ML20245G436

Proposed Tech Specs,Suporting Removal RCS Hot Leg & Cold Leg Resistance Temp Detector Bypass Sys

ML20245G436
Person / Time
Site:	Beaver Valley
Issue date:	04/21/1989
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML19297H613	List: ML20245G404 ML20245G411 ML20245G436 ML20245G505 ML20245G681
References
NUDOCS 8905030147
Download: ML20245G436 (42)

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ATTACHMENT A Revise the Technical Specifications as follows:

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a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 f setpoint. I

b. Above P-6 but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to j increasing THERMAL POWER above 5% of RATED THERMAL  !

POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION may  !

continue.

ACTION 4 - With the number of channels OPERABLE one less than j required by the Minimum Channels OPERABLE requirement l and with the THERMAL POWER level:

a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 setpoint. I

b. Above P-6, operation may continue.

ACTION 5 - With the number of channels OPERABLE one less than 3 required by the Minimum Channels OPERABLE requirement, {

verify compliance with the SHUTDOWN MARGIN requirements j as applicable of Specification 3.1.1.1 or 3.1.1.2, ]

within 1 hour, and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

l ACTION 6 - Not Applicable l I

ACTION 7 - With the number of OPERABLE channels

• one less than I the Total Number of Channels and with the THERMAL POWER J level- I

a. Less than or equal to 5% of RATED THERMAL POWER, place j the inoperable channel in the tripped condition within 1 hour; restore the inoperable channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.  ;

b. Above 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within i hour; operation may continue until performance of.the next required CHANNEL FUNCTIONAL TEST.

ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

• An OPERABLE hot leg channel consists of: 1) three RTDs per hot leg, or 2) two RTDs per hot leg with the failed RTD disconnected and the required bias applied.

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'. ATTACHMENT B Safety Analysis Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 163 '

Description of knendment Request: DCP-698 will incorporate a plant design change during the 7th refueling outage (September 1989) to remove the existing Reactor Coolant System (RCS) hot leg and cold leg Resistance Temperature Detector (RTD) bypass system and replace it j with fast respense thermowell mounted RTD's installed in the reactor l coolant loop piping. The revised design will affect the FSAR and i UFSAR response time design basis for the Overtemperature and '

Overpower delta T and loss of flow reactor trip functions.

WCAP -12058 entitled "RTD Bypass Elimination Licensing Report for l Beaver Valley Unit 1 " (Proprietary) describes the extensive l analyses, evaluation and testing performed to ensure the new design meets all safety, licensing and control requirements necessary for the safe operation of the plant. The proposed technical-specification amendment would support this design change by i incorporating the revised response times into the applicable reactor trip functions. As a result of the calculations described in the WCAP, the following protective function technical specification  ;

requirements must be revised: l ITEM CHANGE

1. Table 2.2-1 Item 7 In Note 3 change the maximum allowable l Overtemperature AT value not to be exceeded from 4 '

percent to 3.3. percent.

2. Table 2.2-1 Item 8 Replace the Note specified for the overpower AT allowable value from Note 3 to Note

4. Add a new Note 4, similar to Note 3 but with a maximum allowable value not to exceed 2.9 percent of the channels maximum trip point.

3. Table 2.2-1 Item 12 Replace the allowable value of > 89%

Loss of Flow of design flow per loop with > 88.9%.

f. Table 3.3-1 Action 7 Add a

• note to identify the use of two hot leg RTDs per loop provided the failed RTD is disconnected and the required bias is applied.

5. Table 3.3-2 Item 7 Replace the response time of 4 seconds Overtcmperature AT with 6 seconds.

6. Table 3.3-2 Item 8 Add the response time of 6 seconds.

Overpower AT l

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The FSAR accident analyses are affected due to the different response time characteristics and instrumentation uncertainties between the fast response thermowell RTD system and the current RTD bypace system. However, as shown in the WCAP Table 2.1-1, the total re"ponse time remains the same.

RTD Fast Response Bypass System (sec) Thermowell RTD System (sec)

RTD Bypass Piping and Thermal Lag 2.0 N/A RTD Response Time 2.0 4.0 Electronics Delay 2.0 2.0 Total Response Time 6.0 6.0 As described in UFSAR Section 7.2.1.3, the hot and cold leg RTD's are inserted into reactor coolant tipass loops. A bypass loop from upstream of the steam generator to downstream of the steam generator is used for the hot leg RTD's and a bypass loop from downstream of the reactor coolant pump to upstream of the pump is used for the cold leg RTD's. The RTD's are located in manifolds within containment and are inserted into the reactor coolant bypass loop flow without thermowells. UFSAR Section 7.2.2.3.2 describes the monitoring and testing used to demonstrate the accuracy of the RTD temperature measurements. UFSAR Figure 4-1 illustrates the existing RTD bypass piping configuration. These and other sections of the UFSAR will be revised as shown in Attachmeat D to address the design modifications as well as describe how the new thermowell mounted RTDs will be used in the control and protection functions of the plant.

The plant design modifications will involve replacing the current RTD bypass system with RTD's installed in each reactor coolant loop to obtain the individual loop temperature signals for input to the reactor control and protection system. The design change will eliminate the bypass line components which have been a major source of plant outages as well as occupational radiation exposure. Three fast resoonse, narrow range, single element thermowell mounted RTD's will be installed on each hot leg. The thermowells will be located within the three existing RTD bypass manifold scoops and measure the hot leg temperature to calculate the reactor coolant loop differential temperature and average temperature. The hot leg connection for the Reactor Vessel Level Instrumentation System (RVLIS) will be mounted in a new boss mounted at the same elevation as the existing connections on the same two hot legs. One fast response, narrow range, dual-element RTD will be located in each cold leg to replace the cold leg RTD located in the bypass manifold. One element of the RTD will be considered active and the other element will be a spare. The RTD bypass manifold return line will be capped j at the nozzle on the crossover leg. The new thermowell mounted RTD's will be used for both control and protection. The average  ;

temperature and differential temperature signals used in the control-grade logic will be input into a median signal selector which will select the signal which is in between the highest and lowest values of the three inputs to avoid any adverse plant response as a result of a single signal failure. The 7100 process electronics that require modification will be qualified to the same level as the i existing 7100 electronics.

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The RTD manufacturer, WEED Instruments, Inc. will perform time response testing of each RTD and thermowell prior to installation.

Response time testing of the RTD's will also be performed in-situ to demonstrate that the RTD's can satisfy the response time requirement as installed in the plant. Westinghouse has evaluated data taken from other operating plants and determined the appropriate temperature error to account for the effects of temperature streaming and incorporated this error into the safety analysis and calorimetric flow calculations. The spare cold icg RTD element provides sufficient spare capacity to accommodate a single cold leg RTD failure per loop. Provisions in the RTD electronics allow for operation with only two hot leg RTD's in service. Failure of a hot leg RTD would require manual action to defeat the failed signal and rescale the electronics to average the remaining two hot leg signals. The procedure for using the actual plant bias data is provided in Appendix A of the WCAP.

