D. D. Yue - ORNL
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ond 0/JJ< H'DGE NATIONAL LABORATORY LIBRARY 3 4i45b DDSllflfi 4 NUREG/CR-2182 (ORNL/NUREG/TM-455/V1) Station Blackout at Browns Ferry Unit One—Accident Sequence Analysis D. H. Cook S. R. Greene R. M. Harrington S. A. Hodge D. D. Yue OAK RIDGE NATIONAL LABORATORY CENTRAL RESEARCH LIBRARY CIRCULATION SECTION 4500N ROOM 175 LIBRARY LOAN COP DO NOT TRANSFER TO ANOTHER PERSON If you wish someone else to see this report, send in name with report and the library will arrange a loan. Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Interagency Agreements DOE 40-551-75 and 40-552-75
Transcript of D. D. Yue - ORNL
ond0/JJ< H'DGE NATIONAL LABORATORY LIBRARY
3 4i45b DDSllflfi 4 NUREG/CR-2182
(ORNL/NUREG/TM-455/V1)
Station Blackout at Browns
Ferry Unit One—AccidentSequence Analysis
D. H. Cook S. R. Greene
R. M. Harrington S. A. HodgeD. D. Yue
OAK RIDGE NATIONAL LABORATORY
CENTRAL RESEARCH LIBRARY
CIRCULATION SECTION
4500N ROOM 175
LIBRARY LOAN COPDO NOT TRANSFER TO ANOTHER PERSON
If you wish someone else to see thisreport, send in name with report andthe library will arrange a loan.
Prepared for the U.S. Nuclear Regulatory CommissionOffice of Nuclear Regulatory Research
Under Interagency Agreements DOE 40-551-75 and 40-552-75
Printed in the United States of America. Available fromNational Technical Information Service
U.S. Department of Commerce5285 Port Royal Road, Springfield, Virginia 22161
Available from
GPO Sales ProgramDivision of Technical Information and Document Control
U.S. Nuclear Regulatory CommissionWashington, D.C. 20555
This report was prepared as an account of work sponsored by an agency of theUnitedStates Government. NeithertheUnitedStatesGovemment noranyagencythereof, nor any of their employees, makes any warranty,express or implied,orassumes any legal liability or resoonsibility for the accuracy, completeness, orusefulness of any information, apparatus, product, or process disclosed, orrepresents that its use would not infringe privately owned rights. Reference hereinto any specific commercial product, process, or service by trade name, trademark,manufacturer, or otherwise, does not necessarily constitute or imply itsendorsement, recommendation, or favoring by the United States Government orany agency thereof. The views and opinions of authors expressed herein do notnecessarily state or reflect those of the United States Government or any agencythereof.
NUREG/CR-2182, Vol. IORNL/NUREG/TM-455/VIDist. Category RX, IS
Contract No. W-7405-eng-26
Engineering Technology Division
STATION BLACKOUT AT BROWNS FERRY UNIT ONE
ACCIDENT SEQUENCE ANALYSIS
D. H. Cook S. R. Greene
R. M. Harrington S. A. HodgeD. D. Yue
Manuscript Completed — October 6, 1981
Date Published — November 1981
Notice: This document contains information of a preliminarynature. It is subject to revision or correction and thereforedoes not represent a final report.
Prepared for theU. S. Nuclear Regulatory Commission
Office of Nuclear Regulatory Research
Under Interagency Agreements DOE 40-551-75 and 40-552-75
NRC FIN No. B0452
Prepared by theOAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830operated by
UNION CARBIDE CORPORATION
for the
DEPARTMENT OF ENERGY
OAK RIDGE NATIONAL LABORATORY LIBRARY
3 445b DDEllfifl 4
XXI
CONTENTS
Page
SUMMARY v
ABSTRACT 1
1. INTRODUCTION 1
2. DESCRIPTION OF STATION BLACKOUT 5
3. NORMAL RECOVERY 7
4. COMPUTER MODEL FOR SYSTEM BEHAVIOR PRIOR TO LOSS OF
INJECTION CAPABILITY 17
5. INSTRUMENTATION AVAILABLE DURING STATION BLACKOUT AND
NORMAL RECOVERY 30
6. OPERATOR ACTIONS DURING STATION BLACKOUT AND NORMAL
RECOVERY 33
7. COMPUTER PREDICTION OF THERMAL-HYDRAULIC PARAMETERS FOR
NORMAL RECOVERY 36
7.1 Introduction 36
7.2 Conclusions 36
7.2.1 Normal recovery — conclusions 367.2.2 Loss of 250 vdc — conclusions 37
7.3 Normal Recovery 37
7.3.1 Normal recovery — assumptions 377.3.2 Normal recovery — results 38
7.4 Loss of 250 vdc Batteries 50
7.4.1 Loss of 250 vdc batteries —assumptions 507.4.2 Loss of 250 vdc batteries — results 50
8. FAILURES LEADING TO A SEVERE ACCIDENT 59
8.1 Induced Failure of the HPCI and RCIC Systems 628.2 Stuck-Open Relief Valve 658.3 Loss of 250-Volt DC Power 68
9. ACCIDENT SEQUENCES RESULTING IN CORE MELTDOWN 71
9.1 Introduction 71
9.2 Accident Phenomenology 71
9.2.1 Accident progression resulting in coremelt 71
9.2.2 Containment failure modes 72
9.3 Event Trees 74
9.3.1 Event trees for accident sequences resultingin core melt 74
9.3.2 Core damage event tree 749.3.3 Containment event tree 74
9.4 Accident Sequences 789.4.1 MARCH computer code 789.4.2 Accident progression signatures 78
9.5 Containment Responses 1239.5.1 Drywell responses 1239.5.2 Wetwe11 responses 149
10. PLANT STATE RECOGNITION AND OPERATOR MITIGATING
ACTION 161
10.1 Introduction 161
10.2 Plant State Recognition 161
XV
10.3 Operator Key Action Event Tree10.4 Operator Mitigating Actions
11. INSTRUMENTATION AVAILABLE FOLLOWING LOSS OF 250 VOLT
DC POWER
12. IMPLICATIONS OF RESULTS
12.1 Instrumentation
12.2 Operator Preparedness12.3 System Design
13. REFERENCES
ACKNOWLEDGEMENT
APPENDIX A. Computer Code used for Normal RecoveryAPPENDIX B. Modifications to the March Code
APPENDIX C. March Code Input (TB')APPENDIX D. Pressure Suppression Pool Model
D.l Introduction
D.2 Purpose and Scope
D.3 Description of the SystemD.4 Identification of the Phenomena
D.5 Pool Modeling ConsiderationsAPPENDIX E. A Compendium of Information Concerning the
Browns Ferry Unit 1 High-Pressure CoolingInjection System
E.l PurposeE.2 System DescriptionE.3 HPCI Pump SuctionE.4 System InitiationE.5 Turbine TripsE.6 System Isolation
E.7 Technical SpecificationsAPPENDIX F. A Compendium of Information Concerning the
Unit 1 Reactor Core Isolation Cooling SystemF.l PurposeF.2 System DescriptionF.3 RCIC Pump SuctionF.4 System InitiationF.5 Turbine TripsF.6 System IsolationF.7 Technical Specifications
APPENDIX G. Effect of TVA-Estimated Seven Hour BatteryLife on Normal Recovery Calculations
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SUMMARY
This study describes the predicted response of Unit 1 at the BrownsFerry Nuclear Plant to a hypothetical Station Blackout. This accidentwould be initiated by a loss of offsite power concurrent with a failure ofall eight of the onsite diesel-generators to start and load; the only remaining electrical power at this three-unit plant would be that derivedfrom the station batteries. It is assumed that the Station Blackout occurs
at a time when each of the Browns Ferry units is operating at 100% powerso that there is no opportunity for use of the batteries for units 2 or 3in support of unit 1.
The design basis for the 250 volt DC battery system at Browns Ferryprovides that any two of the three unit batteries can supply the electrical power necessary for shutdown and cooldown of all three units for aperiod of 30 minutes with a design basis accident at any one of the units.It is further provided that the system voltage at the end of the 30 minuteperiod will not be less than 210 volts, and that all DC equipment suppliedby this system must be operable at potentials as low as 200 volts.
It is clear that the 250 volt system was not designed for the case ofa prolonged Station Blackout, and the period of time during which the DCequipment powered by this system could remain operational under these conditions can only be estimated. It would certainly be significantly longerthan 30 minutes since all three batteries would be available, the equipment is certified to be operable at 200 volts, and there would be no design basis accident. In response to AEC inquiry in 1971, during the period of plant construction, TVA estimated that the steam-driven High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC)systems, which use DC power for turbine control and valve operation, couldremain operational for a period of four to six hours. A period of fourhours has been assumed for this study.*
Within 30 seconds following the inception of a Station Blackout, thereactor would have scrammed and the reactor vessel would be isolated be
hind the closed main steam isolation valves (MSIV's). The initial phaseof the Station Blackout extends from the time of reactor vessel isolation
until the time at which the 250 volt DC system fails due to battery exhaustion. During this period, the operator would maintain reactor vesselwater level in the normal operating range by intermittent operation of theRCIC system, with the HPCI system available as a backup. Each of thesewater-injection systems is normally aligned to pump water from the condensate storage tank into the reactor vessel via a feedwater line.
*It should be noted that the events subsequent to failure of the 250volt DC system are relatively insensitive to the time at which this failure occurs. This is because the only parameter affecting the subsequentevents which is dependent upon the time of failure is the slowly-varyingdecay heat.
As part of a requested review of the results of this study, TVA performed a battery capacity calculation which shows that the unit batteriescan be expected to last as long as seven hours under blackout conditions.In Appendix G, the effect of a battery lifetime of seven vice four hoursupon study results is shown to be limited to timing.
vx
The operator would also take action during the initial phase to control reactor vessel pressure by means of remote-manual operation of theprimary relief valves; sufficient stored control air would remain available to permit the desired remote-manual valve operations for well overfour hours. The Control Room instrumentation necessary to monitor reactor
vessel level and pressure and for operation of the RCIC and HPCI systemswould also remain available during this period.
There is no Emergency Operating Instruction for a Station Blackout atBrowns Ferry. However, the existing written procedure for operator actionfollowing reactor isolation behind closed MSIV's provides for the reactorvessel level control and pressure control described above. The primaryrelief valves would actuate automatically to prevent vessel over pressuri-zation if the operator did not act; the purpose of pressure control byremote-manual operation is to reduce the total number of valve actuationsby means of an increased pressure reduction per valve operation and topermit the steam entering the pressure suppression pool to be passed bydifferent relief valves in succession. This provides a more even spacialdistribution of the transferred energy around the circumference of thepressure suppression pool.
The initial phase of a Station Blackout has been analyzed in thisstudy by use of a relatively simple computer code developed specificallyfor this purpose. This coding uses the Continuous Systems Modeling Program (CSMP) language of the IBM computer system to simulate the responseof Browns Ferry Unit 1 to postulated operator actions. The analysisshows, because of the loss of the drywell coolers, that it is necessaryfor the operator to begin to reduce the reactor vessel pressure to about0.791 MPa (100 psig) within one hour of the inception of the StationBlackout. This depressurization reduces the temperature of the saturatedfluid within the reactor vessel and thereby decreases the driving poten
tial for heat transfer into the drywell, yet keeps the vessel pressurehigh enough for continued operation of the RCIC system steam turbine.With this action, the drywell average ambient temperature can be kept below 149°C (300°F) throughout the initial phase of a Station Blackout;tests have shown that both the drywell structure and the equipment located
therein can be expected to survive temperatures of this magnitude.The analysis also reveals an important second reason for operator ac
tion to depressurize the reactor vesel early in the initial phase of aStation Blackout. This depressurization removes a great deal of steam andthe associated stored energy from the reactor vessel at a time when theRCIC system is available to inject replacement water from the condensatestorage tank and thereby maintain the reactor vessel level. Subsequently,when water injection capability is lost for any reason, remote-manual relief valve operation would be terminated and there would be no furtherwater loss from the reactor vessel until the pressure has been restored to
the setpoint [7.72 MPa (1105 psig)] for automatic relief valve actuation.Because of the large amount of water to be reheated and the reduced levelof decay heat, this repressurization would require a significant period oftime. In addition, the subsequent boiloff* would begin from a very high
*The term "boiloff" is used to signify a monotonic decrease in reactor vessel water level due to intermittent loss of fluid through theprimary relief valves without replacement.
vxx
vessel level because of the increase in the specific volume of the wateras it is heated and repressurized. Thus, an early depressurization willprovide a significant period of valuable additional time for preparativeand possible corrective action before core uncovery after injection capability is lost.
One design feature of the HPCI system logic was brought into questionby the analysis. The existing logic provides for the suction of the HPCIsystem pump to be automatically shifted from the condensate storage tankto the pressure suppression pool upon high sensed suppression pool level.During a Station Blackout, this would occur after about three hours whenthe average suppression pool temperature has reached about 71°C (160°F).*Since the lubricating oil for the HPCI turbine Is cooled by the water being pumped, this would threaten the viability of the HPCI system.
The rationale for the automatic shift in HPCI pump suction on highsensed pool level is not explained in the literature available to thisstudy. There is no corresponding provision for a shift of the RCIC pumpsuction on high pool level, and a separate logic is provided for an automatic shift of the HPCI pump suction should the supply of water from thecondensate storage tank become exhausted. For these reasons, it is recommended that the desirabilty of the automatic shift of HPCI pump suction onhigh sensed suppression pool level be reexamined.
The plant response during the initial phase of a Station Blackout canbe summarized as an open cycle. Water would be pumped from the condensatestorage tank into the reactor vessel by the RCIC system as necessary tomaintain level in the normal operating range. The injected water would beheated by the reactor decay heat and subsequently passed to the pressuresuppression pool as steam when the operator remote-manually opens the re
lief valves as necessary to maintain the desired reactor vessel pressure.Stable reactor vessel level and pressure control is maintained during thisperiod, but the condensate storage tank is being depleted and both thelevel and temperature of the pressure suppression pool are increasing.However, without question, the limiting factor for continued removal ofdecay heat and the prevention of core uncovery is the available of DC
power.
The sequence of events used for the fission product release analysisfor a prolonged Station Blackout was established by this study under theassumption that no independent secondary equipment failures would occur.Dependent secondary failures, i.e., those caused by the conditions of aStation Blackout, were included in the development of the sequence, butwith the assumption that the operator does take action to depressurize thereactor vessel and thereby prevent drywell temperatures of the magnitudethat could severely damage the equipment therein. The point is importantbecause the operator would probably be reluctant to depressurize; currenttraining stresses concern for high suppression pool temperatures (based onLOCA considerationst) and the operator would recognize that suppressionpool cooling is not available during a Station Blackout.
*It is expected that any accident sequence resulting in high suppression pool level would also produce an associated high pool temperature.
tThere is currently no written procedure for the case of StationBlackout.
viii
This study has established that the possible dependent secondaryfailure of a stuck-open relief valve would not reduce the reactor vesselpressure below that necessary for operation of the RCIC steam turbine during the period in which DC power would remain available. However, duringthis period all of the energy transferred to the pressure suppression poolwould pass through the tailpipe of this one relief valve and would be concentrated near its terminus. This would reduce the steam-quenching capacity of the pool to that provided by the effective volume of water surrounding this one tailpipe. Local boiling in this area would pass most ofthe relief valve discharge directly to the pool surface and could resultin early containment failure by overpressurization.
The question of suppression pool effectiveness in quenching reliefvalve steam discharge in cases which involve highly localized energytransfer from the reactor vessel is important to a complete analysis ofany severe accident in which the pressure suppression pool is used as aheat sink. This question has not been resolved, at least in the non-proprietary literature, and has been recently adopted as a dissertation topicfor a University of Tennessee doctorial candidate working with the SevereAccident Sequence Analysis (SASA) project. Pending the results of thiswork, pool-averaged temperatures are used in this study.
The MARCH code has been used in support of the analysis of the secondphase of a Station Blackout, i.e., the period after the 250 volt DC systemfails because of battery exhaustion. The existing versions of MARCH aretoo crude to permit modeling plant response to a series of postulated operator actions such as those previously discussed for the initial phase ofa Station Blackout. Therefore, in the event sequence modeled by the MARCHcode, the reactor vessel remains pressurized with pressure control byautomatic relief valve actuation and level control by automatic operationof the HPCI system. As before, averaged suppression pool temperatures areused, and it is assumed that injection capability is lost after fourhours, when the unit battery is exhausted.
In the MARCH event sequence, the reactor vessel water level is in thenormal operating range and the vessel is pressurized at the four-hourpoint when boiloff begins due to loss of injection capability.* The MARCHresults predict core uncovery 62 minutes after the beginning of boiloff,followed by the inception of core melting 53 minutes later. The modelprovides that the melted core slumps down to the bottom of the reactorvessel and this results in a predicted failure of the reactor vessel bottom head at approximately three hours after injection capability is lost.The subsequent breaching of the primary containment because of failure ofthe electrical penetration modules by overtemperature is predicted atabout four and one-half hours after the inception of boiloff. The detailed
*These conditions at the beginning of boiloff are similar to thosepredicted by the previously discussed sequence which models the plant response to operator actions during the initial phase because in that sequence, the reactor vessel would have to repressurize before boiloff couldbegin. The difference is in the timing. If injection capability werelost at the four-hour point, the boiloff would begin immediately if thevessel is pressurized, but would be delayed if the vessel were depressur-ized.
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thermo-hydraulic parameters needed for the analysis of fission product release from the fuel rods and the subsequent transport of fission productsto the environment were taken from the MARCH results for this sequence.
An estimate of the magnitude and timing of the release of the noblegas, cesium, and iodine-based fission products to the environment is provided in Volume 2 of this study. Under the conditions of a Station Blackout, fuel rod cladding failure would occur while the reactor vessel waspressurized. The distribution of both solid and gaseous fission productswithin the Browns Ferry Unit 1 core at the present time has been obtainedthrough use of the 0RIGEN2 code; the results show that internal gas pressure under Station Blackout conditions could not increase to the magnitude necessary to cause cladding failure by rod burst. Accordingly, thecladding has been assumed to fail at a temperature of 1300°C.
STATION BLACKOUT AT BROWNS FERRY UNIT ONE -
ACCIDENT SEQUENCE ANALYSIS
D. H. Cook S. R. Greene
R. M. Harrington S. A. HodgeD. D. Yue
ABSTRACT
This study describes the predicted response of Unit 1 atthe Browns Ferry Nuclear Plant to Station Blackout, defined asa loss of offsite power combined with failure of all onsiteemergency diesel-generators to start and load. Every efforthas been made to employ the most realistic assumptions duringthe process of defining the sequence of events for this hypothetical accident. DC power is assumed to remain availablefrom the unit batteries during the initial phase and the operator actions and corresponding events during this period aredescribed using results provided by an analysis code developedspecifically for this purpose. The Station Blackout is assumed to persist beyond the point of battery exhaustion andthe events during this second phase of the accident in whichDC power would be unavailable were determined through use ofthe MARCH code. Without DC power, cooling water could nolonger be injected into the reactor vessel and the events ofthe second phase include core meltdown and subsequent containment failure. An estimate of the magnitude and timing of theconcomitant release of the noble gas, cesium, and iodine-basedfission products to the environment is provided in Volume 2 ofthis report.
1. INTRODUCTION
The Tennessee Valley Authority (TVA) operates three nearly identicalreactor units at the Browns Ferry Nuclear Plant located on the TennesseeRiver approximately midway between Athens and Decatur, Alabama. The General Electric Company and the Tennessee Valley Authority jointly participated in the design; TVA performed the construction of each unit. Unit 1began commercial operation in August 1974, Unit 2 in March 1975, and Unit3 in March 1977.
Each unit comprises a boiling water reactor steam supply system furnished by the General Electric Company. Each reactor is designed for apower output of 3440 MWt, for a corresponding generated 1152 MWe; themaximum power authorized by the operating license is 3293 MWt, or 1067net MWe. The primary containments are of the Mark I pressure suppression pool type; the drywells are light bulb shape and both torus and dry-well are of steel vessel construction. The three units share a reactor
building-secondary containment of the controlled leakage, elevated releasetype.
Safety systems for each unit include a Reactor Protection System, aStandby Liquid Control System for Poison injection, and the Emergency CoreCooling Systems: High-Pressure Coolant Injection (HPCI), Automatic Depressurization (ADS), Residual Heat Removal (RHR), and Core Spray (CS).The Reactor Core Isolation Cooling (RCIC) system is also provided for theremoval of post-shutdown reactor decay heat as a consequence limiting system.
Several components and systems are shared by the three Browns Ferryunits. A complete description of these shared features is given in theFinal Safety Analysis Report;! the shared Safeguards systems and theirsupporting auxiliary equipment are listed in Table 1.1. With the assumptionthat the interfaces with the other two units do not interfere with the operation of any shared system as applied to the needs of the unit understudy, the existence of the shared systems does not significantly complicate the analysis of an accident sequence at any one unit.
The results of a study of the consequences at Unit 1 of a StationBlackout (loss of all AC power) at the Browns Ferry Nuclear Plant are presented in this report. Section 2 provides a description of the event anddiscussion of the motivation for consideration of this event. The normalrecovery from a Station Blackout is described in Sect. 3, the computermodel used for the normal recovery analysis is discussed in Sect. 4, andthe instrumentation available to the operator is described in Sect. 5.The actions which the operator should take to prolong the period of decayheat removal are discussed in Sect. 6, and computer predictions of the behavior of the thermal-hydraulics parameters during the period when a normal recovery is possible are displayed in Sect. 7.
A Severe Accident by definition proceeds through core uncovery, coremeltdown, and the release of fission products to the surrounding atmosphere. The equipment failures which have the potential to extend a Station Blackout into a Severe Accident are discussed in Sect. 8, and theSevere Accident sequences which would follow during a prolonged StationBlackout are presented in Sect. 9. The consequences of each of these sequences after the core is uncovered differ only in the timing of events;the actions which might be taken by the operator to mitigate the consequences of the Severe Accident are discussed in Sect. 10. The instrumentation available following the loss of injection capability during the period in which severe core damage occurs is described in Sect. 11.
The conclusions of this Station Blackout analysis and the implications of the results are discussed in Sect. 12. This includes consideration of the available instrumentation, the level of operator training, theexisting emergency procedures, and the overall system design.
Appendix A contains a listing of the computer program developed tomodel operator actions and the associated system response during the period when normal recovery is possible. The MARCH code was used for analyses of the severe accident sequences; the modifications made to this codeare described in Appendix B and an input listing is provided in AppendixC.
The pressure suppression pool is the key to the safe removal of decayheat from an isolated Boiling Water Reactor, but no satisfactory method
Table 1.1 Shared safeguards systems and their auxiliary support systems
System Quantity Function
Standby AC Power Supply System Four diesel generators each coupled as an alternate source ofpower to four Independent shutdownboards for Units 1 and 2. Fouradditional dtesel generators serveas alternate power sources to fourUnit 3 shutdown boards.
Three batteries, one per unit.The 250V DC power supplies for theDC-powered redundant services of
each unit are normally derivedfrom separate batteries. A fourthbattery is provided for commonstation service.
A spare control rod drive hydraulic pump Is shared between Units 1and 2.
Two pumps and two heat exchangersper unit with one common sparepump and heat exchanger for allthree units
System is unitized except for acommon discharge plenum, ducting,and the 600-ft stack
One common drain header Into which
each of the three unit reactor
building floor drainage pumps discharge
Supply emergency power during loss ofoffsite power conditions
250V DC Power Supply System
Control Rod Drive HydraulicSystem
Reactor Building Closed CoolingWater System
Gaseous Radwaste System
Station Drainage System
Standby Coolant Supply System Completely shared system
RHR Service Water System
Emergency Equipment CoolingWater System
Standby Gas Treatment System
Completely shared system
Completely shared system
Completely shared system used onlywhen an abnormal activity releaseoccurs
Supply DC power when battery chargersare not functioning
Provide driving force for normal rodmovement
Provide cooling water to reactorauxiliary equipment
Obtain elevated discharge from theStandby Gas Treatment System
Pass reactor building drainage to receiver tanks
Provide means for core cooling following a complete failure of the Residual Heat Removal (RHR) coolingcomplex
Provide an assured heat sink forlong-term removal of decay heat whenthe main condensers are not availablefor any reason
Distribute raw cooling water to equipment and auxiliary systems which arerequired for shutdown of all threeunits
Filter and exhaust the air from aunit zone and the refueling zone, therefueling zone only, or the entiresecondary containment
for the analysis of the response of the torus and pool under severe accident conditions currently exists, at least in the non-proprietary literature. The nature of the problem and the measures being taken at ORNL toimprove the analysis capability are discussed in Appendix D.
In the event of a Station Blackout, the reactor vessel will be isolated behind the closed Main Steam Isolation Valves (MSIVs) with pressuremaintained by periodic blowdowns through the relief valves to the pressuresuppression pool. Makeup water to maintain the reactor vessel level mustbe injected either by the High Pressure Coolant Injection (HPCI) or theReactor Core Isolation Cooling (RCIC) systems. The operation of these important systems is discussed in Appendices E and F.
As a portion of their review of the draft of this report, the Electrical Engineering Branch at TVA performed a battery capacity calculationwhich shows that the unit batteries at the Browns Ferry plant can be expected to provide power for as long as seven hours under station blackoutconditions. Since a battery life of four hours was assumed for the calculations of this work, a new Appendix G has been added in which all of thepossible failure modes other than battery exhaustion that were consideredin Sections 3, 7, and 8 are re-examined for applicability to the periodbetween four and seven hours after the inception of a Station Blackout.
The magnitude and timing of the release of the fission product noblegases and various forms of iodine to the atmosphere are discussed inVolume 2. This includes the development of release rate coefficients forthe escape of these fission products from fuel as a function of temperature, specification of the various chemical forms, determination of theoptimum strategy for the modeling of precursor/daughter exchange, and thestudy of the particular auxiliary systems involved to determine the applicable fission product release pathways. These results are incorporatedinto a vehicle for the calculation of the transport of the individual fission products from the reactor core, through the primary and secondarycontainments to the atmosphere, control volume by control volume.
The primary sources of information used in the preparation of thisreport were the Browns Ferry Nuclear Plant (BFNP) Final Safety AnalysisReport (FSAR), the Nuclear Regulatory Commission BWR Systems Manual, theBFNP Hot License Training Program Operator Training Manuals, the BFNP Unit1 Technical Specifications, the BFNP Emergency Operating Instructions, andvarious other specific drawings, documents, and manuals obtained from theTennessee Valley Authority*. Additional information was gathered by meansof one visit to the BFNP and several visits to the TVA Power OperationsTraining Center at Soddy-Daisy, Tennessee. The excellent cooperation andassistance of Tennessee Valley Authority personnel in the gathering of information necessary to this study are gratefully acknowledged.
*The setpoints for automatic action used in this study are thesafety limits as given in the FSAR. In many cases these differ slightlyfrom the actual setpoints used for instrument adjustment at the BFNP because the instrument adjustment setpoints are established so as to providemargin for known instrument error.
2. DESCRIPTION OF STATION BLACKOUT
A Station Blackout is defined as the complete loss of AC electricalpower to the essential and nonessential switchgear buses in a nuclearpower plant.2 At the Browns Ferry Nuclear Plant, a Station Blackoutwould be caused by a loss of offsite power concurrent with the tripping ofthe turbine generators and a subsequent failure of all onsite diesel generators to start and load. After these events, the only remaining sourcesof electrical power would be the battery-supplied 250 volt, 48 volt, and24 volt DC electrical distribution systems; AC power would be limited tothe instrumentation and control circuits derived from the feedwater inverter or the unit-preferred and plant-preferred motor- generator setswhich are driven by the DC systems.
The reliability of offsite power at Browns Ferry has been excellent.There are two independent and separated sources of 161 KV offsite power tothe plant, each from a different nearby hydroelectric station. In thenear future, offsite power will become even more reliable. System modifications will permit any of the three reactor units at Browns Ferry, in theevent of a generator trip, to receive reverse power from the TVA 500 KVAgrid which these units normally feed.
However, should all offsite power be lost together with the trippingof the turbine generators, there remain eight diesel generators at BrownsFerry which are designed to automatically start and load whenever normalAC power is lost. By design, all equipment required for the safe shutdownand cooldown of the three Browns Ferry units can be powered by six* ofthese diesel units, even with the assumption of a design basis accident onany one unit.3
Therefore, a Station Blackout is an extremely unlikely event. Nevertheless, the consequences of a Station Blackout should be studied. It isimportant to recall that one Browns Ferry unit suffered a loss of electrical power to all of its emergency core-cooling systems during the March22, 1975 electrical cable tray fire.4* To determine if Browns Ferry andsimilar nuclear plants can recover from the loss of other combinations ofAC powered machinery, it will prove most efficient to first consider theplant response to a loss of all AC powered equipment, as in a StationBlackout.
This study considers the effect on Unit 1 at Browns Ferry of a Station Blackout which begins with a loss of offsite power, associated turbine generator trip, and failure of all diesels to start. Barring furtherequipment failures, the battery powered systems have the capability tosupply the power necessary to operate the high-pressure water injectionsystems to maintain the reactor in a stable cooled state for severalhours. The first seven sections of this report pertain to the methodswhich can be used to prolong this period of water injection capability anddecay heat removal for as long as possible.
*With operator action, any six diesel generators would be satisfactory. However, for adequate short term shutdown and cooldown responsewithout operator action, the six operable diesel generators would have tocomprise three of the four provided for the Unit 1/Unit 2 complex andthree of the four provided for Unit 3.
However, injection capability would ultimately be lost during an extended Station Blackout, either through exhaustion of the installed battery capacity or earlier, through secondary failures of vital equipment.Once injection capability is lost, the Station Blackout would become aSevere Accident with inevitable core uncovery. The remainder of this report proceeds from an assumption that the Blackout does develop into aSevere Accident, involving core meltdown and subsequent fission productrelease to the atmosphere.
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10
As in the case of reactor vessel level control, it is intended that
the most realistic assumptions concerning pressure control be used in thisstudy. Accordingly, it will be assumed that the operator does take theactions in regard to pressure control as described above. However, itshould be noted that without operator action, the pressure would cycle between the limits of 7.377 and 7.722 MPa (1055 and 1105 psig); unfortunately, it is probable that this would occur through the repeated operation ofthe same relief valve causing severe localized heating in the vicinity ofthe tailpipe discharge in the suppression pool.
Short-term Plant Status. Within minutes after the inception of a
Station Blackout, both reactor vessel level control and pressure controlwill be achieved with or without operator action. An open cycle willexist, with water pumped from the condensate storage tank into the reactorvessel, converted to steam there by reactor decay heat, and subsequentlypassed to the pressure suppression pool and condensed. Satisfactory reactor cooling is provided, but both the pressure suppression pool leveland temperature are increasing, as is the drywell ambient temperature dueto loss of the drywell coolers. The reserve DC power stored in the batteries is being dissipated.
There is a 250V DC battery for each unit at Browns Ferry, plus a station battery for common loads. Each unit battery is designed to provide aguaranteed supply of direct current for the first 30 minutes of a designbasis accident. Under the less severe conditions of a Station Blackout,
and assuming prudent actions by the operator to conserve battery potential, DC power should be available for the first 4-6 hours of a StationBlackout.
Long-Term Considerations. Once DC power is lost, the HPCI and RCICsystems will be inoperable and cooling water can no longer be injectedinto the reactor vessel. With prudent and realistic operator actions,this should not occur until 4-6 hours* after the inception of a StationBlackout. However, there are other considerations which must be properlyaddressed during this 4-6 hour period during which all efforts would bemade to accomplish the restoration of AC power.
First, since all plant ventilation is stopped during a Station Black
out, the ambient temperature near the RCIC turbine may reach 93.3°C(200°F), causing an automatic isolation of the RCIC system. This is unlikely since the RCIC system will be operated only intermittently, and theturbine is not located in a closely confined space. If isolation does occur, the isolation signal can be overridden by the placing of rubber insulation between the appropriate relay contacts in the auxiliary instrumentation room, or the HPCI system can be used as a backup.
Second, a relief valve might stick open although it should be notedthat the currently installed two-stage Target Rock valves are much lessprone to this malfunction than their three-stage predecessors. A stuck-open relief valve would produce difficulties with localized suppressionpool heating, but reactor vessel pressure would not decrease to the point
that the HPCI or RCIC system turbines could not be used. This eventualityis discussed in more detail in Section 8.2.
*Recent battery capacity calculations have indicated that DC powermay remain available for as long as seven hours.
11
A third consideration would be the continued availability of conden
sate storage tank water for injection into the reactor vessel. As previously discussed, there is a guaranteed minimum stored supply of 511.0 m3(135,000 gallons) available to the HPCI or RCIC systems. As illustratedin Section 7, only about 359.6 m3 (95,000 gallons) would be used duringthe first five hours of a Station Blackout. It can be concluded that an
adequate supply of condensate storage tank water is provided for injectionduring the period in which DC power will remain available.
A fourth consideration is the availability of sufficient stored dry-well control air to permit the desired number of remote-manual reliefvalve actuations during the period of Station Blackout. The accumulatorsprovided for the six relief valves associated with the ADS system aresized to permit five operations per valve, or a total of 30 actuations.As illustrated in Section 7, less than 25 relief valve actuations will berequired during the first five hours of a Station Blackout. It can beconcluded that there is enough stored air in the ADS relief valve accumulators alone to permit the desired number of relief valve operations during the period in which DC power is available. In addition, the storedair in the Drywell Control Air system receivers provides an ample backupto the accumulator supply.
A fifth consideration concerns the steady heatup of the pressure suppression pool water with no means of pool cooling available during StationBlackout. The pressure suppression pool of approximately 3785 m3 (onemillion gallons) of water is contained within a torus of 33.99 m (111.5ft) major diameter and 9.45 m (31.0 ft) minor diameter. The torus surrounds the drywell and reactor vessel as shown in Fig. 3.1.
The large heat sink afforded by the pressure suppression pool watercan be effectively utilized during a Station Blackout only as long as thesteam being discharged into the pool via the relief valves is condensed.
If the local temperature of the water surrounding the relief valvetailpipe terminus in the pool is excessive, condensation oscillations mayoccur causing gross unstable vibrations of the torus assembly and pressu-rization of the torus airspace due to the escape of saturated steam fromthe water surface. This effect is discussed in detail in Appendix D. For
the "T-quencher" type of discharge header which is installed at the terminus of each relief valve tailpipe at Browns Ferry Unit 1, the condensation oscillations are not expected to occur if the local pool temperatureis limited to 93.3°C (200°F) or equivalent to 87.8°C (190°F) average pooltemperature.6
The T-quencher dischargers on the tailpipes of the thirteen primaryrelief valves are distributed approximately evenly around the torus, asshown in Fig. 3.2. This diagram is posted on the Browns Ferry Unit 1 control room panel which contains the relief valve operating switches. During a Station Blackout, it would be incumbent on the operator in his efforts to manually control reactor vessel pressure to ensure that oppositely located primary relief valves are operated in turn so that theenergy input to the pressure suppression pool is evenly distributed.
As illustrated in Section 7, the average pressure suppression pooltemperature will have reached approximately 82.2°C (180°F) five hoursafter the inception of a Station Blackout. It can be concluded that withoperator action to alternate the relief valve discharges, the pressuresuppression pool temperature will not reach the point when condensation
12
ORNL-DWG 81-8602 ETD
REACTOR VESSEL
DRYWELL
Fig. 3.1 Arrangement of drywell and torus.
oscillations might occur during the first five hours of a Station Black
out.
As a related matter, the HPCI system is equipped with logic whichwill automatically shift the suction of the HPCI pump to the pressure suppression pool if the indicated pressure suppression pool level increasesto a certain point, as discussed in Appendix E. Depending on the initialwater level in the pool, this will occur following an increase in poolvolume of between 257.4 and 370.9 m3 (68,000 and 98,000 gallons), orsometime between two and four hours into the Station Blackout. As previously discussed, the HPCI system is not expected to be utilized but merelyto serve as a backup to the RCIC system. However, to preserve the HPCI
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PCV 1-19 pcv l-»
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191°180°
ORNL-DWG 81-8603 ETD
RELIEF VALVE POSITIONSOPPOSITE VALVES
PCV 1-42 PCV 1-18
PCV 1-41 PCV 1-22
PCV 1-34 PCV 1-4PCV 1-30 PCV 1-5
PCV 1-31 PCV 1-2
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Fig. 3.2 Unit 1 relief valve discharges.
system as a viable backup, the operator should take action to maintain thepump suction on the condensate storage tank. This can be done by rackingout the motor-operated breakers for the suction valves, arid will preventthe excessive HPCI system lubricating oil temperatures which would resultfrom the pumping of the hot pressure suppression pool water.
The sixth and most important consideration for long term stabilityfollowing a Station Blackout is the heatup of the drywell atmosphere. Theten drywell coolers immediately fail on loss of AC power, while the rate
14
of heat transfer to the drywell at that instant from the hot [287.8°C(550°F)] reactor vessel and associated piping is about 0.98 MW (3.35 x10° Btu/h). Under this impetus, the drywell temperature rapidly increases. After a few minutes, the drywell temperature will have significantly increased so that the rate of heat transfer from the reactor vesselis reduced while a substantial rate of heat transfer from the drywell atmosphere to the relatively cool drywell liner has been established, asdiscussed in Section 4.
The design temperature for the drywell structure and the equipmentlocated therein is 138.3°C (281°F). The response of this structure and ofcertain safety components within the drywell to higher temperatures hasbeen previously investigated7 and the results are summmarized below:
It was determined that the drywell liner will not buckle under linertemperatures as high as 171.1°C (340°F) nor would this temperature producehigher than allowable stress intensity.
The drywell electrical penetrations were purchased with a specifiedshort term (15 min) temperature rating of 162.8°C (325°F) and a long termrating of 138.3°C (281°F).
It was determined that the stresses in the drywell piping penetrations would not exceed the allowable stress intensities at a temperatureof 171.1°C (340°F).
A DC solenoid control valve used for the remote-manual operation ofthe primary relief valves was tested at 148.9°C (300°F) for ten hours with118 actuations during the test period. An AC-DC solenoid control valvefor the Main Steam Isolation valves was tested at 148.9°C (300°F) for 7hours with 83 actuations each on AC, DC, and the AC-DC combination. Themaximum temperature achieved during each of these solenoid tests was153.3°C (308°F).
The electrical cable which feeds the safety equipment inside the dry-well has a 600-V rating with cross-linked polyethelene insulation. It hasseven conductors of number 12 wire with no sheath. It is estimated thatthe ten-hour temperature rating is in excess of 160.0°C (320°F).7
The results of these tests and investigations indicate that it can beconfidently predicted that the primary relief valves will remain operableand the drywell penetrations will not fail if the drywell ambient temperature is prevented from exceeding 148.9°C (300°F) during the normal recovery phase of a Station Blackout.
With the operator acting to maintain reactor vessel level by use ofthe RCIC system and to control vessel pressure between 7.688 and 6.309 MPa(1100 and 900 psig) by remote-manual relief valve actuation as previouslydiscussed, the drywell ambient temperature would reach 148.9°C (300°F) inone hour (curve A of Fig. 3.3). Before this occurs, the operator shouldact to reduce the reactor vessel pressure by blowdown to the pressure suppression pool. This will reduce the temperature of the saturated liquidwithin the reactor vessel and thereby reduce the driving potential forheat transfer into the drywell.
The vessel pressure reduction would be by remote-manual actuation ofthe primary relief valves, and should proceed at the Browns Ferry Technical Specifications limit of a rate equivalent to a 55.6°C/h (100°F/h) decrease in saturated liquid temperature. The response of the drywell ambient temperature to such a reactor vessel depressurization begun at onehour after the inception of a Station Blackout is shown by curve B of Fig.3.3.
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The curves of Fig. 3.3 show that the maximum drywell average temperature can be kept below 148.9°C (300°F) provided that the operator takesaction to depressurize within one hour following the loss of drywell coolers concomitant with a Station Blackout.* A depressurization at theTechnical Specifications limit of 55.6°C/h (100°F/h) will produce a gradual reduction of the drywell ambient temperature; a rapid depressurization will produce a faster drywell temperature reduction, but is probablynot necessary. Certainly, if the operator perceives developing difficulties with the in-drywell equipment such as the relief valve solenoids, heshould convert an ongoing 55.6°C/h (100°F/h) depressurization into a morerapid one.
Summary. Assuming no independent secondary failures of equipment,reactor vessel level control and pressure control can be maintained duringa Station Blackout for as long as the DC power supplied by the unit battery lasts. This is expected to be a period of from four to six hours,and this portion of the Station Blackout severe accident sequence isgraphically displayed with commentary in Section 7. If AC power is regained at any time during this period, a completely normal recovery wouldfollow.
If AC power is not regained, and DC power is lost, remote-manual operation of the primary relief valves would no longer be possible and thereactor vessel pressure would increase to 7.722 MPa (1105 psig), the lowest setpoint for automatic actuation of the relief valves. Thereafter,the reactor vessel pressure would cycle between 7.722 and 7.377 MPa (1105and 1055 psig) while the vessel level steadily decreased because injectioncapability would be lost as well. The sequence of events following theloss of injection capability is described in Sections 9 through 11.
*If the backup HPCI system were used for injection, the average reactor vessel pressure after depressurization would have to be maintainedat about 1.14 MPa (150 psig), which corresponds to an average saturatedliquid temperature of 185°C (365°F). This is only 15°C (25°F) higher thanthe average temperature associated with RCIC operation and would notsignificantly increase the drywell temperatures shown in Fig. 3.3.
17
4. COMPUTER MODEL FOR SYSTEM BEHAVIOR
PRIOR TO LOSS OF INJECTION CAPABILITY
The purpose of this chapter is to describe the BWR-LACP (BWR-Loss ofAC Power) computer code that was developed to calculate the effects ofoperator actions prior to loss of injection capability after StationBlackout. Topics covered include: general features, simplifying assumptions, solutions to more important modeling problems, and model validationefforts.
