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Technical Rept 23.1GG Grand Gulf Nuclear Station-Integrated Containment Analysis. Jr Siegel 870313 Release Ltr Encl
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{{#Wiki_filter:- - IDCOR Program Report 1 TechnicalReport 23.1GG Grand Gulf Nuclear Station Integrated Containment Analysis March 1985 c'll).Y O, s ' '. ((p.) - J^ e[0 {) { ggf)O 416 gg PDR 0 /4 The Industry Degraded Core Rulemaking Prograni,5ponsored By the Nuclear Industry _ _..__ ______.._____._..__.-_ _ ___... _ -__ _,-_- -.-__.._.______....,__.----.-_.---.-------_----.--I

IDCOR Arizona Public Serivce Company New wrk Power Authority The Babcxk & IVilcox Company Niagara Mohawk Power Corporation Baltimore Gas and Electric Company Northeast Utilities Service Company Buhtel Power Corporation Northern Indiana Public Service Company Black & Veatch, Consulting Engineers Northern States Ikwer Compa::y Boston Edison Company Pacifc Gas and Electric Company C F Braum & Ce Iknnsylvania ikwer & Light Company The Cincinnati Gas & Electric Company Philadelphia Electric Company The Cleveland Electric illuminating Company Ibrtland General Electric Company Combustion Engineering, Inc. Public Service Company ofOklahoma Commonwealth Edison Company Public Service Electric and Gas Company Consolidated Edison Company ofNew Drk, Inc. Public Service Indiana Consumers Ikwer Company Puget Sound Puwer & Light Company Daniel Construction Company Rochester Gas and Electric Corporation The Detroit Edison Company Sargent & Lundy Duke Ikwer Company South Carolina Elutric and Gas Company Duquesne Light Company Southern Cahfornia Edison Company Ebasco Services Incorporated Southern Company Services, Inc. Exxon Nuclear Company, Inc. Stone & TVebster Engineering Corporation Florida Ibwer & Light Company Swedish State Power Board Fluor Power Services, Inc. Taiwan Ikwer Company General Electric Company Technical Research Centre ofFinland Gibbs & Hsil, Inc. Tennessee Valley Authority Gilbert / Commonwealth Companies Texas Utilities Generating Company GulfStates Utilities Company The Toledo Edison Company Houston Lighting & Ikwer Company Union Elatric Company llhnois Power Company United Engineers & Constructors Inc. Japan Atomic Industrial Forum, Inc. Virginia Electric and Power Company Kansas Gas and Electric Company IVashington Public Power Supply System Long Island Lighting Company IVestinghouse Elatric Corporation Middle South Services, Inc. (Visconsin Elatric Ikwer Company Nebraska Public Power District %nkee Atomic Elatric Company The IDCOR program is a large, independent technical effort sponsored by the nuclear industry. The Program is directed by a Policy Group comprised of representatives of the sponsoring organizations and operates under the corporate auspices of the Atomic Industrial Forum. The Program's purpose is to develop in an expeditious manner a comprehensive,in-tegrated technically sound position to assist in determining whether changes in regulation are needed to reflect degraded core and core melt accidents. Further information on the Program can be obtained by contacting John R. Siegel, Special Licensing Projects Manager /IDCOR, Atomic Industrial Forum,7101 Wisconsin Avenue,liethesda, MD 20814-4805, (301) 654-9260.

IDCOR Technical Report 23.1GG Grand Gulf Nuclear Station - Integrated Containment Analysis March 1985 by: Mississippi Power & Light Co. Gulfport, Mississippi i The Industry Degraded Core Rulemaking Program, Sponsored by the Nuclear Industry

NOTICE This report was prepared on account of work under contract to the Atomic Industrial Forum. Neither the Atomic In-dustrial Forum, nor any of its employees or members, the IDCOR Policy Group or the IDCOR or Atomic Industrial Forum consultants and contractors, makes any warranty, expressed or implied, or assumes legalliability or responsibili-ty for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately-owned tights. The opinions, conclusions, and recommendations set forth in this report are those of the authors and do not necessarily represent the views of the Atomic Industrial Forum, Inc., its employees, or the IDCOR Policy Group, its members, or the Atomic Industrial Forum or IDCOR Policy Group consultants or contractors. Because IDCOR is supported in part by Federal funds, the following notice is required by Federal regulations: The Atomic Industrial Forum's IDCOR activities are subject to Title VI of the Civil Rights Act of 1964, which prohib-its discrimination based on race, color, or national origin. Written complaints of exclusion, denial of benefits, or other discrimination of those bases under this program may be filed with (among others) the Tennessee Valley Authonty (TVA), Office of EEO,400 Commerce Avenue EPB14 Knoxville TN 37902, and must be not later than 90 day 5frem the dare of the alleged diurimination. Applicable TVA regulations appear in part 302 of Title 18, Code of Federal Regulations. Copies of the regulations, or further information, may be obtained from the above address on request. Copyright O 1985 by Atomic Industrial Forum, Inc. 7101 Wisconsin Avenue Bethesda, MD 20814-4805 All rights reserved.

Acknowledgements The lead authors of this report were Michael Lloyd, Middle South Services, Inc., and Robert E. Henry and Jeff R. Gabor of Fauske G Associates, Inc. The support and cooperation of Jim Haley, Mississippi Power and Light Co., is greatly appreciated. The IDCOR Program would like to acknowledge the support and cooperation of the above organizations. In particular, the technical guidance contributed by Dr. Edward-L. Fuller, the i l technical manager for this project assigned to IDCOR from EPRI, is very much appreciated. O I I i i TABLE OF CONTEriTS 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 Nuclear System.................... 3-1 3.1.2 Containment 3-3 3.2 Modular Accident Analysis Program (MAAP) 3-4 3.2.1 MAAP Nodalization.................. 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 3-16 4.0 PLANT RESPONSE TO SEVERE ACCIDENTS................ 4-1 4.1 Plant Response to the T QUV Accident ............4-3 j 4.1.1 Sequence Description................. 4-3 4.1.2 Primary System and Containment Response....... 4-3

- ii - TABLE OF C0llTEtiTS (Continueo) 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 T QW Accident 4-27 23 4.3.1 Sequence Description................. 4-27 4.3.2 Primary System and Containment Response 4-27 1 4.4 Plant Response to the T C Accident. 4-34 l 23 4.4.1 Sequence Description................. 4-34 4.4.2 Primary System and Containment Response 4-38 4.5 Reference.......................... 4-48 5.0 PLANT RESPONSE WITH OPERATOR ACTION 5-1 5.1 Plant Response to the T QUV Accident with Operator Action.. 5-5 3 5.2 Plant Response to the AE Accident with Operator Action... 5-7 5.3 Plant Response to the T 0W Accident with Operator Action.. 5-16 23 5.4 Plant Response to the T C Accident with Operator Action 5-20 23 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION 6-1 6.1 Introduction........................ 6-1 6.2 Modeling Approach...................... 6-1 6.3 Sequences Evaluated..................... 6-3 6.3.1 T QUV Sequence.................... 6-3 j 6.3.2 AE Sequence..................... 6-9 6.3.3 T QW Sequence.................... 6-10 23 6.3.4 T C Sequence.................... 6-10 23

- iii - TABLE OF COMTENTS (Continued) i Page d 6.4 References......................... 6-11 7.0

SUMMARY

OF RESULTS........................ 7-1 .3 1 7.1 WASH-1400 Comparison Case Analyses 7-1 7.2 Operator Action Analyses 7-5

8.0 CONCLUSION

S........................... 8-1 APPENDIX A.1 - Grand Gulf Parameter Fil e.............. A-1 APPENDIX A.2 - MAAP Input Files for Section 4 Sequences...... A-17 l j APPENDIX B - Supplemental Plots for the WASH-1400 Comparison Sequences....................... B-1 i Supplemental Plots for Sequence T)QUV......... B-3 Supplemental Plots for Sequence AE .......... B-17 QW......... B-31 Supplemental Plots for Sequence T23 C......... B-45 Supplemental Plots for Sequence T23 _ _. ~,.. _,. - _ _ _ _.. _

_g. LIST OF FIGURES Figure No. Page 3.1 BWR primary system................... 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 pedestal....... 4-20 4.9 Concrete ablation 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 Compartment B . 4-26 4 4.14 Temperature of the suppression pool . 4-30 4.15 Temperature of gas in Compartment B 4-31 4.16 Pressure in Compartment B . 4-32 4.17 Temperature of gas in the drywell . 4-35 4.18 Average corium temperature in the pedestal....... 4-36 4.19 Concrete ablation depth in the pedestal . 4-37

- vi - LIST OF FIGURES (Continued) Figure No. Page 4.20 Average core power................... 4-40 4.21 Reactor vessel water level............... 4-42 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 5.1 Downcomer water level 5-8 5.2 Suppression pool temperature.............. 5-9 5.3 Compartment B pressure................. 5-10 5.4 Cori um tempera ture................... 5-13 5.5 Suppression pool temperature.............. 5-14 5.6 Compartment B pressure................. 5-15 5.7 Suppression pool temperature.............. 5-18 5.8 Compartment B pressure................. 5-19 5.9 Downcomer water level 5-23 5.10 Suppression pool temperature.............. 5-24 5.11 Compartment B pressure................. 5-25 B.1 T QUV - Total in-vessel H2 generated........... B-4 j B.2 T QUV - Total H2 generated................ B-5 j B.3 T QUV - Reactor vessel water level............ B-6 j B.4 T QUV - Temperature of structure. *F........... B-7 j B.5 T 0VV - Fission product decay power on structure, 3 Btu /hr.......................... B-8

- vii - LIST OF FIGURES (Continued) Figure Mo. Pace B.6 T QUV - Total CO genera ted................ B-9 j B.7 T QUV - Mass of water in the pedestal .......... B-10 j B.8 T)QUV - Mole fraction of H2 in Compartment B....... B-ll B.9 T QUV - Mole fraction of 0 in Compartment B....... B-12 j 3 B.10 T QUV - Mole fraction of CO in Compartment B ...... B-13 j 2 B.ll T QUV - Mole fraction of steam in Compartment B ..... B-14 j B.12 T)QUV - Volumetric flow out of containment........ B-15 i n core region............ B-16 B.13 T)QUV - Mass of U02 B.14 AE - Total in-vessel H2 generated ............ B-18 B.15 AE - Total H2 generated................. B-19 B.16 AE - Reactor vessel water level ............. B-20 B.17 AE - Temperature of structure. *F ,........... B-21 B.18 AE - Fission product decay heat on structure, Btu /hr... B-22 B.19 AE - Total C0 generated................. B-23 B.20 AE - Mass of water in the pedestal............ B-24 B.21 AE - Mole fraction of H in Compartment B ........ B-25 2 B.22 AE - Mole fraction of 0 in Compartment B ........ B-26 2 B.23 AE - Mole fraction of CO in Compartment B........ B-27 2 B.24 AE - Mole fraction of steam in Compartment B....... B-28 B.25 AE - Volumetric flow out of containment ......... B-29 l B.26 AE - Mass of UO in core region ............. B-30 2 B.27 T QW Total in-vessel H2 generated............ B-32 23 B.28 T QW Total H2 gene ra ted................. B-33 23 B.29 T QW Reactor vessel water level............. B-34 23

- viii - LIST OF FIGURES (Continued) Figure No. Page B.30 T 0W Temperature of structure. *F............ B-35 23 B.31 T QW Fission product decay heat on structure, 73 B ta/ hr.......................... B-3 6 B.32 T QW Total C0 generated................. B-37 23 B.33 T QW Mass of water in the pedestal ........... B-38 23 B.34 T QW Mole fraction of H in Compartment B........ B-39 23 2 j B.35 T QW M le fraction of 0 in Compartment B........ B-40 23 2 j B.36 T QW M le fraction of CO in Compartment B ....... B-41 23 2 B.37 T QW M le fraction of steam in Compartment B ...... B-42' 23 B.38 T QW Mass of UO in core region............. B-43 23 2 i B.39 T C Total in-vessel H2 generated ............ B-46 23 1 B.40 T C Total H2 generated................. B-47 23 f B.41 T C Reactor vessel water level ............. B-48 23 B.42 T C Temperature of structure. *F ............ B-49 23 B.43 T C Fission product decay heat on structure, Btu /hr... B-50 23 B.44 T C Total CO generated ................. B - 51 23 f B.45 T C Ma ss of water in the pedestal............ B-52 23 { B.46 T C Mole fraction of H in Compartment B ........ B-53 23 2 B.47 T C Mole fraction of 0 in Compartment B ........ B-54 23 2 B.48 T C Mole fraction of CO in Compartment B........ B-55 23 2 B.49 T C Mole fraction of steam in Compartment B....... B-56 23 B.50 T C Volumetric flow out of containment ......... B-57 23 B.51 T C Mass of U0 in core region ............. B-58 23 2 I 2

-ix-LIST OF TABLES Table No. Pa7e 3.1 Initial Inventories of Fission Products and Structural Materials Released from the Fuel 3-10 4.1 Grand Gulf Nuclear Station, T OUV - WASH-1400 i Comparison Case, Accident Chr6nology.......... 4-4 4.2 Effects of Cepressurization in the T)QUV Accicent 4-16 4.3 Grand Gulf Nuclear Station, AE WASH-1400 Comparison Case, Accident Chronology 4-18 4.4 Grand Gulf Nuclear Station, T 0W - WASH-1400 Comparison Case, Accident Chr$2nology.......... 4-28 4.5 Grand Gulf Nuclear Station, T C WASH-1400 Comparison Case, Accident Chr0$ ology.......... 4-39 7 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 0VV - Operator 3 Action Case, Accident Chronol 6gy.. 5-6 5.5 Grand Gulf Nuclear Station, AE - Operator Action Case, Accident Chronology. 5-12 5.6 Grar.d Gulf Nuclear Station, T 0W - With 73 Operation Action, Accident Chronology . 5-17 5.7 Grand Gulf Nuclear Station, T C - Operator 73 Action Case, Accident Chronology............ 5-21 6.1 Distribution of Csl and Cs0H in Plant and Environment (Fraction of Core Inventory)........ 6-4 6.2 T QUV Fission Product Release 6-5 j 6.3 AE Fi ss on Product Release............... 6-6 6.4 T 0W Fission Product Release 6-7 23 6.5 T C Fission Product Release.............. 6-8 23

-x-LIST OF TABLES (Continued) Table flo. Page i 7.1 Summary of Fractional Radionuclide Releases to the Environment.. 7-2

1-1

1.0 INTRODUCTION

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 containment 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 which are covered in Task 23.4. This study also includes assessments of the results of a limited set of operator interventions in these secuences. 1.2 Relationship to Other Tasks The primary system and containment response analyses of IDCOR Subtask 23.1 are carried out with the Modular Accident Analysis Program. This includes models developed in IDCOR Subtasks 11,12,14,15, and 16 for ther-mal-hydraulic behavior as well as fission product release, transport and deposition within both the primary system and containment. The accident sequences used for the analyses along with the operator interventions were developed by considering the dominant core melt accident sequences identified in Subtask 3.2 (Assess Dominant Sequences) and the physical orocesses occur-ring during these accidents. It should be noted that the analyses developed as part of 10COR Subtasks 16.2 and 16.3 involve the detailed consideration of many different phenomena which are themselves considered in separate 10COR 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 15.2) and core-concrete interactions (Subtask 15.3).

l-2 The ultimate structural capacility of the containnents associated l with the reference plants and other typical designs were assessed in ICCOR Subtask 10.1. Task 10.1 analyses define the containtrent failure pressure and 1 j failure mode assumed in the Subtask 23.1 analyses for tnose sequences result-l ing in containment failure cue to high internal pressure. Calculations of the rate and amount of fission products released i from the containment, for those sequences which result in containment failure, l were supplied to IDCOR Subtask 18.1 (Atmospheric and Liquid Pathway Dose) to l formulate assessments of the health consequences associated with these postu-l lated accident scenarios. These health consequence analyses were then sup-l plied to IDCOR Subtask 21.1 to evaluate the risk reduction potential for possible mitigating operator actions and containment mitigative design features. Detailed considerations for each of the related subtasks can be found in the final reports submitted for the specific task. Individual issues are addressed in this report only as required to understand the specific j behaviors obtained for the accident sequences considered. Operator intervention sequences were developed as part of Subtask 23.1 and applied to the specific accident sequences in the Grand Gulf fluclear I Station design to determine those potential actions which could terminate the accident sequence and result in a safe stable state. These results ware used 4 j in IDCOR Subtask 22.1 (Safe Stable States) which discusses the safety of a BWR l in terminating the various core damage sequences considered for t e Grand Gulf i fiuclear Station design. I i 4 1 ) l I i 1 _. _. - - _ _. - - _ _ - - - _ _, -. - - ~---,--- _ __ _._ - _ _ -. _ _ _

2-1 2.0 STRATEGY AND METHODOLOGY The basic strategy of this subtask was to analyze accident sequences which have been previously identified as dominant or key potential contribu-tors to core melt frequency in previous PRAs. These analyses consisted of plant thermal hydraulic response and fission product transport calculations for accident sequences which were postulated to lead to core degradation and mel ting. These analyses model performance of the ECCS systems and the con-tainment engineered safety systems, such as the suppression pool, decay heat removal system, etc. The MAAP code [2.1] was used to perform the primary system and containment thermal-hydraulic response analyses. This code considers the major physical processes associated with an accident progression, including hydrogen generation, steam formation, 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 product release model to determine the transport of vapors and aerosols throughout the primary system and containment. Fission product deposition processes modeled include vapor condensation, steam condensation and sedimen-tation. For each of the four GGNS accident scenarios selected for analysis, thermal-hydraulic calculations were performed both with and without selected l operator actions during the accident. The WASH-1400 comparison case analyses, which assume only minimal operator response during the accident, establish a reference system response during each of the accident scenarios. The " opera-tor action" analyses are branch calculations of the WASH-1400 comparison cases. These operator intervention cases demonstrate the effect of selected operator actions on the progression 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].

2-2 2.1 References 2.1 "MAAP, Modular Accident Analysis Prcgram 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 15.1B, 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.

3-1

3.0 DESCRIPTION

OF MODELS AND MAJOR ASSMPTIONS The Modular Accident Analysis Program (MAAP), Ref. [3.1] is used to model the Grano Gulf Nuclear Station (GGNS) response to postulatec severe accidents. This code includes containment response, fission product release, and fission product transport. 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 i 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 thermal 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 General Electric Company. Each reactor is housed in a steel-lined reinforced concrete fiark III containment building. 3.1.1 Nuclear System l l 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 perform two general functions: core cooling and contain-ment pressure control. Those systems which perform the first function include the reactor core isolation cooling (RCIC) system, the high pressure and low pressure emergency core cooling systems (ECCS), the automatic depressurization system (ADS), control rod drive (CRD), feedwater, condensate, fire pumps, service water, and the standby liquid control (SLC) system. The containment pressure control function is accomplished via the suppression pool makeup system, the drywell purge system, the post-LOCA vacuum breakers, the

3-2 suppression pool cooling and containnent spray moces of the resicual

r. eat removal (RhR) system, and the hydrcgen igniticn system.

The primary system consists of the equipment and instrumentation necessary to produce, contain, and control the steam power 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 normal operation and in the event of fuel failure. The major RPV internal components consist of the core, the shroud top grid, core plate, steam separator and dryer, jet pumps, control rods, and control rod drive guide tubes. The core is composed of approximately 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 fuel pellets, 2 with the Zircaloy-2 cladding providing both structural support and a fission product barrier between the fuel and the primary system water. The remaining reactor pressure vessel internal components support and align the fuel and provide the water circulation flow paths 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 reactor water recirculation system provides a forced continuous internal circulation of coolant water through the core. Four main steam lines direct 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 reactor vessel overpressure protection, with their discharge being directed to the suppression pool. l

3-3 The majority of the primary system data used in this analysis came from the Granc Gulf FSAR [3.2]. This infcrmation includes initial conditiens, pressures, temperatures, flow rates, enthalpies, masses, system pressure setpoints, control logic, and other parameters. A plant parameter file fer MAAP was prepared based on these data; it appears in Appendix A l. 3.1.2 Containment 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 may be released as a result of a Loss of Coolant Accident (LOCA). The Grand Gulf Mark III containment is a steel-lined rein-forced concrete structure, with a cylindrical shape, topped with a hemispheri-cal dome. The containment foundation is a thick, circular reinforced concrete slab. The Mark III containment building is a pressure-suppression design. Its inner volume and an outer volume are 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 con-crete cylinder called the pedestal. The drywell and outer containment volumes consnunicate 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 provides for steam suppression during postu-lated LOCA events. The outer containment volume 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 additional means of rapidly removing possible post-accident steam and/or fission products from the outer containment atmosphere. In addition to these features, hydrogen igniters are located in both crywell and outer containment volumes to control hydrogen accumulations following postulated severe accidents. 3 MAAP input data, including initial conditions, heat transfer coeffi-cients, exposed surface areas, and flow areas between volumes are based on information from the Grand Gulf FSAR [3.2], and architect / engineer drawings. These data appear in the MAAP parameter file listed in Appendix A.I.

3-4 3.2 Modular Accident Analysis Program (ftAAP) Within the IDCOR Program, the phenomenological models developed in Tasks ll,12,14 and 15 have been incorporatec 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 deposition 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 Nodalization ) 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 l l 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. Modeling of the primary system is used to determine if a given 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 spectrum of accident scenarios considered requires the specification of pump curves, valve set points, system logic, etc. With the specification of the accident sequence, the primary system model determines the vessel water

3-5 SRV J, MAIN STEAM UPPER PLENUM j STEAM " TO ECCS vTURBINES ~ V RPV INJECTION d# UPPER DOWNCOMER (FW,ECCS, / /\\ HPSW, ETC.) r ECCS J SPRAYS II LOWER CORE DOWNCOMER E A ) N'Y RECIRC PUMP LOWER PLENUM , see SCC EW Re CY Com C00biNG SYSTEW "N.i !!$$?[b#,'.'un'"'"" CRD FLOW Fi g. 3.1 BWR primary system.

