ML20125C589

Auxiliary Feedwater Sys Reliability Analyses for B&W Reactors. Revised for Davis-Besse

ML20125C589
Person / Time
Site:	Davis Besse
Issue date:	01/04/1980
From:	Dorman R, Enzinna R, Weaver W BABCOCK & WILCOX CO.
To:
Shared Package
ML20125C587	List: ML20125C584 ML20125C589 ML20125C593
References
BAW-1584, BAW-1584-02, BAW-1584-2, TAC-43516, NUDOCS 8001100482
Download: ML20125C589 (31)

Text

.s Docket No. 50-346 License No. NPF-3 Serial No. 573 January 4, 19 80 AUXILI ARY FEEDWATER SYSTEM RELIABILITY ANALYSIS FOR THE DAVIS-BESSE NUCLEAR GENERATING STATION

, UNIT NO. 1 By W. W. Weaver R. W. Dorman Revision 1, Novecher 1979 1

Babcock & Wilcox '

Power Genera tion Groep Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Va. 24505 l

l 90008114 8001100 IJ L/F 2

_ 1 l

TABLE OF CONTENTS Page

_.Section Executive Summary 1.0 In troduc tion 1.1 Background .

1.2 Objectives

1. 3 Scope 1.4 Analysis Technique

1. 5 Assumptions & Cri teria 2.0 Sys tem Description 2.1 Overall Configuration 2.2 Supporting Systems & Backup Water Source 2.3 Power Sources 2.4 Instrumentation & Control

2. 5 Operator Actions 2.6 Tes ti ng 2.7 Technical Specification Limitations 3.0 Reliabili ty Evaluation 3.1 Faul t Tree Technique 3.2 Comparative Reliabili ty Resul ts 3.3 Dominant Failure Contributors Re fe rences Appendix A 90008115 Appendix B 1

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. LIST OF FIGURES

1. Davis-Besse 1 AFWS

2. Alternate Water Supply (Service Water System)

3. AC Power Distribution to APdS Valves

4. Functional Logic Diagram for Train 1 Initiation and Steam Generator Isola tion

5. Comparison of Davis-Besse 1 AFWS Reliability with NRC Results for Westinghouse Plants 90008116

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' 0 EXECUTIVE

SUMMARY

j l The itRC has requested all operating plants with Babcock & Wilcox (B&W) designed reactors to consider means for upgrading the reliability of their

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i Auxiliary Feedwater Systems (ARIS). As a part of the response to this request, Toledo Edison and the other B&W Owners Group utilities have requested B&W to perform a simplified reliability analysis of existing auxiliary feedwater systems.

This draft report presents the results of that reliability study for the l I Davis-Besse AFWS. l 1

The primary objective of this study was to evaluate Davis-Besse AFWS reliability (defined as " point unavailability") using an approach which would produce results comparable to those obtained by NRC staff analyses for Westinghouse and Combustion Engineering Plants. Another objective was to identify dominant failure contributors affecting system reliability.

ARIS reliability was assessed for three cases: Loss of Main Feedwater (LMR4) l with reactor trip, LMFW with Loss of Offsite Power (LMR4/ LOOP) and LMRI with Loss of all AC power (LMFW/LOAC). System reliability was assessed by the construction and analysis of fault trees.

The results of this study are on the following page. These results indicate the Davis-Besse AFWS reliability, based on the reliabilities obtained by the flRC for Westinghouse plants, is medium for LMFW, medium for LMRl/ LOOP, and low to high for LMFW/LOAC. For the LMFW/LOAC case, the AFWS is unavailable at 5 minutes because of the AC dependency of the AFWS motor-operated valves;  ;

however, for 15 and 30 minutes, the reliability improves because the operator can manually operate these valves.

Dominant failure contributors which were identified in this study include i

1) the AC dependency of the motor-operated valves, and 2) system unavailability resulting from outages for preventive maintenance. -

A similar study will be performed for each Owners Group utility and additional plant specific draft reports will be prepared. At the conclusion of the program, information contained in the plant specific reports will be collected and used to generate an ARIS reliability report comparing all B&W operating pl an ts .

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CASE 1: Lii!FW CASE 2: LOOP CASE 3: LOAC J.~ lI  !

LOW HIGH MED LOW IAED HIGH LOW

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b li!!SSION SUCCESS WITH!N 5 MINUTES O--O RANGE OF W PLANTS ' - ' i i

w O MISSION SUCCESS WITilit! 15 MINUTES

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, o CASES 1 & 2.

I l o @ MISSION SUCCESS V!! THIN 30 MINUTES

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[ , COMPARISON OF DAVIS-BESSE-1 AFWS RELIABILITY WITH NRC ,

,i jl 00 RESlJLTS FOR W PLANTS '

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,' '1.0 In troduction 1.1 Ba'ckground This raport presents the results of a reliability study for the Davis-Besse Unit 1 Auxiliary Feedwater System (AFWS). The NRC is conducting similar analyses for Westinghouse and Combustion Engineering plants. Preliminary results of the NRC study are available (Reference 1) and have been included in this report for comparison with the Davis-Besse 1 A&ls relia-bili ty. The approach employed in this study has been developed in close coordination with the NRC and is therefore expected to yield comparable res ul ts .

1.2 Objectives The objectives of this study are:

e To perform a simplified analysis to assess the relative reliability of the Davis-Besse 1 AnlS. It is intended that the results of this analysis be directly comparable to those obtained by the NRC for Westinghouse and Combustion Engineering plants. This is assured by the use of the same. evaluative technique, event scenarios, assumptions and reliability data used by the NRC.

e To identify, through the development of reliabili ty-based insight, dominant failure contributors to AFWS unreliability.

1.3 Scope The Davis-Besse 1 AnlS was analyzed as it existed on August 1,1979. Three event scenarios were analyzed:

o Case 1 - Loss of Main Feedwater with Reactor Trip (LM&l).

o Case 2 - LMal coincident with Loss of Offsite Power (LMFW/ LOOP). .

o Case 3 - LMB! coincident with Loss of all AC Power (LMFW/LOAC).

These event scenarios were taken as given; that is, postulated causes for these scenario.s and the associated probabilities of their occurrences were not considered. Addi tionally, external common mode even ts (earthquakes ,

fires, etc.) and their effects were excluded from consideration.

For each of the three cases, system reliability as a function of time was eval da ted.

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1.4 Analysis Technique  ;

The evaluation of reliability for the AFWS was based primarily on the construction and analysis of fault trees. This technique encourages the development of insights which permit identification of the primary contributors to system unreliability. Application of this technique is described in detail in Section 3.1.

1.5 Assumptions and Criteria Assumptions and criteria were made in consultation with the NRC staff and  ;

were selected to assure that the Davis-Besse 1 reliability evaluation results 1 l

will be comparable to those obtained by the NRC for the Westinghouse and ,

1 Combustion Engineering analyses. )

1) Criterica f or !!ission Success + In order to evaluate the overall reliability contribution of system components, it is necessary to establish whether or not failure of those components will prevent successful accomplishment of the AFWS mission. Thus, it is necessary to explicitly define the criterion for mission success. The criterion adopted for this study was the attainment of flow from at least one pump to at least one steam generator. Mission success can be alternatively defined as at least one running pump with suction to a source of water and an open flow path to at least one generator without flow diversion.

System reliability was calculated at times of 5,15, sad 30 minutes to allow for a range of operator action. These times were specifically 1

chosen because NRC-supplied operator reliability data for these times l was available; however, these times are reasonable and consistent with LXFW mitigation f or B&W plants. In their study, the NRC staff has used steam generator dryout time as a criterion for successful AFWS initiation, and the 5 minute case represents a comparable result for B&W plants since auxiliary feedwater delivery within 5 minutes will prevent steam generator dryout. However, steam generator dryout itself does not imply serious consequences; a more appropriate criteria is the maintenance of adequate core cooling. Recent analysis (Reference 2) have shown that with no RCS break adequate core cooling can be maintained without AFWS operation, providing that one makeup pump and the startup feed pump is operated and the PORV is opened within 30 minutes of loss of main feedwater.

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.' 2) Power Availability - The following assumptions were made regarding power avail abili ty:

LMFW - All AC and DC power was assumed available with a probability of 1.0.

LMPNLOOP - One diesel generator was unavailable with a probability of i

10-2 . The other generator was assumed available with a probabili ty of 1.0.

LMPNLOAC - DC and battery-backed AC were assumed availableivith a probability of 1.0.

3) NRC-Supplied Data ,NRC-supplied unreliability data for hardware, operator actions and preventive maintenance were assumed valid and ,

directly applicable. These data are listed in Appendix B. l l

4) Small Lines Ianored - Lines on the order of 1-inch were ignored as possible flow diversion paths.

5) Coupled Manual Actions - Manual initiation of valves with identical

- function was considered coupled. Such valves were assumed to be both opened manually or both r.ot opened. The case in which one valve was opened and the other valve was lif t closed was not considered.

6) Deoraded Failures - Degraded failures were not considered; that is, components were assumed to operate properly or were treated as failed.

7) Backuo Water Sources - No credit or penalties were taken for the suction header connections to the startup feed pump, the fire water system or the deaerator storage tank.

8) Relief Valves - It was assumed that 2 relief valycs stuck open would defeat the steam supply to the affected turbine until automatic or manual isolation of the associated steam generator is accomplished.

