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Technical Rept 23.1Z Zion Nuclear Generating Station - Integrated Containment Analysis. W/870313 Release Memo
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Issue date:	06/30/1985
From:	; COMMONWEALTH EDISON CO., INDUSTRY DEGRADED CORE RULEMAKING PROGRAM
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n IDCOR Program Report Technical Report 23.1Z Zion Nuclear Generating Station - Integrated l Containment Analysis junc 1985 I

l ho-7O1 ur:888i:8888;b

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The Industry Degraded Core Rulernaking Prograin, Sponsored By the Nur dear Industry

IDCOR Arizona Public Scrivce Company New wrk ikwer Aushority The Babcock & Wilcox Company Niagara Molawk ikwer Corpration Baltimore Gas and Electric Cc.npny Northeast Utilities Service Compny Buhtel1%wer Corpration Northern Indiana Public Service Company Black & Veatch, Consulting Engineers Northern States 1%wer Compny Boston Edison Company Paajc Gas and Electric Compny C F Braun & Co Iknnsylvania 1%wer & Light Company The Cincinnati Gas & Elutric Company Philadelphia Elutric Company l

The Clenland Elutric illuminating Compny Ibrtland Genaal Elutric Company Combustion Engineering, Inc. Public Service Company ofOklahoma Commonwealth Edison Company Public Service Elutric and Gas Compny Consolidated Edison Company ofNew wrk, Inc. Public Service Indiana Consumers 1%wer Company Puget Scund1%wer & Lighe Company DanielConstruaion Compny Rochester Gas and Electric Corpration The Detroit Edison Company S2rgent & Lundy Duke 1%wer Company South Carolina Elutric and Gas Company Duquesne Light Company Southern Cahfornia Edison Company Ebasco Services incorprated Southern Company Services, Inc.

Exxon Nuclear Company, Inc. Stone & Webster Engineering Corpration Florida 1%wer & Light Company Swedish State Ikwer Board Fluor 1%wer Services, Inc. Taiwan Ibwer Company General Electric Compny Tuhnical Research Centre ofFinland Gibbs & Hill, Inc. Tennessee Valley Authority Gilbert / Commonwealth Compnies Texas Utilities Generating Company GulfStates Utilities Company The Toledo Edison Company Houston Lighting & 1%wer Company Union Elutric Company filinois 1%wer Company United Engineers & Constructors Inc.

Japan Atomic IndustrialForum, Inc. Virginia Electric andIkwer Company Kansas Gas and Elatric Company IVashington Public 1%wer Supply System Long Island Lighting Company IVestinghouse Elutric Corpration Middle South Services, Inc. Wisconsin Elutric 1%wer Compny Nebraska Public 1%wer District %nkee Atomic Electric 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 purrose 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 contactingJohn R. Siegel, Special Licensing Projects Manager /IDCOR, Atomic Industrial Forum,7101 Wisconsin Avenue, Bethesda, MD 20814-4805, (301) 654-9260.

Atomic indu steixl Feeum, Inc.

Plo t Wisconsin Avenue Bethesda.Mo 20814 4091 Telephone. (3o1) 654 9260 TWX 7108249602 Atomic Fon 0C John R Siegel Vice Peesident March 13, 1987 Mr. Harold Denton Office of Nuclear Regulatory Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 1

Dear Mr. Denton:

1 The nuclear 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. Nuc1 car 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 qitestions regarding this issue.

Sincerely, J %7 JRS:hlw cc: Cordell Reed Anthony Buhl

IDCOR l

TechnicalReport 23.1Z Zion Nuclear Generating l Station - Integrated l

Containment Analysis J [

June 1985 l

l by:

l' Commonwealth Edison Company Chicap, Illinois l

l 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, not 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 legal liability or responsibili-ty for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infrins' Privately owned rights.

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 complainti of exclusion, denial of benefits, or other i discrimination of those bases under this program may be filed with (among others) the Tennessee Valley Authority (TVA). Office of EEO,400 Commerce Avenue EPBl4, Knoxville, TN 37902, and must be not later than 90 4,ysfrom the Jare of the e/Irgel Jistrimination. 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.

I Copyright C 1985 by Atomic Industrial Forum,Inc.

- i.-

TABLE OF' CONTENTS

,Page '.o.

1-1

1.0 INTRODUCTION

1-1 1.1 Problem Staterent . . . . . . . . . . . . . . . . . . . . .

1 -1 1.2 Relationship to Other Tasks . . . . . . . . . . . . . . . .

i 1-2 1.3 References ........................

2-1 2.0 STRATEGY AND METHODOLOGY . . . . . . . . . . . . . . . . . . . .

2-2 2.1 References ........................

. . . . . . . . . .- 3-1

3.0 DESCRIPTION

OF MODELS AND MAJOR ASSUMPTIONS 3-1 3.1 Plant Specific Information ................

3-3 3.1.1 Primary System . . . . . . . . . . . . . . . . . . .

3-3 3.1.2 Containment ....................

3.2 Modular Accident Analysis Program (MAAP)- ......... 3-4 3-5 3.2.1 MAAP Nodalization .................

3.2.2 Safety Systems Modeled in MAAP . . . . . . . . . . 3-5

......... 3-8 3.2.3 Fission Product Release from Fuel 3.2.4 Fission Product Release and Aerosol Generation 3-11

Resulting from Core-Concrete Attack ........

3.2.5 Description of the Natural Circulation Model . . . . 3-12 3.2.6 Fission Product Deposition . . . . . . . . . . . . . 3-14 3-14

3.3 References ...................-.....

............. 4-1 4.0 PLANT RESPONSE TO CORE MELT ACCIDENTS 4-1 4.1 Plant Response for TMLB/ Seal LOCA . . . . . . . . . . . . .

Reactor Coolant System Response . . . . . . . . _ . . 4-1 4.1.1 4-7 4.1.2 Containment Response . . . . . . . . . . . . . . . .

- ii -

TABLE OF CONTENTS (Continued)

Page No.

4.2 Plant Respense for TMLE , . . . . . . . . . . . . . . . . . 4-11 4.2.1 Reactor Coolant System Response .......... 4-11 4.2.2 Containment Response . . . . . . . . . . . . . . . . 4-13 4.3 Plant Response for SLFC . . . .-. . . . . . . . . . . . . . 4-17 4.3.1 Reactor Coolant System Response .......... 4-20 4.3.2 Containment Response . . . . . . . . . . . . . . . . 4-25 4.4 Plant Response for ALFC . . . . . . . . . . . . . . . ... . 4-28 4.4.1 Reactor Coolant System Response ........... 4-28 4.4.2 Containment Response . . . . . . . . .-. . . . . . .. 4-28 5.0 REC 0VERY ACTIONS . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 TMLB with Recovery at 2.5 Hours.. . . . . . . . . . . . . . 5-1 5.2 TMLB/ Seal LOCA with Recovery at 1 Hour .......... 5-3 5.3 TMLB/ Seal LOCA with Recovery at 2.5 Hours . . . . . . . . . 5-10 5.4 TMLB/ Seal LOCA Recovery at 6 Hours ............ 5-14 5.5 TMLB/ Seal LOCA with Recovery at 15 Hours ......... 5-22 5.6 SLFC with Recovery at 10 Hours .............. 5-22 5.7 ALFC with 1 Operational Charging Pump . . . . . . . . . . .

5-27 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION ....... 6-1 6.1 Introduction ....................... 6-1 6.2 Modeling Approach . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Containment Failure Modes Summary . . . . . . . . . . . . . 6-3 6.3.1 Failure Modes Identified in the Study Conducted -

for Task 10 .............,...... 6-3 6.3.2 Study to Identify Containment Failure Modes .... 6-4 6.3.3 Other Potential Failure Modes ........... 6-13 6 -

_ gyy _ . -

l i

I TABLE OF CONTEWTS (Continued) l l

4 Page '.o. l 6.4 Sequences Evaluated . . . . . . . . . . . . . . . . . . . .

6-13 TMLB' With a Seal LOCA . . . . . . . . . . . . ... . 6-14 6.4.1 6.4.2 TMLB' Without a Seal LOCA . . . . . . . . . . . . . 6-18

..........,............ 6-18 6.5 References . .

7-1 7.0

SUMMARY

OF RESULTS . . . . . . . . . . . . . . . . . . . . . . .

7.1 Base Case Analysis .................... 7-1 7.2 Operator Actions Analysis . . . . . . . . . . . . . . . . . 7-1

...................... 8-1

8.0 CONCLUSION

S . . . .

APPENDICES A-1 A.1 Zion Parameter File . . . . . . . . . . . . . . ... . . . .

A.2 MAAP Sequence Control Cards . . . . . . . . . . . . .,. . . A-15 B-1 B.1 TMLB' Plots . . . . . . . . . . . . . . . . . . . . . . . .

..................... B-7 B.2 TMLB'/LOCA Plots

...................... B-25 B.3 SLFC Plots . . .

B-33 3.4 ALFC Plots . . ......................

i w-- ,w.-

..-- y- . _- , - - - .

_ iv _

LIST OF FIGURES i

Page No.

Figure No.

Structure of the analytical approach . . . . . . . . . 2-2 2.1 Schematic of MAAP compartmentalization . . . . . . . . 3-6 3.1 MAAP-PWR primary system nodalization . . . . . . . . . 3-7 3.2 MAAP sys' tem description for Zion . . . . . . . . . . . 3-9 3.3 4.1 Primary system pressure - Zion TMLB with a seal LOCA . . . . . . . . . . . .-. . . . . . . . . . . . .

4-3 4-4 4.2 Core water level - Zion TMLB with a seal LOCA ....

4 4.3 In-core hydrogen production - Zion TMLB with a seal LOCA . . . . . . . . . . . . . . . . . . . . . . . . .

4-5 ,

Corium mass - Zion TMLB with a seal LOCA . . . . . . . 4-6 4.4 .

Containment building pressure ............ '4 3 4.5

! 4.6 Reactor cavity water mass - Zion TMLB'with a seal LOCA . . . . . . . . . . . . . . . . . . . . . . ... .

4-9' 4.7 Water level in the lower compartment - Zion TMLB with a seal LOCA ..................... 4 TMLB' primary system pressure ............. 4-12 4.8 TMLB ' core water l evel . . . . . . . . . . . . . . . . 4-14 l 4.9 TMLB' corium in core region ............. 4-15 4.10 Containment building pressure ............ 4 ! 4.11 Reactor cavity water mass - Zion TMLB ... . . . . ._ . 4-18 4 4.12 Water level in the lower compartment - Zion TMLB . . 4-19 4.13 i" 4-21 4.14 SLFC primary' system pressure . . ... . . . . . . . . . i SLFC core water level ................ 22 4.15

............... 4-23 4.16 SLFC core H2 pr duction SLFC corium in core region . . . . . . . . . ... . . . 4-24 4.17 SLFC containment pressure .............. 4-26

. 4.18 SLFC cavity water level ............... 4-27 4.19

l _ _

LIST OF FIGURES (continued)

Figure No. Page No.

4.20 ALFC primary system pressure . . . . . . . . . . . . . 4-29 4.21 ALFC core water level ................ 4-30 4.22 ............... 4-31 ALFC core H2 pr duction 4.23 ALFC corium mass in primary system . . . . . ..... 4-32

t. 24 ALFC containment pressure .............. 4-33 4.25 ALFC containment pressures . . . . . . . . . . . . . . . 4-34 5.1 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - primary system water levels .. 5-2 5.2 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - primary system pressure. . . . . 5-4 5.3 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - steam generator level ..... 5-5 5.4 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - H2 generation ......... 5-6 J 5.5 TMLB' (seal LOCA) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recovery - steam generator level ......................... 5-7 5.6 TMLB' (seal LOCA) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recovery - primary system water l evel s . . . . . . . . . . . . . . . . . . . . . . 5-8 5.7 TMLB' (seal LOCA) I hour recovery - primary system pressure . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.8 TMLB' (seal LOCA) 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months recovery - primary system pressure . . . . . . . . . . . . . . . . . . . . . . . . 5-11 5.9 TMLB' (seal LOCA) 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months recovery - primary system wate r l evel s . . . . . . . . . . . . . . . . . . . . . . 5-12 5.10 TMLB' (seal LOCA) 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months recovery - H2 generation .. 5-13 5.11 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - containment pressure . . . . . . . . . . . . . . . . . . . . . . . . 5-15 5.12 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - lower compartment water level ...................... 5-16 5.13 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - cavity water level ......................... 5-17 5.14 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - containment flame temperatures . . . . . . . . . . . . . . . . . . . . . . 5-18

- vi -

LIST OF FIGURES (centinued)

Figure No. Dage *;c.

5.15 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - steam generator pressure . . . . . . . . . . . . . . . . . . . . . . . . 5-19 I

l 5.16 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - primary system l gas temperatures . . . . . . . . . . . . . . . . . . . . 5-20 5.17 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - primary system volumetric flows . . . . . . . . . . . . . . . . . . . . 5-21 5.18 TMLB' (seal LOCA) 15 hour1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months recovery - containment pressure . . . . . . . . . . . . . . . . . . . . . . . . 5-23 5.19 TMLB' (seal LOCA) 15 hour1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months recovery - containment average corium temperatures .............. 5-24 5.20 TMLB' (seal LOCA) 15 hour1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months recovery - containment flame temperatures . . . . . . . . . . . . . . . . . . . 5-25 5.21 SLFC 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months recovery - core water level . . . . . . . . 5-26 5.22 SLFC 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months recovery - H2 generation ......... 5-28 5.23 SLFC 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months recovery - core mass ........... 5-29 5.24 SLFC 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months recovery - accumulator pressure . . . . . . 5-30 5.25 ALFC RWST refill - core water level .......... 5-31 5.26 ALFC RWST refill - containment pressure ........ 5-32 6.1 Radial displacement of cylinder vs. pressure ratio . . 6-5 6.2 Penetration-pipe assembly .............. 6-7 6.3 Equipment hatch flange rotation ........... 6-8 6.4 Equipment hatch section ............... 6-9 6.5 Containment penetrations elevation . . . . . . . . . . 6-11 6.6 Containment penetrations Section A-A . . . . . . . . . 6-12

LIST OF TABLES Table No. Page No.

3.1 Initial Inventories of Fission Products and Structural Materials Released as Aerosols . . . . . . 3-10 6.1 TMLB' With a Seal LOCA . . . . . . . . . . . . . . . . 6-15 6.2 Distribution of Cs, I for TMLB' With Seal LOCA (Fraction of Core Inventory) . . . . . . . . . . . . . 6-16 7.1 Summary of Base Case Sequences . . . . . . . . . . . . 7-2 7.2 TMLB' With a Seal LOCA . . . . . . . . . . . . . . . . 7-3 l

I

l-1

1.0 INTRODUCTION

1.1 Problem Statement The main objective of this investigation is to calculate the response of the primary system and containment to a selected group of postulated severe accident sequences, i.e., those which have been identified as potentially leading to core degradation and melting. These will be addressed on a best estimate basis with the assessments of uncertainties associated with state-of-the-art modeling being reported in Ref. [1.1]. Also, the study will include assessments of the results of operator intervention in various sequences.

The results of the primary systen, and containment analysis are incorporated into an assessment of the fission product release and deposition within the various regions of the primary system and containment building.

For those sequences in which containment integrity is violated, the release of fission products to the surrounding environment is estimated.

1.2 Relationship to Other Tasi The containment analyses of 10COR Subtask 23 are dependent upon the primary system and containment response models developed in Subtask 16.2 and 16.3 [1.2]. The fission product release, transport, and retention models developed in IDCOR Task 11 [1.3] have not been employed for subtask 23 analyses.

Instead, a series of updated, integrated models, which are based upon large scale experiments, have been incorporated into the Modular Accident Analysis Program (MAAP) [1.2]. The accident sequences used for the analyses along with the operator interventions were developed by considering the accident sequences presented in Subtask 3.2 [1.4] and the results of the Zion Probabilistic Safety Study (ZPSS) [1.5].

The ultimate structural capability of the reference plant contain-ments was assessed in IDCOR Subtask 10.1 [1.6]. This report also includes an evaluation of potential failure modes for the Zion containment building.

These analyses define the containment failure pressure and failure mode employed in the analyses for Task 23.1.

l-2 Calculations of the rate and amount of fission croducts released from the containment, for those sequences which result in containment failure,

. were supplied to IDCOR Subtask 18.1 [1.7] to formulate assessments of tne health consequences associated with these postulated accident scenarics.

These health consequence analyses were then supplied to IDCOR Subtask 21.1

[1.8] to evaluate the risk reduction potential of possible mitigating devices considered for this containment design.

Also, operator intervention sequences were developed and applied to specific accident sequences for the Zion station to determine those potential actions which could terminate the accident sequence and result in a safe stable state. This was considered as part of IDCOR Subtask 22.1 [1.9], Safe Stable States, which discusses both the inherent and intervention means of terminating the various core damage sequences considered for the Zion station.

Finally, it should be noted that the analyses developed as part of IDCOR Subtask 16.2 and 16.3 involve the detailed consideration of many differ-ent phenomena which are themselves considered in separate IDCOR subtasks.

These inc'ude: hydrogen generation, distribution and combustion (12.1, 12.2, and 12.3); steam generation (14.1); core heatup (15.1); debris behavior (15.2); ar.d core-concrete interactions (15.3) as discussed in Ref. [1.2].

Detailed discussions of these topics can be found in the final reoorts submit-ted for each task. Individual issues will only be addressed as required to understand the specific behaviors obtained for the accident sequences con-sidered and the specific design characteristics of the Zion system.

1.3 References 1.1 " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," Technical Report on IDCOR Subtask 23.4, to be published.

1.2 "MAAP, Modular Accident Analysis Program," Technical Report on IDCOR Subtasks 16.2 and 16.3, June 1983.

1.3 " Fission Product Transport in Degraded Core Accidents," Technical Report on IDCOR Subtask 11.3, December 1983.

1.4 " Assess Dominant Sequences," Technical Report on IDCOR Subtask 3.2, October 1983.

1-3 1.S " Zion Probablistic Safety Study, ZPSS," Commonwealth Edisor, Co .pany, September 1981.

1.6 " Containment Structural Capability of Light Water Nuclear Power Plants," Technical Report IDCOR Subtask 10.1, July 1983.

1.7 Technical Report IDCOR Subtask 18.1, to be published.

1.8 Technical Report IDCOR Subtask 21.1, to be published.

1.9 "The Attainment of a Safe Stable State for Damage Core," Technical Report IDCOR Subtask 22.1, to be published.

2.0 STRATEGY AND METHCDCLOGY The basic strategy of this subtask was to analyze accident sequences which have been previously identified as potential significant contributors to core melt frequency or risk. These analyses consisted of models for the plant thermal hydraulic response and, if the progression of the accident sequence led to core degradation and melting, for fission product transport and deposi-tion. These analyses include the performance of the ECCS systems and the containment engineered safety systems, such as the fan coolers, containment spray systems, decay heat removal system, etc.

The principal tool used to perform the containment thermal-hydraulic response analyses is the MAAP [2.1] code. 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, as adapted from Ref. [2.2] was used 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 sedimenta-tion.

With the defined accident sequences, analyses were carried out for the best estimate path of the accident progression including the fission product transport before and after reactor vessel failure and also after containment failure. Flows between the primary system and containment and natural circulation flows within the primary system are included in this analysis. The primary system response following vessel failure including the heatup of the reactor vessel and its structures, is evaluated through the natural circulation models for both primary system and containment. Fission product transport of both vapors and aerosols is determined by these density driven flows. Included in this evaluation is the containment pressurization which would be imposed upon the primary system, and would determine the magnitude and direction of flows between the primary system and containment.

2-2 In addition to the containment analyses discussed in tn's recort, other cases are considered as part of the uncertainty and sensitivity analyses.

One such uncertainty is the vapor pressure of cesium and iodine corpcunds.

The results of analyses in which the vapor pressure was varied are recortec in Ref. [2.4]. Evaluations also have been performed for containrent bypass and failure to isolate scenarios. These are reported under IDCOR Task 23.5 (Ref.

[2.5]).

For each of the four Zion accident scenarios selected for analysis, thermal-hydraulic calculations were performed both with and without selected operator actions during the accident. The " base case" analyses, which assume no, or only minimal, operator response during the accident, establish a reference system response during each of the accident scenarios. The " operator action" analyses are branch calculations of the base 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 opera-ter to take such cr. tion. Additional uncertainty and sensitivity analyses have been performed on several key parameters associated with the accident response.

These are reported in Ref. [2.4].

2.1 References 2.1 "MAAP, Modular Accident Analysis Program User's Manual," Technical Report on IDCOR Tasks 16.2 and 16.3, May 1983.

2.2 "Anslysis of In-Vessel Core Melt Progression," Technical Repcrt on IDCOR Subtask 15.18, 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.

2.5 IDCOR Technical Report on Task 23.5, " Evaluations of Containment Bypass and Failure to Isolate for the IDCOR Reference Plants," to be published.

3-1

3.0 DESCRIPTION

OF MODELS AND MAJCR ASSUMPTIONS This section of the report describes the plant model and rajor assumptions used in the IDCOR Task 23.1 analysis of the Zion station using tne Modular Accident Analysis Program (MAAP).

