ML20095C725
| ML20095C725 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf, Sequoyah, 05000000 |
| Issue date: | 08/07/1984 |
| From: | INDUSTRY DEGRADED CORE RULEMAKING PROGRAM |
| To: | |
| Shared Package | |
| ML20095C674 | List: |
| References | |
| 23.1, NUDOCS 8408230119 | |
| Download: ML20095C725 (175) | |
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8408230119 s4cgo7 p DR ADOCK 050003g7 PDR The Industry Degraded Core Rulemaking Program Sponsored By the Nuclear Industry
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4 DRAFT GRAND GULF NUCLEAR STATION IDCOR Task 23.1 Integrated Containment Analysis (
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TABLE OF CONTENTS Page LIST OF FIGURES ..........................v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . ix
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Statement of the Problem ..................1-1 1.2 Relationship to Other Tasks . . . . . . . . . . . . . . . . . 1-1 2.0 STRATEGY AND METHODOLOGY . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 References . . . . . . . . . . . . . . . . . . . . . . . . . 2-2
3.0 DESCRIPTION
OF MODELS AND MAJOR ASSUMPTIONS . . . . . . . . . . . 3-1 3.1 Plant Specific Information . . . . . . . . . . . . . . . . . 3-1 3.1.1 N u cl e a r Sy s t em . . . . . . . . . . . . . . . . . . . . 3-1 3.1.2 Containment . . . . . . . . . . . . . . . . . . . . . 3-2 3.2 Modular Accident Analysis Program (MAAP) . . . . . . . . . . 3-4 3.2.1 MAAP Ncdalization . . . . . . . . . . . . . . . . . . 3-4 3.2.2 Grand Gulf Systems Modeled in MAAP . . . . . . . . . . 3-6 3.2.3 Fission Product Release from Fuel . . . . . . . . . . 3-9 3.2.4 Description of the Natural Circulation Model . . . . . 3-11 3.2.5 Aerosol Deposition . . . . . . . . . . . . . . . . . . 3-13 3.2.6 Fission Product and Aerosol Release from ' Core-Concrete Attack . . . . . . . . . . . . . . . . . 3-14 3.3 References . . . . . . . . . . . . . . . . . . . . . . . . . 3-15 4.0 PLANT RESPONSE TO SEVERE ACCIDENTS . . . . . . . . . . . . . . . . 4-1 4.1 Plant Response to the T j QUV Accident . . . . . . . . . . . . 4-3 4.1.1 Sequence Description . . . . . . . . . . . . . . . . . 4-3 4.1.2 Primary System and Containment Response . . . . . . . 4-3
DRAFT glE OF CONTENTS (Coltinued) Page 4.1.3 Manual Depressurization Sensitivity Analysis . . . . . 4-15 4.2 Plant Response to the AE Accident . . . . . . . . . . . . . . 4-17 4.2.1 Sequence Description . . . . . . . . . . . . . . . . . 4-17 4.2.2 Primary System and Containment Response . . . . . . . 4-17 4.3 Plant Response to the 23 T 0W Accident . . . . . . . . . . . . 4-27 4.3.1 Sequence Description . . . . . . . . . . . . . . . . . 4-27 4.3.2 Primary System and Containment Response . . . . . . . 4-27 4.4 Plant Response to the 23 T C Acc iden t . . . . . . . . . . . . . . 4-35 4.4.1 Sequence Description . . . . . . . . . . . . . . . . . 4-35 4.4.2 Primary System and Containment Response . . . . . . . 4-39 5.0 PLANT RESPONSE WITH RECOVERY ACTION . . . . . . . . . . . . . . . 5-1 5.1 Plant Respcnse to the T;QUV Accident with Operator Action . . 5-5 5.2 Plant Response to the AE Accident with Operator Action . . . 5-8 5.3 Plant Response to the T23QW Accident with Operator Action . . 5-10 5.4 Plant Response to the T23C Accident with Operator Action . . 5-11 \ 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION . . . . . . . . 6-1 ) 6.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l 6.2 Model ing Approach . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Sequences Evaluated . . . . . . . . . . . . . . . . . . . . . 6-3 6.3.1 Tj QUV Sequ enc e . . . . . . . . . . . . . . . . . . . . 6-3 6.3.2 AE Sequence . . . . . . . . . . . . . . . . . . . . . 6-9 6.3.3 Tg3QW Sequence . . . . . . . . . . . . . . . . . . . . 6-10 6.3.4 AE Sequence . . . . . . . . . . . . . . . . . . . . . 6-10 6.4 References . . . . . . . . . . . . . . . . . . . . . . . . . 6-11
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TABLEOFCONTENTS(Continued) e a page 7.0
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1 Base Case Analyses . . . . . . . . . . . . . . . . . . . . . 7-1 7.2 Operator Action Analyses, . . . . . . . . . . . . . . . . . . 7-5
8.0 CONCLUSION
S . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 APPENDIX A - Grand Gulf Parameter File . . . . . . . . . . . . . . . A-1 APPENDIX B - Supplemental Plots for the Base Accident Sequences . . B-1 Supplemental Plots for Sequencej T QUV . . . . . . . . . B-3 Supplemental Plots for Sequence AE . . . . . . . . . . B-9 Supplemental Plots for Sequence T23@ . . . . . . . . . B-15 Supplemental Plots for Sequence T C . . . . . . . . . B-29 23
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LIST OF FIGURES DRAFI Figure No. Page 3.1 BWR p r ima ry sy s t em . . . . . . . . . . . . . . . . . . . 3-5 3.2 Schematic representation of Grand Gulf Mark III containment and MAAP nodalization . . . . . . . . . . . 3-7 3.3 Schematic representation of Grand Gulf safety and other systems . . . . . . . . . . . . . . . . . . . . . 3-8 3.4 BWR natural circulation model . . . . . . . . . . . . . 3-12 4.1 Pressure in the drywell . . . . . . . . . . . . . . . . 4-7 4.2 Temperature of gas in the drywell . . . . . . . . . . . 4-8 4.3 Pressure in Compartment B . . . . . . . . . . . . . . . 4-9 4.4 Temperature of gas in Compartment B . . . . . . . . . . 4-10 4.5 Average corium temperature in the pedestal . . . . . . . 4-11 4.6 Concrete ablation depth in the pedestal . . . . . . . . 4-12 4.7 Temperature of the suppression pool . . . . . . . . . . 4-14 4.8 Average corium temperature in the cedestal . . . . . . . 4-20 4.9 Cor: rete ablatior depth in the pedestal . . . . . . . . 4-21 4.10 Temperature of gas in the drywell . . . . . . . . . . . 4-23 4.11 Temperature of the suppression pool . . . . . . . . . . 4-24 4.12 Temperature of gas in Compartment B . . . . . . . . . . 4-25 4.13 Pressure in (c.npartment B . . . . . . . . . . . . . . . 4-26 4.14 Ten:eerature of the suppression pool . . . . . . . . . . 4-30 4.15 Tempec:ture of gas in Compartment B . . . . . . . . . . 4-31 4.16 Pressure in Compartment B . . . . . . . . . . . . . . . 4-32 4.17 Temperature of gas in the drywell . . . . . . . . . . . 4-34 4.18 Average corium temperature in the pedestal . . . . . . . 4-36 4.19 Concrete ablation depth in the pedestal . . . . . . . . 4-37
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D~ LIST OF FIGURES (C tinued) Figure No. Page 4.20 Average core power . . . . . . . . . . . . . . . . . . . 4-40 4.21 Reactor vessel water level . . . . . . . . . . . . . . . 4-41 4.22 Pressure in the drywell . . . . . . . . . . . . . . . . 4-43 4.23 Temperature of gas in the drywell . . . . . . . . . . . 4-44 4.24 Pressure in Compartment B . . . . . . . . . . . . . . . 4-45 4.25 Temperature of gas in Compartment B . . . . . . . . . . 4-46 4.26 Average corium temperature in the pedestal . . . . . . . 4-49 4.27 Concrete ablation depth in the pedestal . . . . . . . . 4-50 B.1 Total H2 g enera ted . . . . . . . . . . . . . . . . . . . . B-4 B.2 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-5 B.3 Reactor vessel aater level . . . . . . . . . . . . . . . . B-6 B.4 Temperature of structure, 'F . . . . . . . . . . . . . . . B-7 B.5 Fission oroduct decay power on structure, Btu /hr . . . . . B-8 B.6 To tal C0 genera ted . . . . . . . . . . . . . . . . . . . . B-9 B.7 Mass of water in the pedestal . . . . . . . . . . . . . . B-10 B.8 Mole fraction of H in Compartnent B . . . . . . . . . . B-ll , 2 B.9 Mole fraction of 0 i n Compartmant B . . . . . . . . . . . B.12 2 B.10 Mole fraction of C0 in Compartment B . . . . . . . . . 8-13 2 B.ll Mole fraction of steam in Compartment B . . . . . . . . . B-14 B.12 Volumetric flow out of containment . . . . . . . . . . . .'B-15 B.13 Mass of UO i n core reg ion . . . . . . . . . . . . . . . . B-16 2 B.14 Total H2 g enera ted . . . . . . . . . . . . . . . . . . . . B-18 B.15 Total H2 genera ted . . . . . . . . . . . . . . . . . . . . B-19 B.16 Reactor vessel water level . . . . . . . . . . . . . . . . B-20
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LISTOFFIGURES(Continued) Am Figure No. Page B.17 Temperature of structure, 'F . . . . . . . . . . . . . . . B-21 B.18 Fission product decay heat on structure, Btu /hr . . . . . B-22 B.19 To tal CO g enera ted . . . . . . . . . . . . . . . . . . . . B- 23 B.20 Mass of water in the pedestal . . . . . . . . . . . . . . B-24 B.21 Mole fraction of H i n Compartment B . . . . . . . . . . . B-45 2 B.22 Mole fraction of 0 in Compartment B . . . . . . . . . . . B-26 2 B.23 Mole fraction of CO in Compartment B . . . . . . . . . . B-27 2 B.24 Mole fraction of steam in Compartment B . . . . . . . . . B-28 B.25 Volumetric flow out of containment . . . . . . . . . . . . B-29 B.26 Mass of UO i n core region . . . . . . . . . . . . . . . . B-30 2 B.27 Total H2 g enera ted . . . . . . . . . . . . . . . . . . . . B-32 B.28 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-33 B.29 Reactor vessel water level . . . . . . . . . . . . . . . . B-34 B.30 Temperature of structure, 'F . . . . . . . . . . . . . . . B-35 B.31 Fission product decay heat on structure, Btu /hr . . . . . E-35 B.32 Total C3 generated . . . . . . . . . . . . . . . . . . . . B-37 B.33 Mass of water in the pedestal . . . . . . . . . . . . . . B-38 B.34 Mole fraction of H in Compartment B . . . . . . . . . . . B-39 2 B.35 Mole frsction of 0 in Compartment B . . . . . . . . . . . B-40 2 B.36 Mole fraction of CO in Compartment B . . . . . . . . . . B-41 2 B.37 Mole fraction of steam in Compartment B . . . . . . . . . B-42 B.38 Mass of UO i n core regio n . . . . . . . . . . . . . . . . B-43 2 B.39 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-46 B.40 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-47
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D LIST OF FIGURES (Continued) Figure No. Page B.41 Reactor vessel water level . . . . . . . . . . . . . . . . B-48 B.42 Temperature of s tructure. *F . . . . . . . . . . . . . . . B-49 B.43 Fission product decay heat on structure. Btu /hr . . . . . B-50 B.44 To tal C0 g enera ted . . . . . . . . . . . . . . . . . . . . B-51 B.45 Mass of water in the pedestal . . . . . . . . . . . . . . B-52 8.46 Mole fraction of 2H in Compartment 8 . . . . . . . . . . . B-53 B.47 Mole fraction of 20 in Compartment 8 . . . . . . . . . . . B-54 B.48 Mole fraction of CO2 in Compartment B . . . . . . . . . . B-55 B.49 Mole fraction of steam in Compartment B . . . . . . . . . B-56 B.50 Volumetric flow out of containment . . . . . . . . . . . . B-57 B.51 Mass of 200 i n core region . . . . . . . . . . . . . . . . B-58 h
DRAFT LIST OF TABLES Table No. Page 3.1 Initial Inventories of Fission Products and Structural Materials Released from the Fuel . . . . . . 3-10 4.1 Grand Gulf Nuclear Station. T QUV j - Base Case. Accident Chronology . . . . . . . . . . . . . . . . . . 4-4 4.2 Effects of Depressurization in the T QUV j Accident . . . 4-16 4.3 . Grand Gulf Nuclear Station, AE Base Case, Accident Chronology . . . . . . . . . . . . . . . . . . 4-18 4.4 Grand Gulf Nuclear Station, T 0W - Base Case. 23 Accident Chronology . . . . . . . . . . . . . . . . . . 4-28 4.5 Grand Gulf Nuclear Station, T C Base Case, Accident Chronology . . . . .23. . . . . . . . . . . . . 4-38 5.1 Systems Available for Core and Core Debris Cooling . . . 5-2 5.2 Systems Available for Containment Cooling and Pressure Control . . . . . . . . . . . . . . . . . . . . 5-3 5.3 Operator Response Selection . ........... . 5-4 5.4 Grand Gulf Nuclear Station, T QUV 3
- Operator Action Case, Accident Chronol 6gy . . . . . . . . . . . . 5-6 5.5 Grand Gulf Nuclear Station, AE - Operator Action Case, Accident Chronology . . . . . . . . . . . . 5-9 5.6 Grand Gulf Nuclear Station. T 7-QW - With i Operation Action. Accident Chr$nology . . . . . . . . . 5-12 5.7 Grar.d Gulf Nuclear Station, T C - Operator 3
Action Case, Accident Chrono 189y . . . . . . . . . . . . . s-14 6.1 Distribution of Csl in Plant and Environment (Fraction of Core Inventory) . . . . . . . . . . . . . . 6-4 6.2 Tj QUV Fission Product Release . . . . . . . . . . . . . 6-5 6.3 AE Fission Product Release . . . . . . . . . . . . . . . 6-6 6.4 T23QW Fission Product Release . . . . . . . . . . . . . 6-7 6.5 T23C Fission Product Release . . . . . . . . . . . . . . 6-8 l t i
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DRAFT LIST OF TABLES (Continued) Table No. Page 7.1 Sumary of Fractional Radionuclide Releases to th e Env i ro nme n t . . . . . . . . . . . . . . . . . . . 7- 2 I l
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1.0 INTRODUCTION
DRAFT 1.1 Statement of the Problem The objective of this investigation was to calculate the response of the Grand Gulf Nuclear Station (GGNS) primary system and contairrnent to postulated severe accident sequences which have been identified as potentially leading to core degradation and melting. These analyses include evaluations of the thermal-hydraulic response, the release of fission products from degraded fuel, and the transport of the released fission products within the containment. These calculations were performed on a best estimate basis phenomenologically and include assessments of the major uncertainties associ-ated with state-of-the-art modeling. This study includes assessments of the results of a limited set of operator interventions in these sequences and an assessment of the influence of a specific mitigating feature associated with the Grand Gulf Nuclear Station design. 1.2 _ Relationship to Other Tasks The primary system and containment response analyses of IDCOR Subtask 23.1 are dependent upon the primary system and containment thermal-hydraulic models developed in du~c tasks 16.2 nd 16.3 (Executive Analysis Prograin) and the fission product release and retantion models developed in IDCOR Task 11 (Fission Product Transpot t). The accident sequences used fcr the analyses along with the operator interventions w.re developed by consider-ing the dominant accident sequences identiffed in Subtask 3.2 (Assess Dominant Sequences) and the physical processes oc:urring ouring these accioents. It should be noted that the analyses developed as part of IDCOR Subtasks 16.2 and 16.3 involve the detailed consideration of many different phenomena which are themselves considered in separate IDCOR subtasks. These include: hydrogen generation; distribution and combustion (Subtasks 12.1, 12.2 and 12.3); steam generation (Subtask 14.1); core heatup (Subtask 15.1); debris behavior (Subtask 1.5.2) and core-concrete interactions (Subtask 15.3).
DRAFT Operater intervention sequences were developed as part of Subtask 23.1 and appiied to the specific accident sequences in the Grand G'ulf Nuclear Station design to determine those potential actions which could terminate the accident sequence and result in a safe stable state. These results were used in IDCOR Subtask 22.1 (Safe Stable States) which discusses both the inherent and intervention means of tenninating the various core damage sequences considered for the Grand Gulf Nuclear Station design. The mitigative design feature sequence for GGNS was developed via a review of a list of mitigative and preventative design features identified in IDCOR Task 21 (Risk Reduction Potential). The ultimate structural capability of the containments associated with the reference plants and other typical designs were assessed in IDCOR Subtask 10.1. These analyses define the containment failure pressure and failure made in this analysis. Calculations of the rate and amount of fission products released from the containment, for those sequences which result in containment failure,
. - were supplied to IDCOR Subtask 18.1 (Atmospheric and Liquid Pathway Dose) to formulate assessments of the health consequences associated with these postu-lated accident scenarios. These health consequence analyses were then sup-plied to IDCOR Subtask 21.1 to evaluate the risk reduction pott.ntial for possible raitigatir.g operator actions and containment mitigative design fea tures.
Detailed considerations for each of the related subtasks can Le found in the final reports submitted for the specific task. Individual issues are addressed in this report only as required to understand the specific behaviors obtained for the accident sequences considered.
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2.0 STRATEGY AND METHODOLOGY DRAFT The basic strategy of this subtask was to analyze accident sequences which have been previously identified as potential contributors to core melt frequency. These analyses consisted of plant thennal hydraulic response and fission product transport calculations for accident sequences which led to core degradation and melting. These analyses model perfonnance of the ECCS systems and the containment engineered safety systems, such as the suppression pool, decay heat removal system, etc. The NAAP code [2.1] was used to perform the primary system and containment thennal-hydraulic response analyses. This code considers the major physical processes associated with an accident progression, including hydrogen generation, steam fonnation, debris coolability, debris dispersal, core-concrete interactions, and hydrogen combustion. The FPRAT module for MAAP was adopted from [2.2] to evaluate the fission product release from the fuel. Natural and forced circulation within the primary system is modeled both before and after vessel failure and is integrated with the fission produce release model to determine the transport of vapors and aeroscis thrcughout the pric,ary system and containment. Fission product deposition proce:ses tredeled include vapor condensation, steam condensation and sedimentation. , for eacn of the four GGNS accident sceqarios selected for analysis, ther.aal-hydraulic calculation: were perfonned both with and without selected operator actions during tne accident. The " base case" analyses, which assume only minimal cperator response during the accident, establish a reference sys ter. response during each of the accident scenarios. The " operator action" analyses are branch calculations of the base cases. These operator interven-tion cases demonstrate the effect of selected operator actions on the progres-sion of an accident, based on the time windows available to the operator to take such action. Additional uncertainty and sensitivity analyses have been performed on several key parameters associated with the accident response. These are reported in Ref. [2.4].
DRAFT 2.1 References 2.1 "MAAP, Modular Accident Analysis Program User's Manual," Technical Report on IDCOR Tasks 16.2 and 16.3, May 1983. 2.2 " Analysis of In-Vessel Core Melt Progression," Technical Report on IDCOR Subtask 15.lB, September 1983. 2.3 Richard K. McCardell, " Severe Fuel Damage Test 1-1 Quick Look Report," EG&G Idaho, October 1983, 2.4 IDCOR Technical Report on Task 23.4, " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," to be published, d V 6 1 l
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DRAFT
3.0 DESCRIPTION
OF MODELS AND MAJOR ASSUMPTIONS l The Modular Accident Analysis Program (MAAP), Ref. [3.1] is used to model the Grand Gulf Nuclear Station (GGNS) response to postulated severe accidents. This code includes contaiment response, fission product release, and fission product transport. In addition, both the thermal hydraulic response and the fission product behavior are modeled for the reactor building which surrounds the primary containment. 3.1 Plant Specific Information The Grand Gulf Nuclear Station (GGNS) is a two unit boiling water reactor located in Claiborne County, Mississippi, on the east side of the Mississippi River approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez, Mississippi. The two units are nearly identical; both will be operated by Mississippi Power & Light Company (MP&L). Unit 1 is scheduled to go into comercial operation in early 1985; Unit 2 is scheduled to do so several years later. Each unit is designed with a core themal output of 3833 MWth, a gross electrical power output of 1306 MWe, and a net electrical output of 1250 MWe. Each unit is powered by a BWR-6 water reactor, designed and supplied by &neral Electric Company. Each reactor is housed in a steal-lined reinforced c::ncrete Mark III containment building. 3.1.1 Nuclear System The primary system consists of the equipment and instrumentation necessary to produce, centain, and control the steam pcwer required by the turbine-generator. Principal components of the system are the reactor pres-sure vessel (RPV) and internals, reactor water recirculation system, and the main steam system. Other important systems include the condensate and main feedwater systems which close the primary system flow loop by condensing the steam and water exhausted by the turbines and pumping this condensate back into the RPV. The reactor vessel houses the reactor core, contains the heat, produces steam within its boundaries, and serves as one of the fission product barriers during nomal operation and in the event of fuel failure.
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DRAFI The core is composed of 800 fuel assemblies, each containing 62 fuel rods and two hollow water rods. These fuel rods are sealed Zircaloy-2 tubes, which are loaded with UO'
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2 fuel pelle's, t with the Zircaloy-2 cladding providing both structural suppof t and a fission product barrier between the fuel and the primary system water. The remaining reactor pressure vessel internal compo-nents support and align the fuel and provide the water circulation flow paths
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to distribute , coolant to the fuel. Upper vessel internals also furnish moisture removal for the steam generated within the core, to minimize the moisture content of the exiting steam. The major internal components consist of the core, the shroud top grid, core plate, steam generator and dryer, jet
, pumps, control rods, and control rod ' drives.
The reactor water recirculation system provides a forced continuous internal circulation of coolant water through the core. Four main steam lines dirIct steam to the balance of the plant. During an abnormal event occurring during power operation, main steam isolation valves (MSIVs) on each of these lines provide -isolation of the reactor vessel from the balance of the plant. If their closure is required, a set of 20 safety / relief valves (SRVs) provide
; ,, , ,reacter vessel' overpressure. protection, with their discharge being directed to the suppression, pool. .
The majority of the primary system data used in this analysis came from tho Grand Gulf FSAR [3.2]. This information includes initial conditions, pressures, temperatures, flow rates, enthalples, masses, system pressure setpoints, control logic, and other parameters. A plant parameter file for MAAP was prepared based on these data; it appears in Appendix A.I. 3.1.2 Centa inment The reactor vessel is housed in the containment building. This ~ structure is designed to condense the steam (pressure suppression) and contain the fission products which mcy be released as a result of a loss of Coolant Accident (LOCA).
, The Mark III containment is a steel-lined reinforced con-crete structure. with a cylindrical shape, topped with a hemispherical dome.
