Text

Draft, Peach Bottom Atomic Power Station Integrated Containment Analysis, Technical Rept
	ML20095C773
Person / Time
Site:	Peach Bottom, Sequoyah, 05000000
Issue date:	08/07/1984
From:	; INDUSTRY DEGRADED CORE RULEMAKING PROGRAM
To:	;
Shared Package
ML20095C674	List: ML20095C690; ML20095C701; ML20095C725; ML20095C773;
References
	23.1, NUDOCS 8408230131
	Download: ML20095C773 (75)
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IDC01:

Program Report DRAFT TeclinicalReport n.1 PEACH BOTTOM ATOMIC POWER STATION INTERGATED CONTAINMENT ANALYSIS I

f 4-7-37 B((h O ph The Industry Degraded Core Rulemaking Program Spcasored By the Nuclear Industry

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TABLE OF CONTENTS DllAFT Page LIST OF FIGURES ..........................y LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . ix

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Statement of the Problem ............. . . . . . 1-1 1.2 Relationship to Other Tasks . . . . . . . . . . . . . . . . . 1-1 2.0 STRATEGY AND METHODOLOGY , . . . . . . . . . . . . . . . . . . . . 2-1 2.1 References . . . . . . . . . . . . . . . . . . . . . . . . . 2-2

3.0 DESCRIPTION

OF MODELS AND MAJOR ASSUMPTIONS . . . . . . . . . . . 3-1 3.1 Plant Specific Information . . . . . . . . . . . . . . . . . 3-1 3.1.1 Nuclear System . . . . . . . . . . . . . . . . . . . . 3-1 3.1.2 Containment . . . . . . . . . . . . . . . . . . . . . 3-4 3.2 Modular Accident Analysis Program (MAAP) . . . . . . . . . . 3-6 3.2.1 MAAP Nodalization ............ . . . . . . 3-6 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 3.2.7 Fission Product and Aerosol Release from Core-Concrete Attack . . . . . . . . . . . . . . . . . 3-19 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 Model Inputs . . . . . . . . . . . . . . . . . . . . . 3-23 3.3.4 Influence on Fission Product Release . . . . . . . . . 3-24

1 UHM TABLEOFCONTENTS(Continued)

Page 3.4 References . . . . . . . . . . . . . . . . . . . . . . . . . 3-26 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 . . . . . . . . . . . . . 4-24 4.3.1 Sequence Description . . . . . . . . . . . . . . . . . 4-24 4.3.2 Primary System and Containment Response . . . . . . . 4-25 4.4 Plant Response to the TQVW Sequence . . . . . . . . . . . . . 4-32 4.4.1 Sequence Description . . . . . . . . . . . . . . . . . 4-32 4.4.2 Primary System and Containment Response . . . . . . . 4-32 4.5 References . . . . . . . . . . . . . . . . . . . . . . . . . 4-41 5.0 PLANT RESPONSE WITH REC 0VERY ACTIONS . . . . . . . . . . . . . . . 5-1 5.1 Possible Actions . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Seque nce 1 ( TW) . . . . . . . . . . . . . . . . . . . . . . . 5-6 5.3 Sequence 2 (TC) . . . . . . . . . . . . . . . . . . . . . . . 5-10 5.4 Sequence 3 (S;E) . . . . . . . . . . . . . . . . . . . . . . 5-14 I

5.5 Sequence 4 (TQVW) . . . . . . . . . . . . . . . . . . . . . . 5-19 6.0 FISSION PRODUCT RELEASE. TRANSPORT AND DEPOSITION . . . . . . . . 6-1 l

TABLE OF CONTENTS (Continued)

DRAFT Page 6.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Model ing Approach . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Sequences Evaluated . . . . . . . . . . . . . . . . . . . . . 6-3 l

6.3.1 TW Fission Product Release . . . . . . . . . . . . . . 6-3 6.3.2 TC Fission Product Release . . . . . . . . . . . . . . 6-6 6.3.3 S E j Fission Product Release .............6-8 6.3.4 TQVW Fission Product Release . . . . . . . . . . . . . 6-8 6.4 References . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 7.0

SUMMARY

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

8.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 APPENDIX A - Peach Bottom Parameter File . . . . . . . . . . . . . . A-1 APPENDIX B - Supplemental Plots for the Base Accident Sequences . . B-1 Supplemental Plots for Sequence TW . . . . . . . . . . B-3 Supplemental Plots for Sequence TC ..........B-17 Supplemental Plots for Sequence j S E . . . . . . . . . . B-29 Supplemental Plots for Sequence TQVW . . . . . . . . . B-43

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DRAFT

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DHAT l LIST OF FIGURES Figure No. Page 3.1 Primary systen model . . . . . . . . . . . . . . . . . . 3-7 3.2 Mark I nodalization . . . . . . . . . . . . . . . . . . 3-9 3.3 BWR Mark I system features modeled in MAAP . . . . . . . 3-11 3.4 BWR natural circulation model . . . . . . . . . . . . . 3-15 3.5 Fission product transport paths for the primary system and containment . . . . . . . . . . . . . . . . . 3-17 3.6 Nodalization scheme for the reactor building . . . . . . 3-18 3.7 Peach Bottom secondary containment . . . . . . . . . . . 3 - 21 4.1 Pressure in the drywell . . . . . . . . . . . . . . . . 4-5 4.2 Temperature of gas in the drywell . . . . . . . . . . . 4-6

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4. .? Temperature of the suppression pool . . . . . . . . . . 4-7 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 4.8 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 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 4.17 Average corium temperature in the pedestal . . . . . . . 4-31

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DRAFT LIST OF FIGURES (Continued)

Figure No. Page 4.18 Pressure in the drywell . . . . . . . . . . . . . . . . 4-34 4.19 Temperature of gas in the drywell . . . . . . . . . . . 4-35 4.20 Temperature of the suppression pool . . . . . . . . . . 4-36 4.21 Concrete ablation depth in the pedestal . . . . . . . . 4-37 4.22 Average corium temperature in the pedestal . . . . . . . 4-38 5.1 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-8 5.2 Reactor pressure vessel water level . . . . . . . . . . 5-9 5.3 Reactor water level . . . . . . . . . . . . . . . . . . 5-12 5.4 D rywell p res s u re . . . . . . . . . . . . . . . . . . . . 5-13 5.5 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-17 5.6 Temperature in drywell . . . . . . . . . . . . . . . . . 5-18 5.7 . Reactor pressure vessel water level . . . . . . . . . . 5-21 5.8 Pressure in drywell . . . . . . . . . . . . . . . . . . 5-22 5.9 Temperature of the suppression pool . . . . . . . . . . 5-23 B.1 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-4 B.2 Reactor vessel water level . . . . . . . . . . . . . . . . B-5 B.3 Temperature o f s tructure, *F . . . . . . . . . . . . . . . B-6 B.4 Fission product decay power on structure, Btu /hr . . . . . B-7 B.5 Total C0 generated . . . . . . . . . . . . . . . . . . . . B-8 B.6 Mass of water in the pedestal ........ . . . . . . B-9 B.7 Volumetric flow out of containment . . . . . . . . . . . . B-10 B8 Mass of UO in 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

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LIST OF FIGURES (Continued)

Figure No. Page B.11 Cesium and iodine released from containment, kg . . . . . B-14 B.12 Cesium and iodine released to environment, kg . . . . . . B-15 B.13 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-18 B.14 Reactor vessel water level . . . . . . . . . . . . . . . . B-19 B.15 Temperature of structure. *F . . . . . . . . . . . . . . . B-20 B.16 Fission product decay power on structure, Btu /hr . . . . . B-21 B.17 Total C0 gene ra ted . . . . . . . . . . . . . . . . . . . . B-22 B.18 Mass of water in the pedestal . . . . . . . . . . . . . . B-23 B.19 Mass flow out of containment . . . . . . . . . . . . . . . B-24 B.20 Reactor building gas temperature, 'F . . . . . . . . . . . B-25 B.21 Reactor building steam partial pressure . . . . . . . . . B-26 B.22 Cesium and iodine released from containment, kg . . . . . B-27 B.23 Mass of cesium and iodine released to environment . . . . B-28 B.24 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-30 B.25 Reactor vessel water level . . . . . . . . . . . . . . . . B-31 B.26 Temperature of structure, 'F . . . . . . . . . . . . . . . B-32 B.27 Fissien product decay power on structure, Btu /hr . . . . . B-33 B.28 Total CO generated . . . . . . . . . . . . . . . . . . . . B-34 B.29 Mass of water in the pedestal . . . . . . . . . . . . . . B-35 B.30 Volumetric flow out of containment . . . . . . . . . . . . B-36 B.31 Mass of UO i n core region . . . . . . . . . . . . . . . . B-37 2

B.32 Gas temperature in reactor building, 'F . . . . . . . . . B-38

B.33 Steam pressure in reactor building, Pa . . . . . . . . . . B-39 B.34 Cesium and iodine released from containment, kg . . . . . B-40 i

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- viii - l LIST OF FIGURES (Continued)

Figure No. Page B.35 Costum and iodine released to environment, kg . . . . . . B-41 B.36 Total H2 generated . . . . . . . . . . . . . . . . . . . . B-44 B.37 Reactor vessel water level . . . . . . . . . . . . . . . . B-45 B.38 Temperature of structure, 'F . . . . . . . . . . . . . . . B-46 8.39 Fission product decay power on structure, Btu /hr . . . . . B-47 B.40 Total C0 generated . . . . . . . . . . . . . . . . . . . . B-48 B.41 Mass of water in the pedestal . . . . . . . . . . . . . . B-49 B.42 Mass of UO i n core region . . . . . . . . . . . . . . . . B-50 2

B.43 Gas temperature in reactor building, *F . . . . . . . . . B-51 l

B.44 Steam pressure in reactor building, Pa . . . . . . . . . . B-52 B.45 Cesium and iodine released from containment, kg . . . . . B-53 I

B.46 Cesium and iodine released to environment, kg . . . . . . B-54 1

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LIST OF TABLES DRAFT Table No. Page 3.1 Initial Inventories of Fission Products and Structural Materials Released as Aerosols . . . . . . . 3-13 3.2 Reactor Building Model Inputs . . . . . . . . . . . . . 3-25 f

4.1 Peach Bottom - TW Event Sumary . . . . . . . . . . . . 4-4 4.2 Peach Bottom - TC Event Sumary . . . . . . . . . . . . 4-15 4.3 Peach Bottom - Sj E Event Sumary . . . . . . . . . . . . 4-26 4.4 Peach Bottom - TQVW Event Sumary . . . . . . . . . . . 4-33 5.1 Injection 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 . . . 5-11 5.7 Sj E With Selected Operator Actions - Event Sumary . . . 5-16 5.8 TQVW With Selected Operator Actions - Event Sumary . . 5-20 6.1 Distribution of Csl in Plant and Environment (Fraction of Core Inventory) . . . . . . . . . . . . . . 6-4 6.2 TW Fission Product Release . . . . . . . . . . . . . . . 6-5 6.3 TC Fission Product Release . . . . . . . . . . . . . . . 6-7 6.4 Sj E Fission Product Release ..............6-9 i 6.5 TQVW Fission Product Release . . . . . . . . . . . . . . 6-11 7.1 Sumary of MAAP Results for Base Sequences . . . . . . . 7-2 1

7.2 Sumary of Fission Product Release Fractions . . . . . . 7-3 !

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1-1

1.0 INTRODUCTION

1.1 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. The 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 Risk 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 the WASH-1400 BWR analyses which identified those sequences most likely to lead to core melting. The results of these analyses indicate the time windows available for operators to 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 is sufficient to demonstrate the effects that simple, individual actions would 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 regions of the primary and secondary containment structures. For those sequences in which containment integrity is violated, the release of fission products to the surrounding enviroment is calculated. The influence of a few existing systems with operator action is described in Section 5.

1.2 Relationship to Other Tasks The primary system and containment response analyses of IDCOR l Subtask 23.1 are carried out with the Modular Accident Analysis Program. This includes models developed in IDCOR subtasks ll, 12,14,15 and 16 for thermal-

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hydraulic behavior as well as fission produce release transport and deposition within both the primary system and containment. The accident sequences P analyzed were developed by' considering the dominant core melt accident se-quences presented in Subtask 3.2, Assess Dominant Sequences. Selected primary c.ontainment failure modes were chosen to demonstrate the radionuclide trans-port phenomena for the best-estimate analyses, g

The ultimate structural capability of containments associated with

the reference pidnts and other typical designs was assessed in IDCOR Subtask 10.1. This task defined the containment failure pressure and location assumed

i 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 containment, fdr 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 effects on the severe accident sequences analyzed for Peach Botton -- 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 some of the inherent and intervention means of terminating the various core damage sequences.

It was not the intent of Task 23.1 to address the likelihood of occurrence of the particular sequence:: and operator actions, but rather to assume these situations and analyze the accompanying containment challenges and release of fission pr$ucts 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.

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DRAFT 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 the 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 sequences considered and the specific design characteristics of Peach Bottom Atomic Power Station.

2.0 STRATEGY AND METHODOLOGY The basic strategy of this subtask was to analyze accident sequences which have been previously identified as key potential contributors to core melt frequency. These analyses consisted of plant thermal hydraulic response and fission product transport if the progression of the accident sequence led to core degradation and melting. These analyses include the performance of the ECCS 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 themal-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 fuel. Natural and forced circulation within the primary system is modeled both before and after vessel failure and is integrated with the fission produce release model to determine the transport of vapors and 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, thermal-hydraulic calculations were performed both with and without selected operator actions during the accident. The " base case" analyses, which assume only minimal operator response during the accident, e.stablish a reference system response during each of the accident scenarios. The " operator action" analyses are branch calculations of the base cases. These operator interven-tion cases demonstrate the effect of selected operator actions on the progres-sion of an accident, based on the time windows available to the operator to take such action. Additional uncertainty and sensitivity analyses have been performed on several key parameters associated with the accident response.

These are reported in Ref. [2.4].

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

l 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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DRAFT

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 comercial 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 pressure of 1250 psig. The nominal operating pressure is 1020 psia in the stea::. space above the separators. The reactor core is cooled

- by demineralized 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 dried by 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 lire is provided with two isolation valves in series, one on each side of the primary containment barrier.

DRAFT When a scram is initiated by the Reactor Protection System, the Control Rod Drive system (CRD) 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 from an accumulator 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 abnormal operational transients or accidents.

Although process lines which penetrate the primary containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steam lines, because of their large )

size and large mass flow rates, are given special isolation consideration.

Two automatic isolation valves, each powered by both air pressure and spring force, are provided in each main steam line.

The Reactor Core Isolation Cooling system (RCIC) provides makeup water to the reactor vessel whenever 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 during and af ter plant shutdown.

2. Injection of water into the reactor vessel, following a LOCA, rapidly enough to reflood the core and prevent excessive 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 accomplished by cooling and recirculating the water inside the l

primary containment.

