ML20213F266

Technical Rept 23.1PB Peach Bottom Atomic Power Station- Integrated Containment Analysis. W/870313 Release Ltr

ML20213F266
Person / Time
Site:	Peach Bottom
Issue date:	03/31/1985
From:	ATOMIC INDUSTRIAL FORUM, INDUSTRY DEGRADED CORE RULEMAKING PROGRAM, PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
References
NUDOCS 8705150276
Download: ML20213F266 (200)

Text

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IDCOR Pro,qram Report 1 F i i i

Technical Report 23.1PB Peach Bottom Atomic Power Station -

Integrated Containment Analysis 1

March 1985 l ,

[h00150276850331 ADOCK y 05000171 i PDR s\

60/4

, The industry !)egraded Core Rulemaking Program, Spcnsored liy the Nuclear Industry

IDCOR \

Arizor.a Public Serirce Company New urk ikwu Authority The Babcock & IVikox Company Niagara Mohawk Ibwer Corporation Baltimore Gas and Elutric Company Northeast Utilities Service Company BuhtelIkwer Corporation Northern Indiana Public Service Company Black & Veatch, Consulting Engineers Northern States ikwer Company Boston Edison Company Pacufc Gas and Elutric Company C F Braun & Co Iknnsylvania Ikwer & Light Company The Cincinnati Gas & Electric Company Philadelphia Elatric Company The Clewland Elatric illuminating Company Ibrtland General Elatric Company Combustion Engineering, Inc. Public Surice Company ofOklahoma Commonwealth Edison Company Public Service Elutric and Gas Company Consolidated Edison Company ofNew urk, Inc. Public Service Indiana Consumers Ibwu Company Puget Soundikwer & Light Company Daniel Construction Company Rochester Gas and Elutric Corporation The Detroit Edison Company Sargent & Lundy Duke Ibwer Company South Carolina Elutric and Gas Company Duquesne Lu:ght Company Southern Cahfornia Edison Company Ebasco Services incorporated Southun Company Services, Inc.

Exxon Nuclear Company, Inc. Stone & IVebsta Enginaring Corporation Florida Ibwer & Light Company Swedish State ikwer Board Fluor ikwer Surices, Inc. Taiwan Ibwer Company General Elutric Company Tuhnical Research Centre ofFinland Gibbs & Ilill, Inc. Tennessee Valley Authority Gilbert / Commonwealth Companies Texas Utilities Generating Company GulfStates Utilities Company The Toledo Edison Company flouston Lighting & Ikwer Company Union Elutric Company illinois Ibwer Company United Engineers & Constructors Inc.

Japan A tomic industrial Forum, Inc. Virginia Elutric and Ikwer Company Kansas Gas and Elutric Company (Vashington Public Ikwer Supply System Long Island Lighting Company IVestinghouse Elatric Corporation Middle South Services, Inc. IVisconsin Elutric lbwer Compary Nebraska Public Ikwer District unkee Atomic Elutric Company The IDCOR program is a large, independent technical effort sponsored by the nuclear industry The Program is directed by a Policy Group comprised of representatives of the sponsoring organizations and operates under the corporate auspices of the Atomic Industrial Forum. The Program's purpose is to develop in an expeditious manner a comprehensive, in-tegrated technically sound position to assist in determining whether changes in regulation are needed to re0cct degraded core and core melt accidents. Further information on the Program can be obtained by contacting John R. Siegel, Special Licensing Projects Manager /IDCOR, Atomic Industrial Forum,7101 Wisconsin Avenue, Bethesda, MD 20814-4805, (301) 654-9260.

IDCOR Technical Report 23.1PB Peach Bottom Atomic Power Station -

Integrated Containment Analysis March 1985 by:

Philadelphia Electric Co.

Philadelphia, Pennsylvania 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-Justrial Forum, nor any ofits employees or members, the IDCOR Policy Group or the IDCOR or Atomic Industrial Forum consultants and contractors, makes any warranty, expressed or implied, or assumes legalliability or responsibili-ty for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately-owned 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 complaints of exclusion, denial of benefits, or other discrimination of those bases under this program may be filed with (among others) the Tennessee Valley Authority (TVA). Office of EEO,400 Commerce Avenue EPB14, Knoxville, TN 37902, and must be not later than 90 daysfrone the Jare of the alleged discrimination. 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 Wdress on request.

4 Copyright @ l 985 by Atomic Industrial Forum, Inc.

7101 Wisconsin Avenue Bethesda, MD 20814-4805 All rights reserved.

Acknowledgements The lead authors of this report were George Daebeler, Alan Marie and Greg Kregar of Philadelphia Electric Company, and Robert E. Henry and Jeff R. Gabor, Fauske 4 Associates, Inc.

The IDCOR Program would like to acknowledge the support and cooperation of the above organizations. In particular, the technical guidance contributed by Dr. Edward L. Fuller, the-technical manager for this project assigned to IDCOR from EPRI, is very much appreciated.

I i

TABLE OF CONTEfiTS Page LIST OF FIGURES ..........................v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . ix 1.0 -INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Statement of the Problem . . . . . . . . . . . . . . . . . . 1-1 1.2 Relationship to Other Tasks . . . . . . . . . . . . . . . . . 1-2 2.0 STRATEGY AND METHODOLOGY . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 References . . . . . . . . . . . . . . . . . . . . . . . . . 2-2

3.0 DESCRIPTION

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

2 3.1.1 Nu cl ea r Sys tem . . . . . . . . . . . . . . . . . . . . 3-1

, 3.1.2 Containment . . . . . . . . . . . . . . . . . . . . . 3-4 3.2 Modular Accident Analysis Program (MAAP) . . . . . . . . . .

3-6 4

3.2.1 MAAP Nodalization . . . . . . . . . . . . . . . . . . 3-7 3.2.2 Peach Bottom Systems Modeled in MAAP . . . . . . . . . 3-10 3.2.3 Fission Product Release and Transport . . . . . . . . 3-10 3.2.4 Fission Product Release from Fuel . . . . . . . . . . 3-12 3.2.5 Description of the Natural Circulation Model . . . . . 3-14 3.2.6 Aerosol Deposition . . . . . . . . . . . . . . . . . . 3-16 4

3.2.7 Fission Product and Aerosol Release from

, Core-Concrete Attack . . . . . . . . . . . . . . . . . 3-20 3.3 Analysis of Reactor Building Thermal-Hydraulic Conditions . . 3-20 3.3.1 Reactor Building and Standby Gas Treatment System (SGTS) . . . . . . . . . . . . . . . . . . . . . . . . 3-20 3.3.2 Modeling Approach . . . . . . . . . . . . . . . . . . 3-22 3.3.3 Mod el I n pu ts . . . . . . . . . . . . . . . . . . . . . 3- 2 3 3.3.4 Influence on Fission Product Release . . . . . . . . . 3-25 f

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TABLE OF CONTENTS (Continued)

Page 3.4 References . . . . . . . . . . . . . . . . . . . . . . . . . 3-25 4.0 PLANT RESPONSE TO SEVERE ACCIDENTS . . . . . . . . . . . . . . . . 4-1 4.1 Plant Response to the TW Sequence . . . . . . . . . . . . . . 4-3 4.1.1 Sequence Description . . . . . . . . . . . . . . . . . 4-3 4.1.2 Primary System and Containment Response . . . . . . . 4-3 4.2 Plant Response to the TC Sequence (Without Operator Action to Reduce Power Level) . . . . . . . . . . . . . . . . 4-13 4.2.1 Sequence Description . . . . . . . . . . . . . . . . . 4-13 4.2.2 Primary. System and Containment Response . . . . . . . 4-14 4.3 Plant Response to the S j E Sequence . . . . . . . . . . . . . 4-25 4.3.1 Sequence Description . . . . . . . . . . . . . . . . . 4-25 4.3.2 Primary System and Containment Response . . . . . . . 4-25 4.4 Plant Response to the TQVW Sequence . . . . . . . . . . . . . 4-33 4.4.1 Sequence Description . . . . . . . . . . . . . . . . . 4-33 4.4.2 Primary System and Containment Response . . . . . . . 4-33 4.5 References . . . . . . . . . . . . . . . . . . . . . . . . . 4-42

5.0 PLANT RESPONSE WITH REC 0VERY ACTIONS . . . . . . . . . . . . . . . 5-1 5.1 Possible Actions . . . . . . . . . . . . . . . . . . . . . . 5-1 f 5.2 S eq u e nc e 1 ( TW ) . . . . . . . . . . . . . . . . . . . . . . . 5- 6 l

5.3 S equence 2 (TC) . . . . . . . . . . . . . . . . . . . . . . . 5-10 l

5.4 Sequence 3 (S jE) . . . . . . . . . . . . . . . . . . . . . . 5-14 5.5 Sequence 4 (TQVW) . . . . . . . . . . . . . . . . . . . . . . 5-19 5.6 References . . . . . . . . . . . . . . . . . . . . . . . . . 5-25 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION . . . . . . . . 6-1

- iii -

TABLE OF CONTENTS (Continued)

Page 6.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Model i ng Approa ch . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Sequences Eval ua ted . . . . . . . . . . . . . . . . . . . . . 6-3 6.3.1 TW Fission Product Release . . . . . . . . . . . . . . 6-3 6.3.2 TC Fission Product Release . . . . . . . . . . . . . . 6-4 6.3.3 S Ej Fission Product Release . . . . . . . . . . . . . 6-7 6.3.4 TQVW Fission Product Release . . . . . . . . . . . . . 6-9 6.4 Assumptions and Actions Relating to Fission Product Distribution and Release . . . . . . . . . . . . . . . . . . 6-11 6.5 References . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 7.0

SUMMARY

OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . 7-1

8.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 APPENDIX A.1 - Peach Bottom Parameter File . . . . . . . . . . . . . A-1

. APPENDIX A.2 - MAAP Input Files for Section 4 Sequences . . . . . . A-15 APPENDIX B - Supplemental Plots for the WASH-1400 Comparison Sequences . . . . . . . . . . . . . . . . . B-1 Supplemental Plots for Sequence TW . . . . . . . . . . B-3 Supplemental Plots for Sequence TC . . ... . . . . . . B-19 Supplemental Plots for Sequence S j E . . . . . . . . . . B-33 Supplemental Plots for Sequence TQVW . . . . . . . . . B-51 i

t

. y-LIST OF FIGURES Figure No. Page t

3.1 Primary system model . . . . . . . . . . . . . . . . . . 3-8 3.2 Mark I nodal ization . . . . . . . . . . . . . . . . . . 3-9

'3.3 BWR Mark I system features modeled in MAAP . . . . . . . 3-11 4

3.4 BWR natural circulation model . . . . . . . . . . . . . . 15 3.5 Fission product transport paths for the primary system and contai nment . . . . . . . . . . . . . . . . . 3-17 3.6 Nodalization scheme for the reactor building . . . . . . . 3-19 4

3.7 Peach Bottom secondary containment . . . . . . . . . . . 3-21 1

4.1 Pressure in the drywell . . . . . . . . . . . . . . . . . 4-5 l

4.2 Temperature of gas in the drywell . . . . . . . . . . . . 4-6 4.3 Temperature of the suppression pool . . . . . . . . . . . 4 4.4 Concrete ablation ' depth in the pedestal . . . . . . . . . 4-8 4.5 Average corium temperature in the pedestal . . . . . . . 4-9 4.6 Reactor pressure vessel water level . . . . . . . - . . . . 4-16 4.7 Average core power . . . . . . . . . . . . . . . . . . . .

4-17 4

4.8 Pressure in the drywell . . . . . . . . . . . . . . . . . 4-18 4.9 Temperature of gas in the drywell . . . . . . . . . . . . 4-19 4.10 Temperature of the suppression pool . . . . . . . . . . . 4-20 4.11 Concrete ablation depth in the pedestal . . . . . . . . . . 4- 21 l

4.12 Average corium temperature in the pedestal . . . . . . . 4-22 4.13 Pressure in the drywell . . . . . . . . . . . . . . . . . 4-27 4.14 Temperature of gas in the drywell . . . . . . . . . . . . 4-28 4.15 Temperature of the suppression pool . . . . . . . . . . . 4-29 4.16 Concrete ablation depth in the pedestal . . . . . . . . . 4-30 L 4.17 Average corium temperature in the pedestal . . . . . . . 4-31 i

I, 9

e

.-- . .- - , + . . --,c- . , - . , - . . , -s , - - . . . --- - -,%-,-,u ,-. - - - , , . . . - . - .

- vi -

LIST OF FIGURES (Continued)

Figure No. Page 4.18 Pressure in the drywell . . . . . . . . . . . . . . . . 4-35 4.19 Temperature of gas in the drywell . . . . . . . . . . . 4-36 4.20 Temperature of the suppression pool . . . . . . . . . . 4-37 4.21 Concrete ablation depth in the pedestal . . . . . . . . 4-38 4.22 Average corium temperature in the pedestal . . . . . . . 4-39 5.1 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-8 5.2 Reactor pressure vessel water level . . . . . . . . . . 5-9 5.3 Drywel l pressure . . . . . . . . . . . . . . . . . . . . 5-11 5.4 Reactor water level . . . . . . . . . . . . . . . . . . 5-13 5.5 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-17 5.6 Temp era tu re i n d rywel l . . . . . . . . . . . . . . . . . 5- 18 5.7 Reactor _ pressure vessel water level . . . . . . . . . . 5 - 21 5.8 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-23 5.9- Temperature of the suppression pool . . . . . . . . . . 5-24 8.1 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-4 ,

8.2 Reactor vessel wa ter l evel . . . . . . . . . . . . . . . . B-5 B.3 Temperature of s tructure, *F , . . . . . . . . . . . . . . B-6 B.4 Fission product decay power on structure, MBtu/hr . . . . B-7 1

_B.5 Tota l C0 genera ted . . . . . . . . . . . . . . . . . . . . B-8 B.6 Mass of water in the pedestal . . . . . . . . . . . . . . B-9 B.7 Volumetric flow out of containment . . . . . . . . . . . . B-10 B.8 Mass of UO i n core region . . . . . . . . . . . . . . . . B-ll 2

B.9 Gas temperature in reactor building, 'F . . . . . . . . . B-12 B.10 Steam pressure in reactor building, Pa . . . . . . . . . . B-13

- vii -

LIST OF FIGURES (Continued)

Figure No. Page B.ll Cesium iodide released from containment, kg . . . . . . . B-14 B.12 Cesium iodide released to environment, kg . . . . . . . . B-15 B.13 Tellurium released from containment, kg . . . . . . . . . B-16 B.14 Tellurium released to environment, kg . . . . . . . . . . B-17 B.15 Total H2 g enera ted . . . . . . . . . . . . . . . . . . . . B-20 B.16 Reactor vessel water level . . . . . . . . . . . . . . . . B-21 B.17 Temperature of structure. *F . . . . . . . . . . . . . . . B-22 B.18 Fission product decay power on structure, Btu /hr . . . . . B-23

B.19 Total C0 genera ted . . . . . . . . . . . . . . . . . . . . B-24 B.20 Mass of water in the pedestal . . . . . . . . . . . . . . B-25 B.21 Mass flow out of containment . . . . . . . . . . . . . . . B-26 B.22 Reactor building gas temperature, 'F . . . . . . . . . . . B-27 B.23 Reactor building steam partial pressure . . . . . . . . . B-28 B.24 Cesium iodide released from containment, kg -. . . . . . . B-29 B.25 Mass of cesium iodide released to environment . . . . . . B-30 B.26' Tellurium released' from containment, kg . . . . . . . . . B-31 B.27 Tellurium released to environment, kg . . . . . . . . . . B-32 B.28 Total H2 genera ted . . . . . . . . . . . . . . . . . . . . B-34 B.29 Reactor vessel water level . . . . . . . . . . . . . . . . B-35 B.30 Temperature of structure, F . . . . . . . . . . . . . . . B-36 B.31 Fission product decay power on structure, MBtu/hr . . . . B-37 B.32 To tal C0 g enera ted . . . . . . . . . . . . . . . . . . . . B-38

. B.33 Mass of water in the pedestal . . . . . . . . . . . . . . B-39 B.34 -Volumetric flow out of containment . . . . . . . . . . . . B-40

- viii -

LIST OF FIGURES (Continued)

Figure No. Page B.35 Mass of UO i n core reg ion . . . . . . . . . . . . . . . . B-41 2

B.36 Gas temperature in reactor building, F . . . . . . . . . B-42 B.37 Steam pressure in reactor building, Pa . . . . . . . . . . B-43 B.38 Cesium iodide released from containment, kg . . . . . . . B-44 B.39 Cesium iodide released to environment, kg . . . . . . . . B-45 B.40 Cesium iodide passing through the Standby Gas Trea tment Sys tem . . . . . . . . . . . . . . . . . . . . . B-46 -

B.41 Tellurium released from containment, kg . . . . . . . . . B-47 B.42 Tellurium released to environment, kg . . . . . . . . . . B-48 B.43 Tellurium passing through the Standby Gas Trea tment Sys tem . . . . . . . . . . . . . . . . . . . . . B-49 B.44 Total H2 g enera ted . . . . . . . . . . . . . . . . . . . . B-52 B.45 Reactor vessel water level . . . . . . . . . . . . . . . . B-53 B.46 Temperature of structure. *F . . . . . . . . . . . . . . B-54 B.47 Fission product decay power on structure, Btu /hr . . . . . B-55 B.48 To tal C0 g enerated . . . . . . . . . . . . . . . . . . . . B-56 B.49 Mass of water in the pedestal . . . . . . . . . . . . . . B-57 B.50 Mass of UO 2 in core region . . . . . . . . . . . . . . B-58 B.51 Gas temperature in reactor building. *F . . . . . . . . . B-59 B.52 Steam pressure in reactor building, Pa . . . . . . . . . . B-60 B.53 Cesium iodide released from containment, kg . . . . . . . B-61 B.54 Cesium iodide released to environment, kg . . . . . . . . B-62 B.55 Tellurium released from containment, kg . . . . . . . . . B-63 B.56 Tellurium released to environment, kg . . . . . . . . . . B-64

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LIST OF TABLES Table No. Page 3.1 Initial Inventories of Fission Products and Structural Ma terial s . . . . . . . . . . . . . . . . . . 3-13 3.2 Reactor Building Model Inputs . . . . . . . . . . . . . 3-25 4.1 Peach Bottom - TW Event Summary . . . . . . . . . . . . 4-4 4.2 Peach Bottom - TC Event Summary . . . . . . . . . . . . 4-15 4.3 Peach Bottom - Sj E Event Summary . . . . . . . . . . . . 4-26 4.4 Peach Bottom - TQVW Event Summary . . . . . . . . . . . 4-34 5.1 I nj ec tio n to RPV . . . . . . . . . . . . . . . . . . . . 5- 2 5.2 Tank Capacities and Replenishing Sources . . . . . . . . 5-3 5.3 RPV/ Containment Cooling . . . . . . . . . . . . . . . . 5-4 5.4 Containment Venting Provisions . . . . . . . . . . . . . 5-5 5.5 TW With Selected Operator Actions - Event Summary . . . 5-7 5.6 TC With Selected Operator Actions - Event Summary . . . 5-12 5.7 Sj E With Selected Operator Actions - Event Summary . . . 5-16 5.8 TQVW With Selected Operator Actions - Event Summary . . 5-20 6.1 Distribution of Csl in Plant ano Envirorment (Fraction of Core Inventory) . . . . . . . . . . . . . . 6-5 6.2 TW Fission Product Release . . . . . . . . . . . . . . . 6-6 6.5 TC Fission Product Release . . . . . . . . . . . . . . . 6-8 6.4 S j E Fission Product Release . . . . . . . . . . . . . . 6-10 6.5 TQVW Fission Product Release . . . . . . . . . . . . . . 6-12 6.6 Csl Distribution and Release (Fraction of Initial Core Inventory) . . . . . . . . . . . . . . . . . . . . 6-14 7.1 Summary of MAAP Results for Base Sequences . . . . . . . 7-2 7.2 Summary of Fission Product Release Fractions . . . . . . 7-3 i

l

1-1 l-

1.0 INTRODUCTION

11 Statement of the Problem

! The main objective of this investigation is to calculate the re-sponse of the Peach Bottom Atomic Power Station (PBAPS) primary system and containment for selected postulated low probability, severe accident sequences representative of those which have been identified in Task '3.2 as dominant sequences 'potentially leading to core degradation and melting. This response is' addressed on a best-estimate phenomenological basis. This study includes

. assessments of the effects of selected examples of operator interventions on the progression of these sequences.

This analysis is not intended to be a probabilistic assessment in that no assessment of'the probabilities of Peach Bottom systems or operator failures is included. However, the accident sequences were defined based on a

- review of past BWR PRAs including the WASH-1400 BWR analyses which ' identified those sequences most likely to lecd to core melting with assumptions of multiple system failures and very few operator mitigating actions. The results of these analyses indicate the time windows available for operators to l'

implement mitigative actions. The effects of selected actions on accident progression _ are addressed. No attempt was made to model the variety of operator actions prescribed in the PBAPS emergency procedures. This approach i is sufficient to demonstrate the effects that simple, individual actions would i

have on accident progression.

The results of the containment analysis are incorporated into an

! assessment of the fission product release and deposition within 'the various l regions of' the primary' and secondary containment structures. For those i seq'uences in which core damage occurs and containment integrity is violated, the release of fission products to the surrounding environment is calculated.

l The influence of a few existing systems with operator action is described in Section 5.

l i

1 i

c m . .- . . , . , ,, ., . ._-_ _ ...= ,- - . --. -_ _ __ _-- -

1-2 1.2 Relationship to Other Tasks The primary system and containment response analyses of IDCOR Subtask 23.1 are carried out with the Modular Accident Analysis Program. This includes models developed in IDCOR subtasks 11,12,14,15 and 16 for thermal-hydraulic behavior as well as fission product release transport and deposition within both the primary system and containment. The accident sequences analyzed were developed by considering the dominant core melt accident se-quences presented in Subtask 3.2, Assess Dominant Sequences. Selected primary containment failure modes were chosen to demonstrate the radionuclide trans-port phenomena for the best-estimate analyses.

The ultimate structural capability of containments associated with the reference plants and other typical designs was assessed in IDCOR Subtask 10.1. Task 10.1 also defined the containment failure pressure and location assumed in Subtask 23.1 analyses for those sequences resulting in containment failure on overpressure.

Calculations of the rate and amount of fission products released from the conte.inment, for those sequences which result in containment failure, were supplied to IDCOR Subtask 18.1 to formulate assessments of the health consequences associated with the assumed accident scenarios. These health consequence analyses were then supplied to IDCOR Subtask 21.1 to evaluate effects on perceived risk.

Also, a few examples of operator interventions were analyzed to demonstrate their effect on the severe accident sequences analyzed for Peach

Bottom -- that operator actions can terminate the accident sequence and achieve a safe stable state. The operator actions selected considered IDCOR Subtask 22.1, Safe Stable States, which discusses the inherent safety of a BWR in terminating the various core damage sequences.

