ML20106C439
| ML20106C439 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 07/31/1984 |
| From: | Cybulskis P, Denning R, Gieseke J Battelle Memorial Institute, COLUMBUS LABORATORIES |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| RTR-NUREG-1150-2-V2-B.42 BMI-2104-DRFT, BMI-2104-V05-DRFT, BMI-2104-V5-DRFT, NUDOCS 8410240167 | |
| Download: ML20106C439 (258) | |
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1 DRAFT BMI-2104 Volume V Radionuclide Release Under Specific LWR Accident Conditions Volume V PWR-Large, Dry Containment Design (Surry Plant Recalculations) 4 Prepared by J. A. Gieseke, P. Cybutskis, R. S. Denning, M. R. Kuhlman, K. W. Lee, H. Chen i i Battelle Columbus Laboratories Columbus, Ohio 43201 July 1984 i i 8410240167 840731 PDR ADOCK 05000280 P PDR l
DRAFT BMI-2104 Volume V Radionuclide Release Under Specific LWR Accident Conditions Volume V PWR Large, Dry Containment Design (Surry Plant Recalculations) Prepared by J. A. Gieseko, P. Cybulskis, R. S. Denning, M. R. Kuhlman, K. W. Lee, H. Chen Battello Columbus Laboratories Columbus, Ohio 43201 July 1984 Prepared for Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555
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-v ACKNOWLEDGMENTS Battelle's Columbus Laboratories wishes to acknowledge and !
express appreciation for the computer codes made available for this program by Sandia National Laboratory, Battelle's Pacific Northwest Laboratories, and the Kernforschungszentrum Karlsruhe, and for the compu-tations and consultation providad by Sandia and the Oak Ridge National Laboratories. Further, members of the Peer Review Group have contributed significantly to"this effort by providing cannents, suggestions, and informationonvariousreactorhystemsdesign. Thesupport,oftheU.S.'IuclearRegulatoryCommissionisgrate- ' fully acknowledged, and is the untiring leadership of the NRC staff, particularly Mel Silberberg:and Mike Jankowski. The diligent efforts of many Battelle staff members contributed to the preparation of this report. The following list identifies those staff making major contributions: a~ RJ f.rers GT Brooks EP Bryant R Freeman-Kelly CS Jarrett ' H Jordan , , .c' RG Jung DJ Lehmicke HB Neher DR Rhodes PM Schumacher R0 Wootaa. r a f f
'f l
i e
TABLE OF CONTENTS Pale.
- 1. EXECUTIVE
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . 1 -1 Approach .......................... 1-2 The Surry Plant .................... 1 -3 Accident Sequences Chosen for Study .......... 1-3 Computer Codes Used in the Study . . . . . . . . . . . . 1-4 Surnary of Results ..................... 1-6
- 2. INTRODUCTION ........................ 2-1 2.1 Re fe ren ce s . . . . . . . . . . . . . . . . . . . . . . . 2-3
- 3. GENERAL APPROACH ...................... 3-1 3.1 Plant Selection .................... 3-1 3.2 Selection of Accident Sequences ............ 3-2 3.3 Computer Codes Used in the Study . . . . . . . . . . . . 3-2 3.3.1 As s umpti on s . . . . . . . . . . . . . . . . . . . . 3-3 3.3.2 Uncertainty Considerations ........... 3-6 3.4 References . . . . . . . . . . . . . . . . . . . . . . . 3-7
- 4. PLANT SELECTION AND ACCIDENT SEQUENCES ........... 4-1 4.1 General Plant Description . . . . . . . . . . . . . ... 4-1 4.2 Selection Basis and General Description of Accident Sequences ....................... 4-1 4.2.1 Sequence AB (Loss of Coolant Accident Loss of AC Power) ..................... 4-7 4.2.2 Sequence TMLB' (Transient, Loss of Primary System He a t Re mo v al ) . . . . . . . . . . . . . . . . . . 4-8 4.2.3 Sequence 2S D (Small Pipe Break Failure of ECC System) . . . . . . . . . . . . . . . . . . . . . 4-11 4.2.4 Sequence V (Interfacing Systems Loss of Coolant Accident) . . . . . . . . . . . . . . . . . . . . 4-13
- 5. ANALYTICAL METHODS ..................... 5-1 i
5.1 The rmal Hydra ul ic Be havi or . . . . . . . . . . . . . . . 5-1 l 5.1.1 Overall System Thermal Hydraulics: MARCH 2 . . . 5-1 l 5.1. 2 Primary System Thermal Hydraulics: MERGE . . . . 5-5 t I i 0
TABLE OF CONTENTS (Continued) P_ age, 5.2 Radionuclide Release from Fuel . . . . . . . . . . . . . . 5-6 5.2.1 Source Within Pressure Vessel: CORSOR ...... 5-6 5.2.2 Source from Melt-Concrete Interactions: VANESA . . 5-10 5.3 Radionuclide Transport and Deposition .......... 5-12 5.3.1 Transport in Reactor Coolant System: TRAP-MELT . . 12 5.3.2 Transport in Containment: SPARC ......... 5-17 5.3.3 Transport in Containment: NAUA 4 . . . . . ... . . 5-17 5.4 Re fe re n ce s . . . . . . . . . . . . . . . . . . . . . . . . 5-21
- 6. BASES FOR TRANSPORT CALCULATIONS ............... 6-1 6.1 Plant Geometry and Thermal Hydraulic Conditions ..... 6-1 6.1.1 Sequence AB (Hot Leg) . . . . . . . . . ...... 6-2 6.1.2 . Sequence TMLB' .................. 6-13 6.1.3 Sequence V .................... 6-27 6.1.4 Sequence S 2D ................... 6-28 6.1.5 Other Cases Considered .............. 6-51 6.1.6 General Discussion ................ 6-53 6.2 Radionucli de Sources . . . . . . . . . . . . . . . . . . . 6-56 6.2.1 Source Within Pressure Vessel . . . . . . . . . . . 6-56 6.2.2 Sources Within the Containment .......... 6-64 6.3 Re fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . 6-67
- 7. RESULTS AND DISCUSSION .................... 7-1 7.1 I n t ro d u c t i on . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.2 Transport and Deposition in Primary System . . . . . . . . 7-1 7.2.1 RCS Transport and Deposition for Sequence AB ... 7-1 7.2.2 RCS Transport and Deposition for Sequence TMLB' . . 7-10 7.2.3 RCS Transport and Deposition for Sequence -V . . . . 7-17 7.2.4 RCS Transport and Deposition for the S 2D-c (Col d Leg) Seq uence . . . . . . . . . . . ....... 7-27 7.2.5 RCS Transport and Deposition for the S 2D (Hot Leg)
Sequence ..................... 7-34 I i l l l 1 _ dl . - .--
TABLE OF CONTENTS (Continued) M 7.3 Transport. Deposition and Leakage in Containment . . . . . 7-41 7.3.1 AB Sequence . . . . . . . . . . . . . . . . . . . . 7-42 7.3.2 TMLB' Sequence .................. 7-51 7.3.3 V Sequence .................... 7-66 7.3.4 S2D Sequence ................... 7-72 7.3.5 Results for Groups of Reactor Safety Study .... 7-76 7.3.6 General Observations ............... 7-82 7.4 Di s c us s i on . . . . . . . . . . . . . . . . . . . . . . . . . 7-82 I
^
1.0 EXECUTIVE
SUMMARY
l This is Volume 5, dealing with the Surry nuclear power plant, of a seven-volume report of work done at Battelle's Columbus Laboratories to esti-mate the amount of radioactive material that could be released from light water reactor (LWR) power plants under specific, hypothetical accident conditions. To make these estimates, five power plants were selected that represent the major categories of LWRs: three pressurized water reactors (PWRs) and two boiling water reactors (BWRs). Specifications and data from these plants, along with data from laboratory experiments, were input to computer codes designed to describe various conditions prevailing and physical processes occur-ring inside reactor systems, including the containment and other buildings. Ultimately, these computer codes provide an estinate of how much radioactive material would be able to escape to the environment if a specific series of events (an " accident sequence") took place. This volume reports the results of the analysis of the Surry power station, a Westinghouse PWR housed in a large, dry, high-pressure containment building. The specific accident sequences investigated for the Surry plant were selected to represent cases of high risk, severe consequences, and most importantly, a wide range of physical conoitions. The computer codes used to analyze the accident sequences include the new code MARCH 2. The Surry plant was previously examined in this study using an earlier version of the code, MARCH 1.1; those results were reported in Volume 1 of this series. Other power plants included in the study are Peach Bottom BWR (Volume 2); Grand Gulf BWR (Volume 3); Sequoyah PWR (Volume 4); and Zion PWR (Volume 6). The seventh volume will address technical questions raised during peer review meetings sponsored by the Nuclear Regulatory Comission. The possibility of radioactive material being released to the environ-ment has long been the focus of considerable public concern about the safety of nuclear power plants. Since 1962, several major reports have addressed that concern by using computer codes to estimate the release of fission products l (radioactive material produced during reactor operations) to the reactor containment building, and thence to the environment, during a hypothetical severe accident. Although these analyses have improved over the years, in terms of how realistically they describe what happens during a hypothetical 1-1 l l l
l accident, it has not previously been possible to apply the various codes i sequentially to follow the transport of fission products along their flow path from the core to the environment. This limitation resulted in piecemeal, parametric estimates of release. The research results reported here are intended to provide a system-atic, sequential application of the codes as well as to present analyses performed with improved computational procedures.. It is to be recognized that this report describes an analytical approach for estimating radionuclide trans- i port and deposition which incorporates. individual physical and chemical processes or mechanisms. This approach is being evaluated for use in predict- .ing the amount fission product release (the " source term") to the environment for specific reactors and accident sequences. When verified, these prediction techniques are expected to be more specific and perhaps to supersede generic tables of release fractions provided by previous analyses. The purpose of this report is then to: (1) Develop updated release-from-plant fission product source terms for four nuclear power plant designs and for accident sequences giving a range of conditions. The estimated source terms are to be based on consistent step-by-step analyses using improved computational tools for predicting radionuclide release from the fuel and radionuclide transport and deposition. (2) Determine the effects of major differences in plant design and accident sequences on fission product releases. (3) Provide in-plant time- and location-dependent distributions of fission product mass for use in equipment qualification. Approach This study was conducted by selecting specific plants and accident sequences and then using consistent and improved analyses of fission product release from fuel and radionuclide transport and deposition to predict fission product release'to the environment for these specific cases. The approach comprises a sequence of steps; in the combined analysis, the results are 1-2.
l l specific to a particular set of accident conditions, and each step is based on 1 results from analyses of the previous step. j l ! The Surry Plant Surry Unit 1 was selected to characterize large, high-pressure PWR designs. In some respects, Surry is not an ideal choice: it is not especially j large, the subatmospheric design feature is representative of only a handful of plants and its power output is lower than the U.S. average. In another respect, Surry 'is the natural choice because it was the basis of the WASH-1400 study; by choosing to study the same plant, a direct comparison can be made with results of WASH-1400. Although clearly not a main goal, the comparison of results was judged important enough to guide the plant selection. To compensate for discrepancies between Surry and other large, high-pressure PWR designs, analyses have been performed for two accident sequences for the Zion plant. These results are reported in Volume 6 of the study. Accident Sequences Chosen for Study Three criteria were used in selecting accident sequences for analysis. The set of sequences was chosen to include: sequences which had been found to dominate risk in previous risk studies, sequences which would be expected to have particularly severe consequences, and sequences covering a broad range of physical conditions. AB Sequence: e Loss-of-Coolant Accident (LOCA), loss of AC power. e Emergency Core Cooling injection system does not operate. o Depressurization of the reactor coolant system occurs rapidly following the break. l e Containment safety features (containment sprays and containment cooling systems) are not available. e A break in the hot leg is assumed for this study to examine a minimum pathway and presumably minimum fission product reten-tion within the reactor coolant system. I I 1-3
I r TILB' Sequence: e Transient, loss of primary system heat removal. I e System pressure remains high during core heatup (i.e., core uncovery is delayed for a few hours). e High pressure, essentially stagnant flow, a much longer release path, and interactions between core materials and water in the reactor cavity take place. e Containment safety features (containment-sprays and containment cooling systems) are not available. i e Rapid pressure rises in the containment following vessel failure could threaten containment integrity and lead to a potentially large release of fission products. S2D Sequence: e Small pipe break, failure of Emergency Core Cooling system. e Low flow and intermediate pressure occur in the reactor coolant system during the meltdown period. e Containment safety features (containment sprays and containment cooling systems) are operable up to containment failure. e This is the only sequence analyzed in which the effectiveness of containment safety features is examined. V Sequence: I l e Interfacing systems Loss-of-Coolant Accident (LOCA). e Check valves that separate the low-pressure Emergency Core Coolir.g system from the high-pressure primary system fail, o The pathway for release bypasses the protection normally provided by the containment building, so the amount of reten-tion in the reactor coolant system is particularly important. l e This sequence potentially involves a large fission product release to the atmosphere. Computer Codes Used in the Study 1 Our efforts built on previous computer modeling work performed at i Battelle-Columbus, at Sandia, and in the Federal Republic of Germany, and on experimental and model evaluation studies performed at Oak Ridge, EG&G Idaho, Sandia, and Pacific Northwest Laboratories. In addition to the calculations 1-4
performed at Battelle-Columbus, calculations of. thermal-hydraulic behavior and i i fission product release related to molten core-concrete interactions were performed by Sandia. Research efforts specifically directed toward increasing our understanding of fission product release'and transport under severe acci-dent conditions are under way at the laboratories listed above, as well as at cther research installations around the world. Over the next few years, it is expected that considerable progress will be made in this area. Therefore, j i 'this report must be considered as an expression of current knowledge, with the expectation of future validation or modification of the calculated fission product releases. The first step in analyzing accident sequences was to collect plant design data and perform thermal-hydraulic calculations. Thermal-hydraulic conditions in the reactor over time were estimated with the MARCH 2 code, and detailed thermal-hydraulic conditions for the reactor's primary coolant system were estimated with the MERGE code, developed specifically for this program. The time-dependent core temperatures from the MARCH 2 code were used as input to another code developed for this program, CORSOR, which predicts time- and temperature-dependent releases of radionuclides from the fuel inside the reactor pressure vessel. Releases of radionuclides from the interaction of the melted reactor core with the concrete outside the reactor vessel were estimated by Sandia National Laboratories using their computer code, VANESA. Using the MARCH / MERGE-predicted thermal-hydraulic conditions and the CORSOR-predicted radionuclide release rates as input, a newly developed ver-sion of the TRAP-MELT code was used to predict vapor and particulate transport in the primary coolant circuit. Transport and deposition of radionuclides in the containment were calculated using the NAVA-4 code. In Volume 1, analyses were also performed with the CORRAL-2 code to relate the results to WASH-1400 assumptions. The calculations performed in this study were of a "best estimate" type. Whenever possible, input was derived from experimental measurements. Data employed in these analyses include vapor deposition velocities, aerosol deposition rates, aerosol agglomeration rates, fission product release rates from fuel, particle sizes formed from vaporizing / condensing fuel materials, engineering correlations for heat and mass transfer, and physical properties of various fuel, fission product, and structural materials. 1-5
Summary of Results A number of accident sequences that had been analyzed previously in l Volume 1 of BMI-2104 were reanalyzed in this report. The principal difference in the analyses involved the use of the MARCH 2 computer code rather than the MARCH 1.1 code used in Volume 1. In addition, a gradual fuel slumping model was used in the MARCH 2 analyses which is believed to provide a more realistic characterization of thermal-hydraulic conditions in the reactor coolant system-during slumping. It is interesting to compare the results obtained in the current effort with both the earlier Volume 1 analyses and the results from WASH-1400 for the same plant and accident sequences. The release of fission products from fuel during the' period of fuel heatup and melting in-vessel and during ex-vessel attack of the concrete predicted in this study is not mirkedly different from the results in WASH-1400. Essentially all of the volatile fission products are predicted to be released from the fuel in all cases. Retention of fission products during transport through the reactor coolant system was not credited in the WASH-1400 study. In the current study the amount of retention in the reactor coolant system is in many cases substan-tial (e.g., 85 percent of the iodine in TMLB') depending on the thermal-hydraulic conditions in the system. The uncertainty regarding the ultimate fate of fission products deposited within the reactor coolant system is very great, however. Follow-on analyses of decay heating of reactor coolant system surfaces and the reevolution of fission products are in progress and will be reported in an Appendix at a later time. The results of the containment transport analyses performed for this study indicate that the potential for retention in the containment building and secondary buildings is somewhat greater than predicted in WASH-1400. This is partially because of an underestimation of removal processes in the WASH-1400 CORRAL code and partially because of an increased time to containment failure in the current analyses. 1 In general the results of the current study indicate that for many important accident sequences the source term in WASH-1400 is overestimated by approximately an order of magnitude. The effect of a reduction of this magni-tude on estimated risk would be essentially proportional for the risk of latent i 1-6
fatalities (i.e., an order of magnitude) but would be even more dramatic for the risk of early fatalities because of the threshold behavior of this measure. Of the important sequences analyzed, the one in which a major reduc-tion in consequences was not observed is the interfacing LOCA sequence, V, for the assumption of no overlying pool of water at the entry point into the safe-guards building. If there is an overlying water layer, the source term is considerably reduced. Although this is no longer considered a risk-dominant sequence because of steps taken to suppress its likelihood, the existence of credible sequences with potentially large consequences can have public percep-tion and regulatory implications regardless of the risk significance of the sequence. It is important that the reader recognize the magnitude of the uncer-tainties in the "best estimate" release terms presented in this report. The prediction of fission product release has been shown in this study to be sensitive to accide'nt thermal-hydraulics and core melting and slumping behavior as well as to the mechanisms of fission product release, transport, and deposi-tion. The reader is referred to the companion QUEST study being undertaken by Sandia National Laboratory to obtain a perspective on the magnitude of the uncertainties in these results. 1-7
- 2. INTRODUCTION The possibility of radioactive material being released to the envi-ronment from LWRs has long been the focus on public concern about the safety
.of nuclear power plants and the impetus for safety research. Most reactors in the United States were designed, and their sites were chosen, on the basis of l accidental release assumptions described in research report TID-14844.(2.1)
Published in 1962, TID-14844 prescribes a formula for the release of fission products (radioactive material produced during reactor operations) to the reactor containment area during a hypothetical severe accident. Although these assumptions are representative of the state of knowledge at the time, the behavior of fission products has become better understood in the intervening years. In the period 1972 to 1975, the Nuclear Regulatory Commission conducted the Reactor Safety Study to assess the accident risks in U.S. commercial nuclear power plants. The report of that study, known as WASH-1400,(2.2) provided a more comprehensive and physically accurate description of fission product behavior. The amount of fission product release (the " source term") estimated in WASH-1400 has since been used extensively in planning and evaluating reactor operations. The WASH-1400 source term to the environment for accident sequences has had broad implications for operating LWRs--in licensing, emergency plan-ning, safety goals, and indemnification policy. However, additional research continued to provide even better methods for estimating fission product release and transport. In 1981, the Nuclear Regulatory Conunission issued the report
" Technical Bases for Estimating Fission Product Behavior During LWR Accidents",(2.3) a review of the state of knowledge at the time. As part of I the Technical Bases report, the assumptions, analytical procedures, and avail-able data were evaluated, and new estimates were made. One advantage of the new estimates was that they took into account the fact that some radioactive material would be deposited inside the reactor during an accident and would
! therefore not escape from the reactor core to the environment. On the other ! hand, because of the limitations of the computer codes available at that time, ! the new estimates could not follow the transport of fission products along l their flow path from the core to the environment by applying the various codes sequentially. This resulted in piecemeal, parametric estimates of release. l l 2-1 l
The research results reported here are intended to provide this ! systematic, sequential application of the codes as well as to present analyses performed with computational procedures improved since the " Technical Bases" report. It is to be recognized that in this study, an analytical approach was { developed for estimating radionuclide transport and deposition which incorpor-ates individual physical and chemical processes or mechanisms. This approach is being evaluated for use in predicting fission product source terms for release to the environment for specific reactors and accident sequences. When I verified, predictions made with the approach used here are expected to replace the generic tabular release fractions such as those in Table 6, Appendix V of WASH-1400, where release fractions are given for broad classes of accidents. The purpose of this report is then to: (1) Develop analytical procedures and use them to predict updated release-from-plant fission product source terms for four types of nuclear power plants and for accident sequences giving a range of conditions. The estimated source terms are to be based on consistent step-by-step-analyses using improved computational tools for predicting radionuclide release from the fuel and radionuclide transport and deposition. (2) Determine the effects on "ission product releases associated with major difte'ences in plant design and accident sequences. (3) Provide in-plant time- and location-dependent di:tributions of fission product mass for use in equipment qualification. It is not necessarily the intent of this work to produce an all-encompassing definition of source terms, but rather to make best estimates of source terms for a range of typical plants and several risk-significant sequences covering a wide range of conditions. These analyses are to be made with the best available techniques, in a consistent manner, following along-release pathways for fission products, and at a level of detail consistent ! with current knowledge of pertinent physical processes. Based on state-of-the-art techniques, these best-estimate analyses should provide an indication of the conservatisms inherent in current source term assumptions and guidance for the development of new source terms. The analytical methods and 2-2
l corresponding predictions presented here are based on currently available information and are subject to revision and improvement as better analytical procedures are developed and as a more extensive experimental base evolves. The preparation of this report, therefore, is an evolutionary process which will be carried out over a period of_ time, with verification and possibly_ revision of the procedures continuing over several years. As a part of this evolutionary process, it should be noted that this report is expected to be revised using improved analytical procedures and incorporating predictions of possible fission product reevolution during decay heating of deposits. The results of these calculations are to be published as an appendix to this report. 2.1 References (2.1) DiNunno, J. J., et al, " Calculation of Distance Factors for Power and Test Reactors Sites", TID-14844 (March 23, 1962). (2.2) " Reactor Safety Study--An Assessment of Accident Risks in U.S. Com-mercial Nuclear Power Plants", WASH-1400, NUREG-75/014 (October, 1975). (2.3) " Technical Bases for Estimating Fission Product Behavior During LWR Accidents", NUREG-0772 (June, 1981). 2-3
. ~ . _ .. -
- 3. GENERAL APPROACH The general philosophy behind this stu'dy is that mechanistic predic-tions of radionuclide release _and transport are possible if proper modeling is performed to represent the physical and chemical processes occurring during LWR accidents. The study,:then, represents an attempt'to describe in a reason-ably complete:but tractable fashion the processes influencing radionuclide
! release-to the environment for selected plants and accident conditions.
The objectives of this study originally called for a consistent
~
analysis of radionuclide behavior by following fission product transport along flow paths, starting with release into.the core region and ending with final release to the environment. To meet these objectives, numerous decisions and assumptions were required for the analyses: selection of plants and sequences for consideration; choice of analytical tools to be used or upgraded; evalua-tion and incorporation of experimental data; and determination of major physi-cal effects which would be considered on a parametric variation basis to determine the sensitivity of calculations to such variations. Some of the I major considerations will be reviewed and discussed in this section. The general approach in this study was to select specific plants and accident sequences for consideration and then to use consistent and improved analyses of fission product release from fuel, transport, and deposition to predict fission product release to the environment for these specific cases. The approach consists of a series of steps performed in sequence such that in the combined analysis, the results are specific to an individual set of acci-dent conditions, and each step is based on results from analyses of the pre-vious step. 3.1 Plant Selection
- The first major step in the process was the selection of types of
, nuclear power plant designs to be considered and a specific plant to represent '
each type. The types to be considered were: large, dry PWRs; Mark I BWRs; l Mark III BWRs; and ice-condenser containment PWR designs. The specific plants b chosen to represent each type, respectively, are the Surry and Zion, Peach Bottom, Grand Gulf, and Sequoyah plants. These selections were made on a 3-1 , l l
combined basis of typicality of design and availability of design details l needed for analysis. 3.2 Selection of Accident Sequences Accident sequences were chosen for each plant such that significant contributions to risk and a wide range of physical conditions were represented in the analyses. The selected plants and accident sequences are listed below: PWR: Large Dry . Containment PWR: Large Dry BWR: Mark I (Surry-Volumes 1 Containment (Peach Bottom-and 5) (Zion-Volume 6) Volume 2) AB TMLB' TC SD2 SD 2 AE V TW TMLB' PWR: Ice Condenser BWR: Mark Ill Containment (Grand Gulf- (Sequoyah-Volume 3) Volume 4) TPI S2HF TQUV TMLB' TC TML SE 2 The accident sequences for each plant are described in detail in Section 4.2 of the volume of the report dealing with that plant. j 3.3 Computer Codes Used in the Study (
)
Following the selection of plants and sequences, the required plant design data were collected and thermal-hydraulic analyses performed for each accident sequence. Overall thermal-hydraulic conditions on a time-dependent basis were estimated with the MARCH code,(3.1) and detailed thermal-hydraulic conditions for the primary system were estimated with the MERGE code (3.2) developed specifically for this program. The time-dependent core temperatures were used as input to another code developed for this program, CORSOR(3.3), which predicts time- and 1 3-2
t temperature-dependent mass releases of radionuclides from the fuel within the pressure vessel. Releases during core-concrete interactions of radionuclides remaining with the melt were provided by Sandia National Laboratories using their newly developed model, VANESA(3.4), Using the MARCH / MERGE-predicted thermal-hydraulic conditions and the C0RSOR-predicted radionuclide release rates as input, a newly developed ver-sion of the TRAP-MELT code (TRAP-MELT 2)(3.5) was used to predict vapor and particulate transport in the primary coolant circuit. Transport and deposition of radionuclides in the containment were calculated using the NAVA-4(3.6) code. The basic stepwise procedure described above is illustrated in Figure 3.1, which shows the relationships among the computational models. The calculations were of a "best estimate" type using input derived from experi-mental measurements whenever possible. Types of data employed in the analyses include vapor deposition velocities, aerosol deposition rates, aerosol agglo-meration rates, fission product release rates from fuel, particle sizes formed from vaporizing / condensing fuel materials, engineering correlations for heat and mass transfer, and physical prop 2rties of various fuel, fission product and structural materials. 3.3.1 Assumptions In preparation for performing calculations of thermal-hydraulic con-ditions and radionuclide transport and deposition, it was necessary to make a number of assumptions or to select conditions from among several options. Major assumptions used in this study of the Surry plant, using the MARCH-2 code, are listed below in the categories of geometry, thermal hydraulics, mechanisms, and sequences. Geometry (1) Surfaces within the containment building available for radio-nuclide deposition include only the major geometrical features of the building. 3-3
SELECTION OF TYPES OF PLANTS SELECTION OF SPECIFIC PLANTS Y SELECTION OF ACCIDENT SEQUENCES v SPECIFICATION OF PLANT INVENTORY GEOMETRY AND ACCIDENT -------- SEQUENCE PHENOMENA ORIGEN ir l OVERALL THERMAL HYDRAULICS MARCH i y u o PRIMARY SYSTEM RELEASE FROM FUEL THERMAL EYDRAULICS - CORSOR MERGE i
't if CORE-CONCRETE PRIMARY SYSTEM TRANSPORT INTERACTION
--------~~
TRAP-MELT CORCON y
>f CORE CONCRETE RELEASE CONTAINMENT TRANSPORT VANESA NAUA-4 Modified l v
RELEASE TO ENVIRONMENT FIGURE 3.1. INFORMATION FLOW FOR RELEASE, TRANSPORT, AND DEPOSITION CALCULATION 3-4
(2) There is no attenuation of radionuclides as they pass through Lleak paths in the containment shell. Therma'l' Hydraulics-t (3) Before pressure vessel failure, flow in:the_ primary coolant system is restricted to' direct leak paths. (4) The_ upper plenum geometry is modeled in terms of surface areas, steel thickness, and compartment heights rather than with exact geometries. (5) Decay heating of surfaces by. deposited fission products is neglected in the calculations reported in the main body of this volume but is considered in the Appendix (to be added later). (6) No_ operator intervention occurs that would lead to cooling of the steam generators. Mechanisms (7) Neither deposition nor resuspension of radionuclides occurs during rector coolant system depressurization at the time of pressure vessel melt-through. (8) In the long term (after pressure vessel failure), deposited radionuclides remain in the primary system indefinitely. The effects of this assumption are considered in the Appendix (to be added later). (9) No change in fission product physical or chemical properties results from radioactive decay. Sequences (10) In the V sequence, all fission products released from the fuel after pressure vessel failure are assumed to mix in the contain-ment volume prior to flowing back through the reactor coolant system to the environment. (11) Vaporization release is scrubbed by an overlying layer of water in the S 20 sequence. 3-5 i _____._____________a .
. Some of the above assumptions have been relaxed or changed to accommodate best estimates of conditions and occurrences in specific cases. These are dis-cussed in greater detail for each plant in Section 6.1 of the volume of the report dealing with that plant.
3.3.2 Uncertainty Considerations The computation of radionuclide release and transport using mechanis- l tic models is subject to many uncertainties of various magnitudes and importance. Quantitative estimates of uncertainties in individual parameters, and hence the overall importance of such uncertainties, has been outside the scope of this stWy. Where practical, however, qualitative (and in some cases quantita-tive) estimates of uncertainties have been noted. Some of the uncertainties in the analyses and procedures can be iden-tified that are currently considered significant. The following is a list of some uncertainties that are believed significant and warrant further evalu-ation through more detailed analyses: (1) The simplified fuel melting model in MARCH (i.e., a single melting temperature) could bias the predicted release of material from overheated fuel, particularly regarding the source of inert and low volatility fission product aerosols. (2) The rate coefficients for the release of fission products from overheated fuel are empirical, rather than mechanistically based, and rely largely on scaled, simulant experiments. (3) The model for the release of fission products and inert materials during the attack of concrete has a very limited experimental basis. l (4) The flow patterns in the reactor coolant system are uncertain. l The adequacy of the simple thermal-hydraulic models used in l this study will require experimental verification. ! (5) Primary system transport models used in these analyses have not l been validated against integral experiments. (6) The mode and timing of containment failure in severe accident ! sequences can have a major influence on fission product behavior but are subject to large uncertainty. (7) The calculation methods for water condensation in the contain-ment are based on limited, small-scale experiments and require verification at larger scales. 3-6 a
(8) Deposition velocities for vapor species used in the TRAP-MELT calculations were taken as a mid-points in order-of-magnitude ranges of experimental data. More accurate data would reduce the uncertainty in these parameters and in the resulting rates for deposition by scrption. 3.4 References (3.1) Wooton, R. 0, and Avci, H. I., " MARCH (Meltdown Accident Response Characteristics). Code Description and Users' Manual", NUREG/CR-1711, BNI-2064 (October, 1980). (3.2) MERGE Manual. (3.3) CORS0R Manual. (3.4) VANESA Manual.. (3.5) TRAP-PELT 2 Manual. (3.6) Bunz, H., Koyro, M., and Schock, W., "A Code for Calculating Aerosol Behavior in LWR Core Melt Accidents Code Description and Users' Manual". i l I l l 3-7 i 1 l l L ^ , - - - -_
I I I
- 4. PLANT SELECTION AND ACCIDENT SEQUENCES 4.1 General Plant Description The plant selected to characterize pressurized water reactors with large, high-pressure containment designs is Surry Unit 1. In some respects, the selection is not ideal: the Surry plant is an older design, and in comparison with ave age parameters for U.S. plants, its power output is low, its containment volume is small, and its containment design pressure is low.
However, the Surry plant was chosen because its design was analyzed in the Reactor Safety Study, and, a direct comparison of our predicted source terms and those predicted in the Reactor 4fety Study was judged important enough to guide our selection. Important differences between Surry and the " ideal" choice have been examined by parametric variation. The Surry unit is a pressurized water reactor (157-inch diameter vessel with 157 fuel assemblies) designed by Westinghouse. A detailed descrip-tion of plant data is provided in Table 4.1. The reactor is designed to operate at a nominal power of 2441 MW(t) and a reactor coolant system pressure of 2250 psia. The containment is a steel-lined reinforced concrete structure with a free volume of 1.8 x 106 ft ;3it is operated subatmospheric with initial pressure and temperature of 10 psia and 100 F. Figure 4.1 illustrates the layout of the containment. 4.2 Selection Basis and General Description of Accident Sequences The four accident sequences selected for analyzing the large, high-pressure containment PWR plant design were AB, TMLB', 52 0, and V. (Table 4.2 relates the letter used in the accident identifier to the type of event and to the failure of the engineered safety systems.) Two criteria were used in selecting sequences. First, the selected sequences were to be risk dominant for a number of design variations within the large high-pressure containment category of PWRs. Second, these sequences were to cover a range of accident
- conditions and engineered safety system performance within the reactor coolant system and containment building.
i I 4-1
t TABLE 4.1. PWR DATA 1 Nominal power 2,441 MW(t 8.331 x 10 Btu /hr , Internal energy of water 246.9 x 10 Btu 2.66 x 1010 kg-m) i' Sensible heat in the core 16.35 x 10 Btu 1.76 x 109k Total water in the system - 423,200 lb 192,000 kg) g-m) Avg temperature (Excl. pres.) 571.8 F 300C) Pressure 2280 psia 15.7 MPa) Reactor coolant system volume 10,370 (293.7) , Pressurizer volume, total 1,336 ft3 (37.83 m3) water 816 ft3 (23.1.m3 steam 520 ft3 (14.7 m3) Three accumulators, total volume 4,350 ft3 (123.2 m3) water volume 2.775 ft3 78.58m3) pressure 665 psig 4.585 MPa) temperature 120 F 48.9C) Containment recirculation spray 2 systems, flow each 3,500 gpm 220.81/s) Containment free volume 1.8 x 106 ft3 5.1 x 104 m) Initial temperature 100 F Initial pressure 10 psia 3.778 6.89 x C) 104Pa) l-Dewpoint 80 F 26.7 C) Primary system hot metal 1,686,285 lb (766,000 kg) Temperature 572 F (300C) Containment Heat Sinks Thickness Area Walls inside containment 1.0 ft (0.3048 m) 3,320 ft 2 (308.4 m2 ) Walls inside containment 2.0 0.6096 27.600 2564) Walls inside containment 3.0 0.9144 19,400 1802) Walls inside containment 4.0 1 .21 9) 5,000 464.5) Walls inside containment 6.5 1 .981 ) 2,100 1 95.1) Containment wall 4.5 1.372) 46,747 (4343) Dome 2.5 0.762) 25,000 (2323) Floor above foundation mat 2.0 0.6096) 11,250 (1045) Foundation mat 10.0 3.048) 11,250 (IMS) Containment liner Walls 0.38 in. (0.9652 cm) 46,747 (4343) Dome 0.50 in. (1.27 cm) 25,000 (2323) Floor 0.25 (0.635cm) 11,250 (1045) Miscellaneous metal - 1,200,000 lb (544,308 kg) Core Equivalent diameter 119.7 in. (3.04 m) Active height 144.0 in. (3.658 m) Total cross sectional area 78.3 ft2 (7.27 m2) 4-2
i TABLE 4.1. PWR DATA (CONTINUED) l Core (Continued) No. of fuel assemblies 157 Rods per assembly 204 Pitch 0.563 in. (1.430 cm) Assembly dimensions 8.426 in. square l Fuel rod diameter 0.422 in. (1.072 cm) l Clad (Zr-4) thickness 0.0243 in. (0.0617 cm) Total number of fuel rods 32,028 Core weight 226,200 lb (102,700 kg) UO2 175 600 1b (79,820kg) Zircaloy 36,$00 lb (16,5C0 kg) Misc. 14,300 lb (6,500kg) Fuel pellet diameter, Region 1 0.3669 in. (0.9319 cm) 2 and 3 0.3659 in. (0.9294 cm) l Fuel pellet length 0.6 in. 1.524 cm) Diametral gap, Region. 0.0065 in. 0.01651 cm) 2 and 3 0.0075 in. 0.01905 cm) Fuel density, Region 1 94% 2 92 3 91 Fuel enrichment, Region 1 1.85 w/o 2 2.55 3 3.10 No. of grid spacers 7 Neutron adsorber Ag-In-Cd Clad 304 ss Clad thickness 0.024 in. (0.06096 cm) l No. of control assemblies 53 Full length 48 Part length 5 l Rods per assembly 20 1 Burnable poison rods 816 No. per assembly 12 No. of assemblies 68 Material Borosilicate glass 0.D. 0.4395 in. (1.116 cm) I.D. 0.2365 in. (0.6007cm) Clad 304 ss Boron (natural) loading 0.0429 g/cm Reactor vessel I.D. of shell 157 in. (3.99 m) Belt line thickness (w/o clad) 7.875 in. (0.2000m) Head thickness 5.0 in. (0.127 m) Clad thickness 0.125 in. 0.3175 m) Overall height 40 f t-5 in. ((12.32 m) 4-3 l l
TABLE 4.1. PWR DATA (CONTINUED) Reactor vessel (continued) Inlet nozzles 27.5 in. (0.699 m) tapered to 35.4 in. (0.899 m) Outlet nozzles 29 in. (0.737 m) l Water volume with core and internals in place 3,718 ft3 (1.053 x 102 m3) Core barrel I.D. 133.9 in. (3.401m) 0.D. 137.9 in. (3.503 m) Thermal shield I.D. 142.6 in. (3.622 m) 1 0.D. 148.0 in. (3.759 m) Safety Injection Charging Pumps Number 3 Design pressure, discharge 2750 psig (18.96 MPa) Design pressure, suction 250 psig 1.724 MPa) Design temperature 250 F 121 C) Design flow 150 gpm 9.46 1/s) Maximum flow 600 gpm (37.81/s) Design head 5800 ft (1768 m)
< Low Head Safety Injection Pumps Number 2 Design pressure, discharge 300 psig 2.07 MPa)
Design temperature 300 F 148.9 C) Design flow 3000 gpm 189.2 1/s) Design head 225 ft (68.58 m) Maximum flow 4000 gpm (252.31/s) Containment Spray Pumps Number 2 Design flow 3,200 gpm (201.91/s) Design head 225 ft (68.58 m) Design pressure 250 psig (1.724 MPa) Recirculation Spray Pumps Inside Containment Number 2 Design flow 3500 gpm (220.81/s) Design head 230 gpm (14.51/s) Recirculation Spray Pumps Outside Containment Number 2 , Design flow 3,500 gpm (220.81/s) l Design head 249 ft (75.89m)
- Recirculation Spray Coolers l Number 4 Design duty, each 55,534,520 Btu /hr (16.3 MW) l Refueling Water Storage Tank l Volume 350,000 gal (1.32 x 106 1) l Boron concentration 2,500 ppm l Design pressure Hydraulic head l Design temperature 150 F (65.6 C)
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TABLE 4.2. KEY TO PWR ACCIDENT SEQUENCE SYMBOLS A - Intermediate to large loss of coolant accident (LOCA). B - Failure of electric power to engineered safety features (ESF). B' - Failure to recover either onsite or offsite electric power within about 1 to 3 hours following an initiating transient which is a loss of offsite AC power. C - Failure of the containment spray injection system. D - Failure of the emergency core cooling injection system. F - Failure of the containment spray recirculation system. G - Failure of the containment heat removal system. H - Failure of the emergency core cooling recirculation system. K - Failure of the reactor protection system. L - Failure of the secondary system steam relief valves and the auxiliary feedwater system. M - Failure of the secondary system steam relief valves and the power j conversion system. Q - Failure of the primary system safety relief valves to reclose after opening. 1 R - Massive rupture of the reactor vessel. 51- A small LOCA with an equivalent diameter of about 2 to 6 inches. 52- A small LOCA with an equivalent diameter of about 1/2 to 2 inches. T - Transient event. V - Low pressure injection system check value failure. Containment Failure Modes: a = steam explosion s = containment isolation failure Y = overpressurization due to hydrogen combustion 6, = early overpressure failure due to steam and noncondensible gases 6; = delayed overpressure failure due to steam and noncondensible gases e = basemat melt-through 4-6
i Although the large, high pressure PWR containment design is often j referred to as " dry", a great quantity of steam would be released to the l
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I containment building in each of the sequences analyzed (except V, in which steam would be released to the safeguards building). In these cases, steam condensation on the walls and on aerosols can enhance the natural removal of radionuclides from the containment atmosphere. The containment spray systems would'be even more effective, in those sequences in which they operate; sequences were selected to illustrate the performance of the containment system with and without spray operation. In the Three Mile Island accident, the release of radionuclides to the containment atmosphere was significantly limited by a large quantity of dater in the pathway of release to the environment. This is not characteristic of accident sequences leading to complete core melting of the type selected for the present analysis. In the sequences studied here, fuel heatup would not begin until the water level had dropped below the top of the core, and very hot steam and hydrogen from the melting core would superheat the struc-tures in the pathway to the containment. There are large variations in the volume and design pressure of large high-pressure containments. To some extent, generic accident sequences can be defined which are independent of variations in plant designs; for example, the sequence AB- 6 (large LOCA, loss of all AC power, and failure of containment by overpressurization) could occur in any PWR, because the general safety philoso-phy and safety functions provided to protect the plants are the sa.ae. Different vendors and architect-engineers have chosen different approaches to satisfy l these safety functions, however, so the likelihood of each sequence may vary greatly between plants. Similarly, the timing and order of events in an acci-1 dent sequence may also vary, deoending on the design. 4.2.1 Sequence AB (Loss of Coolant Accident, Loss of AC Power) A large pipe break accident was selected for analysis because it represents one extreme of reactor coolant system conditions during core melt-down. Depressurization of the reactor coolant system is expected to occur rapidly following the break. If all AC power is lost, the accumulators discharge into the v'essel to supply some emergency coolant, but the pumped ECC 4-7
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./t' wayTand, presumablys minimum fission product'etention r within the reactor-coolant l system.--'The flow path for' fission l products from the core,--to the supper plenum,andtothehotlegbrakisillustrated,ikfigurel4.2. 'In this. sequence,
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' vaporization release per.iod-is shown .in Figure 4.3.