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. 2.0 Sys tem Description 2.1 Overall Configuration A diagram of the Davis-Besse 1 AFWS is presented in Figure 1. This safety grade system consists of two interconnected trains, capable of supplying auxiliary feedwa ter to either or both steam generators under automatic or manual initiation and control.

2.1.1 Suction The primary water source for both AFWS trains consists of two interconnected condensate storage tanks. Each of the tanks has a capaci ty of 250,000 gallons; a combined reserve of 250,000 gallons is required by Tecnnical i Specifications for AFWS use. The water level in both tanks is indicated l in the control room.

A single suction heade.r conducts wa ter from the tanks to the normally-open AC motor-operated pump suction valves , HV786 and HV790. This header

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contains several additional normally closed connections which were not addressed in this study.

An alternate water source for AFWS use is available from the service water system; details of this source are provided in Section 2.2.

2.1.2 Pumos and Discharge Cross-connections Each train has a turbine-driven pump rated at 1050 gpm with a design

., recirculation flow of 250 gpm. Thus, each pump is capable of supplying 800 gpm against a steam generator pressure of 1050 psig (safety relief i valve set pressure).

The pumps are protected against cavitation by pump suction pressure, switches; the switches for each pump are interlocked with both the primary and cross-connect steam supply valves for the corresponding turbine (e.g. , the pressure switch for AFWS pump 1 is interlocked with valves HV105 and HV106A) . In the event that low suction pressure should occur for a pump, its steam supply valves will close, stopping steam flow to the pump turbine.

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,- Steam generator level is controlled by varying turbi)ne speed. 1 1

Turbine speed is controlled in part by the turbine governor valves ICS38A and ICS388; operation of these valves is further described i

in Section 2.4.2. Auxiliary feedwater flow is also affected by the opera- I tion of speed-controlled valves HV360 and HV388.. As turbine speed increases above 2800 RPM, these valves open; as turbine speed decreases l below 1100 RPM, the valves close. Both speed-controlled valves are  !

bypassed by restricting orifices which are capable of passing a flow of 20,0 gpm (nominal) when the valves are closed. For the purposes of this analys" it was assumed that each speed-controlled valve must be open to obtain successful auxiliary feedwater delivery via the associated pump.  !

The primary discharge paths for the train 1 and train 2 pumps are through )

normally-closed AC motor-operated valves HV3870 (train 1) and HV3872 (train 2). An alternate cross-connection path, which permi ts each pump ,

i to feed the opposite, steam generator, is available via normally closed AC motor-operated cross-connect valves HV3S69 and HV3871. Au toma ti c selection of the auxiliary feedwater flow path and opening of the associated valves is under the control of the safety grade Steam-Feedwater Rupture Control Sys tem (SFRCS) and is described in more detail in Section~ 2.4.1. The flow of auxiliar !

feedwater to each steam generator from ei ther path must also pass through normally-open AC motor-operated s team generator isolation valves MV608 and MV599. These valves are also under control of the SFRCS.

2.1.3 Steam Supoly for the AFWS Turbines Steam supply for each turbine is available by opening a normally closed AC motor-operated valve (HV105 orHV107) in the primary steam path from the steam generator normally fed by the turbine pump. A cross-connect steam source is available from the opposite steam generator by opening another normally-closed AC motor-operated valve (HV106A andHV107A).

Automatic operation of these steam supply valves is under control of the SFRCS, described elsewhere.

Other valves in the s team admission path include the turbine overspeed stop valve which trips closed on turbine overspeed (and which must be reset locally) and the turbine governor valve. The turbine governor valve controls turbine speed to control steam generator water level. .

Cor. trol for these valves is described in Section 2.4.2.

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Exhaust from the tu. sines passes through check valv.s) into a common exhaust line and is vented to the atmosphere via an exhaust silencer.

2.1.4 Other System Features Normal, recirculation for each pump is provided by a 2-inch line containing.

locked-open valves, a - check valve and a restricting orifice. This line returns " mini-flow" recirculation, 250 gpm, to a condensate storage tank crosstie or drain. Of more significance to reliability is the full flow test recirculation line containing locked closed valves AF21 and AF22, an'd AF23. If these valves are inadvertently left open in certain combi-nations, it is assumed 'that full pump flow could be diverted to the condensate storage tank or drain.

2.1.6 Valve Operation and Indication With the exception of the turbine governor valves (hydraulic valves which make use of DC motor-operated speed changes), all other motor-operated valves in the AFWS are powered by 430 VAC. In the absence of AC power,

• these valves will remain "as is". All such valves are position indicated in the control room. The power for control and position indication for these valves is derived from the power for the valve motor operators.

Control switches for the AC motor-operated valves are provided in the ,

1 con trol room. These switches can be used for manual operation of the ARIS  !

valves whenever the SFRCS is not controlling those valves; however, these switches will not manually override an SFRCS control signal to the SFRCS controlled valves (HV106, HV106A, HV107, HV107A, HV3870, HV3869, HV3871, HV3872, MV608, MV599). Similarly, pressure switch signals to the inter- 1 locked valves cannot be remotely overridden (valves HV106, HV106A, HV107, HV10'/A,HV786,HV1382,HV790,HV1383). Manual control of the speed-controlled valves (HV360 and HV388) and the turbine governor valves (ICS38A and ICS388) is available at all times.

2.2 Succorting Systems and Backup Water Source The AFW turbines and turbine-driven pumps are self-contained enti ties without dependencies on secondary support systems. Circulation of lubri-cating oil for both the turbine and pump is shaf t powered; lube oil cooling is provided by the pumped fluid. ,

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l The only support sy jm of significance is the baci _iwater supply from the

I service water system. A simplified diagram of the portion of this system {

relating to the AFWS is shown in Figure 2. Water can be made available to both trains of the AFWS via normally-closed AC motor-operated alternate suction valves HV1382 (train 1) and HV1383 (train 2). This water is provided by three service water pumps which are on diesel generator-backed power. Normally two of these pumps are kept running

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l at all times. I Automatic switchover to this backup water supply is initiated by the detection of low suction pressure by pressure switches immediately l upstream of suction valves HV786 (train 1) and HV790 (train 2). These pressure switches will cause HV786 and HV790 to close and alternate j

suction valves HV1382 and HV1383 to open, thus making service water available to AFW pumps. (These switches will not cause swi tchover to occur in the event that HV786 or HV790 are inadvertently closed.)

2.3 Power Sources I

A simplified diagram showing the AC power distribution for the AFWS components is provided in Figure 3. As shown, AC power for all AFUS components necessary to achieve auxiliary feedwater flow is derived , _

from diesel generator-backed busses. Normally, (Case 1), power for these busses is obtained from the switch yard. However, in the event of

, LMFW/ LOOP (Case 2), the diesel generators are automatically started and AFWS components will remain operable with no operator action required.

,, As shown in the figure, train 1 valves. (and the backup water supply for train 1, not shown) are powered by diesel generator 1, and, similarly, train 2 is powered by diesel generator 2. Thus, a failure of one diesel generator to start will not prohibit the initiation of auxiliary feedwater flow.

In the event of LMFW/LOAC (Case 3), automatic initiation of auxiliary feedwater flow will not occur because of the AC dependence of AFWS valves; manual actions, described in Section 2.5, are requi red.

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, *2.4 Instrumentation ano Jontrol -

2.4.1 Initiation Logic A functional logic diagram illustrating the means of AFWS initiation for train I is shown in Figure 4. This logic'is a part of the SFRCS and is ,

on battery-backed power. The diagram is greatly simplified and does not l

show actual hardware redundancies, the availability of manual initiation, and some interlocks.

1 As indicated in the figure, the train 1 AFW flow can be initiated by a low water level in either steam generator, low steam generator 2 pressure, loss of all four reactor coolant pumps, or high reverse differential 1 l

pressure across main feedwater check valves. Among other things, this last condition will result from the loss of both main feedwater pumps, in which case reverse MFW AP signals will originate from both steam generators.

SFRCS logic is designed to isolate both main and auxiliary feedwater to l a bad generator (e.g. , a generator with a rupture in the associated steam  !

l piping) and automatically align AFW valves to feed the remaining good s team generator. This capability is reflected in the' AFMS initiation logic shown in Figure 4. If steam generator 1 pressure is not low, then, upon AFW initiation, valves in the primary feedwater and steam supply paths between steam generator 1 and AFWS turbine / pump 1 are opened. If, however, steam generator 1 has low pressure and steam generator 2 does not, the cross-connect valves betneen ARl turbine / pump 1 and steam generator 2 are opened and the primary path valves are closed. SFRCS logic is designed so that actuation signals do not lock on; valve align-ment is automatically changed to suit existing conditions. The logic shown in the figure is duplicated for train 2.

2.4.2 Con t rol Af ter . initiation, control of steam generator level is accomplished by varying turbine speed via control signals to the turbine governor valves.

Signals for control come from level transmitters in each steam generator.

Separate control hardware is provided for each train. This equipment is powered from battery-backed sources and is separate and independent from the Integrated Contrc1 System. Manual control of the turbine speed may be required dapending on' the steam generator level desired.. )

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, . 2.4.3 Ins trumen tation -

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AFWS instrumentation in the control room, in addition to the valve position indications previously described, includes:

e Level Indication for each condensate storage tank.

__ e AFW p, ump low suction pressure alarm for each pump.

e AFW pump discharge pressure for each pump.

o Steam generator outlet pressure.

s AFW flow indication obtained from flow measurement devices located upstream of the AFW steam generator isolation valves.

e Steam generator startup range levels.

All AFWS instrumentation is powered from battery-backed sources.