BWR-LACP is a digital computer code written specifically to predictgeneral Browns Ferry thermal-hydraulic behavior following a Station Blackout event (defined in Sect. 2). The code consists of the differential andalgebraic equations of mass and energy conservation and equations of statefor the reactor vessel and containment. The code was written in the IBM
CSMP (Continuous System Modeling Program)8 Language to minimize programming and debugging time. The listing in Appendix A specifies all requiredinput parameters.
Some of the important variables calculated by the BWR-LACP code include:
1. Reactor vessel levels (above the fuel as well as in the downcomerannulus),
2. Reactor vessel pressure,3. Total injected water volume,4. Safety-relief valve (SRV) flow rates,5. Containment pressures and temperatures, and6. Suppression pool water level and temperature.
These variables may be calculated for arbitrary time periods following the loss of AC power — typically several hours. Run specific input includes:
1. Injection flow vs time or a law to represent operator control of reactor vessel level with injection flow.
2. SRV opening(s) vs time or a law to represent operator control of reactor vessel pressure with SRV opening,
3. Initialization parameters: initial time elapsed since reactor trip,initial reactor vessel pressure and level, and initial containmentpressures, temperatures, and suppression pool level.
Specific characteristics of the Station Blackout event allowed sim
plifying assumptions to be made, or required specific assumptions for adequate description of system response. These assumptions are discussed below:
1. Reactor trip at time zero: heat input from the reactor is assumed to consist of decay heat and is calculated from the ANS standard9.
2. No fuel temperature calculation: the fuel remains covered fortransients considered here and 100% of the decay heat is assumed to betransferred directly to water.
18
3. Density of steam-water mixture calculated by drift-flux correlation: the low steaming rate of decay heat operation requires a model,such as the drift-flux model10 that can allow relative slip betweensteam and water.
4. Phase separation is likely in the standpipes: the low steamingrates of decay heat operation allow phase separation to occur11 if the2-phase level is below the steam separators (in standpipes or core outletplenum).
5. Recirculation pumps trip at time zero: these pumps require ACpower for operation. During the loss of AC power transients, the water inthe recirculation piping is treated as an inactive volume.
6. Steam flow through the SRV's is calculated by ratioing the flcLow
at rated conditions:
W =Wr ^(pP)/(pr Pr)
where,
Wr = rated SRV flow at pressure Pr and density prW = SRV flow at pressure, P, and density p.
This relationship is rigorously true for choked flow of an ideal gas withconstant Cp/Cv ratio12 and should give reasonable results for chokedflow of dry steam.
7. A realistic estimate of SRV relieving capacity is calculated bytaking into account the ASME code rule which requires nominal valve flowto be only 90% of the flow measured at 103% of the nominal actuation pressure. For the two-stage Target Rock SRV's at Browns Ferry, the resultingestimate of actual valve capacity is:
Wr = 121.2 kg/s @ 7.79 MPa
(= 960,000 lb/h @ 1115.0 psig).
8. Steam flowing to RCIC and HPCI turbines is calculated by ratioing the steam flowing at rated conditions;
W = Wr*(P/Pr)*(AHr/AH)
where,
Wr = turbine flow required at conditions Pr and Allr(available enthalpy difference for isentropic expansionbetween Pr and Pexhaust) Lo PumP rated injection flow.
W = turbine flow required at conditions P and AH (isentropic expansion between P and Pexhaust) to PumP the same(rated) injection flow.
19
This equation predicts a. linear relationship between steam pressure andturbine flow providing steam pressure is high enough to establish chokedflow across the turbine nozzles. Turbine steam flow is zeroed when theturbine is tripped.
Several modeling problems had to be solved in order to adequatelysimulate behavior of the system following loss of AC power:1. Reactor vessel steaming rate,2. Reactor vessel natural circulation flow,3. Suppression pool water temperature(s), and4. Drywell air temperature.These particular problems are highlighted because of their importance indetermining the overall results and to point out approximations that weremade in formulating the solutions.
Reactor vessel steaming rate is defined here as the rate at whichsteam flows from the liquid regions in the reactor vessel into the steamvolume in the upper part of the vessel. When pressure is constant thesteaming rate equals steam production rate in the reactor core which is afunction of core inlet enthalpy, flow, and pressure. However, when pressure is changing, the steaming rate is influenced by the significantamount of water that normally exists in the steam-water mixture above thecore in the core outlet plenum, standpipes and steam separators. Thiswater adds to the steaming rate when pressure is decreasing by flashing.It is assumed to subtract from the steaming rate when pressure is increasing by condensing part of the core steam production. In modeling this effect, it was assumed that this water remains at saturation at all times.The net effect on the steaming rate was calculated using the following relationship:
Wn = Wc - —- *dP dhf M,lf w
dt dP (hK - hf)
where
Wn = net steam production rateWc = core steam production rateP = reactor vessel pressure
hf = saturated fluid enthalpyhg = saturated vapor enthalpyMj, = mass of water in core boiling region and above the core in
outlet plenum, standpipes, and steam separators.The inherent natural circulation capability of BWRs is an important
feature that should be considered even in calculation of thermal-hydraulictransients at decay heat level. The natural circulation flow path is fromthe area outside the standpipes and steam separators (referred to here asthe downcomer annulus), down through the jet pump diffusers into the lowerplenum and then into the reactor. As it flows up through the core thefluid becomes saturated and begins to boil. The steam/water mixture flowsup through the outlet plenum and standpipes to the steam separators, wherethe water is returned to the downcomer annulus and the steam flows through
20
the steam dryers to the steam dome. The natural circulation rate is dependent on a number of variables including: water level in the downcomerannulus, water-steam level above the core, core steaming rate, and densityof the water-steam mixture in and above the core. One phenomenon thatmakes the calculation more difficult is that when water level gets lowenough the recirculation of water from the steam separators ceases andwater flows from the downcomer annulus only as required to counter-balancewater lost by boil-off from the core.
The equation used to calculate core inlet flow is:
KfWci2 = Pdc*Ldc - Ptp * Ltp
where
Kf = empirically determined friction coefficientwci = core inlet tiowPdc = downcomer water densityLdc = downcomer water level (ref. to zero at bottom of active
fuel)
Jt = elevation-averaged density of water-steam mixture in andabove core
LtD = level of steam-water mixture above bottom of active fuel.
The value of the friction coefficient, Kf, was calculated from thenatural circulation curve on Fig. 3.7-1 of the BFNP FSAR by forcing agreement with the predictions of this equation at the 30% thermal power point.This procedure resulted in a reasonably good approximation of the flow atother points, as shown on Fig. 4.1. _
The rate of recirculation of water back to the downcomer annulus iscalculated as a function of the water-steam mixture level in the steamseparators:
wrecir = <wci ~ wn> * f<Ltp)
where
= water recirculated back to downcomer annulusW recirf(Ltp) = empirical function of 2-phase levelWn = net steam production rate^ci = core inlet flow.
The function f(Lt ) is 1.0 at normal level in the steam separators,decreasing to 0.0Pwhen level is below the lower edge of the steam separators.
Calculation of local temperature of suppression pool water is necessary for simulation of those events that require the assumption of extended SRV discharge into one local area of the pool. For the normal recovery accident sequences discussed in this report it was assumed that theoperators would follow the Browns Ferry procedures for reactor vesselisolation which require manual cyclic sequential operation of the SRVs
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DIRECTLY FROM FSARFIGURE 3.7-1
A POINT USED FOR BWR-LACP —INPUT PREPARATION
O RESULTING BWR-LACP FLOW —CALCULATION
20 30 40 50
PERCENT RATED CORE FLOW
60 70
Fig. 4.1 Natural circulation flow.
such that the SRV discharge is distributed throughout the pool. Therefore, it wasn't necessary to calculate local temperatures, and the modelwas programmed to calculate only the whole-pool volume-averaged temperature. It was assumed that the pool would quench all of the SRV dischargeas long as the average pool temperature is below the boiling point. TheSRV T-quencher dischargers are located at a depth below the midpoint between pool top and bottom and thus tend to "see" a temperature close to orbelow the average temperature. The possible occurrence and consequencesof non-homogeneous suppression pool temperatures are discussed in AppendixD.
During normal full power operation, plant measurements have shownthat the Browns Ferry Unit 1 drywell coolers remove about 1960 kW[6.7(10)6 Btu/h] of thermal energy from the drywell atmosphere. Thisheat comes from various sources, including heat transfer from the hot reactor vessel and steam lines and from the operation of AC powered equipment inside the drywell. During a Station Blackout event, the AC poweredequipment is lost so only about half of the heat source remains; accordingly, the initial drywell heat load following loss of AC power was takento be 980 kW [3.35(10)6 Btu/h]. This heat load Qdw is assumed to be
22
proportional to the difference between reactor coolant temperature (Trc)and drywell atmosphere temperature (Tdw). Therefore,
Qdw " 980*(Trc - Tdw)/(Trc - Tdw)0 .
To test the validity of the 980 kW heat load, an independent calculation was performed under the following assumptions:
1. Total heated surface area is 1115 m**2 (12,000 sq ft). This includes539 m**2 (5800 sq ft) for the reactor vessel, 372 m**2 (4000 sq ft)for the steam lines, and 204 m**2 (2200 sq ft) for other heated piping inside the drywell.
2. 90% of the heated surface is insulated. This is a rough estimate toaccount both for uninsulated piping and for imperfectly installedinsulation.
3. Total heat loss through insulated surfaces is 0.25 kW/m**2 (80 Btu/hsq ft). This is a good nominal estimate for MIRROR insulation whentotal hot surface to ambient temperature difference is about 220°C(400°F).
4. Radiant heat loss from uninsulated surface is figured using emmis-sivity =0.8 (i.e., unpolished surface)
5. Convective heat loss from uninsulated surface figured using the relation Nu = 0.13 (Gr*Pr)**0.3333 (as discussed in a following paragraph).The results were as follows:
Qins,total " 281 kWQunins,convective = 170 kW
Qunins,radiant = 375 kWQtotal - 826 kW
This shows that the heat losses assumed here for Unit 1 are realistic andperhaps slightly conservative. The Station Blackout analysis presented inthe Browns Ferry FSAR (response to AEC Question 14.2) assumes a much moreconservative heat loss of 2000 kW.
In addition to heat transfer from hot surfaces, the rate of steam escaping directly to the drywell must be known in order to calculate dry-well pressure and temperature. Figure 9.2-2 of the Browns Ferry FSARspecifies a total drywell drain sump input during normal power operationof 22 m**3 (5800 gallons) per day or 4 gpm as measured at the normal sumptemperature of about 38°C (100°F). Recirculation pump seals and controlrod drive flange seals are identified as the major source of this leakage.This 4 gpm leak rate is believed to be realistic for Browns Ferry Unit 1during a Station Blackout. The fraction of the leak that flashes to steamwas calculated in terms of saturation properties:
Xflash - <hf@Pr ~ nf@Pdw)/(hg@Pdw - hf@Pdw)
where hf and hg are saturated liquid and vapor enthalpies and Prand Pdw are reactor vessel and drywell pressures.
Drywell temperature was calculated using a very simple noding scheme— the model assumes that the drywell atmosphere is at a uniform temperature. This is equivalent to assuming that effective natural circulation paths will develop for heat transfer from hot surfaces, to drywell
23
atmosphere, to cold surfaces (such as the drywell liner). Such naturalcirculation probably exists in the lower part of the drywell where themain steam isolation valves and drywell coolers are located. However itis likely that temperatures in certain areas near the top of the drywellcould be much hotter than the single atmosphere temperature this modelcalculates.
Heat transfer between drywell atmosphere and liner was calculatedusing the relation
Nu = 0.13*(Gr*Pr)**0.333
where
Nu = Nusselt Number (hL/k)Gr = Grashof Number (p2ggL3AT/p:,d)Pr = Prandtl Number (u C /K) .
This formula is also used for the same purpose in the MARCH and CONTEMPT-LT codes.13>1H
Internal heat transfer resistance of the steel drywell liner is lowrelative to the resistance between liner and the atmosphere. This allowsthe use of one temperature to represent the whole thickness and it alsomeans that the entire ~364,000 kg (800,000 lbs) of drywell metal is effectively available as a heat sink during drywell heat-up. The heat sinkeffect of the drywell liner is important because without this heat sinkthe drywell atmosphere would very quickly approach primary coolant temperature after loss of the drywell coolers.
Various code comparison activities were pursued in order to test the
validity of the BWR-LACP simulation:1. Comparisons of BWR-LACP results to calculations reported in the FSAR.2. Comparison to results computed by the RELAP-IV computer code.
The first FSAR comparison was the "Loss of Auxiliary Power — AllGrid Connections" transient reported on Figs. 14.5—12 and 14.5—13of the BFNP FSAR. Plant conditions assumed by the FSAR analysis include:1. Loss of all external grid connections at time = 0.2. All pumps tripped at time = 0.3. Reactor trip and main steam isolation at about time = 2 s.
The BWR-LACP calculation was started at time equals 20 s with reactorvessel level and pressure equal to those shown for t = 20 s on FSAR Figs.14.5-12 and 14.5-13. Figure 4.2 compares the code and FSARpredictions of reactor vessel pressure and core inlet flow between 20 and50 s. During this period the core inlet flow is decreasing because steamproduction is falling off rapidly; reactor vessel pressure is being controlled between the assumed 7.59 MPa (1100 psia) and 7.41 (1075 psia) SRVset and reset points. Results are reasonably close. Figure 4.3 showsBWR-LACP and FSAR calculations of reactor vessel level from 20 to 2200 s.
During this period, the vessel level decreases slowly until the mass addition rate of the 0.0378 m**3/s (600 gpm) injection flow exceeds the massloss rate due to steam produced by decay heat. The results show similarbehavior. The FSAR minimum level is 11.76 m (463 in.) at time equals 1600s and the BWR-LACP predicted minimum level is 11.56 m (455 in.) at timeequals 1200 s.
24
ORNL-DWG 81-8605 ETD
(MPa) (psi)100
UJ
O
< 0.6 i-Xo
g 0.5 _ 75
Z3
$ 0.4rra. 50
w 0.300COUJ
> 0.2rr 25O
o 0.1<UJrr
0L 0
5 50O
UJ
CURVE A
CURVE B
L
TAKEN DIRECTLY FROM FSAR
FIGURE 14.5-12
BWR-LACP RESULTS USING FSAR
INPUT ASSUMPTIONS
±
TAKEN DIRECTLY FROM FSAR FIGURE 14.5-12
BWR-LACP RESULTS USING FSAR INPUT
ASSUMPTIONS
TIME (s)
Fig. 4.2 Comparison of FSAR and BWR-LACP results for loss ofauxiliary power transients.
50
0000
(m)
14
13
>
z 12
I"
LU 10>UJ
(in.)
570
520
470
420
370
320
ORNL-DWG 81-7858 ETD
I I |llll|—| I I Mill —LEVEL REFERENCED TO BOTTOM OF VESSEL
| I I |llll| 1 I I | Mil
DIRECTLY FROM FSAR
FIGURE 14.5-13
O CALCULATED BY BWR-LACPUSING FSAR INPUT
TOP OF ACTIVE FUEL
I I I I Mill I I I I Mill I I I I Mill I I I I INI10 5 100 2
TIME (s)
1000 2
Fig. 4.3 Comparison of FSAR and BWR-LACP results for reactorvessel level.
5 10,000
ro
26
Comparison of FSAR and BWR-LACP calculations of drywell temperaturewas performed for the "Loss of All AC Power" results shown on Fig.Q.14.2-1 of the BFNP FSAR. Plant conditions assumed by the FSAR calculations include:
1. Initial power = 100%.2. Loss of all AC power at time = 0.3. Drywell perfectly insulated on outside.4. Heat capacity of interior equipment and structures is negligible.5. Energy input from primary surfaces is initially 2000 kW.6. Heat transfer coefficient between atmosphere and wall is 11.36
w/m**2°C (2 Btu/h ft2 °F) without steam condensation.7. Drywell coolers lost at time = 0 and not restarted.8. Steam leak = 0
The conditions assumed for this analysis are of the same order as thoseassumed for the "Normal Recovery" part of the Station Blackout as discussed in this report. Comparisons of various predictions of drywellatmosphere temperature to the FSAR result are shown on Fig. 4.4. The dry-well temperature climbs rapidly during the first 5 to 10 min, before heat
<°C)
200 I-
180
160
140UJoc
Dl-
<
S5 120
100
80
60
40100
ORNL-DWG 81-8606 ETD
CURVE 3
-L
DATA TAKEN DIRECTLY FROM
FSAR FIGURE Q14.2-1 (T-ATM
WITHOUT LEAK)
CALCULATED BY BWR-LACP
USING FSAR INPUT
SAME AS CURVE 2 EXCEPT DW
LINEAR AREA LARGER BY
FACTOR OF 1.4
J_ J_
15 30 45 60 75
TIME (min)
90 105 120
Fig. 4.4 Drywell atmosphere temperature: comparison of FSAR and0RNL BWR-LACP results.
27
transfer to the drywell liner is established. After the first 10 min, thetemperature climbs much more slowly, with the metal liner acting as theheat sink. Curve 2 on Fig. 4.4 was calculated using as input to the containment model the FSAR assumptions as listed above. The resulting curveis similarly shaped, but about 17°C (30°F) above the FSAR prediction. Inorder to estimate the possible magnitude of input parameter difference,the calculation was repeated with identical input except a 40% increase indrywell liner area. Results are shown on Curve 3 and are very close tothe FSAR prediction.
A boil-off transient was selected for comparison to RELAP-IV andBWR-LACP predictions. The RELAP-IV calculation was prepared and run byIdaho National Engineering Laboratory. It starts at nominal full powerplant conditions. During the first several seconds the reactor trips andthe main steam isolation valves close. Initially, several SRVs are required to control vessel pressure, but after about 40 seconds one SRV issufficient. Injection flow is zero throughout. The BWR-LACP calculationwas intialized at 30 seconds because it is programmed only for decay heatoperation. Figure 4.5 compares the results of the calculations of reactor
(m) (in.)
60015 i—
14
13
12
>ui
11_iUiOJCOUI
> 10
— 550
500
450
400
350
300
250
500
ORNL-DWG 81-7859 ETD
STEAM/WATER ABOVE CORE
BWR-LACP
O RELAP-IV RESULTS
1000 1500
TIME (s)
2000 2500
Fig. 4.5 Comparison of RELAP-IV and BWR-LACP results for reactorvessel level.
28
vessel level. The steam/water level above the core is shown for bothcodes. Results are very similar, with RELAP predicting 33 minutes andBWR-LACP predicting 30 minutes to begin uncovering active fuel. Figure4.6 shows RELAP and BWR-LACP results for reactor steam pressure control.Results are similar although RELAP predicts a longer SRV cycle (about 60seconds) than BWR-LACP (about 38 seconds).
Results presented in this Section show that the BWR-LACP code is capable of providing reasonable predictions of overall thermal-hydraulic
(MPa) (psia)
760 r 1100
UJ<r
V)V)UJ
cca.
2<UJ
t-
7.25 L 1050 -
7.60
7.25 •— 1050 -
1000
500 1000 1500
TIME (s)
ORNL-DWG 81-7860 ETD
2000 2500
Fig. 4.6 Comparison of RELAP-IV and BWR-LACP results for reactorvessel pressure.
29
variables such as reactor vessel levels and pressure and containment temperature for extended periods after reactor trip prior to system damagedue to loss of injection capability. Features of the code that have contributed to its utility in the analysis of the Normal Recovery portion ofStation Blackout include:
1. The code is specifically designed for BWRs; therefore, parameterchanges are straightforward and easily made.
2. The code is designed for flexibility to model the plant response todifferent operator actions.
3. It is locally available and has quick turnaround time.
30
5. INSTRUMENTATION AVAILABLE DURING STATION BLACKOUT
AND NORMAL RECOVERY
For some time following a Station Blackout, the reactor vessel waterlevel can be maintained in the normal range by operation of the HPCI and/or RCIC systems, and the pressure can be maintained in any desired rangeby remote-manual relief valve operation. Restoration of AC power at anytime during this period of stable level and pressure control would permitcomplete and normal recovery without core damage. The Control Room instrumentation available and necessary to monitor the plant status duringthe period of stable control following the inception of a Station Blackoutis discussed in the following paragraphs.
1. Electrical system status. The Station Blackout would be clearlydiscernible by the loss of much of the control room indication and theloss of normal Control Room lighting. Emergency Control Room lighting isavailable. Numerous instruments would indicate the loss of voltage, amperage, and frequency on the electrical supply and distribution boards.
2. Reactor vessel water level. Two channels of Control Room instru
mentation would provide indication of the reactor vessel water level overa range between 13.41 and 14.94 m (528 and 588 in.) above the bottom ofthe vessel. This range includes the normal operating level of 14.15 m(561 in.) and extends over the upper portion of the steam separators, adistance of 4.27 to 5.79 m (14 to 19 ft) above the top of the active fuel.Mechanical Yarway indication available outside of the Control Room wouldprovide an additional range from the low point of the Control Room indication, 13.41 m (528 in.) above vessel zero down to a point 9.47 m (373 in.)above vessel zero, which is 0.33 m (13 in.) above the top of the activefuel in the core.
The level instrumentation derives the reactor water level by comparing the head of water within the downcomer region of the reactor vessel tothe head of water within a reference leg installed in the drywell. Withthe loss of the drywell coolers, the drywell ambient temperature will increase significantly, heating the reference leg water to above-normal temperatures. This will reduce the density of the water in the referenceleg, causing an error in the indicated water level, which will be toohigh. For example, if the drywell ambient temperature increases from itsnormal range of 57.2 to 63.6°C (135 to 150°F) to 171.1°C (340°F), the indicated level can be as much as 0.76 m (30 in.) too high.15 As discussed in Sect. 3, the drywell ambient temperature is expected to reachabout 148.9°C (300°F). However, the lower boundary of reactor vessellevel indication is 4.27 m (168 in.) above the top of the active fuel.Therefore, if the operator maintains the level in the indicating range, a0.76 m (30 in.) error is not significant as far as the goal of keeping thecore covered is concerned.
3. Reactor vessel pressure. Two channels of pressure instrumentation would provide indication of reactor vessel pressure over a range of8.274 MPa (0 to 1200 psig).
4. Main steamline flow. Two channels of steam flow instrumentation
with a range of 2016 kg/s (0 to 16 x 106 lbs/h) would provide verification that steam flow had ceased following Main Steam Isolation Valve(MSIV) closure.
31
5. Feedwater flow. One channel of flow instrumentation with a rangeof 1008 kg/s (0 to 8 x 106 lbs/h) for each of the three feedwater pumpswould indicate the decay of feedwater flow following loss of steam to thefeedwater turbines.
6. Neutron flux. The power range meters fail on loss of AC power.The source range and intermediate range monitors remain operational, butthe detectors for these systems are withdrawn from the reactor core duringpower operation and could not be reinserted under Station Blackout condi
tions. Nevertheless, the operator could verify the reactor scram whichoccurs when the Station Blackout is initiated by observing the decay ofthe indicated levels on these monitors.
7. Control rod position indication. The control rod position indication system remains operational so that the operator could verify thatthe control rods had inserted with the scram.
8. Relief valve position. There is no provision for direct indication of the actual position of any primary relief valve, even under normaloperating conditions. Thus the operator has no indication of automaticactuation of a specific relief valve other than the recorded tailpipe temperatures available on charts behind the control room panels, or the relief valve acoustic monitors, all of which would be inoperable under Station Blackout conditions. However, remote-manual actuation of a reliefvalve is accomplished by energizing its DC solenoid operator; lights onthe control panel for each valve indicate whether or not these solenoidsare energized, and this capability is maintained during Station Blackout.
9. RCIC instrumentation and controls. The RCIC system can be operated and monitored under Station Blackout conditions.
10. HPCI instrumentation and controls. The HPCI system can be oper
ated and monitored under Station Blackout conditions.
11. Condensate Storage Tank Level. The condensate storage tanklevel indication [range: 9.75 m (0 to 32 ft)] remains operational so thatthe operator can determine the remaining amount of water available for reactor vessel injection via the RCIC (or HPCI) system.
12. Drywell Pressure. Drywell pressure instrumentation [range:0.55 MPa (0 to 80 psia)] remains operational so that the operator can monitor the increase in drywell pressure due to the ambient temperature increase following loss of the drywell coolers.
13. Pressure Suppression Pool Water Level. Indication of torus
level [range: —0.63 to 0.63 m (—25 to +25 in.)] remains available allowing the operator to monitor the increasing pool level caused by both therelief valve blowdowns and the increasing pool temperature.
The above instrumentation is powered during Station Blackout eitherby DC power directly from the installed batteries or by AC power indirectly obtained from the battery systems. The sources of AC power duringStation Blackout are the feedwater inverter and the unit-preferred and
plant-preferred systems where single-phase 120V AC power is provided bygenerators driven by emergency battery-powered DC motors.
It is important to note that Control Room temperature instrumentationfor the drywell and the pressure suppression pool does not remain operational under Station Blackout conditions. However, the ambient temperatures at various points in the drywell would be available at an indicator-meter and a recorder mounted on panel 9-47, which is located on the backof the Control Room panels and is accessible from outside of the Control
32
Room. Also, Using a portable self-powered potentiometer, station personnel can monitor local pressure suppression chamber and drywell temperatures which are sensed by installed thermocouple elements.
Following the restoration of normal AC power, either by recovery ofthe offsite power sources or by the delayed but finally successful loadingof the onsite diesel-generators, the Instrumentation and Control buseswould be re-energized, restoring all normal Control Room instrumentationfor use in monitoring the normal recovery.
33
6. OPERATOR ACTIONS DURING STATION BLACKOUT
AND NORMAL RECOVERY
In the event of a Station Blackout, the immediate result would be aturbine trip, reactor scram, and Main Steam Isolation Valve (MSIV) closure, which would all occur automatically with no operator action; the reactor would be shut down and isolated behind the closed MSIVs within 30
seconds following the initiating loss of offsite power. Immediately following the isolation, the reactor vessel pressure would increase to thesetpoints of as many relief valves as are required to terminate the pressure increase. The affected relief valves would open to pass steam directly from the reactor vessel to a terminus beneath the water level in
the pressure suppression pool so that the released steam is condensed.The core water void collapse caused by both the scram and the pres
sure increase following MSIV closure would result in a rapid drop in reactor vessel water level to some point below the lower limit of level indication available in the Control Room under Station Blackout conditions
[13.41 m (528 in.) above vessel zero]. Standard Browns Ferry EmergencyOperating Instructions for immediate action following reactor scram andMSIV closure call for the operator to manually initiate both the HPCI andthe RCIC systems. The combined injection flow of 0.353 m3/s (5600 GPM)would rapidly restore the level into the indicating range and the operatorwould secure the HPCI system when the indicated level reached a pointequivalent to about 13.72 m (540 in.) above vessel zero. The level wouldthen continue to increase due to the remaining 0.038 m /s (600 GPM) ofRCIC system flow and the heatup and expansion of the injected water, butat a slower rate.
For pressure control, standard procedure is for the operator toremote-manually operate the primary relief valves as necessary to cyclethe reactor vessel pressure between 7.69 and 6.31 MPa (1100 and 900 psig).This prevents the automatic actuation of the relief valves [lowest set-point: 7.72 MPa (1105 psig)], and since the pressure reduction followingautomatic actuation is only 0.34 MPa (50 psi), greatly reduces the totalnumber of relief valve actuations. The purpose is to decrease the opportunity for a relief valve to stick open, and to allow the operator to alternate the relief valves which are in effect passing the decay heatenergy to the pressure suppression pool; this more evenly distributes thepool heatup.
Throughout the period of Station Blackout, the operator would operatethe RCIC system intermittently as necessary to maintain the reactor vessel
level in the indicating range, i.e., between 13.41 and 14.94 m (528 and588 in.) above vessel zero. Thus, it is expected that the operator wouldturn on the RCIC system at an indicated level of about 13.61 m (538 in.)and turn the system off at about 14.68 m (578 in.). The periods of RCICsystem operation would become less frequent as the decay heat intensitysubsides during a prolonged Station Blackout; after the first hour, theoperator would find it necessary to run the RCIC system for only aboutone-third of the time. This is illustrated in Sect. 7.
As previously discussed, the operator would control reactor vesselpressure by remote-manual operation of the relief valves. It is importantthat the operator take action to begin to lower the reactor vessel pressure to about 0.79 MPa (100 psig) within 60 min following the inception of
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36
7. COMPUTER PREDICTION OF THERMAL-HYDRAULIC
PARAMETERS FOR NORMAL RECOVERY
7.1 Introduction
The purpose of this chapter is to present the results of BWR-LACPcalculations (see Sect. 4) of system behavior for two different portionsof the Station Blackout:1. Normal Recovery: DC power from the unit battery remains available
during this period; therefore RCIC and HPCI injection flows are assumedavailable throughout, and2. Loss of 250 vdc Batteries: The batteries are assumed to fail after
four hours, causing failure of RCIC and HPCI injection, ultimately leadingto uncovering of the core about eight hours after the blackout.
The normal recovery portion of the Station Blackout is discussed indetail in Sects. 3 and 6. Operator actions assumed here are consistentwith those sections. Results were calculated to five hours* for normalrecovery in order to provide some overlap with the MARCH code severe accident calculations (see Sects. 9 and 10), which assume that the boil-offbegins after four hours from a fully pressurized condition. The Loss of250 vdc Batteries results are given in this section in order to provide anestimate of system behavior possible after loss of dc power if the plantis depressurized during the normal recovery period.
7.2 Conclusions
Major conclusions drawn from the calculated results are given below,succeeded by a detailed discussion of code input assumptions and transientresults.
7.2.1 Normal Recovery — Conclusions
If power were recovered within five hours of the inception of theStation Blackout, a normal recovery would be possible. System parametersare within acceptable ranges after five hours:1. The 250 vdc batteries, by assumption, last the full five hours.2. Reactor vessel level is within the normal control range, about 5.08 m
(200 in.) above the top of active fuel.3. Reactor vessel pressure is being controlled at about 0.69 MPa (100
psia).4. About 354 m3 (93,500 gal.) of water have been pumped from the Con
densate Storage Tanks, which had an assured capacity of 511 m3(135,000 gal.) before the blackout.
5. Suppression pool temperature is about 82°C (180°F), but this shouldnot be a problem for the T-quencher type of SRV discharge header piping.
6. Containment pressures are elevated to about 0.17 MPa (25 psia), wellbelow the 0.53 MPa (76.5 psia) design pressure.
*These results are extended to seven hours in Appendix G.
37
7. Drywell atmosphere temperature is below the 138°C (281°F) design temperature.
7.2.2 Loss of 250 vdc Batteries — Conclusions
The results of this transient show very clearly how fuel damage ispostponed because the reactor vessel was depressurized early in the Sta-ttion Blackout:
1. The repressurization time after loss of the 250 vdc batteries isgreater than one hour, during which time there is no significantcoolant loss from the reactor vessel.
2. When the boil-off does begin, it takes a much longer time to uncoverfuel because of the higher starting inventory of water.Although fuel damage is significantly delayed, the ability to avoid
ultimate fuel damage is compromised because of the elevated drywell temperature experienced after loss of the 250 vdc batteries. As discussed inSect. 3, a containment temperature of 149°C (300°F) would not prevent normal recovery. This temperature is reached about 40 min. after loss of thebatteries. At about four hours after the battery loss, the fuel is beginning to be uncovered, and the drywell temperature is above 191°C (375°F).This elevated temperature may cause failure of the drywell electrical penetrations and may fail the solenoid operators necessary for operation ofthe SRVs, inner isolation valves, and containment cooler dampers (whichfail closed on loss of AC power). Even if electrical power were fully restored at this point, considerable operator ingenuity would be required toeffect a normal recovery if the MSIV and SRV solenoid operators havefailed.
In addition, the chances for a normal recovery are compromised by theelevated suppression pool temperatures experienced at the end of the transient. Condensation oscillation during SRV discharge (see Appendix D),which can damage the suppression pool pressure boundary, becomes morelikely at higher pool temperatures. Steam discharge without oscillationfrom the Browns Ferry T-quencher type discharge piping is assured up to88°C (190°F), but average pool temperature at the end of the Loss of 250vdc Batteries calculation is above 93°C (200°F).
7.3 Normal Recovery
7.3.1 Normal Recovery — Assumptions
Assumptions specific to programming and input preparation for thenormal recovery calculation include:
1. The calculation begins 30 s following the Station Blackout initiationwith the reactor tripped and the main steam isolation valves closed.Values for system parameters at the 30-s point are taken from the results reported by Sect. 14.5.4.4 of the Browns Ferry FSAR. The calculation ends at the 5 hour point.
2. The RCIC system is used in an on-off mode to control vessel level between 13.7 m (538 in.) and 14.7 m (578 in.) above vessel zero. The
38
HPCI system is actuated only when level is more than 0.25 m (10 in.)below the 13.4-m (528-in.) lower limit of control room indicationunder Station Blackout conditions. Suction for both systems is fromthe condensate storage tank (CST). The operators actuate HPCI andRCIC 1 min. after the blackout.
3. Following 1 min. of automatic SRV actuation, reactor pressure is controlled by remote-manual actuation of the SRVs. Control range during the first hour is 7.52 MPa (1090 psia) to 6.31 MPa (915 psia).After 1 hour the reactor is depressurized to about 0.79 MPa (115psia) at a rate consistent with the 55.6°C/h (100°F/h) cooldown ratelimit. After depressurization, pressure is controlled between 0.62MPa (90 psia) and 0.86 MPa (125 psia).Manual sequential SRV action is used to spread SRV discharge aroundthe suppression pool and thereby avoid localized heating of the pool.
4. The assumed reactor system leakage into the drywell is 0.0089 m /s(4 gpm).
7.3.2 Normal Recovery — Results
Results for normal recovery are shown on Figs. 7.1 through 7.9.Each system variable is discussed below.
7.3.2.1 Reactor vessel pressure. Figure 7.1 shows reactor vesselpressure. During the first minute, reactor pressure cycles between 7.72MPa (1120 psia) and 7.38 MPa (1070 psia) on automatic SRV actuation.After 1 min. the operator opens one SRV to lower pressure to 6.31 MPa (915psia). Since the HPCI is running, pressure continues to decrease to below6.21 MPa (900 psia) until the HPCI is shut down. During the first hour asingle SRV is cycled open and shut seven times to keep pressure in the desired range.
After one hour, the depressurization begins. Initially, intermittentoperation of one SRV is sufficient to meet the target depressurizationrate [i.e., 55.6°C (100°F)/h], As pressure decreases, the rate of decrease begins to slow until finally an additional SRV is opened. Forchoked flow of dry steam, the mass flow rate is linearly proportional toreactor pressure . Therefore an extra valve is required at the lower pressure. When depressurization is complete, the operator controls pressurebetween 0.62 and 0.86 MPa (90 and 125 psia).
After depressurization, the steam flow through one SRV is nearlyequal to the core steaming rate, resulting in much slower pressure change.Under these conditions, the most rapid pressure change is caused by theRCIC system. When RCIC comes on the 600 gpm injection of cold water tendsto lower the core steaming rate, causing pressure to decrease. This effect can be seen in Fig. 7.1. At 256 min., the RCIC comes on, reducingpressure until the open SRV is shut. When the RCIC is shut off 23 min.later, pressure increases rapidly until the SRV is reopened.
Steam flow from the reactor vessel is shown in Fig. 7.2. The largespikes are due to SRV actuation. Also included in the total steam floware the smaller amounts of steam flowing to the RCIC and HPCI turbineswhen they are running.
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41
7.3.2.2 Reactor vessel level — normal recovery. Figure 7.3 showsvessel water level (distance above vessel zero) in the region outside thecore outlet plenum, standpipes, and steam separators. This is the levelmeasured by vessel level instrumentation and available in the controlroom. As discussed in Sect. 4, there is a corresponding level of two-phase mixture in and above the core. For an undamaged core in a mildtransient such as Station Blackout, the level of steam/water mixture inand above the core will be higher than the downcomer level whenever downcomer level is below the top of the steam separators.
Downcomer level is initialized at 12.7 m (500 in.) and, at first, decreases rapidly until the operator initiates injection via the RCIC andHPCI systems. When level recovers to 13.7 m (540 in.), the operator shutsoff the HPCI system, and when level reaches 14.7 m (578 in.), the RCICinjection is shut off. Level continues to increase due to the continuingheatup of the large quantity of cold water injected by the RCIC and HPCIsystems and due to formation of additional voids in and above the core asthe steaming rate increases. The level peaks at 15.2 m (600 in.). This isabout 0.30 m (12 in.) above the top of the range of level indicationavailable in the control room during Station Blackout.
Control Room level indication during a Station Blackout is limited toa range of 3.66 m (0 to 60 in.) with instrument zero corresponding to13.41 m (528 in.) above vessel zero, which is the bottom of the reactorvessel. The basic cycle of RCIC actuation by the operator when levelreaches an indicated 0.25 m (10 in.) and shut-off when level reaches anindicated 1.27 m (50 in.) is repeated eight times on Fig. 7.3. The finestructure is caused by action of the SRVs. When the SRV opens, vessellevel first swells due to increased steaming, but then begins to decreasedue to inventory loss. When the SRV shuts, level at first continues todecrease due to decrease in steaming rate, but then begins to recover dueto heatup of subcooled liquid still within the downcomer.
Injection flow from the CST to the reactor vessel is shown in Fig.7.4. The operator controls the RCIC and HPCI systems in an off-on mode inorder to maintain vessel level within the desired control band as speci
fied in Para. 7.3.1.2. The HPCI system is only run during the initial fewminutes. The periods of RCIC system operation become less frequent as thedecay heat subsides with time. Total injected flow (i.e., total amount ofwater taken from the CST) is shown in Fig. 7.5. The injected total in
creases rapidly at first during the brief period of combined HPCI and RCICoperation; the increase is much slower during the subsequent intermittentperiods of RCIC operation.
7.3.2.3 Suppression pool level — normal recovery. When steam flowsfrom the reactor vessel to the suppression pool, it is condensed by thecolder pool water. Suppression pool level increases both because of theextra mass of the condensed steam and because of the temperature increaseof the original pool water. Suppression pool level is shown in Fig. 7.6,referenced to instrument zero of the torus. The torus has a 9.45 m (31ft) inside diameter and instrument zero is 0.1 m (4 in.) below the mid-plane. (A zero level indicates the torus is about half filled withwater.) The pool level is normally maintained between -0.051 and -0.152 m
(—2 and —6 in.) indicated; an initial pool level of -0.102 m (—4 in.)was assumed for this analysis. The pool condenses steam both from the
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HPCI OFF
RCIC OFF
30—r
60—r-
90 120
RCIC
ON OFF ON OFF
{ \ { \
150
TIME (min)
—T-
180 210
ORNL-DWG 81-8492 ETD
240 270 300
Fig. 7.4 Normal recovery after station blackout - injection flow.
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INJECTION BY COMBINED HPCIAND RCIC SYSTEMS
60 90 120 150
TIME (min)
ORNL-DWG 81-8493 ETD
RCIC
OFF-
180 210 240 270 300
Fig. 7.5 Normal recovery after station blackout - total injectedflow.
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ORNL-DWG 81-8494 ETD
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PERIODS OF RAPID INCREASEDUE TO QUENCHING OF SRVSTEAM FLOW
150
TIME (min)
210 240
Fig. 7.6 Normal recovery after station blackout - suppressionpool level.
300
46
SRVs and the RCIC and HPCI turbine exhaust. The effect of RCIC turbineexhaust shows up in Fig. 7.6 as a very slow background rate of increaseagainst the dominant effect of the SRVs.
7.3.2.4 Suppression chamber temperatures — normal recovery. Suppression chamber temperatures are shown in Fig. 7.7. For all practicalpurposes, the average pool temperature responds solely to condensation ofreactor vessel steam. Therefore, this curve has almost the same shape asthe pool level curve. The pool atmosphere can be influenced by mass andheat transfer from the pool water and by mass transfer from the drywellatmosphere.
During the first 30 min. the suppression chamber atmospheric temperature increases rapidly because of mixing with the inflow of hotter gasesfrom the expanding drywell atmosphere. Drywell temperature (Fig. 7.8) increases rapidly at first due to loss of the drywell coolers. As a result,the drywell atmosphere expands into the suppression chamber atmospherethrough the pool water via the eight vent pipes, the single ring header,and the 96 downcomer pipes.
In the long term, mass transfer from the drywell ceases and the suppression chamber temperature continues to increase due to heat transferand evaporation from the suppression pool surface.
7.3.2.5 Containment pressures — normal recovery. The suppressionchamber and drywell pressures are shown in Fig. 7.9. The drywell and suppression chamber atmospheres are coupled in the following manner:1. When suppression chamber pressure exceeds drywell pressure by 0.00345
MPa (0.5 psi), the vacuum breaker valves will open and allow flow ofsuppression chamber atmosphere into the drywell.
2. When drywell pressure exceeds suppression chamber pressure by morethan about 0.0121 MPa (1.75 psi), this will clear the 1.22 m (4 ft)of water from the downcomer pipes and allow drywell atmosphere tobubble up through the pool water into the suppression chamber.Immediately after the Station Blackout, drywell pressure increases
rapidly due to the concurrent rapid heatup of the drywell atmosphere.Even after the drywell atmosphere reaches its peak temperature and startsdown, the drywell pressure continues to increase under the influence ofthe assumed 0.0089 m3/s (4 gpm) leak of hot reactor coolant, some ofwhich flashes to steam. Throughout the first 2 hours, conditions in thedrywell bring the suppression chamber pressure up at about the same rateby expansion (via the downcomers) into the suppression chamber. After thefirst two hours, suppression chamber pressure is increasing faster thandictated by the drywell because of increasing rates of heat transfer andevaporation from the overheated suppression pool. After about 190 min.the pressure is high enough to open the vacuum breakers and cause expansion of suppression chamber atmosphere back into the drywell.