1 In terms of reactor cool'antiystem response, there11silittile' difference b'etween AB and AD'(involv'in'g failure of pumped ECC injection rather O than-loss of all'AC power).- The-conbainmentconditionsfor'AD,would-bevery
- similar to those-of'S D, 2 whichlwas aldo' analyzed. ,It-was -felt that- AB provided ,
an opportunity'to examine a numt,er ofscontainment failure modes in which fission products would be retained by natural deposition./p , Three failure modes were*esamined for the:AB sequence: containment
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s isolation failure (8),. early overpressure failure resulting from hydrogen combustion (y), and'basemdf melt-throbgh failure (c). For 'the Surry plant, the. probability of isolation failureits quite small,p ince the plant's'operat-ing containment pressure is significantly )ess ,than atmospheric; for other PWR designs, however, this may- be a major potential failure mode. The conditions leadingtoearlyoverpres"arefailureare/htexpectedto'occurinAB,but because of'the potentially large conseqlences of this failure mode and the uncertainty in the processes associ'ated with it, the associated fission = product release was evaluated. ' Delayed overpressure failure and melt-through of the l basemat,without aboveground failure (c) are the most likely modes of contain-ment failure. Because of the long delcy to' failure, the consequences of these two modes of failure a're' expected to be similar and to differ primarily in the ow magnitude of noble gas release. 4.2.2 Sequence TMLB' (Transient, Loss of Primary System Heat Removal) s 'S The TP4' sequer.ce was found to be a major risk contributor in' WASH 1400. ~The predicted release fractions for the containment overpressurization y 4-8 ,I + u M .- , .. ._ __ a
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r_ ___ __-- _ -- _____ -----____---- t I failure mode for this sequence were used to characterize release category PWR
- 2. The reactor coolant system behavior of TMLB' contrasts sharply with AB because the reactor coolant system pressure remains elevated (-2500) during core heatup and fission product release. The beginning of core uncovery is also delayed for a few hours, providing some time for the decay heat power to be reduced.
The flow path for fission products through the reactor coolant system is illustrated in Figure 4.4. Prior to core uncovery, the water in the pressur-izer is predicted to be carried from the pressurizer with the steam flow or to
. fall back into the hot leg during periods when the relief valve is closed. If the reactor pump seals were to fail in this sequence, the results are expected to be similar to a cold leg break.
As in AB, containment sprays are assumed to be inoperable in this sequence. One reason for selecting THLB' was to investigate the effect of containment failure time on fission product retention. Early and late failure of the containment by overpressurization (6, c) are considered. Although early failure is considsered unlikely, it was evaluated because of its potentially large consequences. Interaction between molten core material and accumulator water in the reactor cavity could lead to early failure of the containment. The magnitude of the ensuing steam pressure spike depends on mixing and heat transfer processes that are uncertain. 4.2.3 Sequence S2 D (Small Pipe Break, Failure of ECC System) Because containment cooling and the containment spray systems are operable in this sequence, the expected release of fission products to the environment would be quite small. As a result, the contribution to the predicted public health risk would also be expected to be small. This type of event is believed to be comparatively likely, however, and is of interest for this reason. Also, this is the only sequence analyzed in which the effective-ness of the containment safety features is examined. The possibility of hydrogen combustion is of particular interest in this case because steam inert-ing will not be present as in other sequences. Core meltdown occurs with elevated reactor coolant system pressure, as in TMLB', but at slightly lower 4-11 s
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- 5. ANALYTICAL METHODS This section describes the analytical methods used in reassessing the source term to the environment for'the Surry plant, a PWR with a large, high pressure containment. The methods employed here differ significantly from those used to analyze the Surry plant, as described in Volume 1 of this report (5.1) ,
The first major difference between the methods used here and those described in Volume 1 is that the MARCH 2 code (which-includes an improved model for core slumping) is used here for the overall thermal-hydraulic calcu-lations, replacing the MARCH 1.1 code used for Volume 1. The second major change is that in the present analysis, an accounting is made for the effect of Zircaloy cladding oxidation on the release rate for tellurium. Finally, a new code called SPARC-is used here to predict the retention of aerosols in suppression pools. Some other, less significant changes have been made as well. These are discussed in the text. 5.1 Thermal Hydraulic Behavior This section describes the computer code MARCH 2, which, along with the MERGE code, was used to analyze the thermal-hydraulic response of the reactor core, the primary coolant system, and the containment system for the selected accident equences. 5.1.1 Overall System Thermal Hydraulics: MARCH 2 The MARCH 2(5.2) (Meltdown Accident R_esponse C_H_aracteristics) computer code describes the physical processes involved in severe fuel-damage accidents in light water reactors. Version 2 of the code replaces Version 1.1.(5.3) The differences between the two versions include changes in models, code struc-ture, and programing language. The new models in MARCH 2 were developed at a number of institutions, including Battelle, Sandia National Laboratories, Oak Ridge National Laboratories, Brookhaven National Laboratory, and the Tennessee Valley Authority. In many cases, these models are provided as options to 5-1
existing models. .The changes in MARCH were largely undertaken to address recognized deficiencies in the early version related to modeling approximations, time-step control, and transportability of the code to other installations. The MARCH 2 code was developed primarily for use in probabilistic risk assessment. The uncertainties in many of the MARCH 2 models are large, and in many cases the extent to which the models have been validated against experiments is limited. More mechanistic codes are being developed by the NRC, such as SCDAP and HECTR, but they were not available for use in this program. The MARCH 2 code examines the behavior of a large variety of accident processes including depressurization or leakage from the reactor coolant system, core uncovery, core heatup, oxidation of Zircaloy cladding, fuel melting, fuel slumping, fuel-coolant interaction in the lower vessel head, vessel head fail-ure, fuel-coolant-interaction in the reactor cavity, debris bed ccolability, core-concrete interactions, production of combustible gases, gas combustion in the containment, containment heat transfer, intercompartment flows, and the effect of engineered safety features on containment thermal hydraulic behavior. Some of the principal modeling improvements in Version 2 of the MARCH code are described below. 5.1.1.1 Containment Response. The containment response modeling in MARCH 2 includes the following principal changes: provision for expanded blow-down input via subroutine INITIAL, the ability to accept two input terms from the primary system, completely revised treatment of burning of combustibles, addition of a heat sink for radiation heat transfer from the debris in the reactor cavity, and removal of a number of restrictions in the earlicr code. The expanded blowdown table input capability is intended to facili-tate the interfacing of the MARCH cade with more detailed thernal-hydraulic codes that may be used to describe the initial portion of the accident sequence. The containment response subroutine, MACE, has been changed to accom-modate simultaneous break and relief / safety valve flows from the primary system. The two inputs can be directed to different compartments if desired, e.g., break flow to the drywell and relief / safety valve flow to the suppression pool of a BWR. 5-2
5.1.1.2 ' Primary System Response. 'The MARCH 2 treatment of the primary system includes both improvements in the treatment of initial (early) primary system response ~and the addition of several~phenomenological models to treat the processes following core collapse:into the bottom head.' Included are changes in the steam generator model to remove some of the restrictions -and limitations of the earlier version, improved break flow models, changes in
~
the flashing model in response to primary system pressure changes, provisions for simultaneous break.and relief / safety valve flow, changes in the treatment of heat transfer to structures, and consideration of the transport of fission products within the primary system.- 5.1.1.3 Water and Steam Properties. The representation of the properties of water and steam has-been improved in MARCH 2. This has included expansion of the property tables and correlations incorporated in the code as well as inclusion of additional properties required by the new phenomenological models. The input parameters are based on the ASME steam tables. 5.1.1.1 Decay Heat. MARCH 2 incorporates the current American National Standard (b*4) for evaluating fission product decay heating as a function of time after shutdown and time at power, including the contributions from heavy element decay. This replaces the earlier, simplified version incorporated in MARCH 1.1. Alternatively, decay heat as a function of time may be input in tabular form; this approach would be particularly appropriate for transients with failure to scram, where the power history would be provided by more detailed system codes. 5.1.1.5 Core Heat Transfer. MARCH 2 retains the basic model of the core as developed for the earlier version, but incorporates additional models for a more detailed treatment of heat transfer processes. Heat transfer between the fuel rods and the steam-hydrogen gas mixture is now calculated using either the full Dittus-Boelter correlation (5.5) for turbulent flow or a laminar flow correlation. A subroutine has also been added to approximate axial conduction heat transfer in the fuel rods using the Fourier law of heat conduction and the-80Il-calculated node temperatures. The effect of axial and radial thermal radiation heat transfer within the core, as well as between the core and 5-3
surrounding structures and water surfaces, can now be calculated. The heatup of the core support-barrel by thermal radiation is included. Additional changes 1 include corrections in the heat transfer analysis of partially covered core l nodes and improvements in the metal-water' reaction model. j 5.1.1.6 Core Debris. A number of phenomenological models have been added for the treatment of the core debris in the reactor vessel bottom head. These include a flat plate critical heat flux model, a fragmented debris-to-water heat transfer correlation,.and several options that consider formation of debris beds within the vessel head while water is still in the vessel. The
. bottom head heatup model utilizes a calculated heat transfer coefficient between the molten debris and the vessel head.
A major area of concern and controversy in the analysis of core melt-down accidents has been the behavior of core and structural debris upon contact } with water in the reactor cavity. The highly simplified models of MARCH 1.1 have been supplemented with a flat plate critical heat flux model, a part.culate heat transfer model with more mechanistic heat transfer coefficients, and several debris bed heat transfer correlations. If desired, the switchover from one model to another can be based on calculated conditions, e.g., debris temperature. The production of hydrogen from steel-water reactions has been incorporated into these models in addition to the zirconium-water reaction previously avail-able. Also included are the heating of the evolved gases by the debris beds and the effect of hydrogen flow on bed floodings. A heat sink has been provided for the thermal radiation from the top of the core debris as calculated by the INTER subroutine. The decomposition of concrete due to radiated heat flux is treated by an ablation-type model with the resulting gases added to the containment atmosphere. Also, the geometry of the corium-concrete mixture is fixed following solidification of the melt. 5.1.1.7 Burning in Containment. The treatment of combustible gases now includes consideration of the burning of hydrogen and carbon monoxide if { their concentrations exceed flammability limits. Included are explicit consi-derations of inerting due to high steam concentrations and oxygen depletion, ; I direction-dependent compositions for flame propagation between compartments, and burn velocities as functions of composition. Various options are available 5-4 l l
to explore the-effects assumptions about the burning of hydrogen and carbon monoxide. 5.1.2 Primary System Thermal Hydraulics: MERGE When the MERGE (5.6) code was written, the existing computer codes describing the thermal-hydraul h behavior of'a core meltdown accident were not capable of analyzing the flow and temperatures in the individual volumes of the reactor coolant system downstream of the core in the pathway for release to the containment. The report " Technical Bases for Estimating Fission Product Behavior During LWR Accidents"(5.7) , published by the NRC in 1981, indicated that in at least some accident sequences, the retention of fission products in the reactor coolant system (RCS) could be significant. To support more realis-tic analyses of fission product retention with the TRAP-MELT code discussed in Section 5.3.1, an effort was undertaken to write a simple stand-alone code, MERGE, to predict gas temperature, surface temperature, and flow within the reactor coolant system. MERGE calculations are based on the output of MARCH, and the output of MERGE is input to the TRAP-MELT code. The MARCH results used by MERGE are: the primary system pressure, the flow rate of hydrogen leaving the core, the flow rate of steam leaving the core, and the average temperature of gases 1:av-ing the core. The MERGE analysis accounts for conservation of energy and conservation of mass by species. It is assumed that the gases within a volume are well mixed and have the same temperature, and that the pressure differen-tial between volumes is negligible. In MERGE, the equations are solved with an explicit time difference scheme. At a particular time step, conditions within the first volume down-stream of the core are calculated first, and the solution proceeds from each volume to the next downstream volume. Knowing the initial and inlet flow conditions for each volume, MERGE solves for the value of the outlet flow from the volume that yields the known pressure. Heat transfer from flowing gas to structures is accounted for. Forced laminar, forced turbulent, and natural convection heat transfer coefficients are utilized as appropriate, with a radiative term added to the coefficient. In addition, the MERGE-calculated 5-5
radiative heat transfer from the core to the first structure is calculated , based on a MARCH-calculated radiative flux. The MERGE code involves certain approximations and limitations. In { the MERGE analysis, the flow of gases in the upper plenum is assumed to be one-dimensional; in reality, circulation patterns would more probably be established in this region due to the strong temperature gradients. Whether a more detailed analysis is required for this region must be determined by the results of sensitivity studies with the TRAP-MELT code. The need for validation experiments must also be evaluated. 5.2 Radionuclide Release from Fuel 5.2.1 Source Within Pressure Vessel: CORSOR CORSOR(5.8) is a simple correlative code which estimates aerosol and fission product release rates from the core during the period of core melting in a light water reactor. Quantifying the aerosol and fission product release , from the core region is an important first step in determining the radionuclide ) source term to the containment during a hypothetical severe core damage acci-dent. The timing of the release of various materials influences their reten-tion in the reactor coolant system because it determines which species emanat-ing from the core will be able to interact. The timing also determines the l residence time of the released materials and the temperatures in the reactor ' coolant system, since these are both dynamic parameters. Simplistic source terms, such as constant or linearly increasing release rates with concurrent releases for all radionuclides, may therefore lead to unrealistic estimates of radionuclide transport behavior. The reassessment of the Surry plant includes the most recent adjust-ments of CORSOR. In comparison with the initial assessment of Surry reported in Volume 1 of this report, this analysis uses a refined method for calculating tellurium release from fuel and a physically-based mathod for calculating aerosol release from silver-indium-cadmium control rods. For the present analysis, the core has been divided into 240 nodes, 10 radial and 24 axial, which have distinct temperatures as predicted by MARCH. 5-6
The core inventory, determined from the program ORIGEN(5.9) , has been divided equally among the nodes. In an actual reactor, the distribution would vary both. axially and radially and would change with time. Typically, fuel is
. shifted.between three radial' zones during its irradiation history. To flatten the power distribution across the core, the freshest fuel is placed in the outside zone of;the core and the most. highly burned-up fuel is placed in the central region. Thus,'an abrupt change in the spatial distribution of radio-nuclides occurs at the time of refueling but then continues to shift during the cycle.as the fissile inventory is preferentially depleted-in the regions of higher flux.
Alternative distributions of fission products can be used in the CORSOR program, and the effect on fission product release rates of the " flat-flux" assumption can be quantitatively assessed by examining the results of parametric studies.such as those described in Appendix B of Volume 1(5.1) , Uncertainties in the release rate coefficients are expected to have a more significant effect on release rates than will the assumptions regarding fission product distribution among core regions. Temperatures at each of the nodes are obtained from the MARCH code for each of a number of time steps, beginning at the start of the accident and continuing to a user-specified time. An average temperature is computed over each time span during core heatup and melting, and if the temperature is less than 900 C for any node, no release will occur from that node. The average temperature for failure of the cladding of a fuel rod is taken to be 900 C.(5.7) The sensitivity of CORSOR release estimates to the temperature set for cladding f ailure was also discussed in Appendix B of Volume 1.(5.1) When any axial position in a fuel bundle achieves a temperature of 900 C, CORSOR calculates a gap release of certain volatile fission products for all fuel rods in that radial zone. This is intended to simulate the gap release accompanying the bursting of individual fuel rods. This release occurs because certain fission products accumulate in the fuel-cladding gap because of migration within the fuel. The amount of the gap release is taken to be 5 percent of the initial amount present for cesium, 1.7 percent for iodine, 3 percent for the noble fission gases. 0.01 percent for tellurium and antimony, and 0.0001 percent for barium and strontium. Since this emission is very small in comparison with the melt release, and is concurrent with the melt release, it is not treated separately in any of the transport analyses. Clearly, the gap release would 5-7
require more careful analysis if less severe hypothetical accident conditions were considered. Subsequent mass release as the nodes progress toward melting is calculated on a nodal basis as the product of the amount of each species remaining, the release rate coefficient, and the time interval of integration. The mass released is then summed over all the nodes in the core for each species to give the-total mass released during the time step. It should be noted that the MARCH code predictions for core temperatures do not take into account the heat of vaporization of materials released from the core. The computation of the fractional release rate coefficients for fis-sion products is based on empirical correlations derived from experiments performed by Lorenz, Parker, Albrecht, and others.(5.ES.M) The data from these experiments were graphed and curves developed of the releases. A fractional release rate coefficient, K(T), is derived for species by fitting an equation of the form K(T) = Ae8T to each of these curves. The resulting values of A and 8 for three different temperature regions of the graph are given in Table 5.1 and are basically the same as those defined in Appendix 8 of the " Technical Bases Report"N*0) but have, in many cases, been adjusted to account for updated evaluations.(5.17) It should be noted that the fractional release rate is a function of temperature and elemental species only, and any effects of pressure and specific surface area of the melt on the release rate are not considered. Additionally, details of complex phase interactions of various components within the melting core are, for the most part, not known quantitatively; hence the release rates are valid only to the extent that the experiments upon which the release rates are based adequately modeled a core meltdown situation. The release rate coefficients used in CORSOR in the analyses presented in this volume are similar to those presented in Volumes 1-3, except that the method for calculating tellurium release from the fuel elements has been refined. Tellurium release is dependent on the extent of oxidation of the zirconium cladding. In the equation shown above, depending upon the local extent of s zirconium oxidation, different A values are used. For a node where zirconium oxidation is nearly complete, one A value is used, ar.d for a node at which 5-8
zirconium oxidation is less than some specified fraction, the A value is lowered by a factor of 40. For the two-side oxidation of zirconium fuel-rod cladding, the extent of oxidation used as the delimiter between higher and lower release rates is 0.70 in CORSOR. For one-side oxidation this figure would be 0.96-1.0. It is recognized that this is a simplistic treatment of a complex process, but it is believed to be the best representative of the limited available information. Several uncertainties associated with the CORSOR predictions must be mentioned. These uncertainties most strongly impact the predicted release rates for aerosols, rather than for the more volatile materials. One difficulty in predicting aerosol release is that as core melting progresses, the tempera-tures increase throughout the core until, eventually, a loss of geometry would be expected to occur. In the BWR analyses, core slumping occurs in such a way as to remove radial regions from the core in an incremental fashion. Thus, the emission of fission products from the fuel rods which have fallen to the lower structures is included in the calculations of the source to the primary system. This represents a change from the PWR analysis reported in Volume 1, in which emission into the primary system was' halted at the time of core slump-ing. A further difference between the two sets of analyses is that for the BWR sequences, significant periods of time elapse between vessel dryout and bottom head failure. Thus, during a portion of the melt period, the core is emitting fission products into an essentially stagnant volume. The behavior of the control rods during core melting is still a source of uncertainty with respect to aerosol generation, as it was in the first three volumes of this report. In this analysis, however, a physical model for the release of control rod materials has been adopted, based on the recommendations of ORNL. The calculations proceed as follows: silver-indium-cadmium control rods in a radial zone are assumed to fail when, at any axial node within the radial zone a temperature of 1400 C is reached. At this time, 0.05 of the silver, 0.05 of the indium, and 0.50 of the cadmium are released from the control rods. The masses of these species released increases linearly (regardless of heating rate) so that when the melting point (2300 C) is reached, 0.50 of the silver, 0.15 of the indium, and 0.80 of the cadmium has been released. For temperatures greater than 2300 C, the mass released increases linearly with temperature, so that at 2800 C the balance of the local inventory has been 5-9
released from the rod. While this release model is physical, there is no adequate set of experimental observations against which its predictions can be objectively assessed. One further point regarding the calculation of release rate coeffi-cients should be noted. During core melting, the MARCH code predicts some core nodal temperatures above the U02 melting point (approximately 2880 C)(5.19) which are not regarded as being realistic. Using these high values in the' expression for the release rate coefficients would lead to excessively high estimates of release rates for the lower-volatility materials. Thus, the release rates calculated in this work are calculated using a tempera-ture value of 2760 C in place of any values predicted by MARCH in excess of this value. The " Technical Bases Report"(5.8) states that the melting point of UO2 may be lowered by up to 300 C with the addition of Zr02 , and even lower with other compounds, such as control rod material. Thus, it is not yet clear what this maximum achievable temperature should be. The sensitivity of the CORS0R predictions to uncertainties in the maximum temperature, the core temperature, and the release rate coefficients used in the calculations is discussed in Appendix B of Volume 1.(5.1) 5.2.2 Source from Melt-Concrete Interactions: VANESA l The release of fission products and nonradioactive aerosols during i i the interaction of molten core materials with concrete plays an important role in determining the risk of severe reactor accidents and is modeled with the VANESA code. Aerosol production and fission product release from core debris outside the reactor vessel can persist for many hours. The aerosols produced in this way do not usually have to traverse a convoluted pathway before they enter the reactor containment as do aerosols produced in the reactor vessel. The increased inventory of aerosols in the reactor containment brought on by ex-vessel core-debris interacthns could lead to rapid agglomeration and settling of the condensed fission products released during the in-vessel phases of an accident. If containment failure is delayed, the primary source of radioactivity released to the environment would come from ex-vessel sources. Release of fission products from core-concrete interactions can com-pensate for any inhibition in the release of volatile species during the 5-10
in-vessel _ phase of an accident because gases from the thermal decomposition of concrete sparge through the melt and drive the release processes. Ex-vessel
. processes can also lead to the release of fission product elements that are ordinarily quite refractory. .This, again, is because of the strong driving force produced by gas sparging and the unusual melt chemistry that arises
~
during ex-vessel interactions of core debris with concrete. Also of importance is the generation of aerosols.from nonradioactive materials, such as concrete and. steel, during ex-vessel interactions. The additional concentrations of suspended particulates in the containment brought on by these aerosols naturally mitigate the inventory of radioactivity released from the fuel that would tnen be available for release to the environment. This additional material, on the other hand, poses yet another threat to equip-ment in the containment whose performance is degraded by the presence of aerosols. VANESA is a mechanistic model of fission p'roduct release and aerosol generation during core-concrete interactions. This model was based on observa-tion from experiments involving high-temperature melts on concrete and informa-tion from analogous industrial processes. Two broad mechanisms of aerosol formation are considered in the model: vaporization of melt species accentuated by gas sparging, and mechanical formation of aerosols by violent agitation of the molten debris sparged with decomposition gases. Vaporization processes produce the most intense aerosol generation during ex-vessel core debris inter-actions, while mechanical proces as provide a mechanism for aerosol formation that persists even when debris temperatures are so low that little vaporization of species in the debris can occur. Input to this model includes melt temperature, concrete erosion rate, and gas generation rate predicted by the CORCON model of melt-concrete inter-actions, it computes the thermochemical limits of vaporization from the melt, and then compares the extent of vaporization recognizing kinetic barriers, such as mass transport, to the approach to the thermochemical limits for vapori-zation. Mechanical aerosol generation is estimated by analogy to ex'perimental data with simulant systems. More complete descriptions of the model are provided in the users' manual (5.19) and its uncertainties are discussed further in Appendix C of Volume 1.(5.1) a 5-11 1
4 5.3 Radionuclide Transport and Deposition
\
5.3.1 Transport in Reactor Coolant System: TRAP-MELT The TRAP-MELT code that was used for the primary system radionuclide-transport analyses of this study was developed from the published TRAP-MELT code (5.20) used for the " Technical Bases" report (5.8) . Major changes were made in the treatment of aerosol particle transport and behavior and in radio-nuclide condensation on and evaporation from particles. In addition, the internal data base of the code was increased to include physical property data for tellurium and cesium hydroxide. An outline of the code, highlighting these changes, is given below. A more detailed description is given in the TRAP-MELT Users' Manual (5.20) , The TRAP-MELT model is designed to treat radionuclide transport in an arbitrary flow system whose thermal-hydraulic conditions are given as func-tions of time. For this study, the data needed by TRAP-MELT to define the thermal-hydraulic conditions of the primary system were generated by MERGE. In addition, TRAP-MELT requires the definition of source terms for each radio-nuclide; these terms were developed by C0RSOR. Once the flow system is defined, it is subrtivided into a series of control volumes that can, in principle, be arbitrary in number and flow connec-tions and that are chosen on the basis of characteristic geometry, thermal-hydraulic conditions, and suspected significant radionuclide behavior such as change of phase, agglomeration, or deposition. Radionuclides in each control volume are assigned, with uniform distribution, to one of two carriers: the wall surfaces and the gas phase. Each radionuclide is allowed to reside on these carriers in either particulate (liquid or solid) or vapor form so that by combining carrier with fvrm in the concept of " state", the condition of a radionuclide in a given control volume is completely determined by its state. TRAP-MELT thus considers five states: e Radionuclide vapor carried by gas e Radionuclide particle carried by gas e Radionuclide vapor carried on wall surface 5-12 i
e Radionuclide particle carried on wall surface e Radionuclide vapor chemisorbed on wall surface. This list of states is not exhaustive (for instance, in two-phase flow, the carrier water must be considered) and the logic of the code has been chosen to accept an arbitrary number of states readily . Radionuclide transport can occur among the five states of an indi-vidual control volume or between certain states of different control volumes are connected by fluid flow. The former types of transport are modeled or correlated in the code itself. The latter are assumed to occur in phase with the fluid flow (as developed by codes such as MERGE) and are imposed on the system. Sources of radionuclides to the system may occur in any volume and any state, and they must be input to the code as mass rate functions of time. At present, the intravolume transport mechanisms contained in TRAP-MELT are: o Competitive condensation on, or evaporation from, wall surfaces and particles of cesium iodide, cesium hydroxide, and tellurium o Irreversible sorption of molecular iodine, cesium hydroxide, and tellurium on stainless steel surfaces e Particle deposition on surfaces due to
- Settling
- Diffusion from laminar and turbulent flow
- Inertial impaction from turbulent flow
- Thermophoresis.
Particle transport (and evaporation or condensation from or on particles) depends on particle size. TRAP-MELT takes this into account by considering a discretized particle size distribution that is subject to change, in each volume, by the deposition processes themselves, by possible particle sources, by flow of particles from other volumes, by flow of particles out of the volume in question, and by agglomeration. The last can be due to many mechanisms. TRAP-MELT considers the following agglomeration mechanisms: e Brownian e Gravitational e Turbulent (shear and inertial). 5-13
~
Considerations of stiffness and linearity split the system of first-order differential equations resulting from the above-listed transport mechanisms into three classes. Most of the deposition mechanisn.s (transfer from gas to wall surface) are taken as first order in the concentration of ; radionuclide species on the carrier (gas, particle, o'r wall) from which the transfer occurs. They constitute the first class, whose transport scheme can be written in the form: 5 = S + MC, (5.1) dt where C is the concentration vector of the species in question for each state and volume, ! is the source rate vector for each state and volume, and M is the transport matrix between all states and volumes. Because the deposition terms are taken as first order, M is independent of C and depends, with S, on time only. It is thus possible to solve Equation (5.1) as a set of first-order differential equations with constant coefficients by standard techniques. This is done in TRAP-MELT for the class of linear mechanisms. Condensation and evaporation, which have a much shorter time constant than the linear pro-cesses, constitute the second class and are treated outside this framework but parallel to it, as is particle agglomeration, which constitutes the third class of mechanisms in the TRAP-MELT code. The approach to this parallel treatment is as follows: Equation (5.1) is taken as the master time-translation operation of the radionuclide system. Time steps are adjusted so that S and M change little over a time step and so that the time step does not exceed one-third of the smallest flow residence time for any control volume. The latter assures that the system does not translate excessively between couplings to the other two classes of l mechanisms. In addition, the characteristic coagulation time for the aerosol in each volume is evaluated and compared to the master time step. If the former is short compared to the latter, the master time step is appropriately I reduced. At the beginning of each time step, phase transitions of radio- ! nuclides are modeled by examining each control volume in turn and solving the molecular mass transport equations for' vapor transport among the gas phase, particles, and wall surfaces. Because of the low heats of vaporization of the 5-14
~ ^
radionuclides in question,.this transport is assumed to be isothermal. Transfer to the walls assumes the Dittus-Boelter correlation (5.5) for pipe
-flow and transfer to the particles occurs by diffusion based on the size distribution at the beginning of the time step. Redistribution of the vapor phase occurs in a time that is small compared to the master time step; there-fore, this redistribution is essentially decoupled from the other processes considered which justifies the use of a time parallel solution treatment.
Once redistribution of the vapor phase has been effected, its effect on the existing particle size distribution (in the volume in question) is cal-culated by assuming that each size class gains (or loses) mass in proportion to the rate of vapor transfer to (o'r from) that size class. Conservation of number for each size class then dictates redistribution between, in general, two new contiguous size classes, the number in each size class being determined by mass conservation. At the end of a time step, the particle size distribution in each volume is reevaluated over that time step to account for possible particle agglomeration, sources, and flow terms. The a glomeration algorithm has been excerpted from the QUICK aerosol behavior code 5.22) , which is based on a size discretization scheme. The approximations inherent in this parallel treatment are minimized by relegating mass redistribution and conservation to the master Equation (5.1), except for redistribution due to radionuclide phase change. Agglomeration and particle evaporation / condensation serve only to modify the particle size distri-bution and therefore affect particle deposition indirectly through mass-distri-bution-averaged deposition velocities. Thus the aerosol aspect is solved (cver a master time step) completely in parallel to Equation (5.1), using all sources, flow terms, and particle removal terms evaluated for each size class considered. The resultant distribution is used to evaluate average particle deposition terms for use in the master equation only. Similarly, reevaluation of the particle size distribution due to radionuclide phase change affects these average deposition terms only. In addition to the time-dependent thermal-hydraulic conditions and ! mass input rates by species, the TRAP-MELT code requires input information on the initial particle size distribution of the source, the control volume i geometry, and the physical properties of species (including deposition veloci-5-15
~ .
ties on surface materials). The code provides output in terms of time- and location-dependent mass by species and state, as well as size distribution of suspended particulate material. There are a number of uncertainties which affect the TRAP-MELT code predictions of primary system retention of materials. Any errors or impreci-sions in the input to the code will clearly affect the quality of the results, both for the primary system thermal-hydraulics provided by MERGE and for the core release rates determined by CORSOR. The extent of interaction among the materials released from the melting core is determined largely by the timing of their releases, and this represents a less straightforward, but no less important, potential effect on the code's results due to input inaccuracies. The experimentally determined vapor deposition velocities for Te, Cs0H and 12 on hot surfaces may not represent an accurate description of the process as it occurs in the reactor coolant system (RCS) because of the impre-cision in the available data and because the experimental systems may differ from the actual RCS conditions. Nevertheless, what data are available have been incorporated, since these analyses are intended to reflect the state of the art. Additional uncertainties affecting vapor and aerosol deposition arise from possibly inadequate specifications of primary system geometry and flow patterns. The disposition of materials suspended in the coolant system at the time of core slumping or at depressurization of the pressure vessel can have significant impact on retention calculated for some of the sequences analyzed. This is because some fission products and aerosols emitted from the core have not escaped the RCS at the time of core slumping and are still available for injection into the containment. The large burst of steam which accompanies core slumping or depressurization when the pressure vessel fails will rapidly sweep out the coolant system, and the very short transit time to the contain-ment is expected to lead to minimal retention of these materials. Thus, in the analyses in this document, the material suspended in the RCS at the time of core slump or pressure vessel failure is assumed to be injected into the containment as a " puff" release, with no further retention in the primary system. The analyses in the main body of this document are subject to some uncertainties which may overpredict retention in the primary system. One 5-16 I
mechanism not included in the current analysis is the structure heatup due to decay heat from the deposited fission products. Heatup of surfaces where species of intermediate volatility (e.g., Csl and Cs0H) are deposited would lead to reevolution and transport of the previously deposited materials through the reactor coolant system to regions of lower surface temperature or to the containment. Thus, the deposition of these species may be self-limiting to some extent. This effect is analyzed in the Appendix to this volume. 5.3.2 Transport in Containment: SPARC Mar.y BWR accident sequences involve a fission-product flow path which passes through a suppression pool. Although the importance of the suppression pool in removing fission products has long been recognized, comprehensive analytical models that consider all pertinent parameters (such as particle size, bubble size, pool dimensions, and pressure and temperature conditions) have not been available. The SPARC code was written specifically for the analy-sis reported here and was used to calculate removal of particulate matter by a suppression pool. The SPARC code was developed by Owczarski, et al( 23) . The model includes particulate removal due to steam condensation, gravitational settling, inertial impaction both inside the gas bubbles and in the gas injection regime, diffusion deposition, and mechanical entrainment of pool liquid at the pool surface. In addition, a mechanism which retards particle deposition due to evaporation of steam is considered. Details of the model are provided in the SPARC Users' Manual (5.23) , Recognizing that the role of the suppression pool in removing parti-culates was neglected in the past, or a fixed removal efficiency or decontami-nation factor was arbitrarily assumed, use of the SPARC code in this program represents an improvement over previous source term analyses by accounting mechanistically for the effects of the suppression pool. 5.3.3 Transport in Containment: NAUA 4 The NAUA code was developed at the Kernforschungszentrum Karlsruhe, West Germany, for calculating aerosol behavior in LWR core melt accidents.(5.24) It is based on mcchanistic modeling of aerosol agglomeration and deposition 5-17
~
)
within a containment vessel where a condensing steam atmosphere may exist. The model for steam condensation on particles was validated by small-scale experimental measurements (5.25), and larger-scale validation is being planned. The NAVA code calculates physical processes, excluding chemical changes and radioactive decay. The removal processes considered include gravi-tational settling and diffusional plateout. Interactive processes include Brownian and gravitational agglomeration and steam condensation. Aerosol sources and leakage are also included. Compositional changes resulting from time-dependent compositions for the input aerosol are tracked by the code. The particle size distribution is defined by a number of monodisperse fractions. In this approach, the governing integro-differential equation is transformed into a system of coupled first-order differential equations. In effect, the particle size fractions interact and deposit according-to the included mechanisms, generating a time-dependent distribution of mass among the various size fractions. Steam condensation is handled in a separate inte-gration. Output from the code includes mass concentrations of condensed water and dry aerosol materials (airborne and on surfaces), as well as particle size distributions at various times throughout the calculation. Since the original version of the NAUA code has no provision for engineered safeguards, calculations were made to account for removal of aerosol particles by sprays, as follows: 2
= -cwR N (Vg
- vg )n, (5.2) where n is the aerosol particle concentration, e is the collision efficiency, Vo and vq are settling velocities of the spray drops and aerosol p5rticlet, respectively, R is the radius of the spray drop, and N is the water drop concentration.
Due to hydrodynamic interaction between a falling water drop and airborne particles, only a small fraction of the particles within the cross-sectional , area of the water drop is removed by spraying. To account for this hydrodynamic l effect, the collision mechanisms due to inertial impaction, interception, and ! 5-18 l l t n.