3-6 inventory, including the boiled-up level in the core, to evaluate the poten-tial for core uacovery. 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 CRD flow. If available, this flow can limit core damage for long-term heat removal failure events. The Grand Gulf Mark III 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 Compartment 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 r.od drive (CRD) flow from the condensate storage tank, main steam lines, main steam isolation valves (MSIVs), turbine bypass, feed-water, reactor core isolation cooling (RCIC), high pressure core spray (HPCS), low pressure coolant injection (LPCI) and other residual heat removal (RHR) system modes, low pressure core spray (LPCS), and standby liquid control system (SLCS). In addition to these plant systems, MAAP nodalizes both the l

3-7 / ' g. j':',i' M/ pi I,, /. / / /, ,.,j / / '/< COMPARTMENT B Y F (, b / a

AE - GPAND GULF 7- ? 6 p.. .[ 5_ y 4 p. / W w4, w C-3_ w M 4, w 'N 2_ ....,.. fl.. _.. -s ~ 3/ y e ,,,,,,,,,,,,,,,,,,,,i 6 10 20 30 40 -50 60 78 TIME (HOURS ) Fig. 4.9 Concrete ablation depth in the pedestal.

4-22 CRD flow keeps the debris quenched until the CST runs out of water at '22.3 hours. Without replenishment, the pedestal water boils away ar.d. by 2C hours the debris begins to reheat. Concrete ablation in the pedestal resumes at 30 hours. The thermal decomposition of the pedestal concrete floor and walls l 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 continuous burn-off prevents high concentrations of combustible gases from occurring. By 43 br, 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 compartments, burning ceases and the containment becomes self-inerted. Drywell temperatures rise to about 1000 F af ter the core debris-concrete interaction resumes in the pedestal, as shown on Fig. 4.10. The suppression 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 B remain relatively low due to the cooling e#fect of the suppression pool (as shown in Fig. 4.12). At 58 hours into the event, the GGNS containment reaches 71.3 psia (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 25 psid of atmospheric pressure in about 10 hours. And, the suppression pool remains intact following the containment failure event. Since the pool temperature is nearly 280 F at the time of the containment failure, about 2% of the pool inventory is calculated to boil away within 10

4 e. AE - GRAND GULF .i t i f },2@@ i: 1 i 1.900-b l c s%i r / { .. f... ........_t .r.

. g a _
/ = l c' O 600- ,G l ,1..,@t'4 = 2 m W 9 d.._. .-j..-..........r .- v p j 4@h P g..a, ,n' l 1.f .-.-.i-e -F i 200-1 1 I l l o e is 20 30 49 50 60 78 4 tit E ' (HOURS 1 Fig. 4.10 Temperature of gas in the drywell. 4 4 4
._m b i 4 4 L AE - GRAND GULF I-i t 1 300-i t 1 __ s 256 _ /. / 1 a ,200- .. -. ~ G 1 c: -t w - 150 1 / ~ -s x., E* a r .t 1hh L... _ N =~ i ,,4 A A 4 d as = -, 3, g,,,,. i i g1 i i i.... ,i i o 10 20 30 40 50 ~60 78 TIME (HOURS) i r, Fig. 4.11 Temperature of the suppression pool. I 1 6 h l 1 i i
AE - GRAND GULF 3gg_ [4 f-250-f~~ t ._ q ~ _-- I ,.~. _ 200-r' - x / / e. W a
150i,
,/- C ./ / x a l 1 g g b;', - +. a 1 1 i J saq -t 4 0. i.. ,,,,,,,,,,,,,i ,,,i e 10 20 30 40 50 60 70 tit E IHOURS) Fig. 4.12 Temperature of gas in Compartment B.
._.~ _.. _ - _ _ _. _ _. i t i i i. AE GRAND GULF 1 i j 88-, 7.. 1 i e 78- + / -- / / i 6 0,- - - - + - - -- /-- i / ~ /p - 3ed .....f ~ ~ / . ;\\..... _. .e. / . s\\ j
/
.o . - sf ' -..._, ~.-..._ -..-. L.
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- -+ .i. J" i I 2eq. i J 0 ...,....,....i....,...,, 0 le 28. 30 40 50 - 60 79 i TIK (HOURS 1 Fig. 4.13 Pressure in Compartment B. t 1 0 w 4
4-27 l l l hrs following failure. Appendix B includes acditicnal plots of results for this sequence. QW Accident 4.3 Plant Response to the T23 4.3.1 Sequence Description The T 0W accident is assumed to occur during full-power operation. 23 4 It is assumed to be initiated by inadvertent main steam isolation valve (MSIV) closures (Event T23). The main feedwater and main condenser are assumed to be unavailable (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 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. llo credit is taken for any operator action other than to start the containment igniter system at the accident initiation and to manually depressurize the RPV when the suppression pool temperature exceeds 145*F. The T QW accident chronology is provided on Table 23 4.4. i 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 suberitical by 3.7 sec into the event. The core power remains at decay heat levels for the remainder of the event. The HPCS system autoactuates on low RPV level at 28 sec and the RCIC autoactuates at one minute into the event. Both systems transfer suction to the suppression pool at 1.1 hr. At 2.35 hours into the accident the suppression pool temperature reaches 145*F and the reactor is modeled to be manually depressurized via the automatic depressurization system (ADS). RCIC trips off after this depressurization. n ne
4-28 Table 4.4 GRAND GULF UUCLEAR STATION T 0W - Wash-1400 COMPARISON CASE 23 ACCIDENT CHRONOLOGY i i 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 4 1.1 hr HPCS and RCIC systems transfer suction from CST 1 to SP I 2.35 hr Suppression pool temperature exceeds 145*F, manual ADS 4.1 hr High DW pressure LOCA setpoint reached; DW purge system actuates; LPCS and LPCI actuate 4.6 hr SPMU actuates 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 l l 48.8 hr Core begins to uncover ) 54.1 hr Fuel mel ting begins 56.2 hr Core plate failure followed by vessel failure 4 [ ,,,,.,,-.,-,,.---a +,.,, - = -, - - -,. n.,- w
1 4-29 l At 4.1 hr, steam pressurization of the containment building causes a high drywell (DW) pressure LOCA signal. This signal is a permissive signal for the DW purge system, the SP makeup (SPMU) system, and the automatic depressurization system (ADS); it is an actuation signal for the low cressure core spray (LPCS) and low pressure coolant injection (LPCI) systems. The DW purge system actuates af ter a 30 sec time delay and the SPMU system actuates the upper containment pool dump following a programmed 30 min delay. The RPV 1 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 the early part of this event (Fig. 4.14). One i exception to this trend occurs at 4.6 hr, when the SP makeup systen releases relatively cold upper containment pool water into the SP. After the upper I pool dump, the SP water temperature co ltinues 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 temperatures. After the loss of the RCIC, the HPCS and the CRD flow continue to maintain adequate RPV inventory. At 22.4 hr the CST empties and the CRD flow ceases. From this point on, only the HPCS is available to maintain 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 definition assumes that the RHR heat exchangers are unavailable. Thus, the operation of containment sprays removes no heat from the containment; it merely homogenizes tempera-tures in the outer containment. The effect of this homogenization can be observed in Figs. 4.15 and 4.16. The containment spray actuation causes the outer containment air temperature to increase, and, consequently, the outer containment pressure to increase 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 af ter the vessel fails. At that time, trains A and B of the residual heat removal (RhR) system auto-matically switch into their spray mode, i
1 ) .1 i I i T230W - GRAND GilLF f a e O i X 2 i e 1 2 l
l kn L
f', i /\\/ __ N i 1g n 4 w O 1 E f .........I,........l ........I,........l........l.........l.........l.........l.........l.........I o. o.20 0 40 0.60 0 80 1 1.2 1.4 1.6 18 20 l Tite llR xlO ' 4 Fig. 4.14 Temperature of the suppression pool. 1 t 1
I! .E ., i. [. i-
,i t
ot.w-I
0. 3
" 2O l x - 8 I 1 B t - 6 n I e 1 m t r a - 4 m p o I F 1 C L n U i G " 2 s D I ag 1 R N f t A o R .s 1 G I E e r t u I T a t r W = e n I 8 " 0 p 0 m 3 e 2 ~ " 0 T T ~ - 0 5 1 - 6 .p - 0 I 4 r g i " 0 F v > 4 l f > 0 } o 0 7 o!. d. u.0 ,s.,
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/e e ]. e i e u i. a m r i i 0 = H u ,3 a O 1 I ') _e ~ e t (N 1 s- ? i 2 1
^
E .e.,n 2 LA. .y, ~
o I
N., ,o i 2 s
R
.., ~ 2 i o 4 2 i I,,1,,,,,,1,,,,,,,,,1,,,,,,1., t...,,,, I,,,,,,i,,a. 4o t 11 6' 4 5 C I I i Olx VISd 83d i i t i t t E l i .m m ,,,-------m---m--.-----,.-,#---.,~,w,~,v--,e.., .r-.w,vvm,er,g-++rf*-wem,<---~
~. 4-33 t j At 40 hr into the event, the GGl15 containment pressure reaches 71.3 f psia. The containment is modeled to fail at this pressure at a location just i below the junction between the cylincer and the come. The failure cause is 2 steam overpressurization. A containment breach area of 0.1 f t was rodeled. For this containment failure size and location, the containment pressure remains in the 53 psia range after containment failure. The containment pressure begins to drop again much later. The suppression pool remains intact l following the containment failure event. Suppression pool boiloff maintains an elevated containment pressure af ter the containment fails. Gas tempera-4 I tures in all containment compartments are relatively constant at about 300 F until long (60 hr) af ter containment failure. l 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 i containment failure location, pressure, fluid flow loading, and ECCS pump suction temperature [4.1], pressure, and NPSH limitations [4.2] indicates that at least one ggt 45 ECCS train should survive a containment failure event. jl However, for this analysis, the conservative assumption that all ECCS equip-1 ment fails on containment failure was made, f Without vessel makeup, the RPV water level falls. The decrease is j relatively slow in comparison with the T QUV and AE events, since decay heat j j levels in the T 0W accident are relatively low. Core uncovery takes place 23 ] about 9 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. f Af ter 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 l l
4-34 i slumping into the lower plenum. The core debris tnen falls to the bottom of l the RPV and, about 30 sec later, vessel failure is assumed to occur at a welceo RPV penetration. At vessel failure, the molten fraction of the lower j plenum core debris falls onto the pedestal floor followed by the lower plerum j water. l 2 Since the containment failure size was 0.1 ft, 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. At 100 hr into the event, the water covering the core debris in the pedestal boils away. As a result, the containment pressure drops to the 30 psia range and the drywell gas temperature rises from about 400'F to the 800*F range. These effects are shown on Figs. 4.16 and 4.17. The gas temperatures in the outer containment I compartments remain relatively constant at about 300 F during the event. Since the containment has such large amounts of steam, it is effec-tively inerted when the hydrogen leaving the vessel enters the wetwell (prior l to vessel failure) and the drywell (after vessel failure). Hence, no burning I is predicted to occur inside the containment. 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 C Accident 23 4.4.1 Sequence Description The T C accident is assumed to occur during full-power operation. 23 ] It is assumed to be initiated by an inaovertent main steam isolation valve (MSIV) closure (Event T23). The accident sequence specifies that the control rod drive (CRD) system fails to automatically bring the reactor subcritical and all other means of inserting negative reactivity fail (i.e., manual scram, SLCS, etc.), (Event C). This analysis assumes that no control rods were l inserted into the core. All other plant systems, except the normal feedwater
- -. - _ - - -. -... -... ~ -. i o i i .l i T230W - GRAND GULF [ l O~ x 1 m i i
\\
4 4 k 1 L. 6 i f E i 3 o J / j / D ~- ? w 1 m j%~cer'[ 2 n f'- i / ./ 11111.1 Mitt 11e e a t ! r a n i t s e a n l n a a a e a e a a lg t a e e a a a n ! n a e n e a a n l n a t a e s e e s l n a e a e i a n l: e e a e a e a n l a a n a a a a a n l } m J 0 0.20 0.40 0.60 0 80 I l.2 14 1.6 18 2.0 2 i Tite tR xlO 1 1 l Fig. 4.17 Temperature of gas in the drywell. A f i i
1 1230W - GRAND G4K.F m-o x i j t-La In g g s N n __- =. G l ] Ts........i.........l.........l.........l........l........l........I........l.........l.........I 0 0.20 0 40 0.60 0 80 1 1.2 1.4 1.6 18 20 TIME IR x10* Fig. 4.18 Average corium temperature in the pedestal.
1 I T230W - GRAND LULF O-ra L 4 p.- w W l O 2 UM i e L sw N 4 [ 1 N f I........l.........l.......= !=--- s ....l.........l.........l.........I.........I 0 O.20 0.40 0.60 0 80 1 1.2 1.4 1.6 18 20 i 2 TIE IIR xlO l l Fig. 4.19 Concrete ablation depth in the pedestal. I
4-38 system, are assumed to be available. No credit is taken for any operator ] action other than to start the containtrent igniter system at the accicent initiation and to manually decressurize the reactor pressure vessel when the suppression pool temperature exceeds 145 F. The T C accident chronclogy is 23 provided on Table 4.5. 4.4.2 Primary System and Containment Response 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 i core power is assumed to decrease from its initial full-power level to about 20% of full power level within seconds. This power reduction simulates the i thermal-hydraulic reactivity feedback effects which are expected to occur as a j result of the initiating MSIV closure event, the resultant recirculation pump j and feedwater trips, and the ensuing high pressure core spray (HPCS) and i reactor core isolation cooling (RCIC) systems actuations. The estimate of 20% j 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 20% to 6% of full power as the downcomer water level decreases from 7.2 ft above the j active core to the top of the jet pumps. However, the event analysis still assumes HPCS and RCIC maintain a high water level with correspondingly high j power. Decay heat power levels are assumed for uncovered fuel nodes. The l T C core power history is provided in Fig. 4.20. 23 l 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 I (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 l generate automatic activation of 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, j suction for these systems is automatically transferred from the condensate storage tank (CST) to the SP on a high SP water level signal. At 8 min, when l
. =. 4-39 Table 4.5 GRAriD GULF fiUCLEAR STATI0t1 1 T C WASH-1400 COMPARIS0:1 CASE ] 23 ACCIDEflT CHR0tiOLOGY l I Time Event g; 0 sec Initiating events: MSIV closures; Failure to scram; Loss of main feedwater 1 33 sec RPV Level 2 LOCA setpoint reached 49 sec HPCS begins operating l 52 sec RCIC begins operating j 4.5 min HPCS/RCIC systems transfer suction from CST to SP l 8 min Depressurization manually initiated 18.3 min RCIC pump assumed to fail on high suction j temperature l 23.0 min High DW pressure LOCA setpoint reached; Post-LOCA J DW vacuum breakers open j 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 j l 53.1 min SPMU actuates i 1.0 hr Containment failure and subsequent ECCS failure l 1.3 hr Core begins to uncover I 3.0 hr fuel melting begins I 3.8 hr Core plate failure followed by vessel failure i i l _ _ _. - _. _,, _. -. _. _ _ _ _ _. - _. _ _ _ _.. _. - -. _ ~. _ _ _ -,. _.,
T23C - GRAND GULF z. ;.z, ;, 1 l t l l 2.'E +
3.J T-
--i l J l 4._ g.13 a i 1 = s '3E.33-4 e x_ a ;:.. > ; -. .a a. e = o 4 ,c , ;;c.;3. rq.'.. \\ YA. d .i i i ,,,,,,,,,,,i., i, 3 re c 53
aJ 1.50

2. c0
.. 50 3.00 3.50
4. 00 4.50

5. 00
'I!E LHOURS) Fig. 4.20 Average core power.
4-41 l the suppression pool temperature reacnes 145*F, the RPV is manually cepres-j surized according to emergency procedure guidelines. Because the core power f generation rate is much greater than the cecay heat level, the SP water temperature rises very rapidly. When the SP temperature reaches 200*F at 13.3 min, the RCIC pump is assumed to fail. As seen in Fig. 4.21, water level is maintained with the combination of HPCS, LPCS, and LPCI until containment l failure. The SP reaches saturation conditions at about 20 minutes into the i event and afterward is no longer able to completely quench the steam exiting j the RPV; a steam-pressurization of the containment ensues. The rising sup- ] pression pool water temperature and the resulting rise in ressures and j temperatures in both the drywell and outer containment can be seen in Figs. I 4.22 through 4.25. The rising pressure actuates the 1.89 psig high DW pres-sure LOCA signal at 23.0 min. This signal is a permissive signal for the DW l purge system, the post-LOCA DW vacuum breakers, and the SP makeup (SPMU) system; it is an actuation signal for the low pressure core spray (LPCS) and low pressure coolant injection (LPCI) systems. Since the post-LOCA DW vacuum breaker permissive requiring a 0.87 psid drywell vacuum relative to the i wetwell is already satisfied, the vacuum breakers open imediately. The DW t { purge system actuates af ter a 30 see time delay and the SPMU system actuates the upper containment pool dump following a programmed 30 min delay. The continuing HPCS injection maintains RPV level. At 26.2 min into the event, I j the containment pressure reaches the 9 psig containment spray actuation pressure setpoint. At that time, trains A and B of the residual heat removal 4 f (RHR) system automatically switch into their spray mode and eight minutes l later begin to spray SP water into the upper containment volume. Since the l secondary side of the RHR heat exchangers require manual alignment, which was not modeled in this analysis, the containment spray system is unable to effect j a containment pressure reduction. j 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. 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 1.0 hr into the accident, I
t t T23C - GPAND GULF e i t e 4 i s' t a ~. _ _ q [ .s t 1 TOP OF ACTIVE FUEL t
? ;3 _ _ _ _ _ _ _ _.; _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _
m 1 i N'
m. N
? N I
... ?.4 a
2 N ~ BOTTOM OF ACTIVE. FUEL I 16: _ 6 .i i 1 .g... g .g.. .g.. .g. ..g....y
20 c. t.'
aJ id
2. 00
'2.50
3. ed