3.1 Plant Specific Information The two-unit Zion nuclear station is owned and operated by the Commonwealth Edison Company was conceived in the late 1960's. The units are four loop Westinghouse pressurized water reactors. Construction permits for these units were granted by the U.S. Atomic Energy Commission in 1968.

The Zion commercial service dates were:

e Zion Unit 1,12/31/72 e Zion Unit 2, 9/19/74 The site for the Zion Station is located on the western shore of Lake Michigan about 40 miles north of Chicago, Illinois, and 42 miles south of Milwaukee, Wisconsin. The centers of the cities of Waukegan, Illinois, and Kenosha, Wisconsin, lie about 6 miles south and 8 miles north, respectively, of the site. The 250-acre site is situated on a predominantly sandy area.

The site is well ventilated and generally not subject to severe inversions.

The site is not particularly susceptible to tornadoes or other severe mete-orological phenrmena. The site and the region within 100 miles of the site are considered an area of minor seismic activity. The seismic design basis for the plant is a safe shutdown earthquake (SSE) with a 0.179 horizontal ground acceleration and a 0.llg vertical acceleration acting simultaneously.

The nuclear steam supply system for each of the Zion units consists of four loop Westinghouse pressurized water reactors. Reactor coolant loop stop valves permit the isolation of individual loops from the reactor during periods of reactor shutdown. Each reactor is rated at 3,250 megawatts (thermal). The emergency core cooling systems (ECCS) are totally independent

3-2 for each of the two units and consist of redundant high pressure injection trains, redundant low pressure injection trains, and passive accumulatcrs for each unit. Hot leg as well as cold leg injection capability exists. All ESFs are located in the auxiliary building. The ECCS takes sucticn from the containment sump through special pipes to the inlet of the low pressure injection pumps during recirculation.

Each reactor system is housed in an individual containment building.

These structures consist of post-tensioned concrete shells over 1/4-inch thick steel liners. Each containment is served by both fan cooler and containment spray systems. These systems provide redundant and diverse containment heat removal capability. Five fan cooler units are provided fcr each containment with each being rated at one-third required capacity for design base accident conditions. Three completely redundant containment spray system trains are provided for each ur.it, each rated at 100 percent capacity for accident conditions. One train has a diesel engine driven spray pump for added diver-sity. The liner plate welds in each containment are covered by seal welded channels with air pressure in selected sections maintained above the contain-ment peak accident pressure. Containment penetrations are similarly pressur-ized. These features, along with an isolation valve seal water system, provide added assurance of containment leak-tight intt grity.

The plant auxiliary cooling systems consist of a shared component cooling water system (a closed system) and a shared service water system.

Extensive redundancy, separation capability, and extra capacity exist for these systems. The auxiliary feedwater system, serving the secondary side of the steam generators, is separate for each unit. Each unit has three pumping trains, each capable of feeding all four steam generators. Two trains are fed by separate, 100 percent capacity, motor-driven pumps while the third train is fed by a 200 percent capacity steam turbine-driven pump. Electrical power is supplied through multiple off-site power sources, with backup diesel generators available for safety related loads in the event off-site power is lost. Batteries are available for supplying DC power in event of such loss.

The diesel generator arrangen. ant consists of five machines, two per unit, with the fifth being a swing diesel capable of tying into a third bus on either

unit as demand arises. Safeguards actuation systems consist generally of standard Westinghouse logic networks with sequential diesel generator loacing.

The balance of plant equipment is not unique from a safety stanc-point. The turbine-generators are Westinghouse tandem compound,1,800 rpm units rated at 1,085 Mwe gross output. Six stages of feedwater heating are provided. Each unit uses a single pass, deaerating type condenser. Once-through cooling is provided using Lake Michigan as a source of cooling water.

The fuel handling, fuel storage, waste management, and miscellaneous auxiliary and support systems are of secondary interest from a safety standpoint in this evaluation.

3.1.1 Primary System The majority of the primary system data comes from the Zion FSAR

[3.1] and the Zion Probabilistic Safety Study [3.2]. This information in-cludes initial condition pressures, temperatures, flow rates, enthalpies and masses. Additional information describing system pressure set points, relief valve flow rates, control logic and emergency core cooling system (ECCS) parameters, are also drawn from these two sources. Geometrical input data such as flow areas and relative component elevations are taken from Westinghouse and architect-engineer (AE) drawings. Much of the critical information was field checked during the performance of the Zion Probabilistic Safety Study [3.2]. Volumes representing components such as the cold leg, pump " bowl" and steam generator plena are based on Westinghouse information.

Pump curves for the ECCS pumps are based on vendor information. Reactor coolant pump coastdown data is taken from Zion FSAR [3.1] analyses.

For the core heatup calculations, the core is represented by 70 nodes; the axial and radial power peaking factors are computed for a represen-tative core consisting of the 10 axial regions and 7 radial regions by collapsing a more detailed core power distribution map.

3.1. 2 Containment The containment model consists of nodes or regions for the upper compartment, the lower (steam generator and coolant loop) compartment, the

j 3-4 annular compartment (located outside the crane wall and below the operating deck), and the reactor cavity. Input data, including initial conditions, free volumes, heat sinks, heat transfer coefficients, exposed surface areas and flow areas between volumes are based on information in the Zion FSAR [3.1], AE drawings, and the Zion Probabilistic Safety Study [3.2]. The " sump" (or water

, collection) areas modeled in the containment include the lower conpartment and annular compartment floors, the recirculation sump and the reactor cavity.

The keyway leading from the lower compartment floor to the reactor cavity is surrounded by a 6" high curb causing water to flow into the recirculation sump. If the water level rises above this level, excess water will flow into the reactor cavity.

Containment heat removal and pressure suppression is accomplished by the passive heat sinks, the operation of the containment spray system and/or the reactor containment fan cooler units (RCFCs). Only containment spray injection is considered for active fission product removal. Reactor contain-j ment fan coolers (RCFC), and containment spray system data as well as the containment safeguards actuation and containment isolation setpoints were supplied by the Westinghouse Electric Corporation.

Hydrogen is considered to be well mixed in individual primary system and containment regions due to the intra-compartmental flows resulting from natural circulation. Concentration differences between regions are reduced by intercompartmental gas flows. Steam and non-condensable gas flows are based on pressure and density differences between the regions.

The containment ultimate failure pressure (149 psia) was determined in the Zion Probabilistic Safety Study [3.2] and reviewed in IDCOR Task 10

[3.3]. Also, potential containment failure modes are considered in Section 6.5 of this report.

3.2 Modular Accident Analysis Program (MAAP)

Within the IDCOR Program, the phenomenological models developed in Tasks 11,12,14 and 15 have been inccrporated into an integrated analysis

3-5 code (MAAP) [3.4] to analyze the major degraded core accident scenarios fer both pressurized water reactors (PWRs) and boiling water reacters (SWPs).

MAAP is designed to provide realistic assessments for severe core carage accident sequences, including fission product release, transpcrt anc ceposi-tion, using first principle models for the major phenomena that govern the accident progression. The following sections describe the primary system nodalization and containment nodalization, the safety systems modeled in the MAAP-PWR code as applied to the Zion large dry containment design, the fission product release model and the fission product deposition models. A complete Zion parameter file is given in Appendix A.l.

3.2.1 MAAP Nodalization The MAAP plant model for a large dry PWR is divided into several nodes as shown in Fig. 3.1, Nodesexistfortheuppercompartment(compart-ment A), lower compartment (compartment B), annular compartment (compartment D), reactor cavity (compartment C), quench tank. and primary system. This nodalization provides detailed tracking of containment gas temperature, wall temperatures, and steam / hydrogen concentrations as shown in Fig. 3.1.

The primary system is divided into ten nodes as shown in Fig. 3.2.

Nodes exist for the core region, upper plenum, downcomer, broken loop cold leg, broken loop hot leg, unbroken loop cold leg, unbroken loop hot leg, pressurizer, and both the broken or unbroken loop steam generator secondary sides. This primary system nodalization permits a detailed accounting of the water which is available for cooling the core and for reacting with the zircaloy fuel cladding. In addition, this scheme allows the user to track hydrogen and fission product concentrations through the primary system and thereby calculate release rates to the containment. The core is further divided into a user selected number of subnodes; a 7 radial x 10 axial nodaliza-tion is used for the Zion analysis.

3.2.2 Safety Systems Modeled in MAAP The safety systems considered in the Zion reference analysis include the charging pumps, safety injection pumps, low pressure injection (or RHR)

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3-8 pumps, passive safety injection accumulators, auxiliary feedwater, centainrent sprays, and containment fan coolers. These are shown in Fig. 3.3 along witn other systems important to accident progression such as the pressurizer and steam generator safety and power operated relief valves. All of these systers can be enabled or disabled by the use of " event codes" in MAAP at the ciscretion of the user. The MAAP User's Manual [3.4] gives a complete description of the use of MAAP and also compares the physical models with pertinent experiments.

Unless otherwise stated, it was assumed that if a particular safe-guards system was available (e.g., LPI), only the minimum r, mber of such units were operational (e.g., only 1 LPI pump). Credit was taken for only one train of the ECCS, the containment spray injection system, auxiliary feedwater system, and three reactor containment fan coolers.

3.2.3 Fission Product Release from Fuel The FPRAT module in MAAP, as adapted from Ref. [3.5] was used to calculate the release rates of fission products from the fuel matrix. These rates are dependent upon the fuel temperature history during heatup and upon characteristics of the atmosphere within the vessel which e ~fect saturation of the chemical species as discussed in IDCOR Task 11.1 [3.6]. Fuel temperature histories for the seventy regions in the core were tracked to determine the release characteristics for the fission products and inert materials. The initial inventories of the various fission products were obtained from Ref.

[3.7]andaregiveninTable3.1. .

The FPRAT calculation considers evaporation and condensation characteristics of various chemical species. Several key assumptions, consistent with the recommendations of IDCOR Task 11.1 were made regarding the physical form of release fission products. These are:

1. Cesium and iodine combine to form Cs! upon entry to the fission product release pathway. The excess cesium forms Cs0H. Both chemical species exhibit similar physical behavior, hence the source rate for the Cs, I fission product group is assumed to be the sum of the Cs and I release rates. The form of this source is assumed to be vapor.

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Kr 16.5 Xe 320 Cs 161 I 14.8 Te 30.8 Sr 59.1 Ru 128 La 76.8 Mo 1 91 Sn 347 Mn 196 Ag 2232 In 419 Cd 140

3-11

2. Tellurium is assumed to enter the release pathway as vaporized Te02 '

3. Inert aerosol generation rate is the combined release rates for volatile structure materials (Cd, In, Ag, Sn, and Mn).

4. Cesium, iodine, and tellurium are completely released during fuel heatup.

5. Strontium and ruthenium are assumed to represent their respec-tive nonvolatile fission product groups as defined in WASH-1400.

Both were assumed to enter the release pathway in aerosol form.

The melt release for strontium and ruthenium was assumed to cease as each node melted and dropped into the cavity. Releases in the cavity are calculated separately (see next section).

3.2.4 Fission Product Release and Aerosol Generation Rasulting from Core-Concrete Attack The release of aerosols due to core-concrete attack was determined using a model based on the concrete ablation rates from MAAP. The mass of low volatility fission products and inert aerosols released from core debris is based upon a vapor stripping model assuming the melt constituents follow Raoult's law. This c.alculation is dependent upon the amount of gas sparging through the core debris, the molar concentration of fission products in the core debris, the vapor pressure of the chemical species of interest, and the temperature of the core debris.

The key assumptions are:

1. The masses of CO and water vapor released per cubic meter 2

ablated for the limestone concrete used at Zion are 577 kg and 130 kg respectively.

2. Stripping only occurs when the corium is molten.

3. The gases rel' eased by' the ' downward attack pass thrcugh the molten pool and cause stripping. Gases generated by sidewall !

attack are assumed.to bypass the pool.

I- 4 .The predominant form of Sr is Sr0, of Ru is elecental Ru,, and

f. of La is La23 0'

! 5. Inert aerosols of Ca0 may be generated during core-concrete

! attack. This chemical form is used as 'a surrogate for the 1

various concrete melt constituents that could be added to the corium pool.

3.2.5 Description of the Natural Circulation'Model MAAP models the primary system thermal-hydraulics, prior- to and

' after vessel failure, including .the, effects of volatile fission product release. If large amounts of volatile fission . products are retained in the I primary system after vessel failure, which is generally the case, the feedback mechanisms between fission product behavior and the thermal-hydraulics must be e

j modeled.

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The natural circulation model calculates the primary system fission- ,

f product transport and thermal-hydraulics after reactor vessel failure, and includes models for the following phenomena:

a. Natural circulation- flows due to temperature and concentration ,

(cesiumiodide)differencesaroundtheprimarysystem.

> b. Heat transfer between gas and structures in the primary system.

i i c. Heat transfer between the primary system and the steam genera-tor shells,

d. Heat transfer to containment through reflective insulation.

This treatment includes degradation of insulation performance L

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e. Fission product transport due to re-volatilization and subse-4 quent condensation and sedimentation in cooler nodes.

The chemical state of the fission products represents an uncertainty in the calculations. IDCOR Subtask 11.1 [3.6] identified the dominant chemical species for cesium and iodine to be cesium iodide and cesium hydroxide.

Recent experiments [3.9] show this may characterize much of the material, but significant quantities have also been observed to be irreversibly plated-out on steel surfaces above the fuel region. For these analyses, the cesium iodine and cesium hydroxide fission products are assumed to have a vapor pressure characteristic of cesium hydroxide. Because of this assumption, and the molar dilution of Cs! by Cs0H, releases of Cs and I to the environment quoted in this report should be regarded as consisting mainly of Cs0H. If surface reactions between fission products and steel surfaces occurred, the vapor pressure could be greatly reduced. Experience with MAAP indicates that even large vapor pressure reductions delay but do not ultimately effect the mobility of cesium and iodine in the primary system. Thus, the basic behavior is the same as for the analyses reported here, but the releases to the environ-i ment are lower since less material is gaseous at the time of containment failure. If, on the other hand, vapor pressure reductions were so large as to prevent the movement of cesium and iodine before the melting temperature of steel was reached, the steel on which the fission products were deposited would melt, and the fission products would drop to the reactor cavity and be added to the melt there. This would also be expected to result in a net reduction in the ultimate release to the environment since less material would

, be in a gaseous state at the time of containment failure. Other analyses reported in Ref. [3.10] investigate the sensitivity of the release fractions to the details of the representation of the Cs and I compounds and include the use of recent experimental data for the vapor pressure of Cs0H. This recent l

data indicates that the Cs0H vapor pressure used in the calculations reported here is conservatively high.

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3.2~.6 Fission Product Deposition 10COR Task 11.3 evaluated models for aerosol agglomeration and deposition processes of fissicn products [3.8]. These removal processes reduce the magnitude of radionuclide release to the environnent. The correspond-ing MAAP models depict physical mechanisms for vapor condensation on structures and aerosol retention due to steam condensation and gravitational settling.

The agglomeration and sedimentation are represented as a removal rate that can be correlated as a function of the aerosol cloud density [3.11]. This formula-tion is consistent with the available large scale experimental results. Vapor retention is governed by vapor condensation / evaporation on aerosol surfaces and walls. Mechanisms considered for aerosol retention are sTaam condensation and sedimentation. The MAAP nodalization scheme for fission poduct transport is identical to that used for the thermal-hydraulic models in MAAP.

3.3 References 3.1 Zion Final Safety Analysis Report 3.2 " Zion Probablistic Safety Study, ZPSS," Commonwealth Edison Company, September 1981.

3.3 " Containment Structural Capability of Light Water Nuclear Power Plants," Technical Report 10COR Subtask 10.1, July 1983.

3.4 "MAAP, Modular Accident Analysis Program Users' Manual," Technical

. Report on IDCOR Tasks 16.2 and 16.3, May 1983.

3.5 " Analysis of In-Vessel Core Melt Progression," Technical Report on IDCOR Subtask 15.1B September 1983.

3.6 EPRI/NSAC, " Technical Report 11.1,11.4 and 11.5, Estimation of Fission Product and Core-Material Source Characteristics," October 1982.

3.7 J. A. Gieseke, et al., "Radionuclide Release Under Specific LWR Accident Conditions, Volume 1 - PWR Large Dry Containment," Draft report BMI-2104, July 1983.

3.8 IDCOR Technical Report on Task 11.3, " Fission Product Tr'ansport in Degraded Core Accidents," December 1983.

3.9 Richard K. McCardell, " Severe Fuel Damage Test 1-1 Quick Look Report " EG&G Idaho, October 1983.

3-15 j

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3.10 IDCOR Technical Report on Task 23.4, " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," to be published.

i 3.11 IDCOR Technical Report, " Fission Product Deposition Models in PAAP,"

to be published.

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.4-1 4.0 PLANT RESPONSE TO CORE MELT ACCIDENTS This section describes the results of the key accident sequences selected for the Zion statien for IDCOR Task 23.1. Four sequences were selected for evaluation. In brief, the four accident sequences are,

1. TMLB/ seal LOCA - Loss of all AC power and auxiliary feedwater with a 50 gpm/ pump seal LOCA imposed at 45' minutes.

2. TMLB - Loss of all AC power and auxiliary feedwater.

3. SLFC - Small (2 inch) cold leg break, failure of ECCS recircu-lation, fan coolers and containment sprays available.

4. ALFC - Large cold leg break, failure of ECCS recirculation, fan coolers and containment sprays available.

4.1 Plant Response for TMLB/ Seal LOCA In this section, the plant response to a TMLB sequence with a 50 gpm/ pump seal LOCA will be discussed. This highly improbable sequence involves a loss of all AC power and loss of auxiliary feedwater capability. No Emergency CoreCoolingSystems(ECCS)withtheexceptionofthepassivesafetyinjection accumulators would be available due to the loss of AC power. The reactor coolant pumps seals are assumed to be degraded, due to the lack of both seal injection and component cooling water, resulting in a 50 gpm/ pump seal LOCA 45 minutes into the accident. The containment safeguards systems (fan coolers andcontainmentsprays)wouldalsobeunavailable.

4.1.1 Reactor Coolant System Response The initiating event involves a total loss of primary and active secondary heat sinks at time zero. The steam generators act as a heat sink by relievingsteamatthesteamgeneratorsafetyvalvesetpoint(1065 psia)until the secondary side inventory is depleted (about 1.7 hrs). As the steam generators relieve energy, the primary side pressure quickly drops to about

4 2075 psi and remains there until the 50 gpm/ pump seal LOCA occurs at 45-minutes. 'At this~ time the primary system pressure drops to approximately 2C50 psia as shown in Fig. 4.1.

After the secondary side has boiled dry, the primary-system pressure

.quickly rises, due .to ' decay heat generation, to the pressurizer safety valve

.setpoint (2500 psia). The pressurizer safety valve intermittently opens and

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this time due to pump seal leakage. From the time that the seal LOCA occurs

(.75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months ) and the pressurizer safety valve first opens (about 1.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) until vessel failure, the primary system loses inventory and the core water level begins to drop as shown in Fig. 4.2, When' the core water level reaches approximately 14 ft (as measured from the bottom of the reactor vessel) the zircaloy fuel cladding begins to react with the steam to produce' hydrogen.

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Due to decay heat generation and the heat of reaction from zircaloy/

steam reaction, the fuel begins to melt at approximately 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months as shown in Fig. 4.4. When approximately 50 percent of the uranium dioxide fuel has i melted it is assumed to slump onto the lower core support plate failing this ;

support in a short time and slumping into the lower head of the reactor. vessel causing the primary system pressure to spike to approximately 2000 psia as shown in Fig. 4.1. The core debris will thermally attack the welds around the reactor vessel penetrations causing the penetration to fail at about 3.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Blowdown of the primary system occurs from a pressure of approximately 2000 psia into the reactor cavity. The molten core debris is ejected first, ablating the penetration failure to several times its original size of 1.5 inches diameter. The water / steam / hydrogen mixture then follows with' the safety injection accumulators discharging into the primary system near the end of the depressurization.

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4-7 4.I.2 Containment Response The containment temperature and pressure remain at normal operating levels until the pump seal LOCA is imposed and the primary system begins to relieve steam and water. This occurs at approximately 0.75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months as shown in Fig. 4.5. At approximately 1.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the pressurizer quench tank rupture disk fails and continued containment pressurization is due to both pump seal leakage and flow through the safety valve. The containment pressure and temperature increase to approximately 35 psia and 230 F just before reactor vessel failure. When the reactor vessel fails, the resultant blowdown increases the pressure to approximately 63 psia as shown in Fig. 4.5. At the time of vessel failure, approximately 112,000 lbm of core debris is ejected from the reactor cavity and dispersed onto the floor of compartment B. The remainder of the core slowly melts and drops into the reactor cavity.

