and why they are hazardous to human health. 2.2.1 Source of Fission Products There are many mechanisms by which radioactive materials are created in a reactor core. The term "fission product" as used in this document will include not only the isotopes produced directly in fission (primary fission products) but also those produced indirs:tly through primary and fission product decay (secondary fission products) and other methods. 2.2.2 Inventory of Fission Products in the Core The first consideration in determining the contribution of a particular fission product to the overall source term is how much of the radionuclide is available in the core at the time of the accident. This is a difficult question to an-swer because the fission product inventory is influenced by a number of factors as shown in Table 2.1. However, as suggested by Table 2.1, the inventory of short-lived radionuclides is affected primarily only by reactor thermal power. Since short-lived radionuclides are the principal contributors to early health effects, reactor thermal power will be the only factor used to adjust assumed core inventory. Average power densities of operating reactors of a given size , and type would be very similar; therefore, power density considerations would not significantly change core inventories if core power and type are known. Burnup, although unimportant in considering the inventory of short-lived fis-sion products, is an important factor in determining the inventory of long-lived fission products. 2.2.3 Fission Products Important to Offsite Consequences Many fission products in the core do not need to be considered in source term estimation because they contribute little or nothing to offsite consequences. Many studies have been performe.1 to determine which fission products are the most important in terms of offsite consequences during severe core damage NUREG-1228 2-1 l
l l l Table 2.1 Factors that have the greatest effect on radionuclide l inventory Factor Effect Burnup Inventory of the long-lived radionuclides (e.g., ; Cs-137) is proportional to burnup. I Inventory of short-lived isotopes is not sensitive to burnup after the initial buildup (several weeks of full power operation). Power density Inventory of short-lived isotopes (those that reach equilibrium) is directly proportional to power density. Inventory of long-lived isotopes is not sensitive to power density at a given exposure (burnup). Reactor power Power reactors will produce fission product inventories proportional to the long-term thermal power level. Reactor type Fission product generation (inventory) is similar for BWRs and PWRs. accidents. Perhaps the most comprehensive study performed in this area is the Reactor Safety Study (WASH-1400). In this study the contributions of selected radionuclides to various organ doses as the result of a severe (core damage) accident were estimated. The results for early health effects are represented in Table VI 13-1 of WASH-1400. A scale from 0 to 2 was established in the Reactor Safety Study to delineate the contribution of a radionuclide to early and late health effects. A value of 2 indicates that the radionuclide contributed significantly to the specified effect, and a value of 1 indicates that the radionuclide had a small contribu-tion. A rough ranking of the importance of each group of radionuclides (e.g., iodine) from a health effects perspective can be obtained by summing the assigned scale values for each radionuclide. Those fission products that had a total score of 2 or more in WASH-1400, Table VI 13-1, in the areas that contributed to early health effects, have been selected to be considered in source term assessment. These are listed in Table 2.2. From the ranking in WASH-1400 ' can be seen that most of the noble gases (xenon, krypton) make a small contribution to health effects. However, noble gases are the most likely group of fission products to be released to the environment following a severe accident because they are chemically inert, are available in large quantities, and would not be affected by the various reduction mechanisms that would remove other fission products before they could be released. In addition, if all the noble gases in the core were released promptly, whole-body doses of about 100 rem are possible 1 mile from the plant. Therefore, noble NUREG-1228 2-2
Table 2.2 Fission product inventories (Ci/MWe) Fission Inventory product (Ci/MWe) l Kr-85 560 Kr-85m 24,000 Kr-87 47,000 Kr-88 68,000 Sr-89 94,000 Sr-90 3,700 Sr-91 110,000 Y-91 120,000 Mo-99 160,000 Ru-103 110,000 Ru-106 25,000 Te-129m 5,300 Te-131m 13,000 Te-132 120,000 Sb-127 6,100 Sb-129 33,000 I-131 85,000 I-132 120,000 1-133 170,000 I-134 190,000 i I-135 150,000 Xe-131m 1,000 Xe-133 170,000 Xe-133m 6,000 Xe-135 34,000 Xe-138 170,000 Cs-134 7,500 Cs-136 3,000 Cs-137 4,700 Ba-140 160,000 La-140 160,000 Ce-144 85,000 Np-239 1.64x108 Source: WASH-1400 j 1 gases will be included in the list of radionuclides to consider in source tom estimation. The importance of the various radioactive elements in terms of contribution to bone marrow dose can be seen in Figure 2.1. This figure shows the contribution assuming the most serious accident (e.g., BWR/PWR-1). NUREG-1228 2-3
- = ._
i l Te ! l A Y I Others Ba Sr Kr Cs
'l Figure 2.1 Relative importance to bone marrrow dose of the radioactive elements found in the core of a reactor given a major release Source: NUREG/CR-4467 NUREG-1220 2-4
l 2.3 Inventory Assumptions,To Be Used for Source Term Estimation For each of the radioisotopes listed in Table 2.2, a specific inventory ex-pressed in Ci/We is provided. These data are the standard starting inventory cited in many source term studies, such as NUREG-0956, and are in a1reement with other computer codes, such as CINDER results (NUREG/CR-3108). The com-puter codes used to estimate core inventory are considered to be te most accurate of all the codes used in estimating source term. These t. odes are considered to be accurate only within 25% for the given core burnup assump-tions. The values for specific activity correspc... to the end-of-equilibrium cycle and should generally provide an upper bound for fission product inven-tory. A rough approximation of the inventory (in curies) of a particular isotope can be obtained by multiplying the values in the "inventory" column by the electrical rating of the plant (in megawatts). Generally, the fission product activity would increase as burnup increased for isotopes with relatively long half-lives as illustrated by Figure 2.2 for Kr-85 (half-life = 10.7 years). The decreases in fission product activity, shown in " Figure 2.2, were the result of decay and unloading of spent fuel. Therefore, it should be noted that if the above assumptions are used, the resulting inventory estimates for a new core could greatly overestimate the quantities of long-lived fission products, such as Kr-85, Cs-134, and Cs-137. 0.8 i i E l j 0.6 - 1 ~ 1 8 5 e E 0.4 - - E ' E b 0.2 - - O I l 0 10K 20K 30K RELOAD EXPOSURE (MWD /MTu) Figure 2.2 Inventory of Kr-85 in core of 1000 We PWR Source: HUREG/CR-3108 l NUREG-1228 2-5 l
3 RADIONUCL10E TRANSPORT MECHANISMS UNDER NORMAL OPERATING CONDITIONS 3.1 Objective To enable the reader to describe the fission product barriers describe the changes that occur in fuel during normal operation and what effect these changes have on fission product transport from the core identify the fission products most likely to be released from the fuel and plant during normal operation recognize normal versus accident release rates specify the coolant concentration typical of an operating reactor that could be used to estimate the source term for a coolant release assuming no fuel damage and how these concentrations can change during rapid power changes (spikes) 3.2 Barriers to Fission Product Release For the fission products generated within the core to reach the environment, they must pass through four fission product barriers. These barriers are illustrated in Figure 3.1 for a pressurized-water reactor (PWR). The first barrier is the fuel pellet often referred to as the fuel matrix. The second barrier is the fuel pellet cladding. The reactor coolant system provides a third barrier to fission product release. The final and ultimate barrier to fission product release is the reactor containment. The purpose of this section is to describe the transport of fission products that routinely fission producttakes place under normal conditions (not accident) past these barriers. It is necessary to understand fission product transport during normal operation to recognize those release rates outside of the norm that may indicate an accident. Section 4 discusses the transport of fission products during accident conditions. 3.3 Radionuclide Transport From the Fuel Matrix Into the Fuel Cladding Gap The first barrier is the fuel pellet. After a reactor reaches full power, the fuel used in commercial light-water reactors (LWRs) undergoes thermal
- stor-tion and cracking because of the large temperature differences that e. .st between the center line and the surface of the fuel pellet. Figure 3.2 shows a cross section of a fuel pellet and the cracking that forms during operation.
As gaseous fission products are formed, they will move to cracks where they can escape the fuel pellet. The rate at which fidion products are released from the pellet will increase with increasing fuel temperatures. NUREG-1228 3-1
l STEAM GENERATOR TUBES, PART OF BARRIER 3
- 2. FUEL PELLET CLADDING
- 4. CONTAINMENT FUEL PIN l~ 50,000 I A COREl STRUCTURE pFISSION G AS PLENUM
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- .,,.s Figure 3.1 Fission product barriers to release of radioactive material for a PWR Source: NUREG-1210 NUREG-1228 3-2 l
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Noble and volatile fission product gases are the most likely to escape the fuel during normal operation. The volati b fission products include iodine, cesium, and tellurium (see Table 3.1). The fission product gas that escapes the fuel pellet is released to the gap formed by the fuel cladding or the fuel pin plenum specifically designed to accommodate the fission gas pressures over the life of the fuel. The collec-tion of fission products in the voids and plenum is often referred to as just "gap" activity. A cross se:: tion of a PWR fuel pin showing the fission gas plenum is shown in Figure 3.3. 3.4 Radionuclide Transport From the Fual Into the Coolant The second barrier to fission product release, fuel cladding or pins, prevents the fiFrion products that leave the fuel pellet from entering the coolant. The coolant refers to the water contained in the reactor coolant system that sur-rounds the fuel pins that form the core. Thus, an important objective in designing fuel pins is to preclude cladding failures. However, a small frac-tion of fuel pin cladding will leak during normal operation because of manu-facturing flaws, irradiation-induced creep, and other mechanisms. Current LWR fuel designs have demonstrated cladding failure rates of less than 0.1% over the in-core time of the fuel pins. The American Nuclear Society (ANS) Standards Committee Working Group has pre-pared a set of typical radionuclide concentrations for estimating the nor-accident radioactivity in the principal fluid streams of an LWR over its life time [American National Standards Institute (ANSI)/ANS-18.1-1984). The expected coolant concentravions for the reference plant types in the ANS standard are shown in Table 3.2 and can form the basis for source term estimation for coolant releases where actual coolant samples are not available. Actual coolant concen-tration levels can be several orders (10-100) higher than the ANS standard in plants with poor fuel performance; but generally these levels are in reasonable agreement (i.e., within a factor of 5) with actual measured coolant levels (NUREG/CR-4245). Coolant concentrations under accident conditions are discussed in Section 4.4. Rapid increases in the iodine and other fission products concentration as high as 3 orders of magnitude (i.e., a factor of 1000) may be seen following shutdowns, startups, rapid power changes, and reactor coolant system depressurization. Such increases are referred to as iodine spikes. Iodine spikes are increases that may not be a result of additional cladding failures. The NRC has estab-lished standard technical specifications for primary coolant iodine concentra-tions that make allowances for iodine spikes by permitting temporary excursions above the equilibrium concentration limit, as long as they do not exceed 48 hours. The failure to recognize the potential for an iodine spike may lead to con-fusion. It is important to recognize the fact that iodine spikes are possible and likely during an accident. Figure 3.4 illustrates a typical iodine spike following a reactor shutdown. Figure 3.5 illustrates a typical cesium spike l and the fact that iodine may not be the only fission product to spike. Spikes have been measured in a wide rango of isotopes (NUREG/CR-4245). In examining ( NUREG-1228 3-4
l Table 3.1 Melting point and boiling point (*F) of selected materials Material Melting point Boiling point Volatile fission products , 12 237 365 Csl 1158 2336 Cs0H 599 1814 Te 842 1810 ; Refractory fission products i Ba0 3493 5086 Ru 4082 7502 Sr0 4406 5880 La203 4109 >7232 Control rods Ag 1761 '3925 In 314 3763 Cd 609 1412 BC4 4478 >6332 Hf 4031 8042 Zircaloy Zr 3365 7968 Sn 449 4717 Stainless Steel Fe 2795 5183 Cr 3434 4841 Ni 2647 5277 Mn 2271 3743 Fuel UO 2 5144* 5959
*0xidized Zr will form a liquefied two phase mixture with UO2 at about 3484*F.
i NUREG-1228 3-5
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,, i C LA D D:N G ---* -
i ,,, (A) SPACER-Tt (B) Figure 3.3 Typical PWR fuel assembly (A) and fuel pin (B) NUREG-1228 3-6
Table 3.2 Typical coolant concentrations (pci/g) Radioisotope BWR PWR Kr-88 - 2.8x10 1 Xe-133 - 2.6 Xe-135 8.5x10 1 1-131 2.2x10 8 4.5x10 2 1-132 2.2x10 2 2.1x10 1 1-133 1.5x10 2 1.4x10 1 I-134 4.3x10 2 3.4x10 1 1-135 2.2x10 2 2.6x10 1 Cs-134 3.Ox10 5 7.1x10 8 Cs-137 8.Ox10 5 9.4x10 8 Ru-106 3.0x10 8 9.0x10 2 Te-132 1.0x10 5 1.7x10 8
; Ce-144 3.0x10.s 4.0x10 8 Source: ANSI /ANS-18.1-1984.
i t i 2 4 NUREG-1228 3-7 ! i
Pcwcr (%) o 6 0 1_ , g4' lig i i -
- W0;l:3 itlid :
5 :5 - Eh O l 131 EG&G. data ~ e 1131 CP&L data j _ g -
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e ,_,/ 10--3 i i i e i i e i i i ii 0600 1800 0600 1800 0600 1800 0600 1800 1200 2400 1200 2400 1200 2400 1200 2400 4/22 4/23 4/24 4/25 4/26 Date/ time Figure 3.4 Typical iodine spike following shutdown Source: NUREG/CR-4245. HUREG-1228 3-8
Pcwcr (%) 0 6 7 0 10 - 1 _ '
, l1ll 8 _
NC;S 41 IO$ g:g _
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O CS-137 EGEtG data M g _ ~ - e CS-137 CPuL data y eos e1 g _ O 10 - 2
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e o g _ _ cc 10 - 4 , 2 g ~ i , , i , , , i i i i ii 10 - 5 i 0600 1800 0600 1800 0600 1800 0600 1800 1200 2400 1200 2400 1200 2400 1200 2400 4/22 4/23 4/24 4/25 4/26 Date/ time Figure 3.5 Typical cesium spike following shutdown Source: NUREG/CR-4245, il NUREG-1228 3-9 i
Figures 3.4 and 3.5, note the rapid increase in coolant concentration when the power was reduced. These increases were observed during normal conditions. Fission products and other radionuclides released into the coolant are removed by cleanup systems to maintain the reactor coolant activity at equilibrium levels. The standard technical specification limits for reactor coolant activ-ity are 1.0 pCi/g I-131 dose equivalent for PWRs and 0.2 pCi/g for boiling-water reactors (BWRs). Licensees routinely collect and analyze coolant samples. Such samples would provide the best estimate of actual coolant levels (assuming no iodine spike). However, the concentrations shown in Table 3.2 also could be used because actual differences would be well within the uncertainties if these were to be i used to estimate the release resulting from a coolant leak. 3.5 Routine Effluent Releases Radionuclides are routinely released in nuclear power plant effluents during nor-mal operation or as a result of anticipated operationa7 occurrences. Releases during normal operation can be planned or unplanned, but they do not result in consequences that would warrant consideration of the event as an accident. The ranges of airborne effluent releases (curies) from BWRs and PWRs for the par 1980 are summarized in Table 3.3. Releases are not necessarily a function of plant size. Releases from a particular plant are most sensitive to factors such as the number of fuel cladding defects, design features of the plant radioactive waste treatment systems, the number of operational occurrences, and equipment performance. The major constituents of the airborne release for the PWR and the BWR are isotopes of the noble gases, xenon and krypton. Total releases, however, do not give any insight into which release rates (i.e., Ci/sec) are notmal and which require action. Table 3.4 shows the release rate benchmarks required (1) to exceed Environmental Protection Agency (EPA) and Food and Drug Administration (FDA) Protective Action Guides (PAG), (2) to reach the maximum reported annual airborne releases, and (3) to warrant further exami-nation by the NRC Headquarters Operations Officer (H00). These last levels were established for the NRC H00 to indicate when further health physics (HP) expertise and information on cause and corrective action should be obtained for events called into the NRC Operations Center. , Therefore, it is clear that release rates at least 1000 times normal are re-quired to exceed the offsite PAG. See NUREG-1210, Volumes 2 and 4, for a dis-cussion of release rates relative to health effects and PAG. NUREG-1228 3-10
Table 3.3 Range of total BWR/PWR airborne effluent i release (curies) for 1980 l l BWR PWR I-131 and I-131 and ;
> Range Gases particulate Gases particulate l Maximum 3.8x104 2.2 3.8x104 8.4x10 1
. Minimum 7.0x101 1.3x10 2 7.0x101 5.4x10 5 J t Source: NUREG/CR-2907, Vol. 3. l i
- Table 3.4 Release rate (Ci/sec) needed to meet benchmarks l
Benchmarks i l 1 Maximum reported EPA thyroid FDA thyroid NRC H00 , i annual EPA whole- PAG inhala- PAG inges- evaluation l i Release release in body PAG in tion in tion in levels - type 1 year (1) I hr (2)(4) I hr (3)(4) 1 hr (3)(4) (rate) (5) ! Noble gas 1x10 3 3x102 . . lx10 1 ! l i I-131 1x10 7 - 3x10 1 3x10 4 1x10.s i t (1) Assunies a constant release rate for 1 year to reach maximum reported, f Source: NUREG/CR-2907. L ! (2) 5 rem whole body. Source: NUREG-1210. 1 (3) 15-25 rem thyroid. Source: NUREG-1210. (4) Oose at 1 mile, E stability, 4 mph wind speed. Source: NUREG-1210. (5) Lovel if reported to the NRC Headquarters Operations Officer (H00) would warrant further evaluation by the NRC. Source: NUREG-1210. l
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4 ESTIMATING RADIONUCLIDE TRANSPORT MECHANISMS UNDER SEVERE ACCIDENT CONDITIONS 4.1 Objective To enable the reader to describe the phenomena that occur as a result of core heatup, beginning with normal operating conditions and ending with melting of the fuel pel- , lets, and the relationship of these phenomena to release of fission prod- - ucts from the core describe pathways that allow radionuclides to be transported from the core to the environment and explain what effect the pathway has on the fraction of radionuclides released describe the dominant mechanisms that act to reduce an offsite release and how to characterize those mechanisms during an event
- describe the procedure for bounding the source term for a potential release during an accident 4.2 Major Considerations Section 5 will outline a very simple method for estimating source terms for various severe accident conditions. This method is based on two assumptions:
(1) that there is a limited number of accident conditions that characterize the i size of source terms and (2) that these accident conditions can be estimated during an accident. This section will identify these conditions and character-ize their effects on source terms and discuss how these accident conditions can be estimated based on observable plant conditions. The hope is to relate actual plant conditions to the range of possible source terms. This section will be the bridge between source term science and actual source term estination during severe accidents. Section 5 will show how this can be applied to compare various possible accident sequences in terms of consequences or how the range of possible offsite consequences can be estimated. i To make the first approximation of a severe accident source term the analyst must: 2 (1) estimate the inventory of fission products in the core (2) estimate the fraction of the fission product inventory released from the Core i (3) estimate the fraction of the fission product inventory released from the core that is removed on the way to the environment NUREG-1228 4-1
i (4) estimate the amount of the available fission product invent. ry with poten-tial for release to the environment An estimate of the amount of material released from the core, if available (e.g., containment monitor readings), can be the starting point of the assess-ment. If this is the case, only steps 3 and 4 above would be performed. However, care must be taken because an unmonitored and unanticipated release pathway may exist. This section will identify the basic fission product inventory (FPI) in the core, core release fraction (CRF), reduction fraction (RDF), and escape frac-tion (EF) for various plant conditions. A severo accident source term can be estimated by: Source termj = FPI4 x CRF4x RDF(4, x EF j I i for radionuclide i and n reduction mechanisms where FPI 4 = element i core or coolant inventory CRF , i = element element i iinventory releasedin from core core RDF i = element i available for for release after reduction mechanisms i element i available release before reduction mechanisms element i released to environment [ EFi= element i available for release l
- The assumption that the total effectiveness of the RDFs can be estimated by multiplying them together is obviously suspect. The various reduction mech- l anisms (e.g. , pools or sprays) may be acting on the same form of a radioiso- l tope. Therefore, except for filters, a maximum total reduction factor of 0.001 will be assumed for the combined action of any set of reduction factors (except l for filters) on non-noble fission products. This maximum RDF (0.001) appears to be the maximum value observed in experiments involving any of the mechanisms ;
studied. The effectiveness of a filter in a release path will not be so limited l since under appropriate conditions filters are assumed to act on a wide range of i the forms of the fission products. : r l 4.3 Estimating Core Fission product Inventory To estimate the inventories of fission products in the core, Table 2.2 can be ' used. Specific plant inventory can easily be estimated by multiplying Table 2.2 f' t values b/ the long-term steady-state power level (MWe) at the time of the accident. f i l NUREG-1228 4-2 ! i
i When using the fission product inventories specified in Table 2.2, keep in mind that these values will greatly overestimate long half-life fission product (e.g., cesium) inventory in a new core. t 4.4 Estimatina Fission Product Inventory Released From the Core I The second step in estimating source term is to estimate the fraction of core ,
- inventory released following the failure of the first and second fission prod-
! uct barriers (fuel pellets and fuel pin cladding). Fuel and/or cladding fail-ures and subsequent release of fission products will primarily be a function of t temperature. For the purpose of severe accident source term estimation, four core damage temperature regimes will be discussed: (1) normal fuel pin leakage (normal operating temperature 600'F) (2) fuel cladding rupture release (gap release) (1300'-2100'F) I (3) grain boundary release (>3000*F) (4) melt (in-vessel) release (>4500 F) , However, because of the difficulties in specifically identifying core tempera-4 tures and conditions during a severe accident, only three of the levels of dam- { age would normally be used to represent toe full range of core damage: normal ! fuel pin leakage, fuel cladding rupture, and melt. Licensees have established specific procedures to assess the degree of core damage. In addition, the major reactor vendors have established guidelines for relating instrument readings (e.g., containment monitor, coolant levels, exit thermocouple temperatures) to the level of core damage. These procedures an; discussed generally in NUREG-1210, Volume 2. Table 4.1 shows the fraction of fission products that can be assumed to be re-leased from the core for each of the fuel damage states. Recall that these fission products were identified in Section 2 because of their importance to early health affects. Average operating fuel temperatures for light-water reactors (LWRs) are shown in Table 4.2 for comparison. The basis for these release fractions will now be discussed. l 4.4.1 Normal Fuel Pin Leakage (600'F) The first fuel damage regime corresponds to releases from fuel pins (the second fission product barrier) during normal operation. As discussed in Section 3.4, small quantities of fission products escape : through small holes in the fuel pin cladding into the reactor system coolant ; during normal operation. Typical primary coolant fission product concentra- . tions are shown in Table 3.2. Table 3.2 concentrations will be used to esti- ' mate source terms from normal coolant leaks. Results from plant coolant sample ' analysis could be used to scale (up or down) any results based on Table 3.2 concentrations. This obviously simple method provides sufficiently accurate results when considering all of the uncertainties. I i ' It is important to note, as discussed in Section 3.4, that following plant shut- l downs or other rapid changes in power levels, there can be rapid changes (as i high as a factor of a 100) in iodine, cesium, and other fission product coolant 1 i NUREG-1228 4-3 l
Table 4.1 Core release fraction assumptions (1-hour release) Fuel cladding Fission Assumed release Core condition temperature product fraction from fuel Fuel pin cladding 600'F Normal, use intact - normal Table 3.2 leakage Gap release 1300'F-2100'F Xe, Kr 0.03 (cladding failure) I 0.02 Cs 0.05 Te, Sb 1x10 4 s 3 Grain boundary >3000*F Xe, Kr 0.5 release I, Cs 0.5 ,' Te 0.1 Sb 0.02 j Ba 0.01 1 Mo 0.01 l Sr 1x10 8 i Ru 1x10 4 i Core melt >4500'F Xe, Kr 1. 0 (in-vessel)(1)(2) Cs 1.0 l I
- 1. 0 Sb 0.02 Te(3) 0.3 Ba 0.2 Sr(3) 0.0/
Mo 0.1 i Ru 7x10 8 La 1x10 4 Y 1x10 4 Ce 1x10 4 l Np 1x10 4 ; (1) Based on Tables 4.8 and 4.9 of NUREG-0956. i (2) For La, Y, Np, and Ce, the Zr release fraction was used, based on l 4 BMI-2104, Vol. VI, page 6-24 grouping, Battelle Columbus Laboratories,
! 1984. I i (3) Ex-vessel (melt-through) melt release fractions may be much larger !
i (0.4 to 0.8). i
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?
i i 4 , NUREG-1228 4-4 [
l l 1 Table 4.2 Average LWR fuel, cladding, and coolant operational temperatures Temperature-(*F) Areas PWR BWR ! I Inlet core coolant 569 532 l Outlet core coolant 625 547 : j Outer surface cladding 670 565 ! l Inner surface cladding 810 615 ! Fuel pellet surface 1235 1000 1 Fuel centerline ~3700 1650-3300 i L
- concentratiens. These cesium and iodine spikes are seen during normal opera- !
L tion. Therefore, under accident conditions any coolant samples taken before ;
! the plant shutdown or accident must not be considered as representative of ; ) actual coolant concentrations. Similarily, during an event, elevated coolant !
l concentrations can result from spikes that do not indicate accident conditions. ' f The good news on coolant concentrations is that even if a very large fraction of i.he coolant inventory at 100 times normal coolant concentrations were re-j leased into the atmosphere, doses in excess of the upper limit of whole-body Environmental Protection Agency (EPA) Protective Action Guides (PAG) (5 rem) t are not possible 1 mile from the plant (NUREG-1210 Vol. 2). Doses to the thy- l 1 roid in excess of the upper limit of the thyroid EPA PAG (25 rem) are possible, i i but would require the release of a large fraction of the total iodine in the l l coolant combined with adverse meteorological conditions. This is very unlikely. L I ' j Tables 4.3 and 4.4 show the calculated coolant concentrations for selected iso- i topes that can be assumed if various levels of core damage take place and the ! j core is reflooded trapping the majority of the fission products released from ! i the core in the coolant. The tables also show for comparison the American i Nuclear Society (ANS) typical concentration (Section 3.4) and those measured at : i Three Mile Island (TMI) following the accident. The TMI concentrations are ! l discussed in Section 4.5.2.1.6. J ; { 4.4.2 Fuel Cladding Rupture / Gap Release (1300*-2100'F) ! The second fuel damage regime, cladding rupture (gap release), corresponds to i l fuel cladding temperatures sufficient to result in the failure of the fuel ( l pins / walls (second fission product barrier). l As discussed in Section 3. gaseous fission products are released from the fuel f i during operation. Because the fuel is encapsulated in the cladding, fission l l } i g NUREG-1228 4-5 I i
E
- =
5 0 oo Table 4.3 PWR baseline coolant concentrations for various levels of core damage Core inven- Gap coolant Melt cool- TMI concen-tory Normal con- Gap concentra- Melt ant concen- tration (1000 MWe) centration release tion release tration +48 hrs
- Nuclide (Ci) (pCi/g) fraction (pCi/g) fraction (pCi/g) (pCi/g)
Kr-85 5.60x105 4.30x10 1 3.00x10 2 6.72x101 1.00 2.24x103 - Kr-85m 2.40x107 1.60x10 1 3.00x10 2 2.88x108 1.00 9.60x104 - Xe-133 1.70x108 2.60 3.00x10 2 2.04x104 1.00 6.80x105 - Xe-135 3.40x107 8.50x10 1 3.00x10 2 4.08x103 1.00 1.36x105 - I-131 8.50x107 4.50x10 2 2.00x10 2 6.80x108 1.00 3.40x105 1.30x104 I-133 1.70x108 1.40x10 1 2.00x10 2 1.36x104 1.00 6.80x105 6.50x108 i' I-135 1.50x10s 2.60x10 1 2.00x10 2 1.20x104 1.00 6.00x105 - Cs-134 7.50x10' 7.10x10 3 5.00x10 2 1.50x103 1.00 3.00x104 6.30x101 Cs-137 4.70x10' 9.40x10 3 5.00x10 2 9,40xio2 1.00 1.88x104 2.80x102 Sr-90 3.70x10' 1.20x10 5 0.00 1.2x10 5 7.00x10 2 1.04x108 5.30 Mp-239 1.60x10S 2.20x10 3 0.00 2.2x10 3 1.00x10 4 6.40x102 _ l Assumptions: WASH-1400 inventory. ANSI /ANS 18.1-1984 normal coolant concentrations. 2.5x105 kg primary coolant inventory.