3. 50 4.00 4.50

5. 3d Tli'E IHOURSI Fig. 4.21 Reactor vessel water level.
4
l 1 l T23C - GRAND GULF 1 l l l ua,. !) -.._ -. _ _. -..._ l ,i t Ia i I 2 a q i; .. _._3
I
,,is 1 60-i:' .. I T J8 % i,l l i ? $ <e,: 1 t e; m i
o;;
-9 .l ~. -~~.. ~. . ~--- e , -l _ pL.- & ~ ? J 5 10 'E 40 25 30 35 40 45 Su TITE (HOURS) Fig. 4.22 Pressure in the drywell.
1 1 4 i i t T23C - GRAND GULF i i i 1 1 9 0_, a ~ .s Wl%5"w) ~ng.l \\. l* [ '): g.j m y,,jg-- h % S j J - ;2 3 ~}.,[ 7 a d t 4 2 no p,j ,~ + _i s 1 i $ uO] i 4 .N 4 a j .\\ =aq 4 t 4
d a
y g ga g5 g5 gW gu gW g5 gW g 1 5 z s s 4 4 y a W W u 3 5 5 v v v W W 5 3 5 W W W W o 5 le '8 25 30 35 49 45 50 i 1 TITE tHOURSI 1 i 1 ] Fig. 4.23 Temperature of gas in the drywell. I Ji l j 4 9 r l l 4
__.m.___.._._.__.- i i i e, { T23C - GRAND GULF 7 i 1JN. e i t 2. ,i j ) i. i i
.a q lg
.;_...__.....,.._...._.._p._.._.;._._j_. ..i l I l i l.. i t. 3__. 4.. ~ j -i s g q. * ' ,1 1 1 .o I d_. 4 4, t. .j p a g ae g } a is ui i if 4 J: j 89 is i I
a _.li 4.....,_..;
?* ~.% u f g t 4 1 1 A l . t i i....i....i....g....i....i....i...., 2 f t 2 5 !b 15 23 05 38 35 48 45 Se 4 J TIRE tHOURSI i Fig. 4.24 Pressure in Compartment B. i i i M
I i f f I i i, i T23C - GRAND GULF 't 1 ~, = d,.- 4 .. ~.,... h 4 J ~,, s,,.,g ... p .._.._.._.._.g.._.9., f "I t 4 19\\ I .7 4 l H 3 25 N,j i _.g..%.. ....__..a.....;..._...2..__. <l \\n- .4 k q Jt v *fo 4, ;l
^ VArt,_.w, i
~-S'%NM p s .L - [* d O .., -.. -..,,.....- 4 ,t,f ~ T = ,8 a 44 i 15 t) aIy i. .,..._..r_.....4 4 r y .,i e 7t , O +71 -s 4 _.. _. 4 a '.3] .L J J j o g.. 3,,,,i,... ....i....,,,,,,,,,,,,,,,,,,,,3 6 5 is 15 28 25 38 35 de 45 58 i TITE tHOURS) l 4 Fig. 4.25 Temperature of gas in Compartment B. i 9 r 1 l { 4 4 k
4-47 only minutes af ter the renewed pressurization, the containment is meceled to fail at this pressure at a failure location just below the jurction between the cylincer and the come. The failure cause is steam overpressurization. A 2 containment breach of 1.5 f t was modeled. In order for the T C sequence to result in core damage, it is 23 necessary that systems supplying or capable of supplying water to the RPV become unavailable. A realistic mechanism which could cause such failure has l not been identified. The accounting of containment failure location, pres-l sure, fluid flow loading, and ECCS pump suction temperature, pressure, and flPSH limitations indicates that at least one GGNS ECCS train should survive a containment failure event. However, for this analysis, the conservative l assumption that all injection sources are unavailable at 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 to the top of active fuel, the power level decreases to 6% of full power. The core begins to uncover at 1.3 hrs into the event. As each fuel noce uncovers, its power level is modeled to decrease to its decay heat level. Fual temperatures in the uncovered regions of the core begin rising i 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. Af ter 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 slumping into the lower plenum. The core debris (nen falls to the bottom of the RPV; shortly thereafter, the vessel is assumed to fail 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.
~ 4-48 Since the vessel had been depressurized previously, the cebris does not disperse from the pedestal to the drywell upon vessel failure. Further-more, the remainder of the core material gradually enters the pecestal from the vessel and also stays in the pedestal. The debris attacks the pecestal concrete as it is being quenched (see Fig. 4.26) until about three irches of concrete have been ablated. Once the core debris bed in the pedestal is cooled to below concrete ablation temperatures by the lower plenum water and the continuing CRD flow, it remains quenched until af ter 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. Consequently, no appreciable quantities of noncon-densable gases are generated. Subsequent to vessel failure steam flows steadily from the pedestal, 6 3 to the drywell, to the suppression pool at a rate of roughly 2 x 10 f t /hr. The flow is due to the fact that the CRD water is continuing to que6ch 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 discharged 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 B includes additional plots of results for this sequence. 4.5 Reference 4.1 H. A. Morewitz, " Leakage of Aerosols from Containment Buildings," Health Physics, Vol. 42, No. 2, pp.195-207,1982. .=_ --
f l T23C - GRAND GULF 4 i f, dQ J,. f (, 2 J D t.. r..._._, l cd. u. , ], @ y @,'. .i .~..i...._.....;.......i 1: A i l-I i i i 4 t .i + 2. n e s i _m. i i.. .1 ) i l n i ) 1~ g l 8 e ....,.g.. 4...... i. a, i e c. a 5 10 15 26 25 38 35 de 45 SS i TITE (HOURS) 1 J t Fig. 4.26 Average corium temperature in the pedestal. + 4 4 4 i b + r
l T23C - GRiiND GULF ^ J. 3 0 0,- I J ,i 6 i 4 l- \\ I
3. 25 6 q a :
4 t r 1 y,, J il y;.... j 4 g .y........ ? i i w j e 5 m i 1 e i i a naq t 4 t = i l ui w ,)., { f @ g . +. 4 ^ .s 4 o 4 l I 1 i t + e
s e-- a sa ai u,,....
l -I i. 0 4'0 0 l ~ 5 10 '5 'O 25 38 35 48 45 Se TITE (HOURS) ' Fig. 4.27 Concrete ablation depth in the pedestal. 4
5-1 5.0 PLANT RESPONSE WITh OPERATOR ACTIO!; In most accident situations, including those identified as the dominant GGNS plant and public-health risk scenarios, the operator has many means at his disposal to terminate it. Whatever the means, the purpose of any operator action is to place the plant into a safe stable state. Such a state exists if: (1) the core or core-debris is being cooled and (2) the contain-ment integrity is maintained. A partial list of GGNS plant systems available to the operator to cool the core or core-debris is provided on Table 5.1. A similar list of containment pressure control systems is provided on Table 5.2. The interventions selected for the operator action analyses with MAAP [5.1] are listed on Table S.3. Their selection is based on the accident sequence definition and include actions which lead to a safe stable state. For simplicity, operator actions associated with accidents involving reactor pressure vessel makeup system cemand-failures (Events UV and E) were modeled as the successful recovery of a single ECCS train. For diversity, both high and low pressure RPV makeup system recoveries were modeled. In addition, the effect of the operator response timing on the degree of core damage was also investigated. The one hour operator response time for the recovery of core cooling in the first operator action case approximates the longest time delay which will avoid significant fuel cladding damage. The assumed five hours delay in this operator response in the AE case occurs too late to prevent core melting and vessel failure, but demonstrates the potential for establishing a coolable debris bed. Since all of the WASH-1400 comparison case accidents are assumed to lack containment cooling, operator actions which recover this function are necessary in all these accidents. There are actions that the operator could take, however. The operator could vent the containment or terminate the heat input to the containment by restoring a path to the main j condenser, for example, to mitigate these events. For the scenarios in which the core is at decay heat levels (cases one through three), the effect of significant delays in the recovery of the suppression pool cooling mode of the residual heat removal (RHR) system was investigated. For the fourth case, an ATWS accident, the 10 minute standby liquid control (SLC) system actuation time approximates a time delay which still avoids containment over-pressuriza-tion. Other operator actions, also some things which aren't operator actions
__2 Table 5.1 SYSTEMS AVAILABLE FOR CORE AND CORE DEBRIS COOLING Pressure Number Rated Flow Water Source Partial List of Ra System of Pumps (gpm per pump) (gal) System Dependencies [p 5 HPCS 1 0/1435 1650 0 1147 psig 2.4 x 10 Dedicated Diesel (CST) 5 7115 0 300 psig 1.3 x 10 (SP) 5 RCIC 1 0/1190 800 2.4 x 10 Main Steam (CST) 5 1.3 x 10 (SP) 5 LPCS 1 0/520 7115 1.3 x 10 Emergency AC (SP) 5 LPCI 3 0/300 7400 1.3 x 10 Emergency AC E (SP) 5 Feedwater 2 0/1525 17800 1.1 x 10 Normal AC, (Condenser) Condensate Booster 5 Condensate 3 0/600 7170 1.1 x 10 Normal AC, Booster (Condenser) Condensate Operation 5 Condensate 3 0/224 7200 1.1 x 10 Normal AC (Condenser) 5 CRD 2 0/1950 100 '2.4 x 10 Emergency AC (CST) SLC 2 0/1400 43 4500 Emergency AC (SLC Tank) 6 SSW* 1 0/96 12000 13.9 x.10 Emergency AC (SSW Basins)
Via RHR-SSW intertie.
Table 5.2 SYSTEMS AVAILABLE FOR CONTAINMENT COOLING AND PRESSURE CONTROL Partial List of Number of Capacity System Pumps / Fans System Dependencies 6 RHR - Suppression Pool 2 Trains Each With: 75 x 10 Btu /hr per Train Emergency AC Cooling Mode 1 Pump 1 Heat Exchanger 6 RHR - Containment 2 Trains Each With: 75 x 10 Btu /hr per Train Emergency AC Spray Mode 1 Pump 1 Heat Exchanger Emergency AC SSW - Containment 1 Pump Spray Mode
Drywell Cooler 2 Trains Each With:
Normal AC 1 Cooling Coil 54.6 Ton / Coil Y' 1 Fan 12,000 SCFM/ Fan Containment 3 Trains Each With: Normal AC Cooling Cooler 1 50% Cooling Coil 63.8 Ton / Coil 1 Fan 72,500 SCFM/ Fan Steam Tunnel Cooler 1 Train With: Normal AC 2 100% Cooling Coils 14.7 Ton / Coil 2100% Fans Containment Purge 1 Train With: Normal AC 2 Pumps 3000 SCFM l Charcoal Filter Train Hydrogen Recombiner 2 - 100% Recombiners 100 SCFM-Air /Recombiner Emergency AC Hydrogen Igniter 90 Igniters Emergency AC
Via RHR-SSW intertie.
5-4 Table 5.3 OPERATOR RESPONSE SELECTION Core / Core Containment Expected Safe Case Accident Debris Cooling Pressure Control
Stable State 1
T)QUV HPCS recovered RHR-A in SPC 0 5 hr NCD/NCF 1 hr i 2 AE CRD RHR-A in SPC 0 5 hr VF/NCF 3 T 0W HPCS** RhR-A in SPC 0 30 hr NCD/NCF 23 4 T C HPCS/LPCS/LPCI** SLC 0 10 min NCD/NCF 23 RHR-A 0 10 min
SPC - Suppression pool cooling.
NCD - No core damage. VF - Vessel Failure NCF - No containment failure. iNormal CRD flow but refilling of CST required before 22 hours.
Automatic actuation, no operator response necessary, depressurization required for low pressure systems to be effective.
i I f 7 y -.r----,, y
5-5 (i.e., partial rod insertion) if taken, can extend the time available for SLC actuation. Because of the large number of aitigating mechanisms, both systems and operator actions, only a limited number were investigated here. Very slight variations in early event times compared to the cases in Section 4.0 can be observed. These are due to recent modificaticns to the plant parameter file and ao not affect the objective of Section 5 to show plant response to recovery actions. 5.1 Plant Response to the T)QUV Accident with Operator Action l A description of the GGNS plant response during the WASH-1400 This comparison case T)QUV accident scenario is provided in Section 4.1. 4 sequence is initiated by a loss of off-site power event and assumes that both high and low pressure reactor pressure vessel (RPV) makeup systems are un-available when initially demanded. When the operator realizes that none of these systems are functioning, he will attempt to restore at least one RPV makeup system to operation. Since the faults in the RPV makeup systems are assumed to be such that they are unavailable in any mode of operation, the operator must also re-establish containment heat removal. A partial tabula-tion of systems which are available to the operator and capable of cooling the core and/or cooling the containment is provided in Tables 5.1 and 5.2, respec-tively. These are specified in the plant emergency procedures. The specific operator actions modeled in this analysis are as follows: (1) the recovery'of the high pressure core spray (HPCS) one hour af ter the initiating event and j (2) the recovery of train A of the residual heat removal (RHR) system in its j suppression pool (SP) cooling mode five hours into the event. The former demonstrates the effect of re-establishing core cooling prior to significant core damage; and the latter cemonstrates the effect of re-establishing con-tainment cooling prior to containment overpressurization. The T)QUV operator action case accident chronology is provided in Table 5.4. Prior to the first of the above operator actions, the accident definition and chronology are similar to those of Section 4.1. At 1.~0 hr into the event, the operator is modeled to actuate HPCS injection to the reactor pressure vessei (RPV). Within about four minutes of this action, the core is
5-6 Table 5.4 GRAND GULF NUCLEAR STATION T)QUV - OPERATOR ACTION CASE ACCIDENT CHRONOLOGY Time Event O.0 sec Initiating Events: Loss of off-site power; Loss of main feedwater; MSIV, TSV/TBV closures 6.3 sec Reactor scram completed 45.0 sec RPV Level 2 LOCA setpoint reached t 50.0 sec Recirculation pumps trip off 32 min RPV Level 1 LOCA setpoint reached, ADS l manually activated .i 33.0 min DW purge system actuates 33.0 min Core begins to uncover 1.0 hr HPCS recovered 1.0 hr SPMU actuates 1.1 hr Core covered 1.2 hr RPV Level 8 reached and maintained 5.0 hr RHR-A placed in SP cooling mode 5.1 hr SP water temperature begins to decrease 6.0 hr Containment temperatures and pressures declining; Safe stable state achieved .e...- -.rw -r
5-7 l completely covered with water and fuel temperatures in the previously un-covered core regions drop from the 1000 F range to below 500 F. Only minor damage to the fuel cladding is predicted. Once covered with water, and maintained covered and thus, cooled, the core is in a safe stable state. The downcomer level for this recovery sequence is shown in Fig. 5.1., At 5.0 hr into the accident, the operator is modeled to recover the RHR-Train A in its SP cooling mode. The SP temperature drops monotonically afterward as observed in Fig. 5.2. Containment pressure (Fig. 5.3), which also rises slightly in the early period of the accident is calculated to be stabilizing by 6.0 hrs. Beyond that time, the containment is in a safe stable state. This sequence demonstrates the potential for recovery prior to core damage and results in a safe stable state for both reactor core and contain-ment. Since only limited core damage occurs and containment integrity is maintained, only a negligible fission product release from fuel occurs. This release is retained by the containment building. 5.2 Plant Response to the AE Accident with Operator Action A description of the GGNS plant response during the WASH-1400 comparison case AE accident scenario is provided in Section 4.3. This se-quence is initiated by a large break LOCA event and assumes that both high and low pressure reactor pressure vessel (RPV) makeup systems are unavailable when initially demanded. When the operator realizes that none of these systems are functioning, he will attempt to restore at least one RPV makeup system to operation. Since the faults in the RPV makeup systems are assumed to be such that they are unavailable in any mode of operation, the operator must also re-establish containment heat removal. A partial tabulation of systems available to the operator and capable of cooling the core and/or cooling the containment is provided in Tables 5.1 and 5.2, respectively. To demonstrate the potential for establishing a coolable debris bed, the accident was allowed to progress beyond vessel failure before any successful operator recovery action. The specific operator actions modeled in this analysis are as fol-lows: (1) filling of the condensate storage tank within 22 hours after the
5-8 T10UV.W ADS. REC -- GRAND GULF 8: e o e : H 2 5o E f W E e E TOP OF ACTIVE FUEL E ag r BOTTOM OF ACTIVE FUEL g g.o s g 5 5 0 ~r = O~ 0 2 4 6 8 TIME HR Fig. 5.1 Downcomer w&ter level.
5-9 TlOUV.W ADS. REC oo L N - GRAND GULF a Q 1 n E W8 =- = 5 L', 9 ~ J A 1 f i O. 2 4 6 TIME HR 8 Fig. 5.2 Suppression pool l \\ temperature. \\ l 1
5-10 TIOUV.W ADS. REC -- GRAND GULF o m 3$ T w E E mo g: $~ T v g-O. 2 4 6 8 TIME HR Fig. 5.3 Compartment B pressure.
5-11 initiating event and (2) recove (RHR) system in its suppressio ry of Train A of former establishes core debri n pool (SP) cooling mode at fi tainment cooling prior to cs bed cooling; and the latter est ve hours. The action case accident chronology iontainment overpressurization ablishes con-s provided in Table 5.5 The AE operator Prior to the first of . definition and chronology are similar to ththe above o the accident, most of the co actions, the ose of Section 4.2. accident the pedes ta l. Due to the release of the lowere has At 5.0 hr into ing vessel failure and to the r plenum rom the vessel into is assumed to be placed in its suppreventually q
g. 5. 4
, the core debris temperature of the suppressio ession pool cooling mode.At 5.0 hr, RH \\ n pool falls. As a result, the t water, thereby establishing a ciThe CRD flow is ca rculatory cooling water flow p thl core debris and the suppression p with pool is removed by the RHR A t ool. Decay heat rejected to the supbetween a cooling loop is established rain. , the SP water temperature decrea pression about 150 F at 8 hr into the eve on pool trates the containment pressnt as shown in Fig. 5.5. es, reaching hours the pressure has stabiliure for this sequence and showFigu zed at about 3 psig. s that beyond 5 containment failure with a coolablThis sequence d e debris bed in the pedestal and a safe stable state. Core melting occurs but the oa tainment integrity are maint i results in are highly attenuated by the supp Fission products released fsuppression a ned. I 50% ofremains intact, no significant r l ression pool and since the rom the fuel thtre for the duration ofthe total Csl is deposited ie ease f ( n the upper plenum of Approxima tely parature the fission product dthe sequence. ecay heat in the RPV is convect drywell and removed by supp tem-ression pool cooling. out into the
5-12 Table 5.5 GRAND GULF HUCLEAR STATICh AE - OPERATOR ACTION CASE ACCIDENT ChRON0 LOGY i Time Event 4 0.0 see Initiating Event: A large break in suction side of a recirculation loop 0.2 sec High DW pressure setpoint reached 4.0 sec Reactor scram completed 5.0 sec RPV Level 2 LOCA setpoint reached 6.0 sec Recirculation pumps trip off 12.0 sec RPV Level 1 LOCA setpoint reached; MSIVs close; Main feedwater pumps trip off (due to MSIV closure) 40 sec Core begins to uncover 30.0 min SPMU actuates 1.0 hr Fuel melting begins 1.3 hr vessel failure 5.0 hr RHR-A placed in SP cooling mode 10 hr Containment temperatures and pressures estimated to be declining; Safe stable state achieved 't
5-13 AE.OP INTER -- GRAND GULF s i 6 i i i b i y e H ~ E F E ~ 3 i u o g i l'N C o ? 8 2 ,g: i O. 5 10 15 20 TIME HR Fig. 5.4 Corium temperature.
5-14 i l AE.OP INTER -- GRAND GULF 9W h :: w g w LB ~ r-M g E@ b 5 E n m 9 F i L. i i e-O. 5 10 15 20 TIME HR Fig. 5.5 Suppression pool temperature. --n
5-15 l 1 AE.0P INTER -- GRAND GULF o i e i n n t.0 n v o ^ .m ay llt; r Oc i 7 ~_ E o y f-i i i O. 5 10 15 20 TIME HR Fi g. 5.6 Compartment B pressure.
5-16 5.3 Plant Response to the T QW Accident with Ocerator Action g A description of the GGh5 plant response during the WASH-1;CC QW accident scenario is provided in Section 4.3. This comparison case T23 sequence is initiated by inadvertent main steam isolation valve (MSIV) cic-sures and assumes that containment heat removal is unavailable.- When the operator realizes that containment cooling has failed, he will attempt to restore it. A partial tabulation of systems available to the operator which are capable of cooling the containment is provided in Table 5.2. Tnese are specified in the plant emergency procedures. The specific operator action modeled for this analysis is the recovery of Train A of the residual heat removal (RHR system in its suppression pool (SP) cooling mode thirty hours after the ir.itiating event. This action demonstrates the effect of re-establishing containment cooling and thus, containment pressure control prior to containment overpressurization. The T 0W perator action case chronology 23 is provided in Table 5.6. Prior to the operator action, the accident definition and chronology are similar to those of Section 4.3. At 30.0 hr into the accident, the operator is modeled to successfully place RhR-Train A in its SP cooling mode. Within a minute of this action, the SP temperature (Fig. 5.7) and all contain-ment temperatures and pressures (Fig. 5.8) are calculated to begin falling. A rapid containment air temperature decrease is due to the automatic operation of RHR-Train B in its containment spray mode. Even though it has no heat removal capacity of its own, RHR-Train B provides a large heat exchange surface area for cooling the containment atmosphere. At 30.0 hr into the event, the ccntainment is in a safe stable state. This sequence demonstrates the potential for recovery prior to containment damage and results in a safe stable state for both core and containment. No core damage occurs since RPV makeup was available throughout the event, and containment integrity was preserved.
5-17 i Table 5.6 GRAND GULF NUCLEAR STATION T 0W - WITH OPERATOR ACTION 23 ACCIDENT CHRONOLOGY s Time Event l 0 sec Initiating Event: MSIV closures; Loss of main feedwater 4.7 sec Reactor scram completed 15 sec RPV Level 2 LOCA setpoint reached 35 sec Recirculation pumps trip off 1.0 min HPCS and RCIC systems begin operating 1.1 hr HPCS and RCIC systems transfer suction from CST to SP 2.4 hr Suppression pool temperature exceeds 145*C, manual depressurization 2.5 hr RCIC pump tripped off on low RPV pressure 4.1 hr High DW pressure LOCA setpoint reached; DW purge system actuates; LPCS and LPCI actuate 4.6 hr SPMU actuates 12.3 hr High wetwell pressure setpoint reached; Containment sprays actuate 30.0 hr RHR-A placed in SP cooling mode; Containment temperatures and pressures begin dropping; Safe Stable state achieved i l
5-18 T230W. REC - GRAND GULF k.:' ~ 9 ~ _~ N w i E' 5 9 :- 1 EE : E ~ h =_ r. 5 9 5 i i o-0 10 20 30 40 50 60 TIME HR Fig. 5.7 Suppression pool temperature.
5-19 T230W. REC GPAND via p .C. u ~t a b 3 4 .E' r /g / \\ 5 ./ 3, -\\ w 3 / 2 r : / i ji 5aE / \\ n b \\, t s -1 E 3 2 3.s ( A c E 5 B 3 er 1 l 3 3 o '- 3 0-10 20 30 40 50 60 II.ME HH Fig. 5.8 Compartment B pressure. l l
5-20 6.4 Plant Response to the T C Accident with Operator Action g A cescription of the GGNS plant response during the WASH-14;C comparison case T C accident scenario is provided in Section 4.4. This 23 sequence is initiated by inadvertent MSIV closures followed by a failure to scram. When the operator realizes that the reactor has not automatically scrammed, he will attempt a manual scram. This is also assumed to fail in this analysis. he then, is expected to take two actions to reduce core power: (1) actuate the standby liquid control (SLC) system and (2) take manual control of the emergency core cooling systems (ECCS) to lower the water level in the reactor pressure vessel (RPV). The former will bring the reactor subcritical via the injection of borated water into the RPV. The latter will reduce the core pcwer from an estimated 20% to approximately 8% of full power for the postulated scenario. For simplicity, only the SLC actuation was modeled in this analysis. It is noteworthy to state that the latter will delay containment failure and thus, will increase the amount of time available for additional operator intervention. In addition to reducing core power to a decay heat level, in order to prevent containment overpressurization, the operator must also establish containment cooling. In this analysis, this was modeled by assuming the operator places Train A of the residual heat removal (RHR) system in its suppression pool (SP) cooling mode. Both SLC and RHR-A actuations were modeled to occur at 10 minutes after the initiating event. These actions establish containment pressure control prior to containment overpressurization. The T C operator action case chronology is provided in 23 Table 5.7. Prior to the operator actions, the accident definition and chronolo-gy are identical to those of Section 4.4. At 10 min into the event, the operator is modeled to actuate SLC and to place RHR-A in its SP cooling mode. The SLC actuation was conservatively modeled to bring the core subcritical twenty minutes af ter its actuation. Despite this decreasing core power, the SP water temperature rises to 200'F at 19 min into the event. At this temperature the reactor core isolation cooling (RCIC) system is modeled to fail. After its failure, the high pressure core spray (HPCS) system continues to deliver makeup flow to the
5-21 Table 5.7 GRAND GULF NUCLEAR STATION T l 23C - OPERATOR ACTION CASE ACCIDENT CfiRON0 LOGY Time Event I O sec Initiating Events: MSIV closures; Failure to scram; Loss of main feedwater 20 sec Recirculation pumps trip off 22 sec RPV Level 2 LOCA setpoint reached 51 sec HPCS begins operating 53 sec RCIC begins operating 4.6 min HPCS/RCIC systems transfer suction from CST to SP 8.0 min ADS manually open 10.0 min SLC system actuated and RHR-A placed in its SP cooling mode 19 min RCIC pump fails on high suction temperature 24 min 1.73 psig high DW pressure LOCA setpoint reached 25 min 1.89 psig high DW pressure LOCA setpoint reached 26 min RPV Level 8 reached and maintained 1 30.0 min Reactor subcritical and reactor power at decay heat level 30.4 min i 1 SP water temperature begins to decrease 55 min SPMU actuates 2.0 hr Containment temperatures and pressures decreasing; Safe stable state achieved
5-22 vessel. At 24 min, the continuing steam pressurization of the containrer.t actuates the 1.73 psig high drywell LCCA signal. The reactor scrar signal which follows is assumed to fail. At 25.0 min, the continuing steam pres-surization of the containment building causes a 1.89 psig high drywell cres-sure LOCA signal. This signal is a permissive signal for the CW purge system, and the SP makeup (SPMU) system; it is an actuation signal for The low pres-sure core spray (LPCS) and low pressure coolant injection (LPCI) systems. The SPMU system actuates the upper containment pool dump following a programmed 30 min delay. Without the drywell-wetwell pressure difference permissive signal, the DW purge system does not actuate. The RPV water inventory is maintained by the HPCS systems. The high DW pressure LOCA signal is modeled to switch the HPCS level control logic to maintain the RPV water level about the RPV Level 8 setpoint. At 26 min into the event, the decreasing core power level and thereby decreasing steam production rate allows the HPCS to fill the reactor pressure vessel (RPV) to its Level 8 setpoint. At 30 minutes into the event, the core is modeled to become subcri-tical and its power level is modeled to be at decay heat levels thereaf ter. The SP water temperatures is calculated to begin decreasing at the same time. The downcomer water level for this sequence is shown in Fig. 5.9. Suppression pool temperature and Compartment B pressure are illustrated in Figs. 5.10 and 5.11 respectively. This sequence demonstrates the potential for recovery prior to containment damage and results in a safe stable state for both core and containment. No core damage occurs since adequate RPV makeup was available throughout the event, and containment integrity was preserved.
5-23 T23C. REC -- GRAND GULF ~ o u: :- i l l m 3 a, o. 50 f i 6 J i TOP OF ACTIVE FUEL W rK :_________ l 3 g ~ BOTTOM OF ACTIVE FUEL yO i- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, - i i h t I I i o-0 5 10 15 20 25 TIME HR Fig. 5.9 Downcomer water level. .-,e,- -,.3 --,,,..,-m g, w ,y ,,.y ._e p+.--.-
5-24 T23C. REC -- GRAND GULF 8_ e- ~ g _ a bM: ~h
'}
o - pR -. E h d 8_ g: i 0 5 10 15 20 25 TIME HR Fig. 5.10 Suppression pool temperature.
o 5-25 T23C. REC -- GRAND GULF on 29 E m d 1 < N ~ '.1 " o 1 n, W i GE r* Di e F - e; E E Co i 5 a f = V d b 5 o-0- 5 10 15 20 25 TIME HR Fig. 5.11 Compartment B pressure. l I l
i 6-1 6.0 FISSION PRODUCT RELEASE, TRANSPCRT ALD :EPOSITION 6.1 Introcuction The phenomena of fission product release from the fuel matrix, its l transport within the primary system, their release from the primary system into the containment, their deposition within the containment anc 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. 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 roble 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 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 fission products can be discharged into the containment before vessel failure through relief valves or through breaks in the primary system. This is also
6-2 sequence specific.) This set of inter-related processes are treatec in :%;,P and essentially result in a release of all airborne aercsol and vapor from tr.e primary system into containment immediately following vessel failure. As a result of the dense aerosols formed when fission procucts are released from the fuel, considerable ceposition occurs within'the pricary 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 operating pressure. For celayed containment failure, this small fraction of material is depleted by in-containment mechanisms. Therefore, 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 ts 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 casium hydroxide ty 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 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.
6-3 Deposition within the containment is calculated using thermal hydraulic conditions determined 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 cue to long term, heatup, and concrete attack. At the time of containment failure, the airborne aerosol and vapor in the containment atmosphere can be released to the environment. The resuspension of deposited aerosols following containment failure has been assessed [6.6] 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 3 failure of injection (AE); transient followed by loss of containment heat i removal (T QW); and transient with failure to scram (T23 ). Thermal-hydrau-23 C lic behavior for these sequences is described in Section 4. In this section { it is shown that, for T QW and T C 23, the containment is taken to fail before 23 the core is uncovered. Hence, for these sequences the cesium and iodine are still in the fuel matrix. l 6.3.1 T)QUV 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 nete-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
6-4 Table 6.1 DISTRIBUTIO:4 0F Csl AND Cs0H Ii1 PLANT AWD 4 ENVIRONMENT FOR WASH-1400 COMPAP.ISCN CASES (FRACTI0il 0F CORE INVEilTORY) At Vessel Failure T QUV AE T 0W T C j 23 23 RPV .98 .98 .90 .68 Drywell 0.0 .02 0.0 0.0 Suppression Pool .02 0.0 .10 .32 Primary Containment 0.0 0.0 5.3 x 10-5 2.2 x 10-5 Environment 0.0 0.0 3.2 x 10-5 2.6 x 10-4 At Containment Failure T QUV AE T 0W T C j 23 23 i RPV .46 .91 1.00 1.00 l Drywell .20 .03 0.0 0.0 Suppression Pool .34 .06 0.0 0.0 Primary Containment 0.0 0.0 0.0 0.0 Environment 0.0 0.0 0.0
0. 0 1
Ultimate Distribution T QUV 'AE T 0W T C j 23 23 RPV .33 .90 .50 .26 Drywell .02 .03 .12 .05 Suppression Pool .64 .07 .38 .69 Primary Containment 7.3 x 10-4 5.1 x 10-4 2.1 x 10-4 1.1 x 10-4 Envi ronment 7.3 x 10-5 < 1 x 10-5 2.6 x 10-4 7.6 x 10-4 1 q,--,w--- a m -w y .c g w w
6-5 Table 6.2 T QUV FISS10fl PRODUCT RELEASE j Assumptions Containment Failure Location - Compartment B, 237' 9" 2 Containment Failure Size .1 ft i i Fission Product Ultimate Release Fraction l Group to Environment Noble Gases 1.0 Cs, I 7.3 x 10-5 Te, Sb 3.2 x 10-5 Sr, Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 I i I i I D e, ---s y e--e -w n-.* a
6-6 Table 6.3 AE FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Compartment B, 237' 9" 2 Containment Failure Size .1 f t Fission Product Ultimate Release Fraction Group to Environment floble Gases 1.0 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 l
6-7 Table 6.4 T QW FISSION PRODUCT RELEASE 23 Assumptions Containment Failure Location - Compartment B, 237' 9" 2 Containment Failure Size .1 ft i Fission Product Ultimate Release Fraction Group to Environment Cs, I 2.6 x 10-4 1 Te, Sb 2.2 x 10~4 Sr Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 1 l l I i i Y i 5 ,..c-
6-8 Ta bl e 6. 5 T C FISSION PRODUCT RELEASE 23 As sumptions Containment Failure Location - Compartment B, 237' 9" 2 Containment Failure Size - 1.5 ft Fission Product Ultimate Release Fraction Group to Environment Cs, I 7.6 x 10~4 Te, Sb 7.5 x 10-4 Sr, Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 l \\ l
6-9 products. A decontamination factor of 600 is associated with passage from the drywell to the pool through the suppression pool vents [6.7]. Table 6.1 also shows the Csl and Cs0h fission product inventories in the various compartments at the time of containment failure. Only the air-borne material in the upper compartment and that portion of the Eiaterial still to be revolatized in the vessel that is not scrubbed in the suppression pool 2 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 term 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. i 6.3.2 AE Secuence I 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 airborne. 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 l inferred from Table 6.1 about 1". of cesium and iodine are relocated from the vessel to the suppression pool during the period following containment fail-Of this, only one part in 600 escapes the pool to the outer containment ure. [6.7]. Release fractions to the environment are very low, as can be seen in
6-10 Table 6.3. As for the T 0L", sequence, however considerat.le ccccrete at,13:icn j occurs, although it does not occur fcr the first 20 hr of tne evert. h.C -r the ablation cepth is approximately 5 f t. 6.3.3 g30W Secuence s Table 6.1 shows the distribution of the volatile fission products (cesium and iodine) through the various regions, at vessel failure and at 150 hr when the calculation was terminated. At vessel failure, nearly all 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 has already failed prior to core uncovery there is no " puff release" of contamination to the environment on containment failure as in the T QUV and AE sequences. Thus the ultimate fission product j distribution is such that the release to the environment is very small, as indicated in Table 6.4. 6.3.4 k3C Sequence The use of MAAP leads to the release fractions shown in Tables 6.1 and 6.5. The MAAP thermal-hydraulic analysis is described in Section 4.4. Table 6.1 shows the distribution of cesium and iodine through the various regions both at vessel failure and at 50 hr, when the calculation was terminated. At vessel failure 139 kg are deposited in the upper plenum,10 kg are in the downcomer,14 kg are in the core region, and 76 kg have lef t the vessel through the SRVs 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
l 6-11 reflective insulation is very effective in transferring a considerable pcetion of the heat to the drywell as temperatures rise. Since the containment is already failed prior to core uncovery there is no " puff release" of contamination to the environment on containment failure. 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 I l i 6.1 MAAP - Modular Accident Analysis Program, User's Manual, August, 1983. l 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 Draf t IDCOR Technical Report, "FAI Aerosol Correlation," July,1984. 6.4 IDCOR Technical Report on Task 11.3, " Fission Product Transport in l Degraded Core Accidents," December,1983. 6.5 IDCOR 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 Foll; wing Primary System or Containment Failure," July, 1984. 6.7 K. Holtzclaw, Per;onal Communication,1984. 4 {
7-1 7.0 SUM.'QRY OF RESUL.TS As outlined in Section 2 of this report, the 10CCR Subtask 23.1 Integrated Containment Analysis of the Grand Gulf Muclear Station (GG35) consisted of three parts: (1) WASH-1400 comparison case accident analyses, (2) operator action case accident analyses, and (3) uncertainty and sensitiv ty analyses. The accident sequences selected for analysis represent a majori-ty of previously-assessed risk and demonstrate a variety of initiating events a variety of system failures combinations, and a diversity of accident pheno-menology. The primary system and containment thermal-hydraulic response j analyses and fission product transport were performed 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, respectively. This section of the report summarizes the major results of each of these analyses. 7.1 WASH-1400 Comparison Case Analyses The WASH-1400 comparison case analyses establish a reference sys 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 an The WASH-1400 comparison case fission product transport results are ses. summarized on Table 7.1. The BWR-4 release magnitudes are provided for comparison purposes. A discussion of these results follows. Accidents involving demand-type failures of all automatically-actuated high and low pressure reactor pressure vessel (RPV) makeup system namely those accident sequences containing events UV or E, result in core damage unless an appropriate operator response is taken. 7 For accidents which involve relatively small RPV coolant inventory loss rates and decay power levels, such as T QUV, the core is predicted to begin to uncover within abou j 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 hcurs af ter the initiating event. Vessel failure will follow within i
Table 7.1
SUMMARY
OF FRACTIONAL RADIONUCLIDE RELEASES TO THE ENVIRONMENT Fission Product Group i Accident Xe and Kr Cs and I Te . Sr and Ba Ru and Mo a i -5 -5 T QUV 1.0 7.3 E-5 3.2 E-5 < 1 x 10 < 1 x 10 j -5 -5 -5 AE 1.0 < 1 x 10 1.1 E-5 < 1 x 10 < 1 x 10 -5 -5 T QW l.0 2.6 E-4 2.2 E-4 < 1 x 10 < 1 x 10 23 T C 1.0 7.6 E-4 7.5 E-4 < 1 x 10'b < 1 x 10-5 23 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.
Cesium release fraction is 5.0 E-3. 't l' w m-
7-3 another half-hour. AE, these events occur soonerFor accidents with large RPV i fuel melting was predicted toFor the large-break LO
accident, ating event and was closely followed b occur in about one hour ofnalyzed, the A y vessel failure.
the initi-cooling, such as TAccidents involving successful RPV 23, will result in containment failu 230W and T appropriate operator action is take C all ECCS injection into the RPV will f n. Previous studies have postulat d unless t With this assumption, and witho t ail as a result of containment fa e tha t will inevitably follow. u The results of this study indicatappro ure. tion that all ECCS equipment f il e ting basis and postulated dominant GGNS accidentthus is ex a Thus, many of n stic lead to core damage following contais sequences do the previously-nment failure.not on a best-estimate basi ~ in the core was calculated to bThe mass of hydroge e significantly lower than that prv the NRC in the interim rule ng The MAAP predictions demonstrat on hydrogen control for Mark III co t escribed by oxidation prior to fuel melting fo e that less than about n ainments. { oxidation rule specifies a 75% claddir severe GGNS accidents.10% to progress unmitigated to vesng reaction. The NRC cladding oxidized is predicted at only 35%sel failure, the maximum necessary to generate cladding r Several very complicated action cladding a low vessel makeup flow or an orcheactions of higher magnitudes. s would be gency core cooling would be nece estrated termination and r j calculated for the GGNS severe accid ssary. i The rate of hydrogen productio emer-i than those used in previous st di \\ ent analyses is also substantially l n u es. ower debris is calculated to fallFor accidents which proceed b onto the pedestal floor.eyond vessel failure calculated , tha pedestal floor and walls.to exit the pedestal volume core No core debris is Without core-debrisThus, concrete attack is . erosion of the pedestal floor and walls i e to cooling, substantial s calculated to occur. \\ -...,, - -,, - - -.. -. - - - -. _.. _. _ ~ _,. 7
7-4 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-celayed containment failure events for tne GGNS accidents analyzed, the MAAP code predicts no steam explosions large enough to fail either the rea'Ctor 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 j this study. For the GGNS Mark III design, the suppression pool was observed to l the exert a dominant influence on the accident progression. There are a number of reasons that the suppression pool displays this behavior.
First, overpressurization of the containment by steam can occur only if the sup-pression pool is teated to high temperatures or if the suppression pool is by-passed. The fonner requires a substantial energy deposition and inadequate suppression pool hc;t 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. Lastly, 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 ex-hausted through a containment breach without first being highly decontaminated i by the suppression pool unless a drywell to wetwell bypass is postulated. With the low pressure difference between the drywell and wetwell, any openings due to seal leakage or degradation.would be readily plugged by aerosols generated during core-concrete attack.
8-1
8.0 CONCLUSION
S Based on the results of the severe accident analyses performed in this study, a number of conclusions can be drawn regarding the progression anc consequences of such postulated severe accidents for plant designs similar to that of the Grand Gulf Nuclear Station. 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. The MAAP code is a viable means of analyzing both the thermal-e hydraulic and the radiological response of the Grand Gulf Nuclear Station primary systen and containment to severe accident scenarios. For selected accidents postulated to lead to core damage, fuel e melting, and/or containment failure, there are sufficient time and means available to the operating staff to place the plant into a safe stable state. Containment failure should no longer be considered a cause for e the failure of all ECCS flow to the reactor vessel.
Thus, containment failure should no longer be considered a cause for core melt.
The mass and rate of hydrogen calculated to be produced in the e vessel prior to fuel melting is substantially less than that predicted by previous studies, If successful fuel cooling is delayed beyond the point of o significant core damage and/or vessel failure, the core debris coolability is possible.
8-2 e The suppression pool exerts an important thernal-hydrau'.ic ar.d l radiological influerce on the containment resporse to a severe accident. e The GGNS Mark III ccntainment failure modes are overpressuriza-tion via steam, noncendensable gas generation, and/or hydrogen combustion. Containmbnt failure times are long compared to previous studies. No prompt containment failures due to steam explosions or steam spiking were calculated. e Through continuous burning of the containment combustible gas, the GGNS containment hydrogen igniters can significantly delay containment failure during a severe accident. e Decontamination of the fission product releases by the suppres-sion pool and sedimentation of aerosols in the drywell were found to be the two most important fission product removal mechanisms. e The public health consequences of the severe accidents are substantially less than those of previous assessments. e
A-1 APPENDIX A.1 Grand Gulf Parameter File s'* MARK III BWR PLANT PARAMETER VALUES-- TYPICAL OF GRAND GULF I
SI UNITS (M-KG-SEC-DEGK)