It was not th intent of Task 23.1 to address the likelihood of occurrence of the particular sequences and operator actions, but rather to assume these situations and analyze the accompanying containment challenges

1-3 and release of fission products utilizing the models developed within the IDCOR program.

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 include: hydrogen generation, distribution and combustion (12.1,12.2 and 12.3), steam generation (14.1), core heatup (15.1), debris behavior (15.2), and core-concrete interactions (15.3). Detailed considerations for each of these related subtasks can be found in the final reports submitted for the specific task. Individual issues will only be addressed in this report as required to understand the specific behavior obtained for the accident se-quences considered and the specific design characteristics of Peach Bottom Atomic Power Station.

_ _ _ _ ,- - ,y

.2-1 2.0 STRATEGY AND METHODOLOGY

. The basic strategy of this subtask was to analyze accident sequences which have been previously identified as cominant or key potential contribu-It ' tors to core melt frequency in other probabilistic risk assessments. Task 23.1. analyses consisted of plant thermal hydraulic response and fission i- product. transport if the progression of the accident sequence led to core

degradation 'and melting. These analyses include the performance of the ECCS

j. systems and the containment engineered safety systems, such as the suppression pool, containment inerting, decay heat removal system, etc.

The MAAP code [2.1] was used to perform the primary system and containment thermal-hydraulic response analyses. This code considers the major physical processes associated with an accident progression, including hydrogen generation, steam formation, debris coolability, debris dispersal,

, core-concrete interactions, and hydrogen combustion. The FPRAT module for

! MAAP, as adopted from [2.2] to evaluate the fission product release from the i fuel. Natural and _ forced circulation within the primary system is modeled both before and af ter 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

sedimentation. '

For each of the four PBAPS accident scenarios selected for analysis, l thermal-hydraulic calculations were performed both with and without selected

-operator actions during the accident. The WASH-1400 comparison case analyses,

{ which assume only minimal operator response during the accident, establish a

! reference system response during each of the accident scenarios. The " opera-tor action" analyses are branch calculations of the WASH-1400 comparison i

cases. These operator intervention cases demonstrate the effect of selected

_ operator actions on the progression of an accident, based on the time windows available to the operator to take such action. Additional uncertainty and f sensitivity analyses have been performed on several key thermal hydraulic l

pararreters associated with the accident response. These are reported in Ref.

, [2.4].

l

{

l

2-2 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 "FPRAT User's Manual".

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.

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3.0 DESCRIPTION

OF MODELS AND MAJOR ASSUMPTIONS The Modular Accident Analysis Program (MAAP'), Ref. [3.1] is used to model the Peach Bottom response to postulated severe accidents. This code includes models for the primary system and containment response, fission

~

. product release. and fission product transport. In addition, the thermal hydraulic conditions as well as the fission product behavior are modeled for the reactor building which surrounds the primary containment.

3.1 Plant Specific Information Each of Peach Bottom Units 2 and 3 is a single cycle, forced circu-

-lation, 3293 MW(t) General Electric BWR-4 producing steam for direct use in the steam turbine. Each unit has a Mark I primary containment housed in a secondary containment (reactor building). Both units went into commercial operation in 1974.

3.1.1 Nuclear System The reactor vessel contains the core and supporting structure, the steam separators and dryers, the jet pumps, the control rod guide tubes, distribution lines for the feedwater, core spray, and standby liquid control, the in-core instrumentation, and other components. - The main connections to the vessel include the steam lines, the coolant recirculation lines, feedwater lines,' control rod drive housings, and core standby cooling lines.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure is 1020 psia in the steam space above the separators. The reactor core is cooled by domineralized water which enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is directed through steam separators and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steam lines. Each steam line is provided with two Main Steam Isolation Valves (MSIVs) in series, one on each side of the primary containment barrier.

3-2 When a scram is initiated by the Reactor Protection System, the Control Rod Drive system (CRO) inserts the negative reactivity necessary to shutdown the reactor. Each control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high pressure water for each rod forces each control rod rapidly into the core. There are 185 control rods which enter through the bottom of the reactor vessel.

A pressure relief system, consisting of relief and safety valves mounted on the main steam lines, prevents excessive pressure inside the nuclear system following either operational transients or accidents.

Process lines which penetrate the primary containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities. In addition, the main steam lines, because of their large size and large mass flow rates, are given special isolation considera-tion. Two automatic isolation valves, spring loaded to close and requiring air pressure to remain open, are provided in each main steam line.

The Reactor Core Isolation Cooling system (RCIC) can provide makeup water to the reactor vessel even when the vessel is isolated. RCIC uses a steam driven turbine pump unit and operates automatically, in time and with sufficient coolant flow, to maintain adequate reactor vessel water level.

The Residual Heat Removal system (RHR) is a system of pumps, heat exchangers, and piping that fulfills the following functions:

1. Removal of residual heat from containment during and af ter plant shutdown (manual operation).

2. Automatic injection of water into the reactor vessel, following a LOCA, rapidly enough to reflood the core and prevent exces-sive fuel clad temperatures, independent of other core cooling systems.

3. Removal of heat from the primary containment following a LOCA to limit the increase in primary containment pressure. This is

3-3 accomplished by cooling and recirculating the water inside the primary containment. The redundancy of the equipment provided for containment cooling is further extended by a separate part of the RHR system which sprays cooling water into the drywell .

4. Remove heat from the RPV at low pressure in the shutdown cooling mode.

A number of Core Standby Cooling (CSC) systems are provided to prevent excessive fuel clad temperatures in the event a breach in the nuclear system process barrier results in a loss of reactor coolant. The four CSC systems are:

1. High Pressure Coolant Injection system (HPCI)

2. Automatic Depressurization System (ADS)

3. Core Spray System (LPCS)

4. Low Pressure Coolant Injection (an operating mode of the RHR system)(LPCI)

Although not intended to provide rapid reactor shutdown, the standby liquid control (SLC) system provides a redundant, independent, and diverse method from the control rods to bring the reactor subcritical and to maintain it subcritical as the reactor cools. The system makes possible an orderly and safe shutdown in the event that insufficient control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition.

The standby AC power supply consists of four diesel generator sets.

The diesel generators are sized so that three diesels can supply all necessary power requirements for one unit under postulated design basis accident condi-tions plus necessary power requirements for the safe shutdown of the second unit. The diesel generators start and attain rated voltage and frequency

3-4 within 10 seconds. The diesel generator system is arranged with four indepen-dent 4-kV buses for each unit, each bus being connected to one diesel genera-tor. Each diesel generator starts automatically upon undervoltage on the emergency bus or detection of symptoms characteristic of postulated accidents.

The necessary engineered safeguard system loads are applied on a preset time sequence. Each generator operates independently without paralleling.

Two independent sets of 125/250-V batteries are provided for each reactor unit. The sets are not interconnected. In addition, a separate 250-V battery is provided for each main turbine generator emergency bearing oil pump. One battery charger'is provided for each battery.

The 125/250-V DC system is designed to provide an adequate power source for supplying the engineered safeguard loads of one unit, and the shutdown loads of the second unit, with concurrent loss of off-site power and j any single failure in the DC system.

i 3.1.2 Containment-The primary containment is a pressure suppression system and houses the reactor vessel, the reactor coolant recirculation systems, and other primary system piping. The primary containment system consists of a drywell, a pressure suppression chamber which stores a large volume of water, a con-necting vent system between the drywell and the suppression pool, isolation valves, vac.uum breakers, containment cooling systems, and other service

! equipment.

In the event of a primary system piping failure within the drywell,

, reactor water and steam would be released into the drywell atmosphere. The resulting increased drywell pressure would force a mixture of drywell at-mosphere, steam, and water through the downcomer vents into the suppression

[ pool, resulting in a pressure reduction in the drywell due to steam condensation.

Vacuum breakers are provided in the vent headers and located in the

j. suppression chamber, to equalize the pressure between the suppression chamber I

3-5 and the irywell . A vacuum breaker system is also provided between the sup-pression chamber and secondary containment. Cooling systems are provided to remove heat from the drywell and from the water in the suppression chamber.

Appropriate isolation valves are provided to ensure containment of radioactive materials.

The downcomer system conducts flow from the drywell to the suppres-sion chamber and distributes this flow uniformly in the suppression pool. The suppression pool condenses the steam portion of this flow and the suppression chamber contains the noncondensable gases and fission products. The suppres-sion chamber-to-drywell vacuum breakers and the secondary containment-to-suppression chamber vacuum breaker system limit the pressure differential between the drywell and torus so as not to exceed the design limit of 2 psi.

The suppression chamber is designed for the same leakage rate as the drywell.

The suppression pool also provides for steam condensation during the 4 actuation of a safety relief valve and the subsequent blowdown through the discharge piping. The dynamic suppression pool loads resulting from a safety relief valve discharge are reduced by a sparger (T-quencher) on the discharge end of the safety relief valve piping. The sparger also provides for uniform and stable condensation of steam in the suppression pool.

The stiffened pressure suppression chamber is a steel pressure vessel in the shape of a torus. It is located below and encircles the dry-well, with a centerline diameter of approximately 111 ft. and a cross-3 sectional diameter of 31 f t. It contains approximately 123,000 ft of water 3

and has a gas space volume of approximately 132,000 ft . The drywell vents are connected to a 4 f t. 9 inch diameter vent header, in the form of a torus, which is contained within the airspace of the suppression chamber. Projecting downward from the header are 96 downcomer pipes, nominally 24 inch in diameter and terminating 4 ft. below the design water level of the pool.

1 The downcomer vent system outside the torus consists of eight circular vent pipes, each having a diameter of 6 ft. 9 inches. These down-comer pipes are connected to the vent header located inside the torus. The downcomer pipes and header have the same temperature and pressure design

3-6 requirements as the containment. Jet deflectors are provided in the drywell at the entrance of each vent pipe to prevent damage to the downcomer pipes from jet forces which might accompany a pipe break in the drywell.

Pressure suppression pool temperature and pool level are continuous-ly indicated in the main control room.  ;

4 i

i The RHR system can be placed into operation in the suppression pool cooling mode to limit the temperature of the water in the suppression pool'.

In this mode of operation, the RHR system pumps take suction from the suppres-sion pool and deliver the water through the RHR system heat exchangers, where cooling takes place by transferring heat to the service water. The fluid is then discharged back to the suppression pool.

Another portion of the RHR system is provided to spray water into i the primary containment as a means of reducing containment drywell temperature i

and pressure following a LOCA. This capability is in excess of the required energy removal capability and can be placed into service at the discretion of

! the operator.

i 3.2 Modular Accident Analysis Program (MAAP)

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

package in Subtask 16.3, while Subtask 16.2 provides a computer code (MAAP) l [3.1] to analyze the major degraded core accident scenarios for both Pres-5 surized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The MAAP code is designed to provide realistic assessments for severe core damage ,

j accident sequences using first principle models for the major phenomena that

, ' govern the accident progression, the release of fission products from the fuel

] matrix, the transport of these fission products and their deposition within '

[ the primary system and containment. The following sections describe the i primary system and containment nodalization and include a description of the l . safety systems modeled in the MAAP. A complete Peach Bottom parameter file is given in Appendix A.l.

3-7 3.2.1 MAAP Nodalization The BWR primary system nodes are illustrated in Figure 3.1 and include the lower plenum, downcomer, core, and uppe.r plenum. Also indicated are the flow entry locations for CRD flow, feedwater, HPCI, RCIC, LPCI and LPCS as well as the standby liquid control system (SLCS), which is only modeled as an additional water source since MAAP does not have a neutronics model. Individual mass and energy equations are written for each of these nodes using the water addition locations and the appropriate connecting flow paths. The primary system model also represents the main steam isolation valves and the main steam safety and relief valves which exhaust into the suppression pool.

Modeling of the primary system is used to determine if a given sequence (1) leads to core uncovery, (2) results in core damage, (3) yields Zircaloy clad oxidation and hydrogen formation, (4) leads to core melt and vessel failure, (5) can be recovered before vessel failure, and predicts the time of these occurrences. The transient response to the spectrum of accident scenarios considered requires the specification of pump curves, valve set points, system logic, etc. With the specification of the accident sequence, the primary system model determines the vessel water inventory, including the boiled-up level in the core, to evaluate the potential for core uncovery. If the collapsed water level decreases below the top of the core, the HEATUP subroutine calculates the temperature increases of the fuel and cladding, including steam cooling and the oxidation of the Zircaloy clad and fuel channel cans as determined by the appropriate rate laws and oxygen starvation.

The model permits incorporation of CRD flow to evaluate the potential of specific sequences, such as TW, being terminated with limited core damage.

The Mark I (Peach Bottom) containment nodalization scheme as shown in Fig. 3.2 separates the containment into: the pedestal, the drywell, and the wetwell regions. MAAP evaluates the behavior of the various compartments during the entire progression of the accident sequence by calculating the mass and energy flow rates between these compartments.

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3-10 Individual compartment (region) pressures and gas temperatures are

! derived from the mass and energy balances. MAAP models the transport of water

( throughout the containment due to drainage, vaporization, condensation and l mass addition to assess the potential for cooling core debris should the vessel have failed. Separate water and corium temperatures are calculated for each containment compartment.

3.2.2 Peach Bottom Systems Modeled in MAAP In general MAAP characterizes the response of the primary system, the containment and many of the balance of plant systems to user specified event sequences. Figure 3.3 illustrates the plant systems modeled in the code  !

including the various water sources available and the valve line-ups which would allow this water to be injected into either the primary system and/or containment during a postulated sequence. Particular systems of importance include, the control rod drive (CRD) flow from the condensate storage tank, main steam lines, MSIVs, turbine bypass, feedwater, reactor core isolation cooling (RCIC), high pressure coolant injection (HPCI), low pressure coolant injection (LPCI) and other RHR system modes, low pressure core spray (LPCS),

standby liquid control system (SLCS), and high pressure service water (HPSW). l In addition to these plant systems, MAAP nodalizes both the primary system and containment to model their response to postulated core damage and recovery scenarios.

3.2.3 Fission Product Release and Transport The rate of release of fission products from the fuel matrix was calculated with the FPRAT module in MAAP. The FPRAT code was developed as part of the 10COR program and is described in the report for Subtask 15.18

[3.2]. FPRAT was integrated into the MAAP coding structure such that the fission product release and transport from the core is evaluated at each time step.

The release of fission products due to corium-concrete thermal attack and ablation was calculated as described in Section 3.2.7. Transport of fission products through the primary system and containment was calculated

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3-12 with the CIRC module in MAAP and includes models for the fission product ource terms, primary system compartment temperatures, primary system heat losses, gas flow due to forced and natural circulation, and steam condensation for steam and fission products.

Estimates of thermal-hydraulic characteristics of the flow of the containment effluent through the reactor building (secondary containment) were developed as described in Section 3.3.

3.2.4 Fission Product Release from Fuel i The initial fission product inventories were obtained from Ref.

[3.3] and are given in Table 3.1. Fission product release rates depend on ,

fuel temperature history during heatup, and on the flow through the core. The gas flow through each node is assumed to be saturated with the vapor of each constituent. If the flow cools as it is transported to higher nodes, the gas cools and creates aerosols of each species to remain saturated. This flow provides the aerosol and vapor source for the upper plenum.

For the regions in which blockage has occurred, it is assumed that sufficient flow exists to remove the volatile fission products as saturated vapor. Once this flow is determined, the removal of the remaining less volatile, species are evaluated basoJ upon saturation of this calculated flow.

The required FPRAT input for MAAP is given in the parameter file in Appendix A.l.

The calculations consider evaporation and condensation characteris-tics of chemical species. Several key assumptions coasistent with the recom-mendations of IDCOR Subtask 11.1 were made regarding the physical and chemical forms of released fission products. These are:

1. Cesium and iodine combine for form Csl upon entry to the fission product release pathway. The excess cesium forms Cs0H. I These two fission product groups are assumed to be liberated in vapor form.

3-13 Table 3.1 INITIAL INVENTORIES OF FISSION PRODUCTS AND STRUCTURAL MATERIALS l

Fission Products Initial Inventory (kg)

Kr 25.7 Xe 387 Cs 207 I 16.6 Te 34.9 Sr 62.7 Ru 172 La 98.3 Mo 237 Sn 1050 Mn 432 i

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3-14

2. Tellurium is assumed to be released as vaporized Te0

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3. Inert aerosol generation rate is the combined release rates-for volatile structure material (Mn and Sn).

4. Strontium and ruthenium represent their respective nonvolatile fission product groups as defined in WASH-1400. They.are also calculated to be released as vapor which quickly forms aerosols when they exit the core.

5. Release of volatile fission products (Cs, I, Te) and the noble gases (Xe and Kr) is allowed to continue until complete, even if the vessel has already failed.

3.2.5 Description of the Natural Circulation Model For certain transient-induced accident sequences substantial quanti-ties of fission products are calculated to be released during core degrada-  !

tion, but before reactor vessel failure. Gas flow through the primary system I determines the aerosol transport and deposition throughout the reactor vessel.

Following reactor vessel failure, fission products could remain within the primary system and subsequently heat the adjacent structures. As the struc- )

ture and gas temperatures increase, density differences within the primary system would result in natural circulation flows that could distribute both heat and mass throughout the primary system. Prior to vessel failure circula-l tion among the core, upper plenum and downcomer is not credited due to the water blockage within the jet pumps.

The BWR-CIRC module models natural circulation flows within the primary system. This includes descriptions for fission product heat genera-tion, material vaporization, condensation and deposition. Also, this nodali-zation allows for a representation of the structural heatup in each node as well as the heat losses from these nodes to the containment environment. The circulation for the BWR system af ter vessel failure is graphically represented in Fig. 3.4. As illustrated, the throat area for the jet pumps controls the

3-15 H

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3-16 circulation rate and the containment pressurization /depressurization influ-ences the flow from the primary system.

Since natural circulation flows are driven by the gas density differences between various regions, and since the volatile fission products are dense vapors, the gaseous flows must have a detailed accounting of the gas mixture properties in the various nodes. In addition, with the reflective insulation used on the Peach Bottom reactor vessel the heat losses from the vessel must also include the magnitude of heat losses as a function of the primary system temperature and the potential for oxidation of the stainless steel layers in the reflective insulation. Heat addition to the containment from pumps, motors or hot piping is not modeled by MAAP. ,

These analyses have been coupled with models for aerosol deposition and heatup to evaluate the primary system flows af ter reactor vessel failure.

Such assessments provide the rate and amount of material released from the primary system as a result of the subsequent heatup of primary system struc-tures. In this analysis, the difference between the primary systen and containment pressurization determines the flows between these two systems which govern the release of fission products to the containment environment.

3.2.6 Aerosol Deposition IDCOR Task 11.3 evaluated models for aerosol agglomeration and deposition processes of fission products [3.4]. These removal processe's reduce the magnitude of radionuclide release to the environment. The cor-responding 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 [3.5].

This formulation 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 steam condensation and sedimentation. The MAAP nodalization scheme for fission product transport is identical to that used for the thermal-hydraulic models in MAAP. The specific transport paths are illustrated in Fig. 3.5 are

3-17 UPPER INTERNALS  ;

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3-18 for the primary system and containment and in Fig. 3.6 for the reactor build-ing and the SGTS.

The assumptions in the aerosol modeling are:

1. Cesium and iodine are assumed to be released as Csl with excess cesium as Cs0H.

2. The decontamination factor associated with the wetwell suppres-sion pool is estimated to be 1000 for release through the spargers and 600 for releases through the downcomers (Ref.

[3.6]).

3. Compartments representing the release pathway are: three regions in the reactor vessel, pedestal, drywell, wetwell, reactor building, standby gas treatment system (SGTS) ducts, and SGTS charcoal filter (physical removal mechanisms only).

Figures 3.5 and 3.6 depict the release pathways and compart-ments for the analyses. Specific release paths and the flow rates are dependent on the thermal-hydraulic conditions in the reactor building as well as the flow capacity of SGTS and percent steam composition of the carrier gas as determined by the thermal hydraulic models in MAAP (see Section 3.3).

4. Steam carrying fission products out of the containment and along the release pathway would condense in the cooler reactor building. Steam condensation rates, volumetric gas flows through the reactor building and temperatures are calculated as described in Section 3.3.

5. Hygroscopic aerosols, such as cesium hydroxide, are assumed to accumulate an equilibrium concentration of water as determined by the steam partial pressure and temperature.

6. Deposition of fission products in the SRV discharge lines was neglected.

3-19 ENVIRONMENT d .

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FROM DRYWELL

=

• Flow path for postulated Station Blackout accident sequence, also portion greater than SGTS flow for the other sequences.

k Fig. 3.6 Nodalization scheme for the reactor building.

3-20

7. -Chemical reaction between Cs and the reactor vessel steel is neglected..

3.2.7 . Fission Product and Aerosol Release from Core-Concrete Attack The release of aerosols due to core-concrete attack was not included in the Peach-Bottom analysis. This omission leads to an underprediction of-

' the overall fission product removal in the primary and secondary containments.

3.3 Analysis of Reactor Building Thermal-Hydraulic Conditions

-3.3.1 Reactor Building and Standby Gas Treatment System (SGTS)

Each of Peach Bottom Units 2 and 3. primary containments is housed in a multilevel reactor building. Under accident conditions, the reactor build-ing atmosphere is isolated from the normal ventilation system and exhausted through the Standby Gas Treatment System (SGTS) HEPA and charcoal filters at a rate which maintains the building pressure negative relative to the environ-ment. The reactor building and SGTS comprise the secondary containment system at Peach Bottom.

As shown-in Fig. 3.7, the reactor building is divided into five major levels with gaseous flow communication between them through open hatches. An equipment transfer shaf t from the 135' level up to the refueling floor is the major pathway for communication between the various volumes of the reactor building. The lowest elevation (s 92' to 133') contains the torus room and torus, and the RHR and core spray pump rooms located in the four corners of the building separated from the torus room by concrete walls and water tight doors. The next elevations (135' to 163') contain the main steam pipe tunnel, components of the CR0 hydraulic system, neutron monitoring system and other instrumentation partially separated by several partition and shield walls. Elevation 165' to 193' contains auxiliary pumps, heat exchanger. ;nd instrumentation separated by many partitions and shield walls creating sub-stantial interior surface area. Much of the space in elevation 195' to 232' is occupied by the spent fuel pool and steam separator and dryer storage pool.

This level also contains the standby liquid control system and reactor

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3-22 building fan room. Therefore, much of the volume is closed off by major walls and available surface area is less than that available in the lower eleva-tions. The top elevation (234' to roof at 296') is essentially a wide open I area comprising the refueling floor. Most of the exterior wall area is insulated corrugated sheet metal on this elevation as compared to concrete on all lower elevations.

The SGTS takes suction from multiple intakes located on all major elevations in the reactor building. Tests have indicated that reactor build-ing inleakage at Unit 2 is approximately 5500 cfm when the building is iso-lated and SGTS is operating. System design capacity is 10,500 cfm. Another design feature is the fusible links which close the fire dampers on the SGTS j if the temperature exceeds a specified limit. This temperature limitation can eliminate the SGTS transport path for many accident sequences.