The water remaining in the reactor vessel at the time of failure is ejected from the reactor cavity and onto the floor of compartment B. Also, the safety injection accumulators discharge during the primary system depres-surization. Due to decay heat generation the reactor cavity dries out at 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months as shown in Fig. 4.6 and concrete attack begins after the core debris-heats up to the melting point of concrete at about 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Dry out in compartment B occurs at approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , as shown in Fig. 4.7. At this time, the containment pressurization rate decreases due to reduced steam generation as shown in Fig. 4.5 and further containment pre'ssurization from 120 psia to ultimate containment pressure (149) psia is due to non-condensable gas and steam generation from core-concrete attack and heating of the contain-ment atmosphere. Concrete attack occurs in the reactor cavity but is precluded in B compartment. Due to the large surface area of the floor in B compartment over which the core debris is dispersed, decay heat from this portion 'of the debris can be removed by natural convection to the containment atmosphere, radiative heat transfer to surrounding walls, and conduction through the concrete floor such that the debris temperature remains below the melting l point of concrete.

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For this sequence, the containment failure pressure of 149 esia is attained at approximately 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months . The containment is expected to remain intact until this time.

4.2 Plant Response for TMLB In this section, the plant response to and expected fission product release from a TMLB sequence will be discussed. The TMLB sequence involves a loss of all AC power and loss of auxiliary feedwater capability. No Emergency Core Cooling Systems (ECCS) with the exception of the passive safety injection

' accumulators would be available due to the loss of AC power. The containment safeguards systems (fan coolers and containment sprays) would also be unavail-able. Comparisons between this sequence and that with a seal LOCA show only minor differences in the primary system behavior and essentially no difference for the containment behavior.

4.2.1 Reactor Coolant System Response The initiating event involves a total loss of primary and active secondary heat sinks at time zero. The steam generators act as a heat sink by

! relieving steam at the steam generator safety valve setpoint (1065 psia) until the secondary side inventory is depleted (about 1.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

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inventory is depleted as shown in Fig. 4.8.

l After.the secondary side has boiled dry, the primary system pressure quickly rises, due to decay heat generation, to the pressurizer safety valve setpoint (2500 psia). The pressurizer safety valve intermittently opens and closes relieving water and/or steam. The primary system pressure remains near the 2500 psia setpoint until reactor vessel failure as shown in Fig. 4.8.

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4-13 From the time the pressurizer safety valve first opens (abou; 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ) until vessel failure, the primary system inventory decreases and tne core is ur.ceverec at about 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months as illustrated by the core water level history shown in Fig. 4.9. When the core water level reaches approximately 14 ft (as measured from the bottom of the reactor vessel) the zircaloy fuel cladding begins to react with the steam to produce hydrogen.

Due to decay heat generation and the heat of reaction from zircaloy/

steam reaction, the fuel begins to melt at approximately 3.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as shown in Fig. 4.10. When approximately 50 percent of the uranium dioxide fuel has melted it is assumed to slump onto the lower core support plate failing this support in a short time and slumping into the lower head of the reactor vessel causing the water level in the core to boil up at about 3.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months as shown in Fig. 4.9. The core debris will thermally attack the welds around the reactor vessel penetrations causing the penetration to fail at about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Blow-down of the primary system occurs from a pressure of approximately 2500 psia into the reactor cavity. The molten core debris is ejected first, ablating the penetration failure to several times its original size of 1.5 inches

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diameter. The water / steam / hydrogen mixture then follows with the safety injection accumulators discharging into the primary system near the end of the depressurization.

4.2.2 Containment Response The containment temperature and pressure remain at normal operating :

IcVels until the quench tank rupture disk fails due to lifting of the pressur-izer safety valves. This occurs at approximately 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months as shown in Fig.

4.11. At about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the reactor vessel fails and the resultant blowdown ,

j increases the pressure to approximately 60 psia as shown in Fig. 4.11. At the time of vessel failure, approximately 112,000 lbm of core debris (51 percent of core) is ejected from the reactor cavity and dispersed onto the floor of l compartment B. The remainder of the core slowly melts and drops into the reactor cavity.

l The water remaining in the reactor vessel at the time of failure is ejected from the reactor cavity and onto the floor of compartment B. The 1 i

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Due to decay heat generation the reacter cavity dries out at 8.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , Fig.

4.12 and concrete attack begins after the core debris heats up to the melting point of concrete at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Dry out in -compartment B occurs at approx-imately 15.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as shown in Fig. 4.13.

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At this time, the containment pressurization rate decreases due to reduced steam generation as shown in Fig.

4.11_ and further containment pressurization to the ultimate containment ,

pressure (149 psia) is due to non-condensable gas and steam generation from core-concrete attack and heating of the containment atmosphere. Concrete attack occurs in the reactor cavity but is precluded in B compartment. Due to the large surface area of the floor in B compartment over which the core debris is dispersed, decay heat from this portion of the debris can be removed by natural convection to the containment atmosphere, . radiative heat transfer to surrounding walls, and conduction through the concrete floor such that the debris temperature remains below the melting point of concrete.

The flame temperature criteria for a global burn in dry air'(1310*F) is exceeded in only the reactor cavity where only small amounts of hydrogen exist.

For this sequence, the containment failure pressure of 149 psia is attained at approximately 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months . - The containment is expected to remain intact until this time. The best-estimate containment break area which develops at 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months (approximately .01 'ft2 ) is that which is necessary to relieve the pressurization rate. The effect of a larger break area is addressed within IDCOR Task 23.4. It should be noted that the analysis assumed a direct release to the environment; a release to the auxiliary building would be more likely and would further reduce the environmental release.

4.3 Plant Response for SLFC The SLFC sequence involves a small~ LOCA initiator with success of ECCS in the injection mode only and ECCS recirculation failure at switchover.

Auxiliary feedwater is operational throughout the accident as are the reactor containment fan coolers. One train of containment spray is available in the injection phase.

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-l 4.3.1 Reactor Coolant' System Response )

1 The initiating event involves _ a small (2 inch) cold leg break at time zero. The primary system depressurizes very quickly to approximately 900 psia as shown in Fig. 4.14. The ECCS maintains the core in a cooled and ~

covered configuration until the low-level in the RWST is reached at approxi- l mately 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . At this time, the attempt to switch ECCS to the recircula- l tion phase is assumed to fail. The vessel water level begins to drop at approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months with core uncovery occurring at 7.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .as shown in Fig. 4.15. At approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the passive accumulators begin to inject water causing the pressure to fluctuate as shown in Fig. 4.14. At approxi-mate' < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the water level drops to the 14 - ft. level and the zircaloy cladding begins to react with the steam producing hydrogen. As shown in' Fig.

4.16, approximately 1200 lbm (65 percent clad reaction) of hydrogen is producedd The high production of hydrogen is due to two very conservative assumptions.

First, it is assumed that clad failure effectively doubles the surface area available for zircaloy-water reaction. In addition, a model is used for hydrogen production in recovered, overheated nodes which' assumes intact geometry. As seen in Fig. 4.16, this second assumption .results in a large amount of hydrogen produced when the core slumps into the lower plenum and swells the water level up over the lower half of the core. In reality, the lower half of the core would be largely disrupted, if not completely removed by the time of core slump. These assumptions- were made to conservatively bound the possible hydrogen production, and they demonstrate the insensitivity of the containment response to the details of the core melt process. '

The U0 fuel begins to melt at approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months as shown in 2

Fig. 4.17 with core slump into the lower plenum occurring at 13.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months causing the core water level' to boil up to 15 ft as shown in Fig. 4.15.

Shortly after core slump, at 13.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months the reactor vessel fails and primary system blows down from a pressure of approximately 800 psia as shown in Fig.

4.14. The core debris will be ejected first ablating the original penetration to several times its original size with the water / steam / hydrogen mixture 1 following.

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i Due to the initial blowdown, the pressure and temperature in the containment rise to 25 psia and 205'F as shown in Fig. 4.18. The reactor-containment fan coolers are activated by a pressure of 22 psia at approximate-ly 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Due to fan cooler operation the containment pressure drops- to about 20 psia at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months as shown in Fig. 4.18. At this-time .ECCS fails and increased steam flow begins to repressurize the containment. However, as

the water level in the core subsequently drops the steam flow out the break-decreases and the fan coolers again bring the containment pressure down to L about 16 psia at 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months shown in Fig. 4.18. At vessel failure, the rapid primary system depressurization causes the containment pressure and the temperature to spike as shown in Fig. 4.18. At vessel failure, approximately 112,000 lbm of core debris is ejected from the cavity and dispersed onto -the ~

floor of compartment B. The remainder of the fuel -slowly. melts and drops into the reactor cavity.

The water remaining in the vessel as well as the water in .the reactor cavity is ejected out of the reactor cavity by the blowdown .of the vessel. However, spillover from the lower compartment quickly refills the

cavity. Due to continued fan cooler operation, the debris remains water covered and cooled in the reactor cavity and containment pressure continually drops as shown in Fig. 4.18 and 4.19. With the exception of a very small

( amount of concrete attack, less than 1 inch, in the reactor cavity and 'ower l compartment at vessel failure, concrete attack is preclu'ded.

Due to the large amount of calculated hydrogen production, a global burn is initiated at reactor vessel failure. As seen in Fig. 4.18 the peak containment pressure (32 psia) is reached at this time. It should be noted that the use of a 1300 F flame temperature value to predict global burns is conservative under these conditions (steam mole fraction of .3 and higher).

While the containment sprays are available in the injection phase, l the pressure setpoint of-37.7 psia is never reached so the spray system-is not functioning at any time in this sequence. The spray system could be operated

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4-28 manually from the control room to suppress the containment pressure and temperature further and also aid in fission product removal.

4.4 Plant Response for ALFC The ALFC sequence involves a large LOCA initiator with success of ECCS in the injection mode only. ECCS recirculation fails at switchover.

Auxiliary feedwater is operational throughout the accident as are the _ reactor containment fan coolers.

One train of containment sprays is operational in the injection phase.

4.4.1 Reactor Coolant System Response The initiating event involves a double-ended cold leg break at time The ECCS zero. The primary system depressurizes as shown in Fig. 4.20.

maintains the core in a cooled and covered configuration until the low-level At 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , the attempt to switch ECCS to in the RWST is reached (0.5 hrs).

the recirculation phase fails. The vessel water' level begins to drop at approximately .8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with core uncovery occurring at .83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months as shown in Fig. 4.21. At approximately 1.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the water level drops to the 14 ft level and the zircaloy cladding begins to react with the steam producing hydrogen. As shown in Fig. 4.22, approximately 950 lbm (49 percent clad reaction) is produced. The hydrogen production is considered to be conserva-tive for the same reasons stated in the discussion of the previous case.

The UO fuel begins to melt at approximately 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months as shown in 2

Fig. 4.23 with core slump into the lower plenum occurring at 2.2 - hours.

Shortly after core slump, at 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , the reactor vessel fails and the core debris will flow out first, ablating the original penetration to several times its-original size, with remaining water in the primary system following.

4.4.2 Containment Response Due to the initial blowdown, the pressure and temperature in the containment rise to 45 psia and 330*F as shown in Figs. 4.24 and 4.25. The reactor containment fan coolers (RCFC) and containment spray injection are

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9 10 O[f .{ j 3 4 5 6 7 53 TIME HR Fig. 4.25 ALFC containment temperatures.

.4-35 actuated at essentially time zero and serve to drop the containrent pressure j to about 16 psia at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as shown in Fig. 4.24. Because of tne large calculated hydrogen production during core slump, two global burns occur in the lower compartment which momentarily increases the temperature there as shown in Fig. 4.25. At vessel failure the core debris drops into a ficoced reactor cavity and the resulting steam spike and quench raise the containment l pressure to 33 psia. Due to the low primary system pressure _ at vessel failure no core debris is ejected from the reactor cavity.. The remaining fuel slowly melts and drops into the reactor cavity.

Concrete attack occurs in the reactor cavity for a short period of time after vessel failure during the quenching process. Concrete attack in the reactor cavity is limited to less than 5 inches. Due to continuous fan cooler operation, the reactor cavity remains flooded and the debris remains cooled precluding further concrete attack.

Since RCFC operation keeps the containment oressure low for the long-term 23 psia at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , this accident sequence presents virtually no threat to containment integrity.

5-1 i

i 5.0. RECOVERY ACTIONS I

I Section 4 discusses plant response to the identified accident-l sequences. The results demenstrate that there is substantial tirre for ~

operator actions to be implemented to prevent core degra'dation or' mitigate its.

l.

L consequences. This section discusses the effects of some examples of possible-operator actions that could be taken during these accident sequences to

! achieve a safe stable state. The limited set of operator actions considered

-in this section should be viewed as an example of the capabilities of the plant with only limited operator response.

Operators are trained and emergency procedures are written so that a certain function is achieved by utilizing whatever relevant. systems are available. The Zion emergency procedures specify a wide variety of operator actions which can be taken to provide adequate core cooling and maintain containment integrity. These actions include many options for the required functions. Table 5.1 summarizes some of the various means of injecting water-i into the reactor pressure vessel thai, are available to'an operator. Included

in this table are sources of water, system flow ranges and power requirements for system oper.stion. Similarly, Table 5.2 summarizes some of the various l means for containment cooling.

l The operator actions analyzed in this section were selected from the

! possibilities resulting from the availability of these systems 'for each-sequence and tre included in existing emergency procedures at Zion. The scht1on is not all inclusive nor necessarily the preferred actions from an operational viewpoint. However, they provide examples of the effects on accident sequence progression that can be achieved through the utilization of the availchle systems.

Seven sequences uere selected for evaluation:

1. TMLB with recovery of one vital bus at 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

2. TMLB/ seal LOCA with recovery of one vital bus at:

5-2.

Table 5.1 ZION ECCS PUMP DESIGN PARAMETERS i

Charging SI LPI (RHR)

! Pumps Pumps i Pumps ,

3 2 2 Number 1.93E7- 1.17E7 ~4.14E6 Design Pressure (Pa) 422- 422 477-Design Temperature (K) 3 9.5E-3 2.52E-2 1.9E-1 Design Flowrate (m /s)*

1617.8 762 107 Oesign Head (m) 3 2.56E-2 4.lE-2 2.8E-1 Maximum Flowrate (m /s)*

396 457 91 Head at Maximum Flowrate (m) 1.58E7 1.05E7 1.17E6 ShutoffPressure(Pa) 447 298 -298 Motor. Power (kw)

RWST RWST RWST Injection Source 3 3 (1325 m ) (1325 m )

RHR outlet RHR outlet Containment Recirculation Source ' Sump

Design flowrate for a single pump.

d n- -. ~ -. -

Table 5.2 ZION FAN COOLERS AND CONTAINMENT SPRAY-Fan Coolers i Number 5 ',

Fan Type Centrifugal Speed (rpm) 1200 (normal)/900 (LOCA) l-Capacity (m3 /s) 40 (normal)/25 (LOCA)

Heat Removal (kw) 923 (normal)/23,733 (LOCA)

Containment Spray System Pumps ,

Number 3 Type 2 motor-driven Centrifugal 1 diesel-driven Centrifugal 3

Flowrate (m /s) 1.90 E-1 Spray Additive NaOH(30w/oinwater) 3 SprayAdditiveTank(m) 18.9 3

Injection Source RWST (1325 m )

Recirculation Source Containment Sump

5-4 A. I hour B. 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months C. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D. 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months

3. SLFC with recovery of recirculation capability at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

4 ALFC with one charging pump operational and limited RWST refill capability.

5.1 TMLB with Recovery at 2.5 Hours This sequence is initially identical to the transient accident without seal LOCA described in Section 4.2. At 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , just before the onset of core melting, it is assumed that one vital bus is recovered. This makes operational one train of ECCS (one charging pump, one high pressure injection pump, and one low pressure injection pump), one motor-driven auxiliary feedwater pump, one containment spray pump, and 2 fan coolers. Service water for the recirculation cooling heat exchangers is also recovered so that injection can continue indefinitely.

As shown in Fig. 5.1, the core is uncovered just prior to the time of AC power restoration. The primary system pressure at this time is equal to the pressurizer safety valve pressure as shown in Fig. 5.2. This-pressure is above the shut-off head of the charging pumps. Feedwater addition to the steam generators, shown in Fig. 5.3, while limited in rate by the capacity of the single AFW pump, causes the primary system pressure to decrease relatively rapidly. As shown in Fig. 5.1, this allows the core to be recovered by the charging pump and high pressure injection pump within about 30 minutes of the time of power restoration. Consequently, hydrogen generation, Fig. 5.4, is limited to a negligible amount. The operators are further assumed to manually open the pressurizer PORVs at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in an attempt to vent the (unknown to them)quantityofhydrogenpresumedtoexistinthesystem. This time is just

d i

1 1

1 l

i

~

ZION TMLB/NO LOCA/RECOV AT 2 5 HR 4

o G VESSEL - PRI SYS COLLAPSDgPRESSLRIZER a

o~

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) TIME HR 4

! Fig. 5.1 TMLB'2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - primary system water levels.

.1 I

ZION TMLBeND LOCA/RECOV AT 2.5 HR l "

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) 0 1 2 3 4 5 I TIME HR Fig. 5.2 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - primary system pressure.

l

i l ZION TMLB/NO LOCA/RECOV AT 2.5 HR

, . . i.j iei iiii.i_q i ii i iiiii; i ii ii iiiig ii i iii iiig iii iiiiii;iiiii iiiigiiii iiiiig iiiiiiiigiii iiii sig.

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TIME HR i Fig. 5.3 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - steam generatur level.

i i

l

i ZION TMLB/NO LOCA/RECOV AT 2.5 HR N j iii . iiii g iiiiiiii. g iiii . iii. g i . iiii ii g iiiiiiiii g i s iiiii .i j iiiiiiiii g i . i . iiiiil 4 iiiiiiiiiiiiiiiiit j

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O- 1 2 3 4 5 l TIME HR Fig. 5.4 TMLB' 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months recovery - H2generation.

9 after the system has gone effectively solid (Fig. 5.1), and drastically reduces the primary system pressure.

Late in the accident. 'the primary system is solid (except for tne small amount of trapped hydrogen) due to continued injection of ECCS water and removal of water through the PORVs. As shown in Fig. 5.2, calculated primary system pressure eventually goes unstable; this is merely an artifact of the code due to large time steps, subcooled primary system, and the small hydrogen

" bubble" trapped in the primary system. This artifact does not alter the conclusion that the system is recovered in a safe, stable state.

5.2 TMLB/ Seal LOCA with Recovery at 1 Hour This sequence is initially identical to the transient accident with seal LOCA described in Section 4.2. At I hour, restoration of 1 vital bus is achieved which recovers the engineered safeguards features listed in the previous section. At 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the operators manually open the pressurizer PORVs to depressurize the primary system.

Recovery at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is prior to complete loss of steam generator inventory and occurs when the primary system inventory is still intact except for the small seal LOCAs which ocurred at .75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months . Figure 5.5 shows the increase in steam generator level. The primary system response (Fig. 5.6) is a shrinkage of coolant which just barely drains the pressurizer for awhile but introduces no significant voids into the primary coolant system. Primary system pressure, Fig. 5.7, decreases somewhat but is more stable than the previous case due to the much greater coolant inventory at the time of power restoration.

The system is stable long-term with decay heat being removed by the steam generators and the injection of ECCS water in the recirculation mode.

l 5.3 TMLB/ Seal LOCA with Recovery at 2.5 Hours l

I

This sequence is defined identically to the recovery sequence described in Section 5.1 except that the main coolant pump seals are assumed

i a

ZION TMLB/LOCA/RECOV AT 1 HR

~

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1 2 3 4 5 l

TIME HR I fig. 5.5 TMLB' (seul LOCA) I bour recovery - ste6m generator level.

4 4

I

4 1

i t

i ZION TMLB/LOCA/RECOV AT 1 HR I - 0 VESSEL aPS COLLPSD vPRESSlRIZER t

4 O

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i TIME HR

! Fig. S.6 TMLB' (seal LOCA) I hour recovery - primary system water levels.

l

i t

i ZION TMLB/LOCA/RECOV AT 1-HR i

n D f i. o u n g s i n i o u g o s iiii n g iiiii n o g i s i s ii s u i s ii o ii n ii n i s o i u i s iiiiii n l + > > i s ii u i s i a i > > "t

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4 5 6 7 8 9 10 O. I 2 3 i TIME HR t

Fig. S.7 TMLB' (seal LOCA) I hour recovery - primary system pressure.

1, I

4

5-13 to fail at .75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months and the operators are assumed to manually ocen tne pressurizer PORVs within 10 minutes of power restoration.

The primary system pressure is reduced rapidly by the opening of the PORVs at 2.66 hours7.638889e-4 days 0.0183 hours 1.09127e-4 weeks 2.5113e-5 months (Fig. 5.8). The resulting increase in injection flow allows the core to be recovered by 3.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as shown in Fig. 5.9 and rapidly terminates hydrogen production (Fig. 5.10). The plant remains coolable long-term in a state identical to that described in Section 4.1.

If the PORVs are not opened, the behavior of the plant after recovery is initially similar to the case described in Section 5.1. However, the effect of the seal LOCAs is to accelerate the loss of inventory and core uncovery occurs somewhat sooner than in the earlier case. Consequently, the zircaloy-water reaction is proceeding at a more rapid rate at the time of power restoration and accelerates further when injection water reaches the core. Because of the low feedwater flowrates to the steam generators, there is limited heat transfer area available, and this area is easily blocked by the hydrogen produced in the core. The end result is that primary system depressurization is slowed, injection rates reduced, and the accident progresses to vessel failure. Ocening the PORVs is consistent with procedural guidelines and training which dictate that the core be recovered as soon as possible.