7.3.2.6 Drywell temperature — normal recovery. Drywell atmospheretemperature response is shown in Fig. 7.8. The initial rate of increaseis very rapid because the drywell coolers are lost at time zero, and abouthalf of the 100% power drywell heat load is assumed to remain. The rateof increase becomes much slower when drywell temperature is high enough
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Fig. 7.8 Normal recovery after station blackout - drywelltemperatures.
270 300
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Fig. 7.9 Normal recovery after station blackout - containmentpressures.
270 300
50
to establish natural convection heat transfer to the drywell liner (seediscussion in Sect. 4). Drywell temperature begins decreasing after about60 min. because a reactor coolant system (RCS) depressurization is begunat this point. The depressurization reduces saturation temperature withinthe RCS, hence surface temperatures of the reactor vessel and piping.
7.4 Loss of 250 vdc Batteries
7.4.1 Loss of 250 vdc Batteries — Assumptions
Assumptions specific to programming and input preparation for theLoss of 250 vdc Batteries calculation include:
1. When the batteries are lost, injection capability is lost; in addition, the SRVs can be opened only by automatic actuation.
2. Under automatic actuation, the lowest set SRV will open at about 7.72MPa (1120 psia) and close at about 7.38 MPa (1070 psia).
3. The calculation begins 4 h after the Station Blackout and proceeds toabout 8 h.
4. Initial values of system parameters, except reactor vessel level, areequal to those calculated at the 4 h point in the normal recovery results (see Figs. 7.1 through 7.9). Reactor vessel level is assumedto begin at 14.2 m (558 in.), the midpoint of the control range assumed for the normal recovery calculation.
7.4.2 Loss of 250 vdc Batteries — Results
Results for the Loss of 250 vdc Batteries case are shown in Figs.7.10 through 7.16. Each system variable is discussed below.
7.4.2.1 Reactor vessel pressure. Reactor vessel pressure is shownin Fig. 7.10. During the first 90 min., pressure climbs slowly from the0.86 MPa (125 psia) initial value to the 7.72 MPa (1120 psia) actuationpoint of the lowest set SRV. The rate of increase is slow because the decay heat after 4 hours is small compared to the thermal inertia of liquidwater within the vessel.
After repressurization, the pressure is controlled between 7.72 MPa(1120 psia) and 7.38 MPa (1070 psia) by one SRV, actuating automatically.As shown in Fig. 7.11, no steam flows from the vessel until pressure reaches the SRV automatic actuation setpoint.
7.4.2.2 Reactor vessel level. Reactor vessel level is shown in Fig.7.12. No injection is provided at any time in this transient, since DCpower is assumed lost.
The initial response of level is a brief but rapid decrease caused byclosing of the previously open SRV at time zero. Within several minuteslevel begins slowly rising and peaks at about 15.88 m (625 in.). Levelswells during repressurization because the large mass of water in the vessel is heated from about 165°C (330°F) to about 288°C (550°F) with about a25% increase in specific volume, and there is no coolant loss (other thanthe assumed leakage) during repressurization.
360 390 420
TIME (min)
ORNL-DWG 81-8498 ETD
Fig. 7.10 Loss of 250 VDC batteries - reactor vessel pressure.
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AUTOMATIC ACTUATION OFLOWEST SET SRV
330 360 390 420
TIME (min)
450
ORNL-DWG 81-8499 ETD
CALCULATIONS
STOPPED
480 51C 540
Fig. 7.11 Loss of 250 VDC batteries - vessel steam flow.
ro
240 270 300 330 360 390 420
TIME (min)
ORNL-DWG 81-8500 ETD
450 480 510 540
Fig. 7.12 Loss of 250 VDC batteries - vessel level.
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54
After repressurization, the SRVs begin to discharge steam so leveldecreases throughout the remainder of the transient. Fuel is beginning tobe uncovered at the end of the transient.
7.4.2.3 Suppression pool level. The suppression pool level is shownin Fig. 7.13. During the first 90 min., there is no SRV discharge, sopool level is constant. After SRV discharge resumes, the pool level begins increasing steadily.
7.4.2.4 Suppression chamber temperatures. Figure 7.14 shows suppression pool and suppression chamber atmosphere temperature. Pool temperature, like pool level, is constant for the first 90 min. and then begins a steady increase. Atmosphere temperature continues to rise duringthe first hour due to heat transfer and evaporation from the pool surface. Atmosphere temperature rise accelerates slightly between the firstand second hours due to an influx from the expanding drywell atmosphere.After about 2 hours, the mass interchange ceases, and the suppressionchamber atmosphere temperature is responding solely to the increasing pooltemperature.
7.4.2.5 Containment pressures. Figure 7.15 shows containment pressures. During the first 90 min., drywell pressure increases rapidly dueto heatup of the drywell atmosphere during repressurization. This bringsdrywell pressure from slightly below suppression chamber pressure to about0.0125 MPa (1.8 psi) above, allowing some drywell atmosphere to expandinto the suppression chamber. After about 2 hours, the drywell and suppression chamber pressures are increasing independently at about the samerate, with little mass interchange.
7.4.2.6 Drywell atmosphere temperature. Drywell atmosphere temperature is shown in Fig. 7.16. During the repressurization, the coolant temperature within the reactor vessel increases by over 111 C° (200 F°),thereby increasing the drywell heat source. Atmosphere temperature risesto new quasi-equilibrium values, with increased natural circulation heattransfer from the atmosphere to the drywell liner. After repressurization, the coolant temperature becomes approximately constant, and the temperature of the drywell atmosphere rises more slowly at a rate limited bythe thermal inertia of the liner.
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TIME (min)
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Fig. 7.13 Loss of 250 VDC batteries - suppression pool level.
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ORNL-DWG 81-8502 ETD
-POOL TEMPERATURE INCREASES WHENSRV DISCHARGE RESUMES
HEATUP ACCELERATES DUE TO INFLUX"OF HOTTER DRYWELL ATMOSPHEREVIA DOWNCOMERS
8-240 270 300 330 360 390 420
TIME (min)
450 480
Fig. 7.14 Loss of 250 VDC batteries - suppression chambertemperatures.
510 540
LnON
ORNL-DWG 81-8503 ETD
510 540
Fig. 7.15 Loss of 250 VDC batteries - containment pressures.
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ORNL-DWG 81-8504 ETD
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Fig. 7.16 Loss of 250 VDC batteries - drywell atmospheretemperature.
510 540
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59
8. FAILURES LEADING TO A SEVERE ACCIDENT
As discussed in Sect. 3, control can be maintained over reactor vessel pressure and water level at Browns Ferry Unit 1 for a significantperiod of time during a Station Blackout provided the installed equipmentoperates as designed and the operator takes the required actions. Thefirst objective of this Section is to briefly discuss the other sequenceswhich might occur given significant secondary independent equipment fail
ures or improper operator actions. Subsequently, the equipment failureswhich have the potential to directly convert the relatively stable periodof level and pressure control of the isolated reactor vessel during theearly stages of a Station Blackout into a Severe Accident will be discussed in detail.
The several sequences which are considered most probable in a StationBlackout during the period in which DC power remains available are displayed in the "event tree" of Fig. 8.1. Each of these sequences is discussed below.
Sequence 1; Following the complete loss of AC power, the operatortakes remote-manual control of the RCIC system and maintains the reactorvessel water level within the limits of the available Control Room indica
tion. Since the RCIC System is being used for level control, the statusof the HPCI system is not a factor in this sequence. The operator alsocontrols the reactor vessel pressure by remote-manual operation of theprimary relief valves and takes action to begin depressurization of thereactor vessel within one hour; this lowers the temperature of the saturated liquid within the vessel and thereby decreases the driving potentialfor heat transfer into the drywell atmosphere. As a result of this depressurization, the maximum average ambient temperature in the drywell islimited to 148.9°C (300°F), as discussed in Sect. 3. The operator will beable to maintain reactor vessel level and pressure control for as long asDC power is available from the Unit Battery. This is considered to be themost realistic sequence should a Station Blackout occur, and served as the
basis for the plots of thermohydraulic parameters as functions of timewhich were presented in Sect. 7. This sequence will also serve as thestarting point for the degraded accident behavior discussed in Section 9in which the Station Blackout develops into a Severe Accident when DC
power is lost so that cooling water can no longer be injected into the reactor vessel.
Sequence 2; This sequence differs from Sequence 1 only in that theoperator does not act to depressurize the reactor vessel. The averageambient drywell temperature continues to increase beyond the one-hourpoint, reaching 176.7°C (350°F) about four hours after the inception ofthe Station Blackout (cf. Curve A of Fig. 3.3). This excessive temperature existing over a period of several hours can cause serious damage to
the equipment located within the drywell and a breaching of the primarycontainment due to failure of the drywell electrical penetration assemblyseals. If the DC solenoid operators for the primary relief valves failbecause of the excessive temperature, the operator will lose remote-manualrelief valve control. If the solenoid operators for the inboard MainSteam Isolation Valves also fail, the operator will be unable to depres
surize the reactor vessel through the primary relief valves even after AC
ORNL-DWG 81-1544A ETD
COMPLETE LOSS
OF AC POWER
OPERATOR ACTIONTO CONTROL LEVELAND RCIC OPERABLE
HPCI OPERABLEOPERATOR ACTIONTO DEPRESSURIZE
WITHIN 1 h
HPCI SUCTION SHIFTS TO PSP AT
PSP LEVEL = +7"
Fig. 8.1 Event tree for initial phase of a station blackout inwhich DC power remains available.
©
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61
power is restored. There is no reason to accept these risks to the dry-well integrity and installed equipment; the operator should take action todepressurize, as in Sequence 1.
Sequence 3; Either the RCIC system is inoperable or the operatordoes not choose to employ it for long-term reactor vessel level control.The HPCI System is operable, and either automatically functions as necessary to cycle the reactor vessel level between 12.1 and 14.8 m (476 and582 in.) above vessel zero, or is remote-manually controlled by the operator to maintain the water level within the indicating range, 13.4 to 14.9m (528 to 588 in.). The operator controls the reactor vessel pressure byremote-manual operation of the relief valves and takes action to begin de-pressuration within one hour. Up to this point, this sequence is similarto Sequence 1, except that the injection of condensate storage tank waterinto the reactor vessel is by the HPCI System instead of the RCIC System.
When the indicated water level in the pressure suppression pool increases to 0.18 m (+7 in.), the HPCI pump suction is automatically shiftedfrom the condensate storage tank to the pressure suppression pool. Depending on the initial suppression pool level, the necessary amount ofwater to cause this shift will have been transferred from the reactor vessel via the primary relief valves at some time between two and four hoursafter the inception of a Station Blackout. Since the lubricating oil forthe HPCI turbine is cooled by the water being pumped, and the average suppression pool temperature will be about 71.1°C (160°F) at this point, anearly failure of the HPCI system is threatened.
Sequence 4; This sequence is the same as Sequence 3, except that theoperator has recognized the difficulties with the elevated suppressionpool temperature and has taken action to prevent the shift in HPCI pumpsuction. The operator will be able to maintain reactor vessel level andpressure control for as long as DC power remains available from the UnitBattery. The continuing increase in suppression pool level will not causeany significant problems during the remaining period in which DC powerwould remain available.
It should be noted that pressure suppression pool temperature instrumentation would be inoperable during a Station Blackout. Operator training and the Station Blackout Emergency Operating Instruction (which doesnot now exist) must provide for operator understanding of the need to prevent a shift of the HPCI pump suction to the overheated pressure suppres
sion pool.Sequence 5: This sequence represents the case of Station Blackout
with no operator action and no secondary independent equipment failures.The HPCI System automatically cycles the reactor vessel water level between the HPCI system initiation point of 12.1 m (476 in.) and the HPCIturbine trip point of 14.8 m (582 in.) above vessel zero. Over the longterm, one primary relief valve would repeatedly operate as necessary tomaintain the reactor vessel pressure between about 7.722 and 7.377 MPa(1105 and 1055 psig).
There are three problem areas with this case of no operator action:1. the difficulties with high drywell ambient temperature as discussed in
connection with Sequence 2,2. the increased probability of loss of injection capability after the
undesirable shift of the HPCI pump suction to the overheated pressuresuppression pool as discussed in connection with Sequence 3, and
62
3. the decrease in suppression pool effectiveness caused by the highlylocalized heating of the pool water surrounding the discharge of theone repeatedly operating relief valve. The possible consequences arediscussed in Appendix D.Sequence 6: This sequence differs from Sequence 5 only in that the
HPCI pump suction is maintained on the condensate storage tank, either byoperator action, or by fortuitous failure of the pertinent HPCI systemlogic. This reduces the probability of HPCI system failure due to inadequate lube oil cooling during the period of Station Blackout in which DCpower is still available. The difficulties with high drywell ambient temperature and localized suppression pool heating as discussed in connection with Sequence 5 remain.
Sequence 7: The RCIC system is not available because of equipmentfailure, or lack of operator action. The HPCI system has failed becauseof a system malfunction. The represents the worst case of loss of injection capability while DC power remains available. A boiloff of the initial reactor vessel water inventory begins immediately after the completeloss of AC power, leading to a relatively quick core uncovery and subsequent meltdown. In this sequence, the core uncovery is hastened by a depressurization of the reactor vessel either by means of a stuck-open relief valve or because of inappropriate operator action; the depressurization increases the rate of loss of the irreplaceable reactor vessel waterinventory.
Sequence 8: This sequence is the same as Sequence 7, except thatthere is no reactor vessel depressurization. The uncovery of the corewould begin about one-half hour after the loss of AC power. The plant response to Sequences 7 and 8 will be discussed further in Section 9.
The preceding discussion of alternate sequences again shows that aStation Blackout at the Browns Ferry Nuclear Plant will not evolve into aSevere Accident as long as reactor vessel water injection capability ismaintained. The remainder of this section contains a discussion of the
methods by which this injection capability might be lost. There are threemain areas of concern for the viability of the injection systems: StationBlackout induced direct failure of the HPCI and RCIC systems, a stuck-openrelief valve which might reduce reactor vessel pressure below that necessary for HPCI and RCIC turbine operation, and the ultimate loss of DCpower. Each of these will be discussed in turn.
8.1 Induced Failure of the HPCI and RCIC Systems
The design and the principles of operation of the HPCI and RCIC systems are discussed in Appendices E and F, respectively. Each of thesesystems has the capacity to provide the necessary reactor vessel waterlevel control during a Station Blackout; both must fail to cause a totalloss of injection capability.
The failure modes considered in this Section will be limited to those
induced by the conditions of a Station Blackout. Independent secondaryfailures, whose occurrence during a Station Blackout would be purely coincidental, have been considered in other studies16 and will not be discussed here.
63
The first threat to the continued operation of the HPCI and RCIC systems under Station Blackout conditions is due to the loss of the ReactorBuilding cooling and ventilation systems. The major components of theHPCI and RCIC systems are located in the basement of the Reactor Building,where the average temperatures during a Station Blackout would be significantly affected by the temperature of the pressure suppression pool. TheHPCI and RCIC systems are designed for continuous operation at an ambienttemperature of 64.4°C (148°F). As was shown in Section 7, the averagepressure suppression pool temperature would reach 71.1°C (160°F) afterabout three hours under Station Blackout conditions. However, it is expected that the ambient temperature in the vicinity of the HPCI and RCICsystems would significantly lag the increasing suppression pool temperature; it is estimated that the design temperature of 64.4°C (148°F) wouldnot be exceeded for at least four hours. In any event, exceeding the design ambient temperature would not assure failure of the HPCI and RCICsystems.
A more probable cause of HPCI or RCIC system failure is the automaticisolation of these systems due to the tripping of the temperature sensingcircuits designed to detect steam leaks in the system piping. As described in Appendices E and F, each of these systems has a set of four triplogics with four temperature sensors per logic. The 16 sensors are physically arranged in four groups placed near the HPCI or RCIC equipment, withthe four sensors in each group arranged in a one-out-of-two taken twicetrip logic. If the trip setpoint of 93.3°C (200°F) is reached, an automatic closure signal is sent to the RCIC or HPCI primary containment inboard and outboard steam supply valves, and the system turbine is tripped.Since the inboard steam supply valve for each of these systems is AC-motor-operated, it would not close, but the system would be effectivelyisolated due to closure of the outboard (DC-motor-operated) steam supplyvalve.
As shown in Section 7, it is expected that the RCIC turbine would beoperated only intermittently during a Station Blackout, while the HPCIsystem would serve only as a backup in the event of RCIC system failure.Also, the RCIC equipment is not located in a closely confined space, andsome natural convection within the reactor building will certainly occur.Nevertheless, it is conceivable that the local temperature in the vicinityof the steam leak detection sensors could reach 93.3°C (200°F) during RCICsystem operation after the average space temperature has increased to over60°C (140CF). If this occurs, the resulting RCIC system isolation signalcan be overriden in the auxiliary instrumentation room and the steam supply valves reopened and the turbine trip reset. Thereafter, the HPCI system and the RCIC system, which are located on opposite sides of the suppression pool torus, might be alternately used for reactor vesssel levelcontrol so as to reduce the heatup in the vicinity of the RCIC system.*Thus, the tripping of the temperature sensing circuits is not expected tolead to a total loss of injection capability.
*The HPCI room has not been designed to permit natural circulationand the HPCI system has much more surface area for heat transfer Into itssurrounding space. Therefore, it is expected that the heatup associatedwith HPCI operation would be more severe than that associated with RCICoperation.
64
Another challenge to the viability of the HPCI and RCIC systems wouldbe posed by the loss of control air pressure during a prolonged StationBlackout. The plant control air system is supplied from three air receivers with a combined capacity of 22.60 m3 (798.0 ft3). Under StationBlackout conditions, the air compressors which normally run intermittentlyas necessary to maintain the receiver pressure in the range of 0.66 to0.86 MPa (80 to 105 psig) would be inoperable; the stored inventory ofpressurized air would gradually be lost due to valve operations and leakage.
Control air is used in the HPCI and RCIC systems to operate the steamsupply line drain isolation valves. In each of these systems, the twoprimary containment isolation valves in the steam line to the turbine arenormally open so that the steam supply piping is kept at elevated temperatures; this permits rapid turbine startup on demand. The water formed bythe steam which condenses in this line when the turbine is not operating
is removed to the main condenser via a thermostatic steam trap. The twoair-operated drain isolation valves close to prevent the flow of waterfrom the steam trap to the main condenser when the primary containment isto be isolated; the closing signal is automatically generated on low reactor water level [12.10m (476.5 in.) above vessel zero].
The steam supply line drain isolation valves fail closed on loss ofcontrol air, which would eventually occur during a prolonged StationBlackout. This should not cause serious difficulties for the RCIC system
which would be run intermittantly so that the accumulation of water in thesteam supply piping between runs would not be excessive. However, a substantial amount of water would collect in the HPCI steam supply piping ifthis system is not operated over a long period following the loss of control air pressure. Subsequent initiation of the HPCI system might causeturbine damage. However, it should be noted that the HPCI and RCIC turbines are two-stage, non-condensing Terry turbines, which are of very rugged construction and are designed for use under emergency conditions, andthe operator could take action to run the HPCI turbine for short periodsto clear the lines of water as necessary. Thus the loss of control airduring a prolonged Station Blackout should not lead to a total loss of injection capability.
A third challenge to the viability of the water injection systemsshould affect the HPCI system only. This challenge would occur because ofthe HPCI system logic which provides for an automatic shifting of the HPCIpump suction from the condensate storage tank to the pressure suppressionpool when the indicated pressure suppression pool level reaches 0.18 m (+7in.). As previously discussed, this would occur between two and fourhours after the inception of a Station Blackout, when the pressure suppression pool temperature has increased to about 71.1°C (160°F). Sincethe turbine lubricating oil is cooled by the water being pumped, and theoil cooler is designed for a maximum inlet water temperature of 60.0°C(140°F), the oil would become overheated, possibly leading to failure ofthe turbine bearings. This can be avoided if the operator takes action tomaintain suction on the condensate storage tank for any HPCI system operation. There is an ample supply of condensate storage tank water avail
able for injection, and the continuing increase in pressure suppressionpool level will in no way threaten the continued operation of the
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Table 8.1 Unit preferred system loads
1. Containment Isolation panel
2. Feedwater Control panels
3. Reactor Manual Control panel
4. Drywell Ventilation and Reactor Building ClosedCooling Water System Control panel
5. Feedpump Turbine Control panels
6. Rod Position Information System
7. EHC Control Unit
8. Unit Computer and Rodworth Minimizer panels
9. Reactor Water Cleanup panel
Table 8.2 Major unit 250-volt DC loads
1. Turbine Building Distribution Board
2. 480-volt Shutdown Board Control Power
3. Emergency DC Lighting
4. Unit Preferred AC Motor-Generator
5. Circuit Breaker Board 9-9 (Feedpump TurbineControls)
6. Reactor Motor-Operated Valve Boards
a) Primary Relief Valve DC Solenoids
b) Main Steam Isolation Valve DC Solenoids
c) Recirculation Motor-Generator Set Emergency OilPumps
d) Backup Scram Valves
e) RHR Shutdown Isolation Valves
f) Engineered Safeguards Logic Power Supplies
g) HPCI System Controls, Valves, and Auxiliaries
h) RCIC System Controls, Valves, and Auxiliaries
70
"The engineered safeguards systems that are supplied from the 250 volt DC system shall be designed tooperate at a minimum of 200 volts."
It has previously been estimated5 that under the less severe conditions of a Station Blackout, with all unit batteries avaiable, and assuming prudent actions by a well-trained operator to conserve battery potential by minimizing DC loads, the necessary 250 volt DC power for HPCIor RCIC system operation would remain available during the first four tosix hours of a Station Blackout.
In summary, it is reasonable to expect that control can be maintainedover reactor vessel pressure and water level at each of the Browns FerryUnits during a Station Blackout for as long as the 250 volt DC power remains available. This period will depend on the operator's ability toconserve the unit battery potential by minimizing the DC loads, but is expected to last from four to six hours.* During this period the operatormay have to take other actions to maintain the operability of the HPCI andRCIC systems as -Hscussed in subsection 8.1.
*A battery life of seven hours is considered in Appendix G.
71
9.0 ACCIDENT SEQUENCES RESULTING IN CORE MELTDOWN
9.1 Introduction
This section deals with the various accident sequences which mightoccur during a complete Station Blackout (CSB) at the Browns Ferry NuclearPlant which can lead to core meltdown.18 Event trees are presented forwhat are considered to be the six most probable sequences. The characteristics and event timing for each of these sequences were determined by useof the MARCH code, and include consideration of the operator's role in reactor vessel pressure and level control. The progression of core damageand containment failure following core uncovery will be discussed in thissection; an event tree for operator key actions in mitigating the accidentprogression will be discussed in Sect. 10.
Of the sequences modelled by MARCH and presented in this section, thesequence TB'* is deemed to be most representative of the events whichwould occur after core uncovery following a four-hour period during whichthe availability of DC power permitted reactor vessel level and pressurecontrol. For this analysis, it is assumed that no independent secondaryequipment failures occur, and this sequence was chosen to serve as thebasis for the fission product transport analysis discussed in Volume 2 ofthis report.
9.2 Accident Phenomenology
9.2.1 Accident Progression Resulting in Core Melt
Upon a loss of offsite and onsite AC power, a number of reactorsafety systems respond immediately and automatically. First of all, afull load rejection (i.e., fast closure of the turbine control valves) occurs, followed by the tripping of the recirculation pumps and the maincondenser cooling water pumps. With the load rejection, the scram pilotvalve solenoids are deenergized and the control rods start to move towardthe fully inserted position. The main steam isolation valves (MSIVs) begin to close, resulting in a rapid reactor vessel pressure increase. These events are closely followed by a main turbine trip (i.e., closure ofthe turbine stop valves) and the tripping of the feedwater turbines.
After the automatic actions described above, core flow is provided bynatural circulation and excess reactor vessel pressure is relieved bysteam blowdown through the safety/relief valves (SRVs) into the pressuresuppression pool. If any SRV fails to reclose after actuation, the reactor vessel will continue to depressurize with an accompanying loss of
*The nomenclature TB' follows that used in the Reactor SafetyStudy,19 where T denotes a transient event and B' denotes failure torecover either offsite or onsite AC power within about 1 to 4 hours. Ineffect, TB' is a loss of decay heat removal (TW) sequence with the lossof both offsite and onsite AC power as an initiating event.
72
water inventory similar to that occurring in a small break LOCA. According to the Reactor Safety Study,ly the probability of a stuck-open relief valve (SORV) event is estimated to be 0.10 with an error spread of 3.However, this estimate was based upon consideration of three-stage SRVssuch as those originally installed at the Browns Ferry site, with thetwo-stage relief valves of improved design which have a recently been installed at the Browns Ferry Plant, the probability of a SORV event is substantially reduced and is estimated to be less than 10-z. The occurrence of a sequence involving a SORV, designated sequence TPB', is included in the event trees of this section.
The reactor vessel water level decreases rapidly during the initialmoments of a Station Blackout due both to the void collapse caused by theincreased pressure and to the water inventory lost through the SRVs. Whenthe water level sensed by the wide range detector reaches the low waterlevel setpoint (Level 2), the DC-powered HPCI and RCIC systems are automatically initiated, with their turbine-driven pumps supplied by steamgenerated by the decay heat. The HPCI and RCIC systems are assumed to remain operational until the unit battery is exhausted at four hours intothe Station Blackout. Following the loss of these injection systems, thereactor vessel level decreases until the core is uncovered about one hour
later. Subsequently, the core begins to melt.The failure probability for HPCI has been estimated19 to be 9.8 x 10~2
with an error spread of 3 and for RCIC has been estimated to be 8 x 10~2with an error spread of 3. The overall failure probability of both HPCIand RCIC to provide makeup water during the sequence TB', is designatedTUB' and has been estimated19 to be 2 x 10-3 with an error spread of about4. The failure of both HPCI and RCIC together with a SORV occurrence isdesignated TUPB'.* Sequences TUB' and TUPB' have been included in theevent trees of this section.
9.2.2 Containment Failure Modes
Any sequence resulting in core melt will eventually lead to containment failure if electrical power is not restored before the reactor vesselfails. According to the Reactor Safety Study,19 BWR containment failurecould occur in the following categories:a — Containment failure due to steam explosion in vessel,0 — Containment failure due to steam explosion in containment,Y — Containment failure due to overpressure - release through reactor
building,
Y' — Containment failure due to overpressure-release direct to atmosphere,
j — Containment isolation failure in drywell,e — Containment isolation failure in wetwell,5 — Containment leakage greater than 2400 volume percent per day.
Containment failure by overpressurization was considered to be thedominant failure mode for a BWR inerted containment in the Reactor Safety
*TUB' and TUPB' are in effect TQUV and TPQUV sequences with theloss of both offsite and onsite AC power as an initiating event.
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74
9.3 Event Trees
9.3.1 Event Trees for Accident Sequences Resulting in Core Melt
It has been concluded in this study that the sequence of events during a prolonged complete Station Blackout (CSB) at the Browns Ferry Nuclear Plant (BFNP) would most probably be described by one of the following titles:
1. CSB + HPCI/RCIC (TB')2. CSB + HPCI/RCIC + SORV (TPB')3. CSB + Manual RCIC & SRV (TyB')4. CSB + Manual RCIC & SRV + SORV (TVPB")5. CSB + No HPCI/RCIC (TUB')6. CSB + No HPCI/RCIC + SORV (TUPB')
As indicated in this listing, a stuck-open relief valve (SORV) correspondsto a small break LOCA. The terminology "Manual RCIC & SRV" refers to theassumption that the operator controls level and pressure by remote-manualoperation of the RCIC system and the relief valves, respectively. Thephrase "No HPCI/RCIC" means that both of these water-injection systems areassumed to be unavailable due to mechanical failure.
The event tree identifying these six accident sequences is shown inFig. 9.1. Sequence No. 1, traced by the dotted line in this figure, isthe sequence TB' which involves no independent secondary equipment failures and was selected as the basis for the fission product transport analysis discussed in Volume 2. Other possible but much less likely sequencessuch as failure of reactor scram are shown on Fig. 9.1 but were not considered in this study.
9.3.2 Core Damage Event Tree
Without restoration of offsite or onsite AC power and with the consequent inevitable loss of DC power, any of the numbered sequences shownon Fig. 9.1 will lead to core uncovery and subsequent core melting, reactor vessel failure, and corium-concrete interaction. A tree for theevents subsequent to core uncovery is shown in Fig. 9.2, with the pathused for the fission product transport analysis of Vol. 2 identified by adotted line.
9.3.3 Containment Event Tree
A containment event tree for the occurrences following core melt isgiven in Fig. 9.3. The particular path which has been chosen for the fission product transport studies is again indicated by the dotted line path.
It should be noted that if AC power is restored after core, melt, butbefore a failure of the reactor vessel occurs, the low-pressure EmergencyCore Cooling Systems (ECCS) would function and a breach of the containmentin the drywell or wetwell areas may be averted.
INITIATING EVENT
LOSS OF OFFSITE
AND ONSITE
ac POWER
YES
A
VNO
SYSTEM
AUTOMATIC
RESPONSES
r-
I
STUCK OPEN
RELIEF VALVE
HPCI AND
RCIC
OPERABLE
MANUAL SRV
DEPRESSURIZATION
WITHIN 15 min.
LOSS OF dc POWER -
HPCI/RCIC INOPERABLE
ORNL-DWG 81-8164A ETD
SEQUENCE CONSEQUENCE
•© ITVPB') CORE MELT
POTENTIAL DELAYED
CORE DAMAGE
•(?) (TPB'l CORE MELT
POTENTIAL DELAYED
CORE DAMAGE
.(6)(TUPB) CORE MELT
•(5) <TVB') CORE MELT
POTENTIAL DELAYED
CORE DAMAGE
.-Q(TB') CORE MELT
POTENTIAL DELAYED
CORE DAMAGE
•©(TUB') CORE MELT
POTENTIAL CORE
DAMAGE - ATWS
Fig. 9.1 Event tree for accident sequences resulting in core melt.
-J
INITIATING EVENT -CORE UNCOVERY
YES
REACTOR VESSEL
DEPRESSURIZED
dc POWER RESTORED
BEFORE EXTENSIVECORE DAMAGE -
HPCI/RCIC OPERABLE
ORNL-DWG 81-8200 ETD
OFFSITE ac POWER RESTORED
BEFORE EXTENSIVECORE DAMAGE-
LOW PRESSURE ECCS
OPERABLE
dc POWER RESTORED
AFTER SIGNIFICANTCORE MELT, BUTBEFORE VESSEL
FAILURE -
HPCI/RCIC OPERABLE
OFFSITE ac POWER RESTORED
AFTER SIGNIFICANT
CORE MELT, BUTBEFORE VESSEL
FAILURE-
LP ECCS OPERABLE
OFFSITE ac POWER RESTORED AFTER VESSELFAILURE - LP ECCS OPERABLE
Fig. 9.2 Core damage event tree.
SLIGHT CORE DAMAGE,LITTLE OR NO CORE MELT
CORE MELT, DEBRISCONTAINED IN VESSEL
CORE MELT, RPVFAILURE, NO CONCRETE
ATTACK
CORE MELT, RPVFAILURE, AND CONCRETE
ATTACK
SLIGHT CORE DAMAGE,LITTLE OR NO CORE MELT
CORE MELT, DEBRISCONTAINED IN VESSEL
CORE MELT, RPVFAILURE, NO CONCRETE
ATTACK
CORE MELT, RPVFAILURE, AND CONCRETE
ATTACK
C*
INITIATING EVENT-CORE MELT
YES
A
LARGE AMPLITUDECONOENSTION
OSCILLATIONS INWETWELL
T-QUENCHERFAILURE
WETWELLRUPTUREFAILURE
77
VESSELFAILURE
DRYWELLEPA VENTING
DRYWELL
EPA FAILURE
DRYWELL
RUPTUREFAILURE
ORNL-DWG 81-0201 ETD
HIGH RELEASE (WETWELL FAILUREl
VERY HIGH RELEASE (DRYWELL FAILUREl
SOME RELEASE (DRYWELL VENTINGI
VERY HIGH RELEASE (DRYWELL FAILUREl
NO CONTAINMENT FAILURE
NO VESSEL FAILURE. NO CONTAINMENTFAILURE
HIGH RELEASE (WETWELL FAILUREl
VERY HIGH RELEASE (DRYWELL FAILUREl
SOME RELEASE (ORYWELL VENTING I
VERY HIGH RELEASE (DRYWELL FAILUREl
NO CONTAINMENT FAILURE
NO VESSEL FAILURE, NO CONTAINMENTFAILURE
VERY HIGH RELEASE (DRYWELL FAILUREl
SOME RELEASE (DRYWELL VENTINGI
VERY HIGH RELEASE (DRYWELL FAILUREl
NO CONTAINMENT FAILURE
NO VESSEL FAILURE. NO CONTAINMENTFAILURE
VERY HIGH RELEASE IDRYWELL FAILUREl
SOME RELEASE (DRYWELL VENTINGI
VERY HIGH RELEASE (DRYWELL FAILUREl
NO CONTAINMENT FAILURE
— NO VESSEL FAILURE, NO CONTAINMENT
FAILURE
Fig. 9.3 Containment event tree.
78
9.4 Accident Sequences
9.4.1 MARCH computer code
Detailed calculations for the accident sequences have been performedwith the MARCH computer code , version 1.4B. This version is based onversion 1.4 at Brookhaven National Laboratory, but contains a differentsubroutine ANSQ for the decay heat power calculations and modifications tothe subroutine HEAD which were necessary to improve the prediction of vessel failure time. 18
In the MARCH code, the fission product decay heat source term isbased on ANS Standard ANS-5.1 (1973)za which does not account for thedecay of U-239 or Np-239. The new standard ANS-5.1 (1979) zl* providesdecay parameters for U-239 and Np-239, but does not account for the decaypower generated from these or the numerous other heavy nuclides (acti-nides) which exist in power reactors. The new subroutine ANSQ used inthis study (Cf. Appendix B) is based on the actinide decay heat source ina BWR following a depletion of 34,000 MWd/t. The actinide heating calculations Zb were performed using the EPRI-CINDER code and include all significant actinides from Tl-208 through Cm-246. Furthermore, the initialfission energy release from the fuel is assumed to last 3 seconds ratherthan 5 seconds as used originally in the MARCH code. This has been foundto yield better agreement with test data in implementing the REDY code atGE.2b
A comparison of MARCH calculations for the sequence TB" as performed by versions 1.4 and 1.4B are given in Figs. 9.4 through 9.11 forthe time- dependence of (1) the maximum core temperature, (2) the water-steam mixture level, (3) the containment wall temperature, and (4) thecontainment volumetric leak rate. It should be noted that the times predicted by version 1.4B for core uncovery and core melt are less by approximately 18 and 26%, respectively, and that the maximum containment walltemperature is increased by more than 120%. Other comparisons have shownthat the 1.4B version produces results in better agreement with the coreuncovery time calculated by the RELAP4/Mod 7 Code for the sequence TUB "at INEL. 2/
9.4.2 Accident progression signatures
This section contains the accident progression signatures for thesix most probable sequences following complete station blackout (CSB).The early part of the plant transient characteristics follow that contained in the Browns Ferry Nuclear Plant Final Safety Analysis Report.All other accident signatures have been based on the results calculated bythe MARCH computer code.13
9.4.2.1 CSB + HPCI (TB'). The accident progression signature forthe sequence TB' is given in Table 9.1. This sequence has been chosenfor the fission product transport studies presented in Volume 2.
9.4.2.2 CSB + HPCI + SORV (TPB'). The accident progression signature for the sequence TPB' is given in Table 9.2. The SORV event hasthe same effect as a small break LOCA, ensuring that the reactor vessel
3000.0
2500.0-
3 2000.0\-<cc
H 1500.0UJ
cc
oo
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0.0
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ORNL-DWG 81-8515 ETD
CORE MELTING BEGINS
DECREASE FOLLOWING SCRAMCORE UNCOVERY BEGINS
50.0
dc POWER LOST
—I 1 I100.0 150.0 200.0
TIME (min)
1250.0 300.0
Fig. 9.4 Maximum core temperature (MARCH/MOD 1.4B).
I
350.0 400.0
3000.0
2500.0-
LU
3 2000.0
<rrLUa.
2
h 1500.0UJ
cc
Oo
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<
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0.0
ORNL-DWG 81-8516 ETD
CORE MELTING BEGINS
DECREASE FOLLOWING SCRAM
CORE UNCOVERY BEGINS
dc POWER LOST
T T1
T T50.0 100.0 150.0 200.0 250.0 300.0
TIME (min)
350.0 400.0 450.0 500.0
Fig. 9.5 Maximum core temperature (MARCH/MOD 1.4)
oso
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TIME (min)
ORNL-DWG 81-8517 ETD
LOSS OF dc POWERBOILOFF BEGINS
CORE SLUMPS-
—I 1 1
250.0 300.0 350.0 400.0
Fig. 9.6 Water-steam mixture level (MARCH/MOD 1.4B)
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TIME (min)
LOSS OF dc POWER
BOILOFF BEGINS
Fig. 9.7 Water-steam mixture level (MARCH/MOD 1.4)
ORNL-DWG 81-8518 ETD
CORE SLUMPS-
T T
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VESSEL BOTTOM HEAD FAILS
T
100.0 200.0 300.0 400.0
Fig. 9.8 Containment wall temperature (MARCH/MOD 1.4B).
GROSS EPA FAILURE
EPA VENTING
T 1 1
500.0 600.0
TIME (min)
I I
700.0 800.0
ORNL-DWG 81-8519 ETD
—I 1^ 1
900.0 1000.0 1100.0
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TIME (min)
VESSEL BOTTOM HEAD FAILS
Fig. 9.9 Containment wall temperature (MARCH/MOD 1.4)
ORNL-DWG 81-8520 ETD
HEATUP DURING
CORIUM-CONCRETE
REACTION
oo
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700.0 800.0 900.0 1000.0 1100.0
Fig. 9.10 Containment leak rate (MARCH/MOD 1.4B)
00
400.0-1
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200.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0 1000.0 1100.0 1200.0TIME (min)
Fig. 9.11 Containment leak rate (MARCH/MOD 1.4).
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87
Table 9.1
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + HPCI/RCIC
(TB')
Time
(sec) Event
0.0 Loss of all AC power and diesel generators. The plantis initially operating at 100% power.
Initial drywell temperature = 66°C (150°F)Initial wetwell temperature = 35°C (95°F)
0.2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0.2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0.2 Reactor pressure increases suddenly due to load rejection.
0.3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0.5 Turbine bypass valves start to open due to load rejection.
1.0 Neutron flux starts to decrease after an initial in
crease to over 100% rated power level.
1.0 Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2.0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawnposition.
88
Time
(sec) Event
3.0 Turbine trips off (turbine stop valves fully closed).
3.5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5*0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Seven (7) out of thirteen (13) S/RVs start to open inresponse to pressure rise above the setpoint.
5.2 Water-steam mixture level recovers 0.51 m (20 in.)from the previous momentary 1.02 m (40-in.) drop.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 7 S/RVs are completely closed.
15.7 Four out of 13 S/RVs start to open.
17.0 Neutron flux drops below 1% of initial full powerlevel.
21.0 Narrow range (NR) sensed water level reaches low alarm(Level 4), i.e., 5.98 m (235.50 in.) above Level 0, or5.00 m (196.44 in.) above TAF.
22.0 Suppression pool water average temperature risesto 35.13°C (95.24°F) in response to the first S/RVpops.
29.0 All 4 S/RVs are completely closed.
89
Time
(sec) Event
29.7 Two out of 13 S/RVs start to open.
47.0 All 2 S/RVs are completely closed.
47.7 One out of 13 S/RVs starts to open.
56.0 Suppression pool water average temperature is approximately 35.3°C (95.54°F).
56.0 NR sensed water level reaches low level alarm (Level3), i.e., 5.50 m (216.00 in.) above Level 0, or 4.50 m(176.94 in.) above TAF.
90.0 Suppression pool water average temperature is approximately 35.4°C (95.72°F).
101.0 The S/RV is completely closed. The same S/RV continues to cycle on and off on setpoints throughout thesubsequent cyclings of HPCI and RCIC injections.
625 Wide range sensed water level reaches low water levelsetpoint (Level 2), i.e., 4.18 m (164.50 in.) aboveLevel 0 at 2/3 core height, or 2.96 m (116.50 in.)above TAF.
625 HPCI and RCIC systems are automatically turned on.The HPCI and RCIC turbine pumps are driven by steamgenerated by decay heat. System auxiliaries arepowered by the 250 V dc system.
655 HPCI and RCIC flows enter the reactor pressure vesselat 315 Jt/s (5000 gpm) and 38 £/s (600 gpm), respectively, drawing water from the condensate storagetank.
12.5 min. Narrow range sensed water level reaches Level 8 set-point, i.e., 6.86 m (270.00 in.) above Level 0, or5.64 m (222.00 in.) above TAF.
Time
(sec)
12,>5 min.
20 min.
26..5 min.
90
Event
HPCI and RCIC trip off.
Drywell and wetwell temperatures exceed 70°C (158°F) and42°C (108°F), respectively.
Wide range sensed water level reaches Level 2 setpoint andHPCI automatically restarts. (RCIC does not automaticallyrestart.)
27.0 min. HPCI flow enters the RPV.
29.0 min. Narrow range sensed water level reaches Level 8 setpoint andHPCI trips off again. The HPCI system, driven by steam generated by decay heat, turns on and off automatically betweensensed Levels 2 and 8 until the batteries run out.
80 min. Auto-isolation signal initiates as increase of drywellpressure exceeds 13.8 KPa (2.0 psi). The HPCI/ RCIC systemsare not isolated.
240 min. The HPCI pump stops when the batteries run out.