7-22-83

PRIMARY SYSTEM PS 01 13.521DO AFLCOR FLOW AREA 0F REACTOR CORE PS
ALSH IS CALCULATED BY TAKING THE VOLUME OF WATER IN THE LOWER

DOWNCOMER+ JET PUMP AND DIVIDING BY (ZTOAF-ZBJET) 02 8.57D0 ALSH FLOW AREA IN LOWER SHROUD PS 03 5 695D0 AFLBYP CORE BYPASS FLOW AREA PS

AUSH IS CALCULATED BY TAKING THE VOLUME OF WATER IN THE UPPER

DOWNCUMER ABOVE TOAF AND DIVIDINO BY THE WATER HEIGHT ABOVE TOAF 04 2.64401 AUSH FLOW AREA IN UPPER SHROUD PS 05 1 116D5 HCRD SPECIFIC ENTHALPY OF FLOW IN CRD TUBES PS 06 9.248DS HFW SPECIFIC ENTHALPY OF FEEDWATER PS 07 1.65697D5 MU2PS TOTAL M45S OF UO2 IN CORE PS 08 8.0D2 NASO NUMBER OF FUEL ASSEMBLIES IN REACTOR CORE PS 09 6.2D1 NPINS NUMBER OF FUEL RODS IN A FUEL ASSEMBLY PS 10 1.93D2 NCRD NUMBER OF CRD TUBES PS 11 4.5D0 NOFPS SENSIBLE ENERGY STORED IN FUEL (FULL POWER SECONDS)PS 12 4.0D0 TDMSIV DELAY TIME FOR MSIV CLOSURE PS 13 3.5DO TDSCRM DELAY TIME FOR FULL SCRAM PS 14 5.976D7 TIRRAD TOTAL EFFECTIVE IRRADIATION TIME FOR CORE PS

ALL PUMP CURVES ASSOCIATE THE FIRST FLOW RATE WITH THE FIRST PRESSURE

FOR THAT SPECIFIC PUMP 15 7.D-3 WVCRDI CRD FLOW RATE PUMP CURVE FOR CRD FLOW PS 16 1.12D-2 WVCRDI CRD FLOW RATE PPS VS WVCRDI PS 17 1.12D-2 WVCRDI CRD FLOW RATE M3/S PS 18 1.12D-2 WVCRDI CRD FLOW RATE PS 19 1.12D-2 WVCRDI CRD FLOW RATE PS 20 1.12D-2 WVCRDI CRD FLOW RATE PS 21 1.12D-2 WVCRDI CRD FLOW RATE PS 22 1.120-2 WVCRDI CRD FLOW RATE PS 23 6.894D6 PCRD PPS FOR CRD PUMP PS 24 1.0134D5 PCRD PPS FOR CRD PUMP PS 25 1.0134D5 PCRD PPS FOR CRD PUMP PS 26 1.0134D5 PCRD PPS FOR CRD PUMP PS 27 1.0134D5 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 1.0134D5 PCRD PPS FOR CRD PUMP PS 31 3.333D3 WFWMAX MAXIMUM FEEDWATER FLOW RATE (RUN OUT FLOW)
PS 32 6.85D2 WBPMAX MAXIMUM TURBINE BYPASS FLOW RATE PS 33 1.63D-1 NXCORE EXIT CORE QUALITY AT TIME ZERO PS 34 5.264D0 XDCORE REACTOR CORE DIAMETER TO INNER SHROUD WALL PS 35 2.206D1 XHRV INTERIOR HEIGHT OF REACTOR VESSEL PS 36 3.188DO XRRY INTERIOR RADIUS OF REACTOR VESSEL PS 37 41.01D0 ZBJET ELEVATION AT BOTTOM OF JET PUMPS PS 38 38.77D0 ZBRDT ELEVATION AT BOT 10M OF CRD TUBES PS 39 50.44D0 IBSEP ELEVATION AT BOTTOM OF STEAM SEPARATORS PS 40 37.4100 ZBV ELEVATION AT BOTTOM OF REACTOR UESSEL PS 41 42.73D0 ZCPL ELEVATION AT CORE PLATE PS 42 45.48DO ZTJET ELEVATION AT TOP OF JET PUMPS PS l 43 1.33D0 AJET TOTAL AREA 0F JET PUKPS PS 44 46.74DO ZTOAF ELEVATION AT TOP OF ACTIVE FUEL PS 45 52.65DO ITSEP ELEVATION AT TOP OF STEAM SEPARATORS PS 46 51.91DO ZWNORM ELEVATION AT NORMAL SHROUD WATER LEVEL PS 47 41.8200 ZLOCA ELEVATION AT BREAN PS 48 .2919D0 ALOCA AREA 0F BREAK F3 49 52.32500 ZWLB ELEVATION AT LEVEL 8 TRIP PS 50 0.000 NOT USED 51 51.2600 ZSCRAM LOW WATER LEVEL SCRAM PS 52 7.4435D6 PSCRAM HIGH PRESSURE SCRAM SETPOINT PS l
A-2 53 .20D0 FOATWS ATWS CONSTANT POWER ASSUMPTION PS 54 1.2D3 TDSLC TIME FOR SCRAM WITH SLC FS 55 0.D0 TIRR(1) TIME VS. FRACTICN OF TOTAL FLCW FOR RECIRC PUMP FE 56 2.D0 TIRR(2) PS 57 4.D0 TIRR(3) FS 58 6.D0 TIRR(4) PS 59 8.D0 TIRR(5) PS 60 10.D0 TIRR(6) PS 61 15.01200 TIRR(7) PS 62 40.D0 TIRR(8) PS 63 1.D0 FWRR(1) PS 64 .67D0 FWRR(2) PS 65 45D0 FWRR(3) PS 66 .30D0 FWRR(4) PS 67 .20D0 FWRR(5) PS 68 .135D0 FWRR(6) PS 69 .050D0 FWRR(7) PS 70 0.D0 FWRR(B) PS 71 1.18393D5 HSLC INLET ENTHALPY OF SLC PS 72 0.D0 PSLC(1) PRESSURE POINTS FOR SLC FLOW CURVE PS 73 7.93D6 PSLC(2) PS 74 7.93D6 PSLC(3) PS 75 7.93D6 PSLC(4) PS 76 7.93D6 PSLC(5) PS 77 7.93D6 PSLC(6) PS 78 7.93D6 PSLC(7) PS 79 7.93D4 PSLC(8) PS 80 2.713D-3 WVSLC(1) SLC FLOW RATE AT PSLC(1) -- M3/S PS i 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 1 86 2.713D-3 WVSLC(7) PS 87 2.713D-3 WVSLC(8) PS 88 .2D0 TDRPT DELAY TIME FOR RECIRC PUMP TRIP PS 89 47.16D0 ZLMSIV LOW WATER LEVEL FOR MSIV CLOSURE PS 90 49.92D0 ZLRPT LOW WATER LEVEL FOR RECIRC PUMP TRIP PS 91 7.850D6 PHRPT HIGH VESSEL PRESSURE FOR RECIRC PUMP TRIP PS 92 1 1327D5 PDWSCM HIGH DRYWELL PRESSURE FOR SCRAM PS 93 .032D0 FENRCH NORMAL FUEL ENRICHMENT PS 94 2.04 EXPO AVERAGE BURNUP IN MWD / TONNE PS 95 .600 FCR PRODUCTION OF U239 TO ABSORBTION IN FUEL PS 96 1.3D0 FFAF RATIO OF FISSILE ABSORBTION TO TUTAL FISSIUN P3 97 5.D-1 F0FR1 FISSION POWER FRACTION OF U235 AND PU 41 PS t 98 4.2D-1 F0FR2 FISSION POWER FRACTION OF PU239 PS 99 8.D-2 F0FR3 FISSION POWER FRACTION OF U238 FS 100.305100 XPCRDT PITCH OF CRD TUBES PS 101.2755D0 XDCRDT OUTER DIAMETER OF CRD TUBE PS 102 58.D0 NINST NUMBER OF INSTRUMENT TUBES PS 103.0044D0 XTHCRD THICKNESS OF CRD TUBE WALL PS 104.0508D0 XDINST OUTER DIAMETER OF INSTRUMENT TUBE PS 105.0818D0 XDRIVE LOWER CRD DRIVE OUTER DIAMETER PS 106 1 0040-3 VWCRD SPECIFIC VOLUME OF CRD WATER PS 107 1.0040-3 VWCST SPECIFIC VOLUME OF SLC WATER PS 108 5.8167D4 MEOPS MASS OF UPPER PLENUM HEAT SINK PS 109 1 016D3 AEDPS SURFACE AREA 0F UPPER PLENUM HEAT SINK PS 110.241D0 XTRV THICKNESS OF LOWER VESSEL HEAD PS 111 0.00 TIFWCD TIME SINCE MSIV CLOSURE SIGNAL VS. FEEDWATER PS 112 0.D0 COASTDOWN MASS FLOW RATE PS 113 0.D0 PS 114 0.D0 PS 115 0.D0 PS
A-3 116 0.D0 PS 117 0 D0 PS 118 0.D0 PS 119 0.D0 WFWCD PS 120 0.D0 PS 121 0.D0 PS 122 0.00 PS 123 0.D0 PS 124 0.D0 PS 125 0.D0 PS 126 0.D0 PS 127 5.86D6 PLMSIV LOW RPV PRESSURE FOR MSIV CLOSURE PS 128 53.9D0 ZMSL ELAVATION AT CENTER LINE OF MAIN STEAM LINE PS HE
CIRC
THIS SECTION IS FOR INPUTS TO THE DETAILED PRIMARY SYSTEM T/H