This sequence demonstrates that recovery of the plant in high pressure sequences after hydrogen production commences is relatively sensitive to the details of the actions taken. Injection and auxiliary feedwater j flowrates must be made large enough to terminate hydrogen production before l the steam generators' effective heat transfer area is lost.

5.4 TMLB/ Seal LOCA Recovery at 6 Hours This case is similar to the previous TMLB/ seal LOCA cases except that recovery is postponed until 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor vessel failure.

As shown in Fig. 5.11, fan coolers and the one containment spray pump are immediately effective in reducing containment pressure. Sprays are

5-14 i

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ZION TMLB/LOCA/RECOV AT 2 5 HRS o O VESSEL ,. PS CLLPSD

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Fig. S.9 TMLB' (seal LOCA) 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months recovery - primary system water levels.

1 3

4 i

r - - -

ZION TMLB/LOCA/RECOV AT 2 5 HRS O u.u.a.i.no..oi..o....pui..... ion....i...on.ii....o...i..o.....i.o.. .... m ...

7 1

to c.

i . . ~l 1

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I 2 3 4 5 6 7 8 9 10 0

1 TIME HR Fig. 5.10 TMLB' (seal LOCA) 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months recovery - 112 generation.

i 1

i ZION TMLB/LOCA/RECOV AT 6 HR i

) em

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O. 1 2 3 4 5 6 7 10 TIME HR Fig. 5.11 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - containment pressure.

5-18 lost at 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months due to depletion of the RWST, but the fan ecolers remain effective. Through this time in the accident, the debris beds in the lower compartment and cavity have remained water-covered, (see Figs. 5.12 and 5.13),

and containment cooling by either sorays or fan coolers ensures continued cooling of the debris thereafter. Figure 5.14 shows that while the contain-ment atmosphere does become relatively more flammable due to the removal of steam by the fan coolers, flammability limits are not approached. Given that further concrete attack and non-condensable gas generation is prevented, there is no long-term risk of hydrogen or carbon monoxide combustion.

Steam generator pressures and temperatures are also immediately reduced by feedwater addition as shown in Fig. 5.15. At this time, the primary system still contains virtually the entire volatile fission product mass; this represents roughly 20 percent of the decay heat. Figure 5.16 indicates that feedwater addition to the steam generators is only partly effective in cooling the primary system. As modeled in MAAP, only the 4 hot and cold leg nodes are directly coolable by the steam generators. After vessel failure, cooling of the upper plenum, for example, depends on the establishment of inter-loop natural circulation flows. Figure 5.17 shows the magnitude of these flows. After recovery of ECCS, the pump bowls fill with injection water, sealing the cold legs and essentially eliminating the internal flows. Since the upper plenum heating by deposited fission products exceeds the heat which can be lost through adjacent insulation, the upper plenum begins to heat. Although not modeled in MAAP, hand-calculations indicate that with nominal values of vapor pressure, sufficient counter-current natural circulation flows would be developed in the hot legs to transport gaseous fission products to the steam generators. If surface reactions with steel greatly reduced the vapor pressure, the upper plenum would be expected to melt eventually unless a hole develops in the hot leg to allow natural circulation flows between the primary system and containment. At any rate, since contain-ment failure is prevented by the fan coolers (RHR sprays are also available thoughnotcredited),releasestotheenvironmentwouldbenegligible.

In this particular case, long term cooling is provided by fan coolers, auxiliary feedwaters, and ECCS injection in the recirculation mode.

Alternative sequences where recirculation is not credited show similar re-

t i

l ZION TMLB/LOCA/RECOV AT 6 HR

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i TIME HR

! Fig. 5.12 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - lower compartment water level.

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! LION TMLB/LOCA/RECOV AT 6 HR

= , OIFPER CEDFT AIDER CtNFT

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TIME HR Fig. 5.14 TMLB' (seal LOCA) 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months recovery - contairunent flame temperatures.

i l

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5-25 sponses, since the fan coolers alone are easily capable of cooling the ccntain-ment.

5.5 TMLB/!eal LOCA with Recovery at 15 Hours b

This case is qualitatively similar to the _6 hour recovery sequence.

As in the previous case, containment pressure (Fig. 5.18) is quickly reduced l

by the spray pump and the two fan coolers. Addition of water to the lower

! compartmentandcavityquicklyquenchesthedebristhere(Fig.5.19),terminat-ing concrete attack. While condensation of steam increases the flammability of the' containment atmosphere somewhat (Fig. 5.20), flammability limits are not reached.

3 As was mentioned before, there is the potential for ultimately

} over-heating and melting the uncooled upper plenum area. This would have no

public risk consequences since containment integrity is maintained. ,

5.6 SLFC with Recovery at 10 Hours l

This sequence is initially identical to the SLFC accident described

in Section 4.3. It is assumed that at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (just prior to the onset of

! core melting), recirculation capability is restored. This results in the

{ recovery of I charging pump,1 HPI pump and 1 LPI pump. As shown in Fig. 5.21,

! water level in the core is rapidly recovered, which terminates hydrogen l production, Fig. 5.22. The core mass, Fig. 5.23, shows a slight increase due

! to oxygen absorption, but melting is prevented due to the quick recovery of j water level.

j The relatively long time available between loss of ECCS at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and recovery at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is due in part to accumulator discharge (see Fig. 5.24).

It should be noted that this situation depends sensitively on the assumed j break size.

l 5.7 ALFC with 1 Operational Charging Pump This sequence is defined as a double-ended cold leg break with only one charging pump available for injection. It is further assumed that, while

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recirculation is lost, RWST refill at a flowrate of at least 400 gpc is rede available from one of the several sources capable of such flow.

Figure 5.25 demonstrates that the flow from one charging cump is sufficient to keep the core covered at all times. Sprays are lost'at one hour after the switch to recirculation fails, but the containment fan coolers are capable of removing all the decay heat and containment pressure (Fig. 5.26) remains controllable.

While this sequence is not, strictly speaking, a true recovery sequence since recirculation cooling (RHR) will have to be initiated eventual-ly, it does demonstrate that there are means by which to delay core uncovery for an extended period. During this period, which is limited only' by the containment volume and the ability to supply refill to the RWST at greater than the charging pump runout flow, actions could be taken to bring the RHR system on-line or to initiate other emergency response actions for mitigating the accident consequences.

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1 6.0 FISSION PRCDUCT RELEASE, TRANSPORT A*40 DEDOSITI0t.

6.1 Introduction

. The phenomena of #ission product release frcm the fuel, tFeir transport and deposition within the primary system, their release .from the primary system. into the containment, their deposition'within the containment and the subsequent 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 l . subroutine in MAAP which is based on the FPRAT code -[6.2]. Transport of fission products outside the core boundaries is treated by. fission product models in MAAP described in Ref. [6.1]. Fission product behaviorcis evaluated-for the best estimate transport, deposition and relocation processes as j discussed in -Refs. [6.1] and [6.4]. The influence of alternative' assumptions for the vapor pressure of cesium hydroxide and other uncertainties are consider-I ed in subtask 23.4 [6.3]. The best estimate calculations, assuming cesium iodide and cesium hydroxide are the chemical state of cesium ^ and iodine, are

!- discussed below.

6.2' Modeling Approach j Evaluations of the dominant chemical species-in Ref. [6.5] show the states of the radionuclides (excluding noble gases) which dominate the public

health risk to be cesium iodide and cesium hydroxide, tellurium oxide and

strontium oxide. These and others are considered in the code when calculating the release of fission products from the fuel matrix. Vapors of these dominant.

species form dense aerosol clouds in the upper plenum, in some cases approach-ing 100 g/m3 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 would remain suspended as airborne aerosol- and vapor and would be discharged from the primary system into the containment. The rate'of dis-charge 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 9

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6-2 relief valves or through breaks in the primary system. This is also sequence

- specific.) This set of inter-related processes are treated . in ~ PAAP and essentially result in a release of all airborne aerosol and vapor. from the t

- primary system into containment immediately following vessel failure, As a result of the dense aerosols formed when fission products are 4

released from the fuel, considerable deposition ' occurs within. the primary system prior to vessel failure. For some accident sequences, the primary system may be at an elevated pressure at the time of core slump and reactor vessel failure. Resuspension of these aerosol deposits during.the primary-system blowdown is assessed in Ref. [6.6] in terms of the available experi '

mental results and basic models. It is concluded that resuspension immediate-ly following reactor vessel failure would not be significant (less .than li of the deposited materials) even for depressurizations initiated from the nominal

! operating pressure. For delayed containment failure, this small fraction of ,

material is depleted by in-containment mechanisms. l t

4 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-b circulation through the primary system after 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 primary system heat losses as I dictated by the reflective insulation. Material transport as aerosols and vapors after vessel failure is governed by the heatup of structures due to radioactive decay of deposited fission products. This heatup .is principally l determined by the transport of cesium iodide and cesium hydroxide by the natural circulation flows. In this regard, the vapor pressure of cesium-

! hydroxide is applied to both the cesium iodide and cesium hydroxide chemical

species. In carrying out these calculations, the pressurization of the primary system is dependent upon the pressurization of the containment and the h' eating within the primary system. These determine the in- and out-flows between the primary system and containment.

Deposition within the containment is calculated using thermal hydraulic conditions determined- by MAAP. The major potential sources of

_ _ _ ~ .._ _ ,_ _ _ _ . - _

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' aerosols are the releases prior to vessel failure (sequence specificI, the l airborne aerosols and vapors transferred from the primary system at the time l

l of vessel failure, the subsequent releases from the primary system.due to lon;

! term heatup, and aerosol produced as a result of core-concrete attack. At the l time of containment failure, the remaining airborne aerosol and vapor can be released to the environment at a rate determined largely by the containment leak rate. Assessments of the potential for resuspension of deposited aerosols following containment failure [6.6] show this to be negligible.

Containment failure is discussed below in Section 6.3. This discus-sion addresses, in summary form, a containment failure modes evaluation prepared for Zion which goes beyond the work of Task 10. It should be noted that both this work and Task 10 assume, for simplicity, that the containment failure leads directly to the environment. In reality, the most likely pathways lead to the auxiliary building or the mainsteam tunnels.

6.3 Containment Failure Modes Summary 6.3.1 Failure Modes Identified in the Study Conducted for Task 10 Potential containn.s..t failure modes have been addressed in the study j on the " Primary Containment Ultimate Capacity of the Zion Nuclear Power Plant for Internal Pressure Loads," prepared for IDCOR Task 10, dated April 15, 1982

[6.7].

The containment failure was defined at such levels of stress, strain i or deformation beyond which the leaktightness on the structural integrity of !

the containment could no longer be assured. The containment itself was f evaluated as a axisymmetric structure without considering any disturbances.

l The containment penetrations and the equipment hatch were evaluated separately and it was found that they did not limit the containment ultimate capacity.

The governing failure mode was determined to be hoop tendon yielding at 1 percent strain. Other potential containment failure modes were reviewed and found to have the margin factors as listed below.

6-4 Margin to Failure

1. Concrete Shear in Basemat 1.27

2. Concrete Shear in Containment Wall 1.28

3. Concrete Shear at Equipment Hatch 1.29

4. Concrete Compression in Basemat 3.85

5. Concrete Compression in Containment Wall 5.83

6. Reinforcing Stress at Equipment Hatch 1.32

7. Equipment Hatch (conservative) 1.0

8. Soil Pressure 1.95

9. Liner Fiber Strain (upper bound) 2.10

10. Reinforcing Strain 4.74

11. Hoop Tendon Strain 1.0 The failure pressure was determined to be 134.4 psig at a radial displacement of the containment wall at midheight of approximately 4.8".

The study projected that if the containment pressure is increased further, the hoop tendons would rupture at a strain of 4 percent. The contain-ment pressure at this point was determined to be 143 psig at a radial displace-ment of the containment wall at midheight of approximately 30".

Figure 6.1 illustrates the relationship between containment pressure and radial displacement.

6.3.2 Study to Identify Containment Failure Modes The purpose of this additional study is to identify local or gross containment failure modes at levels beyond the one percent strain limits set in the above mentioned study [6.7]. A failure is dafined as the opening of a leakage path from the containment to the environment with the further considera-tion for this study that it must be sufficiently large to terminate the building pressurization. The study was performed to define the location and nature of the failure and provides an estimate of the possible leakage areas for each location. In general, these studies show that. the failures take place at containment pressures and deformations smaller than the one causing gross containment failure due to the hoop tendon rupture.

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6-6 Due to the nature of the problem the failure modes are evaluated by

[

engineering judgment. The calculations and results are approximate, l

l The following is a summary of the study and lists the nature of tre

[

failure, the location and the estimate of the leakage area. The estimated leakage area is compared to a vent area of 16 sq. inches. It was determined in Ref. [6.7] that a vent area of 16 sq. inches would prevent any further rise in containment pressure under the assumed accident scenarios. In particular, this value corresponds to steam pressurization by 1". decay, which generally overestimates the long term pressurization rate. For the TMLB' case of interest here, a vent area of only 1.3 square inches would suffice. The opening of larger leakage areas will therefore cause a depressurization of the containment.

1. Gross containment failure due to the hoop tencon rupture at the midheight of the containment.

The failure takes place at a radial displacement of 30".

The maximum leakage area is estimated to be in the order of 100 sq. ft.

2. Failure of bellows seals on the penetrations for the steam and feedwater lines.

A typical configuration is shown in Fig. 6.2.

The failure is estimated to take place at a radial displacement of less than 30". The maximum leakage area is estimated to be a maxitum of 220 sq. inches (much greater than 16 sq. inch reference vent area) per each of 8 penetrations.

3. Failure of the seals of the equipment hatch.

The configuration is shown in Figs. 6.3 and 6.4. The failure is estimated to take place at displacements of less than 30",

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' 6-10 the high containment temperature contributing to the cegrada-tion of the seals. The maximum leakage area is esticated to be less than 16 sq. inches.

4. Failure of the collar adjacent to the equipment hatch penetra-tion sleeve.

The configuration is shown in Fig. 6.3. The failure is es-timated to take place at displacements of less than 30". The maximum leakage area is estimated to be greater than 16 sq.

inches.

5. Liner failure between the electrical and purge line penetra-tion.

The layout is shown on Figs. 6.5 and 6.6. The failure is estimated to take place at displacements of less than 30". The maximum leakage area is estimated to be _ larger than 16 sq.

inches.

6. Other potential failure modes are:

liner failure in areas where adjacent structures obstruct a) the containment growth.

b) liner failure at piping penetrations where the piping -

containment interaction causes additional liner strains.

These potential failure modes are not evaluated at this time.

Based on the results from this additional study it is concluded that it is likely that local leakage paths will open prior to a gross containment failure. The most likely vent area is one that just limits further pressuriza-tion. Such a value is assumed here for the best-estimate case.

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s

6-13 I The local failures are listed as' Item Mumbers 2 Lthrough 6. -

The gross failure is listed as Item 1.

, 6.3.3 Other Potential Failure Modes In addition to the foregoing failure modes, two additional postulat-

, ed failures have been considered. . These are: " bypass failures" and " failure

! to isolate". The bypass failures include steam generator tube rupture events c which are not risk significant and the' interfacing systems LOCA (event "V")

j which will be discussed in a separate report [6.8]. Also, bypass events l include potential openings in the containment. pressure boundary which result i from maintenance errors, early component failures, or similar causes and which l involve areas not equipped with isolation features. Examples might include a j liner weld failure or personnel hatches being left open and going undetected.

These latter failures are judged to have a negligible probability of occur-rence at Zion-due to the multiplicity of barrier systems, and alarms associated i

therewith, installed at Zion. Such systems include weld channel pressurization, l penetration pressurization. .and isolation valve seal water systems and 'the

! interlocks and pressurization systems for access hatches.

1 j Failure to isolate events involve systems which are normally or l might be normally open to the containment atmosphere or the RCS at one end and l which do not terminate in a closed system at the other end. This narrows down

! quickly to the 10 inch diameter containment pressure and vacuum relief system l which is used about 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months a week for atmospheric purging. This line is j equipped with two, series, power operated isolation valves actuated from separate safeguards trains. The conditional probability of this line being l open given an accident is 1 x 10-0 .

l l

Evaluations of these failure modes, using MAAP, are the subject of IDCOR Subtask 23.5 [6.8] and are covered in a separate report on that subtask.

6.4 Sequences Evaluated

-m< --e, e.. y- 3,..,ee.- -, -w= -w-g+ -.--w. ee-- -m, 9- , -,,,gmy--+--m., .cp-*-- m- - - - , - - e- ,-%,-ev -r--r.-.y-- -

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6-14

6.4.1 TMLB' With a Seal LOCA-The use of MAAP leads to 3 source terms, or release fraction character '

izations for the various containment failure situatio'ns. -The failure to isolate the containment' situation is treated, along with containment bypass scenarios, in a separate task, 23.5 [6.8] (refer to Fig. 2.1). A source terrr for the best estimate case is presented in' tabular form in Table 61 and is supplemented by more detailed discussions below.

l ,

This analysis considers retention of volatile ~ fission products

within the primary system. as a function of species vapor pressure, coupled with long term overpressure failure of the containment:1 This case constitutes a best estimate or most likely case. Volatile fission products are distributed - -

l j throughout the primary system prior to vessel failure as modeled by MAAP.

1 These fission products are largely in the form of airborne and deposited I

aerosols, but deposited material will quickly dominate material' in aerosol form once the release of volatile material from the fuel-creates a dense l aerosol cloud. At and subsequent to vessel failure, the removal'of these I fission products from the primary system to the containment is determined by-

! that fraction of the fission products which is airborne aerosol or' vapor

! within the primary system at any instant.

I The key element of this evaluation is the amount of material deposite6 within the primary system at the time of reactor vessel failure'. For this f

TMLB seal LOCA sequence, this material deposition -in relationship to the amount of volatile products released is illustrated in Table 6.2. As shown, nearly 100% of the material released from the fuel matrix is deposited within the primary system prior to vessel failure. This deposited material could 4

4 potentially undergo chemical reactions. However, the material is considered here to be able to revolatize with a vapor pressure characteristic of cesium hydroxide, j With the heat generating material deposited, the structures are j

heated and natural circulation flows are developed as a result of density differences between primary system regions. As circulation begins, energy and t- material are transported into the hot leg and cold leg side of the steam i

I

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6-15 i-Table 6.1 TMLB' WITH A SEAL LOCA i

Assumptions:

I Containment failure elevation; l annular area, 579' to 614' Containment failure size; sufficient to terminate the pressurization' Results (at 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months ,13 hours after containment failure):

Isotope Release Fraction Cs, I .0017

! Te, Sb 2 x 10-5

, Sr. Ba < 1 x 10-5 l Ru, Mo < 1 x 10-5 i

I

Time of release: 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months l

l

- - - - - - * - r - -- N n -- v- w

6-16 Table 6.2 DISTRIBUTION OF Cs, I FOR TMLB' WITH SEAL LOCA (FRACTION OF CORE INVENTORY)

At Vessel Failure Primary System 0.995 Containment 0.005 Environment 0.0 At Containment Failure Primary System .975 Containment .025 Environment 0.0 13 Hours After Containment Failure Primary System .93 Containment .07 Environment .002 e

l

__ . .__ _ __-_. _ -~_ _ . . _ ___ _ - . ._ _ _ . _ _

17 l

generators and eventually into the downconer. The heatup of the upper plenum initiates a flow which convects heat through the primary system resulting in structural heatup. With continued heating of the upper plenum, caterial is l transported as a' vapor and airborne aerosol into the hot leg side of the stear l generators. Eventually, the hot leg side temperature reaches a level suffi-i cient to revolatilize vapor and transfer vapor and aerosol into the cold legs and the downcomer. At temperatures below 900K, little vapor is available for transmission from the primary. system into the containment at the time of containment failure or anytime thereafter. As a result of this extensive retention with the primary system and the limited concrete attack the inert

} aerosols resulting from such attack have only a secondary influence on the

! release fraction. Thus, these aerosols were neglected in this analysis which

l is a conservatism in the calculations.

I l Released from the primary system vessel failure is small and subse-l quently releases are driven only by the flow from the primary system into the containment, which is very low until containment failure (about 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ). At :

4 that time the flow from the primary system to the containment increases somewhat and sweeps airborne aerosol and vapor from the primary system into '

l the containment. Very little of the airborne material is aerosol at this time j and the releases from the primary system to the containment are principally j the result of vapor flow. With the localized containment failure and slow j depressurization, these releases are limited and holdup within the containment-

} provides for substantial retention of the material by in-containment mechanisms.

Principle mechanisms for material retention are vapor condensation, and

, gravitational settling of aerosols formed by the vapor as it flows into the containment.

l t

) With the localized failure of penetrations, liner tears or seal -

j degradation, the radionuclide releases.from the containment would be into the

! auxiliary building where additional holdup could occur. Due to the very small I releases calculated in the best-estimate analysis of the primary system and l containment behavior, this additional holdup was not calculated. The overall j influence of the release into the auxiliary building and the extent of fission i product rotation by walls and structures within the building will be evaluated j as part of the uncertainty analyses reported in IDCOR subtask 23.5.