DRAFT The redundancy of the equipmer.t provided for containment cooling is further extended by a separate part of the RHR jstem which sprays cooling water into the drywell.

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 ifquid control (SLC) system provides a redundant, independent, and different way 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 not enough 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 within 10 seconds. The diesel generator system is arranged with four incepen-dent 4-kV buses for each unit, each bus being connected to one diesel genera-

! tor. Each diesel generator starts automatically upon loss of off-site power cr detection of a nuclear accident. The necessary engineered safeguard system l

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i DRAFTloads are applied on a preset time sequence. Each generator operates indepen-dently 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 any single failure in the DC system.

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, vacuum 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 l

resulting increased drywell pressure would force a mixture of drywell at-mosphere, steam, and water through the vents into the suppression pool, resulting in a pressure reduction in the drywell due to steam condensation.

l l Vacuum breakers are provided in the vent headers and located in the suppression chamber, to equalize the pressure between the drywell and the.

I suppression chamber. A vacuum breaker system is also provided between the suppression 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.

BRAFT The vent system conducts flow from the drywell to the suppression chamber and distributes this flow unifonnly 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 suppression chamber-to-secondary containment vacuum breaker system limit the pressure differential 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 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-sectional diameter of 31 ft. It contains approximately 123,000 ft3 of water and has a gas space volume of approximately 132,000 ft3. The drywell vents are connected to a 4 ft. 9 inch diameter vent header, in the form of a torus, which is contained f thin the airspace of the suppression chamber. Projecting downward from the header are 96 downcomer pipes, nominally 24 inch in diameter and terminating 4 f t. below the design water level of the pool.

The vent system outside the torus consists of eight circular vent pipes, each having a diameter of 6 ft. 9 inches. These vent pipes are con-nected to the vent header located inside the torus. The vent pipes and vent header have the same temperature and pressure design requirements as the containment. Jet deflectors are provided in the drywell at the entrance of each vent pipe to prevent damage to the vent 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.

DRAFT 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 the primary containment as a means of reducing containment pressure following a LOCA. This capability is in excess of the required energy removal capabili-ty and can be placed into service at the discretion of the operator.

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)

[3.1] to analyze the major degraded core accident scenarios for both Pres-surized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The FAAP code is designed to provide realistic assessments for severe core damage accident sequences using first principle models for the major phenomena that govern the accident progression, the release of fission products from the fuel matrix, the transport of these fission products and their deposition within the primary system and containment. The following sections describe the primary system and contairrnent nodalization and include a description of the safety systems modeled in the MAAP. A complete Peach Bottom parameter file is given in Appendix A.I.

3.2.1 MAAP Nodalization l

l The BWR -primary system nodes are illustrated in Figure 3.1 and include the lower plenum, downcomer, core, and upper plenum. Also indicated are the flow entry locations for CRD flow, feedwater, 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

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DRAFT 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 specificatic,n 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.

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 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.

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3-10 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, lor pressure core spray (LPCS),

standby liquid control system (SLCS), and high pressure service water (HPSW).

In addition to these plant systems, MAAP nodalizet both the primary system and containmer.t 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 IDCOR 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 with the CIRC module in MAAP and includes models for the fission product j source 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.

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DRAFT 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 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 sufficie.1t flow exists to remove the volatile fission products as saturated vapor. Once this flow is detemined, the removal of the remaining less volatile, species are evaluated based upon saturation of this calculated flow.

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

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

1. Cesium and iodine combine for form Cs! upon entry to the fission product release pathway. The excess cesium forms Cs0H.

Both chemical species exhibit similar physical behavior, hence f.he source rate for the Cs,I fission product group is assumed to be the sum of the Cs and I release rates. As stated above, it is assumed to be liberated in vapor form.

2. Tellurium is assumed to be released as vaporized Te0

2

Table 3.1 DRAFT INITIAL INVENTORIES OF FISSION PRODUCTS AND STRUCTURAL MATERIALS RELEASED AS AEROSOLS 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

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

4. Strontium and ruthenium represent their respectivc 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 Substantial quantities of fission products are released during core degradation, but before vessel failure. Gas flow through the primary system detennines the aerosol transport and deposition throughout the reactor vessel.

Following vessel failure, fission products could remain within the primary system and subsequently heat the adjacent structures. As the structure and gas temperatures increases, density differences within the primary system would result in natural circulation flows that could distribute both heat and mass throughout the primary system.

u 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 nadali-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 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

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RAFTmixture 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.

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 system 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 applied state-of-the-art fission product behavior models to produce the RETAIN code, which describes the aerosol agglomeration and removal processes based upon an assumed log normal distribution [3.4].

Both vapor and aerosol forms of fission products are considered. MAAP repre-sents the aerosol removal rate due to settling as a function of the aerosol cloud density [3.5]. This is consistent with the general behavior predicted by detailed descriptions, such as RETAIN and also large scale experiments.

MAAP models physical mechanisms for vapor condensation on structures and aerosol retention due to steam condensation in addition to gravitational settling. These removal processes substantially reduce the magnitude of the release to the environment.

The primary system and containment nodalization for fission product transport are the same as those used for the therral hydraulic calculations.

The specific transport paths are illustrated in Fig. 3.5 are for the primary system and containment and in Fig. 3.6 for the reactor building and the SGTS.

The key assumptions in the aerosol modeling are:

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FLOW PATH FOR TOVW SEOUENCE, ALSO PORTION OF FLOW GREATER THAN SGTS CAPACITY FOR THE OTHER SEOUENCES B-Fig. 3.6 Nodalization scheme for the reactor building.

1.

DBAFT Cesium and iodine are assumed to be released as Cs! with excess cestian 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.

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 tne 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 PAAP (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 and 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.

3.2.7 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.

1 DRAFT 3.3 Analysis of Reactor Building Thennal-Hydraulic Conditions 3.3.1 Reactor Building and Standby Gas Treatment System (SGTS)

Each of Reach Sottom 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 comunication between them through open hatches. An equipment transfer shaf t from the 135' level up to the refueling floor is the major pathway for comunication 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 CRD hydraulic system, neutron monitoring system and other instrumentation partially separated by several partition and shield walls. Elevation 165' to 193' contains auxiliary pumps, heat exchangers and 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 build-ing 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 elevations.

The top elevation (234' to roof at 296') is essentially a wide open 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.

l The SGTS takes suction from multiple intakes located on all major elevations in the reactor building. Tests have indicated that reactor

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3-22 Un building inleakage at Unit 2 is approximately 5500 cfm when the building is isolated 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 if the temperature in a given region of the building exceeds 74*C (165'F). This temperature limitation can control 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-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 or containment failure, and the imposed flows resulting from the SGTS.

The SGTS, which provides a suction flow at each elevation,is normal-ly fed by inleakage from the outside, will then flow passing through the l

respective volume and into the ducts and filters in the system. This flow is partitioned between the various elevations and is represented in the computer model by a specific suction flow at each elevation. The inleakage at each

! elevation is determined by the strength of the source at that elevation, including flow from other compartments. For example, following containment l failure, the region at the equator of the drywell has a substantial source l

I 1

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MAFT from the drywell which can be greater than the suction flow for that eleva- j i

tion. As a result. the inleakage from the environment to this volume would be reduced to zero with any excess flow going to the higher elevations. 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 refueling floor has a direct connection to the' environment representing the leakage through the sheet metal siding or the opening of blowout panels.

The physical processes occurring in each volume, including heat losses to the structural surfaces, thennal 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 shaf t, 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. The relative rates of single phase and two-phase energy transfer determine the response of each reactor building region and subse-quently determine the flow between the three compartments. 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 associ-ated 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 As discussed in Section 3.3.2 and illustrated in Section 3.4.3, the reactor building is characterized by three major volumes which are intimately coupled. The first volume (or compartment) represents the lower three major elevations of the reactor building for those sequences in which SGTS is

assumed to be operating. The second compartment represents the volume above elevation 195', and the third compartment represents the volume above the refueling floor, elevation 234'. For sequences in which the SGTS would be unavailable, 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 each of these compartments are listed in Table 3.2.

The SGTS exhaust rates from each compartment used in the model are also presented in Table 3.2. These rates are based on a total SGTS flow of 10,500 cfm assuming that the initial pressure spike following containment failure causes increased building leakage are:, (i.e. blowout panels) result-ing in maximum SGTS flow as the system tries to maintain a negative pressure in the reactor building. The flows are proportioned for each compartment in accordance with plant data for the actual flows from the elevations. In accordance with the design, if a compartment temperature exceeds the 74*C limit on the fusible links for the fire dampers, the SGTS flows in that compartment are zero thereafter.

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 the fission product retention for this reactor design. In particular, it has a substantial influ-l ence on retaining the fission products within the plant and limiting the l 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 l condensation, gravitational settling, and steam condensation. Flow through I 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.

l

f Table 3.2 DRAFT REACTOR BUILDING MODEL INPUTS Compartment Vol m Surface Area SGTS Exhaust *

(m ) (m) (m/sec) 1 36.500 15,800 2.83 1 (Wf thout 9,940 10,200 -

SGTSFlcw) 2 4,600 2,500 0.47 3 27,450 4300 (Steel) 1.42

SGTS exhaust is zero for the station blackout sequence. Also, the SGTS flow from each compartment is zero if the compartment temperature exceeds the 75'C limit for the fusable links on the fire dampers.

l

r l .

Inclusion of the SGTS with the associated suction flows for each major volume increases the flow through the reactor building volumes, thereby decreasing somewhat the material deposited in these volumes 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 10COR Technical Report 15.18 " Analysis of In-Vessel Core Melt Progression," Vol. IV (User's Manual) and Modeling Details for the Fission Product Release and Transport Code (FPRAT), September,1983.

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

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.

L7 4-1 D

4.0 PLANT RESPONSE TO SEVERE ACCIDENTS Four base accident sequences were analyzed for Peach 30ttom using MAAP to determine plant response and temperature and pressure challenges to containment. These sequences, described below, in general are based on the sequences identified in Subtask 3.2.

Transient initiated sequences requiring a reactor shutdown 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.

The base sequances are:

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

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

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

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

The sequences analyzed in this section are low probability core damage events and include no, or minimal, recognition of operator actions that would significantly delay the progression toward core melt or mitigate conse-quences. This approach was taken to produce results which bound or are at the high end of the range of possible consequences for the four selected se-quences. Generally, only minimal operator actions to control selected plant systcms are assumed for these events. For example, it is assumed that the

~

DRMT operators regulate low pressure injection to maintain water level at the high level trip.

Consequently, the results presented here do not represent what would be expected to occur for the defined equipment fa11eres and are extrcmely 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 Procedure Guidelines. As a result of these actions the operator is able to terminate the event prior to core melt or significantly mitigate its consequences. Section 5 considers only some examples of the many actions available to the operator to prevent or mitigate the accident.

A major objective of excluding mitigating operator actions in this 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 ultimate pressure capacity is based on the evaluation for the Browns Ferry Mark 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 j calculated to be 132 psia with the defined failure condition (twice the l 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, I

. Q 4-3 I this value is assumed to represent Peach Bottom. In cases where high tempera-tures in contaiment were reached before ultimate pressure, the containment was assumed to fail when the containment atmosphere temperature reached 1200*F. The Peach Bottom contalment is made from high strength carbon steel.

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.

A containment break size of 0.1 f t2 (0.2 ft2 for TC) is assumed because it pemits 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 isolating the reactor from the power conversion system. High pressure injection (HFCI and RCIC) are initially available until high suppression pool temperatures cause loss of these systems. Low pressure injection (LPCI and LPCS) are available as long as NPSH requirements are met *. Control rod drive (CRD) flow remains on until the available inventory of water in the condensate storage i 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 sumarized in Table 4.1. Plots of key parameters are presented in Figs. 4.1 through 4.5.

In MAAP, the available NPSH was calculated but the requirement was set tn zero. The resulting time dependent behavior was then reviewed to determine if cavitation had occurred and if it would have been sufficient to fail the pumps.

9

~

DRAFT Table 4.1

. PEACH BOTTOM - TW EVENT

SUMMARY

Time . Event 0 Transient (MSIV closure) 4 sec Reactor scramed 4.5 min HPCI, RCIC on 8.0 hr High SP temperature failure assumed for HPCI (200'F) 10 hr RCIC lost 14 hr CR0 flow ceases 15 hr ADS on LPCI and LPCS injecting 25 hr ADS valves close 32 hr Containment failure (overpressurization); LPCI and LPCS lost 34 hr Top of core uncovered 37 hr ADS valves open 39 hr Start of core melt 40 hr Vessel failure l

TW -- PEACH BOTTOM i4a . _. . . - . . - - . . . - . . . . . . - . . -

.._.7 .-....

7 .

4 ;

~

2 . e .

128- --._--: . - - - - - - - + - - . - - r.--.-.- .-.-- --- + +

2 e

198- --,.-.-+--+------+-.- F-- i - ---- V- - :--- - F -- --- i

_ t

<t . . ; , ;

- 86-

.--. i- t - -- r -+ +-------+ t a m t C.

~ .

-- gg_ 4. .

+ . .. ._ 1 j , . - . -

...+- .

- . ... - i .

d T, a

Q. .

\ $

4g_ . . _.. .

i-.. j -. . . ._ .. .q

N. . . .

. / ! j ! !

e .

pg_ f , ,

- .. f . ;- .- . . . . . si

/

~

e g g y 5 y I E E 5 5 5 3 g5 5 5 5 g5 5 5 5 g5 5 5 5 g5 5 5 5 g4 5 5 5 g5 5 5 5 g5 5 5 5 g5 4 5 5 g5 5 5 5 g 0 10 20 30 40 50 60 70 90 90 100 110 120 TIME (HOURSI n g. 4.i pressure w e arw eii. )l$3

---l

~JO TW -- PEACH BOTTOM m

1.800- . . . . . _ . . _ . . _ _ .

W

I 1.600-

_ ..-. . .$.. ... ; . ... . . . . p. . .,.

v.

rf _

s

' .?

i,400- - . _ . . .I - !.. } _ .__ _ _. . ._ . . _ . , . . _. _, ..__...._,.,_,,_._,,,,(,

- /

j .

_ 1.200-g o

w 1, O @

D - i ._ ..j_. . ; . ...g.... . ; .- ..,.,p_,._,,,,,,,,_,,,

C2 -

f. . I ! .

600- . ,. -

. . ; . - . .. ,i .

7 o

/ a 6- c a r.

==wq

/- - -. ; _. . ..

a f ..

t i !

400- _ .

~

/ t" _.g -

. .. r u

l '

t....i....i....i ... ....i.... .... .........,....,....,....,

O 10 20 30 46 50 60 70 go 90 100 110 120 l TITE (HOURS)

I Fig. 4.2 Temperature of gas in the drywell.