Brownian diffusion of aerosol particles were used by defining c in Equation (5.2) as: c=c g+cR + *D'
- where cI, cR and CD are the collision efficiencies due to intertial impaction, interception, and Brownian diffusion, respectively. The following collision ,
efficiency models were utilized for-the three mechanisms: cy = Stk (Stk + 0.35)Z (5.4) c
,1.5(r/R)2 -
R (5.5) (1+r/R) c D
= 3.5 Pe-2/3 ,
(5.6) where Stk is the Stokes number for aerosol particles based on a characteristic length of water drop with radius R; r is the particle radius; and Pe is the Peclet number. The Stokes number and the Peclet number.are defined as 2 2rpVC -' Stk = gj and (5.7) Pe = (5.8) D where D is the diffusion coefficient of aerosol particle vg is the settling velocity of water drop C is the Cunningham slip correction far. tor op is the particle density y is the gas viscosity. , In general, for relatively large particles, the inertfal effects on the overall collision efficiency are larger than the interception term because the water drops are much larger than the aerosol particles. As particle size becomes smaller, the Brownian diffusion term will become increasingly important. It should also be mentioned that Equation (5.4) is given by Hetsroni(5.26) and Equations (5.5) and (5.6) are based on the work of Lee and Gieseke(5.27, 5.28), 5-19 ; t r
Another particle deposition mechanism, diffusiophoresis, was added to the NAVA code. Diffusiophoresis results from steam condensation onto containment walls and involves two mechanisms: a net flow of gas toward the wall surface (known as Stephan flow), and a molecular weight gradient caused by the steam concentration gradient. In general,.the effects of Stephan flow are much larger than those of the molecular weight gradient and result in deposition of particles on the wall surface. The condensation rate toward wall surfaces calculated by the MARCH code has been used to calculate deposi-tion due to diffusiophoresis. In utilizing the NAUA computer code for calculating aerosol behavior during various accident sequences, it was noted that in certain cases the code requires a long computing time to calculate the rate of condensation of water vapor onto particles. This type of problem takes place when a large amount of condensible water vapor was used as an input. It was noted that a supersatura-tion ratio of much greater than 1.0 was frequently encountered even after the condensation calculation was completed. Some literature suggests that pure water vapor at 20 C will spontane-ously form water droplets in the absence of condensation nuclei when the super-saturation ratio exceeds 3.5, and at 0 C a saturation ratio of 4.3 is required for homogeneous nucleation. This mechanism has been implemented in NAVA in addition to the existing condensation calculation. As the critical supersatura-tion for the homo eneous nucleation, the following correlation equation given by Green and Lane 5.29) was used: S = exp(0.557-(o/T)3/2,g), where S is the critical supersaturation o is surface tension T is the temperature in *C M is the molecular weight of water. No nucleation or self-condensation rates are calculated in the code. Rather, if critical supersaturation is realized at a given time, the excess water vcgor is assumed to form water particles of a uniform size spontaneously. Of co6rse, these small pure water droplets are subsequently subject to NAVA's ' usual condensation and coagulation processes both among themselves and with 5-20 o s L -
.; +; ;; . ;
1 ik l other particles containing solids. Although the effects of this mechanism on ; the overall aerosol. concentration change are insignificant, the computational l time it reduced considerabiy by this implementation. N References (5.1) Gieseke, J. A., et al, "Radionuclide Release Under Specific LWR Acci-dent Conditions,< Volume 1", BMI-2104 (July,1983).
' Wooton, R. O., et al, " MARCH 2 Code Description and' Users' Manual",
(5.2) DraftIDecember,1982). (5.3) Wooton', f. O. and Avci, H. I., " MARCH (Meltdown Accident Response Characteristics) Code Description and Users' Manu'al", NUREG/CR-1711, BMI-2064 (October, 1980). (5.4) "American National Standard for Decay Heat Power in Light Water Reactors", ANSI /ANS-5.1-1979. (5.5) Geankoplis, C. J., " Mass Transport Phenomena", Holt, Rinehart, and Winston (1972). (5.6) Freeman-Kelly, R., "A Users' Guide for MERGE", Batte11e's Columbus Laboratories, October, 1982. (5.7) " Technical Basis for Estimating Fission Product Behavior During LWR Accidents", NUREG-0772 (June, 1981). (5.8) CORS0R Manual. (5.9) Bell, M. J., "0RIGEN, The ORNL Isotope Generation and Depletion Code", ORNL-4628 (1973). (5.10) Lorenz, R. A., et al, "l'ission Product Release from Highly Irradiated LWR Fuel", NUREG/CR-0722 (1980). (5.11) Lorenz, R. A.. Collins, J. L., and Malinauskas, A. P., " Fission l Product Sou - Terms for the LWR Loss-of-Coolant Accident", l NUREG/CR-1288(1980). l (5.12) Lorenz, R. A., et 31, " Fission Product Release from Highly Irradi-ated LWR Fuel Heated to 1300-1600 C in Steam", NUREG/CR-1386 (1980). l l (5.13) Lorenz, R. A., " Fission Product Release from BWR Fuel Under LOCA Conditions", NUREG/CR-1773 (1981). (5.14) . Parker, G. W., Martin, W. J., and Creek, G. E., "Effect of Time and Gas Velocity of Distribution of Fission Products from U02 Melted in 5-21
a Tungsten Crucible in Helium", Nuclear Safety Program Semi-Annual Report for period ending June 30, 1963, ORNL-3483, 19-20 (1967). (5.15) Albrecht, H., Matschoss, V., and Wild, H., " Experimental Investiga-tion of Fission and Activation Product Release from LWR Fuel Rods at Temperatures Ranging from 1500-2800 C", proceedings of the Special-ists' Meeting on the Behavior of Defected Zirconium Alloy Clad Ceramic Fuel in Water Cooled Reactors, 141-146 (September, 1979). (5.16) Albrecht, H., Matschoss, V., and Wild, H., " Release of Fission and Activation Products During Light Water Reactor Core Meltdown", Nuclear Technology, 46, 559-565 (1979). (5.17) Niemczyk, S. J. and McDowell-Boyer, L. M., " Technical Considerations Related to Interim Source Term Assumptions for Emergency Planning and Equipment Qualification", ORNL/TM-8275 (1982). (5.18) Weast, Robert C., Ed., CRC Handbook of Chemistry and Physics, 59th Edition,(1978). (5.19) VANESA Manual. (5.20) Jordan, H., Gieseke, J. A., and Baybutt, P., " TRAP-MELT Users' Manual", NUREG/CR-0632, BMI-2017 (February, 1979). (5.21) TRAP-MELT 2.1 Users' Manual. (5.22) Jordan, H., Schumacher, P. M., and Gieseke, J. A., " QUICK Users' Manual", NUREG/CR-2105, BMI-2082 (April, 1981). (5.23) Owczarski, P. C., Postma, A. K., and Schreck, R. I., " Technical Bases and Users' Manual for SPARC -- Suppression Pool Aerosol Removal Code", report to the U.S. NRC, NUREG/CR-3317 (May, 1983). (5.24) Bunz, H., Koyro, M., and Schock, W., "A Code for Calculating Aerosol Behavior in LWR Core Melt Accidents Code Description and Users' Manual". (5.25) Schock, W., Bunz, H., and Koyro, M., "Messungen der Wasserdampfkonden-sation an Aerosolen unter LWR-un-falitypischen Bedingungen", KfK 3153 (August, 1981). (5.26) Hetsroni, G., " Handbook of Multiphase Systems", McGraw Hill Book Company and Hemisphere Pub. Co. (1982). (5.27) Lee, K. W. and Gieseke, J. A., J. Aerosol Science, 1, 335 (1980). (5.28) Lee, K. W. and Gieseke, J. A., Environ. Sci. & Technol., 13, 446 -- (1979). (5.29) Green, H. L. and Lane, W. R., Particulate Clouds: Dust, Smokes and Hists, D. Van Nostrand Co., Princeton, New Jersey (1957). 5-22
- 6. BASES FOR TRANSPORT CALCULATIONS 6.1 Plant Geometry and Thermal Hydraulic Conditions MARCH 2 code calculations were performed for several variations of each of the four accident sequences considered. The results of the MARCH analyses are used as input for three aspects of the fission product release and transport calculations:
(1) The predicted time-dependent temperatures of the fuel are used by CORSOR to calculate fission product release. (2) The primary system pressure and flow of steam and hydrogen from the core are input to MERGE to calculate primary system thermal-hydraulic input to TRAP-MELT. (3) The thermal-hydraulic conditions in the contain-ment building as well as leak rates out are input to the containment transport codes. A sumary of MARCH modeling options utilized in the analyses is presented in Table 6.1*. The design parameters for the Surry plant were summarized in Table 4.1. One of the many areas of modeling uncertainty for thermal-hydraulic analysis for this study has been the behavior of the flow in the reactor cool-ant system in the pathway of release to the containment. In particular the conditions in the upper plenum and upper dome region are quite uncertain and could have a significant effect on the transport of radionuclides. The first problem in describing flow behavior in this region is in obtaining an adequate characterization of the structures. These structures are not described in detail in publicly available reports because of proprietary design features. This problem was alleviated by input from Westinghouse Electric Corporation, the reactor manufacturer for the Surry plant. In addition, even if the geometries of these structures had been well known, there is significant uncertainty as to the nature of the flow patterns in the upper plenum and dome during accident conditions, and as a result how much of the flow reaches the
*All tables in this section of the report have been placed at the end of the section.
6-1 6 1
available surfaces. The current version of TRAP only considers one structure with a given surface to mass ratio within a volume. While this limitation can be circumvented by subdividing the volume of interest into a number of smaller subvolumes, each with a particular structure, such an approach would require information on the possible series / parallel flow splits among the subvolumes, i.e., how much of the flow is seen by each of the structures. Such fine detail was not warranted by the existing level of understanding of flow patterns in the upper plenum. As a practical matter the gases leaving the core were assumed to flow in series through the upper core support structures, past the control rod guide tubes and support columns, past the top support structure, along the upper core barrel, and out the hot legs. Clearly alternate flow paths are possible, e.g., after passing through the upper core plate some of the flow could go directly to the hot legs without passing past the other structures. The sensitivity of results to this approximation will be investigated later in c the program. In the following sections of the report, the results obtained with the MARCH and MERGE codes are described for each of the accident sequences. In Section 6.1.6, some of the uncertainties in the analyses and sensitivities to assumptions are discussed. h 6.1.1 Sequence AB (Hot Leg) A large pipe-break accident with failure of the ective emergency core coolant injection system, as would result from total loss of AC power, would be expected to result in comparatively rapid core meltdown. This is because core uncovery would occur essentially at the start of the accident with the decay heat level relatively high. The luss of electric power would also preclude the operation of containment safety features. Table 6.2 indicates the times of key events as predicted by the MARCH code for the input and model-ing assumptions utilized. Table 6.3 provides details of the core and primary system conditions for this sequence. Core uncovery, heatup, and melting would occur at low primary system pressure, corresponding to the pressure of the containment. The temperature of selected fuel regions is illustrated as a function of time in Figures 6.la and 6.lb. In this, as well as in subsequent plots of core no4 temperatures, the designation R0D (X,Y) denotes the core 6-2 l n - - - - - - - -
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node in axial position X and radial region Y. For these analyses the core was divided into 24 axial nodes and 10 radial regions. Prior to the accident, the piping and structures in the reactor cool-ant system would be in the temperature range of 290-315 C. Because heatup of the fuel and the release of fissicn products would occur at about 172 kPa (25 psia), these surfaces would be expected to be considerably superheated. In addition, because of the high boiloff rate (high decay heat level) and low density of gases in the primary system (low primary system pressure), the velocity of gases passing through the reactor coolant system would be high in comparison to other accident sequences. At the time of core uncovery, the velocity of steam in the upper plenum is estimated to be approximately 1/2 meter /sec (2 feet /sec). The total residence time in the system, from leaving the core to exiting the break in the hot leg, would be less than 1 minute. As the water level in the core drops, the production of steam decreases accordingly. Just prior to slumping into the lower plenum, most of the steam that is being generated is predicted to be reduced to hydrogen. Temperatures of the structure in the reactor coolant system are illustrated in Figure 6.2. Structure 1 represents the upper core plate, Structure 2 represents the control rod guide tubes and support columns, and Structure 3 represents the top support structures and core barrel. (In this figure, time is measured from the start of core uncovery.) The gas temperatures leaving the core rise rapidly as the core melts and begins to slump. When the i core collapses into the bottom head of the reactor vessel, large quantities of 1 l steam are generated, resulting in a sharp decrease in gas temperatures as well l as cooldown of the structures in the upper plenum. The maximum temperature of ! gases leaving the hot leg is estimated to be in the range of 650 C (1200 F), but these persist only a short time. A schematic of the gas flow path for the AB sequence is illustrated in Figure 6.3. Several possible times and modes of containment failure have been investigated for this sequence: failure to isolate (s), early failure due to hydrogen burning (y), and basemat melt-through-(c). Table 6.4 presents the details of the containment res.ponse for the various cases considered. For this analysis, the containment isolation failure is assumed to exist at the start of the accident and to be in one of the penetrations from the containment ; 6-5 1
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.to the safeguards or auxiliary buildings. . Thus, both these volumes were consi-
'dered:in'the analyses for the AB-8 sequence'.
~The containment pressure and temperature histories for this sequence are illustrated in Figures 6.4_and 6.5. In these figures, Compartments 1 and 2 represent.the containment and the safeguards building, respectively. As a result of leakage through the isolation failure, the containment pressure decreases-fairly rapidly after primary system blow <iown. The large steam genera-tion associated with the collapse of the core into the bottom head leads.to a temporary increase in the pressure. The pressure in the safeguards building is seen to stay near atmospheric except for a' sharp increase at about 200 minutes, when a hydrogen burn is predicted to take place. At other times,.the leakage from the safeguards building is able to keep the pressure low. It should be noted that a pressure rise of the type calculated could lea'd to the failure of the safeguards building.
The timing of the burn is controlled by the combination of increasing hydrogen concentration due to leakage from the containment and decreasing steam concentration due to condensation. (High steam concentration tends to inert the containment atmosphere early in the sequence.) The early overpressure containment failure (Y) considered was associated with a hydrogen been taking place during the concrete-attack phase of the accident. ,The hydrogen was allowed to burn when flammable conditions were reached. The timing of.the burn was determined by the increasing hydrogen concentration from concrete attack and decreasing steam concentration due to condensation. The containment pressures and temperatures for the AB-Y sequence are illustrated in Figures 6.6 and 6.7. Different containment pressure responses would be predicted if the assumptions regarding hydrogen. generation and burning were varied. The likelihood of containment failure due to a hydrogen burn or other event would, of course, depend on the failure pressure as well as the magnitude of the pressure; however, quantifying the probability of containment failure.is not a part of this' effort. If the containment is able to withstand earlier challenges, long-term overpressure failure or basemat melt-through may be the eventual. outcome for this sequence. The current MARCH calculations, using as input a concrete composition
-representative of that actually used at Surry, indicate that melt-through of the basemat would take place before long-term overpressurization. This is the 6-8
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case that was analyzed as representative of delayed failure modes. The contain-
. ment pressure and temperature histories for this sequence are shown in Figures 6.8 and 6.9. Containment melt-through is predicted to take place at about 1450 minutes into the accident. Since upon melt-through the containment depressurizes through the ground, only a small change in the containment pres-sure is seen for the conditions encountered here.
It should be noted that there is substantial uncertainty regarding the progression of concrete attack, and the timing of the occurrence of melt-through could vary considerably. It is also possible that overpressure failure could precede melt-through. In the Surry design, the reactor cavity and the containment sump are not connected; thus sump water evaporation is not a source of containment pressurization. After reactor vessel penetration, the principal driving force for pressurization would be the' release of gases from the decom-position of the concrete. The time required to reach the assumed failure level of 0.69 MPa (100 psia) would be longer than the melt-through time predicted for the particular case considered here. If, however, lateral attack of the concrete by the core debris leads to the ingress of sump water into the reactor cavity, the rate of containment pressurization could be increased. Table 6.5 summarizes the containment leakages for the various cases considered that were derived from the MARCH results and used in the evaluation of the fission product release from containment. 6.1.2 Sequence TMLB' In the transient sequence TMLB', the ability to remove heat from the reactor coolant system is lost, and containment safety features are not avail-able due to loss of all electric power. Decay heating following reactor shut-down boils off the water in the secondary side of the steam generators. After steam generator dryout, the primary system pressure rises to the relief valve setpoint, and reactor coolant is discharged through the relief line to the discharge tank and ultimately to the containment building. Table 6.2 indicates the times of key events as predicted by the MARCH code. Core and primary system conditions are given in Table 6.3. The temperature transient of selected fuel regions is illustrated in Figures 6.10a and 6.10b. Core uncovery, heatup, and melting occur with the primary system pressure at approximately 17.24 MPa (2500 6-13
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i psia). Because'of the high density of steam at this pressure, the flow velo-city .in the primary: system would be quite small -- generally less than 1/2' cm/sec (1 foot / min) -- until the start of core slumping. The Reynolds Number in the upper plenum is predicted to be in the laminar regime until core slumping begins, as for the AB case. The Rayleigh Number, however,'is substantially~ larger in the range of 1011,1014 Thus, significant mixing could occur in the upper plenum could be driven by tempera-ture gradients and the buoyancy of hydrogen. The temperatures of the gas and structures in the volumes of the primary system are !!1ustrated in Figures 6.11a and 6.11b. Volume 2 represents the reactor vessel upper plenum with its struc-tures and Volume 3 represents the piping and the pressurizer. Within Volume 2 (the upper plenum), Structure 1 is the top core plate, Structure 2 represents , the control rod guide tubes and support columns, and Structure 3 represents the top support structure and the core barrel. In Volume 3, Structure 1 is the pressurizer and Structure 2 represents the piping. In these figures, the time is measured from the start of core uncovery. A schematic of the gas flow path for TMLB' is illustrated in Figure 6.12. Two specific containment failure mode possibilities were considered for the TMLB' sequence, an early and a late failure. 6.1.2.1 Early Failure: TMLB' (6 ) i The early failure was assumed to be the result of the rapid steam generation from the interaction of the core debris with accumulator water in the reactor cavity. The failure of the vessel bottom head releases the high pressure steam from the primary system to the containment and discharges the core debris into the reactor cavity; the resulting drop in the primary system , pressure allows the accumulators to discharge onto the top of the core debris. Figures 6.13 and 6.14 give the containment pressure and temperature histories for TMLB-6, in which the containment is assumed to fail early due to the pres-sure rise from the debris water interacting in the reactor cavity. MARCH calculates the rate of steam and hydrogen production resulting ; j from the debris-water interaction in the reactor cavity. Several user-selected ' options in MARCH can be used to describe these interactions, ranging from a l simple quenching model to several debris-bed models; combinations of models l l 6-18
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are also possible. All the models require user inputs for which there is little basis. Perhaps the " lost important input parameter is the assumed debris-particle size; additional inputs that may be required relate to such items as debris-bed porosity, bed area, and criteria for transition from particulate to debris bed heat transfer. The rate and magnitude of the calculated contain-ment pressure rise can obviously be sensitive to the input and modeling assump-tions utilized. The use of the simple quench models in conjunction with small particle sizes tends to maximize the amount as well as rate of steam production from debris-water interaction; since in such cases the debris are quenched rapidly, only small amounts of hydrogen are predicted to be generated under these assumptions. The debris bed models can also predict large amounts and l rates of steam generation, depending on the input parameters, but generally lead to slower rates of containment loading. In some cases, associated with marginally coolable debris beds, large amounts of hydrogen generation can be l predicted. 6.1.2.2 Late Failure: TMLB' (c) The containment pressures and temperatures for a representative TMLB-c sequence are given in Figures 6.15 and 6.15. The effect of the debris-water interaction is seen in the sharp pressure increase at about 160 minutes in Figure 6.15. In the particular case illustrated, the pressure rise would probably not challenge containment integrity; higher pressures could be predicted if alternate assumptions were used. If the containment maintains its integrity through the above early pressure transient, the containment pressure will decline somewhat due to 1 condensation of steam on internal structures, but will later increase again due to the attack of the concrete basemat by the hot core and structural debris. Since the gas and vapor input rates from concrete decomposition are low, except when the debris is very hot, a long time would be required for the pressure to build up to levels at which the likelihood of failure is significant. While the pressure is slowly building up, basemat melt-through could precede and preclude a long-term overpressure failure. Melt-through of the concrete basemat, with venting of the gases through the ground, is reflected in the sharp pressure decrease at about 740 6-24 o - -
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minutes in Figure 6.15. It should be noted that the progression of concrete attack by the core debris, and hence the possible timing of melt-through, are highly uncertain, and that the high partial pressures of steam in the con-tainment atmosphere throughout most of this sequence are predicted to preclude hydrogen burning. It may be noted that some of the input parameters for the TMLB 6 sequence were changed from those in TMLB-c, leading to a higher containment pressure peak. In considering the occurrence of an early containment failure, no representation is made as to the likelihood of such failure; the quantifi-cation of the probability of containment failure due to such interactions is beyond the scope of this study. It is suggested, however, that the magnitude of the pr_ssures resulting from debris-water interactions may be sufficiently high that the possibility of failure should be considered. Table 6.5 summarizes the containment leak rate information derived from the MARCH results and used in the evaluation of the fission product release from the containment. 6.1.3 Sequence V The V sequence, or interfacing systems LOCA, is initiated by the failure of the check valves separating the low pressure emergency core cooling system from the primary coolant system. The release of the high pressure primary coolant inventory to the low pressure piping would not only lead to the failure of the emergency core cooling system but also provide a path for I the release of radioactivity that bypasses the primary containment. It is also possible that the primary system blowdown would result in the failure of the safeguards or the auxiliary building. The interfacing systems LOCA sequence is of particular interest because the containment building is bypassed for much of the sequence, and the primary system could represent the principal location for the retention of fission products released from the core. The possibility of retention in the safeguards (or auxiliary) building also exists if it does not also fail; this possibility would be quite design dependent. The thermal-hydraulic behavior of the system during the period of fuel melting is similar to that of the Sequence AB. Following a period of 1/2 6-27 1
- - - . _ - _ - - _ __o_________ _ _ - _ _ _ _ _____ _ _
{ hour, during which blowdown and loss of reactor coolant invenwcy lead to the l l point of core uncovery, melting of fuel would occur over an interval of another 1/2 hour. The temperature histories of selected core nodes are given in Figures 6.17a and 6.17b. The timing of key events is presented in Table 6.2. The core and primary system characteristics at key times during the sequence are given in Table 6.3. In this case, the primary system pressure would be slightly more-elevated (0.68 MPa), and the velocity in the upper plenum would be reduced to approximately 10 cm/s. The velocity of the steam and $ hydrogen flowing back through the ECC injection line to the auxiliary building wrald be greater than 1 m/s. The residence time in the reactor coolant system , from core exit to the atmosphere of the auxiliary building would be on the order of 1 minute, with the majority of the time spent in the upper plenum. The predicted temperatures of gases and structures in the flow path to the auxiliary building are illustrated in Figures 6.18a-d. In these figures, , Volume 2 represents the reactor vessel upper plenum, Volume 3 the hot leg piping, Volume 4 the steam generator, and Volume 5 the emergency core cooling system piping. The thret structures in the upper plenum (Volume 2) are as previously described. Figures 6.19a-d show the results for an alternate breakdown in which the emergency core cooling system piping was represented as four connected volumes rather than a single volume. In both sets of figures, the time is measured from the start of core uncovery. The results are similar to those obtained for the AB sequence. A schematic of the gas flow path for V is illustrated in Figure 6.20. The auxiliary building and containment characteristics at various times during the sequence are given in Table 6.4. Table 6.5 summarizes the containment leakage flows derived from the MARCH analyses and used to evaluate fission product release to the environment. 6.1.4 Sequence S D_ 2 This small-pipe-break accident sequence with failure of ECC injec-tion involves conditions intermediate to the high pressure meltdown sequence TMLB' and the low pressure sequences AB and V. In the S 2D sequence, the containment sprays and heat removal systems are operating; thus the contain-ment' pressure would be maintained at low levels, except perhaps for brief 6-28
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i l I periods when rapid transient loadings may be encountered. Two containment i failure modes were considered for this sequence: an early overpressure failure resulting from hydrogen combustion (S2 D-Y), and basemat melt-through failure with no direct atmospheric failure of the containment (S2 0-c). The timing of significant events is given in Table 6.2. Core and primary system parameters are summarized in Table 6.3. Selected core node temperatures are illustrated in Figures 6.21a and 6.21b. The predicted temperatures of gases and struc-
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tures in the primary system are illustrated in Figures 6.22a, 6.22b, and 6.22c. Volume 2 again represents the upper plenum with its associated structures, Volume 3 represents the hot leg piping, and Volume 4 represents the steam generator. In these figures, the time is measured from the start of core uncovery. A schematic of the gas flow path for 2S D is illustrated in Figure 6.23. Table 6.4 sumarizes the containment response at key times during the accident sequence. Since the rate of primary coolant loss is restricted by the break size and the containment sprays are operational, pressure in the containment would remain relatively low during most of this accident sequence. This is illustrated by the pressure history for the melt-through case in , Figure 6.24; the corresponding temperature history is given in Figure 6.25. Substantial containment pressure increases would require the accumulation and ' subsequent coherent burning of large amounts of hydrogen. For the case of failure due to a hydrogen burn, the latter was assumed to take place following vessel failure when the hot core debris entered the I reactor cavity. The ejection of the hot core debris into the containment would provide the ignition source for the burn. The containment pressure and tempera-ture histories for such a hydrogen burn are given in Figures 6.26 and 6.27. This burn produced a peak pressure of about 0.62 MPa (90 psia); the calculated peak pressure can be sensitive to the timing of the assumed burn. In assuming containment failure due to such a burn, no representa-tion is made as to the likelihood of failure. If the containment maintains its integrity through challenges such as hydrogen burning, it is likely that the basemat will eventually be penetrated due to the attack of the concrete by the core and structural debris. The time required for melt-tnrough would provide ample opportunity for fission product removal by the sprays. If suffi-cient spray water can continuously rea:h toe reactor cavity, a ccolable debris 6-40
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bed could form.- In such'a case, the S 2D sequence would be terminated without containment breach. Table 6.5 summarizes the containment leakages derived from the MARCH calculations that were used in the evaluation of fission product releases from the containment. 6.1.5 Other Cases Considered The primary containment in the above cases was modeled as a single compartment with its inherent assumption of being well mixed. For the AB-S and V sequences, the releases from the primary containment were input into the safeguards building before their release to the atmosphere. Thus the potential for fission product deposition in the safeguards building was included in the analyses for these two cases. In the other sequences, the release froin the primary containment was directly to the environment. To investigate the possible effects of containment compartmentaliza-tion on the release of radioactivity to the environment, the AB-8 sequence was reformulated to model the primary containment as a system of four interconnected compartments. While the Surry containment has many internal structures and partitions, the flow paths between adjacent subcompartments are typically very large, and multiple flow paths are possible. Thus there is no obvious unique approach to containment compartmentalization. The containment model in MARCH is restricted to handling only compartments connected in series. The four-volume representation utilized was a practical compromise between the actual containment design and the code modeling capabilities. With the four-compartment containment model, the melt release took place into a different compartment than the vaporization release, i.e., the melt release was assumed to enter one of the steam generator cubicles and the vaporization release was into the reactor cavity. The location of the containment leak was in still another compartment. In the V sequence, the rupture of the coolant system takes place in the low pressure emergency core cooling system piping outside the primary containment. There is a long length of pipe between the core where the fission products leave the fuel and the safeguards building where they leave the piping. In the base-case analysis, this piping was represented as a single volume. To 6-51
~
4 gain insight'on'the possible. sensitivity of the predictions ~this analysis was repeated wit'h the piping being divided into four equal volumes.- The analyses for this sequence as discussed up to this point have included consideration of fission product deposition in the safeguards building bu't have'taken the safeguards building as an open' structure. A number of. variations of the V sequence are possible. The effect-of.the primary system blowdown on the-secondary structures may result'in signi- l
- ficant changes in secondary building integrity. Other. questions relate to the '
operability of various engineered safety feature-systems and/or components:- e.g., does the failure of the low pressure; emergency. core cooling system piping , necessarily imply failure of the emergency core cooling system pumps?. IfLsome of these pumps are operable,.they could serve to delay the time of core uncovery and subsequent overheating, or in the least they could flood the area in which the break in the system is-postulated and thus provide a possiole mechanism ! for fission product removal.
- ~
While many variations of the V sequence can be postulated, it has i been suggested by Stone & Webster, the Architect-Engineer for the Surry plant,- l that a realistic consideration'of the building layout in the area where the emergency core cooling piping is located and where the break in the system is , l postulated, would lead to the conclusion that this area would be. flooded by { the water from the refueling water storage tank. Representatives of Stone & j Webster have indicated that the compartment where-the system break is likely l to occur, given the other assumptions related to the definition of this j sequence, would be flooded to a depth of 5 feet cbove the floor of the compart- ! ment. The water depth would be limited to this height by an opening in the l I wall at this level; additional water would spill through this opening to other i parts of the plant. Since the piping is aproximately 2 feet off the floor, i the effluent from the primary system would have to pass through about 3 feet j of water before being released to the safeguards building atmosphere and.subse-i quently to the environment. The volume of water contained on the floor of l this compartment is understood to be approximately 30,000. gallons. An additional analysis was performed to investigate the possible fission product attenuation by passage through the 3 feet of water if the sequence of events is as described above. The water in the Surry refueling water storage tank is maintained at a temperature of 45 F to improve its 6-52 i i- _ _ . , . , , , , . . , _ , , . _ . . _ . . _ - - . _ _ _ . _ - , _ . . _ - - , - , . . . _ _ _ , . _ _ . . _ , - , , _ . _ . _ _ . _ . - . . , _ - - . ~
4) 4r gh O o 9 IMAGE EVALUATION [M (([/pNNY kj77p \[Nkh/ gf/ TEST TARGET (MT-3) 6 f[4'4e
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effectiveness for emergency core cooling purposes. After spilling to the floor of the compartment, this water would be heated by the passage of steam generated during core uncovery and subsequently by the high temperature steam and hydrogen during core melting. It is possible that some of the water on the floor would be boiled off during these stages of the accident, and this has been calculated by Stone & Webster staff to give a reduction in water depth of a few inches. For purposes of the additional fission product attenuation analyses, it was assumed that the water was heated to saturation at atmospheric pressure and - that the. submergence of 3 feet was maintained throughout the duration of the accident sequence. The releases from the primary system were then input into the SPARC code to determine aerosol attenuation by this pool of water. An additional calculation was performed accounting for water spilled on the floor of the safeguards building for the V sequence. The water was assumed saturated and a submergence of 3 feet was used as discussed in Section 6.1.5. Table 7.18b shows the calculated locational distribution of fission products in various compartments as predicted by the additional calculation. When compared with 7.18a, the additional fission product attenuation due to the water present in the safeguards building is found to be significant. It should be noted that for a rigorous calculation, it is necessary to perform a complete thermal-hydraulic _ analysis for this sequence to obtain the accurate temperature conditions of water and of the safeguard atmosphere. 6.1.6 Gener,1 Discussion The release and transport of fission products are strongly influ-enced by the thermal-hydraulic behavior of the accident. The computer codes MARCH 2 and MERGE that have been used to predict the thermal-hydraulic condi-tions treat various aspects of accident behavior with different degrees of confidence. In the following paragraphs, the principal areas of uncertainty in the analyses, simplifying assumptions and approximations, and the implica-tions for fission product transport will be discussed. In the MARCH 2 analy es of fuel heatup, the reactor core was sub-divided into 24 axial and 10 radial mesh regions. The variation that would occur in the timing of heatup and fission product release across the core is well characterized. Up to the point of cladding melting and fuel / cladding 6-53
liquefaction,-the theoretical treatments of the thermal behavior of the fuel and oxidation of the cladding are supported by experimental data. Reasonable agreement has been obtained in the past between different computer codes in analyzing this behavior. The MARCH code makes the simplifying approximation that the fuel would melt at a single characteristic temperature which is input. The selected input melting temperature of 2550 K (4130 F) has been chosen to be between the temperature at which the fuel would dissolve into molten zirconium and the melting point of uranium dioxide. In the actual system, melting would occur over the above range of temperatures up to the melting i point of uranium dioxide, with the actual melting or liquidus temperature being i a function of local composition and state of oxidation. As a result of the single melting point approximatien, the peak fuel temperatures predicted by l MARCH may be underestimated for some quantity of fuel. The time for which fuel stays at elevated temperature is also very dependent on modeling uncertain-ties. These uncertainties will have little effect on the predicted release of volatile fission products but could affect the vaporization of involatile materials, most likely by underprediction. The MARCH analyses for the present study utilized meltdown model "A" wifh no movement of fuel out of the core until the bottom node in any radial region was molten. At that point, the molten nodes in that region were allowed to slump to the core _ support plate. As the bottom nodes in successive radial regions melted, these regions were allowed to slump. When 75 percent of the l- Ecore was molten, the entire core was assumed to slump into the lower head of the reactor vesscl. The assumptions regarding fuel slumping and ultimate collapse into the vessel bottom head also affect the driving forces for fission product transport, the timing of reactor vessel dryout, and subsequent analyses of head heatup and failure. MARCH does not attempt to describe fuel melting I l and movement mechanistically. There is no actual redefinition of core geometry. The slumping models attempt to treat mathematically what would happen if the fuel were to move in accordance with the selected scenario. These representa-tions are highly intuitive but have some support from core meltdown simulation [ . experiments. The MERGE code was developed specifically for the analysis of reactor coolant system temperatures in this study. There is very little past experience i- in performing this type of analysis. The flow patterns in the system could be 6-54 k ._____ _ _.
quite complex, particularly in the upper plenum region, and are treated approximately. If there is substantial internal recirculation, both the fission product residence time and the heating of structures could be higher than calculated here. The flow within the reactor coolant system is treated as one-dimensional, with well-mixed volumes. Natural convection within the upper plenum is considered in predicting the heat transfer to structures. Although convection patterns are not examined explicitly, the mixing which is expected to result would be consistent with the well-mixed approximation. The one-dimensional treatment of the upper plenum does not take into account the radial temperature profile of gases leaving the core and transporting through the upper plenum. The calculated temperatures are averages across the flow cross section and would be expected to be higher near the center and lower near the periphery. As demonstrated later in this report, the timing of containment fail-ure has a major impact on the predicted release of fission products to the environment. The pressure level at which the containment would be expected to fail is input into MARCH. To the extent that this failure pressure is uncertain (typically it is quite uncertain), it would tend to compound any uncertainties associated with MARCH code calculations. The thermal-hydraulic conditions within the containment can be predicted with relative confidence if the driving forces are well defined. The principal early challenges to containment integrity are the rapid steam generation from core debris interaction with water and the burning of hydrogen. The analysis of steam generation from debris quenching is particularly uncertain and is sensitive to the input and modeling assump-tions utilized. This phenomenology is inherently uncertain and one in which unique answers, except in a bounding sense, cannot be expected. The prediction of the pressures due to hydrogen burning is fairly straightforward if the initial conditions and the timing of the burn are known; however, this is generally not the case, and key assumptions must be made. The amount of hydrogen present in the containment at any point in time is subject to the uncertainties in the prediction of core slumping, vessel failure, ' debris interactions in the cavity, etc. For any set of conditions, the composi-tion of the atmosphere and its potential flamability can be tracked as a func-tion of time. Except in the presence of igniters, however, the occurrence of ignition cannot be predicted and must be assumed. Typically containment integrity 6-55
would be challenged by large coherent burns, but it would not be challenged by extended combustion; the timing of the ignition is the key difference between the two predictions. Tables 6.6 and 6.7 provide the information on which the calculations of radionuclide transport and deposition in the containment were based. Table 6.6 gives containment geometrical data, and Table 6.7 provides contain-ment spray parameters. 6.2 Radionuclide Sources 6.2.1 Source Within Pressure Vessel Inventory The reactor fission product inventory which was used in all four sequences considered in t?is report is based upon ORIGEN calculations for the Surry plant with a three region model, with the maximum burnup corresponding to 33,000 MW days / ton. Table 6.8 contains the inventory of the fission products and control rod structural materials. Since release rate information is not available for all these species, rates for members of the various groups consi-dered in the Reactor Safety Study were taken to be equal when no other informa-tion was available. The nonfission product materials constitute the bulk of the aerosol mass released during core melting. The value for Ag in this table is based on a total of 1060 control rods composed of 81 mol percent Ag,14 mol percent In, i and 4.9 mol percent Cd. The inventory of fission products was distributed
'according to the power peaking factors in Table 6.9, and the control rod and structural materials were distributed according to the volume of each radial node.
Release From Fuel l l The rates for radionuclide release from the fuel were computed using the CORS0R code for the core temperature profiles specific to each accident l sequence. The mass of each species of interest released as a function of time l l l 6-56
is shown in Figures 6.28 through 6.32 for the AB, TMLB', V, and S D2 sequences. Two runs of CORS0R were performed for the S 2 D sequences,c and Y . This was done because the. conditions which give rise to the Y containment failure mode engender somewhat different temperatures in the melting fuel and produce a significantly different time to bottom head failure. In these figures, aerosol materials were considered to be the sum of. fission products Sb, Sr, Ba, Ru, Mo, Zr, and Ap, along with nonfission products Fe, UO 2 , Zr (cladding), Ag, In, Cd,.and Sn. After melt-through of the reactor pressure vessel, the release during the core-concrete interaction was taken as~a release to the containment. The inventory available for release during the core-concrete interaction is listed in Tables 6.10 and 6.11. Regrouping of Released Species In order to track the Reactor Safety Study groups independently, an additional C0RSOR run was performed for the AB and TMLB' sequences which produced release rates for all groups. A description of the makeup and methodo-logy for release of each group follows. e Group 1 (Xe, Kr) -- Xe and Kr releases were sumed and a release rate computed. This group was not previously computed. e Group 2 (I, Br) -- Br release was not considered due to an absence of data concerning Br release and its small inventory relative to I (1:16). e Group 3 (Cs, Rb) -- Thermodynamic and physical r properties of Rb justify treating it identically to Cs. As a result, the Rb inventory was lumped
- into the Cs inventory for treatment by CORS0R.
