ML20212H414

Rev 1 to Beaver Valley Unit 2 Failure of Feedwater Control Channel Used for Protection

ML20212H414
Person / Time
Site:	Beaver Valley
Issue date:	11/30/1986
From:	Butler J, Osborne M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20212H397	List: ML20212H393 ML20212H414
References
NUDOCS 8701210393
Download: ML20212H414 (36)

Text

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l BEAVER VALLEY UNIT 2 FAILURE OF FEEDWATER CONTROL l CHANNEL USED FOR PROTECTION Revision 1 J. C. Butler l

November 1986 Approved: LLLLENht M. P. Osborne, Manager Transient Analysis !!

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Westinghouse Electric Corporation l Nuclear Energy Systems P.O. Box 355  ;

l Pittsburgh, Pennsylvania 15230 0701210393 070115 PDR ADUCK 05000412 l A PDR ses00.10/030644 <

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T l' Rev. 1

• TABLE OF CONTENTS l Section Title i 1.0 Purpose of Analysis l

2.0

Background

3.0 Description of the Event 4.0 Transient Results I 5.0 Summary and Conclusions 6.0 References 1

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. LIST OF TABLES Table Title i

1 Initial Conditions l

i 2 Time Sequence of Events for a Feedwater Control l

Malfunction With Reactor Trip ,

3 Time Sequence of Events for a Feedwater Control i Malfunction Without a Reactor Trip l

1 4 Time Sequence of Alarms and Annunciators for a Feedwater Control Malfunction With a Reactor Trip 5 Time Sequence of Alarms and Annunciators for a Feedwater Control Malfunction Without a Reactor Trip t

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Rev. 1

. LIST OF FIGURES Figure Title i

1 Steam Generator 1 Level Logic 2 Steam Generator 1 Initiating Event 3 Steam Generator 1 Case 1 Single Active Failure 1

4 Steam Generator 1 Case 2 Single Active Failure 5 Feedwater Control Malfunction Nuclear Power and Core Heat Flux versus Time No Reactor Trip (Beginning of Core Life) 6 Feedwater Control Malfunctin RCS Average Temperature, Delta T and Pressurizer Presst.re versus Time No Reactor Trip (BeginningofCoreLife) 7 Feedwater Control Malfunctin Steam Generator Secondary Side Volume and DNB Ratio versus Time No Reactor Trip (Beginning of Core Life) 8 Feedwater Control Malfunction Nuclear Power and Core Heat Flux versus Time No Reactor Trip (End of Core Life) 9 Feedwater Control Malfunction RCS Average Temperature, Delta T and Pressurizer Pressure versus Time No Reactor Trip (EndofCoreLife) 10 Feedwater Control Malfunction Steam Generator Secondary Side Volume and DNB Ratio versus Time No Reactor Trip (EndofCoreLife) 56500,10/030684

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.. LIST OF FIGURES (Cont)

Figure Title 11 Feedwater Control Malfunction Nuclear Power and Core Heat Flux versus Time Reactor Trip on Lo-Lo Steam Generator Level (Beginning.of Core Life) 12 Feedwater Control Malfunction RCS Average Temperature Delta T and Pressurizer Pressure versus Time Reactor Trip Lo-Lo Steam Generator Level (Beginning of Core Life) 13 Feedwater Control Malfunction Steam Generator Secondary Side Volume and DNB Ratio versus Time Reactor Trip on Lo-Lo Steam Generator Level (Beginning of Core Life) 14 Feedwater Control Malfunction Nuclear Power and Core Heat Flux versus Time Reactor Trip on Lo-Lo Steam Generator Level (End of Core Life) 15 Feedwater Control Malfunction RCS Average Temperature, Delta T and Pressurizer Pressure versus Time Reactor Trip on Lo-Lo Steam Generator Level (End of Core Life) 16 Feedwater Control Malfunction Steam Generator Secondary Side Volume and DNB Ratio versus Time Reactor Trip on Lo-Lo Steam Generator Level (End of Core Life) 56500.1D/030684

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- 1.0 PURPOSE OF ANALYSIS

.4 The Instrumentation and Control Systems Branch of the United States Nuclear Regulatory Commission has questioned Beaver Valley Power Station Unit 2 in regards to feedwater isolation. More specifically, the issue has been raised of strictly applying single failure criteria to two out of three hi-hi steam generator water level logic for feedwater isolation.

The purpose of this analysis is to describe the expected transient performance of Beaver Valley Unit 2 for several postulated feedwater malfunction scenarios.