- Concentrations of sample taken 3/29/79, 1600 counted 3/30/79 (>48 hours after accident).
_ _ _ ___ ___ __ _ _ m.._ _ _ . _ . . _ _ . _ . _ _ . . . _ . . . . _ _ _ _ _ _ _ _ _ . . _ _ _ .
! E
=
a i A , M ao Table 4.4 BWR baseline coolant concentrations for various levels of core damage Core Gap Melt i inventory Nornal Gap coolant Melt coolant (1000 W e) concentration release concentration release concentration 9 Nuclide (Ci) (pCi/g) fraction (pCi/g) fraction (pCi/g) 9 Kr-85 5.60x105 0.00 3.00x10 2 9.88x101 1.00 3.29x103 i ~ Kr-85e 2.40x107 0.00 3.00x10 2 4.24x103 1.00 1.41x105 ' Xe-133 1.70x10s 0.00 3.00x10 2 3.00x104 1.00 1.00x108 i' Xe-135 3.40x107 0.00 3.00x10 2 6.00x103 1.00 2.00x105 I-131 8.50x107 2.20x10 3 2.00x10 2 1.00x104 1 00 5.00x105
, I-133 1.70x10s 1.50x10 2 2.30x10 2 2.00x104 1.00 1.00x108 4 I-135 1.50x103 2.20x10 2 2.00x10 2 1.76x104 1.00 8.82x105
- Cs-134 7.50x10' 3.00x10 5 5.00x10 2 2.21x103 1.00 4.41x104 Cs-137 4.70x108 8.00x10 5 5.00x10 2 1.38x103 1.00 2.76x104 ~
Sr-90 3.70x10' 7.00x10 5 0.00 0.00 7.00x10 2 1,52x103 j Np-239 1.60x10' 8.00x10 3 0.00 0.00 1.00x10 4 9.41x102 1 ,
- ' Assumptions
- idASH-1400 inventory. '
- ANSI /ANS-18.1-1984 normal coolant concentrations.
1.7x105 kg of coolant. , 4 1 i ) 4
..~ .._--. -.. .- . - ...-- , - . - . , . , - - - - . , . - - , - -, - , - . . . . . _ . . . - -. - . . . - _
products are retained in void regions or in the gap between the fuel pin clad-ding and fuel. This collection of fission products is often referred to as the "gap." These fission products can be released very quickly if the cladding fails. If the gas pressure within the fuel pins is considerably less than the primary system pressure, the cladding may buckle or collapse at about 1300*F (NUREG-0900). balloon and oxidation may become extremely rapid between ThecladdingmagF,leadingtorapidfuelpinfailure. 1400 F and 2000 1 After a fuel pin cladding failure occurs, most of the fission gas in the fis-sion gas plenum will be released into the reactor coolant system (third fission product barrier). This release mechanism is often referred to as a burst re-lease. Once the fuel pin plenum fission gas has been vented, shallowly embed-ded gas atomr. in the fuel pellet and interior cladding surface near the rupture ' araa will diffuse into the reactor coolant system. The release mechanisms dis-cussed in this section will be collectively referred to as a "gap release." Table 4.1 shows the assumed gap release fractions. These fractions are those used in the Reactor Safety Study (WASH-1400). When using these release fractions, it is assumed that all the fuel cladding fails, resulting in a total core gap release, and that the core temperature is not substantially above that required for cladding failure. Higher temperatures would result in a release of substan-tially more fission products, as will be discussed. Obviously this is an unreal-istic model of damaged core, but can be useful to bound accident consequences. The amount of radioactive fission products associated with a gap releasa # 11 depend primarily on the number of failed fuel pins and on the plenum fiseb,n , gas inventory. It is important to realize that the variation in fuel tempera-ture throughout the core could be considerable. This is evident from the dif-ferent levels of core damage seen in the TMI core. Portions of the core were melted while other parts were not damaged. It is important also to realize that - once the core is sufficiently uncovered core temperature will increase rapidly J (at about 1F* per second for a pressurized-water reactor [PW]). In a PWR, i i cladding failure could begin in about 15 minutes following core uncovery and
! core melt in less than an hour. 1%refore, if core conditions allow cladding failure, rapid failure of all the tuel pins and even fuel melt could result, if conditions are not improved quickly. ;
The consensus is that even for the worst accident analyzed, if the plant safety ! systems work as designed, less than 20% of the fuel pin cladding will fail, re-leasing a large fraction of the gap in those pins. This failure would occur during the short period of core uncovery before flooding of the core. Under , l these conditions, most of the non-noble fission products released from the fuel ! would be trapped in the reactor system coolant. 1r.is type of accident (within l plant design limits) would result in release of considerably less than 20% of ;
; the gap from the reactor coolant system. Therefore, any accident that releases ;
more than 20% of the gap from the reactor coolant system is considered a severe accident. One of the best indicators of gap release will be the containment mon tor, if f the release is into the containment. The specific relation of containment f I I I NUREG-1228 4-8 i
monitor reading to fuel damage should be contained in the licensee's emergency plans. Another good indicator would be PWR core thermocouple readings or BWR water level. 4.4.3 Grain Boundary Release (3000*F) and Trit Release (In-Vessel) (4500*F) , Once the cladding fails, the release rate of various fission products increases rapidly with temperature. The release rate has been assumed to double approxi-mately with every 180F' increase in temperature. At about 3600*F, the fraction of the remaininc inventory is assumed to be released at about 10% per minute for noble gases, iocine, and cesium (NUREG-0772). By the time the fuel melts at about 4500*F, most of the volatile fission products may have been released from the fuel. Hcwever, these are basic assumptions, and, as was shown at THI, they may not be correct. At TMI up to 20% of various fission products expected to be released during a melt were still retained in the melted core. I It will be impossible to specifically determine the rate of release during core heatup and melt. Consequently, fuel releases during this phase will be charac- ' terized by two discrete pointr in the progression of fuel heatup. As the fuel heats up, following cladding failure, bubbles form and expand causing the fuel a grains to separate creating pathways out of the fuel. This is called the grain j boundary release.
'! The release fractions for grain boundary and core melt shown in Table 4.1 are bascd on a very simple model that relates the release rates to fuel temperature.
l This model is discussed ic NUREG-0772. Figure 4.1 shows the curves that form
'he basis for the release a numptions (NUREG-0772). These curves are based on
'ary scattered data and are estimated to be accurate only within plus or minus 1 a ftctor of 10. So even if the specific accident core temperature distributions could be accounted for (which they could not), the estimate of release rate from the fuel would only be within a factor of 10. [
T
! ' The Table 4.1 grain bour.dary release fractions represent the fraction released !
in 1 hour assuming a core temperature of 3000*F. Tuis temperature was chosen i because the release rste at this temperature is midway between that at start of l
] fuel damage (cladding failure) and that at fuel melt (4500'F).
[ The Table 4.1 melt release fractions are typical of those projected for various
,' core melt accidents by the Battelle Columbus Laboratories' calculations (BMI-2104) !
performed as part of the NRC reassessment of the technical bases for estimating i source terms (NUREG-0956). This phase of the core damage process is referred to 1 as "in-vessel melting." r Once the fuel is ur. covered, temperatures sufficient to melt fuel could be reached in an hour. It will be very difficult to actually estimate the extent of core damage once it begins. Therefore, if plant / accident conditions indicate pro-j longed core uncovery, a core melt release fraction should be assumed once core i damage beyond cladding failure is expected. Direct indications of this degree of core damage could be thermocouples, neutron monitor, or area and process i radiation monitors (e.g., containment monitor readings). ; 1 1 1 NUM.G-1228 4-9
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l l 4.4.4 Reactor Vessel Helt-Through (Ex-Vessel Melting) l There is one additional release fraction that is often discussed, vessel melt-through. If the core melt is allowed to continue (the core is not reflooded/ cooled), it may eventually melt through the bottom of the reactor vessel (the third fission product barrier) and fall on the concrete containment floor. This phase of the core melt process is often referred to as "ex-vessel melting." If the floor is covered with water (e.g. , from a leak), it may provide adequate cooling, stopping further release from the core. However, if the molten core I contacts dry concrete, or if, in some cases, the flooded corditions do not pro-l vide cooling, the concrete will rapidly decompose, yielding steam and carbon dioxide. Such processes increase the access of the fission products to the surface of the molten mass leading to the formulation of aerosols. Thus, there is a potential for additional fission products to be released. , s i Current information (e.g., THI core examination) indicates that there may be relatively high retention of volatile fission products (e.g., iodine, cesium) in a melted core. However, this material may be released after vessel melt-through. The melt case release fractions for cesium and iodine, shown in Table 4.1, may be representative for the worst-case melt-through, but may under-estimate the release of strontium during the ex-vessel phase of a core melt. The release fractions for strontium have been estimated (NUREG-0956) to be between 0.4 and 0.8 for some melt-through cases versus the 0.07 assumed in J
- Table 4.1 for the in vessel melt case. The release of a large fraction of this additional strontium to the atmosphere could increase the whole-body dose by '
as much as 50% above that estimated for the release on the basis of the core ' melt fractions shown in Table 4.1. M ver, the melt release fraction will be used to represent the core melt and vessel melt-through cases. ! l 4.4.5 Form of Release From the Fuel j In the next sections the transport of fission products from the fuel to the en-vironment will be discussed. The filter efficiency and other reduction factors used in calculating a source term would vary depending on the chessical and physical form of the specific radionuclides. The current consensus is that, i during a severe core damage accident, most non-noble fission products will form 1
- 1 particulates and aerosols. Therefore, it is assumed that all non-noble fission products will form a homogenous mass of aerosols and particulates. Filter effi-ciencies and removal coefficients will be applied equally to all non-noble fis-sion products. Noble gases are assumed not to be reduced by any of these mech-l anisms once released from the fuel. The primary way to reduce noble gas effects 4
is to contain them to allow time for decay or to control the conditions under which they are released. 4.5 Movement of Fission Products From the Core to Atmosphere (Reduction Factors) ! To estimate the amount of fission products released from the fuel that reach i the atmosphere, one must: (1) estimate the pathway the fission products will t l follow through the plant and (2) estimate the effectiveness of the various fis-sion product removal mechanisms encountered. On the basis of this information, l the reduction factor (RDF) for that particular release pathway can be estimated. L Table 4.5 lists those that have been selected to represent the dominant removal i mechanisms.
; NUREG-1228 4-11 i
il Table 4.5 Summary of particulate / aerosol gaseous (unless noted) reduction mechanisms < Release mechanism Reduction factor i Standby Gas Treatment System Filters: Dry-low pressure flow 0.01 Wet-high pressure flow (blowout) 1.0 , Other Filters: Dry-low pressure flow 0.01 Wet-high pressure flow (blowout) 1.0 1 Suppression Pool Scrubbing: Slow steady flow (decay heat)
, Pool subcooled 0.01 ' ; Pool saturated 0.05 Pool bypass 1.00 t Removal of Suspended Aeror.ols and Particulates: ! , i Natural processes (no sprays) !
0.5-hour holdup time 0.40 ' 1 2- to 12-hour holdup time 0.04 ; i 24-hour holdup time 0.01 ! Sprays on
! 0.5-hour holdup time 0.03 1 2- to 12-hour holdup time 0.02 24-hour holdup time 0.002 l Ice Condenser:
One pass through condenser (no recirculation) 0.5 Continual recirculation through condenser (1 hr or more) 0.25 Ice bed exhausted before core damage 1.0 Primary System Retention (Plateout): j Bypass accidents only 0.40 9 SG Partitionf.no (Liquid Release from RCS): l Normal partially filled "U" tube SG (liquid release) 0.02 , Water solid secondary side "U" tube SG (liquid release) 0.50 Normal once-through SG 0.50 SG = steam ganerator.
- RCS = reactor coolant system.
1 \ NUREG-1228 4-12
These issues will be addressed in two steps. First, each of the major removal mechanisms that may be encountered by the fission products as they travel through plant systems will be discussed in general terms. Second, the release pathways from the reactor coolant system (third fission product barrier) through the containment (fourth fission product barrier) and other possible barriers will be discussed along with the removal mechanisms that should be considered. 4.5.1 Basic Fission Product Reduction Mechanisms Two groups of fission product removal media that may be encountered on the way ! to the atmosphere are discussed below; these are gaseous and liquid. Table 4.5 summari:es the various removal mechanisms and associated RDFs. Al-though effects will be discussed in terms of specific plant systems, the con-cepts also may be applicable to related conditions. For example, the boiling-water reactor (BWR) suppression pool scrubbing effects also may be applicable for any large pool of water through which gaseous fission products pass. 4.5.1.1 Gaseous Release Reduction Mechanisms The aerosols and particulates will be carried through plant systems by steam and other gases released as a result of the core damage, if the core is not N flooded with water. This section discusses only those mechanisms that will be effective in removing aerosols and particulates from the gaseous / steam flow. 4.5.1.1.1 Plateout in tha Reactor Coolant System Depending on the conditions in the system through which the fission products pass, large quantities of the aerosols and particulates could condense and/or plate out on system surfaces (e.g., piping). Plateout is principally applied to movement of material through the reactor coolant system (RCS). There are many factors influencing this effect including surface areas and temperatures, flow rates, and aerosol or particle size. As discussed in NUREG-0956, a comparison of available computer projections shows dramatically different plateout (RCS retention) for different chemical forms and events. NUREG-0956 also states that "primary retention factors cannot be used rigorously as a multiplier of accident source term, nor can they be combined linearly with other retention factors (e.g., suppression pools or containment sprays)." Therefore, RCS retention generally will not be considered. However, for bypass accidents, system retention is the only reduction mechanism, and a reduction factor typical of those predicted by computer codes for this accident will be assumed (ROF of 0.4). This will be discussed in Section 4.5.2.1.5. 4.5.1.1.2 Removal of Aerosols and Particulates Suspended in large Volumes The fission product aerosols and particulates may be released into the contain-ment or another large, closed structure or volume (e.g., the auxiliary building). If these structures hold for a sufficient time to allow either natural processes (e.g., gravitational settling) or sprays to act, there will be a substantial reduction in the amount of material airborne and thus available for release through any existing pathways to the environment. There are two basic aerosol and particulate removal rates, as shown in Table 4.5, one for natural conditions NUREG-1228 4-13
and a second for sprayed volumes. If there is a major failure, it is possible that some of this material may become airborne and be released. However, this will not be considered. 4.5.1.1.2.1 Natural Removal Reduction Factors There are a large number of natural processes that work to reduce the airborne concentrations of aerosols and particulates. These processes have been examined by various codes and experiments. Figure 4.2 shows a typical result. Figure 4.2 shows natural processes reducing the concentrations by about a factor of 20 in the first 2 hours, followed by a dropoff in reduction rate. The con-centration is reduced by a factor of about 100 in 24 hours. This agrees with other experimental data and is within a factor of 10 of various computer pre-l dictions (NUREG/CR-4081) for the early timeframe. The codes generally predict reduction by a factor of 500 to 1000 in 24 hours (NUREG/CR-4081). For the pur-pose of source term estimation, a set of reduction factors as a function of time for large volumes was developed based on Figure 4.2. This is a very simple , assumption. In most accidsnts there will be continual release from the core > that could be replacing the fission products airborne in the containment. This effect may be compensated by the core release assumption.
; 4.5.1.1.2.2 Spray Removal Reduction Factors i
4 Reactor containments have sprays designed to remove airborne fission products and to condense steam to prevent overpressurization following ar accident. Other large structures, such as the auxiliary building, also may have spray systems i (e.g., fire suppression) that could be used to remove airborne fission products. Some reactor containments also have fire protection sprays that may be helpful
. in removing airborne fission products, but these are very site specific. '
l Sprays can be very effective in reducing airborne fission production concentra- ; l tions. Figure 4.2 shows typical results for a well-designed system. As in the j case of natural processes the initial effectiveness lessens with time. A reduc-J tion factor of about 20 is observed for sprays in the first hour and a factor l 2 of 500 in 24 hours. Table 4.5 lists reduction factors for sprays based on Figure 4.2. : I Airborne concentrations under accident conditions in the containment will not be as well behaved as shown in Figure 4.2. These predictions assume a well-4 designed and fully operational system. Variations as a result of accident-1 specific conditions, such as the rates of fission product release from the reactor coolant system must be expected. j 4.5.1.1.3 Filters Aerosol and particulate fission products released from the core may encounter a
- nun,ber of filter systems. These systems can be very effective for relatively ;
l small and dry flows. However, they can be blown out under high differential j pressure or can be clogged with aerosols. , y I I NUREG-1228 4-14 l 1 :
i l 100 TOTAL IODINE 2 O
-NATURAL EFFECTS ONLY Q .
l
=
NATURAL "".. SPRAY ON 10 - 1 - EFFECTS i ONLY o u; k E - 4._
- 3 > o ,.
10 - 2 _,
$ 3 SPR Y ON 8
- \ m i : %
NOTE SCALE 10 - 3 il _
, l , l l ; ( CHANGE O 1 2 3 4 5 10 1 30 TIME AFTER RELEASE, HUNDREDS OF MINUTES Figure 4.2 Comparison of removal of airborne aerosols and particulates by natural effects and sprays Source: Hanford Engineering Development Laboratory, llEDL-SA 2254 FP,1980 l
NUREG-1228 4-15
4.5.1.1.3.1 BWR Standby Gas Treatment System (SBGTS) The purpose of the SBGTS is to collect and filter any release from the BWR pri-mary containment--in some cases from the downstream leakage of the main steam isolation valves (MSIVs)--and filter the release to the plant stack. Typical SBGTS suction locations for various containment types are shown in Table 4.6. Table 4.6 Typical BWR standby gas treatment system suction locations Containment type Compartment (Mark) Drywell All Suppression chamber I, II HPCI gland seal exhaust blower I Reactor building I, II Refueling zone I, II Fuel building III Auxiliary building III, IIIA* Shield building annulus III Enclosure building IIIA* Mainsteam isolatten valve leakage II, III control system
- Alternative Mark III design.
Source: NUREG/CR-2940 The SBGTS can be very effective (99% efficiency) in removing aerosol and par-ticulate fission products. However, the SBGTS will perform effectively only for an accident where there is a minimal aerosol loading (dry) and limited pres-sure and temperature conditions with low flows; if the accident causes a high pressure differential, the filter or ductwork would be expected to rupture. In addition, the SBGTS will be bypassed during severe prisery containment failure accidents where the secondary containment blowout panel falls as a result of excessive secondary containment pressures (NUREG/CR-2672). Therefore, no credit should be given for the SBGTS for major p.'imary system leakage or failure, and a reduction factor of 0.01 (994 efficiency) should be assumed for small primary containment leaks in removing aerosol and particulate (non-noble) fission products. 4.5.1.1.3.2 PWR Containment Recirculating Filter System Many PWRs have a containment air recirculating system to trap fission product iodine following an accident. These systems have moisture separators to remove water droplets; however, they are not designed to remove the large quantities of aerosols expected in the containment atmosphere as a result of a severe core damage accident (NUREG-0772). These systems are expected to operate effectively c.J,y for cases of limited core damage (e.g., gap release). NUREG-1228 4-16 l
For a core melt accident, no credit should be given for these systems. These systems only complement and back up the containment sprays, and their effective-ness would be difficult to predict. These systems will not be considered in the method presented in this document. 4.5.1.1.3.3 PWR Auxiliary Building Filter Systems These filters are intended to treat exhaust air from equipment areas and vol-umes outside containment where there is a potential for the release of small l quantities of airborne fission products as a result of leaks in systems process-ing primary coolant system water that has been contaminated. As with other filter systems, these systems use high-efficiency particulate air (HEPA) and activated carbon absorbers and can be very effective in removing particulates and aerosols. As was the case during the THI accident, if filters are exposed to only small amounts of leakage, they should be assumed to be very effective (99% efficiency) with an assumed reduction factor of 0.01. However, for accidents that result in large releases into the auxiliary building, the flow rates are expected to result in pressures that would cause a major leak to develop, thereby creating a direct pathway to the environment. Under these conditions with high aerosol loads in the filtered material, the system might also plug and fail. Therefore, for major releases into the auxiliary building, it should be assumed that these filters will not be effective. 4.5.1.1.4 BWR Suppression Pool (NUREG/CR-3727) All BWRs have pressure suppression pools as part of their containments, which are designed to condense steam following loss-of-coolant accidents. The suppres-sion pool is a large mass of water through which the steam released from a break
; in the BWR reactor coolant system (third fission product barrier) blows down.
Although the suppression pool is not designed as a fission product removal system, a byproduct of the pressure suppression process is the scrubbing of fission products. In a severe accident involving overpressures caused by hydrogen or the accumulation of noncondensible gases, the failure or bypass of the suppres-sion pool is possible. Accidents involving vessel melt-through have been pos-tulated that involve the core melting through the containment liner, providing a suppression pool bypass. Obviously, the suppression pool is effective only if releases from the reactor coolant system leakage pass through the pool. The ability of the system to remove fission products will depend on (1) the energy of the blowdown and (2) the pool temperature. Tests have shown that the pools would have a very low decontamination efficiency (~ 15%) for violently flashing steam that may be present if more than decay heat (reactor not shut down) was being removed by the blowdown. However, there will be some retention resulting from condensation of steam and other factors. As a result, an RDF of 0.05 (50% retention) is estimated. However, for most postulated BWR accidents, the release of fission products is expected to be relatively slow with a steady flow of steam generated by decay heat. In this case, the removal rate could be very high depending on the temperature of the pool. If the pool were allowed to heat to the saturation temperature (e.g., because of inoperability of the resi-dual heat removal system (RHRS]), the scrubbing efficiency would be reduced. The assumed scrubbing efficiencies for various cases are shown in Table 4.5. Obviously, if the pool is bypassed it will not affect the release. l NUREG-1228 4-17
4 There may be other cases, such as a pipe break in a flooded area, here the gaseous release flows are scrubbed by a pool of water. If the pool is large enough, considerable scrubbing could be assumed. 4.5.1.1.5 PWR Ice Condenser (NUREG/CR-3727) An ice condenser containment was designed as an alternative to the large-volume containment traditionally used for PWRs. Like the large, dry containment, the ice condenser containment is designed to accommodate the large steam quantities associated with a loss-of-coolant accident (LOCA). However, the ice condenser feature permits a much smaller containment (50% less volume). In addition to suppressing the rise in containment pressure following a LOCA, the ice condenser also would tend to reduce the fission products in the containment air by entrap-ment and dissolution. A typical ice condenser system consists of an annular compartment that contains about 2.4 x los 1b of flaked borated ico at the outer circumference of the con-i tainment vessel. Following a large LOCA, the blowdown steam, fission products, 1 and reactor compartment air will flow through the ice condenser, where the steam l l will be condensed and fission products will be attenuated. The circulation of i the postaccident containment atmosphere through the Ice condenser is maintained by two axial fans, each with a capacity of about 40,000-ft /3 min. These fans
- j transfer air from the upper to the lower containment compartments, thereby inducing a flow through the ice compartments.
l Preliminary computer analysis concluded that the ice condenser has significant ! potential for removing fission products that pass through it. If the air return '
' fans were not available for circulation, the radioactivity released from the reactor coolant system would make only a single pass through the ice condenser; j even then, approximately one-half (RDF of 0.5) of the fission products released j to the containment are predicted to be removed by the ice bed. If the air l return fans are available to continually recirculate the containment atmosphere '
- ~ for an hour or more, even greater retention by the ice bed is predicted (e.g., j 75%). Therefore, for the purpose of source term estimation, a reduction factor [
I of 0.25 for recirculation cases can be assumed. However, accidents have been - l postulated during which the ice is exhausted before major core damage; under E I these conditions, no credit should be given for removal of fission products by l ice, j 4.5.1.2 Liquid Release Reduction Mechanisms i Fission products may be carried by reactor coolant or other contaminated water ,
. to a point where they can be released to the atmosphere. The levels of contami- i i nation could be very high; for example, a severely damaged core could be reflooded l
! resulting in vastly increased coolant concentration. During the TMI accident, the reactor coolant, after core damage, had 300,000 times the normal levels of l 4 fission product contamination (NUREG-0600). Another source of contaminated i water could be the water in the containment following spray operations to scrub l fission products from the containment atmosphere. In these cases, the quantity ; J of contaminants released to the atmosphere will be a function of the rate the
- fission products escape from the liquid. Releases of fission products in water !
4 that are not released to the atmosphere directly (i.e., basemat melt-through) l are not covered. l 1 NUREG-1228 4-18 , i
The following release fractions for liquid transport cases will be discussed (Clinton,1984): (1) boiling water , (2) hot coolant (water) vented directly into the atmosphere that flashes to steam or is atomized As shown in Figure 4.3, steam generator tube ruptures in a PWR can provide three major release pathways to the environment. If the main steam line does not isolate and the condenser is available, it would be preferred to direct the reactor coolant system leakage to the condenser. Under these conditions, it can be assumed (NUREG-0909) that only the noble gas in the coolant will be released through the steam jet-air ejector to the atmosphere. If the main steam line isolation valve is closed (as shown in Figure 4.3) the safety valve and auxiliary feedwater turbine exhaust provide pathways to the environment. However, in most cases such releases would be greatly reduced as a result of partitioning in the steam generator. The steam from boiling uter has a con-siderably lower concentration of contaminants than the water being boiled. ; This process of decontamination is measured in teres of a partitioning factor where: Mass partition factor = mass fissi n products entrained in unit mass of water mass fission products entrained in unit mass of steam f In cases such as normal "U" tube steam generator operation where a clear boundary
; exists between the boiling wate* on the secondary side and steam generated (Figure 4.4, Case A), a mass partitioning factor of 20 to 250 has been measured. ,
In this case, it will be assumed that only about 1/50 (2%) of the iodine and - other non-noble fission products in the mass of boiled water (release) are
! passed on to the steam, based on a mass partition factor of 50. Therefore, where a water / steam interface exists in a release pathway, a filter efficiency of 98% (RDF of 0.02) can be assumed for dry steam. However, if contaminated water is carried with the steam, the partition factor decreases. This can result from the steam generator design or size and condition of the steam gener- , ator tube rupture. The Babcock & Wilcox (B&W) steam generators, referred to as "once-through steam generators," do not produce a clear steam / water boundary, i and considerable amounts of contaminated water leaking into the secondary side l could be carried with the steam. For N" tube steam generator designs, the leak should be under the secondary side water as shown in Figure 4.4, Case A.