The new thermowell mounted RTD's have a response time equal to, or better than, the current bypass piping RTD. This will allow the total RCS temperature measurement response time to remain unchanged at 6.0 seconds. As stated in the WCAP, the differences in response time characteristics and instrumentation uncertainties associated witn the fast response thermowell RTD system have been analyzed for ,

those UFSAR accident analyses that may be affected by the proposed design change. It was concluded that the applicable UFSAR non-LOCA safety analyses conclusions and acceptance criteria continue to be

.ac t and it was determined that the UFSAR LOCA analyses were unaffected and did not require reanalysis. The WCAP provides a discussion of the effects on the plant instrumentation and control  ;

functions and concludes that compliance with IEEE 279-1971, 1 applicable general design criteria and industry standards and regulatory guides will not be changed. The mechanical effects are also discussed and it is concluded that the integrity of the reactor coolant piping as a pressure boundary component is maintained by adhering to che applicable ASME Code sections and the pressure retaining capability and fracture prevention characteristics of the piping will not be compromised by these modifications.

The method for using fast-response RTD's installed in the reactor coolant loop piping as a means for RC3 temperature indication has undergono extensive analyses, evaluation and testing as described in WCAP-12058. Incorporating this system into the plant design meets all safety, licensing and control requirements necessary for the safe operation of the plant. The analytical evaluation has been supplemented with in-plant and laboratory testing to further verify system performance. The fast response RTD's to be installed in the reactor coolant loop piping will adequately replace the present hot and cold leg temperature measurement system and enhance ALARA efforts as well as improve plant reliability. Other nuclear plants similar to Beaver Valley Unit 1 have replaced the RTD bypass system with the fast response thermowell RTD system, including H. B. Robinson and Salem Units 1 and 2. Therefore, based on the above, the proposed changes have been determined to be safe since the design meets all safety, licensing and control requirements necessary for the safe operation of the plant.

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ATTACHMENT C No Significant Hazard Evaluation Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 163 Basis for proposed no significant hazards consideration determination: The Commission has provided standards for determining whether a significant hazards consideration exists in accordance with 10 CFR 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or cifferent kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

The proposed changes do not involve a significant hazard consideration because:

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1. A plant design change is scheduled for the 7th refueling outage to remove the existing Reactor Coolant System (RCS) hot leg and cold leg Resistance Temperature Detector (RTD) bypass system and replace it with fast response thermowell mounted RTD's installed in the reactor coolant loop piping. The revised design will affect the FSAR and UFSAR response time design basis for the Overtemperature and Overpower delta T and loss of flow reactor trip functions. The transients which assume protection from  !

these functions are:

l (1) Loss of Electrical Load / Turbine Trip (2) Uncontrolled RCCA Bank Withdrawal at Power (3) Accidental Depressurization of the RCS (4) Partial Loss of Forced Reactor Coolant Flow (5) Steamline Break for EQ Outside Containment.

Westinghouse has prepared WCAP-12058 entitled "RTD Bypass l Elimination Licensing Report for Beaver Valley Unit 1" '

(Proprietary) which describes the extensive analyses, evaluation and testing performed to ensure the new design meets all safety, l licensing and control requirements necessary for the safe operation of the plant. Attachment D provides revisions to applicable UFSAR Sections to reflect the bypass piping elimination and replacement with the fast response thermowell-RTD system.

These UFSAR changes are provided as background information for this technical specification change and will be included in a future UFSAR update. The UFSAR accident analyses are affected due to the different response time characteristics and I instrumentation uncertainties between the fast response thermowell RTD system and the current RTD bypass system, however, the total response time remains the same. The new thermowell mounted RTD's will be used for both control and protection. The average temperature and differential temperature signals used in the control-grade logic

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will- be input into a median signal selector which will select the signal which is in between the highest and lowest values of the three inputs to avoid any adverse plant response as a result of a ,

single signal failure. The 7100 process electronics that require i modification will be qualified to the same level as the existing 7100 electronics.

The differences in response time characteristics and instrumentation uncertainties associated with the fast response thermowell RTD system have been analyzed for those UFSAR accident analyses that may be affected by the proposed design change. The applicable UFSAR non-LOCA safety analyses conclusions and acceptance criteria were determined to be met and the UFSAR LOCA analyses were unaffected by the proposed changes and did not require reanalysis. The effects on the plant instrumentation and control functions were evaluated and found to comply with'IEEE )

279-1971, applicable general design criteria and industry standards and regulatory guides. The mechanical effects were evaluated and it is concluded that the integrity of the reactor coolant piping as a pressure boundary component is maintained by adhering to the applicable ASME Code sections and the pressure retaining capability and fracture prevention characteristics of the piping will not be compromised.

Incorporating this system into the plant design meets all safety, licensing and control requirements necessary for the safe operation of the plant. Therefore, the proposed changes will not introduce any adverse safety considerations or involve a {

significant increase in the probability of occurrence or the consequences of an accident previously evaluated.

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2. The design change will eliminate the bypass line components which have been a major source of plant outages as well as occupational radiation exposure. The new design will continue to perform the functions satisfied by the current system. The proposed change will be performed in a manner consistent with the applicable .!

I standards, preserve the existing design bases, and will not adversely impact the qualification of any plant systems. This will preclude adverse control / protection systems interactions.

The design, installation, and inspection of the new thermowell RTD system will be done in accordance with the applicable ASME Boiler and Pressure Vessel Code criteria. Therefore, the probability for an accident or malfunction of a different type than previously evaluated will not be created.

3. The applicable design bases have been evaluated to ensure the new design will provide the overall reliability, redundancy and diversity assumed available in the plant design for the protection and mitigation of accident and transient conditions.

The integrated operation of the new thermowell RTD's is consistent with the assumptions used in the accident analyses.

The applicable surveillance requirements will continue to ensure the system functional capability is maintained comparable to the original design standards. The response time measurement

'. I' provides assurance that the protective functions satisfy the time limits assumed in the accident analyses. The integrity of the j reactor coolant piping as a pressure boundary component is i maintained by adhering to the applicable ASME Code sections and I the pressure retaining capability and fracture prevention 1 characteristics of the piping will not be compromised. 1 Therefore, the proposed changes will not involve a significant  !

reduction in the margin of safety.

I The method for using fast response RTD's installed in the reactor coolant loop piping as a means for RCS temperature indication has undergone extensive analyses, evaluation and testing as described in WCAp-12058. Incorporating his system into the plant design meets all safetye licensing and control requirements necessary for the safe operation of the plant. The analytical evaluation has been supplemented with in-plant and laboratory testing to further verify system performance. The fast response RTD's to be installed in the reactor coolant loop piping will adequately replace the present hot and cold leg temperature measurement system and enhance ALARA efforts as well as improve plant reliability. Other' nuclear plants similar to Beaver Valley Unit 1 have replaced the RTD bypass system with the fast response thermowell RTD system. Based on the considerations addressed above, the design meets all safety, licensing and control requirements necessary for the safe operation of the plant, therefore, it is proposed that this amendment application does not involve a significant hazards consideration.

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4 ATTACHMENT D UFSAR Changes Beaver Valley Power Station, Unit No. 1 I

Proposed Technical Specification change No. 163 i

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BVPS-1-UPDATED FSAR Rev. 0 (1/82)

During refueling, tests are also conducted to confirm condition of stator windings.

3.2.3.5 Instrumentation Applications 1

Instrumentation for determining reactor coolant average temp- l erature (Tavg) is provided to create demand signals for moving  !

groups of full length rod cluster control assemblies to provide i load follow (determined as a function of turbine impulse l

pressure) during normal operation and to counteract operational transients. The hot and cold leg resistance temperature detectors (RTD's) are described in Section 7. 2 . in the reactor Coolant- I bypacc loopc. The location of the RTD's in each loop is shown on the flow diagrams in Section 4. The Reactor Control System which controls the reactor coolant average temperature by regulation of )

control rod bank position is described in Section 7.3.