TW -- PEACH BOTTOM 35 8 . -. .-- --

- r- - - . . - . - -

- \s . .

v4

. 3 333 ~ . . - . .

, - - .h --- --------.I-...L_-... ~ - - - - - .

i .

2502 --;- , -- . - ---- e - -- - t- h.

I -

- -- - i --- - -t --- -- t --- --i E

o 260- -- ----/ /':-

t t s t :

w -

--+--~--+- - - - - - - - + - -

+ +-

2 t

i c.=n l

,/ ,

t j

a.136-- - -i -- L - - - - - - - -

-=

[ .; ,

i ,

i ;

- . e

+

y 180--l? --

t t-

-- - 1

- 3 2

q i Sg . . . - ; - . . ._

- . . .. ; . .. _ . . .q

! i o

0 10 26 30 40 50 66 i

70 80 j

90 100 110 124 i

Q-TIME (HOURSI Fig. 4.3 Temperature of the suppression pool.

1

)

H

23 TW -- PEACH BOTT0r1 m

l 5 ,--

7..._._..__

_...._r__._,

-l'.

J i i . .

,T h- .. ,i-.......4......._...* .. .. . y _. ,.... j.

. , . , . . . , , ., ,_,,;,,_,,,_.,,_,_g N

~

! ) i i

g

! i i ,,. -- / . .

3_ .

.}. ..... g.'. _ ._,.....j....._._.._.;

w -

i ./

. .i

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i a . /

9'2- . . . . - q ! . . ;. - ..- . ; . . ..; _ .;

y g

- CD

{

. ._ - . /. _ - - . . , . . . . . . . . . _ . . . . , . . . . .

14 /

4

,/

._/ ; , ..'

g ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,. ,,,,,,,,,,,,,,,,,,,,

0 10 20 30 40 50 60 70 98 90 100 110 120 tit 1E tHOURSI Fig. 4.4 Concrete ablation depth in the pedestal.

TW -- PE ACH BOTT0t1 l

5, 8 0 8 , - --- -.- . .-.----- .___ . - -- . 7- . _ - - - . . . . . . - - - - - - - - . . . . - - - . - - - - . _ . . . - -

1 i i i i ! !

4,gg8_ . - . . + . . . -...;... . . . - . . ...,

4 ... g . . .. . - . p.. . --_-. j . . - . . ; . - . _.-. ; . - _ ._ _ ;

! ,/

- a-u.

,3,gQg_ - . . . j. j-..._.,. f_j -. . .-. . 3 . - ---j . . . _h i -p j d

i

. i i ,

c3 f g 2.800- -

1 -

i

- 4- - - - - - - - ; - - -- ; -- - i .

N _

E

18. .

~

i 1,000- -- - - -- h- - - ---:---- -r -

- ---- --t-- -- - i' i i

}

L. 8 a u au lsa s a gsy e e gs s e e gu u E E l5 s v s lu aw sgae 5 sgI 5 s ega s u a gu aa sga s ay g

0 10 20 30 40 50 60 70 90 90 100 110 128 TITE (HOURSI .

Fig. 4.5 Average coriun temperature in the pedestal.

""=4 1

BRMI 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 after 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.

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 l control rod drive (CRD) 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . RCIC suction remains from the CST until a low CST level signal is received at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months when it is automatically transferred to the suppression pool. At 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . 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 0.00389 hours 2.314815e-5 weeks 5.327e-6 months 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 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months . This permits low pres-sure systems (LPCI and LPCS) to maintain reactor water inventory.

O i.

4-11 D

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 CRD 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, this volume is not restricted from CRD pump use. An average CST inventcry 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 lifts 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.

Af ter the vessel is repressurized and the SRVs open and pemit 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 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months . Containment failure is assumed to fail the low pressure injection systems. This could potentially result from mechanical failures (piping movement and rupture) induced by containment failure, from electrical failures due to a steam envirorsnent in the reactor building, or possibly due to insufficient NPSH and large scale cavitation in the low pressure pumps. Knowledge of the actual failure mechanism is not

~.

required for this anslysis but may be necessary for the assessment of core

) damage probabilities and public health risk.

, As the containment depressurizes, the ADS-SRVs 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 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

With the assumption of no further injection after loss of low pressure injection, the gradual boil down of reactor water inventory results in the top of the core being uncovered at about 34 hours3.935185e-4 days 0.00944 hours 5.621693e-5 weeks 1.2937e-5 months . 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 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months . At approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months when 20% of the core inventory is molten and has collected on the core plate, it is assumed to fail. This is equivalent to 153 fuel assemblies (764 assemblies in the core) achieving a molten state. Since the assemblies are supported from below by the CRD 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 UO and 2

the associated Zircaloy is assumed to fail this structure 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.

The reactor pressure vessel fails within a few minutes af ter the core plate fails due to rapid melting of the instrument and CRD tubes that penetrate the bottom vessel head. Vessel failure and the subsequent genera-tion of : team 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 0.0172 hours 1.025132e-4 weeks 2.3591e-5 months .

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 environment resulting in oxidation. During the period from 36 to 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months into the sequence, approximately 430 lbm of hydrogen are generated.

i

BRAFT 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 contairsnent integrity. A number of alternative means of arrest- l ing 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. Appendix B includes several plots showing results for this sequence. ,

l 4.2 Plant Response to the TC Sequence (Without Operator Action to Reduce Power Level) 4.2.1 Sequence 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 ano LPCS) injection systems are available until either high suppression pool temperatures or insufficient NPSH cause loss of these systems *. CRD flow remains on until the inventory in the CST is depleted. The condensate pump is 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 overpressurization. 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 af ter wetwell venting is initiated. The low pressure ECCS pumps have been tested in a steam environment and have performed satisfactorily. Therefore, it should be noted this is a conserva-tive assumption regarding pump performance and that the core would not be damaged if the pumps continued to operate.

NPSH was treated in a similar manner as discussed for the TW sequence.

DMFT 4.2.2 Primary System and Containment Response The timing of the key events for this sequence is sumarized 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 loss of 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 liquid 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 normal. 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 confimed to be in the correct range with a neutronics model, RETRAN, and assumes no operator action is taken to throttle injection to reduce power level.

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

An approximately 26 minutes, the suppression pool reaches 200*F resulting in the assumed degradation of the HPCI pump bearings causing loss of this system.

RCIC injection is insufficient to maintain water level while the core is at 18% 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 decreases from a level 20 ft above the core inlet down to the top of the core which is reached at 37 minutes. Af ter this time the power level continues to be a function of I

I

4-15 .. l Table 4.2 T PEACH BOTTOM - TC EVENT

SUMMARY

Time Event 0 Transient (MSIVclosure) 3 sec Failure to scram 1.7 min HPCI, RCIC on 26.4 min HPCI assumed lost (SP at 200*F) 38 min ADS on 40 min LPCI, LPCS on (reduced flow) 54 min RCIC lost 1.3 hr ADS valves close 1.3 hr Top of core uncovered 1.3 hr Open wetwell vent 1.3 hr LPCI, LPCS assumed lost 1.5 hr ADS valves reopen 3 hr Start of core melting 3.9 hr Vessel failure 6.9 hr CRD flow ceases 1

l

TC -- PEACH BOTTOM :::X3 50 -.

m M

l l 4em .g .. . . . . _ . . . . . _ . ..,

t l

v7 >

. _. N g 3e_ . . . ; .. . . . .

.._. .. N { . , ,. . . . _ . . . . . . . _ . , _ _ ,_ .,

~

N 1 .N '

a a

g 20- i i i. j.- . - .. ;- -- .j x _ . .

.L

- t 10 4 4 .

i 1 .

~

O ....,....,....,.... .... ....

0. 00 0.50 1. 60 1.50 2.00 2.50 3. 00 3.50 4.00 4.50 5. 00 TITE (HOURS)

Fig. 4.6 Reactor pressure vessel water level.

4

TC.W/WW VENT --

PEACH BOTTOt1

. 4 a a iiia a g ia e a a a s .. e i a a a . s 6 a i g 4. i 'Iii a 3,. .#ssie O ' g.4I s . i' 4 6 g e ii a i .a e a gi e i ei e i e a g i a a i a iIs giaiia is i "x .

~.

e h L O : -

e s

3& :

I -

g -

to o _ -

us -

8.9O ;

m O .

! N g ~

0 j

Q _ -

O -

o. o.no i 1.5 2 2.5 3 3.5 4 45 5 T INE liR Fig. 4.7 Average core power.

l

_s 5

i l

g' 3

ae$

2 o*

u l

4.:.:_:: .;- .[

5 : 7. t 0

. 1 o

x u

q 9

. 8 H

O .

T T .

_ O , 7 l

l B e w

y H .

r C d A 6 e E ) h

- P G t

_ - R n

- U i 5 O e r

T H u C

N s s

E E e V r 4 M I P

H T W'

8 4

W .

3 .

g C i T F

- 2 n

I o

o i% .

l :- ::L ; r -; ,' p} , :. . >- O

, n ~ 1. . - - ,X d .

m < 1.L

2c0- -

TC .W/lM vel 4T -- PEACH f30TTOM

, In

.... ..g... . ... i. . .i2 3 ** i.e ai"as"lii"i'i x : .

/ %,'- k n

e /,,-r + # ^' : .

a sn : #

' 4 ,- / .- -

J l-O -

1 :

E-V F [

. /

/

1

. si .! 4

' I o !.%/ :

o

c 1 2 3 4 5 a 7 8 9 10 TIME C HOtJRS ) x10 '

\

Fig. 4.9 Temperature of gas in the drywell. 3lma l

1 i

1

TC.W'WW VENT -- PEACH BOTTDH

D o ' ' ' " '

'"I'"'"""'"""'4""""i M l

v L q m :}.; 4

.i .- I"1 g x

==

.i f, ,

'q_ _

5 k i ?

m o

t ::

g..

T

o. I 2 3 4 5 6 7 8 9 to TIME ( flouiiS ) x10*

Fig. 4.10 Temperature of the suppression pool.

TC.W'WW VENT -- PEACH UOTTOM o- . ,- ... ,.......,,;. ...,,,,,,,,,,,, ,

i E

~

elar 4

s -

/

7 -

~ -

f w

O 3 :_ .

T 2 s- .

A.

v t .

m x ; -

l 1

- i

/ -

. - m

l -

- .) ..

j 0 1 2 3 4 5 6 7 8 9 go T IME (1100RS ) x 10 ,

i l Fig. 4.11 Concrete ablation depth in the pedestal.

l

l l

l IC .MWW VE NT -- PF ACil GOT fDM a g:

c i

- es m :,

1 .-. L :

C ;

w _

w ,

5NE i 1 E

v h-I' F- ,. : y

~

-[- g i a

6 a sH :

O. 1 2 3 4 5 6 7 8 9 to 1IME ,HOURG1 x 10 '

Fig. 4.12 Average corium temperature in the pedestal.

si1 AFT 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 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , 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 ft2 ) at a pressure of 115 psia which is reached at 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 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 depres-surizes, the ADS valves reopen when the pressure decreases to 90 psia (1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ), 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 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . When 20% of the core has melted core plate failure is assumed resulting in vessel failure at 3.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reaching a temperature of 1500*F at approximately 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . This occurs as a result of radiative and convective heat transfer from the core debris which lies on the pedastal and drywell ficors in a relatively thin flat geometry with lerge surface area. Appendix B includes other plots for this sequence.

The influence of natural circulation within the primary system is illustrated by Figs. B.15 and B.16 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 is transported throughout the primary system and eventually into the contain-ment. It is this release into the containment that is eventually transported 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 af ter 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 22 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 CR0 flow.

4.3 Plant Response to the S)E Sequence 4.3.1 Sequence Description This sequence is assumed to be initiated by a 0.1 f 2t 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 condensate pump is also assumed to be unavail-4 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 i

l

m%"

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 the key events for this sequence is sumarized 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.

This sequence is initiated by a 0.1 ft 2break 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 rcactor 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-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 into the core. Post-scram CRD flow (177 gpm) is sufficient to keep the core ,

l covered until automatic depressurization (ADS) is actuated at 1.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months due to l high drywell pressure and low reactor water level, further depressurizing the l primary system. There is a low pressure pump permissive signal for ADS l becaust. 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . As reactor water level boils down, the fuel overheats and oxidation of cladding l occurs. The core is heated up leading to fuel melting at 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and vessel l failure at 3.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Approximately 240 lbm of hydre en are generated between 1.5 and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4-26

.. v r,. P S

{ Table 4.3 PEACH BOTTOM - Sj E EVENT

SUMMARY

Time Event 2

0 Break in steam lina (0.1 ft )

6.8 sec Reat.: tr scrammed 84 sec MSIVs closed, feedwater tripped 10 min Suppression pool cooling on 1.05 hr Automatic depressurization on (ADS) 1.13 hr Top of core uncovered 2.5 hr Start of core melt 3.4 hr Vessel failure 15 hr CRD flow ceases 23 hr Containment failure (overtemperature) l I

l

S1E -- PEACH BOTTOM 6 8 ..- - - . - . -- -. . . . . . . . . . _ .._ _- - - . . . . - - - - - . -_ - .

4 i

. i ! i !

! i i i i i i

$$ .. .. f . }.. .. i ._ j ._ .. . .... 4 . . ;.. . . . . .f . . i

-i i -; i i

- / ! > !

,8 } ! !

4g_ _ ..._i._-. . - - - . - - . ..

f .. . ..;- - . - 4

+

a 6/b 38- .- 4 -- - - - - --i- -------------i 4- i

. i. i O i- i a D. -

e 2g_ s . ;, . - . . . . . . . _ _ . _ . y t

<dN%, ,

1 I

fh. . .. .- .. - . . . . . . , . . .........[ . . . . .

e- ....,....,.... .... .... .... .... ........................,

0 5 10 15 28 25 30 35 48 45 58 55 68 TITE (HOURSI Fig. 4.13 Pressure in the drywell.

- + ,

N f

S1E -- PEACH BOTTOM E

l 2, S J e . .. _ . . . _ _ - _ . , . . .

"'W"I '

t

1. 500-- -; - - - . --

-;p.!

- j---- . - - . --. j j-r f .

u. /~

o na -

i i i j hhh. ... .- ,.....5,--...... . . . - . - ,

h- -- .= h , ,

i* e i ,

d -

i, o a a.

Y

~

N co 3gg .. . .., .;. ..,. . - p .- , -.~ .p..

, 9,.....

L.. J l 0

.... ....i.... .... ....i....i.... .... .... ....i....i...rj 0 5 10 15 20 25 30 35 40 45 50 55 60 TILE (HOURSI Fig. 4.14 Temperature of gas in the drywell.

SIE -- PEACH CCTTOM 11e . .. .. _ . , - . . . . . _ _ . . .,

g

\ , ;

s .

4 128- ..

4 . ,-. . . . . . . . . . . . . . . .. . 3 . __... . . ._ v. .. . . .j

, _ j i. .i .