3.3.2 Modeling Approach A separate computer code was constructed from MAAP subroutines to model the reactor building, which can be divided into many nodes, to represent the major regions in the building. As noted earlier, the equipment transfer shaf t provides a path for natural circulation between the major volumes.

Estimates of the compartment temperature differences and the resulting natural circulation flows under accident conditions show the circulation flows between volumes to be large compared to the through flows. As a result, the building can be represented as a single volume. This provides for some conservatism in the analysis, since this somewhat overestimates the aerosol concentration in the upper region of the building and thus overestimates the release to the environment. In addition to the natural circulation flows between compart-l ments, the circulation flows within a compartment due to temperature differ-ences between the gas and the compartment walls can also be important. These effects are also included in the analysis. The building is assumed to be pressure equilibrated throughout the accident. As a result of this equilibra-tion, flow is driven through the reactor building as determined by the source coming from the primary containment following wetwell venting or containment failure, and the imposed flows resulting from the SGTS.

3-23 The SGTS, which provides a suction flow at each elevation is normal-ly fed by inleakage from the outside, with flow passing through the respective volume and into the ducts and filters in the system. The inleakage is deter-mined by the strength of the source. For example, following containment failure, the volume has a substantial source from the drywell which can be greater than the suction flow. As a result, the inleakage from the environ-ment to the volume would be reduced to zero with any excess flow leaking through the sheet metal siding or the opening of blowout panels. If the source flow is less than the required inleakage from the environment, then the inleakage is set to be the difference between that required by the SGTS suction from the volume and the source flow.

The physical processes occurring in each volume, including heat losses to the structural surfaces, thermal profile within the structures and steam condensation are treated in the same manner as the primary containment compartments in the MAAP code. These processes as discussed in detail in the MAAP User's Manual Volume 2 under the subroutine titles of PTCAL and HTWALL.

Since the reactor building volumes are coupled by the equipment shaft, small open hatches, etc., water accumulation on each elevation is assumed to drain to the lowest part of the reactor building and is neglected in the remainder of the calculation. It should be noted that with significant condensation, excess flow can be required from the environment back into the building in addition to the normal inleakage associated with the SGTS.

With the models for flow between compartments and condensation within individual compartments and the source term from the drywell following containment failure, the resulting aerosol agglomeration and removal can be assessed. This is also evaluated using the aerosol deposition model in MAAP discussed in Section 3.2.

3.3.3 Model Inputs For all sequences, whether or not SGTS is available, the reactor building is analyzed as a single node due to the coupling between compartments by natural circulation in the equipment shaf t. The parameters and values used to characterize the compartment are listed in Table 3.2.

3-24 Table 3.2 REACTOR BUILDING MODEL INPUTS Compartment Volume Surface Area SGTS Exhaust

• 3 (m ) (,2) (,3/sec) l 1 50,000 17.000 4.72

• SGTS exhaust is zero for the station blackout sequence. Also, the SGTS flow is zero if the temperature exceeds the limit for the fusible links on the fire dampers.

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3-25 3.3.4 Influence on Fission Product Release The reactor building completely surrounds the primary containment for the Mark I configurations, and as a result, would receive the fission products released following containment failure. Since this building is in a direct path for the release, it is an important part of tne fission product retention for this reactor design. In particular, it has a substantial influ-ence on retaining the fission products within the plant and limiting the

subsequent release to the environment. The large volume represented by the reactor building provides a substantial residence time for materials released from the primary containment and significant deposition occurs due to vapor condensation, gravitational settling, and steam condensation. Flow through the building is particularly important for sequences in which there is no SGTS flow, since the release to the environment is determined by this flow.

Inclusion of the SGTS with the associated flow increases the flow through the reactor building, thereby decreasing somewhat the material de-posited in the volume but provides for deposition within the SGTS filter system if it has not been overburdened by moisture. For sequences in which the SGTS is available, the principal release to the environment is determined by the flow through the filters and through the stack to the environment.

The influence on each individual sequence considered is discussed in Section 6. However, these all illustrate that the reactor building has a substantial influence on retention of fission products for the Mark I contain-ment design.

3.4 References 3.1 MAAP - Modular Accident Analysis Program, User's Manual Volume II, August,1983, 3.2 IDCOR Technical Report 15.lB " Analysis of In-Vessel Core Melt Progression," Vol. IV (User's Manual) and Modeling Details for the Fission Product Release and Transport Code (FPRAT), September,1983.

3.3 J. A. Gieseke, et al. , "Radionuclide Release Under Specific LWR Accident Concitions," Draf t Version of BMI-2104, Battelle Columbus Laboratories Report, July,1983.

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3-26 3.4 IDCOR Technical Report on Task 11.3, " Fission Product Transport in Degraded Core Accidents," December,1983.

3.5 IDCOR Technical Report, "MAAP Models for Aerosol Deposition,!' to be published.

3.6 Letter from H. E. .Townsend (GE) to E. L. Fuller (IDCOR), dated August 9, 1984.

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I 4-1 4.0 PLAllT RESP 0flSE TO SEVERE ACCIDErlTS Four base accident sequences were analyzed for Peach Bottom using MAAP to determine plant response and temperature and pressure challenges to i containment. These sequences, described below, in general are based on the sequences identified in Subtask 3.2.

Transient initiated sequences requiring a reactor shutdown, coolant makeup and subsequent decay heat removal have been identified in WASH-1400

[4.1] as potential dominant contributors to the core melt frequency. These types of sequences may have a broad spectrum of possible outcomes due to the wide variety of possible system performance characteristics and operator actions. For the Task 23.1 MAAP evaluation one specific set of boundary conditions and assumptions has been postulated for each sequence, j The base sequences are:

1. TW - Transient followed by loss of containment heat removal.

2. TC - Transient followed by failure of the reactor to scram and effective standby liquid control injection (without operator action to reduce power level).

3. Sj E - Medium break loss of coolant accident with failure of 'l emergency core cooling injection.

4. TQVW - Loss of offsite and onsite AC power.

The sequences analyzed in this section are low frequency core damage events and include no, or minimal, recognition of operator actions that would significantly delay the . progression toward core melt, prevent core melt entirely or mitigate the consequences of a core melt. This approach was taken to produce results which reflect multiple system failures similar to previous-ly identified dominant core melt sequences. Generally, only minimal operator actions to control selected plant systems are assumed for these events.

4-2 Consequently, the results presented here do not represent what would be expected to occur for the defined equipment failures and are extremely improbable. A more probable plant response to the specified equipment fail-ures is evaluated in Section 5. This later section includes in the sequence definition some of the actions which the operator would be expected to take in accordance with the emergency procedures. As a result of these actions the operator is able to terminate the event prior to core melt or significantly mitigate its consequences. Section 5 considers only some examples of the many actions available to tne operator to prevent or mitigate the accident.

A major objective of excluding mitigating operator actions in this analysis and allowing the events to progress unchecked was to provide the added perspective of defining the time windows available for operator inter-vention. The results clearly demonstrate that the operator has an extensive time period to implement primary or alternative actions that will successfully terminate or mitigate postulated severe accidents.

The plant parameters utilized to characterize Peach Bottom in these analyses are listed in Appendix A.

The following subsections discuss plant response for each severe accident sequence analyzed. In these analyses the containment ul tima te pressure capacity is based on the evaluation for the Browns Ferry fiark I design contained in the IDCOR Task 10.1 report [4.2], Containment Structure Capability of Light Water Nuclear Power Plants, which concludes that "it is felt that the Browns Ferry results are a sufficient representation of the containment capability" of Peach Bottom. The ultimate pressure capability was calculated to be 132 psia with the defined failure condition (twice the elastic strain) occurring at the " knuckle" between the cylindrical and spheri-cal parts of the drywell. (It should be noted that a detailed assessment of penetration behavior under high strain conditions was not part of the analy-sis.) Given the similarity between the Peach Bottom and Browns Ferry designs, this value is assumed to represent Peach Bottom. In cases where high tempera-tures in containment were reached before ultimate pressure, the containment was assumed to fail when the containment atmosphere temperature reached 1200'F. The Peach Bottom containment is made from high strength carbon steel.

4-3 The properties of this material will limit its ability to carry load at 1200*F. At this temperature the material strength is reduced to approximately 30-40% of its nominal value and will exhibit a significant creep rate under load [4.3]. Also, penetrations and/or penetration seals could have failed at these temperatures. Failure mechanisms and locations are addressed in Task 17.5.

A containment break size of 0.1 f t2 (0.2 ft2 for TC) is assumed because it permits depressurization of containment enabling airborne fission products to be transported out the break. This assumption is consistent with the concept of yield leading to rupture resulting in diminishing yield as the containment depressurizes.

4.1 Plant Response to the TW Sequence 4.1.1 Sequence Description This sequence is assumed to be initiated by MSIV closure followed by loss of all heat removal systems. High pressure injection (HPCI and RCIC) are initially available until high suppression pool temperatures are assumed to cause loss of these systems. Low pressure injection (LPCI and LPCS) are assumed to fail at containment failure. Control rod drive (CRD) flow remains on until the initial available inventory of water in the condensate storage tank (CST) is depleted. No operator actions to either prolong injection or utilize alternate means of injection are assumed to occur.

4.1.2 Primary System and Containment Response The timing of the key events for this sequence is summarized in Table 4.1. Plots of key parameters are presented in Figs. 4.1 through 4.5.

This sequence is characterized by heatup of primary containment since adequate containment heat removal is unavailable. This results in containment failure due to overpressurization followed by core melting and vessel failure which occur af ter the coolant injection systems are lost. Operator actions to utilize alternate heat removal paths or means of injection which draw from sources other than the suppression pool were neglected.

. -- .- . .. . . = . ,. __ . - .- , .

4 4-4 Table 4.1 PEACH BOTTOM - TW EVENT

SUMMARY

Time Event 0 Transient (MSIV closure) 4 sec Reactor scrammed J

4.5 min HPCI, RCIC on >

8.0 hr High SP temperature failure assumed for HPCI (200*F) 10 hr RCIC switched to CST on low level and assumed failed 14 hr CRD flow ceases due to assumed depletion of CST 15 hr ADS on, LPCI and LPCS injecting 25 hr ADS valves close

. 32 hr Containment failure (overpressurization); LPCI and J

LPCS lost 6

! 34 hr Top of core uncovered

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4-10 The sequence is assumed to be initiated by main steam isolation valve (MSIV) closure on all four main steam lines isolating the reactor from the power conversion system. The initial reactor power level is assumed to be at 100%. This results in a reactor scram signal followed by successful reactor scram within four seconds. The reactor stored energy and decay power are transmitted to the suppression pool through the safety relief valve (SRV) lines. This results in a continuous heatup of the suppression pool because it is assumed that the pool cooling mode of the RHR system is unavailable.

Reactor water level decreases due to boil-off which cannot be made up by the control rod drive (CRO) flow rate (111-177 gpm) due to the high reactor decay power at this time. High pressure injection systems (HPCI, RCIC) successfully come on in about 4.5 minutes and maintain required reactor water inventory.

HPCI suction is automatically transferred from the condensate storage tank (CST) to the suppression pool on pool high water level signal at 7.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. RCIC suction remains from the CST until a low CST level signal is received at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> when it is automatically transferred to the suppression pool. At 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the suppression pool reaches 200'F resulting in the assumed loss of the HPCI pump due to bearing degradation. RCIC injection is assumed to be lost for the same reason when its suction is transferred to the suppres-sion pool at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. At this time the post-scram CRD flow rate is suffi-cient to keep the core covered until the water source in the condensate storage tank is depleted at 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> assuming a CST inventory of 156,000 gallons. This inventory is conservative in light of the discussion below.

After CRD flow ceases, reactor water level boils down actuating the automatic depressurization system (ADS) at 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. This actuation is possible due to the existing high drywell pressure signal. This action permits low pressure systems (LPCI and LPCS) to maintain reactor water inventory.

The CRD pumps are normally aligned to take suction from the main condenser hotwell via the reject line. Should this suction source be unavail-able for whatever reason, the CR0 pumps will be provided suction through its connection with the condensate storage tank with no operator action required.

The volume of each CST reserved for ECCS use is 135,000 gallons. However,

4-11 this volume is not restricted from CRD pump use. An average CST inventory is estimated to be about 156,000 gallons. Although there are no specific plant procedures or operating limits governing the alignment of these various tanks they are arranged such that they are easily cross-connected. For example, the two CSTs are frequently intertied such that their water levels " float" to-gether. This mode of operation effectively doubles the condensate inventory available to the CRD pumps without operator actions. In addition, simple operator actions can be taken to interconnect the inventory of other various tanks increasing capacity to approximately 750,000 gallons.

As the containment pressurizes, the differential pressure between the ADS-SRV actuating gas and containment atmosphere decreases. When the drywell pressure reaches 110 psia, the differential pressure falls below the 5 psid required to hold the valves open, and the SRVs close. The operators are assumed to take no action to increase pneumatic supply pressure. This stops steam flow from the primary system to the suppression pool until the reactor vessel repressurizes and lif ts the SRVs on high vessel pressure (approximately 1100 psia). Therefore, the containment pressurization is essentially halted during this period of reactor vessel repressurization. LPCI and LPCS injec-tion ceases as the primary system pressure rises above their injection capability.

After the vessel is repressurized and the SRVs open and permit steam to flow to the suppression pool, the continuing heatup of the suppression pool results in pressurization of the containment until the assumed failure pres-sure of 132 psia is reached at 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. The influence of containment failure size on fission product distribution is addressed in Section 6. Containment failure is assumed to fail the low pressure injection systems. This conserva-tive assumption is consistent with WASH-1400 and several previous PRAs.

However, no mechanistic failure has been identified which would cause such a link between containment failure and core melt, as such, an appropriate probability has been assigned to this progression in the risk analysis (Task 21). Possible failures could potentially result from mechanical failures (piping movement and rupture) induced by containment failure, from electrical failures due to a steam environment in the reactor building, or possibly due to insufficient NPSH and cavitation in the low pressure pumps. Knowledge of i

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4-12 l

the actual failure mechanism is not required for this analysis but may be l necessary for the assessment of core damage probabilities and public health l risk. The low pressure ECCS pumps have been tested in a steam environment and have performed satisfactorily. Therefore, it should be noted this is a conservative assumption regarding pump performance and that the core would not be damaged if. the pumps continued to operate.

l As the containment depressurizes, the ADS-SRVs are assumed to reopen when the actuating gas-drywell atmosphere differential pressure reaches 25 psid. This occurs when the drywell pressure decreases to 90 psia at 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

With the assumption of no further injection af ter loss of low pressure injection, the gradual boil down of reactor water inventory results j in the top of the core being uncovered at about 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />. As the water level continues to boil down further uncovering the core, melting in the upper region of the core begins at about 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />. At approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> when 20% of the core inventory is molten and has collected on the core plate, the core plate is assumed to fail. This is equivalent to 153 fuel assemblies (764 assemblies in the core) acnieving a molten state. Since the assemblies are supported from below by the CR0 guide tubes, melting of the material and the subsequent flow into the bypass region will begin to load the core support plate which is only designed for transverse loads. Accumulation of this molten mass of U0 2 and the associated Zircaloy is assumed to fail this struc-ture and allow the molten debris to flow into the lower plenum. The influence of this assumption on the overall effects is discussed in the uncertainty analysis report for Subtask 23.4.

f The reactor pressure vessel fails wichin i few minutes after the core plate fails due to rapid melting of the instrument and (RD tubes that penetrate the bottom vessel head. Vessel failure and. the subsequent genera-tion of steam as the core debris mixes with water on the base mat results in a secondary pressure rise in containment from 90 psia to 96 psia. Heatup of the drywell occurs with temperatures reaching 1500*F at approximately 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br />.

The slow boil down of reactor water results in a relatively long period during which the fuel cladding is at high temperature in a steam

4-13 environment resulting in oxidation. During the period from 36 to 40 rours into the sequence, approximately 430 lbn of hydrogen are generated.

This analysis demonstrates that plant operators have a significant amount of time (nearly two full days) to take action to prevent fuel melting and preserve containment integrity. Appendix B includes several plots showing results for this sequence.

A number of alternative means of arresting this sequence exist which were not included in this analysis but which are explicitly called for in the Peach Bottom plant specific procedures. Section 5 discusses some of these alternatives and their impact on the course of the accident.

4.2 Plant Response to the TC Sequence (Without Operator Action to Reduce Power Level) 4.2.1 Secuente Description This sequence is assumed to be initiated by MSIV closure followed by failure of the reactor protection system to scram the reactor as well as failure to initiate standby liquid control (SLC). Successful recirculation pump trip occurs and both high pressure (HPCI and RCIC) and low pressure (LPCI and LPCS) injection systems are assumed to be available until either high

~

suppression pool temperatures are reached or wetwell venting ~is initiated.

CRD flow remains or. until the inventory in the CST is depleted. The conden-sate pumps are assumed to be unavailable. No operator actions are assumed to either control power level by reducing reactor water level or to utilize alternate means of injection. The operator is assumed to follow the written procedures for venting the wetwell to protect against containment overpres-surization. For this sequence, the steam release would exceed the capacity of the SGTS system and would be discharged to the reactor building. Temperatures within the building would exceed that required to fail' the fusible links on the fire dampers of the SGTS, thereby isolating the system. In this sequence, it is assumed that pumping capacity is lost after wetwell venting is initi-ated. The low pressure ECCS pumps address NPSH, and their motors have been tested in a. steam environment and have performed satisfactorily. Therefore.

4-14 it should be noted that this is a conservative assumption regarding pump performance and that the core would not be damaged if the pumps continued to operate.

A similar sequence without wetwell venting was analyzed to provide an additional conditional consequence input to the risk analysis in Task 21.

The timing of iey events is identical to those of the venting case up to and including the time of core uncovery. Primary containment fails as a result of overpressurization at 1.35 hrs. At this time the low pressure injection systems an a assumed to fail. The core begins to melt at 3 hrs. and subsequent vessel failure occurs at 3.9 hrs. A comparison of the venting and non-venting TC sequences in terms of fission product release and distribution can be found in Section 6.

4.2.2 Primary System and Containment Response The timing of the key events for this sequence is summarized in Table 4.2. Plots of key parameters are presented in Figs. 4.6 through 4.12.

This sequence is characterized by rapid heatup of the suppression pool result-ing in the assumed loss gf.,the high pressure injection systems, wetwell venting and loss of low pressure injection systems. This is followed by core melting and vessel failure.

As in TW, this sequence is assumed to be initiated by MSIV closure isolating the reactor from the power conversion system. However, it is assumed that the reactor fails to scram and that subsequent initiation of standby 11guld control is not attempted or is unsuccessful. It is also assumed that condensate flow is unavailable. Successful recirculation pump trip followed by initiation of HPCI and RCIC at approximately 1-1/2 minutes results in an estimated reactor power level of 18 percent of rated. This is based on the power level required to boil off reactor water at a rate equal to the total injection rate at this time. If the power level were greater than this, the reactor water level would boil down resulting in a power level reduction until this balance was achieved. This power level was confirmed to be in the correct range by running RETRAN with a neutronics model, and assumes no operator action is taken to throttle injection to reduce power level.

4-15 Table 4.2 PEACH BOTTOM - TC i

EVEf4T

SUMMARY

Time Event O Transient (MSIV closure) 3 sec Failure to scram 1.7 min HPCI, RCIC on '

26.4 min HPCI assumed unavailable (SP at 200*F) 38 min ADS on 40 min LPCI, LPCS on (reduced flow) but held at the Level 8 trip set point 54 min RCIC assumed unavailable 1.2 hr ADS valves close 1.3 hr Top of core uncovered 1.3 1r Open wetwell vent 1.3 hr LPCI, LPCS assumed lost 1.5 hr ADS valves reopen 3 hr Start of core melting i

3.8 hr Vessel failure 6.9 hr CRD flow ceases 12.8 hr Containment failure due to high temp. in the drywell

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4-22

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4-23 Steam flow through the SRV lines results in suppression pool heatup.

At approximately 26 minutes, the suppression pool reaches 200 F and is assumed to result in the unavailability of the HPCI system.

RCIC injection is insufficient to maintain water level while the core is at 187, power. Therefore, the level boils down resulting in low pressure system (LPCI, LPCS) initiating signals at 36 minutes, ADS actuation at 38 minutes and effective low pressure injection at 40 minutes. Before low pressure injection can be established, reactor water level drops to the top of the core, but no fuel overheating occurs. Reactor power level is linearly reduced to six percent of normal [4.4] as the water level in the downcomer region decreases from a level 20 ft abov6 the core inlet down to the top of the jet pump core which is reached at 37 minutes. The boiled-up height in the core at this time is near TAF. After this time the power level continues to be a function of reactor water level and is balanced by the primary system pressure, the resulting injection rates of LPCI and LPCS and the relief capacity through the ADS valves. This results in the water level hovering near the top of the core.

RCIC suction remains from the CST until 54 minutes when it is automatically transferred to the suppression pool and assumed lost due to bearing degradation. However, low pressure injection is sufficient to main-tain reactor water level near the top of the core.

As the containment pressure rises due to suppression pool heatup, the SRVs previously actuated by ADS close when the drywell pressure reaches 110 psia for the reasons discussed under the TW sequence. This occurs at 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, causing rapid repressurization of the reactor vessel and loss of injection by the low pressure systems (LPCI, LPCS). The rapid vessel repres-surization and lifting of the SRVs on high reactor pressure result in con-tinued containment pressurization until the wetwell is vented through a vent area of'0.18 m2 (1.98 ft )2 at a pressure of 115 psia which is reached at 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The vent size corresponds to the opening of a 2 in., 6 in. and 18 in.

vent lines from the wetwell. This depressurization and steam flow is assumed to cause loss of the low pressure injection systems as a result of the same possible mechanisms discussed for the TW sequence. As the containment l  :

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4-24 depressurizes, the ADS valves are assumed to reopen when the pressure de-creases to 90 psia (1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />), as discussed for the TW sequence.

With only CRD flow remaining, reactor water level boils down result-ing in the start of core melting at about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. When 20% of the core has melted core plate failure is assumed resulting in vessel failure at 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Approximately 300 lbm of hydrogen are generated from cladding oxidation.

Following reactor vessel failure, the core debris disperses over the pedestal and drywell floors. Drywell heatup begins at about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> reaching a temperature of 1200*F at approximately 12.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This occurs as a result of radiative and convective heat transfer from the core debris which lies on the pedestal and drywell floors in a relatively thin flat geometry with large surface area. The rate of drywell heatup is dependent on both core debris dispersal geometry and the mass of heat sinks in the drywell. Recent data indicate that drywell heat sink mass, particularly for gas temperatures greater than 500*F could be significantly greater than that used in this analysis. Sensitivity to this and core debris geometry is explored in Task 23.4 with the conclusion that the timing of the sequence is changed but the ultimate fission product release to the environment is not. Appendix B includes other plots for this sequence.