The containment foundation is a thick, circular reinforced concrete slab. Major eleme6ts of this pres,sure-suppression design are an inner volume and an l
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DRAFT outer volume,. separated by a large heat capacity suppression pool. The inner region, the drywell, is a cylindrical volume containing the reactor pressure vessel, which is supported by a hollow concrete cylinder called the pedestal. The drywell and outer containment volumes communicate via horizontal vent openings located below the suppression pool surface. A water seal between the inner and outer volumes is accomplished by the drywell weir wall. The pool which provides 1 r steam suppression during postulated LOCA events. The outer containment volurre consists of the annular space above the suppression pool and the dome. The upper containment pool, located in the outer containment volume, provides a post-LOCA source of makeup water to the suppression pool. Containment sprays, also located in the outer compartment, provide an addi-tional means of rapidly removing possible post-accident steam and/or fission products from the outer containment atmosphere. In addition to these fea-tures, hydrogen igniters are located in both drywell and outer containment volumes to control hydrogen accumulations following postulated severe accidents. The GGNS BWR-6/ Mark III design, like that of other nuclear plants, is based on a defense-in-depth principle. Thus, if an abnormal event were to occur, backups to the normal systems are designed to maintain the integrity of the fuel cladding, the reactor pressure vessel, and the containment barriers. These backup systems perfonn two general ful.ction:,: core co:lin; and contain-merit pressure control. Those systems which perfonn the first function include the reactor core isolation coolir.g (RCIC) system, the high pressure and low pressure emergency core cooling systems (ECCS), the automatic derrcssurization system (ADS), and the stendby liquid control (SLC) systs. The containneat pressure control function is accouplished via the suppression pool makeup system, the drywell purge system, the post-LOCA vacuum breakers, the suppres-sion pool cooling and containment spray medes of the residual heat removal I (RHR) system, and the hydrogen ignition system. MAAP input data, including initial conditions, heat transfer coeffi-cients, exposed surface areas, and flow areas between volumes are based on l information from the Grand Gulf FSAR r3.2], and architect / engineer drawings. These data appear in the MAAP parameter file listed in Appendix A.I. l
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DilAFT 3.2 Modular Accident Analysis Program (MAAP) Within the IDCOR Program, the phenomenological mode's developed in Tasks 11,12,14 and 15 have been incorporated into an integrated analysis package in Subtask 16.3, while Subtask 16.2 provides a computer code (MAAP) to analyze the major degraded core accident scenarios for both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The MAAP code is designed to provide realistic assessments for severe core damage accident sequences using first principle models for the major phenomena that govern the accident progression, the release of fission products from the fuel matrix, the trans-port of these fission products and their depos,ition within the primary system and containment. The following sections describe the primary system and containment nodalization and include a description of the safety systems modeled in MAAP. 3.2.1 MAAP Hodalization The BWR primary system nodes are illustrated in Figure 3.1 and include the lower plenum, downcomer, core, and upper plenum. Also indicated are the flow entry locations for CRD flow, feedwater, HPCS, RCIC, LPCI and LPCS as well as the standby liquid control system (SLCS). The SLCS is only modeled as an additional water source since MAAP does not have a neutronics model. Individual mass and energy equations are written for each of these nodes using the water addition locations and the appropriate connecting flow paths. The primary system model also represents the main steam isolation valves and the main steam safety and relief valves. The latter exhaust into the suppression pool. l Modeling of the primary system is used to determine if a given l sequence (1) leads to core uncovery, (2) results in core damage, (3) yields Zircaloy clad oxidation and hydrogen formation, (4) leads to core melt and vessel failure, or (5) can be recovered before vessel failure. The code predicts the times of these occurrences. The transient response to the l spectrum of accident scenarios considered requires the specification of pump l 1 curves, valve set points, system logic, etc. With the specification of the l accident sequence, the primary system model determines the vessel water
DRAFT SRV o
. MAIN STEAM UPPER PLENUM ] -
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DRAFT inventory, including the boiled-up level in the core, to evaluate the poten-tial for core uncovery. If the collapsed water level decreases below the top of the core, the HEATUP subroutine calculates the temperature increases for the fuel and cladding. Steam cooling and the oxidation of the Zircaloy clad and channels are determined by the appropriate rate laws and oxygen starva-tion. The model accounts for the cooling effect of CR0 flow. If available, this flow can limit core damage for long-term heat removal failure events. The Mark III (Grand Gulf) containment nodalization scheme, as shown in Fig. 3.2, separates the containment into five compartments: the pedestal, the drywell, wetwell, Compartment A (annulus above the wetwell), and Compart-ment B (above the operating deck) regions. MAAP evaluates the behavior of the various compartments during the entire progression of the accident sequence by calculating the mass and energy flow rates between these compartments. Individual compartment (region) pressures and gas temperatures are derived from the mass and energy balances. MAAP models the transport of all
'! material throughout the containment due to drainage, vaporization, condensa-tion and mass addition to assess the potential for cooling core debris.
Separate water and corium temperatures are calculated for each containment compartment. 3.2.2 Grand Gulf Systems Modeled in MAAP In general, MAAP characterizes the response of the primary system, the containment, and many of the balance of plant systems to user specified event sequences. Figure 3.3 illustrates the plant systems modeled in the code including the various water sources available and the valve line-ups which would allow this water to be injected into either the primary system and/or containment during a postulated sequence. Particular systems of importance ! include, the control rod drive (CRD) flow from the condensate storage tank, main steam lines, MSIVs, turbine bypass, feedwater, reactor core isolation cooling (RCIC), high pressure core spray (HPCS), low pressure coolant injec-tion (LPCI) and other RHR system modes, low pressure core spray (LPCS), standby liquid control system (SLCS), and high pressure service water (HPSW). l In addition to these plant systems, MAAP nodalizes both the primary system and
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l l DRAFT containment.to model their response to postulated core damage and recovery i scenarios. 3.2.3 Fission Product Release from Fuel The FPRAT module in MAAP, as adapted from Ref. [3.3] was used to calculate the release rates of fission products from the fuel matrix. These rates are dependent upon the fuel temperature history during heatup and upon characteristics of the atmosphere within the vessel which effect saturation of the chemical species as discussed in IDCOR Task 11.1 [3.4]. Fuel temperature histories for the thirty regions in the core are tracked to determine the release characteristics for the fission products and inert materials. The initial inventories of the various fission products were obtained from Ref. [3.5]andaregiveninTable3.1. The gas flow through each node is assumed to be saturated with the vapor of each constituent. If the flow cools as it is transported to higher nodes, the gas cools and creates aerosols of each species to remain saturated. This flow provides the aerosol and vapor source for the upper plenum. For the regions in which blockage has occurred, it is assumed that sufficient flow exists to remove the volatile fission products as saturated vapor. Once this flow is determined, the removal of the remaining less volatile species is evaluated based upon saturation of this calculated flow. The required FPRAT input for MAAP is given in the parameter file in Appendix A.l. The calculations consider evaporation and condensation characteris-tics of chemical species. Several key assumptions consistent with the recom-mendations of IDCOR Subtask 11.1 were made regarding the physical and chemical forms of released fission products. These are:
- 1. Cesium and iodine combine to form Csl upon entry to the fission product release pathway. The excess cesium forms Cs0H. Both chemical species exhibit similar physical behavior, hence the source rate for the Cc,I fission product group is assumed to be the sum of the Cs and I release rates. As stated above, it is assumed to be liberated in vapor form.
l l
DRAFT Table 3.1 INITIAL INVENTORIES OF FISS:' ?RODUCTS AND STRUCTURAL MATERIALS RELEASED FROM THE FUEL Fission Products Initial Inventory (kg) Kr 27.3 Xe 412 Cs 220 I 17.7 Te 37.1 Sr 66.7 Ru 183 Mo 252
.,. Sn , 1190 Mn 268 l
l t .__
DRAFT 1 2.- Telluritsn is assumed to be released as vaporized Te0 ' j 2 l
- 3. Inert aerosol generation rate is the combined release rates for volatile structure material (Mn and Sn).
- 4. Strontium and ruthenium represent their respective nonvolatile fission product groups as defined in WASH-1400. They are also calculated to be released as vapor which quickly forms aerosols when they exit the core.
- 5. Release of the volatile fission products (Cs I, Te) and the noble gases (Xe and Kr) is allowed to continue until complete, even if the vessel has failed first..
3.2.4 Description of the Natural Circulation Model Substantial quantities of fission products are released during core degradation, but before vessel failure. Gas flow through the primary system determines the aerosol transport and deposition throughout the reactor vessel. Following vessel failure, most fission products remain within the primary system and subsequently heat the adjacent structures. As the structure and gas temperatures increases, density differences within the primary system would result in natural circulation flows that could distribute both heat and mass throughout the primary system. The natural circulation model determines flows within the primary system, and includes descriptions for fission product heat generation, mate-rial vaporization, condensation and deposition. Also, the nodalization allows for a representation of the structural heatup in each node as well as the heat losses from these nodes to the containmcat environment. The circulation for the BWR system af ter vessel failure is graphically represented in Fig. 3.4. As illustrated, the throat area for the jet pumps controls the circulation ! rate and the containment pressurization /depressurization influences the flow from the primary system. 1
DRAFT . H GAS FLOW _L
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AND MATERIAL- % -- -- TRANSPORT f SEPARATORS a - 4 ORYERS DOWNCOMER Ag 'O CME O 1 l 1 1 ll Fig. 3.4 BWR natural circulation model. i
Since' natural circulation flows are driven by the gas density MMT differences between various regions, and since the volatile fissio'n products are dense vapors, calculation of the gaseous flows within the primary system must account for the gas mixture properties in the various nodes. In addi-tion, with the reflective insulation used on the Grand Gulf reactor vessel, the heat losses from the vessel must also include the magnitude of heat losses as a function of the primary system temperature and the potential for oxida-tion of the stainless steel layers in the reflective insulation. These analyses have been coupled with models for aerosol deposition and heatup to evaluate the primary system flows af ter reactor vessel failure. Such assessments provide the rate and amount of material lost from the primary system as a result of the subsequent heatup of primary system structures. In this analysis, the difference between the primary system and containment pressurization determines the flows between these two systems which govern the release of fission products to the containment environment. 3.2.5 Aerosol Deposition IDCOR Task 11.3, Ref. [3.6], applied state-of-the-art fission product behavior models to produce the RET,AIN code, which describes the aerosol agglomeration and removal processes based upon an assumed log-normal distribution [3.6]. Both vapor and aerosol forms of fission products are considered. MAAP represents the aerosol removal rate due to settling as a function of the aerosol cloud density [3.5]. This is consistent with the general behavior predicted by detailed descriptions, such as RETAIN, and more importantly, is in good agreement with the results of large scale experiments. MAAP models physical mechanisms for vapor condensation on structures and aerosol retention due to steam condensation in addition to gravitational settling. These removal processes substantially reduce the magnitude of the release to the environment. The primary system and containment nodalization for fission product transport are the same as those used for the thermal hydraulic calculations. The specific transport paths were earlier illustrated in Fig. 3.2 for the primary system and containment. l
DRAFT The key assumptions in the aerosol modeling are:
- 1. Cesium and iodine are assumed to be released as Csl with excess cesium as Cs0H.
- 2. The decontamination factor associated with the wetwell suppres-sion pool is estimated to be 1000 for releases through the spargers and 600 for releases through the horizontal vents
[3.R].
- 3. Prior to vessel failure any fission products that may enter the drywell (such as from a LOCA pathway from the primary system) are available to enter Compartment A via the slight design-basis drywell leakage. These pathways are assumed to be closed off following vessel failure due to plugging by aerosols [3.9].
- 4. Fission products reaching the SRV discharge lines were treated as having reached the suppression pool.
- 5. Hygroscopic aerosols, such as cesium hydroxide, are assumed to accumulate an equilibrium concentration of water as detemined by the steam partial pressure and temperature.
- 6. Release of volatile fission products (Cs, I, Te) and the noble gases (Xe and Kr) is allowed to continue until complete, even if the vessel has already failed.
l 3.2.6 Fission Product and Aerosol Release from Core-Concrete Attack i The release of aerosols due to core-concrete attack was determined using a model based on the concrete ablation rates from PAAP. The mass of low volatility fission products and inert aerosols released from core debris is based upon a vapor stripping model assuming the melt constituents follow l Raoult's law. This calculation is dependent upon the amount of gas sparging ! through the core debris, the molar concentration of fission products in the i l
DRAFT core debrisi the vapor pressure of the chemical species of interest, and the temperature of the core debris. The key assumptions are:
- 1. The masses of CO and water vapor released per cubic meter 2
ablated for the limestone concrete used at Grand Gulf are 572 kg and 130 kg respectively.
- 2. Stripping only occurs when the corium is calculated to be moiten.
- 3. The gases released by the downward attack pass through the molten pool and cause stripping. Gases generated by sidewall attack are assumed to bypass the pool.
- 4. The predominant form of Sr is Sr0, of Ru is elemental Ru, and of La is La230.
- 5. Inert aerosols of Ca0 may be generated during core-concrete a ttack. This chemical form is used as a surrogate for the various concrete melt constituents that could be added to the corium pool.
- 6. Deposition of fission products in the SRV discharge lines was neglected.
- 7. Concrete aerosol generation was not incorporated into the overall fission product removal calculations but was used to make an assessment of the extent of plugging of the drywell to compartment A pathway.
3.3 References 3.1 "MAAP, Modular Accident Analysis Program," Technical Report on 10COR Subtasks 16.2 and 16.3, 1983.
DilAFT 3.2 Final Safety Analysis Report, Grand Gulf Nuclear S tation, Mississippi Power and Light Company,1979. 3.3 " Analysis of In-Vessel Core Melt Progression," Technical Report on IDCOR Subtask 15.18, September 1983. 3.4 " Estimation of Fission Product and Core-Material Source Characteris-tics," Technical Report on 10COR Subtasks 11.1, 11.4, and 11.5, 1983. 3.5 Radionuclide Release Under Specific LWR Accident Conditions -- Volume III BWR, Mark III Design, Battelle Columbus Laboratories, 1984. 3.6 " Fission Product Transport in Degraded Core Accidents," Technical Report on IDCOR Subtask 11.3, December 1983. 3.7 " Aerosol Deposition Model," FAI report to be published. 3.8 Personal Communication, K. Holtzclaw (GE) to R. Henry (FAI), March 1984. 3.9 H. A. Morewitz, " Leakage of Aerosols from Containment Buildings," Health Physics, Vol. 42, No. 2, pp. 195-207, 1982. i
4.0 PLANT RESPONSE TO SEVERE ACCIDENTS DRAFI This section provides the results of plant thermal-hydraulic analy-ses of four base case accident sequences, using the MAAP code. The accident scenarios are specific cutsets of each accident sequence and, as such, are not necessarily representative of all cutsets of these sequences. The accident scenarios are defined below, followed by descriptions of the reactor coolant system response and the containment response. The time dependence of the most significant MAAP-generated thermal-hydraulic parameters associated with each scenario are presented in Appendix B. The plant parameters utilized to characterize Grand Gulf in these analyses are listed in Appendix A. The base sequences are:
- 1. Tj QUV - Transient with failure of injection.
- 2. AE - A large LOCA with failure of injection.
- 3. T23QW - Transient followed by loss of containment heat removal.
- 4. T23C - Transient followed by failure of the reactor to scram and standby liquid control (without operator action to reduce powerlevel).
The T)QUV was analyzed both with and without manual activation of the ADS in order to determine if this action would play a significant role in the overall containment response and fission product release. The sequences analyzed in this section are low probability core damage events. The sequences exclude all, or nearly all, operator actions that could prevent or significantly delay core melt or that could mitigate its consequences. Operator actions which would prevent the accident are consid-ered in the determination of the sequence probabilities. Those which would mitigate the consequences are not considered. This approach was taken to produce results which bound or are at the high end of the range of possible consequences for the four selected sequences. Generally, only minimal l
DRAFT operator actions to control selected plant systems are assumed for these events. For example, it is assumed that the operators regulate low pressure injection to maintain water level at the high level trip. As a result of the minimal operator response models employed in this analysis, the results presented here do not represent what would be realis-tically expected to occur for the specified equipment failures and are ex-tremely improbable. The more probable expected plant response to the speci-fied equipment failures is evaluated in Section 5. This later section in-cludes in the sequence definition some examples of actions which the operator would be expected to take in accordance with the Emergency Procedure Guide-lines. As a result of these actions the operator is able to terminate the event prior to core melt or significantly mitigate its consequences. Section 5 considers only some examples of the many actions available to the operator to prevent or mitigate the accident. A major objective of excluding mitigating operator actions in this
- - . Sanalysis'and allowing the events to progrese unchecked was to provide the added perspective of defining the time windows available for operator inter-vention. The results clearly demonstrate that the operator has an extensive time period to implement primary or alternative actions that will successfully terminate or mitigate postulated severe accidents.
The following subsections discuss plant response for each severe accident sequence analyzed. In these analyses the containment ultima te pressure capacity is based on the evaluation contained in the IDCOR Task 10.1 report [4.1], Containment Structure Capability of Light Water Nuclear Power Plants. The ultimate pressure capability was calculated to be 71.3 psia with the defined failure condition (twice the elastic strain) occurring at the
" transition" between the cylindrical and spherical parts of the containment.
(It should be noted that a detailed essessment of penetration behavior under high strain conditions was not part of the analysis.) A containment break size of 0.1 f t2 (1.5 ft2 for TC) is assumed because it permits depressurization of containment enabling airborne fission products to be transported out the break. This assumption is consistent with l
DRAFT the concept of yield leading to rupture resulting in diminishing yield as the containment depressurizes. 4.1 Plant Response to the T j00V Accident 4.1.1 Sequence Description The Tj QUV accident is assumed to occur during full-power operation. It is initiated by a loss of off-site power event (Tj ). During the accident, all systems not automatically transferred to the emergency busses are assumed to be unavailable. Thus, both the main feedwater and main condenser systems are assumed to be unavailable (Event Q) for the entire accident. The accident also specifies that neither the high-pressure nor the low-pressure emergency core cooling systems (ECCS) are available at any time during the accident (EventsUandV,respectively). The faults in these makeup systems are taken to be such that the systems are unavailable in any of their modes of opera-tion. In addition, the control rod drive (CRD) flow to the reactor pressure vessel (RPV) is modeled as being lost due to the accident initiator. Thus, for this event, no water makeup to the RPV occurs; and, neither primary system nor containment heat removal is assumed available. All other plant systems, including emergency diesels, are modeled to be available. No credit is taken for any operator action other than to energize the containment igniter system at the accident initiation and to manually depressurize the vessel when the water level drops to the RPV Level 2 LOCA setpoint. The T)QUV base case accident chronology is provided on Table 4.1. 4.1.2 Primary System and Containment Response The loss of off-site puwer, the loss of feedwater, the turbine stop valve (TSV) closures, and the turbine bypass valve (TBV) closures are modeled to occur simultant.asly. Loss of off-site power and the TSV closures actuate a reactor scram which is modeled to bring the reactor suberitical by 7.8 sec. The core power remains at decay heat levels for the remainder of the event. The TSV and TSB closures cause a RPV pressure excursion which is relieved by thesafety/reliefvalves(SRV). Steam released from the RPV through the SRVs is routed to the suppression pool (SP), where it is quenched. By 95 sec,
DRAFT Table 4.1 GRAND GULF NUCLEAR STATION TjQUV - BASE CASE ACCIDENT CHRON0 LOGY Time Event 0.0 see Initiating Event: Loss of off-site power; Loss of main feedwater; TSV/TBV closures 7.8 sec Reactor scram completed 95 sec RPV Level 2 LOCA setpoint reached 26.0 min RPV Level 1 LOCA setpoint reached; Vessel depressurization manually initiated 26.5 min DW purge system actuates 28.0 min Qore begins to uncover
, 57.0 min
.. SPMU actuates 2.0 hr Fuel melting ~begins 2.35 hr High DW pressure LOCA setpoint reached 2.35 hr Core plate failure followed by vessel failure 47.0 hr Containment failure i 1
DRAFT sufficient RPY water inventory has been lost through the cycling SRVs to lower the RPV water level to the RPV Level 2 LOCA setpoint. At the Level 2 set-point, signals are automatically generated to trip off the recirculation pumps and to actuate the high pressure core spray (HPCS) and reactor core isolation cooling (RCIC) systems. Since both HPCS and RCIC are unavailable, the RPV water level continues to drop, reaching the RPV Level 1 LOCA setpoint at 26 min. At this point, it is assumed that manual depressurization of the vessel is initiated. In addition, pennissive signals are automatically generated for the drywell (DW) purge, the suppression pool makeup (SPMU), the low pressure core spray (LPCS), and low pressure coolant injection (LPCI) systems. The DW purge system is modeled to actuate af ter a programmed 30 sec delay. Then, the DW purge compressors pressurize the DW atmosphere to the 1.89 psig High DW Pressure LOCA setpoint by 2.35 hours into the event. The SPMU system actuates the upper containment pool dump following a programed 30 min delay. Since neither LPCS nor the LPCI are available, the RPV level continues to fall, and the core begins to uncover at 28 min. Temperatures in the uncovered fuel regions begin to rise, and begin to reach 2000*F in about .5 hour af ter core uncovery. The cladding oxidation rate increases rapidly above the 2000'F fuel temperature point. Oxidation of the Zircaloy cladding increases the fuel heatup rate and thus tends to promote further cladding oxidation. Cladding oxidation within a channel is limited, however, by refreezing of molten cladding in lower, cooler portions of the channel. The steam trying to enter the channel is diverted around the block-age, thus preventing further oxidation and hydrogen formation within the channel (seeRef.[4.2]). Hydrogen generated by the Zircaloy-stea:a reaction in the core is released to the wetwell via the cycling SRVs. The amount released in the vessel is insufficient to cause burning. The maximum release rate being approximately 0.05 lb/sec. Fuel melting is predicted to begin at 2.0 hr. Molten fuel is modeled to relocate to the core plate. By 2.35 hr, sufficient molten core material is accumulated on the core plate (20% of total) to cause it to fail. The core debris then flows to the bottom cf the RPV, initiates thennal attack of the vessel wall and fails the vessel at a welded penetration. Following vessel failure, the molten core debris is discharged onto the pedestal floor.