\i . i. i.

l . . . .

t98- L. --- . - . ; . .-- - .- I ------s. - 1-.-.- - . - - . - - .

t. .--4.-----.-...- ;

g -

o 86-w

-! 4 . -4 -+- - - - - - - - . .

E i

4 4 A

5 60_ .. . .. .. . . _ . . . . . . . .. . :

3 -

A LO

+ r

.* ..t. .....f.

i i i i' i 20- . i- -.

j. . - ,

0 ....,....,....,....,....,....,....,....,....,....,....,....,

0 5 10 15 20 25 30 35 40 45 50 55 68 TITE tHOURSI Fig. 4.15 Temperature of the suppression pool.

-a a

k S1E -- PEACH BOTT0f1 33" 3.90, ,

t.

i 2.5B- - - - - + - - - - + - - - .. -.---6 '

. -+--..--.--.--.---.J-; ;- - 2

~

/ i ;

/ :

!/ !

) i 2.00l - - - - - - . - - - - - - . - .- . .- . - - - . . . . - ,l.. . -- - .. - . -- .r -- . -. j - i s -

/ .

hJ -

/ -

W . 6 O -

.j 1.50- --r-- +--.--.--..- +--.----.e----- 4 -i n

4 a : "

/

l 9- - '

i N.03{

1 - -

h i _ <

,/ . .

j ,

0.50- - - 4 b---

/ --

j -

~

i

/

rnl.i....i....i....i....i....

n u . d. .. u .

0 5 10 15 20 25 30 35 40 45 50 55 60 TITE tHOURSI Fig. 4.16 Concrete ablation depth in the pedestal.

S1E -- PEACH BOTTOM 6, C$ $m ..y......p.. ..

. . . . , . . . . _ . , _ .. . , . . . . . . . . . _.l, . _ ,

i 5.000- --

-4 4- --- -

i l - - - -+--i + 4

~

i i

4.898- +- - - + - - - - + r- 4 +

( 4 t- - -

su !

1 ; ! :

~

3.800- - ----- - - -

v --- 4 L - - --- -- - -- - --! i- ! 4 O 1 1 . .

i

2. 80B g q 1 fI i- I !----- --

---! S ,

~

t t *

t J

t l

t ! 8 1.000- L- -- - i -

4 - -- i - + --}- L i . .

~

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

0 5 10 15 20 25 30 35 48 45 50 55 6B TIME (HOURSI

-__m Fig. 4.17 Average corium temperature in the pedestal. " 'l anoman-e

l 4-32 !

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-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 comences, after the loss of CRD flow, at about 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months .

When the temperature reaches 1200*F at 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months the containment is assumed to fail with o 0.1 f t2 break. The containment pressure is approxi-mately 55 psia when the failure occurs, dropping to atmospheric pressure at about 28 h:urs. Appendix B includes several plots for this sequence.

The analysis of this sequence demonstrates that operators have approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to take action to preserve containment integrity and mitigate releases.

4.4 Plant Resoonse to the TQVW Sequence 4.4.1 Sequence Description Inis 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 loss of the power conversion system. It is assumed that DC power is available for a period of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , no further injecticn is assumed available and no operator action to utilize alternate sources are assumed to occur.

4.4.2 Primary System and Containment Response The kay events for this sequence are sumarized in Table 4.4. Plots of key parameters are presented in Figs. 4.18 through 4.22. In general, this

4-33 Table 4.4 PEACH BOTTOM - TQVW

, A T. l 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 i 11.4 hr Start of core melt l

i 12.4 hr Vessel failure 18 hr Containnent failure (overtemperature)

x:3 TOVW -- PEACH BOTTOM p 12 8 m . . _ , . . .. .- .. . , . . . ,

.7 , _ .... .. _, . . . . . .

7...

m f *

}Q$- ,

. , . . . . . .. . . . . .......4 .;. .. . .. ..i .. )

i /- ~

e z

{

i i

{t @ _ ... 7

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

d ! i a i i.

eg 1 -

9 eh 6$ , . . . .

....9., . - . . . . . .

7 . _ . - . .. ... j - -, _ q y - . . d. . - .

~

I A

a. 4 i

4h_ . . . :. .. . . . . . . . . . . . . . . . . . .

k 5

~

i i r

'- %u,,, - -

./..

~

.Ws %sh e,sm%)

, 2 t) . . . .

9. .

j. .. .

p- ;

E

-h b ai sa ge s ,s gie i e gsa ee gie ae ga : as gia e ig,e s gie a i gs a e e giei a ga: aa g 6 5 10 15 20 25 38 35 40 45 50 55 68 TIME (HOURSI Fig. 4.18 Pressure in the drywell.

TQVW -- PEACH BOTTOM

.i 2,800 -.--. . ~ . _ . . . . . .

i i [H h hhk i

e igY ,

, h- ._-...j-..-...,..

... . j _. .. ..j . .._..-9...... .......)_..g

u.

o ! ,

i sa -

i ca -

- 1,eee_ ...1._._.2...._1._.-.e...... .;

j v .

4 d -

! a i R -

! O

, i !

580- ,

/. .. t i- +- - , . i - --- . 3 - i j 1

LJ J .

+

r l

'1, -- ! !

0

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

0 5 10 15 20 25 30 35 48 45 58 55 60 TIME (HOURSI Fig. 4.19 Temperature of gas in the drywell.

--=4

TOVW -- PEACH BOTTOM i 2 5 0 .. .

3mm

i ! !

i .

l 268_ ..j . . ._ j .._j..._...j...._. . j . .. 3 ..j.. .... . _ .. ;..._ .. _ 4 . _p ..___.j i i ! ! !

/ i

~

c>

150- l-- ! .- - -- 4. . -+ + - --i - -,

F.--- -i- - 4 ~ - -i..

+

l, kJ f

i i i :: _~i i e 200_p f 5 .

_.i i .. -

9 .. .p . . .l. ,

i 3 -

i i '

i' ! ! S 50 l '

i 4 . .

i t

O

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

O 5 to 15 20 25 30 35 40 45 50 55 60 TINE (HOURSI Fig. 4.20 Temperature of the suppression pool.

1 e

TOVW -- PEACH BOTTOM 1 5, . . . . - . _ . _ . . . _ , . , . . .

4_ .....__....i_.. .... _i. .i. . .; . /......_.....__.....__.

/ .

' 1 p _- !

w 3_ _._.4.. 4 4 i.

... ,..._..,...._4 i ;

~

D $#

; /l !

i k2 -

. j. . . . . .

. . p . .... .j

{/

y .

u i

i n J . . . _ , . . . _ . _ . - . _ . _ _ , . _ . _ . .

i

1) . .l. . .

/

l . s' f

~

f

. /

d 0

0 5 10 15 20 25 30 35 48 45 58 55 68 TITE (HOURSI Fig. 4.21 Concrete ablation depth in the pedestal.

W

TOVW -- PEACH BOTTOM %

6. e e e, . - . - - - . - . . . . - - - - . - - . . - - . -- --. - . - - - - . - - - . . - - - .

5.800- --l- ----.-]- --l ---- -l- -. -- - L ---- l - -- - l i i

- +

-+---4 4 u.

4.800- -- - + i- - - - - - - - --

c- -

+- 4 o -

s 3,gg@_ . . . .. . 3 ...-- .. j.- ..;--....... ...i..-...- -..--.l--.- . 4 A i .

$ i i N 2. 00 0 - -

+-- +- - : - - --- i M i i 1 i i 1, 000 & l -- -

--- - - -- i -- -- ' - -- - - ]

h 0 ,,,,,,,,,,,,,,,,,,,y,,,,,,,,,,,,,,,,,,,,,,,,,,,,,g,,,,,,,,,,

0 5 10 15 20 25 30 35 40 45 50 55 68 TINE (HOURSI Fig. 4.22 Average corium temperature in the pedestal.

DRAFT 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months later.

Loss of all AC power results in an immediate reactor scram signal followed by successful reactor scram within 4 secor.ds. The power conversion system is lost 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 DC power required for control of these systems is lost. This is assumed to occur at 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months into the accideot. It is assumed that heatup of the rooms containing the HPCI and RCIC systems does not cause system failure because of the reasons discussed below. After HPCI and RCIC are lost, the reactor water level boils down uncovering the top of the core at about 8.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During the boil down, approximately 500 lbm of hydrogen are generated due to cladding oxidation.

The core begins to melt at approximately 11.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and the melting pro-gresses until core plate and vessel failure occur at approximately 12.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 actua-tion does not occur due to the lack of a low pressure pump permissive signal.

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 noncondensable gases from initial concrete ablation and addi-tional cladding oxidation. After the residual water from the vessel is vaporized there is no water available to quench the ccre debris. Thus con-crete 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 comencing af.Jr 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . The assumed failure temperature of 1200'F 1s reached at approximately 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months when the containment pressure is

DRAFT at 105 psia. Following the failure, the containment pre aure decreases to about 25 psia at 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months . 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months assumed in this analysis.

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 effect s" stem operation. In additien, 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 were lost, additional means of vessel makeup are available which do not raly on plant AC or DC 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 conservative analysis of this sequence described in this section indicates that the operators have over 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to take action to maintain containment integrity to mitigate releases.

I

~.

4-41 .

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

4.2 IDCOR Technical Repcrt 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 0511 and ASTM DS11 Supplement 1.

4.4 L. Chu, " Power Suppression and Boron Remixing Mechanism for General Electric Boiling Water Reactor Emergency Procedures Guidelines,"

NEDC 22166 August 1983.

- 4 E

5-1 DAh

/

5.0 PLANT RESPONSE WITH RECOVERY ACTIONS

-- TO BE SUPPLIED LATER --

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

MAfT 6.0 FISSION PRODUCT RELEASE, TRANSPORT AND DEPOSITION 6.1 Introduction l

The phenomena of fission product release from the fuel matrix, its transport within the primary system, their release from the primary system into the containment, their deposition within the containment and the subse-quent release of some fission products frcm the containment are treated through the use of MAAP [6.1]. Release of fission products from the fuel matrix and their transport to the top of the core are treated by a subroutine in MAAP which is based on the FPRAT code [6.2]. Transport of fission products outside the core boundaries is determined by the natural and forced convection flows modeled in MAAP with the gravitational sedimentation described in Ref.

[6.3] and other deposition processes described in Ref. [6.4]. Fission product behavior is considered for the best estimate transport, deposition 'and reloca-tion processes. Influence of surface reactions between chemically active substances like cesium hydroxide and other uncertainties are considered in subtask 23.4. The best estimate calculation, assuming cesium iodide and cesium hydroxide are the chemical state of cesium and iodine, is discussed below.

6.2 Modeling Approach Evaluations of the dominant chemical species in Ref. [6.5] show the states of the radionuclides (excluding noble gases) which dominate the public health risk to be cesium iodide and cesium hydroxide, tellurium oxide and strontium oxide. These and others are considered in the code when calculating the release of fission products from the fuel matrix. Vapors of these domi-nant species form dense aerosol clouds in the upper plenum, in some cases 3

approaching 100 g/m for a very short time, which agglomerate and settle onto l surfaces. Depending upon the chemical compound and gas temperature, these I deposited aerosols can be either solid or liquid. At the time of reactor vessel failure, some material remains suspended as airborne aerosol or vapor l

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

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

As a result of the dense aerosols formed when fission products are released from the fuel, considerable deposition occurs within the primary system prior to vessel failure. For some accident sequences, the primary system may be at an elevated pressure at the time of core slump and reactor vessel failure. Resuspension of these ae osol deposits during the primary system blowdown is assessed in Ref. [6.6) in terms of the available experi-niental results and basic models. It is concluded that resuspension imediate-ly following reactor vessel failure would not be significant, less than 1% of I

the deposited materials, even for depressurizations initiated from the nominal operating pressure. For delayed containmr.nt failure, this small fraction of

, material is depleted by in-containaient mechanisms.

Therefore, a major fraction of the volatile fission products are retained within the primary system following vessel failure, the distribution t being determined by the MAAP calculations prior to vessel failure. Natural circulation through the primary system after vessel failure is analyzed using MAAP which allows for heat and mass transport in various nodes of the reactor vessel and the steam generators including heat losses from the primary system as dictated by the reflective insulation. Material transport is due to aerosols and vapors as governed by the heatup of structures due to radioactive decay of deposited fission products. This heatup is principally determined by the transport of cesium iodide and cesium hydroxide by the natural circulation flou. In this regard, the vapor pressure of cesium hydroxide is applied to both the cesium iodide and cesium hydroxide chemical species. In essence, this assumes that the solution of cesium iodide and cesium hydroxide has a vapor pressure close to that of cesium hydroxiae, which is a conservatism in the calculations. In carrying out these calculations, the pressurization of the primary system is dependent upon the pressurization of the containment and the heating within the primary system. These determine the in- and out-flows between the primary system and containment.

MMT Deposition within the containment is calculated using thennal The major aerosol sources are the ,

hydraulic conditions detennined by MAAP.

releases prior to vessel failure (sequence specific), the airborne aerosols and vapors transferred from the primary system at the time of vessel failure, the subsequent releases from the primary system due to long term heatup, and concrete attack. At the time of containment failure, the remaining airborne aerosol and vapor can be released to the environment. Assessments of the potential for resuspension of deposited aerosols following containment failure

[6.6] show this to be negligible.

6.3 Sequences Evaluated 6.3.1 TW Fission Product Release As previously described the TW sequence is very extended. Essen-tially a day is available to take corrective action to prevent containment failure. Also, failure of the ECCS systems following containment failure is by definition, not the result of a mechanistic process. With the assumption of no corrective action and loss of all injection results in containment failure prior to melt-through of the reactor pressure vessel. Table 6.1 shows the Csl distribution at vessel failure. Due to the relatively low ficws to the suppression pool and the large settling area in the upper plenum 93% of the Csl 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 l

temperatures to begin transporting the fission products out into the contain-ment. At this time (* 60 hrs) the fission products begin to be discharged into the reactor building. The ultimate Csl distribution at 120 hrs into'the I accident is shown in Table 6.1 and the fraction release of all fission pro-ducts at this time is shown in Table 6.2.

6-4 Table 6.1 DISTRIBUTION OF Cs! IN PLANT AND ENVIRONMENT (FRACTION OF CORE INVENTORY)

At Vessel Failure TW TC SEj TQVW RPV .93 0.81 .47 .997 Drywell 0 0 .30 0 Suppression Pool .07 0.19 .23 .003 Secondary Containment 3 x 10-5 4 x 10-4 0 0 Environment 2 x 10-5 0 0 0 At Containment Failure TW TC S;E TQVW RPV 1.0 0.45 0.42 .80 Drywell 0 0.04 .03 .17 Suppression Pool 0 0.51 .55 .03 Secondary Containment 0 5 x 10'4 0 0 Environment 0 0 0 0 Ultimate Distribution TW TC SE j TOVW F,PV .01 0.01 0 0 Drywell 0 0 0 0 Suppression Pool .07 0.51 .57 .034 Secondary Containment .79 0.45 .42 .92 Environment .13 0.03 .01 .05 l

6-5 g Table 6.2 E TW FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Drywell, El 165' 2

Containment Failure Size .1 ft Fission Product Release Fraction Group to Environment Cs, I 0.13 Te, Sb 0.13 Sr. Ba 9E-5 Ru, Mo 4E-4 Time of Release: 34 hr.