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Rod position indication instrumentation is provided to sense the actual position of each full length control rod so that the actual position of the individual rod may be displayed to the operator. Signals are also supplied by this system as input to the rod deviation comparator. The rod poM tion indication system )

is described in Section 7. 1 1

The reactor makeup control system, whose function is to permit >

adjustment of the reactor coolant boron concentration for reacti-( vity control (as well as to maintain the desired operating fluid inventory in the volume control tank), consists of a group of instruments arranged to provide a manually preselected makeup composition that is borated or diluted as required to the charging pump suction header or the volume control tank. This system, as well as other systems including boron sampling pro-visions that are part of the Chemical and Volume Control System, are described in Section 9.1.

When the reactor is critical, the normal indication of reactivity status in the core is the position of the control bank in relation to reactor power (as indicated by the Reactor Coolant System loop T) and coolant average temperature. These parameters are used to calculate insertion limits for the control banks to give warning to the operator of excessive rod insertion.

Monitoring of the neutron flux for various phases of reactor power operation as well as of core loading, shutdown, startup, and refueling is by means of the Nuclear Instrumentation System.

The monitoring functions and readout and indication character-istics for the following means of monitoring reactivity are included in the discussion on safety related display instrumentation in Section 7.5:

1. Nuclear Instrumentation System

( 2. Temperature indicators

a. T average (Measured) i 3.2-67

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BVPS-1-UPDATED FSAR Rev. 0 (1/82)

b. AT (Measured) fled.s art l

c. Auctioncered T average

d. T reference

3. Demand Position of Rod Cluster Control Assembly Group

4. Actual Rod Position Indicator 1

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, I BVPS-1-UPDATED FSAR Rev. 6 (1/88) develop due to rapid changes in fluid temperature during normal operational transients. These points include:

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l. Charging connections from the chemical and volume control system

2. Both ends of the pressurizer surge line

3. Pressurizer spray nozzle.

Therma 1 sleeves are not provided for the remaining injection connections of the ECCS since these connections are not in normal use.

All piping connections from auxiliary systems are made above the horizon' al centerline of the reactor ccolant piping, 'with the exception of:

1. Residual heat removal pump suction, which is 45 degrees down from the horizontal centerline. This enables the-water . level in the RCS to be lower in the reactor coolant pipe while continuing to operate the residual heat removal system should this be required for maintenance.

2. Loop drain lines and the connection for temporary level measurement of water in the. RCS during refueling and maintenance operation

3. The differential pressure taps for flow measurement {

which are downstream of _the steam generators on the 90 degree elbow.

Penetrations into the coolant flow path are limited to the following:

1. The pressurizer spray line connections extend into the cold-leg piping in the form of a scoop so that the velocity head of the reactor coolant loop flow adds to the spray driving force.

2. The reantor coolant sample system taps are inserted into the mair Stream to obtain a representative sample of the reactor Oe,olant.

3. The Resistance Temperature Detector, RTD, hot leg bypacc .

connectienc arc- scoops which extend into the reactor coolant , -t-o--ec' lect a reprccentati/c tc=perat-urc campic yf-cp the RTO mani-f-eld,-

4. The wide range temperature detectors are located in RTD wells that extend into the reactor coolant pipes.

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• BVPS-1-UPDATED FSAR Rsv. 0 (1/82) '

. TABLE 4.1-11

[ REACTOR COOLANT SYSTEM CODE REQUIREMENTS ASME Code Code Component Code / Standard

• Class Addenda Case Reactor Coolant System Reactor Vessel ASME III, 69 A Thru W 68 1332-3 1335-2 1336 1514 Full Length CRDM Housing ASME III, 68 A Thru W 69 NA' Part Length CRDM Housing ASME III, 68 A Thru W 69 1337-3 4 Part Lengt.a CRDM Lock-up (Roller-Nut Type) ASME III, 77 A NA NA Steam Generator (tube side) ASME III, 65 A Thru S 67 NA (shell side) ASME III, 65 A Thru S 67 NA i

)

Reactor Coolant Stop Valves ASME III, 68 A Thru W 68 NA 4 (Body Bonnet) .j PRESSURIZER .

ASME III, 65 A Thru W 66 1401 )

Reactor Coolant Piping, Fit- )

tings and Fabrication ANSI B31.1, 67 NA NA NA )

Surge Pipe, Fittings, and Fabrication ANSI B31.1, 67 'NA NA NA Loop Dypass Line ANSI B31.1, 67 NA NA NA S ; ;cr ."r i hl d frSI S31.3, 57 'I M;.  ;; l NA NA I i

Reagtw _Coolang Thermowells ANSI B31.1, 67 NA Efety ValT/bs ASME III NA Thru S 68 NA Relief Valves ASA 16.5 NA NA NA Valves to Reactor Coolant System Boundary ASA 16.5, NA NA NA l, MSS-SP 66A Pressurizer Relief Tank ASME VIII NA Thru S 68 NA CRDM liead Adapter Plugs ASME III, 68 NA NA NA-Reactor Coolant Pump Standpipe Orifice No Code NA NA NA

• ASA, ANSI: American National Standards Institute (Under USAS B31.1-1967 there is no class as such)

ASME III: American Society of Mechanical Engineers, Boiler and Pressure vessel Code,Section III ASTM: American Society for Testing and Materials TEMA: Tubular Exchangers Manufacturer's Association NA - Not Applicable d t. N .W2. rQ 1 of 2 he fe C/or hac fe,' (,fof $5 m' y J ,sig gg duw.o day Te"j'e/al tut be~lc c)ct i

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BVPS-1-UPDATED FSAR Rev. 3 (1/85) 1 7.2.1.2.2 Blocks of Reactor Trips at Low Power j 1

Interlock P-7 blocks a reactor trip at low power (below l '

approximately 10 percent full power) on a low reactor coolant flow or reactor coolant pump open breaker signal in more than one loop, reactor coolant pump undervoltage, reactor coolant pump underfrequency, pressurizer low pressure, pressurizer high water .

level. See Figure 7.2-1, Sheets 6 and 16, for permissive 1 applications. The low power signal is derived from l three-out-of-four power range neutron flux signals below the l setpoint in coincidence with two-out-of-two turbine impulse chamber pressure signals below the setpoint (low unit load).

The P-8 interlock blocks a reactor trip when the unit is below 30 percent of full power, on a low reactor coolant flow in any one loop. The block action (absence of the P-8 interlock signal) occurs when three-out-of-four neutron flux range signals are below the setpoint. Thus, below the P-8 setpoint, the reactor will be allowed to operate with one inactive loop, and trip will not occur until two loops are indicating low flow. See l Figure 7.2-1, Sheet 4, for derivation of P-8, and Sheet 5 for j applicable logic.

The P-9 interlock blocks a reactor trip on a turbine trip below l 10 percent power. The block action (absence of the P-9 interlock signal) occurs when three out of four neutron flux range signals are below the 10 percent power level set point. Thus below the P-9 set point, the reactor will b'e allowed to operate and ride 1 out the turbine trip transient, with load rejection with reactor power dissipated by steam dump.

See Technical Specification for the list of Reactor Trip System Interlocks.