DRAFT Due to the depressurized state at the time of vessel failure, no core debris is dispersed from the pedestal onto the drywell floor. The discharge of molten core debris from the vessel is followed by the lower plenum water. Some of this water spills from the pedestal to the drywell. After vessel failure, about 50,000 lb of water remains in the lower downcomer region of the vessel. Following vessel failure, the pedestal and drywell volumes are filled with steam; and, the air in these compartments is exhausted through the SP vents into the outer containment compartments. The pressures and tempera-tures in the drywell and in the outer containment are shown on Figs. 4.1 through 4.4. Drywell leakage flow paths bypassing the suppression pool are modeled to plug with aerosols. These aerosols are released from the vessel when it fails and from the core-concrete interaction in the pedestal. All flow exiting the drywell to the outer containment is af terward forced to pass through the suppression pool. Within ten minutes af ter vessel failure, the core debris beri in the pedestal is cooled to below concrete ablation tempera-tures by the lower plenum water; an ablation depth of 0.3 f t is predicted to this point in the accident sequence. The core debris temperature and concrete penetration depth in the pedestal are provided on Figs. 4.5 and 4.6. The debris remains quenched until its blanket of water is boiled away, which occurs at 4.0 hr into the event. Corium within the pedestal re-heats, renew-ing its attack on the pedestal concrete floor and wall *, at about five hours. The thermal decomposition of the pedestal concrete floor and walls causes significant ablation (see Fig. 4.6), and produces large volumes of carbon dioxide and steam. As these two gases pass through the partially molten corium debris bed in the pedestal, they oxidize the zirconium in the bed to produce hydrogen gas and elemental carbon. The hydrogen production resulting from the core-concrete interaction in the pedestal raises the hydrogen concentration to ignitable levels; and within minutes af ter vessel failure, the igniters start the hydrogen burning. The igniters, which are powered by the emergency bus, provide for an almost continuous controlled burn-off of all combustible gases being evolved during the accident. The burnoff prevents the accumulation of high concentrations of combustible gases. By about 13 hr, 100% of the zirconium has been oxidized. About 1800 lb of i
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DRAFT hydrogen have been produced and burned to this point. Afterward, the endo-thermic reactions of elemental carbon with steam and with carbon dioxide begin in the corium debris bed, and hydrogen and carbon monoxide are evolved. At about 15 hours, when the oxygen concentration falls below a combustible level in all containment compartments, burning ceases. At about 38 hr, the corium inventory of elemental carbon is exhausted and combustible gas production ceases; steam and carbon dioxide gas production continues. A total of 3000 lb of hydrogen and 75,000 lb of carbon monoxide is calculated to be produced during this accident. Note, however, that less than 100 lb of the hydrogen came from in-vessel production. Primarily because the primary system was depressurized prior to vessel failure, debris did not disperse from the pedestal to the drywell. Consequently, the temperature and pressure in the drywell behave as shown in Figs. 4.1 and 4.2. Note the rapid pressure rise in the drywell af ter vessel failure to about 26 psia due to debris entering the pedestal. The drywell temperature rise following vessel failure is due to the corium/ concrete attack in the pedestal. At 47 hr into the event, the GGNS containment reaches 71.3 psia (see Fig. 4.3). The contaiment is assumed to fail at this pressure at a location just below the junction between the cylinder and the dome [4.1]*. The failura cause is overpressurization by noncondensable gases. A containment breach 2 area of 0.1 f t was selected for modeling the containment depressurization. For this containment failure size and location, the containment depressurizes to within about 0.5 psid of atmospheric pressure in about 10 hours. The suppression pool remains intact following the containment failure event. As can be seen in Fig. 4.7, the pool temperature is less than 200'F at the time of the containment failure. Note that the suppression pool remains subcooled throughout the accident. Appendix B includes additional plots of results for this sequence.
- This is consistent with the analyses reported in Ref. [4.1] which only addressed the ultimate capacity. Consequently, failure modes were not addressed, specifically the effects on penetrations under large deflections.
\
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4.1.3 Manual Depressurization Sensitivity Analysis DRAFT In order to assess the sensitivity of the accident response to the assumption of manual vessel depressurization prior to vessel failure, the Tj QUV accident scenario was reanalyzed without vessel depressurization. No major variations in the sequence resulted, although some of the details differed. Key differences between this analysis and the base case are shown in Table 4.2. For the most part, differences from the base case prior to vessel failure are small, and are due core degradation occurring at an elevated or reduced pressure. The only significant differences are the longer time to vessel failure, the increased in-vessel hydrogen production, and the higher primary system gas temperatures. The first two are due to the slower boiloff of primary system water, and the latter is due to the higher hydrogen genera-tion rates. Following vessel failure, most of the molten core debris exiting the vessel is dispersed from the pedestal to the drywell, in contrast. No such dispersion occurs into the base case. Despite this difference. Table 4.2 shows that the difference in drywell pressurization from the dispersal is not large between the two cases. Since the core debris in the drywell is well-dispersed, the heat losses are too large for the debris to reach concrete ablation temperatures. The gas and structural temperatures in the drywell rise more quickly than in the base case, however. There is less concrete attack in the pedestal than in the base case due to the smaller corium inventory in the pedestal. This results in a slower ablation rate, less noncondensable gas generation, and a longer time to containment failure. In sumary, while there are minor differences in the accident progression, these would not substantially alter the overall accident response. l l l l 1
MAFT Table 4.2 EFFECTS OF DEPRESSURIZATION IN THE T jQUV ACCIDENT Depressurization No Depressurization Quantity at 0.43 hr Until Vessel Failure Core Uncovery Time, hr 0.47 0.62 Vessel Failure Time, hr 2.35 3.4 Containment Failure Time, hr 47.0 60 In-Vessel Hydrogen 10 430 Production, lb Mass of Core Debris in Dry-well Following Vessel 0 48,000 l Failure, lb l l Pressure in Drywell Follow- 26- 45 ing Vessel 7ailure, psia *. Gas Temperature in Drywell at ' Vessel Failure, 'F . - 370 550 Concrete Ablation in 7.6 Pedestal at 50 hr 7.2 Total Hydrogen Produced, lb 3,000 3.200 Total Carbon Monoxide 75,000 66,000 Produced, Ib l l l ss m -
4.2 Plant Response to the AE Accident DRAFT 4.2.1 Sequence Description The AE accident is assumed to occur during full-power operation. This accident is a large-break loss of coolant accident (LOCA). It is initi-ated by a 3.14 f 2t liquid line break (Event A) in the suction side of the recirculation loop. The accident sequence specifies that neither the high-pressure nor the low-pressure emergency core cooling systems (ECCS) are available at any time during the accident (Event E). The faults in these makeup systems are taken to be such that the systems are unavailable in any of their modes of operation. Thus, for this event, the only water makeup to the reactor pressure vessel (RPV) is due to the control rod drive (CRD) flow; neither the primary system nor containment heat removal is assumed available. All other plant systems are modeled to be available. No credit is taken for any operator action other than to start the containment igniter system at the accident initiation. The AE accident chronology is provided on Table 4.3. 4.2.2 Primary System and Containment Response The loss of coolant through the primary system break causes a rapid depressurization of the RPV and a rapid pressurization of the drywell (DW). The DW pressure reaches the 1.73 psig and 1.89 psig high drywell pressure LOCA setpoints by 0.2 sec into the accident. The former generates a reactor scram signal; the latter generates actuation signals for the high pressure core spray (HPCS), the low pressure core spray (LPCS), and the low pressure coolant injection (LPCI) systems. The reactor scram is modeled to bring the reactor subcritical by 3.9 sec. The core power remains at decay heat levels for the remainder of the event. Since the HPCS, LPCS and LPCI are assumed unavail-able, the RPV l'evel drops to the RPV Level 2 LOCA setpoint. At this point, 5.2 sec into the event, the recirculation pumps are signaled to trip off and the reactor core isolation cooling (RCIC) system is signaled to start. The recirculation pump trips are comphttd by 5.6 sec; RCIC is assumed unavail-able. The RPV water level falls to the RPV Level 1 LOCA setpoint at 6.5 sec. At this point, the main feedwater system trips off and the main steam isola-tion valves (MSIV) close. In addition, permissive signals are generated for l
DRAFT Table 4.3 GRAND GULF NUCLEAR STATION AE BASE CASE ACCIDENT CHRONOLOGY Time Event 0.0 see Initiating Event: A large break in suction side of a recirculation loop 0.2 sec High DW pressure LOCA setpoints reached 3.9 sec Reactor scram completed 5.2 sec RPV Level 2 LOCA setpoint reached
, 6.5 sec RPV Level 1 LOCA setpoint reached; MSIVs close; Main feedwater pumps trip off 45.0 sec . Core begins to uncover 11.6 min DW purge system actuates 30.4 min SPMU actuates 1.1 hr Feal melting begins 1.4 hr Core plate failure followed by vessel failure l 22.3 hr C3T drained and CRD flow to vessel ceases 58.0 hr Containment failure
DRAFT the suppression pool makeup (SPMU) and the drywell (DW) purge systems. The SPMU system releases the upper containment pool following a programed 30 minute delay. DW purge actuation is delayed until other pennissives are satisfied. Without sufficient water inventory makeup, the core begins to uncover at 45 sec. Temperatures in the uncovered fuel regions begin to rise and begin to reach 2000*F at about 13 min. The cladding oxidation rate increases rapidly above this point. Oxidation of the Zircaloy cladding, in turn, increases the fuel heatup rate and tends to promote further cladding oxida-tion. Since the boiloff time is short for the large-break LOCA response, in-vessel Zircaloy oxidation is minimal. Fuel melting is predicted to begin at 1.1 hr, and relocates to the core plate. By 1.4 hr, sufficient core material is calculated to have fallen onto the RPV core plate to cause it to fail. The core debris then falls to the bottom of the RPV and thirty seconds later, vessel failure occurs at a welded RpV penetration point. At vessel failure, the molten fraction of the lower plenum core debris falls onto the pedestal floor followed by the saturated lower plenum water. A small steam spike occurs at this point, causing a pressure rise in the pedestal and drywell to about 26 psia. Since the vessel was depressurized j prior to failure, no debris is dispersed from the pedestal to the drywell. , Drywell leakage flow paths bypassing the suppression pool are modeled to plug ! with aerosols. These aerosols are released from the vessel when it fails and ; from the core-concrete interaction in the pedestal. All flow exiting the drywell to the outer containment is afterward forced to pass through the suppression pool. The debris attacks the concrete until it is cooled below concrete j ablation temperatures by the lower plenum water at about two hours. The concrete is ablated to a depth of 2.5 inches up to this time. The remaining water in the pedestal, plus that continually added by the CR0 flow, is boiled away while slowly quenching the debris, as can be seen on Figs. 4.8 and 4.9. By about seven hours, the. debris and the water are at about the same
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AE - GRAND GULF 7- - 6_ . _ ;._ _.. . _ . . _
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- Fig. 4.9 Concrete ablation depth in the pedestal.
1 3lll== 1 4
1 4-22 temperature., From this point on, the continuing CRD flow into the pedestal refills it t6 the pedestal doorstep level. Excess water spills into the drywell . The CRD flow keeps the debris quenched until the CST runs out of water at 22.3 hours. Without replenishment, the pedestal water boils away and, by 26 hours the d is begins to reheat. Concrete ablation in the pedestal resumes at 30 hours. The thermal decomposition of the pedestal concrete floor and walls produces large volumes of carbon dioxide and steam. As these two gases pass through the partially molten corium debris bed in the pedestal, they oxidize the zirconium in the bed to produce hydrogen gas and elemental carbon. The igniters provide for an almost continuous controlled burn-off of all combusti-ble gases being evolved during the accident. The first burn begins at about 35 hours; thereafter, their ccntinuous burn-off prevents high concentrations of combustible gases from occurring. By 43 hr, 100% of the zirconium has been oxidized. At this point, the endothermic reactions of elemental carbon with steam and with carbon dioxide begin in the corium debris bed. Hydrogen and carbon monoxide are evolved in these reactions. At about 45 hours, when the oxygen concentration falls below a combustible level in all containment compartmants, burning ceases and the containment becomes self-inerted. Drywell temperatures rise to about 900*F af ter the core debris-concrete interaction resumes in the pedestal, as shown on Fig. 4.10. The j suppress ~ ion pool water temperature, shown on Fig. 4.11, reaches saturation due to the large amount of steam generated by quenching the debris in the pedestal prior to dryout. Temperatures in compartment 8 remain relatively low due to the cooling effect of the suppression pool (as shown in Fig. 4.12). l l At 58 hours into the event, the GGNS containment reaches 71.3 psia l 1 (see Fig. 4.13). The containment is modeled to fail at this pressure at a location just below the junction between the cylinder and the dome. The cause is overpressurization by steam and by noncondensable gases. A containment 2 breach area of 0.1 ft was selected for modeling the containment depressuriza-tion. For this containment failure size and location, the containment depres-surizes to within 0.5 psid of atmospheric pressure in about 10 hours. And, the suppression pool remains intact following the containment failure event.
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i l DRAFT Since the pool temperature is nearly 280'F at the time of the contairnent failure, about 2% of the pool inventory is calculated to boil away within 10 hrs following failure. Appendix B includes additional plots of results for this sequence. 4.3 Plant Response to the T,3QW Accident 4.3.1 Sequence Description The T23QW accident is assumed to occur during full-power operation. It is initiated by inadvertent main steam isolation valve (MSIV) closures (EventT23). The main feedwater and main condenser are assumed to be unavail-able (Event Q) for the entire accident. The accident sequence also specifies that containment heat removal is not available for the entire accident (Event W). Control rod drive (CRD) flow to the reactor pressure vessel (RPV) is , 1 modeled to be available. All other plant systems are assumed to be available. However, all emergency core cooling systems (ECCS) are assumed to fail on containment failure. No credit is taken for any operator action other than to start the containment igniter system at the accident initiation and to manual-ly depressurize the RPV when the suppression pool temperature exceeds 145'F. The T230W accident chronology is provided on Table 4.4. 4.3.2 Primary System and Containment Response The initiating event, which is inadvertent closure of the MSIV, causes a reactor pressure vessel (RPV) pressure excursion which is relieved by the safety relief valves (SRV). The exiting RPV steam is routed to the suppression pool (SP), where it is quenched. The MSIV closures actuate a reactor scram which is modeled to bring the reactor subcritical by 3.7 sec into the event. The core power remains at decay heat levels for the remainder of the event. At 2.35 hours into the accident the suppression pool tempera-ture exceeds 145'F and an operator intervention occurs to manually initiate l ADS. At 4.1 hr, steam pressurization of the containment building causes a high drywell (DW) pressure LOCA signal. This signal is a permissive signal l l l l
~
DRAFT _ Table 4.4 GRAND GULF NUCLEAR STATION T23 QW - BASE CASE ACCIDENT CHRONOLOGY Time Event 0 sec Initiating event: MSIV closures; Loss of main feedwater 3.7 sec Reactor scram completed 28 sec RPV Level 2 LOCA setpoint reached 1.0 min HPCS and RCIC systems begin operating 1.1 hr HPCS and RCIC systems transfer suction from CST to SP
, 2.35 hr . Suppression pool temperature exceeds 145*C, manual ADS 4.1 hr . High DW pressure LOCA setpoint reached; DW purge system actuates; LPCS and LPCI actuate (but can-not provide makeup without RPV depressurization) 4.6 hr SPMU actuates 6.3 hr RCIC pump fails on high suction temperature 22.4 hr CST empties 23.5 hr High wetwell pressure setpoint reached; Contain-ment sprays actuate 40.0 hr Containment failure; All ECCS assumed to fail 48.8 hr Core begins to uncover 54.1 hr Fuel melting begins 56.2 hr Core plate failure followed by vessel failure
DRAFT for the DW purge system, the SP makeup (SPMU) system, and the automatic i depressurization system (ADS); it is an actuation signal for the low pressure l core spray (LPCS) and low pressure coolant injection (LPCI) systems. The CW l purge system actuates after a 30 sec time delay and the SPMU system actuates ! the upper containment pool dump following a programed 30 min delay. The RPV water inventory is maintained by the HPCS and RCIC systems. The high DW pressure LOCA signal is modeled to switch the HPCS and RCIC systems' level control logic to maintain the RPV water level about the RPV Level 8 setpoint. Because of the assumed unavailability of containment cooling, the SP temperature rises during most of this event (Fig. 4.14). One exception to this trend occurs at 4.6 hr, when the SP makeup system releases relatively cold upper containment pool water into the SP. After the upper pool dump, the SP water temperature continues to rise again. When the SP temperature reaches 200*F at 6.3 hr, the RCIC pump is modeled to fail due to high bearing tempera-tures. After the loss of the RCIC, the HPCS and the CRD flow continue to maintain adequate RPV inventory. Driven by the steam produced in the core, the containment pressure reaches the 9 psig containment spray actuation pressure setpoint at 23.5 hr into the event. Note that the accident defini-tion assumes that the RHR heat exchangers are unavailable. Thus, the opera-tion of containment sprays removes no heat from the containment; it merely homogenizes temperatures in the outer containment. The effect of this homo-genization can be observed in Figs. 4.14, 4.15 and 4.16: the suppression pool temperature decreases, the outer containment air temperatures increase, and, consequently, the outer containment pressure increases slightly. The latter pushes water from the wetwell to the drywell side of the suppression pool and results in a large spill of suppression pool water onto the drywell and pedestal floors. This water plays a key role in quenching the core debris after the vessel fails. At that time, trains A and B of the residual heat removal (RHR) system automatically switch into their spray mode. At 22.4 hr the CST empties and the CRD flow ceases. From this point on, only the HPCS is available to maintain inventory. At 40 hr into the event, the GGNS containment pressure reaches 71.3 psia. The containment is modeled to fail at this pressure at a location just below the junction between the cylinder and the dome. The failure cause is
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steam overpressurization. A containment breach area of 0.1 ft was 2 MAFTmodeled. For this containment failure size and location, the containment depressurizes to within about 0.5 psid of atrnospheric pressure in about 10 hours. The suppression pool remains intact following the containment failure event. Suppression pool boiloff maintains an elevated containment pressure after the containment fails. Gas temperatures in all outer containment compartments are relatively constant at about 300*F after containment failure. The drywell air temperature is shown on Fig. 4.17. In order for the T QW sequence to result in core damage, it is 23 necessary that all systems supplying or capable of supplying water to the RPV fail at or before containment failure. A realistic mechanism which could cause such a simultaneous failure has not been identified. The accounting of containment failure location, pressure, fluid flow loading, and ECCS pump suction temperature [4.l], pressure, and NPSH limitations [4.2] indicates that at least one GGNS ECCS train should survive a containment failure event. However, for this analysis, the conservative assumption that all ECCS equip-ment fails on containment failure was made. Without vessel makeup, the RPV water level falls. The decrease is relatively slow in comparison with the T)QUV and AE events, since decay heat levels in the T QW23 accident are relatively low. Core uncovery takes place about 8 hours af ter containment failure, and fuel heatup begins thereaf ter. Fuel temperatures in the uncovered region of the core begin rising above 2000*F at 51 hr. The clad oxidation rate increases rapidly above the 2000*F fuel temperature point. Since the oxidation of the Zircaloy fuel cladding is an exothermic reaction, its occurrence increases the fuel heatup rate and thus tends to promote further cladding oxidation. About 5*, of the total Zircaloy was oxidized at vessel failure. Fuel melting is predicted to begin at about 54 hr into the event. After melting, the fuel moves to the core plate. By 56.2 hr, sufficient core material is calculated to have fallen onto the RPV core plate to cause it to i fail. The core debris then falls to the bottom of the RPV and, about 30 sec later, vessel failure occurs at a welded RPV penetration. At vessel failure,
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BRAFT the molten fraction of the lower plenum core debris falls onto the pedestal floor followed by the flashing high-pressure lower plenum water. Since the containment failure size was 0.1 ft ,2 the suppression pool remains saturated at about 280'F, passing the steam entering it through to the upper compartment. The containment pressure remains high, gradually diminish-ing as the heat load diminishes, as shown in Fig. 4.16. The gas temperatures in all of the containment compartments are relatively constant at about 300'F , , during the period of interest. The drywell temperature variation is shcwn on Fig. 4.17. Since the containment has such large amounts of steam, it is effec-tively inerted when the hydrogen leaving the vessel enters the wetwell (prior to vessel failure) and the drywell (after vessel failure). Hence, no burning is predicted to occur. For the same reason, any noncondensable gases that may be generated at very late times (beyond 100 hours) from core debris-concrete attack would not burn. The average corium temperature and penetration depth histories are shown in Figs. 4.18 and 4.19. Appendix B includes additional plots of results for this sequence. 4.4 Plant Response to the T.,3C Accident 4.4.1 Sequence Description The T23C accident is assumed to occur during full-power operation. It is initiated by inadvertent main steam isolation valve (MSIV) closures (EventT23). The accident sequence specifies that the control rod drive (CRD) system fails to automatically bring the reactor subcritical (Event C). This analysis assumes that no control rods were inserted into the core. All other plant systems are assumed to be available. No credit is taken for any opera-tor action other than to start the containmer.t igniter system at the accident initiation and to manually initiate ADS when the suppression pool temperature exceeds 145'F. The T23C at:.ident chronology is provided on Table 4.5. I I i
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i Fig. 4.18 Average corium temperature in the pedestal. 1
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"""'"I 1 __ _ _ _ _ _ _ _
DRAFT Table 4.5 GRAND GULF NUCLEAR STATION T C BASE CASE 23 ACCIDENT CHRONOLOGY Time Event
. O see . Initiating events: MSIV closures; Failure to scram; Loss of main feedwater 33 sec RPV Level 2 LOCA setpoint reached l 49 sec HPCS begins operating l 52 sec RCIC begins operating 4.5 min HPCS/RCIC systems transfer suction from CST to SP 8 min ADS manually initiated 18.3 min RCIC pump fails on high suction temperature !
23.0 min High OW pressure LOCA setpoint reached; Post-LOCA 3 ,- OW vacuum breakers open 23.6 min Drywell purge system actuates 23.8 min LPCS and LPCI actuate 26.2 min High wetwell pressure setpoint reached 33.8 min Containment sprays actuate 53.1 min SPMU actuates 1.0 hr Containment failure and subsequent ECCS failure 1.3 hr Core begins to uncover 3.0 hr Fuel melting begins 3.8 hr Core plate failure followed by vessel failure
4.4.2 Primary System and Containment Response DRAFT The MSIV closures are modeled to actuate a reactor scram which fails to insert the control rods into the core. Despite this failure to scram, the core power is modeled to decrease from its initial full-power level to about 20% of full power level within seconds. This power reduction simulates the thermal-hydraulic reactivity feedback effects which are expected to occur as a result of the initiating MSIV closure event, the resultant recirculation and feedwater trips, and the ensuing high pressure core spray (HPCS) and reactor core isolation cooling (RCIC) systems actuations. The estimate of 20% of full power is based on the assumption that the core power will equilibrate at a level which just equals the power needed to boil all incoming coolant flow. In addition, core power is assumed to linearly decrease from 18% to 6% of full power as the downcomer water level decreases from 7.2 ft above the active core to the top of the jet pumps. Decay heat power levels are assumed for un-covered fuel nodes. The T 23C core power history is provided in Fig. 4.20. The MSIV closures cause a reactor pressure vessel (RPV) pressure excursion which is relieved by the SRVs. The vessel remains at the SRV relief setpoint pressure. The exiting RPV steam is routed to the suppression pool (SP), where it is quenched. By 33 sec into the event, sufficient RPV water inventory has been lost through the cycling SRVs to drop the RPV water level to the RPV Level 2 LOCA setpoint. At that point, signals are automatically i generated to actuate the HPCS and RCIC systems. The HPCS begins injecting water into the RPV at 49 sec; the RCIC begins at 52 sec. These systems maintain RPV inventory between RPV Levels 2 and 8. At 4.5 min, suction for these systems is transferred from the condensate storage tank (CST) to the SP on a high SP water level signal. At 8 min, when the suppression pool tempera-ture reaches 145'F, the RPV is manually depressurized according to emergency procedure guidelines. Because the core power generation rate is much greater l than the decay heat level, the SP water temperature rises very rapidly. When the SP temperature reaches 200*F at 18.3 min, the RCIC pump is assumed to fail due to high bearing temperatures. The HPCS is unable to maintain sufficient RPV inventory at SRV setpoint pressures and at a 20% of full power level. As a result, the RPV water level decrease to a new equilibrium state. These can be seen in Fig. 4.21.