Duration of Release: 80 hr.

~

DRAFT 6.3.2 TC Fission Prnduct Release The MAAP analysis of this sequence shows an initial deposition of volatile fission products in the reactor vessel and a subsequent redistribu-tion among the different vessel regions after vessel failure as these fission products revaporize due to decay heat.

As indicated in Table 6.1, prior to vessel failure, most of the inventory of volatile fission products 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 CRD water, which flows into the pedestal and cools the debris by vaporization.

After the CST is depleted, the drywell temperature increases to levels which could threaten the integrity of electrical penetrations 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 transferred to the reactor vessel which ultimately comes into thermal equilibrium with the drywell. As the vessel heats up, all of the volatile fission products retained in the vessel are revaporized and are convected out of the vessei 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. 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 signifi-cant amount of concrete aerosols are being released due to core-concrete i attack in the pedestal region. The volatile fission products condense and i

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. Consequently a relatively small frac-l tion of the volatile fission products are released to the environment as indicated in Table 6.3.

~

I Table 6.3 DRAFT TC FISSION PRODUCT RELEASE Assumptions Containment Failure Location - Drywell, El 165' 2

Containment Failure Size .2 ft l

Fission Product Release Fraction Group to Environment Cs, 1 0.034 Te, Sb 0.066 Sr. Ba SE-5 Ru, Mo 2E-4 Time of Release: 13 hr.

Duration of Release: 50 hr.

6.3.3 S.jE Fission Product Release This sequence was analyzed to detemine the time dependent distribu-tion of volatile fission products within the vessel, the rate of release from the vessel to the drywell and, after 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 tenn 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 after containment failure. Revaporization in the reactor vessel is also occurring, 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 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after containment failure.

This material is passed through the drywell to the reactor building. As indicated in Table 6.1, none of the volatile fission products are ultimately retained in the reactor vessel or drywell. Aerosols initially released from the containment would be sucked ir.to the SGTS system and retained in the fil ter. Af ter about 80 kg were were accumulated in the filters, the reduction in flow area would cause the SGTS fans to trip on low flow. With the SGTS system shutdown, the building through flow is determined by the flow from the primary containment. This results in a long residence time in the reactor building.

Gravitational settling of fission product aerosols in the reactor building results in substantial retention in the building. The volatile fission products released to the environment are given in Table 6.4.

6.3.4 TQVW Fission Product Release Since SGTS is unavailable, the path te the environment is through the reactor building with direct leakage to tne atmosphere. The reactor building flow races are governed solely by the containment break flow and the ensuing thermal hydraulic conditions in the reactor building. Therefore,

Table 6.4 BRAFT S;E FISSION PRODUCT RELEASE Assumptions Containment Failure location - Drywell, El 135' Containment Failure Size - 0.1 ft 2 Fission Product Release Fraction Group to Environment Cs I 0.01 Te, Sb 0.01 Sr. Ba 2E-5 Ru, Mo 6E-5 Time of Release: 23 hr.

Durat::n of Release: 30 hr.

i o

DRAFT there is no forced convection or fission product removal resulting from SGTS operation.

7 b As in the other sequences, drywell heatup contributes to the reactor

[ vessel heatup. As indicated in Table 6.1 most of the volatile fission prod-5 ucts are in the reactor vessel and the suppression pool at the time of con-1 tainment failure. The inventories of cesium, iodine and tellurium are some-L what less than those calculated for other sequences since the primary system 6 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. Conse-h g quently, most of these fission products are out of the vessel by about 25

? hours after containment failure. These volatile fission products pass through g the drywell as vapors and into the comparatively cool reactor building where

- they condense to form aerosols and result in substantial gravitational set-tling. The effectiveness of this removal mechanism is enhanced by the low I temperature and the long residence times in the reactor building because of

) the absence of forced convection from SGTS operation. Consequently, only 0.05% of the Cs and I is released to the environment as indicated in Table

}

6.5.

6

$ 6.4 References s

b 6.1 MAAP - Modular Accident Analysis Program, User's Manual, August, p 1983.

6.2 IDCOR Technical Report 15.lB, " Analysis of In-Vessel Core Melt i_ Progression," Vol. IV (User's Manual) and Modeling Details for the g Fission Product Release and Transport Code (FPRAT), September,1983.

6.3 Draft 10COR Technical Report, "FAI Aerosol Correlation," July,1984.

L 6.4 IDCOR 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 f Fission Product and Core-Material Source Characteristics," October, g 1982.

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

i

i 6-11 -

Table 6.5 g g TQVW FISSION PRODUCT RELEASE Assumption',

Containment Failure Location - Drywell, El 135' Containment Failure Size - 0.1 ft 2 9

Fission Product Release Fraction Group to Environment Cs. I 0.05 Te, Sb 0.04 Sr Ba SE-5 Ru, Mo 2E-4 Time of Release: 18 hr.

Duration of Release: 30 hr.

1

nnAPT 6-12 6.7 GESARII, 238 Nuc. Island, Appendix 15-D, Severe Accidents - Rev. 2 -

General Electric Company,1982.

1 l

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7.0

SUMMARY

OF RESULTS MAFT Four severe accident sequences were analyzed for Peach Bottom.

These sequences were identified in Task 3.2 as dominant sequences that could potentially lead to core melting. The analyses assumed the accidents pro-ceeded with minimal operator intervention in order to determine the timing and magnitude of the major phenomenological events. The results of NAAP produce time estimates for core melting, vessel failure, and containment failure as l

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. As seen from Table 7.1 all sequences led to core melting. Assuming the pressure and temperature failure criteria utilized in these analyses, to sequences resulted in con-tainment failure due to pressure. Containment failure for the other two sequences resulted due to high temperature which occurred prior to the over-pressurization criteria. Table 7.2 compares the results of the analyses with WASH-1400. The release fractions of fission 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 frcm several hours to over a day af ter initiation of the accident.

The base sequences were reanalyzed with MAAP to demonstrate the effectiveness of selected operator actions in mitigating the consequences of thase severe accidents. The examples presented in Section 5 demonstrate that proper operator actions are extremely beneficial. There are several alterna-tives available to operators with present systems and procedures at various stages of the accident sequences to bring the plant to a safe stable state.

The assumption that these actions are not taken in the base sequences is unrealistic and makes these cases very low probability events.

Section 5 describes the capabilities that exist at Peach Bottom for venting primary containment. As indicated, venting capacity from the wetwell and drywell is extensive. The gases released from the small line: through the Standby Gas Treatment System and all lines connected to the wetwell are effectively filtered or scrubbed prior to release. The effectiveness of

7-2 9

Table 7.1

SUMMARY

OF MAAP RESULTS FOR BASE SEQUENCES

Event TW TC SE j TQVW ECCS Start (hrs) 0 0 NA 0 ECCSStop(hrs) 32 1.8 NA 6 Core Uncovered (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 2.5 11.4 Vessel Failure (hrs) 40 3.9 3.4 12.4 Fuel Melting Complete (hrs) 75 22 30 35 ContainmentFailure(hrs) 32 NA 23 18 Drywell Temp. at 600'F (hrs) 45 9 19 14 Orywell Temp. at 1500*F (hrs) 62 14 27 21 Max. Containment Pressure (psia) 132 115 b5 105 In-Vessel Zirc 0xidation (lbm H2) 430 300 250 500 Containment Failure Mode Pres NA Temp Temp (BasedonAssumedCriteria)

These sequences assume minimal operator intervention.

1 l

Table 7.2

SUMMARY

OF FISSION PRODUCT RELEASE FRACTIONS (a)

Sequence WASH-1400 F.P. Group TW TC SEj TQVW BWR2(b) BWR(c)

Cesium Iodine 0.13 0.034 0.01 0.05 0.50, 0.90 0.10 Tellurium 0.13 0.066 0.01 0.04 0.30 0.30 Strontium 9 x 10-5 6 x 10-0 2 x 10-5 5 x 10-5 0.10 0.01 Ruthenium 4 x 10 ~4 2 x 10 -4 6 x 10-5 2 x 10~4 0.03 0.02 7

(a) fraction of core inventory released to the environer.ent.

, (b) Containment failure prior to vessel failure; can be compared with l

(TW,TC).

(c) Failure to scram or remove decay heat; can be compared with (TC, Sj E, TQVW).

2::7 .

.bn m

i M

~

DRAFT venting in reducing pressure is also demonstrated in Section 5 for the TW analyses for which it was assumed operators vented from the wetwell in ac-cordance with existing emergency procedures.

It is apparent that a large margin exists in suppression pool venting capacity. Thus, mitigating features such as additional containment

, vent filters (FVCs) are of considerably diminished incremental value, in the unlikely event that venting would be required.

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

l l

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8.0 CONCl.USIONS BRAFT 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. Additionally, fission product releases could be substantially reduced from those calculated in the base cases through the use of existing primary containment venting capabilities and procedures.

4. The containment floor (pedestal) concrete ablation depths at the time of containment failure illustrated in the graphs of Section 4 indicate that base mat penetration is not a likely mode of Mark I containment failure for severe nuclear power accidents.

So2 ram "5. Heatup of containment from radiative and convective heat transfer from core debris on the containment floor may result in a reduction of the ultimate pressure capability of the containment for some sequences.

6. The reactor building (secondary containment) is extremely l effective in retaining aerosol and condensed volatile fission products released from primary containment. ,

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. 1

APPENDIX A Peach Bottom Parameter File BRAFT PEACHEP.DAT;15 6-JUL-1984 14:24 Page 1 AAMARK I BWR PLANT PARAMETER VALUES-- TYPICAL OF PEACH BOTTOM !

AASI UNITS (M-KG-SEC-DEGK) !

AA 7-22-83 !

AA !

APRIMARY SYSTEM PS 01 8.86200 AFLCOR FLOW AREA 0F REACTOR CORE PS A4 ALSH a VOLUME WATER IN LOWER SHROUD BELOW TOP OF ACTIVE FUEL DIVIDED AA BY (ZT0AF-ZBJET) 0 26h FfBYP 0$Bh$5SEkW REA P AA AUSH = VOLUME OF WATER IN UPPER DOWNCCMER ABOVE T0AF DIVIDED BY AA THE NORMAL WATER HEIGHT ABOVE TOAF 04 20.0D0 AUSH FLOW AREA IN UPPER SHROUD PS 05 1.35D5 HCRD SPECIFIC ENTHALPY OF CRD INLET PS 06 8.171D5 HEW SPECIFIC ENTHALPY OF FEEDWATER PS 07 1.586D5 MU2COR TOTAL MASS OF UO2 IN CORE PS 08 7.64D2 NASS NUMBER OF FUEL ASSEMBLIES IN REACTOR CORE PS 09 6.4D1 NPIMS NUMBER OF FUEL RODS IN A FUEL ASSEMBLY PS 10 1.85D2 NCED NUMBER OF CRD TUBES PS' 11 5.0D0 NGEPS SESNIBLE ENERGY STORED IN FUEL (FULL POWER SECONDS)PS 12 3.000 TDMSIV DELAY TIME FOR MSIV CLOSURE PS 13 3.500 TDSCRM DELAY TIME FOR FULL SCRAM PS 14 3.007 TIRRAD TOTAL EFFECTIVE IRRADIATION TIME FOR CORE PS Ap ggg PUMP CURVES ASSOCIATE THE FIRST FLOW ENTRY WITH THE FIRST PRESSURE 15 7.D-3 WVCRDI CBD FLOW RATE PUMP CURVE FOR CRD VOL FLOW PS 16 1.12D-2 WVCRDI CRD FLOW RATE PRESSURE PRIM SYS. VS WVCRDI PS 17 1.12D-2 WVCRDI CRD FLOW RATE PS 18 1.129-2 WVCRDI CRD FLOW RATE PS 19 1.120-2 WVCRDI CRD FLOW RATE PS 20 1.12D-2 WVCRDI CRD FLOW RATE PS 21 1.12D-2 WVCRDI CRD FLOW RATE PS 22 1.12D-2 WVCRDI CRD FLOW RATE PS 23 6.89406 PCRD PPS FOR CRb PUMP PS 24 1.0134D5 PCRD PPS FOR CRD PdMP PS 25 1.0134D5 PCRD PPS FOR CBD PUMP PS 26 1.0134D5 PCRD PPS FOR CRD PUMP PS 27 1.0134D5 PCRD PPS FOR CRD PUMP PS 28 1.0134D5 PCRD PPS FOR CBD PUMP PS 29 1.0134D5 PCRD PPS FOR CRD PUMP PS 30 1.013405 PCRD PPS FOR CRD PUMP PS 31 2275.00 WFWMAX MAXIMUM FEEDWATER ELOW RATE (RUN OUT) PS 32 4.42D2 WBPMAX MAXIMUM TURBINE BYPASS FLOW RATE PS 33 1.4D-1 NXCORE EXIT CORE QUALITY PS 34 5.2600 XDCORE REACTOR CORE DIAMETER TO INNER SHROUD WALL PS 35 22.23D0 XHRV INTERIOR HEIGHT OF REACTOR VESSEL PS 36 3.188D0 XRRV INTERIOR RADIUS OF REACTOR VESSEL PS 37 48.11D0 ZBJET ELEVATION AT BOTTOM OF JET PUMPS PS 38 46.2900 ZBRDT ELEVATION AT BOTTOM OF CRD TUBES PS 39 58.19D0 ZBSEP ELEVATION AT BOTTOM OF STEAM SEPARATORS PS 40 45.19D0 ZBV ELEVATION AT BOTTOM OF REACTOR VESSEL PS 41 50.6500 ZCPL ELEVATION AT CORE PLATE PS 42 53.4000 ZTJET ELEVATION AT TOP OF JET PUMPS PS 43 .65D0 AJET TOTAL JET PUMP AREA PS e4 54.4300 ZT0AF ELEVATION AT TOP OF ACTIVE FUEL PS 45 60.35D0 ZISEP ELEVATION AT TOP OE STEAM SEPARATORS PS 46 59.4900 ZWNORM ELEVATION AT NORMAL SHROUD WATER LEVEL PS 47 62.0000 ZLOCA ELEVATION AT BREAK PS 48 .0093D0 ALOCA AREA 0F BREAK PS 49 60.00D0 ZWL8 ELEVATION AT LEVEL 8 TRIP PS 50 0.0D0 NOT USED

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DRAFT '