However, if the tube leak is massive, the secondary water could be boiling violently, and the secondary side steam would carry considerable contamination. This is called carryover. For large multitube failures or for the once-through ' steam generator (B&W), an RDF of 0.5 should be assumed. L During the Ginna accident (NUREG-0909), the secondary side filled with coolant following isolation as shown in Case B of Figure 4.4. The primary coolant was then ejected into the atmosphere through the safety valve. Therefore, the
; iodine concentration in the release was considerably above what would be ex- ; pected. lhis is an example of the contaminated vater and all of its contami-i nants being ejected into the atmosphere, the water flashing to steam and stomizing, carrying much of its contamination into the atmosphere. However, it is expected that contaminated water will be diluted by the water in the secon-dary side. During the Ginna accident, secondary side dilution reduced the NUREG-1228 4-19 l
RELEASE TO i i ATMOSPHERE V STEAM GENEPATOR
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, LEAK PATHWAY N CPEN VALVE
>4 CLOSED VALVE Figure 4.3 Typical pathways for release of radionuclides to the atmosphere from a steam generator tube rupture NUREG-1228 4-20
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' LEGEND GINNA ACCIDENT CONDITION CASE A TYPICAL CONDITION STEAM FOR A STEAM GENERATOR U" TUBI 2 RUPTURE PRIMARY SECONDARY WATER Figure 4.4 Conditions in radionuclide release pathways following a "U" tube steam generator tube rupture to the environment
release concentration in half. Therefore, a RDF of 0.5 is assumed. This situation also may be encountered when a pipe breaks or safety valve opens directly into the atmosphere and releases hot and high pressure water. Such cases also have the potential of leading directly to severe accidents that bypass the containment and releasing fission pr Mucts dissolved in the coolant directly into the atmosphere. 4.5.2 Basic Release Pathways and Characteristics Section 4.4 discussed the release of fission products into the reactor coolant system from the core. This section will show the major pathways the fission products can take through the remaining fission product barriers. The fission product reduction mechanisms encountered will be listed. Two simplified dia-grams that highlight the basic release pathways of PWR dry and ice condenser containments (Figures 4.5 and 4.6, respectively) and three simplified diagrams } that highlight the BWR Mark I, Mark II, and Mark III containment release path-1 ways (Figures 4.7, 4.8, and 4.9) will be referenced in the discussion. The _) basic release pathways are designated on the figures by letters and numbers. The letters refer to a specific barrier and the numbers specify a specific path-way through the barrier. Tables 4.7 and 4.8 list the specific PWR and BWR path-ways and indicate the specific reduction mechanism that should be considered. Table 4.9 provides a key to the release pathways used in the figures. b 4.5.2.1 Reactor Coolant System (Third Fission Product Barrier) Once the core is damaged, the fission products are initially released into the . reactor coolant system. Without a breach of the retctor coolant system, major fuel damage is not possibit. Therefore, major fuel damage indicates that there was at some time a break in the reactor coolant system. Reactor coolant system breaches are shown as "A" on Figures 4.5, 4.6, 4.7, 4.8, and 4.9. 4.5.2.1.1 PWR Reactor Coolant System Breaks / Leaks A break in the reactor coolant system (RCS) is shown as A-1 in Figures 4.5 and 4.6. The PWR RCS is contained within the containment. The discussion in this section will only deal with breaks that result in RCS breaks directly into the containment. Most analyses of this event assume that the break in the RCS starts the accident. In this case, it is assumed that the break is followed by a failure of the sys-tems designed to reple:e the coolant lost as a result of the break, which leads to core uncovery and damage. Therefore, the break exists before core damage and provides a direct pathway for fission product release from the fuel into the containment atmosphere. For all the PWR containment designs, except for the ice condensers, the gas flow from the damaged core and through the break does not pass through any filter mechanisms. For those plants without ice condenser containments, the only reduction mechanism that is important is plateout in the RCS, and it will not be considered (as discussed in Section 4.5.1.1.1) except for bypass accidents. In the ice condenser containment design, the flow from a break in the RCS flows through an ice condenser, where . team will be condensed and fission products NUREG-1228 4-22
z
$ %/ .
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b , r M ,3 CONDENSER d ' # y AUXIUARY fg s FEEDWATER LOW PRESSURE T TURBINE Og SYSTEM AUXIUARY BUILDING CONTAINMENT TURBINE BUILDING N REUEF VALVE M SPRAY { CLOSED VALVE +g FILTER Zi CHECK VALVE RELEASE PATHWAY REFERENCE Q PUMP M ISOLATION VALVE Figure 4.5 PWR dry containment simplified release pathways i
s 88 ' M tis P 4 CONTAINMENT UXIUARY FEEDWATER TURBINE EXHAUST SAFETY
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/{ CHECK VALVE RELEASE PATHWAY REFERENCE h PUMP M ISOLATION VALVE Figure 4.6 PWR ice condenser containment simplified release pathways
_ _ _ _ _ _ _ - _ . - _ - _ _ . _ _ _ _ _ _ _ _ _ _ ~ _ _ - _ _ _ - z h m ! ? SECONDARY CONTAINMENT h. REACTOR BUILDING i N BLOW OUT PANEL PRIMARY - i CONTAINMENT . p,,,,,- ; Nr it 5 3
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I WET WELL SUPPRESSION POOL U TUR8INE BulLDING l I Figure 4.7 BWR Mark I simplified release pathways
d SECONDARY d " CONTAINMENT .
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LEGEND SUPPRES$10N POOL s A.3 -TO SPRAY W .SBOTS Q FILTER d CHEIK VALVE RELE Alf PATHWAY REFERENCE SBGTS SUCTION La1 ISOLATION VALVE BETWEEN M56VS . M CtOSED VALVE g rump i Figure 4.9 BWR Mark III tieplified release pathways J NUREG-1228 4-27
.._. _ _ . _ _ _ .- __._.__.__._..m.. . _..._ ._ _ .. _ _ .___ _ ._ _ ___ _. ___ _ _____= _ _ _ - - _ _
g Table 4.7 Summary of major PWR release pathway reduction mechanisms A o a Fissicn product Reference Fig- Reduction y barrier breached Pathway Non noble m ures 4.5, 4.6 mechanism Condition assumed RDF Reactor coolant Break / leak A-1 Ice condenser system: Single pass fans 0.50 and open PORY A-2 fail I hour of recir- 0.25 culation through ice condenser Condenser bypass 1.00 or ice exhausted before core damage Steam generator A-3 SG "U" tube secondary
, tube rupture 0.02 partitioning side boiling (dry 4
e steam)
"U" tube secondary 0.50 side solid, no boiling Normal OTSG (88W 0.50 steam generator)
Dry steam generator 1.00 Bypass (failure A-4 Primary into low pressure For bypass accidents 0.40 system only system) i retention vy containment: Design leakage 8-1 Containment (0.1%-0.25%/ day) 0.5-hour holdup in 0.03 sprays containment or Small isolation B-2 valve seal failure 2- to 12-hour holdup 0.02 (1005/ day) in containment or
- -- - _ _ , _ _ . , _ _ _ . _ _ _ . _ _ _ _ . - - . ~ . _ _ . - - , - _ - - _ . _ - ~ ~ - . . . - . . . . _ _ _ _ - . _
l E Table 4.7 (continued)
=
"3 c Non-noble Fission product Reference Fig- Reduction 4 Pathway mechanism Condition assumed ROF
, Z barrier breached ures 4.5, 4.6 t ao l Primary containment Catastrophic 8-3 24-hour holdup 0.002
- (continued)
- failure (1005/hr) in containment l
Nat. oral proc- 0.5-hour holdup in 0.40 J esses in containment I containment (no spray) 2- to 12-hour holdup 0.04 in containment l 24-hour holdup 0.01 in containment
, Bypass B-4 Same as A-4 Same as A-4 Same as A-4 Other: Secondary side C-1 None relief / safety valves and turbine exhaust , Building C-2 Natural 0.5-hour holdup 0.40 failure / leakage processes in building (not filtered) (no spray) 2- to 12-hour holdup 0.04 in building 24-hour holdup 0.01 in building i Through- C-3 Filters Small release into 0.01 ! building building (filters I filters effective)
Large release into 1.0 building (filters blow out)
- . . _ . . - _ _ . . _ . _ . . . . . _ _ _ . _ , - . . _ . ~ , . . - = - - . - -_ - . , _..-- -. . ,. - - . . . - . , . - - . . . . ._. - , _ - - . . _ _ _ . . _ _ --.._______I
5 Table 4.7 (continuad) A c> 4 Fission product Reference Fig- Reduction Non-noble
% barrier breached Pat.hway ures 4.5, 4.6 mechanism Condition assumed RDF m
Other (continued): Steam-jet air- C-4 Condenser All noble gases are 0.00 ejector assumed to be released (no non-noble)
?
8
E Table 4.8 Summary of major 8WR release pathway reduction mechanisms c) L Reference Non-noble M Figures 4.7, Reduction assumed
- Fission product barrier breached Pathway 4.8, 4.9 mechanism Condition RDF Reactor coolant Break / leak A-1 None None 1.00 system: bypasses suppression pool Break / leak A-2 Suppression Slow flow - decay heat 0.01 through pool and pool subcooled suppression scrubbing pool Slow flow - decay heat 0.05 and pool saturated Pool bypassed 1.00 Through ADS / A-3 Same as A-2 Same as A-2 Same as A-2 SRV Bypass of A-4 System For bypass accidents 0.40 containment retention only Primary containment: Design leakage 8-1 Dry well 0.5-hour holdup in 0.03 (0.5%/ day containment dry well or sprays (ON)
Isolation valves 8-2 2- to 12-hour holdup 0.02 seal failure in dry well
- 1005/ day or Catastrophic B-3 24-hour holdup in 0.002 failure dray well (11%/hr)
E Table 4.8 (continued) N,
?
M O Reference Non-noble Fission product Figures 4.7, Reduction assumed barrier breached Pathway 4.8, 4.9 mechanism Condition RDF Primary containment Dry well 0.5-hour holdup in 0.40 (continued): and wet dry wil well natural processes 2- to 12-hour holdup 0.04 (no spray) in dry well 24-hour holdup in 0.01 dry well Bypass B-4 Same as A-4 Same as A-4 Same as A-4 a Controlled 8-5 Same as Same as 8-1, 2, or Same as 8-1, venting staney gas 3 and C-2 2, 3 and C-2 treatment system (58GTS) Secondary Building C-1 Natural 0.5-hour holdup in 0.40 contain n t: failure / leakage processes building (not filtered) (no spray) 2- to 12-hour holdup 0.04 in building 24-hour holdup 0.01 in building
i i " 4 ! .g Table 4.8 (continued)
- =
I E' i .. Non noble
~
m Iteference ,
= Fission product Figures 4.7, Reduction assumed barrier breached Pathway 4.3, 4.9 mechanism condition IIDF !
Secondary containment Through 58GTS C-2 58GTS Small release 0.01 (continued) filters (in (minimal aerosol addition to loading - filters natural effective) I process) ; Filter failure - 1.00 rupture (heavy aerosol loading / , 7 large high pressure i release) , i t , t: t I i I l i 1 a
. vem. , , - - - - , - - - - _ . -_.-.- _ e- w w. - - - , er--. v,v.~.-- ew,,--m..m.-.,---..---.--------.-r-, -- . . . . - - - - - - - - - - _ _ _ _ _ _ _ --
Table 4.9 Key to release pathway references on Figures 4.5 through 4.9 Pathway Description PWR - Finures 4.5 and 4.6 A Reactor Coolant System A-1 Breaks and leaks - A-2 Power operated relief valves (PORVs) A-3 Steam generator tube rupture A-4 Bypass (failure into low-pressure steam) B Containment B-1 Design leakage B-2 Small isolation valve seal failure B-3 Catastrophic (>l sq ft) B-4 Bypass C Other C-1 Secondary side relief / safety valve or turbine exhaust C-2 Building leakage - unfiltered C-3 Building leakage - filtered C-4 Condenser steam-jet air-ejector BWR - Figures 4.7. 4.8. and 4.9 A Reactor Coolant System A-1 Breaks and faaks bypassing suppression pool A-2 Breaks and leaks through suppression pool A-3 Automatic depressurization system (ADS) and safety relief valves (SRV) B Containment B-1 Design leakage B-2 Small isolation valve seal failure B-3 Catastrophic B-4 Bypass C Other C-1 Building leakage - unfiltered C-2 Standby gas treatment system (SBGTS) NUREG-1228 4-34
l i f i will be attenuated. For cases where air return fans are not available (fail), . an RDF of 0.50 should be assumed. If the RCS release is circulated through the i i ice bed, an RDF of 0.25 should be assumed until the ice is exhausted. In some i j accidents analyzed, the ice is exhausted before core damage. Under those con- 6 ditions no credit should be given for the ice bed (RDF of 1.00). l 4.5.2.1.2 PWR Power-Operated Relief Valves j PWR RCS have relief valves designed to prevent the system from overpressurizing. [ If the systems designed to remove heat from the RCS fail, the RCS pressure will increase until the relief valves on the pressurizer open. If the rate of cool- , ant loss through the relief valves exceeds the rate of coolant makeup (i.e., ; j via safety injection), the core can become uncovered resulting in core damage. !
- This same release pathway could become the release path from the RCS (A-2 on !
l Figures 4.5 and 4.6). This was the primary pathway for the fission products ; l released from the RCS during the TMI accident. l ! As shown in Figures 4.5 and 4.6, the relief valves do not release directly into l l the containment, but through piping into a relief tank. The relief tank is de- ! l signed to condense the steam released during normal operations. As happened i
- during the TMI accident, large quantities of water fill the tank, resulting !
- in the rupture of a disk and allowing a path for fission products into the l q containment. :
- l l
} Although some scrubbing may take place as the coolant passes through the relief [ i tank, it will be difficult to estimate the extent to which fission products are ; j retained. It is generally assumed that the relief tank rupture disk would open. l , providing a dit-ect pathway before core damage. Consequently, no credit will be t given for scrubbing. l 4.5.2.1.3 PWR Steam Generator Tube Rupture When the steam generator (SG) tubes fail, a pathway directly to the atmosphere i (i.e., bypasses containment) may be provided. This is shown on Figure 4.3 and as pathway A-3 on Figures 4.5 and 4.6. When a steam generator tube rupture (SGTR) occurs, the fission products from the higher pressure primary system i pass into the secondary side (A-3) and possibly into the atmosphere through the ! I secondary relief valves or turbine exhaust (C-1 on Figures 4.5 and 4.6) or by the condenser steam-jet air-ejector exhaust, j Also as discussed f.1 Section 4.5.1.2 and shown on Figure 4.3, if the condenser i is available and the main steam line does not isolate, the release would be ( j through the steam-jet air-ejector exhaust. In this case only the noble gas in j l the contaminated water is assumed to be released (C-4 ori Figures 4.5 and 4.6). l Generally, this is thought to be a release pathway for fission products dis- I solved in the primary coolant. This also could provide a dry (bypass) release ! pathway for a gaseous and aerosol releases from the core. Dry releases by SGTR ! will be considered as part of bypass accidents (Section 4.5.2.4). j ! If primary coolant leaks into the secondary side of the steam generator and the I f secondary side is allowed to boil, the resultant steam that escapes to the l l atmosphere by the relief valves (C-1) will contain considerably lower fission I 1 l 1 I i y NUREG-1228 4-35 I t ___ _ _ - _ . . _ _ _ _ _ _ _ - _ _--. - _ _ _ ~ _ _ - _
product concentrations as a result of partitioning, as discussed in Sec-tion 4.5.1.2.2. In this case an RDF of 0.02 should be assumed. In some designs, if sufficient primary coolant leaks into the secondary side of the steam generator, the generator will fill and not allow boiling. The entire secondary side could fill, pushing coolant directly into the atmosphere through the relief valve (C4) as happened during the Ginna accident (see Figure 4.4). There is considerabl: dilution when the coolant passes through the secondary side before release. Jnder these conditions, a 50% reduction would be assumed (ROF of 0.5). As noted in Section 4.2, the last factor in estimating a source term is the esti-mated amount of fission products available for release that are actually re-leased. This is referred to as the "escape fraction" (EF). In the case of an SGTR, there are some benchmark EFs that can be precalculated. A full rupture of one steam generater tube could result in leakage of about 500 gpm (7s efm) or about 35% of the total primary coolant volume in 1 hour, assuming the reactne coolant system pressure remains near operating pressure and lost volume is replac.id with the charging pumps. An EF of 0.35 should be used for a 1-hour release by SGTR. This escape rate would fall rapidly as the primary system pressure falls. As soon as the reactor coolant system pressure equals that of the secondary system, the leak will cease. A second benchmark for the escape fraction could be the primary system makeup, provided by a single charging pump at 50 gpm (EF of 0.03), which is being used to make up for coolant leakage at the break. Escape fractioas are summarized in Table 4.10. 4.5.2.1.4 PWR and BWR Bypass Accidents There are a number of low pressure systems that connect to the high-pressure reactor coolant system (RCS). In some cases, low pressure lines and components (e.g., pumps) are located outside the containment and separated from the RCS by check valves. If these valves fail, the high pressure coolant from the RCS could rupture the low pressure system outside the contaiwent as shown by pathway A-4 of Figures 4.5, 4.6, 4.7, 4.8, and 4.9. This failure would provide a pathway from the reactor coolant system to the environment that bypasses the containment. This type of accident is often called a bypass accident, interfacing LOCA, or an event V. In fact, the steam generator tube rupture accident discussed above is a type of bypass accident. Low pressure lines providing a potential direct leakage path from the RCS to locations outside the containment have redundant isolation valves inside and/or outside of the containment so that leakage of coolant from pipe leaks or rup-tures outside of the containment would be limited by closing the check valves or other isolation valves. Failure of these valves to close is possible however, resulting in loss of coolant and a potentially unisolable direct leakage path from the RCS to the auxiliary building or other buildings and from there to the environment (Figures 4.5 and 4.6, C-2 or C-3; Figures 4.7. 4.8, and 4.9, C-1 or C-2). Release by this pathway from the primary system would not be filtered unless the fission products would pass through pools of water. The primary reduction mechanism is reactor coolant / secondary system plateout/ retention. An RDF for this pathway is assumed to be 0.4 for system retention, on the basis of results NUREG-1228 4-36
l L Table 4.10 Assumed containment and steam generator tube rupture escape fraction for 1 hour Escape Release pathway fraction
- Primary containment failu*e/ leakage Typical design leakage:
PWR - large dry (0.1%/ day) 4x10 5 PWR - subatmospheric (0.1%/ day) 4x10 5 PWR - ice condenser (0.25%/dav) 1x10 4 BWRs (0.5%/ day) 2x10 4 Failure to isolate (100%/ day): Failure of isolation valve seal 0.04 Castastrophic failures: 1-br puff release 1 Steam generator tube rupture 1 tube at full pressure 0.35 (coolant leak) - 1 tube at low pressure single 0.03 charging pump flow (coolant leak)
- Fraction of containment volume or primary system coolant inventory released in 1 hour, described in NUREG-0956. In addition, the re13ase is expected to be into another building (e.g. , turb%e building); therefore, consideration should be given to any removal mechani.uis (e.g. , natural processes or filters) that may be encountered as discussed in Section 4.5.2.3.
4.5.2.1.5 PWR and BWR Vessel Melt-Through As discussed in Sectioa 4.4.4, a core melt accident in the absence of any rea flooding could lead to core melt debris penetration of the reactor vessel bottom , head. For the reasons discussed in Section 4.4.4, the release pathways active during the melt phase will be used to characterize the release from the primary system melt-through. NUREG- 108 4-37
4.5.2.1.6 PWR and BWR Leakage From Process Lines Fission products can be transported out of the RCS through connected lines that are run outside the containment building ano into other areas of the plant. Because some of these processes are required even after an accident, thesa path-ways exist after containment isolation. A leak or rupture in one of these lines would allow reactor coolant and the dissolved fission product gases to escape into at ta outside the reactor containment building. Noble gases and the more vole ..t fission products would be released immediately to the area of the plant w..ere the leak occurred and eventually could be released to the atmosphere. This type of containment bypass was the major source of release during the THI accident. As shown in Figure 4.10, there were two release pathways out of the containment during the TMI accident (NUREG-0600). The first resulted from the sump pump automatically starting and pumping the reactor coolant that had collected in the sump as a result of the PORV coolant release (see Sec-tion 4.5.2.1.2). This coolant was pumped to a waste holdup tank in the auxil-iary building. This tank eventually filled, causing a rupture disk to open, which allowed noble gases and iodine contained in the coolant to be released to the auxiliary building. This transfer of coolant took place during the first hour of the accident and well before the first signs of cladding failure. Therefore, the amount of fission products released was small. The cecond and major source of releases from THI resulted from reactor coolant flow through the makeup and purification systems during the accident. Because these processes continued during the course of the accident, this coolant was highly contaminated. For example, a coolant sample taken on March 29, 1979, indicates an I-131 concentration of 1.3x104 pCi/cc (NUREG-0600). A typical I-131 level (see Table 3.2) is 4.5 x 10 2 pCi/cc. Tables 4.3 and 4.4 show the coolant concentrations for selected nuclides if either the total gap or melt inventories were released into a typical coolant inventory. These concen-trations are based on Table 4.1. The TMI coolant I-131 cc,ncentrations were about 300,000 times normal. As shown in Table 4.3, the THI I-131 concentrations fall between the gap and melt concentrations, as would be expected. The cesium concentrations are smaller than projected because this was a new core and the cesium had not yet built up, as discussed in Section 2.2.2. The TMI releases from the containment systems were into the auxiliary building, and the natural processes (discussed in Section 4.5.1.1) removed aerosols /particulates. The release also was filtered before escaping to the environment. Additionally, as was discussed in Section 4.5.1.2.1, only a small fraction of the iodine eve,1ved (was released) from the coolant. All of these factors workeo together to result in a small release to the environment. The key was that the primary containment did not fall or leak in a major direct dry pathwa) to the environment. This experience indicates that leakage from process line pathwer should not provide a major source of release. If it were a major pathway, it would be con-sidered a bypass accident (see Section 4.5.2.1.4). Therefore, this pathway is not considered in the method developed here because it should not provide major offsite consequences. NUREG-1228 4-38
O LEGEND m GASEOUS RELEASE
- LIQUID RELEASE g CONTAINMENT G@ PLUME EPORV PRESSURIZER PLANT STACK h
. .., (
l= : g M ::::::,
- w r
RELIEF TANK TEMS SUMP PUMP v _ M'RIF CAtl N )
" p/ ~
~-
^^ gj ^ ^ ^ ^ ^ ^ - @ AuxtuARY BUttoiNo V
$ L 1 5' ' h If h kE h h h gh*h g 'g. g g
- 1. Early (first 38 min) before core damage, normal coolant only, sump pump operation.
- 2. Af ter core damage, leakage from makeup and purification system was the source of most of the release.