I i

n

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i 6-18 l

The release fractions associated with the best-estimate calculation The for the TMLB' seal LOCA accident. scenario are listed 'in Table 6.1.

release fractions for the volatile fission products are low, reflecting the extensive retention of the materials by both the primary system and the containment. Release fractions for the less volatile materials are also small_

and this reflects the limited release during core melt and also the limited amount of concrete attack for this sequence at Zion. With_ limited concrete attack, little material release would be experienced due to sparging of the melt by the gases liberated from the concrete.

6.4.2 TMLB' Without a Seal LOCA Given the results noted in Section 6.1 and the phenomenology influ-encing those results, there are no significant differences in the system' response or the release fractions for these two sequences.. .The time to containment failure is sufficiently similar to insure comparable behavior.-

Hence, there are no significant differences between the sequences. .Thus, the release fractions listed in Table 6.1 are applicable to this sequence.

6.5 References 6.1 "MAAP, Modular Accident Analysis Program User's Manual," Technical Report on IDCOR Tasks 16.2 and 16.3, May 1983.

6.2 FPRAT Users Manual 6.3 IDCOR Technical Report on Task 23.4, " Uncertainty and Sensitivity Analyses for the IDCOR Reference Plants," to be published.

6.4 " Fission Product Transport in Degraded Core Accidents," Technical Report on IDCOR Subtask 11.3, December 1983.

6.5 EPRI/NSAC, " Technical Report 11.1,11.4 and 11.5, Estimation of Fission Product and Core-Material Source Characteristics," October 1982.

6.6 IDCOR Technical Report on Task 11.6, "Resuspension of ' Deposited Aerosols Following Primary System or Containment Failure " July, 1984.

6.7 IDCOR Technical Report on Task 10.1, " Containment Structural Capabil-ity of Tight Water Nuclear Power Plants," July 1983.

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t_ . 6-19 .-... ,

p , >

1 1 .- T6 .8 ~ IDCOR;Tdchnical l ReportLo'n;:Ta'sk f3.5, "Evaluatic'ns of Centainrent

~

Bypass fand- Fail _ure' to.: Isolate Sequences. for the -.IDCOR Feference "'

Plants, to,be published.:

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7-1 7.0

SUMMARY

OF RESULTS This'section of the -report _ summarizes the results of the various

. analyses performed for the Zion reference plant analysis.

7.1 Base Case Analysis Four base case accident sequences were analyzed for the Zion refer-ence plant. These four sequences are:

1. TMLB without a pump seal LOCA

2. TMLB with a 200 gmp total seal LOCA~

3. SLFC - small cold leg break (2 inches), failure of the ECCS in -

the recirculation phase, and full containment safeguards available

4. ALFC - large double-ended cold leg break, failure of -ECCS .in the recirculation phase, and full containment safeguards available.

As expected, only the two TMLB sequences lead to overpressure of the containment building and fission product release. Due to containment safeguards

! operation, the SLFC and ALFC sequences do not threaten containment integrity and the core debris remains in a safe stable state for .the long term. Table 7.1 provides values for parameters of interestffor the four base case sequences.

Table 7.2 provides the timing, energy, and magnitude' of the atiected fission product release for the best estimate behavior in a TMLB (Squm e.

L

Table 7.1

SUMMARY

OF BASE CASE SEQUENCES TMLB/wo TMLB/w SLFC ALFC Steam Generator Dryout (hr) 1,76 1.7 -NA NA' Core.Uncovery (hr) 2.26 2.2 7.2 .83 3.1 3 12 1.7 Core Melt Initiation (hr) 4.0 3.8 13.9 2.3

' Vessel Failure (hr)

Pressure Spike at Vessel Failure (psia). 61 63 32 31 300 280 1280 950-H2 at Vessel Failure (1bm)

Flame Temperature in Upper Comp. at 470 470 1200 850 Vessel Failure (*F) 32 32 NA NA

' Containment Failure (hr) 730 720 NA NA H2 at Containment Failure (1bm) l 500 500 NA NA Flame Temperature in Upper Comp. at Containment Failure (*F) 1.60 1,71 NA NA Concrete Penetration at Containment Failure (ft)

-e

7-3 ;

l. l f Table 7.2 i

l TMLB' WITH A SEAL LOCA i

l l Assumptions:

Containment failure elevation; annular area, 579' to 614' Containment failure size; sufficient to terminate the pressurization Results (at 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months ,13 hours after containment failure):

Isotope Release Fraction Cs, I .0017 Te, Sb 2 x 10-5 Sr Ba < 1 x 10-5 Ru, Mo < 1 x 10-5 Time of release: 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months l

l l

t

8-1 8.0- CONCLUSIONS From the analyses presented in this report, several conclusions may be-drawn about class 9 accident progression for plant designs of -the Zion type.

Steam spikes at core slump and vessel failure do not threaten containment integrity.

Hydrogen accumulation is not a threat to containment integrity for the sequences analyzed in this study.

Given a loss of all containment heat sink, the most likely form of containment failure is 'long term overpressure due to steam and non-condensible gas generation.

Generally, there is sufficient time and numerous systems available to terminate or mitigate the class 9 accidents studied in this report.

For cases where loss of ECCS recirculation capability would lead to a class 9 accident, refilling the RWST and-continuing ECCS injection is a viable method for delaying the need for recirculation capability.

Resuspension of deposited aerosols at the time of containment i failure is unlikely.

The most likely form of containment failure is a small break area which will terminate the containment pressurization and allow the containment to slowly depressurize in the long term.

A large fraction of the volatile fission products will be retained in the primary system in most sequences. This serves to limit the amount and rate of releases to the environment.

8-2.

- Concrete (inert) aerosol prcduction rate and timing has a small effect on fission product deposition and hence on the release from containment. However,=the retention due to_pri,Tary system retention and settling are the principal processes of interest.

- The Modular Accident Analysis Program (MAAP) is a useful tocl for predicting plant response to class 9 accidents, particular-ly for evaluating the effects of operator actions -on plant response.

- MAAP provides a sufficiently detailed description .of the e

primary system and containment response to conclude that .

environmental releases of cesium, iodine and tellurium will be less than 1% of the core inventory.

l l

l 4

A-1 APPENDIX A.1 Zicn Parameter File AA2IGN PARAMETER FILE-An AASI UNITS (n-KG-SEC-DEGA)

AA A*CERTAIN PROPRIETARr DATA NUT.3M0WN AA AUPPER L0nPAkinENT *ACOMPT-01 5.89D4 VOLunE 02 552.0 AREA UE REFUELING FOOL 03' 43.6 HEIGHT OF 5 FEAT MEAD AB0VE BOTTun UF COMPAPInENT 04 502 FLOW AkEA FROM VPPER COnPARIMENT INTO ANNULAF C0mFT 05 1195 CHAkAC. CF055-SEC AFEA 0F COnPT F0F BUPN IInE CALC 5' 06 0.0 CURB HEIGHT IN REFUELING FOOL TO ALLOW OVERFLOW

- 'N0knALLI 0 UNLE5S A55unE DRAINS BL0thED 07 5807.0 SURFACE AREA 0F OUTER WALLS 08 .011- LINEF InICLNESS ON OUTER WALL

09 .00802 OUTER WALL LINER GAP RESISTANCE 10 1.068 OUTER WALL IHIChNESS 18 1.38 THERNAL CONDUCTIVIT10F 00TER' WALL 12 878.6 SPECIFIC HEAT OF OUTEk WALL 13 2291.0 DEN 51IY10F QUTEk WALL 14 0 ENTER A 1 IF THE OUTER WALL IS SOLID STEEL A* INTERNAL OR INNER WALL 3 ARE WALLS TOTALLI CONTAINED IN A CONFT -

A*FROPS OF INNER WALLS IN A AND P ARE ASSunED THE SAME AND Ah5 A* ENTERED IN BCOnPT SECTION 15 283.3 HALE AREA 0F INTERNAL WALLS 16 0 LINER THICKNESS ON INNER WALL 17 0 LINER GAP RESISTANCE IN INNER WALL.

18 .305 THICKNESS OF INTERNAL WALLS 19 552.0 DECK (UFFEk COnPAP.TnENT FLOOR) AREA 20 0 LINER THICKNESS ON DECK 21 0 LINER GAP RESISTANCE ON DECK l b$ I bh k khnk$ bdNbbbTIVIT10F DECK 24 878.6 SPECIFIC HEAT OF DECK 25 2291.0 DENSITY UE DECK 26 0 ENTER A 1 IF THE DECK IS SOLID dTEEL 27 4.206D5 nETAL EGPT nASS .

, 28 e280.0 EQPT HEAT TRANSFER AREA 29 0. NUnBER OF IGNITION 5OURCES In A 30 0. AVERAGE DISTANCE OF THESE FROM THE CEILING of A i AAFOLLOWING PARAnETERS ARE USED TO DEIEknINE WHICH luNITEkle IGN SOUFCL:

AACAN PROPAGAIE BURNS INTO A 31 0. NO. OF IGNITERS /IGN SOURCES IN B WHICH Cas FE SEEN FF0a a 32 0. NO. OF IGNITERS /IGN S0ukCES IN D WMICH CAN PE SEEN FF0n a 33 0. DISTANCE FROM THE TOP 0F A TO THE DECA 34 1. FRACTION OF SPRAY WATER THAT FUNS INTO PEFUELIaG POOL ssS.

t AA CONTINUING ON DIRECTLY INTO B) 1 35 1. FRACTION OF WATEP DRAINING OUT OF PEFUELING F0OL THAT l AA GOES INTO LOWER COMPI 8.VS INTO CAVITY) l 36 1.034D6 FAILURE FRESSURE OF CONTAINMENT .

37 1. ENTER A 1 IF CONTMI FAILS IN A: 0 FOR FAILUFE IN D 38 1.D0 EQUIVALENT APEA 0F FAILUkE 39 0.D0 EGUIVALENT AktA IO CALCULATE CONTAINnENT N0nINAL LEAAAGE 40 0.D0 MASS OF WATEP IN NEUTFON SHIELD BAGS--wHEN BAGS FUPTUPE AA .

THEY DROP THEIR CONIENT3 INTO REFUELING F00L 41 1500. SEDIMENTATION AREA AA AA ALOWER COMPARTMENT (BCOMPT) 01 12.56 DISTANCE FROM FLOOR TO TOP OF COMPARTMENT 02 382.0 AkEA 0F CORIUM POOL (TYP INSIDE CRANE WALLS 03 .152 HEIGHT OF CURB ON ELOOR t0VER WHICH WATER OVERFLOWS TO C) i l

l

A-2 04 382 CHARAC. CROSS-SEC AREA 0F COMP FDP BuPN TIME CALCtLATI:N 05 1.21E4 V0 Lune 06 11. VERTICAL DISTANCE Ekun THE CAVITI BtF ASS FLGW A&EA AA (EG AREA AkCUND VESSEL NO2*LES buT 5EE DEFINITiuN AA IN CAVITY SECTION BELOW) TO THE CEN!ER OE IME CAv1Tr END AA DE THE TUNNEL FLOW AREA 07 15. DISTANLE Ekun THE ELOUR OF A TV TWE CPENING ER0n B INTO D

- NOTE THAT CORIUn IN B IS ASSUnED TO SEE ONLt ONE EACE DE THE INNEk WAL 08 0.D0 EOR CASES WHERE THE 00TEP B00NDARI GE CONTnt IS A STEE; irELL AA SEPERATED FF0m A CONERETE SHIELD WALL AA THE TWO AND TPEAT THE STEEL SHELL LINER AS =, ENTER

'AC0rPI LISTANCE PE!

AND :C:nP*

- OUTEP WALLS '--ENTER 'J OTHERWISE 09 1165.0 AREA CE OUTER WALL 10 0.0 00TER WALL LINER THICKNESS 11 0.0 GAP RESISTANCE CE OUTER WALL LINER 12 0.3048 THIChNESS OE GUTER WALL 13 1.J8 THERMAL CUNDULTIVITY OE OUTER WALL 14 870.6 SPECIEIC HEAT OF OUTER WALL 15 2291.0 DENSITY UE OUTEk WALL 16 0 ENTER A 1 IF THE OUTER WALL IS SOLID STEEL

17 1370. HALF SURFACE AREA OE INTERIOR WALL 10 0 INNEk WALL LINEk THICLNESS 19 0 GAP RESISTANCE OF BUILDING INNER WALL LINER 20 .3048 THICKNESS OF INTERIOR WALLS 2 21 1.38 IHERMAL CONDUCTIVITY OE INNER WALLS 22 878.6 dPECIFIC HEAT OE INNEk WALLS 23 2291 DENSITY OF INNER WALLS 24 545. AREA 0F FLOOR (USE WATER POOL AREA IF LESSs b *.hbbb bah hESkShkNfE bh hfbOR LINER 27 1.067 THICKNESS OF ELOOR 20 1.38 THERMAL CONDUCTIVITY DE FLO0k 29 878.6 SPECIFIC HEAT OE ELOOR 30 2291 DENSITI 2? ELOOR 31 1.517D6 MASS OF E0PT 32 1.48D4 AREA 0F E0PT AAQUANTITIES 33-36 ARE USED EOR ALL WALLS WHERE APPLICABLE 33 50.0 HEAT TRANSEER COEEFICIENT ON QUISIDE WALLS 34 294 TEMPERATURE ON OUTSIDE OF WALLS 35 0.D0 ERACIl0NAL AREA AVAILABLE FOR REVERSE FLOW ON B-I ELOWPATH AA COMPARED TO THE FORWARD DIRECTION (EG DUE TO ICE

- CONDENSER DOOR (S) SHUTTING)--THIS NO. MUST A* BE NONZERO AND POSITIVE E0k ICE CONDENSER PLANTS 36 1.00 FRACTIONAL AREA AVAILABLE FOR REVERSE FLOW 09 A-D ELOw?ATH

- (EG AIR RETURN FAN ELOW DAMPERS)--ENTER 1 IE NO DAnPER 37 10. AG ELOW AREA EROM B INTO D 38 100. AG ELO6 AREA EROM B TO A 39 0. NUMBER DE IGNITERS / IGNITION SOUFCES IN B 40 0 AVG DISTANCE OF THESE FR0n IHE CEILING OF B 41 16. HE!GHTOFELOOk0FBABOVErLOORDEC 42 400. SEDIMENTATION AREA ACAVITY (CCOMPT)--INCLUDES ALL THE VOLUME BELOW THE REATOR NO~~LES INSIDE A*THE BIOLOGICAL SHIELD AND ALL THE VOL OUT TO WHERE THE TUNNEL SLOPES UP AA AANOTE THAT THE CAVITT HAS TWO ELOWPATHS- ' TUNNEL' REFERS TO A WATER AAAND PERHAPS CORIun ELOW PATH THAT ENTERS NEAk THE FASE OF THE CAVITY:

THIS COULD BE THE AA'BfPASS' REFERS TO A ELOWPATH HIGHER IN THE CAVITY; PLANTS BLOWOUT PANELS AAAREA AkOUND THE RV N0ZZLES: Ok IN THE CASE DE SOME 1 AAHIGHER IN THE CAVITY--bfPAbS AREA IS ASSUMED TO EMPIf INTu B AA AAIN SOME PLANTS WATER CAN FLOW DOWN FROM THE REEUELING POOL TO THE CAVIT!

AAAND IN SOME CORIUM CAN BE ENTRAINED UP TO THE UPPER COMPARTMENT AROUND AATHERVANNU(US--ATPRESENTGASISNOTEXCHANGEDBETWEENCANDAHOWEVER

A-3 ,

l ff!NMANYSEQUENCES.NAT. CIRC.ISSETUPWHEREBrCOLDGASENTEPSTHE AACAVITY THROUGH THE TUNNEL, IS HEATED BY PASSING OVER C0klun, AND LEAVES AATHROUGH THE BYPASS AREA 01 0.5 FLOW AREA THAT BfPASSES TUNNEL LEG FLOW AREA THAT PASSES AA. THkOUGH THE N022LES NEAR THE RX. VESSEL FLANGE) 02 37.07 AREA 0F CAVITY POOL--THIS INCLUDES KE1WAT ETC WHERE APPLI 03 37.07 CHARAC. CROSS-SEC AREA 0F COMPT FOR BukN TIME CALCULATION 04 3.7 HEIGHT OF VESSEL AB0VE BOTION OF CAVITY 05 5.92 TOTAL INSTRUMENT TUNNEL CkOSS-SECINL AkEA 06 12 LARGEST CHARAC CROSS-SECINL AREA THAT CORIUn MUST AA TRANSVERSE ON ITS WAY TO THE OPENING WHERE IT MAY BE AA ENTRAINED OR FLOODED TO CCMPTS A OR B--IN PLANTS WITH AA BOTTOM HEAD PENETRATIONS THIS WILL TYPICALLY BE THE AA 'KEYWAr* AREA (USED TO CALCULATE ENTRAINMENT VELOCITT) 07 217 CAVITY VOLUME 08 3.12 HEIGHT OF TOP OF TUNNEL ABOVE' CAVITY FLOOR'(MEASUkED AT AA CAVITY END OF THE TUNNEL IF IT SLOPES) 09 45.4 AREA 0F CAVITY OUTER WALLS 10 0.0 LINER THICKNESS 11 0.0 ~ LINER GAP RESISTANCE 12 1.068 THICKNESS OF WALL (OR DEPTH CREDITED FOR HEAT TkANSEER) 13 1.38 THEknAL CONDUCTIVITT 14 878.6 SPECIFIC HEAT 15 2291. DENSITY 16 0.- NUMBER OF IGNITION SOURCES IN C 17 0. AVG DISTANCE OF THESE FROM THE CEILING

18 37. SEDIMENTATION AREA 1 ACONCRETE AAFOLLOWING QUANTITIES ARE USED FOR ALL CONCRETE DECOMP CALCS.-

01 .6 n0LECULAR WEIGHT OF CONCRETE 02 5)00.

1s MELTING TEMPERATURE OF CONCRETE 03 61.86 MASS / UNIT VOLUME OF FREE WATEk 04 45.8 MASS / UNIT VOLUME OF BOUND WATER 05 485.0 MASS / UNIT VOLUME OF C0 06 1.D6 ENERGY ABSORBED IN END5' THERMIC CHEMICAL REACTIONS

, AA DURING CONCRETE DECOMPOSITION r 07 .8D6 LATENT HEAT OF MELTING AA AA AA ACONTROL CARDS 01 1 ENTER A 0 TO USE CHEAP STEAM TABLES IN PS WHEN POSSIELE 02 1 ENTER A 0 TO USE CHEAP STEAM TABLES IN CONT WHEN POSSIBLE 03 1 RUNGE-KUTTA ORDER (1 OR 2) (2ND ORDER INACTIVATED AT -PRESENT)

! 04 29 UNIT NUMBER TO WRITE ON FOR RESTART FILES (MAIN PROG) 06 30 UNIT NUMBER TO WRITE ON FOR RESTART FILES (HEATUP) 07 100 MAX NUMBER OF RETAIN FILE OUTPUT POINTS--BEST FOR ITEn 14 AA TO BE AN INTEGER MULTIPLE OF THIS NUMBER l 08 09 UNIT NUMBER TO PUT PRI SYSTEM OUTPUT ON l

09 09 UNIT NUMBER TO PUT CONTAINMENT OUTPUT ON

10) 31 UNIT NUMBER FOR THE FIRST PLOT FILE (OTHERS SEQUENTIAL)-

11 39 UNIT NUMBER FOR SCENARIO FILE 12 150 NON-PEAK NUMBER OF POINTS CONTROL NUMBER (SEE USEk'S MANUAL) l 13 10 PEAK NUMBER OF POINTS CONTROL NUMBER 14 500 max! MUM NunBER OF PLOT POINIS THAT ARE DESIRED 15 1 ESF PUMP LINEUP IN RECIRC (1 FOR ZION, 2 FOR SE000f AH' 16 1 ESF PUMP /ACCun DISCHARGE SETUP t1 FOR ALL TO COLD LEG 5) i 17 0 ENTER A 1 FOR B AND W PLANTS 0 OTHERWISE 18 -2 ENTERA2FORFISSIONPRODUCtRELEASETOBECOMPUTED AA BY THE IDCOR/EPRI STEAM 0x!DATION MODEL; 1 FOR

l

-A-4 AA NUREG-0772 MODEL; NEGATIVE NOS. ACTIVATE THE SAME MCDEL AA AS POSITIVE NunBERS BUT ALSO TURN ON A BLOCtAGE n0 DEL

- WHICH REDUCES THE RELEASE DE NONVOLATILE FISSION PRODE AA WHEN THE NODE IS BLOCLED E0R GAS IRANSPOPT AA 4 A4 J

CORE 01 .009459 EVEL PIN DIAMETER hh hhhh . h0khEh h huEb f1NS 04 98456. INITIAL 002 MASS AAITEM 5 nUST BE ABOVE THE ELEVATION SUPPLIED FOR THE CORE SUPPORT PLATE

- THE ELEV OE THE CORE SUPPORI PLATE nuit BE GREATEP THAN THE PADIUS OF A*THE BOTTOM HEAD 05 - 3.57 ELEVATION OF BOTT0n 0F EUEL AB0VE BOTT0m 0F VE5SEL 06 7.23 ELEVATION OE IOP OF EUEL 07 3.007 TIME DE IRRADIATION 08 3.236D9 EULL POWER IN ADDITION NO MORE THAN AATHE AA20 POWS COREnAY BENODALI2ATION USED AND NO MOREADnITS UP TO 70 NODES *5 OR COLunNS THAN 7 RING A AWHATEVER NODAL 12ATION IS USED. INSERI PEAKING EACTORS 'INTO APPROPPIAT AAENThY NUMBEPS (EG SECOND RING ER0n INSIDE RADIAL PEAKING EACTOR IS AAALWAfS ITEM 32 N0 NATTER HOW MANY AXIAL N0 DES) 09 7 NunBER GE RINGS 10 10 NunBEk OE kOWS 11 0.753 AXIAL PEAhlNG EACTOR BOTT0n 12 1.195 AXIAL PEAKING EACT0k 13 1.30b AXIAL PEAKING EACTOR

. h fALPE Na CT 16 1.127 AXIAL PEAKING EAUTOR 17 1.0b0 AXIAL PEAKING EACTOR 18 0.949 AXIAL PEAKING EACTOR 19 0.720 AXIAL PEAKING EACTOR -

20 0.390 AXIAL PEAKING EACTOR TOP AAENTRIES 21-30 AXIAL PEAKING EACTORS NOT USED bb k.0h hk hhk h$kk!