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2.0 BACKGROUND

A safety analysis of Feedwater System Malfunction Causing an Increase in Feedwater Flow is presented in the Beaver Valley Unit 2 Final Safety Analysis Report. It demonstrates that the Departure from Nucleate Boiling (DNB) design basis is met for the event.

While the FSAR analysis credits feedwater isolation on a hi-hi steam generator level signal, the DNB ratio (DNBR) reaches its minimum and begins to increase prior to feedwater isolation. Therefore, even without taking credit for the hi-hi level signal, the DNB design basis for the event will be met.

The single random failure requirement of IEEE-279 stipulates that where a random failure can result in a control system action that produces a plant condition requiring protective action, and simultaneously prevents the proper action of a protection chanel designed to protect against that plant condition, the remaining redundant channels shall be capable of protecting the plant even when degraded by a second random failure. As regards the steam generator level signal, if the transmitter in the level channel used for control purposes fails in such a way as to cause high feedwater flow (and increasing level), a subsequent failure in one of the two remaining channels might prevent the actuation of feedwater isolation.

Feedwater isolation is normally actuated by a hi-hi steam generator water level signal in any one of the three steam generators. In each steam generator, hi-hi level signal is based upon receiving the indication in 2 out of 3 channels.

Referring to Figure 1, 2 of the 3 steam generator water level channels in each steam generator have bistables for each of the following three functions:

lo-lo steam generator water level reactor trip, low steam generator water level signal for low feedwater flow reactor trip, and hi-hi steam generator water level turbine trip and feedwater isolation. The third channel replaces low steam generator water level signal with input to the appropriate feedwater control valve. If for some reason the transmitter in this third channel were 56500:10/1120M

. .- R:v. 1 to fail low, the feedwater control valve would begin to open, to keep the steam generator water level near its setpoint. Strictly applying the single active failure criteria (failure in one of the other two SG water level channels), the hi-hi steam generator water level signal could not be generated in that loop. Only one channel is available to indicate water level above hi-hi, but 2 are needed for the logic. Thus this function, hi-hi steam generator water level turbine trip and feedwater isolation which was assumed in the FSAR, is not available.

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3.0 DESCRIPTION

OF EVENT This excessive feedwater flow transient is initiated by a feedwater control failure. It is exacerbated by a subsequent protective system failure. This failure precludes the actuation of feedwater isolation on hi-hi steam generator level, which is assumed in the Final Safety Analysis Report.

Figure 1 displays the logic of the level signals in steam generator 1 (used as anexample). Four functions are provided: lo-lo level reactor trip, low level for low feedwater reactor trip and hi-hi level turbine trip and feedwater isolation, protective functions, and feedwater control, a control function. Each protective function requires two out of three bistables actuated to perform. (Low level must be in coincidence with steam flow / feed flow mismatch, but requires only one out of two channels). A dedicated channel is used in feedwater control. It continuously indicates position, rather than a range.

Figure 2 shows the same logic after the initiating event. The transmitter in channel III falls low. A lo-lo level signal is generated in that channel; hi-hi level is not. The Feedwater Control System tells the valve to open.

Figures 3 and 4 take this one step further - the single active failure is incorporated. Figure 3 assumes that the failure causes another channel to believe its level is also at the bottom. Therefore a second lo-lo signal is generated and the reactor is tripped. This is the first case to be analyzed.

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Figure 4 assumes that the failure restrains channel I from generating any signal. The third channel (II) will operate properly above nominaf-(single failure already assumed). However, no other channels will be able to indicate level above the hi-hi setpoint. Channel I has no signal and Channel III indicates below lo-lo level. Therefore, none of the three protective functions will be actuated. This is the second case.

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R:v. 1 If the failure were to produce a hi-hi level signal in that channel, turbine ,

i trip and feedwater isolation would occur when the level in the third l (unfaulted) channel reaches the hi-hi level setpoint. This is consistent with-the FSAR analysis. l l

The excessive heat removal due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code LOFTRAN (Burnett 1972). -

This code simulates a multi-loop system, the ncutron kinetics, the l pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.

A control system malfunction is assumed to cause a feedwater control valve to open fully. Two cases are analyzed as follows:

1. Opening of one feedwater control valve with the reactor at full power.

Reactor trip is generated by a lo-lo steam generator water level in 2 out of 3 channels. (One channel failir.g low initiates the transient; the second channel failing low is the single active failure.)