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4-42 The SP reaches saturation conditions and is no longer able to completely quench the steam exiting the RPV through the cycling SRVs; a steam-pressurization of the containment ensues. The rising suppression pool water temperature and the resulting rise in pressures and temperatures in both the drywell and outer containment can be'seen in Figs. 4.22 through 4.25. The rising pfessure actuates the 1.89 psig high DW pressure LOCA signal at 23.0 min. This signal is a permissive signal for the DW purge system, the post-LOCA DWiacuum breakers, and the SP makeup (SPMU) system; it is an actuation signal for the low pressure core spray (LPCS) and low pressure coolant injec-tion (LPCI) systems. Since the post-LOCA DW vacuum breaker permissive requir-ing a 0.87 psid drywell vacuum relative to the wetwell is already satisfied, the vacuum breakers open imediately. The DW purge system actuates after a 30 see time celay and the SPMU system actuates the upper containment pool dump following a programed 30 min delay. The continuing HPCS injection maintains RPV level. At 26.2 min into the event, the containment pressure reaches the 9 psig containment spray actuation pressure setpoint. At that time, trains A and 8 of the residual heat removal (RHR) system automatically switch into their spray mode and eight minutes later begin to spray SP water into the upper ' containment volume. Since the containment spray water cooling requires manual alignment, which was not modeled in this analysis, the containment spray system is unable to effect a coritainment pressure reduction. At 53.1 min into the event, the SPMU system releases, as designed, approximately half of the upper containment pool volume into the suppression pool. This action brings the suppression pool to a subcooled state. Conse-quently, the containment steam pressurization ceases and, in fact, reverses.
.The former is due to the renewed ability of the~ suppression pool to quench the SRV steam discharge. The rapid outer containment depressurization is due to the action of the containment sprays which draw suction from the suppression pool. Within 15 minutes of the upper pool release, the continued core power generation reheats the suppressiun pool to a saturated state and outer con-tainment prassurization resumes. The additional pool inventory begins to spill onto the drywell and pedestal floors at that time. This spill has a large mitigative effect if this accident proceeds beyond vessel failure. At i
1.0 hrs into the a'ccident, only minutes af ter the renewed oressurization, the containment is modeled to fail at this pressure at a failure location just 7
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DRAFT below the junction between the cylinder and the dome. The failure cause is 2 steam overpressurization. A containment breach of 1.5 ft was modeled. In order for the T C23 sequence to result in core damage, it is necessary that all systems supplying or capable of supplying water to the RPV fail at or before contairment failure. A realistic mechanism which could cause such a simultaneous failure has not been identified. The accounting of containment failure location, pressure, fluid flow loading, and ECCS pump suction temperature, pressure, and NPSH Ifmitations indicates that at least one GGNS ECCS train should survive a containment failure event. However, for this analysis, the conservative assumption that all ECCS equipment fails on containment failure was made. The CRD flow was assumed to continue, at the rate of approximately 90 gpm. Given that all ECCS fail on containment failure, the RPV water level begins to fall sharply as shown in Fig. 4.21. As the water level continues to fall, the power level decreases to 6% of full power. As a fuel node is uncovered, its power level is modeled to decrease to its decay heat level. Fuel temperatures in the uncovered regions of the core begin rising above 2000*F at about 1.9 hr. The oxidation of the Zircaloy fuel cladding by steam increases rapidly above the 2000*F point. About 530 lb of hydrogen is produced in the vessel. Fuel melting is predicted to begin at 3.0 hr. After melting, fuel moves from the core to the core plate. By 3.8 hr, sufficient core material is calculated to have fallen onto the RPV core plate to cause it to fail. The core debris then falls to the bottom of the RPV; shortly thereafter, the vessel fails at a welded penetration. At vessel failure, the molten fraction of the lower plenum core debris falls onto the pedestal floor followed by the lower plenum water. Since the vessel had been depressurized previously, the debris does not disperse from the pedestal to the drywell upon vessel failure. Further-more, the remainder of the core material gradually enters the pedestal from the vessel and also stays in the pedestal. The debris attacks the pedestal l
DRAFT concrete as.it is being quenched (see Fig. 4.26) until about three inches of concrete have been ablate'd. Once the core debris bed in the pedestal is cooled to below concrete ablation temperatures by the lower plenum water, it remains quenched since its blanket of water is boiled away. As can be seen from Fig. 4.27, this would not occur for a very long time, if ever. Conse-quently, no appreciable quantities of noncondensable gases are generated. Subsequent to vessel failure steam flows steadily from the pedestal, 6 to the drywell, to the suppression pool at a rate of roughly 2 x 10 ft3 /hr. The flow is due to the fact that the CRD water is continuing to quench the debris in the pedestal, and producing steam. No hydrogen burning was predicted to occur in this sequence. By the time the hydrogen produced from Zircaloy oxidation in the core reached the wetwell, all of the oxygen had been depleted from the wetwell atmosphere, as well as from the upper containment atmosphere. Furthermore, there are no appreciable quantities of hydrogen or carbon monoxide generated from core debris-concrete attack. Appendix 8 includes additional plots of results for this sequence, i 1 i l
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BRAFT Grand Gulf Section 5 to be supplied later. l 1 l
DRAFT 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION I 6.1 Introduction The phenomena of fission product release from the fuel matrix, its transport within the primary system, their release from the primary system into the containment, their deposition within the containment and the subse-quent release of some fission products from the containment are treated through the use of MAAP [6.1]. Release of fission products from the fuel matrix and their transport to the top of the core are treated by a subroutine in MAAP which is based on the FPRAT code [6.2]. Transport of fission products outside the core boundaries is determined by the natural and forced convection flows modeled in MAAP with the gravitational sedimentation described in Ref. [6.3] and other deposition processes described in Ref. [6.4]. Fission product behavior is considered for the best estimate transport, deposition and reloca-tion processes. Influence of surface reactions between chemically active substances like cesium hydroxide and other uncertainties are considered in subtask 23.4. The best estimate calculation, assuming cesium iodide and cesium hydroxide are the chemical state of cesium and iodine, is discussed below. 6.2 Modeling Approach Evaluations of the dominant chemical species in Ref. [6.5] show the states of the radionuclides (excluding noble gases) which dominate the public health risk to be cesium iodide and cesium hydroxide, tellurium oxide and strontium oxide. These and others are considered in the code when calculating the release of fission products from the fuel matrix. Vapors of these domi-nant species form dense aerosol clouds in the upper plenum, in some cases 3 approaching 100 g/m for a very short time, which agglomerate and settle onto surfaces. Depending upon the chemical compound and gas temperature, these deposited aerosols can be either solid or liquid. At the time of reactor I vessel failure, some material remains suspended as airborne aerosol or vapor and would be discharged from the primary system into the containment. The rate of discharge is determined by the gaseous flow between the primary system and containment which is sequence specific. (It should be noted that some
E DRAFT fission products can be discharged into the containment before vessel failure through relief valves or through breaks in the primary system. This is also sequence specific.) This set of inter-related processes are treated in MAAP and essentially result in a release of all airborne aerosol and vapor from the primary system into containment imediately following vessel failure. As a result of the dense aerosols formed when fission products are released from the fuel, considerable deposition occurs within the primary system prior to vessel failure. For some accident sequences, the primary system may be at an elevated pressure at the time of core slump and reactor vessel failure. Resuspension of these aerosol deposits during the primary system blowdown is assessed in Ref. [6.6] in terms of the available experi-mental results and basic models. It is concluded that resuspension immediate-ly following reactor vessel failure would not be significant, less than 1% of the deposited materials, even for depressurizations initiated from the nominal l operating pressure. For delayed containment failure, this small fraction of material is depleted by in-containment mechanisms.
'. T'herefore,'a major fraction of the volatile fission products are retained within the primary system following vessel failure, the distribution being determined by the MAAP calculations prior to vessel failure. Natural circulation through the primary system af ter vessel failure is analyzed using MAAP which allows for heat and mass transport in various nodes of the reactor vessel and the steam generators including heat losses from the primary system as dictated by the reflective insulation. Material transport is due to aerosols and vapors as governed by the heatup of structures due to radioactive decay of deposited fission products. This heatup is principally determined by the transport of cesium iodide and cesium hydroxide by the natural circulation flows. In this regard, the vapor pressure of cesium hydroxide is applied to both the cesium iodide and cesium hydroxide chemical species. In essence, this assumes that the solution of cesium iodide and cesium hydroxide has a vapor pressure close to that of cesium hydroxide, which is a conservatism in the calculations. In carrying out these calculations, the pressurization of l the primary system is dependent upon the pressurization of the containment and the heating within the primary system. These determine the in- and out-flows between the primary system and containment.
l
Deposition within the containment is calculated using thermal DRAE hydraulic conditions detennined by MAAP. The major aerosol sources are the releases prior to vessel failure (sequence specific), the airborne aerosols and vapors transferred from the primary system at the time of vessel failure, the subsequent releases from the primary system due to long term heatup, and concrete attack. At the time of containment failure, the remaining airborne aerosol and vapor can be released to the environment. Assessments of the potential for resuspension of deposited aerosols following containment failure [6.6] show this to be negligible. 6.3 Sequences Evaluated The use of MAAP in the manner indicated above leads to the release fractions shown in Tables 6.1 through 6.5. Four sequences are analyzed, including: transient with failure of injection (T;QUV); large LOCA with failure of injection (AE); transient followed by loss of containment heat removal (T23QW); and transient with failure to scram (T 23C ). Thermal-hydraulic behavior for these sequences is described in Section 4. In this section it is shown that, for T QW23 and T 23C , the containment fails before the core is uncovered. Hence, the cesium and iodine are still in the fuel matrix. 6.3.1 ,T,30VV Sequence As indicated in Table 6.1, two percent of the volatile fission product inventory is swept from the vessel to the suppression pool via the SRV lines prior to vessel failure. Of the remainder, 2% is still in the fuel matrix, 95% is in the upper plenum area,1% is in the downcomer. During the time between vessel breach and containment failure, revaporization and relocation of material within the primary system occurs, due to the continuing natural circulation flows. Some material continually flows to the pedestal and drywell as vapor, and from there some of the mate-rial flows to the suppression pool. Af ter about a day, the drywell is hot enough that revaporization begins there, and flow to the suppression pool is increased. The pool itself is highly effective in scrubbing the fission
~
DRAFT Table 6.1 DISTRIBUTION OF CsI IN PLANT AND ENVIRONMENT (FRACTION OF CORE INVENTORY) At Vessel Failure T23QW T23C AE T jQUV RPV .90 .68 .98 .98 Drywell 0.0 0.0 .02 0.0 Suppression Pool .10 .32 0.0 .02 Primary Containment 5.3 x 10 -5 2.2 x 10-5 0.0 0.0 Environment 3.2 x 10-5 2.6 x 10~4 0.0 0.0 At Containment Failure
, T23QW T23C AE T jQUV RPV 1.00 1.00 .91 46 Drywell 0.0 0.0 .03 .20 Suppression Pool 0.0 0.0 .06 .34 Primary Containment 0.0 0.0 0.0 0.0 Environment 0.0 0.0 0.0 0.0 Ultimate Distribution T23 QW T23C AE T jQUV RPV .50 .26 .90 .33 Drywell .12 .05 .03 .02 Suppression Pool .38 .69 .07 .645 Primary Containment 2.1 x 10-4 1.1 x 10~4 5.11 x 10~4 7.3 x 10-4 Environment 2.6 x 10-4 7.6 x 10~4 < 1 x 10-5 7.3 x 10-5 l
__J
Table 6.2 DRAFT T jQUV FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Compartment B. 237' 9" 2 Containment Failure Size .1 ft Fission Product Release Fraction Group to Environment Cs. I 7.3 x 10-5 Te, Sb 3.2 x 10-5 Sr. Ba < 1 x 10-5 Ru. No < 1 x 10-5 l l
DRMI Table 6.3 AE FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Compartment 8, 237' 9" 2 Containment Failure Size .1 ft Fission Product Release Fraction Group to Environment Cs, I < 1 x 10-5 Te, Sb 1.1 x 10-5 Sr. Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 l l i l 1 \
\
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Table 6.4 DRAFT T 23 QW FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Compartment 8, 237' 9" 2 Containment Failure Size .1 ft Fission Product Release Fraction Group to Environment Cs, ! 2.6 x 10-4 Te, Sb 2.2 x 10~4 Sr. Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 l 1
.m 6-8 MMEFT UliA Table 6.5 T C FISSION PRODUCT RELEASE 23 Assumptions Containment Failure Location - Compartment B, 237' 9" 2
Containment Failure Size - 1.5 ft Fission Product Release Fraction Group to Environment Cs. I 7.6 x 10-4 Te, Sb 7.5 x 10~4 S r ,~ Ba . - < 1 x 10-5 Ru, Mo < l x 10-5
DRAFT products. A decontamination factor of 600 is associated with passage from the drywell to the pool through the vents [6.7]. - Table 6.1 also shows the volatile fission product inventories in the various compartments at the time of containment failure. Only the airborne raterial in the upper compartment and that portion of the material still to be revolatized in the vessel that would not be scrubbed in the suppression pool is available for release to the environment. As can be seen in Table 6.2, the release fractions to the environment for this case are low. Long tenn re-leases subsequent to containment failure occur but at extremely slow rates. Considerable concrete ablation takes place in the pedestal following vessel failure and subsequent flowing of molten core debris into the pedestal. By 24 hr the ablation depth is more than 5 ft. 6.3.2 AE Sequence The use of MAAP leads to the release fractions shown in Tables 6.1 and 6.3. The thermal-hydraulic analysis is described in Section 4.2. Table 6.1 shows the distribution of cesium and iodine through the various regions, at vessel failure and 70 hr, when the calculation was termi-nated. Due to the very low steam flow in the vessel af ter the initial LOCA blowdown, nearly all of the material is initially deposited in the upper plenum. Hence, very little material enters the suppression pool through the break (less than 1 kg by the time of vessel breach). At the time of vessel breach, only about 1 kg is airborr.e. This material can leave the vessel. The deposited material (about 229 kg) remains in the vessel at this time. Following vessel failure, the remainder of the volatile fission products are released from the fuel as it melts. This material, along with that already deposited, moves around the vessel, being deposited, heating up, revaporizing, moving to cooler regions and redepositing, etc. Drywell pres-surization from the very hot gases in the pedestal cavity prevents materials from escaping the vessel until containment failure at 58 hr. As can be inferred from Table 6.1 about 1% of cesium and iodine are relocated from the l l l l
DRAFT vessel to the suppression pool during the period following containment fail-ure. Of this, only one part in 600 escapes the pool to the outer containment [6.7]. Release fractions to the environment are very low, as can be seen in Table 6.3. As for the T QUV j sequence, however considerable concrete ablation occurs, although it does not occur for the first 30 hr of the event. By 50 hr the ablation depth is approximately 5 ft. 6.3.3 T 0W Scquence 23 The use of MAAP leads to the release fractions shown in Tables 6.1 and 6.4. The thermal-hydraulic analysis was described in Section 4.3. i Table 6.1 shows the distribution of the volatile fission products l (cesium and iodine) through the various regions, at vessel failure and at 150 hr when the calculation was terminated. At vessel failure, nearly ell of the volatiles (90%) in the vessel are deposited in the upper structures. The remainder (10%) are in the suppression pool. Only negligible quantities are present elsewhere. The decontamination factor associated with passage through the SRVs and spargers, and subsequent pool scrubbing, is 1000 [6.7]. Since the containment is already failed prior to core uncovery there is no rapid depressurization as in the T QUVj and AE sequences. Furthermore, there is no large scale concrete attack in the pedestal. Thus the ultimate fission product distribution is such that the release to the environment is very small, as indicated in Table 6.4. 6.3.4 T C Sequence 22 The use of MAAP leads to the release fractions shown in Tables 6.1 j and 6.5. The MAAP thermal-hydraulic analysis is described in Section 4.4. i I Table 6.1 shows the distribution of cesium and fodine through the various regions both at vessel failure and at 50 hr, when the calculation was tenninated. At vessel failure 139 kg are deposited in the upper plenum,10 kg l l l l
DRAFT are in the downcomer,14 kg are in the core region, and 76 kg have left the vessel throurn the SRYs to the suppression pool. Only negligible , quantities are present elsewhere. The decontamination factor associated with passage through the SRVs and spargers is 1000[6.7]. The fission products tend not to exit the vessel but rather transfer their heat to gas and structures and move about the primary system. The reflective insulation is very effective in transferring a considerable portion of the heat to the drywe'l as temperatures rise. Since the containment is already failed prior to core uncovery there is no rapid depressurization. Furthermore, there is no large scale concrete attack in the pedestal. Thus the ultimate fission product distribution is such that the release to the environment is very small, as indicated in Table 6.5. 6.4 References 6.1 MAAP - Modular Accident Analysis Program, User's Manual, August, 1983. 6.2 IDCOR Technical Report 15.18. " Analysis of In-Vessel Core Melt Progression," Vol. IV (User's Manual) and Modeling Details for the Fission Product Release and Transport Code (FPRAT), September,1983. 6.3 Draft IDCOR Technical Report, "FAI Aerosol Correlation," July,1984 6.4 IDCOR Technical Report on Task 11.3, " Fission Product Tra:.., port in Degraded Core Accidents," December,1983. 6.5 IOCOR Technical Report on Tasks 11.1,11.4 and 11.5, " Estimation of Fission Product and Core-Material Source Characteristics," October, 1982. 6.6 IDCOR Technical Report on Task 11.6, "Resuspension of Deposited Aerosols Following Primary System or Containment Failure," July, 1984. 6.7 K. Holtzclaw, Personal Communication,1984. i
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7.0
SUMMARY
OF RESULTS DRAFT As outlined in Section 2 of this report, the IDCOR Subtask 23.1 Integrated Containment Analysis of the Grand Gulf Nuclear Station (GGNS) consisted of base case accident analyses w:1 operator action case accident analyses, . ,, The accident sequences selected for analysis represent a majority of previously-assessed risk and demonstrate a variety of initiating events, a variety of system failures combinations, and a diversity of accident phenomenology. The primary system and contatrinent thermal-hydraulic response analyses and fission product transport were per-formed via the MAAP code. Fission product release was performed via the FPRAT code which has been integrated into MAAP. Detailed descriptions of each of these analyses are provided in Sections 4 through 7 of this report, respec-tively. This section of the report sumarizes the major results of each of these analyses. 7.1 Base Case Analyses The base case analyses establish a reference system response during these accidents by assuming a minimum of operator intervention during the accident progression. As such, these analyses do not realistically account for the mitigative response of the trained operating staff and, thus, should not be considered as representative of realistic plant response analyses. 'The base case fission product transport results are sumarized on Table 7.1. A discussion of these results follows. Accidents involving demand-type failures of all automatically-actuated high and low pressure reactor pressure vessel (RPV) makeup systems, namely those accident sequences containing events UV or E, result in core damage unless an appropriate operator response is taken. For accidents which involve relatively small RPV coolant inventory loss rates and decay power levels, such as T QUV j and T23PQE, the core is predicted to begin to uncover within about half an hour of the initiating event. Within about one hour, significant fuel cladding degradation is predicted, and fuel melting is calculated to begin about two hours af ter the initiating event. Yessel
- lRD Table 7.1 SU M ARY OF FRACTIONAL RADIONUCLIDE RELEASES TO THE ENVIRONMENT Fission Product Group Accident Xe and Kr Cs and I Te Sr and Ba Ru and Mo T jQUV 1.0 7.3 E-5 3.2 E-5 < 1 x 10-5 < 1 x 10-0 AE 1.0 < 1 x 10-5 1.1 E-5 < 1 x 10-5 < 1 x 10-5 T23C 1.0 7.6 E-4 7.5 E-4 < 1 x 10-5 < 1 x 10-5 T23 QW 1.0 2.6 E-4 2.2 E-4 < 1 x 10-5 < 1 x 10-5 BWR-4 0.6 5.0 E-3* 4.0 E-3 6.0 E-4 6.0 E-4
- Iodine release fraction is 0.8 E-4. 7 Cesium release fraction is 5.0 E-3. "
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failure will follow within another half-hour. MAFT For accidents with large RPV inventory loss rates, such as AE, these events occur sooner. For the large-break LOCA case analyzed, the AE accident, fuel melting was predicted to occur within 0.7 hours of the initiating event and was closely followed by vessel failure. Accidents involving successful RPV makeup but inadequate containment cooling, such as T23QW and T 23C , will result in containment failure unless appropriate operator action is taken. Previous studies have postulated that all ECCS injection into the RPV will fail on containment failure. With this assumption, and without appropriate operator action, fuel melting will in-evitably follow. The results of this study indicate that the assumption that all ECCS equipment fails on containment failure has no mechanistic basis and thus is extremely conservative. Without the containment-failure-induced ECCS failure assumption, many of the previously-postulated dominant GGNS accidents sequences do not lead to core melt and, thus, can no longer be considered risk significant. The T QW23 and T C 23 sequences are all among these accidents. The mass of hydrogen produced via steam oxidation of fuel cladding in the core was calculated to be significantly lower than that prescribed by the NRC for interim rule on hydrogen control for Mark III containments. The MAAP predictions demonstrate that less than about 10% fuel cladding oxidation prior to fuel melting for severe GGNS accidents. The NRC cladding oxidation rule specifies a 75% cladding reaction. Even if the accidents were to pro-gress unmitigated to vessel failure, the maximum fraction of cladding oxidized is predicted at only 35%. Judicious misaction is necessary to generate cladding reactions of higher magnitudes. Specifically, a low vessel makeup flow or an orchestrated termination and restart of emergency core cooling would be necessary. The rate of hydrogen production calculated for the GGNS severe accident analyses is also substantially lower than those used in previous studies. The maximum average sustained rate observed in the PAAP calculations was less than 0.5 lb/sec lasting for about less than twenty minutes. l For accidents which proceed beyond vessel failure, the molten core debris is calculated to fall onto the pedestal floor. No core debris is
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DRAFT calculated to exit the pedestal volume. Thus, concrete attack is limited to the pedestal floor and walls. Without core-debris cooling, substantial erosion of the pedestal floor and walls is calculated to occur. Three containment failure modes were observed in the GGNS Mark III containment analysis. They were overpressurization by steam, by noncon-densable gases, and/or by hydrogen combustion. The dominant failure mode was found to be accident dependent. All three modes result in long-delayed containment failure events for the GGNS accidents analyzed, the MAAP code predicts no steam explosions large enough to fail either the reactor pressure vessel or the containment. Thus, no prompt containment failures were ob-served. It is noteworthy to state that the containment failure times pre-dicted in this study are long compared to those of previous studies. This is primarily due to the higher ultimate containment capacity (56.6 psig) used in this study. For the GGNS Mark III' design, the suppression pool was observed to
*9 "
the exert a dominant influence on the accident progression. There are a
' number of reason's that"the s'uppression pool displays this behavior. First, overpressurization of the containment by steam can occur only if the sup-pression pool is heated to high temperatures or if the suppression pool is by-passed. The former requires a substantial energy deposition and inadequate suppression pool heat removal. The latter has been evaluated to be a very low probability occurrence. Secondly, the suppression pool controls the tempera-ture of the noncondensable gases which are calculated to be evolved in se-quences heading to core degradation, core melting and core-concrete attack.