7.2.1.3 Coolant Temperature Sensor Arrangement T- hot and cold leg resistance temperature dete ors are inse d into reactor coolant bypass loops. A byp loop from

%5bObupstream of the steam generator to downstrea of the steam 4 generator used for the hog leg resi. nce temperature detectors an a bypass loop from downst am of the reactor coolant pump to pstream of the pump i sed for the cold leg resistance tempera ure detectors. resistance Jh temperature detectors are locate 'n manifoldsy ithin the containment and are directly inserted into the reactor coolant bypass loop flow without thermowells. Ther wefls are not used in order to keep the detector thermal lag sg m'a The bypass arrailgement permits replacement of defectiye temperh re elements while the unit is at hot shutdown with'out draining o depressurizing the reactor coolant loops. f/

/

Three sampling probes are installed in a crol sectional plane of each ho / leg at approximately 120 degrees ap . Each of the samplini,t 3 probes, which extends several inches in os the hot leg coolant stream, contains five inlet orifices distributed along 7.2-9

, l 1

INSERT A The hot and cold leg temperature signals required for input to

' the protection and control functions are obtained using thermowell mounted RTDs installed in each reactor coolant loop.

The hot leg-temperature measurement in each loop is accomplished using three fast response narrow range RTDs mounted in thermowells. The hot leg thermowells are located within the j three scoops previously used for the RTD bypass manifold. 'J The scoops were modified during the seventh refueling outage by drilling.a flow hole in the tip of the scoops so water will flow in through the existing holes-in the leading edge of the scoop, pass the RTD and out through the new drilled hole in the tip of the scoop. j The cold leg temperature measurements in.each loop are accomplished by one fast response narrow range. duel element RTD. The existing cold leg RTD bypass penetration nozzle was modified to accept the thermowell and RTD. Temperature streaming in the cold leg is not a concern due to the mixing action of the reactor coolant pump.

Due to temperature streaming, the three fast response hot leg RTDs are electronically averaged to generate the hot leg temperature.

In the event one of the three hot leg RTDs fails, the failed RTD will be disconnected and the hot leg temperature measurement will be obtained by averaging the remaining two RTD measurements q in that loop. A bias adjustment will be applied'to correct.for i the temperature offset. The bias adjustment will be based on the most recent periodic temperature measurement obtained at ,

full power prior to the RTD failure. Subsequent measurements-l obtained from the remaining RTDs in that loop and the other loop RTDs may be used to (1) confirm the correct bias adjustment or (2) define changes required to the bias adjustment. In the event a cold leg RTD fails, the failed-RTD should be disconnected from the logic cabinets and the-installed spare cold leg RTD would then be connected in the failed RTDs place.

Operation with less t.han two hot leg RTDs per loop or with >

both cold leg RTD elements per loop failed is not permissible.

This channel is considered inoperable and should be placed in trip.

The basis for operation utilizing the thermowell mounted RTDs is r eesented in Reference 24.

_______--_-_m- ___-___________.m__ _ _ . _

. . 1

. ~- 1 BVPS-1-UPDATED FSAR Rev. 3 (1/85)

- s length. In this way, a total of fifteen locations in the t' (

ystream are sampled providing a representative cj01 t temperat;ure measurement. The 2 inch diameter pipe leadp g to the resistancgtemperature detectors manifold provides p'11ng of the samples to q ve representative temperature measu nie nt .

Care has been tak n to distribute the flo evenly among the five orifices of each obe by effecti y restricting the flow through the orifices. his has bep done by designing a smaller overall orifice flow area th ( that of the common flow path within the probe. This arr ment has also been applied to the flow transition from th ree ow sources to the pipe leading ,

I to the temperature e ont manif . The total flow area of the three probes has erefore been desi'dned to be less than that of the 2 inch pi connecting the probes' to'the g nifold. ]

s N

The cold leg reactor coolant flow is well mixed by'the reactor cool. arf t pump. Therefore, the cold leg sample is taken 'directly froin a 2 inch pipe tap off the cold leg downstream of the puinp.

7.2.1.4 Pressurizer Water Level Reference Leg Arrangement The design of the pressurizer water level instrumentation  ;

includes a slight modification of the usual tank level arrangement using differential pressure between upper and a lower tap. The modification shown in Figure 7.2-4, consists of the use ( l of a sealed reference leg instead of the conventional open column. \

of water. Refer to Section 7.2.2.3.4 for an analysis of this arrangement.

7.2.1.5 Analog System.

The process analog system is described in Reference 1. ,

7.2.1.6 Digital Logic System The solid state protection logic system takes binary inputs (voltage /no voltage) from the process and nuclear instrument channels corresponding to conditions (normal / abnormal) of unit parameters. The system combines these signals in the required logic combination and generates a trip signal (no voltage) to the undervoltage coils of the reactor trip circuit breakers when the necessary combination of signals occur. The system also provides annunciator, status light and computer input signals which l indicate the - condition of bistable input signals, partial trip and full trip functions and the status of the various blocking, permissive and actuation functions. In ' addition, the system (

includes means for semi-automatic testing of the logic circuits.

A detailed description of this system is given in Reference 3.

7.2.1.7 Isolation Amplifiers and Isolation Devices In certain applications, Westinghouse considers it advantageous )

to employ control signals derived from individual protection channels through isolation amplifiers contained in the protection 7.2-10

BVPS-1-UPDATED FSAR Rev. 3 (1/85) irregularities that would leave the unit in an unsafe condition /

even though some trips were initiated. The references also show (

that the typical protection system racks and cabinets were i tested. The equipment that was installed at BVPS-1 is of the j same type and materials as that which has been seismically tested i l

and qualified. The seismic design and qualifications of the reactor protection system and ESF system imposed by Westinghouse on its suppliers were regulated using the Westinghouse quality assurance program discussed in Appendix A.4.

A summary listing of equipment, applicable seismic considerations and test results for the Stone & Webster supplied ESF systems and emergency power systems is shown in Table 7.2-4. l I

All applicable Stone & Webster equipment specifications included Attachment No. 6 which contained " Seismic Design Requirements" (SDR). The equipment supplier is required to perform a static analysis or a dynamic analysis or a test to demonstrate that the  ;

equipment meets the SDR. The SDR contains guidelines for static analysis, or requires approval by the engineers of calculation techniques used for a dynamic analysis, or approval by the engineers of test procedures. The SDR contains guidelines for the preparation of dynamic analysis or vibration test procedures.

When the supplier completed the necessary work to satisfy the SDR, the data was submitted to the engineers for apprcval.

Main control room boards,within Stone & Webster scope of supply [

were designed so that the gross structural section as well-as l local plate sections, including the effects of mounted equipment, exhibit a minimum natural frequency above the " cutoff frequency"  ;

(i.e.,. rigid range of the amplified response curve) for the I control room. Seismic Class I equipment is qualified, as a I

minimum, to acceleration levels applicable to the installed location on the boards in accordance with the procedures outlined in Section B.2.2. These procedures meet or exceed the requirements of Std. 344-1971. As noted in Section B.2.2, the response.of racks, panels, cabinets and consoles is considered in assessing the capability of instrumentation and electrical equipment. Mounted equipment is tested, as a minimum, to acceleration levels consistent with those transmitted by their supporting structure. A design objective is to minimize amplification of floor accelerations by making members of mounted equipment more rigid.