The influence of natural circulation within the primary system is illustrated by Figs. 8.17 and B.18 which show the structure temperatures and decay power associated with the various primary system nodes. As illustrated by these figures, the volatile fission products are initially released into the upper plenum and deposited there as a result of both vapor condensation and gravitational sedimentation. Over an extended time interval, the upper plenum structure temperature increases and circulation is set up between the upper plenum, the downcomer and the core. As the temperature continues to increase, material is transported from the upper plenum into the downcomer, where it is again retained and remains in this locale until all of the water is vaporized in the downcomer. Following vaporization of the water, the material deposited in the downcomer heats the structural mass, vaporizes and i is transported throughout the primary system and eventually into the contain-ment. It is this release into the containment that is eventually transported

4-25 into the reactor building with a small fraction being released to the environment.

The analysis of this sequence, which assumes no early operator action to reduce power level, indicates that operators have approximately 1/2 hour after the core is uncovered to recover the core and prevent fuel melting.

The analysis also indicates that if fuel melting and vessel failure did occur, operators would have approximately 12 additional hours to prevent containment heatup above 1500 F and mitigate those releases resulting from revaporization in the drywell through the use of such systems as HPSW for flooding or con-tainment sprays, condensate pumps, and CRD flow.

4.3 Plant Response to the S Ej Sequence 4.3.1 Sequence Description 2

This sequence is assumed to be initiated by a 0.1 ft break in the main steam line inside containment (drywell). High pressure injection (HPCI and RCIC) and low pressure injection (LPCI and LPCS) are assumed to be un-available. Injection from the condensete pump is also assumed to be unavail-able, but CRD flow is available until the inventory in the CST is depleted.

It is assumed that suppression pool cooling is manually initiated at 10 minutes into the sequence. No actions by the operator to establish alternate means of injection to the core are assumed.

4.3.2 Primary System and Containment Response The timing of tne key events for this sequence is summarized in Table 4.3. Plots of key parameters are presented in Figs. 4.13 through 4.17.

In general this sequence is characterized by loss of makeup to the core resulting in fuel melting and vessel failure. However, suppression pool cooling is available preventing the containment from overpressurizing on steam. Containment failure occurs due to an overtemperature condition in the drywell before sufficient noncondensable gas generation has occurred to overpressurize the containment.

4 i

4-26 Table 4.3 PEACH BOTTOM - S;E EVENT

SUMMARY

Time Event 2

0 Break in steam line (0.1 ft )

6 sec Reactor scrammed 80 see MSIVs closed, feedwater tripped 10 min Suppression pool cooling on -

l.07 hr Automatic depressurization on (ADS) .

1.15 hr Top of core uncovered 2.6 hr Start of core melt 3.5 hr Vessel failure 15 hr CRD flow ceases 23 hr Containment failure (overtemperature)

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4-32 This ' sequence is initiated by a 0.1 ft2break in the primary system at the elevation of the main steam lines. This causes rapid containment

! pressurization to above the 2 psig set point for reactor scram, resulting in a

^

successful scram within 7 seconds of initiation of the break. As the primary system is rapidly depressurized through the break, a low reactor pressure signal for MSIV closure is received, and closure occurs by 84 seconds isolat- l ing the reactor from the power conversion system and shutting down feedwater.

It is assumed that the condensate pumps fail to inject through the feedwater pumps. As reactor water level boils down the high pressure (HPCI and RCIC) and low pressure (LPCI and LPCS) systems are assumed to fail to inject water 4

into the core. Post-scram CRD flow (177 gpm) is sufficient to keep the core covered until automatic depressurization (ADS) is actuated at 1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> due to high drywell pressure and low reactor water level, further depressurizing the primary ' system. The emergency procedures would also direct the operator to depressurize. There is a low pressure pump permissive signal for ADS because the residual- heat removal (RHR) systems in suppression pool cooling mode were manually initiated at 10 minutes. This depressurization causes reactor water level to drop to near the bottom of the core at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. As reactor water level boils down, the fuel . overheats and oxidation of cladding occurs. The

core is heated up leading to fuel melting at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and vessel failure at 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Approximately 240 lbn of hydrogen are generated between 1.5 and 2

! hours.-

- The pressure in the drywell rises initially to about 40 psia as a result of the steam line break. The steam is quenched in the suppression pool and the pressure is about 45 psia at the time of vessel failure. Noncon-e densable gas generation from cladding oxidation and initial ablation of concrete in the pedestal and drywell results in a further pressure increase to about 55 psia. Heatup of the drywell atmosphere and structure from radiation

! and convective heat transfer from the core debris commences, after the loss of CRD flow, at about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />.

i a

When the temperature reaches 1200'F at 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> the containment is 2

assumed to fail with a 0.1 f t break. The containment pressure is approxi-mately 55 psia when the failure occurs, dropping to atmospheric pressure at about 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. Appendix B includes several plots for this sequence. l I

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4-33 The analysis of this sequence demonstrates that operators have approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to establish alternate injection to the reactor prior to fuel melting. If vessel failure did occur, operators would have an additional 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to take action to preserve containment integrity and mitigate releases. .

4.4 Plant Response'to the TQVW Sequence 4.4.1 Sequence Description This sequence is assumed to be initiated by loss of all off-site and on-site AC power (station blackout). This results in reactor scram and foss of the power conversion system. It is assumed that DC power is available for a period of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> permitting control of the steam driven HPCI and RCIC systems to maintain injection to the core for this duration. After 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, no further injection is assumed available and no operator action to utilize alternate sources are assumed to occur.

4.4.2 Primary System and Containment Response The key events for this sequence are summarized in Table 4.4. Plots of key parameters are presented in Figs. 4.18 through 4.22. In general, this sequence leads to core melt and vessel failure due to lack of coolant injec-tion followed by containment failure approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> later.

Loss of all AC power results in an immediate reactor scram signal followed by successful reactor scram within 4 seconds. The power conversion system is not available and the stored energy and decay power are transmitted to the suppression pool through the SRV lines resulting in suppression pool heatup. The only coolant injection assumed available to the reactor is through the HPCI and RCIC systems, because these pumps are steam turbine driven. All other injection pumps require AC power.

HPCI and RCIC maintain reactor water inventory until 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the accident when control of these systems is lost due to the assumption of battery depletion. Within the time frame of 2 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following a station

4-34 Table 4.4 PEACH BOTTOM - TQVW EVENT

SUMMARY

Time Event 0 Loss of off-site and on-site AC power 8 sec Reactor scrammed 5 min High pressure injection on (HPCI, RCIC) 6 hr HPCI, RCIC off (loss of DC power) 8.4 hr Top of core uncovered 11.4 hr Start of core melt 12 hr Vessel failure (high pressurc) 18 hr Containment failure (overtemperature)

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4-40 blackout, there may be a number of challenges to continued successful coolant injection, including:

1. High temperature in the areas of ECCS pumps, instrumentation, power, and the control room;

2. Depletion of the batteries;

3. High suppression pool temperatures;

4. Depletion of the CST.

In this analysis it is assumed that heatup of the rooms containing the HPCI and RCIC systems does not cause system failure because of the reasons dis-cussed bele.4. Af ter HPCI and RCIC become unavailable, the reactor water level

' boils down uncovering the top of the core at about 8.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During the boil down, approxiciately 500 lbm of hydrogen are generated due to cladding oxida-tion. The core begins to melt at approximately 11.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and the melting progresses until core plate and vessel failure occur at approximately 12.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Prior to vessel failure the primary system is at a pressure of 1100 psia, due to the assumption that the operator does not act to open SRVs and ADS actuation does not occur due to the lack of DC power. The pressure in the primary containment rises sharply at vessel failure from 30 psia to 90 psia due to flashing of residual water in the vessel and the generation of noncon-densable gases from initial concrete ablation and additional cladding oxida-tion. After the residual water from the vessel is vaporized there is no water available to quench the core debris. Thus concrete ablation is initiated, but at a slower rate, generating additional gases which continue to pressurize the containment.

The core debris during this time is dispersed over the pedestal and drywell floors in a geometry that results in substantial thermal radiation to the drywell atmosphere and structure. There is a significant temperature rise in the drywell commencing af ter 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. The assumed failure temperature of 1200*F is reached at approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> wnen the containment pressure is at 105 psia. Following the failure, the containment pressure decreases to

4-41 about 25 psia at 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />. Appendix B includes several plots for this sequence.

The indicated progression of this sequence, is not likely to occur if one considers possible operator actions due to the following reasons:

e Explicit plant procedures exist for the conservation of DC capacity during a loss of AC power. Such actions would extend the availability of DC power considerably beyond the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> assumed in this analysis.

s e For loss of AC power events, plant procedures require that the HPCI and RCIC systems be placed in a manual mode of operation such that any instrumentation failures which could result from elevated room temperatures will not adversely affect system operation. In addition, it is expected that opening of the ECCS compartment doors would provide sufficient room cooling to prevent equipment failure.

e Even if HPCI and RCIC become unavailable, additional means of vessel makeup are available which do not rely on plant AC power, (i.e. fire trucks or diesel driven pumps through HPSW/RHR), Plant emergency procedures call for the use of this type of equipment under appropriate conditions.

e The analysis of the sequence described in this section indi-cates that the operators have over 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to restore power or establish an alternative means of injection prior to fuel melting.

e If fuel melting and vessel failure did occur, the operator would have an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to take action to maintain contaimnent integrity to mitigate releases.

4-42 4.5 References 4.1 Reactor Safety Study, WASH-1400, NUREG/75-Oll4, 1975.

4.2 IDCOR Technical Report on Task 10.1, " Containment Structural Capa-bility of Light Water Nuclear Power Plants," July,1983.

4.3 "An Evaluation of the Elevated Temperature Tensile Creep Properties of Wrought Carbon Steel," ASTM 0S11 and ASTM 0S11 Supplement 1.

4.4 " Reducing BWR Power by Water Level Control During an ATWS, A Quasi-Static Analysis," NSAC 69, May, 1984.

e J

5-1 1-5.0 PLAflT RESP 0fiSE WITH REC 0VERY ACTIONS Section 4 ' discusses plant response to the identified accident sequences. The results demonstrate that there is substantial time (several hours to days) for oper. '~ actions to be implemented- to prevent core degrada-tion or mitigate its consequences. This section discusses the effects of some

-examples of possible operator actions that could be taken during these. acci-dent 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.

4 5.1 Possible Actions I'

Operators are trained and emergency procedures are written so that a certain function is achieved by utilizing whatever relevant systems are

! available, (e.g. provide sufficient water injection to the reactor with

, control rod drive flow). The BWR Emergency Procedure Guidelines (EPGs) and the plant specific implementation of these guidelines at Peach Bottom 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. Tables 5.1 and 5.2 summarize some of t..:-

various means of injecting water into the reactor pressure vessel (RPV) as well as sources of water that are available to an operator. Included in this ,

table are sources of water, system flow ranges and power requirements for system operation. Similarly, Table 5.3 summarizes some of the various means of reactor and containnent cooling, and the type of power required for these

, systems. Table 5.4 summarizes the provisions for venting of containment to prevent overpressurization.

The operator actions analyzed were selected from the possibilities resulting from the availability of these systems for each sequence and are included in existing emergency procedures at Peach Bottom. The selection 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 available p

systems.

i l

, - , r = . ,~ - . - - , , - - -- - -, -

Table 5.1

!!!JECTION TO RPV System Pressure Flow Source Power Required Feedwater High 13 mlb/hr Condensate Pump Main Steam and AC HPCI High 5,000 gpm CST /SP Reactor Steam and DC l RCIC High 600 gpm CST /SP Reactor Steam and DC CRD High 55 - 210 gpm CST /Hotwell AC (Diesel or Offsite)

SLC High 43 gpm SLC Tank AC (Diesel or Offsite)

HPSW* < 250 psig 0 - 18,000 gpm River /Emerg. AC (Diesel or Offsite) T Cooling Tower N LPCS < 280 psig 0 --16,000 gpm- SP/ CST AC (Diesel or Offsite)

LPCI < 240 psig 4,000 - 40,000 gpm SP AC (Diesel or Offsite)

Condensate Pump < 650 psig 0 - 30,000 gpm Hotwell AC (Offsite)

• Intertie between Units 2 and 3.

SLC 5,000 gal.

CST,: Tank: 156,000 gal. (nominal water volume)

Hotwell: 90,000 gal.

NOTE: This' volume can De replenished from various sources including the other unit CST, the refueling water storage tank, the torus dewatering storage tank, and the fire protec-tion water system (see Table 5.2).

5-3 Table 5.2 TANK CAPACITIES AND REPLENISHING SOURCES Tank Capacities

• Unit 2 Condensate Storage Tank 200,000 gallons Unit 3 Condensate Storage Tank 200,000 gallons Refueling Water Storage Tank 450,000 gallons Torus Dewatering Tank 1,200,000 gallons Hotwell 90,000 gallons Replenishing Sources Makeup Demineralizer 120 gpm Torus Water Cleanup Pump 100 gpm Portable Pumps Variable (river water or other sources)

• Although there are no specific plant procedures or operating limits governing the alignment of these various tanks they are arranged such that they are easily cross-connected. For example, the two CSTs are frequently intertied such that thcir water levels " float" together. This mode of operation effectively doubles the conden-sate inventory available to the CRD pumps without operator actions.

In addition, simple operator actions can be taken to interconnect the inventory of all of these tanks (normally approximately 750,000 gallons).

B s.

Table 5.3 fiPV/ CONTAINMENT COOLING

-i System Capacity Power Required PCS 25% Power AC (Offstte)

(Turbine Bypass).

RHR - Shutdown' Cooling RHR - WW or-DW Sprays (Combined) 4 x 20 MW AC (Diesel or Offsite)

.RHR - SP Cooling HPSW - WW or DW Sprays AC (Diesel or Offsite)

Containment Venting Variable None.

(Various Paths from WW or DW)

DW Coolers 2 MW AC (Diesel or.Offsite)

Suppression Pool Water Exchange 100 gpm AC (Offsite) i RWCU Hx 2 x 4 MW AC (Diesel or.0ffsite)

RWCU Letdown to Condenser < 190 gpm

---

._x _ _ _ - - - - _ _ - - - _ _ _ _ _ _ . --__-__.__---________--x

5-5 Table 5.4 CONTAINMENT VENTING PROVISI0!is (AVAILABLE VEtiT PATHWAYS) 2" Torus Vent Line (Through SGTS) 6" Torus ILRT Connection Deflated Seals on Torus Vent Line Deflated Seals on Torus Supply Line 18" Torus Vent Line 18" Torus Supply Line 2" Drywell Vent Line (Through SGTS) 6" Drywell ILRT Connection 2-3" Drywell Sump Drain Lines Deflated Seals on Drywell Vent Line Deflated Seals on Drywell Supply Line 18" Drywell Vent Line 18" Drywell Supply Line

5-6 5.2 Sequence 1 (TW)

From Section 4.1, it is evident that for this sequence there is a substantial length of time (at least 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />) to recover suppression pool cooling or the power conversion system before the low pressure coolant injec-tion pumps are assumed lost at containment failure. If either system is recovered before this time, pool heatup would not continue, low pressure injection would be maintained and the core would be covered by water. There-fore, no core overheating or damage would occur.

However, assuming that suppression pool cooling and the power conversion system are not recoverable, and high and low pressure injection are lost as indicated in Section 4, other systems are available to provide makeup to the reactor, and vent the primary containment. The post-scram CRD flow is sufficient to keep the core covered several hours into the sequence when decay power is reduced, and the condensate storage tank (LST) can be refilled from various sources. This sequence was evaluated assuming that the CST water volume was sufficiently replenished to provide continuous makeup to the core, and that the contaimnent was vented to prevent overpressurization. Table 5.5 summarizes the events for the TW sequence with these actions.

The containment is vented from the wetwell with an effective area of 2

0.22 ft . This is done in accordance with procedures when the drywell pres-sure reaches 70 to 115 psia (115 psia is used in this analysis) at 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

With the high drywell pressure of > 100 psia before wetwell venting is initi-ated, the ADS valves close as shown in Figs. 5.1 and 5.2. As the containment depressurizes, the ADS valves reopen when the pressure decreases below 90 psia or as directed by the emergency procedures, as discussed in Section 4.1. The containment pressure decreases over several hours as indicated in Fig. 5.1, and the CRD flow is sufficient to keep the core covered and no core degrada-tion occurs. A plot of reactor pressure vessel water level is illustrated in Fig. 5.2. As the containment depressurizes through venting, the heated suppression pool flashes and the steam flows through the venting lines being u tilized. This boil off prevents the primary containment from becoming

' overfilled due to the continuous makeup from an outside source. The water j

5-7 Table 5.5 TW WITH SELECTED OPERATOR ACTIONS EVENT

SUMMARY

Time Event

.0 Transient (MSIV closure) 4 sec Reactor scrammed 4.5 min HPCI, RCIC or.

8.0 hr High SP temperature failure assumed for HPCI (200"F) 10 hr RCIC assumed failed after auto switchover to suppression pool 18 hr ADS on, LPCI and LPCS injecting 29 hr ADS valves close 35 hr Containment venting begins; LPCI and LPCS are assumed unavailable for demonstration purposes 36 hr ADS valves open

- No core degradation No containment failure

TlJ -- PEACH BOTTOM (INTERVENTIONI 70 , --....---.. . :- ~ - -.-.--..-. .. .... . .- .-. - _ , y--.-...

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Fig. 5.1 Pressure in drywell.

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Fig. 5.2 Reactor pressure vessel water level.

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5-10 gradually depressurize the reactor as the suppression pool heats up, as per the emergency procedures. With injection throttled and the primary system depressurized, reactor power level (steaming rate) fluctuates between 6 and 10 percent depending on water level [5.1].

As the primary system depressurizes, the low pressure injection systems are overridden to prevent RPV reflooding. At approximately 42 min-utes, the suppression pool reaches 200 F resulting in the assumed degradation of the HPCI and RCIC pump bearings causing loss of these systems. However, since the reactor is depressurized at this time, the LPCI system is manually initiated and throttled to maintain core water level near the top of active fuel (TAF). This is regulated as water level in the downcomer region, and the boiled-up height in the core may vary somewhat.

The sustained power generation of about 4% results in suppression pool heatup causing containment pressure to rise rapidly as indicated in Fig.

5.3. It is assumed that the operator increases instrument air pressure in order to keep the SRVs open during containment pressurization. The EPGs require containment venting if the main condenser cannot be restored and suppression pool cooling is inadequate. It is assumed that the containment is vented from the wetwell with an effective area of 2.0 ft2. In accordance with procedures the venting occurs when the drywell pressure reaches 70 to 115 psia (115 psia is used in this example) at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in order to reduce the pres-sure loading and prevent structural failure. It is assumed that makeup to the vessel and flow regulation to maintain level at the top of active fuel (TAF) continues through the use of LPCI, LPCS, HPSW, or other available sources.

The event summary is shown in Table 5.6. Reactor water level in the shroud is shown in Fig. 5.4.

This modified sequence indicates that these operator actions prevent fuel damage and containment failure. A safe stable state is achieved as long I as the alternate means of injection is maintained providing substantial j additional time to ultimately shut down the reactor and recover a heat sink.

The initial actions to reduce power level by lowering reactor water level affects the rate of suppression pool heatup early in the sequence. The time

9 TC -- PEACH BOTTOM (INTERVENTION) i i

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5-12 L Table 5.6 -

TC WITH SELECTED OPERATOR ACTIONS.

EVENT

SUMMARY

Time Event 0 Transient (MSIV closure) 3 sec Failure to scram 1.5 min HPCI, RCIC on 5 min Suppression pool cooling initiated 8.2 min Injection flow throttled to lower level to TAF 10 min Manual reactor depressurization begins 42 min HPCI, RCIC assumed failed (SP at 200*F) 42 min Throttled low pressure injection flow established 2.5 hrs Containment venting begins

- r No core degradation I

i

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i l Fig. 5.4 Reactor water level.

5-14 to reach 200 F in the pool is calculated to be 27 minutes for the base and 42 minutes for the case with the interventions described in this section.

The EPGs emphasize a number of actions for the operator if these functions are not already accomplished:

e Trip the recirculation pumps, o Manually scram the reactor, e flanually insert individual control rods.

e Maintain or re-establish the main condenser as a heat sink.

e As the suppression pool heats up, depressurize the RPV to maintain RPV pressure and suppression pool temperature below the heat capacity temperature limit.

e Minimize the power generated in the reactor by virtue of lowering the reactor water level, e Maintain adequate coolant injection with the systems identified in Table 5.1.

e Vent containment when required to maintain containment integrity.

e Initiate SLC flow with the SLC pumps or inject water into the vessel with other available pumps.

5.4 Sequence 3(S)El As defined, the base sequence analyzed in Section 4.1 did not take credit for condensate pump makeup to the reactor because operator interaction is required to regulate reactor water level. However, with this operator interaction, the condensate pumps can easily makeup reactor water inventory i

5-15 because the reactor has depressurized through the break and manual regulation of level is straightforward. Since suppression pool cooling is also operat-ing, the reactor and containment are maintained in a coolable nondegraded state. However, if one assumes that the condensate pumps and all other low pressure injection pumps are not or cannot be utilized, the following analysis demonstrates the effectiveness of another operator action to mitigate the

. consequences.

The analysis of the S jE base sequence indicates containment failure due to very high drywell temperatures which were calculated due to heatup from the corium on the drywell floor. These high temperatures (1200*F at 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />) could result in containment failure in the drywell. Therefore, this sequence was reanalyzed assuming an operator action of turning on drywell sprays at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> into the event (approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> af ter vessel failure).

The event summary for this sequence is listed in Table 5.7. The timing of the events through the loss of CRD flow are the same as those presented in Section 4.1. However, initiation of the sprays at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> results in a pressure and temperature decrease in containment. The spray water accumulates on the containment floor where it quenches the core debris.

Any steam formation or overflow into the downcomer pipes is directed back to the suppression pool where it is cooled and recirculated to the drywell sprays. A permanently coolable configuration is achieved. Drywell pressure is reduced to less than 25 psia and the drywell gas temperature drops to less than 130*F very rapidly. Drywell pressure and temperature are plotted in Figs. 5.5 and 5.6 respectively. Maximum temperature and pressure are 950*F and 35 psia respectively and occur before drywell spray initiation. These results indicate that the challenges to containment can be mitigated and containment integrity will be maintained if this operator action is taken.

1 This sequence was chosen to identify that the operator action of initiating drywell sprays could substantially alter the course of the accident sequence by reducing temperatures and pressures inside containment.

The alternate operator actions contained in the EPGs include:

_ _ _ - - -,, _,-,,__,_.,n

5-16 Table 5.7 S jE-WITH SELECTED OPERATOR ACTIONS EVENT

SUMMARY

Time Event 2

0 Break in steam line (0.1 ft )

6.8 sec Reactor scrammed 84 sec MSIVs closed, feedwater tripped 10 min Suppression pool cooling on 1.05 hr Automatic depressurization on (AOS) 1.13 hr Top of core uncovered 2.5 hr Start of core melt 3.4 hr Vessel failure 15 hr CRD flow ceases 22 hr Drywell sprays on Containment does not overpressurize Containment does not overheat 1

l l

S1E -- PEACH BOTTOM (INTERVENTION) 60- -

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I Fig. 5.5 Pressure in drywell.