MODULE SUIROUTINE CIRC AND ITS ANCILLARY SUBROUTINES

TO DEVELOP INPUTS FOR THIS SECTION, CONSULT THE APPROPRIATE FIGURE

FOR YOUR PLANT TO DELINEATE NODAL BOUNDARIES

INSERT NUMBERS IN THE SPACES SHOWNi THE CODE COMPUTES THE MISSING

NUMBERS FROM THESE AND OTHER INPUTS

t
CIRC ALLOWS EACH NODE TO HAVE 1 OR TWO STRUCTURESI EACH NODE HAS A

' STEEL' MASS AND MAY IN SOME CASES ALSO HAVE A ' HEAT SINX' MASSi

THE HEAT SINK IS DISTINGUISHED FROM THE STEEL MAINLY BY TWO DIFFERENCES:
1. MAY BE AT A DIFFERENT TEMPERATUREI THIS MAY BE DUE IN PART TO THE HEAT SINK HAVING LOSSES TO CONTNT WHEN THE STEEL DOESN'T (EG S/G SHELLS VS TUBES)

2. HEAT SINKS ARE ASSUMED NOT TO HAVE FISSION PRODUCTS PLATED ON THEM
t
AT PRESENT, THE ENERGY EXCHANGES IN A NODE MAY INCLUDE ONE OR MORE OF

THE FOLLOWING, DEPENDING ON THE INPUT PARAMETERS SUPPLIED:
1. HEAT SINK AND STEEL EXCHANGE ENERGY RADIATIVELY (GAS ASSUMED TRANSPARENT)

2. STEEL EXCHANGES ENERGY WITH PRIMARY SYSTEM GAS VIA INTER-NODAL OR INTRA-NODAL NATURAL CIRCULATION

3. HEAT SINK MAY EXCHANGE ENERGY WITH PRIMARY SYSTEM GAS VIA INTER-OR INTRA-NODAL NATURAL CIPCULATION
ITEMS 1-3 ARE HT AREAS COUPLING THE 2 NODAL HEAT SINK MASSES (STEEL AND

HEAT SINK) 01 0.D0 ACSHS(1) CORE + LOWER PLENUM CARBON STEEL-HEAT SINK HEAT TRANSFER AREA l
02 140.00 ACSHS(2) UPPER PLENUM 03 0.D0 ACSHS(3) DOWNCOMER
ITEMS 6-8 ARE ' STEEL' (INTERNAL) MASSES 06 50.D3 MCS(1)
CORE + LOWER PLENUM CARBON STEEL MASS 07 100.D3 MCS(2) UPPER PLENUM 08 350.D3 MCS(3) DOWNCOMER
ITEMS 11-13 ARE THE ' HEAT SINK' MASSES 11 0.D0 MHS(1)
CURE + LOWER PLENUM HEAT SINK MASS 12 100.D3 MHS(2) UPPER PLENUM 13 0.D0 MHS(3) DOWNCOMER
ITEMS 16-18 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

STEEL MASSES AND CONTAINMENT 16 0.D0 ACSX(1) CORE + LOWER PLENUM CARBON STEEL TO DRYWELL HEAT TRANSFER ARE 17 0.D0 ACSX(2) UPPER PLENUM 18 240.D0 ACSX(3) DOWNCOMER

ITEMS 21-23 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

HEAT SINK MASSES AND CONTAINMENT 21 0.D0 AHSX(1) CORE + LOWER PLENUM HEAT SINK TO DRYWELL j
A-4 HEAT TRANSFER AREA 22 140.D0 AHSX(2) UPPER PLENUM 23 0.D0 AHSX(3) DOWNCOMER
ITEMS 26-28 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

STEEL MASSES AND PRIMARY SYSTEM GAS 26 100.D0 AGCS(1) CORE + LOWER PLENUM GAS TO CARBON STEEL HEAT TRANSFER AREA 27 5.D3 AGCS(2) UPPER PLENUM 28 240.D0 AGCS(3) DOWNCOMER

ITEMS 31-33 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE PETWEEN

HEAT SINK MASSES AND PRIMARY SYSTEM GAS 31 0.00 AGHS(1) CORE + LOWER PLENUM GAS TO HEAT SINN HEAT TRANSFER AREA 32 140.D0 AGHS(2) UPPER PLENUM 33 0.D0 AGHS(3) DOWNCOMER

APPROX. NODE HEIGHT FOR EACH COMPARTMENT USED FOR NAT CIRC i
36 8.0D0 XL(1) CORE + LOWER PLENUM LENGTH 37 5.D0 XL(2) UPPER PLENUM LENGTH 38 10.D0 XL(3) DOWNCOMER LENGTH
GAS FLOW AREA 41 11.D0 AG(1)
CORE + LOWER PLENUM FLOW AREA 42 11.D0 AG(2) UPPER PLENUM FLOW AREA 43 10.D0 AG(3) DOWNCOMER FLOW AREA
HYDRAULIC DIAMETER USED TO COMPUTE HT COEFF 46 5.D0 DH(1) HYDRAULIC DIAMETER FOR CORE REGION 47
.15DO DH(2) HYDRAULIC DIAMETER FOR UPFER PLERUM 48 .4D0 DH(3) HYDRAULIC DIAMETER FOR DOWNCOMER 51 0.D0 QC0 RPU CONVECTION LOSSES AT TIME ZERO 52 8.D0 FINPLT NUMBER OF LAYERS IN REFLECTIVE INSULATION l 53 32.5 ASEDPS(1) AEROSOL SEDIMENTATION AREA FUR CORE + LOWER PLENUM 54 20.0 ASEDPS(2) AEROSOL SEDIMENTATION AREA FOR UFFER PLENUM 55 10.0 ASEDPS(3) AEROSOL SEDIMENTATION AREA FOR DOWNCOMER HE
HEATUP HE 01 3.8100 XIFUEL LENGTH OF ACTIVE FUEL HE 02 5.21D-3 XRFUEL RADIUS OF FUEL PELLET HE 03 8.13D-4 XTCLAD THICKNESS OF CLADDING HE 04 5.033D4 MZRCAN TOTAL MASS OF ZR IN ASSEMBLY CAN HE 05 1.7D4 MBCR TOTAL MASS OF CONTROL BLADES IN REACTOR CORE HE 06 3.048D-3 XZRCAN CAN WALL THICKNESS HE
NODE 1,1 IS BOTTOM-CENTER, 1,10 IS TOP-CENTER, 2,1 IS SECOND RADIAL

RING OUT FROM CENTER AT THE BOTTOM UF THE CORE,ETC 07 6.680D-1 FPEAK(1,1) PEAKING FACTOR FOR NODE (1,1)
HE 08 7.910D-1 FPEAK(2,1) PEAKING FACTOR FOR NODE (2,1) HE 09 4.700D-1 FPEAK(3,1) PEAKING FACTOR FOR NODE (3,1) HE 15 1.145D0 FPEAK(1,2) PEAKING FACTOR FOR NODE (1,2) HE 16 1.466D0 FPEAK(2,2) PEAKING FACTOR FOR NODE (2,2) HE 17 9.290D-1 FPEAK(3,2) PEAKING FACTOR FOR NODE (3,2) HE 23 1.019D0 FPEAK(1,3) PEAKING FACTOR FOR NODE (1,3) HE 24 1.343D0 FPEAK(2,3) PEAKING FACTOR FOR NODE (2,3) HE 25 8.960D-1 FPEAK(3,3) PEAKING FACTOR FOR NODE (3,3) HE 31 1.029D0 FPEAK(1,4) PEAKING FACTOR FOR NODE (1,4) HE 32 1.281D0 FPEAK(2,4) PEAKING FACTOR FOR N0DE (2,4) HE 33 8.67D-1 FPEAK(3,4) PEAKING FACTOR FOR N0JE (3,4) HE 39 1.223D0 FPEAK(1,5) PEAKING FACTOR FOR NODE (1,5) HE 40 1.414D0 FPEAK(2,5) PEAKING FACTOR FOR NODE (2,5) HE 41 9.430D-1 FPEAK(3,5) PEAKING FACTOR FOR NODE (3,5) HE 47 1.235D0 FPEAK(1,6) PEAKING FACTOR FOR NODE (1,6) HE 48 1.37300 FPEAK(2,6) PEAKING FACTOR FOR NODE (2,6) HE 49 9.03D-1 FPEAK(3,6) PEAKING FACTOR FOR NODE (3,6) HE 55 1.19800 FPEAK(1,7) PEAKING FACTOR FOR NODE (1,7) HE M 1.269D0 FPEAK(2,7) PEAKING FACTOR FOR NODE (2,7) HE 57 8.09D-1 FPEAK(3,7) PEAKING FACTDR FOR NODE (3,7) HE
A-5 63 1.235D0 FPEAK(1,8) PEAKING FACTOR FOR N0t'E (1,8) HE 64 1.243D0 FPEAK(2,8) PEAKING FACTOR FOR NODE (2,8) HE l 65 7.11D-1 FPEAK(3,8) PEAKING FACTOR FOR NODE (3,8) HE 71 1.33100 FPEAK(1,9) PEAKING FACTOR FOR NODE (1,9) HE 72 1.107D0 FPEAK(2,9) PEAKING FACTOR FOR NODE (2,9) HE HE 73 5.53D-1 FPEAK(3,9) PEAKING FACTOR FOR NODE (3,9) 79 7.40D-1 FPEAK(1,10) PEAKING FACTOR FOR NODE (1,10) HE 80 5.64D-1 FPEAK(2,10) PEAKING FACTOR FOR N00E (2,10) HE 81 2.69D-1 FPEAK(3,10) PEAKING FACTOR FOR NOLE (3,10) HE 87 0.3D0 XCHIM UNHEATED FUEL LENGTH AT TOP OF CORE HE 88 1.D-7 XIZROX INITIAL CLADDING OXIDE THICKNESS HE ES ES
ENGINEERED SAFEGUARDS ES 01 1.D0 NLPCIl NUMBER OF LPCI PUMPS IN LOOP 1 ES 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 1.00 NLPCSP NUMBER OF LPCS PUMPS ES 05 0.0D0 NOT USED 06 1.4D1 VMNCST MIN. WATER UOLUME IN CONDENSATE STORAGE TANK ES FOR HPCI AND RCIC SUCTION SWITCH OVER ES 07 1.008D-3 VWCST SPECIFIC VOLUME OF CS1 WATER ES
ALL PUMP CURVES ARE ARRANGED SO THAT THE FIRST FLOW ENTRY CORRESPONDS

TO THE FIRST PRESSURE ENTRY 08 1.D10 DRYWELL PRESS WHICH WILL CLOSE ADS VALVES 09 0.D0 DRYWELL PRESSURE WHICH WILL ALLOW ADS TO RE-OFEN IF CLOSED 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 1.641D6 PLPCI(4) ES 28 1.462D6 PLPCI(5) ES 29 1.165106 PLPCI(6) ES 30 .84106 PLPCI(7) ES 31 .4964D6 PLPCI(8) ES 32 0.0D0 WVLPCI(1) ES 33 .1262D0 WVLPCI(2) ES 34 .1893D0 WVLPCI(3) ES 35 .3155D0 WVLPCI(4) ES 36 .3786D0 WVLPCI(5) ES 37 .4417D0 WVLPCI(6) ES 38 .5048D0 WVLPCI(7) ES 39 .5641D0 WVLPCI(8) ES 40 3.584D6 PLPCS(1) LPCS PUMP CURVE ES 41 3.378D4 PLPCS(2) ES i 42 3.06D6 PLPCS(3) ES 43 2.889D6 PLPCS(4) ES 44 2.67D6 PLPCS(5) ES 45 2.392D6 PLPCS(6) ES j 46 2.068D6 PLPCS(7) ES 1 47 1.572D6 PLPCS(8) ES 48 0.D0 WVLPCS(1) ES 49 .1262D0 WVLPCS(2) ES 50 .2524D0 WVLPCS(3) ES 51 .3155D0 WVLPCS(4) ES 52 .3786D0 WVLPCS(5) ES 53 .4417D0 WVLPCS(6) ES 54 .504800 WVLPCS(7) ES 55 .5742D0 WVLPCS(8) ES 56 9.892D6 PHPCS(1) HPCS PUMP CURVE ES 57 8.886D6 PHPCS(2) ES 58 7.521D6 PHPCS(3) ES 59 6.749De PHPCS(4) ES
A-6 60 5.667D6 PHPCS(5) ES 61 4.226D6 PHPCS(6) ES 62 2.296D6 PHPCS(7) ES 63 0.000 PHPCS(8) E"$ 64 0.00 WVHPCS(1) E 65 .1262D0 WVHPCS(2) E5 66 .2524DO WVHPCS(3) ES 67 .3155D0 WVHPCS(4) ES 68 .3786D0 WVHPCS(5) ES 49 .4417D0 WVHPCS(6) ES 70 .5048D0 WVHPCS(7) ES 71 .5742D0 WVHPCS(8) ES 72 10.341D6 PRCIC(1) RCIC PUMP CURVE ES i 73 10.340D6 PRCIC(2) ES 74 6.894D6 PRCIC(3) ES 75 3.447D6 PRCIC(4) ES 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 .0505D0 WVRCIC(2) ES 82 .050500 WVRCIC(3) ES 83 .0505D0 WVRCIC(4) ES 84 .0505D0 WVRCIC(5) ES 95 .0505D0 WVRCIC(6) ES 86 .0505D0 WVRCIC(7) ES 87 0.D0 WVRCIC(8) ES 88 50.D0 ILHPCI LOW WATER INITIATION FOR HPCI ES 89 1.D10 PSHPCI HIGH DRYWELL PRESSURE SET POINT FOR HPCI ES 90 1.D10 TDHPCI TIME DELAY FOR HPCI ES 91 1.D10 PHHPCI MINIMUM PRESSURE FOR HPCI TURBINE ES 92 49.9200 ILHPCS LOW WATER INITIATION FOR HPCS ES 93 1 14405 PSHPCS HIGH DRYWELL PRESSURE SET POINT FUR HPCS ES 94 27.D0 TDHPCS TIME DELAY FOR HPCS ES 95 47.16D0 ZLLPCI LOW WATER INITIATION FOR LPCI ES 96 1 144D5 PSLPCI HIGH DRYWELL PRESSURE SET POINT FOR LPCI ES 97 40.D0 TDLPCI TIME DELAY FOR LPCI ES 98 -1.D10 PLLPCI RPV-WETWELL PRESS DIFFERENCE TO KEEP ADS OPEN 99 47.16D0 ZLLPCS LOW WATER INITIATION FOR LPCS ES 100 1 144D5 PSLPCS HIGH DRYWELL PRESSURE SE1 POINT FOR LPCS ES 101 37.D0 TDLPCS TIME DELAY FOR LPCS ES 102 -1.D10 PLLPCS RPV-WETWELL PRESS DIFFERENCE TO CLOSE ADS VALVES 103 49.92D0 ZLRCIC LOW WATER INITIATION FOR RCIC ES 104 1 0D10 PSRCIC HIGH DRYWELL PRESSURE SET POINT FOR RCIC ES 105 30.D0 TDRCIC TIME DELAY FOR RCIC ES 106 5.15D5 PHRCIC MINIMUM VESSEL PRESSURE FOR RCIC TURBINE ES 107 1.70D5 HCST ENTHALPY OF CST ES 108 498.00 WSWHX SERVICE WATER FLOW RATE (KG/S) THRU EACH RHR HTX ES 109.0119D0 ASRV1 FLOW AREA 0F RELIEF VALVE TYPE 41 ES 110.0119D0 ASRV2 FLOW AREA DF RELIEF VALVE TYPE 42 ES 111.0119D0 ASRV3 FLOW AREA 0F RELIEF UALVE TYPE 43 ES 112.0119D0 ASRV4 FLOW AREA 0F RELIEF VALVE TYPE 44 ES
IF THE AREA 0F GROUP 95 IS INPUT AS A NEGATIVE NUMBER THEN THE VALVE

WILL DISCHARGE DIRECTLY INTO THE DRYWELL, IF POSITIVE IT WILL

DISCHARGE INTO THE SUPPRESSION POOL 113.0D0 ASRV5 FLOW AREA DF RELIEF VALVE TYPE 05 ES 114 1.0D0 NSRV1 NUMBER OF TYPE 41 RELIEF VALVES ES 115 1.000 NSRV2 NUMBER OF TYPE 02 RELIEF VALVES ES 116 9.0D0 NSRV3 NUMBER OF TYPE 93 RELIEF VALVES ES 117 9.000 NSRV4 NUMBER OF TYPE 04 RELIEF VALVES ES 118 0.D0 NSRV5 NUMBER OF TYPE 85 RELIEF VALVES ES 119 0.D0 NADS1 NUMBER DF ADS VALVES IN GROUP 1 ES
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A-7 120 1.D0 121 3.D0 NADS2 NUMBER OF ADS VALVES IN G 122 4.D0 NADS3 NADS4 123 7 122006 PSRV1NUMBER OF ADS VALVES IN GROUP 4 124 7 398D6 PSRV2 ES PRESSURE SETPOINT FOR 92 125 7.674D6 PSRV3 ES 126 7.74306 PSRV4 PRESSURE SETPOINT FOR 43 RELIEF VA E ES 127 1.D10 PSRV5 PRESSURE SETPOINT FOR 84 RELIEF VA E ES 128 47 16D0 ILADS VE ES LOW WATER LEVEL FOR INITI 129 E ES 114.37D3 PSADS 130 115 00 TDADS ES HIGH DRYWELL PRESSURE SET POINT FO
HPCI AND RCIC WILL TRIP OFF ON USE** LPCI, ES TIME DELAY FOR ADS ACTUATION ADS ES 131 373.D0 S
ES TCHPCI INLET TEMP LIMIT FOR HPCI 132 31 400 ICLLPI PUMP CENTER LINE E ES 133 29 3DO 134 30.500 ZCLLPS PUMP CENTER LINE ELAVATION FO 135 366.300 HPCS ES TCRCIC INLET TEMP LIMIT FOR RCIC 136 305.00 LPCI ES TWSW 137 13.D0 LPCS ES HPCS LOAD DELAY TIME FOR DI TDDG1 138 13.D0 ES 139 13.D0 TDDG2 LPCI LOAD DELAY TIME FOR DIESEL ERS.TCOLD) ES 140 2 3D-4 TDDG3 LPCS LOAD DELAY TIME FOR DIESEL ES XDDROP SRRAY DROPLET DIAMETER FOR C 141 19.6D0 ES XHSPWW SPRAY FALL HEIGHT IN WETWEL 142 10.D0 ES
THE HPSW SYSTEM CAN BE USED TO MODXHSP ONTAINMENT SPRAYSES