Releases for I and Cs were combined to produce l release rates for Csl and Cs0H which were the forms assumed to be transporting through the primary system. This assumption is based on the l predicted temperatures and gas compositions com-bined with consideration of the 1 yc 1,6.gmical j thermodynamic equilibrium states. l e Group 4 (Te, Se, Sb) -- Se and Sb were not consi-dered based on their small inventory and lack of l data concerning their behavior. Their inclusion l in this group would be further complicated by the dependence of Te release on Zircaloy oxidation. 6-57 L
4 i
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- t l
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- , i i i o 500 1000 1500 2000 2500 l TIME (sec) i l FIGURE 6.29. CORS0R PREDICTIONS OF SPECIES MASS RELEASED FROM THE CORE DURING THE i TMLB SEQUENCE FOR THE SURRY PLANT l
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o i I I I I o 1000 2000 3000 4000 5000 sooo , i TIME (sec) I i FIGURE 6.31. CORSOR PREDICTIONS OF SPECIES MASS RELEASED FROM THE CORE j DURING THE S29-c SEQUENCE FOR THE SURRY PLANT l i l l 1 1 < i
1 l 30 1 l Cs1
^
G / Te i 5 / 20- - l C e' Aerosol x .01 y i m / 7~~- i < W ' ! Cs0H x .1 1 J ' j a m 2 10 -
/
l
/
i/ 0 i , i 0 500 1000 1500 2000 l TIME (sec) l i l FIGURE 6.32. CORSOR PREDICTIONS OF SPECIES MASS RELEASED FROM THE CORE DURING THE S 0 y SEQUENCE FOR THE SURRY PLANT 2 i I
= e Group 5 (Ba, Sr) -- Ba and Sr were released separately and their releases sumed to form the release rates for this group. Further, their releases were not included in the aerosol materials sum. e Group 6 (Rh, Pd, Tc, Ru, Mo) -- Rh, Pd, and Tc inventories were added to the Ru inventory for purposes of release. The releases of Ru and Mo were then sumed to produce release rates for this group. The aerosol materials sum does not include Mo or Ru releases. e Group 7 (La, Y, Eu, Nd, Np, Sm, Pm, Pu, Zr, Ce, Nb, Pr) -- All members of Reactor Safety Study Group 7 with the exception of Zr were treated identically for purposes of release, using U02 release rate coefficients. Their release and the release of Zr were sumed to produce release rates for this group. Table 6.8.B lists initial inventories of Group 7 members not included in Table 6.8. e Aerosol Materials (Fe, U02 , Zr (cladding), Sn, control rod: Ag, Cd, In) -- The release rate for this group includes only nonfission products. Table 6.8.C lists initial inventories and final CORSOR releases for the Reactor Safety Study groups and compares the results from AB and TMLB' sequences. It is necessary to select an initial particle size for those materials forming the aerosol species. It has been shown I6' I that when significant agglomeration occurs, the initial aerosol size has a negligible effect on subse-quent aerosol behavior after agglomeration has proceeded for a very short time. Nevertheless, initial particle sizes were chosen to correspond to the best available information. Numerous reviews of experimental mean aerosol sizes from vaporizing and condensing fuel will be from slightly below 0.01 um to about 0.1 um with the most likely size being about 0.05 um.(6.4,6.5) A number median radius of 0.05 um and a geometric standard deviation of 1.7 were assumed for the primary particles in the current analyses, and a bulk density of 3 g/cc was assumed for the particles. 6-63
6.2.2 Sources j hin the Containment Radionuclides enter the containment as they are transported through the primary system and on melt-through of the reactor pressure vessel. Material that is still suspended in the RCS is transported into the containment as the RPV and containment pressures equalize. The final source considered is material that is released during the core-concrete interaction. Because of a lack of release information or, even more generally, a lack of evidence that they are of potential importance, sources sometimes postulated as arising from steam explosions (oxidation release) or from jet emission of hot, mo? ten corium at the time of RPV failure were not included in these analyses. r Release from Primary System The source to the containment of material penetrating the prinary system is defined in mass input rate by species of interest and on a time-dependent basis by the output from the TRAP-MELT calculations. Also provided in the TRAP-MELT output is the size distribution of the particulate material. This calculated information is included in Chapter 7. Volatile metals, leaving the primary system are assumed to be condensed as they enter the containment to particles having the same size distribution as the particles otherwise predicted to be released from the primary system. Release from Core-Concrete Interaction VANESA was used to predict aerosol and gas release rates and composi-tions as a function of time. The composition of the core materials contacting the concrete was, as determined with the CORSOR code, to be the materials remaining in the melt at the time of head melt-through. These compositions for the various sequences are given in Tables 6.10 and 6.11. The concrete was taken to be basaltic (CACO 3 = 0.05, Ca0H = 0.09, SiO2 0 = 0.60, H2O = 0.04, Al 02 3 = 0.22), and the initial temperature of the molten material was as calculated with the MARCH code. The total release rates and composition of 6-64
the release are given in Tables 6.12 through 6.16.- These rates and compositions define the source:to the.drywel1 after vessel failure.
. Source Term for Volatile Iodides In a previous section it has been shown that the thermodynamics of the cesium-iodine-hydrogen-oxygen system indicate that iodine will be present primarily as a nonvolatile. iodide in the primary coolant system. After release from'the primary system, a small fraction of the iodine inventory in the con-tainment-is believed to be present;as volatile' iodides.I6*I) The presence of volatile iodide species in containment-type systems has been observed in experimentsI6*0) and in the TMI-2 post-accident. containment atmosphere.(6.7)
At present, the mechanisms responsible for the generation of these volatile iodides are not well understood. Since a theoretical model is not available, an empirical approach has been selected for the formulation of a source term for. volatile iodides. This source term consists of two components. One compo-nent represents the fraction of the containment iodine inventory which is present as volatile iodides Defore containment failure. The second component represents a generation rate for volatile iodides after containment failure. The containment inventory of volatile iodides present prior to containment-failure was estimated from levels observed in TMI-2(6.7) and from estimates of the probable detection limits in relevant experiments.I6*8) The volatile iodide. generation rate was estimated from a conservative evaluation of the measurements of the airborne iodine levels in the TMI-2 containment over the time' period from 100-2000 hours after reactor trip. Based on these estimates it has been assumed for this study that 0.05 percent of the containment iodine inventory will be present as volatile iodides prior to containment failure and after containment failure, additional volatile iodides will be generated at a rate of'2 x 10~7 fraction / hour of the containment iodine inventory. Of this volatile iodine source, it is believed that a fraction of the iodine inventory in a reactor containment will be present as volatile organic iodides (predominantly CH 3 1).I6*II (Other volatile species may also be present.) Therefore, in the analysis of reactor accidents involving a radionuclide release from the reactor system and containment failure, formation in the containment and subsequent release of organic iodides should be considered. 6-65
Unfortunately, the mechanism responsible for the generation of organic iodides has not yet been elucidated. As a result, it is not yet possible to establish a definitive source of organic iodides. Early estimates of the organic iodine source terms were based on a conservative inurpretation of experimental systems studies.(6.6,6.9) Early thermodynamic studies predicted that organic fodides should be present in much smaller concentrations than observed in experi-ments.(6.10) These calculations predicted that CH 3 I would comprise only 10'4 percent of the total gaseous iodine inventory modeled. Experimental data (6.6) and " chemical species specific" measurements of the TMI-2 airborne iodine inventory (6.7) im ly that the concentration uf organic iodides present in a reactor containment during and following an accident may be higher than the concentrations predicted by thermodynamic calculations for an equilibrium system. Additionally, observations of the airborne iodine behavior at TMI-2(6.7) $,pj y the presence of competing sources and sinks for volatile iodine species. In light of these data, a kinetic description may be required to adequately quan-tify the time dependence of the organic iodide concentration in reactor containments during and following reactor accidents. Pending results of studies, such as those which are currently under way,I0* ) use of a general source term for volatile iodides rather than separate source terms for CH3 I. 12 , etc., has been assumed as noted above. I 1 1 l l 6-66 i I
References _(6.1) Technical Base for Estimating Fission Product Behavior During LWR Accidents,NUREG-0772,(June,1981). (6.2) Torgerson, D. F., et al., Fission Product Chemistry Under Reactor Accident Conditions, presented at the International Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, U.S.A. (September 1982). (6.3) Jordan, H., Schumacher, P. M., and Gieseke, J. A., " Comparison of QUICK Predictions with Results of Selected, Recent Aerosol Behavior Experiments", NUREG/CR-2922, BMI-2089 (September, 1982). (6.4) Gieseke, J. A., et al, " Aerosol Source Term for Fast Reactor Safety Analysis", BMI-X-637 (August 11,1972). (6.5) Nuclear Aerosols in Reactor Safety, CSNI/ SOAR No. 1 (June, 1979). (6.6) Postma, A. K. and Zavodoski, R. W., Review of Organic Iodide Forma-tion Under Accident Conditions in Water-Cooled Reactors, WASH-1233 (1972). (6.7) Pelletier, C. A., et al., Preliminary Radioiodine Source Term and Inventory Assessment for TMI-2, SAI-139-82-12-RV (September, 1982). (6.8) Lin, C. C., Chemical Effecs of Gamma Radiation on Iodine in Aqueous Solutions, J. Inorg. Nucl. Chem., g , pp. 1101-1107 (1980). (6.9) Reactor Safety Study: An Assessment of Accident Risks in U.S. Comercial Nuclear Power Plants, WASH-1400 (1974). (6.10) Barnes, R. H., Kircher, J. F., and Townley, C. W., Chemical Equili-brium Studies of Organic Iodide Formation Under Nuclear Reactor Accident Conditions, BMI-1816 (1966). 6-67
v TABLE 6.1. REACTOR CHARACTERISTICS, CONTAINMENT PARAMETERS, AND MARCH OPTIONS FOR LARGE DRY PWR CONTAINMENT ECC storage and injection tanks Accumulator RWST 6 Weight of water 171,300 lb 77,700 kg 2.92 x 10 lb 1.3245 x 106 kg Initial pressure 665 psia 45.9 MPa 14.7 psia 0.1 MPa Temperature 120 F 48.89 C 45 F 7.22 C Fractional value of RWST to start ECC recirculation: 0.01 Fractional value of RWST to start spray recirculation: 0.143 Large LOCA blowdown Time, min Enthalpy Blowdown Rate Btu /lb J/kg lb/ min - ko/s 0 602.7 288,400 2.115 x 100 1.599 x 104 6 4
.20 602.7 288,400 2.115 x 10 1.599 x 10
.201 89.73 42,930 2.770 x 10 5 2.094 x 10 3
.401 89.73 42,930 2.770 x 10 5 2.094 x 10 3 Calculated model input Core heatup section:
Number of radial zones: 10 Number of axial zones: 24 Meltdown model: BOIL model A Core melting temperature: 4130 F (2277 C) Core slumping: Starts when lowest node in a zone has melted Core collapse: Occurs when 75 percent of core has melted Zircaloy - water reaction: Urbanic-Heidrick reaction rate data, hydrogen blanketing, steam limited, continues for melted nodes, reaction of molten Zirca19y in the bottom head calculated. 6-68
TABLE 6.1. (Continued) End of blowdown conditions for large LOCA: Water in vessel: core covered Peak core temperature: 1700 F (927 C) Accumulators: empty Bottom head failure section: Head melting temperature: 2800 F (1538 C) Debris melting temperature: 4130 F (2277 C) Heat loss from top of debris: Radiation to core barrel Debris thermal conductivity: 8 Btu /hr ft F (0.1384 w/cm/C) Tensile strength of vessel: o = min (80,000, 1.49 x 1016 TEMP-3.9105), lb/in2 Reactor cavity processes, debris fragmentation: Particle diameter: 0.5 inch (1.27 cm) Particle thermal conductivity: 2.0 Btu /hr ft F (0.0346 w/cm/C) Reactor cavity processes, concrete decomposition: Metal-concrete interface heat transfer coefficient: HIM = 0.01 w/cm2 g Oxide-concrete interface heat transfer coefficient: HIO = 0.01 w/cm2 g Top surface emissivity: E = 0.5 Heat to cover water: film boiling plus 50 percent of area radiating at internal temperature of top layer. Containment Section: Atmosphere-wall heat transfer coefficient: h = he (TSAT-TWALL) + 0.19 (T-TWALL)4/3 /(T-TWALL) hc = 0 if TSAT < TWALL i 2.0 < h c= Uchida data < 280 Btu /hr ft2 F 2 Containment break area: 7.0 ft2 overpressure failure (0.65 m ) 2 0.349 ft2 isolation failure (0.0324 m ) 6-69
TABLE 6.1. (Continued) Failure of safety systems: (1) Containment failure fails the containment sprays (2) Containment failure fails ECR if sump is saturated. i i I l L l l l. l l l l 6-70 l l
m TABLE 6.2 ACCIDENT EVENT TIMES Event Time, minutes Surry AB-c Core Uncover 9.4 Start Melt 24.8 Core Slump- 42.1 Core Collapse 43.3 Bottom Head Dry 64.4 Bottom Head Fail 110.1 Start Concrete Attack 110.1 Containment Leak 1450.6 End Calculation 1639.6 Surry AB-Y Core Uncover 9.4 Start Melt 24.8 Core Slump 42.1 Core Collapse 43.5 Bottom Head Dry 64.4 Bottom Head Fail 110.1 Start Concrete Attack 110.1 Hydrogen Burn 268.9 Containment Fail 268.9 End Calculation 710.7 6- 71
\ TABLE 6.2 (Continued)
Event Time, minutes Surry AB-8 4 Volumes 2 Volumes Core Uncover 7.1 9.4 Start Melt 24.6 27.2 Core Slump 41.5 44.6 Core Collapse 42.5 45.6 Bottom Head Dry 62.5 65.8 Bottom Head Fail 97.9 104.3 Start Concrete Attack 97.95 104.4 Hydrogen Burn 108.1 198.2 Hydrogen Burn 129.1 Hydrogen Burn 145.4 Hydrogen Burn 226.9 End Calculation 698.3 704.6 Surry TMLB'-6 Steam Generator Dry 67.5 Core Uncover 95.5 Start Melt 118.3 Core S, lump 146.3 Core Collapse 148.0 Bottom Head Fail 152.8 Containment Fail 152.9 Reactor Cavity Dry 177.2 Start Concrete Attack 254.2 End Calculation 1073.4 6-72
TABLE 6.2 (Continued) Event Time, minutes Surry TMLB'- c Steam Generator Dry 67.5 Core Uncover 95.5 Start Melt 118.3 Core Slump 146.3 Core Collapse 147.3 Bottom Head Fail 157.3 Reactor Cavity Dry 214.9 Start Concrete Attack 289.9 Containment Fail 738.2 End Calculation 1100.0 Surry 52D-c Containment Spray Injection On 20.0 Containment Spray Recirculation On 25.0 Core Uncover 27.8 Accumulators Empty 91.5 Containment Spray Injection Off 114.7 Start Melt 134.6 Core Slump 147.3 Core Collapse 148.8 Bottom Head Fail 227.5 Reactor Cavity Dry 325.7 Start Concrete Attack 407.7 End Calculation 2210.4 6-73
l 4 TABLE 6.2 (Continued) Event Time, minutes Surry S2D-Y Containment Spray Injection On 20.0 t Containment Spray Recirculation On 25.0 Core Uncover 27.8 Accumulators Empty 91.4 Containment Spray Injection Off 114.7 Start Melt 134.0 Core Slump 146.6 Core Collapse 150.8 Bottom Head Fail 163.6 Containment Fail 163.7 Reactor Cavity Dry 264.9 Start Concrete Attack 336.9 End Calculation 1114.6 Surry V Containment Fails 0.0 Core Uncover 20.6 L Start Melt 39.7 Core Slump 56.6 l Core Collapse 60.5 Bottom Head Fail 149.9 Start Concrete Attack 149.9 End Calculation 750.2 6-74
TABLE 6.3 CORE AND PRIMARY SYSTEM RESPONSE Primary Primary System System Water Average Core Peak Core Fraction Fraction Accident Time, Pressure, Inventory, Tempera ture, Temperature. Core Clad Event minutes psia lhm F F Melted Reacted Surry TMLB-c Core Uncover 95.5 2369 8.58 x 10 4 669 675 0. O. 4 Start Melt 118.3 2366 5.65 x 10 1990 4130 0.00 0.06 Start Slump 146.3 2362 5.37 x 10 4 3709 4147 0.55 0.33 4 Core Collapse 147.3 2364 4.79 x 10 4130 --- 0.82 0.58 4 Bottom Head Fail 157.3 2368 1.95 x 10 3820 --- --- 0.59 m Surry TMLB-6 Core Uncover 95.5 2369 8.68 x 10 4 669 675 ~ 0. O. 4 Start Melt 118.3 2366 5.65 x 10 1990 4130 0.00 0.06 Start Slump 146.3 2362 5.37 x 10 4 3658 4150 0.55 0.33 Core Collapse 148.0 2367 4.52 x 10 4 4130 --- 0.79 0.89 Bottom Head Fail 152.8 2369 9.33 x 10 4 4130 --- --- 0.93 Surry S2D-Y Core Uncover 27.8 1164 1.01 x 10 5 577 585 0. O. Start Melt 134.0 292 7.03 x 10 4 2067 4130 0.00 0.08 4 Start Slump 146.6 131 6.37 x 10 3688 4139 0.58 0.48 4 i Core Collapse 150.8 337 5.88 x 10 4207 --- 0.77 0.89 Bottom Head Fail 163.6 617 2.46 x 10 5 3925 --- --- 0.89 1 i k
s
.r TABLE 6.3 (Continued)
Primary Primary System System Water Average Core Peak Core Fraction Fraction Accident Time, Pressure, Inventory, Tempera ture, Tempera ture, Core Clad Event minutes psia Ibm F F Melted Reacted Surry S20-c Core Uncover 27.8 1164 1.01 x 10 5 577 585 0. O. Start Melt 134.6 293 7.04 x 10 4 2046 4130 0.00 0.07 Start Slump 147.3 129 6.38 x 10 4 3658 4147 0.58 0.48 Core Collapse 148.8 213 6.20 x 10 4 4130 --- 0.76 0.59 Bottom Head Faii 227.5 18 2.46 x 10 5 4130 --- --- 0.60 Surry V [ Core Uncover 20.6 129 1.11 x 105 567 380 o, o, Start Melt 39.7 24 7.97 x 10 4 2094 4130 0.00 0.05 Start Slump 56.6 17 7.74 x 10 4 3597 4135 0.49 0.22 Core Collapse 60.5 4 122 7.31 x 10 4130 --- 0.75 0.41 Bottom Head Fail 149.9 15 0. 4130 --- --- 0.41 Surry AB-B (2 Volumes) Core Uncover 9.4 44.2 8.32 x 10 4 288 2% 0. O. Start Melt 27.2 36.3 5.64 x 10 4 1952 4130 0.00 0.05 Start Slump 44.6 30.2 5.34 x 10 4 3629 4132 0.54 0.24 Core Collapse 45.6 31.7 5.05 x 10 4 3729 --- 1.0 0.40 Bottom Head Dry 65.8 36.4 0. 3234 --- --- 0.40 Bottom Head Fail 104.3 24.9 0. 4130 --- --- 0.40
.p TABLE 6.3 (Continued)
Primary Primary System System Water Average Core Peak Core Fraction Fraction Accident Time, Pressure, Inventory. Tempera ture, Temperature, Core Clad Event minutes psia Ibm F F Melted Reacted Surry AB-8 (4 Volumes) Core Uncover 7.1 41.3 8.40 x 10 4 286 295 0. O. Start Melt 24.6 29.8 5.62 x 10 4 1962 4130 0.00 0.05 4 Start Slump 41.5 24.0 5.34 x 10 3623 4131 0.54 0.24 4 Core Collapse 5.08 x 10 0.76 0.39 42.5 25.2 4207 --- i Bottom Head Dry 62.5 28.4 0. 3236 --- --- 0.39
% Bottom Head Faii 97.9 18.7 0. 4130 --- ---
0.?9 Surry AB-c/v Core Uncover 9.4 40.2 7.50 x 10 4 296 357 0. O. 4 Start Melt 24.8 35.6 5.70 x 10 1960 4130 0.00 0.05 4 Start Slump 42.1 31.8 5.33 x 10 3644 4139 0.54 0.24 4 Core Collapse 43.5 33.5 5.02 x 10 3732 --- 1.0 0.40
- Bottom Head Dry 64.4 41.2 0. 3299 --- ---
0.40 Bottom Head Fail 110.1 33.9 0. 4130 --- --- 0.40 l l t . ,_
TABLE 6.4 CONTAIMENT RESPONSE Reactor Cogartment Compartment Sump Reector Cavity Steam Accident RWST or CST Sump Water Water Cavity Water Cond. Time, Pressure. Temperature, Event Water Mass. Mass. Temp. Mater Mass. Temp., on Malls minutes psia F llum Iba F 1ho F IbsVoin Surry TEB-c Steam Generator Dry 67.5 13.0 136 3.0 x 10 6 3.39 x 10 4 138 0. --- 1.051 Core Uncover 95.5 28.8 219 3.0 x 106 2.07 x 10 5 197 0. --- 2,498 Start Melt 118.3 25.7 209 3.0 x 106 2.47.x 10 5 200 0. --- 994 Start 51 map 146.3 22.5 197 3.0 x 10 6 2.69 x 10 5 198 0. --- 684 Core Collapse 147.3 3.0 x 106 24.9 205 2.70 x 10 5 197 0. --- 1.378 Bottom Head Fail 3.0 x 106 157.3 45.9 253 2.82 x 10 5 198 1.71 x 105 120 11.810 5 tart Debris / Water Interaction 157.3 46.0 3.0 x 106 2.82 x 10 5 253 198 1.71 x 10 5 120 12,400
? Cavity Dry 214.9 58.6 272 3.0 x 106 4.06 x 10 5 5 Start Concrete 221 0. --- 1,405 Attack 289.9 3.0 x 106 46.0 253 4.54 x 10 5 225 0. ---
653 Containment Fail 738.2 53.7 249 3.0 x 10 6 4.85 x 105 227 0. --- 0 End Calculation 1100.0 14.8 263 3.0 x 106 4.66 x 105 100 0. --- 0 Surry TK B-6 Steen Generator Dry 67.5 3.0 x 10 6 13.0 136 3.39 x 10 4 138 0. --- 1051 Core Uncover 95.5 28.7 219 3.0 x 106 2.07 x 10 5 197 0. --- 2498 Start Melt 118.3 25.8 209 3.0 x 10 6 2.47 x 10 5 200 0. --- 990 5 tart 51tro 146.3 22.3 1% 3.0 x 10 6 2.68 x 10 5 196 0. --- 668 Core Collapse 148.0 26.5 210 3.0 x 106 2.68 x 105 196 0. --- 0
TABLE 6.4 (Continued) l l Reactor Sump Reactor Cavity Steam Compartment Ccapartment RWST or CST Sump Water Water Cavity Water Cond. l Accident Time, Pressure. Temperature Water Mass, Mass. Temp., Water Mass. Temp., on Walls minutes psia F 1ha 1ha F 1ha F lbm/ min Event (Surry TMLB A Continued) l 6 $ 12,300 Botton Head Fail 152.8 46.5 252 3.0 x 10 2.75 x 10 197 0. --- 5 tart Debris / Water Interaction 152.8 46.6 252 3.0 x 106 2.75 x 105 197 1.71 x 105 120 12,510 5 Containment Fail 152.9 86.0 336 3.0 x 106 2.76 x 10 197 8.82 x 10 3 109 0 Cavity Dry 177.2 14.7 202 3.0 x 106 3.16 x 105 202 0. --- 238 Start Concrete 3.0 x 106 3.30 x 105 l l Attack 254.2 14.6 187 187 0. --- 278 End Calculation 1073.4 14.8 207 3.0 x 106 3.33 x 105 159 0. --- 0 l I i 3 sorry s20-c l Containeent Spray g $ g Injection On 20.0 23.4 201 2.90 x 10 2.14 x 10 194 3.% x 10 1% 3,184 l Centaineent Spray 6 5 4 25.0 20.9 2.77 x 10 3.58 x 10 191 4.14 x 10 189 1,301 l Fecirculatica On 191 4 6
- Core Uncover 27.8 19.4 183 2.70 x 10 4.00 x 105 184 6.53 x 10 185 823 6 6 5 Accu,ulate-s E ety 91.5 12.6 126 1.04 x 10 2.08 x 10 131 2.46 x 10 128 0.
5 I Start Pelt 134.6 13.5 136 4.12 x 10 5 2.75 x 106 131 2.46 x 10 128 113 Cere 51tra 147.3 14.2 132 4.12 x 10 --- --- 2.46 x 105 128 27 l l Ccre Collapse 148.3 14.4 132 4.12 x 10 5 --- --- 2.46 x 105 128 45 6 l Bott e Head Fai1 227.5 13.6 144 4.12 a 10 5 2.81 x 10 141 2.46 x 105 128 355 l l
TABLE 6.4 (Continued) Reactor Sump Reactor Cavity Steen Compartment Compartment RM5T or CST Samp Mater Mater Cavity Mater Cand. , Accident Time, Pressure. Temperature. Mater Mass. Mass. Temp., Mater Mass. Temp., en Malls Eweet minutes psia F lies ilus F 1hm F ltm/ min (Surry 52D-c Continued) Start Dettris/ s Water Interaction 227.5 13.6 144 4.12 x 10$ --- --- 2.46 x 10' 128 375 Reacter Cavity Dry 325.7 16.1 162 4.12 x 10 5 3.06 x 10 6 164 0. --- 15 Start Concrete Attack 407.7 13.7 141 4.12 x 105 3.06 x 106 147 0. --- 0 End Calculation 2210.4 13.8 131 4.12 e 105 2.88 x 106 132 2.39 x 10 5 212 12 Sorry 52D-Y I Containment Scray E Injectica Cn 20.0 23.4 2.90 x 106 2.14 x 10$ 201 194 4.76 x 102 1% 3,185 Contaimment Spral Recirculation On 25.0 20.9 190 2.77 x 106 3.58 x 105 191 4.15 x 10 4 189 1,259 Core Uncover 27.8 19.4 183 2.70 x 106 3.93 x 105 184 6.52 x 104 185 824 kcurulatars Doty 91.4 12.6 126 1.04 x 106 2.08 x 105 131 2. M x 105 131 0 5 tart Melt 134.0 13.5 4.15 x 105 2.75 x 106 136 130 2.46 x 105 128 114 Core Slump 4.15 x 105 146.6 14.2 132 2.76 x 100 132 2.46 x 10 5 128 24 Core Collapse 4.15 x 105 150.8 15.4 136 2.76 x 106 132 2.46 x 10 5 128 188 Ection Heat Fail 163.6 23.1 4.15 x 105 2.77 x 106 186 134 2.46 x 10 5 128 6,254 Start Debris / Ester Interaction 163.6 24.0 4.15 x 10$ 2.77 x 106 191 134 2.46 x 10 5 128 6,561
TABLE 6.4 (Continued) 4 Reactor Sump Reactor Cavity Steam Compartment PWST or CST Sump Water Water Cavity Water Cond. Compartment on Walls Time, Pressure, Temperature, Water Mass, Mass, . Temp., Water Mass. Temp., Accident lbm F lbm 'F lbm/ min Event minutes psia F lbm (Surry S2D-Y Continued) 5 6 0 88.6 1696 4.15 x 10 2.78 x 10 134 0. --- Containment Fail 163.7 5 6 491 203 4.15 x 10 2.88 x 10 137 0. --- Reactor Cavity Dry 264.9 14.6 Start Concrete 5 2.89 x 10 6 137 0. --- 0 Attack 336.9 14.7 184 4.15 x 10 5 6- 0. 0 End Calculation 1114.6 14.8 228 4.15 x 10 2.90 x 10 137 --- Surry AB-c 3.0 x 10 6 3.45 x 10 5 247 0. --- 3,878
. Core Uncovers 9.4 40.2 246 6 5 1,797 24.8 35.5 236 3.0 x 10 3.78 x 10 237 0. ---
b Start Melt 6 3.0 x 10 --- 0. --- 0 Start Slump 42.1 31.8 225 --- 6 0. 0 Core Collapse 4.1. 3 33.6 249 3.0 x 10 --- --- --- 6 1,793 Bottom Head Dry 64.4 41.1 245 3.0 x 10 --- --- O. --- Bottom Head Fail 110.1 34.0 205 3.0 x 10 6 4.42 x 10 5 227 0. --- 777 Start Concrete Attack 110.1 34.0 205 3.0 x 106 4.42 x 10 5 227 0. --- 0 6 5 Containment Leak 1450.6 45.8 239 3.0 x 10 4.93 x 10 227 0. --- 0 i End Calculation 1639.6 43.0 288 3.0 x 106 4.93 x 10 5 227 0. --- 0 Surry AB-Y* i 6 5 Containment Fail 268.9 87.4 1152 3.0 x 10 4.52 x 10 227 0. --- 0 0 End Calculation 710.7 14.9 271 3.0 x 10 4.45 x 10 5 213 0. --- 0
- Containment response same as Surry AB-c out to Start of Concrete Attack.
l
w . J ,, - -
<j TABLE ' 6.4 (Continued) ., Compartment Compartment Pres re, Temperature, Sump Reactor Cavity Steam RWST or CST Sump Water Water Cavity Water Cond.
Accident Time, D Water Mass, Mass, Event Temp., Water Mass, Temp., on Walls minutes 1 2 1 2 lbm lbm F lbm F lbm/ min Surry V Containment Fails 0.0 10.0 15.7 100 161 3.0 x 10 6 0. O. O. --- 0/1163* Core Uncover 20.6 10.1 15.8 101 216 3.0 x 10 6 --- ---
- 0. ---
0/0 Start Melt 39.7 10.1 14.8 100 210 3.0 x 10 6 3.38 x 10 4 45 0. --- 0/225 Start Slump 56.6 10.1 14.9 100 428 3.0 x 10 6 3.38 x 10 4 45 0. --- 0/0 Core Collapse 60.5 10.1 15.0 100 739 3.0 x 10 0 3.38 x 10 4 45 0. --- 0/0 Bottom Head Fail 149.9 12.3 12.3 114 141 3.0 x 10 6 3.70 x 10 4 50 0. --- 1084/0 Start Concrete [ Attack N 149.9 12.8 12.8 116 131 3.0 x 10 6 3.89 x 10 4 54 0. --- 1413/0 End Calculation 750.2 14.9 14.9 135 150 3.0 x 10 6 6.62 x 10 4 81 0. --- 0/0 Volume 1/ Volume 2
TABLE 6.4(Continued) Compartment Compartment Sump Reactor Cavity Steam Pressure, Temperature, RWST or CST Sump Water Water Cavity Water cond. Time, psia F Water Mass, Mass, . Temp., Water Mass, Temp., on Walls Accident lbm F- Ibm / min Event minutes 1 2 1 2 lbe lbm F SurryAB-8(2Volumesl 6 Containment Fail 0.6 54.2 15.7 260 99 3.0 x 10 2.24 x 10 5 258 0. --- 3026/0* 6 5 9.4 44.2 15.0 244 155 3.0 x 10 3.40 x 10 245 0. --- 3815/319 Core Uncover 6 5 Start Melt 27.2 36.3 15.0 230 170 3.0 x 10 3.76 x 10 231 0. --- 1525/250 Start Slump 44.6 31.7 15.0 241 171 3.0 x 10 6 3.89 x 10 5 218 0. --- 0/0 6 5 Core Collapse 45.6 31.9 15.0 244 171 3.0 x 10 3.89 x 10 218 0. --- 0/0 Bottom Head Fail 104.3 24.9 15.0 211 181 3.0 x 10 6 4.18 x 10 5 212 0. --- 521/0 Start Concrete Attack 104.4 24.9 14.8 211 181 3.0 x 10 6 4.18 x 10 5 212 0. --- 534/0 i' End Calculation 704.6 14.9 14.8 184 186 3.0 x 106 4.03 x 10 5 176 0. --- 0/0
- Volume 1/ Volume 2
TABLE 6.4 (Continued) Compartment Comparteent Pr re. Temperature Susp Reactor ' C vit Accident RidST or C$f Sump Idater idater Cavity Idater on W11s Time. Event minutes 1 2 3 4 h ter Mast, mss. Temp., Idater Mass. Temp., Itun/ min 1 2 3 4 Itum ibn F Itum F 1 2 3 4 SurryA_B-p(4VoluneM Containment Fall 0.0 15.7 15.7 15.7 15.7 110 109 90 174 3.0 x 106 --- Core Uncover O. --- 0/0* 0/6155' 7.1 41.3 41.3 41.3 41.3 126 193 255 268 3.0 x 106 9.80 x 10 2 120 0. --- 18/1530 2082/1694 Start Melt 24.6 29.9 29.8 29.8 29.8 103 163- 235. 246 3.0 x 10 6 1.008 x 10 3 104 0. --- 0/310 697/0 start Slump 41.5 23.9 23.9 23.9 23.9 98 154 220 377 3.0 x 106 1.006 x 10 3 98 0. --- 0/262 453/0 Core Collapse 42.5 25.2 25.2 25.2 2E.2 109 163 239 614 3.0 x 10 6. 1.008 x 10 3 90 0. --- 0/584 0/0 Bottom Head Fall 97.9 18.8 18.8 18.8 !B.8 124 187 221 215 3.0 x 106 1.139 x 10 3 99 0. .--- 50/0 0/172 Start Concrete Attack 98.0 18.8 18.8 18.8 6 18.8 127 187 221 215 3.0 x 10 1.143x10 3 100 0. --- 66/0 0/156 End Calculation 698.3 14.8 14.8 14.8 14.8 468 148 -279 173 3.0 x 10 6 2.141 x 10 118 0. 0/42 0/0 i
= *
- Volume 1/ Volume 2 Volume 3/ Volume 4 l
l l l l
f
. TABLE 6.5
SUMMARY
OF CONTAINMENT LEAK RATES Leakage _CSIS CSRS Time Leak Start, End, Start, End, Interval, Rate. g Pressure Tm Subsequence min min min min min v/hr MPa psia *F *C Remarks TMLB-6 95.5 4.2E-4 0.20 29 219 104 Core uncovers 95.5-118 4.2E-4 0.19 28 217 103 Core heats 118-146 4.2E-4 0.16 24 201 94 Core melts 146-148 4.2E-4 0.17 25 204 96 Core slumps and collapses 148-152.8 4.2E-4 0.20 29 216 102 Reactor vessel'heatup 152.8 4.2E-4 0.31 46 251 122 Reactor vessel fails 152.8-152.9 4.2E-4 0.51 75 327 164 Boiloff of H p0 152.9 4.2E-4 0.58 86 337 169 Containment fails p 152,9-164.4 7.5 0.26 38 245 118 Boiloff of H 2O E 164.4-166.4 3.8 0.11 16 204 95 Bolloff of H 2O 166.4-168.4 1.9 0.10 15 203 95 Boiloff of H 2O 168.4-173.2 0.9 0.10 15 203 95 Boiloff of H 2O 173.2-254.2 0.01 0.10 15 194 90 Boiloff of H 2O 254.2-272 4.2E-4 0.10 15 186 86 Concrete decomposition 272-325 0.02 0.10 15 184 84 Concrete decomposition 325-567 0.09 0.10 15 180 82 Concrete decomposition 567-627 0.19 0.10 15 204 96 Concrete decomposition 627-1073 0.08 0.10 15 222 106 Concrete decomposition TMLB-c 95.5 4.2E-4 0.20 29 219 104 Core uncovers 95.5-118.3 4.2E-4 0.19 28 217 102 Core heats 118.3-146.3 4.2E-4 0.16 24 203 95 Core melts (a) Normalized to a containment free volume of 1.8 x 106ft . Units 3 are volume fractions per hour.
TABLE 6.5
SUMMARY
OF CONTAINMENT LEAK RATES CSIS CSRS '* ^9' Time Leak Start, End, Start, . End. Interval, Rate.(a) __ Pressure Temp. Subsequence' min min min min min v/hr MPa psia *F *C Remarks TMLB-c (Continued) 146.3-147.3 4.2E-4 0.15 22 197 92 Core slumps and collapses 147.3-157.3 4.2E-4 0.20 29 216 102 Reactor vessel heatup 157.3 4.2E-4 0.31 45 252 122 Reactor vessel fails
- 157.3-290 4.2E-4 0.37 55 267 130 Bailoff of H 2O 290-620 4.2E-4 0.32 48 248 120 Concrete decomposition 620-738.2 4.2E-4 0.35 52 248 120 Concrete decomposition 4
738.2 4.2E-4 0.36 54 249 120 Containment fails 738.2-742.8 7.7 0.26 38 231 111 Concrete decomposition 742.8-770 1.97 0.11 16 236 114 Concrete decomposition 770-1100 0.10 0.10 15 275 135 Concrete decomposition ! ? 8( S20-c 20 2210 20 2210 27.8 4.2E-4 0.13 19 183 84 Core uncovers 27.8-134.6 4.2E-4 0.09 14 140 60 Core heats ! 134.6-147.3 4.2E-4 0.09 14 133 56 ~ Core melts l 147.3-148.3 4.2E-4 0.10 14 132 56 Core slumps and collapses i 148.~3-148.8 4.2E-4 0.17 25 580 305 Hydrogen burns 148.8-148.82 4.2E-4 0.31 46 1424' 773 Hydrogen burns l 148.82-148.83 4.2E-4 0.37 55 1789 976 Hydrogen burns 148.83-149 4.2E-4 0.32 48 1337 725 Hydrogen burns 149-149.2 4.2E-4 0.23 34 659 348 Hydrogen burns 149.2-227.5 4.2E-4 0.10 15 153 67 Reactor vessel heatup (a) Normalized to a containment free volume of 1.8 x 106 ft . Units are volume fractions per hour.
TABLE 6.5 (Continued) Lea age CS CSIS.. v.. . n e,RS
- w. Time Leak Pressure Temp.