PEACHEP.DAI;15 6-JUL-1984 14:24 Page 2 51 58.87D0 ISCRAM LEVEL 2 TRIP PS I 52 7.342D6 PSCRAM HIGH PRESSURE SCRAM SETPOINT PS 53 .2000 EGATWS ATWS CONSTANT POWER ASSUMPTION PS 1 54 3.6D3 IDSLC TIME FOR SCRAM WITH SLC PS AA RR PUMP COASTDOWN CURVE 55 0.D0 TIRR(1) IIME VS. ERACTION OF TOTAL ELOW EOR RECIRC PUMP PS 56 2.D0 TIRR(2) PS 57 4.D0 TIRR(3) PS 58 6.D0 TIRR(4) PS 59 8.D0 TIRR(5) PS 60 10.D0 TIRR(6) PS 61 12.D0 TIRR(7) PS 62 14.00 TIRR(8) PS 63 1.D0 FWRR(1) PS 64 .77D0 EWRR(2) PS 65 .62D0 FWER(3) PS 65 .5300 FWRR(4) PS 67 .45D0 FWRR(5) PS 68 .34D0 FWRR(6) PS 69 .34D0 FWRR(7) PS 70 0.D0 EWRR(8) PS 71 1.46D5 HSLC INLET ENTHALPY OF SLC PS 72 0.D0 PSLC(1) PRESSURE POINTS FOR SLC ELOW CURVE PS ,

73 1.07 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 ELOW 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 1.72D-3 WVSLC(5) PS 85 1.72D-3 WVSLC(6) PS 86 1.72D-3 WVSLC(7) PS 87 1.72D-3 WVSLC(8) PS 88 0.00 TDRPT DELAY TIME FOR RECIRC PUMP TRIP PS 89 54.79D0 ZLMSIV LOW WATER LEVEL FOR MSIV CLOSURE PS 90 57.64D0 ZLRPT LOW WATER LEVEL FOR RPT PS 91 7.886D6 PHRPT HIGH VESSEL PRESSURE EOR RPT PS 92 1.1513D5 PDWSCM HIGH DRYWELL PRESSURE SCRAM SIGNAL PS 93 .029900 EENRCH NORMAL EUEL ENRICHMENT PS 94 20000.D0 EXPO AVERAGE EXPOSURE IN MWD / TONNE PS 95 .6D0 FCR PRODUCTION OF U239 TO ABSORBTION IN EVEL PS 96 1.300 EFAE RATIO OF FISSILE ABSORBTION TO TOTAL EISSION PS 97 6.96D-1 EQFR1 FISSION POWER ERACTION OF U235 AND PU241 PS '

98 2.2380-1 E0ER2 FISSION POWER FRACTION OF PU239 PS

' gg 30 ?584300 lhEllflillHBE815"T5ES""v238 101 .275500 XDCRDI OUTER DIAMETER OF CRD TUBES PS 102 55.00 NINST NUMBER DE INSTRUMENT TUBES PS 103 .0041900 XTHCRD THICKNESS OF CRD TUBE WALL PS 104 .050800 XDINST OUTER DIAMETER OF INSTRUMENT TUBE PS 105 .075D0 XDRIVE LOWER CRD DRIVE OUTER DIAMETER PS 106 1.0051D-3 VWCRD SPECIFIC V0MUME OF CRD WATER PS

~

18 .55 MY0hs MkkS OE U hER hLk E UN$kNK h 109 1.016D3 AEOPS AREA 0F UPPER PLENUM HEAT SINK PS 110 .21437D0 XTRV THICKNESS DE LOWER VESSEL HEAD PS l

I A-3 7

~J PEACHEP.DAT;15 6-JUL-1984 14:24 Page 3 111 0.D0 TIEWCD TIME SINCE MSIV CLOSURE SIGNAL VS. EEEDWATER PS 112 0.D0 COASTDOWN MASS ELOW RATE PS 113 0.D0 8 IIME POINTS,8 ELOW RATES PS 114 0.D0 PS 115 0.D0 PS 116 0.D0 PS 117 0.D0 PS 110 0.D0 PS

"'c

H2 8:88 in 121 0.DO PS 122 0.D0 PS 123 0.00 PS 124 0.D0 PS 125 0.00 PS 126 0.00 PS 127 5.96D6 PLMSIV LOW RPV PRESSURE FOR MSIV CLOSURE PS 62.35D0 ZMSL ELAVATION AT CENTER LINE DE THE MAIN STEAM LINE g hCIRC 01 0.D0 ACSHS(1) CORE + LOWER PLENUM AA CARBON STEEL-HEAT SINK HEAT TRANSEER AREA 02 140.00 ACSHS(2) UPPER PLENUM 03 0.00 ACSHS(3) DOWNCOMER 04 0.D0 ACSHS(4) 05 0.D0 ACSHS(5) 06 50.D3 MCS(1) CORE + LOWER PLENUM CARBON STEEL MASS 07 100.D3 MCS(2) UPPER PLENUM DOWNCOMER 0[h k50

88 D3MC!(3)

Ef W 11 0.D0 MHS(1) CORE + LOWER PLENUM HEAT SINK MASS 12 100.D3 MHS(2) UPPER PLENUM

" ""c "5" 11 15 8:88 0.00 still!

MHS(5) 16 0.D0 AC!r%(1) CORE + LOWER PLENUM CARBON STEEL TO DRYWELL AA HEAT TRANSEER ARE 17 0.00 ACSX(2) UPPER PLENUM 18 240.00 ACSX(3) DOWNCOMER 19 0.D0 ACSX(4) 20 0.D0 ACSX(5) 1 21 0.D0 AHSX(1) CORE + LOWER PLENUM HEAT SINK TO DRYWELL AA HEAT TRANSEER AREA '

22 140.00 AHSX(2) UPPER PLENUM 23 0.D0 AHSX(3) DOWNCOMER 24 0.00 AHSX(4) '

25 0.00 AHSX(5) 26 100.00 AGCS(1) CORE + 10WER PLENUM GAS TO CARBON STEEL AA HEAT TRAnSEER AREA 27 5.D3 AGCS(2) UPPER PLENUM 28 240.D0 AGCS(3) DOWNCOMER N

31

.O 0.D0

!b AGHS(1) CORE + LOWER PLENUM GAS TO HEAT SINK AA HEAT TRANSEER AREA 32 140.00 AGHSf2) UPPER PLENbM !

33 0.00 AGHS(3) DOWNCOMER '

34 0.D0 AGHS(4) 35 0.D0 AGHS(5) 36 8.000 XL(1) CORE + LOWER PLENUM LENGTH

DHAFT PEACHEP.DAT;15 6-JUL-1994 14:24 Page 4 37 5.D0 XL(2) UPPER PLENUM LENGTH 38 10.D0 XL(3) DOWNCOMER LENGTH 39 0.D0 XL(4) 40 0.D0 XL(5) 41 11.D0 AG(1) CORE + LOWER PLENUM ELOW AREA h N.*N AG( 0 R 0 EA 44 0.D0 AG(4) 45 0.D0 AG(5) 46 5.D0 DH(1) HYDRAULIC DIAMETER FOR CORE REGION 47 .1500 DH(2) HYDRAULIC DIAMETER EOR UPPER PLENUM 48 .400 DH(3) HYDRAULIC DIAMETER FOR DOWNCOMER 49 0.00 DH(4) 50 0.D0 DH(5) 51 0.D0 0C0 RPV CONVECTION LOSSES AT TIME ZERO 52 8.00 FINPLT NUMBER DE LAYERS IN REELECTIVE INSULATION AHEATUP HE 01 3.8100 XZEUEL LENG:H OF ACTIVE FUEL HE 02 5.21D-3 XREUEL RADIUS DE FUEL PELLET HE 03 8.13D-4 XTCLAD THICKNESS DE CLADDING HE 04 3.5D4 MZRCAN TOTAL MASS OE ZR IN ASSEMBLY CANS HE 05 1.7D4 MBCR TOTAL MASS OF CONTROL BLADES IN REACTOR CORE HE 06 3.048D-3 XZRCAN CAN WALL THICKNESS HC AA NODE 1,1 IS CENTER BOTTOM,1,10 IS CENTER TOP, 2,1 IS SECOND RADIAL AA REGION AT THE BOTTOM 07 1.061D0 EPEAK(1I)ETC. PEAKING FACTOR FOR NODE (1 1) HE 08 .898D0 EPEAK(21) PEAKING EACTOR FOR H0DE (21) HE 09 .32600 EPEAK(3 1) PEAKING EACTOR EOR NODE (3 1) HE 15 1.43800 EPEAK(1 2) PEAKING EACTOR FOR NODE (12) HE 16 1.217D0 EPEAK(2 2) PEAKING EACTOR FOR NODE (2 2) HE 17 .44200 EPEAK(3 2) PEAKING EACTOR FOR H0DE (3 2) HE 23 1.55300 EPEAK(1 3) PEAKING EACTOR E0R NODE (1 3) HE 24 1.315D0 EPEAK(2 3) PEAKING EACTOR FOR NODE (2 3) HE 25 .477D0 EPEAK(3 3) PEAKING EACTOR FOR N0DE (3 3) HE 31 1.592D0 EPEAK(1 4) PEAKING EACTOR FOR NODE (1 4) HE 32 1.34800 EPEAK(2 4) PEAKING EACTOR FOR NODE (2 4) HE 33 .489D0 EPEAK(3 4) PEAKING EACTOR FOR NODE (3 4) HE 39 1.60500 EPEAK(1 5) PEAKING FACTOR FOR NODE (1 5) HE 40 1.35800 EPEAK(2 5) PEAKING FACTOR FOR NODE (2 5) HE 41 .493D0 FPEAK(3 5) PEAKING EACTOR E0R NODE (3 5) HE 47 1.63100 EPEAK(1 6) PEAKING EACTOR EOR N0DE (1 6) HE 48 1.381D0 EPEAK(2 6) PEAKING EACTOR FOR NODE (2 6) HE 49 .50100 EPEAK(3 6) PEAK 1NG EACTOR EOR NODE (3 6) HE 55 1.567D0 EPEAK(1 7) PEAKING EACTOR FOR NODE (1 7) HE 56 1.32600 EPEAK(2 7) PEAKING Et,CTOR FOR NODE (2 7) HE 57 .482D0 EPEAK(3 7) PEAKING EA2 TOR FOR NODE (3 7) HE 63 1.54100 EPEAK(1 8) PEAKING EACTOR FOR NODE (18) HE 64 1.304D0 EPEAK(2 8) PEAKING EACTOR FOR NODE (2 8) HE 72 i

.98600 0 bhAb h!$ hA bh EPEAK(2 9) PEAKING EACTOR FOR NODE (2 9) b !

HE 73 .358D0 EPEAK(3 9) PEAKING EACTOR FOR NODE (3 9) HE 79 .77700 EPEAK(1 1C) PEAKING EACTOR EOR NODE (1 10) HE 80 .658D0 HE 81 .23900 EPEAK(2',10)

EPEAK(310) PEAKING PEAKING EACTOR EACTOR EUR NODE E0R NODE (310) (2',10) HE 87 0.33D0 XCHIM UNHEATED EUEL LENGTH AT TOP OF CORE HE 88 1.D-7 XIZROX INITIAL CLADDING OXIDE THICKNESS HE AA ES AA ES AENGINEERED SAFEGUARDS ES 01 2.00 NLPCIl hUMBER DE LPCI PUMPS IN LOOP 1 ES l

BRMT PEACHFP.Def;15 6-JUL-1984 14:24 Page 5 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 0.000 NOT USED 06 100.D0 VMMCST MIN. WATER VOLUME IN CONDENSATE STORAGE TANK ES 07 1.0051D-3 VWCST SPECIFIC VOLUME OF CST WATER ES AA PUMP CURVES ARE DEFINED SO THE FIRST PRESSURE ENTRT CORRESPONDS WITH l AA THE FIRST VOLUMETRIC FLOW ENTRY 08 8.124D6 PHPCI(1) PUMP CURVES FOR ECCS ES U

11 2:8k 5.1D6 inhlf3!

PHPCI(4)

U!i"*"'""""c'"" !!

ES l

12 3.06D6 PHPCI(5) ES

! f.' kD6 hff f ES 15 6.17D5 PHPCI(8) ES 16 .31567DO WVHPCI(1) ES 17 .31567D0 WVHPCI(2) ES 18 .31567D0 WVHPCI(3) ES 19 .31567D0 WVHPCI(4) ' ES 20 .31567D0 WVHPCI(5) ES 11 : lit?!8 00 3 hl!!! 11 23 0.0D0 WVHPCI(8) ES 24 2.17206 PLPCI(1) LPCI ES 12 27 1:41282 htlelf!! !!

1.66D6 PLPCI(4) ES 28 1.478806 PLPCI(5) ES 29 1.065D6 PLPCI(6) ES 30 7.894D5 PLPCI(7) ES 31 1.01342D5 PLPCI(8) ES 32 0.D0 WVLPCI(1) ES hk f.*b b~f WV bbffhf kk 35 2.5250-1 WVLPCI(4) ES 38 5.05D-1 C

WVLPCI(7) kk ES 39 6.310-1 WVLPCI(8) ES 40 2.09906 PLPCS(1) LPCS ES 41 2.0306 PLPCS(2) ES 42 1.961D6 PLPCS(3) ES 43 1.892D6 PLPCS(4) ES 44 1.824D6 PLPCS(5) ES

.bbh h h) kh 47 1.01342D5 PLPCS(8) ES 48 0.D0 WVLPCS(1) ES 49 .041D0 WVLPCS(2) ES 50 .063D0 WVLPCS(3) ES 51 .078DO WVLPCS(4) ES 3 *b5h

. WUbbSkk E!