Figure 4.10 TMI accident radioisotope release pathways NUREG-1228 4-39
r 4.5.2.1.7 BWR Reactor Coolant System 'oreaks/ Leaks The BWR RCS (e.g., steam lines) extends beyond the primary containment. During a severe accident, these systems are isolated by valves that confine all reac-tor coolant to the primary containment. This section will assume that this isolation has taken place successfully and that the releast from the RCS break is into the BWR primary containment. There is another big difference between the PWR and BWR. In the event of a large break in the RCS (shown as pathway A-1 on Figures 4.7, 4.8, and 4.9), the steam blowdown from the break flows down into the dry well where it will be directed into and through a suppression pool (shown as pathway A-2 en Figures 4.7, 4.8, and 4.9). As discussed in Section 4.5.1.1.4, this pool can be very effec-tive in removing fission products. Therefore, if the RCS break is such that it does not fail the dry well before core damage--thus ensuring the major portion of the release is through the suppression pool--the reduction factors for the various pool conditions listed in Section 4.5.1.1.4 (Table 4.5) should be assumed. Obviously, if the drywell fails before major release from the fuel, allowing the suppression pool to be bypassed, no credit should be given for the suppression pool. 4.5.2.1.8 BWR Automatic Depressurization System and/or Safety Relief Valves BWRs are equipped with an automatic depressurization system (ADS) that is de-signed to reduce the pressure in the RCS in the ovent of an accident so that the low pressure emergency cooling system can inject water into the core to provide cooling. Automatic initiation of the ADS requires the following: (1) high dry well pressure, (2) Icw water level, (3) confirmation that a core spray train or residual heat removal (RHR) train is operable, and (4) a 120-second delay. The ADS is a subset of the safety relief valves (SRVs) that are designed to prevent overpressurization of the primary system. The SRVs/ ADS can open early in an event. If the systems designed to replace the water fail after the SRVs open, core damage could result. If the SRVs open following core damage, the products contained in the steam release to the suppression pool can be assumed to be scrubbk 'he SRVs discharge into the suppression pool (see Figures 4.7, 4.8, and 4.9, pathway A-3). The flow rate in the suppression pool is limited by the capacity of the SRVs, and the slow-flow condition should be assumed. Under these conditions, the RDF of the pool is a furetion of pool temperature; for example, a subcooled pool would assume an RDF of 0.01 and a saturated pool would provide an RDF of 0.05. 4.5.2.2 Primary Containment (Fourth Fission Product Barrier) In many PWR accidents the primary containment would be the last fission product barrier. Most BWRs and some PWRs have the equivalent of a secondary containment or confinement that encloses the primary containment structure. In some cases the leakage out of the containment may be into other structures such as the auxiliary building. These structures will be discussed later. NUREG-1228 4-40
l This section will discuss primary containment failures only; these failures are l shown as the B pathways on Figures 4.5, 4.E, 4.7, 4.8, and 4.9. The full range of containment failures will be considered. At the lower end of the consequence scale is leakage less than design limits, and at the upper end is catastrophic failure allowing direct releases to the atmosphere. The fission products suspended in the containment atmosphere, thereby available for release following containment failure, will be reduced with time by natural processes and by the actions of containment spray (if available). The RDFs specified in Table 4.5 should be assumed. There are a number of containment configurations. Containments are designed to withstand the pressure resulting from a depressurization of the RCS (blowdown). They also are used to contain any radioactive material released from the RCS. A large, dry PWR containment building is generally free standing and is designed to withstand the blowdown resulting from a failure of the RCS. These containments vary in size from about 1.5 to 3.5 Mft .3 Some of the smaller containments are maintained at pressures below atmospheric (subatmospheric containments). The other type of PWR containment is the ice condenser containment, which uses a large bed of ice to absorb the energy from an RCS blowdown. This ice bed can provide substantial reduction in fission product in the blowdown as discussed in Section 4.5.1.1.5. An important point is that a large radionuclide release does not require that ' the top of the containment be "blown off." Releases are insensitive to hole sizes larger than about 2 ft2. A hole of this size or larger would be considered catastrophic containment failure. For the large-LOCA sequence, the pressures in the containment would rise at a much greater rate than for the small-LOCA sequence. A larger hole size would allow for a rapid pressure decay in the containment; thus, the driving pressures would be substanticily reduced when fission products were released to the containment following initial blowdown. In summary, the consequence of an open containment penetration will vary, not necessarily with hole size but with accident sequence. 4.5.2.2.1 Design Leakage The design leakage release is shown in Figures 4.5, 4.6, 4.7, 4.8, and 4.9 as pathway B-1. Each plant is allowed leakage rates between 0.1 and 0.5% a day of containment volume at design containment pressure. For lower pressure, contain- : ment leak r % should be much less than design as was the case Curing the THI accident. i...,le 4.10 shows the EF that can be assumed for various containment failure / leakage cases. Under accident conditions, the leakage could be expected to increase with containment pressure. Some likely sources of containment leak-age are penetrations such as process piping or air locks. Containment penetra-tions and seals are not designed to withstand the environment conditions in the , containment following a core damage accident. Over an extended period of time (hours), the adverse conditions within the containment building may cause the l containment penetrations and seals to deteriorate, thereby allowing the leak : rate from the containment to increase. Large leakage is more appropriately treated as failure to isolate. l 1 NUREG-1228 4-41 I
4.5.2.2.2 Isolation Failures Because the containment building is the final Darrier to fission product release, pipes that penetrate the containment building are considered an extension of the containment boundary and must be capable of isolation. Systems penetrating the containment that serve a safety function, however, are not automatically isolated when containment isolation is initiated. A PWR and BWR release pathway attributed to a failure to isolate is shown as B-2 in Figures 4.5, 4.6, 4.7, 4.8, and 4.9. Containment isolation failures (failure to isolate one or more penetrations and failures involving the signal that actuates containment isolation) can provide release pathways for fission products retair.ed in water or suspended in the j atmosphere. Some of these pathways lead to other areas of the plant outside the containment building where the fission products may be held up temporarily or indefinitely; however, other pathways lead directly to the environment. There are a wide range of possible isolation valve failure sizes. This analysis uses the type of failure assumptions used in various consequence studies. Isolation failure typically refers to a failure of the isolation valves to go to their required (closed) position. In this assessment, it is assumed that an isolation failure is the result of the failure of the valve seals. In Table 4.10, it is assumed that a failure to isolate is equivalent to 100%/ day leakage. This approximates leakage of purge and vent system isolation valve seals. These valves are typically butterfly valves ranging in size from 20 to 40 in. in diam-eter. The metal-to-metal clearance between the valve disk and body is normally between 1/16 and 1/8 in. Therefore for a 40-in, diameter valve, a total seal failure would correspond to a 6-in.5 hole (NUREG-1037). In WASH-1400, the assumption was that a failure to isolate corresponded to 1000 times design leakage (100%/ day). This equals about an 8-in.2 hole. Therefore, tne 100%/ day assumption for the failure of the isolation valve seals appears to be reasonable. However, this does not characterize a major failure of an isolation system that results in a 1-ft2 hole or larger, which should be assumed to be a catastrophic failure. 4.5.2.2.3 Catastrophic Containment Failures A catastrophic containment failure is one that results in release of a large fraction of the fission products in the containment atmosphere in a short period (1-2 hours). The 1-hour EF for this type of failure is 1 (Table 4.10). This pathway is shown as B-3 in Figures 4.5, 4.6, 4.7, 4.8, and 4.9. This type of failure could be a very violent event. In such a containment failure, the fis-sion products in the containment atmosphere would be carried to the outside atmosphere along with the pressurized gases through the breach in the contain-ment shell. Some settling and plateout would be expected to prevent some of the fission product aerosols from being released. However, the turbulence and pressure reduction in containment could result in resuspension of some of the aerosols that had previously settled out, thus offsetting, somewhat, the pre-viously mentioned effects. A brief description of those accident cot itions that result in this type of failure will now be discussed. There shwld be indicators in the control room (e.g., containment hydrogen levels or pressure) that indicate conditions with NUREG-1228 4-42
i j the potent al to cause catastrophic containment failure. However, the actual timing or iven occurrence of catastrophic failures would be very difficult to predict du*ing a severe accident. This is further complicated because this type cf failure has little warning. The great uncertainties associated with containmer t response, given core damage, are shown in Figure 4.11. This figure is the la',est (NUREG-1150) estimate of the range of probability of early con-tainment failure. This figure also shows that early containment failure cannot be ru1PJ out, given core damage. 4.5.2.2.3.1 Hydrogen Detonation / Burns (NUREG/CR-2726) The hydrogen producing phenomenon that occurs during severe accidents is the reaction between the fuel cladding (Zircaloy) and steam (water). The extent to which hydrogen would be produced from cladding / steam reactions depends on the particular accident sequence, although any accident that results in severe core damage will generate substantial amounts of hydrogen. Zircaloy-steam reac-tions were the primary source of hydrogen generated during the TMi-2 accident. If the accident progresses to the point of whole or partial core melt, the molten core will slump downward, melting as much as 200,000 pounds of steel. When the molten steel comes into contar; with the water retained in the reactor vessel's lower plenum, the steel will oxidize and produce hydrogen. Any molten steel that enters the reactor cavity may oxidize and produce hydrogen. For conditions in which the fuel debris bed is uncoolable, melted core-concrete in-teractions will yield further hydrogen and the combustible gas carbon monoxide. The triangular diagram, Figure 4.12, shows the standard assumption for the relationship between air, hydrogen, and steam required for combustion. This diagram is based on low-temperature and pressure data. There is some experi-mental data indicating that at the high temperatures and pressures found in a containment under some accident conditions that Figure 4.12 is incorrect. These experiments indicate that much lower concentrations of hydroger may be required for burns and detonations under high pressure
- d " , % .es. How-ever, the following discussions will be based on Figure 4.12.
Hydrogen that accumulates in the containment cannot ignite until it reaches a concentration of at least 4% (with an oxygen concentration greater than 5%). However, the reaction between hydrogen and oxygen will not be complete unless the hydrogen concentration is above 8%. In addition, sufficiently high steam concentration can prevent hydrogen detonation. Therefore, under some accident conditions, actuation of the containment spray could reduce the steam concentra-tions in the containment resulting in conditions that could support combustion. Hydrogen ignition will produce a large flare in the containment that would be sustained for a time on the order of minutes, depending on the hydrogen concen-tration. An ignition source could be provided by sparks from electrical equip-ment in the containment. The early presence of an ignition source would allow , the hydrogen present in the containment to burn, potentially precluding hydrogen accumulation to detonable levels. Hydrogen detonation would require hydrogen concentrations in excess of 13%, l depending on the concentration of steam in the containment atmosphere. It should be noted that a hydrogen detonation, either global or local (as a result l i NUREG-1228 4-43
1.00
-WifM OIRECT wtTH tsNgn
_ HE ATING ugLT THmouGH w ~ g 0 90 - d = " NO LINE R $ g gg .
- . MELTTHROUGH ,
- 5 b NO OIRECT 3 z OR -
= HE AfiNo w 0 60 -
a - r - - E 8 _ E a , o so -
~
e = _ E ~ - Ox - - 5 3 - g a 0 30 -
=
~
a ~ ~ ~ ~ an 0 20 - as - - a s -
- I : :
~
E o.to .- 3 5 E
- _
6 g -
=
o -.- E s 1 , SURRY ZION SEQUOYAH PEACH BOTTOM GRAND GULF Figure 4.11 Comparison of early containment failure probabilitics Source: Draft NUREG-1150, Vol. 1 NUREG-1228 4-44
i i 100% AIR E % 4
/+ 5%.
s
# ASSUMED
/ \
DETONATION LIMITS .
#A 9 v i' gg
/\ /\j .
FLAMM/ JILI LIMITSf ' 4 ,--N p . . 100 % 80 60 40 20 100 % H2 PERCENT H2 STEAM FLAMMABILITY LIMITS (limits may be lower for high tempgrature and pressure conditions)
- 68'F 187'F at O psig C) . - 300* F O psig
(
- 300'F 100 pela Figure 4.12 Low pressure and -temperature flammability and detonation limits of hydrogen, air, and steam mixtures Source: NUREG/CR-2726 NUREG-1228 4-45
of stratification), will not necessarily fail containment. Generally a hydrogen detonatica would need to take place in a high pressure environment to be able to fail a containment. The probability of containment failure will depend on the ultimate pressure capacity of the containment and pressure as well as the shock wave characteristics of the detonation. Since hydrogen can be a threat to containment, there are provisions for its control and monitoring during accidents. Most PWR containments have recombiners or igniters that burn the hydrogen before dangerous amounts can accumulate. Most BWR containments are filled with nitro-gen gas so that the hydrogen cannot burn. However, the containments are not normally filled with nitrogen during the first month of operation. 4.5.2.2.3.2 Steam Explosions, Direct Heating, and Missiles In-vessel steam explosions have been postulated to cause the reactor vessel head to becoine detached with enough energy to penetrate the containment. Many recent studies performed on the potential for steam explosion-induced missiles have concluded that containment failures as a result of such explosions may be physically impossible and are considered highly unlikely. Steam explosion-induced missiles are mentioned here primarily because they have been consid> red in risk assessments (i.e., WASH-1400) and not because they are a likely contain-ment failure mode. In recent years a new concern has been raised. In certain reactor accidents, such as those initiated by station blackout or a small-break LOCA, degradation of the reactor core can take place while the reactor coolant system reinains pressurized. Left unmitigated, core melt will slump and collect at the bottom 3 of the reactor verst.l. After boiling off the remaining water in the vessel, the molten core materials will start attacking the bottom head of the reactor. When the bottom head of the reactor vessel is breached in such accidents, the core melt will be ejected under pressure. The ejected materials are likely to be dispersed out of the reactnr cavity into surrounding containment volumes as fine particles, quickly transferring thermal energy to the containment atmo-sphere. More importantly, metal contents of the ejected core debris, mostly zirconium C 1 steel, can react with oxygen and steam in the atmosphere to generate a 15rge quantity of chemical energy, heating and pressurizing the containment further. The hydrogen produced by this process results in burns or detonations that could result in very high loads on the containment. The term "direct containment heating" (DCH) is used in the present discussion to describe this complicated physical / chemical p.ocess. It has been postulated that this type of event could result in containment failure. Because the BWR automatic depressurization system (ADS) is considered very reliable and because of BWR geometry, DCH is usually considered only for PWRs; however, this is under review. For a further discussion of the current position on this issue, see Draft HUREG-1150, Vol. 2, Appendix J, Section 5. 4.5.2.2.3.3 Containment Overpressurization Containments are designed to accommodate the pressure resulting from a blowdown of the primary system. ilotever, if the primary system continues to release energy into the containment, operation of safety systems are required to remove heat and provent an overpressurization and possible failure of the containment. Therefore, failure of the safety system required to control containment pressure could result in c tastrophic failure. NUREG-1228 4-46
l j The response of a specific containment above design pressures to include point ' of failure cannot be predicted. However the technical basis f. estimating source,term, as part theof the NRC effort Containment to reassess Performance Work-ing Group (CPWG) was established. Thir, group concluded (NUREG-1037), for risk-dominant sequences, that the accident environment (pressure / temperature) inside the containment does not challenge its integrity in most cases until several hours after the reactor vessel failure. Major reasons for this conclusion are (1) the more detailed accident progression modeling following the reactor vessel failure and (2) industry-wide studies of containment capability pressures Containment capability pressures used by the CPWG are presented in Table 4.11. It is important to note from Table 4.11 that the estimated failure pressures are 2 to 3 times the design pressures. Table 4.11 Sample containment design pressures and failure pressures Allowable Total con- Estimated leak rate tainment Design failure volume free volume Pressure pressure Plant / type (%/ day) (103 ft3 ) (psig) (psig) Zion /PWR large, dry 0.1 2,600 47 134 Surry/PWR subatmospheric 0.1 1,800 45 119 Sequoyah/PWR ice condenser 0.25 1,200 12 50 Peach Bottom /BWR Mark I 0.5 280 62 117* Limerick /BWR Mark II 0.5 410 55 140 Grand Gulf /BWR Mark III 0.4 1,670 15 60
*The capability pressure predicted for Browns Ferry was used.
Source: NUREG-1037, 4.5.2.2.4 Containment Bypass As discussed in Section 4.5.2.1.4, there are accidents that can result in fail-ures that allow releases of fission products from the primary system so that they bypass the containment. None of the contairment reduction mechanisms can be assumed to affect this release. However, if the release is into tnother structure (e.g., auxiliary building), the reduction mechanism associ.ited with this path should be considered (e.g., holdup). 4.5.2.2.5 Controlled Venting of Containment As discussed in NUREG-1210, Vol. 3, the licensees have emergency opereting pro-cedures (EOPs) to be used by the control room staff to ensure that critical safety functions are maintained during severe accidents. In many cases these procedures call for venting of the containment as a last resort action to pre-vent catastrophic containment failure resulting from overpressurization. Some plants also may use venting to control hydrogen concentrations. 1 Plant conditions that would warrant venting would most likely also result in a highly contaminated containment atmosphere. Containment venting provides a , pathway to the atmosphere for any suspended fission products. l NUREG-1228 4-47
In BWRs venting is from the wet well (pathway B-5, Figures 4.7, 4.8, and 4.9) through the standby gas treatment system (pathway C-2, Figure 4.7). The RDFs for this pathway would be those associated with releases through the suppression pool (Section 4.5.2.1.7). The effectiveness of the standby gas treatment system was discussed in Section 4.5.1.1.3.1. Many plants have E0Ps that instruct the operators to consider venting before pressures in containment reach the ultimate capacity point. The major vent paths for a BWR Mark I containment are the wet-well and dry-well 18-in. vent and purge lines. There are smaller diameter lines (i.e., 2-in. and 6-in. lines) coming off the 18-in. headers that also can be used. The E0Ps will generally instruct the operators to use smaller vent paths first, to ensure that the re-lease path is no larger than necessary (NUREG/CR-4696). If core damage has occurred, suppression pool vent paths are to be used before dry well paths to take advantage of suppressioa pool scrubbing of fission products. Although provisions for venting are included in the operator's E0Ps, it is highly uncertain whether or not initiation of venting would be effective during a severe accident. Becausi only a short section of the vent path is schedule 40 steel pipe and the rest is duct work similar to that used in standard heat-ing and ventilating systens, the vent path may rupture under high pressure con-ditions. Because the vent and purge valves are containment isolation valves,
~
they will get a signal to close on high dry well pressure. To open these valves, technicians will have to go down to the cabinet ar.d jumperout the containment isolation signal. In the event of a station blackout, where there will be no power available to open the valves, the equipment operator will have to manually control the valve at its location. Opening the valve may require more than just turning a handwheel. For example, operators may have to connect a bottle of compressed gas to the valve operator with copper tubing and control the valve position by manual manipulation of the regulator on the gas bottle. Because of the extreme environment and high radiation levels that may be present (i.e., heat, "adiation, steam, etc.), it may be impossible for the operator to stay at the valve location very long, if at all. Venting under an anticipatec-transient-without-scram (ATWS) condition is a spe-cial case. Because of the large pressures that may exist in containment, it might be necessary to use all four 18-in. vent paths. Thus, most of the fission products released through venting would be unfiltered. PWR venting would be directly to the atmosphere with no filtering (filter fail-ure) and should be treated either as a failure to isolate (B-2, Figures 4.5 and i 4.6) or catastrophic failure (B-3, Figures 4.5 and 4.6) depending on the release rates. Thus, the only advantage of venting in a PWR is that the release may be controlled. For a further discussion of containment venting, consult Vol. 3 of NUREG-1150. i 4.5.2.3 Other Barriers BWR primary containments are enclosed in a secondary building (containment) that is designed to confine and filter leakage. This structure is not designed to withstand the pressure fiom a major failure of the primary containment. However, studies suggest that this secondary structure would reduce the release fraction ! by a factor of 2 to 4 (Denning and Cybulskis, 1986). NUREG-1228 4-48
Some PWRs also have an enclosure that acts to collect and filter any containment l 1eakage. In addition, any release pathway from the primary containment may re- , lease into other structures such as the auxiliary or turbine buildings. These l structures most likely would not withstand a major release, but could confine l 1eaks long enough for significant reduction to take place. In some cases, the structures also may have filtered vents. If the release is into a structure and it does fail, the fission products will be reduced by processes discussed in Section 4.5.1.1.2. If the structure has filter systems, the removal of fission products by the fil-ter will be in accordance with that discussed in Section 4.5.1.1.3. In the PWR (Figures 4.5 and 4.6) this is shown as C-3. If the structure fails or there are no filters, the release is shown as C-2. In the BWR there is a standby gas treatment system (SBGTS) as discussed in Sec-tion 4.5.1.1.3.1. The SBGTS takes suction in many areas in the plant basically as shown in Figures 4.7, 4.8, and 4.9. If the flow through a filter system (e.g., SBGTS) (C-2, Figure 4.7) is slow and dry, it can be expected to have an RDF of 0.01 (99% or greater efficiency). However, if the flow is heavily loaded with steam or aerosols or if it has a very high pressure flow rate, the filters would be expected to rupture. In general, this would be the assumption if there < were a major failure of the primary containment. 4.6 Plant Instrumer.tation The method for determining the source term associated with a given release re-quires proper interpretation of only a minimum set of radiation monitors and key plant parameters to characterize a limited set of accident conditions as they were described previously. However, under accident c mditions instru-ments can respond in unexpected and confusing ways. Radiacion monitors can be offscale, be responding to unanticipated sources of radiation, be bypassed, be isolated or not be calibrated for the conditions. Temperature pressure or other key plant parameters also may provide confusing or conflicting informa-tion because they are operating under severe conditions for which they are not designed. These difficulties must be recognized. However, if all the informa-tion is considered without focusing on a single parameter or instrument, the severity of the accident should be apparent. This can be seen from the response of selected area radiation monitors during the THI accident shown in Figure 4.14. While any specific monitor response may be confusing, all of the monitors are showing a factor of 100 or more increase in a few hours. Licensees have developed procedures to relate some instrumentation to plant con-ditions. Typically these procedures show the relationship between containment monitor readings, water level, or thermocouple readings to core conditions. These relationships must be used with caution (NUREG-1210, Vol. 2). 4.6.1 Radiation Monitors During a major accident, radiation monitors can provide valuable information on release pathway, m s release rates, and gross levels of core damage. Unfor-tunately, the TM1 a.cident demonstrated that these monitors also can be a source NUREG-1228 4-49
of confusion. Each of the three types of radiation monitors (process, contain-ment, and area), with the potential to provide the most useful information, con ~ fused the issue instead. Figure 4.13 shows the response of some of the process monitors during the TMI accident. The figure is clear, although the information during the accident was not. Figure 4.13 is a summary of 30 in. of tape from the recorder chart with some monitors not shown. Initially most of the monitors showed little response to the accident. As Figure 4.14 illustrates, several instruments, containment, incore instrument tank, and south refueling bridge monitors all responded to the high radiation released at the start of fuel fail-ure. The trend of the containment monitor shows two discrete steps of about a factor of 10 increase. The first step is at the start (6:25) of fuel failure (gap release) and the second at about 7:15. It is clear from the relative in-creases over 45 minutes that major releases from the core (and fuel dam >1e) were in progress. Figure 4.14 shows the area radiation ;:enitor response during the TMI accident; again, this is a summary of 30 in, of tape i.em the recorder chart. The opera- 1 tors had to glean this information from 30 in. of the tape from the recorder chart with twice the number of traces, all of which were printed poorly. The containment monitor also was a major source of puzzlement because the con-tainment dome monitor, located in a 2-in. lead shield, was yielding almost an identical dose rate as the unshielded south refueling bridge monitor. After considerable effort, it was determined that two errors were involved: (1) the wrong scale was on the recorder and (2) the monitor was not calibrated to l account for the shielding. This was compounded durin2 the actual event by the operator misreading the manitor. The monitor readings in millircentgen were re-ported as roentgen. Calculations made after the accident show that the true radiation level was 700 R/hr by 7:30. Derived relationships betwee. containment monitor levels and postulated core damage states may be key indicators of the level of threat in many accident
! sequences (but not for containment bypats accident sequences) Kecause of the basic accident scenario (release) assumptions used for the est.aates, monitor-ing efficiency, shielding, location in the containment, or other site-specific factors, there is a large variation in the plant-specific estimates associated with levels of core damage. Nevertheless, radiation monitor readings will in- '
crease by several orders of magnitude for progressively more severe core damage levels and the plant operator should be able to recognize the difference be-twfen a big and a little problem. For purposes of comparison, a 1- to 100-R/hr " reading should be expected following a large-break LOCA. Further increases following this spige should be considered as an indication of core damage. ! i The nuclear service cooling monitor response shows a second point of confusion I during the TMI accident. These raonitors are in the auxiliary building and are ; intended to detect leaks into nonradioactive systems. These monitors, like many other monitors, were not responding to their ' nameplate" source, but to other sources in the vicinity. In this case they were re:ponding to highly , radioactive gas in a vent header. If the operators had believed the "name- I plates" on the monitors, they would have concluded erroneously that radioactive material was leaking into the nuclear service water. I i 1 NUREG-1228 4-50
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i Finally, the way the radiation monitors fail during an accident can be very con-fusing. There were failures of radiation monitors at TMI and during other plant , events that left the monitors on scale. In fact, some monitors may fail and ' show mid-scale readings. In summary: The radiation monitoring system has been designed generally for normal cperation. In the event of an accident, many of the monitors would be isolated and not available to aid in the assessments. The area radiation monitoring system (ARMS) and exhaust monitor may provide limited information on the movement of material, pathway location, and gross level of problem. The containment monitoring system could be useful in determining the type of accident and aross degree of core damage for accidents involving releases into the containment, provided the pre-calculated relationships between monitor response and damage state are based on similar assumptions. As discussed earlier, the radiation monitors during the TMI accident were more a source of "puzzlement than of enlighteent" (Babcock & Wilcox, June 1981). Many of the monitors responded to sources of radiation that were not intended to be monitored (e.g., contaminated water being pumped through nearby lines). Therefore, many of the "nameplates" on the monitors did not Indicate what was actually being monitored. In addition, the fact that a monitor does not indicate a problem does not mean that it is not being bypassed.