Start, End, Start, End, Interval, Rate.(a) Subsequence min min min min min v/hr MPa psia *F 'C Remarks S2D-c (Continued) 227.5 4.2E-4 0.10 14 142 61 Reactor vessel fails 227.5-227.6 4.2E-4 0.13 19 153 67 Initial boiloff of H O-2 227.6-233 4.2E-4 0.21 32 226 108 Boiloff of H 20 233-407.7 4.2E-4 0.11 17 164 73 Boiloff of H O 2 407.7-1248 4.2E-4 0.10 14 145 63 Concrete decomposition 1248-2210 4.2E-4 0.10 14 134 57 Concrete decomposition 52D-Y 20 91.4 20 163.7 27.8 4.2E-4 0.13 19 183 84 Core uncovers { 27.8-134.0 4.2E-4 0.09 14 140 60 Core heats 134.0-150.8 4.2E-4 0.09 14 133 56 Core slumps and collapses 150.8-163.6 4.2E-4 0.11 16 142 61 Reactor vessel heatup 163.6 4.2E-4 0.16 23 186 85 Reactor vessel fails 163.6-163.7 4.2E-4 0.34 50 825 441 Hydrogen burns 163.7 4.2E-4 0.60 87 1795 980 Hydrogen burns 163.7 4.2E-4 0.60 88 1726 941 Containment fails 163.7-164.5 11.1 0.50 73 1308 709 Hydrogen burns 164.5-166.7 8.7 0.31 46 588 309 Hydrogen burns 166.7-201.5 1.7 0.16 17 207 97 Boiloff of H O 2 201.5-407.7 0.2 0.10 15 190 88 Boiloff of H O 2 407.7-477.6 4.6 0.14 20 478 248 Hydrogen burns 477.6-479.1 11.8 0.24 36 1599 870 Hydrogen burns 6 (a) Normalized to a containment free volume of 1.8 x 10 ft3. Units are volume fractions per hour.
TABLE 6.5 (Continued) CSIS CSRS Leakage Subsequence Start, End, Start, End, Int al, R e (a) Pressure Temp. 3 min min min min min v/hr MPa psia 'F 'C Remarks S20-Y (Continued) 479.1-480.7 5.1 0.11 16 694 368. Hydrogen burns 480.7-591.3 0.1' O.10 15 256. 124 Concrete decomposition-591.3-591.5 9.2 0.23 34 1260 682 Hydrogen burns 591.5-592.3 12.3 0.26 38 1785 974 Hydrogen burns 592.3-593.3 9.6 0.14 20 973 523 Hydrogen burns 593.3-750.7 0.1 0.10 15 250 121 Concrete decomposition 750.7-750.8 9.0 0.23 34 1249 676 Hydrogen burns 750.8-751 13.2 0.34 50 2252 1233 Hydrogen burns 751-752.2 11.3 0.22 32 1470 799 Hydrogen burns 752.2-758 1.2 0.10 14 425 218 Concrete decomposition
? 758-1115 0.07 0.10 15 238 114 Concrete decomposition E
AB-c -- -- 9.4 4.2E-4 0.28 40 246 119 Core uncovers 9.4-24.8 4.2E-4 0.27 39 244 118 Core heats l 24.8-42.1 4.2E-4 0.23 33 230 110 Core melts 42.1-43.3 4.2E-4 .23 33 238 114 Core slumps and collapses 43.3-110.1 1.2E-4 0.25 37 238 114 Reactor vessel heats 110.1 4.2E-4 0.23 34 229 110 Reactor vessel fails 110.1-1451 4.2E-4 0.29 42 232 111 Concrete decomposition 1451 4.2E-4 0.31 46 236 113 Containment leaks 1451-1452.4 5.8 0.28 41 230 110 Concrete decomposition 1452.4-1639.8 4.2E-4 0.29 42 295 146 Concrete decomposition h) Normalized to a containment free volume of 1.8ft x. Units 3 106 are volume fractions per hour. I
TABLE 6.5 (Continued). CSRS L'*k*9' C S l_S_ Time Leak Start, End, Start, End, Interval, Rate,(a) essure Tem Subsequence min min min min min v/hr MPa psia *F C Remarks AB-Y -- -- -- -- 9.4 4.2E-4 0.32 46 246 119 Core uncovers 9.4-24.8 4.2E-4 0.27 40 245 118 Core heats 24.8-42.1 4.2E-4 0.23 34 230 110 Core melts 42.1-43.5 4.2E-4 0.23 33 243 117 Core slumps and collapses 43.5-110.1 4.2E-4 0.25 37 238 114 Reactor vessel melts 110.1 4.2E-4 0.23 34 230 110 Reactor vessel fails 110.1-268.9 4.2E-4 0.25 37 230 110 Concrete decomposition 268.9 4.2E-4 0.46 68 770 410 Containment fails /H burns 268.9-269.6 14.7 0.93 137 2567 1408 Hydrogen burns 269.6-271 12.9 0.56 82 1809 987 Hydrogen burns 271-271.9 11.1 0.33 48 1211 655 Hydrogen burns [ 271.9-274.2 9.5 0.19 ' 28 769 409 Hydrogen burns 274.2-277.4 4.4 0.11 16 443 228 liydrogen burns 277.4-299 4.2E-4 0.11 15 345 174 Concrete decomposition 299-710.7 0.07 0.11 15 281 138 Concrete decomposition (a) Normalized to a containment free volume of 1.8 x 10 ft . Units are volume fractions per hour.
TABLE 6.5 (Continued) CSIS CSRS Leikage in C mpartment 1 - Leakage in Compartment 2 Start. End, Start. End. Int al. Ra e a) _ Pnssum Tm l g,[*(b) Pressure Temp. Subsequence min ein min ein min v/hr Wa psia 'F 'C v/hr Wa psia *F- 'C Remarks AC-e(2V) -- -- -- -- 9.4 0.32 0.30 44 244 118 4.2E-4 0.10 15 86 30 Core uncovers 9.4-27.2 0.70 0.29 42 241 116 2.3 0.10 ~15 156 69 Core heats 27.2-44.6 0.50 0.22 33 223 1% 2.2 0.10 15 170 77 Core melts 44.6-46.0 0.59 0.22 32 244 118 2.3 0.10 15 170 77 Core slumps and collapses 46.0-104.3 0.57 0.21 31 226 108 2.3 0.10 15 179 82 Reactor vessel heats 104.3 0.45 0.17 25 212 100 2.3 0.10 15 181 83 Reactor vessel falls 104.3-104.4 1.96 0.17 25 212 100 2.3 0.10 15 181 83 Bolloff e' 'f 0 2 104.4-198.2 0.42 0.15 22 215 102 2.3 0.10 15 181 83 Concrete decomposition 198.2-198.3 94.1 0.13 19 219 104 0.44 0.14 20 1345 730 Hydro 9en burns T 198.3-201.5 4.2E-4 0.13 19 219 104 2.3 0.10 15 540 282 Hydrogen burns 8 201.5-257.4 0.29 0.12 18 228 109 1.95 0.10 15 210 99 Concrete decomposition 257.4-442.1 0.13 0.11 16 238 115 1.0 0.10 15 192 89 Concrete decomposition 442.1-704.6 0.05 0.10 15 125 87 0.43 0.10 15 125 87 Concrete decomposition V (2V) -- -- -- -- 0.04-4.0 4.2E-4 0.07 10 100 38 169.0 0.12 18 232 111 Core uncovers 4.0-40.0 4.2E-4 0.07 10 100 38 14.7 0.10 15 214 108 Core heats 40.0-57.0 4.2E-4 0.07 10 100 38 1.6 0.10 15 253 123 Core melts 57.0-61.0 4.2E-4 0.07 10 100 38 13.0 0.10 15 583 3G6 Core slumps and collapses 61.0-149.9 4.2E-4 0.07 10 100 38 3.3 0.10 15 248 lio Reactor vessel melts (a) Normalized to a containment free volume of 1.8 x 106 ft . Units are volume fraction /hr. (b) Normalized to a compartment free volume of 2.746 x 10 ft3 . Units are volume fraction /hr.
TABLE 6.5 (Continued) CSIS a age n er unt 1 leakage in Concartment 2 CSRS _ Start. End. Start. End. Int al. Ra e I*I """ Ra e (b) Pressure Teen Subsequence min ein ein ein sin v/hr Wa psia 'F *C w/hr Wa psia 'F 'C Remarks V (2V)*(Continued) 149.9 4.2E-4 0.08 12 113 45 4.2E-4 0.08 12 142 61 Reactor vessel fails 149.9-159.R 4.2E-4 0.10 15 118 48 4.2E-4 0.10 15 127 53 Concrete decomposition 159.R-371.5 0.11 0.10 15 126 52 0.76 0.10 15 165 74 Concrete decomposition 371.5-750.2 0.09 0.10 15 133 56 0.64 0.10 15 158 70 Concrete decomposition A8-s(4V) -- -- -- -- 7.1 24.6 0.9 0.26 38 121 49 0.01/0.18/ 0.26 38 185 85 Core uncovers and heats-(Compartments 1 & 2 1.03 are given first. 24.6-41.5 m 0.7 0.18 26 100 38 0.01/ 0.18 26 156 69 Core melts a e Compartments 3 & 4 are on page 9) 1.6E-3/ 1.00 41.5-42.5 0.1 0.17 25 104 40 0.2/ 0.17 25 159 71 Core slumps and collapses 4.2E-4/ 1.01 42.5-97.9 0.9 0.16 24 107 42 0.04/0.55/ 0.16 24 185 85 Reactor vessel melts 0.69 97.9 52.3 0.13 19 122 50 4.2E-4/ 0.13 19 187 86 Reactor vessel falls 2.8/0.64 97.9-97.95 4.2E-4 0.13 19 124 51 0.11/ 0.13 19 197 86 Bolloff of H O 2 4.2E-4 0.64-(a) Normalized to a containment free volume of 4.199 x 104 ft . Leakage, in volume fraction /hr, is from Compartment I to Compartment 2. (b) Normalized to a compartment free volume of 5.380 x 105ft . Leakages 3 are respectively. Compartment 2 to compartment 1. Compartment 2 to Compartment 3. and Compartment 2 to the environment. See footnotes on Page 6 for V(2V).
TABLE 6.5 (Continued) CSIS-teakage in Cowartment 1 Leakage in Compartment 2 CSRS l Start, End, Start, End. Ib al, Ra e (a) P nssum Tm g, (b) Pressure Tema Subsequence min ein ein ein ein v/hr Wa osia *F *C w/hr MPa psia *F 'C Remarks AB-e(4V)(Continued) 97.95 79.6 0.13 19 129 54 4.2E-4/ 0.13 19 187 86 Initial concrete attack 3.8/0.61 97.95-108.1 6.9 0.12 18 174 79 0.32/0.17/ 0.12 18 178 81 Concrete decomposition 0.90 108.1 173.4 0.12 18 225 107 4.2E-4/ 0.12 18 175 79 Hydrogen burns 9.1/0.88 108.1-129.1 9.6 0.11 16 409 209 0.%/0.09 0.11 16 175 79 Concrete decomposition 0.74 129.1 89.5 0.11 16 493 256 5.88/151.1/ 0.11 16 189 87 Hydro 9en burns ? 0.62 8 129.1-145.4 10.5 0.11 16 508 264 4.2E-4/ 0.11 16 219 104 Concrete decomposition 0.14/0.56 145.4 76.9 0.10 15 518 270 6.07/155.4/ 0.10 15 197 92 Hydro 9en burns 0.25 105.4-226.9 8.7 0.10 15 528 276 2.4E-3/ 0.10 15 188 87 Concrete decomposition 0.43/0.13 226.9 56.5 0.10 15 530 277 1.61/ 0.10 15 172 78 Hydrogen burns 4.2E-4/ 0.26 226.9-698.3 5.2 0.10 15 502 261 0.05/0.16/ 0.10 15 161 72 Concrete decomposition-0.17 (a) Normalized to a containment free volume of 4.199 x 10* f t .3 Leakage, in volume fraction /hr, is from Compartment I to Compartment 2. 5 3 (b) Normalized to a compartment free volume of 5.380 x 10 gg . Leakages are respectively, Compartment 2 to Compartment 1. Compartment 2 to Compartment 3 and Compartment 2 to the environment.
TABLE 6.5 (Continued) a age n er en 3 Leakage in Compartment 4 CSIS CSRS Start. End. Start. End. In al. Ra ICI' -' " " " " " Ra (d) Pressure Teen Subsequence min - min min min ein v/hr MPa psia 'F 'C w/hr MPa psia 'F 'C Remarks AB-e(4v) (Continued) 7.1-24.6 0.89/0.11 0.26 38 250 121 5.8 0.26 38 263 128 Core uncovers and heats 24.6-41.5 0.43/0.46 0.18 26 227 108 2.2 0.18 26 253 121 Core melts 41.5-42.5 2.67/ 0.17 25 230 110 30.3 0.17 25 521 272 Core slumps and collapses 4.2E-4 42.5-97.9 0.86/0.40 0.16 24 232 111 4.6 0.16 24 253 123 Reactor vessel melts 97.9 4.2E-4/ 0.13 19 221 105 4.2E-4 0.13 19 215 102- Reactor vessel fails 0.46 97.9-97.95 0.51/0.20 0.13 19 221 105 0.35 0.13 19 215 102 Bolloff of H O 2 Ch 221 105 97.95 4.2E-4/ 0.13 19 4.2E-4 0.13 19 215 102 Initial concrete attack E 0.54 97.95-108.1 0.46/0.13 0.12 18 219 104 0.23 0.12 18 212 100 Concrete decomposition 108.1 2.19/1.2 0.12 18 217 103 1.09. 0.I2 18 210 99 Hydro 9en burns 108.1-129.1 0.39/0.17 0.11 16 213 101 0.32 0.11 16 205 96 Concrete decomposition 129.1 4.2E-4/ 0.11 16 213 101 4.2E-4 0.11 16 202 95 Hydrofen burns 15.0 129.1-145.4 1.71/0.13 0.11 16 299 149 1.18 0.11 16 199 98 s rete decomposition 145.4 2.21/15.94 0.10 15 288 142 1.67 0.10 15 196 'I Hydrogen burns 145.4-226.9 0.6WO.18 0.10 15 302 150 0.78 0.10 15 189 R7 Concrete decomposition 226.9 12.9/ 0.10 15 298 148 155.4 0.10 15 236 113 Hydro 9en burns 4.2E-4 226.9-698.3 0.15/0.13 0.10 15 289 143 0.37 0.10 15 182 83 Concrete decomposition (c) Normalized to i compartment free volume of 9.136 x 105 ft Leakages are respectively. Corrpartment 3 to Compartment 2 and Compartment 3 to Compartment 4. (d) Normalized to a compartment free volume of 1.731 x 10 ft Leakage is from Compartment 4 to Compartment 3.
TABLE 6.6. DIMENSIONS OF PWR USED FOR CALCULATIONS Containment Volume Wall Area Floor Area . Design. Compartment 7t3 m3 T m 2- 7 m2 Large, high pressure Containment 1.80x10 6 5.097x1'0 4 2.36x10 5 2.19x10 4 1.374x10 4 1.277x10 3-General, for V sequence 6 4 5 4 4 3 Containment 1.80x10 5.097x10 2.36x10 2.19x10 1.347x10 ' 1.277x10 - 5 3 3 2 3 2 Aux Building _1'. 50x10 4.248x10 5.25x10 4.88x10 1.875x10 1.742x10 m TABLE 6.7. PWR CONTAINMENT SPRAY PARAMETERS E , i l Flow Rate Height Temperature Droplet Diameter, ! Pumps lb/ min kg/s ft m F C um 4 Injection 2.60x10 - 1.%6x10 2 90 27.4 120 48.89 400-4 2 Recirculation. 5.30x10 4.385x10 90 27.4 120 48.89 400
TABLE 6.8. . INVENTORIES OF RADIONUCLIDES AND STRUCTURAL MATERI ALS
-FOR SURRY DURING IN-VESSEL. PERIOD OF MELTING Fission Products Actinides / Structural Element Mass (kg) Element Mass (kg)
Kr 13.4 U 70,210 Rb 14.7 Pu 469 Sr 47.6 Zr 16,460 Y 22.9 Sn 262 Zr 179 Ag 2,750 Mo 155 In 505 Tc 37.1 Cd 173 Fe 6,486 Ru 104 Rh 20.9 Pd 52.5 Te 25.4 I 12.4 Xe 260 Cs 1 31 Ba 61.2 La 62.3 Ce 1 31 Pr 50.7 Nd 171 Sm 34.0 6-95
TABLE 6.8.B. INITIAL INVENTORIES OF ADDITIONAL SPECIES INCLUDED IN GROUPED RELEASE CALCULATIONS
. Element Mass (kg)
Eu 8.9 Nb 2.7
'Np 26.0 Pm 7.2 t
TABLE 6.8.C. CORSOR RESULTS-WITH AND WITHOUT RSS GROUPING Wit.hout RSS Grouping With RSS Grouping Ini ti al Final CORSOR Initial Final CORSOR Inventory Releases (kg) Inventory Releases 'kg) Group (kg)- AB TMLB (kg) AB TMLB Xe- E73 273 272 273 273 2 71 I 12.4 12.4* 12.3 12.4 12.4* 12.3 Cs 131 1 31
- 130 146 146* 145 Te 25.4 9.3 9.0 25.4 9.3 8.8 Sr NA NA NA 109 71.1 35.7 Ru NA NA NA 369 34.7 16.7 La' NA NA NA 986 0.82 0.29 i Aerosol 106900 2330 1750 106300 2220 1710
? Not Grouped Grouped AB TMLB AB TMLB i Csl 25.4 25.2 25.4 25.2 Cs0H 133 132 150 148
*CORS0R releases these species in the form of Cs! and Cs0H.
6-96 l l O zd
- TABLE 6.9. GE0 METRIC AND POWER PEAKING FACTORS FOR CORE CONFIGURATION OF THE SURRY PLANT Radial Power Axial Power Fraction of Core Peaking Factor Peaking Factor in Radial Zone 1.03 0.09 0.047 1.055 0.26 0.062 1.04 0.43 0.083 1.04 0.59 0.062 1.055 0.74 0.062 1.05 0.89 0.062 1.03 1.02 0.083 1.04 1.14 0.124 0.95 1.24 0.166 0.71 1.32 0.249 1.39 1.43 1.44 1.45 1.43 1.41 1.35 1.27 1.18 1.07 0.94 0.80 0.63 0.49 6-97 v
-v v-- v
TABLE 6.10. MATERIAL PRESENT IN MELT AT TIME OF RPV FAILURE FOR THE SURRY PLANT m . -
'AB-6 TMLB-6 S2 0-c S2D-y V Melt Melt. Melt Melt Melt Content Content Content Content Content Species (kg) (kg) (kg) (kg) (kg)
Cs 0.0 0.7 0.0 0.0 0.0 I 0.0 0.1 0.0 0.0 0.0
'Xe 0.0 1.5 0.0 0.0 0.0 Kr 0.0 0.1 0.0 0.0 0.0 Te .16.4 16.4 5.9 0.61 14.0 Ba 38.6 49.1 38.2 49.1 31.0 Sn 38.0 152 44.0 123 20.0 Ru 102 103 102 103 1 01 UO 79594 79630 79591 79624 79559 2
Zr 9981 6690 6367 1783 9846 Zr0 8760 13210 13640 19830 8942 2 Fe 40520 34140 42110 34050 51880 Mo 125 140 125 142 117 Sr 39.8 43.7 39.7 44.1 36.5 Ag 1410 1460 1410 1400 1400 Cd 36.0 38.0 36.0 36.0 36.0 In 4 31 433 431 430 430 Sb 0.11 0.31 0.11 0.60 0.05 6-98
TABLE 6.11. MELT CONTENT FOR SPECIES NOT SPECIFICALLY INCLUDED IN CORS0R CALCULATIONS FOR THE SURRY-PLANT
.h AB-6 TMLB-6 S2D-c S2D-y V Melt Melt Melt Mel t Melt Content Content Content Content Content Reference Species (kg) (kg) (kg) (kg) (kg) Species Rb 0.0 0.1 0.0 0.0 0.0 Cs Y 22.9 22.9 22.9 22.9 22.9 UO 2
-Tc 36.4 36.7 36.4 36.7 36.0 Ru Rh 20.5 20.7 20.5 20.7 20.3 Ru Pd- 51.5 52.0 51.5 52.0 51.0 Ru
, La 62.3 62.3 62.3 62.3 62.2 UO 2 Ce. 131 1 31 131 1 31 1 31 U0 2 Pr -50.7 50.7 50.7 50.7 50.6 U0 ' 2 Nd 1 71 1 71 1 71 1 71 171 UO 2 Sm 34.0 34.0 34.0 34.0 34.0 U0 2 Pu 469 469 469 469 468 U0 2 Cr 6351 5411 6734 5411 8130 (a) Mn 153 157 153 157 1 51 Fe Ni 3528 3006 3741 3006 4517 (a) (a) These values are taken from the MARCH code predictions. , 5-6-99 i-- - . _ _ _ _ _ _ _ _ _ _ __ . - _ _
l TABLE 6.12. AEROSOL COMPOSITION AND TOTAL RELEASE RATE FOR' SURRY AB EARLY, CORE-CONCRETE INTERACTION Species. 1 Time, sec 0 1200 2400 3600 4800 6000 7200 8400 % 00 10800 12000 13200 Fe0 -- 12.1 9.93 7.56 9.67 14.0 18.9 1.01 0.90 Cr 0 - - 0.90 0.82 0.72 23 - 4 x 10 1 x 10 1 x 10 Mi 2.43 0.44
' 2 x 10 - 2 x 10 1.89 4.76 2.46 0.26 0.10 0.25 0.20 0.21 ab 3 x 10 -6 2 x 10'7 2 x 10 -6 1 x 10 -5 8 x 10 -8 0.18 0.16 4 x 10-6 2 x 10 -8 4 x 10 -5 3 x 10 -5 4 x 10 -5 Ru 3 x 10 -5 2 x 10 -6 2 x 10' I x 10'4 4 x 10'0 5x10 -5 3 x 10-5 7 x 10'I 1 x 10'I 2 x 10~7 1 x 10'I 1 x 10 ~7 sn 0.19 0.06 1 x 10' 9 x 10'"
0.14 0.24 . 0.16 0.04 0.02 0.18 0.17 0.18 0.17 Sb 3 x 10 -5 1 x 10-5 2 x 10 -5 3 x 10 -5 8 x 10 -6 0.17 2 x 10-5 6 x 10 -6 2 x 10-5 2 x 10-5 2 x 10 -5 2 x 10 -5 2 x g5 Te 0.95 0.49 0.62 0.66 0.60 0.35 0.30 1.04 1.05 1.05 1.07 1.09 Ag 24.0 21.3 19.1 14.4 18.5 14.8 8.61 25.0 22.8 23.4 22.2 20.9 Mn 8.89 2.94 5.96 7.32 6.14 1.78 1.11 3.37 3.14 '3.20 3.09 Ca0 -- 12.3 2.96 10.6 8.87 10.7 14.3 12.4 0.55 0.52 0.46 Al 0 23 6xM 4.50 6.79 5.46 1 x 10 2xd 8 x 10 d 9 x 10 0.01 0.44 0.01 0.49 Na 0 7.22 0.01 2 6.05 4.63 5.90 8.33 11.0 5.34 5.46 5.48 5.49 5.41 p K0 -- 12.6 10.6 8.04 10.3 2 14.8 19.1 55.5 59.9 59.3 i SiO2 61.7 64.3 26.0 21.4 16.5 20.87 30.1 27.9 6.98 5.26 5.24 4.22 3.23 $ 00 2 3.24 0.35 1.43 3.95 1.67 0.13 0.06 0.63 0.53 0.54 Zr0 0.51 0.48 2 0.04 6 x 10'3 0.02 0.08 0.03 6 x 10'3 8 x 10'3 0.03 0.03 0.03 0.03 0.03 Cs 0 -- -- -- -- 2 -- -- -- -- -- -- -- -- Ba0 3.46 1.30 1.03 0.90 0.79 0.44 0.20 0.03 0.02 0.02 Sr0 8.53 2.05 0.02 0.02 2.14 2.20 1.59 0.52 0.19 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10~3 2 x 10'3 La 0 3.81 2 x 10~4 2.04 6.62 23 '2.52 1 x 10'4 Ce0 7.38 0.70 1.x 10'4 4 x 10'4 5 x 10-4 4 x 10'4 5 x 10 -4 5 x 10 -4 2.55 6.45 2.60 0.15 '0.02 6 x 10'4 9 x 10-4 6 x 10'4 7 x 10'4 7 a 10'4 Nb 4 x 10 2 x 10 0 2 x 10 I x 10 0 25 1 x 10 4 x 10 2 x 10 4 7 x 10 4 7 x 10 7 x 10 7 x 10 7 x 10 Cs! -- -- -- -- -- ' Cd 36.9 -- -- -- -- -- - -- -- -- -- -- Source pate 170 97.1 33.1 123 230 183 g/sec 98.3 17.5 10.6 11.2 12.8 10.3
f-i' L TABLE 6.12. (Continued) Species. Time, sec 1 14400 15600 16800 18000 19200 20400 21600 22800 Fe0 0.64 0.54 0.40 0.32 0.29 0.30 0.31 . ~0.32 Cr 0 23 3 x 10 4 x 10 5 x 10 6 x 10 0 7 x 10 7 x 10 7 x10 7 x 10 Ni 0.14 0.12 0.09 0.07 0.07 0.07 0.07 0.07 Mo 7 x 10 -5 9x10 -5 1 x 10 -4 2 x 10 -4 6 x 10 -4 1 x 10'3 1 x 10 ~3 1 x 10 -3 Ru -- -- -- -- -- -- -- -- Sn 0.18 0.18 0.17 0.18 0.21 0.24 0.25 0.26
-5 Sb 2 x 10 -5 2 x 10 -5 2 x 10 2 x10
-5 2x10 -5 2 x 10 -5 3 x 10 -5 3x10 -5 Te 1.10 1.12 1.14 1.16 1.19 1.20 1.20 1.20 Ag 20.0 18.5 15.7 14.4 14.1 14.4 14.5 14.4 Mn 2.88 2.72 2.43 2.32 2.36 2.46 2.50 2.50 Ca0 0.50 0.49 0.45 0.46 0.51 0.56 0.58 0.58 A1 0 0.01 0.02 0.02 0.02 0.02 0.02 0.02 0.02 23 7 Ma 0 2
5.33 5.22 5.00 4.71 4.44 4.31 4.26 4.24 o K0 2 66.0 68.7 73.0 75.1 75.6 75.2 75.0 75.1 510 2 2.58 1.88 1.12 0.75 0.59 0.55 0.54 0.53 UO 2 0.48 0.46 0.42 0.44 0.51 0.58 0.60 0.60 Zr0 0.03 0.04 0.04 0.04 0.05 0.05 0.04 0.04 2 Cs 0 -- -- -- -- -- -- -- -- 2 Ba0 0.02 0.03 0.03 0.03 0.04 0.04 0.04 0.04
-3 Sr0 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10 -3 2 x 10'3 2 x 10~3 2 x10 -3 2 x 10 La 0 23 5 x 10'4 5 x 10'4 6 x 10 -4 6 x 10'4 7 x 10'4 7 x 10
~4 7 x 10'4 6 x 10'4
-4 Ce0 7 x 10'4 8 x10 9 x 10'4 9 x 10'4 1 x 10~ 1 x 10'3 1 x 10'3 9 x10' Nb 0 8 x10 4 8 x 10 9 x 10 9 x 10 1 x 10 0 1 x 10 0 1 x 10 1 x 10 25 Cs! -- -- -- -- -- -- -- --
Cd -- -- -- -- -- -- -- -- Source Pate 9.1 8.5 7.1 5.5 3.8 3.0 3.0 3.0 g/sec
r y TABLE 6.13. AEROSOL COPPOSITION AND TOTAL RELEASE RATE ~FOR-SUP.RY AB LATE, CORE-CONCRETE INTERACTION Species Time, sec
% 0 1200 2400 3600 4800 6000 7200 8400 % 00 10800 12000 13200 Fe0 --
13.6 12.0 10.7 .9.20 10.5 12.2- 13.2 14.2 15.5 22.1 1.31 ~ Cr23 0 - - - - - - - - - - - 0.12 Ni 2.43 0.22 0.47 1.20 2.60 1.55- 0.52 0.38 .0.26 0.20 0.21 0.39 Mo 3 x 10-6 6 x 10-8 2 x 10~7 1 x 10-6 5 x 10-6 2 x 10-6 3 x 10'7 2 x 10~7 8 x 10 -8 -8 5 x 10
~
4 x 10-8 8 x 10-5 Ru 3 x 10-5 5 x 10'I 2 x 10-6 1 x 10-5 4 x 10-5 1 x 10-5 2 x 10-6 1 x 10 -6 7 x 10~7 5 x 10 ~7 4 x 10'I 6 x 10~7 Sn 0.19 0.04 0.06 0.11 0.17 0.12 0.06 0.05 0.04 0.03 0.04 0.23 Sb 3 x 10 4 1 x 10-5 1 x 10-5 2 x ID'S 3 x 10 -5 2 x 10-5 1 x 10-5 ' 1 x 10 8 x 10-6 8 x 10-6~ 9 x 10-6 2 x 10-5 Te 0.95 0.44 0.48 0.58 0.65 0.58 0.43 0.39 0.35 0.34 0.42 0.98 Ag 24.0 14.2 22.3 20.6 17.7 20.2 22.8 18.4 14.4 12.6 14.2 30.0 Mn 8.89 2.14 3.04 4.85 6.90 5.21 2.81 2.36 1.89 1.68 1.93 4.17 Ca0 -- 13.80 12.3 11.2 10.1 11.2 12.6 13.6 14.5 15.3 3.83 0.70 A1 0 - 6 x 10 7 x 10 3.04 5.92 3.73 1 x 10 1 x 10 23 1 x 10 1 x 10 2 x 10 6 x 10 Na 0 8.09 [ 2 7.20 6.50 C.62 6.38 7.32 7.89- 8.38 9.09 12.5 4.80-o K0 14.2 12.8 11.4 9.79 11.20 13.0 13.9 14.7 15.6 2 20.2 44.1 510 -- 29.4 25.8 23.1 19.9 22.6 26.2 2 28.4 30.5 29.0 24.3 12.2 00 3.24 0.20 0.34 0.87 1.98 1.02 0.28 0.18 0.11 2 0.09 0.09 0.97 Zr0 0.04 0.007 0.006 0.015 0.035 0.018 0.006 0.006 2 0.006 0.006 0.009 0.02 Cs 0 0 0 -- -- -- -- -- 2 -- -- -- -- -- Ba0 3.46 1.36 1.03 1.03 0.97 0.84 0.57 0.40 0.29 0.19 0.05 0.03 Sr0 8.63 1.86 1.65 1.% 2.14 1.62 0.85 0.55 0.36 0.23 0.05 0.03 La 0 23 3.81 2 x 10'4 2 x 10 ~4 1.15 3.01 1.44 1x' 1 x 10'4 1 x 10'4 1 x 10'4 1 x 10'4' 3 x 10'4 Ce0 7.38 0.36 0.63 1.59 3.39 1.71 0, 0.20 0.10 0.05 t 0.009 5 x 10'4 Nb 0 2$ 4 x 10 3 x 10 2 x 10 4 1 x 10 4 x 10 2 x 10 3 x 10* 2 x 10 4 1 x 10 4 2 x 10 4 2 x 107 5 x M Csl -- -- -- -- -- -- -- -- -- -- -- -- Cd 36.9 -- -- -- -- -- -- -- -- -- -- -- Source Rate 158 80.5 205 214 73 136 119 84 73.1 62.2 45.3 17.6 (above pool) g/sec
TABLE 6.13. (Continued) Species sec
% 14400 15600 16800 18000 Time 19 $if 20400 71600 22800 24000 Fe0 1.22 1.01 0.91 0.80 0.69 0.59 0.43 0.36 0.33 Cr23 0' 6xM 4 1xM 2xM 2xd 3xd 4xM SxM 5xM 6xM Ni 0.33 0.27 0.22 0.18 0.16 0.13 0.09 0.08 0.07
-5 -5 -5 -5 -4 -4 Mo 7 x 10 5 x 10 6 x 10 7 x 10 8 x 10-5 1 x 10 1 x 10 2 x 10'4 3 x 10' Ru 4 x 10'I 2 x 10'I 2 x 10'I 1 x 10'I B x 10-8 ,, ,, ,, ,,
$n 0.22 0.20 0.20 0.19 0.19 0.19 0.17 0.18 0.20
-5 Sb 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10
-5 2 x 10 -5 2 x 10-5 2 x 10-5 2 x 10 Te 1.01 1.06 1.09 1.11 1.13 1.15 1.17 1.19 1.21 Ag 28.3 26.0 23.8 22.2 20.7 19.3 15.9 14.9 14.7 Mn 4.01 3.78 3.55 3.38 3.22 3.06 2.66 2.58 2.60 Ca0 0.60 0.52 0.52 0.54 0.53 0.52 0.46 0.47 0.50 A1 0 7 x 10~3 8 x 10-3 0.01 0.01 0.01 0.01 0.018 0.02 0.02 23 5.24 5.05 4.82 4.64 Na 0 5.06 5.30 5.38 5.36 5.33 m 2
--* K0 48.6 54.1 58.8 61.8 64.5 67.2 72.4 73.9 74.4 2 w 510 9.71 7.03 4.98 3.78 2.89 2.14 1.21 0.89 0.74 2 UO 0.82 0.67 0.58 0.54 0.50 0.48 0.42 0.44 0.48 2 2r0 0.03 0.03 0.03 0.03 0.03 0.04 0.04 0.04 2 0.02 Cs20 -- -- -- -- Ba0 2 x 10'3 0.02 0.02 0.025 0.026 0.027 0.03 0.03 0.03 Sr0 2 x 10~3 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10-3 0.002 0.002 7 x 10 -3
~4 ~4 La23 0 4 x 10~4 4 x 10 -4 4 x 10'4 5 x 10-4 5 x 10 5 x 10 6 x 10'4 6 x 10'4 7 x 10'4
~4 -4 7 x 10'# 9 x 10'4 9 x 10'4 9 x 10'4 Ce0 5 x 10'" 6 x 10'4 6 x 10 7 x 10'4 7 x 10 Nb 0 25 2
5 x 10 6 x 10 7 x 10 7xM 7x 8xM 9xd 9 x 10 4 1 x 10 4 Csl -- -- -- Cd 95.7 37.8 29.1 27.9 -- Source 8.5 6.3 4.3 3.6 Rate 18.4 17 8 13.2 10.3 8.9 g/sec
TABLE 6.14. AEROSOL C0f90SITION ANC' TOTAL RELEASE RATE FOR SURRY TMLB, CORE-CONCRETE INTERACTION Species. 1 Time. sec 0 1200 2400 3600 4800 6000 1200 8400 9600 10000 32000 13200 Fe0 -- 12.5 10.3 10.8 13.8 15.2 16.2 16.9 18.2 21.0 Cr 0 1.04. 1.15 23
-- -- 0.03 0.01 0.01 -- -- -- -- -- 0.11 8.18 Ni 2 a 10'3 6 x 10 -2 0.14 0.08 0.15 0.06 0.08 0.09 0.10 0.13 0.37 0.44 Mo --
1 a 10'8 4 a 10'I 2 a10 -8 4 x10 -8 2 x 10-8 1 a 10
-8 2 a 10
-8 2 x 10-8 3 a10 -8 9 a 10 2 a 10'4 av --
9 x 10'8 4 a 10'I 2 x 10'I 3 a 10*I I a 10'I 1 x 10'I 1 x 10'I 2 x 10'I 2 a 10'# 7 x 10',# I a 10-6 Sn 0.05 0.12 0.16 0.12 0.18 0.13 0.13 0.14 0.15 0.18 1.16 1.29 Sb 1 x 10-5 3 a 10 -5 3 x 10-5 3 x 10-5 4 x 10-5 3 x 10-5 3 a 10-5 3 a 10-5 3 a 10-5 4 x10 -5 1 a 10-4 1 x 10'4 Te 0.25 0.39 0.43 0.37 0.51 0.44 0.43 . 0.44 0.48 0.54 1.38 1.26 Ag 1.20 7.94 12.2 8.96 14.0 9.71 9.43 10.3 11.6 13.99 38.2 40.0 Mn 0.29 1.41 2.02 1.53 7.34 1.69 1,64 1.77 1.98 2.36 6.34 6.46 Ca0 -- 1.14 4.28 6.99 ' l3.6 12.9 13.0 12.6 11.0 5.26 0.39 A1 0 0.51 g 23 5 a 10-5 1 a 10-4 3 a 10 6 m 10'4 8 a 10~4 I a 10-3 I s 10~3 1 s 10-3 2 x 10 ~3 4 x 10'3 4 a 10'3 L Ma 0 2 7.06 5.92 6.25 8.09 8.92 9.47 9.89 10.6 12.1 4.01 3.69 g K02 11.2 9.16 10.7 14.0 15.8 16.7 17.3 18.1 20.0 35.9 31.6 510 -- 15.4 21.0 23.1 2 29.7 32.8 31.3 29.3 26.8 24.0 U0 9.90 11.9 0.10 0.10 0.15 0.09 0.14 0.08 0.07 0.07 0.07 0.08 1.13 1.37 2 2 0.02 9 a 10~3 8 x 10~3 7 a 10'3 9 a 10~3 9 a 10'3 9 a 10'3 9 a 10~3 9 a 10'3 0.01 0.02 0.02 Cs 0 0.25 0.10 0.08 0.07 - 2 -- -- -- -- -- -- -- Sao 0.59 1.99 1.68 1.30 1.42 1.03 0.72 0.55 0.40 0.16 0.03 0.04 Sr0 1.27 2.20 2.23 1.55 1,79 1.14 0.78 0.60 0.44 0.18 4 a 10'3 5 a 10'3 ta y3 0 4 x 10'8 2 x 10' 2 x10 -4 2 x 10-4 2 a 10-4 2 a 10'8 2 x 10'8 2 m 10 2 a 10'4 2 x 10 4 a10 -4 3 a 10'4 Ce0 6 a 10'8 0.13 0.27 0.13 0.20 0.08 0.05 0.04 0.04 0.02 6 a 10 5 a 10 Nb 0 25 6m M 3xM 3mM 2mM 3aM 3 a 10 2aM 2 a 10 2 a 10 2 a 10 5 a 10 4xM Cs1 0.16 0.40 0.24 0.28 0.01 1 x 10'3 2 a 10'8 7 a 10'$ 4 x 10'$ 2 a 10 3 a 10'$ 2 a 10 -5 Cd 95.7 37.8 29.1 27.9 -- -- -- -- -- -- -- -- Source pate I.9 3.1 50.0 58.2 75.8 85.4 55.8 47.8 (above 43.6 40.3 12.6 7.9 cool) g/sec l
~
.J TABLE 6.14. (Continued)
Time, sec Species. . 25200 27600 cisisuu 16800 18000 19200 20400 21600 22800 24000 26400 1 14400 15600 - 1.47 1.47 1.44 1.38 1.28 1.17 1.06 0.95 0.86 Fe0 1.26 1.33 1.40 1.45 0.32 0.33 0.31 0.28 0.24 - 0.20 0.16 2 s 10'3 2 s 10'3 2 x 10'3 Cr23 0 0.23 0.27 0.30 0.65 0.65 0.61 0.56 0.50 0.44 0.38 0.34 0.30 0.27 Nt 0.52 0.58 0.62 5 s 10'" 5 a 10'" 5 a 10'" 5 a 10'" 5 a 10 5 a 10"" 5 a 10'" $ s 10 6 a 10'" 8 x 10'" No 3 s 10
- 3 s 10'" 4 a 10'"
3 a10 -6
-6 2 a10 -6 2 a10 -6 1 a 10 -6 1 x 10 -6 7 a 10'I 5 a 10'I 4 s 10'I 3 x 10'I Ru 2 a10 -6 2 a 10-6 3 a 10-6 3 , yg 1.63 1.63 1.63 1.62 1.60 1.61 I.64 1.68 Sa 1.40 1.47 1.54 1.58 1.62
-5 1 a 10' 1 a 10'" 1 : 10'#
Sb 1 10 '9 s10 -5 9 s10 -5 9 x10 9s10' 9 a 10'$ 9 a 10-5 9 a 10'I 1 m 10'" 1 a 10*# 0.92 0.96 4.02 1.08 1.16 1.23 1.29 1.33 1.37 Te 1.14 1.04 0.97 0.93 42.1 41.6 40.8 39.6 37.9 35.5 35.1 33.8 A9 41.4 42.0 42.3 42.4 42.3 6.27 6.28 6.25 6.17 6.05 5.92 5.79- 5.67 mi 6.48 6.44 6.36 6.29 6.27 0.74 0.75 0.76 0.76 0.75 0.14 0.73 0.73 0.73 0.70 0.72 L Ca0 0.63 0.67 3 3 s 10' 4 a10' 5 s 10'3 6 s 10'3 7 a 10~3 8 a 10'3 9 a 10'3 Aly3 0 3 a10' 3 a 10'3 3 a10 2 a 10'3 2 x 10'3 3 a 10'3 $ 3.10 2.93 2.83 2.87 3.03 3.26 3.50 3.77 3.98 4.13 4.21 4.25 us20 3.35 21.7 23.5 26.3 29.6 33.6 37.6 41.0 43.9 46.4 ft 0 27.4 24.4 22.4 21.3 2 6.10 4.81 3.83 16.7 18.4 19.4 19.0 17.3 15.0 12.51 9.97 7.76 510 14.5 2 n.48 1. 33 1.24 1.17 1.14 UO 1.65 1.87 2.04 2.14 2.13 2.01 1.84 1.67 2 0.02 0.02 0.02 0.02 2roy 0.02 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.02 Csy 0 -- -- -- - 0.04 0.04 0.04 0.04 0.04 0.04 0.05 0.05 0.05 Ba0 0.04 0.04 0.04 0.04
-3 Sr0 6 a 10'3 6 a 10-3 5 s 10'3 5a 10 -3 5 a10 ~3 5 m 10'3 5 a 10 -3 5 a 10 3 a 30-3 5 a 10'3 5 s 10'3 5 m 10~3 5 m 10'3 2: 10'" 2 a 10 "
~
10"" 3 s 10'" 3 a 10'" 3 s 10 4 a 10'" 4 s 10~" Lay3 0 2 s 10'" 2 s 10'" 2 s 10'" 2 a 10'" 2 a 10'" 2
-8 6 a10 4 s 10"" 4 a 10'" 4 x 10'" 4 s 10'" 4 a 10'" 4 10'" 4 x 10'" 4 s 10 5 a 10"" 5 a 10'# 5 s 10'"
Cc0 5 a 10'" 2 7 a 10~" 10-8 10'3 1 a 10'3 1 a 10'3 1 10' 8 s 10
-6 3 a 10'0 4 a 10 -6 4 s 10
-6 5a10 -6 5 s 10 -6 5 a 10-6 Nby5 0 9 1 Csl 2 s 10-5 2 s 10-5 1 s 10'$ 1 T0'S -- -- -- --
C4 -- -- Source 15.0 15.7 10.6 9.0 11.4 14.6 18.8 23.4 24.9 23.4 21.1 18.1 Rate 8.4 (above Pool) gh*C
TABLE 6.15. AEROSOL COMPOSITION AND TOTAL RELEASE RATE FOR SURRY S2D, CORE-CONCRETE INTERACTION Species. 1 Tine. sec 0 1200 2400 3600 4800 6000 7200 8400 9600 10800 12000 13200 Fe0 -- 13.7 12.1 11.3 16.6 19.9 20.0 20.1 20.7 22.1 0.96 1.10 Cr 0 -- -- 0.02 0.01 23 -- -- -- -- -- -- 2 x 10 4 x 10-4' Ill 8 x 10'4 0.04 0.11 0.12 0.09 0.04 0.04 0.05 0.06 0.08 0.31 0.39 pts -- 2 x 10 -8 2 x 10-8 1 x 10-8 4 x 10 -5 6 x 10 -5
-8 av 3 x 10-8 2 x 10~7 2 x 10*7 1 x 10'I 3 x 10 -8 4 x 10 -8 5 x10 -8 3 x 10 7 x 10-8 J x 10'I 5 x 10'I '
Sa 3 x 10'3 0.01 0.02 0.02 0.02 0.01 0.01 0.02 0.02 0.02 0.22 0.26 Sb 6 x 10'I 2 x 10'0 4 x 10 -6 4 x 10-6 4 x 10 -6 3 x 10'0 3 x 10-6 3 x 10'0 3 x 10-6 3 x 10-6 2 x 10-5 2 x 10-5 Te 0.05 0.11 0.13 0.13 0.15 0.13 0.13 0.14 0.14 0.15 0.54 0.50 A9 0.46 4.66 8.98 8.99 8.41 5.69 5.69 6.41 1.17 8.23 31.3 34.0 les 0.12 0.87 1.54 1.52 1.49 1.07 1.06 1.18 1.30 1.48 5.54 5.83 Ca0 -- 0.47 2.56 5.30 8.58 9.22 9.% 10.2 9.35 6.72 0.36 0.45 m Alp 3 -- 3 x 10-5 1 x 10'4 2 x 10'4 5 x 10~4 9 x 10'4 1 x 10' I x 10'3 1 x 10'3 2 v 10'3 6 x 10~3 5 x 10'3 g Ita 20 -- 7.70 6.91 6.50 9.64 11.6 II.6 11.7 12.0 12.8 4.86 4.56
- K0 --
12.1 11.2 10.8 16.5 20.3 20.6 20.4 2 20.7 21.6 47.4 42.1 SiO -- 15.4 17.7 21.8 29.8 2 30.1 29.5 28.7 27.6 26.2 7.54 9.63 UD 0.03 0.08 0.14 0.12 0.11 0.08 0.06 0.06 0.06 0.06 0.91 1.12 2 7 0.02 0.01 9 x 10'3 8 x 10'3 0.01 0.01 0.01 0.01 0.01 0.01 0.04 0.03 Cs 0 -- -- -- -- -- 2 -- -- -- -- -- -- -- Be0 0.22 1.16 1.36 1.10 1.22 0.73 0.52 0.42 0.33 0.21 0.03 0.03 Sr0 0.69 1.60 1.95 1.61 1.56 1.03 0.72 0.57 0.43 0.26 4 x 10'3 4 x 10'3 La 0 4 x 10'4 2 x 10'4 2 x 10'4 2 x 10'4 2 x 10'4 2 x 10'4 2 x 10 -4 2 x 10
