Text

Proposed Tech Specs 1.1/2.1,3.1/2.1 & 3.2/4.2 for Facilities
	ML18033A262
Person / Time
Site:	Browns Ferry
Issue date:	06/24/1988
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18033A261	List: ML18033A260; ML18033A262;
References
	NUDOCS 8807070099
	Download: ML18033A262 (28)
	v • d • e

ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATIONS REVISIONS BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFN TS 241) 8807070099 880624 PDR ADOCK 05000259 P

PDC g!'

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1.B.

Power Transient To ensure that the Safety Limits established in Specification 1.1.A are not exceeded, each required scram shall be initiated by its expected scram signal.

The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

2.1.B.

2.

3.

4.

Power Transient Tri Settin s

Scram and isola-

> 538 in.

tion (PCIS groups above 2,3,6) reactor vessel low water level zero Scramturbine 10 per-stop valve cent valve closure closure Scramturbine

> 550 psig control valve fast closure or turbine trip (Deleted) 5.

Scram main 10 percent steam line valve isolation closure 6 ~

Main steam

> 825 psig isolation valve closure nuclear system low pressure C.

Reactor Vessel Hater Level C.

Hater Level Tri Settin s

Hhenever there is irradiated fuel in the reactor vessel, the water level shall be greater than or equal to 378 inches above vessel zero.

2.

Core spray and LPCI actuation reactor low water level HPCI and RCIC actuation reactor low water level 378 in.

above vessel zero 470 in..

above vessel zero 3.

Main steam isolation valve closure reactor low water level 378 in.

above vessel zero BFN Unit 1

1.1/2,1-5

'T.l BASES (Cont'd)

The safety limit has been established at 378 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling.

This point is the lower reactor low water level trip.

REFERENCE 1.

General Electric BNR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10938 and NEDE 10938.

BFN Unit 1

1. 1/2. 1-10

2.1 BASES (Cont'd) decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

C.

Reactor Hater Low Level Scram and Isolation (Exce t Main Steam Lines)

D.

The setpoint for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease.

The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure

barrier, because MCPR is greater than 1.07 in all

cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating range to avoid spurious scrams.

Turbine Sto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

Hith a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.

No significant change in MCPR occurs.

Relevant transient analyses are discussed in References 2

and 3 of the Final Safety Analysis Report.

This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first state pressure.

BFN Unit 1

3.1/2.1-15

'3.2 BASES

. In addition to reactor protection instrumentation which initiates a

reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.

The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a

single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

Hhen necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a

level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.

Such instrumentation must. be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RflR System, Drywell and Suppression Chamber exhausts and drains and Reactor Hater Cleanup Lines (Groups 2 and 3 isolation valves).

The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems.

The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low water level instrumentation set to trip at 378 inches above vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Hater Sample Valves (Group 1).

Details of valve grouping and required closing times are given in Specification 3.7.

These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

BFN Unit 1

3.2/4.2-65

t

~ 4

'I

3.2 BASES (Cont'd)

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray

Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation

and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation;

thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction wi th the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

T>e setting of 200'F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200'F.

The temperat.ure increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident.

Hith the established nominal setting of three times normal background and main BFN Unit 1

3.2/4.2-66

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT 1.1.B.

Power Transient 2.1.B.

Power Transient Tri Settin s

LIMITING SAFETY SYSTEM SETTING To ensure that the Safety Limits established in Specification 1.1.A are not exceeded, each required scram shall be initiated by its expected scram signal.

The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

2.

3.

5.

Scram and isola-

> 538 in.

tion (PCIS groups above 2,3,6) reactor vessel low water level zero Scramturbine 10 per-stop valve cent valve closure closure Scram turbine

> 550 psig control valve fast closure or turbine trip (De 1 e ted)

Scrammain 10 percent steam line valve isolation closure 6.

Main steam

> 825 psig isolation valve closure nuclear system low pressure C.

Reactor Vessel Water Level C.

Water Level Tri Settin s

Whenever there is irradiated fuel in the reactor vessel, the water level shall be greater than or equal to 378 inches above vessel zero.