Resistance temperature detectors used to sense the temperature in

the main coolant loops are rigid, ruggedly built devices designed l to withstand the high temperature, high precrure and flow vibration induced acceleration forces which they are subjected to when installed in the coolant loops. The natural frequency of thi2 device ic decignec to be higher thar che rcquencicc cf the

-f4ew-i-ndueed-v+br-e t ion c so as to minimiz e-eny-empl-i-f-rca tion of

-thcce f ibr-at4 c n c . The- flow induced vibration $ga rc ucP Sigher thar the frequencice acccciated ;;ith tpc cci ..i c disturbance. i The resistance temperature detectors .cre sei ically qualified by enelyric performed by the detector suppl' r/. Thy analycic fesis 7.2-12 y g[ g' q g,p,h //y G b cLar s keso n i kage w e ffecY c,1

&yl*y rda h45f edea,4ce e A tru.

o BVPS-1-UPDATED FSAR Rev. 3 (1/85)

'olcarly indicates resonant frequencies between 300HZ and 500 HZ '

abovea seismic disturbances. A Westinghouse Pressur Water l 7

Reactor Syste ivision analysis, which was com in December 1972, confirmed tha e resonant frequenc between 300 HZ and 500 HZ. In addition to t alysi stinghouse has completed a seismic vibration test, submi e WCAP Report, which confirmed that the resonant frequen were higher uency than the above calculated numbers unctional verification as done at the resistance p6rature detector resonant frequency for a ations l that ee any possible "g" loading that could occur in

_ap ication. .

l The nuclear instrumentation system power range neutron detector has been vibration tested in both the transverse (horizontal) direction and the longitudinal (vertical) direction at acceleration levels greater than those expected during a seismic disturbance at BVPS-1.

Neutron current measurements were made during the tests and current, rasistance and capacitance. checks were made after the tests. No significant changes were seen. There was no mechanical damage to the detector. The nuclear instrumentation racks in the control room were seismically qualified by testing for the maximum acceleration  !

expected at BVPS-1.

Equipment for BVPS-1 is procured on a similar basis co that which was qualified. Any major design change in the equipment would require an evaluation to determine if the changes were of a nature so as not to affect the results of the seismic tests or would require the equipment to be requalified for seismic immunity. The seismic design

( discussed above and the references meets the requirements of GDC 2.

7.2.2 Analysis 7.2.2.1 Evaluation of Design Limits The reactor trip system automatically keeps the reactor parameters operating within a safe, stable region by tripping the reactor when the limits of the region are approached by abnormal transients. The region defined by the trip allows a certain margin before protective action is actually required to constrain the energy releases (See Section 14). This design meets the requirements of GDC 14.

The nuclear power unit reactor trip system design employed by Westinghouse was evaluated in detail with respect to common mode failure and is presented in References 6 and 7. This design meets ,

the requirements of GDC 21.

Preoperational testing was performed on reactor trip system l

components and systems to determine equipment readiness for startup.

This testing serves as a very real evaluation of the system design.

7.2-13 i

I BVPS-1-UPDATED FSAR Rsv. 5 (1/87)

( Also, the control system will respond only to rapid changes in indicated neutron flux; slow changes or drifts are compensated by the temperature control signals. Finally, an overpower signal from any nuclear channel will block automatic rod withdrawal. The setpoint for this rod stop is below the reactor trip setpoint. A negative reactivity insertion in excess of Technical Specifications implies a dropped rod. The reactor will be' tripped to assure automatic rod withdrawal does not cause a DNBR of less than 1.30..

7.2.2.3.2 Coolant Temperature The accuracy of the resistance temperature detector bypacc loop l temperature measurements is demonstrated during unit startup tests by comparing temperature measurement from all -bypass-loop-resistance l temperature detector with one another as well as with the temperature  ;

measurements obtained from the resistance temperature detector  !

located in the hot leg and cold leg piping of each loop. The comparisons are done with the reactor coolant system in an isothermal condition. The linearity of the aT measurements obtained from the hot leg and cold leg-bypaca loop-resistance temperature detectors as l a function of unit power is also checked during unit startup tests. l The absolute value of aT versus unit power is not important as far as reactor protection is concerned. Reactor trip system setpoints are based upon percentages of the indicated aT at nominal full power rather than on absolute values of AT. For this reason, the linearity of the aT signals as a function of power is of importance

( .

rather than the absolute values of the 6T. As part of the unit startup tests, the loop resistance temperature detector signals were ,

compared with the core exit thermocouple signals. ncactor

-temperetame-centrol--has-separete-channels for-proteet-ion-and-cont-rel r

))g$$g ce control is based on the highest average temperature of-the <

1 the control rods are always moved based upon th ( most 25 -->- pess , tic temperature measurement with respect to margins to DNB.

A spur us low average temperature measurement,-from any loop l temperature ontrol channel will cause no contrcl-action. A spurious high average mperature measurement will cause rod insertion (safe direction). ,/

Individual low flow a' arms with 'l'ndividual status lights for each reactor coolant loop byp' ass - flow is provided on the main control board. The alarm and staNut xlights provide the operator with immediate indication of a low flow condition in the bypass loops associated with any- feactor coolant loop'.

Local indicatIrs are provided to monitor t'otal flow through the resistance' temperature detector bypass manifolds fbr~cach loop. The indicators are located. inside the containment but aresaccessible during' power operations. N y 'm s 7.2-27 i____________.________________

INSERT B The input signals to the Reactor Control System are obtained from electronically isolated protection T avg and Delta-T signals, (one per loop).

A Median Signal Selector (MSS) is implemented in the Reactor Control System, one for T avg and one for Delta-T. The MSS receives three signals as input and selects the medium signal for input to the appropriate control systems. Any single failure (high or low) in a calculated temperature will not result in-adverse control system behavior since the failed high or low temperature signal  !

will be rejected by the MSS.

Hence, the' implementation of a MSS in the Reactor Coolant System in conjunction with the two out of three protection logic satisfies the requirements of IEEE 279-1971, Section 4.7, " Control and Protection System Interaction".

a  ;

The response time allocated for measuring RCS hot and cold leg j%ge,nt t=p.cature-using thermowell mounted fast response RTDs is four seconds. This response (

time does not include the process electronics.~

l

{

l 4

O

BVPS-1-UPDATED FSAR Rev. 5 (1/87)

' Flow will be locally monitored: .

Nl. (

Prior - ( restoring temperature channels to normal service following re' opening __of bp as [ loop stop valves whenever a bypass loop has been ut-6_f ervice

2. On a peri i asis 3,3 -F lowing any bypass loop low flow alarm (see above).

Dieg i avi In addition, channel deviation signals in the control ~ system will give an alarm if any temperature channel deviates significantly from the -auet-loneer-ed 'hi est+ value. Automatic rod withdrawal blocks l will also occur if any two of the temperature channels indicate an overtemperature or overpower condition.

7.2.2.3.3 Pressurizer Pressure The pressurizer pressure protection channel signals are used for high and low pressure protection and as inputs to the overtemperature AT trip protection function. This unit uses separate channels for protection and control.

A spurious high pressure signal from one channel can cause decreasing pressure by actuation of either spray or relief valves. Additional redundancy is provided in the low pressurizer pressure reactor trip logic and in the logic for safety injection to ensure low pressure protection.

The pressurizer heaters are incapable of overpressurizing the reactor coolant system. Overpressure protection is based upon the positive surge of the reactor coolant produced as a result of turbine trip under full load, assuming the core continues to produce full power.

The self-actuated safety valves are sized on the basis of steam flow from the pressurizer to accommodate this surge at a setpoint of 2,500 psia and an accumulation of 3 percent. Note that no credit is taken for the relief capability provided by the power-operated relief valves during this surge.

In addition, operation of any one of the power-operated relief valves can maintain pressure below the high pressure trip point for most transients. The rate of pressure rise achievable with heaters is slow and ample time and pressure alarms are available to alert the operator of the need for appropriate action. The power-operated relief valves are designed to operate automatically, assuming no operator intervention for the first 10 minutes of a transient.