By cooling these gases, as they enter the outer containment volume, the suppression pool substantially slows the rate of pressurization within the containment building. Thirdly, for accident sequences which have proceeded past vessel failure, the suppression pool water can, in general, be supplied to the debris to provide either temporary or potentially long term debris bed cooling. t.astly, it is significant to recognize that the suppression pool can retain substantial quantities of noninert fission product material which would be released by the fuel during a core meltdown event. With the location of the suppression pool in the Mark III design, these materials cannot be
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DRAFT exhausted through a containment breach without first being highly decon-taminated by the suppression pool. Fission product release and transport calculations were performed with FPRAT and MAAP for the Tj QUV, T23QW, AE, and T23C base case sequences. A sumary of the final airborne fission product releases to the environment for the accident sequences analyzed are presented in Table 7.1. The BWR-4 release category from the Reactor Safety Study is also presented for comparison. The data presented on this table shows that for the accidents analyzed the frac-tional fission product releases to the environment were generally significant-ly less severe than those associated with the BWR-4 release category. Since the accidents analyzed represent a majority of public health risk, the present analysis indicates that the risk associated with the operation of GGNS is substantially lower than that previously assessed. The lower fission product release terms produced in this study as compared to previous studies are principally due to the higher suppression pool decontamination factor and the relatively late containment failure time. Other factors which were found to influence the amount of fission product escaping the containment system during the severe accident scenarios analyzed were the duration of the melt releases, the time of the vessel failure, the fission product transport pathway, and the assumed fraction of fission product resuspension at the time of containment failure. A specific finding of these analyses is that accidents which involve rapid core heatups or which display a high RPV pressure until the vessel failure result in rapid releases of vola-tile fission products from the fuel imediately af ter the vessel fails. Another finding is that nonvolatile fission product release rates due to core-concrete interaction are small beyond about 20 hours af ter vessel fail-ure. Lastly, the majority of fission product retention was calculated to occur in the suppression pool and in the drywell. 7.2 Operator Action Analyses The major results of the operator action case thermal-hydraulics analyses are sumarized in Section 5. They demonstrate that a safe stable state can be achieved in the vessel if injection can be restored prior to core l
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MAFT plate failure , There are many means available to the operator for providing sufficient makeup flow to the reactor vessel. The time available for aligning and actuating these RPV makeup systems prior to core damage and/or fuel melting was evaluated in the base case analyses to be accident dependent. Once actuated, the operator case analyses indicate that these systems are capable of reflooding the core within minutes. These analyses also demon-strate that given the existence of a safe stable state for the core, a safe stable state for the containment can be achieved by restoring adequate con-tainment cooling. Peak containment temperatures and pressures occur from minutes to hours after such restoration, depending on the core heat level and on the mode and magnitude of containment heat removal. Debris coolability and the maintenance of containment integrity was demonstrated as possible via the restoration of an emergency core cooling system to flood the pedestal and a containment cooling system to cool the suppression pool. Sh a O 4
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8.0 CONCLUSION
S DRAFT Based on the results of the severe accident analyses performed in this study, a number of conclusions can be drawn regarding the progression and consequences of such severe accidents for plant designs similar to that of the Grand Gulf Nuclear Station. The analytical tools employed in this study, namely MAAP. is a viable means of analyzing both the thennal-hydraulic and the radiological response of the Grand Gulf Nuclear Station primary system and containment to severe accident scenarios. The most significant conclusions which can be drawn from this integrated containment analysis of the Grand Gulf Nuclear Station are itemized below. The first refers to the analytical tools used in this study. The next set are thermal-hydraulic related conclusions. And, the last and probably most significant conclusion relates to the radiological results of this study. e The MAAP code is a viable means of analyzing both the thermal-hydraulic and the radiological response of the Grand Gulf Nuclear Station primary system and containment to severe accident scenarios. e For accidents postulated to lead to core damage, fuel melting, and/or containnent failure, there are sufficient time and means available to the operating staff to place the plant into a safe stable state. e Containment failure should no longer be considered a cause for the l failure of all ECCS flow to the reactor vessel. Thus, containment I failure should no longer be considered a cause for core melt. e The mass and rate of hydrogen calculated to be produced in the vessel prior to fuel melting is substantially less than that pre-dicted by previous studies.
w c v -
,.) '
8-2 j e If successful fuel cooling is delayed beyond the point of signifi-
' cast core damage and/or vessel failure, the core debris coolability l is possible.
a ,
, The suppression pool exerts a dominant thermal-hydraulic and radio- I
' logical influence on the containment response to a severe accident.
e The GGNS Mark III containment failure modas are overpressurization via steam, noncondensable gas generation, and/or hydrogen combus-tion. Containment failure times are long compared to previous studies. No prompt containment failures due to steam explosions or steam spiking were calculated. e The overall containment response is much more sensitive to whether continuous hydrogen combustion occurs than to the details of how incomplete combustion progresses within the containment.
' ,e
- Through continuous burning of the containment combustible gas, the CGNS containment hydrogen igniters can significantly delay contain-ment failure during a severe accident.
e Decontamination of the fission product releases by the suppression pool and their condensation and gravitational settling in the drywell were found to be the two most important fission product rcoval mechanisms, e The public health consequences of the severe accidents are substan-tially less than those of previous assessments.
A-1 APPENDIX A = " Grand Gulf Parameter File GULEEP.DAT;14 6-JUL-1984 14:29 Pa9e 1 AAMARK !!! BWR PLANT PARAMETER VALUES-- TYPICAL OF GRAND GULF ! (M-KG-SEC-DEGK) gSj, ! AA I APRIMARY SYSTEM PS 01 13.52100 AFLCOR FLOW AREA 0F REACTOR CORE PS AA ALSH IS CALCULATED af TAKING THE VOLUME OF WATER IN THE LOWER 5 L WA bWER Obb PS 0 5.69500 AFLBYP CORE BYPASS FLOW AREA PS AA AUSH IS CALCULATED BY TAKING THE VOLUME OF WATER IN THE UPPER AA 00WNCOMER AB0VE T0AF AND DIVIDING BY THE WATER HEIGHT AB0VE TOAF 04 2.644D1 AUSH FLOW AREA IN U?PER SHkOUD PS 05 1.11605 HCRD SPECIFIC ENTHALPY OF FLOW IN CRD TUBES PS 06 9.248D5 HEW SPECIFIC ENTHALPf 0F FEEDWATER PS 07 1.65697D5 MU2PS TOTAL MASS OF 002 IN CORE PS 08 8.0D2 NASS NUMBER OF FUEL ASSEMBLIES IN REACTOR CORE PS 09 6.201 NPINS NUMBER OF FUEL RODS IN A FUEL ASSEMBLY PS 10 1.9302 NCED NUMBER OF CRD TUBES PS 11 4.500 NGEPS SENSIBLE ENERGY STORED IN FUEL (FULL POWER SECONDS)PS 12 4.000 TDMSIV DELAY TIME FOR MSIV CLOSURE PS 13 3.500 TDSCRM DELAY TIME FOR FULL SCRAM PS 14 5.976D7 TIRRAD TOTAL EFFECTIVE IRRADIATION TIME FOR CORE PS AA TS bl
. 2D-2 Vbb k k. k hPh W !D hk
"S l? 1:lli:! tO!!B1 ElB!!8 lill 11 13 1: lib:1 Pviill ElB It8:lill 11 21 1.12D-2 WVCEDI CRD FLOW RATE PS 33 24 1.ARD3 Wil" PFPr5WRNMP 1.0134D5 PCRD PPS FOR CRD PUMP fl PS 25 1.013405 PCRD PPS FOR CRD PUMP PS 26 1.013405 PCRD PPS FOR CRD PUMP PS 27 1.013405 PCRD PPS FOR CRD PUMP PS 28 1.0134D5 PCRD PPS FOR CRD PUMP PS 29 1.0134D5 PCRD PPS FOR CRD PUMP PS 30 1.013405 PCRD PPS FOR CRD PUMP PS 31 3.333D3 WFWMAX MAXIMUM FEEDWATER FLOW RATE (RUN OUT FLOW) PS 32 6.8502 WBPMAX MAXIMUM TURBINE BYPASS FLOW RATE PS 33 1.63D-1 NXCORE EXIT CORE QUALITY AT TIME ZERO PS 34 5 XDCORE REACTOR CORE DIAMETER TO INNER SHROUD WALL PS 35 2 26400 20601 XHRV INTERIOR HEIGHT OF REACTOR VESSEL PS 36 3.18800 XRRV INTERIOR RADIUS OF REACTOR VESSEL PS 37 41.0100 ZBJET ELEVATION AT BOTTOM OF JET PUMPS PS
. ZB E A 0!!$ S A hARATORS S 40 37.4100 ZBV ELEVATION AT BOTTOM OF REACTOR VESSEL PS 41 42.73D0 ZCPL ELEVATION AT CORE PLATE PS 42 45.4800 ZTJET ELEVATION AT TOP OF JET PUMPS PS l 43 1.3300 AJET TOTAL AREA 0F JET PUMPS PS <
44 46.7400 ZT0AF ELEVATION AT TOP OF ACTIVE FUEL PS I 45 52.6500 ZTSFP ELEVATION AT TOP OF STEAM SEPARATORS PS 4 51*9100 ZWNORM ELEVAfl0N AT NORMAL SHROUD WATER LEVEL PS 4h 41 0200 ZLOCA ELEVATION AT BREAK PS 48 .2919D0 ALOCA AREA 0F BREAK PS 49 52.32500 ZWL8 ELEVATION AT LEVEL 8 TRIP PS 50 0.000 NOT USED
DRAFT . GULEEP.DAT;14 6-JUL-1984 14:29 Page 2 51 51.26D0 ZSCRAM LOW WATER LEVEL SCRAM PS 52 7.4435D6 PSCRAM HIGH PRESSURE SCRAM SETPOINT PS 53 .2000 EGATWS ATWS CONSTANT POWER ASSUMPTION PS 54 1.2D3 IDSLC TIME FOR SCRAM WITH SLC PS PS 55 0.00 TIRR(1) TIME VS. ERACTION DE TOTAL ELOW EOR RECIRC PUMP PS , 56 2.D0 TIRR(2) TIRR(3) PS 57 4.00 PS 58 6.D0 TIRR(4) 0 0 ffRR ) PS 61 15.01200 TIRR(7) PS 62 40.D0 TIRR(8) 65 .45D0 D0 h2 EWRR(3) PS 66 .3000 EWRR(4) PS 67 .2000 EWRR(5) PS 68 .13500- EWRR(6) PS 69 .05000 EWRR(7) PS 70 0.D0 EWRR(8) PS h.*b Pkb(1) RS T E R SLC ELOW CURVE h5 73 7.93D6 PSLC(2) PS 74 7.9306 PSLC(3) PS 75 7.9306 PSLC(4) PS 76 7.93D6 PSLC(5) PS 77 7.93D6 PSLC(6) PS 78 7.93D6 PSLC(7) PS 79 7.93D6 PSLC(8) PS 80 2.713D-3 WVSLC(1) SLC ELOW RATE AT PSLC(1) -- M3/S PS 81 2.713D-3 WVSLC(2) PS 82 2.713D-3 WVSLC(3) PS 83 2.713D-3 WVSLC(4) PS 84 2.713D-3 WVSLC(5) PS 85 2.713D-3 WVSLC(6) PS 86 2.713D-3 WVSLC(7) PS 87 2.713D-3 WVSLC(8) PS 88 .2D0 TDRPI DELAY TIME FOR RECIRC PUMP TRIP PS 89 47.16D0 ZLMSIV LOW WATER LEVEL E0R MSIV CLOSURE PS 90 49.9200 ZLRPT LOW WATER LEVEL FOR RECIRC PUMP TRIP PS 91 7.858DG PHRPT HIGH VESSEL PRESSURE FOR RECIRC PUMP TRIP PS 92 1.132705 PDWSCM HIGH DRYWELL PRESSURE EOR SCRAM PS 93 .03200 EENRCH NORMAL EUEL ENRICHMENT PS 94 2.D4 EXPO AVERACE BURNUP IN P.WD/ TONNE PS h I h0 hE h hI I R ORNk0 3 0 L SSION S 97 5.D-1 E0ER1 FISSION POWER ERACTIJN OE U235 AND PU241 PS 98 4.2D-1 EQER2 EISSION POWER ERACTION DE PU239 PS 99 8.D-2 EQER3 EISSION POWER ERACTION OE U238 PS 100 .305100 XPCRDT PITCH OF CRD TUBES PS 101 .275500 XDCRDI OUTER DIAMETER DE CRD TUBE PS 102 58.00 NINSI NUMBER DE INSTRUMENT TUBES PS 103 .004400 XTHCRD THICKNESS DE CRD TUBE WALL PS 104 .050000 XDINST OUTER DIAMETER DE INSTRUMENT TUBE PS 105 .0818D0 XDRIVE LOWER CRD DRIVE OUTER DIAMETER PS 106 1.0040-3 VWCRD SPECIFIC VOLUME DE CRD WATER PS 107 1.0040-3 VWCST SPECIE!C VOLUME OF SLC WATER PS 108 5.8167D4 MEOPS MASS DE UPPER PLENUM HEAT SINK PS 109 1.016D3 AEOPS SUREACE AREA 0F UPPER PLENUM HEAT SINK PS 110 .24100 XTRV THICKNESS OF LOWER VESSEL HEAD PS 111 0.D0 IIEWCD TIME SINCE MSIV CLOSURE SIGNAL VS. EEEDWATER PS
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DRAFT GUttfP.DAf;14 6-JUL-1984 14:29 Page 3 112 0.D0 COASTDOWN MASS ELOW RATE PS 113 0.D0 PS 114 0.D0 PS 115 0.D0 PS 119 8:88 IS 119 0.D0 PS 119 0.00 WEWCD PS 120 0.D0 PS 121 0.D0 PS 122 0.00 PS 123 0.D0 PS 124 0.D0 PS 125 0.00 PS 126 0.D0 PS 127 5.86D6 PLMSIV LOW RPV PRESSURE FOR MSIV CLOSURE PS 128 53.900 ZMSL ELAVATION AT CENTER LINE DE MAIN STEAM LINE PS AA HE ACIRC C1 0.D0 ACSHS(1) CORE + LOWER PLENUM AA CARBON STEEL-HEAT SINK HEAT TRANSEER AREA 02 140.00 ACSHS(2) UPPER PLENUM 03 0.D0 ACSHS(3) DOWNCOMER 04 0.D0 ACSHS(4) 05 0.00 ACSHS(5) 06 50.D3 MCS(1) CORE + LOWER PLENUM CARBON STEEL MASS 07 100.D3 MCS(2) UPPER PLENUM 08 350.D3 MCS(3) DOWNCOMER 09 0.00 MCS(4) 10 0.D0 MCS(5) 11 0.D0 MHS(1) CORE + LOWER PLENUM HEAT SINK MASS 12 100.D3 MHS(2) UPPER PLENUM 13 0.00 MHS(3) DOWNCOMER 14 0.00 MHS(4) 15 0.00 MHS(5) 16 0.D0 ACSX(1) CORE + LOWER PLENUM CARBON STEEL TO DRYWELL AA HEAT TRANSEER ARE 17 0.D0 ACSX(2) UPPER PLENUM 18 240.D0 ACSX(3) DOWNCOMER 19 0.D0 ACSX(4) 20 0.00 ACSX(5) 21 0.D0 AHSX(1) CORE + LOWER PLENUM HEAT SINK TO DRYWELL AA HEAT TRANSEER AREA 22 140.D0 AHSX(2) UPPER PLENUM 23 0.00 AHSX(3) 00WNCOMER 24 0.D0 AHSX(4) 25 0.00 AHSX(5) 26 100.00 AGCS(1) CORE + LOWER PLENUM GAS TO CARBON STEEL N 28 5.D3 240.D0 AGCS(2) NR AGCS(3) DOWNCOMER 29 0.00 AGCS(4) 30 0.00 AGCS(5) 31 0.D0 AGHS(1) CORE + LOWER PLENUM GAS TO HEAT SINK AA HEAT TRANSEER AREA 32 140,DO AGHS(2) UPPER PLENUM 33 0.D0 AGHS(3) DOWNCOMER 34 0.D0 AGHS(4) 35 0.00 AGHS(5) 36 8.000 XL(1) CORE + LOWER PLENUM LENGTH 37 5.00 XL(2) UPPER PLENUM LENGTH
DRAFT 60LIEP.Def;14 6-JUL-1984 14:29 Page 4 38 10.D0 XL(3) DOWNCOMER LENGTH 39 0.D0 XL(4) 40 0.00 XL(5) 41 11.00 AG(1) CORE + LOWER PLENUM FLOW AREA 42 11.00 AG(2) UPPER PLENUM FLOW AREA 43 10.D0 AG(3) DOWNCOMER ELOW AREA . 44 0.D0 AG(4) HYDRAULIC DIAMETER FOR CORE REGION 47 .1500 DH(2) HYDRAULIC DIAMETER FOR UPPER PLENUM 48 .4D0 DH(3) HYDRAULIC DIAMETER FOR DOWNCOMER 49 0.00 DH(4) 5f ($*N bf) RPV CONVECTION LOSSES AT TIME ZERO 52 8.00 EINPLT NUMBER OF LAYERS IN REELECTIVE INSULATION AA HE AHEATUP HE 01 3.8100 XZFUEL LENGTH OF ACTIVE FUEL HE 02 5.210-3 XRFUEL RADIUS DE FUEL PELLET HE 03 8.130-4 XTCLAD THICKNESS OF CLADDING HE 04 5.033D4 MZRCAN TOTAL MASS OF 2R IN ASSEMBLY CAN HE