.' Information from radiation monitors should be considered along with all other information to determine what is going on. If a monitor reading is not consis-tent with accident conditions, the following questions should be asked:
Is the monitor being influent.ed I,y some other source of radiation or other plant conditions? Is the flow / sample, etc., representative? Is the monitor shielded? What is the offect of different nuclide mixes? Hew was the instrument calibrated; what assumptions were used? Has the monitor failed? , l Is the monitor being read correctly? Finally, the release may be by a pathway that is not being monitored. 4.6.2 Grab Samples Under some circumstances, it may be desirable to take grab samples from selected locations within or surrounding the plant. Licensees are required to have pro-visions to sample and analyze effluent, primary cociant, and containment atmo-sphere samples within 3 hours. The capability to quantify radionuclides, hy-drogen levels, and boron levels is provided. Detection of hydrogen in samples l NUREG-1228 4-53
indicates core damage and may be useful in estimating the possibility of contain-ment failure. Analysis of boron concentrations is important to ensure the core will remain subcritical. Grab samples can be analyzed using gamma-ray spectros-copy and a multichannel analyzer to identify the various radioisotopes that were included in the sample. Identification of radioisotopes will help identify the degree of core damage. The largest drawback associated with a grab sample is that it frequently requires a considerable amount of time to collect and anal-yze and does not provide the prompt feedback that may be necessary for emer-gency action decision-making purposes. In addition, the gaseous samples may not be representative of the area because of plateout in sample lines. As discussed in Section 1.4.1, samples may be the only method to fully character-ize the release out of the effluent pathway. g Coolant samples taken before an accident are often used as the basis to calcu-late coolant release. The problems associated with this are discussed in Sec-tion 3.4. The actual concentration at the time of a release could be orders of magnitude off as a result of iodine and cesium spikes or further core damage. 4.6.3 Key Plant Parameters Radiation monitors can provide very valuable information about the location, pathway, and magnitude of a release. However, radiation monitors may not be accurate indicators of release trends. Changes in the release rate occur pri-marily as a result of changes in the accident scenario that affect fission pro-duct release mechanisms. Thus, these changes in the accident scenario would provide the most valuable information relative to fission product release trends. Changes in the accident scenario are determined primarily by monitoring and interpreting indic '"ons of key plant safety parameters. Although emphasis is placed on the key safety parameters, other parameters also may provide valuable information to the operator that would better enable the operator to correctly assess the progression of the accident. A brief discussion of some key plant parameters and their relationship to the accident follows. 4.6.3.1 Core Temperature 4.6.3.1.1 Pressurized-Water Reactor The PWR core exit thermocouples are typically located 10 inches or more above the top of the active core. They respond to the temperature of the coolant when the core is covered or to steam and hydrogen if the core is uncovered. Generally the thermocouples are considered the best and most direct means of assessing the degree of core damage. Some information relative to the accuracy of thermocouples during degraded acci-dent conditions has been gleaned from the THI-2 accident (see Figure 4.15). A temperature profile of the THI-2 core was manually recorded on March 28, 1979 between 8:00 and 9:00 a.m. (NUREG-0600) from the readings taken by instrument and control technicians directly off the wires that lead to the in-core thermo-couples. This action was taken when the thermocouple readings from the plant computer printout consisted of only question marks (????). Unbeknown to the operators, the plant computer was overburdened and underranged for the core temperatures that existed. As shown in Figure 4.15, very hipF temperatures NUREG-1228 4-54
l i 1 2 3 4 5 6 7 8 9 10 11 12 13 14 16 A A B 281 326 B C 469 967 325 C D 1196 500 217 D E 599 1926 2500 326 E F 80 2366 2378 323 566 F G 375 2272 1774 1806 1875 234 G H 260 2462 1296 2176 1862 H
- 4. .-
K 1811 632 1760 K r L 373 1566 382 296 1174 L M 326 2171 2327 348 252 M N 413 578 2167 N O 366 462 1138 308 O P 291 P i R 362 475 R ! 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 i Figure 4.15 THI-2 accident: map of core exit temperature 1 (*F), 8:00 a.m.-9:00 e..m., March 28, 1979 Source: NUREG-0600 , NUREG-1228 4-55 l
(i.e., greater than 2200*F) were recorded in se.*ral locations (e.g., E9, F7, M9, G5, and H5), whereas very low temperatures (i.e., less than 600*F) were mea-sured in other locations (H1, L6, and P6). The reasons for these discrepancies are complicated and beyond the scope of this report. It should suffice here to say that the in-core thermocouple readings were rejected by plant personnel dur-ing the TMI-2 incident, primarily because of the discrepancies that existti Following the TMI-2 experienceg it was concluded that thermocouple data can be relatively accurate below 1000 F and are most likely within 125F* of the sensed temperature. On the other hand, at temperatures above 2500 F, the data are more likely to be inaccurate. However, the important lesson from TMI-2 is that, even if major discrepancies exist, core thermocouplcs readings should not be ignored and may provide confirmation of fuel damage when considered with other indica-tions such as the radiation levels already discussed. 4.6.3.1.2 Boiling-Water Reactor The BWR temperature instrumentation is found in the recirculation loops and on the outside of the reactor vessel. During accidents, the flow in the recircula-tion loops may be interrupted and the instruments outside the vessel would take a very long time to respond. For these reasons, the instruments are of little value in determining core temperature (NUREG/CR-2726). 4.6.3.2 Reactor Coolant Temperature and Pressure l The pressure and temperature af the reactor coolant are measured in various locations within the reactor co01 ant boundary. Typically, the hot-leg (reactor vessel outlet) temperature is measured by several sampling probes located inside the hot-leg piping of each reactor coolant loop. Similarly, the cold-leg (reac-tor vessel inlet) temperature is measured by probes located inside the colit-leg piping of each reactor coolant loop. In a PWR, an average loop temperature is calculated by adding the hot-leg and cold-leg temperatures and dividing by two. Reactor enolant system pressures are measured in selected locations with pres-sure transducers. Together, reactor coolant pressure and temperature indications l can provide very valuable information relative to the thermodynamic state of i the reactor coolant. If the temperature in the reactor coolant system is greater l than the saturation temperature corresponding to the coolant pressure, then the I reactor coolant has turned into steam. An indication of the presence of super-heated steam is an indication of an uncovered core. 4.6.3.3 PWR Pressurizer Level l In a PWR, the pressure is maintained at a constant level by means of a pres-surizer. Under normal steady-state operation, the pressurizer is filled with about 60% water and 40% steam. Th;s mixture is maintained at saturation tem-perature by an electric immersion heater (located in the bottom of the pres-surizer vessel) and a spray system (located at the top of the pressurizer ves-sel). In addition, the spray system and electric immersion system control minor increases and decreases in system pressure, respectively. Large pressure in-creases that cannot be controlled by the pressurizer spray system are accommo-dated by relief valves that open when a predetermined set point is reachea. The steam from the pressurizer is vented through the relief valves to the pres-surizer relief tank. The pressurizer relief tank is vented to the containment if the design pressure is exceeded. NUREG-1228 4-56
1 Pressurizer-level readings provide an indication of the water level in the pres-surizei. They may not provide an accurate or reliable indication of reactor vessel water level. An important case in point is the TMI accident. Stuck-open relief valves led the TMI operators to believe that the primary coolant systere was full of water when, in fact, the core was in the process of becoming uncovered. 4.6.3.4 Reactor Vessel Water Level 4.6.3.4.1 BWR Water Level Indicator I For a BWR the water level indicator is often considered the best indicator of potential core damage. If the core is sufficiently uncovered (>2/3) for a period of time, fuel damage is expected. There are several reactor vessel level ; instrument ranges. The narrow- and wide-range instruments measure the reactor vessel water level in the annulus during normal operation. The shutdown-range ! instrument measures the reactor vessel level during shutdown and refueling. The shroud-range instrument, unlikre the other instrumentation, measures the water level within the core shroud and would be utilized primarily during accident conditions. One problem with old BWR designs is that a break into the dry well could cause the level reference leg 1o heat up. The result is that the instruments indicate a higher water level thari actually exists. New plants do not have this problem. However, are should be taken and reliance should not be placed solely on BWR water level to predict core damage (NUREG/CR-2726). 1 4.6.3.4.2 PWR Vessel Coolant level Detectors Before the TMI accident, reactor vessel coolant level detectors were not used with PWRs. Currently, some type of level detector within the reactor vessel must be installed at all LWRs. The pressurizer-level detector provides a diverse means of determining whether the reactor core is covered with watar. t 4.6.3.5 Engineered Safety Feature and Critical Safety Function Status Indication ! The operability of engineered safety features (ESFs) such as the emergency core ; cooling system (ECCS) and the residual heat removal system (RHRS) may be very , i important following the initiation of an accident. For example, during the ! TMI accident when a relief valve stuck in an open position and consequently [ caused the reactor coolant pressure to drop te 1600 psig, the high pressure ; safety injection system (part of the emergency core cooling system) was auto- l matica11y initiated. Had the operators not hypassed this system (i.e., placed [ the system on manual control) and throttled flow to the reactor, the core would i < have remained cooled. For LOCAs like that at THI, the unavailabi'aity of an , ESF may have a substantial effect on the magnitude of the release. The control ! ) panels of the ESFs are kept separate from other controls in the control room, l The operability of the ESFs can be verified from this panel. l Since the TMI accident, the procedures used by the control room staff to respond i to an accident have been revised to concentrate on maintaining a set of critical ! safety functions. The control room staff has procedures for monitoring the status of these critical safety functions. These functions have a direct j i l ! i ) NUREG-1228 4-57 l _ . _ _ _ . _ , - . - _ . - _. _ _ _ , .._,y _ - - , _ . . - . . . _ . . . . _ _ - _ , __,
relationship to maintaining fission product barriers as shown in Figure 4.16. These barriers can fail only if the appropriate critical safety function is lost. Obviously, critical safety function status can provide good indications of fission product barrier status. This is discussed further in NUREG-1210, Vol. 2, 3, and 4. 4.6.3.6 Containment Isolation Status As discussed previously, two primary reasons exist why containment may fail to isolate. The first involves a failure of the containment isolation signal that precludes initiation of isolation valve closure. This type of failure, which would allow all the containment penetrations to remain open, constitutes the largest release path attributable to containment isolation failures. The second involves the failure of one or more penetrations to close as a result of valve or instrument failure. The consequences of containment isolation failures of the second type would depend on the size and the nature of the line penetrating the containment. For example, failure to isolate the containment ventilation exhaust duct would lead to much greater consequences than failure to isolate a demineralized water return line. The problem is that in many cases it will be difficult to determine directly if complete isolation has taken place. Failure to isolate would be indicated by unexpected temperature, pressure, or radiation outside the containment. 4.6.3.7 Containment Atmosphere Temperature and Pressure Containment atmosphere temperature and pressure readings would be very valuable in determining the accident progression path. During the THI accident, in-creases in containment building pressure and temperature were indicators that the pressurizer relief tank rupture disks had blown, thus allowing primary cool-ant to spill onto the containment building floor. Later during the accident, a large pressure spike of 28 psig was recorded following ignition of the accumu-lated hydrogen gas in the containment. This was a clear indication that there must have been extensive core damage. In general, abnormally high pressure or temperature readings in the containment would indicate a loss of reactor coolant boundary integrity. However, care must be taken in assuming that containment pressures near design limits indicate imminent containment failure. Pressures 2 to 3 times design raay be required for failure. NUREG-1228 4-58 _ _ _ _ _ _ _ . )
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l 1 5 SOURCE TERM DETERMINATION 5.1 Objective To enable the reader to [ 1
- work through a source term calculation using the approach discussed in i this section and the information provided in the previous section i l
5.2 Introduction The first step in determining the source term 's to establish the origin of the release or the potential release and its charicteristic. This determination is made by utilizing the information from plant parameters, radiation monitors, or samples. The plant parameters are most helpful in determining the extent of damage to the reactor core, whereas the radiation monitors are most helpful in locating the origin of the radioactivity and later in tracking the release from i the origin through the plant and out into the environment. Enough information - may be available to allow an estimate to be made of the amount of fission pro-ducts for release from the containment atmos)here. If this is the case, this estimate should form the starting point of tie analysis. The pathway removal mechanism and release rates from this point to the environment would be analyzed. If this information is not available, an analysis starting with the core would be required. The second step is to estimate the fission product inventory avail- < able for release based on the core temperature or other indications .of the ap-propriate source regime. This initial inventory can be determined based on methods described in Section 4. The next step is to trace the source back through the plant to account for all pathways and to apply the appropriate fis-sion product reduction factors. The basic method for estimating source term is ' shown by the "ollowing: , Source Termg = FPI 4 x Power Level (We) x CRFjx RDF(jg)xEFj for radionuclide i and n reduction mechanisms ! where FP!j = eiement i core or coolant inventory (Ci/We) (Table 2.2 or 3.2) , CRF ! i = element element 1 iinventory released from core (Table 4.1) in core ! RDF g _ element i available for release after reduction mechanisms (Table 4.7 element i available for release before reduction mechanisms or4.8) i EF i = element element 1i released to the available for environment release $ ' Table 4.10) I The referenced tables provide a summary of reasonable assumptions discussed in previous sections. As discussed in Section 4.2. a maximum total reduction NUREG-1228 5-1
factor of 0.001 should be assumed for the product of all non-filter reduction mechanisms. Following is an example of a release, or a potential release of radioactive materials to the environment. ThestepsinvolvedinestimatIngthesourceters are described. 5.3 Sample Source Term Calculation A severe flood caused loss of offsite power at the Perkins Point Nuclear Power Station. Perkins Point has an electric capacity of 1000 MW and has been operat-ing continuously for 6 months. Perkins Point, which is a PWR with an ice con-was first placed on line 3 years ago. Diesel generators A denser and B, thecontainment,s station source of onsite ac power, failed to start automat l when offsite ac power was lost. Subsequent attempts by the operators to start j the diesel generators proved unsuccessful because the flood water entered the : diesel generator room and disabled the vital buses. Additionally, the turbine-l driven feedwater pump failed to supply emergency feedwater to the steam genera- ! tors because a flow controller malfunction prohibited heat removal from the primary system. About an hour into the incident, while the operators were still attempting to restore ac power, the area radiation alarm in the cell housing reactor coolant pump A alarmed. About 20 minutes later, the area radiation monitors in the cells housing reactor coolant pumps B and C triggered an alarm. The operators observed that the reactor coolant system pressure was decreasing quickly despite the fact that the heat sink had not been regained. The containment building . pressure was rising rapidly, i The operators postulated that the power-operated relief valves (PORVs) stuck open--as at THI-2--despite the fact that the relief valve solenoid status lights indicated that the relief valves were closed and that the pressurizer level indic1 tion had bottomed out. These are indications the block valves have failed. The containment dome monitor reached 1000 R/hr and was rising rapidly. There were other numerous radiation alarm Thermocouple readings of greater than 1500'F were recorded at some locat* as. 5.3.1 Step 1: Gather / Assess Plant Information From the information given, the decreasing system pressure and increasing con-tainment building pressure indicate that a LOCA has occurred. Furthermore, the radiat.on alarms sounding in the containment should have been a good indication j that the primary system integrity had been compromised. Because the area radiation monitors were within the containment, this should . have told the operators that a release to the containment had occurred. The ' dome monitor in the containment was reading 1000 R/hr. Does this mean that fuel deJage of noble hasgasesoccurred, frce ruptured or could the hig?h cladding Onradiation the basislevel correspond of available to a release information, i ', it is reasonable to assume that the plant cannot maintain the ability to remove l decay heat or keep the core covered. < As discussed in Section 4.4.2, once the core is uncovered, cladding failure fol-lows in about 15 minutes. Cladding failure is confirmed by the containment monitor reading of 1000 R/hr and the thermocouple readings. NUREG-1228 5-2
.. -. --_ . . = . .-
1 b l
~
'? Step 2: Estimate the Fission Product Inventory Released From the Core t Although the extent of core damage could not be ascertained from the informa-I tion given, it is certain that a core melt release from the fuel would take place unless plant conditions improve. Thus, utilizing the assumed core melt release fractions frem the core that were provided in Tables 4.1 and 2.5 and ,
assuming a 1000-MWe operating history, the following calculations can be made: i l Core
- Inventory, Power level release Curies released Isotope C1/MWe x (1000 MWe) x fraction = from core Kr-88 6.8x104 x 1x103 x 1.0 = 6.8x107
!' I-131 8.5x104 x 1x103 x 1.0 = 8.5x107 Cs-134 7x103 x 1x103 x 1.0 = 7x10e ;
l 3 .
; 5.3.3 Step 3: Estimate the Fission Products Available for Release t i ; The obvious release path is into the containment building. However, small j ! branch lines attached to the reactor coolant system may penetrate the contain- ;
ment and connect to support or safety systems in the auxiliary building or tur- 1
; bine building. ;
c I
! The following are typical important questions:
Are there indications of containment failures?
- What is the containment pressure now and what will it be an hour from now? !
l How does the measured or anticipated pressure compare to the overall design pressure and to the ultimate capacity? t I + What is the hydrogen concentration? ( i i
- Will operation of containment spray cause the hydrogen concentration to ,
j approach detonable limits? ! i t i - Are the sprays operational? I - Are the recirculation fans operational? l
- l j - Has the ice bed been exhausted before the core damage?
l
! The answers to many of these questions will be unknown. Accurate indication of I plant parameters and knowledge of accident phenomena in containment may be the 1 only means for making a decision regarding the fission products available for i release from the containment and the containment leak rates. It will be assumed i for this example that the containment integrity is maintained and leaks are at
] design rates. It also is assumed that the fans and sprays are no'c available i .i NUREG-1228 5-3 1 I
- - - _ - - - - _ , _ _ _ _ _ _ . - - _ - - - _ _ - - - - _ _.. . _ _ _ _ _ _ - - _ - - _ _ _ _ _ _ _ _ ____ __~ __
I as a result of the loss of ac power and that the ice was not exhausted. Now it < will be necessary to estimate the inventory of fission products in the contain-ment atmosphere that could be released if containment failure did occur. The release path is illustrated in Figure 5.1. It will be assumed that all the fission pr: ducts released from the core are released from the primary system. i Continuing with the assumption that the air return fans are unavailable as a ' result of a loss of ac power, the fission product aerosols released are assumca to make only one pass through the ice condenser bed. Thus, based on Table 4.7, about 50% (RDF of 0.5) of the fission product aerosols may be assumed to be removed by the ice beds. Noble gases are not filtered. The following estimate ! can be made of the inventory of fission products in the containment. ' i l I r Curies released RDF Inventory assumed ; Isotope into containment x (ice bed) = airburne ' Kr-88 6.8x107 x 1. 0 = 6.8x107 i 1-131 8.5x107 x 0.50 = 4.2x107 ,
, Cs-134 7.5x108 x 0.50 = 3.7x108 i !
After several hours these inventories will decrease, as discussed in Sec- { tion 4.5.1.1.2.1, by the factors shown in Table 4.7. l To estimate the release 2 hours after the material has been released into the , containment, a 2-hour holdup time with no spray and only natural processes ' working to remove airborne fission products in the containment will be assumed (RDF of 0.04, Table 4.7), i 1 Inventory assumed Curies airborne RDF airborne and available following passa.ge (natural for release after 2-hr Isotope through ice x process) = holdup i' Kr-88 6.8x107 x 1.0 = 6.8x107 I-131 4.2x107 x 4x10 2 = 1,7xios 3.7x10s Cs-134 x 4x10 2 = 1.4x105 J 5.3.4 Step 4: Estimate 1-Hour Release 1 j The final step is to calculate the fission prodJcts actually released from the containment. Table 4.10, which summarizes typical escape fractions (EFs) re-
- leased in 1 hour, shows that in 1 hour 1x10 4 of a PWR ice condenser containment volume would be released at the design leak rate. Remember this is the upper limit; pressures below design limits would produce much lower leak rates. Thus
l l NUREG-1228 5-4 i
= 88 J- ?
w b >
$s.
CONTAINMENT UXILIARY FEEDWATER TURBINE EXHAUST STEAM h RECIRCUL 7710N t
/ , STEAM LINE 9 ISOLATION OC-2 k,
( VALVE L n j[ PORV: E p , >< 8 ' RELIEF "P GE RA OR
- y ICE l CONDENSER
-Q asar r' - -
s LOW
- PRESSURE SYSTEM \V ,
8 d TURBINE BUILDING _ AUXILIARY BUILDING h RELsEF VALVE 3 0 SPRAY l CLOSED VALVE .g FILTER Zi CHECK VALVE % RELEASE PATHWAY REFERENCE h PUMP l-l ISOLATION VALVE Figure 5.1 Release pathway for sample calculation case
1 ? Curies airborne 1-hr escape in containment fractions Curies released available for x (design = from containment Isotope release leak rate) in 1 hr i Kr-88 6.8x107 x 1x10 4 = 6.8x108 i 1-131 1.7x108 x 1x10 4 = 1.7x102 Cs-134 1.4x105 x 1x10 4 = 1.4x101 5.4 Use of Event Trees To Estimate Release Using the above method of calculation for a long list of isotopes would be very
, time consuming and still would not provide a direct estimate of possible offsite consequences. However, calculation of the amount of each isotope is not re-quired to estimate offsite consequences (dose). A set of event trees has been developed that provides estimates of the whole-body and thyroid doses at 1 mi',e.
Event trees have been calculated and are grou,Jed in the following figures by l release pathway type with the letters designating the level of core damage: . l (1) PWR large dry or subatmospheric containments (Figures 5.2A, 28, 2C)
< (2) PWR ice condenser centainment (Figures 5.3A, 3B, 3C)
(3) PWR steam generator tube rupture (Figures 5.4A, 48, 40, 4D) (4) BWR containment, dry-well leak / failure (Figurer 5.5A, SB, SC) (5) BWR containment, wet-well leak / failure (Figures 5.6A, 6B, 60) (6) BWR/PWR bypass (Figures 5.7A, 78, 7C) In each, the noble gas and particulate release fractions have been calculated i for gap, grain boundary, and molten-core damage states. The release fraction in the trees is the fraction of the material released from the core that is released i to the atmosphere. For example, to calculate the curies of I-131 released to the atmosphere you must multiply the curios released from the core times the r i release fraction shown on the trees. For the steam generator tube rupture, i normal coolant and coolant with 100 times iodine spike also were considered. P Doses were calculated for a 1000-MWe plant, and the release fractions were cal-culated using the reduction fac'or summarized in Tables 4.7 and 4.8, as appro-priate for the release pathway, fransport was estimated assuming a ground level j release and average meteorological conditions (4-sph wind speed and D stability). i The conditions provide dose estimates that are within a factor of 10 of the range of reasonable meteorological condition given a specific source term. Doses , were calculated using MESORAD computer code (NUREG/CR-4000), which is the dose ; assessment code used at the NRC Operations Center. The whole-body doses include '
- cloud shine, inhalation dose, and 3 hours of ground shine. The thyroid dose is for an adult and is for inhalation only. Oose factors for the release of 100%
of gap, grain boundary, and core melt particulate release fractions and 100% M i the noble gases were calculated and are shown in Table 5.1, L'e dose for a specific accident can be estimated by: l 1 j NUREG-1228 5-6
hcido release release s%sle-tody thyrc.d c core contaisent tame se containeest fraction fraction icse treet dose tres) M o toncition cesdattoas contaa r.sent Int ra+e nos-es)1es ectle gas at 1 mile at 1 elle a [ N 20.5 hour - Itr0Utour
- : 4.00E41 1.0M+00 4.24E+01 4.0M+03
- :- 10nthour spay : : 4.00E-02 1.0M40 4.6M+00 4.0M+02
- - - c41 :2-12 hours- :- 10CUtay
- : : 1.60E-03 4.0Cf-02 1.64E-01 1.6M41
- : :- desigs rate
. : 1.60E-06 4.00E-05 1.E4E44 1.6M-02
. : :- 100Uhour
- : : 1.00E-02 1. M +00 1.21E40 1.0M42 l24 hours- l - IDCUtay
- : 4.00E-04 4.0M-02 4.E4E-02 4.0M+00 Er.P RELUSE :- desige rate FPM EcRE 4.00E-07 w 4.00E-05 4.84E-05 4.ori-03 s
M .
- :0.5 hour 100Uhour
- : 3.00E-02 1.0M40 5.4M+00 3.0M +02
- : :- 100Uhour
- spray : : 2.00E-02 1.00E4J 2.60E+De 2.00E+02
- on :2-12 tours l- 100Ugay
- : 8.00E-Os 4.0M-02 1.04E-01 E.0M +00
- - design rate
- B.00E-07 4.COE-05 1.04E-04 E.0M-03
- - 100Uhour
- : 2.00E-03 1.00E40 4.10E-01 2.00E*01 24 hours :- 100Uday
- 8.00E-05 4.00E-02 1.64E-02 B.0M-01 l- destgs rate s.00E-08 4.0M-05 1.64E-05 8.00E-04 Figure 5.2A PWR large, dry or subatmospheric containment release event tree for a gap release from the core
. 7
Pcidup release release shale-tody thyrsti fraction dese tree)
$ tcer contateerat tree 's containeest fraction dese (reel at 1 a Ie at 1 elle y :endition condittees contalmeest leat rate non-selles netle las Cn E l0.5 hour 100Uhes B.4X+02 E.00E+04 1.00E40 M =
- 4.00E-01
- - 100U bour spray : : 4.00E-02 1.00E+00 1.DM42 8.00E43 l- of f l2-12 hours - 100Uday
- : . 1.60E-03 4.0X-02 4.00E+00 3.74 42
- : :- desip rate I : : 1.6M-06 4.0M-05 4.00E-03 L2M-01
- - 100Ubour
- : : * .0M-02 1.00E+00 2.3Mel 2.0M+03
- :24 hours - 100Uday l
- : 4.00{ 4 4 4.0M-02 9.4M-01 B.00E41 Pals SQLa*/JY RELEASE :- destp rate F W CCSE 4.00E-07 *.,00E-c5 9.40E44 8.00E-02 e :
W .
- l0.5 hour 100Unour
- 1.0M+02 6.00E43
' 00E-02 1.00E+00
- : - 100Uhour
- spray : : 2.00E-02 1.0X40 6.00E41 4.00E43
- a :2-12 hours l- 100Uday
- : B.00i44 4.00E-02 2.4X40 1.6M+C2
- :- desip rate
- B.ME-07 4.00E-05 2.4M-03 1.60E-01
- :- 1MUhour
- : 2.00E43 t.0X40 7.5M+00 4.0M42
- 24 hours :- 100Uday
- B.00E-05 4.00E-42 3.00E-01 1.6M+01 l- dessp ate s.00E-Os 4.00E-05 3.0cE-64 1.6M-02 Figure 5.28 PWR large, dry or subatmospheric containment release event tree for a grain boundary release from the core
l i. holiz:p release release sAele-tof y thre d c core contaissent taee in contasement fraction frattles esse treel dose tree) M testation cond:tsons cztatteret leak rate ecs-*=tles solle gas at I a21e at 1 elle O e
- w l0,5 hour 10C1/ hour y
- 4.00E-01 1.00E+00 2.08E+C3 2.00E+05 w .
l l l :- IXUScar i spray : : 4.0CE-02 1.00E+". 2.2CE+S2 2.00E+C4 ;
- - of f - :2-12 tears *- 100Uday [
- : : 1. RE-03 4.00E-02 8.RE40 8.00E+02 -
- dest p rate
- : 1.6cE-06 4.M45 8.SeE-03 8.0Ci-01
- : :- 10CUhcar ,
- : : 1.00E-02 1.00E40 5.70E+01 5.0Cf43
- :24 hours :- 100Uday
- : 4.00E44 4.00E-02 2.2ti+00 2.DC(+02 rn i RELE E : - destte rate 1 l u, FR:M ETE 4.00E-07 4.00E-05 2.!IE-03 2.00E-01 l l J, : !
- :0.5 hour 100Uhour
- : 3.00E-02 1.00E40 2.30E+02 1.50E44 ,
- : :- 100Unser
- spr y : 2.00E-02 1.00E40 1.20E42 1.00E44 ;
- e-. :2-12 hours :- 100Usay '
j : : 8.00E-04 4.00E-02 4. tee +00 4.00E+02
- :- desip rate l : s.00E-07 4.00E-05 4.00E-03 4.00E41 i :
- :- 100Ubour '
1 : 2.00E-03 1.00E40 1.70E41 1.00E*03 ?
- 24 heers-- - 100Uday
- B.00E-05 4.00E-02 6.SeE-01 4.eeE+01
- - seu p rate s.00E-Os 4.0eE-e5 6.seE-04 4.00E-02
, Figure 5.2C PWR large, dry or s14 atmospheric containment release event tree for a melt release from the core - - _ --- -,.--.-.--.._--.,.,-.,,-..-,----n. - , - . - - - - - . . . , , - , . . , - .. . . - - _- . , , , . , - . ,_ . _ _ _ _ _ _ _
-. - - . . .- . . - . - . .. ~~- -- - - . - - - - - .- .- -
E m release shele-body thyrest release shole-tesy thyread E holdes fracties release deee treal done treal fractsen deer treal deae treet d rentatneset time is contaissent ase-naties fractise at I eile at i site sea-nable at I este at I sale y conditises costasseent leek rate esce thre ice ookle tas once thre ice este thre ace estt recir sitt recir eith recir ce
- 9.5 tour 1stuhour
- 2. tee-01 1.00E46 2.24E41 2.00E+ct 1. tee 41 1.24E41 1.00E43
- - 200Uhour sprays : : 2. tee-82 1. tee 40 2.64E49 2. tee 42 I. tee-91 1.6eE40 1. tee +02
----off :2-12 hours :- 199Uday
- : 8.%E44 4. tee-02 1.0eE41 S.00E+49 4. tee-44 6.40E-42 4. tee *00
! :- desty rate 2.0eE-M 1. tee 44 2.NE44 2. tee 42 1. tee-06 1.6aE-te 1. tee-42
- :- 190Uheur
- : 5.0K43 1.00E+49 7.1K41 5. tee +01 2.5K-43 4.68E-91 2.5eE41
- 24 hours l- It0Utay ue : 2.00E44 4.0eE-02 2. tee-42 2. tee +00 1. tee-04 1. tee-92 1. tee *00 O GAP PELEASE :- desige rate l
FPtw EDPE 5. tee-07 1.00E-04 7.1K-05 5.tet-93 2.50E-47 4.68E-95 2.5eE-43
- 0.5 beer 200Uheur
- 1.5tE42 1.00E+00 3.90E40 1.50E42 7.5eE43 3.15E49 7.5eE41
- :- '90Unser sprays : : 1.00E-02 1.00E40 1.6tE+00 1.00E42 5. tee 43 1.1K40 5.00E*41
;. *2-12 hours :- 100Utay I : 4.0K-04 4.00E-02 6.4K42 4.00E+00 2.00E44 4.4E-02 2.00E40
- - design rate
- 1.00E-M 1.00E-04 1.6tE-04 1. tee-02 5. tee-97 1.!K-94 5.oet-43
- :- 100Unsur
- 1.00E43 1.0K+00 3.!K41 1.00E+01 1.etE-03 3.1K-01 1.00E41
- 24 6.urs : - 10cUtay
- 4.0K-05 4.00E-02 1.24E-02 4.0t[-01 4. tee-95 1.24E-42 4.00E-01 l- Jesip rate 1.0K47 1.00E-04 3.1K-05 1.00E-03 1.00E-47 3.1K-95 1.00E43 Figure 5.3A PWR ice candenser containetent rclease event tree for a gap release freet the core
b stoleiody 2 release innsti release ahole-tedy thyrea d c) holts; fractros releaw tese tree) dose (reel fraction dose treet dose treel O containeret itse as containment non-notles fraction at 1 sale at 1 male ses-estle at 1 mile at 1 male y conettices contasseret teak rate once thre ice noble gas once thre see once thrs see alth rettr esth rectr ettt rectr co
- 0.5 hour 10CUhour
! 2.00E-01 1.00E+00 4.43E*02 4.00E+04 1.00E-01 2.40E+02 2.00E44
- :- 100Uhour sprays ! 2.00E-42 1.00E+00 5.00E+0! 4.00E+03 1.etE-t2 3. tee +01 2.00E*03 ef- :2-12 hours : - 100Utay
- : 8.00E44 4.00E-02 2.00E+00 1.60E+02 4.00E44 1.20E+00 B.00E+01
- - tessge rate
- 2.00E46 1.00E44 5.00E42 4.00E-01 1.00E-06 3.00E-93 2.00E-0!