7.2.2.3.4 Pressurizer Water Level Three pressurizer water level channels are used for reactor. trip.

Isolated signals from these channels are used for pressurizer water ,

level control. A failure in the level control system could fill or j empty the pressurizer at a slow rate (on the order of half an hour er ,  !'

(

more).

i 7.2-28 l

Rev. 5 (1/87)

BVPS-1-UPDATED FSAR References for Section 7.2 (Cont'd) L.

12. J. Locante and E. G. Igne, " Environmental Testing of Engineered Safety Features Related Equipment (NSSS Standard Scope)",

WCAP-7744, Volume I, Westinghouse Electric Corporation (August 1971). .

13. " Criteria for Protection Systems for Nuclear Power Generating Stations", IEEE Std. 279-1971, The Institute of Electrical and Electronic Engineers, Inc.

14. "IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems", IEEE Std. 338-1971, The Institute of Electrical and Electronic Engineers, Inc.

15. "IEEE Standard Criteria for Class IE Electric Systems for Nuclear Power Generating Stations", IEEE Std. 308-1971, The Institute of Electrical and Electronic Engineers, Inc.

16. "IEEE Trial-Use Standard,- General Guide for Qualifying Class I Electric Equipment for Nmclasr Power Generating Stations", IEEE Std. 323-1971, The Institr:e of Electrical and Electronic Engineers, Inc.

17. "IEEE Trial-Use Guide Type Tests and Continuous-Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations", IEEE Std. 334-1971, The Institute of ( .

Electrical and Electronic Engineers, Inc.

18. "IEEE Trial-Use Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations", IEEE Std. 344-1971, The Institute of Electrical and Electronic Engineers, Inc.

19. " Proposed General Design Criteria for Nuclear Power Plant Construction Permits", Federal Register, (July 11, 1967).

20. J. J. Carey, " Environmental Qualification of Class IE Equipment",

Letter to NRC, Duquesne Light Company (October 15, 1981).

21. " Westinghouse Protection System Noise Tests", Westinghouse Electric Corporation (December 1974) Filed with the NRC for review on the Diablo Canyon Project, docket nunbers 50-275 and 50-323.

l l 22. " Equipment Qualification Data Package -

Process Protection System"; EQDP-ESE-13, Rev. 3, 7/81; Westinghouse Electric Corporation.

23. " Westinghouse 7300 Series Process Control System Noise Tests",

WCAP-8 892-A -dated (June 1977).

( n M rh 6yp s.s u ,:.,,a di os Lich,sig spcd & &ma i Valley UnrY I " WC Af- nors (De cem bu /Wf),

7.2-36

BVPS-1-UPDATED FSAR Rev. 0 (1/82) 6, 7 and 8 of Figure 7.2-1. Tables 7.3-1 and 7.3-2 give

' additional information pertaining to logic and function.

The interlocks associated with the ESF actuation system are outlined in Table 7.3-3. The interlocks satisfy the functional requirements discussed in Section 7.1.2.

The transfer from the safety injection mode to the recirculation mode will automatically take place on two out of four low level signals from the refueling water storage tank (RWST) coincident

~

with a safety injection signal.

7.3.1.1.2 Devices Requiring Actuation The following are the actions which the ESF actuation system I initiates when it is called on to perform its function:

1. Safety injection ,

i

2. Reactor trip 1

3. Feedwater line isolation by closing all main control l valves, feedwater pump trip and closure of main feedwater pump discharge valves

4. Auxiliary feedwater system actuation

5. River water (pump start and system isolation)

6. Containment depressurization system

7. Containment isolation

8. Control room ventilation system isolation and pressurization

9. Emergency diesel startup

10. Main steam line isolation. ,

7.3.1.2 Design Bases: IEEE Std. 279-1971(2)

The unit conditions which require protective action are given in Section 7.3.1.1. The unit variables that are required to be  !

monitored in order to provide protective actions are also I summarized in Section 7.3.1.1.

The only variable sensed by the ESF actuation system which has spatial dependence is reactor coolant temperature. The effect on  ;

the measurement is negated by taking multiple samples from the  !

reactor coolant hot leg and averaging these samples -by-mi-x-ing-4n- l

( -the-resistance-te reture-detectpr4ypan 1 ^^ p . - l elec $r % 'cally h e fNeeu fRTee?wn sy,cie m , l The parameter values that will require protective action are  !

i given in the Technical Specification. l 7.3-3

_ _ -- _ ---- _ _ - _ -- - ~

l

BVPS-1-UPDATED FSAR Rev. 0 (1/82)

(- It also restores the steam generator water level to within i predetermined limits at unit trip conditions by regulating the feedwater flow rate. Steam generator water inventory control is manual or automatic through use of feedwater control valves.

Steam Dump Control The steam dump control permits the nuclear plant to accept a sudden loss of load without incurring reactor trip. Steam is dumped to the condenser as necessary to accommodate excess power generation in the reactor during. turbine load reduction transients.

It also ensures that stored energy and residual heat are removed following a reactor trip to bring the unit to equilibrium no load conditions without actuation of the steam generator safety valves, maintains the unit at.no load conditions, and permits a '

manually controlled cooldown of the unit.

In-Core Instrumentation The in-core instrumentation provides information on the neutron flux distribution and on the core outlet temperatures at selected core locations.

7.7.1.1 Reactor Control System

(

The reactor control system enables the nuclear plant to follow load changes automatically including the acceptance of step load increase or decreases of 10 percent and ramp increases or decreases of 5 percent per minute within the load range of 15 percent to 100 percent without reactor trip, steam dump or pressure relief, subject to possible xenon limitations. The system is also capable of restoring coolant average temperature to within the programmed temperature deadband following a change in load. Manual control rod operation may be performed at any time.

The reactor control system controls the reactor coo] ant average temperature by regulation of control rod bank position. The-reactor coolant loop average temperatures are determined from hot leg and cold leg measurements in each reactor' coolant loop.

There is an average coolant temperature (Tavg) computed for each loop, where:

T,yg , That + Tcold (7,7_y) mebibtn d[ ]bt The error between the programmed reference te erature (based on turbine impulse chamber pressure) and the highcst of the l average measured temperatures (which is then processed through a f lead-lag compensation unit) from each of the reactor coolant

\ loops constitutes the primary control signal as shown in general on Figure 7.7-1 and in more detail on the. functional diagrams 7.7-3

1' BVPS-1-UPDATED FSAR Rsv. 0 (1/82) shown in Figure 7.2-1, sheet 9. The system is capable of restoring coolant average temperature to the programmed value following a change in load. The programmed coolant temperature increases linearly with turbine load from zero power to the full power condition. The T also supplies a signal to pressurizer level control and steaE9 dump control and rod insertion limit monitoring.

The temperature channels needed to derive the temperature input  !

signals for the reactor control system are -phycical-1-y-separated-  ;

-from--th e t-emperatu e 'qbannelc use c derive ap -opr-ie te-

-pr9teet4on-signalsv- Med 4 Mb gmhc M s t ideAT L 7/M a med6, $

ud sT syds used m ia eche/ g tak li an ;+ysfem )

An additional control input signal id derived from the reacto power versus turbine load mismatch signal. This additional control inpu .si improves system by enchancing s aa/ gnal se'/edfd Wlilck se hesf ad'r lunt valas,shka/ cf.s 'uperformance response.

ah . 7%s a wus 7.7.1.2 Fu TLength'R6d Control system ulay/s vse, oc / f ad acfc,uedW ca ase[

dva s m /e she/QAfv l The full length rod control system is descrbed' in Ifeference 3.