- 3. D-3 AN A H K AA NODE 1,1 IS BOTTOM-CENTER 1 10 IS TOP-CENTER 21 IS SECOND RADIAL AA RING 07 08 OUT FROM 6.6800-1 FPEAK(1 1) PEAKINGCENTER EACTOR FOR AT TEE NODE 1 1)BOTTOM OE HE THE CORK (ET 7.9100-1 EPEAK(2 1) PEAKING EAC70R FOR NODE (2 1) HE 09 4.7000-1 EPEAK(3 1) PEAKING FACTOR FOR N0DE (3 1) HE 15 1.14500 EPEAK(1 2) PEAKING FACTOR EOR NODE (1 2) HE 16 1.466D0 EPEAK(2 2) PEAKING FACTOR FOR NODE (2,'2) HE 17 9.290D-1 EPEAK(3 2) PEAKING FACTOR FOR N006 (3 HE 23 1.019D0 EPEAK(1 3) PEAKING FACTOR FOR NODE (1,2) HE 24 1.34300 EPEAK(2 3) PEAKING FACTOR E0R3)NODE (2'3) HE 25 8.9600-1 FPEAK(3 3) PEAKING EACTOR FOR NODE (3,3) HE 31 1.029D0 EPEAK(1 4) PEAKING EACTOR E0R NODE (1 4) HE 32 1.28100 EPEAK(2 4) PEAKING FACTOR FOR NODE (2 4) HE 33 8.670-1 FPEAK(3 4) PEAKING EACTOR FOR NODE (3 4) HE 39 1.22300 FPEAK(1,'5) PEAKING EACTOR E0R N0DE (1,'5) HE 40 1.414D0 EPEAK(2 5) PEAKING EACTOR EOR NODE (2 5) HE 41 9.430D-1 FPEAK(3,'5) PEAKING FACTOR E02 N0DE (3,'5) HE 47 1.23500 l
FPEAK(1 6) PEAKING FACTOR FOR N0DE (1 6) HE 48 1.373DO EPEAK(2 6) PEAKING EACTOR FOR N0DE (2,,6) HE 49 9.03D-1 FPEAK(3 6) PEAKING EACTOR FOR NODE (3 6) HE 55 1.198D0 EPEAK(1 7) PEAKING EACTOR FOR NODE (1 7) HE 56 1.26900 EPEAK(2 7) PEAKING FACTOR FOR NODE (2 7) HE 57 8.09D-1 FPEAK(3 7) PEAKING FACTOR E0R N0DE (3 7) HE 63 1.23500 EPEAK(1 8) PEAKING EACTOR EOR N0DE (1 8) HE 64 1.243D0 EPEAK(2 8) PEAKING FACTOR FOR N0DE (2 8) HS 65 7.110-1 FPEAK(3 8) PEAKING FACTOR FOR NODE (3 8) HE 71 1.33100 EPEAK(1 9) PEAKING EACTOR FOR NODE (1 9) HE 72 1.10700 EPEAK(2 9) PEAKING EACTOR FOR NODE (2 9) HE 73 5.53D-1 EPEAK(3 9) PEAKING EACTOR FOR NODE (3,9) HE 79 7.400-1 FPEAK(1 10) PEAKING FACTOR FOR NODE (1 10) HE 80 5.64D-1 EPEAK(2,'10) PEAKING EACTOR EOR NODE (2,'10) HE 81 2.69D-1 10) HE 87 0.3D0 FPEAKt3,10) XCHIM UNHEATEDPEAKING FACTOR EUEL LENGTH FOROFNODE (3, CORE AT TOP HE g 1.D-7 XIZROX INITIAL CLADDING OXIDE THICKNESS g AA ES AENGINEERED SAFEGUARDS ES 01 1.00 HLPCII NUMBER DE LPCI PUMPS IN LOOP 1 ES
DRAFT GULITP.DAf;14 6-JUL-1984 14:29 Pa9eh 02 1.0D0 NLPCI2 NUMBER OF LPCI PUMPS IN LOOP 2 ES 03 1.0D0 NLPCI3 NUMBER OF LPCI PUMPS IN LOOP 3 (INJECTION ONLY) ES 04 1.D0 NLPCSP NUMBER OF LPCS PUMPS ES 05 0.0D0 NOT USED 06 1.4D1 VMNCSI MIN. WATER VOLUME IN CONDENSATE STORAGE TANK ES AA FOR HPCI AND RCIC SUCTION SWIICH OVER ES 07 1.008D-3 VWCST SPECIFIC VOLUME OF CST WATER ES AA ALL PUMP CURVES ARE ARRANGED SO THAT THE FIRST FLOW ENTRY CORRESPONDS AA TO THE FIRST PRESSURE ENTRY 24 2.0906 PLPCI(1) PUMP CURVES FOR ECCS -- LPCI ES 25 2.D6 PLPCI(2) PPS-PDW VS VOLUMETRIC FLOW ES 26 1.896D6 PLPCI(3) ES 27 1.64106 PLPCI(4) ES 28 1.462D6 PLPCI(5) ES 29 1.1651D6 PLPCI(6) ES 30 .84106 PLPCI(7) ES 31 .4964D6 PLPCI(8) ES 32 0.0D0 WVLPCI(1) ES 33 .126200 WVLPC I(2) ES 34 .189300 WVLPCI(3) ES 35 .3155D0 WVLPCI(4) ES 36 .378600 WVLPCI(5) ES 37 .4417D0 WVLPCI(6) ES 38 .504800 WVLPCI(7) ES 39 .564100 WVLPCI(8) ES 40 3.584D6 PLPCS(1) LPCS PUMP CURVE ES 41 3.378D6 PLPCS(2) ES 42 3.06D6 PLPCS(3) ES 43 2.889D6 PLPCS(4) ES 44 2.6706 PLPCS(5) ES 45 2.39206 PLPCS(6) ES 46 2.068D6 PLPCS(7) ES 8 .'ho WhbPS1) k 49 .126200 WVLPCS(2) ES 50 .252400 WVLPCS(3) ES 51 .315500 WVLPCS(4) ES 52 .3786D0 WVLPCS(5) ES 53 .441700 WVLPCS(6) ES 54 .504800 WVLPCS(7) ES 55 .574200 WVLPCS(8) ES 56 9.89206 PHPCS(1) HPCS PUMP CURVE ES 57 8.886D6 PHPCS(2) ES 58 7.521D6 PHPCS(3) ES 59 6.749D6 PHPCS(4) ES 60 5.667D6 PHPCS(5) ES 61 4.226D6 PHPCS(6) ES 62 2.296D6 PHPCS(7) ES 63 0.000 PHPCS(8) ES kk Ik200 hbSb ES 66 .252400 WVHPCS(3) ES 67 .315500 WVHPCS(4) ES 68 .3786D0 WVHPCS(5) ES 69 .441700 WVHPCS(6) ES 70 .5048D0 WVHPCS(7) ES 71 .5742D0 WVHPCS(8) ES 72 10.34106 PRCIC(1) RCIC PUMP CURVE ES 73 10.34006 PRCIC(2) ES 74 6.894D6 PRCIC(3) ES 75 3.447D6 PRCIC(4) ES
DRAFT GULEEP.DAT;14 6-JUL-1984 14:29 Pa9e 6 76 2.758D6 PRCIC(5) ES 77 2.068D6 PRCIC(6) ES 78 4.144D5 PRCIC(7) ES 79 4.137D5 PRCIC(8) ES 80 0.D0 WVRCIC(1) ES 81 .050500 WVRCIC(2) ES 82 .0505D0 WVRCIC(3) ES 83 .0505D0 WVRCIC(4) ES 84 .0505D0 WVRCIC(5) ES 85 .0505D0 WVRCIC(6) ES 86 .0505D0 WVRCIC(7) ES 87 0.D0 WVRCIC(8) ES N 6b f kH E R S E ffP0fNTFORHPCI !S 90 1.010 TDHPCI IIME DELAY FOR HPCI ES 91 1.D10 PHHPCI MINIMUM PRESSURE FOR HPCI TURBINE ES 92 49.9200 ZLHPCS LOW WATER INITIATION FOR HPCS ES 93 1.144D5 PSHPCS HIGH DRYWELL PRESSURE SET POINI FOR HPCS ES 94 27.00 TDHPCS TIME DELAY FOR HPCS ES 95 47.1600 ILLPCI LOW WATER INITIATION FOR LPCI ES g gD5 gg g gLgR RE SET POINT FOR LPCI g l 98 1.010 PLLPCI LOW VESSEL PRESSURE PERMISSIVE FOR LPCI ES I 99 47.16D0 ZLLPCS LOW WATER INITIATION FOR LPCS ES l 100 1.144D5 PSLPCS HIGH DRYWELL PRESSURE SET POINT FOR LPCS ES 101 37.D0 IDLPCS TIME DELAY FOR LPCS ES AA THE NEXT PARAMETER IS A LOCA PERMISSIVE SIGNAL AND IF ONE DOES NOT AA EXIST THEN ENTER VERY LARG2 NUMBER (1.D10 PA) 102 1.D10 PLL!'S LOW VESSEL PRESSURE PERMISSIVE FOR LPCS ES
' 103 49.92D0 ZLRCIC LOW WATER INITIATION FOR RCIC ES 104 1.0D10 PSRCIC HIGH DRYWELL PRESSURE SET POINT FOR RCIC ES 105 30.00 TDRCIC TIME DELAY FOR RCIC ES 106 5.15D5 PHRCIC MINIMUM VESSEL PkESSURE FOR RCIC TURBINE ES 8 498 109 .0119D0 X WE LOW RATE (KG/S) THRU EACH RHR HTX N ASRV1 FLOW AREA 0F RELIEF VALVE TYPE 01 ES 110 .0119D0 ASRV2 FLOW AREA 0F RELIEF VALVE TYPE 42 ES 111 .011SD0 ASRV3 FLOW AREA 0F RELIEF VALVE TYPE 93 ES 112 .011900 ASRV4 FLOW AREA 0F RELIEF VALVE TYPE 94 ES AA IF THE AREA 0F GROUP 45 IS INPUT AS A NEGATIVE NUMBER THEN THE VALVE AA WILL DISCHARGE DIRECTLY INTO THE DRYWELL, IF POSITIVE IT WILL AA DISCHARGE INTO THE SUPPRESSION POOL 113 .0D0 ASRV5 FLOW AREA 0F RELIEF VALVE TYPE 45 ES 114 1.0D0 NSRV1 NUMBER OF TYPE t1 RELIEF VALVES ES 115 1.0D0 NSRV2 NUMBER OF TYPE 02 RELIEF VALVES ES 116 9.0D0 NSRV3 NUMBER OF TYPE 43 RELIEF VALVES ES 117 9.0D0 NSRV4 NUMBER OF TYPE 44 RELIEF VALVES ES l 118 0.00 NSRV5 NUMBER OF TYPE 95 RELIEF VALVES E3 1
119 0.D0 NADS1 NUMBER OF ADS VALVES IN GROUP 1 ES 120 1.00 NADS2 NUMBER OF ADS VALVES IN GROUP 2 ES 121 3.00 NADS3 NUMBER OF ADS VALVES IN GROUP 3 ES 122 4.D0 NADS4 NUMBER OF ADS VALVES IN GROUP 4 ES 123 7.122006 PSRV1 PRESSURE SEIPPINT FOR t1 RELIEF VALVE ES
$ .*k hkhk hRfkkhk kkbkf kN REhf F vh hk 126 7.743D6 PSRV4 PRESSURE SEIPOINT FOR 94 RELIEF VALVE ES 127 1.D10 PSRV5 PRESSURE SETPOINT FOR 45 RELIEF VALVE ES 128 47.1600 ILADS LOW WATER LEVEL FOR INITIATION OF ADS ES 129 114.3703 PSADS HIGH DRYWELL PRESSURE SET POINI FOR ADS ES 130 115.D0 TDADS TIME DELAY FOR ADS ACTUATICH ES AA LPCI,LPCS,HPCS HAVE NPSH REQUIRMENTS l
NAFT GULEEP.DAT;14 6-JUL-1*84 14:25 Page '7 AA HPCI AND RCIC WILL TRIP OFF ON USER SUPPLIED TEMPERATURE DE SUPP POOL 131 373.00 ICHPCI INLET TEMP LIMIT FOR HPCI ES 132 31.400 ICLHPS PUMP CENTER LINE ELAVATION FOR HPCS ES 133 29.300 ZCLLFI PUMP CENTER LINE ELAVATION FOR LPCI ES 134 30.5D0 ICLLPS PUMP CENTER LINE ELAVATION FOR LPCS ES 135 366.3DO TCRCIC INLET TEMP LIMIT EOR RCIC ES 136 305.D0 TWSW SFRVICE WATER TEMP (RHR HEAT EXCHANGERS,TCOLD) ES 137 13.D0 TDDG1 HPCS LOAD DELAY TIME FOR DIESEL ES 138 13.D0 TDDG2 LPCI LOAD DELAY TIME FOR DIESEL ES 139 13.D0 TDDG3 LPCS LOAD DELAY TIME FOR DIESEL ES 140 2.3D-4 XDDROP SRRAY DROPLET DIAMETER FOR CONIAINMENT SPRAYS ES 141 19.600 XHSPWW SPRAY EALL HEIGHT IN WETWELL ES 142 10.00 XHSPDW SPRAY FALL HEIGHT IN DRYWELL ES AA THE HTSW SYSTEM CAN BE USED TO MODEL ANY INJECTION MODE SUCH AS THE SYSTEM IS TOTALLY DEFINED BELOW AA 143SERVICE 1.837D5 WATER ORENTHALPY HWHPSW EIRE WATER,OF HIGH PRES dERVICE WATER (MARK IES CI) 144 1.009D-3 VWHPSW SPEC VOL OF HIGH PRES SERVICE WATER (MARK I CI) ES 145 6.525D5 PHPSW(1) PPS VS. VOLUMETRIC ELOW EOR HPSW CODE INJECTION ES 146 6.52405 PHPSW(2) (MARK I CORE INJECTION) ES 147 6.523D5 PHPSW(3) ES 148 6.522D5 PHPSW(4) ES 149 6.521D5 PHPSW(5) ES 150 6.52005 PHPSW(6) ES 151 6.519D5 PHPSW(7) ES 152 0.D0 PHPSW(8) ES 153 0.00 WVHPSW(1) ES 154 .75700 WVHPSW(2) ES 155 .757D0 WVHPSW(3) ES 156 .757D0 WVHPSW(4) ES 157 .75700 WVHPSW(5) ES 158 .757D0 WVHPSW(6) ES 159 .757D0 WVHPSW(7) ES 160 .757D0 WVHPSW(8) ES fh f.' 33b5 NNk N NT ! hh!S kET I R 163 600.00 Kff! TDSPR TIME DELAY EOR MARK III CONTAINMENT SPRAYS NNI 5S ES 164 7.38D5 PDSRV1 DEAD BAND FOR CLOSURE DE SRVt1 ES 165 9.45D5 PDSRV2 DEAD BAND FOR CLOSURE DE SRVt2 ES 166 1.151D6 PDSRV3 DEAD BAND FOR CLOSURE CE SRV03 ES 167 5.17D5 PDSRV4 DEAD BAND FOR CLOSURE DE SRV04 ES 8 0 TS ARE HR T E E NIC N I TURBINE STEAM ELOW 185 8.22D6 PTURRI(1) PPS-PWW VS. STEAM ELOW TO RCIC TURBINE ES 186 1.3406 PTURRI(2) ES 187 1.34D6 PTURRI(3) ES 188 1.34D6 PTURRI(4) ES 189 1.34D6 PTURRI(5) ES 190 1.34D6 PTURRI(6) ES 191 1.34D6 PTURRI(7) ES 192 1.34D6 PTURRI(8) ES , 193 4.8300 WSTRCI(1) ES l l 194 1.5600 WSTRCI(2) ES
- l
- 195 1.56D0 WSTRCI(3) ES 1 196 1.56D0 WSTRCI(4) E; 197 1.56D0 WSTRCI(5) ES 198 1.5600 WSTECI(6) ES 199 1.5600 WSTRCI(7) ES
. 5 hN GH TURBINE EXHAUST PRESSURE FOR HPCI S 202 1.72D5 PHTURR HIGH TURBINE EXHAUST PRESSURE FOR RCIC ES 203 4.916D5 PCEAIL CONTAINMENT EAILURE PRESSURE ES
DRAFT .. GULEEP.DAT;14 6-JUL-1984 14:29 PaSe 8 AA THE SHUT OEE HEAD SHOULD APPEAR IN THE PUMP CURVE DEFINITION FOR ECCS AA THE NEXT TWO PARAMETERS ARE PERMISSIVE SIGNALS FOR TRIPPING SYSTEMS 204 1.D10 PHLPCI HIGH VESSEL PRESSURE TRIP EOR LPCI ES 205 1.D10 PHLPCS HIGH VESSEL PRESSURE TRIP EOR LPCS ES 206 34.237D0 ZHISP HIGH SUPP. POOL LEVEL TRIP EOR HP SUCTION ES 207 47.16D0 ZLSPR NOT USED AA ALL OF THE HEAT EXCHANGER DATA MAY BE OMITTED WITH THE EXCEPTION AA DE NTUHX1,NTUHX2,NHX1 NHX2 208 0.D0 NTHX NUEBER OF TUBES IN RHR HTX ES 209 0.00 NBHX NUMBER OF BAEELES IN RHR HTX ES 210 0.D0 XIDTHX TUBE ID FOR RHR HTX ES 211 0.00 XTTHX TUBE WALL THICKNESS FOR RHR HTX ES 212 0.00 XTCHX TUBE CENTER TO CENTER SPACING EOR RHR HTX ES 13 0.00 XSHX SHELL LENGTH EOR RHR HTX ES 3'14 0.00 RGE00L FOULING EACTOR FOR RHR HTX ES 215 0.00 KTHX THERMAL CONDUCTIVITY FOR TUBE WALL (RHR HTX) ES 216 0.D0 XBCHX BAEELE CUT LENGTH ECR RHR HTX ES 217 0.D0 XIDSHX SHELL ID EOR RHR HTX ES 218 0.00 XSTHX BUNDLE TO SHELL GAP LENGTH EOR RHR HTX ES AA NTU VALUES NOT NEEDED IF ABOVE INFORMATION IS DEFINED 219 1.2D0 NTUHX1 NTU FOR RHR HTX 41 ES 220 1.200 NIUHX2 NIU FOR RHR HTX 42 ES 221 2.00 NHXI NUMBER OF RHR LOOP 41 HTX ES 222 2.D0 NHX2 NUMBER OF RHR LOOP 42 HTX ES 223 2.D0 EHX TYPE DE RHR HTX(1= STRAIGHT TUBE 2=U TUBE) ES TH ibOWIN N V E SbM!R IS EM AA CORRESPONDS TO THE FIRST ELOW RATE LISTED AB0VE EOR THAT PUMP 225 .518D0 ZHDHPS HPCS NPSH EOR GIVEN ELOWS ES
. 226 .510D0 (METERS) ES 227 .51800 ES 228 .549D0 * ' ~
ES 229 .61D0 " ES 230 .762D0 ES 231 1.37200 ES 232 2.226D0 ES 233 2.073DO ZHDLPI LPCI NPSH EOR GIVEN ELOW ES 234 .671D0 ES 235 .335D0 ES 236 .335D0 ES 237 .335D0 ES 238 .33500 ES 239 .36600 ES 240 .396D0 ES 241 2.457D0 ZHDLPS LPCS HPSH EOR GIVEN ELOW ES 242 1.52400 ES 243 .85400 ES 244 .85400 ES 245 .854D0 ES 246 .85400 ES 247 .854D0 ES 248 .85400 ES 249 31.4D0 ICLRCI PUMP CENTER LINE ELAVATION FOR RCIC ES 250 28.3500 ZCLHPI PUMP CENTER LINE ELAVATION E0R HPCI ES 251 .0093D0 ACVENT AREA DE CONTAINMENT VENT ES 252 0.0D0 ICEAIL ELEVATION OF CONTAINMENT VENT IN WETWELL (MII ONLY)ES 253 28.35D0 ISRVD AVERAGE ELEVATION OF SRV DISCHARGE IN SUPP POOL ES 254 0.D0 TGDWHX(1) COOLING CURVE E0R DRYWELL COOLERS 255 0.D0 TGDWHX(2) TEMP IN DRYWELL VS. HEAT LOSS RAIE (J/S) 256 0.D0 TGDWHX(3) 257 0.00 TGDWHX(4)
DIfAFT SULffP.DAf;14 6-JUL-1994 14:29 Page 9 254 0.D0 TGDWHX(5) 259 0.D0 TGDWHX(6) 260 0.00 TGDWHX(7) 261 0.00 TGDWHX(8) 262 0.00 OGDWHX(1) HEAT LOSS RATE FOR DRYWELL COOLERS (J/S) 263 0.D0 OGDWHX(2) 264 0.D0 OGDWHX(3) 265 0.D0 QGDWHX(4) 266 0.00 GGDWHX(5) 267 0.00 OGDWHX(6) 268 0.D0 QGDWHX(7) 269 0.00 QGDWHX(8) AA DW AA DW ADRYWELL DW 01 5.D-1 RELHDW RELATIVE HUMIDITY IN DRYWELL DW 02 7650.2D0 VOLDW VOLUME OF DRYWELL DW 03 30 7100 TDWF ELEVATION AT DRYWELL ELOOR DW 04 31$.4D0 ADWF AREA DE DRYWELL ELOOR DW 05 35.77D0 ZWDWWW ELEVATION OF WEIR WALL BETWEEN DRYWELL AND WETWELL DW 06 24.D0 NIGDW NUMBER DE IGNITERS IN THE DRYWELL DW 07 9 27D0 X IGDW AVERAGE DISTANCE EROM IGNITER TO CEILING DW 08 3i8.4DO ACHDW CHARACTERISTIC ELOOR AREA FOR BURN CALCULATION. DW AA WW AA WW AWETWELL WW 01 28.35D0 ZWWE ELEVATION AT WETWELL FLOOR WW 02 5.070-2 AVB FLOW AREA THROUGH VACUUM BREAKERS WW 03 3.000 NVB NUMBER DE VACUUM BREAKERS WW 04 6.D3 PSETVB PRESSURE SETPOINT E0R VACUUM BREAKERS WW 05 6.D3 PDVB DEAD BAND EOR VACUUM BREAKERS WW 06 7.9856D3 VOLWW TOTAL VOLUME OF WETWELL (PLUS SUPP POOL) WW 07 6.D-1 RELHWW RELATIVE HUMIDITY IN WETWELL WW 08 6.D0 hlGWW NUMBER OF IGNITERS IN THE WETWELL WW 09 .6D0 XIGWW AVERAGE DISTANCE FROM IGNITER TO CEILING WW 10 619.300 ACHWW CHARACTERISTIC ELOOR AREA EOR BURN CALCULATION WW 11 0.00 AWWF AREA 0F WETWELL ELCOR (MARK II) WW AA PD i AA PD APEDESTAL PD 01 3.269D1 APDF AREA 0F PEDESTAL FLOOR PD 02 3.951D0 APDVT AREA 0F PEDESTAL-DRTWELL OPENING PD 03 2.67802 VOLPD VOLUME OF PEDESTAL PD 04 31.7200 ZWPDDW ELEVATION OF WIER BETWEEN PED AND DRYWELL PD 05 28.80D0 2PDF ELEVATION AT PEDESTAL ELOOR PD 06 5.D-1 RELHPD RELATIVE HUMIDITY IN PEDESTAL PD 07 0.00 NIGPD NUMBER OF IGNITERS IN THE PEDSTAL PD 08 0.00 XIGPD AVERAGE DISTANCE FROM IGNITER TO CEILING PD 09 32.69D0 ACHPD CHARACTERISTIC FLOOR AREA FOR BURN CALCULATION PD 10 0.00 XWPDVT WIDTH Of PEDESTAL 000R (MARK II ONLY) PD AA NOTEI THE NEXT PARAMETER-ADCPD-CAN BE USED TO MODEL THE NORMAL AA LEAK AREA BETWEEN THE DRYWELL AND COMPARTMENT A DE A MARK III 11 0.004300 ADCPD AREA DE PEDESTAL 00WNCOMERS 12 0.D0 NDCPD NUMBER OF PEDESTAL DOWNCOMERS 13 2.D0 XHPDDW DISTANCE BETWEEN UPPER AND LOWER VENTS FOR AA PED-DRTWELL NATURAL CIRCULATION
.AA SP AA SP ASUPPRESSION POOL (MARKIII ONLY) SP 01 5.15D1 ASPDW AREA 0F DRYWELL SIDE OF SUPPRESSION POOL SP 02 6.193D2 ASPPC AREA 0F CONTAINMENT SIDE OF SUPPRESSION FOOL SP
~
BRAFT GULFEP.DAT;14 6-JUL-1994 14:29 Page 10 03 4.5D1 NYT1 NUMBER OF VENTS OF TYPE il -- TOP SP 04 4.5D1 NVT2 NUMBER OF VENTS OF TYPE 42 -- MID SP 05 4.5D1 NVT3 NUMBER OF VENTS OF TYPE 43 -- BOTTOM SP 06 7.10-1 XDIAVT DIAMETER OF ONE SUPPRESSION POOL VENT SP 07 33.449D0 2LLSP ELEVATION DE SUPP. POOL LOW LEVEL SETPOINT SP 08 32.16D0 IVT1 ELEVATION OF IOP OF VENT TYPE 41 SP 09 30.89D0 IVT2 ELEVATION OF TOP OF VENT TYPE 42 SP 10 29.6200 IVI3 ELEVATION OF IOP OF VENT TYPE 43 SP AA IN AA IN AINITIAL CONDITIONS IN 01 3.833D9 GPOWER CORE POWER IN 02 7.17D6 PPSO INITIAL PRESSURE IN PRIMARY SYSTEM IN 03 1.005 PPD 0 INITIAL PRESSURE IN PEDESTAL IN 04 1.005 PDWO INITIAL PRESSURE IN DRYWELL IN 05 1.005 PWWO INITIAL PRESSURE IN WETWELL IN 06 34.01D0 2SPDWO INIT.ELEV. OF WATER LEVEL IN DW SIDE OF SUPP. POOL IN 07 34.01D0 ZSPWWO INII.ELEV. OF WATER LEVEL IN PC SIDE OF SUPP. POOL IN 08 3.3D2 IPD0 INITIAL TEMPERATURE IN PEDESTAL IN 09 3.3D2 TDWO INITIAL TEMPERATURE IN DRYWELL IN 10 3.08D2 TWWO INITIAL TEMPERATURE IN WETWELL IN 11 3.08D2 TWSPO INITIAL TEMPERATURE OF SUPPRESSION POOL WATER IN 12 51.91DO ZWSHO INITIAL ELEVATION OF WATER IN THE SHROUD IN 13 2.05D6 MWCB0 MASS OF WATER IN UPPER POOL (MARKIII ONLY) IN 14 896.600 VCSTO VOLUME OF WATER IN CONDENSATE STORAGE TANK IN 15 297.00 TAMB AMBIENT TEMPERATURE IN 16 1.D5 PAMB . AMBIENT PRESSURE IN AA CC
!. ONTROL CARDS .