33 1.14 RADIAL PEAKING EACTOR 34 1.06 RADIAL PEAKING EACIOk 35 1.05 RADIAL PEAKING EACTOR 36 0.88 RADIAL PEAKING EACTOR 37 0.66 RADIAL PEAKING EACTOR OUTSIDE 38 0.020 AREA OR VOLUME FRACTIONS INSIDE 39 0.061 AREA Ok VOLUME FRACTIONS 40 0.102 AREA OR VOLUME ERACTIONS 41 0.143 AREA OR VOLUME FRACTIONS l 42 0.184 AkEA OR VOLUME FRACTIONS 43 0.224 AREA OR VOLUME FRACTIONS OUTSIDE 44 0.265 AkEA OR VOLUME ERACTIONS 45 33000 EVEL EXPOSURE AT SCRAM (nEGAWATT-DATS/nETPIC ION) 46 .3 EUEL ' ALPHA' AT SHUTDOWN (EISSILE ISOTOPE CAFTUhES/EISS10 47 .032 INITIAL ENRICHMENT OF EUEL IN ATOM EhACTION 48 .6DO CONVEFS10N RATIO (PkODUCTION DATE DE U-;39/ABSORPT10N FAT AA IN FISSILE ISOTOPES) AT SHUTDOWN 49 .5D0 ERACTION OF EISSION POWER MADE DUE TO FISS10NS (N 0-235

$0 .42 SAEN543EbR$hbbh$

51 .08 SAME AS 43 E0R U-238 (EAST FISSIONS) 52 1.D-4 EkACTIONAL 2R02 nASS (COMPARED TO ZR nASS) AT TIME O 53 .004647- EVEL PELLET RADIUS AA AA IE LARGE. DRY SIMULATION, SPECIEY ZERO VOLUME FOR ICE-CONDENSER-INIS

e A-5

- CAUSES ALL UPPER PLENUM AND ICE CONDENSEk PAFAnETEPS TO BE v280 FED AA AA AICE CONDENSER 01 0.0 VOLUME AA AA AANNULAR COnPARTMENT (DCOMPT)

AATHIS COnPARInENI PEPRESENTS THE VOLunE BETWEEN THE CRANE WALL 'IF ANri AAAND THE CONTMT WALL AND BETWEEN THE DECL AND THE LOWEk COnPI FLOOF--

AAIFNOCLEARDISTINCfl0N. AkBITFARILY DIVIDE THE SPACE BELOW -InE UPPER

- COMPT AND USE LARGE FLOW AREAS TO LEEP THE GAS WELL nInED--AT PRESENT.

- CORIUM IS ASSUMED NOT TO GEI INTO THIS COMPARInENT 01 9.71E03 VOLUME 02 579. AREA 0F WATER POOL 03 0. DISTANCE THE FLOOP OF D IS ABOVE THE FLOOR OF B 04 579. CHAhAC. CROSS-SEC AREA 0F ConPI FOR BUPN TINE CALCULATION 05 1482. AREA 0F EXTERIOR WALLS 06 0.00954 WALL LINER THICKNESS 07 0.00882 GAP RESISTANCE OF WALL LINER 08 THICKNESS OF WALL 09 1.0b7

1. 8 THERnAL CONDUCTIVITY OF WALL 10 879. SPECIFIC HEAT OF WALL 11 2291. DENSITY OF WALL 12 0. ENTER A 1 IF THE OUTER WALL IS MADE OF STEEL 13 0. HEIGHT OF CukB SEPERATING D AND B MEASURED FROM B'S FLOOF 14 0. NUMBER OF IGNITERS OR IGNIIION SOURCES IN D 15 0. AVG DIETANCE OF THESE FROM THE' CEILING 16 1160. SEDIMENTATION AREA- ,

AA AA AUPLENUM (UPPER PLENUM Ui ICE CONDENSER) 01 0.0 VOLunE A*

AA AENGINEERED SAFEGUARDS AAIN METk!C UNITS

- FLOWRATES SPECIfIED TO BE VOLUMETRIC SHOULD FE n**3/SEC: OTHER FLOWPATES AA!E ALL THOSE NOT EXPLICIILY STATED TO BE V0LunETRIC AASHOULD BE KG/SEC: HEADS SHOULD BE IN n: PRESSURES IN PA: IN ENGLISH THE AjuNITS AkE RESPECTIVELY GPM,LBM/HR,FT, PSIA AAIN THE FOLLOWING ' FANS' REFER TO FAN COOLERS--tAIR RETURN FANS IN

. AACONDENSERPLANTS$

01 0.127 ACCUMULATOR PIPE DIAMETER 02 1.276D6 PRESSURE SETPOINT FOR LPI

, 03 1.172D7 PPESSURE SEIPOINT FOR HPI

! 04 4.24D6 INITIAL PRESSURE OF ACCunULATORS I

05 310 TEMP OF RWST l 06 325 TEMP OF ACCUMULATORS l 07 1.316D6 INITIAL MASS IN RWST l 08 2.382D4 INITIAL MASS PER ACCUMULATOR l 09 100. AREA 0F BASE OF RWST 10 18.3 LENGTH OF AN ACCUMULATOR PIPE 11 0.26D6 PRESSURE SETPOINT OF BLDG SPRAYS 12 0.152006 PRESSURE SETPOINT OF BLDG FANS 13 3 NUMBER OF OPERATING FAN COOLERS 14 VOLUMETRIC FLOW THROUGH A FAN COOLER 15 1.0D-3 NOMINAL DIANETER OF COnPARTMENT SPPAY DROPLETS 16 30.23 VOLUME OF ONE ACCUnULATOR 17 4 NUMBER OF OPERATIONAL ACCUMULATORS 18 1 NUMBEk 0F OPERATIONAL HPI PUMPS 19 1 NUMBER OF OPERATIONAL LPI PUMPS

A-6 20 5 NUMEEP OF ENTPIES USED IN kP1 PunP-wD CUPvE TAbl6*. mAi-21- 1044.5 HIGHEST HEAD IN TABLE (UNITS ARE METERS:

22 925.0 NEXT HIGHEST HEAD IN HPI PUNP-HEAD CUFVE !AELE 23 833.5 NEXT HIGHEST HEAD IN HPI PUNP-MEAD CURVE table 24 615.5 NEAT HIGHEST HEAD IN HPI funP-HEAD CUh.'E TABLE '

25- 0.0 LOWEST MEAD IN HPI PunP-NEAD CURVE TABLE A* ALL ELOWhATES ARE IN n**3/SEC 26 0.0 ELOWRATE C0kESPONDING TO EIPST ENTRY IN PRESSUPE TAELE 27 3.43D-3 NEXT ELOWPAIE 28 1.37D-2 NEXT ELOWRATE 29 2.170-2 NEXT FLOWRATE 30 3.75D-2 NEXT EL0wkATE 31 5 NunBER OF ENTRIES USED IN LPI TABLE 32 119.6 HIGHEST HEAU IN LPI TABLE 33 112.5 NEXT HEAD 34 102.0 NEXT HEAD 35 59.8 NEXT HEAD:

36 0.0 NEXT HEAD 37 0.0 - FIRST ELOWRATE IN TABLE 38 6.86D-2 NEXT ELOWRATE NEXT ELOWRATE 39 40 .}.14D-1 06D-1 NEXT ELOWRATE 41 2.74D-1 NEXT ELOWRATE 42 12.62D6 CHARGING PUMP PRESSUPE SETPOINT 43 1.0 NunBER OF WOPNING CHARGING PUMPS 44 5 NUMBER OE ENIBIES IN CHARGING PunP HEAD CUkVE TABLE 45 1617.8 EIkST HEAD 46 lb47.4 NEXT HEAD 47 1301.1 NEXT HEAD 48 879.2 NEXT HEAD 49 0.0 NEXT HEAD 50 0.0 ElkST ELOWRATE 51 4.57D-3 NEXT ELOWRATE 52 9.150-3 NtXI ELUWkATE 53 1.60D-2 NEXT ELOWRATE 54 2.56D-2 NEXT ELOWRATE 55 10 AREA 0F BASE DE CONTMT SUMP 56 2 DEPTH OF CONTMT SunP 57 1 NUMBEk OE USED ENIRIE5 1N SPkAf PUMP HEAD CUkVES #5 max)

, 58 1000

- 58 98.1 FlkST ENTRY IN SPRAT PUMP HEAD TABLE 63 1.65D-1 EIRST FLOW ENTRY IN SPRAT PUMP HEAD CURVES AA FOR NPSH TABLES THE SANE ELOWS AS WERE GIVEN E0R HEAD CUPVES AFE AAASSUMEDTOC0kkESPONDTOTHENPSHHEADSGIVEN 68 3.00 NPSH (M UNITS) RE0'D FOR CHAkGING FunP AT FIRST ELOW IN 69 3.05 NEXT NPSH ENTRY EOR CHAPGING PUnFS 70 3,05 NEXT NPSH ENTRY E0R CHARGING PUMPS 71 3.05 NEXT NPSH ENTRY E0k CHARGING PUNPS 72 3.05 NEXT NPSH ENTRf E0R CHARGING PUnPS 73 3.05 FIRST NPSH ENTRY E0R LPI 74 3.05 NEXT ENTRY FOR LPI 75 3.05 NEXT ENTRY EOR LPI 76 3.05 NEXT ENTkT E0k LPI 77 3.05 NEXT ENTRY E0k LPI 78 3.05 ElkST NFSH ENTRY EOR HPI 79 3.05 NEXT ENTRY E0k HPI 80 3.05 NEXT ENTRY EOR HPI l 81 3.05 NEXT ENTRY FOR HPI 82 3.05 NEXT ENTkY FOR HPI 83 3.05 FIRST NPSH ENTkY EOR SPRAY PUMPS 84 3.05 NEXT ENTkY FOR SPkAf PUMPS 85 3.05 NEXT ENTRY EUR SPRAY PUMPS 86 3.05 NEXT ENTRY EOR SPkAY PUMPS

-_ . -- . . - . .= ..

A-7 t

87 3.05 NEXT ENTRY F0k SPRAY PunPS 88 1 NUMBEk 0F OPEkATING SPRAY PunPS FOR UPPEk (UNFAsIFENT 89 0 NUMBER OF OPEPATING 5 PEA: PunPS F0F L0wEP :OnPARIaENT i 90' 15.2 HEIGHT OF BOTT0n 0F RWST ABOVE THE ENG 5AFE PuePS

91 7.3 HEIGHT OF BOTT0n UF CONTAIN SunP ABOVE THE ENG SAFE PPS 92 9.75 ELEVATION OF THE PV INJECTION NO:2LES AB0VE THE S1 PPS 93 163.8 FLOW THR0 UGH ONE SPRAI Pump WHEN ITEN 95 NEASUPED

.94 2.75D5 DIFFERENTIAL PRESSURE ACROSS THE SPkAt N0::LES 95 31.5 ELOWkATE OF EtTERNAL RWST REPLACEnENT WATEE. IF 'ANY 4

5.0 96 TIME DELAT FOR HP1 i

97 5.0 TIME DELAT FOR LPI

, 98 5.0 T!nE DELAf FOR CHARGING PUMPS 99 30.0 TIME DELAT FOR UPPER COMPARTMENT SPRAYS 100 30.0 TIME DELAT FOR LOWER COMPARTMENT 3PkAr5 101 5.0 TIME DELAY FOR FAN C00 LEKS 102 NUMBER OF TUBES IN A FAN COOLER

, 103 '0UTSIDE AREA 0F ALL TUBES IN A FAN C00 lek i 104 AREA 0F ALL FINS IN A FAN. COOLER 105 FAN COOLER FIN EFFICIENCY 106 FAN COOLER INSIDE FOULING FACTOR 107 FAN COOLER FIN DIAMETEk

, 100 FAN COOLER IUBE THICKNESS

109 FAN COOLER TUBE THEkmAL CONDUCTIVITY 4

110 MINIMUM FLOW AREA THROUGH FAN COOLER 111 FAN COOLER TUBE ID 4 112' 5 NUMBER OF NODES USED TO MODEL FAN COOLER (5 MAX) 113 INLET COOLING WATER TEMP TO FAN LOOLER

114 INLET COOLING WATER ELUW TO A FAN COOLER 115 1 NUMBER OF LPI PUnFS USED FOR RHR SPRAYS WHEN VALVE OPEN i -116 0 ENTER A 1 IF FANS / COOLERS DISCHARGE TO B;0 TO D AAESE HX'S

, AACALCULATIONS CONTROLLED BY HEAT EXCHANGER ItPE i

ER TYPE t AA -1 AAHEATEXCHANihTOUILEf'TEMPOFHXTORWSTTEMPERATURE-S

A 0 IS N0 HX--0UTLET TEMP IS CONTMT SUMP TEnP AA 1 STRAIGHT TUBE HA .

! A* 2 U-TUBE HX

! AAFOR TYPES 1 AND 2 EITHER SUPPLY ALL GE0 METRIC PARAMETERS A*0R THE NTU (NUMBER OF IRANSFEF UNITS) PER HX A*LATTER IS AVAILABLE BY CONSULTING NAnEPLATE DATA AND USING GRAPHS

. **IN FOR EXAMPLE HOLMAN, HEAT TRANSFER

! A*AL(PARAnETERSdREONAPEkHXBASIS

- 117 1.D0 TYPE OF HX FOR SPRAY

118 0.00 NUMBEk 0F TUBES IN SPEAT HXS f 0.b0 SfRAfH ThBhkD i 121 0.00 SPRAf HX TUBE THICKNESS '

l 122 0.00 TUBE TO TUBE SEPARATION IN SPRAf Hx h.*hh HkhMAbhhNUCI 0F SPRAY HX TUBES 125 0.D0 LARGEST PERP DISTANCE FROM SHELL TO BAFFLE-(' BAFFLE CUT')

126 0.00 SHELL TO TUBE CLEARANCE AT OUTSIDE OF SPRAT HX TUBE BDL i 127 311.9 SPRAI HX COOLING WATER ELOWRATE '

AA128 NOT USED 129 1 TYPE OF HX FOR kHR 130 0.D0 NUMBER OF TUBES IN RHR HXS 131 0.D0 NUMBER OF BAFFLES IN RHR HXS i 132 0.00 TUBE ID IN RHR HXS 133 0.00 TUBE THICKNESS IN kHk HXS 134 0.00 TUBE TO TUBE SEPERATION IN kHR HXS 135 0.D0 SHELL LENGTH IN RHk HXS 136 0.00 TUBE THERMAL CONDUCTIVITY IN RHR HXS 137 0.D0 BAFFLE CUT DISTANCE IN kHR HXS (SEE 125s l

. . - . - -- - - . - - . . . - . - - . . -. . . - .. .~.

~

JA-8 L

.138 0.D0' SHELL TO Tube CLEARANCE AT OUTSIDE CF PHR. HX TUBE kUNDL" 139 311.9 RHR HX COOLING WATER FLOWRATE

, =140 0.999 SPkAt HX NTO-3' 141 0.989 kHk HA NTU 142 0.D0 SHELL ID OF SPkAY PECIRC HX

143 0.D0 SHELL ID DE kHk kEclkC Hx. <

. AAENTER ZERO V0 Lune FOR -! TEM 148 IF NU UNI SYSTEM A*144 INITIAL MASS IN THE UHI WATER ACCunULATOR ,

AA145 LENGTH OF THE UNI PIPE TO THE RV AA146 DIAeETEk 0F THE UHI PIPE A*147 INTIAL PRESSURE OF THE UNI ACCUMULATOR 148 ' O.D0 TOTAL

WATER + GAS) V0 Lune IN THE UHI ACCUnULATORS AA149 FAILURE PRESSUkE OF THE UHI PIPE puPTURE DISK

150- 0.00 MASS =IN THE CAVITY INJECTION SYSTEn < !F ANf) '

151 0.D0 FLOWRATE OF THE CAV INJ SYSTEM WHEN ACTIVATED

. AA USER HAS THE OPTION TO THROTTLE ESF SYSTEMS AT LESS THAN-1

AA THEIk FULL FLOW GIVEN THE CONDITIONS EXISTING--TO DO THIS AA ENTEkFORTHEAPPROPRIATESYSTEM(ANDFORTHEAFW-INTHE$Tn-GENERATOR SECTIONS A TOTAL FLOWRATE DESIRED

THE CODE WILL.USE i AA

( AA. THEMINIMun0FTHISFLOWANDTHATCALCULATE6FROMTHE'HEADCURVES AA AA' AND THE NO. OF OPERATIONAL PURPS'IF OPERATOR ISN'T ENTERALAkGENO.:!EMECHANGES.fHEDEGREEOFTHROTTLING, CARDSTHROTTLIN -

i AA' PARAMETER CHANGES USING INTEkVENTION NO.-1000 IN CONTROL j {.{g {gggTjggb{LOWFORLPISfSTEM(TOTAL)$j 14 1.E6 SAMI FOR CHARGING PUMPS i 155 1.E6 SAME FOR UPPER COMPT NORMAL' SPRAYS 6- AkE OWRCOMh AS u j ** l Ah AINITIAL CONDITIONS ' 1- 01 577 NOMINAL FULL POWER PRIMARf SfSTEM WATEk TEMPERATURE 4 02 15.6206 NOMINAL FULL POWER PkIMAkY SYSTER PRESSURE i- 03 8.55 PRESSUkIZER WATER LEVEL (ABOVE BOTTOM) 04 .101D6 CONTAINMENT PUILDING PPESSURE $ bk h0.0 khE bChhEk$ k hEkPINkkbhE 07 1. LOWER CONTAINMENT DUILDING COMPARTMENTS REL. HunILITY (0-a 08 0 INITIAL ICE MASS kb $2 kN AL EkfEh!Tbhk Oh h0NE RN ET L STRUCTURES

                                .                        11      5.066D6                 INITIAL PRESSURE ON SEC. SIDE OF S/G'S -

4 - la 1.00 UPPER COMPARTMENT REL HUMIDITY 40-1# 4 INITIA IMAk S EM W TM RATURE

 !.                                                      15       15.62D6 INITIAL PRIMARY SYSTEM PRESSURE -                                                                             '

i ** , AA i i APRIMARY SYSTEM l 'AAUNLESS OTHERWISE NOTED ALL ELEVATIONS IN THIS SECTION SHOULD BE

 !                                                       AAREFERENCEDTOTHEBOTT6M0FTHERVHEAD

! 01 4 NUMBEk 0F COLD LEGS i 02 .0D0 DIAMETER OF A HOT LEG PIPE' l 03 2.24 RADIUS OF PkIMARY HEAD 04 2.99 ELEVATION OF SUPPORT PLATE

i. 05 4.78 FLOW AREA 0F CORE +9YPASS AREA

}                                                        06       3.1                     VOLUME OF HORIZONTAL RUN OF PIPE IN COLD LEG OUT TO
 !                                                        AA                              PUMP 90WL j                                                          07      .0192                   RADIUS Of VESSEL PENETRATION--IE NO VESSEL PENETRATION AA                              (EG SOME CE PLANTS) USE ASSUMED RADIUS OF FAILURE WHEN HEAD l                                                         08      3.175D6                 ENERGY INPUT FROM A PRIMARY SYSTEM PUMP (WHEN RUNNING) t                                                          09      0.                      TOTAL MAKEUP FLOW TO THE' PRIMARY SYSTEM--UNDER NORMAL i

4 J

 . - - . . - . ~ . , . , - , . . - . , - - - - _ _                     .._m,---.--,.-..-v                  -.---..----.,_--.~-,.--.-----.~#                             - - - - -