The full length rod control system receives rod speed and direction signals from the T control system. The rod speed demand signal varies over thye 9 corresponding range of 3.75 to 45 inches per minute (6 to 72 steps / minute) depending on the magnitude of the error signal. The rod direction demand signal is determined by the positive or negative value of the error signal. Manual control is provided to move a control bank in or out at a prescribed fixed speed.

When the turbine load reaches approximately 15 percent of rated load, the operator may select the " AUTOMATIC" mode, and rod motion is then controlled by the reactor control systems. A permissive interlock C-5 (see Table 7.7-2) derived from measurements of turbine impulse chamber pressure prevents automatic control when the turbine load is below 15 percent. In the " AUTOMATIC" mode, the rods are again withdrawn (or inserted) in a predetermined programmed sequence by the automatic programming equipment. The manual and automatic controls are further interlocked with the control interlocks (see Table 7.7-2).

The shutdown banks are always in the fully withdrawn position during normal operation and ate moved to this position at a constant speed by manual control prior to criticality. A reactor trip signal causes them to fall by gravity into the core. There are two shutdown banks.

The control banks are the only rods that can be manipulated under automatic control. Each control bank is divided into two groups to obtain smaller incremental reactivity changes per step. All RCCA in a group are electrically paralleled to move simultaneously. There is individual position indication for each RCCA.

7.7-4

l j BVPS-1-UPDATED FSAR Rev. 0 (1/82) uses parameters for each control rod bank as follows:

Z = A* (AT) au^t -+C l LL Medh + B * (Tavg)a &

where: Z LL

= Maximum permissible insertion limit for affected control bank 6

l l

an (AT) y~tr-# = 11 est- T of all loops l )

(4edron 1 avg of all loops

= -Righest- T (T,yg) gc l A,B,C = Constants chosen to maintain Z > actual limit based on physics cdciilations The control rod bank demand position (Z) is compared to Zg as follows:

If Z - Zgg i D a low alarm is actuated 1

If Z - Zg i E a low-low alarm is actuated 1&ne e highest values of T and AT are chosen by auctioneering, -

- er_ hMresentation of power is

._used-in-thu imiertion limit ca utat-ion Actuation of the low alarm alerts the operator of .an approach to a reduced shutdown reactivity situation. Administrative procedures require the operator to add boron through the CVCS.

Actuation of the low-low alarm requires the operator to initiate emergency boration procedures. The value for "E" is chosen such that the low-low alarm would normally be actuated before the insertion limit is reached. The value for "D" is chosen to allow the operator to follow normal boration procedures. Figure 7.7-2 shows a block diagram representation of the control rod bank l insertion' monitor. The monitor is shown in more detail in the functional diagrams shown in Figure 7.2-1, sheet 9. In addition to the rod insertion monitor for the control banks, an alarm system is provided to wacn the operator if any shutdown RCCA leaves the fully withdrawn position.

Rod insertion limits are determined by:

1. Determining the allowed rod reactivity insertion at full power consistent with the purposes given above

2. Determining the differential reactivity worth of the control rods when moved in normal sequence

3. Determining the change in reactivity with power level by relating power level to rod position

4. Linearizing the resultant limit curve. All key nuclear parameters in this procedure are measured as part of 7.7-8

.f a

BVPS-1-UPDATED FSAR Rev. 3 (1/85) ]

1 l

I and generated in the reactor following a reactor trip and turbine i trip. An override signal closes the feedwater valves when the average coolant temperature is below a given temperature and the reactor has tripped. Manual control of the feedwater control 1 system is available at all times.

At low power levels between 0 percent and 20 percent, the steam generator water level can also be controlled automatically by

)

c using the feedwater bypass valve which parallels the main feedwater valve. The bypass valve control scheme uses existing steam generator water level, reference water level, and nuclear power signals as inputs. A steam generator level error signal is generated and fed into a proportional-plus-integral controller. ,

The controller's output is added to a fraction of the nuclear l power signal; and the resulting signal modulates the position of the bypass feedwater valve to maintain the desired steam generator water level. j l

Additional monitoring of steam generator water level is possible l by using the Backup Indicating Panel (BIP) in conjunction with ]

local pressure indicators on main steam instrument lines. Refer j to Section 7.8.3 for further description of the BIP and its j functions. ,

l A block diagram of the steam generator water level control system j is shown in Figure 7.7-6. 1 1

i 7.7.1.8 Steam Dump Control ]

The steam dump system is designed to accept 85 percent of full load steam flow at full load steam pressure.

The automatic steam dump system is able to accommodate this i abnormal load rejection and to reduce the effects of the transient imposed upon the RCS. By bypassing main steam directly  ;

to the condenser, an artificial load is thereby maintained on the  ;

primary system. The rod control system can then reduce the reactor temperature to a new equilibrium value without causing overtemperature and/or overpressure conditions.

If the difference between the reference T (T based on gem turbine impulse chamber pressure and the SUd/la#[f) c compensated

--+ -au c t i e n c c r e d-- T exceeds a predetermined amount, and the I interlock men ti^o"n%d below is satisfied, a demand signal will actuate the steam dump to maintain the RCS temperature within control range until a new equilibrium condition is reached.

To prevent actuation of steam dump on small load perturbations, an independent load r' ejection sensing circuit is provided. This circuit senses the rate of decrease in the turbine load as detected by the turbine impulse chamber pressure. It is provided 7.7-12

a BVPS-1-UPDATED FSAR Rev. 3 (1/85)

( to unblock the dump valves when the rate of load rejection exceeds a preset value corresponding . to a 10 percent step load i decrease or a sustained ramp load decrease of 5 percent / minute. 1 l

A block diagram of the. steam dump control system is shown on

-Figure 7.7-7.

~ '

7.7.1.8.1 Load Rejection Steam Dump Controller This circuit prevents large increase in . reactor coolant temperature following a large, sudden load decrease. The error I g/m signal is a difference between the lead / lag compensated l w nctieneered- T and the reference T is based on turbine l -. 1 impulse chambe/#lessure.

p "V9 ,

The T signal is the same as that used in the RCS. The lead / V9 lag compensation for the T signal is to compensate for ,

lags in the unit thermal reap 8Me and in valve , positioning. I Following a sudden load decrease, T is immediately decreased and T tends to increase, thus ge@ bating an immediate demand signaf"for steam dump. Since control rods are available in this situation, steam. dump terminates as the error comes' within the maneuvering capability of the control rods.

7.7.1.8.2 Turbine Trip Steam Dump Controller ,

3 f Following a turbine trip, as monitored by the ' turbine trip signal, the load rejection steam dump controller is defeated and 1 the turbine trip steam dump controller becomes active. *Since j control rods are not available in this situation, the demand med,;, signal is the error signal between the lead / lag compensated

% -=ctiencored T a When the error l signal exceed P %predetermined nd the no loadsetpoint reference TtAedump9 valves are j tripped open in a prescribed sequence. As . the error - signal  ;

reduces in magnitude indicating that the reactor coolant system 1 T is being reduced toward the reference no-load value, the dOEnh valves are modulated by the, unit trip controller to regulate the rate of removal of decay heat and thus gradually establish .;

the equilibrium hot shutdown condition. 1 The error signal determines whether a group of valves is to be tripped open or modulated open. In either case, .they are '

modulated when the error is below'the trip-open setpoints.-

7.7.1.8.3 Steam Header Pressure Controller l

Residual heat removal is maintained by the steam generator '

pressure controller (manually selected) which controls the amount of steam flow to the condensers. This controller operates -a portion of the same steam dump valves to the condensers which'are 1 used during the initial transient following turbine / reactor trip i on load rejection.