Ek 01 3 IBWR CONTAINMENT TYPE (MARK 12 OR 3 ) CC 02 IRSIW UNIT NUMBER TO WRITE RESIAltf FILE (MAIN) 46 CC 03 47 IHUW UNIT NUMBER TO WRITE RESTART FILE (HEATUP) CC h b blhT FibE Ib bb 06 500 IPTSMX MAXIMUM NUMBER OF PLOTTED POINTS CC 07 4 IPISPK MAXIMUM HUMBER OF PLOT POINTS TRACED FOR FULL CC AA SCALE SPIKE CC 08 80 IPISAV NUMBER OF POINTS SAVED FOR NON-CHANGING PLOT CC 09 1 ISUMM
SUMMARY
DATA (0=ALL EVENIS,1= SHORTER LIST) CC 10 48 ISUM
SUMMARY
FILE NUMBEP CC 11 1 IRUNG 1=1ST ORDER R-K 2=2ND ORDER R-K CC l 12 1 IFREE21=DOFREEZEFRdNTCALC.(0=NOCALC.) CC 13 5 INPGRP NUMBER OF TRACE GAS TYPES (FISSION PRODUCTS) CC 14 1 IRET 1= WRITE RETAIN FILE (0=NO FILE) CC 15 49 IFPPLT RETAIN PLOT FILE UNIT NUMBER CC AA TD AA TD kIIMING DATA ID 01 20.00 TDMAX MAXIMUM ALLOWED TIME STEP TD 02 1.0-3 TDMIN MINIMUM ALLOWED TIME STEP ID 03 5.D-2 EMCHMX MAXIMUM MASS CHANGE (%) FOR INTEGRATION TD 04 5.D-2 FUCHMX MAXIMUM GAS TEMP CHANGE FRACTION FOR INTEGRATION TD i 05 4.D2 MAXMST MAXIMUM MASS OF STEAM CHANGE PER IIME STEP IN PS TD l AA CA l AA CA l ACOMPIA (MARKIII-MIDDLE WETWELL COMPARTMENT) CA l 01 41.2500 ICAF ELEVATION OF HCU DECK CA l 02 1.1589D4 VOLCA VOLUME OF COMPARTMENT A CA 03 6.D-1 RELHCA RELATIVE HUMIDITY IN COMPT. A CA 04 325.D0 ACAF AREA 0F COMPI. A FLOOR CA
DRAFT GULTTP.94T;14 6-JUL-1984 14:29 Page 'll 05 348.D0 ACACB FLOW AREA BETWEEN COMPT. A AND COMPT. B CA 06 139.D0 AWWCA FLOW AREA BETWEEN WETWELL AND COMPT. A CA 07 41.25D0 ZWCAWW CURB HEIGHT ON MIDDLE DECK CA 08 58254.00 PPUR(1) DRYWELL PURGE PRESSURE VS FLOW (MAA3/KG) CA J 09 87140.00 PPUR(2) CA 10 1.05D5 PPUR(3) CA 11 1.128D5 PPUR(4) CA i 12 1.162D5 PPUR(5) CA j 13 1*162D5 PPUR(6) CA 14 1.162D5 PPUR(7) CA 15 1.162D5 PPUR(8) CA 16 .55500 WVPUR(1) CA l 17 .51500 WVPUR(2) CA 18 .453DO WVPUR(3) CA l 19 .40100 WVPUR(4) CA 20 .33000 WVPUR(5) CA 21 .33000 WVPUR(6) CA 22 .33000 WVPUR(7) CA 23 .33000 WVPUR(8) CA 24 2.D0 NPURP NUMBER OF DRYWELL PURGE PUMPS CA 25 47.16D0 2LPUR LOW WATER (LOCA) SIGNAL FOR DRYWELL PURGE CA i 26 1.143705 PDWPUR HIGH DRYWELL PRESSURE (LOCA) FOR DRYWELL PURGE CA I 27 6.D3 PDPUR PRESSURE DIFFERENTIAL SET POINT FOR DRYWELL PURGE CA l 28 30.00 IDPUR TIME DELAY FOR DRYWELL PURGE CA I 29 45.D0 NIGCA NUMBER OF IGNITERS IN THE COMPT A CA 30 8*34D0 XIGCA AVERAGE DISTANCE FROM IGNITER TO CEILING CA 31 6.D0 NIGBCA NUMBER OF IGNITERS IN WETWELL SEEN BY COMPT. A CA 32 325.00 ACHCA CHARACTERISTIC FLOOR AREA FOR BURN CALCULATION CA AA CB AA CB ACOMPTB (MARKIII-UPPER WETWELL COMPARTMENT) CB C1 63.69DO ICBF ELEVATION OF OPERATING DECK CB 02 23766.500 VOLCB VOLUME OF COMPT. B CB 03 6.D-1 RELHCB RELATIVE HUMIDITY IN COMPT.B CB 04 64.00 ZWCBWW CURB HEIGHT ON UPPER DECK CB 05 302.200 AWCB WATER AREA ON CB DECK CB 06 7.5834D4 PCPUR(1) PRESSURE VS FLOW FOR CONTAINMENT PURGE CB 07 0.00 PCPUR(2) CB 08 0.D0 PCPUR(3) CB 09 0.D0 PCPUR(4) CB 10 0.00 PCPUR(5) CB 11 0.00 PCPUR(6) CB 12 0.00 PCPUR(7) CB 13 0.D0 PCPUR(8) CB 14 1.416D0 WVCPUR(1) CB ; 15 1.416D0 WVCPUR(2) CB i 16 1.416D0 WVCPUR(3) CB 17 1.416D0 WVCPUR(4) CB 18 1.416D0 WVCPUR(5) CB f.*4 21 1.416D0 WVCPUR(8) CB 22 1.03D3 VOLUPD VOLUME OF WATER IN UPPER POOL DUMP CB 23 47.1600 ZLUPD LOW WATER (LOCA) SIGNAL FOR UPPER POOL DUMP CB 24 1.1437D5 PDWUPD HIGH DRYWELL PRESSURE (LOCA) FOR UPPER POOL DUMP CB 25 225.00 TDDUMP TOTAL TIME FOR UPPER POOL DUMP CB 26 1.8D3 TDUPD TIME DELAY FOR UPPER POOL DUMP CB 27 56.2500 ZCDF1 ELEVATION OF UPPER POOL FLOOR CB 28 1031.D0 VLAUPD VOLUME OF WATER REMAINING IN UPPER POOL AFTER CB AA UPPER POOL DUMP CB 29 18.D0 NIGCB NUMBER OF IGNITERS IN THE COMPI B CB l
OO A-12 U l .-- GULTTP.DAT;14 6-JUL-1984 14:29 PaSe 12 30 .3D0 XIGCB AVERAGE DISTANCE FROM IGNITER TO CEILING CB 31 16.D0 NIGBCB NUMBER OF IGNITERS IN COMPT A SEEN IN COMPT B CB 32 1121.9D0 ACHCB CHARACTERISTIC FLOOR AREA FOR BURN CALCULATION CB AA HS AA . HS IN fE ER TO DRAWING IN VOL II 0F MAAP USERS MANUAL ON MARKIII HEAT SINKS 01 186.600 AHS1 AREA 0F WALL 61 HS 02 1074.8D0 AHS2 AREA 0F WALL 42 HS 03 492.300 AHS3 AREA 0F WALL 03 HS 04 883.8DO AHS4 AREA 0F WALL 64 HS 05 27J$.100 AHS5 AREA 0F WALL 65 HS
. 06 3411.000 AHS6 AkEA 0F WALL 66 HS 07 371.100 AHS7 AREA 0F WALL 67 HS 08 289.6DO AHS8 AREA 0F WALL 48 HS 09 2.07700 KHS1 THERMAL CONDUCTIVITY OF WALL 01 HS 10 2.077D0 KHS2 THERMAL CONDUCTIVITY OF WALL 02 HS 11 2.077D0 KHS3 THERMAL CONDUCTIVITY OF WALL 43 HS 12 2.077D0 KHS4 THERMAL CONDUCTIVITY OF WALL 64 HS 13 2.07700 KHS5 THERMAL CONDUCTIVITY OF WALL 65 HS 14 2.077D0 KHS6 THERMAL CONDUCTIVITY OF WALL 66 HS 15 2.077DO KHS7 THERMAL CONDUCTIVITY OF WALL 67 HS 16 2.077D0 KHS8 THERMAL CONDUCTIVITY OF WALL 48 HS 17 1.753DO XHS1 THICKNESS OF WALL 01 HS
! 18 1.524D0 XHS2 THICKNESS OF WALL 62 HS i 19 1.52400 XHS3 THICKNESS OF WALL 03 HS 20 1.067D0 XHS4 THICKNESS OF WALL 64 HS 21 1.067D0 XHS5 THICKNESS OF WALL 65 HS 22 . .762D0 *XHS6 THICKNESS OF WALL 66 HS 23 .6100 XHS7 THICKNESS OF WALL 47 HS 12 4 1.29500 XHS8 THICKNESS OF WALL 48 HS 25 0.D0 XLHSIl INNER LINER THICKNESS FOR WALL 81 HS 26 0.00634D0 XLHSI2 INWER LINER THICKNESS FOR WALL 02 HS ( h h.*hhk 4hh XhHhf f N R hfNh2 fHfCK WAhl 29 0.00634D0 XLHSI5 INNER LINER THICKNESS FOR WALL 45 HS , 30 0.00634D0 XLHSI6 INNER LINER THICKNESS FOR WALL 66 HS l 31 0.00 XLHSi? 1NNE: LINER THICKNESS FOR WALL 67 HS 32 0.00634D0 XLHSIS INNER LINER THICKNESS FOR WALL 48 HS 33 0.00634D0 XLHS01 OUTER LINER THICKNESS FOR WALL el HS 34 0.D0 XLHS02 OUTER LINER THICKNESS FOR WALL 42 HS 35 0.0063400 XLHS03 OUTER LINER THICKNESS FOR WALL 43 HS 36 0.D0 XLHSO4 OUTER LINER THICKNESS FOR WALL 64 HS 37 0.00 XLHS05 0 UTER LINER THICKNESS FOR WALL 65 HS 3 . 12D0 LHS0h O k b hh kbkN S OR WA 4 40 0.0063400 XLHS08 OUTER LINER THICKNESS FOR WALL 68 HS 41 2300.00 DHS1 DENSITY OF WALL el HS 42 2300.00 DHS2 DENSITY OF WALL 42 HS l ' 43 2300.00 DHS3 DENSITY OF WALL 03 HS 44 2300.D0 DHS4 DENSITY OF WALL 64 HS 45 2300.00 DHS5 DENSITY OF WALL 65 HS 46 2300.D0 DHS6 DENSITY OF WALL 06 HS 47 2300.00 DHS7 DENSITY OF WALL #7 HS 48 2300.00 DHS8 DENSITY OF WALL 68 HS 49 880.D0 CPHS1 SPECIFIC HEAT FOR WALL el HS 50 880.D0 CPHS2 SPECIFIC HEAT FOR WALL 82 HS . 51 880.D0 CPHS3 SPECIFIC HEAT FOR WALL 63 HS ! 52 880.00 CPHS4 SPECIFIC HEAT FOR WALL 04 HS 53 880 D0 CPHS5 SPECIFIC HEAT FOR WALL #5 HS 54 880.D0 CPHS6 SPECIFIC HEAT FOR WALL 66 HS
MAFT GULFFF.DAT;14 6-JUL-1984 14:29 Pa9e 13 55 880.D0 CPHS7 SPECIFIC HEAT FOR WALL 67 HS 56 880.00 CPHS8 SPECIFIC HEAT FOR WALL 60 HS AAALL OF THESE EQUIPMENT HEAT SINKS ARE LOCATED IN GAS VOL. OF COMPARTMENT 57 0.00 MEOPD MASS OF EQUIPMENT IN PEDESTAL HS 58 151000.D0 MEODW MASS OF EQUIPMENT IN DRYWELL HS 59 0.00 ME0WW MASS OF EQUIPMENT IN WETWELL HS 60 342462.D0 MEOCA MASS OF EQUIPMENT IN COMPT A HS 61 1.9581D6 MEOCB MASS OF EQUIPMENT IN COMPT B HS 62 0.D0 AEOPD AREA 0F EQUIPMENT IN PEDESTAL HS 63 4153.D0 AEODW AREA 0F EQUIPMENT IN DRYWELL HS 64 0.00 AE0WW AREA 0F EQUIPMENT IN WETWELL HS 65 9.7177D3 AEOCA AREA 0F EQUIPMEHT IN COMPI A HS 66 1189.2D0 AEOCB AREA 0F EQUIPMENT IN COMPT B HS 67 50.D0 HIOUIW HEAT TRANSEER COEFF. AI OUTER WALL HS 68 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE 01 HS 69 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE #2 HS 70 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE $3 HS 71 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE 44 HS 72 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE 45 HS 73 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 06 HS 74 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 47 HS 75 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 48 HS 76 0.D0 RGAP OUTER LINER TO WALL GAP RESISTANCE 41 HS 77 0.D0 RGAP OUTER LINER TO WALL GAP RESISTANCE 92 HS 78 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE 43 HS 79 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE 44 HS 80 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE $5 HS 81 0.00 kGAP OUTER LINER TO WALL GAP RESISTANCE 96 HS 82 0.D0 RGAP OUTER LINER TO WALL GAP RESISTANCE 47 HS 83 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE #8 HS 84 0.D0 ME0WWS MASS OF EQUIP. HEAT SINK WETWELL (SUBMERGED) HS 85 0.D0 AE0WWS AREA 0F EQUIP. HEAT SINK WETWELL (SUBMERGED) HS 86 1.45D2 AHSPSI AREA 0F RPV IN GAS SPACE OF VESSEL HS 87 1.7502 AHSPS2 AREA 0F RPV IN DOWNCOMER REGION HS 88 6.500 AHSPS3 AREA 0F RPV IN LOWER PLENUM HS 89 2.2405 MRPV1 RPV WALL MASS IN UPPER DOME REGION HS 90 2.22D5 MRPV2 RPV WALL MASS IN DOWNCOMER REGION HS 91 9.D4 MRPV3 RPV WALL MASS IN LOWER PLENUM REGION HS AA HS AA HS AMODEL PARAMETERS FOR BWR 01 .005D0 FRCOEF FRICTION COEFFICIENT FOR CORIUM IN VFAIL MO 02 2.0D-1 FMAXCP FRACTION OF TOTAL CORE MASS WHICH MUST MELT AA TO FAIL THE CORE PLATE 03 50.D0 HTBLAD FUEL CHANNEL TO CONTROL BLADE HEAT TRANS. COEFF MO 04 300.00 HIFB FILM BOILING HEAT TRANS. COEEF. NO 05 0.D0 FBLOCX FUEL CHANNEL BLOCKAGE PARAMETER MO AA 0= BLOCKAGE AT T200FF,1=NO BLOCKAGE NO 06 2300.D0 TT00FF OXIDATION CUT-OFF TEMPERATURE NO 07 .300 FACPF ERACTION OF CORE PLATE AREA THAT FAILS MO 08 5.00 CDBPD ELAME BUOYANCY DRAG COEEFICIENT IN THE PEDESTAL MO 09 5.D0 CDBDW FLAME BUOYANCY DRAG COEFFICIENT IN THE DRYWELL MO 10 5.00 CDBWW ELAME BUOYANCY DRAG COEFFICIENT IN THE WETWELL MO 11 5.00 CDBCA FLAME BUOYANCY DRAG COEFFICIENT IN COMPARTMENT A MO 12 5.D0 CDBCB FLAME BUOYANCY DRAG COEEFICIEHI IN COMPARTMENT B MO 13 .1000 XCNREF CORIUM REFERENCE THERMAL BOUNDARY LAYER THICKNESS N0 14 1.D3 HTCMCR CORIUM-CRUST HEAT TRANSE. COEFF. USED IN DECOMP N0 5"' "'"* 11 16 0.0100 ilBll'"n!9'115"l?i"'SS " "'"' "
"^"""s8 XDCMSP PARTICLE SIZE (DIAMETER) FOR CORIUM AS IT FALLS M0 AA INTO SUPPRESSION POOL (MARK II ONLY) NO
DRAFT GULEEP.DAT;14 6-JUL-1984 14:29 Page 14 17 903.D0 TCELAN CRITICAL ELAME TEMPERATURE NO 18 1.53D0 ECHTUR CHURN-TURBULENT CRITICAL ELOW PARAMETER MO 3.7D0 EDROP DROPLET CRITICAL ELOW PARAMETER MD 19 20 3.D0 EELOOD ELOODING ELOW PARAMETER MO 21 1.D0 ESPAR PARAMETER FOR BOTTOM-SPARGED STEAM VOID ERACTION MO 2.D0 EVOL PARAMETER FOR VOLUME SOURCE VOID ERACTION MODEL MO 22 M0 23 5.D-1 ITENTR ENTRAINMENT EFFECTIVE EMPTTING IIME f*.kh0 2h 85D0 ECM L i"1ii1 1? s lE EMISSIVITY OF CORIUM M0 h6 .8 0 EO k IVIf N EQt IPMENT H 29 0.500 E0VER ERACTION OF CORE SPRAY ELOW ALLOWED TO BYPASS COREMO 30 1.D0 NPE NUMBER OF PENETRATIONS EAILED IN LOWER HEAD MO 31 2.00 ECDCDW DOWNCOMER PERIMETER PER MEIER FROM PEDESTAL DOOR M0 MO AA (MARK II ONLY) 32 0.14D0 ECHE COEEFICIENT FOR CHE CORRELATION IN PLSIM MO h *.kDO EE B k I L k b b RION BY TO REPRESENT DIEFICULTY (GT 1.00) OR EASE (LI 1.DO) M0
$h AA AA FOR MAIERIAL TO BE BLOWN OUT OF CAVITY MO 35 1.00 SCALU SCALING EACIOR FOR ALL BURNING VELOCITIES MO 36 1.00 SCALH SCALING EACTOR E0R HI COEEFICIENTS TO PASSIVE NO AA HEAT SINKS MO 37 2.0D0 EUMIN CLADDING SUREACE MULTIPLIER M0 AA N0 ACONCRETE PROPERTIES CN 01 56.00 MOLWCN MOLECULAR WEIGHT DE CONCRETE CN 02 1743.D0 ICNMP MELTING TEMPERATURE DE CONCRETE CN 03 .8D6 LHRCN REACTION ENERGY FOR CONCRETE DECOMPOSITION CN 04 65.D0 DCEWCN FREE WATER DENSITY IN CONCRETE CH 05 65.00 DCCWCN COMBINED WATER DENSITY IN CONCRETE CN 06 572.00 DCC2CN CO2 DENSITY IN CONCRETE CN 1.06 LHCM LAIENT HEAT IQ MELT CONCRETE CN
{ FI AFISSION PRODUCTS 01 .028D0 E0P(1) PERCENT POWER IN EISSION PRODUCT GROUP 1 FI 02 .17600 E0P(2) PERCENT POWER IN FISSION PRODUCT GROUP 2 FI 03 .01900 E0P(3) PERCENT POWER IN EISSION PRODUCT GROUP 3 FI 04 0.00 EQP(4) PERCENT POWER IN FISSION PRODUCT GROUP 4 EI 05 0.00 E0P(5) PERCENT POWER IN FISSION PRODUCT GROUP 5 FI 06 439.3 MEP(1) MASS DE FISSION PRODUCT GROUP 1 -NOBLES El 07 237.7 MEP(2) MASS OF EISSION PRODUCT GROUP 2 -CS+1 FI 08 37.1 MEP(3) MASS DE FISSION PRODUCT GROUP 3 -TE El N ISSf R QUP 5 - hf 11 1190.00 MSM0(1) MASS OF SN IN CORE REGION El 12 268.D0 MSM0(2) MASS DE MN IN CORE REGION EI 13 0.0D0 EDSP(1) SPRAY REMOVAL LANDA EOR EP GROUP 1 El 14 0.0D0 EDSP(2) SPRAY REMOVAL LANDA EOR EP GROUP 2 FI 15 0.000 EDSP(3) SPRAY REMOVAL LANDA E0R EP GROUP 3 El 16 0.000 EDSP(4) SPRAY REMOVAL LAMDA FOR EP GROUP 4 FI d 0 fD WWV TNkh0 EkT R EP GROUP 1 bl 19 600.00 EDESP(2) DRYWELL VENTS DECON. EACTOR EOR EP GROUP 2 FI 20 600.0D0 EDESP(3) DRYWELL VENTS DECON. FACTOR FOR EP GROUP 3 El 21 600.000 EDESP(4) DRYWELL VENTS DECON. FACTOR EOR EP GROUP 4 FI 22 600.000 EDESP(5) DRYWELL VENTS DECON. EACIOR E0k EP GROUP 5 EI 23 1000.00 EDERV(1) SRV DECON. EACTOR FOR EP GROUP 1 El 24 1000.D0 EDERV(2) SRV DECON. FACTOR EOR EP GROUP 2 FI 25 1000.0D0 EDERV(3) SRV DECON. EACIOR FOR EP GROUP 3 EI
l I DRAFT GULFEP.DAT;14 6-JUL-1984 14:29 Page 15 26 1000.0D0 EDERV(4) SRV DECON. FACTOR FOR FP GROUP 4 FI 27 1000.0D0 EDERV(5) SRV DECON. FACTOR FOR EP GROUP 5 FI AA 28 - 31 ARE ECUR CONSTANTS IN HENRY-EPSIEIN MODEL AA EI ABR I i f I
DRAFT APPENDIX B Supplemental Plots for the Base Accident Sequences l l 1
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- llC3 T10UV - GRAND GULF >
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T10UV - GRAND GULF 2,000..-...-..----....--..-.-.._-. _
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\ . l l l l T10UV - GRAND GULF so, ass-. . - - , r 10.800 - - l- - -
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I l T10UV - GRAND GULF \ n e. e es._ [ i i 15?.000- -
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Fig. B.7 Nss of water in the pedestal.
T10UV - GRAND GULF O.070- - - , D.66D - -~~
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T10UV - GRAND GULF - 259,808 , -
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+. .