2.5 Operator Actions l Assum'ing no component failures have occurred and the system is correctly configured, no operator actions are required to achieve AFWS mission success in Cases 1 and 2. In Case 3, it will be necessary for the operators to manually open valves in the field.

These valves are HV106, HV3870, and HV360 for train 1, and/or HV107, HV3872 and HV383 for train 2.

2.6 Tes ti ng

. The AFW turbines and pumps are tested monthly, using the mini-flow recirculation line, to assure operability. Valve position for active j valves are also checked monthly to assure correct positions. Level control i equipment and pressure switch interlocks are tested monthly to assure operabili ty.

Every 18 months, the AFWS is tested extensively. This testing includes a full flow test of the pumps recirculating via AF21 or AF22 to the condensate storage tanks,, and a demonstration that valves actuate to the correct position and the pumps start on receiving an AFWS actuation signal.

I Level control and pressure switch interlock equipment is calibrated on an 18-month basis.

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Technical Soecifica ;n Limitations ))

' '2. 7 Te,chnical Specifications require that both AFWS trains be operable; however, if one train is determined to be inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are allowed for restoring operability of the train. After that, the plant must be taken to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Technical specifications also require that 250,000 gallons of water be reserved in the condensate storage tanks for AFWS use. Should this source become unavailable, continued operation for up to 7 days is permitted by relying on the service water system. In this case, operability of the service water system must be demonstrated every 12 hou rs .

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. 3.0 Reliability Evaluation ,

3.1 Faul t Tree Technique The Davis-Besse 1 AFWS reliability was evaluated by constructing and analyzi~ng a faul t tree. The fault tree developed during this study is contained in Appendix A. The top level event in this tree is failure to achieve mission success; from this point, the tree branches downward to a level of detail corresponding to NRC-supplied data. This level is generally indicated by basic event circles.

For construction of the first tier of the tree (page A-1), the AFWS components in each train were grouped into three categories - Suction, Pump and Discharge. System failure can result from Suction 1 - Suction 2, Pump 1 - Pump 2, or Discharge 1 - Discharge 2 failures or from failures within one train when the other train is out of service for preventive maintenance. The fault tree also accounts for system failures resulting from combination failures such as Pump 1 - Discharge 2 with the appropriate discharge cross-connection inoperable. All combinations considered are indicated by the first tier.

The techniques used in fault-tree construction and the symbols shown in Appendix A are consis tent wi th those used in WASH-1400 .(Reference 3).

Following completion of the tree, hand calculations were performed to

'obtain system unavailability for 5,15 and 30 minutes for each of the three event scenario cases.

3.2 Compara tive Reliabili ty Results i The resul ts of the an~alysis are presented in Figure 5. Indicated in this figure are the system reliability results for each of the three cases and for each time 5,15 and'30 minutes. The basic format for this figure, including the characterization of Low, Medium, and High reliability, was adopted from information presented by the NRC in Reference 1. Because the NRC-supplied input data were of ten unverified estimates of component and human reliability, absolute values of calculated system reliability mus t be de-emphasized; results have significance only when used on a relative basis for purposes of comparison. Accordingly, the intent of Figure 5 is to show the relative reliability s tanding of the Davis Besse 1 AFWS for each of the three cases and also to compare these results to the l 1

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1 NRC results for West..nghouse plants. The Westinghouse results and

.' numerical values permitting construction of Figure 5 were all obtained. l from Reference 1. It should be noted that there is a scale change for the Case 3 results; reliability results for Case 3 cannot be cmss-compared with Cases 1 and 2.

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As shoTin~ in Figure 5, relative to Westinghouse, Davis-Besse 1 has medium reliability for Cases 1 and 2; Case 3 has low reliability for success in 5 minutes, but high reliability for success in 15-30 minutes. The under-lying causes for these reliability results are described in Section 3.3 Some general observations may be made regarding the resul ts in Figure 5.

As the time for operator action increases from 5 to 30 minutes, the probability of mission success improves. Most of the improvement occurs between 5 and 15 minutes, reflecting a significant difference in the NRC-supplied operator reliability data for these times. On the other hand, there was little difference in the operator reliability data between 15 and 30 minutes and this is reflected in the system unavailability results.

The small difference in tha results for Cases 1 and 2 indicates the effect of using turbine-driven ' umps in both trains as well as relatively small

'effect associated with tae improbable loss of one diesel generator. The Case 3 results stem from the AC dependence of all AFWS motor-operated valves.

3.3 Dominant Failure Contributors 3.3.1 Case 1 - LMFW This reliability evaluation did not reveal any outstanding design deficiencies for either Case 1 or 2 that would individually make a significant contribution to overall system unavailability. This results in part because of the excellent separation between the two trains of the .

AFWS. Dominant failure contributors, therefore, tend to represent combi-nations of the more probable random component failures.

The dominant failure contributor for Case 1 is turbine / pump failure in one train coupled with preventive maintenance related outage of the other i train. The second most important contributor is turbine / pump failures l involving both trains. Such combined failures could involve, for example, ,

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s.- o the failures of 'a va' ' innediately upstream of one rbine (such as the l turbine stop valve) coupled with a failure in the other train (such as a mechanical problem with the other pump).

l 3.3.2 Case 2 - LMFW/ LOOP' ,

The absence of offsite power has no effect on the dominant failure con tribu tors . Loss of one diesel generator reduces overall system availability because the valves in one train can no longer be remotely ope ra ted. Nevertheless, system success will still depend largely ci the same dominant failure contributors (i.e. , component failures) for the ,

other train as in Case 1.

s 3.3.3 Case 3 - LMFW/LOAC i The dominant failure contributor for Case 3 is the AC dependence of the AFWS mo to,r-cpe ra ted val ves . The system will be unavailable for the 5-minute case. For the 15 and 30 minute cases, there is a very good chance that the operator will open the three valves needed to restore at least one AFWS train to operability. Operation of the sys tem there-after depends on battery-backed power sources and remaining failure 1 1

contributors will resemble those of Cases 1 and 2. '

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REFERENCES 0

. 1) " Auxiliary Feedwater Reliability Study", an NRC staff presentation to the ACRS at the ACRS meeting of July 26,1979,1717 "H" Stree t, Room 1046, Washington, D.C.

2) MghMellt *p_ff]E45 t=c@f3_g2ii-)p~_(ly_Qtg.lr .,lo %_iQ ~

! ' 2* I N.E3_L7__s{c)5.cl J5 5___(5,_19).9._l)_C_C kf5__N.__9_53 - SLt6,

3) WASH-1400 (NUREG-75'/014), " Reactor Safety Study (Appendix II),"

USNRC, October 1975.

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'. l Figure 2 DAVIS-BESSE UNIT l-ALTERNATE WATER SUPPLY (SERVICE WATER SYSTEM)

TO TO TRAIN 1 TRAIN 2 (HV1382) (HV1383)

A A

/; s;  ;+l lt;

~j ~j Z Z Z r r >

/ \ / \ / \

., PUMP PUMP PUMP l-1 13 1-2 PDFERED POWERED PO#ERED FROM FROM FROM DIESEL GEN. 1 EITHER DIESEL GEN. DIESEL GEN. 2 a -

a a 90008134 i

SUCTION FROM INTAKE STRUCTURE l

I i

_ . . - _ ~. . - .__ _ _____ - __ _ _ _ __ _

~

_.:.i..___._.. ._ _ - . . . . . . . - -- - - - - -- ~ ~~~ ~~~ ~ ~ ~

. .' -)

FROM FROM 13.8KV BUS 13.8KV BUS GENERATOR GENERATOR n;n ugn 3r "C1" qr y "D1" v WW WW mm mm 480V 3r BUS "El" 480V y BUS "Fl" 97 9P L J 1F V Y L 2 y

TO MCC'S, BATTERY TO MCC'S, BATTERY CHARGERS & INVERTERS CHARGERS & INVERTERS AFWS VALVES:

AFWS VALVES:

HV380 HV388 MV808 MV599 HV1382 HV1383 HV786 HV790 HV3889 HV3871 HV3870 HV3872 HV108 HV107 HV108A HV107A Figure 3 DAVIS-BESSE-l AC POWER DISTRIBUTION TO AFWS VALVES

~

90008135

~

, ' EGEN0:

L e

[ AND CLOSE < -

MV608 d OR 3g,3 CLOSE 4 @ NOT HM 06 ISOLATED (FEED CLOSE 2

SG 1 LO PRESSUR SG-2 IF M HV3870 (AND NOT IT'S OK) BLOCKED)

~

OPEN 4 '

HV106A p] /% SG.2 L0 PRESSUR OPEN <

s HV3869 E Z e

CLOSE <

HY106A i

CLOSE < DECAY HEAT g

HV38S9 VALVE OH 11 l CLOSED - 'l l

< SG-1 MFW' LOSS O '

SG-1 -< OPEN < < HI REV. AP BOTH OK HV106 4 -

< SG 2 MFW MINE (FEED SG-1) 4 b HI REY.

AP OPEN 4 < SG-1 LOW LEVEL MV608 OPEN < < SG-2 LO# LEVEL I u HV3870 l

) ,

< - LOSS OF ALL RCP'S '

SFRCS i Figure 4 DAVIS-BESSE IgFUNCTIONAL LOGIC l

OlAGRAM FOR TRAIN 1 INITIATION AND S/G ISOLATION FOR AUXILI ARY l FEE 0*ATER I NOTES 1 REQUNDANCIES, MANUAL INITI AT!0N, AND INTERLOCKS NOT SHO#N.