- :- ICCUhour
- : 5.00E-03 1.00E+s 00E+03 2.5eE-43 B.50E+00 5.00E42
- 24 hours :- 10cuday m : 2.00E-04 4.00E42 4-61 4.00E41 1.00E-04 3.40E-41 2.00E41 O
D A!u BOUN'. arf PELEASE :- sesage rate FPP ELTE 5.00E47 1.00E44 1.35E-03 1.00E-St 2.5eE-87 8.54E-64 5.00E-42
- 0.5 hour 100Uheur
- 1.50E-02 1.00E+00 7.00E41 3.00E43 7.50E-03 5.50E+41 1.5tE43 !
- :- 200Unsur sprays : : 1.00E-02 1.00E40 3.00E41 2.00E43 5. tee-43 2. tee 41 1. tee 43
*2-12 tourt
- :- It0Uday
- 4.90E-04 4.00E-02 1.20E40 8. tee +01 2.00E44 B.00E-91 4.00E*01
- :- desip rate
- 1.00E46 1.00E-04 3.00E-43 2.00E-01 5.00E-47 2.00E-43 1.00E41
- :- 190Uhour
- : 1.00E-03 1.00E*00 5.5eE40 2. tee 42 1. tee-43 5.50E40 2.00E+42
- 24 me rs :- 1eousay
- 4.osE-e5 4.seE-o2 2.2eE-91 B.00E40 4. tee-45 2.2eE-et 5. tee 46
- - e ip rate 1.00E47 1.00E-94 5.50E-te 2. tee-02 1.05E47 5.5eE-44 2.90E42 Figure 5.38 PWR ice condenser containment release event tree for a grain boundary release from the core
_ _ - . - . . - . . _ _ . . . . - .._ _ - - - . - ~ - .- .- . . . . .w-. - . - _ _ _ . - _ _ - - . - . - - _ _ _ _ ., , . - . - 4 3: t C
~
JD release seale-tesy tayrsae releene abole-hedy thyreta c) heldup fractane release dose tree) esse treel fracttee sees troal esee treel e '. containeret tier is containeset een-mobles fractine at 1 male at 1 este see-as:Je at 1 mile at 1 este N m comettaess contateerat test rate once thre ice noble gas sece thre ste esce ttre ace sitt recir eith recir eith recir os
- *e.3 hour 100Uhour
- 2.00E-01 1.00E40 1.00F+03 1. tee 45 1. tee 41 LIeE42 3.etE44
- : - to0 U no.r sprare : : 2.00E42 1. tee +00 1.20E42 1.e0E44 1. tee 42 7. tee 41 3. tee 43
----off 12-12 h vis :- 100Usay
- : E.00E44 4.00E42 4.90E40 4. tee +22 4. tee-te 2. bee +00 2. tee +02
- :- ersige rate
- 2.00E-M 1.00E-04 1.2tE12 1.90E+00 1. tee-M 7. tee-03 3. tee-41 i
1
- :- IOGUhour
- 100E43 1. tee +te ? 20E41 2.30E43 2.SeE-43 1.TSE41 1.23E43
*24 hours :- 100Usay y : 2.00E44 4. tee 42 1.2E+tt 1. tee 42 1.tK44 7.NE-41 LtK41 l w ELT SELEASE :- desige rate FA3M CDRE 100E47 1.00E44 3.2e[-83 2. Set-t! 2.3E-07 1.T1-03 1.2M41 l0.1 hour- 19tuhour 1.SeE42 1. tee +M 1.35E42 f.SeE43 7.30E-03 1.17E42 3.75E43
! :- 199Unser sprays : : 1.ceE-62 1. tee 40 7.00E41 3.00E+03 190E-43 4.30E41 2.30E43
- en :2-12 hours :- It0U6sy 4.00E44 4.00E42 2.00E40 2.00E+02 2.90E-te 1.00E+4e 1. tee 42
- :- desige rate
- 1.00E-M 1.00E-94 7.00E-43 100E41 3.0K47 4.30E-43 2.30E-41 I
- *- It0Uhour
, : 1 1.00E-03 1.00E40 1.2M+01 100E+02 ' 1.00E-93 1.20E41 3.0M42
- 4 hours - 100Usay
- 4.00E-05 4.00E-02 4.00E41 2. tee +01 4. tee 45 4. tee 41 2.00E+01 l--- desage rate i 1.00E-07 1.00E44 1.2K43 100E42 1.tK-07 1.2eE-03 10eE42 Figure 5.3C PWR ice condenser containment release event tree for a melt release from the core
l , l l I , i tne release deleiety thyreld i toelaat leet life lfettien deSe treal do6e freel ' conhtloal lleetratel lltenditions seese6tes at i elle et I alle I
- l*~ solid U tde I tone f alls at i 1.7M41 1.7M 01 1.7M*M r high presure I non no6te Spite (M0 spel !
IH : sw ul ----- * -- l~-- awul Of M i i 1.7M41 1.?M 01 1.7M'M i i I l- normal U td, I 7.M( 03 7.00( 4 3 7.00( 4 2 -
; I !
l i I , I l- ulid U tube i i 1.WC2 1. W 02 1.W68 i Itharglat peep i { lilee LH geel i L j l............ .l sw ul 0156 l 1.Ht42 1.30( 4 2 1.W01 ; i
! k W AL C00 A41 1-~ norulUide i nel IMI spilt 6.ME-H 4.00C44 i 6. W O3 l I i !
t ] la- 141..Utone l tibet f ails a l I.7M41 1. 7M-03 1.7M42 + i high pressure I i 1 norul (M4 3p1 I ] conc entr ation - --- - - -----~~ t - sw ul CIM l e l i 1.7M41 8.7M 03 1.7)(42 {
!' ! I
} I l- awul U ide ( i 1 7.0M-03 7.0M41 7.WH l l 1 '
, I i ; I l - utid U tabe i 1 1. W C2 1.50( 4 4 1. W O3 (
2 ithargingpeep I Iflee (M gpel i I
- l. .... .. ...l..~ not ul Oill i
i {
- 1. W C2 1.50(-M 1. W 63 J
)j i 1 1-- norul Utde I 6.M( H 6.0M45 6.W45 ! Figure 5.4A A PWR steam generator tube rupture event tree for a I release of normal and 100X spike coolant J i
; NURF.G-1228 5-13 1
E A
? -
0
~
00 7 ~ l l l l l T steenjet j y safety valve safety valve air ejector f releew releew shale-body thyread shele-body coolant tube fractice fractism dose treet dese treet gone tres) condattens leak sire SE condatsons non-nob!rs solle gas at I sale at 1 este at 1 eile 1 tote f a21s at high pressre *- solid 5 tote EAP (500 spe) : 1.73E41 3.50E41 1.8X41 1. M +03 8.40E-41 CDu2minATIDE : l- noraal GTSE
- : 1.75E-01 3.5CE-01 1.51 41 1.75E+03 8.40E-01 1 l- esrael 5 tube 7.00E43 3.50E-01 1.5eE46 7. tee +et I.40E41 l charging peep :- setid U tute flos (50 gpol : 1.50E42 3.00E-62 1.52E40 1.5eE+02 1.00E-42
- :- mereal GTSE
- 1.50E42 3.00E-02 1.52E+00 1.5eE+02 1.00E42 l- moreal 5 tabe 6.00E-64 3.00E-02 7.00E-02 6.00E4e 1. tee 42 Figure 5.48 MNt steam generator tube rupture event tree for a release of coolant contaminated with a gap release from the core
E A
?
M M 7 , stesejet 4 m safety valve safety value air eJetter release releew shole-body thyread shele4edy coolant tube fraction fraction done tree) deee tree) done tree) test:tions leat site E tsaditions een-nobles sette gas at I ele at i este at i eile I tube fails at ligt presere : solid 5 tube 9AII 30"Je94RY (500 ye) : 1.75E41 3.50E41 3.64E+02 3.50E+94 1.40E41 CLICEmikATIDL : :- eerest BTW 3 : 1.75E-01 3.50E41 3.64E+02 3.ME+04 1.40E41
! : sereal C tube
'7.00E43 3.50E-91 2.90E41 1.40E43 1.40E+41
! charging peep :- solid 5 tube Ifles (30 gpe) I 1.50E42 3.00E-92 3.03E+41 3.90E43 3.00E41
- : - mereal 875 1.50E-02 3.00E42 3 tX41 3.00E+03 3.00E41
- sereal I tube 6.00E-04 3.00E-92 1.50E*40 1.20E42 3.00E41 Figure 5.4C PWt steam generator tube rupture event tree for a release of coolant contaminated with a grain boundary release from the core
a
=
c 0
~
CD u' steae Jet 4 m safety valve safety velse air eje: tor release release m4cle-t ody thyroid whole-body ' coolant tute fractice frattson dose tree) dose treal dose tree) conditions leak size 55 c:rditions nee-nobles noble gas at 1 elle at 1 e le at 1 otte 1 tde f a:!s at nigh presure : - selti !! ter Pai (500 ;;s) : 1.75E-01 3.50E-01 9.02 42 B.7M+04 2.90E41 t>2CPATIM - l l- screal CTSS
- : 1.75E-0! 3.50E41 9.0X42 B.75E44 2.90E+01
- :- screal U tde
- 7.00E-03 3.50E-01 6.3M+01 3.50E+03 2.SOE*01 l charging pne; *~ selii U tale
- 41cm (50 g;o) : 1.50E-02 3.00E-02 7.56E+01 7.5C{+03 6.00E-01
- :- soreal 0T55
- 1.50E-02 3.00E-02 7.56E41 7.50E+03 6.00E-01 l- screal tf tute 6.00E-04 3.00E-02 3.6M+00 3.00E*02 6.0M-01 Figure 5.4D PWR steam generator tube rupture event tree for a release of coolant contaminated with a melt release from the core
- . _ _ _ - .- . _ - ~ _ . . . - _ _ _ _ - . _ _ - _ . _ . _ _ _ _. -
release wholeindy thyroid release shale 4edy thyroid g holdsp fractics release dose treet done treel fracties dose treal done tree) y dry or:1 time is dry sc!! mon-nobles fraction at 1 eale at I mile aos-estle at 1 male at 1 male leat rate act it!tered eatle gas est filtered act filtered filtered filtered it!tered 9 w conditions dry me!! y :0.5 hour 10cubour CD : 4.00E-01 1.00E40 4.24E+01 4.0K43 **+e ++++ nu i :
- 1- 100Uhour j sprays : : 4.00E42 1.0K+00 4.60E+00 4.00E+02 **** **** ****
efi 2-12 hours :- '00Uday
- 1.M 43 4.00E-02 1.84E41 1. W +01 1.60E-05 2.56E-02 1.60E-01
- :- dessy leak
- rate
- 5.0Cf46 2.00E-04 9.20E44 8.00E42 L90E46 1.20E-04 s.00E44
- - 100Uhour
- : 1.00E42 1.0E+00 1.21E+00 1.00E+02 ++++ *** nu
- 24 hours :- 100Uday
- 4.00E44 4.00E-02 4.04E-02 4.00E+00 4.00E-06 8.90E-03 4.00E-02 GAP K LEASE :- ;esip leak y Ftan EDRE rate 9 2.00E-06 2.00E44 2.42E44 2.00E-02 2.00E-00 4.40E-05 2.00E-04 w
l0.5 hour 100Uhour
- 3.00E-02 1.00E40 5.40E40 3.00E+02 **** *** **w
- - 100Unsur sprays : : 2.00E42 1.00E*00 2.60E40 2.00E42 **** **** ****
'2-12 hours - : - 100Uday
- : 3.00E44 4.00E42 1.04E-01 3.00E+00 8.00E46 2.40E-02 B.00E-62
- - desip leet
- rate 4.00E46 2.00E44 5.20E44 4.00E-02 4.00E-08 1.24E-04 4.00E-04
- :- 190Uhour
- ! 2.00E-03 1.00E40 4.10E-01 2.00E+01 ++n **** e+++
24 hours :- 100Uday
- s.0E-05 4.00E-02 1.64E-02 3.00E-01 8.00E-07 B.40E-03 B.00E-03
- - desi p l ut rate 4.00E-07 2.0E-04 8.2K-05 4.00E-03 4.00E49 4.24E-05 4.00E-05 Figure S.SA BWR contafrument, dry-well leakage / failure release event tree for a gap release from the core
. . _. -. . - . ~ . . . _ . _ _ , _ _ _ _ _ .
__- ___ .. _ . . _ _ . - - -. - _ - . -- - , -- _ _ - - = . - - -- shale 4edy thyrsti release ahole-body thyrsad release fraction release dose tree) dose treel fractise dose treal dose tree) y held up nonsoeles fraction at 1 mile at 1 este non-noble at 1 este at 1 otte so dry sell tier in dry sell leak rate est filtered echte gas est filtered not faltered (11teerd filtered (t!!ered oe conditions dry set! e Z m
- .5 hour 10cvhour 4.00E41 1.00E+00 8.4E42 2.00E+04 eeee e*+e use
- : - 100U hour 4.00E-02 1.00E+00 9.0E41 8.00E+03 +++e **** **ee sprays :
sif :2-12 hours - l- 10CUday
- : 1.60E-03 4.00E42 3.60E+00 3.20E+02 1.6K45 4.32E-01 3.2E+00
- :- desi p teak
- rate
- 8.00E-M 2.00E44 1.9E42 1.60E+00 LetE-et 2.1E-03 1.HE42
- :- 100Uheur
- : 1.00E-02 1.00E*00 2.3K41 2.00E+03 **** **** **e
- 24 hours - : - 100U day
- 4.00E-04 4.00E-02 9.40E-01 0.0E+01 4.00E-06 1.40E-01 8.00E41 i 59AIN 90VuWtf RELEASE :- desige test rate T
ea F#0M ECRE 2.00E-06 2.00E-04 4.7E-03 4.00E-01 2.00E-00 7.40E-04 4.00E-03 e
- .5 mour 100Uhour 3.00E-02 1.00E*00 1.0E+02 6.00E+03 eeee **** e**e
- :- 100Unour sprays : 2.00E42 1.00E+00 5.00E41 4.00E+03 ee _1 ***e noe
*2-12 hours- : - 100U day
- : B.00E44 4.00E-02 2.00E+00 1.6K+02 I.00E-M 4.1E41 1.60E*00
- : - desige test
- rate
- 4.0E-06 2.00E-04 1.00E-c2 8.0E41 4.00E48 2.00E-03 B.00E43
- 1- 100Uheur
- 2.00E-03 1.00E+00 7.5aE+00 4.00E+02 e+ee **** ***e ,
!24 hours : - 100U day
- B.0E-05 4.00E-02 3.00E-01 1.60E+01 3.00E-07 1.42E-01 1.60E-01
- - dnip tui rate 4.0E-07 2.00E-04 1.5K-03 8.00E-02 4.00E-09 7.00E-M 8.00E-04 l
Figure 5.58 BWR containment, dry-well leakage / failure release event tree for a grain boundairy release from the core
__ _ _ . . _ _ _ _ . _ __ -_ .____m -. _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ . _ _ _ . _ _ _ _ . . . _ . _ - _ . - - _ - - _ _ _ . _ release e%cle-togy thyroid release shcle-tody thyraid k hold g fraction release dose (reel sase tremi fracttog dose tros! dose tree) at 1 elle
$ dry sell time is dry sell con n:Cles fraction at 1 etle at 1 mile non-notles at 1 mile T* conditions dry sell leak ra*e act it!!ered ashle gas not faltered act f altered (11tered filtered filtere0 w
U :.5 kocr 100Uth.;:- CD ! 4.00E-01 1.00E40 2.0E43 2.00E45 wu e een
- :~- IM Uhour sprays : : 4.00E-07 :.00E40 2.2M42 2.00E+04 ++++ ++++ e+++
I eff :2-12 hours - :- 100Uday
- : 1.6K43 4.00E-02 9.90E40 8.00E42 1.60E-05 B.00E-01 8.00E40
- - desip leet
- rate
- 5.00E46 2.0K44 4.4X-02 4.00E+00 8.90E-Of 4.4K-03 4.00E-42 1 ! :- 100Uhour
- : 1.00E-02 1.00E+00 5.70E41 5.00E+c3 ene ++n un 124 hours :- 200Usay
- 4.0C(-64 4.00E42 2.2EE+00 2.00E42 4.00E-06 3.0K-01 2.00E+00
*RT PREASE !~ desip leak y ;py Est rate g 2.0M-C 2.00E-04 1.14E-02 1.00E+00 2.00E-es 1.50E-03 1.00E-42 1.5 hour - 100Ukour
- 3.00E-02 1.00E40 2.3X+02 1.50E44 em e+n n++
j : :- 100Unsur sprays : : 2.00E42 1.0C(+00 1.2M42 1.0K+04 o+++ n++ ++++ .I - ---on :2-12 hours :- 100Uday
- : 8.00E-04 4.00E-02 4.90E+00 4.00E+02 3.00E-06 3.40E-01 4.0eE*00 i
- :- desip leat
- rate
- 4.0CE46 2.00E44 2.40E-02 2.0M40 4.0M48 4.2X-03 2.00E-02
- - 100Uneur
- : 2.00E-03 1.00E40 1.7K41 1.00E+03 ++H +n* ++++
l24 hears :- 100Uday
- B.00E45 4.00E42 6.90E-01 4.00E+01 3.00E-07 2.94E41 4.00E-01
- - desige leat rate 4.0E-07 2.00E-04 3.4K-03 2.00E-01 4.00E49 1.42E-03 2.0K-03 Figure 5.5C BWt containment, dry-well leakage / failure release event tree for a melt release from the core
. _ _ - _ _ . - . . . - _ - - . . - _ _ _ _ __~ _- _ - _ _ _ _ _ _ _ - - . _ _ __ - -. . - . . . -. - -. _ . - _ - __ _ _ . - .
r z shele- body j releaw e6sle-body thyrend few treal release fracties done treal thyread dose tree) rn soppresten told up fractice releaw dew treet i past tier la set sell ese eseles fracties at 1 este at 1 eile sea-sektes at 1 elle at 1 etle U conditions dry / set mell leat rate est filtered esele gas est filtered set filtered (titered filtered filtered N co
* .5 tour 100Unsur
- 2.0X-02 1.00E40 4.40E40 2.00E42 **** *** ***
- : - 1991/hsur
- : 2.0E-03 1.00E*00 0.0K41 2.00E41 m HM **H
----sat :2-12 hours :- 100Uday
- ; 9.0M45 4.0E-02 L20E42 0.0E-01 0.00E-07 2.41E-02 0.00E-03
- :- destge rate
- 4.00E-07 2.00E-04 1.60E44 4.00E-03 4.00E-09 1.20E-04 4.00E-05
- - 2001/ hour
- : 1.00E-03 1.0M40 3.1K-01 1.00E+01 **** **** ****
224 hours :- 1991/ day
, ! 4.00E-05 4.00E-02 1.24E-02 4.00E41 4.00E-07 0.44E-03 4.00E-03 4
o EAP RELEf6E FR2M CORE
- - desip rate 2.00E-07 2.00E-04 6.20E-05 2.00E-03 2.00E-09 4.22E-05 2.0E-05
- .5 neur 100Utour
- 4.00E43 1.0M+00 2.00E+00 4.00E+01 un uw wu
- !- 100Uhour
! : 1.00E43 1.00E+00 7.00E-01 1.00E+01 *** **** ****
----s,.t :2-12 hours :- 1001/ day
- : 4.00E-05 4.00E-02 2.00E42 4.0K-01 4.00E-07 2.40E-02 4.0K-03
- :- desip rate
- 2.00E47 2.00E44 1.40E-04 2.00E43 2.00E-09 1.20E44 2.00E-05 I
- :- 1001/ hour
- 1.00E-03 1.0K+00 LIK-01 1'.00E41 *** *** ***
- 24 hours :- 19011 day 1 4.00E-05 4.00E-02 1.24E42 4.00E-01 4.00E-07 0.44E-03 4.00E-03 1 - desi p rate 2.0 s-07 2.00E-04 6.20E-05 2.00E-03 2.00E49 4.21-05 2.00E-05 Figure 5.6A OWt containment, wet-well leakage / failure release event tree for a gap release from the core
. _ _ _ . _ . _ . - . -* - . _ yv. _ - - - - - ,_ _ r -
._._. _ _ _._ ___.m __ _ _ _ _ . ..____ __ __ _ ____ _ . _ _ . _ _ _ _ _ _ _ _ _ _ __
l i
)
i i releau shele-body thyrest relesee ahete-body thyrsai
.$ dow treal deee treal g suppreston telt op fractic roleser dose (reel dose (reel fracties ;
y pool t2ee is set me!! sen-not!rs fracties at 1 etle at I este een-estles at 1 mile at 1 male H conditiers try/ set sell leak rate act filtered eetle gas set filtered est f attered (altered filtered filtered to tw
* :.5 hour it0Uhour
- 2.00E-02 1.00E+00 8.0M41 4.00E+t3 +++e **** ++w l : :- 100Uhoor
- : 2.00E43 1.0K+00 1.4141 4.00E+02 ++++ +++e ++++
-sat :2-12 hours - : - 10CD say
- : 1.CX-05 4.30E42 5.60E-01 1.60E41 8.00E47 4.92E-01 1.60E-01
- :- d6 sip rate
- 4.00E-07 2.00E44 2.50E-e3 8.00E-02 4.00E-09 2.01E-03 8.00E-04
- : :- 1scubour
- : 1.00E43 1.00C+00 5.50E40 2.00E+02 **+e one ++w I
! :24 hours :- 10cDear 4.0'4-05 4.00E-02 2.2X-01 B.00E*00 4.00E-07 1.41E-01 B.00E-02 57 E8A:s IOJGRT RELEASE :- dessp rate N FR3 CPE 2.00E-07 2.00E44 1.10E43 4.00E-C2 2.00E-09 7.04E-04 4.00E44 W
i :.5 hour 100Uhear
- 4.00E43 1.00E+00 4.90E41 8.00E42 +++e **+e +++e
- :- 100Uhour
- : 1.0M43 1.00E40 1.2K41 2.00E42 *** +++a ++w
.mb :2-12 hears- :- 100Usay
- : 4.00E-05 4.00E42 4.9M41 8.0X+00 4.00E-07 4.01E-01 5.0M-02
- :- desip rate l
. . 00E47 2.00E44 2.4X43 4.00E-02 2.00E49 2.00E-03 4.0K44 !
- i
- : - 100U bour i : : 1.0X43 1.00E+00 5.5X+00 2.00E+82 ++++ ++n n++
- 24 hours :- 100Utay
- 4.00E-05 4.00E-02 2.2K41 3.00E40 4.00E-07 1.41E-91 B.00E-02
- - seup rate -
2.0M-07 2.00E-04 1.1E-03 4.0K-02 2.00E-09 7.04E-04 4.0M-04 i Figure 5.68 IMt containment, wet-well leakage / failure release event tree for a grain boundary release from the core _ _ _ _ _ _ _ _ . . - . - -_ , _ _ _ , _ . _ - . , _ _ ...._ _ _.. __ ___ ..___ , _._ _~__ _ _ ._ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ ___ __ _
stoleisfy thyread release uhole body tb 9read release release ecse treal dose treal fracties dote tree) done itsel m suppresien tolg up fractise fraction at 1 atte at 1 mile een setle gas at I mile at 1 este i 12ee se set sell ese-nobles sool motte gas ect filtered set faltered filtered filtered filtered leak rate not filtered U eo
- ondztions der / set mell 1
e 2." hour 2001/ hour 1.90E42 1.00E44 ++++ **** ++++
- 2.00E42 1.00E40 1
- :- 1001/ hour ++++ *++e 3.00E41 1.0M+33 ++++
- : 2.00E-03 1.0M+00
---sa t :2-12 hours - 2001/ day 1.2M+00 4.00E+0! 8.00E-07 8.04E-01 4.00E41
- : 8.0E-05 4.00E-02
- :- desa p rate 2.00E43 4.00E-07 2.00E-04 4.00E4! 2.00E41 4.00E-09 4.025-03
- :- loci / hour ++++ ++++
1.2M41 5.30E+02 ++++
- : 1.00E-03 1.0M+00
- 24 hours :- IDC11daF 2.0M41 4.00E-05 4.0aE-02 4.90E-01 2.00E+0! 4.00E47 2.52E41 T PELT RELEASE
- - desip rate 2.0M47 2.00E-04 2.40E-03 1.00E41 2.00E49 1.4:E-03 1.00E-03 U FPyt EDPI
.5 hour loci / hour 4.0M-03 1.00[40 1.00E42 2.00E+03 *** ** nu
- - 1001/ neur 2.5M+01 3.00E*02 +++e +++e ++w
- 1.00E-03 1.0M+00
-ss3 l2-:2 hours :- 1001/ day
- 4.0M-05 4.00E-02 1.00E+00 2.0C41 4.00E-07 8.0X41 2.00E-01
- - destge rate 2.00E47 2.00E-04 5.0M43 1.00E-01 2.00E-09 4.01E43 1.00E43 1
- := 1001/ hour 1.00E*00 1.2M+01 5.00E42 *** **** ++++
- : 1.00E-03
- 24 hours -:- 1001/ day
- 4.00E-05 4.00E-02 4.90E-41 2.00E+01 4.00E-07 2.83 41 2.00E41
- - deur rate 2.00E44 2.4X-03 1.00E41 2.0M-09 1.41E-03 1.00E-03 2.00E-07 Figure 5.6C BWR containment, wet-well leakage / failure release event tree for a melt release from the core
l reltale releste enole-tedy llyteli tore release release frattles fraction dose treet dose tres) tessitical tonittnant rate non-ent es notte gas at I atle at I site t--- 1001/ hour EAP tittAtt i 4.0M41 l Kl+00 4.24t+0! 4.ME+03 F e 05 C0+[ ------ ---l e n t f il t er e d ---~ l --- I Mllt a y
- 1 1.6M42 4.M(42 1.62140 1.60142 I l-- typical testge rate 1 4.0M-C5 1.Nt44 4.06( 4 3 4.ME-01 1
1 1 l -- IM1/nour 1 i este eiese tese seese Ifiltered ----~!--- IM1/ day 1 1.6M C4 4 CCE-02 4.M( C2 1.60t+00 l-- tipital desige rate 4.0M47 1.M( 04 l .M(-44 4.00(-03 Figure 5.7A PWR and BWR containment bypass release event tree for a gap release from the core NUREG-1228 5-23
, reluse releete stole 4edy thyrend core release rein w fruttee frettien does troal 6ees trn) cesitttpe toedittene rate non noblee sette gas et i site at 1 elle l~ IHt/new H41u 90L*M81 1 4.00( 4 1 1.00(+44 8.44( 4 2 8.H(+H litt A$( f tM ~~~! set litterH~~l-- IMt/dat C:kt i I l.Ht42 4.00( 4 2 3.M 41 3. M *t3 1 l~ typical doelpe rate 1 4.00(-95 1.00( 4 4 0.14( 4 2 8.H(+M i
1 i 1- lH1/her i I eeee seen pH weet Illiterd ~~~~1 ~ lHildst i 1.H(44 4.00( 4 2 7.20( 4 1 3.M*01 1 - typical deelee rate 4.00( 4 7 1.00(94 1.00( 4 3 B.00(42 Figure 5.78 PWR and BWR containment bypass release event tree for a grain boundary release from the core NUREG-1228 5-24
l I i I l i r i i j i, roleete release shele4edy thyrsti , care relNeo release frattles fractlen due(tml fest (ree) . coadttleas - toditteet rete nonestles netteget at I elle at i elle (~ IM11bour MELT MLt4It i 4. M 91 1.64DM 2.00t +03 2. M
- M j Ft0R CORE ~~~~ leet filtered--l- IMt/de, 1 I I l.60E 02 4.60( 4 2 1.00t+01 1.00D03 +
1 l~ typical 6etite rate 1 4.0W-03 1.WM 2.0X 01 2. 2 01 1 1 I I ! ~ I N t/heer ' l I este e++ee stee essee , liittered -l-~ IMt/6er i 1- 1.60t44 4.Nb02 1.60t+M I % 01 l l~ typical 6esite rate 4.00E 47 1.M M 4.00E-03 2.00E 01 , 1 Figure 5.7C PWR and BWR containment bypass release event tree for a melt release from the core 1 NUREG-1228 5-25
1 N l g Table 5.1 1-elle doses for release of various core and coolant inventoriesII) m 1 c)
,', Total whole-body dose, res Thyroid inhalation dose, rem
$ (hours after shutdown) (2, 3) (hours after shutdown) (4)
Type of release 0 hr 1 hr 6 hr 12 hr 24 hr 0 hr 6 hr 100% gap noble gases 3x100 2.4x10o(5) 6x10 (5) 3x10 1 2.1x10 1 U)MC NC 100% grain boundary noble gases 5x101 4x101(5) 1x101(5) 5x100 3.5x10o(5) NC NC I 100% melt no W gases 1x102 8x103(5) 2x103(5) 1x101 7x10o(5) NC NC i 100% gap particulates and aerosols 2x102 NC 1x102(5) MC NC 1.7x10* 1x104(5) S' 100% grain boundary g particulates and aerosols 3x103 NC 2x103(5) MC NC 3x105 2x10s(5) 100% melt particulates
! and aerosols 8x103 NC 5x103(5) NC NC 6x105 5x105Ib) l 100% coolant normal 1x10 2(5) NC NC NC NC 1x10 (5) 9x10 2 100% coolant 100X spike (6) 1x10o(5) NC NC NC NC 1x102(5) NC i
l (1) 1-mile dose, 4 aph, D stability, ground level release, MRC NESORAD code. (2) Hours after shutdown release is projected to start. (3) Includes cloud shine, inhalation, and 3 hours of ground shine. 1 (4) Inhalation and adult thyroid only. (5) Assumed in reactor event trees. I
- For 0.5 , 6 , and 24-hour containment holdup cases,1, 6 and 24-hour noble gas factors were used, respectively.