~4 2 x 10'4 2 x 10'4 8 x 10~4 8 x 10 Ce0 6 x 10'4 0.07 0.19 0.18 0.11 0.04 0.03 0.03 0.02 0.02 8 x 10'4 8 x 10'4 feb 0 6 x 10 -6 4 x 10 -6 3 x 10 -6 4 x 10 -6 3 x 10 -6 3 x 10 -6 3 x 10 -6 2$ 3 x 10-6 3 x 10-6 3 x 10-6 8 x 10 -6 8 x 10'4 Csl -- -- -- -- -- -- -- -- -- -- -- --
Cd 98.3 42.0 35.0 30.4 5.68 -- -- - -- -- -- -- Source Rate 0.8 8.1 20.0 32.2 51.4 59.6 45.0 40.1 38.2 38.4 8.8 5.3 (after pool) g/sec
TABLE 6.15. (Continued) Species. Time, see 14400 15600 16800 18000 19200 20400 21600 22800 24000 25200 26400 27600 1 1.24 1.37 1.47- 1.55 1.62 1.67 1.71 1.72 1.71 1.65 1.56 1.44 Fe0 0.20 0.26 0.31 0.36 0.39 0.42 0.43 0.41 0.3P 0.33 0.28 0.23 Cr230 ut 0.48 0.57 0.66 0.72 0.77 0.80 0.80 0.76 0.70 0.63 0.55 0.47
-4 ~4 ~4 -4 N 1 x 10'4 2 x 10'4 2 x 10 3 x 10'4 4 x 10 4 x 10'4 5 x 10'4 4 x 10'4 4 x 10~4 3 x 10 3 x 10'4 3 x 10 au 8 x 10*# 1 x 10 -6 2 x10-6 2 x10
-6 3 x10 -6 3 x 10-6 3 x 10-6 3 x 10-6 2 x 10-6 2 x 10-6 1 x 10 -6 8 x 10'I Sa 0.29 0.32 0.35 0.37 0.38 0.39 0.40 0.40 0.40 0.39 0.38 0.37
-5 -5 -5 -5 -5 -5 -5 -5 5 Sb 2x10 -5 2 x10 2 x10 2x10 2 x10 2x10 2 x 10-5 2x10 2 x10 2x10 2 x107 2 x 10-5 Te 0.46 0.41 0.38 0.34 0.32 0.31 0.30 0.32 0.34 0.36 0.39 0.42 Ag 36.1 37.8 38.6 39.0 39.0 38.9 38.8 38.7 38.4 37.7 36.6 35.2 m 6.00 6.07 6.06 5.98 5.80 5.70 5.66 5.80 5.82 5.82 5.77 5.65 Ca0 0.59 0.67 0.71 0.74 0.77 0.79 0.81 0.82 0.82 0.02 0.81 0.19 A1 0 4 x 10'3 4 x 10'3 3 x 10'3 3 x 10'3 2 x 10'3 2 x10 ~3 2 x10 -3 3x10 -3 3 x 10~3 4 x 10~3 5 x 10'3 6 x 10'3 23 4.03 4.31 e ma 0 4.17 3.80 3.48 3.23 3.06 2.% 2.% 3.13 3.40 3.72 2
36.5 31.4 27.5 24.6 22.7 21.6 21.6 23.4 26.3 30.3 34.5 38.9 gK02 16.4 13.4 10.7 N $102 12.4 15.5 18.4 20.9 22.8 23.9 24.0 22.2 19.6 UO 1.40 1.70 1.% 2.18 2.33 2.42 2.42 2.27 2.06 1.81 1.58 1.38 2 2r02 0.02 0.02 0.02 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.02 0.02 Cs 0 -- -- -- -- -- -- 2 Ba0 0.04 0.04 0.04 0.04 0.04 0.03 0.03 0.03 0.03 0.03 0.04 0.04
-3 -3 Sr0 5 x 10 6 x10
~3 6 x 10'3 6 x 10-3 6 x 10'3 6 x 10'3 5 x 10'3 5 x 10-3 5 x 10~3 5 x 10'3 5 x 10~3 5 x 10 4
La23 0 4 x 10 3 x10 4 2 x10 4 2 x10 2 x 10 2 x 10 2 x10 2 x 10 2 x 10 2 x 10 2 x 10 3 x 10
-4 ~4 5 x 10 Ce0 6 x 10 5 x10 ~4 5 x 10-4 5 x 10 5 x 10'4 5 x 10'4 4 x 10'4 4 x 10 4 x 10'4 4 x 10 4 x10
-6 4 x 10 -6 2
Mb 0 5 x10
-6 4 x 10 -6 9 x 10'4 1 x 10 -3 1 x 10'3 1 x 10'3 1 x 10'3 1 x 10'3 1 x 10'3 8 x 10'4 4 x 10 25 -- -- -- -- --
Cs1 -- -- Cd Sourte Rate 4.9 6.2 8.4 11.2 14.2 17.0 21.9 24.6 23.8 21.0 18.0 15.2 (above pool) s/sec
TABt.E 6.15. (Continued) Species. Time, see 1 2560s 30000 31200 32400 33600 34800 MNJn Fe0 1.31 1.19 1.08 0.98 0.89 0.78 0.69 Cry30 2 a10 2 m 10 O ml 0.41 3 a 10 3 a 10 3 a10 4mM 4 a 10 0.36 0.32 0.26 0.26 0.22 0.19 Me 3 x 10'" 3 a 10 3 a 10 3 x 10'4 4 x 10 4 x 10'4 6 x 10'4 Au 6 a 10'I 4 x 10'I 3 a 10'I 2 x 10' 2 x 10'# 1 x 10' I a 10'I Sn 0.35 0.35 0.34 0.34 0.35 0.35 0.35 56 2 a 10-5 2 x10 -5 2 a 10-5 2 a 10 -5 2 m 10 -5 2 a 10 -5 Te 2 a 10 -5 0.44 0.46 0.47 0.48 0.49 0.50 0.52 A9 33.6 32.2 30.7 29.5 28.4 26.9 25.4 Its 5.51 5.37 5.21 5.08 4.97 4.79 4.62 Ca0 0.77 0.75 0.74 0.73 0.72 0.71 0.70 A1 0 7 m 10 8 =10 9 a10 23 0.01 0.01 0.01 0.01 may0 4.55 4.69 4.78 4.82 4.83 4.82 4.77 m K0 43.3 46.6 49.8 2 52.2 54.3 57.1 59.6 560' 8.47 6.84 5.48 4.49 8 3 71 2.86 2.19 18 1.21 1.11 1.02 2 0.97 0.94 0.89 0.86 2r0 0.02 0.02 0.02 2 0.03 0.03 0.03 0.03 Csp -- -- -- -- -- -- -- Be0 0.04 0.04 0.04 0.04 0.04 0.04 0.04 Sr0 5 m 10~3 4 a 10'3 4 s 10'3 4 x 10'3 4 a 10'3 4 a 10'3 4 a 10'3 La 0 3 a 10 4 a 10 te02 23 5 a 10'" 5 a 10'" 4 x 10 4 s 10 4 s 10 5 a 10 5xM 6 a 10'" 6 s 10'4 6 x 10'" 7 x 10'4 7 a 10' ab 0 2$ 5 a 10 5 a 10 -6 5 a 10 6 a 10-6 6 x 10 -6 6 x 10 O 7 a 10-6 Cs1 -- -- -- -- -- -- -- C4 Sevrce Rate 12.6 10.6 9.1 (above 7.9 7.5 7.0 6.2 poc!) g/sec
TABLE 6.16. AEROSOL COMPOSITION AND TOTAL RELEASE RATE FOR SURRY V, CORE-CONCRETE INTERACTION Species. Time, sec
% 0 1200 2400 3600 4800 6000 7200 8400 % 00 10800 12000 13200 Fe0 -- 14.2 12.7 11.0 9.58 13.6 15.3 18.4 23.7 1.11 1.02 0.86 Cr 0 - - - - - - -- - - 4 x 10 1 x 10 4 2 x 10 23 Ni 2.57 0.24 0.51 1.27 2.78 0.40 0.16 0.12 0.11 0.28 0.24 0.18
-8 -8 -5 Mo 3 x 10 -6 5 x 10 -8 2 x 10
-7 9 x 10'I 4 x 10 -6 1 x 10"I 2 x 10-8 1 x 10 1 x 10 3 x 10'$ 3 x 10-5 2 x 10
-6 -6 Ru 2 x 10
-5 4 x 10'I 2 x 10
-6 8 x 10 3 x 10 -5 1 x 10 2 x 10'I 1 x 10'I 1 x 10"I 2 x 10'I 1 x 10"I ' 9 x 10-8
- Sn 0.08 0.02 0.03 0.04 0.07- 0.02 0.01 0.01 0.01 0.08 0.08 0.07 i Sb 1 x 10-5 3 x 10-6 4 x 10 -6 6 x 10-6 8 x 10 -6 3 x 10
-6 2 x 10-6 2 x 10-6 2 x 10-6 8 x 10-6 7 x 10-6 7 x 10-6 Te 0.70 0.a2 0.36 0.43 0.48 0.32 0.25 0.25 0.28 0.88 0.89 0.91 Ag 25.5 12.1 19.3 21.2 18.4 16.2 8.% 7.89 8.04 22.6 21.0 18.8
- 7.55 1.84 2.66 4.14 5.95 2.19 1.28 1.16 1.21 3.49 3.31 3.03 Ca0 -- 14.4 13.0 11.6 10.6 13.9 15.5 13.4 4.53 0.59 0.48 0.42 m A1 23 0 --
6 x 10'4 8 x 10'4 2.53 4.97 1 x 10 -3 1 x 10~3 2 x 10~3 2 x 10'3 8 x 10 ~3 9 x 10~3 0.01
-* Ma - 8.40 7.60 6.65 5.81 8.10 9.01 10.8 13.5 5.55 5.70 5.77 2
- K02 --
14.7 13.4 11.6 10.0 14.3 15.9 18.6 22.1 56.7 60.2 64.9 510 -- 30.6 27.3 23.7 20.7 29.2 32.3 28.8 26.3 7.94 6.36 4.48 2 UO 3.43 0.20 0.35 0.89 2.06 0.22 0.08 0.06 0.06 0.68 0.58 0.47 2 0.04 2 x 10'3 7'n-10'3 0.01 0.03 6 x 10'3 7 x 10~3 8 x 10~3 9 x 10'3 0.03 0.03 0.03 2 Cs 0 -- -- -- -- -- -- -- -- -- -- 2 8a0 2.64 1.00 0.99 0.74 0.71 , 0.49 0.28 0.18 0.06 0.02 0.02 0.02 Sr0 7.48 1.59 1.42 1.64~ 1.81 0.81 0.37 . 0.23 0.07 3 x 10'3 2 x 10'3 2 x 10'3 La23 0 3.82 2 x 10 2 x 10 i.12 2.37 1 x 10 1 x 10 1 x 10 4 1 x 10 5 x 10 5 x 10 5 x 10 Ce3 6.% 0.33 0.59 1.44 3.10 0.30 0.07 O.03 8 x 10'3 6 x 10'4 7 x 10 7 x 10'4 2
! Mb 0 4 x 10'4 1 x 10 -5 3 x 10
-5 1 x 10'* 5 x 10'4 2 x 10
-5
' 2 x 10 -6 2 x 10
-6 -6 7 x 10-6 7 x 10 -6 8 x 10 -6 25 '
2 510 l Csl -- -- -- -- y -- -, -- -- -- -- -- l Cd 39,2 C /- - -- -- -- -- -- -- -- -- -- Scurce [ , 13.3
, Rate ,
156.0 81.4 20.7 24.6 147.0 157.0 ,5 , 100.0 66.0 49.9 14.4 16.0 g/sec
-/
# y, 5~
. ,r d i "U" s 2
3 4 t ~,' '
?. .
r-
9 - , ,
s, ii M ,n . -
?J4 ;. f ~/
f j' s ., V. 4 f4 , ' TABLE 6.16. (Continued) w z- ,
+ ;
, ' /?
Specie g/~' < Time, sec # y, n
% 14400 15600 16800 18000 19/00 20400 21600 22800 24000 25200 Fe3 0.79 0.72 0.66' O.66 0.56 0.'43 0.38 0.36 0.32 0.32 Cr 0 23 2 x 10'3 3 x 10'3 3 x 10'3 4 x 10 -3 4 x 10'3 5 x 10'3 6 x 10'3 6 x 10-3 7 x 10-3 7 x 10'3 Ni 0.17 0.15 0.14 0.14 0.12 0.08 0.08 0.01 0.07 0.07 Mo 3 x 10 -5 3 x 10 -5 3 x 10 -5 5 x 10 -5 5 x 10 -5 5 x 10-5 7 x 10 -5 1 x 10'4 1 x 10'4 2 x 10'4 Ru 7 x 10 -8 ,, ,, ,, ,, ,, ,, ,, ,, ,,
Sm 0.07 0.07 0.07 0.07 0.01 0.06 0.06 0.07 0.07 0.08 Sb 7 x 10-6 7 x 10-6 7 x 10-6 7 x 10-6 7 x 10 -6 7 x 10 -6 7 x 10-6 7 x 10 -6 7 x 10 -6 8 x 10 -6 Te 0.91 0.92 0.92 0.93 0.93 0.93 0.95 0. % 0.97 0.99 Ag 17.9 17.0 16.3 16.9 15.3 12.9 12.4 12.2 11.6 12.4
- 2.92 2.82 2.73 2.82 2.61 2.30 2.25 2.25 2.19 2.34 Ca0 0.48 0.48 0.47 0.51 0.50 0.42 0.42 0.44 0.44 0.50 A1 0 0.01 0.01 0.01 0.01 0.02 23 0.02 0.02 0.02 0.02 0.02 Ma 0 5.70 5.66 5.60 5.56 2 5.48 5.28 5.13 5.00 4.83 4.76 mK0 2 66.9 68.5 70.0 69.3 72.1 75.9 76.9 77.3 78.3 77.2 d $102 3.69 3.11 2.60 2.60 1. % 1.24 1.01 0.88 0.72 O 0.72 U0 0.45 0.43 0.41 2 0.44 0.40 0.35 0.36 0.37 0.38 0.44 Zr0 0.03 0.03 0.04 0.03 2 0.04 0.04 0.04 0.04 0.04 0.04 Cs 0 -- -- -- --
2 -- -- -- -- -- -- Ba0 0:02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.03 Sr0 2 x 10'3 2 x 10~3 2 x 10-3 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10~3 2 x 10'3 2 x 10~3 2 x 10'3 La 0 23 5 x 10 5 x 10 # 5 x 10 5 x 10 5 x 10 6 x 10 6 x 10 6 x 10 7 x 10 6 x 10 Ce0 7 x 10'4 7 x 10'4 8 x 10' 7 x 10'4 8 x 10'4 9 x 10'4 9 x 10'4 2 9 x 10'4 9 x 10'4 9 x 10'4 Nbg$0 8 x 10 -6 8 x 10-6 8 x 10-6 7 x 10 -6 8 x 10 -6 9 x 10 -6 9 x 10'0 9 x 10'0 1 x 10' 9 x 10-6 Cs1 -- -- -- -- -- -- -- -- -- -- Cd -- -- -- -- -- -- -- -- -- .- Source Rate 9.4 8.6 6.4 5.7 7.2 g/sec 5.8 4.1 3.6 3.2 3.3
- 7. RESULTS AND DISCUSSION 7.1 Introduction Results of calculations for the transport and deposition of radio-nuclides are presented and discussed in this section. The plants and sequences selected for consideration were discussed in Chapter 4, the analytical and calculational methods were described in Chapter 5, and the assumptions and bases for the calculations were described in Chapter 6. Results presented in this chapter include the deposition and release from the reactor coolant system of radionuclides leaving the core region. These results are based on TRAP-MELT code calculations. Also included as results are the masses of radionu-clYdesairborneanddepositedinthecontainmentaswellastheairborne materials leaked to the environment, based on NAUA-4 calculations.
Four basic system sequences were considered in the analyses: AB, TMLB', S D, and V. Combined with these sequences are several variations in 2 the analyses, allowing for different assumptions of containment failure, compartmentalization of the containment, compartmentalization of the ECC injec-tion piping for the V sequence, and heating of primary system structures by decay of deposited radionuclides. 7.2- Transport and Deposition in Primary System 7.2.1 RCS Transport and Deposition for Sequence AB The AB sequences are characterized as low pressure, generally high flow rate scenarios with respect to RCS behavior. These sequences, considered here as hot leg breaks, offer the minimum volume of the RCS to the fission products being transported to the containment atmosphere. This, of course, -usually minimizes the opportunity for retention in the primary system. A variety of containment failure modes have been examined for this sequence in the Surry plant and are reported in this volume. The TRAP-MELT predictions of fission product retention for the AB sequence are found in Tables 7.1 and 7.2. The extent of retention is seen to be different for the various species, and the cause of the differences is 7-1
TABLE 7.1. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE AB SEQUENCE FOR THE SURRY PLANT ' (Tines Measured from Start of Core Melting) Csl Cs0H Te Aerosol Time Ret Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) -(kg) (kg) (kg) (kg) 300 .2. 9 12.4 15.9 67.7 0.3 0.4 31 9 663 600 5.3- 18.9 28.3 100 0.8 1.0 523 974 ,_ 900 6.9 22.7 37.0 120 1.5 .1. 7 682 1235 1200 0.5 24.6 3.9 130 2.4 2.9 71 3 1810 1510 0.5 25.3 3.9 133 2.4 3.6 714 1875 2110 0.5 25.6 3.9 135 2.4 4.4 714 1900 2710 0.5 25.8 3.9 135 2.4 4.9 721 1920 3310 0.6 25.9 4.6 136 2.9 5.7 785 1990 3910 0.7 25.9 4.9 136 4.1 6.9 911 2120 4520 0.7 26.0 5.0 136 5.4 8.2 1030 2235 5120 0.7 26.0 5.1 136 6.5 9.3 1135 2340 7-2
- n. .__ . .
TABLE 7.2. TRAP-MELT PREDICTIONS OF PRIMARY _ SYSTEM RETENTION FACTORS (RF) AND VOLUME SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME FOR THE AB SEQUENCE FOR THE SURRY PLANT Csl Cs0H Te Aerosol Time Core Core Core Core (s) RF Plate RF Plate RF Core Plate RF Core Plate 3 00 .23 .22 .23 .23 .82 0 .78 .48 .32 .16 600 .28 .27 .28 .27 .86 0 .83 .54 .36 .17 900 .30 .29 .31 .30 .93 0 .91 .55 .37 .18 5 1200 .02 --
.03 .01 .85 --
.77 .39 .25 .13 1510 .02 --
.03 .01 .67 --
.61 .38 .25 .13 2110 .02 --
.03 .01 .55 --
.51 .38 .24 .12 2710 .02 --
.03 .01 .49 --
.45 .38 .24 .12 3310 .02 --
.03 .01 .51 .09 .39 .39 .27 .12 3910 .03 --
.04 .01 .60 .25 .32 .43 .31 .11 4520 .03 --
.04 .01 .66 .36 .27 .46 .35 .11 5120 .03 --
.04 .01 .70 .44 .24 .48 .38 .10 i
4 apparently.in the RCS gas flow rates as a function of time. Examination of
'the gas flow rates leaving the core shows that for this sequence there is low-flow for approximately the first 1000 s after the start of core melting. This low flow rate of gas through the upper plenum permits very efficient retention of Te vapor due..to its rapid reaction with the structural surfaces. This 4
' behavior'is illustrated in Figure 7.1. At about t = 1000 s, the RCS gas flow rates' increase dramatically as portions of the core begin to slump into the
-water remaining in the lower plenum. Thus, while Te continues to be emitted by the melting fuel, it moves so rapidly through the upper plenum control volumes that practically no additional Te vapor is retained. After about
- 2500 s, the gas flow rate decreases to very low levels again because the water has been completely exhausted, and the additional release of Te is effectively scavenged.
The. influence of the RCS thermal-hydraulics is also clear in 1 ' Figures 7.2 and 7.3. For these species, the surface reaction is not nearly so rapid, and the mechanism of their retention is condensation on the particles E
-followed by particle' settling in the lower portion of the upper plenum. This f is seen to occur until the gas flow rate increases greatly, at which time the temperature of the upper grid plate increases from 950 F to 1340 F, resulting
.in re-evaporation of the Cs1 and Cs0H into the swiftly flowing gas stream.
The temperatures of the gas and the upper plenum surfaces quickly decrease to values which would permit condensation, but because the inventory of Cs and I l has been nearly expended by this time, there is almost no retention of these 1. vapors in the RCS for.this sequence. The aerosol retention factor for the AB sequence is seen in Table 7.2 to be approximately 50 percent. Figure 7.4 illustrates that about one-half of ( the ultimately retained aerosol mass accumulates in the core and upper plenum during the first 1000 s, characterized by low flow and high aerosol generation. [ As one would expect, the high flow rate during the in-vessel phase of this
- accident sequence permits little added retention because of the low aerosol concentrations and residence times. After vessel dryout, nearly all the emitted aerosol is retained by means of gravitational settling.
The TRAP-MELT' code was executed in a mode which permitted tracking j' of the WASH-1400 radionuclide groups (as described in Chapter 6) for this sequence, and the results are sumarized in Table 7.2a. The correspondence 7-4 L
10 Legend Vol 1 / [ g_ Vol 1 + Vol 2, Structure 1 g Vol 1,,+ Vol 2. Structures 1 and 2 g Emitted _ , en
/
/ 6 (n 6-5 /
/g
'I 4- -
/
3 / y '
- ... .../
2- [ 0
/ 1000 i
2000 i i 3000 4000 i i 5000 6000 0 TIME (sec) FIGURE 7.1. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE AB SEQUENCE (Vol 1 = Core; Vol 2, Structure 1 = 'Jpper Grid Plate; Vol 3, Structure 2 = Guide Tubes). Times Measured from Start of Core Melting.
1 15 0 = l i 4 A CD
.M
*W10 0 - /
/
i M i i
< /
1 , I ! O ! y m Ld Z [ f i < 50- / I C E / Legend
/
i Vol 2. Structure 1
/ Emi tted l
( 0 i , i i , 0 1000 2000 3000 4000 5000 6000 TIME (sec) FIGURE 7.2. MASSES OF Cs0H EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE AB SEQUENCE (Vol 2. Structure 1 = Upper Grid Plate). Times Measured
; from Start of Core Melting.
l 1 30 r m G
.M i 20- /
1 m ( i M 4 2
?
O La.1 b l [ 1 10 - / W< l $ Legend Vol 2. Structure 1 ) / Emitted l 0 i , , i i j 0 1000 2000 3000 4000 5000 6000 j TIME (sec) l FIGURE 7.3. MASSES OF Csl EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE AB SEQUENCE (Vol 2. Structure 1 = Upper Grid Plate). Times Measured l from Start of Core Melting.
2500 Legend Vol 1 y- ! Vol 1 + Vol 2. Structure 1 ,/,- ! 2000- Emitted _ _ ,y'
- m i
6 l, 2 ---- m m 1500-4 j 2 / O ! W 1000-
/ /
/
? Z ,/
/
l m q ' i
/
F / i W T
- tr -
500- ,
/
1
.'/
0 , , , , , O 1000 2000 3000 4000 5000 6000
- (
TIME (sec) I FIGURE 7.4. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS l OF TIME FOR THE AB SEQUENCE (Vol 1 = Core; Vol 2 Structure 1 = Upper Grid Plate). Times Measured from Start of Core Melting. 1
r-TABLE 7.2a. CORSOR PREDICTIONS OF RELEASE FROM CORE AND TRAP-MELT PREDICTIONS OF PRIMARY SYSTEM RETENTION OF WASH-1400 GROUPS FOR AB-y SEQUENCE Released Retained Group (kg) (kg)
-I 12.4 0.4 Cs 146 5.4 Te 9.3 6.5 Sr 71.1 51.1 Ru 34.7 24.4 La 0.8 0.7
- Groups are defined in Chapter 6.
1 I 7-9
f between the results cited for CsI, Cs0H, and Te and Groups 2, 3, and 4 is to be expected since'this merely represents a regrouping of the species already ~ considered. The results obtained for Groups 5, 6, and 7, however, are interesting since they display different overall RCS retention than does the total aerosol. While approximately 50 percent of the total aerosol mass released from(the core is retained in the RCS, about 70 percent of Groups 5
.and 6 and 90 percent of the Group 7 material is predicted to be retained. This
.is the case _not because of any difference in physical or chemical properties of these groups, as they all behave identically in the code. Rather, the differing extents of retention are due to the relative timing of the releases of the various groups. Thus, the Group 7 species,'which are emitted late in the in-vessel portion of the accident, are injected into the RCS during a period in which conditions, especially gas flow rates, are much more conducive to aerosol retention than is the case ectly in the sequence.
There are major. differences in the input to the TRAP-MELT code for [ _this. analysis from that used in the initial analysis of Surry (presented in Volume 1.of this report). In this analysis, the upper plenum has been subdivided into four control volumes, and the geometries and masses of the structures in the' upper plenum have been described more accurately. Another
- difference is the inclusion of control rod material in the aerosol inventory, leading to higher masses of aerosol being released from the core. The most significant difference in these results, however, is the lack of retention of fission products; this is a direct consequence both of the gas flow rates pre-t dicted by the MARCH
- code and e,f the timing of fission product release relative to the. gas flow rates. The use of a different description of the core mciting and slumping behavior appears to be a dominant influence on the predicted l . retention.
l 7.2.2 RCS Trans) ort and Deposition g for Sequence TML P This sequence involves high primary system pressure with generally low flow rate, yielding long residence times of the emitted materials in the RCS. These long residence times lead to quite high retention factors for all the species considered, as seen in Tables 7.3 and 7.4. There is a significant difference in the flow rates through the RCS used in this analysis and those l 7-10
+ "
'si
- TABLE 7.3. .CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS'(RET) DURING THE TMLB SEQUENCE FOR THE SURRY PLANT
-(Times Measured from Start of Core Melting) .
Csl Cs0H~ Te Aerosol Time Ret Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg) 106 0.2 2.9 1.8 22 1- -- 0.1 218 265 306 0.9 8.2 6.1 52.2 -- 0.2 436 470 611 1.7 14.0 11.9 85.5 -- 699 719 1020 2.5 18.8 17.6 112.5 0.1 1.3 941 955
- 1426 3.7 21.7 23.8 129 0.1 2.3 1129 1139 1629 5.4 23.0 33.9 137 1.3 3.5 1241 1291 1934 21.1 24.7 125 147 6.5 6.9 1558 1688 2340 21.8 25.6 128 151 7.6 8.9 1612 1733 4
k 7-11 ,
+
TABLE'7.4. TRAP-MELT PREDICTIONS OF. PRIMARY SYSTEM RETENTION FACTORS (RF) 'AND VOLUME SPECIFIC RETENTION FACTORS AS FUNCTIONS .0F TIME FOR THE TMLB SEQUENCE ~ FOR
.THE SURRY PLANT i
Cs1 Cs0H Te Aerosof Time Core Pres- Core Pres . Core Core (s) Plate Piping' surizer RF RF Plate Piping surizer RF fore Plate 'RF Core Plate 106 .07 .02 -- --
.02 .02 -- --
.04 0 .04 .82 .81' .01 i 306 .11 .04 -- --
.12 .05 -- --
.05 0 .05 .93 .91 .01
- 611 .12 .04 -- --
.14 .05 -- --
.05 0 .05 .98 .% .01 i
! 1020 .13 .06 -- --
.16 .07 -- --
.04 0 .04 .99 .98 .01 1426 .17 .06 -- --
.18 .07 -- --
.04 0 .04 .99 .98 .01 1630 .24 .23 -- --
.23 .23 -- --
.37 0 .37 .% .95 .01 1934 .85 .43 .18 .16 .86 .45 .18 .16 .93 .92 .77 .04 1 2340 .85 .43 .18 .16 .86 .45 .18 .16 .85 .17 .58 - .93 .78 .04
_m
used in the initial analysis of Surry (presented in Volume 1 of this report). In the previous analysis, the flow rate was so low as to prevent all but a few percent of the emitted materials from entering the containment until the time of bottom head failure, with nearly 75 percent of even the early emitted materials still residing in the core region, where gas and surface temperatures prohibit any condensation, at the tin,e of vessel failure. This fact may well have led to large underestimates of the extent of the vapor species' retention. 1 In the present analysis, the low flow rates retain the emitted materials in the core region until just after t = 1500 s. Figure 7.5 shows that no significant retention of Te occurs until after this time. This is because the TRAP-MELT code, as used here, does not permit the surface reaction of Te to occur in the core region. As the gas flow rate from the core increases at 1500 s, the suspended vapors and particles are swept into the upper plenum control volumes, where the Te reaction leads to nearly complete scavenging by the structure surfaces. This period of high flow rates due to core slumping also leads to increased Zr cladding oxidation, which in turn increases the Te release rate from the melting fuel. As the gas flow rate continues to increase, eventually the material is transported through the RC3 too quickly for the surf ace reaction to be completely effective, and the retention factor decrease ~s, although it remains at a fairly high value. The Cs0H and Cs! behavior indicated in Figures 7.6 and 7.7 is similar to that of Te. The cause of the small early retention of the Cs species is condensation of the vapors on particles which subsequently settle out in the upper plenum volumes. There is no retention of these species in the core volume because the high gas temperatures there prevent condensation of the vapors on the particles. As soon as the aerosol and vapors are transported into the relatively cool upper plenum, condensation and retention of these species occurs at a rapid rate. The process does not, however, go "to comple-tion", being curtailed by the large velocities which are achieved and by the low emission rate of Cs and I after about 1400 s. The aerosol retention factor reaches 90 percent for this transient sequence, which is similar to values obtained for such sequences in other plants. The explanation for this is found in the long RCS residence time the low flow rate affords the aerosol. This results in the particle mass median diameter being greater than 3 um when the aerosol reaches the upper plenum, where there 7-13
4 I 10 l W Vol 1 Vol 1 + Vol 2, Structure 1 /
/s' 3- 7 i
Q 3C Vol1_+lol2. Structures 1and2 _ / ~~~ Vol 1 + VolStructures Voll.I1ol2. 2, Structures 1, 2 and 3 + Vol 3, Structure 1,2and3+Vol3 C [I i M Emi tted Structures 1 and 2
/ #
y) 6-i 4 i 2 / l g / i.a /
- Z 4-
/
1
- Y Q
W /
/
i % LaJ / i & / i 2-
/, -
'/
'/
O , , , , 0 500 1000 1500 2000 2500 ! TIME (sec) l 1 FIGURE 7.5. MASSES OF Te EMITTED FROM CORE AND RETAINED IN lHE RCS CONTROL VOLUMES AS FUNCTIONS 1 0F TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core; Vcl 2 Structure 1 = Upper Grid Plate; j Vol 2 Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3,
! Structure 1 = Pressurizer; Vol 3, Structure 2 = Hot Leg). Times Measured from Start j of Core Melting.
l 4 l
4 150 a i Legend -
, Vol 1 '
i A vol 1 + Vol 2 Structure 1 /, ,___________ G Vol 1 + Vol 2 Structures 1 and 2 - i 3 Vol l+ Tol 2, Structures 1, 2 a_nd 3Iv/ ,
- ol 3. Structu.re 2
) 10 0 - YR1.1.+..yp1 2. Structures,1, 2 and 3 + Vol 3, M Emitted / Structures 1 and 2 , $ M / y / )
,/ 3-- --_
4 o w /
/
Z / ! ~ 50- / 1 L H< 7 \ i w w / CE / I / i
/
/
i - / - t l 0_ i i i i i 0 500 1000 1500 2000 2500 ) TIME (sec) FIGURE 7.6. MASSES OF CsCH EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core; Vol 2 Structure 1 = Upper Grid Plate; Vol 2 Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3, Structure 1 = Pressurizer; Vol 3, Structure 2 = Hot Leg). Times Measured from Start of Core Melting. i
- 30 Legend Vol 1 Vol 1 + Vol 2. Structure 1 -
~
Vol 1 + % 1 2. Structures 1 and 2 f', 3 ~ Vol l"+ Tol 2. Structures 1, 2 and 3 + Vol 3. Structure 2 ~~'~~~~~~~~~~~ d 20-Vol.,1,,f,lol2. Structures 1,2and3+fol3 Emitted / f / Structures 1 and 2 / _
/
ln 4 /
/
/
///
1 / a w /
,/
l l- - - Z / g y R H 10 -
/
/ f W /
W /
/
/
/
/
o , , , i o 500 1000 1500 2000 2500 TIME (sec) FIGURE 7.7. MASSES OF Csl EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core; Vol 2, Structure 1 = Upper Grid Plate; Vol 2 Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3, Structure 1 = Pressurizer; Vol 3, Structure 2 = Hot Leg). Times Measured from Start of Core Melting.
is a greater amount of horizontal surface area available for settling of the particles. This effect is clear in Figure 7.8. Significantly, the aerosol generation rate decreases greatly during the high flow rate oeriod. This results in a lower amount of aerosol mass being injected into the containment than would occur for the case where aerosol generation proceeded at a high rate along with the high gas flows. While the aerosol retention factors for this sequence as analyzed here and in Volume 1 are similar, the location of the retention is different because of the different flow rates used in the two TRAP-MELT analyses, and the total mass injected into the containment is nearly three times as great here (165 kg) because of the higher aerosol generation rates used in the current analyses. TRAP-MELT predictions for the RCS retention of the WASH-1400 groups were obtained for this sequence and are summarized in Table 7.4a. As for the AB sequence above, the extent of retention of Groups 5, 6, and 7 vary somewhat due to the differences in the timing of their releases, but the varia*, ion seen here is less than in the AB case and is closely related to the total aerosol rettation predicted by the code. 7.2.3 RCS Transport and Deposition for the Sequence V The V sequence, from the perspective of the primary system, is much the same as an intermediate size break in the cold leg of a steam generator. The difference, which is very significant from a risk standpoint, is that the primary system comunicates directly with the auxiliary building (s) through a long pipe. The gas flow through the primary circuit is fairly low during the first 1000 s after the start of core melting, and then reaches very high values until approximately 2400 s, at which time vessel dryout occurs. Bottom head failure is not predicted by MARCH to occur until about 4000 s after vessel dryout--resulting in a long period during which the melting core is emitting into a stagnant primary system. The total masses of the species of interest emitted from the core and retained throughout the primary system are given in Table 7.5, and the retention factors for the primary system and specific control volumes are given in Table 7.6. 7-17
2000 W voi i , Vol 1 + Vol 2. Structure 1 ' 9 1500- V 1 i +T 12, structures 1, 2 and 3;Vol 3, structure 2 - - ~ y Vol 1 + Vol 2. Structures 1, 2 and 3; Vol 3 Structures 7 1 and 2 V Emitted] ,/ ' , , ~ _ - - - D m *,, ', 4 I .. '~ 1000- - a La.s ,, . r z '.. 5 W A , E 500-
,o' 0
r l s / , , , , , 0 500 1000 1500 2000 2500 TIME (sec) FIGURE 7.8. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 = Guide Tubes, Vol 4 = Upper Plenum Annulus, Vol S = Hot Leg, Vol 6 = Pressurizer). Times measured from start of core melting.