2.

Core spray and LPCI actuation-'-

reactor low water level HPCI and RCIC actuation reactor low water level 378 in.

above vessel zero 470 in.

above vessel zero 3.

Main steam isolation valve closure reactor low water level 378 in.

above vessel zero

-BFN Unit 2

1. 1/2. 1-5

'l.l BASES (Cont'd)

The safety limit has been established at 378 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling.

This point is the lower reactor low water level trip.

REFERENCE 1.

General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10938 and NEDE 10938.

BFN Unit 2 1.1/2.1-10

~P if)

J

'2.1

'ASES (Cont')

C.

decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

Reactor Water Low Level Scram and Isolation (Exce t Main Steam Lines)'.

The setpoint for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease.

The results reported kn FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure

barrier, because MCPR is greater than 1.07 in all

cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating range to avoid spurious scrams.

Turbine Sto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates (he pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50 percent'reater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.

No significant change in MCPR occurs.

Relevant transient analyses are discussed in References 2

and 3 of the Final Safety Analysis Report.

This scram is bypassed when turbine steam flow is below 30 percent of'ated, as measured by turbine first state pressure.

BFN Unit 2 3.1/2.1-15

3.2 BASES In add1t1on to reactor protection instrumentation which initiates a

reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.'his set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.

The objectives of the Specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a

single failure of any component of such'systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequ'ate performance.

Hhen necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a

level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.

Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RllR System, Drywell and Suppression Chamber exhausts and drains and Reactor Hater Cleanup Lines (Groups 2 and 3 isolation valves).

The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems.

The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low water level instrumentation set to trip at 378 inches above vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Hater Sample Valves (Group l).

Details of valve grouping and required closing times are given in Specification 3.7.

These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

BFN Unit 2 3.2/4.2-65

'3.2 BASES (Cont'd)

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray

Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break. of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation;

thus, the results given above are'applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200'F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200'F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident.

Hith the established nominal setting of three times normal background and main BFN Unit 2 3.2/4.2-66

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT 1.1.B.

Power Transient 2.1.B.

Power Transient Tri Settin s

LIMITING SAFETY SYSTEM SETTING To ensure that the Safety Limits established in Specification 1.1.A are not exceeded, each required scram shall be initiated by its expected scram signal.

The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

2.

3.

4.

Scram and isola-

> 538 in.

tion (PCIS groups above 2,3,6) reactor vessel low water level zero Scramturbine 10 per-stop valve cent valve.

closure closure Scramturbine

> 550 psig control valve fast closure or turbine trip (Deleted)

Scram main 10 percent steam line valve isolation closure 6.

> 825 psig Main steam isolation valve closure nuclear system low pressure C.

Reactor Vessel Hater Level C.

tl

~1I 5 tti Hhenever there is irradiated fuel in the reactor vessel, the water level shall be greater than or equal to 378 inches above vessel zero.

2.

Core spray and LPCI actuation reactor low water level HPCI and RCIC actuation--

reactor low water level 378 in.

above vessel zero 470 in.

above vessel zero 3.

Main steam isolation valve closure reactor low water level 378 in.

above vessel zero BFN Unit 3 1.1/2.1-5

1.1 BASES (Cont'd>

The safety limit has been established at 378 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling.

This point is the lower reactor low water level trip.

REFERENCE l.

General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10938 and NEDE 10938.

BFN Unit 3 1.1/2.1-10

~

2.1 BASES (Cont'd)

C.

decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

Reactor Hater Low Level Scram and Isolation (Exce t Main Steam Lines)

D.

The setpoint for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease.

The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure

barrier, because MCPR is greater than 1.07 in all

cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating range to avoid spurious scrams.

Turbine Sto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

lith a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.

No significant change in MCPR occurs.

Relevant transient analyses are discussed in References 2

and 3 of the Final Safety Analysis Report.

This scram is bypassed when turbine steam flow is below 30 percent of'ated, as measured by turbine first state prcssure'FN Unit 3

3. 1/2. 1-15

3.2 BASES In addition to reactor protection instrumentation which initiates a