54 .189500 WVLPCS(7) ES 55 .246D0 WVLPCS(8) ES 72 7.006 PRCIC(1) RCIC ES

! .0Dk htCIhff ES 75 4.0 D6 PRCIC(4) ES 76 3.006 PRCIC(5) ES l

1

)

DRAFT

, PEACHEP.DAT;15 6-JUL-1984 14:24 Page 6 77 2.006 PRCIC(6) ES 78 1.134D6 PRCIC(7) ES 79 6.17D5 PRCIC(8) ES 80 .0378800 WVRCIC(1) ES l

8j :8??!iB8 001 5 8

.03788D0 WVRCIC(4) 11 ES 84 .03788D0 WVRCIC(5) ES 85 .0378800 WVRCIC(6) ES 86 .03788D0 WVRCIC(7) ES 87 0.000 WVRCIC(8) ES 88 57.6600 ZLHPCI LOW WATER INITIATION FOR HPCI ES 89 1.1513D5 PSHPCI HIGH DRYWELL PRESSURE SET POINT FOR HPCI ES 90 25.00 TDHPCI IIME DELAY FOR HPCI ES 91 6.205D5 PHHPCI MININUM VESSEL PRESSUES FOR HPCI TURBINE ES 92 50.D0 ZLHPCS LOW WATER IMITIATION EOR HPCS ES 93 1.1513D5 PSHPCS HIGH DRYWELL PRESSURE SET POINT EOR HPCS ES 96 5b9D0 ZbC$ Wf[Ib!!bINFORLPCI 1.151305 PSLPCI HIGH DRYWELL PRESSURE SET POINT FOR LPCI ES 97 24.00 TDLPCI TIME DELAY FOR LPCI ES

. 98 3.2057D6 PLLPCI LOW VESSEL PRESSURE PERNISSIVE EOR LPCI LOCA SIGNALES 99 54.79D0 ZLLPCS LOW WATER INITIATION FOR LPCS ES 100 1.1513D5 PSLPCS HIGH DRYWELL PRESSURE SET POINI FOR LPCS ES 101 12.00 TDLPCS TIME DELAY FOR LPCS ES fhh b kh Lh hkER I A N RI ES 104 1.1513D5 PSRCIC HIGH DRYWELL PRESSURE SET POINT FOR RCIC ES 105 30.D0 TDRCIC TIME DELAY FOR RCIC ES 106 6.20535 PHRCIC MINIMUM VESSEL PRESSURE FOR RCIC TUR91NE ES 107 1.35.405 HCST ENTHALPY OF CST ES 100 290.00 WSWHX SERVICE WATER FLOW RATE (KG/S) PER RHR HTX ES AA THE SAFETY RELIEF VALVES MUST BE ENIERED IN ORDER OF INCREASING AA PRESSURE ACTUATION SET POINTS AA I.E. GROUP $1 PSRVl=7.7D6 AA GROUP 02 PSRV2=7.75D6........ETC 109 .8605D-2 ASRV1 ELOW AREA 0F RELIEF VALVE TYPE 01 ES lif:128?h:1thlO3ft8:tili81litfllOttOIfill!!

112 .85970-2 ASRV4 ELOW AREA 0F RELIEF VALVE TYPE $4 11 ES AA IF THE AREA FOR GROUP 05 IS INPUT AS A NEGATIVE NUMBER THEN THESE AA VALVES WILL DISCHARGE DIRECTLY INTO THE DRYWELL IF THE AREA IS AA POSITIVE THEN THESE VALVES DISCHARGE INTO THE SUPPRESSION POOL 113 .8659D-2 ASRV5 ELOW AREA 0F SAEETY VALVE TYPE 95 ES 114 1.000 NSRV1 NUMBER OF TYPE il RELIEF VALVES ES

.000 N RV N BEROFfff R LIE VAL E ES 117 3.000 NSRV4 NUMBER OF TYPE $4 RELIEF VALVES ES 118 2.D0 NSRV5 NUMBER OF TYPE 65 RELIEF VALVES ES f h.*hh 121 2.00 NA N OF hh iEf NADS3 NUMBER OF ADS VALVES IN GROUP 3 kh ES 122 3.00 NADS4 NUMBER OF ADS VALVES IN GROUP 4 ES 123 7.717906 PSRV1 PRESSURE SETPOINT FOR el RELIEF VALVE ES 124 7.717906 PSRV2 PRESSURE SETPOINT FOR 92 RELIEF VALVE ES 125 7.7730D6 PSRV3 PPESSukE SEIPOINT FOR 93 RELIEF VALVE ES 126 7.8557D6 PSRV4 PRESSURE SETPOINT FOR 44 RELIEF VALVE ES 127 8.600306 PSRV5 PRESSURE SETPOINT FOR 95 RELIEF VALVE ES 128 54.7900 ZLADS LOW WATER LEVEL FOR ADS INITIATION ES 129 115.13D3 PSADS HIGH DRYWELL PRESSURE SET POINT FOR ADS ES 130 105.00 TDADS TIME DELAY EOR ADS ACTUATION ES l

DRAFT _

PEACHfP.DAT;15 6-JUL-1994 14:24 Page 7 AA LPCS.LPCI HAVE NPSH REQUIRMENTS, RCIC AND HPCI TRIP ON HIGH SUPP POOL AA TEMPERATURE ES 131 366.33D0 TCHPCI INLET TEMP LIMIT FOR HPCI ES 132 27.88D0 ZCLHPS PUMP CENTER LINE ELAVATION FOR HPCS ES 133 27.88D0 2CLLPI PUMP CENTER LINE ELAVATION FOR LPCI ES 134 27.8800 ZCLLPS PUMP CENTER LINE ELAVATION EOR LPCS ES 12 366.3300 TCRCIC INLET TEMP LIMIT FOR RCIC SERVICE WATER TEMP (RHR HEAT EXCHANGERS,7 COLD) ES 136 300.D0 TWSW ES 17 L.00 TDDG1 HPCS LOAD DELAY TIME FOR DIESEL jf.*D DD h LhADD fTfEf0RDIESEL ES ES 140 1.D-3 XDDROP SPRAY DROPLET DIAMETER ES 141 3.66D0 XHSPWW SPRAY EALL HEIGHT IN WETWELL ES 142 14.0200 XHSPDW SPRAY EALL HEIGHT IN DRYWELL AA THE EOLLOWING PUMP CURVES CAN BE USED TO DEFINE ANY INJECTION SYSTEM ES 143 1.35D5 HWHPSW ENTHALPY Of HIGH PRES SERVICE WATER (MARK I )

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

kI O $H El I O RATE FOR HA PU P 145 1.D10 PHPSW(1) PPS VS. VOLUMETRIC ELOW EOR HPSW CORE INJECTION ES 146 1.D10 PHPSW(2) - (MARK I CORE INJECTION) ES ES 147 1.010 PHPSW(3)

ES 148 1.D10 PHPSW(4)

ES 149 1.D10 PHPSW(5)

ES 150 1.D10 PHPSW(6)

ES 151 1.010 PHPSW(7)

ES 152 1.D10 PHPSW(8)

ES 153 3.79D-2 WVHPSW(1)

ES 154 3.790-2 WVHPSW(2)

ES 155 3.79D-2 WVHPSW(3)

ES 156 3.79D-2 WVHPSW(4)

ES 157 3.790-2 WVHPSW(5)

ES 158 3.79D-2 WVHPSW(6) 161 1.D10 h- H ff PDWSPR DRYWELL PRES SET PT EOR MARK III CONTAINMNT SPRAYS ES kh 162 1.D10 PWWSPR WETWELL PRES SET PT EOR MARK III CONTAINMNT SPRAYS ES ES 163 0.D0 TDSPR TIME DELAY EOR MARK 111 CONTAINMENT SPRAYS ES 164 2.413D5 PDSRV1 DEAD BAND EOR SRVel ES 165 2.413D5 PDSRV2 DEAD SAND EOR SRV62 166 2.413D5 PDSRV3 DEAD BAND EOR SRVt3 ES ES 167 2.41305 PDSRV4 DEAD BAND EOR SRVt4 ES 168 2.413D5 PDSRV5 DEAD BAND E0R SRV45 ES 169 7.48D6 PTURHP(1) PPS-PWW VS. STEAM ELOW TO HPCI TURBINE 170 7.928D5 PTURHP(2) ES ES 171 7.92805 PTURHP(3)

ES 172 7.92805 PTURHP(4)

ES 173 7.928D5 PTURHP(5)

ES 174 7.92805 PTURHP(6)

ES 175 7.92805 PTURHP(7)

ES 176 7.928D5 PTURHP(8)

ES 177 23.00 WSTHPI(1)

ES 178 12.00 WSTHPI(2)

ES 179 12.00 WSTHPI(3)

ES 100 12.00 WSTHPI(4)

ES 181 12.D0 WSTHPI(5)

ES

> 182 12.D0 WSTHPI(6) 18 11:88 185 7.7D6

!!Ill!il PTURRI(1) PPS-PWW VS. STEAM ELOW TO RCIC TURBINE ES ES 186 1.013D6 PTURRI(2)

- - - - , , -,----,--.,-,--n----a -,,,- - -,-,-- ,,

DRAFT PEACHEP.DAT;15 6-JUL-1984 14:24 Page 8 187 1.013D6 PTURRI(3) ES 188 1.013D6 PTURRI(4) ES 189 1.013D6 PTURRI(5) ES 190 1.013D6 PTURRI(6) ES ,

191 1.013D6 PTURRI(7) ES l 192 1.013D6 PTURRI(8) ES 193 3.500 WSTRCI(1) ES 194 1.000 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 1.000 WSTRCI(9) ES 201 1.1355D6 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 3.202D6 PHLPCS HIGH PRESSURE TRIP FOR LPCS ES 206 33.64D0 ZHISP HIGH SUPP. POOL LEVEL FOR HP/RCIC SUCTION SWITCH ES 207 0.00 ZLSPR LOW WATER LEVEL FOR AUTO WETWELL SPRAYS (M-III) ES AA IF THE DETAILED HEAT EXCHANGER IS NOT USED ONLY SUPPLY THE AA NIU VALUE AND NUMBER OF HTXS 200 0.D0 NTHX NUMBER OF TUBES IN kHR HTX - ES 209 0.00 NBHX NUMBER OF BAFFLES IN RHR HTX ES 210 0.00 X1DTHX TUBE ID FOR RHR HTX ES 211 0.00 XTTHX TUBE WALL THICKNESS FOR RHR HTX ES 212 0.00 XICHX TUBE CENIER TO CENTER SPACING FOR RHR HTX ES 213 0.00 XSHX SHELL LENGTH FOR RHR HTX ES 214 0.D0 RGEOUL FOULING FACTOR FOR RHR HTX ES 215 0.D0 KTHX THERMAL CONDUCTIVITY FOR IUBE WALL (RHR HTX) 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 .65400 NTUHX1 NTU FOR RHR HTX 91 ES 220 .654D0 NTUHX2 NTU FOR RHR HTX $2 ES 221 2.00 NHX1 NUMBER OF RHR LOOP t1 HTX ES 222 2.00 NHX2 NUMBER OF RHR LOOP 42 HTX ES 223 1.D0 FHX TYPE OF RHR HTX(1= STRAIGHT TUBE 2=U TUBE) ES 224 21.6D3 TDBATT3ATTERYOPERATIONTIMEFORSTATIONBLACK-0UT ES AA THE NPSH POINTS CORRESPOND WITH THE ABOVE FLOW RATE FOR THAT PUMP 233 0.D0 ZHDLP1 NPSH CURVE FOR LPCI VS. FLOW (ABOVE) ES 234 0.00 (METERS) ES 235 0.00 ES 236 7.80800 ES 237 7.576800 ES 238 8.,213D0 ES 239 8.37500 ES 240 8.79600 ES 241 0.00 ZHDLPS NPSH CURVE FOR LPCS VS. FLOW (ABOVE) ES 242 7.969D0 ES 243 8.02400 ES l 244 0.07600 ES t 245 8.134D0 ES l 246 8.38100 ES

! 247 9.116D0 ES 2 .'8800 CENTER LINE ELEVATION FOR RCIC PUMP bh 250 27.88D0 CENTER LINE ELEVATION FOR HPCI PUMP ES

, 251 .009300 ACVENT AREA 0F CONTAINMENT VENT ES l 252 0.00 2CFAIL ELEVATION OF CONTAINMENT VENT IN WETWELL (MII ONLY)ES 1

I

1

. 1 EAFT PEACHfP.DAT;15 6-JUL-1994 14:24 Pagn 9 233 03.8D0 ZSRVD AVERAGE ELEVATION DE SRV DISCHARGE IN SUPP POOL ES 254 0.D0 IGDWHX(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 0.00 TGDWHX(4) 258 0.D0 TGDWHX(5) 259 0.D0 TGDWHX(6) 260 0.D0 TGDWHX(7) lh 8.*N hb k HEAT LOSS RATE FOR DRYWELL COOLERS (J/S) ih!8:88 265 0.D0 8HE!!

OGDWHX(4) 268 0.D0 GGDWHX(7) 269 0.00 OGDWHX(8)

AA DW AA DW ADRYWELL DW 01 .500 RELHDW RELATIVE HUMIDITY IN DRYWELL DW 02 4841.D0 VOLDW EREC VOLUME OF DRYWELL DW 03 36.5500 ZDWF ELEVATION AT DRYWELL FLOOR DW 04 84.00 ADWF AREA 0F DRYWELL ELOOR DW 05 37.2400 ZWDWWW ELEVATION OF DRYWELL-WETWELL WALL DW

. G E E ISI E hR I Ek bCEILING 08 0.D0 ACHDW ELOOR BURN AREA DW A

UU AWETWELL WW 01 28.6500 ZWWF ELEVATION AT WETWELL TLOOR WW

[ N E Oh U B S 04 3.447D3 PSETVB PRESSURE SETPOINT FOR VACUUM BREAMERS WW 05 2.757D3 PDVB DEAD BAND FOR VACUUM BREAKERS WW 06 7419Do VOLWW EBEE VOLUME OF WETWELL (MARK II AND MARK III ONLY) WW 07 1.D0 RELHWW RELATIVE HUMIDITY IN WETWELL WW 08 0.00 MIGWW NUMBER OF IGNITERS IN THE WETWELL WW 09 0.D0 XIGWW AVERAGE DISTANCE FROM IGNITERS TO CEILING WW 10 0.00 ACHWW ELOOR BURN AREA WW 11 0.00 AWWE AREA DE WETWELL ELOOR (MARK II) WW AA PD AA PD APEDESTAL PD 01 2.91701 APDF AREA 0F PEDESTAL FLOOR PD 2.kb2 h0L hbM 0 NEhA b 06 5 h

.500 f S 0 b RELHPD RELATIVE HUMIDITY IN PEDESTAL PD 07 0.D0 NIGPD NUMBER OF IGNITERS IN THE PEDESTAL PD 08 0.00 XIGFD AVERAGE DISTANCE EROM IGNITERS TO CEILING PD 13 8.R ^DMTMDisBREtiliAtDOOR(MARX!!ONtT)

A B 11 0.D0 ADCPD AREA Of PEDESTAL DOWNCOMERS 12 0.00 NDCPD NUMBER OF DOWNCOMERS IN PEDESTAL 13 2.00 XHPDDW DISTANCE BETWEEN UPPER AND LOWER VENTS FOR i