260 min. Wide range sensed water level reaches Level 2 setpoint.Drywell and wetwell temperatures are 85°C (185°F) and 87°C(188°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 27.82 3.68 * 103 7.68 * 107 4.37 x 106Hydrogen 0 0 0 0
302 min. Core uncovery time. Steam-water mixture level is at 3.58 m(11.73 ft) above bottom of the core.
320 min. Average gas temperature at top of core is 485°C (904°F).Drywell and wetwell temperatures and pressures are 101°C(213°F) and 0.21 MPa (31 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 18.75 2.48 x io3 5.40 * 107 3.07 x io6Hydrogen 6.65 x io"7 8.80 x i(f5 3.34 1.90 x io"1
91
Time
(sec) Event
340 min. Average gas temperature at top of core is 821°C (1509°F).Drywell and wetwell temperatures and pressures are 103°C(218°F) and 0.23 MPa (33 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 11.75 1.56 x 103 3.76 x IO7 2.14 x ify*Hydrogen 1.52 x io"3 2.00 x io"1 1.10 x io*1 6.25 x io2
355 min. Core melting starts.
389 min. Water level in vessel drops below bottom grid elevation.
390 min. Bottom grid fails and temperature of structures in bottomhead is above water temperature.
392 min. The corium slumps down to vessel bottom.
394 min. Debris starts to melt through the bottom head.
426 min. Vessel bottom head fails, resulting in a pressure increaseof 0.34 MPa (49 psia).
426.04 min. Debris starts to melt the concrete floor of the containment
building. Temperature of debris is 1433°C (2611°F) initially. Internal heat generation in metals and oxides are 1.05x IO7 and 1.95 x io' watts, respectively.
503.27 min. Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 204°C (400°F) and startto vent through the primary containment at a leak rate of118 SL/s (250 ft3/min).
513.59 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate
(kg/s (lb/min) (w) (Btu/min)
Steam 4.61 610 1.59 x io5 9052Hydrogen 0.11 15 0 0
C02 1.01 133
CO 2.35 311
The leak rate through the drywell penetration seals is-3.04 x io1* jt/s (6.44 x 104* ft3/min).
92
Time
(sec) Event
613 min. Drywell and wetwell pressures are at 0.10 MPa (~14.7 psia)and temperatures are 661°C (~1222°F) and 98°C (~209°F),respectively. The leak rate through the containment failedarea is ~2.96 x 101* A/s (~6.27 x io** ft3/min).
695 min. Drywell and wetwell temperatures are 623°C (1154°F) and 97°C(207°F), respectively. The leak rate through the containment failed area is ~6.47 x io1* i/s (~1.37 x io5ft3/min).
1028 min. Drywell and wetwell temperatures are 614°C (~1138°F) and97°C (~207°F), respectively. The leak rate through thecontainment failed area is ~1.34 x io3 l/s (~2.83 xIO3 ft3/min).
93
Table 9.2
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + HPCI/RCIC + SORV (Small Break LOCA)
(TPB')
Time
(sec) Event
0.0 Loss of all AC power and diesel generators. The plantis initially operating at 100% power.
Initial drywell temperature - 66°C (150°F)Initial wetwell temperature - 35°C (95°F)
0.2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0.2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0.2 Reactor pressure increases suddenly due to load rejection.
0.3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0.5 Turbine bypass valves start to open due to load rejection.
1.0 Neutron flux starts to decrease after an initial in
crease to over 100% rated power level.
1.0 Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2.0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawn
position.
94
Time
(sec) Event
3.0 Turbine trips off (turbine stop valves fully closed).
3.5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5.0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Seven (7) out of thirteen (13) S/RVs start to open inresponse to pressure rise above the setpoint.
5.2 Water-steam mixture level recovers 0.51 m (20 in.)from the previous momentary 1.02 m (40-in.) drop.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 6 S/RVs are completely closed. One S/RV is stuckopen (SORV); this has the same effect as a small breakLOCA of equivalent break area of 0.015 m2 (0.1583ft2).
15.7 Four out of 13 S/RVs start to open.
17.0 Neutron flux drops below 1% of initial full powerlevel.
21.0 Narrow range (NR) sensed water level reaches low alarm(Level 4), i.e., 5.98 m (235.50 in.) above Level 0, or5.00 m (196.44 in.) above TAF.
95
Time
(sec) Event
22.0 Suppression pool water average temperature risesto 35.13°C (95.24°F) in response to the first S/RVpops.
29.0 All 4 S/RVs are completely closed.
29.7 Two out of 13 S/RVs start to open.
47.0 All 2 S/RVs are completely closed.
47.7 One out of 13 S/RVs starts to open.
56.0 Suppression pool water average temperature is approximately 35.3°C (95.54°F).
56.0 NR sensed water level reaches low level alarm (Level3), i.e., 5.50 m (216.00 in.) above Level 0, or 4.50 m(176.94 in.) above TAF.
90.0 Suppression pool water average temperature is approximately 35.4°C (95.72°F).
101.0 The S/RV is completely closed. The same S/RV continues to cycle on and off on setpoints throughout thesubsequent cyclings of HPCI and RCIC injections.
625 Wide range sensed water level reaches low water levelsetpoint (Level 2), i.e., 4.18 m (164.50 in.) aboveLevel 0 at 2/3 core height, or 2.96 m (116.50 in.)above TAF.
625 HPCI and RCIC systems are automatically turned on.The HPCI and RCIC turbine pumps are driven by steamgenerated by decay heat. System auxiliaries are
powered by the 250 V dc system.
655 HPCI and RCIC flows enter the reactor pressure vesselat 315 l/s (5000 gpm) and 38 l/s (600 gpm), respectively, drawing water from the condensate storagetank.
12.5 min. Narrow range sensed water level reaches Level 8 set-point, i.e., 6.86 m (270.00 in.) above Level 0, or5.64 m (222.00 in.) above TAF.
12.5 min. HPCI and RCIC trip off.
96
Time
(sec) Event
20 min. Drywell and wetwell temperatures exceed 70°C (158°F) and50°C (122°F), respectively.
Mass and energy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 35.04 4.63 x io3 9.87 x io7 5.61 x io6Hydrogen 0 0 0 0
25.5 min. Auto-isolation signal initiates as increase of drywell pressure exceeds 13.8 KPa (2.0 psi). The HPCI/ RCIC systems arenot isolated.
26.5 min. Wide range sensed water level reaches Level 2 setpoint andHPCI automatically restarts. (RCIC does not automaticallyrestart.)
27.0 min. HPCI flow enters the RPV.
29.0 min. Narrow range sensed water level reaches Level 8 setpoint andHPCI trips off again. The HPCI system, driven by steam generated by decay heat, turns on and off automatically betweensensed Levels 2 and 8 until the batteries run out.
240 min. The HPCI pump stops when the batteries run out.
261 min. Wide range sensed water level reaches Level 2 setpoint.Drywell and wetwell temperatures are 100°C (212°F) and 96°C(205°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 10.45 1.38 x io3 2.92 x io7 1.66 x io6Hydrogen 0 0 0 0
315.07 min. Core uncovery time. Steam-water mixture level is at 3.53 m(11.58 ft) above bottom of the core.
321.5 min. Average gas temperature at top of core is 211°C (412°F).Drywell and wetwell temperatures and pressures are 106°C(223°F) and 0.24 MPa (35 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 22.66 3.00 x io3 6.33 x io7 3.60 x io6Hydrogen 0 0 0 0
97
Time
(sec) Event
384 min. Average gas temperature at top of core is 1046°C (1915°F).Drywell and wetwell temperatures and pressures are 1148C(237°F) and 0.28 MPa (41 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 2.46 325.0 8.30 x io6 4.72 x IO5Hydrogen 0.021 2.78 1.14 x io5 8.19 x io3
387.77 min. Core melting starts.
418.77 min. Water level in vessel drops below bottom grid elevation.
420.77 min. Bottom grid fails and temperature of structures in bottomhead is above water temperature.
421.67 min. The corium slumps down to vessel bottom.
422.12 min. Debris starts to melt through the bottom head.
515.18 min. Vessel bottom head fails, resulting in a pressure increaseof 0.34 MPa (49 psia).
515.20 min. Debris starts to boil water from containment floor.
515.20 min. Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 204°C (400°F) and startto vent through the primary containment at a leak rate of105 i/a (222 ft3/min).
515.21 min. Debris starts to melt the concrete floor of the containment
building. Temperature of debris is 1771°C (3219°F) initially. Internal heat generation in metals and oxides are 1.01x IO7 and 1.86 x IO7 watts, respectively.
579.24 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s (lb/min) (w) (Btu/min)
Steam 0.63 83.33 1.59 x io5 9052Hydrogen 0.22 29.10 0 0
C02 2.08 275.13
CO 4.63 612.44
The leak rate through the drywell penetration seals is~2.53 x 105 i/s (5.35 x IO5 ft3/min).
98
Time
(sec) Event
681.88 min. Drywell and wetwell pressures are at 0.10 MPa (~14.7 psia)and temperatures are 674°C (~1245°F) and 98°C (~209°F),respectively. The leak rate through the containment failedarea is ~3.61 x IO4 n/s (-7.65 x 10^ ft3/min).
810 min. Drywell and wetwell temperatures are 1006°C (1843°F) and97°C (207°F), respectively. The leak rate through thecontainment failed area is ~3.10 x io1* SL/s (~6.57 xIO4 ft3/min).
1128 min. Drywell and wetwell temperatures are 591°C (~1095°F) and97°C (~207°F), respectively. The leak rate through thecontainment failed area is ~1.27 x io3 X./s (~2.70 x103 ft3/min).
99
is depressurized during core uncovery and melt. However, the pressure remains high enough so that the HPCI turbine can be operated for as long asthe battery lasts. Compared with the sequence TB', the SORV results ina ~30 min. delay in the core melt, ~90 min. delay in vessel failure, and~65 min. delay in the containment failure.
9.4.2.3 CSB + Manual RCIC & SRV (TVB'). The accident progressionsignature for the sequence TVB' is given in Table 9.3. In this sequence, it is assumed that the HPCI system is inoperable. The operator re-mote-manually opens one SRV after 15 minutes of Station Blackout to depressurize the vessel in order to lower the vessel temperature and to prepare for use of the low pressure ECCS injection systems upon restorationof AC power. At the same time, the operator attempts to maintain a constant vessel water level by manually controlling the RCIC injection. Itis noted that the fluid loss through the SRV causes the core to becomemomentarily uncovered, but it refloods immediately with the continued RCICwater injection. This short period of core uncovery is believed not tocause any core damage at that stage.
9.4.2.4 CSB + Manual RCIC & SRV + SORV (T„PB"). The accident progression signature for the sequence TVPB' is given in Table 9.4. Inthis sequence, it is assumed that an SORV event occurs at the beginning ofthe transient, and the operator remote-manually opens a SRV which causesthe reactor vessel to depressurize at a faster rate. Compared with thesequence TVB', the two sequences have produced very similar resultswith respect to the times of core melt, vessel failure, and containmentfailure.
9.4.2.5 CSB + No HPCI/RCIC (TUB"). The accident progression signature for the sequence TUB' is given in Table 9.5. The events of thissequence as calculated by MARCH have been compared with results from theRELAP4/MOD7 code2/ up to the time of core uncovery. With the modifications incorporated in MARCH version 1.4B, the core uncovery times predicted by the two codes are in agreement.
9.4.2.6 CSB + No HPCI/RCIC & SORV (TUPB"). The accident progressionsignature for the sequence TUPB' is given in Table 9.6. This is theseverest of the six Station Blackout signatures studied, and includes aprediction of core uncovery at 17 minutes after the initiating loss of ACpower. The mitigating action which might be taken by the operator in response to this and the other accident progression signatures will be discussed in Section 10.
9.4.2.7 Summary of key events. A comparison of the predicted timesto key events following core uncovery for each of the six sequences ascalculated by MARCH version 1.4B is given in Table 9.7.
9.4.2.8 Key results for sequence TB '. Key results from the MARCHcalculations for the sequence TB1 which is used as the base case for thefission product transport calculations in Volume 2 are presented in Figs.9.12 through 9.21.
100
Table 9.3
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + Manual RCIC & SRV
(TVB-)
Time
(sec) Event
0.0 Loss of all AC power and diesel generators. The plantis initially operating at 100% power.
Initial drywell temperature - 66°C (150°F)Initial wetwell temperature - 35°C (95°F)
0.2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0.2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0.2 Reactor pressure increases suddenly due to load rejection.
0.3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0.5 Turbine bypass valves start to open due to load rejection.
1.0 Neutron flux starts to decrease after an initial in
crease to over 100% rated power level.
1.0 Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2.0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawnposition.
101
Time
(sec) Event
3.0 Turbine trips off (turbine stop valves fully closed).
3.5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5.0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Seven (7) out of thirteen (13) S/RVs start to open inresponse to pressure rise above the setpoint.
5.2 Water-steam mixture level recovers 0.51 m (20 in.)from the previous momentary 1.02 m (40-in.) drop.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 6 S/RVs are completely closed. One S/RV is stuckopen (SORV); this has the same effect as a small breakLOCAL of equivalent break area of 0.015 m2 (0.1583ft2).
15.7 Four out of 13 S/RVs start to open.
17.0 Neutron flux drops below 1% of initial full powerlevel.
21.0 Narrow range (NR) sensed water level reaches low alarm(Level 4), i.e., 5.98 m (235.50 in.) above Level 0, or5.00 m (196.44 in.) above TAF.
22.0 Suppression pool water average temperature risesto 35.13°C (95.24°F) in response to the first S/RVpops.
102
Time
(sec) Event
29.0 All 4 S/RVs are completely closed.
29.7 Two out of 13 S/RVs start to open.
47.0 All 2 S/RVs are completely closed.
47.7 One out of 13 S/RVs starts to open.
56.0 Suppression pool water average temperature is approximately35.3°C (95.54°F).
56.0 NR sensed water level reaches low level alarm (Level 3),i.e., 5.50 m (216.00 in.) above Level 0, or 4.50 m (176.94in.) above TAF.
90.0 Suppression pool water average temperature is approximately35.4°C (95.72°F).
101.0 The S/RV is completely closed. The same S/RV continues tocycle on and off on setpoints throughout the subsequent RCICinjections.
625 Wide range sensed water level reaches low water level set-point (Level 2), i.e., 4.18 m (164.50 in.) above Level 0 at2/3 core height, or 2.96 m (116.50 in.) above TAF.
625 Operator manually controls RCIC injection to maintain constant vessel water level. The RCIC turbine pump is drivenby steam generated by decay heat. System auxiliaries arepowered by the 250 V dc system.
655 RCIC flows enter the reactor pressure vessel at 38 H/s(600 gpm) drawing water from the condensate storage tank.
15 min. Operator manually opens one SRV to depressurize the vessel.
20 min. Drywell and wetwell temperatures exceed 76°C (169°F) and50°C (122°F), respectively. Mass and energy addition ratesinto the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 829.75 1.10 x 105 2.32 x IO8 1.32 x IO7Hydrogen 0 0 0 0
103
Time
(sec) Event
21.14 min. Core uncovery time.
22.0 min. Core refloods.
30 min. Auto-isolation signal initiates as increase of drywellpressure exceeds 13.8 KPa (2.0 psi). The RCIC system is notisolated.
240 min. The RCIC pump stops when the batteries run out.
266.3 min. Wide range sensed water level reaches Level 2 setpoint.Drywell and wetwell temperatures are 99°C (210°F) and 100°C(212°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate Energy Rate
(kg/s) (lb/min) (w) (Btu/min)
Steam 19.16 2.53 x IO3 5.20 x IO7 2.96 x io6Hydrogen 0 0 0 0
347 min. Core uncovers again.
366 min. Average gas temperature at top of core is 491°C (916°F).Drywell and wetwell temperatures and pressures are 113°C(236°F) and 0.28 MPa (40 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 9.26 1.22 x 103 2.97 x IO7 1.69 x io6Hydrogen 4.09 x io"5 5.41 x io"3 222.28 12.64
386 min. Average gas temperature at top of core is 855°C (1571°F).Drywell and wetwell temperatures and pressures are 115°C(239°F) and 0.29 MPa (41 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 5.05 6.68 x io2 1.81 x IO7 1.03 x IO6Hydrogen 1.68 x IO-2 2.23 1.35 x IO5 7.70 x 103
Time
(sec)
395.3 min.
449 .3 min.
451,.2 min.
452 min.
452.,9 min.
539..3 min.
539.,3 min.
539.,3 min.
104
Event
Core melting starts.
Water level in vessel drops below bottom grid elevation.
Bottom grid fails and temperature of structures in bottomhead is above water temperature.
The corium slumps down to vessel bottom.
Debris starts to melt through the bottom head.
Vessel bottom head fails, resulting in a pressure increaseof 0.0047 MPa (0.68 psia).
Debris starts to boil water from containment floor.
Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 204°C (400°F) and startto vent through the primary containment at a leak rate of118 Z/s (250 ft3/min).
539.3 m^n. Debris starts to melt the concrete floor of the containmentbuilding. Temperature of debris is 1750°C (3182°F) initially. Internal heat generation in metals and oxides are 9.99x 10° and 1.84 x io7 watts, respectively.
601.05 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s (lb/min) (w) (Btu/min)
Steam 4.70 621.51 1.59 x io5 9052Hydrogen 0.14 18.27 0 0C02 1.29 170.23CO 2.88 381.21
The leak rate through the drywell penetration seals is-5.33 x io* X,/s (1.13 x io5 ft3/min).
718.8 min. Drywell and wetwell pressures are at 0.10 MPa (-14.7 psia)and temperatures are 700°C ( 1293°F) and 98°C (~209°F),respectively. The leak rate through the containment failedarea is -5.18 x io4 l/s (-1.10 x io5 ft3/min).
105
Time
(sec) Event
821.5 min. Drywell and wetwell temperatures are 737°C (1359°F) and 93°C(199°F), respectively. The leak rate through the containment failed area is -4.23 x io4* Jl/s (-8.96 x io1*ft3/min).
1127.5 min. Drywell and wetwell temperatures are 468°C (~875°F) and86°C (~188°F), respectively. The leak rate through thecontainment failed area is -4.79 x io1* SL/b (-1.02 xIO1* ft3/min).
0.0
106
Table 9.4
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + Manual RCIC & SRV + SORV (Small Break LOCA)
(TVPB')
Time
(sec) Event
Loss of all AC power and diesel generators. The plantis initially operating at 100% power.Initial drywell temperature - 66°C (150°F)Initial wetwell temperature = 35°C (95°F)
°«2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0.2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0*2 Reactor pressure increases suddenly due to load rejection.
0.3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0.5 Turbine bypass valves start to open due to load rejection.
1*0 Neutron flux starts to decrease after an initial increase to over 100% rated power level.
1«° Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2*0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawnposition.
107
Time
(sec) Event
3.0 Turbine trips off (turbine stop valves fully closed).
3.5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5.0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Operator manually opens one S/RV to depressurize thevessel.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 6 S/RVs are completely closed. One S/RV is stuckopen (SORV). This has the same effect as a smallbreak LOCA of equivalent break area of 0.0147 m(0.1583 ft2).
17.0 Neutron flux drops below 1% of initial full powerlevel.
30 Operator manually controls RCIC injection to maintainconstant vessel water level. The RCIC turbine pump isdriven by steam generated by decay heat. Systemauxiliaries are powered by the 250 V dc system.
60 RCIC flows enter the reactor pressure vessel at 38l/s (600 gpm) drawing water from the condensatestorage tank.
108
Time
(sec) Event
10.52 min. Core uncovery time. Steam-water mixture level is at 3.56 m(11.68 ft) above bottom of the core.
11.52 min. Core refloods.
20 min. Auto-isolation signal initiates as increase of drywellpressure exceeds 13.8 KPa (2.0 psi). The RCIC system isnot isolated.
240 min. The RCIC pump stops when the batteries run out.
270 min. Wide range sensed water level reaches Level 2 setpoint.Drywell and wetwell temperatures are 100°C (212°F) and 102°C(216°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 19.50 2.58 x io3 5.29 x io7 3.01 x io6Hydrogen 0 0 0 0
337 min. Core uncovers again.
356 min. Average gas temperature at top of core is 313°C (595°F).Drywell and wetwell temperatures and pressures are 113°C(236°F) and 0.28 MPa (40 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 12.25 1.62 x io3 3.64 x io7 2.07 x io6Hydrogen 3.05 x io~8 4.03 x io~6 0.115 6.53 x io"3
376.4 min. Average gas temperature at top of core is 650°C (1202°F).Drywell and wetwell temperatures and pressures are 115°C(239°F) and 0.29 MPa (41 psia), respectively. Mass andenergy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 7.20 9.52 x io2 2.43 x io7 1.38 x io6Hydrogen 1.60 x io~3 0.21 1.07 x up 6.09 x io2
109
Time
(sec) Event
396.36 min. Core melting starts.
450.4 min. Water level in vessel drops below bottom grid elevation.
452.3 min. Bottom grid fails and temperature of structures in bottomhead is above water temperature.
453.1 min. The corium slumps down to vessel bottom.
454. min. Debris starts to melt through the bottom head.
542.5 min. Vessel bottom head fails, resulting in a pressure increaseof 0.005 MPa (0.7 psia).
542.5 min. Debris starts to boil water from containment floor.
542.5 min. Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 240°C (400°F) and startto vent through the primary containment at a leak rate of104 i/s (221 ft3/min).
542.5 min. Debris starts to melt the concrete floor of the containment
building. Temperature of debris is 1766°C (3210°F) initially. Internal heat generation in metals and oxides are 1.00x IO7 and 1.83 x io' watts, respectively.
596.4 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment. Mass and energy addition rates intothe drywell are:
Mass
(kg/sRate
(lb/min)Energy
(w)Rate
(Btu/min)
Steam
Hydrogen
co2CO
7.69
0.038
2.58
0.80
1017
5.08
342
160
1.59 x io50
9052
0
The leak rate through the drywell penetration seals is-9.20 x io1* SL/s (1.95 x io5 ft3/min).
0.0
110
Table 9.5
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + No HPCI/RCIC
(TUB*)
Time
(sec) Event
Loss of all AC power and diesel generators. The plantis initially operating at 100% power.Initial drywell temperature = 66°C (150°F)Initial wetwell temperature - 35°C (95°F)
0*2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0*2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0*2 Reactor pressure increases suddenly due to load rejection.
0*3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0*5 Turbine bypass valves start to open due to load rejection.
!*0 Neutron flux starts to decrease after an initial increase to over 100% rated power level.
i'O Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2.0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawnposition.
3.0 Turbine trips off (turbine stop valves fully closed).
Ill
Time
(sec) Event
3»5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5.0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Seven (7) out of thirteen (13) S/RVs start to open inresponse to pressure rise above the setpoint.
5.2 Water-steam mixture level recovers 0.51 m (20 in.)from the previous momentary 1.02 m (40-in.) drop.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 7 S/RVs are completely closed.
15.7 Four out of 13 S/RVs start to open.
17.0 Neutron flux drops below 1% of initial full powerlevel.
21.0 Narrow range (NR) sensed water level reaches low alarm(Level 4), i.e., 5.98 m (235.50 in.) above Level 0, or5.00 m (196.44 in.) above TAF.
22.0 Suppression pool water average temperature rises to35.13°C (95.24°F) in response to the first S/RV pops.
29.0 All 4 S/RVs are completely closed.
29.7 Two out of 13 S/RVs start to open.
112
Time
(sec) Event
47.0 All 2 S/RVs are completely closed.
47.7 One out of 13 S/RVs starts to open.
56.0 Suppression pool water average temperature is approximately35.3°C (95.54°F).
56.0 NR sensed water level reaches low level alarm (Level 3),i.e., 5.50 m (216.00 in.) above Level 0, or 4.50 m (176.94in.) above TAF.
90.0 Suppression pool water average temperature is approximately35.4°C (95.72°F).
101.0 The S/RV is completely closed. The same S/RV continues tocycle on and off on setpoints throughout the sequence.
625 Wide range sensed water level reaches low water levelset-point (Level 2), i.e., 4.18 m (164.50 in.) above Level 0at 2/3 core height, or 2.96 m (116.50 in.) above TAF.
625 HPCI and RCIC systems are not turned on because they areassumed to be unavailable.
20 min. Suppression pool water average temperature reaches 46°C(114°F).
33 min. Core uncovery time. Steam-water mixture level is at 3.54 m(11.61 ft) above bottom of the core.
40 min. Auto-isolation signal initiates as increase of drywellpressure exceeds 13.8 KPa (2.0 psi). The HPCI/RCIC systemsare not affected. Drywell and wetwell temperature are 72°C(162°F) and 55°C (130°F), respectively. Mass and energyaddition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 33 4.36 x io3 9.25 x io7 5.26 x io6Hydrogen 6 x 10 9 8.62 x io-7 2.92 x io~2 1.66 x io"3
60 min. Mass and energy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 15.6 2.07 x io3 5.06 x 107 2.88 x 106Hydrogen 2.8 x io~3 3.76 x io"1 2.15 x iq1* 1.22 x io3
Time
(sec)
70
80
min.
min.
113
Event ^^^^
Core melting starts.
Drywell and wetwell temperatures are 75°C (167°F) and 63°C(145°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate
(kg/s) (lb/min)Energy Rate
(w) (Btu/min)
Steam 5.68 7.51 x io2Hydrogen 0.19 2.53 x io1
2.22 x io7 1.26 x io62.29 x io6 1.30 x io5
96 min. Water level in vessel drops below bottom grid elevation.
97 min. Bottom grid fails and temperature of structures in bottom
head is above water temperature.
99 min. The corium slumps down to vessel bottom.
101 min. The debris is starting to melt through the bottom head.Drywell and wetwell temperatures are 97°C (207°F) and 71°C(159°F), respectively. Meanwhile, local pool water temperature at the discharging bay exceeds 149°C (300°F). Steamcondensation oscillations could accelerate due to the con
tinuous discharge of superheated noncondensable gases intothe suppression pool. Mass and energy addition rates intothe wetwell are;
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 18.6 5.46 x io3 5.42 x io7 3.08 x io6Hydrogen 6.8 x io-2 8.93 3.59 x io5 2.04 x io**
129 min. Vessel bottom head fails, resulting in a pressure increaseof 0.34 MPa (49 psia).
129.03 min. Debris starts to melt the concrete floor of the containment
building. Temperature of debris is 1546°C (2815°F) initially. Internal heat generation in metals and oxides are1.36 x 107 and 2.50 x io7 watts, respectively.
114
Time
(sec) Event
165 min. Drywell and wetwell temperatures are 141°C (286°F) and 74°C(166°F), respectively. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 5.46 722.83 1.59 x io5 9052Hydrogen 3.3 x io-2 4.38 0 0CO2 2.58 341.88
CO 0.69 91.35
190 min. Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 204°C (400°F) and startto vent through the primary containment.
193 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment.
219 min. Drywell and wetwell pressures are at 0.10 MPa (14.7 psia).Drywell and wetwell temperatures are 598°C (1109°F) and 78°C(173°F), respectively. Mass and energy addition rates intothe drywell are:
Mass
(kg/s)Rate
(lb/min)Energy Rate
(w) (Btu/min)
Steam
Hydrogen
CO2CO
0.70
0.24
2.32
5.03
92
32
307
666
1.59 x IO5 90520 0
The leak rate through the containment failed areas is-2.90 x io5 X./b (-6.15 x io5 ft3/min).
250 min. Drywell and wetwell temperatures are 675°C (1247°F) and 78°C(173°F), respectively. Mass and energy addition rates intothe drywell are:
Mass
(kg/s)Rate
(lb/min)Energy Rate
(w) (Btu/min)
Steam
Hydrogen
C02CO
6.84
0.25
1.53
5.25
905
33
203
695
1.59 x 105 90520 0
115
Time
(sec) Event
The leak rate through the containment failed area is-4.91 x io4* l/s (-1.04 x io5 ft3/min).
309 min. Rate of concrete decomposition is -4.65 x io1* gm/s.Rate of heat added to atmosphere is -1.20 x 10* kW.
367 min. Drywell and wetwell pressures are at 0.10 MPa (-14.7 psia)and temperatures are 854°C (1570°F) and 77°C (171°F), respectively. The leak rate through the containment failedarea is -3.94 x \if i/s (~8.35 x io4* ft3/min).
733 min. Drywell and wetwell temperatures are 546°C (~1014°F) and77°C (170°F), respectively. The leak rate through the containment failed area is -2.12 x io3 A/s (-4.50 x103 ft3/min).
116
Table 9.6
Browns Ferry Nuclear Plant: Complete Station BlackoutSequence of Events
CSB + No HPCI/RCIC & SORV (Small Break LOCA)
(TUPS' )
Time
(sec) Event
0.0 Loss of all AC power and diesel generators. The plantis initially operating at 100% power.Initial drywell temperature =• 66°C (150°F)Initial wetwell temperature =• 35°C (95°F)
0.2 Full load rejection (i.e., fast closure of turbinecontrol valves) occurs.
0.2 Recirculation pumps and condenser circulatory waterpumps trip off. Loss of condenser vacuum occurs.Core flow is provided by natural circulation.
0.2 Reactor pressure increases suddenly due to load rejection.
0.3 Scram pilot valve solenoids are deenergized due toload rejection. Control rod motion begins.
0.5 Turbine bypass valves start to open due to load rejection.
1.0 Neutron flux starts to decrease after an initial in
crease to over 100% rated power level.
1.0 Reactor power starts to decrease slowly after an initial rise.
2.0 Control rods are 40% inserted from fully withdrawnposition.
2.0 Main steamline isolation valves (MSIVs) start to close(relay-type reactor trip system), resulting in a rapidsteam-line pressure rise.
2.0 Turbine bypass valves are tripped to close.
3.0 Control rods are 75% inserted from fully withdrawnposition.
3.0 Turbine trips off (turbine stop valves fully closed).
117
Time
(sec) Event
3.5 Power generation due to delayed neutrons and fissionproduct decay drops to 10% of initial rated power generation.
4.0 Feedwater turbines trip off.
5.0 MSIVs are fully closed, resulting in a momentary 0.69MPa (100 psi) pressure increase and 1.02 m (40-in.)drop of water-steam mixture level due to collapsing ofvoids.
5.0 All control rods are fully inserted.
5.0 Reactor vessel pressure exceeds the lowest setpoint at7.52 MPa (1090 psi) of safety/relief valves (S/RVs).
5.0 Seven (7) out of thirteen (13) S/RVs start to open inresponse to pressure rise above the setpoint.
5.2 Water-steam mixture level recovers 0.51 m (20 in.)from the previous momentary 1.02 m (40-in.) drop.
5.5 S/RV steam blowdowns into the pressure suppressionpool through the T-quenchers begin.
7.5 Feedwater flow drops below 20%.
9.0 Feedwater flow decreases to zero.
10.0 Power generation due to fission product decay drops toapproximately 7.2% of rated power generation.
15.0 All 6 S/RVs are completely closed. One S/RV is stuckopen (SORV); this has the same effect as a small breakLOCA of equivalent break area of 0.015 m2 (0.1583ft2).
15.7 Four out of 13 S/RVs start to open.
17.0 Neutron flux drops below 1% of initial full powerlevel.
21.0 Narrow range (NR) sensed water level reaches low alarm(Level 4), i.e., 5.98 m (235.50 in.) above Level 0, or5.00 m (196.44 in.) above TAF.
22.0 Suppression pool water average temperature rises to35.13°C (95.24°F) in response to the first S/RV pops.
29.0 All 4 S/RVs are completely closed.
29.7 Two out of 13 S/RVs start to open.
118
Time
(sec) Event
47.0 All 2 S/RVs are completely closed.
47.7 One out of 13 S/RVs starts to open.
56.0 Suppression pool water average temperature is approximately35.3°C (95.54°F).
56.0 NR sensed water level reaches low level alarm (Level 3),i.e., 5.50 m (216.00 in.) above Level 0, or 4.50 m (176.94in.) above TAF.
90.0 Suppression pool water average temperature is approximately35.4°C (95.72°F).
101.0 The S/RV is completely closed. The same S/RV continues tocycle on and off on setpoints throughout the sequence.
625 Wide range sensed water level reaches low water levelset-point (Level 2), i.e., 4.18 m (164.50 in.) above Level 0at 2/3 core height, or 2.96 m (116.50 in.) above TAF.
625 HPCI and RCIC systems are not turned on because they areassumed to be unavailable.
17.2 min. Core uncovery time. Steam-water mixture level is at 3.62 m(11.88 ft) above bottom of the core.
20 min. Auto-isolation signal initiates as increase of drywellpressure exceeds 13.8 KPa (2.0 psi). The HPCI/RCIC systemsare not affected. Drywell and wetwell temperature are 73°C(163°F) and 55°C (130°F), respectively. Mass and energyaddition rates into the wetwell are:
Mass rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 60 7.94 x io3 1.69 x io8 9.61 x io6Hydrogen 8.53 x io-13 1.13 x io~10 3.19 x io"6 1.81 x io~7
40 min. Mass and energy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 10.30 1.36 * 103 3.41 x io7 1.94 x io6Hydrogen 2.38 x IO-1* 3.15 x io"2 1.61 x io3 91.56
119
Time
(sec) Event
56.6 min. Core melting starts.
60 min. Drywell and wetwell temperatures are 75°C (167°F) and 63°C(145°F), respectively. Mass and energy addition rates intothe wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 2.73 3.61 x io2 1.23 x io7 6.99 x io5Hydrogen 0.49 6.48 x io1 7.05 x io6 4.01 x io5
78 min. Water level in vessel drops below bottom grid elevation.
79 min. Bottom grid fails and temperature of structures in bottomhead is above water temperature.
81 min. The corium slumps down to vessel bottom.
81.5 min. The debris is starting to melt through the bottom head.Drywell and wetwell temperatures are 82°C (180°F) and 71°C(159°F), respectively. Meanwhile, local pool water temperature at the discharging bay exceeds 149°C (300°F). Steamcondensation oscillations could accelerate due to the con
tinuous discharge of superheated noncondensable gases intothe suppression pool.
101 min. Mass and energy addition rates into the wetwell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 1.25 165.35 3.47 x io6 1.97 x io5Hydrogen 4.45 x io"* 0.06 1.03 x io3 58.58
142.5 min. Vessel bottom head fails, resulting in a pressure increaseof 0.34 MPa (49 psia).
152.5 min. Debris starts to boil water from containment floor.
162.5 min. Debris starts to melt the concrete floor of the containment
building. Temperature of debris is 2013°C (3655°F) initially. Internal heat generation in metals and oxides are2.43 x io7 and 1.26 x io7 watts, respectively.
120
Time
(sec) Event
162.5 min. Drywell and wetwell temperatures are 128°C (262°F) and 74°C(166°F), respectively. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 0.057 7.54 1.59 x io5 9052Hydrogen 0 0 0 0
167.8 min. Drywell electric penetration assembly seals have failed asthe containment temperature exceeds 204°C (400°F) and startto vent through the primary containment.
175.2 min. Containment failed as the containment temperature exceeds260°C (500°F) and all electric penetration modules are blownout of the containment.
185.3 min. Drywell and wetwell pressures are at 0.10 MPa (14.7 psia).Drywell and wetwell temperatures are 314°C (598°F) and 78°C(173°F), respectively. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
Steam 1.65 218.26 1.59 x io5 9052Hydrogen 0.025 3.31 0 0C02 5.79 765.88CO 0.526 69.58
The leak rate through the containment failed areas is~3.00 x io* A/s (-6.36 x io* ft3/min).
206 min. Drywell and wetwell temperatures are 610°C (1130°F) and 78°C(173°F), respectively. Mass and energy addition rates intothe drywell are:
Mass Rate Energy Rate(kg/s) (lb/min) (w) (Btu/min)
C02 1.83 242
Hydrogen 0.20 26 0 0Steam 1.36 180 1.59 x io5 9052CO 4.15 549
121
Time
(sec) Event
The leak rate through the containment failed area is-2.94 x 10* A/s (-6.24 x io* ft3/min).
222.5 min. Rate of concrete decomposition is -4.46 x io* gm/s.Rate of heat added to atmosphere is -3.71 x 10* kW.
254.5 min. Drywell and wetwell pressures are at 0.10 MPa (-14.7 psia)and temperatures are 746°C (1375°F) and 77°C (171°F), respectively. The leak rate through the containment failedarea is -5.54 x io* A/s (-1.17 x io5 ft3/min).
501 min. Drywell and wetwell temperatures are 815°C (-1500°F) and77°C (170°F), respectively. The leak rate through the containment failed area is -2.34 x io* A/s (-4.96 x10* ft3/min).
Table 9.7. Predicted times to key events
Sequence Event
1 CSB + HPCI/RCIC2 CSB + HPCI/RCIC + SORV3 CSB + Manual RCIC/SRV4 CSB + Manual RCIC/SRV + SORV5 CSB + iNo HPCI/RCIC6 CSB + No HPCI/RCIC + SORV
Core Core
uncover reflood
302
315
21 22
11 12
33
17
Core
uncover
again
347
337
Accident progression time(min)
Start
melt
355
388
395
396
69
57
Core
slump
392
419
449
453
95
78
Vessel
failure
426
515
539
543
128
143
Wetwell
failure
130
145'/
DrywellEPA vent
503
515
539
543
190
168
Drywell
failure
514
580
601
596
193
175
e
Wetwell failure is due to forces of steam jet impingement and condensation oscillations resulting from excessive thermalstratification in the suppression pool.
Drywell electric penetration assembly seals start to vent when the ambient temperature exceeds 204°C (400°F).
Drywell electric penetration assembly seals become decomposed and are blown out of the containment when the ambient temperature exceeds 260°C (500°F).
TJest estimate value.
/,Containment would fail at 288 min. using WASH-1400 failure criterion of static pressurization at 190 psia.Best estimate value.
toro
123
Figure 9.12 shows the reactor vessel pressure distribution; it remains approximately constant during the first 240 minutes when the steammass flow through the SRV discharge into the suppression pool (Fig. 9.13)is balanced by HPCI injection (Fig. 9.14). The pressure increases afterthe HPCI injection has stopped upon the assumed loss of dc power at 4 hinto the transient.
The mass of water and the mass of steam within the reactor vessel are
shown in Figs. 9.15 and 9.16 respectively. The relatively rapid decreasein steam inventory beginning at about 305 minutes as shown in Fig. 9.16 isdue to depletion in the production of hydrogen (Fig. 9.17) by theZirconium-water reaction (Fig. 9.18) before the onset of core melting(Fig. 9.19). The large amount of energy released by the Zr-H20 reaction(Fig. 9.20) before the core slumps into the lower head causes an increasein the steaming rate and the reactor vessel pressure.
The large amount of energy release in the predicted corium-water interaction after reactor vessel failure would cause a significant containment pressure spike. The vertical concrete penetration resulting fromcorium attack is shown in Fig. 9.21.
9.5 Containment Responses
As previously discussed, containment failure could occur either inthe drywell or in the wetwell if a core meltdown accident were to occur atthe Browns Ferry Nuclear Plant. The radiological consequences of a wetwell failure would be less severe than those of a drywell failure, becausea significant fraction of the released fission products would be dissolvedor deposited in the suppression pool.
9.5.1 Drywell Response
Drywell failure would occur in the EPA when the elastomeric sealingmaterials undergo degradation and lose sealing integrity at ambient temperatures greater than 204°C (400°F)i8» 22. This does not preclude thepossibility of an earlier failure of the wetwell by overpressurization.An EPA typical of those installed in the Browns Ferry unit 1 drywell lineris shown in Fig. 9.22.
The masses of steam and hydrogen accumulated in the drywell and wetwell following core melt in the base case TB' are shown in Figs. 9.23through 9.26. For this case, wetwell failure by overpressurization is notpreducted to occur.
The drywell temperature responses for the sequences TB' and TUB ' areshown in Figs. 9.27 and 9.28 respectively; the general trends are verysimilar in the two sequences, except that high temperatures are reachedmuch sooner in sequence TUB', as would be expected.* The correspondingdrywell pressure responses for these two sequences are shown in Figs. 9.29through 9.36.
*In sequence TUB', it is assumed that the HPCI and RCIC systemsare unavailable from the inception of the Station Blackout. Thus boiloff,core uncovery, and the subsequent events would occur sooner.
(x 105)
77.0-
76.0-
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70.0
69.0
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PRESSURE INCREASE FOLLOWING MSIV CLOSURE
50.0 100.0
dc POWER LOST ,
150.0I
200.0
TIME (min)
250.0
Fig. 9.12 Reactor vessel pressure.
ORNL-DWG 81-8523 ETD
r-CORE UNCOVERY
300.0—I
350.0 400.0
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600.0 -r
500.0
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CORE SLUMPS INTO LOWER PLENUM
dc POWE R LOST
HPCI INJECTION STOPS-,
SRV FLOW AND HPCI INJECTIONBECOME BALANCED
100.0
—I
150.0 200.0
TIME (min)
250.0
CORE UNCOVERY
300.0T
350.0
Fig. 9.13 Steam mass flow rate from primary system.
400.0
to
120.0
100.0-
> 80.0-
cc
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or-
60.0-
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100.0 150.0I
200.0
TIME (min)
250.0
ORNL-DWG 81-8525 ETD
-dc POWER LOST
HPCI INJECTION STOPS
300.0 350.0—f
400.0
Fig. 9.14 HPCI flow rate into pressure vessel.
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(x 104)
30.0
25.0-
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I
250.0
ORNL-DWG 81-8526 ETD
LOSS OF dc POWER
BOILOFF BEGINS
CORE SLUMPS-
I
300.0I
350.0 400.0
Fig. 9.15 Mass of water in primary system.
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(x 103)
20.0-
18.0
16.0
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cc0_
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CORE UNCOVERY
dc POWER LOST
BOILOFF BEGINS
150.0I
200.0
TIME (min)
—I
250.0
Fig. 9.16 Steam mass in primary system.
ORNL-DWG 81-8527 ETD
CORE SLUMPS-
300.0—T—
350.0 400.0
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300.0
250.0-
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x 100.0
50.0-
0.0
0.0 50.0
H2 BUILDUP DUE TO Zr-H20 REACTION
100.0 150.0
Fig. 9.17 Hydrogen mass in primary system.