For SGTR high pressure and low pressure cases, 1-hour and
+ noble gas factors were used, respectively.
- For bypass 100%/
(6) Spike of all non-noble fission products. hour and other cases, the 1-hour and 6-hour nc.,le gas factors were used, respectively. NOTE: NC = not calculated.
. - - , - . _ _ . _ _ _ _ _ . - _ _ - - - , _ _ _ . _ _ _ _ _ _ _ _ _ _ - . _ . . . - _ _ - - ~ _ _
Dose = (tree particulate / aerosol release fraction x 100% appropriate (e.g., gap) particulate / aerosol dose factor) + (tree noble gas release fraction x 100% noble gas dose factor x noble gas core release fraction) (Table 4.1). As can be seen from Table 5.1, decay is not important except for noble gases. Therefore, only noble gas dose factors differ depending on decay time. For the particulates/ aerosols, the 6-hour decay dose is used for all cases. For noble gases 1, 6 , or 24-hour decay cases were used, as appropr', ate. As an example, the plant conditions during the Perkins Point accident were as follows: (1) Core melt was projected with release into the containment through the ice condenser. (2) Sprays were not functional (once-through ice). (3) Ice was available. I (4) The containment was assumed to leak at design leakage. (5) A dose assessment was requested 2 hours after core damage. Figure 5.3C is the event tree for a PWR ice condenser plant with core melt. These conditions correspond to the branch of the tree that is fourth from the top. The whole-body dose is shown for these conditions as 1.2x10 2 rem, and the thyroid dose is 1.0 rem. Note also that the non-noble release fraction is 2x10.a for the once-through ice condenser case. From the example of the 7000 curies of Cs-134 being re-leased from the core, 7x108 x 2x10 6 = 1.4x101 curies would be released to the atmosphere. This is the same answer as that calculated in the sample. As dis-cussed in Section 1.3, source term can best be used to compare the possible consequences of various accident scenarios. As example, the following issue may be raised: Should a BWR containment be vented through tne suppression pool or should the containment be allowed to fail in a few hours? One of the related questions is how do the possible consequences of these two accident courses com-pare? Insight to answer this question can be gained by looking at Figures 5.6C and 5.5C for a core melt accident. If it were assumed that there are no sprays, a catastrophic failure of the drywell would occur in 2 to 12 hours from the start of the accident. Figure 5.5C indicates (second branch from top) that the whole-body dose would be in the range of 220 rem (early health effects are possible). While Figure 5.6C shows that if the release is through a subcooled suppression pool 2 to 12 hours after the start of the accident and a 100%/ hour release rate (sixth branch from bottom), the whole-body dose would be about 25 rem (early health effects are not possible). Therefore, it appears that there is a substantial reduction in possible consequences if the containment is vented through the suppression pool. Obviously, this is not the only con-sideration. There are many other questions that must be answered before venting should be started; these questions include: NUREG-1228 5-27
(1) Can the venting system withstand the pressure and other conditions? (2) Can it be stopped? (3) What will be the impact of venting on other plant systems and areas? 5.5 Comparison With Current Source Term Results Release fractions estimated from figures 5.2A through 5.7C were examined against NUREG-1150 release estimate ranges illustrated by the figures in Section 5 of NUREG-1150, Vol. 1. A typical result is shown in Figure 5.8. In this case, the 0.5-hour holdup case with no sprays was assumed from Figure 5.2C. Fig-ure 5.2C indicates an estimated release of 100% of the noble gases and 40% of the non-noble fission products. The multiple of these release fractions and the release fractions from the core for a core melt are shown on Figure 5.8. The release fractions calculated using Figure 5.2C fell well within the range of releases estimated by NUREG-1150 as did the other cases examined. This also shows a major problem for the analyst: the problem of how to show the uncer-talnties when presenting results. The 1-mile dose projections using the trees for ronditions similar to the WASH-1400 PWR release categories (see Section 1.4.3) also were compared with the CRAC code dose projections for the PWR WASH-1400 release categories. The CRAC code results are for weather conditions similar to those assumed in the tree calculations. The CRAC dose whole-body crojections assumed 4 hours of ground shine, and the tree calculations assumed only 3 hours of ground shine. The comparison is shown in Table 5.2, As can bo seen, the results were always within a factcr of 5. 5.6 Estimate the Uncertainties This final step is extremely important and the most difficult. Consider first the uncertainties surrounding some of the assumptions made in this example. The initial core release fraction assumed was based on the ref-erenced reactor core with similar burnup but with an entirely different operat-ing history. All of the fission product release fractions are based on limited experimental data and will most likely correspond to a single temperature. Not only will the actual temperature conditions most likely be different fram those of the experimentally obtained release fractions, but the temperatures of the fission products and their environment will continually change with time and with accident conditions. The transport and deposition of fission products in the reactor coolant system and in the containment also will be affected by pres-sures, viscosities, densities, hole sizes, material melting and boiling points, heat capacities, flows, time, surface areas, vrlumes, system availabilities, and chance. The bottom line is that any source term assessment made using this approach will be very crude. It has been estimated that the current codes used
~~
to estimate source term have an uncertainty ranging between 100 and 1000 (NUREG-0956) depending on the accident sequence. Because these ranges are for accidents for which all conditions are known, the source term uncertainties during an actual accident would be much greater. The single greatest source of uncertainties during a major accident is the containment leak rate, which could range to 107 All other effects after core damage are in the 102.to.10s range of uncertainty. NUREG-1228 5-28 )
a I j E COMPARISON OF RESULTS
- STATION BLACKOUT, EARLY FAILURE SCENARIO 1 1EO bO o
- OA A
~
9 O Ba V 1E-1 r O VA A I E O
- O 3, f m
~
VA O O
$ 1E-2 r
1 y i
~
V O
~
& O T f g z 1E-3 w
r 2 : z -
~
V O 1E-4 RodeonucEdo Group 3 r O Reactor Safety Study f* O MARCH / CORRAL
' * * * **8* ri m 1E-5 e s derium. .trontium 5 l NUREG-1150 (Bin 1) 6 ruthenium B
. A NUREG-1210 Vol. 6 Figure 5.2 7 lanthanidos. =ctinid.e 1E-6 i i 2 3 4 5 6 7 RADIONUCLIDE GROUP Figure 5.8 Comparison of source term projection results for station blackout scenarios at Surry Source: NUREG-1150
Table 5.2 Comparison of event tree doses (1) with CRAC whole-body dose projections for WASH-1400 PWR release categories at 1 mile WASH-1400 Tree WASH-1400 whole-body dose I2) whole-body release category (rem) at 1 mile dose at 1 mile PWR 1 5x102 2x103 PWR 4 1.5x102 2.6x101 , PWR 5 4x101 1.4x10lI3) l (1) Figure 5.20. 1 (2) Source: NUREG-1062, Case 4. (3) Assumed a 3-hour release. As an illustration of uncertainties, Figure 5.8 shows the source term projection , for various studies (e.g., WASH-1400) and the range projected by NUREG-1150. l 1 l l P L i i NUREG-1228 5-30
L I 6 REFERENCES l J American National Standards Institute /American Nuclear Society, ANSI /ANS 18.1, !
"Radioactive Source Tera for Normal Operation of Light Water Reactors," !
La Grange Park,1111nois, December 1984. ; American Nuclear Society, "Report of the Special Committee on Source Terms," ;
- La Grangs Park, Illinois, September 1984.
i Babcock & Wilcox, Study Guide "Operator Training - Degraded Core Recognition ; l and Mitigation," Vo',. 3. TRG-81-3, 77-1125866-00, Lynchburg, Virginia, June 1981. , I Battelle Columbus Laboratories, J. A. Gieseke, P. Cybulskis, et al., "Radio- !
- nuclide Release Under Specific LWR Accident Conditions - PWR-Large, Dry Contain-j ment Design (Zion Plant)," Vol. VI BMI-2104, Columbus, Ohio, July 1984. l Clinton, S. D., "Monthly Report for Experimental Determination of Iodine Trans- !
i port From a Failed PWR Generator Tube (80832)," Oak Ridge National Laboratory, 1 February 1984. ' l 1 Denning, R. S. and P. Cybulskis, "Secondary Containment Buildings Performance ! j in Severe Accidents," Third Workshop on Containment Integrity, Washington. 0.C., ! j May 21-23, 1986. i Hanford Engineering Development Laboratory, R. K. Hilliard, A. K. Postma, I
"Large-Scale Fission Product Containment Test," HEDL-S.t 2254 FP, American {
Nuclear Society Winter Meeting, November 16-21, 1980. : l i Postma, A. K., Benton City Technology, "Iodine Evolution From Spray Water," i j BCT-LR-101, Benton City Washington, October 1978, i
- i U.S. Department of Energy (originally AEC), J. M. Harrer, J. G. Beckerley, "Nuclear Power Instrumentation Systems Handbook," Vol. 2. TID-25952 P2, ;
Technical Inforestion Center, Washington. 0.C., 1974. J -- , W. J. Bailey, M. Tokar "Fuel Performance Annual Report for 1982." Nuclear Safety, Vol. 264, page 367, Office of Scientific and Technical Informa-l I tion, May-June 1985. ! / l' U.S. Environmental Protection Agency "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," EPA-520/1-75-001 Washington, D.C., June 1980. , U.S. Nuclear Regulatory Commission, NUREG-)S/014 (WASH-1400), "Reactor Safety I Study," Washington, D.C., December 1975. 1 l i i -- , NUREG-0396 H. E. Collins, B. K. Grimes, et al. , "Planning Basis for the ! Technical Development of State and Local Government Radiological Emergency l Res;gnse Plans in Support of tight Water Nuclear Power Plants," December 1978. 1 l 1 j NUREG-1228 6-1
--- NUREG-0600, "Investigation Into the March 28, 1979 Three Mile Island AccIJentbyOfficeofInspectionandEnforcement,"August 1979.
-- , NUREG-0737, "Clarification of TMI Action Plan Requirement," November 1980.
NUREG-0772, "Technical Basis for Esti nting Fission Product Behavior durIngLWRAccidents,"June 1981.
-- , biRIG-0800, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants," July 1981.
NUREG-0900, J. T. Larkins, M. A. Cunningham, "Nuclear Power Plant Severe AccIdentResearchPlan,"Janu.ry1983.
-- , NUREG-0909, "NRC Report on the January"25,1982 Steam Generator Tube Rupt :re at R. E. Ginna Nuclear Power Plant, April 1982, i -- , NUREG-0956, M. Silberberg, J. A. Mitchell, et al., "Reassessment of the Technical Bases for Estimating Source Ters," July 1985.
-- , NUREG-1037, "Containment Performance Working Group Report," Draf t Report j for Comment, May 6, 1985.
-- , NUREG-1062 T. S. Magrulies, J. A. Martin "Dose Ct.lculation for Severe
! LWR Accident Scenarios," May 1984.
-- , NUREG-1150, "Reactor Risk Reference Document," Draft for Comment, Vol. 3 "Reactor Risk Reference Document: Appendices J-0," February 1987.
-- , NUREG-1210, J. A. Martin, T. J. McKenna et al., "Pilot Program: NRC Severe Reactor Accident Incident Resp'onse Training Manual," Vol. 2. "Severe Reactor Accitient Overview " Val. 3 Response of Cicensee and State and Local Officials,"Vol.4,"Pub 1IcProtectiveAction"February 1987.
i J
-- , NUREG/CR4?39, D. C. Aldrich, J. L. Sprung, et al. , "Technical Guidance for Si'ing Criteria Devel o sent," SANDIA 81-1549, Sandia National Laboratories, December 1982. i
! -- , NUREG/CR-2'72, d R. P. Wichner, C. F. Weber, et al., "SBLOCA Outside ! Containment at Browns Ferry Unit One,' Vol. 2 "Iodinr, Cesium, and Noble Gas Distribution and Release," ORNL/TM-8119/V2, Ock Ridge National Laboratory, October 1983.
-- , NUREG/CR-2726, A. L. Camp, J. C. Cummings, et al., "Light Water Reactor Hydrogen Minual," SANDIA 82-1137 Sandia National Laboratories, August 1983.
t -- , NUREG/CR-2907, J. Tichler, K. Nonder, "Radioactive Material Released From Nuclear Pover Plants - Annual Report 1982," BNL-NUREG-51581, Vol. 3, Brookhaven < National Laboratory, February 1986. 1
-- , NUREG/CR-2940, S. R. Greene, "Realistic Simulation of Severe Accidents in BWRs - Computer Modeling Requirements " ORNL/TM-8517, Oak Ridge National Laboratory, April 1984.
I NbREG-1228 6-2
-- , NUREG/CR-3108, W. B. Wilson, T. R. England, R. J. La Bauve, "Extended Burnup Calculations for Operating Reactor Reflood Reviews," LA-9563-MS, Los Alamos National Laboratory, February 1983.
-- , NUREG/CR-3287, G. G. Killough, C. L. Begovich, A. L. Sjoreen, "A Guide for the TACT III Computer Code," ORNL/TM-S763, Oak Ridge National Laboratory, May 1983.
-- , NUREG/CR-3727, F. R. Zalowdek, A. K. Postma, W. K. Winegardner, "Fission Product Removal in Engineered Safety FeatJres (ESF) Systems, PNL-5050, Pacific Northwest Laboratory, April 1984.
-- , NUREG/CR-4000, R. I. Scherpelz, T. J. Bander, G. F. Athey, J. V. Ramsdell, "The MESORnD Dose Assessment Model," Vol.1, Technical Basis, PNL-5219, Pacific Northwest Laboratory, March 1986.
-- , NUREG/CR-4081, M. F. Albert, "The Absorption of Gaseout Iodine by 2 ater Droplets," ORNL/TM-9488, Oak Ridge National Laboratory, July 1985.
, NUREG/CR-4467, D.S. Alpert, D. I. Chanin, L. T. Ritchie, "Relative Importance of Individual Elements to Reactor Accident Consequences Assuming Equal Release Fraction," SAN 85-2575, Sandia National Laboratories, March 1986.
-- , NUREG/CR-4245, S. W. Duce, S. T. Croney, "In-Plant Source Term Measurements at Brunswick Steam Electric Station," EGG-2392, EG&G Idaho Inc. , June 1985.
-- , NUREG/CR-4722, T. McKenna, A. L. Sjoreen, "Source Term Estimation Using MENU-TACT; ORNL-6314," Oak Ridge National Laboratory, to be submitted in draf t.
-- , Re2ulatory Guide 1.3, Rev. 2, "Assumptions Used for Evaluating the Potential P.adiological Consequences of a loss-of-Coolant Accident for Boiling-Water Reactors, June 1974.
l l
-- , Regulatory Guide 1.4, Rev. 2, "Assumptions Used for Evaluating the Potential Ra*)iological Consequences of a loss-of-Coolant Accident for Pressurized-Wrer Reactors," June 1974.
NUREG-1228 C-3
l APPENDIX A PROCEDURE FOR ESTIMATING 0FFSITE CONSEQUENCES I BASED ON PLANT CONDITIONS i NUREG-1228
l l This appendix uses a set-by-set system of tabs to walk the user through the analysis of a large ranga of accident conditions and reactor types. The figures and tables referenced in this appendix may be found in Sections 4 and 5 of NUREG-1228 and are called out as they are numbered in the main report. When l this appendix is being built as a stand-alone procedure, these figures and tables would be taken from the main report and included here. Instruction on where material should be inserted if building a stand-alone procedure is given in brackets ([]) throughout. NVREG-1228 1 Appendix A l
DOSE ESTIMATION FROM PLANT CONDITIONS Procedure: Rev. uted: 1.0 OBJECTIVE The objective of this procedure is to estimate offsite consequences and re-lease source terms based on the projected status of a limited number of plant conditions. 2.0 RESPONSIBILITY This procedure will be used by the source term specialist and the accident sequence analyst with the results provided to the appropriate radiological assessment manager. 3.0 GUIDANCE STEP 1 See Tab G for a discussion of the acumptions used in development of this method, if required. NOTE This procedure provides c W e estimates accurate within a factor of 10-100, only if plant conditions are accurately re) resented. Results are for a 1000 MWe plant and should be adjusted if tie plant is much smaller (<1/2). ACTION Select apprcpriate tab for containment / leakage type. Tab A PWR Large, Dry or Subatmospheric Containment Leakage Tab B PWR Ice Condenser Containment Leakage Tab C PWR Steam Generator Tube Rupture Tab D BWR Containment Dry-Well Leakage / Failure Tab E BWR Containment Wet-Well Leakage / Failure Tab F BWR/PWR Containment Bypass (Event V) Tab G General Description of Assumptions Tab G.1 Core release fraction assumptions Tab G.2 Core inventories (Ci/MWe) Tab G.3 Summary of major PWR release pathway reduction mechanisms Tab G.4 Summary of major BWR release pathway reduction mechanisms Tab G.5 Assumed containment and steam generator tube rupture esca)e fraction for 1 hour (release rates) Tab G.6 PWR aaseline coolant concentration. for various levels of core damage Tab G.7 BWR baselina coolant concentrations for various levels of core damage Tab G.8 1-mile dose factors used in the event tree NUREG-1228 2 Appendix A
TAB A PWR LARGE, DRY OR SUBATHOSPHERIC CONTAINMENT RELEASE EVENT TREES FOR A GAP (1), GRAIN BOUNDARY (2), OR MELT (3) RELEASE FROM CORE NOTE A general description of the assumptions used in the event trees follows Step 3. ACTION STEP 2 Gather following plant information: Plant name Core condition - check gap release (1300-2000*F) grain boundary (>3000"F) melt (>4500'F) Containment sprays - check on off Holdup time before release 0.5 hour 2-12 hours 24 hours Containment leak rate design (0.1%/ day) 100%/ day , 100%/ hour Time Analyst .__ STEP 3 Use the event trees, based on information above, to estimate dose and source term. Select appropriate event tree based on core condition. Tab A-1 Gap Tab A-2 Grain Boundary Tab A-3 Melt NUREG-1228 3 Appendix A
FWR LARGE, DRY OR SUBATHOSPHERIC CONTAINMENT EVENT TREES GENERAL DESCRIPTION: The core is uncovered and release fractions for a 1000 MWe plant typical of a gap (1300-2000 F), grain boundary (3000*F), or melt (4500 F) release from the core are assumed. For the vessel melt-through cases (ex-vessel melt), the strontium release fractions could be substantially increased. This could increase the projected whole-body dose by 50% for cases' involving releases of large fractions of the non-noble fission products. The release system into the containment atmo-passes sphere asbyshown a dry inpathway through Figure 4.5 theA-1 pathway primary (primary leakage) or A-2 (PORV or SRV). Particulates and aerosols airborne in containment are reduced to account for the actions of sprays or natural processes for 0.5, 2-12, or 24 hours. Re-leases from the containment are estimated for 0.1%/ day, Figure 4.5 pathway B-1 (design leakage); 100%/ day, aathway B-2 (isolation valve seal failure); or 100%/ hour, pathway B-3 (catastroplic fa.ilure >2 sq ft). Noble gas release fraction is a fanction of conta^nment release rate only. The doses are calcu11' for a 1-hour ground level release, 4 mph, and D stabil-ity wind conditicas. r.nle-body dose includes cloud shine, 3 hours of ground shine, and inha.iation. Noble gas decay is accounted for. Thyroid dose (adult) is for inhalation only. A further description of assumptions is contained in Tab G. [ Insert Figure 4.5 (NUREG-1228) after this page.] NUREG-1228 4 Appendix A
TAB A-1 GAP STEP 4 Record from event tree (see Figure 5.2A): Event de'scription Time Analyst . Whole-body dose @ l mi ' Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi 9 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = 0 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >1 rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager, l (Insert Figure 5.2A (NUREG-1228) after this page.] NUREG-1228 5 Appendix A
TAB A-2 GRAIN B0UNDARY SJEP 4 Record from event tree (see Figure 5.2B): Event description Time Analyst Whole-body dose @ l mi Thyroid dose @ 1 mi STEP 5 l I Calculate baseline dose at 2, 5, and 10 mi. 1 ! Whole-body dose from event tree: Whole-body dose: @ 1 mi ,
@ 1 mi
- 0.40 = @ 2 mi ~ l
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 ,
Present results. NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined Mlow. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >N rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose '
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the )tojected dose. Combine this assessment with the general descripti n (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.2B (NUREG-1228) after this page.] NUREG-1228 6 Appendix A rT =- -.+ _ _
TAB A-3 MELT STEP 4 Record from event trec (see Figure 5.2C): Event description Time Analyst Whole-body dose @ l mi Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi 9 1 mi
- 0.09 = 0 5 mi
@ 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = 0 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results, ;
HOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: ! Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), inpat information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager.
- [ Insert Figure 5.2C (NUREG-1228) after this page.)
NUREG-1228 7 Appendix A
TAB B PWR ICE CONDENSER CONTAINMENT RELEASE EVENT TREES FOR A GAP (1), GRAIN B0UNDARY (2), OR MELT (3) RELEASE FROM CORE NOTE A general description of the assumptions used in the event trees follows Step 3. ACTION STEP 2 Gather following plant information: Plant name Core condition - check gap release (1300-2000*F) grain boundary (>3000*F) melt (>4500 F) Containment sprays - check on l off Holdup time before release 0.5 hour 2-12 hours 24 hours s Containment leak rate , design (0.25%/ day) 100%/ day 100%/ hour Recirculation fans
- on (with recirculation) off(once-throughice)
Analyst STEP 3 Use the event trees, based on information above, to estimate dose and source tern. Select appropriate event tree based on core condition. Tab B-1 Gap Tab B-2 Grain Boundary , Tab B-3 Melt
*If the ice condenser is bypassed or the ice is exhausted before the release from core, use PWR large, dry containment release trees, Tab A.
8 Appendix A NUREG-1228
PWR ICE CONDENSER CONTAINMENT RELEASE EVENT TREES FOR A GAP (1), GRAIN B0UNDARY (2), OR MELT (3) RELEASE FROM CORE GENERAL DESCRIPTION: The core is uncovered and release fractions typical of a gap (1300-2000 F), grain boundary (3000 F), or melt (4500 F) release from the core are assumed. For the vessel melt-through cases (ex-vessel melt) the strontium release fractions could be substantially increased. This could in-crease the projected whole-body dose by 50% for cases involving releases of large fractions of the non-noble fission products. The release passes by a dry pathway through the primary system into the containment atmosphere tr. rough the ice condenser, shown in Figure 4.6 as pathway A-1 (primary leakage) or A-2 (PORV or SRV). Particulates and aerosols airborne in the containment are reduced j to account for the actions of sprays and natural processes for 0.5, 2-12, or ' 24 hours. Ice condenser removal for once-through and recirculation is estimated. It is assumed that the ice is not depleted before core damage. If the ice con-denser is bypassed or the ice is exhausted before the release from the core, the PWR large, dry containment release trees should be used. Releases from the con-tainment are estimated for 0.25%/ day, Figure 4.6 pathway B-1 (design leakage); 100%/ day, pathway B-2 (isolation valve seal failure); or 100%/ hour, pathway b-3 (catastrophic failure >2 sq ft). Noble gas release fraction is a function of containment release rate only. Doses at 1 mi are calculated for a 1-hour ground level release, 4 mph, and D stability wind conditions. Whole-body dose includes cloud shine, 3 hours of ground shine, and inhalation. Noble gas decay was accounted for. Thyroid dose (adult) is for inhalation only. [ Insert Figure 4.6 (NUREG-1228) after this page.] l t NUREG-1228 9 Appendix A
TAB B-1 GAP STEP 4 Record from event tree (see Figure 5.3A): Event description Time Analyst Whole-body dose @ 1 mi Thyroid dose @ 1 mi STEP 5 Calculate baselina dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi 0 1 m!