TABLE 7.4a. CORSOR PREDICTIONS OF RELEASE FROM CORE AND TRAP-MELT PREDICTIONS OF PRIMARY SYSTEM RETENTION OF WASH-1400 GROUPS FOR TMLB' SEQUENCE Released Retained Group (kg) (kg) I 12.3 10.6 Cs 145 125 Te 8.8 ' 7.5 Sr 35.7 31.9 Ru 16.7 13.6' La 0.29 0.26
- Groups are defined in Chapter 6.
7-19
TABLE 7.5. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE V SEQUENCE FOR THE SURRY
- PLANT i
- s! Cs0H Te Aerosol Time Ret Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (k9) (kg) (kg) 300 5.4 12.0 30.1 65.4 0.3 0.4 340 687 6 00 10.3 18.2 54.8 97.0 0.9 0.9 588 999 900 13.6 21.7 71.9 115 1.5 1.6 805 ,1240 1200 14.0 23.5 73.6 124 2.3 2.3 941 1520 1500 13.0 24.4 69.4 128 2.6 3.0 948 1680 t 2100 13.0 25.3 69.4 133 2.6 3.7 948 1690 2690 13.0 25.5 69.4 134 2.6 4.0 955 1710 3290 13.0 25.6 69.5 135 2.6 4.8 1005 1760 4190 13.0 25.8 69.6 136 2.7 6.5 1150 1910 5090 13.0 25.9 69.6 136 2.9 8.5 1330 2090 6585 13.1 25.9 69.7 136 3.2 11.4 1565 2320 7-20 L __
g n i 1 2 1 1 9 9 9 8 7 6 4 p 2 2 2 2 1 1 1 1 1 1 1 i . . . . . P E M U LR OO tg 8 0 0 9 7 7 7 6 5 4 2 VF l o oe 1 2 2 1 1 1 1 1 1 1 1 HL . . DE s NC o AN r E e
)U A e FQ r 3 0 9 8 8 8 1 8 5 2 RE o 0 7 1 0 0 0 0 1 1 2 3 (S C SV R
OE TH 0 9 5 2 6 6 6 7 0 4 7 CT F 5 5 6 6 5 5 5 5 6 5 6 A R FR O NF O IE e TM et 3 6 9 0 8 4 8 9 6 9 3 NI ra 8 8 8 9 7 6 5 4 3 2 2 ET ol . . T CP EF e RO T MS EN 0 1 3 8 6 1 4 5 2 4 8 TO F 9 9 9 9 8 7 6 5 4 3 2 SI R YT SC N YU RF g A n MS i p 1 3 4 5 5 4 4 3 3 3 3 IA 2 2 2 2 2 2 2 2 2 2 2 R i PS P R FO OT H C 0 SA s tg 8 1 3 4 3 2 2 2 2 2 2 NF C oe 1 2 2 2 2 2 2 2 2 2 2 O HL IN TO CI ITT DNN 6 7 3 0 4 2 2 2 1 5 1 5 1 5 EEA F 4 5 6 6 5 5 5 '5 RTL R PEP R T Y LCR g EIR MFU n 3 3 IS i 0 3 4 6 5 4 4 4 3 PC p 2 2 2 2 2 2 2 2 2 2 2 AEE i RPH P TST
. s tg 8 2 4 4 3 2 2 2 2 2 2 6 C oe 1 2 2 2 2 2 2 2 2 2 2 HL 7
E L 0 0 0 B 5 6 3 0 3 1 1 1 5 A F 4 5 6 6 5 5 5 5 5 5 T R e 0 0 0 0 0 0 0 0 0 0 5 m) 0 0 0 0 0 0 9 9 9 9 8 i s 3 6 9 2 5 1 6 2 1 0 5 2 2 3 4 5 6 T( 1 1 d i5
i The behavior of the Te illustrated in Figure 7.9 can be explained entirely in terms of the gas flow through the RCS. During the initial period of low but non-zero flows, the Te is scavenged by the upper plenum surfaces very efficiently due'to the long residence time in that region and Te's high rate of reaction. Since this reacted material is not liberated, except at very high temperatures, it is retained throughout the sequence. The lack of additional retention of the Te emitted beyond t = 1200 s is caused by two factors. Until 2400 s, the flow rate through the system is too high for the reaction with the surfaces to occur at a significant rate. Beyond this time (i.e., af'ter vessel dryout), the lack of flow out of the core region results in the vapor remaining suspended in the core region until the bottom head fails, at which time it is injected into the containment atmosphere. The primary system retention of Cs0H and Cs! is depicted in Figures 7.10 and 7.11. In this sequence, the materials which are retained in the first 1000 s (principally via condensation on particles which are subse-quently deposited), are partially liberated by the surface heatup that occurs at about 1200 s, when the gas flow increases dramatically. After this time, the surfaces cool sufficiently to permit condensation, but the inventory of Cs and I has been nearly exhausted so that no further retention is possible. Approximately two-thirds of the aerosol mass emitted from the core in this sequence is retained within the RCS. It is clear in Figure 7.12 that from t = 1200 s to t = 2400 s, the high flow rates (or very short RCS residence times) lead to no aerosol retention on the surfaces. After vessel dryout, however, almost all further emission of aerosol is retained. As a result, there is but little aerosol injected into the containment at the time of vessel failure. Figure 7.12 also illustrates that the ECC piping--Volume 5 in the figure--is an important, though not dominant, contributor to aerosol retention in this sequence. The total mass retained in this piping is predicted by TRAP-MELT to be 323 kg at the time of bottom head failure. This is equivalent to a deposition of just over 50 g of aerosol per cm of pipe, leading to questions concerning whether this is physically possible. As the aerosol deposit in the pipe increases, one would expect the constriction in the flow path to lead to higher gas flow velocities. These higher velocities could, in turn, result in resuspension of deposited aerosol, reducing retention. A plug of deposited 7-22
S N 0 O 0 I ; T e 0 Ct N ag 7 Ul n FPi
' t
=- _ Sdl Aie p 0 SG rM E e i 0 Mrr 0 U eo
/ 6 L pC Op VUf
/
o L = _ O t _ R1 r T a
/ 0 N et 0 ors i Cu
' 0 S co tm 5 C ur R rf
/ ..
. ESd t
_ H e T .r
/ 0 N 2u s
/
,0 0
)c Il a oe DVM 4 e E
' (s N ;s I ee Arm 2
__ Toi d n a
' i 0
0 E M I ECT R D N1)
=
1
- 0T A l e s
1 s 3 Eob e r r e _ RVu u u - __ O(T C - t t _ Ee c c MCd u u 0 ONi r r - RE u t t i 0 FUG Q S S 0 DE = 2 2 - 2 ES T TV 2 l l ___ I MEr e V / EH u Tt d oMode
+, 0 e TR u c
n1 1 1 t i0 Or t 0 FFt el go l l i O S o o Em 1 E
/
eVV V SM , L EI2 - ST S l AF o - MOV 0 . 9 5 0 5 0 7 1 1 E R U G nG V mD<1 OwzR>wm I F
~6
4 15 0 s 7 Legend n Vol 2 Structure 1 CD / - Vol 2. StructureT1 and 2 d 10 0 - / Vol 2. Structures 1, 2 and 3 + Vol 3
~
voi 2, Structures 1, 2 and 3 + Voi=3 + Voi 4 m f Vol 2, Structures 1, 2 and 3 + Vol 3 + Vol 4 + Vol 5 4 , Emitted 2 / a ,r~ ~N. I w i
~ Z /
h
- R 50- I i F-w ct I/ P%% ^
'!/
y % ,_ _ _ _ _ O- --- --- --- --- O 1000 2000 3000 4000 5000 6000 7000 TIME (sec) FIGURE 7.10. MASSES OF Cs0H EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE V SEQUENCE (Vol 2, Structure 1 = Upper Grid Plate; Vol 2 Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Steam Generator; Vol 5 = Piping). Times Measured from Start of Core Melting.
30 Legend Vol 2. Structure 1
~-
Vol 2.,' Structures 1 and 2, __________ Vol 2. Structures 1. 2 and 3 + Vol 3 i n Vol 2. Structures 1. 2 and 3 + Vol 3 + Vol 4 CD Vol 2. Structures 1. 2 and 3 + Vol 3 + Vol 4 + Vol 5 d 20-Emi tted I (A M
< /
i 2 / o y / ,-- - -
. . . . . . = .
j s* z /
\ R to - I i v \
y I
/
/ /*%t --- -= ---
I .l'/ l __ l
\ !!/ .
1 i O (/~ i i i i i i l 0 1000 2000 3000 4000 5000 6000 700c
) TIME (sec) i ; FIGURE 7.11. MASSES OF Cs! EMITTED FFOM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS l OF TIME r0R THE V SEQUENCE (Vol 2. Structure 1 = Upper Grid Plate; Vol 2 I Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Stean Garerator; Vol 5 = Piping). Times Measured fro- Start of Core Melting.
t 4
2500 p L*9end Vol 1 ' Vol 1 +Tol 2. Structure 1 ' Vi 12. Structures 1. 2 and 3 + Vol 3 ' A 2000- - Vol 1 + Vol 2. Structures 1. 2 and 3 + Vol 3 + Vol 4 Vol 1 + Vol 2. Structures 1. 2 and ' 3 + Yof 3 + Vol 4 + Vol 5 V - Emitted _ _ ~ 1500- '
/ /*
2 / f f .f - CY. s -
/
500- l ,/ 7',/
// ___ _ _ _ -
/
/
0 , , , , , i 0 1000 2000 3000 4000 5000 6000 7000 TIME (sec) FIGUEE 7.12. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FU.NCT OF TIME FOR THE V SEQUENCE (Vol 1 = Core; Vol 2. Structure 1 = Upper Grid Plate; Vol 2. Structure 2 = Guide Tubes; Vol 2. Structura 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Stea-t Generator; Vol 5 = Piping). Turnes Measured from Start of Core Melting.
s material could form in the pipe, but the existence of system pressures in . - 4 excess of 100 psi make this open to question also. , s 7.2.4 RCS Transport and Deposition ' l ~for the 5 2D-c (Cold Leg) Sequence . This sequence exhibits low flow and intennediate pressure in the RCS. The principal contributors to fission product retention in the primary system are the core region, the upper plenum annulus, and the steam generators. Tables'7.7 and 7.8 illustrate the release o'f the species of, interest N fromthecoreand'theretentionanddistributionofretai[1edmassesthroughout the RCS during the in-vessel phase of the sequence. The high overall retention factors for each species are characteristic <of a low ficw, high residence time sequence such abthis one.
- The ratention of Te in this sequence is near.ly complete; the largest portionisretainedonthefirstpost-coresurface,whichistheupperjrid plate. This retention is duelexclusively to the chemical reaction'of tellurium with system surfaces. Thehighretentiondecreasesslightlystartingat\4100 seconds because of vessel dryout and the coupled loss of flow of em nted vapors from the core.' Figure 7.13 illustrates the behavior of tellurium in the RCS, Csl and Cs0H are retained throughout the system in a common pattern, as shown in Figures 7.14.,and 7.15. At 840 seconds, each displays a drop in retention in the core plate and guide tubes. The core has slumped at this time, and the resultant increase in surface temperature to 4.900 F in these volumes causes re-evaporation of the two'Cs species and their transport to s downstream volumes. Emission of Csl and Cs0H from the cere is essentially complete at 1200 s, and the retention of the two species undergoes little change after that time. The steam generators are seen to be tha site of the bulk of - - 8 the retention of the Cs species.
z Aerosol retention, as illustrated in,the tables and Figure 7.16, . shows a less pronounced response to the core slump.g Petention continues to increase, but at a'slightly reduced rate. The increased thw rates acco'mpanying steam generated when,the core slumps serve to reduce agglomeration by . educing aerosol concent' rations and residence times. This is evident in the aerosol massmediandiameterscalcul$tedbytheTkAP-MELTcode. In the core region, the particle diameter fall. from 3.6 un at 560 s to 1.0 un at 840 s. 7-27 - 3 4
*h, b '
s* _e- -
-w -J- +, s.,,__9
U' . l", l Y.')j~ up I 3
~a. :
ya o
~
s>, '
?/l- , jy ;
J /r TABLE 7.7. TRAP-MELT PREDICTIONS OF PRIMRY SYSTEM RETENTION FACTORS (RF) AND VOLUME. SEECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME FOR THE 2S 0-c (COLD LEG) j, 's SEQUENCp FOR THE_SURRY PLANT Cs! Cs0H Te Aerosol Upper Upper Upper Upper Upper Time Plenum Steam Plenum Steam Grid Grid Plenum Steam
'(s) RF Annulus Gen RF Annulus Gen RF Plate RF Core Plate Annulus Gen 280 .48 .17 . 06 .54 .17 .07- .74 .72 . 71 .24 .18 .16 .07 560 .73 .22 .11 .75 .21 .11 .80 .78 .82 .30 .19 .18 .08 840 ,81 .21 .35 .83 .20 .30 .99 .93 .85 .24 .16 .19 .19 1400 .77 .19 .34 .79 .19 .29 .97 .91 .78 .22 .14 .16 .20 1960 .76 .19 .33 .78 .18 .29 .97 .90 .77 .22 .14 .16 .20 2800 .75 .19 .33 .77 .18 .28 .96 .90 .76 .21 .13 .16 .19 3630 .74 .18 .33 .77 .18 .28 .96 .89 .73 ,20 .13 .15 .19 5591 .74 .18 .32 .76 .18 .28 .90 . 84 .73 .29 .11 .13 .16
TABLE 7.8. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE S D-c 2 (COLD LEG) SEQUENCE FOR THE SURRY PLANT Csl Cs0H Te Aerosol Time Ret . Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg) 280 3.9 8.0 24.3 45.3 1.0 1.4 351 492 560 12.5 17.2 68.6 91.8 5.0 6.2 771 935 840 18.2 22.3 97.9 117 12.9 13.1 1150 1360 1400 18.6 24.1 100 126 14.4 14.8 1240 1590 1960 18.6 24.5 100 128 14.8 15.3 1240 1600 2800 18.7 24.9 101 1 31 15.8 16.5 1250 1650 3630 18.8 25.3 101 132 16.9 17.7 1280 1760 5591 18.9 25.6 102 134 17.4 19.3 1570 2140 4 7-29
l i i 20 Legend '- Vol 2. Structure 1 -
~
Vol 2. Structure 2 ,,- ' m Emitted ,, y tn 15 - - x v w Ir' w <
< 9 2 10 -
o 4 w
- z 5
& l w
g 5-I I
~
l l I I I O 1000 2000 3000 4000 5000 6000 TIME (sec) FIGURE 7.13. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2 S D-c (COLD LEG) SEQUENCE (Vol 2, Structure 1 = Upper Grid Plate; Vol 2. Structure 2 = Guide Tubes). Times Measured from Start of Core Melting.
15 0 Legend Vol 2, Structure 1 Vol 2 Structures 1 and 2 , ___ l Vol 2, Structures 1, 2 and 3 V5ET,"5Eructures 1, 2 and 3 + Vol 3
^ Vol 2, Structures 1, 2 and 3 + Vol 3 + Vol 4 CD y Emitted v ~ ~ ~ ~ ~ ~ ~ --
10 0 - / ,. -- -- - M / m
- <t I!
2 Q // w l: , -.
? Z f S /
. R p 50- , W / / Cr /
/ _ _ _ _ _ _-
l ,"
- / s__ ___ ___ ___ ___ ___
i f 0 , , , , , O 1000 2000 3000 4000 5000 6000 TIME (sec) FIGURE 7.14. MASSES OF Cs0H EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2 S D-c (COLD LEG) SEQUENCE (Vol 1 = Core; Vol 2, Structure 1 = Upper Grid Plate; Vol 2, Structure 2 = Guide Tubes; Vol 2. Structure 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Steam Generator). Times Measured from Start of Core Melting.
30
- Legend Vol 2, Structure 1 Vol 2',"SE uctures 1 and 2 I .V.0.I..2,3,,S,t,ructures 1, 2 and 3 _ _ -------
i Vol 2. Structures 1, 2 and 3 + Vol 3 O) Vol 2. Structures 1, 2 and 3 + Vol 3 + Vol 4 3 Emitted "M 20- f m . _ .-- ... ...- _. ---
.= ... =.- . ... ..
4 1
/
1: a La l
/
z l/ e, 10 - / - - s<
~
~
l , N,.............................................................
/
//
/ _ _ _ _ _ _
o ' i i , , , o 1000 2000 3000 4000 5000 sooo TIME (sec) FIGURE 7.15. MASSES OF Csl EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTION OF TIME FOR THE2 S D-c (COLD LEG) SEQUENCE (Vol 1 = Core; Vol 2, Structure 1 = Upper Grid Plate; Vol 2 Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Steam Generator). Times Measured from Start of Core Melting.
2500 Legend Vol 1 Vol 1 + Vol 2 Structure 1 p 2000-- Y9I Ii V I 2, Structures 1, 2 and 3 n Vof 1; Vol 2. Structures 1, 2 and 3 + Vol 3 - Vol 1; Vol 2. Structures 1, 2 and 3 + Vol 3,+ Vol 4 v - Emitte_d __ ~
~
(n 7 (n 1500- '*..
< /
-~~~"*'...
1 ..._...-_-... - -
,(.._..= /',.. . .
a tu jogo_ f/ - . . r w z l ,'f -,.... ........ ......-.....-- --..-- w 4 ', L'i /. - e ___ __- 500- , ,
/
//
'l 0- , , , , i 0 1000 2000 3000 4000 5000 6000 TIME (sec)
FIGURE 7.16. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE S2D-c (COLD LEG) SEQUENCE (Vol 1 = Core; Vol 2, Structure 1 = Upper Grid Plate; Vol 2, Structure 2 = Guide Tubes; Vol 2, Structure 3 = Core Barrel; Vol 3 = Hot Leg; Vol 4 = Steam Generator). Times Measured from Start of Core Melting.
Gravitational settling in the core region continues at a reduced rate, and transport to cooler volumes allows continued retention. After this time, retention shows little change as aerosol emission from the core slows consider-ably, until 4100 s. At this time flow has stopped, the freshly emitted aerosol staying in the core region, where gravitational settling effects an increase in overall retention. 7.2.5 RCS Transport and Deposition for the S D 2 (Hot Leg) Sequence This sequence is characterized by low flow and intermediate pressure in the RCS. Retention occurs primarily in the core region and the upper plenum annulus. Use of the hot leg as the site of the small pipe break removes the steam generators from analysis. The resulting decrease in overall retention, except for tellurium, is the most noticeable feature when this sequence is compared to the previous one. Tables 7.9 and 7.10, along with Figures 7.17 through 7.20, illustrate the release, retention, and distribution of species of interest.' Tellurium shows increased retention, when compared to the cold leg case, because the flow in this sequence does not subside after core slump. This results from the shorter duration of the in-vessel portion of the accident, 1750 s versus 5600 s. The Cs species again show the decrease in retention attributable to core slump at 720 s, with its resultant temperature increase from 1530 F to 1980 F. Csl has a negligible surface reaction rate compared to that of Cs0H, and it shows the more abrupt decrease in retention at this time. Retention for both species has recovered by 1100 s, at which time Cs emission from the core has slowed considerably. Retention is essentially unchanged from this time to the end of the sequence. The aerosol retention pattern is similar to that of the previous sequence with one exception, as illustrated in Figure 7.20 at 1100 s. While emission from the core continues, retention has stopped. This is due to the increased flow beginning at 1070 s, which results from the core slump and subse-quent steam generation and pressure increase. The mass median diameter drops from 3.27 um to 1.79 um in the core region, and from 5.1 um to 2.5 um in the 7-34
TABLE 7.9. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RC5 (RET) DURING THE S D 2 (HOT LEG) SEQUENCE FOR THE SURRY PLANT Csl Cs0H Te Aerosol Time Ret Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg) 1 03 1.0 3.8 7.5 24.4 0.1 0.3 224 336 206 3.1 7.6 19.9 43.5 0.8 1.2 34 8 494 309 5.8 11.5 34.1 63.2 1.7 2.3 481 652 412 8.5 14.9 47.8 80.4 3.0 4.2 629 81 5 514 11.2 17.9 61 .9 95.3 5.3 7.0 776 974 617 13.6 20.3 75.9 108 8.4 9.9 907 1130 720 11.7 22.0 76.9 116 11.3 12.1 1040 1290 1029. 12.6 24.5 78.1 129 17.0 18.1 1190 1610 1337 12.3 25.5 76.7 1 34 22.6 23.0 1240 1790 1749 12.3 25.7 -76.5 135 23.8 24.3 1240 1870 7-35
TABLE 7.10. TRAP-MELT-PREDICTIONS OF PRIMARY SYSTEM RETENTION FACTORS (RF) AND VOLUME SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME2 FOR THE S D (H0T LEG) SEQUENCE FOR THE SURRY PLANT Cs1 Cs0H Upper Te ~ Aerosol Upper Upper Upper Time Grid Upper Upper Upper Guide Plenum Hot Grid Guide Plenum Hot (s) RF Plate Tubes Annulus Grid Grid Plenum Leg RF Plate Tubes Annulus. Leg RF- Plate RF Core Plate Annulus 103 .25 .18 --
. 06 .02 .31 .21 --
.07 .02 .46 .45 .64 .40 .16 .06 206 .41 .22 --
.15 .05 .46 .25 .01 .;5 .05 .66 .64 .70 .37 .18 .11 309 .51 .25 --
.20 .06 .54 .27 .01 .20 .06 .74 .72 .74 .37 .19 .14 412 .57 .27 .01 .22 .d7 .59 .30 .01 .22 .07 .71 .70 .77 .39 .19 .14 g 514 .63 .30 .02 .23 .08 .65 .33 .02 .22 .08 .76 .74 .80 .40 .20 .15 617 .67 .32 .05 .22 .09 . 71 .36 .04 .22 .09 .85 .83 .80 .40 .20 .16 -
720 .52 .12 .09 .21 .11 .66 .28 .08 .21 .10 .93 .91 .81 .38 .20 .17 1029 .51 0 .15 .21 .15 .61 .14 .13 .21 .13 . 94 .91 .74 .32 .17 .18 1337 .48 --
.13 .20 .15 .57 .14 .12 .19 .12 .98 .93 .69 .30 .16 .17 1749 .48 --
.13 .20 .15 .57 .14 .12 .19 .12 .98 .92 .67 .28 .15 .16
- 30 1
1 Legend 1 3 Vol 2, Structure 1 ,[' b , Vol 2, Structures 1 and 2 M Emitted---_ , i i
; y 1
- v 8
z
/
/'
g e 10 - , i
- w
/
i J
- 0 i ,
0 500 1000 1500 2000 f TIME (sec) l FIGURE 7.17. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2 S 0-y (HOT LEG) SEQUENCE (Vol 2 Structure 1 = Upper Grid Plate; Vol 2, Structure 2 = Guide Tubes). Times Measured from Start of Core Melting.
15 0 Legend ~~,,_ Vol 2, Structure 1 l Vol 2, Structures 1 and 2 A Vol 2, Structures 1, 2 and 3 g) -_-. ,,
.x Vol 2, Structures 1, 2 and 3 + Vol 4
" ~~ '
! 10 0 - Emitted tn ' 4 ,' \ 1 ,
/ -y- - -
y , s' /m L
= R H
50- ,' s D x l' ,'
/'
/ N-l' y /
l 0 , , , 0 500 1000 1500 2000 TIME (sec) FIGURE 7.18. MASSES OF Cs0H EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2 S 0-y (HOT LEG) SEQUENCE (Vol 2, Structure 1 = Upper Grid Plate; Vol 2. Structure 2 = Grid Plate; Vol 2, Structure 3 = Guide Tubes; Vol 3 = Hot Leg). Times Measured from Start of Core Melting.
30 -- - Legend i V61 2, Structure 1 l Vol 2, Structures 1 and 2 ,, G Vol 2, Structures 1, 2 and',3
.X v V 1 2, Structures 1,'2 and 3 + Vol 4 j 20-Eg!tted,____,
4 a' 1 ,'
)
g j , _/
\
a 10 - ,' t; ,'
,- ,/ '
, j j
/
1 ,',
? \ 8 i 0
t,' N , , 1 0 500 1000 1500 2000 TIME (sec) FIGURE 7.19. MASSES OF Cs! EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2 S 0-Y (H0T LEG) SEQUENCE (Vol 2, Structure 1 = Upper Grid Plate; Vol 2. Structure 2 = Grid Plate; Vol 2. Structure 3 = Guide Tubes; Vol 3 = Hot Leg). Times Measured from Start of Core Melting.
2000 _ Legend Vol 1 ~~~~~___...... --- - Vol 1 + Vol 2. Structure 1 ,' n CD 1500- vol 1 + Vol 2. Stiuctures 1, 2 and 3 X Vol 1 + Vol 2 Structures 1, '2 and 3 + Vol 3 v M Emitted ,' w ............. f,
<t '
2 1000- ,' y a / / o Ld ' e /, ---- Z I' // / is LJ
" 500-
,I /
//
l 1 1 0 i i i 0 500 1000 1500 2000 TIME (sec) FIGURE 7.20. MASSES OF AEROSOL EMITTED FROM CORE AND r-RETAINED IN THE RCS CONTROL VOLUMES AS FUNCT OF TIME FOR THE S2 D-y (HOT LEG) SEQUENCE'(Vol 1 Core; Vol 2. Structure 1 = Upper Grid Plate; Vol 2, Structure 2 = Grid Plate; Vol 2 Structure 3 = Guide Tubes; Vol 3 = Hot Leg). Times Measured from Start of Core Melting.
hot leg, from 1030 s to 1130 s. The shorter residence times and reduced agglomeration halt the gravitational settling in all volumes. 7.3 Transport, Deposition and Leakage in Containment Calculated results are presented in this section for analyses performed to examine the transport and retention of various fission products in the containment. The NAUA code that was described previously was utilized in the analyses. In general, the NAVA code used here needs information on the thermal-hydraulic conditions of an accident of interest. The conditions provided by the MARCH computer calculation were used. The typical required thermal hydraulic conditions are time-dependent containment temperature, pressure, and wall temperature, and the rates at which steam enters the containment, condenses on the containment structure, and leaks from the containment. Perhaps the most important and critical input that the NAUA code also needs is the fission product source term for particulates. The source rates calculated as release from the primary system (TRAP-MELT code) and the VANESA calculations for release during the core-concrete interaction were taken for the melt and vaporization releases, respectively. For the NAUA calculations, CsI, Cs0H, and Te were considered separately. All these species were assumed to be in the particulate form in the containment atmosphere because the tempera-ture and pressure under the containment conditions indicate that these species will remain as particulates for all practical circumstances. Although it is assumed in the calculation that individual species are distributed evenly over all sizes of particulates, the different amounts of these species at a given time due to different source timings were taken into consideration in the calculations. The aerosol behavior mechanisms considered in the present calculations include agglomeration due to Brownian diffusion and due to sequential gravita-
'l' tional settling, particle size change due to steam condensation / evaporation, sedimentation, Brownian diffusion, and diffusiophoresis. Additionally, homo- - geneous nucleation of water vapor was included to model the water droplet formation in the containment. It should be noted that diffusiophoresis and 7-41
homogeneous nucleation of water vapor were not considered in the initial calcul-ations for Surry (presented in Volume 1 of this report). Four different accident sequences--AB, TMLB', S2 D, and V--were consi-dered in the present calculations. The timing of the fission product source for the containment calculations was permitted to coincide with the prescribed accident sequence that was listed in Chapter 6. Thus, the melt release of aerosol mass occurs as the core starts melting, and the vaporization release takes place as the core-concrete interaction begins. The time-dependent source rates in mass per unit time were provided as input to the NAUA calculations. In certain cases, the effects of multiple compartment modeling (notably for the AB-8 and V sequences), were further examined. The timing of accident events for each accident sequence considered in the present calculations is summarized in Chapter 6. The formation of or reevolution as volatile iodine species in the containment is noted where this effect may be important. In general, if the fraction of the iodine inventory released to the environment for a sequence is in excess of about 0.001, then the volatile iodine is probably an insignificant contributor to total iodine release based on the procedures for estimating this source discussed in Chapter 6. For the final released fractions lower than 0.001, the effect of volatile iodine species is noted on the release
- summary tables.
7.3.1 AB Sequence l This accident represents the sequence in which a minimum retention of fission products is expected to occur in the primary system due to the presence of relatively rapid steam flow conditions combined with a short trans-port pathway. For analysis of the AB sequence, three different containment failure modes were applied and these are identified using WASH 1400 nomenclature as AB-8, AB-y, and AB-c. The containment failure times corresponding to these sequences are shown in Chapter b. The safeguard building was included in the l analysis of the AB-B sequence. As mentioned previously, calculations based on l a multi-compartment model were additionally made to compare with those based on a reduced volume assumption for the AB-S sequence. 7-42 i
7.3.1.1 AB-8. This containment isolation failure mode was analyzed using two compartment models. The NAUA code was run sequentially to calculate the aerosol behavior in the containment and in the safeguard building. Figures 7.21 and 7.22 show the airborne mass in the containment and safeguard building, and Figure 7.23 shows the accumulated mass released to the environ-ment from the safeguard building. Figures 7.24 and 7.25 present the particle size distribution in the containment and safeguard building. The location of
'each species after the accident is completed is shown in Table 7.12.
Table 7.12 shows that a core inventory fraction less than 5 percent for the Csl and Cs0H species is retained in the primary system due to relatively high flow velocity. As a reruit, nearly 80 percent of the Csl and Cs0H inventories are retained in the containment and the safeguard building. The distribution of Te is seen to be substantially different from that of Csl and Cs0H in the table. As already discussed in Section 7.2, the timing of the Te release during the melt release is considerably delayed, and a substantial portion of thc Te core inventory is released directly into the containment during the core-concrete interaction, causing the distribution of Te to differ from that of Csl and Cs0H. To examine the effects of multi-compartment modeling on the ca?culated release fraction, the containment was divided into four separate volumes and similar calculations were performed. The calculated results showing the loca-tional distribution of the CsI, Cs0H, and Te species are shown in Table 7.13. It is interesting to note that the fraction of core inventory for each species that is released to the environment is smaller than the corresponding fraction shown in Table 7.12. While it is rather difficult to assess precisely the effect of the multi-compartment calculation on radionuclide behavior, the observed increase in retention appears to result primarily from altered flow patterns and thermo-hydraulic conditions, such as recirculated and periodic reverse flows, as modeled by the MARCH calculation. 7.3.1.2 AB - y. This containment failure mode represents an earl.v overpressure failure due to hydrogen burning. The containment failure takes place at 269 minutes, which is about 159 minutes after the vessel f ailure. The VANESA calculation results shown in Table 6.13 were used as the source during the core-concrete interaction for this sequence and for the AB 6 sequence. 7-43 i
b .
~
RIRBORNE
- LERKED -
b: - i ),/
- ./
cs _ s m eb: r- . a - c H o . . H b_ . f \ l l b- , , , 10' 10 10 10' TIME, MIN FIGURE 7.21. AIRBORNE AND LEAKED MASSES, AB-8 (Safeguard Building'). l l 7-44 t i l ' t . _ _ _ , . _ - . ._ . , - - - - -
b.. RIRBORNE O
~
LERKED
/ \
( I
- b. . /
l -[ .
- cs .
m m es-r- l e g l o - H _
~
I b_ . l b, 10' 10 10 .....10' TIME, MIN FIGURE 7.22. AIRBORNE AND LEAKED MASSES, AB-s (Safeguard Building) 7-45
l b
- CSI
~
- CSOH - - , '*__________ i b TE
/ '
-1 OTHERS ,.-
- a w A o, l ___- -- -
?
l
/.
b_ ' l l i c.o -
.b, m -i m -
~
C / r -
~
a C H o~ o. i-- -
- I
- f b_
- . J
~
l r o
-i
- I 7
o_
?o 10' 10 10 10' TIME, MIN FIGURE 7.23. ACCUMULATED LEAKED NUCLIDES, AB-S 7-46
'o-: .
.6 HR.
4.2 HR. - - 6.8 HR. -- ['
- 14. 4 HR . ----------- / .
- 20. O HR. - -
\.
r
/ '
,e s \
. s o , ....
s - ox '
\
CS: en : ll
,: '. s\,
\-
. *i :. ,
t z - : i
,/ :f \ '
- \ .
o. e
+ -
*i x -
l! .
\
x a: . HTa <
\ ',',
z: w - l l! g
,: ; i.
o : : z . e
< :i : ',,
o . i o - ,! ;! i. . w >: . z - < ,
\
x l l! \ i ota ,: : i
\.
m: x - > si 1 ', m - I :: : ' x : ,! :! i i
- : i
- i.
! i.
a,o _ . i t l I'
'o, 10-* 10-' 10 10 10 RADIUS, MICRONS FIGURE 7.24 PARTICLE SIZE DISTRIBUTION AB-8 (Containment) 7-47
?o
~
.7 HR.
- 4.3 HR. - -
. 6.7 HR. --
y 13. 8 HR . -----------
,- 18.9 HR. -
- O. .f
?o o ._. -: -
/
a . x : cs .
/
z ,
's \
oro_ - H -4 : / ,. .
\, ,
C : . x - l' - s - i ,: : , z . ,
, :.e i w
i
. i g
,//
O - i i z , :/ i: o?o-l i i o- -
,/ :/ i: i
/
, u - ,e : i z : ,i i: ,
e .
,/ ! i g o -
- i CD . ,! I ',,
.M? . 8 :: * .
C o- -* : l l! i i
- , : i 1
~
l l!