CORE UNCOVERY
1200.0
TIME (min)
I
250.0 300.0
ORNL-DWG 81-8528 ETD
CORE-JSLUMPS
350.0 400.0
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0.20
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0.0 50.0 100.0 150.0 200.0
TIME (min)
250.0
Fig. 9.18 Fraction of clad reacted.
ORNL-DWG 81-8529 ETD
o
300.0 350.0 400.0
00 50.0 100.0 150.0 200.0 250.0
TIME (min)
Fig. 9.19 Fraction of core melted.
300.0
ORNL-DWG 81-8530 ETD
00
350.0 400.0
(x 107)10.0
o 8.0-I-o<UJ
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200.0
TIME (min)
CORE UNCOVERY
250.0 300.0
Fig. 9.20 Energy generated from Zr-^O reaction.
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350.0 400.0
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800.0 -T
700.0-
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100.0-
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T T
100.0 200.0 300.0 400.0 500.0 600.0
TIME (min)
ORNL-DWG 81-8532 ETD
CONCRETE PENETRATION COMPLETED
I I
700.0 800.0—I 1 !900.0 1000.0 1100.0
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OJOJ
END SHIELDFOR RADIATION
PROTECTION
FIELD WELD
PRIOR TOINSTALLATION
SHIELDSUPPORT
FIELD WELD
CONTAINMENT
HEADER PLATES
ORNL-DWG 81-8533 ETD
CONNECTORMOUNTING
HOLES
FIELD WELD
VALVE ANDPRESSURE TAP
Fig. 9.22 Typical electrical penetration assembly canister.
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CORE UNCOVERY
100.0 200.0 300.0 400.0 500.0 600.0
TIME (min)
ORNL-DWG 81-8534 ETD
DRYWELL EPA SEALS FAIL
DRYWELL VENTING THROUGHFAILED EPA SEALS
—\r\l\ JLT T
700.0 800.0 900.0 1000.0 1100.0
Fig. 9.23 Mass of steam in drywell.
HOJOn
250.0
200.0-
5 150.0ooccQ>-
COCO
<
100.0-
50.0
DRYWELL H2 SWEPT INTO WETWELLAS VESSEL BOTTOM HEAD FAILS
Zr-H20 REACTION BEGINS-
0.0- T T
0.0 100.0 200.0 300.0 400.0
ORNL-DWG 81-8535 ETD
-SPIKE DUE TO H2 RETURN FROM WETWELL
FAILURE OF DRYWELL EPA SEALS
500.0 600.0
TIME (min)
i
700.0 800.0 900.0 1000.0 1100.0
Fig. 9.24 Mass of hydrogen in drywell.
OJON
9000.0
8000.0
7000.0
_ 6000.0-
< 5000.0
HC/3
O 4000.0-coCO
<
3000.0-1
2000.0-
1000.0
0.0
VESSEL BOTTOM HEAD FAILS
CORE SLUMPS
CONTINUED RELIEFVALVE ACTUATION
I CORE UNCOVERY
ORNL-DWG 81-8536 ETD
-FAILURE OF DRYWELL EPA SEALS
-n
1000.0 1100.00.0 100.0 200.0I I
300.0 400.0 500.0 600.0 700.0
TIME (min)
—I 1
800.0 900.0
Fig. 9.25 Mass of steam in wetwell.
OJ^J
800.0
700.0-
600.0-
UJ
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500.0
>I 400.0u.
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200.0-
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DRYWELL H2 SWEPT INTOWETWELL AS VESSELBOTTOM HEAD FAILS
Zr-H20 REACTION BEGINS-
1 I
0.0 100.0 200.0 300.0 400.0
M
T
500.0
TIME (min)
ORNL-DWG 81-8537 ETD
H, RETURNED TO DRYWELL AS EPASEALS FAIL AND CONTAINMENT VENTS
I I
600.0 700.0—I I
800.0 900.01
1000.0 1100.0
Fig. 9.26 Mass of hydrogen in wetwell.
OJoo
1600.0
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~ 1200.0cc
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<
£ iooo.osUJ
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800.0-
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INCREASE DURING
CORIUM-CONCRETEREACTION
VESSEL BOTTOM HEAD FAILS
T T
ORNL-DWG 81-8538 ETD
FAILURE OF DRYWELL EPA SEALS
T T T T T100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0 1000.0 1100.0
TIME (min)
Fig. 9.27 Drywell temperature distribution (TB').
r-1OJVO
1600.0-1
1400.0-
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INCREASE DURING CORIUMCONCRETE REACTION
h-vT
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HEAD FAILS-
FAILURE OF DRYWELL EPA SEALS
0.0T
100.0 200.0 300.0 400.0 500.0TIME (min)
Fig. 9.28 Drywell temperature distribution (TUB')
T I
600.0
ORNL-DWG 81-8539 ETD
700.0 800.0
o
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VESSEL BOTTOM HEAD FAILS-FAILURE OF DRYWELL EPA SEALS
100.0
ORE UNCOVERY
—T r 1 i ^ 1 1—200.0 300.0 400.0 500.0 600.0 700.0
TIME (min)
Fig. 9.29 Drywell total pressure (TB'),
T T
800.0 900.0 1000.0 1100.0
(x 105)8.0
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CC
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100.0i r i i^ i r
200.0 300.0 400.0 500.0 600.0 700.0 800.0
TIME (min)
Fig. 9.30 Drywell steam partial pressure (TB').
CORE UNCOVERY
T
ORNL-DWG 81-8541 ETD
FAILURE OF DRYWELL EPA SEALS
I I I900.0 1000.0 1100.0
ro
ORNL-DWG 81-8542 ETD
(x 104)10.0
COQ.
8.0-
uj 6.0-rra.
r-
CC
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4.0
2.0
0.0
H, SWEPT INTO WETWELL ASVESSEL BOTTOM HEAD FAILS-
Zr-H20 REACTION BEGINS-
FAILURE OF DRYWELL EPA SEALS
0.0 1000 200.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0TIME (min)
Fig. 9.31 Drywell hydrogen partial pressure (TB').
1000.0-I1100.0
H>OJ
(x 104)30.0-
« 25.00-
0.0
NON-CONDENSABLES SWEPT INTO WETWELLAS VESSEL BOTTOM HEAD FAILS
ORNL-DWG 81-8543 ETD
•FAILURE OF DRYWELL EPA SEALS
I 1 I I
0.0 100.0 200.0 300.0 400.0I I
500.0 600.0
TIME (min)
700.0 800.0 900.0 1000.0 1100.0
Fig. 9.32 Drywell non-condensables partial pressure (TB')
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^
100.0 200.0
-FAILURE OF DRYWELL EPA SEALS
300.0 400.0
TIME (min)
500.0 600.0
Fig. 9.34 Drywell steam partial pressure (TUB'),
ORNL-DWG 81-8545 ETD
700.0 800.0
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(x 104)8.0
roQ_
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H2 SWEPT INTOWETWELL AS
VESSEL BOTTOM
HEAD FAILS-
Zr-H20REACTION
BEGINS
200.0
FAILURE OF DRYWELL EPA SEALS
300.0 400.0
TIME (min)
500.0 600.0
Fig. 9.35 Drywell hydrogen partial pressure (TUB').
ORNL-DWG 81-8546 ETD
•vj
700.0 800.0
(x 104)25.0
£ 20.0-3COCOUJccQ.
^ 150
Zr-H20REACTION/BEGINS-
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<
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0.0 -f0.0 100.0
-NON-CONDENSABLES SWEPT INTO WETWELLAS VESSEL BOTTOM HEAD FAILS
I
200.0
FAILURE OF DRYWELL EPA SEALS
300.0 400.0
TIME (min)
I
500.0—T
600.0
Fig. 9.36 Drywell non-condensables partial pressure (TUB'),
ORNL-DWG 81-8547 ETD
I
700.0 800.0
oo
149
As the drywell ambient temperature exceeds 204°C (400°F), the EPAelastomeric sealing materials start to deteriorate and venting of thedrywell begins. When the ambient temperature has exceeded 260°C (500°F),the EPA elastomeric sealing materials have decomposed to such an extentthat the EPA seals are blown out of the containment wall by the elevatedpressure within the drywell. This greatly increases the containment leakage (Fig. 9.10).
9.5.2 Wetwell response
The condensation of the SRV steam discharge within the pressure suppression pool is generally accomplished without undue pressure stresses onthe containment wall. If, however, conditions are such that complete condensation of the steam discharge does not occur, pressure loads significantly in excess of design limits can result, leading eventually to wetwell rupture.
Relief valve steam discharge into the pressure suppression pool isaccompanied by pressure oscillations of varying characteristics which arefunctions of the steam mass flux and local pool water temperatures as wellas the type of sparger device installed at the discharge line terminus.
It has been observed that condensation instability can occur when asubmerged pipe vent discharges steam at flow rates higher than critical(sonic discharge) with a sufficiently high ambient pool temperature- theso-called "Wurgassen effect". The threshold of instability is characterized by an increase in the amplitude of pressure oscillations which normally accompany condensation at supercritical flow rates.
At present, ramshead-type sparger devices are installed on the reliefvalve tail pipe discharge lines at the Browns Ferry Plant. The ramsheaddevice consists of two 90° elbows welded back-to-back to form a modifiedT-junction, and provides an improvement in condensation performance overthat of a straight vertical pipe. The horizontal discharge from the ramshead allows the rising convection currents and induced secondary flows tocirculate cooler water around the steam plumes rather than stagnateagainst a downward flow of steam. However, small scale tests have shownthat it also has the potential to produce unstable condensation and concomitant large pressure stresses when sufficiently high pool water temperature is reached during supercritical (sonic) discharge. For this reason,certain limitations on plant operation are established in the Browns FerryTechnical Specifications to preclude any possible steam discharge throughthe ramshead devices at pool temperatures greater than 77°C (170°F).
In the near future, the ramshead devices at the Browns Ferry Plantwill be replaced by T-quencher sparger devices similar to that shown inFig. D-2 of Appendix D. Tests have shown that the T-quencher spargersproduce lower loads during the initial air-clearing upon relief valve actuation and permit smooth condensation of the steam discharge at pool temperatures up to 93°C (200°F). For this study, it was assumed that the SRVdischarge lines terminate in T-quencher spargers.
Although the use of T-quenchers considerably improves the steam condensation characteristics related to SRV steam discharges, the constrictedflows introduced by the T-quencher have caused insufficient thermal mixingin the suppression pool. The resultant thermal stratification in the pool
150
can produce much higher local water temperature in the vicinity of the T-quenchers although the average pool temperature remains quite low.
The problem of thermal stratification has been found to be signific
ant for the sequences TUB' and TUPB', in which the HPCI and RCIC systems are assumed inoperative. Without injection capability, core uncoveryand subsequent degradation occur early in the Station Blackout when thelevel of decay heat is relatively high so that a large amount of superheated steam and noncondensibles is discharged from the T-quenchers intothe suppression pool in a short period of time. The average suppressionpool temperature is plotted as a function of time for sequences TB' andTUB* in Figs. 9.37 and 9.38 respectively.
In sequence TUB', it is realistically assumed that without operatoraction, reactor vessel pressure control over the long term would be by repeated cycling of the same relief valve.* Because of the high steam massflux into the suppression pool bay in which the discharging T-quencher islocated, significant thermal stratification would be expected. MARCH computations show that the difference between the local and average suppression pool temperatures can be estimated to increase from about 5°C at thebeginning of the transient to about 40°C 100 minutes later. This meansthat the suppression pool would lose its condensation effectiveness; theresulting pressure loads from the SRV discharge of steam and noncondensibles would rapidly increase, leading to a possible rupture of the wetwellwhich could occur before the overtemperature-induced failure of the dry-well.
The drywell pressure signatures for the sequences TB' and TUB' areshown in Figs. 9.29 through 9.36. The general trends are again very similar, except for the timing of events. The pressure peaks for the sequenceTB' are generally higher than those for the sequence TUB'; this is attributed to the presence of a larger water inventory as a result of the four-hour period of HPCI injection during sequence TB'.
The wetwell temperature signatures are given in Figs. 9.37 and 9.38,and the average pressure distributions are given in Figs. 9.39 through9.46 for sequences TB-* and TUB'. The plotted pressures do not include theeffect of pressure waves resulting from condensation instabilities or thevapor jet plumes emanating from the T-quencher spargers.
*A1though four of the thirteen SRVs share the lowest nominal set-point of 7.72 MPa (1105 psig), practical considerations dictate that therewould be one lowest-set valve.
550.0
500.0-
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2 400.0 Hr-
CC
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300.0
FAILURE OF VESSEL BOTTOM HEAD
Zr-H20 REACTION BEGINS
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-i 1 1 1 1 I 10.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0
TIME (min)
ORNL-DWG 81-8548 ETD
900.0 1000.0 1100.0
Fig. 9.37 Wetwell temperature distribution (TB').
550.0 -r
500.0
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cc
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cc 450.0UJo.
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400.0-
350.0-
300.0
0.0
FAILURE OF VESSEL BOTTOM HEAD
-VENTING BY FAILURE OF DRYWELL EPA SEALS
Zr-H20 REACTION BEGINS
100.0 200.0—1
300.01
400.0
TIME (min)
500.0—I
600.0
Fig. 9.38 Wetwell temperature distribution (TUB').
ORNL-DWG 81-8549 ETD
to
700.0 800.0
(x 105)10.0-
9.0
« 8(HCDQ.
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Zr-H20 REACTION BEGINS
0.0 100.0 200.0 300.0 400.0
-VENTING BY FAILURE OFDRYWELL EPA SEALS
T
ORNL-DWG 81-8550 ETD
T T
500.0 600.0
TIME (min)
700.0 800.0 900.0 1000.0 1100.0
Fig. 9.39 Wetwell total pressure (TB'),
0/1OJ
(x 10425.0-1
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VENTING BY FAILURE OF
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ORNL-DWG 81-8551 ETD
ATMOSPHERIC PRESSURE
CORE UNCOVERY
T
100.0 200.0 300.0 400.0
—I I
500.0 600.0
TIME (min)
—I 1
700.0 800.0
Fig. 9.40 Wetwell steam partial pressure (TB')
1 I
900.0 1000.0 1100.0
On
(x 104)25.0
20.0-
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£ 15.0-cc
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Zr-H20 REACTION BEGINS
400.0 500.0
-VENTING BY FAILURE OF
DRYWELL EPA SEALS
ORNL-DWG 81-8552 ETD
0.0 100.0 200.0 300.0T
600.0
TIME (min)
—I 1 I I
700.0 800.0 900.0 1000.0 1100.0
Fig. 9.41 Wetwell hydrogen partial pressure (TB'),
On
(x 105)
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FAILURE OF VESSEL BOTTOM HEAD
Zr-H20 REACTION BEGINS-
ATMOSPHERIC PRESSURE
ORNL-DWG 81-8553 ETD
VENTING BY FAILURE OF
DRYWELL EPA SEALS
ALL NON-CONDENSABLESREPLACED BY STEAM
i 1 1 1 I I I I
0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0
TIME (min)
Fig. 9.42 Wetwell non-condensables partial pressure (TB').
—I
1000.0 1100.0
On
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200.0 300.0 400.0
TIME (min)
500.0
Fig. 9.43 Wetwell total pressure (TUB').
ORNL-DWG 81-8554 ETD
600.0 700.0 800.0
(x 104)6.0
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FAILURE OF
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0.0 100.0 200.0
VENTING BY FAILURE
OF DRYWELL EPA SEALS
ORNL-DWG 81-8555 ETD
WETWELL ATMOSPHERE IS NOT PURE STEAM
T TI
500.0300.0 400.0 500.0 600.0
TIME (min)
Fig. 9.44 Wetwell steam partial pressure (TUB').
I
700.0 800.0
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FAILURE OF
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ORNL-DWG 81-8556 ETD
VENTING BY FAILURE OF DRYWELL EPA SEALS
SOME HYDROGEN REMAINS IN WETWELL ATMOSPHERE
0.0
0.0• •II
100.0 200.0 300.0 400.0TIME (min)
500.0 600.0
Fig. 9.45 Wetwell hydrogen partial pressure (TUB')
I
700.0 800.0
On
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FAILURE OF
VESSEL
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0.0 100.0 200.0
ORNL-DWG 81-8557 ETD
VENTING BY FAILURE OF DRYWELL EPA SEALS
SOME NON-CONDENSABLES REMAIN IN WETWELL ATMOSPHERE
300.0I
400.0
TIME (min)
500.0 600.0 700.0 800.0
Fig. 9.46 Wetwell non-condensables partial pressure (TUB')
o>o
161
10. PLANT STATE RECOGNITION AND OPERATOR MITIGATING ACTIONS
10.1 Introduction
The accident signatures developed in Sect. 9 provide baseline information for six possible sequences leading to core meltdown that might occur during a Station Blackout. The objectives of this chapter are to examine these same sequences from the operator's standpoint in order toidentify and evaluate potential preventive and corrective actions keyed tothe time windows available.
As discussed in Sect. 9, a number of plant safety systems would function automatically during a Station Blackout. The automatic safety systemresponses include reactor scram, vessel pressure control by SRV operation,and level control through cycling of the HPCI system. Thus, assuming noindependent secondary failures, the reactor could be maintained in a safestable state without operator action until DC power is lost at an assumedfour hours into the blackout. After the loss of DC power, operator actionwould be crucial in attempting to avert the impending core degradation.
10.2 Plant State Recognition
To enable the operator to better identify and evaluate potentialmitigating actions during a Station Blackout, the accident progression iscategorized into a number of plant states. The available time windows foreach plant state are determined from the baseline information provided inSect. 9. The location of the plant states within each sequence is shownin Fig. 10.1, with the sequence chosen for the fission product transportanalysis of Volume 2 indicated by a dashed line.Plant State 1 (0-s). This is the initial state at the moment of loss ofoffsite and onsite power, and is common to all of the blackout sequences.Plant State 2 (0-625 s). This state represents the period of automaticplant safety system response and is common to the six sequences consideredin detail in this report.Plant State 3 (625 s-240 min). This state is applicable to sequencesTB-* and TPB'. It is characterized by the availability of adequate decay heat removal by relief valve operation and the HPCI and RCIC injectionsystems, which are dependent on DC power from the unit battery.Plant State 3A (625 s-240 min). This state is applicable to' sequencesTVB' and TVPB'. It is characterized by operator action to control levelby operation of the RCIC system (only) and to rapidly depressurize the reactor vessel by remote-manual relief valve actuation. The loss of fluidthrough the relief valve exceeds the capacity of the RCIC system for ashort period of time, which causes a momentary core uncovery early in thesesequences.
Plant State 3B (625 s-28 min). This state is applicable to sequences TUB'and TUPB'. In these sequences, the HPCI and RCIC systems are assumed to beinoperable from the inception of the blackout, because of equipment failure. This state is characterized by decreasing water level.Plant State 4 (240-295 min). This state is applicable to sequences TB' andTPB'. This is the period immediately following the loss of DC power,
INITIATING EVENT -
LOSS OF OFFSITEAND ONSITE
ac POWER
YES
A
VNO
SYSTEM
AUTOMATICRESPONSES
HPCI ANDRCIC
OPERABLE
ONLY RCIC
OPERABLE
MANUAL SRV
DEPRESSURIZATION
WITHIN 15 min.
LOSS OF dc POWER -HPCI/RCIC INOPERABLE
SRV
DEPRESSURIZATIONBEFORE LP ECCS
RESTORATION
| STATE 4
I
ISTATE(
LOW PRESSURE ECCS
OPERABLE - OFFSITE
ac POWER RESTORED
STATE 8A
STATE 4A
STATE 7ASTATE 3A
STATE 8B
STATE 5B
STATE 3B
Fig. 10.1 Operator key action event tree.
ORNL-DWG 81-8163 ETD
NO CORE DAMAGE (TPB')
CORE MELT ITPB)
CORE MELT (TB'l
NO CORE DAMAGE
NO CORE DAMAGE (T,B' AND T,PB')
CORE MELT IT„B' ANDTVPB')
NO CORE DAMAGE
POTENTIAL CORE DAMAGE
NO CORE DAMAGE (TUPB')
CORE MELT (TUPB'I
CORE MELT (TUB')
POTENTIAL CORE DAMAGE
163
power, characterized by the unavailability of the HPCI and RCIC systemsand a steadily decreasing reactor vessel water level.Plant State 4A (240-305 min). This state is applicable to sequencesTVB' and TVPB', and follows the loss of DC power. This state is characterized by a depressurized reactor vessel, unavailability of the HPCIand RCIC systems, and a steadily decreasing vessel water level.Plant State 5 (295-305 min). This state is applicable to sequence TPB'and is characterized by a stuck-open relief valve, so that the reactorvessel is depressurized.Plant State 5B (28-38 min). This state is applicable to sequence TUPB'.This sequence involves the combined secondary failures of inoperative HPCIand RCIC systems and a stuck-open relief valve (SORV) from the inceptionof the blackout. It is characterized by a rapidly decreasing water level.Plant State 6 (295-355 min). This state is applicable to sequence TB'only. It is characterized by boiloff of the reactor vessel water inventory due to automatic relief valve actuation with the reactor vessel fullypressurized, and terminates with the beginning of core melt.Plant State 6B (28-70 min). This state is applicable to sequence TUB'.It is characterized by a pressurized reactor vessel, core uncovery, andthe beginning of core melt.
Plant State 7 (305-360 min). This state is applicable to sequence TPB'.It is characterized by a depressurized reactor vessel, unavailability ofthe low pressure ECCS injection systems, decreasing water level, core uncovery, and the beginning of core melt.Plant State 7A (305-360 min). This state is applicable to sequencesTVB' and TVPB' is identical to state 7.Plant State 7B (38-360 min). This state is applicable to sequence TUPB'and is identical to state 7.
Plant State 8 (305-360 min). This state is applicable to sequence TPB'.It is characterized by a depressurized reactor vessel and the recovery ofAC power, which permits the restoration of normal vessel water level.Plant State 8A (305-360 min). This state is applicable to sequencesTVB' and TVPB' and is identical to state 8.Plant State 8B (38-360 min). This state is applicable to sequence TUPB'and is identical to state 8.
10.3 Operator Key Action Event Tree
For sequences TUB' and TUPB', and after the loss of DC power in thecase of sequences TB', TPB', TVB', and TVPB', mitigating action by theoperator is essential if core degradation is to be avoided and the plantbrought to a safe stable state.
Upon restoration of AC power, the operator's primary responsibilityis to ensure vessel depressurization and operation of the low pressure injection systems.
The operator key action event tree is shown in Fig. 10.1.
10.4 Operator Mitigating Actions
In addition to continued efforts to restore electrical power, therecommended actions after the loss of water injection capability when
164
boiloff and core damage become inevitable include the use of portablepumps, perhaps provided by fire engines, to flood the drywell in an attempt to preclude melt-through of the reactor vessel. If successful, thiswould keep the degraded core inside the vessel and prevent gross containment failure and the corresponding major releases of radioactivity. Forthe sequences TUB' and TUPB', in which injection capability is assumed lostat the inception of the blackout, a drywell flooding rate of 3,200 H/s(50,000 gpm) would be necessary to flood the drywell to the level of thereactor vessel core prior to reactor vessel melt-through.
165
11. INSTRUMENTATION AVAILABLE FOLLOWING
LOSS OF 250 VOLT DC POWER
Reactor vessel level and pressure control can be maintained during aStation Blackout for as long as 250 volt DC power from the unit batteryremains available, as discussed in Sect. 8. The instrumentation availableduring this initial phase of a Station Blackout was discussed in Sect. 5.It is the purpose of this section to discuss the instrumentation whichwould remain operational after the unit battery is exhausted; this finalphase of a Station Blackout would constitute a Severe Accident becausethere would be no means of injecting water into the reactor vessel tomaintain a water level over the core.
In addition to the 250 volt unit battery system, there are two smaller battery systems which supply power to Control Room instrumentation andalarm circuits. The first of these is a 24 volt DC system which suppliespower to the Source Range and Intermediate Range neutron flux monitors aswell as radiation monitors for the off-gas, RHR Service Water, Liquid Rad-waste, Reactor Building Closed Cooling Water, and Raw Cooling Water systems, none of which would be operational during a Station Blackout. The24 volt batteries for this system are designed to supply the connectedloads for a period of 3 hours without recharging. Since the 250 volt system batteries are expected to last for a period of four to six hours, itis unlikely that the 24 volt system would remain available after the 250volt DC unit distribution system has failed.
The second of the smaller battery systems is a 48 volt DC power supply and distribution system for the operation of the plant's communicationand annunciator systems. This system comprises three batteries, one ofwhich supplies the plant communications system while the remaining twobatteries are for the annunciator system. However, the system design provides that the total station annunciator load can be supplied from onebattery. The 48 volt DC system batteries are capable of supplying theconnected loads for a period of eight hours without recharging. This iswell beyond the period of expected operation of the 250 volt DC systemsand assures continued availability of the plant communications system.The efficacy of the annunciator system depends upon the availability ofthe power supplies to the signal transmission systems of the various sensors as well as the 48 volt DC system. Most of the alarm annunciator capability will be lost when the unit battery fails.
There is reason to believe that plant preferred 120 volt AC singlephase power would remain available for a significant period of time afterfailure of the unit battery. Plant preferred power is obtained from theplant 250-volt DC system during a Station Blackout by means of a DC-motor,AC generator combination. The plant battery is similar to each of the
unit batteries, but would be more lightly loaded during a Station Blackout. The major loads on the plant battery are the turbo-generator oilpumps and seal-oil pumps for the three Browns Ferry Units. These pumpscould be stopped within one hour after the inception of a Station Blackoutsince the turbo-generators will have stopped rolling and there is no jacking capability under Station Blackout conditions.
Assuming that plant preferred power does remain available during theperiod of core uncovery following the loss of the unit battery, two
166
sources of valuable information powered by this system and located outsideof the Control Room would remain. The first is an indication of the tem
peratures at various points within the drywell provided both by a recorderand an indicator-meter for which the displayed drywell temperature can beselected by a set of toggle-switches. *
The second information source provided by plant-preferred powerconsists of the temperatures at various points on the surface of the reactor vessel as provided by 46 attached copper-constantan thermocouples.Unfortunately, the indicating range for the instruments reporting thethermocouple responses is only 315.6°C (0-600°F). However, when the instrument responses for the thermocouples attached near or on the bottom ofthe reactor vessel are pegged at the high end, the operator can be surethat the core is uncovered, and the only fluid within the vessel is superheated steam. Both the drywell temperature indication and the reactorvessel thermocouple recorder are mounted on panel 9—47, which is locatedon the back of the Control Room Panels.
Information concerning the reactor vessel water level would also beavailable outside the Control Room from the mechanical Yarway indication(which requires no electrical power) located at the scram panels on thesecond floor of the reactor building; this would provide direct indicationof the reactor vessel water level over the range from 14.94 to 9.47 m (588to 373 in.) above vessel zero. The lower boundary of this Yarway indicating range is 0.33 m (13 in;) above the top of the active fuel in thecore.
There would be no indication of reactor vessel pressure after loss ofthe unit battery. However, it is expected that the vessel pressure wouldbe maintained in the range of 7.722 to 7.377 MPa (1105 to 1055 psig) byrepeated automatic actuation of a primary relief valve for as long as thereactor vessel remains intact.
It should be noted that the emergency lighting for the Control Roomis supplied from the 250 volt DC system; after failure of this systemhand-held lighting for the Control Room would be necessary. The door security system is supplied by plant preferred power and would remain operable as long as the plant 250 volt DC system is functional. The arearadiation monitors located throughout the plant are powered from theInstrumentation and Control buses and would not be operational from theinception of a Station Blackout.
167
12. IMPLICATIONS OF RESULTS
The purpose of this section is to provide a discussion of the present state of readiness at the Browns Ferry Nuclear Plant to cope with aStation Blackout. The discussion will include consideration of the avail
able instrumentation, the level of operator training, the existing emergency procedures, and the overall system design.
12.1 Instrumentation
The availability of Control Room instrumentation during the period ofa Station Blackout in which 250 volt DC power remains available has beendiscussed in Sect. 5. The most important parameters while injection capability remains are the reactor vessel level and pressure, and these wouldbe adequately displayed in the Control Room during this period. This instrumentation available after the loss of DC power was discussed in Sect.11.
The existing instrumentation at Browns Ferry Unit 1 which would beimportant during a Station Blackout is summarized below, with additionalinformation concerning the power supplies. The instrumentation is discussed in the order given in reference 28, and is located in the Control Roomunless otherwise indicated.
I. Core
A. Core Exit Temperature — not available at Browns Ferry.B. Control Rod Position— the rod position indicating system is
powered by the 120V AC unit-preferred system, which would beavailable during a Station Blackout for as long as 250V-CCpower remains available from the unit battery.
C. Neutron Flux _ The Source Range and Intermediate Range monitors are powered by the 24 volt DC battery system, which is designed to supply the connected loads for a period of 3 hoursunder the conditions of a Station Blackout. The Source and Intermediate Range detectors are withdrawn from the core duringpower operation to a point 0.61 m (2 ft) below the bottom of theactive fuel to increase their active life, and these detectorscould not be reinserted during a Station Blackout. However, therelative changes in neutron flux following reactor scram willproduce a proportional change in Source Range meter (range: 0.1to IO6 CPS) response when the detector is fully withdrawn.
II. Reactor Coolant System
A. RCS Pressure — one channel of reactor vessel pressure poweredby the feedwater inverter and another channel powered by theunit-preferred system, each with a range of 8.274 MPa (0 to 1200psig), would be available during a Station Blackout for as longas the unit battery continues to supply 250 volt DC power.
B. Coolant Level in the Reactor — two channels of reactor vessellevel instrumentation, one each powered by the feedwater inverter and the unit-preferred system would be available duringStation Blackout for as long as the 250 volt DC unit battery
168
system remains functional. These two channels each have a rangeof 1.524 M (0 to 60 in.) and provide indication of the reactorvessel water level between 13.41 and 14.94 m (528 and 588 in.)above the bottom of the vessel, a range which extends over theupper portion of the steam separators. Additional level information would be available outside of the Control Room, as follows:
o Two channels of Yarway level indication over the range from9.47 to 14.94 m (373 to 588 in.) above the bottom of the vessel are available at the backup control panel located in theBackup Control Room. This instrumentation is powered fromthe unit-preferred system; the lower limit of this indicationis 0.33 m (13 in.) above the top of the active fuel in thecore.
o The Yarway instruments are located at the scram panel on thesecond floor of the reactor building and mechanically displaythe reactor vessel water level over the same range as thatelectrically transmitted to the backup control panel. Thislevel indication at the Yarway instruments requires no electrical power, and would remain available even after all DCpower was lost.
C. Main Steamline Flow — two channels of steam flow instrumentation powered by the feedwater inverter would provide an indication of total steam flow [range: 2016 kg/s (0 to 16 x IO6lbs/h)] during a Station Blackout while the unit battery remainsfunctional.
D. Main Steamline Isolation Valves' Leakage Control System Pressure— not applicable to Browns Ferry.
E. Primary System Safety Relief Valve Positons, Including ADS —There is no provision for indication of the actual position ofany primary relief valve, including those assigned to the ADSsystem. Under normal operating conditions, the operator canidentify a leaking relief valve by means of the recorded tailpipe temperatures available on charts behind the Control Room
panels, or by the recently installed (December, 1980) acousticvalve monitors. Both of these leakage detection systems wouldbe inoperable under Station Blackout conditions.
On the other hand, remote-manual actuation of a reliefvalve is accomplished by energizing the associated DC solenoidoperator, and each valve has lights on the control panel whichindicate whether or not the solenoid for that valve is energized. The capability for this indirect indication of successful remote-manual valve actuation is maintained during the period of a Station Blackout in which 250 volt DC power is available from the unit battery.
F. Radiation Level in Coolant — not available at Browns Ferry.III. Containment
A. Primary Containment Pressure — Drywell pressure instrumentation with a range of 0.55 MPa (0-80 psia) powered by theunit-preferred system will be available during a Station
169
Blackout while the unit battery remains functional. A set of 12vacuum breakers ensures that the pressure in the pressure suppression pool torus cannot exceed the drywell pressure by morethan 0.003 MPa (0.5 psi).
B. Containment and Drywell Hydrogen Concentration — Instrumentsfor this purpose are provided for each Browns Ferry Unit, butwould not be available during a Station Blackout.
C. Containment and Drywell Oxgyen Concentration — Instruments
for this purpose are provided, but as in the case of the hydrogen monitors, are powered from the Instrumentation and Controlbuses which would not be available during a Station Blackout.
D. Primary Containment Isolation Valve Position — During a Station Blackout, valve position indication is maintained for theMain Steam Isolation Valves (MSIV's) and for the systems whichremain operable, i.e., the HPCI and the RCIC systems. However,valve position indication for the primary containment isolationvalves in the low-pressure injection systems which are poweredby sources not available during a Station Blackout would belost.
E. Suppression Pool Water Level — This instrument, powered bythe unit-preferred system, would indicate the water level in thetorus with a range of 1.27 m (—25 to +25 in.) as long as theunit battery is functional.
F. Suppression Pool Water Temperature — This instrumentationwould not be available during a Station Blackout.
G. Drywell Pressure — This is synonymous with "Primary Containment Pressure" — see "A" above.
H. Drywell Drain Sumps Level — not indicated. The frequency ofdrain pump operation for each of the two 1.893 m^ (500 gal.)drain sumps in the drywell is indicated in the Control Roomunder normal conditions, but would not be operable during a Station Blackout.
I. High-Range Containment Area Radiation — not available atBrowns Ferry.
IV. Power Conversion Systems
A. Main Feedwater Flow — one channel of flow instrumentation
with a range of 1008 kg/s (0 to 8 x IO6 lbs/h) and poweredby the unit-preferred system is provided for each of the threefeedwater pumps. This instrumentation would remain operable aslong as the unit battery is functional.
B. Condensate Storge Tank Level — The condensate storage tanklevel indication [range: 9.75 m (0 to 32 ft)] is powered by theunit-preferred system and would remain operational until theunit battery is exhausted.
V. Auxiliary SystemsA. Steam Flow to RCIC — This instrumentation is powered by the
Instrumentation and Control buses and would not be available
during a Station Blackout.
170
B. RCIC Flow — The RCIC pumped flow is indicated over a rangeof 0.044 m3/s (0-700 GPM) on the flow controller, which ispowered by the unit-preferred system and would be available aslong as the unit battery is functional.
VI. Radiation Exposure Rates — The permanently-installed radiationmonitors located throughout the plant are powered from the Instrumentationand Control buses and would not be operational during a Station Blackout.
The available Control Room instrumentation is adequate to monitor theplant response to a Station Blackout during the estimated four-hour periodduring which 250 volt DC power would remain available from the unit battery. After the unit battery is exhausted, virtually all Control Room instrumentation would be lost and water could no longer be injected into thereactor vessel to make up for that lost to the pressure suppression poolthrough relief valve actuation. Consequently, the reactor vessel waterlevel would slowly decrease until the core became uncovered.
The most important single parameter during the period of decreasingwater level following loss of the unit battery is the reactor vessellevel, which could be monitored at the Yarway instruments on the secondfloor of the Reactor Building as described in part III B above. In addition, the drywell ambient temperatures and the reactor vessel surface temperatures would remain available as long as the plant-preferred 250 volt-DC system is functional, as described in Sect. 11.
12.2 Operator Preparedness
There is currently no specific training provided Browns Ferry Operators in regard to plant response or required operator actions in the event of a complete Station Blackout.
A Station Blackout casualty can be run at the TVA Browns Ferry Unit 1simulator by selecting the pre-programmed loss-of-offsite-power casualtyand then having the simulator operator immediately punch the diesel-gener-ator manual trips in the control room. Since the instrumentation in thesimulated control room is powered from several diverse sources representative of those at the actual plant, the resulting effect closely models aStation Blackout. This procedure was developed and tested by TVA simulator operating personnel as an effort in cooperation with this study, buthas not been used for operator training.
There is no Emergency Operating Instruction to cover a Station Blackout at the Browns Ferry Nuclear Plant, but there are Emergency OperatingInstructions for the loss of individual subsections of the overall AC system.
With two exceptions, the current level of operator training and theexisting Emergency Operating Instructions are believed adequate to preparethe operator to cope with a Station Blackout during the initial phase inwhich 250 volt-DC power would remain available. The first exception involves the need to depressurize the reactor vessel within the first hour.The Technical Specifications provide that the reactor vessel must be depressurized to less than 1.48 MPa (200 psig) if the pressure suppressionpool temperature exceeds 48.9°C (120°F); this is based on a maximum permissible temperature of 76.7°C (170°F) for the pool so that complete condensation of steam is ensured in the event of a LOCA. The Emergency
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recommended order of removal of loads if the loss of AC power is perceived to be long-term.
3. Recovery procedures should be developed to provide a detailed methodfor recovery of vital power supplies and equipment in a safe, efficient manner upon restoration of AC power following a Station Blackout.
12.3 System Design
The existing system design provides sufficient instrumentation andequipment to maintain decay heat removal capability for several hours during a Station Blackout. The only questionable feature of the existing design with regard to the Station Blackout sequence is the provision forautomatic shifting of the HPCI pump suction to the pressure suppressionpool on high sensed pool level. As discussed in Sect. 8.1, this can leadto failure of the HPCI system because of the high temperature of the pressure suppression pool water at the time the shift in pump suction occurs.
Separate provision is made for an automatic shift of the HPCI pumpsuction if the normal condensate storage tank water source becomes exhausted. Thus it appears that the automatic high pool water level shiftmust have been straight-forwardly based on a concern for the effect ofhigh water level in the pressure suppression pool. The basis is not givenin the Technical Specifications, and it should be noted that there is nocorresponding provision for the RCIC system. It is recommended that thedesirability of an automatic shift in HPCI pump suction on high suppression pool water level be reexamined and if found necessary, that the basisbe included in operator training.
173
13. REFERENCES
1. Browns Ferry FSAR, Appendix F.
2. Nuclear Regulatory Commission Memorandum from H. R. Denton to Chairman Ahearne, dated September 26, 1980: Subject: STATION BLACKOUT.
3. Browns Ferry FSAR, Paragraph 8.5.2.
4. R. L. Scott, "Browns Ferry Nuclear Power-Plant Fire on March 22,1975," Nuclear Safety, 17-5, 592-611 (1976).
5. Browns Ferry FSAR, Response to AEC Question 4.8.
6. Staff Report, Anticipated Transients Without Scram for Light WaterReactors, NUREG-0460, Vol. 2, p. XVI-74.
7. Browns Ferry FSAR, Response to AEC Question 12.2.16.
8. Continuous System Modeling Program III (CSMP III) Program ReferenceManual, IBM Corporation, 4th Ed. (December 1975).
9. "Decay Energy Release Rates Following Shutdown of Uranium-FueledThermal Reactors," ANS-5.1, American National Standards Institute(1971).
10. R. T. Lahey, Jr. and P. S. Kamath, "The Analysis of Boiling Water Reactor Long-Term Cooling," Nuclear Technology, Vol. 49 (June 1980).
11. J. F. Wilson, R. J. Grenda, and J. F. Patterson (A-C), "Steam-WaterSeparation Above a Two-Phase Interface," Trans, Am Nucl. Soa.,4,356 (1961).
12. R. T. Lahey, Jr. and F. J. Moody, "The Thermal Hydraulics of a Boiling Water Nuclear Reactor," American Nuclear Society (1977).
13. R. 0. Wooton and H. I. Avci, MARCH Code Description and User's Manual, Battelle Columbus Laborataories/USNRC Report NUREG/CR-1711 (October 1980).
14. D. W. Hargroves and L. J. Metcalfe, CONTEMPT-LT/028: A ComputerProgram for Predicting Containment Pressure-Temperature Response to aLoss of Coolant Accident, EG&G Idaho, Inc./USNRC Report NUREG/CR-0255 (TREE-1279) (March 1979).
15. Browns Ferry Emergency Operating Instruction No. 36, Loss of CooTs-ant Accident inside Drywell*
16. S. E. Mays, et al, Second IREP Status Report, Risk Assessment forBrowns Ferry Nuclear Plant Unit 1, January 1981 (Preliminary).
174
17. Browns Ferry FSAR, Section 8.6.
18. D. D. Yue and W. A. Condon, "Severe Accident Sequence Assessment ofHypothetical Complete Station Blackout at the Browns Ferry NuclearPlant," 1981 International ANS/ENS Topical Meeting on ProbabilisticRisk Assessment, Port Chester, N.Y., September 20-24, 1981.
19. Reactor Safety Study, WASH-1400 (NUREG-75/014), U.S. Nuclear Regulatory Commission, 1975.
20. F. E. Haskin, et al, "Analysis of a Hypothetical Core Meltdown Accident Initiated by Loss of Offsite Power for the Zion I PWR," NUREG/CR-1988, Sandia National Laboratories, 1981.
21. M. L. Corradini, "Analysis and Modelling of Steam Explosion Experiments," NUREG/CR-2072, Sandia National Laboratories, April 1981.
22. D. D. Yue, "Electric Penetration Assembly," U.S. Patent 4, 168, 394,September 1979.
23. American Nuclear Society Proposed Standard ANS-5.1, "Decay EnergyRelease Rates Following Shutdown of Uranium — Fueled ThermalReactors," October 1971, revised October 1973.
24. ANSI/ANS-5.1 (1979), "Decay Heat Power in Light Water Reactors,"August 1979.
25. W. B. Wilson, et al, "Actinide Decay Power," LA-UR 79-283, Los AlamosScientific Laboratory, June 1979.
26. T. E. Lobdell, "REDYV01 User Manual," NEDO-14580, August 1977.
27. D. D. Christensen, "Interim Browns Ferry RELAP4/MOD7, Station Blackout Calculation," DDC-2-80 Idaho National Engineering Laboratory,October 31, 1980.