2. Accidental opening of one feedwater control valve with the reactor at full power without consideration of reactor trip.

Each of these cases is analyzed for both beginning of life and end of life core conditions.

The following assumptions have been made:

1. One indicated steam generator water level signal used for control is assumed to fail in such a way as to indicate zero level and demand full feedwater flow.

2. Feedwater flow rate is automatically controlled through the Steam Generator Level Control System using indicated steam flow, feedwater flow, steam generator water level and a programmed level setpoint.

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3. Steam flow at its full load value until turbine trip (one second after reactortrip).

4. The Pressurizer Pressure Control System functions normally. j l

5. The Steam Dump Control System functions.

6. No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown.

7. Feedwater isolation on hi-hi steam generator water level signal is defeated.

8. The feedwater flow is isolated after reactor trip by a low T,yg signal in two out of three loops.

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9. Initial operating conditions are assumed at values consistent with steady-state operation. Refer to Table 1.

No other reactor control systems or engineered safety feature (ESF) systems are required to function. The reactor protection system (RPS) will function to trip the reactor due to overpower or over temperature conditions. No single active failure will prevent operation of the RPS.

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R v. 1 4.0 TRANSIENT RESULTS The first case analyzed proceeds in the following manner. The steam generator level transmitter u' sed for level control fails low. This causes the control system to open the feedwater control valve in an attempt to restore level to its programmed valve. Also, the failed transmitter generates a lo-lo level reactor trip signal in that channel.

A subsequent single active failure of a second level channel produces lo-lo and low level signals in one of the other two channels. A reactor trip is generated on a 2 out of 3 coincidence of lo-lo steam generator level (Figure 3).

At this point, reactor trip initiates turbine trip and the Steam Dump Control System is actuated to reduce primary temperature to the no-load valve.

The increasing saturation pressure and decreasing temperature in the steam generator due to reduced heat transfer causes the secondary side steam generator mixture to collapse. This " shrink" results in a reduced mixture volume and level of the steam generator secondary side.

When the average RCS temperature in two out of three loops reaches the low T,yg set point (no load plus 7'F) in coincidence with the P-4 permissive (tripped reactor) all feedwater control valves begin to close. This prevents further addition of main feedwater.

Transient results (Figures 5 through 10) show the nuclear power, core heat flux, average RCS temperature, loop delta-T, pressurizer pressure, steam generator water volume and DNB ratio for this case. The steam generator water level reaches a peak of only 40 percent of the narrow range span which is less than the initial value. Therefore, the steam generator will not overfill.

Table 2 presents a sequence of events for this transient.

The second case is initiated exactly as.the first case is. However, its subsequent single failure is assumed to be a failure of'the transmitter at its 56500:1o/030684 7 m . _ _ _ __ ..

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- previous value. Reactor trip does not occur (Figure 4). The purpose of this case is to determine the amount of time available for the operator to terminate this event prior to overfill.

This transient has a very minor impact upon the plant. The only parameter that significantly changes is steam generator water volume, which steadily increases.

Transient results (Figures 11 through 16) show the nuclear power, core heat flux, average RCS temperature, loep delta-T, pressurizer pressure, steam generator water volume, and DNB ratio. The steam generator water volume exceeds the capacity of the secondary side, 5760 cubic feet, at approximately 400 seconds for the BOL case and 450 seconds for the EOL case.

From Figures 13 and 16, one can see that approximately 6.5 minutes are available for the operator to isolate feedwater before stram generator overfill occurs. Table 5 contains a listing of alarms and annunciators which would actuate as a result of this transient.

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5.0 CONCLUSION

S Section 4 presents best estimate results for two postulated feedwater malfunction scenarios. Each scenario assumes a steam generator level transmitter used for level control fails low resulting in an increase in feedwater flow to the affected steam generator. A subsequent single active failure of a second level channel occurs in the same steam generator and fails in a manner which prevents termination of feedwater flow on steam generator high high water level.

The first scenario assumes the second level channel fails in the low direction producing lo-lo and low level signals. The second scenario assumes the second level channel fails at its previous value producing no level deviation signal.

The results of analyses for both scenarios demonstrate that protection against DNB is provided throughout the event. This result is consistent with the excessive feedwater flow analyses presented in the Beaver Valley Unit 2 FSAR.

The results for the first scenario show that feedwater addition to the affected steam generator will be terminated on coincidence of P-4 and Low RCS Tavg prior to steam generator overfill.