. t r ,
o_ l l i
=
s o ,,,,,,,,'
.4 , , ,
l 10 10 10 10 10' RADIUS, MICRONS FIGURE 7 25 PARTICLE SIZE DISTRIBUTION, AB-8 (Safeguard Building) 7-48
7
' TABLE 7.12. LOCATIONAL DISTRIBUTION OF SFECIES AFTER ACCIDENT IS COMPLETED, AB-8 Fraction of Core Inventory Safeguard Species RCS Containment Building Environment Csl 2.7 x 10-2 0.59 0.29 8.7 x 10-2 Cs0H 3.8 x 10-2 0.59 0.29 8.5 x 10-2 Te 0.26 0.19 0.12' 7.0 x 10-2 i
7-49
c: TABLE 7.13. LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED, AB-8 (4 Volume) Fraction of Core Inventory l Species RCS Vol.1 Vol. 2 Vol. 3 Vol. 4 Environment l CsI 2.7 x 10-2 0.33 0.36 0.23 1.8 x 10-3 5.0 x 10-2 Cs0H 3.8 x 10-2 0.33 0.36 0.23 2.2 x 10-3 4.9 x 10-2 Te 0.26 4.3 x 10-2 5.3 x 10-2 o,19 4,9 x 10-2 4.2 x 10-2 i 7-50 7 w- -
Figures 7.26 through 7.28 show the airborne mass, the accumulated mass leaked, and the particle size distribution for this case. Table 7.14 is the locational distribution of each species after a long period of time elapsed. Table 7.14 shows that compared with the calculated Csl and Cs0H fractions released into the environment (as shown in Table 7.12) for the AB-S sequence, slightly lower fractions are found to escape the containment in the AB-y sequence. However, the Te release fraction is substantially higher because more Te is available in this sequence than in the AB-8 sequence. 7.3.1.3 AB-c. Figures 7.29 through 7.31 show the airborne mass, the accumulated mass leaked to the environment, and the airborne particulate size distribution at various times for the AB-c sequence. As shown in Chapter 6, the containment does not fail until a time of 1450 minutes in this sequence. The first peak in the airborne mass shown in Figure 7.29 represents the source from the RCS prior to the bottom head failure; the second peak repre-sents the source during the core-concrete interaction. The particle size distribution shown in Figure 7.31 attains a bimodal distribution, apparently due to the condensation of steam vapor onto particulates. However, when the thermal-hydraulic conditions in the containment become such that there is no further condensation, the size distribution becomes a unimodal distribution which seems to fit approximately the lognormal function. Table 7.15 shows the results of the fraction of core inventory for each species located in various compartmants. Note that for the failure mode the calculated release of fis-l sion products is into the ground rather than to the outside atmosphere. A l substantial additional decontamination factor would be expected for the source l term to the atmosphere. 7.3.2 TMLB' Sequence l Two containment failure modes were examined for this accident l sequence. The first, designated TMLB'-6, is an early overpressure failure resulting from steam and noncondensible gases. In this failure mode, the containment fails 152 minutes after the core uncovers; this coincides with the l time at which the bottom head fails. At this point, the containment pressure l reaches 89 psi. The second containment failure mode considered here, designated l 7-51
*b
~
. RIRBORNE
. LERKED . .-
l f
.a _
en . . . / , m m - T l r . _a C H . O n -
"b _
o 10' 10 10 10' i TIME, MIN FIGURE 7.26. AIRBORNE AND LEAKED MASSES, AB y
- 7-52 I
k
___ _ .~ . - __ .__ __ . . _ . _ _ _ _ . _ _ _ _ _ . _ _ _ . b .
~
~
CSI -
. CSOH l'-~~~~~~
. TE --
OTHERS -l---------
. I I
,I b_ .
l cs .
~
m m eb-r- - g . A ___.__ . _ . c .I s -
. O .
s . __ f ~ 4
. .I
. A_ b_ . l l l - l l b- . . . . . . . . . . . . . ... 10
' 10, 10' TIME, MIN ,
FIGURE 7.27. ACCUMULATED LEAKED NUCLIDES, AB-y 7-53 1 I
?o ,
- .6 HR.
4.2 HR. - [.mN 6.4 HR. --
- 14. 0 HR . ----------- -
N. g_
-. ~
- 19. 9 HR. -- -- - -
[
/
\.\
/
\ \.
- c. , , .
>O, , ' . . . . . ',
/
,% a .
es ,.,~- ./ ,o' :l '.
' \ -
. .' os :! .
z . >/
- i o
- . // i, ' , ,
a : s-. ~ : . E f- *l : s
!! 1 ' \
e- <: : ' M.% z ,_, : i W -
,l/
s O ~
\ ' -
z ,o i: ' o f : O
- *i \ '.
,/ :: :
- i. t w ,o i 1 '
z - M -
\ '.
ob CD M * -: f l! s! i i m - s: i 5 e : ll 1 i i .
,e :: i.
1-i.
. )
,8 :
- I
- i k.
f/
,: : i e '
t ,u :: s,
- o. _. , .
. g
,I f
f
.. ,o
*Q t \
10 10' 10 10 10' RADIUS, MICRONS FIGURE 7.28. FARTICLE SIZE DISTRIBUTION AB y ! 7-54
i ~y
- k. r s y g s .t
~
, ~~g % ".; '
,? --
h
,s . .1.-
c- . ,
? %
, ' ' , TABLE./.14 LOCATIONAL DISTRIMTION OF SPECIES AFTER s 8 ACCIDENT IS C059PLL}TED, AB y '
w k[ , q g C
- s
.. Fraction of Core Inverstory s ' 2,
*n Species "% ,, 'RCS Containment, Environment i+
Csl . 2.7 x'10-2 _ 0.92 5.7 x 10-2 R-Cs0H 0,,90 5.9 x 10-2 ;g . s 3.8 x)10-2 s
' s a' .3
( - 1 Te 0.26\ 0.31 ' O.14 ', ,
.q ,
3 jf s. f
* ,* 3 s , I \ ;{' s, -
. s '
r,
*'*g
$% 'r., '
T' ' #
> !r4 K
4 i i \ T ,4
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s' s - 4 [ ks
,"} k' S
f g % / I gl l
,+ j % 3
)4 g
- - .s 1 .. #.
g i.
\ t {> . ' , ,
',* (
".. # \ 4 '
{ j .t t a I a g , '
, .f
% f s
%\ '
k (\ \ I l, , 3 g > e , ,f ', 4 e g 4 N
^
4 t 4'
) .
\
5 '1 Tg
-4 4
~
g
).
- k 4 4 b. 3
\
- 4. ]
8 7-ss<
's k
)
o k, NL _ , _ _ , J. .1_ _ . _ __- _ _ . - - . - . - -- . .
b-:
- AIRBORNE 1 LEAKED - -
3 .-b_: b_-:
's .
3. -: s . , ., - 4 % .. - 'v ~ h. 4t :
.f to -
s ,c - J7 x - tn .
<n
- g. s cb_.g z
o 3x ; : C - t H - O g -
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/.
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- h. . .
'fa.-:
r O
~ , , ,
,,,,,i, , , . . . . .
s i
'''5 ,
10 10 10 10' s TIME, MIN Y
.n FIGURE 7.29. AIRBORNE AND LEAKED MASSES, AB-c 7-56 i '
l
. _k_ .
b-5 CSI CSOH - - b TE --
~l: DTHERS -----------
~
o
~1 '
l l ././,, b i , i l
~
/
'o w ~1:
- m - /
m, - C 'o r -3 c : H I on " s 'o, , 1 T I t o
-1 f I
l
'o
~i
_' I
~
f O, t O . .
- i i i iiiiii' i .iisi i i i iiiii ,
10' 10 10 10 TIME, MIN FIGURE 7.30. ACCUMULATED LEAKED NUCLIDES, AB-c l 7-57
'o
- .7 HR. !
4.2 HR. . .
)
6.5 HR. [.N \ 1
- 14. 3 HR . -------- -- -
v 20.0 HR. -
'N
\.
s
/ .
. ,-~,
, s o -
/ g \
l ".
%o.
x - l.
\, ,
cm - -: ,; .
. ,: s
\ ' '
z ,:
. 'i m
o -
,/ :/
I
- s s ,i : i c -
,e i i N?o z
,I #
,f \ 'i -
w :- -
, .i i-O -
- \
z o . ll i
'.,i o .
1 ,: s s u i ' l! , . z -
,! i 1 x . ,: :
i s \ of ,: : i m-o : ' i i g E ,i :;! ; 1 I. : , C : '
,! :! i: 1
{t i
- i
- i, l
i i Y I ! O.
/ i f
- 1
. /
'o 10, 10., ......,' . . , ,,,,,
10 10 10' RADIUS, MICRONS FIGURE 7.31. PARTICLE SIZE DISTRIBUTION, AB-c l 7-58
- l. .. .
r TABLE 7.15. LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED, AB-c Fraction of Core Inventory Species RCS Containment Envi ronment* Csl 2.7 x 10-2 0.97 4.8 x 10-5 Cs0H 3.8 x 10-2 0.96 4.7 x 10-5 Te 0.26 0.45 4.0 x 10-5
- Released to the ground below the containment building. Values for Csl is as an aerosol and does not reflect possible iodine release as a volatile iodide.
7-59
TMLB-6, considers basemat melt-through. In this failure mode, the containment fails at a time of 738 minutes with a containment pressure of 55 psi. Unlike the AB accident sequences, in which fissicn products have a minimum retention in the primary system due to a relatively short pathway, 90 percent of the Csl and Cs0H core inventories are retained in the TMLB' sequence (as discussed previously in Section 7.2). 7.3.2.1 TML B ' - 6. In TMLB'-6, the containment fails shortly after the reactor vessel is ruptured. Compared with the TMLB'-c sequence, the source particulates released from the primary system and from the core-concrete inter-action tend to escape the containment before encountering significant natural retention. No engineered safety features are assumed available in this sequence. The calculated results for the airborne mass and accumulated leaked mass for the containment for the TMLB'-6 sequence are shown in Figures 7.32 and 7.33. The locational distribution of species after a long period of time elapsed is shown.in Table 7.16. The calculated core inventory fractions shown in the table are comparable to those obtained for the AB-y sequence shown in Table 7.14 but are somewhat lower than those for the AB-8 sequence shown in Table 7.12. 7.3.2.2 TML B ' - c. The calculated results for the airborne mass and the accumulated leaked mass in the containment for the TMLB'-c sequence are shown in Figures 7.34 and 7.35. In this accident sequence, the containment melt-through occurs at a time of 738 minutes. The containment pressure reaches about 55 psi at this time. As traditionally assumed, no engineered safety features such as sprays are operating due to the loss of all AC power. Details on the accident event timetable and the time-dependent pressure and tempera-ture conditions can be referred to in Section 6.1. Figure 7.34 shows the mass of the total airborne particulates as a function of time. Again, the first peak of the curve represents the melt release prior to the bottom head failure and the subsequent increase is due to the release, calculated from VANESA, during the core-concrete interaction. It should be noted that in this accident sequence, a water mass of 1.117 x 105 kg at 326 K is assumed present at the time of head failure (t = 157 min), and it is further assumed that the mass evaporates over a time period of approximately 7-60
"o E RIRBORNE LE i'E3 - - - - -
/
.f ~
/
/
./
/
/
I'
/
=a
~
~ /
l
~
l 0 . i
~
w . /
.m '
/
C r _J f . ----. C - H o H o_ 1 ~. 1 a_ n . . . . . . . . i 10* 10, 2*10, TIME, MIN l l-FIGURE 7.32. AIRBORNE AND LEAKED MASSES, TMLB'-6 7-61
o v CSI CSOH -- TE ._ _ ,- _ OTHERS ----------- ,'
/
l
/
/
- o. ,/
#. l l
l l
- ,1 w
m
<n -
c c
._3 e
s,. o o_ s_ f.
/
i
'b -
r .- - 7 I 10' 2*10' TIME, MIN FIGURE 7.33. ACCUMULATED LEAKED NUCLIDES, TMLB'-6 7-62 I
~ TABLE 7.16 LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT. IS COMPLETED, TMLB'.6 Fraction of Core Inventory Species RCS Containment Environment Csl 0.85 0.11 4.6 x 10-2
-2 Cs0H 0.86 0.10 3.9 x 10
-Te 0.30 0.16 0.11 7-63
l 4 b-E_' AIR 80RN!: LERKED - .- b_ I b_-: w : I l _ s . O m 1 . e . r ab c ! H _ C " s :
"o _
l b_ i : 1 t t _ b- . . . . . . . . , 10' 10' 2*10' TIME, MIN FIGURE 7.34. AIRBORNE AND LEAKED MASSES, TMLB'-c 7-64
CSI CSOH - -
- TE --
OTH,\ERS _ _ _ . =====- - - - - - - l i
- l 1 o_
=
l ;
= ,.
lw- - - l e"o _ l i l- --: (D - cn : c _ r . I _J - c H"o o- : H
\r
- I
- x. - . . . .
~
o_ ! b_ -: t o_ - 1 3 l 7*10' 10' 2*10' TIME, MIN FIGURE 7.35. ACCUMULATED LEAKED NUCLIDES, TMLB'-c 7-65
1 hour. The presence of water, of course, delays the vaporization release substantially. Figure 7.35 shows the accumulated leaked mass for CsI, Cs0H, and Te as a function of time. Table 7.17 summarizes the calculated distribu-tion of the core inventory fractions. For the c failure mode release is to the ground rather than directly to the outside atmosphere. A substantial additional attenuation factor would be expected for the atmospheric source term. 7.3.3 V Sequence This sequence represents one of the accidents that potentially permits a large amount of fission products to escape to the atmosphere. In this event, the check valves that provide a barrier between the RCS piping and the low pressure ECC system fail, and fission products can bypass the containment safety features a:1d be released to the auxiliary building. If the auxiliary building is treated as a containment for modeling purposes, calculational methods iden-tical to those used for the other accident sequences can be applied. Results calculated with the TRAP-MELT code were used for the melt release source for NAVA containment calculations. In treating the vaporization release, it was assumed that no attenuation of particulates occurs in their passage from the , reactor cavity through the primary system. This simplification is expected to overestimate the amount of fission products released to the environment. To simulate this sequence more rigorously, it may be necessary to use a multi-volume approach to assess the effects of retention of particulate material in the primary system for the vaporization release term. Two separate TRAP-MELT calculations would be necessary in that approach. The fractions of core inventory of various species released to the atmosphere as predicted by the NAUA code are listed in Table 7.18a. As expected, the calculated results show that this accident sequence is still one of the most important in terms of total release, with the release fractions being about 40 percent of the Cs and I inventories. The airborne mass, accumu-lated leaked mass of each species, and the size distribution in the safeguard building are shown in Figures 7.36 through 7.38. If scrubbing in a water pool on entry into the safeguards building is considered, the results are as shown in Table 7.18b. It is seen that the 7-66
TABLE 7.17. LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED, TMLB'-c Fraction of Core Inventory Species RCS Containment Environment
- Csl 0.85 0.15 2.8 x 10-3 Cs0H 0.86 0.14 1.7 x 10-4 Te 0.30 0.19 8.1 x 10 -2
- Released to the ground below the containment building.
t 7-67
TABLE 7.18a. LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED, V Fraction of Core Inventory Safeguards Species RCS Containment Building Environment Cs! 0.50 1.8 x 10-2 6.9 x 10-2 0.41 Cs0H 0.51 1.7 x 10-2 7.1 x 10-2 0.40 Te 0.13 0.71 4.4 x 10-2 0.12 TABLE 7.18b. LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED, V WITH WATER s Fraction of Core Inventory Safeguards Species RCS Containment Building
- Environment Cs! 0.50 1.8 x 10 -2 0.40 7.9 x 10 -2 Cs0H 0.51 1.7 x 10-2 0.40 7.3 x 10-2 Te 0.13 0.71 0.14 2.5 x 10-2
*A water depth of 3 feet is assumed.
7-68
b- RIRBORNE
- LERKED - --
[.~. j
/
b_ _ C.D v u ,, C - r 1 C w - o _ s b_
~
i l h- , , , , , , , , , . , , , 4*10' 10' 9*10' TIME, MIN FIGURE 7.36. AIRBORNE AND LEAKED MASSES, V (Safeguard Building) 7-69 l
-,-v ---w---ww=,we,*'e----- w ----- w--,-,----o-w,
- g- -yy---v-----w--wer--- ,-m,e,,------e----'e+ - ? ----tv--- +-w-
." CSI y- CSOH - -
- - TE --
- O T H E R S ,,' '
=
,1 l
, ,g. . * * * * . . . - -
- o. l
-s ,
- / --
l b-
-5
/-
e c.9 : s . - l U") - u, . x cb- = _.3 : I C - H : I o n - o, I b
-1 I
r o-
"i l
?o-
-3
, i i i i i * ' ' ' ' ' '
2 4*10' 10 9 10 2 TIME, HIN FIGURE 7.37. ACCUMULATED LEAKED NUCLIDES. V (Safeguard Building) 7-70
1
)
l 1 l 8 o 1.0 HR. 3.5 HR. - - i 6.2 HR. -- 9 . 2 HR . ---- -------
'o . .
1 4 , o H g , ............... ~ ....
.~
o a 5' / N ze o o_
-. - ,s '
w , C s N, \
- x. -
t i-z . s, ' w ,' .x i o . z , , a o .
,o
\,;.
i s, y o . . w I ,' . 's z - s x / 1 ' o ,# *
\
i , D - o : s xn ,' :
\ \ ~ =a - :
- x. .
t,a, o I r o' l '
\:
to : :
,o : . :
l l \ \
.,o : .
....,' ..i' 10 10 10 2*10' RADIUS, MICRONS FIGURE 7.38. PARTICLE SIZE DISTRIBUTION, V (Safeguard Building) 7-71 1
L
presence of gas flow through a hot pool only 3 feet deep has a significant effect with the fractional release to the environment being about 0. for Csl and Cs0H as' compared with values of about 0.4 if.the water pool is ignored. 7.3.4 S2D Sequence The S2 D sequence is a small pipe break accident. The containment failure modes examined were overpressure due to hydrogen burning (52 D-y) and basemat melt-through (S2 0 c). Unlike the AB and TMLB' accident sequences, containment spray systems operate during this sequence; this allows steam to I condense and centainment pressure to be reduced. Since the original NAVA code had no provision for accounting for the role of engineered safeguard, calcula-tions were made by adding the removal mechanism of aerosol particles due to spraying. Details of the modification of the NAUA code were discussed in Chapter 3. In general, spraying systems are highly effective in removing aerosol particles, and it is expected that the spraying mechanism will dominate all natural retention mechanisms in the case of the S D2 sequence. 7.3.4.1 Sp D v. In this sequence, containment failure results from overpressure due to hyrirogen burning. The spray systems were assumed to cease operating when the containment failure occurs. Figure 7.39 shows the time-dependent airborne mass for this failure mode. In Figure 7.39, it is seen that, as expected, the airborne mass remain; very low during the melt release period due to the high removal effectiveness of spraying coupled with the relatively high retention of particulates in the primary system. It should be noted that in the S D 2 calculation shown in Figure 7.39, a drop size of 400 m was used. Figure 7.40 shows the leaked , mass, and Table 7.19 shows the fraction of core inventory leaked to the environ-ment. ! 7.3.4.2 S2 b. This accident sequence is similar to the 5 20-Y sequence except that containment failure does not occur until the basemat melts I through and that containment sprays continue to operate throughout all the ) accident events. Due to the quantity of water in the reactor cavity durinn 7-72
- b. -
E.
~'
RIRBORNE . _E9KEL .-- - - -- I b_ .
. .l
. I
. I l
C,9 I . u;%. . D **4 I e - E - a e " e . O H . b_ w. 1
]
b~ , , , , ,
,,,i' 10 10 2*10' TIME, MIN l
FIGURE 7.39. AIRBORNE AND LEAKED MASSES, SpD-y 7-73
~ -
C,SI. ux- t
. ,t TE __
JTHERS ----------- ,. - b_ -.
,f
,I
~
.-e l
,r
- e -
e : - r : -
. ~
l ! H . o H . l l h_ i
. /
b_ i 18 10' 2*10' TIME, NIN i l l FIGURE 7.40. ACCUNULATED LEAKED NUCLIDES, S2 0-y 7-74
I' TABLE 7.19 LOCATIONAL DISTRIBUTION AFTER ThE CALCULATION IS COMPLETED, $2 D-y Fraction of Core Inventory Containment Environment Species RCS Csl 0.48 0.52 4.2 x 10-5. 0.57 0.43 6.4 x 10-5 Cs0H Te 0.91 5.8 x 10-2 3.3 x 10-2 ,
*This value is for Csl as an aerosol and does not indicate iodine release if volatile iodides are formed in the containment. Estimated release for volatile iodides is about 10-3 4
I 7-75
the core-concrete interaction, the VANESA calculation accounts for the presence of water; these calculations were utilized as the source term during the vapori-zation phase. Figures 7.41 and 7.42 show the airborne mass and accumulated leaked mass of species for the S 2D-c sequence, and Table 7.20 presents the locational distribution of the CsI, Cs0H, and Te species. For the c fai!ure mode the release is into the ground rather than to the outside atmosphere. 7.3.5 Results for Groups of Reactor Safety Study As discussed, the results of calculations presented so far distin-guished Cs!, Cs0H, and Te. In general, one can keep track of an increased number of species in the calculation. In order to demonstrate this capability, additional NAUA calculations were performed for the AB and TMLB sequences. The groups and the corresponding species accounted for in these additional calculations were already discussed in Section 6.2.1. Since the VANESA calcu-lations for the release during the core-concrete interaction did not include all the species to be tracked, it was assumed that the species in a group not included in the original VANESA calculation are released at a rate identical to that for the species in the same group. For example, the release rate for Tc, Rh, and Pd in Group 6 was assumed to be the same as that for Ru and Mo combined after accounting for their core inventories. It should also be noted that Group 1 was not included in the NAVA calculation since Kr and Xe are in the vapor form. Table 7.4 lists the calculated fraction of core inventory released to the environment as a function of both time and group for AB-8, AB-y, AB-c , TMLB '-6, and TMLB '-c. The calculated fraction for AB-8 is based on a two-volume model. It is seen from the table that the containment failure time shows pronounced effects on the amount of release to the environment as expected. Compared with the results of the calculations for Cs!, Cs0H, and Te given earlt tr in Tabits 7.12, 7.14, 7.15, 7.16, and 7.17, the results listed in Table 7.21 for G oups 2, 3, and 4 are slightly different because of the regrouping of the species. In general, the release fraction for Group 6 is seen to be the lowest because of its relatively low release from the fuel. l t 7-76 I
RIR80RNE _,yg . . . . . . - . - b_ b_ I
- {
~_
o
.-e :
C.D
~ $b_
e -? - x- . a - x . H - o H 'o _
~ -
'o _ ;
J - i
?o _ !
i -: # 1 : t i O 2 10' TIME, MIN FIGURE 7.41. AIRBORNE AND LEAKED itASSES, S 2 0-r l , 7-77 i l
1
' ' ~
! CSI -
CSCH - - - - - - --
'o TE -- ,
~,3 JTHERS 8 9 o1 ,I
- ,a
- o
,8
??
- ,'l' a, .-
/
f - r- ! ,- - a'
- i#
f
- I cs //
vfo (n * =, I a- : . r : '
\ I Uo, s-o i s -
t o_ i
-3 t
o, E I t - o,
=
I s-8 o, 10' 10' 2*10' TIME, MIN FIGURE 7.42. ACCUMULATED LEAKED NUCLIDES, S2 D-c 7-78
TABLE 7.20- LOCATIONAL DISTRIBUTION AFTER-THE CALCULATION IS COMPLETED, S2 D-c Fraction of Core Inventory Species RCS Containment Environment
- Csl 0.74 0.26 1.5 x 10-8 CsGH 0.76 0.24 1.4 x 10-8 Te 0.69 0.15 7.7 x 10-8
- Released to the ground below the containment building. Value for Csl is as an aerosol and does not reflect possible iodine release as a volatile iodide.
4 7-79
l TABLE 7.21. FRACTION OF ';0RE INVENTORY RELEASE 0 TO THE ATMOSPHERE Time I Cs Te Sr Ru La (hr) Group 2 Group 3 Group 4 Group 5 Group 6 Group 7 AB-8
- 0. 5 - 4.2 x 10~4 4.5 x 10 ~4 6.4 x 10 -8 1.7 x 10-5 2.4 x 10-6 1.0 x 10 -8 1 2.4 x 10-2 2.4 x 10 -2 8.5 x 10 ~4 3.1 x 10-3 4.1 x 10-4 2.6 x 10-6 2 5.8 x 10-2 5.7 x 10-2 4.9 x 10-3 1.1 x 10-2 1.5 x 10-3 4.8 x 10~4 4 8.1 x 10-2 8.0 x 10-2 3.4 x 10-2 5.1 x 10-2 2.5 x 10-3 2.6 x 10-2 7 8.7 x 10-2 8.6 x 10-2 6.0 x 10-2 7.2 x 10-2 2.9 x 10-3 4.3 x 10-2 10 8.7 x 10 -2 8.7 x 10-2 6.3 x 10-2 7.4 x 10-2 2.9 x 10-3 4.4 x 10-2 15 8.7 x 10-2 8.7 x 10-2 6.6 x 10-2 7.5 x 10-2 2.9 x 10-3 4.5 x 10-2 20 8.7 x 10-2 8.7 x 10-2 6.6 x 10-2 7.6 x 10-2 2.9 x 10~3
, 4.5 x 10-2 AB-y 0.5 1.9 x 10 ~7 2.0 x 10 -7 3.0 x 10 ~II l.6 x 10~9 1.6 x 10 -10 7.3 x 10~II 1 1.3 x 10-5 1.3 x 10~0 5.5 x 10~7 4.2 x 10~7 4.0 x 10 -8 2.5 x 10-8 2 2.6 x 10-5 2.6 x 10-5 2.2 x 10-6 1.5 x 10-6 1.0 x 10~7 5.0 x 10'I 4 4.0 x 10-5 4.0 x 10-5 1.4 x 10-5 2.4 x 10-5 1.9 x 10-7 1.7 x 10-5 7 5.6 x 10-2 5.8 x 10-2 1.2 x 10~I 9.4 x 10-2 2.3 x 10 -3 4.5 x 10-2 10 5.7 x 10-2 5.9 x 10~2 1.3 x 10"I 9.6 x 10-2 2.3 x 10-3 4.6 x 10-2 15 5.7 x 10-2 6.0 x 10-2 1.4 x 10~I 9.7 x 10-2 2.4 x 10-3 4.7 x 10-2 20 5.7 x 10-2 6.0 x 10-2 1.4 x 10 ~I 9.7 x 10-2 2.4 x 10-3 4.7 x 10-2 A_.8 _5.*
0.5 ~7 1.9 x 10 2.0 x 10~7 3.0 x 10~II 1.6 x 10~9 1.6 x 10 -10 7.3 x 10 ~II
-5 ~7 1 1.3 x 10 1.3 x 10-5 5.5 x 10~7 4.2 x 10 4.0 x 10-8 2.5 x 10 -8 2 2.6 x 10-5 2.6 x 10-5 2.2 x 10-6 1.5 x 10-6 1.0 x 10-7 5.0 x 10~7 4 4.0 x 10-5 4.0 x 10-5 1.4 x 10-5 2.4 x 10-5 1.9 x 10~7 1.7 x 10-5 7 4.5 x 10-5 4.5 x 10-5 2.8 x 10-5 4.2 x 10-5 2.3 x 10~7 3.0 x 10-5 10 4.7 x 10-5 4.6 x 10-5 3.4 x 10-5 4.6 x 10-5 2.4 x 10~7 3.4 x 10-5 15 4.8 x 10-5 4.7 x 10-5 3.8 x 10-5 4.9 x 10~0 2.4 x 10~7 3.5 x 10-5 20 4.8 x 10-5 4.7 x 10-5 4.0 x 10-5 4.9 x 10-5 2.4 x 10~7 3.6 x 10-5 7-80 .
e a **t g
# '9,,
UNITED STATES !T ' o NUCLEAR REGULATORY COMMISSION w AsHINGTON, D. C. 20555 i ,
,I
- % ..... /
July 27, 1984 ERRATA SHEET for BMI-2104, Volume V DRAFT Radionuclide Release Under Specific LWR Accident Conditions Prepared by Batelle Columbus Laboratories for the U.S. Nuclear Regulatory Comission Please replace page 7-81 with the attached revised page. DIVISION OF TECHNICAL INFORMATION AND DOCUMENT CONTROL
L TABLE 7.21. (Cor,tinued) Nme ! Cs Te Sr Ru La (hr) Group 2 Group 3 Group 4 Group 5 Group 6 Group 7 TML8-8
'2 -0 0 0 0 0 0 4 3.9 x 10-2 3.7 x 10-2 3.0 x 10-2 1.8 x 10-2 4.8 x 10-3 1.4 x 10-5 7 4.1 x 10-2 3.8 x 10-2 3.5 x 10-2 2.5 x 10-2 4.8 x 10~3 2.1 x 10~4 10 4.4 x 10 -2 3.9 x 10-2 5.7 x 10-2 4.3 x 10-2 4.9 x 10-3 7.0 x 10~4 15 4.6 x 10-2 3.9 x 10-2 9.1 x 10-2 5.3 x 10-3 5.0 x 10~3 9.4 x 10~4 20 4'.6 x 10-2 3.9 x 10-2 9.7 x 10-2 5.4 x 10-2 5.0 x 10-3 9.6 x 10~4 50 4.6 x 10-2 3.9 x 10-2 1.0 x 10'I 5.5 x 10-2 5.0 x 10~3 9.9 x 10'4 l
TML B-s* 10 0 0 0 0 0 0 15 2.7 x 10~3 3.7 x 10~4 8.1 x 10-2 1.8 x 10-2 3.1 x 10-6 4.6 x 10~4 20 2.d x 10-3 3.8 x 10~4 8.3 x 10-2 1.8 x 10-2 3.2 x 10-6 4.7 x 10'4 I 70 2.8 x 10-3 3.9 x 10~4 8.5 x 10-2 1.8 x 10-2 3.3 x 10~0 4.7 x 10'4 4 CReleased with gas flow to ground beneath rer.ctor building. Va'uesforiodine(Group 2) releasu could be increased by approximately an additional ' 10-3 if accounting were made for volatile iodide fermation. l L l 7-81 I ~
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- A I- Cs Te Sr Ru La
- -y a-(hr). Group 2 Group 3 Group 4 Group 5 Group 6 Group 7 T
,/ 'h 3 TMLB-6 -. q 2 JO O O O O O 4 6.M x 1072 5.9 x 10
-2 3.0 x 10
-2 1.9 x 10
-2 5.0 x 10 -3 1.5 x 10 -5 6.0 x 10 -2 3.5 x 10 -2
-2 5.1 x 10 -3
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6 44 x 10-2 2.6 x 10 2.2 x 10
-2 -3 -4 i 10 6.8 x 10 -2 6.1 x 10 -2 5.9 x 10-2 4.6 x 10 5.2 x 10 7.3 x 10
-4 15 7.3 x.10 -2 6.2 x 10-2 9.3 x 10
-2 5.6 x 10 -2 5.3 x 10 -3 9.7 x 10
-2 -2 -2 -3 -3
' [?0 7.0 x 10 -2 6.2 x 10 9.8 x 10 5.7 x 10 5.3 x 10 1.0 x 10
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-4 8.3 x 10-2 1.7 x 10-2 2.3 x 10-5 4.4 x 10 -4 70, 2.8 x 10-3 6.0 x 10 -4 8.5 x 10-2 1.7 x 10 -2 2.4 x 10-5 4.4 x 10-4 1
Va iReleased dith gas flow to ground beneath reactor building. releases if accounting werecould made be in for volatile: iodide formation.
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l 7.3.6 General Observations Based on the results of nine NAVA calculations, the following observa-tions can be made on transport and retention of fission products in the contain-ment. The amount of particulates escaping the containment is largely dependent i upon the timing of the two releases from the fuel (melt and vaporization) and upon containment failure time. As demonstrated in calculations for the i sequences involving an early containment failure (AB-8, AB-y, TMLB-6, and V), I these accidents lead to the highest release fractions. It is worthwhile to note that the retention of fission products in i the flow path of the primary system for sequences involving early containment , failure reduces the source term to the containment and will directly reduce I the amount of fission products that escapes the containment. However, the i retention effects of the primary system on the release to the environment would
~not be so apparent for accidents involving a late containment failure mode.
I In such accident sequences, the fission products that are not retained in the primary system will be retained in the containment due to prolonged residence times. l 7.4 Discussion In July, 1983, Volume 1 of BMI-2104 was issued in draft form. Volume 1 provided analyses for essentially the same sequences as this report. l A number of improvements have been made in the methods of analysis in the intervening period between the reports, such as use of the MARCH 2 rather than ' l the MARCH 1.1 code, a more realistic fuel slumping model, more accurate repre-sentation of upper plenum structures, and inclusion of diffusiophoretic removal j terms in the containment transport analysis. It is important to compare the results obtained in Volume 1 with those of Volume 5 to determine the effects , of these modeling changes. Such a comparison may also provide answers to other critical questions. Rapid advancement has been made in the last few years in ! understanding the processes that influence severe accident source terms. How I mature is our calculational methodology? Will future improvements in the methodology dramatically affect the results we are obtaining with the current l t 7-82 l l
methods? Comparisons between Volume 1 and Volume 5 cannot provide definitive answers to these difficult questions but they may provide insights. In this section of the report results obtained with the current methodology will be compared to those obtained in WASH-1400 for the same sequences and the same plant design.. In Table 7.22 results obtained in this study for the release of fission products from the fuel are compared with the WASH-1400 release terms. The comparison between the results is quite close. In both cases essentially all of.the volatile fission products are predicted to be completely released from the fuel, either while the core is heating and melting or during the I period of concrete attack. The current release estimates during the core melt-down phase are higher for the strontium and ruthenium groups than the WASH-1400 fractions but are substantially less for the least volatile Group 7 which includes lanthanum, uranium, and plutonium. The results obtained in Volume 1 were also quite similar to the current results. Higher release fractions were obtained for Group 7 in the Volume 1 analyses because higher fuel temperatures l were obtained in the MARCH 1.1 analyses. It should be recognized in the inter-4- pretation of these results that the treatment of fuel melting in the MARCH codes is overly simplified. Indeed the manner in which the single melting temperature in the MARCH model has been used in the BMI-2104 analyses may tend to underpredict the release of low volatility elements from the fuel. In WASH-1400 no credit was assumed for the retention of 'ission products during transport through the reactor coolant system. In Volume 1 ! analyses of these accident sequences predicted retention of the important volatile species, iodine and cesium, varied over a range depending on the acci-l dent sequence, from approximately 60 percent in the V and S D2 sequences down to'approximately 20 percent in TMLB' and essentially zero in AB. In the more i- .recent analyses the principal difference is in the TMLB' sequence where the I predicted retention is substantially higher (-85 percent). This difference is primarily due to the prediction of very low flow rates during core melting in the MARCH 1.1 analyses which did not enable the fission products to be trans- ' ported out of the core region prior to vessel failure. The reader should be particularly cautious in the interpretation of results presented in this report for retention in the reactor coolant system. These analyses did not include the effect of the feedback of decay heat from 7-83
TABLE 7.22. FISSION PRODUCT RELEASE FROM FUEL Group 1 2 3 4 5 6 7 P ncipal Xe I Cs Te Sr Ru La Current Study
- Gap /mel t ' l.0 1.0 1.0 0.35 0.33 0.045 2.9 x 10-4 Vaporization -- -- --
0.22 0.10 -- 2.8 x 10 -3 Total 1.0 1.0 1.0 0.59 0.43 0.045 3.1 x 10 -3 WASH-1400 Gap / melt 0.90 0.90 0.81 0.15 0.10 0.03 0.03 Vaporization 0.10 0.10 0.19 0.85 0.01 0.003 0.01 Total 1.0 1.0 1.0 1.0 0.11 0.033 0.01 3
*TMLB' results are used in the example, i
l I I 7 'B4
deposited fission products leading to the heating of surfaces and the reevolu-tion of deposited fission products. This effect has been demonstrated to be potentially important in this study. Reevolution is being investigated in follow-on analyses that will be provided at a later time as an Appendix. In Table 7.23 the overall results for the current study are compared with results from WASH-1400. In some cases where analyses were not performed for the same specific case in WASH-1400, the results of similar sequences are substituted. In many important sequences, the releases predicted with the current methodology are substantially lower than the WASH-1400 results. In TMLB'-6 for example, the predicted release is more than an order of magnitude lower. Most of the difference can be attributed to the predicted retention in the reactor coolant system in this study. Greater retention is also predicted within the containment volume with the NAUA code than was predicted for elemental iodine retention using the WASH-1400 CORRAL code. The high release value obtained in Volume 1 resulted from a combination of low retention on surfaces in the reactor coolant system and release to the containment at the time of containment failure. Under those conditions there was little opportu-nity for removal in the containment prior to release. For the AB-g sequence the increased retention in the current analyses comes from additional retention in the secondary building into which containment leakage occurs. Most but not all isolation failures would be expected to release into a building of this type. The reduced release in the AB-y sequence is the result of longer time to containment failure. In the MARCH 2 analysis a 2-1/2-hour delay was predicted from the time of lower head failure to the time of the global hydrogen burn which was predicted to fail the containment. l Aerosol deposition in this time period was effective in reducing the eventual l release of fission products at the time of containment failure. The important sequence for which the predicted release was not reduced dramatically is V. It has been agreed that the V sequence in Surry should actually involve scrubbing of fission products through water which has been discharged to the safeguards building. No credit has been assumed in this analysis for decontamination in a water pool because of major uncertainties regarding the location of the failure and the structural condition of the safeguards building. As the result of procedural changes introduced subsequent to WASH-1400, the likelihood of event V is believed to have been reduced 7-85
TABLE 7.23. RELEASE FRACTIONS FOR IODINE
- Current. Draft Sequence Estimate Volume 1 WASH-1400 AB-s 9 x 10-2 0.4 0.4 AB-y- 6 x 10-2 0.3 0.7 AB-c 5 x 10 -5 3 x 10 ~4 3 x 10 -3 TMLB'-6 5 x 10-2 0.7 0.7 TMLB'-c 3 x 10-3 1 x 10 -3 3 x 10 -3 V 0.4** 0.2 0.7 2 x 10 -8 -5
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S2D-c 6 x.10 2 x 10 Does not include consideration for volatile iodide forma-tion.
** Release fraction reduced to 0.09 if scrubbing by water in safeguards building is considered.
7-86 L
I l l 1 substantially so that it is no longer considered a principal contributor to the risk. Thus, if the release fractions in Table 7.23 were used to reestimate the risk to the puLiic, a substantially smaller risk would be predicted. In terms of the risk of latent fatalities, the reduction would be essentially proportional to the reduction in the source term of the volatile fission products (e.g., an order of magnitude reduction). Because of the threshold nature of early fatalities, the reduction in the predicted risk would be more dramatic. Again the reader is cautioned to recognize the magnitude of the uncertainties in the methods of analysis. Work is continuing to improve and validate the methods which could lead to changes in source term estimates in the future. In particular, the analysis of fission product reevolution from reactor coolant system surfaces should be recognized as a potential source of nonconservatism in the analysis. l l I l I r ! 7-87
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