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Fig. 5.6 Temperature in drywell.

i I

i l

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5-19 e Operator actions to align the backup CRD pump which will ensure approximately 200 gpm. This flow rate is adequate for core cooling beyond 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, e Operator action to use the SLC pumps to provide added coolant injection to augnent other coolant injection sources, e Align and use HPSW and/or diesel fire pumps to provide water from outside sources when the reactor can be depressurized.

e When RPV water level has dropped to approximately 1/4 core height depressurize the RPV in the manner specified in the EPGs to maximize steam cooling in the core and lengthen the time to restore injection systems prior to core degradation.

5.5 Sequence 4 (TQVW)

This sequence was reanalyzed assuming the diesel generators are recovered prior to the initiation of fuel melting. A time of 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after loss of power was selected for power recovery based upon the analysis in Section 4.1 which indicates that no action is necessary for at least 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

Table 5.8 summarizes the events for this sequence with this power restoration assumption.

As in the sequence discussed in Section 4.1, it is assumed that DC power becomes unavailable at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following station blackout. This results in the loss of high pressure injection also at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Two hours later the top of the core becomes uncovered and the water level drops to near four feet below the top of active fuel when power is assumed to be restored. The steady boil down and flow of steam to the suppression pool between 6 and 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> results in heatup of the pool to approximately 225'F at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Therefore, when power is recovered, HPCI and RCIC are assumed to fail due to degradation of the pump bearings. The sequencing of the low pressure makeup systems and the time to depressurize the reactor result in core flooding within 5 minutes following power restoration. Reactor pressure vessel water level for this sequence is illustrated in Fig. 5.7.

5-20 Table 5.8 TQVW WITH SELECTED OPERATOR ACTIONS EVENT

SUMMARY

Time Event 0 Loss of off-site and on-site AC power 4 sec Reactor scrammed 5 min High pressure injection on (HPCI, RCIC) 6 hr HPCI, RCIC off (loss of DC power) 8.4 hr Top of core uncovered 9 hr Onsite AC and DC power restored HPCI, RCIC on and assumed lost on high SP temperature Low pressure ECCS pumps on (1 loop in SP cooling mode) 9 hr, 2 min ADS Initiated 9 hr, 4 min LPCI, LPCS injecting 9 hr, 6 min Core flooded and level maintained No core melt No containment failure

TOVid -- PEACH BOTTOM (INTERVENTIONI EB- 7 r- - 7r t 7 7 q

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$$ .._2. _ . -

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i i '

i i i

\

m u w -

l i .

E 40-i.

I I -

1,i i f m

• i d h

[*g j

-

t

}O- ------.;-..---..---

'g ! -..

t- .+-- -

i q T OP, O F A C.T I V E F.U E L q

1 PDuER REI T DRE D i

FUEL

,30 ....9-.,.

._..._.7 9

B _U.T,10 M OF9A C 1 1 V E.

. p.._._...- ,

,, .,...,.. ,...,.. ,.. , ..i

...i...i..

0 2 4 6 8 18 12 14 16 19 28 TIE (HOURS)

Fig. 5.7 Reactor pressu e vessel water level.

1

5-22 While the core is uncovered, an insignificant amount of oxidation of cladding occurs and none of the core no' des heatup sufficiently to cause fuel mel ting.

Containment pressure, illustrated in Fig. 5.8, peaks at 26 psia at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, then decreases af ter restoration of suppression pool cooling. Suppres-sion pool temperature peaks at 225'F at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> and is illustrated f', Fig.

5.9.

Results of this analysis clearly demonstrate that a safe stable state is achieved even if power is nct restored until 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> af ter' station blackout from an initial reactor power of 100%. The integrity of the primary containment boundary is not challenged.

Operator actions called for in the emergency procedures that could occur during this event are:

e Nonessential loads on the DC buses would be shed in order to preserve the DC batteries for as long as possible into the accident sequence, e Operation of HPCI or RCIC in manual mode.

e The diesel driven fire pump or other pumps that do not require AC or station DC power, (e.g. fire trucks) can be aligned for injection into the HPSW lines. From that point the diesel fire pump can: (a) spray the containment or (b) inject to the RPV for coolant makeup. (The effectiveness of containment spray was shown in the Sj E accident sequence of Section 5.4.)

e Restoration of off-site and on-site AC power sources would proceed in parallel with other actions. The likelihood of not being cble to recover any AC power within the specified time period is extremely low [5.2].

~ - - - - - -

4 l

l l

TQV(J -- PEACH BOTTOM (INTERVENTION) m

se; r r , r -r r ,

PDWER RESTORED g 25- -

/D +-- f

/

i/

~

28- e F 4-- 4 + 4 4

., i

. .e

<t .

i i i un -

i  : . .

a.15- . .

e

- +- t +- .r i +

--- e i

~

. u.s ro w

8 a.

-  ! i i i i 9_- . .p.........+.. . . . .

5: .

. ...+. .....-......;.....-.- 4 ,

- i l  !

i i i i 0' ...,.. ,...,...,...,...,...,...i...,...,

0 2 4 6 8 is 12 14 16 18 28 TIME (HOURS)

Fig. 5.8 Pressure in drywell.

i

k TOVlJ -- PEACH BOTTOM (INTERVENTION) i 252- r --

r - r -- r 7 ,

I

~

e PDuER R E S.J O R E D g;

200- 4 t 7 t- h i-  ; i 1 C -

.; ,158- -

. ---r. . .

g '

ca . . .

~ *  !

/ .

.  ;  ;. m a.

~

A a

un100- (/.-~~~-t------~.-------i-r  ;.- -

-i e i e i -.;

3 -

8

i. i i

i  ;  ;  ;

50- - - -

-.----~4~..~ 4 i-  !

~

0 ...,...i. ,,.. ,...,...,...,...,...,...

0 2 4 6 8 10 12 14 16 10 20 TIE (HOURS)

Fig. 5.9 Temperature of the suppression pool.

5-25 5.6 References 5.1 " Reducing BWR Power by Water Level Control During an ATWS, A Quasi-Static Analysis," NSAC 69, May,1984.

5.2 " Losses of Off-Site Power at U.S. Nuclear Power Plants, All Years Through 1983," NSAC 80, July,1984.

6-1 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION 6.1 Introduction The phenomena of fission product release from the fuel matrix, its transport within the primary system, release from the primary system into the containment, 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 subroutine in MAAP which is based on the FPRAT cooe [6.2]. Transport of fission products outside the core bound-aries is determined by the natural and forced convection flows modeled in MAAP with the gravitational sedimentation described in Ref. [6.3] and other deposi-tion processes described in Ref [6.4]. Fission product behavior is consid-ered for the best estimate transport, deposition and relocation processes.

Influence of surface reactions between chemically active substances like cesium hydroxide and other uncertainties are considered in subtask 23.4. The best estimate calculation, assuming cesium iodide and cesium hydroxide are the chemical state of cesium and iodine, is discussed below.

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

. health risk to be cesium iodide and cesium hydroxide, tellurium oxide and strontium oxide. These and others are considered in the code when calculating the release of fission products from the fuel matrix. Vapors of these domi-nant species form dense aerosol clouds in the upper plenum, in some cases 3

approaching 100 g/m for a very short time, which agglomerate and settle onto surfaces. Depending upon the chemical compound and gas temperature, these deposited aerosols can be either solid or liquid. At the time of reactor vessel failure, some material remains suspended as airborne aerosol or vapor and would be discharged from the primary system into the containment. The rate of discharge is determined by the gaseous flow between the primary system and containment which is sequence specific. (It should be noted that some fission products can be discharged into the containment before vessel failure i

- _ _ _ _ , . - _ _ . -. . . _ - _ - - - - - , - .-. --r - - - - - - -

6-2 4

through relief valves or through breaks in the primary system. This is also sequence specific.) This set of inter-related processes are treated in MAAP and essentially result in a release of all airborne aerosol and vapor from the primary system into containment immediately following vessel failure.

1 As a result of the dense aerosols formed when fission products are released from the fuel, considerable deposition occurs within the primary l system prior to vessel failure. For some accident sequences, the primary {

system may be at an elevated pressure at the time of core slump and reactor vessel failure. Resuspension of these aerosol deposits during the primary system blowdown is assessed in Ref. [6.6] in terms of the available experi-mental results and basic models. It is concluded that resuspension immediate-ly following reactor vessel failure would not be significant, less than 1% of the deposited materials may be resuspended, even for depressurizations initi-ated from the nominal operating pressure. For delayed containment failure, this small fraction of material is depleted by in-containment mechanisms.

Therefore, a major fraction of the volatile fission products are retained within the primary system following vessel failure, the distribution being determined by the MAAP calculations prior to vessel failure. Natural circulation through the primary system af ter vessel failure is analyzed using MAAP which allows for heat and mass transport in various nodes of the reactor vessel including heat losses from the primary system as dictated by the reflective insulation. Material transport in aerosol and vapor form is governed by the heatup of structures due to radioactive decay of deposited fission products. This heatup is principally determined by the transport of cesium iodide and cesium hydroxide by the natural circulation flows. Cesium 4

iodide and cesium hydroxide are modeled with separate vapor pressure curves; the cesium hydroxide vapor pressure coming from recent Sandia data [6.7]. In carrying out these calculations, the pressurization of the primary system is dependent upon the pressurization of the containment and the heating within the primary system. These detennine the in- and out-flows between the primary system and containnent.

Deposition within the containment is calculated in MAAp for sedimen-tation, steam condensation and vapor diffusion to the various surfaces. The A

9

,----mr,,:e ,---ww- --n--- ,--m,-- .- ---.--,e , - . - - ~ - , , - - , - +

r - ,~ ~,.--,,n,-- ,- ,---- -- ,-. ,-v_-,, .

6-3 major aerosol sources are the releases prior to vessel failure (sequence specific), the airborne aerosols and vapors transferred from the primary system at the time of vessel failure, the subsequent releases from the primary system due to long term heatup. and concrete attack. At the time of contain-ment failure, the remaining airborne aerosol and vapor can be released to the environment. Assessments of the potential for resuspension of deposited aerosols following containment failure [6.6] show this to be negligible.

Information was gathered for the Standby Gas Treatment System (SGTS) and according to plant design the fire dampers on the SGTS would close at a specified temperature limit. All of the WASH-1400 comparison case sequences (except TQVW) were run as if the SGTS system was operating and the compartment gas temperatures were found to exceed this criteria for the TW and TC se-quences. As a result of this finding the SGTS system was assumed not to be available for the TC and TW reactor building analysis. Also, a high aerosol loading on the system filters as witnessed in the S Ej sequence was found to possibly cause the SGTS filters to be torn from their mountings. For the sequences analyzed, the operation of the SGTS is influential in decreasing the residence times in the reactor building, and in the event of a loss of the system filters, accelerates the fission product release. A correct under-standing of the reactor building conditions and proper SGTS operation is essential for calculating the environmental release fraction for volatile fission products.

6.3 Sequences Evaluated 6.3.1 TW Fission Product Release As previously described the postulated loss of effective containment heat removal following a reactor isolation (TW) occurs over an extended time period, i.e. 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> before containment pressure exceeds the ultimate capa-bility. Essentially a day is available to take corrective action to prevent containment failure. Also, the assumed failure of the ECCS systems following containment failure is not the result of a mechanistic process. With the assumptions of no corrective action and loss of all injection the accident is assumed to result in containment failure prior to melt-through of the reactor

6-4 pressure vessel. For this sequence with an assumed containment failure size of 0.1 f t , the steam release would exceed the capacity of the SGTS system and would be discharged to the reactor building. Temperatures within the building as a result of the steam discharge would exceed that required to fail the fusible links on the fire dampers of the SGTS, thereby isolating the system.

Table 6.1 shows the Cs! distribution at vessel failure. The Cs0H  !

distribution is calculated separately but is not included due to its similari-ty to the Cs! distribution presented. Due to the relatively low flows to the suppression pool and the large settling area in the upper plenum 93% of the Cs! remains in the primary system. Af ter vessel failure the core debris begins to heat the drywell. Fission products deposited within the primary system heat the structures, vaporize and move within the primary system to colder surfaces. This material transport is illustrated in Figs. B.3 and B.4 of Appendix B. Eventually the entire primary system achieves sufficient temperatures to begin transporting the fission products out into the contain-ment. At this time (s 60 hrs) the fission products begin to be discharged into the reactor building. The ultimate Csl (Cs0H is similar) distribution at 120 hrs into the accident is shown in Table 6.1 and the fraction release of all fission products at this time is shown in Table 6.2.

6.3.2 TC Fission Product Release The MAAP analysis of this sequence shows an initial deposition of volatile fission products in the reactor vessel anu a subsequent redistribu-tion among the different vessel regions af ter vessel failure as these fission products revaporize due to decay heat. SGTS would be isolateo for the same reasons as discussed for TW in Section 6.3.1.

As indicated in Table 6.1, prior to vessel failure, most of the volatile fission product inventory is retained in the vessel upper plenum, but significant quantities are transferred to the suppression pool at vessel failure and for several hours thereafter.

The drywell temperature is maintained at a moderate level by the CR0 water, which flows into the pedestal and cools the debris by vaporization.

l

)

6-5 1

Table 6.1 I DISTRIBUTION OF Cs! IN PLANT AND ENVIRONMENT (FRACTION OF INITIAL CORE INVENTORY)

Vessel Failure

, TW TC SEj TQVW l .

(40 hrs.) (4 hrs.) (3.5 hrs.) (12 hrs.)

RPV .98 . 54 .80 1.0 l l Drywell 0.0 0.0 .11 0.0 l Suppression Pool .02 . 46 .09 0.0 i Reactor Building 2 x 10-5 1 x 10~4 0.0 0.0 Environment < 1 x 10-5 3 x 10~4 0.0 0.0 1

i 1

Containment Failure TW TC SE j TQVW l 2

(32 hrs.) (12.8 hrs.) (23 hrs.) (18 hrs.)  !

l RPV 1.0 .38 .51 .76  !

Drywell 0.0 .02 .05 .20  ;

Suppression Pool 0.0 .60 .44 .04 i Reactor Building 0.0 0.0 0.0 i

.005

Environment 0.0 4 x 10-4 0.0 0.0 l i l l

i End of Evaluation 1 TW TC SE j TQVW j (120 hrs.) (60 hrs.) (60 hrs.) (60 hrs.) l i

l RPV .01 .01 .18 .09

Drywell 0.0 0.0 .05 0.0 Suppression Pool .02 .60 .44 .04 I i Reactor Building .78 .36 .29 .82 1 Environment .19 .03 .04 .05  ;

I i

t i

mw s.w w -,,n,r- > r,- ne,.rn.n,-m,x-,-,,,,,-----,wn-,,w ,en, ---+,m---we--v-. wm - - .mm-w- g-- --n--,- , - - ---, , ,,,

6-6 Table 6.2 TW FISS10ft PRODUCT RELEASE Assumptions Containment Failure Location - Drywell, El 165' Containment Failure Size .1 f t 2 Fission Product Release Fraction Group to Environment Cs! 0.19 Te, Sb 0.11 Sr Ba 4 x 10-4 Ru, Mo 6 x 10'4 Cs0H 0.19 Time of Release: 42 hr.

Duration of Release: 80 hr.

6-7 Af ter the CST is depleted and assuming no operator action, the drywell tem-perature increases to levels which could threaten the integrity of the primary containment and allow a bypass of the suppression pool. As the drywell heats up after vessel failure due to the core debris on the floor, heat is trans-ferred to the reactor vessel which ultimately comes into thermal equilibrium with the drywell. As the vessel heats up, all of the volatile fission prod-ucts retained in the vessel are revaporized and are convected out of the vessel at a low flow rate. These fission products are released from the vessel and transported into the drywell where some are deposited and others are transported to the suppression pool. The drywell is assumed to fail at a temperature of 920 K (1200 F), which allows the airborne fission products to bypass the suppression pool. (The influence of the assumed containment failure temperature is evaluated in Ref. [6.8] and it is demonstrated to have no significant effect.) As a result of the elevated drywell temperature the material is transported mostly as vapor and little deposition occurs in the drywell . At this time a significant amount of concrete aerosols are being re-leased due to core-concrete attack in the pedestal region. The volatile fission products condense and form aerosols as they flow into the reactor building along with the inert aerosols. Most of these materials are removed due to gravitational settling and condensation within the building. Conse-quently a relatively small fraction of the volatile fission products are released to the environment as indicated in Table 6.3.

6.3.3 S E Fission product Release j

This sequence was analyzed to determine the time dependent distribu-tion of volatile fission products within the vessel, the rate of release from the vessel to the drywell and, af ter containment failure, the release to the reactor building and subsequently to the environment. It can be seen that drywell heatup, which occurs from the core debris on the floor, influences the long term heatup of the entire reactor vessel.

Drywell heatup results in the revaporization of the volatile fission products which have been retained in the drywell. Most of this material is convected from the drywell to the reactor building within five hours af ter containment failure. Revaporization in the reactor vessel is also occurring,

6-8 Table 6.3 TC FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Drywell El 165' Containment Failure Size - 2 ft 2 Fission Product Release Fraction Group to Environment Csl 0.03 Te, Sb 0.06 Sr, Ba 1 x 10~4 Ru, Mo 2y 10~4 Cs0H 0.03 Time of Release: 13 hr.

Duration of Release: 50 hr.

I f

. , . - . . - - . - - , . - s ...n. -. - --- -, -. , ,, . . .. . ---.

6-9 but due to low flows from the vessel to the drywell most of the release from the vessel is not complete until about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> af ter containment failure.

This material is passed through the drywell to the reactor building. As I indicated in Table 6.1, none of the volatile fission products are ultimately retained in the reactor vessel or drywell. (It should be noted that chemical reactions between the volatile fission products and the steel surfaces of the primary system could reduce the material vapor pressures. These phenomena are currently being studied experimentally, but are not credited in this analy-sis.) Aerosols initially released from the containment would be sucked into the SGTS system and retained in the filter. After approximately 145 kg of aerosol accumulates in the first set of HEPA filters, the plugged filters are -

then postulated to rupture due to the pressure differential created by the SGTS fans. The second set of HEPA filters would then begin to accumulate aerosol until 145 kg of aerosol is collected, again, the filters are postu-lated to tear, at which time all HEPA filtration is assumed to be lost. The filter efficiency is assumed to be 99.97% (design) from containment failure at 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> to approximately 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> at which time 290 kg has accumulated. The fans continue to operate af ter 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> with an assumed system filter effi-

] ciency of zero.

Gravitational settling of fission product aerosols in the reactor l

i building is also occurring at this time resulting in some retention in the building. The volatile fission products released to the environment are given in Table 6.4.

1 6.3.4 TQVW Fission Product Release Since SGTS is unavailable, the path to the environment is through the reactor building with direct leakage to the atmosphere. The reactor building flow rates are governed solely by the containment break flow and the ensuing thermal hydraulic conditions in the reactor building. Therefore, there is no forced convection or fission product removal resulting from SGTS operation.

As in the other sequences, drywell heatup contributes to the reactor I

vessel heatup. As indicated in Table 6.1 most of the volatile fission

- - - - , , - , - , ~ - - - - - . - - - - , ----,-...----,---,,,-,---.-,---.---,-m-,----- , - , ,me----,--- .------,--,-r----------m ----r--+-

6-10 Table 6.4 SjE FISSION PRODUCT RELEASE I Assumptions Containment Failure Location - Drywell, El 135' Containment Failure Size - 0.1 f t 2 Fission Product Release Fraction Group to Environment Csl 0.04 Te. Sb 0.06 Sr. Da 1 x 10-5 Ru No 2 x 10-5 Cs0H 0.04 Time of Release: 23 hr.

Duration of Release: 30 hr.

6-11 products are in the reactor vessel and the suppression pool at the time of containment failure. The inventories of cesium, iodine and tellurium in the suppression pool are somewhat less than those calculated for other sequences since the primary system does not depressurize until the vessel fails. As the drywell and reactor vessel heatup, revaporization of the volatile fission products in the vessel occurs and they are convected out of the vessel with a low flow rate. Consequently, most of these fission products are out of the vessel by about 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> af ter containment failure. These volatile fission products pass through the drywell as vapors and into the comparatively cool reactor building where they condense to form aerosols and result in substan-tial gravitational settling. The effectiveness of this removal mechanism is enhanced by the low temperature and the long residence times in the reactor building because of the absence of forced convection from SGTS operation.

Consequently, only 5% of the Cs! is released to the environment as indicated in Table 6.5.

6.4 Assumptions and Actions Relating to Fission product Distribution and Release The sequences analyzed in Section 4.0 included certain assumptions aimed at obtaining comparisons with existing studies. These assumptions relating to containment failure modes and operator actions were not selected based on a most probable course of events. This section investigates the affect that the containment failure area has on the fission product releases in the TW sequence. Also, several sequence variations were studied relating to the TC sequence.

The TW WASH-1400 comparison case was analy.ted using a containment failure area equal to .1 f t .2As described in Section 4.1.2, a nonmechanistic assumption was made which led to the failure of all ECCS at the time of containment failure. In order to threaten the integrity of the emergency systems in the reactor building a larger containment failure may likely be required. An analysis of the fission product release associated with a I f t failure area was performed to demonstrate the influence of the assumed failure area. As illustrated in Table 6.6 the larger failure area results in more Cs!

being transported to the suppression pocl. This is due to the re-opening of I

6-12 l

Table 6.5 TQVW FISSION PRODUCT RELEASE l

Assumptions Containment Failure Location - Orywell, El 135' l Containment Failure Size - 0.1 f t 2 Fission Product Release Fraction Group to Environment l

Cs! 0.05 Te, Sb 0.06 Sr, Da 8 x 10-5 i Ru, Mo 1 x 10~4 Cs0H 0.05 Time of Release: 18 hr.

Duration of Release: 30 hr.

+ l

6-13 the ADS values as the containment is depressurizing. At the time the valves open, fission products are available to be transported to the supcression pool. Due to a slower containment depressurization for the .I f t failure area case, a smaller quantity of fission products were available to be trans-ported to the suppression pool at the time the ADS valves re-opened. Another significant difference between the .1 ft and I ft cases is the environmental release fraction. Also, the larger failure area depressurized the containment more rapidly such that later in the accident, when fission products are released from the reactor vessel, the flow through the reactor building is considerably less than that for .1 f t failure area case. As stated earlier, the flow through the reactor building controls the residence times for fission products and significantly impacts the time available for deposition, and hence the environmental release fraction.