SERVICE WATER OR FIRE WATER, THE S L
ES 143 1.83705 EL ANY INJECTION MODE SUCH AS ES HWHPSW ENTHALPY OF HIGH PRES SERVIC 144 YSTEM IS TOTALLY DEFINED BELOW 1 009D-3 ES VWHPSW SPEC VOL OF HIGH PRES SERVICE 145 6.525D5 PHPSWC1) PPS VS. VOLUMETRIC FLOW F 146 6.524D5 E WATER (MARK I CI) PHPSW(2) 147 6 523D5 WATER (MARK I CI) ES PHPSW(3) (MARK I CORE INJECTION)OR H 148 6 522D5 ES PHPSW( 149 6.521D5 PHPSW(4) ES 150 6 520D5 PHPSW(5) ES 151 6 519D5 PHPSW(6) ES 152 0.D0 7) ES 153 0 00 PHPSW(8) ES 154.757D0 WVHPSW(1) ES 155.75700 WVHPSW(2) ES 156.757D0 WVHPSW(3) ES 157.757D0 WVHPSW(4) ES 158.757D0 WVHPSW(5) ES 159.75700 WVHPSW(6) ES 160.757D0 WVHPSW(7) ES 161 1 144D5 WVHPSW(8) ES PWWSPR WETWELL DPRES SET PT F 162 1.4339D5 ES 163 600 00 i ES TDSPR 164 7 38D5 K III CONTAINMNT SPRAY TIME DELAY FOR MARK III CONTAINMENT ES PDSRV1 DEAD BAND FOR CLOSURE OF SRV 165 9.45D5 PDSRV2 DEAD BAND FOR CLOSURE OF SRVt 166 1.151D6 SPRAYS PDSRV3 DEAD BAND FOR CLOSURE OF SRV 167 5.17D5 01 ES PDSRV4 DEAD BAND FOR CLOSURE OF SRV 168 0.00 2 ES
8 POINTS ARE USED HERE TO DEFINE TPDSRV5 D 03 ES 185 8.22D6 t4 ES PTURRI(1) 186 1.3406 t5 ES PPS-PWW VS. STEAM FLOW TO RCIC 187 1 34D6 PTURRI(2)
ES 188 1.34D6 PTURRI(3) NE 199 1 34D6 PTURRI(4) ES 190 1.34D6 PTURRI(5) ES 191 1.34D6 PTURRI(6) ES 192 1.34D6 PTURRI(7) ES 193 4 8300 PTURRI(8) ES WSTRCI(1) ES ES ES ES \\
A-8 i 194 1 56D0 WSTRCI(2) ES 1 195 1.56D0 USTRCI(3) ES 196 1.56D0 WSTRCI(4) ES 197 1 5600 WSTRCI(5) ES 198 1.5600 WSTRCI(6) ES 199 1.56D0 WSTRCI(7) ES 200 1.56DO WSTRCI(8) ES ~ 201 2.737D5 PHTURH HIGH TURBINE EXHAUST PRESSURE FOR HPCI ES 202 1.72D5 PHTURR HIGH TURBINE EXHAUST PRESSURE FOR RCIC ES 203 4.916D5 PCFAIL CONTAINMENT FAILURE PRESSURE ES
THE SHUT OFF HEAD SHOULD APPEAR IN THE PUMP CURVE DEFINITION FOR ECCS

THE NEXT TWO PARAMETERS ARE PERMISSIVE SIGNALS FOR TRIPPING SYSTEMS 204 1.D10 PHLPCI HIGH VESSEL PRESSURE TRIP FOR LPCI ES 205 1.D10 PHLPCS HIGH VESSEL PRESSURE TRIP FOR LPCS ES 206 34.237DO 2HISP HIGH SUPP. POOL LEVEL TRIP FOR HP SUCTION ES 207 47.16D0 ZLSPR NOT USED

ALL OF THE HEAT EXCHANGER DATA MAY BE OMITTED WITH THE EXCEPTION

OF NTUHX1 NTUHX2:NHX1,NHX2 208 0.D0 NTHX NUMBER OF TUBES IN RHR HTX ES 209 0.D0 NBHX NUMBER OF BAFFLES IN RHR HTX ES 210 0.D0 XIDTHX TUBE ID FOR RHR HTX ES 211 0.D0 XTTHX TUBE WALL THICKNESS FOR RHR HTX ES 212 0.D0 XTCHX TUBE CENTER TO CENTER SPACING FOR RHR HTX ES 213 0.D0 XSHX SHELL LENGTH FOR RHR HTX ES 214 0.D0 RGFOUL FOULING FACTOR FOR RHR HTX ES 215 0.D0 KTHX THERMAL CONDUCTIVITY FOR TUBE WALL (RHR HTX)
ES 216 0.D0 XBCHX BAFFLE CUT LENGTH FOR RHR HTX ES 217 0.D0 XIDSHX SHELL ID FOR RHR HTX ' ES 218 0 00 XSTHX BUNDLE TO SHELL GAP LENGTH FOR RHR HTX ES
NTU VALUES NOT NEEDED IF ABOVE INFORMATION IS DEFINED 219 1 2D0 NTUHX1 NTU FOR RHR HTX 91 ES 220 1 200 NTUHX2 NTU FOR RHR HTX #2 ES 221 2.00 NHX1 NUMBER OF RHR LOOP #1 HTX ES 222 2.00 NHX2 NUMBER OF RHR LOOP #2 HTX ES 223 2.D0 FHX TYPE OF RHR HTX(1= STRAIGHT TUBE,2=U TUBE)
ES 224 14.4D3 TDBATT BATTERY OPERATION TIME FOR STA1 ION BLACK-0UT ES
THE FOLLOWING ARE NPSH CURVES AND THE FIRST ENTRY FOR THAT SYSTEM

CORRESPONDS TO THE FIRST FLOW RATE LISTED ABOVE FOR THAT PUMP 225.518D0 ZHDHPS HPCS NPSH FOR GIVEN FLOWS ES 226.518D0 (METERS)
ES 227.518D0 ES 228.549D0 ES 229.61D0 ES 230.762D0 ES 231 1.37200 ES 232 2.22600 ES 233 2.073D0 ZHDLPI LPCI NPSH FOR GIVEN FLOW ES 234.671D0 ES 235.335D0 ES 236.335D0 ES 237.335D0 ES 238.335D0 ES 239.36600 ES 240.39600 ES 241 2.457D0 ZHDLPS LPCS NPSH FOR GIVEN FLOW ES 242 1 524D0 ES 243.85400 ES 244.854D0 ES 245.85400 ES 246.854D0 ES 247.854D0 ES 248.85400 ES 249 31.400 ZCLRCI PUMP CENTER LINE ELAVATION FOR RCIC ES
A-9 250 28 3500 251 ACVENT AREA 0F CONTAINMENT .0093D0 252 0 0D0 253 28 35D0 ZCFAIL ELEVATION OF CONTAINMENT N FOR HPCI 254 0.D0 ZSRVD NT AVERAGE ELEVATION OF SRV DISC 255 0.D0 TGDWHX(1) ES VENT IN WETWELL (MII ON COOLING CURVE FOR D 256 0.D0 TGDWHX(2) ES HARGE IN SUPP P0OL TEMP IN DRYWELL VS.KYWELL COOLERS 257 0.D0 TGDWHX(3) 258 0.D0 TGDWHX(4) 259 0.D0 TGDWHX(5) HEAT LOES RATE (J/S) ~ ES 260 0.D0 TGDWHX(6) 261 0.D0 TGDWHX(7) 262 0.D0 TGDWHX(B) 263 0.D0 OGDWHX(1) HEAT LOSS RATE FDR DPYWELL CO 264 0.D0 OGDWHX(2) 265 0 00 OGDWHX(3) 266 0.D0 OGDWHX(4) RS (J/S) 267 0.D0 OGDWHX(5) 268 0.00 DGDWHX(6) 269 0.00 DGDWHX(7) OGDWHX(8)
DRYWELL 01 5.D-1 02 RELHDW RELATIVE HUMIDITY IN DR 03 7650.200 VOLDW 04 30 7100 DW VOLUME OF DRYWELL YWELL DW ZDWF 05 318.400 AREA 0F DRYWELL FLOORELEVATION 06 35.7700 ADWF DW IWDWWW ELEVATION OF WEIR WALL B 07 24.E0 DW NIGDW 9 27D0 DW AVERAGE DISTANCE FROM IGNIT 08 XIGDW 09 318.400 DW 318.00 ACHDW WELL DW CHARACTERISTIC FLOOR AREA FOR ASEDDW AEROSOL SEDIMENTATION R TO CEILING OW AREA BURN CALCULATION