AA PED-DRYWELL NATURAL CIRCULATION i

AA TO AA TO

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

I A-10 CI A

ghj' $}k *$ u PEACH!P.DAT;15 6-JUL-1984 14:24 PaSe 10 ATORUS AND MARK II WETWELL TO 01 4.72D0 XRTOR MINOR RADIUS OF TORUS (MI ONLY) TO 02 106.700 XLTOR CIRCUMFERENCE OF TORUS (MI ONLY) TO 03 .292D0 ADC AREA 0F DOWNCOMER (MI AND MII ONLY) TO 04 96.00 NDC HUMBER OF DOWNCOMERS (MI AND MII ONLY) TO AA THE NEXT PARAMETER HAS BEEN REPLACES WITH MEQWW AND MEQWWS 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) 70 08 28.800 ZBTOR ELEVATION AT BOTTOM Of TORUS (MI ONLY) TO 09 0.0168D0 XTOR THICKNESS OF TORUS SHELL (MI ONLY) TO 10 5072.D0 ATR AREA 0F TORUS ROOM WALL (MI ONLY) TO 11 101342.D0 PTR PRESSURE IN TORUS ROOM (MI ONLY) TO 12 1.5D4 VOLTR VOLUME OF TORUS ROOM (MI ONLY) TO 13 34.74700 ZVBTOR CENTER LINE ELEVATION OF VACUUM BRKRS(MI AND MII) TO 14 .015DO XTHDC IHICKNESS OF DOWNCOMER PIPE (MII ONLY) TO AA N1TIAL CONDITIONS IN 01 3.293D9 QPOWER CORE POWER IN 02 7.033D6 PPS0 INITIAL PRESSURE IN PRIMARY SYSTEM IN 03 1.04105 PPD 0 INITIAL PRESSURE IN PEDESTAL IN 04 1.04105 POWO INITIAL PRESSURE IN DRYWELL IN 05 1 04105 PWWO INITIAL PRESSURE IN WETWELL IN 06 33.2200 ISPDWO INIT.ELEV. OF WATER LEVEL IN DW SIDE OF SUPP. POOL IN AA (FOR MI AND M11 THIS IS ELEVATION IN 00WNCOMER) IN 07 33.2200 ZSPWWO INIT.ELEV. OF WATER LEVEL IN WW SIDE OF SUPP. POOL IN 00 3.3D2 TPD0 INITIAL TEMPERATURE IN PEDESTAL IN 0 2 NT L TEM E N A TW L M 11 3.05D2 TWSPO INITIAL TEMPERATURE OF SUPPRESSION POOL WATER IN 12 59.49D0 ZWSHO INITIAL ELEVATION OF WATER IN THE SHROUD IN 13 0.00 MWCB0 MASS OF WATER IN UPPER POOL (MARK!!! ONLY) N f$ .khl.' h T B AkBIEI EfERkTURE IN 16 1.0134D5 PAMB AMBIENT PRESSURE IN AA AA AHTSINKS HS 01 189.00 AHS1 AREA 0F WALL 61 PEDESTAL-DRYWELL WALL HS 02 1507.D0 AHS2 AREA 0F WALL 62 DRYWELL WALL HS 03 0.00 AHS3 AREA 0F WALL 43 DRYWELL FLOOR HS 04 5073.D0 AHS4 AREA 0F WALL 04 TORUS ROOM WALL (Mt ONLY) HS 09 1.300 KHS1 THERMAL CONDUCTIVITY OF WALL 61 HS 10 1.300 KHS2 THERMAL CONDUCTIVITY OF WALL 92 HS 11 1.300 KHS3 THERMAL CONDUCTIVITY OF WALL 63 HS 12 1.300 KHS4 THERMAL CONDUCTIVITY OF WALL 64 HS 17 1.3200 XHS1 THICKNESS OF WALL 01 HS 18 1.8300 XHS2 THICKNESS OF WALL (2 HS 19 1.D0 XHS3 THICKNESS OF WALL v3 HS 20 1.07DO XHS4 THICKNESS OF WALL 04 HS 25 0.00 XLHS!1 INNER LINER THICKNESS FOR WALL el HS 26 2.713D-2 XLHS!2 INNER LINER THICKNESS FOR WALL 02 HS 33 8:88 0.00 lisil!I"silElsflIslE!lli18!!!tt!!

XLHS01 OUTER LINER THICKNESS FOR WALL 41

1 HS 34 0.00 XLHS02 OUTER LINER THICKNESS FOR WALL 42 HS 35 0.00 XLHS03 OUTER LINER THICKNESS FOR WALL 63 HS 36 0.D0 XLHSO4 OUTER LINE2 THICKNESS FOR WALL 64 HS 41 2300.00 DHS1 DENSITY OF WALL 61 HS 42 2300.00 DHS2 DENSITY OF WALL 02 HS

~

PEACHFP.DAT;15 6-JUL-1984 14:24 BRMT Page 11 43 2300.D0 DHS3 DENSITT OF WALL 03 HS 44 2300.D0 DHS4 DENSITT OF WALL 64 HS 49 880.D0 CPHS1 SPECIFIC HEAT FOR WALL 01 HS 50 880.00 CPHS2 SPECIFIC HEAT FOR WALL 02 HS 51 880.D0 CPHS3 SPECIFIC HEAT FOR WALL 03 HS 52 880.D0 CPH84 SPECIFIC HEAT FOR WALL 04 HS 57 0.D0 MEOPD MASS OF EQUIPMENT IN PEDESTAL HS

. 2D4 kA 0 E T HS 13 64 fA 20.D0 M 85!

AEQWW

!!!! 81 15 1f:! M !! i! n!E '

AREA 0F EQUIPMENT IN WETWELL HS 67 50.D0 HTOUTW HEAT TRANSFER COEFF. AT DUTER WALL HS 68 0.D0 RGAPIl INNER LINER TO WALL GAP RESISTANCE 01 HS 69 1.D0 RGAP!2 INNER LINER TO WALL GAP RESISTANCE 02 HS h

72 0.

0.D0 I ff I ER L R TO k hh SfSf 9 RGAP15 INNER LINER TO WALL GAP RESISTANCE 45 HS 73 0.D0 RGAPI6 INNER LINER TO WALL GAP RESISTANCE 06 HS 74 0.D0 RGAPl7 INNER LINER TO WALL GAP RESISTANCE 47 HS 75 0.D0 RGAP18 INNER LINER TO WALL GAP RESISTANCE 48 HS h

78 h.*

0.D0 R P0 T Lf R I A S E 02 RGAP03 OUTER LINER TO WALL CAP RESISTANCE 43 HS 80 . 05 TER L NER TO kkhhRkSk!kAN4 B1 0.D0 RGAP06 0 UTER LINER TO WALL GAP RESISTANCE 46 HS 82 0.00 RGAP07 CUTER LINER TO WALL GAP RESISTANCE 47 HS h

85 5.b32D4 !!

1.D2

$ hfk Ebu kTNkbBkkkbkb Nk AE0WWS AREA EQUIP. WETWELL (SURMERGED) HS 86 36.D0 AHSPS1 AREA 0F RPV IN GAS SPACE OF VESSEL HS 87 44.00 AHSPS2 AREA 0F RPV IN DOWNCOMER REGION HS N h $ND5 $ WkL'NSdI fE E REGION H 90 2.0213D5 MRPV2 RPV WALL MASS IN DOWNCOMER REGION HS g 1.5176D5 MRPV3 RPV WALL MASS IN LOWER PLENUM REGION AA HS AMODEL PARAMETERS FOR DWR 01 .00500 FRCOEF FRICTION COEFFICIENT FOR CORIUM IN VFAIL MO 02 2.00-1 FMAXCP FRACTION OF TOTAL CORE MASS WHICH MUST MELT A4 TO FAIL THE CORE PLATE 03 50.00 HTSLAD FUEL CHANNEL TO CONTROL BLADE HEAT TRANS. COEFF MO 1

1 D6 X FU hANN X E1AR EhEIt $

A 0=BLOCXAGE AT TT00FF 1=NO BLOCKAGE NO 06 2300.00 TZOOFF OXIDATIONCUT-OFFTEEPERATURE 07 MO

.3DO FACPF FRACTION OF CORE PLATE AREA THAT FAILS MO 0! hM kNbW flN!bhf$CfNAGC FFIC !N NT 0 10 5.00 CDBWW FLAME BUOTANCT DRAG COEFFICIENT IN THE WETWELL M0

. hD B h Bh C RAGkE EN Nb AR MEN BM 13 .1000 XCHREF C0k!UM REFERENCE THERMAL BOUNDART LAYER THICKNESS MO 14 1.03 HTCMCR CORIUM-CRUST HEAT TRANSF. COEFF. USED IN DECOMP M0 15 0.05D0 XCMX MINIMUM CORIUM THICKNESS ON DRYWELL FLOOR AND PED M0 i

N 0.01D0 XDCMSP fab $Cb$52fbO$AMETER)FORCORIUMASITFALLS $O AA INTO SUPPRESS 10H POOL (MARK 11 ONLT) MO 17 983.00 TCFLAM CRITICAL FLAME TEMPERATURE MO

DRAFT PEACHFP.DAT;15 6-JUL-1994 14I24 Page 12 18 1.53D0 FCHTUR CHURN-TURBULENT CRITICAL FLOW PARAMETER M0 N 3.D0 D 0 G R E 21 1.00 FSPAR PARAMETER FOR BOTTOM-SPARGED STEAM VOID FRACTION M0 2 5.D-1 TR NT FFEC E II $1 N 24 .9000 EW EMISSIVITY OF WATER MO 25 .85D0 EWL EMISSIVITY OF WALL M0 it .85D0 20

in li" EEG

!alillOlli8lEn'""

EMISSIVITT OF EQUIPMENT N

MO 29 0.500 F0VER FRACTION OF CORE SPRAY FLOW ALLOWED TO BYPASS COREMO 30 1.00 NPE NUMBER OF PENETRATIONS FAILED IN LOWER HEAD N0 g 2.00 FCDCDW OgRgIMETERPERMETERFROMPEDESTAL000R g 32 0.1400 FCHF COEFFICIENT FOR CHF CORRELATION IN PLSIM MO 33 .7500 FCDBRK DISCHARGE COEFFICIENT FOR PIPE BREAK M0 AA A

hR Eh fT T D FOR MATERIAL TO BE BLOWN OUT OF CAVITY 08 L 1 0) NO MO 35 1.00 SCALU SCALING FACTOR FOR ALL BURNING VELOCITIES MO 36 1.00 SCALH SCALING FACTOR FOR HT COEFFICIENTS TO PASSIVE NO N 2.0D0 EUMIN ADD URFACE MULTIPLIER

$CONCRETEPROPERTIES N 02 k D0 $ $ RE M T G M R h j g% !";lSEN jiillud"!!Hiji'IE E8EllT!'c "' S'" " E" 0 572 0 DCC CN C I O EE b 07 1.D6 LHCN LATENT HEAT FOR CONCRETE MELTING CH AFISSION PRODUCTS Ft 01 .02800 FOP (1) PERCENT POWER IN FISSION PRODUCT GROUP 1 FI h *NNO

. hh $ P!RC Th0 ! h kh b h!!!bbk bbOh h h!

04 0.00 F0P(4) PERCENT POWER IN FLSSION PRODUCT GROUP 4 FI 05 0.00 FOP (5) PERCENT POWER IN FISSION PRODUCT GROUP 5 FI 06 412.7 MFPt!) MASS OF FISSION PRODUCT GROUP 1 -NOBLES F{

MfP3 MASSOfFfShf0N OD CT C -E I 09 167.7 MFP(4) MASS OF FISSION PRODUCT GROUP 4 -SR FI 060 1) S F CORE I h 12 432.0 MSM0(2) MASS OF MN IN CORE REGION FI 11 8:888 181 Fill!!!!ii!!80tt!!!!i8!If!!!#2 !!

12 8:888 17 0.000

[811111111Ilis80!tt:8:181ff8881!

FDSP(5) SPRAY REMOVAL LAMDA FOR FP GROUP 5 IlFI 18 600.00 FDFSP(1) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 1 FI 19 600.00 FDESP(2) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 2 FI 20 600.000 FDFSP(3) DP.YWELL VENTS DECON. FACTOR FOR FP GROUP 3 FI l 21 600.000 EDFSP(4) DRYWELL VENTS DECON. FACTOR FOR FP GROUP 4 FI i 6 b bbkk0 hkC Nh OR h0W1 hk 24 1000.00 FDFSP(2) SRV DECON. FACTOR FOR FP GROUP 2 FI 25 1000.000 EDFSP(3) SRV DECON. FAF wR FOR FP GROUP 3 FI 26 1000.000 FDFSP(4) SRV DECON. FM TOR FOR FP GROUP 4 FI

~

DRAFT PEACHEP.DAT;15 6-JUL-1994 14:24 Page 13 27 1000.000 FDFSP(5) SRV DECON. FACTOR FOR FP GROUP 5 FI AA FI AA FI ACONTROL CARDS CC 01 1 IBWR CONTAINMENT TYPE (MARN 12 OR 3 ) CC 02 46 IRSTW UNIT NUMBER TO WRITE RESIAIrf FILE (MAIN) CC 03 47 INU'J UNIT NUMBER TO WRITE RESTART FILE (HEATUP) CC 04 40 IPOUT UNIT NUMBER TO WRITE PROGRAM OUTPUT FILE CC 05 41 IPLT1 FIRST PLOT FILE NUMBER (TOTAL OF 4 FILES) CC h

AA ff k Xf M MBEROfk0 01N TRACED FOR FULL SCALE SPIKE CC 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=lST ORDER R-N 2=2ND ORDER R-N :C 12 1 IFREEZ1*D0FREEZEFRhNTCALC.(0=NOCALC.) CC 13 5 INPGRP NUMBER OF TRACE GAS TYPES (FISSION PRODUCTS) CC 14 0 1 RET WRITE RETAIN PLOT FILE CC 15 49 IFPPLT RETAIN PLOT FILE UNIT NUMBER CC

. 36 0 IFPRAT 1=NUREG-0772 FP RELEASES ; 0=CUBICC10TTI CC AA TD AA TD ATIMING DATA TD 01 20.00 TDMAX MAXIMUM ALLOWED TIME STEP TD 02 1.0-3 TDMIN MINIMUM ALLOWED TIME STEP TD 03 4. I - 2 FMCHMX MAXIMUM MASS CHANGE FRACTION FOR INTEGRATION TD 04 5.D J FUCHMX MAXIMUM ENERGY CHANGE FRACTION E0W INTEGRATION TD g 1.010 MAXMST MAXIMUM MASS OF STEAM CHANGE PER TIMF STEP IN PS TD 9

l l

i B-1

~

BRAFT i

APPENDIX B Supplemental Plots for the Base Accident Sequences 4

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=.

+

B-3 B. RAFT 1

I SUPPLEMENTAL PLOTS FOR SEQUENCE TW 1

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DRAFT O

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Fig. B.3 Temperature of structure. *F.

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i 4 .

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1 TIME m xlO Fig. B.12 Cesium and fodine released to environment, kg.

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SUPPLEMENTAL PLOTS FOR SEQUENCE TC

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TIME m xlO

Fig. B.34 Cesium and iodine released from containment, kg.

1 4

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W

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I I

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i rig. B.43 cas temperature in reactor butiding. F.

4 m

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TIME m xto

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i l Ffg. 8.44 Steam pressure in reactor butiding, Pa.

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Fig. B.46 Cesium anti fodine released to environment, kg.

.

ML20095C773

Contents

Text

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

8.0 CONCLUSION

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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