2. LOGIC DUPLICATED FOR TRAIN 2 #!TH VALVE SUBSTITUTIONS:

TRAIN 1 TRAIN 2

)

HV106 HV107 HV106A HV107A $QQQ@j36 HV3870 HV3872 HV3869 Hy]871

• MV608 MV599 OHil DH12 .

3 VALVES ARE AVAILABLE FOR MANUAL CONTROL UNLESS THE SFRCS M69 QCCW A r Tit t Tr n

. r CASE 1: LMFW CASE 2: LOOP CASE 3: LOAC ,.'

LOW MED HIGH LOV' MED HICll LOW

• ME0* HIGH* -

5 d b N

• DAVIS- g [] ,

BESSE UNIT I

. 30 @ Q g i All W PLANTS g ,g g g g g  !

l -

b MISSION SUCCESS WITHIN 5 MINUTES H RANGE OF W PLANTS j e MISSION SUCCESS WITHIN IS MINUTES *THE SCALE FOR CASE 3 IS NOT THE SAME AS FOR O

CASES I & 2.

co Q MISSION SUCCESS WITHIN 30 MINUTES C; Figure 5 COMPARISON OF DAVIS-BESSE-1 AFWS RELIABILITY WITH NRC N

RESULTS FOR W PLANTS

1 l

-. g ,,. . . rs - m .a - :_ _. .,,es an - - . -

_ . _ . ., -- w-- - - - .r.~ - . - l

- . . )

. v -

J APPENDIX A

/

"~ ~

DAVIS-SESSE 1 FAULT TPEE 90008138 l 1

1 l

1 1

i 1

S

/

I

,e INSUFFICIENT -

FLOW FROM AUXILLARY '

FEEDWATER SYSTEM -

n SUCTION DISCHARGE PUMP PUMP PE BELATED SUCTION FAILURES FAILURES FAILURES DISCHARGE FAILURES . PUidP

s FAILURES E

/\ A /\ P 4\ .

A 4\

TRAIN I TRAIN 2 IN PM IN PM w

O O

O

-s

_ _ _ _ _ _ . _ _ - ____._-______1

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

e, e p ,.

I

$UCfl0N (A lltE

. l

. . J i

CONDENSATE  !!AVICE FAILURE BATER FAILURE I

LDSS of CONO(NIATE a

!!0 RACE SUP Ptf I

$1 $2 lhACEVA NV185 CHECK WANUAL gyppty ,,i!gg All AF2 yyygg

-VALVE PtVG5 CCN0t%3 Aft PLUG: CLOSE3 pty;;

p(y;; STCR101 TANKS CLOSE3 80TH INA0E00 ATE SUPPLY INA0V[Rf(NTLY saquAL IN TAhK5 AND LEFT CLOS [0 QUTLET VL- QNE QUTL(i PLUE3 ANO i

PLUGS VALVI PLUC5 NOT OPENED j O

=

QUrur vn

,NAmum am 1,,N,,

g;?

,, A m urr

,ms PLUG 3 CLO;E3 pgggg 73 g,gy 90008140 un< 6 ,o INDICAflCN FAIL 3 yaig ti g l

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

~

.?

, **. )'

l SERVICE EATER

. FAILURE l l

NO O MANUAL ACT 0 OPEN SVC -

WATER l NO FLOW FRC4 SVC FRC4 SVC WATER 1 TRAIN WATER TRAIN 2

SW5 HVl382 SUPPL 3*3 PLUGS CR FAILS TO FROM SERV.

PLUGS INAD. WATER SYS, GPEN fag (3 CLOSED

(%

ECHANICAL WANUAL PLUGS OPENihG FAILS 90008141 WOTCH CONTROL PuvER OPERATCR CIRCUlf FAILS FAILS

.s em AAAAA

HV360 88 FAILS TO FAILS TO OPEN OPEN

-w

{\

MECHANICAL PLUGS NO MANUAL I

REMOTE ^

OPENING NO YANUAL AUTO OPENING SIGNAL FAILS TO OPEP IML TOR CONTROL PONER R CMT OPERATOR -

, i circuli CLC20 IAll3 FAILS 90008142

a Q Q "- * * *

$g

• i =

h. m u w M

.=

. . 0.

==

==

=

=.

=_

= _

=

.=

==

E=

!! -G ^

. == . _

.E. =

.! =.

-CC G-  !!!

.-= _=

=.

$f5

= _ .

= t

= V.

= s'

_=

=. sS i

1 l

1 j

==

!5

! n

==

u --

_=

i

• a*

43- "E 2

. =

=,

-=

.= .

E =.

-.)

s =

=f C

5E \

@  ;* / {

. _= .

i~5 N!= 90008143 i , "" -- _ -- . ,_ _ -_- - - - - - -

. o .

-<![

e3 .

l 1

[\ [h l

,_] -

AF MV608 AF43 3g UV599 PLUGS FAILS pggg3 W

(

INADVERTENTLY ptUG: LEFT CLOSED AND NOT OPENED o

MV6CS [NAO, FAILS TO LEFT OPEN CLg3[0 o

,- J em SL INITIAft;N PLUGS 90008144 TO OPEN WANUAL CP!:ll.*lG FAILS

. s0In NOTCR gggg;;gt P0 s t '. .opEtATOR CIRCulf FAILS 112 7

p , a a __ s y y ., e _a, , , _

e t

., .2

) ,

e e 439 thd S

=

5-t M

m

=s.f -

[r 2

j

=

=r l E: ,

• r

( L _.

U

. .E We 232 E*2 w w

$ Y 69 s

~O yIe C .3

. i .r I W D 44

2. ~.; .:'

_ E w em u

_ IOE O

5- V  ;

E 2

-" 3 5'

z 4

h

., 5 a

• 3 5

o

=5 f

-C- _ 90008145 2 -

= .

• ,o

• ~' '

.. .e

E 2d O C. 5 .

Q C5 m

2g

.=

.. "gG 2d ""E

== .

O$ i 0 t e ~.

r-t E=5 C (

N E r iS .

M*"

o M

O .

8"

O. I e f ) 1 E 1:t 5,5 2 -

~

C T5u C ==~:

0 5 E_

.e w ..

- J

/ EO5 h.e

<A D 55 [F  !

s = =.

v ss=  ;

, U

- =. = $.

== - -

r~- r .

' d E

- =

90008146

%,, =. .=

a

_ -, - u .T

e e O

.% I Fua9

,

• FaltuRIl /

I I L813 OF :suon pygp g att 5fg i ust g,,

g

. SUPPLT pus, f Fif t

.L OCKA GE

~1 I

A 1 l

. . . . ,,,,,,,,,L, ,,,,, ,,B, ,

C:IE ! ST5fts f4Cu suff (5 FAtt Sfs GEM i (Alts filt3

'3 GT F Allt 2177

IE .1 M7

IE 2 I

$ffis !"PPtf Ifsits Facs -

a a WTf:0 fal' 'JFIT:13 TAIL Sfs :141 73 EIAf 73 T2Af F A tts Il ss b '

I l Ill I3 ptggs piggs If0P 1 EPI" Vatf! Ct:$10

_ 1 i i a'ca ,,,,. ,,,,, r . . ..

8 0 ptyg5 ptyg; $f 7 I I"

Vatet Ctt!!S I

ss,)J -

,1,,,

...... ,,,,, ,,,,,4 90008147

+m-A o- 1--+< a- o- a a am nL Jm ----e -- m- ~6-L- en i-a A

a PUSP I FAlt!

FLCl PuuP 8tAAING NO Sf!AM gpgg giugntgg tuR8thE (CN AN IC AL C3OLING gg,yggt AVAILAllt ptygg THR3UEM FAltuRE II' TO CRME WR81NE I

• IAIfIg g CIRCUt m

I

}

I

$ TEAM FRCW , 337:3 IU88'"I SIN.I NOT $107 ptygg AVAltABLE NO FLCf 10 THRCUO:4 CR033 fit VALVE I Siu EthiRATCR 1 TO PRC30CE AP

( ST[Ay

-J f

10 enN Ptuss-4>DYttPEC Jt

<0CA<

BANUAL gg CVICPit3 CDCitt H OPENING TRIPS VALVI FAILS WECHANICAL v -

3ANUAL QP!NING FAIL 3 N0 1 30fC2 CCMfROL I L--J Nrg' IO'<A CPER A TCR CIRCULI

PiNis: nits ,Alu 90008148 W

CHAn(L "I ' SUCfl04 g 3(y

, a,, Ct:s n "umi l

l

e q

e n

.)

e N

$ E

• 2 i.

=-

as B

a mm

= a E

~a W.