WETWELL DW 01 28 35D0 DW 02 03 5.07D-2 ZWWF ELEVATION AT WETWELL FLOOR 3.0D0 AVB WW FLOW AREA THROUGH VACUUM BREAK 04 6.D3 NVB WW PSETVB PRESSURE SETPOINT FOR VHUM 05 6.D3 WW ERS 06 7.9 PDVB WW TOTAL VOLUME OF WETWELLDEAD BA 07 6.D856D3 VOLWW WW 1
08 6.D0 RELHWW RELATIVE HUMIDITY IN WE (PLUS SU WW 09 NIGWW .6DO WW AVERAGE DISTANCE FROM IGNIT 10 XIGWW 619 3D0 TWELL OOL) WW 11 0.D0 ACHWW WELL WW CHARACTERISTIC FLOOR AREA FORR TO C WW 12 619.D0 AWWF WW AREA 0F WETWELL FLOOR ASEDWW AEROSOL SEDIMENTATION (MARK II) BU WW
PEDESTAL AREA WW 01 02 3 269D1 WW i
03 3.95100 APDF AREA 0F PEDESTAL FLOOR 04 2 678D2 APDVT PD VOLUME OF PELESTALAREA 0F PEDESTAL-05 31.72D0 VOLPD PD ZWPDDW ELEVATION OF W1ER BETWE 06 28 80D0 NG PD ZPDF 5.D-1 PD ELEVATION AT PEDESTAL FLOOREN PED AND 07 RELHPD RELATIVE HUMIDITY IN P 08 0.D0 PD NIGPD 0 00 PD AVERAGE DISTANCE FROM IGNITE 09 XIGPD EDESTAL PD 32 6900 10 0 00 ACHPD PD CHARACTERISTIC FLOOR AREA FORR TO CE PD STAL THE NEXT PARAMETER-ADCPD-CANXWPDV PD NOTEt LEAK AREA BETWEEN THE DRYWELL A BURN CA s 11 0.0043D0 (MARK II ONLY)LCULATION PD ND COMPARTMENT A 0FBE USED TO MO ADCPD PD AREA 0F PEDESTAL DOWNCOMERS PD A MARK III m
A-10 I i 12 0.D0 NDCPD NUMBER OF PEDESTAL DOWNCOMERS 13 2.D0 XHPDDW DISTANCE BETWEEN UPPER AND LOWER VENTS FOR PED-DRYWELL NATURAL CIRCULATION 14 32.5D0 ASEDPD AEROSOL SEDIMENTATION AREA SP st SP SP
SUPPRESSION POOL (MARKIII ONLY) l 01 5.15D1 ASPDW AREA 0F DRYWELL SIDE OF SUPPRESSION POOL SP l
02 6.193D2 ASPPC AREA 0F CONTAINMENT SIDE OF SUPPRESSION POOL SP i 03 4.5D1 NVT1 NUMBER OF VENTS OF TYPE 01 -- TOP SP t 04 4.5D1 NVT2 NUMBER OF VENTS OF TYPE 42 -- MID SP 05 4.5D1 NVT3 NUMBER OF VENTS OF TYPE 43 -- BOTTOM SP i 06 7.1D-1 XDIAVT DIAMETER OF ONE SUPPRESSION POOL VENT SP 07 33.449D0 ZLLSP ELEVATION OF SUPP. POOL LOW LEVEL SETPOINT SP 08 32.16D0 ZUT1 ELEVATION OF TOP OF VENT TYPE il SP 09 30.89D0 ZVT2 ELEVATION OF TOP OF VENT TYPE 02 SP 10 29.6200 ZVT3 ELEVATION OF TOP OF VENT TYPE 63 SP IN IN IN 1
INITIAL CONDITIONS 01 3.833D9 QP0bER CORE POWER IN 02 7.17D6 PPSD INITIAL PRESSURE IN PRIMARY SYSTEM IN 03 1.005 PPD 0 INITIAL PRESSURE IN PEDESTAL IN 04 1.0D5 PDWO INITIAL PRESSURE IN DRYWELL IN 05 1.005 PWWO INITIAL PRESSURE IN WETWELL IN 06 34.01D0 ZSPDWO INIT.ELEV. OF WATER LEVEL IN DW SIDE OF SUPP. POOL IN 07 34.01D0 ZSPWWO INIT.ELEV. OF WATER LEVEL IN PC SIDE OF SUPP. POOL IN 08 3.3D2 TPD0 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 TWSPD INITIAL TEMPERATURE OF SUPPRESSION POOL WATER IN 12 51.91D0 ZWSRO INITIAL ELEVATION OF WATER IN THE SHROUD IN 13 2 05D6 MWCBD MASS OF WATER IN UPPER POOL (MARKIII ONLY) IN 14 896.6D0 VCSTO VOLUr.E OF WATER IN CONDENSATE STORAGE TANK IN 15 297.D0 TAMB AMBIENT TEMPERATURE IN 16 1.D5 PAMB AMBIENT PRESSURE IN CC $3 CC st 3 CONTROL CARDS CC 01 3 IBWR CONTAINMENT TYPE (MARK 1,2e OR 3 ) CC 02 46 IRSTW UNIT NUMBER TO WRITE RESTART FILE (MAIN) CC 03 47 IHUW UNIT NUMBER TO WRITE RESTART FILE (HEATUP) CC 04 40 IPOUT UNIT NUMBER TO WRITE PROGRAM OUTPUT FILE CC 05 41 IPLT1 FIRST PLOT FILE (5 FILES) CC 06 500 IPTSMX MAXIMUM NUMBER OF PLOTTED POINTS CC 07 4 IPTSPK MAXIMUM NUMBER OF PLOT POINTS TRACED FOR FULL CC SCALE SPIKE CC 08 80 IPTSAV NUMBER OF POINTS SAVED FOR NON-CHANGING PLOT CC 09 1 ISUMM
SUMMARY
DATA (0=ALL EVENTSe1= SHORTER LIST) CC 10 48 ISUM
SUMMARY
FILE NUMBER CC 11 1 IRUNG 1=1ST ORDER R-Ke2=2ND ORDER R-K CC 12 1 IFREEZ 1= DO FREEZE FRONT CALC. (0=NO CALC.) CC 13 6 INPGRP NUMBER OF TRACE GAS TYPES (FISSION PRODUCTS) CC 14 0 IRET 1= WRITE RETAIN FILE (0=NO FILE) CC 15 49 IFPPLT RETAIN PLOT FILE UNIT NUMBER CC TD TD 88 TD
TIMING DATA 01 20 00 TDMAX MAXIMUM ALL0btB TIME STEP TD 02 1.D-3 TDMIN MINIMUM ALLOWED TIME STEP TD 03 5.D-2 FMCHMX MAXIMUM MASS CHANGE (1) FOR INTEGRATION TD 04 5.D-2 FUCHMX MAXIMUM GAS TEMP CHANGE FRACTION FOR INTEGRATION TD 05 4.D2 MAXMST MAXIMUM MASS OF STEAM CHANGE PER TIME STEP IN PS TD
l A-17
COMPTA (MARKIII-MIDDLE WETW 01 02 41 25D0 1.1 ZCAF ELEVATION OF HCU DECKELL COMPA 03 6.D589D4 UOLCAVOLUME OF 1
CA 04 RELHCA RELATIVE HUMIDITY I 05 325.00 ACAF COMPARTMENT A CA 348.D0 AREA 0F COMPT. A FLOORN COMPT. A CA CA 06 ACACB 07 139.D0 FLOW AREA BETWEEN COMPT 08 41 25D0 AWWCA FLOW AREA BETWEEN WETWELL AN. A CA ZWCAWW 09 58254 00 PPUR(1) CURB HEIGHT ON MIDDLE DECK CA 10 87140.D0 DRYWELL PURGE PRESSURE VS FLD C CA PPUR(2) 11 1 05D5 CA 12 1 128DS PPUR(1) CA OW (M**3/KG) CA 13 1 162D5 PPUR(4) 14 1 162D5 PPOR(5) 15 1 16205 PPUR(6) CA 16 1 162D5 PPUR(7) CA CA 17 555D0 PPnR(8) CA 18 .51500 WVPUR(1) CA 19 .453DO WVPUR(2) CA 20 .40100 WVPUR(3) 21 .330D0 WVPUR(4) CA CA 22 .330D0 WVPOR(5) WVPUR( CA 23 330D0 WVPUR(4) CA 3 24 2.30D0 WVPUR(8) CA 7) 25 47.D0 NPURP CA 16D0 NUMBER OF CA 26 27 1 1437D5 ZLPUR PDWPUR HIGH DRYWELL PRESSLOW WA CA 28 6.D3 LOCA) CA PDPUR SIGNAL FOR BRYWELL 29 30.D0 PRESSURE DIFFERENTIAL SET CA TDPUR 30 45.D0 TIME BELAY FOR DRYWELL PUR CA 31 B.34D0 NIGCA NUMBER CF IGNITERS IN THE POINT FOR DRYWELL PURGE CA 32 6.D0 XIGCA NIGBCA NUMBER OFAVERAGE DISTANCE FROM GE CA 33 325 00 COMPT A CA ACHCA 543.D0 CHARACTERISTIC FLOOR AREA FIGNIT CA ASEDCA AEROSOL SEDIMENTATI CA ON AREA OR BURN CALCULATIONCOMPT CA
COMPTB (MARKIII-UPPER WETWE 01 CA 63 6900 02 237 ZCBF CA LL COMPARTMENT) 03 6 066 5D0 VOLCBELEVATION DF 0?ERATING DECK 04
64. 1 VOLUME OF C0F9T CB RELHCB RELATIVE Pustt) '. B 05 302.D0 CB ZWCBWW CURB HEIGHT 9a : "Y IN COMPT.B 200 06 AWCB CB i
07 7 5834D4 PCPUR(1) PRESSURE VS FLOWWATER AR CB 08 0.D0 CB PCPUR(2) 09 0.D0 1 CB PCPUR(3) 0 00 10 0 PCPUR(4) FOR CONTAINMENT PURGE CB 11 0 D0 PCPUR(5) 00 CB i 12 0.D0 PCPUR(6) CB t 13 14 0.D0 PCPUR(7) CB 15 1 41600 PCPUR(8s CB P" CPU *((2)WVCPLT !) CB 16 1 414D0 CB 17 1 416D0 18 1 416D0 WVCPUR(3) CB 19 1 416D0 WVCPUR(4) CB CB 20 1 416D0 WVCPUR(5) CB 21 1 416D0 WVCPUR(6) CB 8 WYCPUR(7 1 416D0 WVCPUR(8)) 22 1. CB 23 47.03D3 VOLUPD CB 16D0 VOLUME OF WATER 24 CB 1 1437D5 ZLUPD LOW WATER (LOCA)IN UPPER POOL DUMP CB 1 PDWUPD HIGH DRYWELL PRESSURE (L FOR UPPER SIGNA CB CB LOCA) FOR UPPER POOL DUMP MP CB CB =\\ CB
A-12 J 25 225.D0 TDDUMP TOTAL TIME FOR UPPER POOL DUMP CB 26 1.8D3 TDUPD TIME DELAY FOR UPPER POOL DUMP CB 27 56.2500 ZCBF1 ELEVATION OF UPFER P0OL FLUUR CB 28 1031.D0 VLAUPD YOLUME F WATER REMAINING IN UFFER PGOL AFTER CB UPPER POOL DUMP CB 29 18.D0 NIGCB NUMBER OF IGNITERS IN THE COMPT B CB 30 .3D0 XIGCB AVERAGE DISTANCE FROM IGNITER TO CEILING CB 31 16.D0 MIGBCB NUMBER OF IGNITERS IN COMPT'A SEEN IN COMPT B CB 32 1121 9DO ACHCB CHARACTERISTIC FLOOR AREA FOR BURN CALCULATION CB 33 857.D0 ASEDCB AEROSDL SEDIMENTATION AREA HS HS
HTSINKS HS
REFER TO DRAWING IN YOL II 0F MAAP USERS MANUAL ON MARKIII HEAT SINKS 01 184.6D0 AHS1 AREA 0F WALL 61 HS 02 1074.890 AHS2 AREA 0F WALL 62 HS 03 492.3D0 AHS3 AREA 0F WALL 43 HS 04 883.8D0 ANS4 AREA 0F WALL 04 HS 05 2735.100 AHS5 AREA 0F WALL 65 HS 06 3411 0D0 AHS4 AREA 0F WALL 66 HS 07 371.1D0 AHS7 AREA 0F WALL 47 HS 08 289.6D0 AHS8 AREA 0F WALL 48 HS 09 2.077D0 KHS1 THERMAL CONDUCTIVITY OF WALL 41 HS 10 2.077D0 KHS2 THERMAL CONDUCTIVITY OF WALL 02 HS 11 2.077D0 KHS3 THERMAL CONDUCTIVITY OF WALL 03 HS 12 2.077D0 KHS4 THERMAL CONDUCTIVITY OF WALL 44 HS 13 2.077D0 KHS5 THERMAL CONDUCTIVITY OF WALL 45 HS 14 2.077D0 KHS6 THERMAL CONDUCTIVITY OF WALL 46 HS 15 2.077D0 KHS7 THERMAL CONDUCTIVITY OF WALL 47 HS 16 2.077D0 KHS8 THERMAL CONDUCTIVITY OF WALL 48 HS 17 1.753D0 XHS1 THICKNESS OF WALL 41 HS 18 1.524D0 XHS2 THICKNESS OF WALL 42 HS 19 1.524D0 XHS3 THICKNESS OF WALL 43 HS 20 1.067D0 XHS4 THICKNESS OF WALL 64 HS 21 1.067D0 XHS5 THICKNESS OF WALL 45 HS 22
.762D0 XHS6 THICKNESS OF WALL 46 HS 23 .61D0 XHS7 THICKNESS OF WALL 47 HS 24 1.295D0 XHS8 THICKNESS OF WALL 48 HS 25 0.D0 XLHSI1 INNER LINER THICKNESS FOR WALL 61 HS 26 0.00634D0 XLHSI2 INNER LINER THICKNESS FOR WALL 42 HS 27 0.00634D0 XLHSI3 INNER LINER THICKNESS FOR WALL 43 HS 28 0.00634D0 XLHSI4 INNER LINER THICKNESS FOR WALL 44 HS 29 0.00634D0 XLHSI5 INNER LINER THICKNESS FOR WALL 45 HS 30 0.00634D0 XLHSI6 INNER LINER THICKNESS FOR WALL 46 HS 31 0.D0 XLHSI7 INNER LINER THICKNESS FOR WALL 47 HS 32 0.0063400 XLHSIB INNER LINER THICKNESS FOR WALL 48 HS 33 0.00634D0 XLHS01 DUTER LINER THICKNESS FOR WALL 41 HS 34 0.D0 XLHS02 OUTER LINER THICKNESS FOR WALL 02 HS 35 0.00634D0 XLHS03 RbTER LINER THICKNESS FOR WALL 43 HS 36 0.00 XLHSO4 NTER LINER THICKFiSS FOR WALL 04 HS 37 0.D0 XLHS05 00TER LINER THICKhESS FOR WALL 45 HS 38 0.D0 XLHS05 OUTER LINER THICKNESS FOR WALL 66 HS 39 0.061200 XLHS07 OUTER LINER THICKNESS FOR WALL 47 HS 40 0.00634D0 XLHS08 OUTER LINER THICKNESS FOR WALL 48 HS 41 2300.00 DHS1 DENSITY OF WALL 61 HS 42 2300.D0 DHS2 DENSITY OF WALL 02 HS 43 2300.D0 DHS3 DENSITY OF WALL 03 HS 44 2300.D0 DHS4 DENSITY OF WALL $4 HS 45 2300.D0 DHS5 DENSITY OF WALL 45 HS 46 2300.D0 DHS6 DENSITY OF WALL 66 HS 47 2300.D0 DNS7 DENSITY OF WALL 47 HS 48 2300.D0 DNS8 DENSITY OF WALL 48 HS 49 800.D0 CPHS1 SPECIFIC HEAT FOR WALL 61 HS
A-13 50 880.D0 CPHS2 SPECIFIC HEAT FOR WALL 62 HS 51 880.00 CPHS3 SPECIFIC HEAT FOR WALL 63 HS $2 880 00 CPHS4 SPECIFIC HEAT FOR WALL 44 HS 53 880.00 CPHS5 SPECIFIC HEAT FOR WALL 65 HS t 54 880.00 CPHS6 SPECIFIC HEAT FOR WALL 46 HS 55 880.00 CPHS7 SPECIFIC HEAT FOR WALL 67 HS 56 880.00 CPHS8 SPECIFIC HEAT FOR WALL 48 HS
ALL 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 MEDDW MASS OF EQUIPMENT IN DRYWELL HS 59 0.00 ME0WW MASS OF EQUIPMENT IN WETWELL HS 60 342462.00 MEOCA MASS OF EQUIPMENT IN COMPT A HS 61 1.958106 MEOCS MASS OF EQUIPMENT IN COMPT B HS 62 0.D0 AEOPD AREA 0F EQUIPMENT IN PEDESTAL HS 63 4153.D0 AEDDW AREA 0F EQUIPMENT IN DRYWELL HS 64 0.00 AEQUW AREA 0F EQUIPMENT IN WETWELL HS 65 9.7177D3 AEOCA AREA 0F EQUIPMENT IN COMPT A HS 66 1189.200 AEOCB AREA UF EQUIPMENT IN COMPT B HS 67
$0 00 HTOUTW HEAT TRANSFER COEFF. AT GUTER WALL HS 68 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 41 HS 69 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 12 HS 70 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 43 HS 71 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE $4 HS 72 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE 45 HS 73 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 66 HS 74 0.00 RGAP INNER LINER TO WALL GAP RESISTANCE 67 HS 75 0.D0 RGAP INNER LINER TO WALL GAP RESISTANCE 6B HS 76 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE 41 HS SfS Wh A E 79 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE 94 HS 80 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE $5 HS I2I 0.D0 RGAP OUTER LINER TO WALL GAP RESISTANCE $6 HS 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE #7 HS 83 0.00 RGAP OUTER LINER TO WALL GAP RESISTANCE 48 HS 84 0.00 ME0WWS MASS OF EQUIP. HEAT SINK WETWELL (SUBMERGED) HS g 0.00 AE0WWS AREA 0F EQUIP. HEAT SINK WETWELL (SUBMERGED) g HS
MODEL PARAMETERS FOR BWR 01
.00500 FRCOEF FRICTION COEFFICIENT FOR CORIUM IN VFAIL MO 02 2.00-1 FMAXCP FRACTION OF TOTAL CORE MASS WHICH MUST MELT TO FAIL THE CORE PLATE 03 50 00 HTSLAD FUEL CHANNEL TO CONTROL BLADE HEAT TRANS. COEFF MO 04 300 00 HTFB FILM BOILING HEAT TRANS COEFF, MO 05 0.00 FBLOCK FUEL CHANNEL BLOCKAGE PARAMETER MD 0= BLOCKAGE AT TZOOFF,1=ND BLOCKAGE MO 06 2300.D0 T200FF OXIDATION CUT-OFF TEMPERATURE MO 07 .300 FACPF FRACTION OF CORE PLATE AREA THAT FAILS MO 08 5.00 CDSPD FLAME BUOYANCY DRAG COEFFICIENT IN THE PEDESTAL MO 09 5.D0 CD9DW FLAME BUOYANCY DRAG COEFFICIENT IN THE DRYWELL MO 10 5.00 CD9WW FLAME BUOYANCY DRAG COEFFICIENT IN THE WETWELL MD 11 5.00 CDSCA FLAME BUOYANCY DRAG COEFFICIENT IN COMPARTMENT A N 12 5.D0 CD9CB FLAME BUOYANCY DRAG COEFFICIENT IN COMPARTMENT B HD 13 .1000 XCNREF CORIUM REFERENCE THERMAL BOUNDARY LAYER THICKNESS MD 14 1.03 HTCMCR CORIUM-CRUST HEAT TRANSF, COEFF. USED IN DECOMP MO g 0.05D0 XCMX guCgig TgNESSONDRYWELLFLOORANDPEDg 16 0.0100 XDCMSP PARTICLE SIZE (DIAMETER) FOR CORIUM AS IT FALLS M0 INTO SUPPRESSION POOL (MARK II ONLY) NO 17 983.D0 TCFLAM CRITICAL FLAME TEMPERATURE MO 18 1.5300 FCHTUR CHURN-TURBOLENT CRITICAL FLOW PARAMETER MO 19 3.7D0 FDROP DROPLET CRITICAL FLOW PARAMETER MD .--._-r-_. ,n,,
1 A-14 1, 20 3.D0 FFLOOD FLOODING FLOW PARAMETER 21 1.D0 FSPAR PARAMETER FOR BOTTOM-SPARGED STEAM VOID FRACTION MD 22 2.00 FVOL MD PARAMETER FOR VOLUME SOURCE VOID FRACTION MODEL 23 5.D-1 TTENTR MO ENTRAINMENT EFFECTIVE EMPTYING TIME 24 .900 EW MD EMISSIVITY OF WATER 25 .8500 EWL EMISSIVITY OF WALL MD 26 .85D0 ECM EMISSIVITY OF CORIUM MD 27 .600 EG EMISSIVITY OF GAS MD 28 .8500 EED EMISSIVITY OF EOUIPMENT MD 29 0 5D0 F0VER FRACTION OF CORE SPRAY FLOW ALLDED TO BYPASS COREMO MO 30 1.D0 NPF NUMBER OF PENETRATIONS FAILED IN LOWER HEAD 31 2 00 FCDCDW MO 4 DOWNCOMER PERIMETER PER METER FROM PEDESTAL DOOR MO i (MARK II ONLY) 32 0 1400 FCHF COEFFICIENT FOR CHF CORRELATION IN PLSTM MO 33 .75D0 FCDBRK MO DISCHARGE COEFFICIENT FOR PIPE BREAK 34 .33D0 FENTR MD NUMBER TO MULTIPLY KUTATELADZE CRITERION BY TO MO REPRESENT DIFFICULTY (GT 1.D0) OR EASE (LT 1.DO) M0 i FOR MATERIAL TO BE BLOWN OUT OF CAVITY 35 1.00 SCALU SCALING FACTOR FOR ALL BURNING VELOCITIES MO 36 1.00 SCALH MO SCALING FACT 00 FOR HT COEFFICIENTS TO PASSIVEMO HEAT SINKS 37 2.000 FUMIN CLADDING SURFACE MULTIPLIER MO 38 .022 FHE1 FIRST COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 39 .003 FHE2 SECOND COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 40 .590 FNE3 MO THIRD COEFF IN HENRY-EPSTEIN AEROSOL MODEL 41 .330 FHE4 MO FOURTH COEFF IN HENRY-EPSTEIN AEROSOL MODEL 42 1.00-5 DHE MO DENSITY LIMIT IN HENRY-EPSTEIN AEROSOL MODEL 43 1.0 FCSIVP MO GROUP 2 (CSI) AND GROUP 6 (CSOH) VAPOR PRESSURE l MULTIPLIER - NUMBER USES ANL CSOH VAP PRESSe+ USES SANDIA CSON VAP PRESS 3 CONCRETE PROPERTIES MO 01 56.D0 MOLWCN MOLECULAR EIGHT OF CONCRETE CN 02 1743.D0 TCNMP MELTING TEMPERATURE OF CONCRETE CN 03 .806 LHRCN REACTION ENERGY FOR CONCRETE DECOMPOSITION CN 04 45.D0 DCFVCN FREE WATER DENSITY IN CONCRETE CN 05 65.D0 DCCWCN COMBINED WATER DENSITY IN CONCRETE CN 04 572.D0 DCC2CN C02 DENSITY IN CONCRETE CN 07 1.De LHCN LATENT HEAT TO MELT CONCRETE CN CN
FISSION PRODUCTS 01
.020D0 FOP (1) PERCENT POWER IN FISSION PRODUCT GROUP 1 FI 02 .151D0 F0P(2) PERCENT POWER IN FISSION PRODUCT GROUP 2FI 03 .018D0 FOP (3) PERCENT POWER IN FISSION PRODUCT GROUP 3 FI 04 0.D0 FOP (4) PERCENT POE R IN FISSION PRODUCT GROUP 4FI 05 0.D0 FOP (5) PERCENT POWER IN FISSION PRODUCT GROUP 5 FI 06 .008D0 FOP (6) PERCENT POWER IN FISSION PRODUCT GROUP 6 FI 07 439 3 FI MFP(1) MASS OF FISSION PRODUCT GROUP 1 -NOBLES 08 36.2 MFP(2) MASS OF FISSION PRODUCT GROUP 2 -CSI FI 09 37.1 FI MFP(3) MASS OF FISSION PRODUCT GROUP 3 -TE 20 178 7 FP(4) MASS OF FISSION PRODUCT GROUP 4 -SR FI 11 435 0 MFP(5) MASS OF FISSION PRODUCT GROUP 5 -RU FI 12 227 3 NFP(6) MASS OF FISSION PRODUCT GROUP 6 -CSOH FI 13 1190.D0 FI MSM0(1) MASS OF SN IN CORE REGION 14 268 00 MSM0(2) MASS OF MN IN CORE REGION FI 15 0.000 FDSP(1) SPRAY REMOVAL LAMDA FOR FP GROUP 1 FI 16 0.000 FDSP(2) SPRAY REMOVAL LANDA FOR FP GROUP 2 FI 17 0.000 FDSP(3) SPRAY REMOVAL LAMDA FOR FP GROUP 3 FI 18 0.000 FDSP(4) SPRAY REMOVAL LAMDA FOR FP GROUP 4 FI 19 0.000 FDSP(5) SPRAY REMOVAL LAMDA FOR FP GROUP 5 FI 20 0.000 FDSP(6) SPRAY REMOVAt LAMDA FOR FP GROUP 6 FI 21 600.D0 FDFSP(1) DRVELL VENTS DECON. FACTOR FOR FP GROUP 1 FI FI ,w- ,v w,e-. - -~ .y,-,-.,--
A-15 22 600.D0 FDFSP(3) DRYWELL VENTS DECONF 23 600.0D0 24 600.0D0 . FACTOR FOR FP GROUP FDFSP(4) DRYWELL VENTS DECON 25 600.0D0 FDFSP(6) DRYWELL VENTS DEC 26 600.000 FI . FACTOR FOR FP GROUP 4 27 1000.D0 FI FDFRV(1) SRV DECON. FACTOR FOR FP G 28 FOR FP GROUP 5 FI 1000.D0 FDFRV(2) SRV DECON. FACTOR FOR FP 29 ROUP 1FOR FP GROUP 6 ~ FI 30 1000.0D0 FDFRV(3) SRV DECON. FACTOR FOR FP 31 1000.0D0 FI FDFRV(5) SRV DECON. FACTOR F GROUP 2 FI 1000.0D0 32 GROUP 3 FI 1000.0D0 FDFRVf6) SRV DECON. FACTOR FOR FP GR OUP 4 FI
IF DESIRE TABULAR 00TFUT AND OPERATOR ROUP 5 FI

INSERT A *BR HERE (EG FOR OUP 6 FI

INTERVENTIONS, AND TABULAR OUTPUT)
A P;aAMETER DUMP IN BRITISH, OPERATIN FI
BR FI OR
A-17 APPENDIX A.2 MAAP Input Files for Section 4 Sequences R 1 F _ T230W - GRAND GULF 1 1 25 T23C -- GRAND GULF 0 1 1 25 10,3 36 0 10,3,,37 10 1 1 30 1, ,441.0 l i 10, 31 1, ,722.0 ,722.0 i 10, ,38 10, 1,, ,722.0 0,0,,39 1 ,722.0 1, ,722.0 0 1, ,722.0 0 1, ,722.0 80. 1, .1015.4 2.0 1, ,' 29.01 215 1, ,29.01 1 l' ,29.01 0 10 1 145 1, ,29.01 ,29.01 0 1, ,29.01 225 1, ,29.01 1, 0 3, .18 1 ,1.5 12 0,,0,0 199 0 0 0 278 30 1 .5 205 215 1 1 207 257 1 1 211 0 1 10 203 145 1 0 232 225 1 1 260 0 1 12 0 199 12 0 8 205 0 1 1000 207 1 0' 211 0,,b* $03 1 232 60 1 0 12 8 0 1000 6,11,0.0 0 0,0 0,
A-18 1 1 T10tN,W ADS -- GkAND GULF AE -- GPAND GULF 1 1 25 25 0 0 1 1 l 10.2,46 1,48,3.14 00,0,0 10, 47 ~ 10. 40 10 ', 41 0 10 ,48 70 7 10 ,49 0,, 256 0 1 0 211 1 60. 5. 205 250 1 1 203 213 1 1 232 211 1 1 260 205 1 1 207 203 1 1 0 232 12 1 8 260 0 1 1000 207 6,11 0 1 0,0,d 0 0 8 0 154.72 0 225 1 0 12 8 0 1000th** 0 t
8-1 I 1 APPENDIX B Su lement&1 Plots for the WASH-1400 C om arison se uences B m
B-3 SUPPLEMENTAL PLOTS FOR SEQUENCE T QUV j
T10UV - CiRAHD GULF 1 d 10-4 / t // i 8_ i.. 4 j: / w I 03 6-i j-/ J t l oa / 4 R4. /. 7 f / 4 o 2-
/
l 0,,,,,,,,,,',,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, O. 00 0.20 0.40 0.66 0.60 1.00 1.20 1.40 1.60 1.80 2.00 2,20 TITE LHOURS) 1 Fig. B.1 Total in-vessel H2 generated. 1 1 1 e m -.a
i l T10UV - GRAND GULF i 1 t 3, 000, t L d ~ f' ] I -.4 s/ .. b u e -- ~. .-r i f.1 i i i i t -} /' i / 2, egg- ..7.. / .i. .-j ca ~ '". / i i d i i i a / l I -j
1. 500 -
i-o o / i i = ~ { 1.b00 _ .l. E 7- .'( I i r 500-4 .j. /' i i i t i i 4 L s (. .,e. _,__._.7,. __,q_,__...,,._,_,_,_g_, _r,_,,,,,, O 10 29 30 40 58 M 70 8e l IItE (HOURSI I J ]- Fig. B.2 Total H2 generated. I 1 t i en
-e--
{ l L i T10UV - GRAHD GULF 50, l J l _;\\ [- - .;...-....i...+ f .h Q _ e 6-e s- \\ \\ ) \\- TOP OF ACTIVE FUEL )
,, __ - - _ - -)- _.- - - _ - - - - - - - - - - - - - - - - - - -
m e i i x \\ en 'e g N i. a Y l ) m BOTTOM OF ACTIVE FUEL 10- - + - 0 ,,,,,,,,,i,,,,,,,i,,,,,,,i,,,,,,,,,,,,,,,i,,, 0.00 0.20 0.40 0.60 0.00 1.00 1.20 1.49 1.60 1.00 2.00 2.20 TITE tHOURS) I Fig. B.3 Reactor vessel water level. 1
T10UV GRAND GULF 2,000-, /% u. [ i 1,500-e . I m...-.;-.../s-l-+, ... -.. + -.... -. + - - s 1 N ureta ettuun I o 3m 1.000- ...-/: * * " " * " T I u. O ~ u i m a l l 3 i d .i i p. 1 S EI D dl j ..t.. IL 2m F LJ = 0 ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, +,,,, 0 10 20 30 40 50 60 78 88 TIME LHOURS) Fig. B.4 Temperature of structure, F.
l T10UV GRAND GULF E 3g ees,sas-, _ --. .r....... r.- 3 g W 25.h90. 990 - t~- t- ~ b t-- ? i O 3 EV 20, p(sg, g A A $ ..j. g Z ~ O ~ E ~ 15, ggg, eeg N s.--..-.... ..+. k U W ~ m -l / \\ O 16. ^t3, 20* - t - - - + i O ~ i ~ 5 o ~ f C t:R L D O W H C h rt E li 3 s a 3, eee, see _. 3...... ....~U P P E R P L E N U N ""+"" [ O l% W E fi, \\.,v ' li \\ / z .. !.' l. O e 4 s g i i.. ,,..,,....i e 0 10 20 30 40 50 60 70 80 is. TIME (HOURS) Fig. B.5 Fission product decay power on structure, Btu /hr.
- -. - - =. _ - i I I T10UV -- GRANil GULF e i, l 80, 000-. 7c. 0e0}. ,f - i t f.0. 00 0 -- / - i- -i i / 530.000f .,/ d .? + -l~- - -4 \\ 40.000f l l-a .f :. J 5 o ./, .i. .L ...a ao i i O 30 0Gf-e E j ,a l ./ .i. 20,000- / i 1 ,s 4 i 10.000-~ t j / ~ j r ) ) 0g, ,_,-.,,_.,y_,_,._,_,.- _,,,, r,_,_, _,_,_, 0 10 20 30 40 50 GO 70 90 i 'l1ME 0100RSI e-i Fig. B.6 Total C0 generated. A l. -w-e w
T'lOUV - GRAND GULF 1 2ee. coe, I. i i t 155,000-i- -i i 1 d fez, soc _ I. c3 n 5 8* o d i i se,esel. ~ .i. - --4 s J i .., _..___ _,,,, __ i _g,,, _, j, 7,..,._; _ i _.,.,.,.. p, _,, q o... ,.,,....,___..,__i. 0 10 20 30 40 50 60 70 90 1IK til0 URSI / 4 Fig. B.7 Mass of water in the pedestal. =
f T10UV - GP41D GULF i i
0. 070 s
a 0, 0 6 0 d. / g- .i j. / i / O. 650-- ,/ ...i. t .j/ J. 9, /[ '\\ ./ ~ f l ( l 0 040- -[ * % s ;]- -h - --- J m l M x 1 l 4 ~ . }'. i 0 030 c .1. g. . i... m. \\ a m 0 020-h... i i \\! C 010 / N....,j .t. r .i. . i 1 i / i 0 600 T --/--,. y,----,--,,- -,,, i m,,,,,,,,_,4 ,,q,,,,, 0 10 70 3d in 50 60 78 88 I It1E 0100RSI I 3 / Fig. B.8 Mole fraction of H in Compartment B. 2 i j l }
T10UV - GRAND GULF l i 0, 200, ? i _,s._%, i 1 I i \\ I ) i 0.150- ..i. ~ .._3 1 i l. i. i m 5: O I " 0.200- ..): - i j ..j Y N \\ s. 1 '5 i i 1 \\ ro s N U. 85 04. 4 ' ~. N...... 'l. i. '"j i i g i ~ i i i i N'-8 Ba+4'emmmmu,.am-uh w a,mm em .,p .m m O 10 20 30 40 50 60 70 se IINE (HDURS) i Fig. B.9 Mole fraction of 0 in Compartment B. 2 I
1 '110UV - GRAND GULF f e.60-t 50 4. .c n n =? 0 A.; ~ i i i ,/ i. i. ~ .. /.'.
g. 4 0
..p. .v. ..p. ... j / o .( m o E. 3D- /.i + I -f a t W, a I l'i I ? 9.20- ..t. i t i l i .l & 10-- q i 0.00 - , - e',i 1-,, cy, ;' ,,,,,,, g-,,,,,,,,, ,, _ -, -p 0 10 20 30 40 50 60 7J 90 Ilt1E tHOURS) / Fig. B.10 Mole fraction of CO in Compartment B. 2
T10UV - GRAND GULF e.se. 4 i i / f'. .{-. [._f 6,405 /. 4 .i l O.30N ~ - /. -l- .4 cn i/ i i j S i i if i i m i / i 'i ) 9.20- ,/! . /.... i r s i / i i 1 / 0.10-- !.. I. I .i i x - {\\.--s ? %.a. s 0.00 -, ;- r,.., 4,,,- r-),-, v,,, r,, - j -, -,,,j 0 30 20 30 40 50 69 70 GB Ili1E tHDURS) i Fig. B.ll Mole fraction of steam in Compartment B.
t F i i T10UV GRAND GULF i i 250.000, r -s i -\\e-208, Set - t i-i -i- -i u TN2 150,200-- -}- =y- - -j S 4 i [! - 100, 800-m .i. --j i 1 SQ. 000-E h... g...... ,g g g.- y , pg - O 10 20 30 40 50 50 73 86 IIN 'HOURSI i Fig. B.12 Volumetric flow out of containment. J
T10UV - GRAHD GULF 4ee, ees _. _q 350, 000 - .+. ~ i i i 300,000 ~ l + -i. ~ ~ } . --j m.'30,909 d I a 200,000-~ I j ..j g O s 1, i i i i. e .i.
4..
m 7 158.009-l 100,000-~ 1 l. .i. s I I a I.' 50, uso-7 4..
4..
i \\,.._ a 0 ,,--r,, ,-- r,,,,-,, j, y n ;, ,q,,,,j 0 10 20 30 40 50 60 79 88 Iliti tilDURS) / Fig. B.13 Mass of UO in core region. 2 i
B-17 i SUPPLEMENTAL PLOTS FOR SEQUENCE AE f
i e b AE - GRAND GULF i 48 d 35-E- -.. 30 En 25 ca t 201 l--
1..
o e m 15: I -{ .,4. ( 5__ 6 .. j. ; j y / i' .i...i...i...,'...g O i
0. 00 0.20