In addition, several sequences were run to investigate the sensi-  !

tivity to containment failure modes in the TC sequence. The case analyzed in Section 6.3 included wetwell venting according to the operator procedures, yet due to the lack of water on the debris af ter vessel failure, the drywell was I calculated to reach the assumed overtemperature failure condition at 12.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. An operator action to refill the CST and increase CRD flow or finding l an alternate method of putting water on the debris af ter vessel failure would l keep the temperatures in the drywell below that which could potentially fail the shell. This sequence would result in venting of the containment followed by continued fission product scrubbing in the suppression pool. Since the drywell would not fail, the fission products released from the vessel would pass through the suppression pool before entering the reactor building and considerable decontamination would occur. Also, due to the low drywell temperatures, a significant fraction of the deposited fission products could remain in the reactor vessel with their decay power being transferred to the associated heat sinks. Table 6.6 shows the distribution and release of Cs!

for the TC case analyzed in Section 6.3 and the venting case without drywell failure, i

The last two cases run for TC were aimed at investigating the releases associated with no wetwell venting. The first case was much like that addressed in Section 6.3 except without wetwell venting. For these l

Table 6.6 Csl DISTRIBUTION AND RELEASE (FRACTION OF INITIAL CORE INVENTORY)

TW TW TC - Vent TC - Vent TC - No Vent TC - tio Vent

• I ft 2

1 ft 2 No Debris Debris fio Debris Debris Cooling Cooling Cooling Cooling (Tice, hrs.) 120 100 60 60 60 15*

RPV .01 0.0 .01 .25 .04 .27 Dryw11 0.0 0.0 0.0 0.0 0.0 .02 Suppression Pool .02 .17 .60 .75 .27 .55 Secondary Contairrient .78 .79 .36 2 x 10-4 .56 .13 $

Environment .19 .04 .03 6 x 10 -4 .13 .03*

• At 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> the major fission product release has occurred.

i iiote that the current MAAP modeling does not account for fission product retention due to drywell sprays which clay represent a significant conservatism in the ATWS source term evaluation.

i 6-15 conditions the drywell fails cue to overpressure at 1.35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />, and contain-ment failure area of .8 f t 2was used to relieve the drywell pressurization.

l

{ Since there is no water on the debris af ter vessel failure the drywell tem- j peratures increases, causing the temperatures in the reactor vessel to in-crease and thus causing the revaporization of deposited fission products.

! This combination of no wetwell venting, small drywell failure area and no l water on the debris is the least probable of the TC sequences and results in  !

l the largest releases. The second case also excluded wetwell venting but did  !

l allow for water to be injected onto the debris af ter vessel failure. The high l l pressure service water was aligned to the drywell sprays and cycled in order  !

! to minimize the water addition to the suppression pool. As seen before this ,

{ causes the drywell temperatures to remain low, thereby allowing for hold-up of i

) fission products within the reactor vessel, which substantially reduces the <

environmental release fractions. The distribution and release of Cs! for f these two ATWS cases, are presented in Table 6.6 and compared to the base [

} case.

i i I  !

j 6.5 References f

i I j 6.1 MAAP - Modular Accident Analysis Program, User's Manual, August,  !

1 1983.  !

i 6.2 10COR Technical Report 15.1B, " Analysis of In-Vessel Core Melt i Progression " Vol. IV (User's Manual) and Modeling Details for the  !

! Fission Product Release and Transport Code (FPRAT), September,1983. L 6.3 Draf t IDCOR Technical Report, "FAI Aerosol Correlation " July,1984.

i;

! 6.4 10COR Technical Report on Task 11.3, " Fission Product Transport in Degraded Core Accidents " December,1983.  !

6.5 10COR Technical Report on Tasks 11.1,11.4 and 11.5, " Estimation of i Fission Product and Core-Material Source Characteristics," October, l 1982.

, 6.6 IDCOR Technical Report on Task 11.6. "Resuspension of Deposited ,

i Aerosols following Primary System or Containment Failure," July,  ;

1984.

1 6.7 Sandia Cs0H vapor pressure data.

! 6.8 58 LOCA Outside Containment at Browns Ferry Unit 1. Vol. !!, [

{ NUREG/CR-2672, ORNL, September, 1983.

l J

i G .

7-1 7.0

SUMMARY

OF RESULTS l

Four severe accident sequences were analyzed for Peach Bottom.

These sequences were identified in Task 3.2 as cominant or key sequences that could potentially lead to core melting. The analyses assumed the accidents proceeded with minimal operator intervention in order to determine the timing and magnitude of the major phenomenological events. The results of MAAP produce time estimates for core melting, vessel failure, and containment failure as well as estimates of fission product release. The results of the analyses of the four sequences are shown in Tables 7.1 and 7.2. Assuming the pressure and temperature failure criteria utilized in these analyses, one sequence resulted in containment failure due to pressure. Containment failure for two sequences was assumed due to high temperatures which occurred prior to l reaching overpressurization failure criteria. One sequence failed the drywell l due to overtemperature af ter the wetwell had been vented. Table 7.2 compares l the results of the analyses with WASH-1400. The release fractions of fission l products are considerably less than those reported in WASH-1400 due to the more realistic modeling of fission product behavior as well as the transport paths from the containment to the environment. In addition the results indicate that the release of fission products to the environment occurs at longer times than postulated in WASH-1400, i.e. from several hours to over a day af ter initiation of the accident.

The WASH-1400 comparison sequences were reanalyzed with MAAP to demonstrate the effectiveness of selected operator actions in mitigating the consequences of these severe accidents. The examples presented in Section 5  !

demonstrate that proper operator actions are extremely beneficial. There are I several alternatives available to operators with the existing Peach Bottom j systems and the plant specific emergency procedures at various stages of the l accident sequences to bring the plant to a safe stable state. Simple operator actions are judged to be easily performed and very probable. The assumption that these actions are not taken in the WASH-1400 comparison sequences is judged to make these cases very low probability events.

Section 5 describes the capabilities that exist at Peach Bottom for l l

, venting primary containment. As indicated, venting capacity is substantial.

l

7-2 t i

Table 7.1 i

SUMMARY

OF MAAP RESULTS FOR WASH-1400 COMPARISON SEQUEriCES*

f Event TW TC SEj TQvW ECCSStart(hrs) 0 0 NA 0 ECCSStop(hrs) 32 1.3 tiA 6 CoreUncovered(hrs) 34 1.3 1.1 8.4 Cladding Temp. at 2000*F (hrs) 36 2.5 1.6 9.8

, Fuel Melting Begins (hrs) 39 3.0 2.6 11.4 VesselFailure(hrs) 40 3.8 3.5 12 j Fuel Melting Complete (hrs) 75 22 30 35 Containment Failure (hrs) 32 12.8 23 18 Orywell Temp, at 600*F (hrs) 45 8 19 14 i Orywell Temp. at 1500*F (hrs) 62 16 27 21 i j Max.ContainmentPressure(psia) 132 115 58 110 4

l i

In-VesselZirc0xidation(IbmH2) 400 240 240 550 Containment Failure Mode Pres Temp Temp Temp (BasedonAssumedCriteria) ,

• These sequences assume minimal operator intervention.

l

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

_____-.._......,_-,4 _ _ _ , _ _ . _ _ _ , _ , _ _ _ . _ _ _ . _ _ _ .

Table 7.2 SIM ERY OF FISSION PRODUCT RELEASE FRACTIONS (a)

Sequence WASH-1400 F.P. Group TW TC SE j TQVW BWR2(b) BWR3(c) l Cesium Iodide 0.19 0.03 0.04 0.05 0.50, 0.90 0.10 Telluriu:: 0.11 0.06 0.06 0.06 0.30 0.30 Strontiun 4 x 10-4 1 x 10-4 1 x 10-5 8 x 10-5 0.10 0.01 Ruthenium 6 x 10-4 2 x 10'4 2 x 10-5 1 x 10-4 0.03 0.02 Cesium hydroxide 0.19 0.03 0.04 0.05 - -

1

! Time Release (hr) 42 13 23 18 - -

l Duration of Release (hr) 80 50 30 30 - -

I"} fraction of core inventory released to the envirorment.

(b)Contaircent failure prior to vessel failure; can be compared with (TW).

ICI Failure to scram or resove decay heat; can be compared with (TC, S E, TQVW).

j

7-4 The gases released from the small lines through the Stancby Ga: Treatment System and all lines connected to the wetwell are effectively filtered or scrubbed prior to release. The effectiveness of venting in reducing contain-ment pressure is also demorstrated in Section 5 for the TW and TC analyses for which it was assumed operators vented from the wetwell in accordance witn existing emergency procedures.

Review of the base sequences, as well as those with operator inter-ventions, indicates that through realistic assessment of phenomenology, radionuclide releases are reduced and delayed relative to WASH-1400 and safe stable states are achievable.

8-1

8.0 CONCLUSION

S Task 23 analyses for Peach Bottom have demonstrated several key items relevant to nuclear power plant severe accident analysis for BWRs with Mark I containment designs similar to Peach Bottom.

1. The viability of the Modular Accident Analysis Program (MAAP) in analyzing challenges to containment resulting from degraded core accidents has been demonstrated. This provides an inde-pendently developed alternative to the models available prior to the IDCOR Program.

2. The use of MAAP to more realistically determine the release of fission products to the environment following a set of selected low probability, degraded core nuclear power accident sequences indicates that, in general, radionuclide releases would be smaller fractions than those previously estimated in WASH 1400 for similar accident sequences, in addition these releases would occur much later in time.

3. Based on the sequences analyzed it is clear that reasonable actions by trained operators using existing systems and proce-dures could effectively mitigate the accident consequences by bringing the plant to a safe stable state. Fission product releases could be substantially reduced from those calculated in the WASH-1400 comparisun cases through the use of existing capabilities and procedures.

4. For those postulated sequences which are assumed to progress to core melt and RPV failure, the containment floor (pedestal) concrete ablation depths at the time of containment failure illustrated in the graphs of $ection 4 indicate that base mat penetration is not a likely mode of Mark I containment failure.

5. MAAP analyses of heatup of containment from radiative and convective heat transfer from core debris on the containment

8-2 floor indicate that temperatures could becorre high enough to result in a reduction of the ultimate pressure capability cf the containment for some sequences.

l i

4 4

i i  !

i 5

)

I i i

A-1 APPENDIX A.1 Peach Bottom Parameter File 88 MARK ! DWR PLANT PARAMETER VALUES-- TYPICAL OF PEACH BOTTOM  !

84St UNITS (M-MG-SEC-DEGK)  !

$8 10-5-R4  !

oh AFLCOR FLOW AREA 0F REACTOR CORE 88 AL!H = V01.UME WATER IN LOWER SHROUD BELOW TOP OF ACTIVE FUEL DI at BY (ZTOAF-ZSKT) 02 11 000 ALSH FLOW AREA IN LOWER SHROUD PS 03 2.00 AFLBYP CORE BYPASS FLOW AREA PS

$8 AUSH = VOLUME OF WATER IN UPPER DOWNCOMER ABOVE TOAF DIVIDED BY 88 THE NORMAL WATER HEIGHT ABOVE TOAF 04 20 000 AUSH FLOW AREA IN UPPER SHROUD PS 05 1.3505 HCRD SPECIFIC ENTHALPY OF CRD INLET PS 06 0.17105 HFW SPECIFIC ENTHALPY OF FEEDWATER PS 07 1.584D5 MU2COR TOTAL MASS OF 002 IN CURE PS h!

10 4.'k 1.9502 NS NCRD N bhhb!L NUMBER OF CRD TUBES S F L hL h PS

,1 5.000 NOFPS SESNIPLE ENERGY STORED IN FUEL (FULL POWER SECONDS)PS L2 3.000 TDMSIV DELAY TIME FOR MSIV CLOSURE PS L3 3.500 TDSCRM DELAY f!ME FOR FULL SCRAM PS 14 3.007 TIRRAD TOTAL EFFECTIVE 1RRADIATION TIME FOR CORE PS 88 ALL PUMP CURVES ASSOCIATE THE FIRST FLOW ENTRY WITH THE FINST PRESSURE 88 ENTRY 6 . 20 2 C h Aih F R R S VCRDI 19 i h'!l1

1. 52 -2 28:8 WVCRD 8Phk8"!!!!

CR FLOW RATE hi PS 20 11 2l1-2 WVCRD:' CRD FLOW RAIE PS 1 1.' 2 1-7 WVCRD PS

' CRD[LOWRATE 2

24 119 D! b

. 01.3405 PCR0 P F0 R PPS FOR CRD PUMP MP Ph PS 25 .0134D$ PCR0 PPS FOR CRD PUMP PS 26 L .0L3405 PCR0 PPS FOR CRD PUMP PS 27 1. 0 ' 3405 PCR0 PPS FOR CRD PUMP PS hR b 30 1 01J40S PCR0 PPS FOR CRD PUMP PS 11 2275 00 KWMAX MAXIMUM FEEDWATER FLOW RATE (RUN OUT) PS 37 4.4202 W9PMAX MAXIMUM TURb!NE SYPASS FLOW RAIE PS

! 33 1.40-1 NXCORE EXIT CORE QUALITY PS hi R R h hA Ok!!Sf hk 3.thW@0 XRRV fMTER :0R RAI) US UF REAbiOR VESSEL PS 37 40 1100 JET ELEVA" 0N AT DOTTOM OF JET PUMPS PS

. ZDSh k Af A ff hhS AM AFATCRS 40 45. VD0 lDV ELEVAT ON AT 60ff0M OF REACTOR VLSSEL PS 2

. f kVAT0 fk 0 FUMPS hk 43 4500 AKT TOTAL tET PUMP AREA PS 44 Pf0AF LLEVAT ON AT TOP OF ACilVE FUEL PS

!2 47 e f,L A MEb:VA"'ON gg OAfN SkEAKgg l kOUD $$I$$$hEL g h PS 40 ALOCA AREA Of DREAK PS gg.0093 g g8 LEVAfl0N AT LEVEL 9 1 RIP PS 11 m il82 11E1 # hf P elEli a SCRA. SEfroiar f!

A-2 53 .20D0 FOATWS ATWS CONSTANT POWER ASSUMPTION PS 54 3.6D3 TDSLC TIME FOR SCRAM WITH SLC PS 88 RR PUMF COASTDOWN CURVE 55 0.D0 TIRR(1) TIME VS. FRACTION OF TOTAL FLOW FOR RECIRC PUMP PS 54 2.D0 TIRR(2) PS 57 4.00 TIRR(3) PS 58 6.00 TIRR(4) PS 59 8.00 TIRR(5) PS 60 10.00 TIRR(6) PS 61 12.00 TIRR(7) PS 62 14.03 TIRR(8) PS 63 1 00 FWRR(1) PS 64 .77D0 FWRR(2) FS 65 .6200 FWRR(3) PS 66 .5300 FWRR(4) PS 67 .4500 FWRR(5) PS 48 .3400 FWRR(6) PS 69 .3400 FWRR(7) PS 70 0.D0 FWRR(8) PS 71 1.46D5 HSLC INLET ENTHALPY OF SLC PS 72 0.00 PSLC(1) PRESSURE POINTS FOR SLC FLOW CURVE PS 73 1.D7 PSLC(2) PS 74 1.D7 PSLC(3) PS 75 1.D7 PSLC(4) PS 76 1.D7 PSLC(5) PS 77 1.D7 PSLC(6) PS 78 1.D7 PSLC(7) PS 79 1.D7 PSLC(8) PS 80 1.72D-3 WVSLC(1) SLC FLOW RATE AT PSLC(1) -- M3/S PS 81 1.72D-3 WVSLC(2) PS 82 1.72D-3 WVSLC(3) PS 83 1 72D-3 WVSLC(4) PS 84 1 720-3 WVSLC(5) PS 95 1.72D-3 WVSLC(6) PS 86 1.720-3 WVSLC(7) PS 87 1.72D-3 WVSLC(8) PS 98 0.00 TDRPT DELAY TIME FOR RECIRC PUMP TRIP PS 89 54.7900 ZLMSIV LOW WATER LEVEL FOR MSIV CLOSURE PS 90 57.64D0 ZLRPT LOW WATER LEVEL FOR RPT PS 91 7.886D6 PHRPT HIGH VESSEL FRESSURE FOR RPT PS 92 1.151305 PDWSCM HIGH DRYWELL PRESSURE SCRAM SIONAL PS 93 .029900 FENRCH NORMAL FUEL ENRICHMENT PS 94 20000.00 EXPO AVERAGE EXPOSURE IN MWD / TONNE PS 95 .600 FCR PRODUCTION OF U239 TO ABSORBTION IN FUEL PS 96 1.300 FFAF RATIO 0F FISSILE ABSORBTION TO TOTAL FISSION PS 97 6.96D-1 F0FR1 FIS$10N POWER FRACTION OF U235 AND FU241 PS 98 2.2300-1 F0FR2 FISSION POWER FRACTION OF PU239 PS 99 8.D-2 FDFR3 FISSION POWER FRACTION OF U238 PS 100 .3048D0 XPCRDT PITCH OF CRD TUBES PS 101 .2755D0 XDCRDT OUTER DIAMETER OF CRD TUBES PS 102 55.00 NINST NUMBER OF INSTRUMENT TUBES PS 103 .00419D0 XTHCRD THICKNESS OF CRD TUBE WALL PS L04 .0508D0 XDINST DUTER DIAMETER OF INSTRUMENT TUPE PS 105 .075D0 XDRIVE LOWER CRD DRIVE OUTER DIAMETER PS L06 1 00510-3 VWCRD SPECIFIC V0MUME OF CRD WATER PS 107 1.00510-3 VWSLC SPECIFIC VOLUME OF SLC WATER PS 108 1 0505 MEOPS MASS OF UPPER PLENUM HEAT SINK PS

'109 1 01603 AEOPS AREA 0F UPPER PLENUM HEAT SINK PS 10 2143700 XiRV THICKNESS OF LOWER VESSEL HEAD PS 11 .00 TIFWCD TIME SINCE MSIV CLOSURE SIGNAL VS. FEEDWATER PS L12 .D0 00ASTDOWN MASS FLOW RATE PS 113 0 00 8 TIME POINTSe8 FLOW RATES PS 114 0.00 PS l

A-3 115 0.D0 PS 116 0.00 PS 117 0.00 PS 118 0.D0 FS

"'"c" l!

113 8:88 121 0.00 PS 122 0.D0 PS 123 0.D0 PS 124 0.D0 PS ll!8:88  ??

fh8 f2hh0 M E V ON TE L$NE Oh H MAIN STEAM LINE

!f!RC

$8THIS SECTION IS FOR INPUTS TO THE DETAILED PRIMARY SYSTEM T/H

$8M000LE SUBROUTINE CIRC AND ITS ANCILLARY SUBROUTINES

• • TO DEVELOP INPUTS FOR THIS SECTION, CONSULT THE APPROPRIATE FIGURE 88FOR YOUR PLANT TO DELINEATE N0DAL BOUNDARIES

$8 INSERT NUMBERS IN THE SPACES SHOWNI THE CODE COMPUTES THE MISSING 88 NUMBERS FROM THESE AND OTHER INPUIS

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

$$ ' STEEL' MASS AND MAY IN SOME CASES ALSO HAVE A ' HEAT SINK' MASSI 88 THE HEAT SINK IS DISTINGUISHED FROM THE STEEL MAINLY BY TWO DIFFERENCES 1 88 1. MAY BE AT A DIFFERENT TEMPERATUREI

$$ THIS MAY BE DUE IN PART TO THE HEAT SINK HAVING LOSSES TO CONTM

$$ WHEN THE STEEL DOESN'T (EG S/G SHELLS VS TUBES)

$8 2. HEAT SINKS ARE ASSUMED NOT TO HAVE FISSION PRODUCTS 88 PLATED ON THEM

$8 88 AT PRESENT, THE ENERGY EXCHANGES IN A NODE MAY INCLUDE ONE OR MORE OF 88 THE FOLLOWING, DEPENDING ON THE INPUT PARAMETERS SUPPLIED:

• • 1. HEAT SINK AND STEEL EXCHANGE ENERGY RADIATIVELY (GAS 88 ASSUMED TRANSPARENT) 88 *hR NThANOA NA DR RU ON 88 3. HEAT SINK MAY EXCHANGE ENERGY WITH PRIMARY SYSTEM GAS VIA 88 INTER- OR INIRA-NODAL NATURAL CIRCULATION 88 88 ITEMS 1-3 ARE HT AREAS COUPLING THE 2 N0DAL HEAT SINK MASSES (STEEL AND 88 HEAT SINK) 01 0.00 ACSHS(1) CORE + LOWER PLENUM 88 CARBON STEEL-HEAT SINK HEAT TRANSFER AREA 02 140.D0 ACSHS(2) UPPER PLENUM 03 0.00 ACSHS(3) DOWNCOMER

• • ITEMS 6-8 ARE ' STEEL' (INTERNAL) MASSES bh !0D3 Mbhlhl UhhfRP UM 08 350.D3 MCS(3) DOWNCOMER 88 ITEMS 11-13 ARE THE ' HEAT SINK' MASSES 11 0 00 MHS(1) CORE t LOWER PLENUM HEAT SINK MASS 12 100.D3 MHS(2) UPPER PLENUM 13 0 00 MHS(3) DOWNCOMER 88 !TEMS 14-18 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

• • STEEL MASSES AND CONTAINMENT 16 0.D0 ACSX(1) CORE + LOWER PLENUM CARBON STEEL TO DRYWELL 88 HEAT TRANSFER ARE 17 0.00 ACSX(2) UPPER PLENUM 18 240.D0 ACSX(3) DOWNCOMER 88 ITEMS 21-23 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN 88 HEAT SINK MASSES AND CONTAINMENT I

A-4 e

21 0.00 AHSX(1) CORE + LOWER PLENUM HEAT SINK TO DRYWELL

• • HEAT TRANSFER AREA 22 140.D0 ANSX(2) UPPER PLENUM 23 0.90 ANSX(3) DOWNCOMER

• • ITEMS 26-29 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

• • STEFL MASSES AND PRIMARY SYSTEM GAS 26 100.D0 AGCS(1) CORE + LOWER PLENUM GAS TO CARBON STEEL

• • HEAT TRANSFER AREA 27 5.D3 AGCS(2) UPPER PLENUM l 28 240.D0 AGCS(3) 00WNCOMER  !

• • ITEMS 31-33 ARE THE AREAS ASSOCIATED WITH ENERGY EXCHANGE BETWEEN

• • HEAT SINK MASSES AND PRIMARY SYSTEM GAS g 0.D0 AGHS(1) g g PL g GAS TO HEAT SINK 32 140.00 AGHS(2) UPPER PLENUM 33 0.00 AGHS(3) DOWNCOMER

• • APPROX. NODE HEIGHT FOR EACH COMPARTMENT USED FOR NAT CIRC 34 8.000 XL(1) CORE + LOWER PLENUM LENGTH 37 5.D0 XL(2) UPPER PLENUM LENGTH 38 10.D0 XL(3) DOWNCOMER LENGTH ,

88 GAS FLOW AREA 41 11.90 AG(1) CORE + LOWER PLENUM FLOW AREA 42 11.D0 AG(2) UPPER PLENUM FLOW AREA 43 10.00 AG(3) DOWNCOMER FLOW AREA

• • HYDRAULIC DIAMETER USED TO COMPUTE HT COEFF 46 5.00 DH(1) HYDRAULIC DIAMETER FOR CORE REGION '

47 . 15D0 DH(2) HYDRAULIC DIAMETER FOR UPPER PLENUM 48 .400 DH(3) HYDRAULIC DIAMETER FOR DOWNCOMER 51 3.D5 0C0 RPV CONVECTION LOSSES AT TIME ZERO 52 0.D0 FINPLT NUMDER OF LAYERS IN REFLECTIVE INSULATION 53 32.5 ASEDPS(1) AEROSOL SEDIMENTATION AREA FOR CORE + LOWER PLENUM 54 20.0 ASEDPS(2) AEROSOL SEDIMENTATION AREA FOR UPPER PLENUM 55 10 0 ASEDPS(3) AEROSOL SEDIMENTATION AREA FOR DUWNCOMER

• HEATUP HE 01 3.8100 XZFUEL LENGTH OF ACTIVE FUEL HE 02 5.210-3 XRFUEL RADIUS OF FUEL PELLET HE  !