7.7-13 1

BVPS-1-UPDATED FSAR Rev. 3 (1/85)

The pressurizer pressure channels needed to derive the control signals are physically isolated from the pressure channels used to g~) derive protection signals.

\

j I

/

Channels of the nuclear instrumentation that are used in the {

protective system are combined to provide non-pretective functions '

such as signals to indicating or recording dev.:es, the requir signals are derived through isolation amplifiers. e ff C l

7.7.2.2 Response Consideration of Reactivity I i

Reactor shutdown with control rods is completely independent of the control functions since the trip breakers interrupt power to the full length rod drive mechanisms regardless of exinting control signals.

The design is such that the system can withstand accidental .

1 withdrawal of control grdups or unplanned dilution of soluble boron without exceeding acceptable fuel design limits. Thus, the design meets the applicable requirements of 1967 GDC 31. No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single.RCCA from the partially inserted bank at full power operation. The operator could deliberately withdraw =. single RCCA in the control bank, this feature is necessary in order to retrieve a rod, should one be accidentally dropped. In the extremely unlikely event of simultaneous electrical failures which could result in single withdrawal, rod deviation would be displayed or. the unit annunciator, and the rod position indicators would indicate the relative positions of the rods in the bank.

,_s Withdrawal of a single RCCA by operator action, whether deliberate or a combination of errors, would result in activation of the same

(' -) by alarm and the same visual indications. .

Each bank of control and shutdown rods in the system is divided into two groups of 4 mechanisms each. The rods comprising a group operate in parallel through multiplexing thyristors. The two groups in a i bank move sequentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation or deactuation of the stationary gripper, movable gripper and lift coils of a mechanism is required to withdraw the RCCA attached to the mechanism. Since the four stationary gripper, movable gripper and lift coils associated with the RCCA's of a rod group are driven in parallel, any single failure which could cause rod withdrawal would affect a minimum of one group of RCCA.

Mechanical failures are in the direction of insertion or immobility.

The identified multiple failure involving the least number of components consists of open circuit failure of the proper two out of sixteen wires connected to the gate of the lift coil thyristors. The probability of open wire (or terminal failure) is 0.016 x 10-5 per hour by MIL-HDB217A. These wire failures would have to be accompanied by failure, or disregard, of the indications mentioned above. The probability of this occurrence is therefore too low to  ;

have any significance. j I

(~)

v 7.7-18

. .; /

s INSERT C The loop Tavg and Delta T channel required inputs to the steam dump system, reactor control system, the control rod insertion monf!.:)r- and the pressurizer level control system are electrically isolated prior to being routed to the control cabinets. A median signal is then calculated for Tavg and Delta-T in e the control cabinets utilizing a Median Signal Selhiivn--(MSS) for input to the l appropriate control systems. 6e/ec10/

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Notes: 1. Temperatures are measured at steam generator's inlet and reactor coolant pump outlet.

2. Pressure is measured at the pressurizer.

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FIGURE 7.7-1 SIMPLIFIED BLOCK DIA6 RAM OF REACTOR CONTROL SYSTEM BE AVER. VALLEY POWER STATION vNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT

V E

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R

- A R L U A 2 O L F W A O L L LS

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_ M L R NO A OR L L MT A A MN OO W C O

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l PRESSURIZER .

LEVEL SIGNAL LEVEL PROGRAMMER I

(+) I (-) ,

I

( REMOTE MANUAL CONTROL ir y PI C.ONTROLLER 1r TO BACKUP HEATER CONTROL 1r ,

CHARGING FLOW CONTROL VALVE POSITION AND/0R CHARGING PUMP SPEED l

( FIGURE 7 7-5 BLOCK DIAGRAN OF .

PRESSURIZER LEVEL C0liTROL SYSTEM  !

BE AVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPGRT

• p REV. 0 (1/82)

(' STEAM DUMP CONTROL IN MANUAL fitT)IAsl -

-AUCT!M EE"E0 (STEAM PRESSURE CONTROL) TAyg iAVG REFERENCE TURBINE IMPULSE NO-LOAD T gyg STAGE PRESSURE if l

If RATE / LAG COMPENSATION LFAD/ LAG TURBINE COMPEi!SATI ON TRIP LOAD REJECTION BI STABLE

] ( -) E (+) (+) E (-)

DEFEAT LOAD REJECTION STEAM DUMP CONTROL: l ALLOW PLANT TRIP STEAM DUMP CONTROL BISTABLES BISTABLES STEAM HEADER PRESSURE U

SET PLANT TRIP PRESSURE CONTROLLER V

LOAD REJECTION E (-) CONTROLL ER y y i

i y -> LOAD REJECTION

! CONTROL OR PLANT PI CONTROLLER TRIP CONTROL V U if LOAD REJECTION CONTROL OR PLANT U TRIP CONTROL TRIP OPEN STEAM DUMP VALVES HIGH CONDENSER BACK PRESSURE

- , AUTO (TAVG OR ALL CIRCULATING WATER MANUAL CONTROL) PUMP BREAKERS OPEN (STEAM PRESSURE O AIR SUPrLY TO CONTROL)

JUMP $lALVES y C0 DEN P AV$

FIGURE 7 7-7 BLOCK DIAGRAM OF STEAM DUMP CollTROL SYSTEM BE AVER VALLEY POWER STATION UNIT N0. 1 UPDATED FINAL SAFETY ANALYS!! REPORT

-t* .

BVPS-1-UPDATED FSAR Rev. 6 (1/88)

/ TABLE 14D-3 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED'IN ACCIDENT ANALYSE!i Limiting Trip Trip Point Assumed Time Delay Function In Analyses (seconds)

Power Range High Neutron 118% 0.5 Flux, High Setting Power Range High Neutron 35% 0.5 Elux, Low Setting Overtemperature AT Variable l6.0(1)

(See Technical Specifications)

Overpower AT Variable 6.0(1)

(See Technical Specifications)

High pressurizer pressure 2,410 psig(2) 2.0 1 1

Low pressurizer pressure 1,845 psig(3) 2.0  !

[

Low reactor coolant flow 87% loop flow 1.0 (from loop flow detectors)

.Undervoltage Trip 68% nominal (4) 1.2 l Turbine Trip -

Not applicable 1.0 i Low-Low steam generator 0% of narrow range 2.0 level level span I

High steam generator 75%ofnarygyrange 2.0 level trip of the level spant i feedwater pumps and closure of feedwater I system valves, and i turbine trip i (1) Total time delay from the time the temperature difference in the l coolant loops exceeds the trip setpoint until the' rods are free l to fall. Tuc cccendc Orc cllowed for-RTD bypcss-loeP--f4uld- I l - transport-delay-and--thermal-capacity effects. A maximum of 4 i scccnds is alicwcd for-the sum of the sensor response time, the channcl the reacter trip b- - delay, and the , ripper

(

time,

-r-e1cacc timc . feu secmds is a//scahlcaksNe

% AfD M5fms e bne a d l

}w seeeds os alleca h/ k -g e cu, & fk eks n ,a / delays, -A < tea ch -

h,y breaker delsy , a~l % pQps telease h*-c, 1 of 2 i

_ _ _ _ - - _ _ _ -