28. Table 3, BWR VARIABLES, of Revision 2 to Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to AccessPlant and Environs Conditions During and Following an Accident"(December 1979).
29. Browns Ferry FSAR, Paragraph 5.2.1.
30. Safety Evaluation Report — Mark I Containment Long-Term Program,NUREG-0661, July 1980.
31. Three Dimensional Pool Swell Modeling of a Mark I Suppression System,EPRI NP-906, October 1978.
32. H. Nariai, I. Aya, M. Kobayaski, "Thermo-Hydraulic Behavior in aModel Pressure Suppression Containment During Blowdown," Topics inTwo-Phase Heat Transfer and Flow, pp. 89-98, Winter ASME Meeting,1978.
175
33. W. G. Anderson, P. W. Huber, A. A. Sonin, "Small Scale Modeling ofHydrodynamic Forces in Pressure Suppression Systems," NUREG/CR-0003,March 1978.
34. B. D. Nichols, C. W. Hirt, "Numerical Simulation of Boiling WaterReactor Vent-Clearing Hydrodynamics," Nucl. Sci. &Engr., 73,196-209.
35. D. Brosche, "Model for the Calculation of Vent Clearing Transients inPressure Suppression'System," Nucl, Engr. & Des., 38, 131-141,1976.
36. J. S. Marks, G. B. Andeen, "Chugging and Condensation OscillationTests," EPRI NP-1167, September 1979.
37. L. E. Stanford, C. C. Webster, "Energy Suppression and Fission Product Transport in Pressure-Suppression Pools," ORNL-TM-3448, April1972.
38. M. Okazaki, "Analysis for Pressure Oscillation Phenomena Induced bySteam Condensation in Containment with Pressure Suppression System,"Nucl. Sci. and Tech. iA» PP« 30-42, January 1979.
39. J. S. Marks, G. B. Andeen, "Chugging and Condensation Oscillation,"Condensation Heat Transfer, pp. 93-102, 18th National Heat TransferConference, San Diego, California, August 1979.
40. W. Kowalchuk, A. A. Sonin, "A Model for Condensation Oscillations ina Vertical Pipe Discharging Steam into a Subcooled Water Pool,"NUREG/CR-0221, June 1978.
41. D. A. Sargis, J. H. Stubhmiller, S. S. Wang, "A Fundamental ThermalHydraulic Model to Predict Steam Chugging Phenomena," Topic in Two-Phase Heat Transfer and Flow, pp. 123-133, Winter ASME Meeting,1978.
42. P. A. Sargis, et al., "Analysis of Steam Chugging Phenomena, Vol. 1,"EPRI-NP-1305, Janauary 1981.
43. B. J. Patterson, "Mark I Containment Program: Montice-llo T-QuencherThermal Mixing Test Final Report," NEDO-24542, August 1979.
44. "TRAC-P1A An Advanced Best-Estimate Computer Program for PWR LOCAAnalysis," Los Alamos Scientific Laboratory, NUREG/CR-0665, LA-7777-MS, May 1979.
45. U.S. Rohatgi and P. Saha, "Constitutive Relations in TRAC-P1A,"Brookhaven National Laboratory, NUREG/CR-1651, BNL-NUREG-51258,August 1980.
176
46. "Application of the Coupled Fluid Structure Code PELE-IC to PressureSuppression Analysis — Annual Report to NRC for 1979," LawrenceLivermore Laboratory, NUREG/CR-1179, UCRL-52733.
47. C. W. Hirt, B. D. Nichols, and N. C. Romero, "SOLA — A NumericalSolution Algorithm for Transient Fluid," Los Alamos ScientificLaboratory, LA-5852 (1975).
48. Browns Ferry Startup Test Instruction No. 15, High Pressure Injection System.
49. Flow Diagram Condensate Storage and Supply System, DWG No. 47W818-1RIO.
50. Browns Ferry System Operating Instruction No. 64, Primary Containment Unit I, II, or III.
51. Browns Ferry FSAR Section 7.3, Primary Containment and ReactorVessel Isolation Control System.
52. Browns Ferry System Operating Instruction No. 73, High PressureCoolant Injection System Unit I, II, or III.
53. Browns Ferry Startup Test Instruction No. 14, Reactor Core IsolationCooling System.
54. Browns Ferry Nuclear Plant Hot License Training Program, Volume 4,Reactor Core Isolation Cooling System.
55. Browns Ferry FSAR Section 4.7, Reactor Core Isolation Cooling System.
56. Browns Ferry FSAR Figure 4.7-3, Reactor Core Isolation Cooling System Process Diagram.
57. Browns Ferry System Operating Instruction No. 71, Reactor Core Isolation Cooling Unit I, II, or III.
58. B. J. Patterson, "Mark I Containment Program: Monticello T-QuencherThermal Mixing Test Final Report," NEDO-24542, August 1979.
177
ACKNOWLEDGEMENT
Mr. William A Condon made a significant contribution to this report byperforming the MARCH code calculations and preparing the associatedplots for the six station blackout accident sequences discussed inChapters 9 and 10. This accomplishment is especially noteworthy becauseall MARCH computer runs in support of this project were by necessitymade remotely on the CDC computer facilities at Brookhaven NationalLaboratory. Mr. Condon's work was in partial fulfillment of the requirements for the degree of Master of Science with major in NuclearEngineering at the University of Tennessee, and his lone hours ofdedicated effort are gratefully acknowledged.
179
APPENDIX A
LISTING OF PLANT RESPONSE CODE BWR-LACP
181
**$CONTINUOUS SYSTEM MODELING PROGRAM III V1M3 TRANSLATOR OUTPUT$*$
LABEL BWR-LACP CCJOE
LABEL CONTAINMENT MODEL WITH i-NOOE POOL*
* MACRO SUBPROGRAM FOR POOL WATER TEMP CALCULATIONMACRO XNQI,TI,WEI.MWI,LWI-«W(TL,TRILWL,LWR,TWIO,FWI,FSVI,FTEItFOI,FSIJ
* MACRO TC CALCULATE MASS ENERGY BALANCES FOR SP WATER NOOE I* INPUTS :
* TL»TR«TEMPS OF WATER IN ADJACENT NODES* LWL,LWR=LEVELS OF ADJACENT NODES
* TWIO-INITIAL WATER TEMP CF NCDE I* FWl-FRACTION OF TOTAL SP WATER VOL IN NCDE I(CONSTANT)* FSVI-FRACTION OF SAFETY VALVE FLOW DIRECTED TO NODE I* FTEI"FRACTION OF TURBINE EXHAUST DIRECTED TO NOOE I* FDI-*FRACTION OF PUMP OUSCHARGE DIRECTED TG NODE I* FSI*FRACTlON OF PUMP SUCTION TAKEN FROM NODE I
* MASS BALANCE*
* WSSPW*PUMP SUCTION FLCWCTOTAL FROM SP)* WOSPW-PUMP OISCHARGE FLOWCTOTAL TO SP)* OUTPUTS:
* XNQI»FRACTIN
* XNQI-FRACTGON OF STM DISCH TC NOOE I NOT QUENCHED* XNQINFRACTION OF STM DISCH TO NOOE I NOT QUENCHED* TI*AVG. WATER TEMP OF NODE I (DEG-F)* WE I-RATE OF EVAPORATION FROM NOOE I (LB/SEC)* MWI»MASS OF WATER IN NODE I (LB)* LWI-WATER LEVEL OF NODE I (IN)*
* VWSPOINITIAL TOTAL SP WATER VOLUME (CU. FT. )MWIO-»FWI*VWSPO*(63.9-.019*TWI0)
CMWI-MNTGRL(0.,XQI*(FSVI«WSSV*FTEI*WSTE)*FINI-F$I*WSSPW-WEI)FINI*ta£CL+WEQR+FWI*WCSPG+FDI*WOSPWMWI-*MWIO+CMWI
VWI*MWI/(63.9-.0l9*TI)
LWI*AFGEN(SPLEV,VWI/FWI)*
* CALC. QUENCH FRACTION FROM ASSUMPTION OF NO QUENCH WHEN NOOE VAPOR* PRESSURE EQUAL TC TOTAL GAS PRESSURE
PVAPI-NLFGEN(SPFOT.TI)
XQI*LIMIT(0.,l.,(PTSPG-PVAPI-OPQZ)/DPQR)XNQIM1.-XQI)
* SPFOT(TI)=SAT. PRESS AS F OF Tl
* DPQZ*DIFFERENCE(PSI) OF PTSPG OVER PVAPI REQUIRED FOR ANY* QUENCHING TO TAKE PLACE* DPZR»RANGE(PSI) CF PTSPG OVER PVAPI CVER WHICH QUENCHING* GOES FROM 0.0 TO 100PERCENT*
* EVAPORATION RATE BASED ON ECUN.13-33, KREITH'S 'HEAT TRANSFER'* PSSPG=STM. PART PRESSURE ON SP GAS* ASSPW-SURF AREA OF SP WATER(FT.SQ)
WEI*(WEIN/WEID>*CCMPAR(PVAPI,PSSPG)
WEIN*ASSPW*FWI*HSI*12.3*(PVAPI-PSSPG)*PTSPG*VGSPWEID-(PNSPG*2.-PVAPI)*TGSPR*(MSSPG+MNSPG)HSI»5.83E-05*((ABS(TI-TGSP))**.333)
*
* EQUALIZATION FLOWS FROM ACJACENT NODE(S)* FWE«EQ. FLOW CONSTANT (LB/SEC/(INCH LEVEL DIFF) )
WEQL~FWE*(LWL-LWI)WEQR-»FWE*(LWR-LWI )
* ENERGY BALANCE* HEAT TRANSF RATES TO METAL ANO ATMCSPHERE NEGLEGIBLE
182
* WITH RESPECT TO MASS-MIXING ENERGY TRANSFERCMHWI=INTGRL{0.,X«I*<FSVI*WSSV*HST+WSTE*FTEI*HSTE)*E1I+E2I>
EII*(TLX-32.)*WEQL+(TRX-32.)*WECR+WML*(TL-TI)*WMR*(TR-TI)E2I=FWI*WCSPG*HFGSP+FDI*WOSPW*HD-FSI*WSSPW*(TI-32.)-WEI*1105.TLX=TL*COMPAR(LWL,LWI)*TI*COMPAR(LWI,LWL)TRX=TR*C OMPAR(LWR,LWI) +TI*COMPARILWI,LWR)
* CALC OF MIXING FLOWStTHEY DID NOT APPEAR IN THE MASS BALANCE BECAUSE* ZERO NET MASS TRANSFER IS ASSUMED
WML*KMIX*SQRT(ABS(TL-TI))WMR*KMIX*SQRT(A8S(TR-TI)>
*
* TEMPERATURE CALCULATION ASSUMES CONSTANT SPECIFIC HEAT=1.3 FOR WATERTI=32. ♦ UTWI0-32.)*MWI0 ♦ CMHWI)/MWI
*
ENDMACRO
*
INITIAL
*
* INITIALIZATION FOR REACTOR VESSEL CALCULATION*
* CONSTANTS FOR REACTOR MODEL*
* ACOP*CORE OUTLET PLENUM FLOW AREA(FT**2)* ACOR=CORE FLOW AREA(FT**2)» ART-RISER TUBE FLOW AREA(FT**2)* CPO-PRE-TRIP CORE POWER(TOTAL) (MWTH)* EQXO-FRACTION OF WAY TO SATURATION THAT STEAM CONTACT RAISES* INJECTION WATER IF DC LEVEL IS AT LEVEL OF JET PUMP SUCTION* FFLASH-FRACTION FLASHED/SEC PER BTU/LB ABOVE SATURATION* HCIO-INITIAL CORE INLET ENTHALPY(BTU/LB)* HINJIN-ENTHALPY OF INJECTION FLUID(BTU/LB)* LBOT-HEIGHT OF STM SEP BOTTOH(FT)* LOCO-INITIAL DOWNCOMER LEVELIFT) (HEIGHT ABOVE BOT. OF ACT. FUEL)* LHEDER-HEIGHT OF FW HEADER ABOVE BOAF(FT)* LOP»AVG. LENGTH OF CORE OUTLET PLENUM(FT)* LRT*AVG. LENGTH OF RISER TUBES(FT) (STANDPIPES)* PCOR-CORE HEAT TRANSFER PERIMETER(FT)* PO-INITIAL REACTOR VESSEL PRESSURE(PSIA)* RCICfX=NCMINAL RCIC FULL FL0WIL8/SEC)* TAULEN-STABILITY TIME CONSTANT FOR REGICN AVERAGE HEAT FLUX CALC(SEC)* TCFUEL*TIME CONSTANT TO ACCOUNT FOR RESIDUAL HEAT IN CORE FOR* INITIALIZATION CLOSELY FOLLOWING SCRAM
* TO-TIME AFTER TRIP THAT TRANSIENT INITlATES(SEC)* VOLP*LOWER PLENUM VOLUME(FT**3J
* V1DC-VOLUME OF DOWNCOMER BETWEEN BOAF AND PUMP DIFFUSER EXIT(FT**3)* WINJO'INITIAL INJECTION FLOWUB/LEC)*
* REFERENCE PARAMETERS FOR NATURAL CIRC LOSS COEFF CALCULATION* CPR-RATEC CORE POWER(MWTH)
* HREF»REFERENCE CORE INLET ENTHALPYIBTU/LB)* LDCR-REFERENCE CCWNCCMER LEVELd.E. NORMAL LEVEL AT STM SEP MIDDLE)* PRELR-REFERENCE RELATIVE CORE PCWER* PRR*REFERENCE REACTOR VESSEL PRESSURE* RHOO-DENSITY OF STEAM!LB/FT**3) IN R.V. AT INITIAL PRESSURE=PO* WGUESS*GUESS ON FLCWUB/SECJ FOR ITERATIVE SOLUTION FOR INITIAL FLOW* WREF-REFERENCE FLOW(LB/SEC)* XREF«REFERENCE FRACTIONAL QUALITY*
* PARAMETERS FCUNO IN FUNCTION SUBPROGRAMS* VOIFF-JET PUMP VOL BELOW BOAF(FT**3)* VFREE*TOTAL STM. VOL. IN R.V.(LESS LOWER PLENUM,CORE,CORE OUTLET PLENUM* AND STM. SEPARATORS) AND MAIN STEAMLINES TO ISOLATION VALVES (FT*->3)* VJET-VOL BETWEEN JP SUCTION ANO BOAF(FT**3)* VANN-VOL BETWEEN TOP OF CORE OUTLET PLENUM AND JP SUCTION(FT**3)* VSSOP-VOL BETWEEN TOP OF CO PLENUM AND BOTTOM OF STM SEP(CU.FT.)* WRATED-STM. FLOW THUR 1 SRV (LB/SEC) WHEN PRESSURE»PRATED (PSIA)* XKVGJ-EMPIRICLE CONSTANT USEO IN CALCULATION OF DRIFT VELOCITY
183
* XL1=0=ZER0 PCINT FOR DC HEIGhT(FT ): COINCODES WITH BOAF
* XL2=HEIGHT AT JET PUMP SUCTICN(FT)* XL3=HE1GHT AT TOP OF CORE OUTLET PLENUM(FT)* XL4-HEIGHT AT BOTTOM OF STEAM SEPARATORS!FT)
CONSTANT ACCR=82., AC0P=234., ART=42.3CONSTANT CP0=3293., EQXO=.5, FFLASH=.001CONSTANT LHEDER=23.4, LOP=5.58, LRT=1C0CONSTANT LB0T=24.85
CONSTANT PC0R=5518., PRATED=1095.CONSTANT TAULEN=.75, TCFUEL=9.5CONSTANT VFREE=13000., VCLP=3350., V1DC=192.,WRATED=259.
*
* CARD TO INITIALIZE AT ARBITRARY TIME POINTCONSTANT HC10=312.05,LDC0=28.63,PO=125.,T0=14430.,RH0G0=.2788
* INITIAL CORE FLOW GUESS DEPENDS ON LCCO AND PRELO
CONSTANT WGUESS=8310.* REFERENCE PARAMETERS FOR CALCULATION OF NATURAL CIRCULATION* FLOW RESISTANCE COEFFICIENT
CONSTANT CPREF=3293., PRELR=.32, PRR=1020LDCR=27.58, XREF=.133, WREF=9111., HREF=522.
* RUN CONTROL PARAMETERS
CCNSTANT HINJIN=58.,RC ICMX=82.9876*
* FUNCTION SUBPROGRAMS
* CAVOID(XIN,XCUT,WTOTAL,TSAT,RHOF,RHOG)* DKFUN(T)=P/PO AS A FUNCTION OF TIME AFTER SHUTDOWN* QREGAVJBCTTGM,TOP,KAPPA,PEKAVG)=(AVG HT. FLUX FROM A TO B)/(CORE AVG)* RH0TP(RH0F,RHOG,V0ID)=2-PHASE DENSITY* XLENDC(V0L0CI=00WNC0MER LEVEL AS FUNCTICN OF LIQUID VOLUME* VOID(QUALITY,TOTAL FLOW,FLOW AREA,TSAT.RHOF,RHOG)=POINT VOID FRACTION* VOLDC(LCC)=DCWNCCMER HEIGHT ABOVE BOT CF ACT. FUEL*
* STEAM TABLE CSMP INTERPOLATION FUNCTIONS* VSATF(PRESS)=SAT FLUIO SPECIFIC VOL(FT**3/LB )
FUNCTICN VSATF=15.,.0167, 50.,.0173, 100.,.0177, 200.,.0184,...400.,.0193, 600.,.0201, 800.,.0209, 1000.,.0216, 1203 ., .0223,.. .1400.,.0231
*
* VSATG(PRESS)=SAT GAS SPECIFIC VCL(FT**3/LB)FUNCTION VSATG=15.,26.3, 50.,8.51, 75.,5.81,100.,4.43,...
150.,3.01,200.,2.29,400.,1.16, 600.,.77,800.,.569,...
1000.,.446,1200.,.362,1400...302*
* HSATF(PRESSURE)-SAT FLUID ENTHALPY(BTU/LB)FUNCTION rSATF=15.,181., 50.,250., 100.,298., 2C0.,355.,...
400. ,424., 600.,472., 800.,510., 1000.,543., 1200.,572.,...1400.,599.
*
* HSATG(PRESSURE) = SAT GAS ENTHALPY IBTU/LB)FUNCTION HSATG=15.,U51., 50.,1174., 100.,1187., 200.,1198. ,...
400. ,1205., 600.,1204., 800.,1199., 1000.,1193., 1203 .»1185 . , . ..1400.,1175.
*
* TSATM(PRESSUREJ=SAT MIXTURE TEMPERATURE(DEG-F)FUNCTION TSATM=15.,213., 50.,281., 100.,328., 2C0.,382.,...
400.,445., 600.,486., 800.,518., ICOO.,545., 1200.,567.....1400.,587.
*
* VSC(ENTHALPY)=SUBCOOLED FLUIC SPECIFIC VOL(FT**3/LB )FUNCTICN VSC=21.4,.016, 121.,.0163, 221., .0169, 272.,.0174,...
376.,.0185, 431.,.0193, 488.,.0203, 549.,.0217, 562. , .0221,.. .575.,.0224
* CALCULATION CF INITIAL SATURATION PROPERTIESRH0F0=1./(AFGEN(VSATF,P0))
HFO=AFGEN1HSATF,PO)HGO=AFGEN(HSATG, PO)TSAT0=FUNGEN(TSATf,2,P0)
* SUBCOOLED WATER DENSITYRHOSC0=l./(AFGEN(VSC,HCI0))
184
* CALCULATION OF FLOW RESISTANCE COEFFICIENT* NEEDS: REFERENCE CONDITIONS FOR NATURAL CIRCULATION:CPREF,PRELR,PRR,XREF,* WREF, HREF, LDCR* RETURNS: LOSS COEFFICIENT: KLOSSU* CALCULATION OF REFERENCE DENSITIES AND SATURATION PROPERTIES
HFR*AFGEN <HSATF,PRR)TSATR«FUNGEN(TSATM,2,PRR)RHOFR=l./(AFGEN(VSATF,PRR))RhOGR=l./(FUNGEN(VSATG,2,PRR)>RHOSCR=l./(AFGEN(VSC,HREF))QT0TR=CPREF*PRELR*947.8
* THIS EXP FOR LSC ASSUMES AVG POWER=CORE AVG IN SC REGIONLSCR*12.*WREF*(HFR-HREF)/QT0TRLBR«12.-LSCR
RSCAVR=RHOFR/2.+RHOSCR/2.
* FUNCTION SUBPROGRAM CAVOIO RETURNS AVERAGE CORE BOILING REGION* VOID FRACTION FRCM THE GIVEN INPUTS
CAVR-CAVOID(0..XREF,WREF,TSATR,RHOFR,RHOGR)RHOBR=RHOTP(RHOFR,RHOGR,CAVR)RCC0RR=RSCAVR*LSCR/12.+RHCBR*L8R/12.RHOPR=RHOTP(RHOFR,RHOGR,OPVR)
* FUNCTION VOID RETURNS POINT VOID FRACTION FROM GIVEN INPUTSOPVR=»VCID(XREF,WREF, ACOP.TSATR, RHOFR, RHOGR)RGRTR=RHOTP(RHOFR,RHOGR,RTVR)RTVR* VOIDUREF, WREF, ART, TSATR, RHOFR, RHOGR)RDP*LCCR*RH0SCR/144.-R0RTR»LRT/144.-...
RH0PR*L0P/144.-R0C0RR*12./144.KLOSSU=ROP/(WREF*WREF)
*
* INITIAL RELATIVE POWER*
PREL0=(DKFUN(T0)*.94*EXP(-T0/TCFUEL))
QT0T0=PREL0*CP0*947.8*
* CALCULATION CF INITIAL TOTAL FLOW BY ITERATIVE SOLUTION, WITH* STARTING FLOW GUESS INPUT BY USER
WCIQ-IMPHWGUESS, .OOl.WCAL)
XOU»(CTOTO/WCIO+HCIO-HFO)/(HGO-HFO)XO*LIMIT(0.001,1.00O,X0U)
* THIS EXP FOR LSC ASSUMES AVG POWER=CORE AVG IN SC REGIONLSC0»LIMIT(0.,12.,12.*WCI0*IHF0-HCI0)/QT0T0)LBO-12.-LSCO
* CALCULATION OF INITIAL DENSITIESRSCAV0=RHCFO/2.+RHOSCO/2.waoo~wcio*xo
* VOID AT BOTTOM AND TOP OF BOILING REGION CALC. BY DFVOID* FUNCTION SUBPROGRAM, THEN AVERAGED TO GIVE CORE AVERAGE VOID
CAVINO=DFVOID(ACOR,WCIO,0.,TSATO,RHOFO,RHOGO)CAVEXO=DFVOIO(ACOR,WCIO-WBOC,WBOO,TSATO,RHOFO,RHOGO)CAVO=(CAVINO+CAVEXO)/2.RHOBO=RHOTP(RHCFO,RHOG0,CAVO)ROC0RO=RSCAV0*LSC0/l2.'>RHCBO*LBO/12.ROPO-RHOTP(RHOFO.RHOGO.OPVO)OPVO=CFVOIO(ACOP,WCIO-WBOO,WBOO,TSATO,RHOFO,RHOGO)RCRT0=RHQTP(RHOFO,RHOGO,RTVO)RTVO-DFVOID(ART,WCIO-WBOO,WBOO,TSATO.RHOFO,RHOGO)
* NET GRAVITY PRESS DROP!PSI) AVAILABLE FOR UNRECOV. DROP ACROSS CORELDCX=LIMIT(0.,LOCR,LDCO)UNREC0=LDCX*RH0SC0/144. + (12.+LRT+L0P-LDCX)*RH0G0/144.-...
RORT0*LRT/144.-R0P0-»LOP/l44.-ROCORC*12./l44.DPO=LIMIT(.0001,1000..UNRECO)WCAL-»SQRT(DP0*RHOB0/( KLOSSU*RHOBR) )
* ENO ITERATIVE LOOP
* OOWNCOMER AND LOWER PLENUM INITIALIZATION* VOLUME CF WATER IN DOWNCOMER NODE GIVEN BY FUNCTION
185
* VOLDCl ) AS A FUNCTICN OF WATER LEVELMOCO» VOLDC(LDCO)*RHOSCOMHOCO=MDCO*HCIO
MLPO*VOLP*RHCSCOMHLPO=MLPO*HCIO
*
G=4.75*ART*RH0G0MTOTO*(LSCO*RSCAVO+LBO*RHOBO)*ACOR-t-LCP*ACCP*ROPO+LRT*ART*RORTO*G
*
* PRESSURE CALCULATION INITIALIZATIONRHOSVR=l./FUNGEN(VSATG,2,PRATED)UCSRV»WRATED/SQRT(PRATED*RHOSVR)MSTO-IVFREE-VOLDCUDCO) )*RHOGOUMST0»MST0*(HG0-P0*144./(778.*RH0G0))
*
* INITIALIZATION FOR SUPPRESSION PCOL CALCULATION*
*
* CONSTANTS FCR CONTAINMENT MODEL
*
* AOMET=HEAT TRANS AREA BET DW MET AND CW ATMOS(FT**2J* APMET=HEAT TRANSF AREA BET SP MET AND SP ATMOS* ASSPW=AREA OF POOL WATER SURFACE(SQ.FT.)* BDWSPO*CU.FT/SEC/PSI FLOW WHEN DOWNCCMERS CLEARED* BSPDWOCU.FT/SEC/PSI OF FLOW WHEN VALVE OPEN* CDMET=MASS*SPEC HEAT OF OW METAL(BTU/DEG-F)* CPAIR=MASS*SPEC HEAT OF AIR IN SP CHAMBER ROOM* CPMET=MASS*SPEC HEAT OF SP METAL IN CONT WITH GAS* DM-AMOUNT OF STM. QUENCHED!UNIFORMLY ARCUNO POOL) IF THERE* IS ELAPSED TIME BETWEEN NOMINAL START AND INITIALIZATION OF THE RUN* DPQR=RANGE(PSI) OVER WHICH QUENCH FRACTION GOES FROM 1 TO 0.* WITH INCREASING VAPOR PRESSURE* DPQZ=MARGIN(PSI) ABOVE SATURATION FOR COMPLETE QUENCHING* FDI,FSI=FRACTION OF POOL COOLING DISCHARGE AND SUCTION TO POOL NODE I* FWE=FLOW BETWEEN S.P. NODES(LB/SEC) PER INCH OF WATER LEVEL OIFFEKENCE* FW INFRACT ION OF POOL WATER CONTAINED IN POOL NODE I* FSVI*FRACTIQN OF TOTAL SRV FLOW DISCHARGED TO POOL NODE I* FTEI=FRACTION OF TURBINE EXHAUST DISCHARGED TO POOL NODE I* FLSPG, FLOWG=FRACTION OF ATM. LEAKED TO RX-BLDG. PER SECOND* GCH=GAS CONSTANT OF H2* GCM*GAS CONSTANT OF MIMISC.)* GCN~GAS CONSTANT OF N2* GCS-GAS CONSTANT OF H20 VAP(PSI*FT**3/LB*DEG-R)* HSRREF=OELTA TO REREFERENCE STM ENTHALPY FROM ASME STEAM TABLES* TO PERFECT GAS EXP. THAT HAS H=0.0 AT 0 DEG-R* HSTE=ENTH OF TURBINE EXHAUSTIBTU/LB)* HJMDWO- INITIAL DW GAS HUMIDITY (PERCENT)* HUMSPO-INITIAL SP GAS HUMIDITY (PERCENT)* KMIX-EMPIRICLE CONSTANT FOR NAT. CIRC MIXING FLOW BET. POOL NODES* LBASE=»NCMINAL STARTING LEVEL OF POOLdN. FROM INST. 0.)* MHSPGO,MHOWGO=INITIAL MASSES OF H(LBS)* MMSPGO,MMOWG0=INITIAL MASSES OF M (LBS)* POCVP*PRESS. DIFF NECESSARY TO CLEAR THE VENT* PIPES FOR FLOW FRCM DW TO SP* PTDWGO- INITIAL TOTAL PRESS OF DW GAS (PSIA)* PTSPGO-INITIAL TOTAL PRESS OF SP GAS (PSIA)* QRVHLO=REACTCR(+PIPING) HEAT LOSSES (MW) FOR TEMP DIFF=OTRVHL(DEG-F)* TAUFDW=AVERAGE RESIDENCE TIME OF FOG IN DW(SEC)* TAUFSP=AVERAGE RES. TIME OF FOG IN SP* TAUVRV=TIME CONSTANT FOR SP VAC RELIEF VALVE TO LIFT(SEC)* TBASE-NCMINAL STARTING TEMP CF POOL(CEG-F)* TBII-PROVISION FOR BIASING STARTING TEMP CF POOL NODE I* TD=OISCH. TEMP(F) OF POOL COOLING FLOW* TDMETO-INITIAL DRYWELL METAL TEMP IF)* TGDWO*INITIAL DW GAS TEMPtDEG-F)* TGSPO*INITIAL SP GAS TEMP (DEG-F)* TPAIROMNITIAL SP CHAMBER ROOM TEMP
186
* TPMETO=INITIAL TEMP : SP METAL IN CONTACT WITH SP GAS* VGOW=TOTAL FREE VOLUME OF DW (FT**3)* VTSP=TOTAL FREE VOLUME OF SP (FT**3)* WOLEAK=LEAK RATE(LB/SEC) OF SAT. WATER FROM RV TO DRYWELL* WDSPW=DISCHARGE FLOW OF POOL COOLING(LB/SEC)* WSSPW=SUCTION FLOW OF POOL COOL ING!LB/SEC)*
*
* GEOMETRY AND PHYSICAL CHARACTERISTICSCONSTANT ACMET=1.65E04,APMET=1.7E04,ASSPW=10860.CONSTANT BDWSP0=2500.,BSPDWO=2000.CONSTANT CDMET=8.33E04,CPAIR=3200.,CPMET=5.82E04CONSTANT DPQR=2. ,DPQZ=2.
CONSTANT GCH=5.361,GCM=.2436,GCN=.3829,GCS=.5955CONSTANT HSTE=915.,HSRREF=-854.5,P0CVP=1.75CONSTANT TAUVRV=3.,TAUFDW=30.,TAUFSP=15.CONSTANT VGDW=159000.,VTSP=267600.
*
* POOL NQDALIZATICN
CONSTANT FD1=1.,FS1=1.,FWI =1.,FSV1 =1.,FTE1 =1.,FWE=>5000.,KMIX=950.*
* INITIALIZATION
CONSTANT DM=0. ,HUMDW0=18. 4,HUMSP0=79.4 ,LBASE=1C23CONSTANT MHSPGO=0.,MHDWGO=0.,MMSPGO=0.,MHDWG0=0.CONSTANT PTDWG0=23.15,PTSPG0=23.65,TBASE=169.1,TBI1=0.CONSTANT TGDV«0=268. ,TGSP0=161.,TDMET0=212.,TPMET0=145.1,TPAIRO=155.
»
* RUN CONTROL
CONSTANT FLDWG=0.,FLSPG=0.,TD=90.,GRVHL0=1.,DTRVHL=404.CONSTANT W0LEAK=.68,WSSPW=C.,W0SPW=0.HD=TC-32.
* INITIALIZATION CALC. FOR DW AND SP MASS AND ENERGY BALANCES* ASSUME ZERO INITIAL HYDROGEN AND MISC. GAS*
* FUNCTION TABLES
* FUNCTION SPLEV : SP LEVEL!INCHES FROM INSTRUMENT ZERO) AS A FUNCTION A* FUNCTION CF VOLUME(CU.FT)
FUNCTION SPLEV=0.,-182.,20222.,-134.,74646.,-62.,...117344.,-14.,130129.,0.,139300.,10182563.,58.,242176.,130.,267611.,190.
*
* FUNCTION STFOSV: SATURATION TEMP AS A FUNCTION OF SPECIFIC* VOLUME
FUNCTION STF0SV=2.83,363.5,4.65 ,324.1,7.65,288.2,12.2,257.6, ...20.1,228.,31.2,204.,44.7,185.6,76.4,160.5,158.9,129.6,. ..
333.6,101.7,641.5,79.6,1235.,59.3
* FUNCTION SPFOT : SATURATION PRESSURE AS A FUNCTION OF TEMPERATUREFUNCTICN SPF0T=59.3,.25,79.6,.5,1C1.7,1.0,132.9,2.4,...
160.5,4.8,18 5.6,8.5,203.9,12.5, ...228.,20.,250.34,30.,288.2,56.,324.1,95.,363.5,160.,...
381.8,200.,417.3,300.,444.6,400.,467.,500.,486.,600.,...503.,700.,525.,850.,544.6,1000.,556.,1100.,567.,1200.
* PRELIMINARY CALCULATIONS
SPDWO=NLFGEN(SPFOT,TGDWO)SPSPO=NLFGEN(SPFOT,TGSPO)PSDWGO=(HUMDWO/100.)*SPDWOPSSPG0=(HUMSP0/100.)*SPSPO
PNSPGO=PTSPGO-PSSPGOPNOWGO=PTDWGO-PSDWGO
MNSPG0=PNSPG0*!VTSP-VWSP0)/(GCN*(TGSP0+460.) )MNDWG0=PNOWG0*(VG0W)/(GCN*(TG0W0+460.))
187
*
* VWSPO=INITIAL TOTAL WATER VOL IN SP
MSSPG0=PSSPG0*(VTSP-VWSPO)/(GCS*(TGSPO+46O.))MSDWG0=PSDWG0*VGDW/(GCS*(TGDW0+460.))
*
UMSPG0=(TGSPO+460.)*(MNSPGO*(.2475-.1851*GCN)+...MSSPG0*(.45-4.89/(TGSP0+460.)-.1851*GCS) )
UMDWGO=(TGOWO*460.)*(MNDWGO*(.2475-.1851*GCN)+...MSDWGO*(.45-4.89/(TGDWO+460.J-.1851*GCS) )
*
* DEPTH LBASE SPECIFIED: INCHES FROM INST ZER0(4 IN BELOH CENTER)* TBASE IS NOMINAL STARTING TEfP CF PCOL* TBKI) PROVIDES FOR BIASING INITIAL TEMPS IN MULTI-NODE MODEL
VII=2.462*AIIAII=(LBASE-4.)*SQRT(34580.-LBASE*LBASE+8.*LBASE)*...
34596.*ARSIN( (LBASE-4. )/186. ) + 54343.MW1II=FW1*VII*(63.9-.019*(TBASE*TBU))MII=MW1IIDTBASE=(1190.-(TBASE-32.))*0M/(MII+0M)TIC1=TBASE+0TBASE+TBI1VWSP0=(MWlII+FWl*0M)/(63.9-.019*TICl)VGSPO=VTSP-VWSPO
DYNAMIC* DYNAMIC PORTION OF REACTCR VESSEL CALCULATION#
* RUN CONTROLS FOR INJECTION CONTROLLDCVZ=L0C*12.+216.
LDDEL=REALPL(500.,2.OO.LDCVZ)
RCICC=0.0
HPCID=0.0WINJ=RCICMX*(RCICD*8.33*HPCID)*STEP(30.)
* RUN CONTROLS FOR VESSEL PRESSURE CONTROL
P0PEN=1120.
PSHUT*1080.
* CALCULATION OF DYNAMIC SATURATION PROPERTIESRH0F=1./(AFGEN(VSATF.P))RhOG=l./(FUNGEN(VSATG,2,P)»HF=AFGEN(HSATF,P)
HG=AFGEN(HSATG,P)
TSAT=FUNGEN(TSATM,2,P)* INTERMEDIATE CALCULATIONS: AVERAGE HEAT FLUXES* 'DKFUN' RETURNS DECAY HEAT AS A FUNCTION OF TIME SINCE SCRAM* 'QREGAV RETURNS AVERAGE NORMALIZED POWER, GIVEN LOCATION OF TOP* AND BOTTCM OF REGION
T=TO+TIMETMIN=T/60.
PREL=(DKFUN(T)+.94*EXP(-T/TCFUEL) )
QT0T=PREL*CP0*947.8QFCCR=QTOT/(12.*PCOR1AVMSC=QREGAV(0.,LSCD)AVMB=QREGAV(LSCD,LSCD+LBD)
QFSC=AVMSC*QFCORQFB=AVMB*QFCORLSCD=REALPL!LSCO,TAULEN,LSC)L8D=REALPL(LB0,TAULEN,LB)
* DOWNCOMER ANNULUS CALCULATION* NEEDS: CENSITIES: RHCOCRHOLP* FLOWS INTO DCWNCOMER: WINJ, WRECIR, WINJAS* INJECTION ENTHALPY: HINJIN* FLOWS OUT OF DOWNCOMER: WCI, WFLDC, WFLLP* REACTCR VESSEL PRESSURE: P
188
* RETURNS: DOWNCOMER HEIGHT!ABOVE BOTTCM CF ACTIVE FUEL) : LDC* ENTHALPY INTO LOWER PLENUM: HDC
* INTERMEDIATE CALCULATIONS: MASS RATE OF ASPIRATION FROM STEAM SPACE* DUE TO INJECTION FLOW; FLASHING RATE FRCM DC MASS; RECIRC FLOW RATE
EQXX=EQXO*(LHEDER-LDC)/LHEDEREQX=LIMIT(0.,1.,ECXX)HINJFI*HINJIN+(HF-HINJIN)*EQXGP=7.23*WINJ
WINTCT=INTGRL(0.,WINJ)GT=(WINT0T*7.481)/62.11
WINJAS=(HF-HINJIN)*EQX*WINJ/(HG-HF)WFLDCX=MOC*FFLASH*(HDC-HF)WFLDC=LIMIT(0.,9999..WFLDCX)
*
* MASS AND ENERGY BALANCES(81 IS A LOGIC SIGNAL TO STOP THE* INTEGRATION WHEN LDC.LT.O(BELOW BCTTCM OF ACTIVE FUEL)
CMDC=INTGRL(0.,DMOC)MOC=MCCO*CMOC
DMDC*Bl*(WRECIR*WINJ+WINJAS-RHOCC*(WCI*WFLLP)/RHOLP-WFLDC)+WEXPCCMHDC=INTGRL(0.,DMHOC)MHDC=MHDCO*CMHDC
DMHOC*Bl*(HF *WRECIR*W INJ*HINJIN+HG*WINJAS-...HDC*(WCI+WFLLP)*RHODC/RHOLP-WFLCC*FG)
HDC=MHDC/MDCRHOOC=l./AFGEN(VSC,HDC)
* DC LEVEL IS CALCULATED BY FUNCTION SUBPROGRAM XLENOCIVOLUME)VOC-»MCC/RK)OC
LDC=XLENDC(VOC)
LDL=LIMIT!0.,LDCR,LDC)
* LOWER PLENUM CALCULATION
* NEEDS: SAME AS DCWNCCMER* RETURNS: ENTHALPY AT CORE INLET(HLP)* INTERMEDIATE CALCULATIONS: LOWER PLENUM FLASHING RATE, LP WATER VOL* (VOLPW—USED FOR LOGIC CONTROL ONLY SINCE THE LOWER PLENUM IS* MODELED AS A CONSTANT VOLUME UNLESS COWNCOMER IS EMPTY) :
HLP=MHLP/MLPVOLPW=MLP/RHOLP
RHOLP=l./AFGEN(VSC,HLP)XWFLLP*(HLP-HF)*MLP*FFLASHWFLLP=LIMIT(0.,9999.,XWFLLP)
* LOGIC CCNTROL*
X1B1=LDC-.001
X2Bl"«V0LPW-l.01*VCLPB1N=N0R(X1B1.X281)B1=N0T(B1N>
*
* MASS AND ENERGY BALANCES*
DMLP-*(WCI +WFLLP)*{RHOCC/RhOLP)*(Bl)-WCI-WFLLP*...B1N*(WINJ+WINJAS>
CMLP*INTGRL(0., OMLP)MLP=MLPO+CMLP
DMHLP=H0C*(WCI+WFLLP)*RH0DC*B1/RH0LP+...BlN*(WINJ*HINJIN*HG*WINJAS)-WCI*HLP-WFLLP*HG
CMHLP=INTGRL(0.,DMHLP)MHLP=MHLPO+CMHLP
* CORE INLET FLOW CALC.(FORCE BALANCE)* NEEDS: DENSITIES: RHODC,RHOLP,RSCAV,RHCB,ROP, RORT* REACTOR VESSEL PRESSURE : P* FLOW OUT OF BOILING REGION: W80* REGION LENGTHS: LDC,LSC,LB* RETURNS: FLOW INTO CCRE: WCI*
CMTOT«INTGRL(0. ,(WCI-WRECIR-WTOST+WFLLP))MTOT-MTOTO+CMTOT
WB0=(QFB*PCOR*LBD)/(HG-HF)
189
WRECIR=L IMIT(0.,10000.,(WCI-WTOST)*FRECIR)♦WCOLL
WCOLL-LIMIT(0.,10000.,(LINT-30.)*3000.)FRECIR=LIMIT(C.,10.,(LINT-LB0T)/(LDCR-LB0T))LINT=LSCD*LB+LOPTPH.RTTP
WFLOP=0.0
WFLBR=0.0
WFLRT=0.0MC0REM=12.*ACCR*RHOGMOPEM=LOP*ACOP*RHCGMRTEM = (LR T+4.75)*ART*RHOGWBRAV=WCIG
RSCAV»(RH0F+RH0LP)/2.WCIG*REALPL(WCI0,3.0,WCI)
* CALCULATE DENSITIESCAVIN*DFVOID(ACOR,WCIG,0.0 ,TSAT.RHOF,RHOG )CAVEX=DFVOID(ACOR,WCIG-W8G,HBO ,TSAT,RHOF,RHOG)CAV=(CAVIN+CAVEX)/2.RHOB=RHOTP(RHOF,RHOG,CAV)OPV=0FVOID(ACCP,WCIG-WBO,WB0 ,TSAT.RHOF,RHOG)RHOOP=RHOTP(RHOF,RHOG,0PV)RTV»CFVOID(ART,HClG- WBO.WBO .TSAT.RHOF,RHOG)RHORT=RHOTP(RHOF,RHOG.RTV)BUVIN=DFVCI0(ACOR,WCIG.WFLLP.TSAT.RHCF,RHOG)BUVEX-«0FVOID(AC0R,WCIG-WB0,WT0ST, TSAT.RHOF,RHOG)BUAVG-*(BUVIN+BUVEX)/2.ROBUB=RHOTP(RHOF,RHOG,BUAVG)
*
DELHl=LIMIT(l.,400.,HF-HLP)DLSCX=(-2.)*(LSCX*QFSC*PCCR-WCI*LIMIT(0.,400.,HF-HLP) )/...