A-9 i AA OPERATION SHOULD liOUAL LETDOWN ELOW BELOW 10 326. TEMPEkATukE DE mA/EUP WATER 11 .7 DIAnETER OF A COLD LEG PIPE 12 9.00 ELEVATION OF SURGE LINE Ar0VE BOTT0a CE kV 13 4 ENTER FREAK LOCATION FEr:

   **             1--COLD LEG NODE (PunP FOWL Ok VEPTICAL 5ECTION LEADINu AA                 TO S/G OR S/G COLD LEG SIDE 'SEE USER'S mANUAli AA             2--ALL COLD LEGS tEG PunP SEAL LOCA--ENTEF TOI BK AFEe
   $$             3-bbWN SEh NODE (INClubEb H0kIZ PART DE CULD LsG5 AA                 PUNNING OUT TO THE PUMP BOWL AA             4--HOT LEG NODE 14   0.61      PRIMAkY SYSTEM BkEAK AREA (COLD uk HUT LEGS) 15     9       ELEVATION OF BEEAK (THIS SHOULD bE CUNSISTENT WITH NO.13, 16             VOLUME IN A COOLANT LOOP WHICH IS UNDER A LINE DRAWN Tas0uGn AA             THE BOTT0n 0F A COLD LEG NO2ZLE- 'PunP POWL' VOLUME IN A AA             WESTINGHOUSE PLANT 17             MAA VULUnE OF WATEk [N A COLD LEG E0k GAS TRANS IO OCCUR 18             VULunE OF A COLD LEG 19             V0 Lune OE A HOT LEG 20             ELUID VOLUME OF THE RX VESSEL AA21          GAS ELOWRATE OE RX HIGH POINT VENT (S) AT N0n SIS PPESS AADOWNCOMER IS MODELLED AS ENDING AT THE POINT WHERE THE LOWER HEAD AA0F THE kV MEEIS THE CYLLINDRICAL SECTION 22             VOLUME DE DOWNCOMER 23             VOLUME Of DOWNCOMEk BELOW THE COLD LEG N0!!LE 24       0     ENTER A 0 E0R P!R TO BE IN UNELN LOOP; 1 TO BE IN BROKEN AA             LOOP hh     Of50    Uh$bhkbf$hNA        HICH RCP'S TRIP Ok EAIL 27   1.25D7    LOW PRESSURIZER PkESSURE TRIP POINT 28   1.65D7    HIGH PRESSURIZEk PRESSURE TRIP POINT 29   100       HIGH LOOP DELTA-T SCRAn SETFOINT 30 -100.0      LOW PkESSURIZER LEVEL TRIP (EICTIONAL) 31   13.11     HIGH PRESSURIZEk LEVEL TRIP 32   2.0       REACTOR TRIP DELAY TIME 33   -1.00     LOW S/G WATER LEVEL SCRAn SETPOINT 34   4         NunBER OF POINTS IN MCP COAST-DOWN CURVE (5 MA^t) 35   4259.00   FIRST MASS ELOWFATE IN MLP C0AST-DOWN CURVianUST Bd THE AA             NOMINAL ONE PunP ELOW) 36   2011.D0   SECOND ELOWEATE 37   2044.00   NEAT ELOWRATE 38   1363.00   NEXT ELOWRATE AA39           NEXT ELOWRATE 4

40 0.00 FIRST TIME IN C0AST-DOWN CURVE 41 t.00 NEXT TIME 42 10.D0 NEXT TIME 43 20.D0 NEXT TIME AA44 NEXT TIME 45 10.00 ELEVATION OF S/G TUBESHEET ABOVE BOTT0n Of kV i 46 .15 THICKNESS OF RV HEAD 47 8.33 DISTANCE FROM BOTTOM OF N0ZZLES TO BOTTON OE kV HEAD Ak hhfbMbEThEbhfbbEkN E 49 VOLunE DE THE HORIZONTAL kUN OF A HOT LEG PIPE 50 0 TOTAL LETDOWN ELOW 51 .2E6 NORMAL DELTA-P EROM CORE INLET TO HOT LEG SIDE OF OUILET AA N0!!LES WHEN MAIN COOLANT PUMPS ARE ON AA AA APRESSUkIZER 01 50.97 PRESSURIZER VOLUME 02 PRESSURIZER CROSS-SECTIONAL AREA

                                                            -4.
  }n                                                    $                                4 A-10 03 15.6206              PkESSURIZER aEATER PRESSURE SETPOINT 04 15.68D6              PRESSuklIER SPkAt PRES 5UrE SETPOINT 05           2.38       WATER LEVEL AT WHICH HEA!ERS TPIP 06           1.806      HEATER OUTPUT 07           38.9       SPhAf SYSTEn ELOW RATE 08          52.92       FLOW RATE OF SAEETr VALVE AT ITS SETPOINT 4

09 17.24D6 LOWEST SETPOINT OF A SAFE!! VALVE 10 17.24D6 HIGHEST SE! POINT OF A SAEEf f VALVE 11 0.284 DIAnETEP 0F THE SURGE LINE 12 14.26 ELEVATION OF SPPAT HEAD AbOVE BOTT0n OE P2h 13 19.5 LENGTH OF THE SURGE LINE 14 3 NunBEE OF SAFETY VALVES 15 1.0D-3 N0nINAL P!h SPRAT DiOPLET 16 16.206 LOWEST SET POINT OF PORv 17 16.206 HIGHEST SET POINT OF P0kV 18- 2 NunBER UF P0kVS 19 26.46 N0nINAL ELOWRAIE UE A PORV AT ITS SETPOINT 20 1./Ds EMPTY MASS UE PZk 21 0.00 ENTEk A 1 IF THE SUkGE LINE HAS A LOOP SEAL 22 J.5 SEDIMENTATIUN AREA AA AA ASTEAM GENERATOR (VALUES REEER TO ONE UNIT) 01 166.2 SEC SIDE VOLunE 02 DOWNCOMER CROSS-SECTIONAL AREA 03 TUBE BUNDLE (SECONDARY SIDE) ELOW AREA 04 0.D0 B AND W ONLf--ELEVATION OF AUX FEED SPRAY HEAD ABOVE AA LOWEk TUBESHEET 05 1.D6 INITIAL MASS IN CSI--Ok A LARGE NO. IF NO Lin!T ON

'         AA                    AEW SUPPLY 06                    N0kMAL OPTN 2-PHASE WATER LEVEL IN TUBE BUNDLE 07       493.3         MAIN EEEDWATER TEMPERATURE 00       7.34306       LOWEST SETPOINT OF SEC SAFETY VALVES 09       7.68806       HIGHEST SETPOINI OE SEC SAEETY VALVES l

10 5 NUnBER OF SAEETI VALVES PER S/G ' 11 88.2 NOMINAL ELOWRATE OF A SAFETY VALVE 12 7.24D6 SETPOINT OF SEC RELIEF VLV (ASSunED SAME FOR ALL RELIEFS) A*lF NO 'RELIEE VALVES'--SUPPLf A SET POINI PRESSURE HIGHER THAN THE AASAFETIES AND USE THE RELIEES AS MANUALLt CONTPOLLED STEAM DUMPS 13 1 NunBER OF RELIEF VALVES PEP S/G 1 14 112.1 NOMINAL ELOWRATE OF A RELIEF VALVE

15 661.0 nAX EERDWATER ELOWkATE PER S/G 16 9. PRIMARY HEAD VOLUnE 17 10.0 TIME DELAT EOR ACTIVATION OF Aur FEED 10 IIMEDELAYTOJ HUT NSIV i- 19 *301.2 I0TAL FRIMAki iDE V0 Lune 20 100.00 nAX AUX EEED ELOWkATE ,

l 21 310.D0 AUX FEED TEMPERATURE ' 2 3388 NunBER OF IUBES IN A STEAM GENEhA10s

           '33 0.00127 THICKNESS DE STEAM GENEhAT0k TUBE

. 24 0.0197 ID OF STEAM GENEkAT0k TUBES 25 18.4 THEFnAL CONDUCTIVITY OF STEAM SENERAT0k TUBE 5 26 1.E6 THROTTLED ELOW EOR AFW SfSTEM PEP STEAn GENERATOP OR A LAEGE AA NunBER IF ELOW NOT THR0 TILED LSEE DISCUSSION AFTEk ENGIN. AA SAFEGUARDS ITEM 151) l 27 1. ERACTIONAL AkEA USED E0k STEAM DUnPS IN BLN LOOP SiG 28 1 FRACTIONAL AREA USED EOR STEAn DUMPS IN UNBhN LOOP S/GS 29 16.6 00WNCOMER PROGRAM Wik LVL EOR SGWLC SYSTEn IN b N LOOP S/G 30 10.6 DOWNCnk PROG WTR LVL FOR SGWLC SYSTEM IN UNBhN LOOP SiGS AA31 STEAM GENERATOR TUBESHEET DIAMETER AA .i AA ATIMING DATA t

     , e    -   r          ,

w w .

A-ll 03 20.0 MAX TIME STEP l 04 .005 MINIMUM TIME STEP 05 .05 RELATIVE MASS CHANGE USED TO SELECT TIME 3TEP 06 .02 REL OXIDATION FILn THICKNESS CHANGE USED TO SELECT TIME i 07 .02 REL GAS TEMPERATURE CHANGE USED TO SELECT !!ME STEP 08 .1 REL MASS CHANGE FOR FISSION PRODUCTS USED TO SELECT TIME A* STEP IN FISSION PRODUCT POUTINES AA A* AQUENCH TANK 01 50.97 VOLUME 02 2.02D4 INITIAL WATER mA33 N AA05 IN Nfk0bkhThE SEDIMENTATION AREA K VE BCOMPT FLOOR AA MODEL PARAMETERS ' 01 .005 C0kIUM EklCTION 00 EFFICIENT 02 .5 FRACTION OF CORE MASS THAT MELTS bet 0Ri CORIUM PILE N 60. hkE 0 AihVkbbL E ETER CONTACT WITH CM 04 120. TIME TO FAIL SUP. PLATE AFTER CM PILE HAS REACHED IT 05 2.0 MULTIPLIER OF NOFMAL CLAD SURFACE AREA TO ACCOUNT FOR AA POTENTIAL CLAD FUP!URE (MUST BE BETWEEN 1 AND 2) f' 06 983.D0 CRITICAL FLAME TEMP AT 2Ek0 STEAM MOLE EkACTION AA { USED IF NO IGNITION SOURCES: THIS IS MULTIPLIED Bf THE l AA WESTINGHOUSE FLAME TEMPERATUPE MULTIPLIER CORRELATION l N 08 2300 300.0 C O!kNIbu -b CHANNEL BLOCKING TEMP NON-RADIATIVE FILM B0IL. HT TRANS COEFF FROM CM TO POOL 09 850. NAT. CIRC. (MCP'S OFF) S/G PRIMARY SIDE FILM RESISTANCE 10 .1 REFERENCE THERMAL BOUNDARr LATER THICKNESS IN CONCRETE 11 .1 B AND W ONLY: ERACTION OF S/G TUBES STRUCK BE AEW 12 1.D3 HT TRANSFER COEFF BETWEEN MOLIEN CORIUM AND FR00EN CRUST l 13 0.D0 ENTER A 0 FOR ENTRAINMENT EkOM C 10 B' 1 FOR C TO A 15 1.D0 hE Ibl hE6 ubfEktkk Wk Abb bA DRAG COEFFICIENT OF RISING PLUME DURING BURNS IN A COMPT 16 1.00 SAME F0k B COMPT 17 i 1.D0 SAME FOR C COMPT I 18 1.00 SAME FOR D COMPT ' 19 1.D0 SANE FOR U COMPI 20 1.53 CHURN-TURBULENT CRITICAL VELOCIT1 COEFFICIENT 21 3.7 DROPLET FLOW CRITICAL VELOCITY COEFFICIENT 22 1. SPARGED POOL VOID FRACTION COEFFICIENT 23 2. VOLUMETRIC STEAM GENERATION VOID FRACTION COEFFICIENT 24 .5 ENTRAINMENT TIME CONSTANT si :15 14111101118f Mi 27 .85 EMISSIVITY OF EQUIPMENT 28 .85 EMISSIVITY OF CORIUM SURFACE 29 .6 EMISSIVITY OF GAS 30 .3 CURE HfDkODTNAMIC LIMIT KUTATELADIE NO. FOR hiFL00 DING HT AA AND OXIDAfl0N CALCULATION 3 31 .3300 NUMBER TO MULTIPLf KUTATELAD2E CRITERION Bt TO REFFESENT AA AA DIFFICULTY (GT 1.00) OR EASE (LT 1.D0) 70R MATERIAL TO GE' OUT OF CAVITY 32 3.0 FLOODING CRITICAL VELOCITY COEFFICIENT 33 .14 FLATE PLATE CHF CRITICAL VELOCITY COEFFICIENT b b5 D RE E kCNb IMA SNMBREAK 36 1.00 SCALE FACTOR FOR BC)N VELOCITY CORRELATION

A-12 37 1.D0 SCALE FACTOR FOR HEAT TkANSFER COEFFICIENTS 70 PAS 31VE AA HT SINKS

 **38-42 ARE PARAMETERS IN THE GRAVITATIONAL AGGL0nERA! ION AND FALLOUI
 ** CORRELATIONS IN SUBROUTINE FFTPANS h       *bhikb0 sob!khbkkR bF Exhkkkkkb k0 $0b0kk$ff'Akk0h5 40      .5900     DENSITt EtPONENI FOR HIGH-DENSITY AEFOSOLS 41      .3300     DENSITY EXPONENT FOR LOW-DENSITI AEF050LS 42      1.D-5     DENSITY AI WHICH WE SWIICH FR0n THE HIGH-DENSITr A*                EXPRESSION TO IHE LOW 43      1.D0     ABS (nULTIPLIEki uh CSI AND C50H VAP0k PFE55URE--ENTEF A*                                                       POS FOP SANDIA 44      .100      A NEG NUMBER FRACTION        TOGsIDIZED OF CLAD  SELECT JANAF CSUH VP;S WHICH CAust   OPE TO COLLAPiE ON AA                REFLOOD (GIVES SMALLER tu FOR HEAT IPANdFEP THAN INTACT AA               MODEL) 45     0.00       FUk B AND W UNITS ONLY ERACTION OF PERFECT CONDENSATION AA                0FSTEAnENTERINGDOWNdCMERTHROUGHFLAPPEkVALVES AA AFISSION PRODUCTS AAFISSION PRODUCT GROUPING SCHEnE:

AAGROUP 1: NOBLE GA5SES AND ' INEPT' (NON-PAD 10 ACTIVE) AEROSOLS AAGROUP 2: CSI A* GROUP 3: TE02 AND TEH AAGROUP 4: SRO (BA LUMPED IN) AAGROUP 5: RU (n0 LUMPED IN) AAGkOUP b: CSOH AA

 *ASTRUCTURAL MATERIAL G50VPING SCHEME A AUSED IN CORE N0 DES (TRACT.ED IN CONTAINMENT AS GPOUP 1 AEROSOLS)

AAGROUP 1: CD AAGROUP 2: IN AAGROUP 3: AG AAGROUP 4: SN AAGROUP 5: nN AA 01 .028 FRACTION OF FISSION PRODUCT POWER IN GROUP 1 02 .151 SAME FOR GROUP 2 03 .0194 SAME FOR GROUP 3 AANOTE: CALCULATIONS USUALLY SHOW VERY LITTLE SR OR RU RELEASE--T0 ENSURE kk$kb! Ib NEkk kkhuhdI- E N kESUIhLbBEkb$bfkEfLUL I AAAS IF THE SR AND kV WERE LEFT IN THE DEERIS THE CODE WILL CONTINUE TO AACALCULATETHEIRMASSTRANSPORTASIFTHEYC6'ULDBERELEASED,HOWEVER. AASO THAT THE ACCURACY OF THIS APPRO)IMATION CAN BE CHECKED , 04 0. SAME FOR GROUP 4 (ACT .062) 05 0. SAME FOR GROUP 5 (ACT .05471 06 .011 SAME FOR GkOUP 6 07 337. INITIAL nASS OF FISSION PRODUCTS IN GFOUP 1 (NOBLES ONLt ' 08 29. INITIAL MASS IN GROUP 2 09 30.8 GROUP 3 10 135. GROUP 4 11 319. GROUP 5 12 147. GROUP 6 13 140. INITIAL MASS OF CD IN CORE (STRUC MAIERIAL 6kDUP 1) 14 419. INITIAL MASS OF IN IN CORE 15 2232. INITIAL MASS OF AG IN CORE 16 347. INITIAL MASS OF SN IN CORE 17 196. INITIAL MASS OF MN IN C0kE 18 6. NO. OF MAXIUM GROUP 100 WISH TO MODEL LNORMALLY 6e AA ACIRC AATHIS SECTION IS FOR INPUIS TO THE DETAILED PRIMART SISTEM T/H

A-13 A*n0DULE SUBROUTINE CIRC AND ITS ANCILLARY SUBROUTINES AA A*TO DEVELOP INPUTS FOR THIS SECTION CONSULT THE APFROPRIATE FIGUPE A*FOR TOUR PLANT TO DELINEATE NODAL $0VNDAPIE5 AAINSERI NunBERS IN THE SPACES SHOWN' IHE CODE COnPUIES THE MIS 5ING AANUMBERSFROMTHESEANDOTHERINPUT5 AA AAIN GENERAL THE NODES ARE SUFFICIENTLT COARSE AND THE HEAT TPANSFER IS AASO GOOD WIIHIN A N0DE L AT HIGH TEMPEPATURES AT LEASTi THAT ONE SHOULD AAWORRt MORE ABOUT GETTING THE TOTAL AREA AND, MASS COPPECT THAN WORPtING AAA LOT ABOUT THE DETAILS OF WHEkE THE BOUNDAkIES BETWEEN AND WITHIN AAN0 DES ARE AA AAAAAAAAANOTE SPECIAL DEFINITIONS FOR SOME ENTRIES FOR B AND W NODALI ATION AA CIRC ALLOWS EACH NODE TO HAVE I OR TWO MASSES; EACH N0DE HAS A ' STEEL' AA MASS AND MAY IN SOME CASES ALSO HAVE A ' HEAT SINH' MASS

THE HEAT SINH AA IS DISTINGUISHED FRan THE STEEL MAINLY Br TWO DIFFERENCES:

AA 1. nAY BE AT A DIFFERENT TEMPERATURE' AA THIS nAf BE DUE IN PART TO THE HEdT SINF HAVING LOSSES TO CONTMT

  **              WHEN JHE STEEL DOESN T (EG 5/6 SHELLS VS TUBES)
  **          2. HEAT SINKS ARE ASSUnED NOT TO HAVE FISSION PRODUCTS
  *A              PLAffD ON THEn AA AA AT PRESENT, IHE ENERGY EXCHANGES IN A NODE nAt INCLUDE ONE Ok n0RE OF DEFENDING ON THE INPUT PARAMETERS SUPPLIED:

AA THE 1.FOLLOWINuf AA AA HEA SINK AND STEEL EXCHANGE ENERGY RADIATIVE A5SUMED TRANSPARENT) AA 2. STEEL EXCHANGES ENERGt WITH PRIMARY StSTEn GAS VIA INTER-NODAL AA OR INTRA-N0DAL NATURAL CIRCULAI!ON , AA 3. HEAT SINK MAY EXCHANGE ENEPGr WITH PRIMARY SYSTEM GAS VIA AA INTER- OR INTRA-NODAL NATURAL CIRCULATION AA AA ITEMS 2-4 ARE HT AREAS COUPLING THE 2 NODAL HEAT SINK MASSES (STEEL AND AA HEAT SINK)

02 LIMITING AREA FOR RADIATIVE HEAT EXCHANGE BETWEEN UPPER AA PLENUM EQUIPMENT AND THE AD3ACENT RV SHELL 03 ENTER HALF THE AREA 0F THE S/G SHELL (B AND W: ENTER 0) 04 ENTER HALF THE AREA 0F THE S/G SHELL LB AND W: TOTAL AREA)

AA AA ITEMS 11-15 ARE ' STEEL' tINTERNAL MASS) MASSES 11 STEEL MASS FOR CORE NODE: ENTER THE Sun 0F THE MASSES OF AA THE INSIDE HALF 0F THE CORE BARREL (OTHER PART IS IN AA DOWNCOMER NODE) AND THE SUPPORT PLATE 12 nASS OF kV UPPER PLENun INTERNAL STRUCTURE 13 SUM OF HALF THE MASS OF ONE S/G'S TUBES PLUS HALF THE AA LOWER HEAD PLUS THE MASS OF THE HOT LEG PIPE AA (B AND W: ENTER MASS OF HOT LEG ONLY) 14 SUN OF HALF THE MASS OF ONE S/G'S TUBES PLUS HALF THE LOWER AA HEAD PLUS THE MASS OF THE COLD LEG BETWEEN THE 9/G AND AA THE PUMP B0WL (CE: INCLUDE BOTH COLD LEGS IN THE LOOPS AA (B AND W: ENTER TOTAL S/G MASS ASSOCIATED WITH PPINARY SIDE, AA IE TUBES +TUBESHEET+ PRIMARY HEADS, PLUS THE MASSES OF THE AA 2 COLD LEGS UP TO THE PUMP BOWL) l 15 ENTER THE TOTAL MASS OF THE HORI;0NTAL PART OF ALL COLD AA LEGS (IE THAT PART BETWEEN THE RV AND THE PUMPS) PLUS AA THE MASS OF THE RV SHELL CONTAINED WITHIN THE DOWNCOnER 'IE AA THE MASS WHICH LIES BELOW THE NOIZLE ELEVATION) AA PLUS THE MASS OF THE CORE BARREL NOT INCLUDED IN NO.ll AA (DOWNCOMER NODE STEEL MASS) AA AA ITEMS 22-24 ARE THE ' HEAT SINK' MASSES 22 ENTER THE MASS OF THE RV SHELL NOT INCLUDED AS PART OF THE