+ -.
- \ !
s .
'-- ! i 0- ,
, - , ,-- , , ,;,.r, i r, ,,.. ,,,...,....,
0 10 20 30 40 50 60 70 88 Ilt1E til0 URSI Fig. B.13 Mass of UO 2 in core region.
DRAFT SUPPLEMENTAL PLOTS FOR SEQUENCE AE 1 l 1 i l l l
2 AE - GRAND GULF m 40 , . _ _ 35 U ._..___.___..___.-__ __.g.____ 30- . i g 2 5 -.- --. __ m a
~
i
' i 202 - . . . . _ _ ; _.. .. . _
o - a - o - n - g i n -- w
.._.._.+... ...j L i m 1e-- ,
- p. . - . . .;. _ _. ; . _ _.. ;
- i i i
5 --. . . . . p-
. ; __ , ___.. _. fj
. t .. .
'/ i ! -
I g
/
i . . . . i . i .
..i... . . .
j i . . .
- 0. 00 0.20 6.40 0.60 0. 80 1.00 1.20 1.40 TIME
( (HOURSI l fig. B.14 Total H2 generated. l
DRAFT o L i i 4.. . . . . . . . . u*w
's l.
\g 1 i
s.N I
.s% 1 ,-
. .. s . . . - . . -
s,s a
% i * ' t. s. m J ~s . , -s -
b. u e . . . . , N~s ., , . . . . _.s r.n l-
. , 2.2 o
o 2 ,.,...N. .,,-
. en N
q s .
. . - , e "o
r _ m Ld :! ai c n
. ._ c., .
en
- 6 h
\
. I rrn.i.,..,,,,,, . , ,
-r r ,7 ,1 --- i; a o o c., au -, a o o, o > no >
a u c ., .n , I $ $ % @- L 8 'u 010ZHW
l CI lllllI:3 l AE - GRAND GULF 3l2=
* * - . . . - ,. - 0
. - . _; ,_. _._y._. s 4e}
_ i F 30_ j......_.. . .-....j.. . . . . , .....;... _ . j. . . w _ b - i i i I - i 4 20_ .i. i. , - - - . -- - - - -
-- - - - -- i m n -
A, O 10 -
.----+--. -
5 (
~
o . . . . . . , . .. ., . . . i i .
. . . . i . . .
i
- 0. 00 6.20 6. 40 0.60 0.80 1.09 1.20 1.40 tit 1E LHOURS)
Fig. B.16 Reactor vessel water level.
AE - GRAND GULF 2, # 0e g~. . ..--_...._7 _ ., . . . _ . . . _ . . . _ , . _ . . . . _ _ , . . _ + u.. 2. 000-
. . . l.. .
d D O Wr4 C u tt.E R m , , g , n, ~ w \ ! ! o - N --; l g 1.500- . - - - - i
.-"u P P E. R PLEHUR ;
5 g .
! ! 7 N
- u. -
O _ g 1. 00 6 l . . ./. , -
.7....-...
e t. o -l / I I 6.c - ,
. i-I gr _
w .
- a. -
I ! ! 1 500- H./ .
-+------ - --- :
W - 6- . [ , 0 . . . .,....i....,....i....i... ,,.. , , 0 10 20 30 40 50 60 70 TINE (HOURSI Fig. B.17 Temperature of structure, "F.
- Zl3 3lll2 AE - GRA"E GULF M E
I g 35.8A8.680 ,. . . . - - . . - - . _ . . . . . - _ . . . . _ . . _ - . . . , . _ . . . . . _ _ . . . . , . a - i i i ! ! na - !* E 3s. 8Jw &;;- _ . . _ _ _ . . _ . . . . . . . . . . . 3
>- I :
o 4 \ o E as.m.c;;d l .. .-.__ 9- 4_ _ _ _ .. _ . .. _ p . . _ _ . ....;...___....._..i.. ;_ j I e -1 z - O - g i.. .. --.+_ -- . -- 4 4
< 2c.eee.eee j -
Ai cn E 1
/ i A
N g 15.see,eee_ . - _. - - ; . . F .. v._. 4 --;. _.-.i o - 3 , -; nu o
~ f -\ i i i e :..m.u._
4 g
. . . . . .\,. ^. r!uouncun,ea .. .;_..,_. ; ; ;
f
' " 'i'
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E
- s. na. su .- .
[o econon [ % f ,-uerra\* + -
- i
! A s i r R, \ g is ,$C..\l,,,.
-i
,,... ,... ,.. ,, ...i i
' 0 10 20 30 40 50 60 70 TIME 010URS)
Fig. B.18 Fission product decay heat on structure, Btu /hr.
DRAFT o
,7....-....... f
. 7. . . . - . . . . . - . . -
7,
=, ,
m
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- 2
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!. : 7 7
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~.
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.l lllll illlll, I I I, I I
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o si
~IiD 11i L 'll' .:1 '31 3 $ #
4 s i I
1 l l h
- Xll3 AE I . R ANI] GULF .
1 0 90- .
~
1 i ! O S0 g ? [ .. O 'd-i - LJ. t 0 . I
. .n U 50 i g
- f. ;
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, .,____-_s ,, 8
, %.~ s 4
~
m
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'~. ,- ... .,
/ !
. i l .n'g- , , , ,
,, 7 , _ , - - . m _,._,_y_,._,_., ,_q i
in a c.o -; w t, o 7p 1111 ill . UPS) Fig. B.24 Hole fraction of steam in Compartment B. 1 1 J l __
i i 1 i AI: i, R ANil GULF iso. :co i i
'oo. ::o T -
- i d
- . /[.
' I
- 250.2:0 _;-
I ;
*- 3 l u u t _. .
l ** m w
.: en1 ~
e, , 1 - to e
- $ . pas ;- ,
j ..ou.* , . , II i . . .
.g . , . - .
g . - v . 3 - - , . - j , , _, rq l i' to 9 at 4' to ( (i 70 1 Ill. 'H.uPSI Fig. B.25 Volumetric flow out of containment. 1
---l
i f 1 Ob IsI!Allu 6ULF -, i I l 108. he-35u,OZu .
. ._ j 4 .
! 100.'.90 _ , i .- l ;; 250. I;0-} . . ~- -. a
.. . oo: :
O _ u i a _ cn 1 p 25?.one-- i.
^- g i i 100.dd's" .
?- i
) t!. uu9 i-
- i
\
4 a
. . - .. . , . . . g -- i , - > g i --- i - - ;
, i- r g e i - iq i . 10 : at 4: SC t. D 10 i i 11 1. sit ;UF31 I
i fig. B.26 Mass of UO 2 in c re region. 1 i l i
O DRAFT 4 I SUPPLEMENTAL PLOTS FOR SEQUENCE 23 T 0N I i } I 1 l r l \ l l
um m B-32 UH
. ....... .... . . ......,i......... . .. .,, g
~
L. t J 3 3
" N 2 s ~-~.. ---. _
v g
= m
] N u _
y 2. z l - m "I S o n ~
~
m* (N
& c2
.b u
I ! ! I ! " I I I I I $ . 1 I I E *f [ g ) g g g 009 00& OOC 0 02 001 d* Gl (TIDUlW
4 4 1 1 i I230W - GRAND GULF i O ....... . . . . . . . . . . . . . . - i , 8, . - o - 8 : o : : i p - i
- ' / :
! = / : 39b s T
- n. -
? -
E - . cn
. 8 W
1 1 w l - 1 i o _ - ! 8 - /. : r 1 ( . l' . - l . .l g, . 1 0 50 100 150 i
- TIHE HH Fig. B.28 Total H2 generated.
-es H
DRAFT 4 a6 45 56 444 i3448 3 I i4 ga i 6 33 I 3 4 giei4 46 1 63 e 6 W D Q lJ. L
- G s
*J LJ e 3
l3 v
;Z *
~
< e rz a L2 - E >
~
L i ." O g +J
/ U
,i m
.1 . , o 3
L7 ! W / m i= / m.
/
- CO b
4
,f.
, i ri. . ...L , , , , , , , , ,,! , , 1 , , , ! , , 1 o
09 Oh Ok OC 04 0l' LJ llSMX
4 i i ) ~ T230W - GRAND GULF n in g CORE g UPPER PLENUM y DOWNCOMER ,
,o g........................................................................................3 x :
t n - _
- u. : -
=
- w _
em i . j 3 e p o "
- 7
> a e
- h :
t H :
= : .
m o .,
- a j w : I I E -
1 3 i e : - 1 4 -
- e -
1 w l,- - i n. - i 2 w o l i i r _(-
._i- - -
l i f i . . . . . . . . l . . . . . . . . , 1 . . . . . . , , , l . . . . . . . . . l . . . . . . . . . l . , , , , , , , , l . . . . . . . . . l . . . . . . . . l . , , , , , , , , l . . . . . . . . .:
- o. o. 2o o. 4o o.so 080 t 1.2 1.4 1.6 18 202 I' TIME HR x10 ,
I j o ,. e. , e , _ ., _ _ . s. , -i E
i i , 3l38 T230W - GRAND GULF e E CORE h UPPER PLENUM y DOWNCOMER , iiiiiiiigi . ... g ..... . iiiiiii...g ..... . g . ....igini ... ig . iiigiiiiiiiiigi. ... it i x : EN s
~
s : - r _ - m - u -
- m
, e : - o p tn _- - . o : : t a _ g _ W - e _- m i zO n
- 3 r -
/ _# :
4 y - i m _ - i . 1 - y - -
; < 1 2 o
; ,5 Ji _
3 _ i 87 ; . . . . . . . . l . . . . . . . . . l . . . . . . . . . l . . . . . . . . . l . . . . . . . . . l . . . . . . . . l . . . . . . . . l . . . . . . . . . l . . . . . . . . . l . . . . . . . . .: , E O. O.20 0.40 0.60 0 80 1 1.2 1.4 1.6 18 20 . g TIME HR x10 ' e m Fig. B.31 Fission product decay heat on structure, Btu /hr. i
/
T230W - GRAND GULF e m o ,,,,,,,,,,, , . . > > x j 4 i to - _- i m - - 1 J _ : av -
)
,i C . -
a E l - l' - m
- ~ a I %
i ~ , - r - n, _ i -
,/ 7
/
J
/ :
i.
* *I : ' '
fF
" goo 150 TIME ilH
,,,. e. , m e n o eeereteo.
s i p I H l s
llll23 T230W - GRAND GULF
. . .. . . . . . ... ....._ H K : W1 i r
l
,/ \
n
\'
M - l t - 3
~
t j i -
- i N -
! r s
- n. - ,. _
2 ss _ E . .i s : i i I
\ : ?
S E i [
- . I _
_- t _ i :- i
.\ i O
O. 50 100 150 ] l 1 lite HH Fig. B.33 Mass of water in the pedestal, f i b r
'I J
- , - - -- - ,A-d T230W - GRAND GULF i ~
v O ~ . . . . . . . . . . . . . . . . . . . . . . . ....._ -
, X : .
J 1
~
=
( - I I m : f - v : I i
- n. _
l i
.I. n ~
I j _ i t
. co y'
_ e *
=
a - e . i ._ j ! .'t l I 5
~
! i :
! x , i :
O 1 O. 50 100 150 Tine HR ng. B.x Me trutim e H 2 la W*" eat 8. m l H l
1
- 23 T230W - GRAND GULF g R _
,,,,,,l X :' :
I '7 i , g - . I i C ~
\ .
i y - g
~
.s -
l 0o '- s ! 2~ _ s I
- 1 -
m e s 1 - u o l i
}-
'N -
t i-
- i. '_
~
I
~
x O N ' i O. 50 100 150 TlHE HR Fig. B.35 Hole fraction of 0 in2Compartment B.
DRAFT grini .ig .. inti ng6 sieassigi>+ii' o
.- n.
1 a {_3's.
- Jl m l
~ . , ' e
% ~~ 5 9
u e g - c. a d _ o o g .- - o v C
~ -
O z .
$ o"
< a g
o W
- I
- e i
H .O 3 - a o n _ U e u tN *
- c o o
. n x m
_ cn w _1 1? I t t I .
. I I 1 1 1 I i r t t I t t . . I i . 1 1 1 : i1 1. t L I io 01 8 9 k f. 'O
,.Ogx sl30.N f
P a
j lll2:3 3D= l T230W - GRAND GULF %
/
q
/ :
i ~ 7 o : /~ E 1, _ 1
/ -
Qc / 1 Tc' f _ l Nd m
, 3 t
t-Z ' t :
~
l L
- $. / 7 m o _
/ : A
~
i : : j ; : .i o - i A - O _ , ; l I
. s - ..
l 0 SO 100 t 150 j IINE HH Fig. B.37 Hole fraction of steam in Compartment B. l t i 1
r-
^
. 6w H
- : 0
~__ _ _
_: . - , - ~~ 5
'._ 1
. ~
n
' i o
F g L ' - e U
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o H c N l A i n R ' - i G _ - E 2 M O U
- '~ I I f W '
L o 0 ' s 3 ;i 0 .1 s 2 '.i*.,;\
- t' i:
- a M
T O 8 3 S . B
' g i
' F
- _,7- _ .
y 0 n n - .O n O_x m %Cm::E
O O l I 1 6 ' i I l ) l i l i [ i
DRAFT 9 SUPPLEMENTAL PLOTS FOR SEQUENCE T C 23 l l l
- xD T23C - GRAND GULF
-i 680-- . - - . - . ..- .--.-- -- .--
500I .-.t - - - - - - . .
. - - -. . ./
- i / i
- /;
/ '
_ 4002 - - - - . - - - - -- -
. ,/*- - - - - - + - - - - . - - - - + - --
$ i i j i d - '
/ ;
3 g 0_- - - . - - . - _ . . . _ . . . - - . . . _.3 . - . . . . o J .. I a f a Y 200-- j - -
---+------- m 1g0A - . . - . . . . .-.
- I
- / i ,
0- ...,g .,,.g ....g.M.,y, ,.,g ,.. ,,, ,,,,,,j
- 0. 00 0.50 1.00 ?. 50 2. 00 2.50 3.00 3.50 4.00 TIIE tHOURSI Fig. B.39 Total H2 generated.
T2.3C - GRAND GULF
~
688-- . r.. - 500- -
- I- .
i 3- 4- +- J
! ! i i i i !
~
400- - ! ! - !- -i m -
! i d -
o 300-
- - - - !- + -j J _ -
o a - m f 200- e C I
. ! i i
_ i i i i i i i 100- -i i- -i 0 --
,----------- ,' ,,,t,,,,l,,,
0 5 10 15 20 25 3D 35 4D 45 50 T it'E tHDuRSI Fig. D.40 Total H2 generated. H
- Zl3 T23C - GRAND GULF m
50- --- .- - 7.-.._--- -. .,. -
,s
/
.0_ - .. ; __ _ _ _ _ ;. __. -. .; .+.......;
/
b i 30 .
. 4 ... .
4 . ___ .j. .j . . 7--__.p.___~~__.4.... j w . b - i i i i i i x - g 20
. j.. . p.. . - ..,..,....?.
x 1 oo iob .. . . . . . . . _
. .. _ . .)
0 , , , . . . . i i .,i . i,,,,..,,,, ,,,,, 0.00 0. 50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 TIME (HOURS) Fig. B.41 Reactor vessel water level.
- _ _ _ _ _ = _ .
3r E.
,.4 tj: ::I i:!i- i !i: : j :S:i. i;i4j: ,i' ,0 5
.i t;i!i i.: i: :; ,+!, j5
. 4 -
.. :: 3- .:i i;i: ,i ! i0 N
4 U - . N . E . N!I3i L . . P.:i iS:i I;i+; ,+ ! ;5 F F ]: t-
. 3 L R ,
E . e U . P . r u G
. P .
t
.r;:ij U
.- 0
. . l:
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. 3 S u
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.+'
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. h. 5: :t;
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._ j/. . .
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- _ , 0 o, O s g 0 0 b 0 0 0 2 h 0 Q 0 0 0 0 8
h 4 7-0 8 Q 4 2 1 1 1 j 1 u..Wg Osa># gC )wg3F -ehA2W
. t a .
l ! a :I
3* l l 4 X2 l
" ,9uo i = 2 "
_ * . . * ,i:* :I;iii + . :: :i:i:i i.. ;3 ,8
. 5
- 5: - . l J :! 3 i
5 4
. r
; - .I: .i i 0 h 4 /
. u t
B F .
,5 e L -
.: 5: ii+. .ji !
3 r
- 3 u U - .
t c G R . u r E . t h i0 s D .
!: i. .i 0 i .
.+. I n
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- H R A -
u
. U O
t a R . Ga 5 H h e G - 5: -t. 4 - l - :! :$. i2 (
- . y
. 6
- c E
I d e I C . l- .i. M 4:! 9. i
. 2 0
T t c 3 . l l u
- l
. d 2 -
l o E. r . r T - t P u
. 5 p
!: :! o. i n
c 1 o i l E C,/ - i s s P -
/ .
i P
. )
F
.: .i: ?: ii,.
t i0 4 . t. .
._. 1
.if,. F ( .
3 4
. B
/ ,f i: i! e .
.i: , ;
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. . i
. F
- / .
- - . y 1
- - O 0 s e .. e : ; ,
8 e e .. e : ;
& d s .. n : ;
i a a 2. n 3, d i e :. e : 3 e d d e e : 3 4 2. 8. s
- 6. 4, 2, 1 1 1 E Z 3% : > m .W g 33 F o : E p. )c z )C p mI ><Omg s)(3aogg, zh,,E
l T23C - GRAND GULF l 8.60 ,. . . .,- - - . i~ i D.4B- -
. . l. --<
i i ! ! ! i ! g.20_ . . . . , 4 CI3 ! ! d ' a 0.00
.a o
o - to M -0.20- - i S i t
- 0. 4 0_ -
-4
-0.60 - - - - - - - - - - - - - - - - - - - - - -
,- ,ln,,j,,,-, ,,,,,,,,,,,,,,
0 5 10 15 20 25 30 35 40 45 50 IIME 'HOURSI Fig. B.44 Total C0 generated.
=O m
M
Z T23C -- GRf,ND GULF 1
-* s K
JS 0, 0 0 0._ . l
- i i i 300,800-~ . !. I- i-
- i i i 250,000} -
r ~!-- "I- "-i m 2gg,oea l .
..i.,
...J Q -
n, 150.000- - - - . m I : ui m 100, gee j. .j- . .s . .. - q d 50, 003 - 4
+- 1- -1 i
y
~
O , --------,- ,. p ,iryr,,--j ,. fi,,,,,,,,, 0 5 10 15 ?0 2S 30 35 40 45 50 11t-E tH0uRSl Fig. B.45 Mass of water in the pedestal.
mfu
>-iH
,: . ;' !4!;: j i . ji'
;0
. ,i !.
. _ 5
, . l!: '.
1 i: .;: , ,i' j 5 4
, .! lji
.i' i: .;: ;i. ii
;0 .
4 B t n
, e
!I .
l: .i: . . ;5 t m F , 3 r L , a p U , m o G l! . .l l: .;: ;i- ; i,0 I C
, 3 S n D , R i N r 0 u 2 H
A .I . 5 Ht R l2 f o G n E o N I i
-I. . ;0 t
.i: .
2 I c a C f r 3 2 e
,5 l T 1 o H
6 4
, :I iI . .
701 B g
- i F
l j: 5 hl-.l .i 1 l i ( ' ) j l 1
,.1 1 [- aJ, 1igid- J O -
0 D 8 o 0 0 c 0 0 6 4 2 e 9 6 4 2 0 1 1 1 i 0 0 o o 0 8 0 0 e 0 0 n o 0 0 0 0 s 0 0 s o. 0 O p 9~
llll2ll3 T23C - GR AND GULF
- 2a- - . . _.,
i H
-t i i O.153- .;. ! .I! .....
~
m ca . . . L : i ..;.. . _ . g O.10 0 . i i h 4 cn 4
- u. o3eJ .
..,. ..i. ..l. ..,. ... . i i
~
3 i i 0 030 ; - - ;- , ,....i t . ..i... i....j....i....l 0 5 10 15 20 25 30 35 40 45 58 TIPE (HOURSI Fig. B.47 Mole fraction of 0 in2 Compartment B.
I T23C - GRAND GULF
- a. e058 , . . . . . - -----.
-l . .
! t
- i f i i
S.t946- - - - - * > j- 4-i- i- -* G.3030- - i i . .
- l- ~ + g- -<
; i m,
- c. - . , i 0 -i .
H i
) e. med.
- r- :> - - . - - - + =
u. i , v. 1
- e. eetel I --.
J
~
s
- i 0.t00?- , ---H - - , , , . . r-j i .,-j i . . g . . . . j ..,,...,j 0 S 2C 15 2D 25 3D 35 40 45 58 TIPE tHOURS)
Fig. B.48 Mole fraction of CO in 2 Co...partment B. 3l::=
*i
'"=4
)
llllC T23C G R ANil GULF 892lm i 1 l
- 1. 0 0 -, -- - --
mer_
~
d rl J i
-l .
i $Q_ *
- I l
1 t' --1
.z t._,J w
4<-
- s. , m .: , +
m, 4 t.71, o i J
- .y o J: -
3 1
+ +
1
, , , ;,,,.j.,. i ,-- 7n,rirnrj
, . .. n. :. a. ;;. n. ,o 45 50 ill L til3uRS)
Fig. B.49 Hole fraction of steam in Compartment B.
T23C - GRAND f.ULF f.,808.600 i .
- 30. so - A'1 . . . 4 -
.- ) ;
. i. : :
i .. .... .. 3-' b, ! p, t .f;' i I e
; 3. J6 J. 3; d -- 1
-+
. . - 4 -
l
,1 l
5 -
$1 i ' '" '" l '!fh((1%
'I v (Q1 sy , _
r,
, t. l
- a. vos. vo. - l.,..
-l I
. , . . _ _ . ; .. . , . , , . ,_, _ 3_, __, p _,_ q _, ,,q i .
.. I5 / c- .25 ., 35 G 45 50 1 !! E I ill'_:URS )
Fig. B.50 Volumetric flow out of containment. I
-ul
- 23 T23C -
G R ANil GUi_F m 400,000 ,. l : , i ! J, a J , U U .' - . I. '
~
l i i l 3*??,000- .J '
. .i rIa .3 0, 000-~ , .
.. f
.J -
t.
.' . , 4 D -
i
-{ . .
i- . , g ,.
*** $i. ', j
- L
-l
{ us
; i
- I IJU. OCJ- ',
i e 7 ., .. ; t ', J ', 00,LisSO } * *, e
^
.o
. -l ' ~ -
g ,..__ ..,..,y,_,.,,_.__,_,_, . c
., . , 4., , , . q
- e r* l 'J [b f [e 3h I, h 5 a.* 4h #h j
I il L~ illD:.:l.'s i Fig. B.51 Mass of UO 2 in core region.
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