(RH0LP*AC0R*DELH1)LSCX-»INTGRL(LSCO,CLSCX)LSC=LIMIT(0.,12.,LSCX)
MSC»RSCAV*LSCD*ACOR*
* FINO THE TWO PHASE LENGTHS AND REGION MASSES
LB*LINITIO.,12.-LSCD.LBU)LBU=(MT0T-MC0REM+MSCDUM-M0PEM-MRTEM-MSC)/(AC0R*(RH08-RH0G))LC0V=LIMIT(0.,12.,LBC0VU+LSCD)LBCOVU=LBU*(RHOB-RHOG)/(ROBUB-RHOG)MSCDUM=ACCR*RHOG*LSCDMB«LB*RHOB*ACORMC0R*MSC*MB*(12.-LB-LSCD)*RH0G*AC0R
LOPTP-LIMITIO..LOP.(MTOT-MCCR-MRTEM-MOPEM)/(ACOP*(RHOOP-RHOG)))M0P-(RHOOP*L0PTP*(L0P-LOPTP)*RHCG)*ACCPMRT=MTOT-MOP-MCORLRTTP=(MRT-MRTEM)/(ART*(RhORT-RHOG))MPOO=(LB*(1.-CAV)*ACOR+LOPTP*(1.-CPV)*ACOP+...
LRTTP*(l.-RTV)*ART)*RHOF*
* TABULATE PRESS DROPS AND CALCULATE WCI*
PC0B0T-«(MC0R/AC0R+M0P/AC0P-!-LRTLIM*RHCRT-i-(10.-LRTLIM)*RH0G)/144.LRTLIM=LIMIT(0.,10.,LINT-17.52)L0SSIN=tDL*RH0DC/144.+(L0CR-LDL)*RH0G/l44.-PCOBOTX0UM=LIMIT(1.E-9,5.,L0SSIN)WCI=SCRT(XDUM*RH08/(KL0SSU*RH0BR) )WCIO-REALPL!WCI0,10.,WCI)
*
* REACTOR VESSEL PRESSURE CALCULATION*
* STEAM MASS BALANCECMST*INTGRL(0.,OMST)MST=MSTO+CMSTOMST=WTOST*WEVAP«-WFLDC-WEXPC-WCCNO-WINJAS-HSTCHTOST*LIMIT(0.,10C0.,HBO*WFLLP-WP0O)WCONO»0.0
WEVAP*0.0WEXPC=LIMIT(0.,100.,1.67E-04*MST*(HGST-HST>)HGST*AFGEN(HSATG,PST)
190
WPOO=MPOO*DHF/(HG-HF)HFD=REALPL(HF0,2.0,HF)DHF=(hF-HFD)/2.
*
VST=VFREE-V0CDVST=CERIV(O..VST)VSVST=VST/MST
* VST=T0TAL VAPOUR SPACE VOLUME* VSVST=SPECIFIC VOLUME OF STEAM*
*
* TOTAL ENERGY BALANCE
CUMT 0= IN T GRL 10 .. DUMTO )UMT0=UMST0+CUMTO
DUMTO=WTOST*HG+(WFLDC+WEVAPJ*HG-...WEXPC*HF-WC0N0*HG-WINJAS*HST-WSTC*HST-0VST*P*144./778.HST=UMT0/MST+PSTG*VSVST*144./778.
* STEAM PRESSURE CALCULATED BY FUNCTICN SUBPROGRAM PFVH--STEAM PRESSURE* AS A FUNCTION OF SPECIFIC VOLUME AND ENTHALPY* PFVH REQUIRES AN INITIAL GUESS OF PRESSURE, PSTG
PST=PFVH(HST,VSVST,PSTG)PSTG=REALPL(P0,2.0,PST)P=PST
*
* SAFETY RELIEF VALVE MODEL—TWO VALVES MCDELED~WSSV=TOTAL SRV FLOWL0GREL=L0GV1+L0GV2LOGVl=REALPLI0.,.62,L0GPVl)
L0GV2=REALPL(0.,.62,L0GPV2)LCGPV1=RST(X1V1,X2V1,0.)LCGPV2=RST(X1V2,X2V2,0.)XIV1=C0MPAR(PSHUT, P)X1V2=C0MPAR(P0PEN-25.,P)X2V1-CCMPAR1P.P0PEN)
X2V2=C0MPAR(P,P0PEN*25.)
WSSV=UCSRV*SQRT(PSTG/VSVST)*LOGREL
* WSTE=STEAM TURB EXHAUST FLOW(RCIC)* ENTHALPY IS ASSUMED=915. BUU/LB
WSTE=(RCICD*(.00613*PST+1.1)+HPCI0*(.0387*PST*7.81))*STEP130.)* WSTC IS TOTAL STEAM FLOW RATE FROM REACTOR VESSEL TO CONTAINMENT
WSTC=WSTE*WSSV
* CYNAMIC PORTION OF CONTAINMENT CALCULATION*
* CALLS TO SP WATER MACRO—ONLY ONE NECESSARY FUR THE SINGLE NOOE PCOL MODELXNQ1 ,TP1 ,WE1 ,MWl ,LW1 =W(TP1,...
TP1 ,LW1,LW1 ,TIC1 ,FW1 ,FSV1 ,FTE1 ,FD1 ,FS1 IVWSP=MWl/(63.9-.019*TPl)MWSP=MW1
GWSP=7.481*VWSP
TWSPAV=(63.9-MWSP/VWSP)/.019LWSPAV=AFGEMSPLEV,VWSP)
* INTERFACE THE SP WATER TO THE SP GAS
XNQDW=XNQ1WSSVNQ=WSSV*FSV1*XNQ1WSTENC=WSTE*FTE1*XNQ1WTESPW=WE1
*
*
* INTERFACE VARIABLES : RV TO SUPPRESSION POOL* WSSV=STM. FLOW FROM RV THRU RELIEF VALVES(LB/SEC)* WSTE=STM FLOW FROM RV THRU TURBINES!RCIC+HPCI)* WHSV=HYDROGEN FLOW FROM RV ThRU RELIEF VALVES!LB/SEC)* HHM=HYDRCGEN ENTHALPY AT MIXED RV TEMP(BTU/LB) (REF TO 0. DEG R)* HSTE=ENTHALPY OF TURBINE EXHAUSE(ASME)* TM=MIXTURE TEMP OF RV STEAM SPACE
*
*
*
191
INTERFACE VARIABLES : RV TO DRYWELLWSDWR=STEAM FLOW, RELEASE TO DRYWELL(LB/SEC)HSDWR=ENTHALPY OF WSDWR(ASME)
WHDWR-HYDROGEN FLOW, RELEASE TO DRYWELL!LB/SEC)HHDWR-ENTHALPY OF WHOWRIREF TO 0.0 DEG R)WMDWR=MISC. FLOW RELEASEO TO DRYWELLHMDHR-*ENTHALPY OF WMDWRIO.O DEG R)
INTERFACE VARIABLES : SP WATER TO SP GASWSSVNQ-TCTAL NCN-QUENCHED RELIEF VALVE FLOW FROM RV TO SP VIA RV'SWSTENQ-TOTAL NON-QUENCHED TURBINE EXHAUST FLOWWCSPG=TCTAL CONSECSATE FLOW, SPG TO SPWWTESPW-TOTAL EVAPORATION RATE, SPW TC SPGHSTSPW-ENTHALPY OF STEAM(REF 0. AT 0. OEG) AT MASS-WEIGHTED SP WATER
TEMP, TWSPHCSPG*ENTHALPY OF CONDENSED SP STEAM(ASME)
MASS BALANCES FOR SP GAS AND DW GAS
INPUT FROM OTHER CALCULATIONS:VWSP*SP TITAL WATER VOL (FT**3> (VARIABLE)WSSNQ* TCTAL NON-QUENCHED STEAM FLOW (LE/SEC)WHSV- H FLOW FROM THE RV (THROUGH RELIEFS)
WSOWR* FLOW RATE OF STEAM RELEASED DIRECT TO DRYWELL
WHDWR- FLOW RATE CF HYDROGEN RELEASED TO DRYWELL
WMDWR* FLOW RATE UF MISC. RELEASED OIRECT TO DRYWELL
DEFINITIONS
MS SPG
MHSPG
MNSPG= MASS N IN SPG
MMSPG'MASS M IN SPG
MASS STM IN SPGMASS H IN SPG
FLOWS » LB/SEC UNLESS OTHERWISE SPECIFIEDWSSVNG* TOTAL NON-QUENCHED STM FROM SV'S ANT TURB, EXHAUST
WHSV- TCTAL HYDRCGEN FLOW FRCM RELIEF VALVESDWSP* DESIGNATES FLOW FROM DW TO SPSPOW- DESIGNATES FLOW FROM SP TO DW
SPL» SP GAS LEAKAGEB* BULK FLOW RATE (FT**3/SEC)
WCSPG- TOTAL CONDENSATE FLOW FROM SP GASWCOWG* TOTAL CONDENSATE FLOW FROM DW GAS
SP GAS MASS BALANCE:MSSPG«INTGRL(MSSPGO,WSSVNQ+WSDWSP*WSTENQ+WTESPW-WSSPDW-WSSPL-WCSPG)
MHSPG=INTGRL (MHSPGO.WHSV+WHOWSP-WHSPDW-WHSPL)MNSPG- INTGRL(MNSPGO.WNDWSP-WNSPL-WNSPDW)MMSPG=INTGRL (MMSPGO.WMDWSP-WMSPL-WMSPDW)
BSPL=VGSP*FLSPG
WSSPL *BSP L*(MSS PG/VGSP)
WHSPL'BSPL*(MHSPG/VGSP)
WNSPL=BSPL*(MNSPG/VGSP)
WMSPL=BSPL*(MMSPG/VGSP)
BSPL*BULK LEAKAGE FRCM SP GASFLSPG-FRACTION OF TOTAL SP GAS VOLUMES LEAKED PER SECONOVGSP»VOLUME OF SP GAS, VWSP-SP WATER VOLUME
VGSP-VTSP-VHSP
NOTE: 0.5 PSI IS NECESSARY TO OPEN VAC RELIEF VALVEBSPCWX=LIMIT(0.,1.E06,BSPDW0*(PTSPG-PTDWG-.5)I8SPDW-<REALPL(0.,TAUVRV,BSPDWX)WSSPDW=BSPDW*(MSSPG/VGSP)WHSPDW-BSPDW*(MHSPG/VGSP)
WNSPCW=BSPDW*(MNSPG/VGSP)WMSPDW=BSPDW*(MMSPG/VGSP»
BSPOWO* (FT**3/SEC)/PSI OF PRESS DIFFERENCE WHEN
192
* VAC RELIEF VALVE IS OPEN
BOWSPX=LIMIT (0.,l.E06,BDWSPO*(PTOWG-PTSPG-PDCVP))BOWSP-«REALPL(0.,TAUVRV,BOwSPX)WSDHSP»BDWSP*(MSDWG/VGDH)*XNQDWWHDWSP=BDWSP*(MHOWG/VGDH)WNDWSP-BOWSP*(MNDWG/VGDH)
WMDWSP=BDWSP*(MMDWG/VGDW)
* 80WSP0»(FT**3/SEC)/PSI WHEN DW PRESS IS GREAT ENOUGH* TO CLEAR THE VENT PIPE DOWNCOMERS* POCVP«PRESS DIFF. NECESSARY TO CLEAR THE VENT PIPES*
WCSPG«WWCSPG+WVCSPG
WVCSPG»(MSSPG/TAUFSP)*(1.-100./HUMSP)*C0MPAR(HUMSP,100.)* WWCSPG'RATE OF CQND. ON SPG WALL* WVCSPG-RATE OF COND. FROM SPG VOLUME*
* DRYWELL MASS BALANCE:
MSOWGMNTGRL(MSDWGO,WSDWR+WSSPDW-WSDWSP-WSDWL-WCDWG)MHOWG-INTGRL(MHOWGO.WHOWR-tWHSPCW-WHDWL-WHDWSP)MNDWG-*INTGRL(MNOWGO.WNSPOW-WNOWSP-WNDWL>MMDWG*INTGRL(MMDWGO,WMDWR*WMSPOW-WMDWSP-WMDWL)
*
WCDWG*WWCDWG*HVCDWG
WVCDWG»(MSDWG/TAUFDW)*(1.-100./HUMOW)*COMPAR(HUMOW,100.)*
* HYDROGEN AND MISC RELEASES SET TO ZERO
* FFRACT=FRACTION OF HOT LIQUID LEAK THAT FLASHES TO STEAMFFRACT-{AFGEN(HSATF.PST)-AFGEN(HSATF,PTOWG))/...
(AFGEN(HSATG,PTDWGI-AFGEN(HSATF.PTDWG))WSDWR»WDLEAK*FFRACT
HSDWR*AFGEN(HSATG,PTDWG)WHDWR=0.
WMDWR«0.
HHOWR-0.
HMDWR-O.
WWCDW&*RATE OF COND. ON DWG WALLWVCOWG-fRATE OF VOLUME COND. IN DRYWELL
ENERGY BALANCES FOR CW AND SP GAS
INPUT FRCM OTHER CALCULATIONS:HMDWR=ENTHALPY OF MISC. RELEASED DIRECT TO DRYWELLHHOWR'ENTHALPY OF HYOROGEN RELEASED OIRECT TO DRYWELLHHRV=ENTHALPY OF HYDROGEN IN REACTOR VESSELHSRV*ENTHALPY OF STEAM IN REACTOR VESSEL
HSDWR-ENTHALPY OF STEAM RELEASED DIRECT TO DRYWELLHSTE=ENTHALPY OF STEAM FROM TURBINE EXHAUST
NOTE THAT THE ABOVE 3 ARE REF. TO ZERO AT 32.F WATER(ASME STM TABLES), BUT ARE RE-REF TO ZEROAT ZERO DEG-R FOR CONTAINMENT CALC.QVSPG,QVCWG*VOLUME HEAT SOURCE (FROM F.P.'S)QRVHL-HEAT LOSS (THROUGH INSULATION) FRCM RV TO SP GASQLSPG=HEAT LOSS FROM SP GAS TO SP LINER
QLDWG-HEAT LOSS FROM OW GAS TO DW LINER
DEFINITIONS:
TGSP-TEMF OF HOMOGENIZED SP GAS (DEF-F)
TGDW«TEMP OF HOMOGENIZED DW GAS (DEG-F)
UMSPG"TOTAL INTERNAL ENERGY OF SP GAS (BTU)
UMDWG=TCTAL INTERNAL ENERGY OF DH GAS (BTU)
MASS FLOWS DEFINED IN MASS BALANCE SECTION
SP GAS ENERGY BALANCE:
CUMSPG=INTGRL(0.0,DUMSPG)UMSPG»UMSPGO->CUMSPG
193
DUMSPG-EPSSP-t-EPHSP+EPNSP-e-EPMSP+QVSPG-QLSPG-MWORK-t-QSSPGEPSSP*WSSVNQ*(HSSVNQ)*WSTENQ*(HSTE+HSRREF)+...
WSOWSP*! HSTGDW )-( WSSPDW-t-WSSPL )*HSTGSP-. . .WCSPG*(HFGSP*HSRREF)+WTESPW*HSTSPWMWORK«DVOTPG*PTSPG*.1851DVDTPG-»(VGSP-VGSP0)/5.VGSPD«REALPL(VGSP0,5.,VGSP)
* WSSVNQ-TQTAL SRV FLOH THAT EXITS SURFACE OF POOL* WSTENQ-TOTAL NON-QUENCHED STM FROM TURBINE EXHAUST* HSRREF»OELTA-H TO REREF. STEAM ENTHALPY
* HFGOW»SAT FLUID ENTH AT DW TEMP(ASME)* HFGSP»SAT FLUID ENTH AT SP TEMPIASME)* HSSVNQ*ENTHALPY OF NON-QUENCHED STEAM ENTERING SP GASJREF TO 0.* AT 0. CEG-R)
HSSVNC-HST+HSRREFHSTGSP».45*TGSPR-4.89
HSTSPW*.45*(TWSPAV->460.)-4.89
HSTGDW-.45* TGDWR-4.89
*
HFGSP-TGSP-32.
HFGDW-TGDW-32.*
EPHSP«WHDWSP*HHTGCW+WHSV*HHRV-(HHSPOW->WHSPL)*HHTGSPHHTGSP-«3.466*TGSPR-40.
HHTGDH«3.466*TGDWR-40.*
EPNSP«HNTGDW*WNDWSP-(WNSPCW+WNSPL)*HNTGSP
HNTGSP-.2475*TGSPR
HNTG0W«.2475*TGDWR*
EPMSP«HMTGOW*WMDWSP-(WMSPOW-i-WMSPL)*HMTGSPHMTGSP«.21*TGSPR-20.8HMTGDW».21*TGCWR-20.8
*
* DRYWELL ENERGY BALANCECUMDWG»INTGRL(0.0,OUMDWG)UMDWG«CUMDWG*UMOWGODUMDWG-EPSDW+EPHDW+EPNDW+EPMOW+QVOWG*...QRVHL-QLOWG
*
QSSPG»(TWSPAV-TGSP)*ASSPW*5.3E-05*((ABS(TWSPAV-TGSP))**.33>QLSPG-QPMWET+QPHORY
QVDWG=0.
QVSPG»0.QRVHL«QRVHL0*948.*(TSAT-TGDW)/DTRVHLQLDWG»CDM
EPSDW»(HSOWR*HSRREF)*WSDWR+HSTGSP*WSSPOW-HSTGOW*(WSDWSP-fWSOWL>-HFGOW*WCOWG
*
EPHDW=HHDWR*WHDWR+HHTGSP*WHSPDW-HHTGDW*!WHDWSP*WHOWL)
EPNDW«HNTGSP*WNSP0W-HNTGDW*(WNCWSP->WNOWL)
EPMDW*HMDWR*W.-'DWR-t-HMTGSP*WMSPOW-HMTGCW*(WMDWSP+WMDWL)
*
* SOLUTION FOR DW AND SP GAS TEMPERATURES* ITERATION IS NOT NECESSARY SINCE THE S,N,H,ANO M* ENTHALPIES ARE ASSUMED LINEAR WITH TEMP:* H(N2)*.2475T, H(H20)*.45T-4.89, H(M)».21T-20.8, H(H)*3.466T-40.*
TGSPR*(UMSPG+4.89*MSSPG+20.8*MMSPG+40.0*MHSPG)/...(MNSPG*(.2475-.1851*GCN)*MSSPG*(.45-.1851*GCS)<-...MMOWG*(.21-.1851*GCM)+MHSPG*( 3.466-.1851*GCH) )
TGDWR»(UMDWG+4.89*MSDWG*20.8*MMOWG*40.*MHOWG)/...(MNDWG*(.2475-.185l*GCN)*MS0HG*(.45-.18 51*GCS)*...MMDWG*(.21-.1851*GCM)+MHDHG*(3.466-.1851*GCH)>
194
TGSP=TGSPR-460.
TGDW=TG0WR-460.
* TOTAL AND PARTIAL PRESSURES(PS I A) CALC. FROM TEMP(F)PNSPG=MNSPG*GCN*TGSPR/VGSP
PHSPG=MHSPG*GCH*TGSPR/VGSP
PSSPG=MSSPG*GCS*TGSPR/VGSP
PMSPG=MMSPG*GCM*TGSPR/VGSPPTSPG=PNSPG+PHSPG+PSSPG+PMSPG
*
PNDWG=MNDWG*GCN*TGDWR/VGDW
PHDWG=MHDWG*GCH*TGDWR/VGDW
PSDWG=MSDWG*GCS*TGDWR/VGOW
PMDWG=MMDWG*GCM*TGOWR/VGDW
PTDWG=PNDWG + PHOWG*PSDWG+Pf<DWG*
* HUMIDITIES AND DEHPOINTS
HUMSP=100.*PSSPG/NLFGEN(SPFOT,TGSP)HUMDW=100*PSOWG/NLFGEN(SPFOT,TGDW)
TDEWSP=NLFGEN(STFCSV,VGSP/MSSPG)TDEWOW=NLFGEN(STFCSV,VGDW/MSOWG)
♦
8DWL=VGDW*FLCWG
WSDWL=BOWL*(MSDWG/VGDW)
WHDWL=BDWL*(MHDWG/VGDW)WNOWL=BOWL*(MNOWG/VGDW)
WMOWL=BDWL*tPfOWG/VGOW)
* CALCULATION CF SP ANC DW WALL METAL TEMP ASSUMPTIONS:* 1.CONSIDER ONLY METAL SURFACE IN CONTACT WITH GAS* 2.N0 CONDENSING UNLESS METAL TEMP BELOW DEwPOINT* STM. CONDENSATION ON WALLS IS AIR-LIMITED AND IS CALCULATED* FROM ECUN. III.8.26 CF MARCH MANUALtNUREG/CR-17UJ
QPM=QPMDRY+QPMWET-QPGAIRQPMDRY=(TGSP-TPMET)*5.3E-05.*( I ABS( TGSP-TPMET) )**. 3333)* APMET
QPMWET=CCMPARITCEWSP,TPMET)*(TDEWSP-TPMET)*APMET*...
•0185*(1MSSPG/MNSPG)**.707)TPMET=INTGRL< TPMETO.QPM/CPMET)QDM=CDMORY+QCMWETQDMDRY=!TGDW-TDMET)*5.3E-5*((ABS(TGDW-TDMET))**.3333)*ADMETQDMWET=COMPAR!TDEWDW,TDMET)*(TOEWOW-TDMET)*AOMET*...
•0185*((MS0WG/MNDWG)**.707)
TCMET=INTGRL(TDMETO,QDM/CCMET)*
* CALC OF CONDENSATION RATE ON DW AND SP WALLS. APPROXIMATE VALUE* OF 9C0 BTU/LB IS USED FOR (HG-HF)
WWCSPG=«PMWET/900.
WWCDWG=QDMWET/900.*
* CALCULATION OF POOL ROOM AIR TEMPQPAIR=CPGAIR+GPWAIR
QPWAIR=5.3E-05*((ABS(TWSPAV-TPAIR))**.3333)*APMET*(TWSPAV-TPAIRJ
QPGAIR=5.3E-5*((ABS!TPMET-TPAIR))**.333)*APMET*!TPMET-TPAIR)
TPAIR=INTGRL(TPAIRO,QPAIR/CPAIR)*
* NOTE THAT THE TERM APMET IS THE SAME IN BOTH EQUATIONS* BECAUSE IT IS FOR ONE HALF OF TOTAL METAL SURFACE AREA*
* THESE CALCULATIONS ARE TC BE USED FOR AVERALL THERMO
* CONSERVATION CHECK
OMSP=INTGRL(0.,WSTC)
HEATIN=INTGRL!0.,GTOT)OMHRV»lNTGRL!0.,WINJ*HINJIN-WSTC*HST)
*
LINTVZ=LINT*12.+216.
PRINT L0CVZ,LSC,LB,P,WTOST,WSTC,TWSPAV,LWSPAV,GP,TGSP,TGDW,WCIDTIMER DELT=.25, FINTIM=2C.0C0, PRDEL=10.METHCC RECT
NOSCRT
END
STOP
CALL DEBUG!1,0.0)
CALL CEBUGd,3600.)CALL DEBUG(1,7600.)CALL DEBUG!1.1C80C.)CALL DEBUG!1,14400.)CALL CEBUGd,17999.)
195
197
APPENDIX B
MODIFICATION TO MARCH SUBROUTINE ANSQ
In the MARCH code, the calculation of decay heating is carriedout in subroutine ANSQ. For this study, the calculation was modifiedto include the actinide decay heat source in a BWR following a burnupof 34,000 MWd/x. A listing of the revised subroutine follows:
Subroutine ANSQ (ANS-TIME-TAP)C
C ANSQ calculates the decay heat using ANS correla-C tions.
ANS1=ANS2=0.0
TVAR=TIME*60.0
IF (TVAR-GT-10.0) GO TO 10ANS1=0.06950-0.001592*(ALOG(TVAR))IF (TVAR-LT-3.) ANS1=1.GO TO 40
10 IF (TVnAJi.GT. 150.0) GO TO 20ANS1=0.069241-0.0069355*(ALOG(TVAR))GO TO 40
20 IF (TVAR-GT.4.E6) GO TO 30ANS1=4.0954E-2-2.8324E-3*(ALOG(TVAR))GO TO 40
30 ANS1=4.6893E-3-2.3800E-4*(ALOG(TVAR))40 TVAR=(TIME+TAP)*60.0
IF (TV.AJi-GT-10.0) GO TO 50ANS2=6.950E-2-1.592E-3*(ALOG(TVAR))GO TO 80
50 IF (TVhAhR-GT-150.0) GO TO 60ANS2=6.9241E-2-6.9355E-3*(ALOG(TVAR))GO TO 80
60 IF (TVAR.GT-4.E6) GO TO 70ANS2=4.0954E-2-2.8324E-3*(ALOG(TVAR))GO TO 80
70 ANS2=4.6893E-3-2.3800E-4*(ALOG(TVAR))80 ANS-ANS1-ANS2
RETURN
END
199
APPENDIX C
MARCH INPUT FOR ACCIDENT SEQUENCE TB'
201
BROWHS FERRY CSB ♦ HPCI/RCICSNLMAR
ITRAN-1,IBRK=0,ISPRA=1,IECC=2,IBURN=0,NINTER*100,IPDTL=7,IPLOT=3,I0«3,VOLC-=416700.0,DTIHIT-0.01,TAP*2.62E06,
SEND
SNLIBTL
SEND
STEEL COICBETEDBTHELL1 OBTHELL2 CONC SHELLHISC STBELHISC CONC.
C
SNLSLABHNAT*2,NSLAB=3,NOD*1,4,13,DEN(1)=486.92*.157.481,HC<1)=. 1137,.23817,TC{1)=25.001,.80024,ITL*1,1,2,IYR*1,1,2,NH01*3,9,4,HATI-1,2,1,MAT2*1 2 1SAREA=18684.,5358., 15982.,X(1)*0...01,.02083, U«)«0.,.01,.03,.07,.15,.31,.63,1.27,2.5,X(13)=0. ,.01,.03,.0625,TEHP»12*150.,»*95.,
SEND
SNLECCPDHIO*0.001,0HIO*3.11E04,PACHO*0.001,ACHO-3.11E04.PHH~1120.,WHH1 —5000.0,PSIS~1120.,WSIS1*0.0,PLH*1120.,WLH1*0.0,STPHH*240..RtJSTH*3.11E06,BCCRC*0.64,CSPRC*1.0,DTSOB«-100.,HTCAT*100.,TRWST«95.0,
SEND
SNLECX
SEND
SNLCSXSEND
202
SNLCOOLSEND
INDUCE
NC0B*2,NHPV1=2,NRPV2*1,ICEC0B*-1,DTPNT=20.0,IDRY=-1,I«ET*2,RPOOL*7.801E06,TPOOL*95.0,VDRX=3.839E03,VTOROS=257700.0,BVNAX=5. 146E05,HSHP=-2,NSHP2*2,NCAV=-1,VCAV*4789.1,VFLR=15.0,AVBRK*292.0,CVBRK*4.04,VC(1)*159000.0,257700.0,AREA{1)=1.6399E03,1.098E04,HOH(1)*0.2.1.0,TEHPO(1)*150.,95.,N=10,NS<1)=1,1,1,3,3,2.2.2,2,2.NC(1)*1,1,1,1,1,1,1,2,2.2,NT<1)=1.2,3,-7,-7,-7,-7,-7.-7,-7,C1(1)*1.E6,1.E6,0. ,400.,500., 139.7,189.7, 139.7,189.7, 159.7,C2(1)=0.,1.333E5,7.59106,5.9297,.5 83.5.9297,.583.5.92 97,.583,.583,C3{1)=95.,0..1192.5..00694,20.97,.00694,20.97.6.94E-»,.0833,6.94E-3,C4(1)-0.0,0.0,0.0.14.7,0.0,14.7,0.0,14.7,14.7.14.7,KT(1,2)=1,KT(2,1)=1,STPECC*240..
SEND
SNLBOILNNT*37436,MF=35908,NDZ=50,ISTF=3,ISG=0,iim-o,IHR*1,WDED*3.50E04,Q2ER0=1.1242E10,H*12,HO*28.,DC*15.59,ACOR*104.833,ATOT=287.898.HATBH*97000.,D=.04692.DF*.04058,DH=0.056,CLAD*.005594,XOO=8.33E-06,RHOCD=68.783,TGOO*546.,PSET*1050.0,CSRV=3380.»FDCR=-.5.
203
DP ART=0.0208333,FZHCR*0.05,FZOCR*0.08.FZOS1=0. 1,WFE2*7992.,TFEO0=546.,FOLSG*0.0,PVSL=1000.,TCAV=1210.,ABRK*0.0,TBRK=45.0,DTPNTB*5.0,DTPN=-5.,VOLP*2.459E04,VOLS=9.638E03,WCST*3.11E06,F<1|»0.1,0.25,0.47,0.65,0.84,0.96,1.13,1.27,1.295,1.27.1.24,1.21,1.195,F(14)=1.15,1.11,1.08,1.05,1.03,1.02.1.016,1.017,1.05,1.06,1.061,1.062,F(26) =1.061,1.06,1.059,1.059,1.06,1.07,1.075,1.095,1.11,1.12,1.185,1.215,F(38)*1.25,1.26,1.24,1.21,1.15.1.09,1.0,0.87,0.76,0.6,0.41,0.21,0.1,PF(1)=1.017,1.087,1.093,1.095,1.096,1.094,1.0875,1.128,0.9665,0.408,TF(1)=10*0.1,TT=6*546.,CM=1824.,7992.,8760.,2460.,5550.,24900.,AH=740..263.,9225.,400-,7000.,700.,DD=1.,1.,1.,.17,.02,.546,AB*150.,263.,165.,0.,-10.,-20.,
tEND
SNLHEADWZRC=140397.0,NFBC*30447.73,B002*361837.0,HGBID=66750.,MHEAD=175927.08,DBH=20.915,THICK-0.52198,COND=8.0005,E1*.8,E2*.5,
SEND
SNLHOTIHOT=100,DP=0.25,FLBNC=3360..
SEND
SNLINTBCAYC=0.01524,CPC=1.30,DENSC=2.375,TIC=308.16,FC1=0.441,FC2=0.108,FC3=0.357,FC4=0.027,BBR=0.135,R0=322.6,R=6000.0,
HIN=0.2,HIO=0.09,«ALL=1000.,
SEND
205
Appendix D
PRESSURE SUPPRESSION POOL MODEL
D.l Introduction
The primary containment of each of the Browns Ferry Nuclear Plantunits is a Mark I pressure suppression pool system. The safety objectiveof the Mark I containment is to provide the capability, in the event of anaccident, to limit the release of fission products to the environment.29The key to the safety objective of this system lies in the performanceof the pressure suppression pool (PSP). The PSP is designed to rapidlyand completely condense steam released from the reactor pressure vessel,to contain fission products released from the vessel, and to serve as asource of water for the emergency core cooling systems. If the PSP shouldfail to properly perform during an accident sequence, then there is a highprobability that fission products will be released to the environment.
In the past few years, a major concern of the NRC, the BWR vendor,and the utilities who operate the early generation BWRs has been scenarios30 where it has not been conclusively shown that the PSP will meetthe performance requirements stated above. Specifically, there is a lackof information on condensation oscillations and the resulting loads theyplace on the PSP. In addition, there is the question of pressure suppression pool thermal stratification and the resulting impact of this phenomenon on the condensing ability of the PSP.
The problem of condensation oscillations* includes the analysis ofboth loads which originate at the downcomer exit during a large break lossof coolant accident (LOCA) and loads derived from the safety relief valve(SRV) discharge during normal blowdown to the PSP. The condensation oscillation problem involves coupling an analysis of the momentum transport within the PSP to a stress analysis of the torus walls.
The problem of pool thermal stratification is of interest becausethermal layering of the water near a SRV discharge point can limit theability of the PSP to condense the steam. Subsequently, this can lead toincreased intensity of the condensation oscillation loads and possibly tooverpressurization of the torus. Furthermore, the temperature distribution in the torus determines (to some extent) the distribution of fissionproducts in the PSP during an accident. Thus, an analysis of fission product transport to the environment will depend on an analysis of the temperature distribution in the PSP.
D.2 Purpose and Scope
The purpose of this work is to study the dynamics of the Mark I pressure suppression pool (PSP). Knowledge of the response of the PSP is
*In the literature, when the condensation instability occurs outsidethe relief valve tailpipe it is termed "condensation oscillation." If itoccurs inside the pipe, the phenomenon is called "chugging." Here, "condensation oscillations" includes both of these effects.
206
desired for transients ranging from a normal SRV discharge to a fullblown, large break LOCA. The primary objective of this study is to improve the available BWR pressure suppression pool analysis techniques. Asecondary objective is to learn as much as possible about the complexphenomena involved in each of the transients (phenomena such as poolswell, condensation oscillation, chugging, and thermal stratification).
An analysis of the PSP dynamics will necessarily involve accountingfor the two phase flow of steam jets into subcooled water and the ensuingtransport of mass, momentum, and energy by the mechanism of condensation.Because of the complex geometry of the PSP, (toroidal geometry plus large,submerged, complicated flow obstructions plus an air-water interface) theanalysis tool used in this study will be a state-of-the-art, 3-D thermalhydraulics code.
Pending completion of this work, it will be necessary to assumepool-averaged conditions for accident analyses.
D.3 Description of the System
The Mark I containment system consists of the drywell, the pressuresuppression pool, the vent system connecting the drywell and PSP, a containment cooling system, isolation valves, and various service equipment.Figure D-l shows the arrangement of the drywell, PSP, and vent systemwithin the Reactor Building.
The drywell is a steel pressure vessel with a spherical lower portionand a cylindrical upper portion. It is designed for an internal pressureof 0.531 MPa (62 psig) at a temperature of 138°C (281°F). Normal environment in the drywell during plant operation is an inert atmosphere of nitrogen at atmospheric pressure and a temperature of about 57°C (135°F).
The vent system consists of 8 circular vent pipes which connect thedrywell to the PSP. The vent pipes are designed to conduct flow from thedrywell to the PSP (in the event of a LOCA) with minimum resistance, andto distribute this flow uniformly in the pool. The vent pipes are designed for an internal pressure of 0.531 MPa (62 psig) with a temperatureof 138°C (281°F); they are also designed to withstand an external pressureof 0.014 MPa (2 psi) above internal pressure.
The pressure suppression pool is a toroidal shaped steel pressurevessel located below the drywell. The PSP contains about 3823 m3(135,000 ft3) of water and has an air space above the water pool of 3370m3 (119,000 ft3). Inside the PSP, extending around the circumferenceof the torus, is a 1.45 m (4.75 ft) diameter vent header. The 8 drywellvents connect to this vent header. Projecting down from the vent headerare 96 downcomer pipes which terminate 1.22 m (4 ft) below the surface ofthe water. At 13 unevenly distributed positions around the PSP, dischargelines from the safety relief valves extend through the vent pipes and terminate in a T-quencher device located near the bottom of the pool. FigureD-2 shows a cross section of the PSP and the relative locations of the
vent pipe, vent header, downcomer, SRV discharge line, and the T-quencher,which has been rotated 90° for the purpose of illustration. Near thebottom of the PSP, a 0.762 m (30-in.) suction header (ring header) circumscribes the torus and connects to the pool at four locations. AtBrowns Ferry, the RHR, HPCI, core spray, and RCIC systems are suppliedfrom this header.
SUPPRESSION
CHAMBER
TOROIDAL
HEADER
207
ORNL-DWG 79-7611A ETD
DRYWELL
REACTOR VESSEL
REACTOR BUILDING
Fig. D.l BWR reactor building showing primary containment systemenclosed.
The torus which contains the pressure suppression pool is designed toessentially the same requirements as the drywell liner, i.e., a maximuminternal pressure of 0.531 MPa (62 psig) at 138°C (281°F), but neither thedrywell nor the torus is designed to withstand the stresses which would becreated by a significant internal vacuum. To ensure that a significantvacuum can not occur in the drywell, vacuum breaker valves are installed,which will open to permit flow from the PSP airspace into the drywellwhenever the suppression pool pressure exceeds the drywell pressure bymore than 3447 Pa (0.5 psi). Additional vacuum breaker valves with thesame setpoints are installed to permit flow from the Reactor Building intothe PSP airspace, to prevent a significant vacuum there.
D.4 Identification of the Phenomena
The thermal hydraulic phenomena associated with a BWR pressure suppression pool are mainly those dealing with the PSP response during two
RING GIRDER
ECCS
SUCTION
LINE
pool.
t = 0.756 in.
Fig. D.2 Browns Ferry Mark I containment pressure suppression
MAIN VENT-
BELLOWS
ORNL-DWG 81-8567 ETD
SAFETY RELIEF VALVE
DISCHARGE LINE
roooo
209
types of transients: (1) LOCA-related phenomena and (2) SRV-dischargephenomena.
A. LOCA — Related Phenomena
Immediately following the pipe break in a LOCA, the drywell pressureand temperature increase very quickly. The pressure increase forces waterstanding in the downcomer to accelerate rapidly into the PSP and impingeon the torus wall. Following the slug of water, air that was in the ventpipes and drywell is forced into the PSP. This forms a bubble of air atthe downcomer exit which expands into the suppression chamber and causesthe pool to swell. As the air bubble rises into the torus airspace, thewater will experience a gravity-induced fallback and phase separation willagain occur.
The pool swell transient described above lasts on the order of 3 to 5s.30 It has been studied by several authors, both experimentally31-33 andnumerically.3•+,35 The consensus is that pool swell impingement and dragloads induced during a LOCA are conservatively estimated and acceptable.
Immediately following the pool swell transient, an air/steam mixturewill flow into the PSP. Early in this process, when the mass flow rate ishigh, the injected steam condenses at an unsteady rate causing periodicoscillations in the pressure and flow. However, since the mass flux ishigh enough to maintain the steam/water interface outside the downcomer,the overall condensation proceeds at a regular rate. This phenomenon isknown as condensation oscillation. It is characterized by a steady, periodic variation in the pressure which forces local structures within thetorus to vibrate in phase with the oscillations. Condensation oscillationhas been studied experimentally36»37 and analytically;37>38 however,the basic driving mechanism for the pressure resonance has not been identified.39
When the air/steam flow through the downcomer decreases to the pointwhere the condensation rate outside the pipe exceeds the steam flow exiting the pipe, the steam bubble collapses very rapidly. This results in alarge drop in the steam pressure and the steam-water interface rushes upinto the downcomer. Once there, the interface is warmed by condensing
steam and the condensation rate begins to decrease. At some point, thesteam pressure will rise, and the interface is pushed out of the downcomerto form an irregularly shaped bubble at the pipe exit. The bubble beginsto collapse; and the entire process, known as chugging, repeats. Chugging is characterized by rapid, irregular interface accelerations and pressure oscillations that cause large loads on the torus structure.
The chugging phenomenon is very similar to the condensation oscillation problem. It has also been studied in detail, with analysis methodswhich range from manometer-like models that attempt to predict the grossmotion of the interface, to probabilistic models that attempt to predictinternal chugging."40"1*2 The central problems which plague analysis ofthe chugging phenomenon are (1) high uncertainty in the basic condensationrates involved and (2) lack of understanding of the triggering mechanismfor bubble collapse.
Both of the condensation phenomena (chugging and condensation oscillation) involve transient, stochastic, turbulent, two phase flow. Because
210
of the complexity of these problems, no accurate assessment has been madeof the loads involved.30 Consequently, the analysis of PSP response tocondensation oscillations must rely on data from experiments which modelplant behavior. Thus, there is a need for improvement of the analysiscapability in this area.
B. SRV — Discharge Phenomena
When a SRV actuates, water and air initially in the discharge lineare immediately accelerated into the PSP. This results in air-clearingloads much the same as the pool swell loads discussed earlier. Theseloads are of no major consequence because they can be adequately "scaled"from the results of laboratory experiments.
Following the air-clearing phase, steam is injected at high velocityinto the PSP. Experience has shown that, depending on the discharge device used, condensation oscillations can occur as steam bubbles exit thepipe and collapse in the bulk fluid. As the steam bubbles collapse,severe pressure oscillations are induced on the surrounding structures.Current practice is to limit the severity of these oscillations by using aT-quencher device.^3 The T-quencher is a section of pipe (in the formof a "T") with holes strategically drilled in the arms to enhance localmixing. The T-quencher has been shown to be very effective in eliminatingthe severity of the pressure pulses, provided the local bulk fluid temperature is sufficiently low.
This is the factor that establishes a limit on the effectiveness of
the T-quencher — the local fluid temperature. If it becomes too high,the PSP will not be able to rapidly condense all the steam. Some of itwill escape into the air space above the pool and pressurize the torus.When the suppression pool pressure becomes 3447 Pa (0.5 psi) greater thanthe drywell, the vacuum breakers will open and steam will escape into thedrywell (forcing the temperature and pressure up). As more steam is discharged to the PSP, the drywell and torus pressure will continue to increase.
It is concern for this increase in drywell pressure (possibly resulting in overpressurization) that has led to the establishment of strictprocedures for the sequential ordering of SRV blowdowns to the PSP. Current practice involves sequentially venting SRVs on opposite sides of thetorus; this prevents excessive local temperatures near any dischargepoint.