A-14 AA DOWNCOMER (ITEM 15) IE APPR0x. THE NASS ADOVE THE NO::LE AA ELEVATION (UPPEk PLENUM NODE HEAT SINK NASSe 23 STEEL nASS OF HALF THE S/G SHELL (B AND W: ENTEP Os 24 STEEL MASS OF HALF THE S,G SHELL tb AND W: ENTEF TOTAL MASS OF AA ONE S/G SHELL) AA AA ITEMS 33 AND 35 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHA AA STEEL MASSES AND CONTAINNENT OhTHE

                            $5       ENTER TH     SbFACEA$E Oh E hV N Ef I b bD b AS P AA       DOWNCOMER (SURFACE                             AREA 0F RV BELOW THE ELEVATION WHERE IE APPROX THAT SukFACE AREA WHICH IS AA AA      THEDOWNkOMERSTARTSNOTTLE ELEVATION) bELOW TH AA AA ITEMS 42-44 ARE THE AREAS ASSOCIATED WITH HEAT SINK LOSSES 42 SURFACE AREA 0F THE RV NOT INCLUDED IN THE DOWNCOMER (IIEm 4                              AA     (IE APPROX THAT SUkFACE AREA AbOVE                                                 THEOnNOTTLE ELE'.'ATION)

ENTER 43 SURFACE AREA 0F HALF A S/G SHELL (B AND W: 44 SURFACE AREA 0F HALE A S/G SHELL (B AND W: ENTER TOTAL AREA 0F AA ONE SHELL) AA AA ITEMS 51-55 ARE THE AREAS ASSOCIATED WITH EXCHANGE OF EN

'                               kA NODES AND PRIMAFI SYSTEn GAS                                                                                                                 .

i 51 INTERNAL SURFACE AREA 0F MASS IN ITEM 11 (RV NODE: IE CORE 1 ** BARREL PLUS THE SUPPURT PLATEP 53 INTERNAL SURFACE AREA 0F PASS IN ITEM 12 (UPPER PLENUM E0PT AA AREA) 53 INTERNAL SURFACE AREA UF MASS IN ITEM 13 (BROKEN ' HOT LEG NODE) 5 fNTkAL FAbhAkkA0FMAkfN1 5 lbb N ONEh Ohh AA ITEM 62 IS THE AREA ASSOCIATED WITH EXCHANGE OF ENERGf BETWEE AA HEAT SINK AND THE PRIMARY SYSTEM GAS IN UPPER PLENUM 62 INTERNAL AREA 0F THE P0kT10N OF RV SHELL CONTAINED AA IN UPPEk PLENUM NODE AA IS USED TO COnPUTE GAS VELOCITY IN THE UPPER PLENUM AA B2 ITEM GAS 8'ELOW AkEA IN UPPEk PLENUM AA

AA ITEM 92 IS USED TO COnPUTE HEAT TRANSFER COEFFICIENTS IN 92 HYDRAULIC DIAnETER OF UPPER PLENUM STRUCTURE 4A AA SEDIMENTATION AREA ASSOCIATED WITH STEEL MASSES IN EACH AA SETTLING 101 CORE NODE SEDIMENTATION AREA (IE APPROX SUPPORT PLATE AREA) j 102 UPPER PLENUM SEDIMENTATION AREA 103 BROKEN HOT LEG NODE SEDIMENTATION AREA 104 BROKEN COLD LEG NODE SEDIMENTATION AREA 00WNCOMER SEDIMENTATION AREA (APPROX LOWER HEAD PROJECTE

{05 AAHEAT 111 2.5D5 LOSS INPUTSINOMINAL OPERATION CONVECTIVE (NOT THROU AA HEAT LOSSES l 112 NUMBER OF PLATES IN RADIATIVE INSULATION ON S/G'S(16 NAx) ON PRI Sf 5 113 NUMBER OF PLATES IN RADIATIVE INSULATION (16 MAX) 114 TOTAL AAINSERT A ABk HERE THICKNESS (EG FOR A PARAMETEk OF RADIATIVE DUMP IN BRITISH, INSULATIONAAlf OPERATOR D AAINTERVENTIONS, AND TABULAR OUTPUT) ABR l J d 4

                                                                                      , --- , , - , - - - , - ,         .--.-,--.-s    ~,.r   ,--,----.-,--a.--   - - - -- wg a
    . - . - , , , , , -,--.---a              ,

e ,~.--g , , - - , - , , ,y --

A-15 APPENDIX A.2 MAAP Sequence Control Cards 1 / BATCH UNCERTAINTY (NOT SENSITIVITY) RUN ZION InL8/LOCA 1 25 f 11 8,-2

                                 /nAKE TEMPORARf CHANGES TO SOME OF THESE
                    / SELECT RELEASE n0 DEL 18, 3 -1.       /USE JANAF VAPOR PRESS X 1 FOR CSOH 7$     ,.009     /CONTnT BREAK AREA 1        0   /        I  $SIIP 18             /EA0X 2,4 ,,2.1
                                    /PUT SURGE IN BKN LOOP TO GET PROPER STMMETRY 2    ,.00158                   / TOTAL BNEAK AREA FR0n PUMP SEAL LOCA'S 2, ,25.                          /BkEAK ELEVATION 2,                                 /BkEAK KEY 15' '2              /FCRSLU 18.355 .1.           /DISCH COEFF=1 SINCE WE WANT A GIVEN ELOW 0,0,0                           /N0 nURE CHANGES 0

k. 2. 205 / SHUT OFF DC AND AC POWER 1 0 /NO MORE INITIATING EVENTS 8 / INTERVENE WHEN TIME (CONDITION Os

 .75                        / EQUALS .7$ Hk5 09                      NkIbh$ MAR S$bbbuETDfunPSEALLOCA'S 1

0 /NO MORE ACTIONS 0 /NO MORE INTERVENTIONS 1

A-16 1 .' BATCH UNCEkTAINTY (NOT SENSITIVITY ' EUN IION TMLb/NO LOCA 1 25 0 1 /MAKE TEPPOWARY CHANGES TO SOME UE THESE 11 0,-2 / SELECT PELEASE MODEL 18 3,-1. /USE JANAF VAPOW PRESS X 1 FOR CSOH

                        /N         EA A        $E
'        h' 1_%  , ,20
                  ,6 18     ,2.     /FA0x
            ', ', 00158                      /b   L kkK kbdENPb!P5EA b 'k 2,1 ,25.
 '                                             /BkEAK ELEVATION 2 13,2                               /bkEAK KEY
                                /FCkSLU 0,6,5515,'.1.

18 3 0

                                  /DISCH COEEE=1 SINCE WE WANT A GIVEN ELOW
                                             /NO MORE CHANGES 0

1 6. 3. 205 / SHUT OEF DC AND AC POWEk 0 /NO MORE INITIATING EVENTS 8 / INTERVENE WHEN TIME (CONDITION B)

           .75                         /E00ALS 0700 SECS
                                        /NO OTHER INTERVENTIONS g09
           .                           / FAIL PRIMARY SYSTEM DUE TO PUMP SEAL LUCA'S 0

0 /NO MORE ALT 10NS jo 0 0 0 /NO MORE INTERVENTIONS l J'

A-17 I ZION SLFC 1 25 0 ' 1 11 8,-2 /IDC0k RELEASE MODEL WITH bluCKINb ON 18 -1. /JAtAE VAP0k FEE 5s FOR CSOH 2 J,0218

                  ,.                  /FRIMARY SYSTEM PhEAK SIZE 7'    ,.02                /CONTnT FAILukE HOLE SIZE 7     ,0.             /N0nINAL LONIMI LEAKAGE H0LE SIZE 2,'. 1                      /PUT SURGE IN BKN LOOP TO GET PF0 pes SimmETkt 2       19                    /bkEAh ELEVATION 2'll   1                       /PkEAK KEY 13,,10             / MAX TIME STEP 18      .5     /FCPSLU 18,5,2           / FAUX =2 0                                                                                     '

0,6,6 0. 4 20. 1. 1 242 / MAKEUP OFF 43 / LETDOWN OEF 1 209 / BREAK FAILED 1 227 / MANUAL SCRAM 1 0 8

           .0167 0

215 / MANUAL PUMP TkIP AT .0167 HOUkS 1 0 5. 14.9 /kWST SWITCH-0VER AT LOW LEVEL EAILS 0 216 / SECURE INJECTION--3 PRAT 5 DkAIN GW5T IF ON 4 1

          '17 1

{32 220 0 0 0 l l

A-18

                       .g-ZION ALEC 4                        1 25
                     -0 1

11 18,-2 /USE IDC0k FP RELEASE n0 DEL WITH BLOLklNG 16 3 -l /JANAF VP FOR CSOH 2, ,$. /FkinARf SYSTEM SkEAK SIZE 7, ,.02 .CONInf EAILUhE HOLE SIZE

0. / NOMINAL CONIMI LEAKAuE HOLE SIZE 7,2,. ,1 /PUT SUFGE IN btN LOOP TO GET PEOPER 5thMETRf 2,1 ,19 /BkEAK ELEVATION

                                                              /tkEAK KEt

< 1$,,1 2 1 ,10 / MAX IIME STEP 5 /ECRSLU 10 18 'I 3. /FA0X=' 6,19',1 /0NLY ONE HPI 0,0,0 1 0

                        .5 242                          / MAKEUP OFF 43                        /LEIDOWN Off 1

209 /BkEAK FAILED 27 / MANUAL SCRAM 1 215 / TRIP MCP'S IMMED 1

0 4

5 14.8 /kWST SWITCH-0VER AT LOW LEVEL FAILS 16 / TRIP INJECTION, SPRAYS DkAIN kWST 1 217 32 1 220 0 0 0 1 l } t a- .. ~ -, , - er- -. g -

                                                                  %  - . y--,r . .        , er ~. ..,m-e , - , - vi ,y

A-19 1 ' ERTAINTY iNOT SENSITIVITr ; RUN ZIbTRLB/NOLOCA/RECOVIfh"U$p 1 25 0 1 / mat'E TEMPURAEI CHANGES TO SOME Of THESE 1 6,43,1 / 1 CHP 19.1 / 1 LPI 6,b'

                ,                               b0 6 18 1                / 1 HPI FUmP II,lE.-2                 / SELECT EELEASE MODEL 18.43 1.             /USE SANDIA VAPOR PRESS X 1 FOR CSOH 7,38,.,009              /CONTMT BkEAK WAS .009 7 39,0.             / NOM LEAKAGE AREA 15,3,20
                                                    /PUT SURGE IN BVN LOOP TO GET PROPEh SfMMETRY 2 1 ,,.00158                             / TOTAL 6 PEAK AREA FROM PunP SEAL LOCA'S 2,,1 ,25.                                  /6kEAK ELEVATION f0f'f5 18,3 ,1.
                                      /FCRSLU
                                        /DISCH COEFF=1 SINCE WE WANT A GIVEN FLOW 13 2                        /AFPROX FLOW FROM CNE n0 TOR DRIVEN AEW PUMP PER S/G 0,d,,5.6E4                               /NO MORE CHANGES 0

0. 10. 05 / SHUT OFF DC AND AC POWER 1

 ,       0                                    /NO MORE INITIATING EVENTS 8                                    / INTERVENE WHEN TIME (CONDITION 8)
         .75                                  / EQUALS 2700 SECS 0                                    /NO OTHER INTERVENTIONS a

209 / FAIL PRINAkY SYSTEM DUE TO Punr SEAL LOCA'S 0 0 /NO MORE ACTIONS 8 2.5 / RECOVER AT -- H00ks

         $15 1

205 /POWEk ON 31 /CHP ON 1 1 212 /HPI ON 26 /TRY HEATERS OFF 1 8 /0 PEN P0kV At 4 HOURS 4. 0 211 b i 4.8 / REC 1kC AT kWST LEVEL C0kkESPONDING TO 120000 GAL 0 220 1 0 0 /MO MORE INTERVENTIONS

-v

A-20 1

                                                       '                           H UNCERTAINTY tNOT SENSIT1VITti RUN R

ZION TMLB/LOCA/RECOV AT 1

<            25 l

f / 1 CHP

                                                           /MAKE TEMP 0AAkY CHANGEb 10 50nE Of IHESE 6,        ,1 6,        ,1-          e 1 LPI 6,        ,1            / 1 CONTMT SPRAY PUMP 6        ,.            / 2 EAN C00LEh5 6,         1           / 1 HP !

II,E,-2 / SELECT kELEASE MODEL I 18 3 7,5.1. /USE SANDIA VAPOR PRESS X 1 E0R C50H

                                        /CONIMI BREAK WAS .0V9 009 7 39,0.              / NOM LEAKAGE AREA 13,"3,20 18                     / FAUX 2, $I,,2. 1
                                                                              /PUT EUkGE IN BKN LOOP TO GET PROPER SYMMETRY 4               2,14,.00158                                               / TOTAL EREAL AkEA EROM PUMP 3EAL LOCA'S 2,1 ', , 2 5 .                                                 /BkEAK ELEVATION

- 2 D .J /BhEAL LE!

 '                                         /ECRSLU 10, 18, *5,1.    $.5              /DISCH CUEff=1 SINCE WE WANT A GIVEN ELOW
 '              13                          /1 MOTOR ORIVEN AEW PUMP ELOW PEP S/G 0,0,20,t.6E4
                       .0                                                    /NO MORE CHANGES 0

0. 10. d5 / SHUT OFE DC AND AC POWER 1 0 /NO MODE INITIATING EVENTS ' 8 /lNTERVENE WHEN TIME (CONDITION 8.

                 .75                              / EQUALS 2700 SECS 0                                 /N0 OTHER INTERVENTIONS 209                               /EAIL PRIMARY SYSTEM DUE TO PUMP SEAL LOCA'S 1

0 /NO MORE ACTIONS 8

           . 1.         /RECOVEk AT 1 HOUR 15 1

205 /POWEk ON O 231 /CHAEGING PUMP ON 1 212 /1 HP1 ON 1 0 5 14.8 /nECIRC AT kWST LEVEL CURRESPONDING TO 120000 GAL 20 1 1 0 0 /NO MORE INTERVENTIONS I

1 ' ZION TMLD/LOCA/kECOV AT $ 5 NRgNCERTAINTI (NOT SENSITIV1Tri Peu 1 25 f /MekE TEMP 0kARY CHANGES TO SOME uf THESE 6,43.1 / 1 CHP 6,,' fbbNTNTSPkAYPUMP 13, 8b ' 1h / 2 ,2 FAN COOLERS 6 18 1 < 1 HP! II,15.-2 / SELECT RELEASE F0 DEL 18 3 1. /USE SANDIA VAF0k PkESS X 1 FOR CSOH 7 $ ,.,009 /CONInf &&EAK WAS .009 73 ,0. / NOM LEAKAuE AREA 1 , ,5 18 .2. /FA0x 2, 4,1 /PUT SukGE IN BKN LOOP TO GET PkOPEk SinnETk! 2 . 00158 / TOTAL kkEAN AkEA FROM Puer SEAL LOCA'S 2, 29.25 /bkEAK ELEVAIl0N 9 METEkS VS b.3J AT BOT OF NO2 2' /lFEAK KEY 15,.'2 /FCESLU 18,3$.S.1. /DISCH COEFF=1 SINCE WE WANT a GIVEN FLOW 13 20.5.bE4- /1 nul0k Uk1VEN AFW FuMP FLOW PEP S/G 0,0,0 /NO N0kE CHANGES 0 0. 10. 1. 205 / SHUT OEF DC AND AC POWER 1 0 /NO N0kE INITIATING EVENTS 8 NTERVENE WHEN

         .75                                 /jUUALS2700bEC
                                             /                            INE (CONDITION 8) 0                                    /NO OTHER INTEFVENTIONS 209                                  / FAIL PkIMARY SYSTEM DUE TO PUMP SEAL LOCA'S h                                    /NO MORE ACTIONS 8

2.5 / RECOVER AT -- NOURS 0 215 1 205 /POWE) ON h31 /CHP ON 1 212 / HPI ON 1 0 8 /0 PEN P0kV Bf 10 MINS LATER 2.66 0 211 1 0 5 l 14.8 /RECIRC AT kWST LEVEL CORkESPONDING TO 120000 GAL 0 220 1 0 0 /N0 MORE INTEkVENTIONS l

A-22 1 / BATCH QNCERTAINIT (NOT SENSII!VITT) PUN 2I0N InLB/LOCA/RECOV AT o Hk 1 25 f /MAKE TEMPORAkt CHANGES TU 50nE OF THE5E t 6

          .l 1

3!M

                        / 1 CONIMI SFkAY PUMP 6      ',. 2         / 2 FAN C00 LEKS I

18,$f-2 E CT RELEASE MODEL

                        /USE ANDIA VAFOR PRESS X 1 FOR CSOH 7,3 0.3,hd9.           / CON MT BhEAK WAS .009
                      / NOM LEAKAGE AREA 7

13,, 3 ,20

                                                     /PUT SukGE IN 6KN LOOP TO GET FROPfk SIMMETk1 bi,',I*

2, .00150 eT0TAL BPEAL AkEA EPOM PUMP SEAL LOCA's J. ,25. /BkEAK ELEVATION 2 '2 /B> hat MEf

                               /ECESLU 10.5.S 18,3       ,1.                        /DISCH COEFlsl SINCE WE WANT A GIVEN ELOW 13 '0,5.6E4                   /1 n0 TOR DRIVEN AEW PUMP ELOW PER S/G g,0~0                                            /Nu nukt. CHANGES 0.

10. 05 / SHUT GEE DC AND AC POWEP 1 0 /NO MORE INITIATING EVENTS 8 /INTEkVENE WHEN

    .75                                        / EQUALS 2700 SEC{InE (LONDITION 8) 0                                          /NO OTHER INTERVENTIONS 209                                        /EAIL Pk1 MARY S1 STEM DUE TO PunP SEAL LOCA'S h                                          /NO M0kE ACTIONS 8

6. / RECOVER AT HOUkS kO5 / POWER ON 0

0 4.8 /REClkC AT kWST LEVEL C0kkESPONDING TO 120000 GAL 0 220 1 0 0 /NO MORE INTEkVENTIONS

                                                         .A-23 1

1 / BATCH UNCERTAINTY (NOT SENSITIVITil FUN

*~

IION TMLB/LOCA/kECOV AT 15 HR 1 25 0 1

                                              /MAKE TEMPORARY CHANbEb TO SOME OF THESE 6,      1          / 1 CHP 6     ,1          /1 LPI h,                /       F      00 6      1          / 1 HPI li,       -2          / SELECT RELEASE MODEL 18        1.      /USE SANDIA VAPOR PRESS X 1 FOR CSOH 7

7,3 009 /CONIMI BREAK WAS .009 3 0. / NOM LEAKAGE AREA f h? /FA0X 2,. ',1

                                                /PUT SUkGE IN BKN LOOP TO GET PROPEk SYMMETkt 2       .00158                       / TOTAL BREAK AREA FROM PUMP 5EAL LOCA'S 2 ,' ,' 25.                            / BREAK ELEVATION 2                                      / BREAK KEf l$'.'2 18,3 ,I.
                 .5               /FCRSLU
                                    /DISCH COEFF=1 SINCE WE WANT A GIVEN FLOW 13 2                      /1 MOTOR DRIVEN AEW PUMP FLOW PER S/G 0,d,'5.6E4                          /N0 MORE CHANGES 0

0. QS. 205 / SHUT OFF DC AND AC POWER 1 0 /NO MORE INITIATING EVENT 3 8 /INTEkVENE WHEN TIME (CONDITION 8)

         .75                             / EQUALS 2700 SECS 0                               /NO OTHER INTERVENTIONS 209                             / FAIL PRIMAkT SYSTEM DUE TO PUMP SEAL LOCA 5 1

0 /NO MORE ACTIONS 8

15. / RECOVER AT HOUkS 05 /POWEk ON 0

0 5 ; 14.8 /RECIRC AT RWST LEVEL CORRESPONDING TO 100000 GAL 0 220 i b 0 /NO MORE INTERVENTIONS I 1 l l

~l

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ML20213F819

Contents

Text

Dear Mr. Denton:

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

8.0 CONCLUSION

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

SUMMARY

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ML20213F819

Text

Dear Mr. Denton:

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

8.0 CONCLUSION

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

SUMMARY

Navigation menu

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