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to wr.ich early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which DAG may be exceeded (WB dose >l rem or thyrof. dose l >5 rem) Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.3A (NUREG-1228) &fter this page.] NUREG-1228 10 Appendix A
TAB B-2 GRAIN BOUNDARY STEP 4 Record from event tree (see Figure 5.3B): Event description Time Analyst Whole-body dose @ l m1 Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi 0 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi 9 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
>5 rem)
Assuming that the a'cident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathvay(s). Provide results t7 the Radiological Assessment Manager, a (Insert Figure 5.3B (NUREG-1228) after this page.] NUREG-1228 11 Appendix A
TAB B-3 CORE MELT STCP 4 Record from event tree (see Figure 5.3C): Event description Time Analyst Whole-body dose @ l m1 Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi 0 1 mi
- 0.40 = @ 2 mi 0 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
NOTE Decause of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose >5 rem) Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Mariager. [ Insert Figure 5.3C (NUREG-1228) after this page.] NUREG-1228 12 Appendix A
ll TAB C , ! PWR STEAM GENERATOR TUBE RUPTURE RELEASE EVENT TREES l FOR TYPICAL COOLANT AND SPIKE NOTE A general description of the assumptions used in the event trees follows Step 3. ACTION STEP 2 Gather following plant information:
- . Plant name Coolant concentration !
normal l 100X spike l gap release (1300-2000'F) ! grain boundary (3000'F) melt (4500*F) Tube leak size 1 tube failure at high pressure (500 gpe)
- 1 charging pump flow (50 gpm) l i
Steam generator conditions
- I normal "U" tube solid secondary "U" tube normal once-through steam generator (B&W)
Atmospheric release point safety valves condensersteam-jetair-ejector Time i Analyst STEP 3 Use the event trees, based on information above, to estimate dose and source 1~ term. Select appropriate event tree based on coolant and core conditions. Tab C-1 Normal and 100X spike (safety valve release only) Tab C-2 Gap Tab C-3 Grain Boundary Tab C-4 Helt 1 l I *For a dry (primary and secondary) release pathway use the BWR/PWR containment , bypass trees. l 1 1 i j NUREG-1228 13 Appendix A
-y-,,,y, , , , _ _ _ _ . . , - ,..-----.-m--.--.
, + - - _ - - - - - - ,,,+-_.-_.-.-,,%, , , , . . , - - , - - - - , - , - . _ _ . _ , --
-_,-c, pm - , , , . -r-- , -,,-n-_--,
' PWR STEAM GENERATOR TUBE RUPTURE EVENT TREE FOR A RELEASE OF COOLANT NORMAL AND 100 X NORMAL (SPIKE) CONCENTRATIONS (1), GAP CONCENTRATIONS (2), GRAIN BOUNDARY CONCENTRATIONS (3), AND MELT CONCENTRATIONS (4) GENERAL DESCRIPTION: Two normal range coolant concentrations (normal and 100X non-noble fission products spike) are assumed in addition to three accident con-centrations. The accident concentrations assume that all of the gap, grain boundary, or melt core release fractions are contained in the coolant. The associated coolant concentrations are found in Tab G. The coolant is assumed to be released by a steam generator tube rupture (SGTR) to the secondary side and then to the atmosphere by the safety relief valves or through the condenser and then the steam-jet air ejector (Figure 4.5 or 4.6 pathway C-1 or C-4). The release rate for total failure of one SG tube at full pressure or for coolant (Figures 4.5 and 4.6, pathway A-3) being pushed out by one charging pump is assumed. SG partitioning for normal and a solid secondary side of the SG are assumed. A 50% reduction of non noble fission products, to account for secondary side dilution, is assumed. Release of all the noble gases is assumed for release by the steam-jet air ejector. For a dry primary and secondary side release path-way, the BWR/PWR containment bypass trees should be used. For large and/or multiple failures in a "U" tube SG the once-through steam gen-erator case should be used. The doses at 1 mi are calculated for a 1-hour ground level release, 4 mph, and D stability wind conditions. Whole-body dose includes cloud shine, 3 hours of ground shine, and inhalation. Noble gas decay is accounted for by assuming a 1-hour decay for the high pressure release and a 6-hour decay for the low-pressure release. Thyroid dose (adult) is for inhalation only. [ Insert Figures 4.5 and 4.6 (NUREG-1228) after this page.] NUREG-1228 14 Appendix A
I TAB C-1 NORMAL AND 100X SPIKE STEP 4 Pecord from event tree (see Figure 5.4A): Event description Time Analyst Whole-body dose @ l mi Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whcle-body dose from event tree: Whole-body dose: 9 1 mi
@ 1 mi
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 9 5 mi
@ 1 mi
- 0.03 = 9 10 mi Thyro!d dose from event tree: Thyroid dose: @ 1 mi 0 1 mi
- 0.40 = @ 2 mi 0 1 mi
- 0.09 = 9 5 mi
@ 1 mi
- 0.03 = 9 10 mi ___
STEP 6 Present results. NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thytaid dose >5 rem) Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.4A (NUREG-1228) after this page.] NUREG-1228 15 Appendix A
TAB C-2 GAP STEP 4 Record from event tree (see Figure 5.48): Event description Time Analyst Whole-body dose @ 1 mi Thyroid dose @ 1 mi l STEP 5 Calculate baseline dose at 2, 5, and 10 mi. ' Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi l Thyroid dose from event tree: Thyroid dose: @ 1 mi i
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.05 = @ 10 mi STEP 6 i Present results.
i NOTE - Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION / l Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) __ , Distance to which PAG may be exceeded (WB dose >.1 rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. I Combine this assessment with the general description (following Step 3), input 'i information (Step 2), and a markup of figure showing assumed release pathway (s). t Provide results to the Radiological Assessment Manager. [ Insert Figure 5.48 (NUREG-1228) after this page.] NUREG-1228 16 Appendix A
TAB C-3 GRAIN BOUNDARY STEP 4 Record from event tree (see Figure 5.4C): Event description Time i Analyst l Whole-body dose IP 1 mi l Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi 0 1 mi
- 0.03 = @ 10 mi l Thyroid dose from event tree: Thyroid dose: @ 1 mi .
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
"10TE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below.
ACTION , Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the qeneral descripton (following Step 3), input
- information (Step 2), and a markup of figure showing assumed release pathway (s). '
Provide results to the Radiological Assessment Manager. - (Insert Figure 5.4C (NUREG-1228) after this page.] NUREG-1228 17 Appendix A
TAB C-4 MELT STEP 4 Record from event tree (see Figure 5.40): Event description Time Analyst Whole-body dose @ l m1 Thyroid dose @ 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: @ 1 mi 0 1 mi
- 0.40 = 0 2 mi 0 1 mi
- 0.09 = @ 5 mi 0 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 al
- 0.40 = @ 2 mi 0 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi l STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem o.r thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. l Combine this assessment with the general description (following Step 3), input l information (Step 2), and a markup of figure showing assumed release pathway (s). l Provide results to the Radiological Assessment Manager. [ Insert Figure 5.4D (NVREG-1228) after this page.] NUREG-1228 18 Appendix A i
TAB D BWR CONTAINMENT 'JRY-WELL LEAKAGE / FAILURE FOR A GAP (1), GRAIN B0UNDARY (2), OR MELT (3) RELEASE FROM CORE NOTE A general description of the assumptions used in the event trees follows Step 3. ACTION STEP 2 Gather following plant information: Plant name Core condition - check gap release (1300-2000 F) grain boundary (>3000 F) melt (>4500 F) Dry-well containment sprays - check on off Holdup time in dry well before release 0.5 hour 2-12 hours 24 hours Dry-well leak rate design (0.5%/ day) 100%/ day 100%/ hour Release through filters (SBGTS) yes no Time Analyst STEP 3 Use event trees, based on information above, to estimate dose and source tarm. Select appropriate event tree based on core condition. Tab 0-1 Gap Tab 0-2 Grain Boundary Tab 0-3 Melt i HUREG-1228 19 Appendix A
BWR CONTAINMENT DRY-WELL LEAKAGE RELEASE EVENT TREES FOR A GAP (1), GRAIN BOUNDARY (2), OR MELT (3) RELEASE FROM CORE GENERAL DESCRIPTION: The core is uncovered and release fractions for a 1000 MWe plant (Tab G) typical of a gap (1300-2000'F), grain boundary (3000 F), or melt (4500 F) release from the core are assumed. For the vessel melt-through cases (ex-vessel melt) the strontium release fractions could be substantially increased. This could increase the projected whole-body dose by 50% for cases involving releases of large fractions of the non-noble fission products. The release passes by a dry pathway through the primary system into the containment dry-well atmosphere through a loss-of-coolant accident (LOCA), shown in Fig-ures 4.7, 4.8, and 4.9 as pathway A-1 (primary leakage). Particulates and i aerosols airborne in the containment dry well are reduced to account for the ; actions of sprays and natural processes for 0.5, 2-12, or 24 hours. Releases ' from the containment dry well are estimated for 0.5%/ day (Figures 4.7, 4.8, and 4.9) pathway 8-1 (design leakage), 100%/ day pathway B-2 (isolation valve seal failure), or 100%/ hour pathway B-3 (catastrophic failure >2 sq ft). The release can be filtered by the standby gas treatment system (SBGTS) (oethway C-2) be-fore release or bypass the SBGTS (pathway C-1). Noble gas release fraction is a function of release rate only. The doses at 1 mi are calculated for a 1-hour ground level release, 4 mph, and D stability wind conditicns. Whole-body dose includes cloud shine, 3 hours of ground shine, and inhalation. Noble gas decay is accounted for. Thyroid dose (adult) is for inhalation only. [InsertFigures4.7,4.8,and4.9(NUREG-1228)afterthispage,j l i i l NUREG-1228 20 Appendix A
TAB D-1 GAP STEP 4 Record from event tree (see Figure 5.5A): Event description Time Analyst . Whole-body dose 9 1 m1 Thyroid dose 9 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. ! Whole-body dose from event tree: Whole-body dose: 9 1 mi 9 1 mi
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 9 5 mi 9 1 mi
- 0.03 = 9 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi .
9 1 mi
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 9 5 mi 9 1 mi
- 0.03 = 9 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) , i Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose,
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. ] Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). 1 l Pruvide results to the Radiological Assessment Manager. (Insert Figure 5.5A (NUREG-1228) after this page.] > NUREG-1228 21 Appendix A
TAB D-2 GRAIN BOUNDARY STEP 4 i Record from event tree (see Figure 5.5B): Event description Time Analyst Whole-body dose @ l mi Thyroid dose 9 1 mi ) j STEP 5 ' i Calculate baseline dose at 2, 5, and 10 mi. , Whole-body dose from event tree: Whole-body dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi L
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi l
Thyroid dose from event tree: Thyroid dose: @ 1 mi 9 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi 9 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
4 NOTE ! Because of great uncertainty, dose numbers should not be used in presenting [ ] results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: i Distance to which early deaths are possible (WB dose >200 rem) I Distance to which early health effects are possible (WB dose >50 rem) , Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose !
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could - reasonably vary by a factor of 10 to 100 from the projected dose. ! Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. 4 [ Insert Figure 5.5B (NUREG-1228) after this page.) 1 J i i NUREG-1228 Appendix A 22 ( t
i l TAB D-3 CORE MELT STEP 4 Record from event tree (see Figure 5.5C): Event description Time Analyst Whole-body dose IP 1 mi Thyroid dose @ 1 mi i STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: 9 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi Thyroid dose from event tree: Thyroid dose: @ 1 mi
@ 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi
@ 1 mi
- 0.03 = @ 10 mi i l STEP 6 Present results.
NOTE 4 Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) i Distance to which early health effects are possible (WB dose >50 rem) l Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose l >5 rem) ; Assuming that the accident proceeds as projected, the actual releases could
.l reasonably vary by a factor of 10 to 100 from the projected dose.
Combine this assessment with the general description (following Step 3), input l information (Step 2), and a markup of figure showing assumed reloase pathway (s). t Provide results to the Radiological Assesstent Manager. [ Insert Figure 5.5C (NUREG-1228) after this page.] NUREG-1228 23 Appendix a
TAB E 1 BWR CONTAINMENT WET-WELL LEAKAGE / FAILURE FOR A GAP (1), GRAIN BOUNDARY (2), OR MELT (3) RELEASE FROM CORE NOTE A general description of the assumptions used in the event trees follows Step 3. 1 ACTION STEP 2 Gather following plant infonvation: 1 Plant name ~ - ~ i i Core conditT5'n - check l J gap release (1300-2000*F) ~ ~ ~ " " ~ l grain boundary (>3000 F; * ~~" l j melt (>600*F) ' ~ ~ ~ ' " ~ ~ ' ! , Suppression pool conditfoni ; scturated ~~~ i subcoohd , ! Holdup time in dry / wet well befoi a r%an i 0.5 hcer - ~~~~ f 2-12 hot.rs " ~' ~ ~' i 3 i 24 hours ' - ' ~ ' We'.-well leak rate ; i design (0.5%/ day) , _ . 100%/ day : 100%/ hour -'- ~ r 4 Release through filters (SBGT) yes __ no ~ 4 Time ! j Analyst _ _ _ , , f L STEP 3 ; Use event trees, based on information abwe, to estimate dose and source term. ! Select appropriate event tree based on core condition. l Tab E-1 Gap Tab E-2 Grain Boundary ) Tab E-3 Melt ' l
! i i
3 j *For bypass of the suppression pool, use BWR containment dry-well trees, Tab D. e i E i NUREG-1228 24 Appendix A ! i
1 l l BWR CONTAINMENT WET-WELL LEAKAGE RELEASE EVENT TREES F0r, l A GAP (1), GRAIN BOUNDARY (2), OR MELT (3) RELEASE FROM CORE l GENERAL DESCRIPTION: The core is uncovered and release fractions (Tab G) typical of a gap (1300-2000 F), grain boundary (3000'F), or melt (4500'F) re-lease from the core are assumed. For the vessel melt-through cases (ex-vessel melt) the Sr release fractions could be substantially increased. This could increase the projected whole-body dose by 50% for cases involving releases of large fractions of the non-noble fission products. The release passes through the suppression pool into the containment wet-well atmosphere through a LOCA, shown in Figures 4.7, 4.8, and 4.9 as pathway A-2 (primary leakage) or A-3 (ADS /SRV). Particulates and aerosols airborne in the containment wet well are reduced to account for the actions of the suppression pool under subcooled and saturated conditions. For bypass of the suppression pool, use BWR containment dry-well trees. It is assumed that only decay heat is being released. If the reactor is not shut down, 10 to 20 times as much particulates could be released through the pool. In addition, it is assumed that the particulates are reduced to account for natural depletion while held up in the containment. Releases from the containment wet well are estimated for 0.5%/ day as shown in Figures 4.7, 4.8, and 4.9, pathway B-1 (design leakage); 100%/ day, pathway B-2 (isolation valve seal failure); or 100%/ hour, pathway B-3 (catastrophic failure >2 sq ft). The release can be l'iltered by the SBGTS (pathway C-2) before release or by-pass the SBGTS (pathway C-1). Noble gas release fraction is a function of release rate only. Doses at 1 mi are calculated for a 1-hour ground level release, 4 mph, and D stability wind conditions. Whole-body dose includes cloud shine, 3 hours of ground shine, and inhalation. Noble gas decay is accounted for. Thyroid dose (adult) is for inhalation only. (Insert Figures 4.7, 4.8, and 4.9 (NUREG-1228) after this page.] NUREG-1228 25 Appendix A
TA8 E-1 GAP STEP 4 Record from event tree (see Figure 5.6A): Event description Time Analyst Whole-tody dose 9 1 mi , Thyroid dose 9 1 mi STEP 5 d Calculate baseline dose at 2, 5, and 10 mi. ! Whole-body dose from event tree: Whole-body dose: 9 1 mi , 9 1 mi
- 0.40 = 9 2 mi
. 9 1 mi
- 0.09 = 9 5 mi i 9 1 mi
- 0.03 = 0 10 mi
~
Thyroid dose from event tree: Thyroid dose: 9 1 mi ; 9 1 mi
- 0.40 = 9 2 mi '
9 1 mi
- 0.09 = 0 5 mi j 9 1 mi
- 0.03 = 9 10 mi _
1 STEP 6 I Present results. NOTE Because of great uncertainty, dose numbers should not be used in presenting - results. Use the possible consequences as determined below. l ACTION Based on criteria below, determine: 4 l Distance to which early deaths are possible (WB dose >200 rem) i 4 i Distence to which early health effects are possible (WB dose >50 rem)
~
Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
; >5 rem) l J
Assuming that the accident proceeds as projected, the actual releases could j reasonably vary by a factor of 10 to 100 from the projected dose. ] Combine this assessment with the general description (following Step 3), input ' j information (Step 2), and a markup of figure showing assumed release pathway (s). j Provide results to the Radiological f.ssessment Manager. r
- [ Insert Figure 5.5A (NUREG-1228) after this page.]
.i f NUREG-1228 26 Appendix A \. !
TAB E-2 GRAIN BOUNDARY STEP 4 l Record from event tree (see Figure 5.68): 1 Event descriptior.. l Time l Analyst Whole-body dose 8 1 mi Thyroid dose 9 1 mi i STEP 5 t,alculate baseline dose at 2, 5, and 10 mi. - t Whole-body dose from event tree: Whole-body dose: 9 1 mi 9 1 mi
- 0.40 = 9 2 mi t
0 1 mi
- 0.09 = 9 5 mi 0 1 mi
- 0.03 = 9 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi ;
9 1 mi
- 0.40 r 9 2 mi 9 1 mi
- 0.09 = 9 5 mi 9 1 mi
- 0.03 = 9 10 mi '
STEP 6 l Present results, ,' NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION j Based on eriteria belo.1, determine: Distance to which early deaths are possible (WB dose >?M res) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose l
>5 rem) 2 Assuming that the accident proceeds as projected, the actual rtleases could I reasonably vary by a factor of 10 to 100 from the projected dose, i
Combine this assessment wit) the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager.
! (Insert Figure 5.68 (NUREG-1228) after this page.)
NUREG-1228 27 Appendix A :
TAB E-3 CORE MELT STEP 4 Record from event tree (see Figure 5.6C): Event description Tim? Analyst Whole-body done @ 1 mi Thyroid dose 9 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: 9 1 mi 9 1 mi
- 0.40 = @ 2 mi 0 1 mi
- 0.09 = @ 5 mi 0 1 mi
- 0.03 = 9 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi 9 1 mi
- 0.40 = @ 2 mi
@ 1 mi
- 0.09 = @ 5 mi 9 1 mi
- 0.03 = @ 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Lased on criteria below, determi.e: Distance to which early deaths are possible (WB dose >200 rem) __ Distance to which early health effects are possible (WB dose >50 rea) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose >5 rem) Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.6C (HUREG-1228) after this page.] NUREG-1228 28 Appendix A
. n .
_ _-. - _ . - _ . _ _ ~ . . _ - - - - _ _ - _ - _ . - - _ . _ . - . __ ._. ._ l TAB F BWR/PWR CONTAINMENT BYPASS FOR A GAP (1), GRAIN B0UNDAPY (2), OR MELT (3) RELEASE FROM CORE NOTE A general description of the assumptions used in the event trees folle , Step 3. ACTION STEP 2 Gather following plant information: Plant name i Core condition - check i gap release (1300-2000'F) grain boundary (>3000'F) melt (>4500'F) Release path conditions filtered ! not filtered Release rate .'. typical design (0.1%/ day) 100%/ day 100%/ hour Time Analyst ! l STEP 3 I Use event trees, based on information above, to estimate dose and source term, ' Select appropriate event tree based on core condition. ; Tab F-1 Gap ' Tab F-2 Grain Boundary 4 Tab F-3 Melt j ! i l i , i i l i i 1
! NUREG-1228 29 Appendix A i
_ _- _ _ _ _ _ . . _ _ _ __ -. -- . _ . - - _ _ - = _ = .- - BWR/PWR CONTAINMENT BYPASS RELEASE EVENT TREES FOR A GAP (1), GRAIN BOUNDARY (2), OR MELT (3) RELEASE FROM CORE GENERAL DESCRIPTION: The core is uncovered and release fractions for a 1000 MWe plant (Tab G) typical of a gap (1300-2000 F), grain boundary (3000 F), or melt (4500'F) release from the core are assumed. For the vessel melt-through cases (ex-vessel melt) the strontium release fractions could be substantially increased. This could increase the projected whole-body dose by 50% for cases involving releases of large fractions of the non-noble fission products. The release passes through a line that bypasses the containment, shown in Figures 4.5, 4.6, 4.7, 4.8, and 4.9 as pathways B-4/A-4. Particulates and aerosols airborne in the containment are reduced to account for plateout in the line. The NUREG-0956 results were used as the basis for the amount of reduction. Releases from the containment are estimated for 0.1%/ day (typical design leakage),100%/ day (isolation valve seal failure), or 100%/ hour (catastrophic failure). Noble gas release fraction is a function of release rate only. Doses at 1 mi are calculated for a 1-hour ground level release, 4 mph, and 0 stability wind conditions. Whole-body dose includes cloud shine, 3 hours of
- ground shine, and inhalation. Noble gas decay was considered by assuming 1-hour i
decay fer 100%/ hour release case and 6"hour decay for others. Thyroid dose (adult) is for inhalation only. [ Insert Figures 4.5, 4.6, 4.7, 4.8, and 4.9 (NUREG-1228) after this page.] I L i I 5 1 1 ! i j i l 4 i 1 i NUREG-1228 30 Appendix A
TA8 F-1 GAP STEP 4 Record from event tree (see Figure 5.7A): Event description Time Analyst Whole-body dose @ l m1 Thyroid dose 91 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: 9 1 mi 9 1 mi
- 0.40 = 0 2 mi 9 1 mi
- 9.09 = 0 5 mi 9 1 mi * . 03 = 9 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi 9 1 mi
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 0 5 mi 9 1 mi
- 0.03 = 9 10 mi STEP 6 Present results.
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria beluw, determine: Distance to which early deaths are possible (WB dose >200 ren) Distance to which early health effects are possible (WB dose >50 ren) Distance to which 9AG may be exceeded (WB dose >l rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.7A (NUREG-1228) after this page.] i NUREG-1228 31 Appendix A
TAB F-2 GRAIN BOUNDARY STEP 4 Record from event tree (see Figure 5.78): Event description Time Analyst Whole-body dose 9 1 mi Thyroid dose 9 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: 91 mi 0 1 mi
- 0.40 = 9 2 mi 0 1 mi
- 0.09 = 9 5 mi 9 1 mi
- 0.03 = 9 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi 9 1 mi
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 9 5 mi G 1 mi
- 0.03 = 9 10 mi STEP 6 Present results. ,
NOTE Because of great uncertainty, dose numbers should not be used in presenting results. Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rem) Distance to which early health effects are possible (WB dose >50 rem) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
>5 rea) i Assuming that the accident prcceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose.
l Combine this assessment with the general description (following Step 3), ir.put information (Step 2), and a markup of figure showing assumed release pathway (s). Provide results to the Radiological Assessment Manager. (Insert Figure 5.78 (NUREG-1228) after this page.] NUREG-1228 32 Appendix A
l l l TAB F-3 CORE MELT STEP 4 Record from event tree (see Figure 5.7C): Event description Time Analyst Whole-body dose 9 1 mi Thyroid dose 9 1 mi STEP 5 Calculate baseline dose at 2, 5, and 10 mi. Whole-body dose from event tree: Whole-body dose: 0 1 mi 9 1 ml
- 0.40 = 9 2 mi 9 1 mi
- 0.09 = 0 5 mi 9 1 mi
- 0.03 = 0 10 mi Thyroid dose from event tree: Thyroid dose: 9 1 mi 9 1 mi
- 0.40 = 0 2 at 9 1 mi
- 0.09 = 0 5 mi 9 1 mi
- 0.03 = 0 10 mi STEP 6 s
Present results. NOTE b3cause revilts. of great uncertainty, dose numbers should not be used in presenting Use the possible consequences as determined below. ACTION Based on criteria below, determine: Distance to which early deaths are possible (WB dose >200 rea) Distance to which early health effects are possible (WB dose >50 rea) Distance to which PAG may be exceeded (WB dose >l rem or thyroid dose
>5 rem)
Assuming that the accident proceeds as projected, the actual releases could reasonably vary by a factor of 10 to 100 from the projected dose. Combine this assessment with the general description (following Step 3), input information (Step 2), and a markup of figure showing assumed release patnway(s). Provide results to the Radiological Assessment Manager. [ Insert Figure 5.7C (NUREG-1228) after this page ) NUREG-1228 33 Appendix A
i TAB G GENERAL DESCRIPTION OF ASSUMPTIONS
, Tab G.1 Core .*elease fractions
[ Insert Table 4.1) Tab G.2 Core invertory (Ci/MWe) i linsert Table 2.2] Tab G.3 Suom.ary of major PWR release pathway reduction mechanism [ Insert Table 4.7] Tab G.4 Summary of major BWR release pathway reduction mechanism (Insert Table 4.8) Tab G.5 Assumed containment and steam generator tube rupture escape fractions for 1 hr (release rates) (InsertTable4.10]
; Tab G.6 PWR baseline coolant concentration for various levels of core t i damage
- [ Insert Table 4.3]
l Tab G.7 BWR baseline coolant concentration for various levels of core ; i damage (Insert Table 4.4) ; 1 cile dose factors 6 1 Tab G.8 .; (Insert Table 5.1] f i i 1 i I i i I i l l l j 1 I 1 ; i 4
- NUREG-1228 34 Appendix A [
b, _. f
TAB G BASIC METHOD To make a first approximation' of a severe accident source term, follow the event tree method below. (1) Estimate the amount of fission products in the core. (2) Estimate the fraction of the fission product inventory released from the Core. i (3) Estimate the fraction of the fission product inventory released from the i core that is removed on the way to the environment. (4) Estimate the amount of the available fission product inventory actually , released to the environment. (5) Estimate the dose at 1 mi. 1 The event trees estimate source terms by: Source termg = FPIg x CRFgx RDF(gg) x EF 4 for radionuclide i and n reduction mechanisms where I FPIg = element i core or coolant inventory [ assume a 1000 MWe core] - See Tab G.2 CRF i = element i released from core 'ibased element on core inventory damage state)(Tab G.1) in core RDF i = element i after element reduction i before mechanisms reduction mechanisms available available forfor release (Tab G.3 o release I EF
- element i released (Tab G.5) 4 element i available for release The steam generator tube rupture case was somewhat different in that the starting poirt was an assumed coolant concentration (Tab G 6 or G.7) that 1
assumes all the material released from the core is contained in the coolant. Dose at 1 mi is based on MES0 RAD. Tab G.8 shows the dose factors calculated by MES0 RAD that were used to calculate the coses shown in the event trees. D:ses at 2, 5, and 10 mi are based on h which represents the fall off associated with 0 stability (see NUREG-0396). Dose = [ tree non-noble gas release fraction X 100% appropriate (e.g., gap) non-valuable gas dose factor) + [ tree noble gas release fraction X noble gas release fractico X 100% noble gas dose factors). See Tab G.8 for dose factors ; used. l l NUREG-1228 35 Appendix A l 1 l
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l Source Term Estimation During Incident Response To l Severe Nuclear Power Plant Accidents e O.f4.geo.Tco. Lg?tD i
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The various methods used by tne NRC during their response to a reactor accident for estimating radionuclide release to the environment (source tems) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source tems) as a result of nuclear power plant 3ccidents are described. The quantification of these factors based on plant instrumentation also is discuesed. A rangt of accident conditions from those within l the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site l I is presented. The goal is to present a method of source tem estimation that reflects , the current understanding of source tem behavior and that can be used during an event. ' l l
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