More recently, concern has been expressed over the long term responseof the PSP during a prelonged Station Blackout. During part of thisscenario, the operator loses manual control of the SRVs. The long termresult is that a single SRV will continually open into the PSP. The localfluid temperature will monotonically rise, resulting in pressurization ofthe torus and possible condensation oscillations. The potential existsfor rupture of the torus due to overpressure coupled with violent pressureoscillations.
For example, this potential exists in the particular case of a Station Blackout with loss of the HPCI and RCIC systems. At about 100 mininto the transient, local pool temperatures at the discharge bay are postulated (c.f. Section 9) to be greater than 149°C (300°F). At that time,
211
steam condensation oscillations are expected to accelerate due to the excessive temperature and the continuous discharge of superheated, non-con-densible gases (from hydrogen generated in the melting core) into the PSP.These extreme conditions in the PSP yield a high probability for ruptureof the torus.
Along with the temperature rise assocated with SRV discharge comes agravity induced thermal stratification of the pool. This phenomenon is ofinterest because the layering tends to remain long after SRV discharge iscomplete. Since the stratification remains for a significant period oftime, the validity of the current design basis for Mark I PSPs becomesquestionable. This is because a fundamental assumption in virtually allthe transient analyses is that the pool is thoroughly mixed and at a uniform temperature following SRV blowdown. The impact of thermal stratification on the performance of the PSP remains to be evaluated.
Thermal stratification is also of interest because the temperaturedistribution in the PSP affects the fission product distribution in thetorus. Should a breach of primary containment occur, transport of the fission products to the environment will depend, to some extent, on the thermal stratification in the torus. Thus, an analysis of fission producttransport is coupled with the thermal stratification problem.
To the best of our knowledge, no information is available in the openliterature concerning systematic analysis of the thermal stratificationproblem. There is therefore the need for research in this area.
D.5 Pool Modeling Considerations
Model requirements
The prediction of suppression pool system behavior during the courseof a Severe Accident is an exceedingly difficult task. The previously described phenomena comprise a complex set of physical processes for whichfew detailed analytical models currently exist. These phenomena can beroughly divided into two types, i.e., thermal hydraulic phenomena andfluid-structure interaction phenomena. The initial efforts at ORNL aredirected toward the development of the thermal hydraulic model. A briefsurvey was conducted to identify existing multi-dimensional thermal hydraulic analysis codes which might be applied to the problem. The results of the survey are shown in Table Dl which presents a brief summaryof the candidate codes.
Due to the nature of the previously described phenomena, it was feltthat any model employed for suppression pool analysis should possess thefollowing characteristics:1. Appropriate hydrodynamics
a. two phase
b. non-equilibriumc. free field format (equations and constitutive relationships
independent of hydraulic diameter)d. incorporate a non-condensable gas fielde. employ multi-dimensional geometry (3 dimensions desirable)
2. Employ appropriate constitutive relationships3. Utilize efficient solution techniques4. Readily accessable to ORNL staff.
212
Table Dl. Candidate suppression pool analysis codes
Code Developer Geometry T-H characteristics
TRAC LANL 3D cylindrical 2 fluid
(PIA, PD2) nonequilibrium
COBRA-TF BPNL 3D cartesian 3 field
nonequlibrium
CCMMIX-2 ANL 3D 2 fluid
nonequilibrium
BEAC0N/Mod2 INEL 2D cartesian
cylindricalspherical
2 component vapor2 phasenonequilibrium
Requirements l.a and l.b are imposed by the complex nature of the condensation processes involved in SRV discharge and LOCA blowdown transients.Requirement l.c is a result not only of the nature of the SRV dischargeprocess, but of the geometric dimensions of the suppression pool. The enormous size and complex structural design of the suppression pool systemvirtually mandates that free field hydrodynamics be employed. Requirementl.d is imposed because significant amounts of non-condensable gases willbe released from the fuel and generated in metal-water reactions duringSevere Accidents. Requirement I.e. is imposed by the nature of the hy-drodynamic phenomena and the complex structural relationship of the SRVdischarge lines, vent downcomers, and ring header suction system. Appropriate constitutive relationships (requirement 2) are needed for closureof the hydrodynamic field equations. Many of the relationships employedin existing thermo-hydraulic codes may be invalid for use in the presentproblem. Requirement 3 is imposed by the fact that it will be necessaryto predict the behavior of the suppression pool for periods of time ranging from a few minutes to several hours. It is, therefore, important thatall codes employ fast solution techniques in order to minimize computingcosts. The fourth characteristic is desired because the computer codeselected for the analysis will require local modification as necessary toreflect the unique characteristics of the suppression pool problem.
We are unaware of any existing computer code or model which possesses all of the desirable characteristics outlined above. The initial
effort at ORNL is directed toward development of a pool model utilizingthe TRAC-PIA"*1* vessel module. TRAC (Transient Reactor Analysis Code)— PIA is a best estimate computer code developed at Los Alamos National Laboratory for analysis of large break PWR loss of coolant accidents. Although more recent versions of TRAC are available (i.e., PD2,PF1, and BD1), TRAC-PIA was chosen for this application because it is installed and operational on ORNL's computer system.
While TRAC-PIA does employ a full three dimensional, two fluid, nonequlibrium approach to reactor vessel hydrodynamics, it does not accountfor free flow or non-condensable gas fields. Additionally, the constitutive relationships employed In TRAC-PIA are based upon data which was
213
originally developed for vertical pipe flow and may not be valid for application to steam jet discharge phenomena.1*5 Though these limitationsare significant, it is felt that our access to TRAC-PIA will allow us toimplement any changes in the constitutive relationship package which mightbe necessary to reflect the characteristics of the suppression poolphenomena. LANL is currently modifying TRAC to include non-condensablegas field hydrodynamics (TRAC-PF1). ORNL's utilization of TRAC-PIA willallow us to quickly implement this modified version when it becomes available.
We have tentatively chosen the PELE-IC1*6 code for future use in analysis of the fluid-structure interaction within the suppression pool.PELE-IC (developed by LLL) couples a two dimensional Eulerian fluid dynamics algorithm to a Lagrangian finite element shell algorithm. The codecan couple either a one dimensional or a lumped parameter description ofcompressible gases, and can employ either cartesian or cylindrical coordinates. PELE-IC employs the basic semi-implicit solution algorithm contained in the SOLA code.1*7 The movement of free surfaces is treated ina full donar cell fashion based on a combination of void fractions andinterface orientation. The structural motion is calculated by a finiteelement method, from the applied fluid pressure at the fluid structure interface. The finite element shell structure algorithm uses conventionalthin shell theory with transverse shear and provides the fluid module withthe resultant position and velocity of the interface. The code is capableof analyzing both vent clearing and condensation related phenomena.
TRAC suppression pool model descriptionA block diagram of the ORNL-TRAC suppression pool model is shown in
Fig. D-3. The model is comprised of the TRAC VESSEL module which represents the pool hydrodynamics, and two FILL and PIPE modules, which represent the SRV discharge flow and T-quencher assembly. A detailed view of
ORNL-DWG 81-8150 ETD
FILL
PIPE PIPE
1 1
VESSEL
Fig. D.3 Suppression pool model block diagram.
214
the pool model is shown in Figs. D-4 and D-5. The pool is represented as112 cells in cylindrical geometry, located at five axial, four azimuthal,and eight radial regions. SRV discharge is assumed to occur as shown inFigs. D-4 and D-5. Half of the SRV discharge flow is directed radiallyinward from the fifth radial zone, while the other half is directed radially outward. It should be noted that the T-quencher is not modeled ina strictly physical fashion, since the discharge is assumed to occurthrough a single open ended pipe with a flow area approximately equal tothat of the T-quencher device. This approach was chosen for the initialmodel due to the significant reduction in input preparation time andmodel complexity associated with this approach. The initial water levelis assumed to be 4.47 meters above the pool bottom (i.e., at level 4 in
ORNL-DWG 81-8151 ETD
0, = 0.0165689 rad62 = 2.1053743 rad03 = 4.1942798 rad04 = 6.2831853 rad
Fig. D.4 Suppression pool model top view.
TOP VIEW
R, = 12.3901 mR2 = 12.9311 mR3 = 13.5239 mR4 = 14.7035 m
16.9926 m
19.2877 m
R7 = 20.4643 mR8 = 21.0601 mR9 = 21.6011 m
VESSEL
CENTERR6 R7 R8 Rg
ORNL-DWG 81-8152 ETD
HALF CROSS SECTION VIEW AT AA
B8 NO FLOW REGION• INITIAL FLUID CELLSM INITIAL VAPOR CELLS
Fig. D.5 Suppression pool model cross section.
|S3
H
216
Fig. D-5) with all cells in level 5 occupied by vapor. Internal structures (vent header, down comers, etc.) are not modeled.
Current status of suppression pool modelSeveral preliminary computer runs have been made in which TRAC data
base errors have been identified and eliminated. We are currently experiencing problems related to the limitations of TRAC-PIA in accounting forthe simultaneous existence of steam (vapor) and perfect gas (nitrogen)within the vessel. The TRAC user must specify that either vapor or airproperties be used everywhere within the code. In terms of the pressuresuppression pool, this means that the nitrogen gas (which is located inthe space above the pool) must be treated as subcooled vapor if one is toinject steam into the pool. It is also possible that the constitutive relationships within TRAC are producing instabilities in the solution due totheir dependence upon hydraulic diameter and fluid property information.We are currently working closely with the TRAC User Liaison Section atLANL in an effort to determine whether these problems can be overcome. Inthe event these problems prove to be insurmountable, we will re-evaluatethe remaining code candidates and select an alternative program for thesuppression pool thermal hydraulic analysis.
217
APPENDIX E
A COMPENDIUM OF INFORMATION CONCERNING THE BROWNS FERRY
UNIT 1 HIGH-PRESSURE COOLANT INJECTION SYSTEM
E.1 Purpose
The High-Pressure Coolant Injection (HPCI) System is designed to ensure adequate core cooling to prevent damage to fuel in the event of aloss of coolant accident that does not result in rapid depressurization ofthe reactor vessel. The HPCI System provides water to make up for thatwhich is lost through steam generated by decay heat.1*8
E.2 System Description
The HPCI System (Fig. E-l) consists of a steam turbine driven boosterpump — main pump combination and the associated piping and valves.The booster pump can take suction from either the condensate storage tankor the pressure suppression pool. The HPCI pumped flow enters the reactorvessel feedwater line "A" via a thermal sleeve connection. A test line
permits testing of the HPCI System at full flow while the reactor is atpower, with the main pump discharge routed to the condensate storage tank.A minimum flow line connects the main pump discharge to the pressure suppression pool, as a means of ensuring a flow of at least 0.038 m3/s (600GPM) through the pumps. The normal setpoint for HPCI pumped flow is18.93 m3/s (5000 GPM).
The HPCI turbine is driven by steam extracted from main steam line"B" upstream of the main steam line isolation valves. The two primarycontainment isolation valves in the steam line to the HPCI turbine arenormally open to keep the piping to the turbine at elevated temperaturesto permit rapid system startup (within 25 s of receipt of an initiationsignal). The normally closed DC-motor-operated steam supply valves upstream of the HPCI turbine will open against full system pressure within20 s after receipt of a system initiation signal. Signals from the control system open or close the turbine stop valve. The turbine controlvalve is physically attached to the HPCI turbine and is positioned by theturbine governor as necessary to maintain the pumped flow at the level setby the operator, normally 0.315 m3/s (5000 GPM). The turbine exhauststeam is discharged to the pressure suppression pool. The turbine glandseals are vented to a gland seal condenser. A small water flow divertedfrom the booster pump discharge is used to cool both the turbine lubricating oil cooler and the gland seal condenser. This cooling flow is returned, together with the gland seal condensate, to the booster pump suction. Noncondensible gases from the gland seal condenser are removed viaa DC-motor-operated blower to the Standby Gas Treatment System.
A vacuum breaker line (not shown on Fig. E-l) is installed betweenthe torus airspace and the HPCI turbine exhaust line. Its purpose is toprevent water from the pressure suppression pool from being drawn up intothe turbine exhaust line as the steam condenses in this line followingturbine operation.
ELEVATION 577 ft
CONDENSATE
STORAGE
TANK
218
CONDENSATE
RETURN
HEADER6
CONDENSATE
SUPPLY OHEADER ^
MAIN STEAM
SUPPRESSION
POOL
WATER
LEVEL
ELEVATION
536 ft
•hJvi-
SUPPRESSION
CHAMBER
HEADER
PUMP <t ELEVATIONJS24 ft
BOOSTER
PUMP
MINIMUM FLOW LINE
SYSTEM TEST LINEH^-
lA-
TURBINE
MAIN PUMPA
nnncTto Y J
'X
ORNL-DWG 81-8591 ETD
NORMAL WATER
LEVEL
ELEVATION 625 ft
THERMAL
SLEEVE
rELEVATION 574 ft
FEEDWATER
TOTAL
LEAKOFF
FROM
yy-TURBINE
-LUBE OIL
COOLER
TO STAND-BY GAS
TREATMENT SYSTEM
_t
-GLAND SEALCONDENSER
Fig. E.l High pressure coolant injection system.
All components required for operation of the HPCI System are completely independent of AC power, control air systems, or external coolingwater systems, requiring only DC power from the unit battery. On loss ofcontrol air, the HPCI steam line drains to the main condensers will failclosed; this is their normal position when the HPCI system is in operation.
The principal HPCI equipment is installed in the reactor building, ata level below that of the pressure suppression pool. The turbine-pump assembly is located in a shielded area so that personnel access to adjacentareas is not restricted during HPCI System operation. The only operatingcomponent located inside the primary containment is the normally open AC-motor-operated inboard HPCI steam line isolation valve, which will remainopen on loss of power.
219
E.3 HPCI Pump Suction
The HPCI pump can take suction either from the condensate storageheader or from the pressure suppression pool via the suppression pool ringheader. The normal lineup is for suction of the reactor-grade water in
the condensate storage header.Each Browns Ferry unit is provided with a 1419.4 m3 (375,000 gal
lon) condensate storage tank, which provides a water head to the storageheader for that unit. The storage header, which taps into the bottom ofthe storage tank, feeds the suctions of the high-pressure ECCS Systems,specifically, the pumps for the HPCI and the RCIC systems. The core spraypumps and the RHR pumps can be fed from this source, but are not normallyaligned to it. All other demand for condensate storage tank water is fedvia a standpipe within the tank; the standpipe height is such that 511.0m3 (135,000 gallons) of water is reserved for the ECCS Systems.
It is important to note that the local-manual opening of one normallylocked-shut valve will cross-connect the Unit 3 and the Unit 1 condensatestorage header; the opening of a second such valve will cross-connect allthree condensate storage headers.1*9 Thus the Unit 1 ECCS Systems have aminimum of 511.0 and a maximum of 1419.4 m3 (135,000 to 375,000 gallons)available through normally open valves from the Unit 1 storage tank andthese limits can be increased three-fold by manually opening two normallylocked-closed valves. Two additional condensate storage tanks, each of1892.5 m3 (500,000 gallon) capacity, have recently been installed at theBrowns Ferry Nuclear Plant. Thus it is unlikely that HPCI System operation will ever be limited by the availability of condensate storage headerwater.
The HPCI booster pump suction will be automatically shifted from thecondensate storage header to the suppression pool ring header if either:1. The water level above the booster pump suction falls to an elevation
of 168.4 m (551 ft). This would mean that the condensate storage tankhad completely drained and there remained just sufficient water in thecondensate storage header to effect a transfer before losing net positive suction head (NPSH).
or
2. The pressure suppression pool level increases to an indicated level of0.18 m (+7 in). Since the normal pool level is maintained between0.05 and 0.15 m ( 2 and 6 in.),50 this implies the addition ofbetween 257.4 and 371.0 m3 (68,000 and 98,000 gallons) of water tothe pool.
The 0.76 m (30 in.) diameter suppression pool ring header lies parallel toand beneath the suppression pool. Water flow from the suppression pool tothe ring header is via four 0.76 m (30 in.) diameter downcomer pipesspaced at irregular intervals around the torus. Each downcomer is cappedwith strainers at the torus end.
The change in HPCI pump suction lineup is accomplished by the openingof two DC-motor-operated valves in the line from the pressure suppressionpool to the HPCI booster pump suction followed by the closing of the DC-motor-operated valve in the suction line from the condensate storageheader. A check valve in the line from the suppression pool preventsbackflow from the condensate storage tank into the suppression pool duringthe change.
220
Two pressure switches are used to determine the water head from thecondensate storage tank above the booster pump suction and two levelswitches monitor the suppression pool level. In either case, just one ofthe two available signals is sufficient to initiate the shift in HPCIbooster pump suction. Once the booster pump suction has been automatically shifted to the pressure suppression pool, the operator cannot reposition the valves back to the condensate header suction lineup.
The minimum required net positive suction head (NPSH) for the HPCIbooster pump is 6.40 m (21 ft). This requirement is easily satisfied whensuction is taken from the condensate storage header which is at an elevation of approximately 8.69 m (28 1/2 ft) above the booster pump center-line.
When the HPCI booster pump suction is shifted to the pressure suppression pool due to a high indicated pool level of 0.18 m (+7 in.), thepool water level is 4.04 m (13.25 ft) above the booster pump centerline.Therefore, the NPSH requirement of 6.40 m (21 ft) can be met as long asthe temperature of the pumped water is below 85.0°C (185°F), assuming nocontainment back pressure.
E.4 System Initiation
Either of two signals will cause an automatic start of the HPCI System. These are:
1. Low reactor water level [12.09 m (476 in.) above vessel zero].2. High drywell pressure [0.12 MPa (2 psig)].
With either of these conditions, both the suction valve to the condensate storage header and the minimum flow bypass valve will open if theywere closed. The pump discharge valve between the main HPCI pump andfeedwater header "A" will open, and the two test line isolation valveswill close if they were open. The steam supply valve to the turbine willopen.*
The DC-motor-driven auxiliary oil pump starts and as the oil pressureincreases, the turbine stop and control valves open. Above 1800 rpm, themain oil pump which is driven on the turbine shaft takes over and maintains the oil pressure while the auxiliary oil pump shuts down. The minimum flow bypass valve closes automatically when the increasing HPCI Systempump flow exceeds 0.076 m3/s (1200 GPM).
The time from actuating signal to full flow is less than 25 seconds.The turbine control system will act to maintain a pumped flow of 0.315m /s (5000 GPM) into the feedwater line over a reactor pressure range offrom 1.14 to 7.83 MPa (150 to 1120 psig). If desired, the operator in theControl Room can operate the system in the manual or automatic mode at adifferent controlled flow.
The HPCI turbine operates at between 0.746 to 3.356 MW, (1000 and4500 horsepower) with a steam demand of between 6.17 to 23.18 kg/s (49,000and 184,000 lbs/h). System steam and pumped water flows at operating conditions are given in tabular form on Fig. 6.4.1 of the Browns Ferry FSAR.
*The two normally-open primary containment isolation valves in thesteam supply line will not reopen, if closed.
221
E.5 Turbine Trips
On an HPCI turbine trip, the turbine stop valve closes and the minimum flow bypass valve to the pressure suppression pool closes to precludedrainage from the condensate storage header into the suppression pool.The following conditions will cause turbine trip:1. High reactor vessel water level [14.78 m (582 in.) above vessel zero].2. High HPCI turbine exhaust pressure [1.14 MPa (150 psig)].3. Low HPCI booster pump suction pressure [—0.381 m (—15 in.) Hg].4. HPCI turbine mechanical overspeed (5000 rpm).5. Any HPCI isolation signal.6. Remote manual trip from Control Room.7. Manual trip lever on the HPCI turbine.All turbine trips except high reactor water level and HPCI isolation willreset automatically when the initiating condition clears. The high reactor water level signal can be reset manually, or will reset automaticallywhen the reactor water level decreases to the low reactor water level HPCI
initiation point, 12.09 m (476 in.) above vessel zero. The HPCI isolationsignal must be manually reset.
E.6 System Isolation
The Primary Containment and Reactor Vessel Isolation Control Systeminitiates automatic isolation of appropriate pipelines which penetrate theprimary containment whenever certain monitored variables exceed their preselected operational limits. The system is designed so that, once initiated, automatic isolation continues to completion. Return to normal operation after isolation requires deliberate operator action.
An automatic isolation signal for the HPCI System causes the inboardand outboard HPCI steam line isolation valves to close, trips the HPCIturbine, and closes the two motor-operated valves in the suction line fromthe pressure suppression pool. The inboard steam line isolation valve isAC-motor-operated, the outboard DC-motor-operated. The maximum closingtime for these valves is 20 s.
The following conditions cause a HPCI System isolation signal:521. HPCI System equipment space high temperature. Since high temperature
in the vicinity of the steam supply line or other HPCI equipment couldindicate a break in the turbine steam supply line, an isolation signalis generated if this temperature exceeds 93.3°C (200°F). This temperature is sensed by four sets of four bimetallic temperature switches.These 16 temperature switches are arranged in four trip systems withfour temperature switches in each trip system. The four temperatureswitches in each trip system are arranged in one-out-of-two takentwice logic.
2. HPCI turbine high steam flow. Since high steam flow could indicate abreak in the turbine steam supply line, an isolation signal is generated if the measured steam flow exceeds 150% of design maximum steadystate flow. The steam line flow is sensed by two differential pressure switches which monitor the differential pressure across a mechanical element installed in the HPCI turbine steam pipeline. The tripping of either switch at a differential pressure of 0.62 MPa (90 psi)initiates HPCI System isolation.
222
3. Low reactor pressure. After steam pressure has decreased to such alow value that the HPCI turbine cannot be operated, the steam line isisolated so that steam and radioactive gases will not escape from theHPCI turbine shaft seals into the reactor building. The steam pressure is sensed by four pressure switches from the HPCI turbine steamline upstream of the isolation valves. The switches are arranged in aone-out-of-two taken twice logic. The set point for this isolationsignal is 0.793 MPa (100 psig).
4. High turbine exhaust diaphragm pressure. A line tapping off the turbine exhaust line contains two rupture diaphrams in series, with thespace between vented to the HPCI equipment space through a flow restricting orifice. The diaphrams are designed for 1.138 MPa (150psig). If the pressure in the space between the diaphrams exceeds0.172 MPa (10 psig), a system isolation signal is generated.
5. Manual Isolation. If a HPCI initiation signal is present, the operator can cause HPCI system isolation by pushing a control panel pushbutton.
The low reactor pressure isolation will be automatically reset if reactor pressure is restored; all other isolation signals seal in and theoperator must push the HPCI auto isolation circuit reset push-button afterthe condition has cleared.
E.7 Technical Specifications
1. The HPCI System shall be operablea. Prior to startup from cold conditionb. Whenever there is irradiated fuel in the reactor vessel and the
reactor vessel pressure is greater than 0.945 MPa (122 psig), except
c. If the HPCI System is inoperable the reactor may remain in opera
tion for a period not to exceed 7 days provided ADS, core spray,LPCI mode of RHR and RCIC are all operable.
If these conditions are not met, an orderly shutdown shall be initi
ated and the reactor vessel pressure reduced to 0.945 MPa (122 psig)or less within 24 hours.
2. HPCI testing shall be performed as follows:a. Simulated automatic actuation test — once per operating cycleb. Pump operability — once per monthc. Motor operated valve operability — once per monthd. Flow rate at normal reactor operating pressure — once per
three months
e. Flow rate at 1.138 MPa (150 psig) — once per operating cycle3. Whenever HPCI is required to be operable the piping from the pump dis
charge to the last flow blocking valve shall be filled. Water flowfrom the high point vent must be observed monthly. (Purpose is toprevent water hammer when the system is started.)
223
APPENDIX F
A COMPENDIUM OF INFORMATION CONCERNING THE BROWNS FERRY
UNIT 1 REACTOR CORE ISOLATION COOLING SYSTEM
F.1 Purpose
The Reactor Core Isolation Cooling (RCIC) System is designed to ensure that the core is not uncovered in the event of loss of all AC power.The RCIC System provides water to make up for that lost through steam generated by decay heat during reactor isolation.53 RCIC is a consequencelimiting system rather than an ECCS System; the system design is not predicated on any loss of structure accident.51*
F.2 System Description
The RCIC System (Fig. F-l) consists of a steam-turbine driven pumpunit and the associated piping and valves. The RCIC pump can take suctionfrom either the condensate storage tank or the pressure suppression pooland discharges to the reactor vessel feedwater line "B" via a thermalsleeve connection. A full flow system test line permits testing of theRCIC System while the reactor is at power, with the pump discharge routedto the condensate storage tank. A minimum flow line connects the RCICpump discharge to the pressure suppression pool, as a means of ensuring aflow of at least 0.004 m3/s (60 GPM) through the pump. The minimum flowbypass valve closes at a system flow of greater than 0.008 m3/s (120GPM). The normal setpoint for RCIC pumped flow is 0.038 m3/s (600 GPM).
The RCIC turbine is driven by steam extracted from main steam line
"C" upstream of the main steam line isolation valve. The two primary containment isolation valves in the steam line to the RCIC turbine are nor
mally open to keep the piping to the turbine at elevated temperatures soas to permit rapid system startup (within 30 seconds of an initiation signal). The normally closed DC-motor-operated steam supply valve just upstream of the RCIC turbine will open against full system pressure within15 seconds after receipt of a system initiation signal.
The RCIC turbine exhaust steam is discharged to the pressure suppression pool. The turbine gland seals are drained to a barometric condenser,in which the steam is condensed by a water spray. The water spray is provided by a flow diverted from the RCIC pump discharge to pass through theturbine lube oil cooler and subsequently form the spray. The condensatefrom the barometric condenser is pumped back to the RCIC pump suction. Avacuum pump removes the non-condensibles from the barometric condenser andinserts them into the pressure suppression pool.
A vacuum breaker line (not shown on Fig. F-l) is installed betweenthe torus airspace and the RCIC turbine exhaust line. Its purpose is toprevent pressure suppression pool water from being drawn up into the turbine exhaust line when the remaining exhaust steam condenses after turbineoperation and shutdown.
All components normally required for initiating operation of the RCICSystem are completely independent of AC power, plant service air, and external cooling water systems, requiring only DC power from a unit battery
224
TO SUPPRESSION POOL<-
ORNL-DWG 81-8592 ETD
NORMAL WATER
LEVEL
ELEVATION 625 ft
ELEVATION 574 ft
FEEDWATER
Fig. F.l Reactor core isolation cooling system.
to operate the valves, vacuum pump, and condensate pump.55 On loss ofcontrol air, the RCIC drain lines to the main condensers will fail closed;this is their normal position when the RCIC System is in operation. Thedrain functions of these valves is transferred to overseat drain ports inthe turbine stop valves.
The principal RCIC equipment is installed in the reactor building ata level below that of the pressure suppression pool. The turbine-pumpassembly is located in a shielded area so that access to adjacent areas ofthe reactor building is not restricted during RCIC System operation. Theonly operating component located within the Primary Containment is the
225
normally open AC-motor-operated RCIC inboard steam line isolation valve,which will remain open on loss of power.
F.3 RCIC Pump Suction
The RCIC pump can take suction either from the condensate storageheader through a single normally-open DC-motor-operated valve or from thepressure suppression pool via the suppression pool ring header.
Each Browns Ferry Unit is provided with a 1419.4 m3 (375,000 gallon) condensate storage tank, which provides a water head to the storageheader for that unit. The storage header, which taps into the bottom ofthe storage tank, feeds the suctions of the high-pressure ECCS Systems,specifically, the pumps for the HPCI and the RCIC systems. The core spraypumps and the RHR pumps can be fed from this source, but are not normallyaligned to it. All other demand for condensate storage tank water is fedvia a standpipe within the tank; the standpipe height is such that 511.0m3 (135,000 gallons) of water is reserved for the ECCS Systems.
There is no provision for an automatic shifting of the RCIC pump suction from the condensate storage header to the pressure suppression pool.However, if condensate storage tank water is unavailable for any reason,the control room operator can shift the RCIC pump suction to the pressuresuppression pool ring header. This is done by remote-manually opening thetwo DC-motor-operated suction valves to the pool ring header; when thesetwo valves are fully open, the suction valve to the condensate storageheader will automatically close.
The minimum required net positive suction head (NPSH) for the RCICSystem is 6.10 m (20 ft), which is readily available with suction takenfrom the condensate storage tank. With suction from the pressure suppression pool, the required NPSH is available for suppression pool temperatures up to 85.0°C (185°F) with no containment back pressure.
F.4 System Initiation
An automatic start of the RCIC System is initiated by low reactorvessel water level at 12.10 m (476.5 in.) above vessel zero. The singlenormally-closed valve in the steam supply line opens, and steam is admitted to the turbine. (The turbine stop valve and control valve are normally open when the RCIC System is in standby.) The barometric condenservacuum pump starts, and the condensate pump will act to automatically control the water level in the condenser. The RCIC turbine speed and pumpflow are automatically maintained by the steam flow controller. Turbine
lube oil is supplied by a shaft-driven oil pump. The single .normally-closed valve in the pump discharge line to feedwater header "B" automatically opens, and the minimum flow bypass valve closes automatically whenpump flow exceeds 0.008 m3/s (120 GPM).
If the RCIC System is in an abnormal lineup when an initiation signalis received, the system will realign:
a. If the normally open pump discharge valve is closed, it will open.b. If the normally open pump suction valve from the condensate storage
header Is closed, it will open provided at least one of the suppression pool suction valves is not fully open.
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227
F.6 System Isolation
The Primary Containment and Reactor Vessel Isolation Control Systeminitiates automatic isolation of appropriate pipelines which penetrate theprimary containment whenever certain monitored variables exceed their predetermined operational limits. The system is designed so that once initiated, automatic isolation goes to completion. Return to normal operationafter isolation requires deliberate operator action.
An automatic isolation for the RCIC System causes the normally-openinboard and outboard steam supply isolation valves to close, the turbineto trip, and the two RCIC suction valves from the pressure suppressionpool to close (if they were open). The minimum flow bypass valve to thesuppression pool is interlocked to close whenever the turbine is tripped.
The following conditions cause a RCIC system isolation signal:!• RCIC System equipment space high temperature. Since high temperature
in the vicinity of the RCIC equipment could indicate a break in theturbine steam supply line, an isolation signal is generated if thistemperature exceeds 93.3°C (200°F). This setpoint is based on thecalculated AT with a 0.001 m3/s (15 GPM) steam leak in the space.There are 16 temperature sensors arranged in four trip logics withfour sensors in each logic. The 16 sensors are physically arranged infour groups with four sensors in each group. One sensor in each groupis in each of the four one-out-of-two taken twice trip logics.
2. RCIC System high steam flow. Since high steam flow could indicate aleak in the turbine steam supply line, an isolation signal is generated if the measured steam flow exceeds 150% of design maximum steadystate flow. The steamline flow is sensed by two differential pressureswitches which monitor the differential pressure across an elbow installed in the RCIC turbine steam supply pipeline. The tripping ofeither trip channel at a differential pressure of 11.43 m (450 in.)H20 initiates RCIC System isolation.
3. Low reactor pressure. After steam pressure has decreased to such a lowvalue that the RCIC turbine cannot be operated, and there is no coolant spray to the barometric condenser, the steam line is isolated toprotect against continuous gland seal leakage to the RCIC equipmentspace. The set point for this isolation signal is a reactor vesselpressure of 0.345 MPa (50 psig). The pressure is sensed by four pressure switches at the RCIC turbine steam line upstream of the isolationvalves. The switches are arranged in a one-out-of-two taken twicelogic.
4. High turbine exhaust diaphragm pressure. A line tapping off the turbine exhaust line contains two rupture diaphrams in series, with thespace between vented to the RCIC equipment space through a flow restricting orifice. The diaphrams are designed for 1.138 MPa (150psig). If the pressure in the space between the diaphragms exceeds0.172 MPa (10 psig), a system isolation signal is generated.
5. Manual Isolation. If an RCIC initiation signal is present, the operator can cause RCIC System isolation by pushing a control panel pushbutton.
All isolation signals are sealed in and must be manually reset afterthe condition causing them has cleared. A control panel pushbutton isprovided for this purpose.
228
F.7 Technical Specifications
1. The RCIC System must be operable prior to startup from a cold condition or whenever there is irradiated fuel in the reactor and the reactor vessel pressure is above 0.945 MPa (122 psig).
If the RCIC System is inoperable, the reactor may remain in operationfor a period not to exceed seven days if the HPCI System is operableduring such time.If these conditions are not met, an orderly shutdown of the reactormust be initiated and the reactor depressurized to less than 0.945 MPa(122 psig) within 24 hours.
2. RCIC testing shall be performed as followsa. Simulated automatic actuation test — once per operating cycleb. Pump operability — once per monthc. Motor operated valve operability — once per monthd. Flow rate at normal reactor operating pressure — once per three
months
e. Flow rate at 1.138 MPa (150 psig) — once per operating cycle3. Whenever RCIC is required to be operable the piping from the pump dis
charge to the last flow blocking valve shall be filled. Water flowfrom the high point vent must be observed monthly. (Purpose is toprevent water hammer when the system is started.)
229
APPENDIX G
EFFECT OF TVA-ESTIMATED SEVEN HOUR BATTERY LIFE ON
NORMAL RECOVERY CALCULATIONS
Section 7, Computer Prediction of Thermal Hydraulic Parameters forNormal Recovery, is based on the assumption that the 250 vdc unit batteries will fail in four hours under the conditions of Station Blackout.
After review of the draft results presented in Sect. 7, the Electrical Engineering Branch at TVA performed a battery capacity calculation to determine how long the Browns Ferry 250 vdc batteries would last under theseconditions, and arrived at an estimate of seven hours. This appendix examines the impact of the newly estimated seven hour battery failure timeon the results and conclusions of Sect. 7.
Two basic conclusions are presented in Sect. 7.1. If AC power is recovered any time in the first five hours, then
a normal recovery will be possible if the 250 vdc batteries havenot failed.
2. If the 250 vdc batteries fail at four hours and the AC powerremains unavailable then the time between battery failure andfirst uncovering of fuel will be at least three hours if the reactor has been previously depressurized as recommended by thisreport.
If the unit batteries were to last seven instead of four hours, therewould be a slight lengthening of the time interval from battery failure tocore uncovery due to the slightly lower decay heat. The most importantquestion to be answered regarding the extension of the estimated batteryfailure time from four to seven hours is whether there is some other system failure caused by the increased temperatures and/or pressures duringthis period that would complicate or make impossible a normal recovery ifAC power were restored.
To answer this question calculations similar to those shown in Figs.7.1 through 7.9 of Sect. 7 were performed. These new calculations startat an initial point four hours after station blackout and extend to sevenhours. The batteries are assumed to last throughout this period and operator actions are the same as detailed in Sect. 7.3.1, Normal Recovery-Assumptions.
Due to the elevated pool temperatures experienced during the periodfrom four to seven hours, it has been necessary to modify the calculationof the fraction of relief valve (SRV) discharge that is quenched in thesuppression pool. When the pool temperature is within its normal operational range, the SRV discharge is completely condensed (quenched) in thecool water surrounding the submerged SRV discharge nozzle (T-quencher).However, if the temperature of the water around the T-quencher is sufficiently close to saturation, then the condensation will be less than 100%and some of the steam will reach the suppression pool atmosphere. Monticello tests58 showed that an approximately 28°C (50°F) difference between bulk and local pool temperature can exist during extended dischargethrough a single SRV. In this case, one would expect less than 100%quenching to begin at a bulk pool temperature of about 74°C (165°F), corresponding to 100°C (212°F) near the SRV discharge point (provided the
230
pool is at atmospheric pressure). For the purposes of the calculationsreported in this appendix, the following provisions were made for the calculation of quench fraction in the pool:
1. The temperature of the water surrounding the T-quencher duringSRV discharge is assumed to be 28°C (50°F) higher than the bulkpool temperature.
2. If the vapor pressure of the water surrounding the T-quencher isequal to or greater than the static pressure then 0% of the SRVdischarge is quenched; if the vapor pressure is 5.0 psi or morebelow the static pressure then 100% of the discharge isquenched. Variation of quench fraction is linear between thesetwo points.
This should provide a reasonable upper limit estimate of containment pres-surization due to non-quenched SRV discharge. Consideration of this effect was not necessary for the zero to five hour results reported in Sect.7 because bulk pool temperature does not exceed 82°C (180°F) during thefirst five hours of a Station Blackout.
Results of the four to seven hours after blackout calculations are
shown in Figs. G.1 through G.9. From these results it is concluded thatsystem parameters remain within acceptable ranges during this period:
1. The 250 vdc batteries by TVA estimate last the full seven hours.2. Reactor vessel level is within the normal control range, about
5.08 m (200 in.) above the top of active fuel.3. Reactor vessel pressure is being controlled by operator action
at about 0.69 MPa (100 psia).4. At the seven-hour point, about 414 m3 (109,265 gal) of water
have been pumped from the condensate storage tank, which had anassured capacity of 511 m3 (135,000 gal) before the blackout.
5. At the seven-hour point, average suppression pool temperature isabout 92°C (198.2°F), but this should not be a problem for theT-quencher type of SRV discharge header piping.
6. Containment pressures would be elevated to about 0.23 MPa (33psia), well below the 0.53 MPa (76.5 psia) design pressure.
7. Drywell atmosphere temperature is about equal to the 138°C(281°F) design temperature.
These results show that if power were recovered within seven hours ofthe Station Blackout a normal recovery would be possible.* The 92°C(198°F) bulk pool temperature after seven hours may cause elevated airtemperature in the RCIC and HPCI spaces, but this would not be expected tolead to failure of the HPCI or RCIC, because the lube oil of both units iscooled by the water being pumped — in this case, water from the condensate storage tank which would not exceed 32°C (90°F). The periods of RCICoperation are very infrequent after the first four hours of Station Blackout, with about an hour between actuations.
The supply of control air for remote-manual operation of the SRVs issufficient for the full seven hours. As discussed in Chap. 3 of this report, the accumulators provided for the six relief valves associated withthe ADS system are sized to permit five operations or a total of 30 actuations. As illustrated in Sect. 7 and this appendix, less than 30 actuations are required during the first seven hours of the Station Blackout.
*With the assumption of no independent secondary equipment failures.
a.
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Fig. G.l Reactor vessel pressure - four to seven hours afterstation blackout.
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240 270 360 390
Fig. G.2 Vessel steam flow - four to seven hours after stationblackout.
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OPERATOR CONTROLS LEVEL BY REMOTE-MANUAL
ACTUATION OF RCIC SYSTEM
TOP OF ACTIVE FUEL
300 330
TIME (min)
360 390
Vessel level - four to seven hours after station
420
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Fig. G.5 Total injected flow (pumped by RCIC from condensatestorage tank) - four to seven hours after station blackout.
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EXHAUST IF OPERATING
240 270 300 330
T.ME (min)
360 390 420
Fig. G.6 Suppression pool level - four to seven hours afterstation blackout.
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270 300 330
TIME (min)
ORNL-DWG 81-8599 ETD
SUPPRESSION CHAMBER
ATMOSPHERE
CONDENSATION AND HEAT TRANSFER
FROM ATMOSPHERE ARE DOMINANT
WHEN SRV NOT DISCHARGING
360 390 420
Fig. G.7 Suppression chamber temperatures - four to seven hoursafter station blackout.
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DRYWELL DESIGN TEMPERATURE
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DRYWELL ATMOSPHERE TEMPERATURE — REACTOR COOLANT TEMPERATURE (CAUSED
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oREACTOR COOLANT TEMPERATURE SEE FIG. G.1)
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ORNL-DWG 81-8601 ETD
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240 270 300 330
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Fig. G.9 Containment pressures - four to seven hours afterstation blackout.
rO
OJVO
1. W. J. Armento
2. T. E. Cole
3. W. A. Condon
4. D. H. Cook
5. W. B. Cottrell
6. W. G. Craddick
7. G. F. Flanagan8. S. R. Greene
9. R. M. Harrington10-14. S. A. Hodge
15. T. S. Kress
16. R. A. Lorenz
241
Internal Distribution
NUREG/CR-2182, Vol. IORNL/NUREG/TM-455/VIDist. Category RX, IS
17. A. L. Lotts
18. T. W. Robinson, Jr.19. H. E. Trammell
20. B. J. Veazie
21. C. F. Weber
22. R. P. Wichner
23. D. D. Yue
24. Patent Office
25. Central Research Library26. Document Reference Section
27-28. Laboratory Records Department29. Laboratory Records (RC)
External Distribution
30-31. Director, Division of Accident Evaluation, Nuclear RegulatoryCommission, Washington, DC 20555
32-33. Chief, Severe Accident Assessment Branch, Nuclear Regulatory Commission, Washington, DC 20555
34. Director, Reactor Division, DOE, 0R0, Oak Ridge, TN 3783035. Office of Assistant Manager for Energy Research and Development,
DOE, 0R0, Oak Ridge, TN 3783036-40. Director, Reactor Safety Research Coordination Office, DOE,
Washington, DC 2055541-42. D. B. Simpson, Tennessee Valley Authority, W10C126 C-K, 400 Com
merce Avenue, Knoxville, TN 3790243-44. Wang Lau, Tennessee Valley Authority, W10C126 C-K, 400 Commerce
Avenue, Knoxville, TN 3790245-46. R. F. Christie, Tennessee Valley Authority, W10C125 C-K, 400 Com
merce Avenue, Knoxville, TN 3790247-48. J. A. Raulston, Tennessee Valley Authority, W10C126 C-K, 400 Com
merce Avenue, Knoxville, TN 3790249-50. H. L. Jones, Tennessee Valley Authority, 253 HBB-K, 400 Commerce
Avenue, Knoxville, TN 3790251-52. Technical Information Center, DOE, Oak Ridge, TN 3783053-552. Given distribution as shown under categories RX, IS (NTIS-10)
*U.S. GOVERNMENT PRINTING OFFICE: 1981-740-062/291