05 1.714 hh MDCd MkSOF SSEMBLY CANS TOTAL MASS UF CONTROL BLADES IN REACTOR CORE Hh HE 04 3.049D-3 XZRCAN CAN WALL THICKNESS HE

$$ NODE 1,1 IS CENTER DOTTOM, 1,10 IS CENTER TOP, 2,1 IS SECOND ?ADIAL

• • REGION AT THE BOTTOM, ETC.

07 1 061D0 FP:AK(1,1) PEAKING FACTOR FOR NODE (1,1) HE 00 .99900 FP :AK(2,1) AKING FACTOR FOR NODE (2,1) HE 09 .32600 FPlAK(3,1) AKING FACTOR FOR NODE (3,1) HE 15 1.43800 FPEAK(1e2) AKING FACTOR FOR NODE (1e2) HE 16 1.21700 FPEAK(2,2) PEAKING FACTOR FOR NODE (2,2) HE 17 .44200 FPEAK(3,2) PEAKING FACTOR FOR NODE (3,2) HE

. F h 3) I 25 .47790 FPEAK(3e3) PEAKING FACTOR FOR NODE (3e3) HE 31 1 59200 FPEAK(1,4) PEAKING FACTOR FOR NODE (1,4) HE 4 F !3' ,

39 1 60500 FPEAK(1,5) PEAKING FACTOR FOR NODE (1,5) HE 40 1 35000 FPEAK(2,5) PEAKING FACTOR FOR NODE (2,5) HE h F T )

40 13U F AK(2,6) FAK NG FACTOR FOR N0DE (2,6) HE 49 .5011 F AK(3e6) P.AKLNG  ! FACTOR FOR NODE (3,4) HE

! 55 1. F AK(1,7) PEAKING FACTOR FOR N0DE (1,7) HE

54 1. 324D0 W p0 FP AKt2,7) PEAKING FACTOR FOR NODE (2,7) HE  :

l 57 .48200 FPEAK(3,7) PEAKING FACTOR FOR NODE (3e7) HE i

i

! i

' i I

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

A-5 43 1.541D0 FPEAK(1,8) PEAKING FACTOR FOR NODE (1,8) HE 64 1.304D0 FPEAK(2,8) PEAKING FACTOR FOR NODE (2,8) HE 65 .473D0 FPEAK(3,8) PEAKING FACTOR FOR NODE (3,8) HE 71 1.465D0 FPEAK(1,9) PEAKING FACTOR FOR NODE (1,9) HE 72 .986D0 FPEAK(2,9) PEAKING FACTOR FOR NODE (2,9) HE 73 .358D0 FPEAK(3,9) PEAKING FACTOR FOR N0DE (3,9) HE 79 .777DO FPEAK(1,10) PEAKING FACTOR FOR NODE (1,10) HE 80 .65800 FPEAK(2,10) PEAKING FACTOR FOR NODE (2,10) HE 81 .239DO FPEAK(3,10) PEAKING FACTOR FOR NODE (3,10) HE 87 0.33D0 XCHIM UNHEATED FUEL LENGTH AT TOP OF CORE HE 88 1.D-7 XIZROX INITIAL CLADDING OXIDE THICKNESS HE

• • ES

• • ES

• ENGINEERED SAFEGUARDS ES 01 2.D0 NLPCII NUMBER OF LPCI PUMPS IN LOOP 1 ES 02 2.0D0 NLPCI2 NUMBER OF LPCI PUMPS IN LOOP 2 ES 03 0.0D0 NLPCI3 NUMBER OF LPCI PUMPS IN LOOP 3 ES 04 4.0D0 NLPCSP NUMBER OF LPCS PUMPS ES 05 0.0D0 NOT USED 06 100.D0 VMNCST MIN WATER VOLUME IN CONDENSATE STORAGE TANK ES 07 1 0051D-3 VWCST SPECIFIC VOLUME OF CST WATER ES

• • PUMP CURVES ARE DEFINED SO THE FIRST PRESSURE ENTRY CORRESPONDS WITH

• • THE FIRST VOLUMETRIC FLOW ENTRY 08 8.124D6 PHPCI(1) PUMP CURVES FOR ECCS ES 09 8.D6 PHPCI(2) PPS-PDW VS VOLUMETRIC FLOW ES 10 6.8D6 PHPCI(3) HPCI ES 11 5.1D6 PHPCI(4) ES 12 3.06D6 PHPCI(5) ES 13 1.7D6 PHPCI(6) ES 14 1 134Do PHPCI(7) ES 15 6 17D5 PHPCI(8) ES 16 .31567D0 WVHPCI(1) ES 17 .31547D0 WVHPCI(2) ES 18 .31567D0 WVHPCI(3) ES 19 .31567D0 WVHPCI(4) ES 20 .31567D0 WVHPCI(5) ES 21 .31567D0 WVHPCI(6) ES 22 .31567D0 WVHPCI(7) ES 23 0.0D0 WVHPCI(8) ES 24 2.172D6 PLPCI(1) LPCI ES 25 2.123D6 PLPCI(2) ES 26 1.916D6 PLPCI(3) ES 27 1.66D6 PLPCI(4) ES 28 1.4788D6 PLPCI(5) ES 29 1.045D6 PLPCI(6) ES 30 7.894D5 PLPCI(7) ES 31 1.01342D5 PLPCI(8) ES 32 0.D0 WVLPCI(1) ES 33 6.313D-2 WVLPCI(2) ES 34 1.894D-1 WYLPCI(3) ES 35 2.525D-1 WVLPCI(4) ES 36 3.47D-1 WVLPCI(5) ES 37 4.48D-1 WVLPCI(6) ES 38 5.05D-1 WVLPCI(7) ES 39 6.31D-1 WVLPCI(8) ES 40 2.099D6 PLPCS(1) LPCS ES 41 2.03D6 PLPCS(2) ES 42 1.961D6 PLPCS(3) ES 43 1.892D6 PLPCS(4) ES 44 1.824De PLPCS(5) ES 45 1.479D6 PLPCS(6) ES 46 8.928D5 PLPCS(7) ES 47 1.01342D5 PLPCS(8) ES l

A-6 48 0.D0 WVLPCS(1) ES 49 .041D0 WVLPCS(2) l ES 50 .063D0 WVLPCS(3) ES 51 .078D0 WVLPCS(4) ES 52 .0915D0 WVLPCS(5) ES 53 .1355D0 WULPCS(6) ES 54 .1895D0 WVLPCS(7) ES 55 .246D0 WVLPCS(8) ES

)

56 .75834D6 DRYWELL PRESSURE USED TO CLOSE ADS VALVES IF OPEN 57 .62046D6 DRYWELL PRESSURE USED TO RE-OPEN ADS VALVES AS DRYWELL

• • DEPRESSURIZES 72 7.0D6 PRCIC(1) RCIC ES 73 6.0D6 PRCIC(2) ES 74 5.0D6 PRCIC(3) ES 75 4.0D6 PRCIC(4) ES 76 3.0D6 PRCIC(5) ES 77 2.0D6 PRCIC(6) ES 1

78 1.134D6 PRCIC(7) ES 79 6.17D5 PRCIC(8) ES 80 .03788D0 WVRCIC(1) ES i

81 .03788D0 WVRCIC(2) ES 82 .03788D0 WVRCIC(3) ES 83 .0378800 WVRCIC(4) ES 84 .03788D0 WVRCIC(5) ES 95 .03788D0 WVRCIC(6) ES 86 .03788D0 WVRCIC(7) ES 87 0.0D0 WVRCIC(8) ES 88 57.66D0 ZLHPCI LOW WATER INITIATION FOR HPCI ES 89 1 1513D5 PSHPCI HIGH DRYWELL PRESSURE SET POINT FOR HPCI ES 90 25.D0 TDHPCI TIME DELAY FOR HPCI ES 91 6.205D5 PHHPCI MINIMUM VESSEL PRESSURE FOR HPCI TURBINE ES 92 50.D0 ZLHPCS LOW WATER INITIATION FOR HPCS ES 93 1.1513D5 PSHPCS HIGH DRYWELL PRESSURE SET POINT FOR HPCS ES 94 0.D0 TDHPCS TIME DELAY FOR HPCS ES 95 54.79D0 ZLLPCI LOW WATER INITIATION FOR LPCI ES 96 1.1513D5 PSLPCI HIGH DRYWELL PRESSURE SET POINT FOR LPCI ES 97 24.D0 TDLPCI TIME DELAY FOR LPCI ES 98 -1.D10 PLLPCI RPV-WETWELL PRESSURE TO CLOSE ADS IF OPEN 99 54.79D0 ZLLPCS LOW WATER INITIATION FOR LPCS ES 100 1.1513D5 PSLPCS HIGH DRYWELL PRESSURE SET POINT FOR LPCS ES 101 12.D0 TDLPCS TIME DELAY FOR LPCS ES 102 -1.D10 PLLPCS RPV-WETWELL PRESSURE TO RE-OPEN CLOSED ADS 103 57.66D0 ZLRCIC LOW WATER INITIATION FOR RCIC ES 104 1.1513D5 PSRCIC HIGH DRYWELL PRESSURE SET POINT FOR RCIC ES 105 30.D0 TDRCIC TIME DELAY FOR RCIC ES 106 6.205DS PHRCIC MINIMUM VESSEL PRESSURE FOR RCIC TURBINE ES 107 1.3514D5 HCST ENTHALPY OF CSi ES 108 290.D0 USWHX SERVICE WATER FLOW RATE (KG/S) PER RHR HTX ES

• • THE SAFETY RELIEF VALVES MUST BE ENTERED IN ORDER OF INCREASING

• • PRESSURE ACTUATION SET POINTS

• • I.E. GROUP 61 PSRV1=7.7D6

• • GROUP #2 PSRV2=7.75D6........ETC l

109 .8605D-2 ASRV1 FLOW AREA 0F RELIEF VALVE TYPE 01 ES 110 .8605D-2 ASRV2 FLOW AREA 0F RELIEF VALVE TYPE #2 ES 111 .8607D-2 ASRV3 FLOW AREA DF RELIEF VALVE TYPE 43 ES l

112 .8597D-2 ASRV4 FLOW AREA 0F RELIEF VALVE TYPE #4 ES

• • IF THE AREA FOR GROUP 45 IS INPUT AS A NEGATIVE NUMBER THEN THESE

• • VALUES WILL DISCHARGE DIRECTLY INTO THE DRYWELL, IF THE AREA IS

• • POSITIVE THEN THESE VALVES DISCHARGE INTO THE SUPPRESSION POOL 113 .8659D-2 ASRV5 FLOW AREA 0F SAFETY VALVE TYPE #5 ES 114 1.0D0 NSRV1 NUMBER OF TYPE il RELIEF VALVES ES 115 1.0D0 NSRV2 NUMBER OF TYPE $2 RELIEF VALVES ES 116 6.0D0 NSRV3 NUMBER OF TYPE 63 RELIEF VALVES ES

A-7 117 3.0D0 NSRV4 NUMBER OF TYPE 94 RELIEF VALVES ES 118 2.D0 NSRV5 NUMBER OF TYPE 45 RELIEF VALVES ES 119 0.D0 NADS1 NUMBER OF ADS VALVES IN GROUP 1 ES 120 0.D0 NADS2 NUMBER OF ADS VALVES IN GROUP 2 ES 121 2.D0 NADS3 NUMBER OF ADS VALVES IN GROUP 3 ES 122 3.D0 NADS4 NUMBER OF ADS VALVES IN GROUP 4 ES 123 7.7179D6 PSRV1 PRESSURE SETPOINT FOR #1 RELIEF VALVE ES 124 7.7179D6 PSRV2 PRESSURE SETPOINT FOR 42 RELIEF VALVE ES 125 7.7730D6 PSRV3 PRESSURE SETPOINT FOR 43 RELIEF VALVE FS 126 7.855706 PSRV4 PRESSURE SETPOINT FOR 44 RELIEF UALVE ES 127 8.6003D6 PSRV5 PRESSURE SETPOINT FOR 45 RELIEF VALVE ES 128 54.79D0 ZLADS LOW WATER LEVEL FOR ADS INITIATION ES 129 115.13D3 PSADS HIGH DRYWELL PRESSURE SET POINT FOR ADS ES 130 105.D0 TDADS TIME DELAY FOR ADS ACTUATION ES

• • LPCS LPCI HAVE NPSH REQUIRMENTS, RCIC AND HPCI TRIP ON HIGH SUPP POOL

• • TEMPERATURE 131 366.33D0 TCHPCI INLET TEMP LIMIT FOR HPCI ES 132 27.88D0 ZCLHPS PUMP CENTER LINE ELAVATION FOR HPCS ES 133 27.88DO ZCLLPI PUMP CENTER LINE ELAVATION FOR LPCI ES 134 27.88D0 ZCLLPS PUMP CENTER LINE ELAVATION FOR LPCS ES 135 366.33D0 TCRCIC INLET TEMP LIMIT FOR RCIC ES 136 300.D0 TWSW SERVICE WATER TEMP (RHR HEAT EXCHANGERS,TCOLD) ES 137 1.D0 TDDG1 HPCS LOAD DELAY TIME FOR DIESEL ES 138 11.D0 TDDG2 LPCI LOAD DELAY TIME FOR DIESEL ES 139 11.D0 TDDG3 LPCS LOAD DELAY TIME FOR DIESEL ES 140 1.D-3 XDDROP SPRAY DROPLET DIAMETER ES 141 3.66D0 XHSPWW SPRAY FALL HEIGHT IN WETWELL ES 142 14.02D0 XHSPDW SPRAY FALL HEIGHT IN DRYWELL . ES

• • THE FOLLOWING PUMP CURVES CAN BE USED TO DEFINE ANY INJECTION SYSTEM 143 1.35D5 HWHPSW ENTHALPY OF HIGH PRES SERVICE WATER (MARK I ) ES

• • IF VWHPSW IS INPUT AS A NEGATIVE NUMBER,THE HPSW WILL BE ROUTED TO

• • THE DRYWELL SPRAYS.

144 1.D-3 VWHPSW SPEC VOL OF HIGH PRES SERVICE WATER (MARK I ) ES

• • PUMP CURVES ARE WRITTEN S0 THAT THE FIRST PRESSURE POINT CORRESPONDS

• • TO THE FIRST FLDW RATE FOR THAT PUMP 145 1.D10 PHPSW(1) PPS VS. VOLUMETRIC FLOW FOR HPSW CORE INJECTION ES 146 1.D10 PHPSW(2) (MARK I CORE INJECTION) ES 147 1.D10 PHPSW(3) ES 148 1.D10 PHPSW(4) ES 149 1.D10 PHPSW(5) ES 150 1.D10 PHPSW(6) ES 151 1.D10 PHPSW(7) ES 152 1.D10 PHPSW(8) ES 153 .284D0 WVHPSW(1) ES 154 .284D0 WVHPSW(2) ES 155 .284DO WVHPSW(3) ES 156 .284D0 WVHPSW(4) ES 157 .284D0 WVHPSW(5) ES 158 .284D0 WVHPSW(6) ES 159 .284D0 WVHPSW(7) ES 160 .284D0 WVHPSW(8) ES 161 1.010 PDWSPR DRYWELL PRES SET PT FOR MARK III CONTAINMNT SPRAYS ES 162 1.D10 PWWSPR WETWELL PRES SET PT FOR MARK III CONTAINMNT SPRAYS ES 163 0.D0 TDSPR TIME DELAY FOR MARK III CONTAINMENT SPRAYS ES 164 2.413D5 PDSRV1 DEAD BAND FOR SRV41 ES 165 2.413D5 PDSRV2 DEAD BAND FOR SRV62 ES 166 2.413D5 PDSRV3 DEAD BAND FOR SRV43 ES 167 2.413D5 PDSRV4 DEAD BAND FOR SRV84 ES 168 2.413D5 PDSRV5 DEAD BAND FOR SRV85 ES 169 7.48D6 PTURHP(1) PPS-PWW VS. STEAM FLOW TO HPCI TURBINE ES 170 7.928D5 PTURHP(2) ES 171 7.928D5 PTURHP(3) ES 172 7.928D5 PTURHP(4) ES l

t

A-8 173 7.928D5 PTURHP(5) ES )

174 7.928D5 PTURHP(6) ES 175 7.928D5 PTURHP(7) ES 174 7.928D5 PTURHP(8) ES 177 23.D0 WSTHPI(1) ES 178 12.D0 WSTHPI(2) ES 179 12.D0 WSTHPI(3) ES 180 12.D0 WSTHPI(4) ES 181 12.D0 WSTHPI(5) ES 182 12.D0 USTHPI(6) ES 183 12.00 WSTHPI(7) ES 184 12.D0 WSTHPI(8) ES 185 7.7D6 PTURRI(1) PPS-PWW VS. STEAM FLOW TO RCIC TURBINE ES 186 1.013D4 PTURRI(2) ES 187 1.013D6 PTURRI(3) ES 188 1.013D6 PTURRI(4) ES 189 1.013D6 PTURRI(5) ES l 190 1.013D6 PTURRI(6) ES i 191 1.013D6 PTURRI(7) ES l 192 1.013D4 PTURRI(8) ES 193 3.500 WSTRCI(1) ES i 194 1.0D0 WSTRCI(2) ES 195 1.0D0 WSTRCI(3) ES 196 1.0D0 WSTRCI(4) ES 197 1.0D0 WSTRCI(5) ES 198 1.0D0 WSTRCI(6) ES 199 1.0D0 WSTRCI(7) ES 200 1.0D0 WSTRCI(8) ES 201 1.1355D4 PHTURH HIGH HPCI TURBINE EXHAUST PRESSURE ES 202 3.77D5 PHTURR HIGH RCIC TURBINE EXHAUST PRESSURE ES 203 9.0794D5 PCFAIL CONTAINMENT FAILURE PRESSURE ES 204 3.20206 PHLPCI HIGH PRESSURE TRIP FOR LPCI ES 205 3.202D4 PHLPCS HIGH PRESSURE TRIP FOR LPCS ES 204 33.64D0 ZHISP HIGH SUPP. POOL LEVEL FOR HP/RCIC SUCTION SWITCH ES 207 0.D0 ZLSPR LOW WATER LEVEL FOR AUTO WETWELL SPRAYS (M-III) ES

• • IF THE DETAILED HEAT EXCHANGER IS NOT USED ONLY SUPPLY THE

• • NTU VALUE AND NUMBER OF HTXS 208 0.D0 NTHX NUMBER DF TUBES IN RHR HTX ES 209 0.D0 NBHX NUMBER OF BAFFLES IN RHR HTX ES 210 0.D0 XIDTHX TUBE ID FOR RHR HTX ES 211 0.D0 XTTHX TUK WALL THICKNESS FOR RRR HIX ES 212 0.00 XTCHX TUK CENTER TO CENTER SPACING FOR RHR HTX ES 213 0.D0 XSHX SHELL LENGTH FOR RHR HTX ES 214 0.D0 RGFOUL FOULING FACTOR FOR RHR HTX ES 215 0.D0 KTHX THERMAL CONDUCTIVITY FOR TUBE WALL (RHR HIX) ES 216 0.00 XBCHX BAFFLE CUT LENGTH FOR RHR HTX ES 217 0.D0 XIDSHX SHELL ID FOR RHR HTX ES 218 0.D0 XSTHX BUNDLE TO SHELL GAP LENGTH FOR RHR HTX ES 219 .454D0 NTUHX1 NTU FOR RHR HTX 41 ES 220 .654D0 NTUHX2 NTU FOR RHR HTX #2 ES 221 2.D0 NHX1 NUMBER OF RHR LOOP #1 HTX ES 222 2.D0 NHX2 NUMBER OF RHR LDOP 42 HTX ES 223 1.D0 FHX TYPE OF RHR HTX(1= STRAIGHT TUBE.2=U TUBE) ES 224 21.603 TDBATT BATTERY OPERATION TIME FOR STATION BLACK-0UT ES

• • THE NPSH POINTS CORRESPOND WITH THE ABOVE FLOW RATE FUR THAT PUMP 233 0.D0 ZHDLPI NPSH CURVE FOR LPCI VS. FLOW (ABOVE) ES 234 0.D0 (NETERS) ES e

235 0.D0 ES 236 7.80800 ES 237 7.9768D0 ES 238 8.21300 ES 239 8.375D0 ES 240 8.79600 ES 1

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

A-9 241 0.D0 ZHDLPS NPSH CURVE FOR LPCS VS. FLOW (ABOVE) ES 242 7.969D0 ES )

243 8.024D0 ES 244 8.076DO ES 245 8.13400 Es i 246 8.381D0 ES l 247 9.116D0 ES 1 248 10.64D0 ES '

249 27.88D0 CENTER LI E ELEVATION FOR RCIC PUMP ES I 250 27.88D0 CENTER LINE ELEVATION FOR HPCI PUMP ES I 251 .0093D0 ACVENT AREA 0F CONTAINMENT VENT ES l 252 0.D0 ZCFAIL ELEVATION OF CONTAINMENT VENT IN WETWELL (MII ONLY)ES 253 28.8D0 ZSRVD AVERAGE ELEVATION OF SRV DISCHARGE IN SUPP POOL ES 254 0.D0 TGDWHX(1) COOLING CURVE FOR DRYWELL COOLERS 255 0.D0 TGDWHX(2) TEMP IN DRYWELL VS. HEAT LOSS RATE (J/S) 256 0.D0 TGDWHX(3) 257 0.D0 TGDWHX(4) 258 0.D0 TGDWHX(5) 259 0.D0 TGDWHX(6) 260 0.D0 TGDWHX(7) 261 0.D0 TGDWHX(8) 262 0.D0 QGDWHX(1) HEAT LOSS RATE FOR DRYWELL COOLERS (J/S) l 263 0.D0 OGDWHX(2) l 264 0.D0 OGDWHX(3) 265 0.D0 OGDWHX(4) 266 0.D0 OGDWHX(5) 267 0.D0 OGDWHX(6) 268 0.D0 OGDWHX(7) 269 0.D0 QGDWHX(8)