.m .- -

= , .U E E a

ee O

tef u N E

- em U

5 y

. == = -

C g ,, -

g-s "Sd g.,,,>,,

, =-=

G esu . - tan

=a 4.9 u 5EW IC*$$

s s:

_, :# y '

" W - ,

WE G

=$5b O h5$ C J

.issa

-- - & =:- C- V 555 -

Du* -

J 09

== Mu ,D ,- -

4 D =

w w as NE 8 h*

W w = *-

w %9 m 2

- N N m m.

med -

u a w m 2 =#

- g 2 -u 3 M.3 n D = H d T e as ch == w u =.W -

w G B

D I C" " I E c

(W "

.3*

y M EOO y 2 5 ~~

W se 49 a

.M D em N N

i - M eD o- - ]! =3 mama k a

.h ~h5 r-

-d

\ ___ lhlN ~

u ., ,

OM w n8 M ZM u

2 90008149 A s

STM FROM GEN. 2 NOT AVAILABLE THROUGH CROSSTIE F S T0 i i E OPEN "

n

-m (D (h MECHANICAL PLUGS OPENING FAILS

[SFR r ISOLATION

[Eh INITIATl010F VAlii BY FUNCTION  ! DEFEATit0 SFRCS A?D ROUTE ew OPEllNG OF VALVE V

T i POWER OPERATOR I CIRCUIT 90008150

STEAM SUPPLY SYSTEM FROM STM GEN 1 FAILS

[D l

l S

/

HV106 - -~

FAILS TO MS726 i MS730 IUR8INE OPEN PLUGS STOP VALVE n

-w Fm S 0 MS7 MS OPEN PLUGS PLUGS I STOP VALVE l 90008151

n -

. 1 y _

• f- 5 .3 2

,

? :, ,

=i -c- ,,,

n-i 1-I l

1 05 3.

.

,=_

. . =.

.==

=. si i; Tm

- a: 7 u

I I

i we

__. .= ,U

=N*

E** -

_=

0. .*.

w~

_=_

a, G

=..

= _

-!-G-3_

sy.:

i-! -

.1 C*

C  !!

=.,= .

=

== .

y ,

8 55 )

o _

G n Ei v

.=

,

=

= ,.

=.

9

~

s.g.

90008152 a e

.s - *..e.'

as e

e TEAin l IN 98 I

I ftals i f341s !

CDspontaf3 FAILuf ts i 15 73 Tus li%t 5f37

'I383 NY3f f 8 NV3449 7037 1 1

SUG T l *dt V;eP 300 g g pgggg falls f0 sinus Camratsgrica Out 70 -

CPt4 (af 31

• sv4081*(aF43

• svilt

• C20t 51Fttiti) l l

C2013 fit fil38 gp)9

! NOT FAILS pgggg Pt 24 SLE

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

FAILS f 3 pgggg FAILS ptyt3 OUIPUI f 803 SPt t St4 2 2 gg pj y ,,,,,

38 2 FAIL Pt943  % IAll! IG "J 4.2Af :Pt u 90008153 l

~

[' O 7 3

I

.- - 1 1

A .

TRAIN 2 IN PM (h

l TRAIN 1 TRAIN 2 FAILURES COMPONENTS IN PM

[\ [b i r% r% l I

SINE STOP HV388 I HV3872 HV3871 1 PUMP 2 '

2 PUMP h HY360 SUCTION l j i I

() Ett Il 90008154 A . //

~ ,**.,* .,

SUCTION PUMP FAILURES SUCTION 1 SUCTION 2 AND AND PUMP 2 PUMP 1 (h D SUCTION 1 SUCTION 2 PUMP (S (3 PUMP 2

1 FAILS FAtts S1

\

S2

^

A\ ^

p0Ft0, FROM SVC

, NO Fw, l FRCM SYC l TRAIN 1 90008155.

. .: ] .' . .

,- ..' APPEilDIX B D '

.I D D NRC-SUPPLIED DATA USED FOR PURPOSES OF C0f DUCTI!!G A COMPAPATIVE ASSESSMEtiT OF EXISTit:G

-g,45 DESIGi!S & THEIR POTEt!TI AL REL! ABILITIES Point Value Estimate

., of Probability of' Failure on Demand I. Comconent (Hardware) Failure Data

a. Valvet:

Manual Valves (Plugged) N1 x.10~

Check Valves N1 x 10~

Motor Operated Valves Mechanical Co:rponents 3 Plugging Contribution m1 s1 xx 10-10 4 Control Circuit (Local to Valve) '

w/ Quarterly Tes ts s6 x 10-3 ._

w/ Monthly Tests s2 x 10-3

b. Pumps: (1 Pump) ,

Mechanical Components N1 x 10-3 Control Circui t w/ Quarterly Tests 3 l w/ Monthly Tests N7 44 xx 10-10 3

. c. Actuation Locic N7 x'10-3

• Error factors of 3-10 (up and down) about such values are not unexpected for basic data uncertain ties.

90008156 B-1

II. Human Acts & Errors - Failure Data: ,.'

. +

'" 'I

+ Estimated Human Error / Failure Probabilities + .

+ Modifying Factors & Si tuations + , , ,

With Local Walk- -

. With Valve Position Around & Double *

~

Indication in Control Room Check Procedures w/o Either Point Est on Point Es t' on Point Est on Value Error Value Error Value Error

- -

• Estimate . . Fa c to r Es tima te Factor Es tima te Factor  !

A) Acts & Errors of a Pre- -

Accident Nature 3

1. Valves misposi tioned ) I during tes t/ main tenance. .

a) Specific single 1 10 10 , 10 y valve wrongly selected Ri x 10-2 x1 X 20 120 x 10-2 x.1X x f' N out of a population of .

valves during conduct of a test or maintenance .

act ("X" no. of valves '

in population at choice). .

b) Inadvertently leaves %5 x 10-4 20 %S x 10-3 10 ' . N10-2 10 correct valve in urong posi tion.

-4 -3

~ 2. More than one valve -is %1 x 10 20 N1 x 10 10 N3 x 10 yg ,

affected (coupled errors).

O O

O N

4

ppaildi'x 8 ,

m ,

'li[ Human Acts & Errors - Fa: Iure Data (Cont'd):

1

+ Estimated Human Error / Failure Probabilities +

. Estimated Failure

. Prob. for Primary J Time Actuation Opera tor to Actuate Needed- AFWS Comoonents B) Acts & Errors of a Post- .

Accident 14ature

1. Manual actuation of .s5 min. s5 x 10-2 AFWS from Control N15 min. 41 x 10-2 Room. Considering N30 min. s5 x 10 "non-dedicated" operator to actuate AFWS and possible backup actuation of A FWS .

III. Maintenance Outace Contribution Maintenance outage for pumps and EMOVS:

0 Maintenance  % 0.22 (# hours720 / maintenance act) 90008158 B-3

~

. Docket No. 50-346

' ~

RESPONSETO([UESTION1BOFATTACHME:;rATOv~.F.ROSS License No. NPF-3 (NRC) LETTER DATED 8/21/79 -

Serial No. 556 November 13, 1979 Question . .

Attachment 2

~ '

13. Provide justification of relief and safety valve flou models used in the CRAFT 2 code.

Dp D ,D _9 } A

" l RESPONSE ,

c o,] _1 d 2 The CRAFT 2 code, which is' documented in topica1 report BAW-10092, Rev. 2, 1 does not have any special models for prediction of the fluid discharge through the relief and safety valves. Ra ther ,' they are modeled as leak 1

paths from the pressurizer control volume to the contairacnt. Thus, the i Bernoulli (orifice) equation is used for subcooled cIischarge, while the I Moody correlation is,used for. saturated steam or two-phase discharge.

These models are the same as those used in B&W's ECCS Evaluation Model.2 Since little information exists on the flowrate through pressurizer valves for 'subcooled or two-phase fluid conditions,'it is impossible to ascertain the accuracy of this modeling technique. Since pressurizer l'eaks are in-herently less severe than the brechs in the cold les punp discharge piping analy:cd to demonstrate conpliance to 10 CFR 50.46, a truly realistic. I model for the discharga rates is not necessary. However, the codeling '

technique utilized is expected to reasonably approximate the discharge rates and their subsequent effcet on the RCS. l System response to relief valve actuation have been analyzed and submitted to the Staff in Section 6 of the May 7,19793 , report. The cases speci-fically analyzed were:

1. A loss of =ain feeduater accident which results in actuation and a subsequent sticking open of the pressurizer relief valve was addressed.

Offsite power was assumed to remain available and only one '!DI train was used for emergency core cooling. This analysis is similar to the THI-2 event that occurred on March 23, 1979, and deacustrated that, if one HPI pump remained availabic, no core uncovery would hr.vc oc-curred. ,

2. A stuck' open FORV assuming a loss of of fsite power and only one HPI train availabic was analyzed. Results of this evaluation denonstrated that core uncovery would also not occur.

90008159

o . . .

. , . . i An additional analys.a of the effect of a pressurizi ' break which supple-

. mented those presented in reference 3, was provided to the Staff in a 1cteer from J.H. Taylor (B&W) to R.J. Mattson (NRC) dated May 12, 19794 .

That analysis examined the effect of the stuck open PORV case, Case 2 above,

~ '

except the auxiliary feedwater system was assumed inoperable. The results of that evaluation showed that, even without auxiliary feedwater, one HPI pump can handle the accident provided that realistic decay heat valves are l 1

utilized. In all of these evaluations, the PORV was' modeled via a leak '

~

path representation in the CRAFT 2 code. The; orifice area of the PORV was modeled as the leak area (1.05 in.2) and a discharge coefficient of 1.0 was utilized. - - - -

~  !

The method for modeling the PORV described above does, result in a pre- l dicted secam ffowrate, at the valve rated pressure, which is in excess of the design (rated) flowrate. An ' alternative modeling approach is to use

'a discharge coefficient (C D) which, at the valve rated pressure, would yield the valve rated flowrate. For the 177-FA plants, this is a C D

approximately 0.85. For the first two cases described above, this codel-ling approach would result in a slower system depressurization and a slouer discharge of the RCS inventory. Thus, the use of a C = 0 m ed . -

D in previous evaluations results is a conservative assessment of the tran-sient. For the third case, the use of a smaller C wu result in a D

larger repressurization following the loss of the SG as a. heat sink and the change in the discharge'from steam to'two-phase flow. 'However,'use of a C D f 0.85 would result in an inventory loss less than that calcu-lated in reference 4 and no core uncovery would occur.

Besides the cases involving actuation of the pressurizer relief valves,  !

analyses were perfor=ed for a total loss of SG heat sink and are provided in references 5 and 6. In those evaluations, the pressurizer safety valves were exercised. To model these valves, the leak path representa-tion uas used uith the leak path opening and closing at the opening set- l point of the valve. The valve arca and CD was chosen such that the rated

{

flowrate for the valve would be simulated at the valve rated pressure.

$ecause of the large. relief capacity of the valve, the system pressure oscillated within a few psi of the valve setpoint and the valve was exercised intermittently. 'Thus, any discrepancies between the modeled 90008f60 e ~ ~

, 'e .

l 2 ~

/* and'the actual relic capacity.of the pressurizer si ,!t y valve is not' ex- l

. pected to significantly alter the system response.'

While there is little information available on the discharge rates through ~

'the pressurizer valves, it is also important to note the breaks in the *

~

pressurizer are bounded by breaks in the cold leg pump discharge piping.

Pump discharge breaks are analyzed to shou conformance of the ECCS to meet' the criteria of 10 CFR 50.46. The reason that cold leg breaks bound breaks I

, 'in the pressurizer was discussed in. detail in reference 3. Therefore,.it -

l is not necessary to simulate the . actual relief capacitics of the pressuri-

\

zer valves in order to demonstrate the ability of the ECCS to mitigate the

~

i consequences of a loss of RCS inventory through th'e valves within the' criteric 'of 10 CFR' 50.46. -

REFERENCES 1

BAW-10092, Rev. 2, " CRAFT 2 - FORTRAN Program for Digital Simulation of a Multinode Reactor Plant During LOCA," R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, April 1975.

2 BAW-10104, Rev. 3, "B&W's ECCS Evaluation Model," 3.M. Dun, et al.,

~

August 1977.

3 Letter J.H. Taylor (B&W) to R.J. Mattson, May 7,1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177-FA Plant." (

4 Letter J.H. Taylor (B&U) to R.J. Mattson, May 12, 1979.

5 Letter from R.B. Davi to 177 Owner's Group,

Subject:

" Complete Loss of Feeduater Transient," September 11, 1979.

6 Letter from R.B. Davis to Fe. C.R. Domeck,

Subject:

" Complete Loss of Fecdwater Transient on Davis-Besse," September 11, 1979.

e 90008161 .

e ,

- e 4

I * " -

' Docket No. 50-3 6 <

RESPONSE TL WESTION 2A 0F' ATTACimENT A 0F 21/79 . License No. NPF-3 D.F. ROSS LETTER Serial No.'556 November 13, 1979

~ .. Attachment 3 Question .

2. Provide justification that' the 3 node steam generator model used in the CRAFT 2 analysis of small breaks is adequate for the prediction -

of steam generator heat transfer.

~

' RESPONSE /

The B&W ECCS Evaluation Modell for small breaks utilizes a three-node repre-sentation, in the CRAFT 2 simulation, for the prediction of steam generator heat transfer following a small break. Two o.f tne nodes, stacked verti-cally, are used to model the primary side of the once through steam genera-tor (OTSG). Th'e upper node includes the hot leg piping, from'the center on the 180* U-bend at the top of the vertical section of the hot leg to the SG upper head, the upper head of the SG, and the upper one-half of the tube region. The lower node simulates the lower one-half of the tube region. The third node is used' to model the secondary side of the OTSG. i

, To evaluate the suitability of this modeling technique, the unique charac- 1 l

1 teristics of the' 0TSG and its effects on the small break transient must be j cxamined. As is shown later, for small breal:s evaluated'with the auxiliary feeduater system operable, heat removal via the SG is not necessary for the worst case breaks, i.e., those that result in core uncovery, in order to successfully mitigate the transient. For the smaller breaks, heat re-moval via the SG is necessary. The three-node representation utilized

, appropriately models the heat trancfer characteristics of the OTSG. For the smaller breaks, heat removal via.the steam generator is necessary and the heat transfer' characteristics of the OTSG must be appropriately con- '

sidered. Although the 3-node SG model does not rigorously account for  !

the heat transfer process that will occur, it does provide a reasonabic representation of the effects of these heat transfer processes i'n the OTSG. Since these smaller breaks exhibit large nargin to core uncovery, the CFJsFT2 SG model is adequate for demonstrating compliance to 10 CFR 50.46. - -

In performing small break evaluations, the CRAFT 22 code is used to pre-dict the hydrodynamic response of the primary system including the effcet 90008162

,_; nr -

~ -

.- 2:

, ... . m -

~~~

of'SC heat transfer during the ' transient. The option 2 SG model, which is . explained in' detail.in Section 2.6 of topical report BAW-10092, Rev. 2,

.is utilized to predict heat flow in the SG. The calculation progresses

' basically as follows: +

1. Based upon the initial steady-state heat transfer characteristics of the OTSG and the initial primary and seccndary fluid temperatures, an overall UA for each region of the SG is calculated. 'N

2. The calculated steady-state UA can be modified by user-specified-in-put' options. These include an input multiplier table versus time, multiplied based on the primary side control volume mixture height during the transient, and a multiplier fer reverse heat transfer, i.e.,

heat flow from the' secondary to the primary side of the SG.

3. Using the modified UA and-the calculated primary and secondary side control volume temperatures, the amount of heat transferred is calcu-lated.

In performing the small leak calculations for demonstrating compliance to 10 CFF. 50.46 for the operating B&W plants, no input multiplier versus time is utili cd, nor is the modification based on primary side mixture .

level used. However, a multiplier for reverse heat transfer of 0.1 is utilized. This . multiplier and its basis is explained in the ECCS evalu-ation model topical report and is utilized to reflect the change in heat l

transfer regime on the secondary side of the SG for reverse heat flow.

The OTSG design of the B&W designed operating USSs allows use of a simplis- l l

tic codel for calculation of SG performance during a small LOCA transient. '

With the loss-of-offsite power, assumed in design calculations for small  ;

breaks, and the subsequent loss of main feedwater, the auxiliary feed-water systc= is actuated and will become operable in approximately 40 seconds and control the secondary cide level. The auxiliary feedwater enters the SG very high, approximately 2 feet below the upper SG tube sheet, and is sprayed onto the tube bundics. Thus, heat transfer will occur in the upper portion of the SG independent of the actual' level in the SG. The introduction of auxiliary feedwater to the SG has two ef-fcets on the small LOCA transient. First, it raises the thermal center in the SG during the natural circulation phase of the accident which

. 90008163

. .v. . -

~

. . . '.). .

results in a continuation of circulation through the RCS, for some period of time, even phile ' inventory is lost from the primary system. La'ter in . J the transient, af ter sufficient inventory. has been lost from the system, circulation will be interrupted and the auxiliary feedwater, for a'ecrtain range of small breaks, will condense steam on the primary side of the SG; thereby maintaining the primary system pressure near the secondary . side pressure. The analytical approach utilized for the small break evaluation j is consisten't with' this performance of the auxil'iary feedwater system.

It should be noted that between the time chas circulation through the loops is lost and the time that the primary side SG 1evel'has dropped to i the point where condensation heat transfer will occur, system repressuri-zation can occur as heat removal via the SG will be lost. This phenomena occurs only for the very small sized,small breaks in uhich the SG heat removal is necessary. If simulation of this repressurization phenomena of the very small breaks is desired, an additional node would be needed in the small break model in order to separate the hot leg and SG upper '

plenum volumes from the tube region. This will allow steam to accumulate l

in the upper regions of the RCS without being affected by heat remo al that occurs in the steam generator. In the analyses presc$ted in r2ference 5 for these smaller sized breaks, a model which included the additional node was utilized and showed that the repressurization phenomena does

~

not result in core uncovery.  ;

1 It is also important to note the role of the SG on the small break tran-l sient in order to evaluate the appropriateness of the SG model utilized l

in small break evaluations. Licensing calculations for the operating B&W units have previously been submitted to the Staff in. references 3 and 4.

These evaluations have shown.that the worst case small breaks, i.e., breaks ]

which result in core uncovery, occur for breaks in excess of 0.05. f t2 As

]

demonstrated in the May 7,1979 reports, SG heat removal is not necessary for breaks of this size. For smaller breaks, SG heat removal is necessary as the break alone is not suf ficient to remove enough fluid volume and energy to depressurice the RCS. Houcver, as demonstrated in reference 5, these break's are of no consequence as the SG. heat removal and the slower discharge rate for these breaks easily prevents core uncovery.

90h008164 l

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

. \

' ~

. .- t< .

' . . . . .. , 1 As' demonstrated, the oC model utilized in the' 'small' .deak evaluations. for

~

- ~- -

the operating plants approprist'ely accounts for the eff'ect of the spatial heat removal processes that tfill occur in the OTSG during a small break.

~

It was also 'shown that the SG performance is not important for the worse

- - case small breaks. Thus, the CRAFT 2 SG model is adequate for' demonstrating compliance of the ECCS to 10 CFR 50.46.

. REFERENCES 1

"B&W's'ECCS Evaluation Model," BAW-10104. Rev. 3, Babcock'& Wilcox, August 1977.

- ~

2 R. A. Hedrick, J.J. Cudlin, and R.C. Foltz, " CRAFT 2 - Fortran. Program for Digital Simulation of a Multinode Reactor Plant During Loss-of '

Coolant," BAW-10092, Rev. 2, Babcock & Wilcox, April 1975.

3 Letter, J.H. Taylor (B&W) to S.A. Varga (NRC), July 18, 1979.

4 "Multinode Analysis of Small' Breaks for B&W's 177-Fuel Assembly Nuclear Plants With Raised Loop Arrangement and Internals Vent Valves,"

BAW-10075A, Rev. 1, Babcock & Wilcox, March 1976.

5 " Evaluation of Transient Behavior and Snall Reactor Coolant Systen Breaks in the 177-Fuel Assembly Plant," Babcock & Wilcox, trans=itted via letter from J.H. Taylor to R.J. Mattson, dated >by 7, 1979.

1 90008165 e

, - - , , , - - , - . - , . - - - - - - _, - . . , - - . _ - -- - s., .-,

,' I

, Dockat No. 50-346

~~i -'

' Licenae No. NPF-3 Serial No. 556 CRAFT 2 SIMUIATION OF T11E MARCH 28, 1978' '

- November 13, 1979 THI-2 TRANSICNT

Attachment 4

. I. INTRODUCTION In the May 7,1979 " Blue Book" reports , a CRA:?T2 simulation of the first hour of the TMI-2 transient was presented. That analysis has since been -

modified and updated to include more recent estimates.of the net makeup to the RCS during the event. This report presents the results of the .

latest B&W CRAFT 2 simulation of the TMI-2 event and covers approximately the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 20 minutes of the transient.

The small break ECCS evaluation model, which is described in topical report BAW-10104 and the July 18, 1978 letter report , was used,with some "best estimate" modifications, for the simulation. Actual TMI-2 data were combined with available information about the operator actions to determine estimates of the HPI and ATW injection times and flow rates.

The simulator results (described in detail in the "Results" section) show all the trends and very good comparisons to the actual plant data of system pressure, temperature and pressurizer icvel. The analysis also preucts the time for the start of core uncovery which is in reasonable agreement with the NSAC-1 report. Thus, the CRAFT 2 code is shown to benchrark very well versus the THI-2 data and 4.s suitable for the perfor uances of small break evaluations.

~J, METHOD OF ANALYSIS The CRAFT 2 code which is documented in topical report BAW-10092 , was used to simulate t.he TMi-2 reactor coolant system hydrodynamics. The model uses one node for the reactor building, two nodes for the secondary system, and 23 nodes to simulate the reactor coolant system, including four nodes for the pressurizer. A schematic diagram of the model is shown in Figure 1.

The analytical codel used for this simulation is basicall-/ the same as B6W's ECCS evaluation model. However, certain input assumptions uhich dif fer f rom the evaluation,model approach, were made in order to obtain a "best estimate" simulation. These assumptions are described below:

a. The initial core power level used in the model was 102% of 2772. .

However, foll6 wing reactor trip, the fission product decay heat was adjusted to 98% power operation. The decay heat curve utilined 90008166

, 7. _. - _ _,

. ( -

is 100% ', instead of 120% required by Appendix K to 10 CFR 50, of the INS 5.1 decay heat curve. . '.'.

. , b. A loss of the main feeduster pumps, which is the initiating transient, was assumed at time zero. In order' to account .for potential draining of secondary side fluid from~ the secam generator downcomer into the tube region, a main feedwater coast-down of 10 seconds was'utill:ed. *

c. A turbine trip coincident with the -l'oss of main feedwater was assumed. This results in the secam generator pressure being controlled by a combination of the turbine bypass valves, the atmospheric duep valves and the mai: Lteam. safety valves. For the first 90 minutes, the turbine . bypass valves control the secondary side pressure. In the simulation, these valves ucre set at 1025 psig..,

d. The CRAFT 2 input was adjusted to open the pilot-operated relief valve (PORV) at 8 seconds. This opening time had to be input, and the open valve simulated, since CRAFT 2. code'.does no t have codels for the pressure relief systems of the RCS. Prelimi'na ry

~

TMI-2 data was used to determine the PORV opening time. Present TMI-2 scenarios indicate that the PORV actually opened at 3 seconds.

As will be shown in the results section, if the CRAFT 2 code had an e>711 cit PORV model, it would have predicted the opening at 3 seconds.

e. The reactor scram was chosen to occur at 10 seconds based on preliminary TMI-2 data. Since the CRAFT 2 code does not have provisions for a reactor trip on high pressure, this had to be simulated based on time.

f. The leak area utilized for the PORV is 1.05 in. and represents the orifice area of the valve. The Moody critical discharge l correlation was utili cd to predict the fluid lost through the PORV.

For the first 4h minutes of the simulation, a discharge coef ficient (CD ) f 0.8 was used. For the remainder of the evaluation a C D 1.0 was employed.

90008167

I

g. Actuation of' the' High Pressdre .Injec' tion System (HPI) 'was ba::c'd on ESFAS ' signal of 1615 psia. This resulted in the actuation of the 2 HPI systems'at 1 minute and 45'secends into the transient, as opposed to 2 minutes and 2 seconds which was the make-up flow

, . initiation time at TMI-2. Between 275 and 6100' seconds, the HPI flow was assumed to be throttled by the operator to an average flow of = 34 gpn. This value is based on preliminary assessment on the net makeup flow to the RCS. No explicit modeling of -

letdown was used, only net f. low was simulated.' Af-ar_6100 seconds, an average net makeup (HPI) of 42 gpm was utilized.

h. A four-node pressuri:cemodel was used in the evaluation in order to

' reduce instantaneous artificial condensation in the pressurizer.

1 This phenomenon, which occurs when' the subcooled reactor coolant i 1

fluid mixes with two-phase' pressurizer fluid, results from the equilibrium model limitations of the code. This model is necessary l only to predict the response of the RCS during the initial phase of the loss of' main feeduater event. Also, the pressuriser

, surge line resistance was updated to reflect more realistically the TMI-2 surge line.

i. Steam Generator Modeling - The steam generator model was modified to account for the following phenomena:

1. The overall heat transfer coefficient (between primary and secondary) was assumed to ramp to zero in one minute to account for the delayed auxiliary f eedwater inj ection.

2. Full heat transfer coefficient was reinstated at 500 seconds i

to account for the auxiliary feedwater injection af ter 8 minutes.  ;

3. Auxiliary feedwater was initiated at 500 seconds with half of l

the design AFU flow capacity and with the SG level controlled to 3 feet. With the reactor coolant pumps on, AFW is controlled by the ICS to 3 feet.

4. Steam Generator B was assumed to be isolated at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 41 minutes based upon preliminary TMI-2 data. This was simulated by setting the heat transfer coef.ficient across the 3 steam ,

generator to :cro.

90008168

5.

. The auxiliary feedwater control. level was manually raised to 50% 'on the operating range at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes into

.the transient due to the loss of the RC pumps. ~

.6. Themainsteamsafetyvalveswehemodeledtoopen.at 5400' seconds, and the feedwater flow was increased at'6100 seconds. This was done to simulate the steam generator A dep.cessurization following the operator's attempt to increa'se feedwater flow to steam generator A at about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 34 minutes.

j. ' The RC pumps in the B loop were tripped at 4400 seconds; the A loop RC pumps were tripped at 6060 seconds. These values are consistent with the TMI-2 data.

Table 1 provides a comparison of the assumed times for various system actuations and operator actions to the NSAC scenario. As shown, the values utilized are reasonable compared to the actual performance during the TMI-2 transient.

. . . . 90008.169 a

4

,e . .

3. ,. RESULTS ,

j ,

. 3.1 System Pressure Figures 2 and 3 compare the reactor' coolant system pressure calculated by CRAFT 2

~

to.the TMI-2 data. Following the loss of main feedwater, the pressure in the *

. 'RCS rose sha'rply due to the decreased heat removal across the SG. As shown by. .

I Figure 2, the CRAFT 2 prediction overpredicts the pressure during this phase of the accident due to the delayed opening of the PORV 3 seconds in the transient '

versus 8 seconds for the CRAFT 2 simulation, and the' delayed reactor' trip, 8 seconds

' for the transient versus 10 seconds for the simulation. If the CRAFT code had I

an explicit model for the PORV, opening of the valve would have beed consistent with the data and a better comparison would have- been obtained. After the re-actor tripped, the RC pressure decreased. The calculated pressure drops below the actual data af ter 20 seconds. This is apparently caused by the 10 second ATin feedwa ter coastdown e= ployed in the simulation overpredicting the drainage of secondary downcocer fluid to the SG. Af ter the SG dries out, approximately one minute, the difference between the prediction and the data decreases.

Approximately 5 minutes into the transient, the fluid in the hot leg flashed

~

due to the depressurization of the RCS and'the system pressure increased. As  !

indicated on Figure 3, the CRAFT 2 code properly predicts the system repressuri-ation tine, but overpredicts the actual pressure. The overprediction of sys-tem pressure is probably eTused by the assumed net makeup to the RCS during this time period. Although the EPI was throttled to a net makeup of 34 gpa during this time interval in the simulation, between 4:58 and 6:58, the llSAC r scenario of events indicate.that the letdown flow was'in excess of 160 gpm.

Thus, it is quite probable that there was a decrease in inventory in the RCS due'to the high letdown over this time period.

At 8 minutes and 18 seconds, auxiliary feedwater flow was readmitted to' the SG ~

and primary system pressure decreased (Figure 3) to approximately 1100 psig and was maintained at that value up to approximately one hour and 20 minutes.

4 As shown by Figure 3, the CRAFT 2 prediction is greater over this period by about 100 psi. The coolant pressure was measured in the hot Icg during the accident; the predicted system pressure shown is the core pressure. The ac-tual predicted hot Acg pressure is about 60 psi Iower than the predicted core '

pressure. Also, the pressure in the secondary side was held in the CRAFT 2

~

90'008170

. f

~

simulation at 1025 psig, while the measured value was 1000 psig, resulting in an additional deviation. Thus, the CRAFT 2 prediction reasonably follotts the transicnt behavior'over.this period when the deviations are considered. It d

, , should also be noted that the pri=ary system pressure during this phase of the transient i:5 basically controlled by the SG. The CRAFT 2 prediction did not demonstrate fluctuations in system pressure 'during this period as the sec .

ondary pressure of the SG is assumed to be regulated at 1025 psig. The plant data shows that the secondary side SG pressure wat not held constant over this period,'but fluctuated.

At one hour and 34 minutes, the RCS pressure drcpped due to an apparent attempt by the operator to increase feedwater to the A SG. The analysis attempted to simulate the depressurization effect of the increased auxiliary feedwat er flow by opening the relief valves at 5400 seconds and increasing the auxiliary feed-water flow at 6100 seconds. This modeling technique was utilized as little infor=ation is available on the actual auxiliary feedwater flow delivered to the SC during this period. As shotm by Figure 3, this resulted in an underpredic-tion of the primary system pressure until 7500 seconds and an overprediction for the remainder of the transient analyzed.

3.2 Pressurizer Level A comparison of the CRAFT 2 predicted pressurizer level to the TMI-2 data is provided in Figure 4. As shotm, there a ri:er level predictions given in the figure. The first, entitli . e luel - CRAFT, is the calcu-lated mixture level within the pressurizer. The'second, entitled instrumenta-tion reading - CRAFT, is the calculated liquid level that would be "seen" with-in the pressuri:cr level tops and is directly cc: parable to the TNI-2 data.

The initial pressurizer response and comparison to the loss of main feedwater event (first 4 minutes of the transient) is not easily discernable in Figure

4. It was, however, discussed in the May 7, 1979, report. During this phase of the accident, the pressurizer level responded in a similar manner as the systen pressure. Also, the comparison of the predicted to the actual response of the pressurizer level is similar. That is, the rise in pressuri:er level during the first 10 seconds is overpredicted and the pressuri:cr level after reactor trip is undcrpredicted. The reasons for this are the same an those discussed previously in section 3.1.

90008171 i ,

1 l

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= - - - - - - _ _ _ _ - -~ - . - - __.- -. - -

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' . . .The significant aspect of his comparison is the predicteu mixture level re-

~~

. sponse to,'the predicted' instrument reading response during the transient. As

- (shown by Figure 4, the predicted instrument response and the measured response are in good agreement throughout the simulation. However, as shown by the fig .

j -

ure, although the instrument reading is on scale for portions of the first 101 minutes of the transient, the actual predicted mixture level af ter 6 minutes is j l at the top of the pr2ssurizer. Thus, a two-phase mixture exited through the

, valve during this entire period. Af ter 101 minutes, only secam was entering the j pressuriser .through the surge line (note that the RC pumps have been tripped),

and the pressurizer inixture had reached a sufficient void fraction to allow for l phase separation at the top of the mixture and only steam started to flow out.

3.3 System Flow l Figure 5 shous a comparison of the predicted and transient loop flows. 'As shown, the predicted flow rates do not match well with the actual data. This

. disagreement is caused by two factors. First, loop flow was measured by Cent 111is tubes, which are calibrated based on single phase flow. Their actual performance during two-phase flow is unknown. Secondly, performance of the RC l pu~ps with two-phase flow is not well understood. In performing the evaluation,

a. two-phase pump degradation multiplier based on the semiscale pump tests was utiliaed. This multiplier results in a sharp decrease 15 pump head once any I significant voicing is calculated at the pump inlet. As shown, at 55 minutes, the ' loop flow sharply decreased due to this effect. Although the agreement is j not excellent, the pump flow calculation does not appear to have significantly affected the si=ulation.

3.4 Hot and Cold Leg Temneratures Figures 6 and 7 show a co=parison of the predicted versus actual response of the hot and cold leg temperature measurements during the transient. After 5 minutes and up to the time the core started to uncover, the RCS was saturated,.and the l fluid temperature comparison has'the same deviations previously discussed in scetion 1

l 3.1. ,

After the core starts to uncover, which occurs at approximately 110 minutes, the hot leg temperature measurement indicated superheated steam (Figure 6). How-ever, the CRAFT 2 prediction does not exhibit this behavior. This is due to

, the one-node representation of the core and the equilibrium assumption of the 1 .

90008172

s .

~{.

. n -

.' CRAFT code. As lonF. as f uid is predicted to remain within the core node, re-gardless of the actual' amount of core uncovery, the one-node representation calculates the exiting steam temperature to be saturated. However, the actual -

physical process results in saturated steam at the top of the core mixture

~

level. This steam superheats as it receives energy from the uncovered portion ~ '

of the fuel pins. A multinode representation of the cor would be necessary to predict the hot leg temperature response during this period.

3.5 System Void Frac' tion The average system void fraction evolution for the pri=ary system, excluding the ,pressuriser, is shown on Figurc 9. Due to the continued loss of RCS in-ventory through the PORV and the inadcquate not makeup to the RCS, the system void fraction increases almost linearly from 10 until 101 minutes into the i -

transient.- At 101 minutes all the RC pumps have been tripped. At this time, the RCS liquid inventory is distributed as follows; the RV is filled to slightly above the top of the core; the loop seal in the B loop is full; the A loop has very little inventory. During the subsequent 30 minutes, the RV inventory is boiled-off and the steam is condensed by the A loop steam generator. Because I

of the loucred loop design, this inventory remains trapped uit tin the A loop pucp suction piping and the steam generator. During this peri >d of time, the core becomes uncovered. Thus, since the process is a redistribution of water within the RCS with the only fluid loss being steam vented through the PORV, the system average void fraction does not change significantly.

3.6 Core Mixture Level The calculated core misture level for the transient is given in Figure 8. As shown, no core uncovery was calculated while the RC pumps were operating.

How-ever, closely following the termination of the RC pump flow, the level in the core decreased. Core uncovery vas calculated to start occurring at 105 minutes into the transient. This compares reasonably ucil with the NSAC prediction of

~

approxicately 103 minutes. Thus, the calculated loss rate through the PORV and the net makeup to the RCS must be in reasonable agreement with the actual

~ behavior during the TMI-2 incident. -

As shown by Figure 8, the core was predicted to totally uncover. However, this resulti occurs due to the insufficient spatial detail in the core region. The

)

simulation assumes that all core heat is removed and deposited in the fluid.

This results in an overprediction of the core boil-off once the corc is'un-covered. A more detailed core model is necessary in order to predict how muc'.

core heat is deposited in the liquid region for subscquent boil-of f of ' tite 90008173

. ) ,

core liquid and how much energy is used to superheat'the steam. Howevei, . sinc e l the simulation was basically made to determine how the core uncovery occurred, the refined core model was deemed unnecessary.

~~.4. CONCLUSIONS As demonstrated, the CRAFT 2 code simulation predic.ts reasonably well the system behavior during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 20 minutes cif the THI-2 transient. In

]

fact, the core uncovery time is predicted within a'few minutes of the inferred core level response given in the NSAC report. Therefore, it is apparent that  ;

tha net makeup to the RCS was very loir (approximat.ely 34 gpm) over this period which resulted in uncovery of the core and subsequent core damage. Also, it is shoun that the CRAFT 2 code is able to predict the system hydrodynamics dur-ing a small LOCA and is suitable for licensing calculations.

90008174 1

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i

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. . REFERENCES I

" Evaluation of Transient Dehavior and Small Reactor Coolant System Breaks in the 177.-Fuel Assembly Plant," Babcock & Wilcox, May 7,1979.

2 "B&W ECCS Evaluation Model," BAW-10104, Rev. 3, Babcock & Wilcox, Auguct 1977.

3 Letter J.H. Taylor (B&W) to S. A. Varga (NRC),' July 18, 1978.

4

" Analysis of Three Mile Island, Unit 2 Accident," NSAC-1, July, 1979.

5 R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, " CRAFT 2 - FORTRAN Program for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant,"

j BAW-10092, Rev. 2, Babcock & Wilcox, April 1975.

4 90008175 4

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9 l . .

9 9 1

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Table 1. ' Comparison of CRAFT 2 Assumption to NSAC Scenario .

, Time, hrs: min: sec

~~

~

Event NSAC CRAFT 2

~

Loss of feedwater flow / turbine trip 0:0 " 0:0 PORV opens 0:03 0:08 Reactor trip 0:08' 0:10 HPIs actuated 2:02 1:45 HPI throttled 4:38 4:35

• Auxiliary feedwater block valves opened 8:18 8:20 Reactor coolant pump 2B stopped 1:13:29 1:13:33 Reactor coolant pump 1B stopped 1:13:42 1:13:33 4 l

Steam generator 3 isolated 1:42:00 1:41:40 SG A level raised to 50% on operate range 1:40:00 1:41:40 Reactor coolant pump 2A stopped 1:40:37' 1:41:00

- Reactor coolant pump 1A stopped 1:40:45 1:41:00 i

i 1

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