The results for the second scenario, in which no feedwater isolation signal is generated, indicate that the time to steam generator overfill assuming no operator action is greater than 6.5 minutes.

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6.0 REFERENCES

Burnett, T. W. T., et al 1972. LOFTRAN Code Description. WCAP-7907, June, 1972. Also supplementary information in letter from T. M. Anderson, NS-TMA-1802, May 26, 1978 and NS-TMA-1824, June 16, 1978.

Beaver Valley Power Station Unit 2, Final Safety Analysis Report.

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R:v. 1 TABLE 1 INITIAL CONDITIONS i

Core Power, MWt 2660 Thermal Design Flow, GPM 265500 Reactor Coolant Average Temperature, 'F 576.2 Reactor Coolant System Pressure, psia 2250 Steam Generators Secondary Side Volume, ft 3 5760 E00:1D/030684 I1

Rev. 1 TABLE 2 TIME SEQUENCE OF EVENTS FOR A FEE 0 WATER CONTROL MALFUNCTION WITH REACTOR TRIP Accident Event Time (sec) _

1. Beginning of Life Feedwater Control Valve Core Conditions O begins to open, loop 1 Lo-lo SG 1evel reactor trip 0

Minimum DNBR occurs 0

Turbine trip on reactor trip 1 Low T,yg reached, loops I and 3 7

Feedwater control valves 14 fully closed

2. End of Life Core

( Conditions Feedwater Control Valve 0

begins to open, loop 1

{ Lo-lo SG level reactor trip 0

Minimum DNBR occurs 0

l Turbine trip on reactor trip 1 f Low T,yg reached, loops 1 and 3 8

j Feedwater control valves 15

! fully closed

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. TABLE 3 TIME SEQUENCE OF EVENTS FOR A FEEDWATER

' CONTROL MALFUNCTION WITHOUT REACTOR TRIP Accident Event Time (sec)

1. Beginning of Life Feedwater Control Valva 0 Core Conditions begins to open, loop 1 Minimum DNBR occurs 0 Hi-hi SG 1evel reached, loop 1 ~120 Water reaches top of SG, loop 1 400

2. End of Life Core Feedwater Control Valve O Conditions begins to open, loop 1 Minimum DNBR occurs 0 Hi-hi SG level reached, loop 1 ~120 Water reaches top of SG, loop 1 450 e

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. TABLE 4 l TIME SEQUENCE OF ALARMS AND ANNUNCIATORS FOR 1 A FEEDWATER CONTROL MALFUNCTION WITH REACTOR TRIP Accident Event Time (sec)

1. Beginning of Life Bistable 474 A 0 Core Conditions Bistable 476 A 0 Channel 474, lo-lo SG level 0 Channel 476, lo-lo SG level 0 Reactor tripped 0 Low level deviation alarm 0 Steam dump valves open 2 7

Low T,yg interlock Feedwater Control Valves 14 fully closed

2. End of Life Core Bistable 474 A 0 Conditions Bistable 476 A 0 Channel 474, lo-lo SG level 0 Channel 476, lo-lo SG level 0 Reactor tripped 0 Low level deviation alarm 0 Stea:n dump valves open 2 Low T,yg interlock 8 Feedwater Control Valves 15 fully closed 1

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. TABLE 5 TIME SEQUENCE OF ALARMS AND ANNUNCIATORS FOR A FEEDWATER CONTROL MALFUNCTION WITHOUT REACTOR TRIP Accident Event Time (sec)

1. Beginning of Life Bistable 476 A 0 Core Conditions Channel 476, lo-lo SG level 0 Low level deviation alarm 0 Feedwater Control Valve fully open, loop 1 9 Channel 475, hi-hi SG level ~120 Bistable 475C ~120

2. End of Life Core Bistable 476 A 0 Conditions Channel 476, lo-lo SG 1evel 0 Low level deviation alarm 0 Feedwater Control Valve

fully open, loop 1 9 Channel 475, hi-hi SG level ~120 Bistable 475C ~120 56500

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STEAM GENERATOR 1 LEVEL LOGIC

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FIGURE 2 STEAM GENERATOR 1 INITIATING EVENT

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FIGURE 3 STEAM GENERATOR 1 CASE 1 SINGLE ACTIVE FAILURE'

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FIGURE 4 STEAM GENERATOR 1 CASE 2 SINGLE ACTIVE FAILURE

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j Figure 16. Feedwater Control Malfunction Steam Generator Secondary Side Volume and DNB Ration Versus Time No Reactor Trip ,

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