0. 40 0.60 0.89

1. 09 1.20 1.40 TIE (HOURSI J
Fig. R.14 Total in-vessel H2 generated. 4 6
AE film, Nil GULF J. 00 0, s' / ./ 2.500- . / 2.000-f ? m / a ~ l 1.'s00 / a a / a ~ i co e g a
1. ht3 P _)
./ / A l j ./ et 4 ,i './'
. I.
i i g 0 10 2D 30 40 50 '. n to r I ll l. ill:LRS) J Fig. B.15 Total H2 generated. l
l =4o 0 J ,4 , 1 , 0 I.: ,2 ,1 ~ . 8 l ~ ,0 ev F L , 1 e l E L L-U er U E-F t U-G a F E , 0 w I V.- ,8 S D 'E-R l I T e A T-C , 0 U N V s O I s R C-H e A i v F G A- , 0 r O o E t O- 'M F ,6 T c P-O , 0. I a T e R E O-TT A T-O , 0 6 1 B ,4 B ~ , 6 g i F , 0 ,'/ , 6 , 0 0 0 0 0 0 0 0 0 5 4 3 3~ 1 -[wS I)v2x i
B-21 _a t% s _c l r, /t 1 i J j y _o u_ u 3 J } cI "a D = u x 2 (_b U O.. _ o U) [ Q a. w = O m L <[ ~= a w E o 1 as I )l c.o =- i . _ o "d n s_ i r e LU J ~ <r ~ . _o q m e i / ( .-cn
s. ;
I l N (.% ' - --- ___ _ h. i i i i i i i i i T ' ' d I 'l
i i i O
G o c o o a u:- o o e m c4 .N w g de '38010081S 30 EWn1YW3dM31
B-22 -G I i . 'l: - N, i. ,/ .h t ) N
s a
f C3 C 4 s . 9. - g = 5. i ll: (, {,.L. d i b U3 s. j } ? l. r e i u J.. U) C O LI '") 2 O u,a i l e I o E <T 8 1 C'- i k h f 13 C I C! I /. cN 6 rr h/;-- n = = i .,i F u U j Zi (,- 3 L1J 3! 3 C 3l 5 7 - -S C L...... i ..g.. ....3 - N ,r i -N"> "/ a .m ..y.. ...,,.,, ' ~ ~ '- X'N,- 3 l l
j...
'(N .= T ,.e C) t .h 7 [ i % ___ s
--~_,A
\\s. f grrrT;rrrrirTr.iiii,i... i..iY -o e e e e e e 4 a m o e en e .a e .a .a a. a e e a. .i .i 4 4 4. ed N eN M M N.. d n m n ~ WH/n13 '3WO100H18 NO IVEH AVO3010000Wd NOISSid
B-23 Q l l l, ~.. s, s, ~%, '**s,
s

s,**
ce a* M 7 L
Q L.

, = =.
y w O ".'l b ~ 4" 1 CD ~ M -I, Ov ~- iC v O ~ O H t9 C ~ e i C t~ C G O
w N
6 i I .Ow YTs i , y -o i '77;1
n,, j a
i 2 i ' ' a [ ,D ,D C1 g .) G, g e o o 2 D_ G G ~ A .A Cs Q q. J s*, A J7 .y p. 7 ( E l.! 01C.0:u
4i. !: r i t ,i' ' mh ,D I n E ,0 l 5 a F tse L d l. ep l i f l e i. ,0 l t 4 **. h ? i l l t l n A l i i r P e t G 1 a ,0 l w 3 l l fo l,I ~ ll' ss ,' I l 8l A a n/ M
0 2
2 0 B I g I.i i 1 8,'l' ,0 F II-1I 1 i'i,1I 1'. 1 0 i 0 0 0 0 O 0 0 0 0 0 0 0 0 0 0 0 0 0 S 0 5 1 t 1 cf,tE " J- -',t s ,11 1 l 4 il1 i 4
/M Gl' AND fr!!LF 0.070 U. D 6 D.- j
/
r, J. O '.i 0 - . :sf i -I f I I l I I C 0 1 0 -' tu I 'I-(J l l 't,
d
\\ E \\. et 6. 0 J 0 I \\ .l m l \\ b o' \\ o 0 'a \\ l \\ l I \\
0. 010
^ ~ ' ~ - O UdU 'i i' T' '- rF r -. r i j i 0 10 20 30 90 50
5. 9 7e i II E 4tDURS) 1 Fig. B.21 Mole fraction of H in Compartment B.
2
t .:i j L mm ]0 ~ 7 ~O I N ^V \\ 6 e l 6 B tn p 0 e l 5 m t r F a ( m L o I C i ( s.. l0S n 4 H i J l 2 l l 0 3-l l e A fo l n l 0 t' o 1 5 3I i I tc k.N T ar f ,\\ t e A l 0 o N l 2 M \\ I 2 N 2 B
m 3
4 N 0 g 1 i f 4 5 ,4 _.Ji 0 0 0 0 0 0 5 0 5 2 1 1 0 O 0, 0 O gIF o%-
i
,1 !r j)- 1 <,:J j; l I t114
Al: I, l ' r,NI) tilli F
o. zoo,
f._. .3 J "' ,.r .f \\ a, / 'I i \\ 0.I50 / \\ ~ l .\\, 's l .4 a 0.100 [ v) / cn [ r'o ~ t C.050 / a/l ~ l l l } 0.000 0 10 20 30 40 50 60 le l !! l. 8ll .lR 3 ) / Fig. B.23 Mole fraction of CO in Compartment B. 2
I = v
o=
r ,e r
i; 1
f / >[. .I'(?' ,6 'j,if. B 5 t n r e m ,, '0 tr 5 a p m o F C ~ n L i U i f m i0 s m i 4 l a e e t D l s l Ni o i f ^i s n d i ^ ,0 l i I o l f ~ 3i I tc ra f E ,e, e t l '( ,0 o 2 M /, 4 2 7, B ,0 g 1 i F -i0 0 0 o 0 0 0 0 0 0 0 6 5 4 3 2 1 0 9 9 c u u U 0 0 0 0 0 0 o J. h2 tt. I 4 i; !it 4 < i 4
A!: li R ANil G:iLF sno,200 , #~'^ 311D. OD0 J 2M1,000 5 x n
20q uu? -.
r-250 oua E T ~ s 2 100,un+ S D. n o.* o .,g ,- r r q 0 LO 10 30 +0 5e i6 7 ;3 I Il L O L UP 3 ) 1 Fig. B.25 Volumetric flow out of containment. i
... ~.. -._,. -. - ~. B-30 4 1 4 5 i-r T I 4 t I : t o A J J O _u1 j C Q i 1-C1 2 4' _J_
f. o e~n g
.e y gi o i o I i = C .e N ( I ~~ l m - O i a n -- s o l ~ O M t
o e
N. N d m .D) e L _ C p h l O sa i ii i.ii 4 6 . < e < ia ii - ii ii..i 8 l e gp g tp G t-a fa. x 6-O .53 rJ C4
J C
O C G .g s Q t* O 3 5 2 y US n' W L' V) y e f6 N N e M k l 'e .i l i I 1 ..._--____.....-._m,. .w.- , _. _., -. =.. -. - - -. _ _ ~,
e B-31 ~ SUPPLEMENTAL PLOTS FOR SEQUENCE T 0N 23 I
i T23GW - GRAND GULF 1' o i i i b O 2 I O e cc a o C, b i O N l <l u' l i 4i N 4 o 1 O I ). t. 60 10 IlME HR I l Fig. B.27 Total in-vessel H2 generated. 1 l i i
nIwW e 1 0 _ _ _ _ I i. 5 1 ,/ / / lI F d L e U t 0 t 0 a G 1 ren D e NA R g H 2 R H G E l 1 M a I t T o T W 03 8 2 2 T B 1 0 5 g i F i i F-
c,hVt l-0 o8e S 8.,
9$^ 08-o - _e Us_i
)l 1

i '
i i i T230W -- GRAND CUi5 ou) - \\ o /I ro N 2 N 6 e N x N. T'OR OF ACTIVE FUEL x
--____.______1__________________:
o to e w x w 0, ~ BOTTOM OF ACTIVE FUEL o ~40 60 ~11ME HH I Fig. B.29 Reactor vessel water level. a
l T230W - GRAND GULF l in g CORE UPPER PLENUM DOWNCOMER g y n -j'''l''''l''''l''''l''''l''''l''''l''''l''''l'''I oN x \\ N ~ u. i, m e g o UO I O n i o i r n. m e g ? ~ w o e m E 2 e E 4: W j a. 2 m o /- e a _h. e 1 l 6,,,,,,,,l.......,,1.........l.........I........,I,........I,,,,,,,,,1.........l.........l.,,,,,,,: ~ 1 o. o.20
o. 4o o. so o. 80 1
1.2 1.4 1.6 18 '20 I TIME HR x 10 ' Fig. B.30 Temperature of structure, 'F. 'l 1 l i
I t l T230W - GRAND GULF CORE UPPER PLENUM DOWNCOMER e 5 A V g giiiiiiiiiiiiiiiiiiiiiiiii iiiiiiii iiii iii iii ii i iii ii ii iii ii ii i i i iii i ii i ii i iiii i i i i iiii ii i iii i i i ri x eN z r %3 i F m a i d m 3 LO p o 3 l m q r m gn i I -- _ A i e m i z a i 5" ~ ~ 3 ~ m o j ) g
(
= = J = = = 3 l 7
,....... I.,,...... I........ I........ I,........ l......... l......... l......... l......... l.........

0 E O. O.20 0.10 0 60 0 80 1 1.2 1.4 1.6 18 20 1 TIME HR xlO ' m e l IL l Fig. B.31 Fission product decay heat on structure, Btu /hr.
i I i I T230W - GRAND GULF
o 1
..... i i. i i i i i, i i i O i ] x to cc a -a av O ~ vI i o,o w l H N L r4 J L i f i l- ~ 3 L ] o 1 e t t O. 50 100 ISO TIME flH I Fig. B.32 Total C0 generated. 1 1 I i l
l 1 1 T230W - GRAND GULF i 1 p e j g x i F T l
n
~ _a g Q-N 1 3E w i u m j t-j f 1 i O. 50 100 150 i 11ME HR I Fig. B.33 Mass of water in the pedestal. 1 i i 1 e
T230W - GRAND GULF i 7 e L' 1 i ' * -
i i i -
o. x ~ n ~. ev n T ~ Y~ ~ l ~ c.o w 4 1 4 L 1 0-50 100 150 1IME HH 1 Fig. B.34 Mole fraction of H in Compartment B. 2
i T230W - GRAND GULF 1 o , i i. n x L. t l i = ~ u m h 1 a.2 b O in M k. ..l r \\ i%._ f O. SO 100 150 IIME HH 1 I Fig. B.35 Hole fraction of 0 in Compartment B. 2
1 T230W - GRAND GULF 7 O O-1 4 ) x
D
~ i = t.3 ~ ~ u N i t v t: h t { e i G. ? s a ) E O O. { 50 100 i 150 i 11HE HR i l 1 Fig. B. 36 Mole fraction of CO 2 in Compartment B. i l l 1 i w
l l [m 0 " : : ~ : 5: 5 1 B tne m tra F p m L 0 o U i I 0 C G 1 n D i H m N H a A e t R E s G M f l o I n W o i 0 t 3 f ca 2 r T f I O e i S lo i M i 7 3 B i g i i F O s.O wO d gd oN.o 6 I )Ne
1 I T230W - GRAND GULF ^ v ,, i i.... t o X ) _~ .~. i 3 ~ O y N r-f ~ r N B en i j
w 4
1 h I l I d ~ t i 0-50 100 150 f lME tiH 3 Fig. B.38 Mass of UO in core region. 2
B-45 SUPPLEMENTAL PLOTS FOR SEQUENCE T C 23
B-46 o .. N.. -o .\\ N __ o -n i ? ~.o .n s. LL E g D -o O _D
c"
-n s.s ~ a: m a us o .,, N 3 w <I _c 5 e' CE C .i'i'i'i'i e e e o o e o c e e e e o e in ,r n os 13371 070D44
)l lll I ,. O
3_
i. 0 f I ,5 5 j 4 , D q4 5 F j3 L d U e t G f f q3 , D ar I D S e n R N e U g A D l 2 R 7 5 i G _, 2 t H l a E t 7 t o l T ,0 I 2 C 0 3 4 2 B T _. 5 1 g i F ,0 1 _,5 - _ ~ _ - - ~- 0 ~ 0 h 0 0 0 0 6 0 0 0 0 6 6 5 4 3 2 1 md a J,o~ g
T23C - GRAND GULF a 1 5 0 -, J 1 ,/ / .i o _ s ) - f' 5 TOP OF ACTIVE FUEL 2 ',30_- - - - ;- - - u w 1 E r en'10-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' ~ _ ' _ _ _ _ _ _ _. ' BOTTOM OF ACTIVn FUEL 5 3x w i e._ i O ,,,i,,,i,,,,,,,,,,,,,,,,,,,,,,,,
0. 00

0. 50 1.00 1.50 2.00 2.50 3.98 3.50 4.00 T IIE tHOURSI e
fig. B.41 Reactor vessel water level. 4 i s
T23C - GRAND GULF 1,e2o_. t. J 1,600- . -.. - - - -.. - - +.. - - - -. - + - - -. + -- i 1.400- - +--- c o u p /% 1,?gg- ~ p i Os 1.000- -+----i-. a 4- - F +. +- w lii: P E R 01.ENUH A e w f,....j....- .j. j . j...._ _.......; 8@0-.........[........ ~ g O g (,.}g_. ....il.....4... /[-. R i.. /. '..p. ......[! / oo u rp o rtc u. "E 400_ .. 3 J 3 a. ~ i 2 2 0 C. .-.t. .3..... .......4...... 0 4 ....i....i....i...,,,,,,i....i....i,..,,....,...,i 0 5 10 15 20 25 30 35 40 45 50 TIME (HOURS) Fig. B.42 Temperature of structure, "F.
E mEzgrOo3 F $o Isu Ze x3a>o .>3Zz 1 2 u. 4 ,( 9 a g 0 e n ..a. ,2 .J .s o 3 e e a O \\ ss_ s d_ d ~ .ll i M F f. N, g 5i j
L.
/ .B .~~ A.. [.! 4 3 1 i 0 F i s s p u h C i o C n 1 5 j r i 7 k T o 2 duc 2. 3 t T b C 0 2 I d I e E c ay 2 G ( i 1 h 5 e H u a O n R t U u o R u A 3 N c n S .o ) t h j D s 0i i; E r u R c G t U u 3, ,e L r 5 F B l u / 4. =
i h
0, p r 4, { 5 5, ~ j 0 Ow l1i jll
1 c&~ o i 0 / ,5 3, 5 4 ,-i r e +
D
, 4 r 5 F 3 j L d U e t G j. 0 a - 3 r I D S e R n N e U g O A r H 0 R f5 i C 2 r G l 7L a t M o 1 2 l T ,0 C 4 3 4 2 B T -;t f g i F 0 1 5 0 0 0 0 0 0 0 0 6 4 2 0 2 4 6 e 0 0 0 0 0 0 Ta d a;_8 W
T23C GRAND GULF .ss e. eea _ ~
%\\
see,see - -i 2se. seal + r
l
~~ ~ ). i.. + m, 2ee. eee-i Qg ise. seel t a, m k E + i 100 SS? 7, .1 50.000 + i 1 0 t -~-,--- -p-j rv e, r cr,- g - i,- r l r i rr r rr rrj 0 5 10 15 70 2 t. 30 35 4D 45 50 1It E tHOURSI J Fig. B.45 Mass of water in the pedestal.
1 l T23C GRAND GULF 0.0160, q .f I 0.0140y 0.012B} / i 0.0100 3 ] C Q 0.00902 4 1 ~ -1 0 0060 .l, T g l i q B. 004 0d i'.. r-q g J
0. 0020y I
' ); 3 0.0000 _t ii l_ ___! _r T c r r r y n, y 1 r r y-y 1 1 1 9,.. q 0 5 10 15 20 25 30 35 40 4s 5g IlifE (HOURS 1 1 Fig. B.46 Mole fraction of H in Compartment B. 2 3
T23C - GRAND GULF
e. 2ss..
J a /j-O. 153-- .i -j ~ su og c. ige _ i-4 i u., s N
u. ese l
.i.. ..I. j l l ..., 1 '1 o n e -g-- ,,, ' 'Trr, r j rm i. r rrtr crrj r r n-j e 0 5 10 15 20 25 3 E. 35 40 45 58 T 1iE tHOURS) 1 Fig. B.47 Mole fraction of 0 in Compartment B. 2 1
T23C GRAND GULF e.eese, 4 6.9040-l- i N 0 2030- , f cn o 0 l ' v. avaad m, 5 1 0.0010 3 4 .d ....,ye - . - l r i. m r ' T r' 1 ' r 1 q 0 S 22, 15 2D 25 3D 35 40 45 50 IltE iHOURS) / Fig. B.48 Mole fraction of CO in Compartment B. 2 O
123C C.R AN11 GilLF 1.00, J Jl 1 O. 8 (' - lli 1 -t O. G i-
i.
'-I s. u N i 0 40-f J di b J 0.20d J
1..
r 0.00 l i,..q o 5 to is 20 2S 30 ' r' 4'- 50 l i!t i tlicukS) I fig. B.49 Mole fraction of steam in Compartment B.
T23C - GR.4ND GULF 1 s.e n.see i J Nen .xxj b e 1 u. Y
1. $4 AM i
? -l if g I t- '\\ 't,3.am,ou - I. I's.) i,hMf, Y,l },.., = 5*** i
t. 3 l V
,f V,,.I (. , l,;,7;.8} 3 3 3 a l 8-g l ll!,f,IgfL ig y') I l ' I I h 8,,,l W l 1,***.* H al 0 6 g i 0 5 to 1 :. 20 2 '. 3o n. n. c, o I il I e il!.UPS) I J Fig. B.50 Volumetric flow out of containment.
T23C G RNJII Gil _F 4oo. c oa _, i 2 a... iJ.~ 3?* '3a! ~ l - 2se. cool 200,u 0: <S e c.o 150,000 J m t co I. 132. K 4 4 so,. a, J 4 O, 0 5 to 15 20 2 *. Jo 40 41 'o i si ! il l 3.:1 sl J fig. B.51 %ss of UO in c re region. 2
c A tomic industrial Fetum, Inc. 1101 Wisconsen Avenue Bethesda.MD 20814 489I Telephone- (301) 654 9260 TWX 7108249602 ATOMIC FOR DC I l John R. Siegel l Vece Pressdent March 13, 1987 l Mr. Ilarold Denton Office of Nuclear Regulatory Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
The nuc1 car industry has sponsored the Industry Degraded Core Rulemaking (IDCOR) Program to ensure that industry insights regarding severe reactor plant accidents were made available for use in the regulatory process. IDCOR reports are protected by a copyright held by the Atomic Industrial Forum, Inc. to ensure that the rights of program sponsors are protected. Members of your staff have noted recently that restrictions on duplicating copyrighted materials may inhibit widespread.use of IDCOR results by persons involved in the regulatory process. Since this result would be counter to the intent of the Program, it is obviously not in the interest of our sponsors to allow concerns regarding use of this material to continue. The Atomic Industrial Forum, Inc. hereby grants the U.S. Nuclear Regulatory Commission (NRC) authority to duplicate copyrighted IDCOR materials as necessary for use by NRC and NRC contractor personnel in carrying out their regulatory mission. This grant of authority, however, does not apply to the Modular Accident Analysis Program (MAAP) or any of its associated documentation. Please contact Roger Huston, of our staff, if you have any additional questions regarding this issue. Sincerely, dv ( JRS:hlw cc: Cordell Reed Anthony Buhl}}

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SUMMARY

8.0 CONCLUSION

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8.0 CONCLUSION

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Dear Mr. Denton:

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ML20205K563

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1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

8.0 CONCLUSION

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

8.0 CONCLUSION

SUMMARY

SUMMARY

Dear Mr. Denton:

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