• • DW

• • DW

• DRYELL DW 01 .5D0 RELHDW RELATIVE HUMIDITY IN DRYWELL DW 02 4841.D0 VOLDW FREE VOLUME OF DRYWELL DW 03 36.55D0 ZDWF ELEVATION AT DRYWELL FLOOR DW 04 84.D0 ADWF AREA 0F DRYWELL FLOOR DW 05 37.24D0 ZWDWW ELEVATION OF DRYWELL-WETWELL WALL DW 06 0.D0 NIGDW NUMBER OF IGNITERS IN THE DRYWELL DW 07 0.D0 XIGDW AVERAGE DISTANCE FROM IGNITERS TO CEILING DW 08 0.D0 ACHDW FLOOR BURN AREA DW 09 557.0 ASEDDW AEROSOL SEDIMENTATION AREA DW

$$ WW

$$ WW

• WETELL WW 01 28.65D0 ZWWF ELEVATION AT WETWELL FLOOR WW 02 .169D0 AVB FLOW AREA THROUGH VACUUM BREAKERS WW 03 12.0D0 NUB NUMBER OF VACUUM BREAKERS WW 04 3.447D3 PSETVB PRESSURE SETPOINT FOR VACUUM BREAKERS W 05 2.757D3 PDVB DEAD BAND FOR VACUUM BREAKERS W 06 741990 VOLW FREE VOLUME OF WETWELL (MARK II AND MARK III ONLY) W 07 1.D0 RELHWW RELATIVE HUMIDITY IN WETWELL W 08 0.D0 NIGW NUMBER OF IGNITERS IN THE WETWELL WW 09 0.D0 XIGW AVERAGE DISTANCE FROM IGNITERS TO CEILING WW 10 0.D0 ACHW FLOOR BURN AREA WW 11 0.D0 AWWF AREA 0F WETELL FLOOR (MARK II) WW 12 929.0 ASEDW AEROSOL SEDIMENTATION AREA WW

• • PD

• • PD

• PEDESTAL PD 01 2.917D1 APDF AREA 0F PEDESTAL FLOOR PD 02 4.500 APDVT AREA 0F PEDESTAL-DRYELL OPENING PD 03 2.40D2 UOLPD VOLUME OF PEDESTAL PD l 04 34 55D0 ZWPDDW ELEVATION OF WALL BETWEEN PED AND DRYWELL PD

A-10 05 36.35D0 ZPDF ELEVATION AT PEDESTAL FLOOR PD 06 .5D0 RELHPD RELATIVE HUMIDITY IN PEDESTAL PD 07 0.D0 NIGPD NUMBER OF IGNITERS IN THE PEDESTAL PD 08 0.D0 XIGFD AVERAGE DISTANCE FROM IGNITERS TO CEILING PD 09 0.D0 ACHPD FLOOR BURN AREA PD 10 0.D0 AWPDVT WIDTH OF PEDESTAL DOOR (MARK II ONLY) PD 11 0.D0 ADCPD AREA 0F PEDESTAL DOWNCOMERS 12 0.D0 NDCPD NUMBER OF DOWNCOMERS IN PEDESTAL 13 7.D0 XHPDDW DISTANCE BETWEEN UPPER AND LOWER VENTS FOR

• • PED-DRYWELL NATURAL CIRCULATION 14 32.50 ASEDPD AEROSOL SEDIMENTATION AREA PD

• • TO

• • TO

• TORUS AND MARK II WETWELL TO 01 4.72D0 XRTOR MINOR RADIUS OF TORUS (MI ONLY) TO 02 *?4.7D0 XLTOR CIRCUMFERENCE OF TORUS (MI ONLY) TO 03 .292D0 ADC AREA 0F DOWNCOMER (MI AND MII ONLY) TO 04 96.D0 NDC NUMBER OF DOWNCOMERS (MI AND MII ONLY) IO

• • THE NEXT PARAMETER HAS BEEN REPLACES WITH MEDWW AND ME0WWS 05 848 43D0 VSSTOR VOLUME OF STRUCTURE IN TORUS (MI ONLY) TO 06 32.D0 ZBDC ELEVATION AT BOTTOM OF DOWNCOMER (MI AND MII ONLY) TO 07 34.74D0 ZTDC ELEVATION AT TOP OF DOWNCOMER (MI AND MII ONLY) TO 08 28.8D0 ZBTOR ELEVATION AT BOTTOM OF TORUS (MI ONLY) TO 09 0.016800 XTOR THICKNESS OF TORUS SHELL (MI ONLY) TO 10 5072.D0 ATR AREA 0F TORUS ROOM WALL (MI ONLY) TO 11 101341.D0 PTR PRESSURE IN TORUS ROOM (MI ONLY) TO 12 1.5D4 VOLTR YOLUME OF TORUS ROOM (MI ONLY) TO 13 34.747D0 ZUBTOR CENTER LINE ELEVATION OF VACUUM BRKRS(MI AND HII) TO gj .015D0 XTHDC THICKNESS OF DOWNCOMER PIPE (MII ONLY) TO

• INITIAL CONDITIONS IN 01 3.293D9 QPOWER CORE POWER IN 02 7.033D6 PPS0 INITIAL PRESSURE IN PRIMARY SfSTEM IN 03 1.041D5 PPD 0 INITIAL PRESSURE IN PEDESTAL IN 04 1.041D5 PDWO INITIAL PRESSURE IN DRYWELL IN 05 1 041D5 PWWO INITIAL PRESSURE IN WETWELL IN 06 33.22D0 ZSPDWO INIT.ELEV. OF WATER LEVEL IN DW SIDE OF SUPP. POOL IN

• • (FOR MI AND MII THIS IS ELEVATION IN DOWNCOMER) IN 07 33.22D0 ZSPWWO INIT.ELEV. OF WATER LEVEL IN WW SIDE OF SUPP. POOL IN 08 3.3D2 TPD0 INITIAL TEMPERATURE IN PEDESTAL IN 09 3.3D2 TDWO INITIAL TEMPERATURE IN DRYWELL IN 10 3.05D2 TWWO INITIAL TEMPERATURE IN WETWELL IN 11 3.05D2 TWSPO INITIAL TEMPERATURE OF SUPPRESSION POOL WATER IN 12 59.49D0 ZWSHO INITIAL ELEVATION OF WATER IN THE SHROUD IN 13 0.D0 MWCB0 MASS OF WATER IN UPPER POOL (MARKIII ONLY) IN 14 591.00 VCSTO VOLUME OF WATER IN CONDENSATE STORAGE TANK IN I 15 305.D0 TAMB AMBIENT TEMPERATURE IN 16 1.0134D5 PAMD AMBIENT PRESSURE IN

• HTSINKS HS 01 199.D0 ARS1 AREA 0F WALL 41 PEDESTAL-DRYWELL WALL HS 02 1507.D0 AHS2 AREA 0F WALL 42 DRYWELL WALL HS 03 0.00 AHS3 AREA 0F WALL 43 DRYWELL FLOUR HS 04 5073.D0 AHS4 AREA 0F WALL 44 TORUS ROOM WALL (MI ONLY) HS 09 1.3DO KHS1 THERMAL CONDUCTIVITY OF WALL 41 HS 10 1.3D0 KRS2 THERMAL CONDUCTIVITY OF WALL 42 HS 11 1.3D0 KHS3 THERMAL CONDUCTIVITY OF WALL 43 HS 12 1.300 KHS4 THERMAL CONDUCTIVITY OF WALL 44 HS 17 1.3200 XHS1 THICKNESS OF WALL 41 HS 18 1.83D0 XHS2 THICKNESS OF WALL 42 HS '

19 1.D0 XHS3 THICKNESS OF WALL 43 HS i

A-11 20 1.07D0 XHS4 THICKNESS OF WALL 44 HS 25 0.D0 XLHSI! INNER LINER THICKNESS FOR WALL 01 HS 26 2.713D-2 XLHSI2 INNER LINER THICKNESS FOR WALL 82 HS 27 0.D0 XLHSI3 INNER LINER THICKNESS FOR WALL 43 HS

. X R T ICK h OR Lb H 34 0.00 XLHS02 DUTER LINER THICKNESS FOR WALL 02 HS 35 0.D0 XLHS03 OUTER LINER THICKNESS FOR WALL 63 HS 36 0.D0 XLHSO4 OUTER LINER THICKNESS FOR WALL $4 HS 41 2300.D0 DHS1 DENSITY OF WALL $1 HS 42 2300.D0 DHS2 DENSITY OF WALL 42 HS 43 2300.00 DHS3 DENSITY OF WALL 43 HS 44 2300.D0 DHS4 DENSITY OF WALL 44 HS 49 880.D0 CPHS1 SPECIFIC HEAT FOR WALL 61 HS 50 880.D0 CPHS2 SPFCIFIC HEAT FOR WALL 42 HS 51 880.D0 CPHS3 SPECIFIC HEAT FOR WALL 03 HS 52 880.D0 CPHS4 SPECIFIC HEAT FOR WALL 64 HS 57 0.D0 MEDPD MASS OF EDUIPMENT IN PEDESTAL HS 58 1.9D5 NEODW MASS OF EQUIPMENT IN DRYWELL HS 59 5.5482D4 MEDWW MASS OF EQUIPMENT IN WETWELL HS 62 0.D0 AEDPD AREA 0F EQUIPMENT IN PEDESTAL HS 43 1.b33 AEODW AREA 0F EQUIPMENT IN DRYWELL HS 64 20.D0 AE0WW AREA 0F EDUIPMENT IN WETWELL HS 67 50.D0 HTOUTW HEAT TRANSFER COEFF. AT 00TER WALL HS

' 68 0.D0 RGAPI1 INNER LINER TO WALL GAP RESISTANCE #1 HS 69 1.D0 RGAPI2 INNER LINER TO WALL GAP RESISTANCE #2 HS 70 0.D0 RGAPI3 INNER LINER TO WALL GAP RESISTANCE #3 HS 71 0.D0 RGAPI4 INNER LINER TO WALL GAP RESISTANCE 44 HS 72 0.D0 RGAPI5 INNER LINER TO WALL GAP RESISTANCE #5 HS 73 0.D0 RGAPI6 INNER LINER TO WALL GAP RESISTANCE $6 HS 74 0.D0 RGAPI7 INNER LINER TO WALL GAP RESISTANCE 47 HS 75 0.D0 RGAPI8 INNER LINER TO WALL GAP RESISTANCE $8 HS 76 0.D0 RGAP01 OUTER LINER TO WALL GAP RESISTANCE 41 HS 77 0.D0 RGAP02 OUTER LINER TO WALL GAP RESISTANCE 42 HS 78 0.D0 RGAP03 DUTER LINER TO WALL GAP RESISTANCE #3 HS 79 0.D0 RGAPO4 OUTER LINER TO WALL GAP RESISTANCE #4 HS 80 0.D0 RGAP05 OUTER LINER TO WALL GAP RESISTANCE #5 HS 81 0.00 RGAP06 0 UTER LINER TO WALL GAP RESISTANCE 86 HS 82 0.D0 RGAP07 OUTER LINER TO WALL GAP RESISTANCE 47 HS 83 0.D0 RGAP08 OUTER LINER TO WALL GAP RESISTANCE #8 HS 84 5.3232D4 ME0WWS MASS EQUIP. WETWELL (SUBMERGED) HS 85 1.D2 AE0WWS AREA EDUIP. WETWELL (SUBMERGED) HS

• • HS

• • HS

• MODEL PARAMETERS FOR BUR 01 .005D0 FRCOEF FRICTION COEFFICIENT FOR CORIUM IN VFAIL MO 02 2.00-1 FMAXCP FRACTION OF TOTAL CORE MASS WHICH MUST MELT

• • TO FAIL THE CORE PLATE 03 50.D0 HTBLAD FUEL CHANNEL TO CONTROL BLADE HEAT TRANS. COEFF MO 04 300.D0 HTFB FILM BOILING HEAT TRANS. COEFF. MO 05 0.D0 FBLOCK FUEL CHANNEL BLOCKAGE PARAMETER M0

• • 0= BLOCKAGE AT TZ00FF 1=NO BLOCKAGE MO 04 2300.D0 TZOOFF OXIDATION CUT-OFF TEMPERATURE MO 07 .300 FACPF FRACTION OF CORE PLATE AREA THAT FAILS MO 08 5.D0 CDBPD FLAME BUOYANCY DRAG COEFFICIENT IN THE PEDESTAL MO 09 5.D0 CDBDW FLAME BUOYANCY DRAG C0 EFFICIENT IN THE DRYWELL MO 10 5.D0 CDBWW FLAME BUOYANCY DRAG COEFFICIENT IN THE WETWELL MO

. 11 5.D0 CDBCA FLAME BUOYANCY DRAG CDEFFICIENT IN COMPARTMENT A M0

! 12 5.00 CDBCD FLAME BUOYANCY DRAG C0 EFFICIENT IN COMPARTMENT B MO 13 .10D0 XCNREF CORIUM REFERENCE THERMAL BOUNDARY LAYER THICKNESS MO 14 1.D3 HTCMCR CORIUM-CRUST HEAT TRANSF. COEFF. USED IN DECOMP MO 15 0.05D0 XCMX MINIMUM CORIUM THICKNESS ON DRYWELL FLOOR AND PED M0

• • FLOOR (MARK II ONLY) NO 1

~

A-12 t

16 0.01D0

• • XDCMSP PARTICLE SIZE (DIAMETER) FOR CORIUM AS IT FALLS INTO SUPPRESSION POOL (MARK II ONLY)

M0 17 993 00 MO 18 TCFLAM CRITICAL FLAME TEMPERATURE M0 1.53D0 FCHTUR CHURN-TURBULENT CRITICAL FLOW PARAMETER 19 3.700 FDROP MO 20 DROPLET CRITICAL FLOW PARAMETER M0 3.D0 FFLOOD FLOODING FLOW PARAMETER 21 1.D0 FSPAR MO 22 2.D0 FVOL PARAMETER FOR BOTTOM-SPARGED STEAM VOID FRACTION MO 23 5.D-1 TTENTR PARAMETER FOR VOLUME SOURCE VOID FRACTION MODEL MD 24 .90D0 ENTRAINMENT EFFECTIVE EMPTYING TIME MO EW EMISSIVITY OF WATER 25 .9500 EWL EMISSIVITY OF WALL M0 26 .85D0 ECM MD 27 EMISSIVITY OF CORIUM MO

.6D0 EG EMISSIVITY OF GAS 28 .8500 EED MO 29 EMISSIVITY OF EQUIPMENT MO 0.500 F0VER 30 1.00 NPF FRACTION OF CORE SPRAY FLOW ALLOWED TO BYPASS COREMO 31 2.D0 FCDCDW NUMBER OF PENETRATIONS FAILED IN LOWER HEAD M0 l ** DOWNCOMER (MARK II ONLY)PERIMETER PER METER FROM PEDESTAL DOOR M0 32 0.14DO FCHF NO 33 .7500 COEFFICIENT FOR CHF CORRELATION IN PLSTM M0 34 .33D0 FCDBRK DISCHARGE COEFFICIENT FOR PIPE BREAK MO FENTR

• • NUMBER TO MULTIPLY KUTATELADZE CRITERION BY TO MD

• • REPRESENT DIFFICULTY (GT 1.D0) OR EASE (LT 1.00) MO FOR MATERIAL TO BE BLOWN OUT OF CAVITY 35 1.00 SCALU MO 34 1.00 SCALH SCALING FACTOR FOR ALL BURNING VELOCITIES MO

• • SCALING FACTOR FOR HT COEFFICIENTS TO PASSIVE HEAT SINKS MO 37 2.0D0 FUMIN MG 38 CLADDING SURFACE MULTIPLIER MD

.022 FHE1 39 .003 FME2 FIRST COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 40 .590 FHE3 SECOND COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 41 .330 FE4 THIRD COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 42 1.0D-5 DHE FOURTH COEFF IN HENRY-EPSTEIN AEROSOL MODEL MO 43 1.0 DENSITY LIMIT IN HENRY-EPSTEIN AEROSOL MODEL MO

• • FCSIVP GROUP 2 (CSI) I GRUOP 6 (CSOH) VAPOR PRESSURE MULTIPLIER

• • - NUMBER USES ANL CSOH VAP PRESS, + USES SANDIA CSOH

• • VAP PRESS 2 CONCRETE PROPERTIES MO 01 56.D0 MO 02 MOLWCN MOLECULAR WEIGHT OF CONCRETE CN 1743.D0 TCNMP CONCRETE ELTING TEMPERATURE 03 .9D6 CN 04 45.90 LHRCN REACTION ENERGY FOR CONCRETE DECOMPOSITION CN 05 DCFWCN FREE WATER DENSITY IN CONCRETE CN 45.D0 DCCWCN COMBINED WATER DENSITY IN CONCRETE

' CN i **

/ D1 H ET T E MELTING

• FISSION PRODUCTS 01 .020D0 FI i

• F0P(1) PERCENT POER IN FISSION PRODUCT GROUP 1 FI 2

~

04 05

. BOO 0.D0 0.00 F0Pk Rch  !!h!b!$

F0P(4) PERCENT POWER IN FISSION PRODUCT GROUP 4 h b$ h!

FI FOP (5) PERCENT POWER IN FISSION PRODUCT GROUP 5 FI 08 b b 0F F ON DU GR0bP -

S 34.0 E P(2) MASS OF FISSION PRODUCT GROUP 2 -CSI 09 34.9 FI MFP(3) MASS OF FISSION PRODUCT GROUP 3 -TE FI

' ) F SS 0 GR 5 12 213 8 13 1050.0 E P(6) MASS OF FISSION PRODUCT GROUP 6 -CSON FI 14 MSM0(1) MASS OF SN IN CORE REGION FI 4

432 0 MSM0(2) MASS OF MN IN CORE REGION 15 0.000 FI 16 FDSP(1) SPRAY REMOVAL LANDA FOR FP GROUP 1 FI 0 0D0 FDSP(2) SPRAY REMOVAL LANDA FOR FP GROUP 2 FI I w e-----,-------m.w ----------yg, w , y y-- y

A-13 17 0.0D0 FDSP(3) SPRAY REMOVAL LAMDA FOR FP GROUP 3 FI 18 0.0D0 FDSP(4) SPRAY REMOVAL LAMDA FOR FP GROUP 4 FI 19 0.0D0 FDSP(5) SPRAY REMOVAL LANDA FOR FP GROUP 5 FI 20 0.0D0 FDSP(6) SPRAY REMOVAL LAMDA FOR FP GROUP 6 FI 21 400.00 FDFSP(1) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 1 FI 22 600.D0 FDFSP(2) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 2 FI 23 600.0D0 FDFSP(3) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 3 FI 24 600.0D0 FDFSP(4) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 4 FI 25 600.0D0 FDFSP(5) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 5 FI 26 600.0D0 FDFSP(6) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 6 FI 27 1000.D0 FDFSP(1) SRV DECON. FACTOR FOR FP GROUP 1 FI 28 1000.D0 FDFSP(2) SRV DECON. FACTOR FOR FP GROUP 2 FI 29 1000.0D0 FDFSP(3) SRU DECON. FACTOR FCR FP GROUP 3 FI 30 1000.0D0 FDFSP(4) SRV DECON. FACTOR FOR FP GROUP 4 FI 31 1000.0D0 FDFSP(5) SRV DECON. FACTOR FOR FP GROUP 5 FI 32 1000.0D0 FDFSP(6) SRV DECON. FACTOR FOR FP GROUP 6 FI

• • FI

• • FI

• CONTROL CARDS CC 01 1 IBWR CDNTAINMENT TYPE (MARK 1,2 0R 3 ) CC 02 44 IRSTW UNIT NUMBER TO WRITE RESTART FILE (MAIN) CC 03 47 IHOW UNIT NUMBER TO WRITE RESTART FILE (HEATUP) CC 04 40 IPOUT UNIT NUMBER TO WRITE PROGRAM DUTPUT FILE CC 05 41 IPLT1 FIRST PLOT FILE NUMBER (TOTAL OF 4 FILES) CC 06 600 IPTSMX MAXIMUM NUMBER OF PLOTTED POINTS CC 07 6 IPTSPK MAXIMUM NUMBER OF PLOT POINTS TRACED FOR FULL CC

• • SCALE SPIKE CC 08 150 IPTSAV NUMBER OF POINTS SAVED FOR VARIABLE PLOT CC 09 1 ISUMM

SUMMARY

DATA (0=ALL EVENTS,1= SHORTER LIST) CC 10 48 ISUM

SUMMARY

FILE NUMBER CC 11 1 IRUNG 1=1ST ORDER R-Ke2:2ND ORDER R-K CC 12 1 IFREEZ 1= 00 FREEZE FRONT CALC. (0=NO CALC.) CC 13 6 INPGRP NUMBER OF TRACE GAS TYPES (FISSION PRODUCTS) CC 14 0 IRET WRITE RETAIN PLOT FILE CC 15 49 IFPPLT RETAIN PLOT FILE UNIT NUMBER CC 36 0 IFPRAT 1=NUREG-0772 FP RELEASES i 0=CUBICCIOTTI CC

• • TD

• • TD

• TIMING DATA ID 01 20.D0 TDMAX MAXIMUM ALLOWED TIME STEP TD 02 1.D-3 TDMIN MINIMUM ALLOWED TIME STEP TD 03 4.D-2 FMCHMX MAXIMUM MASS CHANGE FRACTION FOR INTEGRATION TD 04 5.D-2 FUCHMX MAXIMUM ENERGY CHANGE FRACTION FOR INTEGRATION TD 05 1.D10 MAXMST MAXIMUM MASS OF STEAM CHANGE PER TIME STEP IN PS TD

• • BY PUTTING A *BR AT THE END OF THE PARAMETER FILE, THE 0'JTPUT WILL

• • BE IN BRITISH UNITS, OTHERWISE THE OUTPUT WILL BE IN SI UNITS.

• BR

B-1 APPENDIX B Supplemental Plots for the WASH-1400 Comparison Sequences f

l l

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B-3 SUPPLEMENTAL PLOTS FOR SEQUENCE TW l

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Fig. B.12 Cesium iodide released to environment, kg.

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TIME HR i

Fig. B.14 Tellurium released to environment, kg.

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B-19 SUPPLEMENTAL PLOTS FOR SEQUENCE TC J

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M i f f f f f f f f f I f f f f i f f f r if f f f I f f I C E I O DM ESANilW

A tornic industrist Forurn, Inc.

1101 Wesconsen Avanue Bethesda. Mo 20814 4891 Telephone:(301) G54 9260 TWX 7108249602 ATOMIC Fon DC e

John R. Siegel Vice Pres dent March 13, 1987 Mr. Harold Denton Office of Nuclear Regulatory Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 ,,

Dear Mr. Denton:

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 dup'licating .

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 a use of this material to continue.

The Atomic Industrial Forum, Inc. hereby grants the U.S. Nuclear 3 '

Regulatory Commission (NRC) authority to duplicate copyrighted IDCOR materials as necessary for use by NRC and NRC. contractor personnel in carrying out their regulatory mission. This grant of authority, however, does not apply to the Modular Accident Analysis Program (MAAP) or any of its associated documentation.

Please contact Roger Huston, of our staff, if you have any additional questions regarding this issue.

Sincerely, J %7  !

JRS:hlw.

cc: Cordell Reed Anthony Buhl

_ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ . . _ - _